ML20006A351

Cycle 16 Startup. W/900118 Ltr

ML20006A351
Person / Time
Site:	Point Beach
Issue date:	01/09/1990
From:	Bruno R, Fay C, Kurtz P WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-90-007, CON-NRC-90-7 VPNPD-90-035, VPNPD-90-35, NUDOCS 9001260156
Download: ML20006A351 (44)

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Wisconsin fElectnc:

-POWER COMPANY

' 231 W Mchgon.Po. Box 204& MAmukeae W $3201 (4t4) 2012345 VPNPD-9 0- 035 -

NRC-9 0- 007 January-18', 1990

. Document Control Desk t ~U.S. NUCLEAR ~ REGULATORY COMMISSION Mail Station Pl-137 s

wa'hington, D.C. 20555 Gentlemen:-

DOCKET NO. 50-301 CYCLE 16 STARTUP REPORT POINT BEACH NUCLEAR PLANT UNIT 2 In accorance with Point Beach Technical Specification 15.6.9.1.A.l.c and 15.6.9.1.A.2.a, we are submitting the attached summary report-of the startup and power. escalation of Point Beach Unit 2 with-a low-low leakage core design using enhanced Optimized

FueliAssemblies.

'Please contact us if you have any questions-concerning this submittal.-

Very_truly yours

[(./ [

C. w. F y Vice President Nuclear Power

' Attachment Copies to NRC Regional Administrator, Region III NRC Resident Inspector

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7'9001260156 960109-PDR ADOCK 05000301 P_ PDC u,&wmans,>a,v ,n ,a s hj

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POINT BEACH NUCLEAR PLANT

!' UNIT 2 CYCLE 16 STARTUP 1

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n TABLE OF C0NTENTS M

LIST OF TABLES. iii

' LIST OF FIGURES iv i- ', PREFACE V lSECTION 1.0, REFUELING 1 1.1 Summary 1 1.2 Core Design 2

'SECTION 2.0, CONTROL ROD OPERATIONAL TESTING 5 2.1 Hardware Changes / Incidents 5 12.2 Rod Drop Times 5.

.2 . 3 ; -Control Rod Mechanism Timing 5 2.4 Rod Position Calibration 6 SECTION 3,0, THERMOCOUPLE AND RTD CALIBRATION 10-SECTION 4.0, PRESSURIZER TESTS 12 4.1 ~ Thermal Transients 12

<4 . 2 Heater Capacity 12 6

, SECTION S.0, CONTROL SYSTEMS 13 SECTION 6.0, TRANSIENTS 13

.SECTION 7.0, INITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKS 13 7.1 Initial criticality 13

,7 . 2 Reactivity Computer Setup and Checkout 14 o

SECTION 8.0, CONTROL ROD WORTH MEASUREMENT 17' 8.1 Test Description 17 8.2 Data Analysis and Test Results 18 8.3- Evaluation of Test Results 19 i

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e .4 Page SECTION 9.0, TEMPERATURE COEFFICIENT MEASUREMENTS 23 SECTION 10.0, BORON WORTH AND ENDPOINT MEASUREMENTS 24 ,

SECTION 11.0, POWER DISTRIBUTION 26 l SECTION 12.0, XENON REACTIVITY- 29 i SECTION 13.0, SHUTDOWN MARGIN CONSIDERATIONS 29 SECTION 14.0, EXCORE DETECTOR BEHAVIOR - 30 14.1 Intermediate Range Detectors 30 14.1 Power Range Detectors 30 SECTION 15.0, OVERPOWER, OVERTEMPERATURE AND DELTA FLUX  !

SETPOINTS, CALCULATION 33 15.1- overpower and Overtemperature AT Setpoints calculation- 33 15.2 Delta Flux Input 33 SECTION 16.0, FUEL PERFORMANCE 36

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SECTION 17.0, CONCLUSION 36 I

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't LIST 0F TABLES Table gage n

2-1 . cycle 16 Control Rods 7 - -

i 3-1 RTD Calibration Check 11 4-1 Heater Group Power Supply Readings 12  ;

7-1 Reactivity computer Setup 15 <

2 Reactivity Computer Checkout 16 L'. 8-1 Critical Rod Configuration Data 20 L 8-2 Comparison of Inferred / Measured Bank Worths

.with Design Predictions 21 10 Boron Worth and Endpoints 24 ,

1~ 11-1 Initial Power Escalation Flux Map Results 26 l:

13-l' . Excess Shutdown Worth Available for a Full l' Power Trip . 29

', 14-1 Power Range Detector BOL Calibration Currents (105%) 13 1

- 15-1 Overpower AT Constants 34 15-2 -Overtemperature AT Constants 35 i

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'i LIST OF FIGURES -

.s Figure M 1-1 U2C16 Core Loading 3 1-2 BOL Burnup Data 4 2-1 Cold Rod Drop Times (Full-Flow) 8 I' '

2-2 Hot Rod Drop Times (Full-Flow) 9 8-1 Reference Bank Differential Worth 22 10-1 BOL HZP Boron Concentrations 25 11-1 Axial Power Shape at BOL, 28% Power 27 11-2 Axial Power Shape at BOL HFP, ARO, EQXE 28 14-1 Intermediate Range Detector Response to Power Level 32 16 RCS Iodine Levels 37 i

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h PREFACE This report is intended to document in a concise format the results of ~!

the physics testing program and unit systems response during the startup'of '

Unit 2 following Refueling 15. The organization of the report follows that utilized in previous ~startup reports. (

Westinghouse performed the core design calculations for Unit 2 Cycle 16. l' The reactivity coefficients were calculated based on estimated cycle 15 burnup of 10,200 MWD /NTU. Actual cycle 15 burnup was 10,205 MWD /MTU.

Cycle 15 was ended on September 23, 1989, with a peak assembly burnup of 45,996 MWD /MTU and average assembly burnup of 28,622 MWD /NTU. Electrical power was first generated during cycle 16 on November 25, 1989.

This report is intended primarily for the use of Wisconsin Electric Power Company personnel as a readily. accessible, complete compilation of.

reduced data.

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1.0 REFUELING 1.1 Summary f The core was completely unloaded so that the core barrel could be removed from the reactor vessel for a Section XI ten year reactor vessel .

inspection. The unload started on October 10,1989,: seventeen days af ter Cycle 15 end of life shutdown. The unload went smoothly and was completed l on October 12, 1989. A discharge fuel assembly (P56) experienced a torn grid when being lowered into spent fuel pit location SD-13.

• All necessary fuel assembly insert changes for cycle 16 were made in the spent fuel pit from October 12, 1989 to October 17, 1989 using one ,

10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. shift. All plug devices were removed from reload fuel i assemblies.- None of the inserts were damaged. The insert tools

• worked properly for the duration of the moves.

After completion of the insert moves, all of the Cycle 15 fuel assemblies were ultrasonically tested (UT) for leaking fuel rods using the company owned equipment. In addition, sever 61 candidates for reuse, discharged from prior cycles, were tested. No leaking fuel  ;

rods were found. Average time to perform a UT in the SFP was 30 minutes. >

c The core reload started on November 6, 1989 and was completed on November 10, 1989. Several changes were made-to the sequence to park bowed' or twisted fuel assemblies in temporary locations until their locations were boxed. All temporary core configurations conformed to the procedural L requirement that a 2x2 array of new fuel shall not be made, ensuring adequate l shutdown margin. No fuel handling mishaps occurred during the reload. . Once

l. baseline count rates were. established for the excore detectors, the count rates did not' change throughout the remainder of the reload sequence. The introduction of the sources in core locations H-3 and F-11, approximately '

doubled the count rates on source range dete tors N31 and N36 from about 80 l and 70 cps to 210 and 180 cps respectively. The spare detector located away from the sources (wide range detector channel N40), responded with about '

15 cps throughout the reload.

No major equipment breakdowns occurred during the reload. A loose fitting on the fuel gripper air cylinder had to be tightened, resulting in about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of down time. All fuel movement operations and UT testing were performed by PBNP personnel. The final core configuration was verified to match the design core configuration, by using a TV camera to scan the top of the core, reading the fuel assembly ID numbers and insert

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1.2 core Design The final core configuration is shown in Figure 1-1 and matches the Westinghouse designed core layout shown in WCAP-12362, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor, Cycle 16."

The as-loaded burnups for each fuel assembly are shown in Figure 1-2.  ;

New fuel consisted of Vantage-5 type fuel assemblies, twelve with an enrichment of 3.4% U-235 and sixteen enriched to 3.8% U-235. No axial blankets, boron coated fuel pellets or extra grids for flow mixing were used in this core design. However, other Vantage-5 design features were used for the first time, such as removable top nozzles and debris catching bottom nozzles. ,

of the 121 assemblies loaded, 115 are of the Optimized Fuel Assembly (OFA) design (also a Vantage-5 feature) and 6 are of the older standard design with wider fuel rods (0.422 inches diameter vs. 0.400 inches diameter for 0FA's).

Besides the normal array of inserts, part length Hafnium rods were loaded on the " flats" of'the core to decrease neutron fluence to the reactor vessel welds nearest the core. This will help extend reactor vessel lifetime.

All plug devices were removed for Cycle 16.

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All the new features of this core design.were analyzed in WCAP 11872,

" Final Report for Increased Peaking Factors and Fuel Upgrade Analysis, Point Beach Nuclear Plant, Units 1 and 2, June 1988."

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FIGURE 1-1 CORE LOADING A- 4 &- 7 &- 8 053 055 074 2R205 2R207 2R203 8- 4 B- 5 B- 6 B- 7 B- 8 B- 9 B-10 Q79 R74 T65 T52 T68 RSO Q77 R98 4P142 R135 C- 3 C- 4 C- 5 C- 6 - C- 7 C- 8 C C-10 C-11 N23 . T70 873 R54 872 R51 874 776 Q10 R133 R32 R111 D- 2 -D D- 4 D5 D- 4 D- 1 D- 8 D- 9 D-10 'D-11 D-12 Q73 777 Q68 860 T54 K58 T55 861 058 T64 Q60 R53 Spill SP114 R11 5- 2 5- 3 B- 4 .B. 5 B- 6 3- 7 B- 8 B- 9 B-10 E-11 3-12 R81 868 852 R72 R53 870 RSS R78 855 865 D68 R71 R28 R115 R5 F- 1 F- 2 ' F F- 4 P- 5 F- 4 F- 7 P- 8 P- 9 F-10 P-11 F-12 P-13 Q65' T78- R62 T60 R65 R69 854 R70 RS2 T61 R64 .T66 Q66 2H211 R84 BP110 R14 R18 SP113 589 R103 2H208 0- 1 , G- 2 G- 3 6- 4 e- 5 6- 6 G- 7 6- 8 e- 9 G-10 0-11 0-12 0-13 Q56 'T51 876 K54 871 851 M19 862 864- R53 867 T62 Q81-2H212 4P141 R31 R110 R7 4P143 2H216 E- 1 E- 2 E- 3 5- 4 E- 5 E- 6 E- 7 B- 8 E- 9 E-10 E-11 E-12 E-13

, Q71- T73 R55- T53 R59 R77 856 R79 R60 T58 R66 T74 Q82 2H204 R2 -8810 8P109 R114 R8 SP112 R149 2H209 I- 2 I- 3 I- 4> I- 5 I- 6 I- 7 I- 8 I- 9 I-10 I-11 1-12 R75 875 559 RS2 R57 869 R63 R67 857 863 R76 R116 R10 R112 R29 J- 2 'J- 3< Ja 4 4- 5 J. 6 'J7 J- 8 J9 J-10 J-11 J-12 Q64 T71 072: ~ 883 T59 K64 T56 858 @7 T72- Q61 R34 SP108 Spill R17 E- 3 E=-4 E- 5 R- 6 R- 7 E- 0 R- 9 K-10 R-11' 069 T67 >878 RS6 877 R61 866 T69 Q52 R127 R107 R109 L- 4 L- 5 L- 6 L- 7 L= 8 L. 9 L-10 959- R71- T63 T57 T75 R73 062 R126 4P140 R139 M- 6 M- 7 M- 8 Q54 Q76 078 2H213 2H206 28214 UNIT 2 CYCLE 16 FINAL CORE CONFIGURATION 10:06:13 a.m. 12/11/89 7'

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l BOL BURNUP DATA .

PDNP UNIT-2 START OF CYCLE 16 1 2 3 4 5 6 7 8 9- 10 11 12 13 953 955 974 A 34632 36057 35106 215' 235- all 979 R74- T65 T52 T68 . R80 977-B 35741 26300 0 0 0 26550 36601 all' -2168 218s 2184 2108 2165 til M23 T70 873 R64 872 RS1 874 776 970 c 31782- . 0 10002 25765 11911 25409 10871 0 36490 1135 2108 2175 ' 216A 217B 216A 2179 2185 = 215 073 T77 968 560 754 RSS 756 861 . 058 764 960 D 36255 0 32179 12396 0 35759 0 12756 32337 0 35901 215 2185 215' 217A 21th 210 218A 217A 215 2108 215 R81 568 852 R72 R53 870 Alt R10 855 865- R60 t 26712 11043 '12767 24519- 25043 11754 25155 24844 12746 10994 2G437 2165 217B 217A 2165 216A 2173 216A 3168 217A 2175 2168 965 T70 'R62 . T60 R65 R69 854 R10 RS2 T61 . R64 766 946 F 35143 -0 25937 0 24942 23935 12893- 24410 25418- 0 24197 0 35311 215 2100 216A 214A . 216A ,2165 217A- 2168 216h 218A 216A 2108 215 d.

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956 T51 - 576 K'54 sti all M19 e62 - s64 R53 867 T62 941 0 35928- 0 11904 35650 11838 12353 28529 12775 -11739 35460 11986 0 36412 215 218A 2175 210 2178 217A 1128 217A 217s 210 2175 218A 215 071 773 R55 T53 R$9 R77 856 R19 R60 T50 - R66 774 982 H 34933 0 '26498 0 24573 23783 12909 24204 25789 0 25192 0 35515 215 2108 216A .214A 216A 2168 217A 2165 216A 218A 216A 2188 215 R75 875 859 RS2 R$1 869 R63 R67 857 863 R76 .

21 26095' 11020 12061 24428 25243 litet 25287 24763 12010 11000 26307 2165 2175 217A 2168 216A 2178 216A 216s 217A- 2175 2168 064 T71 972 853 T59 R64 T56 558 067 T72 061 J 36065 0 31890 12322 0 35623 0 12638 32202 0 35122 215 2lte 215 217A 218A 210 218A 217A 215 2183 215 069 T67 s7e R56 s77 R61 see T69 067 R 35114 0 10893 25502 11839 26536 10979 0 39950 215 2188 2178 216A 2178 216A 217B 2185 - 215 959 R71 743 T57 T15 R73 062 L 35938 26585 0 0 0 26211 36172 215 2168 2188 218A 2188 2165 215

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2.0 CONTROL ROD OPERATIONAL TESTING 2.1 Hardware Changes / Incidents Table 2-1 shows which control rods were carried over from Cycle 15 and which control rods were replaced. Of the replacements, five were new rods from storage in the new fuel vault (R98, R126, R135, R139, and R149).

The others were cleared for reuse based on eddy current testing in January, 1989. All rods carried over from Cycle 15 had less than 7 cycles of operation.

One replacement control rod (R71) had hairline cracks at the rodlet tips which were found during visual inspections in 1983. At that time Wisconsin Electric committed to not using R71 again. In August of 1989 R71 was picked as a replacement rod based on eddy current testing results showing acceptable wear. After R71 was loaded in the core and the reactor vessel head was in place, the documentation on the crack was found in the file for R71. Westinghouse was contacted to reevaluate the condition of R71. Based on a broadened database of control rod wear, Westinghouse recommended that R71 could be used for one more cycle. Therefore, R71 was left in the core for Cycle 16 and the NRC was notified.

All wiring connections to the RV head for rod position indication and rod control were replaced during the refueling outage. After correcting some wiring mistakes discovered during rod drop testing, the RPI and rod control systems functioned nonmally.

2.2 Rod Drop Times See Figures 2-1 and 2-2 for rod drop times and RCS parameters.

Cold rod drop times for some rods were slightly faster than the drop times for other control rod banks because the rods were dropped from 223 steps instead of 228 steps. Some step counters were off by 5 steps as a result of an oversight during rod stepping tests just prior to rod drops.

All rod drop times were well within the Technical Specification limit of 2.2 seconds (15.3.10.E).

2.3 Control Rod Mechanism Timing Traces of control rod gripper coils currents were obtained for all rods. All traces of the lif t, moveable and stationary coil currents were considered satisfactory after correcting a wiring mistake at the RV head for Rod D4.

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2.4 Rod Position Calibration During hot rod drop testing, LVDT voltages were read at 20 steps and 200 steps to determine if any voltages were abnormal. Additional readings were made on Control Bank D rods every 20 steps to verify the new head connections had not changed the RPI coil characteristics. Each plot of voltage vs. step was normal.

"Zero" adjustments were made with rods at 20 steps under hot zero power full flow conditions.

" Span" adjustments were made at full power after rods were verified to be fully withdrawn using RESP 1.2, " Rod Control System: Rod Position Verification and Rod Position Indicator Alignment."

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, , - 4 TABLE 2-1 CYCLE 16 CONTROL RODS DAscharged Used in From Cycle 15 Replacements Cycle 15 R54 R98 R107 R65 R135 R133 R79 R32 R114 R62 R53 R109 R68 R11 R110 R57 R71 R111 R76 R28 R116 R82 'R5 R127

.R69 R14 R112 R55 R18 R115 R81 R31 R103

_R61 R7 R84 R64- R2 R73 R8 R63 R149 R83 RIO RS6 R29 R72 R34 R66 R17 R77 R126 R80 R139 t

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FIGURE 2-1 1

PBNP U2C16 COLD ROD DROP TIMES-i '2 3 4 5 6 7 8 9 10 11- 12 13-1 1 I

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$A SA B 1.56 1.71 4 2,06 2.17 l CA CD CA C 1.53 1.64 1.67 2.11 2.17 2.16 CC CC

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i. 2.06 2.15 CA 58 $8 CA E 1.60 1.54 1.68 1.61 2.10 2.08 2.1s 2,14' 5A C8 CS SA F- 1.63 1.s7 1.66 1.62 2.11 2.10 2.20 2.06 CD CC CD G- 1.56 1.4s 1,64 2.05 1.87 2.09 SA CB CB SA H- 1.56 1.n 1.6s 1.66 2.03 2.22 2.23 2.22 CA $8 $8 CA

v. I 1.61 1,66 1.69 1.60 2.13 2.16 2.17 2,06

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J 1.57 1.54 2.12 2.09 CA CD CA K 1.70- 1.65 1.55 2.17 2.18 2.02 SA SA

( 1.70 2.18 1.57 2.07 M

LEGEND BANK DATE 11/20/89 '

x.xx - Time To Dashpot (sec) x.xx - Time To Seat (sec) TEMP 180 'F Maximum drop time (dash) = H 6 1.72 FLOW 100 %

Minimum drop time (dash) = G.7 1.45 Average time (dash) = 1.61 PRES 330 PSIA 8 .

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I FIGURE 2-2 i PBNP U2C16 HOT ROD DROP TIMES '

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1.28 1.x 1.82 1.88 CA CD CA **

C 1.P6 1.29 1.23 1.85 1.91 1.86 CC CC D 1.25 1.28 1.82 1.85 CA $8 88 CA E 1.24 1.28 1.28 1.28 1.84 1.90 1.86 1.90 5A CB CB $A F- 1.29 1.26 1.28 1.2r 1.2.3 . 1.90 1.91 1.80 ,

CD CC CD G- 1.24 1.30 1.07 1.82 1.79 1.82

$A CB C6 6A 4 H-1.26 1.s2 1.s2 1.30

. 1.80 1.94 1.96 1.89 y CA se st CA 1 1.29 1.26 1.28 1.25

, 1.86 1.89 1.90 1.80 CC CC J- 1.28 1.86 1.27 1.84 CA CD CA K 1.30 1.29 1.24 1.85 1.89 1.80 5

&A 6A L 1.33 1.28 1.88 1.81 M

LEGEND BANK DATE 11/23/89 m.xx - time To oeshpot (sec) m.Kx - 11me to lest (Sec) TEMP 530 'F Maxinun drop time (desh) e 8 8 1.36 FLOW 100 %-

Minlaun drop time (desh) e C 9 1.23 Averese time consh) = 1.28 PRES 1995 PSIA 9

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3.0 THERMOCOUPLE AND RTD CALIBRhTION During initial RCS heatup for Cycle 16, loop RTD's and incore thermocouples were checked for nomal response throughout the heatup range of about 300*F to HZP. Table 3-1 gives each RTD temperature, steam generator temperature and average core exit thermocouple temperature for eight different measurements during the heatup. All 16 RTDis were within the expected 2*F deviation of each other throughout the heatup. Core exit thermocouple 210 was the only thermocouple not responding.

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LOOP A + COLD LEC R 4010 315.62 350.55 417.67 447.19 474.27 497.55 520.73 530.51 R 4058 315.51 350.40 417.31 4 4 .73 473.75 4M.95 520.09 529.78 W 4028 315.30 350.28 417.63 44.77 473.E 496.90 520.12 529.85 W 4068 315.40 350.4 417.00 4 4 .76 474.01 496.93 520.14 579.85  ;

LOOP A

• NOT LtC i a 401A 315.50 330.43 417.61 447.02 473.84 497.03 520.35 530.14 ,

k 405A 316.16 351.11 418.32 447.55 474.M 497.41 520.M 530.46 W 402A 315.56 350.M 417.92 447.03 474.11 4M.90 520.27 550.07  !

W 406A 315.24 350.29 417.49 44.54 473.63 4M.37 519.72 529.53 ,

t LOOP 8

• COLD LEG B 4038 315.60 350.69 418.01 447.18 474.25 496.80 $20.45 530.34 l B 4078 316.06 351.22 418.55 4A7.62 474.80'497.44 520.86 530.74  ;

Y 4048 316.49 351.72 418.88 447.92 474.98 497.22 520.69 530.75 Y 4088 315.83 351.01 418.24 447.33 474.45 496.69 520.23 530.33 .

LOOP S

• NOT LEG B 403A 315.42 350.52 417.67 446.M 473.81 496.33 519.61 529.54 ,

t 407A 315.35 350.52 417.76 4 4 .82 473.91 496.23 519.78 529.83 Yet4A 317.02 352.28 419.37 448.21 475.37 497.62 520.87 530.06 Y 408A 316.21 351. 4 418.57 447.55 474.91 497,20 $20.M 530.75

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RTO AvtRAGE 316 351 418 447 474 497 520 530 5.G. SAT. TEMP 307 348 41f 44 4 74 496 520 530 CORE EE!V T/c TEMP 319 355 419 449 475 497 520 530 l

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4.0 PRESSURIZER TESTS 4.1 Thermal Transients i Pressurizer pressure increase rate with spray valves indicated shut i and all heaters on was 13 psi / min. This is close to the nominal value of ,

14 psi / min. During the thermal equilibrium test, heater group D was cycled ,

on about 3/4 of the time to maintain pressure with main spray valves shut.  :

Spray valve effectiveness was normal with pressure decreases greater than 110 psi / min. _!

Spray bypass valve positions were not changed as the result of these ,

tests.

4.2 Heater Capacity Pressurizer heater capacity was determined from direct volt / amp readings l

on each group of heaters. Table 4-1 shows that heater capacity is above

l. Technical Specification requirements of 100 KW minimum total. ,

- TABLE 4-1 l HEATER GROUP POWER SUPPLY READINGS I

Heater I-Current V-Voltage KW-Energy Input '

Group (amps) (volts) KW = d x V x I/1000 ,

A 287 484 240 l

B 237 479 196 C 237 484 198 D 220 481 183 E 233 476 192 TOTAL 1009 l

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CONTROL SYSTEMS i . There were no difficulties encountered during heatup or startup of the pressurizer level, pressurizer pressure, and rod control systems.

6.0 TRANSIENTS

! There were no transient tests performed during startup or approach to full power. There were no. violations of the fuel conditioning restrictions on power and rod stepping rates.

7.0 INITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKS 7.1 Initial Criticality The approach to criticality was made in two phases. The first step, which began at 0139 hours0.00161 days <br />0.0386 hours <br />2.29828e-4 weeks <br />5.28895e-5 months <br /> on November 24, 1989, was the withdrawal of control rods until Bank D reached 180 steps. The reactor coolant boron concentration was then decreased by dilution until criticality was achieved.

The dilution rate was 97 pps/hr or 50 gpm. The critical boron concentration of 1267 ppm was close to the predicted value of 1269 ppm. ICRR plots were maintained during each phase of the approach to criticality. All plots turned out to be normal, .;

The reactor conditions at the time of criticality were determined to be as follows:

Date November 24, 1989 Time 0G06 RCS Temperature 530'F RCS Pressure 1985 psig '

l- Rod Position Bank D at 178 steps f

Boron Concentration 1267 ppm i

The intermediate range detector trip block permissive came in with source range counts between 50,000 and 70,000 CPS.

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7.2 Reactivity Computer Setup and Checkout 7.2.1 Setup Table 7-1 shows the reactivity computer setup results. Test 1 is a static test which tests for the reactivity zero point. Test 2 is a dynamic test which inputs an exponentially increasing flux to test for a positive reactivity output.

7.2.2 Checkout Following criticality, acceptable zero power physics testing flux levels were determined. The flux level at which nuclear heat appeared was about 5 microamps. Normal flux levels for physics testing are about j one-third t!.e point of adding heat by procedure. i The reactivity computer's response was also checked using actual core flux. Control Bank D was pulled from a critical position to obtain distinctly different reactivity levels. For each reactivity level, flux doubling time was measured with a stopwatch. Measured reactivity was then compared to design reactivity calculated from the measured doubling time. Table 7-2 shows the results.

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t TABLE 7-1 -

i klul4C OlppWitt MfWP Uulf CTCLE Daft ButtiP NTA I L*tfAR the/ItTU 701AL (MS)

I 16 12 13 99 0 0.006017 0.97 '17.6 f M LAYt0 group 1 2 3 4 5 6 M1A FRACflou 0.000196 0.001264 0.001116 0.002300 0.000871 0.000210 LuseDA 0.0128 0.0315 0.1211 0.3222 1.4M0 3.0565 IWPut PCT Ipestt i1 12 21 22 31 32 StiflNG 1.2164 3.0010 1.3109 3.7192 1.1962 0.7856  ?

u ttn ,i 1.216 3. sol I.sil 3.720 1.186 0.7Jt5- ,

At LtFT 82 l

FitD64tC P01 meettt 13 14 23 24 33 34 StifthG 1.2800 3.1500 1.2110 3.2220 1.4040 3.8565 u Ltn ei 1. 2 F0 3.l50 l.2il 3.221 f.403 3.3 5 7 '

Al LIFT #2 -

I tisf 1 Sif Pot 3610 9.100 (v0Lis). POT 35 tietWLD M 5.0365 As Ltti #15.8 6 a ten ,2 .

ADJWET POT 35 UNill AflPLIFitt 14 (AHO) GUTPUT ll 0.0 VOLit.

• AlePtif tta meeHR 11 12 21 22 31 32 ApePLIFitt v0 Lit 8.65M6 10.90079 9.85093 10.5N13 7.60431 1.05367 L u ttn ei B.65 /0.99 9.8'1 /O.55 7.48 ,lg5 As LIFT 82 l

l i 1851 2 MT Pot 26 To Ataff 0.75 Y l

Pot 25 SEffluc 0.20 0.50 0.00 1.10 1.40 1.70 2.00 2.30 2.60 PEtl0D (SEC) 500.00 200.00 125.00 90.91 71.43 54.81 50.00 43.48 38.46 T DeLO (sic) 346.57 138.63 86.64 63.01 69.51 40.77 34.66 30.14 26.66 0.utno i.o ei 350 /39 88 63 50 '11 35 3,q., 2 7 DesttVED T 0 #2 EXPECit0 Reto (PCM) 13.09 30.03 44.59 57.39 68.82 79.17 48.62 97.32 105.39 Deutno teto ei 13.5 3o.0 ft.5 5Z4 68.7 79.3 68.7 97 F /od.o Oesteyto ano #2 1 Daft Ikifl At t 10/61 '

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TABLE 7-2 .

I

+ REACTIVITY COMPUTER OECKOUT  ;

Measured Design Measured calculated Doubling Doubling Reactivity Reactivity Time (sec.) Time (sec.) (pen) (pem) ,

88.2 83.4 46 44 45.6 53.5 65 73 30.0 38.1 83 98 4

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8.0 , CONTROL ROD WORTH MEASUREMENT 1 1

8.1 Test Description j The rod worth verification utilizing rod exchange (" rod swap") was '

divided into two parts. . In the first part, the reactivity worth of the reference bank was obtained from reactivity computer measurements and boron <

endpoint data during RCS boron dilution. In the second part, the critical height of the reference bank was measured after exchange with each remaining bank.

In the rod exchange technique, the reference bank is defined as that bank which the highest worth of all banks, control or shutdown, when inserted into the core alone. For this cycle the reference bank was control Bank A (CA) as was the case in all prior rod swap tests.

Using the analog reactivity computer, reactivity measurements were ,

made during the insertion of control Bank A from the fully-withdrawn to the fully-inserted position. The average current (flux level) during the measurement was maintained within the physics testing range and temperature was held steady near 530*F. Critical boron concentration measurements (boron endpoints) were made before and after the insertion of control Bank A (see Section 10.0). Figure 8-1 shows the results of the differential worth measurements, r Starting at a critical position with the reference bank fully inserted and Control Bank C at 211 steps, a new critical configuration at constant RCS boron concentration was este.blished with control Bank C fully inserted and Control Bank A at 104 steps. Control Bank C was then withdrawn and control Bank A inserted to one step to establish the initial conditions for l the next exchange. This sequence was repeated until a critical position >

L' was established for the reference bank with each of the other banks individually inserted. Criticality determinations before and after each t- exchange were made with the reactivity computer.

The sequence of events during the rod exchange and a summary of the rod exchange data is presented in Table 8-1.

17

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. j 1

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8.2 Data Analysis and Test Results The integral reactivity worth of the measured bank is inferred from 4 the swapped portion of Control Bank A by the following equations j l

WI = { - ap - (a ) x(aps) + E where g  !

i

'- W. f=TheinferredworthofBankX,pcm

[R = The measured North of the reference bank, control A, from fully withdrawn to fully inserted with no other bank in the core.

1 a = A design correction factor taking into account the fact that the x

presence of another control rod bank is affecting the worth of ,

the reference bank.  :

402 = 2he measured worth cf the~ reference bank from the elevation at which the reactor is just critical with Bank X in the core to the  ;

reference bank fully withdrawn condition. This worth was ,

measured with no other bank in the core.

aps = The measured worth of the reference bank from the fully inserted I condition to the elevation at which the reactor was just critical ,

prior to the worth measurement of Bank X. In this test opg is .

zero.  ;

E = The worth of Bank X from the initial position (before the start '

X of the exchange) to 228 steps. This worth is measured by the ,

normal endpoint worth method.

Final values for the integral worth of control and shutdown banks inferred from the measurement data are tabulated in Tabic 8-2. Values for a,were obtained from the design predictions are also listed in Table 8-2.

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8.3 Evaluation of Test Results L

A comparison of the measured / inferred bank worths with design predictions is presented in Table B-2.

In evaluating the test results, the standard review and acceptance criteria below were used. .

Review Criteria

a. The measured worth of the reference bank agrees with design predictions within 110%.

b. The inferred individual worth of each remaining bank agrees with '

design predictions within 115% or 1100 pcm whichever is greater.

c. The sum of the measured and inferred worths of all control and -

shutdown banks is less than 1.1 times the predicted sum.

Acceptance criteria .

The sum of the measured / inferred worths of all control and shutdown banks is greater than 0.9 times the predicted sum.

As shown on Table 8-2, all review and acceptance criteria were met. ,

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1 TABLE 8-1 1 CRITICAL R0D CONFIGURATION DATA Measured '

RCS CA Bank

[] Bank Tavg Position Position ,

Measured Time (*F) (Steps) (Steps)  ;

CC 1810 530 1 211 CC 1820 530 104 1 CC 1832 530 1 211 SB 1843 530 1 213 ,

SB 1855 530 93 1  :

SB 1910 530 1 212 SA 1950 530 1 214.

SA 2000 530 143 1 SA 2030 530 1 214

- t CB 2035 530 1 216 CB 2055 530 77 1 CB 2112 530 1 216 CD 2134 530 1 214 CD 2147 530 127 1 CD 2206 530 1 214 L Boron concentration was 1126 ppe.

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i TABLE 8-2 COMPARISON OF INFERRED / MEASURED BANK WORTHS

!, WITH DESIGN PREDICTIONS L ,

k7 E

W X

X

( ) x 100 Bank X $ M M _(pem). (%)

n CC 869 1.009 36 854 892 -4.3 SB 967 1.043 33 719 782 -8.0 SA 574 0.897 32 1211 1127 +7.4 CB 697 1.083 33 483 553 -12.7 CD 1149 - 0.982 32 1042 1024 +1.8 CA --- ----- --- 1694 1641 +3.2 TOTAL 6003 6019 -0.3 21

t FIGURE 8-1  ;

1 PBNP UNIT 2 CYCLE 16 BOL HZP ,

REFERENCE BANK DIFFERENTIAL WORTH )

16 ,

i 0

• MEASURED WITHIN 10% OF DEllCN 15 1 >

00 x . MEASURED QUTSIDE s10% OF DEllCN

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0 20 40 60 80 100 120 140 160 180 200 220 l . STEPS WITHDRAWN 22 e

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9.0 ,

TEMPERATURE COEFFICIENT MEASUREMENTS I A near all rods out isothermal temperature coefficient measurement was taken during zero power physics testing. The measurement test conditions and results are given in Table 9-1. The measured values are the average of the recorded reactor coolant system heatups and cooldowns. Reactivity from ,

the reactivity computer and reactor coolant system temperature were recorded on an X-Y plotter and two pen recorder.

Measured ARO temperature coefficient was -0.3 pcm/*F, within the review criteria of 13 pcm/'F of the design isothermal temperature coefficient of

+0.1 pcm/'F.

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i 10.0 BORON WORTH AND ENDPOINT lEASUREMENTS Figure 10-1 shows RCS boron concentration during zero power physics

• testing. Table 10-1 shows results of the endpoint measurements. The '

measured boron worth was obtained by dividing bank worth (pca) into change in boron concentration between endpoints. The review criterion of .

10.5 pcm/ ppm was met.

=

TABLE 10-1 BORON WORTH AND ENDPOINTS Endpoint Bank Worth Boron Worth Bank Design 1 Measured Design Measured Design Measured configuration (ppm) (ppm) (pem) (pem) (pem/ ppm) (pem/ ppm)

ARO 1289 1286 ---- ----

CA in 1129 1126 1641 1694 -10.3 -10.6 3 At measurement conditions (530'F) s F

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4 11.0 POWER DISTRIBUTION Table 11-1 illustrates the margin of hot channel factors to their full power limits during initial power increase to full load. Flux maps were taken using ANSI Standard ANS-19.6.1-1985 as guidance. Allowed power levels .i were calculated using the relationships for FAH and FQ versus power level ,

in Technical Specification 15.3.10.B.1.a. The overpower trip setpoint was e initially set at 83% power to ensure peaking factor limits were not >

excee de d. After the 75% power flux map was taken, the setpoint was raised to its normal value of 107% power.

Measured axial power distribution compared to design is. shown in  !

Figure 11-1 and 11-2. The map taken at 28% power was with rods inserted about 10 steps deeper than the design curve was generated for. This accounts for the difference in shapes of the curves in Figure 11-1.

2 i

TABLE 11-1 INITIAL POWER ESCALATION r FLUX MAP RESULTS

, o Flux Map Power Thimbles Allowed Number Date _1$1, Missing Power (%) Bank D A0 FANN g

! 1 11-25-89 28 5 92 95 165 +3.8 p.

L 2 11-27-89 75 5 113 112 189 +3.6 L 3 11-28-89 95 0 101 110 197 +1.1 L

L 4 11-29-89 100 0 102 112 200 +2.5 L

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i FIGURE 11-1 l POINT BEACH UNIT 2 CYCLE iS  !

CORE AVERAGE NORMALIZED AXIAL POWER DISTRIBUTION  :

28% POWER BOL {

actuoi otsica cave - - - usato cave  !

1.4 ,

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FIGURE 11-2 POINT BEACH UNIT 2 CYCLE 16 CORE AVERAGE NORMALIZED AXIAL POWER DISTRIBUTION HFP BOL EQXE L(GtND D[$1GN CURV( - . - . - M(AguR(D CURy[

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12.0 XENON REACTIVITY Xenon reactivity behavior data for Unit 2 Cycle 16 was supplied by Westinghouse separate from the WATCH data package. Point Beach code Xenon 5 will be run with a TDF1 of 0.95 and TDF2 of 1.2 to remain consistent with i the Xenon Tables. Tables are supplied for BOL, MOL and EOL conditions. .

13.0 H S_UTDOWN MARGIN CONSIDERATIONS ,

Rod swap results were within acceptance criteria and were accepted as valid proof of rod worth for shutdown margin determination. See section 8.0 for rod swap details. Thus WCAP-12362 Table 6.2 was accepted as a valid "

shutdown margin determination. Table 13-1 calculates the excess worth available to Unit 2 Cycle 16.

TABLE 13-1 EXCESS SHUTDOWN WORTH AVAILABLE FOR A FULL POWER TRIP BOL (Dem) EOL (pen)

Shutdown Margin 3850 -3510 i

. From WCAP

- Required Shutdown -1000 -2770

= Excess Worth -2850 -740 1

29

i i .

l 14.0 EXCORE DETECTOR BEHAVIOR i

14.1 Intermediate Rance Detectors In anticipation of a reduction in intermediate range detector currents from the hafnium rods positioned in front of the detectors, the source range trip setpoint was raised from 1 x 105 to 5 x 105 CPS. This gave the operator more time to block SR trip after reaching 1 E-10 amps on the IR channels.

Because actual IR attenuation was less than expected, the original SR trip t setpoint would have given adequate but less margin for blocking the trip.

Intermediate range detector currents versus power level are shown in Figure 14-1. Intermediate range detector trip signals activated at about '

2.2 E 4 amps for N35 and 1.3 E-4 amps for N36. Excore detector power level at the time the trip signals occurred was 26% for N35 and 20% for N36. The hafnium poisons reduced intermediate range detector output by about 20%.

The pre-startup trip setpoint for N36 was 2.4 E-4 amps. After the setpoint was set in, the detector was replaced. The new detector had a lower sensitivity and its level had not reached the trip setpoint by the }

time 28% power was reached. The trip would have occurred just below the ,

Technical Specification limit of 40% power. A new setpoint was entered based on actual detector output for 20% power.

14.2 Power Range Detectors

• Table 14-1 lists the " tilt free" power range detector calibration citrrents corresponding to 105% power at BOL. These currents were ,

calculated using the multi-map method at 100% power. A multi-map calibration was performed to verify that the new changes in core design did

- not significantly change the linear response of the encore detectors.

Output of both the top and bottom detectors was reduced by about 5% because of the hafnium rods and L4P (low low leakage loading pattern). ,

Power range quadrant tilt alarms are designed to alert for rapidly developing tilts. Natural core tilts are eliminated by obtaining calibration currents for the core with a tilt. A tilt is indicated only when actual currents deviate from the calibration currents even though the core already i may have a tilt before the start of the deviation. This practice complies with Technical Specifications and the Westinghouse position on core tilt.

1 l

30

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i l- TABLE 14-1 I

i POWER RANGE DETECTOR BOL CALIBRATION CURRENTS (10$%)

41 42 43 _44 Cycle 14 7 314 - 334 350 305 B 263 297 317 277 i Cycle 15 7 267 280 286 256 B 229 255 265 238 Cycle 16 7 253 260 280 234 B 211 232 259 215 e

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l FIGURE 14-1 ,

UNIT 2 CYCLE 16 BOL NI35 AND NI36 RESPONSE TO POWER LEVEL l 9_ -

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0 10 20 30 40 50 60 70 80 90 100 REACTOR POWER LEVEL (%)

32 t

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15.0 ,

OVERPOWER, OVERTEMPERATURE AND DELTA FLUX SETPOINTS CALCULATION 15.1 overpower and overtemperature AT Setpoints calculatior.

Discussion of the setpoints and equatiens has been sufficiently covered in previous reports.

The equations are:

I Overpower AT ( 3,z,3 )

tS 1ATo [K4 - Ks(s z 3 ,3)( 3fz43) T - Ks (T( IftS)~ 4 I

1 Overtemperature AT( 3,;s3 )

1+T S I I

$ATo (Kg - K (T( 1,z,3 )*T )( 3,z 3 )

• K3 (P-P ) - f(AI))

2

. l See Tables 15-1 and 15-2 for the constants associated with this cycle of operation.

4 i

15.2 Delta Flux Input to overtemperature AT Setpoint The overtemperature AT setpoint is reduced when the excore detectors sense a percent power mismatch between the top End bottom of the core.

The dead band is +5% and -17% before the setpoints are reduced. For each percent (more than 5%) the top detector output exceeds the bottom detector, the setpoints are reduced an equivalent of 2% of.the rated power. For each percent (more than -17%) the bottom detector exceeds the ,

top detector, the setpoints are reduced an equivalent of 2% of rated power. t l

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33 4

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m TABLE 15-1 OVERPOWER AT CONSTANTS AT, = Indicated AT at rated power, 'F T = Average temperature, 'F T1 = 573.9'F

?,v K4 <1.089 of rated power Ks = 0.0262 for increasing T

= 0.0 for decrer. sing T Ks = 0.00123 for T > T

= 0.0 for T < T3 ts = 10 seconds f(AI) as defined in Section 15.2 T3 = 2 seconds for Rosemount or

- . equivalent RTD

= 0 seconds for Sost; nan or equivalent RTD T4=- 2 seconds for Rosemount or equivalent RTD-

= 0 seconds for Sostman or equivalent RTD '

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l TABLE 15-2 OVERTEMPERATURE AT CONSTANTS AT, = Indicated AT at rated power, 'F i-T = Average temperature, 'F T2 = 573.9'F P = Pressurizer pressure, psig P1 = 2235 psig Kt = 11.30 K2 = 0.0200 K3 = 0.000791 it = 25 seconds 12= 3 seconds i

t~ ta = 2 seconds.for Rosemount or l

. equivalent RTD j

'- = 0 seconds for Sostman or equivalent RTD 1 14= 2 seconds for Rosemount or j equivalent RTD ~

= 0 seconds for Sostman or equivalent RTD l

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16.0 FUEL PERFORMANCE  !

UT examination of cycle 16 reload fuel identified no leaking fuel rods. Further evidence of no leakers is shown in Figure 16-1 showing relatively low coolant activity before and after refueling. '

17.0 CONCLUSION

The following results of cycle startup testing should be highlighted. '

1. The bank swap method for measuring rod worth produced acceptable <

results, including a normal differential shape. This shows that the new core design features had little affect on this measurement.

2. The hafnium poisons did not cause operational difficulties with the  !

source, intermediate, or power range encore detectors.

3. Cores with higher enrichments may require more time to escalate to full power during the initial cycle startup. Allowing menon poison buildup helps to lower localized power peaks. *

4. The new RPI head cabling did not affect the coil stack output as seen by the process computer in specific, or the RPI system in general.

The remaining Unit 2, Cycle 16 startup test results were normal. ,

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