ML20245J109
| ML20245J109 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/25/1989 |
| From: | Heyse R, Kurtz P, Zyduck R WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-89-095, CON-NRC-89-95 VPNPD-89-430, NUDOCS 8908170505 | |
| Download: ML20245J109 (42) | |
Text
- _ _ _ _ _ _ _ __-
1 * .
.-e Wisconsin Elecinc PONER COMPANY 231 W. Michgan. PO Box 2046. Mdwoukee WI 53201 (414) 221-2345 VPNPD-89-430 NRC-89-095 August 8, 1989 Document Control Desk U.S. NUCLEAR REGULATORY COMMISSION Mail Station Pl-137 Washington, D.C. 20555 i
Gentlemen:
DOCKET NO. 50-266 l CYCLE 17 STARTUP REPORT POINT BEACH NUCLEAR PLANT UNIT 1 l In accordance with Point Beach Technical Specifications 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 1 with a low-low leakage core design using enhanced Optimized Fuel Assemblies.
Please contact us if you have any questions concerning this submittal.
l Very truly yours, b '
/
(lf' V -
C. W. Fay Vice President Nuclear Power Enclosures Copies to NRC Regional Administrator, Region III NRC Resident Inspector 8908170505 890725 6 fi PDR ADOCK 0500 P
A subunm afhwswEncarGwporathw
l i
I I
)
ERRATA WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT UNIT 1 CYCLE 17 STARTUP l MAY 1989 1 l
l i
Page 2, Section 1.2, Paragraph 2, ;
6th line should be corrected to read:
" Cycle 17 contains 116 optimized fuel assemblies..."
l
r- . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - - . _ _ _ _ . _ _
i .
JULY 25, 1989 r
I l
WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT-t UNIT 1 CYCLE 17 STARTUP MAY, 1989
?
1 i
BY i R. C. ZYDUCK P. N. KURTZ R. P. HEYSE l l
?
_ _ _ _ _ _ . _ . . _ _ _ _ _ . . - - _ - _ _ _ _ _ . _ _ _ _ _ .m
l' . .
TABLE OF C0NTENTS Page LIST OF TABLES iii LIST OF FIGURES iv PREFACE V SECTION 1.0, REFUELING 1
1.1 Fuel Movement and Inspection 1 1.2 Core Design 2 SECTION 2.0, CONTROL ROD OPERATIONAL TESTING 6 2.1 Hardware Changes / Incidents 6 2.2 Rod Drop Times .-
6 2.3 Control Rod Mechanish Timing 6 2.4 Rod Position Calibration 6 SECTION 3.0, THERMOCOUPLE AND RTD CALIBRATION 9 SECTION 4.0, PRESSURIZER TESTS 11 4.1 Thermal Transients 11
' 4.2 Heater Capacity 11 4.3 Auxiliary Spray Testing 11
, SECTION 5.0, CONTROL SYSTEMS 12 SECTION 6.0, TRANSIENTS 12 SECTION 7.0, INITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKS 12 7.1 Initial Criticality 12 7.2 Reactivity Computer Checks 13 SECTION 8.0, CONTROL ROD WORTH MEASUREMENT 14 8.1 Test Description 14 8.2 Bank Swap Methodology 17 i
Page SECTION 9.0, TEMPERATURE COEFFICIENT MEASUREMENTS 21 SECTION 10.0, BORON WORTH AND ENDPOINT MEASUREMENTS 22 SECTION 11.0, POWER DISTRIBUTION 24 SECTION 12.0, XENON REACTIVITY 26 SECTION 13.0, SHUTDOWN MARGIN CONSIDERATIONS 26 SECTION 14.0, EXCORE DETECTOR BEHAVIOR 27 SECTION 15.0, OVERPOWER, OVERTEMPERATURE AND DELTA FLUX SETPOINTS CALCULATION 30 15.1 Overpower and Overtemperature AT Setpoints Calculation 30 15.2 Delta Flux Setpoints Calculation 30 SECTION 16.0, FUEL PERFORMANCE 33 SECTION 37.0, CONCLUSION 35 e
ii
4 4 LIST 0F TABLES Table- Page:
1-1 Fuel Movement Summary l' l 1-2 Uranium Loading 3 3-1 RTD Calibration Check 10 4-1 l' eater Group Power Supply Readings 11 7-1 Reactivity _ Computer Checkout 13 8-1 Rod Worths by Bank Swaps 15 8-2 Rod Worths by Steppihg 15 8-3 Rod Worths by Boron Endpoints 16 8-4 Boron Endpoints 16 8-5 Comparison of Rod Worths' 16 8-6 Critical Bank Swap Configuration Data 18 9-1 Isothermal Temperature Coefficient 21 10-1 Boron Worth and Endpoints 22 11-1 Initial Power Escalation, Flux Map Results 24
^
13-1 Excess Shutdown Worth Available for a Full Power Trip 26 l
l 14-1 100% Currents (p amps) 28 1 14-2 BOL Calibration Currents (100%) 28 15-1 Overtemperature AT Constants 31 15-2 Overpower AT Constants 32 16-1 Typical Isotopic Composition of Primary coolant Activity 33
)
, iii q
1 LIST 0F FIGURES !
Figure Page #
1-1 Core Loading 4 I 1-2 BOL Burnup Data 5-2-1 Cold Rod Drop Times (Full-Flow) 7 2-2 Hot Rod Drop Times (Full-Flow) 8 l 8-1 BOL HZP Reference Bank Differential Worth 19 8-2 BOL' Control C Bank Differential Worth 20 10-11 Boron Concentrations Ddring BOL HZP Physics Tests 23 11-1 Axial' Power Shape at BOL HFP ARO 25 14-1 NI35 and NI56 Response to Power Level 29 l
16-1 Unit One Iodine Levels 34 9
1 J
1 I
$V ,
{
i- -
._._x____ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . ._ _ !
- Tl 1
l l
PREFACE This report is intended to document in a concise format the results of I the physics testing program and unit systems response during the startup of !
Unit 1 following Refueling 16. The organization of the report follows that l
utilized in previous startup reports. 1 Westinghouse performed the core design calculations for Unit 1 Cycle 17. The reactivity coefficients were calculated based on estimated Cycle 16 end of . life burnup of 10,700 MWD /MTU. Cycle 16 was . ended on April 9, 1989, with . a peak assembly burnup of 43,784 MWD /MTU and average assembly burnup of 29,483 MWD /MTU. Actual Cycle 16 burnup was
- 10,735 MWD /MTU. Electrical power was first generated during Cycle 16 on May 20, 1988.
This report . is intended primarily for the use of ' Wisconsin Electric Power Company personnel as a readily accessible, complete compilation of j reduced data.
I 4
j h
V
. i; Section 1.0 REFUELING Section 1.1 Fuel Movement and Inspection A core shuffle was' performed during Unit 1 Refueling 16 because no' work on the core barrel,, lower internals or vessel was required.
The shuffle began at 1930 on April 15,.1989. The shuffle was completed at 0902 on April 22, 1989. 'The'following is a table of the number of steps completed each shift, difficulties encountered during the shift, .and hours -
time' lost due to the' difficulty.
TABLE 1-1
,~
FUEL MOVEMENT
SUMMARY
Date Shift' Steps Difficulties Time-Lost 04-15-89 2 19 Fuel Movement Started 1930 0 04-16-89 1 36 0 2 36 0 04-17-89 1 25 0 2 37 0 04-18-88 1 34~ 0 2 33 0 04-19-89 1 33 0 ,
2 19- SFP Upender Brake Failed 3 04-20-89 !
1 39 0 i 2 28 0 04-21-89 1 6 Shuffle Completed 0900 0' 345 3 l 109.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> = 6,570 minutes 6,570 minutes /345 steps = 19.0 min / step (U2R13 = 23.0, U1R15 = 32.9) .
1 Subtract time lost yields 18.5 min / step (U2R13 = 18.1, U1R15 = 20.2)
.]
l 48 1 i 1 .
1
~
This fuel movement was one of the smoothest in years, with only three f hours of down time from equipment problems.
Ultrasonic _ (UT) ' inspection of reused assemblies was conducted in the containment during the shuffle. This added _ about five minutes to the move
}
of an assembly going from the core to the rod basket or 'upender. About ten minutes were added to the move of an _ assembly being repositioned. The UT exams added about twelve hours to the shuffle. No leaking fuel rods were detected.
One _ assembly, T14, had a grid partially torn during the fuel movement.
Assembly T14 was visually inspected at the periscope. The. grid tear area was photographed and the _ photographs were forwarded to Westinghouse. The i
use of T14 was reviewed and approved by Westinghouse and the Manager's Supervisory Staff prior to its loading.
.~ q The final core loading pattern was verified by scanning the top of the I core with a TV camera to read the' fuel and insert numbers, and noting the l location the camera was suspended over.
Section 1.2 Core Design The core was loaded per the information contained in WCAP-12194, "The N'iclear Design and Core Management of the Point Beach Unit 1 Nuclear Reactor, Cycle 17." New fuel consisted of 16 assemblies with a nominal ;
enrichment of 4.0% and 12 assemblies with a nominal enrichment of 3.%. j Secondary sources are located at H3 and Fil for Cycle 17. Twenty new BPRAs were provided for Cycle 17 to control radial power distribution. Twelve 4P j BPRAs and eight 12P BPRAs were loaded in new fuel assemblies (Region 19).
Four 8D water displacers were also loaded in once burned fuel assemblies (Region 18). All control rods from Cycle 16 were used in Cycle 17.
Cycle 16 contains 116 optimized fuel assemblies and 5 standard fuel ,
assemblies.
Twelve hafnium poison assemblies were provided for Cycle 17. Each assembly ;
consists of 16 rodlets with the halfnium portion extending 6 feet down from '
the center plane of the core. These assemblies are handled with an RCCA tool. The control rod type hubs were shortened to not interfere with upper internals mixing vanes. These poisons were placed along the core'" flats."
Their purpose is to reduce the vessel weld ' neutron fluence. These poisons are expected to be used for about 15 years and then will be discharged.
Table 1-1 lists the uranium weight by region. Figure 1-1 shows the final core load pattern and Figure 1-2 BOL burnup data.
2
.e..
.p.
TABLE 1-2 URANIUM LOADING Region- 8 12B- 16 17A 17B 18A 18B' 19A- 19B
.~
Number of Assemblies 4 p 28 16 16 12 16 ~ 12 - 16' Original U Weight (MTU)': 1.61 0.40 10.04 5.74 5.74 -4.28 5.71
~
4.28 5.71.
Total U Weight 43.51 MTU
)
i 1
l
- i i
., i l
l 1 I
3 !
'l
. 1
- _ _ _ _ - _ - = _ - _ - _ _ - _ _ _ _ _ _
_ _ _ _ _ - _ = _ _ _ _ __ _ _ _ . . _ _ - _ _ _ .-_. _
-[
-e 4 1 FIGURE 1-1 CORE LOADING' PBNP UNIT 1 -START OF CYCLE:17; ,
i 1 2 3: 4- -5 6 ~7 '8 ~ 10-9 11' 12 13 ;
R20 R05 R08 A
' LM0JDT LM0JE3 LM0JDS 116 116- 116 l
R16 'S31 U15 UO2- U18 S18 '1 B R12
.LM0JDX LMOKXG LM0Q9D LM0Q90 LM0Q9G LMOKXD LM0JDK.
116 117B 119B' 119A 119B 117B 116 S19 U20 T28 SOS S17 503 C T16 -U26 S20 LMOKXC 117B LM0Q9J 119B LMONA2 LM0KXS LMOKXE LMOKX8 LMON9Q LM0Q9Q LMOKXB 1188 117A 117B 117A 118B 119B 117B R11 U27 H70 T21 004 )
' S11 U0S- .T22 H76 U14' R18-D )
LM0JDL LM0Q9R LM06DT LMON9V LMdQ92 LMOKXZ LM0Q93 LMON9W LM06DZ LM0Q9C LM0JDF, 116 119B' 208 118B 4 119A '117A 119A 118B 208 119B 116 S24 T15 - T26 Sol R04 - T02 RIS S16 T20 T18 - S301 E
LM0KXP 1178 LMON9P 11BB LHONA0 LMOKY2 LM0JE4 LMON9A LM0JDZ LMOKXW LMON9U LMON98 LMOKXH 118B 117A 116 116A 116 '117A 118B 118B 117B R21 U2B SO9 U10 R25 T10 T04 TOS R07 - Ull S10 U16 R10 F
LM0JDH 116 LM0Q95 119B 117A LMOKY1 119A LM0Q98
'116 -118ALM0JDM118A LMON9J 118A LMON9C LMON9D LM0JDY LM0Q99 L 116 119A. 117A' -119B 116
(
, R28 UCI S28 S14 T03 T01 M16 T12 Til SOS S32- U12 G R27 LM0JDN 116 LM0QBZ 119A 117B LM0KXK 117A LMOKX7 118A LHON9B 118A 112B LMON99 LMOBSK LMON9L LMON9K-LM0KXV LMO 118A 118A 117A 117B 119A 116-
' (
R26 U23 S15 UO3 R13 T09 T06 T08. R17 UOB SO2- U24 R23 H
LM0JDR 116 LM0Q9M 119B 117A LMOKY4 119A LM0Q91 116 118A LM0JE1 118A LMON9H M ON9E LMON9G LM0JDE LM0Q96 LM 118A- 116 119A' 117A 119B 116 S21 T13 T19 S12 R22 T07 I
R03 S06 T17 T25 S23 LMOKXA 117B LMON9M 118B 118BLMON9T 117A LMOKXY 116 LM0JDU LMON9F LM0JE5 LMOKXU LMON9R LMON9 118A 116 117A 118B 118B 117B R02 U21 H55 T27 UO9 SO4 UO6 T24 H69 U22 ROI J
LM0JE6 116 LM0Q9K 119B~ 208LM06DC 118B LMONA1 LM0Q97 LMOKX9 LM0Q94 LMON9N LM06DS LMOQ9L LM 119A 117A .119A 118B 208 119B 116 S27 U17 T23 S13 S22 S0?
K T24 U19 S29 i LMOKXL 117B LM0Q9F LMON9X LMOKXX LMOKXR LMOKXT LMON9Y LM0Q9H LMOKXJ ~
119B }
118B 117A 117B 117A - 118B 119B 117B R06 S25 U13 UO7 U25 S26 L R24 l
LM0JE2 LMOKXN LM009B LM0Q95 LM0Q9P LMOKXM LMCJDJ- j 116 117B - 119B 119A 119B 117Bl 116 i l
R14 R19 R09 - '
M i LM0JE0 LM0JDG LM0JDW j 116 116 116
. . . ASSEMBLY ID # ,
ASSEMBLY ANSI #
ASSEMBLY FUEL REGION 4
-l l
-____.___.________._____.1._______________ _ _ . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ ~ . _ . _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ .._______._i.)
e a FIGURE 1-2 BOL BURNUP DATA PBNP UNIT 1 START OF CYCLE 17 1 2 3 4 5 6 7 8 9 10 11 12 13 R20 R05 R08 A 37117 38065 37478 116 116 116 R16 S31 U15 002 U18 S18 R12 B 37942 26274 0 0 0 26440 37701 116 117B 119B 119A 119B 117B 116 S19 U20 T28 S08 S17 SO3 T16 U26 S20 C 24854 0 11351 25578 24734 26392 11993 0 25191 117B 119B 118B 117A 117B 117A 118B 119B 117B R11 U27 H70 T21 UO4 S11 UO5 T22 H76 U14 R18 D 36953 0 34378 12488, ' O 26328 0 12774 34472 0 37879 116 119B 208 118B* 119A 117A 119A 118B 208 119B 116 S24 T15 T26 S01 R04 T02 R15 S16 T20 TIB S30 E
26259 11801 12423 25397 36411 13652 36246 25575 12314 11719 26427 117B 118B IISB 117A 116 118A 116 117A 118B 118B 117B l
R21 U28 SO9 U10 R25 T10 TO4 TOS R07 Ull S10 U16 RIO F 36594 C 25694 0 36202 13324 14023 13459 36020 0 26044 0 37044 l 116 119B 117A 119A 116 118A 118A 118A 116 119A 117A 119B 116 R28 U01 S28 S14 703 T01 M16 T12 Til SOS 532 U12 R27 G 37230 0 24697 26558 13561 13880 27839 13661 13299 26722 25187 0 38384 116 119A 117B 117A 118A 118A 112B 118A 118A 117A 117B 119A 116 R26 U23 SIS UO3 R13 709 T06 708 R17 UOB SO2 U24 R23
! H 3718B 0 26163 0 36854 13220 13462 13248 36346 0 26359 0 37147 116 119B 117A 119A 116 118A 118A 118A 116 119A 117A 119B 116 S21 T13 T19 S12 R22 707 R03 S06 T17 T25 S23 l
26210 117B 11657 118B 12394 25425 36109 13195 36221 25672 12520 11421 26163 118B 117A 116 118A 116 117A 118B 11BB 117B R02 U21 H55 T27 UO9 SO4 UO6 T14 H69 U22 ROI J 37574 0 34224 12217 0 26553 0 12586 34535 0 37830 116 1198 208 1188 119A 117A 119A 11BB 208 119B 116 S27 U17 T23 S13 S22 S07 T24 U19 S29 K 24487 0 11589 259,92 24706 25757 11958 0 25225 117B 119B 118B 117A 117B 117A 118B 119B 117B R06 525 U13 UO7 U25 S26 R24 L 37492 26141 0 0 0 26083 37255 116 117B 119B 119A 119B 117B 116 R14 R19 R09 M
37016 38204 37314 116 116 116
. . ASSEMBLY ID #
BURNUP (MWD /MT)
ASSEMBLY FUEL REGION
~
5
. o.
Section 2.0 CONTROL ROD OPERATIONAL TESTING Section 2.1 Hardware Changes / Incidents All control rods 'from Cycle 16 were used in Cycle 17. -New RPI and Rod Control System cabling was installed on the reactor vessel head during Refueling 16. Detailed RPI readings of volts versus steps were made on all Bank D ' rods to verify that the . new connections did not change the RPI calibration curves. The curves are similar to those taken in 1971 and 1988. No difficulties were encountered with the cabling or with the rod
drive mechanisms.
Section 2.2 Rod Drop Times Rod drop times . to dashpot in the cold full-flow condition ranged from 1.40 seconds to 1.56 seconds. Data scatter was normal.
Rod drop times to dashpot in the hot full-flow condition ranged from 1.26 seconds to 1.38 seconds. Data scatter was normal.
See Figures 2-1 and 2-2 for rod drop times and RCS parameters.
' All rod drop times were well within the Technical Specification limit of 2.2 seconds (Technical Specification 15.3.10.E).
i Section 2.3 Control Rod Mechanism Timing l Traces of control rod gripper coils currents were obtained on May 5, 1989. All traces of the lift, moveable and stationary coil currents were ;
reviewed and determined to be satisfactory. i Section 2.4 Rod Position Calibration l During hot rod testing, LVDT voltages were read at 20 steps and 200 <
i j'
steps on all banks except Bank D to determine if any voltages were !
abnormal. More detailed readings were taken on Bank D using procedure RESP 1.6.
"Zero" adjustments were made with rods at 20 steps under hot zero power i full flow conditions.
" Span" adjustments were made at full power after rods were verified to be fully withdrawn using procedure RESP 1.2.'
O
1 FIGURE 2-1
- I COLD ROD DROP T2MES (FULL FLOW) !
I
- I
. 1 2 3 4 5 6 7 8 9 10 11 12 13 I m-A . ;
g 1.40 1.43 j 2.00 2.06 1 1.40 1.54 1.56 2.09 2.20
\ 2.16
/
l i
1.40 1.46 #
D 2.00 2.05 1.47 1.48 1.43 1.48 E
2.15 ,2J15 2.15 2.08 1.43 1.50 1.48 1.48 F
2.06 2.13 2.13 2.05 g
1.43 1.40 2.07 1.41 ,, !
1.93 2.02 1.47 1.46 1.53 i
H 2,07 1.45 1
, 2.13 2.16 2.08 1.48 1.42 1.45 1.44 2.08 2.05 2,06 2.08 !
1.45 1.40 J
2.00 1.98 !
1.53 1.50 1.40 2.20 2.17 2.06
~
1.52 1.41 ;
2.10 2.03 M-UNIT 1 so*
- DATE 05-09-89 ;
TIME E DASHPOT (SEC) i TIME 'IO BOT 10M (SEC) TEMP. 193 *F FLOW 100 g PRESSURE 295 psia POINT BEACH NUCLEAR PIANT CONTROL ROD TESTING ROD DROP TIMES RE-D6 (10-78) 7
- - _ _ _ _ _ - ___-__m
_, , FIGURE 2-2 HOT ROD DROP TIMES (FULL FLOW) !
. 1 2 3 4 5 6 7 8 9 10 11 12 13 2ro
- A 1.28 1.33 !
O 1.76 1.81 1.29 1.33 1.38 C
1.82 1.87 1.88 1.27 1.30 #
1.74 1.77 1.30 1.29 1.32 1.31 1.88 1.81 J.47 1.84 1.28 1.31 1.30 1.32 F
1.80 1.85 1.84 1.80 1.29 1.32 1.29 1
1.79 1.75 1.82 1.32 1.31 1.33 H 1.30 1.80 1.85 1.85 '
1.82 1.31 1.29 1.32 1.30 j
, 1.81 1.82 1.81 1.83 1.29 1.26 J l 1.73 1.73 j
1.35 1.33 1.26 K
1.88 1.85 1.81 1.35 1.28 1.83 1.80 M
UNIT ~
1 30*
r TIME 70 DASHPOT (SEC)
TIME 10 BOT 10M (SEC) TEMP. 525 'F !
FLOW 100 g PRESSURE 1998 psia POINT BEACH NUCLEAR PLANT CONTROL ROD TESTING ROD DROP TIMES RE-D6 (10-78) 8
4 Section 3.0 THERMOCOUPLE AND RTD CALIBRATION During initial cycle heatup, thermocouple and loop RTD signals were recorded at different temperature levels under full-flow conditions. See Table 3-1 for the results for full flow conditions. The RTD resistance readings were obtained at the protection racks in the control room using a digital multimeter that subtracted lead resistance. Thermocouple maps were obtained on the PPCS.
Since the core was producing very little heat, the hot and cold leg RTDs were at about the same temperature. Thus both hot leg and cold leg readings were averaged into one temperature for the RTDs. The RTD resistances were converted to degrees Fahrenheit by using the vendor's calibration curves.
All RTDs appeared to be # functioning properly. Due to RCS leakage, thermocouple H-6, I-7, and L-10 have been physically disconnected at the RV head.
l 8
~
9
i
-4 4
]
f TABLE 3-1 RTD CALIBRATION CHECK
{
RTD Element RTD Temperatures from Measured Resistances ('F) '
LOOP A - COLD LEG R 401B 312.46 350.66 410.27 442.42 492.41 523.39 528.35 R 405B 312.51 350.70 420.28 442.37 492.37 523.42 528.24 W 402B' 312.79
)'
350.93 410.79 .442.83 492.67' 523.71 528.53 W 406B '312.75 350.84 410.68 442.79 492.50 523.52 528.29-LOOP A - HOT LEG
{
R 401A 312.69 350.81 410.57 441.96 492.49 523.51 528.49 1 R 405A 312.75 350i92 410.71 443.06- 492.57~ 523.61 528.55 !
W 402A 312.75 350.84 410.63 442.78 492.47 523.39 528.29 '
W 406A 312.96 351.06 410.88 442.49 492.69 523.64 528.51 q
LOOP B - COLD LEG l B 403B 312.76 350.89 410.77 442.94 -492.65 -523.58 '528.51 i B 407B 312.93 351.04 410.87. 442.50' 492.73 523.68 528.57 Y 404B 313.08 351.16 411.07 443.49 492.81 523.87 528.52 -;
Y 408B 313.18 351.2.7 411.20 442.43 492.94 523.88 528.63 '
1 LOOP B - HOT LEG B 403A 313.12 351.23 411.06 443.18 492.83 523.73 528.50 B 407A 313.23 351.34 411.18 442.36 492.91 523.87 528.59 Y 404A l 313.40 351.47 411.37 442.65 492.99 523.91 528.55 1 Y 408A 313.44 351.50 411.42 443.44 493.05 523.94 528.57 j AVERAGE 312.93 351.04 410.86 442.73 492.69 523.67 528.48 SAT. TEMP. 312.00 348.00 411.00 444.00 492.00 524.00 528.00 i TEMP. 315.00 353.00 413.00 444.00 493.00 524.00 529.00 l
l l
i l
10
Section 4.0 PRESSURIZER TESTS Section 4.1 Thermal Transients Pressurizer pressure increase rate with spray valves apparently shut and all heaters on was 14 psi / min which is normal. During the thermal equilibrium test, heater Bank D was on about 50% of the time to maintain pressure. Spray valve effectiveness was normal with pressure decreases greater than 110 psi / min.
Spray bypass valve positions were not changed as the result of the pressurizer tests.
Section 4.2 Heater Capacity a
Pressurizer heater capacity was determined from pewer panel readings for each group of heaters. Table 4-1 shows that heater capacity is above Technical Specification requirements of 100 KW minimum total.
TABLE 4-1 HEATER GROUP POWER SUPPLY READINGS Heater I-Current V-Voltage KW-Energy Input Group (amps) (volts)
. KW = 8 x V x I/1000 A 293 486 247 B 251 486 211 C 226 486 191 D 229 486 193 E 246 486 207 TOTAL 1048 KW Section 4.3 Auxiliary Spray Testing l The auxiliary spray line was used to verify the operability of newly ]
l Installed temperature instrumentation on the line. The instrumentation '
functioned properly.
1 l
l 1
11
r .
Section 5.0 CONTROL SYSTEMS There were no difficulties encountered during heatup or testing in the control systems of pressurizer level or pressurizer pressure control system.
Section 6.0 TRANSIENTS There were no significant transients during the startup or approach to full power. There were no violations of the fuel conditioning restrictions on power and rod stepping change rates.
Section 7.0 INITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKS Section 7.1 InitialCrikicality The approach to criticality was made in two phases. The first step, which began at 1842 hours0.0213 days <br />0.512 hours <br />0.00305 weeks <br />7.00881e-4 months <br /> on May 14,1989 was the normal withdrawal of control rods until Bank D reached 180 steps. Then the reactor coolant boron concentration was decreased by dilution until criticality was achieved.
ICRR plots were maintained during each phase of the approach to
, criticality.
The reactor conditions at the time of criticality were determined to
, be as follows:
Date May 15, 1989 Time 0030 RCS Temperature 529 F RCS Pressure 1985 psig Rod Position Bank D at 188 steps Boron Concentration ~1180 ppm 12
?
r- ,
l i
Section 7.2 Reactivity Computer Checks Following criticality, . acceptable zero power physics testing flux levels were determined. The flux level at which nuclear heat appeared was ~
at 3 x 10~6 amps on N-35, 3 x 10~6 amps on N-36 and 3 x 10 ~6 amps on the Keithley picoammeter. Normal flux levels - for physics testing are about one-third of these values.
A check of the reactivity computer was made by comparing the' computer's calculated reactivity for a certain doubling time versus design reactivity for the same doubling time. Reactor coolant system temperature was near 530 F. Table 7-1 shows the results of this check.
- ~
.: TABLE 7-1 REACTIVITY COMPUTER CHECKOUT Bank D Steps Measured Measured Calculated Doubling Reactivity Reactivity From To Time (Sec.) (pcm) (pcm) 188 196 105.0 38 38 188 201 58.0 59 61 188 207 38.7 81 82 l
1
'I J
J I
l 13 ;
Section 8.0 CONTROL ROD WORTH MEASUREMENT Section 8.1 Test Description Since 1978 the bank swap method has been used at Point Beach for j measuring control. rod worth. See Section 8.2 for methodology.
Normally the rod worth verification utilizing bank swap is divided I into two parts. In the first part, the reactivity of the reference bank is obtained from reactivity computer measurements and boron endpoint data during an RC3 boron dilution. In the second part, the critical height of the reference bank is measured after exchange with each rems.ining bank.
The reference bank reactivity compared to the critical height infers the worth of the bank which is fully inserted.
The NRC has established aIceptance and review criteria for the results 1 of bank swaps in two loop Westinghouse PWRs. The acceptance criteria l requires the total measured bank worth to be at least 90% of designed )
worth. The review criteria requires each measured bank vorth to be within j 16% or 100 pcm of the designed value, whichever is greater.
l 1
During Cycle 17 startup testing of Point Beach Unit 1, the sum of the i bank worths as measured by bank swaps was 89.4% of the design value. In
' addition, the measured bank swap worth for four of the six banks of rods j differed by more than 15% from the design value. Additional testing was '
therefore required.
Initially it was determined that a measured boron worth value needed to be obtained. Boron worth was determined by taking the average of two measured values. The differential worth of Control Bank A, the normal reference bank, and also of Control Bank C, reactivity computer. See Figures 8-1 and B-2. was The measured with the sum of each bank's differential worth , i.e., integral worth, was then divided by the change in the critical boron concentration between the fully withdrawn and fully inserted conditions. The measured boron worth was approximately 10% less than the design worth (9.08 pcm/ ppm measured versus 10.15 pcm/ ppm by design).
Bank worths were then obtained by boron endpoint. The change in critical boron concentration from the fully withdrawn to fully inserted condition was multiplied by the measured boron worth to obtain a bank worth.
Using this method to obtain bank worth, the total bank worth measured value was 100.1% of the design value. Two of the six banks still did not meet the 15% review criteria and so the NRC was contacted and forwarded the results of the testing. Adequate shutdown margin had been demonstrated to exist using the boron endpoint method for determining bank worth. The Manager's Supervisory Staff approved the resolution of the discrepancies between the bank swap measured and design data on May 16, 1989.
14
o ,
j 1
1 l
- i l Point Beach Unit 1 Cycle 17 is a 12-month, five-batch, low-low leakage l core design. Westinghouse does the core design, and fuel and insert fabrication. Low-low leakage was ootained by increasing peaking factors (FDH = 1.70 and FQ = 2.50) and shifting power to the center of the core.
Cycle 17 is the first such cycle. Previously, Unit I was using a four and one-half batch normal low leakage pattern.
The following tables are a consolidation of the data from procedures RESP 4.2 and RESP 4.3. )
TABLE 8-1 !
ROD WORTHS BY BANK SWAPS (PCM)
Control Bank C 4789 - (1.070
- 1025) + 17 = 709 Shutdown Bank B 1789 - (1.063
- 1212) + 13 = 514 Shutdown Bank A 1789 - (0.0796
- 718) + 13 = 1230 Control Bank B 1789 - (1.119
- 1282) + 10 = 364 Control Bank D 1789 - (1.070
- 1158) + 10 = 695 REPEATED BANK SWAP MEASUREMENTS (PCM)
Control Bank C 1789 - (1.070
- 1020) + 11 = 709 Control Bank B 1789 - (1.119
Control Bank A 1789 1006 Control Bank C 844 1103 ARO ----
1199 Boron Worth = [(1789/191) + (844/96)]/2 = 9.08 pcm/ ppm l
i 15 l
__-_____-_-.____-..___-__w
l s .
TABLE 8-3 ROD WORTHS BY' BORON ENDPOINTS (PCM) ENDPOINT (PPM)
~~
Control Bank A (9.08
- 193) - 17 = 1717 1006 i
(
Control Bank C (9 08
- 96) - 3 = ~ 869 1103 Shutdown Bank B (9.08
- 70) - 3 = 633 3129 Shutdown' Bank A (9.08
- 124) - 0 = 1125 1075 Control Bank B (9.08
- 74) - 12 = 660 1125 Control Bank D (9.08 *'103) - 12 = 935 1096 TABLE 8-4 BORON ENDPOINTS MEASURED DESIGN Control Bank A 4 1006 ppm. 1024 ppm Control Bank C 1103 ppm 1102 ppm Shutdown Bank B 1129 ppm- 1127 ppm Shutdown Bank A 1075 ppm 1082 ppm control Bank B 1125 ppm 1143 ppm Control Bank D 1096 ppm 1119 ppm ARO 1197 ppm 1199 ppm Boron Worth Measured = 9.08 pcm/ ppm j
~
Boron Worth Designed = 10.15 pcm/ ppm I
TABLE 8-5 1 COMPARISON OF ROD WORTHS AND BORON ENDPOINTS Design (pcm) Bank Swap Endpoint Stepping Swap End pcm % diff pcm % diff pcm % diff' Control Bank A 1755 1755 N/A N/A 1717 97.8 1789 101.9-Shutdown Bank C 960 959 709 73.9 869 90.6 844 87.9 Shutdown Bank B 711 702 514 72.3 633 90.2 N/A N/A Control Bank A 1165 1164 1230 105.6 1125 96.6 N/A N/A Control Bank B 546 533 364 66.7 660 123.8 N/A N/A.
Control Bank D 794 786 695 87.5 935 119.0 N/A N/A j
~"
Total 5931 5899 5300 89.4 5939 100.7
% diff = 100 * [1 - (D-M/D)]of design swap method 16
= _ - _ - _ - _ _ _ - -
e .
~
Section.8.2 Bank Swap Methodology The integral reactivity worth of the measured bank is ' inferred from the swapped portion.of Control Bank A by the following equation:
Wf = - Ap2 - (a X) (AP2) + X where:
Wf=The.inferredworthofBankX,pcm
= The measured worth of the reference bank, control A, from fully withdrawn to fully inserted with no other bank in the core.
a X = A design. correction factor taking into account the fact that the -
presence of another 'hontrol rod bank is affecting the worth of the reference bank.
AP2 = The measured - worth of 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.
Apr = The measured worth of the reference bank from the fully inserted condition to the elevation at which the reactor was just critical prior to the worth measurement of Bank K. In this test apt is zero.
EX = The worth of Bank X from the initial position (before the start 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 bank swap measurement data are tabulated in Table 8-1.
Values for a, were obtained from t_he design predictions are also. listed in Table 8-1.
17
- . i I
l TABLE 8-6 CRITICAL BANK SWAP' CONFIGURATION DATA 05-15-89 Measured RCS CA Bank Bank .
Tavg Position . Position.
Measured Time (*F) (Steps) (Steps) fj CC 1518 '
529 1 217 CC 1538 +" 529 107.5 1 CC 1552 529 1 217 SB 1609' 529 1 217 SB 1625 529 91 1 SB 1639 529 1 218 SA 1648 529 1 219 SA 1701 529 137.5 1 SA 1718 529 l' 219 j CB 1752 529 l' 222 CB 1802 529 85.5 1 CB 1814 529 1 222 CD 1723 529 1 219 CD 1735 529 96 1 CD- 1745 529 1- 219-Boron concentration was 1007 ppm.
j a
l j
i
-j l
1
-l l
l l
j 18
4
- FIGURE B-1.
I PBNP UNIT i CYCLE 17 BOL HZP REFERENCE BANK DIFFERENTIAL WORTH 15
- 0 - MEASURED WITHIN 10% OF DESIGN
- X - MEASURED OUTSIDE 10% OF DESIGN 14 .-
13 -
X O X 12 -
XEED O XX XX OXX X XX XX XX X*
- X 11 -
X XX XX X X MK O X X X +" X X MK X XMK X X
^ 10 -
~
X X m xXxXxx XX* XX X X X X h.
p -
X X X* X to -
X X X XXX g XX X X Ex xx mx
~
XX X X I
e.
8 -
- d
. E -
X
_s 7 --
5 h
e 6 X
~
LL.
~
E 5 --
4
- -1 '
X 4 -
l x -
3 - j
- x
- l m
2 - x '
X i g- .
e
.- i i f f I f f f f f f I f f f l f f f f I f I f 0 20 40 60 80 100 120 140 160 180 200 220 STEPS WITHDRAWN 19 i.
I,
a .
,, FIGURE 8-2 PBNP UNIT 1 CYCLE 17 BOL HZP CONTROL C DIFFERENTIAL WORTH 7
_ 0 - MEASURED WITHIN 10% OF DESIGN !
X - MEASURED OUTSIDE 10% OF DESIGN
' 00 00 O 00 0 6 -
o o
~
O 00 O X 0 X 0 O
XX 00 s -
- ,- x ,
X X
E
~
X.
W -
R -
x 5 4 -
x
& u 1
. E -
x
? -
. d -
x E3 O
i x E -
2 - *
.I x
- i 4
1 0 !
j n, elilililililililililililililililililililili 3 0 20 40 60 80 100 120 140 $60 280 200 220 STEPS WITHDRAWN 20 l
l
Section 9.0 TEMPERATURE COEFFICIENT MEASUREMENTS-An ARO 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 teraperature were recorded on an X-Y plotter and two-pen recorder.
The ueasured temperature coefficient is within ' the review criteria of 13 pcm/'F.
4 TABLE 9-1 i J
ISOTHERMAL TEMPERATURE COEFFICIENT Control Boron Avg.
B: ink Conc. Temp. Measured Design
- Difference Configuration ppm *
-F pcm/ F pcm/'F pcm/'F (M-D)
ARO 1197 530 -3.8 -2.6 -1.2
- Corrected for actual conditions-21
.. . J j
Section.10.0 BORON WORTH AND ENDPOINT MEASUREMENTS Figure 10-1 shows RCS boron concentration during zero power physics testing. Table 10-1 shows results of the endpoint ' measurements'. All values are for N530'F (control Bank A _ worth ' testing temperature). Both !
design and measured boron worths were obtained by dividing bank worth (pcm) into change in boron concentration between the endpoints.
l Review criterion was not met (10.5 pcm/ ppm), see Section 8.0 for-resolution.
TABLE 10-1 a
BORON WORTH AND ENDPOINTS Endpoint Bank Worth Boron Worth Bank Design Measured Design . Measured Design Measured Configuration (ppm) (ppm) (pcm) ' Ipcm_L (pcm/ ppm) (pcm/ ppm)'
ARO 1199 1199 ---- ----
-10.1 N/A CA in 1024 2006 1755 1789 -10.3 -9.4 CC in 1102 1103 960 844 -10.2- -8.8.
l l
-l l
I l
22 l
I
. 'O N '
I '
N A
S
' o '
I
' o '
B S 'O O
Rg A
'o N
'o I
C
'o C 'O 7
1 X '
'o E X ' oO L S X C N ' oO Y O
- X '
C IP TZ S #
'o 1 T '
- i AH S
' oO 0 R E '
1 T TL T ' o I NO '
E N EB S ' oO R U C C '
U NG I ' oO G H ON S * '
C CI ' o R Y I '
F A NU H ' oO E OD P '
B R ' o '
T O I * '
N B G D ' oO N E '
I I D ' O O L N '
P P E M P ' o' A S '
S U S
' oO i
x 'O x '
o
- x '
x ' o O x '
R x ' O A x '
x ' ' o x ' oO x
X
' oO x I o
x '
x '
- I o
I O
' OO 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 6 5 d 3 2 1 1 1 1 1 1 0 9 1 1 IQS NoU o8(D uW
1 1
l Section 11.0 POWER DISTRIBUTION Table 11-1 illustrates the lowering of maximum L hot ' channel factors j during initial power increase to full load. Flux maps were ' taken using ANSI Standard ANS-19.6.1-1985 as guidance. The first flux map was taken at 28% power. Allowed power levels were calculated using the relationships for l FAH and FQ versus power ~ level in Technical Specification 15.3.10.B.1.a.
Measured HFP axial power distribution compared to design is shown in !
Figure 11-1. Burnup was approximately.85 MWD /MTU.
TABLE 11-1 INITIAL POWER ESCALATION PLUX MAP RESULTS Flux Map Power Thimbles Allowed Number Date (%) Missing Power (%) Bank D A0 FAHN FM 1 05-18-89 28 8 93.3 85.1 176 +14.9 2 05-18-89 48.6 0 89.7 96.4 188 +6.4
~
3 05-19-89 74.7 1 91.8- '99.9 198 +4.3 4 05-19-89 82.9 2 97.2 103.7 201 +2.2 5 05-19-89 92.9 0 95.2 -104.3 -206 +0.5 6 05-20-89 92.9 2 101.5 101.4 195 +0.8 a 7 05-20-89 94.6 l
1 101.3 101.7 195 +0.3 8 05-22-89 94.9 1 102.5 102.3 195 +0.6 l 9 05-22-89 94.9 1 101.7 101.5 201 +3.6 10 05-23-89 98.0 2 103.0 102.8 208 +3.6 11 05-24-89 97.5 5 103.4 108.5 199 -6.3 i 1
i' 12 05-24-89 97.5 5 112.6 103.9 199 -0.2 13 06-08-89 100.0 5 102.8 105.8 220 +2.2 ,
i l,
. i 24 l
- _ _ _ _ _ _ _ _ _ _ -___--___----_--J
1
. .j
- FIGURE 11-1 ,
' POINT BEACH UNIT.1 CYCLE 17 AXIAL POWER SHAPE AT BOL IfP, ARD, EGXE 1.3 I i
I-~s%
1.2 -
/ j
~
'- \/ \ \
\/ \
\
1.1
,' \/I
\
/
/ \
l 1.0 -
l l
1 I
tr - )
,g _ / - '
I
\
/ g 1
.e -
I \ !
/
/ \
m -
s I \
g .7 -
j
- w . / g n I i ti .s -I i d
e _i I n ;
I 1.
i
.5 -l s i
_l
.4 i
.3 - DESIGN <
MEAS. '
i
.2 -
.1 - l i i i l i 1 i i i l i I I i l i I ,
o t 0 10 20 30 40 50 60 70 80 90 200 CORE HEIGHT (PERCENT) l 25 i
.C O
~
Section 12.0 XENON REACTIVITY Xenon reactivity behavior data used for Unit 1 Cycle 17 was supplied by.
Westinghouse in the PCNDR - and - on separate xenon tables. ROD 10 data was obtained from the PCNDR. The PBNP xenon tables used the Westinghouse xenon table data and were run with a TDF1 of 0.95 and; TDF2 of 1.2. Tables are supplied for BOL, MOL, and EOL conditions.
Section 13.0 S'HUTDOWN MARGIN CONSIDERATIONS Bank worth measurement using the boron endpoint method were accepted as valid proof of rod worth for shutdown margin detennination. See Section 8.0 for details. Thus WCAP Table 6.2 was accepted as a valid shutdown margin determination. Table 13-1 calculates the excess worth available to Unit 1 Cycle 17.
TABLE 13-1 EXCESS SHUTDOWN WORTH AVAILABLE
, FOR A FULL POWER TRIP BOL (pcm) .EOL (pcm)
Shutdown Margin From PCNDR -3790 -3640
- Required Shutdown -1000 -2770
= Excess Worth -2790 - 870 26
r., -
]
1 I
~
Section 14.0 EXCORE DETECTOR BEHAVIOR l The upper and lower excore detector currents for each power range channel were recorded and calorimetric were performed at various power ';
levels. The upper and lower detector currents were summed for each channel '
and then normalized to obtain predicted currents for 100% power. These 100% currents are listed in Table 14-1. The effect of the low-low leakage loading pattern is seen in the lower sums for Cycle 17.
In anticipation of a reduction in output of the intermediate range-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 10 10 amps on the IR channels. Because actual IR attenuation was less than expected, the original SR trip setpoint would have given adequate but less margin for blocking the t' rip.
Intermediate range detector currents versus power level are shown in Figure 14-1. Intermediate range det,ector trip signals activated at 'about 1.3 x 10 4 amps for N35 and 1.6 x 10 4 amps for N36. Excore detector power level at the time the trip signals occurred was 21% for N35 and 21% for N36.
The hafni'am poisons reduced intern,ediate range detector output by about 20%.
Power range quadrant tilt alarms were meant for rapidly developing tilts. Natural core tilts are washed out to prevent a bias in alarming of rapidly occurring tilts. This is accomplished by obtaining calibration currents for the core with ' a tilt. A tilt is indicated only when actual currents deviate from the calibration currents even thougn the core already may have a tilt before the start of the deviation. This practice complies-with Technical Specifications and the Westinghouse position on core tilt.
Table 14-2 lists the " tilt free" power range detector. calibration currents corresponding to 100% power at BOL. These currents were calculated using the multi-map method at 98% power. A multi-map calibration was performed to verify that the hafnium poisons did not significantly change the linear response of the excore detectors. No significant affects were observed.
5 27
f to' g i
TABLE 14-1 100% CURRENTS (p AMPS) 41 42 43- '44 Cycle 17 472 572 578 589 Cycle 16 487 621 605 642 Cycle 15 506* 606- 620 633 Cycle _14 503. 634' 631 658 Cycle 13 492 637 668 671 Cycle 12 48R 585 638' 645 Cycle 11 -428 517 564 569-
- New Detector TABLE 14-2
~
~
BOL CALIBRATION CURRENTS (100%) ,
41 42 43 44 Cycle 15 T 268 308 325 331 B 241 304 297 303 Cycle 16 T 254 311 314 332 B 233 310 291 309 Cycle 17 T 251 291 304 308 B 221 281 274 281 28
r
. ~ >
l FIGURE 14-1 ;
UNIT 1 CYCLE 17 BOL NI35 AND NI36 RESPONSE TO POWER LEVEL 7
a l
e ;
o j 6 -
g q 5 - '
O tis i
x m -
s m 4 -
o l~ .
e o O
o u
d 0 2 -
e NI35
[ N36 TRIP O NI36 s -
N35 TRIP
' I ' I ' I ' I ' I ' I ' I ' I ' '
0 I 0 10 20 30 40 50 60 70 80 90 100 DELTA T BASED POWER LEVEL (%)
29
t ' .
~
Section 15.0 OVERPOWER, OVERTEMPERATURE AND' DELTA FLUX SETPOINTS CALCULATION l Section 15.1 Overpower and Overtemperature AT Setpoints Calculation j i
~j Discussion of the setpoints and equations has been sufficiently j covered in previous reports. l The equations are:
Overpower AT ( y, ag )
TE 'I 1 1ATo [K4 -K(73+1 5
3 y.I+73 ' 4) T - . Ks M y ,T4 g ) - $ - f(AI)]
Overtemperature AT( p 3) '4 1+t S 1Afo (K i - K (T( g 3 )-T ){
2 I
g;) + K 3 (P-P1 ) - f(AI))
. 1 See Tables 15-1 and 15-2 for the constants associated with this cycle of operation.
Section 15.2 Delta Flux Setpoints Calculation-l The overpower and overtemperature AT setpoints are reduced when the j excore detectors sense a power mismatch between the top and 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. ]
i e
i l
)
30 i
.I
. s _. y
'. {
=
i TABLE 15-1 :-
OVERTEMPERATURE AT CONSTANTS ATo~= Indicated AT at rated power, *F T = Average temperature, *F -l T1 = 573.9'F P = Pressurizer pressure, psig P1 = 2235 psig K2 <1.30 ,-
K2 = 0.0200 .
K3 = 0.000791 Il = 25 seconds I2= 3 seconds T3= 2 seconds for Rosemount or equivalent RTD
=
0 seconds for Sostman or equivalent RTD T4= 2 seconds for Rosemount or equivalent RTD .
= 0 seconds for Sostman or equivalent RTD l
e 31
. _ - _ - .-___--_- ~
c . .
n TABLE 15-2 OVERPOWER AT CONSTANTS ATo = Indicated AT at rated power, 'F T = Average temperature, *F T1 = 573.9 F K4 < 1.089 of rated power Ks = 0.0262 for increasing T
= 0.0 for decreasing T e
Ks = 0.00123 for T > T1
= 0.0 for T < T1 T3 = 10 seconds f(AI) as defined in Section 15.2 T3= 2 seconds for Rosemount or equivalent RTD
=
0 seconds for Sostman or equivalent RTD T4= 2 seconds for Rosemount or equivalent RTD
=
0 seconds for Sostman or equivalent RTD 32
y
.l e *=' s. .l 1
.e Section 16.0 FUEL PERFORMANCE Reactor . coolant activity is surnmarized in Table 16-1. Total activity-soon after startup was down significantly indicating good. fuel integrity.
See Figure 16-1 for iodine equivalents before and after the refueling. i TABLE 16-1 1
TYPICAL ISOTOPIC' COMPOSITION OF PRIMARY COOLANT ACTIVITY End'of Cycle 16 Start of Cycle 17 Isotope Half Life i>C/cc x 10~1 pC/cc x 10~1 I-131 8.05. days 0.058 0.026 i
I-132. 2.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 0.92 0.37-I-133 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> 0.70 0.28 I-134 53 minutes- 1.45 0.40 I-135 6.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 1.35 0.51 TOTAL 4.50 1.59 33
________.___m_- _ - - - - - - - "-
S N
I MD EE O* ' ,h DS S
o iJ c.
U .
J 5 !:! I. ;
1 g> l 6
iM
+-
S L
E V
E L
1
- E ;.'
.^ ^
6 N 1 I >f )$ $ ) b
- 0b>
l
't ihq ( E D
I E ' 4 < '
T R
U G
I F
O I
E M<(
eA A D
N 0' O C.
N T
I O N
U k --
iM O
' 0 0
^ eF 0
e.
)
Q -
9 8
0 o 0 9 0 o 1 1 1 0 s 0 1 0 0
. o. 0 0 0 o O 0 ACTIVIT
,. :n .s-P Seetion
17.0 CONCLUSION
The following results of cycle startup' testing should be highlighted.
- 1. The bank swap- method for measuring rod worth did not- produce acceptable results. However, this can be explained by the extent of '
core design changes made for Cycle 17.
- 2. The hafnium poisons did not cause operational .' difficulties with the.
source, intermediate, or power range excore detectors.
- 3. Cores with fuel enriched to '4%. or greater may require more .' time to escalate . from 75% to full power. Xenon poison. buildup is needed to-lower localized power peaks.
- 4. The new RPI head cabling did not affect the coil stack output as seen by the process computer 4n specific, - or the RPI system in general.
The remaining Unit 1 Cycle 17 startup test results were normal.
4 I
h J
35