ML20090A706
| ML20090A706 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/30/1991 |
| From: | Kurtz P, Wood R WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20090A701 | List: |
| References | |
| NUDOCS 9203030069 | |
| Download: ML20090A706 (39) | |
Text
' t. -
j lIf WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT UNIT 2 CYCLE 18 STARTUP November,1991 BY P. N. KURTZ R.
P. WOOD 9203030069 920225 PDR ADOCK 05000301 P
PDR J
b TABLE OF._CGUENTS l
Pace
-I iLi
^
LIST OF TABLEE.
LIST OF FIGURIS iv PREFACE v
1.0 BilVJJJjig 1
1.1 Summary 1
1.2 Core Design 3
2.0 CONTPOL ROD OPERATIONAL TESTING 6
2.1 Hardware Changes / Incidents 6
2.2 Rod Drop Times 6
2.3 Control Rod Mechanism Testing 6
2.4 Rod Position Calibration 6
3.0 THERMOCOUPLE AND RTD CALIBRATION VERIFICATION 9
4.0 PRESSURIZER TESTS 11 4.1 Thermal Transients 11 4.2 Heater Capacity 11 5. 't ggETROL SYSTEMS 12 6.0 TRANSIENTS 12 7.0 IFITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKg 12 7.1 Initial Criticality 12
-7.2 Reactivity Computer Setup and Checkout 13 7.2.1 Setup 13 7.2 2 Checkout 13 8.0 CONTROL POL WORTH MEASUREMENT 15 8.1
~ Test Description 15 8.2 Data Analysis and Test Results 15 8.3 Evaluation of Test Results 16 9.0' TEMPERATURE COEFFICIENT MEASUREMENTS 20 10.0 BORON WORTH AND ENDPOINT MEASUREMENTS 20 11.0 POWER DISTRIBUTION 22 i
p.
TABLE OF CONTENTS (Continued) 12.0 M NON REAOTIVITY 25 13.0 SHUTDOWN MAROIN CONSIDERATION.1 25 14.0 EXCORE DETECTOR BEllAVIOR 26 14.1 Intermediate Range Detectors 26 14.2 Power Range Detectors 26 15.0 OvrRPOWER. OVERTEMPERATURE AND DELTA FLUX SETPOINTS CALCULATION 29 15.1 overpower and Overtemperature AT Setpointo 29 15.2 Delta Flux Input to Overtemperature AT Setpoint 29 16.0 TUEL PERFORMANCE 32 17.0 pp1CLUSION 32 11
4 LIST OF TABLES Table EASA 3-1 RTD Calibration Check 10 4-1 Heater Group Power Supply Readings 11 7-1 Reactivity Computer Checkout 13 7-2 Reactivity computer Setup 14 B-1 Critical Rod Configuration Data 17 8-2 Comparison of Inferred / Measured Bank Worths With Design Predictions 18 10-1 Boron Worth and Endpoints 20 11-1 Initial Power Escalation Flux Map Rasults 22 13-1 Excess Shutdown Worth Available for a Full Power Trip 25 14-1 Power Range Detector BOL Calibration Currents (105 percent) 27 14-2 Axial Offent Constants 27 15-1 Overpower AT Constante 30 15-2 overtemperature AT Constants 31 111
i LIST OF FIGUREE E12nta Ean2 L
1-1 Final core Loading Pattern U2C18 4
1-2 DOL Burnup Data 5
2-1 PBNP U2C1B Cold Rod Drop Times (Full-Flow) 7 2-2 PBNP U2C18 Hot Rod Drop TLmes (Full-Flow) 8 B-1 Reference Bank Differential Worth 19 10-1 Boron Concentrations During BOL HZP Physico Testing 21 11-1 Power Distribution at 28 percent Power 23 11-2 Power Distribution at 100 percent Power 24 14-1 Intermediate Range Detector Response to Power Level 28 26-1 PBNP Unit 2 Primary Activity 33 iv
PREFhCE i
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 17 in November 1991.
)
Westinghouse performed the core design calculations for Unit 2 Cycle 18.
The reactivity coefficients were calculated based on estLmated Cycle 17 burnup of 10,750 MWD /MTU. Actual burnup was 10,778 MWD /MTU. Cycle 17 was ended on September 28, 1991, with a peak assembly burnup of 44,478 MWD /MTU and average assembly burnup of 31,050 MWD /MTU.
Electrical power was first generated during Cycle 18 on November 14, 1991.
This report is intended primarily for the use of Wisconsin Electric power company personnel as a readily accessible, complete compilation of reduced
- data, v
l 1
I
~
1.0 REFUILTNQ 1.1 Summary Fuel Hovement A complete core unload was performed to allow the safety
' injection system to be taken out of service for modifications.
The core was unloaded from about 0130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> on October 9, 1991 to 1530 on october 11, 1991 by plant operations personnel. The shifts were slightly shorter than before, with the first shift working 0600-1530 and the second shift 1530-0100. Noteworthy items during the unload were:
1.
SFP t'derload trip setpoint was lowered from 1000 lbs.
to 93 lbs. because some of the optimized fuel assem
.es (F/As) without inserts weighed close to 1000.
4.
2.
F/A H25 from core location 19 experienced an overload when being raised out of the core.
Since this is a standard F/A with a control rod, the overload was insignificant. The tare setting was readjt sted to prevent further spurious overloads.
3.
A white flaky substance was noticed coming off of F/As as they were drawn into the mast of the manipulator.
This substance was not observed on any F/As as they were later ultrasonicly tested (UTd) in the SFP.
4.
The takeup reel for the air supply stopped rotating about a foot from full takeup. At the same time the measuring tape for elevation readings was pulled off its takeup reel. The tape was removed since the primary indication of elevation was still working (digital readout). Operations personnel had to manually assist in the last foot of takeup of the air supply.
After the core unload, UT examinations were perfort.ied in the STP on all reload F/As taken out of the core. There were no indications of any leaking fuel pins. The average time between examinations was about 30 minut.se from October 11, 1991 to October 16, 1991.
From October 16, 1991 to October 18, 1991, inserts were moved to obtain the configuration needed for the next cyclw. All insert moves were performed without incident.
From October 17, 1991 to October 18, 1991 and on October 21, 1991, UT examinations were performed on all the discharge fuel. No indications were found of any leaking fuel pins.
Visual Inepectione After Core Unload 1.
One control rod (R502) was inspected at the periscope to determine the condition of the chrome coated rodlets after one cycle of operation. There were some rub marks, but they were superficial.
Page 1 of 33
. ~
9 2.
F/A R69 was inspected at the periscope because the hot channel factor for the location it was in (C3) during cycle 17, was about 10 percent higher than for symmetric locations. The in-core thermocouple (T/C) at C3 was also reading highsr than normal during cycle 17.
There was no indication of any abnormalities on F/A R69.
Another T/C at core location r6 read lower temperatures than expected. A Maintenance Work Request was submitted to check the T/C wiring for the two T/Cs in question, to look for reversed wires or other problems.
The T/C head i
connections were already removed before they could be checked for reversed connections.
However, the rest of the wiring was checked and was satisfactory.
Westinghouse investigated the abnormal peaking factor at core location C3.
Preliminary conclusions were that the computer codes used to generate the core design may have given a bias to C3 when the core design was folded out j
from quarter core symmetry starting at the diagonal-j intersecting C3.
Newer computer codes were used for Unit 2 Cycle 18 with good symmetry between C3 and its symmetric partners.
The core reload was started on October 23, 1991 at about i
2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br /> and was completed on october 27, 1991 at 2315 hours0.0268 days <br />0.643 hours <br />0.00383 weeks <br />8.808575e-4 months <br />.
The following notes highlight the reload:
1.
Excore detector basell.te count rates taken at..er three F/As were loaded in front of each detector. There were no source assemblies loaded.
The count rates were 75 CPS for N31, 47 CPS for N32 and 11 CPS for the spare detector (wide range de*.ector N40 near core location A7.
The maximum count rates vere 89 CPS for N31, 65 CPS for N32 and 19 CPS for N40.
The administrative limit for count rates was tw'.ce the baseline count rates for two of three channels.
These limits were met for the entire reload.
2.
Cavity visibility was very poor and did not improve substantially for about three days.
3.
F/A V77 was loaded off both core pins.
This was noticed by observing a 1 inch gap at top of core after V77 was released by the manipulator.
4.
F/A T78 was loaded off both core pins.
It was not noticed until F/A S53 was lowered into an adjacen; location with a load loss of about 550 lbs. 15 inches above the core pins.
F/A T78 was inspected with only slight scratches noticed.
F/A S53 was inspected with no signs of damage.
5.
New F/A V60 with new RCCA R148 was moved to temporary core location M7 after it couldn't be loaded into C5 because adjacent F/As were leaning into C5.
When trying to re-engage F/A V60 to return it to location C5, the correct elevation for latching could not be obtained.
After several efforts it was noticed thats F/A V60 top nozzle was displaced into core location a.
M8 about 3 inches with the manipulator mast raised about 1 foot above the top nozzle and indexed the same amount over M8.
Page 2 of 33
a b.
T.* hub of the RCCA was caught in the mast gripper mechanism with the RCCA raised about 1 foot out of the F/A.
F/A V60 was moved back to a vertical position in location M7 by moving the mast over M7 mast with no success. attempts were made to free the RCCA hub by shaking t Several A TV camera was lowered to determined the exact cause o binding.
The gripper was in the diseng&Jed position.The RC fingers.
It was decided to go to the grip position.
the gripper fingers were mechanically prevented fromEven though moving to the grip position, there was enough clearance to obtain slight movement, which was enough to free th RCCA hub after shaking of the mast cables.
e The dummy F/A was aoved into containment to check the
~
operation of the mast gripper.
to the new fuel elevator for closeup inspectionThen V60 was transferred Mr. M. Arlotti (West. NFD) which was witnessed by plant personnel. performed the inspection, and RCCA R148 passed the inspection and were qua/A V60 Both F for reuse.
Only minor scratches were found on the lified outside of the hub of the RCCA and on the top nortle of F/A V60.
was found to be an good Llignment with the F/A and*ihe RC passed a drag test.
1.2 Core Deeign The following fuel design changes were incorporated in th fuel for the first time in Point Beach Unit 2:
e new 1.
Axial fuel blankets consisting of natural uranium dioxide fuel pellete at each end of the fuel stack to reduce neutron leakage and improve uranium utilization 2.
Integral fuel burnable absorbers using burnable poison rodded assem(bliesIFBAs).
Instead of external inserts in F/As, (BPRAs) as fuel pellets coated with a one toil thickness ofselected fuel rode contain tirconium diboride.
moderator temperature coefficientIFBAs provide power peaking and control.
3.
Fuel rod bottom end plugs with increased radius at the tip.
This has no effect on core safety considerations
(
New fuel (VS1 - V78)
I w/o U-235 and 12 OFAs with 4.0 w/o U-235.for Cycle 18 consists of 16 OFAs with 3
}
I This is the first cycle not using secondary source assemblies.
core configuration is shown in Figure 1-1.The as-loaded core matc a
ern.
The
- loaded, 120 are OFAs and 1 is of the older standard design (fro Of the 121 F/As the SFP) in location G-7 The as-loaded burnups for each F/A are shown in Figure 1-2.
m Two control rods used in Cycle 18 new control rods (R506 and R148)
(R2 ani R14) were replaced with Fuel Assembly and Control Rod Tracking, which contr lin acccedance w FC-15, service life of control rods.
o s the Page 3 of 33 Illi
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FIGURE 1-1 FINAL CORE LOADING PATTERN U2C18 l
I A6 A7 A8 RB2
$54 R TO 2H214 2N2D6 2N213 84 B3 86 87 88 B9 B 10
$55 T65 V68 V75 V70 T68
$52 R126 R5 c3 c-4 c5 c6 c7 ca C9 c 10 c-11 571 V52 V60 T54 T72 T60 V62 V54
$70 R148 R135 R133 D2 D3 D-4 D* $
D-6 D7 D8 D9 D-10 D 11 D 12 553 V55 069 568 u65 v57 u68 565 U71 V56 558 R139 R116 E2 E3 E4 E-5 E-6 E7 E8 E9 E 10 E-11 E-12 778 V63 5 73 764 U54 5 72 USS T69 5 74 V64 T66 R127 RIO R149 R110 F-1 F2 F* 3 F4 F5 F6 F-7 F8 F-9 F 10 F-11 F 12 F-13 R69 V71 T59 U78 u60 157 UT7 T51 u61 U66 T61 V72 R 72 2H209 R7 R53 R11 R29 2H204 G-1 G-2 G-3 G4 G5 G6 G7 G-8 G-9 G 10 C 11 G 12 G 13
$51 V76 T76 u62 576 067 N17 UT6
$67 U51 167 V77
$62 2N216 R114 R107 R18 2N212 N1 M2 N3 N4 N5 N6 N7 N8 N9 N 10 M*11 N 12 N 13 R67 V73 753 U73 U53 162 U72 T52 U58 U74 T55 V67 R 79 2N208 R32 R34 R17 t506 2N211 t2 I3 I-4 15 I6 17 I8 I-9 l-10 1 11 1 12 T73 V65
$78 T 70 U59 5 77 U56 T71 566 V59 T74 R31 R112 R28 RID 9 J2 J' 3 J4 J5 J6 J7 J-8 J9 J 10 J 11 J 12 560 V57 U64 575 063 U52 U75 543 U70 V51
$61 R103 R98 K3 K-4 K5 K6 K7 K8 K9 K 10 K 11 569 v58 V66 T58 T 77 T56 V61 V53 564 R111 R8 R115 t-4 L5 L-6 L7 t8 L9 L-10
$57 T63 V74 V78 V69 T75 559 R502 RS4 M6 M7 M8 R 77 556 R78 2H203 2H207 2H205 Page 4 of 33
e FIGURE 1-2 BOL BURNUP DATA
'1 2
3 4
5 6
7 8
9 10 11 12 13 R82 354 a70 A
40688 38295 41512 2168 217A 2168 355 T65 v68 V75 V70 T68 552 B
38482 27130 0
0 0 26581 38511 217A 2188 2208 2200 2208 2188 217A
$71 v52 V60 754 in T60 V62 V54 570 C
38552 0
0 28831 27170 28761 0
0 38317 2173 220A 220A 218A 2188 218A 22CA 220A 217B
$53 v55 U69
$64 065 U57 U68
$65 U71 v56
$58 0
38082 0 13360 32610 13211 14260 13464 32351 13289 0 38169 217A 220A 219s 2175 219s 219A 219s 2175 219s 220A 217A T78 v63 5 73 T64 U54 5 72 U55 169
$74 v64 T66 E
26263 0 32276 27555 15569 38910 16010 27510 32487 0 27141 2188 220A 217B 2188 219A 217B 219A 2188 2178 220A 218a a69 V71 T59 U78 U60 T57 U77 151 061 u66 T61 VT2 R72 F 41689 0 28986 13255 15802 28324 14285 27297 15830 13327 29306 0 40773 2168 220s 218A 2196 219A 218A 219s 218A 219A 2198-218A 2203 216e
$51 V76 T76 D62
$76 067 N17 U76 s67 U51 T67 VT'
$62 C 38004 0 27051 14192 38232 13387 28506 13638 38968 13913 27223 0 37951 217A 2208 2188 219A 217B 2199 108 219s 2173 219A 2188 2208 217A R67 V73 T53 U73 U53 162 U72 T52 058 U74 T55 V67 879 N 40955 0 28405 130a3 15451 27426 13549 27699 15897 13489 29303 0 41149 2168 2208 218A 2199 219A 2184 219s 218A 219A 2199 218A 2200 216e T73 V65
$78 T70 U59 5 77 U56 171
$66 v59 T74 1
26413 0 32523 27432 15442 38434 15478 27459 32007 0 26598 218s 220A 217B 2188 219A 2173 219A 218a 217B 220A 218a
$60 v57 U64 575 U63 U52 U75 543 UTO V51 s61 J
38461 0 13395 31995 13417 14173 13358 32495 14007 0 38860 217A 220A 219s 217B 2199 219A 2198 2178 2198 220A 217A
$69 v58 v66 T58 T77 T56 V61 V53 564 K
38652 0
0 29315 27262 28910 0
0 38449 217B 220A 70A 218A 2188 218A 220A 220A 217B 557 T63 V74 V781 v69 T75 559 L
3 & 98 26693 0
0 0 26923 38627 217A 2188 2200 2208 220s 218a 217A 877 556 R78 M
40875 38334 41278 2168 217A 2168 ASSEMBLY ID #
BURN'# (MWD /MT)
ASSEMBLY FUEL REGION Pago 5 of 33
2.0 CONTROL POD OPERATIONAL TESTING 2.1
~ Hardware Chances / Incidents-Two new control rods were used in cycle 18 at core locations C5 and H12.
The ARO position is 228 steps as specified in the Setpoint Document and Procedure FC-15.
During rod movement for cold rod drop testing, control rod position indication system showed control rod C7 moving out when Bank C was selected. Control rod G7 rod position indication remained at the full in position.
It was assumed that the RPI cables for these two rods were swapped. Subsequent investigation showed that RPI connections for 07 and C7 were corre^t.
It was found that the control rsd drive mechanism connectors were inadvertently swapped. The situation was corrected and testing resumed.
Reference MWR No. 915080.
A new test instrument was used along with and as a replacement to the Visicorder.
Qualification of the new test instrument was made by comparing traces made at the same time using the two instruments in parallel.
.'he new test instrument uses a multiplexer to allow the measurement of rod stepping or rod drops on all 33 rods at once.
However, rod stepping will still be performed one rod at a time to verify that all wiring connectione are correct.
2.2 Rod Droo Times See Figures 2-1, 2-2 and 2-3 showing all the rod drop times and RCS conditions. All rod drop times were well within the Technical Specification Limit of 2.2 seconds to dashpot.
Instead of pulling the stationary gripper fuse to initiate a rod-drop measurement on ono rod, the reator trip breaker was opened to initiate the drop of several rods at once.
There is a slight difference in the shape of the rod drop trace at the start.
'his is because there is no more arcing of current across a fuse as it is pulled from its fuse holder.
Because of the difference in the shape of the rod drop trace, a conservative starting point for the drop was chosen which increased the rod drop times slightly.
Scatter of the rod drop times did not change significantly.
2.3 control Rod Mechanism Testino All lift, moveable and stationary gripper signal trace shapes were normal.
2.4 Rod Position Calibration During hot rod drop testing, LVDT voltages were recorded at 20 steps and 200 steps to verify that the RPI coils were responding normally. Once full power operating conditions were obtained, the fu'l out rod positions for any RPIs reading lower than 228 steps were verified. Then the RPIs were aligned using the SPAN adjustment without changing the ZERO settings.
Page 6 of 33
FIGURE 2-1 PBNP U2C18 COLD ROD DROP TIMES 1
2 3
4 5
6 7
8 9
10 11 12 13 I
I I
A SA SA B
1.37 1.44 1.88 1.93 CA CD CA C
1.39 1.42 1.4s 1.90 1.94 1.91 CC CC D
1.3s 1.41 1.90 1.93 CA SB 56 CA E
1.32 1.u 1.47 1.37 1.78 1.97 2.00 1.84 SA CB CS 5A
-F-1.42 -
1.45 1.46 1.39 1.94 1.95 1.98 1.89 CD CC CD G-1.38 1.41 1.40 1.89 1.85 1.93 SA CB C6 SA H-1.37 1.A2 1.50 1.43 1.87 1.95 2.02 1.98 CA
$8 58 CA 1
1.40 1.42 1.48 1.36 1.90 1.94 2.07
- 1. 86 CC CC J
1.46 1.36 1.99 1.91 CA CD CA K
1.45 1.44 1.36 1.91 t.94 1 32 SA SA L
1.41 1.36 1.92 1.89 M
L GEND DATE 1/7/92 BANK x.xx
- Time To Dashpot (sec)
TIME 11:16 x.xx
- lime lo seat (sec)
Maxissn drop tine (dash) e N 8 1.50 TEMP 128
'F Mininsn drop time (dash) s E 3 1.32 Average time (dash) =
1.41 FLOW 100 PRES 350 PSIA Page 7 of 33 l
_._...._..m_.
_ - ~ _
e rsoun 3-2 PBNP U2CiB HOT ROD DROP TIMES 2
2-
'3 4
5 6
7 8
9 20 12 32 23 1
1 1
A
$A
$4 8
1.3s 1.42 1.90 1.91
]
CA CD CA C
1.42 1.4s 1.46 1.93 1.97 1,91 CC CC D
1.40 1.45 1.95 1.95 CA
$6 SS CA E
i.42 1.40 1.47 i.42 1.83 1.93 2.00 1.90 SA CB CB SA F-1.43 1.44 1.45 1.42 1.95 1.95 1.95 1.90 CD CC CD G-1.'o 1,42 1.40 1.90 1.86 1.92 SA CB CB SA H-1.40 1.47 1.4e 1.43 1.90 1.97 1.99 1.96 l
CA
$8 58 CA 1-1.43 1.43 1.46 1.41 1.93 1.95 2.02 1.91 CC CC J
1.45 1.40 1.96 1.93 CA CD CA K
1.47 1.44 1.38 t
1.93 1.94 1.53 SA SA i
[
1.45 1.38 1.95 1.88 M
L GEND l.
DATE 11/15/91 BANK a.ax
- Time To Dashpot (tec)
TIME 10:37 x.xx
- Tire To Seat (sec)
Maximun drop time (dash) = M S 1.48 TEMP 530
'F Minicun drop time (dash) = K 9 1.38 Average time (dasn) =
1.43 FLOW 1 0 PRES 2000 PSIA
(
Page 8 of 33
.. = _ _. _..
3.0 THERMOCOUPLE AND RTD CALIBRATION VERIFICATION During initial RCS heatup for Cycle 18, loop RTDs and incore thermocouples were checked for normal respnse throughout the heatup range of about 195'F to 530'F (HEP).
Tacle 3-1 shows the results. All 16 RTDs were within the expected 2'T deviation of each other throughout the heatup.
Core exit thermocouples responded normally. One thermocouple (110) is DOS.
l l
Page 9 of 33
.m.. - - >._
1 9
TABLE 3-1 pfD CALIBRailDN CHECK
\\
STD flenent '
RfD Tenoeratures f ree Measured Resistences ('F)
LOOP A COLD LEC a 4018 200.3 249.8 3 06.3 345.9 406.6 453.6 4 99.5 528.9 R 405B 2D0.2 249.7 3D6.2 345.5 4D6.2 453.2 499.1 528.1 W 4028 1 99.9 249.5 306.0 345.1 4D6.1 453.2 499.2 528.3 W 4D68 1 99.9 249.6 3D6.1 345.1 406.2 453.2 4 99.2 528.4 l
f LOOP A NOT LEG a 601A 200.1 249.7 306.2 345.2 406.3 453.3 499.3 528.3 a 405A 200.7 250.4 3D6.8 145.7 4D6.7 453.5 4 99.4 528.3 W 402A 200.1 249.9 306.4 344.9 4D6.3 453.3 4 99.3 528.3
. W 406A 1 99.8 249.6 3D6.1 344.5 405.9 452.9-498.8 527.8 LOOP B COLO 4EG B 4038 200.1 249.9 3D6.5 344.2 4D6.4 453.4 4 99.5 528.5 8 4078 2 00.4 250.4 307.0 345.2 406.9 453.9 500.0 529.0 Y 4D4B 200.7 250.8 307.5 344.7 407.1 454.1 500.1 528.9 Y 4L8a 200.2 250.2 336.8-343.9 4D6.5 453.5 4 99.7 5?8,6 LOOP B Hof LEG B 403A 200.0 249.9 306.3 144.3 4 D6.0 452.9 498.8 527.8 8 407A 199.9 249.8 3D6.3 I43.7 4D6.1 453.0 499.1 528.0 Y 4D4A 201.1 251.5 3DS.1 345.6 407.6 454.5 500.4 529.2 Y 4D8A 1 99.9 250 2 306.8 344.3 4D6.5 453.7 499.9 528.9 r
RTO AVERAGE 200.2 250.1 336.6 344.9 4D6.5 453.4 499.5 528.4 S.C. TEMP 197 246 305 342 4 04 452 4 99 527 CORE Extf 1/C 200 253 309 346 408 455 5C' 529
[.
l l
l i-l' l
l l
Page 10 of 33
=
4 4.0 EEKSSURIEER TESTS 4.1 Thermal Transienta Pressuciter pressure incre0se rate with spray valves indicated shut and all heaters on war. 14.5 psi / min. This is typical and close to the nominal value of 14 psi / min. During the thermal i
equilibrium test, Heater Group A was required to be on 56 percent of the time to maintain pressure with main spray valves shut.
Spray valve effectiveness was normal with the 'A" loop valve j
decreasing pressure at 128 psi / min and the "B'
loop at 150 psi / min.
Spray bypass valve positions were adjusted so that spray line temperatures were maintained above 475'F.
4.2 Heater Caoneity i
Pressurizer heater capacity was determined from direct volt / amp
]
readings on each group of heaters. Table 4-1 shows that heater capacity is above Technical Specification requirements of 100 KW minimum for the heater groups operational during omergency conditions (Groupe A, C and D).
Heater Group A current readings were slightly greater than normal.
TABLE 4-1 HEAIER GROUP POWER SUPPLY READINGS I-Current V-Voltage KW-Energy Heater Input Group (amps)
( volt,s )
KW-/3xVxI/1000 1
d A
301 483 252 B
250 480 208 C
226 478 187 D
223 476 184 E
279 480 190 TOTAL 1,021 l
l l
l Page 11 of 33
+
l 5.0 CONTROL E STEMS There were no difficulties encountared during heatup or startup of the pressurizer level, pressuriser pressure, and rod control systems.
6.0 TRANSIENTS There were no transient tests performed during startup er approach to full power.
The limitation on power ramp less than 3 percent /hr was exceeded during the initial approach to full power.
Tha maximum ramp occurred just above 20 porcant power and was determined to be 7.3 percent /hr Ly RCS AT.
7.0 INITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKS 7.1 Initial CritisAllty The approach to criticality was made in two phases. The first step, which began at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br /> on November 12, 1991, 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 averaged about 110 ppm /hr oc 43 gpm.
When criticality was reached near 0100 on November 13, 1991, actual measured boron concentration was 11 ppm greater than estimated concentration of 1394 ppm.
1CRR plots were maletained during each phase of the approach to criticality. All plots were as expected with a more pronounced " knee" in the dilution phase due to tho absence of the secondary sources.
The reactor conditions at the time of criticality were determined to be as follows:
Date:
May 19, 1991 Timo:
0100 RCS Temperatures 530 'r RCS Pressure:
1985 peig Rod Position Bank D at 180 steps Boron Concentration:
1405 ppm i
i l
l Page 12 of 33
4 7.2 Reactivity comogter Setuo and Checkout 7.2.1 fetuo Table 7-2 shows the reactivity computer setup results.
Test 1 is a static test which tests for the reactivity sero 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 3 10 amps on the Keithley picoammeter. Normal flux levels for physics testing are about one-third the point of adding heat by procedure.
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 i
doubling time was measured with a stopwatch. Measured reactivity was then compared to design reactivity calculated from the measured doubling time.
Table 7-1 shows the results.
Differences were within 5 percent, which is acceptable.
TABLE 7-1 REACTIVITY COMPUTER CHECKOUT Measured Measured Calculated Difference Doubling Reactivity Doubling H:n x 100 Time (sec)
(pem)
Time (sec)
D 86.5 43 88.7
+3%
51.0 63.5 53.9
-5%
40.7 75 43.0
-5%
i l
l Page 13 of 33 l
TABLE 7-2 Cstu 680 11 18 S8 KlulAC COMPUTER SETUP Uhlt CTCLE DATE BURuuP BETA I
L STAR MWD /Miu TOTAL (MS) 2 to 01 09 02 0 0.005921 0.97 21.3 k
DELAYED CROUP 1
2 3
4 5
6 SETA FRACil0N 0.000192 0.001224 0.001D98 0.002341 0.000858 0.000?07 LAM &OA 0.0128 0.0315 0.1212 0.3224 1.4D4 7 3.8557 luPV POT WuMatt 11 12 21 22 31 32
$ttilW3 1.1919 3.7379 1.2909 3.6605 1.1691 0.7742 AS LEFT #1 1.1915 3.T39 1.291 3.662 1.1 70 0.774 As LEFT #2 FEEDBACK PCT WUM8ER 13 14 23 24 33 34 SETTING 1.2SD0 3.1500 1.2120 3.2240 1.4D47 3.8557 AS LEFT #1 1.280 3.149 1.212 3.224 1.405 3.R54 AS LEft #2 7EST 1 Sti POT 16 TO 9.200 (Votis). Poi 35 $Mout0 SE 5.7434 AS LEFT *1 5.706 A5 LEFT #2 ADJUST POT 35 Umill AMPLIFIER 14 (RHO) DUTPUT Is 0.0 Votis.
AMPLIFIER WUMBER 11 12 21 22 31 32 AMPLIFIER VOLTS 8.567D4 10.92/97 9.79855 10.44554 7.65679 1.64727 AS LEFT W1 8.58 11.0 9.99 10.54
- 7. 7D 1.85 AS LEFT #2 l
TEST 2 SET Poi 26 TO ABOUT 0.75 V PCT 25 SElithG 0.20 0.50 0.80 1.10 1.40 1.70 2.00 2.30 2.60 PER100 (SEC) 500.00 200.00 125.00 90.91 71.43 58.82 50.00 43.48 38.44 T DCLG (SEC) 346.57 138.63 86.64 63.01 49.51 40.77 34.66 30.14 26,66 06SERVID T 0 81 319.92 134.16 E3.18 60.28 47.26 38.87 12.75 28.49 25.05 l
085ERVID T D #2 l
EXPECTED RNO (PCN) 12.87 29.52 43.83 56.42 67.66 77.84 87.13 95.69 103.63 OBSERVED Rho #1 14.09 30.65 45.59 59.45 69.16 79.10 88.26 96.68 104.94 OBSERVED RHO #2 1
DATE th!TIALS 11/10/91 Das l
Page 14 of 33
1 a
- 8.0 CONTROL ROD WORT 11 MIASUREHENTS 8.1 Test Descrietion A new method of measuring rod worth, developed by Westinghouse, was tested for accuracy prior to the normal rod swap method.
.This new method, called ' Dynamic Rod Worth" measures rod worth by the rate at which a group of rods shuts down the reactor as they are inserted in a continuous fashion to bottom of core. This method does not require any boron changes or the use of a reactivity computer. The data for all rods was collected in less than two hours. The results are not available as the analysis methodology had not yet been developed. This test was performed on November 13, 1991 per Appendix A to Procedure RESP 4.1 for the U2C18 startup.
The rod worth verification, utiliaing 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 with 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.
'Jaing 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 dif f erential worth measurements.
Starting at a critical position with the reference bank fully inserted and Control Bank C fully out, a new critical configuration at constant RCS boron concentration was established with control Bank C fully inserted and Control Bank A at 136 steps.
Control Bank C was then withdrawn and Control Bank A inserted to one step to establish the initial conditions for the next exchange. This sequence was repeated until a critical position was established for the reference ban' with each of the other banks individually inserted.
Criticality determinations before and after each 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.
8.2 Data Analysis and Teet Results The integral reactivity worth of the measured bank is inferred from the swapped portion of Control Bank A by the following equation:
l
-- A o (a ) ( Ao )
=
+
1 x
2 l
l where:
i l
Page 15 of 33
.i
-Wj - The inferred worth of Bank X, pcm.
WY = The measured worth of the reference bank, coatrol A,.from fully withdrawn to fully inserted with no other bank =in the core.
Oz.= A design correction factor taking into account the fact that
.the presence of another control rod bank is affecting the worth of the reference bank.
A O: = 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.
Api. = The measured worth of the reference bank f rom the fully insertad condition to the elevation at which the reactor was just critical prior to the worth measurement of Bank X.
In this test A O: is zero because Bank A was fully inserted.
W! = The worth of Bank X from the initial position (before the start of the exchange) to 228 steps. This worth is messured 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 Table 8-2.
Values for o, obtained from the design predictions, are also x
listed in Table 8-2.
8.3 Evaluation of Test Pesults
~
A comparison of the measured / inferred bank worths with design predictions is presented in Table 8-2.
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 percent.
b.
The inferred individual worth-of each remaining bank agrees with design predictions within 115 percent 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 le greater than 0.9 times the predicted sum.~
All review and acceptance criteria were met.
Although Control Bank B was outside the-115 percent part of criterion "B",
it was within the.100 pcm limit. This is consistent with recent results from prior cycles.
l Page 16 of 33 l
=
~ - -.
1 TABLE 8-1 CRITICAL ROD CONFIGURATION DATA RCS CA Bank Bank Time Tavg Position Position
('F)
Steps Steps CC 1920
$30 1
229 CC 1955 530 136 1
SB 2018 530 1
229 SB 2035 530 S4 1
SA 2056 530 1
229 SA 2110 530 125 1
CB 2127 530 1
229 CB 2139 530 80 1
CD 2150 530 1
229 CD 2203 530 100 1
Boron concentration was 1296 ppm.
i l:
Page 17 of 33 l
__m...
TABLE 8-2 COMPARISON OF INTERRED / MEASURED 2ANK WORTHS WITH DESIGN PT'DICTIQLi a o2 W[
W)
W (I-P)/P x 100 Bank I a,
pem pem pea pem 4
CC 601 0.982 0
1131 1143
-1.1 SB 995 1.018 0
708 712
-0.6 SA 695 0.967 0
1053 1033
+1.9 CD 1165 1.065 0
792 770
+2.9 CB 920 1.010 0
480 510
-5.9 CA 1721 1751
-1.7 TOTAL 5885 5919
-0.6 r
Page la of 33
FIGURE B-1 PBNP UNIT 2 CYCLE iB BOL HZP REFERENCE BANK DIFFERENTIAL WORTH 17 o MEAFJRED WITHlk s10% of DESICu
~
16 x
- MEATJRED ouTSIDE s10% of DEE?CW
~
15 7 oo 0
00,0 14 1 13 o,
x 12 1 I
m
,a o
y 11 r x
Q h
10 h x
e y
9 b x,
n; j
x 8
x y
I 4
x 8 7 tg x
m n
x a
x xx *x j
{
x x z
ta s
a:
y 6 7 g
g
~
5 I
x 4
x x
3 x
x 2
l 2
x l
1 h x
x l
I O
O 20 40 60 80 100 120 140 160 180 200 220 STEPS WITHDRAWN Page 19 of 33
9.0 TEMPERATURE OOEFFICIENT MEASUREMENTS A near all rods out isothermal temperature coefficient measurement was taken during zero power physics testing. The measured value is the average of the recorded reactor coolant system bestups 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 isothermal temperature coefficient was +0.9 pcm/'F, within the review criteria of 13 pcm/'T of the design isotnermal temperature coefficient of +0.8 pcm/'F for $30'F and 1424 ppm.
10.0 BORON WORTH AND EEDPOINT MEASUREMENT.S Figure 10-2 shows RCS boron concentration during zero power physics testing.
Table 10-1 shows results of the endpoint measurements. The measured boron worth was obtained by dividing bank worth (pcm) it.t o change in boron concentration between endpoints. The review crite.-ion of 2 0.f pcm/ ppm was met.
TABLE 10-1 BORON WORTH AND ENDPOINTS Endpoint Bank Worth Boron Worth i
Design Measured Design Measured Design Mearured configuration PPS ppa ppa ppa pen / ppa pen / ppa ARO 1409 1424 CA in 1235 1246 1751 1721
-10.1
-10.1 At measurement conditions (530'F) l l
Page 20 of 33
o o
o o
0900.,,,,,,,,,,,,,,,,,,,
1000 o
<- ARO ENDPOINT o
1100 o
o 1200 o
1300 o
o 1400 o
0 1500 o
0 1600 o
-4
~
E m
o o
1700 o
O
~
h o
<- BANK A IN Q
m O
o$m 1800 o
-i m
>T c -,.
o II g" m
-4 < g n n
H H w 2 o-I O
1900 2H c
O$c OO I
m mm Z
A H
O2 2000 -
Qm rg H
"o
- m tu i
I>
w -4 Nmydn 2100 O
o 0
2 r
W m
2200 -
CD 2300 -
o 2400 Page 21 of 33
11.0 POWER DISTRIBUTION i
Table 11-1 illustrates the margin of hot channel factors to their full power limite during initial poteer increase to full load.
Flux maps were taken *4 sing ANSI standard ANS-19.6.1-1985 as guidance.
Allowed power levels were calculated using the relationships ten FH and FQ versus power level in Technical specification 15.3.10.n.1.a.
Measured salai power distributien, compared to design, is known in Figures 11-1 and 11-2.
i i
Ili1TJAL POWER IECALATION I
FLUX MAP RESULU ALIAMfED POWER MAP DATg POWER TEIN.
BANK A0 NO.
MI53.
STEPS FDM FQ-r 1
11-14-91 28 0
89 95 187
+16.7 2 18-91 75 0
102 110 193
+7.0 3
11-19-91 95 0
106 113 195
+2.5 L
4 11-20-91 99.7 0
112 114 197
+4.0 1
7 11-2$-91 99.8 0
116 117 208
+4.9 1
r l
l i
I-l Page 22 of 33 l
l.-n
r2ounz 11 1 POINT BEACH UNIT 2 CYCLE 18 CORE AVERAGE NORMALIZED AXIAL POWER DISTRIBUTION 28% POWER BOL LtLIND:
t?t$1CW CVtVI
- +
+ - NEAtuttD Cuevt 1.5 I
(
1,4
. e.
.___{_
m fl
\\/N 1.3
+
j- ---
i l
I I
l r
- -- -} -
p -- - '4- -----+y------
l 1.2
~~
- + - -
~
/
I
\\
1.1 j
p, 4- -- +
I
\\l i
j A
f.!
--t----
1.O I
.i
- A.__., -
m
_1
,9 4
1
/
/
F
-4_
4
-_3 _ _
g
,g b
f l
I i
!.i j
i
- I
__/ !
_i t
w
.7 l.,\\
g
/
l
.6 l
'=
/l
.S
$4 i
-+ -~5
-+--l----*--
I f
i
- /
l l
\\
,4 __ j.
7 H
I
}
i I
i l
l I
_.I _
4__
i
,3
~
h I
t t
I l
l l
I
.2
--M
+-
- -i
~-
~
r t
i I
i I
.I i
.1
=
0 10 20 30 40 50 60 70 80 90 100 CORE HEIGHT (t)
Page 23 of 33
e 4
FIGURE 11=2 POINT BEACH UNIT 2 CYCLE 18 CORE AVERAGE NORMALIZED AXIAL POWER DISTRIBUTION HFP BOL E0XE llCIWD DillCW Cutvt
-*-+- MAwaf0Cutvt 1.3 l
1.2 7
- ---- q l.
\\/
x, i
k---- -
./,g'--
1.1
+-
- } --
1.0
--+
l'
\\_
l t
7
_ y
.8 2
~
l W
t
[
.7 4-P 5
I Wx
.6 d
. l I
-= --t-t i
.5
+-
- + - - -
l l
i
+
l l,
1
_.4
- i _.
i f
I 4
=
,4
{
i j
.3
+---- -
j I
-+
.2 4
I I
e i
O 10 20 30 40 50 00 70 80 90 100 CORE HEIGHT (t)
Page 24 of 33
t 12.0 IrNON REACTIVITY Xenon reactivity behavior data for Unit 2 Cycle 17 was supplied by Westinghouse separate from the WATCH data package.
Point Beach code Xenon will be run with a TDF1 of 0.95 and TDF2 of 1.2 to remain consistent with the Xenon Tables. Tables are supplied for BOL, HOL and Rob conditions.
13.0 SHUTDCNN MARGIN CONSIDEMUpKE i
Rod swap resulte were withic 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-12903 Table 6.2 was accepted as a valid shutdown margin determination. Table 13-1 i
calculates the excess worth available to Unit 2 Cycle 18.
TABLE 13-1 EXCESS._ SHUTDOWN WQEULAYh11,haLg IQR A FULL RQWER TRIP BOL (pen)
EOL (pem)
Shutdown Margin From WCAP 3770 4110 Required shutdown
-1000
-2770 Excess Worth
-2770
-1340 Page 25 of 33 1
,,-,,.m,
14.0 EXCORE DETICTOR BEHAVIQB 14.1 Intermediate Rance Detectors intermediate range detector currents versus power level are shown in Figure 14-1.
Intermediate range detector trip setpoints were estimated from the design power sharing of the closest F/As to the detectors. The trip setpoints were reached within the expected reactor power level range of 20 percent - 25 percent.
This shows that the core design changes for Cycle 18 had the expected impact on its.stmediate range detector response.
14.2 Power RmtLptiectora Table 14-1 lists the " tilt free" power range detector calibration currents corressending to 105 percent power at BOL.
These currents were calculated using the multi-map method at 100 percent power. The multi-map method was used as a conservative measure to ensure that core design changes did not affect the power range detectors unevenly.
The first flux maps for the multi-map calibration were taken on December 4, 1991 at a burnup of 600 MWD /HTU. At this time, Bank D position was at 212 steps to keep delta 21ux at or below +5 lercent. Also, the low-low insertion limit alarm was set at about 192 steps.
These conditions restricted the maneuvering normally done to obtain the desired delta flux and rod position combinations for the flux maps.
The test was canceled to wait for mor' normal equilibrium core conditions with rods further out.
On December 12, 1991 rode were fully withdrawn with delta flux near +6 percent to start another uet of flux maps.
The test was completed on December 13, 1991.
Table 14-2 shows the changes in the installed axial offset constants. The changes are probabil due to statistical variances since the last multi-map calibration, more than cycle 18 design changes.
As shown, the only channel that was changed was H41.
This was the only change that was in the conservative direction.
The other three channels already had axial offset constants that were conservative, compared to the new constants obtained from the test.
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 may have a tilt before the start of the de.lation.
This practice complies with Technical specifications and the Westinghouse position on core tilt.
I Page 26 of 33
- 4 I
l T7DLE 14-1 l
power RANGE DETICIQB ROL CALIBRATION CURRENTS (10$%i Nel N42 N43 Nel cycle 14 T
314 334 350 305 B
263 297 317 277 Cycle 15 T
267 280 286 256 B
229 255 265 238 cycle 16 T
253 260 280 234 8
211 232 259 215 Cycle 17 T
255 260 277 235 8
219 242 262 222 Cycle 18 T
254 255 270 235 B
215 237 254 218 TABLE 14-2 M L.. OFFSET CONSTANIg N-41 N-42 N-43 N-44 Before 1.55 1.55 1.63 1.55 After 1.58 1,55 1.63 1.55 l
Page 27 of 33
...~_ _. - -. - - - _... - -. - -. - - -
e f
4.
n FIGURE 14-1 U!!.'T 2 CYCLE 18 NI35 AND NI35 RESPONSE TO POWER LEVEL 8
4
@/[
t a/
ef r
A e
6 h
w x
p y
~
j 5
~
t
~
~
p.
3:
~.
o
~
W_
4 g
Ci
~
g2 e
Q 0:
3 4
y y
[
a 2
i gy V
i
~
p N36 TRIP 1
f e NI35
/
Q NI35 l
a
-9 0 ' !''
'i'
' 'i'
't
'i
'l' 1'
'l'
'i'
O 10 20 30 40 50 60 70 80 90 100 REACTOR POWER LEVEL (%)
Page 28 of 33 r
- y-+-+c e-=,=----a er - r - v s t
-t#--e
--e*-=*
w,-w r* - * -m+
e
-*r
-+e==-
to-*'-*wes*
"m-ei+--m--+-1 wi>me------
~^*----' --<
- =
. p 15.0 OVERPOWKR. OVERTEMPERATURE AND DELTA FLUX SETPOINTS CALCULATION 15.1 Overoower and Overterroerature AT Setootn(g shown below are the equations from Technical specification 15.2.3.1.B.4/5, effective during cycle 18.
Overpower AT 1
t,S
1 i
i
,1
- t:8
,tgS+1;,1*t.S, 1*tS'
~
4 j
Overtemperature AT
1+tgS' s A T, K,-X,'T i
1 i
~ T' (1*ta8 l
- Ts S,
,,1 + t, S, + K, ( P~ P') - f ( AI) j 4
i See Tables 15-1 and 15-2 for the constants associated with this cycle of operation.
15.2 Eelta riux Inout to overterrocrature AT Setooint The overtemperature AT setpoint is reduced when the excore detectors sense a percent power mismatch between the top and bottom of the core. The dead band is 45 percent and -17 percent before the setpoints are reduced.
For each percent (more than 5 percent) the top detector output exceeds the bottom detector, the setpoints are reduced an equivalent of 2 percent of the rated power.
For each percent (more than -17 percent) the bottom detector exceeds the top detector, the setpoints are reduced an equivalent of 2 percent of rated power.
I Page 29 of 33
-. - -.... -... -. - = -. - -. -. -.
p TABLE 15-1 DVERPMR.. dT_.QQMTANTS l
47,
= Indicated AT at rated power, 'T 7
= Average tettperature, 'r T'
s 573.9'T K,
s 1.089 of rated power K,
= 0.0262 for increasing T
= 0.0 for decreasing T r,
= 0.00123 for T a T'
= 0.0 for T < T' t,
= 10 esconds t,
= 2 secondo,'or Rosemount or equivalent n'TD
= 0 seconds for sostman or equivalent RTD t,
= 2 seconds for Rosemount or equivalent RTD
= 0 seconde for Sostman or equivalent RTD Page 30 of 33
~. - - - ~.. - -. - _ -. -..
. - _. - - - -. -... _ ~.
. p F
e TABLE 15-2 j
QY1RTEMPERATURE AT CONSTANTS AT,
= Indicated AT at rated power, *r T
= Average temperature,
'T T'
s $73.9'r P
= Pressuriser pressure, peig P'
= 2235 poig Ki s 1.30 K
= 0.0200 K
= 0.000791 i
t,
= 26 seconds 3 seconds 1:
=
2 seconds for Rosemount or v3
=
equivalent RTD 0 seconds for sostman or equivalent RTD
=
2 seconde for Rosemount or t.
=
equivalent RTD 0 seconde for sostman or equivalent RTD
=
i 4
Page 31 of 33
16.0 ryIu1RFOR)Q@l%
Figure 16-1 shows relatively low coolant activity just before refueling i
with still lower activity af ter refueling.
"there is indication of iodine spiking during large power transients.
This may be caused by a minor fuel defect.
We were unable to discover this with UT examination, but we are continuing to evaluate RCS chemistry.
17.0 CONC 1.USIM The following results of refueling activities should be highlighted.
1.
The bank swap method for measuring rod worth produced relatively excellent results.
2.
Coro design changes including natural uranium blankets and IFBAs did not significantly change the encore detectors.
3.
During initial power escalation, the magnitude of core power distribution hot channel f actors were typical, compared to those obtained in prior cycles.
The other Unit 2 cycle 18 startup and refueling activity results were normal.
Page 32 of 33
Unit Two Reactor Coolant Iodine Dets e.s:__
i l
CVCS DEMINS w
I A
A OFF LINE
^D
- e.osee--
O - - - - O...... o....... o-
.o
?
r A--
1 o
.Gy 1
u
)
J O
~~.,,,.... o....... o 4
1 o
O l
3 I>
l y>
o i
e u e
~
3 j
w g
1 3
0 0
0 g
R#
O !
i
)
M e.cose--
1991 Unit Two m
)
Refueling O
I-m
)
i i
y
)
i.
- e. sees e
o a
o i
Date 1
Unit One Reactor Coolant I-131 Activity (uCi/cc) l
-A-Unit Two Reactor Coolant I-133 Activity (uC1/cc)
---O -- Unit Two Reactor Coolant dose Equiv I-131 Activity
,r
--