3F0990-07, Cycle 8 Startup Rept
| ML20064A776 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/30/1990 |
| From: | Beard P FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F0990-07, 3F990-7, NUDOCS 9009280072 | |
| Download: ML20064A776 (69) | |
Text
. . - .
.c -
> 1 3 x. '
?
7 >
4 l'
, - C OR P O R AT IO N 1 14 .. 'CrystalLRiver. Unit-3 I
( Docket No. 50-302~ -l u
LSeptember'20, 1990 .
' 3F0990-07z. .i l l 1
'U.S. Nuclear Regulatory Commission 3
-Attention: -Document Control Desk -l
- Washingtoni D.C.; 20555 I L R
Subject:
- Cycle-Eight-Startupl Report i
Dear. Sir.:
-l .
l ll, l Florida Power Corporation hereby submits the ' Crystal: River' Unit 3 ~ Cycle.Eight -
1 ll ', >
- Startup Report.' This report is submitted in accordance with Technical Specification 6.9.1.3. .,
j 1
i Sincerely,.
1; ,
1-
.M. Beard, Jr.1
- Senior'.Vice . President-
[,' ~ Nuclear Operations' d
- x~ "
s LPMB:wla >
' Attachment!
~
n ' xc:'- ' Regional Administrator, Region II J Senior: Resident. Inspector l 1
l 3- j l.
. s a l lj'1 ,
- h ^ t
- q i 9009280072 90093o * '
POR ADOCK 05000302 ,
P PDC l i
. POST OFFICE BOX 219
. . - . . i
,(-/$f g. 1 ig I' 4
.oc - - ,
,:. 3 A
('
j \
DOCKET NO. 50; 800,
}$, f lqF; #
,u:: c;e
,c., : .
gn ;a. ,
k .i 3 Il F
mWi% ,'
Crystal River Unit 3 '
4 1
'!kk ' , ;
y . ;
- CYCLE 8 1 1
START UP REPORT '
,)
4 ,
4 s 3 f.-p\
.- L j.
i s
SEPTEMBER 1990 1
-- masm
, s
"""""umans
""*'am.um 1
w I .#-
" }
, / V V - ' .
Florida 1
- POWO" a CORPOR AT' d i
2
!!! i 1
1
^ '
! lf i NT a Table of Contents g
- TABLE OF CONTENTS............................................................i l
[ LIST 4 0F TABLES............................................................ 111 a
. LIST OF: FIGURES...............................................................iv q
1.0 INTRODUCTION
.......................................................... 1
-2.0 PRECRITICAL TESTING................................................... 4- )q 2.1 Calibration and Neutron Response of Source Range Monitoring .
P System............................................................ 4- R 2.2 Re actor Cool an t Fl ow Te s t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.3 Control Rod Drop-Time Tests.........................'............. 5 2.4 Chemical and' Radiochemical Tests................................ 6 4
- 2.5 . ' Pres surizer Ef fectiveness Test. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3 2.6 In Service Loose Parts and Vibration Monitoring System' Tests.... 6 a o
i ,0 3.0 ZERO POWER PHYSICS TESTING............................................ 11' O L3.1" I n i t i al C r i t i c a l i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 h 3.2' Nucl ear Ins trumentation 0verl ap. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 -
D 3.3 Sens i bl e He at Determinati on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . s . . . . . . 12 3.4 Re a c t i me t e r ' C he c k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
-3.5 Al l Rods, 0ut Cri ti cal ~ Boron . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14 - , ,
e 3.6 -Moderator'and Temperature coefficients' Measurements............ 14 l
'3.7 ' Control Rod Group Worths....................................... 17 s 3.8 Di f ferenti al - Boron Worth Determination. . . . . . . . . . . . . . . . . . . . . . . . . 18 ' ;
-3.9- Ej ected Control Rod Worth Measurement. . . . . . . . . . . . . . . . . . . . . . . . . . 18 ,
j h ;J 3.'10 Biological Shield Survey..... .-............................... 19 .j 3.11- Effluent and Effluent Monito: 19................................~19 l 3.12 Chemical : and Radiochemical :Te=ts. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.
. 3 p - . -- -.
< < ~4.0, FULLEPOWER ESCALATIONtTESTING.........................=............... 34 m ;5 L, 4,1 Turbine / Reactor Trip ' '
- Test.......................... 34 0 _
4.2 . Integral Control System Te st. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4 '
4 h 14 '.' 3 - Uni t Loss o f El ectrical Load. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4 .4 f ' Uni t ' Load Trans i ent Tes t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 -
14.5- -Reactivity Coefficients at Power
Test.......................... 35 4.6 Un i t He a t B al a n ce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 L4-7 . Core Power Distribution Test................................... 37 ac ' 4. 8 - ~ Bi ol og i c al . Shi el d Survey. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 1 RJ L4.9 Pseudo Rod Ej ect i on Test . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 '
4~.10 Shutdown From Outside' the Control Room. . . . . . . . . . . . . . . . . . . . . . . . . 42 4 p 4 .117 L o s s o f O f f s i t e Powe r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 4.12 Power . Imbal ance Detector Correl ation Test. . . . . . . . . . . . . . . . . . . . . . 42 a L 4.~ 13 Nuclear Instrumentation Calibration at Power. . . . . . . . . . . . . . . . . . . 45 l L <
4.14 Emergency Feedwater Fl ow Te st . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 2
1 I
i H
l- 1
,N , -L 't
^'
t-L 'i, r 3.:
Table of Contents.
[- n.g -(Continued)-
EAEA ;
Y 4.15 Turbi ne/ Generator Operation . .'. . . . . . . . . . . . '. . . . . . . . . . '. . . ... .:. . . . . . 46 f' 4.16 Dropped Control Rod Te s t . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . 46 4.17' Incore-Detector Test............................................. 46 4.18 Reactor Cool ant System Hot ' Leakage Test . . . . . . . . . . . . . . . . . . . . . . . . 47 *
\,i' 4.19 Pipe'and Component Hanger Hot Inspection'atlPowert............. 47 4.20 Chemical and Radiochemical Tests. . . . . . . . . . . . . . . . . . . .;. . . . . . . . . . . 47 ;
"5 ' . , 4.21 Ef fl uent - ano Effl uent Montitoring. . . . . . . .:. . .:. . .-, . . . . . . . . . . . . . . . 47 'I qqv : .
Ofl
'l t
i
, =>
e l'.
m I
w r I
i-I l
Li-
'I
(
t ,
_b i
11
<k:b:
i list of Tab'Lg1 !
l' IMdit EASA 2.0-1 Summary of Precritical Testing Resul ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l 2.3-1 Control Rod Drive Drop Time Tests Resul ts . . . . . . . . . . . . . . . . . . . . . . . . . . 8 !
2.3-2 Group Average Control Rod Drop Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 '
3.0-1 Acceptance Criteria Deviation Limits Between Measured !
and Pred i ct ed Val ue s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 3.0-2 Summary of Zero Power Physics Testing Results . . . . . . . . . . . . . . . . . . . . . 21 !
3.2-1 Nuclear Instrumentation Overl ap Test Results . . . . . . . . . . . . . . . . . . . . . . 23 !
3.4 1 Reactimeter and Doubling Time Reactivity Comparison . . . . . . . . . . . . . . . 24 l 3.6-1 Comparison of Measured and Predicted Reactivity 1 Coefficients...................................................... 25 3.6-2 Extrapolated Hot Full Power Moderator Coefficient of j Reactivity ..........s............................................. 26' ;
3.8 1 Differential Boron Reactivity Worth Test Results . . . . . . . . . . . . . . . . . . 27 ;
4.5-1 Summary of Temperature, Moderator, and Doppler Coefficients of Reactivity ........................................ 48 4.7-1 Meximum Linear Heat Rate by Incore Detector Level ................. 49 4.12-1 Summary of Power Imbalance Detector Correlation Test, 75% FP...................................................... 50 4.12 2 Summary of least Squares Linear Regression Analysis of Power Range Channels and Backup Recorder During Power Imbal ance correl ation Test,75% FP . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 4.12-3 Worst Case Minimum DNBR and Maximum Linear Heat Rate i' vs. Full Incore Offset, 75% FP .................................... 52 4.17 1 Comparison of Incore M:nitored Assemblies' Flux L Shapes, 75% FP .................................................... 53 i .
I 1
1 i
1 w
t .
t l
iii
List of Fiaures Fiaure Elat 1.0 1 Core Loading Di agram, Cycl e 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.0 2 Fuel Assembly ar.d Control Assembly Component Identifications.................................................... 3 3.2-1 Nuclear Instrumentation Detector Locations . . . . . . . . . . . . . . . . . . . . . . . . 28 3.2-2 Nuclear Instrumentation Detector Locations . . . . . . . . . . . . . . . . . . . . . . . . 29 '
3.7-1 Incore Moni tor and Control Rod Ma ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 3.7-2 Control Rod Group 5 Integral Wort 1, BOC 8 at HZP.................. 31.
3.7-3 Control Rod Group 6 Integral Worth, BOC 8 at HZP.................. 32 ;
3.7 4 Control Rod Group 7 Integral Worth, BOC 8 at HZP . . . . . . . . . . . . . . . . . . 33 4.7-1 Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axi al Powe r Imbal ance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5 4.7-2 Com)arison of Measured and Predicted Radial Power Pea (ing Factors with 3D Equilibrium Xenon at 75% Full Power.................................................... 56 -
4.7 3 Comparison of Measured and Predicted Maximum :
Total Power Peaking Factors with 3D Equilibrium :
Xenon at 75% Full Power ........................................... 57 4.7 4 Com)arison of Measured and Predicted Rtdial Power t Pea (ing Factors with 3D Equilibrium Xenon at 100% Full Power................................................... 58 4.7-5 Comparison of Measured and Predicted Maximum Total
- Power Peaking Factors with 3D Equilibrium Xencn at 100% Full Power................................................... 59
- 4.7-6 Hot Channel Minimum DNBR vs. Core Power Level . . . . . . . . . . . . . . . . . . . . . 60 i 4.12-1 Average Out-of-Core Offset vs. Full Incore Offset, i 75% FP............................................................ 61 ;
4.12-2 Backup Incore Offset vs. Full Incore Of fset, 75% FP . . . . . . . . . . . . . . . 62 ,
4.12-3 Maximum LHR and Worst Case Minimum DNBR vs.
I Full Incore Offset, 75% FP ....................................... 63 L
i I
l' >
4 f
iv
E 1 4 ;
l 4 )
l :
1.0 INTRODUCTION
l On March 14, 1990 Crystal River Unit '3 was shut down to refuel for !
Cycle 8. The' Cycle 7 fuel'had attained a core average exposure of l 497.9019 effective full power days (EFPD). !
Cycle 8 has been designed for a cycle lifetime of 566 i 10 EFPD and operation in a feed-and-bleed mode. The cycle utilizes burnable poison rod assemblies and gray axial power shaping rods- (APSRs). The final Cycle 8 core loading pattern is shown in Figure 1.0-1. The fuel ;
assembly and control component identifications for Cycle 8 are shown in i Figure 1.0-2.
This' report, prepared and submitted in accordance with Technical Speci- '
fication 6.9.1, describes the; precritical, zero power, and power testing performed during the Cycle 8 start up and also summarizes the results of these tests.
- for' Crystal precritical River Unit 3 (CR3) testing. The'was brought results to Mode of these tests3 are (Hotreporte Standby)d in ,
Section 2.0. Zero power physics tests began after criticality was i achieved on June 21. 1990 at 0520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />. The zero power physics ,
testing results are given in Section 3.0. Power escalation testing '
began after breaker closure on June 23, 1990 at 0542 hours0.00627 days <br />0.151 hours <br />8.96164e-4 weeks <br />2.06231e-4 months <br />. These test results are summarized in Section 4.0. As in Cycle 7, the all rods in *
'(ARI) temperature coefficient, the hot zero power (HZP) ejected rod worth test, and one intermediate power distribution test plateau were ,
not performed during Cycle 8 physics testing in accordance with a program for improying pla.it availability through the elimination of unnecessary tests.
i f i
,t>
i s
I" Reduced Physics Testing, Task Summary Report", document by Babcock and Wilcox, B&W document number 86-1164722-00, dated March 1987 for B&W Owners' Group Performar.ce Committa0.
1
.x . y, CORE LOADING DIAGRAM, CYCLE 8
- s I
i
& .i. Pa. PG4 P e
. i. . . . i. . a .
4 898 F E.8 F t. F tie F mit
- i. . i. . i. . a . i. i.
c .m F et F s. F att F au F
.n ie.s
- i. . i. . . . a . a . a g ii F Pit F L. F eji F Lil F a.t F i l' Ri l' a l' .". i' n. l'- & l' & l'
. . a . . . i. . . .
, ui = F en i.
F c.
i.
F = F i::
i.
F & i.
F = in
. & l' & l' & l' R. & & l' & l' & l' t
. . i. i. . ic . i. . i. .
m F Pp F . .a y . to F
.;> F . -7
. . . i. . a . . . i. . i. . i. .
t F14 F F L. F Fil ett c. F Le7 F a.? F F.t
. . a . . i. . a i.
,a ., .
t Fa F ai. 3 p F c. F .i.;i F . , .i
. a . i. . i. . i. . ,a
. !. en F n. F = F u. F m ,i. m a e a . a a . a . .
R .eii F .it F F.B F .e.I F Fli F .6 F Rol 4
F. . i. . i. . i. . a F.
g $a tt F 8 F lit F ti. F c.
I.t3 .
.. i. i. a a .
P .. F .t F F ie F tit
. aa n. Rt . .
~
l
! I I I , . . . t . . w n u n a n Y
m i a ei t.n... <L.
i . .F . t. .u...n SOURCE: B&W CR-3 C8 Reload Report i
i FIGURE 1.0-1 l
2
. .. . . . - . . . - - - - - . - - - - . . - . . - - ~ . - - . . . - . - . _ . ~ . - - - . . . - - . .
I l
l FUEL ASSEMBLY and CONTROL ASSEMBLY COMPONENT IDENTIFICATIONS f l CoflTAL RIVtt III CYCLE t C#t LOADIE PLM FUtl TRANSF(R CANAL - ) I I I I
l AfC 3VI 3UL 300 aTU A i l
497 4WL 48Y 4tpl 491 4WR 499 4WC 49C <
g CO33 C20V C009 l I
til 41! 486 4FU 49P 4V2 492 4WK 482 4WZ ffN :
C C044 NT! C014 8646 C013 86R2 G043 ;
RTtt RTII RTIE
)
i 490 4WW 3U7 4UW 485 4WH 20V 4VW 4UV 376- 4W 4A6 g CM6 0573 A062 BlTA C046 06RV V 96RI (all ;
Ritt RTII RTII RTII -
out 499 4UY 49M 4W1 448 4VL 448 4VD 49F 4VI 497 4R0 t I C032 BET 7 CM0 Bift C210 BlTB C211 blTT CM7 9674 CMt j Ritt Ritt RTII RTEI RTtt 3fR 4M 4WF 48U eVY 20U 4V7 4Al 4WN 200 4W3 49K 4WT 4tJ 370 C0tl t C031 BET 6 A061 BlTR C060 slRE B6R0 C0tt DETN A0lW tlTH C036 Ritt RTII RTEX RTRI RTIE RTII ^
All en 49W 4WN 4AA 4WE 494 48[ 490 evt 44M 4Wt 49A 4WN 3UA 6 C036 blTI C202 BlRL C200 C20W tlRN C212 Blft C039 !
RTIt RTEI RTII RTIE 3UE 48L 4F6 P0T 4VJ 48R est 3A10 48K 48P 4VM E 70C 4V4 4A9 3UB kN CPOT 864P CMS B6RU C024 CM0 tlT5 C004 3 6151 C20I RTII RTIE R9I RTII .'
l' 3pg 4W7 490 4Vf 4AC 4V9 49T 48F 49M 4VV 492 4VI 49V 4WE 3V3 [
g C023 DSTJ C216 86RJ C205 C20Y 86RF C213 BlTV C026 !
Ritt RTIE RTII RTIE i AfK 48V 4VR em 4WS 20F 4VG 4A3 4VA POR 4WD 49N 4W2 496 3TT I
( C027 DETL A060 BlTU C034 86Rt C037 BlRR C002 BlTM A0lt SSTO C057 i
- RTII RTII RTII Ritt RTII RTII
- 42 4tA 4WO 4A7 4WL 486 4VK 493 4WJ 49J 4UI 496 4W9 m C020 SSTF C0ll BlTC C215 SSRT C214 blTP Colt 8570 C030 ;
RT!! RTII RTIE RTII RTEI t
'~
49I 434 37t 4U2 49L 4WZ E16 4VC 496 4V3 3fH 4WY 4A1
- A C06) BERW A062 BlTW C016 86tY A0lY B579 C007 5 RTII RTII RTII RTII -l 304 4WW 487 4W4 495 4VS 4A0 4VP 498 4WT 228 ..
- 0 C019 titt CM9 86RM C044 0574 Colt J ,% 2 RTIE RTRE RTII -
A*.
498 411 4RD 4WP 440 4K3 48C 4WB 49t ,
P CM3 Cott C20R
(
, 3TN 3U4 3fJ 3V7 3Un i
A '
2 14 15 1
lt 3 4 5 6 7 8 9 10 11 12 13 y w.w U- FUtl 10 (3A10* tat 1120** tat li 2YN 3M* tat 7 3TG 3V7 etat St 400 4ADatat 9; 4UV 414* tat 10)
- CONTROL CWP. 10 (A APERAt BetPRAt CoCRA)
Retainers NDit: NJ0 prefines all except the latch IC Fuel Assembly 10 Numbers.
SOURCE: B&W CR-3 C8 +
FIGURE 1.0-2 Refueling specification 3 ,
hm - - - - - -~~, - , , . ..-+m, .----.-_.-,-.e--. -- --
i I
l
.2.0 PRECRITICAL TESTING l During all precritical testing required by Section 13.4 Table 13-2 of !
the CR3 Final Safety Analysis Report (FSAR), the actual. reactor power was maintained at a suberitical level. These tests'were completed in accordance with Performance Testing Procedure PT-100. The results of :
the precritical tests are summarized in Table 2.01. '
2.1 Calibration and Neutron Response of Source Rance Monitorina System ;
Precritical physics testing began following response check out of the-neutron source range detectors. These detectors' are two. BF3 i proportional counters which measure flux from the suberitical to. start up level. The Source Range Monitoring System was subjected to :
functional testing -and calibration per Surveillance Procedures SP-110
- and SP-112 and Technical Specifications 4.3.1.1.1 and 4.10.2.2. 1 Surveillance Procedure SP-102 verified that the source range .!
instrumentation provided a count rate of more than 2 counts /sec.
t 2.2 Reactor Coolant Flow Test l The Reactor Coolant flow test was performed to measure reactor coolant (RC) flow. The. results were compared with design calculations to verify adequate core coolant flow. The test was executed in accordance with Surveillance Procedure SP-224. ,
L The acceptance criteria of the test are as follows:l ;
i ;
Steady state total reactor coolant system (RCS) flow (three or four ;
pump operation) at 53212 'F and 21551100 psig shall be within the '
p following limits:
l 4 RCS Pumps, flow 2 139.7x10 3 RCS Pumps, Flow 2104.4x10l lbm/hr lbm/hr ,
Additionally, with all 4 reactor coolant pumps in operation, steady 'i state loop flows shall be within 2% of each other. To account for measuremant uncertainty, the following must be true:
Heasured Flow x 1.025 s 411,840 gpm (Max. Flow) o Measured Flow x 0.975 2 374,880 gpm (Min. Flow) ,
I SP-224, Section 3.6, Rev. 5 4
1 2.2.1 Method' During this test, the ACS temperature was maintained at 53212'F with a pressure of 21551100 psig. 1 L The steady state temperature, pressure, and flow of the two coolant l flow loops were taken etery minute for ten minutes. This data was then i used to calculate the average reactor coolant flow for each flow loop I and for the entire s'; stem.
2.2.2 Results The total RCS flow was calculated to be 153.49x106 lbm/hr. The calculation for measurement uncertainty was 409,949 gpm for the maximum flow and 389,951 gpm for the minimum flow. Therefore, the acceptance critoria was met. The flow difference between loops was calculated to _
l be 2.39%, which is greater than the 2% acceptance criteria. After ;
i compliance with the contingency for this situation, as per SP 224 ,
l Section 5.2.3, this flow difference between the loops was deemed ;
acceptable and precritical testing was resumed. j i
L 2.3 control Rod Droo Time Tests I
- The Control Rod Drop Time tests demonstrated that the drop times are in accordance with Technical Specifications requirements. The acce -
criteriaforthesetests,perSurveillanceProcedureSP-102,are:ptance l The individual safety and regulating rod drop times from the fully l withdrawn position shall be 11.66 seconds from power interruption at the control rod drive breakers to three-fourths insertion, 25%
position, with Tavg 2 525'F and either 3 or 4 reactor coolant pumps ;
operating.
In order to check for uncoupled control rods, the individual rod drop '
times were compared to their respective group average drop times. The '
individual rod drop time should not exceed its group average drop time by more than 0.05 sec.
~
L i
2.3.1 Method The Control Rod Drive (CRD) Drop Time tests were performed using strip chart recorders to time the rod drops. Each control rod group was ;
pulled to 100% withdrawn and then dropped into the core using the -
manual trip pushbutton. A zero time signal was furnished to the chart ,
recorders for each control rod assembly from a contact on the manual :
trip switch. A second signal to indicate three-fourths insertion was furnished by a reed switch located on the position indicator tube of each CRD. Both source range instruments were continuously monitored I SP-102, Section 4.0, Rev. 18 5
. _ _ - . _ - . . _ _ ~ _ _ , .- _ . . . - .
w .
f l r i
I during rod withdrawal and provided a count rate of more than 2 i I
counts /sec. ;
2.3.2 Results f The results of the Control Rod Drop Time tests are presented in Table '
2.3 1. The average drop time by group is summarized in Table 2.3 2.
~ By examination of Table 2.3-1, it is found that the shortest drop time ,
is 1.25 seconds for eight different control rods. The longest drop time was 1.33 seconds for CR l-2 and CR 1-4. The weighted average drop :
. time for groups 1 through 7 was 1.277 seconds, i Based on the results shown in Table 2.3-1, the acceptance criteria were [
met by all of the control rods. !
2.4 - Chemical and Radiochemical Tests Chemical and Radiochemical testing was not performet during this start- !
-u p . These tests were conducted at initial startup and no plant modifi- !
cations have been made which would invalidate the results of those !
tests. >
t 2.5 Pressurizer Effectjveness Test l No Pressurizer Effectiveness testing was done this startup since no ;
modifications _have been made to the pressurizer. Therefore, the results of the testing done at initial startup are still valid. ,
2.6 In-Service loose Parts and Vibration Monitorina System Tests The Loose Parts Monitoring Subsystem was calibrated during the course of the outage per Surveillance Procedure SP-152. The normal schedule '
I of neutron noise readings, combined with routine core physics and Loose Parts Monitoring Systems, is sufficient to alert plant personnel of degrading core internals, if it occurs. ,
1 l.
?
I f 6
E y-. . j-
- v. . - _ _
~
TABLE 2.0-1 'SUfflARY OF PREChi 1 CAL TESTING RESULTS .
Precritical Testing -
1-Tert fReference) Units Results Acceptance Criteria Acceptability Za. Reacw, 0:010:t Flow ' lbghr 153.49x106 Flow for four pgs OK (SP-224) mustbegreaterthan 139.7x10 lbghr 2.39% Loop flow within OK*
2% of each other
, b. Reactor Coolant Flow gpm 409,949. Max flow must be OK Measurement Uncertainty < 411,840 389,951 Min flow must be OK
> 374,880 v
- 3. Control Rod Drive Time seconds Tests (SP-102)
Least (eight locations) 1.25 All rod drop times must OK i Greatest (CR l-2, 1-4) 1.33 be less than or equal OK to 1.66 seconds Group 1 Average 1.31 None Group 2 Average 1.29 Group 3 Average 1.29 Group 4 Average 1.27 Group 5 Average 1.26 Group 6 Average 1.26 4
Group 7 Average 1.27 Groups 1-7 Average 1.277 None
- Deemed acceptable after review (see Section 2.2.2 Results)
__ . _ . ._ -._ _ _ ,- . , . _ . _ , . _ _ _., _ . . . ~ . _ . ~ . _ . _ _ _ _ _ _ . . _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ , _ . _ -
wp si g
t-TABLE 2.3 1 CONTROL kOD ORIVE DROP. TIME TESTS RESULTS Precritical Testing 1
Drop Time !
Control Rgd Core Position 1 seconds) ,
CR 1-1 B-10 1.32 -I CR l-2 F 14 1.33 .
CR 1-3. L-14 1.32 i
- c. CR 1-4 P-10 -1.33' !
CR l-5 P-6 1.31 i CR l-6 L2 1.32 '
CR l-7 F-2 1.32
, CR l-8 B6 1.30 l CR 2-1 0-11 1.29 CR 2-2 E-13 1.28 '!'
CR 2 3 M-13 1.28-CR 2-4 0-11 1.32- l CR 2-5 0-5 1.29 :
CR'2-6 M-3 1.29 !
CR 2 7 E-3 1,29 :
CR 2-8 C5 1.27<
CR 3 F-8 1,29 CR 3 2 G-9 1.29 i CR 3-3 H-10 1.29 !
CR 3-4 4-9 1.30 ?
CR 3-5 L8 1.30 i CR 3-6, K-7 1.27 !
CR 3-7 H6 1.28 CR 3-8 G-7 1.32 CR 4-1 E-9 1.29 i CR 4-2 G ll 1.29- :
CR 4 K-11 1.27 CR'4-4 M9 1.27
'CR 4 5 M7 1.28 ;
CR 4-6 K5 'l.27 CR 4-7 G5 1.26
, CR 4-8 E-7 1.27 j a
1 ,
f 1-I 8
- r. , , ,- ,
T J.
- i
,\ ,t Lt
' TABLE 2.3-1 CONTROL ROD DRIVE DROP TIME TESTS RESULTS Precritical Testing-(Continued) i Drop Time Control Rod Core Position (seconds)
' CR 5 1 C-9 1.25
.CR 5-2 E-11 1.27 Li CR 5 3 G-13 1.25 CR 5-4 K-13 1.27
- g!
p- CR 5-5 M-ll 1.30 CR 5-6 0-9 1.25 z i , CR 5-7 07 1.26 !
i, CR 5-8 M-5 1.26 t '
CR 5 9 K 1.25 CR 5-10 G-3 1.25 :
CR 5-11 E-5 1.26 !
CR 5 12- C7 1.25 l r.
CR 61 B-8 1.25 :
CR 6 2 F-10 1.28 '
CR 6 3 H 14 1.25 CR 6 4 L-10 1.27- 1 ~
CR 6 5 P8 1.26 CR 6-6 L-6 1.27 !
CR 6-7' H2 1.27 CR 6-8 F-6 1.27
{
CR 7-1 D-8 1.26
- CR 7-2 D-12 1.30 !
CR 7-3 H 'l.26 i CR 7-4 N 12 1.28 -!
CR 7-5 N8 1.28 CR 7-6 N4 1.27 t CR 7-7 H4 1.28- r
. CR 7-8 D-4 1.29
'I Source: SP-102 i
9
E .. , , vm f;o '
m, i ,
TABLE 2.3-2 GROUP AVERAGE CONTROL ROD OROP. TIME .
Precritical Testing l I
Number of Average Drop Time Group Rods (Seconds) i 1 8 1.31 l 2 8 1.29 l 3- 8 1.29 >
4~ 8 1.27 >
5 12 1.26 6 8 1.26 7 8 1.27 17 60 Average - 1.277 f h
t
[
e f
r i
5
'h 10
._ . _ - _ _ _ ~ _ _ -
w f
3.0 ZERO POWER PHYSICS TESTING l
The zero power physics tests (ZPPT) required by Section 13.4 Table 13 3 of the CR3 FSAR were performed in accordance with Performance Testing :
Procedure PT-110. These tests were done to verify nuclear design parameters used in the . safety. analysis, operational parameters, and limits set in the Technical Specifications. !
Acceptance criteria deviation limits for the ZPPT are given in Table :
3.0-1. A summary of measured and predicted values obtained for the i ZPPT is given in Table 3.0-2. ,
3.1 Initial Criticality t
3.1.1 Method-The reactor coolant conditions at criticality were 53212 'F, 21551100 -
psig, and 201: _ppmB. In order to verify that the source range l instrumentation was operating properly, a Chi Square Test was performed :
which confirmed that a good fit to a Poisson distribution existed. The !
approach to criticality began by withdrawing control rod grou) 8 to 25%
withdrawn and groups 5 and 6 to 100% withdrawn. Control roc groups 1 1 through 4 remained at 100% withdrawn. Rod group 7 was then withdrawn until criticality was achieved (24% withdrawn),
o Throughcut the approach to criticality, curves of inverse i
multi)1ication versus rod reactivity worth removed were maintained.
The c ata was taken from two source range detectors by two people, ,
independently. At the end of each control red group withdrawal, the i count rate was taken from each source range detector using the scaler , i counters. The ratio of the initial average count rate to the average :
count rate at the end of each reactivity addition was plotted and the criticality point determined by extrapolation. ;
3.1.2 Results Initial criticality for Cycle 8 was achieved on June 21, 1990 at 0520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />, t 1
3.2 Nuclear Instrumentation Overlaa '
The nuclear instrumentation (NI) detectors were used throughout testing to provide continuous reactor power information. The instrumentation consisted of eight measuring channels divided into three ranges:
source (subcritical tostartup), intermediate (startup to 150% full power), and power (0 to 125% full power). The location of these NIs is shown in Figures 3.2-1 and 3.2-2.
11
C .i g "-
l gr l 4 l
+ 1 Technical Specifications state that: !
V !
NI Source rgnge must overlap the intermediate range by a factor of o 10 or more l
5 This means that before the source range count rate equals 10 cps, the l intermediate range must be on scale. If the required one decade is not !
observed, the approach to the intermediate range cannot be continued until the situation has been corrected.
3.2.1 Method For this test, reactor coolant conditions were 53212 'F and 21551100 ;
psig. To verify the overlap requirements after initial criticality was !
reached, core power was slowly increased until the intermediate range i channels came on scale. Detector signal response was thus recorded for ,
both the intermediate and source range channels. This was then repeated for another decade. :
3.2.2 Results- ;
The results of 'the Nuclear Instrumentation overlap test are given in !
Table 3.2-1. This table shows that the average overlap between- the r source and intermediate range is 2.424 decades, which is above the minimum one decade specified as the acceptance criterion. ,
3.3 Sensible Heat Determination By determining the intermediate range current level at which the ..
production of sensible nuclear heat occurs, the upper zero power '
physics test current limit is established. Thus, by restricting ;
reactor power operation to a level reduced by a conservatism factor of ;
3.3 below the sensible heat level, the effects of temperature feedback .;
are eliminated-in the measurement of physics parameters. The test for 'i sensible heat was done according to Performance Testing Procedure PT- t 116. .
3.3.1 Method for this test, the intermediate range current level was increased in increments until sensible haat was detected. The production of '
sensible heat in the core is indicated by onc or more of the following:
increase' in RCS- Tav increasc in RCS loop Th ot, increase in pressurizer level, incS, ease in or leveling off of makeup tank level, and opening of the turbine bypass valve.
3.3.2 Results The point at which there was a definite heatup rate was 1.3x10-7 ' amps L measured on channel NI-4. This was defined as the sensible heat point.
l H I PT 110, Section 3.3.1, Rev.15, Tech Specs Table 4.3-1, Note 5 12
l y l l
lt From this, the upper current limit was established at 4.0x10-8 ,,ps, l This was found by dividing the sensible heat point by a 3.3 1
, conservatism factor. i i l An additional part of the measurement was to ensure that all power ,
range channels were indicating between 0.8 and 2.0% full power at the l
. point . of sensible heat. This was done to ensure that the overpower <
trip protection was adequate. At the point of sensible heat, the indication on the power range channels was approximately 1.0% full power. Thus, all power range channels were within their recommended i tolerances, and no calibration was performed. l
- s. I,
)
3.4 Reactivity Check i
The Reactimeter is the reacti ity computer manufactured by Babcock and Wilcox which solves the one dimensional, inverse kinetics equation with ,
six delayed neutron groups for core net reactivity based upon periodic ;
samples of neutron flux. In addition to reactivity and neutron flux, the Reactimeter can record 23 other analog and digital signals from the 1 plant.
After initial criticality and prior to the.first physics measurements, 'f an on-line functional check of the Reactimeter was performed to verify -t its readiness for use in the test program. ,
The Reactimeter check is subject to the following acceptance criterion: f The reactivity values computed from measered doubling times must agree withi the reactivity values measured on the Reactimeter.g 15% of
-3.4.1- Method After steady state conditions with a constant neutron flux were ,
established, approximately 50 pcm of negative reactivity was inserted !
into the core by inserting control rod group 7. Stop watches were used to measure the doubling time of the neutron flux and the inserted >
reactivity was determined from period-reactivity curves. The measurement was repeated for values of reactivity ranging from -44.7 to'
+54.0 percent millirho.
y 3.4.2 Results i
The reactivities determined from doubling time raeasurements were then ;
l compared with the reactivities calculated by the Reactimeter, j l
The results of the Reactimeter verification measurements are summarized in Table 3.4-1 The reactivity calculated by the Reactimeter was within the acceptance criterion limit of 15% of the reactivity determined from doubling times in each case.
I PT-110, Section 3.3.2, Rev. 15
[
13 l
4 , !
hj \
l ,, !
'f $
f.
?
3.5 All Rods Out Critical Boron Test This test was used to provide information relating to core excess reac- !
tivity by determining the amount of soluble boron that must be added to l the coolant water to maintain a critical level with all control rods ;
removed, j This test, performed in accordance with PT-111, is subject to the ;
following acceptance criterion: )
The all rods out, hot zero power, critical boron concentration !
r should be within 150 ppmB cf the value given in the Physics Test Manual. If this value is greater than 1100 ppmB, a plant review !
commit}ee review must be accomplished prior to escalation to i power e ,
3.5.1 Method ,
.L j The test measurements were made at a reactor coolant temperature of
. 532'F and a system pressure of 2155 psig. The portion of control rod i I group 7 which remained following deboration was withdrawn and the .:
excess reactivity measured using the Reactimeter. From this'value and h the differential boron worth given in the Physics Test Manual, the all rods out critical boron concentration was determined.
3.5.2 Results j The excess reactivity was found to be 0.10 %Ak/k. From this and the !
. differential boron worth from the Physics Test Manual, it was ;
determined that the all rods out critical boron concentration was 2105 >
ppmB. Since the predicted value is 2056 ppmB, the acceptance criterion ;
was met. ;
3.6 Moderator and Temperature Coefficients Measurements The temperature coefficient of reactivity is defined as the fractional [
change in the excess reactivity of the core per unit change in core '
temperature. The temperature coefficient is normally divided into two i components as shown below.
aT " OM + OD where: aT = Temperature Coefficient of Reactivity 5 og - Moderator Coefficient of Reactivity aD Doppler Coefficient of Reactivity ,
for this test, performed in accordance with SP-103, the temperature and i moderator coefficients at hot zero power were measured. Furthermore, an extrapolated hot full power moderator coefficient was calculated. '
Each of the above coefficients was determined for an all-rods-out (AR0)
I PT-lll, Section 3.3.1, Rev. 9 14
-p f.
control rod configuration,. with rod bank 8 remaining at 25% withdrawn.
Previously, these calculations were repeated with regulating rod hnks 5-7 ins rted. This all rods-in (ARI) calculation was eliminated because:
- 1. There is good agreement between predicted and measured coefficients.
- 2. . There is a very good correlation between the direction of predicted versus measured deviation for the ARO and ARI coefficients for a given reload cycle.
The acceptance criterion for the hot zero power temperature coefficients is:
The calculated within go.4x10'got Ak/k/'Fzero of thepower temperature value given coefficient...sh in the Physics Test
!- Manual.
The acceptance criteria for the hot zero power moderator coefficients' of' reactivity are:
The hot zero power moderator coefficiegt shall bg greater than -
j 3.0x10'4 Ak/k/'F but less than 0.9x10' Ak/k/* F L For the extrapolated moderator coefficient, the criterion is as follows:
The t full power moderator coefficient shall be less than zero 3.6.1 Method The technique used to measure the 532'F and 2155 psig isothermal' temperature coefficient at zero power was to first establish steady state conditions by maintaining reactor flux, reactor coolant pressure, turbine header pressure and core average temperature cons The reactor was maintained at a critical level between 5 x 10'gnt. amps and l 'the upper -zero power physics test current limit as defined by the sensible heat determination experiment. Equilibrium boron concentration was established in the reactor coolant system, makeup tank,and l pressurizer to eliminate reactivity effects from boron changes during the subsequent temperature swings. The Reactimeter and the brush l
recorders were connected to monitor-selected core parameters .with the reactivity value calculated by the Reactimeter and the core average temperature displayed on a two-pen recorder.
1" Reduced Physics Testing, Task Summary Report", document by Babcock and Wilcox, B&W document number 86-1164722-00, dated March 1987 for the gSP-103,Section3.6.5,Rev.9&W Owners' Group Performance Committee.
3 SP-103, Section 3.6.1 and 3.6.3, Rev. 9 4
SP-103, Section 3.6.2, Rev. 9 1
15 l
r ,
h.
?
Once steady state conditions were established, a positive heatup rate !
was maintained by adjusting the turbine header pressure set point. As !
the reactivity changed, the controlling rod group was moved as r
! necessary to produce an adequate intermediate range signal. After the !
core average temperature increased by about 5'F, coolant temperature l and reactivity were stabilized. This process was then reversed, and ;
core average temperature was decreased by about 10'F. Finally, after i stabilizing coolant temperature and reactivity, the core average ;
temperature was returned to its original yklue. The measurement of the temperature coefficient from the data obtained was performed by 6 dividing the change in reactivity by the corresponding changes in core temperature for a specific time period. l The moderator coefficient cannot be directly measured in an operating reactor because a change in moderator temperature causes a similar l change in the fuel temperature. However, since the moderator coeffi- l cient has safety implications, it is an important reactivity coeffi- !
cient. To obtain the moderator coefficient from the measured j temperature coefficient, a Doppler correction of 0.168x10*4 Ak/k/*F must be subtracted.
The extrapolated hot full power moderator coefficient is calculated '
based-on the hot zero power moderator coefficient. Added to this are j both a calculated control rod effect, a temperature effect, and a boron '
change effect, which include the Doppler and xenon effects. -l 3.6.2 Results l
The results of the hot zero power temperature and moderator :
L coefficients measurements are summarized in Table 3.61 along with the ,
I- predicted values which are included for comparison. In all cases, the measured results' compared favorably with the predicted values. All measured temperature coefficiepts of reactivity were within the accep-u tance criterion of 10.40x10' Ak/k/'F of the predicted value. In .
addition, calculation of the moderator coefficient indicates that it.is .
well within the requirements of the Technical Specification. i The extrapolated hot fulg power moderator coefficient, as given in ;
Table 3.6-2, is -4.507x10' sk/k/'F for all rods out. From this, it is ,
E seen that the acceptance criterion of a hot full power moderator coefficient of less than zero is met. '
1 ,
t l.
16 l
.c 4
]
[y -
r j- j 3.7 Control Rod Group Worths f The layout of the core showing the location of the control rod groups and the location of the 52 incore detector strings is given in Figure 3.7-1. The number of rods in each group and the reactivity control !
function of each group is listed below. ;
Rod Group No. No. of Rods Control Function 1 8 Safety }
2 8 Safety i 3 8 Safety i 4 8 Safety !
5 12 Power Doppler !
6 8 Power Doppler i L 7- 8 Transient i r 8 .1 Axial Power Shaping !'
Total a 68
-The control rod worth tests were run in accordance with Performance i Testing Procedure PT-ll2 to provide information about the reactivity ,
worths of banks 5, 6, and 7. The results are also used to verify rod '
worth curves used during plant operation. The tests.are subject to the <
following acceptance criteria: 1 The predicted reactivity worth of rod groups 5, 6, and 7 in the l Physic >
value.{ Test Manualtheshall Additionally, each total predicted be within 115%-
reactivity worthofofthe measu, i the sum of rod groups 5, 6, and 7 in the Physics Test Manual )
shall be within 110% of the measured total.2 l
3,7.1 Method L l Measurements of control rod group worths for groups 5, 6, and 7 were !
made during zero power physics tests using the boron swap method. This 1 L method consisted of setting up a deboration rate and compensating for i the change in reactivity by small step changes in rod group positions.
The continuous calculation of reactivity was made by the Reactimeter. .
The reactivity in percent millirho (PCM) from the Reactimeter and the :
rod position were recorded on a strip chart and magnetic tape.
3.7.2 Results The results of both the predicted rod group worths and the measured group worths are tabulated in Table 3.0-2 for a reactor coolant system l temperature of 532'F and pressure of 2155 psig. Comparison between
! measured and predicted rod worths shows groups 5-7 were within the s)ecified 15% deviation, and the sum of groups 5-7 was within 0.20% of tie predicted value. Therefore, all acceptance criteria were met.
Integral measured rod worth curves for control rod groups 5, 6, and 7 L at 532*F and 2155 psig are plotted in Figures 3.7-2 through 3.7-4.
o ,
i I PT-ll2, Section 3.3.1, Rev. 10 2
PT-ll2, Section 3.3.2, Rev. 10 ,
17 ,
h 3.8 Differential Boron Worth Determination
. Soluble poison in the form of dissolved boric acid is added to the moderator to provide additional reactivity control beyond that ;
available from the control rods. The primary function of the soluble ;
poison control system is to control the excess reactivity of the fuel
- throughout the life of the cycle. !
The differential boron worth was measured in accordance with i Performance Testing Procedure PT ll2. From PT-ll2, the following acceptance criterion must be met: ,
The predicted differential boron worth in the Physics Test '
Manual shall be within 115% of the measured value.1 .
3.8.1 Method !
The test measurements of the boron differential worth was completed at i reactor coolant conditions of 532'F and 2155 psig. The measured value was determined by summing the incremental reactivity values measured ;
during the rod worth measurements over a known boron concentration ,
range from 2087 to 1773 ppmB.
~
3.8.2 Results V .
L The measured differential boron worth was calculated as -8.0lx10 3
- l. Ak/k/ppmB, as compared to.the predicted value of -7.29x10 3 Ak/k/ppmB. >
i The deviation is therefore -8.9%, which is well within the acceptance t criteria of 115%. The results of the differential soluble poison worth i l measurements are tabulated in Table 3.8-1.
3.9 'E.iected Control Rod Worth Measurement l
In previous cycles, this test was - performed to verify the safety l analysis calculations relating to the assumed accidental ejection of L the most reactive control rod. From the existing B&W data base, there is good agreement between the measured and predicted ejected . rod worths. It has been determined 2 that the HZP ejected rod worth (ERW) test can be eliminated if the aredicted value of maximum HZP ejected rod worth is less than 0.8% A(/k. Since the ' predicted ejected rod worth from the Physics Test Manual is 0.197% Ak/k, the ejected control
. rod worth test has been eliminated from zero power physics testing for
- Cycle 8.
IP 2"T-ll2, Section 3.3.3, Rev. 10 Reduced Physics Testing, Task Summary Report", document by Babcock and Wilcox, B&W document number 86-1164722-00, dated March 1987 for the B&W Owners' Group Performance Committee. -
18
, n-
+. ,
< - ( Jl-. o a.g ,'l h.i! >
4
-3.10 Bioloaical Shield Survey A Biological Shield Survey was not done at hot zero power since _ no' i
, :n plant modifications were made which would invalidate the Biological !
T Shield Surveys made during the Initial Startup Testing Program.
f 3.11 Effluent and Effluent Monitorina )
No Effluent or Effluent Monitoring testing was performed as these [
systems have been performing normally since initial startup. '
Therefore, no further testing was required. -
e ;
3.12 Chemical and Radiochemical Tests ;
Chemical and Radiochemical testing was not performed during this start-These tests were conducted at initial startup and no plant 1
up. -
modifications have been made which would invalidate the results of !
those tests. ,
hm :
hi 1
4 e
r
~
19
.c r
I TABLE 3.0 1 ACCEPTANCE CRITERIA DEVIATION LIMITS BETWEEN MEASURED AND PREDICTED VALUES ,
Zero Power Physics Testing i
I Allowable Deviation Between Core Physics Parameters Measured & Predicted Values '!
o All Rods Out Boron Concentration 1 50 ppmB :
o Temperature Coefficient of Reactivity 10.4x10 4 Ak/k/'F ,
[
o Control Rod Worths l
Individual Group Worths (Groups 5, 6, & 7) i 15% i Total Group Worth (Groups 5-7) i 10%
o Differential Boron Worth i 15% !
l 4
9 s
5 I
20
, = _
a Table 3.0-2 SINiARY OF ZERO POWER PHYSICS TESTING RESULTS _
Zero Power. Physics Testing Physics Parameter (Referencei Units Measured Predicted Acceptance Criteria , Comparison l 1. NI Overlap (PT-Il0) decades '2.424 Overlap must be OK (2.424) greater than 1.0 decade j 2. Sensible Heat (PT-116) Amps NI-3 1.4x10-7 None NI-4 1.3x10-7 l
l
- 3. All Rods Out Critical ppmB 2105 2056 Measured value must OK (49)
Scron (PT-111) -be within 50 ppm 8 y
of predicted value e
- 4. Temperature Coefficient Ak/k/*F of Reactivity (SP-103)
ARD, 2087 ppm 8 +0.1804x10-4 +0.2247x10-4 Measuredvaluemugt OK be within 0.4x10-Ak/k/*F of the pre- '
dicted value
- 5. Mcderator Coefficient- Ak/k/*F of Reactivity (SP-103)
ARO, 2087 ppmB +0.3484x10-4 +0.3927x10-4 Maximum positive OK moderator coefficient mustbggreaterthan OK
-3x10- Ak/k F and less than0.9x10f*Ak/k/*F.
- ^'.
TABLE 3.0-2 Sul9%RY OF ZERO POWER PHYSICS TESTING RESULTS Zero Power Physics. Testing (Continued)
! Physics Parameter (Reference) Units Measured' Predicted Acceptance Criteria Comparison
- 6. Extrapolated Moderator Ak/k/*F Coefficient of Reactivity (SP-103)
ARO -4.507x10-5 Extrapolated hot full OK 4
power moderator coeffi-cient must be less than 0.0
- 7. Control Rod Group Worth % Ak/k (PT-ll2)
Group 7 -0.676 -0.705 Percent deviation of OK (+4.3)
Group 6 -0.664 -0.752 group wod.h must be less OK(+13.3)
Group 5 -1.187 -!.075 than 15% OK (-9.4)
Total -2.527 -2.532 Percent deviation of OK (+0.2) u total worth must be
" less than 10%
- 8. Differential Boron % Ak/k/ppmB Worth (PT-ll2) 1831.5 ppmB -8.0lx10-3 -7.29x10-3 Percent deviation must OK (-8.9) be less than 15%
Percent Deviation is calculated as follows:
i
% Deviation Predicted Value - Measured Value x 100 Measured Value f
3 O
m TABLE 3.2-1 NUCLEAR INSTRUMENTATION OVERLAP TEST RESULTS
-Zero Power Physics Testing Case. Source Range (SR) Indication Intermediate Range (IR) Indication Average SR Average IR Indication Indication Overlap
- Number NI-I (CPS) NI-2 (CPS) NI-3 (Amps) NI-4 (Amps). (CPS) (Amps) (Decades) 1 2x10+4 2x10+4 2x10-II 2x10-II 2x10+4 2x10-II 2.000 2 8x10+4 8x10+4 lx10-It' 1x10-10 8x10+4 1x10-10 2.097 3 2x10+5 2x10+5 3x10-9 3x10-9 2x10+5 3x10-9 3.176:
Average Overlap - 2.424 U *0verlap between the Source (SR) and Intermediate Range (IR) average indications is obtained using the
. following equation:
Overlap - [6 - Log (Average SR)] + [ Log (Average IR) + 11]
Source: PT-110, Rev. 15, Enclosure 4, Attachment 1, Console Values
TABLE 3.4-1 REACTIMETEr. AND DOUBLING TIME REACTIVITY COMPARISON i
Zero Power Physics Testing t
Doubling Average Reactivity Reactimeter Absolute Case Time (DT) from DTs Reactivity Error
1 145.0 -44.7 -43 4.0
?
2 75.7 +54.0 +54 0.0 l
- Absolute error (E%) between the Doubling Time reactivity (#DT) and Reactimeter reactivity (FR) is given by the following equation:
E(%)-100*l(#DT-#R)/(PR)l Source: PT-110, Rev. 15, Enclosure 5, Attachment 1 i
l l l )
L l' l
I l
24
~'
TABLE 3.6-1 COMPARISON OF MEASURED AND PREDICTED REACTIVITY COEFFICIENTS Zero Power Physics Testing Reactivity Coefficients (Ak/k/*F)-
Rod Control Rod Group Average. RCS Boron
'(Group Position, % Withdrawn) Temp Concentration Temperature Coefficient Moderator Coefficient Configuration 1 2 3 4 5 ~6 7 8 (*F) . (ppm) Peasured Predicted Measured Predicted ARO 100 100 100 100 100 100 80 25 532 2087- + .180x10-4 + .225x10-4 + .348x10-4 + . 393x1G'4 i
(
w ARO - All Rods Out m
l Source: SP-103, Rev. 9, Enclosure 1
,y-r N-,w-.w g p, Wg v- 4 -e--,y+ ' - w+wc-a - -
v -,p W esq9--s*Nm -w
_ww---@ m' -+_ w-mf,_ u ___e,t=c_.s-.,,_
w-___Ch-__mm-us'sm-_._mm,a_e..is __,m,,,mm .-.-_a_ma- ,._.mm_----mas _mM
?' l r
d TABLE 3.6-2. EXTRAPOLATED HOT FULL POWER MODERATOR i
! COEFFICIENT OF REACTIVITY j Zero Power Physics Testing l b-Predicted Hot Full Power Moderator ,
Rod Confiouration Coefficient of Reactivity (Ak/k/'F)
All Rods Out 4,507x10-5 L- ,
i
. Source: SP-103, Enclosure 2
?
i I
i P
P I
?
t 4
i 26
~.. .
- . - .-r=
r --- . v. ~
- ,,= . -m
~ "'
_. "y.
.- ~
_ l' ~;- " ~
D s y
~
1- -
_- g gp
_:% 4;_
yhf' 5
~
l
. TABLE-3.8-1 DIFFERENTIAL BORON REACTIVITY WORTH TEST RESULTS w? ,
Zero Power Physics Testing s + . . .
Control Rod Group:. Measured Average Delta Differential ~ Boron Worth (Group Position, % Withdrawn) Boron Boron Boron -Boron (uk/k/ppmB)-
' Conc.: - Conc. Conc. Worth --
1 2 3. 4 5 -6 7 8 ' (ppm) (ppm) (ppm) (Mk/k) Measured Predicted
_ -s 100 100 100 100 100 100 80 25.0 2087-100 100 100 100 0- -O O 25.0 ;1773: 1930. -314. -2.515 -8.01x10-3 -7.7 %20 Source: PT-112, Rev. 10,~ Enclosure 6 -
- . * * 'p yi- ,g. y*,uy-f y O= c r g- y gg
. _ _ _ _ . . _ _ _ _ _ ._ _ _ . . . _ . .. . ~ _ . . , . - .
+ ,
ll
- i
$ q .$
~~
-l NUCLEAR INSTRUMENTATI M DETECTOR LOCATIONS I q
j y in Pmp B ' .
Pury A orsG A -
1 NI.5 UCIC yi.3 I
Cic !
p NI . 7 UC1C .
a . ~ :
- 3*. .
. . .... . +
r
- e
., s -
ui ,
, 'e' . .
u , __ _ ,
a...
- e. w ..
, ?
4-NI I: 1 N1 6
Niid 2 NI 2' i:: y1- Pmp D - CIC pg Pmp C OTSG B PC PROPDRTl0NAL' CDUNTER SOURCE RANGE 1 DETECTOR Clt COMPENSATED 10N CHANB'ER INTERNE0 LATE BANGE DETECTOR ,
l 'UCIC UNCOBPENSATED 10N CHAMBER - PONER s RANGE DETECTOR 1;;
I Source: FSAR, Fig. 7-18 3 1 Figure 3.2-1 28'
t ,
, .3 y )
D
-1 1
NUCLEAR-INSTRUMEMTATION DETECTOR LOCATIONS
= 2l'-4* -
j OTSG A j i- ll 1
I .
,)!} O 'S Y' -
}
1-1- .
ij
., k ;;
N/ ,
Punp B s Papp A
.g,4 .
I
-NI-l 2- 4 NI-3 NI- n-7 ), a l
r /
e lt '
L. ._ F Y
)
i , 60 _ 00 s NI-
-6 .;
NI-4 -- !!I-2 5 l
.1 ll E t .g .l.
Pung D ~ 5
, Pung C 2 l
\ -i j [(
) l'C' J;. ;
%[\I> P
- ,f..,es' s t -) 4
- s'-i H' p 1
, 2
.H l \ o !
- s '
CISG B Source: FSAR, Fig. 4-3 y,,gIl'g -. ~
3 12'- 5 g .
- 11'-7 h , .
Figure 3.2-2
- 29
. u . l. . - - .
)
INCORE MONITOR AND CONTROL ROD MAP 1 1 2- 3 4 5- 6. 7: 8 9 10 -11 12- 13- 14 15 9 NI 9 NI-3 A
9 NI-5 CR)-8 CR6-1 CR) 1
- B- 5 E CRJ-E 3t512 CR CRJ.)
CR7 8 CR7-1 CR7-2
,g g- CU-7 tRS-11 CI CP CI CD-2 CR) 7 CR6-8 CR2-1 CR6-2 CRI-2 F q _
y y G y CRS-10 CP ,
CR1-8 CP CR CP CR6-7 CR7-7 CRS-7 CR2-3 CR7-3 CR6-3
-H N . N ]- & $
, CR5-9 CR4-6 CRA-6 CRA 4 CR4-3 . CR5-4 K ..
q f g CI CR CR2-5 CR6-4 . . . CRI-3 CR3 6 CR5-8 CR4-5 CR4-4 CR5-5 CR2 3
.M g g y [T q CR7,,-f CRZd ,_
CR7-4
.N g , tg n CR - -
CRS-7 CRS-6 j.d4 0
7 7 E W CRI-5 CR6-5 CRI-4 P g 9 N1-6
.9 NI ' R' ' q g 9 N1 4 9 NI-2 Q - Total Core Monitor
- Total Core & Symetry Monitor
{ Symetry Monitor z - Incore Monitor Nunber CRX-Y -Control Rod Y of Control Rod Group X Figure 3.7-1 30 4
. . . . . .i
4 . y 1
1" #
. rolRodgou 5 :
- Conp%?rtte khg- 1
> ,i
=4.30 '!
. l Integral Worth - !
- j. R (Eak/k) r o 0.0 1.187 i
!- 3.0 1.180 ,
. 8.0 > 1.152 !
13.0- '1.089-18.0 1.005 21.5 0.924 1.00 25.0 0.839 l 29.0 0.760- +
,32.7 0.685 3 37.0 0.609 41.5 0.541 "'
46.0 0.479
- 51. 8 .. 0.412
. 58.5 - 0.338 -
66.3 0.261 = -
74.5 0.181' 83.7 0.090 .t 9.N . 100.0 0.000 l l
, <3 p ;u >
t g- 5 t ... . . .
..g .
(
. i d-
- o.m . ,
- l j
P t
- 0.So .
T i..
~
1 q i m e,ge I- I I I o So e ao ao soo Rod Position (X Withdrawn) :
Source: PT-li2, Encl. 5 Figure 3.7-2 -
3; 31
~ - -
T 1:6 T a *=n I'
y' ,' > $
[_ + g .
9
- ? ^
Con 1!' Rod! np SL d a >
rel- t 1 8a i
1
.,- 0.00 l-
-t Inteoral Worth . --
gg' >
a'
- 0.0 .
Isak/k)
-C.664-11.0 0.640= I 26.0. 0.550 !
o- 37.0. 0.455, O'N * '47.5 0.361 '
58.5 0.268-
,' , -70.4 0.174, s 82.2 0.080
~100.0 0.000- !
- :I t
. 0.00 - t c1
.t L ;!
o.se - 3 a
. 4-
+4+j@
.ww j' s p .
,^ - x:
- 0.40 -
o'
- s. ' ;
r t e ,'
L cm -
- j 9.30 -
c >
H ..
s
- , e,ge , ,
.- i
.. w -
o
-i '
i i i i o ao a so ao - too i Rod Position (X Withdrawn) :
Source: PT-lia. Encl. 4 ,-
Figure 3.7-3 32 i
_ _ _ _ _ _ - _ _ _ _ _ _ . ,._--,s . .- , - - - -n-
n; .- -- .
r s ,
3
'l r
I 3, . .
~
'Conprol'Rodpotg7- 1
- WISatYZh -
i
- v 9.e9
. I Integral Worthi .
R '(Eak/K) I is1 0.0- 0.676- ;
11.0 0.667- M 24.5 0.596 '
32.0 0.532.'
- 0' "
39.5 0.461~ L N ,
47.0 0.389: l 55.5 0.314-n ,
63.0 0.248 -
71.0- 0.185-81.0 0.100 -
- 100.0 'O.000 k 0.e9 a .1,
~
e.ee .
4
- s. ,
, s c , ..
l' ,
y o.m. !
".i.--
~ ,
-o. 1 s :3 ,
}
,M '
e-Jb..' ,
1... . .,
- .Hc -- \
l 1 :
e; 9.e9 ,
4 ';
JL .
t 0.00 - i : i o ao e ao so too Rod Position (X Withdrawn)
Source: PT-ii2, Encl. 3' 33
,a ,
l l 4 l
J v
- 4.0 FULL = POWER ESCALATION TESTING Full power escalation tests, required by Section:13.4. Table 13-4 of the l CR3 FSAR,- were performed to verify the validity of the safety analysis 3 assumptions and,- therefore, the acceptability: of the core design, i Testing.was performed at' two major power plateaus; 40-75% (69.6%) and-95-100% - (96.7%) full power (FP).; In previous cycles testing was-performed at three major power plateaus, but one intermediate power.
' distribution test- plateau was eliminated as explained in the '
Introduction ~(Section 1.0).- These tests were completed in accordance- :
with Performance Testing Procedure PT-120.
t t
4.1 Turbine / Reactor' Trio Test l
In' previous cycles, Turbine / Reactor Trip testing -was not performed-during startup_ since no modifications were made which would invalidate -
the original testing results. However, since the generator stator was
, ; changed out during the course of the. outage, a single turbine overspeed-trip test was performed at appro .
All- acceptance criteria were met.$imately 20% full. power as per SP-419.
, s 4.2 Intearal Control System Test ;
L Since no modifications were made to the Integrated Control System (ICS)
L during this outage, no specific ICS testing was done during this start- 1 L up. E Minor adjustments were made to -the ICS under normal maintenance *
'. and calibration procedures.
L -
t
- 4.3- Unit loss of Electrical Load 1
L No Loss of Electrical Load testing was performed since no modifications were made which would invalidate the original testing results. +
L L
L r
.4.4- Unit load Transient Test No specific Unit Load Transient testing was performed since no- 1 L modifications were made to invalidate the results of previous tests.
No major problems were encountered during transient operations throughout the testing program.
l 1 SP-419, Sections 3.6.1 and 3.6.2, Rev. 7 I
34
~_
- 6 ; <
i l
4.5 Reactivity Coefficients'At Power Test-The purpose of this test was to determine reactivity- coefficients at
-100% full power, and to verify that they were conservative _with respect-
-to- the FSAR. . The following: coefficients were either measured: ori calculated from the data obtained:
- a. ' Temperature coefficient of reactivity, defined as the - fractional change' in the reactivity -of the core per unit change in . fuel and moderator temperature.
- b. Moderator' coefficient of reactivity, defined as -the fractional-change in the reactivity of the core per unit change in -moderator temperature.
c.- Power Doppler coefficient of reactivity, defined as the fractional change in the reactivity of the core per unit change in power.
- d. Fuel temperature Doppler coefficient of reactivity, defined as the Tractional change in the reactivity of the core per unit change in fI el temperature.
Acceptance criteria specified for the Reactivity Coefficients at Power =
' Test are listed below:
.l. -The moderator tem shall' be less
-positive -than 0.9x10'gerature Ak/k/*F atcoefficient of reactivity power levels below 95% - full _ power, less positive than 0.0x10'4 Ak/k/*F at powe 1evels . at or~above 95%
full ppwer, and less negative than -3.0x10'{ Ak/k/*F at rated thermal power 2.- The fuel temperature D 1
more n: jative than -0.90x10'gppler Ak/k/*F - coe,fficient of reactivity - shall be E4.5.1 Method Reactivity coefficient measurements were made during the power escalation test program at 100.0% full power.
-t Differential rod worth measurements were performed during the reactivity coefficient measurement in order to generate rod worth data for the ' specific test conditions. For temperature coefficients, average reactor coolant temperature was increased and decreased about
5'F and data recorded. For power Doppler . coefficients, the differential rod worth data was used for the calculation. The usual method is to~ increase and decrease power by- about 5% full power, but two _ channels were very near tripping so this method was not used. From the measured temperature and power Doppler coefficients, the moderator 1
and fuel temperature Doppler coefficients were calculated.
4.5.1.1 Differential Rod Worth at Power. The method by which the differential rod worth was determined at power is the fast insertion / withdrawal 1
PT-120, Section 3.3.7, Rev. 16, Tech Spec 3.1.1.3 2 PT-120, Section o.3.8, Rev. 16
' 35
~
s _
l s
-)
6,
'f :
. method..'In this measurement, the controlling rod group is inserted fc l
)' , six seconds, followed immediately by a withdrawal . for six seconds. - -
Since- the : total elapsed time is on the order off the- primary nloop a
recirculation; time,: the' moderator temperature effects are eliminated o;
+ and the'. reactivity - versus time is essentially -a combination- of the E effects- due to the control rod motion and the fuel power variation.- ,
x, ;
-Determination of the differential rod worth was then found by using the *
- measured reactivity and re1 positions, compensating the n data . by a-e predicted fuel power correction factor. The fuel ' power corection factor accounts for the time delay involved in fuel temperature change- i g during the measurement. ;
4.5.1.2 Temperature and Moderator Coefficients. The temperature coefficient of reactivity is defined.as the fractional change in the reactivity of_the-core per ' unit change in fuel and moderator temperature.- The temperature coefficient is normally divided into two components as 1 1 shown in equation 4.5-1. ,
a L
1 Where: aT = Temperature Coefficient of Reactivity.(Ak/k/'F) ,
E og - Moderator Coefficient of Reactivity (Ak/k/*F)L '
l oD = Doppler Coefficient of Reactivity (Ak/k/'F)
} ,
The moderator coefficient cannot be directly measured in an operating'
, . reactor because a change in the moderator temperature also causes a ;
similar change -in the fuel temperature.- Therefore, the moderator -
coefficient-' must be calculated using equation 4. 5 after the. '
temperature and Doppler coefficients have been determined. >
Temperature, moderator, and Doppler coefficients were _ theoretically ,
predicted as:shown in Table 4.5-1 using the distributed moderator and 4 fuel _ temperatures .instead of. the isothermal values which were used for-- ,
zero-power physics predictions. For_ these predictions, the normal mode f 'of operation. with critical- boron:and rod conditions was assumed-which set the-average core moderator temperatura to 579'F.
The measurement method used at power is to change the : reactor coolant temperature setpoint'at the reactor control station ~with the integrated e control system in-automatic-effecting.an approximate 5'F change in the ,
' ' eactor . coolant temperature.
. The reactivity change caused ~ by the _
t temperature. change of the core was measured by recording the change in-the position of the. controlling control rod group and converting this change. to reactivity using differential rod worth values measured during theitest. Prior to- running the test, steady state equilibrium '
xenon conditions, including a stable boron concentration and no 4
.significant control rod motion during the last 30 minutes prior to taking data, were required as prerequisite system conditions.-
The fuel temperature Doppler coefficient relates the change in core reactivity to a corresponding change in fuel temperature. A ,
theoretical prediction of the fuel temperature Doppler coefficient was !
made using the PDQ code with thermal feedback, and is presented in Table 4.5-1.
36 r
-- _ _ _ . , , - - .__4
^ The usual ~ measurement method used is to change the reactor power level 5%; full - power. This change' in power -level would be -initiated by _ '
. manually decreasing = the reactor power ~ at the reactor master control
< station. After: obtaining approximately ten minutes of steady- state data at the reduced power level, reactor power would be returned to the initial. power. The calculation of the power Doppler . coefficient' would use< the measured change' in the controlling rod group position converted to an equivalent reactivity value and the measured change-in ,
reactor power determined by using the normalized core AT,w' hich is the- '
primary side heat balance. Instead, due - to- two channels being very-
- near tripping, the differential rod worth data and associated power i changes were used to' calculate the power Doppler coefficient This was '
then converted to a fuel temperature Doppler coefficient by multiplying by;a theoretically derived factor, A%FP/A'F. q 4.5.2 Results J - The results of ~ the measured temperature and Doppler coefficients, and calculated moderator coefficient at . power are shown in Table '4.5-1 1 1
, which also shows the predicted temperature, moderator, and Doppler.
coefficient results, p
The calculated moderator coefficient is negative and therefore meets
- j. the acceptance criteria of being non-positive.
The results of the measured and predicted Doppler coefficient L of t reactivity are shown' in Table 4,5-1. The acceptance' criterion for the: ;
measured Doppler' coeffjcient is that the coefficient must be more j negative than -0.90x10- Ak/k/*F. Table 4.5-1 shows that.the measured-coefficient is below this ,value and that the acceptance criterion -is
-adequately-met.
--gi l
4.6 - Unit' Heat Balance Heat. balance calculations were performed using the ModComp computer.
The data' was taken in accordance with ' Surveillance Procedure SP 312A. ,
No modifications were made during this shutdown which would require -
i reverification of the heat balance calculation.
' 4.7 Core Power Distribution Tesi .
The Core Power Distribution ' test was performed at each power plateau (40-75% FP and 95-100%' FP) to measure the core flux and power distri- :
butions. These measured powers were then compared to the predicted .
values. The test data was also used to evaluate core performance and i the departure from nucleate boiling ratio (DNBR). Additionally, test results were used in the power Imbalsnce Detector Correlation test and
, the Incore Detector test.
37
r 4 4 1 The . test, performed in accordance with PT-120, is: subject- to the ,
N following acceptance criteria:
. ,, j 1
=c The core' power distribution and thermal-hydra 31c parameters' must be measured, . evaluated,; and deemed reasonable. ;
.The highest measured radial and total: peaking factors shall'not be~
greaterthan5%and'7.5%,.respectigely.ofpredictedvaluesatthe 40-75% FP and 95-100% FP plateaus. ;
r The minimum DNBR is greater than 2.01-and the measured worst case ~
maximum . linear heat rate (LHR) t1 19.2 kW/ft when.
m extrapolated to the next power plateau.ps than The extrapolated worst case minimum DNBR-is greater than 2.01 and
'I the extrapolated worst case maximum LHR-is less than 19.2 kW/ft, or if-the MLHR is greater than 19.2 kW/ft, the extrapolated:imbal--
ance falls ougside the- power imbalance trip envelopei as'shown in )
Figure 4~.7-1
? Continuous monitoring 'of the core power density. at 364 core locations a was accomplished using the' incore monitoring system.- This system.is ij comprised of 52 detector strings each - having 7 individual neutron j detectors. .These detectors' are equally spaced at seven . axial '
elevations in the center of 52 fuel assemblies. .This system is capable-of producing detailed core power distributions for either eighth core- 1 '
The output the 'incoree or quarter core symmetry conditions.
y detectors was connected to the unit computeri and corrected' ;of for :
H background, fuel - depletion and the as-built dimensions to' provide :
accurate outputs:of relative neutron flux. The computer output of the <
I corrected signals was used to develop core power ' distributions 1which L, provide power peaking information,necessary to determine DNBR and LHR.
.i L Implementation of the core power and core powerL imbalance safety 4 1
% limits,' in cterms of the reactor protection;setpoints, 'is - shown ini "
- Figure 4.7-1.
The .out-of-core nuclear instrumentation provides the core power cand ,
core power imbalance signals: to the reactor protection system, since '
i the incore monitoring system does .not immediately respond. to prompt J
. changes in- the core conditions.. The .out-of-core nuclear l instrumentation (NI) is shown in Figure 3.7-1 as NI-l through 8.
r
- 4.7.1 Method o Computer . printouts of the core power distribution and; thermal hydraulicsr conditions- were obtained after establishing steady state conditions' at the -required power level and rod configurations. . For
, this' test, the APSRs were maintained at a constant position 'and the. ;
<v ,
1'T-120, P Section 3.3.6, Rev. 16 2
m PT-120, Section 3.3.5,.Rev. 16 3
PT-120, Sections 3.3.2 and 3.3.3, Rev. 16 -
4 PT-120, Enclosure 2, Parts 5.0 and 6.2, Rev. 16 L p- y j 38
+ j
. . - . . - . - . ~.
axial incoreEimbalance was1 maintained within 2% FP of the imbalance
. identified in the Physics Test Manual.<
4.7.1.1: Normal Operating, Core Power- Distributions. Normal: operating equilibrium xenon core power distM butions were measured-in-this test.
Data was' taken at each power plateap 69.6% FP, and- 96.7%- FP. -Values were obtained over an eighth core for radial and- maximum total power peaking factors. These results were the.1 compared with the _ values -
predicted in the Physics Test Manual. Based on.the maximum calculated:
, 'and maximum' measured peaks, percent deviations were determined. The percent deviation was calculated using the following equation: .
' Calculated - Measured '
% Deviation = '
Measured x 100% EQ.(4.3-1)_
- 4. , ,
The deviation was calculated for.the radial and for the maximum total
- power peaking factors at each power plateau. These distributions are ,
shown in Figures 4.7-2 through 4.7-5.
- 4.7.1.2 Worst Case Minimum DNBR. To maintain the integrity of the fuel cladding which prevents fission product release, it is necessary to prevent: overheating,of the cladding under normai peaking conditions.
The two primary core thermal limits which are indicativeL of_ fuel thermal- performance' are fuel melting and- departure from nucleate ,
boiling. These zlimits are independent; each must be evaluated - to ;
ensure core safety for.a given power peaking situation. Fuel melting 4 is basically a function of the local power generated in the fuel, which ,
H is a combination of radial and axial peaking. The maximum allowable linear heat rate limit is 19.2 kW/ft.
The-upper boundary of the nucleate boiling region is called " departure-
.from nucleate ' boiling" (DNB). At this point, there is a . sharp -
reduction of the heat transfer coefficient,. which would result in high l
cladding temperatures and the possibility of. cladding failure. :The-local DNB ratio-'(DNBR), defined as the- ratio of the -heat fluxJ that I would cause_ DNB.at a particular core location to the actual _ heat flux 1
at that location, is indicative of the mar _ gin to DNB.
Flow, temperature, and pressure. can be related to DNB through the use-of the B&W-2 or- BWC correlation. The B&W-2 and BWC correlations have
"" been empirically developed by Babcock & Wilcox to predict DNB~and the l location of DNB for uniform and non-uniform . axial heat flux J h distributions. The minimum value of the DNBR during steady-state
[ operation,Enormal operational transients, and anticipated transients is I
limited to . 2.01. A DNBR of 2.01 corresponds to 94.5% probability at i 99% confidence level that DNB will not occur. This is considered a p conservative margin to DNB for all operational conditions.
The worst case minimum DNBR values were calculated by the unit computer for each core power distribution that was taken during the power escal-ation test program.
Two normal operating equilibrium xenon core power distributions, as required by the Core Power Distribution Test, were obtained during the S. 39 E _ _ _ _.
o y
[
1
~
power escalation sequence. .Each - power distribution was' subjected to d the following- analysis. From each core power distribution, -the-worst !
caseLmeasured minimum DNBR..was selected. These -DNBRs - were then a extrapolated to the overpower trip setpoint and corrected for axial--
peak- location- and magnitude. Verification of acceptable ; corer ;
conditions at .the present and- next" power level of escalation was then a performed based on the acceptance criterion that the extrapolated DNBR F must be greater.than 2.01. The extrapolated DNBR: values were ~ supplied by, the plant computer's PD0s based on actual conditions at-the plateia.
Next, the: measured worst case minimum DNBRs were extrapolated to the ,
Loss of Coolant Accident and design overpower power level and corrected for axial peak location and magnitude.
4.7.1.3 Worst Case Maximum Linear Heat Rate Determination. Worst case maximum- !
linear heat rate (LHR) values were calculated using SP-104, " Hot Channel Factors Calculations", for each standard core- ' power-l:
distribution taken as part of the power escalation test program.
4.7.1.4- Quadrant Power Tilt. Quadrant power tilt limits have been established >
in the Core Operating Limits Report. These limits, when used in >
conjunction with the control rod position limits, assure ;that the.
design peak heat rate criterion is not exceeded during normal power ;
operation.
Quadrant power tilts are subject to the following acceptance criteria:
L The'. quadrant power tilts shall ngt exceed the limits specified in g the Core Operating Limits Report.
I Quadrant. power tilt is defined by the following _ equation'and expressed i
.in percent:
t Power in Any Core Ouadrant
< -Quadrant Power =* Average Power of all. Quadrants
._3 Ex'100% EQ.(4.3-2)
During the .startup testing program, ' maximum quadrant power tilt was -
determined' using the corrected signals from the 16 symmetric incore monitoring assemblies. Figures 4.7-2 through'4.7-5 include the maximum quadrant power til_t for each-standard core. power distribution taken as required by the Core Power Distribution Test..
'4.7.2 Results
, 4 lj 4.7.2.1 ' Normal Operating Core Power Distributions. The results of the core power . distributions are snown in Figures 4.7-2 through 4.7-5. .The percent deviation for each symmetric length core assembly is- also a given. The acceptance criteria for this test were met.
,o
'4.7.2.2 Worst Case Minimum DNBR. The results of various worst case minimum .
DNBR ' values calculated at each test plateau under normal rod configurations are plotted in Figure 4.7-6. These results indicate I PT-120, Section 3.3.4, Rev. 16, and Tech. Spec. 3.2.4 40
, q 1
1
[
that all'. measured values 1were above the' design worst case minimum DNBR.
.versus power level,. and= well above the minimum acceptable value. of 2.01.
4.7.2.3 .W orst Case . Maximum Linear, Heat Rate. In all cases, the- acceptance- !
criterion limit: of 19.2 kW/ft was_ met during the power escalation. test i program. These results are summarized .in Table 4.7-2. l
-4.7.2.4- Quadrant Power Tilt. Technical. Specification tilt limits were not- r exceeded at. any of the power plateaus. j
-1 4.7.2.5 Axial Power Imbalance.- Results from standard-core power distributions taken at 69.6% FP- during the performance of -the Power Imbalance . <r Detecto_r Correlation Test show that the -imbalance trip envelope (Figure 1 4.7-1) - of the reactor protective. system- is sufficient to- protect the >
unit from exceeding the DNBR and the LHR limits under all ~ core ,
imbalance conditions. -In addition, analyses indicate that the clargest thermal margins (measured by DNBR and LHR) exist when ainegative 10.0% .;
to positive 5.0% incore axial offset is present as shown. in Figure-4.12-3.
The core imbalances measured in conjunction with the core power distri-bution of this section are included in Figures 4.7-2 through 4.7-5.
9 '4.8 Bioloaical Shield Survey A Biological Shield Survey was not conducted as part of this power testing program - since no plant modifications were made which would
-invalidate the Biological Shield Surveys made -during the initial <
m startup testing program.
l i
4.9 Pseudo Rod Eiection Test t
in. previous cycles, the Pseudo Rod Ejection test. was not performed during the power esc.alation test program sincethis test had been
. performed during zero power- physics testing.- Since the maximum 'HZP -
ejected rod worth from the Physics Test Mgnual is less than 0.8 %Ak/k, it was unnecessary to perform these tests. ;l t
1" Reduced Physics Testing, Task Summary Report", document by. Babcock and Wilcox, B&W document number 86-1164722-00, dated March 1987 for the B&W.0wners' Group Performance Committee.
.41
~ - . . __
4 y_
4
~
4 A5 7 5
- 4.10 Shutdown From Outside the Control- Room No-Shutdown from Outside the Control Room testing was performed since no modifications were made which would- invalidate the results of 1
g testing during the initial start up. t 4.11- Loss of Offsite Power No Loss' of-Offsite Power testing was performed since no modifications j were- made which would invalidate the results of _ the tests performed during initial startup. ,
L$ 4.12 Power Imhalance Detector Correlation Test 4
Power -imbalance is defined as;
- Imbalance = % Power in top of core - % Power in bottom of core -
D EQ. (4.5-1) 3 1
The imbalance of the neutron flux in a- reactor results from temperature distributions,. ~ fuel . depletion, xenon oscillations, or control . rods-
!?"
- " positioned in the core. The amount of imbalance allowed in the core so '
that the DNBR or LHR limits are not exceeded is set in the reactor .
- protection-system. . The imbalance is a function of the power level and s the reactor coolant flow. Since this' imbalance is determined using j ' input signals from-the out-of-core detectors, it is essential that they 1
- are calibrated to read the true imb'alance as determined from the incore:
M -
- detectors, y, -
r The Power Imbalance Detector Correlation- (PIDC)' test was performed in accordance -with Performance Testing Procedure PT-120. The PIDC was done to determine the relationship between core offset as indicated by Ef the out-of-core power range NI.. detectors and core offset - as indicated
- by the full incore monitoring system. Offset is the imbalance divided -
. by the total core power. The PIDC also verified' that an acceptable
- f relationship between the backup and - full incore monitoring systems
- offset was observed. Finally, the-- PIDC verified that' the measured ,
worst case minimum DNBR-and maximum LHR were acceptable. These values :
were verified to. assure that the plant operates within the assumptions
- made in the safety analysis calculations. ,
w , ,
M i ' .
42 t " s
r gj,,
< m
.i T ,
j .
i .Three1 acceptance -criteria are specified: for the Power Imbalance Detector' Correlation Test:--
[
l!
The measured correlation between each out-of-core ' detector offset-J to full incore monitoring system offset lies within the acceptable ;
region shown in Figure 4.12-1. The correlation slope shall . be 1
between 0.96 to 1.10.2 j The measured . relationship between :the full incore- monitoring. 4 s stem offset and' the backup incore recordeg offset lies within-- ;
t e acceptable region shown in Figure 4.12-2 j
' ThemeasuredworstcaseminimumDNBR:isgreaterthan2.gliandithe measured worst case maximum LHR is less.than 19.2 kW/ft 1 1
q
[4.12.1 Method m' ~
During the power escalation sequence at Crystal River 3, imbalance measurements were made to determine the acceptability:of the out-of-- 'l
- core. detectors' ability to measure imbalance and to' establish ~ a basis I
for 4 verifying L that DNBR and LHR limits would - not be exceeded L whiles i operating within the' flux /(delta flux)/ flow envelope set in the reactor j J protection system. These imbalance measurements were made at-69.6% FP. !
In performing the test, the APSRs were positioned to obtain the desired. R
. full incore imbalance with reactivity compensations made by control rod ;
groups and/or 7. - At 69.6% FP,- the offset indicated by the full ;
incore , -system, out-of-core' system, and backup recorder system was -
recorded, and is shown in Table 4.12-1.
Based upon previous startup experience, the relationship between incore
] '
offset and out-of-core offset was determined to be-a linear' equation of1 the form below:
0C0 = (M x-100)'+ B EQ.-(4.52) .
Where: OCO -'Out-of-Core Offset (Percent) ..
1 100 --Incore Offset (Percent) '
M - Slope' of Relationship B = Intercept, when Incore Offset - 0 u
y, f
, s I PT-120, Section 3.3.9, Rev. 16 2 PT-120, Section 3.3.10, Rev. 16 3 PT-120,. Sections 3.3.2 and 3.3.3, Rev. 16 43
{Jy ,
+
m
.i y' :
The experimental slope ~ and intercept could then be = obtained using a linear least squares fit from the data obtained. If the measured slope-of the relationship of 100 to 000 is outside the range of 0.96 to 1.10, then: the gain of the out-of-core power range detectors would: be ,
adjusted.- The ' relationship of measured slope to : gain factor is - as follows.
GF = (M2/M1) x GFo. EQ. (4.5.3) j
.t Where: .GF = Desired Gain Factor !
M2 - Desired Slope ,
M1 = Measured Slope i GFo - Present Gain Factor' ;
3 Verification of the adequacy of the - power imbalance , system _ trip _ !
setpoint was performed in conjunction with the worst case analysis on each minimum DNBR and maximum LHR measured. .Each measured point was extrapolated to the power / imbalance / flow envelope-boundary limits given t in Figure 4.7 '1. In this: way, the adequacy of the imbalance system trip setpoints : to= protect - the unit from exceeding thermal-hydraulic !
limits could be verified. t 4.12.2 : Result:
I The measurement of the offset correlation function between' the full incore system andieach .out of-core detector ~ was determined during imbalance scans by APSRs snd control rod group 6 and/or 7 at 69.6% full <
power, to be a linear relationship on ~a ll power range detectors.
Figure-4.12-1-_shows the average response in offset between the out-of-core power range ' detectors and the full incore system' . Test data indicated that all-measured effsets from the out-of-core power range detectors fell- within the acceptable areas of- the curve during -the performance of the test. For each power range detector, a linear least squares fit was : applied toithe measured data points to- obtain a value E for the slope and intercept of the observed relationship. The- results' '
L of these- calculations are tabulated in' Table '4.12-2. In all; cases, the 1 measured slopes were in the. range of '0.96' to 1.10, which verified < the
-utilization of the gain factors used for the difference amplifiers; ,
L The ability of the backup recorder to follow full incore offset was L also verified as' part of this ' test by collecting backup recorder data and performing the necessary calculations. The results of this >
- analysis are plotted in Figure 4.12-2, and show that -the acceptance '
criterion was met. l
? During .all . phases 'of testing, worst case minimum DNBR and maximum LHR -
were recorded against incore offset. These results are_ given in Table 4.12-3 and Figure 4.12-3. The most limiting value observed for the maximum. LHR and the worst case minimum DNBR was 9.63 kW/ft and 4.365, respectively, which is well within the procedural acceptance criteria. ;
44
p o
/>((
' l. {
e jl ' t
<i 4.13 Nuclear Instrumentation Calibration At Power .
The . Nuclear' Instrumentation Calibration was: performed L"to verify _the ability to calibrate power ' range nuclear instrumentation to- measured rW core-conditions".3 This calibration was performed in accordance with =;
procedures PT-120 and SP ll3, and in- conjunction- with= Surveillance "
Procedure SP-312A.
The acceptance criteria for this calibration are:
The power range nuclear instrumentation (NI) 1s calibrated to- 1 7 within 12% full power of the heat balance power.2 :
~'
r The higg level bistable trip is set to trip within the specified limits.
,, The absolute difference between the out-of-core detector axial- [
power-imbalance and the in-cgre detector axial power imbalance-is within the specified limits
- O Furthermore, the Technical Specifications require that the overlap e
between the intermediate and power range nuclear instrumentation be in Y excess!of one decade. =l c ,4 C 4.13.1 Method J
1, Reactors power =was increased to the specified power level: while- a M continuosly. monitoring all the parameters that - indicate power level l
,, change. A' heat balance was then performed. Based on the results of '
L;
-the heat balance, the sensitivity of the linear amplifiers for_ each I. power range channel was adjusted', if .- necessary, and another - heat h balance; was performed. This process continued until indicated power and heat balance power were within 2% of. each other.
my l g' This calibration was performed at.each of the major power' plateaus; " 75% FP and'95-100% FP. t g
- p l4.13.2 Results i
(
t The results of the NI calibration met:the acceptance criteria.
1 m.. ;
1:l5 m.
0 4.14 Emeraency Feedwater Flow Test Lib .
L ,
No Emergency Feedwater Flow test was performed since no modifications
@ -were made ~which would invalidate the results of the tests performed 't o 'during Cycle 2 startup.
p 5 E I FSAR, Table 13-4, #13 2
, PT-120, Section 3.3.1, Rev. 16 3 SP-113, Section 4.3, Rev. 48 '
4 i 4 SP-113, Section 4.5, Rev. 48 y , -
45 L
v _ @.v , .e. . .
1 4 < 4.15' Turbine / Generator Operation-I TheJgenerator-stator stator was changed out during the course of thp outage, so:in order'"to verify proper operation: of turbine / generator" r chec All acceptance - criteria were: !
- met.gs were performed as per- SP.-325~.
7' l
y i
4 '.16 = Drooped Control Rod Test'
- The . Dropped Rod test was - not performed because sufficient thermal ,
margin exists as indicated by B&W 2 and BWC correlation analyses.;
p
[
4.17- Incore Detector Test The Incore Detector ' Test was performed to verify the adequacy of the system to provide a description of core conditions. The test ,- )
i performed in' accordance with PT-120, is subject to - the following- .
l acceptance criterion: ,
l Allt detector ' outputs must be consistent and reasonable.3
. 4.17;l Method i The .incore detector output' was verified by comparing : the corrected i detector response from similar core locations. All detector outputs
~
L were normalized to the average detector output per assembly - ,
The. values were taken ' from the Performance Data Output . at 69.6% FP. .
- From this, the average ' level " current for each detector, including l background, was found. Then, each current for each ' level was' divided 1by the average current. This process was repeated for each detector string. The groupings in Table 4.17-1 were made based on symmetric or 7
'near-symmetric locations. .
- 4.17.2 Resulti [
.The results of this comparison at 69.6% FP are shown in Table-4.17-1. ,
w I FSAR, Table 13-4, #15 2 SP-325, Section 3.6, Rev. 28 ,
3 PT-120, Section 3.3.6, Rev. 16 t- ,
.______E______________ _ _ -- . - .
^
o q l
[ ,
4.18- Reactor Cool' ant System Hot Leakaae Test !;
- The. Reactor Coolant System -(RCS) Hot Leakage is monitored on a regular :
basis .during- plant operation. as required by Technical Specifications.'
No. additional. RCS leakage testing was performed at this time..
a 4.19 Pioe and Component Hanaer Hot Insocction at Power No' Pipe and. Component Hanger Hot Inspection at Power was done:during i this startup since: no modifications. had - been ~ made which would' invalidate the results of the testing conducted at initial startup. l
[
4.20' Chemical and Radiochemical Tests Chemical and Radiochemical testing was not performed during this- .
startup. These tests' were conducted at initial -startup and 4 no plant (
modifications have been made - which would invalidate. the results of -
-those tests.- ,
4.21- Effluent and Effluent Monitorina No Effluent and Effluent - Monitoring testing was performed - as these systems- have been. performing normally since initial startup.
Therefore, no;further testing was required. .
k I
[
] .
l 47 -
N - . _ , . - . . ,
37 2, . .
, -TABLE 4,5-l?
SUMMARY
OF TEMPERATURE, MODERATOR, AND DOPPLER '
COEFFICIENTS OF REACTIVITY' !
Full Power Escalation Testing i
t 1
m n<-
-. ACCEPTANCE :
COEFFICIENT- MEASURED CALCULATED REQUIREMENT i
-0.6276 a- -0.579,0 --
l Temperagure .
a '(10' Ak/k/'F) .
-0.7805;a -0.428c Less than zero, and ;
Moderatgr i
~
P ~(10" l Ak/k/'F) lessneggtivethani L
-3.0x10- Ak/k/* F j
-1.1095 b -1.53 d~
E Fuel ~Tegp. Doppler Morenegagivethan- ,
' (10' , Ak/k/'F) -0.90x10' ~Ak/k/*F_-
an b Source: APT-120, Encl. 10; Rev. 16 '
L .bPT-120, Encl. 9, Rev. 16 ,
cPhysics Test _ Manual for CR3- C8, Table 12 ;
d '
Reload Report for CR3 C8, Table 5-1, BAW-2102 5,1 ' ,
y .
b r
_h,
?' '
,i 4!h t .'. e of r I
- ? - ,
48 i
.t . - . .
- A
h ..i' n; . v^.
t
.{,
TABLE 4.7-2 MAXIMUM LINEAR HEAT RATE BY INCORE DETECTOR LEVEL-y Full Power Escalation Testing Maximum Linear Heat Rate -
In-Core _ (kW/ft) i Detector LOCA Limit. .J Level- 69.6% FP 98.2%'FP ,
8 6.39- 8.85 15.0 7 '7.52 lo.40 15.4 '
- 6. 8.14 11.02 -16.2-5 8.23. 11.15 16.1 4 8.09 11.09 16.1
- -3 7.79 ~10.97- '14.9 2- ~7.25 10.46 14.5. 4 1 5.49 8.25 14'5 i4 Source: SP-104,. Encl. 1,,Rev. 25-. .
1;)
- n. y i
7 9
- c g f i
t . . :- i , .
A t
[ ['
4 lj 1
t.
i 'I H
i l
c ?: 49
+ , .
- ~ .. . .
, = . , wi - _;- - --
~
g- , , ,
"I
+ -
cTABLE.4.12-1'
SUMMARY
OF. POWER' IMBALANCE.~ DETECTOR CORRELATION TEST
~
~75% FP.
Full, Power Es'calation. Testing :,
l l
Rod Position'(% WD)a Incore _' Offset '(%) Out-of-Core Offset (%)b Worst Case Worst' Case Power Level b MLHR- Mini -
l (% FP) 1-6 7. 8 Full b- Backup c NI-5 -NI-6 NI-7 ~NI-8 (kW/ft)d DNBR -
I 69.70 100 88.13 29.89 -0.73 0.0 -1.99 0.48 1.30 -0.13 8.62 4.7%
i- 69.61 100 89.65 ' 26.32 +1.049 1.893 0.39 2.87 2.72 2.29 8.74 4.593 69.77 100 89.35' -11.16 +6.2 4.73 5.03 7.39 7.19 6.76 9.21 4.460 69.88 100 87.18 0.62 +7.78 6.0 6.15 8.52 8.52 7.81 9.47 .4.365 69.74 100 -85.27 46.63 -4.97 -4.03 -5.99 -3.47 -3.47 -3.78 8.845 4.750 69.75 100 67.29 '46.61 -19.39 -17.87 -21.36 -18.75 -18.~75 -18.20 9.455 4.585 69.59 100 65.52 50.35 -21.98 -7.0. 04 . -23.82 -21.11 -21.35 -20.43 9.63 4.482 o
Sources: a PT-120, Rev 16 b PT-Its, Encl. 6, Part.4,'Rev. 16-c PT-d py_120, 120, Encl.Encl. 4, 7,'Part:1, Part 2,1Rev. Rev. 16 16
(
- ._ _ - s _ _ , - -
s ,. -- . ~ , . . -- r - :. .- _. -. _ . . _ -
_..___-___.____..____.___.m.-_
, / 1 '
i
+
~
,', , , i
' LTABLE!4.12-2
SUMMARY
.0F-LEAST-SQUARES LINEAR REGRESSION ANALYSIS OF PGWER RANGE CHANNELS AND BACKUP RECORDER DURING POWER IMBALANCE 4 b CORRELATIONLTEST-~ ' ;
75W FP Ea Full Power Escalation Testing- :
-I E :
n, _, ,
, Intercept Slope-(m) ,
w' ' '
. Detector . Data .
y System Sets Calculated Allowable Calculated Allowable - 1
- a muummmuummu -
w T Out-of-Core '
NI-5 l
1.2528 12.5 j
7- 1.0243-- 0.96smsl.10- ,
p NI-6 7 1.2031 12.5 1.0139 0.96smsl.101 i
.NI-7 7- 1.259 12.5 1.019 0.96smsl.10 j g NI-8:
'~
~7 0.7482 12.5 -0.9649 0.96smsl.10:
j Q . Backup b 7 1.31 .0.92 in., Sources:.. a PT-i?O, Encl. 6, Part 4, Rev. 11 :[
m, ,
b:PT-120,-Encl. 7, Part'1, Rev. 11 g i
~ /
I I 1
7
- 1
$p .[
U 1
+ _
~
. ;!~
1 f
'i r
k "E :
z .
l a
tQt 51 i.)[ [;. . .
V '
l
{
TABLEJ4.12-3 WORST CASE MINIMUM DNBR AND MAXIMUM LINEAR HEAT RATE' l VS. FULL INCORE OFFSET" :
C 75% FP Full Power Escalation Testing 1
' Worst Case Maximum LHR Worst Case inimum DNBR ~!
Full Incore ,i
.g g g Value Location
.-0.73 ;8.62~ 0-3 4;796- 0-4
+1.049 8.74 D-3 4.698- 0-4 -
+6.2 9.21 D3 4.460- 0-4 ;
-+7.78 9.47 D-3 4.365' 0-4~
-4.97. 18.845 0-3,0-4 4.750- -D-3 ;
'-19.39 9.455 0-4 4.585 N-13
-21.98 9.63 0-3 '4.482 0-4 '
e Source:. PT-120, Encl.~4, Part 2, Rev. 16' l
i l :.
I l
l a
1 52 r>
f j,:
~
~ ~ '
' ' ~
~, _ ,
- m. -
TABLE'4.17-1 COMPARISON OF INCORE MONITORE'D. ASSEMBLIES' FLUX SHAPES-p ..
s.. -.75% FP <
. Full-Power Escalation Testing-Incore. _
(Detector Level Current)/(Average Detector' Current) Back-Group Detector Detector . .
. ground j Number Loc. tion Level 1 Level'2- Level 3.- Level 4 : Level.5 Level 6 Level 7 (Amps) 01 5' E-09 0.863 1.067 1.165 1.063 1.170~ 1.042 iO.629 63 7 E -0.848 1.081 1.138 1.137 -1.143 1.032 0.620 74 -
! 9 G-05 0.854 1.093 1.141 1.131 1.130 1.036 0.617- 59-11 K-05 0.854 1.062 1.134 1.158 1.130 1.049 0.612 73 13 M-07 0.853 1.097 1.126 1.149 1.122 1.032 0.620 69 t 16 M-09 0.859 1.071 1.140 1.156 1.124 1.043 0.608 69 l 19' K-11 0.868 '1.097 1.120 1.159 1.124 1.028 0.603 74 25 G-11 0.854 1.094 1.139 1.162 1.121- 1.031 0.600 58 02 23 F-13 0.788 1.089 1.134 1.166 1.129 1.077 0.616 -7 I
m 28 C-10 0.797 1.071 1.154 1.189 1.148 1.043 0.599 57- .
l u 32 C-06 0.807 1.062 .1.148 1.180 1.153 1.046 0.604 55 '
35 F-03 0.798 1.065 1.119 1.118 1.114 1.016 0.770 66 39 L-03 0.821 1.058 1.152 1.181 1.162 1.018 0.607 56-43 0-06 0.765 1.117 1.145 1.172- 1.150 1.040 0.611 55 47 0-10 0.802 1.086 1.152 1.162 1.144 1.033 0.620 55 50 L-13 0.901 1.087 1.149 1.173 1.153 1.046 0.592 57 03 6 F-07 0.800 1.044 1.120 1.159 1.147 1.075 0.654 50 8 G-06 0.805 1.058 1.134" 1.143 1.146 1.052. 0.662 68 15 'N-09 0.849 1.075 1.133 1.140 -1.143 1.054' O.607 64' "
17 M-10 0.846 1.080 1.118 1.137 1.128 1.049 0.642 58 18 L-11 0.850 1.087 1.124- 1.146 1.129 1.030 0.634 70-
. 20 K-12 0.855 1.089 .l.140 1.162 -1.137 1.029- 0.588 69
! 33 D-05 0.839 1.096 1.162. 1.162 1.141 1.037 0 .563' 69 34 E-04 0.810 1.091 -1.151 1.161 1.144 1.036 0.608 72 04 24 ~F-12 0.888 1.162' 1.123 1.122 1.092 -1.015 0.598 ?72 27 D-10 0.852 '1.121 1.090 1.132 1.099 1.075 0.632 60 05 31 B-07 0.734 1.048 1.138 -1.187 1.215- 1.096 0.582 18 36 G-02 0.719 :1.035 1.086 -1.108 - 1.103 1.021' 0.928- 157
- - .- ~
a-- ~ - . , - -- ., - - - - . . . . . .. - - = - . .-
u y c -
_ g -- c -
_c
.y'.
m
. TABLE:4.17-1 COMPARISON OF INCORE-MONITORED ASSEMBLIES'. FLUX SHAPES-75% FP
? Full Power Escalation Testing (Continued)
Incore ;(Detector. Level Current)/(Average Detector Current) Back-Group Detector Detector .
ground Number Location Level 1 Level 2 Level.3 Level 4- Level 5' Level'6 . Level 7 (Amps) 06 '38 L-02 0.810 .l.086. '1.193- 1.157 1.142 1.047 0.564 .59 44 P-06 0.771 1.072 1.158 1.197 1.151 1.024 0.628 11 07 22- G-13 0.818 1.094 1.157 1.168 1.147 1.037 -0.579 61 29 'C -
0.797 1;086 1.152 1.188 1.144 1.039 P.593 75 08 40 M-03 0.811 1.091 1.153 1.177 1.141 0.993 0.634 0 42 0-05 0.789 1.079 -1.157 1.175 1.155 1.046- 0.600 :17 .
09 1 H-08 0.734 1.064 1.134' 1.125 1.119 1.079 0.746 51 2 H-09 0.788 1.070 1.134 1.137 1.137 1.072 0.662 60 E 3 G-09 0.812 1.060 1.165 1.148 1.129 1.050 0.637- :76 5 10 4 F-08 0.835- 1.065 1.140 1.142 1.126 1.061 0.631 82 10 H-05 0.830' 1.053 1.124 1.161 1.161 1.042 0.630 0 12 L-06 0.801 1.016 1.125 1.145 1.138- 1.076 0.699 46- ,
14 N-08 0.869 1.092 1.158 1.182 1.185 1.071- 0.443 -31 11 37 H-01 0.742 1.016 1.111 1.187 1.173 1.115 0.655 0 45 R-07 0.756 1.084 1.118 1.154 1.183 1.055 0.649 2 !
51 D-14 0.738 1.085 1.180 1.217 1.164 1.057 'O.559 4' 52 C-13 0.738 1.082 1.179 1.203 1.231 1.050. 0.517 9_
12 26 E-11 0.849 1.121 1.125 1.158 1.119 1.024 0.603- 76 30 B-08 0.782 --1.071 1.154 1.187 1.157 1.045- 0.605 4 41 N-04 0.828 1.106 1.207 1.193 1.199 1.074 0.393 15 4 13 21 H-13 0.816 1.088 1.144 -1.191 1.148 '1.033- ~0.579 0:
46 R-10 0.742 1.037 1.123 1.173 1.168 0.912 0.846 ~0 48 0-12 0.713. 1.085 '1.168 1.213 1.219 1.062 0.541 3 49 M-14 0.732~ 1.086 1.164. 1.175 1.205 :1.078 0.560 3 Source: Group 1,,(BKGD1, SPDNRI)
_.__ L _ __ - _-
~
- - . a -- ,-- -. .- a _- . _ _ - _
-.---..,....,, -.--~ -- - . - - ----- - - --
c 7
> ?
l 1
Trip Setpoint for Nuclear Overpower Based !
on RCS Flow and Axial Power imbalance :
Full Power Escalation Testing l f
110 (17,16). "
_ (17,108) i Acceptable M2 ' -1.833 43 = 1.0 4 Pump o 100 f Operation 7
(-34.7,90.3)i - a 90 l
(-17,80.67) (17,80.67) i 50 t
' (35,75)
" 70 ;
Acceptable
(-34.7,62.97), 3 & 4 Pump
,g Operation l l
.h . ,- 50 ;
7 .
> (35,47.67) [
, , j - 40 j l
- l - 30
- g. . En j t .
E. 10 e , _ , , , , , , , , i 1
40 -30 -20 -10 0 10 20 30 40 50 ,
Axial Power Imbalance,1 Source: Toch. Specs.
Figure 2.2-1 i.e. ;
Figure 4.7-1 55
,3
If y sw
// COMPARISON OF MEASURED AND PREDICTED RADIAL POWER 7
// PEAKING FACTORS WITH 3D EQUILIBRIUM XENON AT 75% FULL POWER ,
Full Power Escalation Testing Measured Predicted l
Control Rod Group Positions Groups 1-6 100.0 %WD 100.0 %WD ,
Group 7 87.1 %WD 90.1 %WD Group 8 29.9 %WD 30.4 %WD ;
Core Power Level 69.6 %FP 75 %FP ;
Boron Concentration 1654 ppm -
Core Burnup 2.6 EFPD 3.0 EFPD Axial Imbalance -1.00 %FP -0.98 %FP Maximum Quadrant Tilt 1.69 % -
[
H K L M N O P R 0.95 <
8 0.89-6.30 F 1.25 1.17 9 1.17 1.20
-6.40- 2.56 :
1.19 1.31 0.94 [
10 1.23 1.33 0.93 i 3.36 1.53 -1.06 i 1.23 1.16
- 27
. 1.19 l 11 1.30 1.24 1.30 1.23 5.69- 6.90 2.36 3.36 0.84 1.20 1.14 1.30 0.98 12 0.88 1.26- 1.16 1.27 0.95 ,
L 4.76 5.00 1.75 -2,31 -3,06 i 1.14 1.10 1.23' l.16 1.17 0.40 PT-120, Encl. 2,
- 13 1.20 1.15 1.23 1.16 1.08 0.40 Source:
5.26 4.54 0.00 0.00 -7.69 0.00 Parts 1 & 2 0.88 1.19 0.86 0.95' O.49 ,
i 14 0.93 1.18 0.87 0.88 0.47 .
5.68 -0.84' l.16 -7.37 -4.08 x.xx Measured Results x.xx' Predict'ed Results 0.41 '0.38 0.31 x.xx % Deviation
~ 15 0.32 0.32 C.26
-22.0 -15.8 -16.1
% Deviation = Predicted - Measured x 100 Measured Figure 4.7-2 56
11
- COMPARISON OF MEASURED AND PREDICTED MAXIMUM TOTAL POWER
!: PEAKING FACTORS WITH 3D EQUILIBRIUM XENON AT
- 75% FULL POWER o Full Power Escalation Testing l Measured Predicted Control Rod Group Positions Groups 1-6 100.0 %WD 100.0 %WD Group 7 87.1 %WD 90.1 %WD Group 8 29.9 %WD 30.4 %WD Core Power Level j9.6 %FP 75 %FP Boron Concentration 1654 ppm -
Core Burnup 2.6 EFPD 3.0 EFPD Axial Imbalance -1.00 %FP -0.98 %FP Maximum Quadrant Tilt 1.69 % -
H K L M N O P R 1.08 8 1.05
-2.78 1.42 1.35 9 1.36 1.40
-4.22 3.70 1.35 1.51 1.07 10 1.43 1.55 1.08 5.93 2.65 0.93 1,43 1.32 1.45 1.36 11 1.53 1.45 1.50 1.43 r 6.99 9.85 3.45 5.15 0.98 1.38 1.28 1.51 1.16 12 1.04 1,48 1.33 1.50 1.14 6.12 7.25 3.91 -0.66 -1.72 1.35 1.28 1.44 1.37 1.41 0.48 13 1.43 1.37 1.45 1,38 1.32 0.48 Source: PT-120. Encl. 2, 5.93 7.03 0.69 0.73 -6.38 0.00 Parts 1 & 3 1.04 1.38 1,01 1.14 0.59 14 1.12 1,42 1.04 1.07 0.57 7.69 2.90 2.97 -6.14 -3.39 x.xx Measured Results x.xx Predicted Results 0.48 0.43 0.36 x.xx % Deviation 15 0.38 0.38 0.31 20.8 -11.6 -13.9
% Deviation . Predicted - Measured x 100 Measured Figure 4.7-3 57
COMPARISON OF MEASURED AND PREDICTED RADIAL POWER L PEAKING FACTORS WITH 3D EQUILIBRIUM XENON AT i 100% FULL POWER Full Power Escalation Testing s
Measured Predicted Control Rod Group Positions .
Groups 1-6 100.0 %WD 100.0 %WD Group 7 87.2 %WD 90.1 %WD Group 8 31.4 %WD 30.4 %WD Core Power Level 96.7 %FP 100 %FP Boron Concentration 1543 ppm -
Core Burnup 5.6 EFPD 4.0 EFPD Axial Imbalance -3.33 %FP 4.11 %FP Maximum Quadrant Tilt 1,49 % -
H K L M N O P R 0.96 8 0.91'
-5.21 1.25 1.18 9 1.17 1.20
-6.40 1,69 1.19 1.31 0.95 10 1.22 1.32 0.94 2.52 0.76 -1.05 1.23 1.16 1.27 1.19 11 1.30 1.24 1,29 1.22 5.69 6.90 1.57 2.52 0.85- 1.20 1.14 1.28 0.98 12 0.88 1.26 1.15 1.26 0.95 3.53 5.00 0.88 -1.56 -3.06 1.15 1.10 1.23 1.17 1.16 0.40 13 1.20 1.15 1.22 1.16 1.08 0.40 Source: PT-120, Encl. 8, 4.35 4.55 -0.81 -0.85 -6.90 0.00 Parts 1 & 2 0.89 1.20 0.86 0.95 0.49 14 0.94 1.18 0.88 0.88 0.48 5.62 .t.67 2.33 -7.37 -2.04
. x.xx Measured Results x.xx Predicted Results 0.41 0.38 0.32 x.xx % Deviation 15 0.33 0.33 0.27
-19.5 -13.2 -15.6
% Deviation = E.redicted - Measured x 100 Measured Figure 4.7-4 58
c COMPARISON OF MEASURED AND PREDICTED MAXIMUM TOTAL POWER r
' PEAKING FACTORS WITH 3D EQUlllBRIUM XENON AT 100% FULL POWER Full Power Escalation Testing l
Measured Predicted Control Rod Group Positions Groups 1-6 100.0 %WD 100.0 %WD Group 7 87.2 %WD 90.1 %WD Group 8 31.4 %WD 30.4 %WD Core Power Level 96.7 %FP 100 %FP Boron Concentration 1543 ppm -
Core Burnup 5.6 EFPD 4.0 EFPD Axial Imbalance -3.33 %FP -4.11 %FP Maximum Quadrant Tilt 1.49 % -
H K L $; N O P R 1.10 8 1.08 L -1.82 1.42 1.36-l 9 1.36 1,40 l -4.23 2.94 1.35 1.50 1.07
'10 1.43- 1.55- 1.08 5,93 3.33 0.93 l
1.42 1.31 1.44 1.36 i
11 '.53 1.45 1.52 1.44 7.75 10.7 5.56 5.88 l
. 0.98 1,38 1.31 1,51 1.16
.12 1.05 1.49 1.35 1.51 1.14 7.14 7.97 3.05 0.00 1.72 1.35 1.28 1.42 1.36 1.39 0.47 13 1.43 1.37 1.46 1.39 1.33 0.49 Source: PT-120, Encl. 8, 5.93 7.03 2.82 2.21 -4.32 4.26 Parts 1 & 2 1.04 l'.36 1.01 1.12 0.59 14 1.12 1.43 1.05 1.08 0.57 7.69 5.15 3.96 -3.57 -3.39 x.xx Heasured Results x.xx Predicted Results 0.48 0.43 0.36 x.xx % Deviation 15 0.39 0.38 0.32
-18.8 -11.6 11.1
% Deviation . Predicted - Measured x 100 Measured Figure 4.7-5 59
I t
H0T CHANNEL MINIMUM DNBR VS. CORE POWER LEVEL Full Power Escalation Testing l
2$ . .,,,
........<........................................,u..............................................
. n.......... ... .... . ...........
0 .... .............. .... .............................................................................
..ic...........................................................................
........................r..........................................................
..................................................................r:.....................:.......
i
......................... 1 10 . ................... ..........................................
.......r........ ....... ,.................... ................ ........
c ..................................t.............................t............................................. .......... .. .............. ... .... . .............................
se u
i u
S .... .............
t .... ........................................... .... ........ ................... ,
... . .................. 2...............
........................................... t
.......................................................... ........................... ........r.................. .................
.......z.. ... .... ..<
0 ......................................................... ... .... .......................... ....... ,
t 0 500 1000 1500 2000 2500 ,
CorePowerLevel(MW) t l Source: Group 67 D k
f Figure 4.7-6 60 t.
.-_m._,.,_.m__.:_-.__m_. _ _ , , , . _ _ , , _ _ . , , , , , , ., . , , _ -c.g_. ,., ,._ ,p.,g,,,,__,,n.,,,p.,.p.,77__.,_,g7,..-,.%9.q w g 9,.mg_,9,,
_ m_ . _ , , _ ,,___,,__,__._,,__._.m , _ , , , _ , , _ ,
r
. rt ,
W' N j AVERAGE OUT-0F CORE OFFSET VS. FULL INCORE OFFSET 75% FP Full Power Escalation Testing l' 1 E -
0 E E E
, a E. L E h E L k k p g . M M M % % h h M X E s % s h g g h h M M h h h h I h h s g g
. . m m . . m m m . .
. . m ,
g L k E E E 'L E E L E E m E X ,
. . m - m m . - - m m . . m , ,
-m m .
m m
m m
m m
m m
m m
m m
m , ,m
. , , m l . . . . m . . . ,
(, 430 . m m
- - - m - - E - - -
m
. m m m m m m- - - m . % , , %
- M% W % m E E h w h E h M W n 1 m m m m m m m - m , r m I
L
. , , , , . . ,. ,. , , , ,. , s, k E E E E E h L h E E X M E y K W
- l. . .
m m
m -
K m
W K L M
A % g g W %
- . - .m m s l' R h h M % h h M h I ; M
- M M JF % M %
. m ,. , , , . . . . . , . -- - , ,
L L L E E E h h E E h . F , . E %
m .
. . . A
. m m s % m - 1 - ,
- m M M
. . m %
M % h M N h M i W E J w % D %
L L L E E E h E E h M , E m E W L W K % % % M m M A % % %
W w w w 1 W. A A W ' - M M mm w 1 1 M W M W % % W % % K M A W W , M %
, ,10 m
. . . . h - . _. h . . . .
g h 5 % % % % M M M A E i M I M h % A M X M 1 M M % M . M J Im F. M . M M % % M M W M K W W M M M M ' N . M W M A % %
E E k E h k E M
EE
.,- b .
h E
E E
h h h h M M M M . X ' 1 M - M M % % h
% % % % % % % M TA . V M I M m % % M %
s - s s s L h 1 w w , m I h h % . M %
0 .
m m m . . s m , - - - . m - m m m m . . . s . m W
m
, s,.
M E h h h h M M - M - E k h h %
. . m . A m , - - m . . m m m .
M A K W % %A VW % M M K % % M M w w w % A , A M . M M n K M % M M m
-10
. . E . . - . . .
s s s a s r m u
= = ' - ' - -
. . . . ' ,' m . Incore Avera e
,. . . . - . . . . Offset Out of- o.re
(%) Offset (1)
M M M , W M %
~DT M
-20 ' '
o.oD
'%- m - - ,- . % - =+1.o49 2.or -
', ', " " ' 4 +6.2 6.59
', =
+1.18 1.15 j
'."xs- ,
N.-. m -4,97 4.18 ,
. ,,._ A . -19.39 10.21
--T - ,' ,' '
,. =% -
-h
-21.98 -!!.68
-30 -
- . % . . . . W . . . . .
. , , . , %--h , , %, %
, ., %, s r
JK s
E E
- s. ,
M h
E h
E
% % 1
, M, , %, %, %, .
, %, %, %, A A
i 40 q
-40 -30 20 -10 0 . +10 +20 +30 +40 !
L FULL INCORE OFFSET, t 1
1 Source: PT-120 Encl. 6, Part 4 l
I l
l I
Figure 4.12 1 ;
61
,i 4 ,
BACK UP INCORE OFFSET VS. FULL INCORE OFFSET I 75% FP l Full Power Escalation Testing !,
I I
l-x x x x 60 x x x x < , i
'x s , . x --\ x . x x x x x ,, ,
x x N ,- x, x, ,
x x x x -
1 x x , s s x x x ,,
x x
x s
x s
- x. x _ ,2 x s
x 1
s t
x x
s x
s x
e:. r,. .
x 1
x x
x x
x x,,. ,
x x x ..
_x _.
. s s , x , , y x x Y x 't 't ( gf g -
(
s - s s s r, . 40 s x x , , s t
's t s s, 7 . 't x x; , x-- >
x x x s
, x,
, x x , x x, x x W -
x s 1,,,
, x !
g 3;4 s x x x x m s .
s s - s, < . _ s s , , x x x -
x, x <
x x
x, x. ,
x, x
x, x i-x, x - x , ,
t .
t x x x x , ,
~ ~ x
, x ,, , x x w ,
l s s s _., 3,. x , , x x x i x
x x
x x _ r <
x e r-
,isx x x
x x -
-A--- i X- --h- -- _
A x x -_, e . . :. . !
i
's 'sx 3,< /,
x 1 %... 5 $ ~t x
A s
's, N x 's-x -
t x x
\..N.--h, e - .
../_x x x x._ _ g--
0 to ~~ 30 5 40 -I 50 t 60 !
60C'-to x
x x
s i-401 c-s t
s
-30 i
s, s tob-#J---/- J s s N<s 'sx ix xx <
x
[C~h 'N, , ' ,
s x x ,
,rh, N . N OFF$ti. 5 . . . . - . . . . . . .
% -~~N
--% {,
,C 'yi
', ', ; 20 IA- L INCORE DETECTOR $ (
,s x < s, r ,'t t * * " '
s x, ,
't x 't , , x T x L
t x
x x
x ,
x
\-....*.
x x ,
s < i -
is, 't *30 <x s s Incore-Backup x,r x f f x ,
't A i 't h -b -- Y, W Offset Offset *
(: x , kx x '
x
(%) (%) t 1
'x '
r e
't 't 40
%-- 3c- 3 s
x
't x - ;
s e ._. a x- x ,
s 5 75- ,,
(__ .,
e4c._ -V,., .
N. . N . 3,o4g 3,8g3 - l
--\- y%.
-N A, - (--h-h l6.fs k'b ,
..y .
,f . 't
.3
, . -.g -
g ... 4.sr 19.39 4.o3 ,
-17.87
, ' ', Z ","c ~ . '
k" *.. t- '.~ A ~ .)-21.98 -20.04 l r
.g ,
Source: PT 120 l Encl. 6 and 7, Part 4 and 1 i l
1 l
l Figure 4.12-2 62
-i
_ . . . . . _ _ _ . . __ ~
s MAXIMUM LHR AND WORST CASE MINIMUM DNBR VS FULL INCORE OFFSET 75% FP Full Power Escalation Testing 4 . .
a a
+ . . .
... t
, ._.#... .. _. .-. .m ..~ -.. .. -
. ,. .. a . y
..i.
,6..
a.
3 7
. -.. . _ .,. , . I
. i.
.. -~ . -
- 1. + . . y
, . . . ..=
. ~ .l .
.e..
L .... . ,
.. . ... i
. ,! . 4 1
,-7,.. .. . .
e.
....; e
.. . .l. .....J.
. . 4 1 . ,. ,.
6 .. ,
1.
. ~ .. _. ,,-
.= , . .
. . .s
.. . .1 s
.I..-.
t
-. . <. .. .. =i
=
.l. . .. m.,..
m.. . I .-
.,. . . g . I. ,
j d. . } . , ,.,. ~ : , . j.. . l .- l
. .i . . ..
. . . ,. . .... .g(. -.. ..
. -..ae
, . ... . ... - , . ..- - .. + .. . t
. , . . . . . -.. . . .1
.l. 17...
.t. :
. 3.,6 ~.
l....
e
-- . .... 4.
. .!..t.
.e -
.1
. . ; . ,J *-
! 1 ; . nt_ .; . : *
, , . . . . .. . . . . , . ... 1. -,' .,: t, : >
. . . . . . . ..... _. .. - I_.................L. . . ... . . ... . 4.. .1;,.. . . .
-. . . . ~ .. . .
.. .. _..4 _.. ... ......... . . ... ~.+. . . ..._.
1
. .t. _ . . . .
t
.e . .$..... ,9
...t...
. L. . .. ,
. i . .... . _ ,- , , .- .
' 1 Incore
. . : F. V Mar. Worst 1 i
- ..i .. m . 2.; : . .n a.. U;r: Offset LMt Case i; J ;
. '. 1
+
- 4. . - .. . ', , (1) (kW/ft) Min DNBit . ! I.
.l... 4
_ .i . _ . -
1 +
.. . UW 4.790' -l i .
,g,o4g g.74 .
4
..4. _... :
.3._ '
l.
.- . . L. _. .
r,' +6.2 9.2; 4.460 .
g;..
- _o *
+7.78 d.47 4.365 * ' '
- F !
j -4.97 8.845 4.750
[ . 9.45 -
- i. -. _.__ . _ . _. _ , . . - . . . . . .
y.. 19.39 4. _ ._.
.. -21.98 4.482 '
.'. P.. 9.63 y
'M*.-
- t. ..
4 - - - OW .d ree se
, t.
g.. .
6 q
.. 6 b f.
=
.c .. a. , .
. . -.-.& .-5 ._.e .$. -
f 6. . 6
-7 ' I. ,
b er .-. . r~..e-- .+., w6 ee.
.- 4 5
-. . ,- , . . . - . ~ . ~ . - . . , . - . . . _ . . . ... .. - _ .. . -. .. . .. . . . . ~
' 4 . 1 7 6 ' 8
.b
- w. _.. -
w.. . .......m.,.....e.r-. , e.-~. . . - . . - - + &- .- ;....-. 3 l
- . L
.i }
. 1 a' ,
i-,. .4 I' .\.
>{
.- = . . *. l 6...
......6 t a.. ... .
,.9 . ,
. .~ ..-~ . . ,
...t......<
..+.l..,...._.
. 6
,- ..............4.... ..
. . - . .1. , . . . .. .
....6...-.... . ,.w.m...
... ~ .
. .,-. ..g .. ,
':M :fr'.
~
l . .[ _ * ::_ r. ::t mt.. :a l.
.. - . ....: .m ..:...., -- -
1:. om +
.t. a. - , _: ...t.a........<.. p.. . ."
+
.........m
....J....._a.........................._............
1-
. t. .
=
. . . _ ,.i . .. . -
+ . _
. . . . y . ,
...... . . ..t... .. ..
g.... .. .. .. .. .. ... .......
.m......._..
. . . _ . g. . . . .
7, .
.t . - . .
=,.
.g .
y Source: PT-120 L
1 Encl. 4, Part 2 Figure 4.12 3 63 l
- l. 5
.- _