ML20210C463
| ML20210C463 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 03/31/1986 |
| From: | Lyon W, Morrissey K, Papanic G YANKEE ATOMIC ELECTRIC CO. |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| FYR-86-030, FYR-86-30, NUDOCS 8603260033 | |
| Download: ML20210C463 (44) | |
Text
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- k, Core XVIII Startup Program For The Yankee Nuclear Power Station March 1986 Prepared By: Ah Aw &M'[6 William E. Lyon, 5hfft Technical Advisor (Date)
Reactor Engineering Department Prepared By: bnhm 3-b-N Kevin J.htorri'ssey, S6nlor Engineer (Date) ,
Nuclear Services Division Reviewed By: ~
E Y"8b Frederick N. Williams, Manager (Date)
Reactor Engineering Department Reviewed By: /M. <
d Richard J. acciapout*, Reactor Physics Manager ('Date)
Nuclear rvices Div on i
l Approved By: V 4 7/ M ~ A ---_ 3 /Y/74 !
Ti:nothy K. Hefiiderson, Technical Director (Date) l Yankee fuelear Power Station Approved By:
- 2. dI
~
8///u d!N/'/,
Normand N.'St. Laurent, Plant Superintendent '(Date)
Yankee Nuclear Power Station Yankee Atomic Electric Company.
Star Route Rowe, Massachusetts 01367 8603260033 860331 PDR P ADOCM 05000029 PDR, g
TABLE OF CONTENTS Page LIST OF TABLES................................................... iii LIST OF FIGURES.................................................. iv I. INTRODUCTION..................................................... I II.
SUMMARY
OF RESULTS............................................... 2 III. STARTUP PROGRAM - MECHANICAL..................................... 3 A. Fuel Assemblies.............................................. 3 B. Control Rods................................................. 4 IV. STARTUP PROGRAM - NUCLEAR........................................ 5 A. Physics Testing.............................................. 5 B. Critical Boron Concentration................................. 6 C. Control Rod Group Worths..................................... 6 D. Moderator Temperature Coefficients........................... 6 E. Power Distribution........................................... 7 F. Power Plus Xenon Defect...................................... 7 V. RFLOAD DESIGN REANALYSIS......................................... 9 VI. REFERENCES....................................................... 40~
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I LIST OF TABLES .l l
Humber Title m P
1 Core XVIII Startup Program Physics Testing Results 11 2 Core XVII-XVIII Refueling Control' Rod Inspection Results 12 3 Core XVIII Delayed Neutron Fractions 13 4 Critical Boron Concentrations 14-5 Group C Worth 15 6 Group A Worth 17 7 Group B Worth 18 8 Moderator Temperature Coefficient (Measured) 20 9 Moderator Temperature Coefficient Comparisons 21 10 Power Plus Xenon Defect Data 22
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LIST OF FIGURES
~ Number ~ Title Page 1 Yankee Core XVIII BOL Assembly Burnups (mwd /Mtu) 23 2 Core'XVIII Control Rod Identification -24 3 Group C Differential Worth 25 4 Group C Integral Worth 26 5 Group A Differential Worth 27 6 . Group A Integral Worth 28 7 Group B Differential Worth 29 8 Group-B Integral Worth 30 9 Moderator Temperature Coefficient (Graphical) 31 10 Gross Quadrant Tilt 32 11 Radial Power Distribution - Comparison of Reaction Rates 33 12 Summary of Incore Results 34 13 Core Locations of Modified Assemblies 35 14 Location of Inert Rods in Recycled Assembly B-1002-R 36
, 15 Location of Inert Rods in Recycled Assembly B-688 37-16 Location of Inert Rods in Recycled Assembly A-679 38 17 Lattice Locations of Inert Rods and New Guide Bars 39
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I. INTRODUCTION.
The Core XVII-XVIII refueling at the Yankee Nuclear Power Station began on October 19, 1985 and was completed with the startup of Core XVIII on December 10, 1985. This report provides details of the Startup Program for Core XVIII.
The intent of the Startup Program is to ensure the proper condition of the reactor and fuel from a mechanical as well as nuclear standpoint. During the refueling outage, fuel assemblies and control rods were' inspected, utilizing various methods, to assure their sound physical condition. During the physics testing, various nuclear parameters and coefficients were measured and recorded to verify the design calculations used in analyzing plant transients and accidents. The nuclear parameters also provide a guide for operator understanding of Core XVIII characteristics during routine plant operation.
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i II.
SUMMARY
OF RESULTS Table 1 contains a summary of the Startup Program physics testing results. Predicted values and acceptance criteria tolerances are from Reference Documents 1 and 2. All parameters measured and/or determined were found to meet the Acceptance Criteria with the exception of Control Rod Group A. This was acceptable since the total of all groups measured was
'within expected tolerances. Fuel modifications.in the form of replacing fuel rods with inert rods were performed during the core XVII-XVIII refueling outage. Refer to the section entitled " Reload Design Reanalysis" for an explanation of the analyses done to ensure the fuel modifications were acceptable.
III. STARTUP PROGRAM - MECHANICAL A. Fu_ej Assemblies Yankee C' ore XVIII is loaded with 40 new zircaloy clad.3.7 w/o fuel assemblies around the perimeter of the core, with 36 once-burned zircaloy clad 3.7 w/o fuel assemblies in the center region as shown in Figure 1. Eight of the fresh assemblies have solid zircaloy inert rods in selected positions, 3 per A assembly, and 10 per B assembly, for a total of 52 inert rods. The B assemblies have two additional guide bars and the A assemblies have one additional guide bar (Figure 17). Spacer stiffener strips are. attached to these new guide bars at various positions along the axial length.
These modifications were performed at Combustion Engineering prior to delivery to Yankee as a precaution against flow-induced fretting wear as described in Reference 1.
During the Core XVII-XVIII fuel shuffle, the first-cycle assemblies were inspected ultrasonically and visually to check for leaking
- fuel rods which had been suspected during Core XVII operation from chemistry analysis of main coolant water. One assembly-(B-6961) was found to have fret ting damage on the spacer grids on the side
, adjacent to the core baffle wall. Two fuel rods were also found to have fretting damage but no through-wall wear. This assembly had been in Core Position C-9 which was the site of previously damaged assembly B-636 during Core XVI and B-574 during Core IV.' The assembly was reconstituted into a new cage (B-1002-R) which had 19 i
inert rods. Two additional inert rods to replace the damaged fuel-i rods'and all the undamaged fuel rods from the damaged assembly (B-696I) were transferred to a new case after individual eddy current testing (Figure 14). Two other assemblies'(B-688 and A-679) were found to have single-failed fuel rods WhichLwere i replaced by inert rods prior to their' placement in the core (Figures 15 and 16).
-__--_____-_a__--_-_-____________-__-_--__________--______--_-___-__________
During the Core XVII-XVIII refueling, two baffle spacer plugs.were installed near Core Position C-9. These baffle spacer plugs were
-designed to reduce the flow behind the baffle spacer, thereby
{ reducing the AP between' the core and baffle spacer. This minimizes j' the driving force for jetting' flow anomalies at that core position.
Upon completion of fuel loading, assembly positioning was checked l
by underwater television and video tape. The video tape was then reviewed independently to verify the core loading.
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,. B. Control Rods 1
The Yankee core has 24 Ag-In-Cd control rods with zircaloy 1
followers. The rods are divided into-three shutdown groups and one controlling group as shown in Figure 2.
During the Core XVII-XVIII refueling, five control rods were f
f inspected visually and checked for bowing in a straightness gauge.
Based on the results'of these inspections, one control rod was determined to have excessive bow and was replaced. 'One shim rod f was also replaced due to excessive bowing. All 24 control rods 1
' o were rotated 90 and returned to-the core. Following completion of fuel loading, all rods were checked for excessive drag force and found acceptable. Following completion of reactor vessel upper internals installation, all rods were again checked for excessive ;
- drag force and found acceptable.
Prior to initial criticality, control rod exercises were performed to verify proper functioning of the control rod drive system. The
. exercises involved moving the rods from 0" to 90" and back to 0" 1
again. Additionally, control rod drop times were measured as a final check that there was no binding or obstruction. The drop.
time is the. interval between when the power is cut to the rod j l latching mechanism until the. control rod drive shaft passes the 6" l coil on the indicating' stack. The rod drop times are measured using a calibrated Honeywell Visicorder. A detailed tabulation of the results of control rod inspection data is shown in' Table 2. l
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IV. STARTUP PROGRAM - WUCLEAR A. Phyaics Testint' In general, physics testing data is collected by. intentionally.
varying one core parameter and measuring its response or effect on reactivity while other parameters are held as constant as
! possible. The variable parameters affecting' reactivity include i~
boron concentration,-temperature, and control rod. position. The i correlations derive ( 'com this data include boron worth, moderator r
i temperature coefficient, xenon plus power defect and control rod worths.
Reactivity data is obtained by connecting a plant nuclear
! instrumentation channel into a reactivity computer. .The
- Westinghouse solid-state reactivity computer is an analog computer solving the differential Inhour equation. Delayed neutron j fractions for Core XVIII, as. calculated by Yankee Nuclear Services Division, are programmed into the computer prior to physics
}
testing. Table 3 contains a listing of Core XVIII delayed neutron i fractions used. Dynamic checks of the reactivity computer are l~ performed before, during, and after the data taking to verify 4
proper calibration of the computer.
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l Boron concentration numbers are provided by the plant Chemistry Department based upon titration analysis of main coolant samples
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- taken at selected times during the course of physics testing.
Multiple samplings and repeated titrations provide a high degree of
- reliability in the boron concentration data.
Main coolant temperature is measured by existing calibrated in-plant instrumentation. Incore thermocouples and loop RTDs, which read out in the Main Control Room, provide reliable data.
Control rod position is indicated with LEDs and odometers on the j Main Control Board.
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Power distribution data is obtained.through use of the plant incore flux mapping system in conjunction with the CDC Computer System.
The incore system controls and computer terminals are located in ,
the Main Control Room.
B. Critical Boron Concentrations Just critical boron concentrations were measured as close as possible to the following conditions: t All rods out Group C inserted
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Croups C and A inserted Refer to Table 4 for the results. Note that the predicted values ,
have been adjusted to reflect actual control rod positions at the ! ,
time of the measurement.
C. Control Rod Group Worths Differential rod worths were measured for Groups C, A, and B using the dilution balance technique. A dilution rate of approximately 25 gpm is used to produce a positive reactivity response. Control' 4 rod group motion is then used to compensate or balance this
- effect. Reactivity is allowed to vary plus or minus 20 pcm from a just. critical state, thereby producing a sawtoothed graphical measurement of differential control rod group worth.
9 From this data, differential and integral rod worths are derived.
Tables 5, 6, and 7 provide a tabulation of the results while Figures 3, 5, 7 and 4, 6, 8 provide graphical representation of rod l group differential and integral worths, respectively.
1 D. Moderator Temperature Coefficient (MTC)
NTC data is obtained by varying the moderator temperature and measuring the corresponding reactivity change -for a minimum of' i
d E
4 three heatup/cooldown cycles. A linear least square fit of.
temperature change versus reactivity change yields the mo'erator d temperature coefficient. Data sets were taken at various boron 29 concentrations. Control rods were moved to compensate for boron changes instead of the burnup compensation which occurs during normal operation. Data was taken as close as possible to the following conditions:
All rods out Group C inserted Groups C and A inserted l
Table 8 provides a listing of the NTC numbers as taken whereas
! Table 9 provides a listing of the NTC results compared to the calculated values. The calculated ~alues were derived based on the actual plant conditions at the time of each measurement. Figure 9 provides a graphical representation of these results.
E. Power Distribution An incore flux map (YR-18-003) was taken at approximately 25% power.
to check for gross quadrant tilt. Figure 10 shows the results of the gross quadrant tilt measurement. The maximum tilt was calculated to be within the 5% acceptance criteria.
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1 An incore flux map (YR-18-007) was taken at approximately 70% power to check relative radial power distribution. Figure 11 shows the-
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comparison of measured versus theoretical reaction rates. This-incore flux map (YR-18-007) was also used to= check that the'LHGR, F,and{,(nuclear)werewithinTechnicalSpecification i limits. Figure 12 shows the results of'these measurements. I l
l F. Power Plus Xenon Defect
-The power defect and the xenon defect are negative reactivity effects which are functions of reactor power.- The power defect is j determined by the fuel and moderator temperatures, while the xenon f
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defect is related to xenon concentration. Instead of trying to separate the two their combined effect is calculated as the power plus xenon defect.
Primary system data (temperature, boron concentration, rod position, pressure, etc.) was taken at zero power and at three other power levels (62%, 70%, 94.5%) during power ascension. A i reactivity balance was performed between the zero power data and each of the other power level data sets to determine the reactivity effects of power plus xenon. Table 1 provides the results of these calculations while Table 10 provides the data with which the calculations were performed.
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V. RELOAD DESIGN REANALYSIS During the Core XVII-XVIII refueling outage, three recycled fuel assemblies were found to contain failed fuel rods which necessitated some fuel assembly redesign with solid zircaloy rods. Two of these assemblies required just one fuel rod to be replaced. Assembly B-6961 (now B-1002-R) sustained the most damage and required the replacement of two fuel rods. Also, as a measure of prevention, eight fresh fuel assemblies in the southwest and northwest core quadrants were built with solid zircaloy rods and extra guide bars with spacer stiffener strips. There were 64 fresh fuel pins replaced, resulting in a core total of 127 inert rods and 12 additional guide bars.
Thislowersthetotalnumberoffuelpinsfrom17[518to17,379,thenet effect being a higher core average linear heat generation rate (4.430 versus 4.395) than initially assumed. - Figure 13 is provided to show the locations and number of replaced pins for the revised Core XVIII reload design, with Figures 14 through 16 showing the individual recycled assembly inert rod )
locations, and Figure 17 depicting the modified fresh assemblies.
Another factor which impacts the original core design is the revised core average burnup due to the longer than assumed Core XVII cycle length.
The initial reload design assumed a recycled batch burnup of 12,998 mwd /Mtu, which resulted in a core average burnup of 6,148 mwd /Mtu. The. revised loading has a core average burnup of 6,239 mwd /Mtu, based on a recycled fuel batch average of 13,186 mwd /Mtu. This fact makes the core slightly less reactive and causes a small increase in the fresh fuel power' sharing.
The major effect of the longer Core XVII cycle length and reconstituted fuel assemblies was a change in relative power distribution. The higher recycled fuel burnup results in a slight shif t in power sharing to the fresh fuel batch. The effeci of the inert rods becomes more of a localized effect, with unaffected assemblies changing only on a relative basis. The largest deviation in power distribution was calculated for a BOL core, with the peak fuel pin in the fresh fuel batch showing about a 1.0% increase. No limiting ~
fuel assemblies of either batch which were analyzed in the Reference 1 report were reconstituted. The small deviations calculated are within the assumed
- uncertainties of the original reload design.
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In terms of core reactivity, the effect of the higher core average
- burnup and the insertion of a small number of solid pins is minimal. The original BOL, HZP, ARO critical bcron concentration was calculated to be 2,068 ppm, while the revised concentration was 2,060 ppm. Therefore, no change in~
the BOL moderator temperature coefficient is realized. Since the reactivity is so similar and the EOL power distribution is also similar, no EOL reactivity parameter coefficients or defects are affected. Additionally, cycle-dependent critical boron concentrations and shutdown margin boron concentrations are well within the assumed uncertainties of the original reload design.
Contrcl rod worths and scram reactivities were also evaluated with the initial relori values found to be slightly higher. The revised total control rod worth at BOL was 11.51%df versus the 11.55% AP originally assumed. Scram 4 reactivities are, therefore, slightly lower than the design values assumed due to the lower total control rod worths, but the difference is clearly bounded by the values assumed in the' safety analysis. No assessment of the rod worth change on rodded transients, such as the dropped and ejected rods was performed due to the little change in worth and slight changes in pin power
. peaking.
In conclusion, the evaluation of the physics parameters of the revised reload design has been satisfactorily perforned. All parameters were either less adverse or within the uncertainties of the original design values.
Therefore, the Reference 1 Core Performance Analysis is bounding in terms of the physics parameters assumed.
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1 TABLE 1 i
CORE XVIII STARTUP PROGRAM PHYSICS TESTING RESULTS Predicted Measured Difference or Accept crit.
Parameter Value Value % Difference Tolerance Control Rod Drop Tines 1.94 sec = 2.5 see Critical Boron concen.
ARO 2060 ppm 2089 ppm 1.4% . 10%
Group C In 1828 ppm 1885 ppm 3.1% t10%
-Groups C & A In 1642 ppm 1713 ppm 4.3% 110%
Control Rod Group Worths Group C 1730 pcm 1765 pcm- . +2.02% i7.5%
Group A 1380 pcm 1506 pcm +9.13% .17.5%
Croup B 2560 pcm 2526 pcm -1.33% 7.5%
To'tal 5670 pcm 5797 pcm +2.24% i7.5%
Moderator Temperature Coef.
ARO -2.4 pcm/ 0 F .4 pcm/0 F +2.0 pcm/0 F 15.0 pcm/ 0 F Group C In -4.8 pcm/ 0 F -4.1 pcm/0 F +0.7 pcm/ 0 F 15.0 pcm/ 0 F Groups C & A In -9.8 pcm/ 0F -8.7 pcm/0 F +1.1 pcm/ 0 F 15.0 pcm/ 0 F Gross Quadrant Tilt 1.7% -
15.0%
Radial Power Distribution 4.442% i5.0%
(Reaction Rate Comparison) -3.705%
Power Plus Xenon Defects 0 - 62% Power 3053 pcm 2611 pcm -14.5%
0 - 70% Power 3219 pcm 3134 pcm -2.63%
0 - 94.5% Power 3755 pcm 3643 pcm -2.98% ---
Table 2.
CORE XVII - XVIII REFUELING.
. CONTROL ROD INSPECTION RESULTS Bow Replacement Drop Time.
Rod Original (Inches) Serial No. (Seconds)-
Position Serial No.
-- 1.51 1 A132
-- 1.54 2 A156
.267 A151 1.48 3 A109
-- 1.48 4 A157
-- 1.45 5 A130
- 1.47 6 A142
.233 - 1.54 7 A113
.11/ - 1.52 8 A131
-- 1.51 9 A134
-- 1.59 10 A139
-- 1.47 11 A140 --
-- 1.73 12 A133
-- 1.77 13 A137
-- 1.52 14 A146 --
.225 -- 1.75 15 A108
-- 1.55 16 A135
-- 1.58 17 A138
-- 1.91 18 A141
-- 1.56 19 A143
-- 1.58 20 A148
-- 1.52 21 A136
-- 1.58 22 A144
-- 1.53 23 A149
.210 - 1.94 24 A127
TABLE 3 YANKEE CORE XVIII DELAYED NEUTRON FRACTIONS FRACTION EFFECTIVE LAMBDA GROUP BETA BAR FRACTION (SEC)-1 1 .00019038 .00018962 .01252 2 .00135436 .00135061 .03056
- 3. .00123816 .00123411 .11498 4 .00253835 .00252911 .30906 5 .00087125 .00086880. -1.16288 6 .00031036 .00030952 3.04694 BETA EFFECTIVE = .006482 BETA BAR = .006503 I BAR = .996761 PROMPT NEUTRON LIFETIME = 20.26-MICROSECONDS STARTUP RATE PERIOD REACTIVITY (DECADES / MIN.) (SEC.) (PERCENT)
.100 260.6 .0269
.500 52.1 .0976 1.000 26.1 .1515 2.606 10.0 .2474 TABLE 4 CRITICAL BORON CONCENTRATIONS (PPM)
Control Rod Position Predicted Measured Difference ARO 2060 2089 +1.4%
Group C In. 1828 1885 +3.1%
Groups C and A In 1642- 1713 +4.3%
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TABLE 5 YANKEE ROWE GROUP C WORTH FROM PHYSICS TESTING OF DECEMBER 6 1985 TOTAL INTEGRAL WORTH OF GROUP C IS 1.7646 % zga INITIAL FINAL DELTA AVERAGE DELTA DIFF. INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH WORTH INCHES INCHES INCHES INCHES PCM PCM/ INCH 0 TO 90 90 TO O 90 82.5 7.5 86.25 30.2 4.02667 30.2 1764.6 82.5 81.75 0.75 82.125 5.5. 7.33333 35.7 1734.4 81.75 75.75 6 78.75 59 9.83333 94.7 1728.9 l
75.75 72 3.75 73.875 48 12 8 142.7 1669.9 72 69 3 70.5 40.5 13.5 183 2 1621.9 69 66.375 2.625 67.6875 38.4 14.6286 221 6 1581 4 j 66 375 64.125 2.25 65.25 34.5 15.3333 256 1 1543 j 64.125 61.875 2.25 63 40.5 18 296 6 1508.5 61.875 60 1.875 60.9375 35 18.6667 331.6 1468 60 58.125 1.875 59.0625 37.5 20 369 1 1433 58.125 56.25 1.875 57 1875 39.2 20.9067 408 3 1395.5 56.25 54.375 1.875 55.7125 41.5 22.1333 449.8 1356.3 54.375 52.875 1.5 53 625 34 22 6667 483.8 1314.8
-1280.8 52.875 51.375 1.5 52.125 35 23.3333 518.8 51.375 50.25 1.125 50.8125 27.4 24.3556 546 2 1245.8 50.25 49.125 1.125 49.6875 28.5 25.3333 574.7 1218.4
- 49.125 47.625 1.5 48.375 39 26 613.7 1189.9 47.625 46.5 1.125 47.0625 30.4 27 0222 644.1 1150.9
- 46.5 45 1.5 45.75 42.2 28.1333 686.3 1120.5
- 45 43.5 1.5 44.~25 43.5 29 729.8 -1078.3 43.5 42.375 1.125 42.9375 33 29.3333 762.8 1034.8 42.375 41.25 1.125 41.8125 35 31.1111 797 8 1001.8
! 41.25 40 125 1.125 40.6875 35.5 31.5555 833.3 966.799 40.125 39 1.125 39 5625 36.4 32.3556 869.699 931 299 39 37 875 1.125 38.4375 36.8 32.7111 906.499 894.899 37.875 36.75 1.125 37.3125 37.4 33.2444 943.899 858.1 36.75 36 0.75 36 375 25 33.3333 968.899 820.7 36 34.875 1.125 35.4375 38 33.7778 1006.9 795.7 34.875 33.75 1.125 34.3,125 38 33 7778 1044.9 757.7 33.75 32 625 1.125 33 1875 39.4 35.0222 1084.3 719.7 32.625 31.875 9.75 32.25 26.6 35.4667 1110.9 680.3 31.875 31.125 0.75 31.5 26.6 35.4667 1137.5 653.7 31.125 30.375 0.75 30.75 26.6 35.4667 1164.1 627.1
'. 30.375 29.625 0.75 30 27 36 1191.1 600.5 29.625 28.875 0.75 29 25 26.6 35.4667 1217.7 573.5 28.875 27.75 1.125 28.3125 39.5 35.1111 1257.2 546.9 27 75 27 0.75 27 375 27 36 1284~.2 507.4 27 25.875 1.125 26 4375 39 34.6667' 1323.2 480.4 25.875 24.75 1.125 25.3125 38 33.7778 1361.2 441.4 24 75 23.625 1.125 24.1875 38 33.7778 1399.2 403.4 23.625 22.5 1.125 23.0625 36.6 32.5333 1435.8 365.4 22.5 21.375 1.125 21.9375 35 31.1111 1470.8 328.8 21 375 20.25 1.125 20.8125 32.8 29.1556 1503.6 293.8
[ 20.25 19.125 1.125 19.6875 32 28.4444 1535.6 261 19.125 18 1.125 18.5625 30.5 27.1111 1566.1 229 18 16.5 1.5 17.25 35 23.3333 1601.1 198.5 16.5 14.625 1.875 15.5625 18 20.2667 1639.1 163.5 14.625 12.375 2.25 13.5 39 17.3333 1678.1 125.5
' TABLE 5 (Continued)
INITIAL -FINAL DELTA AVERAGE DELTA DIFF. INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH WORTH INCHES INCHES INCHES INCHES PCM PCM/ INCH 0 TO 90 90 TO O 12.375 8.25 4.125 10.3125 49 11.8788 ~1727.1 86.5 8.25 0 8.25 4.125 37.5 4.54545 1764.6 37 5 GROUP C DEC!IMBER 6 1985 4
5
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TABLE 6 f
YANKEE ROWE GROUP A WORTH FROM PHYSICS TESTING OF DECEMBER 6 1985 TOTAL INTEGRAL WORTH OF GROUP A IS 1.5063 % 2ho INITIAL FINAL DELTA ~ AVERAGE DELTA DIFF. INTEGRAL INTEGRAL ;
- HEIGHT.- HEIGHT HEIGHT HEIGHT RHO- WORTH WORTH WORTH
- INCHES INCHES INCHES INCHES PCM PCM/ INCH 0 TO 90 90 TO O i
l 90 84.625 5.375 87.3125- 31.5 5.86046 31.5 1506.3
! 84.625 77.375 7 25 81L 73.5 10.1379 105 1474.8
{ 77.375 '73.25 4.125 75.3125 50.6 12.2667 155.6 -1401.3 l 73.25 69.875 3 375 71.5625 46.5 13.7778 202.1 1350.7 i 69.875 66.5 3.375 68.1875 51 15.1111 253.1 1304.2 66.5 63.625 2.875 65.0625 49 17.0435 302.1 1253 2 63.625 60.875 2.75 62.25 45.8 16.6545 347.9 1204.2 60.875 58.625 2.25 59.75 42 18.6667 389.9 1158.4 58.625 54.375 2.25 57.5 44 19 5556 433.9 .1116.4 56 375 -54.125 2.25 55.25 46.5 20.6667 480.4 1072.4 54.125 52.25 1.875 53.1875~ 40- 21.3333 520.4 1025.9-52.25 50.375 1.875 51.3125 42 22.4 562 4 985.9 50.375 48.875 1.5 49.625 34.8 23.2 597.2 943.9 48.875 47.375 1.5 48.125 36 24 633.2- 909.1 47.375 45.5 1.875 46.4375 46.8 ~ 24.96 680 873.1 45.5 43.625 1.875 44.5625 49 5 26.4 729.5 826.3 i 43.625 41.75 1.875 42.6875 50.5 26.9333 780 776.8-41.75 40.25 1.5 41 41 27.3333 821 726.3 l 40.25 38.75 1.5 39.5 42 28 863 685.3 38.75 37.25 1.5 38 42.5 28.3333 905.5 643.3 37.25 35.75 1.5 36.5 43 28.6667 948.5 600.8 i 35.75 34.25 1.5 35 42.6 28.4 991.1 557.8 j 34.25 32.75 1.5 33.5 42.6 28.4 1033.7 515.2
- ' 32 75 31.25 1.5 32. 42.6 28.4 1076.3 472.6 31.25 29.75 1.5 30.5 42.5 28.3333 1118.8 430 29.75 28.25 1.5 29 41.2 27.4667 1160~ 387.5 4
28.25 26.75 1.5 27.5 40.6 27.0667 '1200.6 346.3 26.75 24.875 1.875 25.8125 47 25.0667 1247.6 305.7 24.875 23.375 1.5 24.125 35.5 23.6667 1283.'1' 258.7'
- 23.375 21.875 1.5 22.625 34 22.6667 1317.1 223.2 i 21.875 20.375 15 21.125 32.5 21.6667 1349.6 189.2 i 20.375 18.5 1 875 19.4375 35.5 18.9333 1385.1 156.7 18.5 16.25 2.25 17.375 35 15.5555 1420.1 121 2 l 16.25 0 16.25 8.125 86.2 5.30462 1506.3 86 2 L
i GROUP A DECEMBER 6 1985 l:
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, TABLE 7 i
j YANKEE ROWE GROUP'B WORTH FROM PHYSICS TESTING OF DECEMBER 7 1985 TOTAL INTEGRAL WORTH OF GROUP B IS 2.52639 % 2po i
INITIAL FINAL DELTA- AVERAGE DELTA -DIFF. INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT ~ RHO WORTH WORTH WORTH
) INCHES INCHES INCHES INCHES PCM PCM/ INCH 0 TO 90 90-T0 0
- 90 85.125 4.875 87.5625 35 7.17949 35 2526.4
- 95.125 77.625 7.5 81.375 79.4 10.5867 114.4 2491.39 j- 77.625 73.875 3.75 75.75 54 14.4 168.4 2411.99 l 73.875 70.5 3.375 72.1875 54.3 16.0889 222.7 2358 70.5 67.5 3 16 9 54.5 18.1667 277.2 2303.7
- 67.5 65.25 2.25 66.375 44.6 19.8222 321.8 2249.2 j 65.25 63.375 1.875 64.3125 39.3 20.96 361.1 2204 6 l
63.375 61.5 1 875 62,4375 42.9 22.88 404 2165.3 61.5 59.625 1.875 60.5625 44 23.4667 448 2122.39 4
59.625 57.75 1.875 58.6875 45.7 24.3733 493 7 2078.'4 57.75 56.25 1.5 57 38.2 25 4667 531.9 2032.7 l
56.25 54.75 1.5 55.5 36.1 24.0667 568 1994.5 i
54.75 53.25 1.5 54 42.7 28.4667 610.7 1958.4 53.25 52.125 1.125 52.6875 33.4 29.6889 644.099' 1915 7
! 52.125 51 1.125 51.5625 34.7 30.8444 678.799 1882.3 51 49.875 1 125 50.4375 37.7 33.5111 716.499 1847.6 1
49.875 48.75 1.125 49.3125 40 35.5556 756.5 1809.9 48.75 47 625 1.125 48.1875 40.9 36.3556 797.399 1769.9
. 47.625 46.5 1 125 47.0625 42.1- 37.4222 839.499 1729
! 46.5 45.375 1.125 45.9375 42.5 37 7778 ~881.999 1686 9-l 45.375 44.625 0 75 45 29.2 38.9333 911.199 1644.4 1 44.625 43.875 0.75 44.25 29 38.6667 940.199 1615.2
, 43.875 43.125 0.75 43.5 29.8 39.7333 969.999 1586.2 l 43 125 42.375 0.75 42.75 30.0 41.0667 1000 8 1556.4 j 42.375 41.625 0.75 42 32.3 43.0667 1033.1 1525.6 4 41.625 40.875 0.75 41.25 33.1 44.1333 1066 2 1493.3' 40.875 40.125 0.75 40.5 37.5 50 1103.7 1460.2 40.125 39.375 0 75 39.75 33.9 45.2 1137.6 1422 7 39.375 38.625 0.75 39 35 46.6667 1172.6 1388.8 38.625 37.875 0.75 38.25 35.9. -47.8667 1208.5 1353.8 1 37.875 37.125 0.75 37.5 36.1 48.1333 1244.6 1317 9 37.125 36.375 0.75 36.75 37.2 49.6 1281.8 1281 8
. 36.375 35.625 0.75 36 37.8 50.4 1319.6 1244.6 4
35 625 34.875 0.75 35.25 38.5 51.3333 1358.1' 1206.8 34.875 34.125 0.75 34.5 39.4 52.5333 1397.5 1168.3 I
34.125 33.375 0.75 33.75 40.1 53.4667 1437.6 1128.9 33.375 32.625 0.75 33 41.3 55.0667 1478.9 1088.8 1 32.625 31.875 0.75 32.25 41.1 54.8 1520 1047.5 i
31.875 31.125 0.75 31.5 41.5 55.3333 1561.5 1006.4 31 125 30.375 0.75 30 75 43 57.3333 1604.5 964.899
- 30.375 29.625 0.75 30 43.1 57.4667 1647.6 921 899 29.625 28.075 0.75 29.25 43.8 58.4 1691.4 878.799 l t 28.875 28.125 0.75 20.5 44.3 59.0667 1735.7 -834.999 28.125 27.375 0.75 27 75 43.7 58.2667 1779.4 790.699
, 27.375 26.625 0.75 27 43 57.3333 1822.4 746.999 l t 26.625 25.875 0.75 26.25 43 57.3333 1865.4 703.999 I l
} 25.875 25.125 0.75 25.5 42.8 57.0667 1908.2 660.999 l 25.125 24.375 0.75 24.75 42.4 56 5333 1950.6 618.199'
f TABLE 7 (Continued)
INITIAL FINAL DELTA AVERAGE DELTA DIFF. INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH _ WORTH INCHES INCHES INCHES INCHES PCM PCM/ INCH O TO 90 90 TO O 24.375 23.625 0.75 24 42.2 56.2667 1992.8 575.8 23.625 22.875 0.75 23.25 40 53.3333 2032.8 533.6 22.875 22.125 0.75 22.5 38.7 51.6 2071.5 493.6 22.125 21.375 0.75 21.75 38.2 50.9333 2109.7 454.9 21.375 20.625 0.75 21 34 45.3333 2143.7 416.7 20.625 19.875 0.75 20.25 34.9 46.5333 2178.59 382.7 19.875 19.125 0.75 19.5 32.7 43.6 2211.3 347.8 19 125 18.375 0.75 18.75 30.2 40.2667 2241.49 315.1 18.375 17.625 0.75 18 27.9 37.2 2269.39~ 284.9 17.625 16.5 1.125 17.0625 39.1 34.7555 2308.49 257 16.5 15.75 0.75 16.125 23.2 30.9333 2331.69 217.9 15.75 14.625 1.125 15.1875 32 28.4444 2363.69 194.7 14.625 13.125 1.5 13.875 38.3 25.5333 2401.99 162.7 13.125 11.25 1.875 12.1875 39.6 21.12 2441.59 124.4 11.25 8 25 3 9.75 45.2 15.0667 2486.79 84.8 8.25 0 8.25 4.125 39.6 4.8 2526.39- 39.6 GROUP B DECEMBER 7 19E5
_ _ . ~ , _
r TABLE 7 (Continued)
INITIAL FINAL DELTA AVERAGE DELTA DIFF. INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH WORTH INCHES INCHES INCHES INCHES PCM PCM/ INCH 0 TO 90 90 TO O 24.375 23.625 0.75 24 42.2 56.2667 1992.8 575.8 23.625 22.875 0.75 23.25 40 53.3333 2032.8 533.6 22.875 22.125 0.75 22.5 38.7 51.6 2071.5 493.6 22.125 21.375 0.75 21.75 38.2 50.9333. 2109.7 .454.9 21.375 20.625 0.75 21 34 45.3333 2143.7 416.7 20.625 19.875 0.75 20.25 34.9 46.5333 2178.59 382.7 19.875 19.125 0.75 19.5 32.7 43.6 2211.3 347.8 19.125 18.375 0.75 18.75 30.2 40.2667 2241.49 315.1 18.375 17.625 0.75 18 27.9 37.2 2269.39 284.9 17.625 16.5 1.125 17.0625 39.1 34.7555 2308.49 257 16.5 15.75 0.75 16.125 23.2 30.9333 2331.69 217.9 15.75 14.625 1.125 15.1875 32 28.4444 2363.69 194.7 14.625 13.125 1.5 13.875 38.3 25.5333 2401.99 162.7 13.125 11.25 1.875 12.1875 39.6 21.12 2441 59 124.4 11.25 8.25 3 9.75 .45.2 15.0667 2486.79 84.8 8.25 0 8.25 4.125 39.6 4.8 2526.39 39.6 GROUP B DECEMBER 7 1985
TABLE 8 YANKEE ROWE MODERATOR TEMPERATURE COEFFICIENT (MTC)
RUN 4 RODS TEST MODE TEST DATE MTC (PCM/DEG) 1 ARO HEATUP 12 6 85 -1.25767 2 ARO C00LDOWN 12 6 18 5 -0.136178 3 ARO HEATUP 12- 6 85 -0.452587 4 ARO COOLDOWN 12 6 85 0.00104314 5 ARO HEATUP 12 6 85 -0.630434 6 ARD C00LDOWN 12 6 85 0.0577642 7 ARO HEATUP ~12 6 85 -0.53924 8 C90 HEATUP 12 6 85 -3.444 9 C90 C00LDOWN 12 6 85 -4.25045 10 C90 HEATUP 12 6 85 -3.35648 11 C90 COOLDOWN 12 6 85 -4.33542 12 C90 COOLDOWN 12 6 85 -4.13649 13 C90 HEATUP 12 6 85 -4.81234 14 CIA 9 0 HEATUP 12 6 85 -10.3908 15 CSA 9 0 COOLDOWN 12 6 85 -9 41022 16 CSA 9 0 HEATUP 12 6 85 -8.17282 17 CSA 9 0 COOLDOWN 12 6 85 -7.36228 18 CSA 9 0 HEATUP 12 6 85 -8.72532 19 CSA 9 0 C00LDOWN 12 6 85 -8.38632 TABLE 9 MODERATOR TEMPERATURE COEFFICIENT COMPARISONS
> (PCM/0 F)'
(1)
Control Rod Position Predicted Measured Difference ARO -2.4 -0.4 +2 -. 0 Group C In -4.8 -4.1 +0.7-J Groups C and A In -9.8 -8.7 +1.1 l-(1) Average of'all measurements performed.
)
i
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TABLE 10 POWER PLUS XENON DEFECT DATA Boron Rod Power Mwt Concentration Temperature Position 0% 0 19 31 ppm 510 F C @ 25.13" 62% 371.8 1750 ppm 520 F. C @ 90" 70% 419.9 1685 ppm 513 F C @ 83"-
94.7% 568.4 1597 pp-t 535 F C @ 87.5" Refer to Table 1 for power plus xenon defect results.
RGURE1 YANKEE CORE XVill BOL ASSEMBLY BURNUP (MWD /MTU) 4
- 0. O. O. O.
1
- 0. O. 11642. O. O. O.
I
- 0. O. 12313. 11805, 17276. 17022. O. O.
f
- 0. O. 16991. 9493. 9739. 12305. 9688. 1218 3. O. .0.
i i
l 0. O. 17471. 122 48. 17121. 16945. 9889. 11867. 11417. O.
O. 11863. 11681. 9525. U036. 17030. 12114. 17 412. O. O.
l
- 0. O. 12315. 9711. 12137. 9721, 9906. 18872. O. O. I 1
- 0. O. 17003. 17401. 11434. 12138. O. O.
O. O. O. 11666. O. O.
O. O. O. O.
I
(
1
)
i i
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FIGURE 2 YANKEE CORE XVIII CONTROL ROD IDENTIFICATION
-A l
BCI l D r 7 -
l u
1 4 I I 1 I l i i i __i___l______
l l l D i i i i i i i i 17 i
i i i D- C __i--'---2 i i i I
, i 24 14 _,_l_ _ _
i B B D i D
5 6 18
[ ~
B A B 23 9 3 10 C A A C -
1 13 1 2 15 B A B D !
12 ' t* ' l' D B B 22 8 7
. C D ,
, I 16 20 l i
D 21
___________g KEY:
A - Shutdown Group A l
B - Shutdown Group B C - Controlling Group C D - Shutdown Group D l
1 j
1
FIGURE 3 8
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e i a i i O 8 8 O R R S (HONI/WDd) HIMOM lWIIN383JJIO
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8 0
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FIGURE 7 8
a en a
a h -8 O
a o z
o a en c a "ez-o a r a c I
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- M o O 3
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en
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i z
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8 R 8 8 ? 8 R S (HONI/WOd) H180M lWIIN383dJIO
-.z FIGURE 8 8
en Ed D
__.1 C
> 'o co O
Ed D
ed
_o-
~Z E 3 CC I E I
O
- P I E-'
- u. "OH H -I o 3 H E-e LD H E .-4
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ra a e O M CC O. e x L.1 -8Z H O CD Q Ed O.
Cd E-* Z Z Z O O H
O H E E Cd oN
- H CD N Ed M Cd Q_ E-* O M D C Q-Z O I E & O
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. . , i . . , , , . :o N n a a - e -
(WOd) HIMOM 198931NI
i YANKEE R0WE CORE XVIII MODERATOR TEMPERATURE COEFFICIENTS BOL HOT ZERO POWER (515 F) 4
, o - RVERAGE VALUES o - RLL RODS OUT L
2-A - GROUP C IN R + - GROUPS C+R IN E
l b
0-
. N.
.- o 8 g g D A i c
$ 9=
i N +
~8-y
- c
. 5 +
E
-12 , ,
j 1600 1800 2000 2200 i BORON CONCENTRATION'(PPM)
FIGURE 10 GROSS OUADRANT TILT INCORE RUN YR-18-003
. 165. MWT. GROUP C 9 58 INCHES i
STANDARD ORIENTATION 1.0064 .9984
.9830 1.0122 DIRECTIONAL ORIENTATION
.9949 1.0023 1.0113
.9915
\
f I
Maximum Value = 1.70%
r Acceptance Criteria = 1 5.0%
i i
A i
j i
4
l I
! l l
FIGURE 11 COMPARISON OF MEASURED AND PREDICTED SIGNALS INCORE RUN YR-18-007 420.0 MWT. GROUP C AT 82.11NCHES 74. MWD /MTU 0.666 0.696
, -4.4 1.132 1.169
-3.1 1.053 0.971 1.049 0.973 l
0.4 -0.2 0.974 1.107 I 0.973 1.062 0.1 4.2 1.062 1.062
> 1.017 1.051 l 4.4 1.1 1.054 i 1.015 I 3.9 1.103 0.976 1.069 0,986 3.2 -1.0 1.052 1.067
-1.4 0.641 1.146 WEASURED S10NAL 0.682 1.190 PREDICTED SIGNAL
-6.1 -3.7 PERCENT DIFFERENCE i
l AVERAGE ABSOLlJTE DIFFtd<ENCE BETWEEN MEASURED AND PREDICTED 2.643 PERCENT , ?
RMS ERROR 3.225 1
4
,- , - - - - , - - . - - - - , , - ~ ,-, - ,. .
1 FIGURE 12
SUMMARY
OF INCORE RESULTS.
YR-18-007 420.MWT. 74. MWD /MTU FRESH FUEL RECYCLED FUEL Fq (Measured) 2.413 2.400 Fq (Limit) 3.943 3.943
% Margin to Limit 38.8 39.1 F g (Measured) 1.643 1.644 FoH (Limit) 1.908 1.908
% Margin to Limit 13.9 13.8 LHGR (Measured) 7.153 7.114 LHGR (Limit) 10.227 11.124
% Margin to Limit 30.1 36.1 RGURE 13 .
YANKEE ROWE CORE XVill CORE LOCATIONS OF MODIFIED ASSEMBUES 1 2 3 4 CE (4) 5 6 7 8 9 10 CE EXXON
(+) (4) 11 12 13 14 15 16 17 18 CE EXXON (4)' (12) 19 20 21 22 23 24 25 26 27 28 CE (4) 29 30 31 32 33 34 35 36 37 38 EXXON EXXON EXXON EXXON (1) (1) (12) (4) 39 40 41 42 43 44 45 46 47 48 EXXON (21) 49 50 51 52 53 54 55 56 57 58 CE EXXON EXXON EXXON (12) (4) (4) (12) 59 60 61 62 63 64 65 66 CE (12) 67 68 69 70 71 72 CE (12) ,
73 74 75 76 ASSEMBLY NUMBER l CE FUEL TYPE l (12) l OF REPLACED RODS l i
1 1
l FIGURE 14 Location of Inert Rods In Recycled Assembly B-1002-R FUEL ASSEMBLY MAP ae^== YANKEE R0WE-TYPE B ASSEW "*
TOP ASSEMBLY NUMBER Q-iocq,-% MATRIX: gg y gg REACTOR TYPE: PWR MIEJ SIDE A memnarna mass A BC D EFG HK LMNPRST l 1
E 3 l
______ l 4 ______
l 5
e l Il l l l l l l 7 l a l l l l l l l l $
w 6 l l l l l l l l w S 9 ms g 1
11 E _ l~
14 15 ____ _
7 SIDE C l
- w. on=
g
. - mca su maass me a sur a 3 zmt RA l
l FIGURE 15 ,
Location of Inert Rods !
In Recycled Assembly B-688 FUEL ASSEMBLY MAP REMD* aS5= v "*-
YANKEE R0WE-TYPE B TOP ASSEN MMBER: g -(3 gg MATRIX: gg y gg REAC " "-PWR -
MY SIDE A memncarun asem A B C D EF G HK LMNPRST 1 X 2 X X 3
4 5
6 7
- a m W
a e
g Insi X w o
H TU9E H U
10 N .
it 12 13 1A 15 7
SIDE C c amE aus i e - s-AmR 35tIAL maass AfE tu sng 3 l n za uG l L ____ A
FIGURE 16 Location of Inert Rods j In Recycled Assembly A-679 FUEL ASSEMBLY MAP a'^ = =
YANKEE ROWE-TYPE A ^== = v 2 5* TOP Assexsty wuusen: A - c,q.g wArnzx: 16 X 18 REACTOR TYPE: PWR AEM.1 SIDE A menmiza nem AB C D EFG H KLMNP3S T i X e X _____ E _______ _
3 ______________ _
4 5
6 7
- c m W W Q g INST C H TIEE H
- 10
- X 11 12 13 14 15 ie X X X SIDE C c smEnus s - MS MRIAL ltBWs AfE GI II!E B lI ceet. R.ek
-sa-
FIGURE 17 Lattice Locations of Inert Rods And New Guide Bars (CE Fuel Assemblies)
YANKEE ASSEMBLY. TYPE A X x X X
0 ASSEMBLIES s -
n X A703I O A7071 A709I O A711I X X X
YANKEE RSSENBLY TYPE B X X X O
O Q - zusr. neetz O ,
O Z - mIm - i O
W X H - =w wim m D i X Q - 1== an ASSEMBLIES B704I B706I B7081 B710I X X MOOOOO:X.
n
VI. REFERENCES
- 1. YAEC-146,." Core %VIII Performance Analysis".
- 2. Internal Memo, R. Paulson to F. Williams, " Yankee Core XVIII'Startup i Physics Data", RP 85-331, November 27, 1985.
- 3. Plant Refueling and Inspection Procedures: OP-1700, 1704, 1705, and 1706.
- 4. Physics Test Procedures: OP-1701, 1702.
f T*'=pho"'(' ") *0 YANKEE ATOMIC ELECTRIC COMP)lNT TWX 710-380 7619
-'" /
~
'e' FYR 86-030 1671 Worcester Road, Framingham, Massachusetts 01701 2.C2.01
[
, /#
3~ ,
/
, s t .t K fM , '
_ - . - March 7,~ 1986 h
United States Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 631 Park Avenue King of Prussia, PA 19406 Attention: Dr. Thomas E. Murley, Regional Administrator
Reference:
(a) License No. DPR-3 (Docket No. 50-29)
Subject:
Yankee Core XVIII Startup Program Report
Dear Sir:
Enclosed please find two (2) copies of the Yankee Nuclear Power Station Core XVIII Startup Program Report. This report is submitted in accordance with Yankee Technical Specification Section 6.9.1.
If you have any questions or desire additional information, please contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY n'
George panic, Jr.
Senior Project Engineer Licensing GP/gbc Attachments
-l 'I
[4 .