ML20235Q777

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Core Xix Startup Program for Yankee Nuclear Power Station
ML20235Q777
Person / Time
Site: Yankee Rowe
Issue date: 09/30/1987
From: Carpenito F, Lyon W, Papanic G
YANKEE ATOMIC ELECTRIC CO.
To:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
FYR-87-105, NUDOCS 8710070696
Download: ML20235Q777 (42)
v • d • e


Text

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l Core XIX Startup Program For The Yankee Nuclear Power Station September 1987 Prepared By: //

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9 8"7 yr W'illiam E. Lyon, Shift Technical Advisor (Date)

Reactor Engineering Department Prepared By:

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Frederick L. Carpeni\\o, Engineer (Date)

Nuclear Services Division Reviewed By:

N O. M NL 9fE!E7 Kevin' J. M8rrissey, Senior $gineer (Date)

Nuclear Services Division Reviewed By:

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[ '7 R/jd.Cacciap ti, Reactor Physics Manager (Dafe)

NGelear Serv es Division Reviewed By: 7

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Frederick N. Williams, Manager (Date)

Reactor Engineering Department Approved By:

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Timothy K. Henderson, Technical Director (Date)

Yankee Nuclear Power Station

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Approved By:

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Normand N. St. Laurent, Plant Superintendent (Date)

Yankee Nuclear Power Station Yankee Atomic Electric Company Star Route Rowe, Massachusetts 01367 5759R/20.105 8710070696 870930 PDR ADOCK 05000029 F.

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.t TABLE OF CONTENTS Page y

l L I ST 0F TAP ', E S...................................................

iii LIST OF FIGURES..................................................

iv:

I.

INTRODUCTION.....................................................

1 l'

III..

SUMMARY

OF RESULTS............~.............................'......

2 i

-III.

STARTUP PROGRAM - MECEANICAL.....................................

3 A.

Fuel Assemblies..............................................

3 l

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.

RELOAD DLSIGN REANALYSIS.........................................

9-VI.

REFERENCES.......................................................

36

-ii-5759R/20.105 4

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LIST OF TABLES

' Number-Title Page 1

Core XIX Startup Program Physics Testing Results 11 2

Core XVIII-XIX Refueling Control Rod Inspection Results 12

.3-Core XIX Delayed Neutron Fractions 13 4

Crit.ical Boron Concentrations la 15 5-Group C Worth 16 6

Group A Worth 17

~

7 Group B Worth

'8 Moderator Temperature Coefficient (Measured) 19 9

Moderator Temperature Coefficient Comparisons 20 10 Power Plus Xenon Defect Data 21 1

1

-lii-5759R/20.105 i

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LIST OF FIGURES l

Number Title

_Page

'l Yankee Core XIX BOL Assembly Burnups (mwd /Mtu) 22 2

. Core XIX Control Rod Identification 23 1

1

'3 Group C Differential Worth 24 j

'4 _

Group C Integral Worth 25 5-Group A Differential Worth 26-6

- Group A Integral Worth 27 7

Group B Differential Worth 28

.8 Group B Integral Worth 29

.9 Gross Quadrant Tilt

.30 10 Radial Power Distribution - Comparison of Reaction Rates 31 11 Summary of Incore Results 32 12 Core Locations of Modified Assemblies 33 13 Location of Inert Rods in Recycled Assembly ASP 1I 34 14-Lattice Locations of Inert Rods and New Guide Bars 35 4

-iv-5759R/20.105

~ I. 4, INTRODUCTION I

The Core XVIII-XIX. refueling at the Yankee Nuclear Power Station began

..on.May.2. 1987 and was completed with the startup of Core XIX on 1

July 3,~1987.. This report provides details of the Startup Program.for Core XIX.

The. intent of the Startup'Prograr is to ensure the proper condition of the reactor and fuel from.a mechanical as well as nuclear standpoint. During-th:2 refueling outage, fuel assemblies and control rods were inspected, utilizing various methods,.to assure their sound physical condition. During tha 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 oparator understanding of Core XIX characteristics during. routine plant

' i i

operation.-

4 L

5759R/20.105

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l II.

SUMMARY

OF RESULTS i

Table 1 contains a summary of the Startup Program physics testing rssults. Predicted values and acceptance criteria tolerances are from Raference Documents 1 and 2.

All parameters measured and/or determined were fcund to meet the Acceptance Criteria with the exception of Control Rod Group A integral worth. This was acceptable since the total integral worth 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 XVIII-XIX refueling outage. Refer to the section entitled " Reload Design Rsanalysis" for an explanation of the analyses done to ensure the fuel modifications were acceptable.

6 5759R/20.105 ___ _ ________________

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1 i.

x

.l

'III.

STARTUP PROGRAM'- MECHANICAL'

]

A.-

Fuel Assemblies

! Yankee Core XIX is loaded with 36 new zircaloy clad 3.8 w/o fuel assemblies around the perimeter of the core, with 40 once-burned sirealoy 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 rodsLin aelected positions, 6 per A assembly, and 10 per B l

assembly, for a total of 64 inert rods.- The B assemblies have two additional ~ guide bars and,the A assemblies hive one additional

-guide'bar (Figure 14). Spacer stiffener strips are attached to-these new guide bars at various positions along the axial length.

i 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 XVIII-XIX fuel shuffle, the first-cycle assemblies were inspected ultrasonically and visually to check for leaking fuel rods which had been suspected during Core XVIII operation from chemistry analysis of. main coolant water. One assembly (A-731) was found to have fretting damage on the spacer grids on the side adjacent to the core baffle wall. Fourteen fuel rods were also found to have been damaged.

This assembly had been in Core Position H-9, and has been shuffled to Position H-5 for Core XIX.

The assembly was reconstituted into a new cage (A-SP1-1) which has 13 inert rods, and one additional guide bar. Figure 13 presents this new assembly's cross sectional view.

During the Core XVII-XVIII refueling, two baffle spacer plugs were installed near Core Position C-9.

These bcffle spacer plugs were

' designed to reduce the flow behind the baffle spacer, thereby reducing the AP between the core and baffle spacer. This minimizes the driving force for jetting flow anomalies. During the Core XVIII-XIX refueling, eighteen more plugs were installed throughout the core in order to alleviate any fuel damage that may

~~

5759R/20.105 I'L -_- _

l I

I l

be. caused by water jetting.

Baffle plugs are now installed in all baffle spacer locations in the core.

l~

j Upon completion of fuel loading, assembly positioning was checked I

l

.by underwater television and video tape. The video tape was then reviewed independently to verify the core loading.

)

B.

Control Rods The iankee core has 24 Ag-In-Cd control rods with zircaloy followers. The rods are divided into three shutdown groups and one

(

i controlling group as shown in Figure 2.

During the Core XVIII-XIX refueling, all twenty-four control rods were inspected visually. Six were checked for bowing in a straightness gauge. Based on the results of these inspections, one i

control rod was determined to have excessive bow and was replaced.

Two shim rods were also replaced due to excessive bowing. All 24 control rods were rotaced 90 and returned to the core.

l Fo11owir.g completion of fuel loading, all control and shim rods were checked for excessive drag force. One shim rod failed this test and was replaced. 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 i

to verify proper functioning of the control rod drive system.

The exercises involved moving the rods from 0" to190" and back to 0" again. Additionally, control rod drop times were measured as a final check that thete 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" i

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.

5759R/20.105

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IV.o.STARTUP PROGRAM - NUCLEAR A.

Physics Testing In general, physics testing data is collected byfintentionally varying one core parameter and measuringLits response or effect on reactivity while other parameters are held as constant as possible..The variable parameters affecting reactivity include'

' boron concentration, temperature, and control rod position. The

~

correlations derived from this data include boron worth, moderator o

temperature coefficient, xenon plus power defect and control rod worths..

Reactivity data is obtained by connecting a plant nuclear l

instrumentation channel into a-reactivity computer. The Westinghouse solid-state reactivity computer is an analog computer solving the differential Inhour equation. Delayed neutron fractions for Core XIX, as' calculated by Yankee Nuclear Services j

Division, are programmed into the computer prior to physics testing. Table'3 contains a listing of Core XIX delayed neutron fractions used.: Dynamic checks of the reactivity computer are 1

1 performed before, during, and after the data taking to verify proper calibration of the computer.

i Boron concentration numbers are provided by the plant Chemistry Department based upon titration analysis of main coolant samples 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 thermocouple and loop RIDS, which read out in the Main Control Room, provide reliable data.

Control rod position is indicated with LEDs and odometers on the Main Control Board.

57598/20.105-,

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Power distribution data is obtained through use of the plant incore L

~

' flux mapping system in conjunction with the CDC Computer System.

)

The incore system controls and computer terminals are located in i

.the Main Control Room.

B.

Critical Boron Concentrations L

Just critical boron concentrations were measured as close as possible to the following conditions:

All rods out y

Group C inserted Refer to Table 4 for the resulte. Note that the measured values have been adjusted to reflect actual control rod positions to allow one-to-one comparison with predicted values.

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 spm is used to produce a positive reactivity response. Control 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.

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 group differential and integral worths, respectively.

D.

Moderator Temperature Coefficient (MTC)

MTC data is obtained by varying the moderator temperature and measuring the corresponding reactivity change for a minimum of three heatup/cooldown cycles, A linear least square fit of 5759R/20.105

l temperature change versus reactivity change yields the.moderstor j

temperature coefficient'. Data sets were taken at various boron 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 l,

following conditions:

C i

l All rods out Group.C inserted l3 Table 8 provides a listing of the MTC numbers as taken whereas Table 9'provides a listing of the MTC results compared'to the calculated values. The calculated values were derived based on the actual plant conditions at the time of each measurement.

E.

Power Distribution An incore flux map (YR-19-205) was taken at approximately 25% power to check for gross quadrant tilt. Figure 9 shows the results of j

.the gross quadrant tilt measurement. The maximum tilt was calculated to be within the 5% acceptance criteria.

\\

An incore flux map (YR-19-207) was taken at ;3.8% power to check relative radial power distribution. Figure 10 shows the comparison of measured versus theoretical reaction rates. This incore flux-map (YR-19-207) was also used to check that the LHGR, F, and FH.

(nuclear) were within Technical Specification limits.

Figure 11 i

shows the results of these measurements.

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 determined by the fuel and moderator temperatures, while the xenon 5759R/20.105.

g-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 two other 7

power. levels (65%, 99.7%) during power ascension. A 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 I

calculations were performed.

1 I

i I

5759R/20.105 4

V.

RELOAD DESIGN REANALYSIS During the Core XVIII-XIX refueling outage, one recycled fuel assembly was fo'nd to contain failed fuel rods which necessitated fuel assembly u

reconstitution with solid zircaloy rods. Assembly A-731 (now A-SP1-I) sustained damage and required the replacement of fourteen 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 76 fresh fuel pins replaced, resulting in a core total of 141 inert rods and 25 additional guide bars.

This lowers the total number of fuel pins from 17,518 to 17,352, the net offect being a higher core average linear heat generation rate (4.437 versus 4.395) than initially assumed. Figure 12 is provided to show the locations and number of replaced pins for the revised Core XIX reload design, with Figure 13 showing the individual reconstructed assembly inert rod locations,

(

and Figure 14 depicting the modified fresh assemblies. All of the changes from the original reload design were implemented into the current licensing models. Redepletion in full-core geometry was performed, and any parameter changes were evaluated.

A factor which had a very small impact on the original core design is the core average burnup. The initial reload design assumed a recycled batch burnup of 14.427 mwd /Mtu, which resulted in a core average burnup of 7,582 mwd /Mtu. The revised loading has a core average burnup of 7,598 mwd /Mtu, based on a recycled fuel batch average of 14.440 mwd /Mtu. This fact has an insignificant effect on the licensing calculations of the initial reload design.

The combination of fuel reconstitution and a core average burnup change has virtually no effect on the initial reload design. The reconstituted essembly does have a small localized power deviation but was not one of the limiting assemblies in the Reference 1 report.

The small deviations calculated are within the assumed uncertainties of the original reload design.

5759R/20.105. _ _ _ _ _

l l

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 boron concentration was calculated to be 1,851 ppm (520 F), while the revised concentration was 1,873 ppm (515 F).

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.

Control rod worths and scram reactivities were unchanged. No assessment of the rod worth change on rodded transients, such as the dropped l

and ejected rods, was deemed necessary.

In conclusion, the evaluation of the physics parameters of the revised reload design has been satisfactorily performed. 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.

4 5759R/20.105 l l

TABLE 1 CORE XIX STARTUP PROGRAM PHYSICS TESTING RESULTS Predicted Measured Difference or Accept Crit.

Parameter Value Value

% Difference Tolerance 2.5 see l

Control Rod Drop Times 1.98 see Critical Boron Concen.

ARO 1873 ppm 1925 ppm

+2.8%

210%

Group C In 1641 ppm 1708 ppm

+4.1%

i10%

. Control Rod Group Worths

!.5%

7 Group C

.fi'O pcm 1789 pcm

+1.1%

15%

7 Group A 1370 pcm' 1536 pcm

+12.1%

25%

7 Group B 2300 pcm 2417 pcm

+5.1%

Total 5440 pcm 5742 pcm

+5.6%

7.5%

Moderator Temperature Coef.

1 0 pcm/0F ARO

-3.7 pcm/0F

-4.6 pcm/0F

-0.9 pcm/0F 5

1 0 pcm/0F Group C In

-7.2 pcm/0F

-5.6 pcm/0F

+1.6 pcm/0F 5

!.0%

5 2.9%

Gross Quadrant Tilt

+3.512%

5.0%

Radial Power Distribution (Reaction Rate Comparison)

-2.266%

Power Plus Xenon Defects 0 - 65% Power 2870 pcm 3089 pcm

+7.65%

0 - 99.7% Power 3618 pcm 3540 pcm

-2.16%

5759R/20.105 -

TABLE 2 CORE XVIII-XIX REFUELING CONTROL ROD INSFECTION RESULTS Rod

. Original Bow Replacement Drop Time Position Serial No.

(Inches)

Serial No.

(Seconds) 1.52 1

A132 1.50 2

A156 1.45 3

A151 1.53 4

A157 1.43 5

A130 1.45 6

A142 7

A113

.200 1.53 1.50 8

A131 1.72 9

A134

.140 1.75 10 A139

.160 1.48 11 A140 1.73 12 A133 1.77 j

13 A137

.200 1.53 14 A146 15 A108

.270 A150 1.70 1.52 16 A135 1.60 17 A138 18 A141 1.95 19 A143 1.58 1.58 20 A148 1.52 21 A136 1.55 22 A144 1.53 23 A149 1.98 24 A127

.200 5759R/20.105 - - _ - -

TABLE 3 l

YANKEE CORE XIX l

DELAYED NEUTRON FRACTIONS l

FRACTION EFFECTIVE LAMBDA GROUP BETA BAR FRACTION (SEC)-1 1

.00018645

.00018599

.01253 2

.00133418

.00133268

.03054 3

.00121616

.00121410

.11529 4

.00248914

.00248388

.30974 5

.00085633

.00085530 1.16396 6

. 00030459

.00030426 3.04058 BETA EFFECTIVE =

.006376

.006387 BETA BAR

=

.998334 I BAR

=

PROMPT NEUTRON LIFETIME =

20.49 MICROSECONDS STARTUP RATE PERIOD REACTIVITY (DECADES / MIN.)

(SEC.)

(PERCENT)

I

.100 260.6

.0265

.500 52.1

.0961 1.000 26.1

.1490 2.606 10.0

.2434 5759R/20.105

. TABLE 4 CRITICAL BORON CONCENTRATIONS (PPM)

Control Rod Position Predicted Measured Difference ARO 1873 1925

+2.8%

Group C In 1641 1708

+4.1%

$759R/20.105.

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TABLE 5 YANKEE ROWE GROUP C WORTH FROM PHYSICS TESTING OF JULY i 1987 l

l TOTAL INTEGRAL WORTH OF GROUP C IS 1.789 %

INITIAL FINAL DELTA AVERAGE DELTA DIFF.

INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH WORTH i

INCHES INCHES INCHIS INCHES PCM PCM/ INCH O TO 90 90 TO O 90 77.125 12.875 83.5625 82 6.36893 82 1789 77.125 72 625 4.5 74.875 55 12.2222 137 1707 72 625 69 3 625 70.8125 53 14.6207 190 1652 69 65.5 3.5 67.25 53 15.1428 243 1599 l

l 65.5 63.25 2 25 64.375 39 17.3333 282 1546 63.25 61 2 25 62.125 41 18.2222 323 1507 61 58 75 2 25 59.875 43 19 1111 366 1466 58.75 56.5 2 25 57.625 46 20 4444 412 1423 56.5 55 15 55.75 32 21 3333 444 1377 95 53.125 1.875 54.0625 42 22.4 486 1345 53 125 51.25 1 875 52.1875 43 22.9333 529 1303 58 25 49.375 1.875 50.3125 46 24.5333 575 1260 49 375 48.25 1 125 48.8125 27 5 24.4444 602.5 1214 l

48.25 46 75 15 47.5 39 5 26.3333 642 1186.5 46 75 45 25 1.5 46 41 27.3333 683 1147 45.25 44 125 1 125 44.6875 32 28 4444 715 1106 44.125 43 1.125 43.5625 32 5 28 8889 7e7.5 1074 43 41 875 1.125 42 4375 33 29.3333 780.5 1041.5 41 875 40.75 1.125 41.3125 54 30 2222 814.5 1008.5 40.75 39 625 1.125 40.1875 35.5 31 5555 850 974.5 39.625 38 5 1.125 39.0625 36 32 886 939 38 5 37 375 1.125 37.9375 37 32 8889 923 903 37 375 36.25 1.125 36.8125 38 33 7778 961 866 36 25 35 125 1.125 35.6875 38 33 7778 999 828 35 125 34 1.125 34.5625 39 34.6667 1038 790 34 32.875 1.125 33.4375 40 35.5556 1078 751 32.875 31 75 1.125 32 3125 40 35.5556 1118 711 31.75 30.625 1.125 31 1875 40 35 5556 1158 671 30 625 29.875 0.75 30.25 27 36 1185 631 29.875 29.125 0.75 29.5 27.5 36 6667 1212.5 604 29 125 28.375 0.75 28 75 27 36 1239.5 576.5 28 375 27.625 0.75 28 27 36 1266.5 549.5 27.625 26 875 0.75 27.25 27 36 1293.5 522.5 26.875 26.125 0.75 26.5 26 34.6667 1319 5 495.5 26 125 25 1.125 25 5625 39 34.6667 1358.5 469.5 35 23.875 1.125 24.4375 38 33.7778 1396.5 430.5 23.875 22 75 1.125 23.3125 37 32 8889 1433 5 392.5 22.75 21.625 1.125 22 1875 36.5 32.4444 1470 355.5 21 625 20.5 1.125 21 0625 35 31.1111 1505 319 20.5 19.375 1.125 19 9375 32 28.4444 1537 284 19.375 17.875 1.5 18.625 40 26.6667 1577 252 17.875 16.75 1 125 17 3125 36.5 32.4444 1613.5 212 16.75 14.5 2 25 15 625 40 17.7778 1653.5 175.5 14.5 12.625 1.875 13 5625 35 18.6667 1688.5 135.5 12.625 9.25 3.375 10.9375 47.5 14.0741 1736 200.5

,9 2 5 0

9.25 4.625 53 5.72973 1789 53.

TABLE 6 YANKEE ROWE GROUP A WORTH FROM PHYSICS TESTING OF JULY 2 1987 TOTAL INTEGRAL WORTH OF GROUP A IS 1.5362 %

INITIAL FINAL DELTA AVERAGE DELTA DIFF.

INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH WORYH INCHES INCHES INCHES INCHES PCM PCH/ INCH 0 TO 90 90 TO O

'90 81.25 8.75 85.625 59.4 6.78857 59.4 1536.?

81 25 76.75 4.5 79 48.8 10.8444 108.2 1476 8 76 75 73.375 3.375 75.0625 41.5 12.2963 149.7 1428 73.375 70.375 3

71.875 40.6 13.5333 190.3 1386.5 70.375 67 75 2 625 69.0625 39.1 14.8952 229.4 1345.9 67.75 64.75 3

66.25 47.2 15.7333 276.6 1306.8 64.75 62.5 2 25 63.625 37.7 16.7555 314.3 1259.6 62.5 60.25 2 25 61.375 39.3 17.4667 353.6 1221.9 60 25 58 2 25 59 125 41.5 18.4444 395.1 1182.6 58 56.125 1.875 57.0625 36.2 19.3067 431.3 1141.1 56.125 54.25 1.875 55.1875 38.4 20.48 469.7 1104.9 54 25 52.375 1.875 53 3125 40.5 21.6 510.2 1066.5 52.375 50.5 1.875 51.4375 42.3 22.56 552.5 1026 50 5 49 1.5 49.75 35.5 23.6667 588 983.699 49 46.75 2.25 47.875 55.1 24.4889 643.1 948.199 46.75 45.25 15 46 37.3 24.8667 680.4 893.099 45.25 43.75 1.5 44.5 38.8 25.8667 719.199 855.799 43.75 42.625 1.125 43.1875 29.6 26 3111 748.8 816.999 42.625 41.125 15 41.875 41.5 27.6667 790.3 787.4 41.125 40 1 125 40.5625 31 3 27.8222 821.599 745.899 40 38.875 1 125 39.4375 31.7 28.1778 853.299 714.6 38.875 37.75 1.125 38 3125 31.9 28.3555 885 199 682.9 37.75 36.625 1 125 37.1975 32.6 28.9778 917 799 651 36.625 35.5 1.125 36.0625 33.6 29.8667 951.399 618.4 35.5 34 1.5 34.75 44 29.3333 995 399 584.8 34 32.5 1.5 33.25 44 29.3333 1039.4 540.0 32 5 31 1.5 31.75 44 29.3333 1083 4 496.8

-31 29.875 1 125 30.4375 32.9 29.2444 1116.3 452.8 29.875 28.75 1 125 29.3125 32.5 28.8889 1148.8 419.9 28.75 27.25 1.5 28 42 5 28.3333 1191 3 387.4 27 25 26 125 1.125 26.6875 31 1 27.6444 1222 4 344 9 26 125 24.625 15 25.375 40.5 27 1262.9 313 8 24 625 23.125 1.5 23.875 37.4 24.9333 1300.3 273 3 23 125 21 625 1.5 22.375 35.3 23.5333 1335.6 235 9 21 625 19.75 1 875 20.6875 40 21.3333 1375.6 200 6 19.75 17.5 2.25 18.625 41.6 18.4889 1417.2 160.6 17 5 15.25 2.25 16 375 35 15.5555 1452.2 119 15.25 11.875 3.375 13.5625 39.6 11.7333 1491.8 84 11.875 0

11.875 5.9375 44.4 3.73895 1536.2 44.4 _ _ - _ _ _ _ _ _ _ _ _

TABLE 7 YANKEE ROWE OROUP B WORTH FROM PHYSICS TESTING OF JULY 2 1987 TOTAL INTEGRAL WORTH OF GROUP B IS 2 41719 %

INITIAL FINAL DELTA AVERAGE DELTA DIFF.

INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH WORTH INCHES INCHES INCHES INCHES PCM PCN/ INCH 0 TO 90 90 TO O 90 87.25 2.75 88 625 13 4.72727 13 2417.19 l

87.25 82 75 4.5 85 36.6 8 13333 49.6 2404.19 82 75 79.75 3

81 25 31.7 10 5667 81.3 2367.6 79.75 76 3.75 77 875 48.8 13.0133 130.1 2335.9 76 72.625 3.375 74.3125 51 1 15.1407 181.2 2287.1 72 625 69 25 3 375 70.9375 57.4 17 0074 238.6 2236 69 25 66 25 3

67.75 54.3 18 1 292.9 2178.6 66.25 63.625 2 625 64.9375 51.2 19.5047 344.1 2124.3 63 625 61 2.625 62 3125 55 20.9524 399.1 2073.1 61 58.75 2.25 59.875 50.9 22 6222 450 2018.1 58 75 56 5 2 25 57 625 54.8 24.3556 504.3 1967.2 56 5 54 25 2 25 55 375 58.5 26 563 3 1912.4 54.25 52 375 1.875

,53 3125 51.2 27 3067 614.5 1853.9 52 375 50.875 15 51 625 43.4 28.9333 657.9 1802.7 50.875 49.375 15 50.125 45,2 30.1333.

703 1 1759.3 49.375 47.875 1.5 48 625 47 31.3333 750.1 1714.1 47 875 46 375 15 47 125 49.4

'32.9333 799.499 1667.1 46.375 44.875 1.3 45 625 51.9 34.6 851.399 1617.7 44 875 43.375 1.5 44.125 55.3 36.8667 906.699 1565.8 43.375 41.875 15 42 625 57 38 963 699 1510.5 41 875 40.75 1.J25 41 3125 44.9 39.9111 1008.6 1453.5 40.75 39.625 1 125 40 1875 46.9 41 6889 1055.5 1408.6 39.625 38 5 1.125 39.0625 48.1 42.7556 1103.6 1361 7 38.5 37 375 1.125 37 9375 49.3 43.8222 1152.9 1313 6

~37 375 36.25 1 125 36 8125 50 6 44.9778 1203 5 1264.3 36 25 35.5 0.75 35 875 34.7 46.2666 1238.2 1213.7 35.5 34 375 1 125 34.9375 53.5 47.5556 1291.7 1179 34.375 33 25 1 125 33.8125 55 48.8889 1346.7 1125.5 33.25 32 5 0.75 32.875 37.5 50 1384.2 1070.5 32.5 31.75 0.75 32 125 38.1 50.8 1422 3 1033 31.75 31 0.75 31.375 37.3 49.7333 1459 6 994.899 31 29.875 1.125 30.4375 57.9 51 4667 1517.5' 957.599 29.875 29.125 0.75 29 5 39.2 52.2667 1556 7 899.699 29.125 28.375 0.75 28.75 39.3 52.4 1596 860.499 28.375 27.625 0.75 28 39.3 52.4 1635.3 821 199 27.625 26.875 0.75 27.25 39.6 52.8 1674.9 781.9 26 875 26.125 0.75 26.5 39.5 52.6667 1714.4 742.299 26.125 25.375 0.75 25.75 39.1 52.1333 1753.5 702.8 25 375 24.625 0.75 25 38.8 51 7333 1792.3 663.7

'24.625 23.5 1.125 24.0625 56.8 50 4889 1849.1 624.9 23.5 23.125 0.375 23.3125 18 5 49 3333 1867 6 568 1 23 125 22.75 0.375 22.9375 18.5 49 3333 1886 1 549 6 22.75 22.375 0.375 22.5625 18.5 49.3333 1904.6 531.1 22.375 21 625 0.75 22 36.3 48.4 1940 9 512 6 21.625 20.875 0.75 21.25 35.6 47.4667 1976 5 476.3 30.875 20.125 0 75 20.5 34.2 45.6 2010 7 440.7 20.125 19.375 0.75 19.75 33 4 44 5333 2044.1 406.5 19.375 19.625 0.75 19 32 5 43.3333 2076 6 373 1 g,

1 INITIAL FINAL DELTA AVERAGE DELTA DIFF.

INTEGRAL INTEGRAL HEIGHT HEIGHT HEIGHT HEIGHT RHO WORTH WORTH WORTH INCHES INCHES INCHES INCHES PCM PCN/ INCH 0 TO 90 90 TO O 18.625 17.875 0 75 18.25 30.9 41 2 2107.49 340.6 17.875 17.125 0.75 17.5 28.8 38 4 2136 29 309.7 17.125 16 1 125 16.5625 39.9 35 4667 2176.2 280.9 16 14.875 1 125 15.4375 36 9 32 8 2213.1 241 14 875 13.75 1 125 14.3125 33 5 29 7778 2246 59 204.1 13.75 12 625 1 125 13 1875 30 26 6667 2276.59 170.6 12 625 11 125 15 11.875 33 6 22 4 2310.19 140.6 11.125 9.25-1.875 10.1875 32 4 17 28 2342.59 107

'9.25 6.625 2 625 7 9375 34 5 13.1428 2377.09 74.6 6 625 1.375 5.25 4

37 1 7.06667 2414 19 40.1 1.375 0

1 375 0 6875 3

2 18182 2417.19 3

GROUP B JULY 2 1987 k

e----------_

TABLE 8 YANKEE ROWE MODERATOR TEMPERATURE COEFFICIENT (MTC)

RODS TEST MODE TEST DATE MTC (PCM/DEG)

RUN e C00LDOWN 7

1 87

-2 2731 1

HEATUP 7

1 87

-8 6175 ARD 2

COOLDOWN 7*1 87

-2 40499 ARO:

HEATUP 7

1 87

-8 84212 3

ARD 4

COOLDOWN 7

1 87

-3 37803 ARO 5

ARD 7

1 87

-2 14784 HEATUP ARO 6

C90 COOLDOWN 7

2 77

-5 26885 7

HEATUP 7

2 87

-9.0003 8

COOLDOWN 7

2 87

-5.45814 C90 9

HEATUP 7

2 87

-6.36933 C90 10 COOLDOWN 7

2 87

-3.2236 C90 11 COOLDOWN 7

2 87

-4.25538 C90 C90 12 e

-lf.-

,L

TABLE 9 MODERATOR TEMPERATURE COEFFICIENT COMPARISONS (PCM/0F)

(1)

Control Rod Position Predicted Measured Difference ARO

-3.7

-4.6

-0.9 Group C In

-7.2

-5.6

+1.6 S

(1) Average of all measurements performed.

5759R/20.105 -

l i

TABLE 10 POWER PLUS XENON DEFECT DATA 1

L Boron Rod l

Power Mwt Concentration Temperature Position t

05 0

1747 ppa 5160F C 0 23.25" 65%

390.0 1529 ppm 5130F C 0 85.88" 99.7%

598.4 1644 ppm 5310F C 0 81.75" Refer to Table 1 for power plus xenon defect results.

5759R/20.105 _ - _ - - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _

FIGURE 1 YANKEE CORE XIX BOL ASSEMBLY AVERAGE BURNUP l

l 0.

O.

O.

O.

O.

D.

12935.

12220.

O.

O.

O.

O.

19737, 11909.

13060.

17811.

O.

O.

O.

O.

U628.

18398.

10034.

10471.

18171.

19830.

O.

O.

O.

12080.

13101, 10289.

18001.

17806.

9854, 12160.

12964.

O.

O.

12732.

12142.

10160.

17952.

U899.

10278.

13181.

12065.

O.

O.

O.

19875.

18381.

10041.

10047.

18332.

17934.

O.

O.

O.

O.

17977 13228.

12261.

19941.

O.

O.

O.

O.

11812, 12632.

O.

O.

O.

O.

O.

O..

FIGURE 2 Yankee Core XIX Control Rod Identification G

H X

A 3C 3

v i

I g.

i l

i i

j i

_ _ i _ _ _ g _ _ _ g _. _ _

i i

I O

I l

I l

l-I I

17 l

__i--'---2 i

D C

i i

l

,,i---3 I

24 14 i

l

.i B

B_

D i

i

[-

5 6

18 D

B A

B_

23 g

9 3

to

)C A_

A_

C 13 1

2

[15 B

A B

__D.

12 4

,11 19

]

~

D B

B

~

8 7

22

,20 16

___________jo 2'

Key:

A - Shutdown Group A B - Shutdown Group B C - Controlling Group C D - Shutdown Group D - - - - _ - _ _ _ _ _ _ - _ - _ _

09 SEU O

LR 0

,8 V

DE o

RU

)

S a

,0 N R

7 E

W M

o R

R 0

OH a

D T

)

H F

0I T

,6 W R

5 O

O 1

S W

5 O

E o

H

(

L C

A O

R N

0 I

E D

,5 I T

W D

(

N O

a E

P O

N R

O E

O I

F T

R F

,0 I E

I 4

2 S

D O

T D

P G

O E

H K

b P

N U

0 R O

,3 B L

R O

G L

B O

O R

U T

D N

O

,0 O D

2 o

C D

0 O

o 0

1 O

0

~

~

~

0 0

U 1

4 fZ (r Q" rHg2 J :ig5ebo i%:

0 1

9 SEUL R

o d

V DE RU

)

S oN R

'7 E

W M

RR OH o

T

)

F oI d W H

5 T

1 S

R 5

E O

H

[

W C

R N

L o

E R

dI W

(

R O

G P

N E

O T

O I

N R

T I

E oI l

Z S

G O

T P

P O

U H

K O

N RG oR L

dB OB LO RTN oO d

C

'a l

1 o

,~

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 8

6 4

2 0

8 6

4 2

2 1

1 1

1 1

5:@ d$iz" d

3_"

ia ca iN'

FIGURE 5 f

l 8

l D

W J

D i

J W

-8 O

a W

C M

\\

D D

LO C

-R Z W

O z

i Z-C D

M E

O j

O r

E g

^

k O

-89 mg O

l *m O

U")

M O

p*

to B e 0

Z c

O

-@ w w

3 O

g g

Z O

o O

0 0

g w

m

-S y c.

N g

O H

c.

m O

l

=

O O

l I

M O

z 2

O C

_J

-8 m 2

O D

CD O

]

O O

O O

Hz Q

-80 O

o O

O

-3 0

e i

O o

h (HONI/WOd) H180M 19IIN383JJIO

I 1

FIGURE 6 l

i g

m W

D

._J C>

aW M

D V)

C

_O z N

U 3

E C

g O

D I

H L

_g 1 to m

9 W

c' w

I 2

O g

- -Z i

y o

-.- g s 3

O ZO O

M

=~3 H

y

-S

  • N g

O H

O-O I

m u

Z

_1

_Q C n CD O

(D OM HZ

-@ O O

-3

..,..,..,.,,..,...,...:.a (WOd) H180M 198931NI _ _ - _ _ - - _.

D S

U EUL D

AV D

DE D

RUS D

RE o

M O

D D

)

u F

u o

D u

5 os D

1 u

5 D

t

[

e_

u R

D uu E

8

)

W O

\\

d O

J B

P D

0r h

D i

O t

D R

U D

E U

B O

Z D

)

t O

u T

D

]

o O

D t

H b

fl A

d u

Q1 L

U%

\\

O L

B of DOU 0 D

D O

D D

O D

3 a

~_.

~

U 0

0 0

0 0

8 s

4 3

2 1

2OZHDOQ" b S a sHZy E L O

~

i'

SEULRV DE R

US R

E M

0

)

F H'

5 T

1 R

w 5

O

(

W L'

RE N

A W

R O

G P

ET O

N E

R L

I d

E u

r Z

G f

B u

u T

P U

U O

U H

OR G

LOB 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

7 4

1 8

5 2

9 6

3 3

2 2

2 1

1 1

- oS bm@ dgo Z"

)

' FIGURE 9 1

GROSS QUADRANT TILT-INCORE RUN-YR-19-205 150. MWT. GROUP C @ 61.875 INCHES STANDARD ORIENTATION 1.0018

.9793 l.

1.0113 1.0076 o!

DIRECTIONAL ORIENTATION

.970 1.0165

.9938 1.0196 4

Maximum Value = 3.00%

Acceptance Criteria = ! 5.0%

l 5759R/20.105 -

1

)

i l

1 FIGURE 10 COMPARISON OF MEASURED AND PREDICTED SIGNALS INCORE RUN YR-19-207 3

382.8 MWT. GROUP.C AT 85.875 INCHES 71.0 MWD /MTU l

i 0.8242 O.713

-4.3 0.972 0.995

-2.3 1.000 1.010

-1.0

[

1.003 1.150 j

1.022 1.111

- 1. 8 3.5 1.u1 1.116 3.2 1.147 1.113 3.1 1.159 1.016 1.121 1.022 3.4

-0.6 1.022 0.995 1.028 1.003

~0.6

-0.8 0.703 MEASUltED SIONAL 0.747 PftEDICTED St0NAL

-5.9 PE!tCOR DIF7DIENCE

)

AVERAGE ABSOLUTE DIFFERENCE BETWEEN MEASURED AND PREDICTED 2.540 PERCENT RMS ERROR 2.999

.i FIGURE 11-

SUMMARY

=0F INCORE RESULTS YR-19-207 382.8MWT. 70.5 MWD /MTU l

l FRESH FUEL RECYCLED FUEL-Fq (Measured).

.2.309 2.365 Fq (Limit) 4.326 4.326

_i

.% Margin to Limit 46.6 45.3 11 (Measured) 1.559 1.638 FH (Limit) 1.930 1.930

% Margin to Limit 19.3 15.1 LHGR (Measured) 6.246 6.398 l

LHGR (Limit) 10.215 11.125

% Margin to Limit 38.9 42.5 i

5759R/20.105 !

l

I 1

i FIGURE 12 l

YANKEE ROWE CORE XIX

CORE LOCATIONS OF MODIFIED ASSEMBUES i

i 1

2 3

4 4

j 5

8 7

8 9

10 m

.(12) 11 12 13 14 15 16 U

18 M

19 20 21 22 2,3, 24 25 26 27 28 M

M (12) 29 30 31 32 33 34 35 36 37 38 (14) 39 40 41-42 43 44 45 48 47 48 M

-(12)

M M

49 50 51 52 53 54 55 56 57 58 (12) 59 60 61 82 63 64 65 66 (12)

(12) 67 68 69 70 71 72 (12) 73 74 75 78 ASSEMBLY NUNSER (t2)

  1. or REptuso noos.

i

I FIGURE 13 Location of Inert Rods in Reconstituted' Assembly A-SP1-I i

X X

X 1

l O

M X

i 1

OO O QD OX X

X 0 0 000

- Inst. Thimble

- Guide Bar g-se.Guiee.r i

]-InertRod,

4.

o, FIGURE 14 Lattice Locations of Inert Rods and New Guide' Bars i

)

l

. YANKEE ASSEMBLY TYPE A X

X X

Assemblies A775I ll X

A7771

'A773I' A779I-o

+

O O

O

.. o 0

O X

X X.

c YRNKEE'RSSEMBLY TYPE B X

X X

O Q

-. Inst. Thimble O

O Z-outaaSar Ok

[

- New Guide Bar i

D L

- Inert Rod Assemblies

!9 B774I B7721 B778I

" ~

B776I X

X MOOOOOX

_Y.

k p

'1

'l'l

. VI. : REFERENCES 1.

YAEC-l'583 " Core XIX' Performance Analysis".

i

2.

Internal Memo, F. L.,Carpenito to F. Williams, " Yankee Core XIX Startup

/

2 Physics Data",'RP 87 210. June 8, 1987.

3.

. Plant Refueling and Inspection Procedures: OP-1700, 1704, 1705, and 1706.

4.. - Physics Test Procedures: OP-1701, 1702.

I

-i i

l l.

5759R/20.105.

i

l

.a Telephone (617) 872-8100 e -..n..

TWX 710 380-7619 YANKEE ATOMIC ELECTRIC COMPANY y

1671 Worcester Road, Framingham, Massachusetts 01701 m

September 30, 1987 FYR 87-105-United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 l

Reference:

(a) License No. DPR-3 (Docket'No. 50-29)

{

Subject:

Core XIX Startup Program for the Yankee Nuclear Power Station

Dear Sir:

Enclosed is the Core XIX Startup Program for Yankee Nuclear Power Station. This report is submitted in accordance with Yankee Technical Specification 6.9.1.

If you have any questions or desire additional information, please contact us.

Very truly yours, l

YANKEE ATOMIC ELECTRIC COMPANY l

Q) ct ~J George (Papanic, Jt'.

Senior Project Engineer Licensing cc: USNRC Region I USNRC Resident Inspector. YNPS rE24 Ilt

-

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