ML20244D393
| ML20244D393 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/11/1989 |
| From: | May D, Morrissey K, St Laurent N YANKEE ATOMIC ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BYR-89-72, NUDOCS 8904210321 | |
| Download: ML20244D393 (38) | |
Text
{{#Wiki_filter:___________.___-___._, l J 1 1 I Core 20 Startup Program For The Yankee Nuclear Power Station l , March 1989 ! Prepared By: artruo . 4 ' 77 Dennis F. May," Shift Tddhnical Advisor (Date) Reactor Engineering Department Prepared By: bA C\ N /MA> A A>>& 8i KevidJ.Moffissey,' Senior'Nuc1[jlrEngineer '(Date) Nuclear Services Division Reviewed By: 76 M
' J. Cacciapahti, Reactor Physics Manager / (D(te) clear Services Division Reviewed By: /0[87 Frederick N. Williams, Manager (Date)
Reactor Engineering Department Approved By: Y 89 R.A.Mellor,YechnicalDirector (Date) Yankee Nuclear Power Station Approved By: U/Fu i - [n V /N9 I Normand N. St.Laurent, Plant Superintendent Yankee Nuclear Power Station (Date) I Yankee Atomic Electric Company Star Route Rowe, Massachusetts 01367 5759n I 8904210321 890411 PDR ADDCK 05000029 I P PNV l
I IABLE OF CONTENTS PaFe LIST OF TABLES................................................... 11 LIST OF FIGURES.................................................. iii I I. INTRODUCTION..................................................... 1 II.
SUMMARY
OF RESULTS............................................... 2 III. STARTUP PROGRAM - MECHANICAL..................................... 3 A. Fuel Assemblies.............................................. 3 B. Control Rods................................................. 3 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 I V. F. Power Plus Xenon Defect...................................... 8 9 RELOAD DESIGN EVALUATION......................................... VI. REFERENCES....................................................... 33 l I I I I I g
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p LIST OF TABLES Number lille fagn 1 Core 20 Startup Program Physics Testing Results 10 2 Core 19-20 Refueling Control Rod Inspection Results 11 3 Core 20 Delayed Neutron Fractions 12 4 Critical Boron Concentrations 13 5 Group C Worth 14 6 Croup A Worth 15 7 Group B Worth 16 8 Moderator Temperature Coefficient (Measured) 17 9 Moderator Temperature Coefficient Comparisons 18 10 Power Plus Xenon Defect Data 19 I I I I I
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I LIST OF FIGURES I Number Title Eng_e 1 Yankee Core 20 BOL Assembly Burnups (mwd /Mtu) 20 2 Core 20 Control Rod Identification 21 3 Group C Differential Worth 22 4 Group C Integral Worth 23 I- 5 Group A Differential Worth 24 , 6 Group A Integral Worth 25 l 7 Group B Differential Worth 26 8 Group B Integral Worth 27 9 Gross Quadrant Tilt 28 I 10 Radial Power Distribution - Comparison of Reaction Rates 29 11 Summary of Incore Results 30 12 Core Locations of Modified Assemblies 31 13 Lattice Locations of Inert Rods and New Guide Bars 32 I I I I
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I.- INTRODUCTION The Core 19-20 refueling at the Yankee Nuclear Power Station began on November 12, 1988 and was completed with the startup of Core 20 on January 15, 1989. This report provides details cf the Startup Program for Core 20. I 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 the Core 20 physics characteristics during routine plant operation. I I I I l I I 4 I 5759R 1_ I l
l II.
SUMMARY
OF RESULTS l 1 1 Table 1 contains a summary of the Startup Program physics testirig 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 integral worth and the maximum difference in the predicted-to-measured reaction rate in Location C-6. The difference in measured-to-predicted Group A integral worth was acceptable since the total integral worth of all groups measured was within expected tolerances. The difference in I measured-to-predicted reaction rate for Location C-6 was acceptable because this location was not a peak location and still was within the bounds of the current Safety Analysis. I I I I I I I I I I 5759R I I
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I. p . 1 III. STARTUP PROGRAM - MECHANICAL A. Fuel Assemblies Yankee Core 20 is loaded with 36 new zircaloy clad 3.9 w/o fuel assemblies and four new zircaloy clad 3.7 w/o assemblies c.round the perimeter of the core, with 36 once-burned zircaloy clad 3.8 w/o j fuel assemblies in the center region as shown in Figure 1. Sixteen of the fresh assemblies have solid zircaloy inert rods in selected l positions, six per A assembly, and either ten or five per B f I assembly depending on the core location, for a total of 110 inert rods. The B assemblies have one or two special guide bars and the A assemblies have one special guide bar (Figure 13). Spacer stiffener strips are attached to these special 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 I Reference 1. I During the Core 19-20 fuel shuffle, the once-burned cycle assemblies were inspected ultrasonically and visually to check for leaking fuel rods. All assemblies were found to be free of any fuel damage. I Upon completion of fuel loading, assembly positioning was checked by underwater television and video tape. The video tape was then reviewed independently to verify the core loading. I B. Control Rods The Yankee core has 24 Ag-In-Cd control rods with zircaloy followers. The rods are divided into three shutdown groups (A, B, and D) and one controlling group (C) as shown in Figure 2. I 5759R I I _______ _ __. __-_
4 During the Core 19-20 refueling, all 24 control rods were inspected {
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visually. Thirteen rods were checked for bowing in a straightness q 1 gauge. Based on the reaults of these inspections, three control g rods were determined to have excessive bow and wen replaced. All j E 24 control rods were rotated 90 .and returned to the core. Following completion of fuel loading, all control and shim rods were checked for excessive drag force and found acceptable. Following completion of reactor vessel upper internals installation, all control rods were again checked for excessive j 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" to 90" and back to 0" again. Additionally, control rod drop times were measured as a final check that there was no bindinF ar obstruction. The drop l time 1s the interval between when the power is cut to the rod stationary gripper coil until the rod drive shaft passes the 6" coil on the indicating stack. The rod drop times are measured using a calibrated Visicorder. A detailed tabulation of the results of control rod inspection data is shown in Table 2. l I I I I ( I I 4 i I I g
IV. STARTUP PROGRAM - NUCLEAR A. Physics 'Iesting 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 I possible. The variable parameters affecting reactivity include boron concentration, temperature, and control rod position. The correlations derived from this data include boron worth, moderator temperature coefficient, control rod worths, and xenon plus power defect. I 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 I solving the differential Inhour equation. Delayed neutron fractions for Core 20, as eniculated by the Yankee Nuclear Services Division, are programmed into the computer prior to physics testing. Table 3 contains a listing af Core 20 delayed neutron fractions used. Dynamic checks of the reactivity computer are 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 I taken at selected times during the course of physics testing. Multiple samplings and repeated titrations provide a high degree of relir.bility in the boron concentration data. Main coolant temperature is measured by existing calibrated in-plant instrumentation. Incore thermocouple, 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.
Power distribution data is obtained through use of the plant.incore flux mapping system (both moveable and fixed) in conjunction with the CDC Computer System. The incore system controls and computer terminal are located in the Main Control Room. B. Critical Boron Concent. rations
,g Just critical boron concentrations were measured as close as 5 possible to the following conditions:
All rods out Group C inserted I Refer to Table 4 for the results. Note that the measured values I have been adjusted to reflect actual control rod positions to allow one-to-one comparison with predicted values. I 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 rod group motion is then used to compensate or balance this I 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, 6 provide graphical representation of rod group differential and integral worths, respectively. I' I D. Moderator Temperature Coefficient (MTC) MTC data is obtained by varying the moderator temperature and I a measuring the corresponding reactivity change for a minimum of three heatup/cooldown cycles. A linear least square fit of ) l I 5759R l l I __ - - - I l
l temperature change versus' reactivity change yields the moderator I , temperature coefficient. Data sets were taken at various boron concentrations. Control rods were moved to compensate for boron
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changes instead of the burnup compensation which occurs during normal operation. Data was taken as close as possible to the following conditions: I All rods out Group C inserted Table 8 provides a listing of the MTC data as measured, while Table 9 provides a listing of the MTC results compared to the calculated values. Note that the measured values were corrected for control rod positioning to allow for direct comparison to the predicted values. Since the measured HZP ARO MTC was so close to zero (the Technical Specification limit), an administrative limit I on maximum critical boron concentration of 2,025 ppm was imposed to assure compliance to the Technical Specification MTC zero power
-l limit during normal ' operation. This administrative control was reviewed and approved by the Plant Operations and Review Committee (PORC) prior to initial operations.
E. Power Distribution I An incore flux map (YR-20-001) was taken at approximately 25% power to check for gross quadrant tilt. Figure 9 shows the results of the gross quadrant tilt measurement. The maximum tilt was calculated to be within the 5% acceptance criteria. An incore flux map (YR-20-005) was taken at 65.5% power to check relative radial power distribution. Figure 10 shows the comparison of measured versus theoretical integrated fission reaction rates. j One of the measurements in the recycled fuel batch (Core Location C-6) showed a difference of 5.5%, outside the 5.0% i criteria. This assembly was not the peak recycled assembly; and, I therefore, even though it was outside of criteria, it was within 5759R I
the assumptions of the Reference 1 safety analysis. This deviation was assessed by the Nuclear Services Division (Reference 5) and l evaluated by PORC prior to power ascension to a higher power level. Incore flux map (YR-20-005) was also used to check that the LHGR, F , and FAH (nuclear) were within Technical Specification q I limits. Figure 11 shows the results of these measurements which were acceptable. F. Power Plus Xenon Defeqt.
, 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 defect is related to xenon concentration. Plant measurement is made, based on the combined effects of power and xenon reactivity effects and compared to the total predicted defect.
E Primary system data (temperature, boron concentration, rod position, pressure, etc.) was taken at zero power and at two other I power levels (63.6%, 93.8%) during power ascension. A reactivity balance was performed between the zero power data and each of the I 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 that were collected. I I I I I 5759R I I l V. RELOAD DESIGN EVALUATION I As a means of fuel damage prevention, 16 fresh fuel assemblies in the 1 southwest, northwest, and southeast core quadrants were f abricated with solid zircaloy rods and special guide bars with spacer stiffener strips. There were 130 fresh fuel pins and 76 recycled fuel pins replaced by inert rods and gulds bars, resulting in a core total of 174 inert rods and 32 special guide bars. I This lowers the total number of fuel pins from the design value of 17,518 to 17.312; the net effect being a higher core average linear heat generation rate of 4.437 versus the nominal value of 4.395. Figure 12 is provided to show the locations and number of replaced pins for the Core 20 reload design, with Figure 13 depicting the different configurations of the modified fuel assemblies. All of these fuel modifications were assumed in the original reload design analysis and were implemented into the current licensing analysis models. A factor, which had a very small impact on the original core licensing I design calculations, was the core average burnup. The reload design assumed a Core 19 cycle average exposure of 15,750 mwd /Mtu, while the actual value was 15,518 mwd /Mtu. This minor deviation was well within the bounds of the Core 20 licensing calculations performed in support of the reload design. I I I I ' I I 5759R I
TABLE 1 I CORE 20 ETARTUP PROGRAM PHYSICS TESTING RESULTS Predicted Measured Difference or Accept crit. Parameter Value _ygig. 1 Difference Tolerance Control Rod Drop Times 1,91 sec(1) 12.5 see Critical Boron Concen. Aho 2125 ppm 2185 ppm (2) +2.8% 110% Group C In 1894 ppm 1973 ppm (2) +4.2% 110% Control Rod Group Worths Group C 1720 pcm 1696 pcm -1.4% 17.5% Group A 1220 pcm 1314 pcm +7.7% 17.5% Group B 2360 pcm 2353 pcm -0.3% A7.5% Total 5300 pcm 5363 pcm +1.2% 17.5% Moderator Temperature Coef. ARO -1.1 pcm/*F -0.54 pcm/*F(2) +0.56 pcm/'F 15.0 pcm/*F
- Group C In -4.6 pcm/*F -3.04 pcm/'F(2) +1.56 pcm/*F 15.0 pcm/*F Gross Quadrant Tilt 2.5% 5.0%
Radial Power Distribution +5.5% 5.0% (Reaction Rate Comparison) -4.6% Power Plus Xenon Defects 0 - 63.6% Power 2793 pcm 3150 pcm +12.8% 0 - 93.8% Power 3440 pcm 4090 pcm +18.9% I (1) Maximum value. I (2) Corrected for control rod position to allow for direct comparison with predicted values. I 5759R lI. l 1 =__.___-__--___-___-_.
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' TABLE 2 CORE 19-20 REFUELING CDNTROL ROD INSPECTION RESULTS Rod Original Bow Replacement Drop Time Position Serial No. (Inches) _ Serial No. (Seconds)
I 1 A132 .175 - 1.55 2 A156 .125 -- 1.48 3 A151 - - 1.44 4 A157 -- - 1.54 5 A130 -- -- 1.46 6 A142 -- - 1.48 7 A113 .250 -- 1.53 8 A131 -- -- 1.54 A134 .200 1.68 I 9 - 10 A139 .275 A147 1.73 11 A140 -- -- 1.48 12 A133 .340 A145 1.73 13 A137 .200 -- 1.77 14 A146 -- -- 1.53 15 A150 -- -- 1.74 1 16 A135 .300 A152 1.49 17 A138 .160 - 1.56 l 18 A141 .210 1.85 I 19 A143 .060 -- 1.58 20 A148 - - 1.57 21 A136 - -- 1.53 22 A144 -- -- 1.66 23 A149 .100 -- 1.53 l 24 A127 .150 - 1.91 I lI I 5759R 1 'I l
! TABLE 4 CRITICAL BORON CONCENTRATIONS (PPM) Control Rod (1) Position Predicted Measured Corrected Difference I ARO 2125 2174(2) 2185 +2.8% Group C In 1894 1988(3) 1973 +4.2% I (1) Corrected for control rod position to allow for direct comparison with predicted values. (2) Group C @ 76.625 inches withdrawn. (3) Group C @ 15.125 inches withdrawn. I I I I I I I I I 5759R I
TABLE 3 i YANKEE CORE 20 I)ELAYED EUTRON FRACTIONS FRACTION EFFECTIVE LAMBDA GROUP BETA BAR FRACTION (SEC)-1 1 .00019119 .00019150 .01252 2 .00135798 .00136048 .03056 3 .00124172 .00124385 .11495 4 .00254580 .00254988 .30900 5 .00087218 .00087375 1.16258 I 6 .00031084 .00031144 3.04634 iI BETA EFFECTIVE = .006531 BETA BAR = .006520 I BAR = 1.001715 PROMPT NEUTRON LIFETIME = 19.89 MICROSECONDS J STARTUP RATE PERIOD REACTIVITY (SEC.) (PERCENT) _I II)ECADES / MIN. )
.100 260.6 .0271 1 .500 52.1 .0984 e
1.000 26.1 .1527 2.606 10.0 .2493 I I I I 5759R
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TABLE 8 I 1 MODERATOR TEMPERATURE COEFFICIENT (MEASURED) BOL. H2P. Group C @ 76.625 Inches i l Condition MTC (PCM/ Decree Fl. Heatup No. 1 -4.231 Heatup No. 2 1.033 He8 tup No. 3 -1.385 5 g Cooldown No. 1 .481 Cooldown No. 2 -2.333 Cooldown No. 3 2.552 I Average .647 ) I BOL. HZP. Group C @ 15.125 Inches 1q
)
MTC (PCM/ Degree F) I Condition 1 Heatup No. 1 .717 l Heatup No. 2 -6.7
-1.704 I. Heatup No. 3 -
Cooldown No. 1 -3.346 Cooldown No. 2 -3.42 g Cooldown N O -1.649
-2.923 W Average I
I I I I I I 5759R I
I TABLE 9 MODERATOR TEMPERATURE COEFFICIENT COMPARISONS-l (PCM/*F) Control Rod Boron (1) Position Concentration Predicted Measured Corrected Difference ARO 2174 -1.1 .65(2) .54 +0.56 i Group C in 1988 -4.6 -2.92(3) -3.04 +1.56 I I (1) Average of all measurements performed and corrected for control rod position to allow for direct comparison with predicted values. (2) Group C @ 76.625 inches withdrawn. (3) Group C @ 15.125 inches withdrawn. I I I I I I I I I 5759R E
) L. . TABLE 10 POWER ELUS XENON DEFECT DATA Boron Moderator Rod Power My1 Concentration Temperature Position 0.0% 0.0 2013 ppm 517.5'F C @ 23.75" l 63.6% 382.0 1757 ppm 510.0*F C @ 80.63" 93.8% 563.0 1656 ppm 521.0*F C @ 84.50" I Refer to Table 1 for power plus xenon defect results. I i I 8 J I I I I I I I 5759R E
l I- < FIGURE 1 -l YANKEE CORE 20 BOL ASSEMBLY AVERAGE BURNUP I
, e. e. e.
x e. 0. I O. 12172. O. O. O. O. O. 12094. 11943. 17606, 17659. O. O. ; I o- o. -e . , o24- ==. , ~ '. w' >- ,e. o. o. I e. 0.
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I 0. 12061. 1218 0. 9949. 17520. 17061. 12489. 17858. O. x O. I 0. O. 12652. 9979. 12383. 10146. 10002. 17214. O. O. I 0. O. 17446. 18117. 12063. 12663. O. O. N. I x
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. g. Yankee Core 20 Control Rod Identification A 3C D - - G - J K I l I I I I I I I I I I I
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l FIGURE 9 ~I GROSS OUADRANT TILT 1 INCORE RUN YR-20-001 162 MWT. CROUP C 0 53.5 INCHES E,- : STANDARD ORIENTATION 1.0099 .9942
.9923 1.0036 I - \
I DIRECTIONAL ORIENTATION i
- .9904 1.0234 1.0110
.9751 I
I Maximum Value = 2.49% I Acceptance Criteria = 5.0% I E I E 5759R I
I 1 I l FIGURE 10-COMPARISON OF MEASURED AND PREDICTED SIGNALS g INCORE RUN YR-20-005 393.0 MWT. GROUP C AT 80.250 INCHES 50. MWD /MTU E _ 0.702 I 0.742
-5.4 1.192 0.669 I 1.249
-4.6 0.696
-3.9 1.006 I 1.13 3 1 I
1.094 1.005 3.6 0.2 I 1.001 0.997 0.4 1.10 3 1.055 4.6 1.101 1.110
-0.8 1.054 1.070 1.008 1.064 4.6 0.5 1.12 2 1.038 1.064 1.005 3.3 I
5.5 1.0 87 0.999 1.056 1.006 2.9 -0.6 1.065 , 1.0 87
-2.0 ,
MEASURED SIGNAL j 0.680 1.2 17 g 0.725
- 6.2 1.266
-3.8 PREDICTED SIGNAL PERCENT DIFFERENCE l g O.757
' O.770 j T -1.7 AVERAGE At3EOLUTE DIFFERENCE BETWEEN lI MEASURED AND PREDICTED 3.039 PERCENT l RMS ERROR 3.593 lI FIGURE 11 I
SUMMARY
OF INCORE RESULTS YR-20-005 q 393 MWT. 50 MWD /MTU
,I FRESH FUEL RECYCLED FUEL j
'Fq (Measured) 2.470 2.381 Fq (Limit) 4.214 4.214
% Margin to Limit 41.4 43.5 1 FAH (Measure.() 1.621 1.570 FAH (Limit) 1.924 1.994
% Margin to Limit 15.8 18 4 LHGR (kW/ft) (Measured) 6.910 6.660 LHGR (kW/ft) (Limit) 10.129 11.129
% Margin to Limit 31.8 40.2 I
I I I i 1 E 5759R E t
l I FIGURE 12 l- YANKEE CORE 20 l CORE LOCATIONS OF MODIFIED ASSEMBUES I 1' 2 3 4 A-1 M 7 8 9 10 5 6 I A-1 A-1 M M 12 13 14 15 16 # 18 11 I< , A-1 B-1 M (12) 24 25 26 27 28 19 20 21 22 23 A-1 M > 37 38 32 33 34 35 36 ' 29 30 31 B-1 A-1 B-2 M (12) (6) I 39 A-1 40 B-1 41 42 43 44 45 46 47 48 M (12) 58 54 55 56 57 49 50 51 52 53 A-1 B-1 A-2 B-1 A-1
@ (12) M (12) M 62 63 64 65 66 59 60 61 A-2 B-1 g M 5 (12) 68 69 70 71 72 67 A-2 0-1
( M _ 12) 73 74 75 76 ASSEMBLY NUMBER B-1 A-2 B-3 A-2 ASSEMBLY TYPE (12) M (6) M f 0F REPLACED RODS E I __ I
1 . g FIGURE 13 l'
. YANKEE CORE 20 LRTTICE LOCATIONS OF INERT RODS AND NEW GUIDE BARS I N RSSE 2 LY TYPC R 1 YliKKCC MSCnBLY.TYPC 6 2 X l X X X X X I _
W X X
] X J X X X I X XDDDC0CE I x YllNKIC RSSD1DLY TYPC B ]
x x x YANKEC ASSD1DLY YYPC D 2 x x t j I 3-- 8 o a* x + x g X X X X R X Cuaoox x I x _ .S_, X YY,C . , X l D-om.vnmu
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< ~I))))]R
I l VI. REFEREtiCES
- 1. YAEC-1652, " Yankee Nuclear Power Station Core 20 Performance Analysis."
I 2. Internal Memo, M. E. Napolitano to F. Williams, " Yankee Core 20 Startup Physics Data", RP 88-375, November 28, 1988.
- 3. Plant Refueling and Inspection Procedures: OP-1700, 1704, 1705, and 1706.
- 4. Physics Test Procedures: OP-1701, 1702.
I 5. Internal Memo, K. J. Morrissey to F. N. Williams, " Analysis of Incore Run YR-20-005," RP 89-29, January 20, 1989. I I , I I I I I I : lI 5759R i I
F a YANKEE ATOMICELECTRIC COMPANY l
"'fl "*[o".ggo';*;"l" l 580 Main Street, Bolton, Massachusetts 01740-1398 April 14, 1989 j BYR 89-72 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Reference:
(a) License No. DPR-3 (Docket No. 50-29)
Subject:
Core 20 Startup Program for the Yankee Nuc1 car Power Station
Dear Sir:
Enclosed is the Core 20 Startup Program for Yankee Nuclear Power Station. This report is submitted in accordance with Yankee Technical % Specification 6.9.1. If your have any questions or desire additional information, please contact us. Very truly yours,
.A YANKEE ATOMIC ELECTRIC COMPANY j Georg apanic, Jr Senior Project Engineer Licensing GP/ pac /0372u cc: USNRC Region I USNRC Resident Inspector YNPS I
If i}}