ML17320B024

Forwards Summary Rept of Cycle 8 Startup & Power Escalation Testing,Per Tech Spec 6.9.1.1,Items 2 & 3

ML17320B024
Person / Time
Site:	Cook
Issue date:	02/06/1984
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
AEP:NRC:0745K, AEP:NRC:745K, NUDOCS 8405040264
Download: ML17320B024 (107)

Text

INDIANA8 MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 KJ'ILE cgpyi February 6,

1984 AEP: NRC: 0745K Donald C.

Cook Nuclear Plant Unit No.

Docket No. 50-315 License No. DPR-58 STARTUP REPORT FOR UNIT 1

CYCLE 8 PRI IPAL STAFF

, RP D/RA Mr. James G. Keppler, Regional Administrator U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137

Dear Mr. Keppler:

/RA

~AO SGA ORMS DRMA SCS File This letter and its Attachment transmit a summary report of the Unit 1

Cycle 8 startup and power escalation testing.

Th1s submittal is being made 1n compliance with Techn1cal Specification No.

6.9.F 1, Items 2 and 3.

This document has been prepared following Corporate Procedures which incorporate a reasonable set of controls to ensure its accuracy and completeness prior to s1gnature by the undersigned.

Very truly yours, eked

'$1 I<'i MPA/bgs M.P. Alexich Vice President cc: John E. Dolan M'.G. Smith, Jr. - Bridgman R.C. Callen G. Charnoff E.R.

Swanson, NRC Resident Inspector Bridgman i 8~OSO~O2~~

S~O~O~

PDR ADOCK 050003i5 P

PDR FEB 8 1984

0

~

Mr. James G. Keppler AEP:NRC:0745K bc:

J.G. Feinstein/J.M.

Cleveland/W.L.

Zimmermann H.N. Scherer, Jr./S.H.

Horowitz/R.C.

Car ruth R.F. Hering/S.H. Steinhart/J.A.

Kobyra R.F. Kroeger T.P. Beilman - Bridgman J.A. DiBella R.W. Jurgensen B.H. Bennett/F.S.

VanPelt, Jr.

J.F. Stietzel - Bridgman J.B. Shinnock D. Wigginton - NRC AEP:NRC:0856A AEP:NRC:0584A DC-N-6500

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TABLE OF CONTENTS SECTION TITLE IXX INTRODUCTION CORE LOADING AND FUEL INSPECTION INITXALCRITICALITY IVa ZERO POWER PHYSICS TESTING

,LOW LEAKAGE LOADING PATTERN (L P)

INVERSE COUNT RATE RATXO (ICRR)

PLOT SHAPES FOR UNXT 1 CYCLE 8 PROBLEMS ASSOCXATED WITH FLUX MAPPING SYSTEM Va POWER ASCENSION TESTING PLANT CHEMISTRY HISTORY PLANT RADIATION SURVEYS REACTOR COOLANT FLOW MEASUREMENT PLANT THERMAL POWER CALIBRATION

I Introduction The Unit 1 Cycle 7-8 outage began'ith reactor shutdown on July 16, 1983, after a twenty-eight (28) day power coastdown re-sulting in a total cycle 7 burnup of 10,446.8 MWD/MTU and final boron concentration at essentially zero ppm.

One-hundred per-cent (100%)

RTP was again reached on November 7,

1983.

Startup testing was completed on November 8, 1983, with a full power flux map.

The fuel shuffle sequence for this outage was the typical shuffle.

The fuel shuffle began on August 18,

1983, and was completed on August 23, 1983.

Other than typical mechanical problems and one fuel shuffle sequence error, the shuffle was uneventful with no damaged fuel assemblies detected.

A total of eighty (80) Westinghouse Optimized Fuel Assembles (OFA) were placed in the core with sixty-eight (68) of these fuel assemblies containing Wet Annular Burnable Absorbers (WABA).

For the first time in Unit 1, the fresh assemblies were loaded into the core in a low neutron leakage pattern.

Also for the first time, as a

a result of the WABA placement, the irradiated secondary sources were placed in fuel assemblies on the periphery of the core near the source range detectors.

The placement of the secondary sources re-sulted in extensive analysis of their effect on the Inverse

Count, Rate Ratio (ICRR) during the approach to criticality.

The approach to criticality started with shutdown banks with-drawal at 0133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br /> on October 19, 1983.

Dilution toward criticality was halted at, approximately 1700 ppm pending Plant Manager's approval prior to entering Mode,2.

As a result of a problem with the Rod Position Indicators (RPI's), dilution to criticality from 1700 ppm was delayed approximately thirty (30) hours.

Dilution resumed at 1812 hours0.021 days <br />0.503 hours <br />0.003 weeks <br />6.89466e-4 months <br /> October 20,

1983, and criticality was achieved at 2103

hours, October 20, 1983.

After stabilizing the reactor, data was obtained to determine the Zero Power Physics testing range and the point of nuclear heat,.

Zero Power physics testing began at 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on October 21,

1983, and included the usual all rods out (ARO) Isothermal Temperature Coefficent,(ITC), control rod worth, boron worth and boron endpoint tests along with ARO flux map.

The measured ARO ITC was essentially the same as the design value, therefore, no rod withdrawal limits were required to ensure a negative MTC above 70%

RTP.

All rod worth and boron endpoint measurements compared favorably with design values.

Power ascension'esting followed, starting on October 24,

1983, with a 30%

RTP flux map, followed by a 48%

RTP incore/excore cross calibration, Moderator Temperature Coefficient, Doppler Power Co-efficient, hT/hP ratio measurements and several flux maps.

Power ascension went smoothly with the only uplanned hold being at 95%

RTP, as the three-dimensional analytical factors were unavailable

for flux map processing.

On November 4, 1983, the three-dimensional analytical factors were used in processing the 95%

RTP flux map which resulted in an allowable power level (APL) greater than 100%

RTP.

The reactor power level was subsequently increased to 100%

RTP.

h In general, all startup tests were relatively routine.

They were conducted in a timely and expedient manner and resulted in accurate startup information and data that compared favorably with design expectations and met acceptance criteria.

As stated in section 6.9.1.2 of Unit 1 technical Specifica-tions, the tests identified in the FSAR shall be addressed in the Startup Report.

The tests in the FSAR are tests which were performed at the beginning of Unit 1 Cycle 1.

Not all these tests, need to be performed on a reload cycle.

The FSAR tests that were required to be performed on this reload core are addressed in de-tail in this report.

Those FSAR tests not required to be performed on this reload core are addressed in the Unit 1 Cycle 1 Startup Report.

XI CORE LOADING AND FUEL INSPECTION The Unit 1 Cycle 8 fuel assembly shuffle sequence commenced at 1250 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.75625e-4 months <br /> on August 18,

1983, and was completed at 2245 hours0.026 days <br />0.624 hours <br />0.00371 weeks <br />8.542225e-4 months <br /> on August 23, 1983.

Eighty (80) depleted fuel assemblies (sixty-five [65] region G and fifteen [15] region H, all Exxon 15 x 15) were removed from the core.

The depleted assemblies were replaced with eighty (80) Westinghouse 15 x 15 (region K) assemblies; The Westinghouse fuel assemblies are of the Optimized fuel assembly (OFA) design and many contain the recently designed part length Wet Annular Burnable Absorbers (WABA's).

Both the OFA's and the WABA's are designed to increase the neutron economy and the cycle length of the core.

Westinghouse's safety analysis of the reload fuel has shown the OFA's are mechanically and hydraulically compatible with the Exxon fuel assemblies, and with the control rods and reactor internals interfaces.

A more detailed description of the new OFA's and WABA's is in the attached Appendix IX A.

The remaining one-hundred-thirteen (113) assemblies consisted of forty-nine (49) region H and sixty-four (64) region J fuel assemblies, all Exxon 15 x 15.

Core loading diagrams for Unit 1 Cycle 7 and Cycle 8 are shown in Figures II.l and II.2 respectively.

(Core loading pattern diagrams, illustrating the difference between Cycle 7's core de-sign and Cycle 8's new low-leakage core design, are shown in Figures II.3 and II.4 respectively.)

The first fuel shuffle problems occurred at. approximately 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br />, on August 18, 1983.

The "frame down" light would not activate although the upender frame was down. and the winch had shut off.

After several

seconds, the light activated,

however, the winch cable then fell off the winch into the water.

Apparently the cable had unwound to a point such that the "ball" that holds it was able to fall out.

A diver was sent down to repair the winch while repairs were made to the.limit and proximity switches.

The system was declared operable at approximately 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on August 19, 1983.

At approximately 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br /> on August 22,

1983, an error was discovered in the fuel shuffle procedure.

An unusual pattern for the shuffle had been necessary because of the new WABA loading pattern and the addition of two (2) new secondary sources.

This

pattern, however, would not allow core subcritical behavior to be properly monitored on both source range channels while an activated secondary

source, SSS, was absent from the core.

Minor changes to the procedure corrected the error and the shuffle continued.

This problem again occurred in the sequence involving SS6 and was handled in similar manner.

J

'K

At approximately 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> on August 23,

1983, a problem developed with the fuel transfer system.

The pneumatic system air pressure dropped as the transfer cart traversed from containment to the spent fuel pit.

Later, this same problem developed on the re-verse path.

After much discussion, it was decided to resume the shuffle to complete the remaining steps (approximately

10) without repairing the system.

Problems later developed with the solenoid, which had to be repaired.

The final steps consumed a great deal of time as the system further degraded.

At this time, repairs have been made to the transfer system.

The solenoid has been replaced with a new solenoid.

Westinghouse Fuel Assembly Handling Deviation Reports, FAHDR's, numbered approximately the same as the Cycle 7 fuel shuffle (27 versus 24).

With the exception of the previously noted inci-

dents, the majority of the reports involved problems associated with the insertion of assemblies into their specified core posi-tions because of minor bowing or because of in-place assemblies leaning into open core locations.

Some FAHDR's were the result of minor adjustments of the manipulator crane index.

Binocular inspection of new and irradiated assemblies took place both in the containment and in the spent fuel area, thus optimizing inspection conditions.

All irradiated assemblies were observed.from as many angles as possible as they were removed from the core.

Additionally, those assemblies discharged to the spent fuel pit and all new assemblies underwent a more thorough inspection in the spent fuel pit.

Each assembly was slowly rotated, thus allowing the observer a more careful inspection.

All assemblies placed into the core were inspected as they were inserted as a final check for any damage.

The binocular inspection of the assemblies revealed no structural damage or abnormalities.

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Appendix II A Westin house 0 timized Fuel Assemblies (OFA's) and Wet Annular Burnab'le Poison Absorbers (WABA's)

Westinghouse, in order to demonstrate compatibility of the OFA with ENC fuel, chose the following approach.

ENC, in es-tablishing their fuel design, demonstrated compatibility with the Westinghouse LOPAR fuel used in Unit 1 Cycle 1.

Westinghouse has shown compatibility of the 15 x 15 OFA design with the LOPAR design.

Thus, compatitbility between OFA and ENC fuel was shown.

The purpose of the new OFA design was.to improve the utili-zation of Uranium while allowing operating margins to be maintained The 15 x 15 OFA design is similar to the 15 x 15 LOPAR (low parasi-tic) design which was the initial Unit 1 fuel.

The major difference in the two designs is that the five intermediate inconel grid straps of the LOPAR fuel were replaced with zircaloy grid straps.

The very low thermal neutron cross-section of zircaloy makes it less parasitic and hence results in better neutron economy.

The zircaloy straps are wider and thicker than the inconel grid straps in order to duplicate the inconel grid strength.

The ENC grid straps are bi-metallic, consisting of zircaloy-4 straps with inconel springs.

The grid heights of both OFA and ENC assemblies are equal.

A com-parison of Westinghouse 15 x 15 OFA and Exxon 15 x 15 assemblies appear in Figure II.5 and Table II.l.

The OFA rods use the same design as the LOPAR rods.

As the zircaloy grid straps are thicker than the inconel, there has been a reduction in grid strap cell size.

This reduction resulted in the OFA guide thimble tubes inner diameter above the dashpot being smaller than the ENC thimble tube inner diameter.

Below the dashpot, the dimensions are the same.

The reduced dia-meter provides sufficient space for all inserts;

however, the Technical Specification rod drop time limit has been increased to allow for anticipated increases in rod drop times.

All accident analyses used the lengthened rod drop time.

Minor differences between ENC assemblies and OFA's include

1) the overall height of top and bottom nozzles,

2) the adapter plate flow-slot configuration and holddown leaf springs.

These differences have no adverse affect on the interaction of the two (2) types of assemblies.

The OFA's use a 3-leaf holddown spring design (as previously used in LOPAR's) which provides additional holddown force margin compared to the 2-leaf ENC assembly spring.

The OFA bottom nozzle, although similar in design to the ENC bottom

nozzle, has a reconstitutable feature.

This feature allows it to be easily removed.

(See Figuie II.5)

The OFA design utilized a locking cup to lock the thimble screw of a guide tube in place, where as the LOPAR design utilized a lock-wire.

This feature facilitates remote removal of the nozzle and relocking of the thimble screws as the nozzle is reattached.

'a L<

The burnup dependent rod bow for the OFA design was conserva-tively assumed to be the same as for the LOPAR design.

Significant rod bow is mainly caused'y rod-grid and pellet clad interaction forces and wall thickness variation.

OFA fuel rods and LOPAR rods are the

same, thus eliminating the rod consideration.

Grid forces are reduced because of the zircaloy springs in OFA, thus rod bow can be predicted to decrease.,

Fuel rod cladding wear is dependent on grid support and flow environment.

Hydraulic tests between ENC and LOPAR assemblies and OFA and LOPAR assemblies showed similar crossflows between the two sets.

The results showed small cross-flow between assemblies and no significant rod wear due to vibration.

Extrapolation showed that clad wear would not impair fuel rod integrity.

The purpose of the new WABA design was to reduce residual poison penalty at the end of cycle.

This in turn could

1) aid in reducing initial boron concentration and thus maintain a negative beginning of cycle moderator temperature coefficeint and

2) aid in extending cycle length.

Additionally, the. WABA's aid in flattening the power distribution and in controlling power peaking.

The im-proved WABA's are different from the older burnable poison rods in the following ways:

1) aluminum oxide-boron carbide is used instead of Borosilicate glass,

2) tubings are made of zircaloy instead of stainless

steel,

3) the annular plenum contains helium instead of air and

4) the inner tube is open-ended allowing the reactor coolant flow through it instead of being filled with air.

Additionally, the WABA rods are reduced in length to 123 inches and its centerline is positioned 1.5 inches above the fuel rod centerline.

Thus the design benefits of the WABA are better neutron economy from the less parasitic zircaloy tubes, the increased water fraction in the cell which increases the thermal flux in the cell which increases the effect of the poison and the reduction on the end of cycle boron penalty.

II A-2

Table II.'1 Com arison of OFA and EHC Assembl Desi n

Parameter 15xl5 M

15xl5 Optimized Fuel EHC Fuel Fuel'Assembly Length, in.

Fuel Rod Length, in.

Assembly Envelope, in.

Compatible with Core Internals Fuel Rod Pitch, in.

Number of Fuel Rods/Ass'y Number of Guide Thimbles/Ass'y Number of Instrumentation Tube/Ass'y Compatible with Movable In-Core Detector System Fuel Tube Material Fuel Rod Clad OD, in.

Fuel Rod Clad Thickness, in.

Fuel/Clad Gap, mil Fuel Pellet Diameter, in.

Guide Thimble Material Guide Thimble ID, in."

Structural Material Five Inner Grids 159.765 151.85 8.426 Yes 0.563 204 20 1

Yes Zircaloy-4

,0. 422 0.0243 7.5 0.3659 Zircaloy-4 0.499 Zircaloy-4 159.71 152.07 8.426 Yes 0.563 204 20 1

Yes Zircaloy-4 0'24 0.030 7.5 0.3565 Zircaloy-4 0.511 Zircaloy-4 Straps Inconel Springs "Above'ashpot

APPENDIX Table, II.g

,'(coni:inued)

Parameter 15xl5 W

Optimized Fuel Assembl Oesi n

15x15 ENC Fuel Structural Haterial-Two End Grids Inconel Zircaloy-4 Straps, Inconel Springs Grid Height, in., Outer

Straps, Valley-to Valley 2.25 2.25 Bottom Nozzle Top Nozzle Holddown Springs Reconstitutable 3-leaf 2-1 ea f

FIGURE II' SCHEhIATIC OF WESTINGHOUSE 15X16 OFA 159.765 (W) REF.

159.710 (EN C) 161.85 (W) REF.

152.07 (ENCI 2.738 (W) REF.

2.72 (ENC) 3 LEAF SPRING IW) 2 LEAF SPRING (ENC) 3.55 FWI REF.

3.48 (ENCI 15331 (W) REF, 153.26 (ENC) 134.63 REF.

108.44 REF.

82.26 REF 66.06 REF C GRIDTYP 29.87 REF 6.668 (W) REF.

6.645 (ENC)

W - WESTINGHOUSE 15X15 OPTlhIIZEO FUEL ASSEhIBLY (OFA) DlhIENSION ENC ~ EXXON NUCLEAR COMPANY (ENC) 15X15 FUEL ASSEhIBLY OlhIENSION NOTE: OFA AND ENC ASSEhIBLY hIIO GRIOS HAVE IDENTICALAXIALSPACINGS ENC GRID HEIGHT -2.25 WESTINGHOUSE TOP tc BOTTOM GRID HEIGHTS ~ 1.6 WESTTNGHouSE hIID GRID HEIGHT ~ 2.25 Comparison of ENC Fuel Assembly Dimensions With Vlestinghouse 15X15 OFA Schematic G

I

$ s 1

~

t3 4

III INITXALCRITICALITY Unit 1 Cycle 8 achieved initial criticality at 2103 hours0.0243 days <br />0.584 hours <br />0.00348 weeks <br />8.001915e-4 months <br /> on October 20, 1983.

The All Rods Out (ARO) Boron was cal-culated to be 1572.4

ppm, as compared to a design value of 1534 ppm.

The approach to criticality began with the withdrawal of the Shutdown Banks.

Shutdown Bank withdrawal began at 0133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br /> and was completed at 0219 hours0.00253 days <br />0.0608 hours <br />3.621032e-4 weeks <br />8.33295e-5 months <br /> on October 19, 1983.

The next step was to withdraw control banks in overlap, as shown in Figure XII.l.

The withdrawal of the control banks started at 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> and was finished at 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> on October 19,

1983, with Control Bank D-(CBD), at 190 Steps (CBD at.

190 Steps corres-corresponded to approximately 100 pcm of negative reactivity still inserted in the core).

At 0420 hours0.00486 days <br />0.117 hours <br />6.944444e-4 weeks <br />1.5981e-4 months <br />, after the withdrawal of CBD to 190 Steps, RCS dilution began from approximately 2220 ppm to approximately 1700

ppm, at. a dilution rate of 60 gpm.

Because of the anticipated behavior in the Inverse Count Rate Ratio, ICRR, due to secondary source placement in the coze, see Section IVc, the dilution rate was cut to 30 gpm at 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on October 19, 1983.

Dilution was stopped, pending Plant Managers approval to enter Mode 2, at 1032 hours0.0119 days <br />0.287 hours <br />0.00171 weeks <br />3.92676e-4 months <br /> on October 19, 1983, at a boron concentration of 1729 ppm. While awaiting the Plant Managers approval to enter Mode 2, Control and Instrumentation (CGI) personnel performed THP 6030 IMP.038, A.R.P.X. Coil Stack Voltage Data.

During the performance of this procedure, four (4) rod position indicators (RPI's) were determined to be out of specification.

After a hold approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, during which Nuclear Section personnel performed a "glitch" test to verify the actual rod height of the RPI's in question and C&I personnel completed a recalibration of all of the RPI's, permission to dilute to critical was granted.

Dilution to critical was reinitiated at a rate of 45 gpm at 1812 hours0.021 days <br />0.503 hours <br />0.003 weeks <br />6.89466e-4 months <br /> on October 20, 1983.

Mode 2 was achieved at 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br /> and the Unit was subsequently declared critical at.

2103 hours0.0243 days <br />0.584 hours <br />0.00348 weeks <br />8.001915e-4 months <br /> on October 20, 1983.

The Reactor's stable critical conditions were

-8 CBD at 169 steps, flux level at 10

amps, and a boron concentration of 1552 ppm.

During the dilution to critical Xnverse Count Rate

Ratio, ICRR, data vs. Boron Concentration, Primary Water, and

. time were plotted, and is shown in Figures III.2, III.3, and III.4, respectively.

-8 After the Reactor was stabilized at 10

amps, data was ob-tained to determine the Zero Power Physics Testing Range and

, Nuclear Heating Level.

CBD was maneuvered between 169 and 176

Steps, at different flux levels, in order to obtain reactivity

changes, (see Table XXX.1)

From the data obtained, the.Zero Power Physics Testing Range was determined to be from

-8

-7 3 x 10 amps to 3 x 10 amps.

See Figure III.5.

Nuclear Heating was determimed to be at approximately 7 x 10 am'ps.

Nuclear Heating determination is made by observing a decay in the reactivity trace on the, reactivity meter strip chart.

At the same time, an increase in RCS temperature is noted, as the flux is slowly raised.

The physics testing range is typically approximately two (2) decades above the y

background level, and approximately one (1) decade below Nuclear Heating.

The y

background is measured prior to the introduction of positive reactivity for the approach to critical.

At lower flux levels the reactivity changes is affected by small variations in flux.

See Figure III.5.

This is due to the fact. that the background remains constant while the flux can be varied.

At lower flux levels the background is more predominate.

As the flux is increased, the background becomes less significant.

A plateau is reached where the background, does not have a signi-ficant effect on the observed reactivity.

Table III.1 Subcritical Data Induced current with 1000. V appli'ed

= + 0.14 x 10 amp 90$ 'value

= 0.126 x 10 amp

-8 508 value

= 0.07 x 10 'mp Zero Power Ph sics Test Ran e Data Flux Level, am s

CBD initial position 169 steps, CBD final position 176 steps

(~p, pcm, for various compensating currents)

Disconnected 50% value 90% value 5 x 10

-8 1 x 10 x 10 x 10 5x10 x 10 2 x 10

21. 6 24.3 26.1 26.S

26. 5.,

25.7 25.6 24.5 25.5 27.0 26.0 25.7 25.4 26.0

27. 7 26.2 26.0 26.0 25.-2 24.7 25.0 26.25 26.4 CBD initial position 203 steps, final CBD position 211 steps "0.3 x 10 26.1 27.0 27.4

-0.3 x 10 25.8 Nuclear Heatin Level Flux Level =

7 x 10 amps Zero Power Ph sics Testin Ran e*

Flux Range

= 0.03 x 10 amp to 0.3 x 10 amp

• Due to the quality of the data obtained, no compensating current was required during zero power physics testing.

'1 i

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FXGURE XXI.2 1.0 4 UNIT 1 CYCLE 8 ICRR vs. BORON CONCENTRATION p 5- ~ 9 .8 ~ I ~ ~ 7 J ~ 5 4' 3 .2 4I ~ 1 7 2200 2100 2000 1900 1800 BORON CONCENTRATION (pPm) 1700 16 0 ypL ]500 1523 CBD 8 190 Stops

'I ~ 4

FIGURE III.3 UNIT 1 CYCLE 8 ICRR vs. PRIHARY WATER (thosands of gallons) j tj h I l I ~ ~ I I pped Sto dilu io 10 I 1 I 10 18 19 F 7/20 PRIHARY WATER (thousands of gallons) 22 22,9 24 26/26. 1 28 0 h

UI ~4 t Pl 4 P

1.0 P URZ XIX.4 UNIT 1 CYCLE 8 ICRR VS. TINE 0. 0. 0. 0. o0 M 0 cd di 0 2 h urs 0 0. 0. 0. 0400 0500 0 600 0700 0800 0900 1000 00 1900 2000 2100 2200 2300 2400 2500 2600 10/19/83 Dilution time (hours) 10/20/83

PIGURE III.5 Unit 1 Cycle 8 I I )F (-"1 '>(4(- tV Q No compensation current F 50% comp. current 90% comp. current %! Q tt Qi . F'=Itt tt r ~ .F It

':LL

I '", iit ~ ~ I ~ ~

.F=- r(-.=(= -"i= =~

I ~ tt 1 x 10 ~ ~ I I t ~ ~ I 'I I ~ ~ I I I I Nuclear Heat -FIFER( t ! I-.it I I-I I

r. t t

'1 .I .=* ~ '1== ~ I 1 x10 7.. ~ ~ ~ t I'M + tt It w FF I' Pirl ( l -8 1 z 10 tt. r( r I '. ~-- ~ ~ I F I=: I:PW IIF t ~ I I lt ;t F.'I ~ g ('( t ( I t A t 'i; .E i:-1 = I 1 x10- 'I ~ ~ t ~ I I I ~ -:E=~P 20 Reactivity (pcm) ~ t ~ 25 30 35

IVa ZERO POWER PHYSICS TESTING Zero power physics testing commenced at 0600 hours on October 21,

1983, and was completed on October 22 at 1025 hours.

The testing program consisted of 1) ARO Isothermal Temperature Coefficient 2) ARO Flux Map 3) Individual Control Bank Worths 4) Boron Endpoints and Boron Worths 5) Control Banks. Worth in Overlap The testing program was routine and was completed with no problems. The data presented in the following tables and figures shows'compliance with design values and acceptance criteria. The Moderator Temperature coefficient calculated from the ARO ITC tests was 2.81 pcm/ F, therefore, no rod withdrawal limits were required to ensure a negative MTC above 70% RTP. All control bank rod worth data presented here has been adjusted in accordance with Westinghouse's rod configuration g/g correction factors. The design data utilized in this report has been taken from Westinghouse's Unit 1 Cycle 8 Core Physics Characteristics, WCAP-10376. The Zero Power Physics testing is performed for the following reasons: 2) 3) To determine the MTC and therefore, assure the MTC is less than the Technical Specification of 5 pcm/ F below 70% RTP. To measure control rod worths to assure an adequate shutdown margin. To ensure power distribution is close enough to design values to permit power escalation. A summary of the measured data and design predictions is given in Table IVa.l with more detailed data presented in the following tables and figures: Parameter Boron Endpoints Rod Worths Boron Worths Isothermal Temperature Coefficient Power Distribution,

ARO, HZP Relative Errors in Theoretical Factors,

HZP, ARO Nuclear Peaking Factors for Enthalpy Rise,

HZP, ARO Nuclear Peaking Factors for Heat Flux,

HZP, ARO Relative Errors in F H, from Theoretical Factors,

HZP, ARO Table IVa.2 IVa.3 IVa.

4'Va.5 IVa.6 IVa.7 IVa.8 IVa.9 IVa.l0 IVa-1

The graphical results of the integral rod worths and reactivity vs. boron concentration are given as follows: Reactivity Inserted vs. Boron Concentration Integral & Differential Overlap Rod Worth,

HZP, BOC Integral S Differential Worth of Control Bank D, HZP,BOC Integral G Differential Worth of Control Bank C,

HZP, BOC Integral 6 Differential Worth of Control Bank B,

HZP, BOC Integral a Differential Worth of Control Bank A,

HZP, BOC FicCure IVa.l IVa.2 IVa.3 IVa.4 IVa.5 IVa.6 IVa-2

Table IVasl D. C. Cook Unit 1 Cycle 8 Summary of Zero Power Physics Data ARO CBD in CBC in (D in) CBB in (C< D in} CBA in (B, C, D in) BORON ENDPOINT Desi n ( m) 1534 1407 1328 1264 1135 Measured ( m) 1578.7 1450.5 1374.2 1293.0 1168.2 CBD CBC (D. in) CBB (C, D in) CBA (B, C, D in) Control Banks Total. Control Banks Overlap ROD WORTH 1168 725 596 1190 3679 3679 1191. 3 740.1 609.7 1207.0 3748.1 3683.8 lc* Corrected Measured ( cm)* 1201.3 741.3 610..1 1179. 6 3732.3 3668.2 HZP BOL BORON WORTH Desi n ( cm/ m) -9. 22 Measured ( cm/ m) -9. 13 HZP BOL ITC Desi n ( cm/'F) 0.71 Measured ( cm/'F) 0.70 FLUX MAPS HZP ARO F~~~ Desictn Meas. 1.532 1.6296 ~@~* Desicen Meas. 2.027 2.2964 F (Coze av) ~esicen Meas. 1.321 1.3546

• Adjusted for Westinghouse rod configuration 8/X correction factors.
  • Measured, unpenalized.

Table XVa.2 Unit 1 Cycle 8 Startup Tests 1. Boron Endpoint Data ROD CONFIGURATION ARO ARO CB (ppm) INDIVIDUAL 1572.4 1585 CB (ppm) DESXGN 1534 1534 hC (ppm) MEASURED-DESIGN, 38.4 51 ARO avg CBD XN CBC IN CBB IN CBA IN 1578.7 1450.5 1374.2 1293.0 1168.2 1534 1407 1328 1264 1135 44.7 43.5

46. 2

29. 0 33.2 1 The boron endpoint is the just critical boron concentration for the particular rod configuration.

Table IVa.3 UNXT 1 CYCLE 8 STARTUP TESTS ROD NORTH DATA (in pcm) Bank e/~1 Measured Corrected Design %Error Worth Worth Worth CBD CBC CBB CBA 1191.3 740.1 609.7 1207.0 1201.3 741.3 610-1 1179.6 1168 725 596 1190 +2. 77 +2. 20 +2. 31 -0. 88 CONTROL BANKS TOTAL CONTROL BANKS lN OVERLAP / 3748.1 3683.8 3732.3 3668.2 3679 3679 +1.43 -0.29 1. Adjusted for Westinghouse g/X Configuration Correction Factors. Tables A.lb and A.lc, Unit 1 Cycle 8 Design Manual (WCAP-10376). Corrected Worth Desi n Worth Corrected Worth

) ~ ' ~ ~ ~ L ~ ~ ~ ) ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ '0 ~ ~ ~ ~ e I ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ )- o o ~ a ~-

Table IVa.5 Unit 1 Cycle 8 Startup Tests'RO Isothermal Temperature Coefficient DESIGN HEATUP ITC (pcml F) COOLDOWN AUERAGE ERROR* ITC ERROR* ITC ERROR* '0-71, +-3.0 0.67 0.04 0.75 "0.04 0.71 0.0 0.71 + 3.0 0.58 0.13 0.78 -0.07 0:68 /

0. 03

• Error = Design Measured

Table XVa.6 Unit 1 C cle 8 HIP Power Distribution CBD (steps) N BH FT BH Axial Offset (~) 108-01 226-2.2964 2.4835 3.8310 1.6296

1. 9278

18. 933 P

Measured, penalized by 1.05 and 1.03 N

Measured, upenalized T

Technical Specification Limit

Table XVa.7 V 4 10801 10>21/'83 2W RTP BOLi llZPt AROt 3D AtlALYTICAL FACTORS RELATIVE ERRORS AEP - TilltlBLL DATA 6 IM DETECTOR THEORETICAL FACTORS'CALC. -tlCAS. ) illCA. ) FOR AXIAL CONFIGURATION 2 S. L K tttttt tttttt -0.045 tttttt 0.031 tttttt tttttt -0.00V tttttt -0.065 tttttt tttttt tttttt tttttt tttttt tttttt tttttt tttttt tttttt -0.059 tttttt tttttt tttttt 0.010 tttttt 0.001 f 0.008 0.033 tttttt >ttttt tttttt tttttt -0.011 5 tttttt tttttt tttttt tttttt tttt 0.060 tttttt tttttt tttttt 0.036 tthttt -0.009 tttttt -0.+!7 tttttt tttttt 0.073 tttttt tttttt tttttt ttttt>> tttttt tttttt tttttt -0.029 <<tttt 7 tttttt tttttt tttttt 0.009 t ttt tttttt 0.077 tttttt t'ttttt 0.068 tttttt tttttt 0.006 tttttt t r 8 tttttt tttttt -0.061 tttttt tttttt tttttt 0.069 tttttt tttttt 0.071 tttttt -0'26 -0.0'll 035 9 tttttt -0.053 tttttt tttttt tttttt tttttt tttttt tttttt tttttt tttttt 0.0'l'l tttttt tttttt tttttt -0.069 'l0 tttttt tttttt tttttt tttttt 0.0'l2 tttttt 0.036 tttttt tttltt tttttt tttttt 0.0'll ll -0.025 tttttt tttttt tttttt 0.051 tttttt tttttt tttttt tttttt tttttt 0.073 tttttt tttttt tttttt -0 026 13 tttttt tttttt tttttt tttttt tttttt -0.002 tttttt tttttt 0.026 tttttt tttttt 0.005 tttttt 0.018 tttttt tttttt -0.061 0'20 tttttt tttttt tttttt tttttt tttttt tttttt -0.017 tttttt -0.005 tttttt l5 -0.038 tttttt tttttt -0.07V tttttt tttttt tttttt ~ 1 ) THE NFAN VALUE a 0.0029 AND THE STANQQRD DEVIATION s 0.0'139 FOR THE ABOVE 'l6 VALUES THE NEAN OF THE ABSOLUTE lYALUES a 0 0365 'E tlhxltlUN tlAGNITUDE g 0 0765 AT 7 - l THr Hnv Hrr A A'9 ~ ~ ~

Table lVa.a 10801 10/21/83 2'4 RTP &OLI HZPt ARO s 3D -ANALYT?CAL FACTORS AEP THTN&LC tATA BINUCLEAR PEAKTNG FACTORS FOR ENTHALPY RISE FOR ASSEtlBLAGES IN THE PIIMER NORNALIZATION

0. I01 1.0V'I 1.193 1.267 1.216 1.085 0 ~ 't17

0. 2'l8

0. 560

1. 09'I

1. 251 1.'l30

1. 120 1 ~ 'I'l7

1. 319 1 ~ 12'I 0.561 0.25V 0<<250 0 ~ 667

1. 125

'I, 137

1. 321

0. 995

1. 23'I

0. 987

1. 31 &

I. 155

1. I 't7

0. 679

0. 256 0<<565 I ~ 121

1. 'I 12

1. I'I6 I ~ 038 1:170 1.077 1

~ I'l2

1. 056

1. 18't 1.13'I

1. 155 0 576 5

0<<'tO I

1. 106 1

~ I 't& I ~ 178

0. 863

1. 003

0. 833

0. 998

0. 839 I, 016

0. 882

1. 18'I I. 182 1 ~ 126 0 ~ 't12 6

1.0'17 1.260 1.307 1.039 1.0'IO 0.808 1.017 0.786 1.038 0.817 1.002 1.08'I 1.310 1.286 1.073 7 1,156 1.V09 0.989 1.123 0.827 1.03'5 0.910 0.87V 0.92V 1.02V 0.8VS 1.129 0.956 1.V05 1.189 8 1.2'l& I. 113 'IIO 1.096 0.975 0.763 0.876 0.691 0.&77 0.76't 0.966 1 ~ 093 1.215 1.082 1 ~ 257 9 1.196 1.'I'l I ~ 018 I.IV& 0.833 1.0VI 0'25 0.890 0.930 1.030 0.&36 1.126 0'73 1 't37 1.231 10 1 ~ 083 1 <<298 I ~ 318 1 ~ 022 I ~ 029 0 ~ 830 I ~ 052 0.789 1 ~ 07'I 0 ~ 819 1 ~ 017 I ~ 027 1 ~ 288 1 ~ 280 1 ~ 092 11 0,'l10

1. I'51 1

~ 158

1. 150

0. 870 1.025 0.83V 0.996 0.8'l3 1.035

'I.&53 1.1't& 1.1'IO 1.136 0.'lll 0.572 1 V8

1. 122 1 ~ 159 1.031

1. I'IO I

1.099 1<<132 I ~ 0'l7 1<<165 1<<119 1<<132 0<<567 13 0<<252 0.679 1.1V6 1.150 1 ~ 313 0.9&V 1 '38 0'89 1 ~ 319 1 l9 F 1'l9 0'75 0<<255 I IV ~ 0.251 0.56$ 1 ~ 111 1 265 1 ~ 't65 1 ~ 103 1 ~ 't25 'I 272 1 <<121 0.570 0.256 0.'ll6 1 ~ 102 1 ~ 217 1.2&V 1.205 1 ~ 090 0,'V06

Table IVa. 9 10801 10/21/83 2X RTP BOLt HZPt AROt 3D-ANALYTTCAL FACTORS AEP THIIIBLE DATA NUCLEAR Pf AKING FACTORS FOR HEAT FLUXt F SUB Qt FOR ASSEIIBLAGES IN THE POIIER NORNALIZATION R P C B A

0. 5'l9

1. I'I2

1. 636

1. 7'I&

1. 679 1

~ 517

0. 577 I 2 0 ~ 3'IO 0.765 1.'l9'I

'l.713 1.935 l.511 1.9681.809 1.550 0.77'I 0.3'l7 0.339 '0.909 1.533 1.533 1.779 1.331 1.650 1.318 1.79'I-1.593 1.578 0.931 0.350 0 ~ 76 I 1 ~ $20 1 ~ 502 1 @546 1 ~ 388 1 ~ '553 1 e 137 1 ~ 517 'I e 412 1 ~ 636 1 ~ 566

1. 578 0 ~ 782 5

0 ~ '5'I& 1.501 1.550 1.593

1. W3 1.329 1.110 1.339 1.119 1.356 1 ~ 180 1.636 1 ~ 59V 1.527 0.559 1.'l61 1.719 1.750 1.391 1.379 1.082 1.378 1.053 1.393 1.09'I 1.3V6 1.'I'l&

1.752 1.7'l5 1.'l69 1 ~ 610 1.911

1. 325 1.'l96

1. 103

1. 395 1.239

1. 197 1

~ 250 I ~ 383 I ~ 121 I.'I98 1 270

1. &92 1.6P5 1.723 1.507 1.663 1.475 1.3'l5 1.035 1.212 0.941 l.198 1.027 1.29'I 1.V51

1. 62'I 1.'l56

1. 72'I 9

'1.6'I'I 1,952 1.368 1.'533 1.1'12

1. I03

'1.26 I 1.22. 1.2'19 1.380 1.106

1. I&7 1.29V 1.9'Vl 1.700 10 1.'l92 1.772 1.769 1.371 1.369 1.'113 1.'l31 1.07'I t.'l60 1.091 1.3'l&

1.37'I ~1.717 1.735 1.508 1 1 0.560 1.563 '1.568 1.556

1. 171 1 '71

1. 122 1.360

'l. 137 1.390

1. 1'l7 1.561 1.5'l1 1.545 0 '61 12 0.777 1.565 1.524 1.572 1.389 I' 5'l1 1.496 1.530 1.'l20 1.591 1.538

'I ~ 550 0.771 13 0.3'I'I 0.923 1.560 'I.569 1 ~ 776 1.330 1.673 1 3VI 1.791 1.569 1.567 0.92'I 0.3VB '0.3'l6 0.777 1 '27 1 7'l2 2.018 1 ~ 51'I 1.955 1 51 1.537 0.783 0.351 '5 0.577 1.5'l5 1.703 1.806 1.679 1.52V 0.558

1 Table XVa.l0 10801 IO/21i83 2W RTP BOLi HZP) AROr 3D-AIIALYTICALFACTORS AEP 1 IllliBI.E DATA RELATIVE ERRORS IN F SUB DELTA H CALCULATED FROII kCIGHTED THEORETICAL FACTORS'CALC ~ -tlEAS.)<HEAL P N tl L B A 0.001 -0.001 -0.03'I -0.056 -0.052 -0.039 -0.036 0.028 0.015 0.015 0.001 -0.0'I2 -0.063 -0.053 -0.051 -0'12 0.0'12 0 F 005 0.023 0.023 0.032 0.028 -0.029 -0.03I.0.059 -0.026 -0.027 .0.012 0.012 0.005 -0.001 Oo006 0.036 0.032 0.0'I5 0.027 -0.037 -0.015 -0.01'I 0.009 0 '11 0.0'll 0.006 -O.OIH 5 0.002 0.005 0.0'l8 0.016 0.058 0.06'I 0.0'I'I -0'16

0. 036

0. 051

0. 035

0. 01 1 -0.01 1 -0. 013 -0'26 6

-0.00'I -0,006 -0.019 0.025 0.027 0.069 0.070 0.039 0.0'l9 0.057 0.065 -0.017 -0.021 -0.026 -0.028 7 -0.003 -0.027 -0.028 O.OOV 0.051 0.053 0.075 0.067 0.060 0.06'I 0.029 -0.002 0.006 -0 '25 0.031 -0.0'12 -0.057 -0.063.0.032 0.007 0.070 0.065 0.069 0.065 0.069 0 '16 -0.030 -0.0'I'I -0.031 -0+0'l9 -0.037 -0.0'l9 -0.056 -0.019 0.0'l3 0'.0'l6 0.058 0.0'l9 0.052 0.057 O.OVO 0.001 -0'12 -0'V6 -0'6V 10 -0.037 -0,035 -0 '27 0.0'I2 0.038 0.0'll 0.035 0.035 0.01'I 0.05'I 0.050 0'38 -0.005 -0.022 -0.0'I'S ~ 1 1 0 ~ 020 0 ~ 035 0 ~ 010 0 ~ 0 ll 0 ~ 0 l9 0 ~ 0 ti 0 ~ 0 t2 0 01 I 0 ~ 031 0 ~ 032 0 ~ 069 0 ~ 0 I3 0 ~ 026 0 ~ 022 0 ~ 023 12 -0,008 0.012 0.022 0 ~ 033 0.033 -0.012 -0.035 -0.005 0.018 0.027 0'25 0.026 0.001 13 0.011 0.00'I 0 F 013 0.017 -0.02'I -0.023 -0.062 -0.028 -0.028 0 ~ 018 0.011 0.010 -0.000 0.018 0 ~ 001 -0.000 -0.010 -0 065 -0.0'IB -0.038 -0 ~ 016 -0.009 -0.00'I -0.00'LI -0.035 -0.053 -0.053 -0.069 -0.0'l3 "0.0'I'I -0.010 THF NEAN VALUE

• 0 0033 AND THE STANDARD DEVTATION
• 0.0365 FOR THE ABOVE 193 VALUES NEAN OF THE ABSOLUTE VALUES I 0.0307

~ THE NAXINUN IIAGttITUDE > 0.0751 AT 7 -J THE tlAX~ NEG. -0'686 AT 15-H L.

4 V~

FIGURE IVa.2 Unit 1 Cycle 8 Differential and Integral Rod Worth in Overlap I++ --)) 3500 16.0 I C 3000 14 ' L40 12.0 'E Oo i,i, l ~ 2500 10 ' 0 2000 8.0 : 0 0 6.0 . i-1SOO 9 1000 4.0 2.0-500 i ,Li: 0.0 40 0 120 160 200 'CBBA ,:K::L-;iI: 40 80 120 160 200 T1%I,i&f~lZl I Steps Withdrawn CBB 120 160 200 40 8 CBC 0 12O 0 160 200 228 CBD

.;.4000 3800 3600 4 V'IGURE XVa.' -NyB%--'XX Reactivity Inserted44R vs ~ Boron Concentration Unit. 1 3400 3200 gg.. 3000 2800 RNf r fR INK <<j%R '~<<N 4 f+f. B!: +

4. g"g 4

~Z!TH + 2600 '4) j I<<MA=I !f ~'a eLt<< "l i)p! gg 't 4'I 2400 tgt ~+'Xt I ra ~ I 1 LL <<+4 +A~ Measured Slope = - 9. 13 2200 2000 1800 I <<Lj ! <<: ph<<+,~pPj N 1'"Hh~~N VNk N.%+' NSNR 4 P-XI ~ It ler $ 14 'I<<I f ta ~ ~ 4 ~I I-P're f ~ ~ \\ f4 ~ ~i 41 III ~ ~ ~ r<<SZ re<<t HW 1( ~ I 1 ~ I H:!i-Rj,'600 ~Q F@Ti~ X + ~ +t ~rf rra ~- 1 ~ e+ ~ I Pl. 4HX'. 1400 1200 ~X >jR ~t re+ ~ r>>1 It 4<<I 4- ~ ~

4. ~

~ 4+fr ~ I ~ ~ ~- I-tr<< earL ~ rf 1<<fl. <<4 <<f 4%>> 4 4~ !<<I- -<<t-t j~<< <<4<<<< K'.7<<!~4 Design

4 Slope

= -9. 22 ~ ~' ~ I t Ie P I t .m4 <<<<<<I pf I- ~ ~ <<t el I I P I ~ 44>>e Ia-I I l-fl I f-I 4-1 ~ f<< ~.Ii <<4 ~-4 'L. e>>rt farl h>> ~ ~ 4 ~ + r4<<444. <<>>rp PH~ <<p g+ 4 N) ,0 (6 1000 800 f4f << I la soo 400 lf ~ I ttr ~ ~ 1 ~ ~ * ~ ~ P 'I t.~ + I ~ ~ ~ f fe as ~ p 4 ~ I ~ I ~ I e-!. -~ Ie ~ I ~ ~ e ~ ~ 'I ~ 4a ~ ~ I ~ fl ~ t ea ea 1 ~ I ~ ~ ~: ~ I = e 1 I I ~ r ~ ~ ~ ~ 4 ~I Ife e t.a ~ * ~ ~ >> I ~ ~ 4 ~ ~'I le+4 1 >>1 ~ ~ I I I ~ I ! ~ ~ I 41 4th ~ ~ ~ 1f ~ ~ ~ ~ ~ I I ef 4 ~ ~ ~ ~ a- ~ ~ t ~ ~ ' ~ ~ ~ tI ~I<<' 1 ~ ~ ~ 4 ~ 1 tl ~ ~ I I ~ ~ ' 4 ~ ~ ~ 'I ~ ee rf <<I Ia ~ ~ 4 I ~ ~ ~ ~ 4 ~ ~ " ' t I ~ ~ I ~ ee I I- 'I 1<< ~ ~ I ~ ~ ~ 4 l ~ 4 ~ I ~ ~ ~-~ 4 ~ ~ 1 aI 4 ~ l <<>>4 ~ 4 11 ~ I ~ -t>> - I .1 <<e ~l ~ ~ ~ ~ *4 ~. -11 ~, ee ~

4. Pe 4

~'pe I. * ~ ~ ~ ~ ~ a<<<< t 44 ~ ~ I ~ ~ le ~ I I a. 4 4- ~~ ~ Pt<< 44-4 Pt ~ ~ ~ e<<H ~ I ~. t 4-f I ~ << ~ ae ~ ~ II ~ I I Ie 4 ~ a ~ I ~ ~ t+ e+4 LIL NIP ~ ~ ~<< ~ "I ~ I ~ ~ ~ I << l-a ~ ~ ~ ~ ar ~ ta ~ 1 ~ I ~ ~ ~ ~ ~ I ~ ~ ~ I farl ~ 1 ~ ~ he I I 'I ~ ~ I ~ ~ ~ ~ ~ 1 ~ ~ t I ~ ~ 4 ~"t t ~. ~-r ~rr ~ 4 ~ ~ >>14 ~ ~ ~ 4 ~ 'I Pi ~ I ~ ~ i f4 <<! I!I-lPj Vpl! I <<I 4 t ~jj!'j) 4lrf I ~N.i.-:ij j* teal .!i!! ~ltl !ILI 200 iNRh ~ I<<f ~ ~ ~ f tf ~ t 1 ~ ~ a ~ ~ I re ~ 4<< ~ 1 ii~ ~ 4 ~ ee ~ ~ l=t!;I,'. ~ f ~ 4 ~ ~ ~ ~ I a I ~ I ~ I ~ rr ~4 ~ ~ ea' ~ te 4 4 ~ e 't 'I e I ~ 4 ~ +I <<t ~ 1 pe 4 rrp ee ~ f I r<<' ~ !'I-'I Wrf r!-r!4-.f~~t-' ~ I te ~ I ~jf ~ f+ ~ f<<I 4 ~ I!t ~-!- 4 ~ 4 ~ ~<<4 Iel ~ 4 1+ ~+ rf44 ~ f'. Lef 4

e. ~ t f fl~ 1

~ 1 0 100 et' ~ t<< 1 ~ ~ el ~ ~ 1 ~ 4 ~ ~ ~ <<I P ~ iL 4-I ~ ~tf 1 ~ ~ t 4 r<<I '<<tL ,, I4! I&e I ~ I 0 1100 1200 1300 1400 1500 1600 Boron Concentration (oom)

1700,

FIGURE IVa.3 GR10 STRhPS D. C. Cook Unit 1 Cycle 8 Differential and Integr al Rod North of D-Bank, HZP. BOC

37. 7

78. 8 120. 5 162. 4 204. 3 1200 1100

]0 1000 9 CL 0 E 8 CL 0 -o 6 0 5 8 4 3 900 800 700 0 600 500 4oo 300 200 100 a o a o o o o o o o o o o o o o o o a o a o a m PJ P7 W lA (D W Kl CD a ~ M 07 W ln lD W CO CO o ~ OJ CU PJ D-Bank Steps Nithdr awn

V ~h wr e

FIGURE IVa. 4 D. C. Cook Unit 1 Cycle 8 Differential and Integral Rod Worth of C-Bank. HZP.BOC GRID STRhPS

37. 7 7e. e I20. 5 162. 4 204. 3 800 700 600 E

D 5 L0 0 500 0 0 400 300 "O 200 100 0 o o o o o o o o o o o o o o o o o o o o o o o co M co w lA Q3 M G) 0) n ~ M ol w lA co M cD ol o ~ cU cU (U (V N N C-Bank Steps Withdrawn

~T

FIGURE IVa. 5 D. C. Cook Unit 1 Cycle 8 Oifferential and Integral Rod North of B-Bank, HZP. BOC GR10 STRAPS

37. 7

78. 6 120. 5

]62. 4 204. 3 700 600 E CL E0 CL 4 0 'U 0 lY 3 0 r8 8 2 Cl 500 0 400 300 100 o a a o a a o o o o o o o o o o o o o o o o o co R p) s s (0 cD o ~ N Fi rt' (0 co s o ~ (U N PJ N PJ B-Bank Steps Withdrawn

4 FlGURE IVa.-6 D. C. Cook Unit 1 Cycle 8 Differential and Integral Rod North of A-Bank. HZP. BOC GRl0 STRhPS

37. 7 l20. 5

]62. 4 204. 3 1200 1100 10 1000 9 CL CL 8 EL 0 6 0 5 L8 3 900 800 700 0 600 500 400 e; 300 200 100 0 0 D D 0 0 0 0 0 0 0 0 0 0 0 0 0 0 PJ m M lA (0 M CD CD 0 ~ CU m ~ lA U3 0 0 0 03 CO O PJ O Q CO CV (U M A-Bank Steps 'Nithdr awn

a iY

.IV b Low Leaka e Loadin Pattern (L P) 3 The core design for Unit 1 Cycle 8 is characterized as a low leakage design. By comparing the Unit 1 Cycle 7 and Unit 1 Cycle 8 design manual core loading diagrams, the difference becomes apparent.- The core periphery locations in Cycle 7 were composed entirely of fresh assemblies. In

contrast, the Cycle 8 core design includes once and twice burned fuel loaded into periphery locations.

A comparison of the two (2) loading patterns can be seen in Figures IVb.la and IVb.lb. The effect of loading lower "worth" fuel into the core periphery is to cause these locations to run at a lower power as compared to the previous cycle. This is illustrated quite well in Figure IVb.2 (see attachments). This figure was obtained'irectly from power distribution maps from the cycle design manuals. The values listed represent relative assembly powers. A value of 1.00 indicates an assembly producing a power equal to the average assembly power. It follows that a value less than 1.00 is a low power assembly and a value greater than 1.00 a high power assembly. It can be seen from Figure IVb.2 that for the same power level the periphery assemblies for Cycle 8 are producing significantly less power than the same locations in Cycle 7. In Figure IVb.3 a comparison can be seen of actual measured, power distributions for 90% RTP conditions with burnups of 152 and 168 for the U1C7 and U1C8 cores res-MWD MWD MTU MTU pectively. The figure was obtained from a flux map output map of the nuclear enthalpy rise hot channel factor F>H. By N definition, F ~ represents essentially the identical quantity as N the computer generated relative assembly powers shown in Figure IVb A comparison of Figures IVb.2 and IVb.3 show excellent agreement. ~ 2 ~ An assembly producing less power is undergoing fewer fissions per second and, therefore, producing fewer neutrons per second. 'ased on the neutron diffusion length, peripheral assemblies can be shown to be the significant contributors to neutron leakage flux from the core. Based on the core power distribution maps and the two (2) important statements above it, becomes apparent that at the same power level the Cycle 7 core had a much greater leakage flux than the Cycle 8 core. Since all nuclear instrumentation protective functions supplied by the excore instruments rely on the leakage flux, it was important to consider the low leakage core design effects on the excore detectors. 'f particular concern was obtaining a method to generate excore detector calibration factors for initial startup. Work on this problem was pursued independ-ently at D. C. Cook, AEPSC and Westinghouse. IV b-1

All of the approaches were quite similar, using an equation of the general form: I 8 I 7 x N k 3.= 1 N i=1 p3. 8 c7 8/7 p3. 7 I 8 = Calculated Cycle 8 current, at 0.0-o A.O. I 7 = Cycle 7 current from last In/Ex at 0.0$ A.O. 3. k = Assembly i weighting factor P8 = Assembly i Cycle 8 power 3. P = Assembly i Cycle 7 power R8/7 = Weighting ratio This equation essentially represents a ratio of integrated corner power which will consequently represent a ratio of leakage fluxes. The approach actually used for the problem involved using EOC UlC7 excore currents generated by the last Incore/Excore calibration and an applied weighted ratio. Com-paring the results of their independent

studies, D. C. Cook, AEPSC and Westinghouse agreed to 'also apply a geometric, weight-ing factor (k

) based on the number of exposed faces. Assem-blies with two (2) exposed faces were assigned a value of 1, one (1) exposed face a value of 1/3, and 0 exposed faces a value of 1/6. These values were then multiplied by the appropriate computer generated relative assembly power for BOC 8 and by measured power distribution for EOC 7. Then taking the ratio BOCS of these values ( ) the weighting ratio was found to be 0.5645. EOC7 Also of concern for initial startup was the effect of the low leakage core design on trip set points. The initial cali-bration factor obtained from the analysis previously discussed ensured conservative operation of power range (PR) excore detectors with respect -to trip set points. To ensure conservatism of the power range trip set point due to uncertainty of the initial cali-bration, the PR trip was set at 80% RTP until after the 50% IV.b-2

Incore/Excore calibration. The normal setpoint. is 109% RT Based on the expected power distribution on the 0 and 180

flats, the Intermediate Range (IR) trip set points were not expected to be affected significantly by the low leakage core design.

To monitor the affect on the IR detectors, step 8.20 was added to **12 THP 6040 PER.359, Zero Power and Ascens'ion Testing. This step required IR detector current readings taken at various hT power levels. The IR current was plotted versus hT power to ensure our IR trip setpoint of 255 RTP (30% maximum allowed) was not violated. Figure IVb.4 is the data that was taken during power ascension. IV b-3

1 Q+

R P I I I I FIGURE IV b. la Unit 1 Cycle 7 Core Loading Diagram Rl 11 C X 4 11 C f C 0 C 8 h FIGURE IV b:.1b Unit l Cycle 8 Core Loading Diagram ff~ 1804 X ~ X:.".-CX 4 5 6 X 9 1C X XX ~ il y f hlV'0 12 13 +h>>l Cil 1 i Z.i)1 RCPICC h >>.SRS >>C$ 1CII ) PCPICh C H CPI CC 1 ~ Q Rc91on 9 J 2.903 R C 5 R D C 8 Oc Rc5jon 8 H Enr ch n>> 2 905 15 ~ - Region 10R - R RC5'o:1 108 K 3.3 3>>6

FIGURE IV'b-2 Comparision of predicted Relative power Distribution .916 .91 1.085 1.11 .975 .94 1.54 1.08 .884 1.10 .825 1.12 .880 0.97 .908 1.086 .977 1.11 94 1.002 1.220 1.13 1.20 1.216 1.025

1. 20

.96

1. 159 1. 204 0.95 1.13 1-091 1.225 1.14 1.09

.883 0.94 1.111 1.22 1.217 .904 1.23 1.16 .929 .844 1.156 1.08 1.162 .95 1.204 1.13

1. 022

.97 1.200 1.23 .952 1.133 1.05 .885 .825 1.10 1.12 1.14 .94 1.224 1.110 1.09 1.22

1. 197

. 951 1.22 1.15 .949 .888 1.16 1.15 .889 1.025 1.15 .71 .911 .620 .58 .28 1.091 .883

1. 132

. 646 1.05 .40 .911 .58 .620 ~ 28 .880 .908 0.97 1.05 1.217 ,929 1.23 1.01 .904 .844 1.16 .95 1.05 1.01 .95 .40 U1C 7 HFP 150 MWD MTU U1C 8 HFP 100 MND MTU Assemblies considered to affect excore PR detectors

,.FXGURE IV b.3 Comparison of Measured Relative Power Distribution .840 1.0S1 .887 1.073 1.048* 1.033 1.073 1.084 - 994

1. 222

.920 1.182 1.193 1.180 1.068 .934 . 999 1. 194 .916 1.068 1.229 1.186 1.182 .930 1.022 1.179 .944 1.129 1.179 1.045 1.129 .943 .925 1.124 1.160 1.164 1.255 1.079 1.239 1.188 .828 .883 1.170 .999 .901 .175 .950 1.261 1.125 .899 1.249 1.157 .978 1.118 1.152 1.057 .850 1.081, .879

1. 036

. 796 .956 .626 . 398 .922 1.165 1.127 1.155 .824 .901 1.182 .950 .886 1.175 1-006 1.261 .850 .879 1.081 1.030 1.259 1.245 1.079 1.188 1-127 .980 l.'249 1.152 .901 1.120 1.151 1.057 .796 .626 .956 .396 . 978

1. 129

. 905 1.140 .876 .577 .904 .871 1.140 .578 1.007 .603 .712 .277 .604 .277 U1C7 Map 511 90% RTP; 152 MWD NTU F,> measured U1CB Map g8 90% RTP; 168 MWD MTU F H measured N

~ ~ 0 0 ~ '0 a I I A 1 ~ ~ 0 ~ ~ ~ I ~ 0

IVc I Plot Sha es for Unit 1 le 8 Due to constraints imposed by RCCA and BPRA locations in the Unit 1 Cycle 8 core design, there were only a few locations in which secondary neutron sources could be placed. It can be seen from the core loading patterns in the fuel shuffle portion of this report that the sources went from a location two assemblies in from the core edge at, H3 and H15 to the core periphery locations (F15 and Kl) in the Cycle 7 to Cycle 8 ~ , transition. This movement put the secondary sources closer to the source range (SR) nuclear instrumentation. 'he SR detectors are used to monitor subcritical multiplication during plant shutdown and approaches to criticality. When the sources were moved from their Cycle 7 to Cycle 8 locations an increase in counts of 20 to 30 times was observed on the source range instruments. Based on this information, an in-vestigation by D. C. Cook Nuclear Engineering personnel was ,initiated to study the effects of this source range placement on ICRR (1/M) plots used during approaches to criticality. An ICRR plot is simply the ICRR (initial counts counts ) plotted versus any parameter that increases k f toward criticality. Some examples of ICRR plots are given in Figure IVc.l. The plot is considered conservative if it underpredicts criticality and non-conservative if it overprediets critical conditions. In a situation where the source range detector is dominated by artificial source neutrons , a non-conservative shape can be expected. The secondary neutron source dominates the SR detector and the detector can not detect changes in core neutron population accurately. This is sometimes referred to as source shine. The sources in Unit 1 Cycle 8 are situated in such a fashion. A simple relationship was established to calculate the ICRR values for different values of baseline counts. The model was based on data from Unit 1 Cycle 7 initial criticality. The following equation was used to generate predicted curve shapes for UlC8 initial approach to criticality: A C0 ICRR=AC + M 0 A = Constant which the UlC7 baseline is multiplied by (ex. 10 x, 20 x, etc.). C = UlC7 baseline counts M = U1C7 multiplication at some dilution level. (counts baseline) Based on this analysis, the expected curve shapes were quite non-conservative. To compensate for this expected effect, two (2) cuts in dilution rate were added to **12 THP 6040 PER.357 to give operators a slower approach to criticality. In anticipation IVc-1

of future startups following giant trip, a new source range de-tector N-33a was installed 90 from the current detector locations. It is important to note that when criticality is close the counts on N33a will be approximately equal to Channels N31 and N32. Figures IVc.2 and IVc.3 represent actual data taken from (N31 or N32) and N33a respectively. ,IVc-2

FIGURE IVc.l, ICRR Plot Shapes ICRR Non-conservative Ideal Conservative 1 Decreasing Boron Concentration

FlGURE XVc.2 1.0 ~ ~ UNIT I CYCLE 8 ICRR vs. BORON CONCENTRATION ~ 9 7; I .6 -3 .2 2200 2100 2000 1900 1800 BORON CONCENTRATION (PPm) 1700 16 0 ) 500 1523 CDD 8 190 Steps

~ ~ I lg 4'

FIGURE 1.0 SPARE SOURCE RAIICE (Xnfolnntion On)V) UNIT 1 CYCLE 8 ICRR vs. PRIHARY WATER (thousands of gallons) It) I l:Itii .6 .a tati'ti hatt fl)s) } i) i ti) .3 ~ 2 St0 cdI'i ion u't at 17 I5 10 gals L4 I

• $1 8

10 1c 14 16 13 19.7/20 DDTMhDY Whtt$DD lihwtenntltt Itf ntt 1 lnntt1 22 22.9 24 28

IVd Problems Associated with Flux Ma in S stem On October 18, 1983, during Zero Power Physics Testing, the flux mapping system leak detection alarm began to alarm inter-mittently. An,Emergency Job Order was written to investigate whether there was a leaking thimble or a malfunction in the alarm system. Upon entering the seal table area, water was observed flowing from detector C's ten-path rotary,transfer device (see Figures IVd.l and IVd.2). The technicians investigating the situa-tion conducted a temperature determination on the guide tubes running from detector C's ten-path rotary transfer device to the isola-tion valves and. found thimble J-1 to be leaking (see Figure IVd.3). The J-1 isolation valve was subsequently closed and the water flow ceased. An inspection of the ten-path rotary transfer devices asso-ciated with the other five (5) detectors was conducted on Octo-ber 19, 1983, to determine if borated water had "backed up" those ten-path rotary transfer device drains. All six (6) ten-path rotary transfer device drains are connected to a common line which contains the leak detector. Water was found in three (3) of the five (5) ten-path rotary transfer devices. As the potential existed for water to have gotten into the bore of the thimbles associated with these ten-path rotary transfer devices, and the detector C ten-path rotary transfer device, NUS Corporation was contracted to dewater and air-dry these thimbles as an interim step until they could be properly cleaned. On October 26 and 27,

1983, between flux maps of the Power Ascension 50% Incore/Excore Calibration, NUS Corporation dewatered and air-dried the thimbles associated with the four (4) ten-path rotary transfer devices which contained water.

Only the nine (9) thimbles of C's ten-path rotary transfer device had water within their thimble bore. No water was discharged from the thimbles associated with the three (3) other ten-path rotary transfer devices. During the dewatering and air-drying, on October 27,

1983, a problem arose with thimble H-3 (see Figure IVd.3).

During the dewatering/air-drying process the isolation valves of the suspect thimbles were disconnected from the guide tube frame assembly (see Figure IVd.2). When H-3's isolation valve was disconnected, the reactor coolant system pressure forced the thimble approxi-matley four and one-half (4.5) inches out of the core through the seal table middle high pressure seal (see Figure IVd.4). The iso-lation valve was secured to the guide tube frame assembly via rope and the dewatering/air-drying of the remaining thimbles was completed.

H-3's isolation valve remained tied to the guide tube frame assembly until November 30, 1983. Prior to the cleaning of the Unit's thimbles by NUS Corporation, plant personnel reseated the H-3 thimble and tightened the middle high pressure seal. Thimble cleaning was completed on December 1, 1983. Ivd-2

0 ~ I l-s ~ I I-4 4 I ~ I ~8 ~g ~ laQeQaQa <,<ii:"i 'i~ rrl1 r~ 0 ~ s g ~ ~ ) l

To 5 Path Rotary Transfer Devices 10 Path Rotary Trasfer Devices Guide Tube Frame Assembly ~// /i / / Guide Tubes solation Ualves I ]High Pressure Seals Seal Table FIGURE IVd.2 SEAL TABLE AREA

V

~ ~ 0 ) ~ ~ ~ ~ ~ S ~ ~ SL Sl ~ ~ SL Sl '1 ~ jL gL SL 1r ~ L ' P 4 ~'ll SL 1r ~ L ~ r Ah AL rr ~L ~r ~ ~ ~ 1 'Ir ~L ~r Sl ~r S'L ~r IL ~ ~ ~L r ~ IL 'Ir ~ a SL'Ir jL Sl 'I~ S ~ ~l 1r SL ~r SL 1 ~ SL 1 ~ SL 1r ~L 1r ~l ~ 1 ~lr SL 1r ILrr 4L '1F SL ~L ~r S ~ SL '1 ~ SLlr S ~ ~ ~ ~ I ~ ~ a ~ ~ ~ ~ ~ ~ ) ~ ~ S ~ ~ ~ ~ ~ I ~ ~ ~ o I ~ 1

eq

5/16 in. Flare Fitting Union Connector Hex Body (5/16" flare x 5/16") Upper Seal (5/16") 9 I Ch Distance between these compression Compression Hut nuts was approxi-(5/16" ) mately 4.5 inches. Lower Seal (5/8" or 3/4") 'hrink Tubing Middle Seal (5/16") Reducing Union Hex Body (5/16" x 5/8") or (5/16" x 3/4") Compression Hut '(5/8" or 3/4") Seal through which Thimble was forced by RCS Pressure. Thimble (0.301" O.D-X

0. 201" I - D. )

Conduit (5/8" or 3/4") OPERATIONAL POSITION FIGURE IVd.4 SEAL TABLE HIGH PRESSURE SEAL

Va POWER ASCENSION TESTING The Unit 1, Cycle 8 power ascension testing commenced at 1221 hours on October 24,

1983, and, was completed at 1021 hours on November 8,

1983. Figure Va.l displays the power ascension testing program as a function of time. The power ascension testing consisted of the following: 1) Core power distribution measurements 2) Incore/excore detector cross calibration 3)

MTC, DPC, AT/hP testing Flux maps were obtained at various power levels in the course of power ascension.

Maps were taken at approximately 30%, 48%, 68%, 89%, 95% and 100% RTP. The power distributions were calculated by the DETECTOR code, using data obtained with the incore moveable detector system and analytical factors supplied by Westinghouse. For all of the flux maps, the power distribution measurements were in compliance with Technical Specification limits. A summary of the peaking factors at the mentioned power levels is given in Table Va.l. At 30%

RTP, a flux map was taken with control bank D positioned at the 100%

RTP Rod Insertion Limit (182 steps on CBD). The flux map was taken with this configuration to verify that the Rod Insertion Limits need not be revised. Results of this map are given in Table Va.l. At 48% RTP, four (4) full core flux maps, 108-03, 04, 05, and 06, were obtained with axial offsets ranging from -26.88% to 21.30%. The reduced data of the four (4) flux maps was used for the incore/excore detector cross calibration. Figures Va.2 through Va.5 give the excore current vs. the incore axial offset for channels N-41 through N-44 Figure Va.6 shows the incore axial offset, vs. the excore axial offset for channels N-41 through N-44. Table Va.2 gives the calculated detector upper and lower currents, at selected incore axial offsets, using the least square fit program. After CGI calibrated the NIS, making use of the reduced

data, power was raised to approximately 68%

RTP. Power was held constant for 24 hours to condition the fuel and to obtain a flux map. After the 24 hour hold, the MTC,

DPC, and hT/LP tests were conducted.

The MTC and DPC tests were performed by control rod substitution. For the MTC, control bank D, CBD, was inserted or withdrawn while holding power constant'hich resulted in a temperature change. In the DPC

test, the temperature was held constant for CBD insertion or with-drawal which resulted in a power change.

The changes in temperature and power along with the reactivity change due to rod insertion or withdrawal were used to calculate the MTC and DPC. The ~T/~P test was performed by increasing or decreasing turbine load to acquire a temperature and power change simultaneously. Table Va.3 gives a summary of the

MTC, DPC and hT/hP tests results.

The MTC test was performed to assure a negative moderator 'emperature coefficient exists when power is raised above 70% RTP. The Doppler Power Coefficient test was performed to compare measured data,to the design. hT The P test was performed to check data.consistency. By comparing the measured to the measured (- ) and the measured hT MTC QP DPC (gp ) to the measured (- ), a check on the re 1iabi 1ity of the hT'PC measurements can be made. The ETC is derived from the measured MTC. See Table Va.3. Due to the introduction of the WABA's in Cycle 8, the American Electric Power Service Corporation,

AEPSC, detemined that the use of three-dimensional, 3-D, theoretical factors was necessary for providing a better estimate of the measured local heat, flux factors Prior to power ascension, plant personnel and AEPSC personnel detected errors in the Westinghouse generated 3-D theoretical factors.

Con-sequently, the 89% RTP flux map was analyzed using the conventional Westinghouse generated two-dimensional, 2-D, theoretical factors. As recommended by Westinghouse, the following penalties were applied to the heat flux hot channel factors, F , Technical Specification 0'imits when using the 2-D theoretical factors during power ascension: 2) 3) 15% reduction in the P limit for the top and bottom ten, 10, percent. of the core, 2% reduction for the next 10%, and 1% reduction for the center sixty, 60, percent of the core. The Allowable Power Level, APL, of the 89% RTP flux map was sufficient to allow power ascension to 95% RTP. On November 4,

1983, Westinghouse provided a set, of corrected 3-D theoretical factors.

After a thorough review of the 3-D theoretical factors, by AEPSC, the 89% RTP and 95t RTP flux maps were reprocessed using the corrected 3-D theoretical factors. The APL of the 89% RTP and 95% RTP flux

maps, when using the corrected 3-D theoretical factors, was suffi-ciently high enough to allow power ascension to 100%

RTP. Prom an incore power distribution standpoint, power ascension went well, with little time lost due to APL (Allowable Power Level) restrictions. Va-2

Power Ascension Testing Table Va.l Unit 1 Cycle 8 CORE POWER DISTRIBUTION ~Ma 108.02 108-03 108-07 108-08 108-08 108-09 108-10 Power (%) 30 48 68 89 89 95 100 A.O. (0) -5.983 -3.455 1.719 -2.262 -2.292 -1.428 -2.568 FPenalized 0 2.1483 ~ 1.9611 1.7950 1.8006 1-7656 1.7263 1-7238 FL Q

3. 9400

3. 9301 2.8556

2. 1964 2.2184 2.0838 1.9726 APL

(%)4 80.9 88.2 95.5 96.3

99. 2 101.1 102.3 hH 1.5146 1.4455 1.4196 1.3944 1.3639 1.3687 1.3612 FL hH 1.8033 1.7236 1.6348 1.5401 1.5401 1.5144 1.4906 Penalized

= Measured Value x 1.05 x 1.03 L = Technical Specification Limit APL = Allowable Power Level 1 = 2D Factors for 50%, BOL 2 = 2D Factors for HFP, penalized K(Z) 3 = 3D Factors 4 = APL calculated for +5% Target Band 5 = FCFM 108-02 was obtained at the 100% Rod Insertion Limit

C ~ ~ Table Va.2 Unit 1 Cycle.8. INCORE AND EXCORE DET. CAL 10/26/83-10/27/83 BASED ON MAPS 10803, 04, 05 and 06 DETECTOR CURRENTS IN p AMPS CALCULATED FROM THE LEAST SQUARE FIT Incore Axial Offsets UPPER CURRENTS FOR DETECTOR 41 -30.0 69.5917 -20.0 -10.0 0.0 10.0 20.0 . 30.0 74.8268 80.0619 85.2970 90.5322 95.7673 101.0024 UPPER CURRENTS FOR DETECTOR 42 91.0223 98.3774 105.7325 113.0877 120.4428 127.7979 LOWER CURRENTS FOR DETECTOR 42 133.3245 125.3725 117.4205 109.4685 101..5165 93.5645 LOWER CURRENTS FOR DETECTOR 41 113.9109 106.6792 99.4474 92.2157 84.9839 77.7522 70.5205 135.1531 85.6125 UPPER CURRENTS FOR DETECTOR 43

94. 2801.

101.7934 109.3067 116.8201 LOWER CURRENTS FOR DETECTOR 43 133.9680 125.8832 117.7983 109.7135 124.3335 101.6286 93.5438 85.4590 131.8468 139.3602 UPPER CURRENTS FOR DETECTOR 44 85.0729 92.0106 98.9483 105.8860 112.8237 119.7613 126.6990 LOWER CURRENTS FOR DETECTOR 44 130.8142 122.9700 115.1258 107.2816 99.4374 91.5931 83.74

Table Va.3 Summary of MTC, Doppler Coefficient and hT/hP Test. Data at <70% RTP Unit 1 Cycle 8 MTC (pcm/'F) DPC (pcm/%) R ( F/8) Rc 5 (oF/%) -1.14 -10.33 -8.42 -4.73 -2.64 -10.88 -4.49 -2.52 -l. 67 -6.24 -3.50 -2. 59 -2.01 -10.61 -6.38 -3.58 Desicen -2. 96 1 -ll.25 1 -5.28 2 -3.00 2 Deviation 3 -0. 95 -0.64 1.10 0.58 NOTES: 1 ~ Values were calculated from the vendor's design report, ARO condition 2 Values were calculated using available measured test results (i.e., MTC, DPC, etc.) 3 Deviation = Design value Measured data 4 R hT DPC ( Do ler Power Coefficient hP MTC Moderator Temp. Coefficient 5 R, hT'PC Do ler Power Coefficient hP ETC Isothermal Temp. Coefficient ITC = MTC + DPC = -3.54 pcm/'F DPC = -1.53 pcm/'F

~ ~ FIGURE Va. 1 Unit 1 Cycle 8 Power Ascension vs. Time 100 90 70 60 50 0 OldI-Ld A 0 00 Lo L0 0Ld . C La d dC Q0 0L O O 'd C L dE I ~ ~ . C a C I 'U M 'UI L wl A 0 OI I-Mi 0 Itl IOi wl CO CI I0)O L 0 0 v r Ca oL 00 Itl d E a I I IL O I I I I I I I I I I I I I I I C Cn Q Q w ~ col II,47, 4J Ll, L 0 '~' d U C 0 0 0 0 D OA 0 0 L 0 L Ol J3 ' 0 E U Cw K c ao L CP PD P Q~ IOCw a wo od ~ ao ~ oc ~ L 0 C wo ca C <L OZ 0 Q 0 0 OA OV) 0 I w X 0 ~ J3M ~ QJ D NU wo V) 8 DB 0 C I O QW I d CO 0 f O L 0 X aw ca o Ulo I'tl VIdI-0 a o4 ~ U QCIo I I I I I IIII C) .ClI IXI O CA aN a Ul dLp Dl C 0 Ol Cl I CO CI III 0 O0 La. D0 4 CL 0 d C 0 M 0L n O. Ill d Old C~ 0D 0x 0 A I L I 0 L dE p 0LI C0 IL O IC) CI 10 0 I C IIIIII II II E J: 0 D 0 III I E'I LlI I 0 20 21 22 23 24 ,25 26 27 28 29 30 31, 1 2 3 4 5 6 7 8 October 1983 November 1983

~A

FIGURE Va. 2 Excoro Current vs Zncore A.O". N 41 ~ ) ~ ~ ~ ~ ~ ~ ~ ~ti jij It ~ ~I Hi P.! i) ~et Ite jIf'lij( j lj:.IHl (I'. Hi H} I' ie eti ie'<( !I "i! le> 1)0 i)t Upper Excore CurrentQ L le Lowe Ir Excore Curre t Q7 'Ifl.. H! H) !H.'H H( iil ~ 'I ~ 150 fla llj !H ~I~ 'I ~ I !10 II! i)1 j(i !H !H fjj !!i ~ eeIl: i) Hi flj Hi tef Hi!Ii'li !It I!: ;i( ~ I I'li!!f lil iii iil ili ilt iii iii

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Vb Plant Chemis Histo The Unit 1 refueling outage began on July 16, 1983. Prior to'his refueling outage, the reactor coolant system activities in general, and dose equivalent iodine-131 in particular, were quite 'ow. Dose equivalent iodine-131 was approx. 1.00 x 10-'Ci/cc. No fuel defects or leaks were suspected. After the shutdown, mmimum coolant system dose equivalent iodine-131 was 1.20 x 10-pCi/cc. Reactor coolant system degassing was complete in 39 hours in preparation for the refueling and in pm~cular, the addition of hydrogen peroxide for Co-58 solubilization. Once degassed and the reactor coolant system less than 180'F, 30% hydrogen peroxide was added to the primary coolant to "clean up" the core. A total of four (4) gallons of 30% hydrogen peroxide was added with a maximum detected reactor coolant Co-58 activity of 1. 14 vCi/cc. Cleanup via ma~mm> CVCS letdown purification flaw took approx. 52 hours of critical path tim and resulted in the remval of just over 300 curies of Cobalt-58. This standard cleanup operation results in the steam generator area radiation levels remaining at relatively low levels cycle after cycle. Maximum radiation level detected was 8 R/hr. Following this "cleanup", the unit continued into its refueling operation. Reactor coolant system monitoring during the subsequent startup has detected no new source of activity and the fuel integrity seems as sound as the previous cycle. The secondary system chemistry prior to shutdown was fairly good. Minimal adverse affect was noted due to condenser inleakage based on specifications in effect during that cycle. Steam generator limits during this cycle were <2.0 pmho cation conductivity with <100 ppb sodium contamination. Steam jet air ejector flow was 25-30 scfm actual with no specification in effect. When the new secondary cycle specifications were drafted, they were an initial effort to neet all the Steam Generator Owners Group Guidelines published by EPRI. The new specifications included reduction of the steam generator cation conductivity to 1.5 Rnho's, steam generator sodium to 20 ppb and established a 10 scf'm limit on the steam jet air ejector, anting other changes. In order to meet the new, rrore restrictive power operation specifications, limits were written for various steps during the startup. For example, prior to entering mme 2, prior to turbine roll, 30%, 50$, and 75-o power. The 75% power specifications were later amnended to be in effect at. 80% power or greater. Pull power operation, as far as secondary chemistry specifications, was defined at greater than 80% power. As is standard practice during refueling outages, the steam generators were sludge lanced to remve accumulated deposits from the tube sheet. This operation succeeded in xeaaving 445,

140, 445, and 140 lbs. of sludge from steam generators ll through 14.

As a means of achieving these new specifications, a program of steam generator crevice flushing was initiated in made 4. This basically involves rapid depressurization of the steam generators while at 280-300'F, causing agitation (boiling) at the tube sheet, and hopefully in the crevices. The exact procedural method for performing this activity remains in question; several methods were attempted not only at this site but by other utilities and no "standard" m thod exists to date. (EPRI is hoping to.issue a standard nathod for testing in March, 1984.) The crevice flushing activities created a 48 hour hold until the flushing was complete. Startup and power ascension was delayed due to holds caused by the new chemistry specifications. Operational eqmrience gained during this initial application of the new secondary-system specifications should be used to minimize holds in future startups.

V c PLANT RADIATION SURVEYS The adequacy of the shielding design for the reactor and associated primary coolant system was verified during tPe initial plant start-up. Surveys during subsequent start-ups were performed for the following purposes: 1. Verify the adequacy of any radiation and shielding that has been sig-, nificantly modified since the last start-up. 2. Verify the adequacy of radiation shielding systems that have been significantly modified such that the source term may have changed, thus changing the shielding adequacy. 3. Verify that general plant radiation levels have not significantly changed due to variations in sources that may have been affected by feed changes. No significant modifications were made to primary shielding or the primary coolant systems such that the source term for existing shielding would have been changed. As in past refueling periods, surveys of accessible areas were performed during the transfer of fuel from the refueling cavity, through the transfer

tube, and into the spent fuel pool.

No significant differences were found from previous similar surveys.

However, because of the rate at which fuel moves through the transfer tube, there was some doubt that all areas of ra-diation leakage had been identified.

In an effort to locate any radiation leaks that might only exist for a few seconds during normal transfer tube us-

age, a spent fuel element was "parked" in various locations of the transfer tube to allow a more thorough search for radiation leakage points.

Increases were seen inside some of the already posted high radiation ar-

eas, especially in the annulus near the Reactor Coolant Dr'ain Tank."

On the 633'evel, at the junction of the Containment dome and the Auxiliary Building, maximum levels went from 0.2 mR/hr to 50 mR/hr with fuel in the transfer tube. The most significant radiation level increases were noted inside the Con-tainment Building at the 612'rea between the ice condenser walls. During the time the fuel bundle was at rest in the transfer tube, the shield walls on either side of the transfer tube read a maximum of 17 R/hr. Although the time-weighted average exposure rate with fuel moving norma'Ily through the tube would not require posting the area, rope boundaries were established such that outside the boundaries the instantaneous dose-rate was less than 100 mRem/hr. The routine surveys and Shield Survey conducted at the start-up following refueling did not show significant differences from previous start-up surveys.

. "" Vd Reactor Coolant Flow ~sureaent The primary purpose of this test was to determine the total reactor coolant flowrate independent of the reactor coolant flow transmitters. The reactor coolant flowrate was computed from a steam generator heat balance calculation utilizing steam generator secondary side param ters. ln addition, it, was also the purpose of this test to recalibrate the reactor coolant flow elbow tap differential transmitters, as ~red, based on the computed reactor coolant flow rates and'elbow tap differential pressure data. The total coolant flowrate was determined at both the 48%, 68% and 100% levels of reactor thernel power. The table below indicates that the total reactor coolant flowrate fell within the acceptable region of the Technical Specification Graph 3/4.2-11. Table Vd.l Computed Reactor Coolant System Flowrates at 48%, 68% and 100% Reactor Thermal Out ut Loop Nl Loop N2 Loop N3 Loop N4 Total Flowrate Flow 8 48% 3.42 x 108 8 3.94 x 108 3.55 x 108 3.38 x 108 14.28 x 10 Flow 9 68% (I/hr) 3.52 x 108 8 3.92 x 108 3.61 x 108 3.48 x 108 14.53 x 10 Flow 9 100% (I/hr) 3.43 x 108 8 3.82 x 108 3 78 x 108 3.61 x 108 14.40 x 10 Table Vd. 2 Average Value of Plant Parameters Used in Flow Determination Parameter BCS Pressure RCS T Hot (F) RCS T Cold (P) Main Steam Pressure (psia) Feedwater Pressure (psia) Feedwater Temperature (F) Peedwater Plow (PPH) ,Value 9 488 2232 570.4 541. 3 878. 3 898. 1 370.9 1530696. Value 8 68% 2234 580.4 539.7 840.9 867.5 401.0 2292028. Value 9 100% 2232 595.6 537.4 789.6 827.5 433.7 3510713. The computed reactor coolant flowrates and the elbow tap differential pressure data, obtained as part of the 100% power data set, were used to recalibrate the reactor coolant flmr elbow tap differential transmitters.

Each reactor coolant loop flowrate was computed by performing heat balance calculations around the shell and tube side of the loop's associated steam generator. As is shown in Figure Ve-1 steam generator primary and secondary side pressure and tenperature data was simultaneously trended to provide the required inputs to the heat balance equations. Allof the required data was obtained while the unit was in a steady sta'te aude of operation and steam generator bio@down was secured. The feedwater flow, corrected for specific weight of the fluid and associated piping characteristics, was calculated from the venturi differential. The steam generator thermal output was then caqputed frcm the feedmter flow and the associated increase in enthalpy. Finally, this thermal output value was transferred to the primary side where it was used in conjunction with the reactor coolant enthalpy drop to calculate reactor coolant flmr. The total reactor coolant flow was simply the sum of the individual loop flowrates. Table Vd.l sulu.-izes the calculated flowrates. The total HCS flowrate was verified by Nuclear Section to fall within the acceptable region of Technical Specification Graph 3/4.2-11. Table Vd.2 lists the primary and secondary side plant param ter average values used in the reactor coolant flow determinations. The reactor coolant.flow determination was made at 48%, 68%, and 100% power levels. At each level the flow determination and the reactor coolant flow elks tap differential pressure data (trended simultaneously with the pressure and temperature data used in the heat balance equations) were used to evaluate the accuracy of the elbow tap differential transmitters. The current trip point for the indicated RCS flow is 93%. Since the actual BCS flow is greater than the indicated flow the 93% trip point is conservative and a transmitter recalibration was again deemed unnecessary. Vd-2

0 /f gl A

.Ve P ant Thermal Power Calibration The purpose of this test was to determine reactor thermal power by measuring secondary system feedwater flow and steam param ters, and to verify the accuracy of the following computer outputs: 1. Reactor thermal power 2. Feedwater flows 3. Feedwater temperatures 4. Nuclear power range insUmrentation By neasuring the secondary side paranaters the reactor thermal output can be calculated. The parameters that are measured for this power determination are feedwater flow, feedwater temperature, feedwater

pressure, and steam pressure.

The pcarer determination has been completed, at power levels of 48%, 68% and 100%. 8 Power Co uter Value (U-lll8) Calculated Value 48% 68~o 100%

47. 86

67. 90 99.50

47. 37

67. 36 99.09 During the initial poorer ascension-program the reactor thermal output was calculated at various pcwer levels by rreasuring secondary side parameters.

These param ters are suranarized in Table Ve.l. By measuring the feedwater parameters before the steam generators and the steam parameters after, the annunt of energy added by the steam generators was determined. The energy gained by the steam side of the steam generator is the equivalent energy given off by the reactor coolant system. By knowing the heat transferred by each of the four steam generators, the total heat added to the secondary side of the reactor coolant system is determined. This total heat added to the secondary side minus the heat, added by RCP operation and RCS system losses is the actual.reactor power. All measurements taken for this thermal pcwer rreasurerrent were from instruments calibrated specifically for this test independent of computer points. The pressure measurements were. read using dead weight testers, feedwater flows were measured at the local transmitter for each loop and temperatures were read using thexaxeouples installed in test wells for each loop's feedwater piping. Before the test is begun the steam generator blcwdown is isolated and all plant parameters are as stable as possible. These conditions prevailed throughout the test. The computer thermal power is monitored during the actual thermal pcwer test. After the test is completed, a cczrgxrrison is made between the computer value and the actual rraasured value.

THERMAL PGKR CALIBRATIONDATA ~o POKER FK PRESSURE (PSIA) m TEMPERATURE ( F) Sm. PRESSURE (PS') ADULATED CGWVXKR UALUE (U-lll8) ~ 2 47.37 67.36 99.09 47.86 67.90 99.50 897.72 866.42 827.68 897.72 866.52 826.48 898.22 898.82 369.70 866.52 867.52 400.24 825.38 830.58 433.31 371. 20 401.45 434.42 371. 49 370.34 401.79 400.80 434.21 433.00 878.7 841:5 788. 6 877. 9 840.4 785.9 878.1 . 878.3 839.4 842.1 783.2 789.6 TABLE Ue;1

~r FlGURE Veal nc hs Stean Press RC Tc P ess Reac-o <ogle=-" Flow'pete~--'-p 'o~ Peac.or Coolant Out hH ~RC Press Th Steam Generator Steam Out hf F'I FM . Temp Press .Q Q 'Ha0 Reac.or C'oolant In Feedwater In 1 Fa d.iazar Fio:grata = !li=(3B.i)d-'cFa(VY)(Via30) '~i-(alai klhere: . d= throat diameter of venturi {inches) C= coefficient "ef-discharge Fa= venturi thermal expansion fac.or Y= specific weight of feedwat r (-.";/ft-) <H20= di eren ial pressure across venturi (inches) D= pipe diameter at inlet pressure,tap (inches) 2. Stean Cenerator Thermal Ou put = Bs = lI (n -hf) >!here,: hs = enthalpy of steam (BTU/.=.) hf -" enthalpy of feedwater (BTU/.-".) 3. Reac or Coolant Loop Flow = Bs/(hH-hc) , Where: hH = enthalpy hot leg (BTU/'-) hc = enrhalpy cold leg (STU/;":) 4. Total Reactor Coolant Loop Flow = sum of.he loop ilows

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