ML17334B395

DC Cook Nuclear Plant Unit 2,Cycle 8 Startup Rept. W/

ML17334B395
Person / Time
Site:	Cook
Issue date:	02/22/1991
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AEP:NRC:10710, NUDOCS 9103010150
Download: ML17334B395 (55)

Text

ACCELERATED DI~BUTION DEMONSTITION SYSTEM REGULATORY INFORMATION DXSTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9103010150 DOC.DATE: 91/02/22 NOTARIZED: NO DOCKET FACIL:50-316 Donald C.

Cook Nuclear Power Plant, Unit 2, Indiana 05000316 AUTH.NAME AUTHOR AFFXLIATION ALEXICH,M.P.

Indiana Michigan Power Co.

(forme y Indiana

& Michigan Ele RECIP.NAME RECIPIENT AFFILIATION R

MURLEY,T. E.

Document Control Branch (Document Con rol Desk)

SUBJECT:

"DC Cook Nuclear Plant Unit 2,Cycle 8 Startup Rept."

W/

910222 ltr.

DISTRIBUTION CODE:

IE26D COPIES RECEIVED: LTR

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ENCL

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SIZE:

TITLE: Startup Report/Refueling Report (per Tec Specs)

NOTES D

S RECIPIENT ID CODE/NAME PD3-1 LA COLBURN,T.

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NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

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Indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 INDIANA NICMIGAN POWER AEP:NRC:10710 Donald C.

Cook Nuclear Plant Unit 2 Docket No, 50-316 License No.

DPR-74 DONALD C.

COOK NUCLEAR PLANT UNIT 2, CYCLE 8 STARTUP REPORT U,S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.

20555 Attn: T. E. Murley February 22, 1991

Dear Dr. Murley:

In accordance with the requirements in Donald C.

Cook Nuclear Plant Technical Specifications 6.9.1.1 through 6.9.1,3, attached is the Unit 2, Cycl'e 8 Startup Report, Unit 2 Cycle 8 is the first Unit 2 reload supplied by Westinghouse since Cycle 3.

Westinghouse has replaced Advanced Nuclear Fuels (ANF), which supplied the previous four reloads.

The Unit 2, Cycle 8 core consists of 193 fuel assemblies, 77 of which were newly fabricated by Westinghouse.

The remaining 116 assemblies are ANF assemblies that have been used in previous Unit 2 cycles.

Unit 2, Cycle 8 began on November 8, 1990 and the Startup Test Program was completed on November 23, 1990.

The Startup Report is required by Technical Specification 6.9.1.1(3) in the case of "installation of fuel that has a different design or has been manufactured by a different fuel supplier."

The Startup Report is being submitted within 90 days following completion of the startup test program as required by Technical Specification 6.9.1.3(l).

The contents of the report are identified in Technical Specification 6.9.1.2, which requires in part that "the startup report shall address each of the tests identified in the FSAR.

The tests identified in the FSAR are tests which were to be performed at the beginning of Unit 2 Cycle 1.

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

Those FSAR tests not required to be performed for Unit 2 Cycle 8 are addressed in the Unit 2 Cycle 1 Startup Report.

The tests that were performed for Unit 2 Cycle 8 are addressed in detail in the attached Startup Report.

"'9103010150 910222 PDR ADOCK 0500031'>

P PDR

~Kg(

Dr. T. E. Murley AEP: NRC: 10710 This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

Sincerely,

/

M. P. Alexich Vice President ldp cc:

D. H. Williams, Jr.

A. A. Blind J.

R. Padgett G. Charnoff A. B. Davis - Region III NRC Resident Inspector

- Bridgman NFEM Section Chief

ATTACHMENT TO AEP:NRC:10710 DONALD C.

COOK NUCLEAR PLANT UNIT 2, CYCLE 8 STARTUP REPORT

...9 103010150

DONALD C. COOK UNIT 2 CYCLE 7/8 OUTAGE and STA RT UP REPORT February 11, 1991

TABLE F

NTE T SEE%M TITLE I.

Introduction Fuel Related Activities II.1 Core Unload/Reload and Fuel Inspection II.2 Ultrasonic Fuel Examination of Irradiated Fuel II.3 RCCA Examination Flux Mapping System Activities III.1 Thimble Tube Cleaning III.2 Eddy Current Examination of Incore Thimble Tubes III.3 Thimble Tube Replacement Project III.4 Flux Mapping System Five-Path and Ten-Path Refurbishment IV.

Rod Drop Testing V.

Initial Criticality Zero Power Physics Testing VII.

Power Ascension Activities VII.1 Power Ascension Testing VII.2 Reactor Coolant Flow Measurement VII.3 Plant Thermal Power Calibration VIII.

Plant Chemistry History

The Unit 2 Cycle 7/8 outage began with shutdown of the reactor on June 30, 1990 after six days of power coastdown.

The total Cycle 7 burnup was 17,854.7 MWD/MTU. Cycle 8 began with initial criticality on November 8, 1990 and 100%

reactor thermal power (RTP) was reached on November 21, 1990.

Startup testing was completed on November 23, 1990 following the analysis of Flux Map 208-06.

The feel movement sequence for this outage was accomplished by using the total core unload and reload technique. The Unit 2 Cycle 8 core consists of 116 Advanced Nuclear Fuels (ANF) and 77 Westinghouse VANTAGE-5fuel assemblies and is the transitional cycle to VANTAGE-5fuel. The unload took place over July 26-29, 1990 and the reload occurred over September 2-4, 1990. The refueling campaign had no fuel handling sequence errors; a video inspection of the core confirmed proper fuel loading as compared to design.

In addition, the binocular inspection performed during unload/reload revealed no structural damage to any fuel assemblies.

The approach to criticality began with the withdrawal of the Shutdown Banks at 1521 hours0.0176 days <br />0.423 hours <br />0.00251 weeks <br />5.787405e-4 months <br /> on November 7, 1990.

Rod Position Indicators (RPIs) and Bank Demand Counter differences caused a minor delay in the approach to criticality. Dilution began at 2010 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.64805e-4 months <br />, and the reactor was critical on November 8, 1990 at 0236 hours0.00273 days <br />0.0656 hours <br />3.902116e-4 weeks <br />8.9798e-5 months <br />.

After the reactor reached steady state conditions, data was taken to determine the Point ofAdding Nuclear Heat, Zero Power Physics Testing Range, and AllRods Out (ARO) Critical Boron Concentration.

The Zero Power Physics Testing Program began at 0340 hours0.00394 days <br />0.0944 hours <br />5.621693e-4 weeks <br />1.2937e-4 months <br /> on November 8, 1990; the program included ARO Isothermal Temperature Coefficient (ITC) Determination and Rod Worth Measurements.

The ARO ITC was +2.31 pcm/'F as compared to a design value of + 1.54 pcm/'F. The Moderator Temperature Coefficient (MTC) calculated from the ARO ITCwas +4.89 pcm/'F which is less than the Technical Specification Limit of +5.0 pcm/'F below 70% RTP. All rod worth and boron endpoint measurements compared favorably with design values and excess shutdown margin was verified.

Power Ascension Testing began on November 9,

1990, and was completed on November 23, 1990, The testing program included Flux Maps at approximately 34%,

47%, 66%, 88%, 86%, and 99% RTP. The testing also included an Incore/Excore Cross Calibration using the One Point Methodology at approximately 47% RTP.

Power Ascension Testing went smoothly, with the hot channel factors calculated from the incore flux maps being within the required Technical Specification Limits. The 86% RTP Flux Map was taken due to the indication of a Nuclear Instrumentation System Quadrant Power Tilt Ratio in excess of 2%. The Quadrant Power TiltRatio obtained from the analysis of this Flux Map satisfied the Technical Specification Limit. The power range drawers were recalibrated to remove the instrument tilt.

In general, all startup tests were relatively routine.

The tests were conducted in a timely and expedient manner and resulted in accurate startup information.

The collected data and test results compared well with design predictions and satisfied all of the test's Acceptance Criteria and all Technical Specification Limits.

As stated in Section 6.9.1.2 of Unit 2 Technical Specifications, the tests identified in the FSAR shall be addressed in the Startup Report. The tests in the FSAR are tests which were to be performed at the beginning of Unit 2 Cycle 1.

Not all of these tests need to be performed on a reload cycle. Those FSAR tests not required to be performed on this reload core are addressed in the Unit 2 Cycle 1 Startup Report.

The tests that were performed on this reload core are addressed in detail in this report.

Section II FUEL RELATED ACTIVITIES

111 RE NL AD REL AD AND F EL IN PE TI N

The Unit 2 Cycle 7 core unload sequence began at 1035 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.938175e-4 months <br /> on July 26, 1990 and was completed at 0027 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> on July 29, 1990.

The Cycle 8 reload sequence commenced on September 2, 1990, at 0305 hours0.00353 days <br />0.0847 hours <br />5.042989e-4 weeks <br />1.160525e-4 months <br /> and was completed on September 4, 1990, at 0649 hours0.00751 days <br />0.18 hours <br />0.00107 weeks <br />2.469445e-4 months <br />.

The Unit 2 Cycle 7 core consisted of 26 Region 7, 87 Region 8, and 80 Region 9 assemblies.

Regions 7, 8, and 9 were all ANF fuel assemblies.

The new core loading of Cycle 8 consists of 8 Region 7, 28 Region 8, 80 Region 9, 36 Region 10A, 40 Region 10B, and 1 Region 10C assemblies.

Twenty-six (26) Region 7 and 59 Region 8 ANF assemblies were replaced by 77 fresh Region 10 Westinghouse VANTAGE-5 assemblies and 8 twice-burned Region 7 ANF assemblies which were last used during Cycle 6. A more detailed description of the VANTAGE-5fuel assembly is provided in Appendix II.1. Core loading diagrams for Unit 2 Cycle 7 and Cycle 8 are shown in Figure II.1 and II.2, respectively.

Figures II.3 and II.4 illustrate the differences of the core designs for both cycles.

No fuel handling sequence errors (ie., assemblies put in the wrong location) were reported during this campaign, and only 19 Fuel Movement Deviation Reports (FMDRs) were generated.

The majority of the FMDRs were to resolve difficulties with fuel assembly bowing and the transfer system (upender).

For the first time, a "horseshoe" was used at Cook Plant during the reload.

The horseshoe proved its worth by greatly reducing the number of fuel assembly "boxes" necessary to position assemblies in their final core location.

A binocular inspection of the fuel took place at the Spent Fuel Pit (SFP) area for the unload. During the reload, however, the binocular inspection took place at the SFP as well as inside containment.

As each assembly was unloaded, all four sides were inspected for any structural damage or abnormalities. A final check for damage was

'completed as the assembli'es were reloaded into the core. The binocular inspection of the fuel assemblies revealed no structural damage or abnormalities.

At the conclusion of the core reload,.a video inspection of the core and a visual inspection of the SFP were performed, as required by 12 THP 4040 SNM.302 and SNM.304, respectively.

No discrepancies were found during either inspection.

FIGURE Il.l D.C. Cook Unit 2 Cycle 7 Core Map 0

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APPENDIX II.I WESTINGHOUSE 17 x 17 VANTAGE-5 FUEL ASSEMBLY Unit 2 Cycle 8 core presently consists of 77 Westinghouse and,116 ANF fuel assemblies.

Westinghouse's safety evaluation and analyses concluded that the VANTAGE-5and ANF designs are mechanically and hydraulically compatible with each other.

Table II.1 is a list of the design features for the ANF and Westinghouse fuel assembly.

Figure II.5 is a schematic of the two assembly's dimensions.

The major differences in the Westinghouse VANTAGE-5design as compared to the ANF fuel assembly consist of the following:

Integral Fuel Burnable Absorber (IFBA):

The IFBA is a fuel pellet that has its surface coated with a thin layer of zirconium diboride.

IFBAs are used to control power peaking factors and reduce the beginning of cycle moderator temperature coefficient.

Intermediate Flow Mixer (IFM) Grids:

The VANTAGE-5 design incorporates three Zircaloy-4 Intermediate Flow Mixing (IFM) grids.

The three IFM grids are located in the uppermost spans between the three Zircaloy-4 structural grids.

Increased DNB margin is realized by the mid-span flow mixing in the hottest fuel assembly spans.

The increased DNB margin permits an increase in the design basis F'and Fo.

Reconstitutable Top Nozzle/Extended Burnup Capability:

The VANTAGE-5 top nozzle consists of a design feature which facilitates easy removal of the nozzle from the fuel assembly for fuel rod examination or replacement.

Changes in the design of the top and bottom nozzles have allowed for a slightly longer fuel rod.

The increased length extends the burnup margins by providing additional plenum space for fission gas accumulation.

Axial Blankets:

The axial blankets are a nominal six inch stack of natural UO, pellets at each end of the fuel stack.

The purpose of the blankets are to reduce neutron leakage and improve uranium utilization.

TABLE 11.1 Comparison of Westinghouse VANTAGE-5 and ANF Assembly Design Parameter VANTA E-5 Desi n ANF Desi n Fuel Assy Length, in Fuel Rod Length, in Assy Envelope {width), in Compatible with Core'Internals Fuel Rod Pitch, in Number of Fuel Rods/Assy Guide Thimble Tubes/Assy Instrumentation Tube/Assy Fuel Tube Material Fuel Rod Clad OD, in Fuel Rod Clad Thickness, in Fuel/Clad Radial Gap, mil

'Fuel Pellet Diameter, in

~F' P il Enriched Fuel, in Unenriched Fuel, in Guide Thimble Material 159.975 152.285 8.426 Yes 0.496 264 24 Zircaloy-4 0.360 0.0225 3.1 0.3088 0.370 0.500 Zircaloy-4 159.710 152.065 8.426 Yes 0.496 264 24 Zircaloy-4 0.360 0.0250 3.5 0.3030 0.348 N/A Zircaloy-4 Guide Thimble OD, in (above dashpot) 0.474 0.480

FI RE II 5 COMPARISON OF WESTINGHOUSE VANTAGE-5 and. ANF FUEL ASSEMBLY DIMENSIONS I,7AS w 1.660 ANP l59.97S w 159.710 ANF 152.26S w

> 52.06S ANF 2.565 w 2.720 ANF 5 LEAP SP RIN0 5.475 W

%560 ANF 152.64W t 122.$ $ W 122.07 W $111,60W 101.$ 2 W IS2 74 122.70 112.1$

ANF ANF ANF 91.60 AN7 71.05 ANF 91.2$ W 60.97 W 70.70 W

50. IS W CV C5 29.6OW S.O7 W Q 29.9S

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W - WESTINGHOUSE 17x17 VANTAGE S FUEL ASSEMBLY DIMENSION+

ANF ADVANCED NUCLEAR FUEL 17x17 FUEL ASSEMBLY DIMENSION~

ANF RID WIDTH o 2.50 (valley to valley) o 2.96 (peak to peak)

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'imensions are in inches (nominal)

11.2 VLTRA Nl EXAMINATION F IRRADIATEDFUEL A SEMBLIE Advanced Nuclear Fuels (ANF) provided fuel Ultrasonic Testing (UT) services. The UT began at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on August 17, 1990 and was completed at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br /> on August 19, 1990. The 116 irradiated fuel assemblies designated for the Cycle 8 core were tested.

Predictions of two to five fa'iled fuel pins were made during the Cycle, with an end of Cycle prediction of four failures.

The UT campaign revealed five failed fuel rods; all were found in assembly T-24.

No core redesign was necessary based on the results of the UT campaign.

The UT System works by a probe transceiver sending a high frequency sound wave into a fuel pin and measuring the strength of the returning signal, or "ring back". A fuel pin can be determined to have water in it by monitoring the relative strength of the ring back.

A fuel assembly is tested from two sides; however, all four sides of an assembly may be tested if an indication of a failed rod exists.

During the UT campaign, fuel assemblies V16, V13, V12, and V70 had suspicious indications of failed rods due to their low returning signals.

As a result, these assemblies were retested later in the campaign from all four sides at slightly different elevations and they were determined to be sound.

The initial results of the low returning signals were believed to be caused by poor probe alignment. The final UT results showed no failures in all 116 ANF fuel assemblies.

In addition to the 116 reload assemblies, the discharged assembly T-24 was also tested.

Failed rods in the assembly were detected in locations G01, G03, K10, H16, and H17. This assembly had also been tested during the Cycle 6/7 UT campaign in 1988, and rods G01, K10, H16, and H17 were found to be failed at that time. The UT results on assembly T-24 were close to the end of Cycle 7 prediction.

Visual inspections of two sides of assembly T-24 were performed and video taped with the underwater camera mounted on the UT system.

The failed rods on the periphery of assembly T-24 showed signs of hydride blisters.

II.3 R

A EXAMINATION All53 Rod Cluster Control Assemblies (RCCAs) ofUnit 2 were inspected during the Cycle 7/8 outage by Westinghouse and Echoram personnel.

There were 3.5 days of delay prior to the start of the RCCA examination due to repairs and tests on the Spent Fuel Handling Crane.

Eddy current inspection of the Unit 2 RCCAs began at 1940 hours0.0225 days <br />0.539 hours <br />0.00321 weeks <br />7.3817e-4 months <br /> on August 5, 1990, and was completed at 0718 hours0.00831 days <br />0.199 hours <br />0.00119 weeks <br />2.73199e-4 months <br /> on August 8, 1990.-

The RCCA eddy current inspection equipment consists of an eddy current coil assembly guide fixture and a data acquisition/analysis system.

The eddy current coil assembly guide fixture is placed on top of the spent fuel pool racks.

The data acquisition/analysis system is set up on the operating deck of the spent fuel pool.

The eddy current inspection is based on the principle of Electromagnetic Induction.

When a metal object is placed in a varying electromagnetic field, electromagnetic currents are induced in the object which tend to oppose the external field.

The strength of these induced electromagnetic fields depends on the amount of magnetic

, permeability, conductivity, and shape of the metal object to be tested.

The strength of the induced fields causes the electrical impedance of the exciting electromagnetic field source to change. Ifan anomaly such as a reduction in cross-section or a crack enters the electromagnetic field, the impedance of the system would change.

Thus, anomalies such as areas with worn metal or cracks in the cladding can be measured by lowering a RCCA through the eddy current coil assembly guide fixture.

The results showed normal wear patterns on the rodlets near'he guide card region and longitudinal cracks on several rod tips.

These wear patterns were expected based on tests at other plants utilizing Westinghouse RCCAs and willnot affect the absorbing capability of the control rod. The results also indicated that the wear had penetrated the cladding of rodlet E5 of control rod R15.

A new control rod was used to replace R15 prior to reloading the Cycle 8 core.

Section III FLUX iIAPPING SYSTE'5'I ACTIVITIES

III.1 THI'ABLET BE LEAtglbt'PEX Technologies provided services to clean the Unit 2 incore thimble tubes. The cleaning commenced with demineralized water flushing and vacuum drying of the thimble tubes on July 14, 1990. Thimble cleaning was completed on July 21, 1990.

Problems were encountered during the cleaning due to the C-7 thimble leak and ten-path indexer leaks.

The indexer leaks allowed oil to be released from their reservoirs, which settled on the bottom of the ten-path housings.

When the C-7 thimble leak occurred, oil was flushed from the bottom of the ten-paths and down into the thimble tubes.

Fifteen thimble tubes were completely plugged.

Also, the remaining thimbles showed signs of the oil and water mixture. All thimble tubes were cleaned successfully and were verified passable by use of a test cable.

The excessive amount of water also filled the five-path transfer boxes.

Apex assisted in pumping out the water from the five-path transfer boxes.

They also cleaned the interconnecting tubing runs from the ten-paths to the seal table and from the five-paths to the ten-paths.

The general arrangement of the reactor, thimbles, and seal table is presented in Figure III.1.

Figure III.2 shows the equipment layout for demineralized water flushing and air drying of thimble bores, and Figure III.3 shows the isolation valve and thimble orientation of the seal table.

I I

FIGURE III.2 Demineralized Water Flush and Air Drying of 1'l>imhle Bore

~ 4l'El DEMIT.WATER SUPP< Y6 l5O Prig MiH.

( ~g kh~)

c (kz(E.

WC%.R.

~<<<<nrou)

HO~

Ra.a<

(ulLT6,l4 Fa uSg)

WATS Di.~uidqZ A5>Y A PEX a<r>SC Raw~

(aiR O>y) h<klaO~~R TJIB i~5'.

6/

v1 iq>T. 4.1+

Mc~g tpioo p5l+ Mct4

(/q war)

AIR KA4ieOa C PRess.

G ~00' IJbRl(ATOR Ri4avu.tote.

Plta&b C a~C, l 5vpp < HOSC SVPPoczr ~i~ h~ ~ r 59 vT<b<K bRAsM 5,

(@hat)

~ u, gig+0 4

fuiMbuKQ ( COHbvsl 5 To %@he:kg

FIGURF. Ill.3 Isolation Valve and Thimble Orientation at Seal Table TOWARD REACTOR 1

N13 N14 Ll

++ V4 SEAL TABLE Jl H3 N4 G5 N6 F7 J8 L8 G9 L10 H11 K12 L13 J14 H15

+- W + + + M~ W W + W +- ~ 4 DRAIN HO[ E (2)

F1 D3 B3 C5 H6 B6 B8 C8 A9 D10 A11 D12 B13 D14 TYPICAL LOCATION OF ISOLATION VALVE, THIMBLE AND CONDUIT AT SEAL TABLE (58 LOCATIONS)

CORE LOCATION NUMBER

111.2 EDDY RRENT EXAMIViATIOV F IVi RE THI IBLE T BE Cramer and Lindell Engineers, Inc. provided services for the Eddy Current Examination of Unit 2's 58 Incore Thimble Tubes.

The Eddy Current exam of the Incore Thimble Tubes commenced at approximately 1530 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.82165e-4 months <br /> on July 18, 1990 and was completed at approximately 0026 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> on July 19, 1990.

Bubble Hoods were used during the examination due to the need for respiratory protection and effective communication during the exam.

No major delays were encountered.

The setup of the eddy current equipment was quite simple. An eddy current probe that is welded to a helLv cable was manually fed to the top each thimble.

During withdrawal of the probe, eddy current signals were passed to the data aquisition station located just out8ide of containment.

Eddy current data of all 58 thimbles was collected and analyzed by the vendor.

Twenty-nine thimble tubes were determined to have greater than 30'/fo wall loss.

Yine of those thimble tubes had a wall loss of greater than 607o, including the leaking thimble tube, C-7. The majority of the wear was determined to be located at the fuel assembly bottom nozzle.

Due to the severity of the thimble tube wear, ten thimble tubes were replaced and 19 thimbles tubes were repositioned.

111.3 THIMBLET BE REPLA EMEYT PRO E T The Thimble Tube Replacement Project was performed by Apex Technologies in three different phases at the Seal Table and Reactor Cavity. An additional activity of preparing the new thimbles was completed on the Turbine Deck. Ten thimbles were replaced, and 19 thimbles were repositioned.

The project was completed in the order as follows:

Phase I: Seal Table Preparation with Reactor Cavity Drained Phase I had two objectives.

The first objective was to cut off approximately 15 feet of the "cold" (non-irradiated) end of the thimbles to be replaced.

The second objective was to prepare for the removal of the thimbles through the Reactor Cavity.

This was accomplished with slide seal valve assemblies which were installed at the seal table to act as low pressure seals during the removal of the thimbles.

The'unction of the slide seal valve assembly was to allow personnel at the seal table to push the thimbles up through the lower core plate to personnel stationed at the Reactor Cavity while the cavity was filled.

Phase I w'as started on August 8,

1990 and was completed in approximately two hours.

There were no major delays.

New Thimble Tube Preparation

- Cut and Clean This activity was completed on the Unit 1 Turbine Deck.

The new thimbles were unpackaged, measured, cut to length, and cleaned.

The same cleaning technique was used as ifthe thimbles were installed in the reactor.

The thimbles were first flushed with demineralized water and vacuum dried.'

fine coating of neolube was then applied and allowed to air dry.

This phase was started on August 17, 1990 and was completed on August 18, 1990.

Again, there were no major delays.

Phase II:

Removal of Spent Thimbles from Reactor with the Reactor Cavity Flooded The second phase of the project involved the actual removal of the thimbles from the Reactor Cavity.

The basic procedure for the removal involved pushing the thimbles up through the core plate and grappling onto them.

After this was accomplished, the "hot end" was

moved underwater to the upender and the "cold end" was grappled.

The "cold end" was then taken out of the water and cut into pieces until the radiation fields would not permit anymore cutting above water. At this point the "hot end" was cut underwater and placed into a canister (two were used) which was,positioned in the upender.

The canisters were then moved to the Spent Fuel Pool for storage.

The leaking thimble, C-7, and the next most worn thimble, A-9, were placed in a special canister for later retrieval and analysis.

The wear scars on those two thimbles were video taped with an underwater camera.

Phase IIwas started on August 26, 1990 and was completed on August 29, 1990.

Major delays were encountered in this phase mainly due to extra protective measures required for the radiological conditions.

Phase III: Nevv Thimble Installation and Thimble Repositioning The final phase involved the installation of the 10 new thimbles and the repositioning of 19 others.

Also, the guide tubes of the new thimbles were vacuum cleaned.

The new thimbles were passed through a containment penetration and loaded into their respective locations.

New high pressure seals were installed on the thimbles and connected to the seal table.

The repositioning involved removing the old high pressure seal ferrule sets and cutting off a selected amount of thimble tube, to move the wear scare away from its original position.

New high pressure seals were installed and the connections to the seal table were made. After the ten new thimbles were installed and the 19 others repositioned, an additional Eddy Current Examination was performed to provide baseline data for the next refueling outage.

Phase IIIwork commenced on September 11, 1990 and was completed on September 12, 1990.

Only minor delays were encountered during this phase.

III.3-2

III.4 Flux Ma in v tern Five-Path and Ten-Path Refurbi hment The Ten-path refurbishment of the Unit 2 Flux Mapping System was scheduled to be completed by the plant IAE Section to relieve'the problem of indexer oil leaks.

The indexers were originally installed with a plexiglass baseplate for refueling inspections of oil level. After years of operation, the plate and the sealing gasket began to show signs of wear and allowed small quantities of oil to leak from the indexer.

The oil ran down the indexer and onto the small micro-switches which provide detector core location during a Flux Map.

The oil made these switches sticky, and in some cases, give false indication of detector location on the core map.

A replacement baseplate and gasket were installed to relieve the problem.

The inspection of the flux mapping system after the C-7 thimble leak shoived that extensive repairs would, have to be completed on the system.

All 6 ten-paths had been flushed with -water, and all 6 of the t'ive-path had at least 5 to 6 inches of standing water inside them.

Because the water damaged all electrical components inside of the five and ten-paths, they had to be replaced.

Also, after further inspection by ISSUE, it was determined that a power supply would have to be replaced in Detector F's drive box.

The parts replaced during the refurbishment included:

o 12 Transfer Motors o 12 Clutches 0 12 Baseplates and Gaskets o 150 Micro-Switches

~ 1 Power Supply o 1 Volt Meter The project started on July 13, 1990 with the ten-path removal 'off the support rack and was completed on October 6, 1990 with a successful system checkout.

IV, ROD DR P TE TING Rod drop testing commenced at 0715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br /> on October 13, 1990, with the briefing of all personnel involved in the test and was completed with the repeat drops of the slowest rod, H-14, at 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br /> on October 13, 1990.

Rod drop times were all approximately the same, with the slowest drop time being 1.53 seconds to the dash pot for Rod H-14, and the fastest rod drop time being 1.34 seconds to the dash pot for Rod J-13.

The slowest rod, H-14, was dropped four additional times to check for repeatability. The four drop times were 1.52, 1.50, 1.50, and 1.50 seconds.

The rod drop testing proceeded smoothly with no delays.

The results and shapes of the rod drop traces were in agreement with the previous cycle and no discrepancies were identified. However, the rod drop times slightly increased for some of the rods.

As expected, this was predominately seen in the VANTAGE-5fuel assemblies, which have a smaller diameter guide tube than the ANF fuel assemblies.

All rod drop times were well within Technical Specification 3.1.3.4 Limit of 2.7 seconds.

V. INITIAL RITI ALITY Unit 2 Cycle 8 achieved Initial Criticality at 0236 hours0.00273 days <br />0.0656 hours <br />3.902116e-4 weeks <br />8.9798e-5 months <br /> on November 8, 1990. The All Rods Out Boron Concentration (CD) was calculated to be 1654.9 ppm as compared to the design value of 1687 ppm and was well within the Acceptance Criteria of ~ 75 ppm.

On October 22, 1990, Reactor Engineering Section Personnel were requested to start Low Power Phvsics Testing.

At 1411 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.368855e-4 months <br />,

• • 12 THP 6040 PER.357, "Initial Criticalitv, All Rods Out Boron Concentration and Nuclear Hearing Level", began with the briefing of Operations and Chemistry personnel.

At 1435 hours0.0166 days <br />0.399 hours <br />0.00237 weeks <br />5.460175e-4 months <br /> withdrawal of Shutdown Bank "A" commenced followed by the remaining shutdown banks and control banks.

With Control Bank "D" (CBD) at 194 steps, Reactor Engineering Section was notified to halt Low Power Physics Testing until the steam dump valves were completely repaired.

The Shutdown and Control rods were reinserted to the bottom of the core and PER.357 was aborted.

The approach to criticality, again, began with the withdrawal of the Shutdown Banks at 1521 hours0.0176 days <br />0.423 hours <br />0.00251 weeks <br />5.787405e-4 months <br /> on November 7, 1990.

Next, withdrawal of the Control Banks in

overlap, as shown in Figure V.l, began at 1636 hours0.0189 days <br />0.454 hours <br />0.00271 weeks <br />6.22498e-4 months <br />.

Source Range Data was monitored throughout withdrawal, with the Control Banks stopped every 50 steps to plot the Inverse Count Rate Ratio (ICRR) as shown in Figure V.1. Withdrawal of the Control Banks to CBD at 194 steps (approximately 100 pcm of negative reactivity in the core) was completed at 1736 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.60548e-4 months <br /> on November 7, 1990.

On two rods, Rod Bank Demand Counter and Analog Rod Position Indicator differences of greater than 12 steps delayed dilution to criticality approximately two hours.

At 2010 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.64805e-4 months <br />, RCS dilution began w'ith CD approximately equal to 2077 ppm.

Mode 2 was declared at 2342 hours0.0271 days <br />0.651 hours <br />0.00387 weeks <br />8.91131e-4 months <br /> with a boron concentration of 1807 ppm.

At 0232 hours0.00269 days <br />0.0644 hours <br />3.835979e-4 weeks <br />8.8276e-5 months <br /> on November 8, 1990 dilution was stopped and subsequently, the Reactor was declared critical at 0236 hours0.00273 days <br />0.0656 hours <br />3.902116e-4 weeks <br />8.9798e-5 months <br />.

The Reactor's stable critical conditions were CBD at 185 i/zsteps, flux level at 1x10~ amps, and Cn equal to 1636 ppm. During the dilution to critical, ICRR vs. CD, Primary Water, and Time were plotted and are shown in Figures V.2, V.3, and V.4, respectively.

After the Reactor stabilized at 1x10 amps, data was obtained to determine the Nuclear Heating Level ~ The objective of determining Nuclear Heat was to set the upper limit of the Zero Power Testing Range.

The Nuclear Heating Level was observed as a small decrease in reactivity due to feedbacks (Doppler) caused by an increasing neutron flux. At the same time, an increase in RCS temperature was also observed due to the addition of Nuclear Heat.

The Nuclear Heating Level was determined to be 2.2x107 amps.

Prior to withdrawing the Control Banks, y

background noise data was collected for determining the Zero Power Testing Range lower limit. Thus, the Zero Power Testing Range was set to obtain the least amount of error in reactivity measurements.

The Zero Power Testing Range was determined to be 1x10 to 8x10 amps which was 1.17 decades above the y background noise level (6.8 x10' and 0.44 decades below the Nuclear Heating Level. The Subcritical, Nuclear Heating Level, and Zero Power Testing Range test data is summarized in Table V.l.

TABLEV.l Subcritical Data (Shutdown Banks Withdrawn, Control Banks Inserted)

Induced current with 1000 V applied to Detector N41:

Measured

= 6.8x10'mps 90% value = 6.1x10',amps 50% value = 3.4x10'mps Nuclear Heating Level Flux Level = 2.2x10 'amps Zero Power Physics Testing Range'lux Range

= 1x10~ to 8x10 amps Due to the quality of the data obtained, no compensating current was required during Zero Power Physics Testing.

V-2

ICRR VS. CONTROL BANK WITHDRAWALIN OVERLAP 1.0 Figure V.1 MOST CONSERVATiVE CHANNEL I

I Unit 2 C cle 8 0.9 0.8

~

I 0.7 0.8 CLO 0.5 0.4 I

I T

1 I

I 1

I I

i I

l I

I

~

~

I II

~

I I

I

~

4

~

i l

I 0.3 0.2 0.1 0.0 0

20 40 60 80 100 120 140 160 180 200 220 228 0

20 40 60 80 100 120 140 160 180 200 220 228 CM C2M) 0 20 40 60 80 100 120 140 160 180 200 220 228 0

2O 40 60 80 100 120 140 160 180 200 220 228 ROD POSITION (STEPS)

IN OVERLAP V-3

ICRR VS. BORON CONCENTRATION Figure V.2 MOST CONSERVATIVE CHANNEL Unit 2 Cycle 8 CBD O 194 STEPS 1.0 0.9 0.8 0.7 0.6 K

O 0.5 0.4 rL IL CV r

I

E I rL rL I

sO I r+

C5 Wo orLo rQ E

o rL rL raCl I

K C5 Irn Q

XorLoS

. rr0o IL rL Ol lO l

rL 0

I 2:

C5 8

E 5

0.3'.2 0.1 0.0 2100 2000 1900 1800 1700 1600 EIORON CONCENTRATION (PPM)

V-4

ICRR VS. PRIMARY WATER Figure V.3 MOST CONSERVATiVE CHANNEL Unit 2 Cycle 8 CBD O 194 STEPS 1.0 0.9 0.8 0.7 0.6 O

0.5 0.4 IO

,sAO Pl

D

IL IcL 4A

~+

K Oi~

Io O

IAI' IL Om

~E CL' O

I I

I I

I CVO 0)

D 0'.

tL I

OtL Om O

O I

I I

0.3 0.2 0.1.

Note:

Indicated primary water added to the RCS was determined to be greater than actual amount added.

0.0 0

10 12 14 16 18 20 22 PRIMARY III/ATER, GALLONS (x]000)

V-5

ICRR VS. DILUTION TIME Figure Y.4 MOST CONSERVATIVE CHANNEL DILUllONSTARTED 0 2020 HRS Unit 2 C cle 8 1.0 0.9 0.8 0.7 0.6 O

0.5 0.4 0.3 0.2 0.1 0.0 2000 2100 2200 2300 0000 0100 0200 0220 Dll-UTION TIME (HRS)

V-6

VI. ZER P

WER PHY I TE Tlirt Zero Power Physics Testing for Unit 2 Cycle 8 commenced at 0920 hours0.0106 days <br />0.256 hours <br />0.00152 weeks <br />3.5006e-4 months <br /> on November 8, 1990, and was completed at 0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br /> on November 9, 1990.

Zero Power Physics Testing was performed for the following reasons:

To determine the Moderator Temperature Coefficient (MTC) and verify the MTC Technical Specifications are satisfied, and 2.

To determine Control/Shutdown Bank Rod Worths and verify Shutdown Margin Technical Specifications are satisfied.

The Zero Power Physics Testing Program consisted of:

1.

ARO Isothermal Temperature Coefficient (ITC) Measurement and ARO MTC Calculation, and 2.

Control Rod Worth Measurement (Rod Swap) with ARO and CBD Inserted Critical Boron Concentration Measurements.

The Steam Dump Valves had to be taken out of service just prior to measurement of the ITC due to leak in the plant boiler. As a result, the Pressure Operated Relief Valves (PORVs) were used in performing the ITC Test.

The RCS heatup and cooldown rates established with the PORVs were very constant (linear) and provided excellent test data for analysis.

Alltesting was completed with no anomalies, and results were well within the given Acceptance Criteria.

The MTC was calculated to be +4.89 pcm/'F, less than the 70% RTP Technical Specification limit. However, the MTC was more positive than design so the measured value was normalized to Hot Full Power Conditions, which yielded a negative MTC. Therefore, no adjustments to rod withdrawal/boron limits were required to ensure an MTC of less than 5 pcm/'F at 707o RTP or a negative MTC at 100Fo RTP (Technical Specification 3.1.1.4).

The results of this test are presented in Table VI.1

(

Rod worth measurements were completed using the "Rod Swap" methodology.

The highest worth bank (CBD) was diluted into the core and the reactivity worth was directly measured from the reactivity computer strip chart.

See Figure VI.1 for the plots of CBD Integral and Differential Rod Worth vs Rod Position. Following CBD measurement, the remaining seven banks, from predicted )east to highest worth, were swapped with CBD and the previous Test Bank. The configuration of the rod banks after each swap would be:

Test Bank fully inserted, CBD inserted at a discrete position, and all other banks withdrawn. Based on the position of CBD, the inferred rod worth of the Test Bank was calculated.

The measured/inferred rod worths for each bank were all within the Acceptance Criteria.

The results of the measured/inferred rod worth, design rod worth, and Percent Error are presented in Table VI.2. Also, Table YI.3 shows the results of Boron Endpoint Measurements which were determined during this test. The results of this test were used to verify that excess Shutdown Margin (SDM) would exist for beginning and end of cycle conditions.

The beginning of cycle excess SDM was 1477.5 pcm, and the end of cycle excess SDM was 1377.5 pcm.

The results of the calculations satisfied Surveillance Requirement 4.1.1.1.1d of Technical Specification 3.1.1.1

~

VI-2

D.C. Cook Unit 2 Cycle 8 Zero Popover Physics Testing Results'ABLE VI.1 I

THERMALTE PERAT RE EFFI IENT Measured ITC Doppler Coefficient MTC=(ITC-DTC)

Design MTC

+2.31 pcm/'F

-2.58 pcm/'7

+4.89 pcm/'F

+4.12 pcm/'F TABLE VI.2 R

D W RTH MEA REMENT Bank CBD SBA CBA SBC SBD CBB CBC SBB TOTAL Measured Worth (pcm) 1016.2 380.2 318.5 466.5 471.2 716.4 736.1 897.4 5002.5 Predicted Worth (pcm) 1043 351 363 456 456 647 759 889 4964

% Error

~d.

00 Pred.

-2.6

+ 8.3

-12.3

+ 2.3

+ 3.3

+ 10.7

-3.0

+ 0.9

+0.8 TABLEVI.3 B

R N ENDP INT (from **2THP 6040 PER.352)

Bank ARO CBD in Measured (ppm) 1654.3 1538.0 Design (ppm) 1687 1566

'he design data has been obtained from Westinghouse's Nuclear Design Report.

VI-3

FIGURE VI.1 D.C. Cook Unit 2 Cycle 8 Differential and Integral Rod Worth of Control Bank D, HZP, BOC 1100 10

~

~

~ ~

~

~

1000 900 CL E

~

~

~

800 700 0

0K 0

C0 0

V O

600 500 400 300 200 100 0

0 20 40 80 0

100 120 140 160 180 200 220 228 Control Bank 0 (Steps Withdrawn)

~

~

Design

Measured

VII.1 P WER A EW I N TESTING Unit 2 Cycle 8 Power Ascension Testing commenced with entry into Mode 1 on November 9, 1990, and was completed on November 23, 1990, when Full Core Flux Map (FCFM) 208-06 was obtained (at - 99% RTP) and processed.

Figure Vll.l shows the progress of the Power Ascension Testing Program over time and consisted of the following:

1.

Intermediate Range Trip Setpoint Verification 2.

Core Power Distribution Measurements 3.

Incore/Excore Detector Cross Calibration During the Power Ascension to 30/o RTP, intermediate range detector data was collected.

The data divas used to ensure the calculated, installed High Level Trip (HLT) setpoints were set conservatively, less than 25% RTP.

The HLTs occurred at 23.2% and 24.3% IT Power for N35 and N36, respectively.

Also, calculations were performed and the results indicated that the setpoints would remain conservative over the entire cycle.

Flux maps were taken at various power level plateaus during the ascension.

FCFMs were taken at approximately 34%, 47%, 66%, 88%, 86%, and 99% RTP. The power distribution in each case was calculated using the DETECTOR Code (Version DETECT27) and the appropriate Engineering Data Set. Allof the flux maps had peaking factors well withinTechnical Specification Limits, and the power distribution over the entire core met all of the Acceptance Criteria.

The maximum incore quadrant power tilt ratio of all maps was less than 1.009 (0.9%). A summary of the peaking factors measured at various power levels is given in Table VII.1.

At approximately 47% RTP, FCFM 208-02 was obtained and used to calculate Incore/Excore Cross Calibration Data.

After I&E began the Nuclear Instrumentation System (NIS) calibration, it was discovered that the power range drawers could not be calibrated with the provided data.

Upon investigation, it was discovered that due to the very low leakage loading pattern, the expected output currents from the detectors were too small to complete the calibration.

Also, it was known that the Unit 1 Power Range Drawers were previously modified to handle this situation.

Since Unit 1 was shutdown for a refueling outage, it was decided to replace the Unit 2 drawers with the ones from Unit 1 (Unit 1 drawers were subsequently modified). As each drawer was replaced, the calibration was completed with no problems.

The 86% RTP FCFM was taken due to an indicated NIS Quadrant Power Tilt Ratio in excess of 2/o. The flux map confirmed that no incore tiltexisted (.-0.2%) so data was provided to I&Eto calibrate the power range drawers.

Upon completion of the calibration, the "instrument tilt" disappeared, as expected.

TABLE VII.1 FLUX MAP DATA MAP N

Power 4

A.O.

QPTR Fq (z)

APL 4 208-01 33.55 1.5831 1.0076 1.4599 1.6880 1.8680 4.0764 98.54 208-02

46. 65

3. 9144 1.0088

1. 4555

1. 6490 1.9047 4.0970 97.67 208-03

65. 64 1.7377 1.0067 1.4098 1.5924

1. 7643 3.0972 106.06 208-04 208-05 88.30 86.45 1.0506 2.2327 1.0048 1.0029 1.3958 1.5304 1.3884 1.5249 1.7172 1.7254 2.2791 106.28 2.3458 107.60 208-06 98.72 0.4110 1.0006 1.3849 1.4938 1.6800 2.0386 108.64 A.O.

= Axial Offset QPTR

= Maximum Quadrant Power Tilt Ratio H

Fz Fz

= Measured Total Enthalpy Rise Hot Channel Factor

= Technical Specification Enthalpy Rise Limit Fq"(z)

= Measured Penalized Heat Flux Fq (z)

= Technical Specification Penalized Heat Flux Limit APL

= Allowable Power level VII.1-2

FIGURE Vll.1 Unit 2 Cycle 8 Power Ascension VS. Time 110 100 90 80

~o 70 L0 0

0 50 40 20 10 g~

CA

'H

') I I

~r1 t

I I

C%

I Ol CS C7 IC)

I k

O K

I I-0 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 November 7 November 24, 1990 V11.1-3

VII.2 REA T R

LA T FL W 'VIEA REVfENT 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 parameters.

In addition, it was also the purpose of this test to recalibrate the reactor coolant flow elbow tap differential transmitters, as required, based on the computed reactor coolant flow rates and elbow tap differential pressure data.

The total coolant flowrate was determined at approximately 90% and 100% levels of reactor thermal power using the equations of Figure VII.2. Table VII.2shows that the total reactor coolant system flowrate was above the minimum RCS flowrate of 366,400 GPM as defined in Technical Specification 3.2.5.

TABLEVII.2 Com i ed Reactor oolant vstem Flowra e PM

~

Loop ¹ 90% RTP 93,860 98,260 93,280 98,330 100% RTP 93,600 97,570 92,780 98,040 Average Total Flow Rate 95,930 383,730 95,500 381,990

~ Flows are converted to GPM using RCS cold leg conditions.

VII.2-1

FIGURE VII.2 Reactor lant Flow De erminati n

h, RC Press TG Steam Press Reactor Coolan" Out S team Generator Steam Out hg RC Press TH FV Temp FW Press hHzO Reactor Coolant En Feedwater En

1. Feedwater Flowrate = M,= 359.1 x d'- x C x F. x v'y x b,H,O/(1 (d/D)-)

where:

d C

F, EHP D

throat diameter of venturi (inches) coefficient of discharge venturi thermal expansion factor specific weight of feedwater (Ib/ft')

differential pressure across venturi (inches) pipe diameter at inlet of pressure tap (inches)

2. Steam Generator Thermal Output = SGTO = Mx(h,h,)

where:

h,

= enthalpy of steam (BTU/lb) h,

= enthalpy of feedwater (BTU/lb)

3. Reactor Coolant Loop Flow = SGTO/(h(hc 0.315 BTU/lb))

where:

h= enthalpy of hot leg (BTU/Ib) hc

= enthalpy of cold leg (BTU/lb)

4. Total Reactor Coolant Loop Flow = sum of the loop 1 through 4 flows (lb/hr)

VII.2-2

VII.3 PLANT THERMALP WER ALIBRATI The purpose of this test was to determine reactor thermal power by measuring secondary system feedwater flow and steam parameters and to verify the accuracy of the following plant process computer outputs:

1. Reactor thermal power 2.

Feedwater flows

3. Feedwater temperatures

4. Nuclear power range instrumentation During the initial power ascension program the reactor thermal output was calculated at various power levels'y measuring secondary side parameters.

Data was collected at approximately 30%, 90%, and 100% RTP. The parameters that are measured for this calculation were feedwater flow, feedwater temperature, feedwater pressure, and steam pressure.

With the measured values, thermal output was be calculated.

The power determinations were made by measuring the feedwater parameters before and after the steam generators.

The amount of energy added by the steam generators was determined from these measurements.

The energy gained by the steam side of the steam generator was the equivalent energy given offby the reactor coolant system.

By knowing the heat transferred by each of the four steam generators, the total heat added to the secondary side was determined.

The total heat. added to the secondary side minus the heat added by RCP operation and the RCS system losses was the actual reactor power.

The power determination data at 90% and 100% power is shown in Table VII.3.

Alldata taken for the thermal power measurement were from instruments calibrated for this test.

The pressure measurements were made using 4-20 mA pressure transmitters.

Feedwater flows were measured at the local transmitter for each loop.

Feedwater temperatures were read using the installed thermocouples.

Before data taking commenced, steam generat'or blowdown was isolated and all plant parameters are allowed to stabilize.

The computer thermal power was monitored during the actual thermal power test. Upon test completion, a comparison was made between the plant process computer value and the actual measured value, and no adjustments were necessary.

The primary purpose of the power determination at -30% RTP was to adjust the NIS Power Range gains, if necessary, prior to the power escalation to 48%.

The comparison of the measured power and NIS data proved that no adjustment to the NIS was necessary.

VII.3-1

TABLE VII.3 THERMAL POWER CALIBRATION DATA Calculated Power Computer Power Loop Number 88.35 88.30 2

98.80 98.78 Feedwater Pressure (psig)

Feedwater Temperature

('F) 841.74 836.29 835.96 410. 8 410. 5 410. 7 844.27 410. 7 808.0 831.9 421.5 421.3 830.9 421.5 Steam Pressure (p>>g) 806.46 799.34 797.35 804.74 793.6 784.7 783.4 VII.3-2

Section VIII PLANT CHEMISTRY HISTORY

VIII. PLA T HEW[I TRY HI T RY The Unit 2 Chemistry cleanup efforts began June 29, 1990 during the reactor shutdown, with steam generator hideout analysis testing. Samples were taken at 50%

RTP, 25% RTP, and 500'F with no delay in the cooldown rate.

The cooldown was intentionally held for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> at 325'F and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 250'F for hideout sampling.

The hideout sampling program only directly affected the system cooldown since the RCS degassing was a concurrent event.

Following the unit shutdown, reactor coolant system (RCS) dose equivalent I-131 spiked to a maximum of 0.129 pCi/cc.

No problems in cleaning of the spike were noted.

All compensatory Auxiliary Building Vent samples were analyzed with no problems noted.

Following the reactor coolant system degassing, the system was oxygenated by the addition of 30% hydrogen peroxide to solubilize the Co-58.

RCS Co-58 activity spiked to a maximum of 0.81 pCi/cc following the addition of hydrogen peroxide.

The methodology of additions for this evolution was changed per the request of AEPSC. A large initial addition (four gallons) was made to the system instead of several small initial additions, as done in previous outages. A smaller addition (1.25 gallons) was made after the iodine spike, as in previous outages.

Cleanup was delayed several times due to RCS filter changeouts (with subsequent demineralizer isolation) due to high activities that were deposited on the filters. Approximately 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> after the initial Co-58 spike, a second unexplainable spike occurred.

Reactor coolant system Co-58 activity during the second spike peaked at 0.70 pCi/cc. The cleanup progressed for another 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after the second spike and was called complete at 0.095 pCi/cc. A "fresh" H-OH CVCS mixed bed (120 GPM) and the existing cation bed (75 GPM) were used for the cleanup effort. Approximately 112 Curies of Co-58 were removed from the system.

Based on the duration of'outage, outage management decided not to have sludge lancing of the steam generators and the dye checking of the main and feedpump condensers performed.

During the aborted startup in October 1990, the Unit had to be degassed to repair leaking valves in containment.

The Unit was restarted on November 8, 1990 and was at steady state power (-100% RTP) on November 23, 1990. At Full Power, the dose equivalent I-131 was 7.24 x10" pCi/cc,