ML17261A468

Forwards Info Re Installation of 16 Westinghouse Optimized Fuel Assemblies Containing Integral Fuel Burnable Absorber Rods in Plant Cycle 17 Startup Loading Pattern for 1987,per Tech Spec 6.9.1.1

ML17261A468
Person / Time
Site:	Ginna
Issue date:	05/06/1987
From:	Kober R ROCHESTER GAS & ELECTRIC CORP.
To:	Lear G NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
NUDOCS 8705130288
Download: ML17261A468 (6)

Text

REGULATORY ORl'lATION DISTRIBUTION SYS (R IDS)

ACCESSION NBR: 8705130288 DOC. DATE:: 87/05/06 NOTAR I ZED: NO DOCKET FACIL: 50-244 Robert Emmet Ginna Nuclear Plant> U'nit 1. Rochester G 05000244 AUTH. NAME . 'UTHOR AFFILIATION ROBER>R. M. Rochester-Gas Zc Electric Corp.

RECIP. NAME..-- RECIPIENT .AFFILI'ATI'ON LEAR> G. ' ~ Of fice o'f 'Nuclear 'Reactor Regulation..Direc tor (Pos't 87041'1

SUBJECT:

Forwards info re installation, of 16 Westinghouse 'op timized fuel assemblies containing integral fuel burnable absorber rods in plant Cycle 17 loading pattern for 1987> per Tech Spec 6. 9. 1. 1.

DISTRIBUTION CODE: -IE26D COPIES RECEIVED: LTR ENCL SIZE:

TITLE: Startup Report/Refueling Report <per Tech Specs)

NOTES: License Exp date in accordance uith 10CFR2> 2. 109<9/19/72). 05000244 I I REC P ENT COPIES RECIPIENT COPIES ID CODE/MANE LTTR ENCL ID CODE/NANE LTTR ENCL PD1-3 LA 1 0 'D1-3 PD 1 STAHLE> C 2 2 INTERNAL: DV h! RR/PNAS/ Il RB 0 RES SPEIS> T RGN1 FILE 01 RGN2/DRSS/EPRPB EXTERNAL: LPDR 1 1 NRC PDR NSIC 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTlR 13 ENCL 12

May 6, 1987 U.S. Nuclear Regulatory Commission Document Control Desk Att: Mr. George Lear PWR Project Directorate No. 1 Washington, DC 20555

Subject:

Startup Report

Dear Mr. Lear:

The enclosed information is submitted in accordance with Technical Specification 6.9.1.1. "A summary report of plant startup and power escalation testing shall be submitted following the installation of fuel that has a different design." This is in regards to the installation of sixteen Westinghouse Optimized Fuel Assemblies which contain Zntegral Fuel Burnable Absorber (ZFBA) rods into the Ginna Cycle 17 loading pattern for 1987.

Very truly yours, Roger W. Kober RWK/lms Enc.

xco U.S. Nuclear Regulatory Commission Region Z 631 Park Avenue King of Prussia, PA 19406 T. Pollich Ginna Resident Znspector

~4 8705130288 870506 a~8 ADOCN 05000244 PDR

April 28, 1987 I

Startup Physics Testing Program Cycle 17 Startup %basics Test~ Program was conctuchK'during the period'ram.

M2Q ch 9 I 1987 'o March 20 I 1987 B18 stated dates span fmm initial cri.tical ity to the attainment of the 100% flux map. She results of the physics testing program, provided belav, shaws that all measured data was within the baunds of the acceptance criteria. Ghe Ginna Station ET-34 series of procedures was used. to perform the lear pcarer physics testing program and was used in conjunction with the S-'15 series for flux mapping.

The physics testing program consisted of the following parameters:

1. All Rods Out (ARO) critical boran concentration

2. Isothermal Pmpmwture coef ficient (ITC) 3 Control Rad ~rths for Banks IIDII IICIIi and IIBlt

4. Boron 1~points

5. Core symmetry ani pawer distributian measurements Initial criticality was paramete~ as listed in el 1 thru 4 were then measured. ~

achieved an 3/9/87 at 1900 haum. Ghe Hat Zero

'ul*ht ~

flux symmetry mapping and factors was campleted an March 20, 1987.

The falling data suamarizes the results of the test~ ~xgram.

Critical

a. Initial Criticali and Nuclear Heat Detezminati: Initial criticality with all control rods fully withdrawn was acldpred at a boron concentration of 1350 pgm.

Nuclear heat was c9merved at 3 x 10 amps on the reactivity ceo~ter.

d'l ft It was conducted well below this point of adding nuclear heat.

b. Reactivi Chec3Kut: ~ reactor was placed on varying periads by witMrawing control rods to provide a camparisan of indicated reactivity with that derived from period measurements., The following table mnnmarizes the cczaputer checkout:

Control Bank M'easured Reactivity (KK) 0 Difference D Steps Reactor I gM~P x 100 NitMrawn Peri Seconds Measured Predicted P p 6 steps 407.9 16.05 16.0 + .314 10 steps 336.3 19.17 19.10 + .374 20 steps 324.8 19.79 19.6 + .974

c. Isoth~ Coefficient ITC:.,Heatup and cooldown rates of approx-imately 10 F/hr. were established to determine the Isothermal Tenperatuxe .

was ccnqmred to a graph of XXC and HIC an% a calculated HIC was achieved. The follawing azmarizes the results: ".

Measured ITC [ Calculated MZC(. Acceptance J Predicted HZC(

Criteria [Difference PP ~p

+ .53 + 3 '3 + 2.0 + 1.33

d. Control Hcd Worth: Contxol Rad worth was measured by adjusting xod position to compensate for dilution. 'Ibe follawing table suamarizes the intecpal xod worth data:

Difference (P) Acceptance Criteria

~PP x 100 D 904 858 5.094 C 1390 1430.5 + 2.914 B 827 718 -13. 18%'

SUM 3121 3006.5- 3.74 inS~

Criteria:

worth be within

~ acceptance 15%

criteria is that the measured individual bank of the gmdicted values and that the total worth of three bardm be within 134 of the predicted value. If the criterian an individual bank worth is not met, an evaluation will be performed to determine the cause and any potential impacts. If 'tl the criterion in total worth of the three banks is not met, additional l ~tt d will be compared to the value assumed in the Safety Analysis.

As sumaarized above, all contxol banks measured met the + 15% aca~nce criteria the total of the three banks measured met the overall 13Caca~nce criteria.

and rods ttt tt:

to the withdrawn positian

~' tt and measuring the bcoxn concerrt~tion.

tie %he follcaring table summarizes the boxcn endpoint data.

I I f Difference)

Confi tion Predicted P Measured -P Acce D

ARD Inserted 1368 ppm 1280 ppm 1350 ~

1265 ppm

-18 pgn

-15 ppm

+

+

75 75 ppn pgn 1122.4 pgn[ -22.6 ppn(

~B Inserted Inserted 1066 ~

1145 pgn 1051.4 gm( -14.6 pgn)

+

+

75 75 ppm pgm

Acce Criteria: %he.average<critical boron concentration for a given 'canfiguration will be compared to, the predicted cancentzation for that configuratian.;, The'cceptance data will be.reviewed<and'the fuel supplier will"be'asked to;zeview his praBctions. '.";;:

Concurrent with 'thi's. review .the remainder:.of the zero pawer physics 'testing will be.

the cause of not meeting the acceptance criterion a~ance criteria.

still cannot If after the above actians, the be met an evaluatian will be performed an the effect of this difference an parameters used in the accident analysis. If the accident analysis is unaffected by this difference, the core will be allowed to go above 5%

rated power+

f. Flux and Power Distribution: %he flux symmetzy fluxmap was performeD at 24.4% rated power. All locatians were evaluated and met the acceptance criteria as listed below. Ghe fluxmap was in good agreaaent with predictions anD verified the core was properly loaded. All Nuclear Hot Ct~nel factors were we11 within the baunds of Plant Technical Specifications.

During subsa~ power escalatian, fluxmaps were generated at 48.2%, 65.8% with Incoze/Excore calibration perfozmed at 89.5% pcwer. She 100% map was then taken an 3/20/87, with all Hot Channel Factors well within the bounds of Plant Technical Specifications.

Criteria: 'Ibe acceptance criterion for a flux map is that the plant Technical Specificatian an pea3dap factors be met. As an aid in evaluating the power distributian maps, the diffmmxes between measured and predicted assembly power for assemblies with relative power <1.0 be less than 15% and the diffezmxe for assemblies with relative pawer >1.0 be less than 10%. If these differences are exceeded, an evaluation shall be perfozmed to ensure that the peaking factor Technical Specificatians will be met. If such an evaluatian inDicates that the pea3cing factor limits will be met, no further actian is necessary.