ML17263B085

Summary of 950516 Meeting W/Util in Rockville,Md to Discuss PTS Assessment of Ginna Reactor Vessel Documented in SECY-95-119.List of Attendees,Meeting Agenda,Nrc & Rg&E Discussion Matl Encl

ML17263B085
Person / Time
Site:	Ginna
Issue date:	06/05/1995
From:	Andrea Johnson Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9506220184
Download: ML17263B085 (25)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 5,

1995 LICENSEE:

Rochester Gas and Electric Cor por ation FACILITY:

SUBJECT:

Ginna Nuclear Power Plant

SUMMARY

OF MEETING WITH ROCHESTER GAS AND ELECTRIC CORPORATION ON MAY 16, 1995 PRESSURIZED THERMAL SHOCK ASSESSMENT OF REACTOR VESSEL The NRC staff met with Rochester Gas and Electric Corporation (RG&E), in a meeting on May 16, 1995, at NRC Headquarters, to discuss the staff's pressurized thermal shock (PTS) assessment of the Ginna reactor vessel that was documented in Attachment 2

(May 5, 1995 memorandum from Strosnider to Thadani) to SECY-95-119.

The Ginna analysis was performed by the staff as part of a generic assessment of all pressurized-water reactors.

The generic assessment was performed to determine whether large variability of chemistry-(copper and nickel) would result in an immediate safety concern for any

'eactor vessel.

The staff's generic assessment indicated that there was not an immediate safety concern for any reactor vessel.

The staff's generic assessment indicated that the Ginna reactor vessel would be projected to reach the PTS screening criteria in about 7 effective full power years from January 1,

1995, prior to the expiration of its license in 2009.

RG&E was concerned that the projected shorted life of the reactor vessel would reduce the economic benefit of the steam generator replacement program.

The replacement steam generators are scheduled for installation during the 1996 refueling outage.

The NRC staff's concern about the variability in chemistry resulted from data provided by the licensee for the Palisades plant.

As part of its recent PTS assessment, the licensee for Palisades provided data that indicated a large variability in chemistry of the limiting welds'.

The greater the amounts of copper and nickel, the greater the amount of embrittlement and the lower the material's fracture resistance to PTS events.

The staff's generic assessment was performed to determine the impact of large variability of chemistry on other reactor vessels.

The NRC staff's analysis used a generic mean value for chemistry and large margin terms to account for potential variability of chemistry.

The mean values of chemistry and margin terms were determined from generic data.

The staff's assessment of the Ginna reactor vessel did not include the Ginna surveillance data.

RG&E for Ginna presented a

PTS assessment, which included the test results from their surveillance program.

Their assessment indicated that the Ginna reactor vessel would not exceed the PTS screening criteria at the expiration of its license (EOL).

9506220f84 950605 PDR ADQCK 05000244 P

PDR

I

June 5,

1995 The NRC staff indicated that it would review there'TS assessment if it was submitted for staff review.

The staff indicated that the assessment should consider:

(a) the use of best-estimate chemistries less than the generic value and (b) the use of smaller margin terms than the values determined. from generic data.

The NRC staff encouraged RGEE to pursue: (I) vendor documentation supporting reduced best-estimate values of chemical composition, (2) consideration of all data for estimating the initial reference temperature, and (3) consideration of a plant-specific fluence map in the PTS evaluation.

All of the above could potentially have the effect of reducing the amount of embrittlement and pushing the PTS screening criteria date towards or beyond EOL if technically justified.

RGKE indicated that they were going to pursue the additional documentation and would submit a revised PTS assessment for staff review.

Following the technical portion of the meeting, RGSE met with the Associate Director for Projects to discuss future communications.

A copy of a list of meeting attendees is included in Enclosure I.

Enclosure 2

thru 4 are copies of the meeting agenda and discussion material.

Sincerely, Docket No. 50-244 Allen R. Johns n, Project Manager Pro] t Dire orate I-I Division Reactor Projects I/II Office of Nuclear Reactor Regulation

Enclosures:

1. List of Attendees 2,. Meeting Agenda 3:

NRC Discussion Material 4.

RGSE Discussion Material cc w/encls:

See next page

Rochester Gas and Electric Corporation R.

E. Ginna Nuclear Power Plant CC:

Thomas A. Moslak, Senior Resident Inspector R.E. Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontar io, NY 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Hs.

Donna Ross New York State

Energy, Research, and Development Authority 2 Empire State Plaza Suite 1901

Albany, NY 12223-1253 Charlie Donaldson, Esq.

Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Nicholas S. Reynolds Winston I Strawn 1400 L St.

N.W.

Washington, DC 20005-3502 Ms. Thelma Wideman

Director, Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31

Lyons, NY 14489 Ms. Hary Louise Heisenzahl Administrator, Monroe County Office of Emergency Preparedness 111 West Fal'l Road, Room ll Rochester, NY 14620 Dr. Robert C. Hecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649

DISTRIBUTION w/encl s.

Docket File 50-244 PUBLIC PDI-1 Reading AJohnson JLinville RI/DRP WKane RI/ORA WLazarus RI/DRP JWiggins RI/DRS WHehl RI/DRSS DISTRIBUTION w/encl WRussell/FHiraglia RZimmerman AThadani SVarga JZwolinski THarsh SLittle BSheron DE GLainas DE JStrosnider EHCB OCG (f/info only)

EJordan, D/AEOD ACRS (4)

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LIST OF MEETING ATTENDEES NRC MEETING WITH ROCHESTER GAS AND ELECTRIC CORPORATION R.E.

GINNA NUCLEAR POWER PLANT PRESSURIZED THERMAL SHOCK PTS ASSESSMENT OF REACTOR VESSEL NNE TITLE Allen Johnson Tad Marsh Michael Vassilaros Carol n Fair banks John Tsao Glenn Dentel Andrea Wilford Ed Hackett Barr Elliott Jack Strosnider Brian Sheron Robert C. Mecred Geor e Wrobel Thomas A. Marlow Ron Ja uin Matt OeVan ORGANIZATION USNRC DRPE PDI-1 USNRC ORPE PDI-1 USNRC RES DET USNRC DE EMCB USNRC OE EMCB USNRC DE EMCB USNRC DE EMCB USNRC DE EMCB USNRC DE EMCB USNRC DE EMCB USNRC DE RG&E VP Nuclear 0 er.

RGSE M r. Safet

& Lic.

RG&E M r. Nuclear En r.

RG&E Safet

& Lic.

Babcock

& Wilcox TELEPHONE 301 415-1497 301 415-1416 301 415-6000 301 415-2712 301 415-2702 301 415-1321 301 415-2735 301 415-2751 301 415-2709 301 415-2796 301 415-2722 716 724-8069 716 724-8070 716 771-4676 716 724-8072 804 832-3160 Enclosure 1

AGENDA OC GAS ANO E CTR C

A 6

995 S

R Z

T SMOC T

S SSM T

AC 0 SSE 1.

NRC concern with industry data in general 2.

NRC Methodology resulting in concern for Ginna 3.

- Concerns (if any) regarding Ginna-specific data and results 4.

RGKE provide discussion on how EOL RT>>, value is determined 5.

Future actions to resolve concerns Enclosure 2

I CONCERNS REGARDING CHEMISTRY VARjULSILITYI I

S

~ V

~

I I

MARGINTERM" 2 a'REATER-KHANASSUMI<&INDEVIIXDPMENTOF PIS RULE PALISADES EVALUATIONINDICATESMARGINTI94M IS ADEQUATE "

CONSIDERING O'HillCONSERVATISMS IN THE RULE (E.G. PLANT SPECIFIC FLUENCE DISTRIBUTIEPPQ I

~

~

~'

BEST ESTMATE CHEMISTRY A

COPPER CONTINUA'PPEAR NOT TO BE,HEAT DEPENDENT I

BEST XSTP4tkTE COPPER CAN CHANGE SIGNIFIQAN'FLYWITH

,;,.<).",";> MORE DATA-)..<.

I SURVXZLIANCEDATA CRRRSRRV VACIRCAICIIIAIRGUSING SURVIRIAAIICRRASA NET% TO BE ADJUSTED WHEN lT IS SIGNIFICAN'XLYDIFH<RENT

%%I'ANTHE CHIMISTRYFACTOR CALCULATEDUSING BEST ESTIMATECOPPER ANDNICKIX,COXTI~TS INRPV MATImlALS Enclosure 3

I

Flow Chart for Generic RPV Assessment Considering Chemistry Variability Simulate IRT~

N~~IRTndt'IRTnlt~

Simulate Cu N(p~, n~)

gI I~

Simulate Ni

. N(pi. ~<<)

Simulate F

N(p, ~,)

00 I 1,N Calculate Mean MT~(F,Cu,Ni)

ART)~q IRTg)~ + MT)~q Calculate n of Oistrihution of Simulated ART~~ Values Calculate Fluence (p

, p<<)

that gives peggy~

+ LBT~~ + 2a Screening Criteria Based on R.G. 1.99, Rev.

2 Tables for Mean AT~,

PLANT SPECIFIC EVALUATION GINNAWFLDS.

1 V

~ I

~

~

1

"~"

Mgggggp MEANINEIALRTpgyg

'"'1) it

~~ = STANDARDDEVIATIONABOUTTHE MIBLN= 20'F

2. ~ = MIBLNCOPPER CON'H<22T OF LINDE80 WELDS = 0287 I

a~ = STANDARD'DEVIATIONOg COPPER..=,

0.062..

3.

"~

MM =. MEAXNICR1<X COWHAND%ROM GINNAGL-92-Of

'UBMITFAR

= 0.54

= '.

(3)

~ ~

~ ~ v

's<

I

~

- ~

4.

~~~~ = SPECK'11<3) VALUEOF FLUENCE

+maraca = 20% OF %mmmca (1) SEE TABLEL VALUEOF IRT~ POR B&WWI<XBS (2) SEE TABLEIL VALUESOF COPPER FOR B&WWIK;DS (3) SEE TABLEm: VALUES OF NICKELFOR B&WWFLBS

WIKDID

,RT~

SA 1036 SA 1101 WF 112 SA 1526 WF 67, SA 1585 WF 193 WF 182-1 WF 182-1 PQ 3170 PQ 2923 PQ 3443 PQ 3116 PQ 3117 PQ 3229 NBD RVSP RVSP RVSP RVSP

~

~

~

~ ~

~

33

-13

-20

-3

-7 20 15

-15 10 20 10 10

. -10 WF 278 WF 292 10 WF 314 WF 351

10

'F 696

'F 336 SA 2050

~ WF 353 WF 645

'WF 610 10 10

'10 RVSP = REACTOR VESSEL SURVMLI~CEPROGRAM NBD = NOZZLE BREST DROPOUIS 1UBKRKNCE'AW1803r REV 1

'c P'P 1

COPPER VAAORS OSEO TO DETEOEEIE SOE STANDARDDEVIATIONIOR THE INC1UMSE IN REFER%ACE TEMPERATURE ANDTHE GENERIC MhLN VALUEOF COPPER WF 193 WF 25 WF 25 WF 25 SA 1526 WF 70 WF 209-1 WF 209-1 WF 209-1 SA 1101 SA 1769 SA 1036 SA 1135 SA 1585 WF 112 WF 193 WF 67

~

SA 148]i WF 182-1 FILLER WELD HEAT NUMBER 406L44 299L44 299L44 299IA4 72105 72105 72105 72105 71249 71249 61782 61782 72445 406L44 406L44 821T44 SOURCE OF WELDMENT RVSP RANCHO SECO RVSP TMI-1 RVSP OCC 2 RVSP CR-3 RVSP OCC-3 RVSP WEST.

RVSP OCC 1 RVSP ANO-1 RVSP TMI-2 lo OF COPPER

.28

.33

.35

.35

.37

.42

.36

.36

.3Q

.18

.29

.20

.27

.21

.32

.28

.22

.26

.28 WF 182-1 SA 1263 WF 193 821T44 71249 406L44 RVSP DAVIS BAS

.21 RVSP PONT BH1

.22 RVSP PONT BH2

.25 SA 1036 61782 RVSP GINNA

.23

SA 1101 SA 1094 WF 209-1 WF 209-1'A 1526 WF 23$

"'F 25 WF 67 SA 1585 WF 70 WF 112 SA 1135 71249 71249 72105

.72105 72442 72445 72105 61782 RVSP TP 3 RVSP TP 4 RVSP ZION 1 RVSP ZION 2 RVSP SURRY RVSP KORI 1 BWOG BWOG BWOG NBD MDLAND BWOG BWOG

.21

.30

.35

.30

.35

.27

.35

.22

';2i'32

.27 WF70...,....

72105.

MDLANDBELT

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~ ~

~ ~,

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~

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lb ~

rr;

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NIORRL VALOIIIOSEO 1O OEOEOOM 1NE SVANDARODEVIAIIONMR VNE INCREASE INRRXIXUUVCETKVlPERATUREDATAFROM WRY WIRE

, HEATNUMBER 72105 SOURCE MIDLAND "

'ELTLINE"'

% OF NICKEL

.58'56

.57

.57

.57

~

~

.57

58

.58

.58

.61

.63

.57

.S6

.57

.58

.56'57

.56 SOURCE MIDLAND

BELTLINE, WF 70 NBD

% OF NICKEL

.57

.57

.55

.57

.59

.58

.57

.56

.56

.56

.56

.S6

.59

.59

.59

.59

.59

.59

.59

.58

.58

.58

.56

+I

SOURCE WF 70 NBD

% OF NICKEL

.61

.61

.61 F A1I J%

SOURCE OCONEE 3 RVSP i ~ i0

~

~ ~

~

~

~

IVI~i I

% OF NICKEL

.58

.58

.59'-

OCONEE 2 RVSP

.59

.58

.58

.58

.58

.57

.59

.61

~

b% 9 f

ygC! ~

E I 0

~

~

0

.62 OCONEE 3 RVSP

.59

.59

.59

.59

.58

.58

.58

.58

.58

~

~

~

~ ~

(

4

~

ZION 1 RVSP ZION 2 RVSP

.58

.59

.58

~ 58 is 'su~

.58

.58

.57

.58

.57

.58

.58

.58

.57

.55

.57

.59

~ ae ~

~"

SOURCE

% OF NICKEL SOURCE

% OF NICKEL ZION 2 RVSP

.56

~

~ 0

.59

~

~

~ P

~

~ '

~ 0

~

~ i4 t% ~ ~

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Weld Wire Heat 61782 A.

As Generic lRT Values and Table Values for CF RTpTs = "5 + 66 + ( 174.6 ) ( 1-338 )

295' B.

As Measured 1RTValues and Table Values for CF RTpTs = l9 5+ 67+ ( 174.6 ) ( 1 338 )

281'

Ginna Surveillance A.

Using Generic Values and Table Values for CF RTpTs 5+66+ ( 161A) ( 1 338 )

= 277'F B.

Using Measured Values of IRT M 8 Chemistry Factor Determined Using 8u v..'eillance Results RTpys 19 6 + 46 4 +(le)(g98 up<'F

a Qh llJ I

1 Ph 2

June 5,

1995 1

I The NRC staff -indicated that-'it'-would re'view,there PTS assessment if it was submitted for staff. rev'iew;'he staff.-indicated,-,that the assessment should consider:

(a) 'the use,'of best-estimate chemistries less than the generic value and (b) the use of, smaller margin terms, than the values determined from generic data.

lh I

The NRC staff encouraged RG&E to pursue:

(1), vendor documentation supporting reduced best-estimate values of chemical-composition, (2) consideration of all data for estimating the initial reference temperature, and (3) consideration of a plant-specific, fluence map..in the PTS evaluation.

All of the above could potentially have the effect of.'reducing the amount of embrittlement and pushing the PTS screening criteria date towards or beyond EOL if technically justified.

RG&E indicated that they were going to pursue the additional documentation and would submit a revised PTS assessment for staff review.

Following the technical portion of the meeting, RG&E met with the Associate Director for Projects to discuss future communications.

A copy of a list of meeting attendees is included in Enclosure l.

Enclosure 2

thru 4 are copies of the meeting agenda and discussion material.

Sincerely, Original signed by:

Docket No. 50-244 Allen R. Johnson, Project Manager Project Directorate I-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation

Enclosures:

1. List of Attendees

2. Meeting Agenda 3.

NRC Discussion Material 4.

RG&E Discussion Material cc w/encls:

See next page DOCUMENT NAME:

G: iGINNA516.MTS To receive a copy of th4 document, IruQcate In the born C'

Copy without attachment/encheure E

~ Copy whh ettechmentlencheure "N a No copy OFFICE PDI-1: L PDI-1:PH'DI-1:D NAME SLittle DATE Q /

/95

'1 h AJohnson.

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