ML20104B148

Forwards Results of Rt PTS Calculations for Identification & Location of Beltline Region matls,plant-specific Matl Properties & Reactor Vessel Neutron Fluence.Super Low Leakage Pattern Design Developed Since Initial Submittals

ML20104B148
Person / Time
Site:	Point Beach
Issue date:	09/04/1992
From:	Link B WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-92-102 VPNPD-92-299, NUDOCS 9209150010
Download: ML20104B148 (8)

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, September 4, 1992 l Document Control Desk tU'.S. NUCLEAR-REGULATORY COMMISSION Mail Station P1-137 Washington,,D.C; ~20555-Gentlemen:

DOCKETS'50-261.AEp 50-301' 11ESf0NSE TO 10 CFB 50.61

-. FRACTURE TOUGHNESS REOUIREMENTS FOR_fROJECTION_AG633ET

IMESSURIZED THERMAL SHOCK (PTS) EVENTS . .

' POINT' BEACH NUCISAR ~ PLANT

'10)CFRESO.61, " Fracture. Toughness-Requirements-for Protection-

-:Against -Pressurized Thermal Shock Events," requires licensees to submit" projected values of RTrr s for reactor vessel beltline

. materials.~ . Original' RTyn .submittale for our Point' Beach Nuclear.

Plant,.-Units l1_'andf 2,.-were-provided to the NRC.on January 20, 1986, zand Marche 14,'1986.f Since these original ~submittals, we have instituted: a- super, low , leakage loading pattern (L4P) core design with part-length hafnium absorbers in.the guide tubes of the periphera1Lassemblies, instituted a cavity dosimetry monitoring program, joined the'B&W Owners Group React'orLVessel Working Group,

'and refined our beltline material properties.

110 CFRD50.61 was . revised - in 1991 to - change :the procedure for cal'culating the amount.'of radiation embrittlement that a reactor 2

. vessel receives.- Thel revised rule requires each pressurized water reactor? licensee to submit _ projected values of RTrrs for Leach

'# reactor vessel beltline 1 material. This assersment must be submitted within.5: years of the effective date'of1the rule if no smaterialsfin the' beltline region are projected to exceed the-g J fscreening criteria =before the expiratic _of the. operating license.

7 The rule-further. states.that these- ubmittals'must-be_-updated X ~'

?whenever-there.'is a significant-change in the projected values of RTrn. -We provided RT rrs values lfor' Point Beach Nuclear Plant Unit 2

on October'15, 1991, with our Surveillance Capsule S Report. In.

ouriresponse to

Generic Letter 92-01, Reactor Vessel Structural-Integrity," dated' June 25,11992, we reported beltline materiali properties'and':-fluence values which have-been refined over the'past '

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=several years. .Due to these' refinements, we have recalculated RTrts

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$eptember 4, 1992 Page 2 values for our Point Beach Nuclear Plant Units 1 and 2. Although the projected values of RTrrs have not significantly changed, this letter is submitted to provide an update' s u m m a r y o f o u r R T rrs calculations.

Accordingly, plant specific RTvrs calculations have been performed as of August 1, 1992, and for the expiration date of the operating license for all materials in the beltline region of the Point Beach Nuclear Plant Units 1 and 2 reactor vessels. These calculations indicate that the RTers values for all beltline materials in the reactor vessels will not exceed the screening criteria defined in 10 CFR 50. 61(b) (2) through the expiration date of the current operating licenses.

The attachment to this letter provide.3 t h e r e s u l t s o f t h e R T rrs calculations for Point Beach, as well as the bases for the fluence and material properties used in the RTyn calculations. The fluence projections and the material properties used in the calculations are the same as reported in our response to Generic-Letter 92-01,

" Reactor Vessel Structural Integrity," dated June 25, 1992.

We believe that, based on comparison of our calculated RTru values in Tables 1 and 2 of the attachment to the PTS screening criteria contained in 10 CFR 50.61, Point Beach Nuclear Plant Units 1 and 2 are projected to conform to 10 CFR 50.61 for the duration of the current operating licenses.

If you have any questions or require additional information regarding this report, please contact us. _

sincerely,

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-(, // cr

.w Bob Link Vice President Nuclear Power GLM/jg Attachment cc' NRC Regional Administrator, Region III NRC Resident Inspector i

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M7ACHMENT I BTp7p CALQULATIONS AND BhBES FOR POINT BEACE NUCLEAR PLANT, UNITS 1 AND 2 IDENTIFICATION AND LOCATION OF BELTLINL' REGION MATERIALS Figures 1 and 2 identify and indicate the location of all beltline region materials for the Point Beach Nuclear Plant Unit 1 and 2 reactor vessels, respectively. The beltline region is defined in'10 CFR 50 Appendix G to be the " region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."

Since our original PTS submittal dated January 20, 1980, we have decided to include two additional materials in each reactor vessel as part of the beltline region. These two materials are the nozzle belt (NB) forgings and the nozzle belt forging to intermediate shell welds. Our fluence monitoring program indicates that these materials will experience accumulated fluence greater than 1 x 10 17 n/cm 2 , which is the point where Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials" indicates the materials start to experience radiation damage.

PLANT SPECLFIC MATERIAL PBOPERTIES The pertinent chemical and material properties of the beltline region materials for Point Beach Unit 1 and 2 are provided in Tables 1 and.2. These material properties have been refined over the past several years based on our participation in the B&W Owners Group Reactor Vessel Working Group. All of these material properties have been previously reported in BAW-2166, "B&W Owners Group Response to Generic Letter 92-01," which was forwarded to the NRC by P&W Nuclear Service Company on June 17, 1992.

Additionally,-we referenced BAW-2166 in our docketed response to G9neric Letter 92-01 dated June 25, 1992.

REACTOR VESSEL NEUTRON FLUENCE In 1989, a reactor cavity measurement program was instituted at both Point Beach Units 1 and 2 to provide continuous monitoring of-the neutron fluence of the beltline region of the reactor pressure vessel. When used in conjunction with dosimetry from previously withdrawn internal surveillance capsules and with the results of neutron transport calculations, the reactor vessel cavity neutron dosimetry provides neutron exposure data for the

, reactor pressure vessel and the embrittlement gradients through l the vessel wall.

l Addi~tionally, in 1989, we implemented a super low leakage loading pattern (L4P) core design and introduced part length hafnium absorbers in the guide tubes of the peripheral assemblies. This flux reduction approach was designed to. reduce the maximum neutron exposure on the limiting reactor vessel beltline materials. In our fluence projections for Point Beach Units 1 and 2, we assume this core design will be maintained through the expiration date of our operating license.

As of August 1, 1392, Point Beach Units 1 and 2 had been operated for a total of 16.4 and 16.3 Effective Full Power Years (EFPY),

respectively. Assuming an 80% cumulative capacity factor, the reactor vessel neutron fluence can be }rojected for the end of the operating license. The fluence projections for August 1, 1992 and the end of our licensed life are provided in Tables 1 and 2.

BIPTS VALUES FOR POINT BEACH NUCLEAR PLANT. UNITS _1 AND 2 RT p73 calculations have been performed according to the

! requirements of 10 CFR 50.61, " Fracture Toughness Requirements l for Protection Against Pressurized Thermal Shock Events," and the results are provided in Tables 1 and 2. It is concluded that the Point Beach reactor vessel beltline materials wi]l not exceed the applicable screening criteria through the expiration of the current operating licenses.

9 .1 1

I TABLE 1: POINT BEACH UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES .

Beltline Initial Copper 4 N.ckel- Fluence Fluence Margin RTm RTm Screening

. Material- RTw Content content 8/1/92- 32 EFPY- 8/1/92 32 EFPY Criteria

! NB Forging '+50 0.15 0. 8 2 .' '1.75E+18 2.95E+18 +34. 146 F 161 F' 270 F

-122P237 '(3)' (3) (3) (3) (1)

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Intermediate' +1 0.20 0.056 1.59E+19 2.68E+19 +48 148 F 160 F 270'F 4 Shell.(IS) -(3) (3). (3) (2) (2). (1) 4 A9811-1 1 t 270"F Lower Shell +1 0.12 0.065 1.55E+19 2.33E+1%. +4F 111 F 117*F

, (LS)'C1423-1 (3) (3) .(3) (2)  :(2) 1 (1)

l. NB.to IS Weld 0 =0.20 0.55 1.704+18 2 95E+18 +66 1480 F 167 F 300 F 1 SA-1426 (1) '(3)- (3) (3) (3) (1) i '

IS to LS Weld  :+10 0.26 0.60 1.55E+19 2.33E+19 '+56 268 F 287 F 300 F SA-1101 (3) (3) (3) (2) (2) (1)

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IS Longitudinal 0 0.17. - .'O.52 9.87E+18 1.71E+19 +66 204 F 225 F 270'F lg .

i SA-812 (ID 27%) (1) (3) (3) (2) (2)' (1) i i h I 1 XS Longitudinal 0 0.19 0.63 ---! - - - - ---- --- -- '

l SA-775 (OD 73%) '(1) (3) (3) c  !

[ LS Lcngitudinal 0 0.25 0.54 9.71E+18 1.56E+19 +66 232 F 254"F 270 F l i SA-847 (1) '(3) -(3) '(2) -(2) (1) t l (1)- 10 CFR 50.61, Fracture hiughness Requirements for Protection Against Pressurized Thermal Shock

-Events. i (2) UCAP-12794, Rev. 1,' Reactor Cavity Neutron Measurs ment Program for Wisconsin Electric Power  !'

- Company Point Beach Unit-1, March 1992.

- (3) BAW-2166, B&W Owncrs Group Response to Generic Letter 92-01, June 1992. ,

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' TABLE 2: POINT' BEACH UNIT 2 REACTOR VESSEL' BELTLINE REGION MATERIAL PROPERTIES ,

Beltline: Initial. Copper Nickel Fluence' Fluence Margin RTm. RTm. Screening h Material :RT m Content Content 8/1/92 32 EFPY 8/1/92 32 EFPY Criteria

-NB Forging '+40' O.15 0.73 2.06E+18 3.50E+18 +34 139 F 154*F 270 F 123V352 (3) -(3) (3) '(3)- (3) (1) l Intermediate- + 4 0,- 0.09- -0.70: ' 1'.~ 72E+19 2.92E+ +34 141 F 149 F 270 F ShellL(IS)i (3). '(3)- '(3)

(2) (2) (1) I

'123V500 ,

Lower Shell. +40 0.05' 0.72  ! ;69E+19 1 2.66E+19 +34 lib F 113 F 270 6 F

{(LS)'122W195 (3)' -(3) .(3) '(2) (2) (1)

NB to IS Weld' --5 6 0.27'  : 0.90' 2.06E+18 3.50E+18 +66- '

144 F 175 F 300 F' CE. Weld .(1) (3) (3) (3) (3) (1)

IS to LS Weldi .O O.24 0.60 1.67E419 2.56E+19 +66 264 F 283 F- 300 F SA-1484 -(1) (3) (3) (2) (2) (1)

(1) 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Theanal Shock Events.

(2) WCAP-12795,'Rev. 1,VReactor Cavity Neutron Measurement Program-for Wisconsin Electric Power Company. Point, Beach Unit 2, October 1991.

(3) BAW-2166, B&W Owners Group Response to Generic Letter 92-01,-June 1992.

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IDENTIFICATION AND LOCATION DF BELTLINE REG' ON MATERIAL FOR THE POINT BEACH UNIT No I REACTOR ESSEL CIRCVMrEntNTI AL $ TAMS VERTICAL SEAMS I

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