ML12053A333

Areva Document No. 47-9048125-002, Update of Irradiation Embrittlement in BAW-10008 Part 1 Rev 1, Non-Proprietary

ML12053A333
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Site:	Oconee
Issue date:	10/27/2010
From:	Xu H AREVA NP
To:	Office of Nuclear Reactor Regulation
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BAW-10008 47-9048125-002
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47-9048125-002 Page 1 of 25 Update ýof Irradiation Embrittlement in BAw-1 0008 Part 1 Rev. I AREVA NP Inc. Do(cument:No. 47-9048125-002 Prepared for: Electric Power Research 1nstitute,-(EPRI)

/<z 7-

-77 0 /c, Prepared Reviewed by:

4j/

(6. rfltch:

Approved and released by:

ý:M.jAevýn 10 a? /0 Prepared by AREVA NP Inc.

An AREVA and SI.mensqs company 3315 Old Fo6est Ro ad P.O. Box 10935 Lynchburg, Virginia 24506-0935 A

• AREVA

47-904 8125-002.

Page 2:of 25 NOTE:

This document is the non-proprietary version of the AREVA NP Inc. proprietary document 51-903,8244-002. This document is identical to 51-9038244-002 except the proprietary information have been deleted as marked by"(

]"

I Purpose This document is the :deliverable for the Subtask 2.7b"Update BAW-1 0008 Part 1" of AREVA.-

NP Inc. Job 4160812, Revision 3[1[1, "PWR Internals:Components 'Functionality Analysis for :the:

B&W Design" for the. Electric Power Research Institute:(EPRI).

2 Background and Scope.

2.1 BA W-2248ASection 4.5.21 NRC FSER Section 3.-4.29 Subtask 2.7b is'to address part of the commitment made in BAW-2248AR] during the Babcock&

Wilcox Owners Group (B&WOG) Generic License Renewal Program (GLRP) in the:1990s.

BAW-2248 is the B&WOG :GLRP topical report for the reactor vessel internals..Section 4.5.2 of BAW-2248A identifies an action Item, ofupdating Appendix E to:BAW-10008 Part 1, Rev. 14[3]

(hereafter referred to as BAW-1 0008) concerning neutron: irradiation:embrittlement for the license renewal period. Section4.:5.2 of BAW-2248A is quoted below:

"4.:5.2 Ductility -Reduction:

of'Fracture Toughness:

BAW-10008, Part::l, Re.v.: 1,, documents the acceptability of the reactor vessel internals under LOCA and a combnination:of LOCA and seismic iloadings. The effect of irradiation on.

the material properties and deformation limits for the internals is presented in Appendix E where it is concluded that at the end of 40 years, 'the internals will have adequate ductility to absorb local strain at the regions of maximum stress intensity, and that irradiation will.not adversely affect deformation limits.

Existing literature was: reviewed with: regard to irradiation. of the reactor vessel internals.

Literature: reviewed were: (1) BA W-2060, "Project ITopical Report: Oconee Nuclear Power StationhUnitl Reactor lnterhals Life Extension.Project:; (2) EPRI TR-.103838, "PWR Reactor Pressure :Vessel Internals :License Renewal IndustfryReport; Revision 1 ", and (3) Sixth International Symposium onEnvironmental. Degradation of Materials in, Nuclear:Power:

Systems - WatedrReactors. Two: conclusions: were drawn from the review:

1) Additional testing of the physical andnmechahicalfproperty changes of irradiated material and continued surveillance of the reactor internals should-be performed to provide reliab~ledata: otn the irradiated properties of stainless steel.

2) ASME: Section Xl visual (VT-3) examination standards for category B-N-3 is a current:

and effective program for detection of cracks: and repair/replacement for vessel internals components that:are accessible by removal of thelcore and/or other internals.

However as noted: n Section 4.2 2, Examination: Category B-N-3:may:not be adequate to::

detect: reduction: of fracture toughness in comp*onen ts.: A program *is :being: implemented to:

manage the effects of aging due. to the reduction of fracture toughness of the reactor vessel A.

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47-9048125-002 Page 3 of 25 internals. This aging management program is discussed in Section 4.6. Hence, :this TLAA will be resolved:on a plant-specific basis per 10 CFR 54.21 (c)(1)(iii) based on the:resUlts and conclusion of the program."

The final safety evaluation report (FSER) of BAW-2248:by the US Nuclear Regulatory Commission (NRC) was issued on December: 9, 1999. Section 3.4.2 of the NRC FSER provided the following evaluation to Section: 4.5.2 of BAW-2248:

"3.4.2 Ductility - Reduction of Fracture :Toughness Section 4.5.2 of BA W-2248 describes a:TLAA related to the acceptability of the reactor vessel internals under loss-of-coolant-aaccident (LOCA) and seismic loading. The topical report states that Appendix E to BAW-0008, Part 1, Rev. 1, concludes "that at the:endof.

the 40 years, the internals will have adequate ductilitykto. absorb local strain at: the* regions :of maximum stress intensity, and that irradiation will not'adversely: affect deformation limits.."

The topical report indicates that this: TLAA will bei resolved on. a plant-specific basis per 10 CMFR54.2 1(c)(1)(iii).based on the results and conclusions:of the planned RVIAMP.

T he aging management approach proposed for:neutron irradiation:embrittlement (Section:

.3.34) includes examination for cracking as thleprimary management methopdfor:RVI

.components. This examination is effective in managing neutron embrittlemen Oas it relates to:

the fracture resistance of the: RYIPcomponent material/in the presence: of flaws.: The; deformation limits and adequate. ductility-described above in Appendix Eto BA W-0

1008, Part1, Rev. 1, relates to the material behavior:in the absence of flaws. Therefore, no examination looking: for flaws will/be sufficient to demonstrate resolution of this:TLAA, which will requiredetermination of the expected material properties at the end of the-license renewal period..As described in: the topical report, :the planned RVIAMP.program will provide the data necessary to resolve this TLAA.

Therefore, this item:should be. addressed as a renewal applicant actionmitem on amplant-specific-basis.-This is Renewal ApplicantAction Itemn :12."

Note:

TLAA-. time;limited aging analysis RVIAMP - reactor vessel: internals aging: management program 2.2 BA W-lOOO8 Partl,:IRe. V:: 1,Appendix E BAW-10008:used properties of unirradiated Type 304 austenitic stainless steel in determining allowable stresses. Section 3.2.of.BAW-10008 stated that "... In Appendix E - a discussion of the effect of irradiation on material properties -, it is concluded:that: the use of unirradiated properties is conservative."AppendixE - Effect of Irradiation on the Material Properties and Deformation Limits for Reactor Internals is qUoted below.

"All structural materials of the reactor internals (except for bolts) are 304 stainless steel.

Curves showing the ect* of irradiation on nthe yield strength, :ultimate strength, and ductility:

ofW304:SS ar given in: Figures E-I, IE-2, and E-3. (Note that these are typical properties.

Minimum specification properties are used in:selecting design allowable:stress.) These curves show that the strength of 304:SS is:enhanced :by irradiati6n; thus, margins of safety for the occurrence of LOCA and/or earthquake are increased:with:reactor:internals exposure.

A*

AREW:

47-90481,25-002 Page 4 of 25 The maximum

'primarystress permitted in the internals for Case IV (LOCA plusearthquake) islfimited:to two-thirds of the ultimate strength. As irradiation, increases the yield strength to this vatue, plastic strains approach zero, so that the small loss of ductility is inconsequential.

The maximum: fluence in the core barrel in the region near flanges is much less than 1020 nvt

(E: MeV) at.theend:of a 40-year design lifetime. This is the region of maximum stress intensityand the region wherbe a loss of ductility would be detrimental. As noted in Figure E.

3, the uniform elongation is greater than.20% for this fluence. It is concluded, that even at the

.end of a: 40-year lifetime, the internals will have adequate:ductility to absorb local strain at:

the regions of maximum stress: intensity, and the: irradiation will not adversely affect deformation limits.

t 2.3; Scop-e The internals locations: where the: loss of ductility: due to neutron. irradiation was considered detrimental by Appendix E: to:BA.W-1 0.008 are the core barrel flanges. Specifically, Appendix E stated *hat the. maximum :fluence. near the core: barrel: flange region:would be much less than 1020 nVtt E>1 MeV(nol te, the fluence:unit ýnvt' is an archaic form of 1"nlcm2"!):at: the end of a 40-yeardesign lifetime.. Based on Figure 2-1 (Figure E-3 in:BAW 1,000,8),:the typica.uniform::

elongation would be greater than :20% for this fluence and:therefore Appendix Econcluded that, "even at the end of a 40-year lifetime, the internals will:have adequate ductility to absorb local strain at the regions of maximum stress intensity, and that irradiation will not:

adversely affect deformation limits.'"

The scope of this document is to:address the commitment in Section 3.4.2 of the NRC FSER of BAW-2248 related to.-the above conclusion for a 60-year lifetime. Specifically, the commitment isas follows:

'.*. Lresolution: of this, TLAA... will require ý determination of the expected material properties at the end of the license. renewal period."

Therefore, this docdument will:

1. update fluence of the core barrel flangeslto a 60-year lifetime;

2. examine the validity of: Figure E-3 in BAW-1 0008;

3. examine the "deformation.limits' assumed in BAW-10.008;

4. Update AppendiX E to BAW-10008 to a 60-year. lifetime.

A.

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47-9048125-002 Page..5of.25 10 50 40 C

Q C

w C

ROOM reaperiture

.(V2F)

WA SCALEý 20 ID 0

AS RECEIVED

• I 9,0

-Exposure, nvf(t 3 1 May)

+1,022+'

Figure 2-1 Effect of irradiation on uniform elongation of Type 304SA Irradlated at:290 0C0(s540F) and tested at:various temperatures.: (Note, this is Figure E-3 in:Appendix E to BAW 10008 Part 1, R1 ev.

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47-9048125-002 Page 6 of 25 3

Evaluation 3.1 Locations of Maximum Flux for Core Barrel Flanges

] [4]

3.2 Fluence of CoreBarrel Flanges Based on Section 3.1, the maximum neutron flux locations of the: core:barrel flanges, L.e.,

locations closest to the core, are the following:

1. Top corebarrel flange::[

] below the top surface of the top flange on. the i.D. surface

2. Bottom core barrel flange:AlD. surfacef []Iabove the bottom surface. ofthe: bottom flange,on the 1.0. surface.

The maximum. flux values are listed in Table:3-1. These:flux values are based on a hypothetical fuel cycle design that has a conservatively high* neutron leakage. The high-leakage comes from loading: high powered fuel assemblies on the periphery* of the core. The peripheral assembly powers are so high that the leakage represents a bounding value; no actual fuel cyCle design would be expected to have a higher neutron flux leaving the core. Thus, the flux values in Table 3-1: represent: an upper bound for the coreibarrel flanges. in addition, the neutronics model represents an approximation of the axial. material regions between the core and.flange locations.

The material* approximations allow a higher neutron flux to reach:the: barrel flanges:than would be expected.with: a more. detailed model that contained fewer approximations.

The: bounding flux values are transformed to bounding fluence values at the end of a 60-year lifetime: by. utilizing the following conservative assumptions. for the plant's capacity factor:

• [

]

A ARE VA

47-9048125-002 Page 7 of 25

.Based on the above load capacity factor, the 60-year lifetime reepresents 54 effective full power years (EFPY), or.90% overall load capacity factor over the 60-year iifetim e. The 60-year lif time maximum fluence forithe core barrelflanges are listed in Table3-1.

Table 3-1 Maximum Fluence of the Core Barrel Flanges Location of Maximum Fluence Flux, nlcm2/sec E>1 WV 60,Year (54.EF*i)

Fluencen/cm2 E> 1 M......

Top core barrel flange:

(

] below the top surface of the

[

]

[

]

top flange (ID. surface)

Bottom core barrel flange:

]

above the bottom surface of the C

[

bottomflange:(1.D. surface) 3.3 Tensile:Data of Type 304SA Irradiated to Moderate Fluence The [

] maximum 60-Year fluence forthe~top core barrel flange remains below the 1020 n/cm2, E>1 MeV assumned in Appendix E for a 40-year lifetime. However, the:

maximum 60-year fluence for: the bottom, core barrel flange is [:

exceeds the 1020 n/cm2, ED MeOV assumnp: tion.

Figure 2-l shows.a arapid decrease in uniform elongation between fluence:[

],a fluence. level relevant to the bottom core barrel flange. Although the reference: was not listed in BAW-10008, the Figure 2-1 :data are most likely from Type 304SA:Irradiated: infast breeder reactors. In order. to :examine. the validity of Figure 2-1 for the bottom core barrel flange, a search for more recent Type: 304SA. test data between [,

especially from BWRs or PWRs, has: been conducted. The :search results are :summarized below.:

Decommissioned PWR Core Barrel,- MRP-128[11 and MRP-1 2916]

MRP-129 reported tensile tests of Type 304SA stainless::steels removed:from vessel internals:of adecommissioned PWR. A specimen machined from:the removed Type.304SA core barrelhad accumulated

1. The irradiation temperature of the removed core barrel sample was between 535 and'5750 F (279 and 3020C) over the PWR operating histor.y. This Is also. similar to the operating. temperature of 560*F (2930C) for the bottom core barrel flange. The:tensile specimen.

geometry machined from the removed core barrel is shown in Figure 3-3. The tensile test Wass performed at a strain rate -7.6x104Isec. The stress-strain:curve at 608°F (3200C) is shown in Figure 3-4 and the test: results are summarized in: Table: 3-2.

A

47-9048125-002 Page ý8 of 25 Table 3-2 Tensile Properties of Type 304SA Core Barrel from: a: Decommissioned PvW5 6]

Irradiation temperature, 535 to 5T5F (279 to 6620C)

TetTm.Fluence 0.2%:Yield Ultimate Uniform.

Total TestTemp.

Tensile

, Elng.

Elongý.

Specim en ID................

°C

'F n _M2 MPa MPa pa E> 1MeV BWR Riser Pipe -BWRVIP-351.7[1>

In 1997., BWRVIP-35 reported tensile test results of stainless steels removedfromm reactor vessel internals of two Swedish BWRs, [

]. The tested materials included portions ofa riser pipe (148 years in the reactor core) containing Type 304SA base metal from

[. The irradiation temperature for the rser pipejis estimated as the core coolant temperature of 280°C:C(536°F) and the:fluence was up tol[

] The: tensile specimen geometry machined :from. the riper pipe: is: shown in Figure 3-5. The tensil eI test results are summarized in Table 3-3.

Table.3-37Tensile Properties:of Type 304SA from BWRsV1 Bas

Me al I rradiat]i on,;: e.:.:]

e.....

. [ i e 280'C.ii 36*.x"..

U lt*i

"'i =: = :

... i : :...........im a t........

T o......

al.......

resth emp.

Fluence 0.2%rYield Utimateuniform eTotal BWR baht ererm v

,_eo Tensile Elowg.

Elong.

Component (

Specimen ID n

/omt

~C

'F E>1mV MPa MPa

%0/

Riser Pipe Base Metal J I.......

I 3

_ [ 3C

(Type 304[J

]

))

C))

SA)__

,BWR Neutron-Absorber Tube and.Control..Blade Sheath - Chuna et al.181 The 19.93 paper by Chu ng et: al. reported tensile propertycdaita of 304SA neutron absorber tubes and a control blade sheath extracted from: severaln unnamed ýoperating: BWRs,. The. slow strain rate test (SSRT): specimen size machined fromn the control blade: sheath was 57.2mm x 1:2.7mm x 1.22mm. The SSRT stress-strain curves of commercial purity'Type 304SA neutron absorber tubes and the control blade sheath irradiated toa mderate fluence levels are shownlin Figure 3-6.

The :irradiation temperature:of these:BWR comp6nent items was not reported,-but could: be assumed to be approximately 2889C (550,F). Resultsof the SSRT tested in air at:289WC4(5520°):

are summarized in Table 3-4.

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47-9048125-002 Page:9 of 25 Table 3-4 SSRT Results of Commercial Purity Type 304SA from BWRs1 81 Tested in Air at :8C (52"F 1r, Strain:Rate 1.65 x 10h/sec..

Flene 2%Yild

Ultimate, Uniform Total Spce 1 oellý Source SSRT E:1MeV a

Tehnsiile iEiong. 2' Elong.

BL-BWR-2L

3891A, CP304-A. IR42 2x10 2 1 184 NO) 156P 1.0 BL-BWR-2M 389M2D CP304-A IR-3 6 x 1O(.

221 465 30 34.,

C7131W LCS-5 CP304-B IR-17 2.3 x 1020

~360 577

33 34.1 C7M2K LSC-1 1 CP304-B IR-23 5.3 x 1020.

570

6. 4 17 21.4

.R13: 1ix i* ° 1*

ý*.4 Lm...

n r:*

....... 1=

0,2 2.

Ine irradiation tempera~ure 1ins1ide the BAr Wasiu IIIJUI ii, I U, UUtL: *e e.UIU s alou

.2881C (550)F.1 Uniform elongation values were estimated from the SSRT stress-strain. curves in Ref.: 8 i(shown as Figure 3-6).

The stress-strain curve shows that this specirme n e.ith er fractured before: necki ng or the test: was terminated before necking. This.appears: to be the reason why its. uniform: elongtion and total elongation values are, significantly below the: typical values for this fluence level::

Halden: BHWR Reactor - N UREG/CR-68921 ]1 The:2006 NUREG/CR-6892 report listed SSRT results of many heats of solution annealedand water*.quenched Type: 304 and Type 316 materials irradiated in the Halden boiling: heavy water reactor (BHWR) in Norway. The SSRT specimens were machined before being irradiated at

288° (550°F):in Halden. The SSRT specimen geometry* is shown in Figure 3.7. Table:3-5 summarizes the ýSSRT results tested in air of a commercial purity Type 304SA irradiated to:

moderate fluence,.

Table3-Z5:SSRT Results of Commercial: Purity Type:304SA Irradiated In Halden BHWR19 :

.:rradiatlon oTr!,*

ititure.288c (550F). Tested i Air, Strain Rate 1.65 x: 107/sec.

Test.. "_

Temp F.c i

/.

Ultimate Uniform Total Spece Test/ emp.

oluence

0.

Yielda Tensile Elong.

Elong i~s e~jnen S R T N o.

...C "F cm MPa MPa E > 1 JJ M

1 eV....

ii"";

C19-05 HR-58.

23 73 9.x 1

1130 1260 15.50 M1570 ci 9-02 HR-30 289 552 9:x 10.

888 894 6.41 10.21 Advanced Test Reactor"- Jacobs et al.°01 The: 1987 paper by Jacobs et al. listed constani elongation rate tensile (CERT) test results:of several types of austenitic stainless steels :irradiated in:the Advanced Test Reactor:(ATR):in Idaho. The ATR is a pressurized, light-water moderated and cooled with a beryllium: reflector.

The CERT specimens were irradiated at 500-625° Ff(260-329°C) in ATR. The CERT specimen geometry is shown in: Figure 3-8. Table 3-6,summdarizes the CERT results tested in air at 2880C

.(5500F).of a solution: annealed and water quenched Type 304.

1A

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47-9048125-002.

Page I0 of, 25 Table:3-6:'CERT Results Type 304SA. lrradIated in ATR C10]

.rradiation, temperature 500-625*F.(260-329'C), CERT in air at strainlrate 4i x

• O"Isec.

Test Temp.

Fluence 0.2,% oyield Ultimate Uniform Total Tensile Elong.

Elong.

E.. > 1 eV ksi ksi..

30W4 Solution Annealed:......

288.

559 064.3 41.3 48.0:

304, Solution Annealed 288 550:

8x20

.80.2 92.7 12.6 17.5 JMTR-and Halden BHWR -. Navas et: al.Cll]

The.2003 paper by:Navas: et: al. provided SSRT stress-strain:curvees for two heats of Type 304SA. One heat contained: 0.05% carbon (HC) and the. other contained 0.012%:carbon,(ULO).

OBth heats received: a sensitization heat treatment prior to irradiation. The sensitization treatme nt was 100 m. i nutes at 7500C (13820F) fol owed by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 50000C (9320F). After thi s treatmenti the HC heat had a continuous chromiumcarbide precipitation along grain boundaries while:the ULC heat had:few and discontinuous carbide precipitates at grain boundaries. The ASTM: A 262 Practice A screening test for sensitization showeda :"ditch" microstructurejfor the HC heat. and a dua!".:microstructuire for the ULG heat. Th:eSSRT specimenigeometry is.shown:

Figure:3-9.

Neutron irradiation of the SSRT specimens was performed in the Japan Materials Testing Reactor (JMTR) and Halden reactor in. Norway. The JMTR is a light-water type test reactor used for testing materials for BWRs and PWRs. The.SSRT specimens were irradiated at 2900C (5540F) to7.0x. 019* and 1.1.x1:

h0 n/cm 2, E>1 MeV in the: JMTR reactor and *to 5 xlxP0o and1I.2 x01 E>1 MeV inthe Haden reactor. The typical SSRT stress-strantacurves in inert gas at 29000 *(5:5F) are shown in Figure 3-10. The SSRT strain rate was 3,5x1 ONsec. The unifom and:total elongations: estimated from the Figure 3-10 SSRT stress-strain curves. are listed in Table 3-7.

Table W7 Uniform: and Total Elongations of Type 304SA: Irradiated in JMTR: and:Halden

[Reactors Ll]

l..

rdiatlohntemOperatUreý29 0**ý55R f), SSRT i*nInert gas at strain rate 3.5.x ICIOsec.

Test Temp.:

Fluence Uniform Elong.

Total Elong.

Heats.

ni i

2 F

E > IEMeV 0

48 51 7 x 10" 34 37 Type7 304 HC (0.05%/C):

290.54 lx1O 38 40 X

5x1 22:

25 1.2 x10 2115 18 48 454 7x100 9 40 43 Type'304: UC (0.0.1 2%C):

290, 554<

.x0 64 20

... 5 xlol 27 37

i =;S I:0 °::*:.............

.2 3

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47-;90.4.8125-002 PageI I of 25 3.4.

Evaluation Effect of Strain Rate, A major source of the irradiated tensile data reviewed in Section 3.3 is from. SSRT or:CERT testing. When tested in simUlated BWR and PWR water, SSRT (CERT) tests are used:for:

evaluating irradiation-assisted stress corrosion cra*king (IASCC) of stainless steels in BWR and PWR reactor internals. The Type 304SA data in Section 3.3 are from the baseline SSRT specimens tested in airat elevated temperature. ITheeffect of strain rateon the tensile properties of irradiated Typea304SAis shown in Figurel3-11 for roorn temperature, 4500F, and 7000F.12]. Uniform elongation is s.seen nto decreasem::r.oderately with decreasing, strain rate.at elevated temperatures of 450F*F and 700F. Hence, the uniform elongation values from SSRT at I'

1L 71setaOre conservative compared to those obtained from conventional, tensile tests at a strain rate typically ranging: from 4 to -10" 2/sec.

Comparison of BAW-1 0008 Lines with Recent Test.Data Figure 3-.2 compars.the uniform elongationl ines in Figure62-1 (Figure: E-3 ofAppendix E)with more: recent data reviewed in:Section 3.3. The recent Type 304SA data a6reofrn a decommissioned PWR, several BWRs, a Ieavywater reactor test reactor (Halden), a.light Watef test reactor (JMTR), and. a: pressurized light water test reactor (ATR). The neutron flux levels of these reactors are mo representative of PWR internals than fast breeder reactors. Therefore, the reported.fluence and the associated tensile data from these reactors are more representative of PWR internals than those from fast: breeder reactors used for establishing the lines in Figure E-3 of.Appendix E.

The 30000.and 4000C linesl in Figure 2-1 form the lower bound for the recent test data between fluence:1 *Q:..:.and 1021n/cm*,.

MeV. Therefore, the BAW-I10008 lines have been confirmedto be conservative for the bottom core barrel flange. After neutron exposure: of [

], the: anticipated minimum: uniform:elongation of Type 304SAi [ :] per the 3000C line:and[

] perthe:40000 line. Therefore, the minimum uniformelongationof: the: bottom core barrel flange at the end ofa 60-year lifetime, is predicted to be [

fat itsioperating temperature of 560N F (2930C). The typical u niform elongation of the bottom:core barrel flange at the end of a 60-yearlifetime is predicted to: be[ ý] :at operating temperature based on the recent :test data shown:in :Figure 3-12.

Deformation Limits in BAW-10008 The,deformation limits" in Appendix E were examined in Appendix A to BAW-1 0008, which:

provided the bases for the allowable stress limits f.o the BAW-1 0008 analysis. The last two paragraphs of Appendix A contain :the: following discussion on plastic strain limit.

"Stresses compared to the 2/3 SU, allowable: are calculated on an elastic basis If a plastic:

analysis 6f an internalsc*mponent isperformedthe plastic strains are limited to* thestrain.

corresponding to 213: Si,.

Figure A-2 shows a minimum'stress-straintcurve for 304 stainless steel at 6000 F. The strain corresponding to a stress of 2/3 Su is 8.16% which is approximately equal *to 0.20 x uniform strain, or 7.6%."

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47-9048125-002 Page 12 of 25 Figure A-2 in BAW-1:0008, shown: as Figure 3-13 in this document, shows the minimum stress-strain curve of unitrradiated Type 304SA at 600 0F. According to:Appendix.A, a plastic strain

would be limited to the strain corresponding to 2/3 of the ultimate tensile strength (uTs). For unirradiated Type 304SA, this. limit would be 7.6% plastic:strain at 6000 F for a plastic analysis.

The 2/3 UTS is approximately 42: ksi (290 MPa).at 6000F. in Figure 3-13. Due to the irradiation hardening effect,: the uniform elongation limit.forirradiated Type 304SA would be lower than 7.6%. For example, the stress-straincurves in Figure13-10 indicate that the plastic strain corresponding to.42ksi (290 IMPa):would be 3% or less for Type 304SA: irradiated to above. 1020 n/cm'2,E>i MeV. In fact, asithe fluence increases, the uniform plastic strain:requirement is self-limiting, iLe., the plasticstrain to: 2/3 UTS:will always beý within the uniform elongation range. This.

is consistent with: the original statement in Appendix E that "As irradiation increases the yield strength to this value (le., 2/3* UTS), the plastic: strains approach:zero, U pdate of BAW-1 0008 Appendix E The bottom: core barrel: flange is predicted: to have*a :minimum[

]::Uniform :elongation at operating temperature: at:the end of: a:.60-year: lifetime. The typical. uniform elongaUon at the end of:a 60-year lifetime is predicted to: be'[

] at operating temperaturie:.Treforbe,:th6eoriginal::

conc lU sion, in' BAW-101000: coincrn ing thes duct lity for: a 40-YeaOr ifetime remains valid for a 60-year. lifetime. Hence, the last Oaragraph:in Appendix E1to1BAW-. 0008is updated :to the following fora: 60-year lifetime (the unederilne words have been updated):

The: maximum fluence ;in the core barrel in the region near flanges is conservatively estimated: to be [:

]:at the end of a: 60year (54 EFPY) lifetime..

This. is the region of maximum. stress. intensity: andt the. region, wheere :a. loss. of ductility would be detrimental. As.noted in. Fqurei 3,12. the.uniform elongation:is predicted to be

[

J minimum and[ ]tyica for this fluence. it Js. concluded that even at. the end of a.

60-year lifetime, the internals will have adequate ductility: to absorb ilocal strain at the regions of maximum stress, intensity, and the irradiationlwill not adversely'affect:

deformation limits.

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47-9048 1 25-002 Page 113f25 Plenum Cover

Plenumq,

)vAssembly Assembly"\\

o 'E Core Support Shield CL i :* L

Plenum :Cylinder*

K:

Assembly:

Uppoer Grid Assemblj*

3:

V E 0

t LE

~V)

~

I

~

K PTenma Shfineld AssLoermribAsebl

ýEE VMrudeTb Fiue31LWOsge W ntrasGnrlArneet(Nt:sm opnn tm r

rotatedmal forclaity coA AR)V

ýTop Flange Thermal Shield

.Upper: Restraint:

-Thermal Shield Core: Barrel Cylin SSHT Figure 3,-2 Locations of the top and bottom core barrel -flanges. In: the B&W FA-17 assembly.

47-9048125-002 Page: 14 of25 Assembly nder 7 core: barrel.

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47-9048125-0,02-Page,15 of 25' Figure, 343 Tensile specimen machined from Type i304SA removed ifrom:adeommissionod:PWR;.

all dimension unit in "inches'..

Figure 3-4'Stress-straln curve of Type 304SA core ba rre [

TIP]i A

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47-9048125-002 Page 16of:25 Figure 3-5 Ten.sile specimen machined from Typeq304SA riser pipe removed from the

[U I BWR;: all dimension unit in,;mm':*

71 ARE VA

47-9048125-002 Page 17 of 25 500 4W0 0200 100 60 Elongation (%)6 De 4C

.:ý3C

.! 2C IC 0 4C 2C 6 P3D4

(..

.:B) li 0.S x. 0E91: n ecmE-2 290 ppb-Dilssolved Oxygen 01 5

10 1520 25 30 35 40 Elongation (%)

C PCSO4sia.th (F)

O023 X 0221 nanE-2 In Air 10 15

Elongation (%)

0O a

10 15 t o20 (2 Elonglaton :(Y.)

30 15 40:

Figure 3-6 SSRT stress-strain curves of commercial purity Type 304SA removed from BWRs. Top, heat CP304-A BWR absorber tubes; bottom, heatCP304-B BWR control blade sheath. Strain rate 1.65 X 107 Sec.[8 A

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4749048.125-002 Page 180of25 57.201 19.10 k1-i'2.0 27.15 0,762 ALL DIMENSIONS IN MM FIgure 3-7 SSRT specimens:Irradiated in the. Halden boiling: heavy water.test reactor..9.

9..53 t0.25

(0.375t.00 9.53ýO.25 a0.3756 O.0101.

64.01 +/-0.01......

10. 158+/--t0.00,04)Y.

1 7*.1 1:+/-0.10 Figure 3-8 CERT specimens irradiated in the.ATR pressurized light water test reactor. Di.menslons are shown as mm (Inr.)YO1 A

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47-9048125-002 Page 19 of 25 47.

N a) Tube. specimens of HICnmaterial 1 7..............

16 1.

3,

]* *",""

S 'Ci......

[:

......... *S,

"... I T b) :Plate. specimens of ULCi mat.trial Figure 3-9 SSRT: specimens Irradiated In the JMTR light water: test reactor and :Halden BHWR reactor; a l: dime !onsion imi'nmm I.

l AREVA

47-9.048125-002 Page20 of 25

ý116 10.2 87 73 CL (J-

44~

LO-29ý

%°strain i116 102 87 i73 a

0~

Afl*

In.

L

• 0-

58 IAl

%n 44 219 15 1/ strain Figure 3-10 :SSRT stress-straini curves of Type 304SA received a sensitization heat treatment prior to.irradiatioon. Top Type 304 WIth 0.05% carbon heat, bottom Type 304 with 0.01 2% carbon.

Irradiated Inr JMTR and Halden reactors: at 2900 :(554F), testeddin inert gas at2900C (554F), strain rate 3.5 x *'iosec."1]

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47,-9048125-002 Page 2t of-25 touf e -

I.

ý -

21 v I I. I 1 11 1 a' I

I,,,,

I.

I 1.11, 11 11 WV li I

~

140:

130'

W4120, TEST TEMP

.8 450

• . 700 P. (OF)

MTEMP.,

(644::K)

RO EP

'-~450F (505 K)

I0OD f-.

lbi

!700 0 F (644 K) Ir MATERIAL:

ANN. 304 SS.

IRRAD. TEMP - 725-7300r (658-661 K)

PLUENCE - 8..8-10.3x 10o 22 n/cm 2 (E-0.1 MeI) 100 10.3 STRAIN RATE. 1/MIN

ý1'0 "6,

0:

I...........

.... I I

1 1 1 1 1 "

I

".1 TESTTEMP.

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MATERIAL:

ANN. 304 SS I)

IRRAD*. ITEMP.,

1 :725r730 0F (658-661 K):

S:450 (505 K FLUENCE 8.8.-10.3 xi.022 n/cm2 (E>OI. MeV)

700 (644 K)

.4560F (505 K)!:

7000F.(644 K) f 11 1

1

t. L I

I o v 10o-2 10-1 STRAIN RATE, IIMIN:

• 19" I 1.01, Fig u r.e,:3-1 1 :Efc 0~tan*to esl:propertles !f Type6: 304SA. The: material, was:, Irradiated at 725-7300:F inEBR-lI to.8.8-10.3 :x 1022 nicm2, E>0jIMeV.'.

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.47-9048125-002,.

Page 22 of 25"

!Figure3-1*t*2Jionp§r~ispniof!BAW-10'00Si8ýunt orm e!longiti*n lineswith= more recent ~test data Of Typ* 304*A irradiated between 10'0 and I10' nIom2, Em; MeV from water moderated reactors.,

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47-9048125-002 Page 23 of 25 10

50 40

• 30 i0)

Figure 3-13, Minimum stress-strain curve of uniarraiate)d Type 304SA stainless steel 600*C.,

(Figure: A-2 of Appendix E of BAW 1000*:8Part 1, Rewv. 1 3]).

...A...

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47w-9048125-002 Page 24 of 25 4

Summary This documrent has updated the estimated fluence of the :core barrel flanges in BAW-10008, Appendix E to a 60-year lifetime (54 EFPY). The maximum 60-year fluence.is.[

] for theltop core barrel flange and [

] for the bottom core barrel flange. The uniform elongation lines between fluence [

, in Appendix: Eto: BAW-1i0008 have been verified to be conservative:with the recent Type:304SA data. These data are: from a decommissioned PWR, several BWRs, a. heavY water reactor test re.acto~r(Halden), a light water test reactor (JMTR), and a pressurized :light water* test: reactor (ATR). Thei neutron:flux :levels.:of these r*eactors are more representative of PWR internals than fast breeder' reactors. Therefore, the reported: fluence and:the associated tensile data from:

these reactors are more representative: of PWR internals than the data from fast breeder rea0ctorsused for establishing the lines in Figure: E-3 of:Appendix E..

The. bottom core barrel flange is predicted to have a minimum [

] uniform elongation at operating temperature at the end of a 60-year lifetime (iLe., the end of the license renewal period). The typicaludiform elongation at the end of a 60-year lifetime is predicted to. be

] I at:operating temperature. Therefore, th6 original conclusion in BAW-1!0008 concering the.

ductility for a 40-year lifetime remains:valid for a 60-year lifetime. Hence, the last paragraph in Ap.pendix: E to BAW-1 0008 is updated to the following for a :60-year lifetime. (the underlihe words have been updated):

The maximum fluence fin the core barrel in the region near flanges is conservativel' estimated tobb [

1 Iat the end of a 60-year (54 EFPYJlifetime.

This is the region of maximum stress intensity and the region wherea l6ss of ductility would be detrimental. As noted in: Figure 3-lf2, the: uniform: elongation is predicted to be

[

], minimum*

and [

] toica for this fluence. It is 'concluded that evýen at the: end of a 60-year lifetime, the: internals will have'adequate ductility to absorb local strain at the:

regions of maximumnstress intensity, and the irradiation will not adversely affect deformation limits.

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47-90481 25-002 Page 25-of.25 5:

References

1. AREVA:NP Inc. Letter FANP406-0147, January 112, 2006, to H.T..Tang (EPRI),

Subject:

EPRI Project Agreement No. EP-Pi 8230/C8990, Amendment 1, Job 4160812,. Revision 3,:

PWR Internals Components: Functionality Analysis for the B&W.Design"

2. AREVA NP Inic. Document BAW-2248A, "Demonstration of the: Management of Aging Effects for the Reactor Vessel:Internals," March 2000.

3. AREVA NP 'Inc. Document BAW-1 0008, Part 1, Rev, 1,."Reactor Internals:Stress. and Deflection Due to Lossof-Coolant Accident and Maximum Hypothetical Earthquake*" June 1:970.

4. AREVA NPI.nc. Drawings [

]

.5. Materials Reliability Program: Characterization of Decommissioned PWR Vessel Internals Material Samples - Material Certification, Fluence, and Temperature (MRP-1 28), EPRI, Palo Alto, CA, and U.S. Department of Energy, Washington, D.C.: 2004. 1008202.

6. Materials Reliability Program:: Characterization: of, Decommissioned PWRý Vessel Intemals:

Material Samples-Tensileand. SSRT Testing (MRPi1 29).,EPRI :Palo Alto, CA, and U.S.,.

Department of Energy-2004. 1008205.

7. "aWR Vessel and Internals Project: Fracture Toughness:and'Tens.il.e Propertesof Irradiated Austenitic Stainless Steel Components Removed.From.Service (BWRVIP-35),".EPRi TR-108279,:June 1997, Electric Power Research Institute: (EPRI),. Palo Alto, CA.

8. H.M. Chung etat.,: "Stress Corrosion Cracking Susceptibility of Irradiated Type 304 Stainless Steels," Effects ofRadiation on Materials: 16th Inte*n*tionalfSymposium, ASTM STP 11:75:

AS..Kumar et al., Eds., American Society, for Testing and Materials, Philadelphia, 1993.

9.

KHM. Chung& :W.j: Shack, NUREGCR-6892, ANL-04/*10, "Irradiation-AssistediSttess:

Corrosion Cracking Behavior of Austenitic Stainless Steels Applicable to LWR.Core Internals," 2006, U.S.:: Nuclear Regulatory Commissiosi*n,:Washington, DC 20555-000.1.

10. A,.AJ. Jacobs et al., "Radiation:Effects on the Stresýs Corerosion ::and,Other Selected Properties of Type-304 and Type-316 Stainless Steels," the 3rd International Symposium on Environmental Degradation of: Materials in:Nuclear Power Systems-Water Reactors, TMS,

1987.

11. M. Navas et al., "IASCC Susceptibility of AISI 304 SS with Different Carbon Content in BWR Conditions," the 11 th: International. Symposium on Environmental Degradation of: Materials in Nuclear Power Systems-Water ReactorsANS, 2003.

12. R.L. Fish and C.W. Hunter, 'Tensile Properties :of: Fast:Reactor Irradiated Type.,304:

Stainless Steel, " Irradiation Effects of Radiation on the. Microstructure: and: Properties of Metals, ASTM STP 611, American:Society forTesting and Materials 1976,::pp. 1 19-138.

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