ML20196A287

Forwards Response to Generic Ltr 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Matls & Impact on Plant Operations Using Methodology in Rev 2 to Reg Guide 1.99

ML20196A287
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	11/30/1988
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 8812050302
Download: ML20196A287 (14)

Text

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g lill PortlandGeneralElectricConixwly David W. Cockfield Vice President, Nuclear Novemboe 30, 1988 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nucicar Regulatory Commission ATTN: Document Control Desk Washington DC 20555

Dear Sirs:

Response to Coneric Letter 88-11 J

Conoric Letter 88-11 "NRC Position on Radiation Embrittlement of Reactor Vessol Materials and Its Impact on Plant Operations", requires licensoes to reanalyze neutron radiation embrittlemont of the reactor vossol beltline materials using the revised methodology of Revision 2 to Regul? tory Guide 1.99, "Radiation Embrittlement of Reactor Vossol Materials".

The generic letter required licensees to submit their analyses and schedules for corrective ac+1ons within 180 days of the effective date of Regulatory Guido 1.99, Revision 2.

Attachment A to this lottor providos Portland General Elaetric's (PCE's) response to Concric Lottor 88-11.

As discussed in Attachment A, PCE has deviated from the Regulatory Guido with regard to use of Plant survoillance data. Although justification is provided for this doviation, NRC approval of PCE's methodology is requested prior to implementation of correctivo actions.

An estimated schedulo for correctivo action implementation is provided in Attachment A.

This schedulo is contingent on NRC acceptanco of PGE's analysis.

Sinceroly, W

r Attachment e

c:

Mr. John B. Martin p$O Regional Administrator, Region V

]O fL U.S. Nuclear Regulatory Commission

~8 Mr. William T. Dixon Stato of Oregon yg 30 Department of Energy

$b Q4 Mr. R. C. Barr 6

N NRC Resident Inr octor I

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Trojan Nuclear P ant g \\

n o.L m sw sn umawa cw m

s Trojtn Nuc1Scr Plcnt Document C:ntrol D:ck

• Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 1 of 13 RESPONSE TO U.S. NUCLEAR REGULATORY COMMISSION CENERIC LETTER 88-11 Introduction The Nuclear Regulatory Commission (NRC) issued Generic Letter (CL) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vossol Materials and Its Impact on Plant Operations", on July 17, 1988. The purpose of CL 88-11 was to call attention to Revision 2 to Regulatory Guide (RG) 1.99, "Radiation Embrittlement of Reactor Vessel Materials".

The generic letter requires that licensees perform analyses of neutron embrittlement using the revised RC 1.99 methodology, to ensure continued compliance with Appendix G of Title 10 of the Code of Federal Regula-tions, Chapter 50 (10 CFR 50).

The analysos and a schedule of resulting corrective actions, if any, are to be submitted to the NRC within 180 days of the effective date of the revised Regulatory Guido.

Analyt!.s Regulatory Guide 1.99, Revision 2, provides a revised methodology for determination of the effects of neutron embrittlement on reactor beltline materials. The end product from implementation of RG 1.99 is an adjusted reference temperature (ART) for the limiting vessel beltline material.

This ART is then used in determining the Plant heatup and cooldown pros-sure-temperaturo (P-T) limits. The Regulatory Guide providos two methods of calculating the ART.

The first method is for thoso Plants for which no surveillanco data (i.e., incore material capsule data) exists, and the second method is for Plants which have survoillanco data available.

Portland Conoral Electric (PGE) has a Radiation Survol11anco Program for the Trojan Nuclear Plant roactor vessel. Table 4.4-5 of the Trojan Technical Specifications (TTS) providos a list of the six Trojan survoll-lanco capsules designated by letter "U" through "Z" and the removal scho-dule of each survoillanco capsulo. The results of the analysos for specimons from survoillance capsulos "U" and "X" have been reviously submitted as Westinghouso Roports WCAP-8426(1), WCAP-8469(2 and WCAP-10861(3).

Regulatory Guido 1.99 references Paragraph 5.1 of American Society for Testing and Materials (ASTM) E 185-82, "Standard Practico for Conducting Survol11ance Tosts For Light Water-Cooled Nuclear Power Roactor Vessols",

as tho governing codo for selection of materials for incoro survol11ance capsulos. Uso of ASTM E 185-82 would result in hoat B9883-1 as the limiting matorial for the Trojan vessol. However, material for the Trojan survoillanco capsules was chosen in accordance with the guidelines in ASTM E 185-13.

This resulted in selection of heat C5583-1 as the limiting material for the Trojan vossol.

Since the copper and phosphorous contents of the two heats of interest (both lower sho11 platos) woro within tho standard product analysis tolerances as defined by the ASTM codos, and the initial RTNDT.

s Trojen Nucloce Plant Documint Control Disk

' Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 2 of 13 temperatures were within 10*F, the materials were originally judged equivalent in these respects. Selection was thus based on the lowest upper shelf Charpy energy criteria. This resulted in the material with the lower initial RTNDT and lower residual content being placed in the capsulos. Tho substitution in RG 1.99, Hovision 2, of nickel for phos-phorous as a residual of concern does not alter this situation nor the chemical equivalence of the two materials. The nickel contents of 0.60 percent for heat C5583-1 and 0.62 porcent for heat B9883-1 are within the standard product analysis tolerance of 0.03 weight percent as established by the ASTM Code (4,5),

The method usod here to calculato the adjusted referenco temperaturo for the Trojan reactor vessel departs from the method of Revision 2 to Rogu-latory Guido 1.99 only in that linear extrapolation of credible survoll-lanco capsulo data is used to predict the proporties of the limiting material, B9883-1.

This adjustmont of capsule data is done by multiply-ing the chemistry factor of the survoillanco material by the ratios of the average chemistry factors for each heat as provided in Tablo 2 of RG 1.99.

This is justified for two reasons.

Firstly, the concentrations of all residual olomonts of interest either in the past por ASTM E 185-73 (coppor, phosphorous) and now por ASTM E 185-82 (copper, nickol) are well within the normal product analysis tolerances for this class of steel.

Secondly, the difference in initial RTNDT betwoon B9883-1 and C5583-1 is only 10*F, which is loss than the published standard deviation of 17'F for the calculational method.

Identification and Location of Beltline Rer. ion Materials Figure 1 identifies and indicates the location of all bolt 11no region materials for tho Trojan reactor vossol. Tho beltlino region is defined to bo "the region of the reactor vossol (shell material including wolds, heat-affected zones, and plates or forgings) that directly surrounds the j

offectivo height of the activo core and adjacent regions of the coactor i

vessel that are predicted to exporlonco suf ficient neutron irradiation damage to bo considered in the solection of the most limiting matorial with regard to radiation damago"(6),

I l

Definition of Plant-Specific Material Proporties The portinent chemical and mechanical proporties of the bolt 11no region plato and wold materials of the Trojan reactor vossol are given in Tablo 1.

The "woight-porcent coppor" (Cu) and weight-porcont nickel (NL) matorial chemistry values and the initial RTNDT values shown in the tablo are the samo as those used in the Trojan pressurized thermal shock submittal (6),

y_

1 Trojen Nucisar Plant Docum nt Control Dssk "Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 3 of 13 4

The chemistry factors and "margin" (M) terms that are also shown in Table 1 were determined in accordance with RG 1.99, Revision 2 methods.

Chemistry factor and margin values in Table 1 that were based upon credible survelliance measurements are also given for the beltline region materials where this data was available [see Table 2 from the Trojan I

Surveillance Capsule Program (3)). Table 3 gives further information on a

the determination of the chemistry factors from this data in accordance with the applicable procedure from Revision 2 to RG 1.99.

For those

' reactor vessel materials whero credible surveillance data exists, the L

respective chemistry factor and margin terms are lower than the values based on material chemistry measurements.

F Table 4 demonstrates the conservative method of extending the credible r

surveillanco capsule data to the nominally equivalent, but technically limiting lower-shell plate B9883-1.

The fact neutron fluence for the vessol inside surface was adjusted using the RG 1.99, Revision 2 methods l

to determine predicted fluences at 1/4T and 3/4T locations. The fast neutron fluence for the vessel inside surface is shown as a function of full-power service life (EFPY) in Figure 2.

Adjustment of the survoillance capsule chemistry factor for heat C5583-1 is accomplished by multiplying that factor by the ratio of the chemistry factors for heat B9883-1 and heat C5583-1 from Table 2 of Revision 2 to RG 1.99. This adjusted chemistry factor is then used to calculate i

Addition of initial RTNDT for host B9883-1 (10*F) and an ARTNDT.

appropriate margin (17'F) yields the adjusted reference temperature (ART).

Table 5 summarizes the adjusted referenca temperatures dotormined and compares them to the equivalent total RTNDT from RG 1.99, Revision 1.

It is apparent that the adjusted reference temperature for the 3/4T i

i location is higher at both the 5 and 10 EFPY times than the current 3/4T temperature.

Estimated Impact on Plant Operations and Schodulo for Implementation The increano in the 3/4T adjusted referenco temperatures at both 5 EFPY and 10 EFPY roquires that the hoatup and cooldown pressure-temperaturo operating limits be co-evaluated tor applicability.

Schedulos for implementation of correctivo actions are dependent on NRC approval of PCE's analysis methods regarding application of Trojan surveillance capsulo data. The hoatup-cooldown curvos, associated Technical Specifi-cations, and any proceduro and/or hardwaro changos will be implainented by 6.he end of the 1991 Refueling Outage.

The following is an estimated schedule, assuming acceptanco of PCE methods:

(a) Calculation of revised hoatup-cooldown curvos:

3-4 Months.

(b) Submittal of Licenso Chango Application (LCA) to revise Technical Specifications:

2-3 Months.

t s

iT';jon Nuclter Plcnt D;cument Control Dick

-4c-ket 50-344 '

November 30, 1988

• Lf..anS 3 NPF-1 Attachment A Page 4 of 13 (c) NRC approval of LCA:

6 Months.

(d) Changes to procedures and system n.odifications:

6-12 months.

Determination of RTpys Values Using the Regulatory Culde 1.99, Revision 2 methodology represented in Section 2, RTPTS values were recalculated for all beltline region materials of the Trojan reactor vossol at end-of-life (32 EFPY), as shown in Table 6.

The initial RTNDT values, chemistry factors, and margin terms from Tablo 1 were used in the calculation, including the respective data from credible surveillance measurements.

A peak inner surface neutron fluence value of 2.90 x 1019 2

n/cm,

which corresponds to a fluence factor of 1.28, was used for all the vessol materials, includir.g the longitudinal wolds.

The longitudinal welds experience neutron fluences somewhat less than the peak value, but it is not significant.

All of the end-of-life RTPTS values are less than 200*F, which is well below the respectivo PTS screoning criteria of 270*F for longitudinal welds, plates, and forgings and 300*F for circumferential wolds (7).

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Trojan Nuclear Plant Document Control Desk Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 5 of 13 FIGURE 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL

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Trojcn Nuc1ccr Pltnt Document Cintral Deck

' Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 7 of 13 TABLE 1 TROJAN REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Copper Nickle Initial (D)

(Cu)

(NL)

RTNDT M

,)

Beltline Material Wt.%

Wt.%

CF

(*F)

(*F)

Intermediate Shell Plate 0.12 0.58 82.6 0

34 C5582-1 Intermediate Shell Plate 0.15 0.56 107.8 10 34 C5587-1 Lower Shell Plate B9883-1 0.16 0.62 118.5 10 34 Lower Shell Plate C5583-1 0.15 0.60 110.0 0

34 (75.87)(d)

(17)

Longitudinal Welds 0.06 0.97 82.0

-20 56

( 3 9,. 6 )

(28)

Circumferential Weld 0.06 0.97 82.0

-20 56 (39.6)

(28)

(a) CF = Chemistry Factor.

(b) The initial RTNDT values for the plates and welds are measured values.

(c) Margin (M) as por Regulatory Guide 1.99, Rev. 2, the standard devia-tion for the initial RTNDT margin term is assumed to be zero since the initial RTNDT values were obtained from conservative (ie, "upper bound") test results.

(d) Numbers in ( ) correspond to surveillanco capsule data.

DRS/YLB/mr 2675W.1188

F:

s Trojaa Nu21ccr Plcnt Document C;ntesl De k Docket 50-344-November 30, 1988 License NPF-1 Attachment A Page 8 of 13 TABLE 2

SUMMARY

OF TPOJAN REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS(3) 30 Ft Lb 50 Ft Lb Upper Shelf Trano. Temp.

Trans. Temp.

Energy Capsule Fluenge Increase Increase Decrease Material

_Ident.

(n/cm )

(*F)

('C)

(*F)

('C)

(Ft Lb)

(J)

Plate C5583-1 U

3.88 x 10 53 29 55 31 6

8 (Longitudinal)

X

1. 7 7 x 10 '

90 50 100 56 14 19 Plate C5583-1 U

3.88 x 10 44 24 65 36 0

0 (Transverse)

X 1.77 x 10 '

95 53 120 67 8

11 Wald Metal U

3.88 x 10 22 12 32 18 2

3 X

1.77 x 10 50 28 55 31 3*

4(*

Hsat Affected Zone U

3.88 x 1018 19 11 30 17 5

7 Metal X

1. 7 7 x 10 60 33 60 33 14 19 (c) Upper Shelf Energy Increaso i

DRS/KLB/mr 2675W.1188 l

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Tr jan Nuc1Ctr Pirnt Document C;ntral De k Docket 50-344-November 30, 1988 License NPF-1 Attachment A Page 9 of 13 TABLE 3 CALCULATION OF CHEMISTRY FACTOR USING TROJAN SURVEILLANCE CAPSULE DATA Fluence ART 19 2

Fluence NDT ff x ART 2

Material capsule _ (10 n/cm )

Factor (ff)

(*F)

NDT ff Plate C5583-1 U

0.388 0.7378 44 32.46 0.544 (Transverse)

X 1.77 1.1569 95 109.91 1.338 t

Plate C5583-1 U

0.388 0.7378 53 39.10 0.544 (Longitudinal)

I 1.77 1.1569 90 104.12 1.336 I=

285.59 3.764

• Chemistry Factor (CF) (Plate C5583-1) = E(ff x ARTNDT)

=

285.59 75.87 a

2 Z(ff )

3.764 i

Weld Natal U

0.388 0.7378 22 16.23 0.544 X

1.77 1.1569 50 57.84 1.338 i

I en 74.07 1.882 i

• Chemistry Factor (CF)

.al) = f(ff x ARTNDT) 74.07 = 39.36

=

2 Z(ff )

1.882 l

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DRS/KLB/ar 2675W.1188

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Trojcn Nucicer Plant Documsnt Control Dask

-Dockot'50-344 November 30, 1988 Licenas NPF-1 TABLE 4 Attachmsnt A Page 10 of 13 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE (ART)

FOR LIMITING MATERIAL B9883-1 BY EXTRAPOLATION OF SURVEILLANCE CAPSULE DATA ARTNDT = (CF)(ff)(a)

CF (c) for B9883-1 = 75.87(b) ( B9883-1)

(ff) = 75.87 (118.5'F)

ARTNDT (CFC5883-1)(c)

(110,0*F) (ff)

ARTNDT = 81.7 (ff)

19)

• For 10 EFPY f = 0.9 (10 f1/4 = 0.5404 (10 ');

ff

= 0.828(*

j f3/4T = 0.1949 (10 ');

ff3/4T "

1 ART NDT

• ARTNDT "

ART = ARTNDT +

NDT +

+

l ART = 67.6 + 10 + 17 = 94.6*F at 1/4T ART = 46.0 + 10 + 17 = 73*F at 3/4T

19)

• For 5 EFPY f = 0.45 (10 gj g = 0.643(*'

i f

= 0.27 (10 ff j

[

f

= 0.10 (10 ff 3/4T p

l ARTNDT " (

j ART

= (81.7)(0.417) = 34.1 at 3/4T NDT ART = 52.5 + 10 + 17 = 79.5 'F at 1/4T l

ART = 34.1 + 10 + 17 = 61.1 *F at 3/4T

(

l (a) CF = Chemistry Factor f at innor vessel surface; I

f = Fluence from Figure 2; f a

g f1/4T = f at 1/4 thickness; f3/4T = f at 3/4 thickness, i

ff = Fluence factor From Figure 1 of Regulatory Guide 1.99, Hovision 2.

[

t (b) Chemistry factor for surveillanco capsule as calculated in Table 3.

j l

l (c) From Table 2 of Regulatory Guide 1 99, Revision 2.

DRS/XLB/me/2675W.1188

7 Trojan Nuc1Ccr Plcnt Document Crntral De;k Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 11 of 13 TABLE 5 m

COMPARISON

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURE (ART) VALUES Method 5 EFPY("'

10 EFPY "

Re&ulatory Guide 1.99, Rev. 2(a) 1/4T 3/4T 1/4T 3/4T 80'F 61*F 95'F 73*F Regulatory Guido 1.99, Rev. 2(D)

J./ 4T 3/4T 111'F 55'F (a) As calculated by extrapolation of surveillance capsule data.

(b) The present heatup and cooldown P-T curves are based on these values.

(c) EFPY = Effective Full-Power Years.

DRS/XLB/mr 2675W.1188 i

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Trajan Nusic:r Picnt Document C:ntr21 De k Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 12 of 13 TABLE 6 TROJAN REACTOR VESSEL BELTLINE RTPTS VALUES AT END-OF-LIFE (32 EFPY)

USING REGULATORY CUIDE 1.99, REVISION 2 ART

+ "

NDT (*F)

NDT +

"#E " "

PTS CF *)x FF

('F)

(*F)

(*F)

I Material i

Intermediate Shell Plate 82.6 1.28 0

34 140 C5582-1 Intermediate Shell Plate 107.8 1.28 10 34 182 C5587-1 Lower Shell Plate 118.5 1.28 10 24 196 B9883-1 Lower Shell Plate 110.0 1.28 0

34 175 C5581-1 (75.87)

(17)

(114)(c)

Lon& tudinal Wold 82.0 1.28

-20 56 141 l

(39.6)

(28)

(58)(c.d)

Circumferential Wald 82.0 1.28

-20 56 14 ?.

(39.6)

(28)

(53)(c)

(a) CF = Chemistry factor as calculated in Table 3.

19 (b) Fluence factor (ff) based upon peak inner surface fluence of 2.89 x 10 n/cm2 [5, 6).

(c) Numbers in ( ) indicate RTPTS values based on surveillance capsule data.

(d) Neutron fluenco for the lon51tudinal welds was conservatively assumed to bo equal to the peak inner surface valuo.

DRS/XLB/mr 2675W.1188

Trajtn Nus1Ccr Pltnt Document C ntrol Desk

' Docket 50-344 November 30, 1988 License NPF-1 Attachment A Page 13 of 13 REFERFNCES (1) Davidson, J. A.; Phillips, J. H.; and Yanichko, S.

E.,

"Portland General Electric Company, Trojan Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-8426, January 1975.

(2) Davidson, J. A.; Anderson, S. L.; and Kaiser, W. T.; "Analysis of Capsule V from Portland General %1ectric Trojan Reactor Vessel Radiation Surveillance Program", WCAP-9469, May 1979.

(3) Yanichko, S. E. ; Anderson, S. L. ; and Kaiser, W. T. ; "Analysis of Capss10 X from Portland General Electric Company Trojan Reactor Vessel Radiation Surveillance Program," WCAP-10861, June 1985.

(4) Metals Handbook, Ninth Edition "Propertias and Seleccion:

Irons and Steels" Volume 1, 1978.

(5) American Society for Testing ano Materials A533/A533 M-86, "Standard Specification for Pressure Vnssel Plates, Alloy Steel, Quenched and Tempered, Manganese-Molybdenum and Manganese-Nickel",

October 1986.

(6) Trojan Nuclear Plant, "Calculated Values of Pressurized Thermal Shock, Reference Temperatures (RTPTS).

PGE letter:

B. D. Withers to Steven Varga, USNRC, January 12, 1986.

(7) Nuclear Regulatory Commission, 10 CFR 50, "Analysis of Potential Pressurized Thermal Shock Events," /ederal Register, Volume 50 No. 141, July 23, 1985.

DRS/KLB/mr 2675W.1188

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