ML20104B146

Forwards Info Re Ref Temps for PTS (10CFR50.61) Covering Reactor Vessel Beltline Matls,Including Current & Projected Values for Ref Temps for Beltline Matls at Inner Vessel Surface & Copper & Nickel Contents of Matls

ML20104B146
Person / Time
Site:	San Onofre
Issue date:	09/09/1992
From:	Rosenblum R SOUTHERN CALIFORNIA EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9209150009
Download: ML20104B146 (13)

Text

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l Southem Caliromia Edison Company D PAf IF l- M bl Mt L T IHvhJr. C ALW OH5a A 92 7 sit n u nc+r nou,u September 9, 1992 m,,.,-

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m,m 2001.o m..s U. S. Nuclear Regulatory Commis' n Attention: Document Control ' u Washington, D.C. 20555 Gentlemen:

Subject:

Docket Nos. 50-361 and 50-362 Reference Temperatures for Pressurized Thermal Shock (10 CFR 50.61) - Reactor Vessel Beltline Materials San Onofre Nuclear Generating Station Units 2 and 3

Reference:

September 9, 1992, letter from Harold B. Ray (SCE) to Document Control Desk (NRC),

Subject:

Docket No. 50-362 Amendment Application No. 102. Changes to Technical Specifications 3/4.4.8.1, 3.4.8.3.1 and 3.4.8.3.2 San Onofre Nuclear Generating Station, Unit 3 ransmittal provides the information required by paragraph (b)(1) of 10 CFR

, " Fracture Toughness Requirements for Protection Against Pressurized

,hermal Shock Events," for San Onofre Units 2 and 3. The enclosed information is being submitted in conjunction with the above reference which updates the San hofre Unit 3 reactor vessel pressure-temperature limits, included are the for the current and projected values of the reference temperatures (RTpu)lso reactor vessel beltline materials at the inner vessel surface. A included as requi ed, are 1) the bases for the current and assum tions regarding core loading patterns, the 2) projected copper and nickelvalues and the contents, and 3 the fluence values used in the calculation for each beltline material.

The reference temperatures have been calculated by the method given in paragraph (b)(2) of 10 CFR 50.61. The results are that no reference temperature for any beltline material is projected to exceed the Pressurized Thermal Shock (PTS) screening criteria prior to February 16, 2022 for Unit 2 or November 15, 2022 for Unit 3. These dates are 40 years after the issuance of the respective operating licenses, wnich are well beyond October 18, 2013, the current expist. ion date of both the Units 2 and 3 operating licenses.

i l' W 2o'9150w PDR ADOCK 9 05000361 920909 ,

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D'ocument Control Desk 1 Because the reference temperatures for the beltline materials are projected to l remain below the PTS set ening criteria,10 CFR 50.61 requirements other than  ;

those in As l recuired )aragraph

)y paragraph(b)(1)(are (b) 1), the notassessment applicable to San Onofre provided by thisUnits letter2will andbe3. '

upcated by Southern Califrenia Edison if there is a significant change in the '

projected values of RT or upon request for a change to the expiration date for operationofthefaciliYies.

The assessment provided by this submittal is applicable to both Units 2 and 3.

Therefore, this assessment will not be resubmitted in conjunction with the '

submittal of the update to the Unit 2 pressure-temperature limits by October 30, 1992.

If you have any questions regarding the results of this PTS assessment, please let me know.

Very truly yours, u-f l4R Enclosure cc: J. B. Martin, Regional Administrator, NRC Region V C. W. Caldwell, NRC Senior Resident Inspect <2r, San Onof re Units 1, 2 and 3 H. 8. Fields, NRR Project Manager, San Onofre Units 2 and 3

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~i Enclosure i

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RESPONSE TO 10.,_CFR 50.61 l

, E LS.SMRIZED THERMAL SHOCK fPTS) RULE RE0VIREMENTS SAN ONOFRE NQ.l. EAR GENERATINLSTATION. UNITS 2 AND,1 ]

l t 10 CFR 50.61 describes the Fracture Toughness Requirements for Protection  !

- Against Pressurized Thermal Shock Events. Specifically, paragraph (b)(1) l requires the licensee of each pressurized water reactor (PWR) for which an operating license has t,9en issued, to sub.it projected values of reference temperatures for pressuriad thermal shock (RTns) at the inner vessel surface of. reactor vessel materials from the time of submittal to the expiration of ,

the operating license. Paragraph (b)(2) provides the PTS screening criteria e and prescribes the method by which the values of RTns must be calculated.

- Thti PTS screening criteria are 270'F for plates, forgings, and axial weld '

materials, and 300*F_ for circumferential weld materials. For-the purpose of

- comparison with these criteria, the value of RTns for the reactor vessel must be- calculated as follows. The calculation must be made for each weld and plate, or forging,-in the reactor vessel beltline. -

Equation 1: RTns -

- I +M + ARTpg where:

-1 = initial reference temperature (rte 3) of the unirradiated material as defined in the ASME Code, Paragraph NB<2331 M = the margin to be added to account for uncertainties in the initial RTut, copper and nickel contents, fluence, and the calculational procedures -

ARTp13= the mean value of the shift in reference temperature caused by irradiation, calculated as' follows:

Equation 2: ARins. -- (CF) fm2s+ n lo m .

where:.

CF(6F ) = the chemistry factor, a function of nickel and copper content. CF is given in 10 CFR 50.61 Table 1- for welds and Table 2 for base metal (plates and

' forgings).

1

~J f =

the bestg estimate neutron fluence, in units of 10"n/cm (E greater than 1 MeV), at the clad-b!:se-metal interface on the inside surface of the vessel at the location where the material in question receives the highest. fluence for the period of service in

._ question.

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_p.

The current values of RT,33 as of September 11, 1992, are provided in the attached Table 1 for Unit 2 and Table 3 for Unit 3. TH 3rojected values of

, RTp73 at end of life (EOL) are provided in the attached Taales 2 and 4 for ,

Units 2 and 3, respectively. The Heat Affected Zone (HAZ) material is no l longer included in the surveillance program for beltline mnerials pursuant to ASTM Standard E185-92 (Reference 1). Therefore, the attached Tables 1 through 4 do not contain HA2 data, in addition, there are tu sets of Weld 9-203 data f or Unit 3 because two values of initial reference temperatures were calculated .iue to tS two sets of weld wire heat /flt:x lot number combinations used. The bases for the current and projected RTets values in Tables 1 through.4 are discussed below.

H Initial Reference Temperature and Margj.n The initial reference temperature (1) or RTm values for the unirradiated plate and weld materials, which were determined from Charpy V-notch (C,n) absorbed energy tests and drop weight tests in accordance with ASME Section 111, NS-2331,-are documented in Reference 1 and in Appendix A to Reference 2.

The initial reference temperatures are indicated in the column labeled

" Initial RTm *F" of the attached Tables 1 through 4. Because there was no measured RT m_value available for Weld 8-203, a generic RT m value of -56*F was used as required by Reference 3.

The margin (M) that is added to cover uncertainties in the values of RTm >

copper and nickel--content, fluence, and th* calculational procedures is 56af for volds and 34*f lfor base metals when measured values of initial reference temperatures are used as prescribed by Reference 3, paragraph (b)(2)(ii).

Because there was no measured value available for Weld 8-203, a margin of 66aF was used as required by R e rence 3.

Cooper and Nickel Cont nt The coper (Cu)-and nic 'l -(Ni) contents of vessel beltline materials'are

, h.ed 1 updated materia properties evsluated in response to Generic Letter i-9 -0., " Reactor Vesse Jtructural Integrity, 10 CFR 50.54(f)," Revision 1.

l The_opdated material properties are documentad in_ Reference I and attached as Appendix A to Reference 2. Because there was no measured value available for .

' ~ Cu and Ni _ contents of_ Weld 8-203, upper bound values were used as required

.. R i ro.ica 3.

h 91914ty. Factor l-Jh m .cy factor (CF) 1s a f unction of the copper and nickel content of the ves ' belt!ine materials. Cr is obtained from Reference 3, Tables 1 and 2 for Mdv 'od base metal (plates and forgings), respectively. Linear inte~ -1 H or is permitted to determine chemistry factors.

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3-Fluence The fluence (f) for all plates and welds (except Weld 8-203) is assumed to be the peak vessel fluence at the core centerline. The peak fluence is based on eore power distrioutions typical of low leakage checkerboard loading patterns 1d extended cycle lengths-(i.e., 24-imnth nominal fuel cycle) similar to Unit Cycle 5. The San Onofre Units 2 and 3 surveillance program described in eference 4 obtains fast neutrcn flux measurements from threshold detectors inserted into ehrt of the six irradiation capsules. The location and identification of the Units 2 and 3 beltline materials are shown in Reference

5. Figures 5.2-8 and 5.2-9 are attached for your information.

Except for Weld 8-203, upd' ted fluence projections in Tables 1 through 4 were obtained from the test re: Its and analysis of the surveillance capsule removed from Unit 3 at 4.33 EFPI ' Reference 6). A reduced fluence has been I credited for Weld 8-203 because the weld is not actually located in the L -beltline region (References 1 and 4).

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The e.urrent fluence is calculated to be the fluence on September 11, 1992 L scheduled submittal of this report. On this date, Units 2 and 3 will have experienced a vessel exposure of approximately 6.45 EFPY and 6.10 EFPY, respectively .The fluence for all plates and welds (except Weld 8-203) with a Unit 2 F ,1sure of 6.45 EFPY has been computed to be 0.933 X 10" n/cm2 . The equival # "'ence for a Unit 3 exposure of 6.10 EFPY has been computed to be v.889 ) -

• cm2 . The fluence for Weld 8-2032 with a Unit 2 exposure of 6.45 EFPY 'o - m :omputed to be 0.0254 X 10" n/cm . The equivalent fluence for

. Weld 6 - < >s;L a Uni 3 exposure of 6.10 EFPY has been computed to be 0.0242 X-10" m The pro p. a fluence is computed to be the fluence at 32 EFPY (E0L). The maximum design basis integrated fast neutron (E>l Mev) fluence at the vessel inside su_rface, including tolerance, has been computed to be 4.20 X 10" n/cm' for a vessel exposure of. 32 EFPY. A reduced fluence of 0.114 X 10" n/cm' at 32 EFPY has been credited for Weld 8-203.

Current and Proierled Values of RT ers Current values of- RTpis for San _Onofre Units 2 and 3 as of September 11, 1992, are given in Tables 1 and 3. By September 11, 1992, Units 2 and 3 are expected to-ha a experienced approxirataly 6.45 EFPY and 6.10 EFPY of operation, respectively. For these current exposures the maximum values of RTp3s for any beltline material'are 117.8af (Plates C-6404-1, 2, 3, and 4) for Unit.2 and 144.8'F'(Plate C-6802-1) for Unit 3.

Projected values of Units 2 and 3 RTpn in Tables 2 and 4 are given for 32

-EFPY;(EOL). The E0L dates-are February 16, 2022 and November 15, 2022 for Units-2 and 3. o 6ectively. 1hese dates are forty years after the issuance of the- respec unit operating licenses and are well beyond October 18, 2013, the curr u expiration date of both operating licenses. The maximum projected RTpys values at 32 EFPY for any beltline material are ~146.5aF (Plate l C-6404-5) for Unit 2- and 159.6af (Plate C-6802-1) for Unit 3.

t :

9 The Units 2 and 3 projected RTp13 values in this report and th9 projected RT,13 values previously reported in Reference 7 both meet the Reference 3 PTS screening criteria. However, they are different, and as required by 10 CFR 50.61(b)(1), the differeace must be justified. The difference in projected RTp13 values at 32 EFPY is due to the following:

o The final 10 CFR 50.61 PTS rule (Reference 4) primarily changed the method for calculating the shif t in RTpr3 and the " Margin" to be consistent with-Regulatory Guide 1.99, Revision 2, May 1988. The previous method of calculating thc shift in RTets and " Margin" was based on the original-10 CFR 50.61 rule and was inconsistent with Regulatory Guide 1.99.

o Updated fluence projections from the analysis of the survt.illance l capsule removed from Unit 3 in May 1990 at 4.33 EFPY were used l- (Reference 6). Previously submitted projections of RTers were based on the original calculated fluences.

-- o Updated material properties evaluated in resporse to Generic Letter 92-01, " Reactor Vessel Integrity,10 CFR 50.54(f)" were used (Reference 1).

The plate Cu and.Ni content were obtained by averaging two measurements made by ABB-Combustion Engineering when the plates were received and when _the. surveillance program was defined. The weld Cu and Ni content were obtained from the UFSAR except for Weld 9-203 which was obtained from a welding material certiTication (WMC). Where chemistry data was not available, 0.35% Cu and 1.00% Ni were assumed. Initial fracture toughness data (initial RTm) for phtes were obtained from the material certification reports (MCRc) and the baseline surveillance in

accordance with the:most recent version of ASME Section III, NB-2331.

Initial fracture toughness data for the beltline welds were obtained from the UFSAR and WMC. -Previous values did not consider all of the above.

Conclusion-The projected values of RTp13 for all beltline mn trials indicated in Tables 1 through 4 do not-exceed the PTS screening criteria in 10 CFR 50.61(b)(2). The PTS screening criteria are 270*F for plates', forgings, and axial weld -

materials-and 30CaF for circumferential weld materials. In the year 2022 the limiting materials in the reactor vessel beltline region are projected to be-intermediate shell plate C-6404-5 for Unit 2 with the RT ris equal to 146.Saf,

and intermediate shell piate C-6802-1 for Unit 3 with the RTp13 equal to ID.6*F. The results indicate that reactor vessel integrity will be maintained for the San Onofre Units 2 and 3 reactor vessels throughout their service lives. Hence, the actions required by Paragraphs (b)(3) and (b)(4) are not applicable.

1

Beferences

1. July 6,1992 letter from R.-M. Rosenblum (SCE) to Document Control Desk (SCE),

Subject:

Docket Nos. 50-361 and 50-362, Generic letter 92-01, Revision 1 " Reactor Vessel Structural Integrity,10 CFR 50.54(f), San Onofre Nuclear Generating Station, Units 2 and 3.

2. Calculation N 0220-026, Revision 1, " SONGS 2/3 Reference Temperature for Pressurized Thermal Shock," July 31, 1992.

3. 10 CFR 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Eyents," Final Rule Number 22300, Volume 56, No. 94, May 15, 1991.

4. Updated Final Safety Analysis Report, San Onofre Nuclear Generating Station, Units 2 and 3, Sections 5.2 and 5.3.

5. Updated Safety Analysis Report, San Onofre Nuclear Generating Station, Units 2 and 3, Figures 5.2-8 and 5.2-9.

6. Westinghouse R1 port WCAP-12920, " Analysis of the Southern California Edison Company San Onofre Unit 3 Reactor Vessel Surveillance Capsule Removed from the 97* Location," March 1991.

7. January 22, 1986 letter from M. O. Medford (SCE) to G. Knighton (NRC),

Subject:

Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3.

D i

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TABLE I CURRENT VALUES OF RTers FOR REACTOR VESSEL BELTLIfiE MATERIALS - 6.45 EFPY

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SAN Of10FRE UNIT 2 RTg73 = I or RT e7 + M + ARTrrs 6.45 EFPY i

Initial CF M Fluence 2 ARTns T RTns P1 ate or Cu Ni RTwy

(=F) (*F) (10" n/c:n ) (*F) (*F)

Component Seam No. (%) (%) (=F) 65 34 0.933 63.8 117.8 Inter. Shell C-6404-1 0.10 0.56 20 65 34 0.933 63.8 117.8 Inter. Shell C-6404-2 0.10 0.59 20 65 34 0.933 63.8 117.8 Inter. Shell C-6404-3 0.10 0.56 20 65 34 0.933 63.8 117.8 Lower Shell C-6404-4 0.10 0.62 20 75 34 0.933 73.6 117.6 Lower Shell C-6404-5 0.11 0.64 10 65 34 0.933 63.8 87.8 C-6404-6 0.10 0.58 -10 Lower Shell 56 0.933 40.2 36.2 Long. Weld 2-203A 0.03 0.90 -60 41 56 0.933 40.2 36.2 2-203B 0.03 0.91 -60 41 Long. Weld 56 0.933 40.2 36.2 2-203C 0.03 0.95 -60 41 Long. Wel.

40 56 0.933 39.2 45.2 3-203A 0.05 0.12 -50 Long. Weld 30 56 0.933 29.4 35.4 Long. Weld 3-703B 0.04 0.06 -50 42 56 0.933 41.2 47.2 3-203C 0.06 0.11 ~50 Long. Weld 260 66 0.0254 51.7- 61.7 8-203 0.31 1.00 -56 Circum. Weld 69 56 0.933 67.7 63.7 9-203 0.07 0.29 -60 Circum. Weld

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TABlF 2' PROJECTED VALUES OF RT ers FOR REACTOR VESSEL BELTLINE MATERIALS - 32 EFPY ,

SAN ON0FRE UNIT 2 RTers

- I or RT,c + .M +. ART p73 32 EFPY Initial Plate or Cu Ni RT,n CF M Fluence ART ns RTns Component Seam No. (%)- (%)- (*F) (*F) (*F) (10" n/cm*) '(*F) (*F)

Inter. Shell C-6404-1 0.10 0.56 20 65 34 4.20 88.9 142.9 Inter. Shell C-6404-2 0.10 0.59 20' 65 34 4.20 88.9 142.9 Inter. Shell C-6404-3 0.10 0.56 20 65 34 4.20 88.9 142.9 Lower Shell C-6404-4 0.10 0.62. 20 65 34 4.20 88.9 142.9 Lower Shell C-6404-5 0.11 0.64 10 75 34 4.20 102.5 146.5 Lower Shell C-6404-6 0.10 0.58 -10 65 34 4.20 88.9 112.9 Long. Weld 2-203A' O.03 0.90 -60 41 34 4.20 56.0 52.0 Long. Weld 2-203B 0.03 0.91 -60 41 34 4.20 56.0' 52.0 Long. Weld 2-203C 0.03 0.95 -60 41 34 4.20 56.0 52.0 Long. Weld 3-203A 0.05 0.12 -50 40 56 4.20 54.7 60.7 Long. Weld 3-2038 0.04 0.06 -50 30 56 4.20 41.0 47.0 Long. Weld 3-203C 0.06 0.11 -50 42 56 4.20 57.4 63.4 Circum. Weld 8-203 0.31 1.00 -56 260 66 0.114 115.4 125.4 Circum. Weld 9-203 0.07 0.29 -60 69 56 4.20 94.3 90.3 i

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TABLE 3 -

CURRENT VALUES OF RTp13 FOR REACTOR VESSEL BELTLINE MATERIALS - 6.10 EFPY SAN ON0FRE UNIT 3 RT ers = I or'RTy7 + M + ART pis 6.10 EFPY Plate or Cu Ni RTor CF M Fluence ART ns RT ns 2

Component Seam No.- (%) (%) (*F) (*F) (*F) (10" n/cm ) (*F) (*F)

Inter. Shell C-6802-1 0.06 0.58 75 37 34 0.889 35.8 '144.8 Inter. Shell C-6802'2-0.04 0.58 10 26 34 0.889 25.1 69.1-Inter. Shell C-6802-3 0.06 0.58 20 37 34 0.889 35.8 89.8 Lower Shell C-6802-4 0.05 0.56 10 31 34 0.889 30.0 74.0 Lower Shell C-6802-5 0.04 0.55 10 26 34 0.889 25.1 69.1 Lower Shell C-6802-6 0.06 0.62 20 37 34 0.889 35.8 89.8 Long. Weld 2-203A 0.05 1.00 -40 68 56 0.889 65.8 81.8 Long. Weld 2-2038 0.05 1.00 -40 68 56 0.889 65.8 81.8 Long. Weld 2-203C 0.05 1.00 -40 68 56 0.889 65.8 81.8 Long. Weld 3-203A 0.04 0.16 -70 39 56 0.889 37.7 23.7 Long. Weld 3-2038 0.04 0.16 -70 39 56 0.889 37.7 23.7 Long. Weld 3-203C 0.04 0.16 -70 39 56 0.889 37.7 23.7 Circum. Weld 8-203 0.35 1.00 -56 272 66 0.0242 52.5 62.5

' Circum. Weld 9-203 0.06 0.04 -60 34 56 0.889 32.9 28.9

' Circum. Weld 9-203 0.05- 0.04 -50 31 56 l 0.889 30.0 36.0

• For Unit 3 two sets of values were calculated for Weld 9-203 corresponding to the two sets of weld wire heat / flux lot number combinations used.

-TABLE 4  :-

PROJECTED VALUES OF RT FTs FOR REACTOR VESSEL BELTLINE MATERIALS - 32 EFPY SAN ON0FRE UNIT 3 RTris =

I or RTwT.-+. M + ARTris 32 EFPY Initial Plate'or Cu Ni- RTo r CF M Fluence ARTns RT ns Component Seam No. (%) (%) (*F) -(*F) (*F) 2 (10" n/cm ) (.p) (.p)

Inter. Shell C-6802-1 0.06 0.58 75 37' 34 4.20 50.6 159.6 Inter. Shell C-6802-2 0.04 0.58 10 26 34 4.20 35.5 79.5 Inter. Shell C-6802-3 0.06 0.57 20 37 34- 4.20 50.6 104.6 Lower Shell C-6802-4 0.05 0.58 10 31 34 4.20 42.4 86.4 Lower Shell C-6802-5 0.04 0.52 10 26 34 4.20 35.5 79.5 Lower Shell C-6802-6 0.06 0.65 20 37 34 4.20 50.6 104.6 Long. Weld 2-203A 0.05 1.00 -40 68 56 4.20 93.0 109.0 Long. Weld 2-203B 0.05 1.00 -40 68 56 4.20 93.0 '109.0 Long. Weld 2-203C 0.05' l.00 -40 68 56 4.20 93.0 109.0 Long. Weld 3-203A 0.04 0.16 -70 39 56 4.20 53.3 '39.3 Long. Weld 3-2038 0.04 0.16 -70 39 56 4.20 53.3 39.3 Long. Weld 3-203C 0.04 0.16 -70 39 56 4.20 53.3 39.3 Circum. Weld 8-203 0.35 1.00 -56 272 66 0.114 120.8 130.8

• Circum. Weld 9-203 0.06 0.04 -60 34 56 4.20 46.5 42.5

• Circum. Weld 9-203 0.05 0.04 -50 31 56 4.20 42.4 48.4

• For Unit 3 two sets of values were calculated for Weld 9-203 corresponding to the two sets of weld wire heat / flux lot number combinations used.

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