ML20065G425
| ML20065G425 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 02/28/1994 |
| From: | Malone M, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20065G421 | List: |
| References | |
| WCAP-13932, NUDOCS 9404120271 | |
| Download: ML20065G425 (22) | |
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WESTINGHOUSE CLASS 3 (Non Proprietary) WCAP-13932 i EVALUATION OF PRESSURIZED THERMAL SHOCK FOR THE VOOTLE UNIT 1 REACTOR VESSEL , i M. J. Malone l Febmary 1994 1 Work Performed Under Shop Order'GTXP-108 Prepared by Westinghouse Electric Corporation for the Georgia Power Company r) n ' Approved by: At aks W T. . Mey , Managerk Structural Reliability & Plant Life Optimization l t WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Tecimology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
@ 1994 Westinghouse Electric Corporation All Rights Reserved 1
d
i 4 1h
- PREFACE
- This report has been technically reviewed and verified by
J. M. Chicots . M ((be b 9' i
TABLE OF CONTENTS LIST OF TABLES .............................. .........................ili I LIST OF FIG URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . 111
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............... l' 2.0 PRESSURIZED THERMAL SHOCK . . . . . . . . . . . . . . . . ..................... 2 3.0 METHOD FOR CALCULATION OF RT,,n . . . . . . . .... ................... 4 4.0 ' VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES . . . . . . . . . . .... 5 5.0 NEUTRON FLUENCE VALUES ...................................... 11 6.0 DETERMINATION OF RTen VALUES FOR ALL BELTLINE REGION MATERIALS .......... ........ ...... ................. ........ 12
7.0 CONCLUSION
S . . . . . . . . . . ........... .................. .. . ... 17
8.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 18 .
l i 1 i l i 1 l 1i 1
J LIST OF TABLES { Table 1 Calculation of Average Cu and Ni Weight % for Intermediate Shell Plates i Using All Previous Vogtle Unit 1 Chemistry Test Results . . . . . . . . . . . . . . . . . . 7 , Table 2 Calculation of Average Cu and Ni Weight % for Lower Shell Plates Using All Previous Vogtle Unit 1 Chemistry Test Results . . . . . . . . . . . . . . . . . . . . . . . . . , 8 Table 3 Calculation of Average Cu and Ni Weight % for Weld Metal Using All Previous Vogtle Unit 1 Chemistry Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . 9 i Table 4 Vogtle Unit 1 Reactor Vessel Beltline Region Material Properties . ........ 10 $ Table 5 Neutron Exposure Projections at Key Locations on the Vogtle Unit 1 Pressure '
- Vessel Clad / Base Metal Interface for 4.64. 32 and 48 EFPY . . . . . . . . . . . . . . . I1 J
Table 5 Calculation of Chemistry Factors Using Vogtle Unit 1 Surveillance Capsule ' Dam .................................... ...... .......... 13 ? 1 Table 7 RTen Values for Vogtle Unit I for 4.64 EFPY . . . . . . . . . . . . . . . . 14
........ I 1
l Table 8 RTpn Values for Vogtle Unit I for 32 EFPY . . . . . . . . . . . 15 l l l Table 9 RTpn Values for Vogtle Unit I for 48 EFPY . . . . . . . . . . . . . ...... ..... 16 LIST OF FIGURES Figure 1. Identification and location of Beltline Region Materials for the Vogtle Unit 1 i Reactor Vessel . . . . . . ............. .......................... 6 Figure 2. RTpn Versus Fluence Curves For Vogtle Unit 1 Limiting Material - Intermediate Shell Plate B8805-2 . . . . . . . . . ................ ..... . 17 iii
1.0 INTRODUCTION
A limiting condition on reactor vessel integrity known as Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a Loss-Of-Coolant-Accident (LOCA) or a steam line break. Such transients may challenge the integrity of a reactor vessel under the following conditions: severe overcooling of the inside surface of the vessel wall at relatively high pressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect in the vessel wall. In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on PTS. It established screening criteria on pressurized water reactor (PWR) vessel embrittlement as measured by the reference nil-ductility transition temperature, termed RTru l ". RT en screening values were set for beltline axial welds, forgings or plates and for beltline circumferential weld seams for operation to end of plant license. The screening criteria were determined using conservative fracture mechanics analysis techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end of license. The NRC recently amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlement. The revised FTS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14,1991m This amendment makes the procedure for calculating RTru values consistent with the methods given in Regulatory Guide 1.99, Revision 2m, The purpose of this report is to determine the RTen values for the Vogtle Unit I reactor vessel to address the revised FTS Rule. Section 2 discusses the Rule and its requirements. Section 3 provides the methodolo-r for calculating RTen. Section 4 provides the reactor vessel beltline region material properties for the Vogtle Unit I reactor vessel. The neutron fluence values used in this analysis are presented in Section 5. The results of the RTrn calculations are presented in Section 6. The conclusion that all PTS screening criteria are satisfied at 48 EFPY and references for the PTS evalu2 tion follow in Sections 7 and 8, respectively. 1
l 2.0 PRESSURIZED THERMAL SIIOCK The FTS Rule requires that the FTS submittal be updated whenever there are changes in core loadings, smveillance measurements or other information that indicates a significant change in projected RTen values. The rule outlines regulations to address the potential for FTS events on pressurized water reactor vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (NRC). FTS events have been shown from operating experience to be transients that result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may result in the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel. The mle establishes the following requirements for all domestic, operating PWRs: All plants must submit projected values of RTers for reactor vessel beltline materials by giving values for time of submittal, the expiration date of the operating license, and the projected expiration date if a change in the operating license or renewal has been requested. This assessment must be submitted within six months after the effective date of this Rule if the value of RT rn for any material is projected to exceed the screening criteria. Otherwise, it must be submitted with the next update of the pressure-temperature limits, or the next reactor vessel surveillance capsule report, or within 5 years from the effective date of June 14,1991, whichever comes first. These values must be calculated based on the methodology specified in the FTS Rule. The submittal must include the following:
)
i l
- 1) the bases for the projection (including any assumptions regarding core loading patterns), and
- 2) copper and nickel content and fluence values used in the calculations for each beltline material. (If these values differ from those previously submitted to the NRC, justification must be provided.)
2
l l l The RTm (measure of fracture resistance) screening criteria for the reactor vessel beltline region is: 270 F for plates, forgings, axial welds; and 300 F for circumferential weld materials. The equations that ,"ust be used to calculate the RTm values for each weld, plate or ; forging in the reactor vessel beltline are specified (see Section 3), I All values of RTm must be verified to be bounding values for the specific reactor vessel. In doing this, each plant should consider plant-specific information that could affect the level of embrittlement. Plant-specific FTS safety analyses are required before a plant is within 3 years of reaching the screening criteria, including analyses of alternatives to minimize the PTS concern. NRC approval for operation beyond the screening criteria is required. 1
}
l I l l 3
3.0 METIIOD FOR CALCULATION OF RTen In the FTS Rule, the NRC Staff has selected a consetvative and uniform method for determining plant-specific values of RTen at a given time. For the purpose of comparison with the screening criteria, the value of RTen for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as follows: RTen = I + M + ARTen, where ARTru = (CF)
- FF I= Initial reference temperature (RTuo7) in F of the unitradiated material M= Margin to be added to cover uncertainties in the values of initial RTuor, copper and nickel contents, fluence and calculational procedures in F.
M = 66 F for welds and 48 F for base metal if generic values of I are used. M = 56 *F for welds and 34 F for base metal if measured values of I are used. FF = fluence factor = f n23.aio win, where f = Neutron fluence (E>1.0 MeV at the clad / base metal interface), divided by 10" n/cm 2 CF = Chemistry factor in *F from the tabled 2) for welds and base metals (plates and forgings). If plant-specific surveillance data has been deemed credible per Reg. Guide 1.99, Rev. 2 and is significant, it may be considered in the calculation of the chemistry factor. 4
4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the Vogtle Unit I vessel was performed. The beltline region is defined by the r PTS Rule:) 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 irradiation darnage to be considered in the selection of the most limiting material with regard to radiation damage." Figure 1 identifies and indicates the location of all beltline region materials for the Vogtle Unit I reactor vessel. Material property values were obtained from material test certifications from the original fabrication as t well as the additional material chemistry tests performed as part of the surveillance capsule testing I m program . The average copper and nickel values were calculated for each of the beltline region materials using all of the available material chemistry information as shown in Tables 1 through 3. A summary of the pertinent chemical and mechanical properties of the beltline region forgings and weld materials of the Vogtle Unit ! reactor vessel are given in Table 4. All of the measured values of initial RTsor (1) are also presented in Table 4. 1 1 5
i 5 N !
- i x
5 w 2 i N 2 0 z 0 l B8805-2 90 101-124B
! 4 B8805-3 , 1 A 300 e
W ) .-j 0 180 0 i
, 3 )
E
; 5 101-124A 4 30 5
Core l B8805-1 2700 101-124C !' 4 " 101-171 101-142A g0 0 B8606-1 i D %
! = so
- 2
- m 0 0 1 3 0 -
180 i R i a 0 v 101-142B i B8606-3 1 i 0 B8606-2 101-142C 270 1 d i
) ! Figure 1. Identification and location of Beltline Region Materials for the Vogtle Unit 1 j Reactor VesselR 4
l 6 1 1 i l { i l
Table 1 Calculation of Average Cu and Ni Weight % for Intermediate Shell Plates Using All Previous Vogtle Unit 1 Chemistry Test Results Inter. Shell Inter. Shell Inter. Shell Plate B8805-1 Plate B8805-2 Plate B8805 3 Reference Cu (wt%) Ni(wt%) Cu (w'%) Ni(wt%) Cu (wt%) Ni(wt%) Material Cert. Repod83 0.09 0.60 Material Cert. Repod'l 0.08 0.60 t Chemical Analysis'i 0.08 0.59 Surveillance Program"1 0.08") 0.59'" Material Cert. Report" l 0.09 0.64 Material Cert. Repon" 1 0.08 0.60 . Chemical Analysis"" 0.08 0.59 Surveillance Program"1 0.08r2) 0.59'2) Material Cert. Reportn2: . 0.07 0.60 Material Cert. Reportn2) 0.07 0.61-Chemical Analysis"'I 0.06 0.60 Surveillance Program (51 0.058 0.61 Surveillance Program"3 0.068) 0.6(15) Capsule U Reportl 0.053 0.586 Capsule U Report"1 0.06') 0.60) Capsule U Report'l ! 0.058") 0.61") Capsule Y Report") 0.061 0.584 Average 0.083 0.597 0.083 0.61 0.062 0.598 (1) Not used in average calculation since same values as those from Reference 9; reported only for completeness. (2) Not used in average calculation since same values as those from Reference 11; reported only for completeness. (3) Not used in average calculation since same values as those from Reference 13; reported only for completeness. (4) Not used in average calculation since same values as those from Reference 5; repotted only for completeness. 7
Table 2 Calculation of Average Cu and Ni Weight % for Lower Shell Plates Using All Previous Vogtle Unit 1 Chemistry Test Results Lower Shell 1.ower Shell Lower Shell Plate B8606-1 Plate B8606-2 Plate B8606-3 Reference Cu (wt%) Ni(wt%) Cu (wt%) Ni(wt%) Cu (wt%) Ni(wt%) Material Cen. Repon"'i 0.% 0.61 Material Cen. Repon"'I 0.05 0.58 Chemical Analysisos) 0.05 0.59 Smveillance Program"i 0.05") 0.59'" Material Cert. Report"'i 0.07 0.64 Material Cert. Report"'I 0.05 0.58 Chemical Analysis"'I 0.05 0.58 Surveillance Program"1 0.05
- 0.58")
Material Cert. Repon"81 0.07 0.60 Material Cert. Report"'I 0.07 0.63 Chemical Analysis"'i 0.06 0.64 Surveillance Program"1 0.06 O.64") Average 0.053 0.593 0.057 0.60 0.067 0.623 (1) Not used in average calculation since same values as those frorn Reference 15; reported only for completeness. (2) Not used in average calculation since same values as those from Reference 17: reported only for completeness. (3) Not used in average calculation since same values as those from Reference 19; reported only for completeness. 8 L--
l, Table 3 ; Calculanan of Average Ca and Ni Weight % for Weld Metal Using All Previous Vogtle Unit 1 Chemistry Test Results Reference Weld Metal'" Cu (wt%) Ni (wt%) Stuveillance Programtsj 0.03 -- - Surveillance Programl51 0.04 0.10 Surveillance Programi 'l 0.037 0.10 Capsule U Report
- 0.035 0.091 Capsulc Y Report t4j 0.048 0.101 Capsule Y Report W 0.04 0.117 Capsule Y ReportW 0.041
, 0.105 Mas rial Qualificationt2oj 0.04r2) 0.10r23 Material Qualificationt2n 0.03c2) ..m Average 0.039 0.102 (1) The core region (beltline) welds are considered to include the intermediate and lower shcIl plate longitudinal seams and the joinkg intermediate to lower shell girth seam. All core region (beltline) welds were fabricated using Weld Wire Heat No. 83653. Linde 0091 Flux. Iot No. 3536.
(2) Not used in average calculation since same values as those from Reference 5: reported only for completeness. 9 l l I l
)
Table 4 Vogtle Unit 1 Reactor Vessel Beltline Region Material Properties Material Description Cu (%) ") Ni (%) ") I (*F)l'1 *
~
Intermediate Shell Plate B8805-1 0.083 0.597 0 Intermediate Shell Plate B8805-2 0.083 0.610 20 i Intermediate Shell Plate B8805-3 0.062 0.598 30 Lower Shell Pla'e B8806-1 0.053 0.593 20 ! Lower Shell Plate B8806-2 0.057 0.600 20 Iower Shell Plate B8806-3 0.067 0.623 10 W eld M etal ") 0.039 0.102 80 (a) Average values of copper and nickel as indicated in Tables 1 through 3 on the preceding pages. (b) Initial RTm7r values were estimated per U.S. NRC Standard Review Plan [22]. The initial RTm7r values for the plates and welds are measured values. (c) The core region (beltline) welds are considered to include the intermediate and lower shell plate longitudinal seams and the joining intermediate to lower shell girth seam. All core region (beltline) welds were fabricated using Weld Wire Heat No. 83653, Linde 0091 Flux, Iet No. 3536. 6 10
5.0 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>l.0 MeV) at the inner surface of the Vogtle Unit I reactor vessel is shown in Table 5. These values were projected using the results of the Capsule Y radiation surveillance program W The limiting RTp13 calculations were performed using the peak fluence value, which occurs at the 25 azimuth in the Vogtle Unit I reactor vessel. Table 5 Neutron Exposure Projections
- at Key Locations on the Vogtle Unit 1 Pressure Vessel Clad / Base Metal Interface EFPY 0 15 25 30* 35 45 i
4.64 0.I802 0.2678 0.3155 0.2000 0.2622 0.2942 l i 32 1.243 1.847 2.176 1.379 1.809 2.029 i 48* 1.863 2.769 3.262 2.068 2.711 3.041 2 (1) Fluence in 10" n/cm (E>1.0 MeV) (2) Fluence values for 48 EFPY were calculated using the following equation: Fluence = Present fluence + (48 - Present EFPY)
- Flux
- 3.1536E+7 sec/EFPY where, Present fluence = fluence at 4.64 FEPY Present EFPY = 4.64 EFPY Flux = Present flux rate (4.6% EFPY) l l
l 11 L j
6.0 DETERMINATION OF RTen VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTen values were generated for all beltline region materials of the Vogtle Unit I reactor vessel for fluence values at the present time (4.64 EFPY per Capsule Y analysis), end of license (32 EFPY) and 48 EFPY. The FTS Rule requires that each plant assess the RTen values based on plant specific surveillance capsule data whenever: Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2, and RTen values change significantly. (Changes to RTrn values are considered significant I if the value determined with RTen equations (1) and (2), or that using capsule data, or l both, exceed the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.) Although the RT en value changes are not significant for Vogtle Unit 1, plant specific surveillance capsule data for intermediate shell plate B8805-3 and the weld metal is provided because of the following reasons:
- 1) There have been two capsules removed from the reactor vessel, and the data is deemed credible per Regulatory Guide 1.99, Revision 2.
- 2) The surveillance capsule materials are representative of the actual vessel plates and f circumferential and longitudinal weld materials.
{ l The chemistry factors for intermediate shell plate B8805-3 and the weld metal were calculated using j the surveillance capsule data as shown in Table 6. The chemistry factors were also calculated using <
\
Table 2 from 10 CFR 50.61m ' Tables 7 through 9 provide a summary of the RTe n values for all beltline region materials for 4.64 EFPY,32 EFPY, and 48 EFPY, respectively, using the ITS Rule. 1 12
Table 6 Calculation of Chemistry Factors Using Vogtle Unit i Surveillance Capsule DataW l l Material Capsule Fluence FF ARTun[) FF*ARTyn1 FF2 2 (n/cm . E>l.0 MeV{ ( F) ( F) Inter. Shell Plate U 3.437 x 10"' O.706 15 10.585 0.498 B8805-3 (Longitudinal) Y 1.242 x 10 l.060 40 42.4 1.124 Inter. Shell Plate U 3.437 x 10'"') 0.706 0 0 0.498 B8805-3 (Transverse) Y 1.242 x 10 l.060 20 21.2 1.124 Sum: 74.185 3.244 Chemistry Factor = 74.185 + 3.244 = 22.9 F Weld Metal U 3.437 x 10 l "*) 0.706 15 10.585 0.498 Y 1.242 x 10 l.060 0 0 1.124 Sum: 10.585 1.622 Chemistry Factor = 10.585 + 1.622 = 6.5 F (a) Original fluence value has been revised to be based on actual cycle burn up instead of predicted (original value = 3.41 x 10 from capsule U report, WCAP-12256). (b) The weld metal ARTyn7 values contain no adjustment ratio per Regulatory Guide 1.99, Rev. 2 Position 2.1 since the surveillance weld is identical to that used in the core region of the longitudinal seams and the girth seam weld joining the intermediate and lower shells. 13
Table 7 i RT,,n Values for Vogtle Unit I for 4.64 EFPY ! CF Surface Fluence Hi FF" ARTsor I M RT,3 ( F) (n/cm2 , E>l.0 MeV) -( F) ('F) ( F) (*F) Material (CF x FF) Inter. Shell Plate 53.1 3.155 x 10'8 0.6833 36.3 0 34 70.3 B8805-1 l Inter. Shell Plate 53.1 3.155 x 10 O.6833 36.3 20 34 90.3 B8805-2 Inter Shell Plate 38.4 3.155 x 10'8 0.6833 26.2 30 34 90.2 B8805-3 Inter. Shell Plate B8805-3 22.9 3.155 x 10 O.6833 15.6 30 34 79.6 Using S/C datar2) i Inter. Shell Plate 32.8 3.155 x 10 O.6833 22.4 20 34 76.4 B8606-1 Inter. Shell Plate 35.2 3.155 x 10 O.6833 24.1 20 34 78.1 B8606-2 Inter. Shell Plate 41.9 3.155 x 10 28 0.6833 28.6 10 34 72.6 B8606-3 Cire. Weld 33.2 3.155 x 10 O.6833 22.7 -80 56 -1.3 l 101-171 W eld M etal 6.5 3.155 x 10 O.6833 4.4 -80 56 -19.6 > Using S/C data <2) Long. Weld 33.2 1.802 x 10 O.5448 18.1 -80 56 -5.9 101-124A Long. Weld 33.2 2.0 x 10 ' O.5694 18.9 -80 56 -5.1 101-124B j Long. Weld 33.2 2.0 x 10'8 0.5694 18.9 -80 56 -5.1 101-124C Long. Weld 33.2 2.0 x 10 O.5694 18.9 -80 56 5.1 101-142A Long. Weld 33.2 1.802 x 10 ' O.5448 18.1 -80 56 -5.9 101-142B Long. Weld 33.2 2.0 x 10 O.5694 18.9 -80 56 -5.1 101-142C , l 2 (1) FF (Fluence Factor) based on inner surface neutron fluence (n/cm E>1.0 MeV). (2) Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data. l (3) Peak fluence factor which represents most limiting case for weld metal. 14
ll I Table 8 RT,,a Values for Vogtle Unit I for 32 EFPY l CF Surface Fluence W FF'" ARTy n, 1 M RT,,n ( F) (n/cm2 , E>1.0 MeV) ( F) ( F) ( F) ( F) Material (CF x FF) Inter. Shell Plate 53.1 2.176 x 10 l.2110 64.3 0 34 98.3 B88051 Inter. Shell Plate 53.1 2.176 x 10 ' 2 1.2110 64.3 20 34 118.3 B8805-2 Inter. Shell Plate 38.4 2.176 x 10 ' l.2110 46.5 30 34 110.5 B8805-3 Inter. Shell Plate B8805-3 22.9 2.176 x 10" 1.2110 27.7 30 34 91.7 Using S/C datar23 Inter. Shell Plate 32.8 2.176 x 10" 1.2110 39.7 20 34 93.7 B8606-1 Inter. Shell Plate 35.2 2.176 x 10" 1.2110 42.6 20 34 96.6 B8606-2 Inter. Shell Plate 41.9 2.176 x 10" 1.2110 50.7 10 34 94.7 B8606-3 Cire. Weld 33.2 2.176 x 10" 1.2110 40.2 -80 56 16.2 101-171 Weld Metal 6.5 2.176 x 10* 1.2110") 7.9 -80 56 -16.1 Using S/C data r2> Long. Weld 33.2 1.243 x 10" 1.0606 35.2 -80 56 11.2 101 124A Long. Weld 33.2 1.379 x 10" 1.0893 36.2 -80 56 12.2 101-124B Long. Weld 33.2 1.379 x 10" 1.0893 36.2 -80 56 12.2 101-124C Long. Weld 33.2 1.379 x 10" 1.0893 36.2 -80 56 12.2 101-142A Long. Weld 33.2 1.243 x 10" 1.0606 35.2 -80 56 11.2 101-142B Long. Weld 33.2 1.379 x 10" 1.0893 36.2 -80 56 12.2 101-142C (1) FF (Fluence Factor) based on inner surface neutron fluence (n/cm2 , E>1.0 MeV). (2) Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data. (3) Peak fluence factor which represents most limiting case for weld metal. 15 a-____-___--_. -
1 Table 9 RTrTs Values for Vogtle Unit I for 48 EFPY CF Surface Fluence '" FF'" ARTsor I M RTers ( F) (n/cm2 , E>1.0 MeV) ( F) ( F) ( F) ( F) Material (CF x FF) Inter. Shell Plate 53.1 3.262 x 10 l.3104 69.6 0 34 103.6 B8805-1 Inter. Shell Plate 53.1 3.262 x 10 l.3104 69.6 20 34 123.6 B8805-2 Inter. Shell Plate 38.4 3.262 x 10 l.3104 50.3 30 34 114.3 B8805-3 Inter. Shell Plate B8805-3 22.9 3.262 x 10 l.3104 30.0 30 34 94.0 Using S/C data
- Inter. Shell Plate 32.8 3.262 x 10 l.3104 43.0 20 34 97.0 B8606-1 Inter. Shell Plate 35.2 3.262 x 10 ' l.3104 46.1 20 34 100.1 B8606-2 Inter. Shell Plate 41.9 3.262 x 10 l.3104 54.9 10 34 98.9 B8606-3 Cire. Weld 33.2 3.262 x 10 l.3104 43.5 -80 56 19.5 101-171 Weld Metal 6.5 3.262 x 10 l.3104 8.5 -80 56 -15.5 Using S/C data
- Long. Weld 33.2 1.863 x 10 l.1705 38.9 -80 56 14.9 101-124A Long. Weld 33.2 2.068 x 10 l.1978 39.8 -80 56 15.8 101-124B Long. Weld 33.2 2.068 x 10 l.1978 39.8 -80 56 15.8 101-124C Long. Weld 33.2 2.068 x 10 l.1978 39.8 -80 56 15.8 101-142A Long. Weld 33.2 1.863 x 10 l.1705 38.9 -80 56 14.9 101-142B ,
l Long. Weld 33.2 2.068 x 10 ' l.1978 39.8 -80 56 15.8 101-142C (1) Fluence = present fluence + (48 - present EFPY)
- present flux
- 3.1536 x 10'sec/EFPY.
2 (2) FF (Fluence Factor) based on inner surface neutron fluence (n/cm . E>l.0 MeV). (3) Numbers were calculated using a chemistry factor (CF) based on surveillance capsule data. (4) Peak fluence factor which represents most limiting case for weld metal. 16
7.0 CONCLUSION
S As shown in Tables 7 through 9, all RTm values remain below the NRC screening values for PTS using fluence values for the present time (4.64 EFPY), projected fluence values for the end of license (32 EFPY), and 48 EFPY. A plot of the RTpn values versus fluence shown in Figure 2 illustrates the available margin for the most limiting material in the Vogtle Unit I reactor vessel beltline region, Intermediate Shell Plate B8805-2. 350 300 - SCREENING CRITERIA 250 - f._. 200 vs F 150 E
- A~5'-
100 -
+ '
e 4.64 EFPY 50 - A 32.0 EFPY a 48.0 EFPY 0 1 E + 18 2E+18 3E + 18 SE + 18 1E+19 2E + 19 3E + 19 SE + 19 1E+20 2 FLUENCE (neutrons /cm ) INTERMEDIATE SHELL PLATE B8805-2 Figure 2. RTm Versus Fluence Curves For Vogtle Unit 1 Limiting Material - Intermediate Shell Plate B8805 2 17
- __ __ _ _ ___ _ __ _ ______o
3
8.0 REFERENCES
[1] 10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23,1985. [2] 10CFR Part 50.61, " Fracture Toughness Requirements for Protection Against Pressurized l Thermal Shock Events," May 15,1991. (PTS Rule) [3] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988. [4] WCAP-13931, " Analysis of Capsule Y From The Georgia Power Company Vogtle Unit 1 . Reactor Vessel Radiation Surveillance Program", M. J. Malone, January 1994. [5] WCAP-11011, " Georgia Power Company Alvin W. Vogtle Unit No.1 Reactor Vessel Radiation Surveillance Program", L. R. Singer, February 1986. [6] WCAP-12256, " Analysis of Capsule U From The Georgia Power Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program" S. E. Yanichko, et al., May 1989. [7] Combustion Engineering Inc. Drawing No. 8971-161-003, "As Built Location of Weld Seams Vessel and Closure Head Westinghouse Electric Corp.173" ID PWR", Rev. 01, 8/8/78. [8] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification Report, Contract No. 8971, Job No. 708124-001, Code No. B-8805-1, Heat No. C0613-1, dated October 5,1972 and Lukens Steel Company Test Certificate, Mill Order No. 10554-1, l dated April 25,1972. [9] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from W.A. ! House, Lab No. P18955, Inter. Shell Code B88051, dated March 22,1979. { l [10] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification l Repon, Contract No. 8971, Job No. 708124-003, Code No. B-8805 2, Heat No. C0613-2, dated October 5,1972 and Lukens Steel Company Test Certificate, Mill Order No. 10554-1, dated May 9,1972. 18 l
7 l l l [11] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from W.A. l House, Lab No. P185%, Inter, Shell Code B8805-2, dated March 22,1979. [12] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification Report, Contract No. 8971, Job No. 708124-005, Code No. B-8805-3, Heat No. C0623-1, dated October 5,1972 and Lukens Steel Company Test Certificate, Mill Order No. 10554-1, dated May 16,1972. [13] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from W.A. House, Lab No. P18957 Inter, Shell Code B8805-3, dated March 22,1979. [14] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification Report, Revision 1, Contract No. 8971, Job No. 708142-007, Code No. B-8606-1. Heat No. C2146-1, dated March 29,1974 and Lukens Steel Company Test Certificate, Mill Order No. 12517 2, dated March 23,1973. [15] Combustion Engineering Power Systems Interoffice Correspondence to A.B. Harper from W.A. House, Lab No. P15703, Code B8606-1, dated October 30,1978. [16] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification Report, Revision 1, Contract No. 8971, Job No. 708142-013, Code No. B 8606-2, Heat No. C2146 2, dated March 29,1974 and Lukens Steel Company Test Certificate, Mill Order No. 12517-2, dated March 23, 1973. [17] Combustion Engineering Power Systems Interoffice Correspondence to A.B. liarper from W.A. House, Lab No. P13986, Code B8606-2, dated October 30,1978. [18] Combustion Engineering Inc., Metallurgical Research and Development Materials Certification { Report, Revision 1, Contract No. 8971, Job No. 708142-011,0;de No. B-8606-3, Heat No. l C2085-2, dated March 29,1974 and Lukens Steel Company Test Certificate, Mill Order No. 12517 1, dated March 30,1973. [19] Combustion Engineering Power Systems luteroffice Correspondence to A.B. Harper from W.A. House, Lab No. P15704, Code B8606-3, Jated October 30,1978. 19 I l _ )
T [20] Combustion Engineering Power Systems Welding Material Qualification To Requirements of ASME Section III, Job No. D32255, Project Number 960009, dated November 6,1972. [21] Combustion Engineering Power Systems Interoffice Correspondence from P. C. Kiefer, Qualification Code Gl.43, Job No. D32255, dated November 2,1972. [22] " Fracture Toughness Requirements", Branch Technical Position MTEB 5 2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, Rev.1 July 1981. i l 20
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