ML20154K804
| ML20154K804 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 09/30/1998 |
| From: | Christopher Boyd, Laubham T, Spragg S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20154K776 | List: |
| References | |
| WCAP-15103, WCAP-15103-R, WCAP-15103-R00, NUDOCS 9810190029 | |
| Download: ML20154K804 (22) | |
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15103 Evaluation of Pressurized Thermal Shock for V. C. Summer Unit 1 S. Spragg T.J. Laubham September 1998 Work Performed Under Shop Order STMP-108 Prepared by the Westinghouse Electric Company for the South Carolina Electric and Gas Company Approved: *
- C. H. Boyd, Manager Equipment & Materials Technology Approved: / Pot 9/28/38 '
D. M. Trombola, Manager Mechanical Systems Integration Westinghouse Electric Company Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355
@l998 Westinghouse Electric Company All Rights Reserved
til TABLE OF CONTENTS LIST OF TABLES.. . .
. .iv LIST OF FIGURES.. . .. . . .
.. . . .. v PREFACE ,
. . . . . . .. si EXECUTIVE
SUMMARY
.. sii 1 INTRODUCTION. .. . .
. . 1-1 2 PRESSURIZED THERMAL SHOCK RULE.. .
.2-1 3
METHOD FOR CALCULATION OF RTris. . . .. .
. . 3-1 4
VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES.. . . . 4-1 5 NEUTRON FLUENCE VALUES.. . . . .. . . 5-1 6 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIALS. .6-1 7 CONCLUSION . .
. 7-1 8 REFERENCES. . . .. , , . . . 8-1 Revision 0
iv LIST OFTABLES Table 1 V. C. Summer Unit 1 Reactor Vessel Beltline Unitradiated Material Properties. .4-3 Table 2 Fluence (E > 1.0 MeV) on Pressure Vessel Clad / Base Interface for V. C. Summer Unit I at 32 (EOL) and 48 (Life Extension) EFPY.. . 5-1 Table 3 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR 50.61. .6-2 Table 4 Calculation of Chemistry Factors using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1. .6-3 Table 5 RTrrs Calculation for V. C. Summer Unit i Beltline Region Materials at EOL(32 EFPY). . . .
.. .6-4 Table 6 RTrts Calculation for V. C. Summer Unit 1 Beltline Region Materials at Life Extension (48 EFPY) . .. .. . . .6-5 Revision 0
V LIST OF FIGURES Figure 1 Identification and Location of Beltline Region Materials for the V. C. Summer Unit 1 Reactor Vessel. . .- - .
.4-2 Revision 0
vi PREFACE This report has been technically reviewed and verified by: Reviewer: Ed Terek Revision 0
vii l EXECUTIVE
SUMMARY
The purpose of this report is to determine the RTns values for the V. C. Sununer Unit I reactor vessel beltline based upon the results of the Surveillance Capsule W evaluation. The conclusion of this report is that all the beltline materials in the V. C. Summer Unit I reactor vessel have RTns values below the screening criteria of 270 F for plates, forgings or longitudinal welds and 300 F for circumferential welds at EOL (32 EFPY) and life extension (48 EFPY). i Revision 0
1-1 1 INTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A 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 produce a flaw or cause the propagation of a flaw postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel. The purpose of this report is to determine the RTersvalues for the V. C. Summer Unit I reactor vessel using the results of the surveillance Capsule W evaluation. Section 2.0 discusses the PTS Rule and its requirements. Section 3.0 provides the methodology for calculating RTrrs. Section 4.0 pro 5 ides the reactor vessel beltline region material properties for the V. C. Summer Unit I reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0 and were obtained from Section 6 of WCAP-1510lW The results of the RTrrs calculations are presented in Section 6.0. The conclusion and references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively. v
2-1 2 PRESSURIZED THERMAL SHOCK RULE he Nuclear Regulatory Commission (NRC) recently amended its regulations for light-water-cooled nuclear power plants to clarify several items related to the fracture toughness requirements for reactor pressure vessels, including pressurized thermal shock requirements. The latest revision of the PTS m Rule , 10 CFR Part 50.61, was published in the Federal Register on December 19,1995, with an effective date of January 18,1996. This amendment to the PTS Rule makes three changes: 1. The rule incorporates in total, and therefore makes binding by rule, the method for determming the reference temperature, RTmr, including treatment of the unitradiated RTmr v alue, the margin term, and the explicit definition of" credible" surveillance data, which is also described in Regulatory Guide 1.99, Revision 2W 2. The rule is restrumred to improve clarity, with the requirements section giving only the requirements for the value for the reference temperature for end oflicense (EOL) fluence, RTns. 3. Hermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RTns. He PTS Rule requirements consist of the following: e For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTns, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material. The assessment of RTns must use the calculation procedures given in the PTS Rule, and must specify the bases for the projected value of RTrrs for each beltline material. He report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material. This assessment must be updated whenever there is a significant change in projected values of RTrrs or upon the request for a change in the expiration date for operation of the facility. Changes to RTns values are significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewal term, if applicable for the plant. He RTns screening criterion values for the beltline region are: 270 F for plates, forgings and axial weld materials, and 300 F for circumferential weld materials. Pressurized ThermalShock Rule Revision 0
3-1 3 METHOD FOR CALCULATION OF RTns RTrrs must be calculated for each vessel beltline material using a fluence value, f, which is the EOL fluence for the material. Equation 1 must be used to calculate values of RTer for each weld and plate or forging in the reactor vessel beltline. RTvor = RTuor(v> + M + ARTsor (1) Where,
=
RTuortv) Reference Temperature for a reactor vessel material in the pre-service or unitradiated condition M = Margin to be added to account for uncertainties in the values of RTuor(v>, copper and nickel contents, fluence and calculational procedures. Mis evaluated from Equation 2 M = Jcru2 , g;2 (2) a vis the standard deviation for RTuor(v).
=
ou 0 F when RTuor(v)is a measured value.
=
ou 17*F when RTuor(v)is a generic value. c6 si the standard deWation for RTuor. For plates and forgings:
=
c3 17 F when surveillance capsule data is not used.
=
on 8.5*F when surveillance capsule data is used. For welds:
=
c3 28*F when surveillance capsule data is not used.
=
o6 14*F when surveillance capsule data is used. c6not to exceed one half of ARTuor AR7kor is the mean value of the transition temperature shift, or change in ARTuor, due to irradition, and must be calculated using Equation 3. ARTuor = (CF)
- f ca23-mos/) (3)
Method For Calcualtion of RTers Revision 0
3-2 CF ( F) is the chemistry factor, which is a function of copper and nickel content. CF is determined from Tables 1 and 2 of the PTS Rule (10 CFR 50.61). Surveillance data deemed credible must be used to determine a material-specific value of CF. A material-specific value of CF is determined in Equation 5. 2 F is the higher of the best estimate or calculated neutron fluence, in units of 10" n/cm (E > 1.0 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence. The EOL fluence is used in calculating RTm. Equation 4 must be used for determuunguRTm sing Equation 3 with EOL fluence values for determining RTm RTm = RTuorcu> + M + ARTm (4) To verify that RTer for each vessel beltline material is abounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. Results from the plant-specific surveillance program must be integrated into e the RTer stimate if the plant-specific surveillance data has been deemed credible. s A material-specific value of CF for surveillance materials is determmed from Equation 5. 7 = g .po.:s-o tolos A)) go.u-o 2oics j)) (5) In Equation 5, "A,"is the measured value of ARTer and "f"is i the fluence for each surveillance data point. If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of RTer must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld. Method For Calcualtion of RTm Revision 0 L_,....... -
4-1 4 VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the V. C. Summer Unit I vessel was performed. The beltline region of a reactor vessel, per the PTS Rule, is defmed as "the region of the reactor vessel (shell material including welds, heat affected zones and plates and forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage" Figure 1 identifies and indicates the location of all beltline region materials for the V. C. Summer Unit I reactor vessel. The best estimate copper and nickel contents of the beltline materials were obtained from V.C. Summer response to NRC Generic Letter 92-01, Revision 1, Supplement ll aand WCAP-12867I4. The best estimate copper and nickel content is documented in Table I herein. The average values were calculated using all of the available material chemistry information. Initial RTm>rvalues for V. C. Summer Unit I reactor vessel beltline material properties are also shown in Table 1. i l
~
Verification of Plant Specific Material Properties Revision 0
4-2 CIRCUMFERENTIAL WELDS LONGITUDINAL WELDS 08 A9153-2 "S r 13 1" g o CORE 270*- 90* CORE d '- - m 5 ,_r
< F l i 144.0" 5
[ A9154-1 l 180* . MIDPLANE - 3.0" a $ 0* C9923-2 Z_ I i g - 46o CORE _ w 270*
" g _
90* a _ 49.0" o ' '
.C9923-1 '
1808 Figure 1: Identification and Location of Beltline Region Materials for the V. C. Summer Unit 1 Reactor Vessel Verification of Plant Specific Material Properties Revision 0
4-3 Table 1 V. C. Summer Unit 1 Reactor Vessel Beltline Unitradiated Material Properties Material Description Cu(%) Ni(%) Initial RTm/") Closure Head Flange 5297-l*) n/a n/a 10 F*) Vessel Fir.nge 5301-V-1 n/a n/a 0 F*) Intermediate Shell Plate A9154-1 0.10 0.51 30 F Intermediate Shell Plate A9153-2 0.09 0.45 -20 F Lower Shell Plate C9923-1 0.08 0.41 10 F
- . Lower She'i tSte C9923-2 0.08 0.41 10 F Intennediate Shell Longitudinal Welds, 0.05 0.91 -44 F Seams BC & BD l Lower Shell LongitudinalWelds, 0.05 0.91 -44*F
! Seams BA & BB Intermediate to Lower Shell Plate 0.05 0.91 -44*F CircumferentialAB i l Surveillance Program Weld Metal 0.04 0.95 --- ! Notes. (a) ~ The initial RTmvalues for the plates and welds are based on measured data per WCAP-12867W (b) In the past the closure head flange was reported as Heat A9231 with a IRTmof-20*F. Based on a l review of Westinghouse files, the correct data is Heat # 5297-VI with a IRTwrof 10*F. Also, the , vessel flange was reported a IRTer of 10 F., however, again , based on a review of Westinghouse files, the correct IRTerof O'F. l t Verification of Plant Specific Material Properties Revision 0
L l l 5-1 ! i i l 5 NEUTRON FLUENCE VALUES 1 1 i The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the V. C. Sum reactor vessel for 32 and 48 EFPY are shown in Table 2. These values were projected using the results of the Capsule W radiation analysis. See Section 6.0 of the Capsule W dosimetry analysis report, WCAP-15101 14 (48 EFPY values were obtained from interpolation of fluences between 32 EFPY and 54 E l TABLE 2 Fluence (E > 1.0 MeV) on the Pressure Vessel Clad / Base Interface for V. C. Summer Unit I at 32 (EOL) and 48 (Life Extension) EFPY Material Location 32 EFPY Fluence 48 EFPY Fluence Intermediate Shell Plate A9154-1 0* 3.84 x 10 n/cm2 5.70 x 10 n/cm2 Intermediate Shell Plate A9154-2 0* 3.84 x 10 n/cm2 5.70 x 10 ' n/cm2 i Lower Shell Plate C9923-1 0 2 3.84 x 10 n/cm 5.70 x 10 n/cm2 Lower Shell Plate C9923-2 po 3.84 x 10 n/cm2 5.70 x 10 n/cm2 Intermediate Shell Longitudinal 45* 1.43 x 10 n/cm2 2.15 x 10 n/cm2 Weld Seam BC& BD Lower Shell Longitudinal 2 45 1.43 x 10 n/cm 2.15 x 10 n/cm2 Wcld Seams BA &BB Intermediate to Lower Shell Plate 0 3.84 x 10 n/cm2 5.70 x 10 n/cm2 Circumferential 1 l l t i , l Neutron Fluence Values Revision 0
l s.1 6 DETERMINATION OF RTvrs VALUES FOR A'LL BELTLINE REGION MATERIALS i Using the prescribed PTS Rule methodology, RTrrs values were generated for all beltline region materials of the V. C. Summer Unit I reactor vessel for fluence values at the EOL (32 EFPY) and life ex EFPY). Each plant shall assess the RTrrsvalues based on plant-specific surveillance capsule data. For V. C. Summer Unit 1, the related surveillance program results have been included in this PTS evaluation. (See Reference 5 for the credibility evaluation of the V.C. Summer Unit I surveillance dats.) As presented in Table 3, chemistry factor values for V. C. Summer Unit I based on average copper and nickel weight percent were calculated using Tables 1 and 2 froin 10 CFR 50.61N Additionally, chemistry factor values based on credible surveillance capsule data are calculated in Table 4. Tables 5 and 6 contain the RTrrs calculations for all beltline region materials at EOL (32 EFPY) and life extension (48 EFPY). i l 1 l l l l l l l 1 l l l Determination of RTrrs Values For All Beltline Region Materials Revision 0 l l
6-2 TABLE 3 1 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR Part 50.61 Material Ni,wt % Chemistry Factor, 'F Intermediate Shell Plate A9154-1 0.51 65.0*F Given Cu wt% = 0.10 Intermediate Shell Plate A9153-2 0,45 58.0 F Given Cu wt% = 0.09 Intermediate Shell Plate C9923-1 0,41 Sl.0*F Given Cu wt% = 0.08 Lower Shell Plate C9923-2 0,41 51.0*F Given Cu wt% = 0.08 Intermediate Shell Loncitudinal Welds. BC&BD 0.91 68.0 F Given Cu wt% = 0.05 Lower Shell Loncitudinal Welds BA & BB 0.91 68.0'F Given Cu wt% = 0.05 Inter. to Lower Shell Cire. Weld Seam AB 0.91 ! 68.0 F Given Cu wt% = 0.05 Surveillance Program Weld Metal 0.95 54.0 F Given Cu wt% = 0.04 Determination of RTm Values For All Beltline Region Materials Revision 0
~ - _ - . - - - .._ - __. _ _ _ _. .- - . . -. _-..-_. --. . _ _ - - .
6-3 TABLE 4 Calculation of Chemistry Factors using Surveillance Capsub Data Per i Regulatory Guide 1.99, Revision 2, Position 2.1 i Material Capede Capsule f" FF" ARTer(4 FF*ARTer FF2 Intermediate Shell U 0.654 0.881 36.0 31.7 0.776 1 PlateA9154-1 V 1.538 1.119 52.6 58.9 1.252 X 2.543 1.250 37.7 47.1 1.563 (Longitudinal) W 4.664 1.388 65.7 91.2 1.927 Intermediate Shell U 0.654 0.881 14.5 12.8 0.776 Plate A9154-1 V 1.538 1.119 32.4 36.3 1.252 X 2.543 1.250 26.0 32.5 1.563 ! (Transverse) W 4.664 1.388 57.8 80.2 1.927 . 1 SUM 390.7 11,036 2 CFuiwi = Z(FF
- RTm) + Z(FF ) = (390.7) + (11.036) = 35.4'F Surveillance Weld U 0.654 0.881 28.0(* 24.7 0.776 Metal V 1.538 1.119 58.6(* 65.6 1.252 l X 2.543 1.250 28.3(* 35.4 1.563 l
W 4.664 1.388 54.4(* 75.5 1.927 SUM 201.2 5.518 y 2 CFw.ia = Z(FF
- RTm) + Z(FF ) = (201.2) + (5.518) = 36.5 F l
Notes: (a) f = Measured fluence from capsule W dosimetry analysis results la , (x 10* n/cm2 , E > 1.0 MeV).
.(b) FF = fluence factor = f*2"'" 80 (c) ARTm alues v are the measured 30 ft-lb shift values.
. (d) The surveillance weld metal ARTm values have been adjusted by a ratio factor of 1.26. (CFvw + CFsw = 68 'F + 54 F = 1.26) Determination of RTrrs Values For All Beltline Region Materials Revision 0
i-i, 64 l l TABLE 5 RTns Calculation for V. C. Summer Unit 1 Beltline Region Materials at EOL (32 EFPY) Material Fluence CF FF RTmnd') ARTns") Margin RTns*' (a/cm*, ('F) F>l.0MeV) I Intermediate ShellPlate A9154-1 3.84 65 1.35 30 88 34 152 l Using Surveillance Capsule Data 3.84 35.4 1.35 30 48 17 95 Intermediate ShellPlate A9153-2 3.84 58 1.35 -20 78 34 92 l Lower Shell Plate C9923-1 3.84 51 1.35 10 69 34 113 Lower ShellPlate C9923-2 3.84 51 1.35 10 69 34 113 latennediate Shell & Lower Long. 1,43 68 1.10 -44 75 56 87 Weld Materials BC, BD and BA,
- - - - - BB
. - (.15'
_ _ Azimuth) Using Surveillance Capsule Data 1.43 36.5 1.10 -44 40 28 24 Intermediate to Lower Shell Circumferential Weld Seam AS Using Surveillance Capsule Data 3.84 36.5 1.35 -44 49 28 33 Notes. (a) Initial RTer values are measured values. (b) RTns = Initial RTerm> + ARTns + Margin ('F) (c) ARTns = CF
- FF i
l l I. I Determinatien of RTrrs Values For All Beltline Region Materials Revision 0
6-5 TABLE 6 RTns Calculation for V. C. Summer Unit 1 Beltline Region Materials at Life Extension (48 EFPY) i j Material Fluence CF FF RTmyrm/*) ARTnsM Margin RTrrs" 2 l (n/cm , (.p) ! E>1.0MeV) Intermediate Shell Plate A9154-1 5.70 65 1.43 30 93 34 157 j Using Surveillance Capsule Data 5.70 35.4 1.43 30 51 17 98 Intermediate Shell Plate A9153-2 5.70 58 1.43 -20 83 34 97 l Lower Shell Plate C9923-1 5.70 .51 1.43 10 73 34 117 j Lower Shell Plate C9923-2 5.70 51 1.43 10 73 34 117 Intermediate Shell& Lower Long. 2.15 68 1.21 -44 82 56 94 l Weld Materials BC, BD and BA, BB (45* Azimuth) Using Smveillance Capsule Data 2.15 36.5 1.21 -44 44 28 28 Intermediate to Lower Shell Circumferential Weld Seam A5 Using Surveillance Capsule Data 5.70 36.5 1.43 -44 52 28 36 Notes. (a) v Initial RTmn alues are measured values. (b) RTm = Initial RTamu) + ARTns + Margin ( F) (c) ARTm = CF
- FF l
l l-l
- Determination of RTns Values For All Beltline Region Materials Revision 0 L
l
7-1 t t-7- CONCLUSIONS As shown in Tables 5 and 6, all of te beltline region materials in the V. C. Summer Unit I reactor vessel
- l. ; have EOL(32 EFPY)RTers and Lifi. Extension (48 EFPY) RTns values below the screening criteria
. values of 270*F for plates, forgings and longitudinal welds and 300 F for circumferential welds.
i l i f i l-I i l' l Conclusions Revision 0
l l 8-1 1 ! 8 REFERENCES l l 1 10 CFR Part 50.61, " Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events", Federal Register, Volume 60, No. 243, dated December 19,1995. 2 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988. 3 WCAP-9234, " South Carolina Electric and Gas Company Virgil C. Summer Nuclear Plant Unit 1 No.1 Reactor Vessel Radiation Surveillance Program", J. A. Davidson, January 1978. 4 WCAP-12867, " Analysis of Capsule X from the South Carolina Electric and Gas Company V. C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program", J. M. Chicots, et. al., March, l 1991. 5 WCAP-15101, " Analysis of Capsule W from the South Carolina Electric and Gas Company V. C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program", T. J. Laubham, et al., September 1998. l 6 South Carolina Electric and Gas Letter, GJ. Taylor to U.S. NRC, Dated 11/7/95, Subject, " Virgil C. Summer Nuclear Station, Docket No. 50/395, Operating License No. NPF-12, Response to GL. 92-01, Revision 1, Supplement 1" l l l l' l l References Revision 0 l}}