ML20086P273
| ML20086P273 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/30/1991 |
| From: | Chicots J, Ray N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20086P271 | List: |
| References | |
| WCAP-13106, NUDOCS 9112260312 | |
| Download: ML20086P273 (16) | |
Text
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- i WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-13106 EVALUATION OF PRESSURIZED THERMAL SHOCK FOR BEAVER VALLEY UNIT 1 J. M. Chicots N. K. Ray November 1991 Work Performed Under Shop Order DOLP-108 Prepared by Westinghouse Electric Corporation for the Duquesne Light Company Approved by: -
)N O T.A.Meyer,M1 nager Structural Reliability & Plant Life Optimization WESTlHGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1991 Westinghouse Electric Corp.
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c : Ju ! TABLE OF CO ,1[ Nil E121 . Table of Contents i
. List ~ of Tables . 11
-list of Figures- 11 s -
1.- Introduction- 1 i
-2. Pressurized Thermal-Shock 2 ' r; '
3 .- Methods of Calculation of RTPTS 3 4.- Verification of Plant-Specific Material Properties 6
- 5. Neutron Flue'nce' Values 8
- 6. - Determination of RTPTS Values for All Beltline 9 Regicn Materials
- 7. Conclusions 12.
- 8. - - References- 13-T '
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I l . 1 LIST OF TABLES lible lilla EA91
- 1. Calculation of Chemistry Factors Using Beaver Valley Unit 1 5 Surveillance Capsule Data f
- 2. Beaver Valley Unit 1 Reactor Vessel Beltline Region Material 8 Properties
- 3. Neutron Exposure Projections at Key Locations on the B Beaver Valley Unit 1 Pressure Vessel Clad / Base Metal Interface for 32 EFPY
- 4. RTPTS Values for Beaver Valley Unit 1 for 32 EFPY 10 t
- 5. RTPTS Values for Beaver Valley Unit I for 48 EFPY 11 LIST OF FIGURES Fiaure Title Eagg
- 1. Identification and Location of Beltline Region 7 Material for the Beaver Varley Unit 1 Reactor Vessel
- 2. RTPTS versus Fluence Curves for Beaver Valley Unit 1 12 Limiting Material - Lower Shell Plate, B6903-1 '
11
- 1. INTRODUCTION A limiting condition on r' actor vessel integrity known as pressurized thermal shock (PTS) may occur during a severe system transient such as a loss-of-coolant-accident (1.0CA) or a steam line bret.k. Such transients may challenge the integrity of a reactor vessel under the following conditions:
severe overcooling of the inside surf ace of the vessel wall followed by high repressurization ( significtnt degradation of vessel material toughness caused by ( radiation embrittlement the presence of a critical-size defect in the vessel wall Fracture mechanics analysis can be used to evaluate reactor vessel integrity under severe transient conditions, in 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on pressurized thermal shock, it established screening criterion on pressurized wav;r reactor (PWR) vessel embrittlement as measured by the nil-ductility reference t:.mperature, termed RT PTS ill. RTpn screening values were set for beltline axial welds, forgings and plates
^
and for beltline circumferential weld seams for end-of-life plant operation. 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-life. The Nuclear Regulatory Commission has amended its regulations for light water nuclear power plants to change the proc 9 dure for calculating radiation embrittlement. The revised PTS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14,1991[2]. This amendment makes the procedure for calculating RTPTS values consi.tw vith the methods given in Regulatory Guide 1.99, Revision 2[3), l
The purpose of this report is to determine the reference temperatures for pressurized thermal shock (RTPTS) values for the Beaver Valley Unit I reactor vessel ,, address the Pressurized Thermal Snock (PTS) Rule. Section 2 discusses the Rule and its requirements. Section 3 provides the methodology for calculating RTPTS. Section 4 provides the reactor vessel beltline region material properties for the Beaver Valley Unit I reactor vessel. The neutron fluence values used in this analysis are presented in Section 5. The results of the RTpys calculations are presented in Section 6. The conclusions and references for the PTS evaluation follow in Sections 7 and 8, respectively.
- 2. PRESSURIZED THERMAL SH0CK The PTS Rule requires that the PTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that indicates a significant change in projected values.
' The Rule outlines regulations to address the potential for PTS events on pressurized water reactor (PWR) vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC). PTS 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 produce the propagation of flaws postulated to exist near the inner wall surface, theraby potentially affecting the integrity of the vessel.
The Rule establishes the following requirements for all domestic, operating PWRs: All plants must submit projected values of RT PTS for reactor vessel beltline materials by giving values for time of submittal, the expiration cate of the operating license, and the projected expiration date if a change in 4 operating license or renewal has been requested. This as e w st must be submitted by six months after the efective date of this Rule if the value of RT PTS for sny material is projected to exceed the
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l screening criteria. Otherwise, it should be submitted with the next update of the pressure-temperature limits, or the next reactor vessel surveillance report, or 5 years from tt.e effective date of this Rule, whichever comes first. These values must be calculated based on the methodology specified in this rule. The submittal must include the following:
- 1) the bases for the projection (including any assumptions regarding core loading patterns),
- 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 b(; provided.)
- The RTPTS (measure of fracture resistance) Screening Criterions for the recctor vessel beltline region is 270*F for plates, forgings, axial welds 300*F for circumferential weld materials The following equations should be used to calculate the RTpT3 values for each weld, plate or forging in the reactor vessel beltline.
Equation 1: RTPTS = 1 + H + ARTPTS Equation 2: ARTPTS - (CF)f(0.28-0.10 log f) All values of RTp73 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 emDrittlement. Plant-specific PTS safety analyses are required before a plant is within 3 years of reaching the Screening Criterion, including analyses of alternatives to minimize the PTS concern. NRC approval for operation beyond the Screening Criterion is required. 3, METHOD FOR CALCULATION OF RT PTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific valtes of RTPTS at a given time. For the purpose of comparison with the Screening Critierion, the value of RTPTS for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as given below. t
s RTPTS - I + M + ARTpys, where ARTPTS - (CF)f(0.28-0.10 log f) I- Initial reference temperature (RTNDT) of the untrradiated material M- Margin to be added to cover uncertainties in the values of initial RTNDT, copper and nickel contents, fluence and calculational procedures. 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. Reference 3 defines margin as M _ = 2 [o2 + ,2 where, I A og is the standard deviation for the initial RTNDT. If the initial RTNDT is a measured value, og is considered to be i zero. If a generic value is used, og is "F. oA is the standard deviation for the ARTNDT and oA . is 17'F for_ plates and forging *., and 28'F for welds. Also note, oA can be cut into half when surveillance capsule data is used in developing the chemistry factor, Hence for the lower shell plate, B6903-1 surveillance capsule data is available (see Table !
- 1) and the margin is calculated as: '-
M = 2 / 02 + (17/2)2 . 37.p Similarly for. the longitudinal weld, 305424, surveillar.ce capsule data is available and the margin was-calculated as: Ma2/172 + (28/2)2 , 44 7 f= Neutron fluence, n/cm2 (E > IMeV at the clad / base metal interface), divided by 10 19
l E . 3 CF - Chemistry factor from tables [2] for welds and for base metal (plates and forgings), if plant-specife surveillance data has been deemed credible per Reg. Guide 1.99, Rev. 2, it may be considered in the calculation of the chemistry factor. The chemistry factors for the lower shell plate, B6903-1 and the longitudinal weld, 305424 were calculated using the surveillance capsule data as shown in Table 1. TABLE 1 CALCULATION OF CHEMISTRY FACTORS USING BEAVER VALLEY UNIT 1 SURVEILLANCE CAPSULE DATA (5) FLUENC( ART,33 MAttR!AL CAPSULE II (10 nM ) I ff (*F) ff x ARi m ff I Lower $ hell Plate. V 0.291 0.662 130 86.1 0.438 86903 1 0 0.654 0.881 120 105.7 0.776 (Longitudinal) V 0.949 0.985 150 147.8 0.970 Lower Shell Plate. V 0.291 0.662 '40 92.7 0.438 86903 1 U 0.654 0.881 135 118.9 0.776 (Transverse) V 0.949 0.985 185 XZ_ 0.970. I = 733.4 4.368 2 (ff x ART,33) CHEH15TRY FACTOR = = W = 167.9 Z(ff) I 4,368 Longitudinal Weld. V 0.291 0.662 150 99.3 0.438 (Weld Vire Heat. U 0.654 0.881 155 136.6 0.776 305424) V 0.949 0.985 185 182 2 _Lj!2q,, I = 418.1 2.184 I (ff x ART,33) CHEMISTRY FACTOR = = W = 191.4 2(ff) 2 2.184 hitti ff = fluence fattor
J 4'. VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties was performed. The beltline region is defined by the PTS RuleI23 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 activa core-and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron irradiation 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 Beaver Valley Unit I reactor vessel. Material property values were derived from vessel fabrication test , certificate results. Fast neutron irradiation-induced changes in the tension, fracture and impact properties of reactor vessel materials are largely dependent.on-chemical composition, particularly in the copper concentration. The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration with the weldments. A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the Beever Valley Unit I reactor vessel are given in Table 2. . All of the initial RTNDT values (I-RTNDT) are also presented in Table 2. i
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CIRCUMrERINTlit $fAM$ V[RTICAL $[AMS 270' B6607 2 19-714B f 8.2" 45
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g 15' ' q CORI 0' g f 180' 49.1* 20 714A / 90'- B690?-1 Figure 1. Identification and Location of Beltline Region Material for the Beaver Valley Unit 1 Reactor Vessel
- o TABLE 2 l BEAVER VALLEY UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES :
CU Ni l-RTNDT Material Description (%) (%) ('F) Intermediate Shell, B6607-1 0.14 0.62 43 Intermediate Shell, B6607-2 0.14 0.62 73 Lower Shell, B6903-1 0.20 0.54 27 I Lower Shell, B7203-2 0.14 0.57 20 Longitudinal Weld, 305424 0.28 0.63 -56 ! Longitudinal Weld, 305414 0.34 0.61 -56 l Circumferential Weld 0.29 0.07 -56
- 5. NEUTRON FLUENCE-VALUES The calculated fast neutron fluence (E>l MeV) at the inner surface of the Beaver Valley Unit I reactor vessel. is shown in Table 3. - These values
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were projected using the results of the Capsule W radiation surveillance- ! programI43 i TABLE 3 : NEUTRON EXPOSURE PROJECTIONS AT' KEY LOCATIONS ON THE BEAVEP, VALLEY tlNIT 1 : PRESSURE VESSEL CLAD / BASE METAL lNTERFACE FOR 32 EFPY
.0' 15' 30' 45' Fluence'x 1019 n/cm 2 4.07 2,06 1.13 0.75-
_(E > 1 MeV) !
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6'. DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTPTS values were generated for all beltline region materials of the Beaver Valley 'Jnit I reactor vessel as a function of end-of-life (32 EFPY) and 48 EFPY fluence values. The fluence data were generated based on the most recent surveillance capsule program resultsl43 Tables 4 and 5 provide a summary of the RTPTS values for all beltline region materials for the end-of-life (32 EFPY) and 48 EFPY, respectively, using the PTS Rule. The PTS Rule requires that each plant assess the RTPTS values based on plant specific surveillance capsule data under certain conditions. These conditions are:
- Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2, and RTPTS values change significantly. (Changes to RTPTS values are considered significant if the value determined with RTPTS equations (1) and (2), or that using capsule data, or both, exceed the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.)
For Beaver Valley Unit 1, the use of plant specific surveillance capsule data arises beca9se of the following reason:
- 1) There have been three capsules removed from the reactor vessel, hence the data is credible per Regulatory Guide 1.99, Revision 2.
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l . TABLE 4 RT PTS VALUES FOR BEAVER VALLEY UNIT 1 FOR 32 EFPY ARTNDT('F) + Initial R1NDT + Margin - RTPTS Haterial (CF x FF*) ('F) (*F) ('F) Intermediate Shell 100.5 1.36 43 34 214 Plate, B6607-1 Intermediate Shell 100.5 1.36 73 34 244 Plate, B6607-2 Lower Shell 141.8 1.36 27 34 254 Plate, B6903-1 (167.9) 1.36 27 (17) (272) Lower Shell 98.65 1.36 20 34 188 Plate, B7203-2 Circumferential 132.9 1.36 -56 66 191 Weld Seam 11-714 (Wire Ht. 90136) Longitudinal Weld 191.65 0.92 -56 66 186 Seam 19-714 A & B (191.4) 0.92 " (44) (164) (Wire Ht. 305424) Longitudinal Weld 210.45 0.92 -56 66 203 Seam 20-714 A & B (Wire Ht. 305414) () Indicates numbers were calculated using surveilla.; capsule data. Fluence factor based upon peak inner surface neutron fluence of 4.07 x 2 10l9 n/cm [4], except for longitudinal welds. For longitudinal welds, the fluence factor is based on a neutron flutnce of 0.751 x 1019 n/cm2 [4] at the inner surface of the weld.
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TABLE 5 RT PTS VALUES FOR BEAVER VALLEY UNIT 1 FOR 48 EFPY ARTNDT('F) + Initial RT NDT + Hargin - RTPTS Naterial (CF x FF*) ('F) ('F) ('F) Intermediate Shell 100.5 1.44 43 34 222 Plate, B6607-1 Intermediate Shell 100.5 1.44 73 34 252 Plate, 86607-2 Lower Shell 141.8 1.44 27 34 265 Plate, B6903-1 (167.9) 1.44 27 (17) (286) Lower Stell 98.65 1.44 20 34 196 Plate, 87203-2 Circumferential 132.9 1,44 -56 66 201 Weld Seam 11-714 (Wire Ht. 90136) Longitudinal Weld 191.65 1.034 -56 66 208 Seam 19-714 A & B (191.4) 1.034 -56 (44) (186) (Wire Ht. 305424) longitudinal Weld 210.45 1.034 -56 66 228 Seam 20-714 A & B (Wire Ht. 305414) () Indicates numbers were calculate.1 using surveillance capsule data. Fluence factor based upon peak inner surface neutron fluence of 6.11 x 1019 n/cm2 [4), except for longitudinal welds. For longitudinal welds, the fluence factor is based on a neutron fluence of 1.13 x 1019 n/cm2 [4] at the inner surface of the weld. l
t 7, CONCLUSIONS As shown in Tables 4 and 5, all the RTPTS values remain below the NRC screening values for PTS using the projected fluence values for both the end-of-life (32 EFPY) > and 48 EFPY, axcept o the lower shell plate, B6903-1 when surveillance capsule data is considered. n.e lower shell plate, B6903-1 has a RTPTS values at 32 EFPY and 48 EFPY of 272*F and 286*F, respectively when surveillance capsule data is used. A plot of the RTPTS values versus the fluence are shown in Figure 2 for the most limiting material, the lower shell plate, 86903-1, in the Beaver Valley Unit I reactor vessel beltline regiori. . 350 ~ 325 - - - - - - - - 300 - - - - - - - - ~ LOWER SHELL B69031 * ,,,, 275 -
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~ SCREENING CRITERIA " '
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LOWER SHELL B69031 u 200
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- USING SURVElLLANCE CAPSULE DATA 25 -- - " --- - -- - - - -
0 1E+18 2E+18 3E+18 SE+18 1E+19 - 2E+18 3E+19 SE+19 1E+20 8 FLUENCE ( NEUTRONS / CM ) i l Figure 2. RTPTS versus Fluence Curves for Beaver Valley Unit 1 I_ Limiting Material - Lower Shell Plate, B6903-1. w r em ++_..e.- s. ~ ,e....-,w-,,w- s. w. -
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- 8. REFERENCES l
[1] 10CFR Part 50, " Analysis of Potential Pressurized Thermal ; Shock Events," July 23, 1985. [2] 10CFR Part 50, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", May 15, 1991. (PTS Rule) l [3] Regulatory Guide 1.99, Revision 2, " Radiation' Embrittlement of Reactor Vessel Materials,' U.S. Nuclear Regulatory Commission, May 1988. [4] WCAP-12005, " Analysis of Capsule W from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko, et al . , November 1988. [5] MT-SMART-209(88), "Re:ponse to U.S. Nuclear Regulatory Commision Generic Letter 88-11 for the Beaver Valley Unit 1 Reactor Vessel", N. K. Ray, et. al., November, 1988. I I l l
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