ML20092J119
| ML20092J119 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/31/1995 |
| From: | Peter P, Rishel R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20092J085 | List: |
| References | |
| WCAP-14345, NUDOCS 9509210346 | |
| Download: ML20092J119 (18) | |
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- 1
- EVALUATION OF PRESSURIZED THERMAL SHOCK f FOR THE COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) UNIT 2 ;.
1 i P. A. Peter j July 1995 I i Work Performed Under Shop Order TUPP-2032 l l l Prepared by Westinghouse Electric Corporation ; for the Texas Utilities Electric Company Approved by: R D. Rfshel, Manager Metallurgical & NDE Analysis WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1995 Westinghouse Electric Corporation All Rights Reserved .j l I
PREFACE 4 This report has been technically reviewed and verified by: } E. Terek $& 5 l 1 l I 1 l I l l I i 1
f r 4
~ TABLE OF CONTENTS .
Section lidt 4
- East LIST OF TABLES . .. . . . . . . . . . .....................................iii <
LIST OF FIG URES . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii i ; 1
1.0 INTRODUCTION
. . . . . . . . . . . . .................................. l' 2.0 - PRESSURIZED THERM AL SHOCK . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . '2 t
3.0 METHOD FOR CALCULATION OF RTm . . .. .*. - ........... . . 4 5 l 4.0 ' VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES . . . . . . . . . . . . . l l l 15.0- NEUTRON FLUENCE VALUES . . . . . . . . . . . . . .... ................... 8 6.0 . DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERI AL . . . . . . . . . . . . . . . ... ............. ................... 9
7.0 CONCLUSION
S . . . . . . . . . . . . . . . . . . . ...... ... ................... 12 a 1 i 1
8.0 REFERENCES
................ ...... ... .. .................... 13 l
)
i 1 1
-i 1
1 l 11 i
. LIST OF TABLES '
IAhlt Illlt East 1 CPSES Unit 2 Reactor Vessel Beltlire Region Material Properties Used in - Calculations . . . . . . . . . . . . . . . . . . . .. . .......................... 7 2 Neutron Exposure (n/cm2 , E>l.0 MeV) Projections at Peak Locations on the ,
^ CPSES Unit 2 Pressure Vessel Clad / Base Metal Interface .. . ............... 8.
3' PTS Calculations for CPSES Unit 2 at 0.904 EFPY . . . . . . ................... 10
-l 4 FTS Calculations for CPSES Unit 2 at 32 EFPY . . . . . . . . .. ................ 10 5' PTS Calculations for CPSES Unit 2 at 48 EFPY , ..... ................... I1 I
L LIST OF FIGURES Engun Iitic Eags 1 Identification and Location of Beltline Region Material for the CPSES Unit 2 Reactor Vessel . . . . . . . . . . . . . . . . . . . . ....... . . ....... ........ 6 I i
)
2 RTm versus EFPY of Operation for CPSES Unit 2 Limiting Material - l 1 Intermediate Shell Plate R3807-2 . . . . . . . . . . . . . . . . . . . . ................. 12 l i I I 1 1 1 iii i
SECTION
1.0 INTRODUCTION
A limiting condition on reactor vessel integrity known as Pressurized Thermal Shock (PTS) may occur i during a severe system transient such as a Imss-Of-Coolant Accident (LOCA) or a steam line break. ' Such transients may challenge the integrity of a reactor vessel under the following conditions: I
- severe overcooling of the inside surface of the vessel wall followed by high repressurization:
- tiptificant degradation of vessel material toughness caused by radiation embrittlement; and I
- 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 nil-ductility reference temperature, termed RTys tu. RTns screening values were set for beltline axial welds, forgings or plates and for beltline circumferential weld seams for the end-of-license plant , operation. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end oflicense. 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 14,1991 t23 This amendment makes the procedure for calculating RTns values consistent with the methods given in Regulatory Guide 1.99, Revision 2tH. The purpose of this report is to determine the RTns values for the Texas Utilities Electric Company (TU Electric) Comanche Peak Steam Electric Station (CPSES) Unit 2 reactor vessel to address the revised FTS Rule. Section 2.0 discusses the Rule and its requirements. Section 3.0 provides the methodology for calculating RTn3 Section 4.0 provides the reactor vessel beltline region material properties for the CPSES Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RTn, calculstions are presented in Section 6.0. The conclusions and references for the FTS evaluation follow in Sections 7.0 and 8.0. respectively. 1
3 SECTION 2.0 .. j PRESSURan THERMAL SHOCK 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 RTm i values. The Rule outlines regulations to address the potential for FTS events on pressurized water I reactor 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 FTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistara 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 Rule establishes the following requirements for all domestic, operating PWRs: j i
- All plants must submit projected values of RTm for reactor vessel beltline matertal 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 ram _wed This l
i [ assessment must be submitted within six months after the effective date of this Rule if the value of RTrra for any material is projected to exceed the screening criteria. Otherwise, it ! i . must be submitted with the next update of the pressure-temperature limits, or the next reactor t vessel surveillance capsule report, or within five yeart from the effective date of this Rule change, whichever comes first. These values must be calculated based on the methodology specified in this mle. The submittal must include the following:
- 1) the bases for the projection (including any assumptions regarding core loading patterns),
i and 1
- 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
- j. must be provided.
2
=
- v -
f - e- .. _ - _ ,- 2__ _ _ _ _ _ _ _ _ ,
i
- The RTns screening criteria for the reactor vessel beltline region is:
270'F for plates, forgings, axial welds; and 300 F for circumferential weld material.
- The following equations must be used to calculate the RTn, values for each weld, plate or forging in the reactor vessel beltline:
Equation 1: RT,n = I + M + ARTns Equation 2: ARTpn = CF
- f <a2s.aio ws n
- All values of RTn, 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 three years of reaching i the screening criteria, including analyses of alternatives to minimize the FTS concern.
- NRC approval for operation beyond the screening criteria is required.
3
SECTION 3.0 METiiOD FOR CALCULA~I1ON OF RTn s In the P'IS Rule, the NRC Staff has selected a conservative and uniform method for determining plant specific values of RTns at a given time. For the purpose of comparison with the screening ariteria, the value of RTm for the reactor vessel must be calculated for each weld and plate or forging in the beltline region a follows. RTns = I + M + ARTm, where ARTm = CF
- FF I= Initial reference temperature (RTym) in 'F of the unirradiated material Mm Margin to be added to cover uncertainties in the values of initial RT,m, copper and nickel contents, fluence and calculational procedures, per Regulatory Guide 1.99, Revision 2, in 'F.
M = margin = 2 0 3 2 + o,2, op o, = 0*F when I is a measured value c i= 17*F when I is a generic value For plates and forgings: 0 3 = 17'F when surveillance capsule data is not used o, = 8.5'F when surveillance capsule data is used For welds: 03 = 28'F when surveillance capsule data is not used 03 = 14'F when surveillance capsule data is used o3 not to exceed 0.5* ARTns FF = fluence factor = f *23.aio an o, where 2 f = Neutron fluence (10" n/cm . Fal.0 MeV) at the clad / base metal interface CF = Chemistry Factor in 'F from the tabled 21 for welds and base metals (plates or forgings). If plant-specific surveillance data from two or more surveillanm capsules has been deemed credible per Regulatory Guide 1.99, Revision 2, it should be considered in the calculation of the chemistry factor. 4
SECTION 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 CPSES Unit 2 vessel beltline region was performed. The beltline is defined by ASTM E185-821'l to be "the irradiated region of the reactor vessel (shell material including weld regions and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions that are predicted to experience sufficient neutron damage to warrant consideration in the selection of the surveillance material". Figure 1 identifies and indicates the location of all beltline region material for the CPSES Unit 2 reactor vessel. Material property values along with the copper and nickel values used in the calculations were obtained from Reference 5. A summary of the pertment chemical and mechanical properties of the beltline region plates and weld materials of the CPSES Unit 2 reactor vessel are given in Table 1. 5
N W R3807-2 90 101-1248 d E
=
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$ 101-124A
% 30*
CORE G 101-171 R3807-3 01-1240 R3816-1 90[101-142A d : R3816-3 Y w ac 0 i80 30 30* 101-142C 10!-1428 4 3816-2 g Figure 1 Identification and Location of Beltline Region Material for the Comanche Peak Unit 2 Reactor Vessel 6
TABLE 1 CPSES Unit 2 Reactor Vessel Beltline Region Material Properties Used in Calculations Material CuW" nit
- Chemistry Factor ** Initial RTm'*
Intermediate Shell Plate R3807-1 0.06 0.64 37.0 -20 Intermediate Shell Plate R3807-2 0.0625 0.63 38.8 10 Intermediate Shell Plate R3807 3 0.05 0.60 31.0 -20 I.ower Shell Plate R3816-1 0.05 0.59 31.0 -30 Lower Shell Plate R3816-2 0.03 0.65 20.0 0 Lower Shell Plate R3816-3 0.04 0.63 26.0 40 Circumferential Weld 101-171 0.04 0.051 28.8 -60 Inter. & Lower Shell Imgitudinal Welds 0.06 0.065 36.5 -50
,ivi a (a) Average Cu and Ni weight percent values from WCAP-10684, Tables A-1 through A-3. The average Cu and Ni values are reported as integers in order to determine the most accurate Chemistry Factor values possible.
(b) Chemistry Factor calcuhted per Regulatory Guide 1.99, Revision 2. Position 1.1. (c) Initial RTm values are measured values. 7
SECTION 5.0 ! NEUTRON FLUENCE VALUES The calculated peak fast neutron fluence (E>l.0 MeV) values at the inner surface of the CPSES Unit 2 reactor vessel are shown in Table 2. These values were projected using the results of the surveillance Capsule U radiation analysism. The RTm calculations for the base metal and circumferential weld - were performed using the peak fluence values, which occur at the 45" azimuthal angle. The fluence values at the 0 and 30' azimuthal angles were used for the RTrrs calculations for the longitudinal welds of the CPSES Unit 2 reactor vessel. (Additional details on the basis for the fluence projections can be found in Section 6.0 of Reference 7.) l TABLE 2 Neutron Exposure (a/cm2 , E>1.0 MeV) Projections at Peak Locations on the CPSES Unit 2 Pressure Vessel Clad / Base Metal Interface 0* 30' 45' Long. Welds 101 124A & long Welds 101-124BC & Base Metal Plates & Cire. Time 101-142A 101-142BC Weld 101-171 0.904 EFPY 3.919 x 10" 7.195 x 10" 8.011 x 10" 32 EFPY 1.387x 10" 2.548 x 10" 2.836 x 10 48 EFPY 2.081 x 10 3.821 x 10" 4.255 x 10" 8
1 SECTION 6.0 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIAL Using the prescribed PTS Rule methodology, RT mvalues were generated for all beltline region material of the CPSES Unit 2 reactor vessel for fluence values at the present time (0.904 EFPY per Capsule U analysis), end-of-license (32 EFPY), and 48 EFPY. The IrrS Rule requires that each plant assess the RTm 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 l
RTns values change significantly. (Changes to RTm values are considered significant if the 1 value determined with RTm equations (1) and (2), or that using capsule data, or both, exceed l the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.) Plant specific surveillance capsule data is not provided since Capsule U is the first capsule to be removed from the CPSES Unit 2 reactor vessel. Per Regulatory Guide 1.99, Revision 2, Position 2.1, when rwo or more credible sets of surveillance data become available from the reactor in question, they may be used in the calculation of the chemistry factors. (Furthermore, if the surveillance capsule data gives a higher value of adjusted reference temperature (or RTm) than that given by using the procedure of Position 1.1, the surveillance data should be used. If this procedure gives a lower value, either may be used.) Therefore, the chemistry factors for CPSES Unit 2 were calculated using only Tables I and 2 from 10 CFR 50.61. Tables 3 through 5 provide a summary of the RTm values for all beltline region material for 0.9M,32, and 48 EFPY, respectively. 9
TABLE 3 PTS Calculations for CPSES Unit 2 at 0.904 EFPY Material CF f FF I M 3RTm RTm Screening Criteria 0904 EFPY Inter. Shell Plate R3807-1 37.0 0.08011 03740 -20 13.8 13.8 8 270 Inter. Shell Plate R3807 2 38.8 0.08011 03740 10 14.5 14 3 39 270 Inter. Shell Plate R3807-3 31.0 0.08011 0.3740 -20 11.6 11.6 3 270 lower Shell Plate R3816-1 31.0 0.08011 03740 -30 11.6 11.6 -7 270 lower Shell Plate R3816-2 20.0 0.08011 03740 0 7.5 7.5 15 270 tower Shell Plate R3816-3 26.0 0.08011 03740 -40 9.7 9.7 -21 270 Cire. Weld 101-171 28.8 0.08011 03740 -60 10.8 10.8 -38 300 Inter long Weld 101-124 A 363 0.03919 0.2560 -50 93 93 -31 270 Inter long Weld 101-124BC 363 0.07195 03543 -50 12.9 12.9 -24 270 lower long Weld 101 142A 363 0.03919 0.2560 -50 93 93 -31 270 lower leng Weld 101-142B/C 36.5 0 07195 0 3543 -50 12 9 12 9 24 270 ,dQ1L Iruttal RTum values are measured values. TABLE 4 PTS Calculations for CPSES Unit 2 at 32 EFPY Matenal CF f FF I M ARTm RTm Screening Criteria 32 EPPY Inter. Shell Plate R3807-1 37.0 2.836 1.277 -20 34 47.3 61 270 Inter. Shell Plate R3807-2 38.8 2.836 1.277 10 34 49.6 94 270 Inter. Shell Plate R3807-3 31.0 2.836 1.277 -20 34 39.6 54 270 lower Shell Plate R3816-1 31.0 2.836 1.277 -30 34 39.6 44 270 tower Shell Plate R3816-2 20.0 2.836 1.277 0 25.5 25.5 51 270 lower Shell Plate R3816 3 26.0 2.836 1.277 40 33.2 33.2 26 270 Cire. Weld 101 171 28.8 2.836 1.277 40 36.8 36.8 14 300 Inter long Weld 101-124A 36.5 1387 1.091 -50 39.8 39.8 30 270 Inter long Weld 101 124BC 36.5 2348 1.251 -50 45.7 45.7 41 270 lower Long Weld 101142A 36 3 1387 1.091 -50 39.8 39.8 30 270 lower Innr Weld 101 142B/C 36 5 2 548 1.251 -50 45 7 457 41 270 ,R 1ruttal RTum values are sneasured values. 10
TABLE 5 FTS Calculations for CPSES Unit 2 at 48 EFPY Material CF f FF I M ARTm RTm Screening Criteria 48 EFPY Inter. Shell Plate R3807-1 37.0 4.255 1369 -20 34 50.7 65 270 Inter. Shell Plate R3807 2 38.8 4.255 1369 10 34 53.1 97 270 Inter. Shell Plate R38(T7-3 31.0 4.255 1369 -20 34 42.5 57 270 Lower Shell Plate R3816-1 31.0 4.255 1369 -30 34 42.5 47 270 Imer Shell Plate R3816-2 20.0 4.255 1369 0 27.4 27.4 55 270 Lower Shell Plate R3816-3 26.0 4.255 1369 40 34 35.6 30 270 Cire. Weld 101 171 28.8 4.255 1369 -60 39.4 39.4 19 300 Inter Long Weld 101-124 A 36.5 2.081 1.199 -50 43.8 43.8 38 270 Inter Long Weld 101-124B/C 36.5 3.821 1346 -50 49.1 49.1 48 270 lower Long Weld 101 142A 36.5 2.081 1.199 -50 43.8 43.8 38 270 lower Long Weld 101 142B/C 36.5 3.821 1346 -50 49.1 49.1 48 270
,R Imtial RTm values are measured values.
All of the beltline material RTns values in the CPSES Unit 2 reactor vessel are below the screening criteria at 32 and 48 EFPY. The following plot of RTn, versus EFPY (Figure 2) also illustrates the available margin, where RTrn = I + M + ART,n. 11
( i f SECTION
7.0 CONCLUSION
S As shown in Tables 3,4, and 5, all RTm values remain below the NRC FTS screening criteria values using fluence values for the present time (0.904 EFPY), end-of-license (32 EFPY), and 48 EFPY. A plot of the RTm versus EFPY, shown in Figure 2, illustrates the available margin for the most limiting material in the CPSES Unit 2 reactor vessel beltline region, Intermediate Shell Plate R3807-2. soo R61 lEf61 llD6] sao ano l 150 100 M JEE_____________
/ ,,/
,/'
"lsejr o , ,
- o. sos sit 4s EFPYofOposson i
Figure 2 RTm versus EFPY of Operation for CPSES Unit 2 Limiting Material - Intermediate ShcIl Plate R3807 2 12
SECTION
8.0 REFERENCES
[1] 10 CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23,1985. [2] 10 CFR Part 50.61, " Fracture Toughness Requirements for Protection Against Pressmized Thermal Shock Events," May 15, 1991. [3] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988. [4] ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels" E706 (IF). [5] WCAP-10684, " Texas Utilities Generating Company Comanche Peak Unit No. 2 Reactor Vessel Radiation Surveillance Program", L. R. Singer, October 1984. [6] Comanche Peak S.E.S. Final Safety Analysis Report Units I and 2, Figure 5.3-1B, " Unit No. 2 Reactor Vessel Beltline Region Material Identification", Amendment 8, November 30,1979. [7] WCAP-14315, " Analysis of Capsule U from the Texas Utilities Electric Company Comanche Peak Unit No. 2 Reactor Vessel Radiation Surveillance Program", R. Auerswald, et al., July 1995. 13 - ___- _ _ _ _ _ _ _ _ _}}