ML20141D624
| ML20141D624 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/30/1997 |
| From: | Christopher Boyd, Howell D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20141D495 | List: |
| References | |
| WCAP-14638, WCAP-14638-R01, WCAP-14638-R1, NUDOCS 9705200157 | |
| Download: ML20141D624 (18) | |
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Rev ion 1 i
L EVALUATION OF l
PRESSURIZED
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THERMAL SHOCK 4
l FOR PRAIRIE ISLAND L
UNIT 2 I
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14638, Revision 1 Evaluation of Pressurized Thermal Shock for Prairie Island Unit 2 T. J. Laubham April 1997 Work Performed Under Shop Order NLBP-108 Prepared by Westinghouse Electric Corporation for Northem States Power Company Approved:
u C. H. Boyd, Manager Engineering & Materials Technology Approved:
A D. A. Howell, Manager' Mechanical Systems integration WESTINGHOUSE ELECTRIC CORPORATION Nuclear Service Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
@ 1997 Westinghouse Electric Corporation All Rights Resen/ed 4/97
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i t
i PREFACE.
The following changes have been made to this report:
Revised format to fit current WCAP standards.
Revised Tables 3 and 6 per updated fluences given in reference 5.
C
^
wb Vcrified By:
. E. Terek i
i i
Evaluation of PTS for Prairie Island Unit 2.
ii J
TABLE OF CONTENTS LI ST OF TABLE S................................................ iil 1.0 I NTROD U CTION...........................................
1 2.0 PRESSURIZED THERMAL SHOCK.............................
2 l
3.0 METHOD FOR CALCULATION OF RTpTs.........................
3 4
4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES.......
5 i-5.0 NEUTRON FLUENCE VALUES................................
8 i
6.0 DETERMINATION OF RT VALUES FOR ALL BELTLINE REGION ers M ATE R I ALS..............................................
9 7.0 CON C LU SION S..........................................
12
8.0 REFERENCES
13 Evaluation of PTS for Prairie Island Unn 2 L
iii LIST OF TABLES Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Materials..........................
6 Table 2 Prairie Island Unit 2 Reactor Vessel Beltline Region Material 7
P rope rtie s...........................................
Table 3 Peak Fluence (10" n/cm', E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal Interface for Prairie island Unit 2..............
8 Table 4 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 9
CFR Part 50.61................
Table 5 Calculation of Chemistry Factors Using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1...........
10 Table 6 RTers Calculations for Prairie Island Unit 2 Beltline Region Materials at EOL (35 EFPY).............................. 11 Ev luation of PTS for Prairie Island Unit 2 i
1
1.0 INTRODUCTION
A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water recctors (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 ccts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutrun irradiation. Such an event may produce the propagation of flaws i
postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the v:ssel.
The purpose of this report is to determine the RTers values for the Prairie Island Unit 2 reactor v:ssel using the results of the surveillance Capsule P evaluation. Section 2.0 discusses the PTS Rule and its requirements. Section 3.0 provides the methodology for calculating RTers-Section 4.0 provides the reactor vessel beltline region material properties for the Prairie Island Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RTp13 calculations are presented in Section 6.0. The conclusion and references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively.
i i
l i
1 Evaluation of PTS for Prairie Island Unit 2
-+-
- - m
2 210 PRESSURIZED THERMAL SHOCK The 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 revised PTS Rule"1,'10 CFR Part 50.61, was published in the Federal Register on December 19,1995, with an effectivo date of January 18,1996.
This amendment to the PTS Rule makes three changes:
1.
The rule incorporates in total, and therebre makes binding by rule, the method for determining the reference temperature, R T.m. including treatment of the unirradiated 3
RTmy value, the margin term, and the explicit definition of " credible" surveillance data, t21 which is currently described in Regulatory Guide 1.99, Revision 2,
2.
The rule is restructured to improve clarity, with the requirements section giving only the requirements for the value for the reference temperature for end of life (EOL) fluence, l
RTers-3.
Thermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RTers-The PTS Rule requirements consist of the following:
For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTprs, accepted by the NRC.
for each reactor vessel beltline material for the EOL fluence of the material.
The assessment of RTpr must use the calculation procedures given in the PTS Rule, and must spccify the bases for the projected value of RTpTs for each vessel beltline material. The 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 RTp13 or upon the request for a change in the expiration date for operation of the facility. Changes to RTpr, 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.
The RTyr, screening criterion values for the beltline region are:
270 F for plates, forgings, and axial weld materials, and 300*F for circumferential weld materials.
~$v"luation of PTS for Prairie Island Unit 2
3 3.0 METHOD FOR CALCULATION OF RTp73 RTp13 must be calculated for each vessel beltline inaterial using a fluence value, f, which is the EOL fluence for the material. Equation 1 must be used to calculate values of RT or for u
c ch weld and plate or forging in the reactor vessel beltline.
(1)
wo7 reference temperature for a reactor vessel material in the pre-service or RTwoT(u) =
unirradiated condition Margin to be added to account for uncertainties in the values of RT oT(u), copper M'
N
=
and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2.
M-2/o +o'3 (2) o is the standard deviation for RT rcu)-
u wo o = 0 F when RT orcui is a measured value u
N o = 17'F when RT rtu) is a generic value u
No c is the standard deviation for ARTuor.
3 For plates and forgings:
o = 17*F when surveillance capsule data is not used 3
c = 8.5 F when surveillance capsule data is used 3
For welds:
3 = 28'F when surveillance capsule data is not used 0
0 = 14 F when surveillance capsule data is used 3
o, not to exceed one-half of ARTuor.
ARTuor is the mean value of the transition temperature shift, or change in RTuor, due to irr:diation, and must be calculated using Equation 3.
i ARTwo7=(CF) +f pas 41N (3)
Ev;luation of PTS for Prairie Island Unit 2
4 CF (*F) is the chemistry factor, which is a function of copper and nickel content. CF is given in Table 1 for welds and Table 2 for base metal (plates or forgings) 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.
f is the best estimate 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 RTers.
Equation 4 must be used for determining RTp13 using Equation 3 with EOL fluence values for determining ARTp73 t
(4) l RTers=RTnoT(u)+M + A RT p73 l
1 To verify that RT r for each vessel beltline material is a bounding value for the specific uo r=ctor 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 t:mperature and any related surveillance program results. Results from the plant specific surveillance program must be integrated into the RTuor estimate if the plant-specific surveillance data has been deemed credible.
A material-specific value of CF is determined from Equation 5.
CF= E[Apf[ **#]
(5)
E[f[ **#]
In Equation 5, 'A," is the measured value of ARTuor and "f," is 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 essociated with the vessel weld and the surveillance weld, the measure values of ARTso7 must be adjusted ier differences in copper and nickel content by multiplying them by the ratio of the chemistry fac, tor for the vessel material to that for the surveillance weld.
l 1
1 Ev luation of PTS for Prairie Island Unit 2
i 5
4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES 1
Before performing the pressurized thermal shock evaluation, a review of the latest pl nt-specific material properties for the Prairie Island Unit 2 vessel was performed. The beltline region of a reactor vessel, per the PTS Rulo, is defined as "the region of the reactor v:ssel (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 radiation damage to be considered in the j
selection of the most limiting material with regard to radiation damage".
M;terial property values were obtained from material test certifications from the original j
fcbrication as well as the additional material chemistry tests performed as part of the Prairie ist'nd Unit 2 surveillrace capsule testing program ). The average copper and nickel values ts were calculated for each beltline region material using all of the available material chemistry information as shown in Table 1. A summary of the pertinent chemical and mechanical properties of the beltline region forgings and weld material of the Prairie Island Unit 2 reactor v:ssel is given in Table 2.
l l
1 Ev;tuation of PTS for Prairie Island Unit 2
6 Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Materials A533 Gr. B, CL1 Correlation Monitor ints nediate Shell Lower Shell Inter / Lower Material Forging 22f,29 Forging 22642)
Shell Girth Weld *)
(HSST Plate 02)
Ref.
Cu %
Ni %
Cu %
Ni %
Cu %
Ni %
Cu %
Ni %
J 0.085 0.70 0.082 0.072 0.14 0.68 4
0.068 0.694 0.076 0.071 5
0.068 0575 0.094 0.103 5
0.081 0.087 5
0.078 0.081 6
6 0.075 0.75 7
0.070 0.75 0.085 0.700 7
0.085 0.700 Avg.
0.0725 0.75 0.0782 0.6738 0.0822 0.0828 0.14 0.68 NOTES:
(t) Surveillance program base metal material.
(b) The surveillance weld specimens were made of the same wire and fluy as the intermediate to lower shell circular seam (UM 40 Wire Heat 2721 and UM 89 Flux Lot 1263).
l Evduation of PTS for Prairie Island Unit 2
7 Table 2 Prairie Island Unit 2 Reactor Vessel Beltline Region Material Properties Material Description Cu (%) (*)
Ni (%)
RTuonu) (*F) *)
Intermediate Shell Forging 22829 0.0725 0.75
-4 Lower Shell Forging 22642 0.0782 0.6738
-6 Inter / Lower Shell Girth Weld 0.0.822 0.0828
-31 NOTES:
(
(0) Average values of copper and nickel as indicated in Table 1 on preceding page.
l (b) The RTenu) values for the forgings and weld are measured values and were obtained from Prairie l
Island Unit 2 FSAR..
l Evoluation of PTS for Prairie Island Unit 2
8 i
l 5.0 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the Prairie Island Unit 2 reactor vessel are shown in Table 3. These values were projected using the W
results of the Capsule P radiation analysis. See Section 6.0 of WCAP-14613.
j Table 3 Peak Fluence (10" n/cm', E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal interface for Prairie Island Unit 2 EFPY 0*
17.24 2.44 24 3.11 32 3.89 i
I i
i l
i Evaluation of PTS for Prairie Island Unit 2
9 6.0 DETERMINATION OF RTers VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTp13 values were generah.d for all beltline tcgion materials of the Prairie Island Unit 2 reactor vessel for fluence values at the EOL (35 EFPY).
E ch plant shall assess the RTp73 values based on plant-specific surveillance capsule data.
For Prairie Island Unit 2, the related tiurveillance program results have been included in this PTS evaluation. Specifically, the Prairie Island Unit 1 plant-specific surveillance capsule data for the intermediate shell forging C and weld metal is provided for the following reasons:
1)
There have been four 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 forgings and circumferential weld metal.
As presented in Table 4, chemistry factor values for Prairie Island Unit 2 based on average D1 copper and nickel weight percent were calculated using Tables 1 and 2 from 10 CFR 50.61 Additionally, chemistry factor values based on credible surveillance capsule data are calculated in Table 5. Table 6 contains the RTers calculations for all beltline region materials at 35 EFPY, Table 4 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 CFR Part 50.61 Material Ni, wt %
Chemistry Factor, 'F intermediate Shell Foraina 22829 0.75 45.8 Given Cu wt% = 0.0725 Lower Shell Foraina 22642 0.6738 49.7 Given Cu wt % = 0.0782 Intermediate to Lower Shell 0.0828 45.9 Circumferential Weld Metal Given Cu wt % = 0.0822 Evaluation of PTS for Prairie Island Unit 2
10 l
Table 5 Calculation of Chemistry Factors Using Surveillance Capsule Data Per 10 CFR Part 50.61 8
Material Capsule Capsule f(*)
FF*'
ARTuo/4 FF*ARTuoy FF Intermediate Shell V
0.6206 0.866 35.28 30.552 0.750 Forging 22642 (Axial Orientation)
T 1.199 1.051 29.93 31.456 1.105 R
4.376 1.375 84.73 116.504 1.891 P
4.165 1.365 103.87 141.783 1.863 Intermediate Shell V
0.6206 0.866 32.89 28.483 0.750 Forging 22642 (Tangential T
1.199 1.051 55.69 58.530 1.105 Orientation)
R 4.376 1.375 90.02 123.778 1.891
)
P 4.165 1.365 99.91 136.377 1.863 SUM 667.463 11.210 CFm,wm = I(FF
- ARTuo7) + I(FF )
8
= 667.463 + 11.218 = 59.5 "F l
Weld Metal
- V 0.6206 0.866 70.07 60.681 0.750 T
1.199 1.051 57.73 60.674 1.105 R
4.376 1.375 100.31 137.926 1.891 P
4.165 1.365 96.24 131.368 1.863 4
SUM 390.649 5.609 8
- CFw,
., = I(FF
- ARTuo7) + I(FF )
= 390.649 + 5.609 = 69.6 F NOTE 1 8
(1) f = fluence (10" n/cm, E > 1.0 MeV). All updated fluence values were taken from the Capsule W
P analysis. WCAP-14613.
(b)
FF = fluence factor = f 588-M9 W
-(c)
ARTer values were obtained from the Capsule P analysis.
(d)
The reactor vessel intermediate to lower shell circular weld seam was made with the same weld wire and flux as the surveillance weld specimens. These welds were made with UM 40 Wire Heat 2721 and UM 89 Flux Lot 1263. A review of the available Westinghouse records indicates that only copper and nickel data points available for this weld metal are chemical analyses performed on the surveillance weld metal. Therefore, the ratio procedure given in 10 CFR Part 50.61 was not used in these calculations (i.e. the ratio is assumed to be 1.0).
Eviluation of PTS for Praltie Island Unit 2
11 1
Table 6 RT,13 Calculations for Prairie Island Unit 2 Beltline Region Materials at EOL (35 EFPY) mummmmmuumummumummmmmu-mammemammmmmummummmmum Material CF f'"
FFS)
RTwortu/4 M
ART,13 RT,13 intermediate Shell 45.8 F 4.18 1.366
-4 F 34*F 62.6 F 93*F Forging 22829 Lower Shell Forging 49.7*F 4.18 1.366
-6*F 34*F 67.9'F 96 F 22642 Using Surveillance 59.5 F 4.18 1.366
-6 F 17 F 81.3 F 92 F Capsule Data Circumferential Weld 45.9 F 4.18 1.366
-31'.
56*F 62.7 F 88 F Seam Using surv. capsule 69.6 F 4.18 1.366
-31*F 28*F 95.5* F 93 F data l
NOTES:
(2) f = peak clad / base metal interface fluence (10 n/cm', E > 1.0 MeV) at 35 EFPY (b)
FF = 10 2' ' ' """* "
(c)
RTuor, values are measured values.
Evtluation of PTS for Prairie Island Unit 2 m
1 4
12.
-'7 0 CONCLUSIONS-As shown in Table 6, all of the beltline region materials in the Prairie Island Unit 2 reactor vessel have EOL RTm values well below the screening criteria values of 270 F for plates or forgings and longitudinal welds and 300 F for circumferential welds at EOL (35 EFPY).
i Evaluation of PTS for Prairie island Unit 2 -
13
8.0 REFERENCES
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, effective January 18,1996.
2.
Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988.
3.
WCAP-8193, " Northern States Power Co. Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko et. al., September 1973.
4.
WCAP-11343, " Analysis of Capsule R from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko et. al.,
December 1986.
l 5.
WCAP-14613 Rev.1, " Analysis of Capsule P from the Northern States Power Company i
Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program", T.J. Laubham.
April 1997.
6.
Societe Des Forges Et Ateliers Du Creusot Usines Schneider, Report No. 3.5.7, Rev.
1, item C, Heat No. 22829, Order No. 700 814/54, Dated September 8,1970.
7.
Societe Des Forges Et Ateliers Du Creusot Usines Schneider, Report No. 4.5.7, Rev.
1, Item D, Heat No. 22642, Order No. 700 814/54, Dated June 24,1970.
l Evaluation of PTS for Prairie Island Unit 2
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