ML20141D522
| ML20141D522 | |
| 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-14781, WCAP-14781-R01, WCAP-14781-R1, NUDOCS 9705200132 | |
| Download: ML20141D522 (26) | |
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Westinghouse Non-Proprietary Class 3 WCAP-14781 i
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Revision 1 P
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EVALUATION OF
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PRESSURIZED L
THERMAL SHOCK
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FOR PRAIRIE ISLAND UNIT 1 1
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14781, Revision 1 Evaluation of Pressurized Thermal Shock for Prairie Island Unit 1 I
T. J. Laubham April 1997 Work Performed Under Shop Order NLDP-108 Prepared by Westinghouse Electric Corporation for Northem States Power Company Approved:
C. H. Boyd, Managd3 N Engineering & Materials Technology Approved:
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A D. A. HowellTManager \\
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Mechanical Systems Integration WESTINGHOUSE ELECTRIC CORPORATION Nuclear Service Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
@ 1997 Westinghouse Electric Corporation All Rights Reserved 4/97
i PREFACE The following changes have been made to this report:
Revised Tables 3 and 6 per updated flueces given in reference 5.
7 Verified By:
mD E. Terek 1
Evaluation of PTS for Prairie Island Unit 1
i ii i
9a TABLE OF CONTENTS I
i LI ST OF TABLE S................................................. iii
- i i
'1 I NTROD UCTION...........................................
1 4
i l
2 PRESSURIZED THERMAL SHOCK.............................
2 1
j 3
METHOD FOR CALCULATION OF RTers.........................
3 4
VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES.......
5 i
5 NEUTRON FLUENCE VALUES................................
8 6
DETERMINATION OF RTyrs VALUES FOR ALL BELTLINE REGION M ATE R I ALS..............................................
9 7
CON C LU SION S......................................
12 8
REFERENCES.....................................
13 a
1 i
i i
Evaluation of PTS for Prairie Island Unit 1.
i
(
iii LIST OF TABLES
. Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Materials................................
6 Prairie Island Unit 1 Reactor Vessel Beltline Region Material Table 2
]
P rope rtie s...........................................
7 1
y i
Table 3 Peak Fluence (10" n/cm', E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal Interface for Prairie Island Unit 1..............
8 i
Table 4 Interpolation of Chemistry Factors Using Tables 1 and 2 of 10 C F R Pa rt 5 0.61.......................................
9 t
Table 5 Calculation of Chemistry Factors Using Surveillance Capsule Data i
Per Regulatory Guide 1.99, Revision 2, Position 2.1...........
10 i
a Table 6 RTeis Calculations for Prairie Island Unit 1 Beltline Region i
Materials at EOL (35 EFPY).............................. 11 i
I I
-]
4 i
i Evaluation of PTS for Prairie Island Unit 1
l J
1 INTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water r: actors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by I
cignificant 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 neutron irradiation. Such an event may produce the propagation of flaws 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 RTers values for the Prairie Island Unit 1 reactor vessel using the results of the surveillance Capsule S evaluation. Section 2 discusses the PTS Rule and its requirements. Section 3 provides the methodology for calculating RTprs.
Section 4 provides the reactor vessel beltline region material properties for the Prairie Island Unit 1 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5. The results of the RTer, calculations are presented in Section 6. The conclusion and references for the PTS evaluation follow in Sections 7 and 8, respectively.
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I Evaluation of PTS for Prairie Island Unit 1 1
2 2
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,10 CFR Part 50.61, was published in the Federal Register on N
December 19,1995, with an effective date of January 18,1996.
This amendment to the PTS Ru!e makes three changes:
1.
The rule incorporates in total, and therefore makes binding by rule, the method for j
deterrnining the reference temperature, RTuo7, including treatment of the unirradiated RTuor value, the margin term, and the explicit definition of " credible" surveillance data, 5
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, RTp13 3.
Thermal annealing is identified as a method for mitigating the effects of neutron i
irradiatica, 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 RTers, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material.
The assessment of RT must use the calculation procedures given in the PTS Rule, ers and must specify the bases for the projected value of RTprs 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 RTp7s or upon the request for a change in the expiration date for operation of the facility. Changes to RT values are significant if either the previous value or the p1s 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 RTpys screening criterion values for the beltline region are:
270*F for plates, forgings, and axial weld materials, and 300*F for circumferential weld materials.
Evaluation of PTS for Prairie Islana Unit 1
3 3
METHOD FOR CALCULATION OF RT s er RTp7, 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 RT or for u
each weld and plate or forging in the reactor vessel beltline.
RTuo7 RTuo7 ppm + A RTuor (1) reference temperature for a reactor vessel material in the pre-service or RTwortu) =
unirradiated condition Margin to be added to account for uncertainties in the values of RT orcu), copper M
=
w and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2.
i M 2/o +o$
(2) u o is the standard deviation for RTwortu>-
u o = 0 F when RTwoT<u) is a measured value u
o = 17 F when RTwortu) is a generic value u
j c is the standard deviation for ARTuo7 3
For plates and forgings:
o, = 17 F when surveillance capsule data is not used
)
c3 = 8.5 F when surveillance capsule data is used For welds:
o, = 28 F when surveillance capsule data is not used o, = 14*F when surveillance capsule data is used o not to exceed one-half of ARTum.
3 ARTuo7 is the mean value of the transition temperature shift, or change in RTuo1, due to irradiation, and must be calculated using Equation 3.
ARTuo7=(CF)*f $2e4'N (3)
Evaluation cf PTS for Prairie Island Unit 1
4 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.
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 RTp73 Equation 4 must be used for determining RTer, using Equation 3 with EOL fluence values for determining ART ys.
r RT rs=RTuo7g +M + A RT s (4) e FT To verify that RT r for each vessel beltline material is a bounding value for the specific wo 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 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= I[A,+f[ *"]
(5)
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 tha 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 cssociated with the vessel weld and the surveillance weld, the measure values of ART oy s
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.
Evaluation of PTS for Prairie Island Unit 1
5 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 Prairie Island Unit 1 vessel was performed. The beltline region of a reactor vessel, per the PTS Hule, is defined as "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 radiation damage to be considered in the selection of the most limiting material with regard to radiation damage".
Material property values were obtained from material test certifications from the original fabrication as well as the additional material chemistry tests performed as part of the Prairie Island Unit 1 surveillance capsule testing program'l. The average copper and nickel values i
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 1 reactor vessel is given in Table 2.
i j
1
- Evaluation of PTS for Prairie Island Unit 1
6 l
l Table 1 Calvulation of Average Cu and Ni Weight Percent Values for Beltline Region Materials Intermediate Shell Lower Shell Forging Inter / Lower Shell Forging C D
Circumferential Weld ('*)
W Ref.
Cu %
Ni %
Cu %
Ni %
Cu %
Ni %
6 0.06 0.72 6
0.06 0.72 l
7 0.07 0.66 l
7 0.065 0.66 j
3 0.13 4
0.13 0.09 4
i 5
0.078 0.956 0.149 0.138 5
0.138 0.118 l
5 0.143 0.091 Avg.
0.07 0.80 0.07 0.66 0.14 0.11 l
f NOTES:
(a) Surveillance program base metal material.
(b) The surveillance weld specimens were made of the same wire and flux as the intermediate to lower shell circular seam (Heat 1752 and UM 89 Flux Lot 1230).
l Evaluation of PTS for Prairie island Unit 1
7 i
l Table 2 Prairie Island Unit 1 Reactor Vessel Beltline Region Material Properties m
t Material Description Cu (%) (*)
Ni (%) ")
RTuoTcu, ('F) *'
l intermediate Shell Forging C 0.07 0.80 14 Lower Shell Forging D 0.07 0.66
-4 Inter./ Lower Shell Circumferential Weld 0.14 0.11 0'4 t
NOTES:
(a) Average values of copper and nickel as indicated in Table 1 on preceding page.
(b) The RT ortu) values for the forgings and weld are measured values.
w (c) Per NSP's past reactor vettel integrity calculations, an initial RTuor of 0 *F (Rather than -13 F) was used for conservatism.
i 4
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l Evaluation of PTS for Prairie island Unit 1
8 5
NEUTRON FLUENCE VALUES The calculated fast neutron fluenco (E > 1.0 MeV) values at the inner surface of the Prairie Island Unit 1 reactor vessel are shown in Table 3. These values were projected using the results of the Capsule S radiation analysis. See Section 6.0 of the Capsule S analysis report, N
Table 3 Peak Fluence (10 n/cm', E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal Interface for Prairie Island Unit 1 EFPY O'
18.12 2.42 24 2.93 35 3.90 i
4 Evaluation of PTS for Prairie Island Unit 1
- v
9 6
DETERMINATION OF RTers VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTets values were generated for all beltline r:gion materials of the Prairie Island Unit 1 reactor vessel for fluence values at the EOL (35 EFPY).
Each plant shall assess the RTp1s values based on plant-specific surveillance capsule data.
For Prairie Island Unit 1, the related surveillance program results have been included in this PTS evaluation. Specifically, the Prairle 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 (See Reference 5).
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 1 based on average 14 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 RT rs calculations for all beltline region materials j
e l
ct 35 EFPY.
1 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 Fomina C 0.80 44 Given Cu wt% = 0.07 Lower Shell Foraina D 0.66 44 Given Cu wt % = 0.07 0.00 61 Weld Metal Given Cu wt % = 0.14 0.11 70.9 0.20 79 i
Evaluation of PTS for Prairie Island Unit 1
10 Table 5 Calculation of Chemistry Factors Using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1 Material Capsule Capsule f)
FF )
ARTuor")
F F*AP.'i FF' S
nor intermediate V
0.5630 0.839 24.07 20.19 0.704 Shell Forging C (Axial)
P 1.318 1.077 33.98 36.60 1.160 R
4.478 1.380 84.18 116.17 1.904 S
4.017 1.357 74.27 100.78 1.841 Intermediate V
0.5630 0.839 56.36 47.29 0.704 Shell Forging C (Tangertial)
P 1.318 1.077 23.11 24.89 1.160 R
4.478 1.380 95.85 132.27 1.904 S
4.017 1.357 101.46 137.68 1.841 SUM 615.87 11.22 CF,
,wm = I(FF
- ARTuo,) + WM
= 54.9 F Weld Metal
- V 0.5630 0.839 34.38 28.84 0.704 P
1.318 1.077 45.15 48.63 1.160 R
4.478 1.380 122.47 169.01 1.904 S
4.017 1.357 160.43 217.70 1.841 SUM 464.18 5.61 CFw,,m,,, = I(FF
- ARTuor) + I(FF')
= 82.7*F m
NOTES:
(a) f = fluence (10" n/cm', E > 1.0 MeV). All updated fluence values were tsen from the Capsule W
S analysis (WCAP 14779 ).
(b)
FF = fluence factor = f 528'"N (c)
ARTuoy values were obtained from CVGRAPH Version 4.1 plots (See WCAP-14779W).
(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 (Wire Um 89, heat number 1752, UM 89 flux, batch no.1230). The ratio procedure was not used since there is no clear evidence to conclude that the Cu and Ni content of the vessel weld differs from that of the survei!!ance weld material.
Evaluation of PTS for Prairie Island Unit 1
11 Table 6 RTyr Calculations for Prairie Island Unit 1 Beltline Region Materials at EOL (35 EFPY) mummmmmmmmmmmmmumum ummmmm -
mmmmm-mmmmmmmum Material CF fl')
FF*)
RTuoru )
M ARTeTs RT,7s k
(*F)
(F)
(F)
(*F)
('F) inter. Shell Forging C 44.0 3.90 1.35 14 34 59.4 107 Using Surveillance 54.9 3.90 1.35 14 17 74.1 105 Capsule Data Lower Shell Forging D 44.0 3.90 1.35
-4 34 59.4 89 Circ. Weld 70.9 3.90 1.35 0
56 95.2 151 Using Surveillance 82.7 3.90 1.35 0
28 112.0 140 Capsule Data NOTES:
(a) f = peak clad / base rnetal interface fluence (10 n/cm', E > 1.0 MeV) at 35 EFPY (h) pp, f (c as. o to bo t)
(c)
RTuorcui values are measured values.
Evaluation of PTS for Prairie Island Unit 1
_ _ _... ~.... _ _... _ _. _ _ _ _... _. _.
12 i
7 CONCLUSIONS r
As shown in Table 6, all of the beltline region materials in the Prairie Island Unit 1 reactor v:ssel have EOL RTm values well below the screening criteria values of 270'F for plates or l
forgings and longitudinal welds and 300 F for circumferential welds at EOL (35 EFPY).
1 L
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Evaluation of PTS for Prairie Island Unit 1
13 i
8 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-8086, " Northern States Power Co. Prairie Island Unit No.1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko and D.J. Lege, June 1973.
4.
WCAP 11006, " Analysis of Capsule R from the Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program", R.S. Boggs, et. al.,
February 1986.
5.
WCAP-14779 Rev.1," Analysis of Capsule S from the Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program", T.J. Laubham, April 1997.
6.
Societe Des Forges et Ateliers Du Creusot usines Schneider, Chemical Analysis Report No.17-9-2, NSP shell course C, heat 21918/38566.
7.
Societe Des Forges et Ateliers Du Creusot usines Schneider, Chemical Analysis Report No.15-81, NSP shell course D, heat 21887/38530.
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Evaluation of PTS for Prairie Island Unit 1
~-
14 APPENDIX A:
SURVEILLANCE DATA CREDIBILITY EVALUATION Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the methodology for calculating the adjusted reference temperature and Charpy upper shelf energy of reactor vessel beltline materials using surveillance capsule data.
The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
To date, there have been four surveillance capsules removed from the Prairie Island Unit 1 r: actor vessel. This capsule data must be shown to be credible, in accordance with the i
discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.
The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Prairie Island Unit 1 reactor vessel surveillance data and determine if the Prairie Island Unit 1 surveillance data is credible.
l Criterion 1:
Materials In the capsules should be those judged most Ilkely to be controlling with regard to radiation embrittlement.
The beltline region of tha reactor vessel is defined in Appendix G to 10 CFR Part 50" ),
" Fracture Toughness Requirements", December 19,1995 to be:
"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 radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."
The Prairie Island Unit 1 reactor vessel consists of the following beltline region materials:
a)
Intermediate shell forging C, heat number 21918/38566 b)
Lower shell forging D, heat number 21887/38530 c)
Circumferential weld, heat number 1752, um 89 flux, batch number 1230 Evaluation of PTS for Prairie Island Unit 1
f' 15 J
Per WCAP-8086*, the Prairie Island Unit 1 surveillance program was based on ASTM E185-70, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Per Section i
3.1.2 of ASTM E185-70, "A minimum test program shall consist of specimens taken from the following locations (1) base metal of one heat, incorporated in the highest flux location of the reactor vessel, that has the highest initial ductile-brittle transition temperature, (2) weld metal, fully representative of fabrication practice used for the welds in the highest flux location of the reactor vessel, (weld wire or rod, and flux must come from one of the heats used in the highest flux region of the reactor vessel) and (3) the heat-affected zone of the weldments noted above."
Therefore, at the time the Prairie Island Unit 1 surveillance capsule program was developed, intermediate shell forging C was judged to be most limiting based on the above recommendations and was utilized in the surveillance program.
The surveillance program weld for Prairie Island Unit 1 was fabricated using the same heat of weld wire used to fabricate the circumferential weld seam (heat 1752). The results of l
mechanical property tests performed on the surveillance weld are considered to be representative of the property changes expected in the reactor vessel beltline seams.
Therefore, the materials selected for use in the Prairie Island Unit 1 surveillance program were those judged to be most likely controlling with regard to radiation embrittlement according to i
the accepted methodology at the time the surveillance program was developed. The Prairie Island Unit 1 surveillance program meets this criteria, j
t Criterion 2:
Scatter in the plots of Charpy energy versus temperature for the Irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-Ib temperature and upper shelf energy, unambiguously.
i Plots of Charpy energy versus temperature for the unirradiated condition are presented in WCAP-8086*, " Northern States Power Company Prairie Island Unit No.1 Reactor Vessel Radiation Surveillance Program," dated June 1973. Plots of Charpy energy versus temperature for the irradiated conditions are presented in the WCAP reports (Ref.14,15,4 &
- 5) for Capsules V, P, R and S.
Based on engineering judgement, the scatter in the data presented in these plots is small enough to determine the 30 ft-lb temperature and the upper shelf energy of the Prairie Island Unit 1 surveillance materials unambiguously. Therefore, the Prairie Island Unit 1 surveillance program meets this criteria.
i Evaluation of PTS for Prairie Island Unit 1
16 CrRerlon 3:
When there are two or more sets of surveillance data from one reactor, the scatter of 6RTuor values about a best-fit Ilne drawn as described in i
Regulatory Position 2.1 normally should be less than 28'F for welds and 17'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the I
data fall this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy If the upper shelf can be clearly determined, follow' og the definition given in ASTM E185-82.
l The least squares method, as described in Regulatory Position 2.1, will be utilized in determining a best-fit line for this data to determine if this criteria is met.
i Pralrie Island Unit 1 Surveillance Capsule Data Calculation of Best-Fit une as Described Table A1W in Position 2.1 of Regulatory Guide 1.99, Revision 2 i
Material Capsule f*
FF x.iRT.
FF' (x)
(y)
(xy)
(x')
Intermediate Shell V
0.563 0.839 24.07 20.19 0.704 f
Forging C
]
(Axial)
P 1.318 1.077 33.98 36.60 1.160 f
R 4.478 1.380 84.18 116.17 1.904 S
4.017 1.357 74.27 100.78 1.841 V
0.563 0.839 56.36 47.29 0.704 Intermediate Shell P
1.318 1.077 23.11 24.89 1.160 Forging C (Tangential)
R 4.478 1.380 95.85 132.27 1.940 S
4.017 1.357 101.46 137.68 1.841 fg 9.306 493.28 615.87 11.218 V
0.563 0.839 34.38 28.84 0.704 P
1.318 1.077 45.15 48.63 1.160 R
4.478 1.380 122.47 169.01 1.904 Weld Metal S
4.017 1.357 160.43 217.70 1.841 r
4.653 362.43 464.18 5.609 w
te')TES:
(a) f = Fluence (10 n/cm', E > 1.0 MeV)
(b) FF = Fluence Factor = f*""*'*
i (c)
Va!ues of I. FF and ART,, were taken from Table 5 hemn.
Evaluation of PTS for Prairie Island Unit 1
17 Per the 27* Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method of least squares, the values b, and b, are obtained by soMng the normal equations n b, + b, 4 = Iy, and b,Ix,+ b,Ix, = y y, These equations can be re-written as follows:
n n
E y, = an + bE x, tal i=1 and n
n n
E x,y, - aE x, + bE x,2 i=1 l=1 i=1 Intermediate Shell Foraina C:
Based on the data provided in Table A1, these equations become:
1.)
493.28 = 8a + 9.306b or a = 61.66 1.16b l
and 2.)
615.81 = 9.306a + 11.218b
- Thus, by substituting Eq.1 into Eq. 2, b = 107.1. Now, enter b (= 107.1) into Eq.1 and a = -62.9. Therefore, the equation of the straight line which provides the best fit in the ;ense of least squares is:
Y' = 107.1 (X) 62.9 The error in predicting a value Y corresponding to a given X value is: e = Y - Y'.
Evaluation of PTS '~ Prairie Island Unit 1
18 Table A2:
Best Fit Evaluation for Intermediate Forging Base Material (Orientation)
ART.
Best Fit ART.
Scatter of ART.
FF (30 ft-lb)(*F)
(*F)
(*F)
-umummmmmmmmmmmuumum Intermediate Shell 0.839 24.07 27.0
-2.9 1.077 33.98 52.4
-18.4 a
1.340 44.18 84.9
-0.7 1.357 74.27 82.4
-8.1 Intermediate Shell 0.839 56.36 27.0 29.4
)
1.077 23.11 52.4
-29.3 a genti !)
1.380 95.85 84.9 11.0 1.357 101.46 82.4 19.1 The scatter of ART, values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 17*F for base metal. However, even if the fluence range is large, the scattar should not exceed twice this value (34 F).
As shown above, the error is within 34*F of the best-fit line. Therefore, this criteria is met for the Prairie Island Unit 1 surveillance forging material.
Weld Metal:
Based on the data provided in Table A1 the equations become:
1.)
362.43 = 4a + 4.653b or a = 90.61 1.163b and 2.)
464.18 = 4.653a + 5.609b i
Thus, by substituting Eq.1 into Eq. 2, b = 216.7. Now, enter b (= 216.7) into Eq.1 and a = 161.5. Therefore, the equation of the straight line which provides the best fit in the sense of least squares is:
Y = 216.7 (X) 161.5 i
The error in predicting a value Y corresponding to a given X value is: e = Y Y Evaluation of PTS for Prairie island Unit 1 l
l
19 Table A3: Best Fit Evaluation for Weld Metal Base Material ART,
Best Fit ART, Scatter of ART, FF (30 ft-lb)(*F)
('F)
('F) mummmmusammmmu---
I Weld Metal 0.839 34.38 20.3 14.1 1.077 45.15 71.9
-26.8 1.380 122.47 137.5 15.0 1.357 160.43 132.6 27.84 4
The scatter of ART, values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 28#F. However, even if the fluence range is large, the scatter should not exceed twice this value (56 F). As shown above, the error is within 56 F of the best-fit line. Therefore, this criteria is met for the Prairie Island Unit 1 surveillance weld material.
Criterion 4:
The Irradiation temperature of the Charpy specimens in the capsule should match the vescel wall temperature at the cladding / base metalinterface within +/ 25'F.
a The Prairie Island Unit 1 capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25 F.
Critenon 5:
The surveillance data for the correlation mordtor materialin the capsule should fall within the scatter band of the data base for that material.
Correlation monitor material was supplied by the Oak Ridge National Laboratory from plate material used in the AEC-sponsored Heavy Section Steel Technology (HSST) Program. This material, which was obtained from a 12-inch thick A533 Grade B Class 1 plate (HSST Plate 02), was provided to Subcommittee 11 (of ASTM Committee E 10 on Radioisotopes and Radiation Effects) to serve as correlation monitor material in reactor vessel surveillance progra.as.
Tha plate was produced by the Lukens Steel Company and heat treated by Combustion Engineering, Inc.
Figure A1 contains a plot of the residual (measured shift minus Regulatory Guide 1.99, Revision 2 shift) versus capsule fluence data. The plot shows the Prairie Island Unit 1 data as solid points. The data has been shifted such that the mean value is at zero and the two-sigma bound at 45 F. Ah of the Prairie Island Unit 1 correlation monitor material data falls within the two-sigma scatter band of the A533 Grade B Class 1 data per this criterion.
Evaluation of PTS for Prairie Island Unit 1
20 l
~
t 1
2
.\\ '
4 d
t i
i i
i I
i I
i J
1 I
1 1
FIGURE A1:-
Residual vs. Fast Fluence for HSST Plate 02 Materials
]
)
4 7
Evaluation of PTS for Prairie Island Unit 1-
21 i
==
Conclusion:==
Based on the preceding responses to the criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of
. engineering judgement, the Prairie Island Unit 1 surveillance data s credible.
i e
i l
i e
i i
i l
i l
s Evaluation of PTS for Prairie island Unit 1 2
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