ML20101N159
| ML20101N159 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/06/1992 |
| From: | Gates W OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, LIC-92-320R, NUDOCS 9207090201 | |
| Download: ML20101N159 (18) | |
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c Omaha Public Power District 444 South 16th Street Mall Omaha, Nebraska 68102-2247 402/636-2000 July 6, 1992 LIC-92-203R U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station Pl-137 Washington, DC 20555
References:
1.
Docket No. 50-285 2.
NRC Generic Letter 92-01, Revision 1 - " Reactor Vessel Structural Integrity,10CFR50.54(f)" dated February 28, 1992 Gentiemen:
SUBJECT:
Response to NRC Generic letter (GL) 92-01, Revision 1: Rnctor Vessel Structural Integrity Attached is the Omaha Public Power District (0 PPD) response to GL 92-01, Revision 1.
OPPD believes that appropriate actions have been taken to comply with the requirements of GL 92-01, 10 CFR 50.60 and 10 CFR 50.61.
This response is being submitted under oath in accordance with the requirements of GL 92-01, Revision 1.
If you have any questions, please contact me.
Sincer ely,
- b. N /
W. G Gates Division Manager Nuclear Operations Division WGG/grc Attachment c:
LeBoeuf, Lamb, Leiby & MacRae J. L. Milhoan, NRC Regional Administrator, Region IV R. P. Mullikin, NRC Senior Resident Inspector S. D. Bloom, NRC Acting Project Manager I8 i I. ].
a d5-5124 Empioyment with Equa! Opcortun:ty
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O 9207090201 920706 Male /Femate jf v Q
PDR ADOCK 05000285 P
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4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Omaha Public Power District Docket No. 50-285 (Fort Calhoun Station UnitNo.1)
AFFIDAVIT W. G. Gates, being duly sworn, hereby deposes and says that he is the Division Manager - Nuclear Operations of the Omaha Public Power District; that as such he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached information concerning response to Generic Letter 92-01, Revision 1; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information, and belief.
5 54, l
W. G. Gates Division Manager Nuclear Operations l
t STATE OF NE8RASKA MBRSHMtof Remma LPATmenwAtuNG COUNTY OF DOUGLAS Subscribed and sworn to before me, a Notary Public _in and for the State of Nebraska on this-G>*
day of July, 1992.
l 7)
Notary Public
o
(
Attachment.
LIC-92-203R J
Page 1 RESPONSE TO GENERIC LETTER 92-01. Revision 1
References:
1.
LetterfromOPPD(W.C. Jones)toNRC(H.R.Denton) dated January 23,1981(LIC-81-0011)
Letter from 0! tD (W. C. Jones April 25,1984 (LIC-84-124) )_ to NRC (D. G. Eisenhut) dated 2.
3.
Letter from-OPPD (R. L. Andrews) to NRC: (A. C. Thadani) dated January 23,1986(LIC-86-024)_
l 4.
Letter from OPPD (R. L. Andrews) to NRC (A. C. Thadani) datedMay7,1986(LIC-86-178) 5.' -
Letter from NRC (W.- A. Paulson) to OPPD (R. L. Andrews) dated March 5, 1987 6.
Letter from 0 PPD (R.-L.JAndrews) to NRC (Document Control Desk)datedDecember 21,1987(LIC-87-692)
.7.
NRC Generic Letter 88-11 "NRC Position on; Radiation-Embrittlement of Reactor vessel Materials and Its Impact o_n Plant Operations" 8.
Letter from 0 PPD-(R. L. Andrews) to NRC (Document Control-
~
Desk)datedOctober 28,1988(LIC-88-950) 9.
Letter from 0 PPD (W. G. Gates) to NRC (Document Control Desk)datedMay5,1992(LIC-92-175R)
- 10. Report from Combustion Engineering (R. R. Mills) to 0 PPD (W.
i C. Jones) dated March 28, 1984 l
Ouestion 1
- Certain addressees-are requested to provide the following information regarding Appendix H to 10 CFR Part 50:
i
?
Addressees who do not have a surveillance program meeting ASTM E:185-73, l
-79, or and who do not have an integratedLsurveillance1 program approvedbytheNRC_(seeenclosure=2),arerequestedtodescribeactions taken,. or to be taken, to ensure compliance with Appendix H to 10 CFR Part
- 50. Addressees who plan to revise -the L surveillance program - to L meet
-Appendix H to 10 CFR Part 50 are requested to: indicate when the revised 4
- program will be submitted to the NRC staff for review.'JIf tho surveillance
_ program is not to be revised : to-meet - Appendix; H to -10 CFR y Part 50, o
. addressees are requested to - indicatei wmen ? they-plan to request an
- exemption from Appendix-H to-10 CFR Part 50 under 10 CFR 50.60(b).
l QPPD R'esponse:
ASTM E-185 was originally' issued in 1961 and was; revised in 1966,[1970,_1973, 1979,- and 1982. -' Appendix H c to 10 CFR Part 50, as first published in -1973, outlines the requirements:for compliance with the pertinent edition of--ASTM E-'
,[
185. The Fort Calhoun Station reactor vessel was-designed and. fabricated prior to -the-Summer 1972_ Addenda of the 1971. Edition of Section III of the ASME Pressure Vessel Code. ASTM E 185-1966 was the standard in place at the time the surveillance program was developed.
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Attachment LIC 92-203R Page 2 The surveillance program does comply with the 1966 version of ASTM E 185 that was current on the issue date of the ASME Code to which the reactor vessel was purchased, and the testing was conducted using the appropriate ASTM standard requirements. The surveillance capsules tested to date were analyzed consistent with-the latest editions of ASTM E 185. - The W-225~ capsule analysis was based on the 1973 version of ASTM E 185. The W-265 capsule analysis was based on the 1979 version of ASTM E 185. No testing of surveillance capsules after July 26, 1983' has been performed.
The surveillance program was approved-during the FSAR licensing process, l Specifications.
Thus, program has been approved as part of and the capsule testing the plant Technica the _ surveillance program for Fort.
Calhoun Station meets.the requirements of Appendix H to 10 CFR Part 50.
The surveillance weld was fabricated to'be representative of the actual reactor vessel welds, although not made with an exact beltline weld wire heat. To date, a
the surveillance data obtained fren this material appears to be predictable.
OPPD is currently assessing whet;sr or not the - surveillance weld -will be i
effective in monitoring neutron embrittlement of the Fort Calhoun Station reactor vessel. This assessment will include attempts to locate a source of material for use in an integrated or supplemental surveillance program. The assessment will-be completed and results submitted to the NRC by September 18, 1993.
Question 2.a.
Certain addressees are requested to provide the following information-regarding Appendix G to 10 CFR iart 50:
Addressees of plants for w'hich the Charpy upper shelf energy is predicted to be less than 50 foot pounds at the end of their licenses.using the' guidance in Paragraph C.1.2 or C.2.2 in Regulatory Guide 1.99 Revision 2 i
l-are requested to-- provide to the -NRC the Charpy upper -shelf-energy.
predicted for December:16.-1991, and for the end of their current license-I for the limiting beltline weld and the plate or forging, and are re t
-to describe the actions taken pursuant to Paragraphs IV.A.1 or~ quested Appendix G to 10 CFR Part 50.
V.C of t
u OPpD Response:
t TP Charpy upper shelf energy (June 2008),. based upon' standard fuel managemen is projected to be 52 'ft-lbs at the.end of the current Fort Calhoun license i
With low and extreme low leakage core designs implemented in Cycle 8 and Cycle 14 respectively, additional margin exists-above the 50 ft-lb. limit. Thus Fort i
Calhoun Station is-not predicted to fall below the 50 ft-lb limit and no further L
actions are proposed pursuant to Paragraphs =IV.A.1 or V.C of Appendix G to 10 CFR L
Part 50.
1 l
l j
c.
r Attachment 1
LIC-92-203R Page 3 Ouestion 2.b.
Addressees whose reactor vessels were constructed to an ASNE Code earlier than the Summier 1972 Addenda of the 1971-Edition are recuested to describe the considerations given to the following material propert es in their evaluations perforsed pursuant to 10 CFR 50.61 and Paragraph III. A-of-10 CFR Part. 50, Appendix Gt j
(1) the results from all Charpy and drop weight tests for all unirradiated beltline materials, and the unieradiated reference' temperature for each beltline material, and the method for' determining. the unitradiated t
reference temperature from the Charpy and drop weight test;-
(2) the heat treatment-received by all beltline and surveillance materials; (3) the heat number for each beltline plate or forging and the heat number of wire and flux lot number used to fabricate the surveillance weld;-
(4) the heat number for each surveillance plate or forging and heat number of wire and flux lot number used to fabricate the surveillance weld; (5) the chemical composition, in particular the weight in percent of copper,
- nickel, material; phosphorous, and sulphur for each heitline and surveillance and (6) the heat number of the wire used for determining the weld and ch m! al compositionifdifferentthanItem(3)above.
OPPD Responset i
The generic values det cribed in CEN-189 " Evaluation of Pressurized Thermal Shock i
Effects Due to Smail Break LOCAs with Loss of Feedwater~ for Combustion Engineering NSSS", Appendix A, December 1981, Table A6-1, provide the initial reference temperature, RTndt. For the plate materials, the RT was determined i
using transversely oriented Charpy impact specimens or by conver"Ong longitudinal impact ce. using Branch Technical Position MTEB 5-2 For the weld material, the i
RT was determined using the weld qualification test results benchmarked to the ndt i
surveillance weld for the vessel.
The'RT for.the vessel. beltline welds is' basedonagenericval.veasgiveninCEl-Ikt Section 6.
l Compilation of reactor vessel beltline data requested in. question 2.b item (1) through item (6) is provided in the following Tables 1-10.1 This information was obtained from the original. fabrication records for the vessel and surveillance-materials, p
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Attachment LIC 92-203R Page 4 Table 1 Plant Name:
Fort Calhoun Station
==
Description:==
Reactor Vessel Beltline Plate D-4802 1 Hett No.:
C-2585-3 Specification: SA 533 Grade B, Class 1 j
Supplier:
Lukens Fabricator:
Combustion Engineering Heat Treatment:
Run No.
Max Temp.
Min. Temp.
Temp Range Time Quench Run 10 1
IG00' F 1553*F 50*F 4 hrs WC Austenizing 2
1250'F 1200*F 50'F 4 hrs AC Tempering 3
Il75' F ll25'F 50* F 40 hrs FC Stress Relief Composition (w/o)*:
C Mn P
S Si Ni Cr Mo Cu Cb A1
.21 1.27
.011
.015
.23
.56
.08
.49
.12
<.01
.02 I
Fracture Toughness:
Drop Weight T.,
-50*F (CEN-189,AppendixA,TableA6-1)
Initial RT,
0* F (MTEB Position 5-2, paragraph l.l(3)b)
Initial Upper Shelf 75.4 ft-lb (MTEB Position 5-2, paragraph 1.2)
Energy
- Reference 2 0
l
-+,
Attachment LIC-92 203R Page 5 1
i IQhlL2 i
Plant Name:
Fort Calhoun Station i
==
Description:==
Reactor Vessel Beltline Plate 4802 2 Heat No.:
A 1768 1 Specification: SA 531 Grade B, Class 1 i
Supplier:
Lukens fabricator:
Combustion Engineering Heat Treatment:
Run No.
Max Temp.
Min. Temp.
Ten:p. Range Time Quench Run ID 1
1600*F 1550* F 50'F 4 hrs WC Austenizing 2
1250'F 1200*F 50* F 4 hrs AC Tempering 3
Il 75' F ll25' F 50*F 40 hrs FC Stress Relief Composition (w/o)*:
C Mn P
S Si Ni Cr Ho Cu Cb Al
.22 1,43
.009
.014
.23
.48 04 50
.10
<.01 03 Fracture Toughness:
l Drop Weight Tu,
-20*F (CEN-189 Appendix. A, Table A6-1, Surveillance Program-Data)
Initial RTmn 18' F (CEN 189, Appendix A, Table-A6-1, Surveillance Program Data)
Initial Upper 121 ft-lb.
(CEN-189, Data) Appendix A, Table A6-1, Surveillance Shelf Energy Pt 190' F Program
- Reference 2 m
W P
^r*-'r
T M
T
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-'f""
y
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y
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Attachment LIC 92 203R Page 6 labl0_1 Plant Name:
Fort Calhoun Station
==
Description:==
Reactor Vessel Beltline Plate D-4802 3 Heat No.:
A 1768 2 Specification: SA-533 Grade B, Class 1 Supplier:
Lukens Fabricator:
Combustion Engineering Heat Treatment:
Run No.
Max Temp.
Hin. Temp.
Temp. Range Time Quench Run ID 1
1600*F 1550*F 50' F 4 hrs WC Austenizing 2
1250*F 1200*F 50' F 4 hrs AC Tem)ering 3
Il75' F ll25' F 50'F 40 hrs FC Stress lelief Composition (w/o)*:
C Mn P
S Si Ni Cr Ho Cu Cb Al
.29 1.50
.009
.012
.24
.51
.05
.53
.11
<.01
.024 Fracture Toughness:
Drop Weight Tu,
'0' F (CEN 189, Appendix A, Table A61)
Initial RT,,
0* F (HTEB Position 5-2, paragraph 1.l(3)b)
Initial Upper Shelf 77.35 ft-lb (HTEB Position 5-2, paragraph 1.2)
Energy
- Reference 2 l
t
t L
Attachment LIC-92 203R Page 7 Table 4 Plant Name:
Fort Calhoun Station
==
Description:==
Reactor Vessel Beltline Plate D 4812-1 Heat No.:
C 3213 2 Specification: SA 533 Grade B, Class 1 i
Supplier:
Lukens fabricator:
Combustion Engineering.
Heat Treatment:
Run No.
Max Temp.
Min. Temp.
Temp. Range Time Quench Run ID 1
1600*F 1550'F 50'F 4 hrs WC Austenizing 2
12S0' F 1200'F 50' F 4 hrs AC Temaering 3
Il75'F ll.95' F 50* F 40 hrs FC Stress 1elief Composition (w/o)*:
C Mn P
S St Ni Cr Ho Cu Cb A1
.22 1.31
.009
.012
.24
.60
.18
.54
.12
<.01
.029 Fracture Toughness:
Drop Weight T,
-30*F (CEN-189, Appendix A, Table A6-1)
Initial RTo, 0*F (MTEB Position 5-2, paragraph 1.l(3)b)
Initial Upper Shelf 86.45 ft-lb (MTEB Position 5-2, paragraph 1.2)
Energy
- Reference 2 J
Attachment LIC-92-203R Page 8 Table 5 Plant Name:
Fort Calhoun Station
==
Description:==
Reactor Vessel Beltline Plate D 4812-2 Heat No.:
C 3143-2 Specification: SA-533 Grade B, Class 1 Supplier:
Lukens Fabricator:
Combustion Engineering Heat Treatment:
Run No.
Max Temp.
Min. Temp.
Temp. Range Time Quench Run ID 1
1600*r 1550*F 50* F 4 hrs
-WC Austenizing 2
1250'F 1200*F 50*F 4 hr.s AC Tem)ering 3
Il75* F ll25' F 50* F 40 hrs FC Stress Relief Composition (w/0)*:
C Mn P
S Si Ni Cr Ho Cu Cb Al
.26 1.33
.010
.013
.26
.56
.06
.52
.10
<.01
.038 Fracture Toughness:
Drop Weight Tu,
-20*F (CEN-189, Appendix A, Table A6-1)
Initial RT 0* F (MTEB Position 5-2, paragraph 1.l(3)h) mn Initial Upper Shelf 87 ft-lb (MTEB Position 5-2, paragraph l.2)
Energy
- Reference.2 1
-J
1 Attachment LIC-92-203R Page 9 Table 6 Plant Name:
Fort Calhoun Station
==
Description:==
Reactor Vessel Beltline Plate D.4812-3 Heat No.:
C 3143-3 Specification: SA 533 Grade B, Class 1 Supplier:
Lukens Fabricator:
Combustion Engineering Heat Treatment:
Run No.
Max Temp. Min. Temp.
Temp. Rango Time Quench Run ID 1
1600* F 1550'F 50* F 4 hrs WC Austenizing
?
1250*F 1200*F 50*F 4 hrs AC Temperin 3
ll75' F ll25' f 50*F 40 hrs FC Stress Relie Composition (w/o)*:
C Mn P
S Si Ni Cr Ho Cu Cb Al l
.22 1.30
.01
.011
.25
.56 06 51
.10
<.01
.027 l
Fracture Toughness:
Drop Weight T,
-30* F (CEN-189, Appendix A, Table A6-1) m Initial RT,
O' F (MTEBPosition5-2, paragraph 1.1(3)b)
Initial Upper Shelf 89.7 ft-lb (MTEB Position 5-2, paragraph 1.2)
Energy i
[
- Reference 2 w
e--
y--
-1c
-+w.-m--
w y
y w
yr 4
ey
.m y
-*g 9
a-
%i-yv w*
Attachment LIC-92 203R Page 10 Table 7 Plant Name:
Fort Calhoun Station
==
Description:==
Surveillance Weld Material Wire Heat No.: 305414 Wold Type:
Submerged Arc Wire Type:
Mil B-4, 3/16" Flux Type:
Linde 1092 Flux Lot: 3951 Fabricator:
Combustion Engineering Heat Treatment Actual:
Run No, Max Temp.
Min Temp.
Temp Range Time Quench Run ID-1 1175'F ll25' F 50'F 40 hrs FC Stress Relief Composition (Filler wire, w/o)*:
C Mn P
S Si
. Ni Cr Mo Cu Cb Al
.14 1.57
.013
.011
.14
.60
.03-
-. 50 -
.35
<.01
.009 Composition (As welded, w/o)*:
C Mn P
S St Ni Cr Ho Cu
.16 1,57
.013
.011
.14
.60
.03
.50
.35 Fracture Toughness:
Drop Weight T,
-50'F (CEN-189, Appendix A, Table A6-1; Surveillance Program Data)
-Initial RT,-
-50'F (CEN 189. Appendix A Table A6-1; Surveillance Program Data)
Initial Upper 97.5 ft-lb LTR-0-MCD-001, Revision 1, " Evaluation of Shelf Energy 3rradiated Capsule W-225", August 1980)
- Reference 2
I 4
Attachment LIC 92-203R Page 11 Iable 8 Plant Name:
Fort Calhoun Station
==
Description:==
Vessel Weld 2 410 (longitudinal)
Wire Heat No.: bl989 Weld Type:
Submerged Arc Weld Wire Type:
Mil B-4 Modified Flux Type:
Linde 124 Flux Lot:
3687 Fabricator:
Combustion Engineering Heat Treatment:
Run No.
Max Temp.
Min. Temp.
Temp. Range Time Quench Run 10 1
Il75'F ll25' F 50' F 40 hrs FC Stress Relief Composition (Filler wire, w/o):
S P.
Ni Cu
.015
.008 NA NA (NA = Not Available)
Composition (As welded, w/o)*:
C Mn P
S Si Ni Cr Mo Cu
.10 1.45
.012
.010
.33
.17
.08
.50
.17 Fracture Toughness:
Initial Rio, Estimated -56*F (CEN 189 Section 6)
Charpy at-10*F Average 61 f t-lb (CEN 189, Appendix A, Table A6 2)
- Reference 3
1 i
1 j
Attachment LIC-92 203R Page 12 t
Table 9 i
Plant Name:
Fort Calhoun Station
==
Description:==
Vessel Wold 3 410 (longitudinal)
Wire Heat No.: 12008, 13253, 27204 Weld Type:
Submerged Arc Weld Wire Type:
Mil B-4 Modified Flux Type:
Lirie 1092 Flux Lot:
3774 Fabricator:
Combustion Engineering e
Heat Treatment:
l Run No.
Max Temp.
Hin. Temp.
Temp. Range Time Quench Run ID 1
ll75'F ll25'F 50'F 40 hrs FC Stress Relief 4
Composition (Filler wire, w/0):
S P
Ni Cu
.015
.008 NA NA (NA - Not Available)
Composition (As wolded, w/o)*:
S P
Ni Cu
.011
.013 1.02
.22 Fracture Toughness:
Initial RT,
Estimated -56*F (CEN-189, Section 6)
Charpy at 10'F Average 57.3 ft-lb (CEN-189, Appendix A. Table A6-2)
References 3 and 6 f
~
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Attachment LIC 92-203R Page 13 Table 10 i
Plant Name:
Fort Calhoun Station
==
Description:==
Vessel Weld 9 410 Wire Heat No.: 20291 Weld Type:
Submerged Arc Weld Wire Type:
Hil B-4 Modified Flux Type:
Linde 1092 i
Flux Lot:
3833 Fabricator:
Combustion Engineering Heat Treatment:
Run No, Max Temp.
Min. Temp.
Temp. Range Time Quenc!:
Run ID 1
Il75'F ll25' F 50'F 40 hrs FC Stress Relief Composition (filler wire, w/o):
S P
Ni Cu
.015
.008 NA NA Composition (As welded, w/o)*:
S P
Ni Cu
.011
.013
.74
.21 i
Fracture Toughness:
Initial RT.
Estimated 56*F (CEN189,Section6)
Charpy at 10'F Average 44.3 ft-lb (CEN 189, Appendix A, Table A6-2)
- Reference 3
J Attachment i
LIC-92-203R Page 14 Ouestion 3.a.
Addressees are requested to provide the following information regarding comitments made to respond to GL 88-11:
1 How the temperature effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 525'F were considered.
In particular licensees are requested to describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and the Charpy upper shelf energy.
OPPD Resnonse In reviewing operating records, OPPD determined that Fort Calhoun was operated below 525'F for ) art of Cycle 2(1975-1976). The amount of time and accumulated fluence during tiis off normal period has been quantified to be approximately 31%
of this fuel cycle at an average inlet temperature of 522.7'F. Based on the good agreement between predicted and measured reference temperatures from surveillance capsules W-225 and W-265 (0.889 ratio of averaged predicted to measured reference temperatures), OPPD has concluded that operation at a cold leg temperature slightly less than 525'F has not resulted in any unaccounted increase in the RT shift or reduction in the Charpy upper shelf energy.
These " lower ndt temperature" irradiation effects have been accounted for within the surveillance program measurements. The surveillance program will continue to account for the effects of the Cycle 2 limited reduced temperature operation during future j
surveillance capsule evaluations.
The extrapolated value for the Charpy upper shelf energy at the end of life was predicted to be 52 ft-lbs based on the surveillance capsule W-265 results. This extrapolated value was based on standard fuel management practices in use at the time of the W-265 capsule removal. The extrapolated value does not include the beneficial effects associated with the implementation of low and extremely low radial leakage fuel management patterns which have been in use since Cycle 8 (1983).
Question 3.b.
Utilities are requested to provide information on how their surveillance results have been used in response to GL 88-11 and how the predicted amount of embrittlement has been considered.
OPPD Response The surveillance results from the W-225 and W-265 ca)sules were used to prepare the GL 88-11 response to the NRC. Regulatory posit on 2.1 of Regulatory Guide 1.99, Revision 2 was not used for the generation of the surveillance data reported in GL 88-11.
The predicted amount of embrittlement, from DDT 4.3 analyses, has been used to provide a target for flux reduction efforts in the design of the low and extreme low radial leakage cores.
l
Attachment LIC-92-203R Page 15 Ouestion 3.c.
If a measured increase in reference temperature exceeds the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or if a measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.I.2 in Regulatory Guide 1.99, Revision 2, the licensee is requested to report the inforination and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16, 1991 and for its current license.
OPPD Respons_ed, The measured reference temperatures derived from analyzing the surveillance capsule materials are comparable to the predicted reference temperatures (Regulatory Guide 1.99, Revision 2 calculations). Below is a comparison of the measured to the predicted references temperatures:
Table 3.c-1 Comparison of Measured versus Predicted Reference Temperatures for Surveillance Capsule Materials (Reference 10)
Sury. Cap.
Material Chemistry ART ART Difference:
ndt ndt No.
Factor, RG Measured Predicted Measured-1.99, rev. 2 'F
'F Predicted W-225 Weld 212 205 180 25 W-265 Weld 212 221 199 22 W-225 Plate long.
65 60 55 5
W-265 Plate long.
65 74 61 12 l
W-265 Plate trans.
65 70 62 8
As indicated in Table Lc-1, the difference between the measured and the predicted values are comparable. All measured versus predicted ART values were ndt within one standard deviation. The standard deviation, 0, for ART is 28'F for welds and 17'F for base metal per RG 1.99, Revision 2.3The differences in the ndt measured versus predicted values as indicated in Table 3.c-1 for both the surveillance weld and plate material are within o the measured versus predicted increases in reference temperature for,.- Thus, weld and plate surveillance materials are comparable.
A comparison of the surveillance material Charpy upper shelf energy measured versus predicted Yalues is provided. The predicted values were calculated using RG 1.99, Revision 2, Figure 2 data.
Table 3.c-2 provides a comparison of the measured versus predicted upper shelf energy values for the surveillance materials.
i Attachment LIC-92-203R Page 16 Table 3.c-2 j
Comparison of Measured versus Predicted Charpy Upper Shelf Energy Decrease for Surveillance Capsule Materials j
(Reference 10)
Sury. Caps.
Katerial Copper Upper Shelf Opper Shelf Difference No.
Content Energy Energy I
4
- s Decrease Decrease Measured %
Predicted %
W-225
' Weld 0.35 37 38 1
]
W-265 Weld 0.35 43 41 2
1 4
W-225 Plate long.
0.10 13 17 4
{
W-265 Plate long.
0.10 23 18 5
i W-265 Plate trans. 0.10 23 18 5
No data from the Fort Calhoun Station surveillance program has exceeded the j
"mean-plus-two" standard deviation bound predicted by RG 1.99, Revision 2.
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