ML20249B991
| ML20249B991 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/06/1998 |
| From: | Devan M, Shannon King FRAMATOME |
| To: | |
| Shared Package | |
| ML18152B752 | List: |
| References | |
| BAW-2324, FTI-77-2324, FTI-77-2324-00, NUDOCS 9806250150 | |
| Download: ML20249B991 (161) | |
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ANALYSIS OF CAPSULF. X VIRGINIA POWER SURRY UNIT NO.1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM 9906250150 V80618 PDR ADOCK 05000290 P
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BAW-2324 April 1998 pn m-n;d4def Analysis; JCapsuWX'
.a Virginia Power lSurryl Unit No 4
-- Reactor Vessel Material Surveillance Program - 'E L I
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F R AM ATO M E TECHNOLOGIES
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i BAW-2324 April 1998 5
l Analysis of Capsule X l
Virginia Power Surry Unit No.1 l
- Reactor Vessel Material Surveillance Program -
l by M.J.DeVan S. Q. King i
i FTl Document No. 77-2324-00 (Section 0 for document signatures.)
l Prepared for Virginia Power Prepared by Framatome Technologies, Inc.
f 3315 Old Forest Road
- P. O. Box 10935 Lynchburg, Virginia 24506-0935 Nweam
L Y
f Acknowledgment This acknowledges the efforts of Kevin Hour of the McDermott Technology, Inc.
Lynchburg Research Center. His expertise in specimen testing contributed greatly.to the success of this project.
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l-Executive Summary
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This report describes the results of the examination of the fourth capsule (Capsule X) of the Virginia Power Surry Unit No.1 reactor vessel surveillance program. The capsule l
was removed and examined at the end of the fourteenth cycle. The capsule received 2
an average fast fluence of 1.599 x 10 n/cm (E > 1.0 MeV). The objective of the program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materials by the testing and evaluation of tension and Charpy impact specimens. The program was designed in accordance with the requirements of ASTM Specification E 185-73.
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I Table of Contents 1
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1.
I n t rod u ctio n..................................................................................................... 1 2.
B a ck g ro u n d...................................................................................................... 2-1 3.
S urveilla nce Prog ra m Description................................................................. 3-1 4.
Tests of U nirrad iated M ate rial........................................................................ 4-1 l
f 5.
P ost-l rra d lation Te sti ng................................................................................... 5-1 5.1.
Visual Examination and inventory.................................................... 5-1 5.2.
The rma l M o nit o rs.......................................................................... 5-1 5.3.
Chemical Composition Check Analysis............................................... 5-1 i
5.4.
Te n s io n Te st R e s u lts....................................................................... 5-1 i
5.5.
Charpy V-Notch Impaet Results........................................................ 5-2 1
4 5.6.
Wedge Opening Loading Specimens................................................ 5-3 l
3 6.
N e utro n Fl u e n ce...........................................................................................
6.1.
Objective s a nd Proced u re................................................................ 6-1 I
6.2.
Results.............................................................................................6-2 f
7.
Discu ssion of Capsule Res ults...................................................................... 7-1 1
7.1.
Copper and Nickel Chemical Composition Data................................. 7-1 7.2.
Unirradiated Material Property Data................................................. 7-1 7.3.
I rra d iated Prope rty D ata........................................................................ 7-2
- 7. 3.1 Te n sile P rope rtie s................................................................... 7-2 7.3.2 Impact Prope rties.................................................................... 7-2 8.
S u m m a ry of Res u lt s....................................................................................... 8-1 9.
C e rtifi ca t i o n...................................................................................................
10.
R e fe re n ce s................................................................................................. 1 0
.f
- F R AM ATO M E IV
Table of Contents (Cont'd)
APPENDICES A.
Instrumented Charpy V-Notch Specimen Test Results Load Time Traces.... A-1 B.
U nirra diated Te n sile D ata.....................................................
....... B-1 C.
Unirradiated Charpy V-Notch impact Data........................................... C-1 D.
Surry Unit No.1 Charpy V-Notch Impact Surveillance Data Using Hyperbolic Tangent Curve-Fitting Method......................................... D-1 E.
Fluence Analysis Methodology...................................................... E-1 F.
Capsule Dosimetry Measurements......................................................F-1 s
List of Tables Table I
3-1.
Test Specimens Contained in Surry Unit No.1 Capsule X......
... 3-3
{
3-2.
Chemical Composition of Surry Unit No.1 Capsule X Surveillance Material... 3-4 3-3.
Heat Treatment of Surry unit No.1 Capsule X Surveillance Materials....
....3-5 5-1.
Chemical Analysis Results of Selected Base Metal and Weld Metal I rradiated C harpy Specimens.......................................................... 5-4 5-2.
Tensile Properties of Surry Unit No.1 Capsule X Reactor Vessel 2
Surveillance Materials, Irradiated to 1.599 x 10" n/cm (E > 1.0 MeV)..
..... 5-5 5-3.
Charpy V-Notch Properties of Surry Unit No.1 Capsule X Base Metal 2
Plate, Heat No. C4415-1, Irradiated to 1.599 x 10" n/cm (E > 1.0 MeV).
..5-6 5-4.
Charpy V-Notch Properties of Surry Unit No.1 Capsule X Base Metal I
Plate, Heat-Affected-Zone, Irradiated to 1.599 x 10" n/cm (E > 1.0 MeV).... 5-7 2
5-5.
Charpy V-Notch Properties of Surry Unit No.1 Capsule X Weld Metal SA-1526 (299L44 / 8596), Irradiated to 1.599 x 10" n/cm (E > 1.0 MeV)...
.5-8 2
5-6.
Charpy V-Notch Properties of Surry Unit No.1 Capsule X Correlation 2
Monitor Plate Material (HSST Plate 02), Irradiated to 1.599 x 10" n/cm
( E > 1. 0 M eV)............................................................................
5-7.
Instrumented Charpy Properties of Surry Unit No.1 Capsule X Base 2
Metal Plate, Heat No. C4415-1, Irradiated to 1.599 x 10" n/cm (E>1.0MeV).......................................................................
.. 5-10 IWM v
List of Tables (Cont'd)
Table 5-8.
Instrumented Charpy Properties of Surry Unit No.1 Capsule X Base 2
Metal Plate, Heat-Affected-Zone, Irradiated to 1.599 x 10 n/cm
)
(E>1.0MeV)......................................................................
.. 5 11 5-9.
Instrumented Charpy Properties of Surry Unit No.1 Capsule X Weld 2
)
Metal SA-1526 (299L44 / 8596), Irradiated to 1.599 x 10 n/cm (E > 1.0 MeV)..................
.. 5-12 5-10. Instrumented Charpy Properties of Surry Unit No.1 Capsule X Correlation 2
l Monitor Plate Material, (HSST Plate 02), Irradiated to 1.599 x 10 n/cm (E > 1.0 MeV)....
.. 5-13 6-1.
C/M Ratios for Surry Unit No.1 Capsule X..................
.6-3
(
6-2.
Fast Neutron Flux (E > 1 MeV)............
...... 6-4 6-3.
Incremental Fast Neutron Fluence (E > 1 MeV)......................
..6-5 6-4.
Cumulative Fluences Surry Unit No.1 Analysis............................
.6-6 6-5.
Locations of Pea ks........................................................ 6-7 7-1.
Copper and Nickel Chemical Composition Data for Surry Unit No.1 Reactor Vessel Surveillance Weld Metal SA-1526...........,..
.7-4 l
7-2.
Comparison of Surry Unit No.1 Capsule X Tension Test Results........
.. 7-5 7-3.
Summary of Surry Unit No.1 Reactor Vessel Surveillance Capsules
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Tensile Test Results.
.7-6 i
7-4.
Observed Vs. Predicted 30 ft-lb Transition Temperature Changes for 2
Surry Unit No.1 Capsule X Surveillance Materials - 1.599 x 10 n/cm..
.7-7 7-5 Observed Vs. Predicted Upper-Shelf Energy Decreases for 2
i Surry Unit No.1 Capsule X Surveillance Materials.1.599 x 10 n/cm
.7-8 7-6.
Summary of Surry unit No.1 Reactor Vessel Surveillance Capsules Charpy impact Test Results.................
.. 7-9 B-1.
Unirradiated tensile Properties for the Surry Unit No.1 Reactor Vessel Surveillance Materials......................
..... B-2
(
C-1.
Unirradiated Charpy V-Notch Properties for the Surry Unit No.1 Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT).
..C-2 C-2.
Unirradiated Charpy V-Notch Properties for the Surry Unit No.1 Base Metal Plate Heat-Affected-Zone Material.........................
............C-3 C-3.
Unirradiated Charpy V-Notch Properties for the Surry Unit No.1 i
j Weld Metal SA-1526 (299L44 / 8596)..............
.. C-4 l
C-4.
Unirradiated Charpy V-Notch Properties for the Surry Unit No.1 Correlation Monitor Plate (HSST Plate 02), Longitudinal Orientation (LT)..
.C-5 D-1.
Surry Unit No.1 Capsule T Surveillance Charpy V-Notch Data, Base Metal Plate Heat No. C4415-1, l. longitudinal Orientation (LT)............
.... D-2 D-2.
Surry Unit No.1 Capsule T Surveillance Charpy V-Notch Data, Base f
Metal Plate Heat-Affected-Zone Material.
..D-2 1
D-3.
Surry Unit No.1 Capsule T Surveillance Charpy V-Notch Data, Weld l
I
.D-3 Metal Sa-1526 (299L44 / 8596)......
I f--
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I List of Tables (Cont'd)
Table D-4.
Surry Unit No.1 Capsule T Surveillance Charpy V-Notch Data, Correlation Monitor Plate (HSST Plate 02), Longitudinal Orientation (LT)..................... D-3 I
D-5.
Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data, Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT)........................ D-8 D-6.
Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data, Base Metal Plate Heat-Affected-Zone Material.................................................. D-8 D-7.
Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data, Weld Metal Sa-1526 (299L44 / 8596).....................................
.............. D-9 D-8.
Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data, Correlation Monitor Plate (HSST Plate 02), Longitudinal Orientation (LT)......
.... D-9
]
E-1.
Bias Removal Function (E > 1 MeV).......
.............. E-7 F-1.
O ua niifyin g G a mm a Ra ys.........................................................
......F-5 F-2.
Isotopic Fractions and Weight Fractions of Target Nuclides.
..................F-6 F-3.
Specific Activities for Surry Unit No.1 Capsule X Dosimetry......................F-7 l
F-4.
Copper Dosimetry Measurements from Capsule X Surry Unit No.1.....
...F-8 F-5.
Iron Dosimetry Measurements from Capsule X Surry Unit No.1..
............F-9 F-6.
Cobalt / Aluminum Dosimetry Measurements from Capsule X i
S u rry U n it N o. 1..........................................................................
..F-10 F-7.
Shielded / Cobalt / Aluminum Dosimetry Measurements from Capsule X Surry Unit No.1
......................................F-11 F-8.
Uranium-238 Dosimetry Measurements from Capsule X Surry Unit No.1......F-12 F-9.
Neptunium-237 Dosimetry Measurements from Capsule X Surry Unit No.1..F-13 List of Fiaures Fiaure 3-1.
Reactor Vessel Cross Section Showing Location of RVSP Capsules in Surry Unit No.1 Reactor Vessel................................................... 3-6 3-2.
Type ll Surveillance Capsule Assembly Showing Location of Specimens a n d M o n it o rs....................................................................
.3-7 5-1.
Photograph of Thermal Monitor Melt wires as Removed from Surry Unit No.1 Surveillance Capsule X......................................... 5-14 5-2.
Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. C4415-1, Longitudinal Orientation, Specimen No. V13, Tested at 70 F.............. 5-15 5-3.
Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. C4415-1.
Longitudinal Orientation, Specimen No. V14, Tested at 550 F.............. 5-15 vii
<W N
List of Fiaures (Cont'd)
Fiaure 5-4.
Tension Test Stress-Strain Curve for Weld Metal (299L44 / 8596),
Specimen No. W5, Tested at 70*F........................................
......... 5-16 j
5-5.
Tension Test Stress-Strain Curve for Weld Metal (299L44 / 8596),
Specimen No. W6, Tested at 550*F........................................... 5-16 5-6.
Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal Plate Heat No. C4415-1....................5-17 5-7.
Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal SA-1526 (299L44 / 8596).................... 5-18 5-8.
Charpy impact Data for Irradiated Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT).............................................................. 5-19 5 9.
Charpy Impact Data for Irradiated Base Metal Plate Heat-Affected-Zone Material..................................................................................5-20 5-10. Charpy impact Data for Irradiated Weld Metal SA-1526 (299L44 / 8596)... 5-21 5-11. Charpy impact Data for Irradiated Correlation Monitor Plate Material (HSST Plate 02), Longitudinal Orientation (LT)...............
. 5-22 5-12. Photographs of Charpy impact Specimens Fracture Surfaces, Base Metal Plate Heat No. C4415-1..................................
. 5-23 5-13. Photographs of Charpy impact Specimens Fracture Surfaces, Base Metal Plate Heat-Affected-Zone Material....................
.. 5-24 5-14. Photographs of Charpy impact Specimens Fracture Surfaces, Weld Metal SA-1526 (299L44 / 8596).......................
. 5-25 5-15. Photographs of Charpy impact Specimens Fracture Surfaces, 1
l Correlation Monitor Plate Material (HSST Plate 02)..
.. 5-26 6-1.
Longitudinal Weld Flux Profiles at Three Azimuthal Locations......
.6-8 6-2.
3-D Surface Plot at Vessel inside Surface............
......... 6-9 A-1.
Load-Time Trace for Charpy V-Notch Impact Specimen V63..............
.... A-2 A-2.
Load-Time Trace for Charpy V-Notch Impact Specimen V60.
....... A-2 s
A-3.
Load-Time Trace for Charpy V-Notch Impact Specimen V57.....
...... A-3 A-4.
Load-Time Trace for Charpy V-Notch Impact Specimen V59........
.... A-3 A-5.
Load-Time Trace for Charpy V-Notch Impact Specimen V61................... A-4 A-6.
Load-Time Trace for Charpy V-Notch Impact Specimen V64.
... A-4 A-7.
Load-Time Trace for Charpy V-Notch impact Specimen V58.....
.... A-5 A-8.
Load-Time Trace for Charpy V-Notch impact Specimen V62........
.A-5 A-9.
Load-Time Trace for Charpy V-Notch Impact Specimen H17................ A-6 A-10. Load-Time Trace for Charpy V-Notch Impact Specimen H23.
.A-6 A-11. Load-Time Trace for Charpy V-Notch Impact Specimen H21.....
... A-7 l
A-12. Load-Time Trace for Charpy V-Notch Impact Specimen H20......
.... A-7 A-13. Load-Time Trace for Charpy V-Notch impact Specimen H24................A-8 j
A-14. Load-Time Trace for Charpy V-Notch Impact Specimen H19.........
..A-8 A-15. Load-Time Trace for Charpy V-Notch impact Specimen H18..
.A-9 UNM I
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List of Fiaures (Cont'd)
Fiaure A-16. Load-Time Trace for Charpy V-Notch Impact Specimen H22..................... A-9 A-17. Load-Time Trace for Charpy V-Notch Impact Specimen W23................... A-10 l
A-18. Load-Time Trace for Charpy V-Notch Impact Specimen W18............
. A-10 A-19. Load-Time Trace for Charpy V-Notch Impact Specimen W19................. A-11
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A-20. Load-Time Trace for Charpy V-Notch Impact Specimen W24............
.. A-11 l
A-21. Load-Time Trace for Charpy V-Notch Impact Specimen W22............... A-12 A-22. Load-Time Trace for Charpy V-Notch impact Specimen W20................ A-12 A-23. Load-Time Trace for Charpy V-Notch Impact Specimen W21......
.... A-13 A-24. Load-Time Trace for Charpy V-Notch Impact Specimen W17................ A-1 3 A-25. Load-Time Trace for Charpy V-Notch Impact Specimen R50....
... A-14 l
A-26. Load-Time Trace for Charpy V-Notch Impact Specimen R56.....
..... A-14 A-27. Load-Time Trace for Charpy V-Notch Impact Specimen R54...
.... A-15 A-28. Load-Time Trace for Charpy V-Notch Impact Specimen R52.....
. A-15 A-29. Load-Time Trace for Charpy V-Notch Impact Specimen R51.....
.. A-16 A-30. Load-Time Trace for Charpy V-Notch Impact Specimen R53..
.. A-16 A-31. Load-Time Trace for Charpy V-Notch Impact Specimen R55........... A-17 A-32. Load-Time Trace for Charpy V-Notch Impact Specimen R49................ A-17 C-1.
Unirradiated Charpy impact Data for Surry Unit No.1 Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT) (Hyperbolic Tangent Curve-Fitting Method)...........
..C-6 C-2.
Unirradiated Charpy impact Data for Surry Unit No.1 Base Metal Heat-Affected-Zone Material (Hyperbolic Tangent Curve-Fitting Method)..... C-7 C-3.
Unirradiated Charpy impact Data for Surry Unit No.1 Weld Metal SA-1526 (299L44 / 8596)(Hyperbolic Tangent Curve-Fitting Method).........C-8 I
C-4.
Unirradiated Charpy impact Data for Surry Unit No.1 Correlation Monitor Plate (HSST Plate 02), Longitudinal Orientation (LT) (Hyperbolic Tangent Curve-Fitting Method).......................
.C-9 D-1.
Surry Unit No.1 Capsule T Surveillance Charpy impact Data Base Metal 3
Plate Heat No. C4415-1, Longitudinal Orientation (LT) (Hyperbolic Tangent Curve-Fitting Method)........................................................ D-4 D-2.
Surry Unit No.1 Capsule T Surveillance Charpy impact Data Base Metal Heat-Affected-Zone Material (Hyperbolic Tangent Curve-Fitting Method)...... D-5 D-3.
Surry Unit No.1 Capsule T Surveillance Charpy Impact Data Weld Metal SA-1526 (299L44 / 8596) (Hyperbolic Tangent Curve-Fitting Method)........... D-6 D-4.
Surry Unit No.1 Capsule T Surveillance Charpy impact Data Correlation Monitor Plate (HSST Plate 02), Longitudinal Orientation (LT) (Hyperbolic Tangent Curve-Fitting Method)...
........ D-7 D-5.
Surry Unit No.1 Capsule V Surveillance Charpy impact Data Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT) (Hyperbolic Tangent Curve-Fitting Method)............................................... D-1 0 ix DNM
f List of Fiaures (Cont'd)
Fiaure D-6.
Surry Unit No.1 Capsule V Surveillance Charpy impact Data Base Metal l
Heat-Affected-Zone Material (Hyperbe!!c Tangent Curve-Fitting Method)...... D-11 D-7.
Surry Unit No.1 Capsule V Surveillance Charpy impact Data Weld Metal SA-1526 (299L44 / 8596) (Hyperbolic Tangent Curve-Fitting Method)........... D-12
).
D-8.
Surry Unit No.1 Capsule V Surveillance Charpy impact Data Correlation Monitor Plate (HSST Plate 02), Longitudinal Orientation (LT) (Hyperbolic i
Ta ng ent C u rve-Fitting Method)................................................................... D-13 l
E-1.
Fluence Analysis Methodology............................................................... E-8 l
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- 1. Introduction This report presents the examination results of the fourth reactor vessel surveillance I
capsule (Capsule X) removed from Virginia Power's Surry Unit No.1 reactor vessel. The capsule was removed and the contents evaluated after being irradiated in the Surry Unit No.1 reactor as part of the plant's reactor vessel surveillance program (RVSP) as documented in WCAP-7723.0) This report describes the testing and the post-irradiation l
data obtained from the Surry Unit No.1 Capsule X after receiving an average fluence of 1.599 x 10" n/cm (E > 1.0 MeV). The data are compared to previous Surry Unit No.1 2
RVSP results from Capsule T,m Capsule W*' (only dosimetry was evaluated), and g
Capsule V.")
The objective of the program is to monitor the effects of neutron irradiation on the mechanical properties of reactor vessel materials under actual plant operating conditions.
l The Surry Unit No.1 RVSP was designed and furnished by Westinghouse Electric Corporatiori and was based on American Society for Testing and Materials (ASTM)
Standard E 185-73.5) 8 l
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2. Background
The ability of the reactor vessel to resist fracture is a primary factor in ensuring the safety I
of the primary system in light water-cooled reactors. The reactor vessel beltline region is the most critical region of the vessel because it is exposed to the highest level of neutron I
irradiation. The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels used in the fabrication of reactor vessels are well characterized and documented. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. The most significant mechanical property change in reactor vessel steels is the increase in the ductile-to-brittle transition temperature accompanied by a reduction in the Charpy upper-shelf energy (C,USE) value.
.I Code of Federal Regulations, Title 10, Part 50, (10 CFR 50) Appendix G, " Fracture l
Toughness Requirements,"* specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of light water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitatioris for operation of the RCPB. The fracture toughness and operational requirements are specified to provide adequate safety
- I~
margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be I
subjected over its service lifetime. Although the requirements of 10 CFR 50, Appendix G, became effective on August 16,1973, the requirements are applicable to all boiling and
.h pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effectivo date.
10 CFR 50, Appendix H," Reactor Vessel Materials Surveillance Program Requirements," M defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens contained in capsules that are periodically withdrawn from the reactor vessel. These data I
f-,.m 2.,
permit determination of the conditions under which the vessel can be ope v.ith adequate safety margins against non-ductile fracture throughout its service i.
A method for guarding against non-ductile fracture in reactor vessels is described in Appendix G to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Ill, " Nuclear Power Plant Components"* and Section XI, " Rules for Inservice inspection."* This method uses fracture mechanics concepts and the reference nil-ductility temperature, RTuo7, which is defined as the greater of the drop weight nil-ductility transition temperature (in accordance with ASTM E 208-81"*) or the temperature that is 60 F below that at which the material exhibits 50 ft-Ibs and 35 mits lateral expansion. The RTuo7 of a given materialis used to index that material to a reference stress intensity factor curve (K ni curve), which appears in Appendix G of ASME B&PV Code Section ill and Section XI. The K ni curve is a lower bound of dynamic and crack arrest fracture toughness data obtained from several heats of pressure vessel steel. When a given material is indexed to the K ai curve, allowable stress intensity factors can be obtained for the material as a function of temperature. The operating limits can then be determined using these allowable stress intensity factors.
The RTuo1 and, in turn, the operating limits of a nuclear power plant, are adjusted to account for the effects of irradiation on the fracture toughness of the reactor vessel materials. The irradiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which surveillance capsules containing prepared specimens of the reactor vessel materials are periodically removed from the operating nuclear reactor and the specimens are tested. The increase in the Charpy V-notch 30 ft-lb temperature is added to the l
original RTuo1 to adjust it for irradiation embrittlement. The adjusted RTuo1 s used to i
l index the material to the K,n curve which, in turn, is used to set operating limits for the l
nuclear power plant. These new limits take into account the effects of irradiation on the i
reactor vessel materials.
10 CFR 50, Appendix G, also requires a minimum initial Charpy V-notch upper-shelf energy (C,USE) of 75 ft-lbs for all beltline region materials unless it is demonstrated that lower values of upper-shelf fracture energy will provide an adequate margin of safety against fracture equivalent to those required by ASME Section XI, Appendix G. No action is required for a material that does not meet the initial 75 ft-lbs requirement provided that the irradiation embrittlement does not cause the C,USE to drop below 50 ft-Ibs. The 2-2 IINM
regulations specify that if the C,USE drops below 50 ft-Ibs, it must be demonstrated, in a manner approved by the Office of Nuclear Reactor Regulation, that the lower values will
[
provide adequate margins of safety.
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- 3. Surveillance Program Description The reactor vessel surveillance program for Surry Unit No.1 includes eight capsules designed to monitor the effects of neutron and thermal environments on the materials of the reactor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, are positioned near the inside wall of the reactor vessel at the locations shown in Figure 3-1.
Capsule X was removed during the fourteenth refueling outage of Surry Unit No.1. The capsule contained Charpy V-notch (CVN) impact test specimens fabricated from one base metal plate (SA-533, Grade B, Class 1), heat-affected-zone (HAZ) material, a submerged-arc weld metal fabricated using Linde 80 flux, and a correlation monitor plate material. The tensile test specimens and wedge opening loading (WOL) specimens were fabricated from the base metal plate and submerged-arc weld metal. The number of specimens of each material contained in Capsule X is described in Table 3-1, and the location of the individual specimens within the capsule is shown in Figure 3-2. The i
chemical composition and heat treatment of the surveillance materials in Capsule X are described in Tables 3-2 and 3-3 respectively.
All base metal specimens were machined from the %-thickness (%T) location of the plate material after stress relieving. The test specimens represent material taken at least one l
plate thickness (9 inches) from the quenched ends of the original plate thickness. The base metal tensile and CVN specimens were oriented such that the longitudinal axis of the specimen was parallel to the major working direction of the plate (i.e., longitudinal orientation, LT). The weld metal CVN and tensile specimens were oriented with the longitudinal axis of the specimen transverse to the welding direction. The WOL specimens were machined such that the simulated crack in the specimen would l
propagate parallel to the weld direction.
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I Capsule X contained dosimeter wires of pure copper, nickel, aluminum-cobalt (cadmium-shielded and unshielded) and cadmium-shielded neptunium-237 (227Np) and uranium-238 (2"U). The location of these dosimeters within Capsule X is shown in Figure 3-2.
Thermal monitors fabricated from two low-melting alloys were included in the capsule.
The thermal monitors were sealed in Pyrex tubes and inserted in spacers located in Figure 3-2. The eutectic alloys and their melting points are listed below:
2.5% Ag,97.5% Pb Melting Point 579 F 1.75% Ag,0.75% Sn,97.5% Pb Melting Point 590 F l
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Table 3-1. Test Specimens Contained in Surry Unit No.1 Capsule X
' Number of Test Specimens.-
MaterialDescription ;
. Tension.
CVN 1mpact :
WOL Base MetalPlate (Heat No.C44151) l Longitudinal 2
8 2
Heat-Affected-Zone 8
Weld Metal, SA-1526 2
8 2
l (299L44 /8596)*
Correlation Monitor Material, 8
l HSST Plate 02 (Heat No. A1195-1) l Total 4
32 4
l
- Weld wire heat number and flux lot identifier.
I f
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Table 3-2. Chemical Composition of Surry Unit No.1 Capsule X Surveillance MaterialsW ChemicalComposition,wt%
Correlation Plate Weld Metal SA-1526:
. Monitor Plate Eleme' t Heat No. C4415-ic
- (299L44 /8596)?)
Heat No. A1195-1 n
C 0.22 0.245(4 0.10 0.185(d) 0.22 Mn 1.33 1.46k) 1.49 1.47(d) 1.48 P
0.014 0.012k) 0.011 0.011(d) 0.012 S
0.014 0.017(4 0.010 0.017(d) 0.018 Si 0.23 0.42(4 0.37 0.43(d) 0.25 Ni 0.50 0.569(4 0.68 0.643(d) 0.68 Cr 0.078 0.105k) 0.076 0.074(d)
Mo 0.55 0.618(4 0.46 0.405(d) 0.52 Cu 0.11 0.115k) 0.25 0.243(d) 0.14 V
ND(b) 0.004k) 0.001
<0.002(d)
Co 0.015 0.006k) 0.001 0.011(d)
Sn 0.008 Zn ND(b)
At 0.036 0.013 N2 0.007 0.008 Ti NDS)
Zr 0.002 As 0.007 l
B NDS)
NOTES:
(a) Weld wire heat rr aoer and flux lot identifier.
l (b) Not detected.
(c) Analysis performed on irradiated Charpy plate specimen V 25.(d)
(d) Analysis performed on irradiated Charpy weld specimen W-10.(d)
)
3-4 f"<MNRM
Table 3-3 Heat Treatment of Surry Unit No.1 Capsule X Surveillance Materials Material HeatTreatment Base Metal Plt.te 1650-1700 F for 9 hrs., water quenched Heat No. C4415-1 1200 F for 9 hrs., air cooled 1125 F for 15% hrs., fumace cooled to 600 F Weld Metal SA-1526 1125 F for 15% hrs., fumace cooled to 600 F (299L44 / 8596)*
Correlation Monitor Plate, 1675i25 F for 4 hrs., air cooled HSST Plate 02 1600i25 Ffor4 hrs.,waterquenched Heat No. A1195-1 1225125 F for 4 hrs., fumace cooled 1150 i 25 F for 40 hrs., fumace cooled to 600 F
- Weld wire heat number and flux lot identifier.
l l
l I'"D*D'S 3-5
Figure 3-1. Reactor Vessel Cross Section Showing Location of RVSP Capsules in Surry Unit No.1 Reactor Vessel I
i T (Withdrawn)
Z (Withdrawn EOC-12 270
& Relocated)
Z (Inserted EOC-12)
S CAPSULE (TYP) y V (Withdrawn EOC-14 REACTOR VESSEL
& Relocated)
THERMAL SHIELD 25 15 s
10*
l I_1
~
10-l l
t 0*
180' l/
'/
15*
10' 10, V (Withdrawn)
X (Inserted EOC.12 &
25' Withdraw EOC-14) g y (Inserted EOC-14)
U (Withdrawn EOC-12
& Relocated)
[/
W (Withdrawn)
X (Withdrawn EOC-12
& Relocated) 90 U (inserted EOC-12) 3-6
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- 4. Tests of Unirradiated Material Unirradiated material was evaluated for two purposes: (1) to establish baseline data to which irradiated properties data could be compared; and (2) to determine those material properties as required for compliance with 10 CFR 50, Appendices G and H.
The unirradiated specimens were tested by Westinghouse Electric Corporation as part of the development of the Surry Unit No.1 surveillance program. The details of the testing procedures are described in Westinghouse Electric Corporation Report WCAP-7723.m The original unirradiated CVN data are based on hand-fit Charpy curves using engineering judgment; these data were re-evaluated herein using a hyperbolic tangent l
curve fitting program for consistency with the irradiated data. The unirradiated mechanical properties for the Surry Unit No.1 RVSP materials are summarized in Appendices B and C of this report.
i
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I L
L l
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l l
S. Post-Irradiation Testing 5.1. Visual Examination and Inventory After disassembly, the contents of Capsule X were inventoried and found to be consistent with the surveillance program report inventory (WCAP-7732N).
5.2. Thermal Monitors The two low-melting point (579 F and 590 F) eutectic alloys contained in Capsule X were examined, and no indication of melting was observed (see Figure 5-1). Therefore, based on this examination, the maximum temperature the capsule test specimens were exposed to was less than 579'F.
5.3. Chemical Composition Check Analysis One tested irradiated base metal Charpy specimen and one tested irradiated weld metal Charpy specimen were analyzed to determine their chemical compositions. A small sample was removed from the fracture surface of Specimen V58 (base metal) and from the fracture surface of Specimen W22 (weld metal). Each sample was analyzed using the inductively coupled plasma (ICP) method to determine the following chemical constituents: manganese (Mn), phosphorous (P), sulfur (S), silicon (Si), nickel, (Ni),
chromium (Cr), molybdenum (Mo), copper (Cu), cobalt (Co), vanadium (V),. The results of the analyses are presented in Table 5-1.
5.4. Tension Test Results The results of the post-irradiation tencion tests are presented in Table 5-2, and the stress-strain curves are presented in Figures 5-2 through 5-5. Tests were performed on
)
' specimens at room temperature (70 F) and 550 F. The tests were performed using a MTS servohydraulic test machine. All tension tests were run using stroke control with an 5-1 NMD
)
1
initial actuator travel rate of 0.0075 inch per minute through the yield point. Following specimen yielding, an actuator speed of 0.03 inch per minute was used. The test conditions were in accordance with the applicable requirements of ASTM E 8-96a"" and ASTM E 21-92."') Photographs of the tension test specimen fractured surfaces are shown in Figures 5-6 and 5-7.
5.5. Charpy V-Notch impact Results The CVN impact testing was performed in accordance with the applicable requirements of ASTM E 23-960 ) on a Satec S1-1K Impact tester certified to meet NIST ('Watertown")
standards. Impact energy, lateral expansion, and percent shear fracture were measured at numerous test temperatures and recorded for each specimen. In addition, all CVN impact testing was performed using instrumentation to record a load-versus-time trace and energy-versus-time trace for each impact event. The load-versus-time traces were analyzed to determine time, load, and impact energy for general yielding, maximum load, fast fracture, and crack arrest properties during the test. The dynamic yield stress is l
calculated from the three point bend formula:
a, = 3333 * (generalyielding load)
The dynamic flow stress is calculated from the average of the yield and maximum loads, also using the three point bend formula:
l p = 33.33 *
(general yielding load + maximum load)*
o 2
s i
The results of the CVN impact testing are shown in Tables 5-3 through 5-10 and Figures 5-8 through 5-11. The curves were generated using a hyperbolic tangent curve-fitting program to produce the best-fit curve through the data. The individual load-versus-time traces for the instrumented CVN impact tests are presented in Appendix A.
I Photographs of the CVN specimen fracture surfaces are presented in Figures 5-12 through 5-15.
f I
5-2 IfM M M l
5.6. Wedge Opening Loading Specimens The wedge opening loading (WOL) specimens were not tested at the request of Virginia Power. The specimens are to be stored at the McDermott Technology, Inc. (MTI)
Lynchburg Research Center (LRC) for testing at a future data.
1 l
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5-3
'f M %^7R M -
1
f Table 5-1. Chemical Analysis Results of Selected Base Metal and Weld Metal Irradiated Charpy Specimens ChemicalCompos' tion,wt%
. Base Metal....
jWeld Metal l
Element -
Specimen V58 Specimen W22 Mn 1.05 1.14 P
<0.016*
<0.047*
S
<0.09*
0.042 Si 0.21 0.34 Ni 0.41 0.49 Cr 0.07 0.05 j
Mo 0.45 0.30 Cu 0.08 0.18 Co 0.01 0.015 V
<0.0012*
<0.003*
- below minimum detection limit.
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Table 5-3. Charpy V-Notch Properties of Surry Unit No.1 Capsule X Base Metal Plate, Heat No. C4415-1, 2
Irradiated to 1.599 x 10" n/cm (E > 1.0 MeV)
Test
' Impact lateral Shear Specimen Temp.,
- Energy, Expansion,_
- Fracture, No.
F ft-Ib ' '
inch V63 0
11 0.007 0
V60 68 32.5 0.023 20 V57 125 43.5 0.033 50 V59 175 67 0.054 70 V61 200 77.5 0.060 85 V64 250 97.5*
0.080 100 V58 300 95*
0.078 100 V62 400 92.5*
0.075 100 Values used to determine upper-shelf energy value in accordance with ASTM E 185-94.M
.k l
5-6 I'"'W D
j Table 5-4. Charpy V-Notch Properties of Surry Unit No.1 Capsule X Base Metal Plate, Heat-Affected-Zone, 2
Irradiated to 1.599 x 10" n/cm (E > 1.0 MeV)
I
. Test.
_ Impact :
Lateral Shear Specime'n :.
Temp ~,
. Energy,i Expansion,.
- Fracture,
' Noc:
- F ~ '
ft4tr inch l
H17 0
26 0.018 30 H23 68 52 0.037 65 H21 125 59.5 0.034 70 H20 175 53.5 0.040 85 H24 200 80*
0.059 100 H19 250 66*
0.049 100 H18 300 42.5*
0.036 100 H22 400 43.5 0.045 100 l
l Values used to determine upper-shelf energy value in accordance with ASTM E 185-94J")
I 1
5-7 IMTNRM
Table 5-5. Charpy V-Notch Properties of Surry Unit No.1 Capsule X Weld Metal SA-1526 (299L44 / 8596),
2 Irradiated to 1.599 x 10" nIcm (E > 1.0 MeV)
Test impact Lateral Shear Specimen Temp.,
. Energy, Expansion,
- Fracture, No.
- F ft-lb inch W23 0
4.5 0.001 0
W18 68 13.5 0.009 0
W19 125 17.5 0.013 40 W24 200 28 0.024 60 W22 250 36 0.032 90 W20 300 36.5*
0.037 100 W21 350 40*
0.040 100 W17 400 41.5*
0.041 100 Values used to determine upper-shelf energy value in accordance with ASTM E 185-94.M l
1 5-8 I'"'MM D'5
l I
f Table 5-6. Charpy V-Notch Properties of Surry Unit No.1 Capsule X Correlation Monitor Plate Material (HSST Plate 02),
irradiated to 1.599 x 10" nlcm' (E > 1.0 MeV) i
. Testi Ilrhpact; 1.ateral
. Shear l
Specimen : ~
. Temp.,
Energy,L Expansion,.-
- Fracture, I
No PF1 ft-lb ~
' ~ inch l
R50 0
3 0.002 0
RS6 68 5.5 0.004 0
R54 125 16 0.013 35 RS2 200 35 0.028 55 R51 250 49.5 0.040 60 R53 300 76*
0.068 100 l
R55 350 83*
0.071 100 l
R49 400 87.5*
0.076 100 l
)
Values used to determine upper-shelf energy value in accordance with ASTM E 185-94,M 1
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Figure 5-2. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. C4415-1, Longitudinal Orientation, Specimen No. V13, Tested at 70 F P2 Oct.,
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ce.im.. i-o st-.i-Figure 5-3. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. C4415-1, Longitudinal Orientation, Specimen No. V14, Tested at 550 F 22 Oct..
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Figure 5-4. Tension Test Stress-Strain Curve for Weld Metal (299L44 / 8596),
Specimen No. W5 Tested at 70 F P2 Oct.,
1997 818es WS WS feet temp.e 70 Fi 21 C) g
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i-o s,-.i-Figure 5-5. Tension Test Stress-Strain Curve for Weld Metal (299L44 / 8596),
Specimen No. W6 Tested at 550oF F1f.Oct..
22 1997 i
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5-16
Figure 5-6. Pbtographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal Plate Heat No. C4415-1
_ _ ~...
- ;R l't" E
g
- k..;',~h-h
[v.
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Specimen V14 fF R A M AT O M E 5-17
I Figure 5-7. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces -Weld Metal SA-1526 (299L44 / 8596)
I
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4-0
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I I
ffMTNRP.5 l
5-18
Figure 5-8. Charpy impact Data for Irradiated Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT) l 100 0
0 75 l
50 w
25 e,
0 100 0
100 200 300 400 500 600 Termerature, F l
100 f;.
1" g 40 0
l
-100 0
100 200 300 400 500 600 Termerature, F 200 T,or:
M 180 Tasus: 2.121f l
T:
fjZZI 160 T o:
21Hf i
OwuSE INLfub RT,cr: $
l 34o 2
d_120 B
i r 1m w
e 1=
x 1
40 l
Onentaba lagtudiM I
g
{
Fluenz:
1.59k10Em/
HeatfArttier fddlil I
0 i
-100 0
100 2W 300 400 500 600 Temperature, F h
L s.,,
,s
.em l
1 l
Figure 5-9. Charpy impact Data for Irradiated Base Metal Plate Heat.Affected. Zone Material 100 r
- 75 I
50 l
w h 25 0
100 0
100 200 300 400 500 600 Temperature, F I
, 100 E
80 I
60 e
a 40 e
l 20 1:
0 100 0
100 200 300 400 500 600 Temperature, F 200 Tuor:
ft.Q.
Matenat SA-533 Gr. B CL 1 180 Tuuta : M Orientation: BAZ T:
dif Fluence:
1.599x10.n!&m' 160 Tn:
22 1 Heat Number; C44151 CvUSE: 63 ft Ib 140 ke 120 g 100 m
f 80 e
E 60 m
.......9....................................
40 20 0
- 100, 0
100 200 300 400 500 600 Tr,mperature, F 5-20 I!MTNAM
Figure 5-10. Charpy impact Data for Irradiated Weld Metal SA-1526 (299L44 / 8596) 100 r
- g 75 a
e l 60 t
h 16 e
0
^:
100 0
100 200 300 400 500 600 Temperature, F 100
,=
E 80 bg 60
,E 40 0
0 f 20 0
100 0
100 200 300 400 500 600 Temperature, F 200 f uor :
tLfL 180 T sua : +278 F Tso:
N/A T o-
+210 F 160 i
CvUSE: 22.!L!b "D'
140 E
2 120 Pg 100 su U
l g 80 Ji 60 40
?
Matenal:
Wcid Metal 20 Fluence:
1322x10_D/cm' Heat Number: SA-1526 0
-100 0
100 200 300 400 500 600 Temperature, F I "<^ TNM '
5-21
Figure 5-11. Charpy impact Data for Irradiated Correlation Monitor Plate Material (HSST Plate 02), Longitudinal Orientation (LT) 100 p 75 50 25 so 0
100 0
100 200 300 400 500 600 Temperature, F 100 Y
80 h
e g 60 Lg 40 20 0
100 0
100 200 300 400 500 600 Temperature, F 200 Tuor:
NIL 180 Tuute : +221 F T,:
t24g.L 160 Tso:
+188 F CvUSE: 82 1 @
"DT U
140 E
2 120 j
g 100 W
U 80 G
i 60 Material:
HSST PlateJ2 40 Orientation: Lonattudinal Fluence:
1.599x10dGm' 20 Heat Number; A11951 0
0 0
100 200 300 400 500 600 Temperature, F l
fme.e.e.m l
s.22
Figure 5-12. Photographs of Charpy impact Specimens Fracture Surfaces, Base Metal Plate Heat No. C4415-1 r
\\Y:: l cs,
- {,,
. 1. :
3,
\\
" m' - -
Sgwelme n No. V63, Test Temperature 0*r Specimen h V61. Test Ternperaturc Ilia'r r
Sgweimen No. V60, Tcat Temperature 6FF Specimen No.V64, Test Temprature 2.WF i
Spnimen k V57, Test Terngerature 115'E Specimen No.V58. Test Temperature 34Hp*F l
Spetinien k V59, Test Temperature 175'r Specimen h V62, Test Temperature 400'F f M 9o^7R M 5-23
)
)
Figure 5-13. Photographs of Charpy impact Specimens Fracture Surfaces, Base Metal Plate Heat-Affected-Zone Material l
I
(
g g;l'
- .r, '
' ' ~
+ ~.,
' f 'f y,V{ -
l.,I, '
fk i.,
k',
~
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Spuimen Nn. Ill7.Tcsl Temperature O'F Specimen No. Il24. Test Temperature 20df'F
~~
t
' ', ' ]!!M.
Q?~;;i l
- 4h.
,.ra w.,.
f Spetirnen No. If2J, Test Teengeratuit WF Specimen No. Ill9, Test Temperature 250*F 1
4
- 37 g g<
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- .5
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>w r
I
.)
+
r
~
Sgwrimen No.1820,Tmt Temperatu re 175'F 5pecimen No. II22. Test Temperature 400*1 ffM%^JRM's 5-24
l Figure 5-14. Photographs of Charpy impact Specimens Fracture Surfaces, Weld Metal SA-1526 (299L44 / 8596) l I
~...
r.', ' ' i
.kh!h5
< J',Ts' t*e s
l s.r.,
+
p
, _ l l
E.,y, j
Spedinen No. W23, Test Temperature 0*F Specimen No. W22. Test Temperature 2!#F e
l Specimen No. WIW. Test Temperat urc 68'F Specimen No. W20. Test Temperature 300'E Specimen No. W19. Test Temperature 125'r Specimen No. W21. Test Temperature 350* F Spwimen No.W24 Test Temperatart 200*r S pecimen No. W 17. Test Tempe rature 400* r 5-25 IMTNRP.'s
Figure 5-15. Photographs of Charpy impact Specimens Fracture Surfaces, Correlation Monitor Plate Material (HSST Plate 02)
Specimen No. R50. Tot Temperature (FF Spctirnen Na RSI,Tmt Temperature 254PF
=
$3kO
rf: b53[15$)
?
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(
l I
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4..
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w
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e-
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, - y Spetimen No. R54. Tat Temperature 125'r Specimen No. R55. Tot Temperature 35tFF l
Sperimes; No. R52. Tat Temperature 200*f Specimen No R49. Tot Temperature 400*F l
l l
l l
5-26 fMWNRM I
i
- 6. Neutron Fluence 6.1. Objectives and Procedure Over the last thirteen years, Framatome Technologies, Inc. (FTI) has developed a calculational-based fluence analysis methodology that can be used to accurately predict the fast neutron fluence in the reactoi sessel using surveillance capsule dosimetry or cavity dosimetry to verify the fluence predictions."5J5) The methodology was developed through a full-scale benchmark experiment that was performed at the Davis-Besse Unit 1 reactor.05) The results of the benchmark experiment demonstrated that the accuracy of a fluence analysis that employs the FTl methodology would be well within the NRC-suggested limit of 20%.05J7)
A calculational-based fluence analysis has been performed to calculate the neutron fluence exposure to the pressure vessel, certain vessel welds in the beltline region, and surveillance Capsule X of the Surry Unit No.1 reactor vessel. This section briefly describes the analysis that was used to determine the fluence and presents the results of the analysis at allocations of interest (see belon). A more detailed discussion of the methodology is provided in Appendix E.
The fast neutron fluences (E > 1 MeV) at all points of interest were calculated in I
accordance with the requirements of the Draft Regulatory Guide DG-1053,"U as described in detail in the FTl fluence topical report, BAW-2241P.05) i The energy-dependent flux at the capsule was used to determine the calculated activity of l
each dosimeter. The calculated activities were adjusted to account for known biases l
(photofission and U-235 impurity in the U-238), and compared directly to the measured
)
activities. It is noted that the measurements are not used in any way to determine the magnitude of the flux or the fluence. They are used only to show that the calculational results are reasonable and to show that the Surry Unit No.1 results are consistent with i
the FTl benchmark database of uncertainties. This is discussed in more detail in Appendix E.
f I
6-1 INNM
Explicit values of the fast fluence were computed for the following locations:
pressure vessel inside surface maximum location (IS),
Maximum location on the lower shell Maximum location on the intermediate shell Longitudinal welds:
L1, L2, L3, & L4 Circumferential welds:
WO5 and WO6 The multi-cycle-average full power flux at each of these locations was calculated in two stages, first for cycles 1 through 12, and then for cycles 13 through 14. The corresponding fluence at each location was then calculated by comp,uting the product of each flux by the appropriate effective full power time, in seconds. The total fluence on each weld and plate and at the capsule was then calculated by adding the fluences together, point by point.
The calculated full power flux for cycles 13 through 14 was then used to determine the
" extrapolation flux", which is given in Table 6-4, Column 4. The extrapolation flux was used to determine the fluences at 28.8,29.6, and 48 EFPY, which are all reported in Table 6-4.
6.2. Results The results are presented as follows:
l I
Calculation to Measurement Ratios in Table 6-1 Flux and fluence results at all points of interest in Tables 6-2,6-3, and 6-4/Section D for uncertainty discussion.
Fast flux axial profiles at the reactor vessel surface at three azimuthal positions. These plots present the effect of the poison rods in three dimensions (Figure 6-1).
Surface plot of inside surface fast flux (Figure 6-2).
fmm.m l
e.2
Table 6-1. CIM Ratios for Surry Unit No.1 Capsule X I
Measured Calculated Specific Activity,.
' Activity,'.
pCl/g
~'
. pCi/g Ratio
, :., C
.C,
.C,
M' C/M Dosimeter Cycles 1-12 Cycles 13-14 TOTAL Fe 58.44 809.5 867.94 754.294 1.1507 Cu 3.462 2.483 5.945 6.828 0.8707 U-238 13.63 4.235 17.865 16.16 1.1055 Average =
1.0423 l
l l
)
I l
l e.s
< e
.s i
Table 6-2. Fast Neutron Flux (E > 1 MeV)
Cycle Lo'ngth, e Cycles 1-12 ;
Cycles 13-14 (EFPD) 4863.142 966.594 (EFPS) 4.2018E+08 8.3514E+07 (EFPY) 1.3315E+01 2.6464 E-00 Fast Flux Fast Flux Peak Flux Location (n/cm^2/sec)
(nicm^2/sec) -
For Welds Longitudinal L1 & L2:
8.46090E+09 5.32384E+09 L3 & L4:
8.45090E+09 7.05534E+09 Circumferential WOS:
4.65630E+10 2.76254E+10 WO6:
5.47950E+09 5.94072E+09 For Shells intermediate:
4.66450E+10 3.45507E+10 Lower:
4.66450E+10 3.14629E+10 For Capsule X 3.04200E+10 3.84404E+10 1
i l
l I
1 l
l I
l sa r.mem
I I
Table 6-3. Incremental Fast Neutron Fluence (E > 1 MeV)
Cycle' 1 - 12 Cycles 1 - 12 Cycles 13 - 14 Cycles 13 - 14
~
. Cycle Longth s
(EFPD) 4863.142 966.594 (EFPS) 4.2018E+08 8.35137E+07 l
l Fast Flux inc Fluence' l Fast Flux.-
inc Fluence Peak Flux Location n/cm^2/sec (n/cm^2) nlcm^2/sec.
(n/cm^2)
For Welds l
Longitudinal L1 & L2:
8.45090E+09 3.55086E+18 5.32384E+09 4.44614E+17 i
L3 & L4:
8.45090E+09 3.55086E+18 7.05534E+09 5.89218E+17 l
[
Circumferential WOS:
4.65630E+10 1.95846E+19 2.76254E+10 2.30710E+18 WO6:
5.47950E+09 2.30235E+18 5.94072E+09 4.96132E+17
)
For Shells I
Intermediate:
4.66450E+10 1.95991E+19 3.45507E+10 2.88546E+18 Lower:
4.66450E+10 1.95991E+19 3.14629E+10 2.62758E+18 For Capsule X 3.04200E+1')
1.27817E+19 3.84404E+10 3.21030E+18 l
l l
l l
ffMMWA 6-5
8 0
88 98 9 9 Y
+
1 1 1 1 1 1
+ +
+ +
+ +
P E
EE EE EE F
0 6 7 1 5 9 1 7 A
E 1
9 4
7 9 63 1 3 N
2 7 7 66 67 6
3 8 7 7 5 37 9
9 2
2 1 33 7 5 67 35 33 8
0 88 98 99 Y
+
1 1 1 1 1 1 P
E
+ +
+ +
+ +
EE EE EE F
68 E
8 69 0 1 8 4 A
8 8
3 7 9
4 89 N
2 39 6
6 8 7 8
0 59 0
0 4 9 s
8 9
1 9 3
2 64 i
2 s
66 3
5 33 y
la 9 9 0
9 00 n
00 1
0 1 1 A
P
+ +
+
+
+ +
AX EE E
E EE RU 4 4 4
2 7 9 A
1 83 5
7 02 N
TL 3 5 2
0 5 6 o
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9 41
)
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^
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88 9
8 99
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0 9
1 1 1
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+
1
+ +
+
+
+ +
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(
4 E6
+
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E EE E
r E
3 5 7 8 7
8 5 7 u
C C
59 2
4 0 1
4 4 6 9
3 S
N O
5 5 0 7
8 82 9
- 0. 1 9 4 8
9 4 2 E
E s
U 9 1 1
7 22 5
5 1
e L
3 4 2
2 22 c
F ne E
8 8 9
8 99 9
V u
I 1 1 1
1 1 1 1
l T
2
+ +
+
+
+ +
+
1 1
EE E
E EE E
F A
3 66 6
5 1 1 2
e L
C 8 8 4
3 99 3
8 v
U O
00 6
2 99 1
7 M
E 5 5 5
0 5 5 i
la U
5 5 9, 3 99 2
t 1
33 1
2 1 1 u
C mu 4
C 7 7 8
7 88 1
7 0 e
8 o
00 c) 1 1 1 1
1 1 1 n2
+ +
+
+
+ +
t 4 + -
+
4 9EE e^
EE E
E EE E
3 5 7 9 6 3 3 lum 4 8 0
2 68 3
6 1
1 1 1 3 4 5 0
s 1 6 F c 6 2 7 1 5 7 e
6 i
1 b
l 9 5 4 e
36 c n 4 9 06 82 a
y 82 in(
4 8 39 86 2
l c
3 45 2 4 22 T
C 2
)
81 88 98 99
)
1 c
9 2
00 xe 1 1 1 1 1 1
^
o 2
1
+ +
+ +
+ +
+ +
EE lus
+
t 4
EE EE EE E
m 1 5 6 F2 6 6 6 5 1 1 1
1
^
2 c
37 4 t
8 8 4 3 99
/
s 8
sm 0 0 62 99 n
61 1
(
le 7
303 a c 5 5 5 0 5 5 E
c 4 2 3 Fi 5 5 9 3 99 2
y n
1 C
C 4 1 33 1 2 1 1
(
N E
U n
2: 4:
S: 6 er L
o LL O O t
e a w l WW i o F
h it d
t a
L a
eL s
g c
1 3 it A
n n
) ) )
o LL m
DSY L
la n
T e
e r
N L
PPP n
e e
m FF F x
si r
lst ic E
e EEE u dd e
lei e
n M
lc
( ( (
l let m
h p
f F
u E
y Wig u
S S
k R
C a
r n c
C e
r r
r oo i
o o
N P
FL C
F F
I
Table 6-5. Locations of Pcaks Cycles 1 12 -
- Cycles 1314
,R(cm):
Theta (deg)
_ Z (cm) -
1R(cm)
Theta (deg)
Z (cm)
(a)
(b)-'
(c)
~
For Welds Longitudinal L1 & L2:
199.790 45 131.02 199.790 45.000 131.02 L3 & L4:
199.790 45 131.02 199.790 45.000 277.89
)
Circumferential I
WOS:
199.700 0.875 132.84 199.790 0.380 132.84 WO6:
199.790 0.875 365.6 199.790 0.380 365.6 l
For Shells Intermediate:
199.790 0.875 131.02 199.790 0.380 277.89 Lower:
199.790 0.875 131.02 199.790' O.380 57.88 I
For Capsule X 192.4 25 192.4 15 (a) Radiallocation relative to center of the core.
(b) Axial location relative to lower active fuel.
(c) Azimuthallocation relative to core major axis.
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- 7. Discussion of Capsule Results l
7.1. Copper and Nickel Chemical Composition Data l
To date, several chemical analyses have been performed on the Surry Unit No.1 RVSP weld metal (SA-1526). These analyses have been performed on the unirradiated surveillance weld metal, broken weld metal Charpy specimens tested as part of the Surry Unit No.1, Capsule V analysis, and a broken weld metal Charpy specimen tested as part of the Surry Unit No.1, Capsule X analysis. The copper and nickel chemical content results of these chemical analyses are presented in Table 7-1A. The mean of these data represents the best-estimate copper and nickel chemical contents for the Surry Unit No.1 RVSP weld metal, i.e. Cu = 0.23 wt% and Ni = 0.64 wt%.
The mean copper and nickel contents for the base metal plate heat no. C4415-1 is based on three analyses performed on the unirradiated surveillance base metal, a broken base metal Charpy specimen tested as part of the Surry Unit No.1, Capsule V analysis, and a broken base metal Charpy specimen tested as part of the Surry Unit No.1, Capsule X analysis (see Table 7-1B). The mean of these data represents the best-estimate copper and nickel chemical contents for the Surry Unit No.1 RVSP base metal plate heat number C4415-1, i.e. Cu = 0.10 wt% and Ni = 0.49 wt%.
7.2. Unirradiated Material Property Data
{
A weld metal representative of the controlling weld mdal was selected for inclusion in the Surry Unit No.1 surveillance program in accordance with me criteria in effect at the time
)
the program was designed. The applicable selection criterion was based on the unirradiated properties only. A review of the original unirradiated material properties of I
the reactor vessel core beltline region materials indicated no significant deviation from expected properties except in the case of the upper-shelf energy propedies of the weld metal. Based on the designed end-of-service peak neutron fluence value at the %T 7-1 I I< W M
~ vessel wall location and the copper content of the weld metal, it was predicted that the end-of-service C,USE would be below 50 ft-lbs.
The unirradiated mechanical properties for the Surry Unit No.1 RVSP materials are summarized in Appendices B and C of the report.
7.3. Irradiated Property Data 7.3.1. Tensile Properties Table 7-2 compares the irradiated tensile properties from Capsule X with the tensile properties from the unirradiated tensile specimens. At both room and elevated l
l
. temperatures, the ultimate and yield strengths change in the surveillance base metal plate as a result of irradiation and the corresponding changes in ductility are within the l-ranges observed for similar irradiated materials. There is some strengthening, as indicated in the increases in ultimate and yield strengths and decreases in the ductility properties. The changes in tensile properties for the surveillance weld metal at both room and elevated temperatures, as a result of irradiation, are also within the observed ranges for similar irradiated materials. The strengthening in the surveillance weld metal,
. indicated by the increases in ultimate and yield strengths and decreases in ductility, is
- greater than that observed in the base metal.
A comparison of the tensile data from previously evaluated Surry Unit No.1 capsules (Capsule T and Capsule V) with the corresponding data from Capsule X is shown in Table 7-3. The general behavior of the tensile properties as a function of neutron irradiation is an increase in both the ultimate and yield strengths and a decrease in ducti!ity as measured by the total elongation and reduction of area.
7.3.2. Impact Properties Tables 7-4 and 7-5 compare the observed (measured) changes in irradiated CVN impact properties from Capsule X with the predicted changes in accordance with Regulatory Guide 1.99, Revision 2." The radiation-induced changes in toughness of all the Surry Unit No.1 surveillance materials are summarized in Table 7-6.
1 1.
I 7-2 INNM o
The observed 30 ft-lb transition temperature shifts for the surveillance materials are in agreement with the values predicted using Regulatory Guide 1.99, Revision 2. The measured values of the 30 ft-Ib transition temperature (ATx) fall within one or two standard deviations of the AR% predicted by Regulatory Guide 1.99, Revision 2, Position 1.1 (see Table 7-!;. The observed percent reductions in C,USE due to irradiation are also in agreement with the predicted values for the surveillance materials (see Table 7-5).
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Table 7-1 A. Copper and Nickel Chemical Composition Data for Surry Unit No.1 Reactor Vessel Surveillance Weld Metal SA-1526 Cu Ni Analysis Source Wt%
Wt%
Reference RVSP Baseline Chemistry (Uriirradiated) 0.25 0.68 WCAP-7723 (RVSP Description)
CVN Specimen: W-10 0.24 0.64 WCAP-11415 (Capsule V)
CVN Specimen: W-1 0.25 0.66 BAW-230208)(Suppl. Analysis)
CVN Specimen: W-5 0.26 0.67 BAW-2302(Suppl. Analysis)
CVN Specimen: W-9 0.24 0.66 BAW-2302 (Suppl. Analysis)
CVN Specimen: W-16 0.24 0.65 BAW-2302 (Suppl. Analysis)
CVN Specimen: W-1 0.21 0.66 BAW 2302 (Suppl. Analysis)
CVN Specimen:W-5 0.22 0.65 BAW-2302 (Suppl. Analysis)
CVN Specimen: W-9 0.20 0.64 BAW-2302 (Suppl. Analysis)
CVN Specimen: W-16 0.20 0.65 BAW-2302(Suppl. Analysis)
. CVN Specimen: W-22 0.18 0.49 l Capsule X Table 7-18. Copper and Nickel Chemical Composition Data for Surry Unit No.1 Reactor Vessel Surveillance Base Metal Plate (C4415-1)
Cu Ni Analysis Source Wt%
Wt%
Reference RVSP Baseline Chemistry (Unirradiated) 0.11 0.50 WCAP-7723 (RVSP Description)
CVN Specimen: V-25 0.115 0.569 WCAP-11415 (Capsule V) 0.08 0.41 Capsule X CVN Specimen: V-58..
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Table 7-2. Comparison of Surry Unit No.1 Capsule X Tension Test Results
. Room Temp.
Elevated Temp.(550 F)
' Tension Test
= Tension Test Unirradiated(a)
Irradiated Unirradiated(s)
Irradiated Base Metal Plate,C4415-1 Longitudinal Fluence,1018 n!cm2(E>1 MeV) 0 1.599 0
1.599 Ultimate Tensile Strength, ksi 93.8 108.3 89.7 101.3 i
0.2% Yield Strengtn, ksi 71.8 86.0 62.6 76.1 Uniform Elongation, %
13.6 9.8 13.8 8.4 Total Elongation, %
24.4 22.6 24.6 19.1 Reduction in Area, %
69.8 59.3 68.0 60.1 l
l Weld Metal, SA-1526 299L44 /8596 Fluence,1019 n/cm2(E>1 MeV) 0 1.599 0
1.599 Ultimate Tensile Strength, ksi 83.2 111.3 79.0 101.3 0.2% Yield Strength, ksi 69.7 97.9 58.1 85.1 Uniform Elongation, %
15.2 10.5 14.2 7.7 l
Total Elongation, %
26.5 21.7 22.9 16.3 Reduction in Area, %
66.7 51.3 62.0 49.9 (a) Average of data presented in Appendix A.
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- 8. Summary of Results The analysis of the reactor vessel material contained in the fourth surveillance capsule, Capsule X, removed for evaluation as part of the Surry Unit No.1 Reactor Vessel Surveillance Program, led to the following conclusions:
2
- 1. The capsule received an average fast neutron fluence of 1.599 x 10" n/cm (E > 1.0 I
MeV).
- 2. The 30 ft-Ib transition temperature for the base metal plate, heat no. C4415-1, I
2 increased 86 F after the irradiation to 1.599 x 10" n/cm (E > 1.0 MeV). In addition, the C,USE for this material decrease 24%.
I
- 3. The 30 ft-lb transition temperature for the weld metal, SA-1526, increased 234 F 2
after the irradiation to 1.599 x 10" n/cm (E > 1.0 MeV). In addition, the C,USE for this material decrease 44%.
- 4. The correlation monitor plate demonstrated similar behavior after exposure to the I
same fluence,1.599 x 10" n/cm (E > 1.0 MeV). The 30 ft-Ib transition temperature 2
increased 142*F and the C,USE for this material decrease 33%.
- 5. The measured increases in 30 ft-Ib transition temperature for the Surry Unit No.1 RVSP base metal plates and weld metal fall within one or two standard deviations of the ART predicted by Regulatory Guide 1.99, Revision 2, Position 1.1.
- I
'I I
I fmm.m i
e.,
I I
I l
- 9. Certification The specimens obtained from the Virginia Power Surry Unit No.1 surveillance capsule (Capsule X) were tested and evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10 CFR 50, I
Appendices G and H.
)
l b0 1W
[lS~l9?T M. J'. D6 Van (Material Analysis)
Date Materials & Structural Analysis Unit
$fh I!
SY97 S. Q. King'(Fluenchnalysis)
' Ddte Performance Analysis Unit This report has been reviewed for technical c tent a d,accurac.
9
%/u
'3 H/Xu,(Materi61 Analysis)
Date 5
Materials & S ructur Analysis Unit hr
.5f5 5
W
/ J. R. \\Sorsham (Fluence Analysis)
Date Performance Analysis Unit Verification of independent review.
S&h l
~~
k.' E.'Ifoore, Manager Date lg Materials & Structural Analysis Unit i5 This report is approved for release.
D. L. H'owell Date Program Manager I
f~~"
l'
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l
- 10. References
- 1. S. E. Yanichko, " Virginia Electric and Power Co. Suny Unit No.1 Reactor Vessel l
Radiation Surveillance Program,"WCAP-7723, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, July 1971.
i
- 2. J. S. Perrin, et al.. " Final Report on Surry Unit No.1 Pressure Vessel Irradiation Capsule Program: Examination and Analysis of Capsule T to Virginia Electric and l
Power Company," Battelle Columbus Laboratories, Columbus, Ohio, June 24,1975.
,3. J. S. Perrin, et al.,
- Final Report on Surry Unit No.1 NuclearPlant ReactorPressure Vessol Surveillance Program: Examination and Analysis of Capsule W to Virginia
)
Electric and Power Company," BCL-585-8R, Battelle Columbus Laboratories, Columbus, Ohio, March 30,1979.
- 4. S. E. Yanichko and V. A. Perone, ' Analysis of Capsule V from the Virginia Electric and Power Company Suny Unit 1 Radiation Surveillance Program,"WCAP-11415, 7
Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, February 1987.
- 5. ASTM Standard E 185-73,
- Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials, Philadelphia, Pennsylvania.
- 6. Code of Federal Regulation, Title 10, Part 50, " Domestic Licensing of Production r
and Utilization Facilities," Appendix G. Fracture Touahness Requirements.
- 7. Code of Federal Regulation, Title 10, Part 50, " Domestic Licensing of Production and Utilization Facilities," Accendix H. Reactor Vessel Material Surveillance Proaram Requirements.
- 8. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Ill, " Nuclear Power Plant Components," Accendix G. Protection Aaainst Nonductile Failure.1989 Edition.
- 9. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, " Rules for Inservice Inspection of Nuclear Power Plant Components," Appendix G. Fracture Touahness Criteria for Protection Aaainst Failure,1989 Edition.
10-1 IMMM
- 10. ASTM Standard E 208-81, " Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," American Society for Testing l
and Materials, Philadelphia, Pennsylvania.
)
- 11. ASTM Standard E 8-96a, " Standard Methods of Tension Testing of Metallic Materials," American Society for Testing and Materials, Philadelphia, Pennsylvania.
- 12. ASTM Standard E 21-92, " Standard Test Methods forElevated Temperature Tension Tests of Metallic Materials," American Society for Testing and Materials, Philadelphia, Pennsylvania.
- 13. ASTM Standcrd E 23-96, " Standard Methods forNotched BarImpact Testing of Metallic Materials," American Society for resting and Materials, Philadelphia, Pennsylvania.
- 14. ASTM Standard E 185-94,
- Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing l
and Materials, Philadelphia, Pennsylvania.
- 15. J. R. Worsham, et. al., " Fluence and Uncertainty Methodologies", BAW-2241P.
Framatome Technologies, Inc., Lynchburg, Virginia, April,1997.
- 16. L. B. Gross, et. al., " Demonstration of the Management of Aging Effects for the Reactor Vessel," BAW-2251, Framatome Technologies, Inc., Lynchburg, Virginia, June 1996.
- 17. U.S. Nuclear Regulatory Commission Draft Regulatory Guide DG-1053, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", June 1996.
- 18. U. S. Nuclear Regulatory Commission, " Radiation damage to Reactor Vessel Material,"Reaulatorv Guide 1.99. Revision 2, May 1988.
- 19. M. J. DeVan, " Weld Metal (Weld Wire Heat Number 299L44) Chemical Analysis and Evaluation," BAW-2302, Framatome Technologies, Inc., Lynchburg, Virginia, July 1997.
10-2 fMMM5
}
}
f t
l 1
l APPENDlX A Instrumented Charpy V-Notch Specimen Test Results Load-Time Traces 1
A IMTNRM
)'
)
l PROJ. NO. 55761 QA NO, 97004 LOAD - TIME TRACE FOA SPECIMEN V63 000o ll
.4 i-f gj a
1:
f h.-
g..
l
~
~
y:_ - :
w.
I 7.s4=
...gg,_f.. 7
.747.
Figure A-1. Load-Time Trace for Charpy V-Notch Impact Specimen V63 PROJ NO. 65761 GA NO. 97004 LOAD - TIME TRACE FDA SPECIMEN V60 9D00 4490
=
,E g
anco 1400 y
-400
.sgo,
. 548 7.880 p.e47
.esa
.747s a.sa7 Figure A-2. Load-Time Trace for Charpy V-Notch impact Specimen V60 A-2 I'"M ^Na'
PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN V57 amo g,
ql y
E o.
g-W.__
1
...o.
- a. g,,,,,,3..
7 m..
- 7. u.
...a
.w.
Figure A-3. Load-Time Trace for Charpy V-Notch Impact Specimen V57 PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACC FOR SPECIMEN V59 h
,l 1
==
E i
g I
==
b ~~~
l l
l
.a.
l ano
)
-me a
=
.sano
)
m.sae
- 7. san
..esa
.747e a.g
,s.e47 Figure A-4. Load-Time Trace for Charpy V-Notch Impact Specimen V59 r
i I
I A.s
-e.em L
PROJ. NO, 55761 QA NO. 97004 LOAD - TIMC TRACE FOR SPECIMEN V61 econ 4400 Ia E m.
g.
m a
~.952
.7476 2.
3.947 S.548 7.848 Figure A-5. Load-Time Trace for Charpy V-Notch impact Specimen V61 PROJ. NO. 55761 GA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN V64 ocuo deco
,l *I smoo E
.g am 3000 labD M
)
..a.
-8000 m
a a
a i
a i
a a
".952
.7476 2.
3.947 S.345 7.343 f
Figure A-6. Load-Time Trace for Charpy V-Notch Impact Specimen V64 I
t I
n nw.,a
L l
PROJ. NO. 55701 QA NO. 97004 I
LOAD - TIME TRACE FOR SPECIMEN V58 I
I,eE I1
=
g E
.g y
i B-g.
L
=
-.a.
7....
e. gg,,,,_,...
.w.
Figure A-7. Load-Time Trace for Charpy V-Notch Impact Specimen V58 I
PROJ. NO. 55761 GA ND. 97004 LOAD - TIME TRACE FOR SPECIMEN V62 aooo 4400
,E
.p m
2000 g,.
7....
.. gg,,,,..,
.w.
Figure A-8. Load-Time Trace for Charpy V-Notch impact Specimen V62 A-5 I"<MNRM
PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOA SPECIMEN H17 F.
h 44e0 d
. a.
/
ii amm g
.p pore g.
=
=
l O_...
-.=
-1000 J
a a
a r.
= ogm,,,,,.. 7 7....
Figure A-9. Load-Time Trace for Charpy V-Notch Impact Specimen H17 PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN H23 ewo 4400 i
,R
.. g g esto B-g...
i i
M
-son 1
.952
.747S 2.
3.947 5.5 6 7.846 Figure A-10. Load-Time Trace for Charpy V-Notch Impact Specimen H23 1
1 f="<MNRP 5 A-6
)
PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN H2$
oooo h, _'"
~
]
l
,5 g: g y.~
@m g,_
=
j o.eaa v.sna
- a. gay,,,,,,;. e
..,a
.,no i
Figure A-11. Load-Time Trace for Charpy V-Notch impact Specimen H21 l
PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN H2O i s
.ooo n
,l ",
g-- 4g--
g B.=
f,a m
m 7.i.e a. gg,,,,,,y...
.a
., u.
Figure A-12. Load-Time Trace for Charpy V-Notch impact Specimen H2O A-7 I"<MN M
PROJ. NO. 55761 GA NO. 9/004 LOAD - TIME TRACE FOR SPECIMEN H24 o
I g
.cco
- g. R g
sooo
$ 406 aco
-400
. sono a
a a
i i
z.e47 s.eae 7.sas z.sg
.esa
.747e Figure A-13. Load-Time Trace for Charpy V-Notch Impact Specimen H24 l
PROJ. NO. 55761 GA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN Hi9 aoco 440a iI smoo y
i R
g mo um 900o 840o
-400
)
7...e
.,4,.
.. g m,,,,,... 7 Figure A-14. Load-Time Trace for Charpy V-Notch Impact Specimen H19 l'
)
A-8 fe" A T N R h "
.i
PROJ. NO. S5761 QA NO. 97004 LOAD - TIME TAACE FOR SPECIMEN Hf B l a
.ono s
l*
~
5 ln g.
ecco fs.co ano
-400
-8000 e,se7 sa,e7 av.s7 e.g
-a.sz s.ee9 Figure A-15. Load-Time Trace for Charpy V-Notch Impact Specimen H18 i
PROJ. NO, 55761 GA NO. 97004 LOAD - TIME TAACE FOR SPECIMEN H22 eooo
.o
,1 s ll
=
/
g mo l.00
\\
m i
3.947 S.540 7.148
{
.952 7476 2
l l
Figure A-16. Load-Time Trace for Charpy V-Notch impact Specimen H22
)
i fIMTNAM A-9
PROJ. NO. 55761 GA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN W23 el
{
i'i lE
~
N g.=
f g
".oo
- k. - _ ;
.,e
.852
.74ye 3.3 3.947 5.646 7.146 Figure A-17. Load-Time Trace for Charpy V-Notch Impact Specimen W23 l
PROJ. NO. S5761 QA NO. 97004 LOAD - TIME TRACE FOR GPECIMEN WiB De ag
,\\ I i
a
/
aman k-m.
y=
I g
g.
^
-.oo 7...S
.. g,,_,...o
. 7 o.
Figure A-18. Load-Time Trace for Charpy V-Notch impact Specimen W18 l
l L
A-10 I'"<M
PA0J. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN W19 I g
.ooo e
i,
- R
(
ln f _.
g.
=
.952
.747S 2.
3.947 S.546 7.
46 Figure A-19. Load-Time Trace for Charpy V-Notch Impact Specimen W19 i
PROJ. NO. 55761 GA NO. 97004 LOAD -- TIME TRACE FOR SPECIMEN W24 sooo 4 00 "k~
~
ln
(
g_9000 l
g. 0.
~
-400
.. g,,,'_,.'. 7 10D0 7.
4.
.>.7.
Figure A-20. Load-Time Trace for Charpy V-Notch Impact Specimen W24 f"MWai I-A-11'
PA0J. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN W22 noco e
.g lli_
O g==
t me g..
=
=
-soo
".952 7475 2.
3.ed7 S.SAS 7.146 Figure A-21. Load-Time Trace for Charpy V-Notch impact Specimen W22 PROJ. NO. 55761 GA ND. 97004 l
LOAD - TIME TRACE FOR SPECIMEN W20 9nF gg 44co i ~i a==
g Er.
maan t
6 i
lenoo l
g,_
eco
~400
.sooo e.sae 7 ase z.g
,3.e47
.es2
.7475 Figure A-22. Load-Time Trace for Charpy V-Notch Impact Specimen W20 A-12 I"MN=D "
PROJ. NO. 55761 GA NO. 97004 LOAD - TIME TRACE FOA SPECIMEN W21 noco 44 o i
i E
Ia gacao 4400 aco
-aco e.sae 7.sde 2g
,3.e47
.esa 747e Figure A-23. Load-Tima Trace for Charpy V-Notch Impact Specimen W21 PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN Wi7 cooo 440o
,5 1
.p
==
g oo f=
ca 84 0 l
m
-4
.s m
- .952
.7475 2.
3.947 S.546 7.see 1-L Figure A-24. Load-Time Trace for Charpy V-Notch Impact Specimen W17 I-(
A-13 IIMNRM
I PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN A50
- i. !
I ewo i ~l
==
R l
gg
=
y m.
e.-
{ i.=
- =
-40o
_.932
.7475 2.
3.947 S.545 7.Sd5 Figure A-25. Load-Time Trace for Charpy V-Notch Impact Specimen R50 3
PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN A56 4400
,i I asa R
-p g
e.-
g.
~
Sv
.=,v
=.952
. 7 4 7t$
2.
3.947 S.545 7.Sd5 Figure A-26. Load-Time Trace for Charpy V-Notch impact Specimen RS6 A-14 IMTNRM a
PROJ. NO. 55761 GA NO, 97004 LOAD - TIME TRACE FOR SPECIMEN RS4 i s
.oco n
i ~i amo g
~
oo
.ono
$400
~
-<oo e.eme 7.sae a.sj{
,s. co
.esa
.747e Figure A-27. Load-Time Trace for Charpy V-Notch Impact Specimen R54 l
PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN RS2 eum
.ll
,n -
1
== --
E g
g.~
l B,oon g.
1
~
-<en f
.;.m e..;tu,_,=
o
. ea 7 ***
Figure A-28. Load-Time Trace for Charpy V-Notch Impact Specimen R52 i
)
A-15 fMTND
)-
PROJ. NO. 55761 QA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN R52 n
i g
.oco
.o.
,i I anoo gRg y m.
g.ooo g.
.oo
=
-.o.
- 1. San
.. g,_,..o
.w.
Figure A-29. Load Time Trace for Charpy V-Notch Impact Specimen R51 PAGJ. NO. 55761 GA NO. 97004 LOAD - TIME TRACE FOR SPECIMEN R53 auco h
a.
,l *i
==
.R
==
g
~.
acao 14ao ano l
-*e
-loco 3.e47
- s. nae 7.s4e
.esa
.747s e.g7 Figure A-30. Load-Time Trace for Charpy V-Notch Impact Specimen R53
-7
==
r.
A-16 I"N* nom
PROJ. NO. 55761 GA NO. 97004 l
LOAD - TIME TRACE FOR SPECIMEN R55 ol f
I
.==
1
- M gg
=
t g
l
@m g
1
.m.
....a 7....
e. ggm_,...o Figure A-31. Load-Time Trace for Charpy V-Notch Impact Specimen R55 PROJ. NO. 55761 QA Nfl. 97004 LOAD - TIME TRACE FOR SPECIMEN R49 0000
- as 1I mue y
pE
~
i gh g asco E -.
Nsue
~
-ace
. sono
.esa
.747s t.
,s.e47 a.e4e 7.see Figure A-32. Load-Time Trace for Charpy V-Notch Impact Specimen R49 I
A-17 f ="<MN D 5 1
l l
i l
l l
l
)
APPENDIX B Unirradiated Tensile Data l
l C
o6oe as l
Table B-1. Unirradiated Tensile Properties for the Surry Unit No.1 Reactor Vessel Surveillance Materials Test Strength Elong ation Reduction Temp.,
_. Yield,.-
- Ultimate,
, Total,
- Uniform, in Area, -
^ sik
- psl:
Material
- F:
p Base Metal Room 72,050 94,500 24.3 13.6 69.0 Plate C4415-1 Room 71,500 93,050 24.6 13.6 70.6 300 64,650 86,100 21.8 11.4 70.3 300 65,200 86,600 22.8 12.5 70.0 600 62,900 90,100 24.5 13.9 67.0 600 62,200 89,300 24.8 13.8 69.0 Weld Metal Room 70,000 83,200 26.4 15.4 68.2 SA-1526 Room 69,350 83,200 26.6 15.0 65.2 (299L44,8596) 300 60,200 75,400 24.7 14.3 67.1 300 65,450 78,250 22.4 12.8 67.5 600 56,500 77,950 22.7 13.8 62.8 600 59,750 80,000 23.1 14.6 61.2 l
B-2 IMTNRM l
APPENDIX C Unirradiated Charpy V-Notch Impact Data C-1 I'" A T O D '5
i Table C-1. Unirradiated Charpy V-Notch Properties for Surry Unit No.1 i
Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT)
Test:
Impact Lateral' Shear Specimeri Temp.,
Energy,;.
Expansion,'
. Fracture, No.
F:
ft-Ib - '
mils.
-50 9.0 7
9
-50 8.0 4
5
-50 11.0 8
9
-30 15.0 18 15
-30 12.0 18 12
-30 20.0 23 20 l
-10 33.0 27 25
-10 36.0 29 25
-10 33.0 26 18 10 36.0 29 29 10 40.0 30 29 10 44.0 34 34 40 62.0 50 48 40 67.0 55 48 40 66.0 53 43 100 100.0 72 76 100 90.0 70 79 100 102.0 72 81 160 131.0 90 100 160 124.0 88 100 160 117.0 84 100 210 125.0 89 100 210 127.0 90 100 210 124.0 85 100 1
C-2 fMSNRP"
Table C-2. Unirradiated Charpy V-Notch Properties for Surry Unit No.1 Base Metal Plate Heat-Affected-Zone Material Test Impact Lateral Shear Specimen Temp.,
- Energy, Expansion,
- Fracture, No.
- F ft-lb mils
-150 18.5 16 23 l
-150 14.0 11 13
-150 19.0 14 23
-100 53.0 32 27 I
l
-100 13.5 8
9
-100 6.0 4
5
-50 36.5 25 37
-50 31.5 25 27
-50 18.5 16 23 10 82.0 53 59 10 39.5 35 46 10 67.0 51 57 110 104.0 80 100 110 69.0 53 100 110 69.5 63 99 210 76.5 63 100 210 149.0 63 100 210 68.5 64 100 l
C-3
(=" M N = W
Table C-3. Unirradiated Charpy V-Notch Properties for Surry Unit No.1 Weld Metal SA-1526 (299L44 / 8596)
Test impact Lateral Shear
- Specimen, Temp.,
- Energy,
-Exp'ansion,.
. Fracture, No.
'F
. ft-lb mils
-150 23.5 20 18
-150 5.5 3
5
-150 5.5 3
5
-100 19.0 16 1T
-100 9.0 8
13
-100 16.0 15 13
-50 10.0 13 18
-50 17.5 18 14
-50 23.5 20 29
-25 38.0 36 29
-25 31.0 26 25 10 43.0 40 46 10 35.5 35 40 10 37.0 40 37 75 65.0 67 97 75 54.0 49 69 75 55.0 55 71 110 66.0 70 95 110 61.5 62 90 110 69.0 72 100 210 70.0 71 100 210 68.5 70 100 l
210 71.0 76 100 C-4 I"<MM M
i Table C-4. Unirradiated Charpy V-Notch Properties for Surry Unit No.1 l
Correlation Monitor Plate (HSST Plate 02),
Longitudinal Orientation (LT)
Test.
Impact
. Lateral Shear Specimen -
. Temp.,,
- Energy,
. Expansion,,
Fracture,~.
No.
- F -
. ft-Ib.
mils -
-50 5.0 3
9
-50 5.0 5
9
-50 3.0 4
9
-20 6.5 6
9
-20 9.0 10 13
-20 6.0 9
13 10 12.0 15 23 10 14.5 14 23 10 13.5 14 23 40 22.0 23 33 40 36.0 32 29 40 35.0 32 29 85 58.5 51 43 85 41.5 42 41 85 52.0 45 42 110 82.5 60 58 110 85.5 71 67 110 63.5 54 55 160 108.5 72 84 160 81.0 69 85 160 109.0 79 97 210 117.0 84 98 l
210 115.0 88 98 210 121.0 87 100 l
300 125.0 87 100 300 117.5 83 100 300 127.0 84 100 l
l C-5 IMTNMA
l Figure C-1. Unirradiated Charpy impact Data for Surry Unit No.1 Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT)
(Hyperbolic Tangent Curve-Fitting Method) 100 0
75 50 ky 25 m
0 100 0
100 200 300 400 500 600 I
Temperature, F 100 E,
80
[r l
k"O 20 I
O 100 0
100 200 300 400 500 600 Temperature, F q
200 Two,.
Q T,ute. + 12 F 180 3
T,.
- 24 F CvuSE.
ft-Ib 140 hy 120 100 e
{ 80 I
.e 60 Orentation tonoiteoinst Fluence.
None 20 Heat Number. 551 O
I 100 0
100 200 300 400 500 600 Temperature, F I
I fIMTNRf=5 C-6
Figure C-2. Unirradiated Charpy impact Data for Surry Unit No.1 Base Metal Plate Heat-Affected-Zone Material (Hyperbolic Tangent Curve-Fitting Method) 100
. 75 150 1
j 25 0
200 100 0
100 200 300 400 500 600 Temperature, F 100 2
3 e0 g
0 60 i
e
$ 40 j
20 y
0 200 100 0
100 200 300 400 500 600 Temperature, F 200 Iwo,;
fip.,
180 Tasute : M is, H
T.
72 F 3
160 CvuSE 89 ft lb 140 120 100 4010 s
60 9.....
Matenal SA-533 Gr. B Cl 1 40 Onentation. E Fluence M
20 Heat Number. C44151 0
-200
-100 0
100 200 300 400 600 600 Temperature, F L-C-7 u c~~c
=
i Figure C-3, Unirradiated Charpy Impact Data for Surry Unit No.1 Weld Metal SA-1526 (299L44 / 8596)
(Hyperbolic Tangent Curve-Fitting Method) 100 at 75 J:(
g l 50 6j 25
- s a
0
-200 100 0
100 200 300 400 500 600 l
Temperature, F l
100 80 1
60 p
h40
- 20 0
-200
-100 0
100 200 300 400 500 600 Temperature, F 200 Tuoi; N,,,,Q, 180 Tuute - E Tu-
+45 F Tu:
M 160 J
CvuSE: 70 ft-Ib 140 I
120 100 W
k 80 1
60 40 g
Material.
Weid Metal Ruence None 20 Heat Number. SA-1526 0
.?00 100 0
100 200 300 400 500 600 Temperature, F i
i e
F R AM ATO M E C-8
'<~~-'-
i j
Figure C-4. Unirradiated Charpy impact Data for Surry Unit No.1 Correlation Monitor Plate (HSST Plate 02),
Longitudinal Orientation (LT)
(Hyperbolic Tangent Curve-Fitting Method) 100
^
75 50 in.
ky 25 eo 0
100 0
100 200 300 400 500 600 Temperature, F 100 80 i
60 o-20 d
0 100 0
100 200 300 400 500 600 Temperature, F l
200 Tuot :
!i.Q.
T,ute :
+59 F i
180 3
T,(
+76F Tao-
- 46 F 160 CvuSE: 123 ft-lo RTum-R 120 e
j 100 ad E
I 80 t
60 p
40 g
Orentation: tonaitudinal pyng 20 Heat Number: A1195-1 0
100 0
100 200 300 400 500 600 Temperature, F C-9 Ia"<W M
1 I
l l'
l APPENDIX D i
Surry Unit No.1 Charpy V-Notch impact Surveillance Data Using Hyperbolic Tangent Curve-Fitting Method D-1 f.MTNRP.5
Tablo D-1. Surry Unit No.1 Capsulo T Surveillance Charpy V-Notch Data, Baso Motal Plate Heat No. C4415-1, Longitudinal Orientation (LT)
Test Impact lateral Shear Specimen Temp.,
- Energy, Expansion,
- Fracture, No.
- F ft-lb mils V46 0
13.0 13.0 5
V44 35 31.5 27.5 15 V47 77 48.0 41.5 30 V41 120 64.0 52.5 65 V43 150 81.5 63.5 55 V42 212 106.5 82.0 100 V48 295 125.0 85.5 100 V45 350 115 0 68.0 100 Tablo D-2. Surry Unit No.1 Capsuto T Survoillanco Charpy V-Notch Data, Baso Motal Plato Heat-Affectod-Zono Material Test impact lateral Shear Specimen Temp.,
- Energy, Expansion,
- Fracture, No.
F t!-Ib mils 1
i H4 0
21.5 22.5 10 H1 35 16.0 17.5 10 l
H3 35 80.0 56.0 90 H5 77 83.0 56.5 80 HB 120 39.0 39.0 60 H2 212 113.5 81.0 100 H7 295 131.0 86.5 100 H6 350 52.0 55.5 100
)
I f
1 D-2
(" M W *\\
Table D-3. Surry Unit No.1 Capsule T Surveillance Charpy V-Notch Data, Weld Metal SA-1526, (299L44 / 8596)
Test impscb 1 Lateral.
Shear Specimen Temp.,
- Energy,
" Expansion,.
- Fracture, No. -
- F ft-lb -
' mils =
W3 0
8.5 13.5 0
W4 77 20.0 21.0 10 W6 120 23.5 26.0 30 W8 212 41.0 44.0 90 W2 295 49.0 52.5 100 W7 350 53.0 45.5 100 W5 350 54.0 55.5 100 W1 390 51.0 53.0 100 Table D-4. Surry Unit No.1 Capsule T Surveillance Charpy V-Notch Data, Correlation Monitor Plate (HSST Plate 02),
Longitudinal Orientation (LT)
. Test impact Lateral Shear Specimen Temp.,
Energy,.
Expansion,
- Fracture, No'.
- F ft4b '
mils i
R33 15 4.0 7.0 0
R40 77 13.0 17.0 20 R38 120 33.5 37.0 25 l
R36 150 50.0 49.0 45 i
R39 212 68.0 62.0 70 R37 295 107.0 89.0 100 R34 350 114.5 71.0 100 l
R35 390 98.0 74.0 100 l
l t
D.3
- r. w. w.s l
l Figure D-1. Surry Unit No.1 Capsule T Surveillance Charpy impact Data Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT) l (Hyperbolic Tangent Curve-Fitting Method)
=
=
=
. 76 0
8 50 E N0
\\
a 0
00 0
100 200 300 400
$00 800 Temperature, F 100 f
.0 1, e0 e
40 e
20 0
C0 0
100 200 300 400 soo 800 Temperature, F 200 Tuot:
M T aute :
+63 F 180 3
Too
+ 86 F I o'
+47 F 2
160 CvuSE: 116 ft-Ib
- ' Y l
140 8
120 100 uJ 80 60 Matenal.
SA-533 Gr B ca 1 40 Onentat on: L onoitudinal 8
Fluence:
2 81x10 n/cm 20 Heat Number: C44151 g
0
-100 0
100 200 300 400 600 600 Temperature, F D-4 f!ME^JRF.5
Figure D-2. Surry Unit No.1 Capsule T Surveillance Charpy impact Data Base Metal Plate Heat-Affected-Zone Material (Hyperbolic Tangent Curve-Fitting Method) 100 75 50 u.
kj 25 0
100 0
100 200 300 400 500 600 Temperature, F 100 3 60 i
a l1 80
.....9.........................................
g
] 20 J
0 100 0
100 200 300 400 500 600 Temperature, F 200 Two,.
R 10 T, uta. +?4F T,g
+48 F Tag E
160 CvUSE. 99 ft-Ib RTuog: R 4.120 g
100 W10 1
60 Onentation: HAZ Heat Number. C4415-1 0
100 0
100 200 300 400 500 600
{
Temperature, F l
t I
D.s s
Figure D-3. Surry Unit No.1 Capsule T Surveillance Charpy impact Data Wald Metal SA-1526 (299L44 / 8596)
(Hyperbolic Tangent Curve-Fitting Method) 100 75 50 m
4
\\
f 25 un 0
100 0
100 200 300 400 500 600 I
Temperature, F 100 2E 80 g
=
a-20 0
100 0
100 200 300 400 500 600 Temperature, F 200 Tung :
g 180 Tasuta : +160 F Tse-
+306 F T e'
+147 F 3
160 CvuSE: 52 ft4b RTuov. R 120 100 kJ Ug 80 1
60 Matenal.
Weld Metal 20 Heat Number: 5A 1526 0
100 0
100 200 300 400 500 600 Temperature. F D-6 II^TNRM
Figure D-4. Surry Unit No.1 Capsule T Surveillance Charpy impact Data Correlation Monitor Plate (HSST Plate 02),
Longitudinal Orientation (LT)
(Hyperbolic Tangent Curve-Fitting Method) 100
- 76 50 s
} 25 0
100 0
100 200 300 400 500 600 Temperature. F 100 f
.0 tg 60
[ 40 20 0
-100 0
100 200 300 400 500 600 Temperature, F 200 luot;
- tLE, T,ute : +110 F 180 3
Tso-
+160 F Tao
+118 F 160 CvUSE: 106 ft-lb RT"O m
140 120 g
100 m
80 l
60 Matenal.
HSST Plate 02 40 Orientation ~ Longitudinal Fluence:
2 81 10 n'em' 20 Heat Number: A11951 0
100 0
100 200 300 400 500 600 t
Temperature, F l
I i
D-7 "MNRM
Table D-5. Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data, Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT)
Test Impacti Lateral
. Shear
~
Specime~n Temp.,
- Energy, Expanrion,.
Fracture.
No. '
- FL ft-lb -
mils V50 50 11.0 10.0 3
V52 100 37.0 30.5 13 V49 150 50.0 43.0 41 V53 200 72.0 56.5 66 V54 250 117.0 79.5 100 V55 300 116.0 78.5 100 V51 400 115.0 77.5 100 Table D-6. Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data, Base Metal Plate Heat-Affected-Zone Material
. Test
. impact Lateral Shear Specimen Temp.,
- Energy, Expansion, Fracture.
No. -
'F ft-Ib -
mils.
H10 25 12.0 10.0 12 H9 25 28.0 20.5 31 hl3 50 53.0 31.5 51 H11 100 87.0 59.0 87 H15 150 22.0 20.0 58 H14 200 81.0 61.5 100 H12 300 52.0 39.0 100 H16 400 110.0 70.0 100 l
)
l D-8 II M N M 5 L
l l
Table D-7. Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data, Wald Metal SA-1526, (299L44 / 8596)
$ Test,
. impact
_ ; Lateral Shear Specimen:
Temp.,.
Energy, '
. Expansion,
. Fracture, No.'
- F.
ftJb mils.
W12 50 4.0 3.5 0
W14 150 17.0 19.5 12 l
W13 200 22.0 17.5 28 W16 250 39.0 28.5 73 W10 250 33.0 32.0 52 W15 300 11.0 31.0 96 W11 400 47.0 41 0 100 W9 450 52.0 45.5 100 1
Table D-E. Surry Unit No.1 Capsule V Surveillance Charpy V-Notch Data,
. Correlation Monitor Plate (HSST Plate 02),
Longitudinal Orientation (LT)
Test.
Impact lateral.
Shear Specimen Temp.,
- Energy,
. Expansion,
- Fracture, No.
'F ft-lb
- mils, R41 100 10.0 10.0 4
R43 150 21.0 18.5 10 R47 200 33.0 27.5 33 l
I R48 200 33.0 23.0 26 R46 250 73.0 45.0 43 l
R44 300 92.0 69.0 92 R45 400 101.0 69.5 100 f
R42 450 98.0 72.5 100 l
l 1
D-9 IMTNRM 1
Figure D-5. Surry Unit No.1 Capsule V Surveillance Charpy impact Data Base Metal Plate Heat No. C4415-1, Longitudinal Orientation (LT)
(Hyperbolic Tangent Curve-Fitting Method) 100 0
- - 76 50 25 0
100 0
100 200 300 400 600 600 Temperature, F s '::
e 60 40 f 20 0
100 0
100 200 300 400 500 600 Temperature, F 200 NDY*
b I
T ute. +126 F 180 3
Tso'
+144 F 160 Tao'
- 105 F CvuSE: 116 ft Ib "D'
140 W
l 80 1
e j
60 Matenal.
$A.533 Gr. B Cl 1 40 8
Onentation: Lonottudmal Fluence-194x10 n/cm' 1
20 Heat Number: C44151 1
0
-100 0
100 200 300 400 600 600 Temperature, F
)
D-10 f!MTNRM
Figure D-6. Surry Unit No.1 Capsule V Surveillance Charpy impact Data Base Metal Plate Heat-Affected-Zone Material (Hyperbolic Tangent Curve-Fitting Method) 100
?
O 1,5 7
0 u.
h 25 0
100 0
100 200 300 400 500 600 Temperature, F 100 80 i
g.0 a
<0
....................e..........................
20 0
100 0
100 200 300 400 500 600 Temperature, F 200 Inov;
!! Q.
180 Tasuts : +58F T.-
+89 F T.'
+ 1f, 160 3
CvUSE. 81 ft ib RTuog: R 140 a
g 120 100
~
I 80 n
60 l
..........e..
..............9 Matenal.
SA-533 Gr. B Cl 1 40
..................;............ ;" ;;,' y g,,,,,
Heat Number: C4415-1 0
100 0
100 200 300 400 500 600 l
h Temperature, F I
l D-11
(."h?NR?.\\
l
Figure D-7. Surry Unit No.1 Capsule V Surveillance Charpy impact Data Weld Metal SA-1526 (299L44 / 8596)
(Hyperbolic Tangent Curve-Fitting Method) 100 r
O
- 75 e
1.0 u,
&3 25 m
0 0
-100 0
100 200 300 400 500 600 Temperature, F 100 E6 go 60 g
1
=
c 40
.....s.................
20 d
0 100 0
100 200 300 400 500 600 Temperature, F 200 Two,.
Q T sute ; +309 F 180 3
Tse-
+465 F 160 Tso-
+226 F CvUSE: 49 5 f1-lb RTuot-R 140 120 f $00 W
80 l
60 l
n.......
W 40 g
Materet:
Weld Metal Heat Number; SA.1526 0
-100 0
100 200 300 400 500 600 Temperature, F 1
)
l D-12 IMTNPM 1
l Figure D-8. Surry Unit No.1 Capsule V Surveillance Charpy impact Data Correlation Monitor Plate (HSST Plate 02),
Longitudinal Orientation (LT)
(Hyperbolic Tangent Curve-Fitting Method) 100
~
i i
75 1
I I
50 t
h 25 so 0
100 0
100 200 300 400 500 600 Temperature, F 100 EE 80 60 40 20 0
=
100 0
100 200 300 400 500 600 Temperature, F 200
- Tuot, g
180 In.ns : +218 F Tso'
+220 F 160 T ao'
+188F CvuSE: 100 nib "D' ' W-140 120 h
lE 100 W
O10 l
s 60 Mater d..
HSST Plate 02 40 Orienta*en:
Longitudinal Fluenc.
1 94 x10 n/cma I
20 e
Heat Number, A1195-1 o
l
-100 0
100 200 300 400 500 600 j
Temperature, F l
D-13 R.^.TNRPM
i t
l I
I l
APPENDIX E Fluence Analysis Methodology i
i E-1 f!MT.^, ?R y,5 l
l The two-dimensional discrete ordinates transport code, DORT*'), is used to calculate the fluence exposure of the welds, plates, and at the capsule. The fluence analysis methodology is defined in detail in Section 3.0 of BAW-2241,52) and is conveniently summarized in the following discussion.
Figure E-1 provides a global flow chart of the analytical procedure that was used to determine the incremental fluence accumulated over cycles 1 through 14 Referring to the flow chart (Figure E-1), the analysis is divided into seven main tasks: [1]
generation of the neutron source, [2] development of the DORT geometrical models, [3]
calculation of the macroscopic material cross sections, [4] DORT runs, [5] 3-D synthesis of the results, [6] calculation of best-estimate fluxes, and [7] determination of extrapolation fluxes. Each of these topics is discussed below.
E-1. Generation of the Neutron Source l
The time-averaged pin-by-pin relative power density distribution for cycles 1 through 14 was calculated using the SORREL code.*') The effects of burnup on the spatial distribution of the neutron source were accounted for by calculating the cycle average fission spectrum on an assembly-by-assembly basis and by determining the cycle-average specific neutron emission rate. These data were used with the normalized time-weighted-average pin by-pin relative power density (RPD) distribution to determine the time average space-and energy-dependent neutron source. This neutron source was input to DORT as indicated in Figure E-1.
E-2. Development of Geometrical Models l
The Surry Unit No.1 cavity geometry models for the mid-plane (R-0) and tha vertical plane (R-Z) were developed and input to the DORT code as indicated in Figure E-1.
These models were carefully developed in this analysis and will be used in all subsequent Surry Unit No.1 pressure-temperature (P-T) curve and 10 CFR 50, Appendix H analyses.
ff f< W M E-2
E-3. Calculation of Macroscopic Material Cross Sections in accordance with Draft Regulatory Guide DG-1053,5* the BUGLE-935* cross section 5
library was used. The GIP code
- was used to calculate the macroscopic cross sections for all materials used in the analysis, from the core out through the cavity and into the concrete. The P, scattering approximation was used.
E-4. DORT Analyses Two DORT analyses were necessary: one for the cycle 1 through 12 irradiation and another for the cycle 13 through 14 analysis. The necessity for two calculations is based on two considerations: [1] the capsule was moved at the end of cycle 12 from the 25' location to the 15" location, and [2] the cycle 13 through 14 analysis employed partial length poison rods in two peripheral assemblies.
1 The cross sections, geometry, and appropriate source were combined to create the DORT models. The cycle 1 through 12 analysis required one large DORT run in (R-0) l l
geometry and one in (R-Z) geometry. In the cycle 13 through 14 analysis, several DORT runs were required in (R-e) geometry and several in (R-Z) geometry. The DORT results were used in a multi-planar channel synthesis to determine the three-dimensional flux at all locations ofinterest. Each DORT run utilized a cross section LeDendre expansion of three (P ), a minimum of forty-eight directions (S.), and the appropriate boundary 3
conditions. A theta-weighted flux extrapolation model was used for all runs.
E-5. Determine Synthesized Three-Dimensional Results Cycle 1 Throuah 12 Analysis l
The DORT analyse > produced two sets of two-dimensional flux distributions, one in the vertical plane and one in the horizontal plane. The vertical plane, which will be referred to as the "R-Z analysis" is geometrically defined as the plane bounded axially by the upper and lower core reflectors and radially by the center of the core and a ve.tical line located l
approximately one foot into the water tank. The horizontal plane, referred to as the "R-O
)
analysis" is defined as the plane bounded radially by the center of the core and a point located approximately one foot into the water and azimuthally by the major axis and the I
adjacent 45 radius. The vessel flux, however, varies significantly in all three cylindrical l
E-3 IIMNM
{
l
coordinate directions (R, e, Z). This means that if a point of interest is outside the planes of both the R-Z DORT and the R-e DORT, the true flux cannot be determined from either DORT run. Under the assumption that the true three-dimensional flux is separable, the two two-dimensional data-sets can be mathematically combined to estimate the flux at three-dimensional points (R, e, Z), within the planar boundaries. The synthesis procedure outlined in Draft Regulatory Guide DG-1053 serves as a reference for this task.
Cycle 13 Throuah14 Analysis The cycle 13 through 14 irradiation included partial length poison rods in two assemblies on the periphery. This resulted in an asymmetric neutron source distrit,ution in the periphery and, therefore, an asymmetric flux distribution in the vessel.
The method described above cannot be used to calculate the cycle 13 through 14 flux because that method assumes that the 3-D flux,4(R, e, Z), is separable for a single channel. For asymmetric source distributions such as those that exist in cycles 13 through 14 of Surry Unit No.1, the flux is not a single separable function.
Pin-by-pin X-Y relative source density distributions in the core for 22 planes, over the active core elevation, and for three time steps per cycle, were calculated. This amounts to a discretely expressed, but continuous function cf X, Y, and Z, for six time steps (three in cycle 13 and three in cycle 14).
These source distributions were used in a three-dimensional multi-channel synthesis analysis with piecewise continuous planar functions that provided a means of determining the energy-dependent time-average flux at all points of interest.
E-6. Estimat;on of the Best-Estimate Flux As discussed in the uncertainty analysis, there is no significant bias associated with this analysis beyond that identified in the Cavity Dosimetry Program. Accordingly, the energy-dependent benchmark bias function, Table E-1, was used with the DORT-calculated flux to determine the best-estimate flux at each point of interest in the reactor vessel in accordance with the procedures discussed in the Fluence and Uncertainty Topical Report *"
l E-4 If<W< M
E-7. Extrapolation to the End of Life (EOL)
By necessity, extrapolation of neutron fluence to points in the future is an inexact and approximate process. It is impossible to know with certainty the character of future core
- operations or to accurately estimate the effect of any given core operation on the fluence at any given location, before the fact. It is possible, however, to make reasonable estimates of the inside surface maximum flux using near-future fuel cycle design trends.
The " extrapolation flux"is defined as the constant flux used to determine the fluence at points in the future, in the FTl / B&WOG methodology, extrapolation flux is always based j
on the DORT-calculated flux determined in the just-completed fluence analysis, it is the stated intention of Virginia Power to continue operation into the indefinite future with loadings similar to those used in cycle 13 through 14.
It is emphasized that with proper monitoring, the magnitude of the fluence at EOL and the uncertainty in the EOL fluence will be maintained such that the material properties will not exceed their lawful limits.
E-8. Uncertainty Evaluation The Surry Unit No.1 fluence predictions were calculated using the methodology described in the " Fluence and Uncertainty Methodologies" topical report, BAW-2241P.
The time-averaged fluxes, and thereby the fNences throughout the reactor, are calculated with the DORT discrete ordinates computer code using three-dimensional synthesis. Methods for performing synthesis analysis are described in Section 3.0 of the topical The DORT three-dimensional synthesis results are the basis for the fluence predictions using the FTl " Semi-Analytical"(calculational) methodology. As noted in Sections 6.0 and 7.0 of the topical report, the best-estimate fluence predictions are determined by removing all biases from the calculated fluence results. The bias removal function is dependent on the DORT solution procedures, the BUGLE-93 cross sections, and the FTl dosimetry benchmarks, it is independent of the Surry Unit No.1 fluence predictions and of any and all plant-specific comparisons of dosimetry
(
calculations to measurements. The uncertainties (U), determined in the topical report (BAW-2241P), are as follows:
l l
E-5 M E^ N
U(capsule) s 7.00%
U(vessel, cycles 1 - 14) s 10.02%
. indications from the uncertainties in the dosimetry measurements from Surry Unit No.1, l
L
. Capsule X, and the comparisons of calculations to the measurements are that the Surry Unit No.1 results are consistent with the FTl benchmark database of uncertainties.
Therefore, the FTl Semi-Analytical methodology uncertainties are applicable to the Surry Unit No.1 fluence calculations.
- E-9. References E-1.
M. A. Rutherford, N. M. Hassan, et al., "DORT, Two DimensionalDiscrete Ordinates Transport Code", BWNT-TM-107, Framatome Technologies, Inc.,
Lynchburg, Virginia, May 1995.
E-2.
J. R. Worsham, et. al.,
- Fluence and Uncertainty Methodologies", BAW-2241P.
Framatome Technologies, Inc., Lynchburg, Virginia, April,1997.
E-3.
L. A. Hassler and N. M. Hassan, " SORREL, DOTInput Generation Code User's Manual," NPGD-TM-427. Revision 8, July 1992.
E-4.
U.S. Nuclear Regulatory Commission Draft Regulatory Guide DG-1053,
" Calculational ana Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", June 1996.
E-5.
D. T. Ingersoll et. al., " BUGLE-93, Production and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron / photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data", ORNL-DLC-175, Oak Ridge National Laboratories, Oak Ridge, Tennessee, April 1994.
E-6.
L. A. Hassler and N. M. Hassan, " GIP Users ManualforB&W Version, Group Organized Cross Section Input Program,"NPGD-TM 456. Revision 11. August 1994.
E-7.
C. Garat, et al., "B&WOG Cavity Dosimetry Benchmark Experiment Summery Reporf," BAW-2205, Framatome Technologies, Inc., Lynchburg, Virginia, January 1995.
E-6 I?<MNRM
Table E-1. Bias Removal Function (E > 1 MeV)*#3 Energy 1
- y. Upper Energy,;-
Group ~
MeV~
' hg 1
17.33 1.020 2
14.19 1.022 3
12.21 1.022 4
10.0 1.021 5
8 607 1.020 6
7.108 1.020 7
6.065 1.022 8
4.966 1.019 9
3.679 1.014 10 3.012 1.004 l
l 11 2.725 0.988 12 2.466 0.982 13 2.365 0.979 14 2.346 0.977 15 2.231 0.970 16 1.921 0.960 17 1.653 0.952 18 1.353 0.913 19 1.003 I
I L
l L
E-7 "MMWA
Figure E-1. Fluence Analysis Methodology Assembly x assembly RPD Materials of fission spectrum pin x pin l Reactor Geometry _j construction by fissile isotope distribution history II BUGLE-93 If If l DORT models I cross section l SORREL code l lbrary Time-averaged lf If 37 axial sc ;e (GlP codel Time-averaged neutron source 3r Dosimetry So(E,R,0) l Cross sections l counting 3r y
l DORT Analysis 4 m
analysis Re and RZ l
(NESl)
Data to calculate Power history lf lf absolute (saturation)
Synthesized magnitude results y
3-D results Measured dosimetry V
V activities Calculated dosimeter activities
>l C/M l4 B&WOG y
Y benchmark Statistical l Calculational bias l analysis analysis bias V
Confirm V
plant-specific l Vesselfluence l uncertainties E-8
- IfMNAM
l l
l 1
APPENDIX F Capsule Dosimetry Measurements l
l F-1
(.nygyo y,9
F.1 Introduction Three dosimeter sets were located in blocks that were installed in top, middle, and bottom positions of the capsule assembly. Each dosimeter set consisted of dosimeters made up of shielded and unshielded Co/Al wires, and Cu or Ni wires. One dosimeter set included l
shielded 23sU and 237Np fission powders; these were located in the middle of the capsule.
The dosimeters were stored in via!s identified by labels consisting of the position of the dosimeter holder block within the capsule assembly and the location from where the dosimeters where recovered. Upon removal of the dosimeters from their respective holder blocks, discrepancies were found between the capsule engineering drawing and the dosimeters removed from the capsule holder blocks. These discrepancies are presented below:
Capsule X:
Dosimeters Specified Dosimeters Removed l
Location -
in Capsule Drawing From Capsule-Top Cu wire, shielded Co/Alwire, Cu wire, shielded Co/Alwire, unshielded Co/Alwire unshielded Co/Alwire Middle Niwire, chielded Co/Alwire, Shielded Co/Alwire, unshielded Co/Alwire 2-unshielded Co/Alwires Bottom Cu wire, shielded Co/Alwire, 2-Cu wires, shielded Co/Alwire, unshielded Co/Alwire unshielded Co/Alwire Middle zaaU and 237Np 238U and 237Np (238U and 237Np) g Iron dosimeters were obtained by removing two small samples from Charpy specimens that were located in the top, middle, and bottom positions of the capsule assembly.
These samples were then dissolved and analyzed for specific activities.
F.2 Dosimeter Preparation Vials were prepared for the dosimeters by labeling them with identifications that indicated their types and positions in the holder blocks. For example, the one top block shielded Co/Al dosimeter was labeled Surry T (or TOP) Sh Co/Al. The analyte nuclides were verified during gamma scanning.
F-2 IMMM
The fission powder capsules were clamped in a vise which was mounted on two lead bricks in a hood. A flat mill-bastard file was used to file the capsules open. The fission powder was carefully collected in vials with appropriate labels.
l l
The dosimeter wires were washed in reagent acetone and blotted dry with a laboratory towel. Each dosimeter wire was measured with a certified micrometer caliper and weighed on a certified analytical balance. Each wire was then mounted in the center of a PetriSlide with double-sided tape.
The exact oxide composition of the uranium dosimeters was uncertain. It was not possible to correct for self-absorption of the powders, therefore it was necessary to dissolve them and put them into a geometry for which the gamma spectrometer was calibrated. This was the 20cc liquid scintillation vial geometry. The uranium dosimeters were dissolved in 8N HNO acid and diluted to 20 mlin the same acid in a pre-weighed 3
20cc scintillation vhl The total uranium content was measured by inductively coupled plasma spectroscopy (ICP). The neptunium powder was also prepared using similar procedures, but no ICP analysis was performed.
F.3 Quantitative Gamma Spectrometry Each of the dosimeters, in the PetriSlide" (point source), or 20cc vial geometry, was given a 300 second preliminary count on the 31% PGT gamma spectrometer. This provided information to best judge the distance at which to count the dosimeter to obtain a minimum of 10,000 counts in the photopeak of interest while keeping the counter dead time below 15%. It also provided qualitative identification of the dosimeters. This identification was made from the presence of the gamma rays in Table F-1. The spectra confirmed the identities of the dosimeters.
The spectra were then measured quantitatively at the appropriate counting positions and j
for the appropriate count times determined from the preliminary counts.
l l
1 F-3 f."MNRM 1
l
f F.4 Dosimeter Specific Activities The dosimeter specific activities are shown in Table F-2, and the associated elemental weight fractions of the docimeters and the isotopic fractions of the target nuclides are listed in Table F-3.
The weight fraction listed in Table F-4 through F-9 is the product of the isotopic fraction of the target and the weight fraction of the element in the dosimeter. In the case of the 238U dosimeters, the total uranium content was measured by inductively coupled plasma (lCP) atomic emission spectroscopy and that value and the '87Cs activity were used to 237 calculate the specific activity. In the case of the Np dosimeter, the decay product 23'Pa was used to estimate the neptunium activity. It was assumed that there was no 237 alternative source of protactinium daughter, and that Np was secular equilibrium with the 23'Pa so that the Np activity was equal to the 233Pa activity. The neptunium mass 2
was equal to the protactinium activity divided by the specific activity of neptunium,705 pCi/ gram.
The dosimeter specific activities were calculated by dividing the corrected activity of the analyte nuclide by the target nuclide mass. The results were shown in Table F-3, and the detailed calculations can be found in Tables F-4 through F-9.
1 I
l l
l 1
F-4 IMMRM
Table F-1. Quantifying Gamma Rays Dosimder'
- Analyte :
Iron MMn @ 834 keVfrom MFe Co/Al-
- Co @ 1332 kev from 63Cu, very low activity compared to Co wires, wire has coppery cc.'or 237Np 137Cs @ 662 kev 238U 137Cs @ 662 kev l
i l
I F-5 IMTNAM L
Table F-2. Isotopic Fractions and Weight Fractions of Target Nuclides I
Isotopic Weight l
Target Fraction of Fraction of Dosimeter Nuclide Target Target Element Iron 5dFe 0.0585 0.96792 Cobalt 59Co 1.0 0.0015 (unshielded)
Cobalt 59Co 1.0 ICP*
(shielded)
Nickel 58Ni 0.6777 1.0 Copper 63Cu 0.6917 0.999 Neptunium-237 237Np 1.0 1.0 Uranium-238 238U 1.0 ICP*
- Inductistely Coupled Plasma Atomic Emission Spectroscopy.
l l
fmmz.m
Table F-3. Specific Activities for Surry Unit 1 Capsule X Dosimetry Dosimeter.
Specife.
l
- Shielded 1 Targeti Analyte --
Activity
' identification
, (Yes/No)
Nuclide:'
Nuclide?
( CilgmTarget) l SU1, Top Sh Co/Al Yes Co-59 Co-60 3.189E+04 SU1, Top Co/Al No Co-59 Co40 3.999E+05 SU1, Middle Sh Co/Al Yes Co-59 Co-60 3.041E+03" SU1, Middle Co/Al No Co-59 Co40 3.250E+05 SU1, Bottom Sh Co/Al Yes Co-59 Co-60 1.176E+04 SU1, Bottom Co/Al No Co-59 Co-60 2.912E+05 SU1, Top Cu No Cu43 Co-60 6.958 SU1, Bottom CuA No Cu43 Co40 6.674 SU1, Bottom cub No Cu-63 Co40 6.853 SU1, Middle Ni' No Co-59 Co40 3.869E+05 SU1, Top FeA(CVN H24)
No Fe-54 Mn-54 797 Sul, Top FeB(CVN H24)
No Fe-54 Mn-54 747 SU1, Middle FeA(CVN H18)
No Fe-54 Mn-54 756 SU1, Middle FeB (CVN H18)
No Fe-54 Mn-54 874 SU1, Bottom FeA(CVN V58)
No Fe-54 Mn-54 717 SU1, Bonom FeB(CVN V58)
No Fe-54 Mn-54 653 SU1, Sh U-238 Yes U-238 Cs-137 16.16 SU1, Sh Np-237 Yes Np-237 Cs-137 117.7 See Section F-1 for details.
" This value is an order of magnitude lower than the other dosimeters. Dosimeter wire is suspected to be copper wire that has Co-60 as analyte nuclide but at lower activity.
[
F-7
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EVALUATION OF SURRY UNIT 1 SURVEILLANCE CAPSULE X RESULTS AND RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI)
ON GENERIC LETTER 92-01, REVISION 1, SUPPLEMENT 1
1 l
l EVALUATION OF SURRY UNIT 1 SURVEILLANCE CAPSULE X RESULTS AND RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAl)
I ON GENERIC LETTER 92-01 REVISION 1 SUPPLEMENT 1
Background
During the Surry Unit 1 Cycle 14/15 refueling outage, Capsule X was withdrawn from its holder at the 165 azimuthal location, and was transferred to the Surry spent fuel pool.
i Capsule X was subsequently transported to Framatome Technologies' laboratory in Lynchburg, Virginia, where it was disassembled, inventoried, and tested in accordance with applicable regulations and standards.
The resuits of the surveillance capsule analysis are documented in Reference (1), which is presented in Attachment 1 to this letter.
An evaluation of the Surry Unit 1 Capsule X analysis results has been performed to verify the validity and conservatism of the existing Surry Units 1 and 2 Reactor Coolant System (RCS) pressure / temperature (P/T) limits and Low Temperature Overpressure Protection System (LTOPS) setpoints documented in the Surry Technical Specifications.
The existing P/T limits and LTOPS setpoints were transmitted to the NRC by letter dated June 8, 1995 (2) and were approved by letter dated December 28, 1995 (3). Revised Pressurized Thermal Shock (PTS) Reference Temperatures (i.e., RT rs) have also been e
performed herein to demonstrate that all Surry Units 1 and 2 reactor vessel beltline rnaterials remain below the screening criteria presented in 10 CFR 50.61. The evaluation documented herein uses revised reactor vessel fluence values calculated in accordance with the Virginia Power Reactor Vessel Fluence Methodology Topical Report (4), which is presented in Attachment 3 to this letter.
Evaluation of Existing P/T Limits and LTOPS Setpoints Appendix A presents Reactor Vessel Material Data Tables for Surry Units 1 and 2 which include consideration of the Surry Unit 1 Capsule X analysis results (1). The information in the Appendix A data tables has been organized in a manner consistent with the NRC's draft Request for Additional Information on Generic Letter 92-01, Revision 1,
Supplement 1 (5). The calculations which support the information presented in these tables were performed in accordance with Regulatory Guide 1.99, Revision 2 (6), the Pressurized Thermal Shock Rule in 10 CFR 50.61 (7), and the regulatory guidance provided in the meeting minutes from the November 12,1997 NRC/ Industry meeting on reactor vessel integrity (8). As such, these tables constitute the Virginia Power response to the NRC Request for Additional Information (RAl) on Generic Letter 92-01, Revision 1, Supplement 1 documented in Reference (10). The Surry Units 1 and 2 reactor vessel beltline materials are also included in the forthcoming B&W Owners Group (BWOG)
Reactor Vessel Working Group (RVWG) generic response to the NRC request for Page 1 of 5
I t
i additional information (RAl) on Generic Letter 92-01 Revision 1, Supplement 1 (9) The tables presented in Appendix A compare favorably to those presented in the RVWG response (9).
The existing Surry Units 1 and 2 P/T limits and LTOPS setpoints (2)(3) are based on a limiting %-thickness (% -T) RTuor of 228.4 F. This value of RTuor was determined to I
bound all Surry Units 1 and 2 reactor vessel beltline materials at end-of-license (EOL)
I fluences corresponding to 28.8 EFPY and 29.4 EFPY for Surry Units 1 and 2, respectively (2)(3). After consideration of the Surry Unit 1 Capsule X surveillance data for beltline materials SA-1526 (lower shell longitudinal weld fabricated from weld wire heat 299L44) and for plate 4415-1 (lower shell), the most limiting % -T RTwe-value for Surry Units 1 and 2 is 215.7 F. This value of RTuor was determined on the basis of.'luence values corresponding to 29.6 EFPY and 30.1 EFPY for Surry Units 1 and 2, respectively (4). On the basis of the results presented in Appendix A, it is concluded that the existing RCS P/T limits and LTOPS setpoints (2)(3) remain valid and conservative.
Evaluation of PTS Screening Calculations PTS screening calculations were previously transmitted to the NRC by letter dated April 1, 1996 (11). All Surry Units 1 and 2 reactor vessel beltline materials demonstrated margin to the 10 CFR 50.61 screening criteria at end-of-license (EOL) fluences corresponding to 28.8 EFPY and 29.4 EFPY for Surry Units 1 and 2, respectively (11). On the basis of the results presented in Appendix A, it is concluded that all Surry Units 1 and 2 beltline materials continue to meet the 10 CFR 50.61 screening criteria at end-of-license fluence values corresponding to 29.6 EFPY and 30.1 EFPY for Surry Units 1 and 2, respectively (4).
Revised Reactor Vessel Beltline Neutron Fluence Values The revised reactor vessel beltline neutron fluence values were calculated using the Virginia Power Reactor Vessel Fluence Methodology Topical Report (4).
The methodology described in that report was developed in accordance with Draft Regulatory Guide DG 1053 (12). The vessel fluence calculational methodology was benchmarked using a combination of Virginia Power surveillance capsules, pressure vessel simulator measurements, and Surry Unit 1 ex-vessel cavity dosimetry measurements. Capsule X was not used to validate the calculational methodology because the capsule data were not available during the methodology development.
The table below presents the calculated neutron fluences.
Page 2 of 5 L
1 2
Location Fluence (n/cm )
Surry 1 Surry 2 Peak Vessel Fluence, End of License 3.53E19 3.52E19 Peak Vessel Fluence, End of Life Extension 5.40E19 5.34E19 Peak Beltline Weld Fluence, End of License 3.20E19 3.52E19 Peak Beltline Weld Fluence, End of Life 4.70E19 5.34E19 Extension Peak Upper Longitudinal Weld Fluence, End 6.00E18 6.97E18 of License Peak Upper Longitudinal Weld Fluence, End 9.14E18 1.08E19 of Life Extension Peak Lower Longitudinal Weld Fluence, End 5.40E18 6.97E18 of License Peak Lower Longitudinal Weld Fluence, End 7.90E18 1.08E19 of Life Extension Peak Nozzle to Intermediate Shell Circ.
3.07E18 2.98E18 Weld Fluence, End of License Peak Nozzle to Intermediate Shell Circ.
4.96E18 4.71E18 Weld Fluence, End of Life Extension Revised Reactor Vessel Beltline Fluence Values Surry Units 1 and 2 Conclusions After consideration of the additional material properties data from Surry Unit 1 Capsule X for beltline weld material SA-1526 (lower shell longitudinal weld fabricated from weld wire heat 299L44) and plate 4415-1 (lower shell), it is concluded that the existing Surry Units 1 and 2 P/T limits and LTOPS setpoints remain valid for cumulative core burnups up to 28.8 EFPY and 29.4 EFPY for Surry Units 1 and 2, respectively. It is further concluded that all Surry Units 1 and 2 beltline materia!s continue to meet the 10 CFR 50.61 screening criteria for cumulative core burnups up to 28.8 EFPY and 29.4 EFPY for Surry Units 1 and 2, respectively, The evaluations that support these conclusions utilize fluence values l.
determined in accordance with the Reference (4) topical report.
In addition, the evaluations which support these conclusions were performed in accordance with Regulatory Guide 1.99, Revision 2 (6),10 CFR 50.61 (7), and the November 12,1997 NRC/ Industry meeting notes (8). The tables presented in Appendix A constitute the Virginia Power response to the NRC RAI documented in Reference (10).
Page 3 of 5 N
a
F-b References L
'(1)
" Analysis of Capsule X, Virginia Power Surry Unit No.1, Reactor Vessel Material Surveillance Program," BAW-2324, dated April,1998.
(2)
Letter from R. F. Saunders to USNRC, " Virginia Electric and Power Company, Surry Power Station Units -1 and 2, Request for Exemption - ASME Code Case N-514, Proposed Technical Specifications
- Change, Revised p
Pressure / Temperature Limits and LTOPS Setpoints," Serial No.95-197, June 8, 1995.
(3)-
_ Letter from B. C. Buckley to J. P. O'Hanlon, "Surry Units 1 and 2 - Issuance of 8
Amendments Re: Surry Units 1 and 2 Reactor Vessel Heatup and Cooldown Curves (TAC Nos. M92537 and M92538)," Serial No.96-020, dated December 28,1995.
(4)
Virginia Power Topical Report VEP-NAF-3, " Reactor Vessel Fluence Analysis Methodology," dated November,1997.
(5)
Letter from G. C. Lainas (USNRC) to D. J. Modeen (NEI), " Request for Additional information Regarding Reactor Pressure Vessel Integrity," dated February 5,
'1998.,
(6)
Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," dated May,1988.
(7)
Title 10, Code of Federal Regulations, Part 50.61,
" Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."
(8)
Memorandum from K. R. Wichman to E. J. Sullivan, " Meeting Summary for November 12, 1997 Meeting with Owners Group Representatives and NEl Regarding Review of Responses to Generic Letter 92-01, Revision 1, Supplement 1 Responses," dated November 19,1997.
(9)
" Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," BAW-2325, dated May,1998.-
(10)
Letter' from G.' E Edison (USNRC) to J. P. O'Hanlon (Virginia Power), "Surry Nuclear Power Station Units 1 and 2, Request for Additional Information Related -
to Reactor Vessel Structural Integrity (Generic Letter 92-01, ' Revisio: 1 Supplement 1) (TAC Nos. MA0576 and MA0577)," dated April 20,1998.
.(11) L Letter from J. P. O'Hanlon to USNRC, " Virginia Electric and Power Company, Surry and North Anna Power Stations Units 1 and 2, Pressurized Thermal Shock (PTS) Screening Calculations'" Serial No.96-084, dated April 1,1996.
Page 4 of 5
\\
I (12)
Draft Regulatory Guide DG-1053," Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated June 1996.
l I
Page 5 of 5 t
Appendix A REACTOR VESSEL MATERIAL DATA TABLES l
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