ML20205A505
| ML20205A505 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 07/31/1986 |
| From: | Cane T, Ranganath S, Robare D GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML112240869 | List: |
| References | |
| DRF-B13-01320, DRF-B13-1320, NEDC-31166, NEDC-31166-R01, NEDC-31166-R1, NUDOCS 8608110386 | |
| Download: ML20205A505 (121) | |
Text
I EEE8$1" DRF B13-0 JULY 1986 l
I DUANE ARNOLD ENERGY CENTER REACTOR PRESSURE VESSEL SURVEILLANCE MATERIALS TESTING T. A. CAINE hbh kDO I
GENER AL h ELECTRIC
NEDC-31166 DRF B13-01320 Class 1 July 1986 Revision 1 DUANE ARNOLD ENERGY CENTER REACTOR PRESSURE VESSEL SURVEILLANCE MATERIALS TESTING T. A. Caine Approved:
Y. Ranganath, Manager S
Structural Analysis Services
)ll5 b Approved:
D. J. Robare, Manager Licensing Services NUCLEAR ENERGY BUSINESS OPERATIONS
- GENERAL ELECTRIC COMPANY GAN JOSE. CAUFORNIA 95125 GENERAL $ ELECTRIC
NEDC-31166 R.:,v. 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of Iowa Electric Light and Power Company.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the contract. governing Iowa Electric Light and Power Company Purchase Order No. S24610 and nothing contained in this document shall be construed as changing said contract.
The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any duch unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
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NEDC-31166 Rev. 1 CONTENTS Page ABSTRACT ix ACKNOWLEDGEMENTS ix 1.
INTRODUCTION 1-1 2.
SUNkMRY AND CONCLUSIONS 2-1 2.1 Summary of Results 2-1 2.2 Conclusions 2-4 3.
SURVEILLANCE PROGRAM BACKGROUND 3-1 3.1 Capsule Recovery 3-1 3.2 RPV Materials and Fabrication Background 3-1 3.2.1 Fabrication History 3-1 3.2.2 Material Properties of RPV at Fabrication 3-2 3.2.3 Specimen Chemical Composition 3-2 3.3 Specimen Description 3-3 3.3.1 Charpy Specimens 3-3 3.3.2 Tensile Specimens 3-4 3.3.3 Unirradiated Archive Specimens 3-5 l
4.
FLUX WIRE TEST EVALUATION 4-1 4.1 Flux Wire Analysis 4-1 4.1.1 Procedure 4-1 4.1.2 Results 4-2 4.2 Flux Distribution Lead Factors 4-3 4.3 Estimate of End-of-Life Fluence 4-4 5.
CHARPY V-NOTCH IMPACT TESTING 5-1 5.1 Procedure 5-1 5.2 Results 5-2 5.3 Irradiated versus Unitradiated 5-3 Charpy V-Notch Properties 6.
TENSILE TESTING 6-1 t
6.1 Procedure 6-1 6.2 Results 6-2 6.3 Irradiated Versus Unirradiated Tensile Properties 6-3 7.
DETERMINATION OF OPERATING LIMITS AND END-OF-LIFE CONDITIONS 7-1 l
7.1 Evaluation of Radiation Effects 7-1 7.1.1 Measured Versus Predicted Surveillance Shif t 7-2 7.1.2 Modification of the Shift Relationship 7-2 7.1.3 Lim'iting Beltline Material 7-3 7.1.4 Operating Limits Curves 7-3 7.2 Predicted End-of-Life Conditions 7-4 7.2.1 Adjusted Reference Temperature 7-4 l
7.2.2 Upper Shelf Energy 7-4 l
8.
REFERENCES 8-1 111
i NEDC-31166 R:v. 1 CONTENTS (continued)
APPENDICES A.
IRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS A-1 B.
UNIRRADIATED CHARPY SPECIMEN FRACTURE SUAFACE PHOTOGRAPHS B-1 C.
REVISIONS TO THE TECHNICAL SPECIFICATIONS C-1 D.
REVISIONS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT D-1 iv
NEDC-31166 R:v. 1 TABLES Table Title P'WdL 3-1 Material Properties of Vessel Components 3-6 3-2 Plasma Emission Spectrometry Chemical Analysis of 3-7 RPV Surveillance Plate and Weld Materials 3-3 Identification of Charpy and Tensile Specimens 3-8 Removed from Surveillance Capsule 3-4 Identification of Charpy and Tensile Archive Specimens 3-9 4-1 Summary of Daily Power History 4-5 4-2 Surveillance Capsule Location Flux and Fluence for 4-6 Irradiation from 5/13/74 to 2/2/85 5-1 Qualification Test Results Using U.S. Army 5-4 Watertown Specimens (Tested in February 1986) 5-2 Charpy V-Notch Impact Test Results for 5-5 Irradiated RPV Materials 5-3 Charpy V-Notch Impact Test Results for 5-6 Unirradiated RPV Materials i
5-4 Comparison of Unirradiated and Irradiated 5-7 Charpy V-Notch Data l
6-1 Tensile Test Results for RPV Materials 6-4 6-2 Comparison of Unitradiated and Irradiated 6-5 l
Tensile Properties 7-1 EOL Adjusted Reference Temperatures for 7-6 Beltline Materials 7-2 End-of-Life Upper Shelf Energy 7-7 v/vi r
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F NEDC-31166 R:v. 1 ILLUSTRATIONS Figure Title Page 3-1 Surveillance Capsule Recovered from DAEC Reactor 3-10 3-2 Schematic of the RPV Showing Arrangement of Vessel 3-11 Plates and Welds 3-3 Fabrication Method for Base Metal Charpy Specimens 3-12 3-4 Fabrication Method for Weld Metal Charpy Specimens 3-13 3-5 Fabrication Method for HAZ Charpy Specimens 3-14 3-6 Fabrication Method for Base Metal Tensile Specimens 3-15 3-7 Fabrication Method for Weld Metal Tensile Specimens 3-16 3-8 Fabrication Method for HAZ Tensile Specimens 3-17 5-1 DAEC Base Metal Impact Energy 5-8 5-2 DAEC Weld Metal Impact Energy 5-9 5-3 DAEC HAZ Metal Impact Energy 5-10 5-4 DAEC Base Metal Lateral Expansion 5-11 5-5 DAEC Weld Metal Lateral Expansion 5-12 5-6 DAEC HAZ Metal Lateral Expansion 5-13 6-1 Typical Engineering Stress versus Percent Strain 6-6 for Irradiated RPV Materials 6-2 Tensile Specimen Yield Strength Versus Test 6-7 Temperature 6-3 Tensile Specimen Ultimate Tensile Strength 6-8 Versus Test Temperature 6-4 Tensile Specimen Uniform Elongation 6-9 Versus Test Temperature 6-5 Tensile Specimen Total Elongation 6-10 Versus Test Temperature 6-6 Tensile Specimen Reduction of Area 6-11 Versus Test Temperature vii
NEDC-31166 R:v. I 6-7 Fracture Location, Necking Behavior, and Fracture 6-12 1
Appearance for Irradiated Base Metal Tensile Specimens 6-8 Fracture Location, Necking Behavior, and Fracture 6-13 Appearance for Irradiated Weld Metal Tensile Specimens j
6-9 Fracture Location, Necking Behavior, and Fracture 6-14 Appearance for Irradiated HAZ Metal Tensile Specimens 6-10 Fracture Location, Neckitig Behavior, and Fracture 6-15 Appearance for Unirradiated Base Metal Tensile Specimens 6-11 Fracture Location, Necking Behavior, and Fracture 6-16 j
Appearance for Unirradiated Weld Metal Tensile Specimens 7-1 Shift in RT Versus Years of Operation, According 7-8 NDT to Regulatory Guide 1.99 7-2 Pressure Versus Minimum Temperature, Valid to 12 EFPY 7-9 1
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NEDC-31166 Rsv. 1 ABSTRACT A surveillance capsule was removed from the Duane Arnold Energy Center reactor at the end of Fuel Cycle 7, after accumulating 5.9 ef fective full power years of operation.
The capsule contained flux wires for neutron fluence measurement and Charpy and tensile test specimens for material property evaluation.
A combination of flux wire testing and computer analysis was used to establish the vessel peak flux ragnitude and end-of-life predicted fluence. Charpy V-Notch impact testing and uniaxial tensile testing were performed to establish the material properties of the irradiated vessel beltline.
Unitradiated archive specimens were provided by Iowa Electric Light and Power to establish baseline Charpy and tensile data.
The base n.etal Charpy specimen testing results show a larger irradiation shift than predicted, while the weld metal showed no measurable shift.
The predicted end-of-l'ife conditions of reference temperature and upper shelf energy were calculated to be less severe than the limits requiring vessel thermal annealing.
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ACKNOWLEDGMENTS Flux wire testing was performed by G.
C.
Martin.
The Charpy l
V-Notch testing was done by J.
L.
Bennett.
S. B. Wisner and l
G. H. Henderson tested the tensile specimens.
C.
R.
Judd performed the chemical composition testing.
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NEDC-31166 R3v. I 1.
INTRODUCTION Part of the effort to assure reactor vessel integrity involves evaluation of the fracture toughness of the vessel ferritic materials.
The key values which characterize a material's fracture toughness are the reference temperature of nil-ductility transition (RTg ) and the upper shelf energy (USE).
These are defined in 10CFR50 Appendix G (Reference 1) and in Appendix G of the ASME Boiler and Pressure Vessel Code,Section III (Reference 2).
These documents contain requirements that establish the pressure-temperature operating limits which must be met to avoid brittle fracture.
Appendix H
of 10CFR50 (Reference 3) and ASTM E185-82 (Reference 4) establish the methods to be used for surveillance of the reactor vessel materials.
In April, 1985 one of the surveillance specimen capsules required by Reference 3 was removed from the Duane Arnold Energy Center (DAEC) reactor 'after 7
fuel cycles of irradiation, or 5.9 effective full power years (EFPY) of operation.
The surveillance capsule contained flux wires for neutron flux monitoring and Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated from the vessel plate and weld materials nearest the core (the beltline).
The impact and tensile specimens were tested to establish material properties for the irradiated vessel materials.
The results of surveillance specimen testing are presented in this report.
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Unirradiated archive Charpy and tensile specimens of identical fabrication and material composition as those in the surveillance capsule were provided by Iowa Electric Light and Power (IEL&P) for testing to establish the unirradiated properties of the surveillance materials.
The archive specimens were tested on the same equipment, by the same operators, as were the irradiated specimens. The results of unirradiated specimen testing are presented and comparisons to surveillance specimen test results are made, i
1-1
NEDC-31166 R2v. 1 Predictions of the RT and USI at end of reactor life (EOL) are made ET using Regulatory Guide 1.99, Revision 1 (Reference 5), incorporating results from the surveillance tests.
The predictions are compared with allowable values in Reference 1.
Operating limits curves are examined in light of the results of the surveillance testing.
Technical Specification revisions are suggested, as are changes to the Updated Final Safety Analysis Report (UFSAR).
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NEDC-31166 Rsv. 1 2.
SUMMARY
AND CONCLUSIONS 2.1
SUMMARY
OF RESlILTS Surveillance capsule 1 was removed from the DAEC reactor at the l
end of Fuel Cycle 7 and shipped to the General Electric Vallecitos Nuclear Center.
In addition, unirradiated archive Charpy and tensile specimens were sent by IEL&P for testing to establish baseline properties. The flux wires, Charpy V-Notch and tensile test specimens were tested according to ASTM E185-82 (Reference 4).
The methods and results of the fracture toughness testing are presented in this report as follows:
i a.
Section 3:
Surveillance Program Background b.
Section 4:
Flux Wire Test Evaluation c.
Section 5:
Charpy V-Notch Testing i
d.
Section 6:
Tensile Testing e.
Section 7:
Determination of Operating Limits and End-of-Life Conditions Photographs of Charpy specimen fracture surfaces are in Appendices A and B for irradiated and unirradiated specimens, respectively.
Suggested revisions to the Technical Specification are found in Appendix C, and UFSAR changes are recommended in Appendix D.
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NEDC-31166 Rsv. 1 The significant results of the evaluation are summarized as follows:
a.
Capsule 1 was removed from the 288* azimuth position of the reactor.
The capsule contained 6 flux wires:
2 each of pure copper (Cu), nickel (Ni) and iron (Fe).
There were 24 Charpy V-Notch specimens: 8 each of plate material, weld material and heat affected zone (HAZ) material.
The 6 tensile specimens removed consisted of 2 plate, 2 weld and 2 HAZ metal specimens.
All specimen materials were positively identified as being from the vessel beltline, b.
The chemical compositions of the beltline materials were identified through a combination of literature research and testing.
The copper (Cu), phosphorus (P) and nickel (Ni) contents were determined for all heats of plate asterial and for the seam welds in the beltl'ine.
The chemistry for the limiting beltline plate is 0.13% Cu, 0.012% P and 0.47% Ni, i
For the limiting beltline weld, the chemistry is 0.03% Cu, O 017% P and 1.02% Ni.
c.
The flux wires were tested to determine the neutron flux at the surveillance capsule location.
The fast flux 9
(>1.0 MeV) measured was 2.6x10 n/cm -s.
Based on the flux wire data, the surveillance specimens had received a I
2 nominal fluence of 4.9x10 n/cm at removal.
d.
Lead factors for the DAEC vessel geometry and core power distribution were computed in Reference 6.
The lead factors for the surveillance capsule are 0.78 to the peak vessel surface location and 0.99 to the peak 1/4 T depth location.
The EOL fluence, based on 32 EFPY, was calculated with the 1/4 T lead factor and the flux wire results, including a 4%
power uprate at 6 EFPY.
The resulting EOL fluence is 2
3.6x10 n/cm,
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NEDC-31166 Rsv. 1 e.
The irradiated and unirradiated surveillance Charpy V-Notch l
specimens were impact tested at temperatures selected to define the transition of the fracture toughness curves of the plate, weld, and HAZ materials. Measurements were taken of absorbed energy, lateral expansion and percent shear.
Fracture surface photographs of each specimen are presented in Appendix A for irradiated specimens and in Appendix B for unirradiated specimens.
From absorbed energy and lateral expansion results for the plate, weld and HAZ materials, the following values were extracted:
index temperatures for 30 ft-lb, 50 ft-lb, and 35-mil lateral expansion (MLE) values, and USE.
f.
The unirradiated Charpy r.pecimen curves were compared to the irradiated specimen curves to establish the RT shift due g
to irradiation. The base metal test results indicate a 42*F shift, which is greater than the 32*F shift predicted by Reference 5.
The weld metal test results indicate no measurable shift.
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The irradiated tensile specimens were tested at room temperature (76*F) and reactor operating temperature (550*F). Unirradiated tensile specimens were tested at room l
temperature, reactor operating temperature and at the temperature corresponding to onset of upper shelf.
The results tabulated for the unirradiated' and irradiated specimens include yield and ultimate tensile strength (UTS),
unif orm and total elongation, and reduction of area.
The plate, weld, and HAZ specimens behaved similarly, indicating decreasing properties with increasing temperature.
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The irradiated material tensile test results are compared to unirradiated data from the reactor vessel fabrication test program.
The results indicate yield strength and UTS have increased with irradiation, and the uniform and total elongation have decreased with irradiation.
These trends are characteristic of the effects of irradiation.
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NEDC-21166 Rsv. 1 1.
The adjusted reference temperature for each beltline plate and weld was calculated for EOL conditions.
A factor of 42/32 was used to increase the predicted shifts of the beltline plates, based on the surveillance test results'.
Plate 1-19 was identified as the limiting beltline material, because of its high initial RT of 40*F.
The EOL g
reference temperature is 126*F.
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The initial RT and predicted shif t for plate 1-19 were ET compared with the corresponding values used to develop the operating limits curves in Reference 6.
The shift calculation in this analysis incorporated a correction for the surveillance test results, which increases the shift for a given fluence.
However, the EOL fluence calculated is lower than that used in Reference 6.
The net result is that the EOL shift is identical to that reported in Reference 6.
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The USE values for beltline plates 1-20 and 1-21 and for the beltline weld material were predicted using Reference 5 for the EOL condition.
Comparison of the unirradiated and irradiated specimen USE values showed Reference 5 to be conservative. The plate values were corrected to 65% of the longitudinal USE to estimate the transverse USE, according to the recommendation in Branch Technical Position MTEB 5-2 (Reference 7).
The lowest EGL "alue.s calculated were 70 ft-lb and 86 ft-lb for the plate and weld materials, respectively.
2.2 CONCLUSION
S Surveillance testing is done to support or replace calculated information related to cperating conditions and EOL vessel beltline conditions, as outlined in Reference 1.
The results obtained from surveillance testing which impact these conditions are the predicted EOL fluence, the shift in RT fr the beltline materials and the ET decrease in USE.
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NEDC-31166 Rsv. 1 The flux wire testing, combined with the lead factors from the Reference 6 analysis, give an EOL fluence prediction which is about 20% lower than the value predicted in Reference 6.
This fluence, based on a daily power history for seven fuel cycles, is considered more accurate than the fluence based on the flux wire test results from the end of Fuel Cycle 2.
Since the limiting beltline material initial RT and shift ET determined in this report are the same as were used in Reference 6, there is no need to change the operating limits curves developed in that report.
The only changes needed in the Technical Specification and UFSAR relate to the completion of the testing of the first surveillance capsule.
The suggested changes are in Appendices C and D.
The predicted EOL reference temperature and the USE results developed from the surveillance testing 16dicate that the DAEC vessel is in compliance with the fracture toughness requirements of 10CFR50 Appendix G.
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NEDC-31166 Rsv. 1 3.
SURVEILLANCE PROGRAM BACKGROUND 3.1 CAPSULE RECOVERY The DAEC reactor was shut down in February 1985 for refueling and 6
maintenance. The accumulated thermal power output was 3.45x10 mwd or 5.9 EFPY.
The reactor pressure vessel (RPV) originally contained three surveillance capsules at 36*,
108', and 288' azimuths at the core midplane.
The specimen capsules are held against the RPV inside surface by a spring loaded specimen holder.
Capsule 1 at 288* was removed by DAEC personnel on April 9, 1985.
The capsule was cut from the holder assembly and shipped via a 200 Series cask to the General l
Electric Vallecitos Nuclear Center in Pleasanton, California.
Upon arrival at Vallecitos, the capsule was examined for identification.
The reactor code of 35 from Reference 8 and the capsule number were confirmed on the capsule, as shown in Figure 3-1.
The capsule contained two Charpy specimen packets and three tensile specimen tubes.
Each Charpy packet contained 12 Charpy specimens and 3 flux wires.
The three tensile specimen tubes contained a total of six specimens.
The tensile specimen gage sections were protected by aluminum sleeves, and during. removal of the sleeves, the threaded ends of the specimens were slightly damaged. The threads were later chased with a die-hex re-threading tool.
The gage sections of the tensile specimens were not damaged during removal.
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RPV MATERIALS AND FABRICATION BACKGROUND 3.2.1 Fabrication History The DAEC RPV is a 183-inch diameter BWR/4.
It was field fabricated by Chicago Bridge & Iron to Section III of the 1965 ASME Code with Addenda up to and including Winter 1967. The shell and head plates are ASME SA-533 Grade B, Class I low alloy steel (LAS).
The nozzles and closure flanges are ASME SA-508 Class 2 LAS, and the 3-1
NEDC-31166 Rav. I closure flange bolting materials are ASTM A540 Grade B24 LAS.
The fabrication process employed quench and temper heat treatment immediately after hot forming, then shielded metal arc welding (SMAW) and post-weld heat treatment in the field.
The post-weld heat treatment was typically 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> at 1150' 25'F.
The arrangement of platcs and welds relative to the core beltline and various nozzles is shown in Figure 3-2.
3.2.2 Material Properties of RPV at Fabrication A search of General Electric Quality Assurance (QA) records was made to determine the chemical and mechanical properties of the plates and welds in the RPV beltline.
The results were reported in Reference 6, and are summarized in Table 3-1, including the initial RT values calculated.
The Reference 6 work was supplemented with g
more specific information from Chicago Bridge & Iron on the weld materials that went into the beltlir.e longitudinal and girth welds.
The limiting veld materials used in the beltline welds are shown in Table 3-1.
3.2.3 Specimen Chemical Composition Samples were taken from base metal tensile specimens ETJ and ETK and from weld metal tensile specimens EU3 and EU6 af ter they were tested.
Samples from the tensile specimens represent the Charpy specimens as well, since both were taken from the same plate and weld, as discussed in Subsection 3.3.
A total of four samples were analyzed.
Chemical analysis was performed using a plasma emission spectrometer. Each sample was decomposed and dissolved, and a portion prepared for evaluation by the spectroneter.
The spectrometer was calibrated with a standard solution containing 700 ppm Fe, 8 ppm Mn, 2 ppm Cu, 5 ppm Ni, 5 ppm Mo, 5 ppm Cr, 1 ppm Si, 1 ppm Co, and levels of perchloric acid and lithium consistent with the test.
The phosphorus calibration was done by analyzing a series of seven National Bureau of Stand.irds (NBS) steels with known phosphorus contents. The chemical composition results are given in Table 3-2.
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NEDC-31166 R2v. 1 3.3 SPECIMEN DESCRIPTION The surveillance capsule contained 24 Charpy specimens:
base metal (8), weld metal (8), and HAZ (8).
There were 6 tensile specimens: base metal (2), weld metal (2), and HAZ (2).
The 6 flux wires recovered were iron (2), nickel (2) and copper (2).
The chemistry and fabrication history for the Charpy and tensile specimens are described in this section.
3.3.1 Charpv Specimens The fabrication of the Charpy specimens is described in Surveillance Test Drawings T4 through T12 (Reference 9).
All materials used for specimens were beltline materials from the lower intermediate shell course.
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The base metal specimens were cut from plate 1-21 (Heat B0673-1).
The chemical analysis of this heat of low alloy steel is in Reference 6, as summarized in Table 3-1.
The test plate was heat treated for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> at 1150*F plus 25'F or minus 50*F to conservatively simulate the post-weld heat treatment of the vessel.
The method used to machine the specimens from the test plate is shown in Figure 3-3.
Specimens were machined from the 1/4 T and 3/4 T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction).
The identification of the base metal Charpy specimens recovered from the surveillance capsule is shown in Table 3-3.
The weld metal and HAZ Charpy specimens were fabricated from pieces of plate 1-21 that were welded together using "the same welding procedures as used to weld the core area shell plates together" (Reference 9).
The chemical analysis of the surveillance weld material is shown in Table 3-2.
The welded test plate for the weld and HAZ Charpy specimens received a heat treatment of 1150*F plus 25'F or minus 50*F for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> to conservatively represent the heat 1
treatment of the RPV.
The weld specimens and HAZ specimens were 3-3 l
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NEDC-31166 Rw. 1 fabricated as shown in Figures 3-4 and 3-5, respectively.
The base metal orientation in the veld and HAZ specimens was longitudinal.
Contained in Table 3-3 are the identifications of the weld metal and HAZ Charpy specimens from the surveillance capsule.
3.3.2 Tensile Specimens Fabrication of the surveillance tensile specimens is described in the Reference 9 drawings. The chemical composition and heat treatment for the base, weld and HAZ tensiles are the same as those for the corresponding Charpy specimens.
The identifications cf the base, weld, and HAZ tensile specimens recovered from the surveillance capsule are given in Table 3-3.
A summary of the fabrication methods is presented below.
The base metal specimens were machined from material at the 1/4 T and 3/4 T depth in plate 1-21.
The specimens, oriented along the plate rolling direction, were machined to the dimensions shown in Figure 3-6.
The gage section was tapered to a minimum diameter of 0.250 inch at the center.
The veld metal tensile specimen material was cut from the welded test plate, as shown in Figure 3-7.
The specimens were machined entirely from weld metal, scrapping material that might include base metal.
The fabrication method for the HAZ tensile specimens is illustrated in Figure 3-8.
The specimen blanks were cut from the welded test plate such that the gage section minimum diameter was machined at the HAZ centerline.
The finished HAZ specimens are approximately half weld metal and half base metal, oriented along the plate rolling direction.
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NEDC-31166 Rav. 1 3.3.3 Unirradiated Archive Specimens In order to develop meaningful results from the surveillance test
- data, unirradiated specimen test data must be available for comparison.
After examining the Chicago Bridge &
Iron (CB&I) fabrication test records, it was concluded that there had been no adequate unirradiated specimen data collected.
The plate tests done as part of the fabrication test program by CB&I were on a different plate from that used for the surveillance specimens.
There was no CB&I Charpy data located on the weld material used for the surveillance weld.Therefore, unirradiated archive specimens were retrieved from storage at IEL&P.
The archive specimens are extra Charpy and tensile specimens fabricated for the surveillance program.
They have the same geometries and material properties as the corresponding irradiated specimens taken from the surveillance capsule.
IEL&P provided 18 each of archive base, weld and HAZ metal Charpy specimens for testing.
Since these archive specimens are irreplaceable, only those specimens needed to adequately define the Charpy toughness transition curves were tested.
A total of 14 base, 12 weld and 12 HAZ specimens were tested.
These specimen identifications are shown in Table 3-4.
IEL&P also provided 3 base metal and 3 weld metal tensile specimens (HAZ tensiles are no longer required by ASTM E185).
These archive tensile specimens were tested at room temperature, reactor j
operating temperature (550*F) and at the temperature corresponding to
(
each material's onset of upper shelf.
The archive tensile specimen l
identifications are listed in Table 3-4.
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Table 3-1 MATERIAL PROPERTIES OF VESSEL COMPONENTS Chemistry (wt %) Charpy Test Energy Dropweight RTNDT*
Identification Heat / Lot No.
Cu P
Ni Temp. (*F)
(ft-lb)
Temp. (*F)
(
- F)
Lower Plates:
(Shell 1) 1-18 C6439-2 0.09 0.012 0.51 40 36,48,43 40 40 1-19 B0402-1 0.13 0.012 0.47 40 83,85,72 40 40 1
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Lower-Intermed. Plates:
(Shell 2) 1-20 B0436-2 0.15 0.008 0.64 40 57,54,62
-30 10 2
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1-21 B0673-1 0.15 0.011 0.61 40 99,104,121
-30 10 6
Longitudinal Welds Ht. 432Z0471 0.03 0.017 0.91 10 100,102,106
-50 g;
Lot B003A27A Girth Weld Ht. 07L669 0.03 0.014 1.02 10 50,50,54
-50 Lot K004A27A Upper Plates:
(Shell 4) 1-24 C6794-2 10, 37,44,33 10 14 1-25 C7090-1 10 61,67,79 10 10
?
Standby Liquid E20VW 40 38,60,26 40 58 4
Control Nozzle
" RT values are calculated to be equivalent to transverse RTNDT*
NEDC-31166 Rzv. 1 Table 3-2 PLASMA EMISSION SPECTROMETRY CHEMICAL ANALYSIS OF RPV SURVEILLANCE PLATE AND WELD MATERIALS Base Metal Base Metal Weld Metal Weld Metal Element Tensile ETJ Tensile ETK Tensile EU3 Tensile EU6 Mn 1.4 1.3 1.3 1.2 P
0.006 0.006 0.011 0.010 Cu 0.15 0.15 0.02 0.02 Si
- 0.07 0.06 0.32 0.33 Ni 0.70 0.69 1.00 0.90 Mo 0.63 0.62 0.49 0.49 Cr 0.14 0.14 0.04 0.03 Co 0.014 0.013 0.013 0.012 l
a Si results may be low, because of precipitation during dissolution heating.
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NED':-31166 R2v. 1 Table 3-3 IDENTIFICATION OF CHARPY AND TENSILE SPECIMENS REMOVED FROM SURVEILLANCE CAPSULE Charpy Specimens Base Weld HAZ EBJ
- EE2 ELK EBK EE3 ELL EBL EE4 ELM EBP EES ELP EBT EJ1 ELT EBU EJ2 ELY EBY EJ3 EMT EC1 EJ4 EMY Tensile Specimens Base Weld HAZ ETJ "
EU3 EY1 ETK EU6 EY2 8 All specimen identifications include a double-dot over the middle symbol.
3-8
NEDC-31166 Rev. 1 Table 3-4 IDENTIFICATION OF CHARPY AND TENSILE ARCHIVE SPECIMENS Charpy Specimens Base Weld HAZ ECK ^
EED EMI ECL EEE EM2 ECM EEK EM3 ECP EEL EM4 ECT EEM EMS ED1 EET EM6 7
ED4 EEY EMA ED5 EK2 EP2 ED6 EK3 EP3 ED7 EK4 EP4 i
EDA EK5 EP5 EDK EK6 EP7 EDT EDY Tensile Specimens Base Weld l
ET2
- EUM ET3 EUJ ETA EUP e
I
- All specimen identifications include a double-dot over the middle symbol.
l 3-9 i
NEDC-31166 Rev. 1
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Su rve i l 1.m e e Capsuie Recovered f rom DAEC Reactor i
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WEDC-31166 Rev. 1
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Schematic of the RPV Showing Arrangement of Vessel Plates and Welds 3-11
NEDC-31166 Rev. 1 ROLLING DIRECTION 4
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Fabrication Method for Base Metal Charpy Specimens 3-12
NEDC-31166 Rev. 1 ROLLING DIRECTION
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Fabrication Method for Weld Metal Charpy Specimens i
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3-13
NEDC-31166 Rev. 1 ROLLING DIRECTION c
L BELTLINE WELD
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lll HAZ CEN TERLINE Figure 3-5.
Fabrication Method for HAZ Charpy Specimens 3-14
EEDC-31166 Rev. 1 ROLLING DIRECTION
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NOTES:
- 1. D = 0.250 0.001 DIAM AT CENTER OF REDUCED SECTION
- 2. D' = ACTUAL"D" DI AM + 0.002 TO O.005 AT ENDS OF REDUCED SECTION, TAPERING TO"O
- AT CENTER Figure 3-6.
Fabrication Method for Base Metal Tensile Specimens i
3-15
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NEDC-31166 Rev. 1 I
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Fabrication Method for Weld Metal Tensile Specimens 3-16 l
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Fabrication Method for HAZ Tensile Specimens 3-17/3-18 i
i
NEDC-31166 R;v. 1 4.
FLUX WIRE TEST EVALUATION Reference 6 includes a determination of the peak vessel flux and fluence, based on analytical results and results from a test performed on flux wires removed from the vessel af ter Fuel Cycle 2.
The flux wires removed with the current surveillance capsule give more representative values of flux, as explained later, and provide a measurement of the fluence accumulated by the Charpy and tensile surveillance specimens.
Updated values of peak vessel flux and fluence are computed using the latest flux wire results and the lead factors from Reference 6.
4.1 FLUX WIRE ANALYSIS 4.1.1 Procedure The surveillance capsule contained six flux wires: two each of iron, nickel and copper.
Each wire was removed from the capsule, cleaned with dilute acid, weighed, mounted on a counting card, and analyzed for its radioactivity content by gamma spectrometry.
Each iron wire was analyzed for Mn-54 content, each nickel wire for Co-58 and each copper wire for Co-60 at a calibrated 4-or 10-cm source-to-detector distance with an 80-cc Ce(Li) detector system. The gatma spectrometer was calibrated using NBS material.
To properly predict the flux and fluence at the surveillance capsule from the activity of the flux wires, the periods of full and partial power irradiation and the zero power decay periods were considered.
Operating days for each fuel cycle and the average reactor power fraction are shown in Table 4-1.
Zero power days between fuel cycles are listed as well.
4-1
i NEDC-31166 Rsv. 1 From the flux wire activity measurements and power history, reaction rates for Fe-54 (n.p) Mn-54, Ni-58 (n.p) Co-58 and Cu-63 (n a) Co-60 were calculated.
The
>l MeV fast flux reaction cross sections for the iron, copper, ano nickel wires were estimated to be 0.157 barn, 0.0027 barn and 0.204 barn,' respectively.
These values were obtained from measured cross-section functions determined at Vallecitos from more than 65 spectral determinations for BWRs and for the General Electric Test Reactor using activation monitor and spectral unfolding techniques.
Similarly, the >0.1 HeV cross sections l
for the iron, copper, and nickel wires were estimated from the measured 1-to-0.1 cross-section ratio of 1.6.
3 l
4.1.2 Results I
l The measured activity, reaction rate and determined full-power flux results for the surveillance capsule are given in Table 4-2.
The
>l MeV and >0.1 MeV flux values of 2.6x10' and 4.2x10' n/cm -see from 8
the flux monitors were calculated by dividing the reaction rate measurement data by the appropriate cross sections. The corresponding 17 l
fluence results, 4.9x10 and 7.8x10" n/cm8 for >l MeV and >0.1 MeV, respectively, were obtained by multiplying the full-power flux density 8
values by the product of the total seconds irradiated (2.79x10 see) and the full-power fraction (0.671).
Generally, for long-term irradiations, dosimetry results from copper flux wires are considered most accurate because of the length of the half-life of Co-60 (5.27 years) compared to those of iron's Mn-54 (312.5 days) and nickel's Co-58 (70.8 days).
The iron and nickel flux wires, which are more sensitive to fluctuations in reactor power levels and to peripheral bundle power variations, showed reasonable agreement with the copper wires, varying by 10% and 20%,
respectively.
Given this, and the f act that the copper wire results were the most conservative, the copper flux wire results were used to predict capsule fluence and EOL vessel fluence.
4-2 l
- ~.
r l
l NEDC-31166 Rsv. I l
The accuracies of the values in Table 4-2 for a 2e deviation are estimated to be:
1 5% for dps/g (disintegrations per second per gram)
! 12% for dps/ nucleus (saturated) 30% for flux and fluence >l MeV i 40% for flux and fluence >0.1 MeV A set of flux wires from DAEC was evaluated by General Electric 9
2 in 1977. The
>l MeV flux was 3.1x10 n/cm -sec.
The flux from this study of 2.6x10' n/cm -see is about 15% lower.
Accounting for differences in lead factor between the two tests, the current flux is effectively 23% lower than expected, based on the first flux wire test result.
This difference is probably due to the variations in operation typical in the first two fuel cycles.
The most recent flux wires, which have been exposed to years of steady-state operation, provide a more reliable reading of the vessel steady-state flux.
Therefore, the latest measured flux is used in Subsection 4.3 to predict EOL fluence.
4.2 FLUX DISTRIBUTION LEAD FACTORS A lead factor is a calculated ratio between the flux at the surveillance capsule location and the flux at the location of peak flux in the beltline region (at the inside surface or at 1/4 T depth).
The computer analyses required to establish the lead factors for the DAEC vessel were performed for Reference 6, and the results are discussed therein. The lead factors for the 288* capsule are 0.78 to the peak inside surface location and 0.99 to the peak 1/4 T location.
l 4-3
NEDC-31166 R2v. 1 4.3 ESTIMATE OF END-OF-LIFE FLUENCE The EOL fluence (f) is estimated using the upper bound (130%) of the measured flux from Table 4-2 with the lead factor to the 1/4 T location. As discussed in Reference 6, a power uprate from 1593 MW to 1658 MW is assumed at 6 EFPY.
Flux is proportional to thermal power for a given core radial power shape, so the two fluxes are:
9 2.6x10' n/cm -s
- 1.30/0.99 = 3.41x10 n/cm*-s for 6 EFPY, and 3.41x109 * (1658/1593) = 3.55x109 s
n/cm -s for 7 through 32 EFPY.
The EOL condition of 32 EFPY represents 1.01x10' seconds of full power 7
irradiation, or 3.16x10 seconds per EFPY. The EOL fluence is:
f = 3.16x10 sec/EFPY(3.41x10'
- 6 EFPY + 3.55x10'
- 26 EFPY), or 18 2
f = 3.6x10 n/cm,
This EOL fluence is significantly lower than the value used in 10 2
the Reference 6 analysis of 4.4x10 n/cm.
The impact of the lower fluence value is discussed in Section 7.
i I
l 4-4 l
l
NEDC-31166 Rsv. 1 Table 4-1
SUMMARY
OF DAILY POWER HISTORY Operating Percent of Days Between Cycle Cycle Dates Days Full Power Cycles la 5/14/74 - 6/6/75 389 0.525 42 lb 7/19/75 - 2/13/76 210 0.686 61 2
4/15/76 - 3/12/77 332 0.695 62 3
5/14/77 - 3/18/78 309 0.823 39 4a 4/27/78 - 6/17/78 52 0.700 265 4b 3/10/79 - 2/9/80 337 0.808 68 5
4/18/80 - 3/20/81 337 0.806 73 6
6/2/81 - 2/12/83 621 0.540 82 7
5/6/83 - 2/2/85 639 0.651 3226 0.671 (average) l 4-5
f Table 4-2 SURVEILLANCE CAPSULE LOCATION FLUX AND FLUENCE FOR 1RRADIATION FROM 5/14/74 TO 2/2/85 Wire dps/g Element Reaction Rate Full Power Flux
- Fluence a
2 Wire Weight (at end of
[dps/ nucleus (n/cm -s)
(n/cm )
(Element)
(g)
Irradiation (saturated)]
>l MeV
>0.1 MeV
>l MeV
>0.1 MeV 4
-18 Copper 64713 0.3611 1.99x10 7.15x10 0
-18 Copper 64741 0.3588 1.87x10 6.75x10
-18 9
9 I7 I7 Average = 6.95x10 2.6x10 4.2x10 4.9x10 7.8x10 x
k
-16 c.
Iron 64713 0.0694 1.39x10 3.67x10
-16 Iron 64741 0.1663 1.37x10 3.62x10 Average = 3.65x10 2.3x10'
-16
-6 Nickel 64713 0.3234 1.84x10 4.04x10 6
-16 Nickel 64741 0.3211 1.93x10 4.24x10
-16 9
Average = 4.14x10 2.0x10 l
~
- =
E
- Full power of 1593 MW *
~
t l
l l
NEDC-31166 Rav. 1 5.
CHARPY V-NOTCH IMPACT TESTING The 24 Charpy impact specimens recovered from the surveillance capsule and 38 of the unirradiated archive Charpy impact specimens provided by IEL&P were tested at temperatures selected to establish the toughness transition and USE of the RPV beltline materials.
Testing was conducted in accordance with ASTM E23-82 (Reference 10).
5.1 PROCEDURE The testing machine used was a Riehle Model PL-2 impact machine, serial number R-89916.
The pendulum has a maximum velocity of 15.44 ft/see and a maximum available hammer energy of 240 f.t-lb.
The test apparatus and operator were qualified using U.S. Army Watertown standard specimens.
The standards are designed to fail at 74.1 ft-lb and 13.9 ft-lb at a test temperature of
-40*F.
Acccyding to Reference 10, the averaged test apparatus results uust reproduce the Watertown standard values within an accuracy of 5% or 21.0 f t-lb, whichever is greater.
The successful qualification of the Riehle machine and operator is summarized in Table 5-1.
Charpy V-Notch tests were ' conducted at temperatures between
-100*F and 400*F.
For tests between 32*F and 212*F, the temperature conditioning fluid was water.
Dichloromethane was used at temperatures below 32*F.
Above 212*F, a silicone oil was used.
Cooling of the conditioning fluids was accomplished by circulating liquid nitrogen through an immersed coil of copper tubing, and heating by an immersion heater.
The fluids w'ere mechanically stirred to maintain uniform temperatures.
The, fluid temperature was measured by a chromel-alumel thermocouple and a copper-constantan thermocouple.
These were calibrated with boiling water' (212*F), and ice water (32*F).
Once at test temperature, the specimens were manually transferred with centering tongs to the Riehle machine and impacted within 5 seconds.
5-1
NEDC-31166 R2v. 1 i
For each Charpy V-Notch specimen tested, test temperature, energy
- absorbed, lateral expansion, and percent shear were evaluated.
Lateral expansion and percent shear were measured according to Reference 10 methods.
Percent shear was determined with Method 2 of Subsection 11.2.4.3 of Reference 10, which involves a comparison of the fracture surface appearance with the reference fracture surfaces in Figure 15 of Reference 10.
5.2 RESULTS Eight irradiated Charpy V-Notch specimens each of base metal, weld metal and 3AZ vere tested at temperatures selected to define the toughness transition and USE portions of the fracture toughness curve.
Twelve each of the unirradicted base, weld and HAZ specimens were similarly tested.
Subsequently, two additional unitradiated base metal specimens were tested to better define the transition region.
Absorbed energy, lateral expansion and percent shear data are listed in Table 5-2 for the irradiated specimens and in Table 5-3 for the unirradiated specimens. Plots of absorbed energy data for base, weld, and HAZ metal are presented in Figures 5-1, 5-2 and 5-3, respectively.
Lateral expansion plots for base, weld and HAZ metal are given in Figures 5-4, 5-5 and 5-6, respectively.
The plots show the unirradiated and irradiated data side-by-side to facilitate the i
l determination of irradiation shifts.
The data sets were freehand-fit with best-estimate S-shaped cu rves, characteristic of fracture toughness transition curves.
The curves of impact energy and lateral expansion were used to determine the index temperatures shown in Table 5-4.
Longitudinal USE values were taken from the cu rves,
and then corrected to equivalent transverse USE with the methods in Reference 7.
5-2
NEDC-31166 Rsv. 1 The HAZ data typically show the greatest scatter, because the HAZ l
has in effect been uniquely heat treated by the welding process, and l
l because of the uncertainty of the specimen notch location relative to the veld fusion line.
The HAZ results in this case are very consistent, probably because there is little difference in the Charpy curves for the base and weld metals.
An unirradiated base metal specimen was tested at 400*F, resulting in a no-break test. This was unexpected, since the material has an upper shelf of 160 ft-lb and the test machine capacity is 240 ft-lb.
The specimen appeared to have been struck properly, but slight misalignment may have caused the no-break result.
The result is not shown on Figure 5-1.
A retest at 400*F was not. considered necessary because the USE is well defined by the other data shown on Figure 5-1.
Photographs were taken of the Charpy specimen fracture surfaces.
The fracture surface photographs were used to evaluate percent shear.
The photographs and a summary of test results for each specimen are contained in Appendix A for irradiated specimens and in Appendix B for unirradiated specimens.
5.3 IRRADIATED VERSUS UNIRRADIATED CHARPY V-NOTCH PROPERTIES The differences in index temperatures shown in Table 5-4 represent the change in fracture toughness properties due to the amount of irradiation experienced by the surveillance capsule.
Reference 4 states that the change in RT f a ma eria should be NDT characterized by the change in the 30 ft-lb index temperature.
Therefore, the experimental shift in RT f r the base metal is 42*F, NDT and for the weld metal is 0*F.
5-3
NEDC-31166 R2v. 1 Table 5-1 QUALIFICATION TEST RESULTS USING U.S. ARMY WATERTOWN SPECIMENS (TESTED IN FEBRUARY 1986)
Energy Absorbed Qualification Test Test Temperature Mechanical Gage Specimen Identification
(*F)
(ft-lb)
EE3-0427
-40 72.0 EE3-0075 74.5 -
EE3-0233 80.0 EE3-0956 75.8 EE3-0365 74.2 Average 75.3 Allowable
-40 74.1 2 3.7 l
Acceptable DD5-0442
-40 15.0 DD5-0905 13.8 DD5-0486 14.0 DD5-0977 14.0 DD5-0831 14.0 Average 14.2 Allowable
-40 13.9 1.0 Acceptable 5-4
NEDC-31166 Riv. 1 Table 5-2 CHARPY V-NOTCH IMPACT TEST RESULTS FOR IRRADIATED RPV MATERIALS Test Fracture Lateral Percent Shear Specimen Temperature Energy Expansion (Method 2)
Identification
(*F)
(ft-lb)
(mils)
(%)
Base:
EBU
-60 4.0 4
0 EBP
-20 15.0 19 0
EBT 10 15.5 26 40 EC1 20 49.5 45 40 EBK 40 60.0 49 40 EBJ 120 101.5 70 70 EBL 200 144.5 95 90 EBY 400 160.0 98 90 Weld:
EJ1
-100 6.5 11 0
EJ2
-60 15.3 24 10 EJ4
-40 22.0 26 10 EE5
-20 63.5 54 40 EE2 40 75.2 69 60 EE3 120 97.2 86 80 EE4 200 109.0 90 70 EJ3 400 101.0 91 100 HAZ:
ELY
-60 8.C 15 0
ELP
-20 13.5 18 20 EKY 0
15.0 16 30 ELT 10 51.7 43 40 ELK 40 65.0 50 50 ELL 120 90.1 66 70 ELM 200 126.5 84 90 EMT 400 124.5 75 100 5-5
NEDC-31166 Rav. 1 Table 5-3 CHARPY V-NOTCH IMPACT TEST RESULTS FOR UNIRRADIATED RPV MATERIALS Test Fracture Lateral Percent Shear Specimen Temperature Energy Expansion (Method 2)
Identification
(*F)
(ft-lb)
(mils)
(%)
Base:
ED4
-100 6.5 7
0 ED1
-80 4.2 5
0 ECT
-40 9.0 14 10 EDK
-40 24.0 30 5
ED7
-30 38.0 35 5
EDT
-30 49.0 30 5
ED6
-20 47.0 43 10 EDS
-10 43.0 43 20 ECP 0
56.5 49 25 ECM 40 98.5 73 40 ECL 120 134.5 73 80 ECK 200 158.5 93 85 EDY 300 163.5 81 90 EDA 400 no break Weld:
EET
-100 6.5 11 10 EEK
-80 5.5 10 15 EK3
-60 16.8 19 40 EK4
-50 21.5 22 20 EEE
-40 49.0 45 20 EK2
-20 58.0 57 50 EEL 0
52.5 50 30 EEY 40 65.0 61 50 EED 120 102.0 88 80-EEM 200 94.0 65 70 l
EK5 300 88.2 93 60 l
EK6 400 113.7 96 70 i
i HAZ:
EM6
-100 5.8 12 5
(
EMS
-80 8.0 17 10 EM4
-40 23.8 25 20 EM3 0
28.0 28 20 l
EP2 20 37.0 36 30 EMA 40 65.0 62 40 EP3 80 71.5 67 60 EMI 120 90.0 77 80 EP4 160 109.0 90 60 l
EM2 200 128.0 89 85 l
EPS 300 113.0 89 60 EP7 400 107.0 94 85 l
l 5-6
NEDC-31166 RIv. 1 Table 5-4 COMPARISON OF UNIRRADIATED AND IRRADIATED CHARPY V-NOTCH DATA Upper
- Shelf Index Temperature (*F)
Energy (ft-lb)
Material E=30 ft-lb E=50 ft-lb MLE=35 mil L/T Base:
Unieradiated
-35'F
-22*F
-26*F
.164/107 Irradiated 7'F 23*F 16*F 160/104 Difference 42*F 45'F 42*F 4/3 (2.5%)
Weld:
Unirradiated
-33*F
-11*F
-34*F 101/101 Irradiated
-33*F
-11*F
-34*F 101/101 Difference 0*F 0*F 0*F 0/0 (0%)
HAZ:
Unitradiated
-6*F 34*F
-2*F 112/73 Irradiated 3*F 38*F 40*F 126/82 Difference 9'F 4*F 42*F
-14/-9 (-11.1%)
Longitudinal (L) USE is read directly from Figures 5-1, 5-2 and 5-3.
Transverse (T) USE is taken as 65% of the longitudinal USE, according to Reference 7.
L/T USE values are equal for weld metal, which has no orientation effect.
5-7 A
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NEDC-31166 R,v. 1 6.
TENSILE TESTING l
Uniaxial tensile tests were conducted on six irradiated tensile specimens recovered from the vessel surveillance capsule and on six unirradiated archive specimens provided by IEL&P.
Tests were conducted in accordance with ASTM E8-81 (Reference 11) in air at room temperature, at onset to upper shelf temperature and at RPV operating temperature.
6.1 PROCEDURE All tests were conducted using a screw-driven Instron test frame equipped with a 20-kip load cell and special pull bars and grips.
Heating was done with a Satec resistance clamshell furnace centered around the specimen load train.
Test temperature was monitored and controlled by a chromel-alumel thermocouple spot-welded to an Inconel clip that was friction-clipped to the surface of the specimen at its midline.
Before the elevated temperature tests, a profile of the furnace was conducted at the test temperature of interest using an unirradiated steel specimen of the same geometry.
Thermocouples were spot-welded to the top, middle, and bottom of a central 1-in. gage of this specimen.
In addition, the clip-on thermocouple was attached to the midline of the specimen.
When the target temperatures of the three thermocouples were within 15'F of each other, the temperature of the clip-on thermocouple was noted and subsequently used as the target temperature for the irradiated specimens.
All tests were conducted at a calibrated crosshead speed of 0.005 in./ min until well past yield, at which time the speed was increased to 0.05 in./ min until fracture.
A 1-in. span knife edge extensometer was attached directly to each specimen's central gage region and was used to monitor gage extension during the test.
I
{
6-1
1 NEDC-31166 R;v. 1 The test specimens were machined with a minimum diameter of 0.250 inch at the center of the gage length.
Irradiated specimens of base, weld and HAZ metal were tested at room temperature (RT = 76'F) and RPV operating temperature (550*F).
Unitradiated base and weld metal specimens were tested at room temperature, at operating temperature and at the temperature corresponding to the onset of the material's upper shelf (185'F for base metal and 130*F for weld metal).
l The yield strength (YS) and ultimate tensile strength (UTS) were calculated by dividing the nominal area (0.0491 in.2) into the 0.2%
offset load and into the maximum test load, respectively.
The values listed for the uniform and total elongations were obtained'from plots that recorded load versus specimen extension and are based on a 1-in.
gage length.
Reduction of area (RA) values were determined from post-test measurements of the necked specimen diameters using a calibrated blade micrometer and employing the formula:
RA(%) = 100% * (A,- A )/A, g
Af ter testing, each broken specimen was photographed end-on, showing the fracture surface, and lengthwise, showing fracture location and 1
local necking behavior.
6.2 RESULTS Tensile test properties of YS, UTS, RA, uniform elongation (UE) and total elongation (TE) are presented in Table 6-1 for irradiated and unitradiated specimens.
Shown in Figure 6-1 is a stress-strain curve for a 550*F base metal irradiated specimen typical of the stress-strain characteristics of all the specimens tested.
The results presented in Table 6-1 are shown graphically as follows:
YS in Figure 6-2, UTS in Figure 6-3, UE in Figure 6-4 TE in Figure 6-5 and RA in Figure 6-6.
Photographs of fracture surfaces and necking behavior are given in Figures 6-7, 6-8 and 6-9 for irradiated 6-2 4
---,-------n ea
,-g v.------
,rm.-.m-------,w-,-------n,-----------n--- -, - - - - - - -,. - - - -
NEDC-31166 R;v. 1 base, weld and HAZ specimens, respectively. The unitradiated specimen photographs are in Figure 6-10 for base metal and in Figure 6-11 for veld metal.
The base, veld, and HAZ materials follow the trend of decreasing i
properties with increasing temperature.
The three materials behave very similarly, as seen in Figures 6-2 through 6-6.
The third data value for the unirradiated tests shows that the relationship is 4
non-linear.- Straight lines have been drawn only to show the trends.
6.3 IRRADIATED VERSUS UNIRRADIATED TENSILE PROPERTIES Unirradiated tensile test data are compared in Table 6-2 to the irradiated base metal and weld metal specimen data at 76*F and at 550*F to determine the degree of radiation strengthening. The percent difference is expressed relative to the ir, radiated value, as explained by the equation in the footnote to Table 6-2.
The changes in properties expected with irradiation embrittlement are an increase in YS and UTS with a corresponding decrease in UE, TE and RA.
As shown in Table 6-2, the base metal values reflect such a trend, indicating that a measurable chatge has occurred as a result of 1
irradiation.
The weld metal properties show slight increases in YS and UTS, but the changes in UE, TE and RA are inconsistent with embrittlement. Any change in the weld metal properties as a result of irradiation would likely be small.
These conclusions are consistent i
with the RT shi ts rep rted in Section 5.
NDT i
2 i
6-3 i
NEDC-31166 R;v. 1 1
Table 6-1 TENSILE TEST RESULTS FOR RPV MATERIALS Test Yield Ultimate Uniform Total Reduction Specimen Temp Strength Strength Elongation Elongation of Area Number Material
(*F)
(ksi)
(ksi)
(%)
(%)
(%)
Irradiated:
ETJ Base 76 73.4 96.2 9.5 19.7 70.4 ETK Base 550 67.7 91.7 7.6 15.9 65.9 EU3 Weld 76 79.4 92.9 8.2 17.5 70.0 EU6 Weld 550 65.7 83.5 7.1 15.1 64.0 EY1 HAZ 76 74.2 94.8 9.7 20.7 66.8 EY2 HAZ 550 67.7 87.6 6.6 14.4 59.0 Unirradiated:
ETA Base 76 66.5 90.2 10.2 20.8 71.1 ET2 Base 185 63.2 90.5 9.0 19.3 74.3 ET3 Base 550 61.2 86.7 8.2 16.9 70,0 i
EUP Weld 76 75.4 89.7 10.0 21.6 71.3 EUJ Weld 130 80.4 91.4 9.2 21.2 71.6 EUM Weld 550 66.6 83.6 6.1 14.2 63.2 l
l l
6-4
NEDC-31166 R2v. 1 Table 6-2 COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES Yield Ultimate Uniform Total Reduction Strength Strength Elongation Elongation of Area (ksi)
(ksi)
(%)
(%)
(%)
Base Metal at 76*F:
Unirradiated 66.5 90.2 10.2 20.8 71.1 Irradiated 73.4 96.2 9.5 19.7 70.4 C
Difference 9.4%
6.2%
-7.4%
-5.6%
-1.0%
Base Metal at 550*F:
Unirradiated 61.2 85.7 8.2 16.9 70.0 Irradiated 67.7 91.7 7.6 15.9 65.9 Difference 9.6%
5.5%
-7.9%
-6.3%
-6.2%
Weld Metal at 76*F:
Unirradiated 75.4 89.7 10.0 21.6 71.3 Irradiated 79.4 92.9 8.2 17.5 70.0 Difference 5.0%
3.4%
-22.0%
-23.4%
-1.9%
l Weld Metal at 550'F:
Unirradiated 66.6 83.6 6.1 14.2 63.2 Irradiated 65.7 83.5 7.1 15.1 64.0 Difference
-1.4%
-0.1%
14.0%
6.0%
1.3%
- Difference = [(Irradiated - Unirradiated)/ Irradiated]
- 100%
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NEDC-31166 Rzv. 1 7.
DETERMINATION OF OPERATING LIMITS AND END-OF-LIFE CONDITIONS Reference 6 explains the procedure involved in developing operating limits curves for pressure
- testing, non-nuclear heatup/cooldown and core critical operation.
The vessel is divided into three basic regions: the closure flange region, the beltline region and the non-beltline region.
The only region affected by the surveillance test results is the beltline region.
Therefore, the procedures discussed in Reference 6 for the other two regions are unchanged.
The beltline operating limits are affected by the irradiation shift in RT as measured by the surveillance testing.
The methods used to incorporate the surveillance test results into the operating limits are discussed below.
7.1 EVALUATION OF RADIATION EFFECTS The shift in fracture toughness properties in the beltline materials is a function of neutron fluence and the presence of certain elements, such as copper (Cu) and phosphorus (P). The specific relationship from Reference 5 is:
SHIFT (*F) = [40 + 1000(2cu-0.08) + 5000(%P-0.008)]*(f/1019)
(7-1) l vhere:
%Cu = wt % of Cu present,
%P = wt % of P present, 8
f = fluence (n/cm ) at selected EFPY.
l l
The limiting beltline material, either plate or veld, is
(
determined based on the Cu-P content and initial Ri of the g
materials.
These data are presented for the beltline materials in Table 3-1.
7-1 l
t l
NEDC-31166 Rev. 1 7.1.1 Measured Versus Predicted Surveillance Shift i
Table 5-4 presents a measured shif t for the base metal of 42*F and of O'F for the weld metal.
Obviously, the measured shift for the weld metal is less than predicted, but the measured shift for the curveillance plate exceeds the predicted value.
The predicted shift of the surveillance plate, calculated according to Equation 7-1, assumes 0.15% Cu, 0.012% P and an upper bound fluence of:
for capsule)(1.30 uncertainty)=6.37x10 " n/cm,
17 2
2 f=(4.9x10 n/cm Upper bound fluence is used in order to be consistent with the assumptions used to establish the EOL fluence in Subsection 4.3.
The predicted shift is 32'F, versus the measured shift of 42*F.
7.1.2 Modification of the Shift Relationship Since the measured shift exceeds the predicted shift, using Reference 5 methods to predict the limiting beltline plate shift may be non-conservative.
Therefore, the relationship in Equation 7-1 is modified for the plate materials.
Equation 7-1 shows that the shift calculated is proportional to the material characteristics and to the I
square root of the fluence. Assuming that the fluence relationship is correct, the coefficient representing the materials in Equation.7-1 must be increased by the factor (42/32) or 1.31.
Therefore, values of irradiation shift for the beltline plates will be multiplied by the correction factor of 1.31 to predict shif ts for future EFPY and EOL conditions.
7-2 ll
NEDC-31166 Riv. 1 7.1.3 Limiting Beltline Material The limiting beltline' material must be determined in order to calculate the appropriate shift to be factored into the operating limits curves.
The limiting material is found by evaluating each 2
beltline plate and weld material at the EOL fluence of 3.6x10 n/cm.
Table 7-1 shows the four beltline plate chemistries and the worst case weld chemistries.
Equation 7-1, including the 1.31 factor for the plates, was used for the listed chemistries and the EOL fluence to calculate the shifts in Table 7-1.
These shif ts, when added to the initial RT values, result in the adjusted reference temperature g
(ART) for each material.
The highest ART is calculated for plate 1-19.
Since 1-19 also has the highest initial RT
.. it is the limiting beltline material for all values of EFPY.
7.1.4 Operating Limits Curves The ART for the limiting plate in Table 7-1 is 126*F.
This consists of a 40*F initial RT and a shift of 86*F.
These are the g
same values as were calculated in Reference 6 for the limiting material.
In Reference 6, the highest Cu, P and initial RT values g
for the beltline plates were conservatively enveloped.
In Table 7-1, the appropriate chemistry and RT values are considered for each g
plate.
The other differences in Reference 6 are the higher value of fluence and the absence of a surveillance test correction f actor on shift.
As it happens, when these three differences are considered, the resulting ART for the limiting beltline material is the same as l
was reported in Reference 6.
As a result, the operating limits curves generated in Reference 6 are still valid when the results of the surveillance testing are considered.
The plot of irradiation shif t versus EFPY (Figure 4-4 of Reference 6) for the limiting beltline l
material is shown in Figure 7-1.
Operating limits curves, valid to f
12 EFPY (Figure 5-4 of Reference 6) are shown in Figure 7-2.
l 7-3
NEDC-31166 Rsv. 1 7.2 PREDICTED END-OF-LIFE CONDITIONS Reference 1 states several conditions which, if exceeded, require that vessel thermal annealing be performed.
The conditions relate to RT and USE as follows:
g
- a. The ART of the limiting beltline material must be less than 200*F at EOL.
- b. The USE for all beltline materials must be greater than 50 ft-lb at EOL, accounting for reductions due to irradiation.
As described in the following subsections, these two conditions are met for the beltline materials where data were available.
7.2.1 Adjusted Reference Temperature The ART of each beltline material is shown for EOL in Table 7-1.
The highest value is 126*F for plate 1-19.
This is well below the Reference 1 requirement of 200'F.
7.2.2 Upper Shelf Energy The changes in USE due to irradiation are shown in Table 5-4 for the surveillance plate and weld materials. They are decreases of 2.5%
and 0%, respectively.
Predictions of decrease in USE can be made using Reference 5 as well.
For the surveillance materials at a 2
fluence of 4.9x10 n/cm, decreases of 12% for the plate and 9% for the veld are predicted.
The actual surveillance results show the Reference 5 values to be quite conservative.
USE predictions, based on Reference 5, are shown in Table 7-2.
The unirradiated USE values for plate 1-21 and for the beltline weld were taken from Figures 5-1 and 5-2, respectively. The percentages of decrease in USE for these materials were based on Reference 5, rather than on actual surveillance results, for conservatism.
i 7-4
t NEDC-31166 Rsv. 1 In addition to the surveillance specimen data, Charpy data were available for beltline plate 1-20 (Heat B0436-2).
These data were generated by CB&I as part of the fabrication test program. The Charpy tests were performed up to temperatures of 200*F, which provides a reasonable and conservative estimate of the USE. The unirradiated USE measured for plate 1-20 is included in Table 7-2, along with the predicted EOL value of USE.
USE data for beltline plates 1-18 and 1-19 were not available.,
The USE values in Table 7-2 are based on longitudinal Charpy specimens.
Reference 7 recommends a correction factor of 65% to estimate the transverse Charpy USE of plates.
No correction is required for veld metal, which has no orientation effect.
-As seen in Table 7-2, the corrected transverse USE for the plate materials and l
the USE for the weld are above the minimum of 50 ft-lb at EOL.
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NEDC-31166 Rsv. 1 Table 7-1 EOL ADJUSTED REFERENCE TEMPERATURES FOR BELTLINE MATERIALS Initial RTNDT Material Heat
% Cu
%P NDT Shift ART Beltline Plates:
1-18 C6439-2 0.09 0.012 40*F 55'F 95'F 1-19 B0402-1 0.13 0.012 40*F 86*F 126*F 1-20 B0436-2 0.15 0.008 10*F 86*-F 96*F 1-21 B0673-1 0.15 0.011 10*F 98*F 108'F Beltline Welds:
Longitudinal 432Z0471/
0.03 0.017
-50*F 51*F 1*F B003A27A Girth 07L669/
0.03 0.014
-50*F 42*F
-8'F K004A27A
7-6
NEDC-31166 R2v. 1 Table 7-?
END-OF-LIFE UPPER SHELF ENERGY USE (ft-lb) for USE (ft-lb) for USE (ft-lb) for 17 18 f=0 f = 4.9x10 f = 3.6x10 Material (longitudinal)
(longitudinal)
(long./trans.)
Plate 1-21 164
- 144 134/87 Plate 1-20 134
- 118 108/70 Weld 101 *
~92 86 a These are measured values. All others are calculated using methods from References 5 and 7.
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1600 CURVE A - HYDROSTATIC PRESSURE TESTS CURVE 8 - NON-NUCLEAR HEATUP AND COOLDOWN CURVE C - CORE CRITICAL OPER ATION 1400 7
7 1200 LIMITING RTNDT = 58"F 1000 3s
)
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Figure 7-2.
Pressure Versus Minimum Temperature, Valid to 12 EFPY 7-9/7-10
NEDC-31166 Rsv. 1 r
8.
REFERENCES 1.
" Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, May 1983.
2.
" Protection Against Non-Ductile Failure," Appendix G to Section III of the ASME Boiler & Pressure Vessel Code, Addenda to and including Winter 1985.
3.
" Reactor Vessel Material Surveillance Program Requirements,"
Appendix H to Part 50 of Title 10 of the Code of Federal Regulations, May 1983.
4.
" Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards, E185-82, July 1982.
5.
" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials,"
USNRC Regulatory Guide 1.99, Revision 1 April 1977.
6.
"DAEC RPV Fracture Toughness Analysis to 10CFR50 Appendix G, May 1983." General Electric Company, NEDC-30839, December 1984.
7.
" Fracture Toughness Requirements,"
USNRC Branch Technical Position MTEB 5-2, Revision 1. July 1981.
8.
" Drilled Hole Pattern," General Electric Drawing. 117C4942, Revision 2, April 1971.
9.
" Surveillance Test," Chicago Bridge & Iron Drawings T4 - T12.
10.
" Standard Methods for Notched Bar Impact Testing of Metallic Materials," Annual Book of ASTM Standards, E23-82, March 1982.
11.
" Standard Methods of Tension Testing of Metallic Materials,"
Annual Book of ASTM Standards, E8-81.
8-1/8-2
l l
NEDC-31166 Rev. 1 s
APPENDIX A IRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS Photographs of the irradiated Charpy specimens' fracture surfaces were taken to f acilitate the determination of percent shear, and to comply with the requirements of ASTM E185-82.
The pages following show the fracture surface photographs along with a summary of the Charpy test results for each specimen.
The pictures are arranged by increasing test temperature for each material, with the materials in the order of base, weld and HAZ.
A-1/A-2
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NEDC-31166 Rsv. 1 APPENDIX B UNIRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS Photographs of the unirradiated Charpy specimens' fracture surfaces were taken to facilitate the determination of percent shear, and to comply with the requirements of ASTM E185-82.
The pages following show the fracture surface photographs along with a summary of the Charpy test results for each specimen.
The pictures are arranged by increasing test. temperature for each material, with the materials in the order of base, weld and HAZ.
B-1/B-2 i......
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NEDC-31166 R2v. 1 APPENDIX C REVISIONS TO THE TECHNICAL SPECIFICATIONS Appendix C contains suggested revisions to sections 3.6. A and 4.6. A of the DAEC Technical Specifications. Suggested revisions to these sections were recently made in CE report NEDC-30839. The changes listed here are based on sections 3.6.A and 4.6.A as they were revised in Appendix A of NEDC-30839. The changes do not affect the operating limits. They simply report that the first capsule has been tested, and any significant results related to that testing. C.1 SPECIFICATION 4.6.A The only change needed in the specifications is in the surveillance requirements section 4.6. A.2, the last paragraph. The paragraph should be changed to read as fol' lows: Samples from surveillance capsule 1 at 288* were withdrawn at 6 effective full power years and tested in accordance with 10CFR50, Appendix H. Neutron flux wires installed in the surveillance capsule were tested to experimentally determine the flux and fluence at one-fourth of the beltline shell thickness, used to determine the NDTT shift. Results of the surveillance capsule testing, including fluence, material chemistry results and actual material irradiation shift, are reflected in the curves in Figure 3.6.1. The next surveillance capsule shall be withdrawn at 15 effective full power years. C-1 J
NEDC-31166 R:v. 1 C.2 BASES FOR 3.6.A AND 4.6.A The bases for 3.6. A and 4.6. A are essentially unchanged. The tenth paragraph, which discusses the surveillance program should be revised to read as follows: Neutron flux wires and samples of vessel material are installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The first capsule was removed after fuel cycle 7, at 6 effective full power years. The neutron flux wires tested were used to determine the end-of-life fluence at the 8 2 1/4 T depth in the vessel wall of 3.6x10 n/cm. Irradiated and unirradiated Charpy specimens were tested, showing that the plate shift in NDTT was 31% higher than predicted by Regulatory Guide 1.99, Revision 1. The weld showed no measurable shift. The results of the surveillance testing, including a correction factor of 1.31 for :he plate materials, have been incorporated into the curves on Figure 3.6.1. C-2
NEDC-31166 Rsv. 1 APPENDIX D REVISIONS TO THE UPDATED FINA1. SAFETY ANALYSIS REPORT Recommendations are made for changes to the Updated Final Safety Analysis Report (UFSAR) related to reporting the results of the surveillance capsule testing. The UFSAR sections affected are 5.3.1.5.3 on initial RT values, 5.3.1.6 on material surveillance, NDT 5.3.2.1 on irradiation effects on core beltline, and 5.3.3.1 on reactor vessel design. D.1 SECTION 5.3.1.5.3' The paragraph on initial RT values for the vessel shows an NDT initial RT f r the beltline weld of ,6*F. This was based on an ET examination of all weld material in the vessel shell welds. More specific information from Chicago Bridge & Iron on the beltline welds allows the RT f -6*F to be changed to -50*F. This change should ET be incorporated in section 5.3.1.5.3. D.2 SECTION 5.3.1.6 The beginning of this section of the UFSAR discusses the compliance of the surveillance program with ASTM E185. The only change in this portion, which addresses the physical components of the program, concerns the number of specimens, which is revised in Table 5.3-2. The text portion which should be revised begins with the discussion of the withdrawal schedule. That paragraph and the remainder of 5.3.1.6 should be replaced with the following text: D-1 b
NEDC-31166 Rsv. 1 The contemplated withdrawal schedule for the samples is: Withdrawal Period Integrated Flux (Full Power Years) (10 nyt >l MeV) 6 0.7 15 1.7 32 3.6 There are 37 spare impact specimens and 12 tensile out-of-reactor spare specimens. The surplus base metal is approximately 12 by 21 by 4-11/16 in. The surplus weld sample plate is approximately 6 by 33 by 4-11/16 in. The surplus plates, if it becomes necessary, can be made into specimens with the following dimensions: i Charpy V-notch specimen - 2.1 by 0.39 by 0.39 in. Tensile specimen - 0.25 in. in diameter by 3 in. long The first surveillance capsule was withdrawn after Fuel Cycle 7 for testing. It contaf.ned 24 Charpy specimens, 6 tensile specimens and six flux wires. The test results are presented in General Electric report NEDC-31166, Revision 1, and are summarized below. Gamma spectrometry was used to determine the radioactivity content of iron, copper and nickel flux wires. The fast flux reaction cross sections of these metals, based on over 65 spectral determinations for BWRs and the GE Test Reactor, were used along with the power history, shown in Table 5.3-3, to determine the full power flux at the surveillance capsule location. The flux wire test results are shown in Table 5.3-4. 4 D-2
l NEDC-31166 Rsv. 1 Table 5.3-2 NUMBER OF SPECIMENS BY SOURCE Heat-Actual Affected Suggested Specimen Specimen Base Weld Zone Period (EFPY) (EFPY) Tested Unirradiated Baseline C* 14 12 12 b T 3 3 3 In-reactor C 12 12 12 15 T 2 2 2 15 C 8 8 8 6 5.9 T 2 2 2 6 5.9 C 8 8 8 32 T 2 2 2 32 Out of Reactor Spares C 11 13 13 T 3 3 6 a l C is standard Charpy V-notch impact bar. b T is 1/4-in. gauge diameter tensile specimen. D-3
NEDC-31166 R;v. 1 Table 5.3-3 l
SUMMARY
OF DAILY POWER HISTORY Operating Percent of Days Between Cycle Cycle Dates Days Full Power Cycles la 5/14/74 - 6/6/75 389 0.525 42 lb 7/19/75 - 2/13/76 210 0.686 61 2 4/15/76 - 3/12/77 332 0.695 62 3. 5/14/77 - 3/18/78 309 0.823 39 4a 4/27/78 - 6/17/78 52 0.700 265 4b 3/10/79 - 2/9/80 337 0.808 68 5 4/18/80 - 3/20/81 337 0.806 73 6 6/2/81 - 2/12/83 621 0.540 82 7 5/6/83 - 2/2/85 639 0.651 3226 0.671 (average) i l l D-4
Table 5.3-4 1 SURVEILLANCE CAPSULE LOCATION FLUX AND FLUENCE FOR IRRADIATION FROM 5/14/74 TO 2/2/85 Wire dps/g Element Reaction Rate Full Power Flux
- Fluence Wire Weight (at end of
[dps/ nucleus (n/cm -s) (n/cm ) 2 a (Element) (g) Irradiation (saturated)] >l &V >0.1 kV >l &V >0.1 kV -18 Copper 64713 0.3611 1.99x10 7.15x10 i 4 -18 Copper 64741 0.3588 1.87x10 6.75x10 -18 I Average = 6.95x10 2.6x10' 4.2x10' 4.9x10 7.8x10 4 El 5 -16 ? Iron 64713 0.0694 1.39x10 3.67x10 g Iron 64741 0.1663 1.37x10 3.62x10 h 5 -16 -16 9 j 4 Average = 3.65x10 2.3x10 6 -16 Nickel 64713 0.3234 1.84x10 4.04x10 0 -16 Nickel 64741 0.3211 1.93x10 4.24x10 -6 Average = 4.14x10 2.0x10' l 5
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NEDC-31166 R,v. 1 The full power flux calculated was 2.6x10' n/cm -s. The 8 integrated fluence (>l MeV) for the surveillance capsule was 17 2 determined to be 4.9x10 n/m. T h h s W a 20 m u g of 5% for dps/g (disintegrations per second per gram) 2 1 12% for dps/ nucleus (saturated) i 30% for flux and fluence >l MeV 2 40% for flux and fluence >0.1 MeV Unitradiated and irradiated Charpy specimens were tested at temperatures chosen to establish the fracture toughness transition curves for the base, weld and HAZ metals.. The RT NDT shift for each material was determined from the change in the 30 f t-lb energy index temperature. Experimental shif ts of 42*F for base metal and 9'F for HAZ metal were measured. The veld metal test results indicated no measu'rable shift. D.3 SECTION 5.3.2.1 The wording to section 5.3.2.1 in NEDC-30839 should be revised to read as follows: Estimated maximum changes in RT as a function of the NDT end-of-life (EOL) fluence at the 1/4 T depth of the vessel beltline materials are listed below. The updated predicted peak 18 2 EOL fluence at the 1/4 T depth of the beltline is 3.6x10 n/cm after 40 years of service (32 full power years). The E0L fluence prediction is based on a combination of experimental and analytical results. Flux at the 288* surveillance capsule location in the RPV was evaluated by testing flux wires removed with 'the surveillance capsule after Fuel Cycle 7. The relationship between the capsule location and the peak flux location at the 1/4 T depth was determined by a combination of two-dimensional and one-dimensional flux distribution computer analysis. D-6
NEDC-31166 R v. 1 The transition temperature shift due to irradiation was calculated with Regulatory Guide 1.99, Revision 1, taking into account the data from the surveillance testing. Specifically, the plate shift of 42*F exceeds the predicted shif t of 32*F by 31%. Therefore, a factor of 1.31 was applied to the predicted shift of the limiting beltline plate. The surveillance test i results for the weld showed no measurable shift, so the predicted shift was not modified. The results for the core beltline materials are tabulated below: Plate Weld Limiting Material Chemistry: 0.13% Cu, .0.03% Cu, 0.012% P 0.017% P EOL Transition Temperature Shift: 86*F 51*F Initial Reference Temperature: 40*F -50*F EOL Adjusted Reference Temperature: 126*F l'F Since the predicted E0L adjusted reference temperatures are below 200*F, provisions to permit thermal annealing of the RPV in accordance with Paragraph IV.B of 10CFR50 Appendix G are not required. D.4 SECTION 5.3.3.1 The third paragraph of section 5.3.3.1 refers to the EOL fluence. This reference should be revised to state the latest calculated EOL 18 fluence of 3.6x10 D-7/D-8 __}}