ML20035A768
| ML20035A768 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/28/1993 |
| From: | Caine T, Ranganath S, Sidhom A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20035A762 | List: |
| References | |
| DRF-B13-01661, DRF-B13-1661, GE-NE-523-165-1, GE-NE-523-165-1292, NUDOCS 9303290276 | |
| Download: ML20035A768 (98) | |
Text
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I GE-NE-523-165-1292 Revision 1 DRF B13-01661 February 1993 1
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MILLSTONE 1 NUCLEAR STATION VESSEL SURVEILLANCE MATERIALS TESTING RESULTS AND FRACTURE TOUGHNESS ANALYSIS l
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Prepared by:
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T. A. Caine, Proj ect Manager
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.RPV Integrity.
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Verified by:
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A. Sidhom, Engineer
.j RFV Integrity-Reviewed by:
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S. Ranganath,%anager Structural Mechanics Projects i
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1 GE Nuclear Energy l
9303290276 930322 l
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GE-NE-523-165-1292, Rev. 1 DRF B13-01661 t
IMPORTANT NOTICE REGARDING
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CONTENTS OF THIS REPORT 5
PLEASE READ CAREFULLY l
This report was prepared by General Electric solely for the use of the i
Northeast Utilities (NU).
The information contained in this report is believed by General Electric to be an accurate and true representation of the r
facts known, obtained or provided to General Electric at the time this report was prepared.
i The only undertakings of the General Electric Company respecting information in this document are contained. in the contract between the customer and General Electric Company, as identified in the purchase order for t
this report and nothing contained in this document shall be construed as j
changing the contract.
The use of this information by anyone other than the customer or for any purpose other than that for which it is intended, is not
[
authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
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GE-NE-523-165-1292, Rev. 1 DRF B13-01661 CONTENTS Page ABSTRACT viii l
i ACKNOWLEDGMENTS ix 1.
INTRODUCTION 1-1 s
2.
SUMMARY
OF RESULTS 2-1 I
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 Surveillance Plate 3-3 3.3.2 Surveillance Weld 3-4 t
4.
PEAK RPV FLUENCE EVALUATION 4-1 4.1 Flux Wire Analysis 4-1 4.1.1 Procedure 4-1 4.1.2 Results 4-2 4.2 Determination of Lead Factor 4-3 4.2.1 Procedure 4-3 4.2.2 Results 4-4 v
4.3 Estimate of 32 EFPY Fluence 4-5 5.
CHARPY V-NOTCH IMPACT TESTING 5-1 5.1 Impact Test Procedure 5-1 5.2 Impact Test Results 5-3 i
5.3 Irradiated Versus Unirradiated Charpy V-Notch Properties 5-4 5.4 Comparison to Predicted Irradiation Effects 5-4 5.4.1 Irradiation Snift 5-4 5.4.2 Decrease in USE 5-5 6.
TENSILE TESTING 6-1 l
6.1 Procedure 6-1 6.2 Results 6-2 6.3 Irradiated Versus Unirradiated Tensile Properties 6-3 7.
DEVEIDPMENT OF OPERATING LIMITS CURVES 7-1
7.1 Background
7-1 7.2 Non-Beltline Regions 7-1 7.2.1 Ncn-Beltline Monitoring During Pressure-Tests 7-2 7.3 Core Beltline Region 7-3 7.4 Closure Flange Region 7-4_
7.5 Core Critical Operation Requirements of 10CFR50 Appendix C 7-5 111 h
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
l CONTENTS (cont'd)
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7.6 Evaluation of Irradiation Effects 7-5 l
7.6.1 Surveillance CF Adjustment 7-6
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7.6.2 Application of CF Adjustments to Beltline Materials 7-7
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7.6.3 ART Versus EFPY 7-9 7 6.4 Upper Shelf Energy at 32 EFPY 7-9 l
7.7 Operating Limits curves valid to 32 EFPY 7-11 7.8 Reactor Operation Versus Operating Limits 7-11 3.
REFERENCES 8-1 1
APPENDICES A.
CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS A,
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BASIS FOR CONSERVATIVE RTNDT f
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DRF B13-01661 TABLES Table Title Eagt 3-1 Chemical Composition of RPV Beltline Materials 3-5 3-2 Mechanical Properties of Beltline and Other Selected 3-6 RPV Materials l
i 3-3 Chemical Composition of Irradiated Surveillance Specimens 3-7 L
3-4 Surveillance Specimen Identifications 3-8 4-1 Summary of Daily Power History 7 4-2 Surveillance Capsule Flux and Fluence for
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Irradiation from Start-up to 4/7/91 j
4-3 Flux at Full Power for Each Fuel Cycle 4-9 5-1 Qualification Test Results Using NIST Standard Reference 5-6 Specimens i
5-2 Charpy V-Notch Impact Test Results for Irradiated' 5-7
_i Surveillance Materials in Millstone 1 300* Capsule l
l 5-3 Significant Results of Irradiated and Unirradiated Charpy 5-8 i
0 V-Notch Data l
6-1 Tensile Test Results for Irradiated RPV Materials 6-4 6-2 Comparison of Unirradiated and Irradiated 6-5 j
Tensile Properties at Room Temperature i
7-1 Millstone P-T Curve Values
_7-12 7-2 Beltline Evaluation for Millstone at 32 EFPY 7-16 Including Surveillance Adjustment (SA) f
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7-3 Upper Shelf Energy Analysis for Millstone Beltline Materials 7-17 I
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GE-NE-523-165-1292, Rev. 1 i
DRF B13-01661 j
ILLUSTRATIONS Ficure Title Page l
3-1 Millstone 1 300* Surveillance Capsule 3-9 3-2 Schematic of the RPV Showing Identification 3-10 of Vessel Beltline Plates and Welds I
4-1 Schematic of Model for Two-Dimensional 4-10 i
Flux Distribution Analysis j
4-2 Relative Vessel Flux Variation with Angular Position 4-11 4-3 Relative Vessel Flux Variation with Elevation 4-12 t
5-1 Millstone 1 Base Metal Impact Energy
.5-9 5-2 Millstone 1 Weld Metal Impact Energy 5-10 j
5-3 Millstone 1 HAZ Metal Impact Energy 5-11 f
5-4 Millstone 1 Base Metal Lateral Expansion 5-12 5-5 Millstone 1 Weld Metal Lateral Expansion 5-13 5-6 Millstone 1 HAZ Metal Lateral Expansion 5-14 5-7 Plate Surveillance Data vs. i.99, Revision 2 ?rediction 5-15 5-8 Weld Surveille.nce Data vs. 1.99, Revision' 2 Predicion 5-16 6-1 Typical Engineering Stress-Strain for Irradiated 6-6 RPV Materials i
6-2 Irradiated Tensile Specimen Strength 6 ;
6-3 Irradiated Tensile Specimen Ductility 6-8 6-4 Fracture Location, Necking Behavior and Fracture 6-9 Appearance for Irradiated Base Metal Tensile Specimens 6-5 Fracture Location, Necking Behavior and Fracture 6-10 j
Appearance for Irradiated Weld Metal Tensile Specimens 6-6 Fracture Location, Necking Behavior and Fracture 6-11 Appearance for Irradiated HAZ Metal Tensile Specimens 7-1 P-T Curves for Pressure Tests 7-1B
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t 7-2' P-T Curves for Non-Nuclear Heatup/Cooldown 7-19
.i 7-3 P-T Curves for Core Critical Operation 7-20 vi n
.GE-NE-523 165-1292, Rev. 1 DRF B13-01661 ILLUSTRATIONS
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Firure Title Egge 7-4 Potential Benefit of Bottom Head Monitoring 7-21 7-5 ART for Limiting Beltline Material 7-22 7-6 Veld USE Prediction Adjusted for Surveillance Data -
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GE-NE-523-165-1292, Rev. 1 DRF B13-01661 ABSTRACT The second surveillance capsule was removed from the 300* location in the Millstone 1 reactor at the end of Fuel Cycle 13.
The capsule contained flux wires for neutron fluence measurement and Charpy and tensile test 4
specimens for material property evaluation.
Flux wire testing was used to establish the surveillance capsule fluence and vessel peak flux location and i
magnitude.
Charpy V-Notch impact testing and uniaxial tensile testing were I
performed to establish the properties of the irradiated surveillance materials.
These results, and those from the 210* capsule removed in 1984, were compared to unirradiated data to determine the shift in Charpy curves and decrease in upper shelf energy (USE) due to irradiation.
The shift results are within the (ARTNDT + 2a) predictions of Regulatory Guide 1.99, Revision 2 (1.99).
The measured decrease in the upper shelf energy (USE) of the plate is consistent with predictions.
The measured decrease in weld USE is slightly larger than that predicted by 1.99, but is within the upper bound of predicted USE decrease.
The dosimetry results, measured shift and USE decrease results from the two surveillance capsule tests were used, according to 1.99, in the analysis of adjusted reference temperatures and USE for the Millstone beltline I
materials.
Operating pressure-temperature (P-T) curves, per 10CFR50 Appendix C, are provided based on the updated ART values for up to 32 effective full power years of operation.
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GE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
i ACKNOWLEDGMENTS l
O The author gratefully acknowledges the efforts of the people mentioned below.
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Flux wire testing was performed by G.
C.
Martin, with lead factor computations by S. S. Wang and D. R. Rogers.
Charpy testing was completed by G. P. Wozadlo and G.
E. Dunning.
Tensile specimen testing was done by S.
B.
Wisner and G. H. Henderson, and chemical composition analysis was performed by C. R. Judd. The Charpy test evaluations were completed by R. G. Carey.
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GE-NE-523-165-1292, Rev. 1 g
1 DRF B13-01661 I
1.
INTRODUCTION 1
i 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 j
l reference temperature of nil-ductility transition (RTNDT) and the upper shelf T
energy (USE).
These are defined in 10CFR50 Appendix G [1] and in Appendix G d
of the ASME Boiler and Pressure Vessel Code,Section XI [2].
These documents I
i contain requirements used to establish the pressure-temperature (P-T)
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operating limits which must be met to assure ductile behavior.
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f Appendix H of 10CFR50 [3] and ASTM E185 [4] establish the methods to be
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j used for surveillance of the reactor vessel materials.
The first vessel l
l surveillance specimen capsule required by [3] was removed from Millstone l'in f
j late 1984.
Results of testing are summarized in a test report [5].
The l
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second capsule was removed at about 15 EFPY, as recommended in [4].
The j
capsule was sent to the GE Vallecitos Nuclear Center (VNC) for testing after i
e-j exposure to 13 fuel cycles of irradiation. The surveillance capsule contained flux vires for neutron flux monitoring and Charpy V-Notch impact test i
specimens and uniaxial tensile test specimens fabricated from materials in the vessel near the core (the beltline).
The impact and tensile specimens were tested to establish properties for the irradiated materials.
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The results of the surveillance specimen testing are presented in this
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report.
The irradiated material properties are compared to unirradiated I
specimen test results of the same materials to establish irradiation effects t
for comparison to predictions of Regulatory Guide 1.99, Revision 2 [6],
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t hereafter referred to as 1.99.
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1-1 1
GE-NE-523-165-1292 Rev. 1 l
DRF B13-01661 j
Operating limits curves for the vessel were presented in [5].
The f
I curves accounted for current requirements of [1] and [2].
Geometric
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discontinuities and highly stressed regions, such as the feedwater nozzles and j
the closure flanges, were evaluated separately from the core beltline region.
j The operating limits developed considered the most limiting conditions of the discontinuity regions and the beltline region, so as to bound all operating
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i conditions.
The operating limits developed for the beltline region in [5}
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included irradiation shift, based on Regulatory Guide 1.99, Revision 1 methods.
Those methods are modified here to use 1.99 and to account for the availability of two surveillance capsule test results.
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CE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
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SUMMARY
OF RESULTS t
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The 300* surveillance capsule was removed from the Millstone 1 vessel at I
the end of Fuel Cycle 13 and shipped to VNC.
The flux wires, Charpy V-Notch l
and tensile test specimens removed from the capsule were tested according to f
ASTM E185-82 [4].
The methods and results of testing and evaluation are i
presented in this report as follows-t a.
Section 3: Surveillance Program Background l
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Section 4: Peak RPV Fluence Evaluation l
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Section 5: Charpy V-Notch Impact Testing
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Section 6: Tensile Testing I
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Section 7: Development of Operating Limits Curves j
f Photographs of fractured Charpy specimens are in Appendix A.
The significant 1
results of the evaluation are below:
1 The second capsule was removed from the 300* azimuth position of j
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the reactor.
The capsule contained 6 flux vires:
2 copper (Cu),
2 iron (Fe), and 2 nickel (Ni).
There were 24 Charpy V-Notch specimens in the capsule:
8 each of plate material, weld material l
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and heat affected zone (HAZ) material.
The 6 tensile specimens t
a removed consisted of 2 plate, 2 weld and 2 HAZ metal specimens.
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I GE-NE-523-165-1292, Rev. 1 DRF B13-01661
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b.
The chemical compositions of the beltline materials were determined from available data from CE QA records as part of the work when the 210* capsule was tested [5]. In addition here, some other industry sources were used to estimate chemistry for velds
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where Millstone 1-specific data were not available.
The copper (Cu) and nickel (Ni) contents were determined for all heats of plate and weld material.
The values for the limiting beltline
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plate (Heat C1079-1) are 0.19% Cu and 0.51% Ni.
The limiting beltline weld (Heat V5214) values are 0.26% Cu and 1.2% Ni.
c.
Charpy and dropweight test results from the fabrication program materials certification testing were adjusted to be equivalent to test results done to current standards and the initial RTNDT values for locations of interest in the vessel were determined in
[5]. They are 26*F for the limiting beltline plate, -20*F for the limiting beltline veld, 26*F for the closure flange region and 4
40*F for the limiting nozzle and bottom head region.
d.
The flux wires were tested to determine the neutron flux at the surveillance capsule location.
The average measured fast flux 2
(>1.0 MeV) for the last 3 fuel cycles was 1.37x109 n/cm -sec.
Based on the flux wire data, the surveillance specimens received a 2 at removal.
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best estimate fluence of 6.6x1017 n/cm e
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The vessel inside surface lead factor was established using an analysis that combined two-dimensional and one-dimensional finite
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difference transport analysis.
The flux peak occurs at azimuthal l
location 294.5*, and about 102 inches above the bottom of active fuel elevation.
The lead factor for the surveillance capsule is 0.95 to the peak vessel inside surface location.
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GE-NE-523-165-1292, Rev. 1 i
DRF B13-01661 f.
The maximum accumulated neutron fluence at 32 EFPY was determined at the peak 1/4 T locations, using the flux vire test results and i
lead factor analysis results in Section 4, according to the methods of [6].
The maximum 1/4 T vessel 32 EFPY fluence is 1.1x1018 n/cm2 for the lower-intermediate shell and 6.0x1017 n/cm2 f
for the lower shell.
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The surveillance Charpy V-Notch specimens were impact tested at I
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 percentage shear.
Fracture surface photographs of each specimen are presented in Appendix A.
From absorbed energy and lateral 4
expansion results for the plate and weld materials the following values were calculated:
index temperatures for 30 ft-lb, 50 ft-lb, and 35-mil lateral expansion (MLE) values and USE.
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The plate and weld curves of irradiated specimen Charpy impact energy and lateral _ expansion were compared to data for unirradiated specimens to establish the 30 ft-lb index temperature j
irradiation shift and decrease in USE.
The surveillance plate material showed a 78'F shift and a 10% decrease in USE.
The weld I
material showed a 76*F shift and a 21% decrease in USE.
Based on these shift data, and those from [5), it was determined that 1.99 l
ARTNDT predictions were accurate for the weld, but were less than j
the measured plate shifts by about 60%.
This is, however, within I
the 2a margin of 1.99 uncertainty.
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The irradiated tensile specimens were tested at room temperature (70*F) and reactor operating temperature (550*F).
The results l
tabulated for each specimen include yield strength, ultimate j
l tensile strength, fracture strength and stress, uniform and total elongation, and reduction of area.
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GE-NE-523-165-1292, Rev. 1 DRF B13-01661 j.
The irradiated plate tensile test results were compared to unirradiated data from the vessel fabrication test program records.
The results generally showed increasing strength and decreasing ductility, consistent with expectations for irradiation embrittlement, i
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As a part of the development of the pressure-temperature (P-T) operating limits curves, the adjusted reference temperature (ART - initial RTNDT + ARTNDT + Margin) was predicted for each beltline material, based on the methods of 1.99, including the adjustment determined from the surveillance data.
Since surveillance data were factored inte the ART predictions, a Margin term of only la was included. This approach, allowed for in 1.99, j
Position 2.1, reduces the limiting 32 EFPY ART by about 18'F.
1.
The beltline material USE values at 32 EFPY were predicted using the methods in [6], with initial beltline USE values available for all but veld W5214, where a low estimate of USE for Linde 1092 f
flux welds was assumed.
A more conservative estimate of USE decrease was used for the welds, based on the surveillance weld i
test results.
The lowest beltline transverse plate and weld USE values were predicted to be 55 ft-lb and 71.5 ft-lb, respectively, 3
at 32 EFPY.
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P-T curves were developed for-three reactor conditions:
l hydrostatic pressure test (Curve A), non-nuclear heatup and j
cooldown (Curve B), and core critical operation (Curve C).
The curves are valid for 32 EFPY of operation, with intermediate EFPY curves provided for pressure tests.
The bolt preload and minimum permissible operating temperatures were determined to be 86*F.
l The P-T curves are shown in Figures 7-1 through 7-3.
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.i GE-NE-523-165-1292, Rev. 1 DRF B13-01661 3.
SURVEII.IANCE PROGRAM BACKGROUND 3.1 CAPSUI.E RECOVERY The Millstone 1 reactor was shut down in April, 1001 for refueling and maintenance.
The accumulated thermal power output was 1.09x107 mud or 14.8 EFPY.
The reactor pressure vessel (RPV) originally contained three surveillance capsules, at 120*, 210* and 300* azimuths at the core midplane.
The specimen capsules are held against the RPV inside surface by a spring loaded specimen holder.
Each capsule was designed to receive equal irradiation because of core syvnmetry.
During the 1991 outage, the capsule at 300* was removed. The capsule was cut from its holder assembly and shipped by l
cask to the GE Vallecitos Nuclear Center (VNC), where testing was performed.
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Upon arrival at VNC, the capsule was examined for identification.
The drilled hole code on the basket, shown in Figure 3-1, corresponded to the basket drawing, showing Reactor 15, which is GE's designation for Millstone 1, and Group 3, indicating the basket designed for the 300' location.
The capsule contained two Charpy specimen packets and three tensile specimen l
tubes.
Each tensile specimen tube contained two tensile specimens.
The Charpy specimen packets each contained a total of 12 Charpy specimens of
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plate, weld or heat affected zone (HAZ) and 3 flux wires (one iron, one copper and one nickel) in a sealed helium environment.
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3.2 RPV MATERIALS AND FABRICATION BACKGROUND l
3.2.1 Fabrication Historv j
The Millstone 1 vessel fabrication history was described in [5].
The identification of plates and welds in the beltline region is shown in Figure 3-2.
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GE-NE-523-165-1292, Rev. 1 DRF B13-01661 6
1 3.2.2 Material Properties of RPV at Fabrication l
h Material certification records were retrieved from GE Quality Assurance l
l (QA) records to determine chemical and mechanical properties of the vessel j
materials in [5).
However, only selected beltline materials were reported in
[5}. An addendum to [5), provided in June 1986, gave a comprehensive list of beltline materials (those materials which experience fluence greater than l
l 1x1017 n/cm2) and their available chemistry data, as shown in Table 3-1.
The recirculation nozzles are close enough to the bottom of the fuel to be j
considered beltline materials, which is unusual for BVRs.
The chemistry data j
for veld 34B009 reported in the Addendum was for the weld wire only, not the as-welded condition.
In Table 3-1, the chemistry for 34B009 is taken from the l
surveillance veld [7).
Copper content information was not available for the I
longitudinal welds, Heat W5214 Data from the Indian Point 2 response to
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Ceneric Letter 92-01, where the same heat of weld was tested for copper l
content, resulted in the value shown in Table 3-1.
l Results of certification mechanical property tests performed during RPV j
fabrication were examined, specifically Charpy V-Notch and dropweight impact l
i test results.
Properties of the beltline materials and other locations of l
i interest are presented in Table 3-2.
The Charpy and dropweight data collected l
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were used to establish the RTNDT values for each vessel component, using the methods described in [5).
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3.2.3 Specimen Chemical Composition i
Samples were taken from the surveillance plate and weld tensile specimens after they were tested.
Chemical analyses were performed using a l
Spectraspan III plasma emission spectrometer.
Each sample was dissolved in an l
acid solution to a concentration of 40 mg steel per al solution.
The spectrometer was calibrated for determination of Mn, Ni, Mo and Cu by diluting National Institute of Standards and Technology (NIST) Spectrometric Standard i
Solutions.
The phosphorus calibration involved analysis of five reference materials from NIST with known phosphorus levels.
Analysis accuracies are 10.003% (absolute) for phosphorus and 15% (relative) for other elements.
The l
chemical composition results are given in Table 3-3 for the surveillance plate I
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-GE-NE-523-165-1292, Rev. 1
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DRF B13-01661 and weld materials.
The results show good agreement with corresponding plate i
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[3,5,6], also shown in Table 3-3, except for the veld nickel content.
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nickel value varies widely, due to the fact that nickel was added as a separate wire in the submerged arc welding process.
Data on the surveillance weld vary locally from 0.08% to 1.78%.
The mean value, and that which will be assumed for analysis purposes here, is 1.05%.
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3.3 SPECIMEN DESCRIPTION l
The surveillance capsule contained 24 Charpy specimens: base metal (8),
veld metal (8), and HAZ (8).
There were 6 tensile specimens: base metal (2),
The capsule contained 6 flux vires:
2 iron, 2 nickel and 2 copper. The chemistry and fabrication history for the Charpy and tensile specimens were described in [5]. Some supplemental information on,the surveillance materials is provided in this section.
t 3.3.1 Surveillance Plate i
The description in [5] explained that the surveillance plate was taken from a beltline plate (Heat C1079-1), and that the plate experienced heat treatment comparable to that of the beltline plates.
Specimens were cut from the 1/4 T and 3/4 T with a longitudinal orientation.
The specimen chemical composition results in Table 3-3 correspond well with the certified material test report (CMTR) data, also shown in Table 3-3.
The piece of plate used for the surveillance specimens was also used for the fabrication test program l
specimens, where Charpy curves vere generated for both surfaces, the 1/4 T, 3/4 T and center thicknesses.
Thus, the fabrication test program data provide j
a good baseline for the surveillance plate specimens.
The specimen identifications for the plate Charpy and tensile specimens j
are given in Table 3-4.
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3.3.2 Surveillance Veld In [5], it was reported that the heat of weld wire used for the surveillance weld was W5214, because the surveillance weld fabrication procedure specified that the surveillance veld be identical to a longitudinal i
t veld.
Subsequent evaluation, resulting in the 1985 Addendum to [5), showed by j
r examination of the shop travelers that the surveillance weld was made sith the
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beltline circumferential weld heat 34B009.
The identifications of the weld and HAZ Charpy and tensile specimens removed from the surveillance capsule are shown in Table 3-4.
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Table 3-1 CllEMICAL COMPOSITION OF RPV BELTLINE MATERIALS Composition by Weicht Percent Identification lleat/ Lot No.
C Mn P
S Si
. Ni Mo Q_y_.
Lower Shell Plates:
G2001-1R C1339-1 0.23 1.32 0.011 0.027 0.25 0.49 0.46 0.22 G2001 3 B4928-1 0.22 1.39 0.010 0.020 0.26 0.52 0.49 0.23 C2001 5 C1140 2 0.22 1.36 0.010 0.018 0.26 0.44 0.48 0.23 lower. Intermediate Shell Plates:
G2002 4 B5013 2 0.23 1.39 0.011 0.023 0.23 0.49 0.48 0.21 G2002-5 C1079 1 0.22 1.31 0.008 0.025 0.22 0.51 0.49 0.19 G2002 6 C1140 1 0.22 1.37 0.010 0.020 0.25 0.45 0.45 0.21 Surveillance Plate:
G2006 C1079 1 see Table 3-3 Lower Longitudinal Welds:
w 2-073 A,B C W5214, Linde 1092 0.077 1.05 0.021 0.012 0.26 1.2 0.50 0.26" E
Flux lot 3617 Lower-Intermediate Longitudinal Welds:
1 073 A,B,C V5214, Linde 1092 see above (Tandem)
Flux Imt 3617 Lower to Lower-Intermediate Girth Vgid:
3 073 34B009, Linde 1092 0.11 1.34 0.016 0.016 0.28 1.03 0.49 0.18 Flux Lot 3708 8
surveillance Veld:
34B009, Linde 1092 see Table 3 3 7
Flux Lot 3708 Recirculation Inlet Nozzles:
E G2012 EV.7980 0.23 0.62 0.008 0.014 0.31 0.63 0.61 0.22" oY yR Data, for W5214 with a separate nickel wire, presented in Indian Point 2 response to CL 92 01.
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Chemistry for.this weld based on surveillance weld test results in (7}.
E Upper bound of available Cu data for nozzle forgings of that vintage.
6E 3-5 E.*
I GE-NE-523-165-1292 Rev. 1 i
DRF B13-01661 Table 3-2.
f MECHANICAL PROPERTIES OF BELTLINE AND OTHER SELECTED RPV MATERIALS Test Charpy Ident.
Heat Temp.
~
Location Number Number l'El (ft-lb)
(*F)
(*F)
(*F)
{
Beltline:
(
Lower Shell Plates G2001-1R C1359-1 10 37,47,51
-10 6
6 i
C2001-3 B4928-1 40 61,61,66 10 10 10 i
G2001-5 C1140-2 40 44,49,57
-10 22 22
(
t Lower Intermediate G2002-4 B5013-2 10 42,47,59
-30
-4
-4
[
Shell Plates G2002-5 C1079-1 10 27,29,35
-30 26 26 i
G2002-6 G1140-1 10 30,38,53
-20 20 20 l
Longitudinal Welds 1-073 W5214,3617 10 35,39,48 N/A 20
-20 2-073 W5214,3617
.l Girth Weld 3-073 34B009,3708 10 71,84,90 N/A
-50
-50 Recirc Inlet Nozzle G2012 EV7976 40 44,62,73 N/A 22 40 Non-Beltline:
f
(
Upper Shell Plate C2001-2 A0041-1 10 27,33,51
-10 26 26 Vessel Flange G2009 3P0870 10 96,115,123 0
-20 0
1 Head Flange G2010 3P0870 10 69,82,108 10
-20 10 Top Head Torus G2007-1 B4688-1B 10 62,64,70
-10
-20
-10 i
i Bottom Head Torus G2004-1 A0010-1 40 35,44,67-
- 10 40 40-l Closure Bolts G2070 21385 10 44,46,50 N/A LST - 70*F I
I NOTE: N/A - not available 3-6
{
e I
i
O O
O Table 3-3 CHEMICAL COMPOSITION OF IRRADIATED SURVEILLANCE SPECIMENS Composition by Velzht Percent Identification C
Mn P
S Si Ni Mo
_ Q1_
Plate 1.32 0.009 0.50 0.48 0.22 CE4 CDT 1.31 0.011 0.48 0.46 0.21 C1079-1 (Table 3-1) 0.22 1.31 0.008 0.025 0.22 0.51 0.49 0.19 C1079-1 [8]
0.23 1.37 0.017 0.028 0.22 0.52 0.39 0.20 Values Assumed for 0.23 1.32 0.010 0.028 0.22 0.49 0.47 0.21 48 Surveillance Plate u
Weld:
CKT 1.29 0.019 0.59 0.56 0.20 1.09 0.55 0.20 1.28 0.019 CKC 0.17 0.99 0.53 0.19 C4A (Charpy from [5])
1.27 348009 ([7],10 tests) 0.11 1.34 0.016 0.016 0.28 1.03 0.49 0.18 h
M 34B009 [8]
0.09 1.28 0.017 0.016 0.21 1.78 0.40 0.16 f,
0:
Values Assumed for 0.11 1.28 0.019 0.016 0.17 1.05 0.55 0.20 2.
Surveillance Weld g;
Note:
Element measurements are 10.003% absolute for P and 15% relative for other elements.
03 f3 70 a A " " mark denotes an element that was not evaluated.
SQ 8*
3-7
_...., ~
k GE-NE-523-165-1292, Rev. 1 DRF B13-01661-Table 3-4 SURVEILLANCE SPECIMEN IDENTIFICATIONS Irradiated Irradiated Irradiated Base Metal
'w' eld Metal HAZ Metal CHARPY SPECIMENS:
i B34 BK7 C4K B36 BKA C4Y I
B37 BKJ C4L B3J BKB C51 f
B3L BKK C4M B3A BKC C4P
^
B3C BKD C4T B3E BKE C4U i
O TENSILE SPECIMENS:
i 4
i I
h r
r i
I 3-8
i GE-NE-523-165-1292, Rev. I l
l DRF B13-01661 I
i li i
1 t
l C
)
u,,
rp mxn 1
l i
MT C
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d 1
a
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i O
i
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(
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l 3-9 l
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I
GE-NE-523-165-1292, Rev. 1 DRF B13-01661
\\
/ Top Head Enclosure i
'l Top IIead Flange Vessel Flange PI e
~
(9 i
Upper Shell
?
{
Upper-Int Shell
?
Ia"f";'
///}
/f ?
f k
}N Plate Heats B5013-2 N 1 -073 A.B.C C1079-1 g
C1140-1 Core Lower Intermediate Shell Beltline f
y Region 3-073 7
I
.....f.
i:
C
@@ )
Plate Heats C1359-1 Lower I
B4928-1 C1140-2 Shell fN 2-073 A,B,C @ {
Bottom Head Enclosure
\\
/
/
I O
/
Figure 3-2.
Schernatic of the RPV Showing
(..
Identification of Vessel Beltline Plates and Welds 3-10
r GE-NE-523-165-1292, Rav. 1 DRF B13-01661
{
t 4
PEAK RPV FLUENCE EVALUATION Flux wires removed from the 300* capsule were analyzed to determine flux and fluence received by the surveillance capsule.
The lead factor from {S) was used to establish the peak vessel fluence from the flux wire results.
t 4.1 FLUX WIRE ANALYSIS i
4.1.1 Procedure t
The surveillance capsule contained 6 flux wires: 2 iron, 2 copper and l
2 nickel.
Each wire was removed from the capsule, cleaned with dilute acid, I
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.-cm l
or 10-cm source-to-detector distance with 100-ce and 80-ce Ge(Li) detector systems.
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 reactor average power fraction are shown in Table 4-1.
Zero power days between fuel cycles are listed as well. Given the coastdown at the end of the last fuel cycle, Cycle 13 was divided into two periods, in order to improve the predictions of the iron and nickel wires with t
their shorter half-lives.
To further improve the prediction of flux from the wire activities, the exposures of the bundles closest to the surveillance capsule were compared to the core average exposure for each cycle.
The resulting ratio, shown as local power fraction in Table 4-3, reflects the fact that, after the first fuel cycle, the bundles on the periphery of the core are at lower power than the core average.
These peripheral bundles are the primary source of fluence at the vessel wall.
(L 4-1 i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 From the flux wire activity measurements and power history,_ reaction rates for Fe-54 (n p) Mn-54, Cu-63 (n,o) Co-60 and Ni-58 (n,p) Co-58 vere calculated. The >l MeV fast flux reaction cross sections for the iron, copper r
and nickel wires were estimated to be 0.194 barn, 0.00341 barn and 0.252 barn, respectively.
These values were obtained from measured cross section functions determined at GE's Vallecitos Nuclear C' enter from more than 65 I
spectral determinations for BWRs and for the General Electric Test Reactor using activation monitors and spectral unfolding techniques.
These data functions are applied to BWR pressure vessel locations based on water gap r
(fuel to vessel wall) distances.
The cross sections for >0.1 MeV flux were determined from the measured 1-to-0.1 MeV cross section ratio of 1.6.
4.1.2 Results The measured activity, reaction rate and normalized full-power flux i
results for the 300* surveillance capsule are given in Table 4-2.
The >l MeV l
flux values, shown in Table 4-3, were calculated by dividing the copper wire reaction rate measurements by the copper cross section, factoring in the local power history for each fuel cycle.
The copper wire results were used because the copper wires, with their longer half-life (5.27 years) do not reach j
saturation, and are not as dependent on the power history specifics.
The l
agreement of the iron flux wire results in Table 4-2 show that the power history is accurate. The fluence result. 6.6x1017 n/cm2 (>l MeV) was obtained 7
by multiplying the full-power flux value for each cycle by its operating time and full power fraction, shown in Table 4-1, and adding the result for each cycle to get the total.
The accuracies of the values in Tables 4-2 for a 2a deviation are I
estimated to be:
5% for dps/g (disintegrations per second per gram) i 10% for dps/ nucleus (saturated) 25% for flux and fluence >l MeV 35% for flux and fluence >0.1 MeV i
I 1
4-2 e
a e
l GE-NE-523-165-1292, Rev. 1 I
DRF B13-01661 l
Flux vires from the Millstone 1 210' capsule were evaluated by General.
I
(
Electric in 1984 [5].
Those test results were reanalyzed using the same localized power history approach described above.
The 210' capsule fluence l
17 n/cm2 The resulting >l MeV flux values, also l
changed slightly, to-3.9x10 shown in Table 4-3, are about the same as the 300* values.
The flux wire results at 300* are used in Section 4.3 to predict the fluence at 32 EFPY.
The prediction involves extrapolating the fluence based on the full power fluxes for the last three fuel cycles, which have similar values.
j 4.2 DETERMINATION OF LEAD FACTOR j
i The flux wires detect flux at a single location.
The wires will therefore reflect the power fluctuations associated with the operation of the plant.
However, the flux wires are not necessarily at the location of peak
^
vessel flux.
A lead factor is required to relate the flux at the wires' location to the peak flux.
The lead factor is a function of the core and vessel geometry and of the distribution of bundles in the core.
The lead j
I factor was generated for the Unit 1 geometry, using a typical fuel cycle to determine power shape and void distribution.
The methods used to epiculate the lead factor are discussed below.
[
4.2.1 Procedure i
3 Determination of the lead factor for-the RPV inside wall was made using i
a combination of two separate two-dimensional finite difference computer analyses.
The first of these established the relative azimuthal variation of l
fluence at the vessel surface and 1/4 T depth. The second analysis determined the relative variation of flux with elevation.
The azimuthal and axial I
distribution results were combined to provide the ratio of flux, or the lead l
factor, between the surveillance capsule location and the peak flux locations.
}
l The DORT computer program, which utilizes the discrete ordinates method j
f to solve the Boltzmann transport equation in two dimensions, was used to calculate the spatial flux distribution produced by a fixed source of neutrons j
in the core region. The azimuthal distribution was obtained with a model 4-3 i
i f
-t
GE-NE-523-165-1292., kev. 1 DRF B13-01661 specified in (R,#) geometry, assuming eighth-core symmetry with reflective boundary conditions at 270* and 315*.
Calculations were performed using
\\
neutron cross-sections from a 26 energy group set, with angular dependence of the scattering cross-sections approximated by a third-order Legendre polynomial expansion.
A schematic of the (R,#) vessel model is shown in Figure 4-1.
A totai of 123 radial intervals and 45 azimuthal intervals were used.
The model consists of an inner and outer core region, the shroud, water regions insida and outside the shroud, and the vessel wall.
The core region material compositions and neutron source densities were representative of conditions near the elevation of the wires. Flux as a function of azimuth and radius was calculated, establishing the azimuth of the peak flux and its magnitude relative to the flux at the wires' location of 300*.
The calculation of the axial flux distribution was performed in (R,Z) geometry, using a simplified cylindrical representation of the core configuration and realistic simulations of the axial variations of power density and coolant mass density. The core description was based on conditions
)
near the azimuth angle of 298* where the edge of the core is closest to the vessel wall.
The elevation of the peak flux was determined, as well as its magnitude relative to the flux at the surveillance capsule elevation.
a 4.2.2 Results t
The two-dimensional computations indicate the flux to be a maximum 24.5*
past the RPV quadrant references (0",
90*, etc.), at an elevation about 102 inches above the bottom of active fuel. The peak closest to the 300* location of the surveillance capsule removed is at 294.5*, as shown in Fi ure 4-2.
The 5
relative flux distribution versus elevation is shown in Figure 4-3.
The 3
distribution calculations establish the lead factor between the surveillance capsule location and the peak location at the inner vessel wall.
The lead factor is 0.95.
The transport calculation of surveillance capsule flux, 1.26x109 n/cm2, agrees quite closely with the dosimetry result for the last 9
cycle of 1.3x10 n/cm2 4-4 i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 r
t Due to the flux distribution with elevation, the lower shell plates and welds will experience a lower fluence.
The top of the lower shell is about
-j 31 inches above the bottom of active fuel.
As shown in Figure 4-3, the relative flux'at that location is 60% of the peak value.
The lower shell i
fluences are adjusted accordingly.
i l
The fracture toughness analysis is based on a 1/4 T depth flaw in the l
beltline region, so the attenuation of the flux to that depth is considered.
This attenuation is calculated according to 1.99 requirements, as shown in the j
next section.
t 4.3 ESTIMATE OF 32 EFPY FLUENCE The inside surface fluence (fsurf) at 32 EFPY is determined from the flux wire fluence for 14.8 EFPY of 6.6x1017 n/cm2 and the best estimate of _the
[
full power flux, averaged for cycles 11 through 13 in Table 4-3, using the lead factor of 0.95.
The time period 32 EFPY, typically assumed for 40-year operation (80% capacity factor) is 1.01x109' seconds.
The average full power.
flux for cycles 11 through 13 is 1.37x109 n/cm2-s.
The resulting 32 EFPY fluence value at~the vessel inside surface is:
i
-i 9
9 17 + 1.37x10 *1.01x10 *(32-14.8)/32)/0.95 l
fsurf - (6.6x10 fsurf - 1.5x10 n/cm at the peak location.
l 18 2
17 2 at the lower shell.
fsurf - 8.9x10 n/cm The 1/4 T fluence (f) is calculated according to the following equation from [6]:
I f-fsurf(e-0.24x)
(4 1) 3 I
where x -
distance, in inches, to the 1/4 T depth.
l
[
t i
4-5 l
t
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 The vessel' beltline consists of the lower-intermediate shell and the lower shell, with thicknesses of 5.50 and 6.50 inches, respectively.
The l
corresponding depths x are 1.38 and 1.63 inches.
Equation 4-1 evaluated for these values of x gives:
i 18 2
Lower-Int.
f - 0.7181 fsurf, or f - 1.1x10 n/cm 17 2
Lower f - 0.6762 fsurf, or f - 6.0x10 n/cm i
\\
t t
I t
l l
I u
i 4-6 t
'i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
Table 4-1
SUMMARY
OF DAILY POWER HISTORY i
Operating Full Power Days Between Cycle Cvele Dates Days Fraction Cveles 1
12/1/70 - 8/29/72 638 0.702 l
332 2
7/28/73 - 8/30/74 399 0.774 65 3
11/4/74 - 9/12/75 313 0.808 38 4
10/21/75 - 10/1/76 347 0.768 62 5
12/3/76 - 3/10/78 463 0.847 36 6
4/16/78 - 4/27/79 377 0.881 61 7
6/28/79 - 10/4/80 465 0.783 i
255 l
8 6/17/81 - 9/11/82 452 0.933 66 9
11/17/82 - 4/13/84 514 0.941 75 10 6/28/84 - 10/25/85 485 0.954
,1 58 11 12/23/85 - 6/5/87 530 0.932 i
69 12 8/14/87 - 4/7/89 603 0.952 48 13a 5/26/89 - 2/28/91 644 0.901 0
13b 3/1/91 - 4/7/91 38 0.723 6268 0.863 (average) 4-7
[
O O
O Table 4-2 SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM START-UP TO 4/7/91 Full Powgr Flux
- Fluenge Wire dps/g Element Reaction Rate Wire Weight (at end of
[dps/ nucleus (n/cm -s)
(n/cm )
(Element)
(r)
Irradiation)
(saturated)1
>1 MeV
>0.1 MeV
>l MeV
>0.1 MeV 8
Copper 65329 0.4622 2.26x10 7.03x10 8
Copper 65330 0.4141 2.18x10 6.77x10 Average - 6.90x10 2.0x10' 3.2x10' 6.6x10 1.1x10
-18
-16 Iron 65329 0.1809 1.41x10 3.92x10 -16 Iron 65330 0.1725 1.37x10 3.80x10
-16 1
18 Average - 3.86x10 2.0x10 3.2x10' 6.5x10 1.0x10 En 0
-16 Nickel 65329 0.3047 1.78x10 4.58x10 6
-16 Nickel 65330 0.2976 1.80x10 4.64x10 1
Average - 4.61x10' 1.8x10' 2.9x10' 6.0x10 9.6x10 r9 b=
Y-5:
BE Full power flux applicable to Cycle 1, based on thermal power of 2011 MW,
7 and considering localized power history.
Y:n Se:
g;-
~"
4-8
~. -..,.
t i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 Table 4-3 FLUX AT FULL POWER FOR EACH FUEL CYCLE j
v J
300* Local 300* Cycle 210' Local 210' Cycle Power Full tc:*er Power Full Power i
Cycle Fraction Flux (n/cm2-si Fraction Flux (n/cm2.s) 1 1.00 2.0x109 1.00 1.9x109 i
t 2
0.85 1.7x109 0.
s 1.6x109 I
?
3 0.79 1.6x109 0.77 1.5x109 f
4 0.86 1.7x109 0.87 1.6x109 j
t 5
0.78 1.6x109 0.80 1.5x109
~
f l
[
6 0.66 1.3x109 0.62 1.2x109
()
7 0.62 1.2x109 0.60 1.1x109 j
8 0.55 1.1x109 0.54 1.0x109 9
0.53 1.1x109 0.53 1.0x109 i
i i
10 0.55 1.1x109 0.55 1.1x109 l
l 11 0.70 1.4x109 0.70 1.3x109 j
i 12 0.71 1.4x109 0.70 1.3x109 I
13 0.66 1.3x109 0.65 1.2x109 I
1 4-9 I
.i
O O
O 315 deg.
l r
1 1 = CORE INTERIOR FUEL 45 AZIMUTHAL 2 = CORE EXTERIOR FUEL INTERVALS 2
2 h
2 2
2 2
2 2
2 1
1 a
/1 g
CAPSULE 2
2 2
2 1
1 h
/
1 1
2 2
1 1
1 h
/1 2
2 1
1 1
1 1
t h
1/
2 2
1 1
1 1
1 1
1 h
2 2
1 1
f 1
1 1
1 1
1
\\
)
Y 2
i '
270 deg.
Jy\\
fi 8g' g
_ REFLECTIVE o
WATER REGION CORE EXTERIOR
=
35 INTERVALS AND WATER REGION BOUNDARY m
SO INTERVALS g
m E
w CORE
.w VESSEL WALL SHROUD INTERIOR 17 INTERVALS 8 INTERVALS 13 INTERVALS Sh-e.~
"x Figure 1-1.
Schematic of Model for Two-Dimensional Flux Distribution Analysis
'a S-
r nU 1.2 1
5
\\
A 5.8 c
8
?
.6 z
.$li
/
E 'A J
.2 m
-10 0
10 20 30 40 50 Azimuthol Location (degrees)
{
Y E
e, T A5 Figure 4-2.
Relative Vessel Flux Variation with Angular Position 5~
I.I e-
O O
O 1.2 l
1 h
X x?8
~
o>
_o 5.6 aT
.4 f
u 20 40 60 80 100 120 140 h
Distance Above Bottorn of Active Fuel (inches)
~
of Figure 4-3.. Relative Vessel Flux Variation with Elevation ku
$~
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 5.
CHARPY V-NOTCH IMPACT TESTING O'
The 24 Charpy specimens recovered from the surveillance capsule were impact tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials.
Unirradiated surveillance plate and weld metal specimens have been tested previously, and were reported in [5).
Testing of the surveillance capsule Charpy specimens was conducted in accordance with ASTM E23-88 [9].
5.1 IMPACT TEST PROCEDURE The Vallecitos testing machine used for irradiated specimens 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 ft-lb.
The test apparatus and operator were qualified using NIST standard reference material specimens.
The standards consist of sets of high and l
low energy specimens, each designed to fail at a specified energy at the O
standard test temperature of -40'F.
According to [9), the test apparatus averaEed results must reproduce the NIST standard values within an accuracy of 5% or 1.0 f t-lb, whichever is greater.
The qualification of the Riehle machine and operator is summarized in Table 5-1.
The calibration results for the high energy specimens averaged 74.4 ft-lb, compared to the specified energy range of 64.9 to 71.7 f t-lb.
The Riehle result is 8.9%
higher, which exceed the +5% range specified in [9] for the calibration.
However, the low energy specimens were within 0.3 f t-lb, or 2.5% of the specified value.
If the Riehle machine had a problem, the low energy specimen results would also be off.
t
[
5-1 t
m i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 j
r NIST personnel were contacted to check the reliability of the MM16 l
series of high energy specimens used here.
It was determined that 80% of the machines reporting calibration results with the MM16 specimens had also shown results above 71.7 ft-lb.
Since much more than 50% of the results vere above 71.7 ft-lb, it was concluded that the specified energy level for the MM16 specimens was too low.
Based on the successful calibration results with the low energy specimens, testing proceeded.
New calibration specimens were received from NIST in February 1993.
These were tested just prior to issuance of this report, with the Riehle machine in the identical configuration it was in when the surveillance specimens were tested.
The new calibration results are also shown in Table 5-1, and are acceptable.
l Charpy V-Notch tests were conducted at temperatures between -20*F and-300*F.
For tests below 70*F methanol was used as the cooling fluid.
i Between 70*F and 212*F, water was used as the temperature conditioning j
fluid.
The specimens were heated in oil above 212*F.
Cooling of the conditioning fluids was done with liquid nitrogen, and heating by an immersion heater. The fluids were mechanically stirred to maintain uniform temperatures.
The fluid temperature was measured with a calibrated thermocouple.
Once at test temperature, the specimens were manually transferred with centering tongs to the Charpy test machine and impacted within 5 seconds.
For each Charpy V-Notch specimen the test temperature, energy absorbed, lateral expansion, and percent shear were evaluated.
In addition, photographs were taken of the fracture surfaces.
Lateral expansion and percent shear were measured according t.
methods specified in i
[9;.
Percent shear was determined with method one of Subsection 11.2.4.3 of [9], which involves measuring the length and width of the cleavage j
l surface and determining the percent shear value from Tables 1 or 2 of [9).
i 1
l L
i e
5-2 e--.
7 i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 5.2 IMPACT TEST RESULTS
{
Eight Charpy V-Notch specimens each of irradiated base, weld, and HAZ material were tested at temperatures (-20*F to 300*F) selected to define the toughness transition and upper shelf portions of the fracture toughness curves. The ab' sorbed energy, lateral expansion, and percent shear data are listed for each material in Table 5-2.
Plots of absorbed energy data for base, weld and HAZ materials are presented in Figures 5-1 through 5-3, respectively. Lateral expansion plots for base, weld and HAZ materials are presented in Figures 5-4 through 5-6, respectively.
The fracture surface photographs and a summary of the test results for each specimen are contained in Appendix A.
r Unirradiated specimens taken from the surveillance plate and weld were tested previously in other work [10,7).
Plots of absorbed energy data for-t i
base and weld materials are included in Figures 5-1 and 5-2, respectively, along with the corresponding irradiated specimen data.
Lateral expansion plots for the unirradiated base and weld materials are included in Figures 5-4 and 5-5, respectively, along with their corresponding irradiated specimen data.
Also included in these figures are the curves for irradiated data from [5], replotted using the hyperbolic tangent fit t
described below.
This allows for consistent comparisons of the curve shifts of the two sets of irradiated surveillance data.
The plate and weld data sets are fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook [11):
Y - A + B
- TANH [( T - To )/C),
f where Y - impact energy or lateral expansion T - test temperature, and A, B To and C are determined by non-linear j
i regression.
f The TANH function is one of the few continuous functions with a shape characteristic of low alloy steel fracture toughness transition curves.
5-3 I
I
i i
CE-NE-523-165-1292, Rev. 1 DRF B13-01661 l
5.3 IRRADIATED VERSUS UNIRRADIATED CHARPY V-NOTCH PROPERTIES O
The irradiated and unirradiated Charpy V-Notch data were used to
(
estimate the values given in Table 5-3:
30 ft-lb, 50 ft-lb and 35 MLE
^
index temperatures, and the USE for both sets of base and weld metal irradiated material data and for the base and weld metal unirradiated material data.
Transition temperature shift values are determined as the j
change in the temperature at which 30 ft-lb impact energy is achieved. as j
required in [4].
The resulting shifts in Charpy curves and decreases in l
USE are discussed in the next section.
l
{
5.4 COMPARISON TO PR.EDICTED IRRADIATION EFFECTS l
f 5.4.1 Irradiation Shift i
The measured transition tempe:ature shifts for the surveillance l
materials were compared to the predictions calculated according to 1.99.
f l
The inputs for the surveillance materials are as follows:
Plate:
Copper:
0.21% (based on Table 3-3 values)
Nickel:
0.49% (based on Table 3-3 values) 1.99 CF:
140.7
{
fluence:
3.9 x 1017 n/cm2 (210* capsule) fluence:
6.6 x 1017 n/cm2 (300* capsule) j Weld:
Copper:
0.20% (based on Table 3-3 values)
Nickel:
1.05% (based on Table 3-3 values) 1.99 CF:
228.5 f
fluence:
3.9 x 1017 n/cm2 (210* capsule) i fluence:
6.6 x 1017 n/cm2 (300* capsule)
{
r CF shown above is the chemistry factors from Tables 1 or 2 of 1.99.
The fluence factors, calculated according to Equation (2) in 1.99, are 0.2553 l
for the 21D* capsule and 0.3389 for the 300* capsule. Margins are 34*F for i
plate and 56*F for weld, for 2a uncertainty. The predicted values of l
l
?
l 5-4 l
GE-NE-523-165-1292, Rev. 1 DRF.B13-01661 l
1 i
ARTNDT and (ARTNDT + Margin) are compared to the actual 30 ft-lb shifts in Table 5-3.
These comparisons are shown graphically for the plate in Figure
{
5-7 and for the weld in Figure 5-8.
r 1
5.4.2 Decrease in USE The measured decreases in USE are compared to predictions calculated according to 1.99.
Using the copper and fluence data above with Figure 2
.i of 1.99, predicted USE decreases for the plate and weld are determined, as shown in Table 5-3.
The measured values of USE decrease in Table 5-3 for the plate show that the 1.99 predictions are conservative.
However, for the weld, the measured values are slightly higher than the 1.99 predictions.
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i GE-NE-523-165-1292, Rev. 1 l
DRF B13-01661 Table 5-1 O
1 QUALIFICATION TEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS i
Test Energy Acceptable Specimen Bath Temperature Absorbed Range Identification Medium
(*F)
(ft-lb)
(ft-lb)
Vallecitos Riehle Machine (tested 8/24/92):
MM-16 026 Methanol
-40 69.0 l
4 MM-16 335
-40 75.0 MM-16 629
-40 77.5 MM-16 738
-40 73.0 MM-169992
-41 77.5 l
Average 74.4 68.3 3.4 LL-19 029 Methanol
-40 12.0 LL-19 771
-40 12.5 LL-19 834
-41 12.0 LL-191074
-40 12.0 i
LL-191106
-40 1221 Average 12.2 11.9 + 1.0 Vallecitos Riehle Machine (tested 2/17/93):
HH-40 321 Methanol
-40 74.0 HH-40 507
-40 74.5 HH-40 933
-40 71.5 HH-40 1042
-40 75.0 88-40 1198
-40 75.0.
Average 74.0 74.9 3.7 LL-39 087 Methanol
-40 13.5 LL-39 195
-40 13.5 LL-39 793
-40 13.0 LL-39 849
-40 13.5 LL-39 1102
-40 13.1 2
Average 13.4 13.2 1.0 I
I 5-6 b
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GE-NE-523-165-1292, Rev. 1 DRF B13-01661
{
Table 5-2 CHARPY V-NOTCH IMPACT TEST RESULTS FOR I
IRRADIATED SURVEILLANCE MATERIALS IN MILLSTONE 1 300* CAPSULE l
Test Fracture Lateral Percent Shear Specimen Temperature Energy Expansion (Method 1) l Identification
(*F)
(ft-lb)
(mils)
(%)
1 Bike:
B34 20 9.5 7.5 13 B36 60 17.5 18 16 i
B37 100 24.5 23 40 B3J 120 59 37 52 l
B3L 140 47 43 60 i
B3A 150 61 49 59 B3C 200 80.5 63.5 94 B3E 300 100 78 100 l
Weld:
BK7
-20 19 13 18 BKA 20 34 27 38 i
BKJ 30 29 23 35 BKB 60 39 29 4d BKK 100 51.5 47 64 BKC 150 79.5 69 96 BKD 200 84 69 100 BKE 300 87 75 100 l
l Bal-C4K 20 23.5 20.5 28 C4Y 40 40 32 36 C4L 60 71.5 49 53 CSI 80 61 55 61 CAM 100 55 45
'68 C4P 150 102.5 74 100 C4T 200 92.5 73 99 C4U 300 110 77 100
[
6 F
5-7 l
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CE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
Table 5-3 l
SIGNIFICANT RESULTS OF IRRADIATED AND UNIRRADIATED CHARPY V-NOTCH DATA i
Upper Shelfa Index Temperature (*F)
Energy (ft-lb)
Material E-30 ft-lb E-50 ft-lb MLE-35 mil L/T 300* CAPSULE:
Unirradiated Plate 15 57 37 112/ 73 Irradiated Plate 93 112 129 101/ 66 Difference 78 75 83 11/ 7 (10%)
I 1.99, Rev 2 ARTndt:
48 1.99 Rev 2 % Decrease:
(16%)'
1.99, Rev 2 (A+20):
82 Unirradiated Weld
-45
-23
-28 111 Irradiated Weld
.J_1 88
_69 88 Difference 76 111 97 23 (21%)
1.99, Rev 2 ARTndt:
77 1.99 Rev 2 % Decrease:
(18%)
1.99, Rev 2 (A+20):
133 210* CAPSULE:
Unirradiated Plate 15 57 37 112/ 73 Irradiated Plate 76 114 96
_99/ 64 l
Difference 61 57 59 13/ 9 (12%)
i 1.99, Rev 2 ARTndt:
36 1.99 Rev 2 % Decrease:
(14%)
i 1.99, Rev 2 (A+20):
70 Unirradiated Weld
-45
-23
-28 111 Irradiated Weld
.21 4_Z 21 193 Difference 22 65 51 3
(3%)
1.99, Rev 2 ARTndt:
58 1.99 Rev 2 % Decrease:
(16%)
1.99, Rev 2 (A+20):
114 Longitudinal (L) USE is from the data shown in Figure 5-1.
Transverse (T) plate USE is taken as 65% of the longitudinal USE, according to [12].
L/T USE values are equal for weld metal, which has no orientation effect.
5-8
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6.
TENSILE TESTING Six round bar tensile specimens were recovered from the surveillance capsule and tested. Uniaxial tensile tests were conducted on each material in f
air at room temperature (70*F) and RPV operating temperature (550*F).
The tests were conducted in accordance with ASTM EB-89 [13).
i i
6.1 PROCEDURE t'
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 j
i done with a Satec resistance clamshell furnace centered around the specimen monitored and controlled by a load train.
The test temperature was chromel-alumel thermocouple spot-welded to an Inconel clip that was j
friction-clipped to the surface of the specimen at its midliue.
Before the elevated temperature tests, a profile of the furnace was conducted at the test temperature of interest usin5 an unirradiated steel specimen of the same Thermocouples were spot-welded to the top, middle, and bottom of a geometry.
central 1 inch gage of this specimen.
In additfon, the clip-on thermocouple l
was attached to the midline of the specimen. When the target temperatures of f
the three thermocouples were within 5'F of each other, the temperature of the clip-on thermocouple was noted and subsequently used as the target temperature f
for the irradiated specimens.
All tests were conducted at a calibrated crosshead speed of 0.005 inch / min until well past yield, at which time the speed was increased to j
0.05 inch / min until fracture.
Crosshead displacement was used to monitor specimen extension during the test.
I 1
f 6-1
GE-NE-523-165-1292, Rev.-1 DRF B13-01661 j
specimens were machined with a minimum diameter of 0.250 inch The test f
at the center of the gage length.
The yield strength (YS), ultimate tensile strength (UTS) and fracture strength were calculated by dividing the nominal aa2a Ao (0.0491 in ) into the 0.2% offset load, into the maximum test load and 2
into the fracture load, respectively.
The values listed for the uniform and total elongations were obtained from plots that recorded load versus specimen i
extension and are based on a 1.5 inch gage length.
Reduction of area (RA) i i
measurements of l
from Ao to the final area, Ar, was determined from post-test 8
the necked specimen diameters using a calibrated blade micrometer and l
employing the following formula:
j
- A )/A l
RA - 100% * (A f
o o
f a
The necked specimen diameters were also used to calculate the fracture stress, After testing, each broken specimen 1
which is the fracture load divided by Af.
1 was photographed end-on, showing the fracture surface, and lengthwise, showing i
the fracture location and local necking behavior, i
t i
6.2 RESULTS Irradiated tensile test properties of YS, UTS, Fracture Load, Fracture l
Strength, Fracture Stress, RA, Uniform Elongation (UE), and Total Elongation i
presented in Table 6-1.
A stress-strain curve for a 550*F base (TE) are metal irradiated specimen is shown in Figure 6 - 1..
This curve is typical of the stress-strain chsracteristics of all the tested specimens.
Data from f
Table 6-1 are shown graphically in Figures 6 2 and 6-3.
As can be seen from Figures 6-2 and 6-3, the base, weld and HAZ materials generally follow the Photographs of trend of decreasing properties with increasing temperature.
surf aces and necking behavior are given in Figures 6-4 through l
the fracture t
I 6-6.
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l 6-2 1
1 GE-NE-523-165-1292, Rev I 1
DRF B13-01661 J
i 6.3 IRRADIATED VERSUS UNIRRADIATED TENSILE PROPERTIES
~
Unirradiated tensile test data, shown in Table 6-2, were reported in [5]
j for the surveillance plate heat C1079-1.
The unirradiated data provide average values of YS, UTS, RA, and TE at room temperature.
These were compared to the irradiated plate specimen RT data to determine the irradiation effect.
The trends of increasing YS and UTS and of decreasing RA, l
l characteristic of irradiation embrittlement, are seen in the plate data.
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Tabic 6-1 TENSILE TEST RESULTS FOR IRRADIATED RPV MATERIALS Test Yield" Ultimate Fracture Fracture Fracture Uniform Total Reduction Specimen Temp Strength Strength Load Strength Stress Elongation Elongation of Area
__ ksi)
(ksi)
(kio)
(ksi)
(ksi)
(%)
(%)
(%)
(
Number-(*F)
Base:
CDT RT 78.5 98.9 3.21 65.5 178.3 10.9 27.4 63.3' CE4 550 71.0 91.8 3.35 68.3 142.2 7.7 14.4 51.9 e
Weld:
CKC RT 87.0 101.1 3.22 65.9 188.2 12.2 21.3 65.0 CKT 550 71.6 90.4 3.54 71.9 152.0 8.6 14.3 52.7 i
e llAZ:
i CPJ RT 77.3 99.5 3.20 66.1 198.7 8.4 25.1 66.8 CPU 550 74.3 93.3 3.14 64.5 159.7 7.7 15.4 59.6 E$ '
N.
- Yield Strength.is determined by 0.2% offset.
CI' Si
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6-4
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 t
Table 6-2 COMPARISON OF UNIRRADIATED AND IRRADIATED
(
TENSILE PROPERTIES AT ROOM TEMPERATURE Yield Ultimate Total Reduction Strength Strength Elongation of Area (ksi)
(ksi)
(%)
(%)
i Plate:
Unirradiated 69.4 91.7 24.9 65.9 Irradiated 78.5 98.9 27.4 63.3 a
13.1%
7.9%
10.0%
-3.9%
Difference Results from 6.5%
4.3%
-3.6%
-0.8%
210* Capsule i
l i
i a Difference - {(Irradiated - Unirradiated)/Unirradiated]
- 100%
9 f
i 6-5 b
100 MILLSTONE RPV BASE-CE4 90 -
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I i
L 1
i i
I Cr r
_",' +
~
i r
i i
i j
l r
l f,:
I 1
r i
l i
4 m
l l
-=
-.a.
E' t=
CE4 - 550*F Figure 6-4.
Fracture Location, Necking Behavior and Fracture Appearance for Irradiated Base Metal Tensile Specimens l
]
6-9
[
l
,4 GE-NE-523-165-1292, Rev. I 4
DRF B13-01661 9
i i
E_-
f j
.=
I l
/
i l
Q' f_***
k-l
=
?
a, y
3
,;. m o t'.,.
f, :e t
i S4 sa.
j
=
'e
\\
~
~
f it-. %,_.
=::
.i E ;*
f 4
l l
i I'
CKC - RT i
i s
,,7 I
i l
I i
i I
I i
j l
1-I E.-
I CKT. 550*F i
I Figure 6-5.
Fracture Location, Seeking Behavior and Fracture Appearance for Irradiated Veld Metal Tensile Specimens t
6-10 i
, ~..
i l
i
,u.
GE-NE-523-165-1292, Rev. ]
j DRF B13-0165; l
.i d
4 I
s j
0 1
,. ~
I 1
=
l l
d.e.
f -.
=
I f..
l 5
f
,g f
l l
l
~
l l
CPJ. RT I
i
)
I l
1 n
a, vs as I
j b
. IO,
j I
~*
n
~
.1 w l
u
- - - n gs,,
i l
<h-r.
i i
-- s e o.
u, 7,,,.
J
=
m F
i CPU - 550*F i
I i
Figure 6-6.
Fracture Location, Neckim Behavior and Fracture Appearance i
for Irradiated llAZ Metal Tensile S ecimens P
i 6-11 i
m.
t GE-NE-523-165-1292, Rev. 1 l
DRF B13-01661 7.
DEVE14PMENT OF OPERATING LIMITS CURVES i
i I
i
7.1 BACKGROUND
3 F
l l
Operating limits for pressure and temperature are required for three categories of operation:
(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level
[
physics tests, I:ferred to as Curve B; and (c) core critical operation, t
referred to as Curve C.
There are three vessel regions that affect the j
operating limits:
the closure flange region, the core beltline region, and the remainder of the vessel, or non-beltline regions.
The closure flange I
region limits are controlling at lower pressures primarily because of-ll]
[
requirements.
The non-beltline and beltline region operating limits are evaluated according to procedures in [1] and [2), with the beltline region
{
minimum temperature limits increasing as the vessel is irradiated.
The resulting P-T curves for the three conditions above are shown in I
Figures 7-1, 7-2 and 7-3.
The P-T limits are provided in tabular form in f
i Table 7-1.
7.2 NON-BELTLINE REGIONS l
Non-beltline regions are those locations that receive too littic fluence to cause any RTNDT increase.
Non-beltline components include the nozzles (except for recirculation inlet), the closure flanges, some shell plates, top and bottom head plates.
Detailed stress analyses of the l
non-beltline components, considering operating transients with relatively high pressures and low temperatures, were performed for the Bk'R / 6,
specifically for use in developing pressure-temperature (P-T) limits.
The analyses bounded all mechanical loadings and thermal. transients anticipated.
Detailed stresses were used according to {2] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature RTNDT).
These results are applicable to the Millstone vessel (T
I components, since the non-beltline geometries are not significantly
[
4 different from Bk'R/6 configurations and the mechanical and thormal loadings i
are comparable.
7 f
7-1 I
GE-NE-523-165-1292, Rev. 1 l
DRF B13-01661 t
The non-beltline region results were established by adding the highest for the non-beltline discontinuities to the P versus (T_- RTNDT)
RTNDT curves for the most limiting BL'R/6 components, which are the CRD penetration and feedwater nozzle. Table 3-2 has the limiting RTNDT values applicable to t
the feedwater nozzle limits and to the CRD penetration limits.
They are 40*F for the feedwater nozzle limits and CRD penetration limits, based on the nozzle NDT requirement in the vessel purchase specification, j
7.2.1 Non-Beltline Monitorinc Durine Pressure Tests j
L The beltline limits are controlling for pressure test conditions, but i
the non-beltline limits can be applied to other regions of the vessel.
In particular, it is possible that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline.
This condition can occur in the bottom head when th.e recirculation pumps are operating at low speed, or are off, and injection-through the control rod drives is used to pressurize the vessel.
i
?
Monitoring the bottom head separately from the beltline region may I
reduce the required overall pressure test temperature significantly.
20*F colder than j
Consider an example where the bottom head temperatures are the beltline temperatures during a pressure test.
Some hypothetical l
temperatures demonstrating the potential benefit of separate bottom head j
monitoring are shown in Figure 7-4.
The Technical Specifications may
+
currently require that all vessel temperatures be above the limiting conditions on the P-T curve.
That would mean that, for a 1000 psig leak
)
test, the bottom head would have to be heated above 225'F at 32 EFPY, as l
shown in case (a) of Figure 7-4.
The bottom head temperature reading would likely be the limiting reading on the vessel during the test.
If, by using i
i the bottom. head curve, the required temperature for the bottom head were i
only 162*F, the limiting reading would probably be near the beltline, as shown in case (b), and the actual vessel temperatures could be lowered compared to case (a),
?
9 7-2
?
GE-NE-523-165-1292, Rev. 1 i
DRF B13-01661 One condition on monitoring the bottom head separately is that it must be demonstrated that the vessel beltline temperature can be accurately monitored during pressure testing.
An experiment has been conducted at j
a s
Ek'R/4 which showed that thermocouples on the vessel near the feedwater l
nozzles, or temperature measurements of water in the recirculation loops provide good estimates of the beltline temperature during pressure testing.
t This may need to be confirmed before implementing separate monitoring of the bottom head.
First, however, it should be determined whether there are significant temperature differences during the leak test between the beltline region and the bottom head region.
7.3 CORE BELTLINE REGION As the beltline fluence increases during operation, the beltline P-T l
limit curves shift to the right by an amount discussed in Section 7.6,.
Typically, the beltline curves shift to become more limiting than the
)
non-beltline curves at some time during plant life.
The stress intensity l
r factors calculated for the beltline region according to [2] procedures are based on a com'aination of pressure and thermal stresses.
The pressure calculated using thin-walled cylinder equations.
Thermal i
stresses -cre calc.. lated assuming the through-wall temperature distribution i
stresses were The adj us te d of a flat plate subj ected to a 100*F/hr thermal gradient.
RTNDT (ART) values calculated in Section 7.6 for the limiting beltline materials were used to adjust the P versus (T - RTNDT) values from j
l Figure G-2210-1 of [2].
ART is the initial RT due to irradiation.
i r
The vessel thickness is a variable in the stresses as well as the ART l
I values.
In M111 stone's case, the lower-intermediate shell has the smaller thickness, resulting in higher pressure stresses, and the higher ART values.
As a result, P-T limits for the lower-intermediate shell are clearly more limiting.
I t
r r
f 7-3 8
f
i GE-NE-523-165-1292, Rev. 1 l
DRF B13-01661 The recirculation inlet nozzle blend radius is about 5 inches below the f
bottom of active fuel (BAF).
The 32 EFPY fluence at that location is i
estimated to be 9x1016 n/cm2 Using the chemistry in Table 3-1 and the in Table 3-2, the ART was calculated for the nozzle, using the l
initial RTNDT I
methods described in Section 7.6.
The resulting ART was 73*F for 32 EFPY of operation, so the non-beltline feedwater nozzle P-T limits were positioned i
NDT of 73*F to account for the inlet nozzle irradiation through based on a RT t
32 EFPY.
l
}
7.4 CLOSURE F1ANGE REGION r
1 References [1] and [2] have several requirements that affect the P-T 3
i curves, based on the RTNDT values in the closure flange region.
As stated in Paragraph G-2222(c) of {2), for application of full bolt preload and f
reactor pressure up to 20% of preservice hydrostatic test pressure 1
(312 psig), the closure flange region metal temperature must be at RTNDT or t
The GE practice, however, is to recommend (RTNDT + 60*F) for bolt
{
greater.
preload, because the original ASME Code of construction required l
i J
(RTNDT + 60*F) and because boltup is one of the more limiting conditions l
(high stress and low temperature) for brittle fracture.
For Millstone, (RTNDT + 60*F) of the closure region materials is 86*F, because the RTNDT Of I
f
)
Therefore, the an upper shell plate connected to the vessel flange is 26*F..
(
bolt preload temperature used in developing the P-T curves was 86*F.
i I
Reference [1], Paragraph IV.A.2, sets requirements on minimum temperature when pressure is above 312 psig.
The requirements are based on f
i Curve A temperature must be no less than f
the RTNDT of the closure reE on.
f (RTNDT + 120*F).
The (RTNDT + 90*F) and Curve B temperature no less than Curve A requirement causes a 30*F shift at 312 psig on Figure 7-1.
The l
Curve B requirement has no impact on Figures 7-2 and 7-3, because the l
)
i recirculation inlet curves are more bounding.
I c
7-4
.l k
t
i GE-NE-523-165-1292, Rev. 1 DRF B13-01661 e
7.5 CORE CRITICAI. OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is developed from the l
requirements of [1], paragraph IV.A.3.
Essentially paragraph IV.A.3 requires that Curve C be 40*F above any Curve A or B limits.
Curve B is more limiting than Curve A, so Curve C is Curve B plus 40*F.
Curve C initiates at zero pressure at (RTNDT + 60*F), based on an exception for BWRs in Paragraph IV.A.3, allowing critical operation at temperatures below the hydrostatic pressure (Curve A at 1100 psig) test temperature.
This and exception is valid only when water level is within the normal range pressure is below 312 psig.
f t
7.6 EVALUATION OF IRRADIATION EFFECTS The impact on adjusted reference temperature (AFT) due to irradiation in the beltline materials is determined according to the methods in [6), as a function of neutron fluence and the element contents of copper (Cu) and nickel (N1). The specific relationship from [6] is:
(7-1)
ART - Initial RTNDT + ARTNDT + Margin
?
where:
(7 2)
NDT - [CF)*f(0.28 - 0.10 log f)
ART 2
(7 3)
Margin - 2*(a12 + a3 )l/2 j
i CF -
chemistry factor from Tables 1 or 2 of [6], or from surveillance data, f-1/4 T fluence (n/cm ) divided by 1019,
{
2 standard deviation on initial RTNDT.
ay -
i standard deviation on ARTNDT, is 28'F for welds and 17'F og -
for base material, except that og need not exceed 0.50 times the ARTNDT value.
When using surveillance data i
is taken as 1/2 the values above.
for ARTNDT. "A j
i 7-5 l
GE-NE-523-165-1292, Rev. 1 DRF B13 01661 r
Once two sets of surveillance capsule data are available, the CF values in (6] can be modified to reflect the results.
The method recommended in [6]
is described there in Position 2.1, and is summarized below.
i There are two aspects to applying the surveillance data to the beltline materials.
First is the method of determining the adj u s tment to the surveillance CF.
Second is the method of applyin5 the CF adjustment to the I
beltline materials, f
t 7.6.1 Surveillance CF Adiustment f
i-The surveillance CF adjustment is based on a least squares fit of the surveillance data to the ARTNDT equation (7-2), which can be restated as:
j ARTNDT - CF
- FF I
where FF is the fluence factor shown in (7-2)
The least squares approach uses the actual shifts of the 210* and 300*
i capsule Charpy specimens, comb!ned with the fluence factors applicable to those capsule fluences.
i CF - (Shift 210*FF210 + Shift 300*FF300)/(FF2102 + FF3002)
(7-4)
The values for Equation 7-4 are in Section 5.4.1 and Table 5-4:
l i
?
Plate Veld Plate Veld Location FE__
210*
0.2553 61 22 140.7 228.5 1
300*
0.3389 78 76 4
substituting these values into Equation 7-4 give the following CFs based on surveillance data:
Plate CF - 233.3 i
Veld CF - 174.3 a
7-6 i
l I
GE-NE-523-1G5-1292, Rev. 1 l
DRF B13-01661 1
i The surveillance CFs are compared to the 1.99 CFs to establish the adjustment of generic 1.99 predictions to actual plant conditions:
i Plate Adjustment - 233.3/140.7 - 1.66 Weld Adjustment - 174.3/228.5 - M 1
f 7.6.2 Application of CF Adiustments to Beltline Materials l
The assumption made in applying the CF adjustments to the beltline materials is that the plant-specific conditions which affected the surveillance material shifts also affect the beltline material shifts. This is the same assumption made in 1.99, Position 2.1, for the case where the vessel weld chemistry differs from the surveillance weld chemistry.
- There, the shifts are adjusted by the ratios of the beltline and surveillance
~
material CFs before the least squares calculation is done. Here, the ratios of material CFs are considered after the least squares calculation is done, but the underlying assumption and the results are the same.
In Position 2.1 of 1.99, it appears that the CF ratio approach is intended for the case where the beltline and surveillance welds are the same heat, but chemistry results are, for some reason, significantly different.
Here, the CF ratio approach is taken further by assuming that the surveillance CF adjustments apply to other beltline heats, as well as applying to the same beltline heats.
Both approaches are based on the same assumption; that the CFs for different chemistries in the 1.99 tables are correct relative to one another.
The result is that Equation 7-2 from 1.99 is multiplied by the surveillance adjustment (SA):
(7-5)
/2RTNDT - CF
- SA Implicit in this approach is the assumption that there is a variable other than chemistry or fluence that is affecting the t.RTNDT.
This assumption is feasible when evaluating a BWR, because nearly all of the data' in the data base used to develop 1.99, Revision 2 were PWR data, with fluxes and fluences significantly higher than are typical for BWRs. Fluxes may 7-7
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
vary by a factor of 2 to 100, depending on the specific BWR and-PWR i
compared.
Fluences tend to vary less, because the first PWR surveillance
[
]
capsules are withdrawn earlier in operation than the first BWR capsules, but the fluence differences are still large.
There is also a temperature difference between PWR and BWR surveillance capsule irradiation conditions, where BWR irradiation temperatures of 525'F to 535'F may be 15*F to 35*F
}
lower than PWR irradiation temperatures.
One, or a combination of these variables may account for the quantity SA.
i Although the SA value calculated for the surveillance weld was 0.76, for the beltline welds was assumed to the SA value used to calculate ARTNDT be 1.0 for two reasons:
The 300* surveillance weld shift is very close to matching the based on 1.99.
predicted ARTNDT chemistry significantly different The limiting beltline veld has a
(more limiting; than that of the surveillance weld.
Therefore, in calculating ARTNDT values for the beltline materials, SA values of 1.66 and 1.0 were used for the plates and welds, respectively.
[
I NDT values for the beltline are calculated using Equation 7-5.
The ART The plate Margin terms are taken as half the normal values, as permitted in Position 2.1 of 1.99.
For the plates, the approach using SA-1.66 and la j
Margins is more conservative than using SA-i.0 and 2a Margin.
j l
i Position 2.1 of 1.99 allows that, if a surveillance material is I
credible but is not the limiting material, the value of as may be reduced by an amount to be decided on a case-by-case basis.
The surveillance weld meets criteria 2, 3 and 4 fer credible data in the 1.99 discussion I
(criterion 5 could not be met because there was no correlation monitor material in the capsule).
Both measured shifts for the surveillance weld j
were less than or equal to the 1.99 prediction, as shown in Figure 5-8.
I determination of the beltline weld ARTS with a3 reduced by half Therefore, is expected to provide adequate conservatism.
7-8 l
l
f r
CE-NE-523-165-1292, Rev. 1 i
DRF B13-01661 7.6.3 ART Versus EFPY Each beltline plate and weld ARTNDT value is determined by multiplying the CF from 1.99, determined for the Cu-Ni content of the material, by the fluence factor for the EFPY being evaluated and by the appropriate i
surveillance adjustment. The Margin term and initial RTNDT are added to get
{
the ART of the material.
Calculations to determine 32 EFPY ART values, and are summarized in Table 7-2.
The thus the limiting beltline materials, results show that beltline plate C1079-1 is limiting. As a result of adding la Margin to the veld ARTNDT, the welds are not limiting.
One input to the 1.99 calculations not shown in Table 7-2 is that
- I 0*F for the beltline materials, all of which have RTNDT values t
i determined from measured Charpy data.
The basis for using og - O'F is discussed in more detail in Appendix B.
)
]
Figure 7-5 shows the ART for the limiting materials as a function of a
EFPY. The ART resulting in the limiting P-T curves at 32 EFPY is 137'F.
t 7.6.4 Urrer Shelf Enerry at 32 EFPY t
l of Paragraph IV.B of [1] sets limits on the upper shelf energy (USE)
)
the beltline materials.
The USE must be above 50 f t-lb at all times during i
plant operation, assumed here to be up to 32 EFPY.
Calculations of 32 EFPY USE, using 1.99 methods, are summarized in Table 7-3.
The initial longitudinal USE of the plates was taken as the average of all Charp'; test l
t results with 100% shear, except for C1079-1 where the surveillance USE was transverse USE of the plate material is taken as 65%
used.
The equivalent of the longitudinal USE, according to [12]. Although the plate surveillance data show the decrease in USE to be considerably less than the prediction j
the USE decrease prediction values l
for the corresponding copper content, from 1.99 vere used for the beltline plates in Table 7-3.
The lowest plate j
USE shown in Table 7-3 is 55 ft-lb at 32 EFPY.
7-9 i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 The weld metal initial USE values were not taken during vessel fabrication.
For circumferential veld heat 34B009, the surveillance weld f
so the surveillance weld was made with the same heat / flux combination, initial USE is used in Table 7-3.
For longitudinal weld heat V5214, GE's QA I
data was reviewed and the lowest initial USE value found for a Linde 1092 f
weld was assumed for the W5214 initial USE.
The range of Linde 1092 USE values was 98 ft-lb to 145 ft-lb.
Unlike the plate, the veld metal USE has transverse / longitudinal correction because weld metal has no orientation no
-r effect.
i The 300' capsule surveillance veld data showed a larger decrease in USE i
(21%) than predicted by 1.99 (18%).
Following the method in 1.99 Position I
1.99 USE decrease prediction lines were plotted in Figure 7-6 and 2.2, the the surveillance weld USE decrease data point was added to the plot.
t Coincidentally, the surveillance data point falls on the 1.99 line for l'
0.25% Cu.
The surveillance weld, however, has only 0.20% Cu.
Therefore, for the beltline the approach used to determine USE decrease percentages The welds was to add 0.05% to the Cu content values reported in Table 3-1.
decreases for the beltline welds are shown in Figure 7 6, l
I resulting percent and in Table 7-3.
The lowest veld USE shown in Table 7-3 is 71.5 ft-Ib at l
and this value is based on the conservative initial USE taken for 32 EFPY, the W5214 weld. The actual value is expected to be higher.
f Based on the above results, it is expected that the beltline materials Since will have USE values above 50 f t-lb at 32 EFPY, as required in [1].
USE and ART requirements are met, irradiation effects are not severe er,cugh l
to necessitate additional analyses or preparations for RPV annealing before NU is to the uncertainty of weld USE predictions, 32 EFPY.
However, due participating in a BUR Owners' Group program to perform analyses to l
demonstrate equivalent margin in cases where USE drops below 50 ft-lb [14).
This analysis shows equivalent margin at'USE values significantly lower than i
the values in Table 7-3.
l i
n t
l 7-10 i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 7.7 OPERATING LIMITS CURVES VALID TO 32 EFPY The P-T Figures 7-1 through 7-3 show P-T curves valid to 32 EFPY.
curves are developed by considering the requirements applicable to the non-beltline, beltline and closure flange regions. The pressure test curves in Figure 7-1 include a bottom head curve, which does not change with time, and beltline curves for 18, 21, 24, 28 and 32 EFPY.
r 7.8 REACTOR OPERATION VERSUS OPERATING LIMITS For most reactor operating conditions, pressure and temperature are at saturation conditions, which are well into the acceptable operating area (to the right of the P-T curves).
The most severe unplanned transient relative condition consisting of several transients to the P-T curves is an upset which result in a SCRAM.
The worst combination of pressure and temperature in the lower during this postulated event is 1180 psig with temperatures head of 250*F.
In this case, the core is not critical, so the non-nuclear heatup/cooldown curve for the bottom head applies (Curve B).
As seen in Figure 7-2 or Table 7-1, at 1180 psi the minimum transient temperature of of 201*F.
acceptably higher than the bottom head curve value 250*F is Therefore, violation of the P-T curves is only a concern in cases where such as during pressure testing and initiation operator interaction occurs, of criticality.
f f
7-11
i GE-NE-523-165 ?9?; at.v. 3 DRF B13-01661' t
Table 7-1 MILLS 10kE P-T CURVE VALUES i
n. o..n.....................R E QU1 RE D T EMPER ATURE 5""""*"* "*"*""* * """
f e
BOTTDM IB EFPY 21 EFPY 24 EFPY 28 EFPY 32 EFPY BOTTOM 32 EFPY 32 EFPT FRE55URE HEA0 F1kAL F1kAL FIhAL F1kAL F1hAL HEAD F1hAL F1hAL CURVE A CURVE A CURVE A CURVE A CURVE A CURVE A CURVE B CURVE B CURVE C O
B5.0 B6.0 B6.0 B6.0 B6.0 B6.0 B6.0 j
10 66.0 86.0 86.0 86.0 86.0 86.0 86.0 l
20 86.0 B5.0 86.0 86.0 86.0 86.0 86.0 30 86.0 B6.0 86.0 B6.0 86.0 86.0 B6.0' 40 B6.0 B6.0 86.0 B6.0 86.0 66.0 114.0 50 B6.0 86.0 86.0 B6.0 86.0 67.0 127.0 60 B6.0 86.0 66.0 86.0 B6.0 98.0 138.0 I
70 B6.0 66.0 66.0 B6.0 86.0 107.5 147.5 BD B6.0 B6.0 B6.0 86.0 86.0 115.7 155.7 1
90 86.0 66.0 B6.0 B6.0 86.0 122.7 162.7 100 B6.0 B6.0 B6.0 86.0 B6.0 128.8 168.8 -
l 110 B6.0 B6,0 B6.0 66.0 86.0 134.4 174.4' 120 86.0 86.0 86.0 B5.0 86.0 139.3 179.'3
.j 4
130 86.0 86.0 66.0 86.0 B6.0 144.1 164.1 140 B6.0 B6.0 B6.0 86.0 B6.0 148.7 168.7 I
150 86.0 86.0 86.0 B6.0 B6.0 153.0 193.0 160 86.0 86.0 86.0 B6.0 86.0 156.9 196.9 170 B5.0 B6.0 B6.0 66.0 86.0 160.3 200.3 ISO 80.0 86.0 B6.0 86.0 B6.0 163.3 203.3 190 B6.0 86.0 B6.0 B6.0 B6.0 166.1 206.1 200 86.0 B6.0 86.0 86.0 86.0 168.8 208.B j
210 66.0 86.0 B5.0 86.0 86.0 171.5 211.5 1
220 86.0 86.0 B6.0 B6.0 B6.0 174.0 214.0 230 86.0 86.0 B6.0 B6.0 86.0 176.4 216.4 l
240 86.0 86.0 B6.0 86.D B6.0 178.7 218.7 250 66.0 B6.0 B6.0 B6.0 B6.0 180.9 220.9 260 B6.0 B6.0 B6.0 B6.0 B6.0 183.0 223.0 270 B6.0 66.0 86.0 B6.0 BE.0 185.0 225.0 2B0 86.0 B6.0 86.0 B6.0 86.0 187.0 227.0 l
290 86.0 66.0 86.0 B6.0 86.0 188.9 228.9 300 86.0 B6.0 B6.0 B6.0 86.0 190.7 230.7 310 B6.0 B6.0 B6.0 B6.0 B6.0 192.5 232.5 j
312.5 B5.0 86.0 66.0 B6.0 86.0 192.9 232.9 j
312.5 116.0 116.0 116.0 116.0 116.0 192.9 234.5
_{
I 320 116.0 116.0 116.0 116.0 116.0 194.2 234.5 330 116.0 116.0 116.0 116.0 116.0 195.B 235.B 340 116.0 116.0 116.0 116.0 116.0 197.4 237.4 350 116.0 116.0 116.0 116.0 116.0 199.0 239.0 360 116.0 116.0 116.0 116.0 116.0 200.5 240.5 370 116.0 116.0 116.0 116.0 116.0 202.0 242.0 3B0 116.0 116.0 116.0 116.0 116.0 203.5 243.5 390 116.0 116.0 116.0 116.0 116.0 205.0 245.0 400 41.6 116.0 116.0 116.0 116.0 116.0 70.6 206.5 246.5 410 46.9 116.0 116.0 116.0 116.0 116.0 79.6 208.0 248.0 420 51.7 116.0 116.0 116.0 116.0 116.0 B6.6 209.4 249.4 PTFULL Revision 1. 2/93 7-12 i
i J
GE-NE-523-165-l?92, Rev. I DRF B13-01661 Y
E Table 7-1 i
MILLSTONE P-T CURVE VALUES i
............................ * *R 0 0UI R I O T E M P E R AT UR E 5 * * * * * ** * * * * * * * * * * * * * * * * * * * * * *
- p i
BOTTOM 18 EFPY 21 EFPY 24 EFPY 2B [FPY 32 EFPY BOTTON 32 EFPY 32 EFPY PRESSURE HEAD F!hAL F1hAL F1hAL FIhAL F!kAL HEAD FikAL FIhAL CURVE A CURVE A CURVE A CURVE A CURVE A CURVE A CURVE 8 CURVE B CURVE C I
430 56.3 116.0 116.0 116.0 116.0 116.0 92.6 210.8 250.8 440 60.6 116.0 116.0 116.0 116 0 116.0 97.6 212.2 252.2 450 64.6 116.0 116.0 116.0 116.0 116.0 101.5 213.5 253.f -
460 68.4 116.0 116,0 116.0 116.0 116.0 105.1 214.B 254.8
-['
470 72,0 116.0 116.0 116.0 116.0 116.0 108.2 216.1 256.1 4B0 75.4 116.0 116.0 116.0 116.0 116.0 111.1 217.3 257.3 a
490 78.7 116.0 116.0 116.0 116.0 116.0 113.9 218.5 258.5 500 Bl.B 116.0 116.0 116.0 116.0 117.9 116.6 219.7 259.7 510 B4.7 116.0 116.0 116.0 116.8 122.8 119.3 220.8
'260.8 520 67.5 116.0 116.0 116.0 121.4 127.4 121.9 221.9 261.9 530 90.3 116.0 116.0 119.7 125.7 131.7 124.5 223.0 263.0' 540 92.9 116.0 118.7 123.7 129.7 135.7 127.0 224.0 264.0 550 35.5 117.0 122.5 127.5 133.5 139.5 129.4 225.0 265.0 560 97.9 120.1 126.1 131.1 137.1 143.1 131.7 226.0 266.0 570 100.3 123,6 129.6 134.6 140.6 146.6 133.9 226.9 266.9 5B0 102.5 126.9 132.9 137.9 143.9 149.9 136.0 227.8 267.8 590 104.7 130.0 136.0 141.0 147.0 153.0 137.8 228.7 268.7 I
600 106.9 133.0 139.0 144.0 150.0 156.0 139.6 229.5 269.5 610 108.9 135.9 141.9 146.9 152.9 158.9 141.3 230.3 270.3 620 110.9 138.6 144.6 149.5 155.6 161.6 143.1 231.1 271.1 630 112.9 141.3 147.3 152.3 158.3 164.3 144.7 231.8 271.8 640 114.B 143.8 149.8 154.B 160.B 166.8 146.4 232.5 272.5 650 116.5 146.3 152.3 157.3 163.3 169.3 147.9 233.1 273.1 660 118.4 148.6 154.6 159.6 165.6 171.6 149.4 233.8 273.8 i
670 120.2 150.9 156.9 161.9 167.9 173.9 150.9 234.4 274.4 680 121.9 153.2 159.2 164.2 170.2 176.2 152.4 234.9 274.9 E90 123.6 155.3 161.3 166.3 172.3 178.3 153.8 235.5 275.5 l
700 125.2 157.4 163.4 168.4 174.4 180.4 155.2 236.0 276.0 710 126.8 159.4 165.4 170.4 176.4 182.4 156.6 236.5 276.5 720 128.1 161.4 167.4 172..
178.4 184.4 157.9 236.9 276.9 730 129.B 163.3 169.3 174.3 160.3 166.3 159.2 237.4 277.4 740 131.3 165.2 171.2 176.2 182.2 188.2 160.4 237.8 277.8 l
750 132.8 167.0 173.0 17B.0 184.0 190.0 161.6 238.3 278.3 f
760' 134.2 168.7 174.7 179.7 185.7 191.7 162.7 238.9 278.9 770 135.6 170.5 176.5 181.5 187.5 193.5 163.8 240.1 260.1 7BD 136.9 172.1 178.1 183.1 189.1 195.1 164.9 241.2 281.2 790 138,3 173.8 179.B 1B4.8 190.6 196.8 166.0 242.4 2B2.4 800 139.6 175.4 181.4 186.4 192.4 19B.4 167.1 243.5 2B3.5 B10 140.9 177.0 183.0 168.0 194.0 200.0 168.2 244.6 284.6 5
B20 142.2 178.5 164.5 189.5 195.5 201.5 169.3 245.6 285.6 B30 143.4 180.0 106.0 191.0 197.0 203.0 170.3 246.7 286.7 640 144.6 181.5 187.5 192.5 198.5 204.5 171.4 247.7 2B7.7 850 145.8 182.9 168.9 193.9 199.9 205.9 172.4 248.7 288.7 650 147.0 184.3 190.3 195.3 201.3 207.3 173.3 249.8 289.8 B70.
148.1 185.7 191.7 196.7 202.7 208.7 174.5 250.6 290.B l
PTFULI. Revis ton 1, 2/93 7,33
GE-NE-523-165-1292s Rev 1 DRF B13-01661 i
f Table 7-1 1
FILLST0hE P-T CURVE VALUES
.. u.. u.......u... u.u..u.E E 0VIR E D T EWER AT URES""" *"" * """" * """"*
3 BOTTDM 18 EFPY 21 EFPY 24 EFPY 2B EFPY 32 EFPY BOTTOM 32 [FPY 32 EFPY FRESSURE HEAD F1kAL Fl%AL F1ht.L F1hAL FINAL HEAD F1hAL F1hAL CURVE A CURVE A CURVE A CURVE A CUEVE A CURVE A CURVE B CURVE B CURVE C
-l t
BBC 149.3 187.0 193.0 ISB.O 204.0 210.0 175.6 251.7 291.7 890 150.4 188.3 194.3 199.3 205.3 211.3 176.6 252.7 292.7 900 151.5 189.6 195.6 200.6 206.6 212.6 177.7 253.7 293.7 910 152.6 190.9 196.9 201.9 207.9 213.9 178.7 254.6 294.6 920 153.6 192.2 196.2 203.2 209.2 215.2 179.7 255.5 295.5 930 154.7 193.4 199.4 204.4 210.4 216.4 180.7 256.4 296.4 940 155.7 194.6 200.6 205.6 211.6 217.6 161.7 257.3 297.3 950 156.6 195.8 201.B 206.B 212.B 218.8 182.7 258.2 298.2 960 157.8 197.0 203.0 208.0 214.0 220.0 183.7 259.1 299.1 970 158.7 198.1 204.1 209.1 215.1 221.1 164.7 260.0 300.0 980 159.7 199.3 205.3 210.3 216.3 222.3 185.7 260.B 300.8 990 160.7 200.4 206.4 211.4 217.4 223.4 186.6 261.7 301.7 i
1000 161.6 201.5 207.5 212.5 218.5 224.5 187.6 262.5 302.5 1010 162.6 202.5 206.5 213.5 219.5 225.5 1B8.5 263.4 303.4 1020 163.5 203.6 209.6 214.6 220.6 226.6 189.4 264.2 304.2 1030 164.4 204.6 210.6 215.6 221.6 227.6 190.2 265.0 305.0 f
1040 165.3 205.7 211.7 216.7 222.7 228.7 191.0 265.6 305.8 1050 166.2 206.7 212.7 217.7 223.7 229.7 191.8 266.6 306.6 1050 167.1 207.7 213.7 218.7 224.7 230.7 192.6 267.4 307.4 1070 167.9 208.7 214.7 219.7 225.7 231.7 193.4 268.1 308.1
?
1080 168.8 209.6 215.6 220.6 226.6 232.6 194.2 268.9 308.9 1090 169.6 210.6 216.6 221.6 227.6 233.6 194.9 269.7 309.7 1100 170.4 211.5 217.5 222.5 228.5 234.5 195.7 270.4 310.4 1110 171.3 212.5 218.5 223.5 229.5 235.5 196.4 271.1 311.1 1120 172.1 213.4 219.4 224.4 230.4 236.4 197.1 271.9 311.9 1130 172.9 214.3 220.3 225.3 231.3 237.3 197.8 272.6 312.6 1140 173.7 215.2 221.2 226.2 232.2 238.2 198.5 273.3 313.3 i
1150 174.5 216.1 222.1 227.1 233.1 239.1 199.2 274.0 314.0 1160 175.2 216.9 222.9 227.9 233.9 239.9 199.9 274.7 314.7 1170 176.0 217.8 223.8 228.8 234.8 240.8 200.5 275.4 315.4 l
I 1180 176.8 218.7 224.7 229.7 235.7 241.7 201.2 276.1 316.1 4
1190 177.5 219.5 225.5 230.5 236.5 242.5 201.9 276.B 316.8 1200 178.3 220.3 226.3 231.3 237.3 243.3 202.6 277.5 317.5 7
1210 179.0 221.1 227.1 232.1 23B.1 244.1 203.2 278.1 318.1 I
1220 179.7 222.0 228.0 233.0 239.0 245.0 203.9 278.8 318.8 1230 180.4 222.8 22B.8 233.8 239.8 245.B 204.5 279.4 319.4 1240 181.1 223.6 229.6 234.6 240.6 246.6 205.2 280.1 320.1 1250 181.B 224.3 230.3 235.3 241.3 247,3 205.8 260.7 320.7 l
1260 1B2.5 225.1 231.1 236.1 242.1 248.1 206.5 281.4 321.4 1270 183.2 225.9 231.9 236.9 242.9 248.9 207.1 282.0 322.0 1280 1B3.9 226.6 232.6 237.6 243.6 249.6 207.8 2B2.6 322.6 i
1290 1B4.6 227.4 233.4 238.4 244.4 250.4 208.4 283.2 323.2 1300 285.3 228.1 234.1 239.1 245.1 251.1 209.1 283.9 323.9 1310 185.3 228.9 234.9 239.9 245.9 251.9 209.7 284.5 324.5 1320'-
186.6 229.6 235.6 240.6 246.6 252.6 210.4 285.1 325.1 l
PTFULL Revision 1. 2/93 7-14
~
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
Tatie 7-1 MILLSTOWE P-T CtRVE VALUES I
............................ * *RI QUI R E D T E MP E R AT UR E 5* * "* """ * * " * " * * * """"
PRESSURE HEAD F1hAL F1hAL F1hAL F1hAL F1hAL HEAD FINAL F1hAL CURVE A Ct*VE A CURVE A CURVE A CtRVE A CURVE A CURVE B CLEVE B CURVE C 1330 187.2 230.3 236.3 241.3~
247.3 253.3 211.0 285.7 325.7 1340 187.9 231.0 237.0 242.0 248.0 254.0 211.7 2BS.3 326.3 1350 188.5 231.7 237.7 242.7 248.7 254.7 212.3 286.9 326.9
[
1360 189.1 234.B 240.8 245.8 251.8 257.8 213.0 2E9.4 329.4
'I'370 189.8 235.5 241.5 246.5 252.5 25B.5 213.6 290.0 330.0 i
1380 190.4 236.1 242.1 247.1 253.1 259.1 214.3 290.6 330.6 1390 191.0 236.8 242.8 247.8 253.8 259.8 214.9 291.1 331.1 1400 191.6 237.5 243.5 248.5 254.5 260.5 215.6 291.7 331.7 f
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i 7-15
Table 7-2 BELTLINE EVALUATION FOR MILLSTONE AT 32 EFPY INCLUDING SURVEILLANCE ADJUSTMENTS (SA)
Low-int Shell:
Low-int Shell Thickness -
5.5 inches 32 EFPY Peak 1.D. fluence -
1.5E+18 32 EFPY Peak 1/4 T fluence =
1.lE+18 Lower Shell:
Lower Shell Thickness -
6.5 inches 32 EFPY Peak I.D. fluence -
8.9E+17 32 EFPY Peak 1/4 T fluence -
6.0E+17 Initial 32 EFPY 32 EFPY 32 EFPY COMPONENT I.0.
HEAT OR HEAT / LOT
%Cu %Ni CF SA RTndt Delta RTndt Margin Shift ART PLATES:
lower Shell G2001-lR Cl359-1 0.22 0.49 146.2 1.66 6
78.5 17.0 95.5 101.5 2.
Lower Shell G2001-3 84928-1 0.23 0.52 155.4 1.66 10 83.4 17.0 100.4 110.4
.a Lower Shell G2001-5 C1140-2 0.23 0.44 143.8 1.66 22 77.2 17.0 94.2 116.2 o'
Low-int Shell G2002-4 85013-2 0.21 0.49 140.7 1.66
-4 100.3 17.0 117.3 113.3 Low-int Shell G2002-5 C1079-1 0.19 0.51 132.1 1.66 26 94.2 17.0 111.2 137.2 Low-Int Shell G2002-6 C1140-1 0.21 0.45 135.5 1.66 20 96.6 17.0 113.6 133.6 e,
WELDS:
Lower Long.
2-073 W5214, LINDE 1092 0.26 1.2 276 1
-20 89.2 28.0 117.2 97.2 El s,
FLUX LOT 3617 O!
low-Int Long.
1-073 W5214, LINDE 1092 0.26 1.2 276 1
-20 118.5 28.0 146.5 126.5 J.-
S; FLUX LOT 3617
!B.'.
Lower to 3-073 348009,LINDE 1092 0.18 1.03 216.4 1
-50 92.9 28.0 120.9 70.9
'Of ES!"
low-int Girth FLUX LOT 3708 Y:o ST 8;-
GE-NE-523-165-1292, Rev. I DRF B13-01661 P
Table 7-3 UPPER SHELF ENERGY ANALYSIS FOR MILLSTONE BELTLINE MATERIALS INITIAL INITIAL 32 EFPY 32 EFPY LONGIT.
TRANS.
FLUENCE % DECR.
TRANS.
LOCATION HEAT USE USE
%Cu (x10^18)
USE USE PLATES:
Lower C1359-1 103
- 67. 0 0.22 0.6 16.5 55.9 B4928-1 102 66.3 0.23 0.6 17 55.0 C1140-2 115 74.8 0.23 0.6 17 62.0 Low-Int.
B5013-2 110 71.5 0.21 1.1 18 58.6 C1079-1 112 72.8 0.19 1.1 17 60.4 C1140-1 105 68.3 0.21 1.1 18 56.0 WELDS:
Vertical 1-073 &
W5214 (a,b) 98 0.26 1.1 27 71.5 2-073 LINDE 1092 i
Girth 3-073 34B009 (a) 111 0.18 1.1 22.5 86.0 LINDE 1092 (a) USE decreases were adjusted, based on surveillance data, as shown in Figure 7-6 and described in Section 7.6.4 (b) The assumed USE is the lowest in the range of data available for Linde 1092 submerged arc welds I
t 7-17
GE-NE-523-165-1292, Rev. 1 DRF B13-01661-L 1600 I
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32 137 EL 200 BOLTU" CURVES ARE VALID FOR E6*F 32 EFPY OF OPERATION, EXCEPT NON-BELiLINE,
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7-18
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'GE-NE-523-165-1292, Rev. J' DRF B13-01661 I
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Figure 7-2.
P-T Curves for Non-Nuclear Heatup/Cooldown
-l 6
5 7-19 i
GE-NE-523-165-1292, Rev. 1 DRF B13-01661 i
i 1600 f
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400 D
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[6 r 32 EFPY OF OPERATION l
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Figure 7-3.
P-T Curves for Core Critical Operation 7-20 i
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8.
REFERENCES
" Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of
[1]
the Code of Federal Regulations. July 1983.
" Protection Against Non-Ductile Failure," Appendix G to Section XI of
[2]
the 1992 ASME Boiler & Pressure Vessel Code, with 1992 Addenda.
" Reactor Vessel Material Surveillance Program Requirements," Appendix H f
[3]
to Part 50 of Title 10 of the Code of Federal Regulations, July-1983.
j t
[4]
" Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards, E185-82, July 1982.
" Millstone 1 Nuclear Str. tion RPV Surveillance Materials
[5] Caine, T.A.,
I Testing and Fracture Tou ghne s s Analysis," GE Report NEDC-30833, December 1984.
i
" Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory
[6]
Guide 1.99, Revision 2, May 1988.
i
[7] Vang, M.T., " Fracture Toughness of RPV Steel Welds.," GE Report NEDC-30299, October 1983.
on Phase 2 of the BVROG Supplemental
[8]
- Caine, T.A.,
" Progress Report Surveillance Program,* GE Report GE-NE-523 99-0792, DRF Bll-00392-1,
/
January 1992.
[9]
" Standard Methods for Notched Bar Impact Testing of Metallic l
4 Materials," Annual Book of ASTM Standards, E23-88, October 1988.
l
[10) GE VPr# 1444-305-1, " Fabrication Test Program for Millstone Point 224" BVR," August 1968, ill) " Nuclear Plant Irradiated Steel Handbook," EPRI Report NP-4797 September 1986.
8-1 J
- - - =
GE-NE-523-165-1292, Rev. 1 DRF B13-01661
[12) " Fracture Toughness Requirements,"
USNRC Branch Technical Position MTEB 5-2, Revision 1, July 1981.
1 Annual
[13] " Standard Methods of Tension Testing of Metallic Materials,"
Book of ASTM Standards, E8-89, July 1989.
[14]'BUROC Presentation at NRC Meeting, "BVR Owners' Group Upper Shelf Energy Equivalent Margin Analyses," January 5, 1993.
l t
9 k
B-2
GE-NE-523-165-1292, Rev. I DRF B13-01661 APPENDIX A CHARPY SPECIMEN FPACTURE SURFACE PHOTOGPAPHS j
i Photographs of the irradiated Charpy specimen fracture surfaces were taken per 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 with the materials in the order of base, veld and HAZ.
f s
l l
l i
1 i
i
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A-1 I
1
1 GE-NE-523-165-1292, Rev. 1 DRF B13-01661 W
BASE METAL s.
C_s -
.'i'4. h s
-# A t
.a B34
'.7.'.-,
B36
- [
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. f-
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f B3J B37 Test Fracture Lateral Percent Shear f
Specimen Te:rperature Energy Expansion (Method 1) i Identification
(*F)
(ft-lb)
(mils)
(%)
I B34 20 9.5 7.5 13 i
B36 60 17.5 18 16 B37 100 24.5 23 40 i
B3J 120 59 37 52 I
B3L 140 47 43 60 1
l B3A 150 61 49 59 B3C 200 80.5 63.5 94 l
B3E 300 100 78 100 t
f
}
B3A B3L l
l B3E B3C A-2 l
i
GE-NE-523-165-1292, Rev. I j
DRF B13-01661 3
4 j
VELD METAL
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j Test Fracture Lateral Percent Shear Specimen Temperature Energy Expansion (Method 1)
Identification
(*F)
(ft-lb)
(mils)
(%)
i BK7
-20 19 13 18 i
EKA 20 34 27 38 i
BKJ 30 29 23 35 j
BKB 60 39 29 44 EKK 100 51.5 47 64 i
(
BKC 150 79.5 69 96 B1m 200 84 69 100 BKE 300 87 75 100 Exc BKK i
i i
l EKE B13 A-3 i
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l-GE-NE-523-165-1292, Rev. 1 l
DRF B13-01651 J
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HAZ KETAL j
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Test Fracture Lateral Percent Shear l
Specimen Tc:r.pe ra ture Energy Expansion (Method 1) i Identification
(*F)
(ft-1b)
(mils)
(%)
l j
C4K 20 23.5 20.5 28 i
C4Y 40 40 32 36 C4L 60 71.5 49 53 I
C51 80 61 55 61 l
C4M 100 55 45 68 j
C4P 150 102.5 74 100 C4T 200 92.5 73 99 C4U 300 110 77 100 C4P C4M C4U C'4T A-4 l
GE-NE-523-165-1292, Rev. I J
DRF B13-01661 j
APPENDIX B BASIS FOR CONSERVATIVE RTNDT j
The values of initial RTNDT used in this analysis were based on I
30 ft-lb impact energy verification testing, with longitudinal Charpy as was standard practice at the time of vessel specimens used for plate, fabrication. The calculations of initial RTNDT values in [5] are based on a CE procedure which establishes conservative values of RTNDT from the t
fabrication test data.
These RTNDT values are expected to be conservative l
compared to results that would be obtained from current test methods.
For beltline materials, the methods of calculating adjusted RTNDT in l
1.99 include a Margin term to be added to the calculated value, t.RTNDT.
The Margin term includes a component for uncertainty in initial RTNDT.
'I-1 99 discusses determination of al for two categories of initial RTNDT, measured l
~
1 values and generic mean values.
For generic mean values, or is simply the l
standard deviation calculated for the data set used to compute the mean.
For measured values, requirements for determination of al are somewhat for the material f
vague.
Rev 2 states, "If a measured value of initial RTNDT in question is available, or is to be estimated from the precision of the test method. a GE's position for RTNDT values derived from measured data, is the case for the Millstone beltline materials, is that al is zero, as as explained in the next paragraph.
i i
l i
1 a In the 1.99 draft which was circulated after editing to incorporate public i
comments, the text stated, "oI, the standard deviation for the initial-l RTNDT, may be taken as zero if a measured value of initial RTNDT for the material in question is available.*
B-1 i
CE-NE-523-165-1292, Rev. 1 DRF B13-01661 The Charpy curves that were fit to surveillance data, which ultimately data for development of 1.99, were best-estimate fits.
provided the ARTNDT i
An example is provided as Curve 1 in Figure B-1.
However, the ASME Code approach to determining RTNDT is based on the lowest value of three specimens exceeding the required limits of impact energy and lateral A visualization of a Charpy curve drawn on the basis of the Code expansion.
NDT approach is shown as Curve 2 in Figure B-1.
In comparing Curves 1 and RT rather than is clear that Curve 2, which is based on the lowest value 2, it f
the mean value, provides a conservative estimate of initial RTNDT-from measured Therefore, the current ASME Code method of determining RTNDT data is conservative.
Since the method used in [5] to calculate RTNDT values is conservative compared to current ASME methods, or - 0*F is i
appropriate.
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%C' TEST TEMPERATURE.( F) e5 Y
I o.m Figure ' B-1.
Comparison of Surveillance Fil and RTndt Approach
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Docket No. 50-245 B14403 i
Millstone Nuclear Power Station, Unit No. I Upper Shelf Energy and Reference Temperature Nil Ductility Test Results March 1993
i
^
WELD PLATE CAPSULE MEASURED CALCULATED MEASURED CALCULATED f
~
3 16 12 14 210 Deg.
21 18 10 36 l
300 Deg.
i TABLE I l
% DECREASE IN UPPER SHELF ENERGY WELD PLATE
[
~
CAPSULE MEASURED CALCULATED MEASURED CALCULATED 22 114 61 70 l
210 Deg.
i.
76 133 78 82 j
300 Deg.
4 l
i TABLE II INCREASE IN RT, 9 30 ft-lbs
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