ML20085F122
| ML20085F122 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/30/1995 |
| From: | Branlund B, Caarey R, Caine T GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20085F084 | List: |
| References | |
| GE-NE-523-A164, GE-NE-523-A164-1294, NUDOCS 9506190181 | |
| Download: ML20085F122 (83) | |
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{{#Wiki_filter:___ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ r [ GENuclearEnergy TechnicalServices Business GE-NE-523-A164-1294 GeneralElectric Company, DRF 137-0010-7 - c 175 Curtner Avenue, San Jose, CA 95125 April 1995 I l IIOPE CREEK 1 GENERATING STATION ( RPV SURVEILLANCE MATERIALS TESTING AND FRACTURE TOUGIINESS ANALYSIS f i Prepared by: (Adv R. G. Carey, Engineer Engineering Services Verified by.N D%c_ L B. J. Branlund, Principal Engineer RPV Integrity Approved by: T. A. Caine, Principal Engineer Reactor Internals Management a 9506190181 950609 PDR ADOCK 05000354 P PDR
GE-NE-523-A164-1294 b DRF 137-0010-7 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of Public Service Electric { & Gas Company. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared. The only undertakings of the General Electric Company respecting information in this I document are contained in the contract between the customer and General Electric Company, as identified in the purchase order for this report and nothing contained in this document shall be [ construed as 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 infonnation contained in this document. [ [ [ l i i i [ [ [ - ii -
c GE-NE-523-A164-1294 h, DRF 137-0010-7 CONTENTS [$ g ABSTRACT viii [ ACKNOWLEDGMENTS ix
- 1. INTRODUCTION 1
{-
- 2.
SUMMARY
AND CONCLUSIONS 2 r~ : 2.1 Summary ofResults 2 L :: 2.2 Conclusions 4 4 1
- 3. SURVEILLANCE PROGRAM BACKGROUND 5 I
[ 3.1 3.2 Capsule Recovery RPV Materials and Fabrication Background 5 5 l 3.2.1 FabricationHistory 5 r 3.2.2 Material Properties of RPV at Fabrication 6 L 3.2.3 Specimen Chemical Composition 6 3.3 Specimen Description 6 3.3.1 Charpy Specimens 6 [ 3.3.2 Tensile Specimens 7
- 4. PEAK RPV FLUENCE EVALUATION 13 4.1 Flux Wire Analysis 13
{' 4.1.1 Procedure 13 4.1.2 Results 14 4.2 Determination ofLead Factor 14 4.2.1 Procedure 14 4.2.2 Results 15 4.3 Estimate of 32 EFPY Fluence 16
- 5. CHARPY V-NOTCHIMPACT TESTING 23 5.1 Impact Test Procedure 23 r 5.2 Impact Test Results 24 L 5.3 Irradiated Versus Unirradiated Charpy V-Notch Properties 25 5.4 Comparison to Predicted Irradiation Effects 25 5.4.1 Irradiation Shift 25
[ 5.4.2 Change in USE 26
- 6. TENSILE TESTING 43
{ 6.1 6.2 Procedure Results 43 44 6.3 Irradiated Versus Unitradiated Tensile Properties 44 l [ [ - lii -
r GE-hT-523-A164-1294 L DRF 137-0010-7 CONTENTS r L Pace
- 7. DEVELOPAENT OF OPERATING LIMITS CURVES 51
[ 7.1 Background 51 7.2 Non-Beltline Regions 51 7.3 Core Beltline Region 52 F 7.4 Evaluation ofIrradiation Effects 52 7.4.1 ART Versus EFPY 53 7.4.2 Upper ShelfEnergy at 32 EFPY 53 r 7.5 Operating Limits Curves Valid to 32 EFPY 54
- 8. REFERENCES 64 h E
APPENDICES A. CHARPY SPECIAEN FRACTURE SURFACE PHOTOGRAPHS 66 { B. 72 i EQUIVALENT MARGIN ANALYSIS r L E E E E [ [ [ [ [
- iV -
L - GE-NE-523-A164-1294 L DRF 137-0010-7 TABLES L Iahlg Title hgg 3-1 Chemical Composition of RPV Beltline Materials 8 [ 3-2 Mechanical Properties ofBeltline and Other Selected 9 RPV Materials 3-3 Chemical Composition ofIrradiated Surveillance Specimens 10 [ 4-1 Summary of Condensed Power History 18 L 4-2 Surveillance Capsule Flux and Fluence for 19 Irradiation from Start-up to 3/5/94 r L 5-1 Vallecitos Qualification Test Results Using NIST Standard 27 --- - Reference Specimens 5-2 Unirradiated Charpy V-Notch Impact Test Results 28 { 5-3 Irradiated Charpy V-Notch Impact Test Results 29 [ 5-4 Significant Results ofIrradiated and Unirradiated 30 . Charpy V-Notch Data Tensile Test Results for Irradiated RPV Materials 45 [ 6-1 6-2 Tensile Test Results for Unitradiated RPV Materials 45 6-3 Comparison ofUnirradiated and Irradiated 46 Tensile Properties at Room Temperature ( 6-4 Comparison of Unirradiated and Irradiated Tensile Properties at 550*F 46 7-1 Hope Creek 1 P-T Curve Values 55 { 7-2 Beltline ART Values for Hope Creek 1 58 [ 7-3 Upper Shelf Energy Analysis for Hope Creek i 59 Beltline Material [ [ [ [ [
~
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GE-NE-523-A164-1294 ) DRF 137-0010-7 l ILLUSTRATIONS Fjgute Title Page 3-1 Surveillance Capsule Recovered from Hope Creek 1 11 l l 3-2 Schematic of the RPV Showing Identification 12 l of Vessel Beltline Plates and Welds l 4-1 Schematic ofModel for Azimuthal Flux 20 Distribution Analysis 4-2 Relative Vessel Flux Variation with Angular Position 21 , i 4-3 Relative Vessel Flux Variation with Elevation 22 l 5-1 Hope Creek 1 Unirradiated Base Metal Impact Energy 31 , 1 l 5-2 Hope Creek Irradiated Base Metal Impact Energy 32 4 5-3 Hope Creek 1 Unirradiated and Irradiated Base Metal 33 , Impact Energy 5-4 Hope Creek 1 Unitradiated Base Metal Lateral Expansion 34 l 5-5 Hope Creek 1 Irradiated Base Metal Lateral Expansion 35 5-6 Hope Creek 1 Unitradiated Weld MetalImpact Energy 36 5-7 Hope Creek 1 Irradiated Weld Metal Impact Energy 37 5-8 Hope Creek 1 Unirradiated and Irradiated Weld Metal 38 l Impact Energy 5-9 Hope Creek 1 Unirradiated Weld Metal Lateral Expansion 39 l 5-10 Hope Creek 1 Irradiated Weld Metal Lateral Expansion 40 1 5-11 Hope Creek 1 Unirradiated HAZ Metal Impact Energy 41 1 5-12 Hope Creek 1 Irradiated HAZ MetalImpact Energy 42 l 6-1 Typical Enginc;Jng Stress-Strain for Irradiated 47 RPV Materials lI j 6-2 Fracture Location, Necking Behavior and Fracture 48 l Appearance for Irradiated Base Metal Tensile Specimens 6-3 Fracture Location, Necking Behavior and Fracture 49 Appearance for Irradiated Weld Metal Tensile Specimens 6-4 Fracture Location, Necking Behavior and Fracture 50 Appearance for Irradiated HAZ Metal Tensile Specimens
- vi -
GE-NE-523-A164-1294 DRF 137-0010-7 ILLUSTRATIONS (continued) E. int.c Title Page 7-1 Pressure Test P-T Curves for Hope Creek 1 60 7-2 Heatup/Cooldown P-T Curves for Hope Creek 1 61 7-3 Core Critical Operation P-T Curves for Hope Creek 1 62 7-4 ART for Limiting Beltline Plate and Weld Materials 63 I i I l I l g - vii .
L GE-NE-523-A164-1294 7 L DRF 137-0010-7 g ABSTRACT L The surveillance capsule at 30 azimuth location was removed from the Hope Creek 1 r reactor in Spring 1994. The capsule contained flux wires for neutron fluence measurement and Charpy and tensile test specimens for material property evaluation. The flux wires were evaluated l to determine the fluence experienced by the test specimens. Charpy V-Notch impact testing and l uniaxial tensile testing were performed to establish the properties of the irradiated surveillance p materials. Unirradiated Charpy and tensile specimens were tested as well to obtain the ' appropriate baseline data. The irradiated Charpy data for the plate and weld specimens were compared to the unirradiated data to determine the shift in Charpy curves due to irradiation. The results are within [ the predictions of the Regulatory Guide 1.99 Revision 2. The irradiated tensile data for the plate and weld specimens were compared to the { unirradiated data to determine the effect ofirradiation on the stress-strain relationship of the materials. The majority of the changes shown in the materials were consistent with the irradiation embrittlement effects shown by the Charpy specimens. The flux wire results, combined with the lead factor determined from the last fuel cycle, were used to estimate the 32 EFPY fluence. The resulting estimate was about 14% greater than [ the previous estimate of nominal 32 EFPY fluence; therefore, new pressure-temperature curves were generated. b E E E E [ E
- viii -
I L r GE-hT-523-A164-1294 L DRF 137-0010-7 y ACKNOWLEDGMENTS ~ The author gratefully acknowledges the efforts of other people towards completion of the ( contents of this report. Charpy testing was completed by G. P. Wozadio and G. E. Dunning. Tensile specimen [ testing was done by S. B. Wisner and G. E. Dunning, and chemical composition analysis was r performed by P. C. Wall and R. D. Reager. Flux wire testing was performed by R. Kruger and R. D. Reager. c L I [ r L F L F L [ e
- IX -
GE-NE-523-A164-1294 DRF 137-0010-7
- 1. INTRODUCTION Part of the effort to assure reactor vesselintegrity involves evaluation of the fracture
[ toughness of the vessel ferritic materials. The key values which characterize a material's fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the upper shelf energy (USE). These are defined in 10CFR50 Appendix G [1] and in Appendix G of the ASME { Boiler and Pressure Vessel Code, Section XI[2]. These documents contain requirements used to establish the pressure- temperature operating limits which must be met to avoid brittle fracture. j Appendix H of 10CFR50 [3] and ASTM E185-66 [4] establish the design requirements to be met for surveillance of the Hope Creek I reactor vessel materials. Capsule removal and testing were done per the requirements of ASME El85-82 [6] to the extent practical. The first vessel ( surveillance specimen capsule required by 10CFR50 Appendix H [3] was removed in Spring 1994. The irradiated capsule was sent to the GE Vallecitos Nuclear Center (VNC) for testing. The surveillance capsule contained flux wires for neutron flux monitoring and Charpy V-Notch { impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. The impact and tensile specimens were tested to establish properties for the irradiated materials. Unitradiated tensile specimens were sent from site to GE Nuclear Energy (GE-NE) in San Jose and tested using similar testing procedures. The results of the surveillance specimen testing are presented in this report, as required by l [ 10CFR50 Appendices G and H [1 & 3]. The irradiated material properties are compared to the unirradiated propenies to determine the effect ofirradiation on the tensile properties, through tensile testing, and on material toughness, through Charpy testing. Flux wire results and updated { ' lead factor analyses are used to determine the need for changes to the pressure-temperature (P-T) curves. [
a-c- GE-NE-523-A164-1294 L DRF 137-0010-7 c- 2.
SUMMARY
AND CONCLUSIONS L 2.1
SUMMARY
OF RESULTS The 30 azimuth surveillance capsule was removed and shipped to VNC. The flux wires, { Charpy V-Notch and tensile test specimens removed from the capsule were tested according to ASTM El85-82 [6). The methods and results of the testing are presented as follows: b a. Section 3: Surveillance Program Background
- b. Section 4: Peak RPV Fluence Evaluation
[ b. Section 5: Charpy V-Notch Impact Testing Section 6: Tensile Testing { c.
- d. Section 7: Development of Operating Limits Curves The significant results of the evaluation are below;
- a. The 30 azimuth position capsule was removed from the reactor. The capsule
[ contained 9 flux wires: 3 copper (Cu),3 iron (Fe), and 3 nickel (Ni). There were 36 Charpy V-Notch specimens in the capsule: 12 each of plate material, weld j { material, and heat affected zone (HAZ) material. The 6 tensile specimens removed consisted of 2 plate,2 weld, and 2 HAZ metal specimens. b b. The chemical compositions of copper (Cu) and nickel (Ni) for the irradiated surveillance materials were analyzed (see Table 3-3). The values for the irradiated surveillance plate are 0.09% Cu and 0.66% Ni. The values for the irradiated surveillance weld are 0.06% Cu and 0.46% Ni. [
- c. The purpose of the flux wire testing was to determine the neutron flux at the surveillance capsule location. The flux wire results show that the fluence (from
{ E>l MeV flux) received by the surveillance specimens for 6.01 EFPY was 2 1.42x10" n/cm at removal (see Section 4.3). [
p GE-NE-523-A164-1294 L DRF 137-0010-7 p.
- d. A neutron transport computation was performed, based on the performance of the L last fuel cycle. Relative flux distributions in the azimuthal and axial directions were developed. The lead factor relating the surveillance capsule flux to the peak irside
[ surface flux was 1.01 (see Section 4.2.2). j l
- e. The surveillance Charpy V-Notch specimens were impact tested at temperatures selected to define the transition of the fracture toughness curves of the plate, weld, i and HAZ materials. Measurements were taken of absorbed energy, lateral expansion and percentage shear. From absorbed energy and lateral expansion ]
curve-fit results (for plate and weld metal only), the values ofUSE and ofindex l [ temperature for 30 fi-lb, 50 fi-lb and 35 mils lateral exnansion (MLE) were obtained (see Table 5-4). Fracture surface photographs of each specimen are presented in Appendix A.
- f. The curves ofirradiated Charpy specimens and unitradiated Charpy specimens established the 30 ft-lb index temperature irradiation shift. The plate material showed a measured shift of 4 F and the weld material showed a measured shift of
[ 61"F. The measured shifts for the plate and for the weld (for a fluence of 1.42x10" n/cm 2) were within their respective Reg. Guide 1.99 [7] range predictions (ARTNDT 20) of-26 F to 42 F and -41 F to 71*F, respectively (see { Section 5.4.1).
- g. The measured decrease in USE of 14% for the plate material compares to a Reg.
Guide 1.99 prediction of 7%. The measured decrease in USE of 5% for the weld [ material compares to a Reg. Guide 1.99 prediction of 8.5% (see Section 5.4.2). '
- h. The irradiated tensile specimens were tested at room temperature (70 F), and reactor operating temperature (550 F). The results in comparison to unirradiated data were tabulated (see Tables 6-3 and 6-4) for each specimen including yield and ultimate tensile strength, uniform and total elongation, and reduction of area. The results generally showed increasing strength and decreasing ductility, consistent
[ with expectations for irradiation embrittlement.
- i. The 32 EFPY peak vessel inside surface fluence prediction is 7.5x10" n/cm2,
{ based on the flux wire test and lead factor results (see Section 4.3). This is about 14% greater than the previously determined nominal value (6.6x10" n/cm 2) using the first cycle dosimetry [11). [
GE-NE-523-A164-1294 u DRF 137-0010-7
- j. As a part of the development of the pressure-temperature (P-T) operating limits b curves, the adjusted reference temperature (ART = initial RTNDT + ARTNDT+
Margin) was predicted for each beltline material, based on the methods of [ Reg. Guide 1.99. The ART for the limiting material, plate heat 5K3025/1, at 32 EFPY is 72.5'F.
- k. The beltline material USE values at 32 EFPY were predicted using the methods of Reg. Guide 1.99, with initial beltline USE values based on estimated USE values for the plates and 10 F test results for the welds (see Table 7-3). It is expected that the actual 32 EFPY USE will be in excess of 50 ft-lbs for all beltline plates I b and welds. In addition, the results of the USE testing for the surveillance materials show that the BWROG Equivalent Margin Analysis is bounding.
i
- 1. P-T curves were developed for three reactor conditions: pressure test (Curve A),
non-nuclear heatup and cooldown (Curve B), and core critical operation (Curve C). The P-T curves, shown in Figures 7-1,2, and 3, are valid for 32 EFPY ofoperation. b !
2.2 CONCLUSION
S b The requirements of 10CFR50 Appendix G [1] deal basically with vessel design life [ conditions and with limits of operation designed to prevent brittle fracture. Based on the evaluation of surveillance testing results, and the associated analyses, the following conclusions are made: {
- a. The measured 30 ft-lb shifts are within the Regulatory Guide 1.99, Revision 2 b predictions, as is the measured USE decrease for the surveillance weld. The measured USE decrease for the surveillance plate was greater than predicted.
- b. The values of ART and USE for the reactor vessel beltline materials are shown by calculation to remain within limits in 10CFR50 Appendix G [1] for at least 32 EFPY of operation.
b b
GE-NE-523-A164-1294 DRF 137-0010-7
- 3. SURVEILLANCE PROGRAM BACKGROUND 3.1 CAPSULE RECOVERY The reactor pressure vessel (RPV) surveillance program consist of three surveillance capsules at the 30 ,120 , 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 is located at a similar positions relative to the core geometry because of core symmetry; therefore, each capsule is expected to receive equalirradiation. During the Spring 1994 outage, the 30 positioned capsule was removed. The capsule was cut from its holder assembly and shipped by usk to the GE Vallecitos Nuclear Center (VNC), where testing was performed.
Upon arrival at VNC, the capsule basket was examined for identification. The Hope Creek I reactor code number is 57, as specified in GE drawing 105D4714. The holder part number 131C7717G6 as specified in GE drawing 105D4719G7 from master parts list 238XI11 AC was stamped on the basket and provided positive identification as the 30 surveillance capsule materials for Hope Creek 1. The general condition of the basket as received is shown in Figure 3-1. The basket contained three impact (Charpy) specimen capsules and three tensile specimen capsules. Each tensile specimen capsule contained two tensile specimens. Each Charpy specimen capsule contained 12 plate, weld, or HAZ Charpy specimens and 3 flux wires (one iron, one copper, and one nickel) in a sealed helium environment. 3.2 RPV MATERIALS AND FABRICATION BACKGROUND 3.2.1 Fabrication History The Hope Creek 1 RPV is a 251 inch diameter BWR/4 design. Construction was performed by Babcock-Hitachi to the Winter 1968 edition of the AShE Code Section III with Winter 1969 Addenda. The shell and head plate materials are AShE SA533, Grade B, Class I low alloy steel (LAS). The nozzles and closure flanges are AShE SA508 Class 2 LAS, and the closure flange bolting materials are AShE SA540 Grade B24 LAS. The plates and welds in the beltline region of the RPV are shown in Figure 3-2. The vessel plates were heat treated prior to welding: austenitized typically at 1580 F-1635 F for approximately 3.5 hours and tempered I typically at 1200 F-1256 F for approximately 3.5 hours. The post weld heat treat was typically 40.5 hours minimum at temperatures between 1100 F-1130 F. Submerged arc or shielded metal GE-NE-523-A164-1294 DRF 137-0010-7 arc welding of plates was followed by post weld heat treatment typically at 1112 F-1170 F for at least 40 hours. 1 3.2.2 Material Pronerties of RPV at Fabrication The chemical and mechanical properties of the vessel materials were retrieved from information documented in the response to Generic Letter 92-01 [15], the UFSAR [5] and the I Tech . Spec. [9]. Table 3-1 shows the chemistry data for the beltline materials. Properties of the beltline materials and other locations ofinterest are presented in Table 3-2. 3.2.3 Specimen Chemical Composition l Samples were taken from the irradiated surveillance plate and weld tensile specimens after they were tested. Chemical analyses were performed using a Spectraspan III plasma emission spectrometer. Each sample was etched then dissolved in an acid solution to a specific concentration for each gram of metal. The spectrometer was calibrated for determination ofMn, Ni, Mo, and Cu by diluting National Institute of Standards and Technology (NIST) Spectrometric Standard Solutions. The phosphorus calibration involved analysis of four reference materials from NIST with known phosphorus levels. Analysis accuracies are 10.005% (absolute) of reported value for phosphorus and 15% (relative) of reported value for other elements. The chemical composition results are given in Table 3-3 for irradiated surveillance plate and weld I materials. The results show good agreement with corresponding unirradiated surveillance plate data, but the corresponding unirradiated surveillance weld data have higher %Wt content for most elements. 'l 3.3 SPECIMEN DESCRIPTION The surveillance capsule holder contained 36 Charpy specimens: base metal (12), weld metal (12), and HAZ (12). There were 6 tensile specimens: base metal (2), weld metal (2), and IIAZ (2). The holder contained 9 flux wires: 3 iron,3 nickel and 3 copper. The chemistry and fabrication history for the Charpy and tensile specimens are described in this section. All materials used for surveillance were beltline materials. 3.3.1 Charov Specimens The fabrication of the Charpy specimens is described in the Babcock-Hitachi drawing [8] on the surveillance test program.
E GE-NE-523-A164-1294 u DRF 137-0010-7 y The base metal specimens were cut from Heat 5K3238/1. The test plates received a E similar heat treatment as the fabrication plates for Heat 5K3238/l per requirement of GE document 21 A8707 [16], including the post weld heat treatment. Specimens were machined from the 1/4 T and 3/4 T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction) to the dimensions in GE drawing 158B7977. The base metal Charpy specimens { from the surveillance capsule were stamped on one end with the reactor number, 57, and on the other end with "Pl", which designates base metal. E The weld metal and HAZ Charpy specimens were fabricated from two pieces of the E surveillance test plate Heat 5K3238/l are welded together with a weld which was specified to be identical to the beltline longitudinal seam welds. The general welding spec. used was RS G576 [ and the proc. qual. test spec. no. was RS 9551. Actual welding records obtained from Babcock- { Hitachi show the surveillance weld to be submerged arc welding Heat / Lot D53040/1125-02205, except at the root of the weld. However, weld metal specimens were fabricated away from the root of the weld. This heat and tot number is the same as that of a beltline weld. The welded test ' plate for the weld and HAZ Charpy specimens received a similar post weld heat treatment as the vessel material, per GE doucment 21 A8707 [16]. The base metal orientation in the weld and l HAZ specimens was longitudinal. The specimens were machined to the dimensions in GE drawing 158B7977 and stamped on one end with the reactor number and on the other with "P3" for weld metal or "P4" for HAZ. l 3.3.2 Tensile Soecimens Fabrication of the surveillance tensile specimens is also described in the Babcock-Hitachi surveillance specimen drawings [8]; the specimens are machined to the dimensions in GE drawing 166B7062. The materials, and thus the chemical compositions and heat treatments for the base, weld, and HAZ tensiles are the same as those for the corresponding Charpy specimens. The identifications of the base, weld, and HAZ tensile specimens are: reactor number 57 on one end and P1, P2, or P3 on the other end. The base metal specimens were machined from material at the 1/4 T and 3/4 T depth. The specimens were oriented along the plate rolling direction. The gage section was tapered to a minimum 0 0.250 inch at the center. The weld specimens were machined entirely from weld metal, scrapping material that might include base metal. The HAZ specimen blanks were cut from the welded test plates such that the gage section minimum diameters were machined at the weld fusion line. The fmished HAZ specimens are approximately half weld metal and half base metal oriented along the plate rolling direction.
M M M M M M Table 3-1 CllEMICAL COMPOSITION OF RPV BELTLINE MATERIALS Composition by Weight Percent Identification Heat No. C Mn P S Si Ni Mo Cu Lower Plates: 6C45/l 0.18 1.49 0.008 0.010 0.31 0.57 0.50 0.08 Shell Course No. 5 6C35/1 0.20 1.46 0.010 0.011 0.27 0.54 0.51 0.09 SK3230/1 0.19 1.44 0.010 0.012 0.30 0.56 0.50 0.07 Lower-Intermediate 5K2963/1 0.22 1.43 0.009 0.008 0.29 0.58 0.59 0.07 Plates: SK2530/1 0.20 1.43 0.010 0.008 0.30 0.56 0.54 0.08 Shell Course No. 4 SK3238/l 0.20 1.45 0.012 0.008 0.31 0.63 0.56 0.09 Intermediate Plates SK2698/l 0.21 1.41 0.010 0.010 0.30 0.58 0.56 0.10 Shell Course No. 3 SK2608/l 0.19 1.46 0.009 0.014 0.30 0.58 0.53 0.09 SK3025/1 0.17 1.46 0.012 0.009 0.30 0.71 0.52 0.15 Surveillance Plate: SK3238/l 0.20 1.42 0.012 0.012 0.29 0.62 0.54 0.09 7 Welds: 510-01205 0.072 1.2 0.01 0.011 0.42 0.54 0.45 0.09 519-01205 0.051 1.17 0.01 0.007 0.26 0.53 0.45 0.01 504-01205 0.06 1.3 0.011 0.005 0.26 0.51 0.41 0.01 001-01205 0.058 1.26 0.012 0.006 0.25 0.51 0.40 0.02 D53040 (Flux 1810-02205) 0.088 1.69 0.012 0.003 0.34 0.68 0.45 0.10 D53040 (Flux 1125-02205) 0.01 1.68 0.01 0.004 0.3 0.63 0.43 0.08 D55733 (Flux 1810-02205) 0.076 1.57 0.013 0.003 0.34 0.68 0.43 0.1 Surveillance Weld: D53040 (Flux 1125-02205) 0.11 1.69 0.012 0.005 0.31 0.68 0.51 0.09 O si n DU c$ q-h?
- s s
.L'E
GE-NE-523-A164-1294 DRF 137-0010-7 Table 3-2 MECHANICAL PROPERTIES OF BELTLINE AND OTHER SELECTED RPV MATERIALS Initial Heat RTndt Location Number fEl Beltlinea: Lower Plates 5K3230/1 -10 (Shell Course No. 5) 6C35/1 -11 6C45/1 1 Lower Intermediate Plates 5K2963/1 -10 I (Shell Course No. 4) 5K2530/1 5K3238/1 19 7 Intermediate Plates 5K3025/1 19 I (Shell Course No. 3) LPCI Nozzle 5K2608/1 5K2698/1 19468/1 19 19
-20 10024/1 -20 Welds 001-01205 -40 510-01205 -40 D53040/1125-02205 -30 519-01205 -49 504-01205 -31 I D55733/1810-02205 D53040/1810-02205
-40
-49 Non-Beltline b; I Vessel Flange 10 i
Top Head Flange 10 Top Head Torus 19 Feedwater Nozzle -20 Shell Ring Connected to Vessel Flange 19 Bottom Head Dome 30 Bottom Head Torus 30 Closure Studs met 45 fi-lb and 25 mils @l0*F a Test data information from UFSAR [5] ; g bTest data information from Tech. Spec. [9] y l 9 g
GE-NE-523-A164-1294 DRF 137-0010-7 Table 3-3 CHEMICAL COMPOSITION OFIRRADIATED AND UFRRADIATED SURVEILLANCE SPECIMENS Composition by Weight Percent Identification Mn P Cr Ni Mo Cu PLATE: Tensile PIA (RT F) 1.31 0.011 0.14 0.62 0.54 0.08 Tensile PIB (550 F) 1.34 0.011 0.10 0.70 0.58 0.09 Average Tensile 0.66 0.09 Beltline Plate a 5K3238/1 1.45 0.012 -- 0.63 0.56 0.09 Surveillance Plate b 5K3238/1 1.42 0.012 -- 0.62 0.54 0.09 Average Plate Chemistry c 0.64 0.09 WELD: Tensile P2A (RT F) 1.13 0.016 0.10 0.41 0.30 0.05 Tensile P2B (550 F) 1.41 0.017 0.12 0.50 0.34 0.06 Average Tensile 0.46 0.06 Beltline Weld a 1.68 0.010 - 0.63 0.43 0.08 D53040, Flux 1125-02205 Surveillance Weld b 1.69 0.012 -- 0.68 0.51 0.09 D53040, Flux 1125-02205 Average Weld Chemistry c 0.59 0.08 I a Beltline specimen data from Table 3-1 has been added fbr comparison b Ur. irradiated surveillance material from Table 3-1 c Average chemistry = (Average Tensile + Beltline + Surveillance)/3 l I i l GE-NE-523-A164-1294 L DRF 137-0010-7 r L r i I. wak n r s . L ng:::l.Nc. . - : 3. ..
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. . a . . ., .. -{-(,i,9 A hh d i , s,.,a h Figure 3-1. Smveillance Capsule Holder Recovered From Hope Creek (Holder marked as "131C7717G6)
GE-NE-523-A164-1294 ! DRF 137-0010-7 l Vessel Flange ! Upper Shell T Longitudinal Welds
=
Upper Shell UpperInterinediate Shell _ Assembly _f I X y N._- Shell Course 3 I ( : Intermediate Shell : Plate Heats: 5K3025/1 SK2608/1 _f 5K2698/1 I LPCI Nonle ' 5 J , , , - -
~
h ' Shell Course 4 Plate Heats: SK2963/1 Core Lower Intermediate Shell 5K2530/1 Beltline SK3238/1 Region Lower Shell - Assembly l / Lower Shell : Shell Course 5 QInte Heats: 5K3230/1 6C35/1 s 6C45/1
, , , 1 Bottom Head Enclosure I
I Figure 3-2. Schematic of RPV Showing Identification of Vessel Beltline Plates and Welds GE-NE-523-A164-1294 DRF 137-0010-7
- 4. PEAK RPV FLUENCE EVALUATION Flux wires removed from the 30 capsule were analyzed, as described in Section 4.1, to determine flux and fluence received by the surveillance capsule. The lead factor, determined as described in Section 4.2, was used to establish the peak vessel fluence from the flux wire results.
Section 4.3 includes 32 EFPY peak fluence estimates at the surface of the vessel and at the 1/4 T location. 4.1 FLUX WIRE ANALYSIS I 4.1.1 Proc _qdure The surveillance capsule contained 9 flux wires: 3 iron,3 copper, and 3 nickel. Each wire was removed from the capsule, cleaned with dilute acid, weighed, mounted on a counting card, and analyzed for its radioactivity content by gamma spectrometry. Each iron wire was analyzed for Mn-54 content, each nickel wire for Co-58 and each copper wire for Co-60 at a calibrated 4-cm or 10-cm source-to-detector distance with 100-cc Ge(Li) and 170-cc Ge 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. From the flux wire activity measurements and power history, reaction rates for Fe-54 (n,p) Mn-54, Cu-63 (n,a) Co-60 and Ni-58 (n,p) Co-58 were calculated. The E>l MeV I 1 fast flux reaction cross sections were determined from past testing at Browns Ferry 3 [10], also a 251 inch, 764 bundle plant, using multiple dosimeter and spectrum unfolding techniques. The cross sections for the iron, copper, and nickel wires are 0.213 barn,0.00374 barn and 0.274 barn, respectively. These values are consistent with other measured cross section functions determined at VNC from more than 65 spectral determinations for BWRs and for the General Electric Test j Reactor using activation mordters and spectral unfolding techniques. These data functions are applied to BWR pressure vessel locations based on water gap (fuel to vessel wall) distances. The cross sections for E>0.1 MeV flux were determined from the measured 1-to-0.1 MeV cross
!ection ratio of 1.6.
GE-NE-523-A164-1294 DRF 137-0010-7 4.1.2 Besults The measured activity, reaction rate and full-power flux results for the 30 surveillance i capsule are given in Table 4-2. The E>l MeV flux values were calculated by dividing the wire reaction rate measurements by the corresponding cross sections, factoring in the local power history for each fuel cycle. The fluence result,1.42x1017 n/cm2(E>l MeV) was obtained by I multiplying the full-power flux value for the average ofiron and copper by the operating time and full power fraction, shown in Table 4-1. The accuracies of the values in Tables 4-2 for a 2o deviation are estimated to be: 1 5% for dps/g (disintegrations per second per gram) 10% for dps/ nucleus (saturated) 120% for flux and fluence E>l MeV 20% for flux and fluence E>0.1 MeV 4.2 DETERMINATION OF LEAD FACTOR 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 l 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 the ratio of the flux at the surveillance capsule to the flux at the peak inside surface location. The lead factor is a fimetion of the core and vessel geometry and of the distribution of bundles in the core. The lead factor was generated for the vessel geometry, using a typical fuel cycle to determine power shape and void distribution. The methods used to calculate the lead factor are discussed below. 4.2.1 Procedure Determination of the lead factor for the RPV inside wall was made using a combination of two separate two-dimensional finite difference computer analyses. The first of these established the relative azimuthal variation of 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 distribution results were combined to provide the ratio of flux, or the lead factor, between the surveillance capsule location and the peak flux locations. I
~"~
l
{ GE-NE-523-A164-1294 DRF 137-0010-7 The DORT computer program, which utilizes the discrete ordinates method to solve the b Boltzmann transport equation in two dimensions, was used to calculate the spatial flux l distribution produced by a fixed source of neutrons in the core region. The azimuthal distribution ( was obtained with a model specified in (R,0) geometry, assuming eighth-core symmetry with I reflective boundary conditions at 0 and 45* 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,0) vessel model is shown in Figure 4-1. A total of 131 radial intervals and 90 azimuthal intervals were used. The model consists of an inner and outer core region, the shroud, water regions inside and outside the shroud, and the vessel wall. The core region material compositions and neutron source densities were representative of conditions at an elevation 75 inches above the bottom of active fuel, which is 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 30 { The calculation of the axial flux distribution was performed in (R,Z) geome , using s [ simplified cylindrical representation of the core configuration and rei Estic simulations of 6 . dal variations of power density and coolant mass density. The core description was based on b conditions near the azimuth angle of 25 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. [ 4.2.2 Results { The two-dimensional computations indicate the flux to be a maximum 25.25 past the RPV quadrant references @ 'O , etc.), at an elevation about 82 inches above the bottom of active fuel. The peak closest to the 30 location of the surveillance capsule removed is at 25.25 , [ as shown in Figure 4-2. The relative flux distribution versus elevation is shown in Figure 4-3. The distribution calculations establish the lead factor between the surveillance capsule location [ and the peak location at the inner vessel wall. The calculated flux at the capsule (R,0) position along the midplane was modified by an appropriate ratio derived from the (R,Z) model to account for the actual capsule elevation position. The resulting computed surveillance capsule flux is 9.4x108 n/cm 2-s. The peak flux at vessel surface from the transport calculation, incorporating modification due to differences between (R,0) and (R,Z) models, is 9.3x10 8 n/cm2 -s. Therefore the lead factor is 9.4/9.3=1.01. [
GE-NE-523-A164-1294 DRF 137-0010-7 The transport calculation of surveillance capsule flux,9.4x108 n/cm 2-s, is about 25% higher than the measured (flux wire analysis) surveillance capsule result of 7.5x108 n/cm2-s. This may be due to the conservative methods of the transport computation and to uncertainties in [ physical dimensions, such as using nominal rather than actual vessel radius. The uncertainty l associated with vessel radius has little, if any, effect on the lead factor, since for the lead factor the { fluence values at two locations with the same radius are being ratioed. The lead factor computed here was based on fuel cycles 1 through 5 whereas that determined in [11] was based on fuel cycle 1 only. The fracture toughness analysis is based on a 1/4 T depth flaw in the beltline region, so the attenuation of the flux to that depth is considered. This attenuation is calculated according to Reg. Guide 1.99 requirements, as shown in the next section. 4.3 ESTIMATE OF 32 EFPY FLUENCE The inside surface fluence (fsurf) at 32 EFPY is determined from the flux wire fluence for 6.01 EFPY of 1.42x1017 n/cm2, using the lead factor of 1.01. The time period 32 EFPY is based on 40-year operation at an 80% capacity factor. The resulting 32 EFPY fluence value at the peak vesselinside surface is: fsurf = 1.42x1017*(32/6.01)/1.01 2 fsurf = 7.5x10 17 n/cm b This fluence represents the peak value, which occurs in the lower-intermediate shell (No.4) and will be conservatively applied to lower shell (No.5). The peak surface fluence is about 14% greater than the nominal value (6.6 x1017 n/cm2) calculated from the first cycle dosimetry [11]; however, within the 20% accuracy expected as reported in Section 4.1.2. The bottom of the b intermediate shell (No.3) is 143.2 inches from the bottom of active fuel. The fluence at this elevation can be conservatively calculated using information in Figure 4-3 as: 2 fsurf = 7.5x1017*(0.66) = 4.94 x1017 n/cm for the intermediate shell E The fluence at the LPCI nozzle in the intermediate shell (No. 3) has an additional relative decrease in fluence due to the azimuthal variation shown in Figure 4-2. Since a LPCI nozzle is at a relative [ u-___, _ _ _ _ , _ _ _ _ , .-_ _ . _ _ __ _ - . - . - _ . _ _ _ _ . . - _ . - . - . . _ _ _ _ _ _ _ - - _ - _ . _ _ . _ _ _ , - . _ _ - - - - _ _ - - - _ _ - - - . - - - - _ _ - - _ - - - _ _ ---_,.__,-,____,__,.,_,---
GE-NE-523-A164-1294 DRF 137-0010-7 I angular position of 35 , the fluence at the LPCI nozzle is calculated as follows: i i l l 2 fsurf = 7.5x1017*(0.66)*(0.82) = 4.05x1017 n/cm for the LPCI nozzle l The 1/4 T fluence (f) is calculated as follows, according to the Reg. Guide 1.99 [7]: f = fsurf*(e-0.24x) (4 1), where x = distance, in inches, to the 1/4 T depth. The vessel beltline lower-intermediate shell and intermediate shell are 6.10 inches thick. The corresponding depth x is 1.53 inches. Equation 4-1 yields: 2 f = fsurf(0.69) = 5.2x1017 n/cm for the lower-intermediate shell I f = fsurf(0.69) = 3.42x10 17 n/cm 2 for the intermediate shell 2 f=fsurf(0.69) = 2.81x10 17 n/cm for the LPCI nozzle The impact of these revised fluences on the P-T curves is discussed in Section 7. I I I GE-NE-523-A164-1294 DRF 137-0010-7 Table 4-1
SUMMARY
OF CONDENSED POWER HISTORY Operating Full Power Cycle Cycle Dates Days Fraction 1 10/12/86 - 02/13/88 453 0.825 2 04/16/88 - 09/15/89 518 0.892 3 11/20/89 - 12/16/90 392 0.930 4 02/22/91 - 09/12/92 564 0.943 5 11/12/92 - 03/05/94 473 0.977 2400 (total) 0.914 (aveage) I I ' I v r r~m v rm rm . rm . rm rm r t_- r m rm rm Table 4-2 SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR 1RRADIATION FROM START-UP TO 3/5/94 1 Full Power Flux a Full Power Flux b Fluence Fluence b dps/g Element Reaction Rate (n/cm2-s) (n/cm2-s) (n/cm2) (n/cm2 ) Wire (Element) (at end ofIrradiation) [dps/ nucleus (saturated)] >l MeV >0.1 MeV >l MeV >0.1 MeV Copper G9 9.472 E3 2.801 E-18 7.489 E8 11.982 E8 1.42 E17 2.27 E17 Copper G11 9.306 E3 2.752 E-18 7.358 E8 11.773 E8 1.39 El7 2.22 E17 Copper G12 9.152 E3 2.706 E-18 7.235 E8 11.576 E8 1.37 E17 2.19 E17 Iron G9 8.920 E4 1.595 E-16 7.488 E8 11.981 E8 1.42 El7 2.27 E17 Iron G11 9.079 E4 1.624 E-16 7.624 E8 12.198 E8 1.45 E17 2.32 E17 j Iron G12 9.017 E4 1.613 E-16 7.573 E8 12.117 E8 1.44 E17 2.30 E17 I Nickel G9 1.306 E6 1.942 E-16 7.088 E8 11.341 E8 1.34 E17 2.14 E17 Nickel G11 1.342 E6 1.996 E-16 7.285 E8 11.656 E8 1.38 E17 2.21 E17. Nickel G12 1.306 E6 1.942 E-16 7.088 E8 11.341 E8 1.34 E17 2.14 E17 Average c= 1.42 E17 2.27 E17 10.06 E17 10.10E17 O a Full power flux, based on thermal power of 3293 MWt si D 0. b 1.6 times the E >l MeV results @d c Nickel is not included in final average c $.
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- 40 INTERVALS 90 INTERVALS ~~N I ERVALS IN AZIMUTHAL
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- 2. cone ExTEnion rutt Figure 4-1. Schematic of Model for Azimuthal Flux Distnbution Analysis
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Biw GE-NE-523-A164-1294 [ DRF 137-0010-7
- 5. CHARPY V-NOTCH IMPACT TESTING Ti.e 36 Charpy specimens recovered from the surveillance capsule were impact tested at L temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials. In addition, longitudinal unirradiated base, weld, and longitudinal HAZ metal specimens recovered from the Hope Creek site were tested for baseline data. Testing was conducted in accordance with ASTM E23-88 [12].
5.1 IMPACT TEST PROCEDURE The Vallecitos testing machine used for unirradiated and irradiated specimens was a Riehle Model Pl-2 impact machine, serial number R-89916. The maximum energy capacity of the [ machine is 240 fi-lb, and the test velocity at impact is 15.44 ft/sec. The test apparatus and operator were qualified using NIST standard reference material { specimens. The standards consist of sets ofhigh and low energy specimens, each designed to fail at a specified energy at the standard test temperature of-40 F. According to ASTM E23-88 [12], the test apparatus averaged results must reproduce the NIST standard values within an accuracy of 5% or 11.0 fl-lb, whichever is greater. The qualification of the Riehle machine and operator is summarized in Table 5-1. Charpy V-Notch tests were conducted at temperatures between -100 F and 300 F. The cooling fluid used for both irradiated and unirradiated specimens tested at temperatures below 70 F was ethyl alcohol. At temperatures between 70 F and 200 F, water was used as the { temperature conditioning fluid. The specimens were heated in silicon oil above 200 F. Cooling of the conditioning fluids was done by heat exchange with liquid nitrogen; heating was done by an immersion heater. The liquid bath was mechanically stirred to maintain uniform temperatures. The bath temperature was measured with a calibrated thermocouple. After equilibrium at the test b temperature for at least 5 minutes, the specimens were manually transferred with centering tongs to the Charpy test machine and impacted in less than 5 seconds. [ [
m GE-NE-523-A164-1294 L DRF 137-0010-7 , For each Charpy V-Notch specimen the test temperature, energy absorbed, lateral L expansion, and percent shear were evaluated. In addition, for the irradiated specimens, photographs were taken of fracture surfaces. Lateral expansion was measured according to l specified methods [12]. Percent shear was determined using method number 1 of Subsection l 11.2.4.3 of ASTM E23-88 [12], which involved measuring the length and width of the cleavage surface in inches and determining the percent shear value from Table 2 of ASTM E23-88 [12]. 5.2 IMPACT TEST RESULTS E Thirty-five unirradiated Charpy V-Notch specimens of 12 base,1I weld, and 12 HAZ E l materials were tested at temperatures (-100 F to 300 F) selected to define the toughness l transition and upper shelf portion of the fracture toughness curves. The absorbed energy, lateral L expansion, and percent shear data are listed in Table 5-2. Plots of absorbed energy data for unirradiated base, weld, and HAZ metals are presented in Figures 5-1, 5-6, and 5-11, respectively. Lateral expansion plots for base and weld metals are presented in Figures 5-4 and 5-9, { respectively. l ' Twelve Charpy V-Notch specimens each ofirradiated base, weld, and HAZ material were tested at temperatures (-100 F to 300 F) selected to define the toughness transition and V l upper shelf portions of the fracture toughness curves. One HAZ specimen (specimen 619) partially broke by ductile tearing at 233 ft-lb (see Appendix A). Ti;e absorbed energy, lateral expansion, and percent shear data are listed for each material in Table 5-2. Picts of absorbed [ energy data for irradiated base, weld and HAZ materials are presented in Figures 5-2, 5-7, and 5-12, respectively. Lateral expansion plots for base and weld materials are presented in Figures I 5-5 and 5-10, respectively. The irradiated impact energy curves are plotted along with their corresponding unirradiated curve in Figures 5-3 and 5-8 for base and weld specimens, respectively. The fracture surface photographs and a summary of the test results for each specimen are contained in Appendix A. The plate and weld data sets are fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook [13]: Y = A + B
- TANH [( T - To )/C],
where Y = impact energy or lateral expansion, T = test temperature, and A, B, To and C are determined by non-linear regression.
I GE-NE-523-A164-1294 DRF 137-0010-7 The TANH function is one of the few continuous functions with a shape characteristic of low alloy steel fracture toughness transition curves. The curve fits were generated by fixing the lower shelf to 2.2 ft-lb for Cliarpy energy and 1.0 mil for lateral expansion and setting the upper shelffree. I 5.3 IRRADIATED VERSUS UNIRRADIATED CHARPY V-NOTCH PROPERTIES The irradiated and unirradiated Charpy V-Notch data curves were used to estimate the values given in Table 5-4: 30 ft-lb,50 n-lb and 35 MLE index temperatures. Transition i temperature shift values are determined as the change in the temperature at which 30 n-lb impact energy is achieved, as required in ASTM E185-82 [6]. The resulting shins in Charpy curves are discussed in the next section. The USE values are determined by averaging the Charpy data at 100% shear. 5.4 COMPARISON TO PREDICTED IRRADIATION EFFECTS 5.4.1 Irradiation Shift The measured transition temperature shifts for the plate and weld materials were compared to the predictions calculated according to Regulatory Guide 1.99, Revision 2 [7]. The inputs and calculated values for irradiated shift are as follows: Plate: Copper = 0.09 % Nickel = 0.64 % CF= 58 } fluence = 1.42x1017 niem2 Reg. Guide 1.99 ARTNDT = 8 F Reg. Guide 1.99 ARTNDT 2cA(34 F) = -26 F min,42 F max. Measured 30 R-lb shift = 4 F Weld: Copper = 0.08 % Nickel = 0.59% CF = 105 fluence = 1.42x1017 n/cm2 Reg. Guide 1.99 ARTNDT = 15 F Reg. Guide 1.99 ARTNDT 2cA(56 F) = -41 F min,71 F max. Measured 30 R-lb shia = 61 F
GE-NE-523-A164-1294 DRF 137-0010-7 The weigb' pcreents of Cu and Ni are the average of the plate and weld results in Table 3-3. CF s' above is the chemistry factors from Tables 1 or 2 ofReg. Guide 1.s (7). The fluence factor from Figure 1 of[7] is 0.14. The measured shifts of 4 F for the plate ar.d 61 F for the weld are below and above the predicted shifts of 8 F and 15 F, respectively and are within the bounds of the Reg. Guide 1.99 prediction including uncertainty of 2a. 5.4.2 Change in USE Figure 2 of Reg. Guide 1.99 was used with copper and fluence data above to predict decrease in USE of 7% for the plate and decrease in USE of approximately 8% for the weld. The measured decrease in USE is 19 ft-lb (14% decrease) for the plate; this value is greater than the Reg. Guide 1.99 prediction. The weld material shows a measured decrease in USE of 8 ft-lb (5% decrease) which is less than the Reg. Guide 1.99 prediction. I I GE-NE-523-A164-1294 DRF 137-0010-7 Table 5-1 VALLECITOS QUALIFICATION TEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS Test Energy Acceptable Specimen Bath Temperature Absorbed Range Ic entification Medium ('F) .(ft-lb) (ft-lb) Vallecitos HH-40 229 Eth 1 Alcohol -40 75.0 Richle Machine HH-40 384 Eth 1 Alcohol -40 74.5 (tested 6/28/94) HH-40 980 Eth 1 Alcohol -40 70.5 HH-401152 Ethyl Alcohol -40 72.5 I HH-401172 Ethyl Alcohol -40 Average 75.0 73.5 74.9 3.7 pass LL-39 080 Eth 1 Alcohol -40 13.5 LL-39 095 Eth 1 Alcohol -40 13.0 LL-39 631 Eth 1 Alcohol -40 13.5 LL-39 775 Eth 1 Alcohol -40 13.5 LL-39 930 Ethyl Alcohol -40 JJ 0 Average 13.3 13.2 1.0 pass I I I I l
~ GE-NE-523-A164-1294 L DRF 137-0010-7 Table 5-2 UNIRRADIATED CIIARPY V-NOTCH IMPACT TEST RESULTS u - Test Fracture Lateral Percent Shear . [ Specimen Type S ? ecimen
- Ic entification Temperature
("F) Energy (ft-lb) Expansion (mils) (Method 1) (%) F Base: 4 -80 10 15 10 L lleat 5K3238/1, 3 -40 12.5 15 8 Longitudinal 12 -10 52.5 41 29 p 2 0 34.5 36.5 24 L 11 20 62.5 49 43 1 40 30 35 38 9 50 74.5 57 53 10 70 99.5 75 67 5 80 113.5 89 87 6 120 139.5 94 100 7 200 139.5 92.5 100 8 300 141 92 100 [ Weld: lleats I)53040 4 9
-80
-60 11.5 27.5 14 26.5 15 27 Flux 1125-02205 3 -40 36.5 32 21 e I 10 -20 78.5 66 45 L 2 0 89 63 55 11 10 58 49 37 r 1 40 102 70 74 L 5 80 118.5 77 80 6 120 153 89 100 7 200 151.5 96.5 100 8 300 187 84 100 HAZ: 9 -100 16.5 11.5 11
[ Longitudinal 4 3
-80
-40 106.5 13.5 71.5 14.5 54 13 11 -10 136.5 81 74
( 2 12 20 0 137.5 160 88 81 72 100 1 40 98.5 64 82 5 80 160.5 83 100 { 6 120 104 73 100 7 200 205 80.5 100 10 200 169 85 100 [ 8 300 123 71.5 100 arbitrary numbering of specimen I.D. [
c GE-NE-523-A164-1294 L DRF 137-0010-7 Table 5-3 L IRRADIATED CHARPY V-NOTCH IMPACT TEST RESULTS E ' Test Fracture Lateral Percent Shear Specimen Temperature Energy E ansion (Method 1) Identification (*F) (ft-lb) ( ls) (%) Base: 614 -80 7 6.5 3 p Heat 5K3238/1, 604 -50 8.5 10 6 L Longitudinal, 603 -40 32.5 29.9 15 f=1.42x10" n/cm2 606 -20 7 8.5 18 612 0 42 37 16 607 40 55 50.5 52 610 60 97 77 60 611 70 105 84.5 79 609 80 115 78 85 608 120 113.5 80 100 613 200 115.5 81 100 605 300 133 92 100 Weld: 594 -80 5.5 9.5 12 Heats, D53040 597 -40 22.5 20 17 Flux 1125-02205, 595 -10 31 27 26 f=1.42x10" n/cm2 592 0 19.5 22.5 34 r 596 20 59.5 48.5 39 L 599 40 88.5 72 50 601 50 61 44 58 593 80 112.5 69 77 b 598 120 147.5 92 100 591 150 154.5 90 100 602 200 144.5 91 100 [ 600 300 178.5 88 100 HAZ: 616 -100 70 48 33 [ Longitudinal f=1.42x10" n/cm2 624 626
-80
-40 85 9.5 59 10 44 26 623 -10 59 41 43
{ 622 621 40 0 128.5 41 79 34 79 41 l 618 70 153 84 100 617 80 112.5 75 92 E 620 120 80.5 59 100 615 150 172.5 78 100 625 200 85 68 100 E- 619 300 >233 61 100 1 F -
GE-NE-523-A164-1294 b DRF 137-0010-7 g Table 5-4 L SIGNIFICANT RESULTS OF IRRADIATED AlG I UNIRRADIATED CHARPY V-NOTCH DATA l
)
Index Index l l Temperature Temperature Index Upper Shelf a i E ( F) ( F) Temperature Energy Material E=30 ft-lb E=50 fi-lb MLE=35 mil (ft-lb) PLATE: Heat SK3238/1, Longitudinal [ Unitradiated -9.16 19.34 -0.73 140/91 Irradiated (f=1.42x10" n/cm 2
)-1.31 17.78 _2.,l_2 121 /79 Difference 3.85 -1.56 2.91 19 / 12 (14 %) ,
l Reg. Guide 1.99, Rev 2 ARindt b: 8 1.99, Rev 2 % Decrease in USE c: (7%) Reg. Guide 1.99, Rev 2 (A12c) b: -26 to 42 l l WELD: Heat / Lot D53040/1125-02205 Unitradiated -69.62 -29.70 -44.44 164 l Irradiated (f=1.42x10" n/cm2)-8.73 18.57 5.00 156 l Difference 60.89 48.27 49.44 8 (5%) , Reg. Guide 1.99, Rev 2 ARtndt b: 15 1.99, Rev 2 % Decrease in USE c: (8.5%) Reg, Guide 1.99, Rev 2 (A120) b: -41 to 71 a USE values for base are Longitudinal / Transverse orientation; for weld metal are equal. Values determined by averaging the 100% shear Charpy data. Transverse plate USE is taken as 65% of the Longitudinal USE, per USNRC MTEB 5-2 [17]. b Determined in Section 5.4.1 c See Section 5.4.2 on i sU'>gu' Ou .. e$ CpoSOa g
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L c GE-NE-523-A164-1294 L DRF 137-0010-7
- 6. TENSILE TESTING r
I Six round bar tensile specimens were recovered from the surveillance capsule and sent to VNC. Uniaxial tensile tests were conducted in air at room temperature (70 F), and RPV operating temperature (550 F). Six unirradiated specimens, sent from the Hope Creek 1 site to L GE-NE San Jose, were tested at the same temperatures. The tests were conducted in accordance r with ASTM E8-89 [14]. L 6.1 PROCEDURE L All irradiated tests were conducted using a screw-driven Instron test frame equipped [ with a 20-kip load cell and special pull bars and grips. Heating was done with a Satec resistance clamshell furnace centered arcund the specimen load train. The test temperature was monitored and controlled by a chromel-alumel thermocouple spot-welded to an Inconel clip that was { friction-clipped to the surface of the specimen at its midline. Before the elevated temperature tests, a profile of the furnace was conducted at the test temperature ofinterest using an unitradiated steel specimen of the same geometry. Thermocouples were spot-welded to the top, middle, and bottom of a central 1 inch gage of this specimen. In addition, the clip-on thermocouple was attached to the midline of the specimen. When the target temperatures of the three thermocouples were within 5 F of each other, the temperature of the clip-on thermocouple was noted and subsequently used as the target temperature for the irradiated specimens. The tests were conducted at a calibrated crosshead speed of 0.005 in/ min until well past yield, at which time the speed was increased to 0.05 inch / min until fracture. Crosshead displacement was used to { monitor specimen extension during the test. All unirradiated tests were conducted using an MTS servohydraulic testing machine equipped with a 10-kip load cell and pulling bars consistent with the configuration of the test I specimens. Heating was done with a Marshall resistance clamshell furnace centered around the j specimen load train. The test temperature was monitored and controlled by a chromel-alumel [ thermocouple attached to the test specimen. Contact extensometry was used to record the load-strain relationship for each test specimen. A strain rate of 0.005 in/in/ min was used through the yield portion of the test and then increased to 0.05 in/in/ min until fracture. l 1
. _ _ _ . . -_________m _ _ - - - _ _ _ - - - - _ _ _ ---- - - - - - - -
L GE-hT-523-A164-1294 DRF 137-0010-7 7 The test specimens were machined with a minimum diameter of 0.250 inch at the center of the gage length. The yield strength (YS) and ultimate tensile strength (UTS) were calculated by dividing the nominal area (0.0491 in2) into the 0.2% offset load and into the maximum test u load, respectively. The values listed for the uniform and total elongations were obtained from plots that recorded load versus specimen extension and are based on a 1.5 inch gage length. E Reduction of area (RA) values were determined from post-test measurements of the necked u specimen diameters using a calibrated blade micrometer and employing the following formula: ) e L RA = 100% * (Ao - Ar)/Ao After testing, each broken specimen was photographed end-on, showing the fracture surface, and lengthwise, showing the fracture location and local necking behavior. 6.2 RESULTS Irradiated tensile test properties of Yield Strength (YS), Ultimate Tensile Strength (UTS), Reduction of Area (RA), Uniform Elongation (UE), and Total Elongation (TE) are presented in Table 6-1; all but UE are presented in Table 6-2 for unirradiated specimens. A stress-strain curve for a 550 F base metalirradiated specimen is shown in Figure 6-1. This curve is typical of the stress-strain characteristics of all the tested specimens. The surveillance materials generally follow the trend of decreasing properties with increasing temperature. Photographs of i [ the fracture surfaces and necking behavior are given in Figures 6-2 through 6-4. 1 6.3 IRRADIATED VERSUS UNIRRADIATED TENSILE PROPERTIES { l The unirradiated and irradiated plate, weld, and HAZ data at room temperature and at 550 F was compared to determine the irradiation effect. In most cases, the trends ofincreasing YS and UTS and of decreasing TE and RA, characteristic ofirradiation embrittlement, are seen in the data. b
GE-NE-523-A164-1294 l DRF 137-0010-7 Table 6-1: TENSILE TEST RESULTS FOR IRRADIATED RPV MATERIALS I Specimen Test Temp. Yield a Strength (ksi) Ultimate Strength (ksi) Uniform (%) Total (%) Reduction Elongation Elongation ofArea (%) 1
.Eumber LE) l Base: PIA PIB 70 550 70.2 67.3 91.9 85.1 9.8 8.5 17.0 14.2 68.4 54.9 I Weld: P2A 70 82.2 96.0 10.3 21.0 73.1 P2B 550 70.9 86.3 8.5 16.4 68.6 I
HAZ: P3A 70 69.7 91.6 7.2 15.1 69.2 P3B 550 66.8 86.9 6.7 14.0 67.2 a Yield strength is determined by 0.2% offset. Table 6-2: TENSILE TEST RESULTS FOR UNIRRADIATED RPV MATERIALS Test Yield a Ultimate Total Reduction Specimen Temp. Strength Strength Elongation ofArea Number fE) (ksi) (bil_ (%) (%) Base: 26-9378 70 70.1 92.6 26.2 70.9 26-9378 550 62.3 89.3 21.9 62.8 Weld: 26-9379 70 78.8 93.4 29.0 71.7 26-9379 550 72.3 90.6 21.3 67.9 HAZ: 26-9380 70 68.4 91.6 19.3 68.2 26-9380 550 62.9 88.0 16.3 60.3 a Yield strength is determined by 0.2% offset. I l L c GE-NE-523-A164-1294 L DRF 137-0010-7 p Table 6-3: COMPARISON OF UNIRRADIATED AND IRRADIATED L TENSILE PROPERTIES AT ROOM TEMPERATURE F L Yield Strength Ultimate Strength Total Elongation Reduction of Area (ksi) __(kgi)_ (%) (%) L Base: Unitradiated 70.1 92.6 26.2 70.9 Irradiated 70.2 91.9 17.0 68.4 Difference a 0.1 (0.1%) -0.7 (-0.8%) -9.2 (-35.1%) -2.5 (-3.5%) [ [ Weld: Unirradiated 78.8 93.4 29.0 71.7 _ _ _ _ _ _ Irradiated 82.2 96.0 21.0 73.1 r Difference a 3.4 (4.3%) 2.6 (2.8%) -8.0 (-27.6%) 1.4 (2.0%) L a In parenthesis is %DifTerence = [(Irrad. - Unirrad.)/Unirrad.]
- 100%
Table 6-4: COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES AT 550oF r L Yield Strength Ultimate Strength Total Elongation Reduction of Area ( (ksi) (ksi) (%) (%) Base: Unitradiated 62.3 89.3 21.9 62.8 Irradiated 67.3 85.1 14.2 54.9 ( Difference a 5.0 (8.0%) -4.2 (-4.7%) -7.7 (-35.2%) -7.9 (-12.6%) Weld: Unirradiated 72.3 90.6 21.3 67.9
- Irradiated 70.9 86.3 16.4 68.6 - Difference a -1.4 (-1.9%) -4.3 (-4.7%) -4.9 (-23.0%) 0.7 (1.0%)
( a In parenthesis is % Difference = [(Irrad. - Unitrad.)/Unitrad.]
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GE-NE-523-A164-1294 DRF 137-0010-7
- 7. DEVELOPMENT OF OPERATING LIMITS CURVES The aspects of the curves which have changed as a result of the testing presented here and as a result of AShE Code changes are discussed below.
I
7.1 BACKGROUND
The revised fluence values in Section 4, which are about 14% greater than the current nominal fluence value, is used in this section to revise the adjusted reference temperatures (ARTS), which are then used to revise the beltline P-T curves. The P-T curve revision includes consideration of the change to the allowable fracture toughness equation in AShE Code Section XI, Appendix G, which occurred in 1992. The coefficient 1.233 in the KIR/K Ia quation e in Figure G-2210-1, became 1.223. The revision adds about 1/2 F to the calculated temperature for a given pressure on the P-T curves (i.e., all curved portions of the P-T curves shift 1/2 F to the right). The resulting P-T curves for pressure tests (Curve A), non-nuclear heatup/cooldown (Curve B) and core-critical operation (Curve C) are shown in Figures 7-1 through 7-3, respectively. The P-T limits are provided in tabular form in I Table 7-1. 7.2 NON-BELTLINE REGIONS The non-beltline curves are developed for two regions: the upper vessel region, governed by the feedwater nozzle limits, and the bottom head region, governed by the CRD penetration limits. Table 3-2 has these limiting initial RTNDT values which are: 40 F for the upper vessel region, based on the feedwater nozzle (a value of 40 F consistent with the purchase specification was used, since initial RTNDT values were unavailable for many of the vessel I nozzles) and 30 F for the bottom head region, based on the bottom head. The 1/2 F adjustment was made to the curved portions of the non-beltline curves, but not to the straight line and step portions, which are based on 10CFR50 Appendix G. Although bottom head Curve B is not limiting, it is included in Figure 7-2, as there may be transients where the bottom head is cooler than the upper vessel regions. I I
GE-NE-523-A164-1294 7.3 CORE BELTLINE REGION Figures 7-1 through 7-3 show the beltline curves at 32 EFPY. As with the non-beltline curves, the 1/2 F adjustment was made to the curved portions of the beltline curves. 7.4 EVALUATION OF IRRADIATION EFFECTS The impact on adjusted reference temperature (ART) due to irradiation in the beltline I materials is determined according to the methods in Reg. Guide 1.99 [7], as a function of neutron fluence and the element contents of copper (Cu) and nickel (Ni). The specific relationship from Reg. Guide 1.99 [7] is: ART = Initial RTNDT + ARTNDT + Margin (7-1) where: ARTNDT = [CF]'f(0.28 - 0.10 log f) (7-2) Margin = 2*(oI2 + g32)1/2 (73) CF = chemistry factor from Tables 1 or 2 of Reg. Guide 1.99 [7], f= 1/4 T fluence (n/cm2) divided by 1019, al = standard deviation on initial RTNDT. cA = standard deviation on ARTNDT, is 28 F for welds and 17 F for base material, except that og need not exceed 0.50 times the ARTNDT value. I Once two sets of surveillance capsule data are available, the CF values in Reg. Guide 1.99 [7] can be modified to reflect the results. However, this is only the first set of surveillance data from Hope Creek 1, so only the results of the flux wire tests are factored into beltline ART calculations. I Each beltline plate and weld ARTNDT value is determined by multiplying the CF from Reg. Guide 1.99, determined for the Cu-Ni content of the material, by the fluence factor for the EFPY being evaluated. The Margin term and initial RTNDT are added to get the ART of the material. The 32 EFPY ART values are shown in Table 7-2. Results for all of the beltline plates and several of the most limiting beltline welds are shown. I - GE-NE-523-A164-1294 L DRF 137-0010-7 E E The LPCI nozzle is partially in the beltline region; its value ofinitial RTNDT is-20 F per Table 3-2. Comparing the ART of the LPCI nozzle (24.3 F), determined in Table 7-2, with an [ initial RTNDT or used to develope the P-T curves. f non-beltline nozzles of 40 F, the non-beltline nozzles are more limiting and are I 7.4.1 ART Versus EFPY The results in Table 7-2 show that beltline plate SK3025/1 is limiting at 32 EFPY; the most limiting weld is D53040/1125 02205. Figure 7-4 shows the ART as a function ofEFPY for I u the most limiting plate and weld. The resulting ARTS at 32 EFPY are 72.5*F for the plate and 32.8 F for the weld. 7.4.2 Upper Shelf Enerev at 32 EFPY Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy (USE) of the beltline materials. The USE must be above 50 ft-lb at all times during plant operation, [ assumed here to be up to 32 EFPY. Calculations of 32 EFPY USE, using Reg. Guide 1.99 [7] methods, are summarized in Table 7 3. Percent decrease in USE as a function of fluence and { copper content is shown in Figure 2 ofReg. Guide 1.99. Therefore. the values of 32 EFPY fluence at 1/4 T location as determined in Section 4.3 along with the copper content of the 7 beltline materials listed in Table 3-1 provide the necessary information to determine percent ' decrease in USE. The initial USE of the beltline plates and welds were obtained from Table 5 A-19 of Reference [5]. The unirradiated surveillance plate and weld data from Table 5-4 are also included for comparison. [ From Table 7-3, the lowest value of 32 EFPY USE for the beltline plate is 67 ft-lb. Information in [5] indicates that the specimens data were for transverse orientation. E Weld metal initial USE values were taken during vessel fabrication at 10 F for all heats. Unlike the plate, the weld metal USE has no transverse / longitudinal correction because weld metal has no orientation effect. The lowest weld USE at 32 EFPY, as shown in Table 7-3, is 60 fl-lb, which is expected to be quite conservative based on initial data at 10 F. [
GE-NE-523-A164-1294 DRF 137-0010-7 Based on the above results, it is expected that the beltline materials will have USE values above 50 ft-lb at 32 EFPY, as required in 10CFR50 Appendix G [1]. Since USE and ART requirements are met, irradiation effects are not severe enough to necessitate additional analyses or preparations for RPV annealing before 32 EFPY. Moreover, PSE&G is a participant in a BWR Owners' Group program to perform analyses to demonstrate equivalent margin in cases where the USE drops below 50 ft-lb [18]. This analysis shows equivalent margin at USE values as low as 35 ft-lb. The calculations in Tables B-1 and B-2 in Appendix B show that the equivalent margin analysis is applicable. 7.5 OPERATING LIMITS CURVES VALID TO 32 EFPY Figures 7-1 through 7-3 show P-T curves valid to 32 EFPY. The P-T curves are developed by considering the requirements applicable to the non-beltline, beltline, and closure flange regions. For most of each curve, the beltline curves are more limiting. The bottom head has been included on Curve B to provide the appropriate limits for transients where some bottom head stratification might occur. I I I I I I GE-NE-523-A164-1294 DRF 137-0010-7 Table 71 IIOPE CREEK 1 P-T CURVE VALLTS nu.....u.u..u.u.....u...... REQUIRED 7EMPERARIRES""""*"""*"""""" 32 EFPY NON. BOTTOM 32 EFPY UPPEP. NON. 32 EFPY NON. PRf&$URE BELTLINE BELTLINE HEAD BELTilNE VESSEL BEL 7UNE BELTLINE BEL 7 TINE CURVE A CURVE A CURVE B CURVEB CURVE B CURVE B CURVE C CURVE C 0 79.0 79.0 79.0 79.0 I 10 20 30 79.0 79.0 79.0 79.0 79.0 79.0 79.0 79.0 79.0 79.0 79.0 79.0 40 70.0 79.0 79.0 80.6 50 79.0 79.0 79.0 93.6 60 79 0 79.0 79.0 104 6 70 79.0 79.0 79.0 114.1 79.0 I 80 82.3 82.3 122.3 90 79.0 89.3 89.3 129.3 100 79.0 95.4 95.4 135.4 110 79.0 101.0 101.0 141.0 120 79 0 105.9 105.9 145.9 1 130 140 79.0 79.0 110.7 115.3 110.7 115.3 150.7 155.3 150 79.0 119.6 119.6 159.6 160 79.0 123.5 123.5 163.5 170 79.0 126.9 126.9 166.9 180 79 0 129.9 129.9 169.9 190 79.0 132.7 132.7 172.7 I 200 210 220 79.0 79.0 79.0 135.4 138.1 140.6 135 4 138.1 140.6 69.6 78.9 87.0 175.4 178.1 180.6 230 79.0 143.0 143.0 94.4 183.0 240 79.0 145.3 145.3 101.0 185.3 250 79.0 147.5 147.5 107.0 187.5 260 79.0 149.6 149.6 112.6 189.6 270 79.0 131.6 131.6 117.7 191.6 I 280 290 300 79.0 79.0 79.0 94.2 153 6 155.5 157.3 153.6 155.5 157.3 122.5 126.9 134.2 193.6 195.5 197.3 310 79.0 98.1 159.1 159.1 138.1 199.1 312.5 79.0 99.0 159.5 159.5 139.0 199.5 312.5 109.0 99.0 159.5 159.5 139.0 199.5 320 109 0 101.7 160.8 160.8 141.7 2008 330 202.4 I 109 0 105.2 162 4 162 4 145.2 340 109.0 108.5 164 0 164.0 148.5 204.0 350 109 0 111.7 165 6 165.6 151.7 205.6 360 109.0 114 7 167.1 167.1 154.7 207.1 370 109.0 117 6 168 6 168.6 157.6 208.6 380 109 0 120.4 170.1 170.1 160.4 210.1 390 109.0 123.1 171.6 171.6 163.1 211.6 400 109 0 60 6 125.7 173.1 1731 165.7 213.1 410 109.0 69 6 128.1 174 6 174 6 1681 214.6 420 109.0 76 6 130.5 176.0 176.0 170.5 216.0 430 109.0 82 6 132.8 177.4 177.4 172.8 217.4 440 109 0 87 6 135.1 178 8 178.8 175.1 218.8 450 109 0 91 6 137.3 180 1 180.1 177.3 220.1 460 109.0 95.1 139.4 181.4 181.4 179 4 221.4 470 109 0 93.2 141.4 182.7 182.7 181.4 222.7 480 109 0 101.1 143.4 183.9 183 9 183 4 223.9 1 i GE-NE-523-A164-1294 DRF 137-0010-7 Table 71 1IOPE CREEK 1 P.T Cl 1.VE VALUES
..... ... . .... ...... ............. . . REQUIRED TEMPERAIURES* * * * * * * * * * * * * * * * * * * * * * * ** * *
- 32 EFPY NON. BOTTOM 32 EFPY UPPER NON- 32 EFPY NON-PRES 5URE BELTLINE BELTLINE IIEAD BELTUNE VESSEL BELTLINE BELTLINE BELTLINE CURVE A CURVE A CURVEB CURVEB CURVEB CURVE B CURVE C CURVE C 490 109.0 103.9 1453 185.1 185.1 1853 225.1 I 500 510 520 69.6 73.7 77.6 109.0 109.0 109.0 106.6 1093 111.9 147.2 149.0 150.8 1863 187.4 188.5 1863 187.4 188.5 187.2 189.0 190.8 2263 227.4 228.5 530 813 109.0 114.5 152.5 189.6 189.6 192.5 229.6 540 84.8 109.0 117.0 154.2 190.6 190 6 194.2 230.6 550 88.2 109.0 119.4 155.9 191.6 191.6 195.9 231.6 560 91.4 109.0 121.7 157.5 192.6 192.6 197.5 232.6 570 94.4 109.0 123.9 159.0 193.5 193.5 199.0 233.5 580 973 109.0 126.0 100.6 194.4 194 4 200.6 234 4 590 100.1 109.0 127.8 162.1 195.3 1953 202.1 235.3 600 102.8 109.0 129.6 163.6 196.1 196.1 203.6 236.1 I
610 105.4 109.0 1313 165.0 196.9 196.9 205.0 236.9 620 107.9 109.0 133.1 166.4 197.7 197.7 206.4 237.7 I
- 625 109.1 109.0 133.9 167.1 198.0 198.0 207.1 238.0 630 1103 109.0 134.7 167 8 198.4 198 4 207.8 238.4 I
640 112.7 109.0 136 4 169.2 199.1 199.1 209.2 239.1 650 114.9 109.0 137.9 170.5 199.7 199.7 210.5 239.7 660 117.1 109.0 139.4 171.8 200.4 200.4 211.8 240 4 670 119.2 !!0.2 140.9 173.1 201.0 201.0 213.1 241.0 I 680 690 700 710 121.2 123.2 125.2
!!2.0 113.8 115.6 117.3 142.4 143.8 145.2 146.6 174.?
175 6 1768 178.0 201.5 202.1 202.6 203.1 201.5 202.1 202.6 203.1 2143 215.6 216.8 218.0 24 L5 242.1 242.6 127.1 243.1 I 720 730 740 128.9 130.7 132.4 118.9 120 6 122.1 147.9 149.2 150.4 179.2 1803 181.5 203.5 204.0 204.4 203.5 204.0 204.4 219.2 2203 221.5 20.5 244.b 244.4 750 134.1 123.7 151 6 182.6 204.9 204.9 222.6 244.9 760 135.8 125.2 152.7 183.7 2053 205.3 223.7 2453 770 137.4 126.7 1538 184.7 205.7 205.7 224.7 245.7 780 139.0 128.1 154.9 185.8 206.1 206.1 225.8 246.1 790 1405 129.6 156.0 186.9 206.5 206.5 226.9 246.5 800 142.0 130.9 157.1 187.9 206.9 206.9 227.9 246.9 810 143.5 1323 158.2 188.9 2073 2073 228.9 2473 820 144.9 133.6 1593 189.9 207.7 207.7 229.9 247.7 830 146 4 134.9 1603 190.9 208.1 208.1 230.9 248.1 840 147.8 136.2 161.4 191.9 208.4 208 4 231.9 248.4 850 149.1 137.5 162.4 192.8 208.8 208.8 232.8 248.8 860 150.5 138.7 163.5 193.8 209.1 209.1 233.8 249.1 870 151 8 139.9 164.5 194.7 209.5 209.5 234.7 249.5 880 153 0 141.1 165.6 195 6 209.8 209.8 235.6 249.8 I 890 154.3 142.3 166.6 196 6 210.2 210.2 236 6 250.2 900 155 6 143.4 167.7 197.4 210.5 210.5 237.4 250.5 ; I 910 920 930 940 156 8 158 0 159 1 144 6 145.7 1468 168.7 169.7 170.7 1983 199.2 200.I 210.8 211.2 211.5 210.8 211.2 211.5 2383 239.2 240.1 250.8 251.2 251.5 I l 1603 147 8 171.7 200.9 211.9 211.9 240.9 251.9 l I 950 960 970 161.4 162 6 163.7 148.9 149.9 t 51.0 172.7 173.7 174 7 201.8 202 6 203 4 212.2 212.5 212.9 212.2 212.5 212.9 241.8 242.6 243.4 252.2 252.5 252 9 1 980 164 7 152 0 175 7 204.2 213.2 213 2 244.2 253.2
)
g l
GE-NE-523-A164-1294 DRF 137-0010-7 I Table 7-1 llOPECREEK 1 P-TCURVE VALUES
....u....
.....u........u....... REQUIRED TEMPERAR'RES""*"*""*"""""*""
32 ETPY NON. BOTTOM 32 EFPY UPPER NON. 32 EFPY NON. PRESSURE BELTUNE BELTLINE IIEAD BELHlNE VESSEL BELTUNE BELTLINE BELTIJNE CURVE A CURVE A CURVEB CURVE B CURVE B CURVE B CURVE C CURVE C 990 165.8 153.0 176.6 205.0 213.6 213.6 245.0 253.6 I 1000 1010 1020 166.9 167.9 168.9 153.9 154.9 155.9 177.6 178.5 179.4 205.8 206.6 207.4 213.9 214.2 214.6 213.9 214.2 214.6 245.8 246.6 247.4 253.9 254.2 254.6 1030 169.9 156.8 180.2 208.2 214.9 214.9 248.2 254.9 1040 170.9 157.7 181.0 208.9 215.2 215.2 248.9 255.2 1050 171.9 158.6 18L8 209.7 215.6 215.6 249.7 255.6 1060 172.8 159.5 182.6 210.4 215.9 215.9 250.4 255.9 1070 I 173.8 160.4 183.4 211.2 216.2 216.2 251.2 256.2 1080 174.7 1613 184.2 211.9 216.5 216.5 251.9 256.5 1090 175.7 162.2 184.9 212.6 216.9 216.9 252.6 256.9 1100 176.6 163.0 185.7 2133 217.2 217.2 253.3 257.2 1110 177.5 163.9 186.4 214 0 217.5 217.5 254.0 227.5 1120 178.4 164.7 187.1 214.7 2.17.9 217.9 254.7 257.9 1130 179.2 165.5 187.8 215.4 212.2 218.2 255.4 258.2 1140 180.1 1663 188.5 216.1 217.5 218.5 256.1 258.5
!!50 181.0 167.1 189.2 216.8 213.9 218.9 256.8 258.9 1160 181.8 167.9 189.9 217.4 219.2 219.2 257.4 259.2 1170 182.6 168.7 190.6 218.1 219.5 219.5 258.1 2$9.5 1180 183.5 169.5 191.2 218.8 219.8 219.8 258.8 259.8 1190 184.3 170.2 191.9 219.4 220.1 220.1 259.4 260.1 1200 185.1 171.0 192.6 220.1 220.4 220.4 260.1 260.4 1210 185.9 171.7 193.2 220.7 220.8 ?20.8 260.7 260.8
- 1215 1863 172.1 193.5 221.0 220.9 220.9 261.0 260.9 1220 186.7 172.5 193.9 2213 221.1 221.1 2613 261.1 1230 187.4 173.2 194.5 222.0 221.4 221.4 262.0 261.4 1240 188.2 173.9 195.2 222.6 221.7 221.7 262.6 261.7 1250 189.0 174.7 195.8 223.2 222.0 222.0 263.2 262.0 1260 189.7 175.4 196.5 223.8 2223 2223 263.8 2623 1270 190.5 176.1 197.1 224 4 222.6 222.6 264.4 262.6 1280 191.2 176.8 197.8 225.0 223 0 223.0 265.0 263.0 1290 191.9 177.4 198.4 225.6 223.3 2233 265.6 2633 1300 192.7 178.1 199.1 226.2 223.6 223.6 266.2 263.6 1310 193.4 178.8 199.7 226.8 223.9 223.9 266.8 263.9 1320 194.1 179.5 200.4 227.4 224.2 224.2 267.4 264.2 1330 194.8 180.1 201.0 228.0 224.5 224.5 268.0 264.5 1340 195.5 180 8 201.7 228.5 224.8 224.8 268.5 264.8 1350 196.2 181.4 2023 229.1 225.1 225.1 269.1 265.1 1360 196.8 182.1 203.0 229.7 225.4 225.4 269.7 265.4 1370 197.5 182.7 203 6 230.2 225.7 225.7 270.2 265.7 1380 198.2 183 3 2043 230 8 226 0 226.0 270.8 266.0 1390 198.8 183.9 204.9 231.3 226.3 226.3 2713 2663 1400 199.5 184 6 205.6 231.9 226.6 226.6 271.9 266.6
- Values hnearly eterpolated. Curves A mtersect at approximate!y 625 peig. Curves B (except Bottom Head )
mtersect et opproximately 1215 ps+g. and Curves C miersect at apprerimsto!y 1215 pois. I g ! GE-NE-523-A164-1294 DRF 137-0010-7 I Table 7-2 BELT 1DE ART VALUES FOR HOPE CREEK 1 Shell Course H Phase and Weld 3/4 SheII Cowse #3 Piece and Weld 3/4 ' Thickness = 6.1 inches 32 EFPY Peak 1.D. Susace a 4.94E+17 32 EFPY Peak 1/4 T fluence = 3.42E+17 I. Shett Cowse #4 & 5 and Welde4/5 L Shell Course #4 & 5 and Weide4/5 1 a kne - 6.i ino es 32 EFPY Peak LD. fluence = 7.50E+17 32 EFPY Peak 1/4 T fluence = 5.20E+17 [ LPCI Nesale and Weld LPG Neeste and Wald L 1hicirmens a 6.1 inche: 32 EFPY Peak 1.D. fluence a 4.05E+17 32 EFPY Peak 1/4 T fluence = 2.81E+17
!.D. or Initial 32 EFPY 32 EFPY 32 EFPY COMPONEAT Type IIEAT OR HEAT / LOT %Cu %Ni CF RTadt Delu RTndt Margin ShiA ART T T T T T PLATESs is wer No. 5 5K3230 /1 0.07 0.56 44 -10.0 13.2 13.2 26.3 16.3 No. 5 6C35 /1 0.09 0.54 - 58 l1.0 17.3 17.3 34.7 23.7 No. 5 6C45/1 0.08 0.57 51 1.0 15.2 15.2 30.5 31.5 Imwer-Inenwd No. 4 5K2963 /1 0.07 0.58 44 10.0 13.2 13.2 26.3 163 No. 4 5K2530 /1 0.08 0.56 51 19.0 15.2 15.2 30.5 49.5 No. 4 $K3238 /1 0.09 0.64
- 58 7.0 17.3 17.3 34.7 41.7 Internwdiese No. 3 $K3025 /1 0.15 0.71 113 19.0 26.7 26.7 53.5 72.5 No. 3 5K2608/1 0.09 0.5 s 58 19.0 13.7 13.7 27.5 46.5 Na3 SK2698 /1 0.10 0.58 65 19.0 15.4 15.4 30.8 49.8 LPCI Nessie N7 19468/l 0.12 0.80 86 -20.0 18.2 18.2 36.3 16.3 i N7 10024/I 0.14 0.82 105 20.0 22.2 22.2 44.3 24.3
[ VER71 CAL SMAW 510-01205 0.09 0,54 109 -40.0 32.6 32.6 65.2 25.2
%ILDS 4/S: SAW D53040/112542205 0.08 0.59
- 105 -30.0 31.4 31.4 62.8 32.8 E LPCI NOZZLE WELDS: SMAW 001 01205 0.02 0.51 27 -40.0 5.7 5.7 11.4 -28.6 GIRTH MILD: SMAW 51941205 0.01 0.53 20 -49.0 4.7 4.7 9.5 -39.5 3/4: SMAW 5044 1205 0.01 0.51 20 -31.0 4.7 4.7 9.5 21.5 SAW D53040/1810 02205 0.10 0.68 126 -49.0 29.9 29.9 59.7 10.7 SAW D55733/1810 02205 0.10 0.68 126 -40.0 29.9 29.9 59.7 19.7 Note: Deu of Cu and Ni wt% and initial RTndt are taken from HCGS4 TSAR.
- Averste % Cu and %Ni from Table 3-3 is usesi E
~
GE-iJE-523-A164-1294 bs DRF 137-0010-7 Table 7-3. UPPER SHELF ENERGY ANALYSIS FOR HOPE CREEK 1 BELTLINE MATERIAL h INITIAL.* 32 EFPY TRANS. %DECR.6 TRANS. LOCATION HEAT USE %Cu USE USE ( PLATES: Imtr 5K3230/1 121 0.07 8 111 6C35/1 107 0.09 9 97 6C45/1 97 0.08 8.5 89 Im-Int. SK2963/1 102 0.07 8 94 { 5K2530/1 86 0.08 8.5 79 SK3238/l 76 0.09 9 69 Unitradiated
- SK3238/1 91 0.09 9 83 Surveillance Int. SK3025/1 75 0.15 11 67 5K2608/l 75 0.09 8 69 SK2698/l 75 0.10 8.5 69 19468/l >79 0.12 9 72 10024/1 >70 0.14 10 63
[ WELD: Vertical 510-01205 >92.5 0.09 11.5 82 j D53040 135 0.08 11 120 i ( Unitradiated ' D53040 164 0.08 11 146 Surveillance ( LPCI Nozzle 001-01205 >109 0.02 6 102 Girth 519-01205 >109 0.01 5 104 504-01205 >125 0.01 5 119 D53040 >95 0.10 11 85 D55733 >68 0.10 11 60
' Initial USE obtained from HCGS-UFSAR. Transverse plate values are conservatively estimated as described in the UFSAR; test temperatures for plate materials were not available. Weld values are conservatively based on data taken at 10*F. l 6
Values obtained from Figure 2 of [7] for 32 EFPY 1/4 T fluences equal to 5.2x10" n/cm2 , for Low. and Low-int. shells; 3.42x10" n/cm*, for Int. shell; and 2.81x10" n/cm2 , for LPCI Nozzle. A fluence of 5.2x10" n/cm2 was used for the welds identified as vertical and 3.42x10" n/cm2 for the welds identified as girth. I
' USE data taken from Table 5-4 and chemistry data from Table 3-3
r GE-NE-523-A164-1294 ( DRF 137-0010-7 f- 1600 L A - SYSTEM HYDROTEST LIMIT I. I k CURWA l 1400 r i l i I I I {' 1200 Ili 5 fI l I 5 O ll ! fl ( !g 1000 // i H I . d I [ 8 s / o I 800 NON-BELTLINE ( b j LIMITS AND 4 6 f 10CFR50 APP G a: . REQ'MTS [ s f/ t- / E 600 - - - BELTLINE, 53.5 0 F { d / SHIFT 5 '
/
8 [ E 400 312 { _ PSIG - [ 200 BOLTUP ; 79 F CURVES ARE VALID FOR 32 EFPY OF OPERATION { { o 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE ( F) Figure 7-1. Pressure Test P-T Curves for Hope Creek 1 [
GE-NE-523-A164-1294 DRF 137-0010-7 1600 B - NON-NUCLEAR HEAT-UP/ CURVE B COOLDOWN LIMIT 1400 I . I 1200 ' 52 1 N I I : E i !
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79"F CURVES ARE VALID FOR 32 EFPY OF OPERATION I l l ! 0 1 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPCRATURE (*F) Figure 7-2. IIcatup/Cooldown P-T Cu Tes for Hope Creek 1 g - ei -
E GE-NE-523-A164-1294 g- DRF 137-0010-7 r' 1600 L C-NUCLEAR F { L (CORE CRITICAL) LBIIT CWWC i { J 1400 e 'l L , r r I L 1200
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{ 1 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE (DF) [ Figure 7-3. Core Critical Operation P-T Cun es for Hope Creek 1
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GE-NE-523-A164-1294 DRF 137-0010-7 l
- 8. REFERENCES I
[1] " Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, May 1983. I [2] " Protection Against Non-Ductile Failure," Appendix G to Section XI of the 1992 ASME Boiler & Pressure Vessel Code. [3] " Reactor Vessel Material Surveillance Program Requirements," Appendix H to Part 50 of Title 10 of the Code of Federal Regulations, May 1983. [4] " Surveillance Test for Nuclear Reactor Vessels," ASTM E185-66 [5] Hope Creek Generating Station Updated Final Safety Analysis Report, Section 5.3 [6] " Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards, E185-82, July 1982. I [7] " Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988. [8] " Location and Numbering of Surveillance Test Specimens", Babcock-Hitachi (Drawing Numbers KUO-122-203). [9] " Technical Specifications Hope Creek Generating Station, Docket No. 50-354", U.S. Nuclear Regulatory Commission, July 1986 (NUREG-1202) I [10] Martin, G.C., " Browns Ferry Unit 3 In-Vessel Neutron Spectral Analysis," GENE, Pleasanton, CA, August 1980 (NEDO-24793). [11] " Flux Wire Dosimeter Evaluation for the Hope Creek Generating Station," GE Report SASR 89-23, March 1989 [12] " Standard Methods for Notched Bar Impact Testing ofMetallic Materials," Annual Book of ASTM Standards, E23-88 I GE-NE-523-A164-1294 DRF 137-0010-7 [13] " Nuclear Plant Irradiated Steel Handbook," EPRI Report NP-4797, September 1986. [14] " Standard hiethods of Tension Testing ofhietallic hiaterials," Annual Book of ASThi Standards, E8-89. [15] " Response To Generic Letter 92-01, Revision 1," Hope Creek Generating Station, Docket No. 50-354 [16] " Surveillance Test Specimens for Reactor Pressure Vessel," GE Number 21 A8707, December 1971 (17] Fracture Toughness Requirements," USNRC Branch Technical Position AITEB 5-2, Resision 1, July 1981. [18] H.S. hiehta, T.A. Caine, and S.E. Plaxton, "10CFR50 Appendix G Equivalent hiargin Analysis for Low Upper ShelfEnergy in BWR/2 through BWR/6 Vessels," GENE, San Jose, CA, February 1994 (NEDO-32205-A, Revision 1). I I I I I I GE-NE-523-A164-1294 ! DRF 137-0010-7 APPENDIX A CIIARPY SPECIMEN FRACTURE SURFACE PIIOTOGRAPIIS l Photographs of each Charpy specimen fracture surface were taken per the requirements of ASTM E185-82. The pages following show the fracture surface photographs along with a summary of l the Charpy test results for each irradiated specimen. The pictures are arranged in the order of base, weld, and HAZ materials. l I I l I I I I .I lI 'I I I I I g - ee -
I l GE-NE-523-A164-1294 DRF 137-0010-7 !I i
~
1 BASE: 605 - BASE: 613 Temp: 300 F - Temp: 200 F l Energy: 133 ft-lb .f. / Energy: 115.5 ft-lb
- MLE
- 92 mils .,
j MLE: 81 mils l Shear: 100 % , , Shear: 100 % l BASE: 608 _ BASE: 609 l Temp: 120 F " Temp: 80 F l Energy: 113.5 ft-lb . Energy: 115 ft-lb MLE: 80 mils : . MLE: 78 mils ) l Shear: 100 % i ~ Shear: 85 % ! l I I
)
i BASE: 611 .. BASE: 610 1
. s s i l Temp: 70 F , .
Temp: 60 F ' { Energy: 105 ft-Ib . Energy: 97 fl-lb l MLE: 84.5 mils '
?
s MLE: 77 mils ! Shear: 79 % , Shear: 60 % 1 BASE: 607 , BASE: 612 7 Temp: 40 F c . Temp: 0 F Energy: 55 f1-lb . . Energy: 42 ft-lb l MLE: 50.5 mils . $ . MLE: 37 mils ! Shear: 52 % < Shear: 16 % i !I j
- 1
GE-NE-523-A164-1294 i DRF 137-0010-7 ! l
)
l BASE: 606 --- - -
. BASE: 603 l Temp: -20 F .
Temp: -40 F Energy: 7 n -lb Energy: 32.5 n-lb i MLE: 8.5 mils - j MLE: 29.9 mils Shear: 18 % - Shear: :S % BASE: 604 BASE: 614 Temp: -50 'F _-- i Temp: -80 'F
~
) Energy: 8.5 ft-lb Energy: 7 n-lb ! MLE: 10 mils ._ MLE: 6.5 mils Shear: 6%
, Shear: 3% !
l
~
WELD: 600 , - WELD: 602 Temp: 300 F >- i N . Temp: 200 F Energy: 178.5 n-lb
.44 Energy: 144.5 n-lb j MLE: 88 mils .
;. MLE: 91 mils
~~
Shear: 100 % l Shear: 100 % WEl.D: 591 WELD: 598 Temp i.40 F i Temp: 120 F Energy: 154.5 n-lb Energy: 147.5 n-lb MLE: 90 mils MLE: 92 mils Shear: 100 % , . Shear: 100 %
GE-NE-523-A164-1294 DRF 137-0010-7 I WELD: 593 . WELD: 601 Temp: 80 F . Temp: 50 F Energy: 112.5 ft-lb Energy: 61 ft-lb I MLE: 69 mils 7 MLE: 44 mils Shear: 77 % Shear: 58 % I WELD: 599 ; WELD: 596 Temp: 40 F
~
Temp: 20 F
- Energy
- 88.5 ft-lb Energy: 59.5 ft-lb I MLE: 72 mils Shear: 50 %
MLE: 48.5 mils Shear: 39 % s WELD: 592 l WELD: 595 j j Temp: 0 "F I Temp: -10 F , Energy: 19.5 ft-lb Energy: 31 ft-lb l MLE: 22.5 mils l, i MLE: 27 mils j Shear: 34 % L' Shear: 26 % l WELD: 597 _. WELD: 594 Temp: -40 F Temp: -80 F Energy: 22.5 ft-lb '
,. Energy: 5.5 ft-lb MLE: 20 mils -
MLE: 9.5 mils Shear: 17 % . Shear: 12 % l t
- I g . e9
I GE-NE-523-A164-1294 DRF 137-0010-7
/ % .
Temp: 200 F Energy; 85 n-lb
#^
MLE: 68 mils Shear: 100 % l l HAZ: 615 . HAZ 620 Temp: 150 'F . Temp: 120 F Energy: 172.5 n-lb Energy: 80.5 n-lb MLE: 78 mils - MLE: 59 mils Shear: 100 % ,- Shear: 100 % HAZ: 617 . HAZ 618 Temp: 80 F , Temp: 70 F Energy: 112.5 ft-lb - Energy: 153 n-lb MLE: 75 mils ' - MLE: 84 mils l Shear: 92 % < Shear: 100 % 1 . !l HAZ: 621 P HAZ 622
- Temp
- 40 F Temp: 0F Energy: 41 n-lb Energy: 128.5 n-lb MLE: 34 mils .
MLE: 79 mils l Shear: 41 % - Shear: 79 % L i
GE-NE-523-A164-1294 DRF 137-0010-7 HAZ: 623 , HAZ: 626 Temp: -10 F Temp: -40 F Energy: 59 ft-lb ? Energy: 9.5R-lb MLE: 41 mils ,7 , MLE: 10 mils Shear: 43 % 1 Shear: 26 % HAZ: 624 - HAZ 616 l ,. Temp: -80 F , y Temp: -100 *F Energy: 85 ft-lb . Energy: 70 fi-lb MLE: 59 mils l , MLE: 48 mils Shear: 44 % .
.- 4 Shear: 33 %
l HAZ: 619 .f .. l Temp: 300 F l Energy: >233 ft-lb .
'/J 47 gg l
, MLE: 61 mils '?;;jya Shear: 100 %
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ga i5 k*g;g'; g $ % w h. i , yppy i Qf yd5P lI u a . - . - - 4 2- - - - - ,n. 4 I GE-NE-523-A164-1294 DRF 137-0010-7 APPENDIX B EQUIVALENT MARGIN ANALYSIS I I I I I i
GE-NE-523-A164-1294 DRF 137-0010-7 I TABLE B-1 I EQUIVALENT MARGIN ANALYSIS PLANT APPLICABILITY VERIFICATION FORM FOR ITOPE CREEK UNIT 1 - BWR 4/MK I I BWR/3-6 PLATE Surveillance Plate USE:
%Cu = 0.09 I Capsule Fluence = 1.42 x 10" n/cm2 l
Measured % Decrease = J4 (Charpy Curves) R.G.1.99 Predicted % Decrease = 2 (R.G.1.99, Figure 2) Limitine Beltline Plate USE:
~
%Cu = QJji 32 EFPY 1/4-T Fluence = 5.2 x 10" n/cm2
- I R.G.1.99 Predicted % Decrease = 12 (R.G.1.99, Figure 2)
'I Adjusted % Decrease = 12 (R.G.1.99, Position 2.2) 12 % < 21%, so vessel plates are I bounded by equivalent margin analysis I I il I
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_ 1
GE-NE-523-A164-1294 , DRF 137-0010-7 TABLE B-2 I EQUIVALENT MARGIN ANALYSIS PLANT APPLICABILITY VERIFICATION FORM r FOR HOPE CREEK UNIT 1 - BWR 4/MK I L p HWR/2-6 WELD L Surveillance Weld USE: r L %Cu = 0.08 2 Capsule Fluence = 1.42 x 10" n/cm Measured % Decrease = 5 (Charpy Curves) [ R.G.1.99 Predicted % Decrease = HJ (R.G.1.99, Figure 2) Limiting Beltline Weld USE:
%Cu = 0.10 2
32 EFPY Fluence = 5.2 x 10" n/cm ( R.G.1.99 Predicted % Decrease =12 (R.G.1.99, Figure 2) Adjusted % Decrease = N!/A (R.G.1.99, Position 2.2) t 12 % 5 34%, so vessel welds are [ bounded by equivalent margin analysis ( { {
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