B110073, E-mail from Entergy to T. Le & Document - GE-NE-B1100732-01, Rev. 1, Final Report, Class II, February 1998 Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120 Degree Capsule at 13.4 EFPY

E-mail from Entergy to T. Le & Document - GE-NE-B1100732-01, Rev. 1, Final Report, Class II, February 1998 Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120 Degree Capsule at 13.4 EFPY

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GE Nuclear Energy Technical Services Business General Electric Company 175 Curtner Avenue, San Jose, CA 95125 GE-NE-B 1100732-01, Revision 1 Final Report Class II February 1998 PLANT FITZPATRICK RPV SURVEILLANCE MATERIALS TESTING AND ANALYSIS OF 1200 CAPSULE AT 13.4 EFPY Prepared by:

T. J. Griesbacý,(irector ATI Consulting

ýR. G. Carey, Engineer Verified by:

Structural Mechanics and Materials Approved by:

) ?W B. J. Branlund, Project Manager Structural Mechanics and Materials

Page 1 of 1 Pagano, Terry From:

Porter, Anne Sent:

Thursday, November 10, 2005 10:33 AM To:

Pagano, Terry

Subject:

FW: Fitz Doc Terry - can you get this please From: LOYD, LELAND Sent: Thursday, November 10, 2005 10:21 AM To: Porter, Anne Cc: Harrison, Douglas; Herrmann, Terry; BARTON, SANDRA

Subject:

Fitz Doc Good Morning Anne I need GE-NE-B1 100732-01 (JAF RPV surveillance materials testing & analysis of 120 degree capsule at 13.4 EFPY (effective, full power years) - including BJB-9907 and attachment to BJB-9907).

If you send it hard copy send to Sandy ENTERGY CORPORATION 1448 SR 333 Russellville, AR 72802 c/o Sandra Barton N-GSB-45 Or; sbart90@entergy.com Sandy (this goes with the OE programs, Lori will know what to do with it)

Thanks!

Leland Loyd License Renewa[

f[*gO9o@tn tergy.com 479/858/4696 11110/2005

GE Nuclear Energy Technical Services Business General Electric Company 175 Curtner Avenue, San Jose. CA 95125 GE-NE-B 1100732-01, Revision I Final Report Class H February 1998 PLANT FITZPATRICK RPV SURVEILLANCE MATERIALS TESTING AND ANALYSIS OF 1200 CAPSULE AT 13.4 EFPY Prepared by:

T. J. GriesbacR,(6ýirector ATI Consulting 9~4 A4" Verified by:

R. G Carey, Engineer Structural Mechanics and M teriaNEW YORK POWER AUTHORITY DOCUMENT REVIEW STATUS K:

Approved by:

B. J. Branlund, Project Mano Structural Mechanics and M ger terials 4EJ ACCEPTED ACCEPTED AS NOTED RESUBMITTAL NOT REQUIRED ACCEPTED AS NOTED.

RESUBMITTAL REQUIRED NOT ACCEPTED Pwmission to procftd doen not cmsfi~tft aceptance - approval of des~g detalS, MacWionsM uWlnS.S otW nMU~ot Or MWaia dev~elop~ed or sewseL by tIe wop~lief and doces not refiev suip~ir from, full comspitance wz or~Dl" M Qitlt!= l.

00 FRMEWEIYw7Aff0*w-

-C Aow-I I,.

I.dP I

123 Main ttre~t White Plains, New York 10601 914 681,6200 S

ewYork'Power SAuthority February 13, 1998 To: GE NUCLEAR ENERGY 175 Curtner Ave San Jose CA 95125 DCME-98-0086 Contract No.

C95-Z0013 Attn:

Ms.

B.

J.

Branlund, Project Manager The document(s) listed below are being returned to you with the status indicated on each document.

Document No.

Rev.

Status GE-NE-BI100732-01 1

1 ACCEPTED JAP RPV SURVEILLANCE MATERIALS TESTING & ANALYSIS OF 120 DEGREE CAPSULE AT 13.4 EFPY (EFFECTIVE FULL POWER YEARS )

Very truly yours, W. Spataro Consulting Metallurgist copies (transmittal only): G. Grochowski

w. attachment: G. Rorke, J. Ellmers, C.

Doc

GE Nuclear Energy GE-NE-B 1100732-01 Revision I IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of the New York Power Authority. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained, or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting information in this document are contained in the contract 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 information.

contained in this document.

ii

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE OF CONTENTS ABSTRACT viii ACKNOWLEDGMENTS ix

1. INTRODUCTION I

2.

SUMMARY

AND CONCLUSIONS 2

2.1

SUMMARY

OF RESULTS 2

2.2 CONCLUSION

S 5

3. SURVEILLANCE PROGRAM BACKGROUND 6

3.1 CAPSULE RECOVERY 6

3.2 RPV MATERIALS AND FABRICATION 6

3.2.1 Fabrication History 6

3.2.2 Material Properties of RPV at Fabrication 7

3.2.3 Surveillance Capsule Specimen Chemical Composition 7

3.3 SPECIMEN DESCRIPTION 7

3.3.1 Charpy. Specimens 8

3.3.2 Tensile Specimens 8

4. PEAK RPV FLUENCE EVALUATION 17 4.1 FLUX WIRE ANALYSIS 17 4.1.1 Procedure 17 4.1.2, -Results 18 4.2 :DETERMINATION OF LEAD FACTOR 18 4.2.1 Procedure-19 4.2.2 Results 20 4.3 :ESTIMATE OF 32 EFPY FLUENCE 23

5. CHARPY V-NOTCH IMPACT TESTING 31 5.1 -IMPACT TEST PROCEDURE

.31 5.2 JIMPACT TEST RESULTS 32 5.3 IRRADIATED VERSUS UNIRRADIATED CHARPY V-NOTCH PROPERTIES 32 5.4 COMPARISON TO PREDICTED IRRADIATION EFFECTS 33 5.4.1 Irradiation Shift 33 5.4.2 Change in USE 34 iii

GE Nuclear Energy GE-NE-B 1100732-01 Revision I

6. TENSILE TESTING 52 6.1 PROCEDURE 52 6.2 RESULTS 53 6.3 IRRADIATED VERSUS UNIRRADIATED TENSILE PROPERTIES 53

7. ADJUSTED REFERENCE TEMPERATURE AND UPPER SHELF ENERGY 63 7.1 ADJUSTED REFERENCE TEMPERATURE AT 32 EFPY 63 7.2 SURVEILLANCE CF ADJUSTMENT 64 7.3 APPLICATION OF CF ADJUSTMENT TO BELTLINE MATERIALS 65 7.4 ART VS.- EFPY 66 7.5 UPPER SHELF ENERGY AT 32 EFPY 66

8. PRESSURE-TEMPERATURE CURVES 70

8.1 BACKGROUND

70 8.2; P-T CURVE METHODOLOGY 72 8.2.1 Non-Beltline Regions 72:

8.2.2 Pressure Test - Non-Beltline, Curve A (Using Bottom Head) 73 8.2.3 Core Not Critical Heatup/Cooldown - Non Belhine, Curve B (Using Feedwater Nozzle/Upper Vessel Region) 75 8.2.4 Example Core Not Critical Heatup/Cooldown Calculation for Feedwater Nozzle/Upper Vessel Region 76 8.2.5 Core Beltline Region 78 8.2.6. Beltline Region - Pressure-Test 79 8.2.7 Calculations for the Belfline Region - Pressure Test 80 8.2.8 Beltline Region - Core Not Critical Heatup/Cooldown 81 8.2.9 Calculations for the Beltline Region Core Not Critical Heatup/Cooldown 82 8.3 CLOSURE FLANGE REGION 83 8.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G 84

9. REFERENCES 92 APPENDICES A. IRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS A-1 B. P-T CURVES VALID TO 24 EFPY B-1 iv

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE OF TABLES TABLE 3-1: CHEMICAL COMPOSITION OF RPV BELTLINE MATERIALS 9

TABLE 3-2: RTDT OF VESSEL MATERIALS 10 TABLE 3-3: RTNDT OF NOZZLE, WELD AND STUD MATERIALS 11 TABLE 3-4: CHEMICAL COMPOSITION OF FITZPATRICK SURVEILLANCE MATERIALS FROM SURVEILLANCE SPECIMEN CHEMICAL TESTS 13 TABLE 4-1:

SUMMARY

OF FITZPATRICK IRRADIATION PERIODS 25 TABLE 4-2: SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM START-UP TO 1/12/96 (13.4 EFPY) USING EMPIRICAL CROSS SECTIONS (GE CORRELATION) 27 TABLE 4-3: MEASURED FLUX VS. THEORETICAL FLUX FOR DOSIMETER AND FLUX WIRES 27b TABLE 5-I: VALLECITOS QUALIFICATION TEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS 35 TABLE 5-2: IRRADIATED CHARPY V-NOTCH IMPACT TEST RESULTS SECOND CAPSULE 36 TABLE 5-3: SIGNIFICANT RESULTS OF IRRADIATED AND UNIRRADIATED CHARPY V-NOTCH DATA 37 TABLE 6-1: TENSILE TEST RESULTS FOR IRRADIATED RPV MATERIALS 54 TABLE 6-2: COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES AT ROOM-TEMPERATURE 54 TABLE 6-3: COMPARISON OF IRRADIATED TENSILE PROPERTIES AT 18ý5 0F 55 TABLE 6-4: COMPARISON OF IRRADIATED TENSILE PROPERTIES AT 50OOF 55 TABLE 7-1: 32 EFPY ART VALUES 67 TABLE 7-2: PLATE EQUIVALENT MARGIN ANALYSIS 68 TABLE 7-3: WELD EQUIVALENT MARGIN ANALYSIS 69 TABLE 8-1: FITZPATRICK P-T CURVE VALUES FOR 32 EFPY 88 TABLE B-1: FITZPATRICK P-T CURVE VALUES FOR 24 EFPY B-5 V

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE OF FIGURES FIGURE 3-1: SURVEILLANCE CAPSULE HOLDER RECOVERED FROM FITZPATRICK (1200 AZIMUTHAL LOCATION CAPSULE - REMOVED AT 13.4 EFPY) 14 FIGURE 3-1(A): CHARPY SPECIMEN CAPSULE IDENTIFICATION (1200 AZIMUTHAL LOCATION CAPSULE - REMOVED AT 13.4 EFPY) 15 FIGURE 3-2. SCHEMATIC OF RPV SHOWING IDENTIFICATION OF VESSEL BELTLINE PLATES AND WELDS 16 FIGURE 4-1: SCHEMATIC OF VESSEL GEOMETRY 28 FIGURE 4-2: RELATIVE FLUX VS. ANGLE AT RPV INSIDE SURFACE 29 FIGURE 4-3: RELATIVE FLUX VS. ELEVATION AT RPV INSIDE SURFACE 30 FIGURE 5-1: ABSORBED ENERGY VS. TEMPERATURE (BASE) 39 FIGURE 5-2: LATERAL EXPANSION VS. TEMPERATURE (BASE) 39 FIGURE 5-3: ABSORBED ENERGY VS. TEMPERATURE (WELD) 40 FIGURE 5-4: LATERAL EXPANSION VS. TEMPERATURE (WELD) 41 FIGURE 5-5: ABSORBED ENERGY VS. TEMPERATURE (HAZ) 42 FIGURE 5-6: LATERAL EXPANSION VS. TEMPERATURE (HAZ) 43 FIGURE 5-7: COMPARISON OF UNIRRADIATED AND IRRADIATED ENERGY DATA (PLATE) 44 FIGURE 5-8: COMPARISON OF 1 ST AND 2ND CAPSULE ENERGY RESULTS (WELD) 45 FIGURE 5-9: COMPARISON OF IST AND 2ND CAPSULE ENERGY RESULTS (HAZ) 46 FIGURE 5-10: COMPARISON OF LATERAL EXPANSION RESULTS (BASE) 47 FIGURE 5-11: COMPARISON OF LATERAL EXPANSION RESULTS (WELD) 48 FIGURE 5-12: COMPARISON OF LATERAL EXPANSION RESULTS (HAZ) 49 FIGURE 5-13: TANH CURVE-FITTED RESULTS FOR COMBINED BASELINE DATA (PLATE) 51 FIGURE 5-14: AT3, VS. FLUENCE SHOWING PLATE DATA WITH FITTED RESULTS 52 FIGURE 6-1. TYPICAL ENGINEERING STRESS-STRAIN FOR IRRADIATED RPV MATERIALS 56 FIGURE 6-2: FRACTURE LOCATION AND NECKING BEHAVIOR FOR IRRADIATED BASE METAL TENSILE SPECIMENS 57 FIGURE 6-3: FRACTURE LOCATION AND NECKING BEHAVIOR FOR IRRADIATED WELD METAL TENSILE SPECIMENS 58 FIGURE 6-4: FRACTURE LOCATION AND NECKING BEHAVIOR FOR IRRADIATED HAZ TENSILE SPECIMENS 59 FIGURE 6-5: FRACTURE APPEARANCE FOR IRRADIATED BASE METAL TENSILE SPECIMENS 60 FIGURE 6-6: FRACTURE APPEARANCE FOR IRRADIATED WELD METAL TENSILE SPECIMENS 61 FIGURE 6-7: FRACTURE APPEARANCE FOR IRRADIATED HAZ TENSILE SPECIMENS 62 FIGURE 7-1: ART VS. EFPY FOR LIMITING BELTLINE PLATE AND WELD 69b FIGURE 8-1: PRESSURE TEST CURVE (CURVE A) VALID TO 32 EFPY 85 FIGURE 9-2: NON-NUCLEAR HEATUP/COOLDOWN (CURVE B) VALID TO 32 EFPY 86 FIGURE 8-3: CORE CRITICAL OPERATION (CURVE C) VALID TO 32 EFPY 87 vi

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE OF FIGURES (continued)

FIGURE B-I: PRESSURE TEST CURVE (CURVE A) VALID TO 24 EFPY B-2 FIGURE B-2: NON-NUCLEAR HEATUP/COOLDOWN (CURVE B) VALID TO 24 EFPY B-3 FIGURE B-3: CORE CRITICAL OPERATION (CURVE C) VALID TO 24 EFPY B-4 vii

GE Nuclear Energy GE-NE-B 1100732-01 Revision I ABSTRACT The surveillance capsule at the 1200 azimuthal location was removed at 13.4 EFPY from the FitzPatrick reactor in November 1996. The capsule contained flux wires for neutron fluence measurement, and Cbarpy test specimens and tensile test specimens for material property evaluations.

The flux wires were evaluated to determine the fluence experienced by the test specimens. Charpy V-Notch impact testing and tensile testing were performed to establish the properties of the irradiated surveillance materials.

The irradiated Charpy data for the base material specimens were compared to available unirradiated data to determine the shift in Charpy curves due to irradiation. The results indicate a shift lower than the predictions of Regulatory Guide 1.99 Revision 2 [Rev. 2]. Since two sets of credible data sets were available for the plate material, the Adjusted Reference Temperature (ART) calculations for vessel base materials were adjusted in accordance with Rev. 2. For the vessel weld metal, no unirradiated data was available and the predictions of Rev. 2 were used to calculate ART.

The flux wire results combined with the lead factor were used to estimate the 32 EFPY fluence. The fluence calculations included the effects of a 105% power uprate. The resulting estimated fluence showed a reduction of 22 percent compared with the previous nominal 32-EFPY fluence estimate consistent with the fluence used for the Technical Specification Pressure-Temperature (P-T) Curves.

P-T.,Curves were prepared based on the new projected fluence levels for both 32 EFPY and 24 EFPY.

viii

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 ACKNOWLEDGMENTS The author gratefully acknowledges the efforts of other people towards completion of the contents of this report.

Cask shipping & receipt, and capsule disassembly were performed by I. B. Myers and R. D. Rimmer. Charpy testing was completed by G. E. Dunning and B. D. Frew. Tensile testing was performed by S. B. Wisner. Chemical composition analysis was performed by P. S. Wall.

Flux wire testing and analysis was performed by L. Kessler, R_ M. Kruger and R_ D. Reager.

Fluence and lead factor calculations were performed by D. R_ Rogers, H. A-Careway and S. S. Wang.

Assistance with the capsule evaluation was provided by B. N, Burgos of ATI Consulting. Project management was conducted by B. J. Branlund.

ix

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

1. INTRODUCTION Part of the effort to assure reactor vessel integrity involves evaluation of the fracture toughness of the vessel ferritic materials. The key values which characterize a material's fracture toughness are the reference temperature of nil-ductility transition (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].

Appendix H of I0CFR50 [3] and ASTM E185-70 [4] establish the methods to be used for surveillance of the James A. FitzPatrick (FitzPatrick) reactor vessel materials. The second vessel surveillance specimen capsule required by IOCFR50 Appendix H [3] was removed from FitzPatrick in November 1996. 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 and tensile test specimens fabricated using base metal from the beltline region, as well as weld metal from a similar heat of material as the beltline welds. The impact and tensile specimens were tested to establish properties for the irradiated materials.

The results of the surveillance specimen testing are presented in this report, as required per IOCFR50 Appendices G and H [1 & 3]. The irradiated material properties are compared to available unirradiated properties to determine the effect of irradiation on material toughness for the base and weld materials, through Charpy testing.

Irradiated tensile testing results are provided and are compared with unirradiated data to determine the effect of irradiation on the stress-strain relationship of the materials.

Pressure-temperature (P-T) curves are included in this report which have been developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. The P-T curves are established to the requirements of IOCFR50, Appendix G (1] to assure that brittle fracture of the reactor vessel is prevented.

I

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

2.

SUMMARY

AND CONCLUSIONS 2.1

SUMMARY

OF RESULTS The 1200 azimuth position 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 E185-82 [6]. The methods and results of the testing are presented in this report as follows:

Section 3:

Surveillance Program Background

" RPV Materials and Fabrication

" Material Properties

" Surveillance Specimen Chemical Composition

" Specimen Description Section 4:

Peak RPV Fluence Evaluation Section 5:

Charpy V-Notch Impact Testing Section 6:

Tensile Testing Section 7:

Adjusted Reference Temperature and Upper Shelf Energy Section 8:

Pressure-Temperature Curves The significant results of the evaluation are below:

a.

The 1200 azimuth position capsule was removed from the reactor after 13.4 EFPY (Effective Full Power Years) of operation. The capsule contained 2 sets of 3 flux wires: nickel (Ni), copper (Cu), and iron (Fe). There were 24 Charpy V-Notch specimens in the capsule: eight (8) each of plate (base) material, weld material, and heat affected zone (HAZ) material. The capsule also contained eight (8) tensile specimens: three plate material, three weld material, and two HAZ material. (See Sections 3.1 and 3.3) 2

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

b.

The chemical composition of copper (Cu) and nickel (Ni) for the irradiated surveillance materials was determined from a chemical composition analysis. The best estimate values for the surveillance material chemistries were calculated as averages of the available baseline and irradiated data.

The best estimate values for the surveillance plate are 0 11% Cu and 0.60% Ni, and are 0.29% Cu and 0.71% Ni for the surveillance weld. (See Table 3.4)

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 17 2

E >1 MeV flux) received by the surveillance specimens was 5.0 x 10 n/cm at removal (13.4 EFPY-See Section 4.1.2).

d.

A neutron transport computation had been performed based on the first surveillance capsule.

Relative flux distributions in the azimuthal and axial directions were previously developed in Reference 8. The lead factor was 0.79, relating the surveillance capsule flux to the peak inside surface flux. The lead factor was calculated after the second capsule was removed at 13.4 EFPY, and determined to be 0.68. A lead factor of 0.68 was used for all calculations in this report (See Section 4.2.2).

e.

The surveillance Charpy V-Notch specimens were impact tested at temperatures selected to define the upper shelf energy (USE) and the transition of the Charpy V-Notch curves for the plate, weld, and HAZ materials.

Measurements were taken of absorbed energy, lateral expansion and percentage shear. From absorbed energy and lateral expansion curve-fit results, the values of USE and of index temperature for 30 ft-lb, 50 ft-lb and 35 mils lateral expansion (MLE) were obtained (see Table 5-3). Fracture surface photographs of each specimen are presented in Appendix A.

f.

The irradiated tensile specimens were tested at room temperature (70'F), at reactor operating temperature (550'F) and at 1851F as an intermediate temperature. Unirradiated base material results, as well as results from the first capsule, were available for comparison (See Tables 6-1 through 6-4.)

3

GE Nuclear Energy GE-NE-B 1100732-01 Revision I

g.

The curves of irradiated and unirradiated Charpy specimens established the 30 ft-lb shifts. The plate material showed a 15'F shift and a 12 ft-lb decrease in USE (9% decrease).

These values were not calculated for the weld, as no unirradiated data was available (See Table 5-3).

17 2

h.

The measured shift of 15°F for plate material for a fluence of 5.0 x 10 n/cm, was within the Rev. 2 [7] range predictions (ARTDT++/-2a) of -12°F to 56°F. Since two credible data sets are available for the plate material, the surveillance adjustment (Section 7) was applied to the vessel base plates. The measured shift values were not obtained for the weld as no unirradiated data was available. The best estimate chemical composition for the surveillance weld material was used for evaluating the projected shift of the surveillance weld data (See Table 5-3).

Ig 2

i.

The 32 EFPY RPV peak fluence prediction is 1.81 x 10 n/cm at the vessel wall, based on the flux wire test and lead factor. This is 22% less than the previously 18 2

established nominal 32 EFPY fluence prediction (2.32 x 10 n/cm ) [5].

The LS 2

32 EFPY fluence prediction is 1.31 x 10 n/cm at 1/4 T. (See Section 4.3)

j.

The adjusted reference temperature (ART = Initial RTNDT + ARTNDT + Margin) was predicted for each beltline material, based on the methods of Regulatory Guide 1.99, Rev. 2.

The ART for the limiting material, Axial Weld Heat 27204/12008, at 32 EFPY is 109'F and is lower than the 200IF requirement of IOCFR50 Appendix G [1] and Rev. 2 [7]. (See Table 7-1)

k.

An update of the beltline material USE values at 32 EFPY was performed using the Reg. Guide 1.99, Rev. 2 methodology.

The equivalent margin analyses demonstrate that 1 OCFR50, Appendix G safety requirements are satisfactorily met for FitzPatrick. (See Tables 7-2 and 7-3)

P-T curves were developed for three reactor conditions: pressure test (Curve A),

non-nuclear heatup and cooldown core not critical operation (Curve B), and core critical operation (Curve C) curves which are valid for up to 32 EFPY of operation. The beltline curve is more limiting for Curve A at pressures above approximately 550 psig. For Curves B and C, the beltline curves are limiting for pressures above approximately 600 psig. The P-T curves for 32 EFPY are shown in Figures 8-1 through 8-3, and the P-T curves for 24 EFPY are shown in Appendix B, Figures B-1 through B-3 4

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

2.2 CONCLUSION

S The requirements of IOCFR50 Appendix G [1] deal 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 30 ft-lb shift for the base material was less than the Rev. 2 prediction, and therefore the ART values for beltline plates were modified in accordance with Position 2 of Rev. 2. The changes in USE for the survillance plate are bounded by the Regulatory Guide 1.99 Revision 2 predictions and associated deviations.

b.

The values of ART and USE for the reactor vessel beltline materials are expected to remain within the limits of IOCFR50 Appendix G [1] for at least 32 EFPY of operation.

5

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

3. SURVEILLANCE PROGRAM BACKGROUND 3.1 CAPSULE RECOVERY The reactor pressure vessel (RPV) surveillance program consists of three surveillance capsules at 30*, 120', and 3000 azimuths at the core midplane. The specimen capsules are held against the RPV inside surface by a spring loaded specimen holder. Each capsule is expected to receive equal irradiation because of core symmetry. The first capsule (30' azimuth) was removed in April 1985 after 5.98 EFPY. During the November 1996 outage, the second surveillance capsule was removed from the 120' azimuthal location. The capsule was cut from its holder assembly and shipped by cask to the GE Vallecitos Nuclear Center (VNC), where testing was performed.

Upon arrival at VNC, the capsule was examined for identification.

The identification number stamped on the capsule corresponded to FitzPatrick, as specified by GE drawings, 117C3739 (Outline Specimen Holder) and 921D465 (Surveillance Program), for the FitzPatrick 1200 surveillance materials.

The general condition of the capsule as received is shown in Figure 3-1. The specimen holder contained 2 sets of 3 flux wires (iron, copper, and nickel), three Charpy specimen capsules each containing 8 plate, weld, or HAZ Charpy specimens in a sealed helium environment, and four tensile specimen capsules (together containing 3 base, 3 weld and 2 HAZ tensile specimens in a sealed helium environment).

3.2 RPV MATERIALS AND FABRICATION 3.2.1 Fabrication History The FitzPatrick RPV is a 220.75 inch inside diameter BWR/4 design. Construction was performed by Combustion Engineering (CE) under the 1965 edition of the ASME Code through the 1966 Winter Addenda.

The shell and head plate materials are ASME 'SA533, Grade B, Class 1 low alloy steel (LAS). The nozzles and closure flanges are ASME SA508 Class 2 LAS, and the closure flange bolting materials are ASME A540 Grade B24 LAS [8]. Submerged arc or shielded metal arc welding of plates was followed by post-weld heat treatment at 11500F. The fabrication impact test specimens were given a simulated post weld heat treatment at 6

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 1150°F +/-25°F, held 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> followed by furnace cooling to below 6001F, then air cooled. The identification of plates and welds in the beltline region is shown in Figure 3-2.

3.2.2 Material Properties of RPV at Fabrication Material certification records were retrieved from GE Quality Assurance (QA) records to determine chemical and mechanical properties of the vessel materials. The retrieved information for the beltline materials is documented in [5].

Table 3-1 shows the chemistry data for the beltline materials. Properties of the beltline materials and materials at other locations of interest are presented in Tables 3-2 and 3-3.

3.2.3 Surveillance Capsule Specimen Chemical Composition Samples were taken from the irradiated base and weld Charpy specimens after they were tested.

Chemical analyses were performed using a Spectraspan III plasma emission spectrometer. Each sample was dissolved in an acid solution to a concentration of 40 mg steel per ml solution. The spectrometer was calibrated for determination of Mn, P, Ni, Mo, V, Cr, Si and Cu by diluting National Institute of Standards and Technology (NIST) Spectrometric Standard Solutions. The phosphorus calibration involved analysis of five reference materials from NIST with known phosphorus levels. Analysis accuracies are +/-0.005% (absolute) of reported value for phosphorus and +/-5% (relative) of reported value for other elements. The chemical composition results are given in Table 3-4 for both irradiated and baseline surveillance plate and irradiated weld materials.

The baseline plate data was taken from CE material certification records as documented in [5] for the plate surveillance specimens; no baseline data was available for the weld material.

3.3 SPECIMEN DESCRIPTION The surveillance capsule holder contained 24 Charpy specimens: base metal (8), weld metal (g), and HAZ (8). The holder also contained 2 sets of 3 flux wires (iron, nickel, and copper) and eight (8) tensile specimens (three base, three weld and two HAZ). The chemistry and fabrication history for the Charpy and tensile specimens are described in this section.

7

GE Nuclear Energy GE-NE-B 1100732-01 Revision I 3.3.1 Charpy Specimens The fabrication of the Charpy specimens is described in the CE drawings of the surveillance test program. All materials used for specimens were beltline materials taken from the lower intermediate shell course.

The base metal specimens were cut from plate G-3414-2, heat number C3278-2. The test plates received the same heat treatment as plate heat no. C3278-2, including the post-weld heat treatment for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1150'F +/- 25°F. The Charpy specimens were removed from plate heat no. C3278-2 and machined from the 1/4 T and 3/4 T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction). The Charpy specimens had been stamped on one end with the fabrication codes as listed in GE surveillance program drawings for FitzPatrick.

The weld metal and HAZ Charpy specimens were fabricated by welding together pieces of plates G-3414-1 and G-3414-2 with a weld identical to longitudinal seam weld 1-233 in the RPV beltline.

Welding records obtained from CE indicate the surveillance weld to be a submerged arc weld representive of the vessel beltline circumferential weld. The welded test plates received stress relief heat treatment at 1150'F +/-25'F to simulate the RPV fabrication conditions. The weld and HAZ specimens were cut from the material avoiding the volume near the root of the welds.

The base metal orientation in the weld and HAZ specimens was longitudinal.

3.3.2 Tensile Specimens Fabrication of the surveillance tensile specimens is also described in the CE surveillance program drawings. The materials, chemical compositions, and heat treatments for the tensile specimens are the same as the corresponding Charpy specimens. The identifications of the base, weld and HAZ surveillance specimens are described in Reference 8.

8

GE Nuclear Energy GE-NE-B 1100732-01, Revision I TABLE 3-1: CHEMICAL COMPOSITION OF RPV BELTLINE MATERIALSO Comx9osition by Whht Percent Coranosition bv Weight Percent IdentifiQAtIon IleatL* ot No.

Cut Ni 1<C PI Mnj

, S VSi MMo I

I~

t 4-f 4

4-

+

I PLATES:

Lower Shell:

G-3415-IR G-3415-3 G-3415-2 Lower-Intermed, Shell:

G-3413-7 G-3414,2c G-3414-1 WELDS:

Lower Longitudinal:

2-233 A,B,C C3394-1 C3376-2 C3103-2 C3368-1 C3278-2 C3301-1 27204/12008 Flux 1092 Lot 3774 13253/12008 Flux 1092 Lot 3947 305414 Flux 1092 Lot 3947 0,1 1 r 0.1 3b

0. 14 b 0.12b 0.1le

0. 18 b 0.56 0.60 0.57 0.50 0.600 0.57 0.21 0.22 0,23 0.19 0.20 0.18 N/A N/A 0.14 1.32 1.33 1.36 1.30 1.26 1.36 1.16 N/A 1.45 0.015 0.015 0.012 0.015 0.011 0.008 0.013 N/A 0,012 0.017 0.017 0.015 0.017 0.016 0.015 0.007 N/A 0.01 0.26 0.22 0.26 0U22 0,22 0,29 0,21 N/A 0.18 0.47 0.48 0.46 0.45 0.48 0.46 0.46 N/A 0.51 Lower Int. Long.:

1-233 A,B,C Lower to Lower -Int. Girth:

1-240 0.6 0 9d a

b C

d f

Data from CMTR Reports, GE QA Records and [5] except as noted below Cu values taken from Lukens Steel letter to NYPA dated 10/14/85 [19]

Surveillance plate Best estimate Cu and Ni weld values obtained from CE Owners Group report [18]

Average chemistry of surveillance plate from Table 3-4 Cu content from Generic Letter 92-01 response [21 ]

9 1

I I

I I

I I

I I.

I

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 3-2: RTNIT OF VESSEL MATERIALS COMPONENT ID HEAT TEST CHARPY (T.0T-60)

DROP RTNDT TEMP.

ENERGY (0F)

WEIGHT

(-F)

(OF)

(FT-LB)

NDT I(F)

PLATES & FORGINGS:

Top Head & Flange Dollar Plate Top Head Torus Top Head Flange Shell Courses Upper Shell Flange Upper Shell Upper Int. Shell Low-Int Shell Lower Shell Bottom Head Dollar Plate Bottom Head Torus G-3412 G-341 1-1 G-3411-2 G-3402 G-3401 G-3413-4 G-3413-5 G-3413-6 G-3413-1 G-3413-2 G-3413-3 G-3413-7 G-3414-1 G-3414-2 G-3415-1R G-3415-2 G-3415-3 G-3410 G-3407-1 G-3408-1 G-3409 C-2869-5 C-3055-1 C-3055-1 4P-1885 2V595 B-7255-1 C-3229-2 B-7291-1 C-3116-1 C-3121-2 C-3158-2 C-3368-1 C-3301-1 C-3278-2 C-3394-1 C-3103-2 C-3376-2 C-2917-3 C-2851-1 C-3055-2 C-2906-3 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 40 10 10 10 10 83 98 98 66 117 70 50 81 65 31 95 61 60 45 53 41 43 38 83 53 36 70 73 73 87 94 76 68 69 91 48 87 55 63 77 71 48 51 36 72 73 43 72 118 118 96 117 71 82 65 79 35 76 45 49 58 52 49 49 36 75 66 35

-20

-20

-20

-50

-50

-20

-20

-20

-20 18

-20

-10

-18

-10

-20

-2 24 8

-20

-20 10

-10

-10

-10 30 10

-10

-10

-10

-10 10

-10

-50

-40

-30

-10

-10

-10

-10

-10

-10

-10

-10

-10

-10 30 10

-10

-10

-10

-10 18

-10

-10

-18

-10

-10

-2 24 8

-10

-10 10

~

S 10

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 3-3: RTNDT OF NOZZLE, WELD AND STUD MATERIALS COMPONENT ID HEAT TEST CHARPY (Ts-60) DROP RTmeT TEMP.

ENERGY (OF)

WEIGHT (OF)

(OF)

(FT-LB)

NDT (OF )

LNozzles:

70 93 Recirc. Outlet Nozzle Recirc. Inlet Nozzle Steam Outlet Nozzle Feedwater Nozzle Core Spray Nozzle Top Head Instrumentation Nozzle Vent Nozzle Jet Pump Instrumentation Nozzle CRD Hyd. Sys. Return Drain Nozzle G-3419-1 G-3419-2 G-3436-1 G-3436-2 G-3436-3 G-3436-4 G-3436-5 G-3436-6 G-3436-7 G-3436-8 G-3436-9 G-3436-1 0 G-3420-1 G-3420-2 G-3420-3 G-3420-4 G-3421-1 G-3421-2 G-3421-3 G-3421-4 G-3422-1 G-3422-2 G-2921-3 G-2921-4 G-2920-2 G-3424-1 G-3423 G-2085 EV-9781 AV-1872 E21VW-104J 10 E21VW-104J2 E21VW-104J9 E21VW-104J7 E21VW-104J6 E21VW-104J3 E21VW-104J4 E21VW-104J8 E21VW-104J5 E21VW-104J1 EV-9754 EV-9775 EV-9775 AV-1 576 EV-9741 EV-9741 EV-9741 AV-1607 EV-9741 EV-9741 EV-9781 AV-2379 AV-2374 EV-9792 EV-9143 2106172 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 1031111110 82 95 86 84 89

.106 101 73 110*

82 66 62 30 65 92 69 30 40 54 82 117!

94 101 76 109 94 109 114 116 93 105 40 75 52 75 75 67 36 96

107 79 107 77 116 102 118 103 82 36 78 48 73 9O 68 32 71 76 86 72

-20

-20

-20

-20

-20

-20

-20

-20

-20

-20

-20

-20

-20 8

-20 20

-20

-20

-20

-20

-30

-50

-20

-20

-20

-20

-20

-20

-10 0

<40

<40

<40

<40

<40

<40

<40

<40

<40

<40

-10

-10

-10 0

10 10

-20 0

0 10 0

-10 0

0

-20

-10 0

30 30 30 30 30 30 30 30 30 30

-10 8

-10 20 10 10

-20 20 0

10 0

-10 0

0

-20 20 56 65 89 74 69 72 90 108 145 1821185 144 1441144 Y:

1121 94 180 96 1108192

________________________________.1 ______________.3

.L __________.L....3....I.........-J.

II

GE Nuclear Energy GE-NE-B 1100732-01 Revision I COMPONENT ID HEAT TEST CHARPY (TsoT-6 0)

DROP RTNoT TEMP.

ENERGY

('F)

WEIGHT

('F)

(°F)

(FT-LB)

NDT WELDS:

Vertical Welds Lower Shell 2-233 A,B,C 27204112008 10 63 60 49

-48

-48 Lower-lnt Shell 1-233 A,B,C 13253/12008 10 60 64 56

-50

-50 Girth Welds Lower to Lower-lnt Shells 1-240 305414 10 82 66 80

-50

-50 LST STUDS:

G-3134-1 37385 10 39 40 39 70 OK G-3134-2 37677 10 60 55 57 70 OK 12

GE Nuclear Energy GE-NE-B1100732-01, Revision I TABLE 3-4: CHEMICAL COMPOSITION OF FITZPATRICK SURVEILLANCE MATERIALS FROM SURVEILLANCE SPECIMEN CHEMICAL TESTS Metal Metal Mn Ni Cu Mo Si Cr P

Sample ID Sample (wt%)

(wt%)

(wt%)

(wt%)

(wt%)

(wt%)

(wt%)

Type 5CL.

Base 1.40 0.62 0,11 0.48 0.07D 0.11 0.011 5CMa Base 1.30 0.63 0.12 0.50 0.0 6 b 0.11 0.010 29283 Base 1.17 0.58 0.11 0.45 0.36 0.09 0.013 29285 Base 1,25 0.61 0.11 0.46 0.16 0.10 0.013 29286 Base 1.20 0.60 0.11 0.46 0.19 0.10 0.011 LPI-28c Base 1.43 0.62 0.10 0.42 0.24 N/A 0.018 Baseline*

Base 1.26 0.57 0.13c 0.48 0.22 N/A 0.011

DataAvg, 1.29 0.60 0.11 0.46 0.23 0,10 0.012 SO Dev.

0.10 0.02 0.01 0.03 0.08 0.01 0.003 5DL" Weld 1.50 0.72 0.31 0.50 0,06u 0.04 0.015 5DMa Weld 1.40 0.72 0.31 0.51 0.0 6b 0.04 0.014 29289 Weld 1.36 0.70 0.30 0.48 0.38 0.04 0.014 29295 Weld 1.25 0.70 0.23 0.47 0,41 0.04 0.014 29297 Weld 1.39 0.74 0.31 0.49 0,52 0.04 0.012 DataAvg.

1.38 0.72 0.29 0.49 0.44 0.04 0.014 Std. Dev.

0.09 0.02 0.03 0.02 0.07 0.001 0.001 Chemical analysis of tensile specimens from 30' azimuthal capsule location (Ist capsule report) [8].

b Si results may be low due to precipitiation during dissolution heating (Results not used in Average).

C Data taken from the BWROG Supplemental Surveillance Program for the FitzPatrick Plant.

d Taken from original fabrication records (see Table 3-1).

e Cu value taken from Lukens Steel letter to NYPA dated 10/14/85 [191 13 I

I1II 1

1 I

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 FIGURE 3-1: SURVEILLANCE CAPSULE HOLDER RECOVERED FROM FITZPATRICK (120 AZIMUTHAL LOCATION CAPSULE - REMOVED AT 13.4 EFPY) 14

GE Nuclear Energy GE-&NE-B 1 1' 00732-01 Revision i FIGURE 3-1(A): CHARPY SPECIMEN CAPSULE IDENTIFICATION (120- AZIMUTHl.U LOCATION CAPSULE - REMOVED AT 13.4 EFPY)

GE Nuclear Energy GE-NE-B 1100732-01 Revision I FT

• k TOP HEAD ENCLOSURE 1--

CLOSURE FLANGE REGION oL I

w I

© UPPER SHELL di UPPER INTERMEDIATE SHELL PLATE G-3413-7 HEAT C336$-1 CORE BELTLINE REGION BOTTOM HEAD ENCLOSURE LOWER INTERMEDIATE SHELL PLATE G-3414-1 HEAT C3301-1 LONGITUDINAL SEAM WELD 1-233 LOWER SHELL WELD 2-233 PLATET G-3415-I2 HEAT C3103-2 PLATE G-341 5-1 R.

HEAT C3394-1 PLATE G-3414-2

/

HEAT C3278-2 CIRCUMFERENTIAL 7

GIRTH WELD 1-240 PLATE G-3415-3 HEAT C3376-2 Kk I

1

--- 1 FIGURE 3-2. SCHEMATIC OF RPV SHOWING IDENTIFICATION OF VESSEL BELTLINE PLATES AND WELDS 16

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

4. PEAK RPV FLUENCE EVALUATION Flux wires removed from the 120' location 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.

4.1 FLUX WIRE ANALYSIS 4.1.1 Procedure The surveillance capsule contained 2 sets of 3 flux wires: iron, nickel, and copper. Each wire was removed from the capsule, cleaned with dilute acid, weighed, mounted on a counting card, and analyzed for its radioactivity content by gamma spectrometry.

Each iron wire was analyzed for Mn-54 content, each nickel wire was analyzed for Co-58 content, and each copper wire for Co-60 at calibrated source-to-detector distances with 170-cc Ge and 100-cc Ge(Li) gamma detectors used in conjunction with a Nuclear Data 6700 multichannel analyzer system.

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 were derived from records provided by New York Power Authority are shown in Table 4-1.

Zero power days between fuel cycles are listed as well.

From.the flux wire activity measurements and power. history, reaction rates for Fe-54 (n,p) Mn-54, Ni-58 (n,p) Co-58, and Cu-63 (n,ox) Co-60 were calculated. The E >1 MeV fast flux reaction empirical cross sections for the iron, nickel, and copper wires are 0.182 barn, 0.234 barn and 0.00318 barn, respectively. The calculated fluence result from the iron flux wire was used. The fluence result from the iron specimen was confirmed by the Ni and Cu flux wires, with all three results differing by less than 10%. The GE empirical activation cross sections are consistent with other transport code cross sections, and parallel calculations were performed using the both the empirical and transport code cross sections [201. However, the fluence results obtained from the empirical cross sections are recommended since they yield approximately 4%

higher estimates of RPV fluence. These data functions were applied to BWR pressure vessel locations based on water gap (fuel to vessel wall) distances. The cross sections for > 0.1 MeV flux were determined from the measured 0.1 to I MeV cross section ratio of 1.6 [11].

17

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 4.1.2 Results The measured activity, reaction rate and full-power flux results for the 120' location surveillance capsule are given in Table 4-2. The E > I MeV flux values were calculated by dividing the wire reaction rate measurements by the corresponding cross sections, factoring in 17 2

the local power history for each fuel cycle. The fluence result, 5.0 x 10 r/cm (E > I MeV),

was obtained by using the following equation:

(DCu zZDfp tiPi (4-1)

where, DCu ofP 1i PA

= fluence measured by the Cu dosimeters, r/cm 2

= full power flux value for Cu, nrcm'-s

= operating time, s

= full power fraction as shown in Tables 4-1 through 4-3.

The accuracies of the values in Ta-ble 4-2 for a 2a deviation are influenced by the following sources of error:

+ 2%

+ 15%

+ 10%

counting rates power history cross sections The uncertainty in the E > I MeV fluence is approximately +20% (2c).

This analysis is performed using the GE empirical activation cross sections. A parallel analysis using cross sections from a transport code was made, but is not preferred, because its resulting fluences were approximately 4% lower for all three of the flux wires.

4.2 DETERMINATION OF LEAD FACTOR The flux wires from the surveillance capsule are used to determine the fast neutron (E > I MeV) fluence at the location of the capsule as described in Section 4.1. However, the capsule and flux wires are not located where the peak vessel fluence occurs. A calculated lead factor is used to 18

GE Nuclear Energy GE-NE-B 1100732-01 Revision I relate the fluence at the location of the wires to the peak fluence at the vessel. The lead factor is defined as the ratio of the fast neutron fluence at the surveillance capsule to the peak fluence at the vessel inside surface. A neutron transport analysis was performed to determine the effective full power fast neutron flux distribution at the reactor pressure vessel. The lead factor was evaluated as the ratio of the calculated effective full power fast neutron fluxes at the capsule and vessel peak flux locations. Calculation of the fluxes and lead factor requires modeling of the reactor geometry and materials and depends on the distributions of power density and coolant voids in the core. The lead factor was calculated for the FitzPatrick geometry, using data for a typical operating cycle to determine power shape and void distribution. The lead factor was not adjusted for the 105% power uprate, as the fluxes were assumed to increase linearly with power.

The methods used to calculate the lead factor are discussed below.

The NRC is developing Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", which will include guidance concerning acceptable methods and assumptions for determining the pressure vessel fluence. At this time, the draft has not been finalized for issuance as a Regulatory Guide. However, while the specific regulatory requirements are still subject to change, it is believed that the analysis described in this section is consistent with the intent of the draft guide.

4.2.1 Procedure The lead factor for the RPV inside wall was determined by using a combination of two separate two-dimensional neutron transport computer analyses. The first of these established the azimuthal and radial variation of flux at the fuel midplane elevation.

The second analysis determined the relative variation of flux with elevation. The azimuthal and axial distribution results were combined to provide a simulation of the three-dimensional distribution of flux. The ratio of fluxes, or lead factor, between the surveillance capsule location and the peak flux locations was obtained from this distribution.

The DORT computer program, which utilizes the discrete ordinates method to solve the Boltzmann transport equation in two dimensions, was used to calculate the spatial flux distribution produced by a fixed source of neutrons in the core region. The analysis considered neutrons with energies above 0.1 MeV and used 29 energy groups above this threshold. Angular dependence of the neutron scattering cross-sections was approximated by a third-order Legendre polynomial (P-3) expansion. The DORT calculations were run using Sg angular quadrature.

19

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The azimuthal distribution was obtained with a model specified in (R,O) geometry, assuming eighth-core symmetry with reflective boundary conditions at 0' and 45'. In this model, O-30' is symmetrically equivalent to the 1200 capsule location. A schematic of the (R,0) model is shown in Figure 4-1.

The model incorporates inner and outer core regions, bypass water region, shroud, downcomer water region, and a vessel plus liner region. The portion of the core inside a radius of 133 cm was not included because it will not significantly influence the flux distribution at the vessel. The spatial mesh contained 155 steps of varying sizes in the radial dimension. The azimuthal mesh step was specified to be 1/20 and was reduced to 1/4' in the vicinity of the capsule, resulting in a total of 98 azimuthal intervals. The (Re) model used core region material compositions and neutron source densities for the core midplane elevation (75 inches above the bottom of active fuel). This is near the elevation of the capsule, which is centered at 72.31 inches above the bottom of the active fuel. The neutron source densities and coolant mass densities were based on cycle-average values for the selected representative operating cycle. The output of this calculation provided the distribution of flux as a function of azimuth and radius at reactor. midplane. The azimuth of the peak flux and its magnitude relative to the flux at the 30' capsule/flux wire azimuth were determined from this distribution.

The calculation of the axial flux distribution was performed in (RZ) 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 cylinder radius was specified to be equal to the radius of the outermost comer of the core, which is located at an azimuth of approximately 39.3*. The core model contained inner and outer material regions for each of 25 axial fuel nodes (total of 50 core regions). Source densities and coolant densities in these regions were based on cycle-average values for the representative cycle. The elevation of the peak flux at the reactor vessel inside surface and the magnitude of the peak flux relative to the flux at the surveillance capsule elevation were determined from the (RZ) flux distribution results.

4.2.2 Results The relative distribution of flux at the RPV base metal inside surface vs. azimuthal angle obtained from the (RO) calculation is shown in Figure 4-2.

The relative distribution of flux versus elevation at the RPV inside surface from the (R,Z) calculation is shown in Figure 4-3.

The azimuthal distribution (Figure 4-2) indicates that the 8 flux maxima at the vessel base metal inside surface occur at azimuthal locations which are displaced by 42.75' from the RPV quadrant reference axes (0', 900, etc.). From the R,Z results (Figure 4-3), the peak is estimated to occur at 20

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 an elevation about 79 inches above the bottom of the active fuel. The calculated core midplane E > 1 MeV flux at the (RO) coordinates corresponding to the equivalent capsule center position (0 = 30-, R = 109.19 inches) was 1.382 x10 9 n/cm2/s. This was multiplied by the ratio of flux at the capsule elevation to flux at midplane (0.996), as determined from the (RZ) calculation, resulting in a calculated flux at the capsule location which rounds to 1.38 x 109 n/cm 2/s. The peak flux at the vessel surface (R = 110.375 inches) was similarly obtained by multiplying the calculated midplane flux of 2.015 x109 n/cm2 Is at the peak azimuth by an axial adjustment factor of 1.003 from the (RZ) calculation. The resulting peak flux estimate is 2.02 x 1 09 n/cm 2/s.

Consequently, the lead factor is 1.38 x 10 /2.02 x 109 =0.68.

The calculated capsule full power flux of 1.38x109 n/cm2/s obtained with this model is about 16 % higher than the capsule dosimetry result of 1.19x109 nicm2/s.

The indicated agreement between the analytical and experimental results is within the uncertainties associated with those results and is considered good. It is estimated that the la uncertainty in the calculated flux magnitudes is on the order of 25 - 30 %. However, since the lead factor is determined from the ratio of two calculated fluxes which have sources of error in common, the la uncertainty in the lead factor is estimated to be no more than 15 %.

Use of a lead factor calculated on the basis of the model described above is consistent with current GE practice for estimation of the peak vessel fluence. Application of the lead factor to the capsule dosimetry results yields an estimated end-of-cycle 12 peak fluence of 5.Ox1 017 /

0.68=7.4x10 7 n/cm2 and an estimated peak full power flux of 1.19x10 9/0.68 = 1.75 x 109 n/cm2/s at the vessel inside surface. Since the estimated 1ca uncertainty in the dosimetry results is 10 %

and the estimated la uncertainty in the lead factor is 15%, the combined overall 1 uncertainty in the projected peak values is estimated to be about (102 + 15)0.5 = 18%.

The analysis model discussed above did not include the effects of the material specimens and specimen holder on the local neutron flux. A second calculation was performed in (R,0) geometry with a model which incorporated regions which simulated the material specimens and holder. The densely packed material specimens were represented as solid steel in the model. The perforated wall of the specimen holder was modeled as a steel/water mixture.

This model is expected to provide a reasonable upper bound estimate of the effect of the capsule on local fluxes. The results obtained with this model were also used to provide independent confirmation of the reaction rate cross-sections used in the dosimetry analysis described in Section 4.1.

21

GE Nuclear Energy GE-NE-B 1100732-01 Revision I The flux obtained at the capsule midpoint radius With the modified (R,O) model was 1.53x109 r/'cm.2/s. Application of the axial adjustment of 0.996 results in an estimated flux of 1.52x!0 9 n/cm 2/s at the capsule center point. Consequently, the flux calculated at this point with the simulated capsule materials is about 10 % higher than the flux calculated with the base model. The region-averaged flux obtained in the specimen region, 1.5 lx 9 n/cm2/s, differs only slightly from the center point value. These results indicate that the base model under-predicts the flux within the capsule by a few percent and possibly as much as 10 %.

Therefore, a conservative bias exists in the calculated lead factor and projected peak fluences, since underestimation of the lead factor results in overestimation of the vessel peak fluence.

The 29-group neutron energy spectrum obtained at the simulated capsule center point was plotted and applied to ENDF/B-VI library data for the dosimeter activation reaction cross-sections to calculate spectrum-weighted group cross-sections for the reactions. The DORT case was re-run to obtain calculated total reaction rates which, when divided by the E > I MeV flux, yield the effective reaction rate cross-sections for the fast flux.

The cross-sections used in Section 4.1 to analyze the dosimeter data are derived from fits to empirical data which have been used by GE for analysis of surveillance capsule dosimetry for many years. Region-averaged values obtained for the specimen region in the DORT model are compared with the Section 4.1 cross-sections in the table below.

Comparison of Calculated Activation Cross-Sections in Simulated Capsule Region With Semi-Empirical Cross-Sections Used in Capsule Dosimetry Analysis Effective Cross-Section for E > I Mev Flux (barns)

Difference Reaction Empirical Fit Calculated Fe54(n,p)Mn54 0.182 0.1899

+4.34 Ni58(n,p)Co58 0.234 0.2425

+3.63 Cu63(na)Co60 0.00318 0.003305

+3.93 The close agreement between the calculated cross-sections and the fit-derived cross-sections provides confidence that the empirically derived cross-sections are reliable. It also provides confidence that the calculated neutron spectrum is realistic, even though the magnitude of the calculated flux is somewhat greater than the measured flux. In each instance, the calculated

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 cross-sections are slightly higher than the empirical cross-sections.

Consequently, if the dosimeter material reaction rates are predicted purely from the analysis, the difference between calculated and measured reaction rates will be slightly greater than the difference between the calculated and measured fluxes. The reaction rates are compared below.

Comparison of Calculated Reaction Rates in Simulated Capsule Region With Reaction Rates Determined From Capsule Dosimetry Analysis Dosimeter Reaction Rate (reactionsis/nucleus)

Difference Reaction Capsule Dosimeters Calculated Fe54(n,p)Mn54 2.14E-16 2.86E-16

+33.8 Ni58(n,p)Co58 2.70E-16 3.66E-16

+35.4 Cu63(n,a)Co6O 3.91 E-18 4.98E-18

+27.4 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 the Reg. Guide 1.99, Rev. 2 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 at a particular EFPY and lead factor according to:

fsurf = (fcap

• 32 EFPY)/(LF

• CEFPY)

(4-2) where, fsurf = 32 EFPY fluence at the peak vessel inside surface fcap = capsule fluence measured at the CEFPY 32 EFPY = end of life EFPY based on a 40-year operation at an 80% capacity factor CEFPY = the current EFPY for the capsule LF

= lead factor 23

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The surveillance capsule was removed from FitzPatrick at 13.4 EFPY as calculated in 17 Table 4-2. The fluence at 13.4 EFPY was determined to be 5,0 x 10 n/cm using Equation 4-1, and the lead factor was determined to be 0.68 as discussed in Section 4.2. In addition, the fluence over the remaining 18.6 EFPY was increased by 5% to account for the 5% power uprate that began in December 1996. Using this information with Equation 4-2, the resulting 32 EFPY fluence value at the peak vessel inside surface is:

17 l1*8634)l0/06 1.1 18 2

fsurf

[(5.0x 10 )+(5.0x 10"18.6/13.4)*1.05]/0.68 = 1.81x 10 n/cm (4-3) at the peak location.

The peak surface fluence at 32 EFPY is 22% lower than the nominal value (2.32 x 10 1 n/cm )

that was calculated from the first surveillance capsule dosimetry as a result of power uprate as reported in GE report [15].

This variation can be attributed to refinements in the analysis technique since the first capsule was removed.

The 1/4 T fluence (f) is calculated according to the Reg. Guide 1.99 [7] equation:

-0.24x f = fsurf (e0.4),

(4-4) where x = distance, in inches. to the 1/4 T depth. The vessel beltline lower intermediate shell ring thickness is 5.375 inches minimum requirement. The corresponding depth, x, taken from the minimum required thickness is 1.34 inches for the lower intermediate shell. Equation 4-4 18 evaluated for this value of x gives the 1/4 T value of 32 EFPY fluence, f = 1.31 xl 0 n/cm2 for the lower intermediate shell ring.

In the case of the lower shell ring, the axial fluence distribution was also taken into account. The maximum fluence at the top of the lower shell is 0.89 times the peak fluence, or 18 2

1.61 x 10 n/cm.

The minimum plate thickness of the lower shell is 6.375 inches, which corresponds to an x value of 1.6 inches. The resultant 1/4T fluence at 32 EFPY is 1.10 x 1017 n/cm2.

24

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 4-1:

SUMMARY

OF FITZPATRICK IRRADIATION PERIODS On Off Duration Days to eoi MWd Effective Full Full Power (days)

Power Days Fraction 1126/75 12131/77 1071 6874 1301203 5344 0.499 I/1/78 12/31/78 365 6509 539687 221.6 0.607 1/1/79 12/31/79 365 6144 373919 153.7 0.421

/11/80 12/31/80 366 5778 541475 222.2 0.607 1/1/81 12/31/81 365 5413 592405 243.1 0.666 1/1/82 12/31/82 365 5048 630106 258.8 0.709 1/1/83 12/31/83 365 4683 592197 243.1 0.666 1/1/84 12/31/84 366 4317 633307 259.9 0.710 1/1/85 12/31/85 365 3952 532365 218.6 0.599 1/1/86 12/31/86 365 3587 767477 315.0 0.863 1/1/87 12/31/87 365 3222 545590 224.1 0.614 1/1/88 12/31/88 366 2856 557082 228.8 0.625 V1/89 12/31/89 365 2491 781820 320.8 0.879 1/1/90 12/31/90 365 2126 592684 243.5 0.667 1/1/91 1/31/91 31 2095 69083 28.4 0.915 211/91 2/28/91 28 2067 56800 23.3 0.833 3/1/91 3/9/91 9

2058 19191 7.9 0.875 3/17/91 3/18/91 2

2049 116 0.1 0.024 4/13/91 4/30/91 18 2006 34493 14.2 0.787 5/1/91 5/7/91 7

1999 16095 6.6 0.944 8/18/91 8/31/91 14 1883 26087 10.7 0.765 9/1/91 9/30/91 30 1853 72905 29.9 0.998 10/1/91 10/31/91 31 1822 74840 30.7 0.991 11/1/91 11/28/91 28 1794 63288 26.0 0-928 11/29/91 1/2/93 401 1393 0

0.0 0.000 1/3/93 1/31/93 29 1364 14983 6.2 0.212 2/1/93 2/28/93 28 1336 58272 23.9 0.854 3/1/93 3/31/93 31 1305 17725 7.3 0-235 4/1/93 4/30/93 30 1275 51219 21.0 0.701 5/1/93 5/31/93 31 1244 46629 19.1 0.617 6/1/93 6/30/93 30 1214 72730 29.8 0.995 7/1/93 7/31/93 31 1183 72348 29.7 0.958 8/1/93 8/31/93 31 1152 75443 31.0 0.999 9/1/93 9/30/93 30 1122 62975 25.9 0.862 10/1/93 10/31/93 31 1091 55927 23.0 0.741 11/1/93 11/30/93 30 1061 13756 5.6 0.188 12/1/93 12/31/93 31 1030 74988 30.8 0.993 1/1/94 1/31/94 31 999 75300 30.9 0.997 2/1/94 2/28/94 28 971 68114 28.0 0.999 3/1/94 3/31/94 31 940 73706 30.3 0.976 4/1/94 4/30/94 30 910 4546 1.9 0.062 5/1/94 5/31/94 31 879 63588 26.1 0.842 611/94 6/30/94 30 849 71339 29.3.

0.976 7/1/94 7/31/94 31 818 68452 28.1 0.906 25

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 8/1/94 8/31/94 31 1

787 1 61533 25.3 0,815 9/1 1/94 9130/94 30 757 54488 22.4 0.746 10/1/94 10/31/94 31 726 54520 22.4 0.722 11/1/94 11/30/94 30 696 47247 19.4 0.647 1211/94 12/31/94 31 665 0

0.0 1

0.000 1/1/95 1/31/95 31 634 0

0.0 0.000 2/1/95 228/95 28 606 0

0.0 F

0.000 3!1/95 313 1/95 31 575 5960 25 0.079 4/1/95 4/30/95 30 545 69366 28.5 0.949 5/1/95 5/31/95 31 514 72287 29.7 0.957 6/1/95 6/30/95 30 484 49822 20.5 0.682 7/1/95 7/31/95 31 453 75412 31.0 0.999 8/1/95 8/31/95 31 422 75410 31.0 1

0.999 911/95 9/30/95 30 392 53600 22.0 F

0.733 10/1/95 10/31/95 31 361 75437 31.0 0.999 11/1/95 11/30/95 30 331 73014 30.0 F

0.999 12/1/95 12/31/95 31 300 73993 30.4 A

0.980 1/1/96 1/31/96 31 269 75173 30.9 0.995 2'1/96 2/29/96 29 240 51562 21.2 1

0.730 3/1/96 3/31/96 31 209 56448 23.2 0.747 4/1/96 4/30/96 30 179 72990 30.0 0.999 5/1/96 5/31/96 31 148 73629 30.2 i

0.975 6/1/96 6/30/96 30 118 71757 29.5 i

0.982 7/1/96 7/31/96 31 87 75250 30.9 0.996 8/1/96 8/31/96 31 56 73687 30.3 F

0.976 9/1/96 9/30/96 30 26 49799 20.4 1

0.681 10/1/96 10/26/96 26 0

[

56785.

23.3 i

0.897 Note: Full power was taken as the value prior to uprate of 2436 MW, Total Effective Full Power Days= 4907.8 Total Effective Full Power Years = 13A4 26

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I GE Nuclear Energy GE-NE-B1100732-01, Revision I TABLE 4-2: SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM START-UP TO 11/12/96 (13.4 EFPY) USING EMPIRICAL CROSS SECTIONS (GE CORRELATION)

.~Average' Avcraoe.~

Full:Power Flupx Full Nýwer FIuxC Fluence Flu~enccc Wire dps/g EBJentent Reaction Rate (nc

@/ns)(a/ctuZ)

(olealzi (Elemoent)

'(atend vOfrrooiaton)

[dpsinucleUS.(satvrated)l P7.1 MCW Eý0#je K.. E> I W'7V

E>0.1meV Iron 1.07E05 2.14E-16 1.18E09 1.89F,09 5.00E17 7,99E17 Nickel 1.67E06 2.70E-16

1. 16E09 1.86E09 4.90E17 7.85E17 Copper 1.56E04 3.91E-18 1.23E09 1.97E09 5.21E17 8.34E17 a

b C

Obtained by R.D Reager [201 Full power flux, based on thermal power of 2436 MWt 1.6 times the E >1 MeV result 27

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I

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I II GE Nuclear Energy GE-NE-B1100732-01, Revision I TABLE 4-2: SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM START-UP TO 11/12/96 (13.4 EFPY) USING EMPIRICAL CROSS SECTIONS (GE CORRELATION)

Average"

Average, Full Power-Fluxb Fd.

FIPower Pi~xc h unce.

OdF~encec Wire

.p/

lmn ReactioidRae C2-)(1/ms

1Werv.).

(Nn (E.Iempeit)

.(at' epid 4f Irradiatipp)

Eds4ces(aurtd)....1MceW E,*% MOV EA:KO M

.E>0.1

_eV Iron 1.07E05 2.14E-16 1.18E09 1.89E09 5.00E17 7T99E17 Nickel 1.67E06 2.70E-16 1.16E09 1.86E09 4.90E17 7,85E17 Copper 1.56E04 3.91E-18 1.23E09 1.97E09 5.21E17 8.34E17 a

b c

Obtained by R.D Reager [201 Full power flux, based on thermal power of 2436 MWt 1.6 times the E >1 MeV result 27

GE Nuclear Energy GE-NE-B1100732-01, Revision I TABLE 4-3. MEASURED FLUX VS. THEORETICAL FLUX FOR DOSIMETER AND FLUX WIRES E> I MeV Lead Factor EFPY*

Measured Capsule Fluence t EOL (32 EFPY) FLUENCE Capsule to ID Capsule Flux (n/cm2)

(n/cm 2)

Surface (n/cmr -s)

ID Surface I/4T Location 1982 30" A-Amuth Dosimeter I__......

i Sx I09 1985 3r*" Azimuth Flux Wires g079

' 6.0 1.4x109 2.6x10'7 1.8xlO1" 1.35X10l8 Upper Bound (1.25 Factor) 2.2x 1018 I7xl 0 Reg, Guide 1.99 Rev,2 Evaluation, no 0.61 2,32xl0'"

1,7x10'R upper bound factor included.

Tech Spec P-T curve basis.

_7_

5% Power Uprate based on upper 2.44xi0 176x10's bound value.

1996 1200 Azimuth Flux Wires 0.68 13.4 1.2x109 5.0x 017 1.81xx10 1.38x10"8 Includes 5% Power Uprate New P-T curve basis.

I Effective Full Power Years at 2436 Mw, 27b

GE Nuclear Energy GE-NE-B 1100732-01 Revision I REFLECTIVE BOUNDARY CORE

\\

INTERIOR CORE EXTERIOR 66 INTERVALS TOTAL SHROUD: 9 INTERVALS WATER REGION:

59 INTERVALS VESSEL WALL:

21 INTERVALS 98 INTERVALS IN AZIMUTHAL DIRECTION 1 = CORE INTERIOR FUEL 2

CORE EXTERIOR FUEL 01 FIGURE 4-1: SCHEMATIC OF MODEL FOR AZIMUTHAL FLUX DISTRIBUTION ANALYSIS 28

GE Nuclear Energy GE-NE-B 1 100732-01, Revision I 1.00

.o I I i........

0.9 0 i 0.80 w 0.70 S0.60 z,0.50 I--

LU

.I 4 0.30 0.20 I

0.10 0.00 I

0 5

10 15 20 25 30 35 40 45 ANGULAR POSITION (DEGREES)

FIGURE 4-2: RELATIVE FLUX VS. ANGLE AT RPV INSIDE SURFACE 29

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I GE Nuclear Energy GE-NE-B1100732-01, Revision 1 Ax z

0 wzw It 07 0 0 6 0 0.50o 0.70 0.60 0.50 0.40 0.3 0 0.10 0.00 15 25 35 45 55 65 75 85 95 105 DISTANCE FROM BAF (INCHES)

FIGURE 4-3: RELATIVE FLUX VS. ELEVATION AT RPV INSIDE SURFACE 115 125 13!

30

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

5. CHARPY V-NOTCH IMPACT TESTING 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. Testing was conducted in accordance with ASTM E23-94b [12].

5.1 IMPACT TEST PROCEDURE The Vallecitos testing machine used for irradiated specimens was a Tinius Olsen impact machine, serial number 175363. The maximum energy capacity of the machine is 300 ft-lb, which produces a test velocity at impact of 19.3 ftrsec.

The Tinius Olsen machine was qualified using NIST standard reference material specimens. The Standard Reference Materials (SRMs) consist of three sets of specimens which cover the energy range of the apparatus. Each set has a designated failure energy and a standard test temperature. According to ASTM E23-94b [12], the test apparatus averaged results must reproduce the NIST standard values within an accuracy of +5% or +1.0 ft-lb, whichever is greater. The results of the qualification of the Tinius Olsen impact machine are summarized in Table 5-1.

Charpy V-Notch tests were conducted at temperatures between -80'F and 400'F. The cooling fluid used for irradiated specimens tested at temperatures at or below 501F was ethanol.

At temperatures between 50'F and 210 0 F, water was used as the temperature conditioning fluid.

The specimens were heated in silicon oil for test temperatures above 210'F. Cooling of the conditioning fluids was done by heat exchange with liquid nitrogen through a copper coil; heating was done by an immersion heater.

The bath of fluid was mechanically stirred to maintain uniform temperatures. The fluid temperature was measured with a calibrated Type K thermocouple positioned near the impact samples. After equilibration at the test 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.

For each Charpy V-Notch specimen the test temperature, energy absorbed, lateral expansion, and percent shear were determined. In addition, photographs were taken for the 31

GE Nuclear Energy GE-NE-B 1100732-01 Revision I irradiated specimens. Lateral expansion and percent shear were measured according to specified methods [12]. Percent shear was determined using method number I of Subsection 11.2.4.3 of ASTM E23-94b [121, which involved measuring the length and width of the cleavage surface in inches and determining the percent shear value from Table 2 of ASTM E23-94b [123.

5.2 IMPACT TEST RESULTS Eight Charpy V-Notch specimens each of irradiated base, weld, and HAZ material were tested at temperatures (-80°F to 400'F) selected to define the toughness transition and upper shelf portions of the fracture toughness curves. The absorbed energy, lateral expansion, and percent shear data are listed for each material in Table 5-2. Plots of absorbed energy and lateral expansion for base, weld, and HAZ materials are presented in Figures 5-1 through 5-6. These curves are plotted along with the corresponding curves from the first capsule (and unirradiated base material data where appropriate) in Figures 5-7 through Figure 5-12. The fracture surface photographs and a summary of the test results for each specimen are contained in Appendix A.

The unirradiated and irradiated plate and weld energy and lateral expansion data are fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook [13] (HLAZ was not fit due to data scatter):

Y = A + B

• TANH [( T - T0 )/C],

where Y

impact energy or lateral expansion T = test temperature, and A, B, To and C are determined by non-linear regression.

The TANH function is one of the few continuous functions with a shape characteristic of low alloy steel fracture toughness transition curves.

5.3 IRRADIA TED VERSUS UNIRRADIA TED CHARPY V-NOTCH PROPERTIES Ideally, a shift in RTNDT would be established by comparing the irradiated Charpy specimen data to baseline unirradiated Charpy data. For the case of the FitzPatrick base material specimens, data was obtained from the Certified Material Test Report. Additional Charpy test 32

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 data for the FitzPatrick surveillance plate (heat number C3278-2) was available from the BWROG Supplemental Surveillance Program report [17]. This program was useful in providing plant-specific data and information for the FitzPatrick base material to establish baseline properties. The unirradiated data for the base material, as well as the results for both the plate and weld materials from the first and second surveillance capsules, were fit to a TANH function as described in the previous section. The unirradiated properties for the surveillance plate were determined from the combined sets of data, as shown in Figure 5-13. For the weld material, no credible unirradiated baseline data was available.

5.4 COMPARISON TO PREDICTED IRRADIATION EFFECTS 5.4.1 Irradiation Shift The measured transition temperature shifts for the base and weld materials were compared to the predictions calculated according to Rev. 2 [7]. The inputs and calculated values for irradiated shift for the plate and weld materials based upon measurements taken from the 120' azimuth capsule at 13.4 EFPY are as follows:

Plate:

Copper 0.11%

Nickel =

0.60%

CF=

74 17 2

fluence=

5.0 x 10 n/cm Reg. Guide 1.99 ARTNDT =

Reg. Guide 1.99 ARTNDT +/- 2A(34'F) =

Measured 30 ft-lb shift =

21.7 0F 55.7'F max, -I2.3°F min 14.97°F Weld:

Copper =

Nickel CF =

fluence =

Reg. Guide Reg. Guide 0.29%

0.71%

208 17 2

5.0 x 10 n/cm 1.99 ARTNDT =

1.99 ARTNTDT +/- 2aYA(56 0F) =

60.90F 1 16.9°F max, 4.9°F min 33

GE Nuclear Energy Technical Services Business General Electric Company 175 Curtner Avenue, San Jose, CA 95125 GE-NE-B 1100732-01, Revision I Final Report Class II February 1998 PLANT FITZPATRICK RPV SURVEILLANCE MATERIALS TESTING AND ANALYSIS OF 1200 CAPSULE AT 13.4 EFPY Prepared by:

T. J. GriesbacK,64irector ATI Consulting

R. G. Carey, Engineer Structural Mechanics and Materials Verified by:

Approved by:

B. J. Branlund, Project Manager Structural Mechanics and Materials

Page 1 of 1 Pagano, Terry From:

Porter, Anne Sent:

Thursday, November 10, 2005 10:33 AM To:

Pagano, Terry

Subject:

FW: Fitz Doc Terry - can you get this please From: LOYD, LELAND Sent: Thursday, November 10, 2005 10:21 AM To: Porter, Anne Cc: Harrison, Douglas; Herrmann, Terry; BARTON, SANDRA

Subject:

Fitz Doc Good Morning Anne I need GE-NE-B1 100732-01 (JAF RPV surveillance materials testing & analysis of 120 degree capsule at 13.4 EFPY (effective, full power years) - including BJB-9907 and attachment to BJB-9907).

If you send it hard copy send to Sandy ENTERGY CORPORATION 1448 SR 333 Russellville, AR 72802 c/o Sandra Barton N-GSB-45 Or; sbart90@entergy.com Sandy (this goes with the OE programs, Lori will know what to do with it)

Thanks!

LefandLoyd License Renewal floyd9o@.ntergy.com 479/858/4696 11/10/2005

GE Nuclear Energy.

Technical Services Business General Electric Company 175 Curtner A venue. San Jose. CA 9512.5 GE-NE-B 1100732-0 1, Revision I Final Report Class II February 1998 PLANT FITZPATRICK RPV SURVEILLANCE MATERIALS TESTING AND ANALYSIS OF 1200 CAPSULE AT 13.4 EFPY Prepared by:

ATI Consulting R. G. Carey, Engineer Verified by:

Structural Mechanics and Mý B. J. Branlund, Project Mane Structural Mechanics and M; Approved by:

terialEW YORK POWER AUTHORITY DOCUMENT REVIEW STATUS S~ThMhNO:

_,,, yfACCEPTED P

[0 ACCEPTED AS NOTED ger RESUBMITrAL NOT REQUIRED 3

C ACCEPTED AS NOTED aerials RESUBMITTAL REQUIRED 4

(1]

NOT ACCEPTED Pimissw to poceed does not occutft acceptanoce or approval of des:q try the Wpo an does no re*v

.upir from full cornpiance wz" corm-inew 1

A VIEWEDI

.f

123 Main ttret White Plains, New York 10601 914 681.6200 SNe York Power

~ Authorit February 13, 1998 DCME-98-0086 To: GE NUCLEAR ENERGY 175 Curtner Ave San Jose CA 95125 Contract No. C95-Z0013 Attn:

Ms.

B. J.

Branlund, Project Manager The document(s) listed below are being returned to you with the status indicated on each document.

Document No.

Rev.

Status GE-NE-BI100732-01 1

1 ACCEPTED JAF RPV SURVEILLANCE MATERIALS TESTING & ANALYSIS OF 120 DEGREE CAPSULE AT 13.4 EFPY (EFFECTIVE FULL POWER YEARS Very truly yours, W. Spataro Consulting Metallurgist copits (transmittal only) : G. Grochowski

w. attachment: G. Rorke, J.

Ellmers, C. Doc

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of the New York Power Authority. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained, or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting information in this document are contained in the contract 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 information.

contained in this document.

Ii

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE OF CONTENTS ABSTRACT viii ACKNOWLEDGMENTS ix

1. INTRODUCTION 1

2.

SUMMARY

AND CONCLUSIONS 2

2.1

SUMMARY

OF RESULTS 2

2.2 CONCLUSION

S 5

3. SURVEILLANCE PROGRAM BACKGROUND 6

3.1 CAPSULE RECOVERY 6

3.2 RPV MATERIALS AND FABRICATION 6

3.2.1 Fabrication History 6

3.2.2 Material Properties of RPV at Fabrication 7

3.2.3 Surveillance Capsule Specimen Chemical Composition 7

3.3 SPECIMEN DESCRIPTION 7

3.3.1 Charpy. Specimens 8

3.3.2 Tensile Specimens 8

4. PEAK RPV FLUENCE EVALUATION 17 4.1 FLUX WIRE ANALYSIS 17 4.1. 1 Procedure 17 4.1.2 Results 18 4.2 -DETERMINATION OF LEAD FACTOR 18 4.2.1 Procedure-19 4.2.2 Results 20 4.3.ESTIMATE OF 32 EFPY FLUENCE 23

5. CHARPY V-NOTCH IMPACT TESTING 31 5.1 IMPACT TEST PROCEDURE 31 5.2 IMPACT TEST RESULTS 32 5.3 IRRADIATED VERSUS UNIRRADIATED CHARPY V-NOTCH PROPERTIES 32 5.4 COMPARISON TO PREDICTED IRRADIATION EFFECTS 33 5.4.1 Irradiation Shift 33 5.4.2 Change in USE 34 iii

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

6. TENSILE TESTING 52 6.1 PROCEDURE 52 6.2 RESULTS 53 6.3 IRRADIATED VERSUS UNIRRADIATED TENSILE PROPERTIES 53

7. ADJUSTED REFERENCE TEMPERATURE AND UPPER SHELF ENERGY 63 7.1 ADJUSTED REFERENCE TEMPERATURE AT 32 EFPY 63 7-2 SURVEILLANCE CF ADJUSTMENT 64 7.3 APPLICATION OF CF ADJUSTMENT TO BELTLINE MATERIALS 65 7.4 ART VS. EFPY 66 7.5 UPPER SHELF ENERGY AT 32 EFPY 66

8. PRESSURE-TEMPERATURE CURVES 70

8.1 BACKGROUND

70 8.2 P-T CURVE METHODOLOGY 72 8.2.1 Non-Beltline Regions 72 8.2.2 Pressure Test - Non-Beltline, Curve A (Using Bottom Head) 73 8.2.3 Core Not Critical Heatup/Cooldown - Non Beltline, Curve B (Using Feedwater Nozzle/Upper Vessel Region) 75 8.2.4 Example Core Not Critical Heatup/Cooldown Calculation for Feedwater Nozzle/Upper Vessel Region 76 8.2.5 Core Beitline Region 78 8.2.6. Beltline Region - Pressure Test 79 8.2.7 Calculations for the Beltline Region - Pressure Test 80 8.2.8 Beltline Region - Core Not, Critical Heatup/Cooldown 81 8.2.9 Calculations for the Beltline Region Core Not Critical Heatup/Cooldown 82 8.3 CLOSURE FLANGE REGION 83 8.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G 84

9. REFERENCES 92 APPENDICES A. IRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS A-1 B. P-T CURVES VALID TO 24 EFPY B-1 iv

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE OF TABLES TABLE 3-1: CHEMICAL COMPOSITION OF RPV BELTLINE MATERIALS 9

TABLE 3-2: RTNDT OF VESSEL MATERIALS 10 TABLE 3-3: RTNDT OF NOZZLE, WELD AND STUD MATERIALS I 1 TABLE 3-4: CHEMICAL COMPOSITION OF FITZPATRICK SURVEILLANCE MATERIALS FROM SURVEILLANCE SPECIMEN CHEMICAL TESTS 13 TABLE 4-1:

SUMMARY

OF FITZPATRICK IRRADIATION PERIODS 25 TABLE 4-2: SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM START-UP TO 1/12/96 (13.4 EFPY) USING EMPIRICAL CROSS SECTIONS (GE CORRELATION) 27 TABLE 4-3: MEASURED FLUX VS. THEORETICAL FLUX FOR DOSIMETER AND FLUX WIRES 27b TABLE 5-1: VALLECITOS QUALIFICATION TEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS 35 TABLE 5-2: IRRADIATED CHARPY V-NOTCH IMPACT TEST RESULTS SECOND CAPSULE 36 TABLE 5-73: SIGNIFICANT RESULTS OF IRRADIATED AND UNIRRADIATED CHARPY V-NOTCH DATA 37 TABLE 6-1: TENSILE TEST RESULTS FOR IRRADIATED RPV MATERIALS 54 TABLE 6-2: COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES AT ROOM TEMPERATURE 54 TABLE 6-3: COMPARISON OF IRRADIATED TENSILE PROPERTIES AT 185'F 55 TABLE 6-4: COMPARISON OF IRRADIATED TENSILE PROPERTIES AT 500SF 55 TABLE 7-1: 32 EFPY ART VALUES 67 TABLE 7-2: PLATE EQUIVALENT MARGIN ANALYSIS 68 TABLE 7-3: WELD EQUIVALENT MARGIN ANALYSIS 69 TABLE 8-1: FITZPATRICK P-T CURVE VALUES FOR 32 EFPY 88 TABLE B-i: FITZPATRICK P-T CURVE VALUES FOR 24 EFPY B-5 V

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE OF FIGURES FIGURE 3-1: SURVEILLANCE CAPSULE HOLDER RECOVERED FROM FITZPATRICK (120' AZIMUTHAL LOCATION CAPSULE - REMOVED AT 13.A EFPY) 14 FIGURE 3-1(A): CHARPY SPECIMEN CAPSULE IDENTIFICATION (1200 AZIMUTHAL LOCATION CAPSULE - REMOVED AT 13.4 EFPY)

Is FIGURE 3-2. SCHEMATIC OF RPV SHOWING IDENTIFICATION OF VESSEL BELTLINE PLATES AND WELDS 16 FIGURE 4-1: SCHEMATIC OF VESSEL GEOMETRY 2&

FIGURE 4-2: RELATIVE FLUX VS. ANGLE AT RPV INSIDE SURFACE 29 FIGURE 4-3: RELATIVE FLUX VS. ELEVATION AT RPV INSIDE SURFACE 30 FIGURE 5-1: ABSORBED ENERGY VS. TEMPERATURE (BASE) 38 FIGURE 5-2: LATERAL EXPANSION VS. TEMPERATURE (BASE) 39 FIGURE 5-3: ABSORBED ENERGY VS. TEMPERATURE (WELD) 40 FIGURE 5-4: LATERAL EXPANSION VS. TEMPERATURE (WELD) 41 FIGURE 5-5: ABSORBED ENERGY VS. TEMPERATURE (HAZ) 42 FIGURE 5-6: LATERAL EXPANSION VS. TEMPERATURE (HAZ) 43 FIGURE 5-7: COMPARISON OF UNIRRADIATED AND IRRADIATED ENERGY DATA (PLATE) 44 FIGURE 5-8: COMPARISON OF 1ST AND 2ND CAPSULE ENERGY RESULTS (WELD) 45 FIGURE 5-9: COMPARISON OF 1 ST AND 2ND CAPSULE ENERGY RESULTS (HAZ) 46 FIGURE 5-10: COMPARISON OF LATERAL EXPANSION RESULTS (BASE) 47 FIGURE 5-11: COMPARISON OF LATERAL EXPANSION RESULTS (WELD) 48 FIGURE 5-12: COMPARISON OF LATERAL EXPANSION RESULTS (HAZ) 49 FIGURE 5-13: TANH CURVE-FITTED RESULTS FOR COMBINED BASELINE DATA (PLATE) 51 FIGURE 5-14: AT 30 VS. FLUENCE SHOWING PLATE DATA WITH FITTED RESULTS 52 FIGURE 6-1. TYPICAL ENGINEERING STRESS-STRAIN FOR IRRADIATED RPV MATERIALS 56 FIGURE 6-2: FRACTURE LOCATION AND NECKING BEHAVIOR FOR IRRADIATED BASE METAL TENSILE SPECIMENS 57 FIGURE 6-3: FRACTURE LOCATION AND NECKING BEHAVIOR FOR IRRADIATED WELD METAL TENSILE SPECIMENS 58 FIGURE 6-4: FRACTURE LOCATION AND NECKING BEHAVIOR FOR IRRADIATED HAZ TENSILE SPECIMENS 59 FIGURE 6-5: FRACTURE APPEARANCE FOR IRRADIATED BASE METAL TENSILE SPECIMENS 60 FIGURE 6-6: FRACTURE APPEARANCE FOR IRRADIATED WELD METAL TENSILE SPECIMENS 61 FIGURE 6-7: FRACTURE APPEARANCE FOR IRRADIATED HAZ TENSILE SPECIMENS 62 FIGURE 7-1: ART VS. EFPY FOR LIMITING BELTLINE PLATE AND WELD 69b FIGURE 8-1: PRESSURE TEST CURVE (CURVE A) VALID TO 32 EFPY 85 FIGURE 8-2: NON-NUCLEAR HEATUP/COOLDOWN (CURVE B) VALID TO 32 EFPY 86 FIGURE 8-3: CORE CRITICAL OPERATION (CURVE C) VALID TO 32 EFPY 87 vi

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE OF FIGURES (continued)

FIGURE B-I: PRESSURE TEST CURVE (CURVE A) VALID TO 24 EFPY B-2 FIGURE B-2: NON-NUCLEAR HEATUP/COOLDOWN (CURVE B) VALID TO 24 EFPY B-3 FIGURE B-3: CORE CRITICAL OPERATION (CURVE C) VALID TO 24 EFPY B-4

GE Nuclear Energy GE-NE-B 1100732-01 Revision I ABSTRACT The surveillance capsule at the 1200 azimuthal location was removed at 13.4 EFPY from the FitzPatrick reactor in November 1996. The capsule contained flux wires for neutron fluence measurement, and Charpy test specimens and tensile test specimens for material property evaluations.

The flux wires were evaluated to determine the fluence experienced by the test specimens. Charpy V-Notch impact testing and tensile testing were performed to establish the properties of the irradiated surveillance materials.

The irradiated Charpy data for the base material specimens were compared to available unirradiated data to determine the shift in Charpy curves due to irradiation. The results indicate a shift lower than the predictions of Regulatory Guide 1.99 Revision 2 [Rev. 21. Since two sets of credible data sets were available for the plate material, the Adjusted Reference Temperature (ART) calculations for vessel base materials were adjusted in accordance with Rev. 2. For the vessel weld metal; no unirradiated data was available and the predictions of Rev. 2 were used to calculate ART.

The flux wire results combined with the lead factor were used to estimate the 32 EFPY fluence. The fluence calculations included the effects of a 105% power uprate. The resulting estimated fluence showed. a reduction of 22 percent compared with the previous nominal 32-EFPY fluence estimate consistent with the fluence used for the Technical Specification Pressure-Temperature (P-T) Curves.

P-T Curves were prepared based on the new projected fluence levels for both 32 EFPY and 24 EFPY.

viii

GE Nuclear Energy GE-NE-B 1100732-01 Revision I ACKNOWLEDGMENTS The author gratefully acknowledges the efforts of other people towards completion of the contents of this report.

Cask shipping & receipt, and capsule disassembly were performed by I. B. Myers and R. D. Rimmer. Charpy testing was completed by G. E. Dunning and B. D. Frew. Tensile testing was performed by S. B. Wisner. Chemical composition analysis was performed by P. S. Wall.

Flux wire testing and analysis was performed by L. Kessler, R_ M. Kruger and R_ D. Reager.

Fluence and lead factor calculations were performed by D. R-Rogers, H. A-Careway and S. S. Wang. Assistance with the capsule evaluation was provided by B. N. Burgos of ATI Consulting. Project management was conducted by B. J. Branlund.

Ix

GE Nuclear Energy GE-NE-B 1100732-01 Revision I

1. INTRODUCTION Part of the effort to assure reactor vessel integrity involves evaluation of the fracture toughness of the vessel ferritic materials. The key values which characterize a material's fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the upper shelf energy (USE). These are defined in IOCFR50 Appendix G [1] and in Appendix G of the ASME Boiler and Pressure Vessel Code, Section Xl [2].

Appendix H of I0CFR50 [3] and ASTM E185-70 [4] establish the methods to be used for surveillance of the James A. FitzPatrick (FitzPatrick) reactor vessel materials. The second vessel surveillance specimen capsule required by IOCFR50 Appendix H [3] was removed from FitzPatrick in November 1996. 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 and tensile test specimens fabricated using base metal from the beltline region, as well as weld metal from a similar heat of material as the beltline welds. The impact and tensile specimens were tested to establish properties for the irradiated materials.

The results of the surveillance specimen testing are presented in this report, as required per IOCFR50 Appendices G and H [1 & 3]. The irradiated material properties are compared to available unirradiated properties to determine the effect of irradiation on material toughness for the base and weld materials, through Charpy testing.

Irradiated tensile testing results are provided and are compared with unirradiated data to determine the effect of irradiation on the stress-strain relationship of the materials.

Pressure-temperature (P-T) curves are included in this report which have been developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlemerit effects in the beltline. The P-T curves are established to the requirements of IOCFR50, Appendix G [1] to assure that brittle fracture of the reactor vessel is prevented.

I

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

2.

SUMMARY

AND CONCLUSIONS 2.1

SUMMARY

OF RESULTS The 1200 azimuth position 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 El185-82 [6]. The methods and results of the testing are presented in this report as follows:

Section 3:

Surveillance Program Background

" RPV Materials and Fabrication

" Material Properties

" Surveillance Specimen Chemical Composition Specimen Description Section 4:

Peak RPV Fluence Evaluation Section 5:

Charpy V-Notch Impact Testing Section 6:

Tensile Testing Section 7:

Adjusted Reference Temperature and Upper Shelf Energy Section 8:

Pressure-Temperature Curves The significant results of the evaluation are below:

a.

The 1200 azimuth position capsule was removed from the reactor after 13.4 EFPY (Effective Full Power Years) of operation. The capsule contained 2 sets of 3 flux wires: nickel (Ni), copper (Cu), and iron (Fe). There were 24 Charpy V-Notch specimens in the capsule: eight (8) each of plate (base) material, weld material, and heat affected zone (HAZ) material.

The capsule also contained eight (8) tensile specimens: three plate material, three weld material, and two HAZ material. (See Sections 3.1 and 3.3) 2

GE A¢uclear Energy GE-NE-B 1100732-01 Revision 1

b.

The chemical composition of copper (Cu) and nickel (Ni) for the irradiated surveillance materials was determined from a chemical composition analysis. The best estimate values for the surveillance material chemistries were calculated as averages of the available baseline and irradiated data.

The best estimate values for the surveillance plate are 0.11% Cu and 0.60% Ni, and are 0.29% Cu and 0.71% Ni for the surveillance weld. (See Table 3.4)

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 17 2

>1 MeV flux) received by the surveillance specimens was 5.0 x 10 n/cm at removal (13.4 EFPY-See Section 4.1.2).

d.

A neutron transport computation had been performed based on the first surveillance capsule.

Relative flux distributions in the azimuthal and axial directions were previously developed in Reference 8. The lead factor was 0.79, relating the surveillance capsule flux to the peak inside surface flux. The lead factor was calculated after the second capsule was removed at 13.4 EFPY, and determined to be 0.68. A lead factor of 0.68 was used for all calculations in this report (See Section 4.2.2).

e.

The surveillance Charpy V-Notch specimens were impact tested at temperatures selected to define the upper shelf energy (USE) and the transition of the Charpy V-Notch curves for the plate, weld, and HAZ materials. Measurements were taken of absorbed energy, lateral expansion and percentage shear. From absorbed energy and lateral expansion curve-fit results, the values of USE and of index temperature for 30 ft-lb, 50 ft-lb and 35 mils lateral expansion (MLE) were obtained (see Table 5-3). Fracture surface photographs of each specimen are presented in Appendix A.

f.

The irradiated tensile specimens were tested at room temperature (70'F), at reactor operating temperature (550'F) and at 185'F as an intermediate temperature. Unirradiated base material results, as well as results from the first capsule, were available for comparison (See Tables 6-I through 6-4.)

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GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

g.

The curves of irradiated and unirradiated Charpy specimens established the 30 ft-lb shifts. The plate material showed a 15'F shift and a 12 ft-lb decrease in USE (9% decrease).

These values were not calculated for the weld, as no unirradiated data was available (See Table 5-3).

17 2

h-The measured shift of 15'F for plate material for a fluence of 5.0 x 10 n/cm, was within the Rev. 2 [7] range predictions (ARTN*-T+/-2a) of -12°F to 56'F. Since two credible data sets are available for the plate material, the surveillance adjustment (Section 7) was applied to the vessel base plates. The measured shift values were not obtained for the weld as no unirradiated data was available. The best estimate chemical composition for the surveillance weld material was used for evaluating the projected shift of the surveillance weld data (See Table 5-3).

I 2

The 32 EFPY RPV peak fluence prediction is 1.81

• 10 n/cm at the vessel wall,

,based on the flux wire test and lead factor. This is 22% less than the previously established nominal 32 EFPY fluence prediction (2.32 x 10is n/cm2) [5]. The 18 2

32 EFPY fluence prediction is 1.31 x 10 n/cm at 1/4 T. (See Section 4.3)

j.

The adjusted reference temperature (ART = Initial RTNDT + 6RTNDT + Margin) was predicted for each beltline material, based on the methods of Regulatory Guide 1.99, Rev. 2.

The ART for the limiting material, Axial Weld Heat 27204/12008, at 32 EFPY is 109'F and is lower than the 2001F requirement of IOCFR50 Appendix G [1] and Rev. 2 [7]. (See Table 7-1)

k.

An update of the beltline material USE values at 32 EFPY was performed using the Reg. Guide 1.99, Rev. 2 methodology.

The equivalent margin analyses demonstrate that 1 OCFR50, Appendix G safety requirements are satisfactorily met for FitzPatrick. (See Tables 7-2 and 7-3)

P-T curves were developed for three reactor conditions: pressure test (Curve A),

non-nuclear heatup and cooldown core not critical operation (Curve B), and core critical operation (Curve C) curves which are valid for up to 32 EFPY of operation. The beltline curve is more limiting for Curve A at pressures above approximately 550 psig. For Curves B and C, the beltline curves are limiting for pressures above approximately 600 psig. The P-T curves for 32 EFPY are shown in Figures 8-1 through 8-3, and the P-T curves for 24 EFPY are shown in Appendix B, Figures B-1 through B-3 4

GE Nuclear Energy GE-NE-B 1100732-01 Revision I

2.2 CONCLUSION

S The requirements of IOCFR50 Appendix G [1] deal 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 30 ft-lb shift for the base material was less than the Rev. 2 prediction, and therefore the ART values for beltline plates were modified in accordance with Position 2 of Rev. 2. The changes in USE for the survillance plate are bounded by the Regulatory Guide 1.99 Revision 2 predictions and associated deviations.

b.

The values of ART and USE for the reactor vessel beltline materials are expected to remain within the limits of IOCFR50 Appendix G [13 for at least 32 EFPY of operation.

5

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

3. SURVEILLANCE PROGRAM BACKGROUND 3.1 CAPSULE RECOVERY The reactor pressure vessel (RPV) surveillance program consists of three surveillance capsules at 300, 120c, and 3000 azimuths at the core midplane. The specimen capsules are held against the RPV inside surface by a spring loaded specimen holder. Each capsule is expected to receive equal irradiation because of core symmetry. The first capsule (300 azimuth) was removed in April 1985 after 5.98 EFPY. During the November 1996 outage, the second surveillance capsule was removed from the 1200 azimuthal location The capsule was cut from its holder assembly and shipped by cask to the GE Vallecitos Nuclear Center (VNC), where testing was performed.

Upon arrival at VNC, the capsule was examined for identification. The identification number stamped on the capsule corresponded to FitzPatrick, as specified by GE drawings, 117C3739 (Outline Specimen Holder) and 921D465 (Surveillance -Program), for the FitzPatrick 120' surveillance materials.

The general condition of the capsule as received is shown in Figure 3-1. The specimen holder contained 2 sets of 3 flux wires (iron, copper, and nickel), three Charpy specimen capsules each containing 8 plate, weld, or HAZ Charpy specimens in a sealed helium environment, and four tensile specimen capsules (together containing 3 base, 3 weld and 2 HAZ tensile specimens in a sealed helium environment).

3.2 RPV MATERIALS AND FABRICATION 3.2.1 Fabrication History The FitzPatrick RPV is a 220.75 inch inside diameter BWR/4 design_ Construction was performed by Combustion Engineering (CE) under the 1965 edition of the ASME Code through the 1966 Winter Addenda.

The shell and head plate materials are ASME SA533, Grade B, Class I low alloy steel (LAS). The nozzles and closure flanges are ASME SA508 Class 2 LAS, and the closure flange bolting materials are ASME A540 Grade B24 LAS [81-Submerged arc or shielded metal arc welding of plates was followed by post-weld heat treatment at 11 50F. The fabrication impact test specimens were given a simulated post weld heat treatment at 6

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 1150°F +/- 25-F, held 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> followed by furnace cooling to below 600'F, then air cooled. The identification of plates and welds in the beltline region is shown in Figure 3-2.

3.2.2 Material Properties of RPV at Fabrication Material certification records were retrieved from GE Quality Assurance (QA) records to determine chemical and mechanical properties of the vessel materials. The retrieved information for the beltline materials is documented in [5]. Table 3-1 shows the chemistry data for the beltline materials. Properties of the beltline materials and materials at other locations of interest are presented in Tables 3-2 and 3-3.

3.2.3 Surveillance Capsule Specimen Chemical Composition Samples were taken from the irradiated base and weld Charpy specimens after they were tested.

Chemical analyses were performed using a Spectraspan III plasma emission spectrometer. Each sample was dissolved in an acid solution to a concentration of 40 mg steel per ml solution. The spectrometer was calibrated for determination of Mn, P, Ni, Mo, V, Cr, Si and Cu by diluting National Institute of Standards and Technology (NIST) Spectrometric Standard Solutions. The phosphorus calibration involved analysis of five reference materials from NIST with known phosphorus levels.

Analysis accuracies are +/-0.005% (absolute) of reported value for phosphorus and +/-5% (relative) of reported value for other elements. The chemical composition results are given in Table 3-4 for both irradiated and baseline surveillance plate and irradiated weld materials.

The baseline plate data was taken from CE material certification records as documented in [5] for the plate surveillance specimens; no baseline data was available for the weld material.

3.3 SPECIMEN DESCRIPTION The surveillance capsule holder contained 24 Charpy specimens: base metal (8), weld metal (8), and FAZ (8). The holder also contained 2 sets of 3 flux wires (iron, nickel, and copper) and eight (8) tensile specimens (three base, three weld and two HAZ). The chemistry and fabrication history for the Charpy and tensile specimens are described in this section.

7

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 3.3.1 Charpy Specimens The fabrication of the Charpy specimens is described in the CE drawings of the surveillance test program. All materials used for specimens were beltline materials taken from the lower intermediate shell course.

The base metal specimens were cut from plate G-3414-2, heat number C3278-2. The test plates received the same heat treatment as plate heat no. C3278-2, including the post-weld heat treatment for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1150'F +/- 25°F. The Charpy specimens were removed from plate heat no. C3278-2 and machined from the 1/4 T and 3/4 T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction). The Charpy specimens had been stamped on one end with the fabrication codes as listed in GE surveillance program drawings for FitzPatrick.

The weld metal and HAZ Charpy specimens were fabricated by welding together pieces of plates G-3414-1 and G-34i4-2 with a weld identical to longitudinal seam weld 1-233 in the RPV beltline.

Welding records obtained from CE indicate the surveillance weld to be a submerged arc weld representive of the vessel beltline circumferential weld. The welded test plates received stress relief heat treatment at 11501F +/-+25°F to simulate the RPV fabrication conditions. The weld and HAZ specimens were cut from the material avoiding the volume near the root of the welds.

The base metal orientation in the weld and HAZ specimens was longitudinal.

3.3.2 Tensile Specimens Fabrication of the surveillance tensile specimens is also described in the CE surveillance program drawings. The materials, chemical compositions, and heat treatments for the tensile specimens are the same as the corresponding Charpy specimens. The identifications of the base, weld and HAZ surveillance specimens are described in Reference 8.

8

GE Nuclear Energy GE-NE-B1 100732-01, Revision 1 TABLE 3-1: CHEMICAL COMPOSITION OF RPV BELTLINE MATERIALS" Composition by eight Percent entifictio ni fe ot No..

.N Oil Mn.P.

Si Mo PLATES:

Lower Shell:

G-3415-1R C3394-1 0.11f 0.56 0,21 1.32 0.015 0.017 0.26 0.47 G-3415-3 C3376-2 0.13b 0.60 0.22 1.33 0.015 0.017 0.22 0.48 G-3415-2 C3103-2 0. 14 b 0.57 0.23 1.36 0.012 0.015 0.26 0.46 Lower-Intermed. Shell:

G-3413-7 C3368-1 0. 12b 0.50 0.19 1.30 0,015 0.017 0.22 0.45 G-3414.2r C3278-2 0, 1e 0.600 0.20 1.26 0.011 0.016 0.22 0.48 G-3414-1 C3301-1 0.18b 0.57 0.18 1,36 0.008 0.015 0.29 0.46 WELDS:

Lower Longitudinal:

27204/12008 0.219 0.996 N/A 1.16 0,013 0.007 0.21 0.46 2-233 A,B,C Flux 1092 Lot 3774 Lower Int. Long.:

13253/12008 0.210' 0.873' N/A N/A N/A N/A N/A N/A 1-233 A,B,C Flux 1092 Lot 3947 Lower to 305414 0.3 3 7 d 0.609d 0.14 1.45 0.012 0.01 0.18 0.51 Lower -Int. Girth:

Flux 1092 Lot 3947 1-240 1

1 1

A b

C d

r Data trom CMITK Reports, Gt QA Kecords and p1j except as noted below Cu values taken from Lukens Steel letter to NYPA dated 10/14/85 [19]

Surveillance plate Best estimate Cu and Ni weld values obtained from CE Owners Group report [18)

Average chemistry of surveillance plate from Table 3-4 Cu content from Generic Letter 92-01 response [21]

9

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 3-2: RTNDT OF VESSEL MATERIALS COMPONENT ID HEAT TEST CHARPY (Tf0T-60)

DROP RTNDT TEMP.

ENERGY (OF)

WEIGHT (OF)

(OF)

(FT-LB)

NDT (OF)

PLATES & FORGINGS:

Top Head & Flange Dollar Plate Top Head Torus Top Head Flange Shell Courses Upper Shell Flange Upper Shell.

Upper Int. Shell Low-Int. Shell Lower Shell Bottom Head Dollar Plate Bottom Head Torus G-3412 G-341 1-1 G-3411-2 G-3402 G-3401 G-3413-4 G-3413-5 G-3413-6 G-3413-1 G-3413-2 G-3413-3 G-3413-7 G-3414-1 G-3414-2 G-3415-1 R G-3415-2 G-3415-3 G-3410 G-3407-1 G-3408-1 G-3409 C-2869-5 C-3055-1 0-3055-1 4P-1885 2V595 B-7255-1 C-3229-2 B-7291-1 C-3116-1 C-3121-2 C-3168-2 C-3368-1 C-3301-1 C-3278-2 C-3394-1 C-3103-2 C-3376-2 C-2917-3 C-2851 -1 C-3055-2 C-2906-3 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 40 10 10 10 10 83 98 98 66 117 70 50 81 65 31 95 61 60 45 53 41 43 38 83 53 36 70 73 73 87 94 76 68 69 91 48 87 55 63 77 71 48 51 36 72 73 43 72 118 118 96 117 71 82 65 79 35 76 45 49 58 52 49 49 36 75 66 35

-20

-20

-20

-50

-50

-20

-20

-20

-20 18

-20

-10

-18

-10

-20

-2 24 8

-20

-20 10

-10

-10

-10 30 10

-10

-10

-10

-10 10

-10

-50

-40

-30

-10

-10

-10

-10

-10

-10

-10

-10

-10

-10 30 10

-10

-10

-10

-10 18

-10

-10

-18

-10

-10

-2 24 8

-10

-10 10 10

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 3-3: RTNDT OF NOZZLE, WELD AND STUD MATERIALS COMPONENT ID HEAT TEST CHARPY (Ts-60) DROP RTNDT TEMP.

ENERGY

(°F)

WEIGHT (OF)

(OF)

(FT-LB)

NDT

(°F )

Nozzles:

Recirc. Outlet Nozzle Recirc. Inlet Nozzle Steam Outlet Nozzle Feedwater Nozzle Core Spray Nozzle Top Head Instrumentation Nozzle Vent Nozzle Jet Pump Instrumentation Nozzle CRD Hyd. Sys. Return Drain Nozzle G-3419-1 G-3419-2 G-3436-1 G-3436-2 G-3436-3 G-3436-4 G-3436-5 G-3436-6 G-3436-7 G-3436-8 G-3436-9 G-3436-1 0 G-3420-1 G-3420-2 G-3420-3 G-3420-4 G-3421-1 G-3421-2 G-3421-3 G-3421-4 G-3422-1 G-3422-2 G-2921-3 G-2921-4 G-2920-2 G-3424-1 G-3423 G-2085 EV-9781 AV-1872 E21VW-104J10 E21VW-104J2 E21VW-104J9 E21VW104J7 E21VW-104J6 E21VW-104J3 E21VW-104J4 E21VW-104J8 E21VW-104J5 E21VW-104J1 EV-9754 EV-9775 EV-9775 AV-1 576 EV-9741 EV-9741 EV-9741 AV-1607 EV-9741 EV-9741 EV-9781 AV-2379 AV-2374 EV-9792 EV-9143 2106172 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 70 93 103 82 95 86 84 89 106 101 73 110 82 66 62 30 65 92 69 30 40 54 711 76 86 172 1111110 94 101 76 109 94 109 114 116 93 105 40 75 52 75 75 67 36 56 89 96 107 79 107 77 116 102 118 103 82 36 78 48 73 90 68 32 65 74

-20

-20

-20

-20

-20

-20

-20

-20

-20

-20.

-20

-20

-20 8

-20 20

-20

-20

-20

-20

-30

-50

-20

-20

-20

-20

-20

-20

-10 0

<40

<40

<40

<40

<40

<40

<40

<40

<40

<40

-10

-10

-10 0

10 10

-20 0

0 10 0

-10 0

0

-20

-10 0

30 30 30 30 30 30 30 30 30 30

-10 8

-10 20 10 10

-20 20 0

10 0

-10 0

0

-20 20 82 69 72 1171 90 1108 145 11.82 1185 1441144 144 80 92 112 94 96 1108

_____________________________________.1 i

I _____________ ~

II

GE Nuclear Energy GE-NE-B 1100732-01 Revision I COMPONENT ID HEAT TEST CHARPY (TwT-60)

DROP RTNOT TEMP.

ENERGY

('F)

WEIGHT

('F)

(°F)

(FT-LB)

NDT (OF)

WELDS:

Vertical Welds Lower Shell 2-233 ABC 27204/12008 10 63 60 49

-48

-48 Lower-Int Shell 1-233 A,B,C 13253/12008 10 60 64 56

-50

-50 Girth Welds Lower to Lower-lnt Shells 1-240 305414 10 82 66 80

-50

-50 LST STUDS:

G-3134-1 37385 10 39 40 39 70 OK G-3134-2 37677 10 60 55 57 70 OK 12

GE Nuclear Energy GE-NE-B1100732-01-, Revision 1 TABLE 3-4: CHEMICAL COMPOSITION OF FITZPATRICK SURVEILLANCE MATERIALS FROM SURVEILLANCE SPECIMEN CHEMICAL TESTS Metal Metal Mn Ni Cu Mo Si Cr P

Sample ID Sample (wt%)

(wt%)

(wt%)

(wt%)

(wt%)

(wt%)

(wt%)

Type 5CL a Base 1,40 0.62 0.11 0.48 0.070 0.11 0.011 5CMa Base 1.30 0,63 0.12 0.50 0.06' 0.11 0.010 29283 Base 1,17 0.58 0.11 0.45 0.36 0.09 0.013 29285 Base 1,25 0.61 0.11 0.46 0.16 0.10 0.013 29286 Base 1.20 0,60 0.11 0.46 0.19 0.10 0.011 LPI-28c Base 1.43 0.62 0.10 0.42 0.24 N/A 0.018 Baselined Base 1.26 0.57 0.13c 0.48 0.22 N/A 0.011

DataAvg, 1.29 0.60 0.11 0.46 0.23 0.10 0.012 Ski. Dev.

0.10 0.02 0.01 0.03 0.08 0.01 0,003 5DL Weld 1.50 0.72 0.31 0.50 060 0.04 0.015 5DMa Weld 1.40 0.72 0.31 0.51 0.0 6b 0.04 0,014 29289 Weld 1,36 0.70 0.30 0.48 0.38 0.04 0.014 29295 Weld 1.25 0.70 0.23 0.47 0,41 0.04 0.014 29297 Weld 1.39 0.74 0.31 0.49 0.52 0.04 0.012 DataAvg.

1,38 0.72 0.29

0.

4.44 0.04 0.014 Sitd. Dev.

0.09 0.02 0.03 0.02 0,07 0.001 0.001 a

b d

e Chemical analysis of tensile specimens from 3Q0 azimuthal capsule location (1 st capsule report) [8].

Si results may be low due to precipitiation during dissolution heating (Results not used in Average).

Data taken from the BWROG Supplemental Surveillance Program for the FitzPatrick Plant.

Taken from original fabrication records (see Table 3-1).

Cu value taken from Lukens Steel letter to NYPA dated 10/14/85 [1 91 13 I

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GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 FIGURE 3-1: SURVEILLANCE CAPSULE HOLDER RECOVERED FROM FITZPATRICK (1200 AZIMUTHAL LOCATION CAPSULE - REMOVED AT 13.4 EFPY) 14

GE Nuclear Energy GE-NE-B I 100732-01 Revision 1 FIGURE 3-1(A): CHARPY SPECIMEN CAPSULE IDENTIFICATION (120" AZIMUTHAL. LOCATION CAPSULE - REMOVED AT 13.4 EFPY)

GE Nuclear Energy GE-NE-81100732-01 Revision I i _-/v /

TOP HEAD ENCLOSURE CLOSURE FLANGE REGION 4

PLATE G-3413-7 HEAT C3368-1 CORE BELTLINE REGION U ~

f UPPER SHELL

=

UPPER INTERMEDIATE SHELL LOWER INTERMEDIATE SHELL PLATE G-3414-1 HEAT C3301-1

'11\\

A LONGITUDINAL SEAM WELD 1-233 K

BOTTOM HEAD ENCLOSURE LOWER SHELL WELD 2-233 PLATE G-3415-1R3 HEAT C3394-1

/ PLATE G-3414-2 HEAT C3278-2 CIRCUMFERENTIAL SGIRTH WELD 1-240 PLATE G-3415-3 HEAT C3376-2 Kk I/

>44<

L--

I FIGURE 3-2. SCHEMATIC OF RPV SHOWING IDENTIFICATION OF VESSEL BELTLINE PLATES AND WELDS 16

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

4. PEAK RPV FLUENCE EVALUATION Flux wires removed from the 1200 location 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.

4.1. FLUX WIRE ANALYSIS 4.1.1 Procedure The surveillance capsule contained 2 sets of 3 flux wires: iron, nickel, and copper. Each wire was removed from the capsule, cleaned with dilute acid, weighed, mounted on a counting card, and analyzed for its radioactivity content by gamma spectrometry. Each iron wire was analyzed for Mn-54 content, each nickel wire was analyzed for Co-58 content, and each copper wire for Co-60 at calibrated source-to-detector distances with 170-cc Ge and 100-cc Ge(Li) gamma detectors used in conjunction with a Nuclear Data 6700 multichannel analyzer system.

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 were derived from records provided by New York Power Authority are shown in Table 4-1.

Zero power days between fuel cycles are listed as well.

From the flux wire activity measurements and power history, reaction rates for Fe-54 (n,p) Mn-54, Ni-58 (n,p) Co-58, and Cu-63 (n,cc) Co-60 were calculated. The E >1 MeV fast flux reaction empirical cross sections for the iron, nickel, and copper wires are 0.182 barn, 0.234 barn and 0.00318 barn, respectively. The calculated fluence result from the iron flux wire was used. The fluence result from the iron specimen was confirmed by the Ni and Cu flux wires, with all three results differing by less than 10%. The GE empirical activation cross sections are consistent with other transport code cross sections, and parallel calculations were performed using the both the empirical and transport code cross sections [20]. However, the fluence results obtained from the empirical cross sections are recommended since they yield approximately 4%

higher estimates of RPV fluence, These data functions were applied to BWR pressure vessel locations based on water gap (fuel to vessel wall) distances. The cross sections for > 0.1 MeV flux were determined from the measured 0.1 to I MeV cross section ratio of 1.6 [11].

17

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 4.1.2 Results The measured activity, reaction rate and full-power flux results for the 120' location surveillance capsule are given in Table 4-2. The E > 1 MeV flux values were calculated by dividing the wire reaction rate measurements by the corresponding cross sections, factoring in 17 2

the local power history for each fuel cycle. The fluence result, 5.0 x 10 n/cm (E > I MeV),

was obtained by using the following equation:

oCu DfpO tipi (4-1)

where, 0cu

= fluence measured by the Cu dosimeters, n/cm-cID = full power flux value for Cu, n/cm -s Ti

=operating time, s A

= full power fraction as shown in Tables 4-1 through 4-3.

The accuracies of the values in Taole 4-2 for a 2a deviation are influenced by the following sources of error:

+ 2%

counting rates

+ 15%

power history

+ 10%

cross sections The uncertainty in the E > I MeV fluence is approximately +/--20% (2a).

This analysis is performed using the GE empirical activation cross sections. A parallel analysis using cross sections from a transport code was made, but is not preferred, because its resulting fluences were approximately 4% lower for all three of the flux wires.

4.2 DETERMINATION OF LEAD FACTOR The flux wires from the surveillance capsule are used to determine the fast neutron (E > 1 MeV) fluence at the location of the capsule as described in Section 4.1. However, the capsule and flux wires are not located where the pea. vessel fluence occurs. A calculated lead factor is used to 18

GE Nuclear Energy GE-NE-B 1100732-01 Revision I relate the fluence at the location of the wires to the peak fluence at the vessel. The lead factor is defined as the ratio of the fast neutron fluence at the surveillance capsule to the peak fluence at the vessel inside surface. A neutron transport analysis was performed to determine the effective full power fast neutron flux distribution at the reactor pressure vessel. The lead factor was evaluated as the ratio of the calculated effective full power fast neutron fluxes at the capsule and vessel peak flux locations. Calculation of the fluxes and lead factor requires modeling of the reactor geometry and materials and depends on the distributions of power density and coolant voids in the core. The lead factor was calculated for the FitzPatrick geometry, using data for a typical operating cycle to determine power shape and void distribution. The lead factor was not adjusted for the 105% power uprate, as the fluxes were assumed to increase linearly with power.

The methods used to calculate the lead factor are discussed below.

The NRC is developing Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", which will include guidance concerning acceptable methods and assumptions for determining the pressure vessel fluence. At this time, the draft has not been finalized for issuance as a Regulatory Guide. However, while the specific regulatory requirements are still subject to change, it is believed that the analysis described in this section is consistent with the intent of the draft guide.

4.2.1 Procedure The lead factor for the RPV inside wall was determined by using a combination of two separate two-dimensional neutron transport computer analyses. The first of these established the azimuthal and radial variation of flux at the fuel midplane elevation.

The second analysis determined the relative variation of flux with elevation. The azimuthal and axial distribution results were combined to provide a simulation of the three-dimensional distribution of flux. The ratio of fluxes, or lead factor, between the surveillance capsule location and the peak flux locations was obtained from this distribution.

The DORT computer program, which utilizes the discrete ordinates method to solve the Boltzmann transport equation in two dimensions, was used to calculate the spatial flux distribution produced by a fixed source of neutrons in the core region. The analysis considered neutrons with energies above 0. 1 MeV and used 29 energy groups above this threshold-Angular dependence of the neutron scattering cross-sections was approximated by a third-order Legendre polynomial (P-3) expansion. The DORT calculations were run using Sg angular quadrature.

19

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The azimuthal distribution was obtained with a model specified in (R,O) geometry, assuming eighth-core symmetry with reflective boundary conditions at 00 and 451.

In this model, O= 300 is symmetrically equivalent to the 1200 capsule location. A schematic of the (R,0) model is shown in Figure 4-1.

The model incorporates inner and outer core regions, bypass water region, shroud, downcomer water region, and a vessel plus liner region. The portion of the core inside a radius of 133 cm was not included because it will not significantly influence the flux distribution at the vessel. The spatial mesh contained 155 steps of varying sizes in the radial dimension. The azimuthal mesh step was specified to be 1/2' and was reduced to 1/40 in the vicinity of the capsule, resulting in a total of 98 azimuthal intervals. The (R,O) model used core region material compositions and neutron source densities for the core midplane elevation (75 inches above the bottom of active fuel). This is near the elevation of the capsule, which is centered at 72.31 inches above the bottom of the active fuel. The neutron source densities and coolant mass densities were based on cycle-average values for the selected representative operating cycle. The output of this calculation provided the distribution of flux as a function of azimuth and radius at reactor midplane. The azimuth of the peak flux and its magnitude relative to the flux at the 30' capsule/flux wire azimuth were determined from this distribution.

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 cylinder radius was specified to be equal to the radius of the outermost comer of the core, which is located at an azimuth of approximately 393O. The core model contained inner and outer material regions for each of 25 axial fuel nodes (total of 50 core regions). Source densities and coolant densities in these regions were based on cycle-average values for the representative cycle. The elevation of the peak flux at the reactor vessel inside surface and the magnitude of the peak flux relative to the flux at the surveillance capsule elevation were determined from the (RZ) flux distribution results.

4.2.2 Results The relative distribution of flux at the RPV base metal inside surface vs. azimuthal angle obtained from the (R,0) calculation is shown in Figure 4-2.

The relative distribution of flux versus elevation at the RPV inside surface from the (R,Z) calculation is shown in Figure 4-3.

The azimuthal distribution (Figure 4-2) indicates that the 8 flux maxima at the vessel base metal inside surface occur at azimuthal locations which are displaced by 42.75' from the RPV quadrant reference axes (0', 900, etc.). From the R,Z results (Figure 4-3), the peak is estimated to occur at 20

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 an elevation about 79 inches above the bottom of the active fuel. The calculated core midplane E > 1 MeV flux at the (R,0) coordinates corresponding to the equivalent capsule center position (0 = 300, R = 109.19 inches) was 1.382 xl09 n/cm2/s. This was multiplied by the ratio of flux at the capsule elevation to flux at midplane (0.996), as determined from the (RZ) calculation, resulting in a calculated flux at the capsule location which rounds to 1.38 x 109 n/cm Is. The peak flux at the vessel surface (R = 110.375 inches) was similarly obtained by multiplying the calculated midplane flux of 2.015 xlO9 n/cm2 Is at the peak azimuth by an axial adjustment factor of 1.003 from the (RZ) calculation. The resulting peak flux estimate is 2.02 x 109 n/cm 2/s.

Consequently, the lead factor is 1.38 x 16912.02 x 10 =0.68.

The calculated capsule full power flux of 1.38x10 9 n/cm2/s obtained with this model is about 16 % higher than the capsule dosimetry result of 1.19x10 9 n/cm/s.

The indicated agreement between the analytical and experimental results is within the uncertainties associated with those results and is considered good. It is estimated that the l uncertainty in the calculated flux magnitudes is on the order of 25 - 30 %. However, since the lead factor is determined from the ratio of two calculated fluxes which have sources of error in common, the la uncertainty in the lead factor is estimated to be no more than 15 %.

Use of a lead factor calculated on the basis of the model described above is consistent with current GE practice for estimation of the peak vessel fluence. Application of the lead factor to the capsule dosimetry results yields an estimated end-of-cycle 12 peak fluence of 5.Ox101 /

17 299 0.68=7.4x10 n/cm and an estimated peak full power flux of 1.19x 109/0.68 = 1.75 x 10 n/cm 2/s at the vessel inside surface. Since the estimated I a uncertainty in the dosimetry results is 10 %

and the estimated la uncertainty in the lead factor is 15%, the combined overall la uncertainty in the projected peak values is estimated to be about (102 +152)0. = 18%.

The analysis model discussed above did not include the effects of the material specimens and specimen holder on the local neutron flux. A second calculation was performed in (R,O) geometry with a model which incorporated regions which simulated the material specimens and holder. The densely packed material specimens were represented as solid steel in the model. The perforated wall of the specimen holder was modeled as a steel/water mixture.

This model is expected to provide a reasonable upper bound estimate of the effect of the capsule on local fluxes. The results obtained with this model were also used to provide independent confirmation of the reaction rate cross-sections used in the dosimetry analysis described in Section 4.1.

21

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The flux obtained at the capsule midpoint radius with the modified (RO) model was 1.53x109 rnlcm 2/s. Application of the axial adjustment of 0.996 results in an estimated flux of 1.52xl 0 9 n/cm 2/s at the capsule center point. Consequently, the flux calculated at this point with the simulated capsule materials is about 10 % higher than the flux calculated with the base model. The region-averaged flux obtained in the specimen region, 1.5lxl 0 n/cm2is, differs only slightly from the center point value. These results indicate that the base model under-predicts the flux within the capsule by a few percent and possibly as much as 10 %.

Therefore, a conservative bias exists in the calculated lead factor and projected peak fluences, since underestimation of the lead factor results in overestimation of the vessel peak fluence.

The 29-group neutron energy spectrum obtained at the simulated capsule center point was plotted and applied to ENDF/B-VI library data for the dosimeter activation reaction cross-sections to calculate spectrum-weighted group cross-sections for the reactions. The DORT case was re-run toobtain calculated total reaction rates which, when divided by the E > I MeV flux, yield the effective reaction rate cross-sections for the fast flux.

The cross-sections used in Section 4.1 to analyze the dosimeter data are derived from fits to empirical data which have been used by GE for analysis of surveillance capsule dosimetry for many years. Region-averaged values obtained for the specimen region in the DORT model are compared with the Section 4.1 cross-sections in the table below.

Comparison of Calculated Activation Cross-Sections in Simulated Capsule Region With Semi-Empirical Cross-Sections Used in Capsule Dosimetry Analysis Effective Cross-Section for E> I Mev Flux (barns)

Difference Reaction Empirical Fit Calculated Fe54(n,p)Mn54 0.182 0.1899

-i4.34 Ni58(n,p)Co58 0.234 0.2425

+3.63 Cu63(na)Co6O 0.00318 0-003305

-3.93 The close agreement between the calculated cross-sections and the fit-derived cross-sections provides confidence that the empirically derived cross-sections are reliable. It also provides confidence that the calculated neutron spectrum is realistic, even though the magnitude of the calculated flux is somewhat greater than the measured flux. In each instance, the calculated

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 cross-sections are slightly higher than the empirical cross-sections.

Consequently, if the dosimeter material reaction rates are predicted purely from the analysis, the difference between calculated and measured reaction rates will be slightly greater than the difference between the calculated and measured fluxes. The reaction rates are compared below.

Comparison of Calculated Reaction Rates in Simulated Capsule Region With Reaction Rates Determined From Capsule Dosimetry Analysis Dosimeter Reaction Rate (reactions/s/nucleus)

Difference Reaction Capsule Dosimeters Calculated Fe54(n,p)Mn54 2.14E-16 2.86E-16

+33.8 Ni58(n.p)Co58 2.70E-16 3.66E-16

+35.4 Cu63(n,a)Co6O 3.91E-18 4-98E-18

+27.4 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 the Reg. Guide 1.99, Rev. 2 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 at a particular EFPY and lead factor according to:

fsurf= (fcap

• 32 EFPY)/(LF

• CEFPY)

(4-2)

where, fsurf = 32 EFPY fluence at the peak vessel inside surface fcap = capsule fluence measured at the CEFPY 32 EFPY = end of life EFPY based on a 40-year operation at an 80% capacity factor CEFPY the current EFPY for the capsule LF

= lead factor 23

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The surveillance capsule was removed from FitzPatrick at 13.4 EFPY as calculated in 17 2

Table 4-2. The fluence at 13.4 EFPY was determined to be 5.0 x 10 nicm using Equation 4-1, and the lead factor was determined to be 0.68 as discussed in Section 4.2. In addition, the fluence over the remaining 18.6 EFPY was increased by 5% to account for the 5% power uprate that began in December 1996. Using this information with Equation 4-2, the resulting 32 EFPY fluence value at the peak vessel inside surface is:

fsurf= [(5.0x 10 )+(5.Ox 10"18.6/13.4)*1.05]/0.68= 1.81 x 10 n/cm (4-3) at the peak location.

The peak surface fluence at 32 EFPY is 22% lower than the nominal value (2.32 x 10 n/cm) that was calculated from the first surveillance capsule dosimetry as a result of power uprate as reported in GE report [15].

This variation can be attributed to refinements in the analysis technique since the first capsule was removed.

The 1/4 T fluence (f) is calculated according to the Reg. Guide 1.99 [7] equation:

-0.24x f= fsurf(e

),

(4-4) where x = distance, in inches, to the 1/4 T depth. The vessel beltline lower intermediate shell ring thickness is 5.375 inches minimum requirement. The corresponding depth, x, taken from the minimum required thickness is 1.34 inches for the lower intermediate shell. Equation 4-4 18 evaluated for this value of x gives the 1/4 T value of 32 EFPY fluence, f= 1.31x10 n/cm2 for the lower intermediate shell ring.

In the case of the lower shell ring, the axial fluence distribution was also taken into account. The maximum fluence at the top of the lower shell is 0.89 times the peak fluence, or 1.61 x 10's n/cm2.

The minimum plate thickness of the lower shell is 6.375 inches, which corresponds to an x value of 1.6 inches. The resultant 1/4T fluence at 32 EFPY is 1.10 x 1017 2

n/cm.

24

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 4-1:

SUMMARY

OF FITZPATRICK IRRADIATION PERIODS On Off Duration Days to eoi MWd Effective Full Full Power (days)

Power Days Fraction 1/26/75 12/31/77 1071 6874 1301203 534.4 0.499 1/1/78 12/31/78 365 6509 539687 221.6 0.607 1/1/79 12/31/79 365 6144 373919 153.7 0.421 1/1/80 12/31/80 366 5778 541475 222.2 0.607 1/1/81 12/31/81 365 5413 592405 243.1 0.666 1/1/82 12/31/82 365 5048 630106 258.8 0.709 1/1/83 12/31/83 365 4683 592197 243.1 0.666 1/1/84 12/31/84 366 4317 633307 259.9 0.710 1/1/85 12/31/85 365 3952 532365 218.6 0.599 1/1/86 12/31/86 365 3587 767477 315.0 0.863 1/1/87 12/31/87 365 3222 545590 224.1 0.614 i/1/88 12/31/88 366 2856 557082 228.8 0.625 1/1/89 12/31/89 365 2491 781820 320.8 0.879 1/1/90 12/31/90 365 2126 592684 243.5 0.667 I/1/91 1/31/91 31 2095 69083 28.4 0.915 2/1/91 2/28/91 28 2067 56800 23.3 0.833 3/1/91 3/9/91 9

2058 19191 7.9 0.875 3/17/91 3/18/91 2

2049 116 0.1 0.024 4/13/91 4/30/91 18 2006 34493 14.2 0.787 5/1/91 5/7/91 7

1999 16095 6.6 0.944 8/18/91 8/31/91 14 1883 26087 10.7 0.765 9/1/91 9/30/91 30 1853 72905 29.9 0.998 10/1/91 10/31/91 31 1822 74840 30.7 0.991 S1/1/91 11/28/91 28 1794 63288 26.0 0928 11/29/91 1/2/93 401 1393 0

0.0 0.000 1/3/93 1/31/93 29 1364 14983 6.2 0.212 2/1/93 2/28/93 28 1336 58272 23,9 0.854 3/1/93 3/31/93 31 1305 17725 7.3 0.235 4/1/93 4/30/93 30 1275 51219 21.0 0.701 5/1/93 5/31/93 31 1244 46629 19.1 0,617 6/1/93 6/30/93 30 1214 72730 29.8 0.995 7/1/93 7/31/93 31 1183 72348 29.7 0.958 8/1/93 8/31/93 31 1152 75443 31.0 0.999 9/1/93 9/30/93 30 1122 62975 25.9 0.862 10/1/93 10/31/93 31 1091 55927 23.0 0.741 11/1/93 11/30/93 30 1061 13756 5.6 0.188 12/1/93 12/31/93 31 1030 74988 30.8 0.993 1/1/94 1/31/94 31 999 75300 30.9 0.997 2/1/94 2/28/94 28 971 68114 28.0 0.999 3/1/94 3/31/94 31 940 73706 30.3 0.976 4/1/94 4/30/94 30 910 4546 1.9 0.062 5/1/94 5/31/94 31 879 63588 26.1 0.842 6/1/94 6/30/94 30 849 71339 29.3 0.976 7/1/94 7r/3 1/94 31 818 68452 28.1 0.906 25

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 8/1/94 8/31/94 31 787 61533 25.3 1

0.815 9/1194 9/30/94 30 757 54488 22.4 U

0.746 10/1/94 10)31/94 31 726 54520 22.4 0.722 11/1/94 11/30/94 30 696 47247 19.4 0.647 12/1/94 121/31/94 31 665 0

0.0 0.000 1/1/95 1..

1/31,95 31 634 0

0.0 0.000 2/1/95 2/28/95 28 606 0

0.0 I

0.000 3/1/95 3/31/95 31 575 5960 2.5 0.079 4/1/95 4/30/95 30 545 69366 28.5 0.949 5/1/95 "

5/3195 31 3t 514 72287 29.7 0.957 6/1/95 6/30/95 30 484 49822 20.5 "

0.682 7/1/95 7/31/95 31 453 75412 31.0 0.999 8/1/95 8/31/95 31 422 75410 31.0 T

0.999 9/1/95 9/30/95 30 392 53600 22.0 0.733

.0/1/95 10/31/95 31 361 75437 31.0 0.999 11/1/95 11/30/95 30 331 73014 30.0 0.999 12/1/95 12/31/95 31 300 73993 30.4 0.980 1/1/96 1/31/96 31 269 75173 30.9 0.995 2/1/96 2/29/96 29 240 51562 21.2 1

0.730 3/1/96 3/31/96 31 209 56448 23,2 0.747 4/1/96 4/30/96 30 179 72990 30.0 0.999 5/1/96 5/31/96 31 148 73629 30.2 0.975 6/1/96 6/30/96 30 118 71757 29.5 0.982 7/1/96 7/31/96 31 87 75250 30.9 0.996 8/1/96 8/31/96 31 56 73687 30.3 0.976 9/1/96

.9/30/96 30 26 49799 20.4 1

0.681 10/1/96 10/26/96 26 0

56785 j

23.3 0.897 Note: Full power was taken as the value prior to uprate of 2436 MWt Total Effective Full Power Days= 4907.8 Total Effective Full Power Years = 13.4 26

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i GE Nuclear Energy GE-NE-B1100732-01, Revision I TABLE 4-2: SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM START-UP TO 11 /12/96 (13.4 EFPY) USING EMPIRICAL CROSS SECTIONS (GE CORRELATION)

Average Average

.. Full:Powerlu

1. ?ab Full Power Fluxc

. Fluence.

Funce'Y Wir dp/

Bekneat Re 1actioni Rate W2./m-4Q/r~

2/t 2

(Elemegt)

(44 end of irxaociation),

dp cucs 644trated)1 V-I MeV

.p E0A I

V........E>Ot1meV Iron 1.07E05 2.14E-16 1.18E09 1.89E09 5.00E17 7.99E17 Nickel 1.67E06 2.70E-16 1.16E09 1.86E09 4.90E17 7.85Ei7 Copper 1.56E04 3.91E-18 1.231E09 1.97E09 5.21E 17 8.34E 17 a

b c

Obtained by R.D Reager [20]

Full power flux, based on thermal power of 2436 MWt 1.6 times the E >1 MeV result 27

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I GE Nuclear Energy GE-NE-B 1100732-01, Revision I TABLE 4-2: SURVEILLANCE CAPSULE FLUX AND FLUENCE FOR IRRADIATION FROM START-UP TO 11/12/96 (13.4 EFPY) USING EMPIRICAL CROSS SECTIONS (GE CORRELATION)

Average' AVerage PullPQyerpluxb

F41 Po w*er F1 x Fiicc;,

F0uence Wire dps/g Elemnent Reaction R~ate,

("/cni 2"sy (n/

94s (ien/CM) bc

~Eeie~)

(at clid qf iraitn)

(dsnces~stqrated)I.

E>LMk MeW.ilM6 E>1 MW E4.1MeV~

Iron 1.07E05 2.14F,-16 1.18E09 1.89E09 5.00E17 7.99E17 Nickel 1.67E06 2.70E-16 1.16E09 1.86E09 4.90E 17 7.85E117 Copper 1.56E04 3.911E-18 1.23E09 1.97E09 5.2 1E17 8.34E17 a

b C

Obtained by R.D Reager (20]

Full power flux, based on thermal power of 2436 MWt 1.6 times the E >1 MeV result 27

,GE Nuclear Energy GE-NE-B1100732-01, Revision I TABLE 4-3. MEASURED FLUX VS. THEORETICAL FLUX FOR DOSIMETER AND FLUX WIRES E> I McV Lead Factor EFPY*

Measured Capsule Fluence EOL (32 EFPY) FLUENCE Capsule to ID Capsule Flux (n/cm2)

(n/cm 2)

'Surface

.(nlcm2-s)

ID Surface I/4T Location 1982 30' Az,.imuth Dosimeter I

I Ix 109 1

1 1985 30" Azimuth Flux Wires 0.79 6.0 I.4x10 2.6x10 1.8x1011 1.35x10'8 Upper Bfound (1,25 Factor) 2.2x10'"

1.7x 10l Reg. Guide 1.99 Rcv.2 Evaluation, no 0.61 2,32x10

• 1.7X10' upper bound factor included.

Tcch Spec P-T curve basis.

5%, Power Uprate based on tipper 2.44x 10 1,76x 10 bound value.

1996 120' Azimuth Flux Wires 0.68 13.4 1.2x10 9 5.0x:10" 1.81xl]0' 1.38x1O"I Includes 5% Power Uprate New P-T curve basis, Effective Full Power Years at 2436 Mwv, 27b

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 66 INTERVALS TOTAL SHROUD: 9 INTERVALS 457 98 INTERVALS IN AZIMUTHAL DIRECTION 1 = CORE INTERIOR FUEL 2 = CORE EXTERIOR FUEL WATER REGION:

59 INTERVALS VESSEL WALL:

21 INTERVALS 01 FIGURE 4-1: SCHEMATIC OF MODEL FOR AZIMUTHAL FLUX DISTRIBUTION ANALYSIS 28

GE Nuclear Energy GE-NE-B 1100732-01, Revision 1 1.00 0.90~

0.80 W

0.7 0 1 x0-601 Q.-050fl 0.0 z

S0.40 I-I Uj0.30 IIr I

I 02011 0.10 I

0.00i 0

5 10 15 20 25 30 35 40 45 ANGULAR POSITION (DEGREES)

FIGURE, 4-2: RELATIVE FLUX VS. ANGLE AT RPV INSIDE SURFACE 29

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Ax

-J U.

z 0

zw w

I.

1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 t 2

15 25

.~-

35 45 55 65 75 85 95 105 115 125 13:

DISTANCE FROM BAF (INCHES)

FIGURE 4-3: RELATIVE FLUX VS. ELEVATION AT RPV INSIDE SURFACE 30

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

5. CHARPY V-NOTCH IMPACT TESTING 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. Testing was conducted in accordance with ASTM E23-94b [12].

5.1 IMPACT TEST PROCEDURE The Vallecitos testing machine used for irradiated specimens was a Tinius Olsen impact machine, serial number 175363. The maximum energy capacity of the machine is 300 ft-lb, which produces a test velocity at impact of 19.3 ft/sec.

The Tinius Olsen machine was qualified using NIST standard reference material specimens. The Standard Reference Materials (SRMs) consist of three sets of specimens which cover the energy range of the apparatus. Each set has a designated failure energy and a standard test temperature. According to ASTM E23-94b [12], the test apparatus averaged results must reproduce the NIST standard values within an accuracy of +/-5% or +1.0 ft-lb, whichever is greater. The results of the qualification of the Tinius Olsen impact machine are summarized in Table 5-1.

Charpy V-Notch tests were conducted at temperatures between -80'F and 400'F. The cooling fluid used for irradiated specimens tested at temperatures at or below 501F was ethanol.

At temperatures between 50'F and 210'F, water was used as the temperature conditioning fluid.

The specimens were heated in silicon oil for test temperatures above 210'F. Cooling of the conditioning fluids was done by heat exchange with liquid nitrogen through a copper coil; heating was done by an immersion heater.

The bath of fluid was mechanically stirred to maintain uniform temperatures. The fluid temperature was measured with a calibrated Type K thermocouple positioned near the impact samples. After equilibration at the test 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.

For each Charpy V-Notch specimen the test temperature, energy absorbed, lateral expansion, and percent shear were determined. In addition, photographs were taken for the 31

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 irradiated specimens. Lateral expansion and percent shear were measured according to specified methods [12]. Percent shear was determined using method number I of Subsection 11.2.4.3 of ASTM E23-94b [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 E23-94b [12].

5.2 IMPACT TEST RESULTS Eight Charpy V-Notch specimens each of irradiated base, weld, and HAZ material were tested at temperatures (-80'F to 400'F) selected to define the toughness transition and upper shelf portions of the fracture toughness curves. The absorbed energy, lateral expansion, and percent shear data are listed for each material in Table 5-2. Plots of absorbed energy and lateral expansion for base, weld, and HAZ materials are presented in Figures. 5-1 through 5-6. These curves are plotted along with the corresponding curves from the first capsule (and unirradiated base material data where appropriate) in Figures 5-7 through Figure 5-12. The fracture surface photographs and a summary of the test results for each specimen are contained in Appendix A.

The unirradiated and irradiated plate and weld energy and lateral expansion daia are fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook [13] (HLAZ was not fit due to data scatter):

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.

The TANH function is one of the few continuous functions with a shape characteristic of low alloy steel fracture toughness transition curves.

5.3 IRRADIATED VERSUS UNIRRADIA TED CHARPY V-NOTCH PROPERTIES Ideally, a shift in RTODT would be established by comparing the irradiated Charpy specimen data to baseline unirradiated Charpy data. For the case of the FitzPatrick base material specimens, data was obtained from the Certified Material Test Report. Additional Charpy test

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 data for the FitzPatrick surveillance plate (heat number C3278-2) was available from the BWROG Supplemental Surveillance Program report [17]. This program was useful in providing plant-specific data and information for the FitzPatrick base material to establish baseline properties. The unirradiated data for the base material, as well as the results for both the plate and weld materials from the first and second surveillance capsules, were fit to a TANH function as described in the previous section. The unirradiated properties for the surveillance plate were determined from the combined sets of data, as shown in Figure 5-13. For the weld material, no credible unirradiated baseline data was available.

5.4 COMPARISON TO PREDICTED IRRADIATION EFFECTS 5.4.1 Irradiation Shift The measured transition temperature shifts for the base and weld materials were compared to the predictions calculated according to Rev. 2 [7]. The inputs and calculated values for irradiated shift for the plate and weld materials based upon measurements taken from the 120' azimuth capsule at 13.4 EFPY are as follows:

Plate:

Weld:

Copper 0.11%

Nickel =

0.60%

CF 74 17 2

fluence 5.0x 10 n/cm Reg. Guide 1.99 ARTNDT Reg. Guide 1.99 ARTNDT +/- 2aA(3 4°F)

Measured 30 ft-lb shift =

Copper 0.29%

Nickel =

0.71%

CF =

208 17 2

fluence =

5.0 x 10 n/cm Reg. Guide 1.99 ARTNDT =

Reg. Guide 1.99 ARTNTJT +/- 2crA(56°F) 21.7 0F 55.77F max, -12.3)F min 14.97°F 60.9°F 1 16.9 0F max, 4.90F min 33

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The weight percents of Cu and Ni are best estimates based on averaging (see Table 3-4).

The CF values shown above are the chemistry factors for the materials obtained from Rev. 2.

The fluence factor for the Reg. Guide calculation of 30 ft-lb shift may either be calculated according to the Rev. 2 definition fluence factor = f( 0. 2 8 - 0. 10 log f)

(5-)

or it may be obtained from Rev. 2 Figure 1 [7]. Using Equation 5-1, the fluence factor was calculated to be 0.293.

These values are used to calculate the Reg. Guide 1.99 prediction for 30 ft-lb shift and USE decrease for comparison to the measured shift and USE decrease for the irradiated surveillance materials. The predicted 30 ft-lb temperature shift (ARTNDT) was also calculated according to Rev. 2 using the equation ARTND,, = (CF) f (0.28 - 0.10 log f)

(5-2)

The measured 30 ft-lb temperature shift (Table 5-3) of 14.97°F for the plate material is within the bounds of the Reg. Guide prediction. Since two credible data sets are available for the plate material, the ART prediction was modified in a manner consistent with Position 2 of Rev. 2, as described in Section 7.

A least squares fit to the 30 ft-lb shift (AT30) values was performed as shown in Figure 5-14.

This figure shows the comparison of the AT30 vs. fluence relation predicted by Reg. Guide 1.99, Rev. 2 and the actual fitted results for the surveillance plate. It is noted that the fitted curve exhibits less embrittlement than that predicted by the Reg. Guide for this plate material.

5.4.2 Change in USE Using the copper and fluence data above with Figure 2 of Rev. 2, decreases in USE of approximately 9% are predicted for the plate and 19% for the weld material for the first capsule.

For the second capsule, the predicted decreases in USE are 10% and 22% for the base and weld materials, respectively. In the base metal, the USE increased from the unirradiated to the second capsule. (Since unirradiated weld data was not available, no value was used.)

The USE decreased for both the plate material and the weld material from the first to the second capsule.

34

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 5-1: VALLECITOS QUALIFICATION TEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS Specimen Energy Vari~aiie Identification....Absorbed (fIb VNC NIST Vallecitos LL-45 1 12.71 Tinius Olsen LL-45 2 13.75 Machine LL-45 3 14.20 (tested 6/96)

LL-45 4 13.10 LL-45 5 14.10 Average:

13.572 12.836

+0.736 ft-lbs HHJ-46 1 71.0 HH-46 2 75.5 HH-46 3 76.0 HH-46 4 76.5 HH-46 5 75.0 Average:

74.808 74.284

+0.71%

SH-6 1 170.5 SH-6 2 168.0 SH-6 3 154.0 SH-6 4 154.0 SH-6 5 165.0 Average:

162.327 165.831

-2.11%

Allowable Variance is 1.4J (I ft-lb) or 5%, whichever is greater (ASTM STD-E23) 35

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 5-2: IRRADIATED CHARPY V-NOTCH IMPACT TEST RESULTS SECOND CAPSULE

Test.

Fracture Lateral Shear.

VNC Specimen Temperature.

Energy Expansion (Method 1)

ID Identification' (OF)

(fti-b)

(mils)

[12]

Base:

29292 53U

-50 10.0 8

1 Heat C3278-2, 29291 53M 0

35.4 31 29 Longitudinal 29287 53P 24 44.2 39 31 29283 53Y 49 81.0 65 46 29285 53B 103 96.5 77 78 24286 52D 150 117.4 89 100 29293 53L 250 120.4 91 100 29284 527 400 126.7 93 100 Weld:

29288 56A 0

3.2 3

1 29298 565 80 18.9 18 30 29297 563 103 29.4 27 34 29289 56L 120 33.7 32 55 29295 55Y 163 56.8 43 75 29290 55B 202 68.1 61 88 29294 54M 250 72.5 66 100 29295 54T 400 75.0 73 100 HAZ:

29301 5AT

-80 36.0 29 38 29305 5AY

-50 36.6 30 41 29303 5AK 0

44.0 41 22 29306 5AU 48 98.3*

78 30 29302 57P 80 77.6 67 82 29300 575 120 74.1 69 100 29299 5AB 202 102.2 82 100 29304 5A6 400 112.6 95 100 Note: HAZ data exhibits scatter in fracture energy due to material inhomogeneity and location of notch in relation to the fusion line. Because of these effects, the latest version of ASTM E185-94 recommends not testing the HAZ samples. This version of ASTM E 185 has not yet been approved for use by NRC.

36

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 5-3: SIGNIFICANT RESULTS OF IRRADIATED AND UNIRRADIATED CHARPY V-NOTCH DATA atnlIndex

Temp, Index Temp Index TemplF USE Eý730 ft-b F-750 ftlb PLATE: Heat C3278-2 Unirradiated

-21.83 7.91 8.6 133.8 1 st Capsule a

-25.14 16.07

-19.0 133.3 Difference Unirrad. to 1st

-3.31 8.16

-27.6

-0.5 (-0.37%)

2nd Capsule b

-6.86 22.5 1 0.1 121.5 Difference Unirrad. to 2nd 14.97 14.59 1.5 1 -12.3 (-9.2%)

1 st Capsule" 2nd Capsuleb Difference Reg. Guide -1.99, Rev. 2 ARTvDT:

I 5F 220F C 7 OF Reg. Guide 1.99, Rev, 2 (A+/-2o) :

-A 90 F to 490 F

-120F to 560F C Reg. Guide 1.99, Rev. 2 Decrease in USE d:

9%

10%

1%

Ind.x Temp Index Temp Index Temp (iF)

USE.

Mtra(`)(,IF)

ML-E=35 mil (ft-lby:i)

......................= 3 0 ft-lb

.. E 5........ '

SURVEILLANCE WELD: '

Ist Capsule a 44.4 95.7 54.0 85.1 2nd Capsule b 107.7 147,9 130.7 74.8 Difference Ist to 2nd 63.3 52.2 76.7

-10.3(12.1%)

Reg. Guide 1.99, Rev. 2 ARTDT:

Reg. Guide 1.99, Rev. 2 (A+/-2o):

Reg. Guide 1.99, Rev. 2 Decrease in USE d:

1st Capsulea 420F

-14OF to 98°F 199%

2nd Capsuleb 61OF c 5°F to II7 °F c 22%

Difference 190F 3%

a 1st Capsule pulled from 300 at 5.98 EFPY or2.6xl0 n/cm b 2nd Capsule pulled from 120' at 13.4 EFPY or 5.0xl0. 7n/cm2 c Determined in Section 5.4.1 d Determined in Section 5.4.2 e No Unirradiated data available 37

GE Nuclear Energy GE-NE-B1100732-01, Revision 1 150 140 130 120 110 t00 90 80 70 60 50 40 30 20 10 0

-100

-50 0

50 100 150 200 250 300 350 400 450 Temperature (Deg F)

FIGURE 5-1: ABSORBED ENERGY VS. TEMPERATURE (PLATE-1200 CAPSULE) 38

IiI I

I i

I I

i 1

1.

1 I

I I

I GE Nuclear Energy GE-NE-B1100732-01, Revision I (d2 E

0 100 90 80 70 60 50 40 30 20 10 0

-100

-50 0

50 100 150 200 250 300 350 400 Temperature (Deg F)

FIGURE 5-2: LATERAL EXPANSION VS. TEMPERATURE (PLATE-120' CAPSULE) 450 39

GE Nuclear Energy GE-NE-B1100732-01, Revision 1 1 00 90 90 70

.0 a

60 50 40 30 20 10

-100

-50 0

50 100 150 200 250 300 350 400 Temperature (Deg F) 450 FIGURE 5-3: ABSORBED ENERGY VS. TEMPERATURE (WELD-120' CAPSULE) 40

1 1

1 1

1 1

i I

I I

I I

I GE Nuclear Energy GE-NE-B1100732-01, Revision I 0

C' 100 90 80 70 60 50 40 30 20 10 0

-100

-50 0

50 100 150 200 250 300 350 400 Temperature (Deg F) 450 FIGURE 5-4: LATERAL EXPANSION VS. TEMPERATURE (WELD-1200 CAPSULE) 41

GE Nuclear Energy GE-NE-B1100732-01, Revision 1 150

-i 140)1.1 1-.

120

.... i...---

110 0

]2o

[..........

100

b.

i 90.

I

-l 70 P-.

6 0..

50 o 0120 Deg Capsule Data I I

40 Points 30 20

'I l()

--4,

-100

-50 0

50 100 150 200 250 300 350 400 450 Temperature (Deg F)

FIGURE 5-5: ABSORBED ENERGY VS. TEMPERATURE (HAZ-120° CAPSULE) 42

I I

I I

I I

IiI I

I iI I

ii GE Nuclear Energy GE-NE-B1100732-01, Revision 1 i

0 U,

100 90 80 70 60 50 40 30 20 0

-100

-50 0

50 100 150 200 250 300 350 400 Temperature (Deg F) 450 FIGURE 5-6: LATERAL EXPANSION VS. TEMPERATURE (HAZ-120' CAPSULE) 43

GE Nuclear Energy GE-NE-B1100732-01, Revision I 150 140 130 120 110 100 90 80 70 60 50 40 30 20 10 0

-100

-50 0

50 100 150 200 250 300 350 400 Temperature (Deg F) 450 FIGURE 5-7: COMPARISON OF UNIRRADIATED AND IRRADIATED ENERGY DATA (PLATE) 44

IiI I

iii I

I f

I I

I I

II GE Nuclear Energy GE-NE-B1100732-01, Revision f U

100 90 80 70 60 50 40 30 20 10 0

-100

-50 0

50 100 150 200 250 300 350 400 450 Temperature (Deg F)

FIGURE 5-8: COMPARISON OF 1ST AND 2ND CAPSULE ENERGY RESULTS (WELD) 45

GE Nuclear Energy GE-NE-BI100732-01, Revision 1 U

140 140 130 120 110 100 90 80 70 60 50 40 30 20 10 0;

-100 C

I..

U 0

II

....u.

0 a

0 T

.U30 Deg Cap Data So 120 Deg Cap Data[

300 350 400 450 aI

-1~

-50 0

50 100 150 200 Temperature (Deg F) 250 FIGURE 5-9: COMPARISON OF IST AND 2ND CAPSULE ENERGY RESULTS (IIAZ) 46

I N I

GE Nuclear Energy G -I 1

I I

1 I I.

I I

i I

GE-NE-B "1100 732-01, Revision 1 00 0) 0 0

0 100 90 80 70 60 50 40 30 20

-200

-150

-100 50 0

50 1

U I

-Baseline

-...... 30 Deg Capsule a

30Deg Data 120 Deg Capsule o

120DegData j

00 150 200 250 300 350 400 450 Temperature (Deg F)

FIGURE 5-10: COMPARISON OF LATERAL EXPANSION RESULTS (PLATE) 47

GE Nuclear Energy GE-NE-B1 !00732-01, Revision 1 100 90 80 70 60 50

,, 50 40

~30 20 10

-200

-150

-100

-50 0

50 100 150 Temperature (Deg F) 200 250 300 350 400 450 FIGURE 5-11: COMPARISON OF LATERAL EXPANSION RESULTS (WELD) 48

I I

I I

I I

I I

1 1

I i

I I

I GE Nuclear Energy GE-NE-B! 100732-01, Revision I 5

0 100 90 80.

70 60 50 40 30 20 10 00

-100

-50 0

50 100 150 200 250 300 350 400 Temperature (Deg F) 450 FIGURE 5-12: COMPARISON OF LATERAL EXPANSION RESULTS (HAZ) 49

GE Nuclear Energy GE-NE-B1100732-01, Revision 1

,0 150 140 130 120 110 100 90 80 70 60 50 40 30 20 10 0

-100

-50 0

50 100 150 200 250 300 350 400 450 Temperature (Deg F)

FIGURE 5-13: TANH CURVE-FITTED RESULTS FOR COMBINED BASELINE DATA (PLATE) 50

II I

GE Nuclear Energy 5OO 400 3O

<200-I I

I i

i I

I I

I I

R ei I

GE-NE-81 100732-01, Revision I I

3 10E-17 EiM-18 10-E-19 104E-20 Fluence (n/cm2, E > 1 MeV)

FIGURE 5-14: AT30 VS. FLUENCE SHOWING PLATE DATA WITH FITTED RESULTS 51

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1

6. TENSILE TESTING Eight round bar tensile specimens were recovered from the surveillance capsule.

Uniaxial tensile tests were conducted in air at room temperature (70'F) at RPV operating temperature (550'F) for base, weld and HAZ specimens, and at an intermediate temperature of 185°F for an additional base and weld specimen. The tests were conducted in accordance with ASTM E8-89 [14].

6.1 PROCEDURE All tests were conducted using a screw-driven Instron test frame equipped with a 20-kip load cell and special pull bars and grips. Heating was done with a Satec resistance clamshell furnace centered around the specimen load train. The test temperature was monitored by a chromel-alumel thermocouple spot-welded to an Inconel clip that was friction-clipped to the surface of the specimen at its midline.

All tests were conducted at a calibrated crosshead speed of 0.005 in/min until well past yield, at which time the speed was increased to 0.05 inch/min until fracture.

Crosshead displacement was used to monitor specimen extension during the test.

The test specimens were machined with a minimum nominal 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 measured area into the 0.2% offset load and into the maximum test load, respectively. The values listed for the uniform and total elongation were obtained from plots that recorded load versus specimen extension and are based on a 1.5 inch nominal gage length. Reduction of area (RA) values were determined from post-test measurements of the necked specimen diameters using a calibrated blade micrometer and employing the following formula:

RA = 100% * (Ao - Af)/A0 (6-1)

After testing, each broken specimen was photographed end-on, showing the fracture surface, and lengthwise, showing the fracture location and local necking behavior.

52

GE Nuclear Energy GE-NE-B 1100732-01 Revision I 6.2 RESUL TS 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.

A stress-strain curve for a 550'F base metal irradiated specimen is shown in Figure 6-1.

This curve is typical of the stress-strain characteristics of all the tested specimens. Photographs of the necking behavior and fracture surfaces are given in Figures 6-2 through 6-4, and Figures 6-5 through 6-7, respectively.

6.3 IRRADIATED VERSUS UNIRRADIATED TENSILE PROPERTIES Only unirradiated room temperature tensile test data for the base metal was available for comparison.

The data from the first surveillance capsule is also shown. No trend could be identified from the data (see Tables 6-2 and 6-3). The conclusion is that the material properties, especially ductility, have not been significantly degraded.

53

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 6-1: TENSILE TEST RESULTS FOR IRRADIATED RPV MATERIALS Test ViekIa ultimite uniform TotAl

{ Reduction Specimen Temp.

Strength Strength Elogation Elongation of Area iumber

(,)s) ks

(%)

(119 Base:

5CI RT 72.5 94.7 13.4 23.9 71.0 5C7 185 69.6 89.5 11.6 20.3 71.9 5CJ 550 66.7 88.1 10.7 18.2 69.3 Weld:

5D3 RT 92.6 107.8 12.9 21.7 65.7 5134 185 87.9 104.7 12.2 19.8 62.0 5DD 550 83.7 100.4 10.9 16.4 51.9 HAZ:

5EA RT 81.9 104.6 10.2 17.8 67.8 SEE 550 73.9 94.7 10.7 17.4 63.4 a Yield Strength is determined by 0.2% offset.

TABLE 6-2: COMPARISON OF UNIRRADIATED AND IRRADIATED TENSILE PROPERTIES AT ROOM TEMPERATURE Yield Strength.

UltimateStrnt Total Elongation Reduction of Area (ksi)

.(ksi) fA(%

Base:

Unirradiatedb 67.7 89.3 27.0 69.8 1 st Capsule 71.4 93.6 20.6 68.7 2nd Capsule 66.7 88.1 18.2 69.3 Weld 1 st Capsule 88.6 105.0 18.5 64.5 2nd Capsule 92.6 107.8 21.7 65.7 HAZ:

I st Capsule 77.2 99.6 17.3 68.2 2nd Capsule 81.9 104.6 17.8 67.8 b Values taken as average of data in the material certification reports.

54

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 6-3: COMPARISON OF IRRADIATED TENSILE PROPERTIES AT 185OF Yield Strength.

Ultimate Strength Total Elongation Reduction of Area:

(ksi)

(ksi)

(%)

(%)

Base:

Ist Capsule 68.5 88.7 20.7 72.5 2nd Capsule 69.6 89.5 20.3 71.9 HAZ:

Ist Capsule 73,2 94.0 16.3 69A.

Weld:

2nd Capsule 87.9 104.7 19.8 62.0 TABLE 6-4: COMPARISON OF IRRADIATED TENSILE PROPERTIES AT 5501F Yield Strength..

Ultimate Strength Total Elongation I

eduction of Area

.(ksi)

(ksi)::.

(% )

Base:

Ist Capsule 65.1 89A1 17.4 A 65-9 2nd Capsule 66.7 88.1 18.2 69.3 Weld I1st Capsule 76.3 96.2 14.4 44.7 2nd Capsule 83.7 100.4 16.4 51.9 HAZ:

I st Capsule 74.4 98.0 13.9 54.8 2nd Capsule 73.9 94.7 17.4 63.4

GE Nuclear Energy GE-NE-B 1100732-01 Revision I ca (0,

Lu C

2 4

0 8

10 12 14 i

18 30 STRAIN (%)

FIGURE 6-1. TYPICAL ENGINEERING STRESS-STRAIN FOR IRRADIATED RPV MATERIALS 56

GE Nuclear Energy GE-NE-B 1100732-01 Revision I 5C1 RT 5C0 550°F 5C7 185"F FIGUE FIGURE 6-2: FRACTURE LOCATIONAkND NECKING BEHAVIOR FOR IRRADIATED BASE METAL TENSILE SPECIMENS 57

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 5D3 RT 5DD 550:F 5D4 185 0F FIGURE 6-3: FRACTURE LOCATION AND NECKING BEHAVIOR FOR IRRADIATED WELD. METAL TENSILE SPECIMENS 58

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 5EA RTT_

SEE 550"F FIG-ERE 6-4: FRACTURE LOCATION AND NECKLNG BEHLAVIOR FOR IRRADIATED HAZ TENSILE SPECIMENS

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 S!*-,*.==,i 5C1 RT 5CJ 550°F 5Q7 1850.

FIGURE 6-5: FRACTURE APPEARANCE FOR IRRADIATED BASE METAL TENSILE SPECLMENS 60

GE Nuclear Enercy 6N--, NE-' B 100732-01 Revision 1 5Q3 RT 5DD 550TF 5D4 1850F FIGUE 66:

FACTRE PPEAAN FORIRRD WLD ETA FIGURE 6-6: FRACTURE-APPEARANCE FOR IRRADIATED WELD METAL TENSILE SPECIMENS 61

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 SEA RT0F SE.E 550OF FIGURE 6-7: FRACTURE APPEARANCE FOR IRRADIATED HLAZ TENSILE SPECIMENS 62 f

GE Nuclear Energy GE-NE-B1 100732-01 Revision 1

7. ADJUSTED REFERENCE TEMPERATURE AND UPPER SHELF ENERGY 18 2

The 32 EFPY peak fluence value of 1.81xl0 n/cm defined in Section 4.3 is used to calculate the 32 EFPY 1/4 Tpeak fluence value of 1.3 xWO18 n/cm. The 32 EFPY 1/4 T fluence is used in this section to calculate adjusted reference temperatures (ARTs) and upper shelf energy (USE) decrease for the beltline materials.

7.1 ADJUSTED REFERENCE TEMPERA TURE AT 32 EFPY The effect on adjusted reference temperature (ART) due to irradiation in the beltline materials is determined according to the methods in Rev. 2 [7], as a function of neutron fluence and the element contents of copper (Cu) and nickel (Ni). The specific relationship from Rev. 2

[7] is:

ART = Initial RTNDT + ARTNDT + Margin (7-1) where:

ARTNDT= CF f(0.28 - 0.1 10f (7-2)

Margin= 2

+a2 (7-3)

CF chemistry factor from Tables I or 2 of Rev. 2 [7],

2 19 f =

114 T fluence (n/cm ) divided by 10

,=

standard deviation on initial RTND-, which is taken to be 0°F.

96=

standard deviation on ARTNTD, 28°F for welds and 17'F for base material, except that a. need not exceed 0.50 times the ART*,-D value. If 2 or more sets of credible surveillance data are used, a. is 1/2 the above values.

63

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The ART values are calculated based upon chemistry data as described in Table 7.1.

Once two sets of surveillance capsule data are available, the CF values per Rev. 2 [7] can be adjusted to reflect the results. The method is described in Position 2.1 of Rev. 2 [7], and is summarized below.

7.2 SURVEILLANCE CF ADJUSTMENT The surveillance CF adjustment is based on a least squares fit of the surveillance data to the ARTNDT Equation 7-2, restated as:

ARTNDT = CF*FF where FF is the fluence factor shown in (7-2).

The least squares approach uses the actual shifts of the 30' and 120' capsule Charpy specimens, combined with the fluence factors applicable to those capsule fluences.

CF (Sh' 30o

• FF3 0 + Shf 1 20 - *FFJ2oo)

(FFo2 + FF120o2)

The values for Equation 7-4 are in Section 5.4.1 and Table 5-3:

The chemistry for the surveillance plate is 0.11% Cu and 0.60% Ni, which has a Chemistry Factor from Table 2 of Rev. 2 (Rev. 2 CF) of 74.0 (Section 5.4.1).

Location FF Plate Shift CF per Rev. 2 300 2.6 x 10'7 0.205

-3.31 1200 5.0 x 1017 0.293 14.97 74.0 (For the weld material, no unirradiated data was available; therefore the CF could not be adjusted. The ART was calculated for the weld in accordance with Position I of Rev. 2 [7].)

Substituting these values into Equation 7-4 gives a CF of 29.4 based on the surveillance data. The surveillance CF is compared to the Rev. 2 CF to establish the adjustment of generic Rev. 2 predictions to actual plant conditions:

Plate adjustment = 29.4/74.0 =Q.4Q 64

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 7.3 APPLICATION OF CF ADJUSTMENT TO BEL TLINE MATERIALS The assumption made in applying the CF adjustments to the beltline materials is that the plant-specific conditions that affected the surveillance material shifts also affect the beitline material shifts. This is the same assumption made in Rev. 2, Position 2.1, for the case where the vessel weld chemistry differs significantly from the surveillance weld chemistry. Position 2.1 recommends that, when chemistries differ, the measured surveillance shifts be adjusted by the ratios of the belfine and surveillance material CFs, and then the least squares calculation be done to determine the adjusted beitline CF. The same basic approach is followed below.

In Position 2.1 of Rev. 2, it appears that the CF ratio approach is intended for the case where the beltline and surveillance base material and welds are the same heat, but chemistry results are significantly different. Here, in applying the CF ratio approach it is assumed that the surveillance CF adjustments should be applied to other beltline heats, as well as to the same belfine heats. This assumption and the recommendation in Position 2.1 have the same basis; that is that the CFs for different chemistries in the Rev. 2 tables are correct relative to one another.

The result is that Equation 7-2 from Rev. 2 is multiplied by the surveillance adjustmenrt (SA):

ARTNoT = CF* FF*SA (7-5)

Implicit in this approach is the assumption that there is a variable other than chemistry or fluence that is affecting the ARTNDT. This assumption is feasible when evaluating a BWR, because nearly all of the data base used to develop Rev. 2 was PWR data, taken at significantly higher fluxes and fluences than are typical for BWRs. Fluxes may vary by a factor of 2 to 100, depending on the specific BWR and PWR compared. Fluence differences are large as well.

There is also a temperature difference between PWR and BWR surveillance capsule irradiation conditions; BWR irradiation temperatures are 525"F to 5350F, 15°F to 350 F lower than PWR irradiation temperatures. One, or a combination of these variables may account for the quantity SA.

The ARTTrDT values for the beltline plate materials are calculated using Equation 7-5.

The margin terms are taken as half the normal values, as permitted by Position 2.1 of Rev. 2.

65

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 7.4 ART VS. EFPY Each beltline plate and weld ARTvT value is determined by multiplying the CF from Rev. 2 determined for the Cu-Ni content of the material, by the fluence factor for the EFPY being evaluated and the surveillance adjustment, if appropriate. The Initial RTNDT, ARTNDT and Margin are added to obtain the ART of the material. The 32 EFPY ART values for all of the beltline plates and several of the most limiting beltline welds are shown in Table 7-1. The ART for the limiting beltline material, Longitudinal Weld Heat 27204/12008, at 32 EFPY is 109'F. The ART for the limiting beltline plate, heat number C3376-2, at 32 EFPY is 56 'F.

The ART vs. EFPY curve for the limiting beltline weld and plate materials is shown in Figure 7-1.

7.5 UPPER SHELF ENERGY AT 32 EFPY Unirradiated Upper Shelf data were not available for all of the material heats. Due to the lack of specific pre-operational USE data, FitzPatrick has been evaluated to verify that the BWR Owners! Group Equivalent Margin Analyses are applicable. The calculations in Tables 7-2 and 7-3 show that the equivalent margin analyses are applicable.

The Equivalent Margin Analyses demonstrate that the 10 CFR 50, Appendix G safety requirements are satisfactorily met for FitzPatrick. The Owners' Group Program Report [ 16] was submitted to the NRC in December 1993 and approved by SER on December 8, 1993.

66

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I GE Nuclear Energy GE-NE-B1 100732-01, Revision I TABLE 7-1:32 EFPY ART VALUES BEITLINF ART VALUES FOR FITZPATRICK Lower Intermediate 32 EFPY Peak I.D. fluency =

32 EFPY Peak 1/4 T fluence =

Lower Intermediate Thickness -

5.375 inches 1.81E+18 n/cmA2 1,31E+18 n/cm^2 a 1.61E+18 n/cm^2' 1.17E+18 n/cm^2

• I I

n.4-l.n/I A-)

Lower Weld Thickness Plait Thickness-Lower 5.375 inches (Girth) 6.375 inches 32 EFPY Peak I.D. fluence 32 EFPY Ptak 1/4 T weld fluence S..

I

.P.

n Weld Initial 32 EFPY 17 CFA 32 EFPY 32 EFPY COMPONENT Type HEAT OR HEAT/IO1

%Cu

%Ni CF Adjusted RTndt A RTndt Margin Shift ART CF OF oF F

1F OF PLATES:

Lower G-3415-1R C3394-1 0,11 0,56 73.6 29.2

-10 12.8 0.0 6.4b 12.8 26 16 0-3415-3 C3376-2 0.13 0.60 91 36.2 24 15.8 0.0 7.9' 15,8 32 56 0-3415-2 C3103-2 0.14 0.57 99 39.3

-2 17.2 0.0 8.5' 17.0 34 32 Lower-Interm.

G-3413-7 C3368-1 0.12 0.50 81 32,2

-10 15.2 0.0 7.6' 15.2 30 20 G-3414-2 C3278-2 0.11 0,60 74 29.4

-10 13.9 0.0 6,95, 13.9 28 18 G-3414-1 C3301-1 0.18 0.57 131 52.1

-18 24.7 0.0 8.5b 17.0 42 24 WELDS:

Lower Long.

2-233 27204/12008 Flux 3774 0.219 0.996 231 231

-48 101 0.0 28.0 56 157 109 Lower Interined.

1-233 13253/12008 Flux 3774 0.210 0.873 208.7 208.7

-50 99 0.0 28.0 56 155 105

Long, Girth 1-240 305414 Flux 3947 0,337 0.609 209.1 209.1

-50 91A 0,0 28.0 56 150 100 a Includes effects of 105% power uprate b Reduced a, based on use of credible data 67

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 7-2: PLATE EQUIVALENT MARGIN ANALYSIS PLANT APPLICABILITY VERIFICATION FORM FOR FITZPATRICK - BWR 4/MK I - Including Uprated Power Condition BWR/3-6 PLATE Surveillance Plate USE:

%Cu = 0Q.1 1 st Capsule Fluence = 2,6.21 2nd Capsule Fluence = 5.k. 1QULlm 2

Unirradiated to 1 st Capsule Measured % Decrease =3lZ (Charpy Curves)

Unirradiated to 2nd Capsule Measured % Decrease =9.2 (Charpy Curves) 1 st Rev. 2 Predicted % Decrease = 9 (Rev. 2, Figure 2) 2nd Rev. 2 Predicted % Decrease =1. (Rev. 2, Figure 2)

Limiting Beltline Plate USE:

%Cu = 0.18 32 EFPY 1/4 T Fluence = 1,31 X 1

_n-2 Rev. 2 Predicted % Decrease = 1_ (Rev. 2, Figure 2)

Adjusted % Decrease = N/A (Rev. 2, Position 2.2)

TI I

I 18 % ý< 21%, so vessel plates are bounded by equivalent margin analysis j...

II 68

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 7-3: WELD EQUIVALENT MARGIN ANALYSIS PLANT APPLICABILITY VERIFICATION FORM FOR FITZPATRICK - BWR 4/MK I - Including Uprated Power Condition BWRM-6 WELD Surveillance Weld USE:

%Cu = 0.29 1st Capsule Fluence = 2.6 x cI1Bmr 2

2nd Capsule Fluence = 5.0x 'nn/m Unirradiated to 1 st or 2nd Capsule Measured % Decrease =Unknown 1 st to 2nd Capsule Measured % Decrease = 1 (Charpy Curves) 1st Rev 2 Predicted % Decrease 19 (Rev. 2, Figure 2) 2nd Rev 2 Predicted % Decrease =22 (Rev. 2, Figure 2)

Limitin2 Beltline Weld USE:

%Cu = 0.33 32 EFPY 1/4 T Fluence = 1.31 x 10n/cm2 Rev 2 Predicted % Decrease = 29 (Rev. 2, Figure 2)

Adjusted % Decrease = N/A (Rev. 2, Position 2.2)

S29%

< 34%, so vessel welds are bounded by equivalent margin analysis Note: the limiting beltline weld case (0.33 wt/o Cu @ 1.31 x I018 n/cm2) is not physically possible. However, it represents a worst case condition 69

GE Nuclear Energy GE-NE-B 1100732-01, Revision I 120

100, 8 0, 60 40--

Lower Plate G-3415-3 o

-(0.

13 wt%

Cu, 0.60 wt%

Ni) 20-0-

-20

-40

-60 i 1 i

1 0

5 10 15 20 25 30 35 EFFECTIVE FULL POWER YEARS FIGURE 7-1. ADJUSTED REFERENCE TEMPERATURE VS. EFPY FOR LIMITING BELTLINE PLATE AND WELD 69b I

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GE Nuclear Energy GE-NE-B 1100732-01 Revision I

8. PRESSURE-TEMPERATURE CURVES

8.1 BACKGROUND

Nuclear Regulatory Commission (NRC) IOCFR50 Appendix G [1] specifies fracture toughness requirements to provide adequate margins of safety during operation to which the pressure-retaining component pressure boundary may be subjected over its service lifetime. The ASME Code (Appendix G of Section XM of the ASME Code) forms the basis for the requirements of I OCFR50 Appendix G. The limits for pressure and temperature are required by 10CFR50 Appendix G for three categories of operation: (a) hydrostatic pressure tests and leak tests, (b) core not critical heatup/cooldown, and (c) core critical operation. The condition that results in the highest temperature for the limiting material determines the minimum temperature requirement for the vessel.

In all cases, the applicable temperature is the greater of the IOCRF50 minimum temperature requirement and the ASME Appendix G limits. A summary of the requirements is as follows:

Operating Condition and Pressure I. Hydrostatic Pressure Test & Leak Test (Core is Not Critical) - Curve A

1. At < 20% of preservice hydrotest pressure

2. At > 20% of preservice hydrotest pressure Minimum Temperature Requirement. *F Larger of ASME Limits or highest of closure flange region initial RTNDT + 60°F*

Larger of ASME Limits or highest of closure flange region initial RTNDT + 90°F

• 60'F adder is included by GE as an additional conservatism as described in Section 8.3 70

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 Minimum Temperature Requirement. 0F Operating Condition and Pressure, II. Normal operation (heat-up and cool-down),

including anticipated operational occurrences A. Core not critical - Curve B

1. At < 20% of preservice hydrotest pressure

2. At > 20% of preservice hydrotest pressure B.

Core critical - Curve C I. At < 20% of preservice hydrotest pressure with the water level within the normal range for power operation.

2. At > 20% of preservice hydrotest pressure Larger of ASME Limits or highest of closure flange region initial RTNDT + 600 F*

Larger of ASME Limits or highest of closure flange region initial RTNrT + 120TF Larger of ASME Limits + 40°F or A. I Larger of ASME Limits + 40'F or A.2 + 40TF or the minimum permissible temperature for the inservice system hydrostatic pressure test

• 60'F adder is included by GE as an additional conservatism as described in Section 8.3 Note: The core critical operation curve is identical to the core not critical heatup/cooldown curve but shifted by 40'F, as required in IOCFR50, Appendix G [1]. Hence, the methods used for determining the core not critical heatup/cooldown curves apply to the core critical curves, as well.

There are three vessel regions that affect the operating limits: the closure flange region, the core beltline region, and the remainder of the vessel, or non-beltline regions. The closure flange region limits are controlling at lower pressures primarily because of I OCFR50, Appendix G requirements. The non-beltline and beltline region operating limits are evaluated according to procedures in IOCFR50, Appendix G [1], ASME Boiler and Pressure Vessel Code, Section XU, Appendix G [2], and Welding Research Council (WRC) Bulletin 175 [10]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

The P-T curves for the non-beltline region were conservatively developed for a large BWRi6 (nominal inside diameter of 251 inches).

The analysis is considered appropriate for FitzPatrick as the FitzPatrick specific values are bounded by this generic analysis. The generic 71

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 value was adapted to the conditions at FitzPatrick by using the specific RTN.r values for the FitzPatrick reactor pressure vessel (RPV). The presence of nozzles and control rod (CRD) penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.

This was the result of the stress concentrations and higher thermal stresses for certain transient conditions, experienced by the upper vessel and the bottom head.

P-T curves are provided for 32 EFPY. The 32 EFPY curves are effective through the end-of-life (EOL).

The 32 EFPY curves are provided in Figures 8-1, 8-2 and 8-3. The corresponding numerical values for the curves are given in Table 8-1. The P-T curves for 24 EFPY are shown in Appendix B, Figures B-1 through B-3.

Under certain conditions, the minimum bottom head temperature can be significantly cooler than the beltline or closure flange region.

These conditions can occur when the recirculation pumps are operating at low speed (or off), and during water injection through the control rod drives. To account for these circumstances, individual temperature limits for the bottom head were established.

8.2 P-T CURVE METHODOLOGY 8.2.1 Non-Beltline Reaions Non-beltline regions are defined as the vessel locations that are remote from the active fuel such that the neutron fluence is not sufficient to cause any shift of RTNT.

Non-beltline components include the nozzles, the closure flanges, some shell plates, the top and bottom head plates and the control rod drive (CRD) penetrations. Detailed stress analyses of the non-beltline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The analyses took into account all mechanical loading and anticipated thermal transients.

Transients considered included 1000F/hr startup and shutdown, SCRAM, loss of feedwater heaters or flow, loss of recirculation pump flow, and all transients involving emergency core cooling injections. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [21 to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the two most limiting BWR/6 components; the feedwater nozzle and the CRD penetration (bottom head). All other components in the non-beltline regions are categorized under one of these two components.

72

GE Nuclear Energy GE-NE-B 1100732-01 Revision I The non-beltline curves are based on the most limiting (conservative) properties of either the upper vessel region or the bottom head. The non-beltline curves are shifted based on the most limiting initial RTNDT values for the appropriate non-beltline components; the initial RTNDT values are listed in Table 3-2. For the case of Curve A, pressure test for the non-beltline region, the recirculation inlet nozzle (4N2) is the limiting case. Curve B, for core not critical heatup/cooldown of the non-beltline region is also based on the recirculation inlet nozzle being the limiting case.

For pressures below 20% of preservice hydrostatic test pressure (312 psig) and with full bolt preload, the closure flange region metal temperature is required to be at RTDT or greater as described in Section 8.3. At low pressure, the ASME Code [2] allows the beltline and bottom head regions to experience even lower metal temperatures than the flange region RTNDT.

However, temperatures should not be permitted to be lower than 687F for the reason discussed below.

The shutdown margin is calculated for a water temperature of 68°F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod frilly withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 68'F limit, further extensive calculations would be required to justify a lower temperature. However, the boltup temperature as described in Section 8.3 is at 901F. Because the water temperature is currently limited to a minimum of 90 0F, the metal temperature should not fall below this limit while fuel is in the vessel. The 90'F limit applies when the head is on and tensioned, and also, when the head is off. (When fuel has been removed from the vessel, the head is tensioned, and the pressure is below 20 psig, the limiting vessel temperature is equal to the limiting RTNDT of the vessel materials. This limiting RTNDT is 30'F. When the head is not tensioned and fuel is not in the vessel, the requirements of IOCFR50 Appendix G [I] do not apply, and there are no limits on the vessel temperatures.

8.2.2 Pressure Test - Non-Beltline. Curve A aJsin2 Bottom Head)

In the finite element analysis, the BWR/6 CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, KI. The results of that computation were KI = 154.3 ksi-in1/2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T -RTNDT) was 161 °F.

73

GE Nuclear Energy GE-NE-B 1100732-01 Revision I The method to solve for (T - RTNTJT) for a specific KI is based on the curve in Figure G-2210-1 in ASME Appendix G [2):

(T - RT,,ur) = In [K] -26.78)/1.223] / 0.0145 - 160 (T - RTNDT) = In [(154.3 - 26.78) / 1.223] / 0.0145 - 160 (T-RTNDT) = 161°F The generic curve was generated by scaling 154.3 ksi-in"2 by the nominal pressures and calculating the associated (T - RTNDT):

CRD Penetration KI and (T - RTNDT) as a Function of Pressure Nominal Pressure Kr T - RTNDT (pg)

(ksi-m (0F) 1563 154.3 161 1400 138.2 151 1200 118.5 138 1000 98.7 121 800 79.0 99 600 59.2 66 400 39.5 1

The highest RTNDT for the bottom head plates and welds is I OF, as shown in Table 3-2.

The generic curve is applied to the FitzPatrick bottom head by shifting the P vs. (T - RTDT) values above to reflect the RTNDT value of 10°F.

The P-T curve is dependent on the KI value calculated, which is proportional to the stress and the crack depth according to the relationship:

K1 occY.(nra) 1 2 (8-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, KI is 1/2 1/2 proportional to Rit. The generic curve value of R/t, based on the generic BWR/6 bottom head dimensions, is:

74

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 121 12 112_

Generic R / t

= 138.2 / 8

= 48.9 inch (8-2)

The FitzPatrick specific bottom head dimensions are R = 110.50 inches and t = 6.813 inches, resulting in:

1/2 112 I1'2 FitzPatrick specific R/t

= 110.50/6.813

= 42.3 inch (8-3) 1/2 Since the generic value of R/t is larger than that for FitzPatrick, the generic P-T curve is conservative when applied to the FitzPatrick bottom head.

8.2.3 Core Not Critical Heatup/Cooldown - Non-Beltline Curve B (Using Feedwater Nozzle/Upper Vessel Region)

The feedwater nozzle was selected to represent non-beItline components for fiacture toughness analysis because the thermal conditions are the most severe experienced in the vessel. In addition to the more severe pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.

Stresses are taken from finite element analysis done specifically for fracture toughness analysis purposes. Analyses were performed for all feedwater nozzle transients that involve rapid temperature changes. The most severe of these was normal operation with cold 40TF feedwater injection.

The non-beltline curves based on feedwater nozzle limits were calculated according to the methods for nozzles in Appendix 5 of the Welding Research Council (WRC)

Bulletin 175 [101.

The stress intensity factor for a nozzle flaw under primary stress conditions is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:

Klp = SF

• cy * (na) 1/2

• F(a/rl)

(8-4) where: SF is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(a/rn) is the shape correction factor.

75

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 Firifte element analysis of a nozzle comer flaw was performed to determine appropriate values of F(a/r, 1 ) for Equation 8-4.

These values are shown in Figure A5-1 of WRC Bulletin 175 [10].

The stresses used in Equation 8-4 were taken from BWR/6 design stress reports for the feedwater nozzle. The stresses considered are primary membrane, a,, and primary bending, rPb.

Secondary membrane, a, and secondary bending, asb stresses are included in the total K, by using ASME Appendix G [2) methods for secondary portion, KI,:

K's = M.- (a.- +2/3 - Usb)

(8-5)

In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [10). However, the correction was not applied to primary membrane stresses. Kip and KI, are added to obtain the total value of stress intensity factor, K1.

The safety factors applied to primary stresses were 1.3 for pressure test conditions and 1.6 for core not critical heatup/cooldown conditions.

Once K1 was calculated, the following relationship was used to determine (T - RTNDT).

The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.

(T - RTNDT) = In [K1 - 26.78) /1.223) / 0.0145 - 160 (8-6) 8.2.4 Example Core Not Critical Heatup/Cooldown Calculation for Feedwater NozzleUpper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the feedwater nozzle generic analysis, where feedwater injection of 40'F into the vessel while at operating conditions (551.40F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle comer stresses were obtained from finite element analysis. These stresses, and other inputs used in the generic calculations, are shown below:

~pm = 20.49 ksi OSm = 16.19 ksi Cys= 45.0 ksi t = 7.5 inch Gpb = 0.22 ksi CTb, = 19.04 ksi a = 1.88 inch r,,

6.94 inch 76

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 In this case, the total stress, 55.94 ksi, exceeds the yield stress ays, so the correction factor, R, is calculated according to the following equation:

R

[a*y - a,. + ((crt

- ay) 30))/ (esto

- up.)

(8-7)

For the stresses given, the Ratio, R 0.70. Therefore, all the stresses are adjusted by the factor 0.70, except for up.. The resulting stresses are:

CYpm = 20.49 ksi ac1 = 11.33 ksi apb = 0.15 ksi Usb = 13.33 ksi The value of Mm from Figure G-2214-1 [2], was based on a thickness of 7.5 inches, hence, t1/2 = 2.74. The stress to yield ratio, a/Iy, was conservatively assumed to be 1.0. The resulting value obtained was:

M =2.84 The value F(a/rn) is taken from Figure A5-1 of WRC Bulletin 175 for an a/r, of 0.27.

F(a/r,,) =1.6 KIp is calculated from Equation 8-4:

Kip = 1.6 - (20.49 + 0.15).(it 1.88)12.1.6 Kip = 128.4 ksi-in '2 Kt, is calculated from Equation 8-5:

KI, = 2.84 * (11.33+2/3.13.33)

KL, = 57.4 ksi-in112 The total KI is therefore 186 ksi-in The total KI is substituted into Equation 8-6 to solve for (T - RT*DT):

(T - RTNDT) = In[(186 - 26.78) 1.223) / 0.0145 - 160 (T - RT,**DT) = 176°F 77

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 The generic curve was generated by scaling the stresses used to determine the KI. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by temperature difference of the 40'F water injected into the hot reactor vessel nozzle. In the base case that yielded a KI value of 186 ksi-infrl, the pressure is 1050 psig and the hot reactor vessel temperature is 551.4°F.

Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (T~ation - 40) / (551.4 - 40). From the KI the associated (T - RTNrT) can be calculated:

Feedwater Nozzle KI and (T - RTN-T) as a Function of Pressure Nominal Pressure Saturation Temp.

K1 (T-RTNDT)

(psig)

(OF)

(ksi-in":)

(OF) 1563 604 226 191 1400 588 213 187 1200 557 198 181 1050 551 186 176 1000 546 182 174 800 520 166 167 600 489 146 156 400 448 115 135 The highest non-beltline RTNDT for the feedwater region component (nozzle #N2) at FitzPatrick is 30'F as shown in Table 3-2. The generic curve is applied to the FitzPatrick upper vessel by shifting the P vs. (T-RTNDT) values above to reflect the RTNDT value of 30'F.

8.2.5 Core BeItline Region The pressure-temperature (P-T) operating limits for the beltline region are determined according to the ASME Code. As the beltline fluence increases with the increase in operating life, the P-T curves shift to a higher temperature.

The stress intensity factors (KI), calculated for the beltline region according to ASME Code Appendix G procedures [2], were based on a combination of pressure and thermal stresses for a 1/4 T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of 78

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 a flat plate; values were calculated for lO0°F/hr thermal gradient. The shift value of the most limiting ART material was used to adjust the RTIMT values for the P-T limits.

8.2.6 Beltline Region - Pressure Test The methods of ASME Code Section III, Appendix G [2] are used to calculate the pressure test beltline limits. The vessel shell, with an inside radius (R) to minimum thickness (tM) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:

a, = PR/tm (8-8)

The stress intensity factor, KtM, is calculated using Figure G-2214-1 of the ASME Code, Appendix G [2], accounting for the proper ratio of stress to yield strength. Figure G-2214-1 was taken from Welding Research Council (WRC) Bulletin 175 [10], based on a 1/4 T radial flaw with a six-to-one aspect ratio (length of 1.5T). The flaw is oriented normal to the maximum stress direction, in this case a vertically oriented flaw. This orientation is used even in the case where the circumferential weld is the limiting beltline material, as traditionally required by the NRC in the past.

The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [2] for comparison with KrR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.

The relationship between K1r and temperature relative to reference temperature (T - RTNDT) is shown in Figure G-2210-1 of ASME Appendix G [2], represented by the relationship:

KIm

• SF = K1, 1.223 exp[0.0145 (T - RTNDT + 160)] +26.78 (8-9)

This relationship is derived in the Welding Research Council (WRC) Bulletin 175 [101 as the lower bound of all dynamic fracture toughness and crack arrest toughness data. This relationship provides values of pressure versus temperature (from KIr and (T - RTNDT), respectively).

For the pressure test curve, a stress intensity factor, Kh, is added for a heatup/cooldown rate of 200F/hr to consider operating conditions. For the core not critical and core critical condition curves, a stress intensity factor is added for a heatup/cool down rate of 1 00°F/ hr. The K1, calculation for a heatup/cooldown rate of lOOT/hr is described in Sections 8.2.8 and 8.2.9.

79

GE Nuclear Energy GE-NE-B 1100732-01 Revision I 8.2.7 Calculations for the Beltline Region - Pressure Test This sample calculation is for a pressure test pressure of 1128 psig for 32 EFPY. The following inputs were used in the beltline limit calculation:

Adjusted RTNDT (Based on ART values in Table 7-1)

Vessel Height, Bottom of Active Fuel Height, Vessel Radius (to inside of clad),

Vessel Thickness (without clad),

Beltline Material Yield Strength, Operating temperature at P, A =109 1F H= 825.2 inches B = 208.6 inches R= 110.375 inches t

5.375 inches a = 50 ksi T = (calculated) 'F Pressure is calculated to include hydrostatic pressure for a full vessel:

P = 1128 psi + (H - B)

• 0.0361 psi/inch = P psig

= 1128 + (825.2 - 208.6) - 0.0361 = 1150 psig Pressure stress:

= PR/t 1.150 9 110.375 / 5.375 = 23.62 ksi (8-10)

(8-11)

The factor Mm, (=2.23) depends on (a/cay) and t1/2 and is determined from Figure G-2214-1 of the ASME Code, Appendix G [2]. The stress intensity factor for the pressure stress is Kr& = Mm"' e. The stress intensity factor for the thermal stress, Kit, is calculated as described in Section 8.3.8 below except that the value of "G" is 20 'F/hr instead of 1 00 0F/hr.

Equation 8-9 can be rearranged, and 1.5. KIm substituted for Kir, to solve for (T - RTNDT).

Using ASME Appendix G, Fig. G-2210-1 [2], Kim = 52.67, and K1, = 1.76 for a 20°F/hr heatup/cooldown rate:

(T - RTNDT) = In[(1.5

• Klm + KIt - 26.78) / 1.223] / 0.0145 - 160

= ln[(1.5 ° 52.67 +1.76 - 26.78) / 1.223] / 0.0145-160

= 1010F (8-12) 80

GE Nuclear Energy GE-NE-B 1100732-01

/Revision 1

T can be calculated by adding the adjusted RT,4D:

T = 101 + 109 = 210F P = 1128 psig 8.2.8 Beltline Region - Core Not Critical Heatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Appendix G [2]:

Kk = 2.0 - Kim +Kit (8-13) where Klm is primary membrane K due to pressure and Kit is radial thermal gradient K due to heatup/cooldown.

The pressure stress intensity factor Kln is calculated by the method described above, the only difference being the larger safety factor applied. The thermal gradient stress intensity factor calculation is described below.

The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant temperature in heatup or cooldown conditions.

The stress intensity factor is computed by multiplying the coefficient M, from Figure G-2214-2 of ASME Appendix G [2] by the through-wall temperature gradient ATe, given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-3 of ASME Appendix G [2].

The relationship used to compute the through-wall AT, is based on one-dimensional heat conduction through an insulated flat plate:

a 2T(x,t) a a x2 =1/0 (aT(xt)/t)

(8-14) where T(x,t) is temperature of the plate at depth x and time t, and D3 is the thermal diffusivity.

The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that fT(xt)I 0t = dT(t)/dt = G, where G is the heatup/cooldown rate, normally I 00°F/hr. The differential equation is integrated over x for the following boundary conditions:

1.

Vessel inside surface (x = 0) temperature is the same as coolant temperature, T0.

2.

Vessel outside surface (x = C) is perfectly insulated; the thermal gradient dT/dx = 0.

GE Nuclear Energy GE-NE-B 1100732-01 Revision I The integrated solution results in the following relationship for wall temperature:

T = Gx 2/2P - GCx/P + To (8-15)

This equation is normalized to plot (T - To)/ATw versus x/C. The resulting through-wall gradient compares very closely with Figure G-2214-3 of ASME Appendix G [2].

Therefore, AT.

calculated from Equation 8-14 is used with the appropriate Mt of Figure G-2214-2 of ASME Appendix G [2] to compute K1 t for heatup and cooldown.

The M, relationships were derived in the Welding Research Council (WRC)

Bulletin 175 [10] for infinitely long cracks of 1/4 T and 1/8 T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important The stress generated by the thermal gradient is a bending stress that changes sign from one side of the plate to the other. In combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4 T location (inside surface flaw) and the 3/4 T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4 T is assumed to be tensile for both heatup and cooldown. This results in the conservative approach of applying the maximum tensile stress at the 1/4 T location. This approach is conservative because irradiation effects cause the allowable toughness, Kfr, at 1/4 T to be less than that at 3/4 T for a given metal temperature.

This conservatism of the approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

8.2.9 Calculations for the Beltline Region Core Not Critical Heatup/Cooldown This sample calculation is for a pressure of 1128 psi for 32 EFPY.

The core not critical heatup/cooldown curve at 1128 psig uses the same Kim as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational rather than test condition (which includes nuclear boiling) that necessitates a higher safety factor. In addition, there is a KI, term for the thermal stress. The additional inputs used to calculate Kit are:

Heatup/cool down rate, normally 100F/hr, G = 100 'F/hr Vessel thickness, including clad thickness, C = 0.474 ft (5.688 inches)

GE Nuclear Energy GE-NE-B 1100732-01 Revision I Thermal diffusivity at 550'F (most conservative value),

t3 = 0.354 fl2/ hr [16]

Equation 8-15 can be solved for the through-wall temperature (x--C), resulting in the absolute value of AT for heatup or cooldown of AT = GC/213 (8-t6)

= 100 a (0.474)2/(2.0.354) = 31.7 The analyzed case for thermal stress is a 1/4 T flaw depth with wall thickness of C. The corresponding value of MK (0.280) can be found from ASME Appendix G, Figure G-2214-2 [2].

Thus the thermal stress intensity factor, KY,= M=

K AT, can be calculated.

The pressure and thermal stress terms are substituted into Equation 8-9 to solve for (T-RTNoDT):

(T - RT-DT) =

ln[(2-Kb + Ki) - 26.78)/ 1.2231/0.0145 - 160 (8-17)

=

ln[(2

• 52.67 -- 8.9 - 26.78) 1.2231/0.0145 - 160

= 134 OF T can be calculated by adding the adjusted RTN-D:

T=134 +109 243 OF P=

28psig 8.3 CLOSURE FLANGE REGION 10CFR50 Appendix G [I] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNTr. In some cases, the results of analysis for other regions exceed these requirements and closure flange limits do not affect the shape of the P-T curves. However, some closure flange requirements do impact the curves.

The ASME Code [2] requirement for boltup was at qualification temperature (T3oL) plus 600F. Current ASME Code requirements state in Paragraph G-2222(c), that for application of full bolt preload and reactor pressure up to 20%/a of hydrostatic test pressure, the RPV metal temperature must be at RTN-r or greater. The approach used for FitzPatrick for the boltup 83

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 temperature must be at RTNDT or greater. The approach used for litzPatrick for the boltup temperature was based on a more conservative value of (RTNT + 60), or the LST of the bolting materials, whichever is greater. The 607F adder is included by GE for two resaons: 1) The pre-1971 requirements of ASME Code Section III, Subsection NA, Appendix G included the 60°F adder, and 2) Inclusion of the additional 60"F requirement above the RTNDT provides an additional assurance that a flaw size between 0.1 and 0.24 inches is acceptable. As shown in Table 3-2, the limiting initial RTNDT for the closure flange region was the upper shell plate material at 30°F and the LST of the closure studs was 70'F, however, an RTNDT + 601F will conservatively be used; therefore the boltup temperature value used was 90°F. This conservatism is appropriate because boltup is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

10CFR50 Appendix G, paragraph IV.A.2 [1] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT of the closure region.

Curve A temperature must be no less than (RTNDT + 90'F) and Curve B temperature no less than (RTNDT + 1207F). The Curve A requirement causes a 301F shift at 20%

hydrotest pressure of 312 psig. The Curve B shift at 312 psig is not visible in Figure 8-2, as the discontinuity curves are limiting.

8.4 CORE CRITICAL OPERATION REQUIREMENTS OF IOCFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of I OCFR50 Appendix G [1, Table 1]. Table I of [1] requires that core critical P-T limits be 40'F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure.

Curve B is more limiting than Curve A, so limiting Curve C values must be at least Curve B plus 40'F for pressures above 312 psig.

Table 1 of 10CFR50 Appendix G [1] indicates that for BWRs with water level within normal range for power operation, the allowed initial criticality at the closure flange region is (RTN-T + 60'F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 90°F, based on an RTNDT of 30'F. In addition, above 312 psig the Curve C temperature must be at least the greater of RT*DT of the closure region + 160°F or the temperature required for the hydrostatic pressure test (Curve A at 1128 psig). Therefore, this requirement causes Curve C to shift at 20% hydrostatic test pressure or 312 psig. This shift is visible in Figure 8-3.

84

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 a.

0

'U 0.

0:

7-J

'U CL 1400 1300 1200 1100 1000 900 800 700 600 500 400 300 I

BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 32 157 RATEF/COOLDOWN I INITIAL RTndt VALUES ARE

-48oF FOR BELTLINE, 30°F FOR UPPER VESSEL, AND 10°F FOR BOTTOM HEAD UPPER VESSEL AND BELTLINE LIMITS

- BOTTOM HEAD LIMITS 200 100 0

L 0

50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F)

FIGURE 8-1: PRESSURE TEST CURVE (CURVE A) 85

GE Nuclear Energy GE-NE-B 1100732-01 Revision I 0a.

0 uj c-0 z

I-U,.

U,,

Lu 1400 1300 1200 1100 1000 900 800 700 600 500 400 300 200 100 BELTINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ('F) 32 157 HEATUP/COOLDOWN SRATE 10*FHR INITIAL RTndt VALUES ARE

-48'F FOR BELTLINE, 30°F FOR UPPER VESSEL, AND 10OF FOR BOTTOM HEAD

.UPPER VESSEL AND BELTLINE LIMITS

- BOTTOM HEAD LIMITS 0

0 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F)

FIGURE 8-2: NON-NUCLEAR I-EATUP/COOLDOWN (CURVE B) 86

GE Nuclear Energy 1400 C-C 1300 CRITICI F(

FITZPA 1200 1100 GE-NE-B 1100732-01 Revision I 0

ca 0

I-w ILl INITIAL RTndt VALUES ARE

-48OF FOR BELTLINE, 30'F FOR UPPER VESSEL, AND 10°F FOR BOTTOM HEAD BELTINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ('F) 32 157 REATE10COOLORWN RATE 00-IHRj 1000 900 800 700 600 500 400 300 200 100 0

BELTLINE AND NON-BELTLINE LIMITS Minimum Criticality

[Temperature. 90'F!,J*.

0 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (@F)

FIGURE 8-3: CORE CRITICAL OPERATION (CURVE C) 87 I

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE 8 FitzPatrick P-T Curve Values for 32 EFPY Required Temperatures at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A FOR FIGURES 8-1 THROUGH 8-3 BOTTOM RPV &

PRESSURE HEAD 32 EFPY BELTLINE CURVE A CURVE A BOTTOM RPV &

HEAD 32 EFPY BELTLINE CURVE B CURVE B (PSIG) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 310 312.5 312.5 320 330 (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68&0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 (OF) 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 120.0 120.0 120.0 (OF) 68.0 68&0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 (OF) 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 92.7 97.5 101.9 106.1 110.1 113.6 116.8 119.8 122.8 125.6 128.2 130.6 132.9 135.2 137.4 139.4 141.4 143.3 145.1 147.0 148.7 150.3 152.0 152.3 152.3 153.5 155.1 RPV&

32 EFPY BELTLINE CURVE C (OF) 90.0 90.0 90.0 90.0 94.5 1052 113,9 121.1 127.4 132.7 137.5 141.9 146.1 150.1 153.6 156.8 159.8 162.8 165.6 168.2 "170.6 172.9 175.2 177.4 179.4 181.4 183.3 185.1 187.0 188.7 190.3 192.0 192.3 208.7 208.7 208.7 88

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 8-1. FitzPatrick P-T Curve Values for 32 EFPY Required Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 8-1 THROUGH 8-3 BOTTOM RPV &

PRESSURE HEAD 32 EFPY BELTLINE CURVE A CURVE A BOTTrOM RPV &

HEAD 32 EFPY BELTLINE CURVE B CURVE B (PSIG) 340 350 360 370 380 390 400 410 420 430 440 450 460 470 480 490 500 510 520 530 540 550 560 570 580 590 600 610 620 630 640 650 660 670 680 690 (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 69.5 71.8 74.0 76.1 78.2 80.2 82.1 84.0 85.9 87.7 89.4 91.1 92.8 (OF) 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 122.0 125.1 128.0 130.8 133.6 136.2 138.7 141.1 143.5 145.8 148.0 150.1 152.2 154.2 (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 70.3 73.2 76.1 78.8 81.5 84.0 86-5 88.8 91.1 93.3 95.5 97.6 99.6 101.5 103.5 105.3 107.1 108.9 110.6 112.3 113.9 115.5 117.1 118.6 120.1 121.6 123.0 (OF) 156.6 158.0 159.4 160.8 162.1 163.4 164.7 166.0 167.2 168.4 169.6 170.7 171.8 172.9 174.0 175.1 176.1 177.1 178.1 179.1 180.1 181.1 182.0 182.9 184.6 186.3 187.9 189.5 191.1 192.6 194.)

195.6 197.0 198.4 199.8 201.1 RPV &

32 EFPY BELTLINE CURVE C

("F) 208.7 208.7 208.7 208.7 208.7 208.7 208.7 208.7 208.7 208.7 209.6 210.7 211.8 212.9 214.0 215.1 216.1 217.1 218.1 219.1 220.1 221.1 222.0 222.9 224.6 226.3 227.9 229.5 231.1 232.6 234.1 235.6 237.0 238.4 239.8 241.1 89

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 8-1. FitzPatrick P-T Curve Values for 32 EFPY Required Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 8-1 THROUGH 8-3 BOTTOM RPV &

PRESSURE HEAD 32 EFPY BELTLINE CURVEA CURVE A BOTTOM RPV &

HEAD 32 EFPY BELTLINE CURVE B CURVE B (PSIG) 700 710 720 730 740 750 760 770 780 790 800 810 820 830 840 850 860 870 880 890 900 910 920 930 940 950 960 970 980 990 1000 1010 1020 1030 1040 1050 (fF) 94.4 96.0 97.6 99.1 100.6 102.0 103.5 104.8 106.2 107.6 108.9 110.2 111.4 112.7 113.9 115.1 116.3 117.5 118.6 119.7 120.8 121.9 123.0 124.0 125.1 126.1 127.1 128.1 129.1 130.0 131.0 131.9 132.9 133.8 134.7 135.6 (OF) 156.1 158.0 159.9 161.7 163.4 165.1 166.8 168.4 170.0 171.6 173.1 174.6 176.0 177,5 178.9 180.2 181.6 182.9 184.2 185.5 186.7 187.9 189.1 190.3 191.5 192.6 193.7 194.9 195.9 197.0 198.1 199.1 200.1 201.1 202.1 203.1 (OF) 124.4 125.8 127.1 128.4 129.7 131.0 132.3 133.5 134.7 135.9 137.0 138.2 139.3 140.4 141.5 142.6 143.6 144.7 14537 146.7 147.7 148.7 149.7 150.6 151.6 152.5 153.4 154.3 155.2 156.1 157.0 157.8 158.7 159.5 160.4 161.2

(-F) 202.4 203.7 205.0 206.3 207.5 208.7 209.9 211.1 212.2 213.3 214.4 215.5 216.6 217.7 218.7 21937 220.8 221.8 222.7 223.7 224.7 225.6 226.5 227.5 228.4 229.3 230.1 231.0 231.9 232.7 233,6 234.4 235.2 236.0 236.8 237.6 RPV &

32 EFPY BELTLINE CURVE C (OF) 242.4 243.7 245.0 246.3 247.5 248.7 249.9 251.1 252.2 253.3 254.4 255.5 256.6 257.7 258.7 259.7 260.8 261.8 262.7 263.7 264.7 265.6 266.5 267.5 268.4 269.3 270.1 271.0 271.9 272.7 273.6 274.4 275.2 276.0 276.8 277.6 90

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE 8-1. FitzPatrick P-T Curve Values for 32 EFPY Required Temperatures at 100 0Fhr for Curves B & C and 20 OF/hr for Curve A FOR FIGURES 8-1 THROUGH 8-3 BOTTOM RPV &

PRESSURE HEAD 32 EFPY BELTLINE CURVE A CURVE A BOTrOM RPV &

HEAD 32 EFPY BELTL, NE CURVE B CURVE B (PSIG) 1060 1070 1080 1090 1100 1110 1120 1130 1140 1150 1160 1170 1180 1190 1200 1210 1220 1230 1240 1250 1260 1270 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 (OF) 136.5 137.3 138.2 139.0 139.9 140.7 141.5 142.3 143.1 143.9 144.7 145.5 146,2 147,0 147.7 148.5 149.2 149.9 150.6 151.3 152.0 152.7 153.4 154.1 154.8 155.4 156.1 156.8 157.4 158.1 158.7 159.3 159.9 160.6 161.2 (OF) 204.1 205.0 206.0 206.9 207.8 208.7 209.6 210.5 211.4 212.2 213,1 213.9 214.7 215.6 218.8 219.6 220.4 221.1 221.9 222.7 223.4 224.2 224.9 225.6 226.3 227.0 227.7 228.4 229.1 229.8 230.5 231.1 231.8 232.5 233.1 (aF) 162.0 162.8 163.6 164.4 165.1 165,9 166.7 167.4 168.1 168.9 169.6 170.3 171.0 171.7 172.4 173.1 173.8 174.5 175.1 175.8 176.5 177.1 177.8 178.4 179.0 179.6 180.3 180.9 181.5 182.1 182.7 183.3 183.9 184.5 185.0 (OF) 238.4 239.2 239.9 240.7 241.4 242.2 242.9 243.6 244.4 245.1 245.8 246.5 247.2 247.8 250.6 251.2 251.9 252.5 253.2 253.8 254.5 255.1 255.7 256.3 256.9 257.5 258.2 258.7 259.3 259.9 260.5 261.1 261.7 262.2 262.8 RPV &

32 EFPY BELTLINE CURVE C (OF) 278.4 279.2 279.9 280.7 281.4 282.2 282.9 283.6 284.4 285.1 285.8 286.5 287.2 287.8 290.6 291.2 291.9 292.5 293.2 293.8 294.5 295.1 295.7 296.3 296.9 297.5 298.2 298.7 299.3 299.9 300.5 301.1 301.7 302.2 302.8 91

GE Nuclear Energy GE-NE-B 110O0 732-01 Revision 1

9. REFERENCES

[1]

"Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

[2]

" Protection Against Non-Ductile Failure," Appendix G to Section X) of the 1989 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, December 1995.

[41 "Surveillance Test for Nuclear Reactor Vessels," Annual Book of ASTM Standards, El 85-70.

[5]

T. A. Caine, "Implementation of Regulatory Guide 1.99, Revision 2 for the James A.

FitzPatrick Nuclear Power plant, GENE, San Jose, CA, June 1989, (GE Report SASR 89-50).

[6]

"Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," Annual Book of ASTM Standards, E185-82, July 1982.

[7]

"Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

[8]

T. A. Caine, "James A. FitzPatrick Nuclear Power Plant, Reactor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis", GENE, San Jose, CA, April 1986, GE Report MDE-49-0386.

[9]

James A. FitzPatrick Nuclear Power Updated Safety Analysis Report, Section 4.2.

[10]

"PVRC Recommendations on Toughness Requirements for Ferritic Materials", Welding Research Council Bulletin 175, August 1972.

[11]

Martin, G.C., "Fast Neutron Cross-Section Determination for BWR's Using Neutron Dosimeters," November 11, 1993 (FMT Transmittal 93-212-0045).

92

(

Nuclear E--nergy GE-NE-B 1100732-01 Revision 1

[12]

"Standard Methods for Notched Bar Impact Testing of Metallic Materials," Annual Book of ASTM Standards, E23-94b.

[13]

"Nuclear Plant Irradiated Steel Handbook," EPRI Report NP-4797, September 1986.

[14]

"Standard Test Methods for Tension Testing of Metallic Materials," Annual Book of ASTM Standards, E8-89.

[15]

B.J. Branlund, "Reactor Vessel Fracture Toughness Engineering Evaluation for the James A. FitzPatrick 104% Power Uprate," March 1996, GE-NE-B 1301805-05RI.

[16]

H. S. Mehta, T. A. Caine, and S. E. Plaxton, "10 CFR 50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWRI2 through BWR/6 Vessels, Rev. 1," GENE, San Jose, CA, February, 1994, (NEDO-32205-A).

[17]

T. A. Caine, "Progress Report on Phase 2 of the BWR Owners' Group Supplemental Surveillance Program," January 1992, GE-NE-523-101-1290.

[18]

"Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds,"

CEOG Task 902, June 1997, CE NPSD-1039.

[19]

Letter from Jeffrey L. Beck (Lukens Steel, Coatesville, PA)) to Richard Chau (NYPA, White Plains, NY) dated October 14, 1985 on "FitzPatrick Vessel Plate Copper Chemistries."

[20]

R. M. Kruger and R. D. Reager, "Determination of Fast Neutron Fluence in Flux Wires at the FitzPatrick Nuclear Power Station (Cycle 1-12 Irradiations), August 1997, GE-NE Report B 11-00732-RMK1.

[21]

"Response to Genric Letter 92-0 1, James A. FitzPatrick Nuclear Power Plant,"

Submitted to U.S. NRC by New York Power Authority, July 9, 1992.

93

GE Nuclear Energy GE-NE-B 1100732-01 Revision I APPENDIX A IRRADIATED CHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS Photographs of each Charpy specimen fracture surface were taken per the requirements of ASTM El 85-82.

The pages following show the fracture surface photographs along with a summary of the Charpy test results for each irradiated specimen. The pictures are arranged in the order of base, weld, and HAZ materials.

A-1

GE Nuclear Energy BASE: 53P Temp: 24 'F Energy: 44.2 ft-lb MLE: 39 mils Shear: 31 %

BASE: 53Y Temp: 49 TF Energy: 81.0 fl-lb MLE: 65 mils Shear: 46 %

BASE: 53L Temp: 250TF Energy: 120.4 fl-lb MLE: 91 mils Shear: 100 %

BASE: 527 Temp: 400TF Energy: 126.7 ft-lb MLE: 93 mils Shear: 100 %

GE-NE-B 1100732-01 Revision 1 BASE: 53U Temp: -50 F Energy: 10 ft-lb MLE: 8 mils Shear: I %

BASE: 53M Temp: 0 F Energy: 35.4 ft-lb MLE: 31 mils Shear: 29 %

BASE: 53B Temp: 103°F Energy 96.5 ft-lb MLE: 77 mils Shear: 78 %

BASE: 52D Temp: 150TF Energy: 1 7.4 ft-lb MLE: 89 mils Shear: 100%

A-2

GE Nuclear Energy WELD: 563 Temp: 103TF Energy: 29.4 ft-lb MLE: 27 mils Shear: 34 %

WELD: 56L Temp: 120 F Energy: 33.7 fl-lb MLE: 32 mils Sht2r: 55 %

WELD: 54M Temp: 250 'F Energy: 72.5 fl-lb MLE: 66 mils Shear: J00 %

WELD: 54T Temp: 400 TF Energy: 75.0 ft-lb MLE: 73 mils Shear: 100 %

GE-NE-B 1100732-01 Revision I WELD: 56A Temp: 0 F Energy: 3.2 ft-lb MLE: 3 mils Shear: 1 %

WELD: 565 Temp: 80 TF Energy: 18.9 ft-lb MLE: 18 mils Shear: 30/ %

WELD: 55Y Temp: 163TF Energy: 56.8 ft-lb MLE: 43 mils Shear: 75 %

WELD: 55B Temp: 202 TF Energy. 68.1 ft-lb MLE: 61 mils Shear: 88 %

A-3

GE Nuclear Energy HAZ 5AK Temp: 0°F Energy: 44.0 ft-lb VNfE: 41 mils Shear: 22 %

HAAZ: 5AU Temp: 489F Energy: 98.3 ft-lb MLE: 78 mils Shear: 30 %

HAZ: 5AB Temp: 202 9 Energy: 102.2 ft-lb MLE: 82 mils Shear: 100 %

HAZ: 5A6 Temp: 400'F Energy: 112.6 fl-lb MLE: 95 mils Shear: 100 %

GE-NE-B 1100732-01 Revision 1 HAZ: SAT Temp: -80 'F Energy: 36.0 ft-lb MLE: 29 mils Shear: 38 %

HAZ: SAY Temp: -50 *F Energy: 36.6 ft-lb MLE: 30 mils Shear: 41 %

11AZ: 57P Temp: 80 'F Energy: 77.6 ft-lb MLE: 67 mils Shear: 82 %

1JAZ: 575 Temp: 120 'F Energy: 74.1 ft-lb MLE: 69 mils Shear: 100%

A-4

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 APPENDIX B PRESSURE TEMPERATURE CURVES VALID TO 24 EFPY B-1

GE Nuclear Energy GE-NE-B 1100732-01 Revision I APPENDIX B PRESSURE TEMPERATURE CURVES VALID TO 24 EFPY B-I

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 0.

0 I-ixw w

a.-

1400 1300 1200 1100 1000 900 800 700 600 500 400 300 200 100 a

INITIAL RTndt VALUES ARE

-48*F FOR BELTLINE, 30OF FOR UPPER VESSEL, AND 10°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (fF) 24 143 HEATUPICOOLDOWN FRATE 20FIHR UPPER VESSEL AND BELTLINE LIMITS

-- - BOTTOM HEAD LIMITS I

I.

I 0

0 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

FIGURE B-i: PRESSURE TEST CURVE (CURVE A)

B-2

GE Nuclear Energy 1400 GE-NE-B 1100732-01 Revision I

-a 0

~.j 0

w It a.

1300 1200 1100 1000 900 800 700 600 500 400 300 BELTINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 24 143 HEATUP/COOLDOWN RATE 100°F/HR INITIAL RTndt VALUES ARE

-480 F FOR BELTLINE, 30'F FOR UPPER VESSEL, AND 100F FOR BOTTOM HEAD UPPER VESSEL AND BELTLINE LIMITS

- BOTTOM HEAD LIMITS 200 100 0

0 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

FIGURE B-2: NON-NUCLEAR HEATUP/COOLDOWN (CURVE B)

B-3

GE Nuclear Energy 1400 C -

1300 CRITI(

I FITZF 1200 GE-NE-B 1100732-01 Revision 1 0..

0 L1 It LU 0>

I-1100 1000 900 800 700 600 500 400 300 200 100 0

INITIAL RTndt VALUES ARE

-480F FOR BELTLINE, 30'F FOR UPPER VESSEL, AND 10OF FOR BOTTOM HEAD BELTINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 24 143

[HETUP/COOLDOWN RATE 100°F/HR

-BELTLINE AND NON-BELTLINE

_ [ _

_LIMITS Minimum Criticality Temperature 90*F 0

50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (=F)

FIGURE B-3: CORE CRITICAL OPERATION (CURVE C)

B-4 7

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE B-I. FitzPatrick P-T Curve Values for 24 EFPY Required Temperatures at 100 OF/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES B-I THROUGH B-3 BOTTOM RPV &

PRESSURE HEAD 24 EFPY BELTLINE CURVE A CURVE A BOTTOM RPV &

HEAD 24 EFPY BELTLINE CURVE B CURVE B (PSIG) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 310 312.5 312.5 320 330 (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 (OF) 90.0 90.0 90.0 90.0 90.0

.90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 120.0 120.0 120.0 (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 (OF) 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 90.0 92.7 97.5 101.9 106.1 110,1 113.6 116.8 119.8 122.8 125.6 128.2 130.6 132.9 135.2 137.4 1394 141.4 143.3 145.1 147.0 148.7 150.3 152,0 152.3 152.3 153.5 155.1 RPV &

24 EFPY BELTLINE CURVE C (OF) 90.0 90.0 90.0 90.0 94.5 105.2 113.9 121.1 127.4 132.7 137.5 141.9 146.1 150.1 153.6 156.8 159.8 162.8 165.6 168.2 170.6 172.9 175.2 177.4 179.4 181.4 183.3 185.1 187.0 188.7 190.3 192.0 192.3 195.0 195.0 195.1 B-5

GE Nuclear Energy GE-NE-B 1100732-01 Revision I TABLE B-1. FitzPatrick P-T Curve Values for 24 EFPY Required Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES B-I THROUGH B-3 BOTITOM RPV &

PRESSURE HEAD 24 EFPY BELTLINE CURVE A CURVE A BOTTOM RPV &

RPV &

HEAD 24 EFPY 24 EFPY BELTLINE BELTLINE CURVE B CURVE B CURVE C (PSIG) 340 350 360 370 380 390 400 410 420 430 440 450 460 470 480 490 500 510 520 530 540 550 560 570 580 590 600 610 620 630 640 650 660' 670 680 690 (OF) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68-0 68.0 68.0 69.5 71.8 74.0 76.1 78.2 80.2 82.1 84.0 85.9 87.7 89.4 91.1 92.8 (OF) 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 120.0 122.4 125.0 127.4 129..7 132.0 134.2 136.3 138.4 140.4

(°F) 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 70.3 73.2 76.1 78.8 81.5 84.0 86.5 88.8 91.1 93.3 95.5 97.6 99.6 101.5 103.5 105.3 107.1 108.9 110.6 112.3 113.9 115.5 117.1 118.6 120.1 121.6 123.0 (OF) 156.6 158.0 159.4 160.8 162.1 163.4 164.7 166.0 167.2 168.4 169.6 170.7 171.8 172.9 174.0 175.1 176.1 177.1 178.1 179.1 180.1 181.1 182.0 182.9 183.8 184.7 185.6 186.5 187.3 188.2 189.0 189.8 190.6 191.4 191.9 192.3 (OF) 196.6 198.0 199.4 200.8 202.1 203.4 204.7 206.0 207.2 20,8.4 209.6 210.7 211.8 212.9 214.0 215.1 216.1 217.1 218.1 219.1 220.1 221.1 222.0 222.9 223.8 224.7 225.6 226.5 227.3 228.2 229.0 229.8 230.6 231.4 231.9 232.3 B-6

GE Nuclear Energy GE-NE-8 1100732-01 Revision I TABLE B-I. FitzPatrick P-T Curve Values for 24 EFPY Required Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A FOR FIGURES B-I THROUGH B-3 BOTTOM RPV &

PRESSURE HEAD 24 EFPY BELTLINE CURVE A CURVE A BOTTOM RPV &

HEAD 24 EFPY BELTLINE CURVE B CURVE B (PSIG) 700 710 720 730 740 750 760 770 780 790 800 810 820 830 840 850 860 870 880 890 900 910 920 930 940 950 960 970 980 990 1000 1010 1020 1030 1040 1050 (OF) 94.4 96.0 97.6 99.1 100.6 102.0 103.5 104.8 106.2 107.6 108.9 110.2 111.4 112.7 113.9 115.1 116.3 117.5 118.6 119.7 120,8 121.9 123.0 124.0 125.1 126.1 127.1 128.1 129.1 130.0 131.0 131.9 132.9 133.8 134.7 135,6 (OF) 142.1 144.0 145.9 147.7 149.4 151.1 152.8 154.4 156.0 157.6 159.1 160.6 162.0 163.5 164.9 166.2 167.6 168.9 170.2 171.5 172.7 173.9 175.1 176.3 177.5 178.6 179.7 180.9 181.9 183.0 184.1 185.1 186.1 187.1 188.1 189.1 (OF) 124.4 125.8 127.1 128.4 129.7 131.0 132.3 133.5 134.7 135.9 137.0 138.2 139.3 140.4 141.5 142.6 143.6 144.7 145.7 146.7 147.7 148.7 149.7 150.6 151.6 152.5 153.4 154.3 155.2 156.1 157.0 157.8 158.7 159.5 160.4 161.2 (OF) 192.8 193.2 193.6 194.0 194.4 194.8 195.9 197.1 198.2 199.3 200.4 201.5 202.6 203.7 204.7 205.7 206.8 207.8 208.7 209.7 210.7 211.6 212.5 213.5 214.4 215,3 216,1 217.0 217.9 218.7 219.6 220.4 2212 222.0 222.8 223.6 RPV &

24 EFPY BELTLINE CURVE C (OF) 232.8 233.2 233.6 234.0 234.4 234.8 235.9 237.1 238.2 239.3 240.4 241.5 242.6 243.7 244.7 245.7 246.8 247.8 248.7 249.7 250.7 251.6 252.5 253.5 254.4 255.3 256.1 257.0 257.9 258.7 259.6 260.4 261.2 262.0 262.8 263.6 B-7

GE Nuclear Energy GE-NE-B 1100732-01 Revision 1 TABLE B-1. FitzPatrick P-T Curve Values for 24 EFPY Required Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES B-I THROUGH B-3 BOTTOM RPV &

PRESSURE HEAD 24 EFPY BELTLINE CURVE A CURVE A BOTTOM RPV &

HEAD 24 EFPY BELTLINE CURVE B CURVE B (PSIG) 1060 1070 1080 1090 1100 1110 1120 1130 1140 1150 1160 1170 1180 1190 1200 1210 1220 1230 1240 1250 1260 1270 1280 1290 1300 1310 1320 1330 1340 1350 1360 1370 1380 1390 1400 (OF) 136.5 137.3 138.2 139.0 139.9 140.7 141.5 142.3 143.1 143.9 144.7 145.5 146.2 147.0 147.7 148.5 149.2 149.9 150.6 151.3 152.0 152.7 153.4 154.1 154.8 155.4 156.1 156.8 157.4 158.1 158.7 159.3 159.9 160.6 161.2 (OF) 190.1 191.0 192.0 192.9 193.8 194.7 195.6 196.5 197.4 198.2 199.1 199.9 200.7 201.6 204,8 205.6 206.4 207.1 207.9 208.7 209.4 210.2 210.9 211.6 212.3 213.0 213.7 214.4 215.1 215.8 216.5 217.1 217.8 218.5 219.1 (OF) 162.0 162.8 163.6 164.4 165.1 165.9 166.7 167.4 168.1 168.9.

169.6 170.3 171.0 171.7 172.4 173.1 173.8 174.5 175.1 175.8 176.5 177.1 177.8 178.4 179.0 179.6 180.3 180.9 181.5 182.1 182.7 183.3 183.9 184.5 185.0 (OF) 224.4 225.2 225.9 226.7 227.4 228.2 228.9 229.6 230.4 231.1 231.8 232.5 233.2 233.8 236.6 237.2 237.9 238.5 239.2 239.8 240.5 241.1 241.7 242.3 242.9 243.5 244.2 244.7 245.3 245.9 246.5 247.1 247.7 248.2 248.8 RPV &

24 EFPY BELTLINE CURVE C (OF) 264.4 265.2 265.9 266.7 267.4 268.2 268.9 269.6 270.4 271-1 271.8 272.5 273.2 273.8 276.6 277.2 277.9 278.5 279-2 279.8 280.5 281.1 281.7 282.3 282.9 283.5 284.2 284.7 285.3 285.9 286.5 287.1 287.7 288.2 288.8 B-8

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• ~Athority Jam.. KnuWe Svix~ Vics Aasgw.o 3n Chief Nucioe Officei March 9, 1998 JPN-98-008 U.S. Nuclear Regulatory Commission Atm: Document Control Desk Mail Station P1-137 Washington, D.C. 20555

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 REVISED REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PRO-OAM

SUMMARY

REPORT AND IMPLEMENTATION SCHEDULE

1.

NYPA Letter, R. J. Deasy to NRC, -Reactor Pressure Vessel Material Surveillance Program Summary Report and Implementation Schedule,' (JPN-97-0351, dated November 10, 1997

Dear Sir:

The lotter provides a copy of the revised FitzPatrick Reactor Pressure Vessel (RPV) Material Testing and Analysis Report. The changes do not affect the conclusion or technical basis of the report and the requirements of 10 CFR 50 Appendix G co,,inue to be satisfied. A schedule and technical justification for the next capsule withdrawal is submitted for NRC review and approval. In addition, the Authority is providing the results of the ongoing Owner's Group RPV integrity program relative to FitzPatrick. The Authority committed to provide this information in Reference 1. describes the Authority's resolution to the Reference 1 Commitments. is the revised FitzPatrick RPV Surveillance Materials Testing and Analysis report. Attachment 3 is a summary of the commitments made in this letter.

If you have any questions, please contact Ms. C. Faison.

Very Truly Yours, C Chief Nuclear Offi Senior Vice Presid cc:

Regional Administrator U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Office of the Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, New York 13093 Mr. J. Williams, Project Manager Project Directorate I-1 Division of Reactor Projects I1/1 U.S. Nuclear Regulatory Commission Mail Stop 14 B2 Washington, DC 20555 cer and ern

- ý '- ýO 1-1

- ý 0

Attachment I to JPN-98-008 RESOLUTION OF COMMITMENT NUMBERS JPN-97-035-OO1 AND JPN-97-035-002 New York Powier Authoity JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

Introduction and Background The Authority committed (JPN-97-035, Reference 1) to provide the NRC with a copy of the test result report, revised to reflect the resolution of comments and concerns within 120 days. Included in this response is a description of the Quality Assurance (CA) finding regarding specimen test procedures, and its resolution. A schedule and technical justification for the next capsule withdrawal is also submitted at this time for NRC review and approval. In addition, the Authority is providing the results of the ongoing Owner's Group Reactor Pressure Vessel (RPV) integrity program relative to FitzPatrick.

Resolution of QA Concerns The Authority had a concern regarding how the capsule testing was performed based on the results of an audit conducted on 'the General Electric (GE, Company by the Authoritv's OA Department in November 1997. The subject of the audit was reactor vessel surveillance specimen testing for FitzPatrick. As a result of this audit, one finding was identified concerning the availability and use of procedures for conducting charpy impact tests. Specifically, the finding stated the following:

"The test engineer responsible for the testing of NYPA's (JAF) surveillance capsules charpy specimen stated that testing was performed without a procedure.'

The Authority performed an audit in February 1998 to address this finding.. During this audit, the GE personnel that actually performed the FitzPatrick specimen testing were interviewed. The Authority was informed during these interviews that the 1993 version of the charpy test procedure, valid at the time of testing. was available in a file cabinet in the, hot lab test room where the tests were performed. However, the GE personnel did not refer to this procedure during testing. The Authority determined during this audit that the GE personnel performing the testing were familiar with the requirements of the 1993 charpy test procedure and followed these requirements to perform the various tasks necessary to conduct the tests. Each engineer has the requisite education and experience required to qualify them to perform these tests.

At the time the FitzPatrick tests were performed, a draft revision to the charpy test procedure existed. This draft test procedure included a newly purchased 300 ft-lb charpy test machine in the list of applicable apparatus. This machine was used to test the FitzPatrick surveillance capsule specimens. However, at the time of the test, the 1993 test procedure had neither been amended nor revised to Include this test machine in the list of applicable apparatus. Other than the omission of the new test machine, there was no substantive differenco between the two procedures. The draft revision to the procedure was approved in October 1997. This version of the charpy test procedure contains thc requirements that GE followed during the conduct of the FitzPatrick surveillance capsule charpy specimen test.

Based on the above, the Authority is satisfied that the tests were performed properly and in accordance with the newly revised charpy test procedure. The test results are therefore valid for use in GE Report No. GE-NE-B1 100732-01, Revision 1, "Plant FitzPatrick RPV Surveillance Material Testing and Analysis of 120* Capsule at 13.4 EFPY' (Reference 2).

1

ý.,

-VI.-

ý Editorial Comments The following twochanges were made to GE Report No. GE-NE-B1 100732-01 (Reference 3ý and are incorporated in Revision 1 (Reference 21 to this report:

1.

Page 12, Table 3-3 Weld 1-233 charpy energy values of 74ft-lb, 63ft-lb, and 82ft-lb have been corrected and replaced with charpy energy values of 60ft-lb, 64ft-lb, and 55ft-lb, respectively. The original values were incorrectly taken from a previous report.

Since this weld continued,rot to be the limiting weld/plate, this correction does not change the resufts of the original GE report.

2.

Page 17, Section 4.1.1, Third Paragraph, Third Sentence Delete:

"The calculated fluence results for the Fe, Ni, and Cu wires differed by less than 10%, thus, an average fluence value yas used."

Replace with:

"The calculated fluence result from the iron flux wire was used. The Ni and Cu flux wires confirmed the fluence result from the iron specimen, with all three results differing by less than 10%.-

This change has no effect on the results of the original GE report.

Based on the above, the editorial comments and GA concerns did not alter the overall conclusion of the report. The revised report continues to demonstrate i~at the requirements of 10 CFR 50, Appendix G are satisfied.

Schedule and Technical Justification for Next Capsule Removal Based on ASTM E185-82 (Reference 4), the third capsule does not need to be withdrawn until end of license (i.e., 2014). The curves contained in the GE "aport are valid for up to 32 Effective Full Power Years IEFPY) of operation which corresponds to at least the end of license. Vessel fluence is expected to be less than 32 EFPY at that time. The third capsule will be withdrawn at approximately 30 EFPY. This will support operation beyond 32 EFPY, should the operating license be extended. In accordance with 10 CFR 50, Appendix H, Section 111.8.3, the Authority requests NRC approval of this proposed withdrawal schedule.

2

Status of ongoing BWR Owner's Group RPV Integrity Program Relative to FitzPatrick A revised report, 'BWR Vessel and Internals Project Update of Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity Issues fBWRVIP-46),' EPRI. Palo Alto, Ca.,

December 1997, (EPRI TR-109727) was submitted by EPRI to the NRC. This revised report included data not previously reported and concluded that there is no effect on the Pressure-Temperature (P-T) curves due to chemistry variability for the BWR vessels.

References

1.

NYPA Letter, R. J. Deasy to NRC, "Reactor Pressure Vessel Material Surveillance Program Summary Report and Implementation Schedule," (JPN-97-035), dated November 10, 1997

2.

General Electric Company Final Report. *Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120 Capsule at 13.4 EFPY," GE-NE-B1100732-01, Revision 1, Class 11, dated February 1998

3.

General Electric Company Final Report, "Plant FitzPatrick RPV Surveillance Materials Testing and Analysis of 120' Capsule at 13.4 EFPY," GE-NE-B1 100732-01, Class I1, dated October 1997

4.

American Society for Testing and Materials, Standard Practice Regarding Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, ASTM E185-82, approved July 1, 1982 3

to JPN-98-008 GENERAL ELECTRIC REPORT NO. GE-NE-B 1100732-01, REVISION 1 FITZPATRICK REACTOR PRESSURE VESSEL SURVEILLANCE MATERIALS TESTING AND ANALYSIS REPORT OF 120 DEGREE CAPSULE AT 13.4 EFPY New York Power Authority JAMES A. FITZPATRICK NUCf EAR POWER PLANT Docket No. 50-333 DPR-59

1. ý to JPN-98-008 Summary of Commitments Commitment Number Description Due Date Approximately 30 EFPY JPN-98-008-01 Withdraw the third capsule.