ML17292A830

Wppss WNP-2 RPV Surveillance Matls Testing & Analysis.

ML17292A830
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Site:	Columbia
Issue date:	03/31/1997
From:	Branlund B, Chu C, Frew B GENERAL ELECTRIC CO.
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ML17292A829	List: ML17292A828 ML17292A830
References
GE-NE-B1301809, GE-NE-B1301809-01, GE-NE-B1301809-1, NUDOCS 9705050168
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GE Nuclear Energy Lngineering & Licensing Consulting Services GE-NE-B 1301809-01 General Electric Company 175 Curtner Avenue, San Jose, CA 95125 March 1997 WASHINGTON PUBLIC POWER SUPPLY SYSTEM WNP-2 RPV SURVEILLANCE KIATERIALS TESTING AND ANALYSIS Prepared by:

C.L. Chu, Principal Engineer Engineering & Licensing Consulting Services Verified by:

B.D. Frew, Engineer Engineering & Licensing Consulting Services Approved y:

B.J. Branlund, Project Manager, Engineering & Licensing Consulting Services 9705050168 970424 1 PDR ADQCK 05000397 P PDR

GE-NE-B1 301 809-01 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of Washington Public Power Supply System. 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.

GE-NE-BI301 809-Ol TABLE OF CONTENTS ABSTRACT V11 ACKNOWLEDGMENTS vn1 INTRODUCTION

SUMMARY

AND CONCLUSIONS 2.1 Summary of Results 2.2 Conclusions SURVEILLANCEPROGRAM BACKGROUND 3.1 RPV Materials and Fabrication History 3.2 Capsule Recovery 3.3 Specimen Description SURVEILLANCE SPECIMEN CHEMICALCOMPOSITION 15 PEAK RPV FLUENCE EVALUATION 17 5.1 Flux Wire Analysis 17 5.2 Determination of Lead Factor 19 5.3 Evaluation of 32 EFPY Fluence 22 CHARPY V-NOTCH IMPACT TESTING 31 6.1 Impact Test Procedure 31 6.2 Impact Test Results 32 6.3 Comparison of the Measured and Predicted Irradiation Shifts 33 6.4 Change in USE 34 TENSILE TESTING 45 7.1 Procedure 45 7.2 Results 45 ADJUSTED REFERENCE TEMPERATURE AND UPPER SHELF ENERGY 49 8.1 Adjusted Reference Temperature at 32 EFPY 49 8.2 Upper Shelf Energy at 32 EFPY 50 REFERENCES 55 111

GE-NE-B1 301 809-01 TABLE OF CONTENTS APPKNDICKS UNIRRADIATEDAND IRRADIATEDCHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS A1 UNIRRADIATEDAND IRRADIATEDTENSILE SPECIMEN FRACTURE APPEARANCE B1

GE-NE-B1301 809-01 TABLES Table Title ~Pa e Test Code Requirements 3.-1(a) Chemical Composition 8c Initial RTNDr of RPV Beltline Plate Materials From WNP2 FSAR Records 10 3-1(b) Chemical Composition & Initial Rior of RPV Beltline Weld Materials From WNP2 FSAR Records 4-1 Chemical Composition of WNP2 Surveillance Specimens From GE Chemical Analysis 16 5-1 Summary of Daily Power History 24 5-2 Summary of WNP2 Irradiation Periods 25 5-3 Dosimeter Nuclear Parameters 25 5-4 Surveillance Capsule Flux and Fluence For Irradiation from Start-up to 3/2/96 26 6-1 Vallecitos Qualification Test Results Using MIST Standard Reference Specimens 35 6-2 Unirradiated Charpy V-Notch Impact Test Results 36 6-3 Irradiated Charpy V-Notch Impact Test Results 37 6-4 Key Parameter Summary of Irradiated and Unirradiated Charpy V-Notch Impact Data (for 300'zimuth Surveillance Capsule at 7.2 EFPY) 38 7-1 Tensile Test Results for Irradiated And Unirradiated RPV Materials (for 300'zimuth Surveillance Capsule at 7.2 EFPY) 47 8-1 Beltline Base Plate ART for 32 EFPY at 1/4T 52 8-2 Beltline Weld ART for 32 EFPY at 1/4T 53 8-3 Upper Shelf Energy Analysis for WNP2 Beltline Materials Represented by the Surveillance Specimens 54

GE-NE-B1 301809-01 I

FIGURES F~iure Title ~Pa e 3-1 RPV Schematic With Beltline Plate Identification 12 3-2 Recovered Surveillance Capsule Basket 13 3-3(a) Charpy Specimen Container Identification 14 3-3(b) Tensile Specimen Identification 14 5-1 Azimuthal Flux Distribution Analysis Model 27 5-2 Angular Flux Variation At Core Midplane 28 5-3 R-Z Calculation Model 29 5-4 Axial Flux Profile At Vessel I.D. Surface 30 6-1 Irradiated And Unirradiated Charpy Tests WNP2 Base Plate Impact Energy 39 6-2 Irradiated And Unirradiated Charpy Tests WNP2 Base Plate Lateral Expansion 40 6-3 Irradiated And Unirradiated Charpy Tests WNP2 Weld Impact Energy 41 6-4 Irradiated And Unirradiated Charpy Tests WNP2 Weld Lateral Expansion 42 6-5 Irradiated And Unirradiated Charpy Tests WNP2 HAZ Impact Energy 43 6-6 Irradiated And Unirradiated Charpy Tests WNP2 HAZ Lateral Expansion 44 7-1 Typical Engineering Stress-Strain For Irradiated RPV Materials 48

GE-NE-B1 301 809-01 ABSTRACT The surveillance capsule at the 300'zimuthal location was removed at 7.2 EFPY (normalized full power of 3323 MWt) from the WNP2 reactor in Spring 1996 (the reactor was at uprated power level of 3486 MWt when WNP2 was shutdown). The capsule contained flux wires for neutron fluence measurement and Charpy 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 uniaxial tensile testing were performed to establish the properties of the irradiated surveillance materials.

The irradiated Charpy data for the base plate, weld and heat affected zone (HAZ) specimens were compared to the corresponding unirradiated specimen test data to determine the shift in Charpy curves due to irradiation. Both the irradiated and unirradiated Charpy base plate data are of longitudinal orientation. The shift results for the base plate and the weld materials were also compared with the predictions of the Regulatory Guide 1.99 Revision 2 (RG1.99) and were found to be within the predicted values.

The data from the irradiated materials tested at 70'F and 550'F were compared with those from the unirradiated data tested at the same temperatures to determine the effect of irradiation on the stress-strain relationship of the materials.

The flux wire results were used to calculate the 32 EFPY fluence. The resulting fluence is higher than those based on the neutron flux from the first cycle.

0 GE-NE-B1301 809-01 ACKNOWIEDGMENTS The author gratefully acknowledges the efforts of other people towards completion of the contents of this report.

Cask shipment and capsule disassembly was performed by Jon Myers. Lead factor calculations was provided by D. Rogers and S. Wang. Charpy testing was completed by G. E.

Dunning and B. D. Prew. Tensile specimen testing was done by S. B. Wisner and chemical composition analysis was performed by P. Wall. Flux wire testing and analysis was performed by L. Kessler, R. Kruger and R. Reager.

l GE-NE-B1 301809-01 An important part of the effort to assure reactor vessel integrity involves evaluation of the fracture toughness of the vessel ferritic materials. The key parameters which characterize a material's fracture toughness are the reference nil-ductility transition temperature (RT>DT) and the upper shelf energy (USE). Both of these parameters, defined in 10CFR50 Appendix G 'nd in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI 1, are required to be 1

updated from surveillance sample testing and analyses at specified intervals of a reactor's effective full power years. The WNP2 FSARt calls for the first surveillance capsule (at the 1

300'essel azimuth location) to be withdrawn at 8 Effective Full Power Years (EFPYs).

I Appendix H of 10CFR50 and ASTM E185-82 establish the methods to be used for surveillance testing of the WNP2 reactor vessel materials. It should be noted that the use of ASTM E185-82 is required by the Appendix H of 10CFR50 which specifies that, for each capsule withdrawal, the test procedures must meet the requirements of ASTM E185-82 to the extent practicable for the configuration of the specimens in the capsule. For the WNP2 surveillance specimen evaluation, however, there are no significant differences between ASTM E185-82 and its earlier version of E185-73 which was referenced in the WNP2 FSAR. During the scheduled outage of 1996, WNP2 completed the first surveillance capsule removal from the reactor at 7.2 EFPY (normalized full power of 3323 MWt, the reactor was at uprated power level of 3486 MWt when it was shutdown) and the irradiated samples were shipped in May, 1996 to the GE Vallecitos Nuclear Center (VNC) for testing.

The surveillance capsule contained flux wires for neutron flux monitoring and Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated from the same vessel materials as those located within the core beltline region. The impact and tensile specimens were tested to establish properties for the irradiated materials. Prior to this effort, GE had received from WNP2 the unirradiated Charpy impact and tensile baseline specimens, which were similarly tested. The test results from these unirradiated specimens constitute as the baseline for evaluating the irradiation effects on the material properties which include (i) the fracture toughness as measured in terms of- Charpy impact (absorbed) energy, lateral expansion and percent shear area, and (ii) the stress-strain relationship of the vessel materials.

All the test code requirements and procedures are summarized in Table 1-1.

GE-NE-B1301 809-01 Table 1-1 Test Code Requirements Test Procedure Tensile Test Tensile tests are conducted in accordance with (i) ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, and (ii)

ASTM E8-95, "Standard Methods of Tension Testing of Metallic Materials", with the exception that specimen displacement is obtained from crosshead movement of the Instron load frame rather than an extensometer attached to the s ecimen.

Charpy Impact Charpy tests are conducted according to (i) ASTM E185-82, Test "Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, and (ii) ASTM E23-94b, "Standard Test Methods for Notched Bar Impact Testin of Metallic Materials".

Flux Wire The analysis is conducted according to (i) ASTM E185-82, Analysis "Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, and (ii) GE Report NEDE-11373, Method 10.1.6.0, Rev.4, "Determination of Neutron Fluence Rate and Fluence Using Neutron Dosimeters". The dosimetry method is adhered to ASTM E181, E261, E262, E263, E264, E523. E844 and E1005.

Chemical The analysis is conducted according to GE Report NEDE-Composition 11373, Method 10.2.1.3, Rev. 0, "Determination of Element Analysis Concentrations In Solutions by DCP Emission Spectroscopy CM & S Laborato Manual"

0 GE-NE-BI301 809-OI 2.

2.1

SUMMARY

OF RESULTS The 300'zimuth 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. The methods and results of the testing are presented in this report as follows:

Section 3: Surveillance Program Background

~ RPV Materials and Fabrication

~ Capsule Recovery

~ Specimen Description Section 4: Surveillance Specimen Chemical Composition Section 5: Peak RPV Fluence Evaluation Section 6: Charpy V-Notch Impact Testing Section 7: Tensile Testing Section 8: Adjusted Reference Temperature and Upper Shelf Energy The test and analysis summary is provided as follows:

The 300'zimuth position capsule was removed from the reactor after 7.2 EFPY of operation. The capsule contained 6 flux wires: 2 each of copper (Cu), iron (Fe), and nickel (Ni). There were 24 Charpy V-Notch specimens in the capsule:

8 each of plate, weld, and heat affected zone (HAZ) materials. The 6 tensile specimens removed consisted of 2 plate, 2 weld and 2 HAZ metal specimens. (see Sections 3.2 and 3.3)

One box of unirradiated specimens was received with a total of 68 Charpy specimens (24 for base. 21 for weld and 23 for HAZ) and 12 tensile specimens (4 for base, 3 for weld and 5 for HAZ). Twelve Charpy specimens from each material were tested. Two tensile specimens from each material were tested at room temperature and at 550'F. The remaining samples will be returned to WNP2 for future evaluations.

GE-NE-B1301 809-01 The chemical composition of copper (Cu) and nickel (Ni) for the irradiated surveillance materials were determined from the chemical composition analyses.

The average values for the surveillance base plate are 0.11% Cu and 0.49% Ni, and are 0.03% Cu and 0.89% Ni for the surveillance weld. (see Table 4-1)

A neutron transport computation had been performed based on Cycle 10 core data in representing the power shape and void distribution of the core. The lead factor was 0.95, relating the surveillance capsule flux to the peak inside surface flux.

(see Section 5.2.3)

From the flux wire test data, the neutron flux (E>1 MeV) at the surveillance capsule location was determined to be 6.85x10 n/cm -s, which was larger than both the nominal and the upper bound flux values from the first cycle dosimetg data by 43% and 15% respectively. The flux wire measurement shows that'the fluence (E >1 MeV) received by the surveillance specimens was 1.55x10'/cm at removal. The resulting 32 EFPY RPV peak fluence prediction is 7.57x10'/cm at the vessel inside diameter (I.D.) wall surface. At 1/4T from the I.D. surface, the 32 EFPY fluence prediction is 5.14x10 n/cm. The 32EFPY fluence was based on the sum of 6.57 EFPY at 3323 MWt for Cycles 1-10 and 25.43 EFPY at 3486 MWt from Cycle 11 to the end of life. (see Section 5.1.3, Section 5.3 and Table 5-4)

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 impact (absorbed) energy curves for the plate, weld, and HAZ materials.

Measurements were taken of the absorbed energy, lateral expansion and percentage shear. The absorbed energy and lateral expansion values were curve-flit with the hyperbolic tangent function. From these curves, the USE values and the index temperatures for 30 ft-lb, 50 ft-lb and 35 mils lateral expansion (MLE) were obtained (Table 6-4). Fracture surface photographs'showing the shearing characteristics of each specimen are presented in Appendix A.

The curves of irradiated and unirradiated Charpy specimens established the 30 ft-lb shifts. The base material showed an -1.0'F shift (decrease) and 4.6 ft-lb (3%) increase in USE. The weld material showed an -4.9'F shift (decrease) and a 12.1 ft-lb (12%) increase in USE. (see Table 6-4)

GE-NE-B1 301 809-OI

h. The measured shifts (at 7.2 EFPY, End of Cycle 11) of -1.0'F for the base plate and -4.9'F for weld, at the fluence of 1.55x10 n/cm, were well within the Reg.

Guide 1.99, Rev.2 (RG1.99)@ predicted range of -23 to 45'F and -50 to 62'F and 12.0'F for the plate and weld, respectively. (see Section 6.3)

Both the irradiated and unirradiated tensile specimens were tested at room temperature (70'F) and at reactor operating temperature (550'F). As expected, the results show that the irradiated data has an increase in strength and a decrease in ductility, which is a typical indication of irradiation embrittlement. (see Table 7-1)

The 32 EFPY adjusted reference temperature (ART = initial RTgpT + BRTgDT +

Margin) was predicted for each beltline material, based on the RG1.99 methodology. The ART for the limiting plate (Heat No. C1272-1) at 32 EFPY is 83.8'F and is lower than the 200'F requirement of 10CFR50 Appendix G. (see Table 8-1)

k. An update of the beltline material USE values at 32 EFPY was performed using the RG1.99 methodology. The irradiated USE for all beltline materials will remain above 50 ft-lb through 32 EFPY as required in 10CFR50 Appendix G.

(see Section 8.2)

2.2 CONCLUSION

S The requirements of 10CFR50 Appendix G specifies the vessel design life conditions with limits of operation designed to prevent brittle fracture'. From the evaluation of surveillance test results and the related analyses, the following conclusions are made:

ao The 30 ft-lb shifts and changes in USE are well within the values predicted by RG1.99. From the surveillance test results it is clear that after 11 cycles of operation, the 30 ft-lb shifts and the USE of the base plate and thc weld show little change after accumulating a peak irradiation fluence (E>1 MeV) of 1.55x10 n/cm at 7.2 EFPY (normalized full power of 3323 MWt, the reactor was at uprated power level of 3486 MWt for Cycle 11 from 6/9/95 to 3/2/96)

0 GE-NE-B1 301 809-01

b. The values of ART and USE for the reactor vessel beltline materials are expected to remain within limits of 10CFR50 Appendix G (< 200'F and >

50 ft-lb., respectively) for at least 32 KFPY of reactor operation.

GE-NE-B1301 809-01

3. V 3.1 RPV MATERIALSAND FABRICATION HISTORY Material certification records were retrieved from GE Quality Assurance (QA) records to determine chemical and mechanical properties of the vessel materials. The retrieved information is documented in the WNP2 FSAR. Table 3-1 shows the chemistry data for the beltline materials.

The WNP2 RPV is a 251 inch diameter BWR/5 design. Construction was performed by CBI Nuclear Company (CBIN) under the 1971 edition of the ASME Code through the 1971 Summer Addenda. The reactor pressure vessel was primarily constructed from high strength, low alloy (HSLA) steel plate and forging. Plates (for the shell and head plate) were ordered to ASME SA-533, Grade B, Class 1, and forging (for the nozzles and closure flanges) to ASME SA-508, Class 2, and the studs, nuts and washers for the main closure flange were ordered to ASME SA540, Grade B23 or Grade B241 1. The fabrication process of the vessel plates included hot forming immediately followed by a quench and temper heat treatment. During the assembly process for the completed vessel, submerged arc welding and shielded metal arc welding of plates were applied and were followed by post-weld heat treatment at 1150'F. The identification of plates and welds in the beltline region is shown in Figure 3-1.

3.2 CAPSULE RECOVERY The reactor pressure vessel (RPV) surveillance program consists of three surveillance capsules at 30', 120', and 300'zimuths 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. During the current 1996 Spring outage, a surveillance capsule was removed from the 300'zimuthal 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 basket was examined for identification. The identification number stamped on the capsule basket (specimen holder) corresponded to basket number 131C7717G1 and reactor number 55, as specified by GE drawings, 131C7717 (specimen holder) and 105D4719G001 (Surveillance Program), for the WNP2 300'urveillance materials.

The general condition of the'asket as received is shown in Figure 3-2. The basket contained

GE-NE-B1301 809-01 two impact (Charpy) specimen containers (numbers stamped on the containers: 131C7716G4 Ec 131C7716G5, respectively shown in Figure 3-3(a) and three tensile specimen capsules (numbers stamped: Gl, G2 and G3, respectively shown in Figure 3-3(b). During the removal of the Charpy impact specimens from the specimen holder, one specimen container was found to have leaked. The specimens were visually examined for features that could possibly affect test results.

The specimens appeared somewhat darker in appearance than the other specimens. This uniform discoloration was most likely caused by the exposure of the specimen to the high temperature water environment. The surfaces of the discolored specimens were similar to the other specimens, i.e., no defects, pits, or detrimental corrosion was observed. This is expected since the reactor water trapped inside the capsule was expected to be stagnant in an enclosed space where the activities of corrosive agent in the trapped reactor water to be at minimum. Based on these observations, it is was concluded that the specimens were not affected by the exposure to water, and will give credible surveillance results.

3.3 SPECIMEN DESCRIPTION The surveillance capsule holder contained 24 Charpy specimens: base metal (8), weld metal (8), and HAZ (8). There were 6 tensile specimens: base metal (2), weld metal (2), and HAZ (2). The holder contained 6 flux wires: 2 iron, 2 nickel, and 2 copper. The chemistry and fabrication history for the Charpy and tensile specimens are described below.

3.3.1 Charpy Specimens The fabrication of the Charpy specimens is described in the CBIN drawings of the surveillance test program. The test plate used for the surveillance were cut from the same material of the same heat as one of the beltline platest '1, which was Heat B5301-1. Sub-blocks were cut from the test plate for base, weld and HAZ test specimens. Two sub-blocks were welded together for the weld and HAZ specimens, using the weld electrode heat number of 3P4966 pj . The same weld electrode number was also used for one of the vertical welds in the lower No.l shell (between slabs 1 and 2) during the vessel assembly process. Thus, the test plate had gone through the same fabrication process of a quench and temper heat treatment immediately after hot forming, then submerged arc welding. Finally, the base metal, weld metal and HAZ test plates or sub-blocks had gone through the post weld heat treatment for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> at 1150'F + 25'F followed by furnace cooling to below 600'F then air cooled, in a manner that will simulate the actual heat treatment performed on the core region shell plate of the completed vessel. The Charpy and the tensile specimens were then machined from these sub-blocks, as

GE-NE-B1301 809-01 described in the CBIN surveillance test specimen document Pl and the GE surveillance program documents l . Charpy specimens were machined from the I/4 T and 3/4 T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction) and were stamped on one end with the fabrication codes 'nd on the other end the vessel code of 55 (WNP2). The base metal orientation in HAZ specimens was also in longitudinal direction.

3.3.2 Tensile Specimens The surveillance tensile specimens were fabricated of the same materials of the Charpy specimens described in the surveillance specimen drawings '

[7,9]

The materials, and thus the chemical compositions and heat treatments for the base, weld, and HAZ Charpy and tensile specimens are identical.

GE-NE-B1 301 8 -01 TABLE 3-1(a) CHEMICALCOMPOSITION 4 INITIALRTNDT OF RPV BELTLINEPLATE MATERIALS FROM WNP2 FSAR RECORDS I I Com ositionb Wei htPercent Initial Identification Heat/Lot No. CU Ni Mn Si Mo Rior

('I')

Lower Shell Plates:

21-1-1 C1272-'I 0.15 0.60 0.23 1.31 0.013 0.02

0.26 0.53 21-1-2 C1273-I 0.14 0.60 0.23 1.28 0.014 0.02 0.23 0.5? 20

'.50 21-1-3 C 1273-2 0.14 0.60 0.23 1.28 0.014 0.02 0.23 0.57 21-14 C1272-2 0.15 0.60 0.23 1.31 0.013 0.02 0.26 0.55 Lower-Intermediate Shell 22-1-1 B5301-I 0.14 0.20 1.34 0.017 0.01 0.23 0.52 -20

-8

'22-1-2 C1336-'I 0.13, 0;50 0.21 1.36 0.017 0.01 0.22 0.49 22-1-3

'2-Id C1337-I C1337-2 '.

'.15 0.15 0.51 031 0.22

0.22 ':: ...1.32 1.32 0.018 0.018 0.01 0.01 0.21 0.21 0.50 0.50

-20

-20 Note:

]1] Data from WNP2 Letter, Ref.[25].

10

GE-NE-BI3018 -01 TABLE3 1(b) CHEMICALCOMPOSITION & INITIALRTNPT OF RPV BELTLINEWELD MATERIALS FROM WNP2 FSAR RECORDS II Com ositionb Wei htPercent Initial

[t)

Identification PI Mn Si Mo

~

Heat/Lot No. Cu Ni RTm)r

( F)

Lower-Long. BA,BB, BD 04P046 0.06 0.90 0.044 1.04 0,009 0.021 0.40 0.58 0.02 -48 Lower - Long. BA, BB 0.016 0.48 0.54 -50 Lower-Long, BA-BD 07L669 3P4966 0.03 0.02 1.02 0.80 0.05 0.059 '.35 1.24 0.014 0.013 0.013 0.38 0.50 0.005 -30

-48 Lower-Long. BA-BD 3P4966 0.02 0.92 0.077 1.42 0.014 0.013 0.41 0.53 0.005 Lower-Long. BB, BC, BD C3L46C 0.02 0.87 0.063 0.96 0.019 0.017 0.32 0.53 -20 I.ower - I.ong. l3B 08M365 0.02 1.10 0.057 1.23 0.02 0.023 0.47 0.57 -48 l,ower I.ollg. If(. 091.853 0.03 0.86 0.052 1.23 0.023 0.46 0.51 -50 Lower - Lower Interm. BE-BH 3 F4966 0.03 0.88 0.074 1.38 0.010 0.013 0.36 0.49 0.006 -26 Lower - Lower Interm. BE-BH 3 F4966 0.03 0.90 0.067 1.39 0.011 0.014 0.38 0.53 0.008 -6 Long. BF, BH 04P046 0.06 0.90 0.044 1.04 0.009 0.021 0.40 0.58 0.02 Long. BF 05 F018 0.09 0.90 0.057 1.21 0.008 0.021 0.44 0.53 0.01 -38 Long. BG 624063 0.03 1.00 0.041 1.12 0.009 0.018 0.41 0.54 0.01 -50 Long, BH 624039 0.07 1.01 0.060 1.11 0.015 0.025 0,45 0.57 0.02 -50 Long. BG 624039 0.10 0.92 0.041 1.12 0.01 0.02 0.45 0.53 0.01 -36 Lower- Lower Interm. AB 492L4871 0.03 0.98 0.07 1.17 0.02 0.02 0.32 0.51 0.02 -50 Lower- Lower Interm. AB 5P6756 0.08 0.93 0.063 1.27 0.01 0.011 0.57 0.45 0.006 -50 Lower- Lower Interm. AB 5P6756 0.09 0.92 0.078 1.24 0.01 0.012 0.53 0.46 0.006 -50 Girth Weld AB 3P4955 0.025 0.90 0.035 1.33 0.016 0.011 0.56 0.52 0.006 -44 Girth Weld AB 3P4955 0.023 0.95 0.054 1.28 0.016 0.010 0.55 0.54 0.007 -16 Girth Weld AB 04T931 0.03 1.00 0.05 1.03 0.02 0.024 0.28 0.53 0.01 -50 Notes:

[I] Data from WNP2 Letter, Ref.[25]

[2] Weld Location Identification from WNP2 Power Uprate Report, Ref.[10]

11

~'

GE-NE-BI301 809-01 Vessel Flange Longitudinal Welds Upper Shell Girth Welds Intermediate Shell LPCI (3) ¹2 Shell Ring Lower Intermediate (¹2) Shell 22-1-1 B 5301 22-1-2 C1336 Core 22-1-3 C 1337 Beltline 22-1-4 C 1337 Re ion

¹I Shell Ring Lower (¹1) Shell let: 'l Recirc. Inlet (10) 21-1-1 C1272 Reclrc. Jet Pump Inst. (2) 21-1-2 C1273 Outlet 21-1-3 C1273 (2) 21-1-4 C1272 Bottom Head Enclosure FIGURE 3-1 RPV SCHEMATIC WITH BELTLINEPLATE IDENTIFICATION 12

GE-NE-B1301809-01 FIGURE 3-2 RECOVERED SURVEILLANCE CAPSULE BASKET 13

GE-NE-B1301 809-01 v

w FIGURE 3-3(a) CHARPY SPECIMEN CONTAINER IDENTIFICATION

.C FIGURE 3-3(b) TENSILE SPECIMEN IDENTIFICATION

GE-1VE-B1301809-01 4.

Samples were taken from the irradiated base and weld Charpy specimens aAer 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, Ni, Mo, 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 4-1 for both irradiated and baseline surveillance plate and weld materials. The baseline data were taken from CBIN CMTR material certification records for the plate and weld surveillance specimens 15

GE-NE-B 13018 9-0 TABLE 4-1 CHEMICALCOMPOSITION OF WNP2 SURVEILLANCE SPECIMENS FROM GE CHEMICALANALYSIS Specimen ID Metal Type CU (wt%)" Ni (wt%)tu Mn (wt%) Mo (wt%) Si(wt%) Cr (wt%) P (wt%)

(Heat No.)

29141(B5301) 0.12 0.51 1.31 0.53 0.12 0.18 0.010 29143 (B5301) Base 0.11 0.49 1.25 0.50 0.1 1 0.18 0.014 0.48 0.10 0.17 0.010 29146 (B5301)

Baseline"'9148

"'ase Base Base 0.1 0.14 1 0.47 0.50 1.21 1.34 0.52 0.23 NA 0.017 (3P4966) Weld 0.03 1.00 1.40 0.52 0.25 0.08 0.011 29125 (3P4966) Weld 0.03 0.86 1.23 0.42 0.21 0.06 0.008 29153 (3P4966) Weld 0.03 0.80 1.20 0.40 0.13 0.05 0.010 Baseline Weld 0.03 0.90 1.39 0.53 0.38 NA 0.011 Notes: [1] Average GE Compositions:

Base Cu: 0.11% Base Ni: 0.49%

Weld Cu: 0.03% Weld Ni: 0.89%

[2] From WNP2 FSAR, Ref.[3]

16

GE-NE-B1 301 809-01 5.

Flux wires removed from the 300'ocation capsule were analyzed, as described in Section 5.1, to determine flux and fluence received by the surveillance capsule. The lead factor, determined as described in Section 5.2, was used to establish the peak vessel fluence from the flux wire results. Section 5.3 includes 32 EFPY peak fluence estimates.

5.1 FLUX WIRE ANALYSIS 5.1.1 Procedure The surveillance capsule contained two sets of flux wires: each set with one each of iron, copper and nickel, totaling 6 flux wires. Each wire was removed from the capsule, cleaned with dilute nitric acid followed by rinses with water and acetone, weighed, mounted on a counting card, and analyzed for its radioactivity content by gamma spectrometry. Each iron wire was 54 ~ ~ 58 analyzed for Mn content, each nickel wire for Co and each copper wire for Co at a calibrated 4-cm, 10-cm and 20 cm source-to-detector distance with 35-cc and 100-ccGe(Li) gamma spectrometers.

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 Washington Public Power Supply System and are shown in Tables 5-1 and 5-2 respectively.

From the flux wire activity measurements and power history, reaction rates for Fe(n,p) "

58 58 Mn, Ni(n,p) Co and Cu(n,u) 60 Co were calculated. The cross sections for the iron, nickel and 63 copper wires are 0.213 barn, 0.273 barn and 0.00374 barn respectively. The values for Fe and Cu are based on the experimental data correlation which establishes the cross sections as a function of the water gap (between the RPV inside surface and the fuel radius). The cross section of Ni was calculated assuming its ratio to that of Fe is independent of the water gap, based on the 54 58 v~

observation that Fe and i~i are activated in the same energy range. In Browns Ferry Unit 3, which is also a 251 inch, 764 fuel bundle plant, the cross sections of Fe and Ni are 210 mb and 270 mb respectively' Therefore, for WNP 2 the o value of Ni is ratioed according to 213~270/210 which is 273 mb. These cz values are consistent with the cross section data established at GE's Vallecitos Nuclear Center based on more than 65 spectral determinations for 17

GE-NE-B1 301 809-01 BWRs and for the General Electric Test Reactor using activation monitors and spectrum unfolding

)

techniques. The cross sections for 0.1 MeV flux were determined from the measured 0.1 to 1 MeV cross section ratio of 1.6 I 5.1.2 Method of calculation:

I6,17]

The reaction rates can be calculated according to [IS, c1ps M (Eq 5-1) g <<Zp (I- "")( "")

where dps/g is the measured disintegration rate of the sample (S4 Mn, SS Co and 60 Co respectively) at end-of irradiation, M is the atomic weight of the element. A is Avogadro's number, a is the isotopic abundance of the target nuclide, p; is the full-power fraction for period i, using the information in Table 5-1, X, is the decay constant, t,; is the exposure (irradiation duration) time of period i, and td'; is the decay (elapsed) time between the end of period i and the end of irradiation or counting. Data from the cycle 1-11 power generation histories were used (Table 5-1). A summary R/o'Eq of these key parameters is provided in Table 5-3.

Using the reaction rate, the full power flux is calculated according to 4fp = 5-2).

where v is the reaction cross sections shown in Table 5-3.

The E) 1 MeV fluence (total) is then calculated according to 8, = Ctp/P;t; (Eq 5-3).

I where, 4fp = full power flux value from Eq 5-2 t; = operating time PI

= full power fraction where the summation equals 2626.2 days (with p; normalized at 3323 MWt and the normalization affects the results of Eq. 5-1 and Eq. 5-2 but not Eq. 5-3) based on the power history data in Table 5-2.

18

GE-NE-B1301 809-01 5.1.3 Results The fluence calculated according to Eq 5-3 is summarized in Table 5-4 which shows that the three wire fluences differ by less than 10%. For conservatism the maximum fluence (E >

'1 MeV) value of 1.55x10 n/cm for Ni is used. The accuracies of the values in Tables 5-3 for a 2 a deviation are influenced by the following sources of error:

Relative Error: +3%

Counting Rates: +1%

Power History: +15%

Cross Section: +10%

The overall 2o error is(3 +1 +15 + 10 ) =18%.

[18]

of the Cycle analysis,

~

The 11 cycle results can be compared with those 1 which included nominal and upper bound neutron fluxes (E > 1 MeV) of 4.8x10 and 6.0x10 n/cm -s, respectively. Comparing the flux result 6.85x10 n/cm -s of reported here, the difference may be accounted for by the following factors:

1. Cycle 1 fuel load may have included low enrichment of outer bundles.

2. The difference in capsule positions (Cycle 1 flux wire at vessel 30'zimuth and Cycle 11 wire at the 300'ocation) could account for some slight variation.

3. Cycle 1 operation could have had an unusual radial power distribution (averaged over the cycle)

4. Cycle 1 operation could have had an unusual axial power distribution (averaged over the cycle) 5.2 DETERMINATIONOF LEAD FACTOR The flux wires are used to detect neutron flux at the location of the surveillance capsule.

The wires will reflect the power fluctuations associated with the operation of the plant. However, the flux wires are not necessarily at the location of peak vessel flux. A lead factor relating the flux at the flux wire to the peak flux at vessel surface has to be determined in order to assess the irradiation level at the reactor vessel wall. Vessel ID lead factor is defined as the ratio of flux at the surveillance capsule to the peak flux (q~;d) at the vessel inner surface, i.e. (lead factor) x qi;d

=

capsule flux. Lead factor is a function of core configuration as well as vessel geometry. It is also dependent on the power density and coolant density distributions in the core.

19

GE-NE-BI30I 809-OI The lead factor for the WNP2 surveillance capsule was determined using Cycle 10 core data as basis in representing the power shape and void distribution of the core. WNP2 rated power

[19]1 was 3323 MWt until the end of Cycle 10. The two-dimensional transport code DORT1 was used to calculate the spatial flux distribution produced by a fixed source of neutrons in the core region.

DORT is a deterministic code using discrete-ordinates S~ method to solve the integro-differential form of the Boltzmann transport equation.

Two DORT calculations were performed to simulate R-6 and R-Z configurations of the reactor. The resultant flux distributions can be synthesized to obtain an equivalent R-6-Z flux distribution. Lead factor, or the ratio of flux at the capsule location to that at the peak flux location, can be determined from this distribution.

5.2.1 R-6 Calculation Flux distribution in the azimuthal direction was obtained by calculation in the R-6 geometry. The calculation model incorporates inner and peripheral core regions, the shroud, water regions inside and outside the shroud, and the reactor vessel wall. The core region material compositions and neutron source densities were typical of those at the core midplane. Core midplane is chosen for the R-6 calculation because it is representative for beltline flux evaluation, it is also the center elevation of the surveillance capsule specimen holder. Neutron cross-section data used in the calculation is a 26-group, spatial and composition dependent cross section set, which was condensed from the Los Alamos Scientific Laboratory (LASL) 80 group microscopic cross section data library. The spatial mesh of the calculation model consists of 155 radial intervals and 90 azimuthal intervals. Figure 5-1 is a schematic view of the R-6 model. Since WNP2 core has an eight-fold symmetry, only 1/8 of the core was modeled with reflective boundary conditions assumed at the 0'nd 45'oundaries. A reflective boundary is also assumed at R = 167 cm to save computation time. Because the central core region with R < 167 cm has little effect on the flux level at reactor vessel, this portion of the core need not be modeled in detail. Output of this calculation is the flux distribution as functions of azimuthal angle and radial distance. The magnitude of neutron flux at the surveillance capsule center can be read from the output file at 30'nd R = 319 cm. Figure 5-2 shows the angular (R-6) flux variation at the core midplane.

5.2.2 R-Z Calculation Neutron flux profile in the axial direction is determined by the R-Z calculation. The calculation model incorporates similar material regions as in the R-6 model. However, the full 20

GE-NE-B1301 809-01 length of the core was modeled with simulated data corresponding to those at core 30'zimuth, including material composition, radial dimension, coolant density, and power density distribution in the axial direction. Based on operating experiences, the axial profile of fast neutron flux near the core edge is not very sensitive to the radial distance nor the azimuthal direction of the axis.

Therefore other azimuthal directions could be used as representative of the core for R-Z calculations, the results would be similar. In our analysis, azimuth 30'f a 45'odel is chosen because it is equivalent to the surveillance capsule location. Figure 5-3 is a schematic view of the R-Z calculation model. There are 155 spatial meshes in the R-direction and 125 intervals in the Z-direction.

Output from the R-Z calculation provides flux variations as functions of elevation. A relative flux profile normalized to the peak vessel flux is given in Figure 5-4. The ratio of flux at core midplane to the peak flux can be readily estimated.

5.2.3 Results The result of R-e calculation demonstrated in Figure 5-2 shows that along the circumference of the reactor vessel, the flux peaks around 25'ast quadrant reference. The flux level at core midplane, at the location of surveillance capsule is 7.34x10 n/cm -sec, which differs from the measured flux of 6.85x10 n/cm -sec by 7%. The maximum vessel surface flux at core midplane can be read from the same figure as approximately 7.23x10 n/cm -sec.

The axial flux profile shown in Figure 5-4 indicates that along the longitudinal direction of vessel surface, flux peaks approximately 100 inches above the bottom of active fuel (BAF). The peaking factor from core midplane to the peak elevation is 1/0.932, or 1.073. Therefore, peak vessel surface flux is quantified as 7.23 x10 n/cm -sec x 1.073 = 7.76x10 n/cm -sec. Therefore the lead factor at the vessel surface is 7.34 x10 /7.76 x10 = 0.95. This value is very similar to the lead factor of 0.98 previously calculated for the first cycle flux wire measurement A similar procedure is applied to the flux results at vessel 1/4 T. The peak flux at vessel 1/4 T is determined to be 5.16x10 n/cm -sec. Therefore the lead factor at vessel 1/4 T is 7.34 x10 / 5.16 x10 =1.42. This value is comparable to the lead factor of 1.51 previously calculated for the first cycle flux wire measurement 21

GE-NE-B1301 809-01 5.3 EVALUATIONOF 32 EFPY FLUENCE Generally, the inside surface fluence (fsurf) at 32 EFPY is determined from the flux wire fluence at a particular EFPY and lead factor according to f;d~~ = (fcfc

• 32 EFPY)/(LF CEFPY) (Eq 5-5) where f;dm~= 32 EFPY fluence at the peak vessel inside surface f,f, = capsule fluence measured at the current CEFPY (WNP2: 1.55 x 10 n/cm )

32 EFPY = end of life EFPY based on a 40-year operation at an 80% capacity factor CEFPY = the current EFPY for the capsule (WNP2: 7.2 at 3323 MWt)

LF = lead factor (WNP2: 0.95)

Since the WNP2 was uprated for Cycle 11, the method of the fluence calculation of the above general equation has to be modified to include the uprated cycles. The value of CEFPY is the sum of 6.57 EFPY at 3323 rated thermal power (MWt) and 0.6 EFPY at 3486 MWt.

Assuming that WNP2 will be operated at 3486 MWt for the remaining cycles, 32 EFPY is the sum of 6.57 EFPY at 3323 MWt and 25.43 EFPY at 3486 MWt. Therefore the fluence (f>3')

cumulated at 3323 MWt can be ratioed from the effective full power days (EFPD) provided in Table 5-2:

F3+ f,f, x 2398.1/(2398.1+ 228.3) = 1.42 x 10'/cm And the fluence (f>4gg) cumulated at 3486 MWt is ratioed from f>3+.

fg4g6 f33+ x (25.43/6.57) (3486/3323) = 5.77 x 10'/cm Therefore the resulting 32 EFPY fluence value at the peak location of the vessel inside surface is:

f,g= (f33/3 + f34g6)/0.95 = 7.57 x 1Q n/cm The above peak surface fluence is higher than both the nominal fluence of 4.9x10'/cm and the upper bound fluence of 6.2x10'/cm from the dosimetry measurement of the first cycle

'he probable causes for the difference in fluence from the first cycle and the 11th cycle measurements have already been discussed in Section 5.1.3.

22

0 GE-NE-B1 301 809-01 Figure 5-4 shows that the fast neutron flux level drops off significantly near both ends (TAF and BAF) of the active fuel. The top of the WNF2 lower shell is at elevation 230 inches, or approximately 14 inches above BAF. From Figure 5-4, the flux level at this elevation is less than 80% of the peak flux. Since the lower shell has the most limiting RTNDT value, the conventional method of calculating the ART based on the peak fluence is too restrictive. Therefore the fluence level at the lower shell is conservatively adjusted to be 80% of the vessel peak fluence. The ART calculation provided in Section 8 is based on the adjusted flux value to account for the axial location of the lower shell relative to the active fuel region.

The fracture toughness analysis is based on 1/4 T depth flaw in the beltline region, so the attenuation of the flux to that depth is considered. The attenuation is calculated according to the RG1.99 equation:

(e-0.24x) (Eq 5-6) where x is the distance, in inches, from the inner vessel surface to the 1/4 T depth. Based on the vessel beltline plate thickness intermediate (No.2) shell of 7.5 inches for lower (No.l) shell and 6.44 inches for the lower-the corresponding peak 1/4 T fluences at 32 EFPY are 3.86x10'adjusted) and 5.14x10 n/cm, respectively.

23

GE-NE-B1301 809-01 TABLE 5-1

SUMMARY

OF DAILYPOWER HISTORY Cycle Begin End Duration (t) Mwd'ycle. Begin -., End::: ',. DUration {t) Mwd 5/27/84 3/16/86 659 1091606 6/9/95 6/13/95 370 6/4/86 4/10/87 311 723417 6/13/95 6/14/95 802 6/19/87 4/30/88 317 763293 6/14/95 7/3/95 19 6/19/88 4/29/89 315 793865 7/3/95 7/5/95 1875 6/25/89 4/21/90 301 844042 7/5/95 7/9/95 7521 8/4/90 4/13/91 253 736045 7/9/95 7/13/95 13605 9/26/91 4/18/92 206 496456 7/13/95 8/13/95 31 108172 7/5/92 5/1/93 301 764290 8/13/95 8/20/95 16850 6/19/93 6/22/93 673 8/20/95 11/3/95 75 '58193 6/22/93 6/24/93 2270 11/3/95 11/6/95 8555 6/24/93 6/26/93 2020 11/6/95 11/26/95 20 66250 6/26/93 6/27/93 3907 11/26/95 11/30/95 14340 6/27/93 6/28/93 3313 11/30/95 12/2/95 5591 6/28/93 7/16/93 18 59136 12/2/95 12/27/95 25 58252 7/16/93 7/17/93 3023 12/27/95 1/11/96 15 52195 7/17/93 8/3/93 17 54661 1/11/96 1/16/96 13291 8/3/93 8/16/93 6851 1/16/96 1/19/96 10367 8/16/93 1/14/94 151 502247 1/19/96 1/22/96 7429 1/14/94 1/16/94 4686 1/22/96 2/3/96 12 41712

'/16/94 2/27/94 42 137755 2/3/96 2/7/96 10043 2/27/94 3/7/94 26228 2/7/96 2/11/96 12670 3/7/94 3/8/94 647 2/11/96 2/14/96 9655 3/8/94 3/13/94 17332 2/14/96 2/25/96 25185 3/13/94 3/14/94 3445 2/25/96 2/29/96 12981 3/14/94 3/22/94 24974 2/29/96 3/1/96 233 3/22/94 3/26/94 12434 3/1/96 3/2/96 2550 3/26/94 3/27/94 3023 Note: Full power of Cycles 1 - 10 3323 MWt 3/27/94 3/30/94 8474 Full power of Cycle 'l1 3486 Mwt 3/30/94 4/26/94 27 62146 10 7/25/94 8/3/94 10 7136 2/1/95 182 603115 2/1/95 2/5/95 10822 2/22/95 17 43170 2/22/95 2/25/95 2536 2/25/95 2/26/95 5293 2/26/95 3/4/95 5620 3/4/95 4/5/95 32 106638 4/5/95 4/13/95 1686 4/13/95 4/22/95 29290 24

GE-NE-B1301 809-OI TABLE 5-2

SUMMARY

OF WNP2 IRRADIATIONPERIODS Rated .

Full: power Full Power Thermal -.

Cycle:; On. Off Duration Mwt-d Factor Power i days .

'(MW)"

5/27/84 3/16/86 659 1091606 328.5 0.498 3323 6/4/86 4/10/87 311 723417 217.7 0.700 3323 6/19/87 4/30/88 317 763293 229.7 0.725 3323 6/19/88 4/29/89 315 793865 238.9 0.758 3323 6/25/89 4/21/90 301 844042 254.0 0.844 3323 8/4/90 4/13/91 253 736045 221.5 0.875 3323 9/26/91 4/18/92 206 496456 149.4 0.725 3323 7/5/92 5/1/93 301 764290 230.0 0.764 3323 6/19/93 4/26/94 312 940409 283.0 0.907 3323 10 7/25/94 4/22/95 272 815464 245.4 0.902 3323 6/9/95 3/2/96 268 758554 217.6 0.812 3486 Total EFPD = 2615.7 Total EFPY = 7.2 Note: [I] Equivalent full power day for Cycle 11 normalized at 3323 MWt = 217.6 x 3486/3323 = 228.3.

Total EFPD for Cycles 1-11 normalized at 3323 Mwt = 2626.4, EFPY = 7.2 TABLE 5-3 DOSIMETER NUCLEAR PARAMETERS Target Isotopic Half-

) I MeV Dosimeter abundance Radionuclide X(d) cross section nucleus life

(%) (mb), (+ 2s) 54F Mn Fe 5.8 312.3 d 2.2195 E-3 213 +21 Ni 58N. 58C 68.3 70.82 d 9.7874 E-3 273 &27 63( 60(

Cu 69.2 5.271 y 3.6003 E-4 3.74 + 0.4 25

GE-NE-B130180 -01 TABLE 5-4 SURVEILLANCECAPSULE FLUX AND FLUENCE FOR IRRADIATIONFROM START-UP TO 3/2/96 (300'zimuth Capsule at 7.2 EFPY at full power of 3323 MWt)

A'vera'ge',. .','; Full Power,,Flux Full Power Flux ". Flueri'ce". :Fl'uence' "'.

-:"= Average, -':"

Wiie dps/g Element Rate ul eaction

'ri/cm2-'s) (n/cm2-'s)'&lJJHeY (n/cm2) (n/cm2)

E>UJHW F&1JJHeX Fe 6.85E+04 1.45E-16 6.83E+08 1.09E+09 1.55E+17 2.48E+17 Ni 1.13E+06 1.87E-16 6.85E+08 1.10E+09 1.55E+17 2.48E+17 Cu 8.28E+03 2.49E-18 6.65E+08 1.06E+09 1.51E+17 2.42E+17 Value used fiuencel l: 1.55E+17 Notes:

[l] Rounded offto two decimal points 2] Full power flux, based on thermal power of 3323 MWt

[3] 1.6 times the E >l MeV result

[4] Maximum of the three flux wire fluences 26

GE-NE-81301809-01 1 1 1 REFLECTIVE BOUNDARY 1 1 1 2 1 1 1 1 1 CORE IN TE RID 2 2 2 1 1 1 67 2 2 2 2 2 2 CORE INTERVALED EXTERIOR TOTAL 2 2 2 45'ATER WATER REGION CAPSULE SHROUD: 11 INTERVAI REGION:

52 INTERVALS VESSEL WALL:

90INTERVALS 25 INTERVALS IN AZIMUTHAL DIRECTION 1 ~ CORE INTERIOR FUE 2 ~ CORE EXTERIOR FU L po FIGURE 5-1 AZIMUTHALFLUX DISTRIBUTION ANALYSISMODEL 27

la L

GE-1VE-B13018 - 1 WNP2 Angular Flux Variation at Core Midplane (Based on 3323 MWt) 9e+8 Se+8 Capsule Location

= 7.34e8 7e+8 6e+8 E

) 5e+8 T

A x 4e+8 3e+8 2e+8 Capsule Radius Vessel inner Surface 1e+8 10 15 20 25 30 35 40 45 Angular Degree FIGURE 5-2 ANGULARFLUX VARIATIONAT CORE MIDPLANE 28

GE-NE-B1301 809-01 Centerlipe of Core To of Active Fuel I

Core Zone 49 Core Zone 50 I

Core Zone 47 Core Zone 48 I

Core Zone 45 Core Zone 46 I Core Zone 43 Core Zone 44 Water I

I I

I I

I I

I I Reflective I Bounda~

I I

I Capsul I Vacuum I

I I

I I

z Fuel Node I

I Core Zone 5 Core Zone 6 Core Zone 3 Core Zone 4

, ud Core Zone 1 Core Zone 2 Bottom of Active Fuel R=16 7cm FIGURE 5-3 R-Z CALCULATIONMODEL 29

GE-NE-B13018 - 1 WNP2 Axial Flux Profile at Vessel Surface Core Midplane W.932 20 80 Inches above BAF FIGURE 5-4 AXIALFLUX PROFILE AT VESSEL I.D. SURFACE 30

GE-NE-B1 301 809-01 The 24 Charpy specimens recovered from the surveillance capsule as well as the 36 unirradiated Charpy specimens 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 I l.

6.1 IMPACT TEST PROCEDURE The Vallecitos testing machine used for irradiated specimens was a Riehle Model Pl-2 impact machine, serial number R-89916. The maximum energy capacity of the machine is 240 ft-lb, which produces a test velocity at impact of 15.44 ft/sec. The test apparatus and operator were 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, 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 qualification of the Riehle machine and operator is summarized in Table 6-1.

Charpy V-Notch tests were conducted at temperatures between -100'F and 300'F. The cooling fluid used for irradiated specimens tested at 50'F or below was ethyl alcohol. At temperatures between 50'F and 200'F, water was used as the temperature conditioning fluid. The specimens were heated in silicon oil for test" temperatures above 200'F. Cooling of the conditioning fluids was done by heat exchange with liquid nitrogen; 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 thermocouple. 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, for the fractured specimens, photographs were taken of fracture surfaces. Lateral expansion and percent shear were measured.

Percent shear was established by first measuring the length and width of the cleavage surface in inches and then determined from Table 2 of ASTM E23-94b.

31

GE-NE-B1 301809-01 6.2 IMPACT TEST RESULTS Twelve Charpy V-Notch specimens each of unirradiated base, weld, and HAZ materials were used to define the baseline toughness transition and upper shelf portions of the fracture toughness curves. They were tested at temperature ranges of -100'F to 300'F. The absorbed energy, lateral expansion, and percent shear data are listed for each material in Table 6-2. Eight Charpy V-Notch specimens each of irradiated base, weld, and HAZ material were similarly tested in the temperatures range of -60'F to 300'F. During the test of an irradiated weld specimen at 50'F., the Charpy energy was inadvertently not recorded and the test was repeated at the same temperature. As shown later, the resulting shift in bRTNpT of the weld metal was not affected by the loss of one data point. The absorbed energy, lateral expansion, and percent shear data are listed for each irradiated material in Table 6-3. The key parameters of the index temperatures at Charpy Energy of 30 ft-lb, 50 ft-lb, lateral expansion at 35 mils and the USE of each material are tabulated in Table 6-4. The fracture surface photographs revealing shear area characteristics of each specimen as well as the individual test data are contained in Appendix A.

For each irradiated and unirradiated base plate, weld and HAZ material, the test data of absorbed Charpy energy and the lateral expansion are respectively fit with the hyperbolic tangent function developed by Oldfield for the EPRI Irradiated Steel Handbook Y = A + B

• TANH K T - To)/C] (Eq 6-1) where Y = impact energy or lateral expansion T = test temperature, and A, B, To and C are regression curve fit parameters.

The TANH function is one of the few continuous functions with a shape characteristic of low alloy steel fracture toughness transition curves. The usefulness of this particular relationship is apparent in the physical significance of curve fit parameters: A-B = lower shelf, A+B = upper shelf, To = mid-transition temperature and B/C = slope of the transition temperature region. The resulting transition temperature curves are presented in Figures 6-1 through 6-6.

32

h gl

GE-NE-B1301 809-01 6.3 COMPARISON OF THE MEASURED AND PREDICTED IRRADIATION SHIFTS The measured transition temperature shift for the base plate and weld material was compared to the predictions calculated according to RG1.99. The inputs and the calculated shifts are as follows:

MS'opper 0.11% 0.03%

Nickel 0.49% 0 89%

CF 73 41 Measured Fluence, f(at 7.2 EFPY) 1.55 x 10 n/cm 1.55 x 10 n/cm Calculated Fluence Factor:

ft0.28 - 0.10 lOg f) 0.147 0.147 Reg. Guide 1.99 b,RT>pz 11'F 6'F

'Eq Reg. Guide 1.99 ERTgpy+ 20'~ -23 to 45 'F -50 to 62'F Measured shift at 30 ft-Ib -1.0 'F -4.9 'F The weight percents of Cu and Ni are based on the average value of GE chemical composition measurements shown in Table 4-1. The CF shown above is the chemistry factor from RG1.99 Tables 1 and 2. The fluence factor is provided by RG1.99 as:

fluence factor= ft 6-2) which can either be calculated or read from Figure 1 of RG1.99. The predic'ted 30 ft-lb temperature shift (ERTdg was also calculated according to the following RG1.99 equation (Eq 6-3)

The measured 30 ft-ib temperature shifts of -1.0'F for the base plate and -4.9'F for the weld (Table 6-4) are well within the bounds of the RG prediction with the uncertainty of +2'. (o is the square root of the sum of o~ + o~ where a> = Standard deviation on initial RTgpr which is zero 2 2 and ag = Standard deviation on b,RT>pz, which is 28'F for welds and 17'F for base material.

33

e GE-NE-B1 301 809-01 6.4 CHANGE IN USE Based on the copper content and the fluence data provided in the previous sections, RG1.99 predicts decreases in the irradiated Charpy USE of approximately 8% for the base, and 6% for the weld material at the fluence of 1.55x10'/cm . The measured USE values for these materials however show increases of 3% and 12%, respectively for the base and weld material.

Upper Shelf Energy is expected to decrease due to irradiation. The amount of expected

. decrease is related to both copper content and fluence, which is relatively low (about an order of magnitude less than in PWRs) in BWRs. Both the copper content in the WNP2 vessel (0.11%)

and the fluence (7.57x10'/cm ) are relatively low, and therefore, materials may not experience significant decreases in USE. Experience has shown that, at relatively low fluence and considering typical scatter in Charpy data, BWR vessel material USE test results may show an increase. Given the typical scatter in Charpy data and the low fluence of the irradiated specimens, the increase in WNP2 plate and weld material is not unexpected.

34

GE-NE-B1 301 809-01 TABLE 6-1 VALLECITOS QUALIFICATIONTEST RESULTS USING NIST STANDARD REFERENCE SPECIMENS

. Specimen-- Bath

-- ',, Test; A'bsorbed

-Energy,

-.'emperature

.; Acceptable, Range Identification Medium ':=":.F:.':; 't-'.1b ft-ib Vallecitos HH-46 1 Ethyl Alcohol -40 74 Riehle Machine HH-46 2 Ethyl Alcohol -40 72.5 (tested 8/95) HH-46 3 Ethyl Alcohol -40 75.5 HH-46 4 Ethyl Alcohol -40 73 HH-46 5 Ethyl Alcohol -40 77

.Avera e 74:4: .

74.3+3.7 ass LL-45 1 Ethyl Alcohol -40 13 LL-45 2 Ethyl Alcohol -40 13 LL-45 3 Ethyl Alcohol 40 13 LL-45 4 Ethyl Alcohol -40 13 LL-45 5 Ethyl Alcohol -40

':<Avera e.. ': 13

'1:3'i'::::.) :12;8+1.0 ass SH-5 1 Ethyl Alcohol 70 170 SH-5 2 Ethyl Alcohol 70 170 SH-5 3 Ethyl Alcohol 70 162.5 SH-S 4 Ethyl Alcohol 70 161.5 SH-5 5 Ethyl Alcohol 70 156

'Avera e '164;0::.: '.:164.1+ 8.2 ass 35

GE-NE-B1 301 809-01 TABLE 6-2 UNIRRADIATEDCHARPY V-NOTCH IMPACT TEST RESULTS Fracture L'ateral:,

Specimen 'ath  ::

Test Tempera- En'ergy: Expansion Percent Identification.,', Medium  :;ture 'F.: ft-':-lb:>> . iiiils ".. Shear

(%)'ase:

Bl ETOH -60 5.0 11 1 Heat B5301-1, B2 ETOH -20 10.5 12 20 Longitudinal, B3 ETOH 10 17.0 20 34 B4 ETOH 20 38.0 35 34 B5 ETOH 30 27.0 26 37 B6 ETOH 40 41.0 37 38 B7 ETOH 60 77.5 63 45 B8 WATER 80 99.0 81 63 B9 WATER 108 113.0 86  : 69 B10 WATER 120 130.5 83 80 Bl 1 WATER 200 148.0 90 '00 B12 OIL 300 153.5 91 100 Weld:

Heat 3P4966 W1 ETOH -100 7.0 9 12 Lot 1214 W2 ETOM -60 9.0 14 14 Linde 124 Flux W3 ETOH -20 25.5 30 42 W4 ETOH -10 52.0 48 48 W5 ETOH 0 42.0 40 37 W6 ETOH 10 36.0 43 55 W7 ETOH 20 53.0 52 69 W8 ETOH 60 74.5 70 87 W9 WATER 108 89.0 76 97 W10 WATER 120 98.5 76 100 Wl 1 WATER 200 107.0 67 100 W12 OIL 300 97.0 91 100 HAZ: HAZ1 ETOH -60 28.0 17 20 Heat B5301-1, HAZ2 ETOH -50 16.0 16 18 Longitudinal and HAZ3 ETOH -20 48.0 37 37 Weld heat 3P4966 HAZ4 ETOH -10 34.0 35 41 Lot 1214 HAZ5 ETOH 0 44.5 44 70 Linde 124 Flux HAZ6 ETOH 20 83.0 62 65 HAZ7 ETOH 60 103.5 81 88 HAZ8 WATER 80 74.5 73 90 HAZ9 WATER 108 110.5 86 88 HAZ10 WATER 120 118.0 84 100 HAZ11 WATER 200 121.5 76 100 HAZ12 OIL 300 118.0 89 100 36

GE-NE-B1301 809-01 TABLE 6-3 IRRADIATEDCHARPY V-NOTCH IMPACT TEST RESULTS (for 300'zimuth Surveillance Capsule at 7.2 EFPY)

Specimen. Bath

';Test .. ':.-Fr'a'ctur'e..'.,Lateial"',:.';,."-' '.

Temp'era-.- Energy.,

':Expjansio'n','., Percent

Identification Medium .

-ture'F '= ft-'.lb ':-:. -'ils ':: "'-  :,"Shear:(%)

Base: 29146 ETOH 0 16.0 16 22 Heat B5301-1, 29141 ETOH 30 30.0 26 30 Longitudinal, 29143 ETOH 39 56.5 36 30 29145 ETOH 67 66.0 50 35 29140 WATER 100 109.0 73 56 29144 WATER 150 139.5 87 70 29139 WATER 200 154.0 85 96 29142 OIL 300 152.5 91 100 Weld: 29148 ETOH -20 43.5 36 31 Heat 3P4966 29153 ETOH -10 35.0 36 34 Lot 1214 29152 ETOH 22 50.0 48 54 Linde 124 Flux 29147 ETOH 50 NIA 71 96 29157 ETOH 50 72.0 66 67 29151 WATER 85 102.5 87 94 29149 WATER 100 104.0 76 91 29150 WATER 200 108.0 76 100 HAZ: 29157 ETOH %0 7.0 8 1 Heat B5301-1, 29159 ETOH -40 28.0 26 35 Longitudinal and 29158 ETOH -20 58.5 43 39 Weld heat 3P4966 29162 ETOH 8 62.0 45 43 Lot 1214 29156 ETOH 67 88.0 61 76 Linde 124 Flux 29160 WATER 105 105.5 77 91 29155 WATER 130 127.0 82 100 29161 WATER 200 127.0 81 100 37

GE-NE-B1301 809-01 TABLE 6-4 KEY PARAMETER

SUMMARY

OF IRRADIATEDAND UNIRRADIATEDCHARPY V-NOTCH IMPACT DATA (for 300'zimuth Surveillance Capsule at 7.2 EFPY)

",:-Index. Te'mp ('F) Index Temp ('F) Index USE Material '. '=30 ft-Ib E=50 ft-:Ib Temp;('F)'LE=35-mil (ft-Ib)

PLATE (B5301-1):

Unirradiated 22.2 43.7 31.4 150.2 Irradiated 21.2 45.8 41.6 154.8 Difference 2.1 (t) 10.2 (t') 4.6 (+3%t)

RG1.99, b,RTNpT = 11 'F RG1.99, (MTNpT+Ma) = -23 'F to 45 'F RG1.99, Decrease in USE = 8%

WELD (3P4966):

Unirradiated -19.6 15.8 -14.8 102.5 Irradiated -24.5 9.4 -13.1 114.6 Difference - 4.9 (4) 1.7 (1') 12 1 (y1 2% t)

RG1.99, BRTNpT = 6 'F RG1.99, (BRTNpT 220') = -50 'F to 62 'F RG1.99. Decrease in USE = 6%

HAZ:

Unirradiated -37.5 -3.8 -16.9 119.8 Irradiated -43.0 -6.0 -15.3 129.9 Difference -2.2 (4) 1.6 (1') 10.1 (+8% t) 38

GE-NE-8I301809-Ol FIGURE 6-1 IRRADIATEDAND UNIRRADIATEDCHARPY TESTS WNP2 Base Plate Impact Energy 160 140 120 Unirradiated rra iae Irradiated 100

/

Ul 80

/

Lu

/

Q.

E 60

/

40 USE change: + 4.6 ft-ibs.

30 ft-Ib RTndt Shift: - 1.0'F 20

-200 -100 100 200 300 400 Test Temperature, 'F 39

0 GE-NE-B1301809-01 FIGURE 6-2 IRRADIATEDAND UNIRRADIATEDCHARPY TESTS WNP2 Base Plate Lateral Expansion 100 90 80 Unirradiated /

70 Irradiated 60 C

0 g 50 CL X

uj L. 40 30 20 10

-200 -100 100 200 300 400 Test Temperature, 'F 40

GE-NE-BI30I809-OI FIGURE 6Q IRRADIATEDAND UNIRRADIATEDCHARPY TESTS WNP2 Weld Impact Energy 120 110 Irradiated 100 90 narra iae 80 Ch 70 CD 60 LQ 50 E

40 USE change:+ 12.1 ft-lbs.

30 30 ft-Ib RTndt Shift: - 4.9'F 20 10

-200 -100 100 200 300 400 Test Temperature, 'F 41

0 GE-NE-BI30I809-OI FIGURE 6-4 IRRADIATEDAND UNIRRADIATEDCHARPY TESTS WNP2 Weld Lateral Expansion 100 90 Irradiated 80 Unirradiated 70 lO 60 C

0 th m 50 CL X

lu 40 O

C 30 20 10

-200 -100 100 200 300 400 500 Test Temperature, 'F 42

GE-NE-B I 301 809-01 FIGURE 6-5 IRRADIATEDAND UNIRRADIATEDCHARPY TESTS WNP2 HAZ Impact Energy 140 Irradi ted I20 Unirradiate d 100 Ll C 80 U)

C)

C Lll g 60 Q.

E 40 I USE change:+ 10.1 ft-lbs.

20

/ 30 ft-Ib RTndt Shift: - 5.5'F

-200 -100 100 200 300 400 500 Test Temperature, 'F 43

GE-NE-B 130I809-01 FIGURE 6-6 IRRADIATEDAND UNIRRADIATEDCHARPY TESTS WNP2 HAZ Lateral Expansion 100 90 80 Unirra iated 70 th Irradiated 60 e

0 CO m 50 Q.

X uj 40 C$

30 20 10

-200 -100 100 200 300 400 500 Test Temperature, 'F 44

GE-NE-B1301 809-01 v.

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) and at an intermediate temperature of 150'F for the two additional base and weld specimens. The tests were conducted in accordance with ASTM E8-951[22]1.

7.1 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)/Ao (Eq 7-1)

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

7.2 RESULTS Irradiated tensile test properties of Yield Strength (YS), Ultimate Tensile Strength (UTS), Reduction of Area (RA), Uniform Elongation(UE), and Total Elongation (TE) are presented in Table 7-1. A stress-strain curve for a 550'F base metal irradiated specimen is shown in Figure 7-1. This curve is typical of the stress-strain characteristics of all the tested specimens.

45

GE-iVE-B1301 809-01 The surveillance materials generally follow the trend of decreasing properties with increasing temperature. Photographs of the necking behavior and the fracture surfaces are provided in Appendix B. As expected the data shows the general trend of increase in YS and UTS and decrease in TE with irradiation.

46

GE-NE-B1 301 809-01 TABLE 7-1 TENSILE TEST RESULTS FOR IRRADIATEDAND UNIRRADIATEDRPV MATERIALS (for 300'zimuth Surveillance Capsule at 7.2 EFPYj Test ., YIelda Ultima'te Uniform Total Reductio'n SpecIinen Temp. Strength Strength Elongation Elongation of Area Number',. 0 ksi ksi  ::(%)

Irradiated Base: P.1'-.'A.'::-.'.": 70 ':;. ".'."::.,. 59.6"':.i';: ':"'~ ', 84.3  :,. 12:5;- 21.6, 72.7 Pl-B 550 55.7 79.7 10.9 18.4 70.4 Irradiated Weld: . .P2-.A,.;-,. 70 =

. 67.3,, 83,9 12.8 22,3 68.4 P2-B 550 59.0 78.0 12.4 18.6 63.8 Irradiated HAZ: P3.-.A.... 70 56.8 81.3 12.2- 21.1 71.5 P3-B 550 60.3 80.7 11.5 18.1 65.9 Unirradiated Base: . Pl-:A-.;,: 70 ,"-56;5 .,:,': -..:.:.- 80.4 15.2,,  : 26.1.. 73.8 P 1-B 550 53.4 77.6 12.2 19.7 71.2 Unirradiated Weld: ,;-:. P,2-;A... =. 70,. ...'61.4 -.,=-,, '": 80;2. 13.2 ..: .. 22.1 68.2 P2-B 550 58.3 77.0 12.7 21.1 65.3 Unirradiated HAZ: .--. P3-A';,',.;.',70 ,. 53.4.:..'..;.:; -;.;:; -.78.2. 12.5 20.8 72.9, .

P3-B 550 58.1 78.2 10.7 18.3 66.1 a Yield Strength is determined by 0.2% offset.

47

0 GE-NE-BI301809-OI 100 WNP2 BASE 550'F 80 70 60 M 50 UJ L'

co 40'0 20 10 10 12 14 16 18 20 STRAIN, %

FIGURE 7-1 TYPICAL ENGINEERING STRESS-STRAIN FOR IRRADIATEDRPV MATERIALS 48

GE-NE-B1301 809-01 S.

The 32 EFPY peak fluence value of 7.57x10 n/cm in Section 5.3 is used to calculate the 32 EFPY 1/4 T fluence values of 3.86x10 n/cm and 5.14x10 n/cm for the lower shell and lower intermediate shell, respectively. It should be noted that the value of 3.86x10 n/cm for the lower shell represents 80% of the vessel peak fluence as discussed in Section 5.3. Based on these fluence values, the adjusted reference temperatures (ARTs) and upper shelf energy (USE) for the beltline materials are calculated.

S.1 ADJUSTED REFERENCE TEMPERATURE AT 32 EFPY The effect on adjusted reference temperature (ART) due to irradiation in the beltline materials is determined according to the methods in RG1.99, as a function of neutron fluence and the elemental compositions of copper (Cu) and nickel (Ni). The RG1.99 ART equation is:

ART = Initial RT>pz+ bRT>pz+ Margin (Eq 8-1) where b,RT =CF x ft 'pT (Eq 8-2) 2 0.5 Margin=2(a, +cr~) (Eq 8-3)

CF = Chemistry factor from Tables 1 or 2 of RG1.99, f = 1/4 T fluence (n/cm ) divided by 10 a< = Standard deviation on initial Ropy and is equal to zero in this case.

ay = Standard deviation on b,RT>pz, is 28'F for welds and 17'F for base material, except that erg need not exceed 0.50 times the b,RT>pr value.

Since the chemical composition measured by GE as described in Section 4 can be considered as an addition to the existing data, the chemistry used to calculate the ART is based on the average of the GE data and those reported in the FSAR. Therefore the chemistry for plate B5301-1 is 0.13%Cu and 0.49%Ni, which has a chemistry factor of SS. The chemistry for weld 3P4966 is 0.03%Cu and 0.90%Ni, which has a corresponding chemistry factor of 41. The composition for the remaining materials remains unchanged, same as those provided in Section 4.

Each beltline plate and weld b,RT>py value is determined by multiplying the CF from RG1.99 determined for the Cu-Ni content of the material, by the fluence factor for the EFPY being 49

GE-NE-B1301 809-01 evaluated. The Initial Ropy 5 Ropy and Margin are added to get the ART of the material. The 32 EFPY ART values for the beltline plate and weld are shown in Tables 8-1 and 8-2 respectively.

It should be noted that the WNP2 FSAR requires that, in order to satisfy the 10CFR50 Appendix G, initial Ropy is either NDT or transverse Charpy V-notch (CVN) 50 ft-Ib temperature minus 60'F. Since all the tests performed were from specimens with longitudinal orientation, the transverse CVN 50 ft-lb transition temperature has to be calculated from the existing data of longitudinal orientation in the following manner. The lowest longitudinal CVN energy, ifbelow 50 ft-lb, is adjusted to derive a longitudinal CVN 50 ft-lb transition temperature by adding 2'F per ft-lb to the test temperature. Ifthe actual data equals or exceeds 50 ft-lb, the test temperature is used. Once the longitudinal 50 ft-lb temperature is derived, an additional 30'F is added to account for orientation changed from longitudinal 50 ft-lb to transverse 50 ft-lb. For the vessel weld metal, the 30'F addition is omitted, since there is no principal working direction in weld metal. The initial Ropy for plate materials shown in Tables 8-1 are adjusted for the transverse orientation as required by the ASME Code after Summer 1972.

8.2 UPPER SHELF ENERGY AT 32 EFPY Paragraph IV.B of 10CFR50 Appendix G sets limits on the upper shelf energy of the beltline materials. The USE must be above 50 ft-Ib at all times during plant operation; assumed here to be up to 32 EFPY. However, the WNP2 vessel materials did not have USE data taken at fabrication, the calculation of 32 EFPY USE provided in Table 8-3 is for the beltline materials represented by the surveillance specimens only. The equivalent transverse USE of the plate material shown in the table is taken as 65% of the longitudinal USE, according to USNRC MTEB 5-2 P>}. Unlike the plate, the weld metal USE has no transverse/longitudinal correction because weld metal has no orientation effect.

Since the USE evaluation in Table 8-3 is limited to the surveillance materials and in order to show that the remaining beltline materials also satisfy the 10CFR50 Appendix G 50 ft-Ib limit, WNP2 has been evaluated against the BWR Owners'roup Equivalent Margin Analyses. From Tables 3-1(a) & (b), the limiting copper content of the beltline plate and weld materials are 0.15%

and 0.1%, respectively. Based on this, the RG1.99 Figure 2 predicts a decrease in the USE approximately by 12% for both the. plate and the weld at the 32 EFPY 1/4 T Quence of 5.14x10 n/cm . The Equivalent Margin Analysis shows that the USE decreases are bounded by the allowed limits of 21% and 34%, respectively for the plate and weld materials. Thus, the analysis demonstrates that the 10 CFR 50, Appendix G safety requirements are satisfactorily met 50

GE-NE-B1 301 809-01 for WNP2. The Owners'roup Program Report was submitted to the NRC in December 1993 and approved by SER on December 8, 1993.

51

GE-NE-B1 1809-01 TABLE S-1 BELTLINEBASE PLATE ART FOR 32 EFPY AT 1/4 T Lower (No.1) Shell Lower (No.l) Shell Plate Thickness = 7.5 inches 32 EFPY with Peak I.D. fluence = 7.57E+17 n/cm"2 32 EFPY with Adjusted I/4 T fluence = 3.86E+17 n/cm 2 Lower Intermediate (No.2) Shell Lower Intermediate (No.2) Shell Plate Thickness = 6.44 inches 32 EFPY with Peak I.D. flucncc = 7.57E+17 n/cm"2 32 ill'I'Ywith Peak I/4 T flucncc = 5.14E+17 n/cm"2 Girth Weld Girth Weld Weld Thickness= 6.44 inches 32 EFPY Peak I/4 T Weld fluence ~ 5.14E+17 n/cm"2 Fluence Initial 32 EFPY 32 EFPY

%Cu %Ni CF ARTndt Margin Shift ART COMPONENT IIEAT No. Factor RTndt oF OF oF oF oF Lower Shell with Adjusted Fluence[tl 21-1-1 C 1272-I 0.15 0.60 110 0.25 28 27.9 14.0 27.9 55.8 83.8 21-1-2 C1273-I 0.14 0.60 100 0.25 20 25.4 12.7 25.4 50.8 70.8 21-1-3 C1273-2 0.14 0.60 100 0.25 4 25.4 12.7 25.4 50.8 54.8 21-1-4 C 1272-2 0.15 0.60 110 0.25 0 27.9 14.0 27.9 55.8 55.8 Lower-Interm. Shell Peak Fluence 21-1-1 B5301-I I'I 0.13 0.49 88 0.30 -20 26.2 13.1 26.2 52.3 32.3 22-1-2 C1336-I 0.13 0.5 88 0.30 -8 26.2 13.1 26.2 52.3 44.3 22-1-3 C 1337-I 0.15 0.51 105 0.30 -20 31.2 15.6 31.2 62.5 42.5 22-IA C1337-2 0.15 0.51 105 0.30 -20 31.2 15.6 31.2 62.5 42.5 Notes:

[I] Lower shell fluence adjusted to 80% of the peak value to consider axial flux distribution (see Figure 54).

[2] Wt% of Cu and N i is the average of the data measured by GE and those listed in TABLE 3-l(a) 52

TABLE 8-2 BELTLINEWEL RT FOR 32 EFPY AT 1/4 T GE-NE-B1301809-01 Lower (No.1) Shell Lower (No.l) Shell Plate Thickness = 7.5 inches 32 EFPY with Peak I.D. fluence = 7.57E+17 n/cm"2 32 EFPY with Peak I/4 T fluence = 3.86E+17 n/cm"2 Lower Intermediate (No.2) Shell Lower Intermediate (No.2) Shell Plate Thickness = 6.44 inches 32 EFPY with Peak I.D. fluence = 7.57E+17 n/cm"2 32 EFPY with Peak I/4 T fluence = 5.14E+17 n/cm"2 Girth Weld Girth Weld Weld Thickness= 6.44 inches 32 EFPY Peak I/4 T Weld fluence = 5.14E+17 n/cm"2 Fluence Initial 32 EFPY 3? EFPY 32 EFPY COMPONENT HEAT No. %Cu %Ni CF Factor RTndt ARTndt Margin Shift ART oF oF OF oF oF Lower-Long. BA,BB,BD 04P046 0.06 0.90 82 0.25 -48 20.8 10.4 20.8 41.6 -6.4 Lower-Long. BA, BB 07L669 0.03 1.02 41 0.25 -50 10.4 5.2 10.4 20.8 -29.2 I.ower-Long. BA-BD 3P4966 0.025 0.895 34 0.25 -30 S.6 4.3 8.6 17.3 -12.7 I.ower-l.ong. l3A-Ill) ttI

"'P4966 0.025 0.895 34 0.25 -48 8.6 4.3 8.6 17.3 -30.7 I.ower-l.ong. 13B,IIC.BD C31.46C 0.02 0.87 27 0.25 -20 6.9 3.4 6.9 13.7 -6.3 Lower-Long. BB 08M365 0.02 1.10 27 0.25 -48 6.9 3.4 6.9 13.7 -34.3 Lower-Long. BC 09L853 0.03 0.86 41 0.25 -50 10.4 5.2 10.4 20.8 -29.2 Lower-Lower Int. BE-BH 3P49661'I 0.025 34 0.30 -26 10.1 5.1 10.1 20.2 -5.8 Lower-Lower Int. BE-BH 3P49661 I 0.025'.885 0.895 34 0.30 -6 10.1 5.1 10.1 20.2 14.2 Long. BF,BH 04P046 0.06 0.90 82 0.30 -48 24.4 0 12.2 24.4 48.7 0.7 Long. BF 05P018 0.09 0.90 122 0.30 -38 36.3 0 IS.I 36.3 72.5 34.5 Long. BG 624063 0.03 1.00 41 0.30 -50 12.2 ,0 6.1 12.2 24.4 -25.6 Long. BH 624039 0.07 1.01 95 0.30 -50 28.2 0 14.1 28.2 56.5 6.5 Long. BG 624039 0.10 0.92 135 0.30 -36 40.1 0 20.1 40.1 80.2 44.2 Lower - Lower Int. AB 492L4871 0.03 0.98 41 0.30 -50 12.2 6.1 12.2 24.4 -25.6 Lower - Lower Int. AB 5P6756 0.08 0.93 122 0.30 -50 36.3 18.1 36.3 72.5 22.5 Lower - Lower Int. AB 5P6756 0.09 0.92 122 0.30 -50 36.3 I S. I 36.3 72.5 22.5 Girth Weld AB 3P4955 0.025 0.90 34 0.30 A4 10.1 5.1 10.1 20.2 -23.8 Girth Weld AB 3P4955 0.023 0.95 34 0.30 -16 10.1 5.1 10.1 20.2 4.2 Girth Weld AB 04T931 0.03 1.00 41 0.30 -50 12.2 6.1 12.2 24.4 -25.6 Notes:

]I] Lower shell fluence adjusted to 80% of the peak value to consider axial flux distribution (see Figure 54)

]2] The data provided by GE shown in Table 4-1 are considered here as additional data points for the chemical composition provided in Table 3-1(b).

Therefore for 3P4966 weld, the average value of GE and Table 3-l(b) data is used.

53

GE-NE-B1 301 809-01 TABLE 8-3 UPPER SHELF ENERGY ANALYSIS FOR WNP2 BELTLINEMATERIALS REPRESENTED BY THE SURVEILLANCE SPECIMENS Location -'nitial:Long. Initial Trans '~/o:,Cu ';De'crease'in 32:EFPY (Heat '-:USE ft-lb- 'USE (A-'lb SE:1 'rans..USE".

1 Base Plate 150 98 0.11% 10% 88 22-1-1 (B5301-1)

Weld Lower-Int 103 0.03% 9.5% 93 BE-BH (3P4966)

Note: [1] Based on the peak 1/4T fluence of 5.14xl0 n/cm . The percentages were estimated from Figure 2 of RG1.99. The curve for 0.05%Cu was used for the weld.

54

GE-NE-B1301 809-01 9.

, Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

[21 , Appendix G to Section XI of the 1989 ASME Boiler & Pressure Vessel Code.

, with amendments, June 1983

[4] , Appendix H to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

[5]

E185-82, March 1982.

V, Annual Book of ASTM Standards,

[6] , USNRC Regulatory Guide 1.99, Revision 2, May 1988.

[7] CBIN, V , Fabrication Report P.O.

8205-AE023, Contract: 872-2647

[81 V, GE Document 21A8731, Rev. 1

[9] - Manufactures Data reports and Vessel Certifications, P.O.

No. 205-AE023, Contract: 872-2647 (December 1974 & July 1981)

[10] , GE-NE-208-17-0993, Rev. 1

[11] T. A. Caine to D.M. Kelly,

~gghgm, March 31, 1993

[12] T. A. Caine to D.M. Kelly,

~+~, May 5, 1993.

55

0 GE-NE-B1 301 809-01

[13] G.C. Martin, e D~~es, November 11, 1993 (FMT Transmittal 93-212-0045)

[14] Martin, G.C., GENE, San Jose, CA, August 1980, (GE Report NEDO-24793).

[15] Standard Test Methods for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron, Annual Book of ASTM Standards, E263-93.

[16] Standard Test Methods for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel, Annual Book of ASTM Standards, E264-93.

[17] Standard Test Methods for Measuring Fast-Neutron Reaction Rates by Radioactivatiod.of Copper, Annual Book of ASTM Standards, E523-93.

[18] T. A. Caine, t

W'eptember 1986, (MDE-98-0986).

[19] GE DORT Code - a part of Code Package CCC-543: TORT:

dgyad tgdg N C Igd wgCk i gg'Cdg Information Center.

[20] , Annual Book of ASTM Standards, E23-94b.

[21] , EPRI Report NP-4797, September 1986.

[22] , Annual Book of ASTM Standards, E8-95.

[23] , USNRC Branch Technical Position MTEB 5-2, Revision 1, July 1981.

[24] Letter from James T. Wiggins to Mr. Lesley A. England,"

56

GE-NE-B1301 809-01 V," USNRC, Washington, D.C., December 8, 1993.

[25] Letter No. G02-95-235 (Docket No. 50-397) from J.V. Parrish to USNRC, "~~

", November 2, 1995.

57

GE-NE-B1301 809-Ol APPENDIX A UNIRRADIATEDAND IRRADIATEDCHARPY SPECIMEN FRACTURE SURFACE PHOTOGRAPHS Photographs of each Charpy specimen fracture surface were taken per the requirements of ASTME185-82. The fracture surface photographs along with a summary of the Charpy test results for each unirradiated and irradiated specimen are provided in this Appendix. The pictures are arranged in the order of base, weld, and HAZ materials.

GE-NE-BI301 809-01 APPENDIX A - UNIRRADIATEDSPECIMEN FRACTURE SURFACE APPEARANCE

~PM%~ q7 BASE: B1 BASE: B7 Test Temp: -60'F Test Temp: 60'F Energy: 5.0 ft-lbs Energy: 77.5 ft-lbs Lateral Exp: 11 mils Lateral Exp: 63 mils v Shear: 1% Shear: 45%

BASE: B2 BASE: BS Test Temp: -20'F Test Temp: 80'F Energy: 10.5 ft-lbs Energy: 99.0 ft-Ibs Lateral Exp: 12 mils Lateral Exp: 81 mils Shear: 20% Shear: 63%

BASE: B3 BASE: B9 Test Temp: 10'F Test Temp: 108'F Energy: 17.0 ft-lbs Energy: 113.0 ft-lbs Lateral Exp: 20 mils Lateral Exp: 86 mils Shear: 34% Shear: 69%

BASE: B4 BASE: B10 Test Temp: 20'F Test Temp: 120'F Energy: 38.0 ft-lbs Energy: 130.5 ft-lbs Lateral Exp: 35 mils Lateral Exp: 83 mils Shear: 34% Shear: 80%

BASE: B5 BASE: B11 Test Temp: 30'F Test Temp: 200'F Energy: 27.0 ft-ibs Energy: 148.0 ft-Ibs Lateral Exp: 26 mils Lateral Exp: 90 mils Shear 37% Shear: 100%

BASE: B6 BASE: B12 Test Temp: 40'F Test Temp: 300'F Energy: 41.0 ft-lbs Energy: 153.5 ft-lbs Lateral Exp: 37 mils Lateral Exp: 91 mils Shear: 38% Shear: 100%

GE-NE-BI301 809-01 APPENDIX A - UNIRRADIATEDSPECIMEN FRACTURE SURFACE APPEARANCE WELD: Wl WELD: W7 Test Temp: -100'F Test Temp: 20'F Energy: 7.0 ft-lbs Energy: 53.0 ft-lbs Lateral Exp: 9 mils Lateral Exp: 52 mils Shear: 12% Shear: 69%

t I WELD: W2 WELD: WS Test Temp: -60'F Test Temp: 60'F Energy: 9.0 ft-lbs Energy: 74.5 ft-lbs Lateral Exp: 14 mils Lateral Exp: 70 mils Shear: 14% Shear: 87%

WELD: W3 WELD: W9 Test Temp: -20'F liI 4jiil" Test Temp: 108'F Energy: 25.5 ft-lbs j Energy: 89.0 ft-lbs Lateral Exp: 30 mils Lateral Exp: 76 mils Shear: 42% Shear: 97%

WELD: W4 WELD: W10 Test Temp: -10'F Test Temp: 120'F Energy: 52.0 ft-Ibs Energy: 9S.S ft-lbs Lateral Exp: 4S mils Lateral Exp: 76 mils Shear: 4S% Shear: 100%

WELD: W5 WELD: W11 Test Temp: O' Test Temp: 200'F Energy: 42.0 ft-lbs Energy: 107.0 ft-lbs Lateral Exp: 40 mils Lateral Exp: 67 mils Shear: 37% '

Shear: 100%

WELD: W6 WELD: W12 Test Temp: 10'F Test Temp: 300'F Energy: 36.0 ft-lbs Energy: 97.0 ft-lbs Lateral Exp: 43 mils Lateral Exp: 91 mils Shear: 55% Shear: 100%

A-3

GE-NE-B1 301 809-01 APPENDIX A - UNIRRADIATEDSPECIMEN FRACTURE SURFACE APPEARANCE k'AZ'AX1 Test Temp: -60'F Energy: 28.0 ft-lbs HAZ: HAZ7 Test Temp: 60'F Energy: 103.5 ft-lbs Lateral Exp: 17 mils Lateral Exp: 81 mils Shear: 20% Shear: 88%

HAZ: HAZ2 HAZ: HAZS Test Temp: -50'F Test Temp: 80'F Energy: 16.0 ft-1bs Energy: 74.5 ft-lbs Lateral Exp: 16 mils Lateral Exp: 73 mils Shear: 18% Shear: 90%

HAX: HAZ3 @CA>> k rtQ h HAZ: HAZ9 Test Temp: -20'F Test Temp: 108'F Energy: 48.0 ft-Ibs Energy: 110.5 ft-lbs Lateral Exp: 37 mils Lateral Exp: 86 mils Shear'7% Shear: 88%

HAZ: HAZ4 HAZ: HAZ10 Test Temp: -10'F Test Temp: 120'F Energy: 34.0 ft-lbs Energy: 118.0 ft-Ibs Lateral Exp: 35 mils Lateral Exp: 84 mils Shear: 41% Shear: 100%

HAX: HAZS HAZ: HAZ11 Test Teinp: O' 'est Temp: 200'F Energy: 44.5 ft-lbs Energy: 121.5 ft-lbs Lateral Exp: 44 mils Lateral Exp: 76 mils Shear: 70% Shear: 100%

HAZ: HAZ6 HAZ: HAZ12 Test Temp: 20'F Test Temp: 300'F k~ Q y~+ 1*

Energy: 83.0 ft-lbs Energy: 118.0 ft-lbs Lateral Exp: 62 mils Lateral Exp: 89 mils k4 PPk Shear: 65% Shear: 100%

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GE-NE-B130I809-OI APPENDIX A - IRRADIATEDSPECIMEN FRACTURE SURFACE APPEARANCE BASE: 29146 BASE: 29140 Test Temp: O' Test Temp: 100'F Energy: 16.0 A-lbs Energy: 109.0 A-ibs Lateral Exp: 16 mils Lateral Exp: 73 mils Shear: 22% ~gPPyk Shear: 56%

BASE: 29141 BASE: 29144 Test Temp: 30'F Test Temp: 150'F Energy: 30.0 ft-lbs Energy: 139.5 ft-lbs Lateral Exp: 26 mils Lateral Exp: 87 mils Shear: 30% Shear: 70%

BASE: 29143 BASE: 29139 Test Temp: 39'F Test Temp: 200'F Energy: 56.5 ft-lbs Energy: 154.0 ft-lbs Lateral Exp: 36 mils Lateral Exp: 85 mils Shear: 30% Shear: 96%

BASE: 29145 BASE: 29142 Test Temp: 67'F Test Temp: 300'F Energy: 66.0 ft-ibs Energy: 152.5 ft-lbs Lateral Exp: 50 mils Lateral Exp: 91 mils Shear: 35% Shear: 100%

GE-NE-B 1301809-Ol APPENDIX A - IRRADIATEDSPECIMEN FRACTURE SURFACE APPEARANCE a+0 WELD: 29148 WELD: 29151 Test Tem~ -20'F Test Temp: 85'F Energy -lbs Energy: 102.5 ft-Ibs Latenl+ /36 mils Lateral Exp: 87 mils Shear: 31% Shear 94%

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WELD: 29153 WELD: 29149 Test Temp: -10'F Test Temp: 100'F Energy: 35.0 ft-lbs Energy: 104.0 ft-lbs Lateral Exp: 36 mils Lateral Exp: 76 mils Shear: 34% Shear'1%

WELD: 29152 WELD: 29150 Test Temp: 22'F Test Temp: 200'F Energy: 50.0 ft-Ibs Energy: 108.0 ft-lbs Lateral Exp: 48 mils Lateral Exp: 76 mils Shear: 54% Shear: 100%

WELD: 29154 Test Temp: 50'F Energy: 72.0 ft-lbs Lateral Exp: 66 mils Shear: 67%

A-6

GE-NE-B 1301 809-01 APPENDIX A - IRRADIATEDSPECIMEN FRACTURE SURFACE APPEARANCE HAZ: 29157 HAZ: 29156 Test Temp: -60'F Test Temp: 67'F Energy: 7.0 ft-Ibs Energy: 88.0 ft-Ibs Lateral Exp: S mils Lateral Exp: 61 mils Shear'% Shear: 76%

HAZ: 29159 HAZ: 29160 Test Temp: -40'F Test Temp: 105'F Energy: 28.0 ft-Ibs Energy: 105.5 ft-Ibs Lateral Exp: 26 mils Lateral Exp: 77 mils Shear: 35% Shear: 91%

HAZ: 29158 HAZ: 29155 Test Temp: -20'F Test Temp: 130'F Energy: 5S.5 ft-Ibs Energy: 127.0 ft-Ibs Lateral Exp: 43 mils Lateral Exp: 82 mils Shear 39% Shear: 100%

HAZ: 29162 HAZ: 29161 Test Temp: 8'F Test Temp: 200'F Energy: 62.0 ft-Ibs Energy: 127.0 ft-Ibs Lateral Exp: 45 mils Lateral Exp: 81 mils Shear: 43% Shear: 100%

GENE-B 1301 809-01 APPENDIX 8 UNIRRADIATEDAND IRRADIATEDTENSILE SPECIMEN FRACTURE APPEARANCE The necking behavior and the fracture appearance of the tensile test specimens are provided in this Appendix. The pictures are arranged in the order of Base, Weld and HAZ materials at the test temperatures of 70' and 550'F, respectively.

GE-NE-B1301 809-OI APPENDIX B - UNIRRADIATEDTENSILE SPECIMEN FRACTURE APPEAI4&lCE l

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GE-NE-B1301 809-01 APPENDIX B - UNIRRADIATEDTENSILE SPECIMEN FRACTURE APPEARANCE Unirradiated Base at 550'F

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Unirradiated HAZ at 550'F B-3

GE-NE-B1301809-01 APPENDIX 8 - IRRADIATEDTENSILE SPECIMEN FRACTURE APPEARANCE Irradiated Base at 70'F

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OPERATIONAL QUALITYASSURANCE PROGRAM DESCRIPTION PROPOSED

'REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Attachment 1 Page 1 of 8 Section 1.3.3.1 is being changed to correct the reference from the WNP-2 Technical Specifications to Appendix III of the Operational Quality Assurance Program Description (OQAPD) as the location for the requirements for the Plant Operations Committee (POC).

The revision to Section 1 - ORGANIZATION is an editorial change and the section will remain at Revision 25.

Table 2-1 of Section 2 - QUALITY ASSURANCE PROGRAM is changed to list new Site Wide Procedures that implement the Quality Assurance Program. This is an editorial change.

APPENDIX I- QUALIFICATIONREQUIREMENTS APPENDIX 1, Revision 12, defines the minimum qualification requirements for the Manager, Quality and Supervisor, Quality Services as:

Manager, Quality

a. Education: Bachelor Degree or equivalent* in Engineering or a related science.

b. Experience: Ten (10) years experience in the field of quality assurance, or equivalent number of years of nuclear industry experience in a management position or a combination of the two. The requirement that the manager have at least two years of experience in the administration of and adherence to the Quality Assurance Program in a significant management role directly involving nuclear power plants is being deleted.

Supervisor, Quality Services

a. Education: Bachelor Degree or equivalent* in Engineering or a related science.

b. Experience: Four (4) years in the field of quality assurance, or equivalent number of years of nuclear plant experience in a supervisory position, preferably at an operating nuclear plant, or a combination of the two. At least one (1) of these four (4) years of experience shall be nuclear power plant experience in the implementation of the quality assurance program.

~Equivalency will be determined based upon an evaluation of the following factors:

1. High School diploma or GED.

2. Sixty (60) semester hours of related technical education taught at the college level (900 classroom or instructor conducted hours).

3. Qualified as an NRC- senior operator at the assigned plant,

4. Four (4) years of additional experience in his area of responsibility.

5. Four (4) years of supervisory or management experience.

6. Demonstrated ability to communicate clearly (verbally and in writing).

7. Certification of academic ability and knowledge by corporate management.

8. Successful completion of the Engineer-In-Training examination.

9. Professional Engineer License.

10. Associated degree in Engineering or a related science.

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4 OPERATIONAL QUALITYASSURANCE PROGRAM DESCRIPTION PROPOSED REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Page 2 of 8 The Supply System proposes to modify the Qualification Requirement as follows:

"The Manager, Quality or the Supervisor, Quality Services fulfills the posidon described in ANSI/ANS-3.1-1978, Section 4.4.5, Quality Assurance. The qualifications of this position are:

a. Education: Bachelor Degree or equivalent* in Engineering or a related science.

b. Experience: Six (6) years experience in the field of quality assurance, or equivalent number of years of nuclear industry experience in a supervisory/management position or a combination of the two. At least two (2) years of these six years experience shall be nuclear power plant experience in the overall implementation of the quality assurance program. (This experience shall be obtained within the quality assurance organization.)

This proposed revision clearly specifies the Qualification Requirements for these two positions.

This revision would bring the qualification requirements into alignment with NUREG-0800, Standard Review Plan, Section 17.2, Quality Assurance During The Operations Phase, which states, "The qualifications of the QA Manager should be at least equivalent to those described in Section 4.4.5 of ANSI/ANS-3.1-1978, Selection and Training of Nuclear Power Plant Personnel, as endorsed by the regulatory positions in Regulatory Guide 1.8."

Allowing either position to meet the requirements of ANSI/ANS-3.1-1978 provides Supply System management the fiexibility to use one of the positions for management rotation, to expand on the individual's knowledge base. This will assure the Supply System continues to strengthen the knowledge and experience of individuals moving into top management positions. There will always be at least one individual, the Quality Manager or the Supervisor, Quality Services with the required qualifications and the necessary knowledge for the position.

CONCLUSION: The Supply System has concluded that this revision to Qualification Requirements is a reduction of a current commitment. However, the change will continue to meet the requirements of 10 CFR 50, Appendix B and ANSI/ANS-3.1-1978, Section 4.4.5.

Wolf Creek Nuclear Operating Corporation has a similar qualification statement in their FSAR Section 13.1.2.4, which states in part, ".. The Supervisor Quality Evaluation or the Manager Performance Improvement and-Assessment fulfills the position described in ANSI/ANS 3.1-1978, 4.4.5, Quality Assurance..."

APPENDIX II - POSITION STATEMENTS An editorial change is being made to page 1 to correct the reference to the Supply System Quality Department.

OPERATIONAL QUALITYASSURANCE PROGRAM DESCRIPTION PROPOSED REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Page 3 of 8 APPENDIX III- ADDITIONALQUALITYPROGRAM REQUIREMENTS APPENDIX III, Revision 0, identifies additional quality program requirements that were formerly located in the WNP-2 Technical Specifications, Section 6.0, Administrative Controls.

PR P AL 1 REVIEW AND APPR VAL F PR RAM AND PR ED RES The Supply System proposes to modify the POC procedure review responsibilities for nuclear safety related procedures and procedure changes and assign the responsibility to the line organizations. POC will continue to perform safety reviews associated with procedures that are of safety significance.

The proposed changes to the OQAPD, Appendix III reduce the administrative burden on POC by establishing a procedure review and approval process which shifts more responsibility for the review of nuclear safety related procedures and procedure changes from POC to the line organizations. Currently, all Technical Specification required procedures and changes thereto must be reviewed by POC. Instead, these items will be reviewed and approved through a new procedure review and approval process.

The new process for procedure review and approval shall be controlled by administrative procedures. The process requires. that all nuclear safety related procedures and procedure changes be reviewed by two designated technical reviewers, qualified licensing basis impact determination (LBID), including 10 CFR 50.59, preparer and reviewer and approved by a responsible procedure owner. The designated technical reviewers verify the technical accuracy and usability of procedures and revisions including human factor considerations. A qualified preparer performs the 10 CFR 50.59 screening and, as necessary, safety evaluation associated with the procedure or procedure change which is then independently reviewed by a qualified reviewer.

The following changes reflect the implementation of the procedure review and approval process:

a. The proposed change adds the use of "Qualified Procedure Reviewer" in the procedure review process. LBID/technical reviewers will perform reviews of procedures and procedure changes prior to approval by the responsible procedure owner.

b. A new proposed Section (4.0) which sets forth the principal provisions of the review and approval of programs and procedures. The review and approval of programs and procedures is outlined below and shall be controlled by an administrative procedure as required by Appendix A of Regulatory Guide 1.33. OQAPD Appendix III, Section 4.1 requires that an administrative procedure be maintained and established covering procedure review and approval activities.

OPERATIONAL QUALITYASSURANCE PROGRAM DESCRIPTION PROPOSED REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Page 4 of 8

1. Each program and procedure required by Technical Specification 5.4 and other procedures that affect nuclear safety, and changes thereto, shall be reviewed by a Qualified Procedure Reviewer who is knowledgeable in the functional area affected, but is not the individual who prepared the procedure or procedure change. The Qualified Procedure Reviewer shall meet or exceed the qualifications described in Section 4 of ANSI N18.1-1971, for applicable positions with the exclusion of the positions identified in Sections 4.3.2 and 4.5. Individuals whose positions are described in Sections 4.3,2 and 4.5 may qualify as a Qualified Procedure Reviewer provided they meet the qualification described in other positions of Section 4. All required cross-disciplinary reviews of new procedures, procedure revisions or changes thereto shall be completed prior to approval.

2. Each program and procedure required by Technical Specification 5.4 and other procedures that affect nuclear safety, and changes thereto, will be reviewed by a qualified LBID preparer and reviewer. The qualified LBID preparer and reviewer evaluate the procedure from a safety perspective. This reviewer completes the 10 CFR 50.59 safety screening and evaluation.

3. Each new procedure or proposed procedure change required by Technical Specification 5.4 shall be reviewed by the procedure sponsor. The procedure sponsor is to ensure that the proposed activity has been prepared, documented, and reviewed in accordance with the administrative procedure governing the procedure review and approval process. The procedure sponsor is responsible for the procedures technical accuracy and usability including human factor considerations.

4. Ifa safety evaluation is not required, the new procedure or procedure change shall receive final approval by the procedure sponsor prior to implementation. POC review is not required.

5. If a safety evaluation is required, the procedure or procedure change shall be reviewed by POC. The new procedure or procedure change shall receive final approval by the appropriate member(s) of management, as determined by the Plant General Manager.

6. Proposed changes to the Process Control Program and the Offsite Dose Calculation Manual must be reviewed by POC and accepted by the Plant General Manager prior to implementation whether-or not a safety evaluation is necessary.

CONCLUSION: The proposed change to use a Qualified Procedure Reviewer is considered to be an improvement to the process. The approval of procedures and programs by methods other than by POC and the Plant General Manager is considered a reduction in commitments.

The proposed change in OQAPD, Appendix III, Section 4.5 allows for an appropriate responsible member of management to approve procedures rather than the Plant General

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OPERATIONAL QUALITYASSURANCE PROGRAM DESCRIPTION PROPOSED REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Page 5 of 8 Manager. This section is changed to allow for the implementation of the proposed procedure review and 'approval process. Nuclear safety related procedures and procedure changes will be reviewed and approved, prior to implementation, by the appropriate member(s) of management, as determined by the Plant General Manager. If the procedure requires a safety evaluation, then the evaluation shall be reviewed by POC. POC will recommend approval or disapproval of the evaluation to the Plant General Manager.

The proposed change in OQAPD, Appendix III, Sections-3,1 and 3.2 allows for temporary procedure changes to be reviewed per the procedure review and approval process and approved by the procedure sponsor within 14 days. POC review of the temporary procedure change may or may not be required depending on whether or not the procedure change has a safety evaluation associated with it. The procedure sponsor is responsible for ensuring that the intent of the procedure is not changed. The wording in Section 3.2.b. is also being changed to clarify that the approval of a temporary change must be approved by the supervisor in charge of the shift. This clarification is to align the process with the requirements of ANSI N 18.7-1976, Section 5.2.2 Procedure Adherence. The current OQAPD, Appendix III, Section 3.2, required each temporary procedure change to be reviewed by POC and approved by the Plant General Manager.

CONCLUSION: The proposed change for approval of temporary procedure changes is considered a reduction in commitments.

The proposed procedure review and approval process requires that POC review safety evaluations associated with procedures and procedure changes. These changes alter the scope of POC's review function by eliminating reviews of items that do not impact the environment or nuclear safety and, therefore, enhance POC's function by improving its efficiency and effectiveness. POC can then focus its attention on matters which could affect nuclear safety.

The proposed process establishes flow and documentation requirements for the review and approval of all nuclear safety related procedures and procedure changes. The program builds on a safety evaluation process which is implemented through an administrative procedure which meets the requirements of 10 CFR 50.59.

Each procedure or procedure change will be reviewed by a qualified reviewer. The "review" is a two step process comprised of: 1) a Licensing-Basis Impact Determination and 2) a 10 CFR 50.59 Safety Evaluation. The Licensing Basis Impact Determination screen consists of a series of questions whose responses determine whether or not a Safety Evaluation is required.

The results of the screening process and Safety Evaluation will be documented in writing. If the review for a new procedure or procedure change concludes that a safety evaluation is required, then the evaluation(s) shall be prepared and then reviewed by POC. Personnel completing the screening process and safety evaluation will be appropriately qualified and trained to perform this function.

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OPERATIONAL QUALITYASSURANCE PROGR'AM DESCRIPTION PROPOSED REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Page6of 8 The completed procedure or procedure change package is reviewed by a Qualified Procedure Reviewer. The Qualified Procedure Reviewer shall meet or exceed the qualifications described in Section 4 of ANSI N18.1-1971 for applicable position with the exclusion of the positions identified in Sections 4.3.2 and 4.5. Individuals whose positions are described in Sections 4.3.2 and 4.5 may qualify as a Qualified Procedure Reviewer provided they meet the qualifications described in other portions of Section 4. They can be the same individual as the qualified LBID preparer/or reviewer if qualified to perform both functions. The Qualified Procedure Reviewers are knowledgeable in the functional/technical subject matter related to the proposed activity and are designated by the Department Managers. Qualified Procedure Reviewers shall be responsible for reviewing the procedure or procedure change for adequacy, completeness, and accuracy. They shall also be responsible for identifying whether or not cross disciplinary reviews are required.

After all the necessary reviews (LBID and procedure) have been completed, the procedure or procedure change package is reviewed by the procedure sponsor. The procedure sponsor is responsible for ensuring that the proposed activity has been prepared, documented and reviewed in accordance with the administrative procedure that governs the procedure review and approval process. The procedure sponsor is responsible for the technical accuracy and usability of procedures and revisions, including human factor considerations.

If a safety evaluation is not required, the new procedure or procedure change receives final approval from the procedure sponsor. POC review is not required. If however, a safety evaluation is required, POC is responsible for reviewing the safety evaluation and.

recommending to the Plant General Manager approval or disapproval of it. The procedure change still receives approval from the procedure sponso'r.

The Supply System implemented the bases of this procedure review program in January 1996, with the exception that POC has continued review all procedures and procedure changes. All requirements currently specified for POC and procedure review and approval remained in place.

Changes similar to those proposed for the Review and Approval of Programs and Procedures in the OQAPD, Appendix III, Section 2.1.1 were found acceptable by the NRC and approved for Indian Point Nuclear Generating Unit No. 3, for Wolf Creek Generating Station, Unit No.

1, and Palo Verde Nuclear Generating Station, Unit Nos, 1-, -2; & 3.

PR P AL2 PLANT PERATI NS MMITI'EE 0 MP ITI N OQAPD, Appendix III, Sections 2.1.1 and 2.1.2 will be modified to specify the POC composition in terms of the following plant functional areas rather than by designating members by organizations:

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OPERATIONAL QUALITYASSURANCE PROGRAM DESCRIPTION PROPOSED REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Page 7 of 8 Operations Radiation Protection Maintenance Technical Services Engineering Chemistry Quality Planning/Scheduling/Outage Administrative Services The proposed change would require the Plant General Manager appoint, in writing, a Vice Chairman 'and individual members experienced in the designated functional areas. The qualifications of all members shall meet the minimum requirements of ANSI/ANS-3.1-1981, Section 4.7, and cumulatively have expertise in the designated functional areas noted above.

The OQAPD, Appendix III, Section 2.1.1 currently defines the POC composition and designates POC members by specific organizational titles. The proposed changes will delete the position titles and replace them with the functional areas to be represented in POC. POC serves as a multi-disciplinary review and advisory organization to the Plant General Manager on matters related to nuclear and radiological safety. The proposed method of designating POC composition will provide the Plant General Manager with the flexibility to appoint qualified individuals from disciplines within a functional area. The POC composition change will also alleviate the need to process OQAPD changes for future organizational changes involving organizational .position titles, while maintaining the requirement for a multi-disciplinary POC membership.

The proposed change to OQAPD, Appendix III, Section 2.1.2 establishes that the Plant General Manager will appoint, in writing, the POC Vice Chairman, and individual members from each of the designated functional areas. POC members and alternate qualifications specified as Section 4.7 of ANSUANS-3.1-1981 assures that the education and experience level of each member meets or exceeds the current qualification requirements specified in Technical Specification 5.3.1 for those individuals who fill organizational positions considered part of the unit staff in Technical Specification 5.3.1 . This will ensure POC will continue to be staffed by qualified personnel having a variety of expertise and experience.

CONCLUSION: This change to POC composition is not considered to be a reduction of commitments.

Changes similar to those proposed for POC composition in OQAPD, Appendix III, Sections 2.1.1 and 2.1.2 were found acceptable by the.NRC and approved for the Kewaunee Station, for Fermi 2 Station, and for St. Lucie Plants 1 and 2.

PR P AL RB MEMBER HIP AND R P N IBILITIES There are two changes to the Corporate Nuclear Safety Review Board (CNSRB) responsibilities.

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OPERATIONAL QUALITYASSURANCE PROGRAM DESCRIPTION PROPOSED REVISION 26, DESCRIPTION OF CHANGES AND JUSTIFICATION Page 8 of 8 The proposed change to OQAPD, Appendix III, Section 2.2.2 would clarify the number of members required for the CNSRB. The present number of nine would be modified to read, "The CNSRB shall be composed of at least nine and no more that twelve members appointed..." Section 2.2.6 would also be modified to add, "The quorum shall consist of not less than the majority of the members, or duly appointed alternates." This means for:

9members -aquorumis5 10 members - a quorum is 5 ll members - a quorum is 6 12 members - a quorum is 6 Appendix III, Sections 2.2.9.a. & 2.2.9.b. are being clarified to allow the distribution of CNSRB records within 15 working days, rather than 1'4 days. This is considered an administrative change to the process and it does not have an impact on the review results.

These changes will continue to meet the requirements of ANSI N18.7-1976, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, Section 4.3.2. Standing Committees Functioning as Independent Review Bodies.

CONCLUSION: These changes are not considered a reduction of commitments.