ML20248L409

Analysis of Seabrook Stations Unit 1,Reactor Vessel Surveillance Capsules U & Y

ML20248L409
Person / Time
Site:	Seabrook
Issue date:	05/31/1998
From:	Biemiller E, Solan G Duke Engineering & Services
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20248L411	List: ML20248L409
References
DES-NFQA-98-01, DES-NFQA-98-1, NUDOCS 9806110139
Download: ML20248L409 (150)

Text

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DES-NFOA-98-01 l

l ANALYSIS OF SEABROOK STATION UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULES U AND Y l g DukeEngineering&Services l

l E. C. Biemiller G. M. Solan 1 i

l DUKE ENGINEERING & SERVICES - BOLTON, MA May 1998 A KO 43 P PDR l

Prepared by: -a [ 88 E. C. Blemiller Date Metallurgist, Project Manager 11 Technical Services Prepared by: $[8fN G/A. Sol $n / Date Project Manager 11 Reactor Physics Group Reviewed by: < /1/>o / b 6l 1i VAIL. Paliulis Date Jngineer Plant Engineering Services Reviewed by b b P FP

[. Caccicpou[ ' Ddte Manager Reactor Physics Group Duke Engineering & Services 580 Main Street Bolton, Massachusetts 01740

DISCLAIMER OF RESPONSIBILITY l This document was prepared by Duke Engineering & Services (DE&S). The use of information l contained in this document by anyone other than DE&S, or the Organization for which this document was prepared under contract, is not authorized, and with resoect to any unauthorized use. neither DE&S nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained h this document.

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TABLE OF CONTENTS I

L Page DISCLAIMER OF RESPONSIBILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . lil

l. TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv I

1 LI ST O F TABLE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi l

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LI ST O F FIG U R ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xi 1.0 SUMM ARY OF RESU LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 . I NTRO D U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 SURVEILLANCE MATERIALS AND CliEMISTRY . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.0 PRE- AND POST-IRRADIATION TEST RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 l i

4.1 Baseline Data, WCAP-10110 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.2 Irradiated Data, B&W Report BAW-2157 and Framatome Report BAW-2316 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

6 4.3 Tanh Curve Fits . . . ............................................ 6 4.4 Re s u l ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 5.0 RT 7 CALC U LATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 5.1 - Calculated RT.7 Values . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 5.2 Comparison of Surveillance Capsule Results to Regulatory Guide 1.99, Revision 2 Predictions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.3 Effect of Surveillance Results on Current Technical Specifications . . . . . . . 22

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f l TABLE OF CONTENTS (continued)

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l 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY . . . . . . . . . . . . . . . . . . . . . . 24 L 6.1 In troduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 1

6.2 Pressure Vessel and Surveillance Capsule Region Geometry . . . . . . . . . . . . 25 6.3 Discrete Ordinates Analysis Modeling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 6.4 Discrete Ordinates Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 6.5 Neutron Dosimetry and Uncertainties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 6.6 Best-Estimate Reactor Vessel Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 6.7 Projections of Reactor Vessel Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 7.0 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE . . . . . . . . . . . . . . . . . . . . 73 8.0 R E FE R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 APPENDICES A UNIRRADIATED VESSEL PLATE AND WELD DATA B B&W CAPSULE U TEST RESULTS REPORT BAW-2157 C FRAMATOME CAPSULE Y TEST RESULTS REPORT BAW-2316, REVISION 1

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LIST OF TABLES Number Title Page 4-1 Seabrook Surveillance Capsule Data Per ASTM E-185 and Regulatory Guide 1.99. Initial Values (Fluence = 0) .

1 Per WCAP-10110 ............................................... 8 4-2 Seabrook Station Unit 1 Surveillance Program Tensile Properties Base Line Properties Data Per WCAP-10110 and Irradiated Data Per BAW-2157 and BAW-2316. . . . . . . . . . . . . . . . . . . . . . . . . 9 '

l 4-3 A Comparison of Tanh Computed Charpy Values to l Westinghouse and B&W Reported Values . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 G-1 Seabrook Station Unit 1 Calculated Neutron Exposure Parameters at the Center of the Surveillance Capsule Locations . . . . . . . . . . . . . . . . . . . 34 l 6-2 Seabrook Station Unit 1 Calculated Azimuthal Variation of Neutron Exposure Parameters at the Reactor Vessel Clad / Base Metal Interface . . . . . . . . . . . . 35 6-3.1 Seabrook Station Unit 1 Relative Radial Distribution of Neutron Exposure Parameters within the Reactor Vessel Wcil, Cycle 1 Results . . . . . . . . . . . . 36 6-3.2 Seabrook Station Unit 1 Relative Radial Distribution of Neutron Exposure Parameters within the Reactor Vessel Wall, Cycle 4 Results . . . . . . . . . . . . 37 1

6-3.3 Seabrook Station Unit 1 Difference in Relative Radial Distribution of Neutron Exposure Parameters within the Reactor Vessel Wall, Change from Cycle 1 to Cycle 4 Relative Radial Distributions . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38

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l LIST OF TABLES (Continued)

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. Number Title Page 6-4.1 Seabrook Station Unit 1 Calculated Fast Neutron Flux, Fluence and Lead Factors for Capsules U and X at the Reactor Vessel Clad / Base Metal Interface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 ....

i l 6-4.2 Seabrook Station Unit 1 Calculated Fast Neutron Flux, Fluence and l Lead Factors for Capsules V and Y at the Reactor Vessel Clad / Base Metal Interface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 6-5.1.1 Seabrook Station Unit 1 Calculated Fast Neutron Energy Spectrum at the Center of Capsules U and X . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 6-5.1.2 Seabrook Station Unit 1 Calculated Epithermal / Thermal Neutron l

Energy Spectrum at the Center of Capsules U and X . . . . . . . . . . . . . . . . . . . 42  !

! 6-5.2.1 Seabrook Station Unit 1 Calculated Fast Neutron Energy Spectrum at the Center of Capsules V and Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43

6-5.2.2 Seabrook Station Unit 1 Calculated Epithermal / Thermal Neutron Energy Spectrum at the Center of Capsules V and Y . . . . . . . . . . . . . . . . . . . 44 l

6-6 Nuclear Parameters for Neutron Flux Monitors . . . . . . . . . . . . . . . . . . . . . . . . 45 j- 6-7 Seabrook Station Unit 1, Cycles 1 through 5 Operation, Operating Dates, Days and Effective Full-Power Years . . . . . . . . . . . . . . . . . . . . . . . . . . 46 i

6-8 Seabrook Cycle 1 Power History for Capsule U Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 1 . . . . . . . . . . 47

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LIST OF TABLES (Continued)

Number Title Page 6-9.1 Seabrook Cycle 1 Power History for Capsule Y Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 5 . . . . . . . . . . 48 6-9.2 Seabrook Cycle 2 Power History for Capsule Y Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 5 . . . . . . . . . . 49 6-9.3 Seabrook Cycle 3 Power History for Capsule Y Activation )

Ratio of Measured to Saturated Activities at the End-of-Cycle 5 . . . . . . . . . . 50 ,

6-9.4 Seabrook Cycle 4 Power History for Capsule Y Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 5 . . . . . . . . . . 51 6-9.5 Seabrook Cycle 5 Power History for Capsule Y Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 5 . . . . . . . . . . 52 6-10.1 Seabrook Station Unit 1 Capsule U Activation Analysis to Determine Average Fast Neutron Flux (E>1 MeV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 6-10.2 Seabrook Station Unit 1 Capsule Y Activation Analysis to Determine ,

Average Fast Neutron Flux (E>1 MeV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 6-11.1 Seabrook Station Unit 1 Capsules U and Y - Summary of Calculated and Measured Reaction Rates for Fast Flux Nuclides . . . . . . . . . . . . . . . . . . 55 l

6-11.2 Callaway Unit 1 Capsules U, Y and V - Summary of Calculated and Measured Reaction Rates for Fast Flux Nuclides in WCAP-14895 . . . . . 56,57

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L j LIST OF TABLES l

(Continued) l Number Title Page 6-12 Seabrook Station Unit 1 Capsules U and Y - Summary of Fast Neutron Flux and Fluence (E>1 MeV) Results at Capsule Centor -

Calculated and Measured . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 6-13.1 Seabrook Station Unit 1 Best-Estimate Neutron Exposure Rate Parameters at the Surveillance Capsules and within the Reactor Vessel Wall - Average through 0.913 EFPYs for End-of-Cycle 1 using Capsule U Calculated-to-Measured Ratio of 0.896 . . . . . . . . . . . . . . . . 59

'6-13.2 Seabrook Station Unit 1 Best Estimate Neutron Exposure Rate Parameters at the Surveillance Capsules and within the Reactor Vessel Wall - Average through 5.572 EFPYs for End-of-Cycle 5 using Capsule Y Calculated-to-Measured Ratio of 0.893 . . . . . . . . . . . . . . . . 60 6-14.1 Seabrook Station Unit 1 Best-Estimate Neutron Integrated Exposure Parameters at the Surveillance Capsules and within the Reactor Vessel Wall- At 0.913 EFPYs for End-of-Cycle 1 using Capsule U Calculated to-Measured Ratio of0. 896 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6-14.2 . Seabrook Station Unit 1 Best-Estimate Neutron Integrated Exposure Parameters at the Surveillance Capsules and within the Reactor Vessel Wall - At 6.5i2 F9Ys for End-of-Cycle 5 using Capsule Y Calculated-to-Measured Ratio of0.893 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 6-15: Seabrook Station Unit 1 Best-Estimate Neutron Integrated Exposure Parameters at the Surveillance Capsules and within the Reactor Vessel Wall- Projected at 32 EFPYs using Capsule Y Calculated-to-Measured Ratio of 0. 89 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63

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LIST OF TABLES (Continued)

- Number Title Page 6 16 Seabrook Station Unit 1 Best-Estimate Fast Neutron Fluence (E>1 MeV) Projections at the Surveillance Capsules and within the 4 Reactor Vessel Wall - Projected for 18-Month Cycles to 32 EFPYs using Capsule Y Calculated-to-Measured Ratio of 0.893 . . . . . . . . . . . . . . . . 64

.i 1 Seabrook Station Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 l

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l LIST OF FIGURES I

' Number Title Page 4-1 Tensile Properties for Seabrook Station Unit 1 Reactor Vessel .

I Plate R1808-3 Longitudinal Prop. Unirradiated and Surveillance '

Capsules U and Y Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 i

l 4-2 Tensile Properties for Seabrook Station Unit 1 Reactor Vessel Plate R1808-3 Transverse Prop. Unirradiated and Surveillance

. Capsules U and Y Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4 Tensile Properties for Seabrook Station Unit 1 Reactor Vessel Weld Metal Unirradiated and Surveillance Capsules U and Y Data . . . . . . . . 13 l

4-4 Seabrook Station Unit 1 Capsule U and Capsule Y Surveillance Results l 1

Plate R1808-3 Longitudinal Charpy Impact Data . . . . . . . . . . . . . . . . . . . . . . 14 j

'4-5 Seabrook Station Unit - 1 Capsule U and Capsule Y Surveillance Results Plate R1808 3 Transverse Charpy impact Data . . . . . . . . . . . . . . . . . . . . . . . 15 4-6 Seabrook Station Unit 1 Capsule U and Capsule Y Surveillance Results l Weld Metal Charpy impact Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 l

l 4-7 . Seabrook Station Unit 1 Capsule U and Capsule Y Surveillance Results

Heat-Affected-Zone Charpy impact Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 I

5-1 Seabrook Station Unit 1 Capsules U and Y Surveillance Results l  : for Plate and Weld Metal versus Regulatory Guide 1.99 Predictions . . . . . . . 23 L

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I LIST OF FIGUP.ES (Continued)

Number Title Page 6-1 Plan and Elevation Views of the Irradiation Surveillance Test Capsules in the Seabrook Station Unit 1 Reactor Vescel . . . . . . . . . . . . . . . . 65 6-2 Plan View of the Dual Irradiation Surveillance Test Capsules U and V in the Seabrook Station Unit 1 Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . 66 6-3 Seabrook Station Unit 1 - Assemblies for Pinwise Detailed Fission Source Modeling in DORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 6-4 Seabrook Station Unit 1 - Calculated and Measured Assembly Relative Powers - Cycle 1 Average, Assemblies Providing Fission Source Neutrons near Capsule Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 1

6-5 Seabrook Station Unit 1 - Calculated and Measured Assembly Relative Powers - Cycle 1-5 Average, Assemblies Providing Fission Source Neutrons near Capsule Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 6-6 Seabrook Station Unit 1 - Measured Assembly Relative Powers -

Cycle 1-5 Average and Cycle 1 Average, Assemblies Providing Fission Source Neutrons near Capsule Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 I l

6-7 Equation for Nuclide Product Activity in Surveillance Capsules . . . . . . . . . . . 71 6-8 Equations for Saturated Nuclide Product Reaction Rates . . . . . . . . . . . . . . . 72 I

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ANALYSIS OF SEABROOK STATION UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULES U AND Y l

1.0

SUMMARY

OF RESULTS CAPSULE U Capsule U, the first Seabrook Station, Unit 1, Reactor Vessel Surveillance Capsule, was removed from the Seabrook vessel in August 1991 after 0.913 effective full power years (EFPYs) of operation. The analyses of the capsule test specimens and neutron dosimetry led to the following conclusions:

o The capsule received an average neutron fast fluence (E>1 MeV) of 0.28 x 10 n/cm2 from operation of Cycle 1. This is equivalent to the fluence which was received at the reactor vesselinner diameter at approximately 4 EFPYs of operation.

o The reactor vessel lower shell plate material, R1808-3, was included in the surveillance capsule as the limiting plate material. For its Charpy specimens oriented in the longitudinal direction (LT), the 30 and 50 ft lb transition temperatures increased by 36*F and 34*F, respectively. The plate's transversely (TL) oriented Charpy specimens experienced increases in the 30 and 50 ft-lb transition temperatures u: 28'F and 20*F, respectively. The shift in the 35 mils-lateral-expansion (MLE) Index temperature was 24.5'F for LT specimens and 15'F for the TL specimens.

o The weld metalirradiated to 0.28 x 105 n/cm2 e::perienced 30 ft-lb and 50 ft-lb transition temperature increases of 10*F and 15'F, respectively.

o The avercqe upper shelf energy for transversely oriented specimens from lower shell plate R1808 3 decreased from 79 ft-lbs to 72 ft-lbs after irradiation to the fluence of 0.28 x 10

2 n/cm . The weld metal, exposed to the same fluence as the plate material, experienced a decrease in upper shelf energy from 160 ft-lbs to 129 ft-Ibs. The plate and weld materials exhibited upper shelf energies for continued safe plant operation. The upper shelf energy

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for these materials is expected to be maintained above 50 ft-lb throughout vessel life as required by 10CFR50, Appendix G.

o The adjusted RT, values for the plate and weld material, based on the surveillance capsule data, are within the two standard deviations of Regulatory Guide 1.99, Revision 2,

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predictions.

l CAPSULE Y j

. 1 Capsule Y, the second Seabrook Station, Unit 1 Reactor Vessel Surveillance Capsule, was l removed from the Seabrook vessel in May 1997 after 5.572 EFPYs of operation. The analyses of l the second capsule test specimens and neutron dosimetry led to the following conclusions:

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o The capsule received an average neutron fast fluence (E>1 MeV) of 1.15 x 10 n/cm from

. operation of Cycles 1 through 5. At the current rate, this is equivalent to the fluence which will be received at the reactor vessel inner diameter after approximately 15 EFPYs of operation.  !

.o The reactor vessel lower shell plate material, R1808-3, was included in the surveillance capsule as the limiting plate material. For its Charpy specimens oriented in the longitudinal direction (LT), the 30 and 50 ft-lb transP' 7 temperatures increased by 44*F and 46'F, j respectively. The plate's transverst. g/L) oriented Charpy specimens experienced increases in the 30 and 50 ft-lb transition temperatures of 34*F and 32'F, respectively. The shift in the 35 miis-lateral-expansion (MLE) Index temperature was 41.5'F for LT specimens l and 26'F for the TL specimens, o The weld metal Irradiated to 1.15 x 10 n/cmdexperienced 20 ft-lb and 50 ft Ib transition .

temperature increases of 10*F and 18'F, respectively. )

o The average upper shelf energy for transversely oriented specimens from lower shell plate -

R1808-3 decreased from 79 ft-lbs to 66 4-Ibs af:er irradiation to the fluence of 1.15 x 10

n/cm8. The weld metal, exposed to the same fluence as the plate material, experienced a

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L decrease in upper shelf energy from 160 ft-lbs to 144 ft-lbs. This upper shelf energy is higher than that for Capsule U and is a function of the test which requires an average to be reported. The plate and weld materials exhibited upper shelf energies for continued safe

> plant operation. The upper shelf energy for these materials is expected to be maintained above 50 ft-Ib throughout vessel life as required by 10CFR50, Appendix G.

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l o The adjusted RTm values br the plate and weld material, based on the surveillance capsule data, are within the two standard deviations of Regulatory Gulde 1.99, Revision 2, predictions.

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2.0 INTRODUCTION

The Seabrook Station Unit 1 Reactor Vessel Radiation Surveillance Program is described in Westinghouse Report WCAP-10110, dated March 1983." The program utilizes six surveillance capsules. Each capsule contains 60 Charpy V-notch specimens,9 tensile specimens and 12-1/2T compact test specimens. The capsules contain vessel plate material R1808-3 with Charpy specimens oriented in the longitudinal and transverse direction, weld metal and heat-affected zone I (HAZ) material. Each capsule provides accelerated data relative to concurrent reactor vessel inner wall material condition, since the capsules are located in the reactor on the neutron shield pad l between the core barrel and the reactor vessel wall, opposite the center of the core. The surveillance program meets the requirements of ASTM E-185-79, Standard Prac#ce forConduc#ng Surveillance Tests For Light-Water Cooled Nuclear Power Reactor Vessels.

The first surveillance capsule in this program, designated Capsule U, was removed after operation of Cycle 1 in August of 1991. The capsule was irradiated for 0.913 EFPYs of operation. The capsule specimens, with the exception of the compact specimens, were tested by B & W Nuclear Services Co. and reported in BAW-2157 ". The 1/2T compact test specimens are being saved for future test needs. The Capsule U data report is attached as Appendix B to this report.

The second surveillance capsule in this program, designated Capsule Y, was removed after I 1

operation of Cycle 5 in M ty of 1997. The capsule was irradiated for 5.572 EFPYs of operation. The capsule specimens, with the exception of the compact specimens, were tested by Framatome - l l

Technologies, Inc. (formerly B&W Nuclear Services Co.) and reported in BAW-23168 . The 1/2T compact test specimens are being saved for future test needs. The Capsule Y data report is attached as Appendix C to this report.

1 The first analysis of the specimen data and dosimetry for Capsule U was performed by Yankee Atomic Electric Co. and reported in YAEC-1853d. The analysis of the specimen data and dosimetry from both Capsules U and Y was performed by Duke Engineering & Services. These analyses are 1 the subject of this report.

• Superscripts denote references located in Section 8.0 of this report 4

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3.0 - SURVEILLANCE MATERIALS AND CHEMISTRY Based on an evaluation of the vessel plate materials, considering initial RT., values, chemistry and the irradiation prediction methods of Regulatory Guide 1.99, Revision 1, vessel lower shell plate R1808-3 was expected to have the highest end-of-life RTer. This reactor vessel surveillance material was supplied by the vessel fabricator Combustion Engineering, Inc. Additionally, Combustion Engineering, Inc. supplied a weldment made up of sections of Lower Shell Plate

R1808-3 and the adjacent Lower Shell Plate R1808-1. This weldment was made using Weld Wire Heat No. 4P6052 and Linde flux 0091, Lot No. 0145. The reactor vessel beltline wold, intermediate and lower shell longitudinal weld seams, and the intermediate to lower shell girth welds, were all

' f bricated using the above weld wire / flux combination. Therefore, the wold supplied for the surveillance program is the limiting weldment. The chemical analyses, heat treatment history, drop weight and RTervalues for the materials used in the beltline region of the Seabrook Station Unit I 1 are provided in Appendix A to this report. The tables are reproduced from the Westinghouse description of the surveillance program in WCAP-10110'.

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1 4.0 PRE- AND POST-IRRADIATION TEST RESULTS 1

~ 4.1 BASELINE DATA, WCAP-10110 I The surveillance program materials were tested in the unirradiated or baseline condition by

. Westinghouse and these results ce reported in Westinghouse report WCAP-10110'. The Charpy impact test results of the base line data reported in WCAP-10110 are shown in Table 41 as the )

zero fluence values. The tensile data b shown in Table 4-2 as the zero fluence values, l

~ 4.2 IRRADIATED DATA, B&W REPORT BAW-2157 AND FRAMATOME REPORT BAW-2316 '

The Capsule U and Y test specimens and dosimetry were tested by B & W/Framatome, as mentioned previously. The test data are provided as Appendixes B and C to this report. The Charpy impact results from the irradiated capsules are !!sted in Table 4-1 with the unirradiated data. l a

The irradiated Charpy data were analyzed using the EPRI Tanh Curve Fitting Routine, Version 1.8 (see Section 4.3).' The upper shelf energy values were calculated using the averaging technique described in ASTM E-185-945 . The tensile data from the surveillance capsule is listed in Table 4 2.

The effects of irradiation on the tensile properties are shown graphically in Figures 4-1 through 4-3.

All data were analyzed in accordance with the 1994 revision to ASTM E-1858 as specified by 1 10CFR50, Appendlx H, Reactor VesselMaterialSurveillance Program Requirements.

4.3 TANH CURVE FITS ASTM E-185-945 defines the Charpy V-notch impact test transition temperature as "the difference in the 30 ft-lbf (41J) Index temperatures for the best fit (average) Charpy curve measured before and after irradiation." There are two methods employed in the industry for determining the best fit Charpy curve. The first is to " eye" the data and draw a best fit curve through it; the second le to use a hyperbolic tangent function (tanh) and computer fit the data to determine the best curve shape.

EPRI, in conjunction with industry experts, developed a computer routine for fitting Charpy tast data with the tanh function. The advantege to using this computer routine is that the curve fits are performed in a consistent manner. This reduces scatter in tre reported shift data. To generate the irradiated Charpy results reported in Table 4-1, the EPRI Tanh Curve Fitting Routine Version 1.8 was used on the data reported in BAW-2157 and BAW-2316.

1 L 1 I

For the Seabrook Station Unit 1 Reactor Vessel Surveillance Program, the initial (unirradiated)

Charpy values for the 30 ft-lb and 50 ft-lb fixes (T, and T ) and for the 35 mils-lateral-expansion (MLE) are documented in Westinghouse Report WCAP-10110. The irradiated values are those generated by using the tanh computer fit. Table 4-3 provides a comparison of the Westinghouse (baseline) reported values, B & W (irradiated) reported values, and the tanh fit values. The Capsule Y results reported by Framatome in BAW-2316 were generated using a tanh computer routine. No comparison is shown for these results as both the Framatome and EPRI curve ths were very close.

4.4 RESULTS The tensile results are presented in Table 4-2 and graphically in Figures 4-1,4 2, and 4-3. The irradiated yield and tensile strengths increased slightly over the unirradiated values. The properties which measure ductility, reduction in area and elongation, decreased with irradiation. These changes are a result of irradiation induced microstructural changes and were expected. The tensile property changes will not affect reactor vessel operation.

The Charpy data, presented in Table 4-1, showed increases in the Charpy transition temperatures (T., T., MLE) and decreases in the Charpy upper shelf energies of the various materials. These r:sults are shown graphically for the surveillance plate R1808-3 specimens in Figure 4-4 (longitudinal orientation) and Figure 4-5 (transverse orientation). The weld metal results are shown graphically in Figure 4-6 and the heat-affected-zone (HAZ) material is shown graphically in Figure 4-7. All graphs of Charpy data show the curve fits using the tanh function for both the unirradiated end Irradiated data.

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TABLE 4 2 Seabrook Staton Unit 1 Surveillance Program Tenslie Properties Base Line Properties Data per WCAP-10110 and irradiated Data per BAW-2157 and BAW-2316 Fluence Test Temp. UTS .2% YS Red.in Area Total Elong.

Maledal E+19 n/cd Uniform Bong.

• F KSI KSl  %  %  %

PLATE R1808-3 0.00 75 92.0 74 4 t

LT 65.0 24h 14.5 75 91.0 70.0 88.0 27.0 14.5 300 85.0 64.0 68.0 24.0 13.0 30" 85.0 65.0 86.0 24.0 13.0 550 89.0 63.0 61.0 24.0 14.0 550 an.o 83.0 as o 24.0 13.0 0.28 70 94.2 73.8 63.5 23.7 9.8 300 48.8 67.8 65.4 20.5 i

8.3 Eso 01.1 87.s GP 1 19.4 7A 1.15 70 08.5 75.9 82.6 25.3 10.9 300 88.8 88.7 82.8 21 2 0.26 550 93.7 67.5 57.0 21.6 10.4 PLATE 0.00 75 914 71.0 55 4 26.5 15.5 R1808 3 TL 75 91 4 71.0 55.0 26.5 15.5

, 300 88.0 88.0 $3.5 21.0 12.0 300 88.0 86.0 55.0 22.0 12.0 550 88.0 63.0 51.5 214 13.0 Eso 88.0 84 o 47.E 2d 0 1E_E 0J8 70 94.1 73.3 54.3 21.4 8.5 300 85.7 67.2 55.9 17.8 7.8 sso e1.s an a da s 1s.o 7e 1.15 70 95.7 74.9 54.8 23.4 11.1 300 88.1 68.4 50.4 - 19.4 9.91 550 R2.8 67.3 38 8 18.7 10.1

i. WELD 0.00 75 ' 87.0 75.0 7?S 27A 154 MATERIAL 75 88.0 74.0 75.0 28.0 14.0 300 81.0 88.0 73.0 18.0 0.0 300 81 4 87.0 73.0 23.0 10.0 550 85.0 85.0 87.0 22A 10 4 ano ad o as_o 71.0 22.0 11.0 OJ8 70 90.0 76.0 72.5 23.7 8.9 300 83.7 70.5 71.7 21.2 7.5 j

Eso 87 7 80.4 89 3 1a.0 49 1.15 70 91.6 78.6 71.9 272 10.5 300 84.3 70.7 14.3 21.1 i

8.43 j 550 90.0 70.5 69 8 24.5 10.1

.g.

TABLE 4-3

/. Comparison of Tanh Computed Charpy Values to Westinghouse and B&W Reported Values Unirradiated Data:  ;

Material Parametsr Tanh Westinghouse B&W l

Plate Tm -28'F -25'F --

I R1808-3 T. -4'F O'F --

LT 35MLE -6*F O'F --

Plate Tm 9'F 10*F -

R1808-3 T, 58'F 60'F -

TL 35MI.E 46*F 50*F -

-199'F HAZ Tm -160* F --

(a) T, -139'F -120* F --

35MLE -126'F -105'F --

WELD - -Tm -75'F -60* F --

Tg -50*F -45'F --

35MLE -47'F -50* F --

trradiated Data:

l Material Parameter Tanh Westinghouse B&W Plate T. 11'F -- 3'F  :

R1808-3 Tu 34*F -- 40'F LT USMLE 24'F -- 26'F Plate Tm 38'F -

39'F R1800-3 T, 80'F -

78'F TL- 35MLE 65'F --

68'F HAZ T. 104*F -

-85'F l (a) T. -60* F - -41 'F 35MLE -57'F -- -47'F l WELD T, -50*F -

-56

• F l- Tu -30*F -- -38'F j 35MLE -32* F - -40'F Note: (a) HAZ data exhibit significant scatter which is typical of HAZ material.

l Figure 4-1 TENSILE PROPERTlES FOR SEABROOX STATION UNIT 1 REACTOR VESSEL PLATE R 1803-3 LONGrrUDINAL PP.OP.

UNIRR ADIATED AND SURVEILI.ANCE CAPSULES U AND Y DATA 100 -

I b.,' .

650 C -

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60 O 100 200 300 400 500 600 Temperature Degrees F UNIRRAD. UTS UNIRRAD. YS CAP. U UTS CAPU. YS CAP.Y LTTS CAP.Y YS

+ - G-- O C $' O - -%*

TENSILE PROPERTIES FOR SEABROOK STA110N UNIT I REACILR VESSEL PLATE R1808-3 LONGITUDINAL PROP.

UNIRRADIATED AND SURVEILLANCE CAPSULES U AND Y DATA 80 70 -

A 2 m g J L g..._. ........... ...x.....

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' ' ' I ' ' ' I ' ' '

0 200 300 400 500 000 0 100 Temperature Degrees F UNRR. RW UNIRR TOT ELON3 UNtRR.UNIF ELONG CAP. U RA CAP. U TOT ELCMG M -b -h CAP Y TOT ELONG CAP. Y UNIF. ELONG

~._

CAP U UNIF ELONG CAR Y RIA f~ ' _.4., ,.J . _ . . _ .

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TENSILE PROPERTIES FOR SEABROOK STATION UNIT I Figure 4 2 REACTOR VESSEL PLATE R1808-3 TRANSVERSE PROP.

LE4 IRRADIATED AND SURVEILLANCE CAPSULES U AND Y DATA 100 o

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60 600 100 200 300 400 500 0

Temperature Degrees F UNIRR. LTTS UN!RR Y5 CAP. U tJr$ CAR UYS CAP Y UTS CA.R Y Y5

.bdu

---&- - 0 C - 0 * - 5% *

- TENSILE PROPERTIES FOR SEABROOK STATION UNrr i REAC'IOR VESSEL PLATE R1808-3 TTLANSVERSE PROP.

UNIRRADIATED AND SURVr.IT LANCE CAPSULES U AND Y D STA 60 1_

- -.-....... h , ,

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-- *h-+-- C CAP UUNtf flONG CAP Y kl4 CAP Y TOT r.1DNG

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Figure 4-3 TENSILE PROPERTIES FOR SEABROOK STATION UNIT I RSACTOR VESSEL WELD METAL UNIRRADIATED AND SURVEILLANCE CAPSULES U AND Y DATA 100 -

650 b(

~

90 -

• .., ,. 6

h " '

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ea

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450 I ' ' ' I ' ' ' I '

60 O 100 200 300 400 500 600 Temperature Degrees F LNIRR UTS UNIRR. YS CAP. U UTS C- U YS CAP V UTS CAP Y YS

^,

& --9-- _' - [- - ' ENSILE PROPERTIES FOR SEABROOK STATION UNIT I REACIOR VESSEL WELD METAL UNIRRADIATED AND SURVEILLANCE CAPSULES U AND Y DATA 80 A _...s,.

,0 g,_ - . _ . - . - - - - k

-......._,,,,,,,_4> __

x v

60 -

50 -

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0 O 100 200 300 400 500 600 Temperature Desrees F UNIRR RED IN ARE A UNIRR TOTAL ELONG LIN!RR UNIFORM E!DNO CAP U RI A CAP 701 FLhNG

^

---

G-- -F T cAruuwwetown car v miA carvvoi nn~a rArvuun iLoso stiwtd

1 l

Figurs 4-4 SEABROOK STATION NO. I C,APSULE U AND CAPSULE Y SURVEILLANCE RESULTS PLATE R1808 3 LONGITUDINAL CHARPY IMPACT DATA 200 -

250

200 150

.s  : + $  ? o: e 150 3 J ,a % .0...0 ..........

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0 200 400 600 200 Temperature DegreesF UNIRR DATA 0.28E+19 E>lMeV 1. iSE+19, E>lMeV

$ Q O sb-31t

\

SEABROOK STATION NO.1 CAPSULE U AND CAPSULE Y SURVE!LLANCE RESULTS PLATE R1808-3 LONGITUDINAL, MLE, CHARPY IMPACT DATA 200 5

- 4 4.

-  : g 150 -

u

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UNIRR. DATA 0 28E+ 19, E>l MeV I 15E419, E>lMeV

$ ') Q sb 3ltle 14-

Figure 4-5 SEABROOK STATION NO.1 CAPSULE U AND CAPSULE Y SURVEILLANCE RESULTS PLATE R1808-3 TRANSVERSE CilARPY IMPACT DATA 200

- 250 150 -

200 h

~

c: -

150 o

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$$2................... 100 80" 9MQ.y.'. .

........g...o................ O ,

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p. ht , I , t .

g 200 0 200 400 600 Temperature Degrees F CAPSULE U CAPSULE Y

^ ^ 0.2BE+19, E>l MeV 1.15E+19, E>lMeV sb 3tl SEABROOK STATION NO. I CAPSUI.E U AND CAPSULE Y SURVEILLANCE RESULTS PLATE R1608-3 TRANSVERSE, MLE, CHARPY IMPACT DATA 200 5

?

.- 4 150 -

E $

8 -

g -

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$ 5  !

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^ '

O 28E+ 19,1> l McV i 15E+ 19, E>l MeV sb-3ttle

Figure 4-6 SEABROOK STATION NO. I CAPSULE U AND CAPSULE Y SURVEILLANCE RESULTS WELD METAL CHARPY IMPACT DATA g 200 g

a _

250

+  :

V V Q-150

... n .g g g ,. ,.,. g ,. ,. ,.,. ,.,. ,.,.. 200 i O -

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$ Q O SEADROOK STATION NO.1 CAPSULE U AND CAPSULE Y SURVEILLANCE RESULTS WELD METAL, MLE. CHARPY IMPACT DATA 200  ; 5

~  :

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Figurs 4-7 SEABROOK STATION NO. I f CAPSULE U AND CAPSULE Y SURVEILLANCE RESU1;IE HAZ CHARPY IMPACT DATA 200 250 Q -

is0 o...... o , _.  ;

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100 E 50 -

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-200 0 200 400 600 Tempemture Degrees F CAPSULE U CAPSULE Y UNIRR. DATA 0.28E+ 19, E>l MeV l.15E+ 19, E>lMeV

.w + 0 o SEABROOK STATION NO.1 CAPSULE U AND CAPSULE Y SURVEILLANCE RFSULTS HEAT AFFECTED ZONE, MLE, CHARPY IMPACT DATA 00

-5

~

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5.0 RT y CALCULATION

~'

5.1' ' CALCULATED RT.7 VALUES l CAPSULE U DATA ONLY:

Adjusted RTer values of 96*F for the R1808-3 surveillance plate and minus 40*F for the surveillance weld metal (see Table 4-1) were caleJlated from the surveillance capsule test results using the methodology of Regulatory Guide 1.99, Revision 2. Regulatory Guide 1.99, Revision 2,8

. requires that two or more credible surveillance data sets be available before actual surveillance data can be used for licensing purposes. Since this discussion treats the analysis of the first capsule (only one data set), these calculated RT y values are presente'd so a comparison of the two Regulatory Guide methods can be made. The Table 4-1, adjusted RT y values for the Capsule Y data were calculated using the credible surveillance data method. All future capsules and adjusted reference temperature will be calculated with this latter method.

In the absence of two or more credible surveillance data sets, Section C.1 of Regulatory Guide 1.99,

-- Revision 2 requires that the adjusted RTer be based on a delta RT.7 value calculated from fluence

. and derived chemistry factors for the material. To determine the adjusted RT y numbers in j Table 4-1, the delta RTervalue is the T, shift value measured from Charpy specimen testing. This allows the surveillance data from this first capsule to be compared with Regulatory Guide 1.99  !

predictions (discussed in Paragraph 5.2). The adjusted RTer calculation used to develop Table 4-1 values is as follows:

j The adjusted RT rs are determined using R. G.1.99 equations, where:

Adjusted RTer = (initial RT.7 + T Shift + Margin) , t i

and where Margin = 2 x (sigma, + sigma,")

1 For the plate (TL) and weld initial RT p, the initial sigma margin (sigma) is set to zero because the

' initial RT rs were determined in accordance with ASME Code, Section 111, NB2300 which is a i conservative (upper bound) method based on drop weight and Charpy data. Per Regulatory

, _ - _ . _ _ = _ _ -__ -_

Guide 1.99, Revision 2, sigma, is 28'F for welds and 17'F for base metal except that sigma, need not exceed 0.5 times the mean value of Tm Shift. Therefore, sigma, for plate is 14*F (28'F + 2) and sigma, for weld is 5'F (10'F + 2).

Thus, the Table 4-1 values are:

1

! Plate: Adj. RTer = 40*F + 28'F + 2(14'F) = 96*F l

Weld: Adj. RTor = -60*F + 10'F + 2(5'F) = -40*F CAPSULE U AND Y DATA:

With the Capsule Y data, we have two or more " credible" surveillance data. Using these data points, the adjusted RTervalues are calculated for comparison to the Capsule U only results. This method allows the base Regulatory Guide 1.99 predictions to be compared to data fitted in accordance with Section 2.1 of the Regulatory Guide using the credible surveillance data. Regulatory Guide 1.99, Section 2.1 states:

If this procedure (fitting credible data) gives a higher value of adjusted reference temperature than that given by using the procedures of Regulatory Position 1.1 (the Capsule U on!y treatment), the surveillance data should be used. If this procedure gives a lower value, either may be used.

Section 2.1 requires that the surveillance material chemistry be compared to the actual vessel chemistry to determine if differences exist between surveillance material and the average of reported values for the vessel materials. For the Seabrook Unit 1 vessel the copper or nickel levels of the surveillance materials do not differ significantly from that of the average values for the vessel materials (see Tables A-1 through A-3).

i Nste: The key words here are " average of the reported values." Table A-3 shows a weld copper spread of 0.02 to 0.07. The average of three reported values is only 0.037.

' A low copper alloy was specified for the plates and a low nickel, low copper wire was used for all welds. Using the chemistry ratio method of Regulatory Guide 1.99 is not necessary here.

Next, the survelliance data is fitted using Equation 2 of Regulatory Guide 1.99 to determine the chemistry factor (CF) for the best fit of the data.

a RT, = (CF)fS'8"W 0 Equation 2 Stated, the adjusted reference temperature equals the chemistry factor multiplied by the fluence factor.

The Regulatory Guide procedure requires a method which minimizes the sum of the squares of the errors of the fit. The Guide states:

The surveillance data should be fitted using Equation 2 to obtain the relationship of sRTmr to fluence. To do so, calculate the chemistry. factor, CF, for the best fit by multiplying each adjusted sRTwr (Charpy Tx Shift) by its corresponding fluence factor, summing the products, and dividing by the sum of the squares of the fluence factors. E For the transverse plate (orientation required) data, the resulting CF is 35.60.

For the weld data, the resulting CF is 11.24.

The calculation is shown below.

Seabrook Unit 1 Surveillance Data CF Fit Capsule Material Fluence Charpy Fluence Fluence Fluence Calculated E+19 Tw Shift Factor Factor Factor Chemistry n/cm 2 x Shift Squared Factors U Plate TL 0.28 28 0.652591 18.27254 0.425875 Y Plate TL 1.15 34 1.039027 35.32693 1.079578 35.60356 Sum 53.59947 1.505453 U Weld TL 0.28 10 0.652591 6.52591 0.425875 L Weld TL 1.15 10 1.039027 10.39027 1.079578 11.23661 Sum 16.91618 1.505453

~

~

' The resulting calculated adjusted RT y values for plate and weld using the surveillance data cvailable method of Regulatory Guide 1.99 are:

Calc. Shift + Initial RT y + Margin = Adjusted RT y Calculation of Adjusted RT Values Material' Fluence Calculated l initial RT. Margin Adjusted E+19 n/cm' RT Shift - 2Walue' RT.

Plate 1.15 36.99 40 17 93.99 Wold 1.15 11.67 -60 28 -20.33 L

Because " credible" surveillance data are used, the Regulatory Guide allows th6 Guide's margin terms to be halved.' For plate the margin is 8.5'F; for weld 14"F. The margin term used is twice the square root of the sum of the margin terms squared. As mentioned earlier, the margin term for the

!' initial RT.y values is set to zero because the conservative method of the ASME Code was used to determine those values.

5.2 COMPARISON OF SURVEILLANCE CAPSULE RESULTS TO REGULATORY GUIDE 1.99, REVISION 2 PREDICTIONS i

As described in Section 5.0, the adjusted RTervalues reported in Table 4-1 for the Capsule Y data tre based on Regulatory Guide 1.99 delta RTervalues calculated using fitted chemistry factors and th9 fluence function. To compare the capsule results to Reg. Guide 1.99 predictions, chemistry fIctors for the surveillance plate and weld were calculated using Tables 1 and 2 of Reg. Guide 1.99.

Applying the chemistry factors to the Reg. Guide 1.99 fluence function, delta RTer curves were generated to graphically show the Reg. Guide 1.99 predictions against the T shift data from the surveillance capsu'es. Additionally, the figures show, as a dashed line, the predictions using the

- fitted chemistry factors. The CF fit agrees well with the actual data, as it should.

. Figure 5-1 depicts the T, shifts for the plate L-T and T-L oriented Charpy specimens and weld specimens plotted against the Reg. Guide 1.99 prediction curve. The Reg. Guide curve has the form of:

a RTer (shift) = (CF)f****"N where ,

f = fluence (E+19 n/cm ), and CF = chemistry factor

]

l

- For the plate, the Regulatory Guide, tabular CF is 44 (Avg. Cu = 0.07 wt%, Avg. Ni = 0.59 wt%). For the weld, this CF is 28.3 (Avg. Cu = 0.037 wt%, Avg. Ni = 0.057 wt%).

As demonstrated by the CFs, the surveillance capsule, transverse plate data (orientation used for ,

RT y determination) shown at the top of Figure 5-1 shifts slightly below the Reg. Guide 1.99 prediction line. - The longitudinal data points fall around the prediction line. Both points fall below the two sigma margin limit specified by Regulatory Guide 1.99, indicating good agreement with Reg.

Guide 1.99 predictions. The +2 sigma margins are shown as the dashed lines in Figure 5-1.'

The weld metal results are shown at the bottom of Figure 5-1. Here the weld Charpy shift results falls below the prediction line. The test data points fall below the two sigma margin.

5.3 - EFFECT OF SURVEILLANCE RESULTS ON CURRENT TECHNICAL SPECIFICATIONS The current Seabrook Technical Specifications contain heatup, cooldown, and LTOP curves which are based on the predicted RT 7 shift in Regulatory Guide 1.99, Revision 2. No't. that Seabrook has two sets of credible surveillance data, the impact of using these data for Technical Specifications can be evaluated.

l 1

l

)

1

-l 1

Figure 5-1 SEADROOK STATION UNIT I CAPSULES U AND Y SURVEILLANCE RESULTS FOR PLATE VERSUS R.O l.99 PREDICTIONS 100

~

RG 1.99 +2 Sigma 80 -

u,.

-~~_,,,,, O,1,99 Prediction f 80 -

C .

3 **.*',.* ____________

2

-g

• ______________%N

• 40 a , './4

, Fitted CF

/ Prediction 20 -

/

m. ...I . . . . t . . . . . I , . . . l . . . . l . . . . l ....I . . . .

n 0 0.5 1 1.5 2 2.5 3 3.5 4 Fluence Et 19 n/cm2. E>lMcV PLATE PLATE LT TL RO 1.99 CF = 44 SbPrg Fitted CF = 35.6 SEABROOK STATION UNIT I CAPSULES U AND Y SURVEILLANCE RESULTS FOR WELD METAL VERSUS R. G.1.99 PREDICTIONS 100 80 - RG 1.99 + 2 sigma margin E

~

il a -

d e - R. G.1.99 Prediction 2 '**-

y 40 -/

a /

Fitted CF Prediction i p --4 ____________-_-_---------.

0

' ' ' ' ' ' ' ' ' ~' ' ' ' I ' I ' ' ' ' ' ' ' ''I ' ' ' '

O 0.5 1 1.5 2 2.5 3 3.5 4 Fluence E419 n/cm2. E>lMeV WELD SHIFT RG 1.99 CF = 28.3 Fitttd CF = 11.2 23-o

6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

The characterization of the neutron environments within the surveillance capsules and the reactor pressure vessells necessary to properly assess the effects of neutron-induced changes in materials properties on these components. The key parameters of the neutron environments in which these components are exposed include the neutron energy spectrum, the neutron flux levels and the resulting accumulated fluence. The neutron environment in the surveillance capsules is determined by a combination of analysis and measurements of activity from the passivo neutron flux monitors (dosimeters) contained in each surveillance capsule. In order to relate the materials properties changes observed in the test specimens to present and future conditions of the reactor pressure vessel, the differences in neutron environment between the surveillance capsule and locations within the reactor pressure vessel must be determined analytically.

Fast neutron fluence with E>1 MeV has traditionally been used to correlate the measured material properties changes to neutron exposure. In recognition of the differences in neutron environment between the surveillance capsule and various reactor components, additional neutron exposure information is provided, as recommended in ASTM Standard Practice E853-87.7 This additionai neutron exposure information is fast neutron fluence with E>0.1 MeV and displacements per iron atom (dpa). The energy-dependent dpa function used in pressure vessel analyses is specified in ASTM Standard Practice E693-948.

This chapter provides the results of the dosimetry evaluations for Capsule U, which was irradiated in Cycle 1, and Capsule Y, which was irradiated in Cycles 1 through 5. The analysis utilizes current industry-standard analytical methods and evaluated nuclear data, including the most recently released neutron transport and dosimetry cross-section libraries derived from the ENDF/B-VI data base. The differences between the calculated and measured capsule dos! metry reaction rates provide normalization factors to determine best-estimate current and future projected neutron exposure parameters for the pressure vessel. The differences between the calculated and measured capsule dosimetry reaction rates for different nuclear reactions provides data to assess the uncertainties associated'with the determination of these parameters. This report, therefore, provides a self-consistent, up-to-date neutron exposure evaluation for materials properties based on the Capsule U and Y results.

I

.m.

7 6.2 PRESSURE VESSEL AND SURVEIU,ANCE CAPSULE REGION GEOMETRY l The plan and elevation views of the core and the reactor pressure vessel, including the locations of the surveillance capsules, are shown in Figure 6-1. The six irradiation capsules are attached to the outside of the neutron shield pad at the specified azimuths. Given the symmetry of the core, Capsules U, X, W and Z are at the same relative azimuth (31.5') within a one-eighth core region defined by 0* at the core flats (azimuths of 0,90,180 and 270') and 45' at the core diogonallines (azimuths of 45,135,225 and 315'). Ukewise, Capsules V and Y are at the same relative azimuth (29') within a one-eighth core region.

i l

l The plan view of the dual survelliance capsule holders for Capsules U and V are shown in Figure 6-2. Capsules X and Y are in a similar dual holder, while Capsules W and Z are each in single holders. The stainless steel specimen containers within the holders are approximately 56 inches l In height. As shown in Figure 6-1, the containers are positioned axially so that the test specimens are approximately centered near the core midplane of the 12 foot active core region.

l

- Detailed region geometries and material compositions are included in the discrete ordinates analysis modeling. There are explicit regions representing the active core, core baffle, reflector water, core barrel, neutron shield pads, surveillance capsule holders, survelliance capsule specimens, i

downcomer water, pressure vessel with inner stainless steel cladding, cavity insulation, cavity void, concrete liner and concrete shield. The spatial and energy distribution of neutrons at a particular location is determined by factors such as distance and shielding from the active core and local moderation.

6.3 DISCRETE ORDINATES ANALYSIS MODELING 4

. . The discrete ordinates analysis is performed in (r,0) geometry using the DORT (Version 2.8.14) '

I two-dimensional discrete ordinates code and the BUGLE-96* cross section library. The BUGLE-96 library is a 47 neutron energy group ENDF/B-VI based data set, produced specifically for light water reactor shielding and reactor prescure vessel dosimetry applications, in the analysis, anisotropic scattering is determined with a P, expansion of the scattering cross sections. The angular discretization is characterized with an S, order of angular quadrature, which represents 48 discrete engles. Fine radial and azimuthal mesh are used, which allows for modeling of the surveillance capsule holder and specimens as distinct regions. There are 16 radial mesh in all assemblies near 25-

the core periphery (see Figure 6-3), which is one mesh block per fuel pin. There are 74 radial mesh between the core periphery and the OD of the pressure vessel, resulting in an average mesh spacing of less than 1.1 cm in th's region. There is a uniform azimuthal mesh spacing of 0.5* for the O to 45' one-eighth core region, for a total of 90 azimuthal mesh. This detailed modeling results in an azimuthai mesh spacing of approximately 1.9 cm at the pressure vessel inner radius.

The discrete ordinates analysis consists of five separate forward-mode transport calculations, one for each of the five Seabrook operating cycles. Each of the operating cycles has its own discrete ordinates calculation using its specific cycle-average neutron source distribution. Thus, absolute predictions of neutron exposures and reaction rates, within each of the 47 neutron energy groups, in the surveillance capsules and through all azimuths and thicknesses of the reactor pressure l vessel, are obtained in detail for each operating cycle. Thus, there is no need for adjoint solution techniques in these evaluations. 1 l

The discrete ordinates analysis provides saturated reaction rates for the neutron dosimetry at the surveillance capsules to permit a comparison of analytical predictions to measurements. The discrete ordinates analysis also provides the means to relate the dosimetry results to key locations l

at the Inner radius and through the thickness of the reactor vessel wall.

The neutron source distribution is specified each cycle in pinwise detail for each of the assemblies l 1

near she core periphery (the 17 assemblies designated in Figure 6-3) which are the most important neutron flux contributors to the surveillance capsule and pressure vessel analysis. The pinwise (x,y) neutron sources are mapped to the (r,0) geometry of the DORT modeling. The pinwise (x,y) neutron sources are developed from the same detailed pinwise ohysics calculations used in the plant licensing and core monitoring analyses. The cycle average power level and cycle average burnup of each fuel pin is used to develop a cycle average source for each pin. These neutron j source calculations explicitly include, on a pinwise average basis, the change in neutron yield per l fission and the change in fission spectrum associated with the local burnup-dependent depletion of uranium and buildup of plutonium.

l The incore monitoring of the power distribution in the reactor core during operation is based on the I signals from the f!xed incore detectors locatad in the center guide tubes of selected assemblies.

l These measured assembly average relative powers may be compared to the predicted assembly average relative powers from the pinwise neutron sources calculations to evaluate the accuracy of the source distributions in the DORT analyses. Figure 6-4 shows the calculated and measured l

l

Essembly average relative powers (cycle average) In Cycle 1 for the core peripheral assernblies which provide the largest contributions to the surveillance capsule and vessel neutron fluence. The Cycle 1 comparison shows excellent agreement between calculated and predicted average powers, with the largest difference around 0.7%. Thus, the Cycle 1 source distribution, which was the source contributing to the Capsule U activation, is accurate by these comparisons to plant measurements. Figure 6-5 shows the calculated and measured assembly average relative powers

~ covering Cycles 1 through 5 for the core peripheral assemblies. The Cycle 1 through 5 comparison again shows excellent agreement between calculated and predicted average power, with the largest difference around 0.8%.Thus, the source distributions used in Cycles 1 through 5, which contributed l to the Capsule Y activation, are accurate by these comparisons to plant measurements.

b i

i The measured Cycle 1 through 5 average powers (contributing to Capsule Y) are compared to the measured Cycle 1 average powers (contributing to Capsule U) in Figure 6-6. The ratios of the Cycle l 1 through 5 to the Cycle 1 average powers show the results of the low-leakage fuel management impiamented in recent cycles. These core peripheral assemblies show reduced relative powers, with an average reduction of approximately 20%, and reduct!ons of over 30% in key assemblies (Assemblies B-3 and A-5 in Figure 6-6). Thus, the Capsule Y results, relative to the comparable Capsule U results, are expected to show the effects of these fluence reduction measures.

l 6.4 DISCRETE ORDINATES ANALYSIS RESULTS Th3 discrete ordinates analysis calculated results are summarized in this section. The neutron  !

exposure rate parameters of primary interest are the following, which are provided either on a cycle- ,

specific or cumulative (Cycles 1 through 5) basis:

i o Fast neutron flux greater than 1 MeV ( x 10" n/cm -sec) 8 '

o' Fast neutron flux greater than 0.1 MeV ( x 10" n/cm -sec) 8 o Iron Atom Displacement Rate ( x 104 'dpa/sec) I These parameters are typically integrated over time and power (effective full-power years, or EFPYs) to obtain the following integrated neutron exposure parameters:

i 4

o Fast neutron fluence greater than 1 MeV ( x 10 n/cm')

o Fast neutron fluence greater than 0.1 MeV ( x 10 n/cm') ,

o Iron Atem Displacement - ( x 10'8 dpa)

-27

{

l Selected results from the discrete ordinates analysis are provided for the key locations in the problem. These key locations are comprised of:

o Capsule specimen locat!ons in Capsule U (and X) and Capsule Y (and V).

o Selected reactor vessel azimuths, ranging from O' at the core flats to 45' at the core diagonal lines (see Figure 6-1).

o Reactor vessel Inner surface, as defined by the minimum average base metal inner radius (stainless steel clad-base metal interface) at 86.5 inches from the core center. There is a 5/32-inch thick stainless steel cladding on the inner surface of the reactor pressure vessel, assumed to be inside the 86.5 inch inner radius.

o Reactor vessel outer surface, as defined by the minimum average vessel thickness of 8 % inches, o Reactor vessel fractional thicknesses, at %T, %T and %T, assuming the above-mentioned minimum average reactor vessel inner radius and thickness.

The as-built reactor vessel dimensions are not addressed in these calculations. The as-built pressure vessel inner radius and thickness are larger than the assumed minimum average design values, so that the fluence parameters provided here for the pressure vessel locations are conservatively high.

The neutron exposure parameters for the geometric center of the surveillance capsule locations are provided in Table 6-1. The Capsule U and X locations show higher (4-12%) neutron exposures than the Capsule V and Y locations. This is expected, since the Capsule U and X locations are closer to one of the fuel assembly comers than the Capsule V and Y locations (see Figure 6-1). The Cycle 1 through 5 results, when compared to the Cycle 1 results, show the effects of the low-leakage fuel management, consistent with the power distribution differences discussed in the previous section.

The Cycle 1 through 5 neutron exposure parameters are 27-29% below those of Cycle 1.

28-

~

The calculated neutron exposure parameters for the reactor vessel clad / base metal interface are

( provided in Table 6-2 for selected azimuths. The selected azimuths are:

o O' at the core flats, o 19-21.5*, which is the azimuthal range of a broad local maximum

(

! near a fuel assembly comer (see Figure 6-1). The maximum of the fluences at any of the azimuthal mesh at 19,19.5,20,20.5,21 or

( 21.5* are provided in the tables.

o 29*, at the azimuth of Capsules V and Y, behind the neutron shield pad,

)

o 31.5*, at the azimuth of Capsules U and X, behind the neutron shield pad, and f.

o 44-45*, at the core diagonal line, behind the neutron ableid pad.

The values in bold Italles in Table 6-2 are the maximum values for any azimuth. The 44-45' ezimuth at the core diagonal is the maximum exposure rate azimuth for most cycles, and is the Cycle 1 through 5 maximum exposure rate azimuth. The 19-21.5* azimuth, which is near a fuel I assembly comer and not shielded by the neutron shield pad, is a competitive location. The azimuths of the surveillance capsules benefit from the neutron shield pad and additional distance from the active core and are therefore relatively low in neutron exposure. The 0* azimuth at the core flats is relatively low in neutron exposure due to the vessel's increased distance from the active core.

' The relative radial distribution of the neutron exposure parameters within the reactor vessel wall are presented in Tables 6-3.1 and 6-3.2 for the same azimuthal locations. The results for Cycle 1, which <

was the highest fluence rate cycle for the reactor vessel, are shown in Table 6.3-1. In general, the higher neutron energies exhibit a greater falloff through the vessel wall. The fast neutron flux greater than 1 MeV, therefore, shows a greater relative falloff than the fast neutron flux greater than 0.1 i MeV, The dpa falloff rate is between these relative falloff rates. The results for Cycle 4, which was the lowest fluence rate cycle for the reactor vessel, are shown in Table 6.3-2. Relative to Cycle 1 Cycle 4 sepresents a lower fluence rate, a different azimuthal peak location, and a source attribution representative of higher enrichment and bumup assemblies. Despite these differences, the relative radial distribution of the neutron exposure parameters within the reactur vessel wall are  ;

similar, as shown by the small changes from Cycle 1 to Cycle 4 in Table 6-3.3. Thus, projections i

l

for future cycles using these relative falloff factors are not expected to be significantly influenced by the details of the cycle's fuel management. Cycle 4 is considered representative of future cycles and falloff factors from this cycle will be used for future cycle projections.

Lead factors, based on the calculated fast neutron fluxes and fluences (E>1 MeV) for the maximum fluence azimuths, are developed in Table 6-4.1 for Capsules U and X and Table 6-4.2 for Capsules V and Y. The lead factor is the ratio of the fast neutron flux at the surveillance capsule to the maximum fast neutron flux at the location at the reactor vessel clad / base metal interface.

Lead factors are developed for the limiting 19 21.5* and 44-45' azimuths. The cumulative (Cycle 1 through 5) lead factors, which represent the peak fluence azimuth at 44-4S*, are used for future cycle projections. j l

The neutron energy group structure and the calculated fast neutron energy spectra are shown in Table 6-5.1.1 for Capsules U and X and Table 6-5.2.1 for Capsules V and Y. The neutron energy group structure and the calculated spithermal and thermal neutron energy spectra are shown in Table 6-5.1.2 for Capsules U and X and Table G-5.2.2 for Capsules V and Y. The Cycle 1 through 5 average spectra and total neutron fluxes for selected energy ranges are also provided.

6.5 NEUTRON DOSIMETRY AND UNCERTAINTIES l l

The passive neutron dosimetry included in the Seabrook surveillance program is described in WCAP-10110' and identified in Table 6-6. This table also provides the primary nuclear reactions i and constants that were used in the evaluation of the dosimeter reaction ratos. Typical relative i l i i locations of the neutron sensors within the capsule specimen block are shown in Figure 6-2 for the copper, iron, nickel and cobalt-aluminum (bare and cadmium-shielded) wires. The wires are placed in holes drilled in spacers at several axial elevations (designated top, middle and bottom) within the capsules. The uranium and neptunium fission monitors are cadmium-shielded containers placed l axially near the center of the capsules.

l .

The specific activity of each of the neutron dosimeters, corrected to the time of each capsule's end of irradiation, was provided by B & W/Framatome for Capsules U and Y in the reports attached as Appendixes B and C, respectively. These analyses were govemed by the existing ASTM standard 5 7A"'" provide the practices at the time of the analysis. The currant ASTM standard practices nuclear constants (product half-life and fission yield) in Table 6-6 for evaluation of the dosimetry.

30-

' The specihc activity of the neutron dosimeters provides a measure of the time and energy-intsgrated cffect of the neutron flux Incident upon the target nuclides. In order to assess the fluence essociated with the specific activities, the following parameters must be accounted for:

l o The cycle-specific power distributions in the reactor, which determirse the magnitude and spectrum of the neutron source for target activation, and

-o the power operating history of the reactor, which determines the production and L decay time periods for the activation products.

The discrete ordinates analysis provides the detailed energy spectrum associated with survelilance l

capsule locations for each operating cycle. The saturated reaction rates for the target nuclides are l ~ calculated in the same 47 neutron energy-group detail, providing a separate cycle-specific, detailed l . spectrum weighting for sach of the activation reactions. Using this data and the power cperating histories, the measured discharged reaction rates may be corrected to a " saturated" reaction rate,

- assuming no decay of the activation nuclides, using the equations in Figures 6 7 and 6-8. The

! measured saturated reaction rates may then be compared to the ca,culated saturated reaction rates.

A summary of the operating dates, days and effective full-power years (EFPYs) of operation for Cycles 1 through 5 is provided in Table 6-7. The detailed power history for Cycle 1 operation, the

' decay time to end-of-cycle (EOC) and the ratio of measured to saturated activity, using the equation l In Figure 6-7, are provided in Table 6-8 for the Capsule U discharge at EOC 1. The same data for '

Cycles 1 through 5 operation is provided in Tables 6-9.1 through 6-9.5, respectively, for Capsule

Y discharge at EOC 5. The total ratios for each cycle provide the fraction of that cycle's saturated j cctivation level that remains at the time of capsule discharge.

p The Capsule U and Y activation calculations, which result in measured saturated reaction rates and infirred fast neutron flux (E>1 MeV) levels, are provided in Tables 6-10.1 and 6-10.2 , respectively.

f These caleclations use the nuclear constants in Table 6-6 and the results of the discrete ordinates cnalysis for the five operating cycles. For Capsule Y, the cross sections for fast neutron flux (E>1 MeV) and the ratios to saturated actMty in Table 6-10.2 are the five-cycle " effective" values, which include the cycle-specific factors in the equations in Figure 6-7 and 6-8. These cycle-specific factors explicitly account for the differences in neutron spectrum, activation levels and power history during

- Irr:diation. ,

-31

The calculated and measured saturated reaction rates for Capsules U and Y are compared in Table 6-11.1. Similar comparisons for Capsules U, Y and V of Callaway Unit 1 are provided in Table 6-11.2, as taken from WCAP-1489588. The Callaway calculations from Westinghouse also utilize recently released neutron transport and dosimetry cross-section libraries derived from the ENDF/B-Vi data base. Calculated-to-Measured (CM) ratios are provided for each of the activation nuclides, using the average of the top, middle and bottom dosimeters. A standard deviation of the CM ratios is also indicated, l

(

The Callaway measured results include correction factors which reduce the U 238 and Np 237

)

measured reaction rates to account for U-235 impurity fissions, plutonium build-in and gamma ray-l Induced fissions. These corrections are stated in Table 6-11.2. Since the impurity levels are not l

well-defined for Seabrook, such corrections are not applied to the Seabrook measurad results. This

,. effectively decreases the average CM ratios for Seabrook and, upon normalization of calculated parameters to the measurements, results in conservatively higher neutron exposure projections for future cycles.

The calculated-to-measured comparisons show that, on an average basis, the Seabrook average  :

CM ratios are effectively the same for Capsules U and Y (0.906 and 0.897). Including the U-238 and Np-237 corrections results in higher CM ratios for Callaway, which are closer to unity and show i

an increasing trend from Capsules U to Y to V (0.961,0.993 and 1.024). The standard deviation of differences by nuclide for Seabrook Capsules U and Y (0.104 and 0.059) are similar to the Callaway Capsules U, Y and V (0.064, 0.087 and 0.095). The Seabrook and Callaway results are comparable and representative of current industry-standard discrete ordinates methods results.

l The Cu, Fe and Ni activations are most representative of fast neutron flux greater than 1 MeV, as

l. indicated in Table 6-6. The U-238 and Np-237 activations cover lower neutron energy levels and require the corrections stated above, resulting in higher uncertainties. As a result, separate average CM and standard deviations are provided for Cu, Fe and Ni. For Seabrook, the Capsule U results for Cu, Fe and Ni show a CM ratio significantly closer to unity (0.977) and an improved standard deviation (0.050). The Capsule Y results for Cu, Fe and Ni show a CM ratio slightly closer to unity (0.900) and an improved standard deviation (0.038). The Callaway result for Cu, Ni and Fe also show higher CM ratios, but no improvement in standard deviations.

The measured reaction rates for Seabrook Capsules U and Y are translated to fast neutron flux greater than 1 MeV in Table 6-12 for each of the activation nuclides. Average measured fast

neutron fluxes and standard deviations for all nuclides are presented, as well as separate averages and standard deviations for Cu, Ni and Fe. These comparisons are used to establish best-estimates and uncertainties for future cycle projections. The Capsule Y C/M ratio,0.893, will be used to

estabilsh best-estimates for all projected parameters. The maximum standard deviation for all L activation nuclides, 0.104, at the two-sigma level, or 20.8%, is the uncertainty assigned to fast

, neutron flux greater than 0.1 MeV and dpa. The maximum standard deviation for Cu, Fe and Ni, 0.050, at the two-sigma level, or 10.0%, is the uncertainty assigned to fast neutron flux greater than 1 MeV. These uncertainties are marginally greater than, but similar in magnitude, to those applied to Callaway in WCAP-14859, which are based on a more sophisticated analysis of uncertainties.

6.6 BEST ESTIMATE REACTOR VESSEL EXPOSURE Best-estimate neutron exposure rate parameters for the surveillance capsules and within the reactor vessel wall at selected azimuths are provided for the Capsules U and Y in Tables 6-13.1 and 6-13.2, respectively. Best-estimate neutron integrated exposure parameters for the Capsule U and Y are presented in Tables 6-14.1 and 6-14.2, respectively. In addition to fast neutron flux and fluence greater than 1 MeV at %T and %T of the reactor vessel, neutron flux and fluence greater than 1 MeV are also provided utilizing the less-steep dpa fa'loff slope within the reactor vessel.

6.7 ' PROJECTIONS OF REACTOR VESSEL EXPOSURE

' Best-estimate neutron integrated exposure parameters for the surveillance capsules and within the reactor vessel wall at selected azimuths are projected to 32 EFPYs in Tables 6-15. These projections are based on the Capsule Y results and assume exposure rates equal to the Cycle 1 through 5 exposure rates. Best-estimate fast neutron fluences (E>1 MeV) at the surveillance capsules and the maximum at the base metal inner radius (IR) and %T locations are projected for the end of each operating cycle in Table 6-16. The table assumes 18-month operating cycles of 1.3 EFPYs each for Cycles 7 through 24, and extension of Cycle 25 to a total lifetime of 32 EFPYs.

Assuming 18-month cycles, including 40-day refueling outages, the 1.3 EFPY cycle lengths correspond to 93.5% capacity factor operation. The table identifies the cycles after which surveillance capsule withdrawals projected, in accordance with the requirements of ASTM E185-945.

__._____m__ . . _

I Table 61 I Seabrook Station Unit 1 Calculated Neutron Exposure Paramete.1 at the Center of the Surveillance Capsule Locations Parameter Units Cycle Capsules Capsules Ratio U and X V and Y UN or XN Fast Flux (E > 1 MeV) 10"n/cm'-sec 1 0.8840 0.8084 1.093 2 0.6487 0.6258 1.037 3 0.6888 0.6209 1.109 4 0.4937 0.4602 1.073 5 0.5376 0.4948 1.087

_ 1-5 0.6350 0 5865 1.083 Fast Flux (E > 0.1 MeV) 10"n/cm'-sec 1 3.1271 2.8166 1.11U 2 2.2832 2.1651 1.055 3 2.4098 2.1446 1.124 ,

4 1.7088 1.5753 1.085 5 1.8606 1,6940 1.099 15 2.2204 2.0236 1.097 Displacement Rate 10* dpa/sec 1.5473 1.4062 1 1.100 2 1.1341 1.0857 1.045 l 3 1.2005 1.0772 1.114 4 0.8573 0.7959 1.077 =l 5 0.9333 0.8557 1.091 1-5 1.1067 1.0170 1.088 II l l

I 1

I' N

L Table 6-2 Seabrook Station Unit 1 Calculated Azimuthal Variation of Neutron Exposuro Parameters at the Reactor Vessel Clad / Base Metal Interface f

1 Parameter Units Cycle o' 29' l

19-21.5' 31.5' 44-45' at Core Capsults Capsules at Core Flats V and Y U and X Diagonal" Azimuth Azimuth Fast Flux (E > 1 MeV) 10"n/cm'.sec 1 0.1424 0.2688 0.1518 0.1549 0.30fs 2 0.1327 0.2416 0.1211 0.1157 0.2106 3 0.1215 0.1987 0.1164 0.1210 0.236f 4 0.1049 0.7686 0.0887 0.0874 0.1563 5 0.0863 0.1638 0.0935 0.0946 0.1723 1-5 0.1144 0.2018 0.1113 0.1119 0.2099 Fast Flux (E > 0.1 MeV) 10"n/cm'-sec 1 0.2977 0.5648 0.4153 0.4347 0.7538 2 0.2777 0.5061 0.3282 0.3243 0.5242 3 0.2528 0.4168 0.3169 0.3366 0.5856 4 0.2182 0.3524 0.2387 0.2419 0.3853 5 0.1795 0.3414 0.2523 .0.2613 0.4247 15 0.2385 0.4225 0.3019 0.3117 0.5205 Iron Displacement Rate 10" dpa/sec 1 0.2182 0.4045- 0.2471 0.2543 0.4700 2 0.2034 0.3634 0.1967 0.1901 0.3279 3 0.1860 0.3002 0.1893 0.1982 0.3675 i 4 0.1608 0.2545- 0.1440 0.1431 0.2433 5 0.1325 0.2474 0.1518 0.1547 0.2880 15 0.1754 0.3043 0.1809 0.1835 0.3267 Values in bolditacce are the maximum values for any azimuth i

• • Maximura of values at 19,19.5,20,20.6,21 and 21.5' azimuths from 0.5' increment azimuthal mesh Maximum of values at 44,44.5 and 45' azimuths from 0.5* Increment azimuthal mesh

Table 6-3.1 Seabrook Station Unit 1 Relative Radial Distribution of Neutron Exposure Parameters within the Reactor VesselWcil Cycle 1 Results (Highest Fluence Rate Cycle)

Parameter Units Radial O' 19-21.5* 29' 31.5* 44-45' Location in at Core Capsules Capsules at Core Base Flats V and Y U and X Diagonal .

ustar Azimuth Azimuth '"

)

Fast Flux (E > 1 MeV) 10"n/cm'-sec IR 1.000 1.000 1.000 1.000 1.000 1/4 T 0.575 0.561 0.588 0.586 ' O.562 1/2 T 0.284 0.268 0.297 0.298 0.269 1 i

3/4 T 0.135 0.124 0.145 0.143 0.124 l OR 0.063 0.053 0.070 0.069 0.053 Fast Flux (E > 0.1 MeV) 10"n/cm'-sec lR 1.000 1.000 1.000 1.000 1.000 I 1/4 T 0.896 0.853 0.914 0.911 0.858 h

1/2 T 0.679 0.619 0.705 0.700 0.822 l 3/4 T 0.477 0.416 0.500 0.494 0.415 OR 0.303 0.243 0.314 0.311 0.233

)

Iron Displacement Rate 10* dpa/sec IR 1.000 1.000 1.000 1.000 1.000 1/4 T 0.648 0.630 0.696 0.695 0.652 1/2T 0.395 0.370 0.451 0.452 0.394 3/4 T 0.240 0.215 0.286 .0.285 0.233 )

OR 0.141 0.116 0.171 0.171 0.123 There are 20 radial mesh within the clad / base metal reactor pressure vessel region and results are provided for theindicated radiallocations.

The radial locations within the base metal correspond to the following radii in the modeling: 3 Radial Location in Base Metal Radius (in) Radius (cm)

Inner Radius 86.5 219.71  !

1/4 T 88.65625 225.19  !

1/2 T G0.8125 230.66 3/4 T 92.96875 236.14 Outer Radius 95.125 241.62 j Maximum of values at 19,19.5,20,20.5,21 and 21.5' azimuths from 0.5* increment azimuthal mesh l ,

Maximum of values at 44,44.5 and 45' azimuths from 0.5' increment azimuthal mesh

Table 6-3.2 Seabrook Station Unit 1 Relative Radial Distribution of Fast Neutron Exposure Parameters within the Reactor Vessel Wall Cycle 4 Results (Lowest Fluence Rate Cycle)

Parameter Units Radial 0* 19-21.5* 29' 31.5' 44-45' Location in "

at Core Capsules Capsules at Core Base Flats V and Y U and X Diagonal Metal

• Azimuth Azimuth *"

Fast Flux (E > 1 MeV) 10"n/cm'-sec IR 1.000 1.000 1.000 1.000 1.000 1/4 7 0.575 0.564 0.588 0.585 0.562 __

1/2 T 0.284 0.273 0.297 0.297 0.269 3/4 T 0.135 0.126 0.145 0.143 0.124 OR 0.062 0.055 0.070 0.069 0.053 Fast Flux (E > 0.1 MeV) 10"n/cm'sec 1R 1.000 1.000 1.000 1.000 1.000 1/4T 0.894 0.858 0.913 0.909 0.858 1/2 T 0.675 0.625 0.704 0.699 0.624 3/4 T 0.472 0.421 0.499 0.494 0.418 OR 'O.294 0.243 0.314 0.312 0.237 Iron Displacement Rate 10* dpa/sec IR 1.000 1.000 1.000 1.000 1.000 1/4T 0.647 0.632 0.693 0.692 0.650 1/2T 0.392 - 0.373 0.448 0.449 0.392 3/4T 0.237 0.217 0.283 0.283 0.232 OR 0.137 0.117 0.170 0.171 0.124

' There are 20 radial mesh within the cladtbase metal reactor pressure vessel region and results are provided for the indicated radiailocations.

The radial locations within the base tr:etal correspond to the following radiiin the modeling:

Radial Location in Base Metal' Radius (in) l Radius (cm) inner Radius ,86.5 219.71 1/4T 88.65625 225.19 1/2 T 90.8125 230.66 3/4 T 92.96875 236.14 Ou',r Radius 95.125 241.62

• • • Maximum of values at 19,19.5,20,20.5,21 and 21.5' azimuths from 0.5' increment azimuthat mesh Maximum of values at 44,44.5 and 45' azimuths from 0.5' increment azimuthal mesh I

l Table 6-3.3 Seabrook Station Unit 1 Difference in Relative Radial Distribution of Fast Neutron Exposure Parameters .

within the Reactor Vessel Wall Change from Cycle 1 to Cycle 4 Relative Radial Distributions (Highest to 1.owest Fluence Rate Cycles)

Parameter Units Radial O' 19-21.5* 29' 31.5' 44-45' Location in at Core Capsules Capsules at Core Base Flats V and Y U and X Diagonal Metal

• Azimuth Azimuth Fast Flux (E > 1 MeV) 10"n/cm'-sec IR 0.000 0.000 0.000 0.000 0.000 l l

1/4 T 0.000 0.003 0.000 0.001 0.000 l 1/2 T 0.000 0.005 0.000 -0.001 0.000 3/4T 0.000 0.002 0.000 0.000 0.000 OR -0.001 0.002 0.000 0.000 0.000 Fast Flux (E > 0.1 MeV) 10"n/cm'-sec IR 0.000 0.000 0.000 0.000 0.000 l 1/4 T -0.002 0.005 -0.001 0.002 0.000 1/2 T 4.004 0.006 -0.001 -0.001 0.002 3/4 T 0.005 0.005 -0.001 0.000 0.003 OR -0.009 0.000 0.000 0.001 0.004 Iron Displacement Rate 104 'dpa/sec IR 0.000 0.000 0.000 0.000 0.000 1/4T -0.001 0.002 -0.003 -0.003 -0.002 )

1/2 T 0.003 0.003 -0.003 -0.003 0.002 3/4 T -0.003 0.002 -0.003 -0.002 0.001

]

OR -0.004 0.001 -0.001 0.000 0.001

• There are 20 radiel mesh witido the clad / base metal reactor pressure vessel region and results are provided for the indicated radiallocations.

The radial locations within the base metal correspond to the following radil in the modeling: .

Radial Location in Base Metal- Radius (in) Radies (cm)'

inner Radius 86.5 219.71 ,

$/4T 88.65625 225.19 1/2 T 90.8125 230.66 I 3/4 T 92.96875 236.14 Outer Radius 95.125 241.62  ;

Difference of maximum values at 19,19.5,20,20.5,21 and 21.5* azimuths from 0.5' increment azimuthal mesh Difference of maximum values at 44,44.5 and 45' azimuths from 0.S' increment azimuthal mesh i

i '

Table 6-4.1 Seabrook Station Unit 1 Calculated Fast Neutron Flux, Fluence and Lead Factors for Capsules U and X  !

at the Reactor Vessel Clad / Base Metal interface I at the 19-21.5* AzimtfJi l

Cycle Flux, E>1MeV (10"n/cm'sec) Lead Effective Fluence, E>1MeV (10n/cm')

Factor Full-Power '

Capsules 19-21.5* Years Capsules 19-21.5' U and X on Vessel U and X on Vessel 1 0.8840 'O.2608 S.288 0.913 0.2546 0.0774 i 2, 0.6487 l' 0.2416 2.685 0.872 0.1786 0.0685 3 0.6888 0.1987 3.46J 1.207 02623 0.0757 l

4 0.4937 0.768r 2.929 1.208 0.1883 0.0643 {

5 0.5376 0.1638 3.282 1.371 0 2326 0.0709 1-5 0.6350 0.2018 3.147 5.572 1.1164 0.3548 ,  !

l VcJues in boldItalles are the maximum values for any azimuth j i

I at the 44-45' Azimuth l

L.  !

Cycle Flux, E>1MeV (10"n/cm*sec) Lead Effective Fluence, E>1MeV Factor (10'd r bm') ~)

Full-Power  !

Cysules 44 45' Years . Capsubs 44 -45' U and X on Vessel U and X __ on Vsssel i 1 0.8840 0.30f8 2.929 0.913 02546 0.0869 I i

-2 0.6487 0.2106 3.080 0.872 0.1786 0.0580 l 3 0.6888 0.236f 2.917 1.207 0.2623 0.0890 i

I

! 4. 0.4937 0.1563 3.158 1.208 0.1883 0.0596 5 0.5376 0.7723 3.120 1.371 0.2326

0. 'Jg, l 5 0.6350 0.2099 3.025 5.r2 1.1164, _ 0.gt90_J l Values in bolditalles are the maximum values for any azimuth 1

Table 6-4.2 Seabrook Station Unit 1 Calculated Fast Neutron Flux, Fluence and t.aad Factors for Capsules V and Y at the ficactor Vessel Clad / Base Metal interface at tis 19-21.5' Azimuth Cycle Flux, E>1MeV (10"r.' ..".ec) Lead Effective Fluence, E>1MeV (10n/cm') l Factor Full-Power .

Capsules 19-21.5' Years Capsules 19-21.5' J V and Y on Vessel V and Y on Vessel 1 0.8084 0.2688 3.008 0.913 0.2328 0.0774 2 0.6258 0.24f6 2.590 0.872 0.1723 0.0665 3 0.6209 0.1987 3.124 1.207 0.2365 0.0757 l

4 0.4602 0.1886 2.729 1.208 0.1755 0.0643 5 0.4948 0.1638 3.021 1.371 0.2141 0.0709 15 0.5865 0.2018 2.906 f.572 1.0312 0.3548 Values in boMitalics are the maximum values for any azimuth  ;

)

I at the 44-45' Arimuth l Cycle Flux, E>1MeV (10"n/cm'sec) Lead . Effective Fluence, E31MeV (10DNem')

Factor - Full-Power Capsules 44 45' Years Capsules 44 45' V and Y on Vessel V and Y on Vessol i 1 0.8084' O.30f8 2.679 0.913 02328 0.086W 2 0.6258 0.2106 2.971 0.872 0.1723 0.0580 j i

3 0.6209 0.238f 2.629 1.207 - 0.2365 0.0899 4 0.4802. 0.1563 2.944 1.208 0.1755 0.0596 1 5 0.4948 0.1723 2.872 1.371 0.2141 0.0748 i 15 0.5065 - 0.2099 2.794 5.572 1.0312 0.3890 i

I

- Values h hoMitalics are the maximum values for any azimuth l

%- -.~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _

Table 65.1.1 Saabrook Station Unit 1 Calculated Fast Neutron Energy Spectrum at the Corter of C6poules U and X Group Upper Average Flux (10" nf crif-sec )

Energy Ene m (MeV) (MeVL Cycle 1 Cyc$)2 Cycle 3 Cg, Cycle 5 Cycles 1-5

{

1 17.332 15.762 0.000129 0.000iG3 0.000111 0.00C088 0.000094 0.000103 2 14.191 13.203 0.000419 0.000329 0.000353 0.000275 0.000297 0.000329 3 12.214 11.107 0.00186

_ 300144 0.00155 0.00118 0.00128 0.00144 4 10.000 9.3036 I 0.00370 0.00285 0.00306 0.00233 0.00253 0.00284 5 8.6071 8.0077 0.00855 0.00500 0.00537 0.00405 0.00440 0.00498 i 6 7.4082 6.7368 0.0161 0.0123 i

0.0131 0.0099 0.0107 0.0122 7 6.0653 5.5156 0.t.LN2 0.0190 0.0203 0.0151 0.0164 0.0188 8 4.9659 4.0224 0.0!l21 0.0307 0.0413 0.0301 0.0327 0.0381 i

9 3.6788 3.3454 0.043'3 0.0320 0.0340 0.0246 0.0267 0.0314 to 3.0119 2.8686 0.0332 0.0244 0.0260 0.0187 0.0204 0.0240 11 2.7253 2.5957 0.0400 0.0293 0.0312 0.0224 0.0244 0.0288 i2' 2.4660 2.4157 0.0200 0.0147 0.0156 0.0112 0.0122 0.0144 13 2.3653 2.3555 0.0356 0.0041 0.00435 0.0031 0.0034 0.0040 14 2.3457 2.2885 0.0280 0.0205 0.0218 0.0157 0.0171 0.0201 15 2.2313 2.0759 0.0782 0.0573 0.0609 0.0436 0.0475 0.0561 16 1.9205 1.7868 0.0939 0.0687 0.0728 0.0520 0.0566 0.0671 17 1.6530 1.5032 0.1473 0.t076 0.1141 0.0812 0.0885 0.1051 18 1.3534 1.1780 0.286 0.209 0.221 0.157 0.171 0203 19 1.0026 0.91173 0.210 0.15:f 1 0.162 0.115 0.125 0.149 i 20- 0.82086 0.78180 0.105 0.077 0.081 0.057 0.063 0.075 21 0.74274 ! 0.67542 0.291 0.212 G223 0.158 0.172 0.206 k- -.

22 0.608)0 0.56299 0.270 0.196 0.207 0.146 0.159 0.191 23 0.49787 0.43333 0.276 0.201 0.212 0.150 0.163 0.195 L 24 0.368&3 0.33302 0.264 0.193 0.202 L - 0.143 0.155 0.187 25 02G721 0.24019 0.398 0.290 0.305 0.215 0.234 0.281

, 26 0.18316 0.14713 0.372 0.271 C.285 '

0.201 0.218 0.262 27 0.11109 0.08923 0.286 0.208 0.219, 0.154 0.168 0.201

> 1MeV 0.884 0.649 0.609 0.494 0.538 0.635

>0.1 MeV 3.127 2.283 2.410 1.709. 1 361 2.220 I l

Table 6-5.1.2 Seabrook Station Unit 1 Calculatad Epithermalfrhermal Neutron Energy Spectrum at the Center of Capsules U and X

-n:ma Group . Upper Average Flux (10" n/ cm'-sec )

Energy Energy (kev) (kev) Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycles 1-5 27 111.090 89.235 0.286 . 0.208 g 0.219 0.154 0.168 0.201 28 67.379 54.124 0.256 0.186 0.196 0.138 0.150 0.180

' 29 40.868 36.348 0.1022 0.0744 0.0782 0.0550 0.0599 0.0720-30 31.828 28.943 0.0551 0.0401 0.0421 0.0297 0.0323 0.0388 31 26.058 25.117 0.0788 0.0574 0.0603 0.0424 0.0462 0.0555 32 ^ 24.176 23.026 0.0489 0.0356 0.0374 0.0263 0.0286 0.0344

33. 21.875 18.455 0.1461 0.1063 0.1117 0.0786 0.0856 0.1029 34 15.034 11.068 0.291 0.212 0.223 0.157 0.171 0.205 35 7.1017 5.2282 0.301 0.219 0.230 0.162 0.176 ' O.212 l 1

36~ 3.3546 ?46fd 0.284 0.206 0.217 0.152 0.166 0.200 37 1.5846 1.0193 0.472 0.344 0.361 0.254 0.276 0.333 1

38 0.45400 0.33423 0.266 0.193 0.203 0.143 0.155 0.187 39 0.21445 0.15788 0.283 0.206 0.217 0.152. 0.166 0.199 40 0.10130 0.069283 0.367 0.267 0.280 0.197 0.214 0.258 41 0.037266 0.023972 0.455 0.331 0.347 0.244 0.266 0.320 42 0.010677 0.007860 0.264 0.192 0.202 0.142 0.154 0.186 43 0.005044 G.003449 0.346 0.252 0.264 0.186 0.202 0.243 44 0.001855 0.001366 0.250 0.182 0.191 0.134 0.146 0.176 45 0.000876 0.000645 0.217 0.158 0.165 0.116 0.126 0.152 i I

46

• 0.000414 0.000257 0.329 0.239 0.251 0.176 0.192 0.231 l 47 0.000100 0.000050 1.056 0.762 0.797 0.554 0.602 0.733

< 0.1 6.094 4.427 4.645 3.258 3.647 4.277 l MeV

<0.414 1.385 1.001 1.048 0.730 0.793 0.964 l eV l l

l 1

I L

Table 6-5.2.1 Seabrook Station Unit 1 Calculated Fast Neutron Energy Spectrum at the Center of Capsules V and Y Group Upper Average Flux Energy Energy (10"n/ cm ec)

(McV) (MeV) Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycles 1-5 1 17.332 15.762 0.000124 0.000102 0.000w3 0.000085 0.000091 0.000100 2 14.191 13.203 0.000402 0.000326 0.000334 0.000266 0.000285 0.000317 3 12.214 11.107 0.00178 0.00143 0.00145 0.00114 0.00122 0.00138 4 10.000 9.3036 0.00352 0.00283 0.00288 0.00225 0.00241 0.00272 5 8.6071 8.0077 0.00621 0.00495 0.00502 0.00389 0.00417 0.00475 6 7.4082 6.7368 0.0153 0.0121 0.0123 0.0095 0.0101 0.0116 7 6.0653 5.5156 0.0238 0.0188 0.0189 0.0144 0.0155 0.0179 8 4.9659 4.3224 0.0488 0.0382 0.0381 0.0286 0.0307 0.0360 9 3.6788 3.3454 0.0403 0.0314 0.0311 0.0233 0.0250 0.0295 10 3.0119 2.8686 0.0308 0.0240 0.0238 0.0177 0.0190 0.0225 11 2.7253 2.5957 0.0370 0.0287 0.0285 0.0211 0.0227 0.0269 12 2.4600 2.4157 0.0185 0.0143 0.0142 0.0105 0.0113 0.0134 13 2.3653 2.3555 0.0052 0.004 0.00397 0.0029 0.0032 0.0038 14 2.3457 2.2885 0.0259 0.0200 0.0199 0.0147 0.0158 0.0188 15 2 i 313 2.0759 0.0717 0.0555 0.0550 0.0407 0.0438 0.0520 16 1.9205 1.7868 0.0856 0.0662 0.0655 0.0483 0.0520 0.0618 17 1.6530 1.5032 0.1337 0.1032 0.1022 0.0752 0.0809 0.0964 18 1.3534 1.1780 0.257 0.198 0.196 0.144 0.155 0.185 19 1.0026 0.91173 0.189 0.145 0.144 0.105 0.113- 0.135 20 0.82085 0.78180 0.095 0.073 0.072 0.053 0.057 0.068 21 0.74274 0.67542 0.261 0.200 0.198 0.145 0.156 0.187 22 0.60810 0.55299 0.242 0.185 0.183 0.134 0.144 0.173 23 0.49787 0.43335 0.247 0.190 0.188 0.137 0.148 0.177 24 0.36883 0.33302 0.236 0.181 0.179 0.131 0.141 0.169 25 0.29721 0.24019 0.356 0.272 0.270 0.197 0.212 0.254 26 0.18316 0.14713 0.332 0.254 0.252 0.184 0.198 0.237 27 0.11109 0.08923 0.255 0.195 0.193 0.141 0.152 0.182

> 1MeV 0.808 0.626 0.621 0.460 0.495 0.586 v0.1 MeV 2.817 2.165 2.145 1.575 1.694 2.024

Table 6-5.2.2 Seabrook Station Unit 1 3

Calculated Epithermal / Thermal Neutron Energy Spectrum at the Center of Capsules V and Y Group Upper Average Flux (10" n/ cm'-sec )

Energy Energy (kev) (kev) Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycles 15 ,

27 111.090 89.235 0.255 0.195 0.193 0.141 0.152 0.182 28 67.379 54.124 0.228 0.175 0.173 0.126 0.136 0.14 29 40.868 36.348 0.0910 0.0697 0.0690 0.0503 0.0541 0.0650 30 31.828 28.943 0.0491 0.0376 0.0372 0.0271 0.0292 0.0350 31 26.058 25.117 0.0703 0.0538 0.0532 0.0389 0.0418 0.0502 32 24.176 23.026 0.0435 0.0333 0.0329 0.0240 0.0259 0.0311 i

33 21.875 18.455 0.1301 0.0996 0.0986 0.0719 0.0773 0.0920 )

34 15.034 11.068 0.259 0.198 0.196 0.143 0.154 0.185 35 7.1017 5.2282 0.268 0.205 0.203 0.148 0.159 0.191 36 3.3546 2.4696 0.252 0.193 0.191 0.139 0.150 0.180 37 1.5846 1.0193 0.420 0.321 0.318 0.232 0.249 0.299 38 0.45400 0.33423 0.236 0.181 0.179 0.130 0.140 0.168 39 0.21445 0.15788 0.251 0.192 0.190 0.139 0.149 0.179 40 0.10130 0.069283 0.325 0.249 0.246 0.179 0.193 0.232 41 0.037266 0.023972 0.403 0.308 0.305 0.222 0.239 0.287 42 0.010677 0.007860 0.234 0.179 0.177 0.129 0.138 0.166 43 0.005044 0.003449 0.306 0.234 0.231 0.168 0.181 0.218 44 0.001855 0.001366 0.221 0.169 0.167 0.121 0.131 0.157 45 0.000876 0.000645 0.191 0.146 0.145 0.105 0.113 0.136  !

l 46 0.000414 0.000257 0.290 0.222 0.219 0.160 0.171 0.207 47 0.000100 0.000050 0.923 0.702 0.690 0.499 0.534 0.650 i

< 0.1 5.392 4.119 4.072 2.903 3.185 3.837 MeV

<0.414 1.213 0.923 0.909 0.658 0.706 0.857 eV

n*

od) 2 7 is le% - - - 0 1 - -

ai FY( 6 6 s s s s s r

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up M M M M 8 < 5 es 7 0 0 4 0 V 1 0

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> > > > > 4 <

it n E E E E E 0 E o

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8 7 3 5 8 3 3 8 8 iy t A

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( ,

Table 6-7 Seabrook Station Unit 1 Cycles 1 through 5 Operation Operating Dates, Days and Effective Full-Power Years -

Cycle Startup Shutdown Days of Days of Total Days from Days from

- Date Date Prior Power Days Shutdown Startup to Outage Operation to Cycle 1 Cycle 5 j Startup Shutdown 1 3GNV90 7/25/91 0 492 492 492 2608 2 10/9/91 9/7/92 76 334 410 902 2040 3 11/11/92 4/8/94 65 513 578 1480 1641 4 7/29/94 11/4/95 112 463 575 2055 1016 5 12/9/95 5/10/97 35 518 553 2608 518 Total 288 2320 2608 Cycle 1 date is approximately the date at which measurable bumup accumulation started.

l 1

Cycle As-Built Cycle Bumup Effective Effective Core Loading Full Power Full Power Days Years (MtU) (mwd /Mt) (EFPDs) (EFPYs) 1- 89.431 12715 333.4 0.913 2 89.385 12159 318.6 0.872 3 89.396 16820 440.8 1.207 4 89.393 16842 441.4 1.208 5 88.780 19243 500.8 1.371 Total ,

2035.0 5.572 l

l l

1 i

L- .

Table 6-8 Seabrook Cycle 1 Power History for Capsule U Activation Ratio of Measured to Saturated Activities at the End-of Cycle 1 Day, Cycle Average Power Decay to Month Sumup Power Time EOC1 Co-60 Mn-54 Co-58 Cs 137 and Year (mwd /Mt) (Mwt) (devs) (days) 3/20/90 0 0.0 0 492 0.0000 0.0000 0.0000 0.00000 3/31/90 15 122.0 11 481 0.0001 0.0003 0.0000 0.00002 4/11/90 40 203.3 11 470 0.0002 0.0005 0.0001 0.00004 6/6/90 129 142.1 56 414 0.0007 0.0019 0.0003 0.00014 8/30/90 355 842.1 24 390 0.0018 0.0054 0.0011 0.00037 7/31/90 1085 2106.0 31 359 0.0060 0.0185 0.0048 0.00118 8/31/90 2035 2740.6 31 328 0.0079 0.0258 0.0085 0.00154 9/30/90 3114 3216.5 30 298 0.0091 0.0314 0.0130 0.00175 10/31/90 4153 2997.4 31 267 0.0089 0.0323 0.0169 0.00169 11/30/90 4688 1594.9 30 237 0.0046 0.0178 0.0117 0.00087 12/31/90 5869 3407.0 31 206 0.0103 0.0421 0.0348 0.00193 1/31/91 7050 3407.0 31 175 0.0104 0.0451 0.0471 0.00194 2/28/91 7830 2491.3 28 147 0.0069 0.0318 0.0415 0.00128 3/31/91 8927 3164.7 31 116 0.0099 0.0477 0.0780 0.00180 4/30/91 9829 2688.9 30 86 0.0082 0.0420 0.0864 0.00149 5/31/91 11012 3412.8 31 55 0.0109 0.0589 0.1528 0.00195 6/2/91 11061 2191.1 2 53 0.0005 0.0025 0.0074 0.00008 6/5/91 11110 1460.7 3 50 0.0005 0.0025 0.0076 0.00008 6/27/91 11931 3337.4 22 28 0.0076 0.0438 0.1441 0.00136 7/1/91 11982 1140.2 4 24 0.0005 0.0028 0.0101 0.00008 7/4/91 ~ 12072 2682.9 3 21 0.0008 0.0050 0.0185 0.00015 7/9/91 12120 858.5 5 16 0.0005 0.0027 0.0103 0.00008 7/25/91 '12715 3325.7 16 0 0.0056 0.0340 0.1413 0.00099 Totals 492 0.1119 0.4948 0.8366 0.02083

Table 6-9.1 Seabrook Cycle 1 Power History for Capsule Y Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 5 Day, Cycle Average Power Dectsy to i Month Bumup Power Time EOC 5 Co-80 Mn-54 Co-58 Cs 137 )

and Year (mwd /Mt) (Mwt) (da'/s) (days) '

3/20/90 3 0.0 0 2608 0.0h00 0.0000 0.0000 0.0000 3/31/90 15 122.0 11 2597 0.0001 0.0000 0.0000 0.0000 4/11/90 40 203.3 11 2586 0.0001 0.0000 0.0000 0.0000 6/6/90 129 142.1 56 2530 0.0003 0.0000 0.0000 0.0001 j 6/30/90 355 842.1 24 2506 0.0000 0.0000 0.0000 0.0003 7/31/90 1085 2106.0 31 2475 0.0028 0.0002 0.0000 0.0010 8/31/90 2035 2740.6 31 2444 0.0037 0.0002 0.0000 0.0013 9/30/90 31*.4 3216.5 30 2414 0.0042 0.0003 0.0000 0.0015 10/31/90 4153 299'/.4 ' 31 2383 0.0041. 0.0003 0.0000 0.0015 11/30/90 4688 1594,9 50 2353 0.0022 0.0002 0.0000 0.0008 12/31/90 5869 3407.0 31 2322 0.0048 0.0004 0.0000 0.0017 t/31/91 7050 3407.0 31 2291 0.0049 0.0004 0.0000 0.0017 2/28/91 l 7830 2491.3 28 2263 0.0032 0.0003 0.0000 0.0011 3/31/91 8927 3164.7 31 2232 0.0046 0.0004 0.0000 0.0016 4/30/91 9829 2688.9 30 2202 0.0038 0.0004 0.0000 0.0013 5/31/91 11012 3412.8 31 2171 0.0051 0.0005 0.0000 0.0017 6/2/91 ~ 11061 2191.1 2 2169 0.0002 0.0000 0.0000 0.0001 6/5/91 11110 1460.7 3 2166 0.0002 0.0000 0.0000 0.0001

-m u.

6/27/91 '11931 3337.4 22 2144 0.0036 0.0004 0.0000 0.0012 7/1/91 11982 1140.2 4 2140 0.0002 0.0000 0.0000 0.0001 7/4/91 12072 2682.9 3 2137 0.0004 0.0000 0.0000 0.0001 7/9/91 - 12120 858.5 5 2132 0.0002 0.0000 0.0000 0.0001 7/25/91 12715 3325.7 16 2116 0.0026 0.0003 0.0000 0.0009 lotals 492 0.0523 0.0045 0.0000 0.0182 L

)

i l

l l

(

Table 6-9.2 Seabrook Cycle 2 Power History for Capsule Y Activation Ratio of Measured 1o Saturated ActMties at the End-of-Cycle 5 '

l Day, Cycle Average Power Decay to Month Bumup Power Time EOC5 Co-60 Mn-54 Co-58 Cs 137 j and Year (mwd /Mt) (Mwt) (days) (days) 10/lW91 0 0.0 0 2040 0.0000 0.0000 0.0000 0.0000 10/24/91 155 923.6 15 2025 0.0007 0.0001 0.0000 0.0002 1028/91 304 3329.6 4 2021 0.0007 0.0001 0.0000 0.0002 12/23/91 2421 3379.1 56 1965 0.0097 0.0015 0.0000 0.0031 12/25/91 2464 1921.8 2 1963 0.0002- 0.0000 0.0000 0.0001 l 1/31/92 3873 3403.9 37 1926 0.0066 0.0011 0.0000 0.0021 2/29/92 4980 3412.0 29 1897 0.0052 0.0009 0.0000 0.0016 3/31/92 6163 3411.0 31 1866 0.0057 0.0011 0.0000 0.0017 4/30/97 7307 3408.5 30 1836 0.0055 0.0011 0.0000 0.0017 f i

5/31/92 8490 3411.0 31 1805 0.0058 0.0012 0.0000 0.0017 6/30/92 9633 3405.6 30 1775 0.0057 0.0012 0.0000 0.0017 7/31/92 10815 3408.2 31 1744 0.0059 ~ 0.0014 0.0000 0.0018 i 1

8/31/92 11993 3J96.6 31 1713 0.0060 0.0015 0.0000 0.0018 "'

9/3/92 12098 3128.5 3 1710 0.0005 0.0001 0.0000 0.0002 9/7/92 12159 1363.1 4 1706 0.0003 0.0001 0.0000 0.0001 Totals 334 0.0586 0.0114 0.0000 0.0179

(. ,

i i

l l

3 f

I

Table 6-9.3 Seabrook Cycle 3 Power History for Capsule Y Activation Ratio of Measured to SaNrated Activities at the End-of-Cycle 5 l Day, Cycle Average Power Decay to 3 Month Bumup Power Time EOC5 (days)

Co-60 Mn-54 Co-58 Cs-137 g and Year (mwd /Mt) (Mwt) (dayjs 11/11/92 0 0.0 0 1641 0.0000 0.0000 0.0000 0.0000 11/30/92 425 1999.6 19 1622 0.0022 0.0007 0.0000 0.0006 E 12/13/92 895 3232.0 13 1609 0.0025 0.0008 0.0000 0.0007 1/21/93 1958 2436.6 39 1570 0.0057 0.0018 0.0000 0.0016 2/28/93 3408 3411.2 38 1532 0.0078 0.0027 0.0000 0.0022 3/31/93 4591 3411.5 31 1501 0.0065 0.0024 0.0000 0.0018 4/30/93 5734 3406.0 30 1471 0.0063 0.0025 0.0000 0.0017 5/20/93 6467 3276.4 20 1451 0.0041 0.0017 0.0000' O.0011 5/24/93 6529 1385.6 4 1447 0.0003 0.0001 0.0000 0.0001 6/30/93 7940 3409.1 37 1410 0.0080 0.0034 0.0000 0.0021 7/27/93 8946 3330.8 27 1383 0.0057 0.0026 0.0000 0.0015 7/31/93 8993 1050.4 4 1379 0.0003 0.0001 0.0000 0.0001 8/31/93 10175 3408.6 31 1348 0.0068 0.0033 0.0000 0.0018 9/22/93 10986 3295.5 22 1326 0.0047 0.0024 0.0000 0.0012 10/8/93 11018 178.8 16 1310 0.0002 0.0001 0.0000 0.0000 11/14/93 12413 3370.5 37 1273 0.0083 0.0046 0.0000 0.0021 11/25/93 12654 1958.6 11 1262 0.0014 0.0008 0.0000 0.0004 12/31/93 14010 3367.2 36 1226 0.0082 0.0050 0.0000 0.0021 1/25/94 14942 3332.7 25 1201 0.0057 0.0037 0.0000 0.0014 2/18/94 14983 152.7 24 1177 0.0003 0.0002 0.0000 0.0001 4/3/94 16662 3411.3 44 1133 0.0105 0.0075 0.0000 0.0026 4/8/94 16820 2824.9 5 1128 0.0010 0.0007 0.0000 0.0002 Totals 513 0.0964 0.0472 0.0000 0.0255 I

~

Table 6-9.4 Seabrook Cycle 4 Power History for Capsule Y Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 5 l Day, Cycle Average Power Decay to l Month Bumup Power Time EOC5 Co-60 Mn 54 Co-58 Cs-137 and Year (mwd /Mt) (Mwt) (days) (days) 7/29/94 0 0.0 0 1016 0.0000 0.0000 0.0000 0.0000 8/5/94 120 1532.5 7 1009 0.0008 0.0007 0.0000 0.0002 8/31/94 1111 3407.2 26 983 0.0065 0.0063 0.0000 0.0015 9/30/94 2256 3411.8 30 953 0.0076 0.0078 0.0000 0.0018 10/31/94 3439 3411.4 31 922 0.0080 0.0086 0.0000 0.0018 11/30/94 4585 3414.8 30 892 0.0078 0.0089 0.0000 0.0018 12/31/94 5767 3408.5 31 861 0.0081 0.0098 0.0001 0.0019 1/31/95 6949 3408.5 31 830 0.0082 0.0105 0.0001 0.0019 2/28/95 8019 3416.1 28 802 0.0075 0.0102 0.0001 0.0017 3/31/95 9201 3408.5 31 771 0.0084 0.0120 0.0001 0.0019 4/30/95 10345 3408.D 30 741 0.0082 0.0124 0.0002 0.0018 5/31/95 11527 3408.5 31 710 0.0086 0.0137 0.0003 0.0019 6/18/95 12206 3372.1 18 692 0.0050 0.0083 0.0002 0.0011 7/4/95 12234 156.4 16 676 0.0002 0.0004 0.0000 0.0000 7/7/95 12327 2771.2 3 673 0.0007 0.0012 0.0000 0.0001 7/31/95 13243 3411.8 24 649 0.0068 0.0123 0.0004 0.0015 8/20/95 14006 3410.3 20 629 0.0057 0.0107 0.0004 0.0012 8/24/95 14155 3329.9 4 625 0.0011 0.0022 0.0001 0.0002 9/30/95 15567 3411.4 37 588 0.0107 0.0214 0.0010 0.0023 11/2/95 16826 3410.5 33 555 0.0097 0.0206 0.0012 0.0020 11/4/95 16842 715.1 2 553 0.0001 0.0003 0.0000 0.0000 Totals 463 0.1198 0.1783 0.0041 0.0266

Table 6-9.5 Seat. rook Cycle 5 Power History for Capsule Y Activation Ratio of Measured to Saturated Activities at the End-of-Cycle 5 Day, Cycle Average Power Decay to 3 and Year Month Bumup Power Time (days)

EOC5 (days)

Co-60 Mn-54 Co-58 Cs-137 g (mwd /Mt) (Mwt) 12/9/95 0 0.0 0 518 0.0000 0.0000 0.0000 0.0000 12/31/95 702 2832.9 22 496 0.0055 0.0132 0.0013 0.0011 E 1/30/96 1721 3015.6 30 466 0.0080 0.0202 0.0024 0.0016 2/22/96 2284 2173.2 23 443 0.0045 0.0119 0.0017 0.0009 2/29/96 2553 3411.7 7 436 0.0022 0.0059 0.0009 0.0004 3/31/96 3744 3410.9 31 405 0.0096 0.0271 0.0050 0.0019 4/30/96 4895 3406.2 30 375 0.0044 0.0280 0.0065 0.0018 5/31/96 6086 3410.9 31 344 0.0098 0.0310 0.0090 0.0019 6/30/96 7238 3409.2 30 314 0.0096 0.0321 0.0118 0.0019 7/31/96 8430 3413.7 31 283 0.0100 0.0355 0.0164 0.0019 8/31/96 9620 3408.0 31 252 0.0101 0.0380 0.0222 0.0019 9/30/96 10773 3412.1 30 222 0.0099 0.0394 0.0290 0.0019 10/31/96 11959 3396.6 31 191 0.0103 0.0433 0.0402 0.0019 11/30/96 13111 3409.2 30 161 0.0101 0.0451 0.0520 0.0019 12/31/96 14299 3402.3 31 130 0.0106 0.0497 0.0731 0.0019 1/31/97 15491 3413.7 31 99 0.0107 0.0534 0.0994 0.0019 2/28/97 16567 3411.7 28 71 0.0098 0.0515 0.1197 0.0018 3/31/97 17758 3410.9 31 40 0.0109 0.0609 0.1769 0.0020 4/30/97 18909 3406.2 30 10 0.0107 0.0629 0.2304 0.0019 m 5/8/97 19217 3418.0 8 2 0.0029 0.0176 0.0740 0.0005 5/10/97 19243 1154.1 2 0 0.0002 0.0015 0.0066 0.0000 Totals 518 0.1648 0.6681 0.9789 0.0312 I

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Table 6-11.1 Seabrook Station Unit 1 Capsules U and Y Summary of Calculated and Measured Reaction Rates for Fast Flux Nuclides Capsule U at 0.913 EFPYs Reaction Rates (1048tps/ atom) Ratio Activation Nuclide Calculated Measured C/M Cu-63 0.06022 0.06550 0.919 Fe-54 6.372 6.374 1.000 Ni-58 8.766 8.660 1.012 U-238 31.031 39.502* 0.786 Np-237 247.38 304.70** 0.812 Average -Cu, Fe, NI 0.977 Standard Deviation 0.050 Average - All 0.906 Standard Deviation 0.104 No correction for U-235 Impuity fissions, Pu build-in and y-induced fissions No correction for y-induced fissions Capsule Y at 5.572 EFPYs Reaction Rates (1048rps/ atom) Ratio Activation Nuclide Calculated Measured C/M Cu-63 0.04361 0.04836 0.902 Fe-54 4.430 4.727 0.937 Ni-58 6.064 7.042 0.861 U-238 20.908 21.668* 0.965 Np-237 162.02 198.18** 0.818 Average -Cu, Fe, Ni 0.900 Standard Deviation 0.038 Average- All 0.897 Standard Deviation 0.059 No correction for U-235 impulty fissions, Pu build-in and y-induced fissions No correction for y-induced fissions

Table 6-11.2 Callaway Unit 1 Capsules U, Y and V Summary of Calculated and Measured Reaction Rates for Fast Flux Nuclides in WCAP-14895 Capsule U at 1.05 EFPYs Reaction Rates (10- rps/ atom) Ratio Activation Nuclide Calculated Measured C/M Cu-63 0.0532 0.0596 0.893 Fe-54 6.09 6.09 1.000 Ni-58 8.54 8.42 1.014 U-238 32.0 36.0* 0.889 Np-237 305 302 " 1.010 Average -Cu, Fe, N1 0.969 Standard Deviation 0.066 Average - All 0.961 Standard Deviation 0.064 Measurement corrected by a factor of 0.84 for U-235 impuity resions, Pu build-in and y-induced fusions Measurement corrected by a factor of 0.99 for y-induced fissions i

Capsule Y at 4.6 EFPYs Reaction Rates (10-'8 rps/ atom) Ratio Activation Nuclide Calculated Measured C/M Cu-63 0.0456 0.0513 0.889 Fe-54 5.15- 4.69 1.098 Ni-58 7.22 6.80 1.062 U-238 27.0 29.0* 0.931 Np-237 257 261 " 0.985 Average Cu, Fe, Ni 1.016 Stantiard Deviation 0.112

- Average All 0.993 Standard Deviation 0.087 Measurement corrected by a factor of 0.81 for U-235 impulty fissions, Pu build-in and y-induced fissions Measurement rnrrected by a factor of 0.99 for y-induced fissions l

Table 6-11.2 (continued) l

~

Callaway Unit 1 Capsules U, Y and V Summary of Calculated and Measured Reaction Rates I for Fast Flux Nuclides in WCAP 14895 i ,

-l

)

. Capsule V at 9.85 EFPYs  !

Reaction Rates (10 rps/ atom) Ratio l l

Activation Nuclido Calculated Measured C/M '

Cu-63 0.0421 0.0465 0.905 i Fe-54 4.76 4.30 1.107 l l Ni-58 6.68 5.94 1.125  !

U-238 25.0 26.2* 0.954 i

No-237 237 230" 1.030 +

Average -Cu, Fe, Ni 1.046 j I

Standard Deviation 0.122 Average All 1.024 I

- Standard Deviation 0.095 i

Measurement corrected by a factor of 0.77 for U-235 impulty ficalons, Pu build-in and y-induced fissions l Measurement corrected by a factor of 0.99 for y-induced fissions  !

i t

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C_.__ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ . _

Table 6-12

! Seabrook Station Unit 1 Capsules U and Y l-

Summary of Fast Neutron Flux and Fluence (E>1 MeV) Results at Capsule Center Calculated and Measured i

i Fast Flux, E>1MeV Capsule U . (Cycle 1) Capsule Y (Cycles 15)

(10" n/cm*sec) Value Ratio C/M* Value Ratio C/M*

Calculated 0.8840 0.5865 Measured Co-63 0.9615 0.919 0.6504 0.902 Fe-54 0.8842 1.000 0.6258 0.937 Ni-58 0.8733 1.012 0.6811 0.861 U-238- 1.1253 0.786 0.6078 0.965 Np-237 1.0888 0.812 0.7173 0.818 Measured Average Cu, Fe, Ni 0.9063 0.975 0.6524 0.899 Standard Deviation 0.050 0.038 l

Measured Average- All 0.9866 0.896 0.6565 0.893 Standard Deviation 0.104 0.059 EFPYs at Capsule Discharge 0.913 5.572 Fast Fluence, E >1 MeV 8

(10"n/cm )

Calculated 0.255 1.031 Measured - Cu, Fe, Ni 0.261 ~0.975 1.147 0.899 i 1

Measured All 0.284 0.896 1.154- 0.893 1

The C/M ratio for the measured average fast fluxes is the ratio of the calculated fast flux to the average of the measured fast fluxor, i

Table 6-13.1 Seabrook Station Unit 1 Best-Estimate Neutron Exposure Rate Parameters at the Surveillance Capsules and within the Reactor Vessel Wall Average through 0.913 Effective Fu!!-Power Years for End-of-Cycle 1 using Capsule U Calculated-to-Measured Ratio of 0.896 Parameter Units Capsule O' 19-21.5' 29' 31.5' 4445' or Radial at Core Capsules Capsules at Core Location Flats V and Y U and X Diagonal in Azimuth Azimuth "*

Base Metal' Fr.st Flux (E > 1 MeV) 10"n/cm'-sec Capsule - -

0.9023 0.9866 -

lR 0.1589 0.3000 0.1694 0.1728 0.3368 1/4 T 0.0914 0.1683 0.0996 0.10 3 0.7894 3/4 T 0.0215 0.0371 0.0245 0.0247 0.04f6 1/4 T"" 0.1030 0.1891 0.1179 0.1201 0.2f95 3/4 T*"* 0.0381 0.0646 0.0484 0.0493 0.0785 Fast Flux (E > 0.1 MeV) 10"n/cm'-sec Capsule - -

3.1435 3.4001 -

IR 0.3323 0.6304 0.4635 0.4851 0.84f3 1/4 T 0.2976 0.5380 0.4238 0.4418 0.72f5 3/4 7 0.1580 0.2625 0.2316 0.2397 0.3493 fr:n Displacement Rate 104'dpa/sec Capsule - -

1.5694 1.7268 -

IR 0.2435 0.4515 0.2758 0.2838 0.5245 1/4 T 0.1579 0.2846 0.1919 0.1973 0.34f8 3/4 T 0.0583 _ _ 0.0971 0.0788 0.0810 0.1223 Vilues in bolditalles are the maximum values for any azimuth See note in Table 6-3.1 Maximum of values at 19,19.5,20,20.5,21 and 21.5'szimuths from 0.5' increment azimuthal mesh Maximum of values at 44,44.5 and 45' azimuths from 0.5' increment azimuthal mesh 1/4 T and 3/4 T values for fast fluence (E>1 MeV) using the loss steep dpa falloff slope within the pressure vessel

Table 6-13.2

' Seabrook Station Unit 1 Best-Estimate Neutron Exposure Rate Parameters at the Surveillance Capsules and within the Reactor Vessel Wall Average through 5.572 Effective Full-Power Years for End-of-Cycle 5  !

using Capsule Y Calculated-to-Measured Ratio of 0.893 i

~!

Parameter Units Capsule O' 19-21.5* 29' 31.5* 44-45' or Radial at Core Capsules Capsules - at Core Location Flats V and Y U and X . Diagonal in Azimuth Azimuth '

Base Metal *

' Fast Flux (E > 1 MeV) 10"n/cm'-sec Capsule - - 0.6565 .0.7108 -

IR 0.1280 0.2259 0.1246 0.1253 0.2350 1/4T 0.0736 0.1274 0.0732 0.0733 0.f320 l

3/4 T 0.0173 0.0285 0.0181 0.0179 0.0291 1/4 T*"' O.0828 0.1427 0.0864 0.0867 0.f527 3/4 T*"* 0.0304 0.0491 0.0353 - 0.0355 0.0545 Fast Flux (E > 0.1 MeV) 10"n/cm'-sec Capsule - - - 2.2653 2.4856 -

IR 0.2670 0.4730 0.3380 0.3489 0.5827 1/4 T 0.2387 0.4056 0.3087 0.3170 0.4999 .

l 3/4 T 0.1261 0.1991 0.1687 0.1723 0.2434 Iron Displacement Rate 10"dpa/sec Capsule - -

1.1384 1.2389 -

lR 0.1963 0.3406 0.2025 0.2054 0.3657 1/4 T . 0.1f10 0.2152 0.1404 0.1422 0.2378 3/4 T 0.0465 0.0740 0.0573 0.0582 0.0850 ,

Values in bohi#alles are the maximum values for any azimuth See note in Table 6 3.1 -

" Maximum of values at 19,19.5,20,20.5,21 and 21.5* azimuths from 0.5' increment azimuthal mesh

"' Maximum of values at 44,44.5 and 45' azimuths from 0.5' increment azimuthal mesh 1/4 T and 3/4 T values for fast fluence (E>1 MeV) using the less steep dpa falloff slope within the pressure vessel

l Table 6-14.1 Seabrook Station Unit 1 i Best-Estimate Neutron Integrated Exposure Parameters at the Surveillance Capsules and within the Reactor Vessel Wall At 0.913 Effective Full-Power Years for End-of-Cycle,1 using Capsule U Calculated-to-Measured Ratio of OJJ96 Parameter Units Capsule O' 19-21.5' 29' 31.5* 44-45' or Radial "

at Core Capsules Capsules at Core Location Flats V and Y U and X Diagonal in Azimuth Azimuth "*

Base Metal

• Fast Fluence (E > 1 MeV) 10"n/cm' Capsule - -

0.2600 0.2842 -

IR 0.0458 0.0864 0.0488 0.0498 0.0970 1/4 T 0.0233 0.0485 0.0287 0.0292 0.0546 3/4T 0.0062 0.0107 0.0071 0.0071 0.0120 1/4 T*"* 0.0297 0.0545 0.0340 0.0346 0.0632 3/4 T"" 0.0110 0.0186 0.0139 0.0142 0.0226 Fast Fluence (E > 0.1 MeV) 10"n/cm8 Capsute - -

0.9057 1.0056 -

IR 0.0957 0.1816 0.1335 0.1398 0.2424 1/4 T 0.0858 0.1550 0.1221 0.1273 0.2079 3/4 T 0.0457 0.0756 0.0667 0.0691 0.f 006 fron Displacement 10dpa Capsule - -

0.4522 0.4975 -

IR 0.0702 0.1301 0.0795 0.0818 0.1511 1/4 T 0.0455 0.0820 0.0553 0.0568 0.0965 3/4 T 0.0168 0.0280 0.0227 0.0233 0.0352 V . lues in bolditalles are the maximum values for any azimuth See note in Table 6-3.1 Maximum of values at 19,19.5,20,20.5,21 and 21.5' azimuths from 0.5' increment azimuthal mesh Maximum of values at 44,44.5 and 45' azimuths from 0.5* Increment azimuthal mesh 1/4 T and 3/4 T values for fast fluence (E>1 MeV) using the less steep dpa falloff slope within the pressure vessel

Table 6-14.2 Seabrook Station Unit 1

,Best-Estimate Neutron integiated Exposure Parameters i at the Surveillance Capsules and within the Reactor Vessel Wall At 5.572 Effective Full-Power Years for End-of-Cycle 5 using Capsule Y Calculated-to-Measured Ratio of 0.893 I

)

_\

Parameter Units Capsule 0* 19-21.5* 29' 31.5* 44-45' or l Radial at Core Capsules Capsules at Core l Location Flats V and Y U and X Diagonal in Azimuth Azimuth Base l Metal

• l Fast Fluence (E > 1 MeV) 10'*n/cm' Capsule - - 1.1545 1.2499 -

lR 0.2252 0.3972 0.2191 0.2203 0.4f3f 1/4T 0.1294 0.2239 0.1287 0.1288 0.232f i

3/4 T 0.0305 0.0500 0.0317 0.0315 l 0.05f2 1/4 T*"* 0.1456 0.2510 0.1518 0.1525 ' O.2684 3/4 T*"* 0.0534 0.0863 0.0620 0.0624 0.0960 Fast Fluence (E > 0.1 MeV) 10n/cm' Capsule - - 3.9833 4.3706 -

IR 0.4696 0.8317 0.5943 0.6135 f.0246 -l 1/4T 0.4197 0.7132 0.5428 0.5575 0.8790 3/4 T 0.2218 0.3501 0.2966 0.3030 0.4279 Iron Displacement 10 8dpa ' Capsule - - 2.0018 2.1785 -

IR 0.3452 0.5990 0.3561 0.3611 0.6430 1/4T 0.2233 0.3785 0.2468 0.2500 0.4f78 3/4 T 0.0818 0.1301 0.1008 0.1023 0.f495 Values in bokt Rafics are the maximum values for any azimuth See note in Table 6-3.1 Maximum of values at 19,19.5,20,20.5,21 and 21.5* azimuths from 0.5* increment azimuthal mesh Maximum of values at 44,44.5 and 45' azimuths from 0.5* increment azimuthal mesh 1/4 T and 3/4 T values for fast fluence (E>1 MeV) using the less steep dpa falloff slope within ths pressure vessel

Table 6-15 Seabrook Station Unit 1 Best-Estimate Neutron integrated Exposure Parameters at the Surveillance Capsules and within the Reactor Vessel Wall i

Projected at 32 Effective Full-Power Years using Capsule Y Calculated-to-Measured Ratio of 0.893 i

Parameter Units Capsule O' 19-21.5' 29' 31.5* 44-45' .

or . -

Radial at Core Capsules Capsules at Core Location Flats V and Y U and X Diagonal in Azimuth Azimuth "*

Base usiar .[

Fast Fluence (E > 1 MeV) 10"n/cm' Capsule - - 6.630 7.178 -

IR 1.293 2.281 1.258 1.265 2.3 73 1/4 T 0.743 1.268 0.739 0.740 f.333 3/4 T 0.175 0.287 0.182 0.181 0.294 1/4 T"" 0.836 1.441 0.872 0.876 f.542 3/4 T " " - 0.307 0.496 0.356 0.359 0.552 'j Fast Fluence (E > 0.1 MeV) 10*n/cm' Capsule - - 22.876 25.100 -

IR 2.697 4.776 3.413 3.523 5.884 1/4T 2.411 4.096 3.117 3.202 5.048 3/4 T 1.274 2.011 1.703 1.740 2.458 l Iron Displacement 10*dpa Capsule - - 11.496 12.511 -

IR 1.983 3.440 2.045 2.074 3.893 1/4T 1.282 2.174 1.417 1.436 2.399 3/4 T 0.470 0.747 0.579 0.588 0.858 j Values in boM Alslics are the maximum values for any azimuth

- See note in Table 6-3.1 Maximum of values at 19,19.5,20,20.5,21 and 21.5* azimuths from 0.5' increment azimuthal mesh

{

• " Maxirnum of values at 44,44.5 and 45' azimuths from 0.5' increment azimuthal mesh i

1/4 T ad 3/4 T values for fast fluence (E>1 MeV) using the less steep dpa falloff slope within the pressure  ;

vessel i l.

1. m.r.. '

-l'i.p.i mi

TCble 6-16 Saabrook Station Unit 1 Best-Estimate Fast Neutron Fluence (E> 1MeV) Projections at the Surveillance Capsuies and within the Reactor Vessel Wall Projected for 18-Month Cycles to 32 Effective Full-Power Years using Capsule Y Calculated-to-Measured Ratio of 0.893 Fast Fluence (E>1 MeV) (10"n/cm')

Cycle EFPYs at Capsule Capsules Base Metal 1/4T 1/4 T with i EOC X V and Y IR Maximum Maximum dpa Slope J

5* 5.572 1.25 1.15 0.41 0.23 0.27 6 7.2 1.62 1.49 0.53 0.30 0.35 7 8.5 1.91 1.76 G63 0.35 0.41 8 9.8 2.20 ;t.03 0.73 0.41 0.47 9" 11.1 2.49 2.30 0.82 0.46 0.53 10 12.4 2.78 2.57 0.92 0.52 0.60

,4N1.

11 13.7 3.07 2.84 1.02 0.57 0.66 12"* 15.0 3.36 3.11 1.11 0.62 0.72 13 16.3 3.66 3.38 1.21 0.68 0.79 14 17.6 3.95 3.65 1.31 0.73 0.85 15 18.9 4.24 3.92 1.40 0.79 0.91 16"" 20.2 4.53 4.10 1.50 0.84 0.97 17 21.5 4.82 4.45 1.59 0.90 1.04 18 22.8 5.11 4.72 1.69 0.95 1.10 l

19 24.1 5.41 4.99 1.79 1.00 1.16 20 25.4 5.70 5.Cd 1.88 1.06 1.22 21 26.7 5.99 5.53 1.98 1.11 1.29 l 22 28.0 6.28 5.80 2.03 1.17 1.35 l

l 23 29.3 6.57 6.07 2.17 1.12 1.41 24 30.6 6.86 6.! ' 2.27 1.27 1.47 25 32 7.18 6.63 2.37 1.33 1.54 EOC 5 removal of Capsule Y- Capsule fast fluence approaches the maximum vessel 1/4 T fast fluence at 32 EFPYs EOC 9 removal of Capsule V- Capsule fast fluence approaches the maximum vessel base metal IR fast fluence at 32 EFPYs EOC 12 for Capsule Y data - Capsute fast fluence remains above the maximum vessel base metal IR fast fluence at 15 EFPYs EOC 16 remove' of Capsulo X - Capsule fast fluence approaches a factor of 2 times the maximum l vessel base metal IR fast fluence at 32 EFPYs

Figure 6-1 Plan and Elevation Views of the Irradiation Surveillance Test Capsules in the Seabrook Station Unit 1 Reactor Vessel 0*

REACTOR VESSEL CORE BARREL

' NEUTMON PAD (301.5 *) Z CAPSULE U (58.5 *)

! g r

< ~58.5

• l 58.5*  % -

61*

27@ 90*

1 p

(241 *) y (238.5 *) X .

W (121.5*)

I 180* l l \

PLAN VIEW h[

N VESSEL WALL l s CAPSULE s/ EMBW

/ CORE

\lilllllllll f(

. CORE MIDPLANE j 3 {

f. a s N

! NEUTRON PAD

/ '

CORE BARREL f

I ELEVATION VIEW

Figure 6 2 Plan View of the Dual Irradiation Surveillance Test Capsules U and V l

in the Seabrook Station Unit 1 Reactor Vessel

'l l

(TYPICAL)

Co - 61.0 0 58.50 Fe Cu f// Yf/  !

g \,\/

l

/

34 . . . - -

q.s ,

' s' N ] .( N\[;N

' \,'\ s\ NN' N ,' .

J'x' s'N NJ:', ;xN xNNNs L . -lx.

I ,J I, ;' , , ' NEUTRON PAD \ '- ' s\\ 'd

!\xxN:s.

- 's\

s,f - s \'s,' ,',' \,\,,'kh --

N N' q ,

x N,s %s.c s,'s,'N s.; s s N x'sys ;s . xxxs N x' c s'NN s Q .,s j.- , (j isN' s. N x s Q.Q,N ,' _

l 1

l i

Figure 6-3 l 1

Seabrook Station Unit 1 Assemblier for Pinwise Detailed Fission Source Modeling in DORT I 1

i l

i I

l

. Diagoral  !

Location D-2 17 C-2 18 at 225'

/

/ Capsuls Y ,

/ at 241' )

- - - , /

l D-3 29 C-3 30 B-3 31 DORT /

Model... 11 5 )

/ l D-4 42 C-4 43 B-4 44 / i i

17 12 6 / l l

q I

D-5 56 C-5 57 B5 58 A-5 59 3 i

13 7 1 l

D-6 71 C-6 72 B-6 73 A-6 74 {

l l

14 8 2 D-7 86 C-7 87 B7 88 A7 89 15 9 3 l

i D-8 101 C-8 102 B-8 103 A-6 104 i 16 10 4 ---

l 1 l

( l

)

i

)

o Figure 6-4 Seabrook Station Unit 1 Cabuhted and Measured Assembly Relative Powers - Cycle 1 Average Assemblies Prmiding Fission Soume tieutrons near Capsule Y f'

Diagonal l

i' Locatica D2 17 C-2 18 at 225' Ca;ct.tated... 0.663 /

. Messscred..... 0.667 l / Capsule Y ]

l CM..... ... . . . 0.994 / at 241'

/

t D-3 29 C-3 30 B-3 31

~

1.054 0.663 /

1.057 0.607 0.997 0.994 /

D-4 42 C-4 43 B-4 44 /

0.921 0.926 /

O.995 D-5 56 C-5 57 B-5 58 A5 59 1.062 0.647 1.061 0.646 1.001 1.002 D-6 71 C-6 72 B-6 73 A6 74 0.787 0.785 1.003 D-7 80 C-7 87 B-7 88 A-7 89 0.833 0.839 0.993 D-8 101 C-8 102 B-8 103 A-8 104 0.825 0.827 ---

0.998 l

Figure 6-5 Seabrook Station Unit 1 Calculated and Measured Assembly Relative Powers - Cycle 1-5 Average Assemblies Providing Fission Source Neutrons near Capsule Y Diagonal Location D-2 17 C-2 18 at 225' Calculated.... 0.443 /

Measured..... 0.445 / Capsule Y C/M .............. 0.997 / at 241'

/

D-3 29 C-3 30 B-3 31 1.036 0.445 /

1.028 0.442 1.008 1.003 /

D-4 42 C-4 43 B-4 44 /

0.688 0.689 /

0.998 D-5 56 C-5 57 B-5 58 A-5 59 1.105 0.455 1.096 0.451 1.008 1.007 D-6 71 C-6 72 B-6 73 A-6 74 0.650 0.645 1.008 D-7 86 C-7 87 B-7 88 A-7 89 )

0.684 )

0.687 0.995 D-8 101 C-8 102 B-8 103 A-8 104 l 0.661 0.661 ---

1.000 l

)

~

i i

I Figure 6-6 Seabrook Station Unit 1 Measured Assembly Relative Powers - Cycle 1-5 Average and Cycle 1 Average Assemblies Providing Fission Source Neutrons near Capsule Y Diagonal Location D-2 17 C-2 18 at 225*

C 1 5 ........... 0.445 /

C1................. 0.667 / Capsule Y ,

(C1 5)/(C1)... 0.666 / at 241' D-3 29 C-3 30 B-3 31 I 1.028 0.447 /

1.057 0.667 0.973 0.663 / j D-4 42 C-4 43 B-4 44 /

0.699 0.926 /

0.744 i

D-5 56 C-5 57 B-5 58 A5 59 1.096 0.451 1.061 0.646 1.033 0.699 ,

I D-6 71 ' C-6 72 B-6 73 A-6 74 0.645 0.785 0.822 D-7 86 C7 87 B-7 88 A-7 89 0.687 0.839 1 0.819 l D-8 101 C-8 102 B-8 103 A-8 104 0.661 0.827 ---

0.799 70

)

Figure 6-7 Equation for Nuclide Product Activity in Surveillance Capsule t

A=(NjM)fYR,,f C/PjP,,,)(1 -e *'He **'"

}=1 where:

A induced product activity (disintegrations /sec per gram of target isotope)

N, Avogardo's number (atoms / mole)

M atomic welght of target isotope (grams / mole)

/ weight fraction of target isotope in target material Y yield fraction of product atoms per reaction R,,, saturated product reaction rate for reference cycle at the reference power level (disintegrations /sec)

C, cycle factor for cycle I, relative to the reference cycle, to account for cycle-dependent changes in capsule activation level (F/P,,,) power fractica for time period J, relative to the reference power level A decay constant of product isotope (sec4)

In time of activation for time periodj (sec) to time of decay, from the end of time period j to the end of capsule irradiation (sec) 1

Figure 6-8 Equations for Saturated Nuclide Product Reaction Rates  !

R,=fEa(E)&fE)dE= opp lE>1Mev)

R, saturated product reaction rate for cycle I (disintegrations /sec) o(E) energy-dependent neutron reaction cross section for capsule activation

/E) energy-dependent, cycle-averaged neutron flux at the capsule location for cycle 1 5, spectrum-averaged reaction cross section providing flux greater  ;

than 1 Mev for cycle i

{

/E>7Mev) cycle-averaged flux greater than 1Mev at the capsule location for cycle i C,=RlRg C, cycle factor for cycle I to account for cycle-dependent changes in capsule activation level R, saturated product reaction rate for cycle I (disintegrations /sec) saturated product reaction rate for the reference cycle R,,,,

(disintegrations /sec) l l

l 4

l

)

7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The surveillance capsule removal schedule in Table 7-1 meets the requirements of ASTM E185-94s end is recommended for future capsules to be removed from the Seabrook Station Unit 1 reactor vessel. TFL, recommended removal schedule is based on 32 Effective Full-Power Years (EFPYs)  ;

cf lifetimrv operation. i l

i

)'

l l

l I

73-l-

I

Table 7-1 Seabrook Station Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule Surveillance Vessel Lead Removal Removal after Capsule  !

f Capsule Azimuthal Factor ^ Time 8 Operation of Fluence, ^

Location Cycle E > 1 MeV (degrees) (EFPYs) (10 n/cm )

j U 58.5 2.93 0.913 1 0.284 c k

Y 241 2.79 5.572 5 1.154 C V 61 2.79 11.1 D 9 2.30 0 f

8 E X 238.5 3.03 20.2 16 4.53 W 121.5 3.03 Standby - -

Z 301.5 3.03 Standby - -

Notes A. Updated for Capsules U and Y and projected for Capsules V, X, W and Z.

Projections assume a continuation of the historical fast fluence accumulation rate through 5.572 EFPYs (Cycles 1 through 5).

B. Cumulative Effective Full-Power Years (EFPYs) of operation from plant startup.

1 C. Plant-specific values from removed capsules based on dosimetry results.

I D. Estimated removal of Capsule V near 11.1 EFPYs at End-of-Cycle 9.

I Capsule fast fluence approaches the maximum vessel base metal IR fast fluence at 32 EFPYs.

I E. Estimated removal of Capsule X near 20.2 EFPYs at End-of-Cycle 16.

Capsule fast fluence approaches a factor of 2 times the maximum vessel base metal IR fast fluence at 32 EFPYs.

8.0 REFERENCES

1. WCAP-10110, Pubilo Service Company of New Hampshire Seabrook Station Unit No.1

\ Reactor Vessel Radiation Surveillance Program, L R. Singer, Westinghouse Electric Corporation, March 1983.

k

. BAW-2157, Test Results of Capsule U, Public Service Company of New Hampshire, New Hampshire Yankee DMslon Seabrook Station Unit No.1, Reactor Vessel Material Surveillance Program, A. L Lowe, et al., B&W Document No. 77-2157-00, B&W Nuclear

- Servloe Company, May 1992.

3. BAW-2316, Revision 1, Test Results of Capsule Y, North Atlantic Energy Service Corpuratica, Seabrook Station Unit No.1, Reactor Vessel Material Surveillance Program, M. J. DeVen, FTl Document No. 77-2316-01, Framatome Technologies, Inc., December 1997.

4. . YAEC-1853, Analysis of Seabrook Station Unit 1 Reactor Vessel Surveillance Capsule U, E. C. Biemiller, R. J. Cacciapouti, Yankee Atomic Electric Company, June,1992.

5. ASTM Designation E185-94, Standard Practice for Conducting Surveillance Tests for

^ Light-Water Cooled Nuclear Power Reactor Vessels, E706(IF), in ASTM Standards, Section 12, American Society fer Testing and Materials, West Conshohocken, PA,1997.

6. U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, Radiation Embrittlement of Reactor VesselMaterials, Revision 2. May 1988.

7.

ASTM Designation E853-87 (Reapproved 1995), Standard PrW7ce for Analysis and interpretation of Light-Water Reactor Surveillance Results, E706(IA), in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

8.

ASTM Designation E893-94, StandardPrac6ce IbrCharacteridng Neutron Exposuresin lion and Low Alloy Steels in Terms of Displacements per Atom (DPA), E706(ID), in ASTM Standards, Section 12, American Society of Testing and Materials, West Conshohocken, PA,1997.

9. RSIC Computer Code Collection CCC-543, TORT-DORT Two- and Three-Dimensional

. Discrete Ordinates Transport, Version 2.8.14, Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, June 1994.

10. RSIC Data Library Collection DCL-185, BUGLE-96, Coupled 47 Neutron,20 Gamma Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure VesseIDos/metryApplications, Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, March 1996.

i

11. ASTM Designation E482-89 (Reapproved 1996), Standard Guide for Application of Neutron TransportMethods forReactor VesselSurveillance, E706(IID), in ASTM Standards, Section i

12, American Society for Testing and Materials, West Conshohocken, PA,1997.

l

12. ASTM Designation E560-84 (Reapproved 1996), Standard Practice 'for Extrapolating I Reactor Vessel Surveillance Dosimetry Results, E706(IC), in ASTM Standants, Section 12,

- American Society for Testing and Materials, West Conshohocken, PA,1997.

8

13. ASTM Designation E706-87 (Reapproved 1994), Standard Master Matrix for Ught-Water ReactorPressure VesselSurveillance Standards, E706(0), in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

14. ASTM Designation E261-%, Standard Practice for Determining Neutron Fluence, Fluence l

\

Rate, and Spectra by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

]

l

15. ASTM Designation E262-86 (Reapproved 1991), Standard Test Method for Determ/n/ng Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

16. ASTM Designation E263-93, Standard Tect Method for Measuring Fast-Neutron Reaction Rates by Rad /oactivatlon of /ron, in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

)

l 17. ASTM Designation E264-92 (Reapproved 1996), Standard TestMethod forMeasudng Fast-Neutron Reaction Rates by Radioactivation of Nickel, in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

18. ASTM Designation E481-86 (Reapproved 1991), Standard Test Method for Measuring Neutmn Fluence Rate by Radioacdvation of Cobalt and Silver, in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

t

19. ASTM Designation E523-92 (Reapproved 1996), Standard TestMethod forMeasuring Fast-Neutron Reaction Rates by Radioactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997,

20. ASTM Designation E704-96, Standard Test Method for Measuring Reaction Rates by Radioact/vation of Uranium-238,~ln ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

21. ASTM Designation E705-96,' Standard Test Method for Measuring Reaction Rates by l Radloactivation of Neptunlum-237, in ASTM Standards, Section 12, American Society for . .t Testing and Materials, West Conshohocken, PA,1997.

1

22. ASTM Designation E1005-84 (Reapproved 1991), Standard TestMethod forApplication and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E706(IIIA), in ASTM Standards, Section 12, American Society for Testing and Materials, West Conshohocken, PA,1997.

l

23. WCAP+14Es95, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vesse/ Radiation Surveillance Program, Westinghouse Electric Corporation, July 1997.

i i

l l

• ll*

)

APPENDIX A Unirradiated Vessel Plate and Weld Data

i TABLE A-1 Chemical Analysis of the Lower Shell Plates used in the Core Region I of the Seabrook Station Unit 1 Reactor Pressure Vessel Chemical Composition!*l (weight %)

i Plate R1808-1 Plate R1808-2 PlateSI R1808-3 Element C 0.22 0.22 0.2 l Mn 1.39 1.36 1.45 P 0.005 0.007 0.007 S 0.01 0.012 0.01 Si 0.22 0.21 0.24 Ni 0.58/0.58M 0.57/0.58M 0.57/0.61M Mo 1 0.56 0.55 Cr 0.04 0.03 0.06 Cu 0.05/0.07M 0.05/0.07M 0.06/0.08M Al 0.017 0.021 0.028 Co 0.009 0.009 0.01 Icl Icl Icl g

W <.01 <.01 <.01 Ti <.01 <.01 <.01 Zr <.001 <.001 <.001 V 0.004 0.004 0.003 Sn 0.001 0.002 0.011 As 0.002 0.001 0.006 Cb <.01 <.01 <.01 N, 0.007 0.008 0.008 B <.001 <.001 <.001 M Chemical Analysis by Combustion Engineering, Inc.

M Surveillance Program test plate.

M Not detected.

M' Check analysis by Lukens Steel A-1

TABLE A-2 L

Chemical Analysis of the intermediate Shell Plates used in the Core Region of the Seabrook Station Unit 1 Reactor Pressure Vessel Chemical Composition!'l (weight %)

" ~

Element C 0.25 0.22 0.21 Mn 1.47 1.33 1.33 P 0.012 0.007 0.007 S 0.012 0.009 0.012 SI 0.22 0.21 0.23 N1 0.64/0.58'l 0.05/0.63'l I

0.65/0.61kl Mo 0.59 0.59 0.58 Cr 0.08 0.04 0.03 Cu 0.04/0.05kl 0.05/0.07kl t 0.07/0.08 c)

Al 0.024 0.018 0.027 Co 0.014 0.012 0.012 Pb W 0.02 0.01 0.02 Ti <.01 <.01 <.01 Zr 0.001 0.001 0.001 V 0.006 0.004 0.004 Sn 0.003 0.004 0.004 As 0.006 0.007 0.07 Cb <.01 <.01 <.01 N, 0.01 0.008 0.01 B <.001 <.001 <.001 M

Chemical Analysis by Combustion Engineering, Inc.

M Not detected.

kl Check analysis by Lukens Steel A-2

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TABLE A-4 Heat Treatment History of the Seabrook Station Unit 1 Reactor Pressure Vessel Core Region Shell Plates and Weld Seams Material Temperature ('F) Time (hr) Cooling Lower Shell Piales Austenitizing: 4 Water-quenched R180812-3 1600

• 25 Tempered: 4 Alr-cooled 1225 e 25 Stress Relief: 16 Fumace-cooled 1150

• 50 Intermediate Shell Plates R1806 Austenitizing: 4 Water-quenched 1-2-3 1600

• 25 Tempered: 4 Air-cooled 1225*25 Stress Relief: 16.5 Fumsco-cooled 1150

• 50 Lower Shell Plate Longitudinal Stress Relief: 16 Fumace-cooled Seam Welds 1150

• 50 Intermediate Shell Plate Stress Relief: 16.5 Fumace-cooled Longitudinal Seam Welds 1150

• 50 Intermediate to Lower Shell Girth Stress Relief: 12.50 Fumace-cooled Seam Weld 1150

• 50 Surveillance Program Test Material Surveillance Program Test Plate Austenitizing: 4 Water-quenched R1808-3 1600 25 Tempered: 4 Air-cooled 1225t25 Stress Relief:Id 16.25 Fumace-cooled 1150

• 50 Weldment Stress Relief:Id 17 Fumace-cooled 1150

• SO

[a] The stress relief heat treatment received by the surveillance test plate and weldment have been simulated.

A-4

TABLE A-5  !

Tor, RT.1, and Upper Shelf Energy for the Seabrook Station Unit 1 j Reactor Pressure Vessel Core Region Shell Plates and Weld Metal Upper Shelft'l j

l Ter'1 RT rM Energy Material ('C) (*F) ('C) ('F) (J) (ft Ib)

Lower Shell Plates:

R1808-1 -34 -30 4 401040 1e+08 787778 R1808-2 29 20 . R1808-3M -29 4 i

Intermediate Shell Plates:  ;

R1806-1 -34 -30 4 40010. 1e+08 8e+07 R1806-2 -34 -30 18 R1806-3 -40 -40 ,2 intermediate Shell and Lower Shell Longitudinal -51 -60 -51 -60 212 156 I Seams-Weld Metall*84 and the intermediate to Lower Shell Girth Seam-Weld Metall*I8 M - Data by Combustion Engineering, Inc.

M Surveillance Program test plate.

M Weld Metal Heat No. 4P6052 Flux Type 0091, and Lot No. 0145.

18 Combustion Engineering Surveillance Weld Test Plate "C" Certification Report.

1 A-5 1

. - - + __ - - - -

APPENDlX B B&W Capsule U Test Results Report BAW-2157

BAW-2157 i May 1992 I

l TEST RESULTS OF CAPSULE U PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE NEW HAMPSHIRE YANKEE DIVISION SEABROOK STATION UNIT NO. 1

-- Reactor Vessel Material Surveillance Program --

by A. L. Lowe, Jr., PE R. E. Napolitano W. R. Stagg B&W Document No. 77-2157-00 (See Section 6 for document signatures)

B&W NUCLEAR SERVICE COMPANY Engineering and Plant Services Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 GWM h"%r

SUMARY This report describes the results of the testing of the. specimens from the first capsule (Capsule U) of the Public Service Company of New Hampshire Seabrook .

Station Unit No. I reactor vessel surveillance program. The objective of the j program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materials by the testing and evaluation of tension and Charpy impact specimens. The program was' designed in i accordance with the requirements of 10CFR50, Appendix H, and ASTM Specification .

E185-79.

The results of the tension tests and the Charpy impact test results indicated that the materials exhibited normal behavior relative to the estimated neutron

. fluence exposure.

1 l

l l

- ii -

'SWMwWiki==v

CONTENTS Page

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 ,

2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

-3. POST-IRRADIATION TESTING . . . . . . . . . . . . . . . . . . . . . 3-1 3.1. Visual Examination and Inventory . . . . . . . . . . . . . . 3-1 3.2. Thermal Monitors ......... 4

. . . . . . . . . . . 3-1 3.3. Tension Test Results . . . . . . . . . . . . . . . . . . . . 3-1

- 3.4. . Charpy V-Notch Impact Test Results . . . . . . . . . . . . . 3-2

.4. DOSIMETER MEASUREMENTS . . . . . . . . . . . . . . . . . . . . . . .

4-1

'4.1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2. . Dosimeter Preparation . . - . . . . . . . . . . . . . . . . . . 4-1 4.3. Quantitative' Gamma Spectrometry . . . '.' . . . . . . . . . . . 4-2

5. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1

6. CERTIFICATION . . . . ... . . . . . . . . . . . . . . . . . . . . . 6-1 List of Tables Table 3-1.. Tensile Properties Irradiated Base Metal and Weld Metal from C ap s ul e U . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3-2. Charpy Impact' Results for Irradiated Base Metal, Longitudinal (LT)-

Orientation, from Capsule U . . . . . . . . . . . . . . . . . . . 3-4 3-3. Charpy' Impact Results for Irradiated Base Metal, Transverse (TL)

Orientation, from Capsule U . . . . . . . . . . . . . . . . . ' . . . 3-4 3-4. Charpy. Impact Results for Irradiated Heat-Affected Zone Metal, from Capsule U . . . . . . . . . . . . . . . . . . . . . . . . .

3-5 x 3-5. Charpy Impact Test Results for Irradiated Weld Metal, Transverse .

(TL) Orientation, from Capsule U . . . . . . . . . . . . .-. ... 3-5

=4-1. Copper Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1 .. . . . . . . . . . . ... . . . . . . . . . . . . . . 4-4 1

4-2. Iron Dosimetry Measurements from Capsule U Seabrook Station Unit No.'l . . . . . . .-. . .-. .-.. . .-. . .-. . . . .-. . . .

4-5

- iii -

kBWit&We%r

Tables (Cont'd)

Table Page

. 4-3. Cobalt Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4-4. Cadmium Shielded Cobalt Dosimetry Measu'rements from Capsule U ,

Seabrook Station Unit No. 1 . . ... . . . . . . . . . . . . . . . 4-7 i 4-5. Nickel Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1 ...........................4-8 4-6. Uranium-238 Dosimetry Measurements from Capsule U Seabrook i Station Unit No. 1 . . . . . . . . . . . . . . . . . . . . . . . 4-9 .

i 4-7. Neptunium-237 Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1 . . . . . . . . . . . . . . . . . . . . . . . 4-9 j List of Fiaurss ,

Figure 3-1. Photographs of Tested Tension Test Specimens and Corresponding I Fractured Surfaces - Base Metal, Longitudinal Orientation . . . . 3-6 3-2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal,. Transverse Orientation . . . . 3-7 3-3. Photographs'of Tested Tension Test Specimens and Corresponding -

Fractured Surfaces - Weld Metal, Transverse Orientation . . . . 3-8 3-4. Tension Test Stress-Strain Curve for Base Metal' Plate R1808-3, Specimen No. KL2, Tested at 70F . . . . . . . . . . . . . . . . 3-9 3-5. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Specimen No. KL3, Tested at 300F . . . . . . . . . . . . . . . . 3-9 3-6. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Specimen No. KL1, Tested at 550F . . . . . . . . . . . . . . . . 3-10 3-7. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Specimen No. KT2, Tested at 70F . . . . . . . . . . . . . . . . 3-10 3-8. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Specimen No. KT1, Tested at 300F . . . . . ; . . . . . . . . . . 3-11

.3-9. Tension Test Stress-Strain Curve for Base Metal Plate' R1808-3, Specimen No.- KT3, . Tested at 550F . . . . . . . . . . . . . . . . . 3-11 1 3-10. Tension Test Stress-Strain Curve for Weld Metal, Specimen No. KW3, Tested at 70F . . . . . . . . . . . . . . . . . . . . . . . . . 3-12 3-11. Tension Test Stress-Strain Curve for Weld Metal, Specimen No. KWl, 4 Tested at 300F . . . . . . . . . . . . . . . . . . . . . . . . . 3-12 3-12. Tension Test Stress-Strain Curve for Weld Metal, S Tested at 550F . . . . . . . . . . . . . . . . . .pecimen No. KW2,

. . . . . . . 3-13 3-13. Charpy Impact Data for Irradiated Plate Material, R1808-3,

~ Longitudinal Orientation . . . . . . . . . . . . . . . . . . . . 3-14 3-14. Charpy Impact Data for Irradiated Plate Material, R1808-3, Transverse Orientation . . . . . . . . . . . . . . . . . . . . . 3-15 3-15. Charpy Impact Data for Irradiated Plate Material R1808-3 Heat-Affected Zone . . . . . . . . . . . . . . .,. . . . , . . . . 3-16 iv -

S W # AiWi! M ..

Fioures (Cont'd)

Figure Page l

l 3-16. Charpy Impact Data for Irradiated Weld Metal . . . . . . . . . . 3-17 3-17. Photographs of Charpy Impact Specimen Fracture Surfaces -

j Plate Material Longitudinal Orientation . . . . . . . . . . . . 3-18 i

3-18. Photographs of Charpy Impact Specimen Fracture Surfaces -

Plate Material Transverse Orientation . . . . . . . . . . . . . 3-19 3-19. Photographs of Charpy Impact Specimen Fracture Surfaces -

Plate Material, Heat-Affected Zone . . . . . . . . . . . . . . . 3-20 3-20. Photographs of Charpy Im Weld Metal . . . . . . . pact. Specimen Fracture Surfaces -

. . . . . . . . . . . . . . . . . . 3-21

-v-TAW E WSMS v

1. INTRODUCTION This report describes the results of the testing of the' specimens from the first l capsule (Capsule U) of the Public. Service Company of New Hampshire Seabrook -

Station Unit No. I reactor vessel material surveillance program (RVSP). The i l

capsula was removed and evaluated after being irradiated in the reactor as part of the Reactor Vessel M4terta10 Surve111arce Program. The objective of the l I

program is to monitor the effects of neutron irradiati:n on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillance program for Seabrook Station Unit I was designed and furnished by Westinghouse Electric Corporation (W). The program was planned to monitor the effects of neutron irradiation on the reactor vessel materials for tl. 40-year design. life of the reactor pressure vessel. The surveillance program for Seabrook Unit I was designed in accordance with E185-79' and thus is in I 2 8 compliance with 10CFR50, Appendixes G and 11.

1-1 m ar B W !! # L *co'n'e*

+

l 2. BACKGROUND The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water-cooled ~

reactors. The beltline ' egion of the reactor vessel is the most critical region Gf the vessel because it is exposed to neutron irradiation. The general effects of fast neutron irradiation ott the mechanical properties of such low-alloy ferritic steels as SA533, Grade B, used in the fabrication of the Seabrook Station Unit I reactor vessel, are well characterized and documented-in the literature. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility ifter irradiation. The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the Charpy upper-shelf energy value.

Append!x G to 10CFR50, " Fracture Toughness Requirements," specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of

. water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations on operation of.the RCPB. The toughness and Operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 13, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those und0r construction or in operation on the effective date.

2-1 BWllMWLv

Appendix H to 10CFR50, " Reactor Vessel Materials Surveillance Program Requirements,"8 defines the material surveillance program required to monitor changes in the fracture toughness properties of territic materials in the reactor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the therwal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout .

its service life. -

A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code,Section III, l

" Nuclear Power Plant Components."' This method utilizes fracture mechanics I concepts and the reference nil-ductility temperature, RTwor, which is defined as the greater of the drop weight nil-ductility transition temperature or the. i temperature that is 60F below that at which the material exhibits 50 ft ~1bs and 35 mils lateral expansion. The RTwor of a given material is used to index that  ;

material to a reference stress intensity factor curve (K curve), which appears in Appendix G of ASME Section III. The Km curve is a lower bound of dynamic, static, and cracir arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Km curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

The RTwor and, in turn, the operating limits of a nuclear power plant, can be adjusted to account for the effects of radiation on the properties of the reactor vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which a surveillance capsule containing prepared specimens ' of the reactor vessel materials is periodically removed from the l operating nuclear reactor and the specimens are tested. The increase in the Charpy V-notch 30 ft-lb temperature is added to the original RTuor to adjust it for radiation embrittlement. This adjusted RTwor is used to index the material 2-2 SWif4M5=r

i i

to the K curve which, in turn, is used to set operating limits for the nuclear l power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.  ;

i A

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2-3 N9IINEMM....

i

f

3. POST-IRRADIATION TESTING J

3.1. Visual Examination and Inventory All specimens were visually examined and no signs of abnormalities were found.

The contents of the capsule were inventoried and found to be consistent with the j surveillance program report inventory. There was no evidence of rust or of the

. penetration of reactor coolant into the capsule. The compact fracture toughness specimens and three-point bend bar were stored for future disposition.

3.2. Thermal Monitors Surveillance Capsule U contained temperature monitor sets in each of three holder blocks. The holder blocks'each contained one thermal monitor. The monitors located at the tcp and httom of the capsule are designed to melt at 579F and the monitor located at the midooint of the capsule is designed to melt at 590F. The holder blocks were radiographer for evaluation. None of the three sets of thermal' monitors exhibited any signs of melting. From thesa data, it was concluded that the irradiated specimens had been exposed to a maximum temperature of less' than ' 579F during the reactor vessel operating perio This is not significantly greater than the nominal inlet temperature . of 558F, and is considered acceptable for inclusion of the data in the general pool of irradiated surveillance data. There appeared to be no significant signs of a temperature gradient along the capsule length.

l 3.3. Tension Test Results -

The results of the postirradiation tension tests are presented in Table 3-1.

Tests were performed on specimens at room temperature, 300, and 550F. They were i tested on a computer controlled 55,000-lb load capacity MTS servohydraulic test i

machine at a crosshead speed of 0.005 inch per minute to yield point and thereafter 0.040 inch per minute. A 4-pole extension device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were 3-1 IRUlHEWSW,

in accordance with the applicable requirements of ASTM A370-77.5 for each material type and/or condition, specimens were tested'at room temperature, 300, and 550F to correspond to the unirradiated material test temperatures. The tension-compression load cell used had a certified accuracy of better than +0.5%

of full scale (25,000 lb). Photographs of the tension test specimen fractured surfaces are presented in Figures 3-1 through 3-3.

• In general, the ultimate strength and yield strength of the material increased with a corresponding slight decrease in ductility as compared to the unirradiated -

values; both effects were the result of neutron radiation. The type of behavior observed and the degree to which the material properties changed is within the range of changes to be expected for the radiation environment to which the specimens were exposed.

The stress-strain curves from the irradiated tension tests are presented in Figures 3-4 through 3-12.

3.4. Charov Y-Notch Imoact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 3-2 through 3-5 and Figures 3-13 through 3-16. Photographs of the Charpy specimen fracture surfaces are presented in Figures 3-17 through 3-20. The Charpy V-notch impact tests were conducted in 8

accordance with the requirements of ASTM E23-88 on an Satec SI-lK impact tester certified to meet Watertown standards.

The data show that the materials exhibited a sensitivity to irradiation within the values to be expected from their chemical composition and the fluence to which they were exposed.

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Tab 1: 3-2. Charpy Impact Results for Irradiated Base Metal, Longitudinal (LT) Orientation. from Caosule U Test Impact lateral Shear.

Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch  %

KL13 -40 11.0 0.009 0 KL2 0 26.0 0.025 25 KL15 20 34.0 0.031 25 KL6 40 62.0 0.047 50 KLl4 40 52.5 0.042 30 KL1 70 102.5 0.071 85 KL4 70 60.0 0.049 50 KL7 90 88.0 0.068 80 KL12 100 117.5* 0.077 100 KL5 125 122.0* 0.085 100 KL9 125 120.0* 0.083 100 KL3 150 117.0* 0.084 100 KLil 225 112.5* 0.076 100 KLIO 325 112.0 0.085 100 KL8 550 121.5 0.078 100

• Values used to determine upper-shelf energy value per ASTM E185.7 Table 3-3e Charpy Impact Results for Irradiated Base Metal, Transverse (TL) Orientation. from Caosule U Test I'mpact Lateral Shear Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch  %

KT1 -40 12.0 0.009 0 KIS 0 15.0 0.014 20 KT3 20 19.0 0.016 20 KT4 40 32.5 -0.028 30 KT7 70 45.0 0.036 45 KTil 100 59.5 0.045 65 KT14 120 58.0 0.051 60 KT8 125 71.5*- 0.059 100 KT6 150 71.0* 0.059 100 KT2 175 73.5* 0.062 100 KT12 200 69.5* 0.061 100 1- KT15 200 67.0* 0.062 100 KT13 225 81.5* '0.065 100 i KT3 325 79.5 '

0.071 100 KT10 550- 70.0 O.062 100

• Values used to determine upper-shelf energy value per ASTM E185.7 I

3-4  !

BWMM1MLar w_ _ _ - _ - _ - - - - - --_ -- - - - _ - - - - - - - -- - - - - -. - -- - A

Table 3-4. Charpy Impact Results for Irradiated Heat-Affected Zone Metal. from Cantule U Test Impact Lateral Shear

Specimen Temperature Energy Expansion Fracture ID F ft-lbs Inch  %

KH1 -80 32.0 0.021 45 KH4 -40 89.0 0.056 60 KHIS -40 79.0 0.052 60 KH7 -20 62.5 0.042- 70 i KH8- 0 52.0 0.040 50 '

KH6 20 71.0 0.051 70

! KH14 40 121.5*- 0.074 100 i i

KH12 70 130.0* 0.082 100 I KH3 100 147.5* 0.085 100 KH9 150 142.5* 0.085 100 i KH5 175 111.0* 0.080 100 l KH10 200 138.0 0.079 100 KH13 225 118.5 0.078 100 KH11 325 161.5 0.080 100 KH2 550 132.0 0.082 100

,

• Values used to determine upper-shelf energy value per ASTM E185.7 Table 3-5. Charpy Impact Test Results for Irradiated Weld Metal, Transverse (TL) Orientation. from Caosule U Test Impact Lateral Shear Specimen Temperature Energy Expansion Fracture

ID F ft-lbs Inch  %

KW10 -80 7.0 0.004 10 KWil -50 35.0 0.026 30 '

KW2 -40 48.0 0.036 40 KW5 -20 30.5 0.025 30 KW9 -20 69.0 -

0.051 55 .

KW1 0 101.5* 0.073 100 KW6 0 93.5 0.065 80

_). KW7 40 123.0 0.083 90 i

KW13 70 --- 0.092 100 KW15 70 133.0*' O.090 100 KW12 100 138.0* 0.091 100 KW8 150 143.0* 0.089 100 KW14 225 142.5 '0.087 100 KW3 325 157.0 0.088 100 KW4 550 153.0 0.085 100

• Values used to determine upper-shelf energy value per ASTM E185.7 3-5 B WIfn MIM L a  :

Figure 3-1. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal. Longitudinal Orientation

, 'l'

'[

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I Figure 3-2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal. Transverse Orientation e

Is -

? .

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Specimen KT2 (70F) =

I l

-%p [V~ I

> Specimen KTl (300F)

! I pecimen KT3 (550F)

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pecimen KW3 (70F) pecimen KW1 (300F) Specimen KW2 (550F)

BWMsa?i % r

Figure 3-4. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3. Specimen No. KL2. Tested at 70F

! a Specimen: KL2 Test Temp.: 70 F( 21 C)

Strength Yield: 73645. - res.

- UTS: 94193.

SS. .

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{

i Strength Yleid: 67609.

- UTS: 86650.

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Figure 3-6. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3. Specimen No. KLI. Tested at 550F f 3 3,,

Specimen: KL1 Test Temp.: 550 F( 287 C) i Strength Yield: 67276. 7es.

UTS: 91064 .

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33 70 F( 21 C)

Strength Yield: 73284. 7se.

UTS: 94094. .

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Figure 3-8. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3. Specimen No. KT1. Tested at 300F Specimen: KT1 Test Temp.: 300 F( 148 . C)

Strength Yield: 67210.

- UTS: 85657. _ eso.

g .f / . see. n.

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Strength Yield: 66442. . 7ss.

- UTS: 91294.

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Figure 3-10. Tension Test Stress-Strain Curve for Weld Metal, Specimen No. KW3. Tested at 70F Specimen: .KW3 Test Temp.:

, 70 F( 21 C)

Strength Yield: 76034. 7as.

UTS: 90024. .

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Strength Yield: 70525.

- UTS: 83683. _ ees.

es. e.

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3. 33 . De .12 .18 . 24 . 38 Engineering Strain 3-12 BW!!nEWafe%=r

Figure 3-12. Tension Test Stress-Strain Curve for Weld Metal, Specimen No. KW2. Tested at S50F Specimen: KW2 Test Temp.: 550 F( 287 C)

Strength Yleid: 69366.

- (JTS: 87673. - ses.

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3-13 SW!!ntfaM! y

Figure 3-13. Charpy Impact Data for Irradiated Plate Material, R1808-3. Longitudinal Orientation im , - : : .  : ,

2

", 7s -

5 h 50 - * * -

j 25 - .

0.10 , , , , , .

5 - - -

~

e g0.03 , e e

-.f0.06 -

[0.04 - -

E 5 0.02 - -

E / , , , f , ,

g 0 ' ' ' ' '

- DATA S'U MARY -

200 -T,Not Determined -

Tcy (35 m.s) +26F 180 Tey (50 FT-u) +40F Tgy (30 FT-u) +3F g 160 q.USE(AVG)lIbfA~ibS

$ Rig7 Not Determined

, 140 - _

E E120 - " -

g  :  :

5

• 30 1

~

mygnat SA 533.GrB1(L) 20 -

FLutact To be determined -

HEAT No. R1808-3 0

-100 0 100 , 200 500 860 0 500 W Test Temperature, F 3-14 SW!!!afELv

Figure 3-14. Charpy Impact Data for Irradiated Plate Material, R1808-3. Transverse Orientation _

'

100 . , - : 7 = ' '

2

} 75 3

g E 50 I

j 2s -

g 0 -

0.10 , , , , , ,

l ,

.C.

I 0

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i - -

j0.0.

I j0.04 - - E

! s E a

5 0.02

} E '

O

}

2x . . , , , , m

- DATA SuffiARY -

200 -T,,,Not Determined Tey (35 m.e> +68F 1 60 -Tey (50 rr-u) +78F -

Tey (30 rT-u) +39F -

,g 160 Cy -USE (Avo) 72 ft-lbs

! S RT,,,Not Determined -

. 1@ -

5

$m - -

l

' a s -

g 100 -

4 U

5 E

.m - . - -

l

• g w so -

. g 40 -

MTrain. SAS33.GrB1(T) g 20 - p m cg To be determined - g HEAT No. R1808-3 0 400 600

-100 0 100 , 200 300 500 Test Temperature, F 3-15 BW!!nnfatLv

Figure 3-15. Charpy Impact Data for Irradiated Plate Material, R1808-3. Heat-Affected Zone 100 ,  : : ;  : : ; :  : ,

.:. ,5 -

2 15

~

$ 50 m

6 h 25 -

0 '

O.10 e i e i i i d

~

w 2 '

5_ 0.08 f0.06 -

j0.0a - *e -

E 5 0.02 -

0 ' ' ' ' ' '

220 . i e i i s

- DATA SumARY -

200 -T,Not Determined -

Tcy (35 mu) -47F 180 -

-Tn (50 n-La) -41F Tcy (30 n-La) -85F ,

g 160 ~C -

y -USE (Ava)131 f t-lbs RT,37 Not e

. lai0 -

Determined e -

8 .

• 5m S

-

• e -

e Q

g 100 - ~

5 *

- u -

1 60 -

e

~

partnutSA533.GrB1(HAZ) 20 -

FLutact To be determined -

HCAT No. R1808-3 1 f i i e s 0

-100 0 100 , 200 500 400 500 600 Test Temperature, F 3-16 B WsYsMa h r

I Fioure 3-16. Charoy Imoact Data for Irradiated Weld Metal l

100  ;  ;  ; , , ,

e 2; ,

g 75 B

j 6

25 -

' ' ' i i .

I o 0.10 ' 8 I e a i g0.08 w

h0.06

, E0.04 5

I -

g 0.02 a1 l

' ' i i , e

! o 220 . ,

- DATA SumARY -

200 -T,,, Not Determined Tcy (35 nts) -40F

-Tcy (50 rt-ts) -38F 180 TcyW FT-Ls) -56F -

, 160 cy.USE (avr.)129 ft-Ibs ,

0 S RT,gy Not Determined * -

. 140 -

, g

= -

8 120 -

R S ,

a 100 ,

w 5

., 80 -

~ ,

60

• MAtta:AL Weld Metal 20 - FLutmer To be determined -

HEAT No.

0.-100 0 100 , 200 300 400 WO 20 m

Test Temperature, F 3-17 anasswsuctzen sdWSERVICE COMPANY

_ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ - _ - - ~

Figure 3-17. Photographs of Charpy Impact Specimen Fracture Surfaces -

Plate Material Longitudinal Orientation h ",

j'o

\ . -5, -

l .

y -

te  %*.~ '[?

4  ?

T Vg #

. *:v~

24' pectmen KLt (70f) Specimen KL2 (DF) Specimen KL3 (150F) Specimen KL4 (10F) Specimen KL5 (125F) 9

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pec men (40f) Specimen KL7 (90f) Spectmen KL8 (550F) Specimen KL9 (125f) Spectmen KLIO (325f) r

+.

.,  : 4 a.%

4)f*?, .T Y&.

i

--[ h* ,,-

s. *: ' y "t pecicnen KLll (2257) Specimen KL12 (100f) Spectmen KL13 (-40f) pecimen t. tis (40f) Specimen KLib (20f) 3-18 I3WREEa%"

Figure 3-18. Photographs of Charpy Impact Specimen Fracture Surfaces -

Plate Material Transverse Orientation

+

d

' T h -

i C i g  ; $, '

3* w.

e .

% *t M

g e.

pecimen f) pectmen KI2 (lisf) pectmen K ( ) Specimen K14 (40F) 7 ).' '

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cimen KI9 (20f) 1 ,

p.: .. . r . .! m- -

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('

l 'hi]i Y -

t '. .1[e{ '; ' , [ 1, ' i,

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p .~ . p[. n .; . -

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A 5pecimen Alli (J006) Specimen till (2006) Specimen All3 (225f) Specimen Alle (120fJ Specimen F.ll5 (2006 )

3-19 BllJ!!a*v>%Vc%omv

Figure 3-19. Photographs of Charpy Impact Specimen Fracture Surfaces -

Plate Material. Heat-Affected Zone

. - ~ . . .

i

. e .

• h 5

,~ % *

-m.- t.

+

- c men 1 1 -') Specimen 22 (550F) Spectmen 23 (100f) Spectmen 24 (-40f) specimen KMS (175F) k '

9

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men P'.

Specimen 2 7 (-20f) men. pecimen ) pec MF)

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h%$ n 9

pactmen 1 ) Specimen AM 2 ( 06 ) pecimen Ant) (225F) pecimen KH14 (40f) pectmen EHib ( 40F) 3-20 GWuivEWikfeimr

Figure 3-20. Photographs of Charpy Impact Specimen Fracture Surfaces - Weld Metal

,1

,. s.

[ . br *g' l 'e' , ]s t,

, $r,

~ * , < Q,]$ . ,. s,

. .. p.

. n,, il AM (-

ed.*4c .-

pecimen ) pecimen KW2 ( 4 f) pecimen W ) pecimen KW4 (550f) Specimen KW5 ( 20f) t I .g, ;.. u-f m.

1 '.W,7. > 1 4 a

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a r ! ,-

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Pec imen 8.hli ( bOf ) Specimen Ahlt (1006) Specimen kW13 (70f) ,bpecimen Kvid (llbf ) Spec imen Ahlb (70F )

3-21 B Wi!M fasiar i

4. DOSIMETER MEASUREMENTS 4.1. Introduction One set of dosimetry from the reactor vessel surveillance capsule (RVSP) denoted by the letter "U" was delivered to the Nuclear Environmental Laboratories by the Failure Analysis and Facility Operations Group of same affiliation (NES).

The set consisted of seventeen dosimeters made up of shielded and unshielded Co/A1 wires, unshielded Cu, Ni, and Fe wires, and unshielded 2380 an.d as'Np fission powders.- Cadmium sleeves were used to shield the Co/A1 wires. Each dosimeter was contained'in one of three stainless steel holder blocks that were installed in various positions in the assembly.

The dosimeters were delivered in vials identified by labels consisting of the position of the holder block in the assembly, and the position of the dosimeter item in the holder (see part 4.2 for explanation).

4.2. Dosimeter Preparation Vials were prepared for the dosimeters by labeling them with identifications that indicated their positions in the holder blocks. For example, the first wire in the top block was labeled Sbl, 01-U TOP 1. When the nuclides to be analyzed were determined by gamma scanning, the identifications were appended accordingly. For example, Sbl, 01-U TOP 1 Cu. This identification code stands for Seabrook Unit 1, Cycle 1 Capsule U, Top holder block, first wire, Copper.

, The stainless steel fission powder capsules were clamped in a metal-working vise I

which was mounted by a suction cup in a hood. A flat mill-bastard file was used to file the capsules open.

The cadmium-covered wires had been crimped at the ends so that the wires had to be removed by " nibbling" through the shield with diagonal cutters and removing the wires.

4-1 13 Ill!! E W 15 M u ,v

The dosimeter wires were cleaned by washing in reagent grade acetone, blotting dry with a laboratory towel. Each dosimeter wire diameter was n;sasured with a certified micrometer caliper, and weighed on a ' certified analytical balance.

Each was then mounted in the center of a PetriSlide with double-sided tape.

Wires over 1/2" in length were bent in a Yf before mounting.

The exact oxide compositions of the uranium and neptunium dosimeters were

~

uncertain. It was not possible to correct for self-absorption of the powders, F !' was necessary to dissolve them and put them into geometries for which our .

pma spectrometer was calibrated. This was the 20 cc liquid scintillation vial VM d ry. The uranium dosimeters were dissolved in CH HNO3 and diluted to ca.

20 mL with the same acid in a pre-weighed 20 cc scintillation vial. The neptunium dosimeters were digested in 6H HC1/16H HF with addition of H,0 2 in increments until dissolved. These were also diluted up to ca. 20 mL in.a pre-weighed 20 cc scintillation vials.

4.3. Quantitative Gamma _ Spectrometry Each of the dosimeters, in the PetriSlide* (point source) or 20 cc vial geometry, was given a 300 second preliminary count on the 31% PGT gamma spectrometer. This provided information with which to judge the best distance ,

at which to count the dosimeter to get a minimum of 10,000 counts in the photopeak of interest while keeping the counter dead time below 15%. It also provided qualitative identification of the dosimeters. This identification was made from the presence of the gamma rays in the table below in the spectra.

Dosimeter _

Ans1vte Cobalt "Co 91332 kev from "Co, very high activity Iron "Mn 6 834 kev from "Fe Nickel "Co 9 811 kev from "Ni Copper "Co 91332 kev from "Co, very low activity compared to Co^ wires, wire has coppery color Titanium "Sc 9 889 kev from "Ti

'37 Cs 9 662 kev, ***Pa 91001 kev 2"U as>Np

'87 Cs 9 662 kev, assPa 9 312 kev 4-2 SWit&M1!!dibv ,

The spectra confirmed the identifies of the dosimeters.

The spectra were then measured quantitatively at the appropriate counting positions and for the appropriate count times determined from the preliminary counts.

h 4-3 13IllHEWL%%.~

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5. REFERENCES

1. ASTM Designation E185-79, " Standard Practice for Conducting Surveillance ,

Tests for Light-Water Cooled Nucle,e Power Reactor Vessels," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA,

2. Code of Federal Regulation, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements. .

3. Code of Federal Regulation, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix H Reactor Vessel Ma'terial Surveillance Program Requirements.

4. American Society of Mechanical Enginee~.s (ASME) B"ler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure (G-2000).

5. ASTM Designation A370-77, " Methods and Definitions for Mechanical Testing of Steel Products," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

6. ASTM Designation E23 AA, " Methods for Notched Bar Impact Testing of Metallic Materials," 'in ASTM Standards, American Society for Testing and Materials, i Philadelphia, PA.

7. ASTM Designation E185-XX (to be released), Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, in ASTM Standards, American Society for Testing and Materials, Philadelphia, l PA. I l

5-1 ItllINE M WA..~

1 I

i l l i

6. CERTIFICATION i The specimens were tested, and the data obtained from Public Sarvice Company of New Hampshire Seabrook Station Unit No. 1, reactor vessel surveillance Capsule -

U were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H.

J

/I LIME 7Jr. , P.Et ' Q[ fjiTll72

)lr.' L'.

g ,,

'Date t Project Technical Manager j

' This report has been reviewed for technical content and accuracy. j j

??O b&d GbNR M. J.'Devan (Material Analysis)

Date M&SA Unit Verification of independent review.

Zb M K. E. MooreMa Mger' Date i M&SA Unit  !

This report is approved for release.

4/ /99- l T. L. Baldwin Date Program Manager L

6-1 l l

SW#a*cW:ifw%=r l

~

1 l

l APPENDIX C Framatomo Capsule Y Test Results Report BAW-2316, Revision 1 1

9

g BAW-2316. Revision 1 December 1997 Test Results of Capsule Y North Atlantic Energy Service Corporation Seabrook Station Unit No.1

- Reactor Vessel Material Surveillance Program -

by M.J.DeVan FTl Document No. 77-2316-01 (Section 6 for document signatures.)

Prepared by l Framatome Technologies, Inc.

3315 Old Forest Road 3 P. O. Box 10935 Lynchburg, Virginia 24506-0935 G 6 et O&O# #

i ____________________________o

Acknowledgment This acknowledges the efforts of Kevin Hour of the McDermott Technology, Inc.

l (Lynchburg, VA). His expertise in specimen testing contributed greatly to the success of this project.

ii R.A.TN.9 f.$

Summary This report describes the results of the test specimens from the second capsule (Capsule Y) of the North Atlantic Energy Service Corporation Seabrook Station Unit No.

1 reactor vessel surveillance program. The objective of the program is to monitor the  !

effects of neutron irradiation on the mecha.. sal properties of the reactor vessel materials by the testing and evaluation of tension and Charpy impact specimens. The l program was designed in accordance with the requirements of 10 CFR 50, Appendix H, I and ASTM Standard E 185-79.

This revision corrects the specific activity ( Cl/gm Target) value for the dosimeter ,

1 identified as Seabrook, Sh U-238. 1 I

l I

iii R.A.TM.9Y.5

L Record of Revisions Date Revision No. Description l

November 1997 0 Originalissue December 1997 1 Summary - Revision statement added.

Section 1 - Revision statement added.

Table 4-3. - Corrected Specific Activity value for dosimeter Seabrook, Sh U-238 Section 6 - New signatures added.

iv f!MTNP.P.'i

I I

1 Table of Contents Pace 1.0 I n t rod u c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.0 B a c kg ro u n d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3.0 P o st-I rra diation Testin g . . . . . . . . . . . . . . . . .. . . . . .. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . 3- 1 3.1 Visual Examination and Inventory ......................................................... 3-1 l 3.2 The rm al M onit o rs . . . . .. . . . . . . . . . . . . . . . . . . . ... . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .

3.3 Tension Te st R esult s . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . .; . . . .

3.4 Charpy V-Notch Impact Test Results .................................................... 3 2 j 4.0 D osimete r Measu rements . ... ... . .... . . .... . ... ..... . ...... . . .. .. ... . . . .. . . .. .. . ........ ..... ........ .. . . 4- 1 1

4.1 I n t rod u ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4.2 D osimete r Preparation .. . ..... . . . . ...... ... .... .. .. .. .. .. . .. .. . .. . .... . . .. ........ .. ... ..... . .. . . 4- 1 4.3 Quantitative Gamma Spectrometry ....................................................... 4-2 4.4 Dosimeter Specific Activities.................................................................. 4-2 5.0 R e f e re n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l. . . . .

6.0 C e rti fic ati o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

l List of Tables Table 3 1. Tensile Properties For Irradiated Base Metal and Weld Metal from Seabrook Station Unit No.1 Surveillance Capsule Y....................................... 3-3 3-2. Charpy impact Results for Irradiated Base Metal Plate, R1808-3, Longitudinal (LT) Orientation .... .............. . ..... ..... ... ... . ... ..................... ... ...... .. .. ... 3-4 3-3. Charpy impact Results for irradiated Base Metal Plate, R1808-3, Tran sve rse (TL) Orientation . .. .. . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . .. . . . .. .. . . . . . . . . . .. . . .

. 3-4. Charpy impact Results for Irradiated Heat-Affected-Zone Metal...................... 3-6 3-5. Charpy impact Results for Irradiated Weld Metal (4P6052 / 0145).................. 3 7 V

/

s F. .R.A M.A.T.O.M. .E.

6

I Tables (Continued) l l

l Table Paae 4-1. Q u a n ti fyin g G a m m a R ays . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -4 4-2. Isotopic Fractions and Weight Fractions of Target Nuclides............................ 4-5 4-3. Specific Activities for Seabrook Station Unit No.1 Capsule Y Dosimetry........ 4-6 4-4. Copper Dosimetry Measurements from Capsule Y Seabrook Station UnitNo.1.........................................................................................................4-7 4-5. Iron Dosimetry Measurements from Capsule Y Seabrook Station UnitNo.1.........................................................................................................4-8 4-6. Nickel Dosimetry Measurements from Capsule Y Seabrook Station UnitNo.1.........................................................................................................4-9 4-7. Unshielded Co/Al Dosimetry Measurements from Capsule Y Seabrook S tatio n Un it N o. 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 0 4-8. Shielded Co/Al Dosimetry Measurements from Capsule Y Seabrook S t at io n U n it N o . 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 1 4-9. Uranium-238 Dosimetry Measurements from Capsule Y Seabrook S ta t i o n U n it N o. 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 2 4-10. Neptunium-237 Dosimetry Measurements from Capsule Y Seabrook S tatio n U n it N o . 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 3 List of Fiaures Fiaure 3-1. Th e rmal M onito r P h otog raph s ................... .... ....... .. ..... ............ ....... .......... ........ 3-8 3-2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surface - Base Metal Plate, R1808-3, Longitudinal Orientation....... 3-9 3-3. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surface - Base Metal Plate, R1808-3, Transverse Orientation...... 3-10 3-4. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surface - Weld Metal (4P6052 / 0145)........................................... 3-11 3-5. Tension Test Stress-Strain Curve for Base Metal Plate, R1808-3, Longitudinal Orientation, Specimen No. KL 14, Tested at 70 F..................... 3-12 3-6. Tension Test Stress-Strain Curve for Base Metal Plate, R1808-3, Longitudinal Orientation, Specimen No. KL 13, Tested at 300 F................... 3-13 3-7. Tension Test Stress-Strain Curve for Base Metal Plate, R1808-3, Longitudinal Orientation, Specimen No. KL 15, Tested at 550 F................... 3-14 3-8. Tension Test Stress-Strain Curve for Base Metal Plate, R1808-3, Transverse Orientation, Specimen No. KT 13, Tested at 70 F...................... 3-15 3-9. Tension Test Stress-Strain Curve for Base Metal Plate, R1808-3, Transverse Orientation, Specimen No. KT 14, Tested at 300 F.................... 3-16 vi f!MTNRt%

Fiaures (Continued)

Fioure Pace 3-10. Tension Test Stress-Strain Curve for Base Metal Plate, R1808-3, I

Transverse Orientation, Specimen No. KT 15, Tested at 550 F.................... 3-17 3-11. Tension Test Stress-Strain Curve for Weld Metal (4P6052 / 0145),

Specimen No. KW 14 Tested at 70*F ........................................................... 3-18 l

3-12. Tension Test Stress-Strain Curve for Weld Metal (4P6052 / 0145),

Specimen No. KW 15, Tested at 300 F .................... ..................................... 3-19 3-13. Tension Test Stress-Strain Curve for Weld Metal (4P6052 / 0145),

Specimen No. KW 13, Tested at 550 F .................. ....................................... 3-20 g 3-14. Charpy impact Data for Irradiated Plate, R1808-3, Longitudinal (LT) E O ri e r it at i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3-15. Charpy impact Data for Irradiated Plate, R1808-T., Transverse (TL) g 3

O ri e n t at i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3-16. Charpy impact Data for Irradiated Heat-Affected-Zone Metal........................ 3-23 3-17. Charpy impact Data for Irradiated Wald Metal (4P6052 / 0145) .................... 3-24 E4 3-18. Photographs of Charpy impact Specsmen Fracture Surfaces of Irradiated g!

Base Metal Plate, R1803-3, Longitudinal (LT) Orientation................ ........... 3-25 3-19. Photographs of Charpy impact Specimen Fracture Surfaces of Irradiated g Base Metal Plate, R1808-3, Transverse (TL) Orientation .............................. 3-27 3 3-20. Photographs of Charpy impact Specimen Fracture Surfaces of irradiated H e at- Affe ct e d-Zo n e M e tal . . ... . . . . . . . . .. . . . . .. .. . ... . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . ... . . 3-29 3-21. Photographs of Charpy impact Specimen Fracture Surfaces of Irradiated Weld M etal (4 P6052 / 0145) .. . .. .. . ... .. .. . . . . .. . .. ... . .. . . . . . .. . .. . . ... .. .. . .. . . ... ..... . . . .... . . . .. 3-31 I

l l

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vii

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I

_._---------_-_______________-.___________-___.-____----.-_-____._..-_--______-_-._____.J~~\____s

M M

t 1

l 1.0 Introduction This report describes the specimen test results from the second reactor vessel surveillance capsule (Cap:ule Y) of the North Atlantic Energy Service Corporation Seabrook Station Unit No.1. The capsule was evaluated after being irradiated as part of the Seabrook Station Unit No.1 reactor vessel surveillance program (RVSP).

The objective of the progrrim is to monitor the effects of neutron irradiation on the mechanical properties of reactor vessel materials under actual operating conditions.

The surveillance program for Seabrook Station Unit No.1 was designed and fumished by Westinghouse Electric Corporation (W) as documented in WCAP-10110.N The program was planned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40-year design life of the reactor vessel. The surveillance program for Seabrook Station Unit No.1 was designed in accordance with American Society for Testing and Materials (ASTM) Standard E 185-79(2) and is in compliance with 10 CFR 50, Appendices GW and H.W This revision corrects the specific activity ( Ci/gm Target) value for the dosimeter 1

identified as Seabrook, Sh U-238.

1-1 ECH o&Oe 8

I I

I'

2.0 Background

I The ability of the reactor vessel to resist fracture is a primary factor in ensuring the saf6ty of the primary system in light water-cooled reactors. The reactor vessel beltline region is the most critical region of the vessel because it is exposed to the highest level of neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels used in the fabrication of reactor vessel are well characterized and documented. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yleid strength properties with a corresponding decrease in ductility after Irradiation. The most significant mechanical property change in reactor vessel steels is the increase in the ductile-to-brittle transition temperature accompanied by a reduction in the Charpy upper-shelf energy (CvUSE).

10 CFR 50, Appendix G, ' Fracture Toughness Requirements," ) specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of light water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations for operation of the RCPB. The fracture toughness and operational requirements are specified to provide adequate safety margins during any condition of nonnal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of 10 CFR 50, Appendix G, became effective on August 16,1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation en the effective date.

10 CFR 50, Appendix H, " Reactor VesselMaterial Surveillance Program Requirements,d') defines the material surveillance program required to monitor 2-1 mem I .

l

changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens contained in capsules that are periodically withdrawn from the reactor vessel. These data permit determination of the conditions under which the vessel can be operated with adequate safety margins against nonductile fracture throughout its service life.

A method for cuarding against nonductile fracture in reactor vessels is described in Appendix G to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section lil, " Nuclear Power Plant Components'0) and Section Xl, " Rules for Inservice Inspection of Nuclear Power Plant Components.'*)

This method uses fracture mechanics concepts and the reference nil-ductility temperature, RTwor, which is defined as the greater of the drop weight nil-ductility transition temperature (in accordance with ASTM Standard E 208-81M) or the tempera *ure that is 60 F below that at which the material exhibits 50 ft-Ibs impact energy and 35 mils lateral expansion. The RTuor of a given materialis used to index that material to a reference stress intensity factor curve (Kincurve), which appears in Appendix G of ASME B&PV Code Section 111 and Section XI. The Kincurve is a lower bound of dynamic and crack arrest fracture toughness data obtained from several heats of pressure vessel steel. When a given material is indexed to the Kin curve, allowable stress intensity factors can be obtained for the material as a function of temperature.

The operating limits can then be determined using these allowable stress intensity factors.

The RTworand, in tum, the operating limits of a nuclear power plant, are adjusted to account for the effects of irradiation on the fracture toughness of the reactor vessel materials. The irradiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which surveillance capsules containing prepared specimens of the reactor vessel materials are periodically removed from the operating nuclear reactor and the 2-2

%^MNRP.'i

specimens are tested. The increase in the Charpy V-notch 30 ft-Ib temperature is added to the original RTwor to adjust it for irradiation embrittlement. The adjusted RTwor is used to index the material to the Kin curve which, in tum, is used to set operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.

10 CFR 50, Appendix G, also requires a minimum initial CvUSE of 75 ft-Ibs for all beltline l region materials unless it is demonstrated that lower values of upper-shelf fracture energy will provide an adequate margin of safety against fracture equivalent to those required by ASME Section XI, Appendix G. No action is required for a material that does not meet the initial 75 ft-lbs requirement provided that the irradiation embrittlement does not cause the CvUSE to drop below 50 ft-Ibs. The regulations specify that if the CvUSE drops below 50 ft-lbs, it must be demonstrated, in a manner approved by the Office of Nuclear Reactor l I

Regulation, that the lower values will provide adequate margins of safety.

2-3

" f!MSNRM

)

i 3.0 Post-Irradiation Testing I

3.1 Visual Examination and inventory All specimens were visually examined and no signs of abnormalities were found. The contents of the capsule were inventoried and found to be consistent with the surveillance fabrication report inventory.N There was no evidence of rust or penetration of reactor coolant within the capsule. The compact fracture toughness acimens and three-point bend bar were stored for future disposition.

3.2 Thermal Monitors The two low-melting point (579*F and 590 F) eutectic alloys contained in Capsule Y were examined, and no indication of melting was observed (see Figure 3-1). Therefore, based on this examination, the capsule test specimens were exposed to a maximum temperature no greater than 579 F.

3.3 Tension Test Results The results of the post-irradiation tension tests performed on the base metal and weld metal test specimens are presented in Table 3-1. The tension tests were performed on a MTS servohydraulic computer-controlled universal test machine. All tension tests were run using stroke control with an initial actuator travel rate of 0.0075 inch per minute through the yield point. Following specimen yielding, an actuator speed of 0.03 inch per minute was used. The test conditions were in accordance with the applicable requirements of ASTM Standard E 8-96W and ASTM Standard E 21-92.M Photographs of the tension test specimen fractured surfaces are presented in Figures 3-2 and 3-4.

3-1 ECM e6Oe E

I The stress-strain curves from the irradiated tension test specimens are presented in l

Figures 3-5 through 3-13.

3.4 Charpy V-Notch impact Results The Charpy V-notch impact testing was performed in accordance with the applicable requirements of ASTM Standard E 23-960 ) on a Satec S1-1K Impact tester certified to l

meet the National Institute of Standards and Technology (NIST) standards.UU The test results of impact testing are shown in Tables 3-2 through 3-5 and Figures 3-14 through l

3-17. The curves were generated using a hyper'Ja:!c langent curve-fitting program to produce the best-fit curve through the data.

l Photographs of the Charpy specimen fractured surfaces are presented in Figures 3-18 through 3-21.

I I

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s e 7 0 3

5 5

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5 5 5 0

6 7 0 3

5 5

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t l

a t la P l:

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(

n la t

la t

la t

e e e e m M M M 4 5 3 ic e 4 3 5 e 3 4 5 1 1 1 e 1 1 1 1 1 1 d po

. s s l e

SN B a L L L a T T T W W W K K K B K K K W K K K l'

Table 3-2. Charpy impact Results for irradiated Base Metal Plate, R1808-3, Longitudinal (LT) Orientation l

KL 72 -40 8.5 2 0 i l

KL 73 0 18.5 13 10 KL 70 25 42.0 30 30 l

l KL 62 50 55.0 41 50 KL 74 68 73.5 55 70 KL 63 71 50.0 42 45 KL 71 100 92.0 65 85 KL 68 125 103.0 75 95 KL 65 150 112.5* 76 100 KL 64 175 107.0* 75 100 KL 67 200 111.5* 77 100 KL 66 200 110.5* 81 100 KL 75 250 103.5* 75 100 KL 69 300 104.5* 77 100 KL 61 350 102.0 76 100 )

l i

• Values used to determine0ugper-shelf energy value in accordance with ASTM Standard E 185-94 i

l

)

I l

3-4 i

e a se e6oe s

Table 3-3. Charpy impact Results for Irradiated Base Metal Plate, R1808 3, Transverse ,TL) Orientation

~ '-

} ;pfgN gasta73gg g 89 i Ihk3 1 Elixom a53

! 6FDA e f #;snik . . r5venic2 53 ca KT 69 -40 8.5 1 0 KT 62 0 12.0 6 5 KT 67 25 26.0 17 30 KT 63 50 34.5 24 40 KT 70 68 34.0 28 45 KT 75 71 44.0 36 40 KT 66 100 48.5 43 70 KT 65 125 60.5 48 90 KT 73 150 66.5* 58 100 KT 64 175 67.5 56 95 KT 74 200 70.0* 68 100 KT 68 200 68.5* 56 100 KT 72 250 60.0* 53 100 KT 71 300 64.0* 60 100 KT 61 350 67.0 57 100 ,

• Values used to determine up3per-shelf energy value in accordance with ASTM Standard E 185-940 3-5 3 s se oao 3e e

Table 3-4. Charpy impact Results for Irradiated Heat-Affected-Zone Metal KH 61 -80 15.5 7 15 KH 68 -40 32.5 22 50 KH 63 -25 46.5 30 60 KH 70 0 71.0 46 70 KH 74 25 116.0 68 90 KH 71 50 100.0 66 80 KH 67 68 95.5 59 90 KH 62 72 102.0 67 95 l KH 69 100 102.0* 68 100 KH 73 125 107.0* 69 100 KH 72 175 111.5* 76 100 KH 65 200 148.0* 79 100 KH 75 200 115.5* 75 100 KH 64 250 108.0* 70 100 KH 66 300 120.5 78 100

• Values used to determine ugper-shelf energy value in accordance with ASTM Standard E 185-94.( l l

l 36 f e eeYo s a e

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _-m

i Table 3-5. Charpy impact Results for irradiated l

Weld Metal (4P6052 / 0145)

, 1 81 KW 75 -80 4.5 0 0 KW 66 -40 27.5 17 30 KW 64 -25 44.0 29 50 KW 61 0 104.5 71 85 KW 63 25 102.0 67 75 KW 65 50 132.5 77 100 KW 74 68 132.5 83 90 KWG8 72 104.0 72 75 KW 67 100 119.0 81 90 KW 70 125 126.0 84 90 KW 62 175 149.0* 91 100 KW 69 200 140.0* 87 100 KW 73 200 147.5* 89 100 KW 71 250 135.5* 86 100 KW 72 300 146.5* 87 100

• Values used to determine up)per-shelf energy value in accordance with ASTM Standard E 185-940 3-7 UMYNP.P.'i

I Figure 3-1. Thermal Monitor Photographs l I

I I

TOP MID bot I 573 590 I 5 79-I S EABR 0 0 K erscock 3 witCox LYNCHBURG RESEARCH CENTER l

RA86 101397 I

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I 3-8 R A M AT O M E 58CHW D&OeoE$

I

l' Figure 3-2. Photographs of Tested Tension Test Specimens and Corresponding Fracture Surface - Base Metal Plate, R1808-3, Longitudinal Orientation

. .N' ' S ~:% i

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Specimen KL15 3-9 l

sua e C se oL0e t b______________._________________

I Figure 3-3. Photographs of Tested Tension Test Specimens and Corresponding Fracture Surface - Base Metal Plate, R1808-3, Transverse Orientation l

l ..

.e he. -

_  %{f??S';M.

c .

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a ayn.5.l i f -l,x:"sR:',Wh

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Specimen KT15 3-10 mem I

Figure 3-4. Photographs of Tested Tension Test Specimens and Corresponding Fracture Surface - Weld Metal (4P6052 / 0145) nn, .

4 4 .

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Figure 3-5. Tension Test Stress-Strain Curve for Base Metal Plate, R1808-3, Longitudinal Orientation, Specimen No. KL 14, Tested at 70 F i

31 Oct.. 1997 File: KL- I ti d Specimen: KL-14 Test Temp, 70 FI 21 C1 2 Strength _ dm T1eId: 75850.

UTS: 96509.

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Engineering Strain I

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3-12 I

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Figure 3-6. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Longitudinal Orientation, Specimen No. KL 13, Tested at 300'F 31 Oct 1997 File: KL-13 d Specimen ML-i3 2 - Test Temp.: 300 Fi 148 Cl S treng th Yleid:

UTS:

68668.

88769.

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1 l

l l

1 l

l l

3-13  ;

mus 8 C *e OLOe E

Figure 3-7. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Longitudinal Orientation, Specimen No. KL 15, Tested at 550*F i

1

\

31 Oct. 1997 File: KL-15 Test Temp.: 550 Ff 287 Cl f N '** :

$-15 i 5 - d 5

Tleid: 67506.

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(

l 3-14 r.m.e.em o

Figure 3-8. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Transverse Orientation Specimen No. KT 13, Tested at 70 F 31 Oct. 1997 File: Ki-13 d Speel.en: KT-13 Test Temp.: 70 F t 21 C)

$ S treng th Tleid: 74889.

8 UTS: 95737.

N-53 5d . c f

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.s  :

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n"

i o i i t I '

d 0.00 0.04 0.08 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain 3-15 fl e g ac oaoe s

Figure 3-9. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Transverse Orientation, Specimen No. KT 14, Tested at 300*F l

31 Oct 1997 File: Ki-14 SP*cimen: KT-ly Test Temp.: 300 Fi 148 Cl d

3 Strength d Tield: 68386.

UTS: 88102.

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i l Figure 3-10. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3,

! Transverse Orientation, Specimen No. KT 15, Tested at 550 F l

l 31 Oct. 1997 File KT-15 l d Specimens KT-15 Test Temp.: 550 FI 287 Cl i

Strength .

Yleid: 8

~

67266.

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0.00 0.04 0. 38 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain I

3-17 ff.

M F RfECe A.Me6AO TL O M. 8 &

O O E

D u_.__________..___.______ _ _ _ _ _ _ _ _ _ _ _

i Figure 3-11. Tension Test Stress-Strain Curve for. Weld Metal (4P6052 / 0145), l Specimen No. KW 14, Tested at 70'F 31 Oct.. 1997 file KW-14 Specimen: KW-14 Test Temp.: 70 Ff 21 Cl d

3 S treng th - d m

Yleid: 76635.

UTS: 91647.

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0.00 0.04 0. 38 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain i

i l

l l

1 l

3-18 f E 5 ee oLoe n

Figure 3-12. Tension Test Stress-Strain Curve for Weld Metal (4P6052 / 0145),

Specimen No. KW 15, Tested at 300'F

! 31 Oct. 1997 l File: KW-15 d Specimens KH-15 _ Test Temp.: 300 Fi 148 C1 3 S treng th -

d Tield: 70741.

UTS: 84276.

~

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d 0.00 0.04 18 5.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain 3-19 eCM O4OO E

f Figure 3-13. Tension Test Stress-Strain Curve for Weld Metal (4P6052 / 0145),

Specimen No. KW 13, Tested at 550 F 1

1 31 Oct. 1997 File: KW-13 Specimen: KW-13 Test Temp.: 550 F I 287 C) d

$ Strength - d Tleid: 70465.

UTS: 89980. l 8

d*o 5s !

I'.

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c  :

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o O 0.00 0.04 0 I)8 0.12 0.16 0.20 0.24 0.28 0.32 Engineering Strain i

i l

1 3-20

%SNP.P.'i

Figure 3-14. Charpy impact Data for Irradiated Base Metal Plate, R1808-3, Longitudinal (LT) Orientation 300 -= = = =

Y. 75 -

e

!!i h 50 -

u. E 3

g 25 -

0 _

l 100 0 100 200 300 400 500 Temperature, F 0.10

$ 0.08 - __E -

g -

o 1

3 0.06 -

8.

)

O

- 0 .04 - E j O.02 -

0.00 100 0 100 200 300 400 500 Temperature, F l

200 t 80 - Tcv ME):M Tev(50 fte):2 die i

160- Tev (30 fle): t12f Cv0SE: 108 ft Ibs 140-

= 220 -

a=

100 e = = E E. 80 E

- so

............ ..E.......................................

40 l 10 No.: R1808 3 0

100 0 100 200 300 400 500 Temperature, F 3-21 ff M F Rac AM=aaoeie*

ATO M E I

I i

t________u_______.________ _ _ _ _ _ _ _ _ - _

Figure 3-15. Charpy impact Data for Irradiated Base Metal Plate, R1808-3, Transverse (TL) Orientation 100 = -

=

Y. 75 -

!i U

m

! 50 3

I 25 - E e

0 - -

100 0 100 200 300 400 500 Temperature, F l

l l

0.10 l

I

$ 0.08 - l a

.9 E 0.06 - m g - - g -

0 0 .04 j 0.02 -

0.00

-100 0 100 200 300 400 500 Temperature. F 200 180 Tcv (35 MF 1Z$f l Tcv (50 R4):232f 160 - Tcv (30 he): t.fif Cv0SE: 06 ftes 140 c 120 6

l100 kJ Ea 80 E am E -

M E ' ~

40

..........- M.......................................

20 Material. SA-533 Gr. si (T1_t 10 No.: R1808 3 0

-100 0 100 200 300 400 500 Temperature, F 3 22 M

/l FR iec A M. .AT OME soeees 9

r Figure 3-16. Charpy impact Data for Irradiated Heat-Affected-Zone Metal f

I

,on __ __ _ _

1. . 75 - E e

3 U

e 50 -

u.

1

, 25 -

5 0

100 0 100 200 300 400 500 Temperature, F 0.10 g 0.08 - mE a o a aa =

y 0.06 - g E

5 .040 -

i ......

5 j 0.02 -

, m 0.00 100 0 100 200 300 400 S00 Temperature. F 200 180- cv (35 MLE)::20f_

TM50 hehM Tcv(30 ft@h:dI.E IM' OvuSE: its ft Ibn 140-B 51M - E = N g,00 g. -

5 80 -

I w 40 20 Matettal: SA-533 Gr. 81 (HAZ) 10 No.: R1808 3 0

100 0 100 200 000 430 500 Temperature. F 3-23

/-

F R AM ATO M E ecw=o6oeies 9

Figure 3-17. Charpy impact Data for Irradiated Weld Metal (4P6052 / 0145) iOO = _

= =

, M WE j

e 75 5 m i 7 1 l E E 50 - ,

1 a

h 25 -

0 l

100 0 100 200 300 400 500 l

Temperature. F 0.10

_ - E a E I 0.08 - g =m- I d E

.9 5 g 0.06 -

i i 5 0 .04 -

t 3 ...... ................................................

b j 0.02 -

i 0.00 100 0 100 200 300 400 500 Temperature. F 1 f

200 Tcv (35 MLE)::22f I 180' Tcv (50 he)W i Tev (30 h@)::d2f 160' CvuSE: 144 ft !bs EE 3 i , 140- 5 g I 4 m l

i 51M - E l l-E 80 -

60 40 20 Material: WakUdetal Heat No.:4P60D / 0145 I

l 100 0 100 200 300 400 500 1

Temperature. F 3-24 7

fF R A Mo a o .M vac ATO E

i Figure 3-18. Photographs of Charpy impact Specimen Fracture Surfaces of Irradiated Base Metal Plate, R1808-3, Longitudinal (LT) Orientation

, e , . ,. .

m m ..

Specimen No. KL73, Test Temperature O'F Specimen No. KL74, Test Temperature 68'F l,,* '

.! s ' . , t' ~ -l' ..

t . : 1; i .<

+3.s 1 .

l i. , .

. y. , ,

1'

..l

=

nj ', ,,;  :-  ? '.;

t M M t' l Specimen No. KL70, Test Temperature 25'F Specimen No. KL63, Test Temperature 71'F Specimen No. KL72, Test Temperature 40'F Specimen No. KL71. Test Temperature 100'F t

~

Specimen No. KIA2, Test Temperature 50'F Specimen No. KL68, Test Temperature 125*F 3-25 f! % % ^ 7 P. M 5

Figure 3-18 (cont.). Photographs of Charpy impact Specimen Fracture Surfaces of irradiated Base Metal Plate, R1808-3, Longitudinal (LT) Orientation l

I

w. _.

i, , .

.i.

Specimen No. KIAS, Test Temperature 150'F Specimen No. KL75, Test Temperature 250'F Specimen No. KI44, Test Temperature 175'F Specimen No. KI49, Test Temperature 300'F 4 _

n

_ _- ) -

. ~ . . _._

-G.

$.,'(( .

}

~

'h.: .- .,.k '

l . 'gt f _ l h .? Njef's ' _ ,_x },'

g ,_ ~. . -i E

Specimen No. KI46, Test Temperature 200'F Specimen No. KIA1, Test Temperature 350*F I

Specimen No. KL47, Test Temperature 200'F I

3-26 g

/e_f!MTNRM

~

Figure 3-19. Photographs of Charpy impact Specimen Fracture Surfaces of Irradiated Base Metal Plate, R1808-3, Transverse (TL) Orientation Specimen No. KT62, Test Temperature O'F Specimen No. KT70, Test Temperature 68'F

'. '.*; q * ., , - ..

c';

8 I . .

\Y

'3.

iI'. ')

.,I ':

g'I g

a

{ (i,/;' s

• a

.@, _ ... . . n.s . . .- .;

Specimen No. KT67, Test Temperature 25'F Specimen No. KT75, Test Temperature 71*F M. . . . . . .a ..~.. ~ ..

.c. .

" 'f*

_O.h.. !;n_ r-z..

Specimen No. KT69, Test Temperature 40'F Specimen No. KT66, Test Temperature 100'F 1

Specimen No. KT63, Test Temperature 50'F Specimen No. KT65, Test Temperature 125'F l

3-27 eCM o4oe t

Figure 3-19 (cont.). Photographs of Charpy impact Specimen Fracture Surfaces of irradiated Base Metal Plate, R1808-3, Transverse (TL) Orientation

w ... .---<~"'$, ,

,s?? ^

@?% .

?% ^ ,

.Es&

Specimen No. KT73, Test Temperature 150'F Specimen No. KT72, Test Temperature 250*F l I

l

.}

Specimen No. KT64, Test Temperature 175'F Specimen No. KT71, Test Temperature 300*F j

+

Specimen No. KT68, Test Temperature 200*F Specimen No. KT61, Test Temperature 350*F Specimen No. KT74, Test Temperature 200'F 3-28 5( H O6OO 85

L Figure 3-20. Photographs of Charpy Impact Specimen Fracture Surfaces of irradiated Heat-Affected-Zone Metal I

Specimen No. KIl71. Test Temperature 50'F Specimen No. KH70, Test Temperature O'F Specimen No. KII74, Test Temperature 25'F Specimen No. K1167, Test Temperature: 68'F Specimen No. KH63, Test Temperature 25'F Specimen No. KII62, Test Temperature 72'F 4

Specimen No. KII68, Test Temperature 40*F Specimen No. Klf61, Test Temperature 80'F i

3-29

<se.e m

Figure 3-20 (cont.). Photographs of Charpy impact Specimen Fracture Surfaces of irradiated Heat-Affected-Zone Metal Specimen No. KH69, Test Temperature 100'F Specimen No. Kli65, Test Temperature 200*F Specimen No. KH73, Test Temperature 125'F Specimen No. KH64, Test Temperature 250 F i Specimen No. KII72, Test Temperature 175'F Specimen No. KH66, Test Temperature 300'F

% a. -

j ..i

.. ~"9

i. n; .

'R i *! ' '. ~ .. ,,

.. ,. ~ -

_ 4 Specimen No. KII75, Test Temperature 200'F 3-30 EC H OLOe 8

L Figure 3-21. Photographs of Charpy impact Specimen Fracture Surfaces of Irradiated Weld Metal (4P6052 / 0145) l l

l Specimen No. KW61, Test Temperature O'F Specimen No. KW65, Test Temperature 50'F i

Specimen No. KW64, Test Temperature 25'F Specimen No. KW74, Test Temperature 68'F

~ "

3 i , ;'

. >; ,

• 1;. .

)I - '8 .

i kh(f ',

.l ,_

, g- -

Specimen No. KW63, Test Temperature 25'F Specimen No. KW68, Test Temperature 72*F

+ ;'

pa . . n, c w - .

, .~. .

m sa b-sm 1

'a' sj.

Specimen No. KW66, Test Temperature 40*F Specimen No. KW75, Test Temperature 80*F 3-31 E C se OLOO S

I Figure 3-21 (cont.). Photographs of Charpy impact Specimen Fracture Surfaces of irradiated Weld Metal (4P6052 / 0145)

• %._ s--

?  ;

,s..t t t .

'i l

~

'hki.$ I .k - ' *i f,k ,

[)  ;

W:t .: -

i x.~q Specimen No. KW67, Test Temperature 100*F Specimen No. KW69, Test Temperature 200*F l

Specimen No. KW70, Test Temperature 125'F Specimen No. KW71. Test Temperature 250*F I

1 l Specimen No. KW62, Test Temperature 175'F Specimen No. KW72, Test Temperature 300*F e

'- , ,% e

.;*AI.

n ii'i -

. a .

Specimen No. KW73, Test Temperature 200*F I

3-32 I

f! M Y N P. M I

l 4.0 Dosimeter Measurements 4.1 Introduction Three dosimeter sets wers located in blocks that were installed in top, middle, and bottom positions in the capsule assembly. Each dosimeter set consisted of dosimeters made up of shielded and unshielded Co/Al wires and unshielded Fe, Cu, and Ni wires. One 7

dosimeter set included shleided

• 8Uand Np fiscion powders; these were located in the middle of the capsule.

The dosimeters were stored in vials identified by labels consicting of the position of the dosimeter holder block within the capsule assembly and the local on from where the dosimeters were recovered.

4.2 Dosimeter Preparation Vials were prepared for the dosimeters by labeling them with identifications that indicated their types and positions within the holder blocks. For example, the one top block shielded Co/Al dosimeter was labeled Seabrook T (or TOP) Sh Co/Al. The analyte nuclides were verified during gamma scanning.

The fission powder capsules were clamped in a vise which was mounted on two lead bricks in a hood. A flat mill-bastard file was used to file the capsules open. The fission powder was carefully collected in vials with appropriate labels.

The dosimeters were cleaned by washing in reagent grado acetone, and blotting dry with a laboratory towel. Each dosimeter was measumd with a certified micrometer ca!iper and weighed on a certified analytical balance. Each was then mounted in the center of a PetriSlide* with double-sided tape.

4-1

1. f GE 4e O&Oe AS

The exact oxide compositbn of the uranium docimeters was uncertain.12 was not possible to correct for self-absorption of the powdem, therefore it was necessary to dissolve them and put them into a geometry for which the gamma spectrometer was calibrated. This was the 20cc liquid scintillation vial geometry. The uranium dosimeters were dissolved in 8N HNO3 acid and diluted to 20 miin the same acid in a pre-weighed 20cc scintillation vial. The neptunium powder was also prepared using similar procedures. l 4.3 Quantitative Gamma Spectrometry 4

Each of the dosimeters, in the PetriSlide* (point source), or 20cc vial geometry, was given a 300 second preliminary count on the 31% PGT gamma spectrometer. This provided information to best judge the distance at which to count the dosimeter to obtain a minimum of 10,000 counts in the photopeak of interest while keeping the counter dead time below 15%. It also provided qualitative identification of the dosimeters. This identification was made from the presence of the gamma rays in Table 4-1. The spectra  !

confirmed the identities of the docimeters.

The spectra were ti:en measured quantitatively at the appropriate counting positions and

)

for the appropriate count times detemened from the preliminary counts. I 4.4 Dosimeter Specific Activities The dosimeter specific activities are chown in Tabic 4-2, and the associated elemental l weight fractions of the dosimeters and the isotopic fractions of the target nuclides are l listed in Table 4-3.

l \

The weight fraction !!sted in Table 4-4 through 4-10 !s the product of the isotopic l fraction of the target and the weight fraction of the element in the dosimeter. In the case of the 23eU dosimeters, the total uranium content was measured by inductively coupled plasma (ICP) atomic emission spectroscopy and that value and the ' 7Cs activity were 7

used to calculate the specific activity. In the case of the Np dosimeter, the decay product 233 Pa was used to estimate ths neptunium activity. It was assumed that there 4-2

' M S N R P.'!

u ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

7 was no alternative source of protactinium daughter, and that Np was secular Np activity was equal to the 2 Pa activity. The 7

equilibrium with the Pa so that the l neptunium mass was equal to the protactinium activity divided by the specific activity of neptunium,705 Cl/ gram.

The dosimeter specific activities were calculated by dividing the corrected activity of the analyte nuclide by the target nuclide mass. The results were shown in Table 4-3, and the detailed calculations can be found in Tables 4-4 through 4-10.

4-3

/[-fM%^7P.Wi

Table 4.1. Quantifying Gamma Rays Dosimeter Analyte Iron "Mn @ 834 kev from "Fe 8

Co/Al Co @ 1332 kev from "Co Nickel ssCo @ 811 kev from "Ni l l

Copper 8

Co @ 1332 kev from 8'Cu, very low activity compared to Co wires, wire has coppery color j

'7 l sU Cs @ 662 kev 237 7 4 Np Cs @ 662 kev f

l i

l 1

4-4

%^YNRMS 1

" ^

e , , ,

Table 4-2. Isotopic Fractions and Weight Fractions of Target Nuclides f

Isotopic Weight Target Fraction of Fraction of

, Dosimeter Nuclide Target Target Element l

l Iron "Fe 0.0585 1.0 Cobalt 8'Co 1.0 0.0015 (unshielded) 58 Cobalt Co 1.0 ICP*

(shielded) 58 Nickel Ni 0.6777 1.0 5

Copper Cu 0.6917 0.999 237 Neptunium-237 Np 1.0 1.0 8

Uranium-238 U 1.0 ICP*

Inductively Coupled Plasma Atomic Emission Spectroscopy.

4-5 fMYNRM'i

Table 4-3. Cpecific Activities for Seabrook Station Unit No.1 Capsule Y Dosimetry Specific Dosimeter Shielded Target Analyte Activity identification (Yes/No) Nuclide Nuclide ( Ci/gm Target)

Seabrook, Top Sh C/Al Yes Co-59 Co-60 2.234E+04 Seabrook, Top Co/Al No Co-59 Co-60 4.784E+05 g Seabrook, Middle Sh Co/AI Yes Co-59 Co-60 2.283E+04 g Seabrook, Middle Co/Al No Co-59 Co-60 4.896E+05 Seabrook, Bottom Sh Co/Al Yes Co-59 Co-60 2.451 E+04 Seabrook, Bottom Co/Al No Co-59 Co-60 4.665E+05 Seabrook, Top Cu No Cu-63 Co-60 6.423 Seabrook, Middle Cu Seabrook, Bottom Cu No No Cu-63 Cu-63 Co-60 Co-60 5.502 5.689 l

Seabrook, Top Ni Seabrook, Middle NI No No Ni-58 Ni-58 Co-58 Co-58 1776 1591 l Sea'orook, Bottom Ni No Ni58 Co-58 1565 Saabrook, Top Fe No Fe-54 Mn-54 1166 Seabrook, Middle Fe No Fe-54 Mn-54 1045 Seabrook, Bottom Fo No Fe-54 Mn-54 1030 g Seab' rook, Sh U-238 Yes U-238 Cs 137 10.56 1 Seabrook, Sh Np-237 Yes Np-237 Cs-1:37 99.41 I

I I

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l 4 i

l 5.0 References

1. L. R. Singer, "Public Service Company of New Hampshire Seabrook Station Unit No.

1 Reactor Vessel Radiation Surveillance Program,"WCAP-10110. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, March 1993.

2. ASTM Standard E 185-79, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, Philadelphia, Pennsylvania.

3. Code of Federal Regulation, Title 10, Part 50, " Domestic Licensing of Production I

and Utilization Facilities," Appendix G. Fracture Touahness Requirements.

4. Code of Federal Regulation, Title 10, Part 50, " Domestic Licensing of Production and Utilization Facilities," Appendix H. Reactor Vessel Material Surveillance Proaram Requirements.

5. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel g Code, Section Ill, " Nuclear Power Plant Components," Apoendix G. Protection g Aaainst Nonductile Failure.1989 Edition.

6. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel ~

Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," Appendix G. Fracture Touahness Criteria for Protection Aaainst Failure.1989 Edition.

7. ASTM Standard E 208-81, " Method for Conducting Drop-Weight Test to Determine "

Nil-Ductility Transition Temperature of Ferritic Steels," American Society for Testing and Materials, Philadelphia, Pennsylvania.

8. ASTM Standard E 8-96, " Standard Methods of Tension Testing of Metallic Materials," American Society for Testing and Materials, Philadelphia, Pennsylvania.

9. ASTM Standard E 21-92, " Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials," American Society for Testing and Materials, Philadelphia, Pennsylvania, g 5-1 I

~

M l F. .R.A M. .A.T.O. M. .E.

10. ASTM Standard E 23-96, " Standard Methods for Notched BarImpact Testing of Metallic Materials," American Society for Testing and Materials, Philadelphia, Pennsylvania.

11. Standardized Specimens for Certification of Charpy impact Specimens from i I NationalInstitute of Standards and Technology, Office of Standard Reference Materials, Gaithersburg, Maryland.

12. ASTM Standard E 185-94,

• Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, Philadelphia, Pennsylvania.

\

I I

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6.0 Certification I

The specimens obtained from the North Atlantic Energy Service Corporation Seabrook s Station Unit No.1 surveillance capsule (Capsule Y) were tested and evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10 CFR 50, Appendices G and H. g

@ O IY- lll20 l99 M. J. D'eVan, E 1gineer IV 'Date g Materials & Str actural Analysis Unit l This report has been reviewed for technical content and accuracy.

0W uhin l'./B. Gross, Advisory Enginder Date I

g Materials & Structural Analysis Unit B Verification of independent review.

l- hSL?, // *20 k7

k. E. Moore, Manager Date Materials & Structural Analysis Unit This report is approved for release.

l 11 20 M D. L. Howell D' ate Program Manager 6-1 mm.sm

Revision 1 The revisions to this report were made as stated in accordance with the standard methods and procedures for the original report.

$0 )VE /Rl/flT7 M. J.' D(Vari,' Engineer IV Date Materials & Structural Analysis Unit Revisions have been reviewed for technical content and accuracy.

.W I LfEi. Gross, Advisory Engineer' lf'

?

D' ate Materials & Structural Analysis Unit 1 Verification of independent review.

/ 804L R2dh07 K. E.' Moore, Manager Date Materials & Structural Analysis Unit This report is approved for release.

Oh 12l19 $4 *]

D. L. Howell ' Date Program Manager 6-2 f

'MTNRM

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