Text

Part 1 of 2, H.B. Robinson Steam Electric Plant, Unit No. 2, Report of Analysis of Surveillance Capsule X for Reactor Vessel Radiation Surveillance Program
	ML021190313
Person / Time
Site:	Robinson
Issue date:	04/25/2002
From:	Fletcher B; Carolina Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RNP-RA/02-0033 WCAP-15805
	Download: ML021190313 (102)
	v • d • e

10 CFR 50, Appendix H CP&L A Progress Energy Company Serial: RNP-RA/02-0033 APR 2 5 2002 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 REPORT OF THE ANALYSIS OF SURVEILLANCE CAPSULE X FOR THE REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Ladies and Gentlemen:

In accordance with the provisions of the Code of Federal Regulations, Title 10, Part 50, Appendix H (10 CFR 50, Appendix H), Carolina Power and Light (CP&L) Company is submitting the enclosed report pertaining to the results of the analysis of Surveillance Capsule X for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. This capsule was removed from the HBRSEP, Unit No. 2, reactor vessel on April 29, 2001, during Refueling Outage 20.

This submittal satisfies the 10 CFR 50, Appendix H, requirement to report the results of capsule testing within one year of the date of capsule withdrawal. The enclosed report is titled: WCAP 15805, "Analysis of Capsule X from the Carolina Power and Light Company H. B. Robinson Unit 2 Reactor Vessel Radiation Surveillance Program."

If you have any questions concerning this matter, please contact Mr. C. T. Baucom.

Sincerely, B. L. Fletcher III Manager - Regulatory Affairs Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550

United States Nuclear Regulatory Commission Serial: RNP-RA/02-0033 Page 2 of 2 CAC/cac Enclosure C: Mr. L. A. Reyes, NRC, Region 1I Mr. R. Subbaratnam, NRC, NRR NRC Resident Inspector, HBRSEP

United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/02-0033 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 ANALYSIS OF CAPSULE X FROM THE CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM

Westinghouse Non-Proprietary Class 3 ANALYSIS OF CAPSULE X FROM THE CAROLINA POWER AND LIGHT COMPANY H.B.

ROBINSON UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Westinghouse Electric Company LLC

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15g05 Analysis of Capsule X from the Carolina Power & Light Company H.B. Robinson Unit 2 Reactor Vessel Radiation Surveillance Program T. J. Laubham E. P. Lippincott J. Conermann MARCH 2002 Prepared by the Westinghouse Electric Company for the Carolina Power & Light Company Approved: _ _ _ _ _ __ýý C. H. Boyd, Manager Equipment & Materials Technology Westinghouse Electric Company LLC Nuclear Services Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355

TABLE OF CONTENTS L IST O F TA B L E S ...............................................................................................................................................

LIS T O F FIGURE S ............................................................................................................................................. v P R E FA C E .................................................................................................................................................. v ii EXECUTIVE

SUMMARY

(OR) ABSTRACT ........................................................................................... viii 1 SUM MA RY O F RE SULT S ................................................................................................................ 1-1 2 IN T RO DUC TIO N ............................................................................................................................... 2 -1 3 BA C K GR OU ND ............................................................................................................................... 3 -1 4 DESCRIPTION OF PROGRAM ........................................................................................................ 4-1 5 TESTING OF SPECIMENS FROM CAPSULE X ........................................................................... 5-1 5 .1 OV ERV IE W ........................................................................................................................... 5 -1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS ................................................................ 5-3 5.3 TEN SILE TEST RE SU LTS ................................................................................................... 5-5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY ............................................................. 6-1 6 .1 IN T RO D UC TIO N .............................................................................................................. 6-1 6.2 DISCRETE ORDINATES ANALYSIS ................................................................................. 6-2 6.3 N EU TR ON D O SIM ETRY .................................................................................................... 6-5 6.4 PROJECTIONS OF REACTOR VESSEL EXPOSURE ................................................... 6-13 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE .................................................................. 7-1 8 RE FEREN C E S .................................................................................................................................... 8-1 APPENDIX A LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS APPENDIX B CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING HYPERBOLIC TAGENT CURVE-FITTING METHOD APPENDIX C CHARPY V-NOTCH SHIFT RESULTS FOR EACH CAPSULE HAND-FIT VS.

HYPERBOLIC TANGENT CURVE-FITTING METHOD (CVGRAPH, VERSION 4.1)

APPENDIX D H.B. ROBINSON UNIT 2 SURVEILLANCE PROGRAM CREDIBILITY ANALYSIS Analysis of H.B. Robinson Unit 2 Capsule X

LIST OF TABLES Table 4-1 Chemical Composition (wt %) and Heat Treatment of Material for the H.B. Robinson Unit 2 Reactor Vessel Surveillance Material ........................................... 4-4 Table 5-1 Charpy V-Notch Data for the H.B. Robinson Unit 2 Intermediate Shell Plate W10201-4 Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV)

(Longitudinal Orientation) ................................................................................................ 5-6 Table 5-2 Charpy V-notch Data for the H.B. Robinson Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV) ............................................ 5-7 Table 5-3 Charpy V-notch Data for the H.B. Robinson Unit 2 Heat Affected Zone (HAZ)

Material Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV) ............................. 5-8 Table 5-4 Charpy V-notch Data for the H.B. Robinson Unit 2 Correlation Monitor Material Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV) ............................................ 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Intermediate Shell Plate W10201-4 Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV)

(Longitudinal Orientation) ................................................................................................ 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 4.49 x 1019 n/cm2 (E > 1.0 MeV) ........................ 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV) ........ 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Correlation Monitor Material Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV) ............. 5-13 Table 5-9 Effect of Irradiation to 4.49 x 1019 n/cm 2 (E > 1.0 MeV) on the Notch Toughness Properties of the H.B. Robinson Unit 2 Reactor Vessel Surveillance Materials ............. 5-14 Table 5-10 Comparison of the H.B. Robinson Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions ............................................................................................ 5-15 Table 5-11 Tensile Specimens From Intermediate Shell Course Plate W10201-4 and Weld M aterial ............................................................................................................................. 5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance C ap sule C enter .................................................................................................................. 6-15 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface ........................................... 6-19 Table 6-3 Relative Radial Distribution of Neutron Fluence (E > 1.0 MeV) Within the Reactor V essel Wall .......................................................................................................... 6-2 3 Analysis of H.B. Robinson Unit 2 Capsule X

iii LIST OF TABLES (CONTINUED)

Table 6-4 Relative Radial Distribution of Iron Atom Displacement (dpa) Within the Reactor V essel W all .......................................................................................................... 6 -2 3 Table 6-5 Nuclear Parameters Used in the Evaluation of Neutron Sensors ..................................... 6-24 Table 6-6 Monthly Thermal Generation During The First 20 Fuel Cycles of the H.B. Robinson Unit 2 Reactor (Reactor Power of 2300 MWt) ................................................................ 6-25 Table 6-7 Calculated4(E > 1.0 MeV) and Cj Factors at the Surveillance Capsule Center C ore M idplane Elevation .................................................................................................. 6-29 Table 6-8 Measured Sensor Activities And Reaction Rates ............................................................. 6-30

- Surveillance C apsule S ................................................................................. 6-30

- Surveillance C apsule V ................................................................................ 6-3 1

- Surveillance C apsule T ................................................................................. 6-32

- Surveillance C apsule X ................................................................................ 6-33 Table 6-9 Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance C apsule C enter ............................................................................................ 6-34 Table 6-10 Comparison of Calculated and Best Estimate Exposure Rates at the Surveillance C apsule Center ............................................................................................ 6-36 Table 6-11 Comparison of Calculated/Measured (C/M) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions .................................................................................. 6-37 Table 6-12 Comparison of Calculated/Best Estimate (C/BE) Exposure Rate Ratios ....................... 6-37 Table 6-13 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from H .B . R obinson Unit 2 ....................................................................................................... 6-38 Table 6-14 Calculated Maximum Fast Neutron Exposure of the H. B. Robinson Unit 2 Reactor Pressure Vessel at the Clad/Base Metal Interface at Selected A zim uth al A ngles .............................................................................................................. 6-39 Table 6-15 Calculated Maximum Fast Neutron Exposure of the H. B. Robinson Unit 2 Upper Circumferential Vessel Weld at the Clad/Base Metal Interface at Selected Azim uth al A ngles .............................................................................................................. 6-40 Table 6-16 Calculated Maximum Fast Neutron Exposure of the H. B. Robinson Unit 2 Lower Circumferential Vessel Weld at the Clad/Base Metal Interface at Selected Azim uthal Angles .............................................................................................................. 6-4 1 Analysis of H.B. Robinson Unit 2 Capsule X

iv LIST OF TABLES (CONTINUED)

Table 6-17 Calculated Surveillance Capsule Lead Factors ................................................................. 6-42 Table 6-18 Calculated Fast Neutron Exposure of the H. B. Robinson Unit 2 Nozzle C omp onents ...................................................................................................................... 6-4 2 Table 7-1 H.B. Robinson Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule ........ 7-1 Analysis of H.B. Robinson Unit 2 Capsule X

v LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the H.B. Robinson Unit 2 Reactor Vessel ..... 4-2 Figure 4-2 Typical H.B. Robinson Unit 2 Surveillance Capsule Assembly ...................................... 4-3 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate Wi 0201-4 (Longitudinal Orientation) ......................... 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W 10201-4 (Longitudinal Orientation) ............ 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Longitudinal Orientation) .......................... 5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor V essel Surveillance Weld Metal ...................................................................................... 5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor V essel Surveillance Weld M etal ....................................................................................... 5-21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Surveillance Weld Metal ...................................................................................... 5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel H eat A ffected Zone M aterial ................................................................................. 5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Heat Affected Zone Material ................................................................................. 5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Heat Affected Zone M aterial ................................................................................. 5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor V essel Correlation M onitor M aterial ................................................................................ 5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Correlation Monitor M aterial .................................................................. 5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor V essel Correlation Monitor M aterial ................................................................................ 5-28 Analysis of H.B. Robinson Unit 2 Capsule X

vi LIST OF FIGURES (CONTINUED)

Figure 5-13 Charpy Impact Specimen Fracture Surfaces for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Longitudinal Orientation) .......................... 5-29 Figure 5-14 Charpy Impact Specimen Fracture Surfaces for H.B. Robinson Unit 2 Reactor V essel W eld M etal Specim ens .......................................................................................... 5-30 Figure 5-15 Charpy Inpact Specimen Fracture Surfaces for H.B. Robinson Unit 2 Reactor V essel Heat Affected Zone (HAZ) ................................................................................... 5-31 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for H.B. Robinson Unit 2 Reactor V essel Correlation M onitor M aterial ................................................................................ 5-32 Figure 5-17 Tensile Properties for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W 10201-4 (Transverse Orientation) ...................................................................... 5-33 Figure 5-18 Tensile Properties for H.B. Robinson Unit 2 Reactor Vessel Weld Metal ...................... 5-34 Figure 5-19 Fractured Tensile Specimens from H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Transverse Orientation) ......................................... 5-35 Figure 5-20 Fractured Tensile Specimens from H.B. Robinson Unit 2 Reactor Vessel Weld Metal.. 5-36 Figure 5-21 Engineering Stress-Strain Curves for Intermediate Shell Plate W 10201-4 Tensile Specimens C6 and C7 (Transverse Orientation) .............................................................. 5-37 Figure 5-22 Engineering Stress-Strain Curve for Weld Metal Tensile Specimens WI and W2 ......... 5-38 Analysis of H.B. Robinson Unit 2 Capsule X

vii PREFACE This report has been technically reviewed and verified by:

Reviewer:

Sections 1 through 5, 7, 8, Appendices A, B and C J.H. Ledger Section 6 S.L. Anderson 4.a Analysis of H.B. Robinson Unit 2 Capsule X

viii EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance capsule X from H.B.

Robinson Unit 2. Capsule X was removed at 20.39 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens was performed, along with a fluence evaluation based methodology and nuclear data including recently released neutron transport and dosimetry cross-section libraries derived from the ENDF/B-VI database. The calculated peak clad base/metal vessel fluence after 20.39 EFPY of plant operation was 2.76 x 1019 n/cm2 and the surveillance Capsule X calculated fluence was 4.49 x 1019 n/cm2 . A brief summary of the Charpy V-notch testing results can be found in Section 1 and the updated capsule removal schedule can be found in Section 7. A supplement to this report is a credibility evaluation, which can be found in Appendix D, that shows the H.B. Robinson Unit 2 surveillance weld data, while including all surveillance data for weld heat W5214, is credible. Of the three surveillance plates, only intermediate shell plate W10201-5 was found to be credible.

Analysis of H.B. Robinson Unit 2 Capsule X

1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance capsule X the fourth capsule to be removed from the H.B. Robinson Unit 2 reactor pressure vessel, led to the following conclusions: (General Note: Temperatures are reported to two significant digits only to match CVGraph output.)

The capsule received an average fast neutron calculated fluence (E > 1.0 MeV) of 4.49 x 101' n/cm2 after 20.39 effective full power years (EFPY) of plant operation.

Irradiation of the reactor vessel intermediate shell plate W10201-4 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation), to 4.49 x 1019 n/cm2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 104.73 0 F and a 50 ft-lb transition temperature increase of 98.68'F. This results in an irradiated 30 ft-lb transition temperature of 86.55 0 F and an irradiated 50 ft-lb transition temperature of 116.04-F for the longitudinally oriented specimens Irradiation of the weld metal Charpy specimens to 4.49 x 1019 n/cm 2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 265.93OF and a 50 ft-lb transition temperature increase of 251.74'F. This results in an irradiated 30 ft-lb transition temperature of 179.640 F and an irradiated 50 ft-lb transition temperature of 211.38 0 F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 4.49 x 1019 n/cm 2 (E >

1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 210.13°F and a 50 ft-lb transition temperature increase of 216.59'F. This results in an irradiated 30 ft-lb transition temperature of 100.47'F and an irradiated 50 ft-lb transition temperature of 150.54°F.

Irradiation of the correlation monitor material Charpy specimens to 4.49 x 1019 n/cm 2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 125.2 1'F which resulted in an irradiated 30 ft-lb transition temperature of 188.15'F. The tested specimens did not reach the 50 ft-lb transition temperature.

The average upper shelf energy of the intermediate shell plate W10201-4 (longitudinal orientation) resulted in an average energy decrease of I ft-lb after irradiation to 4.49 x 1019 n/cm2 (E> 1.0 MeV).

This results in an irradiated average upper shelf energy of 94 ft-lb for the longitudinally oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 33 ft-lb after irradiation to 4.49 x 1019 n/cm2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 80 ft-lb for the weld metal specimens.

The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 24 ft-lb after irradiation to 4.49 x 1019 n/cm 2 (E > 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 105 ft-lb for the weld HAZ metal.

Analysis of H.B. Robinson Unit 2 Capsule X

1-2 The average upper shelf energy of the correlation monitor material Charpy specimens resulted in no energy decrease after irradiation to 4.49 x 1019 n/cm 2 (E > 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 42 ft-lb for the correlation monitor material.

A comparison of the H.B. Robinson Unit 2 reactor vessel beltline material test results with the Regulatory Guide 1.99, Revision 2ý'), predictions led to the following conclusions:

- The measured 30 ft-lb shift in transition temperature values of the intermediate shell plate W10201-4 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured 30 ft-lb shift in transition temperature values of the weld metal contained in capsule X (longitudinal) is less than the Regulatory Guide 1.99, Revision 2, predictions.

- The measured percent decrease in upper shelf energy of the Capsule X surveillance material is less than the Regulatory Guide 1.99, Revision 2, predictions.

The peak calculated end-of-license (29 EFPY) neutron fluence (E> 1.0 MeV) at the peak vessel location approximately 4 inches above the core midplane for the H.B. Robinson Unit 2 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (ie. Equation # 3 in the guide; f(depthx) = fsface

e (-24x)) is as follows:

Calculated: Vessel inner radius = 3.67 x 1019 n/cm 2 Vessel 1/4 thickness = 2.10x 1019 n/cm 2 2

Vessel 3/4 thickness = 6.87 x 1018 n/cm

" The credibility evaluation of the H.B. Robinson Unit 2 surveillance program presented in Appendix D of this report indicates that the surveillance results for intermediate shell plate W 10201-5 and the weld metal are credible. Intermediate Shell Plates W10201-4 and W10201-6 were found not to comply with credibility criteria #3 of the Regulatory guide 1.99, Revision 2.

" The beltline is defined as portions of the vessel exposed to a fluence of 1 x 1017 n/cm 2 . Per Table 6-18 the lower portion of one inlet and one outlet nozzle see a fluence of slightly greater than 1 x 1017 n/cm 2 .

Thus, the nozzles are considered in the beltline and will be addressed when evaluating PTS and PT Curves. However, it should be noted here that based on the magnitude of the fluence, the nozzles or the nozzle weld will not become the limiting material in the H.B. Robinson Vessel.

Analysis of H.B. Robinson Unit 2 Capsule X

2-1 2 INTRODUCTION This report presents the results of the examination of Capsule X, the fourth capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the H.B.

Robinson Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Carolina Power and Light Company H.B. Robinson Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Company. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-7373, "Carolina Power and Light co. H.B. Robinson Unit No. 2 Reactor Vessel Radiation Surveillance Program"t'31. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-66, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors". Capsule X was removed from the reactor after 20.39 EFPY of exposure and shipped to the Westinghouse Science and Technology Center Hot Cell Facility, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the testing of and the post-irradiation data obtained from surveillance capsule X, removed from the H.B. Robinson Unit 2 reactor vessel and discusses the analysis of the data.

Analysis of H.B. Robinson Unit 2 Capsule X

3-1 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as A302 Grade B (base material of the Carolina Power and Light Company H.B. Robinson Unit 2 reactor pressure vessel beltline) are well documented in the literature.

Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.

A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Codet41 . The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-2081-]) or the temperature 60°F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (Ka curve) which appears in Appendix G to the ASME Codet41 . The KIa curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIh curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors. Note that Code Case N-641 now allows the use of the Kmc curve as an alternative to the Kia curve.

RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor surveillance program, such as the H.B.

Robinson Unit 2 reactor vessel radiation surveillance programt 31 , in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. This ART (RTNDT initial + M + ARTNDT) is used to index the material to the Kk curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

Analysis of H.B. Robinson Unit 2 Capsule X

4-1 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the H.B. Robinson Unit 2 reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

It should be noted that Capsules "X" and "U" were relocated to the "T" and "S"positions after cycle 8 Capsule X was removed after 20.39 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch impact and tensile specimens made from reactor vessel intermediate shell plate W 10201-4, submerged arc weld metal representative of the beltline region girth weld seam, and heat affected-zone (HAZ) metal. This capsule also contained Charpy V-notch specimens from the 6-inch thick ASTM correlation monitor material (A302 Grade B).

Test specimens obtained from intermediate shell plate W10201-4 (after the thermal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched ends of the plate. All test specimens were machined from the 1/4 thickness location of the plate. All plate Charpy V-notch specimens were oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation). Charpy V-notch impact specimens from the weld metal were oriented with the longitudinal axis of the specimen transverse to the weld with the notch oriented in the direction of the weld.

Tensile specimens from the plate materials were machined with the longitudinal axis of the specimen transverse to the major rolling direction of the plate. Tensile specimens from the weld metal were oriented with the longitudinal axis of the specimen transverse to the weld direction.

Capsule X contained dosimeter pure copper, nickel, and aluminum-cobalt (cadmium-shielded and unshielded).

The capsule contained thermal monitors made from four low-melting-point eutectic alloys and sealed in glass capsules. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the four eutectic alloys and their melting points are:

2.5% Ag, 97.5% Pb Melting Point 579°F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 5901F (310°C)

The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in capsule X is shown in Figure 4-2.

Analysis of H.B. Robinson Unit 2 Capsule X

4-2 T

1800 _ 30 - _ V 3030 30 w

x THERMAL SHIELD REACTOR VESSEL Figure 4-1. Arrangement of Surveillance Capsules in the H.B. Robinson Unit 2 Reactor Vessel Analysis of H.B. Robinson Unit 2 Capsule X

4-3 SPECIMEN NUMBERING CODE C- PLATE W10201-4 W - WELD METAL H - HEAT-AFFECTED-ZONE R - CORRELATION MONITOR MATERIAL SURVEILLANCE CAPSULE X 2 37 Np I

U238 DOSIMETER CHARPYS CHARPYS C46 IR4 TENSILE WOL CHARPYS C43 I R43 WOL CHARPYS BLOCK I411 2W7H7 C41E 73 CHARPYS WOL CHARPYS W6 H6 W5H5 LI WOL TENSILE CHARPYS CHARPYS t

Cu Co Co Co(Cd) - ! -- Ni Co(Cd)

I I I I aSIII LJ IL.i I LJ LJ ri II II 590°F 579 0F II 579°F II MONITOR Ii MONITOR I I a a MONITOR -i-i--

CENTER REGION OF VESSEL TO TOP OF VESSEL TO BOTTOM OF VESSEL Figure 4-2 Capsule X Diagram Showing the Location of Specimens, Thermal Monitors and Dosimeters Analysis of H.B. Robinson Unit 2 Capsule X

4-4 Table 4-1 Chemical Composition (wt %) and Heat Treatment of Material for the H.B. Robinson Unit 2 Reactor Vessel Surveillance Material(')

Chemical Composition Element Plate W10201-4 Plate W10201-5 Plate W10201-6 Weld Metal Correlation Monitor Material C 0.19 0.20 0.19 0.16 0.24 Mn 1.35 1.29 1.32 0.98 1.34 P 0007 0.010 0.010 0.021 0.011 S 0.019 0.021 0.015 0.014 0.023 Si 0.23 0.22 0.19 0.34 0.23 Mo 0.48 0.46 0.49 0.46 0.51 Cu 0.12 0.10 0.09 0.34 0.20 V --- --- --- 0.001 --

Ni --- --- --- 0.66 0.18 Cr --- --- --- 0.024 0.11 Co --- --- --- --- --

Heat Treatment Plate W10201-4, 1550°F to 1600°F, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Water Quench Plate W10201-5, & 1200 0F to 1250°F, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Air Cooled Plate W10201-6 1125-F to 1175-F, 15 1/2 hours, Furnace cooled to 600OF Weld Metal 1125°F to 1175°F, 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , Furnace cooled to 600°F Correlation Monitor 1650°F, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Water Quenched 1200°F - 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , Air Cooled Notes:

a) The data given in this column (originally) is from WCAP-7373 & WCAP-10304.

Analysis of H.B. Robinson Unit 2 Capsule X

5-1 5 TESTING OF SPECIMENS FROM CAPSULE X 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Science and Technology Department. Testing was performed in accordance with 10CFR50, Appendices G and HE'J, ASTM Specification El 85-82E'l, and Westinghouse Procedure RMF 8402, Revision 2 as modified by Westinghouse RMF Procedures 8102, Revision 1, and 8103, Revision 1.

Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master lists in WCAP-7373t31 . No discrepancies were found.

Examination of the two low-melting point 579 0F (304 0 C) and 5901F (3101C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed to was less than 579 0 F (3040 F).

The Charpy impact tests were performed per ASTM Specification E23-9881' and RMF Procedure 8103, Revision 1, on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 930-I instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix A), the load of general yielding (PGY), the time to general yielding (tGry), the maximum load (PM), and the time to maximum load (tM) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA). The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load (Em).

The yield stress (cy) was calculated from the three-point bend formula having the following expression:

c5,=(Pry *L) / [B * (W- a) 2 *C] (1) where: L distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W = height of the specimen, measured perpendicularly to the notch a = notch depth Analysis of H.B. Robinson Unit 2 Capsule X

5-2 The constant C is dependent on the notch flank angle (k), notch root radius (p) and the type of loading (i.e.,

pure bending or three-point bending). In three-point bending, for a Charpy specimen in which O= 450 and p =

0.0 10 inch, Equation 1 is valid with C = 1.21. Therefore, (for L = 4W),

ay= (PY *L) / [B * (W- a) 2 *1.21] = (3.33 *PGy

W) / [B * (W- a) 2 ] (2)

For the Charpy specimen, B = 0.394 inch, W = 0.394 inch and a = 0.079 inch. Equation 2 then reduces to:

cry=33.3 *P,, (3) where ay is in units of psi and PGY is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

The symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:

A = B * (W - a) = 0.1241 sq. in. (4)

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification E23-98 and A370-97aI9 3. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-99°101and E21-92t1 3,and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Extension measurements were made with a linear variable displacement transducer extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-93['2 ].

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures.

Chromel-Alumel thermocouples were positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower tensile machine griper and controller temperatures was developed over the range from room temperature to 550'F. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +20F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

Analysis of H.B. Robinson Unit 2 Capsule X

5-3 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in capsule X, which received a fluence of 4.49 x 1019 n/cm 2 (E > 1.0 MeV) in 20.39 EFPY of operation, are presented in Tables 5-1 through 5-8, and are compared with unirradiated results as shown in Figures 5-1 through 5-12.

The transition temperature increases and upper shelf energy decreases for the capsule X materials are summarized in Table 5-9. These results led to the following conclusions:

Irradiation of the reactor vessel intermediate shell plate W10201-4 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation), to 4.49 x 10'9 n/cm2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 104.731F and a 50 ft-lb transition temperature increase of 98.68'F. This results in an irradiated 30 ft-lb transition temperature of 86.55'F and an irradiated 50 ft-lb transition temperature of 116.04'F for the longitudinally oriented specimens Irradiation of the weld metal Charpy specimens to 4.49 x 10'9 n/cm2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 265.931F and a 50 ft-lb transition temperature increase of 251.741F. This results in an irradiated 30 ft-lb transition temperature of 179.64°F and an irradiated 50 ft-lb transition temperature of 211.38'F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 4.49 x 1019 n/cm 2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 210.13'F and a 50 ft-lb transition temperature increase of 216.59'F. This results in an irradiated 30 ft-lb transition temperature of 100.47'F and an irradiated 50 ft-lb transition temperature of 150.541F.

Irradiation of the correlation monitor material Charpy specimens to 4.49 x 1019 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 125.21 0F which resulted in an irradiated 30 ft-lb transition temperature of 188.15'F. The tested specimens did not reach the 50 ft-lb transition temperature.

The average upper shelf energy of the intermediate shell plate W10201-4 (longitudinal orientation) resulted in an average energy decrease of 1 ft-lb after irradiation to 4.49 x 1019 n/cm 2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 94 ft-lb for the longitudinally oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 33 ft-lb after irradiation to 4.49 x 1019 n/cm 2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 80 ft-lb for the weld metal specimens.

The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 24 ft-lb after irradiation to 4.49 x 1019 n/cm2 (E > 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 105 ft-lb for the weld HAZ metal.

The average upper shelf energy of the correlation monitor material Charpy specimens resulted in no energy decrease after irradiation to 4.49 x 1019 n/cm2 (E > 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 42 ft-lb for the correlation monitor material.

Analysis of H.B. Robinson Unit 2 Capsule X

5-4 A comparison of the H.B. Robinson Unit 2 reactor vessel beltline material test results with the Regulatory Guide 1.99, Revision 2[11, predictions led to the following conclusions:

- The measured 30 ft-lb shift in transition temperature values of the intermediate shell plate W10201-4 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured 30 ft-lb shift in transition temperature values of the weld metal contained in capsule X (longitudinal) is less than the Regulatory Guide 1.99, Revision 2, predictions.

- The measured percent decrease in upper shelf energy of the Capsule X surveillance material is less than the Regulatory Guide 1.99, Revision 2, predictions.

The fracture appearance of each irradiated Charpy specimen from the various surveillance capsule X materials is shown in Figures 5-13 through 5-16 and show an increasingly ductile or tougher appearance with increasing test temperature.

The load-time records for individual instrumented Charpy specimen tests are shown in Appendix A.

The Charpy V-notch data presented in this report is based on a re-plot of all capsule data using CVGRAPH, Version 4.1, which is a hyperbolic tangent curve-fitting program. Hence, Appendix C contains a comparison of the Charpy V-notch shift results for each surveillance material (hand-fitting versus hyperbolic tangent curve-fitting). Additionally, Appendix B presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data.

Analysis of H.B. Robinson Unit 2 Capsule X

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in capsule X irradiated to 4.49 x 1019 n/cm 2 (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results as shown in Figures 5-17 and 5-18.

The results of the tensile tests performed on the intermediate shell plate W10201-4 (transverse orientation) indicated that irradiation to 4.49 x 10 " n/cm 2 (E> 1.0 MeV) caused an approximate increase of 9 to 10 ksi in the 0.2 percent offset yield strength and approximately a 7 to 9 ksi increase in the ultimate tensile strength when compared to unirradiated dataE 11 (Figure 5-17).

The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 4.49 x 1019 n/cm 2 (E > 1.0 MeV) caused a 28 to 30 ksi increase in the 0.2 percent offset yield strength and a 22 to 28 ksi increase in the ultimate tensile strength when compared to unirradiated data (Figure 5-18).

The fractured tensile specimens for the intermediate shell plate W10201-4 material are shown in Figure 5-19, while the fractured tensile specimens for the surveillance weld metal are shown in Figures 5-20, respectively.

The engineering stress-strain curves for the tensile tests are shown in Figures 5-21 and 5-22.

Analysis of H.B. Robinson Unit 2 Capsule X

5-6 Table 5-1 Charpy V-notch Data for the H.B. Robinson Unit 2 Intermediate Shell Plate W10201-4 Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E> 1.0 MeV) (Longitudinal Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm %

C42 25 -4 11 15 5 0.13 10 C44 80 27 37 50 25 0.64 25 C48 100 38 22 30 18 0.46 20 C45 125 52 62 84 40 1.02 50 C46 150 66 73 99 45 1.14 60 C41 250 121 99 134 60 1.52 100 C43 325 163 92 125 57 1.45 100 C47 350 177 91 123 57 1.45 100 Analysis of H.B. Robinson Unit 2 Capsule X

5-7 Table 5-2 Charpy V-notch Data for the H.B. Robinson Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm %

W3 0 -18 4 5 0 0.00 0 W2 100 38 14 19 4 0.10 15 W6 175 79 28 38 16 0.41 35 W4 200 93 38 52 22 0.56 40 W8 250 121 74 100 49 1.24 100 W7 350 177 78 106 51 1.30 100 W5 375 191 85 115 56 1.42 100 Wi 425 218 82 111 54 1.37 100 Analysis of H.B. Robinson Unit 2 Capsule X

5-8 Table 5-3 Charpy V-notch Data for the H.B. Robinson Unit 2 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 4.49 x 10i9 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm %

H2 -50 -46 25 34 12 0.30 15 H7 0 -18 18 24 6 0.15 5 H6 25 -4 13 18 3 0.08 5 H5 100 38 24 33 9 0.23 25 H4 150 66 46 62 28 0.71 55 H3 250 121 90 122 63 1.60 100 H8 375 191 120 163 74 1.88 100 Hi 400 204 106 144 59 1.50 100 Analysis of H.B. Robinson Unit 2 Capsule X

5-9 Table 5-4 Charpy V-notch Data for the H.B. Robinson Unit 2 Correlation Monitor Material Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm %

R41 30 -1 2 3 0 0.00 5 R45 100 38 25 34 14 0.36 40 R48 150 66 18 24 10 0.25 30 R43 200 93 27 37 17 0.43 65 R46 225 107 38 52 28 0.71 95 R47 300 149 43 58 31 0.79 100 R42 325 163 44 60 29 0.74 100 R44 425 218 39 53 30 0.76 100 Analysis of H.B. Robinson Unit 2 Capsule X

5-10 Table 5-5 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Intermediate Shell Plate W10201-4 5-10 Irradiated to a Fluence of 4.49 x 1019 n/cm2 (E>1.0 MeV) (Longitudinal Orientation)

Normalized Energies (ft-lb/in 2 )

Charpy Yield Time to Time to Fast Test Energy Load Yield toy Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY (msec) Load PM Tm Load PF Load PA Stress Sy Stress No. (OF) (ft-lb) ED/A EM/A Er/A (Ib) (Ib) (msec) (Ib) (ib) (ksi) (ksi)

C42 25 11 89 44 45 3745 0.17 3788 0.18 3783 0 125 125 C44 80 37 298 202 97 3589 0.17 4268 0.48 4159 454.73 120 131 C48 100 22 177 32 146 3098 0.16 3102 0.16 3098 1022.55 103 103 C45 125 62 500 219 280 3429 0.17 4218 0.53 3822 1125 114 127 C46 150 73 588 205 384 3469 0.17 4164 0.50 3467 1559 116 127 C41 250 99 798 281 517 3227 0.17 4074 0.67 n/a n/a 107 122 C43 325 92 741 207 534 3225 0.17 4015 0.53 n/a n/a 107 121 C47 350 91 733 185 548 3214 0.17 3862 0.49 n/a n/a 107 118 Analysis of H.B. Robinson Unit 2 Capsule X

Table 5-6 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E>1.0 MeV)

Normalized Energies (ft-lb/in 2 )

Charpy Yield Time to Time to Fast Test Energy Load Yield tGy Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY (msec) Load PM Tm Load PF Load PA Stress Sy Stress No. (OF) (ft-lb) ED/A EM/A Ep/A (lb) (lb) (msec) (lb) (lb) (ksi) (ksi)

W3 0 4 32 17 15 2201 0.12 2214 0.13 2201 0 73 74 W2 100 14 113 70 43 4204 0.17 4615 0.22 4593 0 140 147 W6 175 28 226 71 155 3994 0.17 4380 0.23 4194 707 133 139 W4 200 38 306 213 93 3967 0.17 4582 0.47 4482 1090 132 142 W8 250 74 596 223 373 3898 0.17 4535 0.49 n/a n/a 130 140 W7 350 78 628 231 397 3808 0.17 4450 0.52 n/a n/a 127 138 W5 375 85 685 230 455 3873 0.17 4512 0.51 n/a n/a 129 140 Wi 425 82 661 211 450 3477 0.17 4201 0.51 n/a n/a 116 128 Analysis of H.B. Robinson Unit 2 Capsule X

5-12 Table 5-7 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Heat Affected Zone (HAZ) Material 5-12 Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E>1.0 MeV)

Normalized Energies (ft-lb/in2 )

Charpy Yield Time to Time to Fast Test Energy Load Yield try Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY (msec) Load PM Tm Load PF Load PA Stress Sy Stress No. (OF) (ft-lb) ED/A EM/A Er/A (ib) (lb) (msec) (Ob) (lb) (ksi) (ksi)

H2 -50 25 201 74 128 4351 0.17 4747 0.22 4539 0 145 151 H7 0 18 145 71 74 4172 0.17 4593 0.22 4403 0 139 146 H6 25 13 105 66 39 4326 0.17 4654 0.21 4652 0 144 150 H5 100 24 193 68 126 4068 0.17 4406 0.22 4267 171 135 141 H4 150 46 371 197 173 3860 0.17 4454 0.46 4384 2065 129 138 H3 250 90 725 306 420 3460 0.16 4353 0.67 n/a n/a 115 130 H8 375 120 967 294 673 3477 0.17 4266 0.67 n/a n/a 116 129 HI 400 106 854 301 554 3540 0.17 4246 0.68 n/a n/a 118 130 Analysis of H.B. Robinson Unit 2 Capsule X

Table 5-8 Instrumented Charpy Impact Test Results for the H.B. Robinson Unit 2 Correlation Monitor Material Irradiated to a Fluence of 4.49 x 1019 n/cm 2 (E>1.0 MeV)

Normalized Energies (ft-lb/in 2 )

Charpy Yield Time to Time to Fast Test Energy Load Yield toy Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY (msec) Load PM Tm Load PF Load PA Stress Sy Stress No. (OF) (ft-lb) ED/A EM/A Ep/A (Ib) (lb) (msec) (ib) (lb) (ksi) (ksi)

R41 30 2 16 8 8 1155 0.11 1165 0.1 1155 0 38 39 R45 100 25 201 71 131 3886 0.17 4248 0.23 4092 896 129 135 R48 150 18 145 62 83 3749 0.17 4061 0.22 4057 605 125 130 R43 200 27 218 68 149 3669 0.17 4028 0.23 3917 1858 122 128 R46 225 38 306 121 185 3728 0.17 4108 0.33 3965 2009 124 130 R47 300 43 346 142 204 3231 0.18 4048 0.4 n/a n/a 108 121 R42 325 44 355 153 202 3620 0.17 4156 0.39 n/a n/a 121 129 R44 425 39 314 68 246 3428 0.17 3744 0.24 n/a n/a 114 119 Analysis of H.B. Robinson Unit 2 Capsule X

5-14 Table 5-9 Effect of Irradiation to 4.49 x 1019 n/cm 2 (E>1.0 MeV) on the Notch Toughness Properties of the H.B. Robinson Unit 2 Reactor Vessel Surveillance Materials Average 30 (ft-lb)(a) Average 35 mil Lateralh) Average 50 ft-lb(a) Average Energy Absorption(a)

Material Transition Temperature (*F) Expansion Temperature (*F) Transition Temperature ( 0F) at Full Shear (ft-lb)

Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AE Intermediate -18.17 86.55 104.73 -3.32 120.96 124.28 17.35 116.04 98.68 95 94 -1 Shell Plate W10201-4 (Longitudinal)

Weld Metal -86.29 179.64 265.93 -60.64 219.24 279.89 -40.35 211.38 251.74 113 80 -33 HAZ Metal -109.66 100.47 210.13 -84.08 164.51 248.60 -66.04 150.54 216.59 129 105 -24 Correlation 62.94 188.15 125.21 --- --- --- --- --- 39 42 +3 Monitor Material

a. "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-1, 5-4, 5-7 and 5-10).

b. "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-2, 5-5, 5-8 and 5-11)

Analysis of H.B. Robinson Unit 2 Capsule X

5-15 Table 5-10 Comparison of the H.B. Robinson Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured 2

(x 1019 n/cm ) (OF) (a) (OF) (b) (%) (a) (%)(C)

Inter. Shell Plate S 0.479 45.39 32.51 18 10 W10201-4 X 4.49 78.86 104.73 30 1 (Longitudinal)

Surveillance V 0.530 179.17 209.32 39 38 Program T 3.87 293.68 288.15 52 46 Weld Metal X 4.49 300.64 265.93 54 29 Heat Affected V 0.530 59.21 - - - 26 Zone Material T 3.87 -- (d) --- 24 X 4.49 - - 210.13 -- - 19 Correlation S 0.479 - - 72.79 --- 3 Monitor Material V 0.530 - - 69.39 --- 5 T 3.87 - - 156.83 --- 5 X 4.49 - - 125.21 - - - 0 Notes:

(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

(b) Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 (See Appendix B)

(c) Values are based on the definition of upper shelf energy given in ASTM El185-82.

(d) Only 2 specimens were tested from capsule T to confirm the upper shelf energy, thus, there is insufficient data to determine the measured 30 ft-lb shift.

Analysis of Hl.B. Robinson Unit 2 Capsule X

5-16 Table 5-11 Tensile Specimens From Intermediate Shell Course Plate W10201-4 and Weld Material Sample Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Material Temperature Strength Strength Load Stress Strength Elongation Elongation in Area (OF) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) (%)

C6 PLATE 200 69.3 87.6 2.80 123.4 57.0 9.9 19.3 54 C7 PLATE 550 66.2 89.6 3.80 138.4 77.4 10.5 17.1 44 W1 WELD 275 91.2 105.3 3.75 186.5 76.4 9.5 19.8 59 W2 WELD 550 87.6 101.7 4.10 166.6 83.5 8.3 16.8 50 Analysis of H.B. Robinson Unit 2 Capsule X

5-17 INTERMEDIATE SHELL PLATE W10201-4 (LONG.)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 13:57:18 on 10-23-2001 Results Curve Fluence ISE d-LSE USE d-USE To30 d-To30 To50 d-To50 1 0 2.19 0 95 0 -18.17 0 17.35 5-17 0

2 0 219 0 85 -10 14.33 32.51 602 42.84 3 0 2.19 0 94 -1 86.55 104.73 116.04 98.68 CI) 0 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend ID- 20 ---------- 3 .-------

Data Set(s) Plotted Curve Plant CaDsUle flri l-l*at #

oa4 Malberal r 1 H132 UNIRR PLATE SA302B "LT A-6604-1 2 1182 S PLATE SA302B LT A-6604-1 3 H132 x PLATE SA302B LT A-6604-1 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Longitudinal Orientation)

Analysis of H.B. Robinson Unit 2 Capsule X

5-18 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Longitudinal Orientation)

Analysis of H.B. Robinson Unit 2 Capsule X

5-19 INTERMEDIATE SHELL PLATE W10201-4 (LONG)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 14:18.H3 on 10-23-2001 Results Curve Fluence T o 50x Shear d-T 0 5fi'/_ Rho*r T o 50Y Shear -0K/qn 1 0 25.31 0 2 0 2040 2014.68 3 0 132.12 106.81 4)

C-

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10- 20 ----------

Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat#

1 H12 UNIRR PLATE SA302B LT A-6604-1 2 H82 S PLATE SA302B LT A-6604-1 3 HB2 X PLATE SA302B LT A-6604-1 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Longitudinal Orientation)

Analysis ot H.13. Robmson Umt 2 Capsule X

5-20 SURVELLIANCE PROGRAM WELD MATERIAL CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 0954.57 on 10-24-2001 Results Curve Fluence ISE d-ISE USE d-USE T

30 d-T o 30 T o 50 d-T o 50 1 0 2.19 0 113 0 -8629 0 -40.35 0 2 0 2.19 0 70 -43 123.02 209.32 214.59 254.94 3 0 2.2 0 61 -52 201.86 288.15 205.94 246.3 4 0 2.19 0 80 -33 179.64 265.93 211.38 251.74 0

z

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10-4 _ _

Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat#

I HB2 UNIRR WELD N/A W5214 2 HB2 V WELD N/A W5214 3 H12 T WELD N/A W5214 4 HB2 X WELD N/A W5214 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Surveillance Weld Metal Analysis of H.B. Robinson Unit 2 Capsule X

5-21 SURVEILLANCE PROGRAM WELD MATERIAL CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 095952 on 10-24-2001 Results Curve Fluence USE d-USE T @LE35 d-T o LE35 1 0 91.98 0 -60.64 0 2 0 65.7 -2627 19026 250.9 3 0 41.57 -50.4 2042 264.5 4 0 542 -37.77 21924 279.89 U) co

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10 20 ---------- 3 4" Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat#

1 HB2 UNIRR WELD N/A 15214 2 H132 V WELD N/A W5214 3 HB2 T WELD N/A W5214 4 H132 X WELD N/A W5214 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Surveillance Weld Metal Analysis of H.B. Robinson Unit 2 Capsule X

5-22 SURVIELLANCE PROGRAM WELD MATERIAL CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 1tf.03'28 on 10-24-2001 Results Curve Fluence T o 50/ Shear d-T o 50>'. Shear 1 0 -30.93 0 2 0 168.75 199.68 3 0 22125 252.18 4 0 198.75 229.68 a)

-4 4-j a) a)

a)

-300 -200 -100 0 100 200 300 400 5W0 600 Temperature in Degrees F Curve Legend 10 2 ..........

- 30 --- 4 -

Data Set(s) Plotted Curve Plant Capsule Material Ori. lHeat#

1 H12 UNIRR WELD N/A W5214 2 HB2 V WELD N/A W5214 3 HB2 T WELD N/A W5214 4 HB2 X WELD N/A W5214 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Surveillance Weld Metal Analysis of H.B. Robinson Unit 2 Capsule X

5-23 HEAT AFFECTED ZONE CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10:36:17 on 10-24-2001 Results Curve Fluence ISE d-ISE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1 0 .19 0 129 0 -109.66 0 -66.04 0 2 0 2.19 0 96 -33 -50.44 5921 -.75 6528 3 0 2.19 0 105 -24 100.47 210.13 150.54 216.59 CI bi C.r)

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend to- 20G.------- 30.--

Data Set(s) Plotted Curve Plant Cansule Material N* 14o*f#

& Mateiallri He I HB2 UNIRR HEAT AFFD ZONE SA302B A-66Z1-1 2 H132 V HEAT AFFD ZONE SA302B A-660-I 3 HM2 X HEAT AFFD ZONE SA302B A-6023-1 4 HB2 T HEAT AFFD ZONE SA3O2B A-660-1 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Heat Affected Zone Material Analysis Of H.1. Robinson Unit 2 Capsule X

5-24 HEAT AFFECTED ZONE CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10.3927 on 10-24-2001 Results Curve Fluence USE d-USE T @LE35 d-T o LE35 1 0 93.06 0 -84.08 0 2 0 76.67 -16.38 -5.18 78.9 3 0 66.83 -2622 164.51 248.6 Y) 4-

-300 -200 -100 0 100 2W0 300 4o00 500 600 Temperature in Degrees F Curve Legend 1D- 3 .-----

Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat#

I HB2 UNIRR HEAT AFFD ZONE SA302B A-6623-1 2 HB2 V HEAT AFFID ZONE SA302B A-6623-1 3 HB2 X HEAT AFDD ZONE SA302B A-6623-1 4 HB2 T HEAT AFFD ZONE SA302B A-8Zi3-1 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Heat Affected Zone Material Analysis of H.B. Robinson Unit 2 Capsule X

5-25 HEAT AFFECTED ZONE CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10:42:31 on 10-24-2001 Results Curve Fluence T o 50z. Shear d-T 0 50/ Shear 1 0 -39.84 0 2 0 33.02 72.86 3 0 140.33 180.17

-4 C)

CU

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 1 13 3 .---.-

Data Set(s) Plotted Curve Plant Capsule Material Ori. He~at/!

Ori. Heat#

1 HB2 U'NIRR HEAT AFFD ZONE SA302B 2 HB2 V HEAT AFFD ZONE SA302B A-6620-1 3 HB2 X HEAT AFFD ZONE SA302B A-6623-1 4 HB2 T HEAT AFFD ZONE SA302B A-6623-1 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Heat Affected Zone Material Analysi os 1-o.B. Robinson Umt 2 Capsule X

5-26 CORRELATION MONITOR (SRM)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 13:3915 on 10-24-2001 Results Curve Fluence [SE d-LSE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1 u 22 0 39 0 62.94 0 2 0 2.2 0 38 -1 135.73 72.79 3 0 2.2 0 37 -2 13233 69.39 4 0 2.2 0 37 -2 219.77 156.83 5 0 22 0 42 3 188.15 12521 300F 1 300- 1 1

-')t*

~~~I

'i,~~~~nE t I ___________

200 1501 I I 10 CL) 50-

= -

A I

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 1O - 20---------- 30.---.--- 4 - 5z Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat#

1 HB2 UNIRR SRM SA302B TL 2 ff82 S SRM SA302B TL 3 H112 V SRM SA302B TL 4 HB2 T SRM SA302B TL 5 H12 X SRM SA302B TL Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Correlation Monitor Material Analysis of H.1. Robinson Unit 2 Capsule X

5-27 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Correlation Monitor Material Analysis of H.B. Robinson Unit 2 Capsule X

5-28 CORRELATION MONITOR (SRM)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 13.52:40 on 10-24-2001 Results Curve Fluence T o 50Y. Shear d-T o 50Y Shear 1 0 T o SOz. Shear 37.03 0 2 0 1530 1492.96 3 0 195.46 158.43 4 0 175.78 138.75 5 0 156.73 119.7 4-)

©-

C~)

-300 -2W0 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10O 20---------- 30..... 4- 5v Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat#

I Ht2 UNIRR SRM SA302B TL 2 HB2 S SRM SA302B TL 3 HB2 V SRM SA302B TL 4 H12 T SRM SA302B TL 5 HH2 X SRM SA302B TL Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for H.B. Robinson Unit 2 Reactor Vessel Correlation Monitor Material Analysis of H.B. Robinson Unit 2 Capsule X

5-29 C45, 125° C46, 150° C41, 250°F C43, 325WF C47, 350WF Figure 5-13 Charpy Impact Specimen Fracture Surfaces for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Longitudinal Orientation)

Analysis of H.B. Robinson Unit 2 Capsule X

5-30 W2. lO0F W4, 200°F W8, 250TF W7, 350°F W5, 375-F W1, 425°F Figure 5-14 Charpy Impact Specimen Fracture Surfaces for H.B. Robinson Unit 2 Reactor Vessel Weld Metal Specimens Analysis of H.B. Robinson Unit 2 Capsule X

5-31 r:4yc

--17 flOF H6, 25°F H5, lO0-F H4, 150 0 F ft !,

H3, 2500 F H8, 375 0 F H1, 4000 F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 2 Reactor Vessel Heat Affected Zone (HAZ)

Analysis of H.B. Robinson Unit 2 Capsule X

5-32 5-32 R43, 200°F R46, 225°F R47, 300°F R42, 325-F R44, 425-F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for H.B. Robinson Unit 2 Reactor Vessel Correlation Monitor Material Analysis of H.B. Robinson Unit 2 Capsule X

5-33 (0 C) 0 50 100 150 200 250 300 350 120 I I I ! I I I_

800 110 100 ULTIMATE TENSILE STRENGTH 700 90 A A 600 A2 CS 80 C-La 70 500 I-C-, 60 400 50 0.2% YIELD STRENGTH 40 300 LEGEND:

Ao UNIRRADIATED A

19 2 IRRADIATED TO A FLUENCE OF 4.49 X 10 n/cm (E>1.0MeV) AT 550 F 0

80 70 REDUCTION IN AREA 60 50 -

ý

"'I A 40 30 TOTAL ELONGATION 20 -0 ý 2 10 0

I ELONGATION UNIFORM I II I 0 100 20 0 300 400 500 600 700 TEMPERATURE (OF)

Figure 5-17 Tensile Properties for H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Transverse Orientation)

Analysis of H.B. Robinson Unit 2 Capsule X

I I 5-34 (0C) 0 50 100 150 200 250 300 350 120 II I I I I I " _

800 110 A-A 100 700 90 Lc%

-2 600 80 "S-V- 70 ULTIMATE TENSILE STRENGTH 500 I.

60 0.2%

0.2% YIELD E0 YIELSTRENGTH 400 50 40 300 LEGEND:

Ao UNIRRADIATED 19 Ae IRRADIATED TO A FLUENCE OF 4.49 X 10 n/cm2 (E>1.OMeV) AT 550 0 F 80 70 REDUCTION INAREA 60

,-, 50

-J I

40 C-,

30 TOTAL ELONGATION 20 2 *A 10 UNIFORM ELONGATION A 0 I I I I I 0 100 2010 300 400 500 600 700 TEMPERATURE (OF)

Figure 5-18 Tensile Properties for H.B. Robinson Unit 2 Reactor Vessel Weld Metal Analysis of H.B. Robinson Unit 2 Capsule X

5-35

ý1if. 11 u m! I ttjl Ik1 j q q 1r II I~*I I*I Ij I ý, I 1,J ,* . ...-

Specimen C6 Tested at 200'F Specimen C7 Tested at 550'F Figure 5-19 Fractured Tensile Specimens from H.B. Robinson Unit 2 Reactor Vessel Intermediate Shell Plate W10201-4 (Transverse Orientation)

Analysis of H.B. Robinson Unit 2 Capsule X

5-36

~

li~~~~~ll[.

llIll~l ~ Specimen L

Wi. IATested

~ 1 ; 1,*IIHII.ITt L1!II I.LKr[lll[

at ,11275°F HEOt*[t*Lt8 DO]* M~HU *L LL11DO T

Specimen W2 Tested at 550'F Figure 5-20 Fractured Tensile Specimens from H.B. Robinson Unit 2 Reactor Vessel Weld Metal Analysis of H.B. Robinson Unit 2 Capsule X

5-37 STRESS-STRAIN CURVE ROBINSON UNIT 2 "X" CAPSULE 100 90 80 70 60 C," 50 I-U, 40 30 C6 20 200 F 10 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN STRESS-STRAIN CURVE ROBINSON UNIT 2 "X" CAPSULE 100 90 80 70 60

,, 50 LI n,0 40 30 C7 20 550 F 10 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-21 Engineering Stress-Strain Curves for Intermediate Shell Plate W10201-4 Tensile Specimens C6 and C7 (Transverse Orientation)

Analysis of H.B. Robinson Unit 2 Capsule X

5-38 STRESS-STRAIN CURVE ROBINSON UNIT 2 "X* CAPSULE 100 80 60 LiJ I--

40 Wl 275 F 20 0 0.05 0. 0.15 0.2 0.25 0.3 STRAIN, INAN STRESS-STRAIN CURVE ROBINSON UNIT 2 "X"CAPSULE 100 80 55 60 LU IC, 40 W2 20 550 F 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-22 Engineering Stress-Strain Curves Weld Metal Tensile Specimens W1 and W2 Analysis of H.B. Robinson Unit 2 Capsule X

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates Sn transport analysis performed for the H. B. Robinson Unit 2 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this evaluation, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis for the first 21 reactor operating cycles. In addition, neutron dosimetry sensor sets from Surveillance Capsules S, V, T, and X withdrawn from the H. B. Robinson Unit 2 reactor at the conclusion of fuel cycles 1, 3, 8, and 20, respectively, were analyzed using current dosimetry evaluation methodology. Comparisons of the results of these dosimetry evaluations with the analytical predictions provided a validation of the plant specific neutron transport calculations. These validated calculations were then used to provide projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 50 Effective Full Power Years (EFPY). These projections conservatively account for an assumed plant uprating, from 2300 MWt to 2335 MWt, beginning with the operation of the twenty second fuel cycle.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, "Analysis and Interpretation of Light-Water Reactor Surveillance Results,"'[ 31 recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom."'" 41 The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."E'11 Recently a new standard dpa cross section has been developed based on ENDF/B-VI cross sections. [15] If the new standard was used, minor changes in calculated dpa values would be obtained. As tabulated in Reference 15, this would result in about a 0.3% decrease in dpa at the reactor vessel surface, a 1.7% increase at the 1/4T position, and a 3.7% increase at the 3/4T position for the H. B. Robinson reactor vessel. Since the correlations used for embrittlement extrapolation and Regulatory Guide 1.99 are based on the earlier ENDF/B-IV dpa cross section, the results in this report are also based on the older standard.

All of the calculations and dosimetry evaluations described in this section were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools.

Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance and meet the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."E' 61 Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

January 1996.1171 The specific calculational methods applied are also consistent with those described in WCAP- 15557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology. "[l 8]

Analysis of H.B. Robinson Unit 2 Capsule X

6-2 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the H. B. Robinson Unit 2 reactor geometry at the core midplane is shown in Figure 4-1.

Eight irradiation capsules attached to the thermal shield were included in the reactor design that constitutes the reactor vessel surveillance program. The capsules were located at azimuthal angles of 270' (00 from the core cardinal axis), 2800 (10' from the core cardinal axis), 290' (20' from the core cardinal axis), 30' and 150' (30' from the core cardinal axes), and 40', 50', and 2300 (40' from the core cardinal axes) as shown in Figure 4-1. The stainless steel specimen containers are 1-inch square by 36 inches in height.

The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core.

From a neutronic standpoint, the surveillance capsules and associated support structures are significant. The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.

The fast neutron exposure evaluations for the H. B. Robinson Unit 2 surveillance capsules and reactor vessel were based on a series of fuel cycle specific forward transport calculations that were combined using the following three-dimensional flux synthesis technique:

4(r,0,z) = [(r,)] * [rz)/[r]

where 0(r,O,z) is the synthesized three-dimensional neutron flux distribution, 0(r,0) is the transport solution in r,O geometry, 0(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and 0(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at H. B. Robinson Unit 2.

The synthesis procedure for H. B. Robinson Unit 2 was complicated by the use of partial length shield assemblies (PLSA) beginning in fuel cycle 10. These assemblies are designed to provide a reduction in exposure to the lower circumferential weld. The PLSA are located in the three outer row positions near each of the core cardinal axes where the fluence maximum is located. The fuel pins in the PLSA contain of lower enrichment fuel pellets in the top 8.5 feet of the core region and contain stainless steel rods in the lower 3.5 feet instead of fuel. The PLSA assemblies are irradiated at these locations for a total of six cycles.

To perform the synthesis with the assymetry introduced by the PLSA, the core source is divided into two axial regions and two sets of calculations are performed. For the top region, the flux is given by 0t(r,0,z) = [H(r,0)] * [0t(r,z)]/[4t(r)].

In this region, the axial source is set to zero in the lower 3.5 feet of the core. In the PLSA region, the flux is given by OPp(r,0,z) = [Op(r,0)] * [OP(r,z)]/[d~(r)].

In the PLSA region, the axial source is set to zero in the upper 8.5 feet of the core. The total source is then given by the sum of the two parts:

,0A = 0t(r,0,z) + *V(r,0,z).

Analysis of H.B. Robinson Unit 2 Capsule X

6-3 For the H. B. Robinson Unit 2 calculations, a one octant r,0 model was developed since the reactor core is octant symmetric. This r,0 model includes the core, the reactor internals, the thermal shield -- including explicit representations of the surveillance capsules at 00, 100, 200, 300, and 400, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. The r,0 model was utilized to perform both the surveillance capsule dosimetry evaluations, and subsequent comparisons with calculated results, and to generate the maximum fluence levels at the pressure vessel wall.

The inclusion of all the capsules in a single octant does not represent a real case, since no octant has capsules at all positions, but the perturbation between capsules is negligible. However, the capsules do reduce the fluence to the vessel by 5 to 7% at nearby vessel locations. This perturbation was examined by running several cases with and without capsules, and appropriate corrections were made to obtain the maximum vessel exposure which occurs at locations where no capsules are present.

In developing the analytical model, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis for cycles 8 and 9 which ran at a reduced power. (The power was reduced for part of cycle 8 and all of cycle 9 due to steam generator limitations). The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The r,0 geometric mesh description of the reactor model consisted of 161 radial by 107 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a point-wise basis. The point-wise inner iteration flux convergence criterion utilized in the r", calculations was set at a value of 0.001.

For the top region (fuel cycles 10 to 21) the appropriate radial and azimuthal source distribution was used and included the power produced in this axial region by each fuel assembly. For the PLSA region, the source averaged over the bottom 3.5 feet of the core was used and the source in the PLSA assemblies was set to zero.

The r,z model used for the H. B. Robinson Unit 2 calculations extended radially from the centerline of the reactor core out to a location within the primary biological shield and over an axial span from an elevation 1-foot below the active fuel to 1-foot above the active fuel. As in the case of the r,0 model, nominal design dimensions and full power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of the reactor model consisted of 158 radial by 110 axial intervals. As in the case of the r,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a point-wise basis. The point-wise inner iteration flux convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

The rz neutron source for the cycles with PLSA was handled as follows. In the calculation for the top region, the source was set to zero in the bottom 3.5 feet of the core. In the calculation for the bottom 3.5 feet, the source in the top 8.5 feet was set to zero. The appropriate radial source distribution for each region was used in each case. It should be noted that the r,z geometrical model was similar for each of the two regions with the only difference being a composition change in the annular region corresponding to the outer row of assemblies to represent the substitution of stainless steel for fuel.

The one-dimensional radial model used in the synthesis procedure consisted of the same 158 radial mesh intervals included in the rz model. Thus, radial synthesis factors could be determined on a mesh-wise basis throughout the entire geometry.

Analysis of H.B. Robinson Unit 2 Capsule X

6-4 The core power distributions used in the plant specific transport analysis were mostly taken from an information package supplied by the H. B. Robinson Unit 2 plant stafft 191 and H. B. Robinson Unit 2 fuel cycle design reports.[20' 2'1 For the first fuel cycle, a generic axial power distribution for three loop plants was used. 221 The data extracted from these reports represented cycle dependent fuel assembly enrichments, bum ups, and axial power distributions. This information was used to develop spatial and energy dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and bum-up history of individual fuel assemblies. From these assembly dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3.1 [231 and the BUGLE-96 cross-section library. [241 The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 legendre expansion and angular discretization was modeled with an 516 order of angular quadrature. Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-4. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the Capsules S, V, T, and X irradiation and provide the calculated neutron exposure of the pressure vessel wall for the first 21 fuel cycles. In Table 6-1, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the five different azimuthally symmetric surveillance capsule positions (00, 100, 20', 30' and 40'). These data, calculated at the axial midplane of the active core, are meant to establish the exposure of the surveillance capsules withdrawn to date and to provide data for an absolute comparison of measurement with calculation. Similar information is provided in Table 6-2 for the reactor vessel inner radius. The vessel data given in Table 6-2 calculated at the axial location of the maximum neutron exposure at each of six azimuthal locations. Again, both fluence (E > 1.0 MeV) and dpa data are provided. It is important to note that the data for the vessel inner radius were taken at the clad/base metal interface, and thus, represent the maximum calculated exposure levels of the vessel plates. The axial position of maximum exposure rate varies from cycle to cycle, and therefore the integrated exposure cannot be directly derived from the cycle exposure rates in this table.

However, the difference is very small since the flux level is nearly flat in the axial region encompassing the various cycle maxima and the reactor axial midplane.

Radial gradient information applicable to 4(E > 1.0 MeV) and dpa/sec are given in Tables 6-3 and 6-4, respectively. The data, based on the cumulative exposure through cycle 20, are presented on a relative basis for each exposure parameter at six azimuthal locations. Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-3 and 6-4.

Analysis of H.B. Robinson Unit 2 Capsule X

6-5 6.3 NEUTRON DOSIMETRY 6.3.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the four neutron sensor sets withdrawn to date as a part of the H. B. Robinson Unit 2 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time [EFPY]

S 10 End of Cycle 1 1.28 V 20 End of Cycle 3 3.18 T 0 End of Cycle 8 7.27 X 40, 0 End of Cycle 20 20.39 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules. Capsule X was irradiated at the 40° location for the first 8 cycles and was then moved to the 0' location.

The passive neutron sensors included in the evaluations of Surveillance Capsules S, V, T, and X are summarized as follows:

Reaction Sensor Material of Interest Capsule S Capsule V Capsule T Capsule X 63 Copper Cu(na)6°Co x X X X

54 Iron Fe(n,p)4Mn X X X X 58 Nickel Ni(np)58Co X X X X 23 8 Uranium-238 U(n,f)l Cs 37 none X X 237 Neptunium-237 Np(n,f) 137 Cs none X X 59 Cobalt-Aluminum* Co(n,y)6°Co X X X X

The cobalt-aluminum measurements for this plant include both bare wire and cadmium-covered sensors.

- The fission reaction data for this capsule are regarded as in error and were not used.

The copper, nickel, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several radial locations within the test specimen array. Iron wires were not included in these capsules, but iron activation measurements were made on samples taken from Charpy bars. As a result, it was necessary to apply gradient corrections to these measured reaction rates in order to index all of the sensor measurements to the radial center of the respective surveillance capsules. Since the cadmium-shielded uranium and neptunium Analysis of H.B. Robinson Unit 2 Capsule X

6-6 fission monitors were accommodated within the dosimeter block centered at the radial, azimuthal, and axial center of the material test specimen array, gradient corrections were not required for the fission monitor reaction rates. Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table 6-5.

The use of passive monitors such as those listed above does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.

In particular, the following variables are of interest:

"* the measured specific activity of each monitor,

"* the physical characteristics of each monitor,

"* the operating history of the reactor,

"* the energy response of each monitor, and

"* the neutron energy spectrum at the monitor location.

The radiometric counting of the neutron sensors from capsules S and V was carried out at the Southwest Research Institute [25,261. These data are less reliable than more modem data, and corrections were made to some of the data in the Westinghouse Surveillance Capsule Neutron Fluence Evaluation1 279,carried out in 1994. For capsule S, two Cu reaction rate measurements were made that were inconsistent. In the present analysis, the top Cu measurement was discarded and the bottom measurement, which appears to be consistent with the other data sets, was used. For capsule V, the iron reaction rate was decreased by dividing by 1.15 as recommended in Reference 27. The analysis of capsule T was carried out at the Westinghouse Analytical Services Laboratory at the Waltz Mill Site and results are reported in Reference 28. The radiometric counting of the sensors from capsule X was completed at the Antech Analytical Laboratory, also located at the Waltz Mill Site.t 291 For capsules T and X, the radiometric counting followed established ASTM procedures.

Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by Capsules S, V, T, and X was based on the reported monthly power generation of H. B. Robinson Unit 2 from initial reactor startup through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules S, V, T, and X is given in Table 6-6.

Analysis of H.B. Robinson Unit 2 Capsule X

6-7 Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation:

R =A R- A No F Y Cj [I -e-"][e-"d]

Pef where:

R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref (rps/nucleus).

A = Measured specific activity (dps/gm).

No = Number of target element atoms per gram of sensor.

F = Weight fraction of the target isotope in the sensor material.

Y = Number of product atoms produced per reaction.

Pj = Average core power level during irradiation period j (MW).

Pref = Maximum or reference power level of the reactor (MW).

Cj= Calculated ratio of O(E > 1.0 MeV) during irradiation period j to the time weighted average 4(E > 1.0 MeV) over the entire irradiation period.

,= Decay constant of the product isotope (1/sec).

tj = Length of irradiation period j (sec).

td = Decay time following irradiation periodj (sec).

and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Cj term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another. The fuel cycle specific neutron flux values along with the computed values for Cj are listed in Table 6-7. These flux values represent the cycle dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

Normally, the neutron spectrum at the capsule location is relatively constant and thus a single set of Cj values can be used for all the dosimeter reactions. In the case of Capsule X, however, the change in position creates a change in the spectrum that was taken into account in the analysis. For this capsule, Cj was determined separately for each individual reaction.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma ray induced Analysis of H.B. Robinson Unit 2 Capsule X

1ý.

6-8 fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the H. B. Robinson Unit 2 fission sensor reaction rates are summarized as follows:

235 Correction Capsule T Capsule X U Impurity/Pu Build-in 0.747 0.727 238 U(y,,f) 0.956 0.956 Net 238U Correction 0.714 0.695 237 Np(Y,f) 0.984 0.984 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules S, V, T, and X are given in Table 6-8. In Table 6-8, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 238U impurities, plutonium build-in, and gamma ray induced fission effects.

6.3.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as )(E> 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, g+/-, =(0-g

+/- al )(0g

+/-8g +Sgo g

relates a set of measured reaction rates, R*, to a single neutron spectrum, 4 g, through the multigroup dosimeter reaction cross-section, rig, each with an uncertainty 8. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least squares evaluation of the H. B. Robinson Unit 2 surveillance capsule dosimetry, the FERRET codet 30 ] was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (ý(E > 1.0 MeV) and dpa) along with associated uncertainties for the four in-vessel capsules withdrawn to date.

Analysis of H.B. Robinson Unit 2 Capsule X

6-9 The application of the least squares methodology requires the following input:

1- The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2- The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3- The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the H. B. Robinson Unit 2 application, the calculated neutron spectrum was obtained from the results of plant specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section 6.3.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library[311. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E 1018, "Application of ASTM Evaluated Cross-Section 32 Data File, Matrix E 706 (IIB)".[ ]

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E 944, "Application 331 of Neutron Spectrum Adjustment Methods in Reactor Surveillance."[

The following provides a summary of the uncertainties associated with the least squares evaluation of the H. B. Robinson Unit 2 surveillance capsule sensor sets.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:

Reaction Uncertainty 63 Cu(n,ct) 60Co 5%

54Fe(n,p)54Mn 5%

58 Ni(n,p) 58Co 5%

238 U(n,f)137Cs 10%

237 Np(n,f)137Cs 10%

59 Co(n,y) 60 Co 5%

These uncertainties are given at the 1 level.

Analysis of H.B. Robinson Unit 2 Capsule X

6-10 Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library.

This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as inihe fluence and energy characterization of 14 MeV neutron sources.

For sensors included in the H. B. Robinson Unit 2 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

Reaction Uncertainty 63 Cu(n,M)6°Co 4.08-4.16%

54 Fe(n,p)54Mn 3.05-3.11%

58Ni(n,p) 58 Co 4.49-4.56%

238 37 U(n,f)1 Cs 0.54-0.64%

237 Np(n,f)137Cs 10.32-10.97%

59 Co(n,Y)6°Co 0.79-3.59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Calculated Neutron Spectrum The neutron spectra input to the least squares adjustment procedure were obtained directly from the results of plant specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape).

Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

Mgg,= R~2 +Rg *Rg *Pgg, where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties R. and RP.

specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Pgg, = [fl-O]3,g + 0 e-H where 2

H- (g_2g,)

Analysis of H.B. Robinson Unit 2 Capsule X

6-11 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 5 is 1.0 when g = g', and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the H. B. Robinson Unit 2 calculated spectra was as follows:

Flux Normalization Uncertainty (RJ) 15%

Flux Group Uncertainties (R,, Rk,)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

Short Range Correlation (0)

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 6.3.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the four H. B. Robinson Unit 2 surveillance capsules withdrawn to date are provided in Tables 6-9 and 6-10. In Table 6-9, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates. These ratios of M/C and M/BE illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Table 6-10, comparison of the calculated and best estimate values of neutron flux (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the C/BE ratios observed for each of the capsules.

The data comparisons provided in Tables 6-9 and 6-10 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 6.4 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the 1a level. From Table 6-10, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to 6-8% for neutron flux (E > 1.0 MeV) and 7-9% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the la level.

Analysis of H.B. Robinson Unit 2 Capsule X

6-12 Further comparisons of the measurement results with calculations are given in Tables 6-11 and 6-12. These comparisons are given on two levels. In Table 6-11, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table 6-12, calculations of fast neutron exposure rates in terms of O(E > 1.0 MeV) and dpa/s are compared with the best estimate results obtained from the least squares evaluation of the four capsule dosimetry results.

These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the C/M comparisons for fast neutron reactions range from 0.76-1.13 for the samples included in the data set. The overall average C/M ratio for the entire set of H. B. Robinson Unit 2 data is 0.87 with an associated standard deviation of 11.4%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding C/BE ratios for the four capsule data set range from 0.8 1-0.95 for neutron flux (E > 1.0 MeV) and from 0.82 to 0.95 for iron atom displacement rate. The overall average C/BE ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 0.89 with a standard deviation of 7.9% and 0.89 with a standard deviation of 7.1%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.4 of this report are validated for use in the assessment of the condition of the materials comprising the beltline region of the H. B. Robinson Unit 2 reactor pressure vessel.

Analysis of H.B. Robinson Unit 2 Capsule X

6-13 6.4 PROJECTIONS OF REACTOR VESSEL EXPOSURE The final results of the fluence evaluations performed for the four surveillance capsules withdrawn from the H. B. Robinson Unit 2 reactor are provided in Table 6-13. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the H. B. Robinson Unit 2 reactor. As shown by the comparisons provided in Tables 6-11 and 6-12, the validity of these calculated fluence levels is demonstrated both by a direct comparison with measured sensor reaction rates as well by comparison with the least squares evaluation performed for each of the capsule dosimetry sets.

The corresponding calculated fast neutron fluence (E > 1.0 MeV) and dpa exposure values for the H. B.

Robinson Unit 2 pressure vessel are provided in Table 6-14. As presented, these data represent the maximum exposure of the clad/base metal interface at azimuthal angles of at 00, 100, 200, 300, 400, and 450, relative to the core cardinal axes. The data tabulation includes the plant and fuel cycle specific calculated fluence at the end of the twentieth operating fuel cycle as well as projections for future operation to the end of cycle 21 and to 29, 30, 35, 40, 45, and 50 effective full power years. The projections were based on the assumption that the spatial power distributions averaged over fuel cycles 16 through 21 (these six cycles represent the exposure of a single set of PLSA) will be representative of future plant operation. The future projections also account for a 1.5% power uprate from 2300 MWt to 2335 MWt.

Similar calculated fluence projections are provided in Tables 6-15 and 6-16 for the H. B. Robinson Unit 2 circumferential welds. The upper circumferential weld is located 5.375 inches below the top of the fuel region and fluence and dpa for this weld are in Table 6-15. The lower circumferential weld is located at 21.75 inches above the bottom of the fuel. Fluence and dpa for this weld are in Table 6-16. This weld has a reduced fluence rate near 00 for cycles starting with 10 due to the PLSA. This reduction shifts the maximum fluence rate to this weld to angles between 20 and 30 degrees. However, due to the large fluence acquired in cycles 1 to 9 near 00, the maximum fluence will be at 00 through 50 EFPY Data from Table 6-14 can also be used to determine the maximum exposure to the three intermediate shell vertical welds, which are located at angles of 100, 20', and 400 with respect to the cardinal axes. Similarly, the maximum exposure to the vertical welds in the upper shell (also located at 10', 200, and 40' with respect to the cardinal axes) can be taken from Table 6-15 and the maximum exposure to the vertical welds in the lower shell (located at 00, 30', and 300 with respect to the cardinal axes) can be taken from Table 6-16.

Updated lead factors for the H. B. Robinson Unit 2 surveillance capsules are provided in Table 6-17. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface. In Table 6-17, the lead factors for capsules that have been withdrawn from the reactor (S, V, T, and X) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (U, Y, W, and Z), the lead factors correspond to the calculated fluence values at the end of cycle 20. For these four capsules, future lead factors are also given in Table 6-17 since these can be different from the past exposure.

In addition to the pressure vessel materials located adjacent to the active core region, nozzle components located in the vicinity of the azimuthal peak fluence will experience a neutron exposure (E > 1.0 MeV) greater than 1.OE+ 17 n/cm 2 during the extended operating periods. These components include the inlet and outlet nozzle structures as well as the circumferential welds connecting the nozzles to the vessel shell course. The portions of these materials experiencing fluence values greater than 1.OE+17 n/cm 2 are limited to small areas at the lowest extent of the nozzles and associated welds as well as to the inlet and outlet nozzles located near Analysis of H.B. Robinson Unit 2 Capsule X

6-14 the 0.0 degree azimuth. A summary of the maximum calculated neutron fluence applicable to the nozzle components is given in Table 6-18.

The uncertainty associated with the calculated neutron exposure of the H. B. Robinson Unit 2 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

I- Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2- Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

3- An analytical sensitivity study addressing the uncertainty components resulting important input parameters applicable to the plant specific transport calculations used in the neutron exposure assessments.

4- Comparisons of the plant specific calculations with all available dosimetry results from the H. B. Robinson Unit 2 surveillance program capsules.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters. The overall calculational uncertainty applicable to the H. B. Robinson Unit 2 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with H. B. Robinson Unit 2 capsule measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures.

The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 7.

Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

The net calculational uncertainty was determined by combining the individual components in quadrature which assumes that each component of the uncertainty is random and uncorrelated. Therefore, no systematic bias was applied to the analytical results.

The plant specific measurement comparisons provided in Tables 6-11 and 6-12 support these uncertainty assessments for H. B. Robinson Unit 2.

Analysis of H.B. Robinson Unit 2 Capsule X

6-15 Table 6-1 Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cycle Total Neutron Flux (E > 1.0 MeV)

Length Irradiation [n/crn 2s]

Cycle [EFPY] Time

[EFPY] 0 Degrees 10 Degrees 20 Degrees 30 Degrees 40 Degrees 1 1.279 1.28 1.71E+1 1 1.19E+11 5.33E+10 4.1OE+10 2.84E+10 2 0.814 2.09 1.70E+1I 1.22E+11 5.65E+10 4.37E+10 2.94E+10 3 1.091 3.18 1.45E+11 1.01E+11 4.94E+10 3.98E+10 2.78E+10 4 0.791 3.97 1.92E+1I1 1.35E+11 5.75E+10 4.39E+10 3.05E+10 5 0.812 4.79 1.77E+11 1.26E+11 5.56E+10 4.23E+10 2.83E+10 6 0.840 5.63 1.70E+l 1 1.20E+l1 5.39E+10 4.14E+10 2.77E+10 7 0.810 6.44 1.64E+11 1.17E+1I 5.54E+10 4.29E+10 2.82E+10 8 0.835 7.27 1.67E+11 1.18E+1I 5.31E+10 4.07E+10 2.69E+10 9 0.857 8.13 8.35E+10 6.10E+10 4.09E+10 3.67E+10 2.52E+10 10 0.871 9.00 1.05E+11 7.87E+10 5.49E+10 4.59E+10 2.72E+10 11 0.916 9.92 9.51E+10 7.53E+10 5.73E+10 4.77E+10 2.63E+10 12 0.982 10.90 8.93E+10 7.30E+10 5.98E+10 4.91E+10 2.62E+10 13 0.983 11.88 7.67E+10 6.44E+10 5.61E+10 4.78E+10 2.64E+10 14 0.996 12.88 7.60E+10 6.58E+10 6.02E+10 5.06E+10 2.68E+10 15 1.074 13.95 7.16E+10 6.31E+10 5.98E+10 5.01E+10 2.60E+10 16 1.076 15.03 1.28E+11 9.06E+10 4.01E+10 3.03E+10 2.36E+10 17 1.185 16.21 1.13E+ll 8.15E+10 3.83E+10 2.78E+10 2.02E+ 10 18 1.346 17.56 1.02E+1I 7.41E+10 3.73E+10 2.81E+10 1.92E+10 19 1.414 18.97 9.17E+10 6.72E+10 3.54E+10 2.82E+10 2.16E+10 20 1.420 20.39 8.OOE+10 5.96E+10 3.32E+10 2.73E+10 2.18E+10 21 1.389 21.78 7.02E+10 5.37E+10 3.28E+10 2.71E+10 2.11E+10 Analysis of H.B. Robinson Unit 2 Capsule X

6-16 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cycle Total Neutron Fluence (E > 1.0 MeV)

Length Irradiation [n/cm 2]

Cce [EFPY] Time Cycle Time

[EFPY] 0 Degrees 10 Degrees 20 Degrees 30 Degrees 40 Degrees 1 1.279 1.28 6.92E+18 4.79E+18 2.15E+18 1.65E+18 1.15E+18 2 0.814 2.09 1.13E+19 7.94E+18 3.60E+18 2.77E+18 1.90E+18 3 1.091 3.18 1.63E+19 1.14E+19 5.30E+18 4.14E+18 2.86E+18 4 0.791 3.97 2.11E+19 1.48E+19 6.74E+18 5.24E+18 3.62E+18 5 0.812 4.79 2.56E+19 1.80E+19 8.16E+18 6.32E+18 4.34E+18 6 0.840 5.63 3.O1E+19 2.11E+19 9.59E+18 7.42E+18 5.08E+18 7 0.810 6.44 3.43E+19 2.41E+19 1.10E+19 8.52E+18 5.80E+18 8 0.835 7.27 3.87E+19 2.72E+19 1.24E+19 9.59E+18 6.51E+18 9 0.857 8.13 4.10E+19 2.89E+19 1.35E+19 1.06E+19 7.19E+18 10 0.871 9.00 4.38E+19 3.11E+19 1.50E+19 1.18E+19 7.94E+18 11 0.916 9.92 4.66E+19 3.32E+19 1.67E+19 1.32E+19 8.70E+18 12 0.982 10.90 4.93E+19 3.55E+19 1.85E+19 1.47E+19 9.51E+18 13 0.983 11.88 5.17E+19 3.75E+19 2.03E+19 1.62E+19 1.03E+ 19 14 0.996 12.88 5.41E+19 3.96E+19 2.22E+19 1.78E+19 1.12E+19 15 1.074 13.95 5.65E+19 4.17E+19 2.42E+19 1.95E+19 1.20E+19 16 1.076 15.03 6.09E+19 4.48E+19 2.56E+19 2.05E+19 1.29E+19 17 1.185 16.21 6.51E+19 4.78E+19 2.70E+19 2.16E+19 1.36E+19 18 1.346 17.56 6.94E+19 5.10E+19 2.86E+19 2.28E+19 1.44E+19 19 1.414 18.97 7.35E+19 5.40E+19 3.01E+19 2.40E+19 1.54E+19 20 1.420 20.39 7.71E+19 5.66E+19 3.16E+19 2.53E+19 1.64E+19 21 1.389 21.78 8.02E+19 5.90E+19 3.31E+19 2.64E+ 19 1.73E+19 Analysis of H.B. Robinson Unit 2 Capsule X

6-17 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Iron Atom Displacements Cycle Total Displacement Rate Length Irradiation [dpa/s]

Cycle [EFPY] Time 0 Degrees 10 Degrees 20 Degrees 30 Degrees 40 Degrees

[EFPY]

1 1.279 1.28 2.91E-10 2.03E-10 8.75E-11 6.73E-11 4.61E-11 2 0.814 2.09 2.89E-10 2.09E-10 9.27E- 11 7.17E-11 4.77E-11 3 1.091 3.18 2.46E-10 1.72E-10 8.10E-I1 6.53E-11 4.51E-11 4 0.791 3.97 3.26E-10 2.30E-10 9.44E- I1 7.21E-11 4.95E-11 5 0.812 4.79 3.OOE-10 2.15E-10 9.14E-11 6.95E-11 4.59E-11 6 0.840 5.63 2.88E-10 2.05E-10 8.85E- 11 6.80E- 11 4.50E-11 7 0.810 6.44 2.78E-10 2.01E-10 9.09E- 11 7.06E-11 4.58E-11 8 0.835 7.27 2.82E-10 2.01E-10 8.70E- 11 6.68E-11 4.37E-11 9 0.857 8.13 1.40E-10 1.03E-10 6.63E-11 5.99E-11 4.06E-11 10 0.871 9.00 1.76E-10 1.34E-10 8.96E-I1 7.53E-11 4.40E-11 11 0.916 9.92 1.60E-10 1.28E-10 9.35E-11 7.84E- 11 4.26E- 11 12 0.982 10.90 1.50E-10 1.24E-10 9.75E- 11 8.07E-11 4.25E-11 13 0.983 11.88 1.29E-10 1.09E-10 9.15E-11 7.85E-11 4.28E-11 14 0.996 12.88 1.28E-10 1.1iE-10 9.82E-11 8.32E- 11 4.35E-11 15 1.074 13.95 1.20E-10 1.07E-10 9.74E-11 8.23E- 11 4.22E-11 16 1.076 15.03 2.17E-10 1.54E-10 6.56E-11 4.95E-11 3.82E-11 17 1.185 16.21 1.91E-10 1.39E-10 6.27E-11 4.54E-11 3.27E-11 18 1.346 17.56 1.72E-10 1.26E-10 6.09E-11 4.60E-11 3.11E-11 19 1.414 18.97 1.54E-10 1.14E-10 5.78E-11 4.62E- 11 3.49E-11 20 1.420 20.39 1.35E-10 1.01E-10 5.42E-11 4.47E- 11 3.53E-11 21 1.389 21.78 1.18E-10 9.07E-11 5.35E-11 4.43E-I1 3.42E-11 Analysis of H.B. Robinson Unit 2 Capsule X

6-18 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Iron Atom Displacements Cycle Total Displacements Length Irradiation [dpa]

Cycle [EFPY] Time 0 Degrees 10 Degrees 20 Degrees 30 Degrees 40 Degrees

[EFPY]

1 1.279 1.28 1.17E-02 8.19E-03 3.53E-03 2.71E-03 1.86E-03 2 0.814 2.09 1.92E-02 1.36E-02 5.91E-03 4.56E-03 3.09E-03 3 1.091 3.18 2.76E-02 1.95E-02 8.70E-03 6.81E-03 4.64E-03 4 0.791 3.97 3.58E-02 2.52E-02 1.11E-02 8.61E-03 5.87E-03 5 0.812 4.79 4.34E-02 3.07E-02 1.34E-02 1.04E-02 7.05E-03 6 0.840 5.63 5.11E-02 3.62E-02 1.57E-02 1.22E-02 8.24E-03 7 0.810 6.44 5.82E-02 4.13E-02 1.81E-02 1.40E-02 9.41E-03 8 0.835 7.27 6.56E-02 4.66E-02 2.04E-02 1.58E-02 1.06E-02 9 0.857 8.13 6.94E-02 4.94E-02 2.22E-02 1.74E-02 1.17E-02 10 0.871 9.00 7.43E-02 5.30E-02 2.46E-02 1.94E-02 1.29E-02 11 0.916 9.92 7.89E-02 5.67E-02 2.73E-02 2.17E-02 1.41E-02 12 0.982 10.90 8.35E-02 6.06E-02 3.03E-02 2.42E-02 1.54E-02 13 0.983 11.88 8.75E-02 6.39E-02 3.32E-02 2.66E-02 1.67E-02 14 0.996 12.88 9.16E-02 6.74E-02 3.63E-02 2.93E-02 1.81E-02 15 1.074 13.95 9.56E-02 7.1OE-02 3.96E-02 3.21E-02 1.95E-02 16 1.076 15.03 1.03E-01 7.63E-02 4.18E-02 3.37E-02 2.08E-02 17 1.185 16.21 1.1OE-01 8.14E-02 4.41E-02 3.54E-02 2.21E-02 18 1.346 17.56 1.17E-01 8.68E-02 4.67E-02 3.74E-02 2.34E-02 19 1.414 18.97 1.24E-01 9.19E-02 4.93E-02 3.94E-02 2.49E-02 20 1.420 20.39 1.30E-01 9.64E-02 5.17E-02 4.14E-02 2.65E-02 21 1.389 21.78 1.35E-01 1.OOE-01 5.41E-02 4.34E-02 2.80E-02 Analysis of H.B. Robinson Unit 2 Capsule X

6-19 Table 6-2 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Total Neutron Flux (E > 1.0 MeV) [n/cm 2-s]

Cycle Irradiation Length Time 0 Degrees 10 20 30 40 45 Cycle [EFPY] [EFPY] Degrees Degrees Degrees Degrees Degrees 1 1.279 1.28 6.23E+10 4.38E+10 2.16E+10 1.62E+10 1.17E+10 1.11E+10 2 0.814 2.09 6.07E+10 4.40E+10 2.23E+10 1.68E+10 1.19E+10 1.12E+10 3 1.091 3.18 5.14E+10 3.63E+10 1.94E+10 1.53E+10 1.12E+10 1.06E+10 4 0.791 3.97 6.96E+10 4.92E+10 2.32E+10 1.72E+10 1.25E+10 1.19E+10 5 0.812 4.79 6.37E+10 4.56E+10 2.23E+10 1.65E+10 1.16E+10 1.09E+10 6 0.840 5.63 6.16E+10 4.39E+10 2.17E+10 1.63E+10 1.14E+10 1.07E+10 7 0.810 6.44 5.92E+10 4.28E+10 2.21E+10 1.68E+10 1.16E+10 1.08E+10 8 0.835 7.27 5.90E+10 4.21E+10 2.09E+10 1.56E+10 1.08E+10 1.02E+10 9 0.857 8.13 2.92E+10 2.18E+10 1.55E+10 1.37E+10 9.90E+09 9.32E+09 10 0.871 9.00 3.76E+10 2.89E+10 2.12E+10 1.75E+10 1.11E+10 1.02E+10 11 0.916 9.92 3.44E+10 2.77E+10 2.21E+10 1.81E+10 1.08E+10 9.71E+09 12 0.982 10.90 3.25E+10 2.70E+10 2.30E+10 1.87E+10 1.08E+10 9.63E+09 13 0.983 11.88 2.81E+10 2.39E+10 2.16E+10 1.82E+10 1.08E+10 9.75E+09 14 0.996 12.88 2.78E+10 2.44E+10 2.30E+10 1.92E+10 1.10E+10 9.84E+09 15 1.074 13.95 2.63E+10 2.34E+10 2.29E+10 1.91E+10 1.08E+10 9.54E+09 16 1.076 15.03 4.61E+10 3.30E+10 1.61E+10 1.19E+10 9.58E+09 9.27E+09 17 1.185 16.21 4.07E+10 2.97E+10 1.53E+10 1.09E+10 8.20E+09 7.88E+09 18 1.346 17.56 3.69E+10 2.71E+10 1.48E+10 1.10E+10 7.85E+09 7.32E+09 19 1.414 18.97 3.32E+10 2.46E+10 1.40E+10 1.10E+10 8.74E+09 8.42E+09 20 1.420 20.39 2.93E+10 2.21E+10 1.33E+10 1.08E+10 8.89E+09 8.63E+09 21 1.389 21.78 2.56E+10 1.98E+10 1.29E+10 1.06E+10 8.56E+09 8.27E+09 Analysis of H.B. Robinson Unit 2 Capsule X

6-20 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface I

Total Neutron Fluence (E > 1.0 MeV) [n/cm 21 Cycle Irradiation Length Time 0 Degrees 10 20 30 40 45 Cycle [EFPY] [EFPY] Degrees Degrees Degrees Degrees Degrees 1 1.279 1.28 2.52E+18 1.77E+18 8.71E+17 6.53E+17 4.73E+17 4.49E+17 2 0.814 2.09 4.07E+18 2.90E+18 1.44E+18 1.09E+18 7.78E+17 7.35E+17 3 1.091 3.18 5.84E+18 4.15E+18 2.1lE+18 1.61E+18 1.16E+18 1.10E+18 4 0.791 3.97 7.55E+18 5.36E+18 2.68E+18 2.03E+18 1.47E+18 1.39E+18 5 0.812 4.79 9.17E+18 6.52E+18 3.25E+18 2.46E+18 1.77E+18 1.67E+18 6 0.840 5.63 1.08E+19 7.67E+18 3.82E+18 2.88E+18 2.07E+18 1.95E+18 7 0.810 6.44 1.23E+19 8.76E+18 4.38E+18 3.31E+18 2.36E+18 2.23E+18 8 0.835 7.27 1.38E+19 9.86E+18 4.93E+18 3.72E+18 2.64E+18 2.49E+18 9 0.857 8.13 1.46E+19 1.05E+19 5.35E+18 4.09E+18 2.91E+18 2.74E+18 10 0.871 9.00 1.56E+19 1.12E+19 5.92E+18 4.56E+ 18 3.21E+18 3.02E+18 11 0.916 9.92 1.66E+19 1.20E+19 6.55E+18 5.08E+18 3.52E+18 3.30E+18 12 0.982 10.90 1.76E+19 1.28E+19 7.26E+18 5.66E+18 3.86E+18 3.60E+18 13 0.983 11.88 1.84E+19 1.35E+19 7.92E+18 6.22E+18 4.19E+18 3.90E+18 14 0.996 12.88 1.93E+19 1.43E+19 8.65E+18 6.83E+18 4.54E+18 4.21E+18 15 1.074 13.95 2.01E+19 1.51E+19 9.42E+18 7.47E+-18 4.90E+18 4.53E+18 16 1.076 15.03 2.17E+19 1.62E+19 9.94E+18 7.87E+18 5.22E+18 4.84E+18 17 1.185 16.21 2.32E+ 19 1.73E+19 1.05E+19 8.27E+18 5.52E+18 5.13E+18 18 1.346 17.56 2.48E+19 1.84E+19 1.11E+19 8.73E+18 5.85E+18 5.44E+18 19 1.414 18.97 2.63E+19 1.95E+19 1.18E+19 9.22E+18 6.24E+18 5.81E+18 20 1.420 20.39 2.76E+ 19 2.05E+19 1.24E+19 9.69E+18 6.63E+18 6.19E+18 21 1.389 21.78 2.87E+19 2.14E+19 1.29E+19 1.02E+19 7.01E+18 6.56E+18 Note: At the end of Cycle 20, the maximum fast (E > 1.0 MeV) neutron fluence at the pressure vessel wall occurs at an axial elevation about 10 cm above the midplane of the active fuel for the 00, 100, and 20' azimuths and at about 37 cm below the midplane of the active fuel for the 30', 400 and 45' azimuths.

However, the axial distribution is quite flat around axial midplane and the maximum fluence value is within

-1% of the midplane value.

Analysis of H.B. Robinson Unit 2 Capsule X

6-21 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Total Iron Atom Displacement Rate [dpa/s]

Cycle Irradiation Length Time 0 10 20 30 40 45 Cycle [EFPY] [EFPY] Degrees Degrees Degrees Degrees Degrees Degrees 1 1.279 1.28 1.03E-10 7.32E-11 3.56E-11 2.65E-11 1.92E-11 1.83E-11 2 0.814 2.09 1.00E-10 7.34E-11 3.67E-11 2.75E-11 1.94E-11 1.84E-11 3 1.091 3.18 8.49E-11 6.06E-11 3.19E-11 2.50E-11 1.82E-11 1.74E-11 4 0.791 3.97 1.15E-10 8.21E-11 3.83E-11 2.82E-11 2.04E-11 1.95E-11 5 0.812 4.79 1.05E-10 7.62E-11 3.67E-11 2.70E-11 1.89E-11 1.79E-11 6 0.840 5.63 1.02E-10 7.33E-11 3.58E-11 2.66E-11 1.87E-11 1.76E-11 7 0.810 6.44 9.78E-11 7.15E-11 3.65E-11 2.74E-11 1.89E-11 1.78E-11 8 0.835 7.27 9.73E-11 7.02E-11 3.44E-11 2.55E-11 1.77E-11 1.67E-11 9 0.857 8.13 4.76E-11 3.59E-11 2.52E-11 2.22E-11 1.61E-11 1.52E-11 10 0.871 9.00 6.18E-11 4.80E-11 3.47E-11 2.86E-11 1.81E-11 1.68E-11 11 0.916 9.92 5.67E-11 4.59E-11 3.61E-11 2.97E-11 1.76E-11 1.60E-11 12 0.982 10.90 5.34E-11 4.47E-11 3.75E-11 3.06E-11 1.77E-11 1.59E-11 13 0.983 11.88 4.61E-11 3.95E-11 3.52E-11 2.97E-11 1.77E-11 1.61E-11 14 0.996 12.88 4.58E-11 4.04E-11 3.76E-11 3.14E-11 1.81E-11 1.63E-11 15 1.074 13.95 4.33E-11 3.88E-11 3.72E-11 3.11E-11 1.76E-11 1.58E-11 16 1.076 15.03 7.60E-11 5.49E-11 2.64E-11 1.94E-11 1.56E-11 1.52E-11 17 1.185 16.21 6.70E-11 4.94E-11 2.51E-11 1.78E-11 1.34E-11 1.29E-11 18 1.346 17.56 6.06E-11 4.50E-11 2.43E-11 1.79E-11 1.28E-11 1.20E-11 19 1.414 18.97 5.45E-11 4.08E-11 2.30E-11 1.80E-11 1.42E-I1 1.38E-11 20 1.420 20.39 4.81E-11 3.66E-11 2.17E-11 1.76E-11 1.45E-11 1.41E-11 21 1.389 21.78 4.20E-11 3.28E-11 2.12E-11 1.73E-11 1.39E-11 1.36E-11 Analysis of H.B. Robinson Unit 2 Capsule X

6-22 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Total Iron Atom Displacements [d pa]

Cycle Irradiation Length Time 0 10 20 30 40 45 Cycle [EFPY] [EFPY] Degrees Degrees Degrees Degrees Degrees Degrees 1 1.279 1.28 4.16E-03 2.95E-03 1.44E-03 1.07E-03 7.73E-04 7.37E-04 2 0.814 2.09 6.73E-03 4.84E-03 2.38E-03 1.78E-03 1.27E-03 1.21E-03 3 1.091 3.18 9.65E-03 6.92E-03 3.48E-03 2.64E-03 1.90E-03 1.81E-03 4 0.791 3.97 1.25E-02 8.94E-03 4.42E-03 3.33E-03 2.40E-03 2.29E-03 5 0.812 4.79 1.52E-02 1.09E-02 5.36E-03 4.02E-03 2.88E-03 2.74E-03 6 0.840 5.63 1.78E-02 1.28E-02 6.30E-03 4.72E-03 3.37E-03 3.21E-03 7 0.810 6.44 2.03E-02 1.46E-02 7.23E-03 5.41E-03 3.86E-03 3.66E-03 8 0.835 7.27 2.29E-02 1.65E-02 8.13E-03 6.0*8E-03 4.32E-03 4. 1OE-03 9 0.857 8.13 2.42E-02 8E-03 4.75E-03 4.5 1E-03 10 0.871 9.00 2.58E-02 1.87E-02 9.75E-03 7.4( 5E-03 5.25E-03 4.96E-03 11 0.916 9.92 2.74E-02 2.OOE-02 1.08E-02 8.322E-03 5.75E-03 5.43E-03 12 0.982 10.90 2.90E-02 2.13E-02 1.19E-02 9.26 E-03 6.30E-03 5.92E-03 13 0.983 11.88 3.03E-02 2.25E-02 1.30E-02 1.02 E-02 6.85E-03 6.42E-03 14 0.996 12.88 3.18E-02 2.38E-02 1.42E-02 1.12 E-02 7.42E-03 6.93E-03 15 1.074 13.95 3.32E-02 2.51E-02 1.54E-02 1.22 .E-02 8.01E-03 7.46E-03 16 1.076 15.03 3.58E-02 2.70E-02 1.63E-02 1.29 E-02 8.53E-03 7.96E-03 17 1.185 16.21 3.83E-02 2.88E-02 1.72E-02 1.35 E-02 9.02E-03 8.44E-03 18 1.346 17.56 4.09E-02 3.07E-02 1.83E-02 1.43 E-02 9.56E-03 8.95E-03 19 1.414 18.97 4.33E-02 3.25E-02 1.93E-02 1.51 1.93E-02 15 E-02 1.02E-02 9.56E-03 20 1.420 20.39 4.54E-02 3.42E-02 2.03E-02 1*

t I. 2.03 E-02 15 E-02 1.08E-02 1.02E-02 3 .42E-02 . I ,

21 1.389 21.78 4.73E-02 3.56E-02 2.12E-02 1.66I 1.14E-02 1.08E-02 2E02

2. 1.66,~02 Note: At the end of Cycle 20, the maximum dpa at the pressure vessel wall occurs at an axial elevation about 10 cm above the midplane of the active fuel for the 00, 100, and 200 azimuths and at about 37 cm below the midplane of the active fuel for the 30', 40° and 450 azimuths. However, the axial distribution is quite flat around axial midplane and the maximum dpa value is within -1% of the midplane value.

Analysis of H.B. Robinson Unit 2 Capsule X

6-23 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 100 200 300 400 450 198.039 1.000 1.000 1.000 1.000 1.000 1.000 203.943 0.516 0.520 0.528 0.524 0.531 0.520 209.846 0.234 0.238 0.244 0.241 0.246 0.236 215.750 0.101 0.104 0.108 0.106 0.109 0.104 221.653 0.039 0.042 0.046 0.045 0.047 0.044 Note: Base Metal Inner Radius = 198.039 Base Metal 1/4T = 203.943 Base Metal 1/2T = 209.846 Base Metal 3/4T 215.750 Base Metal Outer Radius = 221.653 Table 6-4 Relative Radial Distribution Of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 100 200 300 400 450 198.039 1.000 1.000 1.000 1.000 1.000 1.000 203.943 0.629 0.639 0.651 0.640 0.648 0.636 209.846 0.378 0.391 0.406 0.393 0.401 0.389 215.750 0.217 0.232 0.248 0.237 0.243 0.233 221.653 0.104 0.119 0.139 0.133 0.137 0.131 Note: Base Metal Inner Radius = 198.039 Base Metal 1/4T = 203.943 Base Metal 1/2T = 209.846 Base Metal 3/4T = 215.750 Base Metal Outer Radius = 221.653 Analysis of H.B. Robinson Unit 2 Capsule X

6-24 Table 6-5 Nuclear Parameters Used In The Evaluation Of Neutron Sensors Target 90% Response Fission Monitor Reaction of Atom Range Product Yield Material Interest Fraction (MeV) Half-life Copper 63 Cu (n,a) 0.6917 4.8- 11.7 5.271 y 54 Iron Fe (n,p) 0.0585 2.1 -8.3 312.3 d Nickel 58Ni 0.6808 1.6-8.1 (np) 70.82 d 238 Uranium-238 U (n,f) 0.9996 1.3-6.8 30.07 y 6.02 237 Neptunium-237 Np (n,f) 1.0000 0.4-4.3 3 0.0 7 y 6.17 59 Cobalt-Aluminum Co (n,y) 0.0015 non-threshold 5.271 y Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the H. B. Robinson Unit 2 surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified, with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.

Analysis of H.B. Robinson Unit 2 Capsule X

6-25 Table 6-6 Monthly Thermal Generation During The First 20 Fuel Cycles Of The H. B. Robinson Unit 2 Reactor (Reactor Power of 2300 MWt)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1970 10 17339 1973 5 325257 1975 848232 1970 11 17339 1973 6 1315814 1976 1356960 1970 12 17339 1973 7 1379687 1976 1519690 1971 1 315994 1973 8 1498106 1976 1593451 1971 2 315994 1973 9 1490895 1976 1513829 1971 3 315994 1973 10 1517185 1976 1425283 1971 4 0 1973 11 1144053 1976 1574338 1971 5 500000 1973 12 1457592 1976 1515096 1971 6 0 1974 1 1504428 1976 1570061 1971 7 0 1974 2 1504428 1976 1554326 1971 8 0 1974 3 1504428 1976 1470691 1971 9 1500000 1974 4 1539067 1976 0 1971 10 1634231 1974 5 237864 1976 772992 1971 11 1486726 1974 6 125088 1977 1586376 1971 12 1781736 1974 7 1547784 1977 775685 1972 1 1497683 1974 8 1499250 1977 1468421 1972 2 1536043 1974 9 1447142 1977 1247136 1972 3 1637212 1974 10 1462454 1977 1618214 1972 4 1565161 1974 11 1554379 1977 1514674 1972 5 294474 1974 12 1625232 1977 1492603 1972 6 962248 1975 1 1494979 1977 1214347 1972 7 1190975 1975 2 1467416 1977 764597 1972 8 1440808 1975 3 1562880 1977 819614 1972 9 1440195 1975 4 577104 1977 205392 1972 10 1537511 1975 5 220968 1977 1571170 1972 11 1348800 1975 6 1259069 1978 1314192 1972 12 1097499 1975 7 1476658 1978 0 1973 1 908362 1975 8 1616842 1978 0 1973 2 806836 1975 9 1493870 1978 157872 1973 3 611522 1975 10 1570219 1978 1528032 1973 4 0 1975 11 0 1978 1571645 Analysis of H.B. Robinson Unit 2 Capsule X

6-26 Table 6-6 Cont'd Monthly Thermal Generation During The First 20 Fuel Cycles Of The H. B. Robinson Unit 2 Reactor (Reactor Power of 2300 MWt)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1978 7 1280875 1981 2 1279260 1983 9 557630 1978 8 1607654 1981 3 1604554 1983 10 1342574 1978 9 1124904 1981 4 1545048 1983 11 89921 1978 10 1616050 1981 5 795101 1983 12 674378 1978 11 1559923 1981 6 877625 1984 1 783840 1978 12 1552214 1981 7 1523299 1984 2 0 1979 1 1503322 1981 8 0 1984 3 0 1979 2 1335576 1981 9 733663 1984 4 0 1979 3 1603906 1981 10 843842 1984 5 0 1979 4 490565 1981 11 182326 1984 6 0 1979 5 0 1981 12 1039692 1984 7 0 1979 6 0 1982 1 1361232 1984 8 0 1979 7 374532 1982 2 1140046 1984 9 0 1979 8 1635466 1982 3 0 1984 10 0 1979 9 1482838 1982 4 0 1984 11 0 1979 10 1556695 1982 5 0 1984 12 0 1979 11 1518386 1982 6 0 1985 1 139435 1979 12 1555205 1982 7 0 1985 2 779534 1980 1 1593955 1982 8 245143 1985 3 1537210 1980 2 1475827 1982 9 1029149 1985 4 1546428 1980 3 875306 1982 10 1331038 1985 5 1614490 1980 4 641866 1982 11 1265350 1985 6 1631878 1980 5 1070935 1982 12 1297145 1985 7 1613938 1980 6 1486150 1983 1 1227869 1985 8 1552003 1980 7 870228 1983 2 1141757 1985 9 1454410 1980 8 193421 1983 3 1309068 1985 10 1534891 1980 9 0 1983 4 943313 1985 11 1579162 1980 10 177910 1983 5 148543 1985 12 1633037 1980 11 1359907 1983 6 1271311 1986 1 952586 1980 12 916210 1983 7 1289251 1986 2 0 1981 1 1458329 1983 8 1330817 1986 3 235042 X

2 Capsule Unit Robinson ll.B.

of Analysis Analysis of H.B. Robinson Unit 2 Capsule X

6-27 Table 6-6 Cont'd Monthly Thermal Generation During The First 20 Fuel Cycles Of The H. B. Robinson Unit 2 Reactor (Reactor Power of 2300 MWt)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1986 4 1575463 1988 11 548743 1991 6 1648658 1986 5 1659478 1988 12 0 1991 7 1688126 1986 6 1537099 1989 1 0 1991 8 1514026 1986 7 1693481 1989 2 48962 1991 9 1209377 1986 8 1315582 1989 3 1523189 1991 10 1707888 1986 9 1344727 1989 4 921674 1991 11 1556198 1986 10 1695634 1989 5 1670738 1991 12 1687850 1986 11 1623266 1989 6 1640710 1992 1 1701650 1986 12 1557413 1989 7 1696130 1992 2 1596936 1987 1 1694474 1989 8 1153790 1992 3 1443259 1987 2 1508395 1989 9 0 1992 4 0 1987 3 1030087 1989 10 0 1992 5 0 1987 4 0 1989 11 0 1992 6 216605 1987 5 0 1989 12 406879 1992 7 1521367 1987 6 596657 1990 1 1635576 1992 8 1170295 1987 7 1378289 1990 2 1460758 1992 9 332138 1987 8 1268110 1990 3 1038809 1992 10 1664004 1987 9 1064311 1990 4 .1289914 1992 11 1574194 1987 10 1676369 1990 5 782846 1992 12 1700767 1987 11 1559897 1990 6 1095941 1993 1 1703969 1987 12 1693757 1990 7 1518110 1993 2 1537817 1988 1 1426699 1990 8 1615483 1993 3 1699663 1988 2 0 1990 9 323914 1993 4 1644463 1988 3 675427 1990 10 0 1993 5 1562215 1988 4 947950 1990 11 0 1993 6 1634251 1988 5 796370 1990 12 0 1993 7 1429901 1988 6 1180342 1991 1 0 1993 8 1705349 1988 7 1698062 1991 2 0 1993 9 549792 1988 8 1634858 1991 3 1063373 1993 10 0 1988 9 402353 1991 4 1638281 1993 11 46478 1988 10 1245809 1991 5 1626468 1993 12 0 X

2 Capsule Unit Robinson of ll.B.

Analysis Analysis of H.B. Robinson Unit 2 Capsule X

6-28 6-28 Table 6-6 Cont'd Monthly Thermal Generation During The First 20 Fuel Cycles Of The H. B. Robinson Unit 2 Reactor (Reactor Power of 2300 MWt)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1994 1 0 1996 8 1676976 1999 3 1710206 1994 2 98532 1996 9 350686 1999 4 1648769 1994 3 422390 1996 10 444967 1999 5 1703362 1994 4 1516454 1996 11 1654896 1999 6 1651363 1994 5 1679846 1996 12 1709930 1999 7 1710317 1994 6 1651032 1997 1 1703858 1999 8 1694585 1994 7 1699111 1997 2 1544220 1999 9 1127681 1994 8 1435145 1997 3 1667978 1999 10 359518 1994 9 1652081 1997 4 1622494 1999 11 1654730 1994 10 1699663 1997 5 1710262 1999 12 1710648 1994 11 1633865 1997 6 1643304 2000 1 1701264 1994 12 1697400 1997 7 1700878 2000 2 1600138 1995 1 1697897 1997 8 1710262 2000 3 1690390 1995 2 1535057 1997 9 1642531 2000 4 1615594 1995 3 1696793 1997 10 1710648 2000 5 1710317 1995 4 1530641 1997 11 1391095 2000 6 1535002 1995 5 0 1997 12 1710372 2000 7 1710372 1995 6 371054 1998 1 1710372 2000 8 1710427 1995 7 1423056 1998 2 1544827 2000 9 1652798 1995 8 1705956 1998 3 318946 2000 10 1704852 1995 9 1618961 1998 4 718980 2000 11 1653350 1995 10 1708054 1998 5 1710482 2000 12 1710096 1995 11 1650259 1998 6 1624094 2001 1 1706508 1995 12 1700105 1998 7 1679902 2001 2 1533346 1996 1 1687133 1998 8 1696351 2001 3 1433875 1996 2 1596660 1998 9 1655172 2001 4 236808 1996 3 1698338 1998 10 1613330 1996 4 1649431 1998 11 1655172 1996 5 1705294 1998 12 1710538 1996 6 1641703 1999 1 1642973 1996 7 1696130 1999 2 1544827 2Casue ni obnsn f l..

Anayss Analysis of H.B. Robinson Unit 2 Capsule X

6-29 Table 6-7 Calculated 4(E > 1.0 MeV) and Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Capsules S, V, and T Fuel 4E > 1.0 MeV) [ncm 2-s] C Cycle Capsule S Capsule V Capsule T S V T 1 1.19E+1l 5.33E+10 1.71E+11 1.000 1.010 1.016 2 5.65E+10 1.70E+ 11 1.067 1.010 3 4.94E+10 1.45E+11 0.938 0.862 4 1.92E+11 1.137 5 1.77E+11 1.046 6 1.70E+l 1 1.007 7 1.64E+11 0.971 8 1.67E+11 0.993 Average 1.19E+11 5.28E+10 1.69E+11 1.000 1.000 1.000 CapsuleX Fuel p(E>1.0 MeV) Relative Reacti Rate (Cj)

Cycle [n/cm 2 _s] Cu(n,oc) Fe(np) Ni(np) 238 U(n,f) 237 Np(n,f) Co(n,y)(Cd) Co(n,y) 1 2.84E+10 0.445 0.369 0.361 0.324 0.290 0.245 0.252 2 2.94E+10 0.461 0.382 0.374 0.335 0.301 0.254 0.261 3 2.78E+10 0.437 0.361 0.354 0.317 0.284 0.240 0.246 4 3.05E+10 0.478 0.396 0.388 0.348 0.312 0.263 0.271 5 2.83E+10 0.445 0.367 0.360 0.322 0.289 0.244 0.251 6 2.77E+ 10 0.437 0.360 0.353 0.316 0.283 0.239 0.246 7 2.82E+10 0.443 0.367 0.359 0.322 0.288 0.244 0.251 8 2.69E+10 0.429 0.352 0.345 0.308 0.275 0.231 0.238 9 8.35E+10 0.921 0.913 0.911 0.904 0.893 0.876 0.882 10 1.05E+1 1 1.088 1.110 1.111 1.121 1.128 1.138 1.136 11 9.51E+10 1.013 1.020 1.020 1.023 1.026 1.029 1.029 12 8.93E+10 0.971 0.966 0.965 0.963 0.961 0.960 0.960 13 7.67E+10 0.851 0.836 0.835 0.829 0.825 0.821 0.821 14 7.60E+10 0.850 0.831 0.829 0.822 0.816 0.811 0.812 15 7.16E+10 0.810 0.786 0.784 0.775 0.769 0.762 0.763 16 1.28E+11 1.299 1.347 1.351 1.372 1.391 1.412 1.409 17 1.13E+l 1 1.179 1.201 1.203 1.212 1.220 1.230 1.228 18 1.02E+11 1.092 1.097 1.097 1.098 1.100 1.102 1.101 19 9.17E+10 1.002 0.994 0.994 0.990 0.987 0.983 0.983 20 8.OOE+10 0.891 0.874 0.873 0.866 0.860 0.853 0.852 Analysis of H.B. Robinson Unit 2 Capsule X

6-30 Table 6-8 Measured Sensor Activities And Reaction Rates Surveillance Capsule S Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Axial Radial Activity Activity Activity Rate Reaction Location Location (dps/g) (dps/g) (dps/g) (rps/atom) 63Cu(n,a) 6°Co Top 190.923 1.66E+05 1.11E+06 Not used Bottom 190.923 8.34E+04 5.57E+05 5.32E+05 8.11E- 17 Average 8.11E-17 4

' Fe (n,p) 54Mn Top 191.923 2.60E+06 4.79E+06 5.55E+06 8.87E-15 Top-Mid 190.923 2.70E+06 4.97E+06 4.74E+06 7.57E-15 Middle 191.923 2.62E+06 4.82E+06 5.59E+06 8.94E-15 Bot-Mid 190.923 3.00E+06 5.52E+06 5.26E+06 8.41E-15 Bottom 191.923 2.50E+06 4.60E+06 5.33E+06 8.53E-15 Average 8.47E-15 58Ni (n,p) 58Co Middle 190.923 4.74E+07 7.18E+07 6.84E+07 9.77E-15 Average 9.77E-15 59 60Co co (n,,y) Top 191.923 5.84E+06 3.90E+07 4.49E+07 2.93E-12 (Cd) Middle 191.923 5.90E+06 3.94E+07 4.53E+07 2.96E-12 Bottom 191.923 7.44E+06 4.97E+07 5.71E+07 3.73E-12 Average 3.20E-12 "95Co (n,y) 60Co Top 191.923 1.83E+07 1.22E+08 1.19E+08 7.75E-12 Bottom 191.923 2.06E+07 1.37E+08 1.33E+08 8.70E-12 Average 8.22E-12 Note: Measured specific activities'are corrected to the end of irradiation date of March 16, 1973.

Analysis of H.B. Robinson Unit 2 Capsule X

6-31 Table 6-8 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule V Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Axial Radial Activity Activity Activity Rate Reaction Location Location (dps/g) (dps/g) (dps/g) (rps/atom) 63 60 Cu (n,cx) Co Top 190.923 1.18E+05 3.70E+05 3.54E+05 5.40E-17 Bottom 190.923 1.12E+05 3.52E+05 3.36E+05 5.12E- 17 Average 5.26E-17 4

Fe (n,p) 54Mn Top 191.923 1.94E+06 2.72E+06 3.13E+06 5.01E-15 Top-Mid 190.923 2.24E+06 3.14E+06 3.OOE+06 4.79E-15 Middle 191.923 2.12E+06 2.97E+06 3.43E+06 5.48E-15 Bot-Mid 190.923 2.08E+06 2.91E+06 2.78E+/-06 4.44E-15 Bottom 191.923 2.34E+06 3.28E+06 3.78E+06 6.04E-15 Average 5.15E-15 58 Ni (np) 58Co Middle 190.923 4.19E+07 5.40E+07 5.15E+07 7.36E-15 238 U (n,f) 137Cs (Cd) 3.95E+08 Not Used Middle 191.155 2.75E+07 237 137 Np (n,f) Cs (Cd) Not Used Middle 191.155 1.83E+07 2.63E+08 59 Co (n,'y) 6 0Co (Cd) Top 191.923 3.23E+06 1.01E+07 1.14E+07 7.43E-13 Middle 191.923 4.83E+06 1.52E+07 1.71E+07 1.11E-12 Bottom 191.923 6.90E+06 2.17E+07 2.44E+07 1.59E- 12 Average 1.15E-12 59 Co (n,,Y) 60 Co Top 191.923 1.68E+07 5.27E+07 5.01E+07 3.27E- 12 Middle 191.923 1.61E+07 5.04E+07 4.79E+07 3.12E-12 Bottom 191.923 1.62E+07 5.08E+07 4.83E+07 3.15 E- 12 Average 3.18E-12 Note: Measured specific activities are corrected to the end of irradiation date of October 31, 1975.

Analysis of H.B. Robinson Unit 2 Capsule X

6-32 Table 6-8 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule T Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Axial Radial Activity Activity Activity Rate Reaction Location Location (dps/g) (dps/g) (dps/g) (rps/atom) 63 Cu (n,ax) 6°Co Top 190.923 3.56E+05 7.76E+05 7.40E+05 1.13E-16 Bottom 190.923 3.72E+05 8.1OE+05 7.73E+05 1.18E-16 Average 1.15E-16 54 Fe (n,p) 54Mn Top 190.923 3.10E+06 8.70E+06 8.29E+06 1.32E-14 Top 191.923 2.58E+06 7.11E-I+06 8.26E+06 1.32E-14 Middle 190.923 3.06E+06 8.58E+06 8.17E+06 1.31E-14 Middle 191.923 2.56E+06 7.19E+06 8.34E+06 1.33E-14 Bottom 190.923 3.35E+06 9.41E+06 8.96E+06 1.43E-14 Bottom 191.923 2.58E+06 7.24E+06 8.41E+06 1.34E-14 Average 1.34E-14 58Ni (n,p) 58Co Middle 190.923 8.53E+06 1.28E+08 1.22E+08 1.74E-14 23 8U (n,f) 137 Cs (Cd) Middle 191.155 2.37E+06 1.63E+07 1.17E+07 7.68E-14 237 Np (n,f) 137Cs (Cd) Middle 191.155 1.25E+07 8.58E+07 8.45E+07 5.30E-13 59Co (ny) 6°Co (Cd) Top 191.923 2.54E+07 5.52E+07 6.35E+07 4.14E-12 Middle 191.923 2.73E+07 5.94E+07 6.82E+07 4.45E-12 Bottom 191.923 2.76E+07 6.OOE+07 6.90E+07 4.50E-12 Average 4.36E-12 59 Co (n,,y) 60Co Top 191.923 5.61E+07 1.22E+08 1.19E+08 7.76E-12 Bottom 191.923 6.71E+07 1.46E+08 1.42E+08 9.29E-12 Average 8.53E-12 Note: Measured specific activities are corrected to a counting date of October 8, 1982.

Analysis of H.B. Robinson Unit 2 Capsule X

6-33 Table 6-8 cont'd Measured Sensor Activities And Reaction Rates a

Surveillance Capsule X Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Axial Radial Activity Activity Activity Rate Reaction Location Location (dps/g) (dps/g) (dps/g) (rps/atom) 63Cu (n,a) 60Co Top 190.923 3.3 1E+05 4.60E+05 4.39E+05 6.70E-17 Bottom 190.923 3.15E+05 4.38E+05 4.18E+05 6.38E-17 Average 6.54E-17 54 Fe (n,p) 54Mn Middle 190.923 2.67E+06 4.28E+06 4.08E+06 6.48E-15 Middle 191.923 2.23E+06 3.58E+06 4.15E+06 6.58E-15 Average 6.53E-15 58 Ni (np) 58Co Middle 190.923 1.45E+07 7.54E+07 7.18E+07 1.03E-14 2 38 137 Cs U (n,f) (Cd) Middle 191.155 2.01E+06 7.15E+06 4.97E+06 3.27E-14 237Np (n,f) 137Cs (Cd) Middle 191.155 1.25E+07 4.50E+07 4.43E+07 2.83E-13 59Co (n,'y) 60Co (Cd) Top 191.923 2.09E+07 2.94E+07 3.37E+07 2.20E-12 Bottom 191.923 1.94E+07 2.73E+07 3.13E+07 2.04E-12 Average 2.12E-12 59Co (ny) 60Co Top 191.923 5.49E+07 7.73E+07 7.53E+07 4.9 1E-12 Bottom 191.923 5.33E+07 7.50E+07 7.31E+07 4.77E-12 Average 4.84E-12 Note: Measured specific activities are corrected to a counting date of September 1, 2001.

a. Saturated activities and reaction rates for capsule X are determined for the 0' location averaged over fuel cycles 9 through 20.

Analysis of H.B. Robinson Unit 2 Capsule X

6-34 Table 6-9 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule S Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 6"3Cu(noC)60Co 8.11E-17 6.59E-17 7.81E-17 1.23 0.96 54 Fe(n,p) 54Mn 8.47E-15 7.92E-15 8.66E-15 1.07 1.02 58Ni(n,p) 58 Co 9.77E-15 1.1OE-14 1.20E-14 0.89 1.22 59Co(n,y)6°Co 8.22E-12 6.19E-12 8.16E-12 1.33 0.99 59Co(n,y)6°Co (Cd) 3.20E-12 2.88E-12 3.21E-12 1.11 1.00 Capsule V Reaction Rate [rps/atom]

Best Reaction Measured Calculated M/C M/BE 63 I Estimate C Cu(n,a)6°Co 5.26E-17 4.05E-17 5.20E-17 1.30 0.99 54Fe(n,p) 54 Mn 5.15E-15 4.15E-15 5.25E-15 1.24 1.02 58 Ni(n,p) 58Co 7.36E-15 5.65E-15 7.22E-15 1.30 0.98 59 Co(n,y)60Co 3.18E-12 2.27E- 12 3.16E- 12 1.40 0.99 59 Co(n,Y) 60 Co (Cd) 1.11 1.15E-12 1.03E-12 1.15E-12 1.00

__ _ _ _ _I_ _ _ _ _ I_ _ _ _ __I_ _ _ _ _

Analysis of H.B. Robinson Unit 2 Capsule X

6-35 Table 6-9 cont'd Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule T Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 6 3Cu(n,a)6°Co 1.15E-16 9.13E-17 1.12E-16 1.26 0.97 54Fe(np)-4Mn 1.34E-14 1.13E-14 1.34E-14 1.19 1.00 5 8Ni(n,p) 5 8Co 1.74E-14 1.56E-14 1.82E-14 1.12 1.05 238U(n,f) 137Cs (Cd) 7.68E-14 5.80E-14 6.84E-14 1.32 0.89 237Np(n,f) 137 Cs (Cd) 5.30E-13 4.61E-13 5.32E-13 1.15 1.00 59Co(nY)60Co 8.53E-12 8.52E-12 8.57E-12 1.00 1.00 59Co(n,Y) 60 Co (Cd) 4.36E-12 3.97E-12 4.35E-12 1.10 1.00 Capsule X a Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate MIC M/BE 63Cu(n,cc)6°Co 6.54E-17 5.60E-17 6.36E-17 1.17 0.97 54Fe(np)5 4Mn 6.53E-15 6.48E-15 6.93E-15 1.01 1.06 58Ni(np)5 8Co 1.03E-14 8.93E-15 9.86E-15 1.15 0.96 238U(n,f)137 Cs (Cd) 3.26E-14 3.23E-14 3.45E-14 1.01 1.06 237Np(n,f)137 Cs (Cd) 2.82E-13 2.51E-13 2.74E-13 1.13 0.97 59Co(nY)60Co 4.84E-12 4.49E-12 4.83E-12 1.08 1.00 59Co(n,y)6°Co (Cd) 2.12E-12 2.11E-12 2.13E-12 1.01 1.00

a. Reaction rates for capsule X are determined for the 0' location averaged over fuel cycles 9 through 20.

Analysis of H.B. Robinson Unit 2 Capsule X

6-36 Table 6-10 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center O(E > 1.0 MeV) [nrcm 2-s]

Best Estimate Uncertainty Capsule ID Calculated (BE) (ICY) C/BE S 1.19E+ 1I 1.25E+11 8% 0.95 V 5.29E+10 6.56E+10 7% 0.81 T 1.69E+11 1.98E+11 6% 0.85 Xa 9.31E+10 9.88E+10 6% 0.94 Iron Atom Displacement Rate [dpa/s]

Best Estimate Uncertainty Capsule ID Calculated (BE) (1iC) C/BE S 2.01E-10 2.12E-10 9% 0.95 V 8.55E- 11 1.05E-10 8% 0.82 T 2.83E-10 3.27E-10 7% 0.86 Xa 1.55E-10 1.64E-10 7% 0.94

a. Exposure rates for capsule X are determined for the 0' location averaged over fuel cycles 9 through 20.

Analysis of H.B. Robinson Unit 2 Capsule X

6-37 Table 6-11 Comparison of Calculated/Measured (C/M) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions C/M Ratio Reaction Capsule S Capsule V Capsule T Capsule X 63Cu(n,o) 60 Co 0.81 0.77 0.79 0.86 54Fe(n,p)54 Mn 0.94 0.81 0.84 0.99 58Ni(n,p) 58 Co 1.13 0.77 0.90 0.87 23 8 U(n,f)137Cs (Cd) 0.76 0.99 237Np(n,f) 137 Cs (Cd) 0.87 0.89 Average 0.96 0.78 0.83 0.92

% Standard Deviation 16.5 2.7 6.9 7.3 Note: The overall average C/M ratio for the set of sensor measurements is 0.87 with an associated standard deviation of 11.4%.

Table 6-12 Comparison of Calculated/ Best Estimate (C/BE) Exposure Rate Ratios C/BE Ratio Capsule ID O(E > 1.0 MeV) dpals S 0.95 0.95 V 0.81 0.82 T 0.85 0.86 X 0.94 0.94 Average 0.89 0.89

% Standard Deviation 7.9 7.1 Analysis of H.B. Robinson Unit 2 Capsule X

6-38 Table 6-13 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from H. B. Robinson Unit 2 Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule [EFPY] [n/cm 2] [dpa]

S 1.28 4.79E+18 8.19E-03 V 3.18 5.30E+18 8.70E-03 T 7.27 3.87E+19 6.56E-02 X 20.39 4.49E+19 7.52E-02 Analysis of H.B. Robinson Unit 2 Capsule X

6-39 Table 6-14 Calculated Maximum Fast Neutron Exposure of the H. B. Robinson Unit 2 Reactor Pressure Vessel at the Clad/Base Metal Interface at Selected Azimuthal Angles Neutron Fluence [E > 1.0 MeV]

Cumulative Neutron Fluence [n/cm 2]

Operating Time

[e 0.0 10.0 20.0 30.0 40.0 45.0

[EFPY] Degrees Degrees Degrees Degrees Degrees Degrees 20.39 a 2.76E+19 2.05E+19 1.24E+19 9.69E+18 6.63E+18 6.19E+18 21.78 b 2.87E+19 2.14E+19 1.29E+19 1.02E+19 7.01E+18 6.56E+18 29.00 3.67E+19 2.73E+19 1.62E+19 1.27E+19 8.98E+18 8.45E+18 30.00 3.78E+19 2.81E+19 1.67E+19 1.30E+19 9.25E+18 8.71E+18 35.00 4.34E+19 3.22E+19 1.89E+19 1.48E+19 1.06E+19 1.00E+19 40.00 4.89E+19 3.63E+19 2.12E+19 1.65E+19 1.20E+19 1.13E+19 45.00 5.44E+19 4.04E+19 2.35E+19 1.82E+19 1.33E+19 1.26E+19 50.00 6.OOE+19 4.45E+19 2.58E+19 2.OOE+19 1.47E+19 1.40E+19 Iron Atom Displacements Cumulative Iron Atom Displacements [dpa]

Operating Time

[e 0.0 10.0 20.0 30.0 40.0 45.0

[EFPY] Degrees Degrees Degrees Degrees Degrees Degrees 20.39 a 4.79E-02 3.42E-02 2.03E-02 1.58E-02 1.08E-02 1.02E-02 21.78 b 4.99E-02 3.56E-02 2.12E-02 1.66E-02 1.14E-02 1.08E-02 29.00 6.30E-02 4.54E-02 2.66E-02 2.07E-02 1.46E-02 1.39E-02 30.00 6.49E-02 4.68E-02 2.73E-02 2.13E-02 1.51E-02 1.43E-02 35.00 7.40E-02 5.35E-02 3.11E-02 2.41E-02 1.73E-02 1.65E-02 40.00 8.31E-02 6.03E-02 3.48E-02 2.70E-02 1.95E-02 1.86E-02 45.00 9.22E-02 6.71E-02 3.85E-02 2.98E-02 2.18E-02 2.08E-02 50.00 1.01E-01 7.39E-02 4.23E-02 3.26E-02 2.40E-02 2.29E-02

a. EtAY at end ot cycle 20.

b. Estimated EFPY at end of cycle 21.

Note: Up to the end of cycle 21, EFPY is calculated based on a full operating power of 2300 MWt. For projections beyond cycle 21, the projections include a 1.5% plant uprating from 2300 MWt to 2335 MWt.

Projections beyond cycle 21 are based on an average of cycles 16 to 21.

Analysis of H.B. Robinson Unit 2 Capsule X

6-40 Table 6-15 Calculated Fast Neutron Exposure of the H. B. Robinson Unit 2 Upper Circumferential Vessel Weld at the Clad/Base Metal Interface at Selected Azimuthal Angles Neutron Fluence [E > 1.0 MeV]

CumulativeO Neutron Fluence [n/cm 2J perating Time 0.0 10.0 20.0 30.0 40.0 45.0

[EFPY]

Degrees Degrees Degrees Degrees Degrees Degrees 20.39 a 1.21E+19 8.93E+18 5.26E+18 4.11E+18 2.83E+18 2.65E+1 8 21.78 b I r 5.26E+/-18 1- 4.1lE÷18 2.83E+l 8 1.25E+19 9.29E+18 5.50E+18 4.30E+18 2.99E+18 2.80E+18 29.00 1.57E+19 1.16E+19 6.82E+18 5.32E+18 3.78E+18 3.5 6E+ 18 30.00 1.62E+19 1.20E+19 7.OOE+18 5.46E+18 3.89E+18 3.67E+18 35.00 1.84E+19 1.36E+19 7.91E+18 6.16E+18 4.44E+ 18 4.20E+18 40.00 2.06E+19 1.52E+19 8.82E+18 6.87E+-18 4.99E+1 8 4.73E+18 45.00 2.28E+19 1.69E+19 9.74E+18 7.57E+ 18 5.54E+18 5.26E+18 50.00 2.50E+19 1.85E+19 1.06E+19 8.27E+18 6.09E+18 5.79E+18 Iron Atom Displacements Cumulative Iron Atom Dis lacements [dpa Operating Time

[EFPY]0. Degrees 10.0 20.0 30.0 40.0 45.0 Degrees Degrees Degrees Degrees Degrees 20.39 a 2.10E-02 1.49E-02 8.65E-03 6.73E-03 4.63E-03 4.36E-03 21.78 b 2.18E-02 1.55E-02 9.04E-03 7.05E-03 4.88E-03 4.61E-03 29.00 2.71E-02 1.94E-02 1.12E-02 8.71E-03 6.18E-03 5.87E-03 30.00 2.78E-02 1.99E-02 1.15E-02 8.94E-03 6.36E-03 6.04E-03 35.00 3.14E-02 2.26E-02 1.30E-02 1.01E-02 7.27E-03 6.92E-03 40.00 3.5 1E-02 2.54E-02 1.45E-02 1.13E-02 8.17E-03 7.79E-03 45.00 3.87E-02 2.81E-02 1.60E-02 1.24E-02 9.07E-03 8.66E-03 50.00 4.23E-02 3.08E-02 1.75E-02 1.36E-02 9.97E-03 9.53E-03

a. EFPY at end of cycle 20.

b. Estimated EFPY at end of cyc:le 21.

Note: Up to the end of cycle 21, EFPY is calculated based on a full operating power of 2300 MWt. For projections beyond cycle 21, the projections include a 1.5% plant uprating from 2300 MWt to 2335 MWt.

Projections beyond cycle 21 are based on an average of cycles 16 to 21.

Analysis of H.B. Robinson Unit 2 Capsule X

6-41 Table 6-16 Calculated Fast Neutron Exposure of the H. B. Robinson Unit 2 Lower Circumferential Vessel Weld at the Clad/Base Metal Interface at Selected Azimuthal Angles Neutron Fluence [E > 1.0 MeV]

CumulativeO Neutron Fluence [n/cm 2]

perating Time 0.0 10.0 20.0 30.0 40.0 45.0

[EFPY] Degrees Degrees Degrees Degrees Degrees Degrees 20.39 a 1.52E+19 1.20E+19 9.53E+18 8.23E+18 5.72E+18 5.35E+18 21.78 1.54E+19 1.22E+19 9.92E+18 8.62E+18 6.05E+18 5.67E+18 29.00 1.67E+19 1.36E+19 1.20E+19 1.07E+19 7.73E+18 7.29E+18 30.00 1.69E+19 1.38E+19 1.23E+19 1.10E+19 7.96E+18 7.51E+18 35.00 1.78E+19 1.48E+19 1.37E+19 1.24E+19 9.12E+18 8.63E+18 40.00 1.87E+19 1.58E+19 1.51E+19 1.38E+19 1.03E+19 9.75E+18 45.00 1.96E+19 1.67E+19 1.66E+19 1.52E+19 1.14E+19 1.09E+19 50.00 2.05E+19 1.77E+19 1.80E+19 1.66E+19 1.26E+19 1.20E+19 Iron Atom Displacements Cumulative Iron Atom Displacements [dpal Operating Time

[e 0.0 10.0 20.0 30.0 40.0 45.0

[EFPY] Degrees Degrees Degrees Degrees Degrees Degrees 20.39 a 2.52E-02 2.01E-02 1.56E-02 1.34E-02 9.34E-03 8.80E-03 21.78 b 2.56E-02 2.05E-02 1.62E-02 1.41E-02 9.88E-03 9.32E-03 29.00 2.78E-02 2.29E-02 1.96E-02 1.74E-02 1.26E-02 1.20E-02 30.00 2.81E-02 2.32E-02 2.OOE-02 1.79E-02 1.30E-02 1.23E-02 35.00 2.96E-02 2.48E-02 2.23E-02 2.02E-02 1.49E-02 1.42E-02 40.00 3.1 1E-02 2.65E-02 2.47E-02 2.25E-02 1.68E-02 1.60E-02 45.00 3.26E-02 2.81E-02 2.70E-02 2.48E-02 1.86E-02 1.78E-02 50.00 3.41E-02 2.97E-02 2.93E-02 2.71E-02 2.05E-02 1.97E-02

a. EFPY at end of cycle 20.

b. Estimated EFPY at end of cycle 21.

Note: Up to the end of cycle 21, EFPY is calculated based on a full operating power of 2300 MWt. For projections beyond cycle 21, the projections include a 1.5% plant uprating from 2300 MWt to 2335 MWt.

Projections beyond cycle 21 are based on an average of cycles 16 to 21.

Analysis of H.B. Robinson Unit 2 Capsule X

6--4 6-42 Table 6-17 Calculated Surveillance Capsule Lead Factors Capsule Location Status Present Exposure Future Exposure Lead Factor a Lead Factor b S 100 Withdrawn EOC 1 1.90 V 200 Withdrawn EOC 3 0.91 T 00 Withdrawn EOC 8 2.80 X 400,00 Withdrawn EOC 20 1.63 U 300, 10 In Reactor 1.41 2.02 Y 300 In Reactor 0.92 1.04 W 400 In Reactor 0.59 0.61 Z 400 In Reactor 0.59 0.61

a. Present exposure lead factors are the ratio of capsule fluence (E > 1.0 MeV) to the maximum vessel fluence (E > 1.0 MeV) at the time of withdrawal or at the end of cycle 20.

b. Future exposure lead factors are the ratio of capsule fluence rate (E > 1.0 MeV) to the maximum vessel fluence rate (E> 1.0 MeV) for exposure after the end of cycle 20, projected using the average of cycles 16 through 21.

Table 6-18 Calculated Fast Neutron Exposure of the H. B. Robinson Unit 2 Nozzle Components Cumulative Neutron Fluence [E > 1.0 MeV) (n/cm 2)

Operating Time Inlet Nozzle Inlet Nozzle Outlet Nozzle Outlet Nozzle

[efpy] Weld Shell Weld Shell 29 2.47E+17 2.19E+17 1.59E+17 1.21E+17 30 2.54E+17 2.25E+17 1.64E+17 1.25E+17 35 2.88E+17 2.56E+17 1.86E+17 1.42E+17 40 3.23E+17 2.86E+17 2.08E+17 1.58E+17 45 3.57E+17 3.17E+ 17 2.30E+17 1.75E+17 50 3.92E+17 3.48E+17 2.53E+17 1.92E+17 Analysis of H.B. Robinson Unit 2 Capsule X

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the intent of ASTM E185-82 and is recommended for future capsules to be removed from the H.B. Robinson Unit 2 reactor vessel. This recommended removal schedule is applicable to 29 EFPY of operation.

Notes:

(a) Updated in Capsule X dosimetry analysis. Lead Factor in Parentheses are for Future Cycles.

(b) Effective Full Power Years (EFPY) from plant startup.

(c) Plant specific evaluation.

(d) Capsule U will reach a fluence of approximately 6.00 x 1019 (50 EFPY Peak Fluence) at approximately 29.8 EFPY.

Thus, it should be pulled at the closest outage to 29.8 EFPY.

(e) If further material data is desired, then it is recommended that these capsules be moved to a higher lead factor location and then removed once their accumulated neutron fluence equals the license renewal (50 EFPY) fluence on the vessel inner surface.

(f) Moved to Capsule "S" Location (280') at Cycle 8.

(g) Capsule Z was inadvertently removed from the H.B. Robinson 2 Reactor Vessel. At this time it is unconfirmed that Capsule Z was re-installed into the vessel or placed in the spent fuel pool.

Analysis of H.B. Robinson Unit 2 Capsule X

RNP-RA/02-0033, Part 1 of 2, H.B. Robinson Steam Electric Plant, Unit No. 2, Report of Analysis of Surveillance Capsule X for Reactor Vessel Radiation Surveillance Program

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RNP-RA/02-0033, Part 1 of 2, H.B. Robinson Steam Electric Plant, Unit No. 2, Report of Analysis of Surveillance Capsule X for Reactor Vessel Radiation Surveillance Program

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6.1 INTRODUCTION

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