ML20012C012

Analysis of Capsule U from VEPCO North Anna Unit 2 Reactor Vessel Radiation Surveillance Program

ML20012C012
Person / Time
Site:	North Anna
Issue date:	01/31/1990
From:	Albertin L, Shaun Anderson, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20012C010	List: ML20012C009 ML20012C012
References
WCAP-12497, NUDOCS 9003190327
Download: ML20012C012 (79)

Text

WESTINGHOUSE CLASS 3 L

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WCAP-12497 g-2 ANALYSIS OF CAPSVLE U FROM THE VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM E. Terek S. L. Anderson L. Albertin a

January 1990 m

Work performed under Shop Order No. VJGP-106 APPROVED:

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T.A.Meyer,.Mankger Structural Materials and Reliability Technology Prepared by Westinghouse for the Virginia Electric and Power Company WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Division g.

P.O. Box 2728 Pittsburgh, Pennsylvania 15230

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PREFACE e, -

This report has been technically reviewed and verified.

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Reviewer Sections 1 through 5 and 7, 8 N. K. Ray A

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Section 6 E. P. Lippincott D= _? @

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Section Title Page N

1

SUMMARY

OF RESULTS 1-1 2

INTRODUCTION 2-1 3

BACKGROUND 3-1 4

DESCRIPTION OF PROGRAM 4-1 5

TESTING OF SPECIMENS FROM CAPSULE U 5-1 5-1.

Overview 5-1 5-2.

Charpy V-Notch Impact Test Results 5-3 5-3.

Tension Test Results 5-4 5-4.

Wedge Opening Loading Tests 5-4 6

RADIATION ANALYSIS AND NEUTRON 0051 METRY 6-1 6-1.

Introduction 6-1 6-2.

.iscrete Ordinates Analysis 6-2 6-3.

Neutron Dosimetry 6-7 7

SURVEILLANCE LAPSULE REMOVAL SCHEDULE 7-1 8

REFERENCES 8-1 l 9 6

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LIST OF ILLUSTRATIONS E

Figure Title Page i

4-1 Arrangement of Surveillance Capsules in the 4-5 North Anna Unit 2 Reactor Vessel j

4-2 Capsule U Diagram Showing Location of Specimens, 4-6 Thermal Monitors, and Desimeters 5-1 Charpy V-Notch Impact Data for North Anna Unit 2 5-12 Reactor Vessel Shell Forging 04, Heat No. 990496/292424 (Tangential Orientation) 5-2 Charpy V-Notch Impact Data for North Anna Unit 2 5-13 Reactor Vessel Shell Forging 04, Heat No. 990496/292424 (Axial Orientation) 5-3 Charpy V-Notch Impact Data for North Anna Unit 2 5-14 l

Reactor Vessel Weld Metal l

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5-4 Charpy V-Notch Impact Data for North Anna Unit 2 5-15 i

Reactor Vessel Weld Heat F 'ected Zone Metal i

5-5 Charpy impact Specimen Fracture Surfaces for North 5-16 Anna Unit 2 Reactor Vessel Shell Forging 04, Heat No. 990496/292424 (langential Orientation) 5-6 Charpy impact Specimen Fracture Surfaces for North 5-17 l

Anna Unit 2 Reactor Vessel Shell Forging 04, Heat l

No. 990496/292424 (Axial Orientation) 5-7 Charpy impact Specimen Fracture Surfaces for 5-18 North Anna Unit 2 Reactor Vessel Weld Metal L

5-8 Charpy Impact Specimen Fracture Surfaces for 5-19 l

North Anna Unit 2 Reactor Vessel HA2 Metal 1

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r LIST OF ILLUSTRATIONS (Cont) l Figure Title Page

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5-9 Tensile Properties for North Anna Unit 2 Reactor 5-20

'i Vessel Shell Forging 04, Heat No. 990496/292424 (Axial Orientation) i i

5-10 Tensile Properties for North Anna Unit 2 Reactor 5-21 i

Vessel Weld Metal 5-11 Fractured Tensile Specimens for North Anna Unit 2 5-22' Reactor Vessel Shell Forging 04, Heat No. 990496/292424 F

(Axial Orientation) 5-12 Fractured Tensile Specimens for North Anna Unit 2 5-23 Reactor Vessel Weld Metal I

5-13 Typical Stress-Strain Curve for Tension Specimens 5-24 6-1 Surveillance Capsule Geometry 6-13

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LIST OF TABLES E

Table Title Page 4-1 Chemical Composition of the North Anna Unit 2 4-3 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the North Anna Unit 2 4-4 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the North Anna Unit 2 5-5 Reactor Vessel Shell Forging 04, Heat No. 990496/292424 18 2

Irradiated at 550'F, Fluence 9.55 x 10 n/cm (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the North Anna Unit 2 5-6 Reactor Vessel Weld Metal and HAZ Metal irradiated at 550*F, Fluence 9.55 x 10 n/cm2 (E > 1.0 MeV) 18 5-3 Instrumented Charpy Impact Test Results for North 5-7 Anna Unit 2 Reactor Vessel Shell Forging 04, Heat No. 990496/292424 Irradiated at 550'F, Fluence 9.55 x 10 n/cm2 (E > 1.0 MeV) 18 5-4 Instrumented Charpy impact Test Results for North Anna 5-8 Unit 2 Reactor Vessel Weld Metal and HAZ Metal 18 2

Irradiated at 550'F, Fluence 9.55 x 10 n/cm 18 2

5-5 The Effect of 550'F Irradiation at 9.55 x 10 n/cm 5-9 (E > 1.0 MeV) on the Notch Toughness Properties of

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the North Anna Unit 2 Reactor Vessel Materials 5-6 Comparison of North Anna Unit 2 Reactor Vessel Surveillance 5-10 Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions 4076s 'C1239010 yj

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1 LIST OF TABLES (Cont)

Table Title Page

,*s 5-7 Tensile Properties for North Anna Unit 2 Reactor Vessel 5-11 18 2

Material Irradiated at 550'F, to 9.55 x 10 n/cm l

(E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Parameters at the 6-14 f

Center of Capsule 6-2 Calculated Fast Neutron Exposure Parameters at the 6-15 Pressure Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux 6-16 (E>1.0 MeV) Within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-17 (E>0.1 MeV) Within the Pressure Vessel Wall 6-5 Relative Radial Distribution of Iron Displacement 6-18 Rate (dpa) Within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-19 6-7 Irradiation History of Neutron Sensors Contained 6-20 in Capsule U 6-8 Measured Sensor Activities and Reaction Rates 6-23 6-9 Summary of Neutron Dosimetry Results 6-25 6-10 Comparison of Measured and FERRET Calculated 6-26 Reaction Rates at the Surveillance Capsule Center eC?$6> 01230010 yjj

h LIST OF TABLES (Cont)

Table Title Page 1

6-11 Adjusted Neutron Energy Spectrum at the Surveillance

.6-27 Capsule Center i

6-12 Comparison of Calculated and Measured Exposure 6-28 Levels for Capsule U i

6-13 Neutron Exposure Projections at Locations on the 6-29 Pressuro Vessel Clad / Base Metal Interface 6-14 Vessel Neutron Exposure Values (E > 1.0 MeV) for 6 l Use in the Generation of Heatup/Cooldown Curves i

6-15 Updated Lead Factors for Unit 2 Surveillance 6-31 Capsules i

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SECTION 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in Capsule V, the second surveillance capsule to be removed from the Virginia Electric and Power Company North Anna Unit 2 reactor pressure vessel, resulted in the following conclusions:

o The capsule received an average fast neutron fluence (E > 1.0 MeV) 18 2

of 9.55 x 10 n/cm,

o Irradiation of the reactor vessel intermediate shell forging 04, Heat 18 2

No. 990496/292424, to 9.55 x 10 n/cm, resulted in a 30 ar.d 50 f t-lb transition temperature increase of 25'F for specimens oriented with the longitudinal axis of the specimen parallel to the major working direction (tangential orientation),

o Irradiation of the reactor vessel intermediate shell forging 04, 18 2

Heat No. 990496/292424, to 9.55 x 10 n/cm, resulted in 30 and l

50 ft-lb transition temperature increases of 60 and 55'F, respectively, for specimens oriented with the longitudinal axis of the l

specimen perpendicular to the major working direction (axial l

orientation).

1 18 2

o Weld metal irradiated to 9.55 x 10 n/cm resulted in 30 and 50 f t-lb transition temperature increases of 13 and 30*F, respectively.

1 18 2

o Irradiation to 9.55 x 10 n/cm resulted in no decrease in the average upper shelf energy of forging 04, Heat No. 990496/292424 and no decrease in the average upper shelf energy of the weld metal.

Both materials exhibit a more than adequate upper shelf level for continued safe plant operation.

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o Comparison of the 30 ft-lb transition temperature increases for the North Anna Unit 2 surveillance material with predicted increases using i

the methods of NRC Regulatory Guide 1.99, Revision 2, demonstrated 5

that forging 04, Heat No. 990496/292424 material.and weld metal-i transition temperature increases were less than predicted.

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SECTION 2 INTRODUCTION I

This report presents the results of the examination of Capsule V, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Virginia Electric and Power Company North Anna Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Virginia Electric and Power Company North Anna Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation.

A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Davidson, Yanichko, and Phillips.II) The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Electric Corporation personnel performed the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens.

This report summarizes testing and the postirradiation data obtained from surveillance Capsule U removed from the North Anna Unit 2 reactor vessel and discusses the analysis of the data.

The data are also compared to results of the previously removed North Anna Unit 2 Capsule V[2),

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SECTION 3 BACKGROUND The ability of the large steel pressure vessel, which contains 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

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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 SA508 Class 2 (base material of the North Anna Unit 2 reactor pressure vessel beltline) are well documented in industry literature. Generally, low alloy ferritic materials demonstrate an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section 111 of the ASME Boiler and Pressure Vessel Code.

The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)*

RT is defined as the greater of either the drop weight nil-ductility NDT transition tenverature (NDTT per ASTM E-208) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material.

The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (Kyp curve) which appears in Appendix G of the ASME Code. The Kyp 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 Kyp curve, allowable strass 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.

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RTNDT and the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material

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properties.

The radiation embrittlement or changes in mechanical properties-l

- of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Virginia Electric and Power Company North Anna Unit 2 Reactor Vessel Radiation Surveillance Program.I13 A

t surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specinens are tested.

Tne increase in the average Charpy V-notch 30 f t ib temperature ( ARTNDT) due to irradiation is added to the original RT to adjust the RT for radiation NDT NDT embrittlement.

This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the material to the Kjg curve and to set operating limits for the nuclear power plant which reflect the effects of irradiation on the reactor vessel materials.

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SECTION 4 DESCRIPTION OF PROGRAM Eight surveillance ca'psules for monitoring the effects of neutron exposure on 5

the North Anna Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup.

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.

Capsule V (Figure 4-2) was removed after 6.07 effective full power years of plant operation.

This capsule contained Charpy V-notch impact, tensile, and wedge opening loading (WOL) te'st specimens from the reactor vessel intermedi-ate shell forging 04, heat no. 990496/292424, weld metal representative of that used fer the beltline region girth welds of the reactor vessel and weld heat-affected zone (HAZ) material.

All heat-affected zone (HAZ) specimens were obtained from the.HAZ of forging 04, heat no. 990496/292424.

The chemistry and heat treatment of the surveillance material are presented in Table 4-1 and Table 4-2, respectively.

The chemical analyses reported in Table 4-1 were obtained from unirradiated material used in the surveillance program.

All test specimens were machined from the 1/4 thickness location of forging 04, Heat No. 990496/292424 Test specimens represent material taken at least one forging thickness (as quenched) from the quenched end of the plate.

Capsule V contained Charpy V-notch impact specimens from forging 04 machined in both the tangential orientation (longitudinal axis of the specimen parallel to the major working direction) and axial orientation (longitudinal axis of the specimen perpendicular to the major working direction).

Capsule V contained core region weld Charpy impact specimens that were machined from the

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weldment such that the long dimension of the Charpy was normal to the weld direction and the notch was machinea such that the direction of crack l

propagation in the specimen was in the weld direction.

Capsule V, also, m w o m o io 43

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contained Charpy impact specimens machined from HAZ metal. ' Tensile specimens I

were machined with the longitudinal axis of the specimen perpendicular to the mejor working direction of forging 04. Capsule V, also, contained tensile specimens made from the core region weldment.

The WOL weld metal test specimens contained in Capsule U were machined such that the crack will propagate in the weld direction.

Capsule U contained dosimeter wires of pure iron, copper, nickel, and aluminum

-0.15% cobalt (cadmium-shielded and unshielded) wire, Neptunium (Np237) and i

Uranium (U238). The dosimeters are used to measure the integrated flux at specific neutron energy levels.

Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex i

tubes were included in the capsule and were located as shown in Figure 4 2.

The two eutectic alloys and their melting points are:

I 2.5% Agi 97.5% Pb Melting Point 579'F (304'C) l 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590'F (310'C)

-The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule V are shown in Figure 4-2.

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TABLE 4-1 CHEMICAL COMPOSITION OF 4

THE NORTH ANNA UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS J

Forging 04 [c]

I Heat No. 990496/292424 Weld Metal "3ICI Element (Wt. %)

(Wt. %)

C 0.19 0.19 0.08 5

0.011 0.015 0.011 N

0.011 2

0.011 Co 0.003 0.011

<0.002 Cu 0.11 0.09 0.088 Si 0.25 0.21 0.25 No 0.60 0.63 0.49 Ni 0.86 0.80 (b) 0.084 Mn 0.76 0.67 1.82 Cr 0.35 0.34 0.042 V

0.031 0.02 0.002 P

0.018 0.010 0.017 Sn 0.016 0.004 A1 0.023 0.017 0.015 (a) Weldment was made from sections of forging 04 and adjoining lower shell course forging 03 using weld wire. representative of that used in the original fabrication (b) Rotterdam Dockyard Analysis.

(All other chemistry values are from WestinghouseAnalysis)

[c] All elements not listed are less than 0.010 weight percent.

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i TABLE 4-2 I

HEAT TREATMENT OF THE NORTH ANNA UNIT 2 i

REACTOR VESSEL SURVEILLANCE MATERIALS t

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Material Temperature ('F)

Time (br)

Coolant Intermediate Shell 1688-1697' 2 1/2 Water quenched i

Forging 04, Heat 1220-1229 6.0 Furnace cooled to 842*F No. 990496/292424 1130+25 14 3/4 Furnace cooled i

. Weld Metal 1130+25 13 1/2 Furnace cooled l

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U 90' Figure 4-1. Arrangement of Surveillance Capsules in the North Anna Unit 2 Reactor Vessel wwonse:it 4-5

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l SPECIMEN NUMBERING CODE:

CAP GT - FORGING 04 ( AXl AL)

GL - FORGING 04 (LONGITUDINAL)

GW - WELD GH - HAZ.

00 slut TENSILE WOL WOL WOL WOL TENSILE CHARPY CHARPY BLOC Gil2 GWl2 GH72 GH70 GH68 GL 48 GWB GW7 GW6 GWS n:

Gill GWil GH71 GH69 GH 67 GL47 h

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CHARPY CHARPY CHARPY CHARPY CH AR PY CHARPY CHARPY CHARPY CHARPY GH66 GL46 GH64 GL44 GH62 GL 42 Gw72 G172 Gw70 GT70 Gw68 G168 Gw66 G1(6 Gw64 G164 Gw63 G162 GH55 GL45 GH63 GL43 GH61 G L41 GW74 G171 G#69 C169 Gw67 G167 Gw65 G165 Gw63 G163 Gw63 G161

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Also Availabic On Aperture Card TO BOTTOM OF VESSEL gram Showing location of Specimens, Thermal Dosimeters 90 o3 I w; 3a7-0 es

SECTION 5 TESTING OF SPECIMENS FROM CAPSULE U 5-1.

OVERVIEW i

The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center Laboratory with consultation by Westinghout Nuclear Energy Systems personnel.

Testing was performed in accordance with 10CFR50, Appendices G and l3), ASTM Specification E185-82 and Westinghouse Procedure MHL 8402, H

Revision 1 as modified by Westinghouse RMF Procedures 8102, Revision 1 and 8103, Revision 1.

Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8772.II) No discrepancies were found.

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Examination of the two low-melting 304'C (579'F) and 310'C (590'F) eutectic alleys indicated no melting of either type of thermal monitor.

Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304'C (579'F).

The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103, Revision 1 on a Tinius-01sen Model 74, 358J machine.

The tup (striker) of the Charpy machine is instrumented with an Effects Technology Model 500 instrumentation system.

With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E ).

From the lead-time curve, the load of general yielding-D (PGY), the time to general yielding (tgy), the maximum load (P ), and M

the time to maximum load (t ) can be determined.

Under some test y

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 (P ), and the load at which fast fracture terminated is p

identified as the arrest load (P )*

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The energy at maximum load (E ) was determined by comparing the energy-time g

record and the lead-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 (E ) is the difference p

between the total energy to fracture (E ) and the energy at maximum lead.

D The yield stress (cy) is calculated from the three point bend formula.

The flow stress is calculated from the average of the yield and maximum loads, also using the three point bend formula.

Percentage shear was determined from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77.

The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests were performed on a 20,000 pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF Procedure 8102, Revision 1.

All pull rods, grips, and pins were made of Inconel 718 hardened to HRC45.

The upper pell rod was connected through a universal joint to improve axiality of leading.

The tests were conducted at a constant crossheed speed of 0.05 inch per minute throughout the test.

Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer.

The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure.

The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67.

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 temperature:

Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip.

In test mwomeo io 5-2

configuration, with a slight load on the specimen, a plot of tpecimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550'F (288'C).

The upper grip was used to control the furnace temperature. During the actual testing the

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grip temperatures were used to obtain desired specimen temperatures.

Experiments indicated that this method is accurate to plus or minus 2*F.

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 postfracture 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.

5.2.

CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials 3

18 2

contained in Capsule U irradiated to approximately 9.55 x 10 n/cm at 550*F are presented in Tables 5-1 through 5-4 and Figures 5-1 through 5-4.

The transition temperature increases and upper shelf energy decreases for the Capsule U material are shown in Table 5-6.

Irradiation of the vessel intermediate shell Forging 04, Heat No. 990496/292424 18 material (tangential orientation) specimens to 9.55 x 10 n/cm2 (Figere 5-1) resulted in a 30 and 50 f t-lb transition temperature increase of 25'F and an upper shelf energy increase of 5 ft-lb when compared to the unirradiated data.Ill Irradiation of the vessel intermediate shell Forgin 04, Heat No. 990496/292424 material (axial orientation) specimens to 9.55 x 10 8,je,2 (Figure 5-2)

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resulted in a 30 and 50 ft-lb transition temperature increase of 60 and 55'F, respectively, and an upper shelf energy increase of 12 f t-lb when compared to the unirradiated data.

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i 18 Weld metal irradiated to 9.55 x 10 n/cm2 (Figure 5-3) resulted in 30 and 50 ft-lb transition temperature increases of 13 and 30'F, respectively, and an upper shelf energy increase of 7 ft-lb.

18 Weld HAI metal irradiated to 9.55 x 10 n/cm2 (Figure 5-4) resulted in 30 and 50 ft-lb transition temperature increases of 55 and 45'F, respectively, and an upper shelf energy decrease of 8 ft-lb.

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasing test temperature.

Table 5-6 shows a comparison of the 30 ft-lb transition temperature (ARTNDT) increases for the various North Anna Unit 2 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2.I43 This comparison shows that the transition temperature 18 2

increase resulting from. irradiation to 9.55 x 10 n/cm is less than the Guide prediction for both Forging 04, Heat No. 990496/292424 and weld metal.

5-3.

TENSION TEST RESULTS 9

The results of tension tests performed on Forging 04, Heat No. 990496/292424 (axial orientation) and weld metal irradiated to 9.55 x 1018 2

n/cm are shown in Table 5-7 and Figures 5-9 and 5-10, respectively.

These results si?ow that irradiation produced a 0.2 percent yield strength increase no greater than 6 ksi for Forging 04, Heat No. 990496/292424 and 2 ksi for the weld metal.

Fractured tension specimens for each of the materials are shown in Figures 5-11 and 5-12.

A typical stress-strain curve for the tension specimens is shown in Figure 5-13.

5-4.

WEDGE OPENING LOADING TESTS The WOL Test specimens will not be tested at this time and will be stored at the Westinghouse Science and Technology Center Hot Cell.

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TABLE 5-1 J

CHARPY V-NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT 2 REACTDR VESSEL SHELL PLATE FORGING 04, HEAT NO. 990496/292424 18 IRRADIATED AT 550*F, FLUENCE 9.55 x 10 n/cm2 (E > 1.0 WeV)

Temperature Impact Euern Lateral Expansion Shear Saanle No. M' 1*gl (ft-lb) M (mils) 1331 (5)

Axial Orientation l

GT69 0

11.0 15.0 7.0 0.18 10 GT65 76 17.0 23.0 15.0 0.38 15 GT61 100 27.0 36.5 27.0 0.69 25 GT68 100 17.0 23.0 15.0 0.38 15 GT62 125 32.0 43.5 29.0 0.74 30 GT66 125 28.0 38.5 27.0 0.69 25 GT67 150 46.0 62.5 36.0 0.91 35 GT64 150 45.0 61.0 37.0 0.94 35 GT63 200 50.0 68.0 41.0 1.04 50 GT71 250 78.0 106.0 58.0 1,47 75 i

GT72 350 82.0 111.0 65.0 1.65 100 GT70 450 85.0 115.0 62.0 1.57 100 Tannential Orientation i.

CL43

-30 11.0 15.0 12.0 0.30 10 GL41 25 41.0 55.5 23.0 0.58 20 GL45 50 82.0 111.0 58.0 1.47 65 GL44 76 35.0 47.5 24.0 0.61 20 GL48 100 79.0 107.0 58.0 1.47 70 l

GL46 150 102.0 138.0 69.0 1.75 90 GL47 250 1

119.0 161.5 75.0 1.91 100 GL42 350 1

121.0 164.0 76.0 1.93 100 l

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407$3/01229010 5-5 l

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550*F FLUENCE 9.55 x 10 n/cm2 (E > 1.0 MeV) 18 Tem Impact Enern Lateral Expansion Shear Q' peraturel'gl (ft-lb) M (alls) 1331

(%)

Samole No.

Weld Metal GW67

- 50

- 46 17.0 23.0 18.0 0.46 20 GW64

- 25

- 32 33.0 44.5 26.0 0.66 35 CW68 0

- 18 39.0 53.0 27.0 0.69 40 GW66 0

- 18 28.0 38.0 25.0 0.64 30 GW63 25 4

62.0 84.0 49.0 1.24 60 GW65 25 4

45.0 61.0 36.0 0.91 45 GW61 50 10 64.0 87.0 46.0 1.17 65 CW71 150 66 82.0 111.0 64.0 1.63 90 GW70 250 121 119.0 161.5 77.0 11.96 100 GW69 350 177 132.0 179.0 82.0 (2.08 100 GW62 450 232 118.0 160.0 81.0 (2.06 100 RAZ Metal GH69

-100

- 73 30.0 40.5 19.0 0.48 25 CH66

- 50

- 46 15.0 20.5 17.0 0.43 15 CH61

- 50

- 46 5.0 7.0 5.0 0.13 5

CH68

- 25

- 32 65.0 88.0 42.0 1.07 60 GH71

- 25

- 32 30.0 40.5 24.0 0.61 30 GH64 0

- 18 62.0 84.0 33.0 0.84 60 CH70 0

- 18 58.0 78.5 37.0 0.94 60 CH65 50 10 72.0 97.5 47.0 1.19 70 GH72 76 24 86.0 116.5 53.0 1.35 90 GH67

-150 66 73.0 99.0 51.0 1.30 80 CH63 250 121 114.0 154.5 62.0 1.57 100 CH62 350 177 103.0 139.5 55.0 1.40 100 9

m

.m.,oum io 5-6

TABLE 5-3 INSTRUNENTED CHARPY INPACT TEST RESUL.TS FOR NORTH ANNA UNIT 2 REACIOR VESSEL SHELL PLATE FORGING 04. HEAT NO. 9904 % /292424 IRRADIAIED AT 550*F, FLUENCE 9.55x1018,7c,2 (E > 1.0 Mev)

Normalised Enerzies Test Charpy Charpy Maximus Prop Tield Time Maxieue Time to Fracture Arrest Tield Flow Sample Temp Energy Ed/A Ee/A Ep/A Load to Tield Load Namieue Load Load Stress Stress Number f*F1 (ft-lb)

(ft-1b/in )

(hips)

(psec)

(hips)

(msec)

(kips)

(kips)

(ksi)

(ksi) 8 Axial Orientation CT69 0

11.0 89 39 50 2.60 80 3.90 135 3.85 0.30 85 107 CT65 76 17.0 137 110 27 3.15 150 4.20 310 4.15 104 121 CT68 100 17.0 137 41 95 3.02 85 3.75 130 3.70 0.45 106 114 l

CTSI 100 27.0 217 156 62 2.50 95 4.40 375 4.35 0.40 83 115 CT66 125 28.0 225 139 87 2.04 80 4.10 345 4.10 1.00 79 107 T

CT62 125 32.0 258 205 53 2.80 75 4.35 450 4.35 0.20 92 118 CT64 150 45.0 362 151 211 3.10 90 4.25 355 4.10 0.70 103 122 CT67 150 46.0 370 191 179 2.85 75 4.20 430 4.10 1.05 95 117 CT63 200 50.0 403 173 230 2.04 105 3.85 445 3.80 2.00 79 103 CT71 250 78.0 628 181 447 2.65 70 4.00 430 88 110 CT72 350 82.0 660 203 457 2.35 45 4.10 460 77 106 GT70 450 85.0 684 188 497 2.30 60 3.95 455 76 103 i

Tanzential Orientation CIA 3

-30 11.0 89 61 27 3.60 45 4.70 130 4.70 0.20 119 137 CL41 25 41.0 330 243 87 3.00 75 4.75 495 4.70 99 128 CL45 50 82.0 660 245 416 2.50 75 4.75 510 3.65 82 120 CL44 76 35.0 282 249 33 2.75 100 4.70 530 4.70 92 124 CL48 100 79.0 636 2312 405 2.35 75 4.55 506 3.30 0.45 78 114 CL46 150 102.0 821 235 586 2.95 75 4.55 500 2.65 1.35 98 124 CL47 250 119.0 958 228 730 2.75 70 4.40 510 90 118 CL42 350 121.0 974 260 714 2.75 130 4.25 620 90 115 t

r

.7=.

~.

- - ~.

~

I l

j TABLE 5-4 INSTRtMENTED CHARPY IMPACT TEST RESULTS FOR NORTH AleIA UNIT 2 REACTOR VESSEL WELD METAL AND HAZ HETAL IRRADIATED AT 550*F, FLUENCE 9.55x1018,fc,2 l

l l

b realized Energies Test Charpy Charpy Maxiome Prop Yield Time hxieue Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Ee/A Ep/A Load to Yield Load Maximus Load Load Stress Stress 1

haber

,[* 1 (f t-lb) fft-lb/in*)

[ kips) fasec)

(hips)

(ssec)

(hips)

(kips)

(ksi)

(ksi)

Weld html CW67

- 50 17.0 137 74 63 2.50 70 3.90 200 3.90 1.10 82 106 CW64

- 25 33.0 266 153 112 3.10 80 4.15 355 4.00 0.50 103 120 L

CW66 0

28.0 225 144 82 2.95 70 4.06 335 4.00 0.70 97 116 CW68 0

39.0 314 219 95 2.85 90 4.15 510 4.15 0.75 94 116 CW65 25 45.0 362 218 145 2.95 80 4.06 505 4.00 1.85 98 116 4'

CW63 25 62.0 499 218 281 2.50 80 4.20 510 4.06 3.45 83 lit t

CW61 50 64.0 515 111 404 1.25 75 2.06 510 2.00 1.35 40 54 CW72 76 55.0 443 211 232 2.75 20 4 10 460 3.90 2.65 91 114 CW71 150 82.0 660 276 384 2.80 75 3.85 665 92 110 L

CW70 250 119.0 958 275 683 2.55 85 3.85 680 83 106 CW69 350 132.0 1063 261 802 2.04 75 3.70 670 79 100

~

CW62 450 118.0 950 237 713 2.04 50 3.55 615 79 99 f

EAZ h tal

(

CH69

-100 30.0 242 151 90 2.25 125 4.75 370 4.70 0.30 75 141 CH61

- 50 5.0 40 35 5

3.10 100 4.30 130 4.03 103 122 t

CH66

- 50 15.0 121 90 31 2.45 85 4.30 235 4.30 0.50 81 112 i

CH71

- 25 30.0 242 152 89 2.90 75 4.40 335 4.35 1.25 97 121 CH68

- 25 65.0 523 298 226 2.60 85 4.90 000 4.55 2.04 86 125 CH70 0

58.0 467 246 221 3.20 80 4.60 505 4.03 1.30 106 129 i

CH64 0

62.0 499 248 251 3.10 80 4.80 505 4.55 1.05 103 130

(

CH65 50 72.0 580 240 340 3.10 90 4.45 520 4.20 2.60 103 126 i

CH72 76 86.0 692 238 454 2.05 90 4.50 525 93 121 l

CH67 150 73.0 588 185 403 2.80 100 4.20 450 93 116

{

4 CH63 250 114.0 918 222 696 2.85 75 4.25 500 94 117 CH62 350 103.0 829 251 579 2.55 105 4.10 605 84 110 l

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=

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m o

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5-9

TABLE 5-6 COMPARISON OF NORTH ANNA UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS ARTNDT('F)

USEDECREASE(%)

Reg. 1.99 Fluence Rev. 2 19 2

Material Capsule 10 n/cm Meas.

Prod.

Meas.

Prod.

Plate 89004-2 V

0.241(a) 9 47 0

14 (Tangential)

U 0.955(b) 25 76 0

20 Plate B9004-2 V

0.241(a) 9 47 18.6 14 (Axial)

U 0.955(b) 60 76 0

20 Wald Metal V

0.241(a) 2 27 20.5 16 0

0.955(b) 13 44 0

23.5 HAZ Metal V

0.241(a) 10 5.7 U

0.955(b) 55 7.5 19 2

(a) Per Reference 2, Table 3-4, the fluence value (0.241x10 n/cm ) reported here is the fluence experienced by the charpy specimens contained in capsule V.

19 2

The geometric center of capsule V experienced a fluence of 0.255x10 n/cm,

19 2

(b) The fluence value of 0.955x10 n/cm reported here is the fluence at the geometric center of capsule U.

L con.mmo io 5-10 l

~

1' l..

TABLE S TENSILE PROPERTIES FOR NORTH ANNA UNIT 2' l

REACTOR VESSEL MATERIAL IRRADIATED AT 550*F 10 9.55 x 1018,fc,2 (E > 1.0 MeV)

Test 0.25 Yield Ultimate Fractere Fracture Fracture

' Uniform Total Reduction Sample Temp. Strength Strength Load Stress Strength Elongation Elongation in Area Material Number M (ksi)

_(ksi)

(kio)

(ksi)-

(ksi)

(%)

(5)

(5)

Forging CT11 74 87.6

-108.0 3.80 158.0 77.4 10.5 20.3 51 l

Forging CT12 550 80.5 99.8 4.20 293.4 85.6 8.3 13.8 48 Weld CW11 74 78.4 90.7 STRAIN CAUGB SLIPPED -

10.3 17.5 68 Weld CW12

' 550 69.3 85.6 0.30 23.7 6.1 9.8 19.8 60 l

u, t

J, t

j l

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i s

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(

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100 200 300 400 500 Temperature ( F) o FIGURE 5-1 CHARPY V-NOTCH IMPACT DATA FOR NORTH ANNA UNIT 2 REACTOR VESSEL

-SHELL PLATE FORGING 04, HEAT NO. 990496/292424 (TANGENTIAL ORIENTATION) 4075s/012290 10

$.12

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FIGURE 5-2 CHARPY V-NOTCH IMPACT DATA FOR NORTH ANNA UNIT 2 REACTOR VESSEL SHELL PLATE FORGING 04, HEAT NO. 990496/292424 (AXIAL ORIENTATION) 4076s'012290 10 5-13 l

(*C)-

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18 2

w 9.55 x 10 n/cm M

30 F 80 13 F 20 0

i i

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-200

-100 0

100 200 300 40 500 Temperature ( F)

FIGURE-5-3 CHARPY V-NOTCH IMPACT DATA FOR NORTH ANNA UNIT 2 REACTOR VESSEL WELD METAL I

es.,cimo 'o 5-14

('C)

-150 -100

- 50 0

50 100 150 200 250 3

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FIGURE 5-4 CHARPY V-NOTCH IMPACT DATA FOR NORTH ANNA UNIT 2 REACTOR VESSEL WELD HEAT AFFECTED ZONE METAL

.m.<oimo :

5-15

n

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FIGURE 5-5 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR NORTH ANNA UNIT 2 REACTOR VESSEL SHELL PLATE FORGING 04, HEAT NO. 990496/292424 (TANGENTIAL ORIENTATION) 4075s/012290 10 5-16 RM-21923

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j 1

FIGURE 5-6 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR NORTH ANNA UNIT 2 l

REACTOR VESSEL SHELL PLATE FORGING 04, HEAT NO. 990496/292424 l

(AXIAL ORIENTATION) l

{

4C75s/01229010 5 17 au.21922

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.GW69 GW62 FIGURE 5-7 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR NORTH ANNA UNIT 2 REACTOR VESSEL WELD METAL 007bs /01229010 5-18 RM-21924

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Y CH72 GH67 GH63 GH62 FIGURE 5-8 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR NORTH ANNA UNIT 2 REACTOP VESSEL WELD HAZ METAL 4075% C'229C 'C R.}O

4

• C

- 50 0

50 100 150 200 250 300 120 I

00 110 Ultimate Tensile Strength 8

_700

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a 5%

g-80 e

V 500 -

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40 I

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80 i

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100 200 300 400 500 600 Temperature ( F)

FIGURE 5-9 TENSILE PROPERTIES FOR NORTH ANNA UNIT 2 REACTOR VESSEL SHELL PLATE FORGING 04, HEAT NO. 990496/292424 (AXIAL ORIENTATION) 4cn outso ie 5-20

C l

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100 200 300 400 500 600 Temperature ( F) 1 FIGURE 5-10 TENSILE PROPERTIES FOR NORTH ANNA UNIT 2 REACTOR VESSEL WELD METAL SQ7$g /Ot J29010

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l FIGURE 5-11 FRACTURED TENSILE SPECIMENS FOR NORTH ANNA UNIT 2 REACTOR VESSEL SHELL PLATE FORGING 04, HEAT NO. 990496/292424 (AXIAL ORIENTATION) l l

6 l.

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FIGURE 5-12 FRACTURED TENSILE SPECIMENS FOR NORTH ANNA UNIT 2 REACTOR VESSEL WELD METAL I

4meom '

5-23 RM-21927

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I 4cns/ol229010 5-24

SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

?

6.1 INTRODUCTION

6 Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons.

First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known.

Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens.

The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.

The latter information is derived solely from analysis.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition.

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 as well as to a more accurate evaluation of damage gradients through the pressure 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,"

recommends reporting displacements per iron atom (dpa) along with fluence nn.,ome w 6-1

(E > 1.0 MeV) to provide a data base for future reference (5).

The energy dependent dpa function to be used for this evaluation is'specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom."

The application of the dpa parameter to l

the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to' the Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials."

This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule V.

Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation-history.

The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided.

6.2 DISCRETE ORDINATES ANALYSIS A plan view of the reactor geometry at the core midplane is shown-in Figure 4-1.

Eight irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 45', 55', 65*, 165',

245', 285', 295*, and 305' relative to the core cardinal axis as shown in Figure 4-1.

A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 6-1.

The stainless steel specimen containers are 1-inch square and approximately 40 inches in height.

The containers are positioned

~

axially such that the specimens are centered on the core midplane, thus spanning the central 3.33 feet of the 12-foot high reactor core.

son,ionm :c g.g

l From a neutron transport standpoint, the surveillance capsule structures are significant, ~They have a marked effect on both the distribution of neutron.

flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel.

In order to properly determine the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters (e(E > 1.0 Mev,) 9(E > 0.1 Mev),

and dpa) through the vessel wall.

The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/e(E > 1.0 MeV), within the pressure vessel geometry.

The relative _ radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations.

The second set of calculations consisted of a series of adjoint analyses relating the-fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core.

The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for the applicable operating period; and established the means to perform similar predictions and dosimetry evaluations for subsequent fuel cycles.

It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects m,omem 6-3

of varying' neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased.

.The' absolute cycle specific data from the adjoint evaluations together with-relative neutron energy spectra and radial distribution information from the forward calculation provided the means to:

1. -Evaluate neutron dosimetry obtained from surveillance capsule locations.

2.

Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.

3.

Enable a direct comparison of analytical prediction with measurement.

4.

Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.

The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R, O geometry using the DOT two-dimensional discrete ordinates code [6] and the SAILOR cross-section library (7].

The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for-light water reactor applications.

In these analyses anisotopic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an 5 rder of angular 8

quadrature.

The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 3-loop plants.

Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery.

Furthermore, for the peripheral fuel assemblies, a 2e uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used.

Since it is unlikely that a single reactor would have a power distribution at the nominal +2e son, ca so s.4

I level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.

.t All adjoint analyses were also carried out using an S8 order of angular l ;

quadrature and the P3 cross-section approximation from the SAILOR library.

Adjoint source locations were chosen at several azimuthal locations along the I

pressure vessel inner radius as well as the geometric center of each surveillance capsule.

Again, these calculations were run in R, O geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, e (E > 1.0 MeV). Having the importance functions and appropriate core source distributions.-the response of interest could be calculated as:

R (r, 0) = /r #0 E I(r, 0, E) S (r, 0, E) r dr de dE l

where: R(r,0) e (E > 1.0 MeV) at radius r and azimuthal angle 0

=

I (r, 0, E)

Adjoint importance function at radius, r, azimuthal

=

l angle 0, and neutron source energy E.

1.

S (r, 0, E)

Neutron source strength at core location r, 0 and

=

energy E.

Although the adjoint importance functions uted in the North Anna Unit 2 i

analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the l

implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order.

Thus, for a given location the ratio of dpa/c (E > 1.0 MeV) is insensitive to changing' core source distributions.

In the application of these adjoint important functions to the North Anna Unit 2 reactor, therefore, calculation of the iron displace-ment rates (dpa) and-the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/e (E > 1.0 MeV) and e (E > 0.1 MeV)/

e (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific e (E > 1.0 MeV) solutions from the individual adjoint evaluations.

1 dC75s C1239010 6-5 m

-m.

.m

Further discussion of the plant specific exposure calculations for North Anna Unit 2 is provided in Reference 8.

Selected results from the neutron transport analyses performed for the North Anna Unit 2 reactor are provided in Tables 6-1 through 6-5.

The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall.

In Table 6-1, the calculated exposure parameters (# (E > 1.0 MeV),

e-(E > 0.1 MeV), and dpa) are given at the geometric center of surveillance capsule U using plant specific core power distributions and averaging over cycles 1-6 and also over cycles 4-6 which are low leakage fuel loading patterns.

These plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis and to provide an evaluation of the more recent loading patterns appropriate for projecting into the future.

Similar data is given in Table 6-2-for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the cycle 1-6 average and the low leakage plant

~

specific power distributions.

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 exposure levels of the vessel wall itself.

Radial gradient information for neutron flux (E > 1.0 MeV), neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations.

Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.

e l

437$s.0m90 to g.g l

For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the O' azimuth is given by:

2 1/4T(0')

e(199.87, O')-F (204.87, O*)

=

1/4T(0')

Projected neutron flux at the 1/4T position where

=

on the O' azimuth e (199.87, O*)

Projected or calculated neutron flux at the

=

vessel inner radius on the O' azimuth.

F-(204.87,O')

Relative radial distribution function from Table

=

6-3.

Similar expressions apply for exposure parameters in terms of e(E > 0.1 MeV) and dpa/sec.

6.3 NEUTRON DOSIMETRY The passive neutron sensors included in the North Anna Unit 2 surveillance program are 1isted in Table 6-6.

Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest (e (E > 1.0 Mev), e (E > 0.1 MeV),

dpa).

The relative locations of the neutron sensors within the capsules are shown in Figure 4-2.

The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules.

The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.

1

-9 un, omw '

6-7

The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level 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:

o The specific activity of each monitor.

o The operating history of the reactor, o

The energy response of the monitor, o

The neutron energy spectrum at the monitor location.

.o The physical characteristics of the monitor.

The specific activity of each of the neutron monitors was determined using established ASTM procedures (5, 9 through 21).

Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer.

The irradiation history of the North Anna Unit 2' reactor during Cycles 1-6 was obtained from NUREG-020, " Licensed Operating Reactors Status Summary Report" for the applicable period.

~

The irradiation history applicable to capsule U is given in Table 6-7.

Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8.

Reaction rate values were derived using the pertinent data from Tables 6-6 and 6"7.

L Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code [22). 'The FERRET approach used the measured reaction rate data and the calculated l

neutron energy spectrum at the the center of the surveillance capsule as input l

and proceeded to adjust a priori (calculated) group fluxes to produce a best e n co m io 6-8

fit (in a least squares sense) to the reection rate data.

The exposure parameters along with associated uncertainties where then obtained from the adjustedspectra.

In the FERRET evaluations, a log normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations.

In general, the measured values f are

~

linearly related to the flux e by some response matrix A:

f (s,a) = I A (s)

(a) i i9

,9 g

where i indexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be simultaneously adjusted.

For example, R$=I o$g g

o 9

relates a set of measured reaction rates R to a single spectrum 9 by the multigroup cross section og.

(In this case, FERRET also adjusts the g

~

cross-sections.) The legnormal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.

In the FERRET analysis of the dosimetry data, the continuous quantities (i.e.,

fluxes and cross-sections) were approximated in 53 groups.

The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [23).

This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions w.here group boundaries do not coincide.

The 620 point spectrum was then easily collapsed to the group scheme used in FERRET.

e' 1

407$s 'C t 2M01C g.g

The cross-sections were also collapsed into the 53 energy group structure using SAND 11 with calculated spectra (as expanded to 620 groups) as weighting functions.

The cross sections were taken from the ENDF/B-V dosimetry file.

Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section.

Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.

For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight.

In some cases, as for the cross sections, a multigroup covariance matrix is used.

More often, a simple parameterized form is used:

gg,=-Rf+R M

R,P g g gg, where R specifies an overall fractional normalization uncertainty (i.e.,

N complete correlation) for the corresponding set of values.

The fractional uncertainties R specify additional random uncertainties for group g that g

are correlated with a correlation matrix:

Pgg, = (1 - 0)'6g9, + 0 exp (- (a-g')2) 2r The first term specifies purely random uncertainties while the second term describes short range correlations over a range r (0 specifies the strength of the latter term.)

For the a priori calculated fluxes, a short range correlation of r = 6 groups was used. This choice implies that neighboring groups are strongly correlated when e is-close to 1.

Strong long-range correlations (or anticorrelations) were justified based on information presented by R. E. Maerker (24).

Maerker's results are closely duplicated when y = 6.

For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.

un..o m" '

6-10

Results of the FERRET evaluation of the capsule V dosimetry are given in

- Table 6-9 The date summarized in Table 6-9 indicated that the capsule received an integrated exposure of 9.55 x 10 n/cm2 (E > 1.0 MeV) with 18 an associated uncertainty of + 8%.

Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa).

Summaries of the fit of the adjusted spectrum are provided in Table 6-10.

In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates.

The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.

A summary of the measured and calculated neutron exposure of capsule V is presented in Table 6-12.

The agreement between calculation and measurement falls within 11% for all fast neutron exposure parameters, whereas, the thermal neutron exposure calculated for capsule U was low by 28% relative to the measured value.

Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13.

Along with the current (6.07 EFPY) exposure-derived from the capsule U measurements, projections are also provided for an exposure period to end of vessel design life (32 EFPY).

The calculated design basis exposure rates given in Table 6-2 were used to perform projections beyond the end of Cycle 6.

i In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the North Anna Unit 2 reactor coolant system, exposure projections to 15 EFPY and 32 EFPY were employed.

Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel. wall are provided in Table 6-14.

In order to access RTNDT vs.

fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations 4' (1/4T) = 4 (Surface) { dpa (1/4T)

)

dpa (Surface)

W 6-11

F'

- j a

't'(3/4T)=t(Surface){ dpa (3/4T)

)

apa (Surface)

Using this approach results in the dpa equivalent fluence values listed in

-Table 6-14.

In Table 6-15 updated lead factors are listed for each of the North Anna Unit 2 surveillance capsules. These data may be used as a guide in es'tablishing future withdrawal schedules for the remaining capsules, f-T 4

9

""'2*'"

6-12

i 1

\\

l~

1 CHARPY

[ SPECIMEN

//

/

/////

l

'////////

1 l

THERMAL-SHIELD Figure 6-1.

Surveillance Capsule Geometry i

6-13 4cn. "" '

o TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE CENTER OF CAPSULE

~

Cycle 1-6 (a)

Cycle 4-6(b)

+ (E) 1.0 MeV)(c) 5.54 x 10 0 5.14 x 1010 1

$ (E) 0.1 MeV)(c) 1.68 x 1011 1.56 x 1011' dpa/sec 9.14 x 10-11 8.48 x 10-11 (a) Averaged over the plant specific exposure for capsule U.

(b)

Low leakage _ operation (c)' n/cm2 - sec 0321D:1d/011990 6-14

i TABLE 6-2 i

CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRES $URE VESSEL CLAD / BASE METAL INTERFACE _

I AVERAGED OVER CYCLES 1-6 1

0' 15' 30' 45' 1

1

+(E) 1.0Mev)(a) 4.65 x 1010 2.32 x 10 0 1.35 x 10 0 9.35 x 109

$(E) 0.1Mev)(a) 1.22 x 1011 6.66 x 1010

?.39 x 1010 2.48 x 1010 dpa/sec 7.57 x 10-1I 3.92 x 10-11 2.19 x 10-Il 1.53 x 10-Il MIBMED OVER LOWlEAKAGE CYCLES (4-6)

O' 15' 30' 45'

$(E) 1.0Mev)(a) 4.48 x 1010 2.24 x 1010 1.30 k 10 0 9.17 x 109 1

$(E) 0,1Mev)(a) 1.17 x 1011 6.44 x 10 0 3.27 x 10 0 2.44 x 10 0 1

1 1

dpa/sec 7.29 x 10-11 3.79 x 10-11 2.11 x 10-Il 1.53 x 10-11 (a) n/cm2 - sec 03210:1d/011990 6-15

TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV)

WITHIN THE PRESSURE VESSEL HALL Radius (cm)

O' 15' 30' 45' 199.87(I) 1.00 1.00 1.00 1.00 200.91 0.916 0.925 0.914 0.921 202.30 0.788 0.804 0.785 0.797 203.74 0.657 0.679 0.656 0.674 205.13 0.545 0.570 0.543 0,565 206.52 0.448 0.477 0.448 0.470 207.91 0.367 0.396 0.368 0.389 209.30 0.300 0.324 0.301 0.320 210.69 0.244 0.267 0.246 0.263 212.07 0.197 0.219 0.200 0.216 213.46 0.160 0.179 0.163 0.176 214.85 0.129 0.145 0.132 0.144 216.24 0.103 0.117 0.106 0.117 217.63 0.0804 0.0938 0.0849 0.0939 218.86 0.0625 0.0757 0.0688 0.0777 219.87(2) 0.0490 0.0615 0.0570 0.0659 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius 0321D:1d/011990 6-16

i i

TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 HeV)

WITHIN THE PRESSURE VESSEL HALL 1

Radius

{

(cm)

O' 15' 30' 45' 199.87(I) 1.00 1.00 1.00 1.00 200.91 1.00 1.00 1.00 1.00 202.30 0.962 0.972 0.972 0.970 203.74 0.901 0.919 0.916 0.917 4

205.13 0.835 0.859 0.854 0.858 1

206.52 0.764 0.796 0.790 0.799 207.91 0.698 0.732 0.726 0.736 209.30 0.632 0.669 0.664 0.673 210.69 0.566 0.607 0.602 0.614 212.07 0.507 0.547 0.542 0.558 213.46 0.447 0.489 0.484 0.502 214.85 0.390 0.432 0.428 0.446 216.24 0.333 0.375 0.375 0.393 217.63 0.276 0.320 0.323 0.343 218.86 0.223 0.268 0.276 0.298 219.87(2) 0.190 0.229 0.240 0.261 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius 0321D:1d/011990 6-17

TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa)

WITHIN THE PRESSURE VESSEL HALL Radius (cm) 0*

15' 30' 45' 199.87(I}

1.00 1.00 1.00 1.00 200.91 0.929 0.947 0.934 0.937 202.30 0.835 0.858 0.838 0.847 203.74 0.734 0.766 0.735 0.753 205.13 0.644 0.681 0.649 0.668 206.52 0.564 0.606 0.570 0.589 207.91 0.493 0.536 0.500 0.522 209.30 0.430 0.473 0.437 0.459 210.69 0.374 0.416 0.384 0.404 212.07 0.324 0.365 0.334 0.355 213.46 0.280 0.319 0.292 0.311 214.85 0.238 0.274 0.253 0.271 216.24 0.200 0.236 0.217 0.235 217.63 0.164 0.199 0.184 0.202 218.86 0.132 0.166 0.157 0.175 I2' 219.87 O.109 0.140 0.134 0.153 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius l

1 03210:1d/011990 6-18

TABLE 6-6 NJCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS o

Reaction Target Monttor of Weight

Response

Product Material Interest FractinD Range Half-Life Copper Cu63(n.a)Co60 0.6917 E> 4.7 MeV 5.272 yrs Iron Fe54(n.p)Mn54 0.058 E> 1.0 MeV 312.2 days Nickel N1 58(n.p)CoS8 0.6827 E> 1.0 MeV 70.91 days Uranium-238*

U238(n.f)Csl37 1.0 E> 0.4 MeV 30.17 yrs 6.0 j

Neptuntum-237' Np237(n.f)Csl37 1.0 E> 0.8 MeV 30.17 yrs 6.5 Cobalt-Aluminum

• CoS9(n,y)Co60 0.0015 0.4ev<E< 0.015 MeV 5.272 yrs Cobalt-Aluminum

• CoS9(n,y)Co60 0.0015 E< 0.015 HeV 5.272 yrs

• Denotes that monitor is cadmium shielded.

I W

i t

03210:1d/011990 6-19

i TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS i

__ CONTAINED IN CAPSULE U Irradiation Decay Month Year PJ (MW)

PJ/PMAX Time (days)

Time (days) 6 1980 3.7 0.0013 19 3369 7

1980 2.3 0.0008 31 3336

+

8 1980 53.0 0.0183 31 3307 9

1980 854.9 0.2955 30 3277 10 1980 2265.8 0.7832 31 3246 11 1980 124.7 0.0431 30 3216 12 1980 1508.8 0.5215 31 3185 1

1981 2373.9 0.8206 31 3154 2

1981 2702.3 0.9341 28 3126 3

1981 2605.2 0.9005 31 3095 4

1981 2619.8 0.9055 30 3065 5

1981 1611.0 0.5569 31 3034 6

1981 1673.8 0.5786 30 3004 7

1981 0.4 0.0001 31 2973 8

1981 753.9 0.2606 31 2942 9

1981 2504.0 0.8655 30 2912 10 1981 2552.4 0.8823 31 2881 11 1981 2711.4 0.9372 30 2851 12 1981 2608.4 0.9016 31 2820 1

1982 2152.0 0.7439 31 2789 2

1982 2581.4 0.8923 28 2761 3

1982 526.4 0.1819 31 2730 4

1982 0.0 0.0000 30 2700 5

1982 0.0 0.0000 31 2669 6

1982 1334.3 0.4612 30 2639 7

1982 517.0 0.1787 31 2608 8

1982 54.3 0.0188 31 2577 9

1982 2616.9 0.9046 30 2547 10 1982 2758.3 0.9534 31 2516 11 1982 2640.8 0.9128 30 2486 12 1982 2678.1 0.9257 31 2455 1

1983 2080.3 0.7191 31 2424 2

1983 2639.3 0.9123 28 2396 3

1983 2709.2 0.9365 31 2365 4

1983 87.5 0.0303 30 2335 5

1983 72.5 0.0251 31 2304 6

1983 1782.0 0.6160 30 2274 7

1983 2631.3 0.9095 31 2243 8

1983 2754.7 0.9522 31 2212 9

1983 2757.8 0.9533 30 2182 10 1983 2763.5 0.9552 31 2151 11 1983 2768.0 0.9568 30 2121 12 1983 2563.4 0.8861 31 2090 1

1984 2630.2 0.9092 31 2059 03210:1d/011990 6-20

TABLE 6 7 (continued)

IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U o

Irradiation Decay i

Month Year PJ (HW)

PJ/PHAX Time (days)

Time (days) 2 1984 2118.4 0.7322 29 2030 3

1984 2548.2 0.8808 31 1999 4

1984 1722.5 0.5954 30 1969 5

1984 2458.5 0.8498 31 1938 6

1984 2663.8 0.9208 30 1908 7

1984 2390.7 0.8264 31 1877 8

1984 164.6 0.0569 31 1846 9

1984 0.0 0.0000 30 1816 10 1984 0.0 0.0000 31 1785 11 1984 1942.6 0.6715 30 1755 12 1984 1974.5 0.6825 31 1724 1

1985 2773.3 0.9586 31 1693 2

1985 1977.6 0.6836 28 1665 3

1985 1943.5 0.6718 31 1634 4

1985 2154.9 0.7449 30 1604 5

1985 2711.7 0.9373 31 1573 6

1985 2761.0 0.9544 30 1543 7

1985 2297.2 0.7940 31 1512 8

1985 2731.2 0.9441 31 1481 9

1985 2737.2 0.9461 30 1451 10 1985 2001.2 0.6917 31 1420 11 1935 2769.5 0.9573 30 1390 12 1985 2762.4 0.9548 31 1359 1

1986 2752.6 0.9515 31 1328 2

1986 1709.2 0.5908 28 1300 3

1986 0.0 0.0000 31 1269 4

1986 1490.0 0.5150 30 1239 5

1986 2580.4 0.8919 31 1208 6

1986 2597.7 0.8979 30 1178 7

1986 2661.3 0.9199 31 1147 8

1986 1576.4 0.5449 31 1116 9

1986 2867.1 0.9911 30 1086 10 1986 2461.1 0.8507 31 1055 11 1986 2836.6 0.9805 30 1025 12 1986 2891.8 0.9996 31 994 1

1987 2890.7 0.9992 31 963 2

1987 1782.9 0.6163 28 935 3

1987 2832.9 0.9792 31 904 4

1987 2889.0 0.9986 30 874 5

1987 2110.7 0.7296 31 843 6

1987 2552.2 0.8822 30 813 7

1987 2342.8 0.8098 31 782 8

1987 1408.8 0.4870 31 751 9'

1987 0.0 0.0000 30 721 0321D:1d/011990 6-21

TABLE 6-7 (continued)

IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U e

Irradiation Decay Month Year PJ (MW)

PJ/PMAX Time (days)

Time (days) 10 1987 0.0 0.0000 31 690 11 1987 1936.7 0.6695 30 660 12 1987 2892.8 0.9999 31 629 1

1988 2892.1 0.9997 31 598 2

1988 2528.6 0.8740 29 569 3

1988 2891.1 0.9993 31 538 4

1988 2887.6 0.9981 30 508 5

1988 2891.0 0.9993 31 477 6

1988 2892.5 0.9998 30 447 7

1988 2892.4 0.9998 31 416 8

1988 2890.5 0.9991 31 385 9

1988 2859.0 0.9882 30 355 10 1988 2895.5 1.0009 31 324 11 1988 2869.4 0.9918 30 294 12 1988 2783.4 0.9621 31 263 1

1989 2260.4 0.7813 31 232 2

1989 1848.5 0.6390 20 212 NOTE:

1) Pmax - 2893 HWt

2) Decay time is referenced to a dosimetry counting date of 9/20/89.

I 1

I 0321D:1d/011990 6-22 1

o TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES 5-Measured Saturated (a)

Reaction Monitor and Activity Activity Rate Axial Location Ba/am-Ba/am (ros/ nucleus) t Cu-63 (n.a) Co-60 5

5 Top-Middle 1.24 x 10 2.62 x 10 5

5 Middle 1.32 x 10 2.78 x 10 5

5 Bottom-Middle 1.30 x 10 2.74 x 10 5

5 Average 1.29 x 10 2.71 x 10 4.14 x 10-I7 Fe-54 (n.0) Mn-54 6

6 Top 1.03 x 10 2.22 x 10 5

6 Top-Middle 9.53 x 10 2.05 x 10 0

6 Middle 1.03 x 10 2.22 x 10 6

6 Bottom-Middle 1.00 x 10 2.15 x 10 6

6 Bottom 1.02 x 10 2.20 x 10 0

6 Average 1.01 x 10 2.17 x 10 3.45 x 10-15 N1-58 (n.0) Co-58 6

7 Top-Middle 2.63 x 10 2.99 x 10 6

7 Bottom-Middle 2.82 x 10 3.21 x 10 6

7 Bottom 2.53 x 10 2.88 x 10 6

7 Average

~

2.66 x 10 3.02 x 10 4.32 x 10-15 1 o l.

l l

03210:Id/011990 6-23 I

TABLE 6-8 (Cont'd)

MEASURED SENSOR ACTIVITIES AND REACTION RATES ii Measured Saturated Reaction Monitor and Activity Activity (a)

Rate Axial Location Ba/gm Ba/am (ros/ nucleus)

Mp-jl37 (n.f) Cs-132 6

7 Middle 2.89 x 10 2.30 x 10 1.39 x 10-13 U-238 (n.f) Cs-137 5

6 Middle 3.53 x 10 2.80 x 10 1.85 x 10-I4 Co-59 (n.v) Co-60 7

7 Top 1.42 x 10 3.12 x 10 7

7 Bottom 1.36 x 10 2.99 x 10 7

7 Average 1.39 x 10 3.05 x 10 1.99 x 10-12 Co-59 (n.v) Co-60 (Cd1 6

7 Top 6.03 x 10 1,32 x 10 6

7 Bottom 5.39 x 10 1.18 x 10 6

7 Average 5.71 x 10 1.25 x 10 8.18 x 10-13 (a) Adjusted to center of surveillance capsule.

l l

[

1 l

1 l

03210:1d/011990 6-24

TABLE 6-9

SUMMARY

OF NEUTRON DOSIMETRY RESULTS b

TIME AVERAGED EXPOSURE RATES 4

+ (E > 1.0 MeV) (n/cm2-sec}

4.98 x 1010 1 8%

$ (E > 0.1 MeV) {n/cm2-sec}

1.62 x 1011 i 15%

dpa/sec 8.19 x 10-11 i 10%

$(E < 0.414 eV) {n/cm2-sec}

4.89 x 1010 1 19%

i INTEGRATED CAPSULE EXPOSURE e (E > 1.0 MeV) (n/cm2}

9.55 x 1018 i 8%

e (E > 0.1 MeV) {n/cm2}

3.11 x 10 9 1 15%

1 dpa 1.57 x 10-2 1 10%

e (E < 0.414 eV) {n/cm2}

9.37 x 1018 1 19%

NOTE: Total Irradiation Time - 6.07 EFPY I

'I 9

03210:1d/011990 6-25

TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER Adjusted Reaction Measured Calculation CIM Cu-63 (n ct) Co-60 4.14 x 10'I7 4.14 x 1(~I7 1.00 Fe-54 (n.p) Hn-54 3.45 x 10-15 3.44 x 10-15 1.00 N1-58 (n.p) Co-58 4.32 x 10-15 4.43 x 10-15 1.03 U-238 (n. y) Cs-137 (Cd) 1.85 x 10'I4 1.71 x 10-I4 0.93 Np-237 (n f) Cs-137 (Cd) 1.39 x 10-13 1.37 x 10'I3 0.99 Co-59 (n,y) Co-60 1.99 x 10-12 1.98 x 10-12 1.00 Co-59 (n,y) Co-60 (Cd) 8.18 x 10-13 8.19 x 10-13 1.00 e

4 l

l l

03210:Id/011990 6-26

TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER O

Energy Adjusted Flux Energy Adjusted Flux 2

2 i-Group (MeV)

(n/cm -sec)

Group (MeV)

(n/cm,3,g) 1 1.733E+01 6.29E+06 28 9.119E-03 6.02E+09 2

1.492E+01 1.40E+07 29 5.531E-03 7.45E+09 3

1.350E+01 5.31E+07 30 3.355E-03 2.29E+09 4

1.162E+01 1.15E+08 31 2.839E-03 2.15E+09 5

1.000E+01 2.49E+08 32 2.404E-03 2.06E+09 6

8.607E+00 4.13E+08 33 2.035E-03 5.95E+09 7

7.408E+00 9.38E+08 34 1.234E-03 5.80E+09 8

6.065E+00 1.30E+09 35 7.485E-04 5.61E+09 9

4.966E+00 2.57E+09 36 4.540E-04 5.44E+09 10 3.679E+00 3.17E+09 37 2.754E-04 5.71E+09 11 2.865E+00 6.33E+09 38 1.670E-04 6.00E+09 12 2.231E+00 8.02E+09 39 1.0135-04 6.02E+09 13 1.738E+00 1.04E+10 40 6.144E 05 6.02E+09 14 1.353E+00 1.03E+10 41 3.727E-05 5.94E+09 15 1.108E+00 1.73E+10 42 2.260E-05 5.84E+09 l

16 8.208E-01 1.82E+10 43 1.371E-05 5.73E+09 l

17 6.393E-01 1.74E+10 44 8.315E-06 5.56E+09 l

18 4.979E-01 1.23E+10 45 5.043E-06 5.27E+09 19 3.877E-01 1.57E+10 46 3.059E-06 5.03E+09 20 3.020E-01 1.79E+10 47 1.855E-06 4.74E+09 21 1.832E-01 1.63E+10 48 1.125E-06 3.87E+09 l

22 1.111E-01 1.28E+10 49 6.826E-07 4.39E+09 23 6.738E-02 9.19E+09 50 4.140E-07 7.51E+09 24 4.087E-02 5.58E+09 51 2.511E-07 7.92E+09 25 2.554E-02 6.19E+09 52 1.523E-07 8.17E+09 26 1.989E-02 3.64E+09 53 9.237E-08 2.53E+10 27 1.503E-02 4.89E+09 NOTE:

Tabulated energy levels represent the upper energy of each group.

0321D:1d/011990 6-27

TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE U 6

Calculated Measured Cig 2

• (E) 1.0 MeV) {n/cm }

1.06 x 10I9 18 9.55 x 10 j,jj 2

• (E) 0.1 MeV) {n/cm }

3.22 x 10I9 3.11 x 10I9 1.04 dpa 1.75 x 10-2 1.57 x 10-2

),jj 2

18 18

• (E< 0.414 eV) (n/cm }

6.74 x 10 9.37 x 10 0.72

+

03210:1d/011990 6-28

TABLE 6-13 ON THE PRESSURE VE EL CLAD / BASE META INTERFACE AZIMUTHAL ANGLE _

0+(a) 15' 30' 45' 6.07 EFPl 1

1 1

1 e (E) 1.0 MeV)(b) 8.02 x 10 8 4.00 x 10 8 2.33 x 10 8 1.61 x 10 8 e (E) 0.1 MeV)(b) 2.26 x 1019 1.23 x 1019 6.28 x 1018 4.59 x 1018 dpa 1.30 x 10-2 6.73 x 10-3 3.76 x 10-3 2.66 x 10-3 31.0 EFPY e (E) 1.0 MeV)(b) 4.47 x 1019 2.23 x 1019 1.30 x 1019 9.11 x 1018 1

1 1

e (E) 0.1 MeV)(b) 1.18 x 1020 6.50 x 10 9 3.30 x 10 9 2.46 x 10 9 dpa 7.27 x 10-2 3.77 x 10-2 2.10 x 10-2 1.52 x 10-2 (a) Maximum point on the pressure vessel (b) n/cm2 0321D:1d/011990 6-29

TABLE 6-14 VESSEL NEUTRON EXPOSURE VALUES (E>1.0 MeV) FOR USE IN THE GENERATION OF HEATUP/COOLDOWN QJRVJS 15 EFPY NEUTRON FLUENCE (E) 1.0 MeV) SLOPE dp_glLQP_E 2

(n/cm )

g,7c,2)

Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T I9 I9 18 I9 I9 18 0*

• 2.06 x 10 1.17 x 10 2.65 x 10 2.06 x 10 1.36 x 10 4.90 x 10 I9 18 18 18 18 18 15*

1.03 x 10 6.08 x 10 1.49 x 10 1.03 x 10 7.17 x 10 2.83 x 10 18 18 18 18 18 18 30*

5.99 x 10 3.38 x 10 7.91 x 10 5.99 x 10 3.99 x 10 1.52 x 10 18 18 37 18 18 18 45' 4.19 x 10 2.45 x 10 6.03 x 10 4.19 x 10 2.87 x 10 1.14 x 10 3

en

.a 0,

o 32 EFPY NEUTRON FLUENCE (E) 1.0 MeV) SLOPE dealLQPE 2

(n/cm )

g,fc,2)

_Sur_fE g_

1/4 T 3/4 T Surface 1/4 T 3/4 T 0,(a) 4.47 x 10 2.53 x 10 5.77 x 10 4.47 x 10 2.95 x 10 1.06 x 10 I9 I9 18 I9 I9 I9 I9 I9 18 I9 18 15*

2.23 x 10 1.32 x 10 3.24 x 10 2.23 x 10 1.56 x 10 6.11 x 10 I9 18 18 I9 18 18 30*

1.30 x 10 7.33 x 10 1.71 x 10 1.30 x 10 8.65 x 10 3.29 x 10 18 I8 18 18 18 18 45' 9.11 x 10 5.34 x 10 1.31 x 10 9.11 x 10 6.24 x 10 2.47 x 10 (a) Maximum point on the pressure vessel s*

t_..

i TABLE 6-15 UPDATED LEAD FACTORS FOR UNIT 2 SURVEILLANCE CAPSULES l

l Capsule Lead Factor

[

V 1.66 (removed)

X 1.72 U

1.19 (removed) 1.19 g H

Y 1.19 T

0.81 Z

0.81 S

0.65 C

03210:1d/011990 6-31

1 t

SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the North Anna Unit 2 reactor vessel:

j i

Estimated y,g3,1 Capsule Removal Fluence Location Lead Capsule (deo)-

. Factor Time (a)

(n/cm )

2 V

165 1.66 1.03 (removed) 2.55 x 1018 (actual)

U

'65 1.19 6.07 (removed) 9.55 x 1018(actual) _

19[b]

[

W 245 1.19 15 2.49 x 10 19IC3

-X 285 1,72 18.6 4.47 x 10 Y

295 1.19 30 4.987 x 1019Idl T

55 0.81 Standby o

2 305 0.81 Standby S

45 0.65 Standby a)

Effective full power years from plant startup b)

Approximate fluence at vessel 1/4 thickness at end of life (32 EFPY) c) Apprcximate fluence at vessel surface at end of life (32 EFPY)-

d).To be removed at a fluence not-less than once or greater than twice the end of life surface fluence (32 EFPY) h

' 407ts C:D9416 7,}

..~.. -

SECTION 8 I

REFERENCES

.O

1. Davidson, J. A., Yanichko, S. E.,

Phillips, J. H., " Virginia Electric and Power North Anna Unit 2 Reactor Vessel Radiation Surveillance Program,*

[

WCAP-8772, November 1974.

2. BAW-1794, " Analysis of Capsule V Virginia Electric & Power Company North Anna Unit No. 2." October 1983.

3. Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements" and Appendix H. " Reactor Vessel Material Surveillance Program Requirements", U. S. Nuclear Regulatory Commission, Washington, D.C.

f 1

4. Regulatory Guide 1.99, Revision 2. " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May,1988.

5. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, o

Philadelphia, PA,1984.

6. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970.

7. "0RNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self-Shieldeo, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Wat'er l

Reactors",

t l.

nn, '"""'

8-1

t

8. Furchi, C. L., et.al, " North Anna Units 1 and 2 Reactor Vessel Fluence and RT Evaluations," WCAP-11016 Rev.1, December 1985 [ Proprietary).

PTS

9. ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, g

Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

10. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Desimetry Results", in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA,1984.

11. ASTM Designation E706-Bla, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

12. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984, i

13. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

14. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 198a.

15 ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

i m, " " "

8-2

i

16. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

17. ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

18. ASTM Designation E523-82, " Standard Method for Determining Fast-Neutron l

Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

19. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American t

Society for Testing and Materials, Philadelphia, PA, 1984 2D. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM Standards, Section a

12, American Society for Testing and Materials, Philadelphia, PA, 1984.

t O

21. ASTM Designation E1005-84, " Standard Method for A;;1ication and Analysis of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM l

Standards, Section 12, American Society for Testing and Materials, l

Philadelphia, PA, 1984.

22. F. A. Schmittroth, FERRET Data Analysis Code, HEDL-THE 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

23. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Itera'tive l

Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-67-41, Vol. 1-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.

s

24. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology l

for LWR Desimetry Applications", R. E. Maerker, et al.,1981.

I un,unn o g.3

-