ML20059K816

Analysis of Capsule U from South Texas Unit 1 Reactor Vessel Radiation Surveillance Program

ML20059K816
Person / Time
Site:	South Texas
Issue date:	08/31/1990
From:	Madeyski A, Terek E, Wrights G WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20059K802	List: ML20059K816 ST-HL-AE-3553, Forwards WCAP-12629, Analysis of Capsule U from South Texas Unit 1 Reactor Vessel Radiation Surveillance Program. Pressure-temp Curves Currently in Use in Facility Tech Specs Are More Conservative than Presented in App a of Rept
References
WCAP-12629, NUDOCS 9009240260
Download: ML20059K816 (87)

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'I WESTINGHOUSE CLASS 3 WCAP-12629 ANALYSIS OF CAPSULE U FROM THE HOUSTON LIGHTING AND POWER COMPANY SOUTH TEXAS UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM E. Terek G. N. Wrights A. Madeyski l

J. M. Chicots August 1990 Work Performed Under Shop Order UGXP-6620 Preparrd by Westinghouse Electric Corporation ist the Houston Lighting and Power Company k

Approved by:

T. A. Meyer, M4 nager Structural Materials and Reliability Technology WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728

@ 1990 Westinghouse Electric Corp.

06490:10/081790 1

~

i i

i s

PREFACE i

This report has been technically reviewed and verified, i

Reviewer N II<

Sections 1 through 5, 7, and 8 J. M. Chicots Section 6 S. L. Anderson Ndd O*?irNL%

Appendix A N. K. Ray 7, (\\ erv/ /)

v i

l 0649D:1D/070190 i

t TABLE OF CONTENTS Section Title Page 1.0

SUMMARY

OF RESULTS 1-1 2.0

!NTRODUCTION 21 i

3.0 BACKGROUND

31

4.0 DESCRIPTION

OF PROGRAM 41 l

l 5.0 TESTING OF SPECIMENS FROM CAPSVLE V 3-1 1

5.1 Ove-view 5-1 5.2 Charpy V-Notch Impact Test Results 53 5.3 Tension Test Results 55 5.4 Compact Tension Tests 55 6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 61 6.1 Introduction 61 6.2 Discrete Ordinates Analysis 62 i

6.3 Neutron Dosimetry 6-8 7.0 SVRVEILLANCE CAPSVLE REMOVAL SCHEDULE 7-1

8.0 REFERENCES

8-1 06490:lD/070390 ii

i LIST OF ILI.USTRATIONS figure Title Page 4-1 Arrangement of surveillance capsules in the South Texas 4-5 Unit I reactor vessel 4-2 Capsule U diagram showing location of specimens, thermal 4-6 monitors and dosimeters i

5-1 Charpy V-notch impact properties for South Texas Unit 1 5 13 reactor vessel intermediate shell plate R1606-2 (transverse orientation) 52 Charpy V-notch impact properties for South Texas Unit 1 5-14 reactor vessel intermediate shell plate R1606-2 (longitudinal orientation) 5-3 Charpy V-notch impact properties for South Texas Unit 1 5-15 reactor vessel core region weld metal 5-4 Charpy V-notch impact properties for South Texas Unit 1 5-16 reactor vessel core region weld heat affected zone metal 5-5 Charpy impact specimen fracture surfaces for South Texas 5-17 Unit I reactor vessel intermediate shell plate R1606 2 (transverse orientation) 5-6 Charpy impact specimen fracture surfaces for South Texas 5-18 Unit I reactor vessel intermediate shell plate R1606-2 (longitudinal orient.ation) 5-7 Charpy impact specimen fracture surfaces for South Texas 5-19 Unit I reactor vessel core region we d metal 0649D:10/070390 tii

I LIST OF ILLUSTRATIONS (Cont)

Figure Title Page 58 Charpy impact specimen fracture surfaces for South Texas 5-20 Unit I reactor vessel core region weld heat affected zone (HAZ) metal 5-9 Tensile properties for South Texas Unit I reactor vessel 5 21 intermediate shell plate R1606-2 (transverse orientation) 5-10 Tensile properties for South Texas Unit I reactor vessel 5-22 intermediate shell plate R1606-2 (longitudinal orientation) 5 11 Tensile properties for South Texas Unit I reactor vessel 5-23 core region weld metal 5 12 Fractured tensile specimens from South Texas Unit 1 5-24 reactor vessel intermediate shell plate R1606-2 (transverse orientation) 5-13 Fractured tensile specimens from South Texas Unit 1 5-25 reactor vessel intermediate shell plate R1606-2 (longitudinal orientation) 5-14 Fractured tensile specimens from South Texas Unit 1 5 26 reactor vessel core region weld metal i

5-15 Typical stress-strain curve for Houston Light and Power 5-27 i

i Company South Texas Station Unit 1 intermediate shell plate R1606-2 tension specimens 6-1 Plan view of a dual reactor vessel surveillance capsule 6 13 6-2 Core power distributions used in transport calculations 6-14 i

for South Texas Unit 1 4

L 06490:1D/070390 iv

LIST OF TABLES Table Title Page 4-1 Chemical Composition and Heat Treatment of the South Texas 4-3 Unit 1 Reactor Vessel Surveillance Materials 42 South Texas Unit 1 Reactor Vessel Toughness Data 4-4 5-1 Charpy V-Notch Impact Data for the South Texas Unit 1 56 Intermediate Shell Plate R1606-2 Irradiated at 550'F, 18 Fluence 2.93 X 10 n/cm2 (E > 1.0 MeV) 52 Charpy V Notch Impact Data for the South Texas Unit 1 5-7 Reactor Vessel Core Region Weld Metal and HAZ Hetal 18 2

Irradiated at 550'F, Fluence 2.93 X 10 n/cm (E > 1.0 MeV) 53 Instrumented Charpy Impact Test Results for South Texas 58 Unit 1 Intermediate Shell Plate R1606-2 Irradiated at IO 550'F, Fluence 2.93 X 10 n/cir,2 (E > 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results for South Texas 59 Unit 1 Reactor Vessel Core Region Weld Metal and HAZ Metal 18 2

Irradiated at 550'F, Fluence 2.93 X 10 n/cm l

(E > 1.0 MeV) 18 2

5-5 Effect of 550'F Irradiation at 2.93 X 10 n/cm 5-10 (E > 1.0 MeV) on Notch Toughness Properties of South Texas Unit 1 Reactor Vessel Surveillance Materials i

06490:10/070390 v

LIST OF TABLES (Cont)

Table Title Page 5-6 Comparison of Seth Texas Unit 1 Reactor Vessel Surveillance 5-11 Material 30 ft lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-7 Tensile Properties for South Texas Unit 1 Reactor Vessel 5-12 18 Surveillance Material Irradiated at 550'F to 2.93 X 10 n/cm2 (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Parameters at the 6-15 Surveillance Capsule Center u

6-2 Calculated Fast Neutron Exposure Parameters at the 6 16 Pressure Vessel Clad / Base Metal Interface 6-3 Relative Radial Distributions of Neutron Flux 6-17 (E > 1.0 MeV) within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-18 (E > 1.0 MeV) within the Pressure Vessel Wall 6-5 Relative Radial Distributions of Iron Displacement Rate 6-19 (dpa) within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-20 l

67 Irradiation History of Neutron Sensors Contained in Capsule U 6-21 6-8 Measured Sensor Activities and Reactions Rates 6-22 6-9 Summary of Neutron Dosimetry Results 6-24 i

06490:10/070390 vi

LIST OF TABLES (Cont)

Table Title Pige 6-10 Comparison of Measured and Ferret Calculated Reaction 6-25 Rates at the Surveillance Capsule Center g~;

6 Adjusted Neutron Energy Spectrum at the Surveillance 6-26 Capsule Center 6-12 Comparison of Calculated and Measured Exposure Levels for 6-27 Capsule v 6-13 Neutron Exposure Projections at Key Locations on the 6 28 Pressure Vessel Clad / Base Metal Interface 6-14 Azimuthal variation of the Nuetron Exposure Projections 1

on the Pressure Vessel CLAD / BASE Metal Interface 6-30 6-15 Neutron Exposure Values for use in the Generation of 6-31 Heatup/CooldownCurves 6-16 Updated lead Factors for South Texas Unit 1 Surveillance 6-32 Capsules l

1 i

i I

.0649D:1D/070190 yij l

l

SECTION 1.0

SUMMARY

OF RESULTS i

The analysis of the reactor vessel material contained in surveillance Capsule U, the first capsule to be removed from the Houston Lighting anJ Power Company South Texas Unit I reactor pressure vessel, led to the following conclusions:

o.

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

of 2.93 X 10 n/cm after 0.78 EFPY of plant operation.

l o

Irradiation of the reactor vessel intermediate shell plate R1606-2 18 Charpy specimons to 2.93 X 10 n/cm2 (E > 1.0 MeV) resulted in I

a 30 ft-lb transition temperature increase of 30'F and a 50 ft-lb transition temperature increase of 25'F for specimens oriented perpendicular to the major working direction (transverse orientation).

o Irradiation of the reactor vessel intermediate shell plate R1F06 2 18 Charpy specimens to 2.93 X 10 n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 10*F and a 50 ft-lb transition temperature increase of 15'F for specimens oriented parallel to the major working direction (longitudinal orientation).

i o

Irradiation of the reactor vessel core region weld metal Charpy 18 specimens to 2.93 X 10 n/cm2 (E > 1.0 MeV) resulted in a 30 I

ft-lb transition temperature increase of 20*F and a 50 ft-lb i

transition temperature increase of 15'F.

I i

o Irradiation of the reactor vessel core region weld HAZ metal Charpy 10 specimens to 2.93 X 10 n/cm2 (E > 1.0 MeV) resulted in no 30 l

and 50 ft-lb transition temperature increases.

o The average upper shelf energy of the intermediate shell plate R1606-2 showed a decrease in energy of 8 ft-lb (transverse 18 orientation) after irradiation to 2.93 X 10 n/cm2 (E > 1.0 MeV).

The core region-weld metal showed no decrease in upper shelf 06490:10/081790 1-1

energy after irradiation to 2.93 X 10 n/cm2 (E > 1.0 MeV).

18 Both materials exhibit a more than adequate upper shelf energy level for' cot.Linued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life of

~the vessel as required by 10CFR50, Appendix G.

o Cor.parison of the 30 ft-lb transition temperature increases for the S'auth Texas Unit 1 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2, demonstrated that the plate R1606-2 material (transverse orientation) and weld metal transition temperature increases were 13'F and 2*F, respectively, greater than predicted.

NRC Regulatory Guide 1.99, Revision 2 requires a 2 sigma allowance, of 34*F for-base metal and 56'F for weld metal, be added to the predicted reference transition temperature to obtain a conservative upper bound value. Thus, the reference transition temperature increases for Plate R1606-2 material and the weld metal are bounded by the 2 sigma allowance for shift prediction.

0649D:1D/070190 1-2

l-SECTION

2.0 INTRODUCTION

This report presents the results of the examination of Capsule U, the first capsule to be removed from the reactor in the continuing surveillance program

-which monitors the effects of neutron irradiation on the South Texas Unit 1-reactor pressure vessel materials under actual operating conditions.

The surveillance program for the South Texas Unit I reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented by Kaiser, Koyama and Davidson.Ill '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 Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule "U" from the reactor and its shipment to the Westinghouse Science and Technology Center where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens were performed at the remote metallographic facility.

This report summarizes the testing of and the post-irradiation data obtained

~ from surveillance Capsule "V" removed from the South Texas Unit I reactor vessel and discusses the analysis of these data.

06490:10/081790 2-1

SECTION 3.0 BACKGR0VND 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 l

pressure vessel is the most critical region of the vessel because it is

{

subjected to significant fast neutron bombardment. The overall effects of L

fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 533 Grade B Class 1 (base material of the Houston Lighting and Power Company South Texas Unit I reactor pressure vessel intermediate shell plate) are well documented in the literature.

Generally, low alloy ferritic materials show an increase in hardness and tensile I

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 have been presented in " Protection Against Nonductile f

Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel CodeI43 The method uses fracture mechanics concepts and is based on the I

reference nil-ductility temperature (RTNDT)*

I RT is defined as the greater of either the drop weight nil-ductility NDT transition temperature (NDTT per ASTM E-208{5]) 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 material.

The RT of a given material 'is used to index NDT that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The KIR 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 Kip curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

0649D:1D/081790 3-1

RT and, in turn, the operating limits of nuclear power plants can be HDT adjusted to account for the effects of radiation on the reactor vessel material properties.

The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Houston Lighting and Power Company South Texas Unit 1 Reactor Vessel Radiation Surveillance Program,bl3 in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested.

The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to to adjust the RT fr irradiation is added to the original RTNDT NDT initial + ARTNDT) radiation embrittlement.

This adjusted RTNDT (RTNDT is used to index the material to the K curve and, in turn, to set IR operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

06490:10/081790 3-2

SECTION

4.0 DESCRIPTION

OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the South Texas Unit I reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron shield pads and the vessel wall as shown in figure 4-1.

The vertical center of the capsules is opposite the vertical center of the core.

I Capsule U was removed after 0.78 effective full power years of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact tension (CT) specimens (figure 4-2) from the intermediate shell plate R1606-2 and weldment made from sections of intermediate shell plates R1606-2 and R1606-3 using weld identical to that used in the original vessel fabrication f

for all core region weld seams and Charpy V-notch specimens from weld heat-affected zone (HAZ) material. All heat-affected zone specimens were obtained from within the HAZ of plate R1606-2.

The chemical composition, heat treatment and toughness data of the surveillance material are presented in Tables 4-1 through 4-2.

The chemical analyses' reported-in Table 4-1 were obtained from unirradiated material used in the survelliance program. Table 4-2 contains the toughness data for the reactor vessel materials.

i All test specimens were machined from the 1/4 thickness-location of the plate. Test specimens represent material ta' ken at least one plate thickness from the quenched end of the forging.

Base metal Charpy V-notch impact and tension specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction (tangential orientation) and also perpendicular to the major working direction (transverse orientation).

Charpy V-notch and tensile specimens from the weld metal were machined such that the longitudinal axis of the specimen was normal to the welding direction. The base metal Compact Tension (CT) specimens in Capsule U were machined in both the transverse and longitudinal orientations. The CT specimens from the weld metal were machined normal to the weld direction with the notch oriented in the direction of the weld.

06490:lD/072690 4-1

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Capsule U' contained dosimeter wires of pure copper, iron, nickel, and aluminum 0.15% cobalt wire (radmium shielded and unshielded).

In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were contained in the capsule.

Thermal monitors made from the two low-melting eutectic alloys and sealed in.

Pyrex tubes were included in the capsule. The composition of the two eutectic alloys and their melting points are as follows:

2.5% Ag, 97.5% Pb Melting Point:

579'F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point:

590'F (310'C)

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

0649D:1D/070190 4-2

TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE SOUTH TEXAS UNIT I REACTOR VESSEL SURVEILLANCE MATERIALS Element Plate R1606 2 Weld Metal (a)

C 0.19 0.12 S

0.013 0.010 N

0.008 0.004 2

Co 0.012 0.009 Cu 0.04 0.02 S1 0.19 0.42 Mo 0.53 0.53 Ni 0.61 0.09 Mn 1.18 1.36 Cr 0.03 0.02 I

V 0.004 0.003 P

0.008 0.009 Sn 0.002 0.003 B

<0.001 0.001 Cb

<0.01

<0.01 Ti

<0.01

<0.01 W

<0.01 0.02 As 0.003 0.004 Zr

<0.001

<0.001 Pb Not Detected

<0.001 Ar-0.017 0.008 (a) Weld Wire Type B4, Heat Number V89476, Flux Type Linde 124, and Flux Lot. Number 1061. Surveillance Weldment is from weld between the Inter-mediate Shell Plates R1606-2 and R1606-3.

i

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Material Temperature Time Coolant 0

Intermediate Shell 1600 1 25 F 4 hrs Water quenched 0

Plate R1606-2 1225 1 25 F 4 hrs Air cooled 0

1150 1 25 F 14 hrs 43 min.

Furnace cooled 0

Weld 1150 1 25 F 13 hrs 15 min.

Furnace cooled 06490:1D/062890 4-3

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i Figure'4-1. Arrangement of Surveillance Capsules in The South Texas Unit 1 Reactor Vessel 0649D:1D/070190 4-5 a

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Capsulo U Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters 4-6 I

- _ _ _ _ _ _ _ ____ _ _ _.i

I SECTION 5.0 TESTING 0F SPECIMENS FROM CAPSULE U L

5.1 Overview-j

[

The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center with consultation by Westinghouse Power Systems personnel.

Testing was performed in accordance with 10CFR50, Appendices G and H,[2] ASTM Specification E185[6]

and Westinghouse Procedure MHL-8402, Revision 1 as modified by RMF Procedures 8102, Revision 1 and 8103, Revision 1.

l 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-9492 Ell. No discrepancies were found.

Examination of the two low-melting point 304*C (579'F) and 310'C (590'F) 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 was less than 304'C (579'F).

The Charpy impact tests were performed per ASTM Specification E23-8?I73 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen 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 load-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 M

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 l

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

identified as the arrest load (P )*

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

record and the load-time record.

The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen.

06490:10/081790 5-1

Therefore, the propagation energy for the crack (E ) is the difference q

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

i D

The yield stress (oy) 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.

Percent shear was determined from post-fracture photographs using(8]

the ratio-of-areas methods in compliance with ASTM Specification A370 The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

l Tension tests were performed on a 20,000-pound Instron, split-console test-l93 and E21(10), and RMF j

machine (Model 1115) per ASTM Specification E8 Procedure 8102, Revision 1.

All pull rods, grips, and pins were made of Inconel 718 hardened to HRC 45. 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.

Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were I

spring-loaded to the specimen and operated through specimen failure.

The extensometer length is 1.00 inch.

The extensometer is rated as Class B-2 I

per ASTM E83 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.

i Because of the difficulty in remotely attaching a thermocouple directly to the.

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

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

grip temperatures were used to obtained desired specimen temperatures.

Experiments indicated that this method is nccurate to +2*F.

0649D:10/072690 5-2

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 j

calculate the fracture stress (true stress at fracture) and percent reduction t

in area was computed using the final diameter measurement.

5.2 Charov V-Notch Impact Test Results The results of Charpy V-notch impact tests performed on the various materials 18 contained in Capsule U irradiated to 2.93 X 10 n/cm2 (E > 1.0 MeV) at 550*F are presented in tables 51 through 5-5 and are compared with unirradiated resultsIU as shown in Figures 5-1 through 5-4.

The transition temperature increases and upper shelf energy decreases for the Capsule U l

materials are summarized in Table 5 5.

Irradiation of the reactor vessel intermediate shell plate R1606-2 material IO specimens to 2.93 X 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-1) resulted in a 30 ft-lb transition temperature increase of 30*F and a 50 ft-lb transition temoerature increase of 25'F for specimens oriented perpendicular

- to the major working direction (transverse orientation). This resulted in'a 30 ft-lb transition temperature of 25'F and a 50 ft-lb transition temperature of 55'F for specimens oriented perpendicular to the major working direction (transverse orientation).

The average upper shelf energy (USE) of the intermediate shell plate R1606-2 l

material resulted in a decrease of 8 ft-lb in energy (transverse orientation) 18 after irradiation to 2.93 X 10 n/cm2 (E > 1.0 MeV) at 550*F. This resulted in an USE of 105 ft-lb (Figure 5-1).

Irradiation of the reactor vessel intermediate shell plate R1606-2 material 18 specimens to 2.93 X 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 10*F and a 50 ft-lb transition temperature increase of 15'F for specimens oriented paral 31 to the major working diccetian (1er.gii.udinal orientation).

This resulted ja a 06490:10/081790 5-3 l

30 ft-lb transition temperature of 25'F and a 50 ft-lb transition temperature of O'F for specimens oriented perpendicular to the major working direction

_(longitudinal orientation).

i The average upper shelf energy (USE) of the intermediate shell plate R1606-2 material resulted in a decrease of 5 ft-lb in energy (longitudinal orientation) after irradiation to 2.93 X 10 n/cm2 (E > 1.0 MeV) at 18 550*F.

This resulted in an USE of 132 ft-lb (Figure 5 2).

Irradiation of the reactor vessel core region weld medal Charpy specimens to 2.93 X 10 n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-3) resulted in a 30 I8 ft-lb transition temperature increase of 20'F and a 50 ft-lb transition temperature increase of 15'F.

This resulted in a 30 ft-lb transition temperature of -30*F and a 50 ft-lb transition temperature of -5'F.

o The average upper shelf enargy (USE) of the reactor vessel core region weld 18 metal resulted in an increase of 3 ft-lb after irradiation to 2.93 X 10 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an USE of 88 ft-lb.

Irradiation of the reactor vessel weld Heat-Affected Zone (HAZ) specimens to 2.92 X 10 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-3) resulted no 30 ft-lb l

18 and 50 ft lb transition temperature increases.

The average upper shelf energy (USE) of the reactor vessel HAZ metal resulted in an increase of 14 ft-lb after irradiation to 2.93 X 10 n/cm2 (E > 1.0 18 MeV) at 550'F.

This resulted in an USE of 120 ft-lb.

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

A compa"ison of the 30 ft-lb transition temperature increases for the various South Yexas Unit 1 surveillance materials with predicted increases using the 5

methods of NRC Regulatory Guide 1.99, Revision 2 is presented in Table 5-6.

This comparison indicates that the plate R1606-2 material (longitudinal orientation) transition temperature increast and the upper shelf energy decreases of all surveillance capsule material resulting from irradiation to 0649D:1D/081790 5-4

i I8 2.93 x 10 n/cm2 (E > 1.0 MeV) are less than the guide predictions. This comparison, also, indicates that the plate R1606-2 material (transverse orientation) and weld metal transition temperature increases vere 13*F and 2*F, respectively, greater than predicted.

NRC Regulatory Guide 1.99, Revision 2 requires a 2 sigma allowance, of 34*F for base metal and 56*F for weld metal, be added to the predicted reference transition temperature to obtain a conservative upper bound value.

Thus, the reference transition temperature increases for Plate R1606-2 material (transverse orientation) and weld metal are bounded by the 2 sigma allowance for shift prediction.

5.3 Tension Test Results The results of tension tests performed on the reactor vessel intermediate shell plate R1606-2 (transverse and longitudinal orientation) and the weld 18 2

metal irradiated to 2.93 x 10 n/cm are shown in Table 5-7 and are compared with unirradiated results as shown in Figures 5-9, 5-10 and 5-11.

Plate R1606-2 test results are shown in Figures 5-9 and 5-10 and 18 2

indicated that irradiation to 2.93 x 10 n/cm caused a less than 5 ksi increase in the 0.2 percent offset yield strength and ultimate tensile i

strength. Weld metal tension tests results shown in Figure 5-11, show that i

I the ultimate tensile strength and the 0.2 percent offset yield strength increased by 2 to 7 ksi with irradiation to 2.93 X 1018 2

n/cm. The small increases in 0.2% yield strength and tensile strength exhibited by the plate material and weld metal indicate that these materials are not highly sensitive I8 2

to radiation at 2.93 x 10 n/cm.

The fractured tension specimens for the reactor vessel intermediate shell plate R1606-2 are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14.

A typical stress-strain curve for the tension tests is shown in Figure 5-15, 5.4 Comoact Tension Tests Per the surveillance capsule testing program with the Houston Lighting and Power Company, 1/2 T-compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science & Technology Center Hot Cell.

06490:10/081790 5-5

s TABLE 5 1 CHARFY V-NOTCH IMPACT DATA FOR THE SOUTH TEXAS UNIT 1 INTERMEDIATE SHELL PLATE R1606-2 IRRADIATED AT 550'F, 18 FLUENCE 2.93 x 10 n/cm2 (E > 1.0 MeV)

Temperature Impact Energy Lateral Expansion Shear Sample No. Q'F l'01 (f t-lb) IJ1 (mils)

(as)

(%)

Longitudinal Orientation 10.0 13.5 10.0 0.25 5

GL7

-75 32.0 43.5 22.0 0.56 15 i

GL9

-50 38.0 51.5 26.0-0.66 20 GL15

-35 53.0 72.0 42,0 1.07 35 GL12

-20 33.0 44.5 24.0 0.61 25 GL14 0

39.0 53.0 30.0 0.76 25 GL11 0

y 80.0 108.5 58.0 1.47 50 GL10 15 48.0 65.0 35.0-0.89 40 GLIS 30 GL1 50 86.0 116.5 60.0 1.52 65 GL5 75 114.0 154.5 78.0 1.98 75 GL8 125.

135.0 183.0 88.0 2.24 100 UL6 150 101.0 137.0 68.0 1,73 100 GL2 200 127.0 172.0 83.0 2.11 100 l

GL3 200 139.0 188.5 83.0 2.11 100 i

GL4 225 1

131.0 177.5 77.0 1.96 100 l

J Transverse Orientation 18.0 24.5 13.0 0.33 15 GT14

-50 14.0 19.0 10.0 0.25 10 GT9

-50 13.0 17.5 11.0 0.28 10.

GT10

-20 15.0 20.5 38.0 0.97 20 GT5 15 40.0 54.0 31.0 0.79 25 GT13 20 GT4 35 38.0 51.5 35.0 0.89 20 GT12 65 72.0 97.5 50.0 1.27 40 GT15 75 58.0 78.5 45.0 1.14 40 -

GT2 100 78.0 106.0 59.0 1.50 65 GT7-150 101.0 137.0 71.0 1.80 100 GT8 200 106.0 143.5 70.0 1.78 100 GT1 200-101.0 137.0 70.0 1.78 100 GT3 225 1

104.0 141.0 72.0 1.83 100 GT6 225 1

109.0 148.0 74.0 1.88 100 06490:1D/062890 5-6

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE SOUTH TEXAS UNIT 1 REACTOR VESSEL CORE REGION WELD METAL AND HAZ METAL IRRADIATED AT 18 550*F, FLUENCE 2.93 x 10 n/cm2 (E > 1.0 MeV)

Sample No.

e$

7a Weld Metal 10.0 13.5' 10.0 0.25 10 CW13

-85 8.0 11.0 9.0 0.23 10-GW15

-50 19.0 26.0 15.0 0.38 15 CW4

-35 GW17 0

58.0 78.5 48.0 1.22 50 GW2 20 63.0 85.5 51.0 1.30 65 GW3 45 67.0 91.0 50.0 1.27 85 CW12 45 73.0 99.0 61.0 1,55 90 GW1 75 83.0 112.5 67.0 1.70 95 GW9 75 84.0 114.0 62.0 1.57 95' GW5 100 86.0 116.5 73.0 1.85 100 GW6' 150 84.0 114.0 73.0 1.85 100 GW11 150 91.0 123.5 70.0 1.78 100 GT10 220 1

87.0 118.0 66.0 1.68 100 GW8 220 1

89.0 120.5 69.0 1.75 100 HAZ Metal GH7

-100 68.0 92.0 34.0 0.86 30 GH9

- 75 82.0 111.0 46.0 1.17 35 GH1

- 75 104.0 141.0 59.0 1.50 55 GH5

- 50

-4 17.0 23.0 13.0 0.33 10 al

1. 8

!!:8

!!:8 8:11 e

GH2 25 106.0 143.5 58.0 1.47 60 GH6 25 129.0 175.0

'72.0 1.53 70 GH15 75 78.0 106.0 59.0 1.50 80 12 125 TBST CTION GH8 125 163.0 221.0 72.0 1.83

.100 T

UNCTION GH14 200 105.0 (142.5) 76.0 (1.93) 100 0649D:10/062890 5-7

TABLE 5-3 l

INSTRUMENTEL' CHARPY IMPACT TEST RESULTS FOR THE SOUTd TEXAS UNIT 1 INTERMEDIATE SHELL PLATE R1606-2 IRRADIATED AT 550*F, FLUENCE 2.93 x 1018,7c,2 (E > 1.0 MeV)

Normalized Energies t

Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A-Load to Yield Load Maximum Ioad Lead Stress Stress 2

,irips L (kips)

(ksil (ksi)

Number (*F)

(ft-lb)

(f t-lb/in )

(kips)

(psec)

(kips)

(psec)

(

Longitudinal Orientation CL7

- 75 10.0 81 59 21 3.55 110 3.75 185 3.75 0.15 117 120 CL9

- 50 32.0 258 217 41 3.35 100 4.15 505 4.15 0.15 111 124 CL15

- 35 38.0 306 300-6 4.70 170 5.90 630 5.90 0.00 151 173 GL12

- 20 53.0 427 395 32 3.95 100 5.85 680 5.75 0.20 131 163 CL14 0

33.0 266 261 4

3.80 90 5.50 475 5.50 0.45 126 154 GL11 0

39.0 314 245 69 3.15 120 4.00 585 3.95 0.00 105 119 CLIO 15 80.0 644 389 256 4.25 110 5.70 690 5.05 0.35 141 165 CL13 30 48.0 387 380 7

4.25 100 5.55 675 5.40 0.70 140 162 CL1 50 86.0 692 376 316 4.35 110 5.60 675 5.00 1.20 144 164 CL5 75 114.0 918 460 458 4.05.

100 5.55 825 3.65 1.85 135 159 1

CL8 125 135.0 1087 447 640 3.80 110 5.35 845

-+

-+

125 151 l

CLS 150 101.0 813 256 557 2.70 110 3.70 680

-+

-+

89 106 CL2 200 127.0 1023 302 721 2.65 130 3.65 815

-+

88 104 i

CL3 200 139.0 1119 405 714 3.70 120 5.10 800

-+

-+

122

.146 CL4 225 131.0 1055 427 628 3.45 105 5.05 840

-+

114 141 Transverse Orientation CT9

- 50 14.0 113 70 42 3.45 120 3.65 220 3.55 0.00 113' 117 CT14

- 50 18.0 145 145 0

4.20' 90 5.40 280 5.30 0.15 139

.159 GTIO

- 20 13.0 105 39 65 3.85 100 4.70 130 4.60 0.25 127 141 GT5 15 51.0 411 383 28 3.80 80 5.60 675 5.45 0.55 126 156 CT13 20 40.0 322 210 112 2.95 100 3.95 525 3.85 0.00 98 -

115 CT4 35 38.0 306 261 45 4.25 100 5.30 485 5.30 0.70 140 158 CT12 65 72.0 580 369 210 4.10 90 5.40 665 5.15 1.65 136 158 f

CT15 75 58.0 467 381 86 3.75 90 5.35 690 5.30 1.60 124 151 GT2 100 78.0 628 362 266 3.65 90 5.35 675 4.45 2.25 121 149 GT11 125 88.0 709 263 445 2.75-100 3.80 675 3.20 2.70 91 108 GT7 150 101.0 813 350 464 3.55' 80 5.25 655

-+

117 145 CT1 200 101.0 813 313 500 3.55 100 5.00 625

-+

-+

118' 142 CT8 200 106.0 854 248 606 2.50 100 3.55 675

-+

-+

83 100-0 C

UTER MA bN

+ Fully ductile fracture, no arrest load.

06490:10/062890 5-8 l

~

L.

'\\

TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS'FOR THE SOUTH TEXAS UNIT 1 REACTOR VESSEL IO CORE REGION WELD METAL AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 2.93 x 10 n/cm2 (E > 1.0 MeV)

Normalized Energies Test Charpy Charpy Maximus Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2

Number (*F)

(ft-lb)

(f t-lb/in )

(kips)

(msec)

(kips)

(msec)

(kips)

(kips)

(ksi)

(ksi)

Weld Metal CW13

-65 10.0 81 38 43 3.80 80 5.05 110 4.95 1.05 125 146 CW15

-50 8.0 64 33 31 3.65 80 4.70 105 4.70 0.25 120 138 CW4

-35 19.0 153 144 9

4.75 120 5.30 285 5.30 0.55 157 166 CW7

-20 43.0 346 217 129 3.25 90 4.00 510 3.90 1.15 107 120 GW14 0

58.0 467 214 253 3.30 110 3.95 520 3.70 1.20 109 120 CW2 20 63.0 507 284 224 4.20 100 5.40 520 4.60 1.85 140 159 CW3 45 67.0 540 202 338 3.05 100 3.80 510 3.35 2.05 102 113 CW12 45 73.0 588 281 307 4.35 110 5.35 520 4.45 2.50 144 160 CW1 75 83.0 668 330 339 4.10 110 5.35 610 3.25 2.50 135 156 CW9 75 84.0 676 232 444 2.90 100 0 RO 585 2.95 2.00 96 111 CW5 100 86.0 692 230 462 2.95

.120

- 3.7ts 605

-+

98 110 CWS 150 84.0 676 311 365 3.60 90 5.05 600 119 143 CW11 150 91.0 733 233 499 2.70 100 3.60 620

-+

-+

89 104 CW10 220 87.0 701 225

.476 2.55 110 3.50 620

-e 85 101 CW8 220 89.0 717 340 377 3.40 90 4.95 670

-+

112 138 BAZ Metal CH7

-100 68.0 548 248 300 3.85 130 4.65 525 4.35 0.00 127 141 CH9

- 75 82.0 660 330 331 3.75 140 4.65 715 4.15 0.25 125 140 i

Clil

- 75 104.0 837 441 396 5.45 100 6.45 670 5.10 0.00 180 196 Cil5

- 50 17.0 137 102 35 4.45 100 5.60 210 5.60 0.45 148 167 Cil13

- 25 45.0 362 317 46 5.05 120 6.15 525 6.10 0.45 168 186 CH11

'O 26.0 209 109 100 3.55 110 4.00 285 3.95 0.85 117 124 Cil2 25 106.0 854 338 515 3.40 120 4.30 770 3.50 1.50 113 128 CH6 25 129.0 1039 412 626 4.80 110 5.95 690

-+

159 178 CHIS 75 78.0 628 216 412 2.80 80 4.10 505 3.80 2.80 93 114 CHIO 75 125.0 1007 451 555 4.80 120 6.10 750 160 181 3

CH12 125 BAD TEST (MACH DE MALFUNCTION)

CH8 125 163.0 1313 364 949 3.05 110 4.20 855 101 120 i

CH4 150 122.0 982 334 648 3.00 100 4.15 780 99 118 Cll3 200 BAD TEST (MACHINE MALFUNCTION)

CH14 200 105.0 845 379 467 4.10 130 5.50 715 135 158

. Fully ductile fracture, no arrest load.

06490:10/062890 5-9

TABLE 5-5

'EFFECT OF $50*F IRP.A0!ATION AT 2.93 x 10 n/cm2 (E > 1.0 MeV) 10 063 NOTCH TOUGHNESS PROPERTIES OF SOUTH TEXAS (MIT 1 REACTOR VESSEL SURVEILLANCE MATERI ALS Average Energy Average 35 mit Average 30 ft-tb Lateral Expansion Average 50 ft-tb Absorptim at Terperature (*F)

Temperature (*F)

Teaperature (*F)

Full Sheer (fe.tb)

Haterlat

-Unirradiated Irradiated At Unirradiated Irradiated AT unirradiated Irradiated AT-Unirradiated Irradiated A(ft-lb)'

Plate R1606-2

-5 25 33 20 30 to 30 55 2S 113 105

-8 (Transverse)

Plate R1606-2

-35

-25 10

-20

-5 15

-15 0

15 137 132

-5 (Longitudinat)

Weld Metal

-50

-30 7,

20 35 15

-20

-5 15 85 88 3

HAZ Metal

-75

-75 0

-40

-40 0

-55

-55 0

105 119 14 5-10 0634D:1D/073190

,- T TABLE 5-6 COMPARISON OF SOUTH TEXAS UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIAL-30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTI0ils

' 30 ft-lb Transition Temo. Dift Upper Shelf Eneroy Decrease Fluence-R.G. 1.99 Rev. 2 Capsule U R.G. 1.99 Rev. 2 Capsule U (Predicted)

(Predicted)

Material 1018 n/cm2

(*F)

(*F)

-(%)

(%)

Plate R1606-2 (Trans.)

2.93 17.0 30 14 7

Plate R1606-2 (Loag.)

2.93 17.0 10 14 4

Weld Metal 2.93 18.0 20 14 0

a) Cu and Ni values from Reference 1 were used to determine R.G. 1.99 predictions.

0654D:lD/070190 5-11 i

4 2,~~.---.m..-.,<=re:>-.

e---m,

.*,--ti---- - -

--.-a.

w

---

~---..-as

^

o.

TABLE 5-7'

.LE PROPERTIES FOR SOUTH TEXAS UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIAL IRRADIATED AT 550*F TO 2.93 x 1018,jc,2 (E > 1.0 MeV)

~

- l t

Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reductaos Sample Temp. Strength Strength Load Stress Strength Blongation Blongation in Area Material Number l'F),

(ksi)

(ksil (kip)

(ksil (ksi)

(5)

(5)

(5)

Plate 1606-2 CL1 73 68.8 86.6

'2.70 214.8 55.0 12.8 28.0 74 (Long.

CL2 300 62.6 P,1.5 2.70 160.2 55.0 10.5 21.9 66 Orient.)

CL3 550 58.1 85.6 2.60

-124.9 53.0 10.5 24.6 58 Plate 1606-2 GT1 73 66.2

-85.6 2.85 160.2 58.1 12.8 26.5 -

64 (Transv.

CT2 300 59.1 80.1 2.25 111.4 45.8 10.5 21.9 59 Orient.)

CT3 550 59.6 85.6 2.70 115.9 55.0 10.5 23.9 53 Weld CW1 73 67.2 85.6 2.70 182.2 55.0 14.2 28.5 70 Weld CW2

~300 65.2 81.7 2.75 168.4 56.0 10.5 22.1 67 Weld GW3 550 65.7 87.6 2.90 147.7 59.1 10.5 23.3 60 l

l 1

- r l

I i

l

\\

l 06490:10/062890 5-12 i

l l.

-curve =758450-A-

-s

(*C)

-150 -100 0 50 100 150 200 250 I

I i

$2_2';2 t

100

~

- f 80 y60 o

.c e

m @

20

-l 0

15 100 i

S 80 E

2.01 T-1,5e

- 60 2

2 fg 1,0 E 10*F o

0.5 20 g

0 i

i i

i i

i i

1 0

200 180 2@

160

~

l@

3 2

j120 Unirradiated

/

160

~~ 100

@ 80 120 C c

Irradiated ( 550 F) 10 60

~

25 F

2. 93 x 10 n/cm e

1 30 F 20 0

i i

i i

i i

i i

0 i:

- 200

-100 0

100 200 300

1. 0 500 Temperature (* F)

Figure 5-1.

Charpy V-Notch Impact Properties for South Texas Unit 1 Reactor Vessel Intermediate Shell Plate R1606-2 (Transverse Orientation) 06490:10/06.28 90

<. h m

curve 758444-A

( S C)

-i + -100

-50

' 0 --

50

-100-150 200 250-i i

i 1

I I

I l-p 100 p.

2 O

f 80 g@

.cw-m e

,o o 20 0

I

_2 i i

i i

i i

100 2.5 i

i i

A L2 I

I

$ 80 2N i.

2.0 E.

2

1. 5 s

- 60 I@

j

. g op 1, 0 "

0, 5 Q-20 0

1 I

i 1

1 i

i i

0 200 180 2M-160 200

_ 10 o

o g120 o

160

~ 100 - Unirradiated 120 0 80 NIrradiated ( 550 F) 18 2

2. 93 x 10 n/cm 80 60 o

15 F o

10*F M

20 0

i i

i I

i i

i i

0

- 200

-100 0

100 200 300 00 500 Temperature (*F)

Figure 5-2.

Charpy V-Notch Impact Properties for South Texas Unit 1 Reactor Vessel Intermediate Shell Plate R1606-2 (longitudinal Orientation) 06490:10/062890

_...._......__....._.._.)

Curve 758W6-A l

?

('C)

-150 -100

-50 0

50 100 150 200 250 1

'2

_'J2

{2 1

I 100 L2 3

$ 80

( 60 2

=

a O

gm m h

20 5

0 I

I *'

100 2.5 i

5 80 2.0 n

3 60

1. 5 E h8 1.0 3 2

15'F R5 r

3 3

0 i

le i

i i

i i

1 0

I E

i i

i i

i i

i i

i 180 2@

160 200 im 3

~ 120 160 g

2 100 g

Unirradiated

,Q 120 '

Z 5

80 60 Irradiated ( S50 F) 5'F g g

N 2

o7I 2.93 x 10 n/cm 20* F

/[ i 20 i

i i

i i

0 0

i l'

- 200

-100 0

100 200 300 00 500 Temperature ('F)

Figure 5-3.

Charpy V-Notch Impact Properties for South Texas Unit 1 1

Reactor Vessel Core Region Weld Metal 06490:1D/062890

Curvo 758445-A' j

(

C)

-150 -100

- 50 0

50 100 150 200 250 I

I I

I I

44I I

I I

I 100

~

o 3 80 g%

g

.c 4

m m

'g

\\

20 0

100 2.5 i

i i

i i

E 80 2.0 E

_ o

-v o

1. 5 e c.

- 60 e

1. 0 ?

j@

i o

20 0.5 3

o O

i i

i i

i i

i 0

l l

l l

l 1

1 I

I i

180 2@

Irradiated ( 550*F) 160 18 2

200 o

/2 n/cm

_ le

" /

. 93 x 10 a

g120 160 o

e Im b

8 lE O Unirradiated

= 80 O

• o 8o 80 g

E m

o 20 O

i i

i i

i i

i 0

- 200

-100 0

100 200 300 00 500 Temperature (* F)

Figure 5-4, Charpy V-Notch Impact Properties for South Texas Unit 1 i

Reactor Core Region Weld Heat Affected Zone Metal

(___.06_49_D;_ID/062890._________.____._._.__

I a

i l

i w :n

-u.

x hj.

.' [p, -

f ~r((]

I /[lh;

?y l

4 a

-f

,e GT14 GT9 GT10 GT5 GT13 GT4 GT12 GT16 GT2 GT11 r

5 m as 4.s ma w

e>

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mx; GT7 GTS GT1 GT3 GTS Figure 5 5.

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f 06490:1D/062890 5-27

i y

SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY i

6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR 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 i

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 t

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 ener y dependentdamagefunctionfordatacorrelation,ASTMStandardPracticeE85d9)

" Analysis and Interpretation of Light Water Reactor Surveillance Results,"

recommends reporting displacements per iron atom (dpa) along with fluence 06490:10/070290 6-1 1

1

(E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693II73, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to 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(3)

" Radiation Damage to 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.

Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillance program.

The capsules are located at azimuthal angles of 58.5', 61.0',

121.5', 238.5', 241.0', and 301.5' relative to the core cardinal axes as shown in Figure 4-1.

l A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1.

The stainless steel specimen containers are 1.182 by l

l-inch and approximately 56 inches in height.

The containers are positioned l-axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 14-foot high reactor core.

l l

06490:1D/081790 6-2 1

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 neutron pad 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 oerforming the fast neutron enposure 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 > 1.0 Mev), $(E > 0.1 Mov), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillanca capsule as well as for the determination of exposure parameter ratios; i.e.,

dpa/$(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 cycle 1 irradiation; and established the means to perform similar predictions and dosimetry evaluations for all 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 06490:10/070290 6-3

of varying neutron yield per fission and fission spectrum introduced by the build up of plutonium as the irrnup 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:

9 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, 0 geometry using the 00T two-dimensional discrete ordinates code (12) and the SAILOR cross-section library (13). 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 P expansion of the cross-sections i

3 and the angular discretization was modeled with an S8 order of angular quadrature.

The reference core power distribution utilized iii the forward analysis was derived _ from statistical studies of long term operation of Westinghouse 4-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.

Furtherniore, for the peripheral fuel assemblies, a 2o 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 +2a 0649D:10/070290 6-4

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

All adjoint analyses were also carried out using an S order of angular 8

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

Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R,0 geometry to provide r.eutron source distribution importance functions for the exposure parameter of interest; in this case, 9 (E > 1.0 MeV). Having the impor-tance functions and appropriate core source distributions, the response of interest could be calculated as:

R (r,0) - 1 I I l(r, 0, E) S (r, 0. E) r dr d0 dE 7

E where:

R(r,0)

$ (E > 1.0 MeV) at radius r and azimuthal angle 0

=

I(r,0,E)

Adjoint importance function at radius, r, azimuthal angle 0, and neutron source energy E.

S (r, 0, E)

Neutron source strength at core location r, 0 and energy E.

Although the adjoint importance functions used in the South Texas Unit I analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shcwn that, while the 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/$ (E > 1.0 MeV) is insensitive to changing core source distributions.

In the application of these adjoint importance functions to the South Texas Unit I reactor, therefore, the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle 4

specific basis by using dpa/$ (E > 1.0 MeV) and 4 (E > 0.1 MeV)/

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

06490:10/072790 6-5 l

The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design report for the first operating cycle of South Texas Unit 1 (14).

The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of the pressure vessel and surveillance capsules are summarized in Figure 6-2.

For comparison purposes, the core power distribution (design basis) used in the reference forward calculation is also illustrated in Figure 6 2.

Selected results from the neutron transport analyses performed for the South Texas Unit I reactor are provided in Tables 6-1 through 6-5.

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

In Table 6-1, the calculated exposure parameters (9 -(E > 1.0 MeV), 9 (E >

0.1 MeV), and dpa) are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions.

The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis.

The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared.

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 design basis and the cycle 1 plant specific power distributions, it is important to note that the data for the vessel inner red?us were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself.

l 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.

06490:10/070290 6-6

(

1

For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by:

$(220.27, 45') F (225.75, 45')

$1/4T (45')

=

t where el/4T (45')

Projected neutron flux at the 1/4T positite on

=

the 45' azimuth Projected or calculated neutron flux at the 6 (220.27, 45')

=

vessel inner radius on the 45' azimuth.

Relative radial distribution function from F (225.75, 45')

Table 6-3.

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

1 l

The DOT calculations were carried out for a typical octant of the reactor.

How(.sr, for the neutron pad arrangement in South Texas Unit 1, the pad extent i

l l

for all octants is not the same.

For the analysis of the flux to the pressure l

vessel, an octant was chosen with the neutron pad extending from 32.5' to 45' (12.5') which azimuthally produces the peak flux.

Other octants have neutron l

pads extending 22.5' or 20' which provide more shielding.

For the octant with the 12.5' pad, the peak azimuthal flux to the vessel occurs near 25' and the values in the tables for the 25' angle are vessel maximum values above the bottom of the neutron pad. The maximum vessel flux below the neutron pad i

occurs at 45' and is a factor of 1.25 (Cycle 1)/1.31 (Design Basis) higher than the 45' values in the tables which detail the azimuthal variation of the vessel flux above the bottom of the neutron pad.

Exposure values for O*, 15',

and 45' can be used for all octants; values in the tables for 25' and 35' are maximum values and only apply to octants with a 12.5' neutron pad extent. -

l 0649D:lD/072690 6-7

i 6.3 Neutron Dosimetry The passive neutron sensors included in the South Texas Unit I surveillance program are listed 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 tne subsequent determination of the various exposure parameters of interest

($ (E > 1.0 Mev), & (E > 0.1 MeV), dpa).

The relative locations of the neutron sensors wit 5in the capsules are shown in Figure 4-2.

The iron, nickel, copper, and cobalt-aluminum munitors, 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.

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.

l.

The specific activity of each of the neutron monitors was determined using l

l established ASTM procedures (15 through 28].

Following sample preparation and I

weighing, the activity of each monitor was determined by means of a l

lithium-drifted germanium, Ge(Li), gamma spectrometer.

The irradiation l

l l

06490:1D/070290 6-8

1 history of the South Texas Unit I reactor during cycle I was obtained from NUREG-0020, " 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 Values of key fast neutron exposure parameters were derived from the measured reactionratesusingtheFERRETleastsquaresadjustmentcode(29).

The l

FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data.

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

In the FERRET evale ';cns, 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 4 by some response matrix A:

7 (s.u) = [ A (S)

$ g(")

gg 9

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

For example, g=[9 gg $g R

o relates a set of measured reaction rates R$ to a single spectrum $g by the multigroup cross section ogg.

(In this case, FERRET also adjusts the cross-sections.) The log normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.

06490:10/070290 6-9

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

fluxes and cross-sections) were approximated in 53 gi aps.

The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code (30). 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 where group boundaries do not coincide.

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

t The cross-sections were also collapsed into the 53 energy group structure using SAND II 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,=Rh+R M

R,P g g gg, where RN specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values.

The fractional uncertainties R specify additional random uncertainties for g

group g that are correlated with a correlation matrix:

Pgg, - (1 - 0) Sgg, + 0 exp (

)

The first term specifies purely random ut.tertainties while the second term describes short-range correlations over a range g (q specifies the strength of the latter term.)

06490:10/070290 6-10

For the a priori calculated fluxes, a short-range correlation of y - 6 groups was used.

This choice implies that neighboring groups are strongly correlated when 0 is close to 1.

Strong long-range correlations (or anticorrrlations) were justified based on information presented by R. E. Maerker (3i].

Maerker's results are closely duplicated when 0 - 6.

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

Results of the FERRET evaluation of the capsule U dosimetry are given in Table 6-9.

The data summarized in Table 6-9 indicated that the capsule received an integrated expo::ure of 2.93 x 10 n/cm2 (E > 1.0 MeV) with an 18 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 fall within 1-3% for all expoeure parameters listed.

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

Along with the current (0.78 EFPY) exposure derived from the capsule U measurements, projections are also provided for an exposure period of 16 EFPY and 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 1.

Table 6-14 shows the azimuthal variation in neutron exposure projections at the reactor core beltline.

06490:10/070290 6-11

.I In the calculation of exposure gradients, applicable to reactor pressure vessel heatup and cooldown curves for the South Texas Unit I reactor coolant system, exposure projections to 16 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 6-15.

In order to access RT vs.

NDT fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T l

and 3/4T positions were defined by the relations

$' (1/4T) = $ (Surface) (dp (S r ace $' (3/4T) = $ (Surface) (dp ($ur ace)) l Using this approach results in the dpa equivalent fluence values listed in Table 6-15. l In Table 6-16 updated lead factors are listed for each of the South Texas-Unit I surveillance capsules. These data may be used as a guide in l establishing future withdrawal schedules for the remaining capsules. i l I i L l l 0649D:10/072h 0 6-12

) i (TYPICAL) - 88.s* -41.08 - 81.825 JN.- @,h 1'Nxx w'x - h'N %NNN NN NEUTRON PAD I I Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule 06490:10/070290 6 13

I I I i i .i ) I 0.80 0.80 0.77 0.62 Cycle 1 i 1.01 1.04 0.96 0.77 Design Basis l l 0.90 1.01 0.90 1.05 0.89 0.63 i 1.02 1.10 1.00 1.05 1.10 0.71 1.14 1.01 1.18 0.97 0.91 1.02 1.05 0.87 0.87 1.07 1.00 1.05 l E 1.06 1.20 1.05 1.13 1.15 1.09 1.06 0.88 1.10 1.04 il 1.21 1.08 1.21 1.06 0.90 1.04 1.12 0.92 f ) } [ t i i Figure 6-2. Core Power Distributions Used in Transport Calculations for South Texas Unit 1 i s 06490:10/070290 6-14 t ,.,a ,~ ~,.,--,...,.--,....---,n -..-,,~.a

1' p i-l 'i ~ TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER DESIGN BASIS CYCLE 1 u = u l 11 ll 10 - 9.87 x 10 10 $ (E > 1.0 MeV), 1.08 x 10 1.15 x 10 9.13 x 10 2 n/cm,3,e I Il / Il ll 4.11 x 10 I 4.44 x 10 $ (E'> 0.1 MeV), 4.84' :: 10 5.19 x 10 n/cm -sec 2 dpa/sec 2.11 x 10-10 2.26 x 10-10 1.79 x 10-10 1.93 x 10-10 i r I 1 c s b '? 1 1 't 06490:1D/070290 6-15 L 1

TABLE 6 2 CALCULA1ED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE i 1 DESIGN BASIS O' 15* 25' 35' 45.(a) 10 10 10 10 10 $ (E > 1.0 Mev), 1.70x10 3.54x10 2.87x10 2.34x10 2.68x10 2 i .n/cm -sec 10 10 10 10 10 $ (E > 0.1 Mev), 3.53x10 5.34x10 7.85x10 6.65x10 6.72x10 l 2 n/cm-sec dpa/ set 2.64x10-II 3.93x10'II 4,81x10'll 3.96x10'II 4.28x10*II CYCLE 1 SPECIFIC d. O' 15' 25' _ 35' 45.(b)- 10 10 10 10 10 $ (E > 1.0 Mev), 1.44x10 2.18x10 2.53x10 2.14x10 2.51x10 2 n/cm -sec 10 10 10 10 10 $ (E > 0.1 Mev), 3.00x10 4.60x10 6.91x10 6.08x10 6.30x10 'n/cm -sec I 2 dpa/sec 2.23x10-ll 3.38x10-11 4.25x10'll 3.62x10'II 3.99x10'II. ; I (a) Increase values by 1.31 for lower shell base metal below neutron pad (b) Increase values by 1.25 for lower shell base metal below neutorn pad 06490:10/070290 6-16 _ = _ _. _ __ _ __: _ ___

~ TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E s 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL-Radius (cm) 0' 15' _2}' 35' 45' i t 220.27(I) 1.00 1.00 1.00 1.00 1.00 [ 220.64 0.976 0.979 0.980 0.977 0.979 l 221.66 0.888 0.891 0.893 0.891 0.k9 E 222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 0.554 0.552 0.543 I 226.95 0.462 0.460 0.465 0.463 0.452 2 6 0.3 039 0 3 03 230.92 0.267 0.265 0.271 0.267 0.257 232.25 0.221 0.219 0.223 0.221 0.211 233.57 0.183' O.181 0.185 0.183 0.174 j 234.89 0.151 0.149 0.153 0.151 0.142 236.22 0.124 0.122 0.126 0.124 0.116 237.54 0.102 0.100 0.104 0.102 0.0945 '238.86 0.0828 0.0817 0.0846' O.0835 0.0762 240.19 0.0671 0.0660 0.0689 0.0679 0.0608 .241.51 0.0538 0.0522 0.0550 0.0545 0.0471 4 -242.17(2) 0.0506 0.0488 0.0518 0.0521 0.0438 l NOTES:

1) Base Metal Inner Radius 1

2) Base Metal Outer Radius 06490:10/070290-6-17 l

~ TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV)- WITHIN THE PRESSURE VESSEL WALL Radius-(cm) 0* _15' _21*_ 35' 45' + 220.27(I)' l.00 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 1.00 221,66 1.00 1.00 1.00 0.999 0.995 222.99 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901 225.03 0.874 0.865 0.874 0.850 0.842 226.95 0.818' O.808 0.818 0.792 0.782 i 228.28 0.761 0.750 0.716 0.734 0.721 229.60 0.705 0.693 0.704 0.677 0.662 230.92 0.649 0.637 0.649 0.621 0.605 232.25-0.594 0.582 0.594 0.567 0.549 233.57 0.540 0.529 0.542 0.515 0.495 234.89 0.487 0.478 0.490 0.465 0.443 236.22 0.436 0.428 0.440 0.416 0.392-237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201 242.17(2) 0.233 0.226 0.237 0.223 0.188 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius 06490:10/070290 6-18

i TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) { WITHIN THE PRESSURE VESSEL WALL Radius (cm) O' 15' _21*_ 35' 45' 220.27(I) 1.00 1.00 1.00 1.00 1.00 220.64 0.984 0.981 0.984 0.983 0.984 221.66 0.912 0.909 0.917 0.921 0.915 222.99 0.815 0.812 0.826 0.833 0.821 224.31 0.722 0.719 0.737 0.747 0.730 225.63 0.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572-228.28 0.497 0.493 0.519 0.533 0.506 229.60 0.439 0.435 0.462 0.475 0.447 I 230.92 0.387 0.383 0.410 0.423 0.394 232.25 0.341 0.338 0.364 0.376 0.347 1 233.57 0.300 0.297 0.322 0.334 0.305 234.89 0.263 0.261 0.285 0.295 0.266 236.22 0.230 0.228 0.250 0.260 0.231 237.54 0.199 0.198 0.218 0.227 0.199 238.86 0.171 0.170 0.189 0.196 0.169 240.19 -0.145 0.144 0.161 0.167-0.140 241.51 0.121 0.119 0.135 0.139 0.113 242.17(2) 0.116 0.113 0.128 0.134 0.106 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius i

l 0649D:10/070290 6-19

TABLE.6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS- ) Reaction Target Fission Monitor of Weight

Response

Product Yield tiitterial Interest Fraction Ranae Half-Life (%) -] . Copper Cu63(n,a)Co60 0.6917 E > 4.7 MeV 5.272 yrs Iron Fe54(n p)Mn54 0.0582 E >.1.0 MeV 312.2 days Nickel NiS8(n.p)CoS8 0.6830 E > 1.0 MeV '70.90 days - Uranium 238* U238(n,f)CsI37 1.0 E > 0.4 MeV 30.12 yrs 5.99 Neptunium-237* Np237(n,f)Cs137 1.0 E > 0.08 MeV 30.12 yrs-6.50 i Cobalt Aluminum CoS9 (n,y>Co60 0.0015 E > 0.015 MeV 5.272 yrs Cobalt-Aluminum

• CoS9(n,y)Co60 0.0015 0.4ev> E >

5.272 yrs 0.015 MeV cDenotes that monitor is cadmium shielded. 4 i 0649D:10/070290 6-20

TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U Irradiation P P Irradiation Decay 3 3 Period (MW ) P Time (days) Time (days) t Ref. 'o/88 6 .002 24 791 4/88 921 .242 30 761 5/88 56 .015 31 730 6/88 616 .162 30 700 7/88 2074 .546 31 669 8/88 269 .071 31 638 9/88 2539 .668 30 608 10/88 2679 .705 31 577 ll/ v8 3481 .916 30 547 12/88 3061 .806 31 516 1/89 1839 .484 31 485 2/89 0 0 28 457 3/89 3363 .885 31 426 4/89 3647 .960 30 396 5/89 3590 .945 31 365 6/89 3801 1.000 30 335 7/89 _3189 .839 31 304 8/89 2615 .688 5 299 NOTE: -Reference Power = 3800 MW t 06490:10/070290 6-21

_ ~ _. q i. ~ TABLE 6-8 MEASVRED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated Reaction Monitor and Activity Activity. Rate Axial location Jdis/see-am) (dis /sec-om) (RPS/NUCLEVSJ F Cu-63 (n,cx) Co-60 i 4 5 l Top 3.63 x 10 4.20 x 10 4 5 Middle -3.55 x 10 4.10 x 10 4 5 Bottom 3.64 x 10 4.21 x 10 4 5 Average 3.61 x 10 4.17 x 10 6.36 x 10-17 i r Fe-54(n,p) Mn-54 5 6 Top 9.06 x 10 4.05 x 10 5 6 Middle 8.69 x 10 3.88 x 10 j 5 6 Bottom 8.75 x 10 3.91 x 10 l 5 6 Average 8.83 x 10 3.95 x 10 6.29 x 10 i Ni-58(n,p)Co-58 6 7 Top 2.61 x 10 6.04 x 10 { 6 7 Middle 2.55 x 10 5.90 x 10 6 7 Bottom 2.44 x 10 5.64 x 10 l 6 7 Average 2.53 x 10 5.86 x 10 8.36 x 10-15 U-238 (n,. Cs-137 (Cd) 5 6 Middle 1.03 x 10 5.88 x 10 3.88 x 10'I4 1 06490:10/070290 6-22 1

t F TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd Measured-Saturated-Reaction. Monitor and Activity Activity Rate Axial location (dis /sec-cm) ,(dis /sec-om) 1RPS/NVCLEUS) Np-237(n,f) Cs 137 (Cd) 6 7 13-Middle 1.05 x 10 6.02 x 10 3.64 x 107 Co-59 (n,y) Co-60 6 7 Top 8.09 x 10 9.35 x 10 6 7 Middle 8.36 x 10 9.66 x 10 6 7 Bottom 8.61 x.10 9.95 x 10 Average 8.35 x 10 9.65 x 10 6.30 x 10-12 6 7 Co-59 (n,y) Co-60 (Cd) 6 7 Top 4.20 x 10 4.85 x 10 6 7 Middle 4.33 x 10 5.00 x 10 6 7 Bottom 4.40 x 10 5.09 x 10 6 7 Average 4.31 x 10 4.98 x 10 3.25 x 10-12 4 i 06490:10/070290 6-23

~ TABLE 6-9

SUMMARY

OF NEUTRON 00SIMETRY RESULTS TIME AVERAGED EXPOSURE RATES 2 ll $(E>1.0MeV),n/cm-sec 1.19 x 10 8% 2 ll $ (E > 0.1 MeV), n/cm -sec 5.29 x 10 15% .l dpa/sec 2.30 x 10-10 11% 2 11 $ (E > 0.414 eV), n/cm -sec 1.26 x 10 21% INTEGRATED CAPSULE EXPOSURE 2 lb o (E > 1.0 McV), n/cm 2.93 x 10 8% 2 I9 $ (E.> 0.1 MeV), n/cm 1.30 x 10 15% dpa 5.66 x 10-3 11% 2 18 @ Ic > 0.414 eV), n/cm 3.10 x 10 21% NOTE: Total Irradiation Time - 0.78 EFPY 1 0649D:lD/070290 6-24

TABLE 6-10 COMPARIS0N OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER Adjusted Reaction Measured Calculation C/3 Cu-63 (n,cx) Co-60 6.36x10-17 6.45x10-17 1.01 Fe-54 (n,p) Mn-54 6.29x10-15 6.24x10-15 0.99 Ni-58 (n,p) Co-58 8.36x10-15 8.42x10-15 1.01 U-238 (n,f).Cs-137 (Cd) 3.88x10'I4 3.64x10-14 0.94 l Np-237 (n,f)'Cs-137.(Cd) 3.64x10-13 3.73x10-13 1.02 Co-59 (n,7) Co-60 6.31x10-12 6.26x10-12 0.99 Co-59 (n,y) C0-60 (Cd) 3.25x10-12 3.26x10-12 1.00 i 4 1 06490:10/070290 6-25

t 5 ( TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT-THE SURVEILLANCE CAPSULE CENTER Energy Adjusted Flux Energy Adjusted Flux 2 2 Group (Mev) (n/cm-sec) Group '(Mev) (n/cm-sec) 1 6 10 I 1.73x10 9.25x10 28 9.12x10-3 2.40x10 l 2 1.49x10 2.08x10 29 5.53x10'3 9.11x10 I 7 10 1 7 9 3 1.35x10 7.99x10 30 3.36::10-3 9.72x10 I 8 9 4 1.16x10 1.78x10 31 2.84x10 9.30x10 1 8 9 5 1.00x10 3.90x10 32 2.40x10-3 8.97x10 0 8 10 6 8.61x10 6.64x10 33 2.04x10-3 2.53x10 0 9 10 7 7.41x10 1.52x10 34 1.23x10-3 2.34x10 0 9 10' 8 6.07x10 2.17x10 35 7.49x10'4 2.18x10 0 9 10 9 4.97x10 4.60x10 36 4.54x10'4 2.08x10 0 9 10 10 3.68x10 6.13x10 37 2.75x10'4 2.25x10 0 10 10 11 2.87x10 1.30x10 38 1.67x10'4 2,44x10 0 10 10 12 2.23x10 1.81x10 39 1.0lx10'4 2.43x10 0 10 10 13 1.74x10 2.57x10 40 6.14x10 5 2.41x10 0 10 0 10 14 1.35x10 2.87x10 41 3.73x10 2.34x10 0 10 4 10 15 1.11x10 5.28x10 42 2.26x10 2.27x10 16 8.21x10'l 6.05x10 43 'l.37x10-5 - 2.20x10 10 10 17 6.39x10'l 6.28x10 44 8.32x10-6 2.10x10 10 10 18 4.98x10'I 4.56x10 45 5.04x10-6 1.93x10 10 10 19 3.88x10-1 6.40x10 46 3.06x10-6~ 1.80x10 10 10 20 3.02x10'I 6.57x10 47 1.86x10-6 1.66x10 10 10 21 1.83x10'l 6.50x10 48 1.13x10-6 1.23x10 10 10 22 1.llx10'l 5.19x10 49 6.83x10'7 1.59x10 10 10 23 6.74x10-2 3.60x10 50 4.14x10-7 2.12x10 10 10 24 4.09x10'2 2.04x10 51 2.51x10-7 2.13x10 10 10 25 2.55x10-2 2.67x10 52 1.52x10-7 2.05x10 10 10 26 1.99x10-2 1.31x10 53 9.24x10-8 6.28x10 10 10 27 1.50x10-2 1.66x10 10 NOTE: Tabulated energy levels represent the upper energy of each group. 06490:10/070290 6-26

h TABLE 6-12 COMPARIS0N OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE U i P Calculated Measured [fd 2 18 I9 e (E > 1.0 MeV) (n/cm ) 2.43 x 10 2.93 x 10 0.83 2 19 I9 $ (E > 0.1 MeV) (n/cm ) 1.09 x 10 1.30 x 10 0.84 dpa 4.75 x 10-3 5.66 x-10'3 0.84 -i +l l 06490:10/070290-6-27 l

TABLE 6-13 . NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE REACTOR VESSEL CLAD /BASEMETAL INTERFACE et (E > 1.0 MeV), et (E > 0.1 Mev), 2 2 Material EFPY n/cm n/cm dpa Intermediate Sheli 0.78 7'.18x10I7 18 1.19x10-3 1.94x10 I9 I9 Basemetal 16 1.45x10 3.96x10 2.43x10-2 I9 I9 32 2.90x10 7.92x10 4.87x10-2 17 17 Intermediate Shell 0.78 4.06x10 8.35x10 6.21x10~4 18 I9 -2 Long. Weld 16 .8.52x10 1.77x10 1.32x10 I9 I9 ~ 2.64x10-2 At 0* Azimuth 32 1.71x10 3.55x10 17 18 Intermediate Shell 0.78 4.63x10 1.27x10 7.71x10-4 18 I9 Long. Weld 16 9.73x10 2.71x10 1.64x10-2 I9 I9 At 120* Azimuth 32 1.95x10 5.43x10 3.28x10-2 O 17 18 -4 Intermediate'Shell 0.78 4.24x10 1.17x10 7.06x10 18 I9 l Long. Weld 16 8.93x10 2.49x10 1.50x10-2 At 240* Azimuth 32 1.79x10 4.98x10 3.0lx10-2 l9 19 06490:10/070290 6-28 .. ?

TABLE 6-13 (Continued) NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE REACTOR VESSEL CLAD /BASEMETAL INTERFACE et (E > 1.0 MeV), +t (E > 0.1 MeV), 2 2-Material EFPY n/cm n/cm dpa 17 18 Intermediate / Lower 0.78 7.18x10 1.94x10 1.19x10-19 I9 -2 Shell Circ. Weld 16 1.45x10 3.96x10 2.43x10 I9 19 -2 32 2.90x10 7.92x10 4.87x10 Lower Shell 0.78 9.31x10 2.31x10 1.46x10-3 17 18 Basemetal 16 1.77x10 4.45x10 2.82x10-2 I9 I9 32 3.54x10 8.89x10 5.63x10-2 I9 19 Lower Shell 0.78 4.27x10 8.79x10 6.53x10-4 17 17 Long. Weld 16 8.54x10 1.78x10 1.32x10-2 18 I9 At 90* Azimuth 32 1.71x10 3.56x10 2.65x10-2 I9 I9' I7 18 -3 Lower Shell Long. 0.78 7.18x10 1.98x10 1.20x10 I9 I9 -2 Weld At 210* and 16 1.39x10 3.88x10 2.34x10 I9 I9 -2 330* Azimuths 32 2.78x10 7.75x10 4.68x10 1 0649D:10/070290 6-29

TABLE'6-14 AZIMUTHAL' VARIATION OF THE NEUTRON EXPOSURE PROJECTIONS ON THE PRESSURE VESSEL CLAD /BASEMETAL INTERFACE.. Azimuthal $t (E > 1.0 MeV)' $t -(E > 0.1 MeV)' 2 2 EFPY Angle n/cm n/cm dpa 17 17 ~4 0* 4.26x10 8.79x10 6.52x10 17 18 15* 6.47x10 1.35x10 9.90x10-4 17 18 -3 0.78 25* 7.51x10 2.03x10 1.25x10 35' 6.35x10 1.78x10 1.06x10-3 17 18 17 18 -3 45*(a) 7.43x10 1.85x10 1.17x10 0* 8.59x10 1.78x10 1.33x10-2 18 I9 19 I9 15* 1.28x10 2L70x10 1.99x10-2 I9 I9 16 25* 'l.45x10 3 23 2.43x10-2. l9 -2 35* 'l.19x10 3.37xP. 2.0lx10 I9 45*(b) 1.5C.gi n 3.41x1N9 2.17x10-2 I9 19 -2 0* 1.72x10 3.57x10 2.67x10 I9 I9 -2 15* 2.57x10 5.40x10 3.97x10 I9 I9 32 25* 2.90x10 7.94x10 4.86x10-2 19' I9 35* 2.37x10 6.73x10 4.0lx10 I9 I9 -2 45*(b) 2.71x10 6.81x10 4.33x10 (a) Increase by l.25 for ' lower shell basemetal neutron pad (b) Increase by 1.31 for lower shell basemetal below' neutron ' pad. Of490:1D/070290 6-30 .m e .~ -.w.. .-vs. n 4 ,m

l TABLE 6-15 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOW CURVES -16 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface. 1/4 T 3/4 T 18 18 I7 18 18 0* 8.59 x 10 4.67 x 10 9.95 x 10 8.59 x 10 5.42 x 10 1.88 x'1018 19 18 18 I9 18 18 15' l.28 x 10 6.93 x 10 1.45 x 10 1.28 x 10 8.03 x 10 2.77 x 10 I9 IO l9 I9 18 18 25* 1.45 x 10 7.91 x 10 1.71 x 10 1.45 x 10 9.41 x 10 3.45 x 10 l9 18 18 I9 18 18 35* 1.19 x 10 6.47 x 10 1.38 x 10 3,ig.x 10 7.88 x 10 2.95 x 10 I9 18 18 I9 I9 18 45*(a) 1.36 x 10 7.27 x 10 1.47 x 10 1.36 x 10 8.70 x 10 2.98 x 10 32 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 2 -(n/cm ) (equivalent n/cm ) Surface _ 1/4 T 3/4 T Surface 1/4 T 3/4 T I9 18 18 I9 I9 18 0* 1.72 x 10 9.34 x 10 1.99 x 10 1.72 x 10 1.09 x 10 3.77 x 10 I9 I9 I9 I9 I9 18 15* 2.57 x 10 1.39 x 10 2.92 x 10 2.57 x 10 1.62 x 10 5.57 x 10 I9 I9 18 I9 I9~ I9 25* 2.90 x 10 1.59 x 10 3.42 x 10 2.90 x 10 1.89 x 10 6.90 x 10 l9 I9 18 I9 I9 18 35* 2.37 x 10 1.30 x 10 2.74 x 10 2.37 x 10 1.57 x 10 5.87 x 10 I9 19 18-I9 I9 18. 45*(a) 2.71 x 10 1.46 x 10 2.93 x 10 2.71 x 10 1.73 x 10 5.94 x 10 (a) Increase values by 1.31 for lower shell basemetal below neutron pad 0649D:1D/062990 6-31

lL' p TABLE 6-16 ~ VPDATED LEAD FACTORS FOR SOUTH TEXAS UNIT 1 SURVEILLANCE CAPSULES ? E I Caosule Lemd Factor l t U-3.14(a) l X 3.29 W 3.29 Z 3.29 I V 3.09 Y 3.09 (a) Plant specific evaluation

t i

f 1 06490:10/062990 6-32

SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the South Texas Unit I reactor vessel: Capsule Estimated Location Lead Fluence 2 Capsule (deg.) Factor Removal Time (a) (n/cm) 18 U 58.5 3.14 0.78 (Removed) 2.93 x 10 Y 241.0 3.09 5.5 1.88 x 1018(b) J V 61.0 3,09 10.5 2.59 x 10I9(c) 19 X 238.5 3.29 15.0 5.49 x 10 W 121.5 3.29 Standby Z 301.5 3.29 Standby p 1 (a) Effective full power years from plant startup. (b) Approximate fluence at 1/4 thickness reactor ves:el wall at end of life (32 EFPY). (c) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY), 06490:10/073190 7-1 ~ .. _ _ _O

i SECTION

8.0 REFERENCES

q 1. Kaiser, Koyama and Davidson, " South Texas Utilities South Texas Project Unit No. l Reactor Vessel Ra>!ation Surveillance Program," WCAP-9492, June 1979. 2. Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", and Appendix H, " Reactor Vessel Material Surveillance 1 Program Requirements," U.S. Nuclear Regulatory Commission, Washington,- D. C. 3. Regulatory Guide 1.99, Proposed Revision 2, " Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, February, 1986. F 1 4. Section III of the ASME Boiler and Pressure Vessel Code, Appendix G, " Protection Against Nonductile Failure." 5. ASTM E208, " Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels." 6. ASTM E 185-82, " Standard Practice for Light-Water Cooled Nuclear Power Reactor Vessels, E 706 (IF)." 7. ASTM E 23-88, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials." 8. ASTM A 370-88, " Standard Test Methods and Definitions for Mechanical Testing of Steel Products." 9. ASTM E 8-89, " Standard Test Methods of Tension Testing of Met &llic Material s. "

10. ASTM E 21-79, " Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."

06490:10/070190 8-1

L

11. ASTM E 83-85, " Standard Practice for Verification and Classification of

'l Extensometers." 12. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear. Rocket Shielding Methods, Modification, Updatina ed Input Data Preparation.. Vol. 5--Two-Dimensional Discrete Orainates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970, 13. "0RNL RSCI Data Library Collection DLC-761 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors". 14. F. A. Pecjak, et. al., "The Nuclear Design and Core Physics Characteristics of the South Texas Unit 1 Nuclear Power Plant Cycle 1," WCAP-11123, May 1986. (Proprietary) 15. ASTM Designation E482 82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984. 16. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984. 17. 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, Philadelphia, PA,1984. o 18. ASTM Designation E706-81a, " 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. 0649D:10/070190 8-2

19. 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.

20. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, i

Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

21. 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,1984.

22. 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.

23. 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.

24. ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, L

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

25. ASTM Designation ES23-82, " 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.

26. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.

i l 0649D:10/070190 8-3

27.- ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM. Standards, Section 12, American Society' for Testing and Materials", Philadelphia, PA, 1984. 28. ASTM Designation-E1005-84, " Standard Method for Application Ead Analysis of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM i Standards, Section 12, American Society for Testing and Materials, i Philadelphia, PA, 1984. 29. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979. 4 30. W. N. McElroy, S. Berg and T. Crocket, A Comouter-Automated Iterative l Method of Neutron Flux Soectra Determined by Foil Activation, i AFWL-TR-7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967. 31. EPRI-NP-2188, " Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981. l 4 1 \\ 1 ) 0649D:10/070190 8-4

APPEN0!X A HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION A-1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The f the material in the core region of the reactor vessel most limiting RTNDT is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced ARTNDT. NDT RT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and.35-mil lateral expansion (normal to the major working direction) minus 60*F. RT increases as the material is exposed to fast-neutron radiation. NDT Therefore, to find the most limiting'RT at any time period in the NOT reactor's life, ART due to the radiation exposure associated with that NDT time period must be added to the original unirradiated RTNDT. The extent of the shift in RI is enhanced by certain chemical elements (such as copper NDT and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)IA'Il. Regulatory GJide 1.99, Revision 2 is used for the calculation of RT values at 1/4T and 3/4T locations (T is the thickness NDT of the vessel at the beltli.2 *:,lon). A-2. FRACTURE TOUGHNESS PROPERTIES i The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan [A-2] The pre-irradiation fracture-toughness properties of South Texas Unit 1 of the reactor vessels are presented in Table A-1. i 1 06490:1D/070190 A-1

'A 3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup' g or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal ~ temperature at that time. K is obtained from the IR reference fracture toughness curve, defined in Appendix G to the ASME Code (A-3) The K'IR curve is given by the following equation: KIR - 26.78 + 1.223 exp (0.0145 (T-RTNDT + 160)) (1) where KIR.= reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code (A-3) as follows: CKjg + KIT s KIR (2) L where KIM = stress intensity factor caused by membrane (pressure) stress l KIT - stress intensity factor caused by the thermal gradients KIR - function of temperature relative to the RTNDT f the material C - 2.0 for Level A and Level B service limits L C = 1.5 for hydrostatic and leak test conditions during which the reactor l. core is not critical 11 l L 06490:10/070190 A-2 l-l

k i At any time during the heatup or cooldown transient,- K is determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed. From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling i location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4 T location for finite cooldown rates than for steady-state IR operation. Furthermore, if conditions exist so that the increase in KIR exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on tempt.rature at the 1/4 T location and, therefore, allowable pressures may unknowingh be violated if the rate of cooling is decreased at various 06490:1D/070190 A-3

intervals along a cooldown ramp. The use of the composin curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-i temperature relationships are developed for steady-state conditions as well as i finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4 T crack during heatup is lower IR than the K f r the 1/4 T crack during steady-state conditions at the same IR time coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K 's do not offset each other, and the pressure-IR temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. s The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the L 06490:1D/070190 A-4 i

allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the 1983 Amendment to 10CFR50(A-4] has a rule which addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RT by at least 120*F for normal operation when the NDT pressure exceeds 20 percent of the preservice hydrostatic test pressure. Table A-1 indicates that the limiting RT f l'F occurs in the closure NDT head flange of South Texas Unit 1, so the minimum allowable temperature of this region is 121*F. These limits are shown in Figures A-1 and A-2 whenever applicable. A-4. HEATUP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System cave been calculated using the methods discussed in section 3.0. Figure-A-1 is the heatup curve for 100*F/hr and applicable for the first 32 EFPY with margins for possible instrumentation-errors. Figure A-2 is the cooldown curve up to-100'F/hr and applicable for the first 32 EFPY with margins for possibic instrumentation errors. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures A-1 and A-2. This is in addition to other criteria which must be met before the reactor is made critical. The leak limit curve shown in Figure A-1 represents' minimum temperature requirements at the leak test pressure specified by applicable codes [A-2, A-3) The leak test limit curve was determined by methods of references A-2 and A-4. 1 06490:lD/070190 A-5

i I .l Figures A-1 and A-2 define limits for ensuring prevention of nonductile failure for the South Texas Unit 1 Primary Reactor Coolant System. il A-5. ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2 [A-1) the adjusted reference temperature (ART) for each material in the beltline is given by the following expression: ART = Initial RTNDT + ARTNDT + Margin (3). Initial RTNDT. is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. If measured values of initial RT for the material in NDT question are not available, generic mean values'for that class of material may -l be used if there are sufficient test results to establish a mean and standard deviation for the class. ' I ART is the mean value of the adjustment in reference temperature caused NDT by irradiation and should be calculated as follows: NDT - [CF]f(0.28-0.10 log f) (4)- ART To calculate ARTNDT ct any depth (e.g., at 1/4T or 3/4T), the following 4 formula must first be used to attenuate the fluence at the specific depth. 2 f(depth X) " Isurface('.24x) i (5) where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface. The resultant fluence is then put into equation (4) i to calculate ART at the specific depth. NDT CF (*F) is the chemistry factor, obtained from reference A-1. All materials in the beltline region of South Texas Unit I were considered for the limiting. material. RT at 1/4T and 3/4T are summarized in Table A-2. From Table NDT 4 A-2, it can be seen that the limiting material is lower shell for heat. p and cooldown curves applicable up to 32 EFPY. A sample calculation for RT is NDT shown in Table A-3. l 06490:1D/072790 A-6

TABLE A-1 SOUTH TEXAS UNIT 1 REACTOR VESSEL TOUGHNESS TABLE (Unirradiated)(A 5) Material RT Comoonent Code No. h_{?). Ni (%) NDT('F)fa) Closure Head Flange (b) R1602-1 0.05 0.72 0 Vessel Flange (b) R1601-1 0.02 0.75 -10 Intermediate Shell R1606 1 0.04 0.63 10 Intermediate Shell R1606 2 0.04 0.61 0 Intermediate Shell R1606 3 0.05 0.62 10 Lower Shell R1622 1 0.05 0.61 30 Lower Shell R1622 2 0.07 0.64 <30 I: Lower Lbell R1622 3 0.05 0.66 -30 Longitudinal Welds 0.03 0.05 -50 circumferential Welds 0.03 0.04 -70 (a) Based on actual data (b) To be used for considering flange requirements for heatup/cooldown curves.(A-4) 0649D:10/070190 A7

TABLE A-2

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURE (ART) AT 1/4T and 3/4T LOCATION 32 EFPY RTNDT AT ~ Lomeonent_ 1/RT ('F) 3/4T ('F) Intermediate Shell - R1606-3 80* 64* i-Lower Shell R1622-2 57 44

• These RTNDT numbers were used to generate the heatup and cooldown t,urves applicable up to 32 EFPY 06490:10/070290 A-8

TABLE A 3 CALCULATION OF AD WSTED REFERENCE TEMPERATURES FOR LIMITING SOUTH TEXAS UNIT 1 REACT 0:t VESSEL K%TERIAL - INTERMEDIATE SHELL (R1606 3) Reaulatorv Guide 1.99 - Revision 2 32 EFPY Parameter 1/LI 2/_4 I Chemistry Factor, CF (*F)) 31 31 l9 n/cm)(a 1.73 .613 2 Fluence,f(10 Fluence Factor, ff 1.15 .863

• • • • • • • • • • • • • • • • • • • • • • 5****.=:3.***********************************************

ARTNDT = CF x ff (*F) (b) 35.7 26.8 Initial RTNDT, I ('F) 10 10 Margin, M ('F) (C) 34 26.8 Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 80 64 ART = Initial RTNDT + ARTNDT + Margin I9 2 (a) Fluence, f, is based upon fsurf (10 n/cm, E >l Mev) = 2.90 at 32

EFPY, The South Texas Unit I reactor vessel wall thickness is 8.63 inches at the beltline region.

(b) The initial RTNDT (1) value for the lower shell is based on actual data. 2 2 (c) Margin is calculated as, M = 2 (o 1 + o 61 5 The standard 0 deviation for the initial RTNDT margin term (al) is assumed to be O'F since the initial RTNDT is a measured value. The standard deviation for ARTNDT, (oA) is 17'F for the plate. 06490:10/070190 A-9

l l MATERIAL PROPERTY BASLS CONTROLLING MATERIAL: INTERMEDIATE SHELL INITIAL RTNOT: 10 4 RT AFTER 32 EFPY: 1/4T, 80*F NOT 3/4T, 64'F E500 m,, m l'i i,f ! ) i ,r i i i l~' il I I I i ce!0 ~~' Leak Test a r 1 e i I r r i ,, L,i,mit. 1 J ' i i 1 i ' i i i ii! l. _ I. i A J c000 ~ i , I i! 1 ! ill ! l F .I i i ii i. 6. i, i ! i. i r 1 i i ' it i i ! ! I i r I I ii ii i i i [ e i i ; j j i 6 t l iiii i I i i ! ! ! i i + !? $ ll, Heatup Rates l Ug to ,,,, i i, 100 F/Hr if ,,,ii i i i i I f r i i i i ! i i i i i i i i i i e ie i r F 0 I 'i 9 ~ ' i i ! i i i J I i i i ..i i i i ! I i i I I' i [ i La ~ r i i i! Unacceptable J r w g 128.0 !!,l'; Operation ,I ,I 5

i i i i 1

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i /! I i. i 1 I I! l .i g + i i.i I i II i i 6 i i t y i i i i i i e ! I i f i ! fl i i l i i iii i ir i ! i i . i i i y i 1 gg = i ! I i i ! IJ t ! i + 1 + ' ' O j j i ! 3 !I i i i i r i l i i i i i i 2 i

7 i

i fi i , i i ~ Criticality Limit Basec i i i on Inservice itydro-e c,9 !I i i i ' i ~~ l,'! l l, Static Test Temp. ll i i (225'F) for the Servic.,

i i

Period Up to 32 EFPY i! i i 2E) M i i i i,i Acceptable -T i Operation '.al ~~' ,i;; i;i i i i ! ! i i t,, i iii! i i 1 i, i i i I O 50 10n 1 50 200 250 300 150 400 450 900 INDICATED TEWPCRATURE (DEC.r) Figure A-1. South Texas Unit 1 Reactor Coolant System Heatup Limitations (Heat up rate up to 100*F/hr) Applicable for the First 32 EFPY With Margins 10*F and 60 psig For Instrumentation Errors) 06490:10/070290 A-10 l

l MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL T INITIAL RTNDT: 10*F RT AFTER 32 EFPY: 1/4T, 80*F NDT 3/4T, 64*f 2500 asa 1 an 1 I 2250 f i I I' 2000 l r 1750 I I J ^ 1500 Unacceptable i 2 Operation I i 1250 l l / W / -Acceptable E 1000 g 0 Mration 750 ' Coo 1N [ . Rates of U erfgg . r

=

vr i 0 /p 500 25 / / 50 / 100 250 0 0 50 100ll 150 P00 250 300 350 400 450 500 INDICATED TD'tRATURE (Oge.r) Figure A-2. South Texas Unit 1 Reactor Coolant System Cooldown (Cooldown rates up to 100*F/hr) Limitations Applicable for the First 32 EFPY (With Margins 10*F and 60 p;ig For Instrumentation Errors) 06490:10/070290 A-ll

A-7. REFERENCES A-1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May,1988. A-2 " Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Aaalysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981. A3 ASME Boiler and Pressure Vessel Code, Section !!!, Division 1 - Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G. Protection Against Nonductile Failure," pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York,1986. A4 Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Require.nents," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No.104, May 27,1983. A-5 Ray, N. K., " South Texas Units 1 & 2 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation", Letter Report MT/SHART 170(88), Revision 1, October 1988, s 06490:10/070290 A-12 M --__-a

ATTACHMENT A DATA POINTS FOR HEATUP AND COOLDOWN CURVES (With Margins 10'F and 60 psig for Instrumentation Errors) 06490:10/070190 A-13

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TGM-THX COOLDOWN CURVES FOR, REG. GUIDE f.99.REV.2 FOE R-1606-3 09/28/88 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 3 (s0 DEG-F / 68t COOLDOWN ) INRADIATION PERIOD

• 32.000 EFP YEAes FLAW DEPTH = AOWIN 7 TNDICATED INDICATES INDICATED INDICAffD INDICATED INDICATfD TEMPERATURE PRESSURE YEMPERATURE PRE 550tf TEMPERaTueE

'# ESSURE i (DEG.F) (PSI) DEG. F ) (PSI) (DES.F) (Pst) I i e5.000 500.06 7 11s.000 534.ss is 148.000 830.73 2 90.000 526.38 8 120.000 641.70 14 150.000 872.77 3 9s.000 s44.82 9 13s.000 690.87 ts tSs.OOO 017.93 4 100.000 564.89 to 130 000 722.01 16 160.000 966.42 i s 95.000 ses.3d et tas.OOO 7ss.ss 17 tss.OOO tote.4e l 6 10.000 609.69 12 140.000 791.80 3= e I 4 l

1GN-THX COctOOWN CURVES FOR REG. GUIDE

1. 99,RE V. 2 F OR R-1606-3 09/28/89 THE FOLLOWING DATA trERE PLOTTED FOR COOLDOWN PROFILE 4 (tOODEG-F / HR CO% DOWN )

32.000 EFP WEARS TRRAO! ATION PERIM)

• AOWIN T FLAW OEPTH

= fle!CATED 198)f CATEO INDICATED 19etCATED TNDICATEO tefCATED T E 8ePE R A TURE PRESSURE T EssPER A TURE PRESSURE YE tePE R ATURE PRESSURE (Dee.F ) (PSI) (OEG.F) (PSI) ( DE S. F ) (PSI) s 98.000 423.57 7 119.000 973.92 12 840.000 792.92 2 90.000 443.94 8 920.000 605.83 13 145.000 810.00 3 95.000 466.08 9 125.000 640.72 to 150.000 960.79 4 100 000 499.86 to 130.000 678 45 15 955.000 995.59 8 105.000 515.49 ft 139.000 719.07 to 160.000 974.27 6 910.000 543.42 s e O I}}