Text

Analysis of Capsule U from Duquesne Light Co Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program
	ML20138S008
Person / Time
Site:	Beaver Valley
Issue date:	09/30/1985
From:	Shaun Anderson, Boggs R, Kaiser W; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20138S009	List: ML20138S008;
References
	WCAP-10867, NUDOCS 8511190379
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WCAP-10867 WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSUL: U FROM THE DUQUESNE LIGHT COMPANY BEAVER VALLEY UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM R. S. Boggs S. L. Anderson W.T. Kaiser September 1985 Work performed under Shop Order No. DH0J-106 APPROVED:

[.6(- b/ At T. A. Meyer, Manager Structural Materials and Reliability Technology.

Prepared by Westinghouse for the Duquesne Light Company Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 h' Nk j4 90308:1b-090585 P

PREFACE This report has been technically reviewed and verified.

Reviewer Sections 1 through 5 and 7 S. E. Yanichko j[h Section 6 A. H. Fero dMMe Appendix A W. K. Ma

/s/m d //[o 90308:1b-072385 iii

TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 1-1 2

INTRODUCTION 2-1 3

BACKGROUND 3-1 4

DESCRIPTION OF PROGRAM 4-1 5

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

Overview

~

5-1 5-2.

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

Tension Test Results 5-4 5-4.

Wedge Opening Loading Tests 5-4 6

RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1.

Introduction 6-1 6-2.

Discrete Ordinates Analysis 6-1 6-3.

Neutron Dosimetry 6-4 6-4.

Transport Analysis Results 6-8 6-5.

Dosimetry Results 6-10 6-6.

Surveillance Capsule Withdrawal Schedules 6-12 7

SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8

REFERENCES 8-1 90308:1b-090585 v

(

TABLE OF CONTENTS (Cont)

Section Title Page Appendix HEATUP AND C00LDOWN LIMIT CURVES FOR A-1 A

NORMAL OPERATION A-1.

Introduction A-1 A-2.

Fracture Toughness Properties A-2 A-3.

Criteria For Allowable Pressure-Temperature A-2 Relationships A-4.

Heatup and Cooldown Limit Curves A-5 90308:1b-072385 vi

LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the 4-5 Beaver Valley Unit 1 Reactor Vessel 4-2 Capsule U Diagram Showing Location of Specimens, 4-6 Thermal Monitors, and Dosimeters 5-1 Irradiated Charpy V-Notch Impact Properties for 5-16 Beaver Valley Unit 1 Reactor Vessel Lower Shell Plate 86903-1 (Transverse Orientation) 5-2 Irradiated Charpy V-Notch Impact Properties for 5-17 Beaver Valley Unit 1 Reactor Pressure Vessel Lower Shell Plate B6903-1 (Longitudinal Orientation) 5-3 Irradiated Charpy V-Notch Impact Properties for 5-18 Beaver Valley Unit 1 Reactor Pressure Vessel Wold Metal 5-4 Irradiated Charpy V-Notch Impact Properties for 5-19 Beavar Valley Unit 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Beaver Valley 5-20 Unit 1 Reactor Pressure Vessel Lower Shell Plate B6903-1 (Transverse Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for Beaver Valley 5-21 Unit 1 Reactor Pressure Vessel Lower Shell Plate 86903-1 (Longitudinal Orientation) 5-7 Charpy Impact Specimen Fracture Surfaces for 5-22 Beaver Valley Unit 1 Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for 5-23 Beaver Valley Unit 1 Weld Heat Affected Zone Metal 5-9 Comparison of Actual versus Predicted 30 ft lb 5-24 Transition Temperature Increases for the Beaver Valley Unit 1 Reactor Vessel Material based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 90308:1b-072385 vii

r LIST OF ILLUSTRATIONS (Cont)

Figure Title Page 5-10 Tensile Properties for Beaver Valley Unit 1 Reactor 5-25 Vessel Lower Shell Plate B6903-1 (Transverse) 5-11 Tensile Properties for Beaver Valley Unit 1 Reactor 5-26 Vessel Weld Metal 5-12 Fractured Tensile Specimens of the Beaver Valley 5-27 Unit 1 Reactor Vessel Lower Shell Plate B6903-1 and Weld Metal 5-13 Typical Stress-Strain Curve for Tension Specimens 5-28 6-1 Beaver Valley Unit 1 Reactor Geometry 6-34 6-2 Reactor Vessel Surveillance Capsule 6-35 6-3 Maximum Fast Neutron (E > 1 MeV) Fluence at the Beltline 6-36 Weld Locations as a Function of Full Power Operating Time 6-4 Maximum Fast Neutron (E > 1 MeV) Fluence at the Center of 6-37 the Surveillance Capsules as a Function of Full Power Operation Time 6-5 Maximum Fast Neutron (C > 1.0 MeV) Fluence at the Pressure 6-38 Vessel Inner Radius as a Function of Azimuthal Angle 6-6 Relative Radial Distribution of Fast Neutron (E > 1.0 MeV) 6-39 Flux and Fluence Within the Pressure Vessel Wall 6-7 Relative Axial Variation of Fast Neutron (E > 1.0 MeV) 6-40 Flux and Fluence Within the Pressure Vessel Wall A-8 A-1 Effect of Fluence, Copper and Phosphorus on ARTNDT for Reactor Vessel Steels per Regulatory Guide 1.99, Revision 1 A-2 Fast Neutron Fluence (E > 1.0 MeV) as a Function A-9 of Full Power Service Life (EFPY)

A-3 Beaver Valley Unit No. 1 Reactor Coolant System A-10 Heatup Limitations Applicable for the first 9.5 EFPY A-4 Beaver Valley Unit No. 1 Reactor Coolant System A-11 Cooldown Limitations Applicable for the first 9.5 EFPY l

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LIST OF TABLES Table Title Page 4-1 Chemical Composition of the Beaver Valley Unit 1 4-3 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the Beaver Valley Unit 1 4-4 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the Beaver Valley Unit 1 5-5 Lower Shell Plate 86903-1 Irradiated (Transverse) at 550*F, Fluence 6.54 x 10" n/cm (E > 1 MeV) 2 5-2 Charpy V-Notch Impact Data for the Beaver Valley Unit 1 5-6 Lower Shell Plate 86903-1 (Len at 550*F, Fluence 6.54 x 10" gitudinal) Irradiated n/cm2 (E > 1 MeV) 5-3 Charpy V-Notch Impact Data for the Beaver Valley Unit 1 3-7 Pressure Vessel Weld Metal Irradiated at 550*F, Fluence 6.54 x 10" n/cm2 (E > 1 MeV) 5-4 Charpy V-Notch Impact Data for the Beaver Valley Unit 1 5-8 Pressure Vessel Weld Heat Affected Zone Metal Irradiated at 550*F, Fluence 6.54 x 10" n/cm 2 (E > 1 MeV) 5-5 Instrumented Charpy Impact Test Results for Beaver Valley 5-9 Unit 1 Lower Shell Plate B6903-1 (Transverse Orientation) 5-6 Instrumented Charpy Impact Test Results for Beaver Valley 5-10 Unit 1 Lower Shell Plate 86903-1 (Longitudinal Orientation) 5-7 Instrumented Charpy Impact Test Results for Beaver Valley 5-11 Unit 1 Weld Metal 5-8 Instrumented Charpy Impact Test Results for Beaver Valley 5-12 Unit 1 Weld Heat Affected Zone Metal 5-9 The Effect of 550*F Irradiation at 6.54 x 10" 5-13 (E > 1 MeV) on the Notch Toughness Properties of The Beaver Valley Unit 1 Reactor Vessel Materials 5-10 Summary of Beaver Valley Unit 1 Reactor Vessel Surveillance 5-14 Capsule Charpy Impact Test Results 5-11 Tensile Properties for Beaver Valley Unit 1 Reactor Vessel 5-15 Material Irradiated to 6.54 x 102' n/cm8 6-1 47 Eneroy Group Structure 6-13 6-2 Nuclear Parameters for Neutron Flux Monitors 6-14 90308:1b-072385 ix

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LIST OF TABLES (Cont)

Table Title Page 6-3 Fast Neutron (E > 1.0 MeV) Exposure at the Pressure 6-15 Vessel Inner Radius - 0* Azimuthal Angle 6-4 Fast Neutron (E > 1.0 MeV) Exposure at the Pressure 6-16 Vessel Inner Radius - 15' Azimuthal Angle 6-5 Fast Neutron (E > 1.0 MeV) Exposure at the Pressure 6-17 Vessel Inner Radius - 30' Azimuthal Angle 6-6 Fast Neutron (E > 1.0 MeV) Exposure at the Pressure 6-18 Vessel Inner Radius - 45' Azimuthal Angle 6-7 Fast Neutron (E > 1.0 MeV) Exposure at the 15' 6-19 Surveillance Capsule Center 6-8 Fast Neutron (E > 1.0 MeV) Exposure at the 25' 6-20 Surveillance Capsule Center 6-9 Fast Neutron (E > 1.0 MeV) Exposure at the 35' 6-21 Surveillance Capsule Center 6-10 Fast Neutron (E > 1.0 MeV) Exposure at the 45' 6-22 Surveillance Capsule Center 6-11 Irradiation History of Surveillance Capsules Removed 6-23 from Beaver Valley Unit 1 6-12 Measured Flux Monitor Activities from Beaver Valley 6-27 Unit 1, Capsule V 6-13 Measured Flux Monitor Activities from Beaver Valley 6-28 Unit 1, Capsule U 6-14 Factors to Adjust Measure Activities to the Radial Center 6-29 of the Surveillance Capsules 6-15 Calculated Neutron Spectra at the Center of Capsules 6-30 V and U

}

6-16 Spectrum Averaged Reaction Cross-Sections at the Center 6-31 of Beaver Valley Unit 1 Surveillance Capsules 6-17 Thermal Neutron Flux Data from Capsules V and U 6-32 6-18 Comparison of Measured and Calculated Fast Neutron 6-33 Fluence for Capsules V and U A-1 Beaver Valley Unit No. 1 Reactor Vessel Toughness Data A-7 (Unirradiated) 90308:1b-090585 x

SECTION 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in Capsule U, the second surveillance capsule to be removed from the Beaver Valley 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 6.54 x 10 n/cm,

o Irradiation of the reactor vessel lower shell plate B6903-1, to 18 6.54.x 10 n/cm, resulted in 30 and 50 ft-lb transition temperature increases of 120*F and 135 F, respectively for specimens criented parallel to the major working direction (longitudinal orientetion) and increases of '135'F and 160*F, respectively for specimens oriented normal to the major working direction (transverse orientation).

18 2

o Weld metal irradiated to 6.54 x 10 n/cm resulted in a 30 and 50 ft-lb transition temperature increase of 155'F and 200*F respectively.

o The average upper shelf energy of the plate B6903-1 (transverse orientation) decreased from 80 to 78 ft-lbs and the limiting weld metal decreased from 112 to 83 ft-lbs. Both materials exhibit a more than adequate shelf level for continued safe piant operation.

o Comparison of the 30 ft-lb transition temperature increases for the Beaver Valley Unit 1 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1, shows that the plate material and weld metal transition temperature increase were I

less than predicted or that the embrittlement was less than predicted.

9030B:1b-082685 1-1 6

SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule U, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Duquesne Light Company Beaver Valley Unit I reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Duquesne Light Company Beaver Valley Unit i reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. Adescript(onofthesurveillanceprogram and the preirradiation mechanical properties of the reactor vessel materials are presented by Davidson and Yanichko.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, " Recommended Practice for Surveillance Tests for Nuclear Reactors".[2] Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes testing and the postirradiation data obtained from surveillance Capsule U removed from the Duquesne Light Company Beaver Valley Unit i reactor vessel and discusses the analysis of the data.

The data are also compared to capsule V(3) which was removed from the reactor in 1979.

t 90308:1b-0826B5 2-1 L

l SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry.

The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA533 Grade B Class 1 plate (base material of the Duquesne Light Unit I reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNOT)*

RT is defined as the greater of either the drop weight nil-ductility NOT transition temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 f t Ib (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material.

The RT f a given material is used to index that NOT 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 i

90308:1b-082285 3-1

the K curve, allowable stress intensity factors can be obtained for this yp material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress inten'sity factors.

RT and, in turn, the operating limits of nuclear power plants can be NDT adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program,[1] 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 ttmperature (ARTNDT) due to irradiation is added to the original RTNDT f r radiation embrittlement. This adjusted RTNDT to adjust the RTNDT (RT initial + ARTNDT) is used to index the material to the KIR NDT curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

90308:1b-082285 3-2

SECTION 4 DESCRIPTION OF PROGRAN Eight surveillance capsules for monitoring the effects of neutron exposure on the Beaver Valley Unit I reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the neutron shielding pads and the vessel wall at locations shown in figure 4-1.

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

Capsule U was removed after 3.58 effective full power years of plant oppration. The capsule contained Charpy V-notch impact and tensile specimens from the lower shell plate B6903-1 and submerged arc -weld metal representative of'the beltline weld seams of the reactor vessel, WOL specimens from the weld metal and Charpy.V-notch specimens from weld heat-affected zone (HAZ) material (Figure 4-2). All heat affected zone specimens were obtained from within the HAZ of plate B6903-1 of the representative weld.

The chemistry and heat treatment of the surveillance material are presented in table 4-1 and table 4-2, respectively. The chemical analyses reported in table 4-1 were obtained from unirradiated material used in the surveillance program. In addition, a chemical analysis was performed on an irradiated Charpy specimen from the weld metal and is reported in table 4-1.

All test specimens were machined from the 1/4 thickness location of the plate after stress relieving.

Test specimens represent material taken at least one plate thickness from the quenched edges of the plate. Base metal Charpy V notch impact specimens were oriented with the longitudinal axis of the specimen parallel and normal to the major working directior of the plate (longitudinal and transverse orientation). Charpy V-notch and tensile

(.

90308:1b-082685 4-1

specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding direction. The WOL specimens in Capsule U were machined such that the simulated crack in the specimen would propagate parallel to the weld direction.

Capsule U contained dosimeter wires of pure iron, copper, nickel, and aluminum-cobalt (cadmium-shielded and unshielded).

In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were contained in the capsule.

Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2.

The two eutectic alloys and their melting points are:

2.5% Ag, 97.5% Pb Melting Point 579 F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590*F (310 C)

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

1 s

90308:lb-082685 4-2

TABLE 4-1 CHEMICAL COMPOSITION OF THE BEAVER VALLEY UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS As Deposited (c)

Plate B6903-1 Weld Metal Element [a]

Analysis Analysis C

0.20 0.110/0,124(b)

Si 0.180 0.270/0.277(b]

Mo 0.550 0.480 Cu 0.200 0.260/0.230(b]

Ni 0.540 0.620/0.637(b]

Mn 1.310 1.370/1.42[b]

i Cr 0.140 0.015/0.029(b]

V 0.001 0.001/0.008(b]

Co 0.014 0.014/0.009(b]

S 0.015 0.006/0.004(b]

P 0.010 0.018/0.008(b]

Al 0.028 0.010/0.028(b)

N 0.004 0.014 2

Sn 0.010 0.008 (a) Elements not listed are less than 0.01 weight percent.

(b] Analysis performed on irradiated weld specimen DW-63.

(c) Surveillance weld used the same heat of weld wire (#305424) and flux lot (3889) as used to fabricate the intermediate shell vertical seams.

90308:1b-090585 4-3 L

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TABLE 4-2 HEAT TREATMENT OF THE BEAVER VALLEY UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Temperature Material

("F)

Time (hr)

Coolant Lower r1550*/1650*

4 Water quenched Shell Plate B6903-1 1225* 1 25' 4

Air cooled 1150' 1 25' 40 Furnace cooled Weldment 1150' 1 25*

12 Furnace cooled 90308:1b-082285 4-4

14817.1 270*

w CAPSULE (TYPICAL) y REACTOR VESSEL I

i I

e THERMAL SHIELD

\\

g, li s' 10' 1

180*

~

O' V

25' 5

T U

90*

\\

Figure 41 Arrangement of Surveillance Capsules in Beaver Valley Unit 1 i

45

SPECIMEN NUMBERING CODE:

DT - LOWER SHELL PLATE B6903-1 (TRANSVERSE DIRECTION),

4' DL - LOWER SHELL PLATE B6903-1 (LONGITUDINAL DIRECTION DW - WELD METAL DH - WELD HEAT-AFFECTED-ZONE METAL CAPSUL Np23 U2 TENSILE WOL WOL WOL WOL TENSILE CHARPY CHARPY Od DTl2 DW12 OHIDH OHIOL 72170 6e140 U

Dwe DW7 Dws Dws g

DTil DWil dL JL A l

.15% Co ---ek ll h k ll he-- Cu NI NI--e R

Al

.15% Co ( Cd ) --adll l l ll l LJ LJ La La I

r, r,

Il l l l

! !=

r.

r,

)

TO TOP OF VESSEL i

" he

)

. ~ -

14447-1

[

k 25' x

/

s' CORE O

.z-#

CAPSULE U THERMAL SHIELD TI VESSEL WALL APERTURE CARD Also Available On tu Aperture Card rER

( CHARPY cHARPY CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY Elk EIS_

ele El_I E.YI ele ElEA ele frei M

M M

n a

h hhhe - Cu N I ---a hhhe-- Cu Al

.15% Co H khhhe yhhr

--a OR IIII IIII Al

.15% Co (Cd) H4III LJ LJ LJ LJ LJ LJ c,

c c,

II II l l

! !=

r.

ER RE0 ION OF VESSEL TO BOTTOM OF VESSEL Figure 4-2.

Capsule U Diagram Showing Location Of Specimens, Thermal Monitors, and Dosimeters 96/H70879-4 4-6

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SECTION 5 TESTING 0F SPECIMENS FROM CAPSULE U 5-1.

OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H, ASTM Specification E185-82 and Westinghouse Procedure MHL 8402, Revision 0 as modified by RMF Procedures 8102 and 8103. The mechanical test data was documented in an R&D report by Lott and Shogan.I43 i

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-8457 Ill No discrepancies were found.

Examination of the two low melting 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 E23-82 and RMF Procedure 8103 on a Tinius-01sen Model 74, 358J machine.

The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E ).

From the load-time curve, the load of general yielding (PGY),the D

time to general yielding (tGY), the maximum load (P ), and the time to g

maximum load (t ) can be determined. Under some test conditions, a sharp y

9030B:1b-082685 5-1

drop in load indicative of fast fracture was observed.

The load at which fast fracture was initiated is identified as the fast fracture load (P ), and the p

load at which fast fracture terminated is 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 approximately equivalent to the energy required to initiate a crack in the specimen.

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

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

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

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

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

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

Tension tests were performed on a 20,000 pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to R 45. The upper pull rod was connected through a univer, sal C

joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test.

Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure.

The extensometer gage length is 1.00 inch. The extensometer is rated as Class 8-2 per ASTM E83-67.

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone.

All tests were conducted in air.

I 90308:1b-082285 5-2

i Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.

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

In test 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 room temperature to 550*F (288 C). The upper grip mas used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures.

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

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve.

The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs.

The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

5.2.

CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials 18 2

contained in Capsule U irradiated at 6.54 x 10 n/cm are presented in Tables 5-1 through 5-8 and Figures 5-1 through 5-4.

The transition temperature increases and upper shelf energy decreases for the Capsule U material are shown in Table 5-9.

Table 5-10 summarizes the Charpy impact test results from Capsule U with the previous Capsule V.

Irradiation of vessel lower shell plate B6903-1 material (transverse 18 orientation) specimens to 6.54 x 10 m/cm2 (Figure 5-1) resulted in both 30 and 50 ft-lb transition temperature increases of 135'F and 160*F respectively, and an upper shelf energy decrease of 2 ft lbs.

Irradiation of vessel lower shell plate B6903-1 material (longitudinal 18 orientation) specimens to 6.54 x 10 n/cm2 (Figure 5-2) resulted in both 30 and 50 ft-lb transition temperature increases of 120*F and 135 F, respectively, and an upper shelf energy decrease of 35 ft-lb.

l l

90308:1b-082285 5-3 i

i Weld metal irradiated to 6.54 x 10 n/cm2 (Figure 5-3) resulted in both 18 30 and 50 ft-lb transition temperature increases of 155* and 200*F respectively and an upper shelf energy decrease of 29 ft-lb.

Weld HAZ metal irradiated to 6.54 x 10 n/cm2 (Figure 5-4) resulted in 18 both 30 and 50 ft-lb transition temperature increases of 35'F and 45'F, respectively, and an upper shelf energy decrease of 23 ft-lb.

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

Figure 5-9 shows a comparison of the 30 ft-1b transition temperature increases for the various Beaver Valley Unit 1 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1.

This comparison shows that the transition temperature increase resulting from 18 2

irradiation to 6.54 x 10 n/cm is less than predicted by the Guide for plate B6903-1 (longitudinal and transverse orientation).

The weld metal

~

18 2

transition temperature increase resu. ting from 6.54 x 10 n/cm is also.

less than the Guide prediction.

5-3.

TENSION TEST RESULTS The results of tension tests performed on plate B6903-1 (transverse 18 2

orientation) and weld metal irradiated to 6.54 x 10 n/cm are shown in Table 5-11 and Figures 5-10 and 5-11, respectively. These results shown that irradiation produced an increase of 18 ksi in 0.2 percent yield strength for plate B6903-1 and approximately a 20 ksi increase for the weld metal.

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

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

5-4.

COMPACT TENSION TEST At the request of Duquesne Light Company, Wedge Open Loading (WOL) specimen will not be tested. CT specimen will be. stored at the Hot Cell at the Westinghouse R&D Center.

90308:1b-082685 5-4

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE BEAVER VALLEY UNIT 1 LOWER SHELL PLATE B6903-1 (TRANSVERSE) 18 IRRADIATED AT 550*F, FLUENCE 6.54 x 10 n/cm2 (E > 1 MeV)

Temperature Impact Energy lateral Expansion Sample No.

'F (*C) ft-lbs (Joules) mils (ma)

% Shear DT64 50 ( 10) 11.0 ( 15.0) 10.5 (0.27) 4 DT70 78 ( 26) 20.0 ( 27.0) 16.5 (0.42) 16 DT63 100 ( 38) 16.0 ( 21.5) 15.0 (0.38) 21 DT61 150 ( 66) 32.0 ( 43.5) 27.0 (0.69) 31 OT66 150 ( 66) 28.0 ( 38.0) 25.0 (0.64) 25 DT67 175 ( 79) 34.0 ( 46.0) 30.0 (0.76) 33 DT62 200 ( 93) 42.0 ( 57.0) 34.0 (0.86) 45 DT68 250 (1?l) 65.0 ( 88.0) 57.0 (1.45) 53 DT72 300 (149) 69.0 ( 93.5) 51.0 (1.30) 59 DT69 350(177) 77.0 (104.5) 64.5 (1.64) 100 DT71 400 (204) 76.0 (103.0) 66.0 (1.68) 100 DT65 450 (232) 80.0(108.5) 68.5 (1.74) 100 i

?

90308:1b-082285 5-5

l TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE BEAVER VALLEY UNIT 1 LOWER SHELL PLATE B6903-1 (LONGITUDINAL )

IRRADIATED AT 550 F, FLUENCE 6.54 x 10 n/cm2 (E > 1 MeV) 18 Temperature Impact Energy Lateral Expansion Sample No.

F (*C) ft-lbs (Joules) mils (mm)

% Shear DL47 50 ( 10) 8.0 ( 11.0) 8.0 (0.20) 7 DL46 78 ( 26) 25.0 ( 34.0) 19.0 (0.48) 15 DL43 100 ( 38) 20.0 ( 27.0) 17.5 (0.44) 21 DL44 150 ( 66) 53.0 ( 72.0) 37.5 (0.95) 38 OL48 200 ( 93) 58.0 ( 78.5) 40.5 (1.03) 54 DL41 250 (121) 91.0 (123.5) 72.5 (1.84) 87 DL45 300 (149) 111.0 (150.5) 77.5(1.97) 89 DL42 400 (204) 99.0 (134.0) 73.6 (1.87) 99 9030B:1b-082285 5-6

TABLE 5-3 CHARPY V-NOTCH IMPACT DATA FOR THE BEAVER VALLEY UNIT 1 PRESSURE VESSEL WELD NETAL IRRADIATED AT 550*F, 18 FLUENCE 6.54 x 10 n/cm2 (E > 1 MeV)

Temperature Impac't Energy lateral Expansion Sample No.

'F ('C) ft-lbs (Joules) mils (mm)

% Shear DW70 0 (-18) 10.0 ( 13.5) 6.5 (0.17) 4 DW67 50 ( 10) 19.0 ( 26.0) 11.5 (0.29) 22 DW72 78 ( 26) 32.0 ( 43.5) 22.5 (0.57) 35 DW61 78 ( 26) 28.0 ( 38.0) 21.5 (0.55) 35 DW65 100 ( 38) 30.0 ( 40.5) 24.0 (0.61) 38 DW71 150 ( 66) 42.0 ( 57.0) 34.0 (0.86) 66 DW63 150 ( 66) 34.0 ( 46.0) 31.5 (0.80) 47 DW69 200 ( 93) 67.0 ( 91.0) 47.5 (1.21) 72 CW68 200 ( 93) 58.0 ( 78.5) 49.0 (1.24) 64 CW62 250 (121) 80.0.(108.5) 70.5 (1.79) 73 CW66 300 (149) 80.0(108.5) 67.0 (1.70) 100 CW64 400 (204) 86.0 (116.5) 64.0 (1.63) 100 l

90308:1b-082285 5-7

TABLE 5-4 CHARPY V-NOTCH IMPACT DATA FOR THE BEAVER VALLEY UNIT 1 PRESSURE VESSEL WELD HEAT AFFECTED ZONE NETAL IRRADIATED AT 550*F, FLUENCE 6.54 x 10 n/cm2 (E > 1 MeV) 18 Temperature Impact Energy Lateral Expansion Sample No.

'F (*C) ft-lbs (Joules) mils (mm)

% Shear DH72

-75 (-59) 23.0 ( 31.0) 14.5 (0.37) 17 DH69

-25 (-32) 31.0 ( 42.0) 19.5 (0.50) 21 DH64

-25 (-32) 28.0 ( 38.0) 17.5 (0.44) 34 DH65

-25 (-32) 28.0 ( 38.0) 17.5 (0.44) 34 DH63 0 (-18) 60.0 ( 81.5) 35.5 (0.90) 55 DH70 0 (-18) 33.0 ( 44.5) 27.5 (0.70) 26 DH71 25 ( -4) 56.0 ( 76.0) 37.7 (0.96) 48 DH62 50 ( 10) 82.0 (111.0) 60.0 (1.52) 81 DH66 78 ( 26) 105.0 (142.5) 71.5 (1.82) 95 DH68 150 ( 66) 100.0 (135.5) 65.5 (1.66) 97 DH61 200 ( 93) 99.0 (134.0) 66.5 (1.69) 98 DH67 300 (149) 114.0 (154.5) 70.0 (1.78) 100 90308:1b-082285 5-8

toS TABLE 5-5 o

.?

INSTRUMENTED CHARPY IMPACT TEST RESULTS k

FOR BEAVER VALLEY UNIT 1 o$

LOWER SHELL PLATE B6903-1 (TRANSVERSE ORIENTATION) 8 U1 Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow SamDie Temp.

Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress No

(*C)

(Joules)

(kJ/m )

(N)

(USeC)

(N)

(USec)

(N)

(MPa)

DT64 10 15.0 186 134 52 14700 100 16700 190 16700 500 757 809 DT70 26 27.0 339 239 100 15300 95 18400 280 17100 400 785 866 DT63 38 21.5 271 179 93 14000 90 16500 230 15700 1100 722 785 T

DT66 66 38.0 175 352 122 13600 100 17700 415 17200 1600 699 806 DT61 66 43.5 542 359 184 13700 95 18000 415 18000 4200 705 815 DT67 79 46.0 576 393 184 13400 95 18100 450 17800 4900 691 812 DT62 93 57.0 712 457 255 12900

- 95 18300 515 16700 8600 664 804 DT68 121 88.0 1102 419 683 11700 90 17200 505 601 744 DT72 149 93.5 1169 372 797 12400 95 16900 455 638 754 DT69 177 104.5 1305 434 871 11500 90 17000 525 593 734 DT71 204 103.0 1288 396 892 13200 100 17600 465 677 791 DT65 232 108.5 1356 398 958 11200 95 15900 510 575 696

S TABLE 5-6

.h INSTRtMENTED CHARPY INPACT TEST RESULTS FOR BEAVER VALLEY UNIT 1

-T LOWER SHELL PLATE B6903-1 (LONGITUDINAL ORIENTATION) 8N8 Normalized Energies 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 Load Load Stress Stress No.

(*Cl (Joules)

(kJ/m )

(N)

(USec)

(N)

(pSec)

(N)

(MPa)

DL47 10 11.0 136 56 80 11900 75 15500 105 15200 600 610 703 DL46 26-34.0 424 392 32 14400 95 19500 425 19500 800 743 873 DL43 38 27.0 339 269 70 14500 90 18000 310 16800 200 T48 838 DL44 66 72.0 898 546 353 13900 95 19200 585 18700 3900 713 850 m

DL48 93 78.5 983 563 420 13600 115 18500 625 18600 7700 702 826 DL41 121 123.5 1542 459 1083 13000 90 18400 510 670 809 DL45 149 150.5 1881 630 1251 12000 95 18100 710 618 774 DL42 204 134.0 1678 562 1115 9400 70 17000 675 486 681

. _ = _ _ _ _ _ _.==_

u)8 TABLE 5-7 o

P' INSTRtMENTED CHARPY IMPACT TEST RESULTS k

FOR BEAVER VALLEY UNIT 1 WELD METAL e

Ri8 Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yleid Flow Sample Temp.

Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2

2 No.

(*C)

(Joules)

(kJ/m )

(N)

(USec)

(N)

(USec)

(N)

(MPa)

DW70

-18 13.5 169 89 81 9500 60 17500 135 17300 600 491 694 Dw67 10 26.0 322 198 124 15000 95 17000 250 16900 2100 772 822 Dw61 26 38.0 475 319 156 15300 100 18800 355 18900 4600 786 877 DW72 26 43.5 542 377 165 14700 95 19000 415 18900 4200 754 865 (n

DW65 38 40.5 508 391 117 14600 100 18600 435 18600 2800 750 853 h

Dw63 66 46.0 576 380 196 13200 100 17600 450 16700 2000 677 792 DW71 66 57.0 712 390 322 14500 105 18200 440 17300 9400 746 842 Dw68 93 78.5 983 449 534 14100 115 18100 515 16900 10600 724 828 Dw69 93 91.0 1136 429 706 13100 135 17400 540 14800 10800 673 783 Dw62 121 108.5 1356 438 918 11900 90 16600 530 610 731 f

Dw66 149 108.5 1356 455 901 12100 90 16700 540 623 741 Dw64 204 116.5 1458 480 978 11500 90 16700 575 593 727 I

i

to S

TABLE 5-8 o

INSTRUMENTED CHARPY IMPACT TEST RESULTS k

FOR BEAVER VALLEY UNIT 1 WELD HEAT AFFECTED ZONE METAL 8M St Normalized Energies 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 Load Load Stress Stress No.

(*C)

(Joules)

(kJ/m )

(H)

(pSec)

(N)

(pSec)

(N)

(MPa)

(MPa) 2 2

DH64

-32 38.0 475 100 375 18600 110 20100 140 19400 100 959 996 DH72

-59 31.0 390 283 107 18500 100 21000 285 21100 100 950 1017 DH65

-32 38.0 475 256 218 17000 95 20100 275 19700 6300 876 955 DH69

-32 42.0 525 351 175 17400 100 20600 355 20600 4800 894 977 l

un DH70

-18 44.5 559 341 218 17200 105 20300 360 20000 4800 883 963 l

DH63

-18 81.5 1017 516 501 16600 95 21100 500 19800 9900 857 971 l

DH71

-4 76.0 949 483 466 16800 105 20600 480 20600 12400 865 963 DH62 10 111.0 1390 574 816 15400 90 19900 575 5400 900 794 908 DH66 26 142.5 1780 619 1160 15200 90 20200 610 780 910 DH68 66 135.5 1695 560 1135 15100 115 19300 600 777 884 OH61 93 134.0 1678 526 1152 13600 95 18200 585 702 820 DH67 149 154.5 1932 612 1320 13400 90 18500 670 687 819 4

to S

TABLE S-9 EFFECT OF 550*F IRRADIATION AT 6.54 x 1018,fc,2 (E > MeV)

T ON THE NOTCH TOUGHNESS PROPERTIES OF THE b

BEAVER VALLEY UNIT 1 REACTOR VESSEL MATERIALS E

sn Average Average 35 ml]

Average Average Energy Absorption 30 f t -I D T erp (

F )

Lateral Expans ton Temp (*F )

50 ft-ID Temp (*F) at Full Shear (ft-ID)

Material Untrradiated Irradiated AT Untrradiated Irradiated AT Untrradiated Irradiated AT Untrradiated Irradiated A(ft-lo)

Plate 20 155 135 45 200 155 65 225 160 80 78 2

B6903-1 (Transverse)

Plate

-5 115 120 25 150 125 25 110 135 134 99 35 4

B6903-1 (Longitudinal)

Weio

-60 95 155

-45 160 205

-35 165 200 112 83 29 Metal HAZ Metal

-65

-30 35

-30 10 40

-40 5

45 128 105 23 4

TABLE 5-10

SUMMARY

OF BEAVER VALLEY UNIT NO.1 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RE'SULTS 68 Joule 41 Joule 50 ft-lb 30 ft-lb Decrease in Trans. Temp.

Trans. Temp.

Upper Shelf Fluence Increase Increase Energy 18 2

Material 10 n/cm

(*F)

(ft-lb)

Plate B6903-1 (Trans) 2.91 150 140 5

Plate B6903-1 (Trans) 6.54 160 135 2

Plate B6903-1 (Long) 2.91 135 130 20 Plate B6903-1 (Long) 6.54 135 120 35 Weld Metal 2.91 180 150 24 6.54 200 155 29 HAZ Metal 2.91 30 0

13 6.54 45 35 23 90308:1b-103085 5-14

e8 TABLE 5-11 b

TENSILE PROPERTIES FOR BEAVER VALLEY UNIT 1 18 2

k REACTOR VESSEL MATERIAL IRRADIATED TO 6.54 x 10 n/cm 8m3 J

Test 2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp.

Strength Strengtn Load Stress Strengtn Elongation Elongation in Area NO.

Material

(*F)

(kS1)

(ksi)

(kip)

(kS1)

(ksi)

(%)

DT12 TRANS 275 75.4 97.8 3.74 165.7 76.2 10.5 18.8 54 DT11 TRANS 550 71.3 96.2 3.'86 166.7 78.6 9.2 16.8 53 DW12 WELD 225 81.5 97.8 3.46 161.8 70.5 10.2 20.3 56 E

DWil WELD 550 76.4 95.7 3.40 169.1 69.3 9.5 21.4 59 l

a

14447-6 TEMPERATURE (*C)

-100

-50 0

50 100 150 200 300 100 I

I I

bO I

^

5 4

80 N

~

60 e

E 40 2

W a

2 20 4

3 0

I00 2.5 80 2.0 3

O

<gg 60 O

1.5 -

e k

Wd 1.0 ~

_J 2 40 155'F - -

$~

d 20 2

0.5 b'

O 2

80 DO e

UNIRRAd!ATED I

O e

50

.f rRRmIgEoAr 60 6.54xio n/em -

40 50

'8

4 30 3 O

h 40 O

30 issar 20 c:

20 O

e 10 r

G j

10 3

i O

O

-150

-50 50 150 250 350 450 TEMPERATURE (*F)

Figure 5-1. Irradiated Charpy V-Notch Impact Properties for Beaver Valley Unit 1 Reactor Pressure Vessel Lower Shell Plate B69031 (Transverse Orientation) 5-16

l 14447-5 TELFERATURE ('C)

-100

-50 0

50 100 150 200 I

I-I 3 I 3d L

i 100 80 3

N 60 40 m

20 -3 0

2.5 o

b 80 2

2 2.0 4%g 60 O

1.5 -

wJ

[O

. O i 2s r-1.0 40 r

20 g

0.5 O

<J O

l 100 140 V

O 9C UNIRRADIATED 120 g

80 100 70

~

Y O

60 F

80 i

IRRADIATED AT 50 ~'

h.

la 2

60 40 O

6.54xto n/cm m O

_a;

,35 3 u.

i20.r

- Vo )#

I I

I

[

O

-100 O

I00 200 300 400 500 J

TEMPERATURE (*F)

Figure 5-2. Irradiated Charpy V-Notch Impact Properties for Beaver Valley Unit 1 Reactor Vessel Lower Shell Plate B6903-1 (Longitudinal Orientation) 5-17

14447-4 TEMPERATURE ('C)

-100

-50 0

50 100 ISO 200 sf 100 I.

I I

2 i

80 2

e 9

2 g

60 O

O e

40 - O z

20 O

2.5 z

O, P

o U

5 80 O

2.0 e

g O

gg 60 1.5 -

Wd O

E zo5'F 1.0 -

_J 2 40 2

$~

20 0.5 e

3 3

O 140 90 120 - UNIRRADIATED O

go O

8 70 100 o

Y o

e-60

~

80 h

50 2 60 O

\\ IRRADIATED AT -

40 0

2OO'F is 2

2 2

6.54XIO n/cm -

30 I

W 40 20 155' F_

w 20 10 I

I I

0

-200

-100 O

100 200 300 400 TEMPERATURE (*F)

Figure 5-3. Irradiated Charpy V Notch Impact Properties for Beaver Valley Unit 1 Reactor Pressure Nesse! Weld Metal 5-18

I I

14447-3 TEMPERATURE ('C)

-100

-50 0

50 100 150 200 I

I I

I 3M I

100 g

_ 80 l

N

~

60 2

tr<

e 40'-

0 2

20 9,

0

.-.m;d i 00 '

2.5 2

80 n

2,o Zg O

O m

60 1.5 -

Z E

E S

40 4o ' F,,

1,0 -

W J

20 0.5 3

i 0

4J 100 140 o

O O

90 120 - UNIRRADIATEDh o

O (4

80 8

$ 100 O

e e

70 Y

60 O

IRRADIATED AT 50

[

.54xio* n/cina 60 2

9 40 0

45'F, E

30 40 b

20 20 2

10 I

I I

O

-200

-100 O

100 200 300 400 j

TEMPERATURE ('F)

Figure 5-4. Irradiated Charpy V-Notch Impact Properties for Beaver Valley Unit 1 Reactor Pressure Vessel Wcld Heat Affected Zone Metal l

5-19

148174 i '

h"l l.

l ?

Mj?

L;e; +}

l'f.a

~y Q

J, ' 4, y. j 1%

{

'_.a 4 1-t

.t 7 22 '

gas.

4.,.

4 ;;"

E..

t --

s IE i

h=4a,

- {j :. -

g_

']

. c T.,Si"gli',etq [;

' f g[' 31. ".

h(,2[,g) l;.($f-]^#

~"

1*

Q# f. e4 M

4 3

: wuI u

w;t l9 :g: gjy 4 4. s, :4~g~

i =- -

4,;.;:.

y

+

e-4 g-s a ;.q,%.

g 33 DT 64 DT 70 DT 63 DT 66 (p;f{:f-N

/ ~;f 4-

,. r

.;1

..w

'lj :j %

L3._L !

w.,

y.. ^*y; 5

, p; y;.

>V; Q,;y v'

_ w.

~

t. e,

,4 7

g.,

m:v.

m

,-T f_

-}.

[*.,;

YN. '

j, 2 l

s

-.f,.,,t

. l

j. >

c g, *g p y "; c.

43 p

A y

>@ M r -

4

.r m

DT61 DT 67 DT 62 DT 68

\\

.t

... ~ 7 i

.m.

+

a

c..w i

.. ;;p,:d,

.a "g

J 'l DT 72 DT 69 DT 71 DT 65 Figure 5-5. Charpy Impact Specimen Fracture Surfaces for Beaver Valley Unit 1 Reactor Pressure Vessel Lower Shell Plate B6903-1 (Transverse Orientation) 5-20

148174 i

l l

I MW_}' l-

{

.r,.

,.,f_g,

2}9jj.#1.

V g6jQ: 4

$$gy[R g!'.

i

?,, h lw%~ 4 YNY$$l, b.

$f:Y 'l 5

^^

k w;

.v : -..:.

v.

4 i

a i

g-3.,.+.

e w, _ _ ;A Y..

sp.

):.,;

.E v

)

fi i., ; Q '

U;.--

6 4

- - it

. e( t,,'ip[m:j n,

f a..

f U *. y,.

.. w :.,,

i gll [::(q,,'1

?Nf,'j$[?,j * {

[>;:.'i- $; '~

', ~.. < '.

I

)

?

s.;:37 pt W

r u

M l

DL 47 DL 46 DL 43 DL 44 la iv..

5

-h s

-1

+

A 1

i I

s s

4 t

4-f

^

9 t

DL 48 DL 41 DL 45 DL 42 i

l 1

Figure 5-6. Charpy impact Specimen Fracture Surfaces for Beaver Valley Unit 1 Reactor Pressure Vessel Lower Shell Plate B6903-1 (Longitudinal Orientation) 5-21 i

14817-2 WlM e

,y 8

~* '

l:_ _ __ l5

'y f

_fljl~ rY_& Aw),

I-bf

)

Y 2//y_lyQQ.?

a.r.s' fl Sga..r4

p [.

y;yy., j.

+

t jj f2,1.rl7f<

pm.

.Gbf

. 9::ga

,?

?

t-f(

Q~; '

)

nX 1

DW 70 DW 67 DW 61 DW 72

~

f wr,

j s

.r, i

)

t kk

$$li L

i DW 65 DW 63 DW 71 DW 68

.,7 ;.

- - ' r' y

l j

1 p

t 1

l l

l j

N.=-

1 DW 69 DW 62 DW 66 DW 64 i

j Figure 5-7. Charpy Impact Specimen Fracture Surfaces for Beaver Valley Unit 1 Weld Metal l

l 5-22

14817-1 I

M ~ME

<T g;

ou^'

g

..s.

p

'] P A

Ph 7

s ;.

;5

_ T t ',

', gu - 1 j-

,a fjl 4

jt9kl ~

1L - 'i}[

it% ',

^ ;

DH 72 DH 64 DH 65 DH 69 g&,

Sj7~

4

(;

f gf}li3j,

{4 ~ ~ ~

j DH 70 DH 63 DH 71 DH 62 DH 66 DH 68 D H (11 DH 67 Figure 5-8. Charpy lulpact Specimen Fracture Surfaces for Beaver Valley Unit 1 Weld Heat Affected Zone Metal

?

l 5-23

14s17.2 500 WELD METAL 400 b

300 N<

200 5g 150 PLATE (B6903) 100 80 Y

60 E

40 B

PLATE O (6903-1 LONGITUDINAL) m_j A B6903-1 PLATE FL (TRANSVERSE) 20 D WELD METAL g

l l

l l'

I i

18 39 10 2

4 6

8 10 2

3 4 56 i

FLUENCE (n/cm2) i l

Figure 5-9 Comparison of Actual versus Predicted 30 ft Ib Transition Temperature Increases for the Beaver Valley Unit 1 Reactor Vessel Material Based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 l

5-24 i

~ - -..-

14447-7 TEMPERATURE (*C)

O 50 100 150 200 250 300

,,g l

I I

100 O

700 m

90 ULTIMATE TENSILE x

STRENGTH 600 -

2 80 g

g m

h 0

2 500 -

cr 70

=

FW O.2% YIELD 60 STRENGTH 400 n

50 LEGEND:

O UNIRRADIATED G IRRADIATED 6.54 x 10 18 n/cm2 4

80 2

b 60 O

/

o e

=

>-[ 40 _ REDUCTION IN AREA d

b

[ TOTAL ELONGATION g

4 n

m 20

.v p

O b

b a

2 UNFORM ELONGAiION O

I I

O 100 200 300 400 500 600 TEMPERATURE (*F)

Figure 5-10. Tensile Properties for Beaver Valley Unit 1 Reactor Vessel Lower Shell Plate B6903-1 (Transverse) 5-25

_ _=

144474 TEMPERATURE (*C)

O 50 100 150 200 250 300

,,g l

I I

700 O

C 90 ULTIMATE TENSILE m

5 STRENGTH 600 g m

O 80 g

g 500 -

g 70 O.2% YIELD k

STRENGTH 400 60

--O g

50 LEGEND:

O UNIRRADIATED 9 IRRADIATED 6.54 x 10 18 n/cm2 80 h

n 60 S

X REDUCTION IN AREA

~

>-[ 40

!p TOTAL ELONGATION O

S 6

e 20 o

C O

D

'1 UNIFORM ELONGATION" I

I I

O O

100 200 300 400 500 600 TEMPERATURE (*F)

Figure 5-11. Tensile Properties for Beaver Valley Unit 1 Reactor VesselWeld Metal 5-26

14817-5 p.7

...,. 7

,my

,7 hy 10ms 1"

/

y 100iiG 3

67 '8

!) ; ; df'4 ' ]-

TENSILE SPECIMEN DT 12 (PLATE)

.1 21314 j b

6dddNthb!mh!4h!b!

I TESTED AT 275 F f W** ',,

47 7-3 z -,3 ; y g - r, m,

.cr.

g g N

.Ag 10iH5

{ *'!

1

, N ?il 1 GOTHS.

7-1 2

3. 4 G 7 8.-- 11

,. s..!.:h.. j ety!3id.dn (j'

[4 TENSlLE SPECIMEN DT 11 (PLATE)

E TESTED AT 550 F

[

1 p

(

p_w];10THSF T '.[ y. r 7-'8'3 -) 17 p

U"n"A N W

(

LW 2S r' __

,...,. 5?5WR'J 7l"'K'

1 - -

l l 10iHS.-

4S'4 QA 100THS.

M

C

. l'- 2' 3 2 4 ' 0'? ' 8 w7 :.

TENSILE SPECIMEN DW 11 (WELD)

TESTED AT 550 F 1

1 i

Figure 5-12. Fractured Tensile Specimens of Beaver Valley Unit 1 Reactor Vessel Lower Shell Plate B6903-1 and Weld Metal 5 27 I

l j

10447-2 oM O

b N

O s

z N

W o

P

.f N

g 6

.5 8

I E

m O

Z

5 C

o N

m C

d O

in I

O y

2 o

o o

.5 m

N Z

b m

a o

m 2

G m

m 8

o

'5.

o H

Y w

s O

E

.o u.

m o

.o l

l l

1 l

I I

I o

o o

o O

o o

m o

+

N o

m o

4 N

N o

m m

b m

+

M N

1sd ssagis l

5-28

SECTION 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 INTRODUCTION Knowledge of the neutron environment within the pressure vessel-surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be established. 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, on the other hand, is derived solely from analysis.

This section describes a discrete ordinates S, transport analysis performed for the Beaver Valley Unit 1 reactor to determine the fast neutron (E > 1.0 Mev) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules; and, in turn, to develop data for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule U is discussed and updated evaluations of dosimetry from Capsule V are presented.

6-2 DISCRETE ORDINATES ANALYSIS A plan view of the Beaver Valley reactor geometry at the core midplane is shown in Figure 6-1.

Since the reactor exhibits 1/8th core symmetry, only a 0*-45' sector is depicted. Eight irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel 3358e:1d/080585 1 6-1 i

surveillance program. The capsules are located at 45*, 55*, 65*, 165*, 245*,

285*, 295* and 305* relative to the reactor geometry 0* as shown in Figure 6-1.

A plan view of a single surveillance capsule attached to the thermal shield is shown in Figure 6-2.

The stainless steel specimen container is 1-inch square and approximately 40 inches in height. The containess are positioned axially such that the specimens are centered on the core midplane, thus spanning.the central 3.33 feet of the 12-foo! high reactor core.

From a neutronic standpoint, the surveillance capsule structures are significant.

In fact, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus between the thermal shield and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. Use of at least a two-dimensional computation is, therefore, mandatory.

In the analysis of the neutron environment within the Beaver Valley Unit 1 reactor geometry, two sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was utilized primarily to obtain spectrum-averaged reaction cross sections and gradient corrections for dosimetry reactions. The second set of calculations consisted of a series of adjoint analyses relating the fast neutron (E > 1.0 Mev) flux at the surveillance capsule locations and selected locations on the reactor vessel inner wall to the power distributions in the reactor core. These adjoint importance functions, when combined with cycle-specific core power distributions, yield the plant-specific fast neutron exposure at the sur-veillance capsule and pressure vessel locations for each operating fuel cycle.

The forward transport calculation was carried out in R,8 geometry using the DOT two dimensional discrete ordinates code [6] and the SAILOR cross-section libraryEU. The SAILOR library is a 47 group, ENOF-81V based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with a P expansion of the cross-sections. The energy 3

group structure used in the analysis is listed in Table 6-1.

3358e:Id/082685 6-2

I l

The design basis core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 3-loop 1

plants.

Inherent in the development of this design basis core power distribution is the use of an out-in fuel management strategy; 1.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2e uncertainty derived f rom the statistical evaluation of plant to plan't and cycle to cycle variations in peripheral power was used. Since it is j

unlikely that a single reactor would have a power distribution at the nominal

+2s level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results.

The adjoint analyses were also carried out using the P cross secdon 3

j approximation from the SAILOR library. Adjoint source locations were chosen i

at the center of each of the surveillance capsules as well as at positions along the inner diameter of the pressure vessel. Again, these calculations were run in R,9 geometry to provide power distribution importance functions for the exposure parameters of interest. Having the adjoint importance functions and appropriate core power distributions, the response of interest

]

is calculated as 1

R I (R,9) F (R,0) R dR de R,9 *

R 9

4 i

where:

RR,9 = Response of interest (e.g., 4 (E > 1.0 MeV)) at radius R and azimuthal angle 9.

I (R,9) = Adjoint importance function at radius R and azimuthal angle 9 l

F (R,9) = Full power fission density at radius R and azimuthal angle 9 l

It should be noted that as written in the above equation, the importance function 1(R,9) represents an integral over the fission distribution so that l

3358e:1d/080185 3 6-3 r

~

the response of interest can be related directly to the spatial distribution of fission density within the reactor core.

Core power distributions for use in the plant specific fluence evaluations for Beaver Valley Unit I were taken f rom fuel cycle design reports for each operating cycle. The specific power distribution data used in the analysis is provided in Appendix A of WCAP 10829[8]

The data listed in Appendix A represents cycle averaged relative assembly powers. Therefore, the adjoint results are in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence.

The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the ORNL PCA f acility as well as against the Westinghouse power reactor surveillance capsule data base The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be conservative by f rom 7-22%. When plant specific power distributions are used with the adjoint importance functions, the benchmarking studies show that fluence predictions are within i 15% of measured values at surveillance capsule locations.

6-3 NEUTRON 00SIMETRY The passive neutron flux monitors included in the Beaver Valley Unit I surveillance program are listed in Table 6-2.

The first five reactions in Table 6-2 are used as fast neutron monitors to relate neutron fluence (E > 1.0 Mev) to measured materials properties changes. To properly account for burnout of the product isotope generated by fast neutron reactions, it is necessary to also determine the magnitude of the thermal neutron flux at the monitor location. Therefore, bare and cadmium-covered cobalt-aluminum monitors were also included.

The relative locations of the various monitors within the surveillance capsules are shown in Figure 4-2.

The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers 3358e:ld/082685 6-4

at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are acconunodated within the dosimeter block located near the center of the capsule.

The use of passive monitors such as those listed in Table 6-2 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 operating history of the reactor o

The energy response of the monitor o

The neutron energy spectrum at the monitor location o

'The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two procedures.

First, the disintegration rate of product isotope per unit mass of monitor must be determined. Second, in order to define a suitable spectrum averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.

The specific activity of each of the monitors is determined using established ASTM procedures.[10,ll,12,13,14] Following sample preparation, the activity of each monitor is determined by means of a lithium-drif ted germanium, Ge(Li),

gansna spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed f rom Beaver Valley Unit 1, the overall 2a deviation in the measured data is 3358e:ld/082685 6-5

determined to be 110 percent. The neutron energy spectra are determined analytically using the method described in Section 6-1.

Having the measured activity of the monitors and the neutron energy spectra at the locations of interest, the calculation of the neutron flux proceecif as follows.

The reaction product activity in the monitor is expressed as:

N P

-At

3) e-At

[c(E)$(E)

(1 - e i

d (6-1)

R=

f Y

g p

max E

j=1 where:

induced product activity R

=

Avagadro's number l

N

=

atomic weight of the target isotope A

=

weight f raction of the t3rget isotope in the target material f

=

g number of product atoms produced per reaction Y

=

energy-dependent reaction cross section c(E)

=

energy-dependent neutron flux at the monitor location with the

$(E)

=

reactor at full power P) average core power level during irradiation period j

=

maximum or reference core power level P

=

max decay constant of the product isotope 1

=

3358e:ld/080185 6 6-6

tj length of irradiation period j

=

td decay time following irradiation period j

=

Since neutron flux distributions are calculated using multigroup transport methods and, further, since the prime interest is in the fast neutron flux above 1.0 Mev, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-1) is replaced by the following relation.

s(E) $(E)dE a $(E > 1.0 Mev)

=

E where:

N

=

[

a(E) + (E)dE

{

a $g g

6=1

=

N

[

$(E)dE

g 1.0 Mev 6= 1.0 Mev Thus, equation (6-1) is rewritten N

N P

-At

-At R=[f Y a $ (E > 1.0 Mev)

(1-e

3) e d

g p

max j=1 or, solving for the neutron flux, 4 (E > 1.0 Mev) =

(6-2) gf P

-At)) '-At 3

d Y#

II~'

A i

P j=1 max 3358e:1d/080185 7 6-7

The total fluence above 1.0 Mev is then given by N

9 (E > 1.0 Mev) = $ (E > 1.0 Mev) {

t)

(6-3)

-x s.,

l where:

N[

t) total effective full power seconds of reactor

=

j =1 max operation up to the time of capsule removal An assessment of the thennal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered CoS9 (n,y) Co60 data by means of cadmium ratios and the use of a 37-barn 2200 m/sec cross section. Thus,

,I D - l ',

4

, bare L

0

.i (6-4) 4

-At)} '-At N

F d

g II~'

A i

d P

5 max where:

Rbare D is defined as gCd covered 6-4 TRANSPORT ANALYSIS RESULTS i

Calculated fast neutron (E > 1.0 Mev) exposure results for 8eaver Valley Unit 1 are presented in Tables 6-3 through 6-10 and in Figures 6-3 through 6-7.

Data is presented at several azimuthal locations on the inner radius of the pressure vessel as well as the center of each surveillance capsule.

3358e:1d/080185 8 6-8

In Tables 6-3 through 6-6 plant specific neutron flux and fluence levels at 0*,15*, 30', and 45' on the pressure vessel inner radius are listed for each operating cycle of Beaver Valley Unit 1.

Similar data for the center of the surveillance capsules located at 15*, 25', 35', and 45' are given in Tables 6-7 through 6-10, respectively.

In addition to the calculated data given for the surveillance capsule locations, measured fluence data f rom withdrawn surveillance capsules are also presented for comparison with analytical results. Capsules were removed from the 15* location at the end of Cycles 1 (Capsule V), and the 25' location at the end of Cycle 4 (Capsule U).

Graphical presentations of the plant specific fast neutron fluence at key locations on the pressure vessel as well as at the surveillance capsule center are shown in Figures 6-3 and 6-4 as a function of full power operating time for Beaver Valley Unit 1.

The pressure vessel data is presented for the 0*

location on the circumferential weld as well as for the 45' longitudinal weld. The corresponding fluence levels at the center of 15*, 25*,'35' and 45' surveillance capsules are also depicted.

In regard to Figure 6-3 and 6-4, the solid portions of the fluence curves are based directly on the plant specific evaluations presented in this report.

The dashed portions of these curves, however, involve a projection into the future. Since Beaver Valley Unit 1 has initiated a form of low leakage fuel management, the average neutron flux at the key locations over the low leakage fuel cycles was used for all temporal projections.

In particular, the neutron flux average over Cycles 4 and 5 was used to project future fluence levels (8)

It should be noted that implementation of a more severe low leakage pattern would act to reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those projections. In any event it would be prudent to update the fluence analysis as the design of each future fuel cycle evolves.

3358e:ld/082685 6-9

In Figure 6-5, the azimuthal variation of seximum fast neutron (E > 1.0 Mev) fluence at the inner radius of the pressure vessel is presented as a function of azimuthal angle. Data are presented for both current and projected end-of-lif e conditions.

In Figure 6-6, the relative radial variation of fast neutron flux and fluence within the pressure vessel wall is presented.

Similar data showing the relative axial variation of fast neutron flux and fluence over the beltline region of the pressure vessel is shown in Figure 6-7.

A three-dimensional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figures 6-5 j

through 6-7 along with the relation

$(R,0,Z) = $(9) F(R) G(Z)

Fast neutron fluence at location R, 9. Z within where: $ (R,9,Z)

=

the pressure vessel wall Fast neutron fluence at azimuthal location 9 on

$ (0)

=

the pressure vessel inner radius $ rom Figure 6-5 Relative fast neutron flux at depth R into the F (R)

=

pressure vessel from Figure 6-6 Relative fast neutron flux at axial position Z from G (Z)

=

Figure 6-7 Analysis has shown that the radial and axial variations within the vessel wall i

are relatively insensitive to the implementation of low leakage fuel management schemes. Thus, the above relationship provides a vehicle for a reasonable evaluation of fluence gradients within the vessel wall.

6-5 00SIMETRY RESULTS The irradiation history of the Beaver Valley Unit I reactor is given in Table 6-11.

The data were obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report". Measured saturated activities of the flux monitors 3358e:1d/080585 10 6-10

contained in Capsules V and U are listed in Tables 6-12 and 6-13, respectively. The data are presented as measured at the actual monitor radial locations as well as adjusted to the capsule radial center. Adjustment factors for each monitor at 15' and 25*, which are given in Table 6-14, were obtained from radial reaction rate gradients at each location as calculated f rom the forward transport calculation described in Section 6-2.

In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The neutron energy spectrum at the radial and azimuthal center of each surveillance capsule, shown in Table 6-15, was taken f rom the forward calculation. The resulting spectrum-averaged cross sections for each of the five fast neutron reactions are given in Table 6-16.

The fast neutron (E > 1.0 Mev) flux levels derived for Capsules V and U are presented in Tables 6-12 and 6-13.

The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in Table 6-17.

Due to the relatively low thermal neutron flux at the capsule locations, no burnup correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than one percent for the NiS8(n.p)CoS8 reaction and even less significant'for all of the other fast reactions.

A comparison of the measured and calculated fast neutron fluence for each flux monitor of Capsules V and U is shown in Table 6-18.

Examination of the data in Table 6-18 shows that neutron fluences corresponding to the average of the i

monitors at each location agree within 11% of the calculated fluences based on the plant-specific power distributions.

In the case of Capsule U, the measared fluences derived from the individual flux monitors agree well with I

the ulculated fluence with the exception of the result from the Fe54(n.p)Mn54 and Ni (n.p) Co reactions. Since Mn

and Co have short half lives relative to the total operating time of the plant, the fast neutron flux derived from the measurement is characteristic of the most recent fuel cycle. Table 6-8 shows that the introduction of low leakage fuel management after the third cycle caused the cycle-averaged fast neutron flux to decrease significantly at the 25' surveillance capsule location. Thus, the 3358e:ld/082685 6-11 L

fast flux derived from the Mn and Co activity of Capsule U is lower than the fast flux derived from the other monitors and yields a fast neutron fluence which is not representative of the entire operating plant history.

The current dosimetry evaluations for Capsule U and updated evaluations for Capsule V provide a consistent set of fluence data for the Beaver Valley Unit I reactor that is based on current state of the art neutron transport methodology as well as on the most recent nuclear data pertinent to the dosimeter materials.

6-6 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES As discussed in Section 6-4, plant specific fluence evaluations for the center of surveillance capsules located at 15', 25*, 35*, and 45* were presented in Figure 6-4 for Beaver Valley Unit 1.

The data presented on those curves represent the best available information upon which to base the future withdrawal schedules for capsules remaining in the Beaver Valley Unit I reactor.

In the past, withdrawal schedules have been based on the assumption of a constant exposure rate at the surveillance capsule center and a constant lead factor relating capsule exposure to maximum vessel exposure. With the widespread implementation of low leakage fuel management neither of these assumptions can be assumed to be universally valid.

It becomes prudent, therefore, to utilize the actual anticipated capsule exposure in conjunction with appropriate materials properties data to establish capsule withdrawal dates that will provide experimental information that is of most benefit to the utility.

In evaluating future withdrawal schedules, it must be remembered that the fluence projections shown in Figure 6-4 assume continued operation with the low leakage fuel management scheme currently in place. The validity of this assumption should be verified as each new fuel cycle evolves and if significant changes occur withdrawal schedules should be adjusted accordingly.

6-12

TA8LE 6-1 47 GROUP ENER6Y STRUCTURE Lower Energy Lower Energy Group (Nev)

Group (Nev) 1 14.19*

25 0.183 2

12.21 26 0.111 3

10.00 27 0.0674 4

8.61 28 0.0409 5

7.41 29 0.0318 6

6.07 30 0.0261 7

4.97 31 0.0242 8

3.68 32 0.0219 9

3.01 33 0.0150

-3 10 2.73 34 7.10 x 10

-3 11 2.47 35 3.36 x 10

-3 12 2.37 36 1.59 x 10

~

13 2.35 37 4.54 x 10 14 2.23 38 2.14 x 10 15 1.92 39 1.01 x 10 l

16 1.65 40 3.73 x 10

~

-5 17 1.35 41 1.07 x 10 18 1.00 42 5.04 x 10

-6 l

19 0.821 43 1.86 x 10

~

20 0.743 44 8.76 x 10

~

21 0.608 45 4.14 x 10

~

22 0.498 46 1.00 x 10 l

23 0.369 47 0.00 t

24 0.298 i

a

The upper energy of group 1 is 17.33 Mev.

3358e:1d/080185 15 6-13

TA8LE 6-2 i

NUCLEAR PARAMETERS FOR NEUTRON FLUX M0hlTORS Target Fission Monitor Reaction Weight

Response

Product Yield Material of Interest Fraction Range Half-Life

(%)

Copper Cu63(N,a)Co60 0.6917 E>4.7 Mev 5.27 years t

Iron Fe54(n p)Mn54 0.0585 E>1.0 Mev 314 days Nickel Ni 0(n.p)CoS8 0.6777 E>1.0 Mev 71.4 days Urenium-238*

U238(n f)CsI3 1.0 E>0.4 Mev 30.2 years 6.3 Neptunium-237*

Np (n f)Cs 1.0 E>0.08 Mev 30.2 years 6.5 60 Cobalt-Aluminum

CoS9(n,y)Co 0.0015 0.4eV<E<0.015 MeV 5.27 years 60-Cobalt-Aluminum CoS9(n,y)Co 0.0015 E<0.015 Mev 5.27 years

Denotes that monitor is cadmium shielded.

3358e:1d/080185 16 6-14

i TA8LE 6-3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE 8eltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm21 Cycle Time Flux Plant Design No.

(EFPS)

(n/cm2-sec)

SDecific Basis (0) __

I 10 18 18 1

3.66 x 10 5.36 x 10 1.96 x 10 2.37 x 10 0

18 18 2

2.26 x 10 5.76 x 10 3.26 x 10 3.83 x 10 10 18 18 3

2.49 x 10 6.00 x 10 4.76 x 10 5.44 x 10 I

4 2.91 x 10 4.35 x 10 6.02 x 10 7.32 x 10 10 2

(a) Design Basis $ = 6.47 x 10 n/cm,,,e 3358e:1d/080185 17 6-15

TA8LE 6-4 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15' AZIMUTHAL ANGLE 8eltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm21 Cycle Time Flyx Plant Design No.

(EFPS)

(n/ce'-sec)

SDecific 8 asis (8) 1 3.66 x 10 2.56 x 10 9.37 x 10 1.09 x 10 10 18 18 2

2.26 x 10 2.82 x 10 1.57 x 10 1.76 x 10 18 3

2.49 x 10 2.89 x 10 2.29 x 10 2.50 x 10 10 18 18 4

2.91 x 10 2.12 x 10 2.91 x 10 3.36 x 10 10 2

(a) Design 8 asis $ = 2.96 x 10 n/cm -sec 4

1 t

3358e:1d/08018518 6-16

TABLE 6-5 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE Beltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm21 Cycle Time Flux Plant Design No.

(EFPS)

(n/cm2-Sec)

SDecific BasisIO)__

1 3.66 x 10 1.36 x 10 4.98 x 10 6.33 x 10 I

10 lI 18 2

2.26 x 10 1.56 x 10 8.50 x 10 1.02 x 10 10 18 18 3

2.49 x 10 1.51 x 10 1.23 x 10 1.45 x 10 I

10 18 18 4

2.91 x 10 1.09 x 10 1.54 x 10 1.96 x 10 10 gfc,2 (a) Design Basis $ = 1.73 x 10

-sec A

l l

l 3358e:1d/080185 19 6-17

TABLE 6-6 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE Beltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm21 Cycle Time Flyx Plant Design No.

(EFPS)

(n/cm4-sec)

Specific Basis (a) 1 3.66 x 10 9.03 x 10' 3.30 x 10 3.85 x 10 10 I

I 2

2.26 x 10 1.06 x 10 5.70 x 10 6.22 x 10 3

2.49 x 10 9.73 x 10' 8.12 x 10" 8.84 x 10 9

18 18 4

2.91 x 10 7.12 x 10 1.02 x 10 1.19 x 10 10 2

(a) Design Basis $ = 1.05 x 1_0 n/cm -sec 3358e:1d/080185 20 6-18

l l

TABLE 6-7 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15* SURVEILLANCE CAPSULE CENTER Beltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

Cycle Time Flux Plant Design Capsule "V" No.

(EFPS)

[n/cm2-sec)

SDecific Basis (4)

Data 1

3.66 x 10 8.85 x 10 3.24 x 10 3.75 x 10 2.91 x 10 0

18 2

2.26 x 10 9.75 x 10 5.44 x 10 6.06 x 10 3

2.49 :t 10 9.97 x 10 7.93 x 10 8.61 x 10 0

4 2.91 x 10 7.28 x 10 1.00 x 10 '

1.16 x 10 '

I (a) Desic Basis + = 1.02 x 10 n/cm -sec 3358e:1d/080185 21 6-19

TABLE 6-8 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 25' SURVEILLANCE CAPSULE CENTER Beltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm2) i Cycle Time Flux Plant Design Capsule "U" No.

(EFPS)

(n/cm2-sec)

Specific Basis (a)_

Data 7

10 18 18 1

3,66 x 10 S.52 x 10 2.02 x 10 2.38 x 10 7

10 18 18 2

2.26 x 10 6.25 x 10 3.43 x 10 3.85 x 10 I

10 18 18 3

2.49 x 10 6.21 x 10 4.98 x 10 5.47 x 10 7

10 18 18 18 4

2.91 x 10 4.55 x 10 6.30 x 10 7.36 x 10 6.54 x 10 10 2

(a) Design Basis + = 6.50 x 10 n/cm -sec l

l l

l 3358e:1d/080185 22 6-20

TABLE 6-9 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 35* SURVEILLANCE CAPSULE CENTER Beltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm2L Cycle Time Flux Plant Design No.

(EFPS)

(n/cm2-sec)

Specific Basis (a) 18 1

3.66 x 10 3.71 x 10 1.36 x 10 1.63 x 10 10 18 18 2

2.26 x 10 4.31 x 10 2.33 x 10 2.63 x 10 0

18 18 3

2.49 x 10 4.09 x 10 3.35 x 10 3.74 x 10 10 18 8

4 2.91 x 10 2.93 x 10 4.20 x 10 5.03 x 10 10 (a) Design 8 asis + = 4.44 x 10 n/cm -sec 3358e:1d/080185 23 6-21

TA8LE 6-10 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45' SURVEILLANCE CAPSULE CENTER 8eltline Region Irradiation Cycle Avg.

Cumulative Fluence (n/cm2L Cycle Time Flyx Plant Design No.

(EFPS)

(n/cm4-sec)

Specific Basis (a) 10 18 1

3.66 x 10 2.91 x 10 1.07 x 10 1.29 x 10 10 18 18 2

2.26 x 10 3.42 x 10 1.84 x 10 2.08 x 10 0

18 3

2.49 x 10 3.13 x 10 2.62 x 10 2.95 x 10 I

10 18 18 4

2.91 x 10 2.27 x 10 3.28 x 10 3.98 x 10 (a) Design Basis + = 3.51 x 1010,fc,2-sec G

3358e:1d/080185 24 6-22

TA8LE 6-11 IRRADIATION HISTORY OF SURVEILLANCE CAPSULES REMOVED FROM BEAVER VALLEY UNIT 1 Irradiation Decay

P,yg P

P,y /

Tine Time ref Month (MW)

(MW)

P (days)

(days) ref 5/76 - 6/76 50 2652

.019 51 3199 7/76 103 2652

.039 31 3168 8/76 127 2652

.048 31 3137 9/76 591 2652

.223 30 3107 10/76 774 2652

.292 31 3076 11/76 265 2652

.100 30 3046 12/76 536 2652,

.202 31 3015 1/77 263 2652

.099 31 2984 2/77 0

2652

.000 28 2956 3/77 960 2652

.362 31 2925 4/77 1411 2652

.532 30 2895 5/77 2267 2652

.855 31 2864 6/77 1467 2652

.553 30 2834 7/77 1955 2652

.737 31 2803 8/77 1337 2652

.504 31 2772 9/77 0

2652

.000 30 2742 10/77 40 2652

.015 31 2711 11/77 1475 2652

.556 30 2681 12/77 2530 2652

.954 31 2650 1/78 2100 2652

.792 31 2619 2/78 2464 2652

.929 28 2591 3/78 2546 2652

.960 31 2560 4/78 1801 2652

.679 30 2530 5/78 0

2652

.000 31 2499 3558e:1d/082685 6-23 1

TA8LE 6-11 (Continued)

Irradiation Decay

P,,,

P P,,g/

Time Time ref Month (MW)

(MW)

P (days)

(days) ref 6/78 225 2652

.085 30 2469 7/78 2222 2652

.838 31 2438 8/78 - 11/78 0

2652

.000 122 2316 12/78 695 2652

.262 31 2285 1/79 1124 2652

.424 31 2254 2/79 2273 2652

.857 28 2226 3/79 485 2652

.183 31 2195 4/79 - 7/79 0

2652

.000 122 2073 8/79 879 2652

.331 31 2042 9/79 2037

, 2652

.768 30 2012 10/79 1061 2652

.400 31 1981 11/79 971 2652

.366 30 1951 CAPSULE V WITHDRAWN 12/79 - 10/80 0

2652

.000 336 1615 i

11/80 202 2652

.076 30 1585 12/80 1278 2652

.482 31 1554 1/81 1490 2652

.562 31 1523 2/81 1337 2652

.504 28 1495 j

3/81 0

2652

.000 31 1464 4/81 1294 2652

.488 30 1434 5/81 2005 2652

.756 31 1403 6/81 2631 2652

.992 30 1373 7/81 1451 2652

.547 31 1342 8/81 2485 2652

.937 31 1311 9/81 2480 2652

.935 30 1281 4

6-24 3358e:1d/080585 26

TA8LE 6-11 (Continued)

Irradiation Decay

P,yg P

P,,g/

Time Time ref Month (MW)

(MW)

P (days)

(days) ref 10/81 2265 2652

.959 30 1250 11/81 2543 2652

.959 30 1220 12/81 1604 2652

.605 31 1189 1/82 - 6/82 0

2652

.000 181 1008 7/82 1268 2652

.478 31 977 8/82 2108 2652

.795 31 946 9/82 1358 2652

.512 30 916 10/82 e 2161 2652

.815 31 885 11/82 2594 2652

.978 30 855 12/82 2437 2652

.919 31 824 1/83 2355 2652

.888 31 793 2/83 2384 2652

.899 28 765 3/83 2596 2652

.979 31 734 4/83 2652 2652 1.00 30 704 5/83 2318 2652

.874 31 673 6/83 817 2652

.308 30 643 7/83 - 9/83 0

2652

.000 92 551 10/83 2265 2652

.854 31 520 11/83 2230 2652

.841 30 490 12/83 2424 2652

.914 31 459 1/84 1968 2652

.742 31 428 2/84 2453 2652

.925 29 399 3/84 1597 2652

.602 31 368 4/84 2368 2652

.893 30 338 5/84 2363 2652

.891 31 307 3358e:1df080585 27 6-25

TA8LE 6-11 (Continued)

Irradiation Decay

P,,,

P P,,g/

Time Time ref Month (MW)

(MW)

P (days)

(days) ref 6/84 1944 2652

.733 30 277 7/84 2127 2652

.802 31 246 8/84 2381 2652

.853 31 215 9/84 2122 2652

.800 30 185 10/84 2572 2652

.970 31 154 l

TOTAL GENERATION = 3.58 EFPY 8

= 1.13 x 10 EFPS

Decay time is referenced to the counting date of Capsule U Dosimeters (4/3/85) l l

l l

l 3358e:1d/080585 28 6-26 l

TABLE 6-12 MEASURED FLUX MONITOR ACTIVITIES FROM BEAVER VALLEY UNIT 1. CAPSULE V Adjusted Reaction Radial Measured Saturated Saturated Measured and Axial Location Activity Activity Activity

$(E>1.0 MeV) 2 (n/cm _3,c)

Location (cm)

(dPS/am)

Cu63(n.a)Co60 4

5 5

10 Top-Middle 190.92 4.24 x 10 3.85 x 10 3.70 x 10 8.24 x 10 4

5 5

10 Middle 190.92 4.24 x 10 4.01 x 10 3.85 x 10 8.57 x 10 4

0 5

10 Bottom-Middle 190.92 4.28 x 10 3.89 x 10 3.73 x 10 8.31 x 10 10 Avera9e 8.37 x 10 Fe$4(n p)Mn54 5

6 6

10 Top 191.42 5.84 x 10 3.69 x 10 3.91 x 10 8.08 x 10 5

6 6

10 Top-Middle 191.42 5.35 x 10 3.39 x 10 3.59 x 10 7.42 x 10 j

5 6

6 10 Middle 191.42 5.62 x 10 3.55 x 10 3.76 x 10 7.77 x 10 5

6 6

10 Bottom-Middle 19'1.42 5.32 x 10 3.36 x l'0 3.56 x 10 7.36 x 10 5

6 6

10 Bottom 191.42 5.37 x 10 3.59 x 10 3.59 x 10 7.42 x 10 10 Average 7.61 x 10 Ni 8(n.p)CoS8 10 Top-Middle 191.92 9.57 x 10' 4.67 x 10 5.37 x 10 7.63 x 10 5

7 I

10 Middle 191.92 9.75 x 10 4.76 x 10 5.47 x 10 7.77 x 10 7

7 10 Bottom-Middle 191.92 9.06 x 10 4.42 x 10 5.08 x 10 7.22 x 10 10 Average 7.54 x 10 U238(n.f)Cs137 5

6 6

10 Middle 191.15 1.20 x 10 4.50 x 10 4.50 x 10 8.37 x 10 Np 37(n,f)Cs137 I

6 10 Middle 191.15 9.70 x 10 3.64 x 10 3.64 x 10 7.84 x 10 Note: U238(n f)Cs137 activities have been corrected by a factor of 0.88 to account for U-235 impurity and Pu-239 buildup.

3358e:1d/080585 29 6-27

TABLE 6-13 MEASURED FLUX MONITOR ACTIVITIES FROM BEAVER VALLEY UNIT 1. CAPSULE U Adjusted Reaction Radial Measured Saturated Saturated Measured and Axial Location Activity Activity Activity 4(E>1 0 MeV) b (n/cm -sec) location (cm)

(dPS/cm)

(dPS/cm) 60 Cu63(n a)Co 10 Top-Middle 190.92 9.44 x 10*

3.21 x 10' 3.08 x 10 6.01 x 10 5

10 Middle 190.92 1.00 x 10 3.40 x 10 3.26 x 10 6.37 x 10 4

5 5

10 Bottom-Middle 190.92 9.30 x 10 3.76 x 10 3.03 x 10 5.92 x 10 10 6.10 x 10 Average Fe54(n.p)Mn54 6

6 6

10 Top 191.42 1.21 x 10 2.48 x 10 2.63 x 10 5.03 x 10 6

6 6

10 Top-Middle-191.42 1.13 x 10 2.32 x 10 2.46 x 10 4.70 x 10 6

6 6

10 Middle 191.42 1.22 x 10 2.50 x 10 2.65 x 10 5.07 x 10 6

6 6

10 Bottom-Middle-191.42 1.16 x 10 2.38 x 10 2.52 x 10 4.82 x 10 6

6 6

10 Bottom 191.42 1.14 x 10 2.34 x 10 2.48 x 10 4.74 x 10 10

' Average 4.87 x 10 Ni$0(n.p)Co$

0 10 Top-Middle 191.92 4.81 x 10 2.52 x 10 2.92 x 10 3.88 x 10 6

10 Middle 191.92 5.22 x 10 2.73 x 10 3.17 x 10 4.21 x 10 6

I 10 Bottom-Middle 191.92 4.92 x 10 2.58 x 10 2.99 x 10 3.97 x 10 10 4.02 x 10 Average 137 U238(n f)Cs 6

6 10 Middle 191.15 2.54 x 10 3.31 x 10 3.31 x 10 6.00 x 10 2

I Mp (n f)Cs 6

10 Middle 191.15 2.14 x 10 2.79 x 10 2.79 x 10 6.18 x 10 Note: U2 0(n.f)Cs137 activities have been corrected by a factor of 0.88 to account for U-235 impurity and Pu-239 buildup.

3358e:1d/080585 30 6-28

1 TABLE 6-14 FACTORS TO ADJUST MEASURE ACTIVITIES TO THE j

RADIAL CENTER OF THE SURVEILLANCE CAPSULES l

l l

Radial Location l

Reaction 190.92 191.15 191.42 191.92 Capsule V Cu63(n.a)Co 0 0.96 1.00 1.05 1.14 Fe54(,,p),,54 0.96 1.00 1.06 1.17 l

Ni 0(n.p)Co 0.95 1.00 1.05 1.15 8(n.f)Cs137 0

0.97 1.00 1.07 1.19 Np (n,f)Cs 0.96 1.00 1.05 1.17

$(E<.414 eV) 1.00 Capsule U Cu (n.a)Co 0.96 1.00 1.05 1.15 Fe *(n.p)Mn 0.96 1.00 1.06 1.17 Ni (n.p)Co 0.94 1.00 1.04 1.16 38(n,f)Cs 0.96 1.00 1.07 1.18 I

0 Np (n,f)Cs 0.97 1.00 1.07 1.16

$(E<.414 eV) 1.00 1

3358e:1d/080585 31 6-29

Table 6-15 CALCULATED NEUTRON SPECTRA AT THE CENTER OF CAPSULES V AND U GROUP

$ (n/cm - sec) 4(n/cm - sec) l Capsule V Capsule U Capsule V Capsule U 7

7 10 10 1

2.40 x 10 1.79 x 10 25 3.63 x 10 2.10 x 10 10 10 2

8.78 x 10 6.56 x 10 26 3.52 x 10 2.03 x 10 8

8 10 10 3

3.04 x 10 2.23 x 10 27 2.80 x 10 1.61 x 10 8

8 10 10 4

5.54 x 10 4.03 x 10 28 2.07 x 10 1.18 x 10 5

9.20 x 10 7.28 x 10 29 6.82 x 10' 3.92 x 10' 8

8 6

2.06 x 10 1.48 x 10' 30 3.82 x 10 2.19 x 10' 9

9 9

9 7

2.83 x 10' 1.99 x 10 31 8.36 x 10' 4.76 x 10 8

5.62 x 10 3.82 x 10' 32 5.03 x 10 2.85 x 10' 8

9 10 9

9 4.92 x 10' 3.23 x 10' 33 1.17 x 10 6.71 x 10 9

9 10 10 10

.4.00 x 10 2.61 x 10 34

.1.79 x 10 1.02 x 10 9

8 10 10 11 4.70 x 10 3.05 x 10 35 2.59 x 10 1.48 x 10 10 10 12 2.34 x 10' 1.52 x 10' 36 2.29 x 10 1.30 x 10 8

8 10 10 13 7.11 x 10 4.57 x 10 37 3.47 x 10 1.98 x 10 9

10 10 14 3.02 x 10 2.24 x 10' 38 1.99 x 10 1.13 x 10 9

9 10 10 15 8.93 x 10 5.73 x 10 39 2.14 x 10 1.21 x 10 10 9

10 10 16 1.14 x 10 7.16 x 10 40 2.89 x 10 1.53 x 10 10 10 10 10 17 1.68 x 10 1.05 x 10 41 3.52 x 10 1.99 x 10 10 10 10 10 18 3.21 x 10 1.96 x 10 42 2.02 x 10 1.14 x 10 10 10 10 10 19 2.29 x 10 1.38 x 10 43 2.45 x 10 1.38 x 10 10 9

10 20 1.11 x 10 6.67 x 10 44 1.62 x 10 9.09 x 10' 21 3.44 x 10 2.02 x 10 45 1.37 x 10 7.66 x 10' 10 10 10 10 10 10 10 22 2.64 x 10 1.53 x 10 46 2.62 x 10 1.47 x 10 10 10 10 10 23 3.10 x 10 1.81 x 10 47 4.83 x 10 2.69 x 10 10 10 25 2.81 x 10 1.63 x 10 1816S/js 6-30

TABLE 6-16 SPECTRUM AVERAGED REACTION CROSS-SECTIONS AT THE CENTER OF BEAVER VALLEY UNIT 1 SURVEILLANCE CAPSULES a (barns)

Reaction CaDsule V CaDsule U Cu (n,a)Co

.000679

.000775 Fe54(n.p)Mn54

.0745

.0806 Ni (n.p)Co

.100

.107 U

(n f)Cs

.354

.363 Np (n f)Cs 2.81 2.73 l' a(E) $(E) dE a

I"1.0 MeV $(E) dE 3358e:1d/080585 32 6-31

Table 6-17 THERMAL NEUTRON FLUX DATA FROM CAPSULES V AND U Axial Saturated Activity (dPs/qm)

$Th Capsule location Bare cd-Covered (n/cm2-sec)

I 10 V

Top 6.57 x 10 2.35 x 10 7.43 x 10 7

I 10 Bottom 6.57 x 10 2.32 x 10 7.51 x 10 7

I 10 0

Top 3.99 x 10 1.45 x 10 4.49 x 10 I

7 10 Bottom 3.94 x 10 1.42 x 10 4.46 x 10 6-32 1816S/js

TABLE 6-18 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUENCE FOR CAPSULES V AND U Irradiation

+ (E > 1.0 Mev)

Capsule Reaction Time (n/cm )

(EFPS)

Measured Calculated 8

V Cu (n.a)Co 3.66 x 10 3.06 x 10 Fe (n.p)Mn 2.79 x 10 8

Ni (n.p)Co 2.76 x 10 U

(n.f)Cs 3.06 x 10 l

18 Np (n f)Cs 2.87 x 10 18 Average 2.91 x 10 3.24 x 10 8

IO U

Cu (n,a)Co 1.13 x 10 6.89 x 10 54(n.p)Mn 5.50 x 10 54 18 Fe Ni (n.p)Co 4.54 x 10

238(n.f)Cs137 18 U

6.78 x 10 8

Np (n,f)Cs 6.98 x 10 8

18 Average 6.54 x 10 6.30 x 10

Not included in average fluence.

i 335Be:1d/080585 33 6-33 i

L

18.153 15 Figure 6-1 Beaver Valley Unit 1 Reactor Geometry 0*

15 (CAPSULES V, X)

/

25 (CAPSULES U, W, Y)

/

35 (CAPSULES T, Z)

/

45 (CAPSULE S)

////

/

/,

/

/ //m l'

/

j

~

/ /

PRESSURE VESSEL j

l, ll /s

't //

/

/,,!

THERMAL SHIELD

,/

/-

7

/

)

l

/

5 l /

/

-- / /

REACTOR CORE

~Il//

_ //

///

?/

6-34

18.152-16 Figure 6-2 Reactor Vessel Surveillance Capsule (15, 25, 35, 45 )

CHARPY

///////// '

SPECIMEN f;j g

./

THERMAL SHIELD 6-35

64 49 *354E.1 1020 9

8-7H 6h 5

4 p#s#

/

3 p#

CIRCUMF E R ENTIAL WELD,00

/

2 p

/

1019

/

9 l

/

p 7

6

=

'h 4

~

/p#p/

5-0 LONGITUDIN AL WELDS. 45

]

0z

/

a 3

3 j/

~

/

2

/

2-

/

5

/

z

/

1018 9

8 7

6 5

4 l

3 PLANT SPECIFIC PROJECTED

= = = = = = =

2 1017 l

0 5

10 15 20 25 30 35 OPER ATING TIME (EFPY)

Figure 6-3 Maximum Fast Neutron (E > 1 MeV) Fluence at the Eeltline Weld Lccations as a Function of Full Power Operating Time 6-36

64 43 12548 2 1021 9

8 7

6 5

4 3

2 20 10 9

8 g

p # 150 7

g 6

p#

h 5

/

=

g

,/

,s 25*

O

/

1

/

1

/

/p

/,, /

's 3 5*

3

~

2

/

/p

/s's/,s 45*

l 5

/

2

/

/p/

3

/

Z'

/

\\

/

,/

s'

/

l 1019

/

9

/

i 8

7 l

/

l 6

l/

l

/

5 I

4

/

3 PLANT SPECIFIC

-- -- PROJECTED 2

j i

1018 I

l l

I l

0 5

10 15 20 25 30 35 OPER ATING TIME (EFPY)

Figure 6-4 Max mum Fast Neutron (E>1 MeV) Fluence at the Center of the Surveillance Cacsales as a Function of Fuil Power Operation Time 6-37 6.

64-49 13548 3 1020 9

8 7

6-SL

~~s%

4

\\ \\

3

\\

2

\\g

\\ % N N%s%

19 N

10 9

\\g'"%'32 EFPY 8

7 r{

6

=

d 5

I 4

y 2y 3

E E

2 32 1018 3.6 EFPY 9

8 7

6 5

4 3

2 PLANT SPECtFIC PROJECTED f

1017

)

0 10 20 30 40 50 60 70 AZIMUTHAL ANGLE toeg )

Figure 6-5 Maximum Fas: Neutron (E>1.0 MeV) Fluence at the PressureVesselInc.er Rac;us as a Function of Azimutnal Angle 6-38

64 49 *2541 4 109h 8t-7&

6 k-5 4

t 3

2-199.39 1.0 204 39 9 f-u 8 L-

$7

$6 INNER RADIUS X34 1/4T 20 3

~

214.39 a

2 2

$=

5 a

C 219.39 0.1 9

8 3/4T 7

450 6

5 00 4

3 OUTER R ADIUS 2

i I

I 1

I I

l l

I 195 199 203 207 211 215 219 223 R ADIUS (cms Figure 6-6. Relative Radial Distributien of Fast Neutron (E > 1.0 MeV) Flux and Fluence Witnin the Pressure Vessel Wall 6-39

13548-10 9

8 7

6 5

4 3

2 1.0 9

8 7

6 5

w y

4 3

t) 2 3

u.

2

@ 0.g _-

8 w

7 Z

6 y

5

{

4 d

3 a:

2 0.01 9

8 7

6 5

4 3

2 CORE MIDPLANE I

I I

O.001 300

-200 100 0

100 200 300 400 DISTANCE FROM CORE MIDPLANE (cml Fig 6re 6-7 Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 6-40

SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule per ASTM E185-82 is recommended for future capsules to be removed from the Beaver Valley Unit i reactor vessel.

Vessel Estimated Location Removal Fluence Capsule (deg)

Time [a]

(n/cm )

2 18[b]

V 165*

1.16 (removed) 2.91 x 10 U

65*

3.58 (removed) 6.'54 x 1018[b]

18 I

W 245*

6 9.40 x 10 19 Y

295*

15 2.30 x 10 19 X

285*

32 (E0L) 7.60 x 10 T

55*

standby [c]

Z 305*

standby [c]

S 45*

standby a.

Effective full power years from plant startup b.

Actual Fluence c.

It is recommended that a standby capsule be relocated in the 165* and 65*

location at the 10 year inspection when the core barrel is removed.

This l

will result in extra capsules that lead the vessel at end of life that l

could be used for plant life extension.

I 90308:1b-091185 7-1

SECTION 8 REFERENCES 1.

Davidson, J. A and Yanichko, S. E., Etal, "Duquesne Light Company Beaver Valley Unit No.1 Reactor Vessel Radiation Surveillance Program" WCAP-8457, October 1974.

2.

ASTM E185-73 " Practice for Surveillance Tests for Nuclear Reactor Vessels" in ASTM Standards, Part 10 (1973), American Society for Testing and Materials, Philadelphia, Pa.,1973.

3.

Yanichko, S. E., Anderson, S. L., and Kaiser, W. T., " Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-9860, January, 1981.

4.

Lott, R. G., Shogan, R. P. and Thomas L. M., " Beaver Valley Unit No.1 Nuclear Pressure Vessel Surveillance Capsule U Test Program," Westinghouse R&D Memo 85-5D25-DLWU-M1, 1985.

5.

Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials, "U.S. Nuclear Regulatory Commission, April 1977.

l f

6.

Soltesz, R. G., Disney, R. K., Jedruch, J., and Zeigler, S. L., " Nuclear l

Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5 - Two-Dimensional Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.

1 i

7.

SAILOR RSIC Data Library Collection "DLC-76, Coupled, Self-shielded, 47

(

Neutron, 20 Gamma ray, P3, Cross Section Library for Light Water Reactors."

l 90308:1b-090585 8-1

8.

Wrights, G. N., and Heinecke, C. C., " Beaver Valley Unit 1 Reactor Vessel Fluence and RT Evaluations," WCAP-10829, August 1985.

PTS 9.

Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -

to be published.

10. ASTM Designation E261-77, Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 76-87, American Society for Testing and Materials, Philadelphia, Pa., 1983.

11. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 88-96, American Society for Testing and Materials, Philadelphia, Pa., 1983.

12. ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron," in ASTM Standards (1983),

Section 12, Nuclear Standards, pp.97-102, American Society for Testing and Materials, Philadelphia, Pa., 1983.

13. ASTM Designation 481-78, " Standard Method of Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 228-235, American Society for Testing and Materials, Philadelphia, Pa., 1983.

14. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel,".in ASTM Standards (1983),

Section 12, Nuclear Standards, pp. 103-107, American Society for Testing and Materials, Philadelphia, Pa., 1983.

90308:1b-090585 8-2

APPENDIX A HEATUP AND C00LDOWN LIMIT CUR /ES F0P NORMAL OPERATION A-1.

INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility ten:perature). The most limiting RT NDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART RT is designated as the higher of either NDT.

NDT the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft lb of impact energy minus 60*F.

RT increases as the material is exposed to fast-neutron radiation. Thus, NDT to find the most limiting RT at any time period in the reactor's life, NDT ART due to the radiation exposure associated with that time period must NpT be added to the original unirradiated RT The extent of the shift in NDT.

RT is enhanced by certain chemical elements (such as copper and NDT phosphorus) present in reactor vessel steels. Design curves which show the ef fect of fluence, copper and phosphorus contents on ART r reactor NDT vessel steels are shown in Figure A-1.

Given the copper and phosphorus contents of the most limiting material, the radiation-induced ART can be estimated from Figure A-1.

Fast neutron HDT fluence (E > 1 MeV) at the vessel inner surface, the 1/4 T (wall thickness),

and 3/4 T (wall thickness) vessel locations are given as a function of full-power service life in Figure A-2.

The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure,that no other component will be limiting with respect to RT NDT*

t l

l l

9094Q:10/083085 A-1

A-2.

FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the Beaver Valley Unit i reactor vessel materials are presented in Table A-1.

The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review I

Plan.

The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the Beaver Valley Unit 1 Vessel Material Surveillance Program.

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, f r the metal temperature at that time. K is btained from the IR reference f racture toughness curve, defined in Appendix G of the ASME Code.E 3 The K curve is given by the equation:

IR K

= 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)]

(A-1)

IR where K is the reference stress intensity factor as a function of the IR metal temperature T and the metal reference nil-ductility temperature Thus, the governing equation for the heatup-cooldown analysis is NDT.

defined in Appendix G of the ASME Code [2] as follows:

CK7g + kit IR I

(A-2) l where IM is the stress intensity factor caused by membrane (pressure) stress K

is the stress intensity factor caused by the thermal gradients It 9094Q:10/083085 A-2

K is a function of temperature relative to the RT of the material gg NDT C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical 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 RTNOT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, kit, f r the reference flaw are computed. From Equation (A-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 Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling 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 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 ecoldown results in a higher value of K at the 1/4 T IR location for finite cooldown rates than for steady-state operation.

4 9094Q:lD/090385 A-3

Furthermore, if canditions exist such that the increase in K exceeds gg 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 temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures 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 j

pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall.

The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the cociant temperature; therefore, the K rWMT IR crack during heatup is lower than the K for the 1/4 T crack during gg steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K

's do not offset each IR other, and the pressure-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 insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of 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 thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of i

9094Q:10/083085 A-4

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 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. Then, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on Figures A-3 and A-4.

Finally, the new 10CFR50[3] rule which addresses the metal temperature of the closure head flange and vessel flange regions is considered.

This 10CFR50 e

e t ma e i R

by t at1 F or 1

rat n

pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Beaver Valley Unit 1). Table A-1 indicates that the limiting RT NDT of 60*F occurs in both the head and vessel flanges of Beaver Valley Unit 1 and the minimum allowable temperature of this region is 180*F at pressures greater than 621 psig.

A-4.

HEATUP AND C00LDOWN LIMIT CURVES i

Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in Section A-3.

The derivation of the limit' curves is presented in the NRC Regulatory Standard Review Plan.E43 9094Q:lD/083085 A-5 m.

Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program. Charpy test specimens from Capsule U indicate that the surveillance weld metal and core region lower shell plate code no.

B6903-1 exhibited shifts in RT of 200*F and 160*F at a fluence of 6.54 x NDT 18 10 nye,2 These shifts are well within the appropriate design curve (Figure A-1) prediction. Therefore, the heatup and cooldown curves in Figures A-3 and A-4 are based on the trend curve in Figure A-1, and these curves are applicable up to 9.5 effective full power years (EFPY). These heatup and cooldown curves are not impacted by the new 10CFR50 rule.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure A-3.

This is in addition to other criteria which must be met before the reactor is made critical.

i The leak test limit curve shown in Figure A-3 represents minimum temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of References 2 and 4.

l Figures A-3 and A-4 define limits for insuring prevention of nonductile l

failure.

l t

9094Q:1D/083085 A-6 L.-,

TABLE A-1 BEAVER VALLEY UNIT NO 1 REACTOR VESSEL TOUGHNESS DATA (UN!RRADIATED)

T RT Component Heat No.

Code No, Material Type Ch h

hI NDT NDT Upper Shelf Energy (Ft-Ib)

I

(*F)

WD NMWD Closure Head Dome C6213-1B B6610 A5338 CL. 1

.15

.010

.54

-40 0*

121 Closure Head Seg.

A5518-2 B6611 A5338 CL. 1

.14

.015

.60

-20

-20*

131 Closure Head Flange ZV3758 B6601-1 A508 CL. 2

.08

.007

.65 60*

60*

>100 Vessel Flange ZV3661 B6603 A508 CL. 2

.12

.010

.71 60*

60*

166 Inlet Nozzle 9-5443 B6608-1 A508 CL. 2

.10

.008

.84 60*

60*

82.5 Inlet Nozzle 9-5460 B6608-2 A508 CL. 2

.10

.010

.83 60*

60*

94 Inlet Nozzle 9-5712 B6608-3 A508 CL. 2

.08

.007

.80 60*

60*

97 l

Outlet Nozzle 9-5415 B6603-1 A508 CL. 2

.008

.79 60*

60*

97 30 Outlet Nozzle 9-5415 B6603-2 A508 CL. 2

.007

.79 60*

60*

112.5 Outlet Nozzle 9-5444 B6603-3 A508 CL. 2

.09

.007

.78 60*

60*

103 Upper Shell 123V339 B6604 A508 CL. 2

.010

.67 40 40*

155 Inter. Shell C4381-2 B6607-2 A5338 CL. 1

.14

.015

.62

-10 73 123 82.5 Inter. SheII C4381-1 B6607-1 A533B CL. 1

.14

.015

.62

-10 43 128.5 90 Lower Shell C6317-1 B6903-1 A5338 CL. 1

.20

.010

.54

-50 27 134 80 Lower Shell C6293-2 B7203-2 A533B CL. 1

.14

.015

.57

-20 20 129.5 83.5 Trans. Ring 123V223 B6602 A508 CL. 2 30 30*

143 Bottom Head Seg.

C4423-3 B6618 A533B CL. 1

.13

.008

.55

-30

-29*

124 Bottom Head Dome C4482-1 B6619 A5338 CL. 1

.13

.015

.55

-50

-33*

125.5 Inter. Shell Long. Weld SAW

.29

.015

.63 0*

0*

Inter. to Lower Shell Girth Weld SAW

.31 015

.07 0*

0*

Lower Shell Long.

Weld SAW

.33

.013

.61 0*

0*

Weld HAZ

-40

-40 136.5

Estimated per NUREG-0800 NRC Standard Review Plan Branch Technical Position MTLB 5-2 MWD - Major Working Direction NMWO - Nonnal to' Major Working Direction

~

m uu m m u u u u.anu m..<.

s_-w -

t s u uu un numm u t uusun r-r 400 A = [40 + 1000 (% Cu - 0.08) + 5000 (% P - 0.008)] [f/1019] 1/2 fr i

j

g j j j g 4

ji

q n,.

g

]

l I

l i

! !E55 i

_ f

~

O! 5EOPP,,[

}

\\ m zzis w f55 L

W p

j ! EEE=z-

" pf,,p$a G,Ej_;, __-_

f

,s 2

d d'

d,

,8

~,8 l j 8

t i

r o

~,

y ;

8 d'

g g

V,,d Ik

,,,I d

__t s

o i

v f

?

d 3?

,,'d' v

j d

iod I V3 i

d rf I

iil

_..i,2I7-b *'T

I /..,a'"'--'

u-1 h

i ai # !

s*".'

P k

i O

,d r

'i 1

M til 34;

-i s #

l,,, * ' ' i

',,I 3

iP '#

1 M Ii!s'#I :

1:

i P s*#

I JIf[,

f' O

1%'#I

- ii l J,,W'1 1

"I HIl $d4 fiW'#"

J,,,

l

,v*

I

!h 55 55

,l"# 1

.1 a

b 363M0#

/'iIM'iIL,d

x

,c',

l0 2 2

- -MP N !

so s.

0.35 0.30

0.25 0.20 % Cul z:0.15% Cu 0.10% Cu T?

-~-

' p J J:] p g;

q:

7

.KP = 0.6551 d

~~

~

2 Z

~

r:::

llN LOWER EIMIT

~

p'

{ {

Wi k i

-i n'j' i l j', i ill b 3 . % Cu = 0.08 ~ i Ollilj Ei fE E:1 3 ~ L I. ~ =" % P = 0.00_8_ ?.-3:2: !' l! EiM G 1; r -lER 7_. ; d ! :.r r_ r r-- u r 2X1017 18 4 6 8 10 2 4 6 8 10 2 4 6 18 FLUENCE (n/cm, E>1MeV) Figure A-1 Effect of Fluence, Copper and Phosphorus Contents on ART for Reactor Vessel Steels per Regulatory Guide 1.99, Revision 1 NDT A WELD METAL 4 SHELL PLATE B6903-1

I , i ci I !!l lIll l lllll ECAF T 7 R 4 4 U / / S 1 3 53 ) Y P ~ F E ( EF ~ 0 I 3 L EC ~ I VR ES 5 g 2 gy 0 9 ~ L ) L Y U P F F 7 s

E 0

(

N g O I g T g C N g UF t 5c A 7 / 1 i S V A R E ) S Ve ,l 0 / E 1 ( / E C N E U L F N 5 O / R / / T s U / / E [ / N / T rt / S A F 2 8

0 0

9 1 I 0 9 1 I 1 9 y 0 ~ 1 1 p F m%o)w Q(),Z 3a k 2 5aW* g g E y G I F 1 y* l

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: WELD METAL COPPER CONTENT: 0.31 WT% PHOSPHORUS CONTENT: 0.015 WT% RT INITIAL: 0*F NDT RT AFTER 9.5 EFPY 1/4T, 274*F NDT 3/4T,137'F CURVE APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 9.5 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 i'l!'I i l i ' l'l;;I I i ll l i j i il l l: i !.: i l l l t !'!'IIi 'l I-l, i i iji l LEAK -.-q' l I '!lt l YN-i i T n i /_i!til;f ii iiii 't I l 7 il i j'! I T'W i 'li l ' '/! l l I /f / II!/ l i ,I l !I'll l I / i i/;i i-

i.

ij ij l if J'll fit 2000 I /i I /, ' ! i, ; I E' i t ,il j /I I /i I ,!'.'il i / / / / ~ .;i:li, i j j f f' ,i:: ;. !i i 'i i [I f i I'i j UNACCEPTABLE /- ./ i ;'.j f OPERATING REGION e / l O_ ACCEPTABLE i' il l1 t d / i OPERATING REGION i! ! l-ii i ii /l i E i 4 i 1000 I i i / i j li.i j HEATUP - p / RATES UP TO lI! i . Y 602F/HR ';I, j l l'y j ~ CRITICALITY LIMIT 4* BASED ON INSERVICE l1'I i ~ i HYDROSTATIC TEST

i ;j l jj j

.. l p ,-- TEMPERATURE (414*F) I 1 .-++q FOR THE SERVICE ! t PERIOD UP TO 9.5 EFP\\ i _, ~.. 7-..., 0 0 100 200 300 400 500 INDICATED TEMPERATURE (*F) FIGURE A-3 BEAVER VALLEY UNIT NO. 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 9.5 EFPY A-10

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: WELD METAL COPPER CONTENT: 0.31 WT% PHOSPHORUS CONTENT: 0.015 WT% RT INITIAL: 0*F RTh AFTER 9.5 EFPY: 1/4T, 274*F 3/4T,137'F CURVE APPLICABLE FOR C00LDOWN RATES UP TO 100*F/HR PERIOD UP TO 9.5 EFPY AND CONTAINS MARGINS OF 10*F AND 6 POSSIBLE INSTRUMENT ERRORS 3000 f ,i.> gg j 4 i, I I I l I i I il l'l I l t il! I. ~. l l l I l i! l l t i i l[ i I ! l/!!' g ,ii i i it!!' i' 4 .i j ti i /! i ei D ii g 2000 j [:li!!ij 8 .j gi i. g

t i

/ E }l e i f !I' e ii I i / l

l ij/

l til as / IIi 'I z ~ i UNACCEPTABLE Y ii!i'

i; OPERATING REGION

/

ijj, i

it l l / t 1000 t j lii j j , jj, l MI C00LDOWN ii ACCEPTABLE RATES *F/HR J7/ OPERATING REGION i, f/ jg r e jj 0~ i i 20 -- rgg/ j;i;i 40 s-- f /

i j

60 i !,p 100 i iF 0 j -i-U 100 200 300 400 500 INDICATED TEMPERATURE (*F) FIGURE A-4 BEAVER VALLEY UNIT NO. 1 REACTOR COOLANT SYSTEM COO LIMITATIONS APPLICABLE FOR THE FIRST 9.5 EFPY A-ll

APPENDIX A REFERENCES 1. " Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants LWR Edition, NUREG-0800,1981. 2. ASME Boiler and Pressure Vessel Code Section III, Division 1 - Appendices, " Rules for Construction of Nuclear Vessels," Appendix G, " Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York, 1983. 3. Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., . Amended May 17, 1983 (48 Federal Register 24010). 4. " Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants LWR Edition, NUREG-0800, 1981. 9094Q:lD/083085 A-12}}

ML20138S008

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