ML17338B115

Reactor Vessel Matl Surveillance Program for Capsule s, May 1979,final Rept

ML17338B115
Person / Time
Site:	Turkey Point
Issue date:	05/31/1979
From:	Lindholm U, Norris E SOUTHWEST RESEARCH INSTITUTE
To:
Shared Package
ML17338B114	List: L-79-268, Forwards Reactor Vessel Matl Surveillance Program for Capsule s, May 1979 Final Rept.Prepared by Southwest Research Inst ML17338B115
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NUDOCS 7909260426
Download: ML17338B115 (91)

Text

SOU I I-I%VEST RESEARCH INSTITUTE 6220 CULEBRA ROAD

+

POST OFFICE DRAWER 28510

+

SAN ANTONIO>> TEXAS 78284 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR CAPSULE S TURKEY POINT UNIT NO.

3 CAPSULE S TURKEY POINT UNIT NO.

4 FINAL REPORT SwRI Prospect No. 02-5131 SwRI Project No. 02-5380 to Florida Power 6 Light Company P. 0. Box 3100 Miami, Florida 33101

>gLimt To e>mWt <stijl iiii'-5 MoilMI May 1979 APPROVED:

Prepared by:

E. B. Norris U. S. Lindholm, Director Department of 'Materials Sciences

>90926 0 q~g

>> g SAN ANTONIO, 'HOUSTON, TEXAS, AND WASHINGTON, D.C

TABLE OF CONTENTS I.

SUMMARY

OF RESULTS AND CONCLUSIONS II.

BACKGROUND XIII'ESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM

~Pa e A.

B.

C.

Introduction Surveillance Capsule Materials Unit No. 3, Capsule S

Surveillance Capsule Materials Unit No. 4, Capsule S

IV.

TESTING OF SURVEILLANCE SPECIMENS 17 A.

B.

C.

D.

E.

Introduction Opening of Surveillance Specimen Capsules and Recovery of Specimens Neutron Dosimetry Mechanical Property Tests Check Chemical Analyses for Copper 17 18 18 22 22 V.

CAPSULE S TEST RESULTS - TURKEY POINT UNIT NO.

3 23 A.

B.

C.

D.

Neutron Dosimetry Thermal Monitors Mechanical Property Test Results Check Chemical Analyses for Copper 23 27 27 29 VI.

CAPSULE S TEST RESULTS TURKEY POINT UNIT NO.

4 33 A..

B.

C.

D.

Neutron Dosimetry Thermal Monitors Mechanical Property Test Results Check Chemical Analyses for Copper 33 37 37 37 VII.

ANALYSIS OF RESULTS A.

B.

C.

D.

E.

F.

Introduction Pressure Vessel Fast Neutron Exposure Vessel Material Notch Toughness Adjusted Reference Temperature Heatup and Cooldown Limits for Normal Operation Capsule Removal Schedule 43 44 45 49 49 52 VIXI. REFERENCES APPENDIX A Charpy V-Notch and Tensile Test Data APPENDIX B Charpy V-Notch and Tensile Test Data 53 55

LIST OF TABLES Table

~Pa e

FPL Turkey Point Unit No.

3 Reactor Vessel Surveil-lance Materials 12 Specimen Identification and Location in Turkey Point Vessel Material Surveillance Capsules 13 FPL Turkey Point Unit No.

4 Reactor Vessel Surveil-lance Materials 15 IV Summary of Neutron Dosimetry Results, Capsule S,

Turkey Point Unit"No. 3 24 Summary of Plant Operations, Turkey Point Unit No.

3 25 VI Fast Neutron, Spectrum and Foil Activation Cross Sections for Capsule S, Turkey Point Unit No.

3 26 VII Notch Toughness Properties of Capsule S Specimens, Turkey Point Unit No.

3 28 VIII Summary of Neutron Dosimetry Results, Capsule S,

Turkey Point Unit No.

4 34 IX Summary of Plant Operations, Turkey Point Unit No.

4 35 X

Fast Neutron Spectrum and Foil Activation Cross Sections for Capsule S, Turkey Point Unit No.

4 36 XI Notch Toughness Properties of Capsule S Specimens, Turkey Point Unit No.

4 38 XII Projected Maximum Pressure Vessel Exposures, Turkey Point Unit No.

3 46 XIII Projected Maximum Pressure Vessel Exposures, Turkey Point Unit No.,4 47

LIST OF FIGURES

~F1 ure

~Pa e Arrangement of Surveillance Capsules in the Turkey Point Units 3 and 4 Pressure Vessels Vessel Material Surveillance Specimens 10 Arrangement of Specimens in Capsule S, Turkey Point Unit Nos.

3 and 4

Location of Capsule S within One-Eighth Segment Used as Model for Discrete Ordinate 'Transport Calculations 20 Tensile Properties of Forging 123S266VA-l, Turkey Point Unit No.

3 30 Tensile Properties of Forging 123P461VA-l, Turkey Point Unit No.

3 31 Tensile Properties of Forging 122S180VA-1, Turkey Point Unit No.

4 Tensile Properties of Forging 123P481VA-l, Turkey Point Unit No.

4 39 40 Comparison of Deere'ase in Shelf Energies of Turkey Point Unit Nos.

3 and 4'Vessel Forging Materials and Surveillance Reference Steels.'to Regulatory Guide 1.99 Trend Curves 48 10 Comparison of Increase in Reference Temperatures of Turkey Point Unit Nos.

3 and 4 Vessel Forging Mate-rials and Surveillance Reference Steels to Westing-house Trend Curves

~

~l Comparison of Increase in Reference Temperatures of Turkey Point Unit Nos.

3 and 4 Vessel Forging Mate-rials and Surveillance Reference Steels to Regulatory Guide 1.99 Trend Curves 50

I.

SUMMARY

OF RESULTS AND CONCLUSIONS The analyses of the reactor vessel material surveillance

program, capsules (coded "S") removed from the Florida Power and Light Company (FPL) Turkey Point Units 3 and 4 nuclear reactor vessels during the 1977-1978 refuelling.outages led to the following conclusions:

K (1)

The intermediate and low'er shell forging materials utilized in the reactor pressure vessels of Units 3 and 4 exhibited a low sensi-tivity to radiation embrittlement.

The shelf energy reductions and the transition tempe'rature shifts were equal to or below those predicted by the minimum response curves given in Regulatory Guide 1.99.

(2),

Based on the surveillance program results to date and trend curves for low-copper materials, the irradiated properties of the inter-mediate'nd lower shell forging materials utilized in the pressure vessels for Units 3 and '4,will be adequate to meet the. current requirements of

10CFR50, Appendix G, through the 40-year design lifetime.

(3)

Capsule S from Unit No." 3 received a fast neutron fluence of 1.41 x 101

- cm E

> 1 MeV.

Bas'ed on a calculated lead factor of 1.76, the peak fast neutron exposure of the Unit No.

3 reactor vessel is pro-jected to 'be 2.3 x 10 cm after 10 Effective Full Power Years (EFPY) of operation.

The peak end-of-life (32 EFPY) fast neutron fluence is predicted to be 7,.4 x 1019 cm

,, in good agreement with the value of

'6.65 x 10, cm projected earlier from the analysis of Capsule T.-

(4)

Capsule S from Unit No.

4 received a fast neutron fluence of 1.25 x 10 cm E

> 1 MeV.

Based on a calculated lead factor of 1.76, the'peak fast neutron exposure of the Unit No.

4 reactor vessel is,

projected to be 2.1 x 1019 cm

'after 10 EFPY of operation.

The peak end-of-life fluence (32 EFPY) fast neutron gluence is predicted to be 6.7 x 10 cm

, in excellent agreement with the value of 6.62 x 1019 cm pro-

)ected.earlier from the analysis of Capsule T.

(5)

Capsule V from each reactor vessel should be removed and tested after approximately 10 calendar years (-

7 EFPY). of operation.

The data ob-tained should provide the information needed to revise the heatup and cool-down limitations for operation beyond 10 EFPY.

II.

BACKGROUND The allowable loadings on nuclear pressure vessels are determined by applying the rules in Appendix G, "Fracture Toughness Requirements,"

of 10CFR50.

~

~

In the case of pressure-retaining components. made of ferritic materials, the allowable loadings depend on the reference stress intensity factor (KIR) curve indexed to the reference nil ductility temperature (RTNDT) presented in Appendix G, "Protection Against Non-ductile Failure," of Section III of the ASME Code.< )

Further, the materials in the beltline region of the reactor vessel must be monitored for radiation-induced changes in"RTNDT per the requirements of Appendix H, "Reactor Vessel Material Surveillance Program Requirements,"

of 10CFR50.

The RTNDT is defined in paragraph NB-2331 of Section III of the ASME Code as the highest of the following temperatures:

Drop-weight Nil Ductility Temperature (DW-NDT) per ASTM E 208;(

~

60 deg F below the 50 ft-lb Charpy V-notch (C ) temperature;'

60 deg F below the 35 mil Cv temperature.

The RTNDT must be established for all materials, including weld metal and heat affected zone (HAZ) material as well as base plates and forgings, which'omprise the reactor coolant pressure boundary.

It is well established that ferritic materials undergo an increase in strength and hardness and a decrease in ductility and toughness when exposed to neutron fluences in excess of 10 neutrons per cm (E > 1 MeV).( )

Also, it has been established that tramp elements, particularly copper and phospho-rous, affect the radiation embrittlement response of ferritic materials.(

Superscript numbers refer to references at the end of the text.

There is some disagreement concerning the relationship between increase in RTNDT and copper content.

For example, Regulatory Guide 1.99(7) proposes an adjustment to RTNDT proportional to the square root of the neutron fl'u-ence.

Westinghouse Electric Corporation, in their comments on Regulatory Guide 1.99( ), feels that the proposed relationship overestimates the shift at high fluences (above 1019) and underestimates the shift at low fluences (below 10

).

On the other hand, Combustion Engineering, in their comments on Regulatory Guide 1.99( ), suggests that the proposed relationship is overly conservative at fluences below 1019 neutrons per cm (E > 1 HeV).

There is also disagreement concerning the prediction of Cv upper shelf response to exposure to neutron irradiation.("

) It is important to re-solve these questions because the analysis of"reactor vessel material sur-veillance program data requires that estimations be made of shifts in RTNDT and Cv upper shelf energy at fluences other than that received by the sur-veillance capsule.

In general, the only ferritic pressure boundary materials in a nu-clear plant which are expected to receive a fluence sufficient to affect RTNDT are those materials which're located in the core beltline region of the reactor pressure vessel.

Therefore, reactor vessel material surveil-lance programs include specimens machined from the plate or forging, weld

metal, and heat affected zone (HAZ) materials which are located in such a

region of high neutron flux density.

ASTM E 185(10) describes the current recommended practice for monitoring and evaluating'the radiation-induced changes occurring in the mechanical properties of pressure vessel beltline materials.

Westinghouse has provided such a surveillance program for the two-unit Turkey Point nuclear power plant.

The encapsulated Cv specimens are attached to the O.D. surface of each, thermal shield where, the fast neutron flux density is approximately twice that at the adjacent vessel wall surface.

Therefore, the increases (shifts) in transition temperatures. of the materials in the pressure vessel are generally less than the corresponding shifts observed in the surveillance specimens.

'owever," because of.azimuthal variations in neutron flux density, some capsule fluences may be less than the maxi-mum vessel fluence in a corresponding exposure period.

For example, the first capsules removed from Turkey Point Units 3 and. 4 were reported to lead the maximum exposure point on the vessel I.D. by a factor of 2.48 while other capsules scheduled 'to be removed later are.calculated to re-ceive less than half of the fluence accumulated at the point of maximum vessel exposure.

The capsules also contain several, dosimeter materials for experimentally determining the average neutron flux density"at each capsule location during the. exposure period.

The Turkey Point, Units 3 and 4 material surveillance capsules also include tensile specimens as,,recommended by ASTM E 1'85..-,At the. present time, irradiated tensile properties are. used to indicate that the materi-als tested continue to meet the requirements of the appropriate material specification and to assist in judging the credibility of the C

data.

In addition, the material surveillance capsules contain'edge opening loading (WOL) fracture mechanics specimens.

Current technology limits the testing of these specimens at temperatures well below the-minimum service temperature to obtain, valid fracture mechanics data per ASTM E 399(

), "Standard Method of Test, for Plane-Strain Fracture Toughness of

I Metallic Materials."

However, recent work reported by Mager and Mitt(>>)

and Loss~1

~ may lead to methods for evaluating high toughness materials with small fracture mechanics specimens.

Currently, the NRC suggests stor-ing these specimens until an acceptable testing'rocedure has been defined.

Capsule T was removed from Turkey Point Unit No.

3 during the 1974 refuelling outage, and the results have been reported;

~ "~

The weld metal was found to be more sensitive to neutron radiation embrittlement than the forging or HAZ material contained in, the capsule, and it was concluded that the weld metal is the most 1imiting beltline material'in 'the Unit No.

3 vessel.

Capsule T was removed from Turkey ',Point Unit No.

4 during the 1975 refuelling outage, and those results have also been reported; ~15)

The re-sults indicated that the vessel beltline weld metal was more sensitive to radiation embrittlement than the lower" shell forging and HAZ materials in-eluded in the capsule.

This report describes, the results'btained from testing the contents of Capsule S removed from Unit No.

3 and Capsule S removed from Unit No. 4.

These data are analyzed.to estimate the radiation-induced chang'es in the mechanical properties of the respective pressure vessel forging materials at the time of the 1977-78 refuelling outage as well as predicting the changes expected to occur at selected times in the future operation of the-Turkey Point nuclear power plant.

III.

DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM A.

Introduction The Turkey Point Units 3 and 4 material surveillance programs are described in detail in WCAP 7656(

) and WCAP 7660(. '), respectively.

Eight materials surveillance capsules (five Type I and three Type II) were placed in each reactor vessel between the therma'1 'shield and the vessel wall prior to startup, see Figure 1.

The vertical'center of each capsule is opposite the vertical center of the core. 'he reported ratios of neutron flux density at the capsule location, to,the'aximum flux den-sity on the vessel I.D. are given by the factors in parentheses follow-ing the capsule identification letter in Figure 1.

The Type I capsules each contain Charpy V-notch, tensile, and WOL specimens

'machined from the two vessel forgings located at the core beltline plus Charpy V-notch spec-imens machined from one of two reference heats of ferritic steel utilized in the Westinghouse surveillance programs.

The Type 'I'I capsules. include specimens machined from.weld metal and HAZ.material which were intended to be representative of those materials in.the core beltline region of each vessel as well as one heat of forging material, and the reference steels.

All test specimens were. machined from the materials at the quarter-thickness (1/4T) location.(

)

The base metal tensile and Cv specimens were oriented with their long axis parallel to the principal working di-rection; the Cv notches were perpendicular to the major forging surfaces.

The WOL specimens were machined with the simulated crack 'perpendicular to the principal working direction and to the forging surfaces.

All mechanical

Z (0. 34) 270' (2.48)

S (1.61)

V (0. 79) 40'0' p ~

180'

~

3O ~

Y (0.49) 9P ~

Reactor Vessel Thermal Shield U (0.49)

W (0.34)

X (0.34)

Core Barrel Note:

Numbers in parentheses indicate lead factors for the vessel L D.

FIGURE l.

ARRANGEMENT OF SURVEILLANCE CAPSULES IN THE TURKEY POINT UNITS 3 and 4 PRESSURE VESSELS

test specimens, see Figure 2, were taken at least one plate thickness from the quenched edges of the forging material.

Capsule S (a Type I capsule) was removed from each of the Turkey Point nuclear reactor vessels during the 1977-1978 refuelling outages.

The arrangement of specimens within these capsules is shown in Figure 3.

Additional details concerning the contents of the capsules from each unit are discussed below.

B.

Surveillance Ca sule Materials

, Unit. No.

3 Ca sule S

Babcock and Wilcox" Company supplied.prolongations from two 7-7/8-in.

thick forged rings (Heat 123P461VA-1 and 123S266VA-.l produced by Bethlehem Steel Company) of SA 508, Class 2 steel" used for the FPL Unit No.

3 reac-tor pressure vessel intermediate and lower, shell course, respectively, and a weldment which joined sections of the 'two forgings.'orrelation monitor material, supplied by U.S. Steel Corporation through Subcommittee II of ASTM Committee E 10, was obtained from a 6-in. "thick A 302 Grade B plate which was melted using fine-grain practice and a transverse-to-longitudinal rolling ratio of one to one.

The chemistries and heat treatments of the vessel surveillance materials contained in Capsule S from Unit 'No.

3 are summarized in Table I.

The capsule contained 28 Charpy V-notch specimens (10 from each of the two vessel forging materials plus 8 from the reference steel plate);

4 ten-sile specimens (2 from each forging material),;

and 6 WOL specimens (3 from each forging heat).

The specimen numbering system and location within Cap-sule S for Unit No.

3 is shown in Table II.

The capsule also was reported to, contain the following dosimeters for determining the neutron flux density:(

9

464 444

~IIR

.009~~M (a) 'Charpy V-notch Impact Specimen

. 256

.246 6005

.99Y 245 l6 Qese l4hglh

.256 395 4.210 Z50 R I.26 Ag l.495 II

.630

.I98

.I9

.790

.786

.395 o.

.3 SECTION A A (b) Tensile Specimen

.375 0.

380 I.45III IS30 I.f25

~765

.745

.499

.43 F

.0473

.0463 0662 o

.0667

~0662

.0667 7 r I.005

.999 SS /%9 I 1

~

g I

~

.SOI A99 (c)

%'edge Opening Loading Specimen FIGURE 2.

VESSEL MATERIAL SURVEILLANCE SPECIMENS, 10

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Aaas aahaaa Aaaaalaaff Iaaaa IJ a~

~r yaaaeaf

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~

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~alt IICIA&II I 1AAC IafIAt 18l ~. AT%fLtoafg AAI

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f aafta IAA I all fat k

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<<ICE lir aafAAAa acaaa tafAIaaf aata111 fIIOT AIAaaIIIA fA

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elfIAalf FIGURE 3.

ARRANGEMENT OF SPECIMENS IN CAPSULE S, TURKEY POINT UNIT NOS ~

3 A.lD 4

TABLE I FPL TURKEY POINT UNIT NO.

3 REACTOR VESSEL SURVEILLANCE MATERIALS(14, 16)

Heat Treatment Histor Intermediate and Lower Shell Forgings 1550 F 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> water-quenched 1210 F 18-hours air-cooled 1125 F 10-1/2 hours furnace-cooled to 600 F Weldment Correlation Monitor 1125 F 1650 F 1200 F 10-1/4 hours - furnace-cooled to 600 F

(

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> water-quenched to 300 F 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> furnace-cooled Chemical Com osition (wt-%

Element C

Mn P

S Si Ni Cr V

Mo Co Cu Sn Zn Al N2Ti Sb As B

Zr Lower Shell 1238266VA-1

0. 19/0. 21 0.61/0.62 0.010 0.008 0.20/0.19 0.68/0.66

0. 38

0. 02 0.58/0.59 0.015/0.016 0.079 0.008

0. 001 0.005 0.003

0. 001*

0.001*

0.005*

0.003*

0.001*

Intermediate Shell

,123P461VA~1

0. 20 0.64/0.64 0.010

0. 010

0. 26

0. 70

0. 40/0. 39 0.02

0. 62 0.011/0.010 0.058 0.010 0.001 0.005 0.003 0.001*

0;001*

0.005*

0.003*

0.001~

Correlation

- Monitor 0;24 1.34 0.011 0.023 0.23 0.18 0.11 0.51

0. 20

• Not detected.

The number indicates the minimum limit of detection.

12

TABLE II SPECIMEN IDENTIFICATION AND LOCATION IN TURKEY POINT VESSEL MATERIAL SURVEIL'LANCE CAPSULES(16717)

Specimen 5P~

Charpy V

Tensile WOL WOL Charpy V WOL WOL Charpy V

Charpy V

WOL WOL oto I

Capsule S

S97 Slo P9, P10 Pl, P2 P3 P2 S7, S8 P7, P8 Pl 3

S3 R7, R8 R5, R6 R3, R4 Rl, R2 S2 Sl S ecimen Code Unit 3 S Forging 123S266VA-1 P Forging 123P461VA-1 R ASTM Correlation Monitors S ecimen Code Unit 4 S Forging 122S180VA-1 P Forging 123P481VA-1 R ASTM Correlation Monitors S ecimen Orientation VEEE EL Charpy V Tensile Charpy V 8o Charpy V S5, S6 P57 P6 Sl, S2 S3, S4 P3, P4 Sl, S2 Pl, P2 S

S P

P CCEE

Tar et Element Form

(~ant~it Copper Nickel Cobalt (in aluminum)

Cobalt (in aluminum)

Bare wire Bare wire Bare wire Cd shielded wire 2

1 3

3 In addition, slices were taken from ten Cv specimens to serve as iron dosimeters.

Three eutectic alloy thermal monitors had been inserted in holes in the steel spacers in the capsule.

Two (located top and bottom) were 2.5%

Ag and 97.5%

Pb with a melting point of 579. F.

The 'third (located at the cneter of the capsule) was 1.75% Ag; 0.75%, Sn, and 97.5% Pb having a mel~t-ing point of 590 F.

C.

Surveillance Ca sule Materials Unit. Zo.

4

" Ca sule S

Babcock and %7ilcox Company supplied prolongations

-from.two 7-7/8 in.

thick forged rings (Heat 123P481VA-1 and 122S180VA-1 produced by 'Bethlehem Steel Company) of SA 508, Class 2 steel:,used for the FPL Unit No'. '4 reactor pressure vessel intermediate and lower shell course, respectively,.and a

weldment which joined sections of the two forgings.

Correlation monitor material was supplied by the Oak Ridge National Laboratory from plate ma-terial used in the AEC-sponsored Heavy..Section Steel Technology (HSST)

Program. 'his material was obtained from a Lukens Steel Company 12-in.

thick A533 Grade B, Class 1 plate (HSST Plate

02) which has been provided to Subcommittee II of ASTM Committee E10 on Radioisotopes.and Radiation Effects to serve as correlation monitor material in reactor vessel sur-veillance programs.

The chemistries and heat treatments of the Vessel surveillance materials contained in Capsule S from Unit No.

4 are sum-marized in Table III.

14

~

TABLE III FPL TURKEY POINT UNIT NO.

4" REACTOR VESSEL'URVEILLANCE MATERIALS(

Heat Treatment Histor Lower Shell (Heat 122S180VA-1) 1550 F 10-1'/4 hours water-quenched 1210 F 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> air-cooled 1125 F 10-1/2. hours furnace-cooled to 600 F Correlation Monitor 1675 1600 1225 1150

+25F-

+25F-

+ 25'F-

+25F-4 hours air-cooled 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> water-quenched 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> furnace-'co'oled 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> furnace-.,cooled'o 600 F Chemical Co osition (wt-%

Element.

Lower Shell 1228180VA-1 Intermediate Shell

'23P481VA-'1 Correlation Monitor C

Mn P

S Si Ni Cr V

Mo Co Cu Sn Zn Al N2 Ti Pb As B

Zr W

Nb Ta 0.21 0.67 0.011 0.009 0.23 0.70 0.31'.

001 0.56 0.015 0.056 0.008 0.001*

0.008 0.002 0.001*

0.001%

0.005 0.003*

0.004 0.001*

0.001 0.002 0.22.-

0. 67..

0.010 0.009 0.20 0.71 0.33 0.002'.56 0.017 0.054 0.008 0.001*

0;008 0.001

'.001*

0.001*

0.004 0.003*

0.005 0'. 001K:

0;002 0.003

0. 22 1.48 0.012 0.018 0.25 0.68 0.52 0.14

• Not detected.

The number indicates the minimum limit of detection.

15

The capsule contained 28 Charpy V-notch specimens (10 from each of the two vessel forging materials plus' from the reference steel plate);

4 ten-sile specimens (2 from each forging material);

and 6 'VOL specimens (3 from each forging heat).=

The specimen'-numbering system 'and location within Cap-sule S for Unit No.

4 is given in Table II.

The neutron flux wires and thermal monitors contained in the capsule were reported to be the same type and"location as those contained in Cap-sule S from Unit No. 3..(

)

In addi'tion, slices were tak'en from six Cv specimens to serve as iron dosimeters.

IV.

TESTING OF SURVEILLANCE SPECIMENS A.

Introduction The capsule

shipment, capsule

opening, specimen testing, and report-ing of results were carried out under Quality Assurance Plans prepared by Southwest Research Institute (SwRI),and approved by Florida Power and Light t

Company (FP&L).

These plans are on file.at SwRI.

Applicable SwRI Nuclear Project Operating Procedures which were,. called out,in'the Turkey Point Unit No.

3 project plan include:

~

XI-MS-1, "Determination of Specific'Activity of Neutron Radiation Detector Specimen" XI-MS-3, "Conducting Tension Tests on Metallic'aterials"

~

XI-MS-4, "Charpy Impact Tests on Metallic Materials"

~

XIII-MS-1, "Opening Radiation Surveillance Capsules and Handling and Storing Specimens"

~

XIII-MS-102, "Shipment'f Westinghouse PWR Vessel Material Surveillance Capsule" 4

The applicable SwRI Nuclear Project Operating Procedures which were called out in the Turkey Point Unit No.

4 project plan include:

~

XI-MS-101, "Determination of Specific Activity and Analysis of Neutron Radiation Detector Specimen"

'I XI-MS-103, "Conducting Tension Tests on Metallic Specimens" XI-MS-104, "Conducting Impact 'Tests on Metallic Specimens"

~

XIII-MS-103, "Opening 'Radiation Surveillance Capsules and Handling and Storing Specimens"

~

XIII-MS-104, "Shipment-of Westinghouse PWR Vessel Material Surveillance Capsule Using SwRI Cask and Equipment" Copies of the above documents

.-are on file at SwRI.

17

B.

0 enin of Surveillance S ecimen Ca sules and.Recover of S ecimens The capsule shell had been fabricated by making two long seam welds to goin two half-shells together.

The capsule ends were sawed off, then the long seam welds were milled away using a vertical milling machine.

The top half of the capsule shell was removed, and the specimens and spacer blocks were carefully, retrieved and placed in an indexed recep-tacle so that capsule location was identifiable.,

After the disassembly had been completed, the specimens were care-fully checked for identification and location as listed in WCAP 7656(1 (16) and WCAP 7660.(

)

No discrepancies were found.

The thermal monitors and dosimeter wires were removed from holes in the spacer blocks and placed in indexed receptacles.

C.

Neutron Dosimetr The specific activities of the dosimeters were determined, with an NDC 2200 multichannel analyzer and an NaI(Th) 3 in. x 3 in. scintillation crystal.

The calibration of the equipment was accomplished'with appropri-ate standards and an interlaboratory. cross check with two independent count-ing laboratories on Co-,

"Mn-and 5 Co-containing dosimeter wires.

All activities were corrected to the time-of-removal (TOR)'at reactor shutdown.

Infinitely dilute saturated activities '(ASAT) were calculated for each of the dosimeters because ASAT is directly related to the integral of the energy-dependent microscopic activation cross section and'the neutron flux density.

The relationship between ATOR and ASAT is given by:

ATOR m-n

-ATm Atm E

(1-e'(e

)

ASAT m=

where:

decay constant for the activation product, day 1; Tm

=

equivalent operating days at 2200 HMt for operating period m; tm

=

decay time after operating period m, days.

The primary result desired from the dosimeter analysis is the total fast neutron fluence

(> 1 NeV) which the surveillance specimens received.

The average flux density at full power is given by:

A>>SAT Vi (2) where:

A SAT energy-dependent neutron flux density, cm 2 sec 1; saturated activity of the ith activation product at full power, dps/target nculeus; oi

=

spectrum-averaged'cross section for the it tar-get nucleus, cm 2

The neutron flux energy and spatial distribution were calculated for the Turkey Point pressure vessels with the DOT 3.5 two-dimensional discrete ordinates transport

code, a 22-group neutron cross section library, a Pl expansion of the scattering matrix and an Sg order of angular quadrature.

Using a one-eighth segment as being representative'because of the symmetry involved, the core, core barrel, thermal shield, specimen

capsules, and pressure vessel were described in R-8 coordinates for this computation, see Figure 4.

The calculations which produced the lead factors given in Figure 1

'ere based on the core power distribution given in the Turkey Point Units 3 and 4 FSAR.

The DOT 3.5 calculations'ade in support of the Capsule S

analyses, utilizing core power distributions based on plant records sup-plied by FPL, indicated that the Capsule S lead factors were higher than

Reactor Vessel Thermal Shield Core Barrel 10'IGURE 4.

LOCATION OF CAPSULE S WITHIN OiK-EIGHTH SEGMENT USED AS MODEL FOR DISCRETE ORDINATE TRANSPORT CALCULATIONS

originally calculated.

This is primarily caused by a reduction in the maxi-mum calculated flux incident on the reactor vessel, resulting from the al-I 1

tered core power'istribution, as shown below:

Source of Core Power Distribution Calculated Flux cm

'sec E > 1 Mev Lead Factor FSAR, Units 3 and 4

Plant Records, Unit 3 Plant Records, Unit 4 1.13 x 1011 1.16 x 1011 1.18 x 1011 7.13 x 1010 5.45 x 1010 5.58 'x 1010 1.6 2.1 2.1 An additional contribution to the difference in these lead factors is the flux perturbation caused by the presence of, a surveillance capsule.

In the calculation based on the FSAR 'power distribution,"'he model did not in-elude iron in the capsule position.

However,"'n the calculation based on the actual core power distribution, th'e model did include iron in the cap-sule position.

An analysis of the latter results indicates that the Cap-sule S lead factor is increased by lOX when the model includes iron in the capsule position.

Therefore, the lead factor based on the FSAR power dis-tribution should be 1.76 instead of 1.6.

The extrapolation of capsule fluxes to the vessel wall described in Section VII of this report utilizes the perturbation-corrected lead factor'of 1.76 since it is more conserva-tive than the 2.1 factor.

Also, the maximum vessel flux might be altered by future core loading schemes.

The spectrum-averaged cross sections needed for Equation (2) were cal-culated for the iron and copper reaction dosimeters as follows:

12.5 MeV Z

cr (E)f(E) dE 0.11 a

(> 1 MeV) 12.5 MeV Z

$ (E) dE 1.00 (3) 21

D.

Mechanical Pro ert Tests I

The irradiated Charpy V-notch 'specimens were tested on an instru-mented SATEC impact machine.

The test temperatures were selected to de-velop the ductile-brittle transition and upper shelf regions for each material.

Impact energy and lateral expansion transition curves were developed for each material using a tanh curve fitting program based on the following equation:

T Tl C

parameter

=

A + B tanh

[

]

T2 where A, B, Tl, and T2 are adjustable constants defined by employing a nonlinear least squares fitting routine.

In this relationship, the tanh function varies from -1 at the low temperature extreme to +1 at the upper f

temperature extreme.

Therefore,'

B is equal to the lower shelf param-

'I I

eter and A + -B is equal to the upper shelf parameter.

I Tensile tests were carried out'n a Dillon 10,000-lb capacity tester l

equipped with 'a strain gage extensopeter, load cell, and autographic record-I ing equipment.

The tensile specimens were tested at 250 F and 550 F.

Testing of the VOL specimens was deferred at the request of Florida

, Power 6 Light Company.

The specimens ash in storage at the SwRI radiation laboratory.

E.

Check Chemical Anal ses for 'Co er Three tested Charpy specime'ns,

'representing the two forging heats and the correlation monitor material,'ere analyze'd for copper content in accordance with ASTM Method E 322.

V.

CAPSULE S TEST RESULTS TURKEY POINT UNIT NO.

3 A.

Neutron Dosimetr Two discrepancies were noted when the gamma activities of the dosim-eter wires were measured.

The Co activity of the cadmium-covered wire from the middle position was over four orders of magnitude below those of the top and bottom cadmium-covered dosimeters.

Also, the nickel dosimeter did not exhibit a 5 Co peak.

Using an energy dispersive x-ray diffraction "

technique, the middle nickel wire from the Unit No.

3 capsule was identi-fied as being a Co-Al material, and the middle cadmium-covered wire from the Unit No.

3 capsule was identified as being a copper-b'ase material.

The specific activities obtained from the remaining dosimeters in Capsule S are presented in Table IV.

Two items are of interest:

(1) the weights of the two dosimeters in question support the qualitative analyses described above; (2) the iron dosimetry results indicate a 20/ decrease in neutron flux across the capsule in the radial direction, and that the cap-sule had been inserted in the vessel 180'ut of phase.

The saturated activities of the'osimeters, also given in Table IV, were based on the summary of plant operations given in Table V.

The neutron spectrum calculated for the Capsule S location is pre-sented in Table VI along with the spectrum-averaged, cross sections computed for the iron and copper dosimeters.

The mean value of fast neutron flux density of the Capsule S location determined from the ten iron and two copper flux monitors was 1.29 x 1011 cm

'sec E > 1 MeV.

Since Unit No.

3 operated for an equivalent 1266.97 full power days up to the 1977-78 refuelling outage, the resulting value of neutron fluence for Capsule S is 1.41 x 1019 cm 2 E > 1 MeV.

23

TABLE IV

SUMMARY

OF NEUTRON DOSIMETRY RESULTS CAPSULE S

TURKEY POINT UNIT NO.

3 Monitor Identification(a)

Fe-S9 (Top)

Fe-S7 (Top-Middle)

Fe-R8 (Middle)

Fe-S5 (Bottom-Middle)

Fe-Sl (Bottom)

Radial Locatl.og in Ca sule Vessel Side Activation Reaction 54Fe(n,p)54Mn Dosimeter Wei ht (m )

54.5 70.8 45.6 59.3, 62.3 Measured Activity (d s/m )

7.21 x 103 6.36 z 103 5.78 x 103

=

6.63 x 103 6.44 x 103 Saturated Activity ds/

)

9.24 x 103 8.15 x 103 7.40 x 103 8.49 x 103 8.26 z 103 Fe-P9 Fe-P7 Fe-R6 Fe-P5 Fe-Pl (Top)

(Top-Middle)

(Middle)

(Bottom-Middle)

(Bottom)

Core Side 220.9 261.5 167.4 233.0 270.8 5.60 x 10 5.46 x 10 5.16 x 103 5.69 x 103 5.29 z 103 7.18 x 103 7.00 x 103 6.61 x 103 7.29 x 103 6.77 x 103 Cu (Top)

Cu (Bottom)

Ni (Middle)

Co (Top)

Co-Cd (Top)

Co (Middle)

Co-Cd (Middle)

Co (Bottom)

Co-Cd (Bottom)

Center Cu(n,c)

Co Ni(n,p)

Co 59Co (n, y) 60Co 51.4 53.0 8.7 9.7 9.4 9.5 30.7 9.9 7.9 2.25'x 10 2 24 x 102 (b)-

2.34 x 107 1.11 x 107 2.07 x 107 (c) 2.01 x 107-9.40 x 106 6.53 x 102 6.53 x 102 (b) 6.79 x 10 3.23 x 107 6.02 x 107 (c) 5.83 x 107 2.73 x 107 (a)

See Table II for location within capsule.

(b)

Monitor material was. identified as being bare Co in Al.

(c)

Monitor material was identified as being a Cu-base alloy (purity undetermined)

TABLE V

SUMMARY

OF PLANT OPERATIONS TURKEY POINT UNIT NO.

3 Oper.

Period Start Dates~dt*

Operating

~D Shutdo6dn

~D Reactor Pover

~Ot t 86)

Equivalent 0 er.

Da s T

Decay Tine after Period, 10 12 13 14 15 16 17 18, 19 20 21 6

22 23 24 25 26 27 28

'29 30 31 11/2/72 11/4/72 11/5/72 11/13/72 11/18/72 11/21/72 12/4/72 12/9/72 12/22/72 1/16/73 1/23/73 3/6/73 3/13/73 4/6/73 4/16/73 7/27/73 8/9/73 10/21/73 10/29/73 12/7/73 12/2'/73 12/27/73 12/30/73 3/19/74 3/31/74 5/12/74 5/13/74 6/8/74 6/14/74 9/14/74 9/23/74 10/5/74 12/15/74 3/4/75 3/9/75 7/16/75 7/20/75 10/26/75 12/25/75 1/2/76 1/3/76 2/22/76 2/28/76 3/7/76 3/12/76 5/1/76 5/3/76 5/28/76 6/3/76 6/20/76 6/26/76 8/13/76 8/17/76

'/25/76 9/1/76 11/14/76 1/19/77 4/24/77 4/29/77 7/20/77 7/22/77 11/4/72 11/5/72 11/13/72 11/18/72 11/21/72 12/4/72 12/9/72 12/22/72 1/16/73 1/23/73 3/6/73 3/13/73 4/6/73 4/16/73 7/27/73 8/9/73 10/21/73 10/29/73 12/7/73 12/21/73 12/27/73 12/30/73 3/19/74 3/31/74 5/12/74 5/13/74 6/8/74 6/14/74 9/14/74 9/23/74 10/5/74 12/15/74 3/4/75 3/9/75 7/16/75 7/20/75 10/26/75 12/25/75 1/2/76 1/3/76 2/22/76 2/28/76 3/7/76 3/12/76 5/1/76 5/3/76 5/28/76

'/3/76 6/20/76 6/26/76 8/13/76 8/17/76 8/25/76 9/1/76 11/14/76 1/19/77 4/24/77 4/29/77 7/20/77 7/22/77 11/Z4/77 2'

25 42 24 102 73 39 6

79 42 26 92 12 79 129 98 8

50 8

50 25 17 48 8

74 95 82 125 13 13 10 13 8

14 12 71 60 66 Total 14 3,500 577 3,035 21,225 49,954 30,845 165,595 112,547 60 '72 9,122 129,361 83,442 53,782 180,930 20,104 152,829 270,408 200,868 13,035 107,391 16,907 106,529 53,240 35,010 103,731 16,556 159,357 196,939 173,054 266 866 2,787,325 0.01 1.59 0.26

1. 38 9.65

22. 71 14.02 75 ~ 27 51.16 27.53 4.15

58. 80 37.93 24.45 82.24 9.14

69. 47 122.9.1 91.30 5.92

48. 81 7.68 48.42 24.20 15.91 47.15

7. 53 72.44 89.52 78.66 116.76 1,266.97 1846 1837 1829 1811 1773 1724 1693 1581 1495 1448 1428 1346 1292 1265 1167 1146 996 862 760 692 641 627 572 545 522 468 456 375 21'4 127 25

TABLE VI FAST NEUTRON SPECTRUM AND FOIL ACTIVATION CROSS SECTIONS FOR CAPSULE S

TURKEY POINT UNIT NO.

3 Energy Range (MeV

10. 00 12. 5 8.18 10.0 6.36 4.96 4.06 8.18 6.36 4.96 3.01 4.06 2.35 3.01 1.83 2.35 1.00 1.11 0.55 1.00 0.11 0.55

. 1.11 1.83 DOT 3.5 Calculated Neutron Flux 3.04 x 10 9.61 x 108 2.57 x 109 4.96 x 109 5.00 x 109 9.21 x 109 1.52 x 1010 1.87 x 1010 5.03 x 1010 0.92 x 1010 7.02 x 1010 1.24 x 1011 54Fe(n,p)54Mn Cross Section barns 0.521 0.578 0.578 0.491 0.352 0.222 0.098 0.024 0.0025 63Cu(n,y) 60Co Cross Section barns 4.78 x 10 2

4 26 x 10-2 1.53 x 10-2 2.85 x 10 2.78 x 10 4 1.12 x 10 4

5.96 x 10 5 3.86 x 10 5 2.09 x 10 5

1.20 x 10 5

5.48 x 10 6

8.46 x 10 7

0.0902 barns (E

> 1 MeV)

Fe oC

=

0.000985" barns (E > 1 MeV)

Cu 26

Much of the early work published on the radiation-induced embrittle-ment of ferritic steels correlated shifts in ductile-brittle transition temperature with neutron fluence calculated on the assumption that the neutron energies were distributed according to a fission neutron spectrum.

To provide information for reference only, the 'Unit No.

3 Capsule S fast neutron flux density based on a fission-spectrum cross section of 98.26 mb (E

> 1 MeV) for Fe(

) and 0.000606 mb (E > 1 MeV) for Cu

, is calcu-lated to be 1.31 x 1011 cm

'sec E > 1 'MeV.

The Unit No.

3 Capsule S

fast neutron flux (E > 1 MeV) computed with the DOT 3.5 code was 1.20 x 1011 cm sec E

> 1 MeV.

B.

Thermal Monitors The thrmal monitors were examined and had not melted.

This indicates that the capsule did not reach 579 F during the exposure period.

C.

Mechanical Pro ert Test Results The Charpy V-notch impact results obtained from specimens contained in Capsule S are given in Table A-1 in Appendix A.

The transition curves developed for each material using a tanh curve-fitting technique 'are also presented in Appendix A.

A summary of the notch toughness properties of the Turkey Point Unit No.

3 surveillance materials contained in Capsule S

are listed in Table VII.

These results indicate that the lower shell forging (123S266VA-1) is slightly more susceptible to radiation embrittle-ment than the intermediate shell forging (123P461VA-1).

This correlates with the reported copper contents of 0.079/ and 0.058X, respectively.

The results of tensile tests on specimens representing the lower and intermediate shell forging materials contained in Capsule S are listed in Table A-2 in Appendix A.

The, stress-strain curves and tensile test data 27

TABLE VII NOTCH TOUGHNESS PROPERTIES OF CAPSULE S

SPECIMENS TURKEY POINT UNIT NO.

3 50 ft-lb Cv Tem (de F)

Forging Forging 123S266VA-1 123P461VA-1 Correlation Monitor Irradiated, 1.41 x 1019 Unirradiated AT 4

-41 45

-6

-29 23 204 65 139 35 mil Cv Tem de F)

Irradiated, 1.41 x 1019 Unirradiated AT U

er Shelf Ener

'(ft-lbs)

Unirradiated Irradiated, 1.41 x 1019(

)

AE

-1

-53 52 154 122 32

-10

-45 35 145 128 17 182 41 141 76 60 16 (a) Neutron fluence, cm 2, E > 1 MeV 28

sheets are also reproduced in Appendix A.

The tensile strength and ductil-ity data obtained on the forging materials are compared 'to the unirradiated properties<16) in Figures 5 and 6.

These data also indicate that the higher copper forging material is slightly more sensitive to neutron radiation embrittlement than the lower copper forging material.

D.

Check Chemical Anal ses for Co er Check chemical analyses for the copper content of four tested Charpy V-notch specimens gave the following results:

Specimen No.

Material Identification Copper

~/)

S 8 P

9 R 5 R 6 123S266VA-1 123P461VA-1 Correlation Monitor Correlation.Monitor 0.06 0.06'.16 0.18 The results on the vessel surveillance materials confirm the copper contents indicated by WCAP 8631~

"~ and WCAP 7656.~

29

100 75 50 0

4J 25 Code:

Open symbols unirradiated Closed symbols irradiated, 1.41 x 1019 cm 2 E > 1 MeV 0

0 100 200 300 Temperature, feg F 400 500 600 75 50 4J 4l V

A 25 0

0 100 200 300 400 Temperature, deg F

500 600 FIGURE 5.

TENSILE PROPERTIES OF FORGING 123S266VA-1 TURKEY POINT UNIT NO.

3 30

100 75 50 OJ 4J 25 Code:

Open symbols unirradiated Closed symbols irradiated, 1.41 x 1019 cm 2y E > 1 MeV 0

75 0

100 200 300 Temperature, deg F

400 500 600 a

cLO 50 A

25,

\\

0 0

100 200 300 Temperature, deg F

400 500 600 FIGURE 6.

TENSILE PROPERTIES OF FORGING 123P461VA-1 TURKEY POINT UNIT NO.

3

32

VI.

CAPSULE S TEST RESULTS TURKEY POINT UNIT NO.

4 A.

Neutron Dosimetr The same anomalies concerning neutron dosimeter identification de-scribed for Unit No.

3 were noted for the Unit No.

4 Capsule S dosimeters, but these materials were not subjected to qualitative analysis identifica-tion check.

The specific activities obtained from the remaining dosimeters in this capsule are given in Table VIII.'he weights of the "middle cad-mium-covered" and the "nickel" wires support the supposit'ion that these dosimeters were made of a copper alloy and an aluminum-cobalt alloy, respectively.

The iron results again show a 20X decrease in neutron flux

- across the capsule in the radial direction, and it appears that the capsule was placed in the vessel in the planned orientation.

The saturated activi-ties of the dosimeters, also given in Table VIII, were based on the summary of plant operations given in Table IX.

The neutron spectrum calculated for the Capsule S location is given in Table X along with the 'spectrum-averaged cross sections computed for the iron and copper dosimeters.

The resulting mean value of fast neutron flux density at the Unit No.

4 Capsule S location was 1.16 x 1011 cm"2 sec 1, E > 1 MeV.

Since Unit No.

4 operated for an equivalent 1249.13 full power days of operation up to the 1978 refuelling outage, the calculated value of neutron fluence received by this capsule is 1.25 x 1019 cm E > 1 HeV.

To provide information for reference only, the Unit No.

4 Capsule S

fast neutron flux density based'n fission-spectrum cross sections~18~

is calculated to be 1.25 x 1011 cm sec E > 1 MeV.

The fast neutron flux density at the Capsule S location computed with the DOT 3.5 code was 1.18 x 1011 cm-2.sec-l E > 1 MeV.

33

TABLE VIII SR&fARY OF NEUTRON DOSIMETRY. RESULTS CAPSULE S

TURKEY POINT UNIT NO.

4 Monitor Identification( )

Fe-P9 (Top)

Fe-R5 (Middle)

Fe-Pl (Bottom)

Fe-S9 (Top)

Fe-R7 (Middle)

Fe-Sl (Bottom)

Cu (Top)

Cu (Bottom)

Ni (Middle)

Co (Top)

Co-Cd (Top)

Co (Middle)

Co-Cd (Middle)

Co (Bottom)

Co-Cd (Bottom)

Radial Location in Ca sule(

Core Side Vessel Side Center Activation Reaction 54Fe (n, p) 54Mn Cu(n,a)

Co 58Ni(n,p)58Co 5 Co(n,y)

Co Dosimeter Mei ht m )

245.2 193.1 174.7 188.7 265.2 206.0 49.8 50.3 9.2 8.9 8.1 9.5

18. 0 8.3 9.6 Measured Activity (d s/m )

5.29 x 103 5.13 x 103 5.25 x 103 4.58 x 103 4.20 x 103 4.48 x 10 2.30 x 102 2.23 x 102 (b) 2.50 x 107 9.36 x 106 1.91 x 107 (c) 2.26 x 107 1.02 x 107 Saturate'd Activity d s/m) 7.37 x 103 7.15 x 103 7.31 x 103 6.39 x 103 5.85 x 103 6.25 x 103 6.91 x 102 6.72 x 102 (b) 7.54 x 10 2.82 x 107 5.75 x 107 (c) 6.80 x 107 3.08 x 107 (a)

See Table II for location within capsule.

(b)

Monitor material was identified as being bare Co in Al.

(c)

Monitor material was identified as being a Cu-base alloy (purity undetermined).

TABLE IX

SUMMARY

OF PLANT OPERATIONS TURKEY POINT UNIT NO.

/4 Oper.

Period Date Shutdo>>n

~D~

Operating

~D

~

Reactor Po>>er

~Out ~ut lAtd~t gquivaleot Oper.

De~a T

Decay Tise after Period t

10 12 13 14 15 16 17 18 19 20 21 22 23 25 26 27 28 29 30 31 32 33 34 36 06/19/73 07/01/73 07/02/73 07/07/73 07/09/73 07/14/73 07/18/73 07/29/73 08/05/73 09/01/73 09/02/73 09/23/73.

09/24/73 11/04/73 11/16/73 01/03/74 02/03/74 04/03/74 04/05/74 04/17/74 04/18/74 05/25/74 05/31/74 08/18/74 09/10/74 10/27/74 11/02/74 12/04/74 12/07/74 01/06/75 01/10/75 03/30/75 06/21/75 08/03/75 03/09/75 09/21/75 10/01/75 10/12/75 10/12/75 01/10/76 01/17/76 04/18/76 06/10/76 06/12/76 06/16/76 06/17/76 06/19/76 09/10/76 09/16/76 09/24/76 09/29/76 10/10/76 10/14/76 10/28/76 12/03/76 01/06/77 01/11/77 01/25/77 Oi/30/77 03/20/77 03/26/77 04/26/77 05/04/77 05/09/77 08/03/77 08/11/77 08/15/77 10/29/77 11/11/77 02/14/78 03/09/78 07/01/73 07/02/73 07/07/73 07/09/73 07/14/73 07/18/73 07/29/73 08/05/73 09/01/73 09/02/73 09/23/73 09/24/73 11/04/73 11/16/73 01/03/74 02/03/74 04/03/74 04/05/74 04/17/74 04/18/74 05/25/74 05/31/74 08/18/74 09/10/74 10/27/74 11/02/74 12/04/74 12/07/74 01/06/75 01/10/75 03/30/75 06/21/75 08/03/75 08/09/75 09/21/75 10/01/75 10/12/75 10/13/75 01/10/76 01/17/76 04/18/76 06/10/76 06/12/76 06/16/76 06/17/76 06/19/76 09/10/76 09/16/76 09/24/76 09/29/76 10/10/76 10/14/76 10/28/76 12/03/76 01/06/77 01/11/77 01/25/77 01/30/77 03/20/77 03/26/77 04/26/77" 05/04/77 05/09/77 08/03/77 08/11/77 08/15/77 10/29/77 11/11/77 02/14/78 03/09/78-08/13/78 1

1 12 31 2

23 83 10 1

7 53 36 86 13 23 12 5

11 27 21 41 48 59 12 37 79 47 32 30 79 43 43 11 89 92 2

1 83 8

11 14 34 14 49 31 5

8 75 95 157 Total 3>269 2,596 3,036 7,056 21>408 28,405

~ 61>074 72>903 103>061 22>345 75,126 159,505 98,345 58>541 63>850 168,707 86,227 90>287 22,450 190,599 193>789 2,289 480 174,409 14>159 23>551 27>692 71,425 30>659 107,197 66>448 10>825 12,603 157>049 202,068 3>4 604 2,748,037 1.49

1. 18

1. 38 3.21 9.73 12.91 27.76 33.14

46. 85 10.16

34. 15 72.50

44. 70 26.61 29.02 76.68

39. 19

41. 04 10.20 86.64 88.09 1.04 0.

22'9.28 6.44 10.70 12.59 32.47 13.94 48.73 30.20 4.92 5.73 71.39 91.85 143.00 1249.13 1869 1863 1856 1841 1807 1785 1743 1683 1593 1579 1541 1456 1386 1348 1315 1232 1106 1057 1036 946 847 792 787 702 688 672 654 584 565 511 474 461 367 238 180 35

TABLE X FAST NEUTRON SPECTRUM AND FOIL ACTIVATION CROSS SECTIONS FOR CAPSULE S

TURKEY POINT UNIT NO.

4 10.00 12.5 8.18 10.0 6.36

= 8.18 4.96 6.36 4.06 4.96 3.01 4.06 2.35 3.01 1.83 2.35 1.11 1.83 1.00 1.11 0.55 1.00 0.11 0.55 Energy Range (MeV)

DOT 3.5 Calculated Neutron Flux 3.07 x 108 9.71 x 108 2.60 x 109'.03 x 109 5.07 x 109 9.35 x 109 1.55.x'1010 1.90' 10'10'.3.2 x 1010 0.94 x 1010 7.14 x 10 1.26 x 1011 54Fe(n,p)54Mn Cross Section (barns 0.521 0.578 0.578 0.491 0.352 0.222

0. 098-0.024 0.0025 Cu(n, y)

Co Cross Section (barns) 4.78 x 10-2 4.26 x 10-2 1.53 x 10-2 2 85 x 10 3 2.78 x 10 4

1.12 x 10 4

5.96 x 10 5

3.86 x 10 5

2.09 x 10 1.20 x 10 5 5.48 x 10 6

8.46 x 10 7

oF

=

0.0900 barns (E > 1 MeV)

Fe aC

=

0.000980 barns (E > 1 MeV)

Cu 36

Thermal Monitors I

The =thermal monitors were examined and had,not melted.

.This indicates that the capsule did not reach 579 F during the exposure period.

C.

Mechanical Pro ert Test Results The Charpy V-notch impact results'btained from specimens contained, in Capsule S are given in Table B-1 in Appendix B.

The transition'curves de-veloped for each material using a tanh curve-fitting technique are also pre-sented in Appendix B.

A summary of the notch toughness properties of the Turkey Point Unit No.

4 surveillance materials contained in Capsule S are listed in Table XI.

These results indicate that forging heat 123P481VA-1 has more sensitivity to neutron radiation embrittlement than forging heat 122S180VA-1, even though their copper contents-are almost identical (see Table III).

The results of tensile.tests on specimens representing the lower and intermediate shell forging materials contained in Capsule S are listed in Table B-2 in Appendix B.

Also included in this appendix are the stress-strain curves and tensile test data sheets.

The tensile strength and duc-tility data obtained on these forging materials are compared to the unir-radiated properties~

~ in Figures 7 and 8.

These results indicate that both forging heats have about the same low 'irradiation sensitivity as would be expected from,the copper contents of these materials.

D.

- Check Chemical Anal ses for Co er=

Check chemical analyses for the copper content of three tested Charpy V-notch specimens, made with an x-ray fluorescence technique, gave the fol-lowing results:

37

TABLE XI NOTCH TOUGHNESS PROPERTIES OF CAPSULE S

SPECIMENS TURKEY POINT UNIT NO.

4 50 ft-lb C Tem (de F

Irradiated, 1.25 x 10 Unirradiated RENT Forging 1228180VA-1 Forging 123P481VA-1 60 25 35 Correlation Monitor 195 80 115 35 mil C Tem (de F)

Irradiated, 1.25 x 1019(

Unirradiated hT

-15ll 46-2 48 174 62 112 U

er Shelf Ener ft-lb Unirradiated Irradiated, 1.25 x 10 AE 132 122 10 135 123 12 122 88 34 (a)

Neutron Fluence, cm 2, E > 1 MeV

100 75 rn 50 4J 25 Code:

Open symbols unirradiated Closed symbols irradiated, 1.25 x 1019 cm E > 1 MeV 0

75 0

100 200 300 Temperature, deg F

400 500 600 50 4J Q

25 0

0 100 200 300 Temperature, deg F

400 500 600 FIGURE 7.

TENSILE PROPERTIES OF FORGING 122S180VA-1 TURKEY POINT UNIT NO.

4 39

100 75 M

50 25 Code:

Open symbols unirradiated Closed symbols irradiated, 1.25 x 1019 cm 2, E > 1 MeV 0

75 0

100 200 300 Temperature, deg F 400 500 600 50

~pl

~pl V

A 25 0

0 100 200 300 Temperature, deg F

400 500 600 FIGURE 8.

TENSILE PROPERTIES OF FORGING 123P481VA-1 TURKEY POINT UNIT NO.

4 40

Specimen No.

Material Identification Copper

~/

S-l P-1 R-1 122S180VA-1 123P481VA-1 Correlation Monitor nil

.02

.08 These results are below those reported in WCAP 7660.~

~

The background radiation resulting from the gamma activity of each irradiated specimen was nearly twice that observed for the Unit No.

3 chemical analysis samples.

As a result, the background count was a much larger fraction of the total count in the copper peak, reducing the accuracy of the, result.

41

42

VII.. ANALYSIS OF RESULTS A.

Introduction The analysis of data obtained from surveillance program specimens has the following goals:

(1)

Estimate the period of time over which the properties of the vessel beltline materials will meet the fracture toughness requirements of Appendix G of 10CFR50.'his requires a projection of the measured reduction in Cv upper shelf energy to the vessel wall using knowledge of the energy and spatial distribution of the neutron flux and the dependence of C upper shelf energy on the neutron fluence (trend 'curves).

(2).-

Develop heatup and cooldown curves to describe the oper-ational limitations for selected periods of time.

This requires a projection of the measured.shift'n RTNDT to the vessel wall using knowledge of the dependence of the shift in RTNDT 'on the neutron fluence (trend curves) and the energy and spatial distribution of the neutron flux.

The capsules removed from the Turkey Point Nuclear Power Plant pressure vessels during the 1977-78 refuelling outages contained specimens represent-ing the intermediate and lower shell course beltline forging materials but

- did not contain any weld metal or HAZ specimens.

Since the weld metal will'ontrol the RTNDT for both units(,

~

), the results of this analysis may not affect the current heatup and cooldown lim'its.

It is anticipated that the, reliability of neutron embrittlement trend curves will be improved as more 'surveillance data become -available and a better understanding of the factors affecting radiation embrittlement has t~

been achieved.

As an example of the latter, Mr. E.

C. Biemiller of Combus-tion Engineering, in a paper(1

).given 'at the 8th ASTM International Sym-posium on Effects of Radiation on Structural Materials held in St. Louis

in May 1976, indicated that a parameter of

(% Ni +

% Si)

(% Mo +

% Cr +

% Mn) may explain the variation in radiation embrittlement observed in fer-ritic materials of nominally the same copper content.

In addition, at the 9th ASTM International Symposium on Effects of Radiation o'n Structural Ma-terials held in Richland, Washington, in July 1978, Mr; J.

D. Varsik of Combustion Engineering presented a related paper entitled "An Empirical Evaluation of the Irradiation Sensitivity of Reactor 'Vessel Materials."

At the s'arne conference, Westinghouse presented information which indicates 3

that neutron embrittl'ement may reach a limiting value when the irradiation is carried out for long times at approximately 550"F in lower -neutron flux environments.. Also, the Metal properties Council is developing new radia-tion damage curves that will be based on more data than those currently in use.

B.

Pressure Vessel Fast Neutron Ex osure 1.

Turke Point Unit No.

3 Based on the dosimetry results obtained from Capsule S,

and using the conservative lead factor of 1.76 calculated for this capsule, the maximum fast flux incident on the Turkey Point Unit No.

3 pressure vessel is calculated to be 7.33 x 101 cm sec y

E > 1 MeV.

The fast neutron flux is attenuated as it penetrates the pressure vessel wall.

Con-servative estimates of the ratio of fast flux at depths of 2'n.

(1/4T) and 6 in. (3/4T) to that incident on the pressure vessel I.D. surface're 0.60 and 0.15, respectively.<

Utilizing these factors, the maximum fast flux at the 1/4T depth in.the Turkey Point Unit No.

3 pressure vessel wall is estimated to be 4.40 x

'4

10 0 cm

'sec 1, and that at the 3/4T depth is estimated to be 1.10 x 1010 cm sec 1, approximately 12% higher than determined from -the analysis of Capsule T.(14)

The predicted neutron exposures for the Turkey Point Unit No.

3 pressure vessel at the I.D. surface, 1/4T and 3/4T positions after 5, 10, and 32 Effective Full Power Years (EFPY) of op'eration,are summarized in Table XII.

2.

Turke Point Unit No.

4 Using the Capsule S dosimetry results, and the conservative lead factor of 1.76 calculated for this capsule, the maximum fast, flux incident on the Turkey Point Unit No. 4'ressure vessel, is calculated to be 6.59 x 1010 cm

'sec E > 1 MeV.

The maximum 'fast flux values calcu-lated for the 1/4T and 3/4T positions within the vessel wall are 3.95 x 1010 and 0.99 x 101 cm

'sec

, respectively.

These values aie within 1%

of those determined from the analysis of Capsule T.(

)

The predicted neu-tron exposures for the Turkey Point Unit No.

4 pressure vessel at the I.D.

surface, 1/4T and 3/4T positions after 5, 10, and 32'FPY are presented in I

Table XIII.

C.

Vessel Material Notch Tou hness A method for estimating the reduction in C upper shelf energy as a

function of neutron fluence is given in Regulatory Guide 1.99.(7)

The results obtained to date on the vessel beltline forging materials and the reference steels contained in Capsules S and T are compared to a portion of Figure 2 of Regulatory Guide 1.99 in Figure 9.

The shelf energy re-sponse of each vessel beltline forging material from Turkey'Point Unit Nos.

3 and 4 was equal to or less than the minimum'base metal response curve (0.10% Cu) given in Figure 2 of Regulatory Guide 1.99.

The shelf 45

TABLE XII PROJECTED MAXIMUMPRESSURE VESSEL EXPOSURES(

TURKEY POINT UNIT NO.

3 Location in Vessel Mall Neutron Flux E>lMeV 5 EFPY Neutron Fluence E > 1 MeV 10 EFPY 32 EFPY I".D. Surface 7 33 x 1010 cm-2.sec-l 1.16 x 1019 cm 3 x 1019 cm-2' 4

- x 1019 cm-2 1/4T Depth 3/4T Depth 4 40 x 1010 cm-2.sec-l 1.10 x 1010 cm-2.sec-l 6.94 x 101 cm 1.4 x 10 cm 4.4 x 101 cm 1 73 x 1018 cm-2 3 5 x 1018 cm-2 1 1 x 1019

-2 (a)

Based on results from Capsule S

TABLE XIII PROJECTED MAXIMUMPRESSURE VESSEL EXPOSURES~

TURKEY POINT UNIT NO.

4 Location in Vessel Hall Neutron Flux E > 1 MeV 5 EFPY Neutron Fluence E > 1 MeV 10 EFPY 32 EFPY I.D. Surface 1/4T Depth

'/4T Depth 6.59 x 1010 cm-2.sec-l 3 95 x 1010 cm-2.sec-l 0.99 x 1010 cm-2.sec-l 1.04 x 101 cm-2 6.24 x 1018 cm 1.56 x 1018 cm 2 2 1 x 1019 cm 2 1.2 x 1019 cm 3.1 x 1018 cm 6.7 x 1019 cm 4.0 x 1019 cm 2 1.0 x 10 cm (a)

Based on results from Capsule S

60 4

40 4J 20 10

.20/ CU' ff f

0.15% Cu, 0.10%

CU

"-:, Htf 8:i f,'ii f/4 4*1'>> -.f f ft=VH

-4 r}

'I'4 4

t " "ifi

- Ti k.'44

'4X L

1 4

4 4

4

.4-.-4 kk 1

$4 ifif

},'1 W

Vf. "

-~ 0 "20

-f '0'~07 4/Cu CU 4

1 o.o5)

-rf i-- t~~

'.e~)

-a -0',056%

kf fk 11 L

W44 4

4 A.i

}aft

~4 c) 6 Jkt i t krf 4 fr

~

.~'~;r.

Code:

-I-t -'=

1tk 4

f/4

~

I*

fk' Jt (kgb 4 4 k.

~

t 4+

4 V

n 0

T SI 0

SA302B ASTM Correlation MonitorUnit 3 Lower Shell Forging 123S266VA-1 Unit 3 Intermediate Shell Forging 123P461VA-1 SA533B ASTM Correlation MonitorUnit 4 Lower Shell Forging 122S180VA-1 Unit 4 Intermediate Shell Forging 123P481VA-1 Unit 3';

4 I

~ t Unit 4

'$L

}g

-f4=

k. '+

8 1018 6

8 10>>

Neutron Fluence, cm E > 1 MeV FIGURE 9.

COMPARISON OF DECREASE IN SHELF ENERGIES OF TURKEY POINT UNIT NOS.

3 AND 4 VESSEL FORGING MATERIALS AND SURVEILLANCE REFERENCE STEELS TO REGULATORY GUIDE 1.99 TREND CURVES

energy response of the 0.20%

Cu A302B reference steel, (Unit 3) was also less than the appropriate (0.20% Cu) trend curve, but the shelf energy response of the 0.14%

Cu A533 reference material (Unit 4) was above the applicable (0.15%

Cu) trend curve.

D.

Ad usted, Reference Te erature A similar approach can be taken to estimate the increase in RTNDT as a

function of fast neutron fluence.

Figure 10, which compares the Turkey Point Unit Nos.

3 and 4 vessel forging material and reference steel results to the appropriate radiation damage trend curves developed by Westinghouse(15),

in-dicates that the responses of the forging materials are well below the 0.10%

Cu trend curve and the responses of the reference steels are in good agree-ment with the appropriate trend curves.

The same data are compa'red to the trend curves of Regulatory Guide 1.99

) in Figure 11.

This shows that the responses of the vessel forging (7) materials are below the 0.08%

Cu trend curve, the response of the A533B reference steel is in good agreement with the 0.14%

Cu trend curve, and the response of the A302B reference steel is below the 0.20%

Cu trend curve.

Although there is considerable scatter in the data, the transition tem-perature shifts determined for the vessel surveillance materials are in rea-sonable agreement with both sets of trend curves.

However, the vessel forg-ing material data appear to follow the slope of the Regulatory Guide 1.99 trend curves (Figure ll) and the correlation monitor material data appear to follow the slope of the Westinghouse trend'urves (Figure 10).

E.

Heatu and Cooldown Limits for Normal 0 eration Heatup and cooldown limit curves were developed for 0-5 and 5-10 EFPY of operation for the Turkey Point Unit Nos.

3 and 4 nuclear power plants after the removal of the first surveillance capsule (T) from each 49

400 3

I.i II I

Code:

V 0

0v 3

3Unit 4

4Unit SA302B ASTM Correlation MonitorUnit Lower Shell Forging 123S266VA-1 Unit Intermediate Shell Forging 123P461VA-1 SA533B ASTM Correlation MonitorUnit Lower-Shell Forging 122S180VA-1 Unit Intermediate Shell Forging 123P481VA-1 1

~

I' t '

'Ij, t]jr

- '.}j"'

It I I

I 4

200 i.

I l

I

.I>

~

~

I I I

~, I I

I~

~'t

~20%+4uC'$

i j CulI--

0.1 100 0.20%

Cu 80 0

0.15%

Cu I

J

~I~ I qjt r>>~

+

tt 60 g

CI 40

g; TI.

ki',Fl Li"t Ii t ILt }I4i, TT'..

~ijt I ~

4j

<<07 0.10%

Cu I ~

~-

~t

~-ti

~ j

-I4

'w33A I

IIj.

hatt

'LI ~>>j

'I

~

I'I

'C}tt}

~

I I

~'

~'

~

20

~ I..Ii t~

tj

~

I I

t ~

I 10 8

1019 2

Neutron Fluence, cm E > 1 MeV 6

S 1020 1018 FIGURE 10.

COMPARISON OF INCREASE IN REFERENCE TEMPERATURES OF TURKEY POINT UNIT NOS.

3 AND 4 VESSEL FORGING MATERIALS AND SURVEILLANCE REFERENCE STEELS TO WESTINGHOUSE TREND CURVES~

4 4

400

200, t

100-80:

Code:

V 0

0 Y

0 0

I'

~

1.

. '1 I

0,'.20%

Cuj

~ 1 SA302B ASTM Correlation MonitorUnit 3 Lower Shell Forging 123S266VA-1 Unit 3 Intermediate Shell Forging 123P461VA-1 SA533B ASTM Correlation MonitorUnit 4 Lower Shell Forging 122S180VA-1 Unit 4 Intermediate Shell For@in 123P481VA-1 Unit 3; Unit 4:

1 S. 20

-+ 0-314%

'Cu' W0 60 Cl A ~

t 4ci-C cC 1 I~

ti e

t0 40

-'.20% Cu '=',. 0.14%

Cu

~

0.011%

P ~ 0.012% P A

~ C I

ct CCtC cr

-'~ 0.08%

Cu

~,0.012%

P I M 20 1

tt =F t

CC.

4

$1 tt I C.c

~C 1

~.

~~

QC AC r -p 44 6

s 1018 2

6 s

1019 Neutron Fluence, cm E ) 1 MeV FIGURE 11.

COMPARISON OF INCREASE IN REFERENCE TEMPERATURES OF TURKEY POINT UNIT NOS.

3 AND 4 VESSEL FORGING MATERIALS AND SURVEILLANCE REFERENCE STEELS TO REGULATORY GUIDE 1.99 TREND CURVES

reactor vessel. (21)

The projected fast neutron exposures resulting from the analyses of the second surveillance capsule (S) from each unit are in good agreement wi.th those reported earlier.(

~

)

Also, since the S capsules did not contain specimens representing the controlling (weld metal) beltline material, there is no basis for revising the projected values of RTNDT used to develop the current set of heatup and cooldown limit curves.

F.

Ca sule Removal Schedule A third capsule is scheduled for removal from each reactor vessel after 10 calendar years of operation.

Based on the past operating histo-ries of 'the Turkey Point nuclear power plants, 10 calendar years of opera-

,tion should correspond to approximately 7 EFPY of operation.

It is recom-mended that Capsule V, a Type II capsule containing weld metal specimens, be removed from each vessel at that'ime.

The projected fast neutron flu-ence for the V capsules after 7 EFPY is 1.3 x 1019 cm (E > 1 MeV), ap-proximately twice the fluence received by the T capsules..(>4>>>>)

The data obtained from the V capsules should provide the information necessary to revise the heatup and cooldown limitations for operation beyond 10 EFPY of operation.

52

VII.

REFERENCES Title 10, Code of Federal Regulations, Part 50, "Licensing of Produc-tion and Utilization Facilities."

ASME Boiler and Pressure Vessel

Code,Section III,-."Nuclear Power Plant Components,"-

1974 Edition.

ASTM E 208-69, "Standard Method for Conducting Drop-Weight Test to Determine Nil-DuctilityTransition Temperature of Ferritic Steels,"

1975 Annual Book of ASTM Standards.

Steele, L. E.,

and Serpan, C. Z., Jr., "Analysis of Reactor Vessel Radiation Effects Surveillance Programs,"

ASTM STP 481, December 1970.

Steele, L. E., "Neutron Irradiation Embrittlement of.Reactor Pressure Vessel Steels," International Atomic Energy Agency, Technical Reports Series No. 163, 1975.

ASME Boiler and Pressure Vessel

Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,"

1974 Edition.

Regulatory Guide 1.99, Office of Standards Development, U.S. Nuclear Regulatory Commission, July 1975.

Comments on Regulatory Guide 1.99, Westinghouse Electric Corporation, obtained from NRC Public Document Room, Washington, D.C.

Position on Regulatory Guide 1.99, Combustion Engineering Power Sys-

tems, obtained from NRC Public Document Room, Washington, D.C.

ASTM E 185-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,"

1975 Annual Book of ASTM Standards.

ASTM E 399-74, "Standard Method of Test for Plane-Strain Fracture Toughness of Metallic Materials," 1975 Annual Book of ASTM Standards.

Witt, F. J.,

and Mager, T. R., "A Proceduie for Determining Bounding Values of Fracture Toughness, I<lc, at Any Temperature,"

ORNL-TM-3894, October 1972.

Loss, F. J., Editor, "Structural Integrity of Water Reactor Pressure Boundary Components,"

NRL Memorandum Report 3782,'ay 1978.

Yanichko, S. E., Phillips, J. H., Anderson, S. L., "Analysis of Cap-sule T from the Florida Power

& Light Company Turkey Point Unit No.

3 Reactor Vessel Radiation Surveillance Program,"

WCAP 8631, December 1975.

15.

Norris, E. B., "Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 4, Analysis of Capsule T," Final Report, Southwest Research Institute Project 02-4221, June 14, 1976.

16.

Yanichko, S. E., "Florida Power 6 Light Company Turkey Point Unit No.

3 Reactor Vessel Radiation Surveillance Program,"

WCAP 7656, May 1971.

17.

Yanichko, S. E., "Florida Power

& Light Company Turkey Point Unit No.'

Reactor Vessel Radiation Surveillance Program,"

WCAP 7660, May'971.

18.

Steele, L. E.,

and Serpan, C. Z., "Analysis of Reactor Vessel Radiation Effects Surveillance Programs,"

ASTM STP 481, December 1970.

19 ~

Biemiller, E. C., and Byrne, S. T., "Evaluation of the Effect of Chemical Composition on the Irradiation Sensitivity of Reactor Vessel Weld Metal," Irradiation Effects on the Microstructure and Properties of Metals, ASTM STP 611, November 1976.

20.

Telcon, E. B. Norris to K. Hoge (NRC Staff), January 19, 1977.

21.

Norris, E. B.,

and Unruh, J. F., "Pressure-Temperature Limita-tions for the Turkey Point Unit Nos.

3 and 4 Nuclear Power Plants,"

SwRI Project 02-4383-039, June 30, 1976.

54

APPENDIX A Charpy V-Notch and Tensile Test Data Turkey Point Unit No.

3 Capsule S

H K

55

TASLE A-1 CHARPY IMPACT DATA CAPSULE S

TURKEY POINT UNIT NO.

3 (Neutron Fluence

= 1.41 x 1019 cm 2, E

> 1 MeV)

Material Forging 123P461VA-1 Spec.

No.

P-9 P-7 P-10 P-8 P-6 P-5 P-1 P-2 P-3 P-4 Temp.

~de F)

=-50

-20

-10 0

10 30 72 140 210 300 Energy'ft-1b 3.5

33. 0

70. 0 72.0 59.5 79.5 77.0 127.5 132.0 125.5 Lat. Exp.

(mils) 5 24 49 53 45 58 59 92 89 94 Fract.

App.

% Sh'ear) nil ~

5 75 50 80 90 10 100 100 100 Forging 123S266VA-1 Correlation Monitor Steel S-9 S-7 S-8 S-6 S-lo S-5 S-l S-2 S-3 S-'4 R-1 R-2 R-5'-6 R-8 R-3 R-7 R-4

-50

-20 0

10 20

'30 72 140 210 300

'72 140 160 180 200 210 250 300

13. 5 31.0

46. 0 48.0 67.0 86.0 117.5 121.5 124.0 122. 0 15.0 21.0 49.5 33.0

'51. 5 53.0 35.5

60. 0 9

27 37 34 51

.74 81 90 95 91 13 19 40 31 50 47 34 51 nil 5

5 5

75 100 50 100 100 100 5

15 40 50 95 90 60 100 56

TABLE A-2 TENSILE PROPERTIES OF SURVEILLANCE MATERIALS CAPSULE S TURKEY POINT UNIT NO.

3 (Neutron Fluence

= 1.41 x 1019 cm 2, E > 1 MeV)

Spec.

No.( )

Test Temp.

~F 0.2% Yield Strength si Tensile Strength

~si Fractur'e Strength

~si Fracture Stress

~si Uniform Elongation(b)

%)

Total Elongation

(%)

Reduction in Area

~%)

S-2 250 64,200 84,000 52,100 157,100 10.3 23.4 66.8 S-l 550 54; 400

-=

78,900 50,900 149,700 7.6 20.9 66.0 P-2 P-1 250 550 61,300 (c) 81,200 50,900 167,100 (c)

(c)

(c) 10.2 (c) 22.5 (c) 69.5 (c)

(a) Material Code:

S = 123S266VA-1; P = 123P461VA-1 (b) Using change of cross-sectional area in unnecked portion of specimen per ASTM E 184-62.

(c) Specimen not tested.

Temperature controller malfunction caused overheating of specimen.

SHELL'ORGING 123P4BIVA-I

.84/83/79

=TURKEY POINT -N0.-3,.UNZRRADIATED--

--A~

73.588 TI~

1.444 T2~

SS.143

)

rn 288 i 168 b 168 0- '

148 2: 128

+o 188 ~-'

-'e4 48 2e 0

8

'88 TEHPERATURE CDEG F>

388 SHEl L FORGING 123P461 VA-I 84/83/79

=-TURKS'..POINT NO. 3. rCAPSULE S - --...--

)A-"

65.588.

~

-~148

-.B~.62:588

,TI 14.396

-T2~. 61...767 "',.....',

0

-128

'Q 188 Cl

,~, Se 08 -- 0 68 o

8 lee TEHPERATURE-CDEG F>

388'8

" 148 t.

(g 128

-Z z 188 0

o.

88 X

K-P QB SHELL FORGING 123P461VA-1 84/83/79'URKEY POINT -NO. 3, - UNIRRADIATED A~

45.888 B~

43.888 Tl~ -27.187 T2~

75.699 t

t G

28 0

8 188

-.TEHPERATURE (DEG F3 288 388 B- ~

~ 45. 888

~

Tl~

~ 12. 864 T2=- 88..811 SHELL FORGING 123P461VA-1':

" 83/84/79

~TURKEY POINT NO. 3.

CAPSULE S A=

47.888 148

, Q128 HZ m 188 OH Z

68

,8

~ - ~

~ 48

~

~

T t

t',

188

~

TEHPERATURE CDEG F>

288 0 a 59

SHELL FORGING 123S266VA-1, 84/83/79

. TURKEY POINT NO. 3.

UNIRRADZATED ..

A~'8.588 B~ - 75.588 Tl 6.228 T2~ 119.868.

hht h

n m

I-4

, 0"

'UJ o

00 168 168 0""

0 se>

48 28 148 128 ~t;.

h htS)'"',0

0. '..

,h t

8 188 TEHPERATURE (DEG F>

. B~

59.588.

" Ti~

14.475

. -T2~.47.245

..." SHELL FORGING 123S266VA-1...., 84/83/79 TURKEY POINT N0...3.'.,CAPSULE S.

-A~,; 62';588 148

= 128.j-0

~

= ~

m 188 I

68 0

~ 68+

'o -"

0 8

lee TENPERATURE (DEG F>

60

n (0

HZ SHELL FORGING 1233266VA-I,

~ - TURKEY POINT NO. 3

-UNIRRADIATED.-

A=

44.588 B~

42.588

.-TIM -44.837

-T2~

37.SBS, 128 T 84/83/79 zOH COz 0.X Gled

-0 ~

0 V

! 83 t.

t Q

0

~

~'

188 TEMPERATURE (DEG F3 SHELL FORGING 1233266VA-I

=

84/83/79

~. - TURKEY POINT-, NO;"3, CAPSULE S --. '--

A~ - 47.888

6=- 45t888 T I ~

12. 282:

T2~ -49,816 x

~ 128m, H

• z".

z 188 O

a.

88 X

W t'

.- '0-t::'-

t-0 0

68 4J

~J'~

x 8

188 TEMPERATURE CDEG Fi 388 e a 61

f 48 t

128-UNIRRADIATED A382 A~

39.S88 S

36.588 Tl l6.223 T2=

64.971 83/28/79

~ 188 I-5 88 5

68 O

'g 0

188 288 TEMPERATURE (DEG F>

128 T'"

IRRADIATED A382 TURKEY POINT NO 3, CAPSULE 3 A~

31.588 B~

28.S88 Tl~ 1S7.819 T2=

59.332 83/28/79 m 188 m

b 88 68 8,

8 188 288 TEMPERATURE CDEG F) e 62

Q 128' z

I z 188 H

o.

88 X

68 UNIRRADIATED A382 TURKEY POINT NO 3 A=

37.888 0~

32.888 Tlo 44.964 T2~

67.674 4/5/79 188 288 TEMPERATURE (DEG F>

IRRADIATED A382.

.TURKEY POINT NO 3 CAPSULE S

A=

38.888 B-25.888 Tl~ 166.849 T2~

77.261 Q 128~

z 188~

0H 68~

X 68 4/5/79 188 288 TEMPERATURE CDEG F>

63

~

~

outhwest Research Institute Department of Materials Sciences TENSILE TEST DATASHEET Est. U. T.S.

Psl Spec.

No.

Temperature ZS

'F Initial G. L. / OO in.

InitialDia.. XS ~

in.

Date X W + 7P Strain Rate Initial Thickness in InitialArea O', D

/

I Initial Width.

lne Top Temperature "',> cS g 0

Bottom Temperature ZSZ Maximum Load 4 (Z5 lb 0.2% Offset Load 3~5 lb

"~l 1ne 0.02% Offset Load 'b Final Diameter Final Area

- O,O/4 Upper Yield Point lb f;es>> 7jle. (Ifu.>>le>>l>>co ~ e.e 4)

Maximum Load Initial Area 0 2 Y S 0.2% Offset Load Initial Area g4 O/0 psi j')

(o4- /E psi I

p p2%%u Y 'S

0. 02% Offs et Load InitialArea PS1 U

er Yield Point Initial Area PS1

\\

I Final G. L. - Initial G. L.

Elongation

~

%%u R A initial Area - Final Area

]

InitialArea

~Ml'0ALSl 5, bIII,f~a>>>> h>>e>>Is -

t I

X Ice

= lo Fish< Qt A. (L3g-Ptfct'-t-b)

Qg$

Signature:

~ Jg~~

64

L A I 0

-godo -;

/

toed tll t

yl 0

.~~

f

~

f

-4

. ~, oo8

~ ~. ~

I I

I 4

~ 4 L

65

southwest Research Institute Department of Materials Sciences TENSILE TEST DATASHEET Test No. T-Est. U. T.S.

I psi Project No. C7P-S/9/OO J'pec.

No.

Ct Temperature 5 5 < 'F

/J Strain Rate Initial G. L.

, <OO in.

InitialDia..

5 ~

in.

Initial Thickness in Machine No. DrZ4oW Date g'-

r'nitial Area 5'0 Initial Width lno Top Temperature Bottom Tempe ratur e OF OF Maximum Load Zd PS lb

0. 2% Offset Load

~~7O lb Final Gage Length Final Diameter Final Area 6'. 4'f4 1no in.

1no 0.02% Offset Load Upper Yield Point lb built. I lit.(got-togo+en Po inst ooI%I+<

Maximum Load InitialArea p

0 2% Offset Load Initial Area 0 psi 0

02%%u Y S 0.

02%%uo Offs et Load Initial Area PS1 U

er Yield Point InitialArea psl

% Elongation Final G. L. - Initial G. L. x Initial G. L.

Initial Area - Final Area lp InitialArea gurAAC.W( A ~

Lj h)tFA '8 8 Q5 Q g Ftagg+ +It A 40+ +~~*&)

/'ignature:

gg,o

><<ee =

g 4 66

I I I 4

I I

I 4+&

~

-hy I

I I

y-~

I

'4 e

I

~

4= --~

I"*-

~-.

I I

I C.' P/0 UU4 I

'I

. ~

~ +

~

I

~

'I

~ ~

-4 I

,oos 4

67

(.~uthwest Research Institute Department of Materials Sciences TENSILE TEST DATASHEET Test No. T-2 Est. U. T.S.

PS1 Project No. 02-Spec.

No.

Temperature ~~'F ll Strain Rate Initial G. L. / QOO in.

Initial Dia.." 2 5 in, Initial Thickne s s le ~

Machine No. ~/~o~

Date ~ J'0- TP InitialArea 0, D Initial,Width in.

Final Gage Length Final Diameter Final Area

.0 Top Temperature WS'2 Bottom Temperature Z5/

'F oF 1ns I

in.

2 Maximum Load

+

~u lb 0.2% Offset Load 3o@~

lb 0.02% Offset Load lb lb Upper Yield Point Fisei Qttgaq (<st itct-aao)

>4~~ iv, Maximum Load Initial Area

0. 2% Y. S.

0. 2 o Offset Load InitialArea 0l 3 a Ps1 InitialArea U

er Yield Point",

Initial Area Final G. L. - Initi'al G. L.

% Elongation I ti 1G.LL I

InitialArea - Final Area InitialArea tmai Pta.

'gt3t FoP H EL t i<6v julg. ivy.Qgt:tet:0) psl PS1 Signature:

Ii' 68

e

~

4+t

~"

I ~

~

~

I ~

I P

/

I

/Ohio I

'T

~- ~-

g F70 0

~t

~

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( I

~ I.

~

I

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0:

0 I

OO+

Wii&4 gN./~IJ, J

I

,'OOPS I '-" '~OO

~

~

~

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I*

I

~ I 69

( ~cuthwest Research Institute Department of Materials Sciences TENSILE TEST DATASHEET Test No. T-Est. U. T.S.

psi Project No. 02-5/3'/

Spec.

No.

Temperature ~++

'F Initial G. L.

Initial Dia.

in.

Date N

Initial Thickness in.

Initial Area Initial.Width oT in.

- sPwr<EJ

@ED Top Temperature Bottom Temperature Final Gage Length.

Final Diameter

'F lne ln.

Maximum Load

0. 2% Offset Load 0.02% Offset Load Upper Yield Point lblb:

lb lb Final Area inc Maximum Load InitialArea PS1

% Offset Load InitialArea PS1 0- 02 0 Offset Load Initial Area PS1 U

er Yield Point

~ Upper Y.S.

In t l Area PS1 Final G. L. - Initial G. L.

% Elongation

Initial G. L.

Initial Area - Final Area Initial Area Signature:

70

APPENDIX B CHARPY V-NOTCH AND TENSILE TEST DATA TURKEY POINT UNIT NO.

4 CAPSULE S

71

TABLE B-l CHARPY IMPACT DATA CAPSULE S~

TURKEY POINT UNIT NO.

4 (Neutron Fluence

= 1.25 x 1019 cm 2, E

> 1 MeV)

Material Spec.

Temp.

Ne.

~{de F

Energy

~{ft-1b Lat. Exp.

~mils Fract.

App.

% Shear Forging 123P4SlVA-1 Forging 122S180VA-1 P-10 P-9 P-8 P-7 P-1 P-2 P-3 P-4 P-5 P-6 S-lo S-9 S-S S-7 S-6 S-l S-2 S-3 S-4 S-5 0

20 40 60 80 110 160 210 235 260

-30 0

20 40 60 80 110 160 210 235 14.2 7.8

51. 2 44.6 76.1 67.5 98.3 124.5 116.7 122.0 49.2 35.6 59.2 91.1 102.1 85.9 89.7 123.6 121.1 121.7 11 7

41 36 65 57 75 94 89 94 41 27 47 69 81 70 72 93 93 87 nil nil nil 10 15 25 95 100 100 100 10 nil

{

5 30 70 25 50 100 100 100 Correlation Monitor Steel R-1 80 R-2 110 R-8 135 R-3 160 R-7 185 R-4 210 R-5 260 R-6 310 14.4 20-. 9 19.1 34.1 45.3 55.4 89.7 87.2 13 17 16 29 40 48 75 68 nil 10 15 20 25 60 95 100 72

TABLE B-2 TENSILE PROPERTIES OF SURVEILLANCE MATERIALS CAPSULE S TURKEY POINT UNIT NO.

4 (Neutron Fluence

= 1.25 x 10 cm E > 1 MeV)

Spec.

No (a)

Test Temp.

( F) 0.2% Yield Strength

( si)

Tensile Strength

~(si)

Fracture Strength

~(si)

Fracture Stress

~(si)

Uniform Elongation(")

(%)

Total Elongation

(%)

Reduction in Area

~%)

S-l S-2 250 550 67,800

'7,500 88,700 91,100 58,000 170,200 57,400 168,200 6.7 5.9 20.7 20.9 65.9 65.8 P-1 P-2 250 550 66, 200 62,300 87,800 88,100 56,700 176,900 61,300 1707300 7.6 7.6 22.5 19.3 68.0 64.0 (a) Material Code:

S = 122S180VA-1; P = 123P481VA-1 (b) Using change of cross-sectional area in unnecked portion of specimen per ASTM E 184-62.

SHELL FORGING 123P481VA-1 TURKEY POINT NO.

4, UNIRRADIATED A~

69.888 B~

66.888 Tl~

-55.181 0

T2~ 181.547 0

128~

84/83/79 e 188 00---

C9 68 O

8 188 TEMPERATURE CDEG F>

P SHELL FORGING 123P481VA-1 TURKEY POINT NO. 4.

CAPSULE S A~

63.888

-148

- B=

68,888 Tl~

82.151 T2M 98.569

-128

.84/83/79 Q 188

Sel 8= --=

68-0

'4e $

8 lee TEMPERATURE (DEG F)

W 74

SHELL FORGING 123P48IVA-I

.TURKEY POXNT NO.

4 UNXRRADXATED A~

45.888 148 ~

B~

43.888 I Tl~

19.785 I T2-93.851 l

n I

y) 128 q H

z, 188--

OH a.

88 y.

84/83/79 5

68'8>

0 0

$8y 8

188 TEMPERATURE CDEG F>

288 388 SHELL FORGING 123P48IVA-I 84/83/79

- - TURKEY-POINT NO. 4, CAPSULE S-A=

47.888 B=

45.888 TI~

71.857

-T2= - 95.145 Q 1Mt z 188 4 QH 88 68 0

8 188 TEMPERATURE (DEG F>

75

128 SHELL FORGING 1223188VA-1 TURKEY POINT NO. 4.

UNIRRADIATED A=

67.588 B~

64.588 Tt~

13.648 T2=

78.367 0

84/83/79 rn 188 ~

88 (9

0 68 28 p

0 8

188 TEMPERATURE CDEG F)

-B=

59.588 Tl~

21.288 T2~ -84.412 128 SHELL FORGING 122S188VA-1 TURKEY POINT NO. 4.

CAPSULE S A~

62.588 148 84/83/79 Q1M)"

88 68

)0 0

0" 0

8 188 TEMPERATURE CDEG F) 76

. 8-44.888 TI~

-5.413 T2~

37.483 Q 128 HK m 188~

OH z

a.

88 X

4J SHELL FORGING 1223188VA-I

~ TURKEY POINT NO.- 4 UNIRRADIATED

~ A~

46.888 148 84/83/79 8

M g

-28 0

8 188 TEMPERATURE

<DEG F3 SHELL FORGING 1223188YA-I 84/83/79 TURKEY POINT NO. 4.

CAPSULE S A~

46.588

-I'48

--6=

44,588 TI~

15.538 T2~ 73. 771 (g 128 M

m. 188 0M

a. -88 4--

X 68 +

8 188 TEMPERATURE

<DEG F>

77

148 128 e 188 1-v 88

~w 68 u

k UNIRRADIATED AS33 A~

63.588'~58.588 Tl~

96. 1,51 T2~ '78.852

P 0

0

" 83/27/79 188 288 TEMPERATURE (DEG F) 148 128 IRRADIATED A533 TURKEY POINT NO 4, CAPSULE S A-"46.588 B~

41.S88 Tl~ 187.874 T2~

92.862 83/27/69 m 188 I-b 88 68 0

0 188 288 TEMPERATURE (DEG F) r a 78'

148 Q 128 HX z 188 O

z a.

88 X

68 UNIRRADIATED AS33 TURKEY POINT NO 4

=

A~

44. 758 B~

48.758 TI~

81.494 TZ~

78.223 4/5/79 188 288 TEMPERATURE CDEG F>

B 33.588..

Tl~ 183.212 TZ~

87.185

-Q 128

""'Z V

z 188 O

o.

88 ~

OC 68 IRRADIATED AS33 TURKEY POINT, NO 4.

CAPSULE 3 A~

38.588 148 4/5/79

-48 28 188 288 TEMPERATURE (DEG F) 79

c Scathe est Research Institutek

,Department of Materials Sciences TENSILE TEST DATASHEET Test No. T-Est. U. T.S.

psl Project No.&2-5 ZP> -C'E7l Spec.

No.

Temperature +g$ O' Strain Rate In t' G. L.

C?OO Initial Dia.

~ ~

in.

Initial Thickness 'n.

Machine No., 47/sco~

Date Y2 5 7F InitialArea P~l+y'

~

InitialWidth ine Top Temperature Bottom Tempe ratur e 7

'F Maximum Load gXSE lb

0. 2%. Offset Load; '&39'b Final Gage Length /. Z O 7 1Qe 0.02% Offset Load lb Final Area Final Diameter x<70 1ne ills Upper Yield Point Fesu 3<h ~ (tie uecl4 C t ace lb M 'o'.

~p Initial Area 0

0 2 o Offset Load Initial Area Cpl 8+

psi 0

02%%u Y S

0. 02% Offset Load InitialArea psl U

Y S U

er Yield Point Initial Area ps,i I

Final G. L. - Initial G. L.

% Elongation x 100 =

0

%%u R A Initial Area - Final Area x 100 =

. InitialArea

/w>nwc SIA ~

idktg<A, I Ou a)QegGO)

YlcCI Signature:

80

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o2.-538o Spec. go. S-I 2<o F

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4

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4 I

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Southwest Research Institut Department of Materials Sciences TENSILE TEST DATASHEET Test No. T-Est. U. T. S.

psi Project No. 4'~-. >~ ~~ +~/

Spec., No.

8 Z

Temperature g 5 5~ 'F Strain Rate Initial Dia.

W&cP'n.

Initial Thickne s s Initial G. L.

r ~+

in.

Machine No. Nl44' ~

Date Initial -Width xn.

Top Temperature + ~S /

~6s t Final Gage Length Maximum Load g4~0 0 lb 0 2% Offset Load 92-4 0 lb 0.02% Offset Load Final Diameter Final Area

. O/45 lne ills Upper Yield Point lb s

F"- @a, (~"a.~a F ~. - )

X.i.q-(

Maximum Load Initial Area 0 2 Y S 0.2% Offset Load Initial Area 0

02%%u' S

02% Offset Load, InitialArea psl U

er Yield Point InitialArea psi Final-G. L. - Initial G.-L.

% Elongation In tial G. R.

Initial Area - Final Area InitialArea J

(Fssc>*.Gin-nscssa)

I tete n Signature:

82

02 -5380

,/pic, Qo. S-Z.

sro F

=-

-m=-=.

I.

I

~.

~

~

l

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t'

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h h.

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0 ',OOD lOOO

]/

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=--,

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h 0 :0 r 002

,oo+

. oo4

&A)irJ r r,N/r'g.

oe8 83

Southwest Research Institut

'e par tme'nt of Mate rials Science s TENSILE TEST DATASHEET Test No. T-Spec.

No.

Temperature 25 0 F

Strain Rate Est. U. T.S.

'I Initial G. L. ', ~OO in.

Machine No. W/44>~

'ate

~~-5 --P7'.

InitialDia.. g 0 in.

Initial Thickne s s in.

InitialArea N ~g psi Project No. <~ S ~'P<-<</

Initial width In+

Top Temperature

'F Maximum Load Y275 lb Bottom Temperature + Q g

'F

~ Final Gage Length P. Z25 0.2% Offset Load 8~2$

lb 1

0-02% Offset Load Final Diameter

' K/

Final Area,.

0 lS'4 Ill I

Upper Yield Point Fi+Aw34A.(g~.ee'c~p QAtlod Ib I

Maximum Load,

+77 InitialArea A

I 0.2% Offset Load.

0. 02 0 Offset Load InitialArea psI U

er Yield Point InitialArea PSI Final G..L. - Initial G. L;

% Elongation

% R-A.

I

'L&IAAt ~IA ~

Wl j

Inltlal G. L-Initial Area -'inal AIea x" 100 =

InitialAlea 100 =

d7 r~

)c. loo

~ Signature:

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02-6'380 SpeC. Ho. P-l 2.50'F-

~

~

6ago l

q(

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. 004 i t4./s Q.

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Southwest Research Institu Department of Materials Sciences TENSILE TEST DATASHEET Spec.

No.

0.

Est. U. T. S.

Initial G. L.

ps1 Machine No.

z //a Temperature 5 5 0 F

Initial Dia.

g xn.

Date Strain Rate i dl Initial Thickness Initial Width in.

InitialArea 1n>>

Top Temperature 7 5 5/

Bottom Temperature WS 5 ~,

'F Maximum Load

+AS lb

,0. 2/o Offset Load

~<Cn4 lb Final Gage Length Final Diameter Final Area dc 7C'xn.

2 0-02% Offset Load lb Upper Yield Point lb Fiultv DA. (()a.>>(c>>>>>>(y s>>~>>4y yl.

~q. (

Maximum Load pp InitialArea 0YS0 2% Offset Load Initial Area 2 8 psi 0

02 Y S

• 0 02% Offset Load InitialArea psl U

er Yield P'oint PP'r

IitialArea Final G. L. - Initial G. L.

% Elongation

1

.x 1 00 =

I xt al G.L, Initial Area - Final Area

% R.A.

xi 100 =

Initial Area

/ f&(M Ac SiA ~

gg((:>>(>>.>>( BL>>4(,, =Q(

g

~~

~)

(

(.

Signature:

gg Og

)0" la() =

86

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c c

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~

c+

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~

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+ '

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c

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c b ~ c-

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loco

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87