ML20027A316

Analysis of Capsule R from Wi Elec Pwr Co Facility Reactor Vessel Radiation Surveillance Program.

ML20027A316
Person / Time
Site:	Point Beach
Issue date:	08/31/1978
From:	Shaun Anderson, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20027A314	List: ML20027A312 ML20027A315 ML20027A316
References
WCAP-9357, NUDOCS 7811140133
Download: ML20027A316 (62)

Text

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WESTINGHOUSE CLASS 3 i

, . CUSTOMER-DESIGNATED DISTRIBUTION N.

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i ANALYSIS OF CAPSULE T FROM THE WISCONSIN  !

ELECTRIC POWER COMPANY POINT BEACH i NUCLEAR PLANT UNIT NO. 2 REACTOR VESSEL  !

RADIATION SURVEILLANCE PROGRAM I

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J. A. Davidson S. L Anderson R. P. Shogan

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. August 1978

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APPROVED: (Myb

( J. N. Dhirigos, Manager i Structural Materials Engineering Prepared by Westinghouse for the Wisconsin Electric Power Company Work Performed Under ElZP 102 l

Mthough the informadon contemed in this report is non-proprietary, no distribution shell be made outsede Westing-house or its t.aoeneses without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 7 P///V4/33 gd -m P So - 3 a ( p

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TABLE OF CONTENTS t Secdon Title Page f

1

SUMMARY

OF RESULTS 11  !

. ~, 2 INTRODUCTION 2-1 -

, 1 3 BACKGROUND 31 i I

4 DESCRIFTION OF PROGRAM 41 i t

5 TESTING OF SPECIMENS FROM CAPSULE T 51 '

51. Charpy V Notch Impact Test Results 5-2 l 5 2. Tensile Test Results 5-20  !

5-3. Wedge Opening Loading Tests 5 20 i i

6 NEUTRON DOSIMETRY ANALYSIS 6-1

61. Description of Neutron Flux Monitors 61 i

, 52. Analytical Procedures 6-4 l S3. Results of Analysis 6-9 l

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l LIST OF ILLUSTRATIONS i Figure I Title Page t i

41 Arrangement of Surveillanca Capsules in the Point Beach  !

Unit No. 2 Reactor Vesset 42 l

'S 42 Capsule T Schematic Showing Designed Arrangement of I j

, .' Specimens, Thermal Monitors,and Dosimeter Placement l

and Orientation with Respect to the Core and Vessel Wall 4-5 5-1 Charpy V Notch Impact Data for the Point Beach Unit No. 2 Pressure Vessel intermediate Shell Forging 123V500VA1 5-3  ;

5-2 Charpy V-Notch Impact Data for the Point Beach Unit No. 2  !

Pressure Vessel Lower Shell Forging 122W195VA1 5-4 53 Charpy V Notch Impact Data for the Point Beech Unit No. 2 ,

Pressure Vessel Wald Metal 55  ?

5-4 Charpy V-Notch impact Data for the Point Beach Unit No. 2 - k

. Weld Heat-Affected Zone Metal 5-6 55 Charpy V Notch Imoact Data for SA533 Grade B Class 1 l ASTM Correlation Monitor Material 57 5-6 Charpy impact Specimen Fracture Surfaces for Point Beach Unit No. 2 Pressure Vessel Intermediate Shell l Forging 123V500VA1 5-14 5-7 Charpy impact Specimen Fracture Surfaces for Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 5 15 5-8 Charpy impact Specimen Fracture Surfaces for Point Beach Unit No. 2 Weld Metal {

5-16 l 5-9 Charpy impact Specimen Fracture Surfaces for Point Beach l

Unit No. 2 Weld Hest Affected Zone Metal 5-17 5-10 Charpy impact Specimen Fracture Surfaces for Point Beach

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Unit No. 2 ASTM Correlation Monitor Material 5-18

, 5-11 Point Beech Unit No. 2 Material 30 ft Ib Transition i

Temperature increases as Compared to Westinghouse i Pmdictions 5 19 l 5 12 Tensile Properties for the Point Beech Unit No. 2 Pressure Vessel Intermediate Shell Forging 123V500VA1 5 22 5-13

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. Tensile Properties for the Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 5-23 j l

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1 LIST OF ILLUSTRATIONS (cont) .

Figure Title p.g.

5-14 Fractured Tensile Specimens from Point Beach Unit No. 2 '

Pressure Vessel Intermediate Shell Forging 123V500VA1 5 24 5-15 Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 5 25 5 16 Typical StressStrain Curve for Tension Specimens (Tension Specimen No. E14) 5-26 61 j Point Beach Unit No. 2 Reactor Geometry 6-3 j 6-2 Calculated Azimuthal Distribution of Maximum Fast  :

Neutron Flux (E > 1.0 Mev) within the Point Beach Unit No. 2 Reactor Vessel q l 6-12 _/

60 Relative Axial Variation of Fast Neutron Fiux (E > 1.0 Mov) Incident on the Point Beach Unit No. 2 Reactor Vessel 6 13 64 Calculated Maximum End-of-Life Fast Neutron Fluence (E > 1.0 Mov) as a Function of Radius within the Point Beach Unit No. 2 Reactor Vessel 6-14 e

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i LIST OF TABLES Table Title Pogs I t

t 41 Chemistry and Hest Treatment of Matenal Representing  !

the Core Region Shell Forgings and Weld Metal from the '

7-s Point Beach Unit No. 2 Reactor Vesset 43

.. 42 Chemistry and Hest Treatment of Surveillance Material Representing 12 inch-Thick A533 Grade B Class 1 Correlation Monitor Material 4-4 51 The Effect of 550*F frradiation at 9.45 x 1018 n/cm2 (E > 1 Mov) On The Notch Toughness Properties of the Point Beech Unit No. 2 Reactor Vessel Impact Test Specimere 58 52 Summary of Point Beach Unit No. 2 Reactor Vessel Surveillance Capsule Charpy impact Test Results 59 l 53 Charpy V-Notch Impact Data for the Point Beach Unit No. 2

Pressure Vessel Intermediate Forging 123V500VA1

'i trradiated at 550*F, Fluence 9.45 x 1018 n/cm2 (E > 1 Mov) 5 10 I;

. 54 Charpy V Notch Impact Data for the Point Sesch Unit No. 2 l Pressure Vessel Lower Shell Forging 122W195VA1 Irradiated at  !

550*F, Fluence 9.46 x 1018 n/cm2 (E > 1 Mov) 5 10 i 55 Charpy V-Notch impact Data for the Point Beach Unit No. 2 i f Pressure Vessel Wald Metal Irradiated at 550*F, Fluence l 9.45 x 1018 n/cm2 (E > 1 Mov) 5-12  !

5-6 Charpy V Notch Impact Data for the Point Beach Unit No. 2  !

Pressure Vessel Weid-Heat-Affected Zone Metal Irradiated at t 550'F, Fluence 9.45 x 1018 n/cm2 (E > 1 Mav) 5-12

. 5-7 Charpy V-Notch Impact Data for the Point Beach Unit No. 2 ASTM SA533 Grade B Class 1 Correlation Monitor Material  !

Irradiated at 550*F, Fluence 9.45 x 1018 n/cm2 (E > 1 Mov) 5-13  !

58 frradiated Tensile Properties for the Point Beech Unit No. 2 Pressure Vessel Sheil Forgings 5-21 6-1 Neutron Flux Monitors Contained Within Capsule T S2 i S2 Irradiation History of Capsule "T" 67 l

- S3 Spectrum-Averaged Reaction Cross Sections Used in Fast '

Neutron Flux Derivation 6-9  :

S4 Results of Fast Neutron Dosimetry for Capsule T 6-10

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6-5 Results of Thermal Neutron Dosimetry for Capsule T 6-11 {

! 6-6 Calculated Fast Neutron Flux and Lead Factors for '

Capsule T 6-15 i

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._ SECTION 1 i

SUMMARY

OF RESULTS The analysis which compared unirradiated with irradiated material properties of the reactor vessel material contained in the second surveillance capsule, designated T, from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 reactor pressure vessel led to 7

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, the following conclusions:

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s The capsule received an average fast fluence of 9.46 x 1018 n/cm2 (E > 1 Mov).

The predicted fast fluence for the capsule was 8.02 x 1018 n/cm2 (E > 1 Mov).

e The fast fluence of 9.46 x 1018 n/cm2 resulted in a 140*F increase in the 50 ft lb reference nil-ductility transition temperature (RTNDT) of the weid i

metal, which is the most limiting material in the core region of the reactor vessel. The intermediate pressure vessel shell forging 123V500VA1 and lower shell forging 122W196VA1 exhibited a 50 ft-lb transition temperature increase of 40*F and 13*F, respectively (specimens oriented parallel to the working  !

direction of the platas). The weld-heet effected zone material exhibited a 50-ft-Ib transition temperature increase of 103*F.

a The average upper-shelf impact energy of the limiting weld metal decreased from 65 to 56 ft Ib, which remains adequate for continued safe operation of the plant.

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i e it was determined that there is an increase of 105'F in the 30-ft Ib t transition temperature for the ASTM A533 Gr8 C11 reference correlation i

monitor material contained in the capsule. Since this transition temperature ,

1 increase falls within the 560*F trend band developed for this material, it is  !

concluded that the capsule was irradiated at approximately 550*F, and that the fluence measurements are reasonably accurate. l e The following end-of-life projected fast neutron fluences for the reactor vessel, based on 32 full-power years of operation at 1518 Mw as derived from both calculated and measured surveillance capsule results wem determined.

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e e Fast Neutron Fluence (n/cm2) .

Vesel Locetion Calculated Measured (Averags)

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Inner surface 3.9 x 1019 4.65 x 1019 .

% Thickness 2.4 x 1019 2.90 x 1019

% Thickness 7.5 x 1018 9.00 x 1018 The difference of approximately 16 percent in the calculated versus the measured fast neutron fluence is due in part to the i 10 percent uncertainty in the measured activities of the fast ,

neutron iron monitors. The remainder of the difference may be attributed to uncertainties in the monitor reaction cross sections, analytical approximations, and radial and axial peripheral core flux levels. On the whole, the agreement between calculation and measurement values '

is considered good. -

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, .. SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule T, the second capsule of the continuing surveillance program, which monitors the effects of neutron irradiation on the Wisconsin Electnc Power Company Point Beach Nuclear Plant Unit No.2 reactor pressure

, vessel materiais under actual operating conditions.

The surveillance progam for the Point Beach Unit No. 2 reactor pressure vessel materials was desipied and recommended by the Westinghouse Electric Corporation, Descriptions of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in WCAP 7712.UI The surveillance program, planned to cover the 40 year life of the reactor pressure vessel, was based on ASTM E-185 66, " Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors,"[21 This report summarizes testing and the postirradiation data obtained from the second material surveillance capsula (Capsule T) removed from the Point Beach Unit No. 2 reactor vessel, and discusses the analysis of these data. The data are also compared to results of the previously removed Point Beach Unit No. 2 surveillance capsule V, reported by Battelle Memorial Institute in 1975,I31 A revised removal schedule for the Point Beach Unit No. 2 radiation surveillance capsules remaining has been developed based on results of the first two capsules, and is included in the report,

1. Yenienko, S E, ans: Zuie, G. C., '9mestmein Michigen Pouver Co., and the Wiesonsin Elastne Posuer Canoeny Point semen Unit No. 2 Reestor Vessel Reeleson surwesIIense Propom." WCAP.7712, June,1971.

2. ASTM F _ 2+ EleHe, "surwessense Tests on Strustwei Metanete in Noteer Reactors," in ASTM stoneerds (197el, Port 31 Physsess ansi Mechen asi Tessing of Metelo - Mosessogreony, Nondamenctive Testing, Petique Effest of Temoo po. e3s442, Am. Sos. for Teenres and Metertess, Phasseshim, PA,1se7

3. Pernei, J. S., Formeio D. R., Lausry, L. M., Wooton, R. D., and Denning, R. S., "*oint Besen Unit No. 2 Pressare Veemet

.-  ! rweitense Program: svoluetton of Coomsie," setteile Research Report, June 10,1s75.

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~ SECTION 3  !

l BACKGROUND The ability of the large steel pressure vessel containing the reactor core and primary coolant to reset 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 s because it is subjected to significant fast neutron bombardment. The overall effects of fast  ;

\ neutron irradiation on the mechanical p:operties of low-alloy ferritic pressure vessel steels such I as SA508 Class 2 (base meterial of the Unit No. 2 reactor pressure vessel beitline) 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 con-ditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non<fuctile Failure," Appendix G to Section ill l

l, of the ASME Soiler and Pressure Vessel Code. Tne method utilizes fracture mechanics I concepts; it is based on the reference nil-ductility temperature, RT NOT-RTNDT si defined as the greater of the drop weight nil ductility transition temperature  :

(NDTT per ASTM E 208) or the temperature 60* F less than the 50 ft-Ib (and 35 mils lateral  !

expansion) temperature as determined from Charpy specimens oriented normal to the rolling l I

direction of the material. The RTNDT of a given material is used to index that material to I 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 stool. When a given material

, is indexed to the KIR curve, allowable strensintensity factors can be obtained for this meterial as a function of temperature. Allowable operating limits con then be determined utilizing ,

these alloweble stress-intensity factors. i i i RTNOT, and in turn the operating limits of nuclear power plants, can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittle-

,- ment or changes in mechanical properties of a given reactor pressure vessel steel can be  :

monitored by a reactor surveillance program such as the Point Beach Nuclear Plant Unit No. 2 l

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Reactor Vessel Radiation Surveillance Program,Illi n which a surveillance capsule is periodically -

removed from the operating nuclear reactor and the encapsulated specimens are tested. The - -

increase in the Charpy V. notch temperature (ARTNDT) due to irradiation is added to the criginal RTNDT to adjust the RTNOT or f radiation embrittlement. This adjusted RTNDT ~

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NOT nitial + ARTNDT) is used to index the material to the KIR curve and in turn to __

set operating limits for the nuclear power plant which take into account the effect of irradia-tion on the reactor vessel materials.

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1. Yanschko. 3. E and Zule. G. C *hn Michigen Power Co., and the Wisconsen Electre Power Company Point Beach unst No. 2 Reactor Veenet Reelenen Survoeitence Prograrn.** WCAP 7712. June,1971.

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DESCRIPTION OF PROGRAM i

Six surveillance capsules for monitoring the effects of neutron exposure on the Point Beach Nuclear Plant Unit No. 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor 3 vesant between the thermal shield and the vessel well at locations shown in figure 41. The

- vertical conter of the capsules is opposite the vertical conter of the core.

Capsule T was removed after approximately five calendar years (3.46 effective full-power years) of plant operation. This capsule contained Charpy V notch impact, tensile, and WOL specimens

! (shown in WCAP-7712) from the intermediate and lower ring. forgings, weld metal representative of the core region of the reactor vessel, and Charpy V notch specimens from weid-heet-effected-i zone (HAZ) material. The capsule also contained Charpy V notch specimens from the 12 ire thick ASTM correlation monitor material A533 Gr8 C11 fumished by Oak Ridge National Laboratory. The chemistry and heat treatment of the surveillance motorial are presented in -

tables 41 and 42.

l All test specimens were machined from the 1/4 thickness location of the forgings. Test  ;

specimens represent material taken at least one forging thickness from the quenched end of i

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the forging. All base metal Charpy V-notch and tensile specimens were oriented with the j longitudinal axis of the specimen parallel to the principal working direction of the forgings.  !

. The WOL test specimens were machined with the simulated crack of the specimen perpendic-l ular to the surfaces and working direction of the forgings. ,

Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of i the specimens tranwerse to the weld direction. Tensile specimens were oriented with the  :

longitudinal axis of the specine persitei to the weld.

Capsule T contained dosimeter wires of cooper, nickel, and aluminum-0.15 Wt. percent (cadmium shielded and unshielded). In addition, the capsule contained cadmium shielded dosimeters of Np237 and U238, located as shown in figure 42.

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! -i Unit No. 2 Reactor Vessel (Lead Factors for the Capsules are Shown in Parentheses) 42 l

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CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE  :

CORE REGION SHELL FORGINGS AND WELD METAL FROM THE POINT  !

,, BEACH UNIT NO. 2 REACTOR VESSEL  !

Chemical Analyses (percent)

Bement Forging 123V500VA1 Forging 122W195VA1 Wold Metal j C 0.20 0.22 0.079 i Mn 0.65 0.59 1.40 i P 0.000 0.010 0.014

\' S 0.000 0.008 0.013 Si O.24 0.23 0.55  ;

Mo 0.59 0.60 0.39  !

Cu 0.088 0.051 0.25 Ni 0.71 0.70 0.59 (

Cr 0.35 0.33 0.07 l Al <0.005 <0.005 <0.005 i

, N2 0.004 0.002 0.010 V 0.010 0.010 < 0.002 I Co 0.004 0.910 0.013  !

Heat Treatment i

Forging 123V500VA1 Hested at 1550*F, 9% hours, water-quenched j Tempered at 1200*F,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, aircooled i Struerelieved at 1125*F,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, furnace-cooled  !

Forging 122W195VA1 Hested at 1550*F,8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, water-quenched Tempered at 1200*F,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, aircooled Stressrelieved at 1125*F,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, furnacecooled i

t Weldment Stress-relieved at 1125*F,11% hours, fumece. cooled  !

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TABLE 4 2 .

CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL .-

REPRESENTING 12 INCH THICK A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL Chemical Analysis C Mn P S Si Ni Mo Cu Ladle 0.22 1.45 0.011 0.019 0.22 0.62 0.53 Check 0.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14

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v Heat Treatment 1675 25* F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Air-cooled 1600 25' F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Water-quenched 1125 25* F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Fumace cooled

' 1150 25* F - 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> - Fumacecooled to 600*F e

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. Thermal monitors mode from two lowmelting eutectic alloys and sealed in Pyrex tubes were i I

.. included in the capsule and were located as shown in figure 4 2. The two eutectic alloys and f their melting points are: {

2.5% Ag, 97.0% Pb Melting Point 579*F I 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590*  !

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SECTION 5 s TESTING OF SPECIMENS FROM CAPSULE T l i

The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was i performed at the Westinghouse Research and Development Labcratory with consultation by l Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with  !

10CFR50, Appendices G and H.

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-7712.Ill No discrepencies were found.

Examination of the two low-melting (579*F and 590*F) eutectic alloys indicated no molting  !

r of either type of thermal monitor. Based on this examination, the maximum temperature to j which the test specimens were exposed was less than 579*F.

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A Tinius Olsen Mode! 74 impact test machine was used to test the irradiated Charpy V-notch specimens per ASTM E23 72, " Notched Bar impact Testing of Metallic Materials." Before initiating the tests, the accuracy of the impact machine was checked against a set of standard specimens obtained from the Army Material and Mechanics Rosserch Center in Watertown, .

' Massachusetts. The results of the calibration testing showed that the machine qualified for certified Charpy V notch impact testing. l i

The tensile tests were conducted on a screwdriven Instron testing machine of 20,000-Ib capacity I per ASTM E8-89, " Tension Testing of Metallic Materials" and ASTM E2170, " Elevated Tem-perature Tension Tests of Metallic Materials." A crossheed speed of 0.05 inches per minute was used. A strain-gage extensometer measured the deformation of the specimen. Before testing, the

extensometer was calibrated against a Sheffield high magnification drum type extensometer

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calibrator.

l Eiavated temperature tensile tests Were conducted in a split tube furnace. Prior to testing, the f

, specimens were held at temperature a minimum of 20 minutes for thermal stabilization.  !

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1. Yenschko, s. E., and Zude. G. C.. "Wisconem Michigan Power Co and trte Wisconsin Electric Power company, Point Semen unst No. 2 Ramster Veemd Moeistion survestieneo Prayern.** WcAP 7712. June,1971.

51

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Temperature was monitored using a thermocouple in contact with the clevis-pin type upper . -

and lower specimen grips. Temperature was controlled within a tolerance of 2 3*F. - -

The lood-extension data were recorded on the testing machine strip chart. The yield strength, ultimate tensile strength, fracture loed, and uniform elongetion were determined from extenso-meter output plotted as load versus deflection. The reduction in area and total elongation were determined from specimen measurements.

51.

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CHARPY V-NOTCH IMPACT TEST RESULTS The irradiated Charpy V notch specimens represented the Point Beach Nuclear Plant Unit No. 2 reactor pressure vessel beltiine forging material, weld, and heat affected zone (HAZ) ,

material, and the ASTM reference correlation monitor material. The results are presented in figures 51 through 5 5. Comperison is made to the unirradiated results and to impact curves

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developed from testing of the first surveillance capsule. A summary of the increase in the 30 '

.___and 50 ft-Ib energy and 35-mil lateral expansion transition temperature and the decrease irr-- -- -

the upper shelf energy resulting from irradiation to 9.45 x 1018 n/cm2 is presented in ~'

table 5-1. Similar results from the first surveillance capsule (4.74 x 1018 n/cm2) are listed for i comparison in table 5 2.

i The test results obtained on the vessel beltline shell forging material are presented in figures 5-1 -

and 5-2, and tables 53 and 5-4. Comparison is made to previously tested surveillance capsule VIII irradiated to a fluence of 4.74 x 1018 n/cm2. Results show a 30*F increase in the -

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30-ft-lb temperature for forging 123V500VA1, the same shift which this forging exhibited after 4.74 x 1018 n/cm2. The shift in the 50-ft-Ib temperature of 40*F is slightly larger than s the shift at 4.74 x 1018 n/cm2. Forging 123V500VA1 exhibited a shift of 42*F in the

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35-mil lateral expansion (LE.) temperature after 9.45 x 1018 n/cm2, comparable to the 50 ft Ib shift at the same fluence level.

Forging 122W195VA1 exhibited shifts of 17*F and 13*F in the 30 and 50 ft Ib temperatures respectively. Once again, these shifts are essentially identical to those after 4.74 x 1018 n/cm2 fluence. The shift in the 35 mil LE. temperature was 17*F, comparable to the 30-ft lb temperature shift.

The upper shelf energy of both irradiated forgings exhibited an increase over the properties as unirradiated.

1. Perre. J. S., Permelo. D. R.. Loewy, L. M., Wooton R. D.. and Denning, R. S., "Pomt sesch Unit No. 2 Preensre Veenet Sueveillense Program. sweeustion of Capasse V." setteile Ressoren Report. June 10.1975.

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$ 0 - b, Q  !

i i

260 '

, 2t$0 -===. UNIRRADI ATED (OPEN PolNTS) t g

220 - 8 "5 * ' 018 N/CH2 (CLOSED POINTS) ,

200 . .. ,,75 x i0 i as few2 (sATTELLE g g '

180 -

............~~~  !

CuaVE, CAPSULE "Y ) y e*

(

) 3 160 - '

f m

IM - O / .*,- ,

j

! 120 -

o , ..

@ 100 5

o / ,@

• l 80 -

60 -

O @/ AB9 i t

- 4 50 FT. L8. TEMP.8 4C'F i 40  :

W 6 30 FT. LS. TEMP.: 30*F t O Q~* i i i

- -200 -100 0 100 200 300 l 1

TEMPERATURE (*F) 1 Figure 51. Charpy V-Notch imcact Data for the Point Beach Unit No. 2 l Pressure Venel Intermediate Shell Forging 123V500VA1 I 53 l

l . .

t 13,028-2 l

l .

I

' 1 lW - l l 2 O

I g l .

n -

o 3  ;

D 60 - 2 j

=

40 -

O 0 1 20 -

e i 0 - W l l

- lM - 2 --

e

,S m -

7 60 -

O

$ go - O , .-

4 20 - '- '

O o -

200

=me - UntRRAOIATED (oPEu PoluT3) 180 -

s.ss : lo ss e few2 (CLosto Points)

• 4.74 1 1088 m/Cy2 (BATTELLE CURVE CAPSULE 'V*)

iso -

ea _2 .

140 - /

~

~ 120 -

s

• / **'

s 100 -

M b ~

g .'

O so -

O e

j A50 FT. L8. TEMP. = 13*F 40 -

fg 20 -

p/

a so n. Ls. TE*P.. irv o I I I I 100 0 100 200 300 400 .

TDfERATURE ('F)

Figure 5-2. Charpy V Notch !mpact Data for the Point Beach Unit '

No. 2 Pressure Vessel Lower Shell Forging 122W195VA1

! 54 r

r .___ ..

13,028-8

. 100 _ l l l ,- g 3

I

_g M -

p Ea y2 g20

_ g *e 0

2

-*/

100 -

4 -

aw

• 0 -

&~R e -e

20 - B.6 C ,, p,,,,, _ _ .*e

s 0 _ 8  : -

lM

= UNIPRA01 Aft 0 (OPEN Points) 90 -

- s.ts x 10 s fenta (cLost0 P0tuis)

+..... 4.7% x 1033 N/CW2 (BA N m t, CAPSu u '1')

80 -

70 -

o / -9

. 60 o,/g =[*

~

O

( ) 30

. _ =

> f 4 50 FT. Ls. TEMP.s 1408F s --

EW p *,,..=**.........

Q/ g .-

30 -

9 Q,$ = ,. 5 9 4o , A30 FT. Ls. TEMP.s 145'F O fo e '..-

/

10 -

/

O O I I I I i

- -100 0 100 200 300 40 TDIPERATURE ('F)

Figure 53. Charpy V Notch impact Data for the Point Beach Unit No. 2 Pressure Vessel Weld Metal 55

1 .

{ . .' -

! 13,028-3 1

'A

{ 100 -1 1 0 -

e .-

g/

7*

A e -

I

.0 _

20 -g=.,2r,.- .

j.....~. _ _ . . .

0 _ 7 . ._

100 -

i E 80 - 0 0 8 E

O y "'"*'* " "

- O d < \ s,

g20 0

-e -f.P'

.)

2M 4

-- UNINNA01 ATED (OPEN POINTS)

O igo _ s.ss ions ,fe,2 (cLosto reints) i e

( 160 --

j .

4

140 - -

i

}120 0 l .

, E lm -

- s)

=

O l 3 80 -

L/* _ _ _ _?,

o -

a -

/ G aso rt. ts itme.. ioser 20 . aso FT. ts. itwe. = i so*r 0

0

-100 0 100 200 300 400 -

TD4PERATURE ('F) 1 Figure 5 4. Charpy V Notch Impact Data for the Point Beach Unit No. 2 Weld Heat Affected Zone Metal

\

~\ "

% +/

l j

)

13,028-9 l l 100 l 1 ,, @ - - - -O  ; e 80 2 j/3 n #

5 60

/(

'-  !" 2 0

3@

. b# j.A 2 ,

i j

100 -

~

h,oS ,sV

~

[ E 1320F g y - O

~

i 0 @@ /

160

== ==== UNIRRADI ATED (OPEN PolNTS )

lq _ 9.45 X 1038 N/CH2 (CLOSED PolNTS)

• * * * * *** 4. 74 X 1018 N/CH2 (8ATTELLE CURVE CAPSULE "V")

120 - C a* *" "

y#C o N e 7 100 -

/

,. S / *

'\

f so -

$0 ..

/

5 / '. '..,,

I 60 -

d -

rh 6 50 FT. LS. TEMP.

• lil *F w -

p o t

s '%

_/ .* 630 FT. LS. TEMP.

• 1050F 20 - *'

f .

O  !

} -100 0 100 200 300 WO TDIPERATURE(*F)

Figure 5-5. Charpy V Notch Impact Data for SA533 Grade B Class 1 ASTM Correlation Monitor Material 5-7

7__---_____-_____

t , I

~

TAILE 6-1 l i THE EFFECT OF 650*F IRRADIATION AT 9.46 x 1018 n/cm2 (E > 1 M0w) i l

ON THE NOTCH REACTOR VESSEL TOUOHNESS IMPACT TEST SPECIMENS PROPERTIES OF THE POINT BEAC I

! ' i I

Average Enerw Ahmeryden Teensielen Temp (*F) et Full meer litel Mneerial ladreadiated arradiated A Tone (*F) laderediesed arredissed & Energy 50 h4 38 h4 3E smile 54 fth 30 ftih 35 smile 50 h ah 30 feeb 35 seGe 123V500VA1 -40 -40 -75 -20 -50 -33 40 30 42 180 223 + 43.0 122WISEVA1 -ee -45 -37 -2 -28 520 13 17 17 145 150.5 + 5.5 Wald Metal 80 0 17 200 145 142 140 145 145 85 58 -0.0

. HAZ Metal -20 -80 -le 85 30 89 105 l

110 105 53.5 194 +110.5I *I 1122W195VA1) 435 de . Coerdelion 81 49 53 192 154 185 105 til 132 123.5 108.5 -15.0 Monisot l *

e. See sont for hah of HAZ test results.

I t

i

? -

"e 1

ai

i I

- '

i .

t , . - i

+

a

• 4 i .

s .. 1 N#

i ,

[

.. ..-  ; i i .  !

.l ..

TABLE 5 2

SUMMARY

OF POINT BEACH UNIT NO. 2 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTSI*I Trans. Temp Trans. Temp Decreens in Fluence increase (*T' increase (*F) Upper Shelf Metenal 1018 n/cm2 50 ft-lb 30 ft-Ib Energy (ftlb)

Forging 123V500VA1 9.45 40 30 NONE Forging 123V500VA1 4.74 35 30 NONE Forging 122W195VA1 9.45 13 17 NONE

( Forging 122W195VA1 4.74 15 10 10.0 Wold Metal 9.45 140 145 9.0 Weld Metal 4.74 -

165 23.0

' - ~ ' ~

~ MAZ Metal (122W195VA1)(b) 9.45 105 110 6

-HAZ- Metal (122W195VA1) 4.74 - -

Correlation Monitor 9.45 111 105 15.0 Correlation Monitor 4.74 110 90 30.0

a. Remsits et 4.74 a 1018 n/em2 oteaused (see feocesee 1. p. 2 23.

, tL See tout for stueussion of HAZ test results t

l..

G l

[

TABLE 5-3 CHARPY V. NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO. 2 PRESSURE VESSEL INTERMEDIATE FORGING 123V500VA1 IRRADIATED AT .

550*F, FLUENCE 9.46 x 1018 n/cm2 (E > 1 Mov) ~

Specimen Test Temp Energy Laterol Expension Sheer kmber (* F) (ft4b) (mils) (%)

V43 -90 7.5 7 1 V41 -60 32.0 26 10 V47 -10 82.0 64 30 V46 25 69.0 52 30 V46 35 96.5 68 40 ,)

V44 50 162.5 73 85

V37 70 198.0 80 100 V39 70 209.0 74 100 V40 100 168.0 86 100 V42 125 234.5 74 100 V48 150 201.0 76 100 V38 210 166.0 90 100 TABLE 54 CHARPY V-NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO. 2 PRESSURE VESSEL LOWER SHELL FORGING 122W195VA1 IRRADIATED AT .-

550*F, FLUENCE 9.45 x 1018 n/cm2 (E > 1 Mov)

Specimen Test Temp Energy Lateral Expansion Sheer k mber (*F) (ft4b) (mils) (%)

E39 -100 6.0 11 1 E41 -60 18.0 16 5 E44 -25 34.0 24 10 E46 -15 46.0 34 20 E37 0 75.0 48 30 E38 25 117.0 83 60 _

E43 40 66.0 50. 30

~

E40 50 98.0 74 50 .

E47 70 126.0 64 70 -

( E42 125 154.0 94 100 l

E45 150 152.0 94 100 E48 210 145.5 92 100 5 10

j i

Test results obtained on the weld metal are presented in figure 5-3 and table 5 5. Once again, I the results are compared to results from the previous capsule, V. The weld metal exhibited shifts in the 30 and 50-ft-Ib temperatures of 145*F and 140*F respectively. The shift in the 35 mil LE. temperature was identical to the 30 ft-Ib shift. The transition temperature shift for the weld metal at 9.45 x 1018 n/cm2 is less than than reported at 4.74 x 1018 n/cm2 fmm results of capsule V. An upper-shelf energy of 56 ft-Ib resulted from the capsule T tests, 9 ft lb lower than the values as unirradiated, but higher than the reported shelf tamperature from results of capsule V: after a fluence of 4.74 x 1018 n/cm2, the upper shelf energy for the capsule V weld metal fell below 50 ft lbs.

• The test results for the HAZ meterial are shown in figure 5-4 and table 5-6. Large scatter is v' shown in the test results, which appear to follow proporties characteristic of either the base or weld material as reflected in the extremely hi'hg and low values at upper-shelf temperatures.

The average 30 and 50 ft-lb shifts for the HAZ material are 110'F and 105*F respectively, which is more closely characteristic of the weld than the plate material. The shift in the 35-mil t

.J E._ temperature was identical to the 50 ft-Ib shift. The upper-shelf energy increased significantly, l which reflects average properties moni tin'::al of the base material than the weld. Test results I

of capsule V HAZ rnatorial displayed similar scatter.

. Figure 5 5 and table 5 7 present the test rest.:lts obtained on the A533 GRADE 8 Class 1 ASTM reference correlation monitor material. Shifts in the 30 and 50-ftlb temperatures are 105'F and 111*F respectively. The 35 mil LE. temperature shifted 132*F. The upper- l shelf impact energy decreased 15 ft-Ibn from unirradiated shelf energy of 123.5 ft-lb. [

( Charpy impact specimen fracture surfaces of the various Point Beach Unit No. 2 vessel material

! and correlation monitor metenal are shown in figures 5-6 through 510.

A summary of the Charpy impact test results for capsule T is given in table 5-1. Comparisen with results of Capsule V is made in table 5 2. Figure 5-11 gives a comparison of thoss l

30 ft-Ib transition temperature increases for capsules V and T with the Westinghouse-predicted l 30 ft-lb transition temperature increases. At a fluence of 9.45 x 1018 n/cm2 (E > 1 Mev),

none of the shifts in the 30 ft Ib transition temperature for tfw Point Beach surveillance {

l l meterial exceed the shifts predicted in the Westinghouse curves.

l Since the surveillance data indicated that the Westinghouse curves adequately predict the 30 ft-lb transition temperature increase, predicted adjusted reference temperatures based on

[ the Westinghouse curves will be used for future vessel analyses.

l l

l l

l l l l l 5 11 .

l

TABLE 5-5 CHARPY V NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO. 2 -

PRESSURE VESSEL WELD METAL IRRADIATED AT 550*F, FLUENCE 9.45 x 1018 n/cm2 (E > 1 Mov) .

Specimen Test Temp Energy Lateral Expansion Sheer Number (* F) (ft4b) (mils) (%)

W31 -50 5.0 6 1 W32 70 17.0 10 10 W26 125 33.0 32 35 ,

W28 30.0 175 22 50 ', j /

W25 210 30.0 30 60 W27 225 56.0 50 100 W29 275 56.0 42 100 W30 350 57.0 47 100 TABLE 5 6 CHARPY V-NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO. 2 -

PRESSURE VESSEL WELD-HEAT-AFFECTED-ZONE METAL IRRADIATED '

AT 550*F, FLUENCE 9.45 x 1018 n/cm2 (E > 1 Mov) .

Specimen Number Test Temp

(* F)

Energy (ft4b)

Lateral Expansion (mils)

Sheer

(%)

f H29 -50 4.5 3 0 H31 -10 11.0 16 20 H30 25 45.5 12 20 H25 7,0 166.0 86 100 H26 70 30.0 22 25 H27 125 52.0 34 50 H32 210 51.0 48 95 H28 300 194.0 86 100 O

5-12

. l TABLE 57 CHARPY V NOTCH IMPACT DATA FOR THE POINT BEACH ' UNIT NO. 2 ASTM SA533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL IRRADIATED 1 AT 550*F, FLUENCE 9.45 x 1018 n/cm2 (E > 1 Mov) {

i i

Specimen Test Temp Energy Lateral Expension Sheer i Number (*F) (ft4b) (mils) (%)

R30 70 7.0 3 5 R32 150 34.0 26 20 R26 200 45.0 34 30

/ R27 210 43.0 34 40 R31 250 88.0 62 70 R28 300 102.0 78 100 R25 300 116.0 84 100

. R29 425 108.0 68 100 9

9 f

l I

5-13

9 13028-16

.9

}

A 1 6 .

o [ .>

(+ri vu v

~T 5

v43 v41 v47 v4s v4s v44

>. -... ~

i m __ . L

- .v, 4

V37 V39 V40 V42 V48 V38 I

Figure 5-6. Charpy impact Specimen Fracture Surfaces for ,

Point Beach Unit No. 2 Pressure Vessel intermediate -

Shell Forging 123V500VA1

,'. 5-14

'A 4.

13028-15 l

, gM?. . e v~ ..

. e p ,- g: ,. m s .:-- -

. m' T ' 'o "i.  !

, .,m c f

"A j. ..

'%.w)M:  ? g.. = 5 E39 E41 E44 E46 E37 E38

'l

.i I

1 l

i l

i

. q m- y -n e m .

~ l

.,n'.,,.,  ?

..' 1

'.- t b*v 1

( h...

-~

w, .,

t A 1

.  ; .ts-l w- I

~

( l a

_- 4.

E43 240 E47 E42 E46 E48 .

9 Figure 5-7. Charpy impact Specimen Fracture Surfaces for Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 i

5-15 l

I 13C28-11 I

J

(

.- l l l 1

- *<.4 v ,

. m i

4 b

.. 17. , . . . .

l

, y% '4... s

. v.,

'\,

W31 W32 W26 W28 l

l l

l

= N -

l

~

l

5. _e.,t y, -

. . - . +

~;

._'_-. j l

W25 W27 W29 W30 ,

• b e

e Figure 5-8. Charpy impact Specimen Fracture Surfaces for .

Point Beach Unit No. 2 Weld Metal 0

9% e 5-16

t. ,

.. , ,1

. . _ , _ . . . . . . . - . _ . . _ _ . , . . . . . . .. [

13028 13 l

. <=

w

g. ... . _ .

.y I .

u 1 ,*

I H29 H31 H30 H25 I

l d

l 1 $

i  ?

I 1f,.

s

)

H26 H27 H32 H28 i

t t

t r

Figure 5-9. Charpy impact Specimen Fracture Surfaces for Point Beach Unit No. 2 Weid-Heat-Affected-Zone Metal l

6 l

5-17

13028 12 f

f i

!  ??.

i, m ..

4 a . . . E, -

3

. ~~ z

. bO.e I

-- .)

RM RM RM R27 i

Gk s.

T l ,~. -- . -

i s

R31 R28 R25 R29 l

Figure 5-10.

~

Charpy impact Specimen Fracture Surfaces for Point Beach -

Unit No. 2 ASTM Correlation Monitor Material

, 5-18

's.

4' L-- - - . . . . . - . - . . . _ ___.. . . . . . . . _ _ _ _ _ . _ . . .

s- *

• y ^m 2*

~

50. -

- UPPER LIMIT (NGCouTROLLED Cu)

__ - 4.30 WIS Cu WELD -

- - 0.2s Wi$ Cu WLS (0.30$ Cu BASE)

.i _ 0.20 W15 C WEta (0.265 Ce sASE) ,

u, - 0.ls WT5 Cu Wets (0.20$ Cu BASE) a20, _

,, - 0.10 WT,C. m <0.1.,C. . m )

_E.

u - 0.06 WT5 Cu WEtA (0.105 Cu BASE)

OR N0mlTOR a 100

/ - 0.065 C. u E Z W -

5 j- 50 -

is -

4

^

d 123V600VA1 20 -

A 122Wis6vAs so I I i lill I l l I l llll lois 2 5 10l e 2 5 1020 2

FLUEllCE (II/CH ), E > l Mrv

! Figure 6-11. Point Beach Unit No. 2 Material 30 ft-lb Transition Temperature i increases as Compared to Westinghouse Predictions A

- - - - , , - , - - - - . - . - - . ~ , - , _ , - - - _ _ _ , , , , - - , . - - --.-_..,nw.-,,w-w.--e-e-,-,--n~,. - , , - _..__,w_-,wm.--__.----s, .__-._~~_-_-_-__.,----n._..- . . _ _ _.__ _

i. '.. .

l Comparison of the shifts in the weld metal transition temperature from results of capsules "V" and "T" (figure 511) indicates that the rate of change of transition temperature with ,-

fluence is well below that predicted by the Westinghouse trend curve.

~

Considering that Charpy impact data in general exhibit scatter of i 25'F, similar behavior -

may be occurring in the correlation monitor and forging materials. The test results from the next Point Beach Unit No. 2 surveillance capsule will confirm this shift " saturation" behavior, as has been observed from results of the first three capsules removed from Point Beach ,

Unit No.1. N t 5-2. TENSILE TEST RESULTS The results of the tensile tests are presented in table 5-8 and figures 5-12 and 5-13. Tests 'iq were performed on specimens from forging 123VE00VA1 and 122W195VA1 at various /

temperatures from room temperature to 550*F. Irradiation resulted in minimal increases in the yield strengths of both forgings.

Photographs of the fractured tensile specimens at approximately actual size are shown in figures 514 and 5-15. A typical stress strain curve for the tensile tests is shown in figure 5-16.

l 5 3. WEDGE OPENING LOADING TESTS Wedge Opening Loading (WOL) fracture mechanics specimens which were contained in the .'.

surveillance capsule have been storeef at the Westinghouse Research Laboratory at the request .

of the Wisconsin Electric Power Company on the recommendation of the U. S. Nuclear Regulatory Commission; they will be tested and reported later. .

- ~ .

l

~.

1. Will be reported in fortftcoming WCAP-9lM7. Yenicnko and Anderson. \

s x, ,y '

TABLE 58 IRRADIATED TENSILE PROPERTIES FOR THE POINT BEACH UNIT NO. 2 ,

PRESSURE VESSEL SHELL FORGINGS Ultimate Test Yield Tensile Fracture Fracture Uniform Total Red.

Specimen Temp Strength Strength Load Stress Elong. Elong. In Area Forging No. Numher (* F) (psil (psi) (ib) (psi) (%) {%) (%) '

123V500VA1 V12 70 60,300 83,800 2520 204,900 11.9 24.8 75 V13 200 56,100 77,300 2360 197,600 10.7 25.1 76 VII 400 64,000 77,300 2540 193,900 10.3 23.1 73 V10 660 68,500 82,800 2710 189,600 9.9 22.7 71 122W196VA1 E13 70 82,800 103,300 2800 210,500 8.7 23.5 73 E14 200 72,800 92,000 2760 180,900 8.3 22.0 69 E11 300 68,500 87,900 2850 19'i,900 7.8 20.1 70 E10 400 67,100 89,000 2800 200,800 8.6 21.7 73 E12 560 G7,800 91,800 2830 196,200 8.7 22.0 70 1

s. '.. . .

13,028-5 100 l I I I I I .

ULTIMATE TEN 31LE STRENGTH O O 32 80 - O e O e O , -f)

- 0.2% YlELD STRENGTH w

=

0 e Ea -

M

• O e cr '. m O uNiar 40iATED (ALs0 sNOWN BY CNNECTille LINES) -

g e lasA01ATED. 9.45 X 10I8 N/CH2 . 5500F 80 g g REDUCTION IN AREA O- O e _

" _g O .

60 -

! =

l s.

=

TOTAL ELONGATION ~,

G e O 20 -

$- UNIFORM ELollGATION p v

e e v

! I I I l l

( 0 0 100 200 300 WO 500 600 TEMPERATURE (*F) _

Figure 5-12. Tensile Properties for the Point Beach Unit No. 2 Pressure -

Vessel Intermediate Shell Forging 123V500VA1 5-22 3

'^ c.l .

13,028-6 l

I I I I i

.I I 100 -

(

g ULTINATE TEN $tLE STRENGTH g

O le 8 6 "

U 80 -

g O 0.2% YlELD STRENGTH t - _

m

( O o

y#

v l l

O UNIRRADIATED (ALSO SHOWN SY C011NECTING LINES) {

$ IRRADIATED, 9.45 X 10I8 N/CM I

, 550*F I 40 N I g REDUCTION IN AREA g l U $ 7 O h

60 - i

('

~

$ I r yn -

l W i i

TOTAL ELONGATION E 20

- .4 1

@ UNIFORM ELONGATION g

. . T . .  ;

0 l f l I  ! I 1

0 100 200 300 400 500 600 l

TEMPERATURE (*F) l Figure 5-13. Tensile Properties for the Point Beach Unit No. 2 Pressure

' Vessel Lower Shell Forging 122W195VA1 l

5-23 l

i .

13028-11

-4 -%  !:?

' .bl . .-

t= w. . au.w i

! V12 70*F I

i I

L

{%&;bhjj((h,.W

• ~.

..c..<- , e q. ,' n.;;l::cf.:N.)ii-f'

- '; c: - -.v .j e.4' A

(g

Ne.w ;a Sw; .* ~j t _ + P. w(,j:.b,;y1 .cg: r:.

. i  :. .

wr. 2 V13 200* F O

se.-

V11 400 F r

EA.k!P .

,, g , d v.*$. $

,, e ~g m - ~

l l

V10 550 F Figure 5-14. Fractured Tensile Specimens from Point Beach Unit No. 2 -

Pressure Vessel Intermediate Shell Forging 123V500VA1 ~

5-24

l r

l 13028 10 f

l i

. .m tf E13 70* F cc - ,ggr.=. .-

,[%h$'$k$fh[')$'ISp'y *?

.. ?8 4

s

.p~

f ,

- Ar $ p.p p; :c x r.y'.tz * -%. '

s.? , j n *m. .Iy *a-+,.%

.*~~e.' .

( '. '4 s 1-ik.5&t'h tr.-sh ..-Q $(f d 1 ]' & 1 & E?

E14 200*F

_ _w gfe. , ,L -

.;-7*: t n.- :: . . * ,,

a 1

~l

s. J

>

• ix . *4.' .

- T- #

• a - . %F.w e, arQ pW ' i% 'f

., ' 4'%g&Q,i$$Jp%%h E11 300* F m -c-m: .:..

.m m..: g n r ~ : m I

( x.,t; .... c .sc@ : - ' '

t #/74t .-: Eq0'r'.7%it. .

?* h' M A c E -n c,c_.' . .. ..L . 2 E10 400 F n.. n .s

'ii,.@n%%%L,_h,@t'5T~M.'W.w ,

E12 550* F

. Figure 5-15. Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 l

l I

l 5-25 l

I - - - . ..

r .

100 80 -

1 1

60 -

_m n

w

+ W g

• 40 -

20 -

3 o l I l l l l l l I

O 0.03 0.06 0.09 0.12 0.15 0.18 0.21 0.24 0.27 STRAIN (IN./IN.) ,

i l C Figure 5-16. Typical Stress-Strain Curve for Tension Specimens *

(Tension Specimen No. E14) 0 4

s e'g .

l l

l8

• N:' ' . .'

'J '

4 ..  : ..

{

l l

, .. SECTION 6 I NEUTRON DOSIMETRY ANALYSIS l St. DESCRIPTION OF NEUTRON FLUX MONITORS To effect a correlation between neutron exposure and the radiation-induced property changes (

observed in the test specimens, a number of neutron flux monitors were included as an f'j integral part of the Reactor Vessel surveillance Program. The particular monitors contained O within Capsule T, along with the nuclear reaction of interest and the energy range of each '

monitor, are listed in table 61.

I The first five reactions listed in table 61 are used as fast neutron monitors to relate neutron I, fluence (E > 1.0 Mov) to the measured shift in RTNDT. To property account for bumout of

~

the product isotope generated by the fast neutron reactions, it is necessary to also determine l i

~N magnitude of the thermal neutron flux at the monitor location. Therefore, bare and  !

s cadmium covered cobalt-aluminum monitors were included within Capsule .T. i i

The rotative locations of the various monitors within Capsule T are shown in figure 4-3; the radial and azimuthal positions of the capsule with respect to the nuclear core, reactor internais

• l and pressure vessel are illustrated in. figure 61. The nickel, copper, and cobalt-aluminum

{

monitors (in wire form) were placed in holes drilled in spacers at several axial levels within j

k the capsule. The imn monitors were obtained by drilling semples from selected Charpy test specimens. The cadmiumshielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule. t t

The use of activation monitors, such as those listed in table 6-1, does not yield a direct j measure of the energy <iependent neutron flux level at the point of interest. Rather, the j activation process is a measure of the integrated effect that the time and energy-dependent i neutron flux has on the target material. An accurate estimate.of the average neutron flux i 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: {

. e The operating history of the reactor {

e The energy response of the monitor I a The nastron energy spectrum at the monitor location '

I e The physical cheistics of the monitor I 61  !

t i

TACLE 6-1 NEUTRON FLUX MONITORS CONTAINED WITHIN CAPSULE T WT% of Target 8" target Monitor MaterW. Reaction of intersat 8" monitor Response Range Product Half-Life f Copper Cu63 (n,a) Co60 0.6917 E > 4.7 Mov 5.27 years 1 Iron Fe54 (n.pl Mn54 0.0585 E > 1.0 Mew 314 days

/ Nickel NiS8 (n p) Co68 0.6777 E > 1.0 Mew 71.4 days i

I

! Uranium 238 *I U238 (n,f) Cs137 1.0 E > 0.4 Mew 30.2 years Neptunium23/*I Np237 (n,f) Cs137 1.0 E > 0.08 Mew 30.2 years Cobalt-aluminum I*I Co69 (n/y) Co60 0.0015 0.4ev < E < 0.015 Mew 5.27 years Cobalt-aluminum Co60 (ofy) Co60 0.0015 E < 0.015 Mew 5.27 years

a. Denotes that monitor is cadmium shielded

=

l 1

b f

w i

., f

.- l

.n i U .. . .

U .

l

~

. 10.1 W 20

. X f/////7 yo

/

Ehy gg *h ti i I, CAPSULE 7

Aq M' '

r , , s r , , s,

% ,o / ,

- /  ;

I

/ <

i , , ,& , , , , ,, '

/ /

/

/ _ /

REACTOR CORE

/ , , , , , , t

- l

/

~

\

/

/

( .

i

/

i

/

l

/

i

/

l l

l

/  !

/

/

~

1

/ \

\

i Figure 6-1. Point Beach Unit No. 2 Reactor Geometry l

l 63

4

62. ANALYTICAL PROCEDURES The analysis of the activation 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.soction, the neutron energy spectrum at the monitor location must be calculated. ,

~

The energy and spatial distribution of neutron flux within the Point. Beach Nuclear Plant

)

Unit No. 2 reactor geometry was obtained with the DOTUI two. dimensional Sn transport code. The radial and azimuthal distributions were obtained from an R,6 computation wherein the reactor core, reactor internals, surveillance capsule, water annuli, pressure vesasi, and primary

(

i shield concrete were described on the analytical model. These analyses employed 21 neutron \ .~ n y  ;

energy groups, an S8 angular quadrature, and a Pj cross-section expansion. The reactor core "

power distributions used in the calculations were representative of time everaged conditions  !

over an equilibrium fuel cycle; they accounted for rod-by-rod spatial variations in the peripheral i fuel assemblies. The analytical geometries described a 46* sector of the reactor, assuming one- l eighth symmetry, Relative axial variations of neutron flux incident on the reactor vessel were f

obtained from R,Z DOT calculations based on the equivalent cylindrical core concept.

(

The specific activity of each of the activation monitors was determined using established ASTM pror,edures.W341 Following sample preparation, the activity of each monitor was determined ,.' l by means of a lithium-drifted germanium Ge (Li) gamma spectrometer. The overall standard ,

dr,viation 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 s removed from Capsule T, the overall 2a deviation in all of the measured data was determined \ t j

to be .10 percent. '

1. Soltmar. R. G., oleney, R. K., Jedrush, J. and ZIesier, S. L, "Nuedeer Rocket Stieldine Meeods Modification, undedng

}

and input Dem Properation. Vol. 5 - Tsue oimonsonsf Diesrete ordinates Transport Technique," WANt..PR(LUQ34, Vol. 5, l Aveast 1970. }

I

2. ASTM Desipwdon E2et.70, stendard Momed far Mesmaring Neutron Flux by Radioactivetion Techniouse," in ASTM i Stenderde (19751, Port 46, Nussear Standards, pp. 748 756, Am. Society for Testing and Meteriais, Philedsephie, PA.,1975. j

3. ASTM Dessywoon E262 70, **Standere Method for Mansurine Thermal Neutron Plum by Radioactivetion Te=:hniopias/* In i ASTM Stenderes (1975), Part 46, Nucesor Standards, pp. 756 7e3 Am. Society for Testing and Meteriefs, Philadeepnie, PA.1975.

j

4. ASTM F : . _ :n E2eS.70, "Standere Meeod for Measuring Feat Neutron Flus bvKadioactivetion of Iron," in ASTM Stenderes (19751, Port 46, Nuesser Standards, pp. 7e4 7W. Arn. Society for Tesdag and Meteriais, Ph6tedelphie, PA,1975. i S. ASTM Doespia*Mn Edel.737, "Tantative Memod of Maomaring Neutron - Flux Doneity by Radioactivation of Cobedt and . f S8ver," in ASTM Standeres (1975), Part As, Nuclear Stendeeds, pp. 887.ses, Am. Society for Tesarg and Meteriais.

Philedstohle PA,1375. '

6. AsTMr'- ,_ - E2ed.70, " Standard Meeod for Mesmarins Feathoutron Flux by Radioactivetion of Nicksi." in ASTM .

Standards (1975), Part 48, Nucteer Standards, pp. 77o.774, Am. Society for Testing and Metensis, Philadeepnia, PA,1975. -

i f

t l

6-4 i 1

- . . . .- - _ _ _ .- _ .__ - _ ._ =.

~

. . . ~

l

- f Having the measured activity of the monitors and the neutron energy spectrum at the location j

of interest, the calculation of the fast neutron flux proceeded as follows. The reaction product t activity in the monitor was expressed as I

.- h

(

0 = f fg y a(E) ((E) (1 - e 1) e' d (61)

E j=1 max  ;

Wh*

l D = induced product activity J

3 No = Avogedro's number f

/  !

A = stomir; weight of the target isotope l f i = weight fraction of the target isotope in the target material y = number of product atoms produced per reaction a(E) = energy-dependent reaction crossaection

((E) = energy-depexiont neutron flux at the monitor location with '

the reactor at full power

/

Pj = average core power level during irradiation period j  ;

. t P

max = maximum or reference core power level  !

,l '

A = decay constant of thg product isotope  !

rj = length of irradiation period j  !

I rg = decey time following inadiation period j  !

Since neutron flux distributions were c.siculated by multigroup transport methods, and since the prime interest was in the fast notitron flux above 1 Mov, spectrum averaged reaction cross-sections were defined such that the irc.toyal term in equaticsn (6-1) could be replaced by the  !

following relation i

{

a(E)'d(E) = iTd (E > 1 Mov) I where n  !

a(5) ((E) ag (g j

v. .

1 0

G=G10 Mov i

r I

6. 5

l t

I .

I Thus, equation (61) was rewritten

( .

._. D = h fg y id (E > 1.0 Mev) - (1 - euff) e'Afd ,.

l J= 1 mu or, solving for the neutron flux - --- -

l 4 (E > 1.0 Mov) = D (6 2) fg y $pmax (1 - e #i) e' Td ,

r .-

t ~

The total fluence above 1 Mov was then given by n p'.

l 4 (E > 1.0 Mov) = $ (E > 1.0 Mov)[ p rj (6-3) j=1 max where n p. -

) [ p 'ax m fj = total effective full-power seconds of reactor operation j=1 up to the time of capsule removal.

An assessment of the thermal neutron flux levels within Capsule T was obtained from the bare and cadmium-covered Co# (n,7) Co60 data by means of cadmium ratios and the use of a 37 bam 2200 m/sec cross-section. Thus; -

D bare R-1)

R (

4th "

No n P .x pfg y a E p y -(1 - e q) e.Ard (6-4)

=1 max where R is defined as O berelDCd-covered-The irradiation history of the flux monitors removed from Capsule T is listed in table 6 2.

The data were obtained from the Point Beach semiannual operating reports.[1I The spectrum- .-

averaged reaction cross-sections derived for each of the fast neutron flux monitors are listed in table 6-3. ,

1. Point asseh Nusteer Unita 1 and 2 SamMamasi operating Reporte.1e70 throush 197s.

6-6 w - -

l . . . , ..

j , ,  ;

f

l TA8LE 62

,. IRRADIATION HISTORY OF CAPSULE "T" i r

p Irradiation DecoyI 'I  !

Pj P max Time Tim .

Month (MW) (MW) man (dsys) (dsys)  !

Prior to 12H2 389 1518 .243 92 1915 1/73 284 1518 .187 31 1884 2/73 278 1518 .183 28 1856 l 3/73 740 1518 .487 31 1825  ;

4/73 743 1518 .489 30 1795 -

[ ., 5/73 1043 1518 .687 31 1764 s i 6/73 1300 1518 .856 30 1734  !

7/73 1325 1518 .873 31 1703 l 8/73 1397 1518 .920 31 1672 9/73 1501 1518 .989 30 1642 ,

10/73 1474^ 1518 .971 31 1611  !

11/73 1479 1518 .974 30 1581  ;

12/73 1424 1518 .938 31 1550 g 1/74 1462 1518 .963 31 1519  ;

2/74 1458 1518 .960 28 -1491  !

3/74 1485 1518 .978 31 1460  !

?

4/74 1495 1518 .985 30 1430 i 5/74 1494 1518 .984 31 1399 ,

. 6/74 1229 1518 .810 30 1389  ;

7R4 1283 1518 .845 31 1338 I

(. 8/74 1494 1518 .984 31 1307 i I 9/74 1464 1518 .964 30 1277 C  :

10/74 750 1518 .494 31 1246  !

11/74 0 1518 0 30 1216  ;

12/74 406 1518 .267 31 1185  ;

1/75 1497 1518 .986 31 1154 r

2/75 1311 1518 .364 28 1126  !

3/75 1464 1518 .958 31 1095  !

4/75 1273 1518 .3 31 30 1065 5/75 1183 1518 .779 31 1034  :

6/75 1270 1518 .837 30 1004 7/75 1430 l

1518 .942 31 973  !

8/75 1006 1518 .663 31 942  !

. 9/75 1385 1518 .912 30 912  :

10/75 1330 ~ 1518 .876 31 881 11/75 1354 1518 .892 30 851 t 12/75 1498 1518 .987 31 820 l

b i

67

TABLE 6 2 (cont) 1RRADIATION HISTORY OF CAPSULE "T" -

Irradiation Decaybl Pj P max Ti m Time Momh (MW) (MW) max (deys) (days) 1/76 1439 1518 .948 31 789 2/76 1335 1518 .879 29 760 3/76 155 1518 .102 31 729 4n6 1351 1518 .890 30 699 5/76 1286 1518 .847 31 668 6US 1400 1518 .922 30 638 (?

706 1380 1518 .900 31 607 8n6 1374 1518 .905 31 576 9/76 1397 1518 .920 30~ 546

~

10/76 1500 1518 .988 31 _

515 11/76 1478 1518 .974 30 485 12n6 1482 1518 .976 31 454 1/77 1482 1518 .976 31 423 2/77 1391 1518 .916 28 395 3/77 135 1518 .089 31 364 .,

a n w u e to s uv7s

~

'(;

J e

e l

( ~- - _- - _ _ , - - , = - - = -

r I

TABLE &3 ,

1

" SPECTRUM AVERAGED REACTION CROSS SECTIONS  :

USED IN FAST NEUTRON FLUX DERIVATION l

=  ?

Reaction 7(borns)

{

t Fe54 (n,p) Mn5 Mal 0.064  !

Fe54 (n,p) Mn54M 0.066 NiS8 (n,p) CoS8 0.088

[ CuS3 (n,a) Co60 0.00040

\

l U238 (n,f) F.P. 0.34 i

\

Np237 (n,f) F.P. 2.9 i

a. AggHeside to sesvisies taken frorn coro esde Charpy specwnens  ;

ik Aeolientde to sesnesles taken frown pressure. weasel-side Charpy spenwnens

{

S3. RESULTS OF ANALYSIS -!

Tha fast neutron (E > 1 Mov) flux and fluence levels denved from the monitors taken from I Capsule T are presented in table 6 4. The results of the fast neutron dosimetry indicate that upon insertion into the reactor, Capsule T was rotated 180*. The subsequent ' discussion assumes

(

d that this rotation did in fact take place. The thermal neutron flux obtained from the cobalt-l aluminum monitors is summarized in table 6-5. Due to the relatively low thermal neutron [

flux at the capsule location, no bumup correction was made to any of the measured activities. [

The maximum error introduced by this assumption is estimated to be less than 1 percent for  !

the N 58 (n,p) CoS8 remedon and even less significant for all of the other fast reactions.  !

l Results of the Sntransport calculations for the Point Beach Unit No. 2 Reactor are summarized

in figures 6 2 througn 54 and in table 6-8. In figure 6-2, the calculated maximum fast neutron {

i flux levels at the pressure vessel inner radius,1/4 thickness location and 3/4-thickness location >

are presented as a function of azimuthal angle. The relative axial variation of neutron flux is shown in figure 6 3. Absolute axial variations of fast neutron flux may be obtained by r multiplying the favels igiven in figure S2 by the appropriate values from figure 6-3. In [

figure 6 4, the calculated maximum end-of-life fast neutron exposure of the Point Beach Reactor l

Vessel is given as a function of radial position within the vessel well. The calculated fast  !

neutron flux levels interior to Capsule T along with the lead factors (LF) relating caosule  !

exposture to vessel sixposure are listed in table 6-6. The lead factor is defined as the ratio of 4

the calculated flux at the monitor location to the calculated peak neutmn flux incident on the I reactor Vessel.

&9

TABLE 64 ,

RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE T .

~

Mamured*I I  !

Reaction and Activity 4 (E > 1 Mov)[bl 4 (E > 1 Mov)Ibl Monitor Locadon (dpe/gm) (n/cm2-esc) (n/cm2) l Fe54 (n,p) Mn54 l Top-center E47I 'l 1.44 x 108 9.80 x 1010 1.07 x 1019 l Bottom V37{c] 1,41 x 106 9.59 x 1010 1.05 x 1019 i Center E37[e] 1.38 x 106 9.39 x 1010 1.02 x 1019 Top-center W32 Idl

.] ,

1.17 x 106 7.72 x 1010 8.41 x 1018 ,j  !

Bottom H25Idl 1.23 x 106 8.12 x 1010 8.85 x 1018 Center R30 Idl 1.11 x 106 7.32 x 1010 7.98 x 1018 l NiS8 (n.p) CoS8 Center 1.02 x 106 7.85 x 1010 8.56 x 1018 CuS3 (n,od Co60 Bottom 1.03 x 105 1.04 x 1011 1.13 x 1019 Bottom-Center 9.50 x 104 9.64 x 1010 1.05 x 1019 _

Top-Center 8.23 x 104 8.34 x 1010 9.09 x 1018 Top 9.11 x 104 9.24 x 101k 1.01 x 1019 .

Np237 (n,f) Cs137 Center 2.83 x 106 8.10 x 1010 8.83 x 1018 U238 (n,f) Cs137 Center 3.69 x 105 8.69 x 1010 9.47 x 1018

e. Activities we reteensed to sesons

b. Derived Flux and Fluence Levels are nabiect to a 10 percent rnoomsrernent error

c. Co o side

d. Veend side m

6-10

~ '

I

, TA8LE 64 i

.. RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE T Bare AcdvityW Cd Covered AcevityW Monitor Lacedon (des /yn) 2 (des /em) de (n/cm . 3N

~' ~~~ ~'

To p 2.01 x 107 1.01 x 107 5.80 x 1010 t Center 2.09 x 107  :

1.07 x 107 5.90 x 1010 Bottom Conter 2.62 x 107 1.17 x 107 8.41 x 1010 Bottom 2.48 x 107 1.12 x 107 7.87 x 1010 j i

[,- a. Ce# Asuvottee are schronsed to 3/30/78

( . Ik Dortwed thsa levels are asWest to a 10 posesnt enessurment error {

I i.

I i r

f t

j ,4 ' k i

\.

I 1

i f

f i

i f

i I

k

~

.. p i

i

?

s.11 i

' 10,194-21 .

108 '

~

8 6

4 s

_2 o

PRES 3uaE VESSEL l.R. \

W s s

y 1/4T LOCATION x 10 80 -

S -

8 E -

m 6 -

.W .,

i 4

3/4T Location

_ f~ ,,

2 -

l in s l' I l l' l

0 10 20 30 4 50 I AZIMITHAL ANGLE (*) I i

1 *

. f Figure 6 2. Calculated Azimuthal Distribution of Maximum Fast.

l Neutron Flux (E > 1.0 Mev) Within the Point .  !

Beach Unit No. 2 Reactor Vesel j

[

6 12

[

e 10.1 M-22 l l capsuu carsvu -

BOTTOM TOP r

-- 1.0 -

- I

g -

[

I 6  !

4 -

l t

f i

2 -

l c

{

' ) = 0.1 f

s

• 8 g 6 t

as g _ ,

W -

. 4 E

3 i W O  !

2 - i

,.. 3 k

E at 0.01 58

- E i 8

\ . - =$ t

== j 6 -

Iz ou i UR

• 4 -

gL $k  !

!!E

=a L

gS 5  !

2 -

og =

i TO VE3SEL i CLOSURE NUD 1 l

0.001 l l l -l  !

i 300 -200 -100 0 100 200 300 l

DISTAllCE FROM CORE MIDPLAltE (CN) l l Figure &3. Relative Axial Variation of Fast Neutron Flux l (E > 1.0 Mov) incicient on the Point Smh  !

Unit No. 2 Reactor Vessel '

I S13 '

i

e 10,194-23 RA01US(lN) .

66 67 68 69 70 71 72 73 ~.

1020

-l l l l l l l l .

~

8

_167.64 CN (64.0 IN. )

4 171.77 CM (67.625 lit. )

VESSEL 10 -"

2 y l

{ l/4T E

l 8 -

gg 180.02 CH (70.875 111.)

8 .

~

g .

184.15 CM W6 -

(72.50 111.) ~.

3/4T

, _ N

.- J VESSEL 00 2

i ini8 I I I I I I I 166 168 170 172 174 176 178 180 182 184 186 .

RADIUS (CM) ,~

Figure S4. Calculated Maximum End of Life Fast Neutron Fluence *

(E > 1.0 Mov) as a Function of Radius Within the Point Beach Unit No. 2 Reactor Vessel

&14 L

TABLE $8

• CALCULATED FAST NEUTRON FLUX AND LEAD FACTORS .

FOR CAPSULE T

, l i

Locetion Within Capsule T 4 (E > 1 Mov) (n/cm 24ec) Lead Factor j Front Charpy 8.21 x 1010 jo {

Dosimeter Block and Flux Wires 6.89 x 1010 1.76 i Back Charpy 8.50 x 1010 1.86  !

{

g. i1 *

\

[

i

. l f -- i t' l

(. .

i i

I l

e 9

&15

.n. - .. ._ _ -. - ._ -

Based on the iron data presented in table 6-4, the average fast neutron fluence incident on the front mw of Charpy specimens is determined to be 1.06 x 1019 n/cm2, while that on the -

back mw of specimens is 8.41 x 1018 n/cm2. These measured values correspond to analytical ,

values of 8.95 x 1018 and 7.09 x 1018 n/cm2, respectively. A comparison of these values -

shows the calculations to be 17- to 19-percent low. Deviations of this order of magnitude are not considered unreasonable. The average measured fluence is 9.46 x 1018 n/cm 2, With the use of the lead factors listed in table 6-6, a comoarison of the end-oflife peak fast neutron exposure of the Point Beach reactors as derived from both calculations and measured surveillance capsule results may be made as follows:

FAST NEUTRON FLUENCE (n/cm2) t

-T Vessel Based on Iron Based on Iron Lacedon Calculeted From Front Charpys From Back Charpys inner surface 3.9 x 1018 4.6 x 1018 4.7 x 1018 .

l 1/4 Thickness 2.4 x 1018 2.9 x 1018 2.9 x 1018

, 3/4 Thickness 7.5 x 1018 8.9 x 1018 9.1 x 1018 j o

i s a

.._J.; l i

I l

l 1

I l

'. l 6-16 l

~

' ~

, ~~~ ~ ~---"~"