ML20027A315

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

ML20027A315
Person / Time
Site:	Point Beach
Issue date:	08/31/1978
From:	Shaun Anderson, Davidson J, Shogan R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20027A314	List: ML20027A312 ML20027A315 ML20027A316
References
EIZP-102, WCAP-9331, NUDOCS 7811140131
Download: ML20027A315 (57)

Text

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8 I WESTINGHOUSE CLASS 3 d CUSTOMER DESIGNATED DISTRIBUTION B

ANALYSIS OF CAPSULE R FROM THE WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT UNIT NO.1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko S. L Anderson August 1978 O

/ APPROVED: s s%

J. N.\Chirigos,Tianager T

, Structural Materials Engineering Prepared by Westingnouse for the Wisconsin Electric Power Company r Work Performed Under E12P-101 i

Althcugh the information contained in this report is nonproprietary, no cistribution shall be made outside Westinghouse or its Licensees without the l

{ customer's approval.

1 WESTINGHOUSE ELECTRIC CORPORATION I Nuclear Energy Systems P. O. Box 356 l Pittsburgh, Pennsylvanic 15230 dP////y'g/g 60 - 2. (.36 r3

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TABLE OF CONTENTS Section Title 1

SUMMARY

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2 INTRODUCTION 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 41 l 5 TESTING OF SPECIMENS FROM CAPSULE R 5-1 51 Charpy V Notch Impact Test Results 52 52 Tensile Test Results , 5-4 5-3 Wedge Opening Loading Tests 5-4 6 NEUTRON DOSIMETRY ANALYSIS 6-1 6-1 Description of Neutron Flux Monitors 6-1

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( 6-2 A.1alytical Procedures 64 6-3 Results of Analysis 6-7 6

/ 3 . .r  !

'. I LIST OF ILLUSTRATIONS j i

t Figure Title Page 41 Arrangement of Surveillance Capsules in the Point Beach

{

Unit No.1 Reactor Vessel (Land Factors for the Capsules Shown in Parentheses) 3

( 42 Capsule R Schematic Diagram Showing Designed Arrangement, -

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j of Specimens, Thermal Monitors, and Dosimeter Placement and i

Orientation With Respect to the Core and Vessel- Wall 4 7/4-8 l 5-1 Charpy V Notch impact Data for the Point Beach Unit  !

No.1 Pressure Vessel Shell Plate A9811 5-5 i 5-2 Charpy V Notch Impact Data for the Point Beach Unit No.1 Pressure Vessel Shell Plate C1423 f

5-6 5-3 Charpy V Notch Impact Data for the Point Beach Unit

.- No.1 Pressure Vessel Weld Metal 5-7 ,

5-4 Charpy V-Notch impact Data for the Point Beach Unit I

• No.1 Pressure Vessel Weld-Heat-Affected Zone Metal 5-8 -

5 Charpy V-Notch Impact Data for SA302 Grade B ASTM j Correlation Monitor Material 5-9  !

56 Charpy impact Specimen Fracture Surfaces for Point  !

f. Beach Unit No.1 Pressure Vessel Shell Plate A9811 5 10 5-7 l Charpy impact Specimen Fracture Surfaces for Point  !

Beach Unit No.1 Pressure Vessel Shell Plate C1423 5-11 l

5-8 Charpy impact Specimen Fracture Surfaces for Point Beach Unit No.1 Weld Metal 5-9 Charpy impact Specimen Fracture Surfaces for Point 5-12 l Beach Unit No.1 Weld Heat Affected Zone Metal 5-13 5 10 Charpy impact Specimen Fracture Surfaces for Point i Beach Unit No.1 ASTM Correlation Monitor Material 5-14

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5-11 Point Beach Unit No.1 Material 30 FT LB Transition  !

Temperature increases as Compared to Westingnouse I Predictions 5-15 [

5-12 Tensile Properties for the Point Beach Unit No.1 +

Pressure Vessel Weld Metal 5-16

, 5-13 Tensile Properties for the Point Beach Unit No.1 Pressure Vessel Shell Plate A9811 l 5-17 5-14 i

, Tensile Properties for the Point Beach Unit No.1 Pressure Vessel Shell Plate C1423 5 18 i

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LIST OF ILLUSTRATIONS (cont)

Figure ' Title Page 5-15 Fractumd Tensile Specimens From Point Beach Unit No.1 Weld Metal 5 19 5 16 Fractured Tensile Specimens From Point Beach -

rN Unit No.1 Pressure Vessel Shell Plate A9811 5-17 Fractured Tensile Specimens From Point Beach 5-20 _)

Unit No.1 Pressure Vessel Shell Plate C1423 5-21 518 Typical Stress Strain Curve for Tension Specimens (Tension Specimen No. A7) 5-22 6-1 Point Beach Unit No.1 Reactor Geometry &2 42 Calculated Aaimuthat Distribution of Maximum Fast Neutron Flux (E > 1 Mov) Within the Point Beech .-

Unit No.1 Reactor Vessel &9 6-3 Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) incident on the Point Beach Unit No.1 Reactor Vesset 6-10 S4 Calculated Maximum End-of Life Fast Neutron Fluence (E > 1 Mew) as a Function of Radius Within the Point Beach Unit No.1 Reactor Vessei 6-It V

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. t LIST OF TABLES fi I

Table Title - Page.

4-1. Chemistry and Heat Treatment of Material Representing  ;

A s the Core Region Shell Plates and. Weld Metal From the

( Point Beach Unit No.1 Reactor Vessel 4-4 i

42 Chemistry and Heat Ireatment of Surveillance Material

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Representing 6-inch-Thick A3028 ASTM Correlation  !

Monitor Material 4-5 i 5-1 Charpy V-Notch Impact Data for the Point Beach Unit No.1 Pressure Vessel Shell Plate A9811 Irradiated at  !

550*F, Fluence 2.22 x 1019 n/cm2 (E > 1 Mev) 5-23 '

5-2 Charpy V-Notch Impact Data for the Point Beach Unit  ;

No.1 Pressure Vessel Shell Plate C1423 Irradiated at  ;

550*F, Fluence 2.22 x 1019 n/cm2 (E > 1 Mev) 5-23 i 5-3 Charpy V Notch Impact Data for the Point Beach Unit No.1 Pressure V i Weld Metal Irradiated at 550 F, Fluence 2.22 x 101 n/cm2 (E > 1 Mov) 5-24 L 5-4 Charpy V Notch Impact Data for the Point Beach Unit No.1 Pressure Vessel Weld-Heat Affected-Zone Metal (

Irradiated at 550*F, Fluence 2.22 x 10.19 n/cm2 l (E > 1 Mev) 5 24 t 55 Charpy V Notch Impact Data for the Point Beach Unit f

No.1 ASTM SA302 Grade B Correlation Monitor Material f irradiated at 550*F, Fluence 2.22 x 1019 n/cm2 i (E > 1 Mev) 5-25 5-6 The Effect of 550*F Irradiation at 2.22 x 1019 n/cm2 (E > 1 Mev) on the Notch Toughness Properties of the  !

Point Beach Unit No.1 Reactor Vessel Impact Test

. Specimens 5-26 ,

57 Summary of Point Beach Unit No.1 Reactor Vessel I Surveillance Capsule Charpy impact Test Results 5-27 58 Irradiated Tensile Properties for the Point Beach Unit

, No.1 Pressure Vessel Materials 5-28 61 Neutron Flux Monitors Contained Within Capsule R 6-3 6-2 Irradiation History of Capsule R 6-12 6-3 Spectrum Averaged Reaction Cross Sections Used in Fast Neutron Flux Derivation 6-15 ,

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l LIST OF TABLES (cont)

Tatdo Title ,

Page 6-4 Remits of Fast Neutron Dosimetry for Capsule R 6 16 6-5 Results of Thermal Neutron Dosimetry for Capsule R 6-17 .m 6-6 Calculated Fast Neutron Flux. and Lead Factors for Capsule R s 6-17 m eB aP*

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P SECTION 1

SUMMARY

The analysis;which compared unirradiated with irradiated material properties of the reactor vessel material contained in the third surveillance capsule, designated R, from the Wisconsin

, Electric Power Company Point Beach. Nuclear P.lant Unit No. I reactor pressure vessel led to

\ the following conclusions:

a The capsule' received an average fast fluence of 2.22 x 1019 n/cm2 (E > 1 Mev). The predicted fast fluence for the capsule was 1.80 x 1019 n/cm2 (E > 1 Mev).

e The fast fluence of 2.22 x 1019 n/cm2 resulted in a 205 F increase in the 50 ft Ib reference nil-ductility transition temperature (RTNDT) of the

~ weld metal, which is representative of the most limiting rnaterial in the core region of the reactor vessel. The intermediate pressure vessel snen plate A9811 and lower shell plate C1423 exhibited a 50 ft Ib , transition

~- temperature increase of 105*F and 50*F, respectively (specimens oriented parallel to the rolling direction of the plates). The weld heat affected zone material exhibited a 50 ft Ib transition increase of 60*F.

a The average upper shelf impact energy of.the weld metal decreased from '

65 to 51 ft-lb.

(

e An increase of 110 F in the 30 ftIb transition temperature was determined for the ASTM A3028 reference correlation monitor material contained in the capsule. ,

a Transition temperature increases for the surveillance materials were essentially -

the same as those obtainee from the second capsule irradiated to 7.05 x 1018 l n/cm2, indicating that a limiting or steady state value has been reached at a '

level well below predicted trend curves.

e The following end-oflife 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 sunteillance capsule results, were determined.

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2 l Fast Neutron Fluence (n/cm ) _

Vessel Locetion Calculsted Measured l Inner surface 3.9 x 1018 4.90 x 1018 1/4 Thicknese 2.4 x 1018 3.05 x 1018 l 3/4 Thickness 7.5 x 1018 9.50 x 1018 l

The difference of approximstefy 25 percent in the calculated versus the measured fast neutron fluence is due in part to the 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 approximation and differences between the -

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actual core spatial power distribution in the peripheral fuel assemblies and that assumed in the analysis.

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SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule R, the third capsule of the continuing surveillance program, which monitors the effects of neutron irradiation on the

, Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No.1 reactor pressure

( vessel materials under actual operating conditions.

The surveillance program for the Point Beach Unit No.1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in WCAP-7513.III The surveillance program, which was 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. -[2]

Post irradiation data have been obtained from the third material surveillance capsule (Capsule R) removed from the Point Beach Unit No.1 reactor vessel. This report summarizes the tests and the results, and discusses the analysis of the data.

t. Yanichko. S. E.. 'Wsconsin Machigan Power Co. Point Beach Unit No.1 Reector Vessel Radiation Survesitance Program." WCAP 7513.1970.

2. ASTM Desqnstion E185-66. "Surve,itance Tests on Structurai Matenais in Nuclear Reacters.** in ASTM Standards itE671. Part 31. Phys cal and Mechanica Testing of Metals - Me'sitograoliv. Nondestruct:ve Test ng. Fatigue.

Effect of Te nos ature, po. 638442. Am. Soc. for Testing and Materiais. Phiieceinhia. Pa 1967.

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

([ vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA302 Grade B (base material of the Unit No. I reactor pressure vessel beltline) are well-documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain condi-tions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has f

been presented in " Protection Against Non-ductile Failure," Appendix G, to Section lli of l

, the ASME Boiler and Pressure Vessel Code. The method, utilizing fracture mechanics concepts, is based on the reference nil-ductility temperature, RT NOT-RTNDT si defined as the greater of the drop weight nil-ductility transition temperature f (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 .

direction of the material. The RTNDT of a given material is used to index that material to a reference stress-intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness ,

results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress-intensity factors can be obtained for this material as a l function of temperature. Allowable operating limits can then be determined baseo on these allowab!e stress intensity factors.

RTNDT, 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-i ment or changes in mechanical properties of a given reactor pressure vessel steel can be i 1

i 3-1

monitored by a reactor surveillance program such as the Point Beach Nuclear Plant Unit No.1

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Reactor Vessel Radiation Surveillance Program,N in 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 '

original RTNOT to adjust the RT NDT or f radiation embrittlement. This adjusted RTNDT (RT i

~ NDT 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 ef*ect of irraciation on the reactor vessel materials.

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t. Yenienko. S. E " Wisconsin Michigan Power Co. Point Seech unit No.1 Reactor Vesse+ Reciecon Survedience -'

Propern. WCAA7513.1970.

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SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Point Beach Nuclear Plant Unit No. I reactor pressure vessel core region material were inserted in the

, reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

Capsule R was removed in the autumn of 1977 after approximately 7 calendar years (5.1 effective full-power years) of plant operation. This capsule eontained Charpy V notch impact, tensile, and WOL specimens (shown in WCAP-7513) from the intermediate and lower shell ring plates, weld metal representative of the core region of the reactor vessel, and Charpy V-notch specimens from weld-heat-affected zone (HAZ) material. The capsule also contained Charpy V-notch specimens from the 6-inch thick ASTM correlation monitor material (A302 Grade B) furnished by the U. S. Steel Corporation. The chemistry and heat treament of the surveillance material is presented in tables 41 and 4 2.

All test specimens were machined from the 1/4 thickness location of the plates. Test speci-mens represent material taken at least one plate thickness from the quenched end of the plate. All base metal Charpy V notch and tensile specimens were oriented with the longitu-dinal axis of the specimen parallel to the principal rolling direction of the plates. The WOL test specimens were machined with the simulated crack of the specimen perpendicular to the surfaces and rolling direction of the plates.

Charpy V-notch specimens from the we!d metal were oriented with the longitudinal axis of the specimens transverse to the weld direction. Tensile specimens were oriented with the longitudinal axis of the speciment parallel to the weld.

Capsule R contained dosimeter wires of copper, nickel, and aluminum-cobalt (caomium-shielded and unshielded), in addition, the capsule contained cadmium-snielded dosimeters of Np 237 and U238, located as shown in figure 4-2.

b 41

Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were placed in the capsule, located as shown in figure 4-2. The two eutectic alloys and their melting points are:

2 2.5% Ag, 97.5% Pb Melting Point 579'F 1.75% Ag, 0.75 % Sn, 97.5% Pb Melting. Point 590*F l

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i Figure 4-1. Arrangement of Surveillance Capsules in the Point Beach Unit No.1 Reactor Vessel (Lead Factors for the Capstles l Shown in Parentheses) l t

4-3  ;

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TABLE 41 CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE CORE REGION SHELL PLATES AND WELD METAL FROM THE POINT BEACH UNIT NO.1 REACTOR VESSEL ,

CHEMICAL ANALYSES (PERCENT)

Element Plate A9811 Plate C1423 Weld Metal C 0.19 0.21 0.09 Mn 1.42 1.37 1.47 P 0.010 0.014 0.019 .

S 0.020

, 0.019 0.024 Si 0.25 0.25 0.49 Mo -0.48 0.46 0.39 Cu 0.19 0.11 0.18 0.24 Ni - -

0.57 Cr -

0.13 Al --

0.035 N2 -

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0.016 V -

0.001 Sn -

~ 0.004 Ti -

0.001 As -

0.004 Co -

0.001l81 .

Zr --

0.001[a] .

Sb -

0.001[al Zn .- -

0.001[al 8 -

0.003[a1 HEAT TREATMENT e

Plate A9811 Heated at 16ro F, 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, water-quenched Tempered at 1225*F, 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, aircooled Stress relieved at 1125 F,111/4 hours, furnace-cooled Plate C1423 Heated at 1650F, 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, water-quenched Tempered at 1225*F, 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, aircooled Stress-relieved at 1125*F,10-1/2 hours, furnace-cooled -

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Not detecte: the ournber indicates the mansmum limit of detectiori. ,

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, TABLE 4-2 i CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL -

REPRESENTING 6 INCH THICK A3028 ASTM CORRELATION '

MONITOR MATERIAL '

CHEMICAL ANALYSIS (PERCENT) j C Mn P S -Mo Si Cu Ni Cr j 0.24 1.34 0.011 0.023 0.51 0.23 0.20 0.18 0.11 HEAT TREATMENT I The 6-inch-thick plate was charged into a fumace operating at 110*F heated at a i maximum rate of 63*F per hour to 1650*F, held at temperature for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and f

water-quenched to 300*F. The plate was then redarged into a furnace operating at 700* to  !

750*F, heated at a maximum rate of 63*F per hour to 1200 F, and held at that temperature f for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. '

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SECTION 5 TESTING OF SPECIMENS FROM CAPSULE R The' post-irradiation mechanical testing of the Charpy V notch and tensile specimens was (

performed ~at the Westinghouse Research and Development Laboratory with consultation by

, Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with

( 10CFR50, Appendices G and H.

Upon ' receipt of the capsule at the laboratory, the specimens and spacer blocks were car: fully  !

removed, inspected for identification number, and checked against the master list in WCAP-7513.N No discrepancies were found.

Examination of the two low-melting (579;F and 590'F) eutectic alloys indicated no melting

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

, A Tinius Olsen Model 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 tests on the irradiated Charpy-V specimens, the accuracy of the impact machine was 6

checked with a set of standard specimens obtained from the Army Material and Mechanics Research Center in Watertown, Massachusetts. The results of the calibration testing shcwed <

that the machine was certified for Charpy V notch impact testing.

The tensile tests were conducted on a screw-driven instron testing machine of 20,000 lb .

capacity per ASTM E8-69, " Tension Testing of Metallic Materials" and ASTM E2170,

" Elevated Temperature Tension Tests of Metallic Materials." The crosshead speed was 0.5 inch per minute. The defornaation of the specimen was measured with a strain gage extensometer, t The extensometer was calibrated before testmg with a Sheffield high magnification drum-type extensometer calibrator.

Elevated-temperature tensile tests were conducted in a split tube furnace. The specimens were a

held at temperature a minimum of 20 minutes to stabilize the temperature prior to testing.

Temperature was monitored with a chromel alumel thermocouple in contact with the clevis pin-type upper and lower specimen grips. Temperature was controlled within :3*F. '

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Yan;cnko, s. E., 'Wsconen weugan Dewer Co. Pcint Seacn unit No.1 Reacto- Vessei Racianon Sune re4ce Program." WCAP 7513,1970.

5-1

The load-extension data were recorded on the testing machine strip chart. The yield strength, ultimate tensile strength, and uniform elongation were determined from these charts. The #

reduction in area and total elongation were determined from specimen measurements.

51. CHARPY V-NOTCH IMPACT TEST RESULTS The irradiated Charpy V notch specimens represented the Point Beach Nuclear Plant Unit No.1 reactor pressure vessel beltline p! ate material, weld and heat-affected zone (HAZ) material, and the ASTM reference correlatien monitor material. The results are presented in figures 5-1

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through 5-5. Table 5-6 summarizes the increase in the 30 and 50 ft ib energy and 35-mil --

lateral expansion transition temperature, and the decrease in the upper shelf energy ruulting from irradiation to 2.22 x 1019 n/cm 2, --

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Figures 51 and 5-2 give the test results obtained on the vessel beltline shell plate material.

Tables 5-1 and 5-2 show a 30 and 50 ft Ib transition temperature increase of 105*F for plate A9811 and 50*F for plate C1423. Plate A9811 showed a 115*F increase in the 35-mit lateral expansion temperature which was in reasonable agreement with the 105*F increase in the 50 ft-lb transition temperature. Plate C1423 showed a 60 F increase in the 35-mil lateral expansion temperature as compared to a 50*F increase in the 50 ft-Ib transition temperature.

Plate A9811 exhibited a 7 ft-lb or 7 percent decrease in upper shelf energy while plate C1423 showed a 15 ft-Ib or 12 percent increase in upper shelf energy.

Figure 5 3 and table 5-3 give the test results obtained on the weld metal. These results show that, in 30 and 50 ft-Ib testing, the weld metal exhibited respective increases of 165*F and 205*F in transition temperature. The 200*F increase in the 35-mil lateral expansion tempera-ture was nearly identical to the 205 F increase in the 50 ft Ib transition temperature. The upper shelf energy of the weld metal decreased 14 ft lb or 21 percent.

The test results for the HAZ material are shown in table 5-4 and figure 5-4. A 70*F and 60*F transition temperature increase was obtained at the 30 and 50 ft-Ib temperatures, rsspectively. The 35-mil lateral expansion temperature increased 90*F as comparert to a 60*F increase in-the 50 ft Ib transition temperature. The upper shelf of the HAZ decreased 23 ft-Ib ,

or 17 percent.

Figure 5-5 and table 5-5 present the test results obtained on the A3028 ASTM reference correlation monitor material. Respective increases of 110* and 123*F in the 30 and 50 ft-Ib transition temperatures were obtained for this material. The upper shelf energy of the correla-tion monitor material increased 5 ft Ib or 6 percent. -

! Charpy impact specimen fracture surfaces of the various Point Beach Unit No.1 vessel material and the correlation monitor material are presented in figures 5-6 through 5-10. 2 5-2

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Table 5 7 summarizes the Charpy impact test results for the first and se<:ond capsules (L2] and the third capsule. The results show that, as a result of additional irradiation from 7.05 x 1018 to 2.22 x 1019 nicm2, (1) essentially no additional transition temperature increase resulted for plate C1423, the weld metal, and the weld HAZ, and (2) only a 15'F increaae resulted for plate A9811 and the correlation monitor material. A comparison of the 30 ft-lb transition '

temperature increases with predicted temperature increasies (figure 511) indicates that long-time irradiations (5 to 7 years) received by the second and third capsules do not result in as much radiation damage as predicted from trend curves developed primarily from experimental data

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from short time irradiations (< 1 year) and first time surveillance capsule removals (< 2 years).

The 110*F increase in the 30 ft Ib transition temperature of the correlation monitor material  !

, - irradiated to 2.22 x 1018 niem2falls below the trend band established for tnis materialW

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. thus also seeming to indicate that'ir' radiation-induced increases in transition temperature reach a limiting or steady state value well below predicted trend curves, ,

i The weld metal used in the surveillance program is representative of but not identical with any of the core region welds in the Point Beach Unit No.1 vessel W . This is because the particular heat of weld wire and lot of flux used to fabricate the surveillance weldment differs from that used in fabricating the vessel core region welds. A recent weld metal study by

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Westinghouse establishes that the surveillance weldment for the Florida Power and Light Co.

Turkey Point Unit No. 3 surveillance program was fabricated using the same heat of weld .

r wire and tot of flux as that used to fabricate the core region girth weld of the Point Beach Unit No. I vessel. Post irradiation tests performed on the Turkey Point No. 3 surveillance weidIM resultad in respective increases of 155* and 190*F in the 30 and 50 ftlb transition temperatures after irradiation to 5.68 x 1018 niem 2, The fact that these increases are in agreement with those obtained on the Point Beach Unit No.1 surveillance weld, and that the weld metal is the controlling vessel material for normal heatup and cooldown operations of the plant, gave a basis for constructing limit curves '

for those operations.W 'The curves, applicable for up to 10 Effective Full Power Years (EFPY),

and prepared with the use of Westingnouse trend curves for adjusting the reference transitiCn l temperature, are considered adequate as bounds for continued safe operation of the plant.

I 1.

8e rin..J. S., et al.. Final Report on "Pomt Beach Nuclear Plant unit No 1 Pressure Vessei Surved:ancer Prog m: a  ;

Eveauation of Capsule v." to Wisconon Electric Power Company by Battette Laboratories. Columous. oNo. June 197* '

2. Yanichko. S. E. and Anderson S. L.."Anolyses of Capsule S from the Wisconsm Eiectric Power Company and Wisconsm M enigan Power Cornoeny Point Seecn Nuclear P' ant unit No.1 Reactor Vesset Radiat:on Surve.Ilance Program.*-  !

WCAP-873e.1976.

3. ASTM OS 54, Ra4ation Effects Information Generated on tne ASTM Reference Correistion Monitor Steels. ASTM.

P'HaoCoh.a.1974 4

P*lhot. J. H.. "Heatup and Cocidown Limit Corves for tne Wisconsin Electric Power Comcany and Wisconsin M cnigan

, Power Company Point Beecn unit No. I Nucisar Power Plant." WcAP 8743,1976.

j

5. Yanichko S. E., et as . Anaiysis of Caosu'e T from tne Flos.de Power and Lynt Company TuArv Pomt unit No. 3 t Reac:or vesse' Raciat?on Survedlance Program /* WCAP-8631,1975.

5-3

' ^ ~

f Because the surveillance data indicate that the Westinghouse curves conservatively predict the -

30 ft-lb transition temperature increase, predicted adjusted reference temperatures based on -

the use of Westinghouse curves will be applied in various vessel analyses.

5 2. TENSILE TEST RESULTS Table 5 8 and figures 512 through 5-14 give the results of the tensile test. The weld metal and plates A9811 and C1423 were tested at a range of temperatures bounded by the ambient and 550*F. Increases in yield strength caused by irradiation were as follows:

Increase in Yield Strength From -

_ . . _ .' 1rradiation at Several Tomoeratures (ksi) ,

Ambient, ,

Material 200, & 400*F 550# F Ambient 300 F Plate A9911 >5 Plate C1523 -20 30 Weld Metal 25 20 These increases in yield strength are essential!y the same as those resulting from irradiation .

at 7.05 x 1018 n/cm2, thus also indicating that a limiting or steady state condition has been reached. -

Photographs of the fractured tensile specimens are shown in figures 5-15 through 517. A typic'al stress strain curve for the tensile tests is shown in figure 518.

5-3. WEDGE OPENING LOADING TESTS Wedge Opening Loading (WOL) fracture mechanics specimens which were contained in the surveillance capsule have been stored, on the recommendation of the U.S. Nuclear Regulatory Commission and at the request of the Wisconsin Electric Power Company, at the Westinghouse R&D Center. They will be tested and the results reported at a later date.

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Figure 51. Charpy V-Notch Impact Data for the Point Beach  !

Unit No.1 Pressure Vessel Shell Plate A9811 55

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TEMPERATURE (oF) i Figure 5-2. Charpy V-Notch Impact Data for the Point Beach '

Unit No.1 Pressure Vessel Shell Plate C1423 5-6 i i

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r TEMPERATURE (F) \

Figure 5-3. Charpy V Notch Impact Data for the Point Beach Unit No.1 Pressure Vessel 5Neid Metal

> 5-7 1

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TEMPERATURE ("F)

Figure 5-4. Charpy V-Notch Impact Data for the Point Beach -

Unit No.1 Weld Heat Affected Zone Metal

(

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Figure 5-5. Cnarpy V-Notch impact Data for SA302 Grade 8 ASTM Correlation Monitor Material 5-9 ,

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i Figure 5-6. Charpy impact Specimen Fracture Surfaces for i j

Point Beach Unit No.1 Pressure Vessel Shell . ,

Plate A9811 -

(

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\

l_- _ _____

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. Figure 5-7. Charpy impact Specimen Fracture Surfaces for '

Point Beach Unit No.1 Pressure Vessel Shell i Plate C1423 t l

5-11 I

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i Figure 5 8. Charpy impact Specimen Fracture Surfaces for .

Point Beach Unit No.1 Weld Metal ~

5-12

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80 ULTIMATE TENSILE STRENGTH E

Si 60 -

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1 0 I I I 'I I I 0 100 200 300 400 500 600 TEMPERATURE ( F) ~

Figure 5-12. Tensile Properties for the Point Beach Unit No.1 -

Pressure Vessel Weld Metal 5-16

i

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, TEMPERATURE (F) i l 1

l. i Figure 5-13. Tensile Properties for the Point Beach Uni: No.1 Pressure Vessel Sheil Plate A9811  !

i 5-17  !

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ETIMATE TENSILE STRENGTH E

g 0.2' YlELD STRENGTH G

40 -

d C UNIRRADIATED A $ IRRADIATED (5500F - 2.22 x 1089 N/CM2 )

20 80 A

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REDUCTION IN AREA 7

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$ TsTAL ELONGATION 8

20 -

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- _ _ g _ ,,,,,,, ,

! O l i I I I I O 100 200 300 400 500 600 700 TEMPERATURE (F) _

Figure 5-14. Tensile Properties for the Point Beacn Unit No.1 '

Pressure Vessel Shell Plate C1423 1

5-18

l l

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l 10, l M-2 6 1

l l l

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Figure 515. Fractureo Tensile Specimens From Point Beach Unit No.1 Weld Metal l

5 19 .

l

. .: :l : ,

10, l M-27 i

s A6 70 F -

A7 300*F i

A5 550'F Figure 5-16. Fractured Tensile Specimens From Point Beach -

Unit No.1 Pressure Vessel Shell Plate A9811

• i l

l 5-20 ,

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1 C6 550*F Figure 517. Fractured Tensile Specirr. ens From Point Beach l

Unit No.1 Presure Vessel Shell Plate C1423 5-21 i

90,000 -

70.000 -

C K'

[ 50.000 -

l3 9 a:

t 0 $ '

30.000 -

10.000 -

I I I I I I I -

0 .03 .06 .09 .12 .15 .18 .21 . 2 84

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I,, STRAIN (INCH / INCH) ,5 .

a Figure 5-18. Typical Stress-Strain Curve for Tension Specimens (Tension Specimen No. A7) [ ,,

=

.( .

I

. e TABLE 5-1 CHARPY V. NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO.1 ,

PRESSURE VESSEL SHELL PLATE p9811 1RRADIATED AT 550*F, FLUENCE 2.22 x 101 n/cm2 (E > 1 Mev)

Specimen Test Temp Energy Lateral Fxpension Sheer Number (* F) (ft.lb) (mils) 8%)

A22 0 7.5 4 5 A18~ -

40 24.0 25 20 A13 70 28.0 20 20 A23 70 49.0 32 35 A16 90 61.0 50 50 A14 11C 77.0 62 60 A21 110 49.0 39 30 A19 125 74.0 57 50 A17 150 97.0 66 90 A15 210 102.0 70 100 A24 210 95.0 64 100 A20 250 104.0 74 100 TABLE 52 CHARPY V. NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO.1 PRESSURE VESSEL SHELL PLATE C1423 IRRADIATED AT 560 F, FLUENCE 2.22 x 1019 n/cm2 (E > 1 Mov)

Specimen Test . Temp Energy Lateral Expansion Shear Number (* F) (ft.ib) (mils) (%)

C23 0 28.0 18 5 C23 0 36.0' 26 10-C19 25 34.0 25 10 C15 40 48.0 34 20 C21 40 47.0 32 20 C18 70 82.0 57

. 50 C13 70 70.0 40 30 C20 110 61.0 46 45 C14 .150 106.0 68 70 C16 210 135.0 84 100 C17 210 137.0

- 88 100 C24 300 128.0 7R 100 l

l 5-23 1

r .

. .  : .' e : s l . .' -

TABLE 53 "

l CHARPY V-NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO.1 '

PRESSURE VESSEL WELD METAL IRRADIATED AT 550*F, .

FLUENCE 2.22 x 1019 n/cm2 (E > 1 Mov) ,

Specimen Test Temp Energy Lateral Expension Sheer Number (*F) (ft-lb) (mils) (%)

WW11 30 11.5 13 20 WW15 70 24.0 16 30 WW10 125 31.0 24 40 WW16 150 38.0 30 60 WW14 210 47.0 40 98 -

WW9 250 50.0 46 100 WW13 300 49.0 42 100

~

WW12 300 58.0 48 100 TABLE 5-4 -'

CHARPY V-NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO.1 PRESSURE VESSEL WELD-HEAT AFFECTED ZONE METAL IRRADIATED AT 550*F, FLUENCE 2.22 x 1019 n/cm2 (E > 1 Mov)

Specimen Test Temp Energy Lateral Expansion Sheer Number ( F) (ft lb) (mils) (%)

WH11 0 34.0 28 30 WH10 40 84.0 63 90 WH15 40 13.0 8 20 WH16 70 118.0 68 90 WH14 70 45.0 28 60 WH9 100 117.0 82 100 WH13 150 112.0 75 100 WH12 210 105.0 76 100 5-24

4 *.

f i

L i

TABLE 5-5

- CHARPY V NOTCH IMPACT DATA FOR THE POINT BEACH UNIT NO.1 ASTM SA302 GRADE B CORRELATION MONITOR MATERIAL IRRADIATED AT 560*F, FLUENCE 2.22 x 1019 n/cm2 (E > 1 Mov) i Specimen Test Temp Energy Lateral Expension Sheer '

Number (* F) (ft lb) (mils)

(%)  ;

RIO 70 9.0 4 '

10 R13 150 26.0 21 30 i R15 160 36.0 27 40 R9 175 36.0 28 40 -

R11 210 60.5 42

(

.' 85  !

R12 250 50.0 42 80  !

R14 300 87.0 58 100 R16 300 80.0 50 100

[

i i

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l i

i l

I f

5-25

=

v

Pf ,

TABLE 5-6 THE EFFECT OF 560 F 1RRADIATION AT 2.22 x 1018 n/cm2 (E > 1 Mev)

ON THE NOTCH TOUGHNESS PROPERTIES OF THE POINT BEACH UNIT NO.1 REACTOR VESSEL IMPACT TEST SPECIMENS Transstion Tenip l'FI Aversee Ensesy ?"-

A Temp ("F) et Full Stener dit hl. _ _ .

Meterial IJnwredeseed lasedeseed Useirredesenst arredented r Energy 60 fa Ils 30 ie les 35snds te f a als 38 ie it- 36 mais 50 ft les 30 ftIh 35 sinals A9818 -10 -46 -30 95 60 85 106 106 116 107 100 7 Cl423 0 -30 -12 60 20 48 50 50 60 119 134 +15 Watki Weal 0 -45 -18 205 120 182' 206 166 700 66 58 14 f f AZ Wo.s 0 -6G -40 60 to 50 60 70 90 135

( A9811l 112 23 9 Corrulation 72 40 60 196 150 190 123

$ Maturist 110 140 78 83 e5 i

O 9

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• O e

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i .

l

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l f

TABLE 5 7

SUMMARY

OF POINT BEACH UNIT NO.1 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS e

.. i i i Trans. Temp Trans. Temp  :

Fluence increase (*F) Increase (*F) Upper Shelf i

_ . _ . _Meterial 1018 n/cm2 30 ft Ib 50 ft Ib Decrease (ft Ib)  :

Plate A9811 3.50 90 90 18

. Plate A9811 7.05 90 90 15 h i

Plate A9811 22.20 105 105 7 i t

i Plate C1423 3.50 50 50 None Plate C1423 7.05 50 50 None [

t Plate C1423 22.20 50 50 None  ;

Weld Metal 3.50 110 140 12 Weld Metal 7.05 165 200 13 Weld Metal 22.20 165 205 14 1 I

HAZ Metal (A9811) 3.50 70 25 28 . I I

HAZ Metal (A9811) 7.05 70 60 25 L r

HAZ Metal (A9811) 22.20 70 60 23 i Correlation Monitor 3.50 95 100 14 i

Correlation Monitor 7.05 95 105 10 '

Correlation Monitor 22.20 110 123 None -

f I

i 5-27

TABLE 5-8 IRRADIATED TENSILE PROPERTIES FOR THE POINT BEACH UNIT NO.1 PRESSURE VESSEL MATERIALS 0.2% Ultimate Test Yield Tensile Fracture Fracture Fractwo Uniform Total Reduction Meterial Specismen Temp Strength Strength Load Stress Elong.

iden.

Strenesh Elone. In Area Number (*F) (ksil (keil Db) (ksil (ksi) (%) (%) (%)

A9811 A6 70 69.5 90.0 3250 181.6 M.5 13.4 24 S 60 A7 300 62 2 86.1 3250 1663 66 5 11 2 21.4 59 AS 550 58.3 88.7 3500 137.8 71.6 11D 18 2 48 C1423 C5 70 75.0 95.1 3200 195.1 66.4 11.6 '

24.0 66 C7 200 71.6 89.8 2860 159.8 58 5 95 22.1 63 i

C4 400 67.5 88 2 3110 161.1 63 6 . 9.6 20.7 61 l C6 550 66.1 91.0 3200 188.2 66.4 9.4 20 5 65

! tn y Wekt WW3, 70 94.9 108.6 3980 203.1 81.4 10 2 212 60 Metal WW4 300 83 8 99.0 3700 165.2 75.7 10 5 20 9 54 l

l. ..

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l l

SECTION 6 NEUTRON DOSIMETRY ANALYSIS

1. DESCRIPTION OF NEUTRON FLUX MONITORS To effect a correlation between neutron exposure and the radiation induced property changes s observed in the test specimens, a number of neutron flux monitors were included as an  ;

integral part of the Reactor Vessel Surveillance Program. Table 6-1 lists the particular monitors contained within Capsule R, along with the nuclear reaction of interest and the energy range  !

of _each monitor.

The first five reactions listed in table 6-1 are used as fast neutron monitors to relate neutron fluence (E > 1.0 Mev) to the measured shift in RTNDT. To properly account for burnout of the product isotope generated by the fast neutron reactions, it is necessary to also determine  ;

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

cadmium-covered cobalt aluminum monitors were included within Capsule R.

Figure 4-2 shows the relative locations of the various monitors within Capsule R. Figure 61 shows the radial and azimuthal positions of the capsule with respect to the nuclear core, reactor internals, and pressure vessel. The nickel, copper, and cobalt-aluminum monitors (in

( wire form) were placed in holes drilled in spacers at several axial levels within the capsule. The iron monitors were obtained by drilling samples from selected Charpy test specimens. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the ,

dosimeter block located near the center of the capsule.

The use of activation monitors, such as those listed in taole 6-1, does not yield a direct measure of the energy-dependent neutron flux level at the point of interest. Rather, ti.e  ;

ac

• tivation process is a measure of the integrated effect that the' time and energy-dependent "

neutron flux has on the target material. An accurate estimate of the average neutron flux level incident on the various monitors may be derived from the activation measurements only  ;

if the irradiation parameters are well known. In particular, the following variables are '

of interest:

a The operating history of the reactor a The energy response of the monitor t

6-1

.'~ s',. ..

10.194-20 X .

13' -

f/////

lll//// P Rgg, VESggg CA PSULE

'///// R 1//ss, Y

NERy g 45*

4gq f//////y,

1. r

/

/

I /

/

f

/ s s s s s s s s, I / .

/

/ -

REACTOR CORE

/

/ 3

/

/

/

/

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

_.._m_._.___

t TABLE 6-1 ,

i NEUTRON FLUX MONITORS CONTAINED WITHIN CAPSULE R . .

Wt% of Target W aarget

Reaction of Monitor Material laterest 9" monitor Response Range PreAd Half-Life .

Copper Cusa (n.n) Co60 0.6917 E > 4.7 Mew 5.27 years Irno Fe64 (n.pl Mn 64 0.0585 E > 1.0 Mew 314 days flickel Niss (n.p) Co6 " 0.6777 E > 1.0 Mev 71.4 days Uranium 2as s. U23a (n,f) Cs'3' 1.0 E > 0.4 Mev 30.2 years (Jeptunium33'l'l Np23i(n.f) Cs'3' 1.0 E > 0.3U Mew 30.2 years

$ Cobalt aluminum l*8 Co68 (n,A) Co60 0.0015 0.4ev ' E : 0.015 Mew 5.27 years Cobalt-aluminum Co68 (n,A) Co8 0.001L E < 0.015 Mew 5.27 years

.a Gmissususi sha kl si smmueoss.

s The neutron energy spectrum at the monitor location .

a The physical characteristics of the monitor

• 6 2. ANALYTICAL PROCEDURES 1 The analysis of the activation monitors and subsequent derivation of the average neutron flux requires completion of two procadures. First, the disintegration rate of product isotope per unit mass of monitor must be determined. Second, in order to define a suitable spectrum-averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.

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

Unit No. I 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 J

the reactor core, reactor intemals, surveillance capsule, water annuli, pressure vessel, and primary shield concrete were described on the analytical model. These analyses employed 21 neutron energy groups, an S8 angular quadrature, and a Pj cross-section expansion. The reactor core power distributions used in the calculations, which were representative of time-1 averaged conditions over an equilibrium fuel cycle, accounted for rod-by-rod spatial variations .

in the peripheral fuel assemblies. The analytical geometries described a 45* sector of the

reactor, assuming one-eighth symmetry. Relative axial variations of neutron flux incident on

~,

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

The specific activity of each of the activation monitors was determined in accordance with established ASTlW procedures.i2.3.4.5.61 Following sample preparation, the activity of each monitor was determined by means of a lithium drifted germanium Ge (Li) gamma spectrom.

eter. The overall standard deviation of the measured data is a function of the precision of

1. Sosten. R. G Disney, R. K., Jedruch, J. and Ziegser, S. L " Nuclear Rocket Shielding Metnoes. Modification, uudeting and input Date Properation. Vol. 5 - Two oimenseones. Discrete ordinates Transport Technique." WANL-PR(LL)034 Voi. 5. August 1970.

4

2. ASTM Desegnation E26170, " Standard Method for Messunne Neutron Flus by Radioactivation Tecnneques." in ASTM l

Standeres (1975), Part 45. Nucseer Stenderes, pp. 745-756. Am. Society for Testing and Motoriess. Pnitadosen.e, Pe.1975.

3. ASTM Dessenerion E262 70. *.Stendard Method for Measuring Thermet Neutron Flus by Radioactivation Techniques." m ASTM Stenderos (1975). Part 45. Mucteer Stendards, pp. 756-783. Am. Society for Testing and Materiais.

Philadelphie Pe.1e75.

4. ASTM Designation E263 70. ** Standard Meined for Messering Fest. Neutron Flux by Radioactiverion of iron," in ASTM sienderde (197s). Port 48. Nuclear standeres, pp. 764 79. Am. Society for Testmg eno Meteriets. Philometonia. Pe 1975.

5. ASTM Desegnation Edel 737. " Tentative Metnod of Meesuring Neutron - Flum Denssey by Radioactivation of Cobeet and .

Salver " in ASTM Standeres (1975). Port 46. Nucteer Standards. pp. 8s7 se4, Am. Society for Testing and Meteness.

Philadesphie Pa.1975. -

6. ASTM Desegnation E264 70. " Standard Meened for Measurmg Fest Neutron Flus by Radioactiverson of Nicket." in ASTM -

Stenderes (1975). Port 46. Nucteer Standeros. pp. 770 774 Am. Society for Testmg and Mete *iess. Phiteoetonia. Pa.1975. .

1' h'

f= '

- - - - - - - - ' " - - - ~' ' - -

. .- - .- =

- *' I

.' i .. . .

I '

sample weighing, the uncertainty in counting, and the acceptable error in detector calibration.  !

For the samples removed from Capsule R, the overall 20 deviation in all of the measured  ;

data was Jetermined to be :10 percent. I Having the measured activity of the monitors and the neutron energr spectrum at the location of interest, the calculation of the fast neutron flux proceeded as follows. The reaction product f

[

activity in the monitor was expressed as 2

I " I

D =k f i y a(E) 6(E) (1 - e -ArIle Ard JE j= 1

[*P.**

(6-1) ;

i

, 't where i D = induced product activity L No = Avogadro's number A =

atomic weight of the target isotope q

=

f; weight fraction of the target isotope in the target material  ;

, y =

number of product atoms produced per reaction a(E) = energy-dependent reaction cross section c(E) =

energy-dependent neutron flux at the monitor location with the i reactor at full power

=

i Pj average core power level during irradiation period j C/. ,

= I P maximum or reference core power level max i

=

A decay constant of the product isotope

= i fj length of irradiation period j i

=

rg decay time following irradiation period j i

Since neutron flux distributions were calculated by means of multigroup transport methods, '

and further, since the prime interest was in the fast neutron flux above 1 Mev, spectrum- ,

averaged reaction cross sections were defined such that the integral term in equation (61) '

could be replaced by the following relation I

o(E) o!E) = oo (E > 1 Mev)  !

E i

65

where

,=

n o(E) c(E) ogog

,, so ,G=1 '

= n p(E) cg f1.0 MevG=G .0 Mev 1 Thus, equation (6-1) was rewritten P- Ar-D = 0 ' # i y Ec (E > 1.0 Mev) I -Ard A

p -(-e le or, solving for the neutron flux c (E > 1.0 Mev) = (6 2)

No P; p f; y a p (1 e Arj) e -Ard max The total fluence above 1 Mev was then given by -

s n .-

+ (E > 1.0 Mev) = c (E > 1.0 Mev) p rj (6-3) m.-

where ~

n ~

P; p r total effective full power seconds of reactor operation up to j.j max the time of capsule removal An assessment of the thermal neutron flux levels within Capsule R was obtained from the bare and cadmium-covered Co38 (n,A) Co80 data by means of cadmium ratios and the use of a 37 barn 2200 m/sec cross section. Thus, R-I cth

• n IO'4I No - P; p f; y o e -Arj) e-Ard p (1 -

max 33 ._

where R is defined as Dbare/DCd-covered- -

6-6

n ,_ - .,

g.

'S . , ,

i The irradiation history of the flux monitors removed from Capsule R is listed in table 6-2.  ;

The data were obtained from the Point Beach semi annual operating reports.III The spectrum- I c averaged reaction cross sections derived for each of the fast neutron flux monitors are listed .

in table 6-3. '

r 6 3. RESULTS OF ANALYSIS Table 6-4 lists the fast neutron (E > 1 Mov) flux and fluence leve!s derived from the monitors )

taken from Capsule R. Table 6-5 summarizes the thermal neutron flux obtained from the  !

cobalt-aluminum monitors. Due to the relatively low thermal neutron flux at the capsule i 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 Nisa (n, p) f Co sa reaction and even less significant for all of the other fast reactions.

Figures 6-2 through 6-4 and table 6 6 summarize results of the Sn transport calculations for the Point Beach Unit No.1 Reactor, in figure 6-2, the calculated maximum fast neutron flux l levels at the pressure vessel inner radius,1/4 thickness location, and 3/4 thickness location are ,

presented as a function of azimuthal angle. Figure 6-3 shows the relative axial variation of  ;

, neutron flux. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6-2 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 Vessel is

^

[

given as a function of radial position within the vessel wall. Table 6 6 lists the calculated fast l

neutron flux levels interior to Capsule R along with the lead factors (LF) relating capsule exposure to vessel exposure. The lead factor is defined as the ratio of the calculated flux at the monitor location to the calculated peak neutron flux incident on the reactor vessel.

~ '

Based on the iron data in table 6-4, the average fast neutron fluence incident on the front row of Charpy specimens is determined to be 2.43 x 1019 n/cm2, while that on the back row of the specimens is 2.02 x 1019 n/cm2. These measured values correspond to analytical values of '

2 1.99 x 1019 and 1.60 x 1019 n/cm , respectively. A comparison of these values shows the  :

calculations to be 22 to 26 percent low.

i f

I. Poent Beech Nuclear umts 1 and 2 Semi Annues operating Reports 1970 enrough 1977.

6-7  !

. , - i i 1 l With the use of the lead factors listed in table 6-6, a comparison of the end of life peak fast l

neutron exposure of the Point Beach reactors as derived from both calculations and measured -

surveillance capsule results may be made as follows:

_ }

l FAST NEUTRON FLUENCE (n/cm2)

! Vesssi Sased on iron Based on Iron Location Calculated From Front Charpys From Back Charpys

~

l Inner Surface 3.9 x 1019 4.8 x 1019 5.0 x 1019 1/4 Thickness 2.4 x 1019 3.0 x 1019 3.1 x 1019 3/4 Thickness 7.5 x 1018 9.3 x 1018 9.7 x 1018 ,

These data are based on 32 full power years of operation at 1518 Mw.

Based on the results of the tests performed to date, the following recommended removal schedule for the remaining surveillance capsules, developed by the Wisconsin Electric Power t

Company, has been approved by the NRC.

Capsule Identification Lead Factor '

Removal Date -

T 1.6 Fall of 1985 ..

P 1.6 Fall of 1989 N 1.4 Standby The actual removal dates should correspond to normal plant refueling and/or major plant shutdowns for the year identified, -

D i

l j

t

~.

l .

I 6-8

~ ~ -

- _: - _-- _. _ _ _ T_ __-*

h r

, 10,194-21 l t

i 4

10 t 8

6 r

- r 4 -

l 1

t 3 - i

&' PRESSURE VESSEL 1.R.

W.

I  !

a 1/47 LOCAT10N

~

x 10 80 -

s. 3 -

h8 i

ac ,

E 6 w -

=

4 -

3/4T LocaTsoM t

2 -

l 10 8  !  !  ! l 0 10 20 30  !

40 50 i

AZiWTHAL ANGLE ( )

Figure 6 2. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Witnin the Point Beach Unit No.1 Reactor Vessel '

i 6-9 i

10.194-22 l l CAPSULE CAPSULE -

80TTOM TOP l.0 -

s 6

4 -

~

2 m 0.l -

3 m

e 5 6 -

G . _.

W -

w 4

_m. -e W -

O .

2 =

a b

5 ind 0.01 23

=,,

a 6 I" vu 4 ~ uE =

3 2,J O  !

k da

-e _

c 2 - g3 3" uW ,

4 TO VESSEL '

CLO!URE HEAD

~ 0.001 300 200- ~ -100 0 100 200 300  !

DISTANCEFROMCOREMIDPLANE(CN)

Figure 6-3. Relative Axial Variation of Fast Neutron Flux '-

(E > 1.0 Mev) incident on the Point Beach -

~

Unit No.1 Reactor Vessel '

i 5

t 6-10

. .- . - ~ ~ .. . .

e . , ,

• - l 10.194-23 i I

.e RADIUS (IN) ,

66 67 68 69 70 71 72 gg20 73 I I l l l l l l 8

_ 167.64 cW (66.0 IN.)

g -

, 171.77 CM (67.62s IN.)

i VESSEL ID

~

2

- l

{ I/4T 5  !

6 iso.o2 cM (7o.s7s in.) t l10's m -

3 a

8 -

184.1s CM ti 6 (72.50 in.)

3/4T 4 -

r 1

VESSEL 00 2

i i

lois l l l l l l l l l 166 168 170 172 174 176 178 180 182 184 186 RADIUS (CM) ,

e Figure 6-4.

Calculated Maximum Enr1 of Life Fast Neutron Fluence (E > 1.0 Mev) as a Function of Racius Within the  :

Nint Beach Unit No.1 Reactor Vesse! >

i 6-11

, , s' s' >

TABLE 6-2 o

IRRADIATION HISTORY OF CAPSULE R '

P P a max trradiation Time Decay Timel 'I Month (MW) (MW) PPjf max (days) (days) 1100 247 1518 0.163 30 2695 1290 610 1518 0.402 31 2664 1/71 971 1518 0.639 31 2633 2H1 978 1518 0.643 28 2605 3/71 1249 1518 0.821 31 4/71 2574 609 1518 0.401 30 2544 SD1 1114 1518 0.733 31 6/71 2513 1233 1518 0.811 30 7/71 2483 1254 1518 0.825 31 891 2452 1372 1518 0.903 31 9/71 2421 1238 1518 0.815 30 10/71 2391 1299 1518 0.855 31 11/71 2360 1071 1518 0.705 30 12/71 2330 1317 1518 0.867 31 2299 4 1/72 6/72 1393 1518 0.916 182 7/72 2117 1375 1518 C.905 31 8/72 2086 -'

1485' 1518 0.977 31 2055 902 1296 1518 0.853 30 10921292 2025 0 1518 0.000 92

~ 1933 1/73-2/73 0 1518 0.000 59 3n3 1874 854 1518 0.562 31 1843 4n3 1084 1518 0.713 30

~

Sn3 1813 1092 1518 0.718 31 6/73 1782 1236 1518 0.813 30 1752 7/73 1439 1518 0.947 31 1721 8/73 1345 1518 0.885 31 9n3 1690 1401 1518 0.922 30 1660 a.

Oscay time is referenced to the counting date of the Fe, Ni, Ca. and Co monitors (4117/78): the No and U momtors were counted on 4/25/78.

• , e e . , - . =+

=

e e

t 4

6-12

__-____________-__n-

,., + -

\'

e TABLE 6-2 (cont)  ;

IRRADIATION HISTORY OF CAPSULE R  !

,, i P P a max irradiation Time Decay Time I*I I Month (W) (MW) Pj/Pmax (days) (days) I 1003 1275 1518 0.839 31 1629 11/73 1136 1518 0.747 30 f

1599  !

12/73 1268 1518 0.834 31 1568  !

• t In4 1360 1518 0.895 31 1537

,' i

' 2/74 1449 1518 0.953 28 1509 I 3/74 1481 1518 0.975 i

' -' 404 260 1518 0.171 31 1478 l 30 1448  !

5/74 0 1518 0.000 31 1417 .

6/74 937 1518 0.616 30 1387

... . 7/74 1451 1518 0.955 31 1356

[

8/74 1385 1518 0.911 31 1325 9/74 1379 1518 0.907 30 1295  ;

10/74 1394 1518 0.917 31 '

1264

, _ _ . . . 11/74 1303 1518 0.857 30 1234 l 12/74 1482 1518 0.975 31

, 1203 1/75 671 '1518 0.442 31 1172  !

, 2/75 1334 1518 0.878 28 1144 3/75  !

0 1518 0.000 31 1113  !

4/75 1136 1518 0.748 30 ' 1083 5/75 1481 1518 0.974 31 1052 l

6/75 1414 l 1518 0.930 30 1022  !

, 7/75 1488 1518 0.979 31 991  !

8/75 v 1488 1518 0.979 31 960 .j 905 1455 1518 '.957 30 930 10/75 1479 i 1518 0.973 31 899 i

11/75 1425 1518 0.937 16 883  !

.. cece, time . reverenced to tne couniin, date of tne =e. Ne, Cu. and Co monitort (4/17/78h tne No and U l monerors were counted on 4/2508. ,

t i

P 6-13 i

, , s- s'c. .

TABLE 6-2 (cont)  ! ,

IRRADIATION HISTORY OF CAPSULE R '

P; P max frradiation Time Decay Time w (g) (g) PJ/Pmax (days) (Jays) 1105-1205 0 1518 0 45 838 1R6 974 1518 0.642 31 807 2/76 1448 1518 0.954 29 778 396 1473 1518 0.970 31 747 4n6 1465 1518 0.965 30 717 Sn6 1395 1518 0.919 31 686 6/76 1468 1518 0.967 30 656 796 1472 1518 0.970 31 625 8/76 1479 1518 0.974 31 594 9n6 1495 1518 0.985 30 564 1096 36 1518 0.024 31 533 1196 190 1518 0.125 30 503 12n6 1346 1518 0.887 31 472 107 1485 1518 0.978 31 441

• 2/77 1404 1518 0.925 28 413 -

397 1478 1518 0.974 31 382 4/77 1477 1518- 0.973 30 352 -

5/77 1437 1518 0.947 31 321 6/77 1016 1518 0.669 30 291 7/77 1413 1518 0.931 31 260 8/77 1495 1518 0.985 31 229 9/77 1372 1518 0.904 30 199 10/77 164 1518 0.108 31 168

a. Oscay tirne is referenced to the counting date of the Fe, Ni. Cu and Co unitors (4'17178); the No and U rnonitors were counted on 4/25/78.

e l 6-14 l

~~

^

~

l TABLE 6-3

~ SPECTRUM-AVERAGED REACTION CROSS SECTIONS ,

USED IN FAST NEUTRON FLUX DERIVATION I t

Reaction 7 (barns)  !

i Fe (n p) Mns4[a] 0.0561 ,

. Fe5 ' (n.p) Mns4 [b] 0.0587 i I. Niss (n,p) Coss 0.0758  !

t Cues (n,a) Coso 0.000427  !

r U23s (n f) F.P. 0.320

{,

Np 237 (n,f) F.P. 3.00

4. Appiscable to sempses taken from Core-s de Charpy speciment

+  !

y

b. Applicable to semples taken from pressure-vessel.sde Charpy specimens

. I i

l l

I I

6-15

TABLE 6-4 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE R I Reaction and MeasuredI *I Monitor Activity 0 (E > 1 Mov)[b] .b (E > 1 Mev)[bl Location -(dps/gm) (n/cm2.sec) (n/cm2)

Fe54 (n.p) Mn

• 5 Core-Side C.13 (Bottom Center) 2.84 x 106 1.46 x 1011 2.34 x 1019 ,

R-10 (Center) 3.11 x 106 1.60 x 1011 2.56 x 1019 WH-16 (Top Center) 2.90 x 106 1.49 x 1011 '

2.38 x 1019 Vesset Side A-13 (Bottom Center) 2.54 x 106 1.25 x 1011 2.00 x 10I9 R 11 (Center) 2.54 x 106 1.25 i 1011 2.00 x 1019 WW-15 (Top Center) 2.60 x 106 1.28 x 101I 2.05 x 1019 Nisa (n,p) Coss ~

Center 1.17 x 107 1.62 x 1011 2.59 x - 1019 Cu83 (n,a) Cos0 Bottom 2.19 x 105 1.86 x 1011 2.98 x 1019 -

Bottom-Center 2.14 x 105 1.81 x 1011 2.90 x 1019 Top Center 1.79 x 105 1.52 x 1011 2.43 x 1019 Top 2.04 x 105 1.73 x 10ll 2.77 x 1019 Np 237 (n,f) Cs'37 Center 7.47 x 106 1.40 x 1011 2.24 x 1019 i U 3s (n,f) Cs'37 2

1 i

Center 9.92 x 105 1.67 x 1011 2.67 x 1019

a. MnN. Co00. and Q60 activities are referenced to 12:00 4/17'78. Cs t37 activities are esferenced to 12:00 4/25/78.

b. Dersved fium and fluence tevels are subject to rio percent measurement error. ,

{ .

O 6-16 4- --

L..

r T '

a *S.

l i k  !

I i

a TABLE SS l 7 ./ RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE R f

?

o>

Bare ActivityI 'l Cd Covered ActivityI 'I Monitor Location (dps/gm) (dps/gm) cth (n/cm2 .,,,)[bl Bottom 5 00 x 10 7 _

2.67 x 107 9.84 x 1010 Bottom Center 5.62 x 107 2.85 x 107 1.17 x 1011 (d j Center 5.24 x 107 2.46 x 107 1.17 x 1011 Top-Center 5.56 x 107 2.27 x 107 1.39 x 10Il i Top 5.77 x 107 2.32 x 10 7 1.46 x 10ll j

a. Co actmtes are referenced to 12:00 4/17/78. ,

b. Derived fium levels are subsect to 210 percent measurement error.

A I

t TABLE S6 5

CALCULATED FAST NEUTRON FLUX AND LEAD FACTORS FOR CAPSULE R '

j, {

a v' i Location Within Capsule R c (E > 1 Mev) (n/cm 2.sec) Lead Factor l

Front Charpy (Core Side) 1.24 x 1011 3.18 i Dosimeter Block and Flux 1.18 x 1011 3.03 Wires j i

Back Charpy (Vessel Side) 9.91 x 1010 2.52 i i

i i

L- l i

I 6 17 l

_ _ _ _ _ _ __ _ _ _ _ _ , _ . __ _.~ .