ML19305E071

Analysis of Capsule R from Util Facility Reactor Vessel Radiation Surveillance Program.

ML19305E071
Person / Time
Site:	Point Beach
Issue date:	12/31/1979
From:	Magar T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19305E069	List: ML19305E068 ML19305E071
References
1021-3, WCAP-9635, NUDOCS 8004220284
Download: ML19305E071 (112)

Text

1 JL h ANALYSIS OF CAPSULE R FROM THE WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM (WCAP-9035)

EPRI RESEARCH PROJECT 1021-3 TOPICAL REPORT December 1979 Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P. O. Box 355 Pittsburg.., Pennsylvania 15230 T. R. Mager, Principal Investigator i

Prepared for .

ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94304 Project Manager T. U. Marston 8 0 04 2 20 QQ

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ANALYSIS OF CAPSULE R FROM THE WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM (WCAP-9635)

EPRI RESEARCH PROJECT 1021-3 TOPICAL REPORT S. E. Yanichko S. l . Anderson R. P. Shogan R. G. Lott December 1979 Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230 T. R. Mager, Principal investigator Prepared for ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94304 Project Manager T. U. Marston I

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LEGAL NOTICE This report was prepared by Westinghouse Electric Corporation (WESTINGHOUSE) as an account of work sponsored by the Electric Power Reseaich Institute, Inc. (EPRI). Neither EPRI, members of EPRI, nor WESTINGHOUSE, nor any person acting on behalf of either:

a. Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

b. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.

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TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 1-1 2 INTRODUCTION 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE R 5-1 5-1. Charpy V-Notch impact Test Results 5-3 5-2. Tensile Test Results 5-13 5-3. Wedge Opening Loading Tests 5-13 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1. Introduction 6-1 6-2. Discrete Ordinates Analysis 6-1 6-3. Neutron Dosimetry 6-4 6-4. Transport Analysis Results 6-9 6-5. Dosimetry Results 6-13 References A-1 l

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LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the Point Beach Unit No. 2 Reactor Vessel 45 4-2 Capsule R Schematic Diagram Showing Designed Arrangement of Specimens, Thermal Monitors, and Dosimeter Placement and Orientation with Respect to the Core and Vessel Wall 4-7 5-1 Charpy V-Notch Impact Data for the Point Beach Unit No. 2 Pressure Vessel Intermediate Shell Forging 123V500VA1 5-15 5-2 Charpy V-Notch Impact Data for the Point Beach Unit -

No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 5-17 5-3 Charpy V-Notch Impact Data for the Point Beach Unit No. 2 Pressure Vessel Weld Metal 5-19

, 5-4 Charpy V-Notch Impact Data for the Point Beach Unit No. 2 Weld Heat-Affected Zone Metal

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5-21 5-5 Charpy V-Notch Impact Data for A533 Grade B Class 1 ASI'M Correlation Monitor Material 5-23 l 5-6 Point Beach Unit No. 2 Material 30 ft Ib Transition I

Temperature Increases as Compared to Westinghouse Predictions 5-25 5-7 Charpy Impact Specimen Fracture Surfaces for Point Beach Unit No. 2 Pressure Vessel Intermediate Sheil Forging 123V500VA1 5-27 5-8 Charpy impact Specimen Fracture Surfaces for Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 5-29 l 5-9 Charpy Impzt Specimen Fracture Surfaces for Point Beach Unit No. 2 Weld Metal 5-31 5 10 Charpy Impact Specimen Fracture Surfaces for Point l Beach Unit No. 2 Weld Heat-Affected Zone Metal 5-33 5-11 Charpy impact Specimen Fracture Surfaces for Point Beach Unit No. 2 ASTM Correlation Monitor Material 5-35 5-12 Tensile Properties for the Point Beach Unit No. 2 Pres-sure Vessel Intermediate Shell Forging 123V500VA1 5-37 v

LIST OF ILLUSTRATIONS (cont)

Figure Title Page 5-13 Tensile Properties for the Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 5-39 5-14 Tensile Properties for the' Point Beach Unit No. 2 Pressure Vessel Weld Metal 541 5-15 Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Intermediate Shell Forging 123V500VA1 5-43 5-16 Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Lower Shell Ferging 122W195VA1 5-45 5-17 Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Weld Metal 5-47 5-18 Typical Stress-Strain Curve for Tension Specimens (Tension Specimen No. VS) 549 6-1 Point Beach Unit No. 2 Reactor Geometry 6-25 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-27 6-3 Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel-Surveillance Capsule Geometry 6-29 6-4 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-31 6-5 Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-33 6-6 Calculated Radial Distribution of Maximum Fast 3 Neutron Flux (E > 1.0 Mev) Within Surveillance Capsules V, R and T 6-35 6-7 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules V and R 6-37 6-8 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule T 6-39 6-9 Comparison of Measured and Calculated Fast Neutron 1 Fluence (E > 1.0 Mev) for Capsules V, T, end R 6-41

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LIST OF TABLES Table Title Page 4-1 Chemistry and Heat Treatmut of Material Representing the Core Region Shell Forgings and Weld Metal from the Point Beach Unit No. 2 Reactor Vessel 4-2 4-2 Chemistry and Heat Treatment of Surveillance Material Representing 12-inch-Thick A533 Grade B Class 1 Correlation Monitor Material 4-3 5-1 Irradiated Charpy Test Results from Shell Forging 123V500VA1 5-5 5-2 Irradiated Charpy Test Results.from Shell Forging 122W195VA1 > 5-5 5-3 Irradiated Charpy Test Results from Weld Material 5-7 5-4 Irradiated Charpy Test Results from Weld Heat-Affected-Zone Material 5-7 5-5 Irradiated Charpy Test Results from A533 Grade B Class 1 Correlation Monitor Material 5-9 6-6 The Effect of 288 C Irradiation at 2.01 x 1019n/cm2 i (E > 1.0 Mev) on the Notch Toughness Properties of

( the Point Beach Unit No. 2 Reactor Vessel Materials 5-9 5-7 Summary of Point Beach Unit No. 2 Reactor Vessel Surveillance Capsule Charpy impact Test Results 5-11 5-8 Tensile Properties for Point Beach Unit go. 2 Pressure l Vessel Material Irradiated to 2.01 x 101 n/cm2 5-14 )

l 6-1 21 Group Energy Structure 6-3 l

6-2 Nuclear Parameters for Neutron Flux Monitors 6-5 )

l 6-3 Calculated Fast Neutron Flux (E ,-> 1.0 Mev) and Lead Factors for Point Beach Unit No. 2 Surveillance Capsules 6 10 6-4 Calculated Neutron Energy Spectra at the Dosimeter Block Location for Point Beach Unit No. 2 Surveillance Capsules 6-11 6-5 Spectrum Averaged Reaction Cross Sections at the Dosi-meter Block Location for Point Beach Unit No. 2 Surveillance Capsules 6-12 6-6 Irradiation History of Point Beach Unit No. 2 Reactor Vessel Surveillance Capsules 6-14 6-7 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule R 6-17 vii

LIST OF TABLES (cont)

Table Title Page 6-8 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule V 6-18 6-9 Comparison of Measured and Calculated Fast Neutron Flux Monitor Activities for Capsule T 6-19 6-10 Results of Fast Neutron Dosimetry for Capsules R, V, and T 6-20 6-11 Results of Thermal Neutron Dosimetry for Capsules R, V, and T 6-21 6 12 Summary of Neutron Dosimetry Results for Capsules V, T, and R 6-22 l

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SECTION 1

SUMMARY

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

The capsule received an average fast fluence of 2.01 x 10 19n/cm2 (E > 1 mev).

The predicted fast fluence for the capsJa was 2.14 x 10 19n/cm2 (E > 1 mev).

The fast fluence of 2.01 x 1019n/cm2 resulted in a 128 C (230 F) increase in the 41 joule (30 ft Ib) transition temperature of the weld metal which is the most limiting material in the core region of the reactor vessel. No increase in the 68 joule (50 ft Ib) transition temperature was determined for the weld metal since the average shelf energy of the material was only 63.5 joule (47 ft Ib). The weld heat-affected-zone material exhibited a 68 joule transition temperature increase of 108 C (195 F).

t The intermediate shell forging 123V500VA1 and lower shell forging 122W195VA1 exhibited a 68 joule transitioi, temperature increase of 39 C (70 F) and 17 C (30 F) respectively after irradiation to 2.01 x 1019n/cm2, ASTM A533 Grade B Class 1 reference correlation monitor material (HSST Plate 02) showed an increase of 88 C (159 F) in the 68 joute transition temperature after irradiating to 2.01 x 1019n/cm2, from Capsule R with earlier capsules irr to A comparison 6.53 of test resulty indicates that irradiation to 2.01 and 8.29 x 1018n/cm x 10 gdiateg did 1 n/cm not result in radiation damage saturation as observed for Point Beach Unit No. I surveillance material.

l End-oflife projected fast neutron fluences for the reactor vessel, based on j 32 full power years of operation at 1518 Mw, are as follows:

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l Fast Neutron Fluence (n/cm2)

Vessel Location Calculated Measured Inner surface 3.96 x 10 19 3.90 x 1019 1/4 thickness 2.64 x 1019 2.60 x 1019 3/4 thickness 7.78 x 1018 7.66 x 1018 The calculated vessel inner wall fluence after 5.1 EFPY of operation is 6.36 x 1018n/cm2 versus 5.96 x 1018 as determined from surveillance capsule fluence measurements. The difference of approximately 6 percent in the calculated versus measured fast neutron fluence is due in part to the i 10 percent uncertainty in the measured activities of the fast neutron iron monitors. On the whole, the agreement between calculated and measurement values is considered good.

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1 SECTON 2 l INTRODUCTON i

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. 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Point Beach Unit No. 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. Descriptions of tha surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented in WCAP-7712.III The survei!!ance program, planned to cover the 40 year life of the reactor pressure vessel, was bascd on ASTM E-185-66, Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors...[2]

This report summarizes testing and the postirradiation data obtained from the third material surveillance capsule (Capsule R) 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 pre-viously removed Point Beach Unit No. 2 surveillance Capsule V, reported by Battelle Memorial Institute in 1975l3I and Capsule T reported by Westinghouse.l4I I

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SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and 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 SA508 Class 2 (base material of the Unit No. 2 reactor pressure vessel beltline) are well documented in the literature. Generally, low-alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyras to guard against fast fracture in reactor pressure vessels is presented in " Protection Against Non-ductile Failure," Appendix G to Section lli of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature, RTNDT-RT NDT s i 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 temperature (or 35 mil lateral expansion temperature if this is greater) as determined from Charpy I

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 functica of temperature. Allowabic operating limits can then be determined utilizing these allowable 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-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 3-1

l Reactor Vessel Radiation Surveillance Program,IU in which a surveillance capsule is ' l periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the Charpy V-notch temperature (6RTNDT) due to irradiation is added to the original RT NDT to adjust the RT NDT or f radiation embrittlement. This adjusted RTNDT (RT NDT initial + ARTNDT) is used to index the material to the KIR curve and in turn to set operation limits for the nuclear power plant which take into account the effect of irradiation on the reactor vessel materials.

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l SECTION 4 DESCRIPTION OF PROGRAM 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 l

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 after approximately 6.5 calendar years (5.1 effective full-power years) of plant operation. This capsule contained Charpy V-notch impact, tensile, and WOL speci- l mens (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 weld-heat-affected zone (HAZ) material. The capsule also contained Charpy V-notch specimens from the 12-inch-thick ASTM A533 Grade B Class 1 correlation monitor material furnished by Oak Ridge National Laboratory from IISST Plate 02. The chemistry and heat treatment of the surveillance material are presented in tables 4-1 and 4-2.

i All test specimens were machined from the 1/4-thickness location of the forginos. Test specimens represent material taken at least one forging thickness from the quenched end of the forging. All base metal Charpy V-notch and tensile specimens were oriented with the longitudinal axis of the specimen parallel to the principal working direction of the forgings.

The WOL test specimens were machined with the crack plane of the specimen perpendicular to the surfaces and working direction of the forgings.

Charpy V-notch specimens from the weld 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 specimen parallel to the weld.

I Capsule R contained dosimeter wires of copper, nickel, and aluminum-0.15 wt. percent cobalt (cadmium-shielded and unshielded). In addition, the capsule contained cadmium-shielded dosimeters of Np-237 and U-238, located as shown in figure 4-2.

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e TABLE 41 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)

Elenwnt Forging 123V500VA1 Forging 122W195VA1 Weld Metal C 0.20 0.22 0.079 Mn 0.65 0.59 1.40 P 0.009 0.010 0.014 S 0.009 0.008 0.013 Si 0.24 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 Al <0.005 <0.005 <0.005 N2 0.004 0.002 0.010 V 0.010 0.010 <0.002 Co 0.004 0.010 0.013 Heat Treatment Forging 123V500VA1 Heated at 1550 F, 9% hours, 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 />, air-cooled Stress-relieved 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 Heated 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 />, air-cooled Stress-relieved 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 l Weldment Stress-relieved at 1125 F,11% hours, furnace-cooled l

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CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL REPRESENTING 12-INCH-THICK A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL l (HSST Plate 02)

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 --

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Check 0.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14 Heat Treatment 1675125 F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Air cooled 1600125 F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Water-quenched 1125125*F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Furnace-cooled 1150125*F - 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> - Furnace-cooled to 600 F l

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

2.5% Ag, 97.0% Pb Melting Point 304 C (579*F) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 310*C (590*F) i l

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i 15,797 28 l

l R (3.37)

REACTOR VESSEL P (1.94)

[ THERMAL SHIELD l

CAPSULE l

(TYP) - 10

._. ~ 10 _.

_ 57 _

Y e h / W g

f

_ _ l j S (1.79)

T (1.94)

X V (3.37)

Figure 4-1. Arrangement of Surveillance Capsules in the Point Beach Unit No. 2 Reactor Vessel (Lead Factors for the Capsules ,

are Shown in Parentheses) '

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t TENSILE TE NS ILE WOL WOL WOL CMARPY C HAR PY CNARPY CMAAPY CHARPY CHARPY CHARPY CHARPY CNAAPY V -4 V6 V6 Nil V15 N13 V17 N15 Vl9 A9 V21 Ril V23 Rl3 El3 All EI5 v -4 v-5 v-e ==== - -- -- -- -- -- -===N V -5 M9 Vl3 NIO Vit #12 V16 Mit Vl0 Mle V20 R10 V22 R12 V2% Rin fl u Rl6 El 6

.6 26 OPPOSITE MIDPLANE 7F me 8M- 3 04*C e4- 310'C ll u

MONITOR ll u

MONITOR 09 On Fe -** I M-Co (Cs) e 18 Cu -d a88888M-Ce (Cd) m to M Co

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TO BOTT04 0F VE33Ei 1

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CORE SPECIMEN NUMBERipG CODE E - FORGING 122WI95 val .

V - F0 AGING 123V500 val R - ASTM CORRELATION MONITOR W - WELD METAL M - MEAT AFFECTED ZONE VESSEL WALL 26 i

APSUM R 9

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>03 tE TE R CHARPY CHARPY CHARPY CH AR PY TEN 31LE WOL WOL WOL TENS id WOL WOL WOL TEN 5 ILE fLOCR W9 E17 wil Elg 13 E21 Wl5 E23 E4 E -6 W-5 y 69 - mnum - E -g E -5 E-6 - W4 W-5 W -6 imisunzu wie EIS W12 E20 14 E22 Wl6 E24 E-5 W4 wa

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7 7"""

I - 304*C 9F e+- 310C o

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u l MONITOR

'lL MONITOR ll u

MON ITolt nn nr Dn NI - q llp Co(Cd) Cu -yll - to (Cd) Co -gllp-Co (Cd) m Ce m to h Co TO TOP 0F VE75EL 4010N OF VESSEL U l

l Figure 4-2. Caosule R Schematic Diagram Showing Designed Arrangement of Specimens, Thermal Monitors, and DosimSter Placemen; and Orientation with Respect to the Core and Vessel Wall l

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SECTION 5 TESTING OF SPECIMENS FROM CAPSULE R The postirradiation mechanical testing of the Charpy V-notch r .d tensile specimens was performed at the Westinghouse Research and Development L.boratory with consu!tation 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 care-fully removed, inspected for identification number, and checked against the master list in WCAP-7712.UI No discrepancies were found.

Examination of the two low-melting 304 C (579 F) and 310'C (590 F) eutectic alloys indicated no melting of either type of thennat monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304 C (579 F).

The Charpy impact tests were performed on a Tinjus-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instrumentation system. With this ;ystem, load time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve, the load of general yielding (PGY), the time to general yielding (tGY), the maximum load (PM), and the time to maximum load (tM) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture j was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture i terminated is identified as the arrest load (PAI-The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is roughly equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (EP) is the difference between the total energy to fracture (ED ) and the energy at maximum load.

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

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Percent shear was determined from postfracture photographs using the ratio-of-areas method in compliance with ASTM Specification A370-74. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tensile tests were performed on a 20,000 lb Instron, split-console test machine (Model 1115) per ASTM Specifications E8 and E21, and MHL Procedure 7604 Revision 2. All pull rods, grips and pins were made of inconel 718 hardened to Rc 45. The upper. pull rod was connected through a universal joint to improve axiality of loading.

The tests were conducted at a constant crosshead speed of 0.05 inchas/ minute throughout the test.

Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inches. The extensometer is rated as Class B-2 per ASTM E83.

Elevated test temperatures were obtained with a three-zone electric resistance split tube fumace with a 9-inch hot zone. All tests were conducted in air.

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

Chromel-alumel thermocouples we:e inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In test configuration, with a slight load on the specimen, a plot of specimen temper-ature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550*F. The upper grip was used to control the furnace temperature.

During the actual testing the grip temperatures were used to obtain desired specimen tamperatures. Experiments indicated that this method is accurate to i 2 F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used

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

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5-1. CHARPY V NOTCH IMPACT TEST RESULTS j The results of tests performed on the various materials in Capsule R are presentec' in i tables 5-1 through 5-6 and figures 5-1 through 5-5. The results of tests performed on intermediate shell forging 123V500VA1 presented in figure 5-1 and table 5-6 show that irradiation to 2.01 x 1019n/cm2 increases the 68-joule (50 ft Ib) and 41 joule (30 ft Ib) temperatures by 39 C (70 F) and increases the 0.9 mm (35 mil) lateral expansion tempera-ture by 47 C (85 F). No decrease in upper shelf energy resulted from irradiation.

The results of tests performed on material from the lower shell forging 122W195VA1 presented in figure 5-2 and table 5-6 show an increase in the 68-joule and 41-joule tempera-ture of 17 C (30 F) and 19 C (35 F), respectively, after irradiation to 2.01 x 1019n/cm2.

The 0.9 mm lateral expansion temperature of this material increased by 29 C (52*F) and the upper shelf energy decreased by 6.5 joules (5 ft ib).

A comparison of the surveillance program test results to date on material from the two core region forgings irradiated to three different neutron fluence levels (6.53, 8.29 and 20.1 x 1018 n/cm2) is presented in table 5-7. The results indicate that both materials are not very sensitive to irradiation, as expected, because of the low copper content levels

(< 0.10% Cu) of the forgings. The data indicate that some additional increase in transition temperature did result from irradiation to a fluence of 2.01 x 10 19n/cm2, Test results obtained on the weld metal presente.d in figure 5-3 and table 5-6 show that irradiation to 2.01 x 1019n/cm2 resulted in a 41 joule transition temperature increase of 128 C (230 F). No increase in the 68-joule transition temperature could be determined I

since the upper shelf energy after irradiation decreased to 63.5 joules. The transition temperature increase measured at the 0.9 mm !ateral expansion level was 128*C (230 F) which is identical to the 41-joule transition temperature increase. A comparison of these results for the weld metal versus earlier surveillance capsule results from irradiations to

, 6.53 and 8.29 x 1018n/cm2 is shawn in table 5-7. This table shows that irradiation to l 2.01 x 1019n/cm2 resulted in a significant additional transition temperature increase rather than a saturation of damage which was inferred from the results of the lower fluence data.

l The test results for the weld HAZ material are shown in figure 5-4 and table 5-6. Large scatter is shown in the test results, which appear to follow properties characteristic of either the base or weld material as reflected in the extremely high and low energy values

, in the upper shelf region. The 68- and 41-joule transition temperature increases for the HAZ material were 108 C (195 F) and 105 C (190 F) respectively. A 92*C (166*F) increase resulted at the 0.9 mm lateral expansion level. The upper shelf energy of the HAZ material after irradiation is higher than the unirradiated shelf. A comparison of the surveillance 5-3

r#

IRRADIATED Ne Test Charpy Lateral Charpy Sample Temp Energy Expansion Shear Ed/A Number (*C) (J) (mm) (%) (KJ/M 23 V23 -57 5.4 .14 0 67 V14 -32 18.3 .19 0 228 V24 -32 32.5 .60 5 406 V21 -18 52.9 .50 21 660 V20 -18 84.1 1.14 19 1050 V22 -4 72.5 .98 19 906 V17 10 88.1 .98 28 1101 V16 10 145.1 1.77 53 1813 V13 24 200.7 2.11 76 2508

- V15 25 280.7 1.75 100 3508 V19 79 235.9 2.07 100 2948 V18 121 252.2 2.11 100 3152 IRRADIATE @

Na Test Charpy Lateral Charpy Sample Temp Energy Expansion Shear Ed/A Number ( C) (J) (mm) (%) (KJ/M2)

E20 -57 14.2 .16 0 177 E24 -32 12.2 .09 0 152 E17 -32 54.2 .81 15 677 E18 -18 10.8 .25 5 135 E14 -12 40.0 .48 40 499 E22 -4 113.9 1.51 38 1423 E23 -4 80.0 1.02 31 999 E13 24 146.4 1.69 57 1830 E21 49 146.4 1.71 72 1830 E16 79 185.7 1.91 100 2321 EIS 96 18& 5 1.97 100 2355 E19 121 193.9 1.97 100 2423 i

l l

C. -

~

TABLE 5-1 4RPY TEST RESULTS FROM SHELL FORGING 123V500VA1 F

ted Energies hximum Prop Yield Time to Maximum Time to Fracture Arrest Yield Flow Ep/A , Load Yield Load Maximum Load Load Stress Stress Em/$)

fJ/M (KJ/M') (N) (Micro Sec) (N) (Micro Sec) (N) (N) (MPa) (MPa) 59 8 14200 90 14200 90 14200 0 167 61 16000 100 16900 200 16900 0 824 847 347 59 15600 110 18700 370 18200 0 801 881 491 169 13300 90 18700 500 18700 0 687 824 647 403 14700 95 18700 660 18200 0 755 858 538 368 12000 100 17800 600 17800 0 618 767 631 469 14200 00 18200 660 17300 0 732 835 647 1166 14200 100 18200 660 13800 0 732 835 662 1845 13300 225 17300 725 8000 2700 687 790 ~

768 2739 12500 100 16900 750 0 0 641 755 ^

601 2347 12000 90 16500 760 0 0 618 732 647 2504 12000 100 17300 775 0 0 618 765 TABLE 5-2 kRPY TEST RESULTS FROM SHELL FORGING 122W195VA1 led Energies eximum Prop Yield Time to Maximum Time to Fracture Arrest Yield Flow Em/g Ep/A Load Yield Load Maximum Load Load Stress Stress KJ/M ) (KJ/M2) (N) (Micro Sec) (N) (Micro Sec) (N) (N) (MPa)

(MPak _

167 10 17300 100 18700 210 17800 0 893 927 92 59 16700 135 16900 140 16500 0 858 864 522 155 15600 100 19300 510 19600 0 801 904 67 68 15600 100 16000 130 15600 0 801 812 459 40 15600 100 18700 450 18700 0 801 881 585 838 15600 130 19600 600 16500 0 801 904 506 492 15600 100 19600 510 18700 0 801 904 600 1229 14200 110 id700 600 12900 3600 732 847 662 1167 13800 100 18200 660 13300 5300 709 824 570 1751 13300 125 17300 675 0 0 687 790 570 1785 13300 100 17800 650 0 0 687 801 647 1776 13300 225 16900 725 0 0 687 778 5-5 n-

k IRRADIAT d

Norma Test Charpy Lateral Charpy I Sample Temp Energy Expansion Shear Ed/A Number ('C) (J) (mm) (%) (KJ/M2)

W11 -4 8. 8 .20 0 110 W16 24 14.9 .17 0 186 W9 66 25.8 .56 33 322 W12 93 33.9 .52 55 423 W15 121 44.7 .93 90 559 W10 149 67.8 1.07 100 847 W14 177 61.0 .98 100 762 W13 204 62.4 .91 100 779 IRRADIATED CHARPY Norms Test Charpy Lateral Charpy l Sample Temp Energy Expansion Shear Ed/A Number ( C) (J) (mm) (%) (KJ/M2)

H10 -4 23.0 .43 5 288 H9 24 52.9 .62 35 660 H16 52 35.3 .93 48 440 H13 66 146.4 1.71 77 1830 H14 93 65.1 .93 63 813 H11 121 130.2 1.31 98 1626 HIS 149 88.1 1.18 100 1101 H12 178 221.0 1.84 100 2762 i

)

TABLE 5 3 D CHARPY TEST RESULTS FROM WELD MATERIAL ed Energi:s simum Prop Yield Time to Maximum Time to Fracture Arrest Yield Flow Ep/A Load Yield Load Maximum Load Load Stress Stress Em/g)

(J/M (KJ/M2) (N) (Micro Sec) (N) (Micro Sec) (N) (N) (MPa) (MPa) 75 34 16500 100 16500 100 16500 0 184 2 15600 100 16900 210 16900 0 801 835 833 88 14200 110 16500 280 16500 4400 732 790 166 157 14200 120 16500 300 16000 5300 732 790 215 244 13800 110 16000 360 0 0 709 767 TG1 516 13300 95 16900 350 0 0 687 778 882 479 13800 105 16000 350 0 0 709 767

$82 496 13300 100 16500 340 0 0 687 767 TABLE 5-4 EST RESULTS FROM WELD HEAT AFFECTED-ZONE MATERIAL x1 Energies ximum Prop Yield Time to Maximum Time to Fracture Arrest Yield Flow Ep/A Load Yield Load Maximum Load Load Stress Stress Em/g)

J/M (KJ/V2) (N) (Micro Sec) (N) (Micro Sec) (N) (N) (MPa) (MPa)

$82 5 16500 110 17800 290 17800 0 847 881 s22 138 13300 100 19600 530 19600 0 687 847 150 190 14200 100 16900 280 16900 3600 732 801 491 1339 14200 100 18200 500 0 0 732 835 507' 306 13800 115 17800 525 16900 8000 709 812 647 979 13800 100 18200 660 0 0 709 824 459 641 13800 100 17300 490 0 0 709 801 723 2038 12500 110 17300 750 0 0 641 767 S-7 i

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Normalized Test Charpy Lateral Charpy Maxin Sample Temp Energy Expansion Shear Ed/A Em)

Number ( C) (J) (mm) (%) (KJ/M2) (KJ/l R14 24 6.1 .04 5 7G 4:

R12 79 32.5 .67 23 406 2 91 R13 99 46.1 .70 34 576 42' R11 107 56.9 .99 38 711 49' RIO 123 71.9 .76 62 898 3 71 R9 134 111.2 1.52 52 1389 45!

RIS 175 130.2 1.68 100 1626 4 51 R16 204 135.6 2.01 100 1694 49' THE EFFECT OF 288 (

ON THE N{

POINT BEACH Transition Temperature Unirradiated Irra 50 ft Ib 30 ft Ib 35 Mit 50 f t Ib 30 68 Joule 41 Joule .9 mm 68 Joule 41 Material C ( F) C ( F) C ( F) C ( F) *C 123V500VA1 -51 (-60) -62 (-80) -59 (-75) -12 (10) 122W195VA1 -26 (-15) -42 (-45) -38 (-37) -9 (15) -8:

Weld Metal 16 (60) -18 (0) -8 (17) -

11

, HAZ Metal -29 (-20) -62 (-80) -27 (-16) 79 (175) 43 (122W195VA1)

Correlation 27 (81) 10 (49) 12 (53) 115 (240) 93 Monitor i

d

\

t

4 TABLE ' 5 l HARPY TEST RESULTS FROM A533 GRADE B D CORRELATION MONITOR MATERIAL mergi:s lm Prep Yield Time to Maximum Time to Fracture Arrest Yield Flow Ep/A Load Yield Load Maximum Load Load Stress Stress g) (KJ/M2) (N) (Micro Sec) (N) (Micro Sec) (N) (N) (MPa) (MPa) 32 12000 80 12000 80 12000 0 107 13800 120 16900 360 16900 5300 709 790 148 13800 100 17800 460 17800 6700 709 812 220 13300 100 18200 510 18200 8000 687 812 518 13300 105 16500 440 0 0 687 767 930 12900 100 17300 500 0 0 664 778 1167 10700 60 16900 490 0 0 549 709 1203 12900 120 17300 520 0 0 664 178 TABLE 5-6 IRRADIATION AT 2.01 x 10 19 n/cm2 (E > 1.0 MEV)

TCH TOUGHNESS PROPERTIES OF THE UNIT NO. 2 REACTOR VESSEL MATERIALS iated A Transition Temperature Average Energy Absorption ft Ib 35 Mil 50 ft Ib 30 ft Ib 35 Mil loula .9 mm 68 Joule 41 Joule .9 mm Unirradiated irradiated A Energy

("F) C ( F) C ( F) C ( F) C (*F) Joule ft Ib Joule ft Ib Joule ft Ib

(-10) -12 (10) 39 (70) 39 (70) 47 (85) 244 180 256 189 +12 +9 3 (-10) - 9 (15) 17 (30) 19 (35) 29 (52) 196.5 145 190 140 -6.5 -5

' (230) 120 (247) -

128 (230) 128 (230) 88 65 63.5 47 -24.5 -18 110) 65 (150) 108 (195) 105 (100) 92 (166) 113 83.5 146.5 108 +33.5 +24.5 l

200) 104 (220) 88 (159) 83 (151) 93 (167) 167.5 123.5 133 98 -34.5 -25.5 5-3

i TABLE 57

SUMMARY

OF POINT BEACH UNIT NO. 2 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS 68 Joule 41 Joule 50 ft Ib 30 ft Ib . Decrease in Trans. Temp Trans. Temp Upper Shelf Fl Increase Increase Energy Material 101 ynce n/cm2 ( C) ( F) ( C) (* F) (Joule) (ft Ib)

Forging 123V500VA1 6.53 19 35 17 30 None 8.29 22 40 17 30 None 20.10 39 70 39 70 None Forging 122W195VA1 6.53 8 15 6 10 13.5 10 8.29 7 13 9 17 None 20.10 17 30 19 35 6.5 5 Weld Metal 6.53 - -

92 165 31 23 8.29 78 140 81 145 12 9 20.10 - -

128 230 24.5 18 HAZ Metal 6.53 - -- - -

None 8.29 58 105 61 110 8 6 20.10 108 195 105 190 None Correlation Monitor 6.53 61 '110 50 90 40.5 30 8.29 62 111 58 105 20.5 15 20.10 88 159 83 151 34.5 25.5 l

l l

5-11

I l

program test results to date on the HAZ material is presented in table 5-7. The comparison of data indicates additional transition temperature increase as a result of the 2.01 x 10 19n/cm2 Irradiation. Because of the considerable scatter in data for both unirradiated and irradiated tests, the reported transition temperature increases should be considered highly questionable.

Figure 5-5 and table 5-6 present the test results obtained on the A533 Grade B Class 1

! ASTM reference correlation monitor material. Irradiation to 2.01 x 1019n/cm2 resulted in a 68- and 41 joule transition temperature increase of 88 C (159 F) and 83*C (151*F)

, respectively, and a 0.9 mm lateral expansion increase of 93*C (167'F). Upper shelf energy l decreased by 34.5 joules. A comparison of data obtained at three neutron fluence levels from surveillance program capsu!es is shown in table 5-7. These results show that irradiation i

to 2.01 x 1019 n/cm2 produced increases in transition temperature greater than those I obtained from irradiations at lower fluence levels.

j in the summary of results shown in table 5-7, the results of tests from the second surveillance l capsule irradiated at the intermediate fluence level indicated that possibly a saturation of radiation damage was occurring. However, a continuation of this saturation was not con-l l firmed by the test results from the third capsule irradiated to 2.01 x 1019n/cm 2 . It should be noted that prior to the testing and analysis of the third capsule, including re-evaluation of neutron fluence measurements for earlier capsules, the fluences reported for the first and second capsules were 4.74 and 9.45 x 1018n/cm2 respectively. Based on these early fluence measurements, the saturation which seemed to be occurring appeared to be real since one would expect significabt additional radiation damage at the higher fluence level. However,

[

based on the new revised fluence measurements of 6.53 and 8.29 x 1018 n/cm2, the dif-ference in fluence between the two capsules is not large enough to result in any noticeable difference in radiation damage at the two fluence levels. It is therefore suspected that the saturation of radiation damage observed from tests performed on the second capsule is not a real phenomenon, especially since test results on the third capsule did show additional transition temperature increases. A compariw.s of the 41 joule transition temperature increases with Westinghouse-predicted 41 joule transition temperature increases is shown in figure 5-6. These results show that none of the 41-joule transition temperature increases exceed the Westinghouse predictions.

l Charpy impact specimen fracture surfaces of the various surveillance materials are shown '

( in figures 5-7 through 5-11.

l I

5-12

5 2. TENSlLE TEST RESULTS The results of tensile tests performed on the two shell forgings,123V500VA1 and 122W195VA1, and weld metal at various temperatures from room temperature to 288*C (550 F) are shown in table 5-8. A comparison of the tensile test results for the two forgings after irradiation to three different fluence lavels is shown in figures 5-12 and 5-13. The results of the 2.01 x 1019n/cm2 irradiation, when compared with the results of lower irradiation levels, tend to indicate that the tensile properties of the two forgings have saturated. A comparison of the weld metal tensi!e test results after irradiation to fluence levels of 6.53 and 20.1 x 1018n/cm2 is shown in figure 5-14. This comparison shows that irradiation resulted in significantly high increases in 0.2-percent yield strength, indicating that the material is highly sensitive to irradiation. The irradiation at 2.01 x 10 19n/cm 2 which resulted in additional increased yield strength over that resulting from irradiation at 6.53 x 1018n/cm2 tends to confirm the additiona! increase in transition temperature which occurred.

Photographs of the fractured tensile specimens from the two forgings and the weld metal are shown in figures 5-15 through 5-17 respectively. A typical stress-strain curve for the tensile tests is shown in figurc 5-18.

5 3. WEDGE OPENING LOADING TESTS Wedge Opening Loading (WOL) fracture mechanics specimens which were contained in the surveillance capsule have been stored at the Westinghouse Research Laboratory at the request of the Wisconsin Electric Power Cor9any on the recommendation of the U.S. Nuclear Regulatory Commission; they will oe tested and reported later.

5-13 I

TABLE 5-8 TENSILE PROPERTIES FOR POINT BEACH UNIT NO.g PREpURE VESSEL MATERIAL IRRADIATED TO 2.01 x 101 n/cm Test YieW Ultimate Fracture Fracture Fracture Ultimate Total Reduction Sample Temperature Strecsth Strength Loed Stress Strength Elongation Elongation in Area Number ( C) (MPa) (MPo) (N) (MPa) (MPa) (%) (%) (%)

E-4 27 541 653 13,300 1270 419 8.7 22.5 67 E-5 149 503 618 13,000 1290 412 8.4 20.6 68 E-6 288 470 625 13,500 1380 427 8.7 20.6 69 V-4 27 441 597 11.900 1340 375 10.7 25.2 72 V-5 149 401 548 11,200 1260 354 9.5 22.8 72 V' V6 288 387 554 11,900 1390 375 10.2 22.5 73 W-4 27 657 738 17,600 1390 555 9.8 20.9 60 W-5 149 597 684 17,000 1110 535 8.7 18.8 52 l

l W-6 288 573 688 17,600 1210 555 8.2 16.7 54 Semples E-4, E-5, and E-6 are 122W195VA1 material Semples V-4, V-5, and V-6 are 123V500VA1 material Semples W-4, W-5, and W-6 are wefd metal l

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e ./

REDUCTION IN AREA N I 3 40 -

p UNIFORM g ELONGATION 7 TOTAL ELONGATION o t- /

20 - W.. .I\

,_ y g- g . . .P .... ... .

3 ,

e- .

4 .... .. 2 o l I I I I I O 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-14. Tensile Properties for the Point Beach Unit No. 2 Pressure Vessel Weld Metal 5-41

15797-3 V4 27 C s: na.y. . . . 2.. . . . -n 4

i V5 149 C l l

l V6 288*C Figure 5-15. Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Intermediate Shell Forging 123V500VA1 l

1 5-43 i

t

15797-2 l

E4 27 C l

l i

l l . . . , . . . _.

L ek u E5 149 C

._w.,3- m._ .-

l l .

E6 288 C Figure 5-16. Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Lower Shell Forging 122W195VA1 5-45

1 l

15797-1 W4 27 C l

l l

W5 14gaC

~. ,,ra.

e W6 288 C l

Figure 5-17. Fractured Tensile Specimens from Point Beach Unit No. 2 Pressure Vessel Weld Metal 5-47

100 80 -

m v5 E

8 UJ m x 8 $ 40 -

20 -

0 l l l l l l l 0 0.03 0.06 0.09 0.12 0.15 0.18 0.21 0.24 STRAIN (IN./IN.)

G Figure 5-18. Typical Stress Strain Curve for Tension Specimens !S (Tension Specimen No. V5) h

e. - __ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _

SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1. INTRODUCTION Knowledge of the neutron environment within the pressure vessel surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons.

First, in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be established. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information, on the other hand, is derived solely from analysis.

This section describes a discrete ordinates Sn transport analysis performed for the Point Beach Unit 2 reactor to determine the fast neutron (E> 1.0 Mev) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules; and, in turn, to develop lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averagt.J reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule R is discussed and updated evaluations of dosimetry from Capsules V and T are presented.

6-2. DISCRETE ORDINATES ANALYSIS A plan view of the Point Beach reactor geometry at the core midplane is shown in t

figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a 0 -45 sectcr is depicted.

i Six irradiation capsules attached to the thermal shield are included in the design to con-stitute the reactor vessel surveillance program. Two capsules are located synimetrically at 13 , 23*, and 33 from the cardinal axis as shown in figure 6-1.

l l

6-1

1 l

l A plan view of a single surveillance capsule attached to the thermal shield is shown in figure 6-2. The stainless steel specimen container is 1-inch square and approximately 63 inches in height. The containers are positioned axially such that the specimens are centered on the core tnidp'ane, thus spanning the central 5.25 feet of the 12 foot high reactor core.

From a neutronic standpoint, the surveillance capsule structures are significant. In fact, as will be shown later, they have a marked impact on the distributions of neutror. INx and energy spectra in the water annulus between the thermal shield and the reactor assel. Thus, in order to properly ascertain the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. Use of at least a two-dimen-sional computation is, therefore, mandatory.

In the analysis of the neutron environment within the Point Beach Unit 2 reactor geometry, predictions of neut}on flux magnitude and energy spectra were made with the DOT [5] two-dimensional discrete ordinates code. The radial and azimuthal distributions were obtained from an R, O computation wherein the geometry shown in figures 6-1 and 6-2 was described in the analytical model. In addition to the R, O computation, a second calculation in R, Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R, Z analysis the reactor core was treated as an equivalent volume cylinder and, of course, the surveillance capsules were not included in the model.

Both the R, O and the R, Z analyses employed 21 neutron energy groups, an S8 angular quadrature, and a Pj cross section expansion. The cross sections were generated via the Westinghouse GAMBITM code system with broad group processing by the APPROPOSI7I and ANISN(8) codes. The energy group structure used in the analysis is listed in table 6-1.

A key input parameter in the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For the analysis, power distributions representative of time-averaged conditions derived from statistical studies of long-term operation of Westinghouse two-loop plants were employed. These input distributions include rod by rod spatial variations for all peripheral fuel assemblies.

It should be noted that this particular power distribution is intended to produce accurate end of-life neutron exposure levels for the pressure vessel. As such, the calculation is indeed representative of an average neutron flux and small (i 15-20%) deviations from cycle to cycle are to be expected.

62

l

(

TABLE 6-1 21 GROUP ENERGY STRUCTURE Group Lower Energy (Mev) 1 7.79' 2 6.07 3 4.72 4 3.68 5 2.87 6 2.23 7 1.74 8 1.35 9 1.05 10 0.821 11 0.388 12 0.111 13 4.09 x 10-2 14 1.50 x 10-2 15 5.53 x 10-3 16 5.83 x 104 17 7.89 x 10-5 18 1.07 x 10 5 19 1.86 x 10-6 ,

20 3.00 x 10'7 21 0.0

' Upper energy of group 1 is i0.0 Mev l

l 6-3 i

1.

i Having the results of the R,0 and R,Z calculations, three-dimensional variations of neutron flux may be approv.imated by assuming that the following relation holds for the applicable agions of the reactor.

l

&(R,Z,0,Eg) = p(R,0,Eg) F(Z,Eg) (6-1) where:

p(R,Z,0,Eg) = neutron flux at point R,Z,8 within energy group g d(R,0,Eg) = neutron flux at point R,0 within energy group g obtained from the R,0 calculation F(Z,Eg) = relative axial distribution of neutron flux within energy group g obtained from the R,Z calculation 6-3. NEUTRON DOSIMETRY The passive neutron flux monitors included in the Point Beach Unit 2 surveljlance program l are listed in table 6-2. The first five reactions in table 6-2 are used as fast neutron monitors to relate neutron fluence (E> 1.0 Mev) to measured materials properties changes. To properly j account for burnout of the product isotope generated by fast neutron reactions, it is nec-essary to also determine the magnitude of the thermal neutron flux at the monitor location.

Therefore, bare and cadmium-covered cobalt-aluminum monitors were also included.

The relative locations of the various monitors within the surveillance capsules are shown in figure 4-2. The nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The iron nonitors are obtained by drilling samples from selected Charpy test specimens. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule.

The use of passive monitors such as those listed in table 6-2 does not yield a direct measure of the energy-dependent flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time- t.nd energy-dependent neutron flux nas on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux' level incident on the various monitors may be

1. .ved from the activation measurements only if the irradiation parameters are well known.

l In particular, the following variables are of interest:

l l

64

(

TABLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Target Fission Monitor Reaction Weight Response Product Yield Material of Interest Fraction Range Half Life (%)

Copper Cu63(n,a)Co60 0.6917 E>4.7 MeV 5.27 years Iron Fe54(n,p)MnM 0.0585 E>1.0 Mev 314 days Nickel NiS8(n,p)Co68 0.6777 E>1.0 Mev 71.4 days Uranium 238' U238(n,f)Cs137 1.0 E>0.4 Mev 30.2 years 6.3

/

Neptunium-237' Np237(n,f)Cs137 1.0 E>0.08 Mev 30.2 years 6.5 Cobalt-Aluminum

• CoS9(n,y)Co60 0.0015 0.4eV<0.015Mev 5.27 years Cobalt-Aluminum CoS9(n,7)Co60 0.0015 E<0.0015 Mev 5.27 years

• Denotes that monitor is cadmium shielded l

9

a The operating history of the reactor a The energy response of the monitor a The neutron energy spectrum at the monitor location a The physical characteristics of the monitor The analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two procedures. First, the disintegration rate of product isotope per unit mass of monitor must be determined. Second, in order to define a suitable spectrum averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated.

The specific activity of each of the monitors is determined using established ASTM procedures.i9,10,11,12,13 ] Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The over-all standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from Point Beach Unit 2, the overall 20 deviation in the measured data is deter-mined to be i 10 percent. The neutron energy spectra are determined analytically using the method described in section 6-1.

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

The reaction product activity in the monitor is expressed as N N p. -At- -At d R = --- f; Y o(E)$(E) l (1 - e I) e (6-2)

A . P max E j.3 6-6 L

where:

R = induced product activity Ng = Avagadro's number A = atomic weight of the target isotope fi = weight fraction of the target isotope in the target material Y = number of product atoms produced per reaction o(E) = energy dependent reaction cross section 0(E) = energy-dependent neutron flux at the monitor location with the reactor at full power Pj = average core pc 'r level during irradiation period j P

max

= maximum or reference core power level A = decay constant of the product isotope tj

=

length of irradiation period j td

=

decay time following irradiation period j Since neutron flux distributions are calculated using multigroup transport methods and, further, since the prime interest is in the fast neutron flux above 1.0 Mev, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the following relation.

o(E) (E)dE =

o 0(E> 1.0 Mev)

"E l

r l

l i

f 67

where:

N

.=

i o(E) & (E)dE 9 9 o G=1 ga _

,a N

& (E)dE pg 1.0 Wv g.g t .0 Mew Thus, equation (6-2) is rewritten N

Ng P j d R = - f; y a p (E>1.0 Mev) (1-e-Atj) e-At A P max i.i or, solving for the neutron flux, R

& (E>1.0 Mev) =

N N p. -At- -At d

- f; y o (1-e )e (6-3) max j.i The total fluence above 1.0 Mev is then given by N p,

1.0 Mev) = $ (E>1.0 Mev) tj (6-4) j.i max 68

where:

N p, t; = total effective full power seconds of reactor operation pmax up to the time of capsule removal M  !

An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered CoS9(n,6)Co60 data by means of cadmium ratios and the use of a 37-barn 2200 m/sec cross section. Thus, 1

R barc ,

'D-1}

D J ,

=

4Th (6-5) ]

No N Pj -Atd

-Atj f; ya p (I-e le A max j=1 where:

R bare D is defined as R Cd covered 6-4. TRANSPORT ANALYSIS RESULTS Results of the nS transport calculations for the Point Beach Unit 2 reactor are summarized in figures 6-3 through 6-8 and in tables 6-3 through 6-5. In figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius,1/4 thick-ness location, and 3/4 thickness location are presented as a function of azimuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident.

In figure 6-4, the radial distribution of maximum fast neutron flux (E> 1.0Mev) through the thickness of the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel is given in figure 6-5. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figures 6-3 or 6-4 by the appropriate values from figure 6-5.

In figure 6-6 the radial variations of fast neutron flux within surveillance capsules V, R, and T are presented. These data, in conjunction with the maximum vessel flux, are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E>1.0 Mev) at the dosimeter block location (capsule center) to 6-9

TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MEV) AND LEAD FACTORS FOR POINT BEACH UNIT NO. 2 SURVEILLANCE CAPSULES Capsule Azimuthal & (E>1.0 Mev) Lead Identification Location (n/cm 2.sec) Factor V 13 1.33 x 1011 3.37 R 13* 1.33 x 1011 3.37 T 23 7.66 x 1010 j ,94 P 23 7.66 x 1010 1,94 S 33 7.06 x 1010 1.79 N 33 7.06 x 10 10 1,79 l

l a

6-10

TABLE 6-4 CALCULATED NEUTRON ENERGY SPECTRA AT THE DOSIMETER BLOCK LOCATION FOR POINT BEACH UNIT NO. 2 SURVEILLANCE CAPSULES Neutron Flux (n/cm2.sec)

Group No. Capsules V&R Capsules T&P Capsules S&N 1 7.52 x 108 5.51 x 108 4.84 x 108 2 2.47 x 109 1.83 x 109 1.61 x 109 3 4.08 x 109 2.83 x 109 2.51 x 109 4 4.58 x 109 2.93 x 109 2.65 x 109 5 7.97 x 109 4.78 x 109 4.37 x 109 6 1.56 x 1010 9.29 x 109 8.52 x 109 7 2.26 x 1010 1.30 x 1010 1.20 x 1010 8 3.25 x 1010 1.81 x 1010 1.68 x 1010 9 4.30 x 1010 2.33 x 1010 2.16 x 1010 10 4.64 x 1010 2.46 x 1010 2.28 x 1010 11 1.54 x 1011 7.97 x 1010 7.39 x 1010 12 1.94 x 1011 9.68 x 1010 8.98 x 1010 13 8.67 x 1010 4.28 x 1010 3.99 x 1010 14 6.54 x 1010 3.24 x 1010 3.02 x 1010 15 5.22 x 1010 2.58 x 1010 2.41 x 1010 16 1.21 x 10 11 5.90 x 1010 5.51 x 1010 17 9.52 x 1010 4.66 x 1010 4.35 x 1010 18 9.75 x 1010 4.73 x 1010 4.43 x 1010 19 7.74 x 1010 3.76 x 1010 3.52 x 1010 20 8.59 x 1019 4.16 x 1010 3.89 x 1010 21 2.73 x 1011 1.39 x 1011 1.25 x 101I l ,

6-11 I

TABLE 6-5 SPECTRUM AVERAGED REACTION CROSS SECTIONS AT THE DOSIMETER BLOCK LOCATION FOR POINT BEACH UNIT NO. 2 SURVEILLANCE CAPSULES 5'(barns)

Reaction Capsules V&R Capsules T&P Capsules S&N Fe54 (n,p) Mn54 0.0595 0.0683 0.0666 NiS8 (n,p) CoS8 0.0811 0.0912 0.0893 Cu63 (n,a) Co60 0.000404 0.000517 0.000494 U238 (n,f) F.P. 0.333 0.345 0.344 Np237 (n,f) F.P. 2.93 2.80 2.82

[o(E)0(E)dE J*

a=

p(E)dE 1 Mew 6 12

the maximum fast neutron flux at the pressure vessel inner radius. Updated lead factors for all of the Point Beach Unit 2 surveillance capsules are listed in table 6-3.

Since the neutron flux monitors contained within the surveillance capsules are not all located at the same radial location, the measured disintegration rates are analytically adjusted for the gradients that exist within the capsules so that flux and fluence levels may be derived on a common basis at a common location. This point of comparison was chosen to be the capsule center. Analytically determined reaction rate gradients for use in the adjustment procedures are shown in figures 6-7 and 6-8 for Capsules V, R, and T. All of the applicable fast neutron reactions are included.

In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center of each of the Point Beach Unit 2 surveillance capsules is gian in table 6-4. The associated spectrum-averaged cross sections for each of the five fast neutron reactions are given in table 6-5.

6-5. DOSIMETRY RESULTS The irradiation history of the Point Beach Unit 2 reactor is given in table 6-6. The data were obtained from the Point Beach semiannual operating reports.Il41 Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsules R, V, and T are listed in tables 6-7, 6-8, and 6-9, respectively. The data are presented as measured at the actual monitor locations as well as adjusted to the capsule center. The measured results for both capsules R and T were obtained by Westinghouse, whereas those for Capsule V were derived from reference 9. All adjustments to the capsule center were based on the data presented in figures 6-7 and 6-8.

The fast neutron (E> 1.0 Mev) flux and fluence levels derived for Capsules R, V and T are presented in table 6-10. The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in table 6-11..Due to the relatively low thermal neutron flux at

) the capsule locations, no burnup correction was made to any of the measured activities.

The maximum error introduced by this assumption is estimated to be less than 1 percent for the NiS8 (n,p) CoS8 reaction and even less significant for all of the other fast neutron reactions.

Using the iron data presented in table 6-10, along with the lead factors given in table 6-3, the fast neutron fluence (E> 1.0 Mev) for Capsules V, T, and R as well as for the reactor vessel inner diameter are summarized in table 6-12 and figure 6-9. The agreement between calculation and measurement is excellent, with measured fluence levels of 6.53 x 1018 , 8.29 x 1018, and 2.01 x 1019 compared to calculated values of l 6-13 l

TABLE 6-6 IRRADIATION HISTORY OF POINT BEACH UNIT NO. 2 REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Decay

• PJ Pmax P3 Time Time Month (MW) (MW) p (days) (days) max 10/72-12/72 369 1518 .243 92 2357 1/73 284 1518 .187 31 2326 2/73 278 1518 .183 28 2298 3/73 740 1518 .487 31 2267 4/73 743 1518 .489 30 2237 5/73 1043 1518 .687 31 2206 6/73 1300 1518 .856 30 2176 7/73 1325 1518 .873 31 2145 8/73 1397 1518 .020 31 2114 9/73 1501 1518 .989 30 2084 10/73 1474 1518 .971 31 2053 11/73 1479 1518 .974 30 2023 12/73 1424 1518 .938 31 1992 1/74 1462 1518 .963 31 1961 2/74 1458 1518 .960 28 1933 3/74 1485 1518 .978 31 1902 4/74 1495 1518 .985 30 1872 5/74 1494 1518 .984 31 1841 6/74 1229 1518 .810 30- 1811 7/74 1283 1518 .845 31 1780 8/74 1494 1518 .984 31 1749 9/74 1464 1518 .964 30 1719 10/74 750 1518 .494 31 1688 CAPSULE V REMOVED 11/74 0 1518 0 30 1658 12/74 406 1518 .267 31 1627 1/75 1497 1518 .986 31 1596-2/75 1311 1518 .864 28 1568

• Decay time is referenced to June 15, 1979.

6-14

s l

TABLE 6-6 (cont) 1RRADIATION HISTORY OF POINT BEACH UNIT NO. 2 REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Decay

• PJ Pmax P j_ Time Time Month (MW) (MW) p (days) (days) 3/75 1454 1518 .958 31 1537 4/75 1273 1518 .839 30 1507 5/75 1183 1518 .779 31 1476 6/75 1270 1518 .837 30 1446 7/75 1430 1518 .942 31 1415 8/75 1006 1518 .663 31 1384 9/75 1385 1518 .912 30 1354 10/75 1330 1518 .876 31 1323 11/75 1354 1518 .892 30 1293 12/75 1498 1518 .987 31 1262 1/76 1439 1518 .948 31 1231 2/76 1335 1518 .879 29 1202 3/76 155 1518 .102 31 1171 4/76 1351 1518 .890 30 1141 5/76 1286 1518 .847 31 1110 6/76 1400 1518 .922 30 1080 7/76 1380 1518 .909 31 1049 8/76 1374 1518 .905 31 1018 9/76 1397 1518 .920 30 988 10/76 1500 1518 .988 31 957 11/76 1478 1518 .974 30 927 12/76 1482 1518 .976 31 896 1/77 1482 1518 .976 31 865 2/77 1485 1518 .978 28 837 3/77 135 1518 .089 31 806 CAPSULE T REMOVED 4/77 293 1518 .193 30 776 5/77 1506 1518 .992 31 745

' Decay time is referenced to June 15, 1979.

6-15

TABLE 6-6 (cont)

IRRADIATION HISTORY OF POINT BEACH UNIT NO. 2 REACTOR VESSEL SURVEILLANCE CAPSULES Irradiation Decay

• PJ Pmax PJ Time Tima Month (MW) (MW) p (days) (days) 6/77 1497 1518 .986 30 715 7/77 1454 1518 .958 31 684 8/77 1322 1518 .871 31 653 9/77 1456 1518 .959 30 623 10/77 1497 1518 .986 31 592 j 11/77 1472 1518 .970 30 562 12/77 1485 1518 .978 31 531 1/78 1398 1518 .921 31 500 2/78 1474 1518 .971 28 472 3/78 972 1518 .640 31 441 4/78 495 1518 .326 30 411 5/78 1500 1518 .988 31 380 6/78 1472 1518 .970 30 350 7/78 1413 1518 .931 31 319 8/73 1428 1518 .941 31 288 9/78 1474 1518 .971 30 258 10/78 1488 1518 .980 31 227 11/78 1491 1518 .982 30 197 12/78 1431 1518 .943 31 166 1/79 1480 1518 .975 31 135 2/79 1480 1518 .975 28 107 3/79 994 1518 .655 31 76

• Decay time is referenced to June 15, 1979.

6-16

TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE R Radial Saturated Activity Adjusted Saturated Activity Reaction and Axial Location (DPS/gm) (DPS/gm)

Location (cm) Capsule R Calculated Capsule R Calculated Fe54(n.p)Mn54 V-13 157.33 ' 6.18 x 106 6.11 x 106 5.27 x 106 E-23 157.33 5.72 x 106 6.11 x 106 4.88 x 106 E-13 157.33 4.96 x 106 6.11 x 106 4.23 x 106 H-9 158.33 4.66 x 106 4.95 x 106 4.90 x 106 R-14 158.33 4.77 x 106 4.95 x 106 5.01 x 106 W-16 158.33 4.61 x 106 4.95 x 106 4.84 x 106 Average 4.86 x 106 5.20 x 106 cn Cu63(n,a)Co60 G Top 158.33 4.28 x 105 3.39 x 105 4.48 x 105 Mid-top 158.33 3.87 x 105 3.39 x 105 4.05 x 105 Mid-bottom 158.33 4.28 x 105 3.39 x 105 4.48 x 105 Bottom 158.33 4.54 x 105 3.39 x 105 4.75 x 105 Average 4.44 x 105 3.55 x 10 'I NiS8(n,p)CoS8 Middle 158.33 7.45 x 107 7.30 x 107 7.86 x 107 7.70 x 107 Np237(n,f)Cs137 Middle 158.10 6.79 x 107 6.45 x 107 6.79 x 107 6.45 x 107 U238(n,f)Cs137 Middle 158.10 9.03 x 106 7.10 x 106 9.03 x 106 7.10 x 106 ,

m-TABLE 6-8 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE V Radial Saturated Activity Adjusted Saturated Activity Reaction and Axial Location (DPS/gm) (DPS/gm)

Location (cm) Capsule V Calculated Capsule V Calculated Fe54(n,p)Mn54 Top 158.33 5.12 x 106 4.95 x 106 5.38 x 106 Mid-top 158.33 4.87 x 106 4.95 x 106 5.12 x 106 Middle 158.33 4.74 x 106 4.95 x 106 4.98 x 106 Mid bottom 158.33 5.49 x 106 4.95 x 106 5.77 x 106 Bottom 158.33 4.90 x 106 4.95 x 106 5.15 x 106 Average 5.28 x 106 5.20 x 106 Cu63(n.a)Co60

$ Top 158.33 3.99 x 105 3.39 x 105 4.18 x 105 C

Mid top 158.33 3.75 x 105 3.39 x 105 3.93 x 10 5 Mid-bottom 158.33 4.09 x 105 3.39 x 105 4.28 x 105 8ottom 158.33 4.14 x 105 3.39 x 105 4.34 x 105 Average 4.18 x 105 3.55 x 10 5 NiS8(n.p)Co68 Middle 158.33 5.84 x 107 7.30 x 107 6.16 x 107 7.70 x 107 '

Np237(n f)Cs137 Middle 158.10 6.58 x 10 7 6.45 x 107 6.58 x 10 7 6.45 x 10 7 U238(n.f)Cs137 Middle 158.10 9.29 x 106 7.10 x 106 9.29 x 106 7.10 x 106

TABLE 6-9 COMPARISON OF MEASURED AND CALCULATED FAST NEsTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE T Radial Saturated Activity Adjusted Saturated Activity Reaction and AxiJ Location (DPS/gm) (DPS/gm)

Location (cm) Capsule T Calculated Capsule T Calculated Fe54(n.p)Mn54 E-47 157.33 4.07 x 106 4.05 x 106 3.43 x 106 V-37 157.33 3.99 x 106 4.05 x 106 3.36 x 106 E-37 157.33 3.90 x 106 4.05 x 106 3.28 x 106 W-32 158.33 3.31 x 106 3.29 x 106 3.43 x 106 H-25 158.33 3.48 x 106 3.29 x 106 3.61 x 106 R-30 158.33 3.14 x 106 3.29 x 106 3.25 x 106 Average 3.39 x 106 3.41 x 1 6 9 Cu63(n.a)Co60 23 Top 158.33 2.99 x 105 2.52 x 105 3.17 x 105 Pl..d. top 158.33 2.70 x 105 2.52 x 105 2.86 x 105 Mid-bottom 158.33 3.12 x 105 2.52 x 105 3.31 x 105 Bottom 158.33 3.38 x 105 2.52 x 105 3.58 x 105 Average 3.23 x 105 2.67 x 105 NiS8(n,p)CoS8 Middle 158.33 4.79 x 107 4.70 x 107 5.05 x 107 4.95 x 107 Np237(n,f)Cs137 Middle 158.10 3.84 x 10 7 3.40 x 107 3.84 x 107 3.40 x 107 U238(n,f)Cs137 Middle 158.10 5.01 x 106 4.23 x 106 5.01 x 106 4.23 x 106

I TABLE 6-10 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULES R, V, AND T Adjusted Saturated Activity 4(E>1.0 Mev) c(E>1.0 Mev)

(DPS/gm) (n/cm 2.sec) (n/cm2)

Capsule Reaction Measured Calculated Measured Calculated Measured . Calculated R Fe54 (n,p)Mn54 4.86 x 106 5.20 x 106 1.25 x 1011 1.33 x 1011 2.01 x 1019 2.14 x 10 19 Cu63(n,a)Co60 4.44 x 105 3.55 x 105 1.66 x 1011 2.67 x 1019 l NiS8(n.p)CoS8 7.86 x' 107 7.70 x 10 7 1.38 x 1011 2.22 x 1019 Np237(n,f)Cs137 6.79 x 10 7 _6.45 x 107 1.40 x 1011 2.25 x 1019 U238(n,f)Cs137 9.03 x 106 7.10 x 106 1.70 x 1011 2.74 x 1019 V Fe54(n p)Mn54 5.28 x 106 5.20 x 10 6 1.36 x 1011 1.33 x 1011 6.53 x 1018 6.38 x 1018 Cu63(n,a)Co60 4.18 x 105 3.55 x 105 1.56 x 1011 7.49 x 1018 cn NiS8(n,p)CoS8 6.16 x 10 7 7.70 x 10 7 1.08 x 1011 5.18 x 1018 ES Np237(n,f)Cs137 6.58 x 10 7 6.45 x 10 7 1.36 x 1011 6.53 x 1018

U238(nACs137 9.29 x 106 7.10 x 10 6 1.75 x 1011 8.40 x 10 19 T Fe54(n,p)Mn54 3.39 x 106 3.41 x 106 7.61 x 1010 7.66 x .1010 8.29 x 1018 8.35 x 1018 Cu63(n,a)Co60 3.23 x 10 5 2.67 x 105 9.45 x 1010 1.03 x 10 19 NiS8(n.p)CoS8 5.05 x 10 7 4.95 x 107 7.87 x 1010 8.58 x 1018 8.30 x 10 10 Np237(n,f)Cs137 3.84 x 107 3.40 x 107 9.05 x 1018 U238(nACs137 5.01 x-106 4.23 x 106 9.11 x 1010 9.93 x 1018

TABLE 6-11 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULES R, V, AND T Saturated Activity (dps/gm) &Th Capsule Axial Location Bare Cd-covered (n/cm 2-sec)

R Top 1.77 x 108 6.88 x 107 1.91 x 1011 Mid-top 1.23 x 108 6.64 x 107 1.00 x 1011 Middle 1.30 x 108 6.40 x 107 1.17 x 1011 Mid-bottom -

7.63 x 107 -

Bottom 1.72 x 108 7.23 x 107 1.77 x 1011 V Top 1.40 x 108 6.25 x 107 1.37 x 1011 Mid-top 1.25 x 108 5.87 x 107 1.17 x 1011 Middle 1.01 x 108 5.18 x 107 8.68 x 1010 9 Mid-bottom 1.23 x 108 5.87 x 107 1.13 x 1011 8ottom 1.25 x 108 5.39 x 107 1.25 x 1011 T Top 1.06 x 108 _ _

Mid-top 6.59 x 107 3.31 x 107 5.79 x 1010 Middle 6.85 x 10 7 3.51 x 107 5.90 x 1010 Mid-bottom 8.59 x 107 3.84 x 107 8.41 x 1010 Bottom 8.13 x 107 3.67 x 10 7 7.87 x 1010

TABLE 6-12

SUMMARY

OF NEUTRON DOSIMETRY RESULTS FOR CAPSULES V, T, AND R Irradiation Vessel Calculated Time (E>1.0 Mev) c(E>1.0 Mev) Lead Fluence Vessel Fluence Capsule (EFPS) (n/cm 2-sec) (n/cm2) Factor (n/cm2) (n/cm2)

V 4.80 x 107 1.36 x 1011 6.53 x 1018 3.37 1.94 x 1018 1.90 x 1018 T 1.09 x 108 7.61 x 1010 8.29 x 1018 1.94 4.27 x 1018 4.31 x 1018 R 1.61 x 108 1.25 x 1011 2.01 x 1019 3.37 5.96 x 1018 6.36 x 1018 c) i l

l t

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6.38 x 1018, 8.35 x 1018, and 2.14 x 1019 n/cm 2for Capsules V, T, and R, recpectively.

Further, the graphical representation in figure 6-9 indicates the accuracy of the transport l analysis for Point Beach Unit 2 and supports the use of the analytically determined fluence trend curve for predicting vessel toughness at times in the future. Projecting to end-of-life, a summary of peak fast neutron exposure of the Point Beach Unit 2 reactor as derived from both calculation and measurement may be made as follows.

Fast Neutron Fluence (n/cm2)

Surface 1/4 T 3/4 T Capsule V 4.05 x 1019 2.70 x 1019 7.96 x 10 18 Capsule T 3.93 x 1019 2.62 x 1019 7.72 x 1018 Capsule R 3.72 x 1019 2.48 x 1019 7.31 x 10 18 Average measurement 3.90 x 1019 2.60 x 1019 7.66 x 10 18 Calculation 3.96 x 1019 2.64 x 10 19 7.78 x 10 18 These data are based on 32 full-power years of operation at 1518 MWt.

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6-23

l 15J9713 PRESSURE VESSEL SURVEILLANCE CAPSULE l 0 1 13 (CAPSULES V,R)

/ 23 (CAPSULES T,P)

THERMAL 33 (CAPSULES S,N)

SHIELD M

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l Figure 6-2. Plan View of a Reactor Vessel Surveillance Capsule 6-27

15,797-15 1012 5 -

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vs 5 -

SURVEILLANCE 2 CAPSULES X

3 u.

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- 3/4T LOCATION 2 -

1 109 l l l l l l l

t 0 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEGREES)

Figure 6 3. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel - Surveillance Capsule Geometry 6-29

15,797 16 10 11 167.64 5

, 171.77 IR 175.90 m 2 -

"E 1/4T M

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$5 w

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OR 2 -

HO2 j PRESSURE VESSEL g 109 160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 RADIUS (CM)

l. Figure 6-4. Calculated Radial Distribution of Maximum Fast Neutron i Flux (E > 1.0 Mev) Within the Pressure Vessel 6-31

15,797-17 l

10 0 _

5 -

2 -

10'1 X -

D -

d -

z 5 -

e S -

z

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CORE MIDPLANE 2 -

TO VESSEL

' CLOSURE HEAD 10-3  !  !  !

300 -200 -100 0 100 200 300 l DISTANCE FROM CORE MIDPLANE (CM) 1 i

Figure 6-5. Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel l

6-33 1

15,797 18 I

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DOSIMETE R -

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MONITORS 2 -

THERMAL CAN CAN SHIELD j HO 2 j j TEST SPECIMENS j f HO 2

10 10 l k  ! h!  !  !  ! $ k 155 156 157 158 159 160 161 RADIUS (CM) l Figure 6-6. Calculated Radial Distribution of Maxirnum Fast Neutron Flux (E > 1.0 Mev) Within Surveillance Capsules V, R, and T 6-35

1 15,797 20 l

l 10 8 -

157.33 158.10 N,P) Co 68

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$ HO 2

g { I, 105

, , , \ l 155 I 157 158 160 161 RADIUS (CM)

Figure 6-7. Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules V and R l

l l

6-37

l SATURATED ACTIVITY (DPS/GM) 1 -- - -. -

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O CAPSULE T DATA 6 CAPSULE R DATA

_ SuCALCULATION 2 -

l l

10 20 13* CAPSULES (V,R) z 5 -

0 -

5 -

2 d

y2 -

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B z

10 19 VESSEL INNER RADIUS 5

2 -

1018 l l l l llll l l l l l lll

! 1 2 5 10 20 50 100 OPERATING TIME (EFPY) t l Figure 6-9. Comparison of Measured and Calculated Fast Neutron Fluence l (E > 1.0 Mev) for Capsules V, T, and R 641

J I

l REFERENCES

1. Yanichko, S. E., and Zula, G. C., " Wisconsin Michigan Power Co., and the Wisconsin Electric Power Company Point Beach Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-7712, June,1971.

2. ASTM Designation E185-66, " Surveillance Tests on Structural Materials in Nuclear  !

Reactors," in ASTM Standards (1976), Part 31, Physical and Mechanical Testing of Metals - Metallography, Nondestructive Testing, Fatigue, Effect of Temperature, pp. 638 642, Am. Soc. for Testing and Materials, Philadelphia, Pa,1967.

3. Perrin, J. S., Farmelo, D. R., Lowry, L. M., Wooton, R. D., and Denning, R. S.,

" Point Beach Unit No. 2 Pressure Vessel Surveillance Program: Evaluation of Capsule V,"

Battelle Research Report, June 10, 1975.

4. Davidson, J. A., Anderson, S. L., and Shogan, R. P., " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9331, August,1978.

5. Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation. Vol. 5 - Two-Dimension Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.

6. Cellier, G., et al, "Second Version of the GAMBIT Code," WANL-TME-19G9, November,1969.

7. Soltesz, R. G., et al, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation - Volume 3, Cross Section Generation and Data Processing Techniques," WANL-PR-(LL)-034, August,1970.

8. Soltesz, R. G., et al, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation - Volume 4 - One-Dimensional Discrete Ordinates Transport Technique," WANL-PR-(LL)-034, August,1970.

9. ASTM Designation E26170, Standard Method for Measuring Neutron Flux by l Radioactivation Techniques," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 745-755, Am. Society for Testing and Materials, Philadelphia, Pa.,1975. j

10. ASTM Designation E262 70, " Standard Method for Measuring Thermal Nuetron Flux by Radioactivation Techniques," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 75G-763, Am. Society for Testing and Materials, Philadelphia, Pa.1975 l l

l A-1

REFERENCES (cont)

11. ASTM Designation E-263 70, " Standard Method for Measuring Fast Neutron Flux by Radioactivation of Iron," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 764 769, Am. Society for Testing and Materials, Philadelphia, Pa.,1975.

12. ASTM Designation E481-73T, " Tentative Method of Measuring Neutron Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 887-894, Am. Society for Testing and Materials, Philadelphia, Pa.1975.

13. ASTM Designation E264-70, " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 770-774, Am. Society for Testing and Materials, Philadelphia, Pa.,1975.

14 Point Beach Nuclear Units 1 and 2 Semi Annual Operating Reports,1970 through 1979.

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