ML20062E548

Analysis of Capsule V from NSP Subj Facil Unit 2 Reactor Vessel Radiat Surveillance Prog, Repts Effects of Neutron Irradiation on Reactor Pressure Vessel Matl Under Operating Conditions

ML20062E548
Person / Time
Site:	Prairie Island
Issue date:	11/30/1977
From:	Shaun Anderson, Davidson J, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20062E543	List: ML20062E540 ML20062E545 ML20062E548
References
WCAP-9212, NUDOCS 7812080142
Download: ML20062E548 (36)

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ANALYSIS OF CAPSULE V FROM NORTHERN STATES POWER COMPANY PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM J. A. Davidson S. E. Yanichko S. L. Anderson November 1977 APPROVED:

k6 u o,m J. N. Chirigos, Manager Structural Materials Engineering

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Work performed under Shop Order No. EJCP 200 Prepared by Westinghouse for Northern States Power Company WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Sys' ems P. O. Box 355 Pittsburgh, Pennsylvania 15230 7 p p g o g g IH4

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.i TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 11 2

INTRODUCTION 21 3

BACKGROUND 31 4

DESCRIPTION OF PROGRAM 41 1

5 TESTING SPECIMENS FROM CAPSULE V 51 5 1.

Background information 51 52.

Charpy V Notch impact Test Results 5-2 53.

Tensile Test Results 52 6

DOSIMETRY ANALYSIS 61 6 1.

Fast Neutron Flux Monitors 61 62.

Analytical Methods 64 6 3.

Results of Analysis 69 64.

Discussion of Results 69 I

Appendix A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1 i

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l LIST OF ILLUSTRATIONS Figure Title Page 41 Arrangement of Surveillance Capsules in the Prairie Island Unit No. 2 Reactor Vessel 45 42 Capsule V Schematic Showing Location of Specimens, Thermal Monitors, and Dosimeters (Prairie Island Unit 2) 46/47 5-1 Irradiated Charpy V Notch impact Properties for the Prairie Island Unit No. 2 Reactor Vessel Lower Shell Forging 22642, Axial Orientation 58 52 Irradiated Charpy V Notch impact Properties for the Prairie Island Unit No. 2 Reactor Vessel Lower Shell Forging 22642. Tangential Orientation 59 5-3 Irradiated Charpy V Notch impact Properties for the Prairie Island Unit No. 2 Reactor Pressure Vessel Weld Metal 5 10 5-4 Irradiated Charpy V Notch Impact Properties for the Prairie Island Unit No. 2 Reactor Pressure Vessel Weld Heat Affected-Zone. Metal 5-11 55 Irradiated Charpy V Notch impact Properties for the A533 Grade B Class 1 Correlation Monitor Material 5 12 56 Charpy impact Specimen Fracture Surfaces for Prairie

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Island Unit No. 2 Pressure Vessel Lower Shell Forging (j

22642, Axial Orientation 5 13 5-7 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 Pressure Vessel Lowr Shell Forging 22642, Tangential Orientation 5-14 5-8 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 Weld Metal 5-15 59 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 Weld Heat Affected-Zone Metal 5-16 5-10 Charpy impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 A533 Grade B Class 1 Correlation Monitor Material 5-17 l

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i LIST OF ILLUSTRATIONS (cont)

Figure Title Page 5 11 Irradiated Tensile Properties for the Prairie Island Unit No. 2 Reactor Pressure Vessel Lower Shell Forging 22642, Axial Orientation 5 18 5 12 Irradiated Tensile Properties for the Prairie island Unit No. 2 Reactor Pressure Vessel Lower Shell Forging 22642, Tangential Orientation 5-19 5-13 Irradiated Tensile Properties for the Prairie island Unit No. 2 Reactor Pressure Vessel Weld Metal 5-20 5 14 Typical Stress Strain Curve for Tension Specimens 5 21 5 15 Fractured Tensile Specimens from Prairie Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Axial Orientation 5 22 5 16 Fractured Tensile Specimens from Prairie Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Tangential Orientation 5 23 5 17 Fractured Tensile Specimens from Prairie Island Unit No. 2 Pressure Vessel Weld Metal 5 24 61 Prairie Island Unit No. 2 Reactor Geometry 63 62 Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Prairie Island Unit No. 2 Reactor Vessel 6 13 63 Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Incident on the Prairie Island Unit No. 2 Reactor Vessel 6 14 64 Calculated Maximum End of Life Fast Neutron Fluence (E >1.0 Mev) as a Function of Radius Within the Prairie Island Unit No. 2 Reactor Vessel 6 15 vi

f LIST OF TABLES Table Title Page 41 Chemistry and Heat Treatment of Material Representing the Core Region Lower Shell Forging and Weld Metal from the Prairie Island Unit No. 2 Reactor Vessel 43 42 Chemistry and Heat Treatment of Surveillance Material Representing 12 inch Thick A533 Grade B Class 1 Correlation Monitor Material

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44 43 Welding Procedure and Associated information for the Prairie Island Unit No. 2 Core Region Weidments 44 51 Charpy V Notch impact Data for the Prairie Island Unit No. 2 Pressure Vessel Lower Shell Forgin 22642 Irradiated at 550*F, Fluence 5.49 x 10 8 n/cm2 (E > 1 Mev) 53 52 Charpy V Notch impact Data for the Prairie island Unit No. 2 Pressure Vessel 550*F, Fluence 5.49 x 10 peld Metal irradiated at 1

n/cm2 (E >1 Mev) 5-5-3 Charpy V Notch impact Data for the Prairie Island Unit No. 2 Pressure Vessel Weld Heat Affecte Metal irradiated at 550*F, Fluence 5.49 x 10 gone n/cm2 (E >1 Mev) 54 54 Charpy V Notch impact Data for the Prairie Island Unit No. 2 A533 Grade B Class 1 Correlation

'j Materjal Irradiated at 550*F, Fluence 5.49 x 10gnitor n/cm (E > 1 Mev) 55 55 The Effect of 550*F frradiation at 5.49 x 1018 n/cm2 (E >1 Mev) on the Notch Toughnass Properties c,f the Prairie Island Unit No. 2 Reactor Vessel impact Test Specimens 56 56 Irradiated Tensile Properties for the Prairie Island Unit No. 2 Pressee Vessel Lo Weld Metal, Fluence 5.49 x 10g Shell Forging and n/cm (E > 1 Mev) 5-7 61 Neutron Flux Monitors Contained Within Capsule V 62 62 Irradiation History of Capsule V Removed from Prairie Island Unit No. 2 68 vii

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't LIST OF TABLES (cont) j Table Title Page 63 Spectrum Averaged Reaction Cross Sections Used in Fast Neutron Flux Derivation 68 64 Results of Fast Neutron Dosimetry for Capsule V 6 10 65 Results of Thermal Neutron Dosimetry for Capsule V 6 11 66 Calculated Fast Neutron Flux and Lead Factors for s

Capsule V 6 12 67 Comparison of Measured and Calculated Fast Neutron Flux Levels within Capsule V 6 12 O

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3 SECTION 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in the first surveillance capsule from the Northern States Power Company Prairie Island Unit No. 2 reactor pressure vessel led to the s

following conclusions:

The capsule received an average fast fluence of 5.49 x 1018 neutrons /cm2 a

(E > 1 Mev). Predicted fa core cycle was 5.31 x 10g fluence for tge capsule at the end of the first neutrons /cm (E > 1 Mev).

e The fast fluence of 5.49 x 1018 n/cm2 (E > 1 Mev) resulted in a 55'F increase in the 50 ft Ib reference nil ductility transition temperature (RTNQ I of the weld metal. The intermediate pressure vessel lower shell forging 22u-.I exhibited a 15*F shift in the 50 f t Ib nil ductility transition temperature (specimens oriented normal to the major working direction of the forging).

The weid heat affected zone (HAZ) material exhibited a shif t of 45"F in transition temperature.

The average upper shelf impact energy of the weld metal exhibited a 3 ft Ib e

decrease in energy during the first core cycle. Both lower shell forging 22642 and HAZ material upper shelf impact energy increased.

The irradiated properties of forging 22642 and the weld metal are adequate to e

provide for continued safe operation of the Prairie Island Unit No. 2 power plant.

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SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule V. the first capsule of the continuing surveillance program which monitors the effects of neutron irradiation on the Northern States Power Company Prairie Island Unit No. 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Prairie Island Unit No. 2 reactor pressure vessel materials

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was designed and recommended by Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessei materials are presented in WCAP 8193. ld The surveillance program was planned to cover the 40 year life of the reactor pressure vessel and is based on ASTM E 185 73, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels."(21 Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the first capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory where the postirradiation mechanical testing of the Charpy V. notch impact and tensile surveillance specimens was performed.

This report summarizes testing and the postirradiation data obtained from the first material surveillance capsule (Capsule V) removed from the Prairie Island Unit No. 2 reactor vessel and discusses the analysis of these data. Using current methods,[3] heatup and cooldown (j

pressure temperature operating limits were established for the Prairie Island Unit No. 2 nuclear power plant. The heatup and cooldown pressure temperature operating limits are presented in appendix A.

1. Yan.chko, S. E.. Lege. D. J.. " Northern States Power Company Prairie Is!and unit No. 2 Reactor Vesset Radiation Surveinance Program," WCAP.8193 September 1973.

2. ASTM Designation E.185 73. "Surve Hance Tests for Nuclear Reactor Vesseis" in " ASTM Stancards (1974). Part 10 pp. 314-320. Am. Soc. for Testing and Meterials, Philaderphea, Pa. 1974

3. Hosenton. W. s.. Anderson. S. L.. and Yan chko. S. E.. " Basis for Hestuo and Coosdown Lim.: Curves." WCAP 7924 A.

.biv 1972.

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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 belt line 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 3 (ba'se material of Unit No. 2 reactor pressure vessel belt line) are well documented in the literature.. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decreast in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non Ductile Failure," Appendix G to Section lil of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics con-cepts and is based on the reference nil-ductility temperature, RT N DT-i

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RTNDT s defined as the' greater of either the dropweight-pil ductility transition temperature (NDTT per ASTM E 203).or tne temperature 60*F less than the 50 ft Ib (and 35 mils lateral expansion) temperature as determined from Charpy specimens oriented normal to the major working direction of the material. The RTNDT of a given material is used to index

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T 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 prersure vessel steel. When a given material is indexed to the KIR curve, allowable stress intenshy factors can be obtained for this material as a function of temperature. Allowable 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

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monitored by a reactor surveillance program such as tne Northern States Power Company 1

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Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program,Ill 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 50 ft Ib temperature (ARTNOT) due to irradiation is added to the original RTNDT o adjust the RTNDT 0'

t f

radiation embrittlement. This adjusted RTNDT (RTNDT nitial +aRTNDT) is used to index i

the material to the KIR curve and in turn to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

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1. Yenicnko. S. E.. Lege, o. J., " Northern States Po,.or Corrweay Prairie Islenc Unit No. 2 Reactor Vesset Rdiatro-Surveillance Program." WCAP 8193. Sootember 1973.

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SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Prairie Island Unit No. 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in figure 41. The vertical center of the capsules is opposite the vertical center of the core.

i Capsule V was removed af ter approximately 15/6 years of plant operation. This capsule con.

tained Charpy V notch impact, tensile, and wedge opening loading (WOL) fracture mechanics specimens from the lower shell ring forging 22642, weld metal from 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 correlation monitor material (A533 Grade B Class 1) furnished by Oak Ridge National Laboratory. The chemistry and heat treatment 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 forging. Test speci-mens 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 longi-tudinal axis of the specimen both normal to and parallel to the principal working (hoop) direction of the forging. The WOL test specimens were machined such that the simulated i

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crack of the specimen would propagate normal to (tangential specimens) and parallel to (axial specimens) the hoop direction of the forging. All specimens were fatigue precracked per ASTM E399 70T.

Charpy V notch specimens from the weld metal chamfer region 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. Table 4-3 lists the weld procedure and information associated with the Prairie Island Unit No. 2 core region l

weldments. Capsule V contained dosimeter wires of pure copper, iron, nickel, and l

aluminum 0.15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmiura :hielded dosimeters of Np 237 and U238 were contained in the capsule and located as shown in figure 4 2.

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Thermal monitors made from two low mel ing eutectic alloys and sealed in Pyrex tubes t

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 percent Ag. 97.5 percent Pb Melting Point - 579'F 1.75 percent Ag, 0.75 percent Sn, 97.5 percent Pb Melting Point - 590'F 4-2

TABLE 41 CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE CORE REGION LOWER SHELL FORGING AND WELD METAL FROM THE PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL Chemical Analyses (Percent)

Element *I Lower Shell 22642 Weld Metal [b]

I C

0.175 0.045 Mn 1.22 1.37 P

0.011 0.019 S

0.013 0.014 Si I

0.47 Ni 0.70 0.072 Cr 0.14 0.020 V

<0.008 0.001 Mo 0.445 0.51 Co 0.026 0.013 Cu 0.085 0.082 Sn 0.011 0.002 Al 0.036 0.007 N2 0.017 0.026 Heat Treatment Forging 22642 Heated at 1652/1715 F for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, water quenched; Tempered at 1175/1238'F for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, furnace cooled; Heated at 1652/1724 F for 51/2 hours, water quenched; Tempered at 1202/1238*F for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, furnace cooled; Stress relieved at 1022 F for 111/2 hours, furnace-cooled; Stress relieved at 1112*F for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, furnace-cooled Weldment Stress-relieved at 1022'F for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, furnace-cooled; Stress relieved at 1112*F for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, furnace-cooled

a. A qualetative spectrogrechic analysis nos made for elements greater than 0 010 weight percent.

b. Applicable weld were and fluu lot numbers are given in table 4-3 for chamfer filling.

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TABLE 4 2 CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL REPRESENTING 12 INCH THICK A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL Chemical Analysis C

Mn P

S Ni Mo Cu Ladle 0.22 1.45 0.011 0.019 0.22 0.62 0.53 Check 0.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14 Heat Treatment 1675 : 25 F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Air-cooled 1600 : 25'F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Water quenched 1125125'F - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - Furnace cooled 1150 2 25*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 TABLE 4 3 WELDING PROCEDURE AND ASSOCIATED INFORMATION FOR THE PRAIRIE ISLAND UNIT NO. 2 CORE REGION WELDMENTS *I I

Top and Bottom of Chamfer (b)

Automatic submerged arc welding with multiple passes.

Preheat of 400*F. Four passes on each side of chamfer.

Wire: UM 40 - 2.5 mm dia - fot: 3049 Flux: UM 89 lot: 1263 Welding Speed: 36 cm/ min Chamfer Filling Automatic submerged arc welding with multiple passes.

Preheat of 400'F.

Wire: UM 40 - 4 mm dia - tot: 2721 Flux: UM 89 lot: 1263 Welding Speed: 40 cm/ min

a. Weld groove - double U configuration D. The initial penetration passes made into the double U groove preoaration land (fsco).

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SECTION 5 TESTING OF SPECIMENS FROM CAPSULE V 5-1.

BACKGROUND INFORMATION The postirradiation mechanical testing of Charpy V-notch and tensile specimens was per-formed at the Westinghouse Research and Development Hot Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. Testing of the WOL test specimens was delayed upon request by Northern States Power Company. Completion of these tests is

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expected at a later date.

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 8193. UI No discrepancies were found except that the cadmium-covered dosimeter wires could not be located (paragraph 61).

Examination of the two low melting (579 F and 590 F) eutectic alloys indicated that the maximum temperature to which the test specimens were exposed during irradiation was less than 579*F.

A TMI Model TM 52004H impact test machine was used to perform tests on the irradiated Charpy V notch specimens. Before initiating tests on the irradiated Charpy V notch specimens,

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the accuracy of the impact machine was checked with a set of standard specimens obtained

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from the Army Material and Mechanics Research Center in Watertown, Massachusetts. The p

results of the calibration testing showed that the machina was certified for Charpy V notch impact testing.

The tensile tests were conducted on a screw-driven Instron testing machine having a 20,000-pound capacity. A crosshead speed of 0.05 inlmin was used. The deformation of the specimen was measured using a strain gage extensometer. The extensometer was calibrated before testing with a Sheffield high magnification drum type extensometer calibrator.

Elevated temperature tensile tests were conducted using a split tube furnace. The specimens were held at temperature a minimum of 30 minutes to stabilize the temperature prior to

1. Yan.cnko, S. E.. Lege. o. J.. " Northern States Power Cornpany Prairie Island unit No. 2 Reactor Vessel Radiation Surveinence Program", WCAP-8193, September 1973.

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testing. Temperature was monitored using a chromel alumel thermocouple in contact with the upper and lower specimen grips which were clevis pin type. Temperature was controlled within :5' F.

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

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

5-1.

CHARPY V NOTCH IMPACT TEST RESULTS The irradiated Charpy V notch specimens represented reactor pressure vessel belt line forging material, weld and heat affected zone (HAZ) material, and the ASTM correlation monitor material. The results are presented in tables 51 through 5-4 and figures 5-1 through 5 5.

The unirradiated data are also shown in figures 51 through 5 5 for comparison with the postirradiation data. A summary of the increase in the 50 ft Ib fix transition temperature and the decrease in the upper shelf energy is presented in table 5 5.

Test results obtained on the vessel belt line shed ring material are presented in table 51 and figure 51. The data showed that ring forging 22642 exhibited sensitivity to irradiation at 5.49 x 1018 2

n/cm based on the RTNDT shift of 30 F for the axist orientation. The upper shelf impact energy appears to have increased by 10 ft Ib energy.

The test results obtained on the weld material are presented in table 5 2 and figure 5 3.

The dasa showed that the fluence of 5.49 x 1018 n/cm2 received by Capsule V increased the RTNDT of the weld metal specimens by 55F and reduced the upper shelf impact energy by 3 ft lb. The test results for the HAZ material are summarized in table 5-3 and figure 5 4. The data showed that the HAZ material was also sensitive to irradiation at 5.49 x 1018 n/cm2 as increase in RTNDT was 45'F. However, the upper shelf impact energy increased by 19 ft Ib. The fracture and specimen appearance of all the Charpy specimens tested in the Prairie Island Unit No. 2 surveillance Capsule V are shown in figures 5 6 through 510.

5-2.

TENSILE TEST RESULTS The results of the tensile tests are presented in table 5 6 and figures 511 through 513.

Tests were performed on specimens from forging 22642 and the weld metal at room temperature, 300 F, and 550*F. Irradiation increased the yield and tensile strength of the forging by less than 10 percent and the weld metal by less than 16 percent. A typical i

load displacement curve obtained for the tensile tests is shown in figure 5-14. The broken I

tensile specimens from the surveillance capsule are shown ir figures 515 through 5-17.

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TABLE 5-1 CHARPY V NOTCH IMPACT DATA FOR THE PRAIRIE ISLAND UNIT NO. 2 PRESSURE VESSEL LOWER SHELL FORGING 22642 IRRADIATED AT 550*F, FLUENCE 5.49 x 1018 n/cm2 (E > 1 Mev)

Specimen Test Energy Lateral Shear Number Temp (*F)

(ft Ib)

Expansion (mils)

(%)

Axial (Weak) Orientation NT5 0

6.0 5.0 5

NT6 25 30.5 15.0 24 NT7 40 41.0 20.0 31 s

NT 11 40 26.0 20.0 21 NT3 60 44.0 30.0 33 NT2 70 52.0 30.0 38 NT4 70 56.0 30.0 38 NT9 110 72.5 50.0 54 NT1 140 89.5 80.0 64 NT 12 210 1

109.0 95.0 82 NT8 300 115.5 100.0 76 NT 10 300 121.5 100.0 72 Tangential (Strong) Orientation t

NL 12 0

6.0 10.0 7

NL1 25 62.5 20.0 47 NL 11 25 46.0 20.0 36 NL8 40 37.5 20.0 29 NL-9 40 80.0 35.0 62 NL2 70 100.5 50.0 69 NL3 70 91.5 40.0 59 NL5 110 129.5 70.0 82 N L 10 140 145.5 85.0 92 NL6 210 152.0 95.0 93 NL4 300 166.0 100.0 85 NL7 300 172.0 100.0 71 l

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TABLE 5 2 CHARPY V NOTCH IMPACT DATA FOR THE PRAIRIE ISLAND UNIT NO. 2 PRESSURE VESSEL WELD METpL IRpADIATED AT 550*F, FLUENCE 5.49 x 101 n/cm (E >1 Mev)

Specimen Test Temp Energy Lateral Shear Number

(* F)

(f t Ib)

Expansion (mils

(%)

NW8 50 8.0 10.0 7

NW4 0

32.5 30.0 33 NW3 25 64.0 40.0 55 NW5 40 64.5 35.0 54 NW2 70 84.0 70.0 64 NW6 140 92.5 100.0 73 NW7 210 iO1.0 100.0 82 NW1 210 98.0 100.0 83 TABLE 53 CHARPY V NOTCH IMPACT DATA FOR THE PRAIRIE ISLAND UNIT NO. 2 PRESSURE VESSEL WELD HEAT-AFFECTgn/cmZONg (METAL IRRADIAT AT 550'F, FLUENCE 5.49 x 10 E > 1 Mev)

Specimen Test Temp Energy Lateral Shear Number

(* F)

(ft Ib)

Expansion (mils)

(%)

NH7 148 9.5 5.0 10 NH3 76 42.0 10.0 31 NH8

-50 55.0 30.0 35 NH6 0

70.0 35.0 53 NH1 30 75.5 50.0 55 NH5 50 97.5 70.0 63 NH4 70 124.5 100.0 72 NH2 140 119.5 100.0 74 5-4

TABLE 5-4 CHARPY V NOTCH IMPACT DATA FOR THE PRAIRIE ISLAND UNIT NO. 2 A533 GRADE B CLASS 1 CORRELATION Mgln/cm TOR MATERIAL IRRADIATED AT 550*F, FLUENCE 5.49 x 10 (E >1 Mev)

Specimen Test Temp Energy Lateral Shear Number

(%)

(ft Ib)

Expansion (mils)

-(%)

R5 70 12.0 5.0 7

R8 140 18.0 30.0 16 R-3 175 30.0 40.0 26 R2 210 66.0 50.0 40 R7 210 52.5 50.0 32 R6 250 84.0 80.0 47 R1 300 104.0 100.0 61 R4 350 99.5 100.0 65 l

..~

P I

5-5

TABLE 5-5 THE EFFECT OF 550 F IRRADIATION AT 5.49 x 1018 n/cm2 (E > 1 Mev) ON THE NOTCH TOUGHNESS PROPERTIES OF THE PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL IMPACT TEST SPECIMENS Transstion Temperature i Fl Average Energy Aaneerposen at Fuel Shear til Ital Unervadeated Irrad ated

[7emperatisee ( F)

Mosersal 50 f t Its 30 f Its 5 mens 50 f t its 30 f Ib 35 no.as 50 ft sen 30 f t Its 3b mets Unerradiated seead.aeed (E nergy Forging 22642 5

-25 5

30 to 30 35 35 35 150 169

+19 (Tangenisat)

Forging 22642 35 0

25 65 30 62 30 30 37 108 118

• 10 (Ameal)

WWeld Mrtal

-40

-15 45 15 15 0

55 GO 45 103 100 3

HAZ Metal

-90

-135

-80 45 90 50 45 45 30 117 122

+5 Cmeelation 80 45 60 200 170 200 120 125 140 123 102

-21 Matereal t

U V

9

,m i

u.-

TABLE S-6 1RRADIATED TENSILE PROPERTIES FOR THE PRAIRIE ISLAND UNIT NO. 2 PRESSURE VESSEL LOWER SHgl FORGING AND WELD METAL, FLUENCE 5.49 x 10 n/cm2 (E > 1 Mev)

]

Ultimate Test Yield Tensile Fracture Fracture Uniform Total Reduction Specimen Temp Strength Strength Strength Stress Elongation Elongation in Area Material Number

( F)

(ksi)

(ksi)

(ksi)

(ksi)

( %)

(%)

(%)

Forging 22642 N L-1 75 73.32 91.65 51.5 217.5 10.8 25.4 75 (Tangential N L-2 300 65.99 77.90 53.5 199.2 11.0 23.8 73 Direction)

NL-3 550 62.12' 84.52 56.5 194.0 8.7 19.9 71 u,

4 Forging 22G42 NT-3 75 71.28 91.04 57.0 169.7 10.2 23.0 66 (Axial NT2 300 65.99 83.10 62.1 184.8 9.1 19.0 66 Direction)

NT-1 550 62.63 86.56 64.2 166.7 8.4 19.2 62

~

Weld Metal NW-1 75 76.17 84.01 53.0 187.0 10.4 22.8 72 NW-2 300 65.17 82.08 54.0 170.4 8.0 19.0 69 NW3 550 67.82 85.74 64.2 168.1 8.4 18.9 64

10.927-9 120 3

3 100 o

2 5

80 j

E

/

t 60

/

/

9

=

S/

3 40 5

20

$~2 0

l l

l 100

-x

[

80 g

]

5 9,- u

}

60 y

h o*,

w 40

/

2 J

Q 20 3

37oSHIFT IN 35-MIL TEMPERATURE 3

l l

l O

140 120 100 f

/

2

-m UNIRRADIATED 80 e

~

/.( IRR ADI ATED Ai 550 F 5

60 f

2 5.49 = 10'8 N/CM 3

(E >l Mn) 40 g

300 SHIFTS IN BOTH

/

30 AND 50 FT L8 20 -

/

TEMPERATURE j

O l

l j

O

-100 0

100 200 300 400 TEMPERATURE (OF)

Figure 51. Irradiated Charpy V. Notch impact Properties for the Prairie Island Unit No. 2 Reactor Pressure Vessel Lower Shell Forging 22642, Axial Orientation l

l l

5-8 1

10,927 il l

l

[

l 120 l

l3

'l l

'00 G,,g - a, 3

3 80 E

e' t

60

/

r 40 O 598 3

5 20 2

0 3

100 g.

E 80 et G

5 E

0 0

60 g

O 40 g

33 o

G W

a 20

,/

5 0

5 a

180 160 UNIRR AD I A t[D 140 120 0

/

100 9

1RR A01 ATED AT 550 F 18 2

b 5.49:10 N/CH 80

@ g/

g 3, g,g O

/

60 o

l 20

/

0

-100 0

100 200 300 400 TEMPERATURE (0F) l Figure 5 2.

Irradiated Charpy V-Notch Impact Properties for the Prairie Island Unit No. 2 Reactor Vessel Lower Shell Forging 22642, Tangential Orientation i

5-9

'l0,927-42 120 I,

l 3

100 G

G f

2 C

2 80 O

p O

/

y O

e e

60

/

2 40

$,/

O m

O O'

2:

20 2

/

8 0

100 g

g a

9 o

58

,To j

60 o

v5o b _ /

40

~

7 O O@

w g

20 f

{O/

f 0

120 O

2 l00 D

n a

$~$

/

5 80

/

60 IRR ADI ATED AT 550 F

/

o 5.49, o u fcs (c 3,,,g la 55 y

O 7

40 20 i

~

O/

I I

I I

-200

-l oo 0

100 200 300 400 TEMPERATURE (OF)

Figure 5 3.

Irradiated Charpy V Notch Impact Properties for the P airie Island Unit No. 2 Reactor Pressure Vessel Weld Metal 5-10

\\

/

10,927 13 120 I2 1

1 l-1 3

3 3

100 o

O y, c--o l

0 O/

80

=

t 60 O

0 O

/

=

l 40 -o 2

/e O2 O2 8 '

/

20 - e

1. • j 0

100 g

d l

3.-

00 2 -G 4

~

r o

60 E

2 40

-O

/

3, 5

e a

20 5

WO 1

U 0

(

3 140 UNIRR AD I ATED b-G Q

120 O

O 100 O

4 S

80 O

h 5

60 O

/e o

inaaoiaTEo av sso r 18 2

s.49st0 N/CM 45 5

40

-O W

(E >l Mev) 5 g)4-us o

20

- *y I

i l

I l

0

-200

-100 0

100 200 300 400 TEMPERATURE (CF)

Figure 5-4.

Irradiated Charpy V Notch impact Properties for the Prairie Island Unit No. 2 Reactor Pressure Vessel Weld Heat-Affected-Zone Metal 5-11

10,927-14 120 l

l l

3l 7

100

,G O -e 80 G

t

/

60

/

E O

W 40 9' 2 3

2 k

20 2

/

)

/

0 2

100 3.

80 d

-}

d O

a g

e0 f.-.

=

.a 40 0

2 tua g

d 20 g'

h 2

g#

3 0

140 0

120 o

100 M7 g

l

[

80

- UNIRR A)l A TE O

8 u.

[

60 IRRADI ATED AT g

120 y 550 F W

40 gO 30

/

5.49:10 i

125 ++

n/cu z 20 (E > i wev) 2g 35 0

-100 0

100 200 300 400 TEMPERATURE (OF)

Figure 5 5.

Irradiated Charpy V Notch impact Properties for the A533 Grade B Class 1 Correlation Monitor Material l

I l

l 5 12 1

10.927-19 I

I 1

NT-5 NT-6 NT-7 NT-l i j

i I

's w

s,,

N

+%*

W.

4 al.

' k 4k 1'.y) 1j t 4% g,.-

,..i

---+

i h

~ n...

J ' S '.'

r u

.7

, ',,,,f i h h' v.

5 NT-3 NT-2 NT-4 NT-9 I

(

. <y

',,c -

..r..g'

. *,, ',' n, v,..

,t'

. s#

Ebi g d dLA

J'r ,

.3 g., t

• b Y,k Ie $3

...;'EIUn' 4^'

',i 4., _ ^&

f

• h,

[ i. h;i.

i

~~

• • i;

• i: q,

.n.r. i.

% sp ~

l l

NT-l NT-12 NT-8 NT-10 Figure 5-6.

Charpy Impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Axial Orientation 5 13

1 10,927-20 l

l as as

. r =---

?.

NL-12 NL-l NL-il NL-8

.4,-

'*q

'#}

!E" g

f 4,

?.,. '%) j

• _,:.,,

NL-9 NL-2 NL-3 NL-5 l

s.v. g

. p<.w.

h

\\,g,

4r '

• i

.y

.1 * { y

'"' 5) Y

$. Y 'sg -

h{$.

h

~

A.

1.~,

.Q a g -N. '.

..c t:

-. ;r

;p*.

rr.

,,=:. ll

},- !

NL-10 NL-6 NL-4 NL-7 Figure 5-7.

Charpy impact Specimen Fracture Surfaces for Prairie i

Island Unit No. 2 Pressure Vessellower Shell Forging 22642. Tangential Orientation i

1 5-14

10.927-21 Md s

i in i

9 j

y l

~.

8 t#

l

E 'ab A

a

's NW-8 NW-4 NW-3 VW-5 i

l

%,, R.

.fe M

. g

.p

,w g

l 34

- }. :, "p; m {~ ; * :}s t z p, [

ey<

l ' -)

s'

~h p 4.{ f,\\ J, ', f [ y (Is l

k~)

$ f'[ k

[

, ~ t., y g'.t i'

Y% s '-

., L ~. Jnts...

ea NW-2 NW-6 NW-7 HW-1 Figure 5 8.

Charpy Impact Specimen Fracture Surfaces for Prairie I

Island Unit No. 2 Weld Metal I

f 5 15

10.927-22 s..,... ; s.

l

. hh,

l

., W,[;

..s;-

1, E

}

k-5~

Nj $.

-. $. 7},g' *i

~l'

.e

.,.. i '.

n ;.

s

-4 c4 l

NH-7 NH-3 NH-8 NH-6 I

f l

l

. * ;7 l_.[, f ' 'Y Nh'

' A,4.g i

[h

• k k

.7, A V

Y ty%

J-

f_i y;.

f t

Vgg JQu '

<yr'-

5 m

r. y y,.: Mr,.

iS,. 3.,

NH-l NH-5 NH-4 NH-2 l

Figure 5 9.

Charpy Impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 Weld Heat-Affected Zone Metal 5 16 l

10.927-23 l

l

,C.

7;.,: ;

C'

~

L.

V p..

I f,.

,d...

j

,( j '

4 7

n&. tk

,g.

\\

...n.

.-,v.

(.

R-5 R-8 R-3 R-2 l

l l

l l

l i

.t j S.

.. 1:

}

g a.

). T,.4

,Is.

O I

(

Q. W g 4'..

i 4-

.d

?) 8 7

l v

'g j.

f 2.,",.,g.s

'y.

c, r,

. & j % ;'-, ;*

,.<v f R-7 R-6 R -l R-4 Figure 5-10.

Charpy impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 A533 Grade B Class 1 Correlation Monitor Material 5-17

l 10.927-15 100 90 A%

Q ~ ~ ~,, - -

- A 80 U

7 U

E 70 A ~ ~.,,,,

- ULTIMATE TENSILE STRENGTN 60

' ' ' ' 'A e"

50 0.2 YlELD STRENGTH 14 0 LEGEND:

O UNIRRADIATED 30 A IRR AD I A TED. 5.i49 s 10 nj e,2 18 (E > 1 t'ev)

I I

I I

I I

~

20 80 70 b-------

s 7

_ _ _ _ yg _e 60

-g

- REDUCTION IN AREA wg 50 t

[ 40 TOTAL ELONGATION N

- UNIFORM ELONGATION 20 I-

---

A 10 -

A-~~----- k 0

-A l

I 1

I I

I g

O 100 200 300 400 500 600 700 TEMPERATURE (OF)

Figure 511. frradiated Tensile Properties for the Prairie Island Unit No. 2 Reactor Pressure Vessel Lower Shell Forging 22642, Axial Orientation l

l 5-18 i

10.927-16 100 A

90 g -[

80 70 A

ULTIMATE TENSILE STRENGTH

~=g og 7

60 6

<ay 50 0.2 YlELD STRENGTH LEGE ND:

i O UNtRR ADI ATED N n/cd IHA l ATED. S a9 IO 30 A

(E > l Me )

l i

I I

I i

<(

20 80 70 g- - - - -

n- - g - - - _

_ _ _ g g

n V

60 RE0ucTiON iN AREA O5 hi 50 t

40

(

')

3 TOTAL ELONGATION 30 a

G=

m

- e-

• • d 3

UNIFORMELONGATkN 20 io A- - - - - - - - -m- - -./

6 a

I I

I I

I l

O j

0 100

'P200 300 400 500 600 700 TEMPERATURE (CF)

Figure 5-12. Irradiated Tensile Properties for the Prairie Island Unit No. 2 Reactor P essure Vessel Lower Shell Forging 22642, Tangential Orientation t

l 5-19

l

=

IG.927-17 l

100 90 TE NS ILE S TRE NGirt 80

~ ~

- _ _ _ _ _ a A

4 p

G

~T a.

70 A

__-f

"~

o 60

~

0.2 flELD STRENGTH en 0

50 cc5 LE GE ND :

40 Q UNIRR ADI Ai[D 30 A IRRADI ATED. 5.49 10'd n/ c,2 (E > I Mew) 20 80 70

~ ~ ~ ~ ~ ~ ~ -h a-

~ ~

4- -g 60 RED ucil 0N IN AREA 7

z U

no

~

"a a.

40 U

TOTAL ELONG Ail 0N d

30 t;

n

___2______A4

=

~

_,,,,,,, <~<o,go;7 20 c,

F o

e 10 A

__J___.g__________A l

l l

l l

l n

0 100 200 300 400 500 600 700 TEMPERATURE ( F)

Figure 513. Irradiated Tensila Properties for :he Prairie Island Unit No. 2 Reactor Pressure Vessel Weld Metal 5 2C1 f,

\\

1 r

/

s

(

o 10.927-68 o

n.o N

=

m.

~

o o x

I w

.J 2

~

N w we O

C

• o<

=

. e_

c 8_

=

- U x m.

m o

o 5

R

'a C

-7 4

o o o

v

=

w w a

. 6 3

o w u

l C

z

5 N O w

-- 5 65 o

a z'

E e <

m 9g 2

o n

.9 s

e e E o,

(

)

o er

_=

~-

o m ex 2

o.

w s

.?

u.

l l

I I

I I

l l

oe d

o o

o o

o o

o e

o e

e

=

~

=

m e

m V38V 7VN19180/0V01 - ISM 'SS381S ON1833N10N3 i

l i

6 5 21

s 10.927-24

. N.,;' }.',

8

I

~

g.er.r ap.,

a '.

• 1 SPECIMEN NT-3 3

5

'sa$$ '

• ji. A l%*w;y ar'

{.Yd'IhfIkh'ND 7.IN

'4

( +

%!U$?8hdbi',"h*,.k;?.3{![j!!D'ig.

'2

-l.

• h.

• k $,

SPEClHEN HT-2 mRIBM h-

&,r "y,Q) gq ls

. (:4.

4h/k* -

l%*

rN'h.tl.

,d.

SPECIMEN NT-1 i

ll Figure 515 Fractured Tensile Specimens from Prairit! Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Axial Orientation I

S 22

i

\\.

1 l0.927-25 N.*

T.VP. y? 'i:

'.l ' ",f' f.$'.f.G Q,^

NNS IN[,DE' h;;W"k If i

i SPECIMEM ML-l

{

l j

s 1

,,t...

wy.y l

.,4 :n. ce F-s. ;. y,;e..,.

.&...gif.d'.%,F' w;' f' fw(af..i c

v.

...o.,.

x e

SPECIMEN NL-2 I

{

h

. ^*

^

/

'y s, y..

-f f. ;.,,) <,

/..<f *,. ep j 9 3.-

? ' S(WA' 5' ' hk k. [t$.';;

. testyt sist,4 Pre

=

wgy 7

,. %.... d

~.

,.g,T 3

, ( 9,' ; if ass.h i'@ 1:., #.

.s...

' wi v

s.

'i N'

. g'

..A

[g :j{ggy.g'pg' h i';[-

SPECIMEN NL-3 Figure 516.

Fractured Tensile Specimens from Prairie is!and Unit No. 2 Pressure Vessel Lower Shell Forging 22642. Tangential Orientation 5 23

- u. w i-e e 1

SPECIMEN W-1 a CA;C W,..v e

" C",

1g* IUD.2$.%$7'"

?36 SPECIHEN W-2 i

'; ~ !%'. '

f

• ',3

• a f,9s..;..:.>% (..

. ns;w

)Nb fd"2D'ib SPECIMEN W-3 Figure 5-17.

Fractured TensS. Specimens From Prairie Island l

Unit No. 2 Pres 3me veve; Weid Metal 5 24

SECTION 6 DOSIMETRY ANALYSIS 6 1.

FAST NEUTRON FLUX MONITORS To effect a correlation between neutron exposure and the radiation induced property changes observed in the test specimens, a number of neutron flux monitors were included as an integral part of the reactor vessel surveillance program. The particular monitors contained within Capsule V, along with the nuclear reactions of interest and the energy range of each

\\

monitor, are listed in table 61. The first five reactions listed in table 61 describe fast neutron monitors which were used to relate neutron fluence (E > 1.0 Mev) to the measured shif t in RTNDT. The bare and cadmium covered Cobalt aluminum monitors were included to measure thermal neutron flux and thus enable an assessment of the effects of burnup of the product isotopes on the response of the fast neutron monitors.

The relative locations of the various monitors within Capsule V are shown in figure 4 2 while the radial and azimuthal position of the capsule with respect to the nuclear core, reactor internals, and pressure vessel is illustrated in figure 61. The iron, nickel, copper, and cobalt aluminum monitors were in the form of wires placed in holes drilled in spacers at several axial levels within the capsule. The cadmium shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the

)

capsule. N

^

x.s The use of activation detectors, such as those listed in table 61, does not yield a direct measure of the energy dependent neutron flux level at the point of interest. Rather, the activation 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 detectors may be derived from the activation measurements only if the parameters of the irradiation are well known. In particular, the following variables are of interest:

The operating history of the reactor a

a The energy response of the monitor

i. Tn. c om um-in..ioeo ooi.m.ter. r s for caniui. v

.. iri.overeenity not insert.o ouring veo,,c i on or in, e.osui..

6-1

TABLE 6-1 NEUTRON FLUX MONITORS CONTAINED WITHIN CAPSULE V Wt% of Target Monitor Reaction of f 9* target i Product Material Interest (9"moniton Response Range Half-Life Copper Cu 63 (n,n) Co6 0.6917 E >4.7 Mev 5.27 years 64 iron Fe 64 (n.p) Mn 0.0585 E >1.0 Mev 314 days Nickel Nisa (n.p) Co 58 0.6777 E > 1.0 Mev 71.4 days Uranium 238 lal U238 (n,f) Cs '3' 1.0 E > 0.4 Mev 30.2 years Neptunium 23r lal Np 237 (n,f) Cs'3' 1.0 E > 0.08 Mev 30.2 years Cobalt-Aluminum "I Co 68 l

(n,y) Co*

0.0015 0.4 Mev < E ' O.015 Mew 5.27 years 68 6

Cobalt-Aluminurr, Co (n,y) Co 0.0015 E < 0.015 Mev 5.27 years

a. o.-notes sh.at 'nonstor es cadmium shielded b

e O

d

...S I

(

10.927 4 oo 130

////I l

\\

j///lll l

'"ES SURE vgssgz r////

E'*Suli

,,////

y f' / / /

'"E*"A

%Y L s f

n gto u

7/I//P

/

l/

l /

ll7//',///'l l

/

/

/

I

/ / ', //

/

/

/

REACTOR

,s I

i CORE

/

/

/

/

/

/

/

/

/

l Figure 61. Prairie Island Unit No. 2 Reactor Geometry 63

The neutron energy spectrum at the monitor location e

The material composition of the monitor a

6 2.

ANALYTlCAL METHODS The analysis of the activistion detectors and the subsequent derivation of the average neutron flux requires that two procedures be completed. First, the disintegration rate of product isotope per unit mass of detector must be determined. Second, in order to define a suitable spectrum averaged reaction cross section, the neutron energy spectrum at the detector loca-tion must be calculated.

The energy and spatial distribution of neutron flux within the Prairie Island Unit No. 2 reactor geometry were obtained with the DOT UI two dimensional S transport code. The n

radial and azimuthal distributions were obtained from an R,0 computation wherein the reactor core, reactor internals, pressure vessel, primary concrete, water annuli, and the sur-veillance capsule itself were described in the analytical model. These analyses employed 21 neutron energy groups, an Sg angular quadrature, and a Pj cross section expansion. The reactor core power distribution used in the calculations was representative of time averaged conditions over an equilibrium fuel cycle and accounted for rod by rod spatial variations in the peripheral fuel assemblies. The analytical geometries described a 45' sector of the reactor, assuming 1/8 core symmetry, Relative axial variations of neutron flux incident on the reactor vessel were obtained from R,z DOT calculation using the equivalent cylindrical core concept.

The specific activity of each of the activation monitors was determined using established ASTM procedures.[2.W,6) Having the measured activity of the monitors and the neutron

1. Soltess R. G., et al., " Nuclear Rocket Sh'eiding Methods. Modification, updating and input Data Preparation.

Volume 5 Two Dirnensional. Descrete ordinates Transport Technique," WANL PR-(LL) 034, August,1970.

2. ASTM oesignation E26170. " Standard Method for Measuring Neutron Flus by Radioactivation Technicues."

in ASTM Standards. Parr 45, (1975), pp. 745 755, Am. Soc. for Testing and Materials. Philadelphia, Pa.,1975.

3. ASTM oes.gnation E262 70. " Standard Method for Measuring Thermat Neutron Flux by Radioactwation Techruques."

in ASTM Srsaddeds. Part 45.11975 pp. 756 763. Am. Soc. for Testing and Materials, Philadelphia, Pa 1975.

4 ASTM Designation E263 70. " Standard Method for Measuring Fast Neutron Flus by Radioactivation of Iron /

in ASTM Standards, Part 45. (1975). pp 764 769. Am Soc. for Testing and Materials, Philadelphia, Pa.,1975 S. ASTM Designation E48173T. " Tentative Method of Measuring Neutron Flus Density by Radioactivation of Cobalt and Saver." in ASTM Srsaderds. Port 45, (1975), pp. 887-894, Am Soc. for Testing and Materiais.

Philadeichia. Pa. 1975.

6 ASTM Designation E264 70. " Standard Metnod for Measuring Fast Neutron Flux by Radioactivation of Nickel,"

.n ASTM Standards. Perr 45. (1975), pp. 770 774 Am. Soc. for Testing and Materials. Philadeichia. Pa.,1975.

64

energy spectrum at the location of interest, the calculation of the fast neutron flux pro-ceeded as follows. The reaction product activity in the monitor was expressed as g),Arj} e Ard J

D=

f; y a(E) o(E)

(6 1)

E max J=1 where D

= induced product activity l

N

= Avagadro's number o

A

= atomic weight of the target isotope f

= weight fraction of the target isotope in the target material t

y

= number of product atoms produced per reaction a(E)

= energy depe:, dent reaction crosssection I

o(E)

= energy dependent neutron flux at the monitor location with the reactor at full power F

= average core power level during irradiation period J J

l P

= maximum or reference core power level max A

= decay constant of the product isotope rj

= length of irradiation period J rd

= decay time following irradiation period J l

~'

Because neutron flux distributions were calculated using multigroup transport r9ethods, and

)

V further, because the prime interest was in the fast neutron flux above 1.0 Mev, spectrum averaged reaction cross sections were defined such that the integral term in equation (61) could be replaced by the following relation.

o(E) c(E) = F o (E > 1.0 Mev)

E where n

t a(E) c(E) o,,,

G=1 y.

n o(E) 4 1.0 Mev G=G.0 Mev 1

65 l

I Thus, equation (61) was rewritten n

D=

f; y J o(E > 1.0 Mev)

(1e'ATJ), Ard p

J=1 max or, solving for the neutron flux D

o(E) > 1.0 Mev) =

N P

o J

(6 2) f O

Ti)U (j. e J) e d

pmax The total fluence above 1.0 Mev was then given by o(E > 1.0 Mev) = o(E > 1.0 Mev) p r;

(6 3) max J=1 where P

P rj = total effective full power seconds of reactor operation up max to the time of capsule removal Beca'use the cadmium covered cobalt aluminum wires were not recovered, a direct assessment of the thermal neutron flux levels within Capsule V was not possible. Thus, it was assumed that the average thermal flux measured within Capsule V of Prairie Island Unit No.1 Ill was applicable to Capsule V of Priairie Island Unit No. 2. Because Units 1 and 2 are identical reactors with the same maximum rated power level, and the surveillance capsules in question were located at the same relative position within the pressure vessel, this assumption should be valid.

The irradiation history of the flux monitors removed from Capsule V is listed in table 6 2.

The data were obtained from the Prairie Island semiannual operating reports and monthly l

data tables (Docket 50282 343, 401, 493, 484, 501, 508, 551, 580, 599, 617, 636). The spectrum averaged reaction cross sections derived for each of the fast neutron flux monitors I

are listed in table 6-3.

I

1. cavidsor. J. A.. Anderson. S. t... and Scott. K. V.. ** Analysis of Caosute V from Northern States Power Company.

Prairie Island Unit No.1 Reactor Vesses Radiation Survsillance Program." WCAP-8916, August 1977.

66

TABLE 6 2 IRRADIATION HISTORY OF CAPSULE V REMOVED FROM PRAIRIE ISLAND UNIT NO. 2 Irradiation Decay *I I

P P,,

P Time Time j

g J

Month (MW)

(MW)

PMax (days)

/ ys) 12'74 125 1650 0.076 15 804 1.'75 797 1650 0.483 31 773 2/75 1429 1650 0.866 28 745 3/75 992 1650 0.601 31 714 4/75 1518 1650 0.920 30 684 5/75 1567 1650 0.950 31 653 6/75 304 1650 0.184 30 623 7/75 1245 1650 0.755 31 592 8/75 1542 1650 0.935 31 561 9/75 1422 1650 0.862 30 531 10/75 1258 1650 0.762 31 500 11/75 1574 1650 0.954 30 470 12/75 837 1650 0.507 31 439 1/76 496 1650 0.301 31 408 2/76 1518 1650 0.920 29 379 3/76 1481 1650 0.898 31 348 4/76 1335 1650 0.809 61 287 v'

5/76 6/76 1091 1650 0.661 30 257 l

7/76 1606 1650 0.973 31 226 l

8/76 1628 1650 0.987 31 195 9/76 1626 1650 0.985 30 165 10/76 1376 1650 0.834 21 144 Oecay time is referenced to the counting date of the neutron fium monitors (3/14/77) a.

6-7

TABLE 6 3 SPECTRUM AVERAGED REACTION CROSS SECTIONS USED IN FAST NEUTRON FLUX DERIVATION Reaction F (Barns)

Fe ' (n.p) Mn '

O.0580 5

5 Nisa (n.p) Co sa 0.0840 Cu 63 (n.n) CdSO 0.000428 U238 (n,f) F.P.

0.329 237 Np (n.f) F.P.

2.98 6 3.

RESULTS OF ANALYSIS The fast neutron (E > 1.0 Mev) flux and fluence levels derived from the monitor $ taken from Capsule V are presented in table 6 4. In examining the data listed in table 6 4, it should be noted that the Fe 54 monitors were positioned within the surveillance capsule at a radius of 158.72 cm relative to the core centerline. The corresponding radius of the U238 237 and Np monitors was 158.95 cm, and that of the Ni sa and Cu 53 monitors was 159.72 cm. Thus, it should be expected that the measured neutron flux levels reflect the flux gradient caused by attenuation within the test specimens.

As stated earlier, the average thermal neutron flux within the surveillance capsule was taken from the dosimetry analysis of Prairie Island Unit No.1, Capsule V llI. The value of thermal flux from WCAP 8916 Ill is 1.51 x 1011 2

n/cm /sec. Due to the relatively low I.

thermal neutron flux at the monitor 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 Nisa (n,p) Co 58 reaction and even less significant for all

(

of the other fast reactions. The measured activities of the bare cobalt aluminum wires are given in table 6 5.

Results of the S transport calculations for the Prairie Island Unit No. 2 reactor are sum-n marized in figures 6 2 through 6 4 and in table 6-6. In figure 6 2, the calculated maximum fast neutron flux levels at the pressure vessel inner radius,1/4 thickness location and

1. oavidson. J. A., Anderson. S. L. and scott. K. V.. "Analys.s of Capsule V f rom NortNern states Power Company.

Prairie Island Unit No.1. Reactor Vessel Radiation Surveillance Program," WCAP 8916. August 1977.

68

s 3/4 thickness loct. tion are presented as a function of azimuthal angle. The relative axial variation of neutron flux is shown in figure 6 3. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6 2 by the appropriate values from figure 6-3. In figure 6 4, the predicted maximum end of life fast neutron exposure of the Prairie Island Unit No. 2 reactor vessel is given as a function of radial position within the vessel wall. The calculated fast neutron flux levels interior to Capsule V along with the lead factors (LF) relating capsule exposure to vessel exposure to vessel exposure are listed in table 6 6. The lead factor is defined as the ratio of the calculated flux at the rrtonitor location to the calculated peak neutron flux incident on the reactor vessel.

In table 6 7, comparisons of calculated and measured fast neutron flux levels at the front monitor, dosimeter block, and rear monitor locations within Capsule V are presented.

6 4.

DISCUSSION OF RESULTS Using the iron data presented in table 6-4, the average fast neutron fluence at the front flux monitor location is determined to be 6.18 x 1018 2

n/cm ; using the nickel data, the average fast fluence at the rear flux monitor location is 4.79 x 1018 2

n/cm. These measured values correspond to analytical values of 5.92 x 1018 and 4.70 x 1018 n/cm2 at the front and rear locations, respectively, and result in an average fluence for the cap-sule of 5.49 x 1018 2

n/cm Agreement between calculation and measurement is excellent.

By employing the lead factors listed in table 6-6, a comparison of the end of life peak fast neutron exposure of the Prairie Island Unit No. 2 reactor vessel, as derived from both calculations and measured surveillance capsule results, may be made as follows:

ccalculated

= 4.3 x 1019 2

n/cm cfront monitors = 4.5 x 1019 n/cm2 crear monitors

= 4.4 x 1019 n/cm2 These data are based on 32 full power years of operation at 1650 MWt.

t l

I 6-9

TABLE 6 4 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE V F

Reaction and Measured *I I

Monitor Activity 3 (E > 1.0 Mev)(b) o (E > 1.0 Mev)lbi 2

2 Location (dps/gm)

(n/cm /sec)

(n/cm )

Fe54 (n p) Mn58 Bottom 2.49 x 106 11 1.48 x 10 6.50 x 1018 6

Bottom Middle 2.43 x 10 1.44 x 1011 6.33 x 1018 6

Middle 2.34 x 10 1.39 x 1011 6.11 x 1018 Top Middle 2.21 x 106 11 1.31 x 10 5.76 x 1018 Top 2.37 x 106 1.41 x 1011 6.20 x 1018 Nisa (n p) Co sa Middle 1.40 x 107 1.09 x 101I 4.79 x 1018 Cu 53 (n a) Coso 4

Bottom Middle 6.46 x 10 1.47 x 10ll 6.46 x 1018 4

Top Middle 5.90 x 10 1.34 x 10ll 5.89 x 1018 Np237 (n f) Cs'37 6

ll Middle 2.25 x 10 1.55 x 10 6.81 x 1018 U238 (n.f) Cs137 l

Middle 2.62 x 105 1.56 x 10ll 6.86 x 1018

a. Monitor act.v. ties are ref erenced to 12 00 pm on 3/14/77.

b.

Derived fium and fluence levels are subsect to 110 percent measurement uncertainty.

i f

I l

6-10

e i

TABLE 6 5 l

l RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE V Bare 'I Cadmium Covered *I g bi l

I I

Monitor Activity Activity 2

Location (dps/gm)

(dps/gm)

(n/cm /sec)

Bottom 2.02 x 107 Not Measured Not Determined 7

Top 2.15 x 10 Not Measured Not Determined a Co60,,,,,,,,,,,,e referenced to 12 00 cm on 3 '14'7 7.

The average thermgi neutron flus measured within Capsule V from Prairie Island Unit No. * *as b

1 $1 a 10 a em /sec.

'w 6 11

a' TABLE 6 6 CALCULATED FAST NEUTRON FLUX AND LEAD FACTORS FOR CAPSULE V Location c (E > 1.0 Mev)

Within 2

Capsule V (n/cm /sec)

Lead Factor Front Montiors 1.35 x 1011 3.18 Dosimeter Block 1.29 x 1011 3.03 Rear Monitors 1.07 x 1011 2.52 TABLE 6 7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX LEVELS WITHIN CAPSULE V 2

c (E > 1.0 Mev) (n/cm /sec)

Front Rear Monitor Dosimeter Monitor Location Block Location Calculated 1.35 x 1011 1.29 x 1011 1.07 x 1011 L,

Fe" (n,p) Mn" 1.41 x 1011 Nisa (n,p,) Cosa 1.09 x 101I Cusa (n,o) Co6 1.41 x 101l Np237 (n.f) Cs'37 1.55 x 101I U23a (n,f) Co'37 1.56 x 1011

a. F'un values are everage for att samples l

6-12

N i

s 10.927-5 10

8 6

4 POESSURE VEbSEL 2

INSID[ RA0luS Cu N e 1/4T LOCATION d.

1010

=a f

8

=

E 6

E W

4 3/'*T LOC ATION i

2 l

l l

9 I

I I

10 O

10 20 30 40 50 AZIMUTHALANGLE(DEGREES)

Figure 6-2. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Prairie Island Unit No. 2 Reactor Vessel 6-13

10.927-6 1.0 8

2 6

4 2

0.1 s

3 S

m 6

E

~

4 wx W

2 5

0.01

~

8 6

m

~

f i4 s

1 2o" 2

TOWARD OPERATING TOWARD SUMP DECK t

l l

0.001

-300

-200

-100 0

100 200 300 AXIAL DISTANCE FROM CORE HIDPLANE (CH)

Figure 6 3.

Relative Axial Variation of Fast Neutrnn Flux (E > 1.0 Mev) incident on the Prairie Island Unit No. 2 Reactor Vessel 6-14 l

l b

• =.

t

. r e

10.927-7 1020 8

VISSEL in$lDF RADIUS 167.64 6

n

{

2 x2 w

E 1019 VCSSEL OUTSIDE d

3 RADIUS 184.15 zg b

E N

W I

y 2

iai8 I

I I

I I

I I

I I

I 166 168 170 172 174 176 178 180 182 184 186 RADIUS (CM)

Figure 6-4.

Calculated Maximum End of Life Fast Neutron Fluence (E > 1.0 Mev) as a Function of Radius Within the Prairie Island Unit No. 2 Reactor Vessel l

6 15

{

e APPENDIX A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A 1.

HEATUP AND COOLDOWN LIMIT CURVES IN ACCORDANCE WITH ASME BOILER AND PRESSURE VESSEL CODE, SECTION lli Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil ductility temperature). The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material preperties and estimating the radiation induced aRTNDT.RTNDT s designated as the i

higher of the dropweight nil ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft Ib of impact energy and 35 mils lateral expansion (normal to the major working direction) minus 60 F.

RT 'DT ncreases as the material is exposed to fast neutron radiation. Thus, to find the i

N most limi+ing RTNDT at any time period in the reactor's life, a ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. The extent of the shift in RTNDT s enhanced by certain chemical elements (such i

as copper) present in reactor vessel steels. Design curves which show the effect of fitence and copper content on ART f

NDT or reactor vessel steels exposed to 550 F are shown in figure A 1. Other elements such as phosphorous are not considered by Westinghouse to have

~

any appreciable influence on the shift in RT s

N OT-Given the copper content of the most limiting material, the radiation induced SRT NDT can be estimated from figure A 1. Fast neutron fluence (E > 1 Mev) at the vessel inner surface, the 1/4T (wall thickness) and 3/4T (wall thickness) vessel locations are given as a function of full power service life in figure A 2. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to assure that no other component will be limiting with respect to RTN DT-A 2.

FRACTURE TOUGHNESS PROPERTIES The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the ASME Boiler and Pressure Vessel Code.N

1. Append.: G

" Protection Against Nonductile Failure" of ASME Boiler and Pressure Vessel Code. Section lil, I

"Nucteer Power Plant Components." and Summer 1975 Addenda. American Society of Mecnanical Engineers, New York,1974 I

A1 l

10.927-i 400 0.30 COPPEP BASE. 0.25 WELD AND OVER 3.25' COPPER BASr. 0.20' WELO 02.0 COPPE'* BASF 0.i5 Wi LO 200

.. s

)

100 g

80 C

~

60 E

z; 0.15 0 0PPr.4 BASE. 0.10 WELO 4

40 O.10 00PPE 3ASE 0.05' WELO 20 I

I I

I I

I I

I 10 20 10'8 2

4 6

8 10'9 2

4 6

8 10 2

Fl_UENCE (N/CH > I.0 Mev) 1 l

l Figure A 1.

Effect of Fluence and Copper Content on Shift of RTNDT fr Reactor Vessel Steels Exposed to 550*F Temperature A2

e 10.927-2 3

2 t/si 10

8

~

3/ui 4

^m S_

g 2

5 5

w h

10'8 3

8 r,

6

\\

..)

4 l'

2 l

l l

10'7 0

5 10 15 20 25 30 SERVICELIFE(EFFECTIVEFULLPOWERYEARS)

Figure A 2.

Fast Neutron Fluence (E > 1.0 Mev) as a Function 1

of Full Power Service Life A-3

I TABLE A 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

Transverse!*I 50 f: Ib/35 mits Cu P

NDTT Lateral Expansion NDT Average Transverse!'I RT Component Material Type

(%)

(%)

( F)

Temp ( F)

F)

Upper Shelf (f t Ib)

Cosure Head Dome A533 Gr. 8, Cl.1 5

52ICI ICI 5

64 ICI ICI Head Flange A508 Cl. 3

-31 18 31 87 ICI ICI Vessel Flange A508 O. 3 22 18 22 88 ICI ICI Injection Nozzles A508 Q. 3 22 114 22 97 ICI ICI Inlet and Outlet Norile A508 C. 3 13 50 10 89 ICI Upper Shell A508 O. 3 13 41 I

13 85 'I ib]

inter. Shell A508 O. 3 0.075 0.010

-4 56

-4 112 Lower ShellIDI A508 O. 3 0 085 0.011 13 54 6

108 Trans. Hing A508 Cl. 3 10 50 10 76 ICI Bottom Head A533 Gr. 8 Cl.1 13 56 4

68ICI Weldment Weld 0.082 0 019

-31 6

-31 103 IfAZ HAZ 31 35 31 117 Specernen oriented norrnal to the rnago workeng detection a.

h. Based on actual transwerw data ttwouqh the surveettasse suurparn Estermted useing "Picssure Tenverature Lerrats.** Sect.on 5.3.2 of Standard Rewcw Frarr. NUREG 75/087 c.

1975. Isom long.tudinal data.

s f e 4

O

~

v

~.. -

/

and the calculation methods of WCAP 7924Ill. The preirradiation fracture toughness properties of the Prairie Island Unit No. 2 reactor vessel materials are presented in table A 1. The post-irradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Prairie Island Unit No. 2 Reactor Vessel Material Surveillance Program.

A 3.

CRITERIA FOR ALLOWABLE PRESSURE TEMPERATURE RELATIONSHIPS Allowable pressure temperature relationships for various heatup and cooldown rates are calculated using methods derived from non mandatory appendix G in section lli of the ASME Boiler and Pressure Vessel Code. and are discussed in detail in WCAP 7924llI.

l The approach specifies that the allowable total stress intensity factor (K ) at any time during g

heatup or cooldown cannot be greater than that shown on the KIR curve for the metal temperature at that time. Furthermore, the approach applies an explicit safety factors of 2.0 and 1.2512) on tre stress intensity factors induced by pressure gradients. Thus, the

[

governing equation for the heatup-cooldown analysis is:

l 2Kim + 1.25 Kit < KIR I A'll where Kim = the stress intensity factor caused by membrane (pressure) stress i

Kit = the stress intensity factor caused by the thermal gradients KIR = provided by the code as a function of temperature relative to the RTNDT of the material

'\\

/

From equation (A 1) the variables that effect the heatup and cooldown analysis can be readily identified. Kim is the stress intensity factor due to membrane (pressure) stress. Kit is the thermal (bending) stress intensity factor and accounts for the linearly varying stress in the vessel wall due to thermal gradients. During heatup Kit is negative on the inside and positive on the outside of the vessel wall. The signs are reversed for cooldown and therefore an ID or an OD one quarter thickness surface flaw is postulated in whichever location is more limiting. K IR is dependent on irradiation and temperature and therefore the fluence profile through the reactor vessel wall and the rates of heatup and cooldown are important. The procedure used to account for these variables is explained in the following text.

1. Heretton. W. S., Anderson. S. t.. and Yanichko. S. E., " Basis for Hestuo and Cooldown Limit Curves /*

WCAP 7924, July 1972.

2. The 1.25 nefety factor on K represents seditional conservat:en ecove code reovirements.

g A-5

Following the generation of pressure temperature curves for both the steady state (zero rate of temperature change) and finite heatup rate situations, the final limit curves are produced in the following fashion. First, a composite curve is constructed based on a point by point comparison of the steady state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the two values taken from the curves under consideration. The composite curve is then adjusted to allow for possible errors in the pres-sure and temperature sensing instruments.

The use of the composite curve is mandatory in setting heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling analysis switches from the OD to the ID location. The pressure limit must, at all times, be based on the most conservative case.

The cooldown analysis proceeds in the same fashion as that for heatup with the exception that the controlling location is always at the ID. The thermal gradients induced during cooldown tend to produce tensile stresses at the ID location and compressive stresses at the OD position. Thus, the ID flaw is clearly the worst case.

As in the case of heatup, allowable pressure temperature relations ar3 generated for both steady state and finite cooldown rate situations. Composite limit curves are then constructed for each cooldown rate of interest. Again adjustments are made to account for pressure and temperature instrumentation (rror.

The use of the composite curve in the cooldown analysis is necessary because system con-trol is based on a measurement of reactor coolant temperature, whereas the limiting pressure is calculated using the material temperature at the tip of the assumed reference flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the b

vessel ID. This condition is, of course, not true for the steady state situation, it follows that the aT induced during cooldown results in a higher calculated KIR for finite cooldown rates than for steady state under certain conditions.

(

A 4.

HEATUP AND COOLDOWN LIMIT CURVES j

Limit curves for normal heatup and cooldown of the primary reactor coolant system have l

been calculated using the methods discussed in the previous section. The derivation of the limit curves is presented in " Pressure-Temperature Limits."III l

1. "P essure Temocrature Limits." Section 5.3.2 of Stenderd Atwow Asn, NuREG 7s/c87.1975.

A-6

." t c.

Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program.

Weld metal Charpy test specimens from Capsule V indicate that the core region weld metal exhibits the largest shift in ATNDT (55*F). This shift at a fluence of 5.49 x 1018 2

n/cm,

is well within the design curve (figure A 1) prediction. Heatup and cooldown calculations were based on the ARTNOT given by this trend curve.

Heatup and cooldown limit curves for normal operation of the reactor vessel are presented in figures A 3 ano A 4 and represent an operational time period of 32 effective full power years.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves.

The reactor must not be made critical until pressure temperature combinations are to the right of the criticality limit line shown in figure A 3, in addition to other criteria which must be met before the reactor is made critical.

The leak test limit curve shown in figure A 3 represents minimum temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods in Appendix G, to the side and " Pressure Temperature Limits."h 21 Figures A 3 and A 4 define limits to assure prevention to nonductile failure.

A 5.

SURVElLLANCE CAPSULE REMOVAL SCHEDULE To date, Capsule V has been removed and the encapsulated specimens have been examined.

Based on the postirradiation test results on Capsule V, tne recommended removal schedule for the remaining capsules in the reactor vessel is contained in table A 2.

~

i l

1. Appendia G

" Protection Against Nonductile Failure" of ASME Boi er and Pressure Vesset Code. Section Ill.

l j

"Nucieer Power Plant Corroonents." and Summer 1975 Addenda. American Society of Meenanical Engineers.

New York.1974.

2. " Pressure-Temperature Limits.** Section 5.3.2 of Standara necew Fr n. NuREG-75/c87.1975.

a

-7

-4,

e TABLE A 2 l

SCHEDULE FOR REMOVAL OF SPECIMEN CAPSULES Factor by which Capsule Capsule Leads Vessel Maximum Identification Exposure tal at the Surface Removal Time [b,1 T

1.6 5 years (postirradiation test)

R 2.5 10 years (postirradiation test)

P 1.6 25 years (postirradiation test)

S 1.4 Standby i

N 1.4 Standby Lead f actors for the Prairie Island Unit No. 2 surveillance caosules have been re evaluated since the original radiation 3

surveillance report (WCAP41931 had been issued. The updated values for the Prairie it'and Unit No. 2 surveillance caosules are presented in this table.

b From date of initial olent startup, b

N

.w 6

l e

T

$' 4 k.l 1

A-8 Js

e g..

's.

e" r

. ?;,*

10,927-27 s

r

\\

e

' r s

3000 2: uA TERI AL PROPERI" ?AS t @

t l

, 2800, _2h HELD HETAL Cs = 0.Ct2 LEaa IES T LIMIT 2600 ' C IN I T I AL R T"D T =.26 i,CONSERV A TivELY i

I ASSUMEDI Z:

AT 32 EFFECTIVE FULL POWER YEMS']

~~

l, 2400 R T,3 7 AT i/ui.'THICaNESS = 110 F i I

~

[

RT T 200 NDT AT 3/uf THICNNESS 2 @ _(

_i j

f

[

L, y

j j

m

• ~I t.2000 1

I

/

1 l

i I

t T

I kJ E 1800

(

. E

_I i I a

E I II l '

f m

s i

T I

1 g 1600 E

i a.

/

~

p1 S I400 l

/

/

i i

w '._I E

i O

u 1200

'JPACCEPTABLE r' i ACCEP T ABLE -~

s OPERATfd*

~

OPER A T ION g

i/i r'

t-i 3

1 I

f,f

!/ i 1000 ilii CRITICALITY LIMIT [

I I

i i I

i 800

, HE ATUP P4IES UP s

600 22: TO 100 F/X' f

l

=

(

400

(

200 l

0 0

50 100 150 200 250 300 350 l

l lNDICATEDTEMPERATURE(F) f i

l l

Figure A-3.

Pra:rie Island Unit 2 Reactor Coolant System Heatup Limitations Applicable for Periods up to 32 Effective Full Power Years. Margins

> of 60 psig and'10*F are included for Passible Instrument Error s

A-9 i

• )

10.027-28 i

s 2600 g

MATERI AL PROPERTY BASIS

• ELD NE TAL C.: 0.082 l

/

i 2200 INITIAL RTNDT: -20 F (CONSERVA TIVELY ASSUMED )

/

, i_

AT 32 EFFECTIVE FULL POWER YE ARS 1

2000 RT AT 1/4T THICNNESS : 110 F j

, ' \\.

C HD T 7

RT AT 3/47 THICKNESS: 74 CF f

/j ',

NOT 1800 y

t.

l l

1600 f

w l

g 1400 f

U a-1200 e

c 1000 8

g 800 C00g00wNRATES d%

600 j h;f, U

20 - -

60 --

400 ion

-C ck

. N.9 200 g

0 0

50 100 150 200 250 300 INDICATED TEMPERATURE (OF)

Figure A 4 Prairie Island Unit 2 Reactor Coolant System Cooldown Limitations Applicable for Periods up to 32 Effective Full Power Years. Margins l

of 60 psig and 10 F are included for Possible Instrument Error l

A 10 i

1

- - -