Text

Analysis of Capsule T from Northern States Power Co Prairie Island 2 Reactor Vessel Radiation Surveillance Program, Mar 1981
	ML19347F376
Person / Time
Site:	Prairie Island
Issue date:	03/31/1981
From:	Shaun Anderson, Kaiser W, Yanichkd S; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML19347F375	List: ML19347F374; ML19347F376;
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WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTI0i:

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ANALYSIS OF CAPSULE T FROM NORTHERN STATES POWER COMPANY PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko S. L. Anderson W. T. Kaiser March 1981 APPROVED:

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m e; 6b J. N, Chirigos, Manager Structural Materials Engineering Work Performed Under Shop Order No. ELEP 949 Prepared by Westinghouse for Northern States Power Company Although the information contained in this report is ncnproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval.

WESTINGl'0USE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 81 O!y19 () I?)

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TABLE OF CONTENTS SECTION TITLE PAGE,,

1

SUMMARY

1-1 2

INTRODUCTION 2-1 3

BACKGROUND 3-1 4

DESCRIPTION OF PROGRAM 4-1 5

TESTING SPECIMENS FROM CAPSULE T 5-1 5-1.

Test Procedures 5-1 5-2.

Charpy V-Notch Impact Test Results 5-2

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5-3.

Tensile Test Results 53 5-4.

Wedge Opening Loading Tests 5-3 I

6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1.

Introduction 6-1 6-2.

Discrete Ordinates Analysis 6-1 6-3.

Neutron Dosimetry 6-3 6-4.

Transport Analysis Results 5-7 6-5.

Dosimetry Results 6-8 APPENDIX A HEATUP AND C00L;0HN LIMIT CURVES FOR NORMAL A-1 OPERATION l

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LIST OF ILLUSTRATIONS FIGURE TITLE PAGE 4-1 Arrangement of Surveillance Capsules in the 4-5 Prairie Island Unit No. 2 Reactor Vessel 4-2 Capsule T Schematic Showing Location of Spect-4-7/4-8 mens, Thermal Monitors, and Dosimeters (Prairie Island Unit 2) 5-1 frradiated Charpy V-Notch Impact Properties for 5-10 the Prairie Island Unit No. 2 Reactor Vessel Lower Shell Forging 22642, Axial Orientation 5-2 Irradiated Charpy V-Notch Impact Properties 5-11 for the Prairie Island Unit No. 2 Reactor Vessel Lower Shell Forging 22647 Tanoential Orientation 5-3 Irradiated Charpy V-Notch Impact Properties for 5-12 the Prairie Isalnd Unit No. 2 Reactor Pressure Vessel Weld Metal 5a Irradiated Charpy V-Notch Impact Properties for 5-13 the Prairie Island Unit No. 2 Reactor Vessel Weld Heat-Affected-Zone-Metal 5-5 Irradiated Charpy V-Notch Impact Properties for 5-14 the A533 Grade B Class 1 Correlation Monitor Material 5-6 Charpy Impact Specimen Fracture Surfaces for 5-15 Prairie Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Axial Orientation 5-7 Charpy Impact Specimen Fracture Surfaces for 5-16 Prairie Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Tangential Orientation 5-8 Charpy Impact Specimen Fracture Surfaces for 5-17 Prairie Island Unit No. 2 Weld Metal 5-9 Charpy Impact Specimen Fracture Surfaces for 5-18 Prairie Island Unit No. 2 Weld Heat-Affected-Zone Metal 5-10 Charpy Impact Specimen Fracture Surfaces for 5-19 Prairie Island Unit No. 2 A533 Grade B Class 1 Correlation Monitor Material O

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LIST OF ILLUSTRATIONS (cont'd)

FIGURE TITLE PAGE 5-11 Irradiated Tensile Properties for the Prairie Island 5-20 Unit No. 2 Reactor Pressure Vessel Lower Shell Forging 22642, Axial Orientation 5-12 Irradiated' Tensile Properties for the Prairie 5-21

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Island Unit No. 2 Reactor Pressure Vessel Lower Shell Forging 22642 Tangential Orientation 5-13 Irradiated Tensile Properties for the Prairie 5-22 Island Unit No. 2 Reactor Pressure Vessel Weld Metal 5-14 Typical Stress-Strain Curve for Tension Specimens 5 23 5-15 Fractu*ed Tensile Specimens from Prairie Island 5-24 Unit No. 2 Pressure Vessel Lower Shell Forging

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22642, Axial Orientation 5-16 Fractured Tensile Specimens from Prairie Isiand 5-25 Unit No. 2 Pressure Vessel Lower Shell Forging 22642. Tangential Orientation 5-17 Fractured Tensile Specimens from Prairie Island 5-26 Unit No. 2 Pressure Vessel Weld Metal 6-1 Prairie Island Unit No. 2 Reactor Geometry 6-23 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-24 6-3 Calculated Azimuthal Distribution of Maximum Fast-Neutron Flux (E> 1.0 Mev) within the Pressure Vessel-6-25 Surveillance Capsule Geometry 6-4 Calculated Radial Distribution of Maximum Fast-Neutron Flux (E> 1.0 Mev) Within the Pressure Vessel 6-26 6-5 Relative Axial Variation of Fast Neutron Flux (E> 1.0 Mev) Within the Pressurt Ves:al 6-27 6-6 Calculated Radial Distribution of Maximum Fast Neutron Flux (E> 1.0 Mev) Within the Surveillanca Capsules 6-28 V and T 6-7 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule V 6-29 I

6-8 Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule T 6-30 6-9 Comparison of Measured and Calculated Fast Neutron l

Fluence (E> 1.0 Mev) for Capsules V and T 6-31 vi l

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LIST OF TABLES TABLE TITLE PAGE 4-1 Chemistry and Heat Treatment of Material 4-3 Representing the Core Region Lower Shell Forging and Weld Metal from the Prairie Island Unit No. 2 Reactor Vessel 4-2 Chemistry and Heat Treatment of Surveillance 4-4 Material Representing 12-Inch-Thick A533 Grade B Class 1 Correlation Monitor Material HSST Plate 02 4-3 Welding Procedure and Associated Information for 4-4 the Prairie Island Unit No. 2 Core Region Weld-ments 5-1 Charpy V-Notch Impact Data for the Prairie 5-4 Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642 Irgadiated at 550*F, Fluence 1.05 x 1019 n/cm' (E > l Mev) 5-2 Charpy V-Notch Imapct Data for the Prairie 5-E Island Unit No. 2 Pressure Vessel Weld and Heat-Affected-Zon Metal Irradiated at 550*F, Fluence 1,05 x 10 9 n/cm5 (E>l Mev) 5-3 Charpy V-Notch Impact Data for the Prairie 5-6 Island Unit No. 2 A533 Grade B Class 1 Ccerela-tion Monitor Mater d il Ir adiated at 550'F, Fluence 1.05 x 1019 n/cm (E> l Mev) l9 5-4 The [ffect of 550*F Irradiation at 1.05 x 10 5-7 n/cm (E>l Mev) on the Notch Toughness Proper-ties of the Prairic Island Unit No. 2 Reactor vessel Impact Test Specimens

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5-5 Summary of Prairie Island Unit No. 2 Reactor 5-8 Vessel Surveillance Charpy Impact Test Results 5-6 Irradiated Tensile Properties for the Prairie 59 Island Unit No. 2 Pressure Vessel Lower Sggli Forg{n and Weld Metal, Fluence 1.05 x 10 n/cm E > l Mev) 6-1 21 Group Energy Structure 6J. '

6-2 Nuclear Parameters for Neutron Flux Monitors 6-12 6-3 Calculated Fast Neutron Flux (E >1.0 Mev) and 6-13

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Lead Factors for Prairie Island Unit No. 2 Surveillance Capsules vii

LIST OF TABLES (cont'd)

TABLE TITLE PAGE 6-4 Calculated Neutron Energy Spectra at the Center 6-14 of the Prairie Island Unit No. 2 Surveillance Capsules 6-5 Spectrum Averaged Reaction Cross Sections at the 6-15 Dosimeter Block Location for Prairie Island Unit No. 2 Surveillance Capsules 6-6 Irradiation History of Prairie Island Unit No. 2 6-16 Reacter Vessel Surveillance Capsules 6-7 Comparison of Measured and Calculated Fast Neutron 6-18 Flux Monitor Saturated Activities of Capsule V 6-8 Comparison of Measured and Calculated Fast Neutron 6-19 Flux Monitor Saturated Activities for Capsule T 6-9 Results of East-Neutron Dosimetry for Capsule 6-20 V and T 6-10 Results of Themal Neutron Dosimetry for Capsule 6-21 V and T 6-11 Sumary of Neutron Dosimetry Results for Capsule 6-22 V and T s

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SECTIOi,1

SUMMARY

The analysis of the reactor vessel material contained in Capsule T, the second surveillance capsule removed from the Prairie Island Unit 2 reactor pressure vessel, led to the following conclusion:

I9 2

The capsule received an average fast-neutron fluence of 1.05 x 10 n/cm I9 2

(E > 1.0 Mev) compared to a calculated value of 1.05 x 10 n/cm.

Based on the fluence measurements for Capsule T, the vessel 1/4-thickness I8 fluence after 4.0 effective-full-power years of operation is 3.65 x 10 2

18 2

n/cm compared to a calculated fluence of 3.65 x 10 n/cm,

I9 2

The fast-neutron fluence of 1.05 x 10 n/cm resulted in the following increases ir transition temperature and decreases in upper shelf energy for the various reactor vessel surveillance materials: (see table 5-4).

50 Ft 1b 30 Ft 1b Temperature Temperature Shelf Energy Increase Increase Decrease.

Material

(*F)

(ft1b';

Forging 22642 65 55 17 (Tangential)

Forging 22642 45 35 15 (Axial) l Weld Metal 60 60 11 HAZ Metal 45 55 18 l

O 1-1

The results of the material surveillance tests indicate that the reactor pressure vessel is tough enough to continue to operate safely.

The probable end-of-life fluences at various locacions through the vessel wall are as follows:

Fast Neutron Fluence Vessel 2

2 Wall 1.ocation Measured (a/cm )

Calculated (n/cm )

I9 I9 Inner Surface 4.35 x 10 4.35 x 10 I9 I9 1/4 Thickness 2.92 x 10 2.92 x 10 I9 l0 3/4 Thickness 8.43 x 10 8.43 x 10 O

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SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule T, the second cap-sule of the continuing surveillance program for monitoring the effects of neutron irradiation on the Northern States Power Prairie Island Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Prairie Island Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation.

A description of the surveillance program and the preirradiation mechanical pro-perties of the reactor vessel materials are presented by Yanichko and Legebl3 The surveillance program was planned to cover the 40-year life of the reactor pressure vessel and is based on ASTM E-185-73, "Recomended Practice for Sur-veillance Tests for Nuclear Reactor Vessels"{2] Westinghouse Nuclear Energy Sys-tems personnel were contracted for the preparation of procedures for reinovina the second capsule from the reactor and its shipment to the Westinghouse Research and Developmen* Penter, where the postirradiation mechanical testing of the Charpy V-notch impact o.1d tensile surveillance specimens was perfortred.

Th s report sumarizes testing and the postirradiation data obtained from the second material surveillance capsule (Capsule T) removed from the Prairie Island Unit 2 reactor vessel and dicusses the analysis of these data. The data are also compared to results of the previously removed Prairie Island Unit No. 2 surveil-lance Capsule V reported by Davidson.E33 Using current methods,E43 heatup and cooldown pressure-temperature operating limits are established for the nuclear power plant. The heatup and cooldown pressure-temperature operating limits are presented in Appendix A.

l 2-1

SECTION 3 BACKGROUND The ability of the large steel pressurt vessel contianing the reactor core and its primary coolant to resist fracture constitutes an important factor in insuring 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-neu-tron bombardment. The overall effects of fast-neutron irradiation on the mechnical properties of low-alloy ferritic pressure vessel steels such as SA508 Class 3 (base materidl of the Prairie Island Unit 2 reactor pressure vessel beltline) are well documented in the literature. Generally, low-alloy ferritic materials show an in-crease in hardness and tensile properties and decrease in ductility and toughness under certain condition of irradiation.

A method of performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against NonDuctile Failure", Appendix G to Section III 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 is the greater of either the drop weight nil-ductility transition tempera-NDT ture (NDTT per ASTM E-208) or the temperature 60*F less than the 50 ft-lb (and 35 mil lateral expansiin) temperature as determined from Charpy specimens oriented normal to the major working direction of the material. The RT of a given material NDT is used to index that material to a reference stress intensity factor curve (K curve)

IR which appears in Appendix G of the ASME Code. The K curve is a lower bound of IR 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 function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

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3-1

RT and, in turn, the operating limits of nuclear power plants, can be adjusted NDT to account for the effects of radiation on the reactor vessel material properties.

The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the one developed for the prairie Island Unit 2 reactor vessel. In this program

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a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 50 ft-lb temperature ( RTNDT) due to irradiation is added to the original RT to adjust the RT for radiation embrittlement. This adjusted RT NDT NDT NDT curve and, in (RT initial + RTNDT) is used to index the material to the KIR NDT turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

l 3-2

SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the drairie Island Onit 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 then al shield and the vessel wall at locations shown in figure 4-1.

The lead factors shown in figure 4-1 are higher than reported in the original surveillance program report. These lead factors represent updated values resulting from improved analytical techniques developed since the original report. These techniques are discussed in paragraph 6-4.

The vertical center of the capsule is opposite the vertical center of the core.

Capsule T was removed after 4.0 effective full power years of plant operation.

This capsule contained Charpy V-notch impact Specimens from the limiting core region shell forging 22642, core region weld metal, weld-heat-affected-zone material and ASTM A533 Grade B Class 1 correlation monitor material (HSST Plate 02).

Also contained in the capsule were tensile specimens and wedge opening loading (WOL) specimens from shell forging 22642 and weld metal. The chemistry and heat treatment of the surveillance material are presented in table 4-1 and 4-2.

In-formation relative to the core region weldment is presented in table 4-3.

All test specimens were machined from the 1/4-thickness location of the forging.

Test specimens represent material taken at least one forging thickness from the quenched end of the forging. Some base metal Charpy V-notch and tensile specimens were oriented with the longitudinal axis of the specimens nomal to (axially) and some parallel to (tangential) the major working direction of the forging. The WOL test specimens we're machined such that the crack of the specimen would pro-pagate normal to (longitudinal specimens) and parallel to (transverse specimens) the major working direction of the forging. All specimens were fatigue precracked per ASTN E399-70T.

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 normal to the weld direction.

4-1

i Capsule T contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15-w/o-c;balt (cadimum-shielded and unshielded).

In addition, cadmium-shielded 237 238 dosimeters of N and U were contained in the capsule at locations shown in figure 4-2.

Thermal monitors made from low-melting eutectic alloys and sealed in Pyrex

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tubes were included in the capsule and were located as shown in figure 4-2.

The two eutectic alloys and their melting points are:

2.5t Ag, 97.5% Pb melting point, 579'F 1.75% Ag, 0.75% Sn, 97.5% Pb melting point, 590*F 6

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4-2

1 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 t

Chemical Analyses (Percent)

I Element *I Lower Shell 22642 Weld Metal (b)

C 0.175 O.045 Mn 1.22 1.37 P

0.011 0.019 i

S 0.013 0.014 j

Si 0,285 0.47 Ni 0.70 0.072 Cr 0.14 0.020 V

<D.008 0.001 Mo 0.445 0.51 I

Co 0.026 0.013 Cu 0.085 0.082 Sn 0.011 0.002 At 0.036 0.007 l

N 0.017 0.026 2

Heat Treatment Forging 22642 Heated at 1652/1715*F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , water quenched:

Tempered at 1175/1238'F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , furnace cooled; Heated at 1652/1724*F for 51/2 hours, water quenched; Tempered at 1202/1238'F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , furnace cooled; Stress-relieved at 1022*F for 111/2 hours, furnace-cooled; Stress-relieved at 1112*F for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , furnace cooled Weldment Stress relieved at 1022*F for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , furnace-cooled:

l Stress relieved at 1112*F for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , furnace cooled G. A Qualitatswe SpecteDg'ADh*C SaalTlil W49 enade for eiements greater inen 0 010 we qme percent.

b. Apphcable weld were and flus lot numbers are g wen en tab's 4 3 for chamfer fiHirs.

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t TABLE 4 2 CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL REPRESENTING 12. INCH. THICK A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL HSST PLATE 02 Chemical Analysis I

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Mn P

S Si Ni Ma Cu Ladle 0.22 1.45 0.011 0.015 0.22 0.62 0.E3 i

Check O.22 1.48 0.012 0.018 0.25 0.68 0.52 0.14 l

Heat Treatment 1675 25'F - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months - Air cooled 1600125'F - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months - Water quenched 1125 2 25*F - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months - Furnace cceled 1150 = 25'F - 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months - Furnace <ooled to 600*F TABLE 4 3 WELDING PROCEDURE AND ASSOCIATED INFORMATION FOR THE PRAlRIE ISLAND UNIT NO. 2 CORE REGION WELDMENTS *I I

Top and Br.,ttom of ChamferIM Automatic submerged arc welding with multiple passes.

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

Wire: UM 40 - 2.5 tr.m dia - fot: 3049 Flux: UM 89 lot: 1263 Welding Speed: 36 cm/ min Chamfer Filling Automatic submerged arc welding with multiple passes. J Preheat of 400*F.

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

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Figure 41. Arrangement of Surveillance Capsules in the Prairie Island Unit No. 2 Reactor Vessel (Updated Lead Factors for the Capsules are Shown in Parentheses) 4-5

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SECTION 5 TESTING OF SPECIMENS FROM CAPSULE T 5.1.

TEST PROCEDURE The postirradiation mechanical testing of the Charpy V-notch and tensile speci-mens was performed at the Westinghouse Research and Development Center Hot Cell Laboratory with consulation by Westinghouse Nuclear Energy Systems personnel.

Testing of the t!0L test specimens was delayed pending clarification of testing orocedures from the Nuclear Regulatory Commission.

Upon receipt of the capsule at the Laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against themasterlistinthereportbyYanichkohl3 No discrepancies were found.

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Exanination of the two low-melting (579'F and 590*F) eutectic alloys showed that melting did not occur in either alloy. Based onthis examination it was con-cluded that the maximum temperature to which the test specimens were exposed during irradiation was less than 579'F.

l A TMI Model TM 52004H impact test machine was used to perform tests on the irra-diated Charpy V-notch specimens. Before initiating tests on the irradiated Chrrpy V specimens, the accuracy of the impact machine was checked with a set of standard specimens obtained from the Army Material and Mechanics Research Center in Water'6wn, Massachusetts. The results of the calibration testing showed that the machine was certified for Charpy V-notch impact testing.

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The tensile tests were conducted on a screw-driven instron testing machine having a 20,000 pound capacity. A crossht d speed of 0.05 in./ min was used.

The deformation of the specimen was measured using a strain gage extensometer, The extensometer was calibrated berore testing with a Sheffield high-magnifi-cation 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 stab 11ze their temperature prior to testing. Temperature was monitored using a chromel-alumel thermocouple in contact with the upper and lower clevis-pin specimen grips.

Temperature was controlled within plus or minus 5'F.

The load-extension data were recorded on the testing machine strip chart. The yield strength, ultimate tensile strength, and uniform elongation were determined from these charts. The reduction in area and total elongation were determined from specimen measurements.

5-2 CHARPY V-NOTCH IMPACT TEST RESULTS The irradiated Charpy V-notch impact test results for the reactor vessel beltline forging material, weld metal and heat-affected zone (HAZ) material and ASTM corrsla-tion monitor material from HSST Plate 02 are presented in tables 5-1 through 5-3 and figures 5-1 through 5-5, respectively. A summary of the increases in transi-tion temperature and the decrease in the upper shelf energy of the various surveil-lance materials is presented in table 5-4.

The test results show that 30 and 50 ft-lb transition temperature increases for the shell forging 22642 and the weld l9 2

metal and HAZ material are small (35 to 60*F) for a fluence of 1.05 x 10 n/cm,

therefore indicating that the materials are not very sensitive to radiation. The correlation monitor material showed transition temperature increases of 150 to 160'F for the same fluence level. The much larger transition temperature increase exhibited by the correlation monitor material is believed to be the result of the higher copper content (.14% Cu) of the material. A comparison of the temperature increases re-Sulting from irradiation tests performed on the two Prairie Island Unit 2 capsules tested to date is shown in table 5-5.

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l 5.3 TENSILE TEST RESULTS The results of tensile tests performed on specimens from shell forging 22642 and the weld metal are shown in table 5-6.

A comparison of the unirradiated versus irradiated tensile propert'es is shown in figures 5-11 through 5-13 for forging 22642 and the weld metal. The small increases in yield strength of approximately I9 2

5 to 10 ksi resulting from irradiation to 1.05 x 10 n/cm tend to confirm that

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the reactor vessel beltline materials are not highly sensitive to radiation as also indicated by the Charpy V-notch tests. A typical load-displacement curve obtained for the tensile tests is shown in figure 5-14.

Photograpbs of broken tensile speci-mens from the surveillance forging and weld metal are shown in figures 5-15 through 5-17.

5.4 WEDGE OPENING LOADING TESTS The Wedge Opening Loading fracture mechanics specimens that were contained in Cap-sule T have been stored at the Westinghouse Research Laboratory on the recommenda-tion of the United States Nuclear Regulatory Commission and will be tested at a later date. The results of these tests will be reported upon their completion.

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5-3

TABLE 5-1 CHARPY V-NOTCH IMPACT CATA FOR'THE PRAIRIE ISLAND UNIT h0. 2

?RESSUP.E VESSEL SHELL FORGING 22642 IRRADIATE 0 AT 550*F, FLUENCE 1.05 x 10 n/cm2 (E > 1 Mev)

I9 Lateral

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Specimen Temp.

Expansion Shear Energy)

(mils)

(". )

No.

(*F)

(ft 1b (AXIAL DIRECTION)

NT-42 0

19.0 16 5

NT-45 25 33.0 29 8

NT-40 50 23.0 22 26 NT 60 43.5 39 21 NT-46 78 61.0 44 38 NT-37 120 56.5 51 36 NT-48 150 70.5 60 50 NT-38 175 60.5 NT-39 200 68.5 61 68 NT-43 250 94.5 74 100 NT-41 300 88.0 72 100 NT-44 350 96.0 73 100 (TANGENTIAL OIRECTION)

NL-42 0

8.0 9

5 NL-40 25 21.5 19 8

NL-45 25 43.0 35 15 NL-38 50 50.5 45 21 NL-46 78 69.5 55 42 NL-41 100 66.5 60 48 NL-37 125 87.0 71 68 NL-48 150 108.0 76 75 NL-47 200 130.5 88 100 NL-43 250 136.5 85 100 NL-39 300 132.0 92 100

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TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE PRAIRIE TSLAND UNfT NO. 2 REACTOR PRESSURE VESSEL WELD AND HEAT-AFFECTED-ZONE METAL IRRADIATED AT 550'F, FLUENCE 1.05 x 10 n/cm2 (E>l Mev)

I9 Lateral Specimen Te!.p.

Energy Expansion Shear No.

(*F)

(ft lb)

(mils)

(%)

(UELDMETAL)

NW-25

-50 14.0 13 8

NW-26

-10 42.5 35 32 NW-28 0

35.0 31 26 NW-29.

25 53.0 45 58 NW-27 75 82.0 65 79 NW-30 150 86.5 75 91 NW-31 250 103.0 74 100 NH-32 300 86.5 86 100 (HEAT-AFFECTED-ZONE METAL)

NH-30

-100 22.0 14 20 NH-29

-75 47.0 35 33 NH-32

-50 53'.5 34 34 NH-26

-25 34.5 28 42 NH-25

-10 59.5 49 51 NH-31 0

102.5 64 100 NH-27 75 106.0 70 100 NH-28 150 87.5 75 100 5-5

TABLE 5-3 CHARPY V-NOTCH IMPACT DATA FOR PRAIRIE ISLAND UNIT NO. 2 A533 GRADE B CLASS 1 CORRELATION MONITOR MATERIAL I9 IRRADIATED AT 550'F, FLUENCE 1.05 x 10 n/cm2 (E> 1 Mev)

Lateral Specimen Temp.

Energy Expansion Shear No.

(*F)

(ft1b)

(mils)

(%)_

R-31 100 7.5 6

14 R-32 200 21.0 18 25 R-25 210 33.5 28 30 R-27 225 56.5 47 53 R-28 250 54.5 41 62 R-30 300 88.5 79 100 R-29 350 90.0 69 100 R-26 400 84.5 79 100 9

O G

S 5-6

.. =

TABLE 5-4 I9 Tite EFFECT OF 550*F IRRADIATION AT 1.05 x 10 n/cm2 (f> 1 Mev) ON Tilf NOTCil TOUGilNESS PROPERTIES OF Tile PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL IMPACT TEST SPECIMENS 4

i O

TRANSIT 10tl TEMPERATURE ( F)

Average Energy Absorption at Full Shear (f t Ib)

Unirradiated Irradiated ATemperature (*F) i Material 50 ft Ib 30 ft Ib 35 mils 50 ft Ib 30 ft Ib' 35 mits 50 ft Ib 30 ft Ib 35 mils unerradiated Irradiated AEnergy Forging 22642

-5

-25

-5 60 30 40 65 55 45 150 133 17 (Tangential)

Forging 22642 35 0

25 80 35 60 45 35 35 108 93 15 (Axial)

Weld Metal

-40

-75

-45 20

-15 0

60 60 45 103 92 II HAZ Hetal

-90

-135

-80

-45

-80

-45 45 55 40 117 99 18 m

da Eorrelation 80 45 60 230 205 225 150 160 165 123 88 35 l

Material I

i 1

.i I

s O

e M

TABLE 5-5 SL99%RY OF PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL SURVEILLANCE CHARPY IMPACT TEST RESULTS 50 ft Ib 30 ft 1b Upper Shelf Trans. Temp. Trans. Temp.

_ Energy Fluence Increase Increase Decrease I8 2

Material 10 n/cm

(.7)

(.F)

(ft 1b)

Forging 22642 5.86 35 35

+ 19 (Tangential) 10.50 65 55 17 Forging 22642 5.86 30 30

+ 10 (Axial) 10.50 45 35 15 Weld Metal 5.86 55 60 3

10.50 60 60 11

~

Weld HAZ 5.86 45 45 5

10.50 45 55 18 Correlation 5.86 120 125 21 Monitor 10.50 150 160 35 5-8

TABLE 5-6 IRRADIATED TENSILE PROPERTIES FOR THE PRAIRIE ISLAND UNIT NO. 2 REACTOR PRESSURE VESSEL LOWER SifELL FORGING AND WELD METAL IRRADIATED AT 550*F, FLUENCE 1.05 x 10 n/cm2 (E> 1 Mev)

I9 Ultimate Test 0.2% Yield Tensile Fracture fracture Fracture Uniform Total Reduction Specimen Temp.

Strength Strength Load Stress Strength Elong.

Elong.

In Area Material No.

(*FL (ksi)

(ksi)

(1b)

(ksi)

(%)

Forging 22642 NL-12 73 73.8 91.7 2800 211.0 57.0 11.3 25.1 73 J

(Tangential)

NL-11 200 70.8 88.1 2650 215.9 54.0 10.2 23.0 75 NL-10 550 64.2 86.6 2750 145.7 56.0 9.8 22.0 62 Forging 22642 NT-12 74 76.4 91.2 3030 171.2 61.6 11.5 24.2 64 (Axial)

NT-11 200 70.8 86.6 '

3000 132.2 61.1 10.5 21.5 54 i

ui NT-10 550 66.2 88.1 3350 139.3 68.2, 9.0 18.2 51 l

Weld Metal NW-12 74 78.3 91.7 2800 195.6 57.0 9.8 21.9 71 NW-11 200 71.3 84.5 2630 139.1 53.5 10.2 22.8 62 NW-10 550 67.2 87.6 2950 1 91.6 60.1 9.0 19.7 69 f

I

)

18,318-3 I

I I

l I

3 100 3

_/ e 80 7

~

60 w

9 58 40

O 20

/

0 iOO G

80 A

m f

.o.-

em

/.,.

a a.

O X

/

39 W

40 O

W o

(35'F) b 20 0

0

~

140 120 a

g 100 h2 UNIRR# Ol ATED n

a 0

_t e

C U

e e

IRRADIATED (550*F)

(45'F) 1.05 X 1019 n/cm2 0

5 40 o

e (35' F) 9 20 g

I I

l l

0

-100 0

100 200 300 400 500 TEMPERATUi4E (*F)

Figure 5-1.

Irradiated Charpy V-Notch Impact Properties for the Prairie Islanci Unit No. 2 Reactor Vessel Lower Shell Forging 22642, Ax:al Orienta on 1

5-10

18.318-4 l

l l

100 3

p_

80

-e

~

l e'e G

40 O'

i

/

20 e

0 100 G

80 o

7

/

o

~

80 J*e N

O w

40 (45* F) a

,e 0

180 160 UNIRRADIATED P

Q 9

O, 140 o

e-p w

_g 120

$ 100 O

b 80 IRRADIATED (550*F)

O i

l 1.05 X 1019 n/cm2 5

m o

(650 F)

(55* F) 20 je l

0 l

l 0

-100 0

100 200 300 400 500 TEMPERATURE (*F)

Figure 5-2.

Irradiated Charpy V-Notch Impact Properties for the Prairie Island Unit No. 2 Reactor Vessel Lower Shell Forging 22642. Tangential Orientation 5-11

18.318-5 I

I I

3

/*J,;

i00 O,2 i

8,/

.0 O

/

e/

=

40 2

20 e

0 100 a

n s

W fM O

e

[/g n

e

~

(45'F)

O 5

2c O/

0 120 9

d 2 g.

109 UNIRRADIATED 0

0 ef*

a0

=

b

[

60 O

IRRADIATED (550*F) 1.05 X 1019 n/cm2 e

e7 (60*F) 5a O

3

(60'F) 20 O

e 0

-200

-100 0

100 200 300 400 TEMPERATURE ('F)

Figure 5-3.

Irradiated Charpy V-Notch Impact Properties fcr the Prairie Island Unit No. 2 Reactor Pressure Vessel Weld Metal 5:12

. - ~. _ _. - -. - - -. _..

\\

18.318-6 1

I l

l l

2 33 3

100 -

og j

80 O

E 60 o

O Oy 40 -O f

0 2

20 2

g 0

100 5

M a

V 5

o#

7 60 A

O.

(40 F)

J 20 0

140 O

h o

i 120 UNIRRADIATED O

O O.

2 100 a

Hg 80 O

IRRADIATED (550*F) 1.05 X 1019 n/cm2 60 O

g

?-(45* F) y W

40 -O (55* F) 0

-200

-100 0

100 200 300 400 TEMPERATURE (*F)

Irradiated Charpy V-Notch impact Properties for the Prairie Island Unit No. 2 Figure 5-4.

Reactor Pressure Vessel Weld Heat-Aff(rted-Zone Mets!

5-13 l

l

18.318-7 l

l l

3

/

100 G

i 2

80 6

g 60 0

3 Ee 40 20 2

0 100 e

a 5"

e e

o 1

60 a6a e,

2 165*F g

5 20 O

O 140 w

6 120 O

UNIRRADIATED 100 5

6 o

e Em g

60 E

150'F w

40 b

o 160*F Q IRRADIATED (550*F) 1.05 X 1018 n/cm2 20 0

0 l

l

-100 0

100 200 300 400 500 TEMPERATURE (*F)

Figure 5-5.

Irradiated Charpy \\/-Notch Impact Properties for the A533 Grade B Class I Correlation Monitor Material 5-14

j l

l 18.318-8 i

i i

l l

i$,az <6

",'n MP

- 3., :,,

,.f}-

.m.~.

f, *

%^,. -)

,1

~

~ *

.e' tut

d.

;.

f...,.

Q,,.

<- w -

.?.

- n.

g

@.d v, s.- -

i NT-42 NT-45 NT-40 NT-47 NT-46 NT-37 l

~

_- g

_y

(,.

y-

-y g, -

,.4

~

,s, t

f,.,.

~ r *,

\\ * * *,~

h:*

.p y ' ;

y.

c NT-48 NT-39 NT-43 NT-41 NT-44 l

l l

Figure 5-6.

Charpy impact Specimen Fracture Surfaces for Prairie Island l

Unit No. 2 Pressure Vessel Lower Shell Forging 22642, l

Axial Orientation I

i I

5-15 l

l J

w--w,--<mre.

--~ ~ -

e

18,318-9 i

r-- E - - g

,. 3

. I kW $' - ?

?)Y.

~-

T c :.

g.'.

~ -l ' >.

-t t

i

.,;.~,;

-7, -

-j :

% ' %, 2

% 't '

2,, J4-

.aQ)y

)y$ $m.

g.~;' -jf R

i [J[

~

It

.c

~un

.w w

R

.a N L-42 NL-40 N L-45 N L-38 N L-46 N L-41 i

l l

l I

^"

' QM

}_~ 9 %

T W

-2 i

..4 i

i s4

,9

[d-1

~

)

v.

af

-. ;4-i I.

- a e I,,

g 7n N L-37 NL-48 NL-47 NL-43 N L-39 i

Figure 5-7.

Charpy impact Specirnen Fracture Surfaces for Prairie Island Unit No. 2 Pressore Vessel Lower Shell Forging 22642, Tangential Orientation l

I 5-16 l

l

18,318'-10 l

t NW-25 NW-26 NW-28 NW-29 i

t NW-27 NW-30 NW-31 NW-32 Figure 5-8.

Charpy impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 Weld Metal 5-17

18,318 11 i

4 C*

g,.

g j s -}' ' U-

,.i (

gt.

~~

c; -: --

.y.

g.-

t r.

+, ; n..

~p.

~a,%.,f w

W-h 4

NH-30 NH-29 NH-32 NH-26 i

. l r,, _.

,,. -..... g --

g q-

'A c..

'.yt 0 [.

~f.s 1.' y u.

A,q

- ' q.

.- ~ - '

w t,:gs,-

s g,g _. -

4 x,;

,>x)

.s I

NH-25 NH-31 NH-27 NH-28 Figure 5-9.

Charpy impact Specimen Fracture Surfeces for Prairie Island Unit No. 2 Weld Heat-Affected-Zone Metal 5-18

-~m--*-----w---.-sm-ww-.

y-ym g-p-y.

-w-y-,---w-%--

-w-,----ew-e,.-

18,318-12 i

i i

I tye9>p m~ smeam W 9

5

. p,,.. )

l

?

)

, [ _'

e n.'

\\:,"

I R '-',

l# "."

~

m h:.

W -

s R-31 R-32 R-25 R-27 l

"""T M'?,

% \\

r y -

~

W 9

~

d.

.~

,p,

m

.~e r

j. ',,~, -

k 4

f..

u i.

h

(,

F R-28 R-30 R-29 R-26

\\

l 1

l Figure 5-10. Charpy impact Specimen Fracture Surfaces for Prairie Island Unit No. 2 f 533 Grade B Class 1 Correlation Monitor Material 5-19 1 - -.

18,318-13 120 l

l l

l 110 LEGEND:

OPEN POINTS - UNIRRADIATED 0

2 100 CLCSED POINTS -lRRADIATED AT 1.05 X 10 n/cm g 90 A%

g

)

g 80

. TENSILE STRENGTH a

N 9 --

~ ~ ~ _ _ _ e

70 0.2% YlELD STRENGTH si 50 40 80 70 a

m g\\

v m

00 REDUCTION IN AREA 9%_

-E 50

--S

>>3 40 P

O3 30 O

&3__

. _2 TOTAL ELONGATION 20

' A ' J-g O

a a

10 9

A UNIFORM ELONGATION 0

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

Figure 5-11. Irradiated Tensile Properties for the Prairie Island Unit No. 2 Reactor Pressure

~

Vessel Lower Shell Forging 22642, Axial Orientation 5-20

=

18.318-14 120 g

i l

110 LEGEND:

OPEN POINTS - UNIRRADIATED 100 CLOSED POINTS -IRRADIATED AT 1.0b X 10IO 2

n/cm 0 90 A%

g

% g A

g 80 TE NSiLE STRENGTH 70 8 ' --- -

60 0.2% WELD STRENGTH

^v 50 40 -

80 g ' * ~ ~ O'%

REDUCTION IN AREA 70 O

m g

N.

60

!!$ 50 b 40 G

S 30 h

a 2

O TOTAL ELONGATION 2

A-g 20

_zm A

O 10 -

G-O UNIFORM ELONGATION l

l l

O O

100 200 300 400 500 600 700 TEMPERATURE FF)

Figure 512. Irradiated Tensile Properties for the Prairie island Unit No. 2 Reactor Pressure Vessel Lower Shell Forgirig 22642, Tangential Orientation 5-21

18.318 15 120 l

l l

l

~

LEGEND:

100 OPEN POINTS - UNIRRADIATED CLOSED POINTS -IRRADIATED AT 1.05 X 10 8 2

n/cm A

g 90 g

E NA 80 g

TENSILE STRENGYH d

E N

g 70 0

0.2% YlELD STRENGTH m

0 s0 40 80

^

. 70 O

n 60 O~

REDUCTION IN AREA 6 50

$ 40 i:

S 30 O

__a TOTAL ELONGATION 20

- ~ ~a b

O O

O e

6-UNIFCgELONGATION O

O 100 200 300 400 500 600 7M TEMPERATURE l'F) l Figure 513. trradiatea Tensile Properties for the Prairic Island Unit No. 2 l

Reactor Pressure Vessel Weld Metal 5-22

l J

100 e

90 -

i 80 -

l 70 -

i 60 i

a

_M g 50

=

G"

'i"

$ 40 ra ta i

i 30

.I 20 SPECIMEN NL-10 (550*F) 10 l

l l

0 0

.02

.04

.06

.08

.10

.12

.14

.16

.18

.2J

.22

.24 STRAIN (INJIN.)

-m l

'S*

Figure 5-14. Typical Stress - Strain Curve for Tension Specimens

18,318-17 ga-4:

4

. c s

'p=

6

M pg --

.~.e..

3_.4 e l.

p sf R p$ 8 ; ',. 4.'- '* ' v ' ;

~

r x

^

.:r-NT-12 1

)h. '._ '

P.4 j

j

-- +, - m h.' ',

s

.. 7 -.. _ - -.

NT-11 i k b

h.$ 5. 4 y

~~

i V

%t NT-10 Figure 5-15. Fractured Tensile Specimens from Prairie Island

~

Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Axial Orientation 5-24

i 18,318-18 W..G iCllli' :*~"MK1; Mil 3CE I 1" lE ~, i T ?c.$.,

k

@h*D**T@P#iMid(-,f '

..,44Ql %

sm;g n,.,...

?

.5W 4r ;' -.

,,. j NL-12

?w?Wi?)3}I??.' L v pge..,.

E,,.

)

%. ow ;;,;;

$?5?*..

. - m,..

, 5' b ;

~

r j

n

'i N L-11

'?

5

Y& ?;

ka-8 m%-....

. ""W I QJ -' T

.f

.[ ' ~. g ',.;

g 9.x Q.

.~,

N L-10 Figure 5-16. Fractured Tensile Specimens from Prairie Island Unit No. 2 Pressure Vessel Lower Shell Forging 22642, Tangential Orientation 5-25

1 18,318 19

h,.$.fs&f_.hkNhN.. 1hh.h{ ' _,.'

4

. _. a c.

2

%1

'e w y,9 y 8,.go. ; >.,.

m.;. (gy+

e

,;.y1 3

y%.yn)g;1

4rQ.,j.' ;g;%Q3 g

.E e,

,3 J5 c.,

?

.;p 6,,. 2

.. e.

v.: p...i..,

q s

ytA..S kMb.[,h. : O;f j~

NW-12 l

-& ij+k0A$h.*f~j'I'#.. ~........ _.

l y'.-l

. w v-s.,, w y

./.. '

.n

. 4 v.;

r ;

.t' W -4 f8ti,psMe,

? '.

jdiibt.4

.;7

7.. r..egf, Q ;- 4; g%.

- s s' q i

. e.. s w. *., x. 2. u 3 e4

,A.

. nu 4

u :ps i

I NW-11 k

?

i

$Ww-. NNE9hfMN.k d.,..-

.mu. v ga...:

l -]' \\.;;

~..

. J'-

e,

...! < s

' 'M,7 v..

-,i.- ?pyrt'f<;-Wg

[vW44 eM.e.- g[= sl.

4

1. ;...

- 9

t 4

u cm--

jW+ g%, ;

g v :.s. ;>_w_ - w : ! pi

.);..

NW-10 i

i i

I I

i Figure 5-17. Fractured Tensile Specimens From Prairie Island Unit No. 2 Pressure Vessel Weld Metal 5-26 I

SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1.

INTRODUCTION Knowledge of the neutron en/ironment within the pressure vessel-surveillance capsule geometry is required as an integral part of LWR pressure vessel sur-veillance programs for two reasons. First, in the interpretation of radiation-induced property changes observed in material test specimens, the neutron en-vironment (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 o'rdinates Sp transport analysis performed for the Prairie Island Unit 2 reactor to detemine the fast neutron (E> 1.0 Mev) l 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-averaged reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule T is dis-cussed and updated evaluations of dosimetry from Capsule V are presented.

6-2.

DISCRETE ORDINATES ANALYSIS A plan view of the Prairie Island reactor geometry at the core midplane is shown in figure 6-1.

Since the reactor exhibits 1/8th core symetry, only a 0*-45' sector is depicted. Six irradiation capsules attached to the themal shield are included in the design to constitute the reactor vessel surveillance program. Two cap-sules are located symmetrically at 13', 23', and 33* from the cardinal axis as shown in figure 6-1.

6-1 l

A plan view of a single surveillance capsule attachid 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 midplane, thus span-ning the central 5.25 feet of the 12-foot-high reactor core.

From a neutronic standpoint, the surveillance capsule structures are signifi-cants In fact, as will be shown later, they have a marked impact on the distributions of neutron flux and energy spectra in the water annulus between the th mal shield and the reactor vessel. Thus, in order to properly ascer-tain' the neutron environment at the test specimen locations, the caosules themselves must be included in the analytical model. Use of at least a two-dimensional computation is, therefore, mandatory.

In the analysis of the neutron environment within the Prairie Island Unit 2 reactor geometry, predictions of neutron flux magnitude and energy spectra were msde with the 00T[5] two-dimensional discrete ordinates code. The radial azimuthal distributions were obtained from an R, e computation wherein the geo-metry shown in figure 6-1 and 6-2 was described in the analytical model. In addition to the R, e computation, a second calculation in R, Z geometry wes also carried out to obtain relative axial variations of neutron flux through-out 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, e and the R, Z analyses employed 21 neutron energy groups, an S 8 angular guadrature, and a P) cross-section expansion. The cross sections

~

were generated via the Westinghouse GAMBIT [6] code system with broad group EO processing by the APPROPGS and ANISN codes. The energy group structure used in the analysis is listed in table 6-1.

i A key input parameter in the analysis of the integrated fast neutron exposure of the reactor vessel is the core power distribution. For this analysis, power distributions representative of time-averaged conditions derived from statis-tical studies of long-term operation of Westinghouse two-loop plants were em-ployed. These input distributions include rod-by-rod spatial variations for all l

peripheral fuel assemblies, t

6-2 l

It sh:uld b2 noted that this particular power distribution is intended to oro-duce accurate end-of-life neutron exposure levels for the pressure vessel. As such, the calculat'on is indeed representative of an average neutron flux and small (i 15-20%) deviations from cycle to cycle are to be expected.

Having the results of the R, e and R Z calculations, three-dimensional varia-tions of neutron flux may be approximated by assuming that the following re-lation holds for the applicable regions of the reactor.

c(R,Z,0,E )

c(R,0,E ) F(Z,E )

(6-1) i-~c

=

g g

g o(R,Z,0,E )

= neutron flux at point R,Z,0 within energy group g g

c(R,0,E )

= neutron flux at point R,9 within energy group g g

obtained from the R.O calculation F(Z,E )

= relative axial distribution of neutron flux within energy g

group g obtained from the R,Z calculation 6-3.

NEUTRON DOSIMETRY The passive neutron flux moniters included in the Prairie Island Unit 2 sur-veillance program 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 measr terials properties changes. To properly account for burnout of the

_.. isotope generated by fast neutron reactions, it is necessary to also determine the magnitude of the thermal neutron flux at the monitor loca-tion. 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 iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accortmodated within the dosimeter block located near the center of the capsule, 6-3

Ths use of passive monitors such as th:se lir,ted 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-and energy-dependent neutron flux has on the target mateiral over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on tha various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

e The operating history of the reactor e The energy response of the monitor e The neutron energy spectrum at the monitor location e The physical characteristics of the monitor The analysis of the passive monitors and sutsequent 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 [9,10,11,12,13]

Following samole preparation, the activity of each monitor is determined by means of a lithium-drifted gemanium, Ge(L1),

gamma spectrometer. The overall standard deviation of the measured data is a function of the precision of templu weighing, the uncertainity in counting and the acceptable error in detector calibration. For the samples removed from i

Prairie Island Unit 2, the overs 112rdeviation in the measured data is deter-mined to be + 10 percent. The neutron energy spectra are determined analyti-cally using the method described in section 6-1.

i l

Having the reasured activity of the monitors and the neutron energy spectra at the locaticns of interest, the calculation of the neutron flux proceeds as i

follows:

l l

6-4

The reaction product activity in the monitor is expressed as N

-At

-At R=

fY o(E)e(E)

U-'

I*

g (6-2)

A ma E

j.

where:

R = induced product activity N

= Avagadro's number o

A = atomic weight of the target isotope fi = weight fraction of tt'e target isotope in the target material Y = number of product atoms produced per reaction a(E) = energy dependent reaction cross section c(E)

= energy-dependent neutron flux at the monitor location with the reactor at full power Pj = average core power level during irradiation period j P

= maximum or reference core power level max A = decay constant of the produi,2 isotope tj = length of irradiation period j td = decay time following irradiation period j r

Since neutron flux distributions ai'e 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 intrJral term in equation (6-2) is replaced by the following relation:

o(E) $(E)dE a $(E> 1.0 Mev)

=

e 1

6-5 i

where:

N

[ a(E) o (E)dE 9 '9 o

o.,

7, N

e (EldE Pb#

9

.m G*Gim Thus, equation (6-2) is written N

P

-Atj)e-At N,

j d

R = - f; y I o (E> 1.0 Mev)

(1-e A

p" 3.,

or, solving for tha neutron flux, R

e (E>1.0 Mev) =

(6-3)

N N,Pj

-Atj

-Ae o

d

~fIV (1*

I' A

g.,

The toal fluence above 1.0 Mev is then given by N

p, 4 (E>1.0 Mev) = 4 (E>1.0 Mev) tj

'F 2) 66

where:

N g

1 tj = total effective full power seconds of reactor operation p

up to the time of capsule removal x

j.,

An assessment of the thermal neutron flux levels within the surveillance cap-sules is obtained from the bare and cadmium-covered CoS9(n.a)Co60data by means of cadmium ratios and the use cf a 37-barn 2200 m/sec cross section.

Thus, I D - 1I oTh (6-5)

=

N N

P

-Atj

-At o

d f ya (1-e le g

p A

max j=1 Rbare where:

D is defined as R Cd covered 6-4 TRANSPORT ANALYSIS RESULTS Results of the S transport calculations for the Prairie Island Unit 2 reactor n

are summarized in figures 6-3 through 6-8 and in tables 6-3 through 6-E.

In figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius,1/4 thickness location, and 3/4 thickness loncion are presented as a function of azimuthal angle. The influence of the surveillance capsules c1 the fast neutron flux distribution is clearly evident. In figure ti-4, the radial distribution of maxinium fast neu-tron flux (E> l.0 Mev) 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 ob-tained 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 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 6-7

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 the max'nuA fast -autron flux at the pressure vessel inner radius. Updated lead factors for all of the Prairie Island Unit 2 surveillance capsules are listed in table 6-3.

Since the neutron flux monitors contained within the surveillsnce 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. Analy-tically determined reaction rate gradients for use in the adjustment procedures are shown in figures 6-7 and 6-8 for Capsules V and T.

All of the applicable fast neutron reactions are included.

In order to derive neutron flux and fluence levels from the measured disinte-1 gration rates, suitable spectrum-averaged reaction cross sections are required The neutron energy spectrum calculated to exist at the center of each af the Prairie Island Unit 2 surveillance capsules is given in table 6-4.

The asso-ciated spectrum-averaged cross sections for eacf: of the five fast neutron reactions are given in table 6-5.

6-5.

00SIMETRY RESULTS The irradiation history of the Prairie Island Unit 2 reactor is given in table 6-6.

Canparisons of measured and calculated saturated activity of the flux moni-tors contained in Capsules V and T are listed in tables 6-7 and 6-8, respectively.

All adjustnents to the capsule center were based on the data presented in figures 4-7 and 6-8.

The fast neutron (E> 1.0 Mev) flux and fluence levels derived for Capsules V and T are presented in table 6-9.

The thermal neutron measurements obtained from the Cobalt-aluminum monitors is summarized in table 6-10. Unfortunately, cadmium covered monitors were not recovered from either Capsule V or T.

Therefore, thermal neutron flux levels were inferred from the bare Cobalt-aluminum activities via comparison with data obtained from Capsule V of Prairie Island Unit 1.

Due to the rclatively low thermal neutron flux at the capsule locations, no burnup cor-rection was made to any of tt.e measured activities. The maximum error introduced 6-8

~

by this assumption is estimated to be less than 1 percent for the Ni 0(n.p)CoS8 reaction and even less significant for all of the otFer fast neutron reactions.

Using the iron data presented in table 6-9, along with the lead factors given in table 6-3, the fast neutron fluence (E > 1.0 Mev) for Capsules V and T as well as for the reactor vessel in,:cr diameter are summarized in table 6-8 and figure 6-9.

The agreenent between calculation and masurerent is excellen' 18 and 1.05 x ?M9 with measured fluence levels of 5.86 x 10 compared to cal-18 I9 2

culated values of 5.86 x 10 and 1.05 x 10 n/cm for Capsules V and T, re-spectively. Further, the graphical representation in figure 6-9 indicates the accuracy of the transport analysis for Prairie Island Unit 2 and supports the use of the cnalytically 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 Prairie Island Unit 2 reactor as derived from both calculation and measurement may be made as follows.

2 FAST NEUTRON FLUENCE (n/cm )

Surface 1/4T 3/4T I9 I9 18 Capsule V 4.35 x 10 2.92 x 10 8.43 x 10

' I9 I9 18 Capsule T 4.35 x 10 2.92 x 10 8.43 x 10 I9 I9 18 Average measurement 4.35 x 10 2.92 x 10 8.43 x 10 I9 I9 18 Calculation 4.35 x 10 2.92 x 10 8.43 x 10 These data are based on 32 full. power years of operation at 1650 MWt.

Based on the fluence measurements for Capsule T, the vessel 1/4 thickness fluence 18 2

after 4.0 effective full power years of operation is 3.65 x 10 n/cm compared to 18 2

a calculated fluer.ce of 3.65 x 10 n/cm,

l e

[

6-9 t

Based on the new caost.le to vessel inner wall lead factors identified in table 6-3 and the new withdrawal schedule identified in ASTM E185-79 it is recomended that future capsules be removed from the reactor per the following schedule Capsule Vessel Lead Removal Capsule Fluence 2

Identity Location Factor Tima_

(n/cm )

18 Y

77*

3.37 1.28 EFPY (removed) 5.86 x 10 I9 T

67' 1.94 4.0EFPY(removed) 1.05 x 10l R

257' 3.37 6.0 EFPY 2.75 x 10 '

I9 P

247' 1.94 15.0 EFPY 3.95 x 10 I9 5

57' 1.79 32.0 EFPY 7.78 x 10 N

237' 1.79 Standby 6

6 m

o E

6-10

s i

TABLE 61 21 GROUP ENERGY STRUCTURE Group Lower Energy (Mov) 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 1.88 x 104 20 100 x 10 7 l

21 0.0

'Upoor energy of group 1 is 10.0 Mev 6-11 1

I TABLE 6-2 NUCLEAR PARAMSTERS FOR NEUTRON FLUX MONITORS Fission Target j

Monitor Reaction Weight

Response

Product

Yield l

Material of interest Fraction Range Half lJfs

(%)

l Copper Cu63(n alCo60 0.6917 E>4.7 Mew 6.27 years p.54(n,p)MnM 0.0585 EM.0 hv 3H %s Nickel nim (n.p)Co68 0.6777 F31.0 Mev 71.4 days

'{

Uranium 238' U238(n.f)Cs137 1.0 E>0.4 Mew 30.2 years 6.3 i

Neptunium 237' Np237(n.f)Cs137 1.0 E>0.08 Mew 30.2 years 6.5 i

Cobalt-Aluminum

Co69(n,7)Co60 0.0015 0.4eV<0.015Mev 5.27 years Cobalt-Aluminum Co69(n,7)Co60 0.0015 E<0.0015 Mew 5.27 years i

j

Denotes that monitor is cadmium shielded l

2 l

l e

j I

TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E >1.0 MEV) AND LEAD FACTORS FOR PRAIRIE ISLAND UNIT NO. 2 SURVEILLANCE CAPSULES i

Capsule Azimuthal c(E>1.0Mev)

Lead Identification Location (n/cm2-sec)

Factor Il 13*

1.45 x 10 3.37 Il R

13' 1.45 x 10 3.37 10 T

23*

8.33 x 10

),94 10 P

23' 8.33 x 10 j,94 10 S

33' 7.67 x 10 j,79 10 N

33' 7.67 x 10

),79

4My, G

e 6

6-13

TABLE 6-4 CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF THE PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULES 2

Group Neutron Flux (n/cm -sec)

No Capsules V&R Capsules T&P Capsules S&N i

8 8

9 1

8.17 x 10 5.99 x 10 5.26 x 10 9

9 9

2 2.68 x 10 1.99 x 10 1.75 x 10 9

9 9

3 4.43 x 10 3.08 x 10 2.73 x 10 9

9 9

4 4.98 x 10 3.18 x 10 2.88 x 10 Y

9 9

5 8.66 x 10 5.20 x 10 A.75 x 10 10 10 9

6 1.70 x 10 1.01 x 10 o.2F x 10 10 10 10 7

2.46 x 10 1.41 x 10 1.30 x 10 10 10 10 8

3.53 x 10 1.97 x 10 1.83 x 10 10 10 10 9

4.67 x 10 2.53 x 10 2.35 x 10 10 10 10 10 5.04 x 10 2.67 x 10 2.48 x 10 ll 10 10 11 1.67 x 10 8.66 x 10 8.03 x 10 U

10 12 2.11 x 10 1.05 x 10" 9.76 x 10 10 10 10 13 9.42 x 10 4.65 y 10 4.34 x 10 10 10 10 14 7.11 x 10 3.52 x 10 3.28 x 10 I0 10 10 15 5.67 x 10 2.80 x 10 2.62 x 10 10 10 16 1.32 x 10" 6.41 x 10 5.99 x 10 i

U 10 10 17 1.03 x 10 5.07 x 10 4.73 > 10 10 10 18 1.06 x 10" 5.14 x 10 4.82 x 10 60 10 10 19 8.41 x 10 4.09 x 10 3.83 > 10 10 10 10 20 9.34 x 10 4.52 x 10 4.23 x 10 21 2.97 x 10" 1.51 x 10" 1.36 x 10" i

6-14

i l

TAS!,E 6-5 SPECTRUM AVERAGED REACTION CROSS SECTIONS AT THE 00SIMETER BLOCK LOCATION FOR PRAIRIE ISLAND UNIT NO. 2 SURVEILLANCE CAPSULES iT(barns)

Reaction Capsules V&R Capsules T&P Capsules S&N Fe64 (n p) Mn54 0.0505 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,1) F.P.

2.93 2.80 2.82 fo(E)c(E)dE j.

4;E)dE 1 Mew e

S 6-15

}

TABLE 6-6 IRRADIATION HISTORY OF PRAIRIE ISLAND UNIT NO. 2 REACTOR Vi.SSEL SURVEILLANCE CAPSULES P

F,,,

P /P Irradiation DE;:ay 3

j max Month (MW)

(MW)

Time Time (days)

(days) 12/74 78 1650

.047 13 2162 1/75 665 1650 403 31 21 31 i

2/75 1399 1650

.848 28 2103 3/75 987 1650

.598 31 2072 4/75 1526 1650

.925 30 2042 5/75 1551 1650

.940 31 2011 6/75 289 1650

.175 30 1981 7/75 1206 1650

.731 31

950 8/75 1487 1650

.901 31 1919 9/75 1365 1650

.827 30 1889 10/75 1223 1650

.741 31 1858 11/75 1544 1650

.936 30 1828 12/75 813 1650

.493 -

31 1797 1/76 432 1650

.262 31 1766 2/76 1323 1650

.802 29 1737 3/76 1292 1650

.783 31 1706 4/76-5/76 1145 1650

.694 61 1645 6/76 929 1650

.563 30 16i5 7/76 1391 1650

.843 31 1584 8/76~

1407 1650

.853 31 1553 9/76 1417 1650

.859 30 1523' 10/76 1195 1650

.724 21 1502 1

^

CAPSULE V REMOVED 10/75-11/76 0

1650

.000 40 1462 12/76 544 1650

.330 31 1431 1/77 1630 1650 988 31 1400 2/77 1526 1650

.925 28 1372 3/77 1620 1650

.982 31 1341 4/77 1637 1650

.992 30 1311 6-16 i

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

o TABLE 6-6 (cont'd) j/

Month j

max max Irradiation Decay (MW)

(MW)

Time Time (days)

(days) 5/77 1569 1650

.951 31 1280 6/77 1548 1650

.938 30 1250 7/77 1601 1650

.970 31 1219 8/77 1518 1650

.920 31 1188 9/77 1589 1650

.963 30 1158 10/77 1599 1650

.969 31 1127 11/77 470 1650

.285 30 1097 12/77 762 1650

.462 31 1066 1/79 1351 1650

.819 31 1035 2/78 1650 1650 1.00 28 1007 3/78 1638 1650

.993 7

31 976 4/78 1627 1650

.986 30 946 5/78 1502 1650

.91 0 31 915 5/78 1411 1650

.855 30 885 7/78 1419 1650

.860 31 854 8/7S 1429 1650

.866.

31 823 9/78 1591 1650

.964 30 793 10/78 1599 1650

.969 31 762 11/7S 1345 1650

.815 30 732 12/7S 738 1650

.447 31 701 1/79 1650 1650 1.00 31 670 2/79 1632 1650

.989 28 642 3/79 1642 1650

.995 31 611 4/79 1648 1650

.999 30 581 5/79 1440 1650

.873 31 550 6/79 1135 1650

.688 30 520 7/79 1530 1650

.927 31 489 8/79 1515 1650

.918 31 458 9/79 1549 1650

.939 30 428 10/79 1622 1650

.983

'll 397 11/79 1539 1650

.933 30 367 12/79 1546 1650

.937 31 336 1/S

! 1257 1650

.762 2

334 CAPSULE 7 REMOVED

Deca.s Time is referenced to 12/2/80 6-17

TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX M0 llTOR SATURATED ACTIVITIES F0P CAPSULE V i

l Reaction Radial Saturated Activity Adusted Saturated Activity and Location (dPs/gn)

(dPs/gn)

Axial I.ocstion (cm)

Capsule V Calculated Capsule V Calculated 54,,p)g,H g

Fe 6

6 6

Top 157.87 5.89 x 10 5.92 x 10 5.62 x 10 6

6 6

Top-Middle 157.87 5.49 x 10 5.92 x 10 5.24 x 10 6

6 6

i Middle 157.87 5.81 x 10 5.92 x 10 5.55 x 10 6

6 6

Bottom-Middle 157.87 6.04 x 10 5.92 x 10 5.76 x 10 6

6 6

Bottom 157.87 6.19 x 10 5.92 x 10 5.91 x 10 6

6 5.62 x 10 5.65 x 10 Average Cu63(n,a)Co60 5

5 5

- 3.28 x 10 4.86 x 10 l

Top-Middle 158.87 4.13 x 10 0

5 5

Bottom-Middle 158.87 4.53 x 10 3.28 x 10 5.33 x 10 5

5 5.10 x 10 3.86 x 10 Average I

Ni O(n,p)CoS8 7

7 7

7 i

Middle 158.87 7.43 x 10 7.1? x 10 8.73 x 10 8.37 x 10 i

1 Np (n f)Cs 7

7 7

Midd1r, 158.10 7.88 x 10 7.01 x 10 7.88 x 10 7.01 x 10 1

U238(n f)Cs137 6

6 6

6 Middle 158.10 9.18 x 10 7.72 x 10 9.18 x 10 7.72 x 10 i

3 k

i i

TABLE 6-8 i

COMPARIS'ON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR r

j SATURATED ACTIVITES FOR CAPSULE T

(

]

Reaction Radial Saturated Activity Adjusted Saturated Activity and Location (dPs/ge)

(dPs/qm)

Axial Location (cm)

Capsule T Calculated Capsule T Calculated I

p,54{g,p)g,54 6

6 6

l Top 157.87 4.01 x 10 3.92 x.10 3.80 x 10 6

6 6

Top - Middle 157.87 3.57 x 10 3.92 x 10 3.38 x 10 6

6 6

Middle 157.87 4.01 x 10 3.92 x 10 3.80 x 10 6

6 6

i Bottom-Middle 157.87 3.91 x 10 3.92 x 10 3.70 x 10 6

6 6

I Bottom 157.87 4.18 x 10 3.92 x 10 3.% x 10 6

6 Average 3.73 x 10 3.71 x 10 Cu63(n,n)Co60 5

5 5

Top - Middle 158.87 3.00 x 10 2.46 x 10 3.54 x 10 i

5 5

i Bottom-Middle 158.87 3.34 x 10 2.46 x 10 3.94 x 10 5

5 l

Average 3.74 x 10 2.90 x 10 NiS8(n.p) Cob 0 7

7 7

7 Middle 158.87 4.95 x 10 4.53 x 10 5.88 x 10 5.38 x 10 i

I Np 3 (n,f)Csl37 7

7 7

7 Middle 158.10 4.82 x 10 3.83 x 10 4.82 x 10 3.83 x 10 U238(n f)Csl37 6

6 6

6 Middle 158.10 5.71 x 10 4.60 x 10 5.71 x 10 4.60 x 10 l

9 I

TABLE 6-9 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULES V AND T I

Adusted Saturated Activity

+(E>1.0Mev) e (E> 1.0 '4ev) l l

Capsule Reaction (dPs/ge)

{,jc,2.sec)

(n/cm )

l Measured Calculated Measured Calculated Measured Calculated 6

6 N

18 18 54,,p)g,54 5.62 x 10 5.65 x 10 1.45 x 10" 1.45 x IG 5.86 x 10 5.86 x 10 g

V Fe I

CuG3(n.a)Co60 5

5 U

I8 5.10 x 10 3.86 x 10 1.91 x 10 7.72 x 10 S8 7

7 0

18 NiS8(n.p)Co 8.73 x 10 8.37 x 10 1.53 x 10 6.18 x 10 37 7

7 N

18 Np (n.f)Cs 7.88 x 10 7.01 x 10 1.63 x IO 6.59 x 10 U238(n.f)CsI37 6

6 U

18 f

9.18 x 10 7.72 x 10 1.73 x 10 6.99 x 10 10 I9 II T

Fe*(n.p)Mn54 6

6 8.37 x 10" 8.33 x 10 1.05 x 10 1.05 x 10 3.73 x 10 3.71 x 10 j

Cu63(n.a)Co60 5

5 U

I9 3.74 x 10 2.90 x 10 1.09 x 10 1.37 x 10 I

Ni 8(n.p)Co 8 7

10 II 5.88 x 10 5.38 x 10 9.16 x 10 1.15 x 10 Np237(n.f)CsI37 7

7 N

I9 4.82 x 10 3.83 x 10 1.04 x IO 1.30 x 10 l

U 38(n.f)Cs137 6

0 U

1.30 x 10 '

5.71 x 10 4.60 x 10 1.04 x 10 i

i t

i 1

4 4

i

TABLE 6-10 81 RESL'LTS OF THEPJML NEUTRON DOSIMETRY FOR CAPSULES V AND T Saturated Activity (dPs/gm)

Th (a)

Capsule Axial 2

location Bare Cd - Covered (n/cm -sec) 8 V

Top 1.42 x 10 Not Measured Not Determined 8

Bottom 1.51 x 10 Not Measured Not Detennined 7

T Top 7.65 x 10 Not Measured Not Determined

~

Bottom 7.49 x 10 Not Measured Not Determined iS (a) The average thermal ncetron flux measured within Capsule V from Il 2

Prairie Island Unit No. I was 1.51 x 10 n/cm -sec.

The Bare Activities listed above imply that the flux within Capsule T was 10 2

approximately 7.8 x 10 n/cm -sec.

1

.l i

TA8LE 6-11 i

SUPMARY OF NEUTRON DOSIMETRY RESULTS FOR CAPSULES V AND T 1

t Irradiation

$(E> 1.0 Mev)

+(E> 1.0 Mev)

Lead Vessel Calculated 2

Factor Fluence Vessel Fluence Capsule Time

(,f,2-sec.)

(n/cm )

2 j

(EFPS)

(,7c,2)

(n/cm )

1 II I8 18 18 j

V 4.04 x 10 1.45 x 10 5.86 x 10 3.37 1.74 x 10 j,74,jn 3

8 10 I9 18 18 T

1.26 x 10 8.33 x 10 1.05 x 10 1.94 5.41 x 10 5.41 x 10

?U t

i i

l I

e e

e

18 318 2C PRESSURE VESSEL SURVEILLANCE CAPSULE 0*

13* (CAPSULES V,R)

'///////

23* (CAPSULES T,P)

/

WM//

THERMAL SHIELD 33* (CAPSULES S,N)

MW WMH/

45a f//////////////////

)

I

/

/-

/

I

/

l l

' mmtwr I

/

i

/ l,

/

lj [/ /

REACTOR CORE l/ /

///

V Figure 61. Prairie Island Unit No. 2 Reactor Geometry 6 23

16.318-21 1

1 4

(13*, 23*, 33')

CHARPY (12*, 22', 32* )

[ SPECIMEN

/

Y

/

filliff //)

THERMAL SHIELD t

l l

I Figure 6 2.

Plan View of a Reactor Vessel Surveillance Capsule 6-24

~

l l

i8,318-22 2

1011 8

6 SURVEILLANCE CAPSULES 4

G Oz 8w m

2 "a

-ES y

PRESSURE VESSEL IR au.

1010 2

8 1/4T LOCATION E

6 4

3/4T LOCATION 2

108 0

10 20 30 40 50 AZIMUTHAL ANGLE (DEGREES)

Figure 6-3.

Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev)

Withing the Pressure Vessel Surveillance Capsule Geometry 6-25

18.318-23 10" 8

~

167.64

~

171.77 lR EO2 1/4T 175.90 8

~

4

~

55 i/2r iaw 5

g 8

g S

3/47 3

184.15 2

4 OR 2

HO 2

g PRESSURE VESSEL i

10' I

f I

I I

l l

l, 164 166 168 170 172 174 176 178 180 182 184 186 RADIUS (cm) i Figure 6-4.

Calculated Radial Distribution of Maximum Fast Neutron Flux I

(E > 1.0 Mev) Within the Pressure Vessel 6-26

_ _... _. _. _ _ _ _ _~.. _..._..

I 18.318-24 100 E

N 5

2 10'l X

3 u.

2 5

o

?

~

Dw z

w 2

2 H4

_awz 10-2 5

CORE MIDPLANE 2

= TO VESSEL CLOSURE HEAD 10-3 I

I l

l

-300

-200

-100 0

100 20u 300 DISTANCE FROM CORE MIDPLANE (CM)

Figure 6-5.

Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-27 1

L

18.318-25 1012 8

6 4

158.10

2 y

n 5

s5 1011 CAPSULE V s

8 zo 6 CAPSULE T ccy CAPSULE CENTER 54 2

THERMAL CAN CAN 0

TER PECIMEM HO 2

SHIELD g g

g.

2 1010 l

l 5 6 l

l l

3{

155 156 157 158 159 180 161 RADIUS (cm)

Figure 6-6.

Calculated Radial D.stribution of Maximum Fast Neutron Flux (E > 1.0 Mev)

Within Surveillance Capsult.s V and T 6-28

18.318 20 2

108 8

Nisa (n P) CoS8 6

4 Np237 (n,f) Cs 37 1

158.10 2

Ea N 10 7

~

3 8

U238 (n,f) Cs137 t-6 5

Q 4

Fe54 (n,P) Mn54 o

4 g

2 h

CAPSULE CENTER 106 8

6 4

Cu63 (n,a) Co60 2

TH "

^L

^

HO TEST SPECIMENS HO HiELD $

2 2

105 l

k l Nk l

l l

ck 15b 156 157 158 159 160 161 RADIUS (cm)

Figure 6-7.

Calculated Variation of Fast Neutron Flux Monitor Saturated Activity With n Capsule V 6-29

18,318-27 108 8

6

~

NiH (n,P) Coss 4

2 Np237 (n,f) Cs137 158.10

- 107 1

8 s

IC 3

6 D

~

E

~

U23s (n,f) Csl37 u

02 Fess (n,P) Mn54 E

$ 106 CAPSULE CENTER 8 2 l

6 4

Cu83 (n,a) Co#

2 HO SHIE

2 g

TEST SPECIMENS HO 5

fg 2

l 105 l

G l 9(

l l

l I'

156 156 157 158 159 160 161 RADIUS (cm)

Figure 6-8.

Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsule T 6-30

18.318 28 1021 8

6 4

2 13' CAPSULES (V) 1020 23' 8

CAPSULES (T)

N 6

~

VESSE L d

4 5

INNER RADIUS 3

u.

$2 (I:

b E

~

1018 8

O CAPSULE V DATA 6

O CAPSULE T DATA 4

SN CALCULATION 2

l 1018 l

l l

! l lll l

l l

l l lI!

l 100 2

4 6

8 101 2

4 6

8 102 l

OPERATING TIME (EFPY)

Figure 6-9.

Coreparison of Measured and Calculated Fast Neutron Fluence l

(E > 1.0 Mev) for Capsules V and T l

6-31

REFERENCES 1.

Yanichko, S. E., Lege, D. J., " Northern States Power Company Prairie Island Unit No. 2 Reactor Reactor Vessel Radiation Surveillance Program", WCAP-8193, Scotember 1973.

2.

ASTM Designation E-185-73, " Surveillance Tests for Nuclear Reactor Vessels" in " ASTM Standards (1974), Part 10. pp. 314-320, Am. Soc. for Testing and Materials, Philadelphia, Pa.1974.

3.

Davidson, J. A., Yanichko, S. E., and Anderson, S. L., " Analysis of Capsule V from Northern States Power Company Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program", WCAP-9212, November.1977.

4.

U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Standard Review Plan, NUREG-75/087, Sect. 5.3.2, " Pressure-Temperature Limits", November 1975.

Soltesz, R.

G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket

~

5.

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.

Collier, G., et. al., "Second Version of the GAMBIT Code", WANL-TME-1969, November 1979.

Soletsz, R. G., et. al., " Nuclear Rocket Shielding Methods, Modification, 7.

Updating and Input Data Preparation - Volume 3, Cross-Section Generation and Data Processing Techniques", WANL-PR(LL)-034, August 1970.

Soletsz, R.

G., e. al., " Nuclear Rocket Shielding Methods, Modification, 8.

Updating and Input Data Preparation - Volume 4 - One-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)034, August 1970 ASTM Designation E261-70, Standard Method for Measuring Neutron Flux by 9.

Radioactivation Techniques", in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 745-755, Am. Society for Testing and Materials, Philadelphia, Pa., 1975

10. ASTM Designation E262-70, " Standard Method for Measuring Thennal Neutron Flux by Radioactivation Techniques", in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 756-763, Am. Society for Testing and Materials, Philadelphia, Pa. 1975.

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.

S 6-32

REFERENCES (cont'd)

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 flaterials, Philadelphia, Pa., 1975.

L O

e O

6-33

APPENDIX A HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION

~

A-1.

INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RT of the NDT material in the core region of the reactor vessel is determined by using the preservice reactor v3ssel material properties and estimating the radiation-in-duced ART:DT. RT is designated as the higher of either the drop weight t

NDT nil-ductility transition temperature (TNDT) r the temperattre at which the material exhibits at least 50 ft 1b of impact energy and 35-mil lateral ex-pansion (normal to the major working direction) minus 60*F.

RT increases as the material is exposed to fast-neutron radiation. Thus, NDT at any time period in the reactor's life, to find the most limiting RTNDT ART due to the icdiation exposure, associated with that time period must NDT The extent of the shift in be added to the original unirradiated RTNDT.

RT is enhanced by certain chtaical elements (such as copper and phosphorus)

NDT present in reactor vessel steels. The Regulatory Guide 1.99 trend curves which

0" show the effect of fluence and cepper and phosphorus contents :,n ARTNDT reactor vessel steeis are shown in figure A-1.

Given the copper and phosphorus contents of the most limiting material, the radiation-induced ARTilDT can be estimated from figure A-1.

Fast-neutron fluence (E > 1 Mev) at the 1/4 T (wall thickness) and 3/4 T (wall thickness) vessel loca-tions are given as a function of full-power s'ervice life in figure A-2. The data for all other ferritic materirls in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RTNDT*

O A-1 I

e l

1

. ! 444444" "4444"._...

' A = [40 + 1000 (% Cu - 0.08) + 5000 (% P - 0.008)] [f/1018l1/2 R

L

- l'i

! i 111) i : i i 1

lifi II i$ ~

i Hi FW I

{

j-5 5 t

H is T'"i i Epg((.-hj h [l }

i l 1 l) t i le I

h[W 3,,

'}!! !

l ll j i l' -

~

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'"' 5')/

8,P 1

3 !!!

l 1f ll!

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.- -}j, h'

"']e{;,Q y

l-

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l 200

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v i, d g y

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[

{

a

, Ts* L

.p I

._~;. ' I.6,

'-t a

,,4 c

l

?

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s

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-l

,7j v

{

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l 100'

--v C

~

r-

~

{whh E

k.

f i

d

}ll

',)-

~~'

~

~ h

' ~

r

=

RT i

g [e,R gi,y x-

~

/

2p

}p*

l#p? tm nii Y

5h

' k f

.,,f

.,,(..".b ?.

5 3

b$b fT

. p Y

gg ank b

f b

{

l N f

g

[ ] ],. I 0.35 0.30 i

U.O jl.!ll!Ll 1 j... Llll}F

~ l'I

%P = 0.Ud 'l f,,.; j.

- hp

% P = 0.008 r.

0.25 0.20 % Cul 0.15% Cu l

0.10% Cu I

,i s

OWER LIMIT

!:,i;!p;ll:iW l

.I l

' il i 'l!.1 !

l-4 Il,...

-l

~

% Cu = 0.08

--- A Weld Metal

. d i

a{

l 3 I$

$lj

/41L l

1 9 Shell Forging 22642 @

I j!!

fif fillillfillll l a llilllilllililillfillllliffl '

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r

~----

+

l I

i,I I

2X1017 18 i

4 6

8 10 2

4 6

8 10 2

4 6

18

}

FLUENCE, n/cm2 (E > IMeV)

Figure A-1 Effect of Fluence, Copper Content, and Phosphorus Content i

on ART 1.99 MT for Reactor Vessel Steels per Regulatory Guide i

s e

i i

I

I i

10.927 2 e

i I

3 i.

2 1/4T 10'9 8

6 3/uT i

i 4

i "E

y 1

g 2

=

o d

=

E 1018

~

O 8

9

=

6 I

~

14 7

I i

2 i

L 10'7 I

0 5

10 15 20 25 30 l ',

SERVICE LIFE (EFFECTIVE FULL POWER YEARS) i t

Figure A 2.

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

of Full Power Service Life 1

A3 i

a A-2.

FRACTURE TOUCHNESS PROPERTIES The preirradiation fracture-toughness-properties of the Prairie Island Unit 2 reactor vessel materials are presented in table A-1.

The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan.bl3 The postirradiation fracture toughness properties of the reactor vessel belt-

~

line material were obtained directly from the Prairie Island Unit 2 Vessel Material Surve111anca Program.

A-3.

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATI0flSHIPS The ASME aporoach for calculating the allowable limit curves for various heat-up and cooldown rates specifies that the total stress intensity factor, K,

7 for the combined ther:al and pressure stresses

.t any time during heatup and cooldovm cannot be greater than the reference :ress intensity factor, K gg, for the ir,etal temperature at that time. K is obtained from the reference IR fracture toughness curve, defined in Appendix G to the ASME Codt.E23 The K IR curve is given by the equation:

Kyg=26.78+1.223exp[0.0145(T-RTNDT+160)]

(A-1) where K is the reference stress intensity factor as a function of the gg metal temperature T and the metal reference nil-ductility temperature RTNDT*

Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G to the ASME Code [2] as follows:

I IM + It 1 IR (A-2) 1.

" Fracture Toughness Requirements", Branch Technical Position MTEB flo. 5-2, Section 5.3.2-14 in Standard Review Plan, ilVREG-75/087,1975.

2.

ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Appendices,

" Rules for Constraction of i;uclear vessels", Appendix G, " Protection Against !!onductile Failure", pp. 461-469, 1980 Edition, American Society of Mechanical Engineers, tiew York, 1980.

{

A-4 l

l l

~

i TABLE A 1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADI ATED)

Transverse *I I

50 ft Ib/35 mils Cu P

NDTT Lateral Espansion NDT Average Transverse *I HT I

l Component Material Type

(%)

{%)

( F)

Temp ( F)

("Fi Upper Shelf (f t ib)

ICI 5

64 ICI Closuse Head Do.ne A533 Gr. B, Cl.1 5

52 Head Flaruje A508 0. 3

-31 laICI

-3I 8'7ICI ICI ICI

-22 88 Vessel Fiange A508 Q. 3

-22 18

!C!

ICI

-22 97 Injection Nonles A508 Cl. 3

-22

-114 ICI 3=

lCI

-10 89 Es intet and Outle Norile A508 O. 3

-13 50 ICI ICI

-13 85 Upper Shell A508 Q. 3

-13 41 ibl A508 C. 3 0.075 0.010

-4 56

-4 112 inter. Shell IDI A508 O. 3 0 085 0.0 e 1

-13 54

-6 108 Lower Shell ICI Trans. Ring A508 C. 3 10 50 10

/G ICI Bottom ficad.

A533 Gr. B, Cl.1

-13 53

-4 68 Weidment Weld 0.082 0.019

-31

-6

-:11 103 HAZ llAZ

-31

-35

-:t1 111 a Specerren orienerd n.=een.se to the mgoe working diret-teon in. Based on ac tual gran <. ew data through the surveillaeu:e gwis asasn Estimated uwe=e "Pww.ee-Tmgw: ature Linues." Section 5 3 2 of Standard #cwew Fran, NUI1EG 75/ Ort?,

c.

1975. from longitudi

.e data

-n.

.n_...-

9 m

where K

is the stress intensity factor caused by membrane (pressure) stress gg K

is the stress intensity factor caused by the themal gradients It K

is a function of temperature to the RTNOT of the material IR C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical

\\

At any time during the heatup or cooldown transient, K is determined by the IR metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The the/inal stresses resulting from temperature gradients through the vessel wall are calculated and then the corres::crding (themall stress intensity fa:: tors, Kgg, for the refer-ence flaw are computed. Fromequation(A-2),thepressurestressintensity factors are obtained and, from these, the allowable p-essures are calculated.

For the calculation of the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pres-sure-temperature relations are generated for both steady-state and finite cooie down rate situations. From these relations, composite limit curves are con-structed for each cooldown rate of interest.

The use of the composite (.urve in the cocidown analysis is necessary because controi of the cooldown procedure is based on measurement of reactor coolant i

temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the. fluid adjacent to the vessei

~

10. This condition, of course, is not true for the steady-state situation.

It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T location for finite cool-IR down rates than for steady-state operation. Furthemore, if conditions exist such that the increase in K exceeds Kgg, De calcdated aHowaMe pressure IR during cooldown will be greater than the steady-state value.

A-6

e The above procedures are needed because there is no direct control on tempera-ture at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period.

Three separate calculatiens are recuired to determine ths limit curves for finite heatup rates.

As is done in the cooldown analysis, allowable pressure-tercera-ture relattensnips are developeo for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The themal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses prednced by in-ternal pressure. The metal temperature at the crack tip lags the coolant tempera-ture; therefore, the X for the 1/4T crack during heatup is lower than the IR K

for the 1/4T crack during steady-state conditiens at the same coolant IR temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of :ompressive thermal stresses and lower K IR's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases i

have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for stear.iy-state and finite heatup rates is obtained.

1 The second portion of the heatup analysis concerns the calculation of pressure-l temperature limitations for the case in which,a 1/4T deep outside surface flaw i

is assumed. Unlike the situation at the vessel inside surface, the themal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These themal stresses are dependent on both the rate of heatuo and the time (or coolant temperature) along the heatup ramp. Since the themal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

I A-7 i

i

Following the generation of pressure-temperature curves for both the steady-stats and finite heatup rate situations, the final limit curves are produced as fol-lows: A composite curve is constructed based on a point-by-point conparison of the steady-state and finite heatup rate data. At any given temperature, the allow-able pressure is taken to be the lesser of the three values taken from the curves under consideration. The un of the composite 'urve is necessary to set conserva-tive heatup limitations because it is possible for qonditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to iihe outside and the pressure limit must at all times be based on analy-sis of the most critical criterion. Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperatt.re sensing instruments by the values indicated on the respective Curvks.

A-4.

HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in paragraph A-3.

The deriva-tion of the limit curves is presented in the NRC Regulatory Standard Review Plan.

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

Charpy test specimens from Capsule T indicate that the core region weld metal and the limiting core region shell fo ging (22642) exhibited maximum shifts in RT cf 60*F and 45"F, respectively as shown by figure A-1.

The shifts are well NDT within the appropriate design curve (figure A-1) prediction. Heatup cnd cooldown limit curves for normal operation up to 32 effective.N11-power ears (EFPY) a re presented in the Capsule V radiation surveillance program report.

The heatuo and cooldown curves were based on the aRTNDT predicted by the Westinghouse trend curves and the core region weld metal which is the limiting vessel material.

1.

" Pressure-Temperature Limits", Secthn 5.3.2 in Standard Review Plan, NUREG-75/087,1975.

2.

Davidson, J.A., Yanichko, S.

E., and Anderson, S.

L., " Analysis of Capsule V from Northern States Power Company Prairie Island Unit No Radiation Surveillance Program", WCAP-5212, November 1977

2. Reactor Vessel A-8

9 However, since the Regulatory Guide 1.99 trend curves predict a larger shift in RT than the Westinghouse trend curves for the Prairie Island Unit 2 core region NDT weld material, the heatup and cooldown curves were reevaluated using Regulatory Guide 1.99 predictions. Based on this reevaluation it was determined that the heatup and cooldown curves are appropriate for use up to 20 EFPY. These heatup and cool-down veves are shown in figures A-3 and A-4, respectively.

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-tecperature cocbinaticns are to the right of the criticality limit line, shown in figure A-3.

This is 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 of reference.[1,2]

Figures A-3 and A-4 define limits for insuring prevention of nonductile failure.

1.

" Pressure-Temperature Limits", S9ction 5.3.2 in Standard Review Plan, NUREG-75/087,1975.

2.

ASME Boiler and pressure Vessel Code,Section III, Division 1 - Appendices,

" Rules for Construction of t;uclear Vessels", Appendix G, " Protection Against Nonductile Failure", pp 461-469, 1980 Edition, American Society of Mechanical Engineers, New York,1980.

I l

e s

A-9

10.327-27 e

3000 C

rd'E4.at c 4 0P i 4 * ' 34 sis:

2800 3 = ELD "ETAL Co = 0.Os2 LEaa Es:.. i 3 INITIAL RT40,= -20*F (CONSERVATIVELY 2600 Assudo) 3 AT 20 EFFECTIVE FULL power YEARS "n

R T,3. AT I 3: iniC anEss = i iO*F

.,,,l 2400 3

RT C

_ 2200 NDT AT 3/47 THICruts3 = 7s F I,

f f

r e,

f f

i 1

m

,z-1 S.2000

f f

s'

/

w i,,

1 3,.1800 e

,f y,

w i600

=

E i

['

,/

8.I400 Q

UNACCEPTABLE ACCEPT ABLE -

i u 1200 OPERATION PERATich C 2

i, i 1

r x

i a

1

- 1000

.,s CRITgCAL:TY L t M i T -=

ii i

i e i s

800 HE ATUP RATES UP

!f 600 T0 ic0'r/nR I

I

'. mm 400 l

i i, i 1

)

I 3

i T

200 t'

~

t I

1 I

t 0

0 50 100 150 200 250 300 350 llI0lCATED TEMPERATURE ( F)

I l

Figure A 3.

Prairie Island Unit 2 Reactor Coctant System Heatup Limitations Applicable for Periocs up tc' 20 Effective Full Power Years.' \\1argm.s of 60 psig and 10 F are included for Possible instrument Error A-10

]

e 0,927-28

4 2600 va E 4.1-, P:0DE;!) 51515 2400 M WELD METAL Cw:0.092

[

i i i 2200 INITIAL RTNDT: -20 c5 (CONSERvATivEL) AS ScMD )

/!

AT20 EFFECTIVE FULL POWER YEA 45 i

2000 RT AT l/uT THICKhESS : 1IC F l

,,f C

N07 7

RTNDT.AT 3/uT THICNNESS 7a'F

'f '

l 1300 1

2 l,','.'

i i,

1600 l

l f,';'

m g~

m 1400 w

1 x

l a-1200

,,i i,'

D w

I

. M i

1000

> s y" '

o

~

1 g

800 koo a RATES

-s,

. M [ ,,

l 600 II' #E'.

i i

0 -- -

2,g y" ii i,

400

+

i i too -

- i 1

i 1

i >

i i

, i

, I

'I 200 n

' 'i 0

i O

50 100 150 200 250 300 INDICATED TEMPERATURE (OF) 1 l

l Figure A-4 Prairie Island Unit 2 Reactor Coolant System Cooldown Limitations Applicable for Periods up to 20 Effective Full Power Years. Margins of 60 psig and 10*F are included for Possible Instrument Error i

(

A-11

-~

^

ML19347F376

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