ML20213A574

Analysis of Capsule R from Prairie Island Unit 2,Reactor Vessel Radiation Surveillance Program

ML20213A574
Person / Time
Site:	Prairie Island
Issue date:	12/31/1986
From:	Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20213A577	List: ML20213A574
References
TAC-42500, WCAP-11343, NUDOCS 8702030366
Download: ML20213A574 (100)

Text

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WCAP-11343 WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSULE R FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM p

1 S. E. Yanichko J. C. Schmertz December 1986 Work performed under Shop Ordsr No NPAJ-108 hIhl44/A APPROVED:

T.A.Heyer.banager Structural Materials and Reliability Technology Prepared by Westinghouse for the Northern States Power Company Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORAi!ON Power Systems Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230 mr.mem l

0702030366 070127 PDR ADOCK 05000306 P

PDR

PREFACE This report (WCAP-11343) has been technically reviewed and verified.

Reviewer Sections 1 through 5 and 7, 8 C. C. Heinecke deerv[ / /cy k j d ',.

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Section 6 S. L. Anderson

.,4.y (1,,;1rnw Appendix A H. Gong 1.. A, w.io4.iio.

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TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 1-1 2

INTRODUCTION 2-1 3

BACKGROUND 3-1 4

DESCRIPTION OF PROGRAM 4-1 5

TESTING OF SPECIMENS FROM CAPSULE R 5-1 5-1.

Overview 5-1 5-2.

Charpy V-Notch Impact Test Results 5-3 5-3.

Tension Test Results 5-5 5-4.

Wedge Opening Loading Tests 5-5

~

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

Introduction 6-1 6-2.

Discrete Ordinates Analysis 6-1 6-3.

Radiometric Monitors 6-4 6-4.

Neutron Transport Analysis Results 6-8 6-5.

Desimetry Results 6-9 7

SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8

REFERENCES 8-1 Appendix HEATUP AND C00LDOWN LIMIT CURVES FOR A-1 A

NORMAL OPERATION A-1.

Introduction A-1 A-2.

Fracture Toughness Properties A-2 A-3.

Criteria for Allowable Pressure-Temperature A-2 Relationships A-4.

Heatup and Cooldown Limit Curves A-5 w.iomim y

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LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the 4-5 Prairie Island Unit 2 Reactor Vessel 4-2 Capsule R Diagram Showing Location of Specimens, 4-6 i

Thermal Monitors, and Dosimeters 5-1 Irradiated Charpy V-Notch Impact Properties for 5-19 Prairie Island Unit 2 Reactor Vessel Lower Shell Forging 22642 (Axial Orientation) 5-2 Irradiated Charpy V-Notch Impact Properties for 5-20 Prairie Island Unit 2 Reactor Pressure Vessel Lower Shell Forging 22642 (Tangential Orientation) 5-3 Irradiated Charpy V-Notch Impact Properties for 5-21 Prairie Island Unit 2 Reactor Pressure Vessel Weld Metal 5-4 Irradiated Charpy V-Notch Impact Properties for 5-22 Prairie Island Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Material 5-5 Irradiated Charpy V-Notch Impact Properties for 5-23 Prairie Island Unit 2 Reactor Pressure Yessel A533 Grade B Class 1 Correlation Monitor Material 5-6 Charpy Impact Specimen Fracture Surfaces for Prairie 5-24 Island Unit 2 Reactor Pressure Vessel Lower Shell Forging 22642 (Axial Orientation) 5-7 Charpy Impact Specimen Fracture Surfaces for Prairie 5-25 Island Unit 2 Reactor Pressure Vessel Lower Shell Forging 22642 (Tangential Orientation) 5-8 Charpy Impact Saecimen Fracture Surfaces for 5-26 Prairie Island Jnit 2 Reactor Pressure Vessel Weld Metal 5-9 Charpy Impact Specimen Fracture Surfaces for 5-27 Prairie Island Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-10 Charpy Impact Specimen Fracture Surfaces for 5-28 Prairie Island Unit 2 Reactor Pressure Vessel A533 Grade B Class 1 Correlation Monitor Material 5-11 Comparison of Actual versus Predicted 30 ft Ib 5-29 Transition Temperature Increases for the Prairie Island Unit 2 Reactor Vessel Material Based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 m e.ie4 iio.

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LIST OF ILLUSTRATIONS (Cont)

Figure Title Page 5-12 Tensile Properties for Prairie Island Unit 2 Reactor 5-30 Vessel Lower Shell Forging 22642 (Axial) 5-13 Tensile Properties for Prairie Island Unit 2 Reactor 5-31 Vessel Lower Shell Forging 22642 (Tangential) 5-14 Tensile Properties for Prairie Island Unit 2 Reactor 5-32 Vessel Weld Metal 5-15 Fractured Tensile Specimens of the Prairie Island 5-33 Unit 2 Reactor Vessel Lower Shell Forging 22642 (Axial Orientation) 5-16 Fractured Tensile Specimens of the Prairie Island 5-34 Unit 2 Reactor Vessel Lower Shell Forging 22642 (Tangential Orientation) 5-17 Fractured Tensile Specimens of the Prairie Island 5-35 Unit 2 Reactor Vessel Weld Metal 5-18 Typical Stress-Strain Curve for Tension Specimens 5-36 6-1 Prairie Island Unit 2 Reactor Geometry 6-24 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-25 6-3 Calculated Azimuthal Distribution of Maximum 6-26 Fast (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel Surveillance Capsule Geometry 6-4 Relative Radial Variation of Fast Neutron (E > 1.0 MeV) 6-27 Flux and Fluence Within the Pressure Vessel 6-5 Relative Axial Variation of Fast-Neutron (E > 1.0 MeV) 6-28 Flux and Fluence Within the Pressure Vessel Wall A-1 Effect of Fluence, Copper Content, and Phosphorus A-8 Content on ART for Reactor Vessel Steels perRegulatorypuide1.99, Revision 1 N

A-2 Fast Neutron Fluence (E > 1.0 MeV) as a Function A-9 of Full Power Service Life (EFPY)

A-3 Prairie Island Unit 2 Reactor Coolant System A-10 Heatup Limitations Applicable for the first 15 EFPY A-4 Prairie Island Unit 2 Reactor Coolant System A-11 Cooldown Limitations Applicable for the first 15 EFPY mr. io,en t..o n u r yggj

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4 LIST OF TABLES Table Title Page o

4-1 Chemical Composition of the Prairie Island Unit 2 4-3 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the Prairie Island Unit 2 4-4 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the Prairie Island Unit 2 5-6 Lower Shell Forging 22642 (Axial) Irradiated k

at 550*F, Fluence 4.42 x 10" n/cm8 (E > 1 MeV) b 5-2 Charpy V-Notch Impact Data for the Prairie Island Unit 2 5-7 Lower Shell Forging 22642 (Tangential) Irradiated at 550'F, Fluence 4.42 x 10" n/cm (E > 1 MeV) 2 5-3 Charpy V-Notch Impact Data for the Prairie Island Unit 2 5-8 Pressure Vessel Weld Metal Irradiated at 550*F, Fluence 4.42 x 10" n/cm2 (E > 1 MeV) 5-4 Charpy V-Notch Impact Data for the Prairie Island Unit 2 5-9 Pressure Vessel Weld Heat Affected Zone Metal Irradiated at 550*F, Fluence 4.42 x 10" n/cm 8 (E > 1 HeV) 1 i

5-5 Charpy V-Notch Impact Data for the Prairie Island Unit 2 5-10 g

Pressure Vessel A533 Grade B Class 1 Correlation Monitor Material Irradiated at 550*F, Fluence 4.42 x 108' n/cm2 (E > 1 MeV) 5-6 Instrumented Charpy Impact Test Results for Prairie Island 5-11 Unit 2 Lower Shell forging 22642 (Axial Orientation) 5-7 Instrumented Charpy Impact Test Results for Prairie Island 5-12 Unit 2 Lower Shell Forging 22642 (Tangential Orientation) 5-8 Instrumented Charpy Impact Test Results for 5-13 Prairie Island Unit 2 Weld Metal 5-9 Instrumented Charpy Impact Test Results for E-14 Prairie Island Unit 2 Weld Heat Affected Zone Metal 5-10 Instrumented Charpy Impact Test Results for Prairie 5-15 Island Unit 2 Correlation Monitor Material (HSST Plate 02) 5-11 The Effect of 550'F Irradiation at 4.42 x 10$' n/cm2 5-16 s

(E > 1 MeV) on the Notch Toughness Properties of The Prairie Island Unit 2 Reactor Vessel Materials sw w wr,munt 4,

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LISr OF TABLES (Cont)

Table.

Title Page 5-12 Summary of Prairie Island Unit No. 2 Reactor Vessel 5-17 Surveillance Capsule Charpy Impact Test Results 1

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Tensile Properties for Prairie Island Unit 2 Reactor Vessel 5-18 N-

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Material Irradiated to 4.42 x 10" n/cm2 c,

. 6-l '

SAILOR 47 Neutron Energy Group Structure 6-11

.6-2 Nuclear Constants for Radiometric Monitors Contained 6-12 in the Prairie Island Unit 2 Surveillance Capsules

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6-3 Calculated Fast Neutron Exposure Parameters for the 6-13 l

Peak Location of the Prairie Island Unit 2 Reactor Vessel s

i 6-4

. Calculated Fast Neutron Exposure Parameters and Lead 6-14 Factors for the Prairie Island Unit 2 Surveillance Capsules 6-5 Calculated Neutron Energy Spectrum at the Center of 6-15 Prairie Islind Unit 2 Surveillance Capsule R 6-6 Spectrum-tseraged. Reaction Cross Sections at the 6-16 tenter ot Prairie Island Unit 2 Surveillance Capsule R 6-7 Irradiation History of Prairie Island Unit 2 6-17 x

- Surveillance Capsule R 6-8.-

Comparison of Measured and Calculated Radiometric 6-18 Monitor Saturated Activities for Prairie Island Unit 2 Surveillance Capsule R 6-91

' Results of Fast Neutron Dosimetry for Prairie Island 6-21 1

n Unit 2 Surveillance Capsule R

_[-

6-10 Results of Thermal Neutron Dosimetry for Prairie Island 6-22 Unit 2 Surveillance Capsule R 6-11 Summary of Prairie Island Unit 2 Fast Neutron 6-23 Fluence Results Based Upon Surveillance Capsule R

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Reactor Vessel Toughness Data (Unirradiated)

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SECTION 1 SIM ERY OF RESULTS The analysis of the reactor vessel material contained in Capsule R the third surveillance capsule to be removed from the Northern States Power Ccmpany Prairie Island Unit 2 reactor pressure vessel, led to the following conclusions:

The capsule received an average fast neutron fluence (E > 1.0 MeV) o 19 2

of 4.42 x 10 n/cm.

Irradiation of the reactor vessel lower shell forging 22642, to 4.42 x o

19 10 n/cm, resulted in 30 and 50 ft-lb transition temperature increases of 85'F and 95'F, respectively for specimens oriented normal to the major working direction (axial orientation) and increases of 100*F and 115'F, respectively for s;iecimens oriented parallel to the major working direction (tangential orientation).

o Weld metal irradiated to 4.42 x 1019 2

n/cm resulted in 30 and 50 ft-lb transition temperature increases of 100*F and 125'F respectively, 19 2

o Due to irradiation to 4.42 x 10 n/cm the average upper shelf energy of forging 22642 (axial orientation) decreased from 108 to 98 ft-lbs and the limiting weld metal decreased from 103 to 91 ft-lbs. Both materials exhibit a more than adequate shelf level for continued safe plant operation.

The A533 Grade B Class 1 Correlation Monitor material (HSST Plate 02) o 19 2

irradiated to 4.42 x 10 n/cm resulted in a 30 and 50 ft-lb transition temperature increase of 180*F and 190*F respectively.

1-1

o Comparison of the 30 ft-lb transition temperature increases for the Prairie Island Unit 2 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1, shows that the forging material and weld metal transition temperature increase were less than predicted or that the embrittlement was less than predicted.

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1-2

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l SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule R, the third l

capsule to be removed from the reactor in the continuing surveillance program l

which monitors the effects of neutron irradiation on the Northern States Power Company Prairie Island Unit 2 reactor pressure vessel materials under actual l

operating conditions.

The surveillance program for the Northern States Power Company 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 properties of the reactor vessel materials are presented by Yanichko and Lege.[1] The surveillance program was planned I

to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-70, " Recommended Practice for Surveillance Tests for Nuclear Reactors".[2] Westinghouse Nuclear Energy Systems personnel were contracted to aid in the preparation of procedures for removing the 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.

1 j

This report summarizes testing and the postirradiation data obtained from surveillance Capsule R removed from the Northern States Power Company Prairie Island Unit 2 reactor vessel and discusses the analysis of the data. The data are also compared to Capsules V and T which were previously removed from the l

reactor.

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2-1

SECTION 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry..The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA508 Class 3 (base material of the Prairie Island Unit 2 reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Co'de. The method utilizes fracture mechanics concepts and is based on the reference nil ~ ductility temperature (RTNDT)*

RT is defined as the greater of either the drop weight nil-ductility NDT transition temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 ft lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. The RT f a given material is used to index that NDT material to a reference stress intensity factor curve (KIR curve) 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 t

= >. " " "

3-1

the K curve, allowable stress intensity factors can be obtained for this IR material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

RT and, in turn, the operating limits of nuclear power plants can be NDT adjusted 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 Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program, in which 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 30 ft lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RT f r radiation embrittlement. This adjusted RT NDT NDT (RT initial + ARTNDT) is used to index the material to the KIR NDT 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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3-2

SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Prairie Island Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in figure 4-1.

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

Capsule R (Figure 4-2) was removed after 8.81 effective full power years of plant operation. This capsule contained Charpy V-notch impact, tensile, and wedge opening loading (WOL) fracture mechanics specimens from the reactor vessel

-lower shell ring Forging 22642, submerged arc weld metal representative of the beltline weld seams of the reactor vessel and Charpy V notch specimens from weld heat-affected zone (HAZ) material. All heat-affected zone specimens were obtained from within the HAZ of Forging 22642 of the representative weld. The capsule also contained Charpy V notch specimens from the 12 inch thick correlation monitor material A533 Grade B Class 1 (HSST Plate 02) furnished by Oak Ridge National Laboratory.

The chemistry and heat treatment of the surveillance material are presented in table 4-1 and table 4-2, respectively.

The chemical analyses reported in table 4-1 were obtained from unirradiated material used in the surveillance program.

In addition, a chemical analysis was performed on irradiated Charpy specimens from the lower shell forging and weld metal and is reported in table 4-1.

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. All base metal Charpy V-notch impact and av.="""

4-1

tensile specimens were oriented with the longitudinal axis of the specimen both normal to and parallel to the principal working (hoop) direction of the forging. 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 specimens parallel to the welding direction. The WOL test specimens in Capsule R were machined such that the simulated crack in the specimen would propagate normal (tangential) and parallel (axial) to the major working (hoop) direction for the forging specimens and parallel to the weld direction. All specimens were fatigue precracked per ASTM E399-70T.[3]

Capsule R contained dosimeter wires of pure iron, copper, nickel, and unshielded aluminum-cobalt.

In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were contained in the capsule.

Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2.

The two eutectic alloys and their melting points are:

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

The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule R are shown in Figure 4-2.

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TABLE 4-1 CHEMICAL COMPOSITION OF THE PRAIRIE ISLAND UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS A533 Gr B, CL1 Correlation Lower Shell Weld Metal [c]

Monitor Material Element Forging 22642 Analysis (HSST Plate 02)

(a)

(b) s C

0.175 0.174 0.045 0.047 0.22 S

0.013 0.011 0.014 0.014 0.018 N

0.017 0.026 2

Co 0.026 0.015 0.013 0.009 Cu 0.085 0.068 0.082 0.076 0.14 Si 0.285 0.26 0.47 0.42 0.25 Mo 0.445 0.47 0.51 0.50 0.52 Ni 0.70 0.694 0.072 0.071 0.68 Mn 1.22 1.21 1.37 1.19 1.48 Cr 0.14 0.147 0.02 0.011 V

<0.008

<0.01 0.001

<0.01 P

0.011

<0.005 0.019 0.015 0.012 Sn 0.011 0.002 A1 0.036 0.007 Ti 0.002

<0.001 (a) Analysis performed on irradiated Charpy forging specimen NL17 (b) Analysis performed on irradiated Charpy weld specimen NW12 (c) Surveillance weld specimens were made of the same wire and flux as the Intermediate to lower shell circular seam (UM 40 Wire Heat 2721 and UM 89 Flux Lot 1263) m.m.n o.

4-3

TABLE 4-2 HEAT TREATMENT OF THE PRAIRIE ISLAND UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS Material Temperature (*F)

Time (hr)

Coolant Intermediate 1652*/1715' 5 hrs.

Water quenched Shell Forging 22642 1175'/1238' 5 hrs.

Furnace cooled 1652*/1724' 5 1/2 hrs.

Water quenched 1202*/1238' 5 hrs.

Furnace cooled 1022*

11 1/2 hrs.

Furnace cooled 1112' 7 hrs.

Furnace cooled Weldment 1022' 5 hrs.

Furnace cooled 1112' 7 hrs.

Furnace cooled Correlation Monitor 1675* 1 25' 4

Air cooled HSST Plate 02 1600* 1 25' 4

Water quenched 1125' 1 25' 4

Furnace cooled 1150' 1 25' 40 furnace cooled to 600*F ner.ao/unn 4_4

t P

REACTOR VESSEL THERMAL SHIELD I

l CAPSULE 10' (TYP) l 57*

10*

180*

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Figure 4-1. Arrangement of Surveillance Capsules in the Prairie Island Unit 2 Reactor Vessel OO7-A-19690-1 4-5

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SPECIMEN NUMBERING CODE:

NT - LOWER SHELL FORGING 22642 (AXIAL ORIENTATION)

NL - LOWER SHELL FORGING 22642 (TANGENTIAL ORIENTATION NW - WELD METAL l

NH - WELD HEAT-AFFECTED-ZONE METAL I

R - ASTM CORRELATION MONITOR 1

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THERMAL SHIELD VESSEL WALL TI APERT M CARD Aleo Avage un Aperture Card

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Figure 4-2. Capsule R Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-6 50003D%b-o\\

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SECTION 5 TESTING OF SPECINENS FROM CAPSULE R 5-1.

OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H, ASTM Specification E185-82 and Westinghouse Procedure NHL 8402, Revision 0

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as modified by RMF Procedures 8102 and 8103.

Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in.WCAP-8193.E13 No discrepancies were found.

Examination of the two low-melting 304*C (579'F) and 310*C (590*F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed waslessthan304*C(579'F).

i The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74, 358J machina. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (E ).

From the load-time curve, the load of general yielding (PGY),the D

time to general yielding (tgy), the maximum load (P ), and the time to M

maximum load (t ) can be determined. Under some test conditions, a sharp y

i mer.."""

5-1 W

d

i drop in load indicative of fast fracture was observed. The load at which fast fracture was initisted is identified as the fast fracture load (P ), and the F

load at which fast fracture terminated is identified as the arrest load (P )*

A The energy at maximum load (E ) was determined by comparing the energy-time M

record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen.

Therefore, the propagation energy for the crack (E ) is the difference p

between the total energy to fracture (E ) and the energy at maximum load.

D The yield stress (cy) is calculated from the three point bend formula.

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

Percentage shear was determined from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77.

The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests were performed on a 20,000 pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test.

Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure.

The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 j

per ASTM E83-67.

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

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5-2

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

i Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip.

In test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550*F (288'C).

The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures.

Experiments indicated that this method is accurate to plus or minus 2*F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the i

original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

5.2.

CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials 19 contained in Capsule R irradiated to approximately 550*F at 4.42 x 10 2

n/cm are presented in Tables 5-1 through 5-10 and Figures 5-1 through 5-5.

The transition temperature increases and upper shelf energy decreases for the Capsule R material are shown in Table 5-11. Table 5-12 summarizes the Charpy impact test results from Capsule R along with the previous capsules.

Irradiation of vessel lower shell forging 22642 material (axial orientation) 19 specimens to 4.42 x 10 n/cm2 (Figure 5-1) resulted in a 30 and 50 ft-lb transition temperature increase of 85'F and 95'F, respectively, and an upper shelf energy decrease of 10 ft-lb compared to the unirradiated data.

m.io 5-3

Irradiation of vessel lower shell forging 22642 material (tangential orientation) specimens to 4.42 x 1019,fe,2 (Figure 5-2) resulted in a 30 and 50 ft-lb transition temperature increase of 100*F and 115'F respectively.

The irradiated upper shelf energy experienced a decrease of 23 ft-lb compared to the unirradiated data.

19 Weld metal irradiated to 4.42 x 10 n/cm2 (Figure 5-3) resulted in a 30 and 50 ft-lb transition temperature increase of 100*F and 125'F respectively, and an upper shelf energy decrease of 12 ft-lb.

19 Weld HAZ metal irradiated to 4.42 x 10 n/cm2 (Figure 5-4) resulted in a 30 and 50 ft-lb transition temperature increases of 125'F and 115'F, respectively, and an upper shelf energy decrease of 29 ft-lb.

The A533 Grade B Class 1 Correlation Monitor material (HSST Plate 02) irradiated to 4.42 x 10 n/cm2 (Figure 5-5) resulted in both a 30 and 50 19 ft-lb transition temperature increase of 180*F and 190*F respectively, and an upper shelf energy decrease of 46 ft-lb.

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-6 through 5-10 and show an increasing ductile or tougher appearance with increasing test temperature.

Figure 5-11 shows a comparison of the 30 ft-lb transition temperature increases for the various Prairie Island Unit 2 surveillance materials with predicted increases using the methods of N'RC Regulatory Guide 1.99, Revision 1.E43 This comparison shows that the transition temperature increase 19 2

resulting from irradiation to 4.42 x 10 n/cm is less than predicted by the Guide for forging 22642 (axial and tangential orientation). The weld 19 2

metal transition temperature increase resulting from 4.42 x 10 n/cm $3 also less than the Guide prediction.

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5-4 i

5-3.

TENSION TEST RESULTS The results of tension t'ests performed on forging 22642 (axial and tangential 19 2

orientation) and weld metal irradiated to 4.42 x 10 n/cm are shown in Table 5-13 and Figures 5-12, 5-13 and 5-14, respectively.

These results show that irradiation produced a 14-17 Ksi increase in 0.2 percent yield strength for forging 22642 and approximately a 11-17 Ksi increase for the weld metal.

Fractured tension specimens for each of the materials are shown in Figures 5-15, 5-16 and 5-17. A typical stress-strain curve for the tension specimens is shown in Figure 5-18.

5-4.

WEDGE OPENING LOADING TESTING Per the Surveillance Capsule Testing Contract with Northern States Power Company, wedge opening loading (WOL) specimens will not be tested. WOL specimen will be stored at the Hot Cell at the Westinghouse R&D Center, mr. mune.

5-5

TABLE 5-1 CHARPY V-NOTCH INPACT DATA FOR THE PRAIRIE ISLAND UNIT 2 LOWER SHELL FORGING 22642 (AXIAL) 19 IRRADIATED AT 550*F, FLUENCE 4.42 x 10 n/cm2 (E > 1 MeV)

Temperature Impact Energy Lateral Expansion Sample No.

'F (*C) ft-lbs (Joules) mils (mm)

% Shear NT20 25 ( -4) 13.0(17.5) 11.0 (0.28) 7 NT17 75 ( 24) 33.0 ( 44.5) 28.0 (0.71) 13 NT15 75 ( 24) 21.0 ( 28.5) 17.5 (0.44) 8 NT19 100 ( 38) 48.0 ( 65.0) 42.0 (1.07) 27 NT16 100 ( 38) 30.0(40.5) 28.5 (0.72) 21 NT14 125 ( 52) 44.0(59.5) 38.0 (0.97) 31 NT18 150 ( 66) 59.0(80.0) 48.5 (1.23) 43 NT24 200 ( 93) 63.0 ( 85.5) 52.0 (1.32) 59 NT13 250(121) 99.0 (134.0) 73.0 (1.85) 100 NT21 350(177) 98.0 (133.0) 76.0 (1.93) 100

't i

1 m7. ion.'"

5-6

TABLE 5-2

~

CHARPY V-NOTCH IMPACT DATA FOR THE PRAIRIE ISLAND UNIT 2 LOWER SHELL FORGING 22642 (TANGENTIAL) 19 IRRADIATED AT 550*F, FLUENCE 4.42 x 10 n/cm2 (E > 1 NeV)

Temperature Impact Energy Lateral Expansion Sample No.

'F ('C) ft-lbs (Joules) mils (mm)

% Shear NL23

-50(-46) 3.0 ( 4.0) 2.0 (0.05) 0 NL19 25 ( -4) 8.0(11.0) 6.0 (0.15) 3 i

NL17 75 ( 24) 35.0(47.5) 30.0 (0.76) 17 NL14 75 ( 24) 64.0 ( 87.0) 50.5 (1.28) 35 NL18 100(38) 42.0 ( 57.0) 37.0(0.94) 39 NL16 125(52) 53.0 ( 72.0) 42.0 (1.07) 47 NL15 175 ( 79) 87.0(118.0) 62.5 (1.59) 59 NL20 200 ( 93) 102.0 (138.5) 69.5 (1.77) 88 NL22 250(121) 131.0(177.5) 86.0 (2.18) 100 NL21 350(177) 129.0 (175.0) 90.0 (2.29) 100 NL24 400(204) 122.0 (165.5)'

82.0 (2.08) 100 m r.. ' "" "

5-7

TABLE 5--3 CHARPY V-NOTCH IMPACT DATA FOR THE PRAIRIE ISLAND UNIT 2 PRESSURE VESSEL WELD METAL IRRADIATED AT 550*F, FLUENCE 4.42 x 1019,fc,2 (E > 1 MeV)

Temperature Impact Energy Lateral' Expansion Sample No.

• F ('C) ft-lbs (Joules) mils (mm)

% Shear NW16

-10 (-23) 21.0 ( 28.5) 20.5 (0.52) 11 l

NW12 25 ( -4) 33.0 ( 44.5) 29.0 (0.74) 43 NW11 25 ( -4) 29.0 ( 39.5) 29.0 (0.74) 23 j

NW15 75(24) 48.0 ( 65.0) 36.0 (0.91) 52 NW13 150 ( 66) 71.0 ( 96.5) 60.5 (1.54) 90 NW9 200 ( 93) 87.0 (118.0) 73.5 (1.87) 100 NW10 250 (121) 97.0 (131.5) 77.0 (1.96) 100 NW14 350 (177) 89.0 (120.5) 75.5 (1.92) 100 m r.. i u. ""

5-8

TABLE 5-4 CHARPY V-NOTCH INPACT DATA FOR THE PRAIRIE ISLAND UNIT 2 PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL 19 IRRADIATED AT 550*F, FLUENCE 4.42 x 10 n/cm2 (E > 1 MeV)

Temperature Impact Energy Lateral Expansion Sample No.

'F ('C) ft-lbs (Joules) mils (mm)

% Shear NH9

-50(-46) 14.0 ( 19.0) 17.5 (0.44) 18 NH11

-25(-32) 20.0 ( 27.0) 15.0 (0.38) 21 NH15 0(-18) 32.0 ( 43.5) 21.0 (0.53) 39 NH12 25 ( -4) 54.5 ( 74.0) 41.0 (1.04) 82 NH14 75 ( 24) 82.0(111.0) 46.0 (1.17) 88 NH16 125 ( 52) 88.0 (119.5) 64.0 (1.63) 100 NH10 200 ( 93) 89.0(120.5) 57.0 (1.45) 100 NH13 300(149) 88.0(119.5) 69.5 (1.77) 100 i

e 4

mr. io'"""

5-9

.=.

TABLE 5-5 CHARPY V-NOTCH IMPACT DATA FCR THE PRAIRIE ISLAND UNIT 2 PRESSURE VESSEL A533 GRADE B CLASS 1 CORRELATION NONITOR MATERIAL IRRADIATED AT 550*F, FLUENCE 4.42 x 10 n/cm2 (E > 1 MeV) 19 Temperature Impact Energy Lateral Expansion Sample No.

'F ('C) ft-lbs (Joules) mils (mm)

% Shear R15 150 ( 66) 8.0 ( 11.0) 9.5 (0.24) 13

~

R16 225(107) 33.0 ( 44.5) 25.5 (0.65) 41 l

R10 225(107) 27.0 ( 36.5) 29.0 (0.74) 33 R11 250(121) 45.0 ( 61.0) 34.0 (0.86) 56 R9 250(121) 31.0 ( 42.0) 24.5 (0.62) 46 R14 300(149) 62.0 ( 84.0) 56.0 (1.42) 100 R12 350 (177) 87.0(118.0) 70.0 (1.78) 100 R13 450 (232) 82.0(111.0) 54.0 (1.37) 100 mr.*"""

5-10

TABLE 5-6 i

INSTRimENTED CI:ARPY IWACT TEST RESULTS FOR PRAIRIE ISLAND UNIT 2 LOWER SHELL FORGING 22642 (AXIAL ORIENTATION) i Norma 11 red Energies Charpy Maximum Prop Test Charpy Ed/A Em/A Ep/A Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp.

Energy (Ft 1bs/

(Ft Ibs/ (Ft 1bs/

Load to Yield Load Maximum Load Load Stress Stress 2

2 2

No.

(*F)

(Ft Ibs) in )

in )

in )

(K1ps)

(pSec)

(Kips)

(pSec)

(Kips)

(Kips)

(Kst)

(Kst)

NT20 25 13.0 105 78 26 3.35 90 3.80 210 3.80 111 119 NT15 75 21.0 169 147 23 3.35 90 4.05 350 4.00 110 123 7'

NT17 75 33.0 266 238 28 3.40 85 4.60 500 4.50 112 132 C

NT16 100 30.0 242 150 91 3.10 85 4.00 500 4.40 102 117 NT19 100 48.0 387 215 171 3.10 90 4.15 505 3.80

.25 102 119 NT14 125 44.0 354 217 137 3.05 95 4.20 510 4.15

.55 101 121 NT18 150 58.0 475 240 235 3.25 100 4.15 550 4.05

.85 107 122 NT24 200 63.0 507 226 279 3.10 85 4.30 510 3.90 1.45 103 123 NT13 250 99.0 797 267 530 2.90 90 4.10 620 95 116 NT21 350 98.0 789 258 532 2.50 85 3.75 650 83 104 i

.i l

I i

)

2097s:10-861106

TABLE 5-7 INSTRUMEMED CHARPY IIRPACT TEST RESULTS FOR PRAIRIE ISLAND UNIT 2 LOWER SHELL FORGhMG 22642 (TANGENTI AL ORIENTATION)

Norma 11 red Energies Charpy staximon Prop Test Charpy Ed/A Em/A Ep/A Yleid Time ataximJe Time to Fracture Arrest Yleid Flow j

Sample Temp.

Energy (Ft Ibs/

(Ft 1bs/ (Ft Ibs/

Load to Yleid Load Maximum Load Load Stress Stress 2

2 2

l No.

(*F)

(Ft Ibs) in )

in )

In )

(K1ps)

(pSec)

(K1ps)

(pSec)

(Kips)

(K1ps)

(Kst)

(Kst)

N23

-50 3.0 24 to 14 2.40 60 2.40 N19 25 8.0 64 40 25 3.55 100 3.65 130 3.50 118 120 N17 75 35.0 282 220 62 3.40 105 4.30 505 4.25 113 127 N14 75 64.0 515 285 231 3.30 105 4.25 640 4.00 108 125 Ni8 100 42.0 318 283 55 3.20 100 4.25 645 4.00

.20 106 123 N16 125 53.0 427 284 143 3.15 100 4.20 650 4.10 45 104 122 N

N15 175 87.0 701 341 359 3.40 100 4.50 720 3.80 1.45 112 131 N2O 200 102.0 821 300 522 3.20 135 4.25 690 3.55 2.15 106 123 N22 250 131.0 1055 296 759 3.10 100 4.20 675 102 120 N21 350 129.0 1039 294 744 2.75 100 3.85 730 91 109 N24 400 122.0

~982 267 716 2.70 105 3.80 675 90 108 i

4 t

2097S:10-861106

TABLE 5-8 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR PRAIRIE ISLAND UNIT 2 WELD NETAL l

Norma 11 red Energies Charpy maximum Prop Test Charpy Ed/A Em/A Ep/A Vield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp.

Energy (Ft Ibs/

(Ft Ibs/ (Ft Ibs/

Load to Yield Load Maximum Load Load Stress Stress 2

2 2

N6.

(*F)

(Ft 1bs) in )

in )

In )

(Kips)

(pSec)

(Klps)

(USec)

(Klps)

(Klos)

(Kst)

(Kst)

W16

-10 21.0 169 157 12 3.50 100 4.30 360 4.20 115 129 W11 25 29.0 234 213 21 3.35 90 3.95 485 3.95

.20 111 121 7

W12 25 33.0 266 179 87 3.35 95 4.10 420 3.95 1.60 110 123 C

W15 75 48.0 387 242 144 3.35 145 4.30 555 4.20 1.90 til 127 W13 150 71.0 572 213 359 3.10 90 4.05 505 103 118 NW9 200 87.0 701 258 442 3.35 90 4.25 570 til 125 W10 250 97.0 781 249 532 2.80 85 3.85 605 93 110 W14 350 89.0 717 247 469 2.80 90 3.80 610 93 109 i

s 2097s:10-861106 1

TABLE 5-9 INSTRUMENTED CHARPY IWACT TEST RESULTS FOR PRAIRIE ISLAND UNIT 2 WELD HEAT AFFECTED ZONE NETAL Norma 11 red Energies

~

Charpy RRaximum Prop Test Charpy Ed/A Em/A Ep/A YleId Time Itax imum Time to Fracture Arrest Yield Flow Sample Temp.

Energy (Ft Ibs/

(Ft Ibs/ (Ft Ibs/

Load to Yleid Load IIaximum Load Load Stress Stress 2

2 No.

(*F)

(Ft 1bs) in )

in )

in )

(K1ps)

(pSec)

(Kips)

(pSec)

(Kips)

(Klps)

(Kst)

(Kst)

NH9

-50 14.0 113 66 47 4.15 110 4.40 170 4.35 45 138 142 i

NH11

-25 20.0 161 118 43 3.80 95 4.30 270 4.20 126 135 NH15 C

32.0 258 206 52 4.05 90 4.85 405 4.75

.25 134 147 NH12 25 54.5 439 224 215 3.80 105 4.55 475 126 138 I

NH14 75 82.0 660 273 387 3.55 120 4.45 595 3.90 2.60 118 133 NH16 125 88.0 709 278 430 3.35 100 4.30 610 til 126 NH10 200 89.0 717 255 462 3.35 110 4.25 575 III 126 i

NH13 300 88.0 709 248 461 3.30 100 4.30 550 109 126 i

2097s:10-861106 i

TABLE 5-10 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR PRAIRIE ISLAND UNIT 2 CORRELATION MONITOR MATERIAL (HSST PLATE 02) 1 1

Normalized Energies Charpy Maximum Prop Test Charpy Ed/A Em/A Ep/A Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp.

Energy (Ft Ibs/

(Ft 1bs/ (Ft Ibs/

Load to Yield Load Maximum Load Load Stress Stress 2

2 2

No.

(*F)

(Ft 1bs) in )

in )

in )

(Kips)

(pSec)

(Kips)

(pSec)

(Kips)

(Kips)

(Kst)

(Kst)

R15 150 8.0 64 39 25 1.20 20 3.55 130 3.35

.30 40 79 RIO 225 27.0 217 141 76 3.35 100 4.20 335 4,30

.85 110 125 7

R16 225 33.0 266 142 124 3.15 100 4.00 355 3.95 1.40 104 118 l

R9 250 31.0 250 160 90 3.35 90 4.35 360 4.35 2.00 111 128 R11 250 45.0 362 219 144 3.15 90 4.20 500 4.20 2.30 103 122 i

R14 300 62.0 499 195 305 3.20 90 4.15 445 105 122 R12 350 87.0 701 211 489 2.95 90 3.95 505 98 114 R13 450 82.0 660 225 435 2.85 95 3.80 560 94 110

)

2097s:10/0317s-870112 4

--.v--

g j

, ;L z.

')

^

TABLE 5-11' 19 THE EFFECT OF 550*F IRRADIATION AT 4.42 x 10 n/cm2 (E > 1 MeV)

ON THE NOTCH TOUGHNESS PROPERTIES OF THE PRAIRIE ISLAND UNIT 2 REACTOR VESSEL MATERIALS Average Average 35 mit Average Average Energy Absorptton 30 f t -I b Temp (

• F )

l_ateral Expansion Temp ('F) 50 ft-Ib Temp (*F) at Full Shear (ft-lb)

Itaterial Untrradiated Irradiated AT Untrradiated Irradiated AT Untrradiated Irrad1ated AT Untrradtated Irradiated A(ft-Ib) l Forging 22642 0

85 85 25 105 80 35 130 95 108 98 10 (Axial) 9' 5

Forging 22642

-25 75 100

-5 90 95

-5 110 115 150 127 23 (Tangenttal)

Weld leetal

-75 25 100

-45 50 95

-40 85 125 103 91 12 HAZ Metal

-135

-10 125

-80 30 110

-90 25 115 117 88 29 Correlatton 45 225 180 60 260 200 80 270 190 123 77 46 IAoni tor HSST Plate 02 i

2097s:10-861106

TABLE 5-12

SUMMARY

OF PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS 41 Joule 68 Joule 30 ft-lb 50 ft-lb Decrease Trans.. Temp.

Trans. Temp.

Upper Shelf Fluence Increase Increase Energy 19 2

Material Capsule 10 n/cm

(*F)

('F)

(ft-lb)

Forging 22642 V

0.586 30 30 10(a)

(Axial)

T 1.05 35 45 15 R

4.42 85 95 10 Forging 22642 V

0.586 35 35 19[a]

(Tangential)

T 1.05 55 65 17 R

4.42 100 115 23' Weld Metal V

0.586 60 55 3

T 1.05 60 60 11 R,

4.42 100 125 12 I

S "3 HAZ Metal V

0.586 45 45 T

1.05 55 45 18 R

4.42 125 115 29 Correlation Monitor V

0.586 125 120 21 (HSST Plate 02)

T 1.05 160 150 35 R

4.42 3d 190 46

[a]

Upper Shelf Energy Increase nr. ioneno.

5-17

TABLE 5-13 TENSILE PROPERTIES FOR PRAIRIE ISLAND UNIT 2 19 2

REACTOR VESSEL MATERIAL IRRADIATED TO 4.42 x 10 n/cm Test 2% Yleid Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp.

Strength Strength Load Stress Strength Elongation Elongation in Area No, Material

(*F)

(ksi)

(ksi)

(kip)

(kst)

(ksi)

(%)

(%)

(%)

NT6 Forging 22642 125 81.0 96.8 3.35 186.3 68.2 9.8 20.7 63 (Axlal)

NT4 Forging 22642 225 77.9 93.7 3.40 182.5 69.3 9.3 20.4 62 f

(Axlal)

E NTS Forging 22642 550 71.3 94.7 3.70 185.4 75.4 9.0 17.3 59 (Axial)

NL6 Forging 22642 125 80.5 95.7 3.05 212.8 62.1 9.8 22.7 71 (Tangenttal)

NL4 Forging 22642 225 77.9 93.7 2.95 176.5 60.1 9.8 21.7 66 (Tangenttal)

NL5 Forging 22642 550 72.3 93.7 3.30 183.5 67.2 9.5 19.4 63 (Tangenttal)

NW6 WELD Metal 75 83.5 93.7 3.10 216.3 63.2 12.0 22.8 71 NW4 WELD Metal 200 75.9 83.5 2.90 202.3 59.1 9.0 26.5 67 NW5 WELD Metal 550 (Test Malfunction) 20975: 10-861206

('C)

-50 0

50 100 150 200 250 l

I I

I l

I i

100

^

80 E

- 60 e

trj 40 I

O m

20 o

O

_ 100 2.5 e

_ 80 g

2.0 E

O,

7 60 1.5 -

ct e

E b 40

' 8

1. 0.

O g

go.

20 7

e 0.5 g

.J O

O 160 200 140

- 120 - WWADI ATED 160 n

n

[100 120 80 3

to E 60 80 LD E,

o d 40 O

TrraADIATED TO 85* e 4,,zxici9 nfc,2 20 - ((

40 O

I O

O

-100 O

100 200 300 400 500 TEMPERATURE (*F)

Figure 5-1. Irradiated Charpy V-Notch Impact Properties for Prairie Island Unit 2 Reactor Vessel Lower Shell Forging 22642 (Axial Orientation) 007-A-19690-3 5-19

(*C)

-50 0

50 100 150 200 250 1

I l

I i

i I

l00 80 E

- 60 e

tr 40 - g WI#

20 o

^

O 100 2.5 e

9-n 2.0

_ 80 n

Q E

7 60

1. 5 -i k

o e

1. 0 --

W 40 g*

95'F 20 0.5

• g 4

O J

O O

180 240 160 o

UNIRRADIATED o

O n

140 o

g e

- 120 160

100 120 80 O

y U

60

,, g,'7

• go f

IRRADIATED TO 4.42xio19n/cm2 W 40 g,,

40 20 o

I I

I I

O O

-100 O

100 200 300 400 500 TEMPERATURE (*F)

Figure 5-2. Irradiated Charpy V-Notch impact Properties for Prairie Island Unit 2 Reactor Pressure Vessel Lower Shell Forging 22642 (Tangential Orientation) oo7.A 19690-4 5-20

l

(*C) l

-100

-50 0

50 100 150 200 250 l

I I

I I

I I

l00

}I 80 0 4 2 x

8 60 tr o

j 40 5

20 O

100 2.5 80 A

)

a 2.0 E

7 60 O

1.5 k

9 LLI 40 o

95'F_

1.0 20

,,.q-O.5 a

• g<J O

O I60 140 200 UNIRRADIATED 160 o

he

100 o

- l20[

[ eO e

o E 60 O

o 80 40 o

125'F L

IRRADIATED TO IOO*F 4.42XIOl9n/cm2 20 o

I I

I I

O O

-200

-100 O

100 200 300 400 500 TEMPERATURE (*F)

Figure 5-3. Irradiated Charpy V-Notch Impact Properties for Prairie Island Unit 2 Reactor Pressure Vessel Weld Metal OO7-A 19690-5 5-21

(*C)

-100

-s0 0

50 100 150 200 250 l

i l

1 1

I I

I 3

3 3

t 100 0

0 80 O

E

- 60 0-o o

4 40 o

2 23 p 20 2

0

_ 100 2.s e

80 k

2.0

_a E

_a a

60 e

I.s-Q.

E E

XW 40

-o l l O'F_8

1. 0 -

o l

- g 20 O.s o

1 J

O O

160 140 UNIRRADIATED 120 o

R g

,gg 2

o I00 o

+

^

?

?

l 2a,

80 n

y c.9

$ 60 O

o 80 Cf i15'F IRRADIATED TO w 40 - O /12s'F~

O 4.42X1019 n/cm2 l

gg _[V I

I I

I I

I O

O

-200

-100 O

100 200 300 400 500 TEMPERATURE ('F)

Figure 5-4. Irradiated Charpy V Notch Impact Properties for Prairie Island Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Material i

l 007-A 19t,90-6 5-22 i

e

,_m..

-n

(*C)

-50 0

50 100 150 200 250 l

I I

I I

I i

100 2

80 b

O 60 tr$ 40 2

b 20 O

_ 100 2.5 e

n n

_ 80 o

2.0 E

o e

7 60 0

1.5 -

k

/

1.0 -

o i

uJ 40 2OO'F 20 0.5 J

O O

160 200 140 UNIRRADIATED

- 120 h

ISO 0

100 G

120 80 O

o e

y 60 80

,gg,p

~

O TOO*F IRRADIATED To-40 4.42XlOI9 n/cm2 I

I I

I O

O

-100 O

100 200 300 400 500 TEMPERATURE ('F)

Figure 5-5. Irradiated Charpy V-Notch Impact Properties for Prairie Island Unit 2 Reactor Pressure Vessel A533 Grade B Class 1 Correlation Monitor Material 007.A-19690-7 5-23

(.

kk 1

- Q; ;

49 %

4 M

. rj

'9~

3,,,

3r W.

~

2 :-c.

/

2 p.y.

l.f^

d.

l

}

4

~

A 1

NT20 NT17 NT15 NT19 s

fl '

{

k

.s NT16 NT14 NT18 NT24 1

't sl

.e.,

\\

j w:

l.?l NT13 NT21 Figure 5-6. Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 2 Reactor Pressure Vessel Lower Shell Forging 22642 (Axial Orientation) 5-24

.:. m

%n,;pJ -

ygtn d',.. t l[ ~g,j%

if[q'f.

kk4 U,$

. $},;

G I i

{

i u

' fif$'

D'Gj :

NL23 NL19 NL17 NL14

, g;-

'y: M i

p,tyJg4.

w -Jg. : -

• - ?./

'r N(f

'N I

[$

4 0

..a_

V O.

A 6 -

.q a

.a NL18 NL16 NL15 NL20

.f ff' P

Lj;g

.3 A

NL22 NL21 NL24 Figure 5-7. Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 2 Reactor Pressure Vessel Lower Shell Forging 22642 (Tangential Orientation) 5-25

-qf o@-:

?

k

, $y NE

n'

.. o y ;-

lQ

.N!

...t i

i (5)V N

NW16 NW12 NW11 NW15

}:

.hI

,,~f e,. ;

+

+ t 2:..

~~

f

.,i v( 1 b'Q

-g.

m NW13 NW9 NW10 NW14 Figure 5-8. Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 2 Reactor Pressure Vessel Weld Metal I

5-26 J

l j.

NH9 NH11 NH15 NH12 B B1E NH14 NH16 NH10 NH13 l

Figure 5-9. Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 2 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-27

dW E'g.'q,,~-

. ~;q ;j

- ::l%'yl

'no 25 R

, R:g;,c, M sg

E d; v

_--- y s;

g.

14 -4 N~

': p --

....,1

';7 gi,.p-q; -

f. * 'i

m[1 ffk) '

e

.gu n g j

R15 R16 R10 R11 R9 R14 R12 R13 Figure 510. Charpy impact Specimen Fracture Surfaces for Prairie Island Unit 2 Reactor Pressure Vessel A533 Grade B Class 1 Correlation Monitor Material 5-28

{ 50 40 g

30 CORRELATION MONITOR 20 FORGING 22642 O

z s

1 WELD METAL

@ 100 H

80 Z

O 60 A

H 5

40 z

<[

~

LEGEND:

O FORGING 22642 (AXIAL)

O FORGING 22642 (TANG )

CD 20 _.

.J A WELD METAL V HAZ METAL

[

O CORRELATION MONITOR O

10 I

I I

I IIIII I

I I

I I I II 10 18 2

4 6 8 IO I9 2

4 6 8 1020 FLUENCE (n/cm 2)

Figure 511. Comparison of Actual Versus Predicted 30 ft lb Transition Temperature increases for the Prairie Island Unit 2 Reactor Vessel Material Based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 007 A 19690 8 5 29

(*C)

O 50 LOO 150 200 250 300 110 1

I I

I I

I I

100 700 W

3

_s 3

90 y

ULTIMATE TENSILE STRENGTH - 600 SO 4

0 70 500 g 0.2% YIELD STRENGTH v

y 60 g-400 50 300 40 80 70 REDUCTION IN AREA g

3 60 0

_1 m

s 50 CODE:

H OPEN POINTS-UNIRRADIATED 40 CLOSED POINTS-IRRADIATED TO H

4.42XIOI 9 n/cm 2 F-30 2

/

TOTAL ELONGATION a

o 20 1

^

~

n 0

lo

=

=

=

UNIFORM ELONGATION I

I I

I I

I O

O LOO 200 300 400 500 600 i

TEMPERATURE (*F)

Figure 512. Tensile Properties for Prairie Island Unit 2 Reactor Vessel Lower Shell Forging 22642 (Axial) 1 007.A 19690-9 5 30

.~

(*C)

O 50 100 150 200 250 300 I10 l

I i

i i

i l

100 700 i

(

i

^

l

~

90 j

2 ULTIMATE TENSILE STRENGTH - 600 80 A

~

a n

70 500 g 0.2% YIELD STRENGTH m

60 4

g-400 50 300 40 80 70 C

a 8

e o

,x 60 REDUCTION IN AREA v

50 CODE 3p OPEN POINTS-UNIRRADIATED 40 CLOSED POINTS-IRRADIATED TO

~

4.42XIOl9 n/cm 2

_j H

H 30

- (

O A

TOTAL ELONGATION

)

5 20 i

4 G

n 8

10 UNIFORM ELONGATION I

I I

I I

I O

O I00 200 300 400 500 600 TEMPERATURE (*F) i Figure 513. Tensile Properties for Prairie Island Unit 2 Reactor Vessel Lower Shell Forging 22642 (Tangential) i co7 A.19690-to 5 31

('C)

O 50 100 150 200 250 300 110 100 700 90 Y

ULTIMATE TENSILE STRENGTH - 600 80 4

^

0 70 n

500 g LLI m

I 60 0.2% YIELD STRENGTH

~

50 00 40 I

80 70 2

60 REDUCTION IN AREA CODE:

50 OPEN POINTS-UNIRRADIATED f--

CLOSED POINTS-IRRADIATED TO t

40 4.42xiOl 9 n/cm2 H

_J

~

F-30 U

A TOTAL ELONGATION 5

20 A

)

O n

g iO UNIFORM ELONGATION I

I I

I I

I O

l 0

100 200 300 400 500 600 TEMPERATURE (*F)

Figure 514. Tensile Properties for Prairie Island Unit 2 Reactor Vessel Weld l

Metal

{

i I

007 A 1969011 5 32

em.

$'~ ( g :.3, 4-'

.f t '

' 3 lofMS ; '

.14

.1 - 2

3. : 4

'I 100I 2.

g '6'!?'82 TENSILE SPECIMEN NT6 TESTED AT 1250F

~

E/i _.._. _

..12..,21.__

1

i ;, ; i +

• ~ 7 i

0 I

i

,t

.1' t

2 g'

s lt

,3,i

, /.1,1 3

..c h.

TENSlLE SPECIMEN NT4 T

TESTED AT 2250F 1

y..

.,w~r,

w,.,

.h l l l l l ;l l ~~ l ' l

.'l t t' I 2 3 4 g #1.7 8 TJ 1

10iH5

't,

. 10016 t5

^E 4

.g

. G lI 5

' L' i aa 4 c7 e,5 2

' h INDlUN TENSILE SPECIMEN NT5 pdi TESTED AT 5500F l L. u.

3 t

Figure 515. Fractured Tensile Specimens of the Prairie island Unit 2 Reactor Vessel Lower Shell Forging 22642 (Axial Orientation) 5-33

agamummmmmmumumumusume t l l sl:. 7 l l

t E3 4 8 <i -

.1 ionn j.*'

19C Ill's 1 2 314 6 7 8J3 s

.W.!ier%)3r%!abbh.

'y t

t

^

TENSILE SPECIMEN NL6 TESTED AT 1250F F

e O.,

g e on

- amammu

!,I i i

l..

t I

1 6 7 8 9-1

) O ? H.,

g *-

l (6 } } {",

s.

3 ;- 4 6 7 M 9.

- '! f P.[1)? #.df.. ' b 1',; ! d F

TENSILE SPECIMEN NL4 TESTED AT 2250F

^

h h',

w. mm -.

i f

l6 l

[

.1 2.3=4 7 89 1046 c.

.. g

-1 lajil6 5

j :j i 2 3'4 G '7 3-9 1

4hh!

TENSILE SPECIMEN NL5 TESTED AT 5500F n;N

..,..~

l Figure 516. Fractured Tensile Specimens of the Prairie Island Unit 2 Reactor Vessel Lower Shell Forging 22642 (Tangential Orientation) 5 34 I - - - -. -

l l

l l

' 5 Q

I

. 6.7. 9; 1 2 3 4

' 7

( q,,. 10fH; a

5 1

3 4 6.7 8 bh!WGS!d TENSlLE SPECIMEN NW6 TESTED AT 750F e.hv hi..dlN A

a.

p- -

sn h o.,

. U 7 M 9 1

p x

1.

2 3 4 ld;hr, '

I' G

,1

..)

4

'I 8

')

1.

e'4b!M.%tbW hhhak%

TENSlLE SPECIMEN NW4 TESTED AT 2000F

[jty+4eww;mymppid.ep.ugd**"

?)

r 1 2 3,. 4 5

7 10iHS g'..

E.

(-

100ilr, 5'

3:

1 3 3 4 6 7 0 < ').

Ldd TENSILE SPECIMEN NWS TESTED AT 5500F s

?.i n l

l l

l Figure 5-17. Fractured Tensile Specimens of the Prairie Island Unit 2 Reactor l

Vessel Weld Metal 5-35

8y l

?

G.

c 100 90 -

1

]

80 i

70 l

m r

j

$ 60 i

m 50 i

m

$ 40 NL5 (550*F)

SPECIMEN l

cn p

m 30 i

i 20 10 I

I I

I I

I I

I

)

O l

0

.03

.06

.09

.12

.15

.18

.21

.24

.27 STRAIN (IN/IN) l Figure 5-18. Typical Stress - Strain Curve for Tension Specimens i,

~

SECTION 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1.

INTRODUCTION Knowledge of the neutron environment within the reactor pressure vessel / surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known.

Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens.

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 is derived solely from analysis.

This section describes a discrete ordinates S transport analysis performed n

for the Prairie Island Unit 2 reactor to determine the fast (E > 1.0 MeV) neutron flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules. The analytical data were then used to develop lead factors for use in relating neutron exposure of the reactor vessel to that of the surveillance capsules. Based on the use of spectrum-averaged reaction cross sections derived from this calculation and the Prairie Island Unit 2 power history, the analysis of the neutron dosimetry contained in Capsule R is presented.

6-2.

DISCRETE ORDINATES ANALYSIS A plan view of the Prairie Island Unit 2 reactor geometry at the core midplane is shown in figure 6-1.

Since the reactor exhibits 1/8th core symmetry, only w,,oes,s.,s ran s g.1

l

~

a zero-to 45-degree sector is depicted. Six irradiation capsules attached to the neutron pad are included in the reactor design to constitute the reactor i

vessel surveillance program. The capsules are located at 13, 23, and 33 degrees from the cardinal axes as shown in figure 6-1.

l A plan view of a single surveillance capsule holder attached to the thermal i

shield is shown in figure 6-2.

The stainless steel specimen container is l

approximately 1-inch square and approximately 63 inches in height. The l

containers are positioned axially such that the specimens are centered on the

[

core midplane, thus spanning the central 5.25 feet of the 12-foot-high reactor core.

From a neutron transport standpoint, the surveillance capsule structures are l

significant. They have a marked effect on both the distribution of neutron

)

flux and the neutron energy spectrum in the water annulus between the thermal shield and,the reactor vessel.

In order to properly determine the neutron j

environment at the test specimen locations, the capsules themselves must be

[

included in the analytical model. This requires at least a two-dimensional calculation.

i In the analysis of the neutron environment within the Prairie Island Unit 2

)

reactor geometry, predictions of neutron flux distributions and energy spectra j

were made with the 00T(5) two-dimensional discrete ordinates transport

)

code. The radial and azimuthal distributions were obtained from an R,8 calculation wherein the geometry shown in figures 6-1 and 6-2 was represented j

in the analytical model.

In addition to the R,8 calculation, a second

[

calculation in R,2 geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest.

In the R,Z l

analysis, the reactor core was treated as an equivalent volume cylinder.

The surveillance capsules were not included in the R,Z model.

j Both the R,8 and R,2 analyses employed 47 neutron energy groups and a P 3

expansion of the scattering cross sections. The cross sections used in the 4

f analyses were obtained frc:n the SAILOR cross section libraryI63 which was l

developed specifically for light water reactor applications.

The neutron l

energy g oup structure used in the analysis is listed in table 6-1.

mer. i 4.""

6-2

i I

l l

)-

A key input parameter in the analysis of the integrated neutron exposure of the reactor vessel is the core power distribution. For this analysis, core i

power distributions representative of time-averaged conditions derived from statistical studies of long-term operation of Westinghouse 2-loop plants were employed. These input distributions include rod-by-rod spatial variations for l

all peripheral fuel assemblies.

[

I i

This generic, design basis, core power distribution is intended to provide a j

vehicle for the long-term (end of-life) projection of reactor vessel

]

exposure. Since plant-specific core power distributions reflect only past

=

l operation, their use for projection into the future may not be justified. The l

l use of generic data which reflects long-term operation of similar reactor i

cores may provide a more suitable approach.

1 l

1 Benchmark testing of these generic core power distributions and the SAILOR l

cross sections against surveillance capsule data obtained from two, three,

and four-loop Westinghouse plants indicate that this analytical approach i

yields conservative results, with calculations exceeding measurements from 10 to 25 percent. W i

One further point of interest regarding these analyses is that the design l

}

hasis assumes an out-in fuel loading pattern (fresh fuel on the periphery).

j Future commitment to low-leakage core loading patterns could significantly reduce the calculated neutton flux levels presented in section 6-4.

In addition, surveillance capsule lead factors could be changed, thereby l

influencing the withdrawal schedule of the remaining surveillance capsules.

l l

Having the results of the R,8 and R,Z calculations, three-dimensional variations of neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor.

e(R,2,8,E ) = e(R,8,E ) x F(Z,E )

(6-1)

{

g g

g l

where i

s mr. ie""

6-3

9(R,Z,0,E ) = neutron flux at point R,2,0 within energy group g g

1. (R,0,E )

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

g obtained from the R,0 calculation F(Z,E )

= relative axial distribution of neutron flux within energy g

group g obtained from the R,Z calculation This analysis is consistent with established ASTM standards.[8,9,10,11,12]

6-3.

RADIOMETRIC MONITORS The passive radiometric monitors included in the Prairie Island Unit 2 surveillance program are listed in table 6-2.

The first five reactions in table 6-2 are used as fast neutron monitors to relate fast (E > 1.0 MeV) neutron fluence to m~easured material property changes.

In order.to address the potential for burnout of the product nuclides generated by fast neutron reactions, it is necessary to also determine the magnitude of the thermal and resonance region neutron fluxes at the monitor location.

Therefore, bare and cadmium-shielded cobalt-aluminum monitdrs are normally included. However, the cobalt aluminum cadmium shielded monitors were inadverten'tly left out of the Prairie Island Unit 2 Capsule R.

The relative locations of the various radiometric monitors within the surveillance capsule 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 accommodated within the dosimeter block located near the axial center of the capsule.

The use of passive monitors such as those listed in table 6-2 does not yield a direct measure of the energy-dependent neutron 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 material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors sw, seiom,,sions s.4

I i

j may be derived from the activation measurements only if the irradiation i

parameters are well known.

In particular, the following variables are important.

i o

The operating history of the reactor l

o The energy response of the monitor l

o The neutron energy spectrum at the monitor location o

The physical characteristics of the monitor i'

l i

The analysis of the passive monitors and the subsequent derivation of the j

average neutron flux requires two operations.

First, the disintegration rate of product nuclide per unit mass of monitor must be determined. Second, in j

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.[13,14,15,16,17,18,19,20,21] Following sample preparation, the activity of each monitor is determined by means of a lithium-drifted germanium, Ge(Li), gamma ray spectrometer. The overall standard deviation of l

the measured data is a function of the precision of sample weighing, the j

uncertainty in counting, and the acceptable error in detector calibration.

j For the samples removed from Prairie Island Unit 2, the overall 2e deviation

{

in the measured data is determined to be plus or minus 10 percent.

The j

neutron energy spectrum at the monitor location is determined analytically l

using the method described in paragraph 6-2.

l I

r l

Having the measured activity of the monitors and the neutron energy spectrum at the monitor locations of interest, the calculation of the neutron flux l

proceeds as follows. The reaction product activity in the monitor is l

expressed as n

A=N,FYfa(s)#(E)dE [

max

-i t ),-A td (6-2)

(1-e j

E 3,1 l

mr.iom m. aron:

6-5

~

where A

= induced product activity (dps per gram)

N,

= number of target element atoms per gram F

= weight fraction of the target nuclide in the target material Y

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

= energy dependent reaction cross section

((E)

= energy dependent neutron flux at the monitor location with the reactor at full (reference) power P

= average core power level during irradiation period j 3

P,,,

= maximum or reference core power level 1

= decay constant of the product nuclide t

= length of irradiation period j j

t

= decay time following irradiation period j d

n

= total number of irradiation periods i

Because the neutron flux distributions are calculated using multigroup transport methods and, further, because the main interest is in the fast (E > 1.0 MeV) neutron flux, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the

+

following relation.

o(E) ((E) dE = 5 ef where 47 o(E) e(E) dE

'g 'g ha o

Q=1 18

1. (E)dE F

'g Z

1 MeV g=1 mr.in ""

6-6

18 f =J1 MeV((E)dE=[e 9

g=1 g = group number from Table 6-1 4

Thus, equation (6-2) is rewritten t

n p

A = N, F Y 3 of [ h (1-e-it ),-itd 3

max j.1 or, solving for the fast (E > 1.0 MeV) neutron flux, 4

A

'f '

n p

f Y 3 [ [ max (1-e

-it ),-itd (6-3)

N j

o j.1 The total fast (E > 1.0 MeV) neutron fluence is then given by n

'f * 'f t

(6-4) j j=1 max ner. ie*""

6-7

s.

,0

?

where i

n[1pht total effective full power seconds of reactor j = operation up to the time of capsule removal max 3,7 An assessment of the thermal neutron flux levels within the surveillance r

capsules is obtained from the bare and cadmium-covered CoS9 (n,r) Co60 data by means of cadmium ratios and the use of a 37-barn, 2,200 m/sec cross section. Thus, D-1 1

R ITI bare (6_5)

'Th

• N F n

P o

Yo

-At ),-itd (1-e j

j a=

max where D is defined as Rbare/ rcd covered

• An assessment of the potential for product nuclide burnout has determined that such an effect is neg.ligible for Prairie Island Unit 2 Surveillance Capsule R.

f 6-4.

NEUTRON TRANSPORT ANALYSIS RESULTS Results of the discrete ordinates transport calculations for the Prairie Island Unit 2 reactor are summarized in this section.

In figure 6-3, the calculated maximum fast (E > 1.0 MeV) neutron flux levels at the radius of the surveillance capsule center, the reactor vessel inner radius, the reactor vessel 1/4 thickness location, and the reactor vessel 3/4 thickness location are presented as a function of azimuthal angle. The local influence of the surveillance capsules on the fast neutron flux distribution is clearly l

evident.

In figure'6-4, the radial distribution of maximum fast (E > 1.0 MeV) r.eutron flux through the thickness of the reactor vessel is shown. The

. relative axial variation of fast neutron flux within the reactor vessel is j

given in figure 6-5.

Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6-3 or 6-4 by the appropriate values from figure 6-5.

Table 6-3 provides the calculated fast neutron exposure parameters for the Prairie Island Unit 2 reactor vessel.

w. iu no.

6-8 i

G Table 6-4 provides the calculated fast neutron exposure parameters and updated l'ead factors for all of the Prairie Island Unit 2 surveillancs capsules. The lead factor is defined as the ratio of the fast (E > 1.0 kV) neutron flux at the dosimeter block location (capsule center) to the maximum fast neutron flux at the reactor vessel inner radius.

In order to derive neutron flux and fluence levels ~from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The calculated neutron energy spectrum at the center of the Prairie Island Unit 2 surveillance Capsule R is listed in table 6-5.

The calculated spectrum-averaged cross sections for each of the fast neutron reactions are given in table 6-6.

6-5.

DOSIMETRY RESULTS The irradiation history of the Prairie Island Unit 2 reactor up to the time of removal of Capsule R is listed in table 6-7.

Comparisons of measured and calculated saturated activity of the radiometric monitors contained in Capsule R based on the irradiation history shown in table 6-7 are listed in table 6-8.

The fast (E > 1.0 MeV) neutron flux and fluence levels derived for Capsule R using the spectrum averaged cross-sections listed in table 6-6 are presented in table 6-9.

Table 6-10 summarizes the thermal neutron flux obtained from the cobalt-aluminum monitors and extrapolations made from Prairie Island Unit 1 observed bare to cd. covered cobalt ratios.

An examination of table 6-9 shows that the average fast (E > 1.0 MeV) neutron flux derived from the five threshold reactions ranges from 1.42 x 11 2

11 2

10 n/cm -sec to 2.09 x 10 n/cm -sec, a total span of 57 percent.

11 2

The calculated flux value of 1.51 x 10 n/cm -see is below the average of the measured values by approximately 5 percent, with calculation to experimental ratios ranging from 0.72 to 1.06.

I a >. ion.""

6-9

A summary of measured and calculated current fast neutron exposures for Capsule R and for key reactor vessel locations is presented in table 6-11.

The measured value is given based on the average of all five threshold reactions listed in table 6-9.

End-of-life (EOL) reactor vessel fast neutron fluence projections are also included in table 6-11. The calculated EOL peak fast fluence at the reactor vessel inner radius is approximately 26% greater 22 than that reported in WCAP-9877 This difference results from the present improved transport methodology, which employs a P3 expansion of the scattering crosssections, instead of the P1 expansions used for WCAP-9877.

This represents a more accurate treatment of the neutron transport.

Based on the data given in table 6-9, the best estimate fast neutron exposure of Capsule R is i = 4.42 x 10 n/cm2 (E > 1 MeV) at 8.81 EFPY.

19 l

l i

l i

20078 10/061106 6-10

TABLE 6-1 SAILOR 47 NEUTRON ENERGY GROUP STRUCTURE Group Group Energy Lower Energy Energy Lower Energy Group (MeV)

Group (MeV) 1 14.19(*)

25 0.183 2

12.21 26 0.111 3

10.00 27 0.0674 4

8.61 28 0.0409 5

7.41 29 0.0318 6

6.07 30 0.0261 7

4.97 31 0.0242 8

3.98 32 0.0219 9

3.01 33 0.0150

-3 10 2.73 34 7.10x10

-3 11 2.47 35 3.36x10

-3 12 2.37 36 1.59x10

-4 13 2.35 37 4.54x10

-4 14 2.23 38 2.14x10

-4 15 1.92 39 1.01x10

-5 16 1.65 40 3.73x10

-5 17 1.35 41 1.07x10

-6 18 1.00 42 5.04x10

-6 19 0.821 43 1.86x10

-7 20 0.743 44 8.7fx10

-7 21 0.608 45 4.14x10

-7 22 0.498 46 1.00x10 23 0.369 47 0.00 24 0.298 a) The upper energy of group 1 is 17.33 MeV.

a.7. in.""

6-11

TABLE 6-2 l

NUCLEAR CONSTANTS FOR RADIOMETRIC MONITORS CONTAINED IN THE PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULES Reaction Target Fission of Weight Product Yield Monitor Material Interest Fraction Half-life

(%)

Iron wire Fe54 (n,p) Mn54 0.0585 314 dy Nickel wire NiS8(n.p)CoS8 0.6777 71.4 dy Copper wire CuS3(n,a)Co60 0.6917 5.27 yr Uranium-238(a) in U 0 U238 (n,f) Cs137 1.0 30.2 yr 6.0 38 Neptunium-237(a) in Np0 Np237 (n,f) Cs137 1.0 30.2 yr 6.5 2

Cobalt-aluminum (*) wire CoS9(n,r)Co60 0.0015 5.27 yr Cobalt-aluminum wire CoS9(n,r)Co60 0.0015 5.27 yr i

i a) Denotes that the monitor is cadmium-shielded but was inadvertently not included in the capsule.

i 1

\\

i I

mr icain.< 7on:

6-12

TABLE 6-3 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS FOR THE PEAK LOCATION OF THE PRAIRIE ISLAND UNIT 2 REACTOR VESSEL Iron Radial Location Fast Neutron Flux Displacement 2

Within the (n/cm sec)

Rate Reactor Vessel (E > 1.0 MeV)

(E > 0.1 MeV)

(dpa/sec) 10 11

-11 Inner Surface 5.15 x 10 1.35 x 10 8.50 x 10 (R = 86.500 inches) 10 11

-11 1/4 Thickness 3.37 x 10 1.21 x 10 6.13 x 10 (R = 88.657 inches) 10 10

-11 3/4 Thickness 1.00 x 10 6.21 x 10 2.51 x 10 (R = 92.970 inches) 9 10

-11 Outer Surface 4.61 x 10 3.35 x 10 1.29 x 10 (R = 95.126 inches) 4 mr. ion.no.

6-13'

TABLE 6-4 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AND LEAD FACTORS FOR THE PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULES Iron Azimuthal Fast Neutron Flux Displacement Capsule Location (a)

(n/cm -sec)

Rate Lead 2

I.D.

[ Degrees)

(E > 1.0 MeV)

(E > 0.1 MeV)

(dpa/sec)

Factor (b) 11 11

-10 V

13*

1.51 x 10 5.85 x 10 2.81 x 10 2.93 11 11

-10 R

13*

1.51 x 10 5.85 x 10 2.81 x 10 2.93 10 11

-10 T

'23*

9.17 x 10 3.22 x 10 1.62 x 10 1.78 10 11

-10 P

23*

9.17 x 10 3.22 x 10 1.62 x 10 1.78 5

33' 8.33 x 10 2.97 x 10 1.49 x 10 1.62 10 11

-10 10 11

-10 N

33' 8.33 x 10 2.97 x 10 1.49 x 10 1.62 l

i a) The radius of the surveillance center is 62.342 inches.

b) The lead factor is the ratio of the fast (E > 1.0 MeV) neutron flux at the center of the surveillance capsule to that at the peak location on the reactor vessel inner surface.

l nn.. i e'"""

6-14 a

w

TABLE 6-5 CALCULATED NEUTRON ENERGY SPECTRUM AT THE CENTER OF PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R Energy Neutron Flux Energy Neutron Flux 2

2 Group (n/cm -sec)

Group (n/cm -sec) 7 10 1

2.63 x 10 25 7.20 x 10 7

10 2

9.59 x 10 26 7.03 x 10 8

10 3

3.42 x 10 27 5.61 x 10 9

10 4

6.30 x 10 28 4.22 x 10 9

10 5

1.07 x 10 29 1.41 x 10 9

9 6

2.41 x 10 30 7.98 x 10 9

10 7.

3.42 x 10 31 1.63 x 10 9

10 8

7.14 x 10 32 1.03 x 10 9

10 9

6.65 x 10 33 2.52 x 10 9

10 10 5.44 x 10 34 3.75 x 10 9

10 11 6.46 x 10 35 5.06 x 10 9

10 12 3.21 x 10 36 4.48 x 10 8

10 13 9.84 x 10 37 6.89 x 10 9

10 14 4.86 x 10 38 3.95 x 10 10 10 15 1.26 x 10 39 4.17 x 10 10 10 16 1.70 x 10 40 5.60 x 10 10 10 17 2.56 x 10 41 6.82 x 10 10 10 18 5.34 x 10 42 3.92 x 10 10 10 19 4.06 x 10 43 4.80 x 10 10 10 20 1.94 x 10 44 3.22 x 10 10 10 21 6.41 x 10 45 2.79 x 10 10 10 22 5.07 x 10 46 6.12 x 10 10 10 23 5.97 x 10 47 1.52 x 10 10 24 5.70 x 10 6-15

TABLE 6-6 SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R Spectrum-Averaged Cross Section *)

I Reaction of Interest (barns)

Fe54 (n,p) Mn54 0.0630 NiS8 (n,p) CoS8 0.0867 CuS3 (n,a) Co60 0.000524 U238 (n,f) Cs137 0.3315 Np237 (n,f) Cs137 3.051 o(E) #(E) dE a) o=

fIMeV o(E)dE

~

i mu. ion '"

6-16

i TA8LE 6-7 IRRADIATION HISTORY OF PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R Time P (c)

P,,,

P /P Irradiation Time Decay Ti j

j max Cycle Period NWt NWt Days Days a

1 1/1/75 -

1447.9 1650 0.878 508.6 3554 10/22/76 2

12/19/76 -

1578.2 1650 0.957 315.5 3169 11/11/77 r

3 12/15/77 -

1562.2 1650 0.947 324.2 2788 11/27/78 4

12/15/78 -

1561.7 1650 0.947 360.1 2386 1/3/80 5

2/19/80 -

1584.5 1650 0.960 338.0 1966 2/81 6

3/81 -

1593.3 1650 0.966 361.7 1496 6/13/82 7

7/17/82 -

1572.1 1650 0.953 370.5 1074 1

8/29/83 8

9/28/83 -

1576.1 1650 0.955 327.5 713 9/4/84 9(b) 10/13/84 -

1560.7 1650 0.946 310.1 312 1

9/6/85 (a)

Measured to capsule counting date 7/22/86 (b) 10 year outage cagsule removed during this outage, total irradiation time is 2.78 x 10 effective full power seconds (EFPS) or 8.81 effective full power years (EFPY)

(c)

P is the average core power level during.the irradiation period 3

mr. '"""

6-17

TABLE 6-8 COMPARISDN OF MEASURED AND CALCULATED RADIOMETRIC l

MONITOR SATURATED ACTIVITIES FOR PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R Radiometric Monitor Saturated Activity Monitor (Disintegrations /Second-Gram)*

Axial o ation(a)

MeasuredId)

Calculated C/E 54 54 Fe (n,P) Mn 6

Top 5.85 x 10 6

Top - Middle 5.38 x 10 6

Middle 5.79 x 10 6

Bottom - Middle 5.59 x 10 6

Bottom 6.31 x 10 0

Bottom 6.18 x 10 6

6 Average 5.85 x 10 6.22 x 10 1.06 58 58 Ni (n,P) Co 7

7 Middle 9.42 x 10 9.23 x 10 0.98 63 60 Cu (n,a) Co 5

Top - Middle 4.84 x 10 5

Bottom - Middle 5.20 x 10 5

5 Average 5.02 x 10 5.25 x.10 1.04

• Basis is one gram of wire.

m7.unno.

6-13

s TABLE 6-8 (Cont)

COMPARISON OF MEASURED AND CALCULATED RADIOMETRIC MONITOR SATURATED ACTIVITIES FOR PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R Radiometric Monitor Saturated Activity Monitor (Disintegrations /Second-Gram)*

Axial ation(a)

Measured (d)

Calculated C/E I

238 137 U

(n,f) Cs (b) 7 6

Middle 1.24 x 10 7.62 x 10 0.72 6

Corrected (c) 1.06 x 10 237 137 Np (n,f) Cs (b) 7 7

Middle 7.37 x 10 7.63 x 10 1.04 59 60 Co (n,r) Co (b)(e) 7 Top 4.09 x 10 7

Bottom 4.14 x 10 7

7 Average 4.12 x 10 6.35 x 10 1.44 59 60 Co (n,r) Co 8

Top 1.03 x 10 8

Bottom 1.12 x 10 8

8 Average 1.07 x 10 1.58 x 10 1.47 rm.m.""

6-19

TABLE 6-8 (Cont) r COMPARISON OF MEASURED AND CALCULATED RADIOMETRIC MONITOR SATURATED ACTIVITIES FOR PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R a) Refer to Figure 4-2 for the locations of the various radiometric monitors, b) This radiometric monitor was cadmium shielded, c) The measured value has been multiplied by 0.85 to correct for the effect 235 239 of 305 ppm U and the build-in of Pu d) Values' geometrically corrected to surveillance capsule center.

e) Data extrapolated from Praire Island #1 data reported in WCAP 11006, Pg 6-24.

Praire Island #2 Capsule R did not contain Cadmium shielded Al-Co Wires.

som mom..mn 6-20 i

~

TABLE 6-9 RESULTS OF FAST NEUTRON DOSIMETRY FOR PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R Current (b)

Radiometric Monitor Fast (E > 1.0 MeV)

Fast (E > 1.0 MeV)

Saturated Activity (a)

Neutron Flux Neutron Fluence 2

2 Reaction (dps/gm)

(n/cm -sec)

(n/cm )

of Interest Measured Calculated Measured Calculated Measured Calculated Fe54 (n.p) Mn54 6

6 11 19 5.85x10 6.22x10 1.42x10 3.94x10 NiS8 (n,p) CoS8 7

7 11 19 9.42x10 9.23x10 1.54x10 4.28x10 Cu63 (n,a) Co60 5

5 11 19 5.02x10 5.25x10 1.45x10 4.03x10 U238 (n f) Cs137 6

6 11 19 1.06x10 7.62x10 2.09x10 5.81x10 Np237 (n,f) Cs137 7

7 11 19 7.37x10 7.63x10 1.45x10 4.03x10 11 11 19 19 Average 1.59x10 1.51x10 4.42x10 4.20x10 a) Refer to Table 6-8.

8 b) Total irradiation time for surveillance Capsule R is 2.78 x 10 effective full power seconds (EFPS).

m.m no.

6-21

TABLE 6-10 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR PRAIRIE ISLAND UNIT 2 SURVEILLANCE CAPSULE R ")

I Saturated Activity ")

I Axial (dps/gm)

'th Location Bare Cd-covered (b)

{3fe,2 3,c) 8 7

11 Top 1.03 x 10 4.09 x 10 1.70 x 10 0

7 11 Bottom 1.12 x 10 4.14 x 10 1.97 x 10 0

7 11 Average 1.09 x 10 4.18 x 10 1.87 x 10 (a)

Values geometrically corrected to surveillance capsule center (b)

Data extrapolated from Prairie Island #1 data reported in WCAP-11006, Pg. 6-24.

Prairie Island #2 capsule R did not contain cadmium shielded Al-Co wires.

m7. ion ""

6-22

TABLE 6-11

SUMMARY

OF PRAIRIE ISLAND UNIT 2 FAST NEUTRON FLUENCE RESULTS BASED UPON SURVEILLANCE CAPSULE R End of Life Current Fast (E > 1.0 MeV)

Fast (E > 1.0 MeV)

Neutron Fluence (a)

Neutron Fluence (b) 2 2

(n/cm )

(n/cm )

Location Measured (c)

Calculated Measured (c)

Calculated 19 19 Capsule R 4.42x10 4.20x10 19 19 19 19 Vessel IR 1.51x10 1.43x10 5.48x10 5.19x10 18 18 19 19 Vessel 1/4T 9.87x10 9.37x10 3.58x10 3.40x10 18 18 19 19 Vessel 3/4T 2.93x10 2.78x10 1.06x10 1.01x10 a) Current fluences are based on operation at 1650 MWt for 8.81 EFPY b) EOL fluences are based on operation at 1650 MWt for 32 EFPY.

c) The measured results of surveillance Capsule R were extrapolated to the reactor vessel locations using the following calculated lead factors:

Inner Radius - 2.93 1/4 Thickness - 4.48 3/4 Thickness - 15.10 w.m. ""

6-23

PRESSURE VESSEL SURVEILLANCE CAPSULE O*

13' (CAPSULES V,R)

/

23' (CAPSULES T,P)

/

33' (CAPSULES S,N)

/

'///

/

'///

49 f//////

2

/

/

p l ////////)

l

/ /

I I

/ /

/ / / //

THERMAL

/

/

SHIELD I

l

/

l

/

/

/

/

I I

/

I,/

/

REACTOR

/

/ /

CORE

/

I / /

/III,/

II;

///

Figure 6-1. Prairie Island Unit 2 Reactor Geometry OO7 A 19690-13 6 24

(13,23,33-)

CHARPY

[ SPECIMEN (12,22,32-)

/

/

///6 I////////

THERMAL SHIELD Figure 6-2. Plan View of a Reactor Vessel Surveillance Capsule 007 A-1969014 6-25

l l

- loI I o

SURVEILLANCE o

m CAPSULES I

N E

o C

PRESSURE

- m VESSEL IR x

D_]

I/4T LOCATION LL z iglo o[

H 3

Lij 3/4T LOCATION Z

io 9 I

I o

Io 20 30 40 50 60 AZIMUTHAL ANGLE (deg)

Figure 6-3. Calculated Azimuthal Distribution of Maximum Fast (E >1.0 MeV)

Neutron Flux Within the Reactor Vessel - Surveillance Capsule Geometry - 1650 MWt oo7.A 19690-15 O*2O

,x.

W I.O 9

W 8

DJ 7

(

6 xa 5

t/4T JL 4

z O

3

[

F-1/2T Wz 2

W>

s Q

3/4T JWx O.l 9

8 7

6 0

2 4

6 8

10 12 14 16 18 20 DEPTH INTO THE PRESSURE VESSEL (cm) 4 Figure 6-4. Relative Radial Variation of Fast Neutron (E >1.0 MeV) Flux and Fluence Within the Pressure Vessel OO7 A-19690-16 6-27

O.01 8

6 4

W OzW 2

D

_J

\\1.

N s

0.01 d

B =

6 Zo 4

?u Z

2 g>

s t-

<t

_J O.01 WT B :

6 4

2 0.001

-300

-200

-100 O

100 200 300 DISTANCE FROM CORE MIDPLANE (cm)

Figure 6-5. Relative Axial Variation of Fast Neutron (E >1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall OO7-A 1969017 6-28

SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Prairie Island Unit 2 reactor vessel:

Vessel Estimated Location Lead Removal Fluence Capsule (deg)

Factor Time [a]

(n/cm )

2 18[b]

V 77*

2.93 1.39 (removed) 5.86 x 10 19(b)

T 67*

1.78 4.00 (removed) 1.05 x 10 19(b)

R 257*

2.93 8.81 (removed) 4.42 x 10 19[c]

P 247*

1.78 18.00 5.20 x 10 19 N

237*

1.62 32.00 8.41 x 10 5

57*

1.62 standby a.

Effective full power years from plant startup b.

Actual fluence c.

Approximate fluence at vessel inner wall at and of life I

4 mr.iomim 71

SECTION 8 REFERENCES 1.

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

2.

ASTM E185-70 " Recommended Practice for Surveillance Tests for Nuclear Reactor" in ASTM Standards, Part 10 (1970), American Society for Testing and Materials, Philadelphia, Pa., 1970.

3.

ASTM E399-70T. " Plane-Strain Fracture Toughness of Metallic Materials," in ASTM Standards Part 10(1970), American Society for Testing Materials, Philadelphia, PA., 1970.

4.

Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission,. April 1977.

5.

Soltesz, R. G., Disney, R. K., Jedruch, 'J., and Ziegler, S. L.,

" Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5 - Two-Dimensional Discrete Ordinates Transport Technique," WANL-PR(LL)-034, Vol 5, August 1970.

6.

"0RNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".

7.

Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology (to be published).

8.

ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,

1984.

w, mom,-iron:

g.1

9.

ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.

10. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.

11. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.

12. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.

13. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,

1984.

14. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.

\\

15. ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron," in ASTM Standards, Section 12, American Society for Testir.g and Materials, Philadelphia, Pa.,1984.

16. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards, Section 12, America Society for Testing and Materials, Philadelphia, Pa.,1984.

me.*"""

8-2

17. ASTM Designation E481-78, " Standard Method for Measuring Neutro,-Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,

1984.

18. ASTM Designation E523-82, " Standard Method for Deterrnining Fast-Neutron Flux Density by Radioactivation of Copper," in ASTM Standards; Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.

19. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.

20. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.

21. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.

22. Yanichko, S. E., Anderson, S. L. and Kaiser, W. T., " Analysis of Capsule T from Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-9877, March,1981.

am. ion no.

8-3

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 RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART RT is designated as the higher of either NDT.

NDT the drop weight nil-ductility transition temperature (TNDT) or the temperature at which the material exhibits at least 50 ft ib of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.

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

RT is enhanced by certain chemical elements (such as copper) present in NDT reactor vessel steels. Design curves which show the effect of fluence and copper conten't on ART f r reactor vessel steels are shown in Figure A-1.

NDT Given the copper content of the most limiting material, the radiation-induced ARTNDT 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 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 insure that no other component will be limiting with respect to RTNDT*

=>.*"""

A-1

A-2.

FRACTURE TOUGHNESS 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.III The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the Prairie Island Unit 2 Vessel Material Surveillance Program.

A-3.

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup g

or cooldown cannot be greater than the reference stress intensity factor, KIR, f r the metal temperature at that time.

K is obtained from the IR reference fracture toughness curve, defined in Appendix G of the ASME Code.[2] The Kig curve is given by the equation:

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

(A-1) where K is the reference stress intensity factor as a function of the IR metal temperature T and the metal reference nil-ductility temperature RT Thus, the governing equatior. for the heatup-cooldown analysis is NDT.

defined in Appendix G of the ASME Code [2],, f,)),,,;

CKIM + KIt KIR (A-2) f where IM is the stress intensity factor caused by membrane (pressure) stress K

is the stress intensity factor caused by the thermal gradients It moer. io/uno.

A-2

K is a function of temperature relative to the RT f the material IR NDT C = 2.0 for Level A and level 8 service limits C = 1.5 for hydrostatic and leck test conditions during which the reactor i

core is not critical At any time during the heatup or cooldown transient, K is determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve.

The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, kit, f r the reference flaw are computed. From Equation (A-2), the pressure stress intensity factors are obtained and, from these, the allowable pressures 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 pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

l During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel 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 KIR at the 1/4 T location for finite cooldown rates than for steady state roer, iemno.

A-3

operation.

Furthermore, if conditions exist such that the increase in K IR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4 T 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 calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The' thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the i

f r the 1/4 T crack tip lags the coolant temperature; therefore, the KIR

~

crack during heatup is lower than the K for the 1/4 T crack during IR steady-state conditions at the same coolant temperature.

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K

's do not offset each IR 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 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal 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 thermal stresses are dependent on both the rate of amoemno.

A-4

heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve 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 three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit mast at all times be based on analysis of the most critical criterion.

Finally, the new 10CFR50[3] rule which addresses the metal temperature of the closure head flange and vessel flange regions is considered.

This 10CFR50 rule states that the metal temperature of the clostre flange regions must exceed the material RT by at least 120*F for normal operation when the NDT pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Prairie Island Unit 2). Table A-1 indicates that the limiting RT f 5'F occurs in the closure head dome of Prairie Island Unit 2, and NDT the minimum allowable temperature of this region is 125*F at pressures greater than 621 psig.

A-4.

HEATUP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in Section A-3.

The derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan.E43 2emamm..mn:

A-5

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 R indicate that both the surveillance weld metal and core region lower shell plate exhibited shifts in RT f 100*F and 85'F respectively at a fluence of NDT 19 2

4.42 x 10 n/cm.

This shift is well within the appropriate design curve (Figure A-1) prediction. Therefore, the heatup and cooldown curves in Figures A-3 and A-4 are based on the trend curve in Figure A-1, and these curves are applicable up to 15 effective full power years (EFPY).

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.

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 References 2 and 4.

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

2C97s 13n317s-870112 A-6

TABLE A-1 REACTOR VESSEL TOUGiNESS DATA (UNIRRADIATED)

Transverse [a]

50 ft Ib/35 mils RT Cu P

N1 NDTT Lateral Expansion NOT Average Transverse [a]

Component Heat No.

Material Type

(*F)

Temo (*F)

(*F)

Upper Shelf (ft Ib)

Closure Head Dome 54429 A533 Gr. B. C1. 1 5

52[c]

5 64[c]

Head Flange 22398 A508 C1. 3 0.08 0.009 0.72

-31 18[c]

-31 87[c]

Vessel Flange 22251 A508 C1. 3 0.06 0.014 0.73

-22 18[c]

-22 88[c]

Injection Nozzles 22398 A508 C1. 3 0.08 0.009 0.70

-22

-114[c]

-22 97[c]

Inlet and Outlet Nozzle 22371 A508 C1. 3 0.085 0.016 0.72

-13 50[c]

-10 89(c]

Upper Shell 2223/

A508 C1. 3 0.065 0.008 0.73

-13 41[c]

-13 85[c]

39088 Inter. She11[b]

22829 A508 C1. 3 0.075 0.010 0.75

-4 56

-4 112 Lower She11[b]

22642 A508 C1. 3 0.085 0.011 0.70

-13 54

-6 108 b

Trans. Ring 22219 A508 C1. 3 0.075 0.014 0.68 10 50[c]

10 76[c]

Dottom Head Dome 53795 A533 Gr.

D.

C1. 1 0.005 0.011 0.64

-13 56[c]

-4 68[c]

i Inter, to Lower Shell Saw 0.082 0.019 0.07

-31

-6

-31 103 Girth Weld HAZ HAZ

-31

-35

-31 117 a) Specimen oriented normal to the major working direction.

b) Based on actual transverse data through the surve111ance program.

c) Estimated using methods identified in USNRC Standard Review Plan NUREG-0800. Section 5.3.2 " Pressure Temperature Limits".

2097s:10/0317s-870112

i 11 u : uuuu:: -

n uu n:-

m m u u.u.

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

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[OpeEB L p T.

5 i :

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2 FLUENCE, n/cen (E > 1MeV) g Neld Metal Figure A-1 Effect of Fluence, Copper and Phosphorus Contents on ART for Reactor Vessel Steels A Shell Ebrging per Regulatory Guide 1.99, Revision 1 NOT

'A

~

k A

k" e

2

~ 7u.

y3 g

O

'N Y

& }U

=

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on an U 5

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3?

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, m. m vernsvon

• A-9

MATERIAL PROPERTY BASTS M ROLLING MATERIAL:

WE1.D ETAL COPPER MENT:

.082 WI5 PHOSPHORUS CONTENT:

.01R_Wr5 INITIAL RTNDr:

-31'T RT AFTER 15 EFPY:

1/4T,917 ygp 3/4T, 36 F CURVES APPLICABLE FOR HEATUP RATES UP TO 60 FAR FOR THg SERVICE PERIOD UP TO 15 EFPY AND N AINS MARGINS OF 10 AND 60 l

PSIG FOR POSSIBLE INSTRUMENT ERRORS 2500 j

f LEAK TEST LIMIT l

l l

2250

/

/

/

/

/

/

i i

2000

/

/

l 1

l

/

/

l 1750 UNACCEPTABLE OPERATION

/

/

I I

l g 1500 f

f l

G l

g

/

/

[

w w 1250 ACCEPTABLE g

n HEATUPgATESUP

/

OPERATION TO 60 F MR -

M

' s

/

l1000

/

o N

U 750

-CRITICALITY LIMIT BASEI) ON INSERVICE E

a HYDROSTATIC TEST

~

TEMPERATURE (223 F) 500 FOR THE SERVICE PERIOD UP TO 15 EFPY 250 0

50 100 150 200 250 300 350 400 450 500 INOlCATED TEMPERATURE (DEC.F)

FIGUrs A-3: PRAIRIE ISLAND UNIT 2 REACIOR CI)0LANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 15 EFPY l

A-10

1

, MATERTAI. PROPERTY RARTS CONTROLLING MATERIAL:

WEI.D ETAL COPPER CONTENT:

.082 WT5 PHOSPHORUS CONTENT:

.01gWT5 INITIAL RTg:

-31 F RT AFTER 15 EFPY:

1/4T, 91 F g

3/4T, 36 F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100 FAR FOR THE SERVICE PERIOD UP TO 15 EFPY AND (X)NTAINS MARGINS OF 10 F AND 60 PSIG FDR POSSIBLE INSTRUMENT ERRORS l

2500 l

1 2250

[

l

/

2000 j

UNACCEPTABLE f

OPERATION

/

1750

/

/

/

g 1500 b

/

ACCEPTABLE f

OPERATION 1250 l1000

[

0 C0QLDOWNRATES

[

4 FMR g

750;; o, 20,1 g

,' 40-w 500. 60C' e 100 250 00 50 100 150 200 250 300 350 400 450 500 IMOICATED TEMPERATURE (DEG.F)

FIGURE A-4: PRAIRIE ISLAND UNIT 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE FOR THE FIRST 15 EFPY A-11

I APPENDIX A REFERENCES 1.

" Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981.

2.

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

Appendices, " Rules for Construction of Nuclear Vessels," Appendix G,

" Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York, 1983.

3.

Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,

Amended May 17, 1983 (48 Federal Register 24010).

4.

" Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

sm. ionen o.

A-12

t.