ML17328A294

Analysis of Capsule U from American Electric Power Co DC Cook Unit 1 Reactor Vessel Radiation Surveillance Program

ML17328A294
Person / Time
Site:	Cook
Issue date:	01/31/1990
From:	Alertin L, Shaun Anderson, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
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ML17328A293	List: ML17328A292 ML17328A294
References
WCAP-12483, NUDOCS 9006260408
Download: ML17328A294 (149)

Text

HCAP-12483 WESTINGHOUSE CLASS 3

---.9207280249 ANALYSIS OF CAPSULE U FROM THE AMERICAN ELECTRIC POWER COMPANY D. C.

COOK UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM E. Terek S. L. Anderson L. Albertin N. K. Ray January 1990 Work performed under Shop Order No. ADZP-106 APPROVED:

.a.

T. A. Meyer, Ma ager Structural Materials and Reliability Technology Prepared by Westinghouse for the American Electric Power Company 1990 Hestinghouse Electric Corp.

4070s/030990:10 WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Division P.O.

Box 2728 Pittsburgh, Pennsylvania 15230 AP P IRQVED "IN GENERAL" NECHAV.iCALEP!G)NZERlKG DIV!SfON AMERICAN EL RIC PO'VitER SERVICE CORP.

PREFACE This report has been technically reviewed and verified.

Reviewer Sections 1 through 5 and 7, 8

Section 6

N. K. Ray E. P. Lippincott 40701/0l 0500:10

TABLE OF CONTENTS Section Titie Page

SUMMARY

OF RESULTS INTRODUCTION BACKGROUND 2-1 3-1 DESCRIPTION OF PROGRAM TESTING OF SPECIMENS FROM CAPSULE U

4-1 5-1 5-1.

Overvi ew 5-2.

Charpy V-Notch Impact Test Results 5-3.

Tension Test Results 5-4.

Compact Tension Tests 5-1 5-3 5-5 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1.

6-2.

6-3.

Introduction Discrete Ordinates Analysis Neutron Dosimetry 6-1 6-2 6-7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE REFERENCES 7-1 8-1 Appendix A - Heatup and Cooldown Limit Curves for Normal Operations 4070s/030900:10

'P

tk LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the D. C.

Cook Unit 1 Reactor Vessel 4-5 4-2 Capsule U Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-6 5-1 Charpy V-Notch Impact Data for D. C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Transverse Orientation) 5-15 5-2 Charpy V-Notch Impact Data for D. C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) 5-16 5-3 Charpy V-Notch Impact Data for D.

C.

Cook Unit 1 Reactor Vessel Weld Metal 5-17 Charpy V-Notch Impact Data for D. C.

Cook Unit 1

Reactor Vessel Weld Heat Affected Zone Metal 5-18 5-5 Charpy V-Notch Impact Data for D. C.

Cook Unit 1 Unit 1

ASTM Correlation Material 5-19 5-6 Charpy Impact Specimen Fracture Surfaces for D. C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) 5-20 5-7 Charpy Impact Specimen Fracture Surfaces for D. C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Transverse Orientation) 5-21 5-8 Charpy Impact Specimen Fracture Surfaces for D.

C.

Cook Unit 1 Reactor Vessel Weld Metal 5-22 l070s/010190:10 iv

LIST OF ILLUSTRATIONS (Cont),

Figure Titie Jl p

Page 5-9 Charpy Impact Specimen, Fracture Surfaces for D.

C.

Cook Unit 1 Reactor Vessel Held HAZ Metal 5-23 5-10 Charpy Impact Specimen Fracture Surfaces For ASTM Correlation Material 5-24 5-11 Tensile Properties for 0.

C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) 5-25 5-12 Tensile Properties for 0.

C.

Cook Unit 1 Reactor Vessel Held Metal 5-26 5-13 Fractured Tensile Specimens for 0.

C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) 0 5-14 Fractured Tensile Specimens for D. C.

Cook Unit 1 Reactor Vessel Held Metal 5-28 5-15 Typical Stress-Strain Curve for Tension Specimens 5-29 Plan View of a Reactor Vessel Surveillance Capsule 6-13 4070s/030090:l0 v"

LIST OF TABLES Tabl e Titie Page 4-1 Chemical Composition of the D; C; Cook Unit 1 Reactor Vessel Surveillance Materials 4-3 Heat Treatment of the D.

C.

Cook Unit 1

Reactor Vessel Surveillance Materials 4-4 5-1 Charpy.V-Notch Impact Data for the D.

C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 Irradiated at 550'F, Fluence 1.88 x 10 n/cm (E > 1.0 MeV) 5-2 5-3 Charpy V-Notch Impact Data for the D.

C. Cook Unit 1 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550'F,,Fluence 1.88 x 10 n/cm (E > 1.0 MeV)

Charpy V-Notch Impact Data for the D.

C.

Cook Unit 1 ASTM Correlation Monitor Material Irradiated at 550'F, Fluence 1.88 x 10 n/cm (E > 1.0 MeV) 5-7 5-8 5-4 Instrumented Charpy Impact Test Results for D.

C.

Cook Unit 1 Reactor Vessel Shell Plate B4406-3 5-9 5-5 Instrumented Charpy Impact Test Results for D.

C.

Cook Unit 1 Reactor Vessel Weld Metal and HAZ Metal 5-10 Instrumented Charpy Impact Test Results for D.

C.

Cook Unit 1 ASTM Correlation Monitor Material 5-11 5-7 The Effect of 550'F Irradiation at 1.88 x 10 n/cm (E

> 1.0 MeV) on the Notch Toughness Properties of The D.

C.

Cook Unit 1 Reactor Vessel Materials 5-12 4070s/030900:10 vi

LIST OF TABLES (Cont)

Table Title Page 5-8 Comparison of D. C.

Cook Unit 1 Reactor Vessel Surveillance Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions 5-13 5-9 Tensile Properties for D. C.

Cook Unit 1 Reactor Vessel Haterial Irradiated to 1.88 x 10 n/cm (E > 1.0 MeV) 19 2

5-14 6-1 Calculated Fast Neutron Exposure Rates at the Center of Surveillance Capsules 6-14 6-2 6-3 Calculated Fast Neutron Exposure Parameters at the Pressure Vessel Clad/Base Hetal Interface Relative Radial Distributions of Neutron Flux (E>1.0 MeY) Within the Pressure Vessel Wall 6-16 6-18 6-4 Relative Radial Distributions of Neutron Flux (E>0.1 MeV) Within the Pressure Vessel Wall 6-19 6-5 Relative Radial Distribution of Iron Displacement Rate (dpa) Within the Pressure Vessel Wall 6-20 6-6 Nuclear Parameters for Neutron Flux Monitors 6-21 6-7 Irradiation History of Neutron Sensors Contained in Capsule 4

6-22 6-8 Measured Sensor Activities and Reaction Rates Summary of Neutron Dosimetry Results 6-28 40708/030990: l0 V11

LIST OF TABLES (Cont)

Table Title Page 6-10 Comparison of Measured and FERRET Calculated Reaction Rates at the Surveillance Capsule Center 6-29 6-11 Adjusted Neutron Energy Spectrum at the Surveillance Capsule Center 6-30 6-12 Comparison of Calculated and Measured Exposure Levels for Surveillance Capsule U

6-31 6-13 Neutron Exposure Projections at Key Locations on the Pressure Vessel Clad/Base Metal Interface 6-32

,6-14 Neutron Exposure Values for Use in the. Generation of Heatup/Cooldown Curves 6-33 6-15 Updated Lead Factors for D.

C.

Cook Unit 1 Surveillance Capsules 6-34 i070@/010690: l0 vi 11

fg

SECTION 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in Capsule U, the fourth surveillance capsule to be removed from the American Electric Power Company D.

C.

Cook Unit 1 reactor pressure

vessel, resulted in the following conclusions:

o Capsule U was pulled at 9.17 EFPY and had a lead factor of 3.43.

Hith a lead factor of 3,43 at 9.17 EFPY capsule U received an average fast neutron fluence (E.> 1.0 MeV) of 1.88 x 10 n/cm at the geometric center of the capsule.

o Irradiation of the reactor vessel intermediate shell Plate B4406-3, to 1.88 x 10 n/cm

, resulted in 30 and 50 ft-lb transition temperature increases of 115 and 120'F, respectively, for specimens oriented normal to the major working direction (transverse orientation) and temperature increases of 115 and 125'F, respectively, for specimens oriented parallel to the major working direction (longitudinal orientation).

o Held metal irradiated to 1.88 x 10 n/cm resulted in a 205 and 245'F increase in the 30 and 50 ft-lb transition temperatures, respectively.

This results in a 30 ft-lb transition temperature of 115'F and a

50 ft-lb transition temperature of 175'F.

o Irradiation to 1.88 x 10 n/cm resulted in an average upper 19 2

shelf energy decrease of 1 ft-lb for plate B4406-3 (transverse orientation) and an average upper shelf energy decrease of 16 ft-lbs for the weld metal.

Both materials exhibit an upper shelf energy level greater than 50 ft-lb for plant operation through 32 EFPY.

i0705/031590:10

Comparison of the 30 ft-lb transition temperature increases for the 0.

C.

Cook Unit 1 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2, demonstrated that the Plate 84406-3 (longitudinal orientation) material transition temperature increase was 9'F greater than predicted.

This increase is bounded by the 2 sigma allowance for shift prediction of 34'F.

The weld metal showed a transition temperature increase that was 37'F less than the prediction.

o Since capsule U received a fluence of 1.88 x 10 n/cm

, which is 19 2

greater than the maximum calculate EOL (32 EFPY) fluence at the vessel inner surface of 1,41 x 10 n/cm (table 6-14),

and based on the results presented in Tables 5-7 and 5-8 it is recommended that the remaining untested surveillance capsules be held on standby and remain in their current locations.

See Section 7 for the current recommended surveillance capsule removal schedule.

4010s/031990:10 1-2

SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule U, the fourth capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the American Electric Power Company D. C.

Cook Unit 1 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the American Electric Power Company D.

C.

Cook Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation.

A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Yanichko a'nd Lege.

The surveillance program was planned 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 on Structural Materials in Nuclear Reactors" Westinghouse Energy Systems personnel

[291 were contracted to aid in the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Science and Technology Center Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes testing and the postirradiation data obtained from surveillance Capsule U removed from the American Electric Power Company D. C.

Cook Unit 1 reactor vessel and discusses the analysis of the data.

The data are also compared to capsules T,

X, and Y

which were previously removed from the reactor.

4010t/030S90:10 2-1

J

)P e

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 SA533 Grade B Class 1 (base material of the D. C.

Cook Unit 1 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 pressure vessels has been presented Failure," Appendix G to Section III Code.

The method utilizes fracture reference ni 1 -ducti 1 ity temperature guard against fast fracture in reactor in "Protection Against Non-ductile of the ASME Boiler and Pressure Vessel mechanics concepts and is based on the NDT)'TNDT i s def ined as the greater of ei ther the drop weight ni l-ducti 1 ity transition temperature (NDTT per ASTM E-208) or the temperature 60'f less than the 50 ft-lb (and 35-mi 1 lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material.

The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code.

The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel.

When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature.

Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

4070s/030900:10 3-1

RTNpT and, in turn, the operating 1 imits of nuc1 ear power pl ants can be adjusted to account for the projected 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 D. C.

Cook Unit 1 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 RTNpT to adjust the RTNDT for radi ati on embrittlement.

Thi s adjusted RTNpT

( RTNpT initial + hRTNpT) i s used to index the material to the K>R curve and, in turn, to set operating limits for the nuclear power plant which take into account the projected effects of irradiation on the reactor vessel materials.

i070t/030990:10 3-2

SECTION 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the D. C.

Cook Unit 1 reactor pressure vessel core region material were inserted in the reactor vessel'rior to-ini'tial'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 U (Figure 4-2) was removed after 9.17 effective full power years of plant operation.

This capsule contained Charpy V-notch impact, tensile, and 1X-HOL fracture mechanics specimens from the reactor vessel intermediate shell Plate 84406-3, weld metal representative of the beltline region weld seams, Charpy V-notch specimens from weld heat-affected zone (HAZ) material and Charpy V-notch specimens from ASTH correlation material.

All heat-affected zone specimens were obtained from the weld heat-affected zone of Plate 84406-3.

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.

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

Test specimens represent material taken at least one plate thickness from the quenched end of the plate.

All base metal Charpy V-notch impact and tensile specimens were oriented with the longitudinal axis of the specimen both normal to (transverse orientation) and parallel to (longitudinal orientation) the principal working direction of the plate.

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 normal to the welding direction.

The 1X-WOL fracture mechanics test specimens in Capsule U were machined such that the simulated crack in the specimen would propagate parallel to the major working direction and the major surfaces of the shell plate.

All specimens were fatigue precracked per ASTM E399-70T.

4070s/031590:10 4-1

Capsule U contai'ned dosimeters of pure iron, copper, nickel, and aluminum-cobalt wire (cadmium-shielded and unshielded),

and Neptunium (Np

) and 237 Uranium (U

) which measure the integrated flux at specific neutron energy levels.

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 1.75% Ag, 0.75% Sn, 97.5%

Pb Melting Point 579'F (304'C)

Melting Point 590'F (310'C)

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

4070t/010590:10 4"2

TABLE 4-1 CHEMICAL COMPOSITION OF THE D. C.

COOK UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Element Plate B4406-3 Wt.

Held Metal Ht.

C S

N2 Co Cu Si Mo Ni Mn Cr V

P Sn Ti AL Zn As B

Sb 0.24 0.015 0.008

<0.001 0.14 0.25 0.46 0,49 1.40 0.068

<0.001 0.009 0.010

<0.001 0.024

<0.001 0.010

<0.003 0.001 0.26 0.014 0.010

<0.001 0.27 0.18 0.44 0.74 1.33 0.022

.0. 001 0.023 0.006

<0.001 0.006 0.002 0.009

<0.003 0.001 40704/010590'.10 4-3

TABLE 4-2 HEAT TREATMENT OF THE D. C.

COOK UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Material Tem erature 'f Time hr Coolant Intermediate Shell Course Plate B4406-3 1600 1225 1150 4

4 40 Hater quenched Air cooled Furnace cooled Meldment 1150 40 Furnace cooled i070s/030S90:l0 4-4

X (220')

?70' (Is)i)

Y (320')

2 (33di)

Lao.

0' (di)

< (L76')

0 (Ido')

I 90'eactor VesseL Ther maI ShLeLck Cote SarzeL 2 (40')

Figure 4-1; Arrangement of Surveillance Capsules in the 0.

C.

Cook Unit 1 Reactor Vessel 4010u01059G: I0

~ssIKss ~

slssaat ssa sss ssa seasan csasrr csasrr cases csasrr csassr csssn csasrr csasn csun csun csasn curn M

caracas s s s+

sr a ~

s

'1 s Is a-ss

~ ~

s ss s.ss

~ ss SA1 sss

s. ~ Sssa sss s-ss s ss ss ws s-ss Ml

~ ss

~ 1 ar NN as s

sus a N

~

4ss 1 1 4'11 s-sr

~-s1 As SAs s-ss

~ sa r ~ I As as

'1 a ss am s 1 S.ss s"ss 14 a 1 1 st a ss SPECIMEN NUMBERING CODE A - PLATE B4406-3 (LONGITUDINALDIRECTION)

H WELD HEAT-AFFECTED ZONE AT - PLATE B4406-3 (TRANSVERSE DIRECTION)

W - WELD METAL R - ASTM CORRELATION-MONITOR Figure 4-2.

Capsule 0 Diagram Showing Location

ecimens, Thermal Monitors, and Dosimeters 4

550:10

SECTION 5 TESTING OF SPECIMENS FROM CAPSULE U

5-1.

OVERVIEH The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center Laboratory with consultation by Westinghouse Energy Systems personnel.

Testing was performed in accordance with 10CFR50, Appendices G and H

ASTM Specification E185-82 and Westinghouse Procedure MHL 8402, Revision 1 as modified by Westinghouse RMF Procedures 8102, Revision 1 and 8103, Revision 1.

Upon receipt of the capsule at the laboratory, the capsule was visually examinated and photographed for identification purposes.

The specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in HCAP-8047.

No discrepancies were found.

Examination of the two low-melting 304'C (579'F) and 310'C (590'F) eutectic alloys indicated'o melting of either type of thermal monitor.

Based on this examination, the maximum temperature to which the test specimens were exposed was less than 304'C (579'F).

The Charpy impact tests were performed per ASTM Specification E23-86 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358'achine.

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 (ED).

From the load-time curve, the load of general yielding (PG>),

the time to general yielding (tG>), the maximum load (PM), and the time to maximum load (tM) can be determined.

Under some test conditions, a sharp drop in load indicative of fast fracture was observed.

The load at which fast fracture was initiated is identified as the fast fracture load (PF),

and the load at which fast fracture terminated is identified as the arrest load (PA).

4070s/030900: l0 5-1

The energy at maximum.load (EM) was determined by comparing the energy-time record and the load-time record.

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

Therefore, the propagation energy for the crack (E

) is the difference between the total energy to fracture (EO) and the energy at maximum load.

The yield stress (ay) 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 A370-88 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

[34l P5]

RMF Procedure

8102, Revision 1.

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.

Oeflection 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 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.

Because of the difficulty in remotely attaching a thermocouple directly to the

specimen, the following procedure was used to monitor specimen temperature.

Chromel-alumel thermocouples 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 4070s/0309SO:10 5-2

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 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 contained in Capsule U irradiated to approximately 1.88 x 10 n/cm at 550'F are presented in Tables 5-1 through 5-6 and Figures'5-1 through 5-5.

The transition temperature increases and upper shelf energy decreases for the Capsule U material are shown in Table 5-7.

All materials contained in Capsule U exhibited an average Charpy upper shelf energy greater than 50 ft-lb at a fluence of 1.88 x 10 n/cm which is 19 greater than the projected fluence of 1,41 x 10 n/cm at 32 EFPY.

Irradiation of the vessel intermediate shell Plate 84406-3 material (transverse orientation) specimens to 1.88 x 10 n/cm (Figure 5-1) resulted in 30 and 50 ft-lb transition temperature increases of 115 and 120'F respectively, and an upper shelf energy decrease of 1 ft-lb when compared to the unirradiated data.

Irradiation of the vessel intermediate shell Plate 84406-3 material (longitudinal orientation) specimens to 1.88 x 10 n/cm (Figure 5-2) resulted in a 30 and 50 ft-lb transition temperature increase of 115 and 125'F, respectively, and an upper shelf energy decrease of 17 ft-lbs when compared to the unirradiated data.

4070'/032700:10 5-3

Weld metal irradiated to 1.88 x 10 n/cm (Figure 5-3) resulted in a 30 19 2

and 50 ft-lb transition temperature increase of 205 and 245'F respectively and an upper shelf energy decrease of 16 ft-lb.

Held HAZ metal irradiated to 1.88 x 10 n/cm (Figure 5-4) resulted in a 19 2

30 and 50 ft-lb transition temperature increase of 175 and 190'F respectively and an upper shelf energy decrease of 9 ft-lb.

ASTM correlation material irradiated to 1.88 x 10 n/cm (Figure 5-5) resulted in a 30 and 50 ft-lb transition temperature increase of 120'F and an upper shelf energy decrease of 10 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.

Table 5-8 shows a comparison of the 30 ft-lb transition temperature (hRTNDT) increases for the various D.

C.

Cook Unit 1 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2.

This comparison shows that the transition temperature increase resulting from irradiation to 1.88 x 10 n/cm is 9'F higher than predicted by the Guide for Plate 84406-3 longitudinal specimens and 4'F higher than predicted for transverse specimens.

The weld metal transition temperature increase resulting from 1.88 x 10 n/cm is less than the Guide prediction.

5-3.

TENSION TEST RESULTS The results of tension tests performed on Plate B4406-3

( longitudinal orientation) and weld metal irradiated to 1.88 x 10 n/cm are shown in 19 2

Table 5-9 and Figures 5-11 and 5-12, respectively.

These results show that irradiation produced a

12 to 15 Ksi increase in 0.2 percent yield strength for Plate 84406-3 and 18 to'20 Ksi increase for the weld metal.

Fractured tension specimens for each of the materials are shown in Figures 5-13 and 5-14.

A typical stress-strain curve for the tension specimens is shown in Figure 5-15.

4070'/030990:10 5-4

5-4.

COMPACT TENSION TESTS Per the surveillance capsule testing contract with the American Electric Power

Company, the 1X-WOL Fracture Mechanics specimens will not be tested and wi 11 be stored at the Hot Cell at the Westinghouse Science and Technology Center.

4070s/030990:10 5-5

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE D. C.

COOK UNIT 1 REACTOR VESSEL SHELL PLATE B4406-3 IRRADIATED AT 550'F, FLUENCE 1.88 x 10 n/cm (E > 1.0 MeV)

Temperature Impact Energy

~Sag le Ne. ~P

~C

~ft-1b

~J Lateral Expansion Shear

~oils

~mm Lon itudinal Orientation h60 h54 h56 h52 hSS h68 h51 h57 h53 h59 82 100 125 180 175 225 250 300 300 380

( 28 16.0 38 37.0 52 43.0 66 34.0

'9 63.0 107 121.0 121 100.0 (149 124.0 149 113.0 177) 110.0 21.8 13.0 80.0 24.0 58.0 33.0 46.0 30.0 85.5 43.0 164.0 78.0 135.6) 77.0 168.0) 82.0 153.0) 81.0 149.0) 73.0 (0.33) 10 0.61) 15 0.84) 25 0.76) 35 1.09) 40 1.98) 100

1. 96) 100 2.08) 100 2.06) 100 1.85) 100 hT60 hT57 hT59 hT58 hT84 hT51 hT56 hT52 hT56 hT53 50 82 100 125 180 200 240 250 300 350 Transverse Orientation 10 19.0 26.0 18.0 28 29.0 39.5) 20.0 38 22.0 30.0}

22.0 82 32.0 43.5) 21.0 66 37.0 50.0) 32.0 93 56.0 76.6) 42.0 116) 82.0 111.0) 65.0 121) 96.0 129.0) 71.0 149) 96.0 129.0) 61.0 177) 103.0 (139.5) 64.0 0.38) 10

0. 51) 16 0.56 20 0.81) 25 1.07) 75 1.65) 100 1.80) 100 1.55) 100 1.63) 100 S070s/010500:10 5-6

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE D. C.

COOK UNIT 1 REACTOR VESSEL MELD METAL AND HAZ METAL IRRADIATED AT 550'F FLUENCE 1.88 x 10 n/cm (E > 1.0 MeV)

Temperature

~Sem le Ne. ~F

~C Impact Energy

~ft-ib

~J Lateral Expansion Shear

~mile

~mm W47 W42 W44 W45 W43

'48 W46 W41 50

( 10) 74

( 23) 82

( 28) 125

( 52) 150

( BB) 200

( 93) 250 (121) 350 (177)

Weld Metal 21.0 28.6 53.0 71.8) 63.0 85.5) 38.0 51.5 48.0 65.0 61.0 82.5 90.0 122.0 95.0 129.0 14.0 35.0 41.0 31.0 35.0 50.0 63.0 68.0 0.36) 15 0.89) 65 (1.04) 60 0.79 65 0.89 70 1.27 100 1.60) 100 1.73) 100 H46 H47 H46 H44 H48 H43 H41 H42 0

(- 18) 50

(

10) 82

(

28) 250

( 121 275

( 135 350

( 177 HAZ 10.0 48.0 75.0 64.0 95.0 121.0 112.0 78.0 Metal 13.5 66.0) 101.5) 87.0) 129.0) 164.0) 152.0) 106.0) 10.0 36.0 42.0 46.0 72.0 67.0 72.0 63.0 0.25) 20 0.91) 35 1.07) 70 1.17) 80 1.83) 100 1.70) 100 1.83) 100

1. 60) 100 40104/010590:10 5-7

TABLE 5-3 CHARPY V-NOTCH IMPACT DATA FOR THE D. C.

COOK UNIT 1 ASTM CORRELATION MONITOR MATERIAL IRRADIATED AT 550'F, FLUENCE 1.88 x 10 n/cm (E > 1.0 MeV)

Temperature Impact Energy

~Sam le No. ~P

~G

~ft-1b

~J Lateral Expansion Shear

~oils

~nun R44 100 R42 175 R43 200 R48 225 R45 250 R46 275 R47 300 R41 350 38) 14.0 79) 26.0 93) 31.0 107) 68.0 (121) 81.0 135) 89.0 149) 112.0 177) 108.0 35.5I

~ 42.0 92.5 110.0 120.5 152.0 146.5) 12.0 20.0 25.0 49.0 59.0 67.0 70.0 74.0

0. 30) 10 0.51) 25

0. 64) 30 1.24) 90 1.50) 100 1.70) 100 1.78) 100

1. 88) 100 0

5070@i010590:10 5-8

TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR D. C.

COOK UNIT 1 REACTOR VESSEl. SHELL PLATE 84406-3 Normalized Ener Test Charpy C arpy Maximum Sample Temp

-Energy Ed/h Em/h Number

~P

~ft-lb ft-1b in'esProp Ep/h Yield Time Maximum Load to Yield Load

~kikse

~sec

~ki i~s Lon itudinal Orientation Time to Fracture hrrest Yield Flow Maximum Load Load Stress Stress

+i~sec

~ki s

~ki i~s

~ks i

~ksi h60 h54 h56 h52 h55 h58 cp h51 h53 h57 h59 hT60 hT57 hT59 hT58 hT54 hT51 hT56 hT52 hTSS hT53 82 16.0 100 37.0 125 43.0 150 34.0 175 83.0 225 121.0 250 100.0 300 113.0 300 124.0 350 110.0 50 19.0 82 29.0 100 22.0 125 32.0 150 37.0 200 66.0 240 82.0 250 95.0 300 95.0 350 103.0 129 102 298 189 346 233 274 301 507 313 974 293 80S 286 910 382 998 284 886 250 153 101 234 164 177 207 258

- 146 298 174 451 220 660 288 765 249 765 240 829 211 26 108 113

-27 195 682 619 628 715 836 62 69

-30 111 124 230 372 516 625 619 3.10 100 4.20 2.75 75 4.25 3.40 125 4.60 4.45 90 6.00 2.85 80 4.55 2.55 100 4.25 2.70 125 4.30 4.05 100 6.75 2.55 80 4.05 2.70 110 4.15 Transverse Orientation 2.65 70 3.90 3.00 85 4.60 4.45 135 5.75 2.95 100 4.25 2.40 90 4.05 2.95 85 4.30 3.90 105 6.65 2.50 95 4.05 2.50 100 3.90 2.60 110 4.10 260 435 530 500 670 690 695 660 676 605 260 370 395 355 445 505 510 600 600 525 4.15 4.15 4.55 6.00 4.50 3.90 4.45 5.65 4.15 3.90 4.20 0.20 0.30 1.40 1.80 0.25 0.25 0.95 1.20 1.90 102 121 91 116

'13 132 148 173 95 122 84 112 90 116 133 162 84 109 89 113 87 108 99 125 151 171 97 118 80 106 98 120 129 159 83 109 82 106 86 110

TABLE 5-5 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR D. C.

COOK UNIT 1 REACTOR VESSEL MELD METAL AND HAZ METAL Test Charpy Sample Temp Energy Number

~F

~ft-1b Normalized Ener C arpy Maximum Ed/h Em/h ft-lb in'esProp Ep/h Yield Time Load to Yield

~kikse

~sec Maximum Load

~ki )~s Time to Rracture hrrest Yield Rlow Maximum Load Load Stress Stress

+i~sec

~ki s

~ki~s

~ksi

~ksi V47 V42 V44 V45 V43 V48 V46 W41 50 21.0 74 53.0 82 63.0 125 38.0 150 48.0 200 61.0 250 90.0 350 95.0 169 427 507 306 387 491 725 765 244 263 308

-2 197 198 244 248 206 518 200 665 3.15 95 4.40 90 2.75 70 4.45 125 2.60 86 2.65 75 Veld Metal 160 9

4.35 80 COMPUTER MhLRUNGTION 5.80 4.60 6.90 4.35 4.70 3.90 3.75 275 620 605 435 450 510 606 5.75 4.35 6.70 4.35 0.15 1.25 2.50 1.35 144 168 103 128 145 171 91'18 148 169 SB 108 85 104 H45 H47 469 H44 H48 H43 H41 H42 0

10.0 50 48.0 82 76.0 150 64.0 200'5.0 250 121.0 276 112.0 350 78.0 81 52 387 384 604 245 515 246 765 307 974 218 902 350 628 211 29 2

369 269 458 766 552 417 5.20 4.8B 3.00 3.2 2.35 2.7 4.3 2.55 HhZ Metal 140 125 45 90 95-76 125 115 5.50 6.60 6.00 4.60 6.10 4.15 6.00 3.80 155 605 460 620 605 505 610 545 6.50 6.25 4.50 4.45 1.60 0.40 2.00 2.25 172 177 161 190 99 132 105 129 144 174 89 113 142 170 84 105

TABLE 5-6 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR D. C.

COOK UNIT 1 ASTM CORRELATION MONITOR MATERIAL Test Charpy Sample Temp Energy Number ~P

~ft-1b Normalized Ener Charpy aximum Prop Ed/h Em/h Ep/h ft-1b in~

ies Yie Time Load to Yield

~kikse ~sec maximum Time to Fracture hrrest Yield Rlow Load Maximum Load Load Stress Stress

~ki i~s +ij sec

~ki i+s

~ki~s

~ksi

~ksi R44 R42 R43 R48 R45 B46 R47 R41 100 14.0 175 26.0 200 31.0 225 68.0 250 81.0 276 89.0 300 112.0 350 108.0 113 98 116 209 108 125 260 54 100 648 126 130 652 435 95 717 436 95 902 539 85 870 624 40 6.06 3.80 5.05 6.60 4.15 5.45 6.36 3.80 220 310 380 776 620 610 665 605 4.95 3.80 4.85 6.60 0.60 3.80 4.85 6.60 132 96 126 141 89 126 123 76 149 110 147 162 113 154 150 101 4nlnerninnnn >n

TADLE 5-7 THE EFFECT OF 550'F IRRADIATION AT 1.88 x 10 n/cm (E > 1.0 MeV)

ON THE NOTCH TOUGHNESS PROPERTIES OF THE D. C.

COOK UNIT 1 SURVEILLANCE CAPSULE MATERIALS Average 30 ft-lb Tem

( F)

Average 35 mil Lateral Ex ansion Tem

('F)

Average Average Ener gy Absor pt ion 50 ft-lb Tem

'F at Full Shear (ft-lb)

Mater ial Unirradiated Irradiated hT Unirradiated Irradiated hT Unirradiated Irradiated hT Unirradiated Irradiated h(ft-lb)

Plate 84406"3 5

(Longitudinal) 120 115 20 140 120 30 155 125 130 113

-17 Plate 84406-3 (Transverse)

Weld Metal HAZ Metal 15

-90 130 115 115 205 70 175 25

-80

-75 140 115 125 205 75 150 65

-70

-65 185 175 120 120 96 245 110 190 126 95 94 117

-16 Correlation 45 Material 165 120 60 190 130 80 120 120 110

-10 Note:

All unirradiated data presented here was taken from WCAP-8047Ill 407 790: 10 0

TABLE 5-8 COMPARISON OF D. C.

COOK UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS MITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS Material Fluence

~Ca sul e 10 n/cm hRTNPT

( F) h USE DECREASE

(%)

R.G 1.99 R.G 1.99 Meas.

Pred.'eas.

Pred.

Plate B4406-3 T

(Longitudinal)

X Y

U 0.18 0.77 1.34 1.88 60 90 105 115 51.4 18 88.4 21 102.7 21

'111.3 13 15.4 21.1 24.5 26.5 Plate B4406-3 T

(Transverse)

X Y

'I U

0.18 0.62 1.06 1.88 70 110 115 115 51.4 14 82.7 19 97.0 21 111.3 1

15.4 20.5 23.0 26.5 Meld Metal 0.18 0.62 1.06 1.88 80 165 200 205 111.5 27 179.6 33 210.5 37 241.5 15 28.0 37.0 41.0'7.0 HAZ Metal T

X Y

U 0.18 0.77 1.34 1.88 120 160 165 170 25 36 38 7

Correlation Material 0.18 0.69 1.20 1.88 60 100 110 120 15 33 26 8

4070s/030990:10 5-,13

TABLE 5-9 TENSILE PROPERTIES FOR D. C.

COOK UNIT 1 REACTOR VESSEL MATERIAL IRRADIATED TO 1.88 x 10 n/cm (E > 1.0 MeV)

Test 0.2% Yield Ultimate-Fracture Fracture Fracture Sample Temp.

Strength Strength Load Stress Strength Material Number

~P

~ksi

~ksi

~ki

~ksi

~ksi Uniform Elongation Total Elongation Reduction in hrea Plate Plate Weld Weld h3 h4 W11 W12 74 600 74 600 83.0 70.8 83.5'9.5 103.9 95.7 97.8 96.8 3.40 199.8 3.20 162.6 3.30 167.7 3.85 171.6 89.3 85.2 67.3 78.4 11.3 10.5 12.8 10.5 23.1 21.6 23.9 19.4 65 60 60 54 4070s/0 I06SO:

0

-150

-100

-50 0

100 80 60 Pn (OC) 50 100 150 200 250 3

3 100 80 F

60 k

~

20 0

0 115oF

~

0 2.5 2.0 1.5 g LO-0.5 0

180 160 14

~ 120

@ 100 80 60 4

20 0

0 0

Unirradiate

,120 F

50t, o

~

Irradiated at 550'F 1.88 x 10 n/cm 80

-200

-100 0

100 200 300 40 500 Temperature ('F)

FIGURE 5-1 CHARPY V-NOTCH IMPACT DATA FOR D.

C.

COOK UNIT 1 REACTOR VESSEL SHELL PLATE B4406-3 (TRANSVERSE ORIENTATION) 5-15

-150

-100

-50 0

(OC) 50 100 150 200 250 100 80 ca 60 Pn 0

2 o ~2 3

3 20 0

100 80 6

60

'S 0

0 8

~

0 120'F

~

2.5 2.0 L5 g LO-0.5 0

180 160 140

~ 120

@ 100 80 60 40 20 0

0nirradiated 1250 F 1554 ~

~

Irradiated at 550 F

1.88 x 10 n/cm 19 80

-200

-100 0

100 200 300 4N 500 Temperature ('F)

FIGURE 5-2 CHARPY V-NOTCH IMPACT DATA FOR D.

C.

COOK UNIT 1 REACTOR VESSEL SHELL PLATE 84406-3 (LONGITUDINAL ORIENTATION) l070s/0106SO:10 5-16

-150 -100

-50 0

100 80 g

60 Pn 40 20 0

(4C) 50 100 150 200 250

~

~

100 80 E

60 20 0

205o F 2.5 2.0 L5 g LG-0.5 0

180 160 140

~ 120

@ 100 80 60 40 20 0

Unirradiated 8

245OF

~ Irradiated at 550'F 1.88x 10 n/cm 80

-200

-100 0

100 200 300 4N 500 Temperature

(

F)

FIGURE 5-3 CHARPY V-NOTCH IMPACT DATA FOR D.

C.

COOK UNIT 1 REACTOR VESSEL MELD METAL 5-17

(4C)

-150

-100

-50 0

50

-100 150 200 250 100 80 60 Pn 20 0

3 00 0

3 3

100 80 6

60 k 40

~

20 0

1904$

~

? 5 2.0 L5 LO 0.5 0

180 160 140

~ 120

@ 100 80 60 40 20 0

Unirradiated 0

0 1904F 1754F 2

0 4

Irradiated at 550'F 1.88 x 10 n/cm 80

-200

-100 0

100 200 300'0 500 Temperature ('F)

FIGURE 5-4 CHARPY V-NOTCH IMPACT DATA FOR D.

C.

COOK UNIT 1 REACTOR VESSEL WELD HEAT AFFECTED ZONE METAL 5-18

34m ~

,h60.

h54" h56'52

h55, h58 h51 h53 h57 FIGURE 5-6 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR D.

C.

COOK UNIT I REACTOR VESSEL SHELL PLATE B4406-3 (LONGITUOINAL ORIENTATION)

<070 @ i01 060 0: 10 5-20

ATGO AT57 AT59 AT58

~ N i

AT51 AT56 AT52 AT55 AT53 FIGURE 5-7 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR D.

C.

COOK UNIT 1

REACTOR VESSEL SHELL PLATE 84406-3 (TRANSVERSE ORIENTATION) 4 070 s/010690: l0 5-21

i~t(+~

t

~At Sy4~g,l W47

=

W42'

" '44 W45 h..

hht.'

t W~h gih@X

'F.

I

g@t"

. 'I W43 W48 W4S' "."'";.'41:

FIGURE 5-8 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR D. C.

COOK UNIT 1 REACTOR VESSEL MELD METAL 401011010890;10 5-22

H45'47...

H46 H48" H43'41 H42 FIGURE 5-9 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR D.

C.

COOK UNIT 1 REACTOR VESSEL WELD HAZ METAL.

~010ss010600.l0 5-23

I, R4'4;.

1

• I, g$~g<

)

'I

-R42

R43'48

't

.'4b

,R46

,.R47

.-,R41 FIGURE 5-10 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR ASTM CORRELATION MATERIAL 4070e/010090 10 5-24

120 110 100 80 70 60 oc 0

50 100 150 200 250 300 Ultimate Tensile Strength 700

0. 2% Yield Strength 80 70 Code:

Open Points Unirradiated Closed Points Irradiated at 550'F

1. 88 x 10 n/cm 60 50

=

40

~

30 Ch 20 10 0-I00 0

Reduction in Area Total Elongation Uniform Elon ation 100 200 300 4I 500 600 Temperature

(

F)

FIGURE 5-11 TENSILE PROPERTIES FOR D.

C.

COOK UNIT 1 REACTOR VESSEL SHELL PLATE B4406-3 (LONGITUDINAL ORIENTATION) 4070m/010890:10 5"25

120 110 100 g)

L 70 60 oC 0

50 100 150 200 250 300 Ultimate Tensile Strength

0. 2% Yield Strength 700 80 70 60 50 40

~

30 20 10 Code:

Open Points Unirradiated Closed Points Irradiated at 550'F

1. 88 x 10 n/cm 0

Reduction in Area Total Elongation Unif rm Elo gation

-100 0

100 200 300 400 Temperature ('F)

FIGURE 5-12 TENSILE PROPERTIES FOR D.

C.

COOK UNIT 1 REACTOR VESSEL HELD METAL 5-26

Specimen A3 74'It' 10tHS ZOOS>S 1

2 jlrJdl I

~"d~>i '

a's&/

i f

Q 4

Gv 89

~lA44JAI~tob4~A~hbbk Specimen A4 600'P FIGURE 5-13 FRACTURED TENSILE SPECIMENS FOR D.

C.

COOK UNIT 1 REACTOR VESSEL SHELL PLATE B4406-3 (LONGITUDINAL ORIENTATION) 4010s/010690:!0 5-27

l 0,"9.

<<-'6, T.'",8 fL

'pochs

< ~ '

-'.S,3.

44 Specimen" Wll

.74 P.

I 8:3 4

6*7 8,

S Nla&l]ljjjj4jbljj4ebl)Jjjlabhlgf~4.'i1~

0 Specimen

W12 600'F'IGURE 5-14 FRACTURED TENSILE SPECIMENS FOR D.

C.

COOK UNIT 1 REACTOR VESSEL WELD METAL

~070s/0 l0b00;10 5-28

j20 iio 80 Ol yo 8

80 C

eo io SPEC h4, 80OP 0.04 0.08 O.i2 STRhIN IN&N O.i8 0.2 FIGURE 5-15 TYPICAL STRESS-STRAIN CURVE FOR TENSION SPECIMENS 40iOs/010490:10 5-29

I

SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6. 1 INTRODUCTION Knowledge of the neutron environment within the reactor pressure vessel and 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.

The use of fast neutron fluence (E ) 1.0 MeV) to correlate measured materials proper ties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition.

In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice

E853, "Analysis and Interpretation of Light Water Reactor Surveillance Results,"

recommends reporting displacements per iron atom (dpa) along with fluence (E ) 1.0 MeV) to provide a data base for future reference.

The energy 4070s/010S90:10 6-1

dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice

E693, "Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom."

The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."

This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule U.

Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0. 1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history.

The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself.

Also, uncertainties associated with the derived exposure parameters are provided.

6.2 DISCRETE ORDINATES ANALYSIS A plan view of the reactor geometry at the core midplane is shown in Figure 4-1.

Eight irradiation capsules attached to the thermal shield-are included in the reactor design to constitute the reactor vessel surveillance program.

Four capsules are located symetrically at azimuthal angles of 4'nd 40'elative to the core cardinal axes as shown in Figure 4-1.

A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 6-1.

The stainless steel specimen containers are 1-inch square and approximately 38 inches in height.

The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core.

From a neutron transport standpoint, the surveillance capsule structures are 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 environment at the test specimen locations, the capsules themselves must be included in the analytical model.

4070s/010590;10

In performing the fast neutron exposure evaluations for. the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out.

The first, a single computation in the conventional forward

mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters (O(E

> 1.0 Hev,) y(E > 0. 1 Mev),

and dpa) through the vessel wall.

The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/p(E

> 1.0 HeV), within the pressure vessel geometry.

The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e.,

the 1/4T, 1/2T, and 3/4T locations.

The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core.

The importance functions generated from,these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement.

These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of,interest for the first 10 cycles of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles.

It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased.

4010s/010590;10 6-3

The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the forward calculation provided the means to:

1.

Evaluate neutron dosimetry obtained from surveillance capsule locations.

2.

Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.

3.

Enable a direct comparison of analytical prediction with measurement.

.4.

Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.

The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R, 8 geometry using the DOT two-dimensional discrete ordinates code [7] and the SAILOR cross-section library [8].

The SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications.

In these analyses anisotopic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an SB order of angular quadrature.

The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop plants.

Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery.

Furthermore, for the peripheral fuel assemblies, a 20 uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used.

Since it is unlikely that a single reactor would have a power distribution at the nominal

+2a level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.

4070sS010590"l0 6-,4

All adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library.

Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule.

Again, these calculations were run in R, 8 geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, 0

(E

> 1.0 MeV).

Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:

R (r, 8)

= f J8 JE I(r, 8, E)

S (r, 8, E) r dr d8 dE where; R(r, 8) p (E

> 1.0 MeV) at radius r and azimuthal angle 8

I (r, 0, E)

=

Adjoint importance function at radius, r, azimuthal angle 8, and neutron source energy E.

S (r, 8, E)

=.

Neutron source strength at core location r, 8 and energy E.

Although the adjoint importance functions used in the D.

C.

Cook Unit 1

analysis were based on a response function defined by the threshold neutron flux (E

> 1.0 MeV), prior calculations have shown that, 'while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order.

Thus, for a given location the ratio of dpa/p (E

> 1.0 MeV) is insensitive to changing core source distributions.

In the application of these adjoint important functions to the D.

C.

Cook Unit 1 reactor, therefore, calculation of the iron displacement rates (dpa) and the neutron flux (E

> 0,1 MeV) were computed on a cycle specific basis by using dpa/y (E > 1,0 MeV) and y (E

> G. 1 MeV)/g (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific p (E

> 1.0 MeV) solutions from the individual adjoint evaluations.

i070ci0305SO:lO

The reactor core power distributions used in the plant specific adjoint calculations were taken from fuel cycle design studies for the first ten operating cycle of 0, C.

Cook Unit 1 (9 thru 12].

Selected results from the neutron transport analyses performed for the 0.

C.

Cook Unit 1 reactor are provided in Tables 6-1 through 6-5.

The data listed in these tables establish the means for abs'olute comparisons of analysis and measurement for the capsule irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall.

In Table 6-1, the calculated exposure parameters (0 (E > 1.0 MeV),

p (E > 0. 1 MeV), and dpa} are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions.

The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis.

The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared.

Similar data is given in Table 6-2 for the pressure vessel inner radius.

Again, the three pertinent exposure parameters are listed for both the design basis and the cycle 1 through 10 plant specific power distributions.

It is important to note that the data for the vessel inner radius were taken at the clad/base metal interface;

and, thus, represent the max'.mum exposure levels of the vessel wall itself.

0 Radial gradient information for neutron flux (E

> 1.0 MeV), neutron flux (E

> 0. 1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively.

The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations.

Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.

4070@/010690;10

For example, the neutron flux (E

> 1.0 MeV) at the 1/4T position on the 45'zimuth is given by:

1/4T 45

=

4(220.27, 45')

F (225.75, 45')

where 01<4<(45')

=

Projected neutron flux at the 1/4T position on the 45'zimuth y (220.27, 45')

=

Projected or calculated neutron flux at the vessel inner radius on the 45'zimuth.

F (225.75, 45')

=

Relative radial distribution function from Table 6-3, Similar expressions apply for exposure parameters in terms of y(E > 0. 1 MeV) and dpa/sec.

6.3 NEUTRON DOSIMETRY I

The passive neutron sensors included in the D.

C.

Cook Unit 1 surveillance program are listed in Table 6-6.

Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest (y (E

> 1.0 Mev),

y (E > 0.1 MeV), dpa).

The relative locations of the neutron sensors within the capsules are shown in Figure 4-2.

The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules.

The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.

The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest.

i010sz010590:) 0

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 may be derived from the activation measurements only if the irradiation parameters are well

~----- --

known.

In particular, the following variables are of interest:

The specific activity of each monitor.

The operating history of the reactor.

The energy response of the monitor.

The neutron energy spectrum at the monitor location.

The physical characteristics of the monitor.

through 10 was tatus Summary Report" The specific activity of each of the neutron monitors was determined using established ASTM procedures

[13 through 25].

Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li),

gamma spectrometer.

The irradiation history of the D. C.

Cook Unit 1 reactor during cycles 1

obtained from NUREG-0020, "Licensed Operating Reactors S

for the applicable period.

The irradiation history applicable to capsule U is given in Table 6-7.

Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8.

Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7.

Relative to the measured reaction rates listed in Table 6-8, it should be noted that the U-238 and Np-237 results were 50-70 present low relative to the other radiometric sensors from capsule U as well as when compared to data from other capsules withdrawn from the 40 degree capsule position in other 4-loop plants.

These low results may have been due to incomplete recovery of cesium (C ) during the radiochemical processing.

However, the actual cause of the s

low results cannot be definitively determined at this time.

Because of this discrepancy, these fission monitor reaction rates were not used in the least squares adjustment of the capsule U dosimetry data.

4070'/032790:10 6-8

Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code (26].

The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data.

The exposure parameters a'long with associated uncertainties where then obtained from the adjusted spectra.

In the FERRET evaluations,

.a log normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations.

In general, the measured values f are linearly related to the flux p by some response matrix A:

where i indexes the measured values belonging, to a single data set s, g

designates the energy group and a delineates spectra that may be simultaneously adjusted.

For example, R,. =E a,.g 9

relates a set of measured reaction rates R. to a single spectrum y

by 1

g the multigroup cross section a

(In this case, FERRET also adjusts the lg cross-sections.)

The log normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.

4070'/010490,l 0

In the FERRET analysis of the dosimetry data, the continuous quantities (i.e.,

fluxes and cross-sections) were approximated in 53 groups.

The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [27].

This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide.

The 620-point spectrum was then easily --

collapsed to the group scheme used in FERRET.

The cross-sections were also collapsed into the 53 energy-group structure using SANO II with calculated spectra (as expanded to 620 groups) as weighting functions.

The cross sections were taken from the ENDF/B-V dosimetry file.

Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section.

Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.

For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight.

In some cases, as for'he cross sections.,

a multigroup covariance matrix is used.

More

often, a simple parameterized form is used:

i =RN+R R

I P

where RN specifies an overall fractional normalization uncertainty (i.e.,

complete correlation) for the corresponding set of values.

The fractional uncertainties R

specify additional random uncertainties for group g that 9

are correlated with a correlation matrix:

P,

= (I - 8) 5,

+ 8 exp [-

'I 2

gg gg

[2 Y ]

i07Gs/010690:lO 6-10

The first term specifies purely random uncertainties while the second term describes short-range correlations over a range Y (8 specifies the strength of the latter term.)

For the a priori calculated fluxes, a short-range correlation of Y = 6 groups was used.

This choice implies that neighboring groups are strongly correlated when 8 is close to 1.

Strong long-range correlations (or anticorrelations) were justified based on information presented by R.

E. Haerker (28].

Maerker's results are closely duplicated when Y = 6.

For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.

Results of the FERRET evaluation of the capsule U dosimetry are given in Table 6-9 The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1.88 x 10 n/cm (E > 1.0 MeV) with an 19 associated uncertainty of + 13%.

Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa).

Summaries of the fit of the adjusted spectrum are provided in Table 6-10.

In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates.

The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.

A summary of the measured and calculated neutron exposure of capsule U is presented in Table 6-12.

The agreement between calculation and measurement is good for all exposure parameters.

Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13.

Along with the current (9.17 EFPY) exposure, projections are also provided for an exposure period of 23 EFPY and to end of vessel design life (32 EFPY).

The time averaged exposure rates for the low leakage fuel cycles (8, 9,

10) were used to perform projections beyond the end of the cycle 1 through 10 exposure period.

4070'/031590:10 6-11

In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the D. C.

Cook Unit 1 reactor coolant system, exposure projections to 23 EFPY and 32 EFPY were employed.

Data based on both a

fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14.

In order to access RTNDT vs.

fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations 1/4T

=

4 (Surface)

(

3/4T

=

4 (Surface)

(

dpa (1/4T)

)

pa ur ace dpa (3/4T)

)

pa ur ace Using this approach results in the dpa equivalent fluence values listed in Table 6-14.

In Table 6-15 updated lead factors are listed for each of the D.

C.

Cook Unit 1 surveillance capsules.

These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules.

I070s/031590:10 6-12

3.49 2.63 2'."48 1.60 FLUX WIRES FLUX VIRES CAPSULE CENTER FLUX VIRES~

0.27 ~

a P

Gh a

Py C

a P

'h a

Py Radius (cm) 213.83 213.01

~

212.68 212.18 211.68 211.41 211.18 210. 68 210.14 209.81 208.97 THERMAL SHIELD Fig'Ure 6-1.

Plan View of a Reactor Vessel Surveillance Capsule 6-13

TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE CENTER OF SURVEILLANCE CAPSULES FLUX...E.>..1.0 MEV N CM2-SEC CYCLE 1 CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 DESIGN 4

DEG 2.06 x 1010 2.33 x 1010 2.26 x 1010 2.20 x 1010 2.22 x 1010 2.22 x 1010 2.24 x 1010 2.18 x',1010 2.18 x 1010 1.84 x 1010 2.49 x 1010 40 DEG 6.45 x 1010 7.94 x 1010 7.73 x 1010 7.41 x 1010 7.80 x 1010 7.29 x 1010 7.35 x 101 4.p8 x lplp 4.11 x 1010 3.87 x lplp 7.54 x 1010 FLUX E > 0.1 MEV N CM2-SEC CYCLE 1 CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 DESIGN 4

DEG 5.96 x 1010 6.74 x 10 6.54 x 1010 6.36 x 1010 6.42 x 1010 6.42 x 1010 6.48 x 1010 6.31 x 1010 6.31 x 1010 5.32 x 1010 7.20 x 1010 40 DEG 2.17 x lpll 2.67 x 1011 2.59 x 1011 2.49 x 1011 2.62 x 1011 2.45 x lpll 2.47 x 1011 1.37 x lpll 1.38 x 1011 1.30 x 1011 2.53 x 1011 6-14

TABLE 6-1 (Continued)

CALCULATEO FAST NEUTRON EXPOSURE RATES AT THE CENTER OF SURVEILLANCE CAPSULES CYCLE 1 CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 OESIGN 4

OEG 3.35 x 10-11 3'8 x 10-11 3.67 x 10 11 3.57 x 10 11 3.61 x 10 11 3.61 x 10 11 3.64 x 10-11 3.54 x 10 11 3.54 x 10 ll 2.99 x 10 11 4.04 x 10-11 40 OEG 1.10 x 10 10 1.36 x 10 10 1.32 x 10 10 1.27 x 10 10 1.33 x 10 10 1.25 x 10 10 1.26 x 10-10 6.98 x 10 11 7.03 x 10 11 6.62 x 10 11 1.29 x 10 10 6-15

TABLE 6-2

~ 8 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD BASE METAL INTERFACE FLUX E ) 1.0 Mev n cm2-Sec CYCLE 1 CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 DESIGN 0

DEG 6.16 x 109 7.04 x 109 6.74 x 109 6.58 x 109 6.62 x 109 6.64 x 109 6.68 x 109 6.36 x 109 6.22 x 109 5.46 x 109 8.02 x 109 15 DEG 9.90 x 109 1.14 x 10>>

1.09 x 1010 1.06 x 1010 1.07 x 1010 1.07 x 1010 1.08 x 1010 9.27 x 109 8.53 x 109 8.16 x 109 1.30 x 1010 30 DEG 1.23 x 1010 1.45 x 1010 1.41 x 1010 1.37 x 1010 1.41 x 10 1 ~ 35 x ]010 1.36 x 1010 8.46 x 109 8.33 x 109 7.91 x 109 1.61 x 1010 45 OEG 1.88 x 1010 2.29 x 1010 2.22 x 1010 2.15 x 1010 2.24 x 1010 2.12 x 1010 2.13 x 1010.

1.22 x 1010 1.22 x 1010 1.16 x 1010 2.49 x 10 0 FLUX E > O. l Mev n cm2-Sec CYCLE 1

CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 DESIGN 0

OEG 1.55 x 1010 1.77 x 1010 1.69 x 1010 1.65 x 1010 1.66 x 1010 1.67 x 1010 1.68 x 1010 1.60 x 1010 1.56 x 1010 1.37 x 1010 2.01 x 1010 15 DEG 2.48 x 1010 2.86 x 1010 2.73 x 1010 2.66 x 1010 2.68 x 1010 2.68 x 1010 2.71 x 1010 2.32 x 1010 2.14 x 1010 2.04 x 1010 3.26 x 1010 30 OEG 3.16 x 1010 3.73 x 1010 3.63 x 1010 3.52 x 1010 3.63 x 1010 3.47 x 1010 3.50 x 1010 2.18 x 1010 2.14 x 1010 2.03 x 1010 4.14 x 1010 45 OEG 5.00 x 1010 6.09 x 1010 5.90 x 1010 5.72 x 1010 5.96 x 1010 5.64 x 1010 5.66 x 1010 3.24 x 1010 3.24 x 1010 3.08 x 1010 6.62 x 1010

TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E ) 1.0 MeV)

WITHIN THE PRESSURE VESSEL WALL

.Radius 00 15'0'5'20.27(1) 220.64 221.66 222.99 224.31 225.63 226.95 228.28 229.60 230.92 232.25 233.57 234.89 236.22 237.54 238.86 240.19 241.51 242.D(2) 1.00 0.977 0.884 0.758 0.641 0.537 0.448 0.372 0.309 0.255 0.211 0.174 0.143 0.117 0.0961 0.0783 0.0635 0.0511 0.0483 1.00 0.978 0.887 0.762 0.644 0.540 0.451 0.373 0.310 0.257 0.212 0.175 0.144 0.118 0.0963 0.0783 0.0632 0.0501 0.0469 1.00 0.979 0.889 0.765 0.648 0.545 0.455 0.379 0.315 0.261 0.216 0.178 0.147 0.121 0.0989 0.0807 0.0656 0.0519 0.0487 1.00 0.977 0.885 0.756 0.637 0.534 0.443 0.367 0.303 0.250 0.206 0.169 0.138 0.113 0.0912 0.0736=

0.0584 0.0454 0.0422 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-18

TABLE 6-2 (Continued)

CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD BASE METAL INTERFACE d a sec CYCLE 1 CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 DESIGN 0

OEG 1.00 x 10-11 1.15 x 10-11 1.10 x 10 ll 1.07 x 10 11 1.08 x 10-11 1.08 x 10 11 1.09 x 10->>

1.04 x 10 11 1.01 x 10 11 8.90 x 10 12 1.31 x 10 11 15 OEG 1.59 x 10-11 1.84 x 10 11 1.75 x 10 ll 1.71 x 10 11 1.72 x 10 11 1.72 x 10 ll 1.74 x 10 11 1.49 x 10 ll 1.37 x 10 11 1.31 x 10 11 2.p9 x ]0-11 30 DEG 1.99 x 10 11 2.35 x 10-11 2.29 x 10 ll 2.22 x 10-11 2.29 x 10 11 2.19 x 10 ll 2.20 x 10 11 1.37 x lp-11 1.35 x 10 11 1.28 x 10 ll 2.61 x 10-11 45 OEG 3.06 x 10 11 3.73 x 10-11 3.61 x lp-11 3.50 x 10 11 3.65 x 10-11 3'5 x ]0-11 3.47 x 10 11 1.99 x 10 11 1.99 x 10-11 1.89 x lp-11 4.05 x ]0-11.

6-17

TABLE 6-4 RELATIVE RAOIAL OISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV)

WITHIN THE PRESSURE VESSEL WALL Radius

~cm po 15'0'5'20.27(1) 220.64 221.66 222.99 224.31 225.63 226.95 228.28 229.60 230.92 232.25 233.57 234.89 236.22 237.54 238.86 240.19 241.51 242.17(2) 1.00 1.00 1.00 0.965 0.916 0.861 0.803 0.746 0.689 0.633 0.578 0.525 0.474 0.424 0.375 0.328 0.283 0.239 0.229 1.00 1.00 0.996, 0.958 0.906 0.849 0.790 0.732 0.675 0.619 0.565 0.513 0.463 0.414 0.367 0.322 0.277 0.232 0.220 1.00 1.00 1.00 0.968 0.919 0.865 0.809 0.752 0.695 0.640 0.586 0.534 0.483 0.433 0.385 0.338 0.292 0.245 0.232 1.00 1.00 0.994 0.953 0.898 0.838 0.777 0.717 0.657 0.600 0.544 0.490 0.437 0.387 0.338 0.291 0.244 0.196 0.183 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-19

TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa)

WITHIN THE PRESSURE VESSEL WALL Radius

~cm 0'5'0'5'20.

27(1) 220.64 221.66 222.99 224.31 225.63 226.95.

228.28 229.60 230.92 232.25 233.57 234.89 236.22 237.54 238.86 240.19 241.51 242.17(2) 1.00 0.983 0.913 0.818 0.728 0.647 0.574 0.510 0.453 0.402 0.356 0.315 0.277 0.243 0.212 0.182 0.155 0.131 0.125 1.00 0.983 0.914 0.819 0.728 0.646 0.573 0.507 0.450 0.399 0.353 0.312 0.275 0.241 0.210 0.181 0.154 0.128 0.122 1.00 0.984 0.918 0.827 0.739 0.659 0.587 0.523 0.466 0.414 0.368 0.327 0.289 0.254 0.222 0.192 0.164 0.137 0.130 1.00 0.983 0.915 0.820 0.730 0.647 0.573 0.507 0.449 0.397 0.349 0.307 0.269 0.233 0.201 0.170 0.141 0.113 0.106 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-20

TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Monitor Material Reaction of Interest Target Weight Fraction

Response

~Ran e

Fission Product Yield Half-Life Copper Iron Nickel Uranium-238*

Cu63(n,a)Co60 0'917 Fe54(n,p)Mn54 0.0582 Ni58(n,p)Co58 0.6830 U238(n,f) Cs137 1.0 E) 4.7 MeV E) 1.0 MeV E> 1.0 MeV E> 0.4 MeV 5.272 yrs 312.2 days 70.90 days 30.12 yrs 5.99 Neptunium-237*

Np237(n,f)Cs137 1.0 E> 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum*

Co59(n,y)Co60 0.'0015 0.4ev

<E< 0.015 MeV 5.272 yrs Cobalt-Aluminum*

Co59(n,y)Co60 0.0015 E < 0.015 MeY*

5.272 yrs

• Denotes that monitor is cadmium shielded.

6-21

TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U

Month 1

2 3

4 5

6 7

8 9

10 11 121.

2 3

4 5

6 7

8 9

10 11 12 1

2 3

4 5

6 7

8 9

10 11 12 1

2 3

4 5

6 Year 1975 1975 1975 1975 1975 1975 1975 1975 1975 1975 1975 1975 1976 1976 1976 1976 1976 1976 1976 1976 1976 1976 1976 1976 1977 1977 1977 1977 1977 1977 1977 1977 1977 1977 1977 1977 1978 1978 1978 1978 1978 1978 PJ 3.1 27.2 768.8 2260.1 2598.8 2365.2 901.0 2485.4 2587.9 2080.2 1339.7 2451.4 2261.2 2420.2 2515.8 1277.8 1478.3 3164.3 2487.4 3226.1 2258.2 3249.0 2657.2 2338.9 0.0 2759.3 164.4 2347.3 1830.0 1675.4 2088.2 2347.0 1743.1 2142.6 1460.4 2780.8 2584.8 2907.2 3129.4 608.7 0.0 320.0 PJ PMAX 0.0010 0.0084 0.2366 0.6954 0.7996 0.7277 0.2772 0.7647 0.7963 0.6400 0.4122'.7543 0.6957 0.7447 0.7741 0.3932 0.4549 0.9736 0.7654 0.9926 0.6948 0.9997 0.8176 0.7197 0.0000 0.8490 0.0506 0.7222 0.5631 0.5155 0.6425 0.7222 0.5364 0.6593 0.4494 0.8556 0.7953 0.8945 0.9629 0.1873 0.0000 0.0985 Irradiation Time da s

14 28 31 30 31 30 31 31 30 31 30 31 31 29 31 30 31 30 31 31 30 31 30 31 31 28 31 30 31 30 31 31 30 31 30 31 31 28 31 30 31 30 Decay Time

~da~s 5366 5338 5307 5277 5246 5216 5185 5154 5124 5093 5063 5032 5001 4972 4941 4911 4880 4850 4819 4788 4758 4727 4697 4666 4635 4607 4576 4546 4515 4485 4454 4423 4393 4362 4332 4301 4270 4242 4211 4181 4150 4120 6-22

TABLE 6-7 (Continued)

IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U

Month 7

8 9

10 12 1

2 3

4 5

6 7.

8 9

10 12 1

, 2 3

45' 7

8 9

10 11 12 1

2 3

4 5

6 7

8 9

10 11 12 1

Year 1978 1978 1978 1978 1978 1978 1979 1979 1979

'979

'979 1979 1979 1979 1979 1979 1979 1979 1980 1980 1980 1980 1980 1980 1980 1980 1980 1980 1980 1980 1981 1981 1981 1981 1981 1981 1981 1981 1981 1981 1981 1981 1982 3032.3 2992.7 2982.4 3078.3 3203.6 2259.6 3090.6 3112.5 2875.5 571.9 0.0 0.0 1024.6 3221.0 3213.1 2779.1 2115.2 2316.9 1318.1 3174.4 3240.2 3072.7 3099.2 0.0 0.0 2018.2 2917.4 3107.6 3185.7 2464.9 2546.0 3238.8 3240.6 3244.7 3030.9 0.0 0.0 2406.1 3240.6 3241.8 1777.3 3022.1 2109.6 PJ PMAX 0.9330 0.9208 0.9176 0.9472 0.9857 0.6953 0.9510 0.9577 0.8848 0.1760 0.0000, 0.0000 0.3152 0.9911 0.9886 0.8551 0.6508 0.7129 0.4056 0.9767 0.9970 0.9455 0.9536 0.0000 0.0000 0.6210 0.8977 0.9562 0.9802 0.7584 0.7834 0.9965 0.9971 0.9984 0.9326 0.0000 0.0000 0.7403 0.9971 0.9975 0.5469 0.9299 0.6491 Irradiation Time da s

31 31 30 31 30 31 31 28 31 30 31 30 31 31 30 31 30 31 31 29 31 30 31 30 31 31 30 31 30 31 31 28 31 30 31 30 31 31 30 31 30 31 31 Decay Time

~da~s 4089 4058 4028 3997 3967 3936 3905 3877 3846 3816 3785 3755 3724 3693 3663 3632 3602 3571 3540 3511 3480 3450.

3419 3389 3358 3327 3297 3266 3236 3205 3174 3146 3115 3085 3054 3024 2993 2962 2932 2901 2871 2840 2809 6-23

TABLE 6-7 (Continued)

IRRAOIATION HISTORY OF NEUTRON SENSORS CONTAINEO IN CAPSULE U

Month 2

3 4

5 6

7 8

9 10 11 12 1

23' 6

7 8

9 10 11 12 1

2 3

4 5

6 7

8 9

10 11 12 1

2 3

4 5

6 7

8 Year 1982 1982 1982 1982 1982 1982 1982 1982 1982 1982 1982 1983 1983 1983 1983 1983 1983 1983 1983 1983 1983 1983 1983 1984 1984 1984 1984 1984 1984 1984 1984 1984 1984 1984 1984 1985 1985 1985 1985 1985 1985 1985 1985 0.0 2720.5 3116.2 2947.5 3235.2 195.7 0.0 17.7 2642.2 3160.3 2930.5 3187.6 3206.1 2978.0 3192.3 2789.7 3031.5 1268.4 0.0 0.0 434.3 2050.4 991.3 2342.8 2787.6 3064.2 2576.9 3224.1 2585.0 2818.5 1814.1 3096.8 2729.0 3038.6 2842.9 940.4 3089.0 3093.1 474.7 0.0 0.0 0.0 0.0 PJ PMAX 0.0000 0.8371 0.9588 0.9069 0.9954 0.0602 0.0000 0.0055 0.8130 0.9724 0.9017 0.9808 0.9865 0.9163 0.9822 0.8584 0.9328 0.3903 0.0000 0.0000 0.1336 0.6309 0.3050 0.7208 0.8577 0.9428 0.7929 0.9920 0.7954 0.8672 0.5582 0.9529 0.8397 0.9349 0.8747 0.2894 0.9505 0.9517 0.1461 0.0000 0.0000 0.0000 0.0000 Irradiation Time da s

28 31 30 31 30 31 31 30 31 30 31 31 28 31 30 31 30 31 31 30 31 30 31 31 29 31 30 31 30 31 31 30 31 30 31 31 28 31 30 31 30 31 31 Oecay Time

~da~s 2781 2750 2720 2689 2659 2628 2597 2567 2536 2506 2475 2444 2416 2385 2355 2324 2294 2263 2232 2202 2171 2141 2110 2079 2050 2019 1989 1958 1928 1897 1866 1836 1805 1775 1744 1713 1685 1654 1624 1593 1563 1532 1501 6-24

TABLE 6-7 (Continued)

IRRADIATION HISTORY QF NEUTRON SENSORS CONTAINED IN CAPSULE U

Month Year PJ

~MW PJ PMAX Irradiation Time da s

Decay Time

~da~s 9

10ll 12 1

2 3

4 5

6 7

8 9

10 11 12 1

2 3

4 5

6 7

8 9

10 11 12 1

2 3

4 5

6 7

8 9

10 11 12 1

2 3

1985 1985 1985 1985 1986 1986 1986 1986 1986 1986 1986 1986 1986 1986 1986 1986 1987 1987 0.0 0.0 413.5 1491.8 2932.9 2927.5 2930.5 2640.9 2837.1 0.0 1100.8 2670.6 2928.8 2931.8 2784.4 2949.2 2889.1 2846.8 1987 1987 1987 1987 1987 1987 1987 1988 1988 1988 1988 1988 1988 1988 1988 1988 2272.0 0.0 0.0 0.0 1782.7 2933.6 2818.3 2791.6 2870.1 2697.8 2902.5 2898.6 2760.1 2933.6 3091.8 2011.5 1988 2493.2 1988 2669.8 1988 2935.2 1989 2481.3 1989 1989 2327.0 1586.1 1987 2821.6 1987 1346.4 1987

.2764.2 0.0000 0.0000 0.1272 0.4590 0.9024 0.9008 0.9017 0.8126 0.8730 0.0000 0.3387 0.8217 0.9012 0.9021 0.8567 0.9074 0.8889 0.8759 0.8682 0.4143 0.8505 0.6991 0.0000 0.0000 0.0000 0.5485 0.9026 0.8672 0.8590 0.8831 0.8301 0.8931 0.8919 0.8492 0.9026 0.9513 0.6189 0.7671 0.8215 0.9031 0.7635 0.7160 0.4880 30 31 30 31 31 28 31 30 31 30 31 31 30 31 30 31 31 28 31 30 31 30 31 31 30 31 30 31 31 29 31 30 31 30 31 31 30 31 30 31 31 28 19 1471 1440 1410 1379

1348, 1320 1289 1259 1228 1198 1167 1136 1106 1075 1045 1014 983 955 924 894 863 833 802 771 741 710 680 649 618 589 558 528 497 467 436 405 375 344 314 283 252 224 205 6-25

TABLE 6-8 MEASUREO SENSOR ACTIVITIES ANO REACTION RATES Monitor and Axial Location Measured Activity dis sec-m Adjusted Saturated Acticity dis sec-m Reaction Rate RPS NUCLEUS Cu-63 n a Co-60 Top-Middle Middle Bottom-Middle Average 1.30 x 105 1.26 x 105 1.33 x 105 1.30 x 105 2.74 x 105 2.66 x 105 2.80 x 105 2.73 x 105 4.17 x 10 17 Fe-54 n

Mn-54 Top Top-Middle Middle Bottom Average 7.10 x 105 7.34 x 105 7.31 x 105 7.10 x 105 7.21 x 105 2.50 x 106 2.59 x 106 2.57 x 106 2.50 x 106 2.54 x 106 4.05 x 10 15 Ni-58 n

Co-58 Top-Middle Middle Bottom-Middle Average 2.30 x 106 2.27 x 106 2 36 x 106 2.31 x 106 4.21 x 107 4.15 x 107 4.32 x 107 4.23 x 107 6.03 x 10-15 U-238 n f Cs-137 Cd Middle 3.67 x 105 2.10 x 106 1.39 x 10 14 6-26

TABLE 6-8 (Continued)

MEASURED SENSOR ACTIVITIES AND REACTION RATES Monitor and Axial Location N -237 n f Cs-137 Cd Measured Activity dis sec-m Ad)usted Saturated Acticity dis sec-m Reaction Rate RPS NUCLEUS Middle 2.87 x 106 1.65 x 107 9.94 x 10 14 Co-59 n

Co-60 Cd Top Bottom Average 8.66 x 106 8.01 x 106 8.34 x 106 2.21 x 107 2.04 x 107 2.12 x 107 1.39 x 10-12 Co-59 n Y Co-60 Top Bottom Average 2.06 x 107 1.93 x 107 2.00 x 107 4.41 x 107 4.13 x 107 4.27 x 107 2.79 x 10 12 6-27

TABLE 6-9

SUMMARY

OF NEUTRON OOSIMETRY RESULTS TIME AVERAGED EXPOSURE RATES y (E> 1.0 MeV) (n/cm2-sec) g (E> 0.1 MeV) (n/cm2-sec}

6.50 x 1010 2.21 x 1011

+ 13K

+ 22K dpa/sec 1.09 x 10 10

+ 16K INTEGRATEO CAPSULE EXPOSURE 4

(E> 1.0 MeV) (n/cm2}

(E> O.l MeV) (n/cm2) 1.88 x 1019 6.40 x 1019

+ 13K

+ 22K dpa 3.16 x 10-2

+ 16K NOTE:

Total Irradiation Time

= 9.17 EFPY 6-28

TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER Reaction Measured Adjusted Calcu1ation Cu-63 (n,a)

Co-60 4.17x10 17

4. 18xlo-17 Fe-54 (n,p) Mn-54 N$-58 (n,p) Co-58 Co-59.(n,y)

Co-60 (Cd) 4.05x10-15 6.03xl0-15 1.39x10-12 4.14x10-15 5.88x10 15 1.37x10 12 Co-59 (n,y) Co-60 2.79x10 12 2.83x10 12 6-29

TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER

~Grou Energy Adjusted Flux Garou Energy Adjusted Flux 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 NOTE:

1.73xlol 1.49xlol 1.35xlol 1.16xlol 1.OOxlOl 8.61xloo 7.4lxlOO 6.07xloo 4.97xloo 3.68x100 2.87x100 2.23X100 1.74xloo 1.35xloo 1.11xloo 8.21xlo 1

6.39xlo 1

4 98xlo-1 3.88xlo 1

3.02xlo"1 1.83xlo-l 1.11xio 1

6.74xlo 2

4.09xlo 2 2.55xlo 2 1.99x10-2 1.5OxlO-2 Tabulated energy 3.47x106 8.50x106 3.92x107 9.98x107 2.42x108 4.40x108 1.07x109 1.56x109 3.23x109 4.14x109 8.31xlo9 1.04xlolo 1.37xlolo 1.37xlolO 2.34xlolo 2.48xlolo 2.43xlolo 1.71xlolo 2.25xlolo 2.48x1010 2.34xlolo 1.84xlo 1.32xlolo 7.94x109 9.49x109 5.15x109 6.95xl09 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 9.12xlo 3

5.53xlo 3

3.36xlo 3

2.84xlo 3

2.40xlo 3

2.04x10-3 1.23xlo 3

7.49xlo 4 4.54xlo "

2.75xlo 4 1.67xlo 4

1.01X10 4

6.14xlo 5

3.73xlo 5

2.26xlo 5

1.37xlo 5

8.32xlo 6

5.04xlo 6 3.06xlo 6

1.86xlo 6 13xlo 6 6.83xlo 7

4.14xlo 7 2.51xlo"7 1.52xlo "

9.24xlo 8 8.92x109 1.12xlolo 3.48xl09 3.32x109 3.22xl09 9.37x109 9.14xl09 8.88xl09 8.68x109 9.20x109 1.O5xlO'0 9.91X109 9.77x109 9.47x109 9.11xl09 8.76x109 8.30xl09 7.68xl09 7.18x109 6.62xl09 5.13x109 4.37xl09 5.26x109 4.81xl09 4.84x109 1.09xlolo levels represent the upper energy of each group.

6-30

TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR SURVEILLANCE CAPSULE U

> (E> 1.0 MeV) (n/cm2}

g (E> 0.1 MeV) {n/cm2)

Ca1cu1ated 1.76 x 1019 5.91 x 1019 Measured 1.88 x 1019 6.40 x 10>>

~CM 0.94 0.92 dpa/sec 3.01 x 10 2 3.16 x 10-2 0.95 6-31

~ TABLE 6-13 NEUTRON EXPOSURE PROJECTINS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAO BASE METAL INTERFACE 9.17 EFPY 15'ZIUTHAL ANGLE 30'5'(E>

1.p Mey) 1.96 x 1p18 3.p4 x 1p18 3.54 x 1p18 5.46 x 1p18 (n/cm2)

@(E> 0.1 Mey) 5 p2 1p18 7 79 x 1p18 9 30 x 1p18 1 48 1p19 (n/cm2) dpa 23.0 EFPY 3.16 x 10 4.84 x 10 5.68 x 10 8.79 x 10-3 y(E> 1.0 MeV) 4..56 x 1018 6.83 x 1018 7.14 x 1018 1.05 x 1p (n/cm2) 43(E> 0.1 MeV) 1.15 x 1019 1.73 x 1019 1.85 x 1019 2.86 x 1019 (n/cm2) dpa 32.0 EFPY 7.42 x 10 1.09 x 10 1.16 c 10-2 1.74 x 10-2 E> 1 p Mey) 6 26 1p18 g 3p 1p18 g 48 1018 1 41 1p19 n/cm2) y(E> p.l MeV) 1.58 x 1019 2.35 x 10 2.46 x 10 3.76 x 101 (n/cm2)

~

dpa 1.p2 x 1p-2 1.49 x 10-2 1.53 x 10-2 2.28 x 1p-2 6-32

TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP COOLDONW CURVES NEUTRON FLUENCE E > 1.0 HeV SLOPE (n/cm2) 23 EFPY

~da SLOPE (equi val ent n/cm2)

Surface

~14 T

~3'4 T Surface

~14 T 34T 00 15 30'5' 56 x 101&

6.&3 x 1018 7.14 x 1018 l.p5 x 1019 2.41 x 1018 3.64 x 1018 3.84 x 1018 5.63 x lpl&

4 96 x 1017 7 51 x 1017 8.06 x 1017 1.12 x 1018 4.56 x 1018 6.83 x 1018 7.14 x 1018 1.07 x 1019 2.92 x 1018 4.37 x lpl&

4.66 x 1018 6.84 x 1018 1.06 x 10 1.57 x 1018 1.73 x 1018 2.36 x lpl8 NEUTRON FLUENCE E > 1.0 MeV SLOPE (n/cm2) 32 EFPY

~da SLOPE (equivalent n/cm2)

Surface 1 4T

~34 T Surface

~14 T 34T po 15'0'5'.26 x 10 8 9.3p x 1018 9.48 x 1018 1.41 x 1019 3'1 X lpl&

4.94 x 1018 5.09 x 1018 7.42 x 1018 6.82 x 1017 1.02 x 1018 1.07 x 1018 1.48 x 1018 6.26 x 1018 9.30 x 1018 9.'48 x 1018 1.41 x 1019 4.00 x 1018 5.94 x 1018 6.18 x 1018 9.00 x 1018 1.45 x 1018 2.13 x 1018 2 29 x lpl&

3.11 x 1018

TABLE 6-15 UPDATED LEAD FACTORS FOR D.

C COOK UNIT 1 SURVEILLANCE CAPSULES Capsule Lead Factor

3. 43 3.43 3.43 3.43

.0.96 0.96 0.96 0.96 Plant specific evaluation integrated through cycle 10 (9.17 EFPY) 6-34

SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE PI The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the D. C.

Cook Unit 1 reactor vessel:

Vessel Location

~Ca sula

~de Lead Factor Removal Time

'Actual

Capsule-Fluence

~n/cm 40 40 40 40 3.43 3.43 3.43 3.43 1.13 (removed) 0.18 x 10 3.48 (removed) 0.77 x 10 4.94 (removed) 1.34 x 10 9.17 (removed) 1.88 x 10 0.96 Standby 0.96 Standby 0.96 Standby 0.96 Standby a)

Effective full power years from plant startup b)

Approximate EOL (32 EFPY) fluence at 1/4T location (at 45'F) and corresponds to fluence value midway between 1st and 3rd capsule.

c)

Approximate EOL (32 EFPY) fluence at vessel inner wall (at 45')

d)

Not less than once or greater than twice the estimated peak EOL (32 EFPY) vessel fluence.

4070s/030NO:10 7-1

SECTION 8 REFERENCES 1.

Yanichko, S. E.,

and Lege, D. J.,

"American Electric Power Service Corp.

Donald C.

Cook Unit No.

1 Reactor Vessel Radiation Surveillance Program,"

HCAP-8047, March 1973.

2.

"Reactor Vessel Material Surveillance Program for Donald C.

Cook Unit No.

1 Analysis of Capsule T," Final Report SHRI Project 02-4770, December 1977.

3.

"Reactor Vessel Material Surveillance Program For Donald C.

Cook Unit No.

1 Analysis of.Capsule X," Final Report SHRI Project 02-6159, June 1981.

4.

"Reactor Vessel Material Surveillance Program for Donald C.

Cook Unit No.

1 Analysis of Capsule Y," Final Report SHRI Project No. 06-7244-001, January 1984.

5.

Code of Federal Regulations,

10CFR50, Appendix G, "Fracture Toughness Requirements" and Appendix H, "Reactor Vessel Material Surveillance Program Requirements",

U. S. Nuclear Regulatory Commission, Washington, D.C.

6.

Regulatory Guide 1.99, Revision 2, "Radiation fmbrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988.

7.

R.

G. Soltesz, R,

K. Disney, J. Jedruch, and S. L. Ziegler, "Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation.

Vol.

5Two-Dimensional Discrete Ordinates Transport Technique",

WANL-PR(LL)-034, Vol. 5, August 1970.

8.

"ORNL RSCI Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Hater Reactors".

4070s/030QSO.'

0 8-1

9.

AEP Letter RBB 88-005/4, R.

B. Bennett (AEP) to H.

C. Wells (Westinghouse),

January 29, 1988.

10.

WCAP-10376, B. Y. Hubbard, et al.,

"Core Physics Characteristic of the Donald C.

Cook Station Nuclear Plant (Unit 1, Cycle 8), July 1983.

11.

WCAP-10862, F. J. Silva, et al.,

"Core Physics Characteristics of the Donald C.

Cook Station Nuclear Plant (Unit 1, Cycle 9), August 1985 12.

WCAP-11586, B. J.

Johansen et al., "Nuclear Parameters and Operations Package for the Donald C.

Cook Station Nuclear Plant (Unit 1, Cycle 10),

October 1987 13.

ASTH 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.

14.

ASTM Designation E560-77, "Standard Recommended Pr'actice for Extrapolating Reactor Vessel Surveillance Dosimetry Results",

in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

15.

ASTM Designation E706-81a, "Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard",

in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

16.

ASTM Designation E853-84, "Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results",

in ASTM

'tandards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

4070m/030090:10 8"2

17.

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 Haterials, Philadelphia, PA, 1984.

18.

ASTM Designation E262-77, "Standard Method for Measuring Thermal Neutron Flux by Radioactivation TecCiniques",

in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

19.

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

20.

ASTH Designation E264-82, "Standard Method for Determining Fast-Neut'ron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

21.

ASTM Designation E481-78, "Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

22.

ASTM Designation E523-82, "Standard Method for Determining Fast-Neutron Flux Density by.Radioactivation of Copper",

in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

23.

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.

24.

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.

4070s/030990:10 8-3

25.

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.

26.

F. A. Schmittroth, FERRET Data Anal sis Core, HEOL-TME 79-40, Hanford Engineering Development Laboratory,

Richland, HA, September 1979.

27.

H. N. McElroy, S.

Berg and T. Crocket, A Com uter-Automated Iterative Method of Neutron Flux S ectra Determined b

Foil Activation, AFWL-TR-67-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.

28.

EPRI-NP-2188, "Development and Demonstration of an Advanced Methodology for LHR Dosimetry Applications",

R.

E. Maerker, et al.,

1981.

29.

ATSM Designation E185-70, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors".

30.

ASTM Designation E208-87, "Standard Test Method for Conducting Drop-Height Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels".

31.

ASTM.Designation E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Hater Cooled Nuclear Power Reactor Vessels, E706

( IF)".

32.

ASTM Designation E23-86, "Standard Test Methods for Notched Bar Impact Testing of Metallic Materials".

33.

ASTM Designation A370-88, "Standard Test Methods and Definitions for Mechanical Testing of Steel Products".

34.

ASTM Designation E8-83, "Standard Test Methods of Tension Testing of Metallic Materials".

4070 s/030990:10 8-4

35.

ASTW Oesignation E21-79, "Standard Practice for Elevated Temperature

~

~

~

~

~

~

Tension Tests of Metallic Materials".

36.

AS'esignation E83-67, "Standard Practice for Verification and Classification of Extensometers".

4070m/030990.'10 8-5

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

INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference ni 1-ducti 1 ity temperature) for the reactor vessel

~

The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced hRTNDT.

RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.

RTNDT increases as the material i s exposed to fast-neutron radi ati on ~

Therefore, to find the most limiting RTNDT at any time period in the reactor's life, aRTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels.

The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev.

2 (Radiation Embr ittlement of Reactor Vessel Materials)

The value, "f", given in figure A-1 is the calculated value of the neutron fluence at the location of interest (inner surface, 1/4T, or 3/4T) in the vessel at the location of the postulated defect, n/cm (E >

1 MeV) divided by 10 The fluence factor is determined from figure A-l.

19 A-2.

FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan 4111'/031290:10 A-1

Fast neutron irradiation embrittlement of reactor vessel materials is largely dependent on the chemical composition, particularly the copper concentration.

The variability in irradiation-induced property changes, which exist in

general, is cdmpounded by the variability of copper concentration within the weldments.

Westinghouse normally uses the method identified in Section C-1.1 of revision 2 of Regulatory Guide 1.99 [1] for determining best estimate values of copper and nickel content of reactor vessel welds and base metal using this method.

The best estimate chemical content for welds is established from the mean of the measured values for weld samples made with the weld wire heat number that matches the critical vessel weld.

0.

C.

Cook Unit 1 Reactor Vessel weld is identical to Kewaunee and Maine Yankee vessel weld.

Kewaunee recently submitted a report to the Nuclear Regulatory Commission about their modified Cu and Ni contents of their critical vessel weld.

NRC evaluated and approved these revised values [3].

Since the Kewaunee vessel weld is identical to 0.

C.

Cook Unit 1 vessel

weld, the values (Cu, Ni and initial RTN0T) will be used for the calculation of RTN0TS Cu, Ni content and initial RTN0T for the 0; C ~

Cook capsule (weld) is different than the actual vessel weld.

These differences in material chemistry has been considered while developing chemistry factors using surveillance capsule data.

This is as per recommendation of Regulatory Guide 1.99, Revision 2.

All material properties for the plates in the belt line region identified in Table A-1 are plant specific.

0 I 11s/032190:10 A-2

TABLE A-1 0.

C.

UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Component Plate No.

Cu (Mt/.)

Ni Initial (Mt/.)

RTNDT 'F),

Intermediate Shell Plate B4406-1

,12

.52 Intermediate Shell Plate B4406-2

,15

.50 33 Intermediate Shell Plate B4406-3

.15

.49 40 Lower Shell Plate B4407-1

.14

.55 28 Lower Shell Plate B4407-2

.12

.59

-12 Lower Shell Plate Longitudina1 Hel ds Circumferential Hel ds B4407-3

.14

.50 28(a) 74(a)

~.28(')

.74(')

38

-56(')

(a)

Closure Head Flange Vessel Flange 60(b) 28(b)

(a)

Reference (3)

(b)

To be used for considering flange requirements for heatup/cooldown curves (5]

Illes/031290:10 A-3

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, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time; KIR is obtained -from-the

.reference fracture toughness curve, defined in Appendix G to the ASME Code The KIR curve is given by the following equation:

KIR 26 ~ 78 + 1 ~ 223 exp [0 ~ 0145 (T RTNpT + 160) ]

where KIR = reference stress intensity factor as a function of the metal temperature T'nd the metal reference nil-ductility temperature RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

IM IT IR (2) where KIM = stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RTN>T of the material C

= 2.0 for Level A and Level B service limits C

= 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical 411 lt/031200llO

'A"-4

At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference fracture toughness curve.

The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed, From equation 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 reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall.

Ouring 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 the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

Ouring cooldown, the I/O T vessel location is at a higher temperature than the fluid adjacent to the vessel IO.

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 operation.

Furthermore, if conditions exist so that the increase in KIR exceeds KIT, the calculated allowable pressure during cooldown wi 11 be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the I/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various ill 1 s/031290:10 A-5

intervals along a cooldown ramp.

The use of the composite curve eliminates this problem and ensures 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 wall that alleviate the tensile stresses produced by internal pressure.

The metal temperature at the crack tip lags the coolant temperature; therefore, the K<R for the 1/4 T crack during heatup is lower than the K<R for'the 1/4 T crack during steady-state conditions at the same time coolant temperature.

During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K>R's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a

lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered.

Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure cal'culated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the 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 therefore tend to reinforce any pressure stresses present.

These thermal stresses are dependent on both the rate of 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'omposite curve based on a point-by-point comparison of the steady-state and finite heatup rate data.

'At any given temperature, the 4111s/031190:10

'A-6

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 must at all times be based on analysis of the most critical criterion.

Finally, the 1983'mendment to 10CFR50 has a rule which addresses the metal temperature of the closure head flange and vessel flange regions.

This rule states that the metal temperature of the closure flange regions must exceed the material RTNpT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure.

Table A-1 indicates that, the limiting RTNpT of 60'F occurs in the closure head flange of D. C.

Cook Unit 1, so the minimum allowable temperature of this region is 180'F.

These limits are shown on Figures A-2 through A9, whenever applicable.

A-4.

HEATUP AND COOLOONN 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.

Figures A2 through A5 are applicable for the first 23 EFPY and Figures AG through A9 are applicable for the first 32 EFPY.

Instrumentation error margin of 10'F and 60 psig are included in developing heatup and cooldown curves for Figures A2, A3, A6 and A7.

No instrumentation error margins are considered for Figures A4, A5, A8 and A9.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in figures A-2 through A-9.'his is in addition to other criteria which must be met before the reactor is made critical.

l111g/031290:10 A-7

The leak limit curve shown in figure A-2, A-4, A-6, and A-8 represent minimum temperature requirements at the leak test pressure specified by applicable codes The leak test limit curve was determined by methods of references 2

[2,4) and 5.

Figures A-2 through A-9 define limits for ensuring prevention of nonductile failure for the D.

C.

Cook Unit 1 Primary Reactor Coolant System.

A-5.

AOJUSTEO REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev.

2 the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:

ART Initial RTNOT + hRTNOT + Margin (3)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-331 of Section III of the ASHE Boiler and Pressure Vessel Code.

If measured values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establisg a mean and standard deviation for the class.

BRTNOT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

CF~f(0.28-0.10 log f)

(4)

NDT f

= Neutron fluence, n/cm (E >

1 MeV), divided by 10 CF

=

Chemistry factor 'from tables for welds and for base metal (plates and forgings) (if no data use 0.35%

Cu and 1.0% Ni) il1 1 s/031290.IO A-8

Margin

= 2 [oI

+ a< ] 'here, 2

2 0.5 (5) aI standard deviation of initial RTNDT. If the initial RTNpz i s measured, aI is to be estimated from the precision of the test method; otherwise, oI is obtained from the same set of data that is used to get initial RTNpT o<

=

Standard deviation of LRTNDT, 28'F for welds and 17'F for base metal

[a< need not exceed 1/2 times aRTNpT]

To calculate LRTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

(

~ 24x

)

(depth X) surface (5) where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface.

The resultant fluence is then put into equation (4)

'o calculate hRTN>T at the specific depth.

CF ('F) is the chemistry factor, obtained using the methods given in reference 1.

The Adjusted Reference Temperatures (ART) for all belt line region materials for 0.

C.

Cook Unit 1 are provided in Table A-2 for 23 and 32 EFPYS.

Samples of calculations for ART are shown in Tables A-3 and A-4.

From Table A-2 the limiting material is found to be the intermediate plate B4406-3.

4111'/031200:10 A-9

TABLE A-2

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURES (ART)

AT 1/4T and 3/4T LOCATION Component 23 EFPY RTNDT At 32 EFPY RT NDl Intermediate Plate, B4406-1 1/4T

'F 110

~3/4T

'F

~1/4T

'F

~3/4T

'F 88 116 94 Intermediate

Plate, B4406-2 159 131 167 138 Intermediate

Plate, B4406-3 165(161)'

137(131)*

173(171)**

145(138)""

Lower Plate, B4407-1 Lower Plate, B4407-2 Lower Plate, B4407-3 148 94 156 121 72 130I'56 101 164 129 TS 137 Longi tudinal Weld Circumferential Weld 169(130) 192(149) 117(83) 186(144)

'31(95) 136(100) 209(164) 151(114)

Number within (

) are using chemistry factor based on surveillance capsule data.

These RTNDT numbers used to generate heatup and cooldown curves applicable up to 23 EFPY

• " These RTNDT numbers used to generate heatup and cooldown curves applicable up to 32 EFPY 4111s/032700:10 A-10

TABLE A-3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR D.

C.

COOK UNIT 1 REACTOR VESSEL MATERIAL-CIRCUMFERENTIAL HELD Parameter Re viator Guide 1.99 - Revision 2

Chemistry Factor, CF ('F)

Fluence, f (10 n/cm

)

Fluence Factor, ff 208.7 (184.7) 208.7 (184.7) 208.7 (184.7) 208.7 (184.7)

.634

.232

.852

.311

.872

.605

.955

.679 dRTNDT -

CF x ff ('F)

Initial RTNDT, I ('F)

Margin, M ('F) 182.1 (161.2) 126.2 (111.8) 199.3 (176.4) 141.8 (125.5)

-56

-56

-56

-56 65.5 (44) 65.5 (44) 65.5 (44) 65.5 (44)

Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 192 (149)

ART = Initial RTNDT + ARTNDT + Margin 136 (100) 209 (164) 151 (114)

(a)

Fluence, f, is based upon f f (10 n/cm, E>1 MeV)

= 1.05 at 23

EFPY, and 1.41 at 32 EFPY [Table 6-14 of the text].

The D.

C.

Cook Unit 1 reactor vessel wall thickness is 8.5 inches at the bel tl inc region.

(b)

The initial RTNDT (I) value for the weld is a generic value (from Table A-l).

(c)

Margin is calculated as, M

= 2 [o1

+ o< ] '

The standard deviation 2

2 0.5 for the initial RTNDT margin term (oI) is assumed to be 17'F since the initial RTNDT i s a generic mean val ue.

The standard devi ation for hRTNDT, (aa) i s 28'F for the weld.

o is 14'F for weld (cut into half) when surveillance data

~

~

is used.

(

) numbers in parenthesis were calculated using surveillance capsule data.

4111'/032790:10 A-11

'ABLE A-4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR D.

C.

COOK UNIT 1 REACTOR VESSEL MATERIAL "

INTERMEDIATE PLATE, B4406-3 Parameter Re ulator Guide 1.99 - Revision 2

Chemistry Factor, CF ('F)

Fluence, f (10 n/cm

)

Fluence Factor, ff 104 (119.55) 104 (119,55) 104 (119.55) 104 (119,55)

.634

.236.

.852

.311

.872

.609

.955

.679 hRTNDT CF x ff ( F)

Initial RTNDT

. I ( F)

Margin, M ('F) 90.7 (104.3) 63.4 (72.8) 99.3 (114.2) 70.7 (81.2) 40 40 40 40 34 (17) 34 (17) 34 (17) 34 (17)

Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 165 (161)

ART = Initial RTNDT + ERTNDT + Margin 137 (131) 173 (171) 145 (138)

(a)

Fluence, f, is based upon f f (10 n/cm, E>1 MeV)

= 1.05 at 23 EFPY, 19 2

and 1.41 at 32 EFPY [Table 6-14 of the text).

The D.

C.

Cook Unit 1 reactor vessel wall thickness is 8.5 inches at the beltline region.

(b)

The initial RTNDT (I) value for the plate is a measured value (from Table A-1).

(c)

Margin is calculated as, M = 2 [a1

+ a< ] '

The standard deviation 2,2 0.5 for the initial RTNDT margin term (aI) is assumed to be O'F since the initial RTNDT is a measured value.

The standard deviation for LRTNDT, (oh) is 17'F for the plate.

a< is 8.5'F for the plate (cut into half) when surveillance data is used.

(

) numbers in parenthesis were calculated using surveillance capsule data.

411 1s/Oll7QO:10 A-12

2.0 1.6

.3 10 2

3 4

6 6

7 991 1010 3

1 6

d 7881 1010 Ruence, n/cm~ (6 > 1MoV) 2 3

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6 7991 10 ~

Figure A-1.

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A-15

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4111 cs011190'10 D.

C.

Cook Unit 1 Reactor Coolant System Cooldown Limitations Applicable for the First 32 EFPY (With Margins for Instrumentation Errors)

A-19

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I CRITICALITY LINIT BASEO ON INSERVICE HYOROSTATIC TEST TEMPERATURE (304 F)

FOR THE SERVICE PERIOO UP TO 32 EFPY I

~

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Figure A-8. D.

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Cook Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 32 EFPY (Mithout Margins for Instrumentation Errors) 41111/011790 10 A-20

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:

INITIAL RTNDT'TNDT AFTER 32 EFPY.

INTERMEDIATE PLATE, 84406-3 40'F 1/4T, 171'F 3/4T, 138'F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100'F/HR FOR'HE SERVICE PERIOD UP TO 32 EFPY.

CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS.

2500 4

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COOLOOMN RATES

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'50 Figure A-9.

D.

C.

Cook Unit 1 Reactor Coolant System Cooldown'imitations Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors) 41tlstOltISO lO A-21

A-6.

REFERENCES

[1] Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988.

[2] "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-800, 1981.

[3] Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to Fast Neutron Fluence for Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 10CFR 50.61, Wisconsin Public Service Corporation Kewaunee Nuclear Power Station Dockett No. 50--305.

[4] ASME Boiler and Pressure Vessel Code,Section III, Division 1-Appendixes, "Rules for Construction of Nuclear Power Plan Components, Appendix G, Protection Against Nonducti le Failure," pp.

558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986.

[5] Code of Federal Regulations,

10CFR50, Appendix G, "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C.,

Federal

Register, Vol. 48 No. 104, May 27, 1983.

4111s/031tQO:10 A-22

A-7.

DATA POINTS FOR HEATUP AND COOLDOWN CURVES

~ 11 Is/03 l?QO'.10 A-23

Data Points for Heatup and Cooldown Curves for up to 23 EFPY and Mith Hargins for Instrumentation Error 4111'/0312SO:10 A-24

AEP CDDLDDWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DATA WERE PLOTTED FOR CDDLDDWN PROFILE 1

(

STEADY"STATE CODLDDWN

)

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAW DEPTH K ADWIN T 01/12/90 INDICATED TEMPERATURE (DEG.F)

INDICATED PRESSURE (PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 2 3

4 6

6 7

8 9

10 11 12 13 14 16 16 17 86.000 90.000 96.000 100.000 106.000 1 10. 000 116.000 120. 000 126. 000 130. 000 136. 000 140. 000 146. 000 150. 000 156.000 160. 000 166. 000 464.83 459.49 464.3$

469.65 475.32 481. 30 487.86 494.89 502.46 510. 59 619. 34 528.62 638.72 549.59 661. 28 573.70 587.20 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 170. 000 175. 000 180. 000 185. 000 190. 000 1S5. 000 200. 000 205. 000 210. 000 215.000 220. 000 225. 000 230. 000 235. 000 240. 000 245. 000 250. 000 601. 72 617. 30 68~

'51. 99 671. 20 692.05 714.40 738.34 764. 15 791. 84 821.50 853.59 887.S2 924.79 964.38 1006. 91 1052.62 35 36 C Z1+

37 38 39 40 41 42 43 44 46 46 47 48 49 50 256. 000 260. 000 266. 000 270. 000 275. 000 280. 000 285. 000 290. 000 295. 000 300. 000 305. 000 310.000 316.000 320. 000 326. 000 330. 000 1101.74 1154.52 1211. 18 1271.92 1336.96 1407. 01 1481.90 1562.30 1648.49 1740.74 1839.56 1945.08 2068. 16 2178. 95 2308.09 2445.60 F L ~ u< E, R.B P lj IRWIN &IT

AEP COOLDDWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DATA WERE PLOTTED FOR CODLDOWN PROFILE 2

(20 DEG"F / HR COOLDDWN

)

IRRADIATION PERIOD 23.000 EFP YEARS FLAW DEPTH

~ AOWIN T 01/12/90 INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

9 10 12 86.000

90. 000 96.000 100. 000 106.000 1 10. 000 116.000 120. 000 126. 000 130. 000 135. 000 140. 000 412. 87. ~

417.65 422.71 428. 15 433.86 440.27 447. 10 454.45 462.38 470. 91 480.01 489.90 13 14 15 16 17 18 19 20 21 22 23 145.000 150. 000 155. 000 160. 000 165. 000 170. 000 175.000 180. 000 185. 000 190. 000 195.000 500. 58 512.05 524.41 537. 60 651.94 567. 34 583.82 601. 68 620.89 641. 43 663.67 24 25 26 27.

28 29 30 31 32 33 34 200. 000 205. 000 2 10. 000 215.000 220.000 225. 000 230.000 235. 000 240. 000 245. 000 250.000 687.46 713. 19 740.76 770.35 802.38 836.69 873.50 913. 11 955.68 1001. 47 1050.68 r ~

q ~,

AEP CODLDDWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN 01/12/90 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 3

(40 DEG"F / HR COOLDOWN

)

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAW DEPTH

= AOWIN T INDICATED INDICATEQ TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

8 10 11 12

85. 000

90. 000

95. 000 100.000 105. 000 1 10. 000 116.000 120. 000 126. 000 130.000 135. 000 140. 000 370.

12'74.94 380. 12 385.76 391.88 398.49 405.65 413. 35 421.68 430.66 440.28 450.74 13 14 15 16 17 18 19 20 21 22 23 145. 000 150. 000 155. 000 160.000 165. 000 170. 000 175. 000 180. 000 185. 000 190. 000 1SS. 000 462.04 474.20 487.22 501. 36 516.62 532. 91 550.63 569.67 590. 10 612. 20 635.90 24 25 26 27 28 29 30 31 32 33 34 200. 000 205. 000 210. 000 215.000 220. 000 225. 000 230.000 235. 000 240. 000 245. 000 250. 000 661. 64 689.02 718. 54 750.57 784.82 821.68 861. 31 904. 19 960.09 999.46 1052.36 I

PO

AEP CDOLDDWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DATA WERE PLOTTED FOR CODLDOWN PRDFII.E 4

(60 DEG"F / HR CODLDOWN

)

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAW DEPTH

~ ADWIN T 01/ 12/90 INDICATED INDI GATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

{PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 2 3

4 6

6 7

8 9

10 11 86.000

90. 000

86. 000 100. 000 105.000 I 10. 000 116.000 120. 000 125. 000

$ 30. 000

$ 36. 000 326.8$ a-'-.

331.3$

336.77

'42.66 349.01 355.93 363.44 37$.55 380.27 389.76 400.04 12 13 14 15 16 17 18

$ 9 20 21 22 140. 000 145. 000 150. 000 155. 000 160. 000 165.000 170.000 175.000 180.000 185.000 190.000 411. 11 423. 11 435.95 449.93 464.99 481. 17 498.71 5$ 7.66 537.93 559.96 683.63 23 24 26 26 27 28 28 30 31 32 33 195.000 200. 000 205. 000 2 10. 000 2 $ 5.000 220. 000 226.000 230. 000 236. 000 240.000 246. 000 609. 13 636.52 666.2$

698.04 732.33 769.21 809. 13 851. 94 898.03 947.58 1000.73

AEP COOLDOWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DATA WERE PLOTTED FOR COOLOOWN PROFILE 5

(

100 DEG-F/HR CODLDOVN

=-

)

IRRADIATION PERIOD 23.000 EFP YEARS FLAW DEPTH

~ AOWIN T 01/12/90 1

2 3

4 5

6 7

8 9

10 11 INDICATEO TEMPERATURE (DEG.F) 85.000

90. 000 85.000 100. 000 105.000 1 10. 000 115.000 120. 000 125.000 130. 000 135.'000 INDICATED PRESSURE (PSI) 235.61,.

241.96 247.04' 253.53 260.63 268.29 276.68 285.77 285.67 306.33 317.87 12

'13 14 15 16 17 18 19 20 21 22 INDICATED TEMPERATURE (DEG.F) 140. 000 145.000 150. 000 155.000 160. 000 165. 000 170. 000 175. 000 180. 000 185. 000 190. 000 INDICATED PRESSURE (PSI) 330.55 344.21 358.91 374.91 392.11 410.83 431.02 452.79 476.37 501. 77 529. 14 23 24 25 26 27 28 28 30 31 32 195. 000 200.000 205. 000 210. 000 215.000 220. 000 226. 000 230. 000 235.000 240. 000 558.86 590.75 625. 18 662.44 702.54 745.69 792.25 842.35 896.34 954.27 INDICATEQ INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

AEP60 F/HR HEATUP CURVE REG.

GUIDE 1.99,REV.2 WITH MARGIN 01/12/90 THE FOLLOWING DATA WERE CALCULATEOFOR THE INSERVICE HYDROSTATIC LEAK TEST.

MINIMUM INSERVICE LEAK TEST TEMPERATURE ( 23.000 EFPY)

PRESSURE (PS I )

TEMPERATURE (OEG.F) 286 2485 307 PRESSURE (PSI)

PRESSURE STRESS (PSI) 1.5 K1M (PSI SQ.RT. IN. )

22088 92529 2485 27288 115366

AEP60 F/HR HEATUP CURVE REG.

GUIDE 1.99,REV.2 WITH MARGIN 01/12/90 COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAW DEPTH

~ (1-AOWIN)T HEATUP RATE(S) (DEG-F/HR)

~

60.0 INDICATED TEMPERATURE (DEG.F)

INDICATED PRESSURE (PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (OEG.F)

(PSI)

INDICATED TEMPERATURE (DEG.F)

INDICATED PRESSURE (PSI) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 85.000 90.000 95.000 100.000 105.000 110.000 1 15.000 120. 000 125. 000 130. 000 135. 000 140.000 145.000 150. 000 155.000 160. 000 165.000 170. 000 444ALL':"

449-.5e (23 IS 4QV-.&5 423. 15 423.57 425.'71 429.21 434.03 440. 10 447.42 455.83 465. 41 476.08 487.83 500.82 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 175. 000 180. 000 185. 000 190. 000 195. 000 200. 000 205. 000 2 10. 000 215. 000 220. 000 225. 000 230. 000 235. 000 240. 000 245. 000 250. 000 255.000 515. 05 530.40 547.22 565. 41 584.96 606. 21 628.97 653.67 680.09 708.68 739.32 772.20 807.76 845.79 886.67 930.58 977.71 36 37 38 39 40 41 42 43'4 45 46 47 48 49 50 51 52 260.000 265. 000 270.000 275.000 280.000 285.000 290.000 295.000 300.000 305.000 3 10. 000 315. 000 320. 000 325. 000 330. 000 335. 000 340. 000 1028.28 1082.53 1140.71 1203. 10 1270.04 1341.72 1418.60 1500.87 1588.97 1683.27 1784.05 1885.89 1983.07 2086.75 2197. 61 2315. 77 2442. 19

Data Points for Heatup and Coo1down Curves for up to 23 EFPY and Without Margins for Instrumentation Error 411 Is/031290;10 A-32

AEP COOLODWN CURVES REG.

GUIDE 1.99,REV.2 01/12/90 THE FOLLOWING DATA WERE PLOTTED FOR CODLOOWN PROFILE 1

(

STEADY"STATE CODLDOWN

)

IRRADIATION PERIOD 23.000 EFP YEARS FLAW DEPTH

% AOWIN T INDICATEO INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATEO INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 85.000 90.000 95.000 100.000 105.000 1 10. 000 115.000 120. 000 125. 000 130. 000 135. 000 140. 000 145. 000 150. 000 155. 000 160. 000 524.39 529.65 535.32 541. 30 547.85 554.89 562.45 570.59 579.34 588.62 598.72 609.59 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 165. 000 170. 000 175.000 180. 000 185.000 190. 000 195.000 200. 000 205. 000 210.000 215.000 220. 000 225. 000 230. 000 235.000 240. 000 0

752.05 774.40 798.34 824. 15 851. 84 881. 50 913.59 947.92 984.79 1024.38 1066.91 1112.62 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 245. 000 250. 000 255.000 260. 000 265. 000 270. 000 275.000 280. 000 285. 000 290. 000 295. 000 300. 000 305. 000 310. 000 315.000 1161.74 1214. 52 1271. 18 1331.92 1396.95 1467.01 1541.90 1622.30 1708.49 1800.74 1899.56 2005.08 2118. 16 2238.95 2368.09

AEP COOLDOMN CURVES REG.

GUIDE 1.99,REV.2 THE FOLLDMING DATA MERE PLOTTED FOR CODLODMN PROFILE 2

(20 DEG"F / HR CODLOQMN

)

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAM DEPTH

~

AOMIN T 01/ 12/90 INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATEO INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

9 10 11 86.000 90.000 95.000 100.000 106.000 110.000 1 16. 000 120. 000 126. 000 130. 000 136.000 482.71 488. 15 493.96 500. 27 607. 10 514.45 622.38 530.91 640.01 549.SO 660.68 12 13 14 15 16 17 18 19 20

'1 22 140. 000 145. 000 160. 000 155. 000 160. 000 165. 000 170. 000 175. 000 180. 000 185. 000 1SO. 000 572.05 584.41 697.60 611.94 648-.88 6~F8 68EH68 723.67 747.46 23 24 25 26 27 28 28 30 31 32 196. 000 200. 000 206. 000 210. 000 216.000 220. 000 226. 000 230. 000 235. 000 240. 000 773. 19 800. 75 830.36 862.38 SS6.68 933.50 973. 11 1015. 68 1061. 47 1110.68

~ g I fg,g t=f A',N4jK QE'QLII<BlvfGA)T

AEP CODLDOWN CURVES REG.

GUIDE 1.99,REV.2 01/ 12/90 THE FOLLOWING DATA WERE PLOTTED FOR CODLOOWN PROFILE 3

(40 DEG"F / HR CODLDQWN

)

IRRADIATION PERIOD 23.000 EFP YEARS FLAW DEPTH

~ AOWIN T INDICATED INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

8 10 11 86.000

90. 000

86. 000 100.000 105. 000 1 10. 000 115.000 120.000 126. 000 130. 000 136. 000 440. 12.

445.76 461.8$

458.49 465.66 473.35 481.69 490.66 500.29 510. 74 622.04 12 13 14 15 16 17 18 19 20 21 22 140. 000 145.000 150. 000 155. 000 160. 000 165. 000 170. 000 175. 000 180. 000 185. 000 190. 000 534.20 547.22 561.36 576.62 592. 91 610. 63 23 24 25 26 27 28 8~7 29 66~[ F 30 31 695.90 32 721. 54 195. 000 200. 000 205. 000 2 10. 000 215.000 220. 000 225. 000 230. 000 235. 000 240. 000 749.02 778.54 810. 67 844.82 881. 68 921. 31 964. 19 1010. 09 1069.46 1112. 35 P I A.AgF QE'PI)1+6 tgC~J~

AEP COOLODMN CURVES REG.

GUIDE 1.99,REV.2 THE FOLLOWING DATA MERE PLOTTED FOR CODLDOMN PROFILE 4 (60 DEG"f / HR CODLDQWN

)

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAM DEPTH

~ ADWIN T 01/12/90 INDICATED INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED.

TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5 6

7 8

9 10 11

85. 000

90. 000 100. 000 105.000 1 10. 000 1 15. 000 120. 000 125.000 130. 000 135. 000 398.7.7~

~

402.66 409.Ch1 415.93 423.44 431. 55 440.27 449.76 460.04 471. 11 483. 11 12 13 14 15 16 17 18 19 20 21 140. 000 145. 000 150.000 155.000 160.000 165. 000 170.000 175. 000 180.000 185. 000 495.95 509.93 524.99 541. 17 558.71 577.66 597.93 619. 96 669. 13 22 23 24 25 26 27 28 29 30 31 190. 000 195. 000 200. 000 205.000 2 10. 000 215.000 220. 000 225. 000 230.000 235. 000 696.52 726. 21 758.04 792.33 829.21 869. 13 911.94 958.03 1007.58 1060.73

AEP COOLOOWN CURVES REG.

GUIDE 1.99,REV.2 01/12/90 THE FOLLOWING DATA WERE PLOTTED FOR CODLDOWN PROFILE 5

(

100 DEG-F/HR COOLOOWN

)

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAW DEPTH

~ AOWIN T INDICATED INDIGATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5 6

7 8

8 10 85.000 90.000 95.000 100.000 105.000 1 10. 000 115.000 120. 000 125. 000 130. 000 307.04>

313. 53 320.63 328.29 336.68 345.77 355.67 366.33 377.87 390.55 11 12 13 14 15 16 17 18 19 20 135. 000 140. 000 145. 000 150. 000 155. 000 160. 000 165. 000 170. 000 175.000 180. 000 404. 21 418.91 434.91 452. 11 470.83 491.02 512. 79 536.37 561.77 589. 14

'21 22 23 24 25 26 27 28 28 30 185. 000 190. 000 195. 000 200. 000 205. 000 210. 000

~215.000 220. 000 225.000 230. 000 618. 86 650.75 685. 18 722.44 762.54 805.69 852.25 902.35 956.34 1014. 27 b

AEP60 F/HR HEATUP CURVE REG.

GUIDE 1.99,REV.2 01/12/90 THE FOLLOWING DATA WERE CALCULATEOFOR THE INSERVICE HYDROSTATIC LEAK TEST.

MINIMUM INSERVICE LEAK TEST TEMPERATURE ( 23.000 EFPY)

PRESSURE (PSI)

TEMPERATURE (DEG.F) 273 2485 294 PRESSURE (PSI)

PRESSURE STRESS (PSI) 1.5 K1M (PSI SQ.RT.IN.)

21444 89745 2485 26645 1 12505

AEP60 F/HR HEATUP CURVE REG.

GUIDE 1.99,REV.2 COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2

IRRADIATION PERIOD

~

23.000 EFP YEARS FLAM DEPTH i (1-AGAIN)T HEATUP RATE(S) (DEG.F/HR)

~

60.0 01/12/90 INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

8 10 12 13 14 16 16 17

86. 000

90. 000 S6. 000 100. 000 106. 000 1 10. 000 1 16. 000 120. 000 126. 000 130. 000 136.000 140. 000 145. 000 150. 000 156. 000 160. 000 166. 000 SO~;

49~

4~

48+~

483. 15 483.57 486.71 489. 21 484.03 500. 10 507.42 515. 83 526. 41 536.08 547.83 560.82 675.05 18

<83.)5 20 21 22 23 24 25 26 27 28 29 30 31 32 33 170. 000 175. 000 180. 000 185. 000 190. 000 195. 000 200. 000 205. 000 210. 000 2 15. 000 220. 000 225. 000 230.000 235.000 240.000 245. 000 590.40 607.22 6&5~

644.96 666. 21 688.97 713.67 740.09 768.68 799.32 832.20 867.76 905.79 946.67 990.58 1037. 71 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 250. 000 255.000 260. 000 265. 000 270. 000 275.000 280.000 285.000 2SO.OOO 295.000 300.000 305.000 3'10.000 315.000 320.000 325. 000 1088.28 1142. 53 1200.71 1263. 10 1330.04 1401. 72 1478.60 1560.87 1648.97 1743.27 1844.05 1945.89 2043.07 2146. 75 2267.61 2375.77 k 52 fS<

F i A.4(~

gS 4U1~~146~-)T

Data Points for Heatup and Cooldown Curves for up to 32 EFPY and With Margins for Instrumentation Error 41 1 1 s /031 290: 10 4

A-40

AEP COOLDOWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN 01/12/90 THE FOLLOMING DATA WERE PLOTTED FOR CODI.DOWN PROFILE 1

(

STEADY STATE CODLDOWN

)

IRRADIATION PERIOD

~

32.000 EFP YEARS FLAM DEPTH

~ AOMIN T INDICATEO INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDIGATED INDICATED TEMPERATURE PRESSURE (OEG.F)

(PSI)

INDICATEO INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

9 10 11 12 13 14

'16 16 17 18 86.000

90. 000 96.000 100. 000 106. 000 110.000 115.000 120. 000 126. 000 130.000 136.000 140. 000 146. 000 150. 000 156. 000 160. 000 166. 000 170. 000 446. 26..

450.69 464.93 459.49 464.39 469.65 475.32 481. 30 487.86 494.89 602.46 510. 59 519. 34 528.62 538.72 549.59 561. 28 573.70 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 175.000 180. 000 185. 000 190. 000 195.000 200.000 205. 000 2 10. 000 215.000 220. 000 225.000 230. 000 235. 000 240. 000 245.000 250. 000 255. 000 587.20 601. 72 617. 30 633.95 651.99 67 1,.20 692.05 714. 40 738.34 764. 15 791.84 821.50 853.69 887.92 924.79 964.38 1006. 91 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 260. 000 265: 000 270. 000 275. 000 280. 000 285. 000 290. 000 295.000 300. 000 305.000 310. 000 315.000 320. 000 325. 000 330. 000 335. 000 340. 000 1052.62 1101. 74 1164.52 1211. 18 1271. 92 1336.95 1407.01 1481. 90 1562.30 1648.49 1740.74 1839.56 1945.08 2058. 16 2178.96 2308.09 2446.60

AEP COOLOOWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DATA'WERE PLOTTED FOR CODLDOWN PROFILE 2

(20 OEG"F / HR CODLOOWN

)

IRRADIATION PERIOD

~

32.000 EFP YEARS FLAW DEPTH

> AOWIN T 01/12/90 INDICATED INDIGATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATEO INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5 6

7 8

9

'10 11.

12

85. 000

90. 000

95. 000 100. 000 105.000 1 10. 000 1 15. 000 120. 000 125. 000 130. 000 135.000 140. 000 404 22ii 408.24 412.58-417. 27 422.34 427.78 433.58 439.90 446.74 454.09 462.02 470.55 13 14 15 16 17 18 19 20 21 22 23 24 145. 000 150. 000 155. 000 160. 000 170. OC 175. 000 180. 000 185. 000 190.000 195. 000 200. 000 479.66 489.55 500.23 511.71 524.08 537.27 551.62 567.02 583. 51 601. 38 620.60 641. 15 25 26 27 28 29 30 31 32 33 34 35 36 205. 000 2 10. 000 215.000 220. 000 225. 000 230. 000 235. 000 240. 000 245. 000 250. 000 255.000 260. 000 663.40 687.20 712. 94 740.51 770. 13 802. 17 836.50 873.33 912.96 955.54 1001. 36 1050.59

AEP COOLDOWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN 01/12/90 THE FOLLOWING DATA MERE PLOTTED FOR CODLDOMN PROFILE 3

(40 DEG"F / HR CODLDOWN

)

IRRADIATION PERIOD

~

32.000 EFP YEARS FLAM DEPTH

~ AOWIN T INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5 6

7 8

9 10 11 12 85.000 90.000 95.000 100.000 1Oe.OOO 110.000 115.000 120.000 125.000 130.000 135.000 140.000 360.75-364.87 36$.35 374. 18 379.36 385.01 381. 15 397.75 404. 91 412.62 420.96 429.95 13 14 15 16 17 18 19 20 21 22 23 24 145. 000 150. 000 155. 000 160. 000 165. 000 170. 000 175. 000 180. 000 185.000 190. 000 195.000 200. 000 439.58 450.04 461. 34 473.51 486.55 500.70 515.97 532.28 550.01 569.08 589.53 611.66 25 26 27 28 29 30 31 32 33 34 35 36 205. 000 2 10. 000 215.000 220.000 225.000 230. 000 235. 000 240. 000 245. 000 250. 000 255.000 260. 000 635.37 661.03 688.53 718.09 750. 15 784.43 821. 33 860.99 903.92 949.86 999.29 1052.23

AEP COOLDDMN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DATA MERE PLOTTED FOR CODLDOMN PROFILE 4

(60 QEG"F / HR COQLDOVN

)

IRRADIATION PERIOD 32.000 EFP YEARS FLAM DEPTH

~ ADWIN T 01/12/90 INDICATED TEMPERATURE (DEG.F)

INDICATED PRESSURE (PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

9 10 11 12 86.000 90.000 85.000 100.000 106.000 110.000 116.000 120. 000 126. 000 130. 000 136. 000 140. 000 316. 28<,

320.52 326. 16.

330. 16 336.62 341.52 347.88 354.80 362.32 370.44 379. 18 388.67 13 14 15 16 17 18 19 20 21 22 23 24 145. 000 150. 000 155.000 160. 000 165. 000 170.000 175.000 180. 000 185. 000 190. 000 1SS. 000 200. 000 398.96 410.06 422.07 434.92 448.92 464.01 480.21

'97.77 516. 74 537.05 659.11 582.72 25 26 27 28 29 30 31 32 33 34 36 206. 000 210. 000 216.000 220. 000 226. 000 230. 000 235.000 240. 000 246. 000 250. 000 255. 000 608.35 635.78 665.51 697.39 731. 73 768.67 808.64 851. 52 897.68 947.30 1000. 75

AEP COOLDDWN CURVES REG.

GUIDE 1.99,REV.2 WITH MARGIN THE FOLLOWING DATA WERE PLOTTED FOR CODLOOWN PROFILE 5

(

100 DEG-F/HR CODLOOWN

)

'IRRADIATION PERIOD>>

32.000 EFP YEARS FLAW DEPTH

~

AOWIN T 01/ 12/90 INDICATEO INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED.

TEMPERATURE PRESSURE (OEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5 6

7 8

9 10 11 12 85.000 90.000 95.000 100.000 105.000 110.000 115.000 120.000 125.000 130.000 135.000 140.000 224.03vÃg.'. ~

228.59 233.62 239.08 245.08 251. 58 258.70 266.37 274.79 283.89 293.82 304.50 13 14 15 16 17 18 19 20 21 22 23 145. 000 150. 000 155. 000 160. 000 165. 000 170. 000 175.000 180. 000 185. 000 190. 000 195.000 316. 17 328.78 342.48 357. 21 373.26 390.50 409.27 429. 51 451. 34 474.98 500.44 24 25 26 27 28 29 30 31 32 33 34 200.000 205. 000 210.000 215. 000 220. 000 225. 000 230. 000 235. 000 240. 000 245. 000 250. 000 527.88 557.68 589.65 624. 18 661. 54 701.73 744.99 791.66 841.89 896.02 954.08

AEP60 F/HR HEATUP CURVE REG.

GUIOE 1.99,REV.2 WITH MARGIN 01/12/90 THE FOLLOWING DATA WERE CALCULATEOFOR THE INSERVICE HYDROSTATIC LEAK TEST.

MINIMUM INSERVICE LEAK TEST TEMPERATURES ( 32.000 EFPY) ie>'~..i~;.

PRESSURE (PSI )

TEMPERATURE {OEG. F )

\\

296 2485 317 PRESSURE (PSI )

PRESSURE STRESS 1.5 K1M (PSI)

{PSI SQ.RT.IN.)

2486 22088 27288 92529 115366

AEP60 F/HR HEATUP CURVE REG.

GUIDE 1.99,REV.2 WITH MARGIN 01/12/90 COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2

IRRADIATION PERIOD 32.000 EFP YEARS FLAW DEPTH

~ (1-AOWIN)T HEATUP RATE(S) (DEG.F/HR) 60.0 INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (OEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16

$ 7

$ 8 85.000 90.000 85.000 100.000

$05.000 110.000 115.000 120. 000 125. 000 130. 000 135. 000 140. 000 145. 000

$ 50. 000 155. 000 160. 000 165. 000 170. 000 l~

44&+4 4&tH@

4~

44 4 $ 2.80 414.3$

417. 11 42$.25 426.49 432.80 440.22 448.72 458.23 468.82 480.35 1;.":c 19 20 q)8,8o 23 24 25 26 27 28 29 30 31 32 33 34 35 36 175. 000

$ 80. 000

$85.000 190. 000 195. 000 200.000 205. 000 210. 000 215.000 220. 000 225. 000 230.000 235. 000 240. 000 245. 000 250. 000 255. 000 260. 000 493. 11 507.00 522. 12 538.37 556. 10 575.09 595.76 617. 99 641.85 667. 61 695.27 724.95 757.06 791.42 828.30 867.91 910. 47 956. 17 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 265.000 270. 000 275. 000 280. 000 285.000 290. 000 295. 000 300. 000 305.000 310. 000 3 $ 5.000 320. 000 325. 000 330. 000 335. 000 340. 000 345. 000 350. 000 1005.26 1057.96 1 $ 14. 56 1175. 14 1239.9$

1309.73 1384. 19 1464. 20 1549.80 1630.38

$ 709.42 1794.44 1885.24 1982.27 2085.80 2196. 69 2314.48 2440.72 I

'V

Oata Points for Heatup and Cooldown Curves for up to 32 EFPY and Hithout Margins for Instrumentation Error ilI 1 s/031290:10 A-48

AEP CODLDDWN CURVES REG.

GUIDE 1.99,REV.2 01/12/90 THE FOLLOWING DATA WERE PLOTTED FOR CODLDOWN PROFILE 1

(

STEADY"STATE CODLDOWN

)

IRRADIATION PERIOD 32.000 EFP YEARS FLAW DEPTH

~

AOWIN T

'1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 85.000

90. 000 100. 000 105. 000 1 10. 000 115.000 120. 000 125.000 130. 000 135. 000 140. 000 145. 000 150. 000 155. 000 160. 000 165. 000 514. 83

.'19.

49 524.39.

529.65 535.32 541. 30 547.85 554.89 562.45 570.59 579.34 588.62 598.72 609.59 INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 INDICATED TEMPERATURE (DEG.F) 170. 000 175.000 180. 000 185. 000 190. 000 195. 000 200. 000 205. 000

'10.000 215.000 220. 000 225. 000 230. 000 235. 000 240. 000 245. 000 INDICATED PRESSURE (PSI) ee~~ j 711.99 731.20 752.05 774.40 798.34 824. 15 851.84 881. 50 913.59 947.92 984.79 1024.38 1066. 91 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 250. 000 255. 000 260. 000 265. 000 270. 000 275. 000 280. 000 285. 000 290. 000 295.000 300. 000 305. 000 3 10. 000 315.000 320. 000 325. 000 1112.62

'I 161. 74 1214. 52 1271. 18 1331. 92 1396.95 1467.01 1541.90 1622.30 1708.49 1800.74 1899.56 2005.08 2118. 16 2238.95 2368.09 INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

AEP COOLOOWN CURVES REG.

GUIDE 1.99,REV.2 THE FOLL0$$ ING DATA WERE PLOTTED FOR CODLO0$$N PROFILE 2

(20 DEG-F / HR CODLDOVN

)

IRRADIATION PERIOD

~

32.000 EFP YEARS FLAM DEPTH

~ AOMIN T 01/12/90 2

3 4

5 6

7 8

8 10 11 12

85. 000

90. 000

95. 000 100. 000 105.000 1 10. 000 1 $5.000 120. 000 125. 000 130. 000 135.000 140. 000 472.59 477.27 482.34 487.78 483.58 499.90 506.74 514.09 522.02 530.55 538.66 549.55 INDICATED INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI) 13 14 15 16 17 18 19 20 21 22 23 INDICATED TEMPERATURE (OEG.F) 145.000 150.000 155.000 160.000 165.000 170.000~

175.000

$80.000 185.000 190.000 195.000 INDICATED PRESSURE (Psr) 560.23 24 571.71 25 584.08 26 597.27 27 61 $.62 28 680.60 32 701. 15 33 723.40 34 INDICATEO TEMPERATURE (DEG.F) 200. 000 205. 000 210. 000 2 15. 000 220. 000 225.000 230.000 235. 000 240. 000 245. 000 250. 000 INDICATED PRESSURE (PSI) 747.20 772.94 800. 51 830. 13 862.$ 7 896.50 933.33 972.96 1015. 54 1061. 36 1110.59

EP COOLDOWN CURVES REG.

GUIDE 1.99,REV.2 01/12/90 THE FOLLOWING DATA WERE PLOTTED FOR CODLDOWN PROFILE 3

(40 OEG"F / HR CODLDOWN

)

IRRADIATION PERIOD 32.000 EFP YEARS FLAW DEPTH

% AOWIN T INDICATED INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

6 6

7 8

9 10 11 12 86.000

90. 000

86. 000 100. 000 106. 000 1 10. 000 1 16.000 120.000 126. 000 130. 000 136. 000 140. 000 429.36 434.18 439.36 445.01 461. 16 457.75 464.91 472.62 480.86 489.95 499.58 510. 04 13 14 15 16 17 18 19 20 21 22 23 145.000 150. 000 155.000 160.000 165.000 170.000 175.000 180.000 185.000 190.000 1S5.000 521.34 533.51 546.55 560.70 575.97 592.28 610.01

&L~

649.53 671. 66 695.37 24 25 26 27 28 29 30 31 32 33 34 200.000 205. 000 210.000 2 15. 000 220.000 225.000 230.000 235.000 240.000 245.000 250.000 721.03 748.53 778.09 810. 15 844.43 881. 33 920.99 963.92 1009.86

. 1059. 29 1112. 23 FL A,~ +E, gQ QUI+QQEA)T I

CJl

AEP COOLDOWN CURVES REG.

GUIDE 1.99,REV.2 01/12/90 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 4 (60 OEG"F / HR COOLOOWN

)

IRRADIATION PERIOD 32.000 EFP YEARS FLAW DEPTH AOWIN T INDICATEO INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (OEG.F)

(PSI)

INDIGATED INDIGATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 2 3

4 6

7 8

9 10 11

86. 000

90. 000 96.000 100. 000 105.000 1 10. 000 1 15. 000 120. 000 126. 000 130. 000 135. 000 3as. sa==;.-:..

390. 16 395.62 401.52 407.88 414.80 422.32 430.44 439. 18 448.67 46a.aa 12 13 14 15 16 17 1a 19 20 21 22 140. 000.

145. 000 150. 000 155. 000 160. 000 165.000 170.000 175. 000 180. 000 185. 000 190.000 470.06 482.07 494.92 508.92 524.01 540. 21 557.77 576.74 597.05 619. 11 642.72 23 24 26 26 27 28 29 30 31 32 33 195.000 200.000 205.000 2 10. 000 216.000 220. 000 225.000 230. 000 235. 000 240. 000 246.000 668.35 695.78 725.61 757.39 791.73 828.67 868.64 911. 52 967.68 1007.30 1060.76

AEP CODLDOWN CURVES REG.

GUIDE 1.99,REV.2 01/12/90 THE FOLLOWING DATA WERE PLOTTED FOR CODLDDWN PROFILE 5

(

100 DEG-F/HR CODLDOWN

)

IRRADIATION PERIOD

+

32.000 EFP YEARS FLAW DEPTH

~ AOWIN T INDICATED INDICATEO TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 1 2

3 4

5

~

6 7

8 9

10 11 85.000

90. 000 95.000 100. 000 105. 000 1 10.000 115.000 120. 000 125.000 130. 000 135. 000 293.62, ~.,

299.08 805.08.

~

311.58 318. 70 326.37 334.79 343.89 353.82 364.50 376. 17 12 13 14 15 16 17 18 19 20 21 22 140.000 145.000 150.000 155. 000 160. 000 165. 000 170.000 175. 000 180. 000 185. 000 190. 000 388.78 402.48 4'17.21 433.26 450.50 469.27 489. 51 511. 34 534.98 560.44 587.88 23 195.000 24 200. 000 25 205.000 26, 2 10. 000 27 215. 000 28 220. 000 29 225. 000 30 230. 000 31 235.000 32 240. 000 617.68 649.65 684. 18 721. 54 761. 73 804.99 851.66 901. 89 956.02 1014.08

AEP60 F/HR HEATUP CURVE REG.

GUIOE 1.99,REV.2 01/12/90 THE FOLLOWING DATA MERE CALCULATEOFOR THE INSERVICE HYOROSTATIC LEAK TEST.

MINIMUM INSERVICE LEAK TEST TEMPERATURE ( 32.000 EFPY)

PRESSURE (PSI)

TEMPERATURE (OEG

~ F) 283 2485 304 PRESSURE (PSI)

PRESSURE STRESS

~

(PSI) 21444 1.5 K1M (PSI SQ.RT. IN.)

89745 2485 26645 112505

AEP60 F/HR HEATUP CURVE REG.

GUIDE 1.99,REV.2 COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2

IRRADIATION PERIOD

~

32.000 EFP YEARS FLAW DEPTH

% (1-ADWIN)T HEATUP RATE(S) (DEG.F/HR) 60.0 01/12/90 INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI)

INDICATED TEMPERATURE (DEG.F)

INDICATED PRESSURE (PSI)

INDICATED INDICATED TEMPERATURE PRESSURE (DEG.F)

(PSI) 2 3

4 6

6 7

8 9

10 11 12 13 14 16 16 17

86. 000

90. 000

96. 000 100. 000 105. 000 1 10. 000 115.000 120. 000 126. 000 130. 000 136. 000 140.000 146. 000 150. 000 155.000 160. 000 166. 000 400e4%

472.80 474.31 477. 11 48$.25 486.49 492.80 500.22 508.72 518.23 528.82 540.35 653.11 18

<7~>'o 21 22 23 24 25 26 27 28 29 30 31 32 33 34 170. 000 175. 000 180. 000 185. 000 190. 000 195. 000 200. 000 205. 000 210.000 215.000 220. 000 225.000 230. 000 235. 000 240. 000 245. 000 250. 000 567.00 582. 12 598.37 616. 10 635.09 655.76 677.99 701. 85 727.61 755.27 784.95 817.06 851.42 888.30 927.91 970.47 1016. 17 35 36 37 38 39 40 41 42 43 44 46 46 47 48 49 50 51 256. 000 260. 000 265. 000 270. 000 275. 000 280. 000 286. 000 2SO. 000 295. 000 300. 000 305. 000 310. 000 316.000 320. 000 326. 000 330. 000 336.000 1065.26 1117.96 1174. 56 1235. 14 1299. 91 1369.73 1444. 19 1524.20 1609.80 1690.38 1769.42 1854.44 1946.24 2042.27 2146. 80 2256.69 2374.48

IaV

)