ML13311A234

Analysis of Capsule F,Reactor Vessel Radiation Surveillance Program

ML13311A234
Person / Time
Site:	San Onofre
Issue date:	05/31/1979
From:	Shaun Anderson, Kaiser W, Yanichko S Westinghouse, Div of CBS Corp
To:
Shared Package
ML13311A235	List: ML13311A234 ML13330A034 ML13330A035
References
TAC 55027, TAC 59980 NUDOCS 8004240397, WCAP-9520
Download: ML13311A234 (67)

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WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSULE F FROM THE SOUTHERN CALIFORNIA EDISON COMPANY SAN ONOFRE REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko S. L. Anderson W. T. Kaiser May 1979 APPROVED:V J.. Chirigos, Manager Structural Materials Engineering Prepared by Westinghouse for the Southern California Edison Company Work Performed Under EZGP 200 Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. 0. Box 355 Pittsburgh, Pennsylvania 15230 Soo 4203%

LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the San Onofre Reactor Vessel 4-2 4-2 Arrangement of Specimens, Thermal Monitors, and Dosimeters in Capsule F 4-5 5-1 Charpy V-Notch Impact Data for San Onofre Pressure Vessel Shell Plate W7601-8 5-3 5-2 Charpy V-Notch Impact Data for San Onofre Pressure Vessel Weld Metal 5-4 5-3 Charpy V-Notch Impact Data for San Onofre Pressure Vessel Weld Heat-Affected-Zone Metal 5-5 5-4 Charpy V-Notch Impact Data for A302 Grade B ASTM Correlation Monitor Material 5-6 5-5 Charpy Impact Specimen Fracture Surfaces for San Onofre Pressure Vessel Shell Plate W7601-8 5-10 5-6 Charpy Impact Specimen Fracture Surfaces for San Onofre Weld Metal 5-11 5-7 Charpy Impact Specimen Fracture Surfaces for San Onofre Weld Heat-Affected-Zone Metal 5-12 5-8 Charpy Impact Specimen Fracture Surfaces for A302 Grade B ASTM Correlation Monitor Material 5-13 5-9 Tensile Properties for San Onofre Pressure Vessel Shell Plate W7601-8 5-17 5-10 Tensile Properties for San Onofre Weld Metal 5-18 5-11 Fractured Tensile Specimens From San Onofre Shell Plate W7601-8 5-20 5-12 Fractured Tensile Specimens From San Onofre Weld Metal 5-21 5-13 Typical Stress-Strain Curve for Tension Specimens (Tension Specimen No. YB-6) 5-22 v

TABLE OF CONTENTS Section Title Page 1

SUMMARY

1-1 2

INTRODUCTION 2-1 3

BACKGROUND 3-1 4

DESCRIPTION OF PROGRAM 4-1 5

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

Test Procedure 5-1 5-2.

Charpy V-Notch Impact Test Results 5-2 5-3.

Tensile Test Results 5-15 5-4.

Wedge Opening Loading Tests 5-15 6

NEUTRON DOSIMETRY ANALYSIS 6-1 6-1.

Description of Neutron Flux Monitors 6-1 6-2.

Analytical Procedures 6-3 6-3.

Results of Analysis 6-6 Appendix A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1 A-1.

Introduction A-1 A-2.

Fracture Toughness Properties A-1 A-3.

Criteria for Allowable Pressure-Temperature Relationships A-6 A-4.

Heatup and Cooldown Limit Curves A-8 Ill

LIST OF ILLUSTRATIONS (cont)

Figure Title Page 6-1 Calculated Azimuthal Distribution on Neutron Flux (E > 1 Mev) at the Core Midplane of the San Onofre Reactor Vessel 6-14 6-2 Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Incident on the San Onofre Reactor Vessel 6-15 6-3 Calculated Maximum End-of-Life Fast Neutron Fluence (E > 1 Mev) as a Function of Radius Within the San Onofre Reactor Vessel 6-16 A-1 Effect of Fluence and Copper Content on ARTNDT for Reactor Vessel Steels Exposed to Irradiation at 5500 F A-2 A-2 Fast Neutron Fluence (E > 1 Mev) as a Function of Full-Power Service Life A-3 A-3 San Onofre Unit No. 1 Reactor Coolant System Heatup Limitations Applicable for the First 16 EFPY A-10 A-4 San Onofre Unit No. 1 Reactor Coolant System Cooldown Limitations Applicable for the First 16 EFPY A-11 vi

LIST OF TABLES Table Title Page 1-1 End-of-Life Projected Fast Neutron Fluences 1-2 4-1 Chemistry and Heat Treatment of Material Representing the Core Region Shell Plates and Weld Metal From the San Onofre Reactor Vessel 4-3 4-2 Chemistry and Heat Treatment of Surveillance Material Representing 6-Inch-Thick A302B ASTM Correlation Monitor Material 4-4 5-1 Charpy V-Notch Impact Data for San Onofre Pressure Vessel Material Irradiated in Capsule F at 550'F, Fluence 5.14 x 1019 n/cm 2 (E > 1 Mev) 5-7 5-2 Effect of 550 0 F Irradiation at 5.14 x 109 n/cm 2 (E > 1 Mev) on the Notch Toughness Properties of the San Onofre Reactor Vessel Surveillance Materials 5-9 5-3 Summary of San Onofre Reactor Vessel Surveillance Capsule Charpy Impact Test Results 5-14 5-4 Irradiated Tensile Properties for San Onofre Pressure Vessel Materials 5-16 5-5 Summary of Tensile Results for San Onofre Plate W7601-8 and Weld Metal 5-19 6-1 Neutron Flux Monitors Contained Within Capsule F 6-2 6-2 Irradiation History of Capsule 1-6-7 6-3 Spectrum-Averaged Reaction Cross Sections Used in Fast Neutron Flux Derivation 6-11 6-4 Results of Fast Neutron Dosimetry for Capsule F 6-12 6-5 Results of Thermal Neutron Dosimetry for Capsule F 6-13 6-6 Calculated Fast Neutron Flux and Lead Factors for Capsule F 6-15 6-7 Fast Neutron Exposure Derived From Calculated and Measured Results 6-17 A-1 Reactor Vessel Toughness Data (Unirradiated)

A-4 vii

SECTION 1

SUMMARY

The analysis which compared unirradiated with irradiated material properties of the reactor vessel material contained in the third surveillance capsule, designated F, from the Southern California Edison Company San Onofre reactor pressure vessel led to the following conclusions:

a The capsule received an average fast fluence of 5.14 x 1019 n/cm2 (E > 1 Mev). The predicted fast fluence for the capsule was 4.92 x 1019 n/cm 2 (E > 1 Mev).

E The fast fluence of 5.14 x 1019 n/cm2 resulted in a 1650 F increase in the 50 ft lb reference nil-ductility transition temperature (RTNDT) of the weld metal, which is representative of the most limiting material in the core region of the reactor vessel. The intermediate pressure vessel shell plate W7601-8 exhibited a 50 ft lb transition temperature increase of 110 0 F (specimens oriented parallel to the rolling direction of the plate). The weld heat-affected-zone material exhibited a 50 ft lb transition temperature increase of 1300 F. These transition temperature increases are significantly less than would be predicted using the methods of Regulatory Guide 1.99, and indicate a possible limiting or steady state condition may be occurring.

An increase of 1309F in the 30 ft lb transition temperature was determined for the ASTM A302B reference correlation monitor material contained in the capsule. The transition temperature increase for the reference material in this capsule and prior capsules is signifi cantly less than predicted from the ASTM trend curve developed for this material and indicates that a limiting or steady state condition has resulted for this material.

0 The end-of-life projected fast neutron fluences for the reactor vessel were determined, based on 27 full-power years of operation at 1347 Mw as derived from both calculated and measured surveillance capsule results, and are presented in table 1-1.

1-1

TABLE 1-1 END-OF-LIFE PROJECTED FAST NEUTRON FLUENCES Fast Neutron Fluence (n/cm2)

Vessel Location Calculated Measured Inner surface 1.0 x 1020 1.05 x 1020 1/4 thickness 4.2 x 1019 4.4 x 1019 3/4 thickness 7.1 x 1018 7.4 x 1018 1-2

SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule F, the third capsule of the continuing surveillance program, which monitors the effects of neutron irradiation on the Southern California Edison Company San Onofre reactor pressure vessel materials under actual operating conditions.

The surveillance program for the San Onofre 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.[1] The surveillance program, which was planned to cover the 30-year life of the reactor pressure vessel, was based on ASTM E-185-62.[ 2 Postirradiation data have been obtained from the third material surveillance capsule (Capsule F) removed from San Onofre reactor vessel. This report summarizes the tests and the results, and discusses the analysis of the data.

1. Yanichko, S. E., "San Onofre Reactor Vessel Radiation Surveillance Program," WCAP-2834-R1, November 1966.

2. ASTM Designation E185-62, "Surveillance Tests on Structural Materials in Nuclear Reactors," in ASTM Standards (1962),

Part 31, Physical and Mechanical Testing of Metals -

Metallography, Nondestructive Testing, Fatigue, Effect of Temperature, Am. Soc. for Testing and Materials, Philadelphia, PA, 1962.

2-1

I 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 irra diation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA302 Grade B (base material of the San Onofre reactor pressure vessel beltline) are well-documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

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

RTNDT is defined as the greater of the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 60'F less than the 50 ft lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal to the rolling direction of the material. The RTNDT of a given material is used to index that material to a reference stress-intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code.

The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress-intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined based on these allowable stress-intensity factors.

RTNDT and, in turn, the operating limits of nuclear power plants, can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the San Onofre reactor vessel radiation surveillance 3-1

program,1] in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the Charpy V-notch tem perature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the material to the KIR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effect of irradiation on the reactor vessel materials.

1. Yanichko, S. E., "San Onofre Reactor Vessel Radiation Surveillance Program," WCAP-2834-R1, November 1966.

3-2

SECTION 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the San Onofre reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The eight 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 F was removed in December of 1978 after approximately 10 calendar years (7.76 effective full-power years) of plant operation. This capsule contained Charpy V-notch impact, tensile, and WOL specimens (shown in WCAP-2834-R1)1 11 from the intermediate shell ring plates, weld metal representative of the core region of the reactor vessel, and Charpy V-notch specimens from weld heat-affected-zone (HAZ) material. The capsule also contained Charpy V-notch specimens from the 6-inch-thick ASTM correlation monitor material (A302 Grade B) furnished by the U. S. Steel Corporation. The chemistry and heat treatment of the surveillance material is presented in tables 4-1 and 4-2.

All test specimens were machined from the 1/4 thickness location of the plates. Test specimens represent material taken at least one plate thickness from the quenched end of the plate. All base metal Charpy V-notch and tensile specimens were oriented with the longitudinal axis of the specimen parallel to the principal rolling direction of the plates. The WOL test specimens were machined with the simulated crack of the specimen perpendicular to the surfaces and rolling direction of the plates.

Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction. Tensile specimens were oriented with the longitudinal axis of the specimen parallel to the weld.

Capsule F contained dosimeter wires of copper, nickel, and aluminum-cobalt (cadmium-shielded and unshielded). In addition, the capsule contained cadmium-shielded dosimeters of Np23 7 and U238, located as shown in figure 4-2.

1. Yanichko, S. E., "San Onofre Reactor Vessel Radiation Surveillance Program," WCAP-2834-R1, November 1966.

4-1

14,420-1 2700 D

REACTOR VESSEL E

THERMAL SH IELD 780 41

.710 19' 280 44 10 320 54' 1800 480 45' 41o I5' 280 44'.

C F

H B

G CAPSULE (TYP) 900 Figure 4-1. Arrangement of Surveillance Capsules in the San Onofre Reactor Vessel 4-2

TABLE 4-1 CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE CORE REGION SHELL PLATES AND WELD METAL FROM THE SAN ONOFRE REACTOR VESSEL Specimen Chemical Analysis (Percent)

Material Identification Data Sources C

Mn Si P

S Mo Cu V

A302B W7601-1 SwRI 1.45 0.47 0.17 0.02 Kawin 0.25 1.40 0.30 0.013 0.016 0.43 0.17 0.03 Lukens 0.22 1.36 0.24 0.013 0.025 0.46 A302B W7601-8 SwRI 1.39 0.47 0.18 0.02 Kawin 0.22 1.35 0.26 0.010 0.018 0.47 0.18 0.03 Lukens 0.20 1.34 0.20 0.012 0.020 0.47 A302B W7601-9 SwRI 1.50 0.48 0.18 0.02 Kawin 0.22 1.45 0.26 0.008 0.020 0.45 0.18 0.03 Lukens 0.19 1.36 0.23 0.014 0.026 0.47 Weld metal Specimen(al SwRI 1.26 0.43 0.19 0.04 Weld metal 2461 PC3[b]

Kawin 0.11 1.50 0.35 0.017 0.013 0.47 0.19 0.03

a. Irradiated C, specimen from Capsule "A".

b. Nozzle cutout.

HEAT TREATMENT Plate Material Heated at 1550-16000 -

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> -

dip quenched Tempered at 1225 0 F -

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> -

furnace cooled Stress relieved at 1150'F -

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -

furnace cooled Weld Metal Stress relieved at 11500 F -

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -

furnace cooled

TABLE 4-2 CHEMISTRY AND HEAT TREATMENT OF SURVEILLANCE MATERIAL REPRESENTING 6-INCH-THICK A302B ASTM CORRELATION MONITOR MATERIAL Chemical Analysis (Percent)

C Mn P

S Mo Si Cu Ni Cr 0.24 1.34 0.011 0.023 0.51 0.23 0.20 0.18 0.11 HEAT TREATMENT The 6-inch-thick plate was charged into a furnace operating at 1100'F, heated at a maximum rate of 630 F per hour to 16500'F, held at temperature for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and water-quenched to 3000 F. The plate was then recharged into a furnace operating at 7000 to 7500 F, heated at a maximum rate of 630 F per hour to 12000 F, and held at that temperature for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

4-4

5790 F 579 0F 590 0 F MONITOR 590 0 F MONITOR Cv TENSILE C

Cv Cv Cv TENSILE Cv H17 DI7 H19 D19 H20 D20 H21 D21 H22 D22 R64 YB YB D6 D5 YB YB YB YB YB7 YB6 R57 4 1 R59 43 R60.

44 R61 45 R62 46 H24 1,D24 Co

-7 Co Co(Cd)

Cu Co(Cd)

CV WOL WOL WOL WOL C,

Np & U H18 DI8 DOSIMETER H23 D23 D5 D6 BLOCK YBl8 YB21 R58 YB R63 YB 4*2 4*74 G~o Cu Ni Cu Co(Cd)

IIfI TO VESSEL BOTTOM VESSEL WALL SIDE.

YB= PLATE W7601-8 H= HAZ D= WELD METAL R= A302B REFERENCE MATERIAL Figure 4-2.

Arrangement of Specimens, Thermal Monitors, and Dosimeters in Capsule F

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

2.5% Ag, 97.5% Pb Melting point 5790 F 1.75% Ag, 0.75% Sn, 97.5% Pb Melting point 5900 F 4-6

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

TEST PROCEDURE The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H.

Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-2834-R1.[1 1 No discrepancies were found.

Examination of the two low-melting (5790 F and 5900 F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579 0 F.

A Tinius Olsen Model 74 impact test machine was used to test the irradiated Charpy V-notch specimens per ASTM E23-72, "Notched Bar Impact Testing of Metallic Materials." Before initiating tests on the irradiated Charpy-V specimens, the accuracy of the impact machine was checked with a set of standard specimens obtained from the Army Material and Mechanics Research Center in Watertown, Massachusetts. The results of the calibration.testing showed that the machine was certified for Charpy V-notch impact testing.

The tensile tests were conducted on a screw-driven Instron testing machine of 20,000 lb capacity per ASTM E8-69, "Tension Testing of Metallic Materials" and ASTM E21-70, "Elevated Temperature Tension Tests of Metallic Materials." The crosshead speed was 0.05 inch per minute. The deformation of the specimen was measured with a strain gage extensometer.

The extensometer was calibrated before testing with a Sheffield high magnification drum-type extensometer calibrator.

Elevated-temperature tensile tests were conducted in a split-tube furnace. The specimens were held at temperature a minimum of 20 minutes to stabilize the temperature prior to testing.

1. Yanichko, S. E., "San Onofre Reactor Vessel Radiation Surveillance Program," WCAP-2834-R1, November 1966.

5-1

Temperature was monitored with a chromel-alumel thermocouple in contact with the clevis pin-type upper and lower specimen grips. Temperature was controlled with plus or minus 3oF.

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

5-2.

CHARPY V-NOTCH IMPACT TEST RESULTS The irradiated Charpy V-notch specimens represented the San Onofre reactor pressure vessel beltline plate material (W7601-8), weld and heat-affected-zone (HAZ) material, and the ASTM reference correlation monitor material. The results are presented in figures 5-1 through 5-4 and table 5-1. Table 5-2 summarizes the increase in the 30 and 50 ft lb energy and 35-mil lateral expansion transition temperature, and the decrease in the upper shelf energy resulting from irradiation to 5.14 x 1019 n/cm2.

The test results obtained on shell plate W7601-8 shown in figure 5-1 and table 5-2 resulted in a 30 and 50 ft lb transition temperature increase of 120' F and 1100 F, respectively, and a decrease in upper shelf energy of 26 ft lb or 27 percent.

The results of tests on the weld metal are shown in figure 5-2 and table 5-2. These results show that a 145 0 F and a 165 0 F transition temperature increase was obtained at the 30 and 50 ft lb levels, respectively. The upper shelf energy decreased 19 ft lb or 19 percent.

Test results for the HAZ material are shown in figure 5-3 and table 5-2. A 30 and 50 ft lb transition temperature increase of 115 0F and 1300 F, respectively, resulted from irradiation.

The upper shelf of the HAZ material decreased 32 ft lb or 30 percent.

Figure 5-4 and table 5-2 present the test results obtained on the ASTM A302B reference correlation monitor material. Respective 30 and 50 ft lb transition temperature increases of 130 0 F and 135 0 F were obtained on this material. The upper shelf energy of the correlation monitor material decreased 16 ft lb or 12 percent.

Charpy impact specimen fracture surfaces of the various San Onofre vessel materials and the correlation monitor material are presented in figures 5-5 through 5-8.

Table 5-3 summarizes the Charpy impact test results for first and second capsules[1, 2] and the third capsule removed from the San Onofre reactor to date.

1. "Analysis of First Surveillance Material Capsule from San Onofre Unit 1," Southern California Edison Company, July 1971.

2. Norris, E. B., "Analysis of Second Surveillance Material Capsule from San Onofre Unit 1, "SWR1 Project No. 07-2892, June 5, 1972.

5-2

14,420-3 120 100 39 80 -

3 3

(100 80 C 60 LU 1000 0

0 IRRADIATED AT 550oF 20 2

5.14 X 1019 N/CM 2 03.:2 80 80 o

60 Lj--

40 IRRADIATED AT 550 0 F 20

-19 5.14 X 109 N/CM 0

120 100 80 UNIRRADIATED 60 I 100 40 IRRADIATED AT 550 0 F 24 Z

12 00 O O 2.36 X 1019 N/CM 20 -

2F 5.14 X 1019 N/CM2 20 0

I I

I I1

-I00 0

100 200 300 400 500 TEMPERATURE (OF)

Figure 5-1.

Charpy V-Notch Impact Data for San Onofre Pressure Vessel Shell Plate W7601-8 5-3

14,420-4 120 I

100 9

• • 3 80 2

U 22 60 1300

-J IRRADIATED AT 550oF 20 2

5.14 X 1019 N/CM2 0

0 80 60 0

IRRADIATED AT 550oF 20 5.14 X

1019 N/C 0

120 I00 M

80 JUN IRRAD IATED I-

/0 2

O0 U 6 0 0 /

1650/

IRRADIATED 550 0F U

0 w=40 2

0/

I19 2

2LJ 01.20 x 10 N/CM 045 O A6 5.14 X 1l9 N/CM 2 20 0

-100 0

100 200 300 400 500 TEMPERATURE (OF)

Figure 5-2.

Charpy V-Notch Impact Data for San Onofre Pressure Vessel Weld Metal 5-4

14,420-5 120 2

100 P

3 80 2

60 0

LJ.J 550

-J A0 40 0 O IRRADIATED AT 550oF 20 2

5.14 X 101 9 N/CM2 00 80 0 -D 60 0O IRRADIATED AT 5500oF 20 5.14 X 1019 N/CM2 0

Ij 20 120 100 4

v00 UNIRRADIATED 80 O

U 60 O

IRRADIATED AT 550 0 F 40

/

0 1.20 X 1019 N/CM2 0

0, o

A 5.14 X 1ol 9 N/CM2

/1150J 20 0S

-100 0

100 200 300 400 500 TEMPERATURE (OF)

Figure 5-3.

Charpy V-Notch Impact Data for San Onofre Pressure Vessel Weld Heat-Affected-Zone Metal 5-5

14,420-6 120 100 IRRADIATED AT 550 0 F

- 80 5.14 X 1019 N/CM 2

=

60 vi 2

1400 C-)

2 20 2

0 2 2

2 80 80 CL CO 40 --

00 IRRADIATED AT 550 0

F 1020 0.0 209

/M 20)2.36 X 1019 N/CM2 80 2

5.14 b

9 N/M 0

I20 IRRADIATED AT 550oF 100

()

1.20 X lo19 N/cM2 03 2.36 X 1019 N/CM2 80

&-...d 5.14 X 1019 N/CM2

-j 60 UNIRRADIATED (0--1350 40 1300 20 00 0

-100 0

100 200 300 400 500 TEMPERATURE (OF)

Figure 5-4.

Charpy V-Notch Impact Data for A302 Grade B ASTM Correlation Monitor Material 5-6

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR SAN ONOFRE PRESSURE VESSEL MATERIAL IRRADIATED IN CAPSULE F AT 5500 F, FLUENCE 5.14 x 1019 n/cm2 (E > 1 Mev)

Lateral Material Specimen Temperature Energy Expansion Shear Identification Number (0 F)

(ft Ib)

(mils)

(%)

W7601-8 YB45 70 20.0 12 10 YB46 120 29.0 24 20 YB48 150 40.0 32 30 YB41 175 37.5 28 30 YB43 200 55.0 44 45 YB44 210 56.0 50 70 YB42 250 72.5 62 100 YB47 300 71.0 62 100 Weld metal D19 25 12.0 7

5 D17 70 30.0 26 20 D22 125 25.0 23 30 D23 160 40.0 36 40 D21 175 45.0 42 50 D20 210 67.0 62 65 D18 250 80.0 68 100 D24 300 79.0 67 100 HAZ metal H21

-10 3.0 5

2 H22 25 24.0 20 10 H20 50 29.5 26 35 H24 70 49.0 40 50 H19 100 33.0 30 40 H18 125 58.0 48 60 H23 150 81.0 61 95 H17 210 67.0 54 100 5-7

TABLE 5-1 (cont)

CHARPY V-NOTCH IMPACT DATA FOR SAN ONOFRE PRESSURE VESSEL MATERIAL IRRADIATED IN CAPSULE F'AT 550aF, FLUENCE 5.14 x 1019 n/cm2 (E > 1 Mev)

Lateral Material Specimen Temperature Energy Expansion Shear Identification Number (0 F)

(ft lb)

(mils)

(o)

Correlation R 47 70 12.0 16 10 monitor R63 150 26.0 24 20 R58 175 29.0 28 25 R61 200 53.0 44 80 R60 212 47.0 40 85 R62 250 66.0 52 100 R64 300 60.0 47 100 R59 350 59.0 51 100 5-8

TABLE 5-2 EFFECT OF 550aF IRRADIATION AT 5.14 x 1019 n/cm2 (E > 1 Mev) ON THE NOTCH TOUGHNESS PROPERTIES OF THE SAN ONOFRE REACTOR VESSEL SURVEILLANCE MATERIALS Average Energy Absorption Transition Temp (0F)

A Temp (oF) at Full Shear (ft Ib)

Unirradiated Irradiated 50 30 35 50 30 35 50 30 35 Material ft lb ft lb mils ft lb ft lb mils ft lb ft lb mils Unirradiated Irradiated AEnergy W7601-8 80 10 190 130 175 110 120 97 71 26 Weld 15

-20 180 125 160 165

-145 99 80 19 metal HAZ

-30

-60 100 55 80 130 115 106 74 32 metal Correlation 72 40 50 207 170 190 135 130 140 78 62 16 material

14.420-7 YB 45 700F YB 46 120 0F YB 48 1 500F YB 41 1750 F YB 43 2000 F YB 44 2100F YB 42 2500 F Y B 47 3000 F Figure 5-5.

Charpy Impact Specimen Fracture Surfaces for San Onofre Pressure Vessel Shell Plate W7601-8 5-10

14, 420-8 DI9 250F DI7 70oF D22 125 0F D23 160 0F D2 I 175 0F D20 210 0F D18 2500 F D24 3000 F Figure 5-6. Charpy Impact Specimen Fracture Surfaces for San 1Onofre Weld Metal 5-11

14,420-9 H21

-IOoF H22 25OF H20 50oF H24 700F H19 100oF H18 125cF H23 150oF H17 2100F Figure 5-7. Charpy Impact Specimen Fracture Surfaces for San Onofre Weld Heat-Affected-Zone Metal 5-12

14 420-10 R57 700F R63 I500F R58 175 0F R61 2000F R60 212 0F R62 2500 F R64 3000 F R59 350oF Figure 5-8.

Charpy Impact Specimen Fracture Surfaces for A302 Grade B ASTM Correlation Monitor Material 5-13

TABLE -3 SUJMMRY OF SAN ONOFRE REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS 30 Ib 50 fti R.G. 1.99 Fi1eh Traition Tp Transition Temp Transition Temp Material 1019 n/cm2 tcs F ncrese (oF) increase (OF)

W7601-1 2.36 140 110 238 W7601-8 2.36 085 246 W7601-8 5.14 120 110 363 W7601-9 1.20 100 85 186 W7601-9 2.36 10 125 261 Weld metal 1.2 80 95 213 Weld metal

.14 165 385 HAZ metal 1.20 80 85 186 HAZ metal 5.14 115 130 385 ASTM correlation 1.20 120 130 192 monitor 2.36 150 175 269 5.14 130 135 385 5-14

The results in table 5-3 show that transition temperature shifts obtained on the San Onofre surveillance materials are considerably less than those which would be predicted from Regulatory Guide 1.99.[i The results of tests on the ASTM A302B reference correlation monitor material when compared to the ASTM trend curves[ 21 developed for this material also showed lesser transition temperature shifts than predicted. It is suspected that the signif icantly smaller transition temperature shifts exhibited by the San Onofre surveillance materials resulted from self-annealing during irradiation in the power reactor which tends to create a limiting or steady state condition.

Because the transition temperature increases for the material in the third capsule irradiated to 5.14 x 1019 n/cm2 are significantly less than predicted and represent shifts beyond the end of life 1/4 thickness of the vessel (calculated fluence of 4.2 x 1019 n/cm2 ), normal heatup and cooldown operating limit curves using Westinghouse trend curves for adjusting the reference transition temperature have been prepared. The operating limit curves are considered adequate for continued safe operation of the plant.

5-3.

TENSILE TEST RESULTS Table 5-4 and figures 5-9 and 5-10 give the results of the tensile tests on Plate W7601-8 and weld metal, respectively. A summary of the tests performed to date is presented in table 5-5 which indicates that irradiation to 5.14 x 1019 n/cm2 results in no additional increase in yield or tensile strength.

Photographs of the fractured tensile specimens are shown in figures 5-11 and 5-12. A typical stress-strain curve for the tensile tests is shown in figure 5-13.

5-4.

WEDGE OPENING LOADING TESTS Wedge opening loading (WOL) fracture mechanics specimens which were contained in the surveillance capsule have been stored, on the recommendation of the U. S. Nuclear Regulatory Commission, at the Westinghouse Research and Development Center. They will be tested and the results reported at a later date.

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

2. ASTM DS 54, Radiation Effects Information Generated on the ASTM Reference Correlation Monitor Steels, ASTM, Philadelphia, 1974.

5-15

TABLE 5-4 IRRADIATED TENSILE PROPERTIES FOR SAN ONOFRE PRESSURE VESSEL MATERIALS 0.2%

Ultimate Test Yield Tensile Fracture Fracture Uniform Total Reduction Material Specimen Temp Strength Strength Strength Stress Elong Elong In Area Identification Number (aF)

(ksi)

(ksi)

(ksi)

(ksi)

(%)

(%)

%)

W7601-8 YB6 74 82.1 101.2 74.1 218.0 10.4 23.2 66 W7601-8 YB7 550 71.9 97.1 79.4 153.5 10.7 18.1 48 Weld metal D6 74 79.6 94.7 65.2 1.69.3 11.3 25.5 62 Weld metal D5 550 71.4 90.2 70.3 159.7 9.9 19.8 56 0

a

14,420-11 120 10 ULTIMATE TENSILE STRENGTH 1.0 C/3 C/3 60

-0.2-'

YIELD STRENGTH 60 80 REDUCTION IN AREA 60 w

CODE:

CLOSED POINTS -

UNIRRADIATED O

40 OPEN POINTS -

IRRADIATED TO 5.14 X 10I 9 N/CM 2 20 ELONGATION 0

I I

I I

I 0

100 200 300 400 500 600 700 TEMPERATURE (OF)

Figure 5-9.

Tensile Properties for San Onofre Pressure Vessel Shell Plate W7601-8 5-17

14,420-12 100 ULTIMAtE TENSLE STRENGTH 2

C/3 80 LU 60 0.212 YIELD STRENGTH 40 80 REDUCTION IN AREA 60 CODE:

CLOSED POIN TS - UIRADIATED OPEN POINTS -

IRRAIATED TO 6.14 X 1019 N/CM 2 40 2

20 ELO T ION 0.

0 100 20 30 00 500 600 700 TEMPERATURE (0 Figure 5-10. Tensil Properties for Sain nore Weld Metal 5-18

TABLE 5-5

SUMMARY

OF TENSILE RESULTS FOR SAN ONOFRE PLATE W7601-8 AND WELD METAL Test Material Fluence Temp 0.2% YS UTS RA Elong Identification (n/cm2)

(OF)

(ksi)

(ksi)

(%)

(%)

W7601-8 0

room 61.7 84.9 63.4 25.4 W7601-8 0

room 63.3 84.8 58.0 26.9 W7601-8 2.36 x 1019 room 90.1 113.0 63.1 26.0 W7601-8 5.14 x 1019 room 82.1 101.2 66.0 23.2 W7601-8 0

600 55.9 82.3 59.8 24.9 W7601-8 0

600 55.1 83.5 60.8 23.6 W7601-8 5.14 x 1019 550 71.9.

97.1 48.0 18.1 Weld metal 0

room 63.7 82.2 66.0 28.6 Weld metal 0

room 63.0 80.8 66.8 28.6 Weld metal 1.20 x 1017 room 83.0 103.2 60.6 28.3 Weld metal 5.14 x 1019 room 79.6 94.7 62.0 25.5 Weld metal 0

600 55.0 78.9 53.2 23.8 Weld metal 0

600 55.6 78.9 57.1 21.8 Weld metal 5.14 x 1019 550 71.4 90.2 56.0 19.8

14,420-13 YB6 74oF Y 87 550OF Figure 5-11.

Fractured Tensile Specimens From San Onofre Shell Plate W7601-8 5-20

14,420-14 74oF 05 550OF Figure 5-12. Fractured Tensile Specimens From San Onofre Weld Metal 5-21

120 100 80 C-,

LU w

60

'10 20I I

I IIII 20 0

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

Figure 5-13.

Typical Stress-Strain Curve for Tension Specimens (Tension Specimen No. YB-6)

SECTION 6 NEUTRON DOSIMETRY ANALYSIS 6-1.

DESCRIPTION OF NEUTRON FLUX MONITORS To effect a correlation between neutron exposure and the radiation-induced property changes observed in the test specimens, a number of neutron flux monitors were included as an integral part of the reactor vessel surveillance program. Table 6-1 lists the particular monitors contained, within Capsule F along with the nuclear reaction of interest and the energy range of each monitor.

The first five reactions listed in table 6-1 are used as fast neutron monitors to relate neutron fluence (E > 1.0 Mev) to the measured shift in RTNDT. To properly account for burnout of the product isotope generated by the fast neutron reactions, it is also necessary to determine the magnitude of the thermal neutron flux at the monitor location. Therefore, bare and cadmium-covered cobalt-aluminum monitors were included within Capsule F.

Figure 4-2 shows the relative locations of the various monitors within Capsule F, and figure 4-1 shows the radial and azimuthal positions of the capsule with respect to the nuclear core, reactor internals, and pressure vessel. The nickel, copper, and cobalt-aluminum monitors (in wire form) were placed in holes drilled in spacers at several axial levels within the capsule.

The iron monitors were obtained by drilling samples from selected Charpy test specimens. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule.

The use of activation monitors such as those listed in table 6-1 does not yield a direct measure of the energy-dependent neutron flux level at the point of interest. Rather, the activation process is a measure of the integrated effect that the time and energy-dependent neutron flux has on the target material. An accurate estimate-of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

m The operating history of the reactor a

The energy response of the monitor 6-1

TABLE 6-1 NEUTRON FLUX MONITORS CONTAINED WITHIN CAPSULE F Wt% of Target Reaction of g

Monitor Material Interest gmmonitor Response Range Product Half-Life Copper Cu63 (n,a) Co60 0.6917 E > 4.7 Mev 5.27 years Iron Fe54 (n,p) Mn54 0.0585 E > 1.0 Mev 314 days Nickel Ni58 (n,p) Co58 0.6777 E > 1.0 Mev 71.4 days Uranium 2 38[al U238 (n,f) Cs137 1.0 E > 0.4 Mev 30.2 years Neptunium237[al Np2 37 (n,f) Cs37 1.0 E > 0.08 Mev 30.2 years Cobatt-Aluminumlal Co59 (n,X) Co60 0.0015 0.4 ev < E < 0.015 Mev 5.27 years Cobalt-Aluminum Co59 (n,X) Co60 0.0015 E < 0.015 Mev 5.27 years

a. Cadmium shielded monitors

m The neutron energy spectrum at the monitor location a

The physical characteristics of the monitor 6-2.

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

The energy and spatial distribution of neutron flux within the San Onofre Unit No. 1 reactor geometry was obtained with the DOT[11 two-dimensional Sn transport code. The radial and azimuthal distributions were obtained from an R,O computation wherein the reactor core, reactor internals, surveillance capsule, water annuli, pressure vessel, and primary shield concrete were described on the analytical model. These analyses employed 21 neutron energy groups, an S8 angular quadrature, and a P1 cross-section expansion. The reactor core power distribu tions used in the calculations, which 'were representative of time-averaged conditions over an equilibrium fuel cycle, accounted for rod-by-rod spatial variations in the peripheral fuel assemblies. The analytical geometries described a 45-degree sector of the reactor, assuming one-eighth symmetry. Relative axial variations of neutron flux incident on the reactor vessel were obtained from R, Z DOT calculations based on the equivalent cylindrical core concept.

The specific activity of each of the activation monitors was determined in accordance with established ASTM procedures.[2,3,4,5 61 Following sample preparation, the activity of each monitor was determined by means of a lithium drifted germanium Ge (Li) gamma spectrometer.

The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration.

1. Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., "Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5 -

Two-Dimensional, Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.

2. ASTM Designation E261-70, "Standard Method for Measuring Neutron Flux by Radioactivation Techniques," in.ASTM Standards (1975), Part 45, Nuclear Standards, pp. 745-755, Am Society for Testing and Materials, Philadelphia, Pa. 1975.,

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

1975.

4. ASTM Designation E263-70, "Standard Method for Measuring Fast-Neutron Flux by Radioactivation' of Iron," in ASTM Standards (1975), Part 45, Nuclear Standards pp. 764-769, Am. Society for Testing and Materials, Philadelphia, PA. 1975.

5. ASTM Designation E481-73T, "Tentative Method of Measuring Neutron -

Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 887-894, Am. Society for Testing and Materials, Philadelphia, PA. 1975.

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

6-3

For the samples removed from Capsule F, the overall 2u deviation in all of the measured data was determined to be plus or minus 10 percent.

Having the measured activity of the monitors and the neutron energy spectrum at the location of interest, the calculation of the fast neutron flux proceeded as follows. The reaction product activity in the monitor was expressed as ng D

y f o(E) 0(E) n (1

- e-'j) e-Xd (6-1)

A Pmax fE j=1 where D = induced product activity No = Avogadro's number A = atomic weight of the target isotope

f. = weight fraction of the target isotope in the target material y = number of product atoms produced per reaction o(E) = energy-dependent reaction cross section O(E) = energy-dependent neutron flux at the monitor location with the reactor at full power P= average core power level during irradiation period j Pmax = maximum or reference core power level X = decay constant of the product isotope

= length of irradiation period j Td = decay time following irradiation period j Because neutron flux distributions were calculated by means of multigroup transport methods and, further, because the prime interest was in the fast neutron flux above 1 Mev, spectrum averaged reaction cross sections were defined such that the integral term in equation (6-1) could be replaced by the following relation J o(E) 0(E) = uO (E > 1 Mev)

E 6 -4

where oo n

(E) O(E)

Ug 09 on0 n

O(E) o 1.0 Mev G=G1.O Mev Thus, equation (6-1) was rewritten D No

=

u (E=> 1.0 Mev) e ed

.Am max or, solving the neutron flux (E > 1.0 Mev)=

D 6-2 No

(

Tj eXTd f iy Y The total fluence above 1 Mev was then given by n

(E > 1.0 Mev)

(E > 1.0 Mev)

(6-3)

. max j=1 where n

7 = total effective full power seconds of reactor operation up j=1 max to the time of capsule removal An assessment of the thermal neutron flux levels within Capsule F was obtained from the bare and cadmium-covered Co59 (n,X) Co60 data by means of cadmium ratios and the use of a 37-barn 2200 m/sec cross section. Thus, bare{RR f

(6-4)

Oth

- n No fj y

(

7Xrj) e-Xrd 4 rmax j=1 6-5

where R is defined as Dbare/DCd-covered The irradiation history of the flux monitors removed from Capsule F is listed in table 6-2.

The spectrum-averaged reaction cross sections derived for each of the fast neutron flux monitors are listed in table 6-3.

6-3.

RESULTS OF ANALYSIS Table 6-4 lists the fast neutron (E > 1 Mev) flux and fluence levels derived from the monitors taken from Capsule F. In examining the data listed in table 6-4, it should be noted that the Fe54 monitors were positioned within the surveillance capsule at a radius of 168.68 and 169.68 cm relative to the core centerline. The corresponding radius of the U238 and Np237 monitors was 168.91 cm and that of the Ni58 and Cu63 monitors was 169.68 cm. Thus, it should be expected that the measured neutron flux levels reflect the flux gradient caused by attenuation within the test specimens. Table 6-5 summarizes the thermal neutron flux obtained from the cobalt-aluminum monitors. Due to the relatively low thermal neutron flux at the capsule location, no burnup correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than 1 percent of the Ni58 (n, p) Co58 reaction and even less significant for all the other fast reactions.

Figures 6-1 through 6-3 and table 6-6 summarize results of the Sn transport calculations for the San Onofre reactor vessel. In figure 6-1, the calculated maximum fast neutron flux levels at the pressure vessel inner radius, 1/4 thickness location, and 3/4 thickness location are presented as a function of azimuthal angle. Figure 6-2 shows the relative axial variation of neutron flux. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6-1 by the appropriate values from figure 6-2. In figure 6-3, the calculated maximum end-of-life fast neutron exposure of the San Onofre reactor vessel is given as a function of radial position within the vessel wall. Table 6-6 lists the calculated fast neutron flux levels interior to Capsule F along with the lead factors (LF) relating capsule exposure to vessel exposure. The lead factor is defined as the ratio of the calculated flux at the monitor location to the calculated peak neutron flux incident on the reactor vessel.

Based on the iron data in table 6-4, the average fast neutron fluence incident on the front row of Charpy specimens is determined to be 5.74 x 1019 n/cm2, while that on the back row of the specimens is 4.53 x 1019 n/cm2 or an average value of 5.14 x 1019 n/cm2. These measured values correspond to analytical values of 5.54 x 1019 and 4.29 x 1019 n/cm2 respectively, or an average value of 4.91 x 1019 n/cm 2. A comparison of these values shows excellent agreement.

6-6

TABLE 6-2 IRRADIATION HISTORY OF CAPSULE F P

pIrradiation Decayfal avg max Time Time Month (Mw)

(Mw) avg max (days)

(days) 6/67-12/67 240 1347 0.178 201 4028 1/68 944 1347 0.701 31 3997 2/68 700 1347 0.520 29 3968 3/68 277 1347 0.206 31 3937 4/68-8/68 0

1347 0.000 153 3784 9/68 254 1347 0.189 30 3754 10/68 1135 1347 0.842 31 3723 11/68 1182 1347 0.877 30 3693 12/68 1077 1347 0.799 31 3662 1/69 852 1347 0.632 31 3631 2/69 1150 1347 0.854 28 3603 3/69 773 1347 0.574 31 3572 4/69 1034 1347 0.768 30 3542 5/69 1210 1347 0.898 31 3511 6/69 789 1347 0.592 30 3481 7/69 0

1347 0.000 31 3450 8/69 587 1347 0.436 31 3419 9/69 1342 1347 0.997 30 3389 10/69 919 1347 0.682 31 3358 11/69 1331 1347 0.988 30 3328 12/69 1329 1347 0.986 31 3297 1/70-7/70 1253 1347 0.930 212 3085 8/70 1325 1347 0.984 31 3054 9/70 1337 1347 0.993 30 3024 10/70.

80 1347 0.059 31 2993 11/70 299 1347 0.222 30 2963 12/70 1342 1347 0.996 31 2932

a. Decay time is referenced to the counting date of the Fe, Ni, Cu, and Co monitors (1/10/79). The Np and U monitors were counted on 1/19/79.

6-7

TABLE 6-2 (cont)

IRRADIATION HISTORY OF CAPSULE F Irradiation Decaylal Pavg max Time Time Month (Mw)

(Mw)

Favg max (days)

(days) 1/71 1328 1347 0.986 31 2901 2/71 1328 1347 0.986 28 2873 3/71 1330 1347 0.987 31 2842 4/71 1317 1347 0.978 30 2812 5/71 934 1347 Q.694 31 2781 6/71 1138 1347 0.845 30 2751 7/71 980 1347 0.727 31 2720 8/71 1124 1347 0.835 31 2689 9/71 1252 1347 Q.930 30 2659 10/71 1079 1347 0.801 31 2628 11/71 1115 1347 0.828 30 2598 12/71 1024 1347 0.760 31 2567 1/72 0

1347

.Q000 31 2536 2/72 181 1347 0.134 29 2507 3/72 1272 1347 0.944 31 2476 4/72 1237 1347 0.919 30 2446 5/72 1019 1347 0.756 31 2415 6/72 1307 1347 0.970 30 2385 7/72 910 1347 0.676 31 2354 8/72 1190 1347 0.884 31 2323 9/72 1165 1347 0.865 30 2293 10/72 1051 1347 0.780 31 2262 11/72 1327 1347 0.985 30 2232 12/72 1342 1347 0.996 31 2201 1/73 1115 1347 0.828 31 2170 2/73 1326 1347 0.984 28 2142 3/73 1340 1347 0.995 31 2111 4/73 1322 1347 p.981 30 2081

a. Decay time is referenced to the counting date of the Fe, Ni, Cu, and Co monitors (1/10/79). The Np and U monitors were counted on 1/19/79.

TABLE 6-2 (cont)

IRRADIATION HISTORY OF CAPSULE F Irradiation Decaylal avg max Time Time Month (Mw)

(Mw)

Pavg' max (days)

(days) 5/73 1333 1347 0.989 31 2050 6/73 43 1347 0.032 30 2020 7/73 121 1347 0.090 31 1989 8/73 1103 1347 0.819 31 1958 9/73 1312 1347 0.974 30 1928 10/73 845 1347 0.627 31 1897 11/73-12/73 0

1347 0.000 61 1836 1/74 369 1347 0.274 31 1805 2/74 1347 1347 1.000 28 1777 3/74 1306 1347 0.970 31 1746 4/74 1182 1347 0.878

-30 1716 5/74 471 1347 0.349 31 1685 6/74 1315 1347 0.976 30 1655 7/74 1180 1347 0.876 31 1624 8/74 1330 1347 0.987 31 1593 9/74 1237 1347 0.919 30 1563 10/74 1193 1347 0.885 31 1532 11/74 1330 1347 0.987 30 1502 12/74 1347 1:347 1.000 31 1471 1/75 1330 1347 0.987 31 1440 2/75 1334 1347 0.990 28 1409 3/75 579 1347 0.430 31 1381 4/75 277 1347 0.206 30 1351 5/75 1304 1347 0.968 31 1320 6/75 1090 1347 0.809 30 1290 7/75 1347 1347 1.000 31 1259 8/75 1337 1347 0.993 31 1228 9/75 1347 1347 1.000 30 1198

a. Decay time is referenced to the counting date of the Fe, Ni, Cu, and Co monitors (1/10/79). The Np and U monitors were counted on 1/19/79.

6-9

TABLE 6-2 (cont)

IRRADIATION HISTORY OF CAPSULE F Irradiation Decay [al Pavg Pmax Time Time Month (Mw)

(Mw)

Pavg max (days)

(days) 10/75 1320 1347 0.980 31 1167 11/75 1347 1347 1.000 30 1137 12/75 1347 1347 1.000 31 1106 1/76 1311 1347 0.973 31 1075 2/76 1306 1347 0.969 29 1046 3/76 1337 1347 0.993 31 1015 4/76 913 1347 0.678 30 985 5/76 1342 1347 0.996

31.

954 6/76 1315 1347 0.976 30 924 7/76 1144 1347 0.849 31 893 8/76 1089 1347 0.808 31 862 9/76 930 1347 0.691 30 832 10/76-12/76 0

1347 0.000 92 740 1/77-3/77 0

1347 0.000 90 650 4/77 669 1347 0.496 30 620 5/77 1340 1347 0.995 31 589 6/77 1332 1347 0.989 30 559 7/77 1335 1347 0.991 31 528 8/77 1313 1347 0.975 31 497 9/77 374 1347 0.278 30 467 10/77 1040 1347 0.772 31 436 11/77 1285 1347 0.954 30 406 12/77 1340 1347 0.995 31 375 1/78 1347 1347 1.000 31 344 2/78 1347 1347 1.000 28 316 3/78 1337 1347 0.993 31 285 4/78 459 1347 0.341 30 255

a. Decay time is referenced to the counting date of the Fe, Ni, Cu. and Co monitors (1/10/79). The Np and U monitors were counted on 1/19/79.

6-10

TABLE 6-2 (cont)

IRRADIATION HISTORY OF CAPSULE F Irradiation Decay [a avg max Time Time Month (Mw)

(Mw) avg max (days)

(days) 5/78 920 1347 0.683 31 224 6/78 896 1347 0.665 30 194 7/78 1164 1347 0.864 31 163 8/78 1238 1347 0.919 31 132 9/78 1033 1347 0.767 15 117

a. Decay time is referenced to the counting date of the Fe, Ni, Cu, and Co monitors (1/10/79). The Np and U monitors were counted on 1/19/79.

TABLE 6-3 SPECTRUM AVERAGED REACTION CROSS SECTIONS USED IN FAST NEUTRON FLUX DERIVATION Reaction a (barns)

Fe54 (n,p) Mn54 0.0664 Ni58 (n,p) Co58, 0.0891 Cu63 (n,a) Co60 0.000472 U238 (n,f) F.P.

0.343 Np23 7 (n,f) F.P.

2.80 6-11

TABLE 6-4 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE F Measured Reaction and Activity b (E > 1.0 Mev)[b]

4> (E > 1.0 Mev)1bI Monitor Location (dps/gm)

(n/cm2-sec)

(n/cm2)

Fe54 (n,p) Mn54 Core Side Charpy D-17 5.86 x 106 2.37 x 1011 5.81 x 1019 D-20 6.20 x 106 2.51 x 1011 6.15 x 1019 D-24 5.30 x 106 2.15 x 1011 5.27 x 1019 Vessel Side Charpy R-57 4.57 x 106 1.85 x 1011 4.53 x 1019 R-60 4.85 x 106 1.96 x 1011 4.80 x 1019 R-64 4.30 x 106 1.74 x 1011 4.26 x 1019 Ni58 (n,p) Co58 Center 2.18 x 107 1.36 x 1011 3.33 x 1019 Cu63 (n,a) Co60 Top 3.40 x 105 2.06 x 1011 5.05 x 1019 Center 3.64 x 105 2.21 x 1011 5.41 x 1019 Bottom 3.69 x 105 2.24 x 1011 5.49 x 1019 Np237 (n,f) Csl 37 Center 8.32 x 106 1.25 x 1011 3.06 x 1019 U238 (n,f) Csl 37 Center 1.96 x 106 2.29 x 1011 5.61 x 1019

a. Mn54, Co58, and Co6 0 activities are referenced to 1/10/79. Cs137 activities are referenced to 1/19/79.

b. Derived flux and fluence levels are subject to +/-10 percent measurement uncertainty.

6-12

TABLE 6-5 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE F BARElal

[b]

Activity Cd Covered Activity[8]

th Monitor Location (dps/gm)

(dps/gm)

(n/cm2-sec)

Center 5.52 x 107 3.11 x 107 8.08 x 1010

a. Co6 0 activities are referenced to 1/10/79.

b. Derived flux level is subject to +/-10 percent measurement error.

TABLE 6-6 CALCULATED FAST NEUTRON FLUX AND LEAD FACTORS FOR CAPSULE F Location Within (E > 1.0 Mev)

Capsule F (n/cm2-sec)

Lead Factor Core Side Charpy 2.26 x 1011 1.93 Dosimeter Block 2.11 x 1011 1.80 Vessel Side Charpy 1.75 x 1011 1.50 6-13

14,420-16 2

0 5

PRESSURE VESSEL INNER DIAMETER IN/T LOCATION 1i010 5

3/4T LOCATION 2

10 0

10 20 30 40 50 AZIMUTHAL ANGLE (DEGREES)

Figure 6-1.

Calculated Azimuthal Distribution of Neutron Flux (E > 1 Mev) at the Core Midplane of the San Onofre Reactor Vessel 6-14

14,420-17 1.00 5

-2 0.10

.. J U

CD 5

U-i w

2 0.01 5

2 TOWARD TOWARD SUMP OPERATING DECK 0.001

-300

-200

-100 0

100 200 300 AXIAL DISTANCE FROM CORE MIDPLANE (CM)

Figure 6-2.

Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev)

Incident on the San Onofre Reactor Vessel 6-15

14,420-18 1 80.3L4 1020 IR 86.63 5

c'J-I /LT 2:

.J LU Z

9 199.22 3/4 T 5/L

205.51 OR 2

178 180 182 184 186 188 190 192 194 196 198 200 202 204 206 RADIUS (CM)

Figure 6-3.

Calculated Maximumi End-of-Life Fast Neutron Fluence (E > 1 Mev) as a Function of Radius Within the San Ohofre Reactor Vessel 6-16

With the use of the lead factors listed in table 6-6, a comparison of the end-of-life peak fast neutron exposure of the San Onofre reactor as derived from both calculations and measured surveillance capsule results may be made as indicated in table 6-7.

TABLE 6-7 FAST NEUTRON EXPOSURE DERIVED FROM CALCULATED AND MEASURED RESULTS Fast Neutron Fluence (n/cm2)

Based on Iron Based on Iron Vessel Location Calculated From Front Charpys From Back Charpys Inner Surface 1.0 x 1020 1.0 x 1020 1.1 x 1020 1/4 Thickness 4.2 x 1019 4.2 x 1019 4.6 x 1019 3/4 Thickness 7.1 x 1018 7.1 x 1018 7.7 x 1018 These data are based on 27 full-power years of operation at 1347 Mw.

6-17

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 nil-ductility temperature). The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material..

properties and estimating the radiation-induced ARTNDT. RTNDT is designated as the higher of 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 600 F.

RTNDT increases as the material is exposed to fast neutron radiation. Thus, to find the most limiting RTNDT at any time period in the reactor's life, a ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT.

The extent of the shift in RTNDT is enhanced by certain chemical elements (such -as copper) present in reactor vessel steels.

Design curves which show the effect of fluence and copper content on ARTNDT for reactor vessel steels exposed to 550aF are shown in figure A-1.

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

A-2.

FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture toughness properties of the San Onofre Unit No. 1 reactor vessel materials are presented in table A-1. The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC A-1

14,420-21 400 300 200 150

.100 80 60 40 0.30%o CU BASE, 0.25% CU WELD 0.25% CU BASE, 0.20% CU WELD 0.20% CU BASE,.0.15/o CU WELD 20" 0.15% CU BASE. 0.10% CU WELD 0.101, CU BASE. 0.5% CU WELD 0"

2 5

101 2

5 0 20 FLUENCE HN/CM2 E :*I MEV)

Figure A-1.

Effect of Fluence and Copper Content on ARTNIDT for Reactor Vessel Steels Exposed to Irradiation at 5500 F A-2

14,420-22 1Q 20 1.0 x 1020 5

SURFACE o~~

4.,2' X 1011 2

I1/4 THICKNESS 101 9 o000 7.1 X, 1018

.5 3/4 THICKNESS LU C-2 0

5 2

017 0

5 10

15.

20 25 30 35 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)

Figure A-2.

Fast Neutron Fluence (E > 1 Mev) as a Function of Full-Power Service Life A-3

TABLE A-i REACTOR VESSEL. TOUGHNESS DATA (UNIRRADIATED)

Minimum Average 50 ft-lb/35 mil Upper Shelf Material Cu P

NDTT Temp (OF)

RTNDT Energy (ft-lb)

Component Code. No.

Type

(%)

(%)

(0 F Long.,

Trans.

(o F)

Long.

Trans.

Cl. Hd. Dome W7604 A302B 1112 132 72 72.5 Peel Segment W7605-1 A302B

-10 114 134 74 70.5 Peel Segment W7605-2 A302B

-10 90 110 50 122 Peel Segment W7605-3 A302B

-10 108 128 68 85 Peel Segment W76054 A302B

-10 120 140 80 74 Peel Segment W7605-5 A302B

-10 26 46

-10 109 Peel Segment W7605-6 A302B

-10 102 122 62 88 Hd. Flange W7602 A336 mod 60[a

[b 60 Ves. Flange-W7603 A336 mod 60[al

[b]

60 Inlet Nozzle W7611-11 A336 mod 6- [-60

[b 60 Inlet Nozzle W7611-2 A336 mod 60[a]

[b]

60 Inlet: Nozzle W7611-3 A336 mod 6 0 [a]

[bi 60 Outlet Nozzle W7610-1 A336 mod 6 0 [a]

[b]

60 Outlet Nozzle W7610-2 A336 mod 60[a]

[b]

60 Outlet Nozzle W7610-3 A336 mod 601

[b]

60 Upper Shell W7601-3 A302B 0.15 0.014

-10 48 68 8

98.5 Upper Shell W7601-6 A302B 0.16 0.012

-30 64 84 24 104 Upper Shell W7601-7 A302B 0.15 0.014

-20 52 72 12 95.5

a. Estimated per NRC Standard Review Plan Branch Technical Position MTEB 5-2.

b. Only 100F Charpy V-notch data available. Conservative estimates for NDTT and RTNDT were used.

TABLE A-1 (cont)

REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

Minimum Average 50 ft-lb/35 mil Upper Shelf Material Cu P

NDTT Temp (OF)

RTNDT Energy (ft-lb)

Component Code No.

Type

(%)

(%)

(OF)

Long.

Trans.

(OF)

Long.

Trans.

Inter. Shell W7601-1 A302B 0.17 0.013 0

57 120[al 60 94 75 Inter. Shell W7601-8 A302B 0.18 0.012 10 93 100[a 40 97 79 Inter. Shell W7601-9 A302B 0.18 0.014 0

64 115 [a]

55 102 72 Lower Shell W7601-2 A302B 0.17 0.013

-20 74 94 34 97 Lower Shell W7601-4 A302B 0.14 0.014

-10 91 111 51 94 Lower Shell W7601-5 A302B 0.14 0.014 10 122 142 82 87.5 Bot. Hd. Peel W7607 A302B

-20 62 82 22 91 Bot. Hd. Dome W7606 A302B 60[b]

99 119 60 86 Weld 0.19 0.017 0 [b]

29[al 0

90 HAZ 0 1b]

14a 0

101

a. Actual not estimated

b. Estimated per NRC Standard Review Plan Branch Technical Position MTEB 5-2.

Regulatory Standard Review Plan.11] The postirradiation fracture toughness properties of the reactor vessel belt line material were obtained directly from the San Onofre Unit No. 1 Vessel Material Surveillance Program.

A-3.

CRITERIA FOR ALLOWABLIE PIESSUF(E-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowabi limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, 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.121 The KIR curve is given by the equation:

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

(A-1) where K1 R is the reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G to the ASME Codel 21 as follows:

C KIM + Kit < KIR (A-2) where KIM is the stress intensity factor caused by membrane (pressure) stress.

Kit is the stress intensity factor caused by the thermal gradients.

KIR is a function of temperature relative to the RTNDT 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 hot critical.

At any time during the heatup or cooldown transient, KIR is determined by the metal tem perature at. the tip of the postulated flaw, the approPriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradientsthrough the

1. "Fracture Toughness Requirements for Older Plants" Branch Technical Position MTEB 5-2 Standard Review Plan, NUREG-75/087, 1975.

2. Appendix G to the ASME Boiler and Pressure Vessel Code, Section 1II, Division 1, Subsection NA, "Protection Against Non-Ductile Failure," American Society of Mechanical Engineers, New York, N.Y., 1977 Edition and Summer 1978 Addenda.

A-6

vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, for the reference flaw are computed. From equation (A-2) the pressure stress intensity factors are obtained, and from these the allowable pressures are calculated.

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

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. 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 KI R at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in KIR exceeds Kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature-at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures 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 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce com pressive stresses 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 KIR for the 1/4. crack during heatup is lower than the KIR for the 1/4T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower KI R's do not offset each other, and the pressure-temperature curve based on steady-state A-7

conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed. in order to, assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at' the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of 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 ire produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is, taken to be the lesser of the 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 such that 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 composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

A-4.

HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in paragraph A-3. The derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan.[1]

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

.1. "Pressure Temperature Limits," Section 5.3.2 of Standard Review Plan, NUREG-75/087, 1975.

A-8

Charpy test specimens from Capsule F indicate that the representative core region weld metal and the limiting core region plate (W7601-8) exhibited shifts in RTNDT of 1650 and 1100F, respectively. These shifts at a fluence of 5.14 x 1019 n/cm2 are well within the appropriate design curve (figure A-1) prediction. Heatup and cooldown calculations were based on the ARTNDT predicted for the core region plate material which is considered to be the limiting vessel material. Heatup and cooldown limit curves for normal operation of the reactor vessel are presented in figures A-3 and A-4 and represent an operational time period of 16 effective full power years.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in figure A-3, in addition to other criteria which must be met before the reactor is made critical.

The leak test limit curve shown in, figure A-3 represents minimum temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of references 1 and 2.

Figures A-3 and A-4 define limits to assure prevention of nonductile failure.

Since plant heatup and cooldown limitations were re-evaluated using the new peak fast fluence values resulting from the analyses performed on Capsule F and updated estimates of RTNDT, sections of the plant Technical Specifications which address plant heatup and cooldown limitations should be amended to reflect the changes in normal plant operation limitations.

A-9

14,420-19 4000 MATERIAL PROPERTY BASIS AT 1/4T LOCATION CONTROLLING MATERIAL: INTERMEDIATE SHELL COPPER CONTENT: 0.18 WTV PHOSPHORUS CONTENT: 0.014 WT RTNDT INITIAL: 550 F RT AFTER 16 EFPY: 217oF MATERIAL PROPERTY BASIS AT 3/4T LOCATION CONTROLLING MATERI AL: LOWER SHELL COPPER CONTENT: 0.14 WT-*

PHOSPHORUS.CONTENT: 0.014 WT-'

3000 RTNDT INITIAL: 820 F RTN AFTER 16 EFPY: 1630F CNDT 14,40-1 I

I E

.UV PLIAL O HEATUP RATES U

~~~~~UPTO 60oF/HFRTE EVIEP R

000 INSTRBASED ONOR INSRI CE LU 2000YDROSTATSC TEST TEMPERATURE (338 0 F) 0 0

100 200 300 400 500 INDICATED TEMPERATURE (oF)

Figure A-3.

San Onofre Unit No. 1 Reactor Coolant System Heatup Limitations Applicable for the First 16 EFPY A-10

14,420-20 3000 MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:

INTERMEDIATE SHELL COPPER CONTENT: 0.18 WTI PHOSPHORUS CONTENT: 0.014 WTI/

RTNDT INITIAL: 550F RTNDT AFTER 16 EFPY: I/NT. 2170F CURVE APPLICABLE' FOR COOLDOWN RATES UP TO 10oF/HR FOR THE SERVICE PERIOD UP TO IE L

IIIIII 16 EFPY AND CONTAINS MARGINS OR 100F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.

'9 2000

-000 Lii C-0 COOLDOWN RATE oF/HR IIl 0

20 40

-60

-100 0

0 100 200 300 400 500 INDICATED TEMPERATURE (OF)

Figure A-4. San Onofre Unit No. 1 Reactor Coolant System Cooldown Limitations Applicable for the First 16 EFPY A-11