ML20079Q428

Analysis of Capsule T from Tva,Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program

ML20079Q428
Person / Time
Site:	Sequoyah
Issue date:	05/31/1983
From:	Shaun Anderson, Mager T, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20079Q430	List: ML20079Q428
References
WCAP-10340, NUDOCS 8402010162
Download: ML20079Q428 (76)

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o WCAP-10340 WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION Analysis of Capsule T From the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program May 1983 S.E. Yanichko S.L Anderson C.A. Cheney W.T. Kaiser Work performed under Shop Order No. UVA-6620 APPROVED:

T. R. Ma'ger, Mg[ager Metallurgical #fid NDE Analysis Prepared by Westinghouse for the Tennessee Valley Authority Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.

Westinghouse Electric ' Corporation Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 8402010162 840124 PDR ADOCK 05000327 P

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Table of Contents Section Title Page 1

Summary of Results 1-1 2

Introduction 2-1

3 Background

3-1 4

Description of Program 4-1 5

Testing of Specimens from Capsule T 5-1 5-1 Overview 5-1 5-2 Charpy V-Notch impact Test Results 5-3 5-3 Tension Test Results 5-18 i

5-4 Wedge Opening Loading Tests 5-18 6

Radiation Analysis and Neutron Dosimetry 6-1 6-1 Introduction 6-1 6-2 Discrete Ordinates Analysis 6-1 6-3 Neutron Dosimetry 6-6 6-4 Transport Analysis Results 6-10 6-5 Dosimetry Results 6 19 7

References 7-1 Appendix A Heatup and Cooldown Limit Curves for Normal A-1 Operation 47 sos:: 07taes iii

List of lilustrations Figure Title Page 4-1 Arrangement of Surveillance Capsules in the Sequoyah Unit 1 Reactor Vessel (Updated Lead Factors for Capsules Shown in Parentheses)

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4-3 4-2 Arrangement of Specimens, Thermal Monitors, and Dos.' meters i1 Capsule T 4-5 5-1

' Charpy V-Notch impact Data for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging (Axial Orientation) 5-8 5-2 Charpy V Notch Impact Data for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging (Tangential Orientation) 5-9 I

5-3 Charpy V-Notch impact Data for Sequoyah Unit 1 Reactor Vessel Weld Metal 5-10 5-4 Charpy V-Notch Impact Data for Sequoyah Unit 1 Reactor Vessel 'Neld HAZ Metal 5-11 5-5 Charpy Impact Specimen Fracture Surfaces for Sequoyah Unit 1 Lower Shell Forging 04 (Axial Orier.tation) 5-13 5-6 Charpy impact Specimen Fracture Surfaces for Sequoyah Unit 1 Lower Shell Forging 04 (Tangential Orientation) 5-14 5-7 Charpy impact Specimen Fracture Surfaces for Sequoyah Unit 1 Weld Metal 5 15 5-8 Charpy impact Specimen Fracture Surfaces for Sequoyah Unit 1 Weld Heat Affected Zone Metal 5-16 5-9 Comparison of Actual Versus Predicted 30 ft Ib Transition Temperature increase for Sequoyah Unit 1 Reactor Vessel Materials Using the Prediction Methods of Regulatory Guide 1.99 Revision 1 5-17 s7 ace a o7tesa v

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d List of lilustrations (cont)

Figure Title Page s

5-10 Tensile Properties for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging (Axial Orientation) 5-20 5-11 Tensile Properties for Sequoyah Unit 1 Reactor Vessel Weld Metal 5-21 5-12 Fractured Tension Specimens from Sequoyah Unit 1 Lower Shell Forging 04 and Weld Metal 5-22 5-13:

Typical Stress-Strain Curve for Tension Specirnens 5-23 6-1 Sequoyah Unit 1 Reactor Geometry 6-2 6-2 Plan View of a Reactor Vessel Surveillance Capsule 6-4 6-3 Calculated Azimuthal Distribution of Maximum Fast Neu-tron Flux (E > 1.0 MeV) Within the Pressure Vesset - Sur-veillance Capsule Geometry 6-11 6-4 Calculated Radial Distribution of Maximum Fast Neutron i

Flux (E > 1.0 MeV) Within the Pressure Vessel 6-12 64.

Relative Axial Variat!an of Fast Neutron Flux (E > 1.0 MeV)

Within the Pressure Vessel 6-13 6-6 Calculated Radial Distribution of Maximum Fast Neutron Flug (E > 1.0 MeV) Within the Surveillance Capsules 6 14 6-7 Calculated Variation of Fast Neutron Flux Monitor Satu-rated Activity Within Capsules Located at 40 Degrees 6-15 6-8 Calculated Variation of Fast Neutron Flux Monitor Satu-rated Activity Within Capsules Located at 4 Degrees 6-16 vi snos:to7:ses

List of Tables

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Table Title Page 4-1 Chemical Composition and Heat Treatment of Material for the Sequoyah Unit 1 Reactor Vessel Surveillance Pn gram 4-7 5-1 Charpy impact Data for S'equoyah Unit 1 Reactor Vessel Lower Shell Forging 04 (irradiated to 2.74 x 10'8 nicm8) 5-4 2 Charpy impact Data for Sequoyah Unit 1 Reactor Vessel Weld Metal arid HAZ Metal (Irradiated to 2.74 x 10's n/cm2) 5-5 5-3 Instrumented Charpy impact Test Results for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging 04 5-6 5-4 Instrumented Charpy impact Test Results for Sequoyah Unit 1 Weld Metal and HAZ Material 5-7 5-5 The Effect of 550*F frradiation to 2.74 x 10'8 n/cm2 (E > 1.0 MeV) on tha Notch Toughness Properties of Sequoyah Unit 1 Reactor Vessel Materials 5-12 5-6 Terisile Properties for Sequoyah Unit 1 Reactor Vessel Material Irradiated to 2.74 x 10 n/cm' 5-19 6-1 47 Group Energy Structure 6-5 l

6-2 Nuclear Constants for Neutron Flux Monitois Contained in the Sequoyah Unit 1 Surveillance Capsules 6-7 6-3 Calculated Fast Neutron Flux (E > 1.0 MeV) and Lead Fac-tors for Sequoyeh Unit 1 Surveillance Capsules 6-17 l

6-4 Calculated Neutron Energy Spectra at the Center of the Sequoyah' Unit 1 Surveillance Capsules 6-18 6-5 Spectrum Averaged Reaction Cross Sections at the Center of Sequoyah Unit 1 Surveillance Capsules 6-19 l

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List of Tables (cont)

Table Title Page 6-6 Irradiation History of Sequoyah Unit 1 Surveillance Cap-sule T 6-20 6-7 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule T 6-21

'6-8 Results of Fast Neutron Dosimetry for Capsule T 6-23

'6-9 Results of Thermal Neutron Dosimetry for Capsule T 6-24 6-10 Summary of Fast Neutron Dosimetry Results for Capsule T 6-25 viii e7aos:1 o7issa

Section 1 Summary of Results The analysis of the reactor vessel material contained in Capsule T, the first sur-veillance capsule to be removed from the Tennessee Valley Authority Sequoyah Unit 1 reactor pressure vessel, led to the following, conclusions:

Capsule T received an average fast neutron fluence (E > 1.0 MeV) of a

2.74 x 10'8 n/cm2 Irradiation of the reactor vessel lower shell forging 04 to 2.74 x 10'8 s

n/cm2 resulted in 30 and 50 ft Ib transition temperature increases of 60*

and 70*F, respect;vely, for specimens oriented normal to the major working direction of the forging and 70* and 80*F, respectively, for speci-g mens oriented in the major working directi.on.

s Weld metal irradiated to 2.74 x 10ie n/cm2 resulted in 30 and 50 ft Ib transition temperature increases of 140* and 175'F, respectively.

Comparison of the 30 ft Ib transition temperature increases for the m

Sequoyah Unit 1 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1 shows that the materials did not embrittle as much as predicted.

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Section 2 Introduction This report presents the results of the examination of Capsule T, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on Sequ,oyah Unit 1 reactor pressure vessel materials under actual operatirig conditions.

The surveillance program for Sequoyah 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 proper-ties of the reactor v%sel materials are presented by Yanichko."8The surveillance program was planneo to cover the 40-year design life of the reactor pressure ves-sel and was based on ASTM E-185-73, " Recommended Practice for Surveillance Tests for Nuclear Reactors".'23 Westinghouse Nuclear Enr.gy Systems personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes testing and the postirradiation data obtained from sur-veillance Capsule T removed from Sequoyah Unit 1 reactor vessel and discusses the analysis of these data.

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===.

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 A508 Class 2 (base material of the Sequoyah Unit 1 reactor pressure vessel beltline) are wall 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 pres-sure vessels has been presented in " Protection Against Non-ductile Failure,"

Appendix G to Section lli of the ASME Bonar and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature RTN DT-l RTNDT s defined as the greater of either the drop weight nil-ductility transition i

temperature (NDTT per ASTM E 208) or the temperature 60*F less than the 50 ft Ib (and 35-mil lateral expansion) temperature as determined from Charpy speci-mens oriented normal (transverse) to the major working direction of the material.

l The RTNOT of a given material is used to index that materid *o a reference stress l

intensity factor curve (KIR curve) which appears in Appendix G of the ASME I

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 l

be obtained for this material as a function of temperature. Allowable operating l

limits can then be determined utilizing these allowable stress imensity factors.

l RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted l

to account for the effects of radiation on the reactor vessel matarial properties.

l The radiation embrittlement or changes in mechanical properties of a given reac-tor pressure vessel steel can be monitored by a reactor surveillance program such as the Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program,l"in 67308:1 071883 3-1 l

which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average

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Charpy V-notch 30 ft Ib temperature (ARTNDT) due to irradiation is added to the NDT or radiation _ embrittlement. This adjusted f

criginal RTNDT to adjust the RT NOT nitial + ARTNDT) is used to index the material to the KIR curve i

- RTNDT (RT and,in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

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Section 4 Description of Program Eight surveillance capsules for monitoring the effects of neutron exposure on the Sequoyah Unit 1 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 41. The vertical center of the capsules is opposite the vertical center of the core.

Capsule T was removed after 1.03 effective full power years of plant operation.

This capsule contained Charpy V notch impact, tensile, and WOL specimens (Fig-ure 4-2) from the lower shell ring forging 04 and submerged arc weld metal rep-resentative of the core region of the reactor vessel and Charpy V-notch speci.

mens from weld heat-affected zone (HAZ) material.

The chemistry and heat treatment of the surveillance material are presented in

~~ ble 4-1. The chemical analyses reported in Table 41 were obtained from unir-a radiated material used in the surveillance program. In addition, a chemical analy-l sis was performed on an irradiated Charpy specimen from the weld metal and is l

reported in Table 4-1.

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

Test specimens represent material taken at least one forging thickness from the quenched end of the.~crging. Charpy specimens were machined from the forging in both the tangential (longitudinal axis of specimen perallel to the major working direction) and axial (longitudinal axis of the specimen perpendicular to the major l

working direction) orientations. Tension specimens were machined from the forg-ing with the longitudinal axis of the specimen perpendicular to the major working direction.

Charpy V-notch and tension specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding direction.

t Capsule T contained dosimeter wires of pure iron, copper, nickel, and aluminum-cobalt (cadmium-shielded and unshieloed). In addition, cadmium shielded dosim-etsrs of Np237 and U238 were contained in the capsule and located as shown in Figure 4-2.

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. Thermal monitors made from two low-melting eutectic alloys and sealed in Pymx tubes were included in the caps"le and were located as shown in Figure 4-2. The two eutectic alloys and their mourng points are:

2.5% Ag,97.5% Pb Melting Point 579'F (304'C)

~ 1.75% Ag,0.75% Sn,97.5% Pb Melting Point 590*F (310*C) -

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Figurc 4-1. Arrangement of Surveillance Capsules in the Sequoyah Unit 1 Reactor Vessel (Updated Lead Factors for Capsesses Shown in Parentheses) 43

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Table 4-1 Chemical Composition and Heat Treatment of Material for the Sequoyah Unit 1 Reactor Vessel Surveillance Program Chemical Composition (W/0)

Lot.er Shell Element Forging 04 Weld Metal C

0.17 0.058 Mn 0.62 1.40 F

0.015 0.021 S

0.016 0.009 Si 0.22 0.42 Ni 0.76 0.17 (0.08)u' Cr 0.37 0.068 Mo 0.56 0.53 Cu 0.13 0.33 (0.41-0.42)98 V

'O.01 0.002 Nr 0.005 0.007 Sn 0.015 0.023 Al 0.009 0.008 Co 0.024 0.010 i

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Heat Treatment

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Lower Shell Forging 04 1580*-1710*F,4 hr, water quenched Heat No. 980919/281587 1200*-1240*F,6 hr, furnace cooled i

1105'-1155'F, 201/2 hr, furnace cooled Weld Metal 1105'-1155'F,14 3/4 hr, l

furnace cooled

a. Results of analysis performed on irradiated Charpy specimen TWS8 anos:icnssa 4-7

9 Section 5 Testing of Specimens from Capsule T..

5-1. OVERVIEW The postirradiation mechanical testing of the Charpy V notch and tension speci-mens was performed at the Westinghouse Research and Development Labora-tory with consultation by Westinghouse Nuclear Energy Systems personnel.

Testing was performed in accordance with 10CFR50, Appendices G and H. ASTM Specification E185 79 and Westinghouse Procedure MHL 7601, Revision 3 as modified by RMF Procedures 8102 and 8103.

Upon receipt of the capsule at the laboratory, the specimens end spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8233."I.No discrepancies were found.

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

The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74,358J machine. 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 (Eo).

From the load time curve, the load of generas yielding (PGY), the time to general yielding (tGY), 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 identi-fied as the fast fracture load (Pp), and the load at which fast fracture terminated is identified as the arrest load (P )-

A 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 roughly equivalent to the energy required to initiate a crack in the specirnen. Therefore, the propagation energy for the crack (E ) is the difference between the total p

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

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The yield strese (cry) is calculated from the three-pent bend formula. The flow

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

Percent shear was determined from postfracture photographs using the ratio of-areas methods in compliance with ASTM Specification A370-77. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests were performed on a 20,000-pound lastron, split-console te' machine (Model 1115) per ASTM Specifications E8-81 and E2179, and R' : Pro-

. cedure 8102. All pull rods, grips, and pins wdre made of Inconel 718 hardened to Rc 5. The upper pull rod was connected through a universal joint to improve axi-4 p

ality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute thraughout the test.

Deflection measurements were made with a linear varidble displacement trans.

. ducer (LVDT) extensometer. The extensometer knife edges were spring loaded to the specimen and operated through specimen failure. The extensometer gage

~ 1:ngth is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67.

p 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 cach end of the gage section of a dummy specimen and in each grip. In test con-

figuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the rcnge 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 elonga-tion 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 moea ones

5-2. CHAAPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials contained in Capsule T irradiated at 2.74 x 10'* n/cm3 are presented in Tables 5-1 through 5-4 and Figures 5-1 through 5-4. A summary of the transition tempera-ture increases and upper shelf energy decreases for the Capsule T material is shown in Table 5-5.

The vessel lower shell forging 04 material specimena oriented normal to the major working direction (axial orientation) resulted in 30 and 50 ft Ib transition temperature increases of 60* and 70*F, respectively, as shown in Figure 51 after irradiation to 2.74 x 10 n/cm. No decrease in upper shelf energy appeared to a

occur due to irradiation.

Lower shell forging 04 material specimens oriented in the major working direc-tion (tangential orientation) resulted in 30 and 50 ft Ib transition temperature 4

increases of 70* and 80*F, respectively, as shown in' Figure 5-2 after irradiation to 2.74 x 10 n/cm2. Irradiation resulted in an 18 ft Ib decrease in upper shelf j

energy.

Weld metal specimens irradiated to 2.74 x 10 n/cm2 as shown in Figure 5-3 resulted in 30 and 50 ft Ib transition temperature increases of 140* and 175'F, respxtively. An upper shelf energy decrease of 33 ft Ib resulted due to irradiation.

Weld HAZ material specimens irradiated to 2.74 x 10 nicm2 resulted in 30 and 50 ft Ib transition temperature increases of 45' and 70*F, respectively, and an i

upper shelf energy decrease of 18 ft Ib as shown in Figure 5-4.

i The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8. The specimens for each material show a more ductile or tough appearance with increasing test temperature.

Figure 5-9 shows a comparison of the 30 ft Ib transition temperature increases for the various Sequoyah Unit 1 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99 Revision 1?'

The increase in transition temperature (.1RT i

NDT n degrees Fahrenheit) of the irradiated weld metal. which contains 0.33 weight percent copper, was less than predicted by the Regulatory Guide. The increase in transition temperature of the irradiated base metal, containing 0.13 weight percent copper, was also less than Regulatory Guide prediction. From these results, the Sequoyah Unit 1 reactor vessel materials irradiated to 2.74 x 10 n/cm2 did not embrittle as much as pre-j l_

dicted by the methods of Regulatory Guide 1.99 Revision 1.

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i Table 5-1 Charpy impact Data for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging 04 (Irradiated to 2.74 x 10'* n/cm2)

Specimen Temperature impact Energy Lateral Expansion Shear No.

('C)

(*F)

(J)

(ft Ib)

(mm).

(mils)

(%)

Axial Orientation VT-59

-46

- 50 6.0

' 4.5 0.05 2.0 1

VT 58

-7 20 18.5 13.5 0.32 12.5 3

VT-57 26 78 41.5

'30.5 0.76 30.0 29 VT-54 38 100 36.0 26.5 0.74 29.0 24 VT-51 52 125 67.0 49.5 1.21 47.5 55 VT-56 52 125 39.5 29.0 0.89 35.0 53 VT-60 66 150 68.5 50.5 1.16 45.5 49 VT-55 93 200 64.5 47.5 1.16 45.5 81 VT-53 121 250 92.0 68.0 1.65 65.0 100 VT-52 163 325 95.0 70.0 1.83 72.0 100 VT-50 204 400 108.0 79.5 1.91 75.0 100 VT-49 232 450 93.5 69.0 1.65 65.0 100 Tangential Orientation VL-35

- 46

- 50 6.0 4.5 0.13 5.0 1

VL 37

-7 20 47.5 35.0 0.71 28.0 16 VL-33 10 50 50.0 37.0 0.76 30.0 22 VL-39 26 78 65.0 48.0 1.07 42.0 39 VL-34 38 100 74.0 54.5 1.63 64.0 49 VL-40 66-150 121.5 89.5 1.82 71.5 69 VL-36 121 250 134.0 99.0 2.06 81.0 100 VL 38 177 350 131.5 97.0 2.11 83.0 100 i

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Table 5-2 Charpy impact Data for Sequoyah Unit 1 Reactor Vessel Weld Metal and HAZ Metal (irradiated to 2.74 x 10" n/cm2)

Specimen Temperature impact Energy Lateral Expansion Shear No.

('C)

('F)

(J)

(ft Ib)

(mm)

(mils)

(%)

Wold Metal.

TW-54

- 46

- 50 21.0 15.5 0.15 6.0 1

TW-53

-7 20 40.0 29.5 0.62 24.5 16 TW 59 10 50 41.5 30.5 0.67 26.5 31 TW-51 26 78 36.5 27.0 0.58 23.0 35 TW-58 38 100 61.0 45.0 0.94 37.0 44 TW-60 52 125 64.5 47.5 1.14 45.0 53 TW 56 66 150 79.5 58.5 1.38 54.5 62 TW-50 93 200 93.5 69.0 1.66 65.5 85 l

TW 55 121 250 105.0 77.5 1.78 70.0 100 TW-49 177 350 100.5 74.0 1.96 77.0 100 l

TW-52 232 450 110.5 81.5 1.80 71.0 100 l

HAZ Metal TH 53

- 46

- 50 23.5 17.5 0.23 9.0 10 TH-51

- 18 0

34.5 25.5 0.37 14.5 22 TH-55

-7 20 43.5 32.0 0.51 20.0 19 TH-5?

-7 20 45.5 33.5 0.60 23.5 34 TH 56 10 50 61.0 45.0 0.86 34.0 62

- TH 50 10 50 55.0 40.5 0.81 32.0 47 TH-58 26 78 73.0 54.0 1.08 42.5 90 TH-49 38 100 102.5 75.5 0.99 39.0 72 TH-57 66' 150 93.5 69.0 1.40 55.0 100 TH-59 93 200 74.5 55.0 1.32 52.0 100 TH-60 121 260 112.0 82.5 1.61 63.5 100 TH-54 177 350 93.5 69.0 1.46 57.5 100 i

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5-5

Table 5-3 Instrumented Charpy impact Test Results for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging 04 Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maalmum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress No.

( C)

(J)

(kJ/m )

(kJ/m8)

(kJ/m')

(N)

(ps)

(N)

(ps)

(N)

(N)

(MPa) (MPa) 2 Asial Orientasioen VT59

-46 60 76 52 24 15100 115 15100 200 V158

-7 18 5 229 145 84 14800 100 16100 200 15900 0

762 795 VI57 26 41.5 517 376 141 15000 115 18200 435 17900 700 770 853 VT54 38 36 0 449 310 139 13800 147 16700 404 16700.

3300 711 786 Y

Vi56 52 39 5 491 228 264 13400 105 15300 310 15300 4700 390 739 V151 52 67.0 839 420 419 12500 105 17300 515 15700 4600 e42 765 TV60 66 68 5 856 431 425 12800 110 17000 525 16000 4700

.i59 766 V155 03 64 5 80u 249 556 12000 105 14400 365 615 678 V153 621 92 0 1152 427 726 13000 130 17200 525 670 778 V152 163 95 0 1186 360 826 10300 100 14800 510 528 644 VISO 204 108 0 1347 412 936 11900 575 15400 585 613 702 VT49 232 93 5 1169 368 801 10500 125 14700 525 538 646 Tangential Orientation

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VL35

- 46 60 76 53 23 14100 105 14400 110 14400 200 727 734 VL37

-7 47.5 593 454 139 14800 110 18500

'510 18500 0

764 859 VL33 10 50 0 627 SF4 73 15200 147 19000 607 18900 500 781 880 VL39 26 65 6 813 644 169 14900 125 19800 680 19200 1200 76 7 892 VL34 38 74 0 924 474 450 14900 115 18600 525 17100 12100 768 863 VL40 66 121 5 1517 589 928 13300 100 17900 670 12700 6100 683 801 VL36 121 1340 1678 544 1134 13000 140 16600 700 668 762 I

O VL38 177 131.5 1644 543 1101 11800 120 16600 685 608 730 89 52 B

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g 9

E Table 5-4 I

instrumented Charpy impact Test Results for Sequoyah Unit 1 Weld and HAZ Material Normalized Energies Test Charpy Charpy Maximum Prop Yield Time Maximean Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Lead Masimum Load Lead Stress Stress No.

[Cl (J)

(kJ/m')

(kJ/m8)

(kJ/m')

(N)*

(ps)

(N)

(ps)

(N)

(N)

(MPa) (MPa)

Eld Metal

~'

TW54

- 46 21 0 263 241 21 16300 110 18300 280 18300 200 840 891 TW53

-7 40 0 500 457 43 14800 10S 18100 510 18100 0

760 846 IW59 10 41.5 517 476 41 15500

Ill, 18700 525 18700 100 796 879 TW51 26 36 5 458 357 100 14600 155 17900

• 430 17900 4800 753 836 TW58 38 61 0 763 613 149 14400 147 18500 680 18000 1400 743 849

?

TW60 52 64 5 805 491 314 13500 115 17300 580 16400 4400 693 792 N

TW56 66 79 5 991 408 583 12800 105 165M 510 14200 8000 657 753 TW50 93 93 5 1169 427 743

!2800 105 167w 525 9700 5400 661 760 IW55 121 105 0 1313 485 829 13100 150 16900 600

'672 772 TW49 177 100 5 1254 485 769 12500 147 16300 625 643 742 IW52 232 110 5 1381 465 916 9900 105 14300 66C 508 621 HAZ Metal Til53 46 23.5 297 212 84 14000 110 15800 290 15700 9

720 769 18 t51 18 34 5 432 343 89 16300 136 19000 384 18700 3100 837 908 18t52 7

45 5 b68 367 201 14900 100 18300 410 18300 5800 766 854 18155 7

43 5 542 485 57 15900 105 19600 505 10600 100 821 914 TitSO 10 55 0 686 288 398 14400 100 16700 355 15900 9700 741 801 Til56 10 61.0 7G3 383 380 15300 125 18300 440 18100 9700 789 865 18158 26 73 0 915 394 521 14>00 125 17900 455 15800 12200 758 839 Til49 38 102 5 1280 559 720 12300 95 17500 655 13400 3400 630 766 18 t57 66 93 5 1169 418 752 13300 110 16900 505 686 778 lits 9 93 74 5 932 327 606 12800 100 15800 415 659 731 11160 121 112 0 1398 582 816 13200 150 17700 680 679 795 18154 177 93 5 1169 380 790 1:100 105 15700 510 571 690 i

5396A 3 TEMPERATURE (OC)

-100

-50 0

50 100 150 200 250 120 l

l l

l 3 l 3 l

l l

4% GQ--4-4 100

/

6 4

G 3

80 M

a y

/.

m e

20 0

0 2.5 100 2.0 2

80 0-1.5 60 2

2m2 i

=

1.0 m/

40 a

40o g-H 0.5 20 0

0.0 90 120 O

g 80

~

2 gg g

e 70 g

UNIRR ADI ATED O

80 l

"j 60

>hg 5

0 70 F 50 60 3

0 IRRADIATED (550 F) e 40 18 2

2.74 X 10 n/cm 40 0

30 60 F g

20 O

20 10 Of 0

0

-200

-100 0

100 200 300 400 500 TEMPER ATUR E (OF)

Figure 5-1. Charpy.V-Notch Impact Data for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging (Axial Orientation) 5-8

5336A4 TEMPERATURE (CC)

-100

-50 0

50 100 150 200 250 l

l l

l 3 l3 l

l l

,=

.-p-.

2 g

M

~

9 a

$m

%*/

/

/g.S 3

20 0

9 100 2.5 G

G 80 2.0 O

E 60 1.5 40 O

4_33op 1.0 g

5 20 0.5 0

0.0 j

160 200 140 0

120 n

n O

160 I

UNIRRADIATED O

O O

=

1#

e.G e

9

~

120 y

80 3

m 0

IRRADIATED (550 F) 80 W

60 0

O 18 80 F

• Q 2.74 X 10 n/cm2 40 0

l 70 F

.g; l

20 :

2 0

0 I

O l

-200

-100 0

100 200 300 400 500 TEMPERATURE (OF)

-Figure 5-2. Charpy V-Notch' Impact Data for Sequoyah Unit 1 Reactor

> Vessel Lower Shell Forging (Tangential Orientation) l 5-9 l

l

1396A-5 1

TEMPERATURE (OC)

~100

-50 0

50 100 150 200 250 120

- l 3

l l

l I

I I

20; O

i=

3 R

80 O

/

e'9'.

=

w 5

40 g.S 2

e/g/

20 O

2.5 100 2

2.0

,80 O

g e

/g-e 1,5 7

60 g

O g*g*g I

r 1.0 m

40 0

g 150 F

• 0.5 5

20 0.0 0

740 160 8

g 120 i

UNIRRADIATED g

c j

l f!

100 120 i

~

g^

e i

g3,9 80 60 l

g 175,F 0

IRRADIATED (550 F) i 2w 40 y

F 2.74 x 1018 n/cm2 40

, j e g + 14 l

20 O

0

-200

-100 0

100 200 300 400 500 TEMPERATUR E (CF)

Figure 5-3. Charpy V-Notch impact Data for 'Sequoyah Unit 1 Reactor Vessel Weld Metal 5 10

5396A o TEMPERATURE (OC)

-100

-50 0

50 100 150 200 250 120 l

l I

I I

I I

I ia 99 g,

3 3

80 6

O e

60

1. ' b 5

40 98

1. 9 20 O/

0 100 2.5 3

80 Og#

2.0 O

'E O4g 60 g

1.5 40 65 F g* O

~

0 ar I

20 0.5 g.

0 0 52 140 120 UNIRRADIATED 160 l

O8 g

100 120 l

7 O

BO*

80 3

l c

O m

60 70 F O

O 80 0

O 40

,9 0

0 45 F IRRADIATED (550 F) 40 20 2.74 X 1018 n/cm2 0

0

-200

-100 0

100 200 300 400 500 TEMPERATURE (CF) l l

Figure 5-4. Charpy V-Notch impact Data for Sequoyah Unit 1 Reactor Vessel Weld HAZ Metal 5-11

Table 5-5 The Effect of 550'F Irradiation to 2.74 x 10 n/Cm2 (E > 1.0 MeV)

On the Notch Toughness Properties of Sequoyah Unit 1 Reactor Vessel Materials

~

Transition Temperature d

Unitradiated Irradiated a Transition Temperature Average Energy Absorpiton 50 ft Ib 30 f t Ib 35 mils 50 ft Ib 30 f t ib 35 mils SGf\\ab 30 MIb 35 mils at Fut) She.r 68 J 41 J C.9 m m 68 J 41 J 0.9 mm 68J 41 J 0.9 mm Unirra-Irradiated a Energy cn diated i

Material ( C) ( F) ( C) ( F) { C) { F) { C) (' F) l C) l'F) ('-C) ('F) ('Cl (~F) ( C) (' F) (-C) ( F) (J) (ft Ib) (J) (ft %) (J) (f t $b) fu Foruing 04 40 105 7

4 ',

27 80 79 175 40 105 49 120 39 70 33 60 22 40 98 72 98 '/2 0

0 (Axial)

Forging 04 18 0

37 35 -23

-10 27 80 2 35 7 45 45 80 39 70 30 55 157 116 133 98 24 18 (Tangen-tial)

Weld Metal 46 50 65 85 - 51

-60 51 125 13 % 32 90 97 175 78 140 83 150 151 111 106 71 45 33 l

HAZ 20 5

34 30 20

-5 19 65 3 15 16 60 39 70 25 45 36 65 117 86 94 69 23 17

$u 5

E 9

h I

4 I

l 5396A 7 2

!g N

gg ow n y

VT59 VT58 VT57 VT54 VT51 VT56 l

VT60 VT55 VT53 VTS2 VT50 VT49 Figure 5-5. Charpy impact Specimen Fracture Surfaces for Sequoyah Unit 1 Lower Shell Forging 04 (Axial Orientation) 5-13

_.. ~

l I

i i

i 5396A-8 i

VL35 VL37 VL33 VL39 l

r i

VL34 VL40 VL36 VL38 Figure 5-6. Charpy impact Specimen Fracture Surfaces for Sequoyah Unit 1 Lower Shell Forging 04 (Tangent!al Orientation) 5-14 l

4 L

.i 5396A 9 I

i I

l P

B.!*.-

d1;.(N

?I$

ah

~

Pr[&=

~

,, s;+cy;*%k$

a 3+1 TW54 TW53 TW59 TW51 TW58 TW60 I

TW56 TW50 TW55 TW49 TW52 l

Figure 5-7. Charpy Impact Specimen Fracture Surfaces for Sequoyah l

Unit 1 Wald Metal 5-15

4

l 5396A 10

~

1 i

TH53 THS1 TH55 TH52 TH56 TH50 l

A TH58 TH49 TH57 TH59 TH60 TH54 l

l l

l t

Figure 5-8. Charpy impact Specimen Fracture Surfaces for Sequoyah Unit 1 Weld Heat Affected Zone Metal 5-16

5336A 11 l

103 8 2

~

- 4 6

_u.

_O o_

o_

W 4 - WELD METAL W

g PREDICTION 2

y

=

- (0.33* Co.

e U

0.021% P)

U z

z

~

~

• g 102 g

k 1

8 3

4m 4*

102 6

w g

k s

w 8

FORGING 04 w

H O

PREDICTION 4

~

6 G

(0.13% Co,0.015% P) r p

i;i O

G l

z 4

z l

4 2

m c

H LEGEND:

E A WELD METAL E

A HAZ METAL 2

O FORGING 04 (TANGENTIAL) 101 9 FORGING 04 (AXIAL) 101 I

I I I I

I I

I

2 4

6 8 1019 2

4 6

8 1020 1018 2

FLUENCE (n/cm )

l l

l Figure 5-9. Comparison of Actual Versus Predicted 30 ft Ib Transition Temperature Increase for Sequoyah Unit 1 Reactor Vessel Materials Using the Prediction Methods of Regulatory i

Guide 1.99 Revision 1 l

5-17 i

5-3. TENSION TEST RESULTS The results of tension tests performed on forging 04 (axial orientation) and the weld metal irradiated to 2.74 x 10" n/cm2 are shown in Table 5-6 and Figures 510 and 5-11, respectively. These results show that irradiation produced an increase in 0.2 percent yield strength of 5 to 8 ksi for the forging 04 material and approxi-mately 12 ksi for the weld metal. Fractured tension specimens for each of the materials are shown in Figure 5-12. A typical stress-strain curve for the tension specimens is shown in Figure 5-13.

5-4. WEDGE OPENING LOADING TESTS Test results for wedge opening loading (WOL) fracture mechanics specimens contained in Capsule T will be reported at a later time.

5-18 e7 m e ses

k

~

m r

E i

Table 5-6 Tensile Properties for Sequoyeh Unit 1 Reactor Vessel Meterial Irradiated to 2.74 x 10 n/can2 Test Viead Ultinnate Fracteste Fracture Fracture Ued5rsue

' Total Rodeaction Samp4e Temp Strength Strength Lead Strees Strengt8: Elongation Eierweien in Area No.

Asatorial "C(F)

RSPalksil RSPe{ksil N(liip)

AAPallisil ASPalksil

(%)

(*al t%)

Eft VT10 Forging 04 107 446 583 13.600 850 428 11.0 18.9 49 (Amial)

(225)

(64.7)

(84 5)

(3 05)

(122 61 (62.1) a to VT9 Forging 04 28a 414 587 16.500 720 520 S.3 11.4 27 (Amial)

(550)

(60.11 (85 2)

(3.70)

(103 8)

( 6.4) i TW10 Weld 107 517 615 14.300 1260 452 10.3 21.2 64 Metal (225)

(750)

(89.2)

(3.22)

(182.21 (65.6) 1 TW9 Weld 288 488 612 16.100 1000 509 94 17.1 49 Metal (550)

(70 8)

(88 81 (3 63)

(145 7)

(738) i

?

f

5396A 12 TEMPERATURE ( C) 0 50 100 150 200 250 300 350 120 I

I l

l I

l l

l-800 700 100 ULTIMATE TENSILE STRENGTH g

A 600 80 O

500 x

a e

-9 400 60 0.2% YlELD g

STRENGTH 300 l

l LEGEND:

i 0 6 OPEN POINTS - UNIRRADIATED e A CLOSED POINTS -IRRADIATED AT 2.74 X 1018 n/cm2 80 REDUCTION IN AREA 60 C

o

-e e

4 40 TOTAL ELONGATION o

UNIFORM ELONGATION g

20 G

Aa o

=

j q b

U b

y

_A w

0 0

100 200 300 400 500 600 700 l

TEMPER ATURE (OF) l Figure 5-10. Tensile Properties for Sequoyah Unit 1 Reactor Vessel Lower Shell Forging (Axial Orientation) l l

5-20

$396A 13

~

TEMPERATURE (OC) 0 50 100 150 200 250 300 350 120 j_

\\

I"

- ULTIMATE TENSILE STRENGTH

~

i

'f g

%0

_A

/

2 80 500 C

~

Q MO

- 0.2% YlELD STRENGTH 300 40 LEGEND:

OPEN POINTS - UNIRRADIATED CLOSED POINTS -IRRADIATED AT 2.74 X 1018 n/cm2 80 2

REDUCTION IN AREA e0 E

e tg 40 TOTAL ELONGATION 20 e

c g

g M

UNIFORM ELONGATION e--O v

i l

l I

I I

o 0

100 200 300 400 500 600 700 TEMPERATURE (OF)

Figure 5-11. Tensile Properties for Sequoyah Unit 1 Reactor Vessel Weld Metal 5-21

i i

5396A 14

--s

.... ~,. _.

T Y f

c. e ;W.'?p.

r 3-f t-J's e s 6.3 ic 3

w t

1:'M. p.

.j.,

2

't e 4 a A I ' 'y 5 "

I i 11 i1 ! ' ;~

i

, he VT10 2250F i

' 7 :.:/ v n ? ~~g., '. a.

N i

e-

,4 n r.w

. 3:

c, x 2 v.a j< -

7 J.

u J

Id,. :,.

%*J..s,e'o ?,e.. s a

.,4 !. i-

~

s j.. 5

. e i:,1g I

9 e

i.a-e a u

-4 VT9 5500F ymu c g. 4,m reer:--

y w * >._,,,

.g

=

.A i : s :,, e-5...

as-3'JGlHi _> ' * " +yC s ?,

• ',-}:

Li t.lil < tJ. I._.lnlI.

_ :qnt :

,),'

.,'. f C,' 7,4 ; b.' E 11 d.3 -

-A i

E;.

TW10 2250F 7:_

,,,,- r n,7,-. r y%>m t,-

-* s n:. ;.

I

• hg l

e m

i

sawe ;:x na k.y,we J

@192;; ~ '" '

Figure 5-12. Fractured Tension Specimens From Sequoyah Unit 1 Lower Shell Forging 04 and Weld Metal l

5-22 iL

120 800 f

105 -

J 90 --

600

'I r

. 75 p*

l

}

l 1

W 60 E

4M $

~

F

• p-vs O

45-30 200 4

15 SPECIMEN TW10 (225 F) 0 l

l l

l 0

0 0.000 0.025 0.050 0.075 0.100 0.125 0.150 0.175 0.200 0.225 0.250 1

STR AIN (in./in.)

5 Figure 5-13. Typical Stress. Strain Curve for Tension Specimens s

9 i

Section 6 Radiation Analysis and Neutron Dosimetry 6-1. INTRODUCTION KnowleQe of the neutron environ' ment within the pressure vess91 - surveillance capc:.:le geometry is required as an integral part of LWR pressure vessel surveil-lance programs for two reasons. First, in the interpretation of radiation induced properties changes observed in materials test specimens, the neutron environ-ment (fluence, flux) to which the test specimens were exposed must be known.

Second, in roiating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the i

environment at various positions within the reactor vessel and that experienced by the test specimens must be established. The former requirement is normally met,by employing a combination of rigorous analytical techniques and measure-ments obtained with passive neutron flux monitors contained in each of the sur-veillance capsules. The latter information is derived solely from analysis.

This section describes a discrete ordinates Sn transport analysis performed for the Sequoyah Unit 1 reactor to determine the fast neutron (E > 1.0 MeV) flux and fluence as well as the neutron energy spectra within the reactor vessel and sur-veillance capsules and, in turn, to devc!ap lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum averaged reaction cross sections derived from this calculation, the anal-ysis of the neutron dosimetry contained in Capsule T is discussed and compari-sons with analytical predictions are presented.

6-2. DISCRETE ORDINATES ANALYSIS A plan view of the Sequoyah Unit 1 reactor geometry at t% core midplane is shown in Figure 6-1. Since the reactor exhibits 1,8th core symmetry only a zero-to 45-degree sector is depicted. Eight irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance pro-gram. Four capsules are located symmetrically at 4 and 40 degrees from the car-dinal axes as shown in Figure 6-1.

e7 ace:i o7:sss 6-1

5396A 16

~

00 4o -

CAPSULES f

l (S,V,W,Z

/

NE4Cro 1[ CAPSULES [

I E4gz.

j T. u, x, y q 4oo V///

l 45o NEg l

4tgz j

S

/

us////,,,,,,,,,,,

I

/

/ 3 t

/ V

/

/

/

/

,/ j '

l

./ /

//

//

/'

//

//

/

,L Figure 6-1. Sequoyah Unit 1 Reactor Geometry 6-2

l L

A plan view of & single surveillance capsule attes:hed to the thermal shield is shown in Figure 6-2. The stainless steel specimen container is 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 neutronic standpoint, the surveillance capsule structures are significant.

In fact, as is shown later, they have a marked impact on the distributions of neu-tron flux and energy spectra in the water annulus between the thermal shield and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimen locations, the capsules theniselves must be included in the analytical model. Use of at least a two-dimensional computation is therefore mandatory.

In the analysis of the neutron enviroriment within the Sequoyah Unit 1 reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOTW two-dimensional discrete ordinates code. The radial and aximu-thal distributions were obtained from an R,8 computation wherein the geometry shown in Figures 6-1 and 6-2 was described in the analytical model, in addition to the R,9 computation, a second calculation in R,Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R,Z analysis, the reactor core was treated as an equivalent volume cylinder and, of course, the surveillance capsules were not included in the model.

Both the R,6 amd R,Z analyses employed 47 neutron energy groups and a P3 expansion of the scattering cross sections. The cross sections used in the anal-yses were obtained from the SAILOR cross section library

• which was developed specifically for light water reactor applications. The neutron energy group struc-ture used in the analysis is listed in Table 6-1.

A key input parameter in the analysis of the integrated fast neutron exposure of

- the reactor vessel is the core power distribution. For this analysis, power distribu-tions representative of time-averaged conditions derived from statistical studies of long-term operation of Westinghouse'4-loop plants were employed. These input distributions include rod-by-rod spatial variations for all peripheral fuel assemblies.

enos:iones:

6-3

L 5396A-17 0

(40 OR 40 )

0 0

(3 OR 39 )

y CHARPY SPECIMEN

/

- x m

THL.: '4AL SHIELD Figure 6-2. Plan View of a Reactor Vessel Surveillance Capsule I

6-4

. J

J Table 6-1 47 Group Energy Structure Lower Energy J.ower Energy Group (MeV)

Group (MeV) 1 14.19

- 25 0.183 2

12.21 26 0.111 3

10.00 27 0.0674 4

8.61 28 0.0409 5

7.41 29 0.0318 6

6.07 30 0.0261 7

4.97 31 0.0242 8

3.68 32 0.0219 9

3.01 33 0.0150 10 2.73 34 7.10x10 -'

11 2.47 35 3.36x10-3 l

12 2.J7 36 1.59x10 - 3 l

13 2.35 37 4.54x 10 -*

i 14 2.23 38 2.14x 10 -*

15 1.92 39 1.01 x 10 -

• 16 1.65 40 3.73x10 - 5 17 1.35 41 1.07x10 - 5 18 1.00 42 5.04x 10 -*

19 0.821 43 1.86x10 -8 20 0.743 44 8.76x 10 -'

21 0.608 45 J.14x10-7 22 0.498 46 1.00x10- 7 23 0.369 47 0.00x 24 0.298 l

a. The upper energy of group 1 is 17.33 Mev.

e730s:107issa 6-5 1'

'It should be noted that this generic design basis power distribution is intended to provide a vehicle for long term (end-of-life) projection of vessel exposure. Since

' plant specific power distributions reflect only past operation, their use for projec-

' tion into the future may not be justified; the use of generic data which reflects 11::ng-term operation of similar reactor cores may provide a more suitable approach.

Benchmark testing of these generic power distributions and the SAILOR cross sections against surveillance capsule data obtained from 2 loop and 4-loop West-

. inghouse plants indicates that this analytical approach yields conservative results

.. th calculations exceeding measurements from 10 to 25 percent.*

One further point of interest regarding these analyses is that the design basis assumes an out-in fuel loading pattern (fresh fuel on the periphery). Future com-mitment to Idw leakage loading patterns could significantly reduce the calculated

neutron flux levela presented in paragraph 6-4. In addition, capsule lead factors could be changed, thus impacting the withdrawal schedule of the remaining sur-veillance capsules.

Having'the results of the R,0 and R,Z calculations, three-dimensional variations of

~

_ neutron flux may be approximated by assuming that the following relation holds for'the applicable regions of the reactor.

(6-1)

' 4(R,Z,0E )g = 4(R,0E ) F(Z,E )

g g

' where 6(R,Z,0,E ) = neutron flux at point R,Z,0 within energy group g g

6(R,0,E ) - = neutron flux at point R,0 within energy group g obtained from g

the R,0 calculation F(Z,E )

= relative axial distrib'ution of neutron flux within energy group g g

obtained from the R,Z calculation 6-3. ~ NEUTRON DOSIMETRY The passive neutron flux monitors included in Capsule T of Sequoyah Unit 1 are listed in Table 6-2. The first five reactions in Table 6-2 are used as fast neutron monitors to relate neutron fluence (E > 1.0 MeV) to measured materials proper-ties changes. To properly account for burnout of the produ t isotope generated by fast neutron reactions,it is necessary to also determine the magnitude of the thermal neutron flux at the monitor location.-Therefore, bare and cadmium-cov-ered cobalt-aluminum monitors were also included.

6-6 usoea onens t

Table 6-2 Nuclear Constants for Neutrc9 Flux Monitors Contained in the Sequoyah Unit 1 Surveillance Capsules Target Fission Weight Product Yield Monitor Material Reaction of Interest Fraction Half-Life

(%)

Copper Cu" (n,ca) Co" 0.6917 5.27 years Iron Fe5* (n,p) Mn5' O.0585 314 days Nickel NiS8 (n,p) CoS8 0.6777 71.4 days Uranium-238tal Um (n,f) Cs'"

1.0 30.2 years 6.3 Neptunium-237tal Npm (n,f) Cs'"

1.0,

30.2 years 6.5 Cobalt-aluminumI*3 Co58 (n,y) Co" 0.0015 5.27 years Cobalt-aluminum Co" (n,y) Co" 0.0015 5.27 years

a. cenotes that monitor is cadmium-shielded The relative locations of the various monitors within the surveillance capsule are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule.

The use of cassive monitors such as those listed in Table 6-2 does not yield a direct measure of the energy-dependent flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the variou.i monitors may be derived from the activation measurements only if the irradiation parameters are well known in particular, the following variables are of interest.

a The operating history of the reactor s

The energy response of the monitor a

The neutron energy spectrum at the monitor location a

The physical characteristics of the monitor The analysis of the passive 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.

67308:1 071883 6-7

. = -.

~

EThe specific activity of each of the monitors is determined using established ASTM procedures?*Sao"' Following sample preparation, the activity of each

. monitor is determined by means of a lithium-drifted germanium, Guili), gamma spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing,'the uncertainty in counting, ano the ac-

cnptable error in detector calibration. For the samples removed from Sequoyah

~ Unit 1, the overall 20 deviation in the measured data is determined to be ; as or minus 10 percent. The neutron energy spectra are determined analytically using tha method described in paragraph 6-1.

HEving the measured activity of the monitors and the neutron energy spectra at

.tha locations of interest, the calculation of the neutron flux promeds as follows.

The reaction product activity in the monitor is expressed as i

I R=

f; y o(Ele (E)dE (1-e - Ati),- Atd (6-2)

A j.1 Pmax wh'ere r.

= induced product activity -

N

= Avogadro's number o

A

= atomic weight of the target isotope f;

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

= number of product atoms prcduced per reaction

-o(E)

= energy dependent reaction cross section e(E).= energy dependent neutron flux at the monitor location with the

.. reactor at full power Pj

= average wre' power. level during irradiation period j Pmax = maximum or reference core power level A

= decay constant of the product isotope -

tj

= length of irradiation period j td

= decay time following irradiation period j 8:cause neutron flux distributions are calculated using multigroup transport methods and, further, because the p-ime interest is in the fast neutron flux above

' 1.0 MeV, spccuum-averaged reaction cross sections are drJined such that the integral term in equation (6-2) is replaced by the following relation.

of E)6(E)dE = a 6 (E > 1.0 MeV)

. 8 ence:i onses i

where N

• <r(El4(E)dE I

erg g

~/~

e o

g=1.

o=

=

/*1.0 MeV

- 4(E)dE N

I 6g 9 " 91.0 MeV Thus, equation (6 2) is rewritten N

N P R=d f Y & 4 (E > 1.0 MeV) I (1.e-Ati),- Atd i

A

.j=1Pmax or, solving for the neutron flux, R

4(E > 1.0 MeV) =

5 fY&

(1-e - Atj),- Atd (6-3) i A

j=1 Pmax l

The total fluence above 1.0 MeV is then given by l

n P-1

$(E > 1.0 MeV) = 4(E > 1.0 MeV) I tj (6-4) j=1 Pmax where n

p.

1 total effective full power seconds of reactor operation up to 3

t j=1 Pmax i = the time of capsule remova!

(

87308:5071883 69 i

1

' An assessment of the thermal neutron flux levels within the surveillance capsules i3 obtained from the bare and cadmium-covered Co" (n, f) Coa data by means of cadmium ratios and the use of a 37-barn,2,200 m/sec cross section. Thus.

R are D_.1 b

_ D.,

6Th "

n p f Ye I (1-e-Atjg g-Atd i

A j=1 Pmax where D is defined as R are/ rcd covered-b 3

6-4. TRANSPORT ANALYSIS RESULYS Results of the Sn transport calculations for the Sequoyah Unit 1 reactor cre sum-marized in Figures 6-3 through 6-8 and in Tables 6-3 through 6-5. In Figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule center-line, pressure vessel inner radius,1/4 thickness location, and 3/4 thickness loca-tion are presented as a function of azimuthal angle. The influence of the surveillance capsules on the fast neutron flux distribution is clearly evident. In Figure 6-4, the radial distribution of maximum fast neutron flux (E > 1.0 MeV)

. through the thickness of the reactor pressure vessel is shown. The relative axial variation of neutron flux within the vessel is given in Figure 6-5. Absolute axial variations of fast neutron flux may be obtained by multipl*;ing the levels given in Figure 6-3 or 6-4 by the appropriate values from Figure 6-5.

t 6-10 anos;i otissa

saswa 1011 SURVEILLANCE 8

CAPSULES R = 211.41 cm 6

4 PRESSURE VESSEL IR 1

2 1/4T LOCATION 3.

X 10 3 10

. u.

8 g

N 6

8

=

4 3/4T LOCATION i

2 108 0

10 20 30 40 50 60 AZIMUTHAL ANGLE (deg)

Figura 6-3. Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel-Surveillance Capsule Geometry 6-11

5396A.19 11 10 8

- 6,.

4 219.71 l

l j

IR 225.19 2

l c

E i

I Q

1/4T 10 3 10 8

g m

6 g

w 236.14 z

4 l

241.62 2

9 10 214 216 218-220 222 224 226 228 230 232 234 236 238 240 242 RADIUS (cm)

Figure 6-4. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 6-12

> + < - ~ -

w--

,y,---,w

,-w.

5396A.23

~

0 10 8

'6 4

2 y 10-1 d

8 2

6 8

g 4

((

y 2

j P

J

~

$ 10-2 8

6 4 -

CORE MIDPLANE 2 -

TO VESSEL CLOSURE HEAD 10-3

-300

-200

-100 0

100 200 300 400 OlSTANCE FORM CORE MIDPLANE (cm)

Figure 6-5. Relative Axi.nl Variation of Fast Neutron Flux (E > 1.0 MeV)

Within the Pressure Vessel 6-13

5336A 21 1012 8

6 4 -

2 -

211.41 0

40 CAPSULES 1011 T

8 8

N O

Ex

-cy 4

40 CAPSULES 3

=

z b

CAPSULE CENTER wz 10 10 8

6 4

2 THERMAL l f $ %

TEST SPECIMEN

N" o

SHIELD ;

3p 9

I U

I 10 207 208 209 210 211 212 213 214 RADIUS (cm)

Figure 6-6. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Surveillance Capsule 6-14

5336A.22 i

108 8

6 4

t i

2 8

NiS8 (n, p) CoS8 6 [

Np237 (n, f) Cs137 l

l O.

4

.=

D N

2 3;

211.41 Pc4 107 8

U238 (n, fl Cs137 Q

6 p,54 (n, p) MnN m

l 5

4 5

2 CAPSULE CENTER 106 8

l 2

CuS3 (n, a) Co60 4

THERMALj l$l5lh TESTSPECIMENS 2

SHIELD :

5 5

j' l lC5 (5lll l

l EW W53:

10

,e 207 208 209 210 211 212 213 214 RADIUS (cm)

~

Figure 6-7. Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules Located at 40 Degresa t 4g 6-15

5396A-23 8

10 8

6 4

NiS3 (n, p) CoS8 2

Np237 (n, f) Csl37 7

10 8

6 211.41 O

4 Fe54 (n.p) Mn54 3

• m 2

U238 (n, f) Cs137 g

aw 106 8

O CAPSULE CENTER 6

g 4

Cu63 (n, d) Co60 2

5 10 8

6 4

2 THERMAL!

! $ hkl TEST SPECIMENS j

h l

l l

9 4

10 207 208 209 210 211 212 213 214 R ADIUS (cm)

Figure 6-8. Calculated Variation of Fast Neutron Flux Monitor Saturated Activity Within Capsules Located at 4 Degrees 6 16

.3 5336A.20 i

100 8

l 6

4 2

s 10-1 u'.

8 2

6 8

g 4

W 2

5 4a E 10-2 8

6 4

CORE MIDPLANE 2

TO VESSEL CLOSURE HEAD 10-3 l

l l

-300

-200

-100 0

100 200 300 400 DISTANCE FORM CORE MIDPLANE (cm)

Figure 6-5. Relative Axial Variation of Fast Neutron Flux (E > 1.0 MeV)

Within the Pressure Vessel 6-13

l 5336A.21 1012 8

6 -

4 -

2 -

211.41 400 CAPSULES 11 10 3

8 8

a 6

g

..e e

4 40 CAPSULES

~

x 3

u.

2 g

2 CAPSULE y

CENTER wz-10 10 8

6 4

2 THERMAL l fC k T EST SPECIMEN 3 g%

b D

SHIELD :

9 l {

] !c ;d

\\

l MC; 0

207 208 209 210 211 212 213 214 RADIUS (cm)

Figure 6-6. Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Surveillance Capsule 6-14

==n a

Table 6-5 Spectrum-Averaged Reaction Cross Sections at the Center of Sequoyah Unit 1 Surveillance C.apsules a (barns)

Reaction Capsules at 4*

Capsules at 40*

Fe5' (n,p) Mn" 0.0980 0.0735 Cu" b:p) Coa 0.00112 0.000659 Ni" (n.pi Cow 0.127 0.0993 Npm (n,f) Cs'"

2.62 2.83 U= (n,f) Cs'"

0.385 0.385

((a(E)e(EldE x 6(E)dE 1 MeV in order to derive neutron flux and fluence levels from the measured disinte-gration rates, suitable spectum-averaged reaction cross sections are required.

The neutron energy spectrum calculated to exist it the center of each of the Sequoyah surveillance capsules is listed in Table J-4. The associated spectrum-averaged cross sections for each of the fast neutron reactions are given in Table 6-5.

6-5. DOSIMETRY RESULTS The irradiation history of the Sequoyah Unit 1 reactor up to the time of removal of Capsule T is listed in Table 6-6. Comparisons of measured and calculated satu-rated activity of the flux monitors contained in Capsule T based on the irradiation history shcwn in Table 6-6 are given in Table 6-7. The data are presented as measured at the actual monitor locations as well as adjusted to the capsule cen-ter. All gradient adjustments to the capsule center were based on the data pre-sented in Figure 6-7.

snos.:oriess 6-19

Table 6-6 Irradiation History of Sequoyah Unit 1 Surveillance Capsule T Irradiation Decay PJ Pmax Pj/

Time Time Month (MW)

(MW)

Pmax (days)

(days) 10/80 18 3,565 0.005 31 803 11/C0 439 3,565 0.123 30

773 12/80 1,814 3,565 0.509 31 742 1/81 2,014 3,565 0.565 31 711 2/81 320 3,565 0.090 28 683 3/81 1,236 3,565 0.347 31 652 4/81 2.653 3,565 0.744 30 622 Se81 1,695 3,565 0.475 31 591 6/81 1,849 3,565 0.519 30 561 7/81 2,263 3,565 0.635 31 530 8/81 2,033 3,565 0.570 31 499 9/81 1,245 3,565 0.349 30 469 10/81 90 3,565 0.025 31 438 11/81 2.634 3,565 0.739 30 408 12/81 2.577 3,565 0.723 31 377 1/82 1,162 3,565 0.326 31 346 2/82 33 3,565 0.009 28 318 3/82 2,236 3,565 0.627 31 287 4/82 2,849 3,565 0.799 30 257 I

5,82 3,477 3,565 0.975 31 223 6/82 3,159 3,565 0.886 30 196 i

7/82 3,447 3,565 0.967 31 165 8/82 3,430 3,565 0.962 31 134 9/82 3,123 3,565 0.876 30 123 NOTE: (1) Decay time is referenced to 1 12133.

(2) Total irradiation time equal to 3.25 x 10' EFPS.

6-20 ence:t ortaas

-Table 6-7

. Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule T Saturated Activity Adjusted Saturated Activity Reaction Radial (disls; (dis /s) and Location 9

9 Axial Position (cm)

Capsule T Calculat,ed Capsule T Calculated Fe" (n.p) Mn" Top 211.68 3.68x10*

3.88x108 Top-Middle 211.68 3.71x108 3.91x10e Middle 211.68 3.68x10e 3.88x105 Bottom-Middle 211.68 3.74x10' 3.94x108 Bottom 211.68 3.74x108 3.94x108 Average 3.71x108 4.30x10' 3.91 x10' 4.53x108 Cu'8 (n.a) Co" Top-Middle 211.18 4.07x105 3.87x105 Middle 211.18 4.06:105 3.86x105 Bottom-Middle 211.18 4.12x105 3.92x105 Average 4.08x105 4.32x105 3.08x105 4.11x105 Ni" (n.p) Co" Top-Middle 212.18 5.15x10' 5.90x10' Middle 212.18 4.99x10' 5.72x10' Bottom-Middle 212.18 5.11x10' 5.86x10' Average 5.08x107 5.75x10' 5.83x10' 6.59x10' NP88' (ri,f) Cs'8' Middle 211.41 4.02x107 4.41x 10' 4.02x10' 4.41 x107 U88*~ (n,f) Cs'8' Middle 211.41 5.73x10' 5.31x10e 5.73x10' 5.31x108 anos:t onses 6-21

The fast neutron (E > 1.0 MeV) flux and fluence levels derived for Capsula T are presented in Table 6-8. The thermal neutron flux obtained from the cobalt-alumi-num monitors is summarized in Table 6-9. Due to the relatively low thermal neu-tron flux at the capsule location, no bumup correction was made to any of the measured activities. The maximum error introduced by this assumption is esti-reaction and even less sig-mated to be less than 1 percent for the Ni"(n,p)Cow nificant for all of the other fast neutron reactions, n

An examination of Table 6-8 shows that the fast neutron flux (d > 1.0 MeV) derived from the five threshold reactions ranges from 8.15 x 105 to 8 90 x 105 n/cm8-sec, a total span of less than 10 percent. It may also be noted that the cal-culated flux value of 9.44 x 10" n/cm8-sec exceeds all of the measured values with calculation to experimental ratios ranging from 1.06 to 1.16. This behavior is

' consistent with prior benchmarking studies.

-Comparisons of measured and calculated current fast neutron exposures for Capsule T as well as for the inner radius of the pressure vessel are presented in

- Table 6-10. Mgasured values are given based on the Fe"(n,p) Mn5' reaction alone as well as for the average of all five threshold reactions. Based on the data given

-in Table 6-10, the best estimate exposure of Capsule T is ~

.b< T =. 2.74 x 10" n/cm8 (E > 1 MeV)

.Since the calculated fluence levels were based on conservative representations of

~

core power distributions derived for long term operation while the Capsule T data are representative only of cycle 1 operation, it is recomm' ended that projec-tions of vessel toughness into the future be based on design br. sis calculated flu-1 ence levels. Withdrawal of future surveillance capsules should further substan-

.tiate the adequacy of this approach.

l i

j 4

i 6-22 e730s:1071ssa

1 t

k_

5 4

Table 6-8 Results of Fast Neutron Dosimetry for Capsule T Adjusted Saturated Activity IdistsI

+ (E > 1.0 MeV) 4 (E > 1.0 MeV) g (n/cm*-sec)

(n/cm*)

Reaction Measured Calculated Measured Calculated Measured Calculated 9

Fe"(n.p)Mn" 3.91x106 4.53x106 8.15x10'*

9.44x10'*

2.65x10

3.07x 10

Ei Cu d(n,u)Co'*

3.88x10$

4.11x10$

8.90x10'*

9.44x10'*

2.89x10'8 3.07x 10'"

a Ni$"(n p)Co*

5.83x10' 6.59x10' 8.34x10;*

9.44x10'* '

2.71 x10'*

3.07x10'"

Np3(n.f)Cs'2' 4.02x10' 4.41 x10' 8.60x10

9.44x15

2.80x10'*

3.07x 10'*

Um(n,f)Cs

S.04x10*

5.31x10*

8.21 x 10'*

9.44x10'*

2.67x10'"

3.07x 10'"

U" adpa>Ged saturdied actively lean beest stusitsg>lsed by 0 88 to corsect for 350 ppm umomsnusty.

i e

e

~

Table 6-9 Results of Thermal Neutron Dosimetry for Capsule T fdis/sg Saturated Activity \\ g /

<DTh Axial Location Bare Cd-Covered (n/cm -sec) a Top 6.41 x 10' 2.56 x 10' 6.77 x 10'o Bottom 6.23 x 10' 2.40 x 10' 6.76 x 10'o Average 6.32 x 10' 2.48 x 10' 6.77 x 10'o Using calculated design basis maximum fluence levels of 1.68 and 3.20 x 10" nicm2 for the vessel 1/4 thickness and inner surface, respectively, at end of life and capsule lead factors presented in Table 6-3, the following revised capsule withdrawal schedule per ASTM E 185-79 is recommended.

Vessel Estimated Location Lead Removal Fluence Capsule (deg)

Factor TimeI*'

(n/cm2)

T 40 3.17 1.03 (removed) 2.74 x 10's U

140 3.17 3

9.0 x 10

X 220 3.17 6

1.79 x 10'**'

Y 320 3.17 11 3.28 x 10'**8 S

4 1.02 34 3.26 x 10

V 176 1.02 standby W

184 1.02 standby Z

356 1.02 standby a, 8ffect!ve full power years from plant startup ts. Approximcte fluence at 14 thickness vessel watt at end of life

c. Approximate fluence at vesselinner wall at end of life 6-24 enceit orissa

?

E Table 6-10 I

Summary of Fast Neutron Dosimetry Results for Capsule T Calculated Irradiation Vessel Vessel A3 Time 4 (E > 1.0 MeV) 4 (E > 1.0 MeV)

Lead Fluence Fluence Basis (EFPS)

(n/cm*-sec)

(n/cm2)

Factor (n/cm2)

(n/cm2)

Fe"(n.p)Mn" 3.25x10' 8.15x10'a 2.65x10

3.17 8.36x10

9.69x 10

'.17 8.64x10" 9.69x10" Dosimeter Avg 3.25x10' 8.44x10'a 2.74x10

3 4

1 4

I i

t

~

Section 7 References 1.

Yanichko, S. E., Lege, D. J., and Phillips, J. H. " Tennessee Valley Authority Sequoyah Unit No.1 Reactor Vessel Radiation Surveillance Program,"

WCAP-823'3, December,1973.

2.

ASTM Standard E185-73," Recommended Practice for Surveillance Tests for Nuclear Reactors" in ASTM Standards, Part 45 (1973), American Society for Testing and Materials, Philadelphia, Pa.1973.

3.

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

4.

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.

5.

SAILOR RSIC. Data Library Collection DLC-76, Coupled Self-Shielded,47 Neu-tron,20 Gamma. Ray, P, Cross Section Library For Light Water Reactors.

2 6.

Benchmark Testing of Westinghouse Neutron Transport Analysis Methodol-ogy (To be published).

7.

ASTM E N177, " Standard Practice for Measuring Neutron Flux, Fluence, and Spectra by Radioactivation Techniques." 1981 Annual Book of ASTM Stan-dards. Part 45 Nuclear Standards pp 915-926 Philadelphia: American Society for Testing and Materials,1981.

8.

ASTM E C22-77,

• Standard Test Method for Measuring Thermal Neutron Flux by Radioactivation Techniques." 1981 Annual Book of ASTM Standards. Part 45 Nuclear Standards. pp 927-935 Philadelphia: American Society for Testing and Materials,1981.

enos:omes 7-1

ASTM E 263-77, " Standard Test Method for Measuring Fast-Neutron Flux by Radioactivation of iron." 1981 Annual Book of ASTM Standards. Part 45 Nuclear Standards, pp 936-941 Philadelphia: American Society for Testing and Materials,1981.

-10. ASTM E 481-78," Standard Method for Measuring Neutron-Flux Density by

- Radioactivation of Cobalt and Silver," 1981 Annual Book of ASTM Standards.

Part 45 Nuclear Standards. pp 1063-1070 Philadelphia: American Society for Testing and Materials,1981 11.' ASTM E 264-77, " Standard Test Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel." 1981 Annual Book of ASTM Standards. Part 45 Nuclear Standards. pp 942-945 Philadelphia: America Society for Testing and Materials,1981.

.e e

7-2 staoe:S attoss

. - - =. -

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 matenal in the core region of the reactor vessel is determined by using the pre-service reactor vessel material properties and estimating the radiation-induced ARTNDT. RTNDT is designated as the higher of either the drop weight nil ductil-ity transition temperature (TNDT) or the temperature at which the material exhibits at least 50 ft Ib of impact energy and 35-mil lateiral expansion (normal to the major working direction) minus 60*F.

RT DT ncreases as the material is exposed to fast neutron radiation. Thus, to N

i find the most limiting RTNDT at any time period in the reactor 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 cer-tain chemical elements (such as copper and phosphorus) present in reactor ves-l sei steels. The Regulatory Guide 1.99 trend curves which show the effect of l

fluence and copper and phosphorus contents on ARTNDT for reactor vessel l

steels are shown in Figure A-1.

Given the copper and phosphorus contents of the most limiting material, the radiation-induced ARTNDT can be estimated from Figure A-1. Fast-neutron flu-l ence (E > MeV) at the 1/4T (wall thickness) and 3/4T (wall thickness) vessel loca-l tions is given as a function of full-power service life in Figure A-2. The data for all l

other ferritic materials in the reactor coolant pressure boundary are examined to ensure that no other component will be limiting with respect to RTNDT. The frac-ture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in a:cordance with the NRC Regulatory Standard Review Plan."8. The pos: irradiation fracture toughness properties of the reactor vessel beitline material were obtained directly from the Sequoyah Unit 1 Vessel Material Surveillance Program.

e730s:i o7issa A-1

~u.

0

$ 400 - A = 140 + 1000 (% Cu - 0.08) + 5000 (% p - 0.008)] If/1G 1 1/2 l8 U

FOR COPPER AND PHOSPHORUS CONTENTS OTHER opFEB tu 300 THAN THOSE PLOTTED, USE THE

/

EXPRESSION FOR A GIVEN ON w

THE FIGURE.

$ 200 u.

m b

i,%

o t

oV un

'ipo nu 2

P a

e 4

O a

4 50

]

o 0.35 0.30 0.25 0.20% Cu D.15% Cu 0.10% Cu

. LOWER LIMIT

% Cu = 0.08 i

N

%r - 0.012 9

% F = 0.008 1

l A WELD METAL

[

$ FORGING 01 I

I I

I I

I i

l l

I I

I I

i 4

0 19' 20 1

?X1017 4

6 8 1018 2

4 6

8 10 2

4 6

810 i

FLUENCE, n/cm2 (E > 1MeV) 4 l

Figure A-1. Predicted Adjustment of Reference Temperature, "A", as a f

Function of Fluence and Copper Content l

4 q

I Y

e

53S6A 25 1020

~

8 g

4 2

1/4T 1019 8 C g

g 6

4 5

4 w

3/4T

~

5 2

1018 8 2 6

4 2

1017 l

l l

l l

l 0

5 10 15 20 25 30 35 EFPY Figure A-2. Fast Neutron Fluence (E > 1 MeV) as a Function of Full Power Service Life A-3

i A-2. CRITERIA FOR ALLOWABLE PRESSURE TEMPERATURE RELATIONSHI*S 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 com-bined thermal and pressure stresses at any time during heatup and cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal t:mperature 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 KlR = 26.78 + 1.223 exp (0.0145 (T-RTNDT + 160))

(A-1) where KIR is the reference stress intensity factor as a function of the metal tem-perature T and the metal reference nil-ductility temperature RTNDT. Thus, the governing equation of the heatup-cooldown analysis is defined in Appendix G to the ASME Codei28 as follows:

C KIM + K!t G Kirl (A-2) where KIM s the stress intensity factor caused by membrane (pressure) stress i

Kit s the m'ress intensity factor caused by the thermal gradients i

KIR s a function of temperature to the RTNDT of the material i

C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K!R s determinea by the i

metal temperature at the tip of the postulated flaw, the appropriate value of RTNOT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, 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.

Icr the calculation of the allowable pressure-versus coolant temperature during 2

cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the control location of the flaw i3 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.

A-4 s7aos:iotissa

- ~ _ _ _ -

I i

\\

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 thta 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 temperatureithe.iT developed during coc.down results in a high r value of KIR at the 1/4T location fer finite cooldown rates than l

for steady state operation.

i Furthermore, if conditions exist such that the increase in KIR exceeds Kit, the cal-culated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on tempera-ture at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals alcng a cooldown ramp. The use of time composite curve eliminates this problem and ensures con-servative 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-tempera-ture 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 compressive 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 1/4T crack during heatup is lower than the KIR for the 1/4T crack during steady state conditions at the same coolant temperature. Dur-ing haatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower KIR's do not offset each othbr, and the pressure temperature curve based on steady state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for_ steady state and finito heatup rates is obtained.

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

assumed. Unlike the situation at ths vessel inside surface. the thermal gradients

. established at the outside surface during heatup produce stresses which are ten-site 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 cool-ant temperature) along the heatup ramp. Since the thermal stresses at the out-snoe:i o7tes:

A-5 i

~

.: side 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 as fol-Inws. A composite curve is constructed based on a point by-point comparison of the steadp-state and finite heatup rate data. At any given temperature, the allow ;

- able pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conser-

. v:tive 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 analy-sis of the most critical criterion. Then, composite curves for the heatup rate data End the cooldown rate data are adjusted for possible errors in the pressure and tImperature sensing instruments by the values indicated on the respective curves.

A-3J 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 previously. The deri-

. vition'of the limit curves is presented in the NRC Regulatory Standard Review Plan.m Transition temperature shifts occurring in the pressure vessel materials d'ue to.

rIdiation exposure have been obtained directly from the reactor pressure vesse!

surveillance program, f

. Charpy test specimens from Capsule T irradiated to 2.74 x-10" nicm2 indicate that the representative core region weld metal and the limiting core region forg-ing 04 exhibited maximum shifts in RTNDT of 140*F and 60*F, respectively, as shown by Figure A-1. The shifts are well within the appropriate design curve (Fig-l

' ure A-1) prediction Heatup and cooldown limit curves for normal operation of

. S:quoyah Unit 1 for up to 7 effective-full power years (EFPY) have been gener-ated and are shown in Figures A-3 and A-4.

l

' A!!owable combinat!6ns of temperature and pressure for specific temperature L

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-tem-perature combinations are to the right of the criticality limir line (Figure A-3). This l'

is in addition to other criteria which must be met before the reactor is made critical.

i The leak-test limit curve shown in Figure A-3 represents minimum temperature riouirements at the leak test pressure specified by applicable codes.'238 i

l A-6 usos:i ones:

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V J

$39 U.26 3000.0 I

I I

l llI Material Property Basis l

l l

", Controlling Material : Base Metal j

g j

-- Copper Content

0.13 w/o

. Phosphorus Content :0.015 w/o i__

g

_, RTNOT nitial

73* F

\\

l l

l l

I

_ RTNDT After 7 EFPY:

(

if

]

1/4T, 150* F If J

l l

3, 4T,110* F r

j j

g i5 I

f f

2000.0 Curve applicable for heatup rates up Y

{

to 100*F/hr for the service seriod up

]

to 7 EFPY and contains marg.n3 of J

J E

10*F and 60 PSIG for possible

/

/

ll

_ l gj instrument errors.

f f

l ll

--l l

)

l l

1 s

/

/

i fl

/

)

l 2 1000.0

/

/

I I

I

~

/

I

---. Criticality i

Heatup Rates

/

l l[

Limit l

l up to 100* F/hr --

f l ll ll l

ll j)(

lilli 11 f

Based on inservice hydro- _

static test temperature (296*F) for the service period up to 7 EFPY.

? f I

t 9

9 l

l ll ll ll llllllllll l

I 0.0 0.0 100.0 200.0 300.0 400.0 500.0 INDICATED TEMPERATURE (*F)

Figure A-3. Sequoyah Unit No.1 Reactor Coolant System Heatup Limita-tions Applicable for the First 7 EFPY A-7

_______________ __ _ _ _____ ____ _ _ A

5336A 27

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3000.0 l

Material Property Basis l'

I i

l I

lllllll1llll1ll!

{

l l

___ Controlling Material : Base Metal Copper Content

0.13 w/o

--- Phosphorus Content :0.015 w/o NDT nitial

73*F

--- RT I

f RTNDT After 7 EFPY:

1/4T.150' F j

3/4T.110' F

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l l

)

.l l

n I

2000.0 0;,

Single curve applicable for cooldown tu

_. rates up to 100* F/hr for the service

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period up to 7 EFPY and contains

].

margins of 10* F and 60 PSIG for

(/)

possible instrument errors.

7 in J

l E

/

I Ill I

a

)

l l

l' \\ \\

l' N

/

I l'

lill il

/

I 1 llll11 til 6 1000.0

/

l l

lll Cooldown Rates. *F/hr

[

l llIIl ll l ', l

.1111 A

l l

lit l I

I A7 I

l l

l

~*h U ff/

I' 1

1 I

i

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l l

I

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_ 504W ll ll

-1 ll ll l

ll li II

!!! I

!!!I II II 100 l

11 I

Ill l

ll I

I it i

il i

1, il i

liti lillill g,g 00 100.0 200.0 300.0 400.0 500.0 INDICATED TEMPERATURE ('F)

I Figure A-4. Sequoyah Unit No.1 Reactor Coolant System Cooldown Lim-itations Applicable for the First 7 EFPY A-8

.)

A-4. RWERENCES

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

2. ASiviE Boiler and Pressure Vessel Code, Section lil, Division 1 - Appendices,

" Rules for Construction of Nuclear Vessels", Appendix G, " Protection Against Nonductile Failure, pp. 461469,1980 Edition, American Society of Mechani-cal Engineers, New York,1980.

3. " Pressure-Temperature Limits", Chapter 5.3.2 in Standard Review P/an for the Review c.' Safety Analyses Reports for Nuclear Power Plants, LWR Edition, NUREG-0E00,1981.

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i enosa anse3 A-9

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