ML20137P807

Analysis of Capsule X from Portland General Electric Co Trojan Reactor Vessel Radiation Surveillance Program

ML20137P807
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	06/30/1985
From:	Shaun Anderson, Kaiser W, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20137P794	List: ML20137P791 ML20137P807
References
TAC-59991, TAC-60677, TAC-63527, WCAP-10861, NUDOCS 8602050349
Download: ML20137P807 (90)

Text

,

WCAP-1086.1

.=

ANALYSIS OF CAPSULE X FROM PORTLAND GENERAL ELECTRIC COMPANY TROJAN REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko S. L. Anderson W. T. Kaiser June 1985 APPROVED: I1 " ^L T. A. Meyer, Manager Structural Materials and Reliability Technology Work performed under Shop Order No. PMVJ 106 Prepared by Westinghouse Electric Corporation for the Portland General Electric Company

. Although the information contained in this report is nonproprietary, a no distnbution shall be made outside Westinghouse or its Licensees without the Customer's approval s WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 seR2188f4898syjg4 P

a

PREFACE This report has been technically reviewed and verified.

• Sections 1 through 5,7 and 8 9 33 R. S. Boggs

• Section 6 0. E h A. H. Foro

• Appendix A T [ '

F. J. Witt S

4 iii

s TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 11 2 INTRODUCTION 21 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE X 51 I 51. Overview 5-1 l- 5 2. Charpy V Notch Impact Test Results 53

, 5-3. Tension Test Results 5-4

. 5-4. Compact Tension Tests 5-4 1

6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1. Introduction 6-1 61 6 2. Discrete Ordinates Analysis 1 6-3. Neutron Dosimetry 6-3

64. Transport Analysis Results 67 6-5. Dosimetry Results 68 j 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8 REFERENCES 81 i Appendix A HEATUP AND COOLDOWN LIMIT CURVES FOR

, NORMAL OPERATION A1 A-1 Introduction A1

. A 2. Fracture Toughness Properties A1 j ! A 3. Criteria for Allowable Pressure.

Temperature Relationships A2 '

A 4. Heatup and Cooldown Limit Curves A5 v

LIST OF ILLUSTRATIONS .

i .*

Figure Title Page 4-1 Arrangement of Surveillance Capsules in Trojan '

Reactor Vessel (Updated Lead Factors for the Capsules Shown in Parentheses) 4-3 4-2 Capsule X Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters 4-6/4 7 5-1 Irradiated Charpy V-Notch Impact Properties for i Trojan Reactor Vessel Lower Shell Plate C5583-1, l Transverse Orientation 5 12 l 5-2 Irradiated Charpy V Notch Impact Properties for Trojan Reactor Vessel Lower Shell Plate C5583-1, Longitudinal Orientation 5 13

5-3 Irradiated Charpy V Notch Impact Properties for Trojan Reactor Pressure Vessel Weld Metal 5 14

, 5-4 Irradiated Charpy V Notch Impact Properties for 5

Trojan Reactor Pressure Vessel Weld Heat-Affected j Zone Metal 5 15 55 Charpy impact Specimen Fracture Surfaces for j-Trojan Pressure Vessel Lower Shell Plate C5583-1, Transverse Orientation 5 16 1 5-6 Charpy impact Specimen Fracture Surfaces for

. . Trojan Pressure Vessel Lower Shell Plate C55831, j Longitudinal Orientation 5-17

5-7 Charpy impact Specimen Fracture Surfaces for l Trojan Weld Metal 5 18 l 5-8 Charpy impact Specimen Fracture Surfaces for Trojan Weld Heat Affected Zone Metal 5 19 59 Comparison of Actual Versus Predicted 30 ft lb I Transition Temperature increases for the Trojan

Reactor Vessel Material Based on the Prediction l Methods of Regulatory Guide 1.99 Revision 1 5-20 5-10 Irradiated Tensile Properties for Trojan Reactor

Pressure Vessel Lower Shell Plate C55831, -

I Transverse Orientation 5 21 5-11 Irradiated Tensile Properties for Trojan Reactor ,

Pressure Vessel Lower Shell Plate C5583-1, l 1.ongitudinal Orientation 5 22 14 5 12 Irradiated Tensile Properties for Trojan Reactor Pressure Vessel Weld Metal 5 23

5 13 Fractured Tensile Specimens from Trojan Pressure

! Vessel Lower Shell Plate C55831, Transverse ,

i Orientation 5 24 l vii ,

i c

. _ - . . __ ._. _ . . . _ , _ _ . . , _ . ~ ~ . - - - - - - - _ , ~ . - , , - . , , , , _ _ _ . _ - . _ _ _ _ . - , - . . - . . _ - . . . _ , . _ _ - . ,

LIST OF ILLUSTRATIONS (cont)

I Figure 'litle Page 5-14 Fractured Tensile Specimens From Trojan Pressure Vessel Lower Shell Plate C5583-1, Longitudinal

!, Orientation 5-25 5-15 Fractured Tensile Specimens From Trojan Pressure Vessel Weld Metal 5-26 5

5-16 Typical Stess-Strain Curve for Tension Specimens 5-27 6-1 Trojan Reactor Geometry 6-9 62 Plan View of a Reactor Vessel Surveillance Capsule 6-10 6-3 Calculated Azimuthal Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel Surveillance Capsule Geometry 6-11 6-4 Calculated Radial Distribution of Maximum Fast .

, Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 6-12 6-5 Relative Axial Variation of Fast Neutron Flux (E > 1.0 MeV) Within the Pressure Vessel 6-13 6-6 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 MeV) Within The Surveillanco

, Capsules 6-14 A1 Predicted Adjustment of Reference Temperature, as a Function of Fluence, Copper and Phosphorus Contents A-8

! A2 Fast Neutron Fluence (E > 1.0 MeV) as a Function of Full Power Service Life (EFPY) A9 A-3 Trojan Reactor Coolant System Heatup Limitations Applicable for the First 10 EFPY A 10 A-4 Trojan Reactor Coolant System Cooldown Limitations

Applicable for the First 10 EFPY A 11

} .

e t

I 6

e viii i

I

- - -- _ - - . . - _ _ _ _ - - - ~ , . . _ _ = _ - - _ . _. ..

LIST OF TABLES Table Title Page 4-1 Chemistry and Heat Treatment of Material Representing

! the Core Region Lower Shell Plate and Weld Metal from Trojan Reactor Vessel 4-4 3

5-1 Charpy V-Notch Impact Data for Trojan Pressure Vessel Lower Shell Plate C55831 Irradiated at 550*F, Fluence 1.77 x 1019 n/cm2 (E > 1.0 MeV) 55 5-2 Charpy V Notch Impact Data for Trojan Pressure Vessel Weld Metal and HAZ Metal Irradiated at 550*F Fluence 1.77 x 1019 n/cm2 (E > 1.0 MeV) 5-6 5-3 Instrumented Charpy impact Test Results for Trojan Reactor Vessel Lower Shell Plate C55831 5-7

!~ 5-4 Instrumented Charpy impact Test Results for Trojan Reactor Vessel Weld Metal and HAZ Metal 5-8 2

5-5 Effect of 550'F Irradiation at 1.77 x 10'8 n/cm (E > 1.0 MeV) on Notch Toughness Properties of Trojan [

) 5-9 l Reactor Vessel Surveillance Material [

56 Summary of Trojan Reactor Vessel Surveillance Capsule i Charpy impact Test Results 5-10 5-7 Tensile Properties for Trojan Reactor Vessel Material f 2 Irradiated at 550*F to 1.77 x 10 19 n/cm (E > 1.0 MeV) 5 11 6-1 47 Group Energy Structure 6-15 6-2 Nuclear Constants for Neutron Flux Monitors Contained in the Trojan Surveillance Capsules 6-16 j

63 Calculated Fast Neutron Flux (E > 1.0 MeV) and Lead

~

Factors for Trojan Surveillance Capsules 6 17 i -

64 Calculated Neutron Energy Spectra at the Center of the Trojan Surveillance Capsule X 6 18 i*

l l

l j ix i

T t

LIST OF TABLES (cont)

J Table Title Page i 6-5 Spectrum Averaged Reaction Cross Sections at the Center of Trojan Surveillance Capsule X 6-19 6-6 Irradiation History of Surveillance Capsules Removed from the Trojan Reactor 6 20 6-7 Comparison of Measured and Calculated Fast Neutron Flux Monitor Saturated Activities for Capsule X 6-23 q

6-8 Results of Fast Neutron Dosimetry for Capsule X 6-24 6-9 Results of Thermal Neutron Dosimetry for Capsule X 6-25

6-10 Summary of Fast Neutron Dosimetry Results for

, Capsule X 6 26

~

{ 6-11 Calculated Current and EOL Vessel Exposure fo- Trojan 6-27 A-1 Trojan Reactor Vessel Toughness Table A6 I

i 0

) r X

t

... SECTION 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule X, the second capsule to be removed from the Trojan reactor pressure vessel, led to the follow-ing conclusions:

E The capsule received an average fast neutron fluence (E > 1.0 MeV) of 1.77 x 1019 n/cm2 while the maximum fluence at the vessel inner wall was 3.87 x 1018 n/c m2, E Irradiation of the reactor vessel lower shell plate C5583-1 to 1.77 x 10'9 n/cm2 resulted in 30 and 50 ft Ib transition temperature increases of 95'F and 120*F, respectively, for specimens oriented normal to the principal rolling direction of the plate and 30 and 50 ft Ib transition temperature increases of 90*F and 100*F, respectively for specimens oriented parallel to the plate principal rolling direction.

A Weld metal irradiated to 1.77 x 1019 n/cm2resulted in 30 and 50 ft Ib transition

~

temperature increases of 50*F and 55'F, respectively.

E Weld HAZ metal showed a 30*F and 50*F transition temperature increase of 60*F, after irradiation to 1.77 x 1019 n/cm2 E Plate C5583-1, weld metal, and HAZ metal all showed upper shelf energy levels well above 50 ft Ib after irradiation to 1.77 x 10 19n /cm2, E The 30 ft Ib transition temperature increases for the weld metal, weld HAZ and plate C5583-1 show that these materials are less sensitive to irradiation than predicted by Regulatory Guide 1.99 Revision 1.

5 New plant heatup and cooldown limit curves were developed for 10 EFPY based

<. on the capsule test results.

[ E The surveillance capsule removal schedule was revised as a result of the Capsulo X evaluation.

11

i i- SECTION 2

INTRODUCTION

This report presents the results of the examination of Capsule X, the second capsule to be removed from the reactor in the continuing surveillance program, which monitors the effects of neutron irradiation on the Trojan reactor pressure vessel materials under actual operating conditions.

i 1

The surveillance program for the Trojan reactor pressure vessel materials was designed

) and recommended by the Westinghouse Electric Corporation. A description of the t surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by Davidson and etal.I'l The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel, and was bas-

- ed on ASTM E-185 73, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels."l21 Westinghouse Nuclear Energy Systems personnel were con-l- tracted 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 l

surveillance specimens were performed.

This report summarizes the testing of and the postirradiation data obtained from surveillance Capsule X removed from the Trojan reactor vessel and discusses the analysis of these data. The data are also compared to results of the previously remov-ed Capsule U.133 1

4 4,

i l

21 i

I i

! 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 irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 533 Grade B Class 1 (base material of the Trojan 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 Ncnductile Failure," Appendix G, to Section Ill of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts is based on the reference nil-ductility temperature, RTNor-RT Nor is defined as the greater of the drop weight nil-ductility transition temperature (NDTT per ASTM E 208) or the temperature of 60*F less than the 50 ft-lb (and 35 mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. The RTno7 of a given material is used to index that material to a reference stress intensity factor curve (Kn curve) which appears in Appendix G of the ASME Code. The Kin curve is a lower bound of dynamie, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Kin curve, allowable stress inten-sity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

RT go7 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 radia-tion embrittlement of changes in mechanical properties of a given reactor pressure vessel 3-1

steel can be monitored by a reactor surveillance program such as the Trojan Reactor -

Vessel Rediation Surveillance Program,lilin which a surveillance capsule is periodical-ly removed from the operating nuclear reactor and the encapsulated specimens are tested. -

The increase in the average Charpy V-Notch 30 ft Ib temperature (ARTgg7) due to ir-radiation is added to the original RTgo7 to adjust the RTug7 for radiation embrittlement.

This adjusted RTNor(RTug7 initial + ARTug7) is used to index the material to the K in 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.

l l

\

i l

O w

32 l

r w-- -

-w. -v--ww- ,.

.m-y, ,, , , " -F "

Y

' - ~ -

F--'r-u-W-'*'- -m- w-w --- -

+r ee +----s-w---

SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Trojan reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron shielding pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

Capsule X was removed from the reactor after 4.28 Effective Full Power Years (EFPY) of plant operation. This capsule contained Charpy V-Notch, tensile, and Compact Ten-sion (CT) specimens from submerged arc weld metal representative of the reactor vessel core region weld metal, and Charpy V-Notch, tensile, CT, and bend bar specimens from the lower shell plate C5583-1. The capsule also contained Charpy V-Notch specimens from weld Heat Affected Zone (HAZ) metal. All heat affected zone specimens were ob-tained from the weld HAZ of plate C5583-1. The chemistry and heat treatment of the program surveillance materials is presented in Table 41.

All test specimens were mat. led from the %-thickness location of the plates. Test specimens represent material taken at least one-plate thickness from the quenched end of the plate. Some base metal Charpy V-Notch and tensile specimens were oriented with the longitudinal axis of the specimens normal to (transverse orientation) and some parallel to (longitudinal orientation) the major working direction of the plate. The CT test specimens were machined so that the crack of the specimen would propagate nor-mal to (longitudinal specimens) and parallel to (transverse specimens) the major work-ing direction of the plate. All specimens were fatigue precracked per ASTM E399 72.

The precracked bend bar was machined in the transverse orientation. Charpy V-Notch specimens from the weld metal were oriented with the longitudinal axis of the specimens

~

normal to (transverse orientation) the weld direction. Tensile specimens were oriented with the longitudinal axis of the specimen normal to (transverse orientation) the weld direction.

41

Capsule X contained dosimeter wires of pure copper, iron, nickel, and aluminum-15% -

cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Np237 and U23s were contained in the capsule. -

Thermal monitors made from two low-melting-point eutectic alloys and sealed in Pyrex tubes were included in the capsult.. The composition of the two allo;/s and their melting points are as follows:

2.5 percent Ag,97.5 percent Pb Melting point: 579'F (304*C) 1.75 percent Ag,0.75 percent Sn,97.5 percent Pb Melting point: 590*F (312'C)

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

G 4

m D

9

~

4-2

g,

- REACTOR VESSEL CORE BARREL NEUTRON PAD z (4.76) i CAPSULE I l (TYP) V (4.06) 56* I 56' J To. <

270* -- -

90*

.# N Y (4.06) l x (4.76) '

w (4.76) 180*

FIGURE 4-1. ARRANGEMENT OF SURVEILLANCE CAPSULES IN THE TROJAN REACTOR VESSEL (UPDATED LEAD FACTORS FOR THE CAPSULES SHOWN INN PARENTHESES) 4-3

TABLE 4-1 ..

CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE CORE REGION LOWER SHELL PLATE AND .,

WELD METAL FROM THE TROJAN REACTOR VESSEL CHEMICAL ANALYSIS (WEIGHT PERCENT)

Lower Shell Plate Weld Metal Element C5583-1 (a) (b)

C 0.21 0.18 0.19 Mn 1.27 1.26 1.52 P 0.011 0.028 0.010 S 0.016 0.015 0.012 Si 0.20 0.54 0.35 Ni 0.60 0.93 0.97 Cr 0.048 0.058 0.09 V 0.002 0.002 < 0.010 Mo 0.53 0.51 0.50 Co 0.019 0.029 0.020 Cu 0.15 0.051 0.06 -

Sn 0.007 0.005 Al 0.021 0.009 0.020 .

N2 0.007 0.010 HEAT TREATMENT Lower shell 1650*/1750*F, for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, water quenched; plate C5583-1 1550*/1650*F, for 4% hours, water qiienched; 1200'/1300*F, for 4% hours, air-ccoled 1100*/1175*F, for 36% hours, furnace-cooled Weldment[c] 1100*/1175'F, for 9% hours, furnaced-cooled a) Onginal anaYs b) Analysis on irradiated Charpy specimen PW50 c) Fabncated with Adcom Wire Heat No. S3986 and Linde 124 Flux LM No 934, which was used to fabncate the vessel beltline region intermediate to lower shell girth weld seams and the .

associated longitudinal weld seams. ,

e 4-4

4 e

t 4

1 MMO CORAPACT CoePACT CouPACT C4er&CT Bam Tma nues0N TEmescal O*Aary CNAAPY Ca8AAPT TWIIM001 TIGEIBON 4 P412 Pwe0 PMoD PW57 PHS7 PW54 PH54 PWM RT4 rwii me %s m,. Pw,3 ms.

P ee Pws. G Pu

% a i. a,s a,. a,3 m

Ce  ;-' 8 l j' l l l - As .15% Co Fe

!  !] l LILJ LJ r.r,

~II Aa .15% CosCd) sn F l 7:

am '

s*

ll lllil 1-l ,l m' l II i

+

e TO TOP OF VESSEL 4

I I \

I e

VESSEL WALL CORE SPECIMEN NUMBERING CODE PH - HEAT AFFECTED ZONE PW - WELD METAL non RT - PLATE CS$83-1 (TRANSVERSE)

RL - PLATE C5583-1 (LONGITUDINAL) as ad CA N ULEE enP' vm D00ABETER C000 PACT C0esPACT lt4 APT Cm SLOCR TEssedLE CseARPT r[ CstanPT CHAAPT CMAAPT CatanPT T13:00086 Tlas000ss Ttassas RAD R'57 % 57 RT54 %S4 RT51 8LS1 RTee get RT12 PH51 PW48 PM48 RLt2 RT90 f s p esso PW47 Ps47 tes Rtii RT5e m.se RTw RLu RT53 Rt5s RT5o R5o RT47 a47 RTi6 RTi5 Rii. RTis RT,, y F

1 B PH40 PW46 PH46 fL10 RT50 RL50 RTM RL55 RT52 RL52 RT4  % 49 RT44 R46 PT10 I

J l f

c.

%;j i ~ '- c-l "ii ; + : . . . , m c.

xlii>>lll7 b

ll l F. "is i dLJ LJ '[LJLI t

= .is ca ece l

l1' ,

590*F M I

' llir Qj n s I i g

! 1t If f l 9 CENVER REGION OF YESSEL TO BOTTOed OF VESSEL FIGURE 4-2. CAPSULE X DIAGRAM SHOWING LOCATION OF SPECIMENS, THERMAL MONITORS, AND DOSIMETERS N

l APER M l CARD

^2g^;"""'d'*f g6,B205D3 @ 01

SECTION 5

~

TESTING OF SPECIMENS FROM CAPSULE X 5-1. OVERVIEW The postirradiation mechanical testing of the Charpy V-Notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with con-sultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H, ASTM Specification E185-82, and Westinghouse Procedure RMF-8402, Revision 0.

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-8426I'l. 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 Pro-cedure 8103 on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Char-py machine is instrumented with an Effects Technology Model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addi-tion to the standard measurement of Charpy energy (Eg). From the load-time curve, the load of general yielding (Pay), the time to general yielding (toy), the maximum load (Py), and the time to maximum load (tM) can be determined. Under some test condi-tions, a sharp dcop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pp ), and the load at s which fast fracture terminated is identified as the arrest load (PA)*

, [ 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 specimen.

5-1

Therefore, the propagation energy for the crack (Ep )is the difference between the total -

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

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

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.

Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-81 and E21-79, and RMF Procedure 8102.

All pull rods, grips, and pins were made of inconel 718 hardened to Rc 45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

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

and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67.

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

Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature. Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and

controller temperatures was developed over the range room temperature to 550*F (288'C). The upper grip was used to control the furnace temperature. During the actual -

testing the grip temperatures were used to obtain desired specimen temperatures. Ex-periments indicated that this method is accurate to i2*F. .

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, 5-2

and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The

.- fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

5-2. CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-Notch impact tests performed on the various materials con-tained in Capsule X irradiated at 1.77 x 1019n/cm are2 presented in Tables 5-1 through 5-5 and Figures 5-1 through 5-4. The fractured surfaces of the impact specimens are shown in Figures 5-5 through 5-8.

Irradiation of Charpy V-Notch impact specimens from the reactor vessel lower shell plate, C5583-1, to 1.77 x 1019 n/cm2as shown in Figure 5-1 resulted in 30 and 50 ft Ib tran-sition temperature increases of 95'F and 120*F, respectively, for specimens oriented normal to the principal rolling direction (transverse orientation) of the plate. Specimens

, oriented parallel to the principal rolling direction (longitudinal orientation) of the plate as shown in Figure 5-2 exhibited 30 and 50 ft Ib transition temperature increases of

. 90*F and 100'F respectively. The upper shelf energy of the shell plate showed a 8 ft Ib decrease in the transverse direction and a 14 ft Ib decrease in the longitudinal direction.

Weld metal specimens irradiated to 1.77 x 10 19 n/cm2 resulted in a 30 ft Ib and 50 ft Ib transition temperature increases of 50*F and 55'F, respectively, as shown in Figure 5-3. Irradiation did not cause any decrease in upper shelf energy.

Weld HAZ specimens irradiated to 1.77 x 1019 n/cm2resulted in a 30 ft Ib and 50 ft

[ Ib transition temperature increase of 60*F as shown in Figure 5-4. The upper shelf energy of the HAZ metal decreased by 14 ft Ib due to the irradiation.

l Table 5-6 shows a summary of the Charpy test results for the two Capsules X and U j tested to date. These results show that the higher fluence 1.77 x 1019 n/cm2received

by Capsule X resulted in some additional increase in transition temperature and an ad-

- ditional decrease in shelf energy when compared with the Capsule U results.

. Figure 5-9 shows a comparison of the actual increase in the 30 ft Ib transition temperature l

versus the predicted increase based on Regulatory Guide 1.99 Revision 11dl prediction methods for the Trojan vessel surveillance materials. These results show that the metals i are less sensitive to radiation than predicted by the Guide.

5-3

(

The upper shelf energy values for the lower shell plate C5583-1 and the weld metal at -

1.84 x 1019 n/cm2 are greater than 50 ft Ib as required by 10CFR50 Appendix G.

Paragraph IV.A.1 and are predicted to exceed 50 ft ib throughout the design life of the .

vessel.

5-3. TENSION TEST RESULTS The results of tension tests performed on material from the reactor vessel lower shell plate C5583-1 and weld metalirradiated to 1.77 x 10 9 n/cm2are shown in Table 5-7 and Figures 5-10 through 5-12. Plate C5583-1 test results are shown in Figures 5-10 and 5-11 and indicate that irradiation to 1.77 x 10"3 n/cm2caused a 15 ksi maximum increase in 0.2 percent yield strength and ultimate tenst;e strength. Weld metal tension test results presented in Figure 5-12 show that the 0.2 percent yield strength and ultimate tensile strength increased ~ 7 ksi with irradition. The fractured tension specimens for the plate material are shown in Figures 5,13 and 5-14, while the fractured tension specimens for the weld metal are shown in Figure 5-15. A typical stress-strain curve for the tension tests is shown in Figure 5-16.

5-4. COM' PACT TENSION TESTS .

The %T compact tension fracture mechanics specimens that were contained in Cap-sute X have not been tested and are stored at the Westinghouse Research Develop-ment Laboratory.

S e

5-4

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR TROJAN PRESSURE VESSEL LOWER SHELL PLATE C5583-1 IRRADIATED AT 550'F, FLUENCE 1.77x 1019 n/cm2 (E > 1 MeV)

Temperature Impact Energy Lateral Expansion Shear Sample No. (* F) (*C) (ft Ib) (J) (mils) (mm) (%)

RL50 0 - 18 10.0 13.5 9.5 0.24 1 RL58 50 10 22.0 30.0 17.0 0.43 15 RL53 60 16 25.0 34.0 19.5 0.50 18 g RL54 75 24 24.0 32.5 18.5 0.47 22 j RL56 75 24 46.0 62.5 29.5 0.75 29 j RL60 76 24 45.0 61.0 29.5 0.75 29

-g RL46 100 38 43.0 58.5 29.5 0.75 24 g RL49 125 52 37.0 50.0 27.5 0.70 43

.5 RL51 125 52 50.0 68.0 33.5 0.85 47 3 RL59 150 66 69.0 93.5 47.5 1.21 58 k RL55 175 79 61.0 82.5 47.0 1.19 65 f RL57 RL52 200 225 93 107 80.0 101.0 108.5 137.0 55.5 67.5 1.41 1.71 87 100 RL47 300 149 102.0 138.5 67.5 1.71 100 RL48 350 177 103.0 139.5 72.0 1.83 100 RT53 0 - 18 8.0 11.0 10.0 0.25 0 RT49 75 24 22.0 30.0 16.0 0.41 16 RT55 76 24 30.0 40.5 22.0 0.56 22 c RT54 100 38 31.0 42.0 20.0 0.51 24

{e RT50 RT56 100 125 38 52 30.0 30.0 40.5 40.5 27.5 27.5 0.70 0.70 24 36 E RT58 125 52 44.0 59.5 37.5 0.95 43 RT46 150 66 44.0 59.5 38.0 0.97 50

- 2 RT51 175 79 38.0 51.5 36.0 0.91 57 lj RT52 RT59 200 225 93 107 55.0 74.0 74.5 100.5 48.5 61.5 1.23 1.56 76 97

. $ RT47 250 121 85.0 115.0 57.5 1.46 98 RT60 300 149 81.0 110.0 61.0 1.55 100 RT48 350 177 65.0 88.0 53.0 1.35 100 RT57 400 204 73.0 99.0 60.0 1.52 100 5-5

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR TROJAN PRESSURE VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550*F, FLUENCE 1.77 x 1018n/cm2 (E > 1 MeV)

Temperature impact Energy Lateral Expansion Shear Sample No. (* F) (*C) (ft Ib) (J) (mils) (mm) (%)

PW59 - 100 - 73 4.0 5.5 6.5 0.17 0 PW52 - 50 - 46 18.0 24.5 12.5 0.32 6 PW53 0 - 18 16.0 21.5 14.0 0.36 16 PW55 25 -

4 27.0 36.5 23.5 0.60 32 PW49 50 10 28.0 38.0 25.0 0.64 36 PW57 50 10 31.0 42.0 26.5 0.67 47

$ PW47 75 24 51.0 69.0 41.0 1.04 70 3 PW50 77 25 38.0 51.5 32.0 0.81 62 3 PW58 100 38 51.0 69.0 38.5 0.98 79 3 PW48 125 52 65.0 88.0 53.0 1.35 87 PW56 150 66 74.0 100.5 68.5 1.74 89 PW54 200 93 58.0 78.5 56.0 1.42 94 -

PW46 250 121 82.0 111.0 71.5 1.82 98 PW51 300 149 84.0 114.0 70.0 1.78 100 -

PW60 350 177 91.0 123.5 69.0 1.75 100 PH60 - 100 - 73 12.0 16.5 11.5 0.29 8 PH52 - 50 - 46 15.0 20.5 18.5 0.47 18 PH55 - 50 - 46 41.0 55.5 25.0 0.64 18 PH54 - 25 - 32 51.0 69.0 36.5 0.93 25 PH59 0 - 18 29.0 39.5 18.0 0.46 30

3 PH56 25 - 4 85.0 115.0 53.5 1.36 74 l j PH58 25 - 4 16.0 21.5 19.5 0.50 24 N PH57 25 -

4 50.0 68.0 36.5 0.93 38

$ PH47 50 10 50.0 68.0 34.5 0.88 54 l PH46 50 10 71.0 96.5 43.5 1.10 77 PH50 75 24 109.0 148.0 69.0 1.75 95 .

PH53 77 25 126.0 171.0 68.5 1.74 100 -

PH49 150 66 97.0 131.5 64.5 1.64 100 ,

PH48 200 93 103.0 139.5 63.0 1.60 100  :

PH51 300 149 120.0 162.5 76.0 1.93 100 5-6

TABLE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR TROJAN REACTOR VESSEL LOWER SHELL PLATE C5583-1 Normalized Energies Test Charpy (ft Ibs/in2) Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Charpy Maximum Prop Load to Yield Load Maximum Load Load Stress Stress Number (* F) (ft Ib) Ed/A Em/A Ep/A (kips) (pSec) (kips) (pSec) (kips) (kips) (ksi) (ksi)

RL50 0 10.0 81 62 18 3 65 95 3.95 170 3.95 120 125 RL58 50 22.0 177 137 40 3.50 95 4.20 320 4.15 116 128 RL53 60 25.0 201 163 38 3 55 100 4.40 370 4.35 117 131

.E RL54 75 24 0 193 115 78 3.50 90 4.20 275 4.20 0.55 115 127 3 RL56 75 46 0 370 285 85 3.45 100 4.45 610 4.30 0.20 113 131 E RL60 76 45.0 362 282 81 3.45 100 4.50 600 4.20 0.40 115 132 8 RL46 100 43 0 346 261 86 3.55 120 4.40 585 4.40 0.55 117 131

% RL49 125 37.0 298 197 101 3 35 100 4.40 445 4.20 1.00 III 128

$ RL51 125 50 0 403 275 128 3.35 110 4.45 605 4.30 1.30 111 129 5 RL59 150 69 0 556 262 294 3 40 120 4.30 605 3 85 2.05 112 127 E RL55 175 61.0 491 225 267 3 25 110 4 25 520 4 05 2.00 107 124 3 RL57 200 80 0 644 264 381 3 05 105 4.20 610 3.75 2.50 101 120 RL52 225 101.0 813 260 553 3.15 90 4.10 560 104 120

? RL47 300 102.0 821 259 562 2.90 95 4.20 595 96 118 N RL48 350 103 0 829 262 567 2.75 95 3 90 630 91 111 RT53 0 80 64 42 22 3.75 100 3.90 130 3 85 124 127 RT49 75 22.0 177 104 73 3.35 100 3 90 270 3.90 0.75 110 120 RT55 76 30 0 242 178 63 3.40 100 4 30 410 4.30 112 127 8 RT54 100 31.0 250 154 96 3.50 115 4 30 370 4.15 0 80 115 129 i RT50 100 30 0 242 137 105 3 40 105 4.10 335 3 90 0 80 113 124 g RT56 125 30 0 242 144 97 3.40 105 4 05 350 3 85 1.25 112 123 y RT58 125 44 0 354 220 134 3 20 105 4 20 510 4.15 1.45 106 122 a, RT46 150 44.0 354 177 177 3 25 115 4 05 435 3 95 1.85 107 120

$ RT51 175 38 0 306 140 166 3 25 95 3 95 345 3 90 2.00 107 119

$ RT52 200 55 0 443 217 226 3 05 95 4.15 505 3 95 2.45 101 119

$ RT59 225 74.0 596 187 408 3.15 130 3 95 485 105 117

• RT47 250 85 0 684 211 473 2.60 85 4 00 520 85 109 RT60 300 81.0 652 225 427 2.80 85 3 95 535 93 112 RT48 350 65.0 523 161 363 2.70 90 3 60 425 89 104 RT57 400 73 0 588 166 422 2.70 120 3 65 455 89 105 l

TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR TROJAN REACTOR VESSEL WELD METAL AND HAZ METAL Normalized Energies Test Charpy (ft lbs/in2) Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Charpy Maximum Prop Load to Yield Load Maximum Load Load Stress Stress Number (*F) (ft Ib) Ed/A Em/A Ep/A (kips) (p Sec) (kips) (p Sec) (kips) (kips) (ksi) (ksi)

PW59 - 100 4.0 32 20 12 2.95 75 3.20 85 3.20 98 102 PW52 - 50 18.0 145 110 35 3.80 105 4.40 260 4.40 126 136 PW53 0 16.0 129 68 61 3.55 95 3.95 185 3.95 0.30 117 124 PW55 25 27.0 217 151 66 3.55 110 4.25 360 4.25 1.00 117 129 PW49 50 28.0 225 150 76 3.30 90 4.20 350 4.20 1.30 110 124 3 PW57 50 31.0 250 110 139 3.45' 95 4.05 275 3.85 1.65 113 123

$ PW47 75 51.0 411 200 211 3.35 95 4.15 460 3.90 2.30 110 123 p PW50 77 33.0 306 157 149 3.25 85 4.00 370 3.90 2.15 107 120 y PW58 100 51.0 411 215 196 3.40 100 4.20 485 4.20 2.40 112 126 PW48 125 65.0 523 195 329 2.90 105 4 05 475 95 114 PW56 150 74 0 596 207 389 3.30 120 4 05 495 110 122 on PW54 200 58 0 467 180 287 3 00 90 3.95 435 100 115 O PW46 250 82.0 660 221 439 2.95 90 4.05 520 97 116 PW51 300 64.0 676 207 470 2.85 90 3 85 510 95 III PW60 350 91.0 733 235 498 2.85 85 3.90 560 95 112 PH60 - 100 12.0 97 79 18 4.30 130 4.55 210 4.55 142 147 PH52 - 50 15.0 121 83 38 3 90 95 4.35 200 4.20 0 20 130 137 PH55 - 50 41.0 330 256 74 3 85 95 4 85 505 4.65 128 144 PH54 - 25 51.0 411 261 150 3 90 110 4.85 525 4.70 0.35 130 145 PH59 0 29 0 234 83 151 3.80 100 4.05 210 4.00 1.25 126 130 3 PH58 25 16 0 129 40 89 3 60 110 3 85 135 3 65 0 90 119 123 j PH57 25 60 0 403 237 165 3.70 100 4 55 505 4.40 1.90 122 137 N PH56 25 85.0 684 270 415 3 85 115 4 60 560 3.75 2.20 127 139 y PH47 50 50.0 403 240 163 3 65 100 4 65 505 4.40 1.85 121 137 PH46 50 71 0 572 246 326 3.60 90 4.55 510 119 135 PH50 75 109 0 878 287 591 3.45 95 4.55 600 115 133 PH53 77 126 0 1015 316 699 3 45 90 4 60 655 115 133 PH49 150 97.0 781 276 505 3.25 95 4.30 610 107 125 PH48 200 103 0 829 264 565 3 30 95 4 30 585 109 125 PH51 300 120 0 966 306 660 2.95 100 4 20 695 98 119

. . , , , l

.., 4 ,

TABLE 5-5 THE EFFECT OF 550*F IRRADIATION AT 1.77x 1018 n/cm2 (E > 1 MeV)

ON NOTCH TOUGHNESS PROPERTIES OF TROJAN REACTOR VESSEL SURVEILLANCE MATERIALS Average Average 35 mil Average Average Energy Absorbtion 50 ft Ib Temp ('F) Lateral Expansior. Temp (*F) 30 ft Ib Temp (*F) at Full Shear (ft Ib)

Material Unirradiated Irradiated AT Unirradiated Irradiated AT Unitradiated Irradiated AT Unirradiated Irradiated (ft$b)

Plate C5583-1 50 170 120 32 142 110 8 103 95 84 76 - 8 Transverse m

Plate C5583-1 20 120 100 15 120 105 - 15 75 90 116 102 - 14 Longitudinal 31 86 55 16 76 60 2 52 50 83 86 +3 e1 "j, - 37 23 80 - 40 25 e5 - 57 3 80 i25 iii -i4

TABLE 5-6

SUMMARY

OF TROJAN REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS 30 ft Ib 50 ft Ib Upper Shelf Trans. Temp. Trans. Temp. Energy Capsule Fluence increase increase Decrease ci Material Ident. n/cm2 (.F) (*C) (*F) (*C) (ft Ib) (J) 5 Plate C5583-1 U 3.88 x 10'8 53 29 55 31 6 8 (Longitudinal) X 1.77 x 10'S 90 50 100 56 14 19 Plate C5583-1 U 3.88 x 10is 44 24 65 36 0 0 (Transverse) X 1.77 x 1018 95 53 120 67 8 11 U 3.88 x 10'8 22 12 32 18 2 3 WeM Meml X 1.77 x 1018 50 28 55 31 3l'1 4l*l HAZ Metal X 1.77 x 10'S 60 33 60 33 14 19 a Upper Shelf Energy increase

s ,

TABLE 5-7 TENSILE PROPERTIES FOR TROJAN REACTOR VESSEL MATERIAL 1RRADIATED AT 550*F TO 1.77 x 10 18n /cm (E > 1.0 MeV)

Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Specimen Temperature Strength Strength Load Stress Strength Elongation Elongation inArea Number Material ('F) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) (%)

RL12 ' ' ' ' ^

Plate C5583-1 3.34 170.4 68.0 10.5 20.4 60 RL11 225 75.4 95.7 RL10 550 73.3 96.8 3.75 171.2 76.4 9.4 18.7 55 m ,

a a

RT12 ' ' '

Plate C5583-1 147.2 71.3 9.6 17.4 52 RT10 225 77.4 95.7 3.50 ansese 71.3 93.7 3.90 160.3 79.5 9.3 16.0 50 RT11 550 PW11 Weld Metal 75 84.0 99.0 3.20 248.7 65.2 9.9 21.9 74 PW12 175 78.9 95.7 3.20 171.8 65.2 9.5 20.4 62 PW10 550 75.9 95./ 3.60 179 0 73.3 9.0 17.9 59

100 -

g3-g g-g

^

s 80

~

j I

T 60 -

O i O

m g 40 2

2 O

I.,e '

20 -

gb2 0 0 =

100 n

$ 80 -

2 l.0

,s p.1. g m -

p 2 -

110*F g,g

< 20 -

/ 'O 0 W 9e/

100 90 2 O ,

80 8-.. h Q

70 UNIRRADIATED O p -

e E 60 -

> 9 e 0 50 - 120*F _ IRRADIATED (5500*F)

C 7 1.77 x 10 'n/cm W

z 40 -

ee w O #

30 _ O g95'F l O /

20 - e 10 O

_ O O

o* I I I I I

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

FIGURE 5-1. IRRADIATED CHARPY V-NOTCH IMPACT PROPERTIES FOR TROJAN REACTOR VESSEL LOWER SHELL PLATE C5s83-1, TRANSVERSE ORIENTATION 5 12 1

100 -

D 9-9

$8o

... g /

$ 60 -

2 e' i"

- 2 S' 20 -

3 2

%s) 34

/

6 0

100 f8 60 - h,e O y .._ e n

m 40 -

/

Oh105'F /

g

<i: 20 -

2m/2 Un gM a*'t a

0

^ ,q' e -

140 120 -

UNIRRADIATED h d100 -

O e 9-8

~

80 -

O'2 e O G IRRADIATED (5500*F) g 60 -

o co*F ,/ O ' 7' * ' U ' '" *

b 40 / 0

/ 90*F j 0'O 20 - O SN

. O O I I I I

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

FIGURE 5-2. IRRADIATED CHARPY V-NOTCH IMPACT PROPERTIES FOR TROJAN REACTOR VESSEL LOWER SHELL PLATE C5583-1, LONGITUDINAL ORIENTATION 5-13

l 100 -

h g-9-9 Q?

OO y 80 -

/

E O g #e/ -

60 -

< /

$ 40 - T " -

20 -

2 /

• d /

0 l 100 80 - 2 2 s ._.

g 60 -

O oly#B_.

Oe l 40 -

O g

2yf' ..S 3 20 -

gg O

100 90 -

0 -

9- e 80 -

UNIRRADIATED Q'g p/O '

E

- 70 8

O/

E O e

> 60 -

o ,

55'F 50 -

Z IRRADIATED (550*F) uJ 50*F 1.77 x 10n/cm2 40 j*

O 30 -

20 -

j , .

• ,2J'I , , , ,

-200 -100 0 100 200 300 400 500 j TEMPERATURE (*F)

FIGURE 5-3. IRRADIATED CHARPY V-NOTCH IMPACT PROPERTIES FOR TROJAN REACTOR

PRESSURE VESSEL WELD METAL 5-14

2 100 - 2 ,33,_,f 3 ,

80 -

OO*l. l g 60 -

< 3 ,/ ,

y 40 -

w 20 -

2g..

S 0

100

= 80 E

60 y0"h # e Q

• k OOo*

O' W 40 65'F g

$20 WS ys

%e--g 0

160 UNIRRADIATED 2 g" - g

_ 120 1.

e

= 100 -

/ , 9 OO

$ 80 - OG IRRADIATED (550*F) z 60 -

60*F 1.77 x 10"n/cm2

. W e.- e e

. 40 -

O O/e 20 60*

@e * ,F/

• , I I I I o ,

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

FIGURE 5-4. CHARPY V-NOTCH IMPACT PROPERTIES FOR TROJAN WELD HEAT-AFFECTED ZONE METAL 5-15

.[

h~

y?i" p N

,p fT4 43 s, ,; r

.: , , -n

$ t

,4*

w - n.

p . J,w. ~

,'3 g,nvj' 7 y <nj' ~ .(]; . . ',

f

[.'Q * ;

< .. gM. , j @A, .!

37,g . r 2 m. -.,- , 2 7,

's

, .  :. ..s.- 1 w._ 2. . . '- d 31 f.. r. .t . ; ,, w < +g e, c

t, ., ,. . ,

, / , s ,y .: ,

[.-.,2

. v . . , ' . * -] . r k. * * = ,/ . p , . ,p mW , .

l,% . l. e ' & .i, .  :

xq.%.1 g ,  ;

, i,r 2 m.. --

s. bd4 :A% t',g:3.a O

y ty

y C%

g,I~ n.

m  !

s .

A4 w/ n. . . y k.,,c[~

m .."a4.? 4 '.e & , * -f , , + ' *i RT53 RT49 RT55 RT54 RT50 j

V # w.

i . a .e

(,,g ,h

A

~ _ . - . , .

k* ',' **i .

- hy4". e'*

t Li i

ia

,- ,. fM; ,, t 1,,

4 F ;. ,

, m., . . I

1. .c ,4

%.~C.t jo p@;;,h . .' IS $

. - po .

L ,L .

,. p.

tS4  :[A$

I b., '

~

['

.l.?{.s

• Q."~.,,;'.

p *. (~l .

+

jf*}~*\

e s-

'~'. -

3

.m .

[Wr"~ ~ ,

~1 n:C

. .f .

El F7

+.

's ,~$q[ s.

i.y l

. s

..f ,

fi Q . _~

-a{e'.y4 (D

f,piQ.i ) , .

MO _ bd, hjkw '

5f RT56 RT58 RT46 RT51 RT52

, yn*-

._ ,. . y %..y } .

a g: n, ,

3. ' ' ;l 'x! m: 2 o .,

N: N M . W' k.f?.&> ,lo, ' .

W f' ..

.jN,'

• hs

. . , . ~.

. s. y-4r.-

. +. ~s . .. .

j p i, 2-q;f ;3z 3 .1 A, l .c,. . , .  %.4, . .. W, .

w << c. ; 3 te ~, .

g<a - :W: '

j ,.,' 5 . .,. m #

' , . ?!

~:.+ . M>. gi.-c Cs" ,2

, a l

nl. n;; w 9

g ' r'". , ~-'lL:. b:).~1 WAf W .. .:

.. ~:

c-r,;,. -

yj

,Q,0 gR=-

St. -

..&-t*. i g(.J i

1-

... in f w '.

RT59 RT47 RT60 RT48 RT57 FIGURE 5 5. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR TROJAN PRESSURE VESSEL SHELL PLATE C5583-1, TRANSVERSE ORIENTATION 5-16

';%{p

,e ,

.~j m "- 1 " ,, ~

'A,

f-

.s s,, f

.c.

j. . .a. ,.! ,'o cas"'! "e :

e

. c. .,+ '

.3.

+ %';$

g.. ..t?

s. g

\ '* 'l-

$ L'$' 'l

[ ' '; . mJ;[n 5" ,.' [ l

.t .i

. . ,i.

,/ \g

',N.t s

~ ~ .

8 .

==

.- 's/j[.., 'd)N[ } (_[c,k '

'h..' . .

h@A.%s.

g 3 s.: . . .

j@ . ,4@.l 3

UQ 8

. g%y.y-

'lp@l fjl u.

RL50 RLS8 RL53 RL54 RL56 r~-

d

,- ~ \

n a.rm .- y bz, ,w . m --

m jr l s a $, , ,

di 2.

\%. r4j' M =;.

g  ;- /

.4, lj i:*

i[e"

!k~,

s _ f..Q .  %.g y fl* Qt

'5 '*

' rk r.Ay( .)'

v.#}'*g - i p*.

,. f. " -4 ,

t .

2

?

m) l N,- - .- :l44, 7 .

'  :\g.

.s P h..y g 4 y : 1- .' c;< U., .-

7. ~. . ?.y -<; y ; -;m_ ; ;

9 s,- -

cy - ,

, y,,,9, ,,'.

t f

?i '

V! * ' ' ,

.g . ,

@s .g. re 4

ai 4 @.o.;h7 w. .-

, @'Am ti\ -.

%&_N 95t_: '  %-%. .

-: N, ;

I RL60 RL46 RL49 RL51 RL59 i

n; . w. am m,

w. , 1 sp % .f"; .tu.)c ,)a <*(,

,i

>,,,- w? e h.1.-l

> ~s ,~W..

'4fa'$'.s. f. ,

'u,..m a

. W, n n.

1 y%

r ~x y*f e-

~ y ',' .1 y Q, .;; .y , y. b y, .  ; Q:*. 37; *i: e j,b .,;

w ._. ~g. _ -

t-  % 4

.7

" c

"?

c.

J t I 4 .

n a

. .r

( 2 4% e% si;; ~

\,4,'y* ,, , +4

.SQ

.yM.>. ' 9.7'>

1 - .,s. , . !

[./,1 -

pa

< T.c

. g,~n

~:

h c. f.4,,7 m<-

.y , / ip; - }

u,f. +,. - .fj ; *O i cf . . .v. - .n.

~ m.7 f.;

3j RL55 RL57 RL52 RL47 RL48 FIGURE 5 6. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR TROJAN PRESSURE VESSEL SHELL PLATE C55831, LONGITUDINAL ORIENTATION 5-17 l __ _. . _ - - _ _ - _ - - -

r .

7tsgr- -

. t " wF W: . ; . .  %' N:

- -~' Rd;], '. i1I'y[n, ) K

.$]??::

Bi;;q

..s W @, h Rfb hh}Y? }h' ll' A * & :

a. -l Q? ':.'

Wt g i Mi,'i :LL , , , . . .. .

+ .

p;?%.

% M;ct: 4

• 6, i A lp %,; - - :1 < j W 0;;> , ; ? ;; g f. jpM ,yj J!-

k_. hQid'-Mid)A $!d0i c pkj Y o

,Y li

A EUYb(

'II!Ubi F h' M&f$7 ~ * ' chYif '

sl.,

i l l

PW69 PW52 PW53 PW55 PW49  !

l l

l

_ f?Nf,

~

_l '; [. .; -- 4 56 '

{'"W'."'N.

rC i 'n,

., >;R

,h

v,* " i v (

.: ..Q s.

'a';yN }t ,

,? p In-

\ .y'y qfy ,

9::::o  ;:nn -

a ,: > v. y <( o, t

. .p - ,

- aa

.t r h k. N '

af,f- * 'g T.

's

~

Y^

(: 4

+

A *"

, , e /.. -

1

' gt 44I

,#

1 [ # .jL e4 5* A h;S;.

wy

@?f L.j' dhT; 6^ ji v .L

's~

v n;9, !y 3 = h -

.._... m m . _ ... u - J PW57 PW47 PW50 PW58 PW48 p 5 J]. , "i

~ , ,,,. 37  %

2r '

. i .;,

3 ,4,

.t b  ::a r g, D .a

~

c:f ; .l

...t-c

.n", ,.t s q,. . j .

'- . [-

+

' pi 4 g 4y

^

_;,}- Y*

/ ' *h Ijl.' , j; j 136 1

?

t *(g,

~

l ..

{

4

[ A m,

W: .

.]'

c, t

.J . .-

PW56 PWS4 PW46 PW51 PW60 FIGURE 5-7. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR TROJAN WELD METAL 5 18

m W v.,j -

9 . . .-'d # J'

~

As;; s .' _.;:,) j 4 .f . -r .. N

t. - ** . y ; ,:e t  % X ;o &

y *l

*. * *> y _

< .% J;;;f s - .

(,s.q

. .k,;;

< *i ' % ,e

.1 14 -4 4 ,1- ,1 ' .'s .

n '> *

~

• hk .~ ..

,'fi -f "' f (d ' '*

• j ;

• , 3a t u

. h. 9

~

i Y,e,.'td -y j;. %.q. '

3 s. .- -

. ~.

'd((1

.- ;QU 1-

):,h y lf, , ,\ v;(

o _3-

,A.s; %p ,-q

. N: eu-4 cc-w-r .. ign s f-t: , (ps  ?' A'

~ ' D j

i x

'y A.Y i , ,;

q' a) .. .Y,  %,'%).y } g?.f g.hq--

dix)1

f. \j t . 2 PH60 PH52 PH55 PH54 PH59 prkNr, .

. w h?' mg'a' 3 , . , , rg < l

( ,V 4 94.s y m' ,) , V

' q, [,C I [. Y 0hf g>kf '.

.. . , -i f on .

y

.' . v-c 'ry 3<,, 3 i' _e .

n=/. , y;'.3  ; w .., en'j ;. r; > , . ,),. ,

. :.- T .

,- - ,. 'Y

~

"g. ._

n. . ... .

i ~

, '!k

_~3 l; ) e- p

\

jY ' f 'u lr ; ' . ?, ' , _ **j)0I

w?. c. .s. '

<.'; n p. 3ll ,l,. .c y!

"v'y

" %y Q g y3i is?.3 '

~

,. ~;g  :

ay

.  %,, '~

PH56 PH58 PH57 PH47 PH46 i

l L"

\ m e y

< 2;L l

%, '  % e $w_.%fg." l -

V4 '- ji, .

l hf h, N~A h., -

l i

- *k

[!l "h, *.

-}, sp, Q .,

f a) he.\

.}.

t Q v ,.

. j :g y lll t

' .;f h)

- 3, w/ y 6 *  !

i:y PH50 PH53 PH49 PH48 PH51 FIGURE 5 8. CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR TROJAN WELD HEAT-AFFECTED ZONE METAL 5-19

500 400 -

[ 300 u) d 200 -

5 z s PLATE C5583-1 us G s@

E e~

100 -

's ei s F*

$ 90 -

@ '* g,o 8 O*'

y 80 -

o eT g cfo@ WELD METAL

$ 50 -

6 e ; e g 2 73 40 -

k E

F 30 -

LEGEND S O PLATE C5583-1 (LONGITUDINAL)

O PLATE C5583-1 (TRANSVERSE) 20 -

o V 6 WELD METAL V HAZ METAL 10 I I I I I I I I i 1018 2 4 6 8 1019 2 4 6 8 1020 FLUENCE (n/cm2)

FIGURE 5-9. COMPARISON OF ACTUAL VERSUS PREDICTED 30 ft Ib TRANSITION TEMPERATURE

INCREASES FOR THE TROJAN REACTOR VESSEL MATERIAL BASED ON THE PREDICTION METHODS OF REGULATORY GUIDE 1.99, REVISION 1

. . .' /

TEMPERATURE (*C) 0 50 100 150 200 250 300 l l l l- I I l -

800 110 -

ULTIMATE TENSILE STRENGTH -

700 e A- A k

b -

600

$ 80 -

N 8 O c.

2 W

E g . -

500 70 -

$ N 400 0 200 WELD STRENGTH 50 -

- 300 40 CODE:

Open Points - Unitradiated Closed Points - Irradiated at 1.77x 1019n/cm2 80 70 -

g a o

\

y o -g C 50

~

>- REDUCTION IN AREA D 40 -

TOTAL ELONGATION 20 -

A% g_ -b b

,o -

O' . - . Q b.

UNIFORM ELONGATION I I ' ' ' I 0

o 100 200 300 400 500 600 TEMPERATURE (*F)

FIGURE 5-10. IRRADIATED TENSILE PROPERTIES FOR TROJAN REACTOR PRESSURE VESSEL LOWER SHELL PLATE C5583-1, TRANSVERSE ORIENTATION 5-21 1

TEMPERATURE (OC) 0 50 100 150 200 250 300 120 l l l l l l l -

800 110 - -

A ULE.' ATE TENSILE STRENGTH 700 00 _

A A ,

$ 90 $

/g e00

,N._ 0 08 E h 8 -

500

' G K\ O 60 -

400 0.2% YlELD STRENGTH 50 -

40 CODE:

Open Points - Unirradiated Closed Points - Irradiated at 1.77x 10 9n/cm2 80 70 -

g Q --

REDUCTION IN AREA g -

60 -

G S _g 50 -

40 -

P 30 -

TOTAL ELONGATION Q

O b- A 20 -

A A 10 -

O S UNIFORM ELONGATION I I ' ' ' I O

o 100 200 300 400 500 600 TEMPERATURE (*F) ,

FIGURE 5-11. IRRADIATED TENSILE PROPERTIES FOR TROJAN REACTOR PRESSURE VESSEL LOWER SHELL PLATE C5583-1, LONGITUDINAL ORIENTAION 5-22

TEMPERATURE (*C) 0 50 100 150 200 250 300

... 12 I I I I I I I -

800 110 -

ULTIMATE TENSILE STRENGTH -

700

_m g NA- # -A 0 -

g -b -

600 ,

$ 80 -

N g $

Q k -

500 F 70 -

k

@- Q m

0.2% YlELD STRENGTH 50 -

- 300 40 CODE:

Open Points - Unirradiated

. Closed Points - Irradiated at 1.77x 1019n/cm2 80 70 -

O - g ---

60 -

REDUCTION IN AREA

~0

> 50 -

b y 40 -

o 3 30 -

TOTAL ELONGATION O b cA1 6 20 -

10 -

@ , O UNIFORM ELONGATION g

i I I I I I

, o o 100 200 300 400 500 600

," TEMPERATURE (*F)

FIGURE 5-12. IRRADIATED TENSILE PROPERTIES FOR TROJAN REACTOR PRESSURE VESSEL WELD METAL

> 5-23

Specimen RT12 125'F 7ww a wiryy; y;7;7~7;y; .

.mc -

I i I I I I I I 1 2 3 4 5 6 7 8 9 0 10 10THS Specimen RT10 225'F m 1 m, mmgw ~

n ,e - :

I I I I I I I I .

i 1 2 3 4 5 6 7 8 9 0 10 10THS Specimen RT11 550*F 4 -,

,, - 9 . ,, . _ , :0,-- 2 e ---

I l

l I I I I I I I .

1 2 3 4 5 6 7 8 9 '

0 10

• i I

10THS

'. i l

FIGURE 5-13. FRACTURED TENSILE SPECIMENS FROM TROJAN PRESSURE VESSEL LOWER SHELL PLATE C5583-1, TRANSVERSE ORIENTATION l l

5-24 )

l i

l

Specimen RL12 125'F M-..me.mw.,

I I I I I I I I 1 2 3 4 5 6 7 8 9 0 10

, 10THS Specimen RL11 225*F e; . . w- .; . .

-; y. g 1

1 I I I I I I I I 1 2 3 4 5 6 7 8 9 0 10 10THS Specimen RL10 550 F a, 7 s , r. .

,m.=. ,,c +;-, - t z m ?_-j ;

• % v

. . , , . .n4 7

vg i< e_, y ,u I I I I I I I I l

1 2 3 4 5 6 7 8 9 0 10 10THS FIGURE 5-14. FRACTURED TENSILE SPECIMENS FROM TROJAN PRESSURE VESSEL LOWER SHELL PLATE C5583-1, LONGITUDINAL ORIENTATION 5-25

Specimen PW11 75*F

. - - _ _ .. ., 7 te e? :y cy i:

.:.e -e +- pcsn7% -

4 r

rjg

,,,,. .k .

I I I I I I I I I 1 2 3 4 5 6 7 8 9 0 10 10THS Specimen PW12 175*F N ,

, 'h

• < ,. e w-1 I I I I I I I -

1 2 3 4 5 6 7 8 9 0 10

• 10THS l

Specimen PW10 550*F

.j; e e,- ->

I I I I I I I I 1 2 3 4 5 6 7 8 9 ~

0 10 ,

10THS FIGURE 5-15. FRACTURED TENSILE SPECIMENS FROM TROJAN l

PRESSURE VESSEL WELD METAL 5-26

e 120 108 -

96 -

84 -

, 72

$ 60 -

E Ga -

. 36 -

Specimen No. RL11 TEST TEMPERATURE 225'F 12 -

0  ! I I  !  !  !  !  !  !

0 0.03 0.06 0.09 0.12 0.15 0.18 0.21 0.24 0.27 0.30 STRAIN (indin.)

e e

FIGURE 5-16. TYPICAL STRESS-STRAIN CURVE FOR TENSION SPECIMENS 5-27

SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1. INTRODUCTION Knowledge of the neutron environment within the pressure vessel / surveillance capsule geometry is required as an integral part of LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced property, changes observ-ed in materials test specimens and the neutron environnient (fluence, flux) to which the test specimens were exposed must be known. Second, in relating the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship must be established between the environment at various posi-tions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techni-ques and measurements obtained with passive neutron flux monitors contained in each

- of the surveillance capsules. The latter information is derived solely from analysis.

This section describes a discrete ordinates Sn transport analysis performed for the Trojcn reactor to determine the fast neutron (E > 1.0 MeV) flux and fluence as well as the neutron energy spectra within the reactor vessel surveillance capsules. The analytical data were then used to develop lead factors for use in relating neutron exposure of the pressure vessel to that of the surveillance capsules. Based on spectrum-averaged reaction cross sections derived from this calculation, the analysis of the neutron dosimetry contained in Capsule X is presented.

6-2. DISCRETE ORDINATES ANALYSIS A plan view of the Trojan reactor geometry at the core midplane is shown in Figure

[ 6-1. Since the reactor exhibits % core symmetry, only a zero- to 45-degree sector is depicted. Six irradiation capsules attached to the neutron pad are included in the design

,' to constitute the reactor vessel surveillance program. Four capsules (U,W,X,Z) are located at 34 degrees and two (V,Y) at 31.5 degrees from the cardinal axes shown in Figure 6-1.

6-1

J A plan view of a double surveillance capsule attached to the neutron pad is shown in -

Figure 6-2. The stainless steel specimen container is 1.182 inches by 1 inch and ap-proximately 56 inches in height. The containers are positioned axially such that the ,

specimens are centered on the core midplane, thus spanning the central 5 feet of the 12 foot high reactor core.

From a neutronic standpoint, the surveillance capsule structures are significant. In fact, they have a marked effect on the dist:ibutions of neutron flux and energy spectra in the water annulus between the neutron pad and the reactor vessel. Thus, in order to proper-ly ascertain the neutron environment at the test specimen locations, the capsules themselves must be included in the analytical model. Use of at least a two-dimensional computation is therefore mandatory.

In the analysis of the neutron environment within the Trojan reactor geometry, predic-tions of neutron flux magnitude and energy spectra were made with the DOT!51 two-dimensional discrete ordinates code. The radial and azimuthal distributions were obtained from and R,0 computation wherein the geometry shown in Figures 61 and 6-2 was describ-ed in the analytical model. In addition to the R,0 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 in-cluded in the model.

Both the R,0 and the R,Z analyses employed 47 neutron energy groups and a P3 expan-i sion of the scattering cross sections. The cross sections used in the analyses were ob-tained from the SAILOR cross section libraryMI which was developed specifically for light water reactor applications. The neutron energy group structure used in the analysis is l listed in Table 6-1.

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

It should be noted that this generic design basis power distribution is intended to pro- -

vide a vehicle for long-term (end-of-life) projection of vessel exposure. Since plant-specific i

l l 6-2 1

{

.- power distributions reflect only past operation, their use for projection into the future many not be justified; the use of generic data which reflects long-tem operation of similar reac-

,- tor 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 Westinghouse plants indicate that this analytical approach yields conservative results, with calculations ex-ceeding measurements from 10 to 25 percent.171 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 commitment to low-leakage loading patterns could significantly reduce the calculated neutron flux levels presented in Section 6-4. In addition, capsule lead factors could be changed, thereby influencing the withdrawal schedule of the remaining surveillance 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.

f(R,Z,0,E g ) = f(R,0,E g ) x F(Z,Eg )

where:

f(R,Z,0,Eg) = neutron flux at point R,Z,0 within energy group g 4 (R,0,Eg ) = neutron flux at point R,6 within energy group g obatined from the R,0 calculation F (Z,Eg ) = relative axial distribution of neutron flux within energy group g obtained from the R,Z calculation 6-3. NEUTRON DOSIMETRY The passive neutron flux monitors included in the Trojan surveillance program are listed in Table 6-2. The first five reactions in Table 6-2 are used as fast neutron monitors to relate neutrca fluence (E > 1.0 MeV) to measured material property changes. To pro-perly account for burnout of the product isotope generated by fast neutron reactions, 6-3 4

it is necessary to also determine the magnitude of the thermal neutron flux at the monitor -

locition. Therefore, bare and cadmium-covered cobalt-aluminum monitors were also

!ncluded. .

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

Th3 use of passive monitors such as those listed in Table 6-2 does not yield a direct miasure of the energy-dependent flux level at the point of interest. Rather, the activa-tion or tission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period.

An accurate assessment of the average neutron flux level incident on the various monitors m y be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

E The operating history of the reactor 5 The energy response of the monitor 5 The neutron energy spectrum at the monitor location 5 The physical characteristics of the monitor Th3 analysis of the passive monitors and subsequent derivation of the average neutron flux requires completion of two operations. 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.

Tha specific activity of each of the monitors is determined using established ASTM procedures.[8.9.10,11,12] Following sample preparation, the activity of each monitor is .

d;termined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The ov;rall standard deviation of the measured data is a function of the pr,ecision of sample w;ighing, the uncertainty in counting, and the acceptable error in detector calibration.

6-4

.- For the samples removed from Trojan, the overall 2o deviation in tha measured data is determined to be plus or minus 10 percent. The neutron energy spectra are determin-

. ed analytically using the method described in paragraph 6-1.

Having the measured activity of the monitors and the neutron energy spectra at the loca-tions of interest, the calculation of the neutron flux proceeds as follows. The reaction product activity in the monitor is expressed as:

N r N p R= I d P (1 - e-At) e-At

__0 f , Y j (6-2)

A seo(E) f(E)dE {max where:

R = induced product activity Ng= Avagadro'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 f(E) = energy dependent neutron flux at the monitor location with the reactor at full power P, = average core power level during irradiation period j P,,,= maximum or reference core power level

. A = decay constant of the product isotope t3 = length of irradiation period j to = decay time following irradiation period j d

6-5

Because neutron flux distributions are calculated using multigroup transport methods .

and, further, because the prime interest is in the fast neutron fiux above 1.0 MeV, spectrum-averaged reaction cross sections are defined such that the integral term in .,

equation (6-2) is replaced by the 'ollowing relation.

e o(E) f(E)dE = H 4(E > 1.0 MeV) sE where:

co N s0 o(E) 4 (E)dE [ogfg g_ _

G=1 r e. N f(E)dE s 1.0 MeV

[ 4g G = G1 .0 MeV Thus, equation (6-2) is rewritten I

R= A,Yf f (E > 1.0 MeV) i .

1 P,,, (1 - e-At ) e-Ato or solving for the neutron flux, 4 (E > 1.0 MeV) = N (64)

I bA f, Y 5 [ P mo, (1 - e-At)j e-Ato The total fluence above 1.0 MeV is then given by .

P N p 5 (E > 1.0 MeV) = $ (E > 1.0 MeV) [ p i

t) (6-4) '.

3.i max 6-6

, ... where:

N Pi total effective full power seconds of reactor operation up to p, i

• the time of capsule removal An assessment of the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered CoS9 (n,y) Co60 data by means of cadmium ratios and the use of a 24-barn temperature corrected 2200 m/sec cross section. Thus,

'O-1' Rb8 D e3= (6-5) b f, Y 5 i A P ,, (1 - e-At ) e-Ato where:

D is defined as Rb *

R cd covered 6-4 TRANSPORT ANALYSIS RESULTS Results of the Sn transport calculations for Trojan reactor are summarized in Figures 6-3 through 6-6 and in Tables 6-3 through 6-5. In Figure 6-3, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius,

% thickness location, and % thickness location 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 vesselis shown. The relative l axial variation of neutron flux within the vesselis given in Figure 6-5. Absolute axial varia-tions of fast neutron flux may be obtained by multiplying the levels given in Figure 6-3

," or 6-4 by the appropriate values from Figure 6-5.

In Figure 6-6, the radial variations of fast neutron flux within the surveillance capsules

~

are presented. These data, in conjunction with the maximum vessel flux, are used to develop lead factors for each of the capsules. Here the lead factor is defined as the ratio of the fast neutron flux (E > 1.0 MeV) at the dosimeter block location (capsule center) 6-7

to the maximum fast neutron flux at the pressure vessel inner radius. The updated lead factors for the Trojan surveillance capsules are listed in Table 6-3. The neutron flux ,

monitors contained within the surveillace capsule are all located at the same radial loca-tion, the capsule center. Had they been located at different radial locations within the ,

capsules it would have been necessary to adjust the disintegration rates for the gradients th:t exist within the capsules. In the present analysis, the point of comparison for all r: action rates is, of course, the capsule center.

In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The neutron energy spectrum calculated to exist at the center of the Trojan surveillance Capsule X is listed in Table 6-4. The associated spectrum-averaged cross sections for each of the fast neutron reactions are given in Table 6-5.

6-5. DOSIMETRY RESULTS Tho irradiation history of the Trojan reactor up to the time of removal of Capsule X is listed in Table 6-6. Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsule X based on the irradiation history shown in Table 6-6 cre listed in Table 6-7. The data are presented as measured at the capsule center.

The fast neutron (E > 1.0 MeV) flux and fluence levels derived for Capsule X are presented in Table 6-8. The thermal neutron flux obtained from the cobalt-aluminum monitors is .

summarized in Table 6-9. Due to the relatively low thermal neutron flux at the capsule locction, no burnup correction was made to any of the measured activities. The max-imum error introduced by this assumption is estimated to be less than 1 percent for the Nisa (n,p)Cosa reaction and even less significant for all of the other fast neutron reactions.

An examination of Table 6-8 shows that the fast neutron flux (E > 1.0 MeV) derived from tha five threshold reactions ranges from 1.14 x 10" to 1.61 x 10" n/cm2sec, a total span of about 41 percent. It may also be noted that the calculated flux value of 1.36 x 10"n/cm 2-sec compares well with the average measured value of 1.31 x 10" n/cm2-sec with individual calculation to experimental ratios ranging from 0.84 to 1.19.

Comparisons of measured and calculated fast neutron exposures for Capsule X and the inn:r radius of the pressure vessel are presented in Table 6-10. Calculated current and .

EOL vessel exposures are presented in Table 6-11, for vessel inner radius, % thickness -

and % thickness. Agreement of 4% between the average measured value and the calcula- ,

tion supports the use of the calculated exposure for vessel embrittlement projections. -

Based on the data given in Table 610, the best estimate exposure of Capsule X is 1.77 x 10" n/cm 2(E > 1.0 MeV).

6-8

l l

9=0*

CORE BARREL 1

7 -

9:25' 9:31.5*

/ 6:34*

NEUTRCN PAD 9:45'

//

\ //

i'/

I /

/

- / 7

/

i '.

- i //

//

s

//

t

'/

' /

//

/>,

l '/

//

I Y

i A X

FIGURE 61. TROJAN REACTOR GEOMETRY 69

4 CHARPY SPECIMEN

- 1, s/1 ////s.<

fff f f f ff .

i 'NNNNNNNNNNNNNNNNN\NN

\ NEUTRON PAD \ ,

ANNN\\\\\\\\\\\\\\\\

[

t FIGURE 6-2. PLAN VIEW OF A REACTOR VESSEL SURVEILLANCE CAPSULE l

6-10

100.0 70.0 -

50.0 -

2 1

S 20.0 -

x SURVEILLANCE CAPSULE R = 207.31 cm 70 -

! 5.0 -

e PRESSURE VESSEL IR 2.0 A ~

_ 1/4t LOCAT W

E 0.7 -

30.5 3/4t LOCATION i 0.2 - -

0.1 l l l I I I I  !

45 50 55 60 65 70 75 80 85 00 AZIMUTHAL ANGLE (DEGREE)

' FIGURE 6 3. CALCULATED AZlMUTHAL DISTRIBUTION OF MAXIMUM FAST NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL - SURVEILLANCE CAPSULE GEOMETRY 3

6 11

M 100.0 _

70.0 -

50.0 30.0 l'

225.19

% 20.0 -

2 10.0 ce =

E 7.0 -

i 5.0 -

l

~

h 3.0 -

^ 2.0 - l 241 62 l

E 3 1.0 _ l

• 0.7 = _

0.5 -

03 -

l l 0.2 - INNER RADIUS OUTER R ADIUS (IR) 1/4T 3/4T (OR) 0.1 I I  ! '

215 220 225 230 235 240 245 R ADIUS (cm)

FIGURE 6-4. CALCULATED RADIAL DISTRIBUTION OF MAXIMUM FAST NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL 6 12

1.000 0.700 -

0.500 -

0.200 -

0.100 x -

3 0.070 -

m -

$ 0.050 -

5 -

a W

s G 0.020 -

3 a

0.010 0.007 -

0.005 -

- - CORE MIDPLANE 0.002 -

r TO VESSEL CLOSURE HEAD 0.001  !  !  ! I L 300 200 100 0 100 200 300 OlSTANCE FROM CORE MIDPLANE (cm)

FIGURE 6-5. RELATIVE AXIAL VARIATION OF FAST NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL 6 13

a 100 70 -

50 -

CAPSULES U. W, X. Z k -

205.73 f 207.31 h 20 208.90

_~ 10 - CAPSULES V. Y a -

2 -

- 7 A -

E 5 -

5a -

w -

2 ~

IR 9. OR i l l l I  !

200.0 202.5' 205.0 207.5 210.0 212.5 215.0 R AOlUS (cm)

FIGURE 6-6. CALCULATED RADIAL DISTRIBUTION OF MAXIMUM FAST NEUTRON FLUX (E > 1.0 MeV) WITHIN THE SURVEILLANCE CAPSULES 6-14

i TABLE 6-1 47 GROUP ENERGY STRUCTURE Lower Energy Lower Energy Group (MeV) Group (MeV) 1 14.19*l 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 t

10 2.73 34 7.10 x 10-3 11 2.47 35 3.36 x 10-3 12 2.37 36 1.59 x 10-3 13 2.35 37 4.54 x 10 -4 14 2.23 38 2.14 x 10 - 4 15 1.92 39 1.01 x 10-4 16 1.65 40 3.73 x 10-5 17 1.35 41 1.07 x 10-5 18 1.00 42 5.04 x 10-6 19 0.821 43 1.86 x 10-6 20 0.743 44 8.76 x 10- 7 21 0.608 45 4.14 x 10- 7 22 0.498 46 1.00 x 10- 7

. 23 0.369 47 0.00

. 24 0.298

[a) Note: The upper energy of Group 1 is 17.33 MeV.

J 6-15

TABLE 6-2 NUCLEAR CONSTANTS FOR NEUTRON FLUX MONITORS CONTAINED IN THE TROJAN SURVEILLANCE CAPSULES Target Fission Monitor Reaction Weight Product Yield Material of Interest Fraction Half Life (%) -

Copper 83 60 0.6917 5.27 years Cu (n.a)Co Iron Fe5d(n,p)Mn54 0.0585 312 days -

Nickel Nisa(n.p)Cosa 0.6777 71.4 days Uranium-238(al U2aa(n,f)Cs137 1.0 30.2 years 6.0 Neptunium 237tal Np237(n,f)Cs137 1.0 30.2 years 6.5 Ccbalt Aluminum [a] C059(n,7)Co60 0.0015 5.27 years Cobalt-Aluminum CoS9(n,y)Co6o 0.0015 5.27 years l1l Denotes that monitor IS Cadmium.sheeM a

e O

6 r >

(

6-16

1

• TABLE 6-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AND LEAD FACTOR FOR TROJAN SURVEILLANCE CAPSULES Azimuthal f(E > 1.0 MeV) Lead Capsule Location (deg) 2 (n/cm -sec) Factor U 56' W 124' 1.36 x 10" 4.76 X 236' Z 304*

~

V 58.5*

1.16 x 10" 4.06

. Y 238.5*

t O

6 17

l TABLE 6-4 j CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF THE TROJAN SURVEILLANCE CAPSULE X Group No. 4(n/cm2.sec) Group No. 4(n/cm2.sec) 1 2.23 x 107 25 7.89 x 1010 2 8.19 x 107 26 8.22 x 10'0 3 2.85 x 108 27 6.61 x 10'0 4 5.22 x 108 28 4.63 x 10'0 5 , 8.72 x 10 8 29 1.39 x 10'0 6 1.93 x 109 30 7.47 x 109 7 2.69 x 109 31 2.05 x 10 0 8 5.56 x 109 32 1.32 x 10 0 .

9 5.18 x 109 33 2.33 x 10'0 10 4.34 x 109 34 3.41 x 10'O .

11 5.23 x 109 35 5.79 x 10'0 12 2.62 x 109 36 5.21 x 10'0 13 8.06 x 10s 37 7.29 x 10'0 14 4.06 x 109 38 3.96 x 10'0 15 1.09 x 10'0 39 4.40 x 10'O 16 1.49 x 10 0 40 5.95 x 10'O 17 2.33 x 10'O 41 7.03 x 10'0 18 5.28 x 10'O 42 3,89 x 10'0 19 4.07 x 10'0 43 4.44 x 1010 20 1.92 x 10'0 44 2.75 x 10'0 21 7.09 x 10 0 45 2.12 x 10'0 22 5.56 x 1010 46 2.99 x 10'0 23 6.95 x 10'0 47 3.81 x 1010 -

~

24 6.78 x 10'0 6-18 i

TABLE 6-5 SPECTRUM AVERAGED REACTION CROSS-SECTIONS AT THE CENTER OF TROJAN SURVEILLANCE CAPSULE X Reaction 6 (barns)

Fe5d(n p)Mn 54 0.0559

CuS3(n,a)Coso 0.000479 Nise(n.p)Cose 0.0779 Np237(n,f)Cs137 3.338 U23a(n,f)Cs 37 0.313 CoS9(n,y) Co60 24 e=

c(E) $ (E)dE J0 0 =

r*

f(E)dE s 1.0 MeV e

e e

I 1

6-19

t l

l TABLE 6-6 IRRADIATION HISTORY OF SURVEILLANCE CAPSULES REMOVED FROM THE TROJAN REACTOR l

l l P aya P ,g, Irradiation Decaytal Month Year (MW) (MW) Paya/ PREF Time (Days) Time (Days) 12 1975 34 3411 0 010 17 3303 1 1976 205 3411 0 060 31 3272 2 1976 495 3411 0 145 29 3243

! 3 1976 92 3411 0 027 31 3212 l 4 1976 819 3411 0 240 30 3182 5 1976 1283 3411 0 376 31 3151

\ 6 1976 THROUGH 8 1976 0 3411 0 000 92 3059 9 1976 1504 3411 0 441 30 3029 10 1976 965 3411 0 283 31 2998 l 11 1976 1627 3411 0 477 30 2968 12 1976 liG*, 3411 0 969 31 2937

• l 1 1977 3049 3411 0 894 31 2906 2 1977 3152 3411 0 924 28 2878 3 1977 THROUGH 6 1977 1422 3411 0 417 122 2756 7 1977 2855 3411 0 837 31 2725 8 1977 3087 3411 0 905 31 2694 9 1977 2333 3411 0 684 30 2664 10 1977 2957 3411 0 867 31 2633 11 1977 3281 3411 0 962 30 2603 12 1977 2875 3411 0 843 31 2572 1 1978 3319 3411 0 973 31 2541 2 1978 2869 3411 0 841 28 2513 3 1978 1460 3411 0 428 31 2482 Capsule U Withdrawn i 1978 THROUGH 12 1978 0 3411 0 000 275 2207 1 1979 3032 3411 0 889 31 2176 ,

2 1979 3411 3411 1 00 28 2148 3 1979 3411 3411 1 00 31 2117 4 1979 2920 3411 0 856 30 2087 ,

5 1979 THROUGH 6 1979 0 3411 0 000 61 2026 7 1979 2742 3411 0 804 31 1995 8 1979 3360 3411 0 985 31 1964  !

9 1979 2965 3411 0 875 30 1934 I 10 1979 1146 3411 0 336 31 1903 11 1979 0 3411 0 000 30 1873 12 1979 7 3411 0 002 31 1842 l l

6 20 l 1

1

,. TABLE 6-6 (Continued) lRRADIATION HISTORY OF SURVEILLANCE CAPSULES REMOVED FROM THE TROJAN REACTOR P aya P ag, Irradiation Decaytal Month Year (MW) (MW) Pgya/ PREF Time (Days) Time (Days) 1 1980 3186 3411 0 93= 31 1811 2 1980 3090 3411 0906 29 1782 3 1980 2214 3411 0 649 31 1751 4 1980 754 3411 0 221 30 1721 5 1980 THAOUGH 6 1980 0 3411 0 000 61 1660 7 1980 938 3411 0 275 31 1629 8 1980 3196 3411 0 937 31 1598 9 1980 3404 3411 0 996 30 1568 to 1980 3363 3411 0 986 31 1537 11 1980 3401 3411 0 997 30 1507 12 1980 3049 3411 0 894 31 1476 1 1981 2974 3411 0 872 31 1445

. 2 1981 1907 3411 0 559 28 1417 3 1981 3411 3411 1 00 31 1386 4 1981 3220 3411 0 944 30 1356 5 1981 68 3411 0 020 31 1325 8 1981 0 3411 0 000 30 1295 7 1981 1088 3411 0 319 31 1264 8 1981 3169 3411 0 929 31 1233 9 1981 33-6 3411 0 981 30 1203 to 1981 2422 3411 0 710 31 1172 11 1981 3350 3411 0 982 30 1142 12 1981 3343 3411 0 980 31 1111 1 1982 2',62 3411 0 751 31 1080 2 1982 3166 3411 0 934 28 1052 3 1982 2470 3411 0 724 31 1021 4 1982 THROUGH 7 1982 0 3411 0 000 122 e99 8 1982 590 3411 0 173 31 868 9 1982 2923 3411 0 857 30 838 to 1982 3411 3411 1 00 31 807 11 1982 2698 3411 0 791 30 777

• • 12 1982 3411 3411 1 00 31 748 1 1983 2231 3411 0 654 31 115 .

• 2 1983 THROUGH 6 1983 0 3411 0 000 150 565 7 1983 467 3411 0 137 3t 534 8 1983 1985 3411 0582 31 503 9 1983 3237 3411 0 949 30 473 6 21

TABLE 6-6 (Continued) .

IRRADIATION HISTORY OF SURVEILLANCE CAPSULES REMOVED FROM THE TROJAN REACTOR P aya P REF Irradiation Decay (41 Month Year (MW) (MW) Paya/Png, Time (Days) Time (Days) t 10 1963 3303 3411 0 970 31 442 11 1963 3411 3411 1 00 30 412 13 1963 3363 3411 0 986 31 381 1 1984 3350 3411 0 982 31 350 8 1964 3114 3411 0 913 29 321 3 1964 3200 3411 0 938 31 290 4 1964 3086 3411 0005 27 263 Cacsuie X W4thdrawn Totd irradiation time = 1.35 x 108 EFPS a) Decay time is referenced to the counting date of the flux monitors (115-85). .

e 4

6 e

6 22

- ,o TABLE 6-7 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX MONITOR SATURATED ACTIVITIES FOR CAPSULE X Reaction Radial Saturated Activity and Location (dps/gm)

Axial Position (cm) Capsule X Calculated Fe54(n,p)Mn 54 Top 207.31 4.04 x 10e Middle 207.31 4.20 x 106 Bottom 207.31 4.24 x 106

- Average 4.16 x 106 4.97 x 106 Cu63(n.o)Co 60 Top 207.31 3.98 x 105 Middle 207.31 4.15 x 105 Bottom 207.31 4.15 x 105 Average 4.09 x 105 4.31 x 105 Nisa(n p)Cosa Top 207.31 6.07 x 107 Middle 207.31 6.37 x 107 Bottom 207.31 6.40 x 107 Average 6.28 x 107 7 47 x 107 Np237(n,f)Cs'37

, Middle 207.31 7.30 x 10 7 7.46 x 107 U 238(n,f)Cs137

, - Middle 207.31 9.03 x 106 6.47 x 106 6-23

TABLE 6-8 RESULTS OF FAST NEUTRON DOSIMETRY FOR CAPSULE X Adjusted Saturated Activity e (E > 1.0 MeV) 9 (E > 1.0 MeV)

(DPS/gm) 2 (n/cm -sec) (n/cm2)

Reaction Measured Calculated Measured Calculated Measured Calculated

$a Fe54(n.p)Mn 54 4.16 x 106 4.97 x 106 1.15 x 10" 1.36 x 10" 1.55 x 10'8 1.84 x 10'8 Cu63(n.o)Co 60 4.09 x 105 4.31 x 105 1.30 x 10" 1.36 x 10" 1.76 x 10'8 1.84 x 10'8 Nise(n p)Coss 6.28 x 10 7 7.47 x 10 7 1.14 x 10" 1.36 x 10" 1.54 x 10'8 1.84 x 10'8 Np237(n,f)Cs'37 7.30 x 107 7.46 x 10' 1.33 x 10" 1.36 x 10" 1.80 x 10'8 1.84 x 10'8 I

U23a(n f)Cs'37 7.68 x 106'I I 6.47 x 108 1.61 x 10"I*I 1.36 x 10" 2.17 x 10'8 'I 1.84 x 10'8

a. U* adjusted saturated activity has beeri multiphed by 0 85 to correct for 350 ppm US impunty.

'l

• ' a

. ,1

TABLE 6-9 RESULTS OF THERMAL NEUTRON DOSIMETRY FOR CAPSULE X Saturated Activity (dps/gm) kTh Axial Location Bare Cd covered (n/cm 2-sec)

Top 1.26 x 10s 6.47 x 107 1.66 x 10" Middle 1.14 x 10e 6.06 x 107 1.45 x 10" Bottom 1.20 x 108 6.31 x 107 1.55 x 10" Average 1.20 x 10s 6.28 x 107 1.55 x 10" 4

m

't e

4 l

6-25

TABLE 6-10

SUMMARY

OF FAST NEUTRON DOSIMETRY RESULTS FOR CAPSULE X Calculated Irradiation Vessel Vessel Time $(E >1.0 MeV) R(E 3 10 MeV) Lead Fluence Fluence g Basis (EFPS) (n/cm a-sec) (n/cm2) Factor 2 (n/cm ) (n/cm2)

Fe54(n.p)Mn 54 1.35 x 108 1.15 x 10" 1.55 x 10'S 4.76 3.26 x 10ia 3.87 x 10ia Avg. of all Dosimeters 1.35 x los 1.31 x 10" 1.77 x IO'S 4.76 3.72 x 10is 3.87 x 10'8 l

.e

• e TABLE 6-11 CALCULATED CURRENT AND EOL VESSEL EXPOSURE FOR TROJAN Current E (E > 1.0 MeV) EOL E (E > 1.0 MeV)

(n/cm2) (n/m2)

Location Calculated Calculated Vessel IR 3.87 x 10'8 2.89 x 10'S Vessel %T 2.22 x 10'8 1.65 x 10'S Vessel %T 4.48 x 10'7 3.35 x 10'8 Note: EOL fluences are based on operation at 3411 MWt for 32 effective fuit power years e

4 e

W

's e

6-27

r SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule is recommended for future capsules to be removed from the Trojan reactor vessel.

Lead Removal Estimated Fluence Capsule Factor Time!*l (10"n/cm 2)

U 4.76 1.23 0.388 (Actual)

. X 4.76 4.28 1.77 (Actual)

V 4.06 8 2.93tbl

~

Y 4.06 15 5.5 W 4.76 Standby Z 4.76 Standby

a. EFPY from plant startup

b. Approximate vessel end of life inner wall location fluence a gD

'O 7-1

SECTION 8 REFERENCES

1. Davidson, J. A., Phillips, J. H., and Yanichko, S. E., " Portland General Electric Company Trojan Unit 1 Reactor Vessel Radiation Surveillance Program,"

WCAP-8426 January,1975.

2. ASTM Standard E185-73, "Recomended Practice for Surveillance Tests for Nuclear Reactor Vessels" in ASTM Standards, Part 10 (1973), American Society for Testing and Materials, Philadelphia, Pa.1973.

3. Davidson, J. A., Anderson, S. L., and Kaiser, W. T., " Analysis of Capsule U from Portland General Electric Company Trojan Reactor Vessel Radiation Surveillance Program," WCAP-9469, May 1979.

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

5. Soltesz, R. G., Disney, R. K., Jedruch, J., and Zeigler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation. Vol. 5 - Two-Dimensional Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.

6. SAILOR RSIC Data Library Collection "DLC-76, Coupled, Self-shielded,47 Neutron e 20 Gamma-ray, P3, Cross Section Library for Light Water Reactors."

-e

. 7. Benchmark Testing of Wesinghouse Neutron Transport Analysis Methodology

- to be published.

8-1 t

i -

}

REFERENCE.S (cont) s

8. ASTM Designation E261-77, Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards (1983), Section 12, .

Nuclear Standards, pp. 76-87, American Society for Testing and Materials, Philadelphia, Pa.,1983.

9. ASTM Designation E262-77, Stande.rd Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards (1983), Section 12. Nuclear Standards, pp. 88-96, American Society for Testing and Materials, Philadelphia, Pa.,

1983.

10. ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron," in ASTM Standards (1983), Section 12, Nuclear Standards, pp.97-102, American Society for Testing and Materials, Philadelphia, Pa.,

1983.

11. ASTM Designation 481-78, " Standard Method of Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 228-235, American Society for Testing and Materials, Philadelphia, Pa.,1983.

12. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards (1983), Section 12, Nuclear Standards, pp. 103-107, American Society for Testing and Materials, Philadelphia, Pa.,1983.

i i

3-e i

8-2