ML20116C970
| ML20116C970 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/31/1992 |
| From: | Shaun Anderson, Meyer T, Munoz Frances Ramirez WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20116C965 | List: |
| References | |
| WCAP-13422, NUDOCS 9211050117 | |
| Download: ML20116C970 (189) | |
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E WESTINGHOUSE CLASS 3 WCAP-1347^
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- ANALYSIS OF CAPSULE U FROM THE TEXAS UTILITIES ELECTRIC COMPANY COMANCHE PEAK UNIT NO. 1 -
REACTOR VESSEL &
RADIATION SURVEILLANCE PROGRAM M .A. Ramirez S. L. Anderson P. L. Strauch
!e A. Madeyski Th p?';
July 1992 Work Performed Under Shop Order WCTP-6620A Prepared by Westinghouse Electric Corporation for the Texas Utilities Electric Company i- -
Approved by:
k T. A. Meyer, Manag&r c5 Structural Reliability and Plant Life Optimization
.a WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1992 Westinghouse Electric Corp. l All Rights Reserved l
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PREFACE ,
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-.This report has.been. technically reviewed and verified,
. Reviewer:
,. r Sections 1.through-5,-7, 8 and E. Terek Appendix A and B
- . .Section 6 E. P. Lippincott z N> '
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TABLE OF CONTENTS 4
S. :Section Title EAga
- " - 1. 0
SUMMARY
OF RESULTS- 1-1
2.0 INTRODUCTION
2-1
+' .
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 ' TESTING OF SPECIMENS FROM CAPSULE U 5-1 5.1 Overview 5 l
'5.2 .Charpy V-Notch Impact Test Results 5-4 -l 5.3- Tension' Test Results 5-6 5.4 Compact Tension Tests 5-7 i
e-6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1
.. 6.1 Introduction 6-1 l
6.2 ' Discrete Ordinates Analysis 6-2 j
'6.3 Neutron Dosimetry 6-7 7.0' SURVEILLANCE CAPSULE REMOVAL' SCHEDULE 7-1
8.0 REFERENCES
8-1 iAPPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS i
' APPENDIX'B - ~HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION O FOR COMANCHE-PEAK UNIT 1 FOR 16 AND'32'EFPY ls ~
-11 M
- 7. =
4 LIST OF TABLES -
c libJg- Title hag .
(
l 4-1 Chemical Composition of the Comanche Peak Unit 1 Reactor 4-3 l
Vessel Surveillance Material 1 4-2 Heat Treatment History of the Comanche Peak Unit 1 Reactor 4-4 Vessel Plate Surveillance Material 4-3 Heat Treatment History of the Comanche Peak Unit " Reactor 4-4 Vessel Weld Surveillance Material 5-1 Charpy V-Notch Impact Data for the Comanche Peak Unit 1 5-8 Lower Shell Plate R1108-2 Irradiated at 550'F, Fluence 3.70 x 1018 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Comanche Peak Unit 1 5-9 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 3.70 x 1018n/cm2 (E > 1.0 MeV)
~
5-3 Instrumented Charpy Impact Test Results for the Comanche 5 Peak Unit 1 Lower Shell Plate R1108-2 Irradiated- at 550'F, 18 Fluence 3.70 x 10 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results for the Comanche 5-11
~
Peak Unit 1 Weld Metal and HAZ Metal, Irradiated at 550*F,
' 18 Fluence 3.70 x 10 n/cm2 (E > 1.0 MeV) 5-5 Effect of 550*F Irradiation to 3.70 x 1018n/cm 2 5-12 (C > 1.0 MeV) on the Notch Toughness Properties of the Comanche Peak Un't 1 Reactor Vessel Surveillance Materials iii
LIST OF TABLES Table Title Egge ,
o l 5- 6 Comparison of the Comanche Peak Unit 1 Surveillance Material 5-13
- 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy .
Decreases with Regulatory Guide 1.99 Revision 2 Predictions -
1 57 Tensile Properties for the Comanche Peak Unit 1 Reactor 5-14 Vessel Surveillance Materials Irradiated at 550'F to 18 3.70 x 10 n/cm2 (E > 1.0 MeVi ;
6-1 Calculated Fast Neutron-Exposure Parameters at the 6-14 Surveillance Capsule Center 6-2 ;- Calcul'ated Fast Neutron Exposure Rates at the 6-15 Pressure Vessel Clad / Base Metal Interface .
'6-3 ' Relative Radia Distributions of Neutron Flux 6-16 ,
(E > 1.0 MeV) within the Pressure Vessel Wall 16-41 -Relative Radial Distributions of Neutron Flux 6-17 e (E > 0.1 HeV) within the Pressure Vessel Wall 6-5 Relative Radial' Distributions.of Iron Displacement Rate 6-18 (dpa) within the Pressure Vessel Wall 6 Nuclear-Parameters for Neutron Flux Monitors 6-19 ,"
6-7 . Honthly Thermal Generation.During the First Fuel Cycle 6-20 of the Comanche Peak Unit 1 Reactor 6-8 Measured Sensor Activities and Reactions Rates 6-21 iv
'f
r LIST OF-TABLES p;
h.) Ithlg Title EARt b 6-9 . Summary 'of Neutron Dosimetry Results 6-23 L6-10 -Comparison of Measured and FERRET Calculated Reaction 6-24 Rates at the Surveillance Capsule Center 6-11 AdjustedNeutronEnergySpectrumattheSurveillance 6-25 -_
Capsule Center-
.6 Comparison of Calculated and Measured Exposure Levels 6-26 for Capsule U 6-13. Neutron Exposure Projections at Key Locations on the 6-27
- , ' Pressure Vessel Clad / Base Metal Interface
.. 6-14 ; Neutron Exposure-Values at the Surface, 1/41 and 3/4T 6-29 Locations for 16 and 32 EFPY h- 6-15 Updated Lead Factors for the Comanche Peak Unit 1 6-30 Surveillance Capsules
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LIST OF ILLUSTRATIONS Fiaure Title Eing .
4-1 Arrangement of Surveillance Capsules in the Comanche Peak 4-5 Unit 1 Reactor Vessel 4-2 Specimen locations-in the Comanche Peak Unit 1 Reactor 4u -
Surveillance ' st Capsule 'U" 5-1 Charpy V-Notch Impact Properties for Comanche Peak Unit 1 5-15 -
Reactor Vessel Lower Shell Plate R1108-2 (Longitudinal Orientatica) 5-2 Charpy V-Notch Impact Properties for Comanche Peak Unit 1 5-16
. Reactor Vessel Lower Shell Plate R1108-2 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for Comanche Peak Unit 1 5-17 Reactor Vessel Surveillance Weld Metal ,
5-4 Charpy V Notch Impact Properties for Comanche Peak Unit 1 5-18 Reactor Vessel Weld Heat-Affected-Zone Met,,1 5-5 Charpy impact Specimen Fracture Surfaces of the Comanche 5-19 Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 (longitudinal Orientation) 5-6 Charpy Impact. Specimen Fracture Surfaces of the Comanche 5-20 ,"
Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 '
(Transverse nrientation) vi
gr q LIST OF ILLUSTRATIONS (Cont)
-a ,
Fiaure Title flagg 5-7 Charpy: Impact Specimen Fracture Surfaces of the Comanche 5-21
_ Peak Unit 1 Reactor Vessel Surveillance Weld ittal 5 Charpy Impact Specimen Fracture Surfaces of the Comanche 5-22
- - Peak Unit 1 Reactor Vessel Weld Heat-Affected-Zone Metal p -- 5-9 Tensile Properties for Comanche Peak Unit 1 Reactor 5-23 Vessel. Lower Shell Plate R1108-2 (longitudinal Orientation) 5-10'~ Tensile Properties for Comanche Peak Unit 1 Reactor 5-24
-Vessel Lower Shell Plate R1108-2 (Transverse Drientation) 5-11 Tensile' Properties for Comanche Peak Unit 1 Reactor 5-25 Vessel Surveillance Weld Metal
,.-- 5-12 Fractured-Tensile Specimens from Comanche Peak Unit 1 5-26
' Reactor Vessel Lo er Shell Plate R1108 (Longitudinal Orientation) 5-13 Fractured Tensile Specimens from Comanche Peak Unit 1 5-27 Reactor Vessel Lower Shell P1 ate R1108-2 (Transverse Orientation)
~
5-14 ~ Fractured Tensile Specimens from Comanche Peak Unit 1 5-28
( Reactor Vessel Surveillance Weld Metal
$ .5 - Engineering Stress-Strain Curves for Lower Shell Plate 5-29
.; f R1108-2' Tensile Specimens TL1 and TL2 (Longitudinal Orientation) y vii 5 ,-
> T LIST-0F ILLUSTRATIONS (Cont) 2 fiqute Title P.iLqa -
5'-l6 - . Engineering-Stress-Strain curve for Lower "keil Plate 5-30 - -
R)'08-2 Tensilo Specimen TL3 (longitudinal Orsintation)
'5 Engineering Stress " rain Curves for Lower Shell Plate 5-31 *
.R1108-2 Tensile Specimens TTI and TT2 -
. sTransverse Orientation) 5-18 Engineering Stress-Strain Curve for Lower Shell flate 5-?2 R1108-2 Tensile Specimen TT3 (Transverse Or, -tat on)
. 5-19 '. Engineering Stress-Strain Curves for Weld Metal 5-33 Tensile Specimens TW1 and TW2 5-20 Engineering Stress-Strain Curve for Weld Metal 5-34 ,
Tensi e Specimen TW3 6-1 Plan View'of a Dual Reactor Vessel Surveillance Capsula 6-13 9
9 m
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viii i.i
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SECTION 1.0 SUMKARY OF RESULTS The analysis of the reactor vessel materials contained in surveillance capsule U, the fir.st capsuls to be removed from the tomanche Peak Unit I reactor pressure vessel, led to the following conclusions:
. o The capsu'e rectived an average fast neutron fluence (E > 1.0 MeV) of 3.70 x 10 18n jen2 af ter 0.91 EFPY of plant operation.
o Irr-diation of th9 reactor vessel lower shell plate R1108-2 Charpy specimens to 3.70 x 10 18 n/cm2 (E > 1.0 MeV) resulted in 30 ft-lb and 50 f t-lb transition temperature increases of 10*F and 5'F, respectively, for Charpy specimens oriented with the longitudinal axis parallel to the major rolling direction of the plate (longitudinal orientation). This results in a 30 ft-lb transition temperature of -10*F and a 50 ft-lb transition
~
temperature of 10'F.
o Ir.adiation of the reactor vessel lower shell plate R1108-2 Charpy specimens to 3.70 x 10 18 n/cm2 (E s 1.0 MeV, resulted in 30 ft-lb Y and 50 ft-lb transition temperature increases of 15'F and 20'F, respectively, for Charpy specim...s oriented with the longitudinal axis normal te .he major rolling riirection of the plate (transverse orientation). This results in c 30 ft-lu transition temperaturc of 15'F and a 50 ft ,b transition temperature of 75'F.
o The weld stal Charpy specimens irradiated to 3.70 x 1018n/cm2 ;E
> 1.0 MeVi resulted in 30 ft-lb and 50 ft-lb transition temperature u.reases of O'F. The 30 ft-lb and 50 ft-lb transition temperatures retnined at -70*F and -35'F, respectively, for
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the weld metal.
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- c. 'o Irradiation of'the reactor vessel weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 3.70 x 10 I8 n/cm2 (E > 1.0 MeV) ru ulted in 30 ft-lb and 50 ft-lb transition temperature increases of 0*F and 3*F, respectively.. Tht: results in a 30 ft-lb transition ;
temperature of -110*F and a 50 ft-lb transition temperature of .
-73*F for the weld HAZ metal, ,
I o The average upper shelf energy of lower shell plate R1108-2 .
(longitudinal orientation) resulted in an energy decrease of 6.5 .
18 l ft-lb after irradiation to 3.70 x 10 n/cm2 (E > 1.0 MeV). This -
results in an upper shelf energy of 123.5 ft-lb for longitudinally oriented specimens. ,
-o The average upper shelf energy of lower shell plate R1108-2 (transverse orientation) resulted in no energy decrease after irradiation to 3.70 x 1018 n/cm2 (E > 1.0 HeV). This results in ;
an upper shelf energy of 78 ft-lb for transversely oriented specimens.
o The (verage upper shelf energy of the weld metal resulted in no '
18 decrease after. irradiation to 3.70 x 10 n/cm2 (E > 1.0 MeV).
This results in an upper shelf energy of 125 ft-lb for the weld metal, o The average uppw shelf energy of the weld HAZ metal resulted in no decrease after-irradiation to 3.70 x 1018n/cm2 (E > 1.0 MeV).
This results in an upper shelf energy of 119 ft-lb for the weld HAZ metal.
~ "
o The surveillance capsule U test results demonstrate that the surveillance material 30 ft-lb transition temperature changes and upper shelf' energy decreases are less than the Regulatory Guide 1.99, Revision 2 predictions.
v 6
1-2 i
r o The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb
> throughout the life (32 EFPY) of the vessel as required by 10CFR50, Appendix G. '
o The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the Comanche Peak Unit I reactor vessel is as follows:
2
. Vessel inner radius * - 3.04 x 1019 n/cm
- 2 Vessel' 1/4 thickness - 1.66 x 10l9 n/cm 2
. Vessel 3/4 thickness - 3.59 x 1018 n/cm
- Clad / base metal interface
+
9
(;
e o'
9.
a N
1 . . ___. .
L SECTION 2.0 INTRODUCTION This report presents the results of the examination of Capsule U, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Comanche Peak Unit 1 !
reactor pressure vessel materials under actual operating conditions.
The surveillance progr.m for the Comanche Peak Unit- 1 reactor pressure vessel materials was designed and' recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties or the reactor vessel materials is presented in WCAP-9475 entitled " Texas Utilities Comanche Peak Unit No. 1 Reactor Vessel Radiation Surveillance Program" by W. T. Kaiser, et al Ill. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule "U" from the reactor and its shipment to the Westinghouse Scicnce and Technology Center llot Cell Facility, where, the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance capsule "U" removed from the Comanche Peak Unit I reactor vessel and discusses the analysis of the data. -
9 j.*
- e.
2-1
SECTION
3.0 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 ruclear 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 allny, ferritic .
pressure vessel steels such as SA533 Grade B Class 1 (base material of the Comanche Petk Unit I reactor pressure vessel core region) 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 Nonductile failure,"
~
Appendix G to Section 111 of the liSME Boiler and Pressure Vessel CodeI43 The method uses-fracture mechanics concepts and-is based on the reference nil-ductility temperature (RTNDT)-
RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208[5]) or the temperature 60'f less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined
'from Charpy specimens oriented normal (transverse) to the major rolling direction of the plate. The RTNDT of a given material is used to index tnat material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G to the ASME Code. The KIR curve is a lower bound of dynamic.-crack arrest, and static fracture toughness results obtained from
- several heats of pressure vessel steel. When a given material is indexed to th( KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be
~
determined using these allowable stress intensity factors, e
3-1
-- =
3 m:~~
f , j'jv
- . , f
' r f I:k ,,,
qq/.}'(
fl " ' sng '.s, in turn, the operating limits of nuclear power plants can be klO .idjusted to account for the effects of radiation on the reactor vessel material properties. The radiation ombrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor vessel surveillance program, such as the Comanche Peak Unit 1 Reactor Vessel Radiation .
Surveillance Programill, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens .
tested. The increase in the average Charpy V-nctch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust ,
~
the RTHD!
for radiation embrittlement. .This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the material to the KIR curve and,
-in turn, to set operating limits for the nucitar power plant which take into account the effects of irradiation on the reactor assel materials.
9' G
e L
l .
3-2
-. sg n<,s- 9my y 4 y = n ---w - --s y m w yg--
r _ _ _ _ _ _ _ _ - _ . _ _ - _ _ _ _ - - _ _ _ _ _ - _ _ _ _ _ _ .
e i ,I i ,
L
-SECTION 4.0 i t DESCRIPTION OF PROGRAM-
- u. Six surveillance capsules for monitoring the effects of neutron exposure on the ,
Comanche Peak Unit I reactor pressure vetdel core region material were inserted
.' in'the reactor vessel prior to initial plant start-up. The six capsules were !
. positioned -in the reactor-vessel between the neutron pads and the vessel wall !
at locations _shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.
9 Capsule U was removed after 0.91 Effective Full Power Years (EFPY) of plant' operation. This' capsule contained Darpy Y-notch, tensile, and 1/2 T compact ;
tension _(CT) test _ specimens from the reactor vessel 9-5/8-inch-thick lower -
shell plate R1108-2 and weld metal representative of the beltline region weld i seams of the reactor vessel. Capsule 0 also contained Charpy V-notch specimens ;
from the weld _ heat-affected-zone (HAZ) of plate R1108-2.
l
-The test specimens included- forty-five Charpy_V-notch, nine tensile, and twelve compact tension specimens (Figure-4-2) from the lower shell plate R1108-2 and
- . from a weldment made/from sections'of lower shell plates R1108-2 and R1108-1 using weld. wire identical to that used in the original vessel fabrication for all. core region weld-seams. The test specimens also' included fifteen Charpy
.V-notch from the weld HAZ of plate R1108-2.
The test' material was obtained from the lower shell' plate R1108-2 after the heat: treatment ano- forming of the plate. - All test specimens were machined from the .1/4 thickness. location of the plate, after performing a simulated postweld stress-relieving treatment on the test material, and also from the weld and _
heat-affected-zone metal of a stress relieved weldment' joining lower shell
- ,p later R1108-2 and R1108-1. _The test specimens represanted material taken at
-least-one plate thickness from the quenched ends -of the plate.
> Base metal Charpy. V-notch'_ impact 'and tensile test specimens were oriented with lthe'longitudinalf axistof_.the specimen parallel to the major rolling direction
~
' of the plate-(1_ongitudi.nal; orientation)-and with the longitudinal axis of the ,
g-ee ,
4-1 il
specimen normal to the major rolling direction of the plate (transverse
, orientation). Charpy V-notch and tensile test specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding direction. The notch in the charpy specimens was machined such that .
the direction of crack propagation in the specimen was in the welding direction. .
Compact tension test specimens from lower shell plate R1108-2 were machined in , l both the transverse and longitudinal orientations. Compact tension test -
l specimens from the weld metal were machined normal to the weld direction with .
the notch oriented in the direction of the weld . All specimens were fatigue
- precracked according to ASTM E399.
The chemistry and heat treatment history of the surveillance materials are
- presented in Tables 4-1 and 4-2, respectively. The chemical analyses reported in Table 4-1 are the results of an independent analysis performed by Combustion Engineering on the unirradiated surveillance materials.
Capsule V also contained dosimeter wires of pure iron, copper, nickel, and ^
aluminum 0.15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium ,
(U238) were placed in the capsule to measure the integrated flux at specific neutron energy levels.
Thermal monitors made from the two low-melting eutectic alloys, sealed in Pyrex tubes, and inserted in spacers, were included in the capsule. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two alloys and their
-melting points are as follows:
2.5% Ag, 97.5% Pb Melting Point: 579'F (304*C) '
1.75%-Ag, 0.75% Sn, 97.5% Pb Helting Point: 590'F (310*C)
The arrangement of the various mechanical- specimens, dosimeters and thermal monitors contained in capsule U is shown in Figure 4-2.
4-2
y -
TABLE 4 f I
CHEMICAL COMPOSITION OF THE COMANCHE PEAK UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIAL Ill
- u. :
Chemical Composition (wt%)(a) _,
- p. , Element Lower Shell Plate R1108-2 Weld Metal (c)
C- 0.210 0.190 ,
.; Mn 1.360 1.330 I
- e. P- 0.006 0.004 _,
- 0.012 S 0.006
'Si 0.230- 0.170 ,
Ni 0.590 0.220(d) l Cr 0.020 0.040 Mo' O.520 0.580 Cu' O.050 0.040 1l 0.002 0.005 N 0.007 0.006 Sn- 0.002 0.001 A1 0.008 0.006 Co 0.012 0.006
- B- <0.001 <0.001
Ti- <0.010 <0.010 W: <0.010 <0.010
.As 0.004 .<0.001 Zr- 0.001- <0.001
$b <0.001 (b) 0.0013 (a)< Results of analysis performed by Combustion Engineering.
~
- " -(b)'Results 'of analysis performed by Westinghouse.
- -(c) Weld / wire type: B4,1 heat- number 88112, flux type Linde 0091, and flux lot number 0145;isurveillance weldment obtained from weld-between lower ,
} shell plates. R1108 and- R1108-1.
.(d) _ Note, this~ value of %Ni- content is obtained from a chemistry analysis Ldone only on th~e weld surveillance material. Thus, this value differs
[
--from the %Ni content per Combustion l Engineering material certifications.
- ;4-3
~
I TABLE 4-2 HEAT TREATMENT HISTORY OF THE CDMANCHE PEAK UNIT 1 -
REACTOR VESSEL PLATE SURVEILLANCE MATERIALIll Material Temperature Time Coolant ,'
Lower Shell- 1600 1 25'F 4 hrs Water Quenched .
Plate R1108-2 1225 1 25'r. 4 hrs Air Cooled 1150 1 50*F 11 hrs Furnace Cooled TABLE 4-3 HEAT TREATMENT HISTORY OF THE COMANCHE PEAK UNIT 1
~
REACTOR VESSEL WELD SURVEILLANCE MATERIAllll Material Temperature Time Coolant Weld 1150 1 25'F 7. hrs 50 min. Furnace Cooled O
e 9
4-4
O' REACTOR VESSE CORE BARREL NEUTRON PAD
~
(301.5') 2 ,
CAPSULE U (58.5')
, p* sg,ge #D
~
58.5' +,
61*
270' 90' (241 * ) Y ]
W (121.5')
(230.5') . X REACTOR vEssa 180' .
PLAN VIEW l 6
~
~
h VESSEL
\ : y ou.
) l]..
(s/ CAPSULE ASSEMBLY llllllllQ : '
>' CORE Q MIDPLANE s i N f , %
i ~
D NEUTRON PAD
^
+ / CORE BARREL kJ ELEVATION VIEW Figure 4-1. Arrangement of Surveillance Capsules in the Comanche Peak Unit 1 Reactor Vessel 4-5 i
- w. -
1 1
4 1
i 3
- i l !
l j .
A 1>
]:*
1 1 <
y
'3- .
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"?
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1 i, a 1. s .. w N l' N1 11. 1.
tu le M.
la 1.*
)..= T 3 O WP the fW) TV7 TW1 to is 11 g *
- Il8 TL3 'lU Ill j
4 u __
rn ro t:-
4, e 7 y
'l !
L(Cf4D: it - lower tvell Piste #1106 ? (N1
- C403 f longitudteell
)
f jj fi . Lower Shell Plate 91108 t (Hi, - (8533-7 treesverse)
Th . Weld Metal
..; 1. . .e.t. A,,, t ,d. r o e ~ 1 e.4. ,
i i
1-i 4
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i i
0 1
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i i er*% w y i
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<an m comi teswntes et=suti c,,a m i cua ns tsant c>am puns co-ans erans trasuts I
"T" T .,
TT n r 't i T6 'T n TT a' w
^
r t E b 2 Y Y Y Y b i-- + fr t Y 7 4 4 1 4 4-i 5" m f.n A ra- p 4p p L-
. 3 4,. y At t- t i 3
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SI APERTURE CARD ..
AN Available On Ap(i tur e Card Figure 4-2 Specimen Locations in the i Comanche Peak Unit 1 Reactor l Surveillance Test Capsule "U"
. 5 4-6
SECTION 5.0 TESTING 0F SPECIMENS TROM CAPSULE U f
- 5.1 Overview The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center hot !
cell with consultation by Westinghouse Power Systems personnel. Testing was performed in accordance with 10CFR!,0, Appendices G and HI23, ASTM <
. Specification E185-82[6), and Westinghouse Procedere RMF 8402, Revision 1, as modified by Westinghouse RMF Proceoures 8102, Revision 1 and 8103, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9475Ill. No discrepancies were found.
Examination of the two low-melting point 579'F (304'C) and 590*F (310*C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579'T (304'C).
The Charpy impact tests were performed per ASTM Specification E23 88l73 and RMF Procedure 8103, Revision 1, on a Tinius Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with a GRC 8301 instrumentation system feeding information into'an IBM XT computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED ), From the load-time curve
- (Appendix A), the-load of general yielding (Pay), the time to general yielding (tcy), the maximum load (PM_ ), and the time to maximum load (tM) can-be determined. Under some test conditions, a sharp drop in load indicative cf fast fracture was observed. The load at which fast fracture was initiated is identified as _ the- fast fracture load (Pp), 'and the load at-which fast fracture terminated is -identified as the arrest load-(PA )'
9 5-1
W The energy at caximua load (Ep,) was determined by comparing the energy-time record 'and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen.
Therefore, the. propagation energy for the crack (E p ) is the difference between the total energy to fracture (E )Dand the energy at maximum load.
The yield stress (oy) was calculated from the three-point bend formula having the following expression: ,
oy - Pay * {L/(B*(W-a)2*C)) (1) .
where L - df '
- ace between the specimen supports in the impact testing machine; B = the width of the specimen measured parallel to the noah; W = height of the specimen, measured perpendicularly to the notch; a - notch depth. The constant C is dependent on the notch flank angle (4), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending).
In three-point bending a Charpy specimen in which p - 45' and p -
0.010', Equation 1 is valid with C - 1.21. Therefore (for L = 4W), ,
oy - Pay * {L/[B*(W-a)2*l.21]) - [3.3PayW)/[B(W-a)2] (2) ,
For the Charpy specimens, B = 0.394 in., W = 0.394 in., and a = 0.079 in.
Equation 2 then reduces to:
oy - 33.3 x Pay (3) where oy is in units of psi and Pay is in units of lbs. The flow '
stress was calcula ed from the average of the yield and maximum loads, also using the three-point bend formula. ,'
- Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in ' compliance with' ASTM Specification A370-89[8),
The _ lateral expansion _was measured using a dial gage rig similar to that shown in the same specification.
5-2
r Tensile tests mre performed on a 20,000-pound Instron Model 1115, split-console test machbe, per ASTM Specification E8-89bl93 and E21-79
.'(1988) (10), and RMF Procedure 8102, Revision 1. 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.
j,; Extension measurements were made with a linear variable displacement transducer
'4 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 E83-85(11),
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 temperatures.
Chromel-alumel thermocouples of a type K with a no. I wire gage (0.030in.
~
diameter) were used. The 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 of room temperature to 550'T (288'C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimcn temperatures. Experiments 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, 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 tne fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5-3
5.2 Charov V-Notch Imoact Test Results The results of the Charpy V-notch impact tests performed on the various materials .
contained in Capsule U, which was irradiated to 3.70 x 1018 n/cm2 (E > 1.0 MeV), in 0.91 EFPY of operation, are presented in Tables 5-1 through 5-4 and are ,
compared with unirradiated resultsIll as shown in Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-5. .'
Irradiation of the reactor vessel lower shell plate R1108-2 Charpy specimens .
oriented with the. longitudinal axis of the specimen parallel to the major rolling direction of the plate-(longitudinal orientation) to 3.70 x 10 18 n/cm2 (E > :
1.0 MeV)Lat 550'F (Figure 5-1) resulted in 30 ft-lb and 50 ft-lb transition I
temperature increases of 10'F'and 5'F, respectively. This resulted in a
-30 ft-lb transition-temperature of -10*F and a 50 ft-lb transition temperature of 10'F (longitudinal orientation). .;
- The t.verage upper shelf energy (USE) of the lower shell plate R1108-2 Charpy
? specimens (longitudinal. orientation) resulted in an energy decrease of 7 ft-lb after irradiation 1to 3.70 x 1018 n/cm2 (E > 1.0 MeV) at 550'F. This results in an average USE of 124 ft-lb (Figure 5-1)..
Irradiation:of the reactor vessel lower shell plate R1108-2 Charpy specimens
-oriented with the longitudinal axis-of the specimen normal to the major rolling direct' ion.of the plate (transverse orientation) to 3.70 x 1018n/cm2 (E > 1.0 MeV) at 550'F (Figure 5-2) resulted in 00 ft-lb and 50 ft-lb transition temperature increases of 15'F and 20*F, respectively. This resulted in a 30 ft-lb transition temperature of 15'F and a 50 ft-lb transition temperature -
of 75'F _ (transverse orientation). .
The average upper shelf. energy .(USE) of the lower shell plate R1108-2 Charpy specimens (transverse orientation) resulted in no energy decrease after irradiation to 3.70 x11018n/cm2 (E >-l.0 Mev) at 550'F. This results in an average USE of 78 ft-lb (Figure 5-2). -
5-4 I:
, . . . . _ . . , . . , , , _ ,_,.,__._._,m,m m .. . , , - , ~ _ . s...,m,_m. ,_,-m..__ . . _ . _ _ . . .
r Irradiation of the reactor vessel core region weld metal Charpy specimens to 3.70 x 1018 n/cm2 (E > 1.0 MeV) at 550*f (figure 5-3) resulted in no 30 ahd 50 ft"1b transition temperature increaser. The 30 ft-1b and 50 ft-lb transition temperatures remained at -70'f and -35'f, respectively.
The average upper shelf energy (USE) of the reactor vessel core region weld '
metal resulted in no energy decrease after irradiation to 3.70 x 10 18n/cm 2
. - (E > 1.0 MeV) at 550'f. This results in an average USE of 125 ft-lb (figure 5-3).
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 3.70 x 1018 n/cm2 (E > 1.0 MeV) at 550'f (figure 5-4) resulted in no 30 ft-lb transition temperature increase r.nd in a 50 ft-lb transition temperature increase of 3'f. This results in a 30 ft-lb transition temperature of -Il0'r and a 50 ft-lb transition temperature of -73'f.
The average upper shelf energy (USE) of the reactor vessel weld HAZ metal resulted in no energy decrease after irradiation to 3.70 x 1018 n/cm2 (E >
l.0 MeV) at 550'F. This results in an average USE of 119 ft-lb (Figure 5-4).
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in figures 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasing test temperature.
A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the various Comanche Peak Unit I surveillance materials
- with predicted values using the methods of NRC Regulato*y Guide 1.99, Revision
- 2(3) is presented in Table 5-6. This compariton indicates that the transition temperature increases and the USE decreases .esulting from irradiation.to 3.70 x 1018 n/cm2 (E > 1.0 MeV) are less than the NRC
' Regulatory Guide 1.99, Revision'2I33, predictions.
9 5-5
p The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and-are expected to maintain an upper shelf energy of no 10ss than 50 ft-lb throughout the life (32 EFPY) of .
the vessel as required by 10CFR50, Appendix G.
The load-ttw rec >rds for the individual instrumented Charpy specimen tests are shown in Appendix A. ,
All broken test specimens will be stored at the Westinghouse Science and .
Technology Center. - i 5.3 Tension Test Results The results of the tension tests performed on the various materials contained in capsulo V irradiated to 3.70 x 10 n/cm2 18 (E > 1.0 MeV) at 550'F are presented in Table 5-7 and are compared with unirradiated resultsill as shown in Figures 5-9 through 5-11.
The results of the tension tests prirformed on the lower shell plate R1108-2 *
(longitudinal orientation) indicated that irradiation to 3.70 x 1018 n/cm 2 (E > 1.0 MeV) at 550'F caused a 2 to 3 ksi increase in the 0.2 percent offset yield strength and-.a 2 to 3 ksi increase in the ultimate tensile strength when compared to unirradiated dataill (Figure 5-9).
The results of the tension tests performed on the lower shell plate R1108-2 (transverse orientation) indicated that irradiation to 3.70 x 1018n/cm2 E > 1.0 MeV)-at 550'F caused less than a 2 ksi increase in the 0.2 percent offset yield strength and a-1 to 3 ksi increase in the ultimate tensile
-streigth when compared to unirradiated dataIll (Figure'5-10).
5-6
- - - -, , -, -w, -
y~, '
,p h s ;.y};)
nn ;
l?s
si, y, m m
%The riisube of the tension-tests performed on the reactor vessel core region I i'~
- weld meul irNicated'that irr3diation to 3.70:x 1018n/cm2 (c > 3,o g4,y) at.
- 550*Fcausef4/to'5ksi~;i,ncrease.inthe0.2percentoffsetyieldstrength l and a 1 W 4 2 ksi increase in the ultimate tensile' strength when coapared to !
..- s. unirra# ah; tatkIII (Figure 5-l% l l
The fractured tension specimens for the lower shell plate R1108-2 material are
[. shown in Figures 5'i2 and 5-13, while the fractured specimens for the weld
.-- - met al . are shoen in figure 5-14. -
.~
- The small increaits in 0.2% yield strength and ultimate tenslie *trength I
exhibited by lower shell plate R1108-2 and the weld metal' indicate that these materials'are not highly sensitive to irradiation to 3.70 x.1018 n/cm2 (E >
.l.0'NY), as issilso indicated by the Charpy impact test results.
The eng'ineering stress-strain -curves -for the tension tests are shown in Figures 15 through 5-20. 4 i
J'- All bNken test ~ specimens will be stored at the Westinghouse Science and Technology Center.. !
< r 15.4C Comoact Tension Tests Per the _ surveillance capsule testing program with Texas Utilities, the 1/2T -
':ompact
. tension fracture mechanics specimens will not be-tested and will'be
. stored at the Westinghouse Science and Technology Center.
t
)
i? l e 5-7:
ap 1
i
, . . , , , , ,,, , . , . ,.a_,.,n..,,., -,. . . - - . , , . , . . . , ~ , , - ,n, .,. , , , , _ , . . , . . , , , ~ . . ,
L TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE COMANCHE PEAK UNIT 1 LOWER SHELL PLATE R1108-2 1RRADIATED AT 550'F, .
18 FLUENCE 3.70 x 10 n/cm2 (E > 1.0 MeV) ,
Sample.NA $ $ $$$ Y m $s$ O $ .
Longitudinal 0 entation -
This WL -CQ 12 11 0.28 10 TL14 -50 -
9 8 0.20 10 TL11 -25 - 13 8 0.20 15 TL9 -10. -
23 14 0.36 20 TL10 0 -
49 32 0.81 35 TL2- 5 -
59 46 1.17 45 TL3 15 -
63 42 1.07 55 TLS 25 -
85 46 1.17 60 TL1 30 -
84 44 1.12 55 TL13 50 116 67 1.70 100 TL4 -100 87 59 1.50 80 TL7 200 125 80 2.03 100 TL8 240 123 72 1.83 100 TL5 275 134 82 2.08 100 Transverse Or ntation ,
TT9 -50 -
17 14 0 36 10 TT4 -25 -
17 15 0.38 10 TT5 0 -
14 13 0.33 15 TT2 15 -
37 32 0.81 20 TT7 25 -
30 29 0.74 25 TT8 50 40 39 0.99 35 TT13 60 38- 32 0.81 35 TT11 75 56 45 1.14 60 TT12 100 54 47 1.19 60 TT1 125 68 57 1.45 70 TT10 150 73 61 1.55 90 -
TT14 200 - 78 65 1.65 100 TT6 225 1 97 69 1.75 100 TT15 260 1 85 61 1.55 100 -
- TT3' NA NA NA NA NA NA ,
-
- Specimen has'been misplaced. However, this has no technical significance'on:the validity of the data presented in_this
- report, since 15 specimens were-tested. When the specimen -
is le.cated, it will bo stored in-the Hot Cell.
5-8
TABLE 5-2 I CHARPY V-NOTCH IMPACT DATA FOR THE COMANL PEAK UNIT 1 REACTOR VESSEL WELD HETAL AND HAZ HETAL 4RRADIATED
- AT 550*F, FLUENCE 3.70 x 10 18 n/cm2 (E > 1.0 MeV)
Sample No. f 0$ ) ) eis$ mm f Vleld Metul TW3 -135 - 93 4 5 1 0.03 1 TW12 -100 - 73 32 18 0.46 25 TW2 - 90 - 68 29 16 0.41 25 TW6 - 75 - 59 43 27 0.69 35 TW4 - 60 - 51 32 22 0.56 40 TW9 - 50 - 46 49 32 0.81 50 TW13 - 35 - 37 53 32 0.81 60 TW7 - 25 - 32 55 34 0.86 60 TW11 0 - 18 88 54 1.37 80 TW1 35 2 99 67 1.70 95 TW15 50 10 122 75 1.01 100 TW10 100 38 123 82 2.08 100
, TW5 150 66 115 78 1.98 100 TY!8 200 93 136 85 2.16 100 TW14 250 121 133 83 2.11 100 HAZ Metal TH6 -125 - 87 16 8 0.20) 10 TH7 -115 - 82 38 24 0.61 25 TR2 -110 - 79 57 38 0.97 40
?B15 -100 - 73 35 23 0.58 30 TD1 - 00 - 68 44 26 0.66 35 T'd10 - 75 - 59 47 33 0.84 35 TH3- - 60 - 51 61 43 1.09 50
'i35 - 50' - 46 38 28 0.71 30 TH8 - 40 - 40 73 51 1.30 60 TH14 - 25 - 32 72 49 1.24 60
, TH9 0 - 13 107 1 72 1.83 100 TH12 TH4 0
50
- 18 137 fl 80 2.03 100 10 134 77 1.96 100
. TH11 100 38 110 74 1.88 100 TH13 150 66 115 80 2.03 100 e
5-9
TABLE 5-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE COMANCHE PEAK UNIT 1 LOWER SHELL PLATE R1108-2 IRRADIATED AT 550*F, FLUENCE 3.70 x 10 18 n/cm2 (E > 1.0 MeV)
Normalised Fuermies Test Charpy Charpy Maximum Prop Yield Time Maximum ' Time to Fracture Arrest Tield Flow Sample Temp Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Strees Strees liumber (*F) (ft-lb) (f t-lb/in N) (Ibel (mece) (Ibe) (weec) (lbs) (Ibel (kel) (kei)
Lonstitudinal orientation TL15 - 75 12 97 84 13 3407 0.63 4274 0.71 4274 144 113 128 TL14 - SO 9 72 45 28 3211 0.13 3390 0.17 3390 50 107 110 TL11 - 25 13 105 36 89 3436 0.33 3749 0.15 3591 252 111 119 TL9 - 10 23 185 146 39 3439 0.15 4298 0.38 4298 228 114 128 TL10 0 49 395 286 109 3169 0.13 4534 0.62 4394 394 los 128 TL2 5 59 475 369 106 3304 0.13 4450 0.01 4450 248 110 129
, TL3 15 63 507 32'4 184 3283 0.13 4532 0.69 4352 831 109 130 TLS 25 65 523 .. .. .. .. .* ** ** *. .. *.
TL1 30 64 515- 262 253 3256 0.13 4465 0.59 4140 826 108 128 TL13 50 58 467 357 110 3314 0.13 4468 0.77 4394 855 110 129 i TL4 100 87 701 355 346 3141 0.14 4414 0.77 4014 2218 104 125 l- TL12 150 112 902 280 622 3033 0.50 4347 0.97 *
- 101 123 TL7. 200 125 1007 398 609 2SS7 0.12 3989 0.95 .
- RS 109
' u, TL8 240 1".3 990 332 S58 2646 0.14 4129 0.79 . . 88 113 8
TL5 275 134 1079 371 708 2002 0.13 4004 0.91 . . 86 110 o
Transverse Orientation TT9 - 50 17 137 110 26 3501 0.13 4191 0.30 4181 64 116 128 l
TT4 - 25 17 137 94 43 3232 0.13 3950 0.27 3950 122 107 119 i
T*5 0 14 113 56 57 3412 0.14 3450 0.20 3450 700 113 114 i TT2 15 37 208 233 65 3131 0.13 4326 0.54 4326 601 104 124
- TT7 25 30 242 157 85 3073 0.14 4077 0.41 4077 852 102 110 i
TT8 50 40 323 186 136 3181 0.17 4200 0.47 4200 1213 106 123 TT13 60 38 306 193 113 2967 0.12 4246 0.47 4246 1104 99 120 TT11 75 56 451 231 220 3167 0.14 4236 0.55 3971 1998 105 123 TT12 100 54 435 240 195 3314 0.22 4076 0.60 4076 20C8 slo 123 i TT1 125 68 548 200 348 2849 0.13 3990 0.54 -
JS 114 TTIO 150 73 588 203 385 2849 0.15 3975 0.54 - -
96 114
. TTit 200 78 628 241 387 2727 0.14 3871 0.62 . . 91 110
! TTS- 225 97 781 334 448 2729 0.14 4149 0.79 . . 91 114 TTIS 260 85 684 228 456 2640 0.14 3875 0.61 . . 88 208
- TT3 .. .. .. .. .. .. .. .. .. .. .. .. ..
l
- Fully ductile fracture; no arrest load.
- Specimen has been misplaced. However, this has no techr.ical significance on the validity of the data presented ii this report, since 15 specimens were tested. When the specimen is located, it will be stored in the Hot Cell.
1
. y e g 8 *
- _ . - -__-___--_-_--__--__----_---_.--------a
. '.* . . , +
TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR THE COMANCHE PEAK UNIT I WELD METAL AND HAZ METAL, IRRADIATED AT 550*F, FLUENCE 3.70 x 10 18 n/cm2 (E > 1.0 MeV)
Normalized Enerzies Test Charpy Charpy Maximum -Prop Tield Time Maximum Time to Fracture Arrest Tield Flow Sample. Temp Energy Ed/A- -Em/A Ep/A Load to Yield Lead Maximum Load Load Stree, Streen Number (9F) Ift-lb) (f t-lb/in 2) (Ibe) (mece) (Ibel (mece) (Ib ) (Ibe) (kel) (keil weld Metal TW3 -135 4 32 14 18 2155 0.09 2255 0.10 2255 45 72 73 TW12 -100 32 2 9 186 71 4492 0.18 5194 0.38 5194 101 149 tel TW2 - 90 29 234 189 45 4246 0.16 5323 0.39 5323 84 141 159 TWS - - 75 43 34S 225 122 4392 0.15 5209 0.44 5070 326 146 159 TW4 - 60 32 258 181 76 4081 0.13 6202 0.38 5202 818 136 154 TW9 - 50 49 395 277 118 4043 0.16 5185 0.54 4732 1485 134 153 TW13 - 35 53 427 289 138 4381 0.14 5296 0.53 6197 1891 146 161 TW7 - 25 SS 443 276 167 4012 0.14 5175 0.52 5070 1739 133 153 TW11 0 88 709 271 437 4409 0.21 5095 0 54 4563 2379 146 158 TW1 35 99 797 206 532 3500 0.13 5029 0.52 3600 2450 126 147-TW15 50 122 982 3.58 724 3611 0.14 4912 0.53 *
- 120 %42 (n TW10 100 122 982 339 643 3S55 0.16 4780 0.e9 *
- 121 140 e TW5 150 115 926 321 605 3430 0.14 4651 0.67 =
- 114 134
(( TW8 200 136 1095 284 811 3503 0.16 4482 0.81 * * '116 132 TW14 250 133 1071 323 748 3378 0.13 4652 0.67 *
- 112 133 HAZ Vetal THS -12E 16 129 88 40 2802 0.10 4692 0.24 4692 73 93 124 TH7 -115 38 306 262 44 3684 0.13 5263 0.52 5263 61 122 149 TH2 -110 57 459 322 13e 4552 0.21 5212 0.61 4900 126 151 162 THis -100 35 282 178 104 4459 0.22 4992 0.38 4000 416 148 152 TH1 - 90 44 354 280 75 4170 0.14 5143 0.54 E143 678 139 155 TH10 - 75 47 378 256 122 3508 0.14 4385 0.53 4935 1603 117 141 TH3 - 60 61 491 281 230 3822 0.15 4857 0.53 3533 1828 127 144 TH5 - 50 38 306 161 125 3908 0.14 4596 0.40 4595 154S 130 141 TH8 - 40 73 588 252 336 3789 0.13 4932 0.54 4479 1760 125 145 TH14 - 25 72 580 283 Sie 3735 0.16 4842 0.55 4352 1139 124 142 TH9 0 107 882 328 534 3865 0.14 4784 0.67 *
- 122 140 TH12 0 137 1103 339 764 3499 0.14 4807 0.69 *
- 116 138 TH4 50 134 1079 240 839 3554 0.13 4585 0.53 *
- 118 135 TH11 100 116 934 270 684 2888 0.11 4521 0.61 *
- 96 123 TH13 ISO 115 936 223 703 2916 0.13 4301 0.54 *
- 97 120
- Fully ductile fractures no arrest load.
a
.. w"'
TABLE 5-5. -
4 II EFFECT OF $50*F IRRADIAT100t.TO 3.70 x IC n/cm2 (E > 1.0 Mef)
OR TE 180TCH TOUGHIIESS PROPERTIES OF TE COMetHE FDK 1811T I REACTUR VESSEL StRVEILLAICE MATERIALS
$ Average 30ft-lbIII '. Average'35 m11III : Average 50 ft-Tb III Averege Eresy III '
Transition . Lateral Expanston . Transit 1on. Absoepticn at Terperature (*F) Temperature (*F) Temperature (*F) Full shear (ft-1b)
Material tMirradiated irradiated AT untreadiated Ir-adiated AT -lmtreadiated -Irradiated Ar unteradtated 'tradiated A(ft-ib)
Plate R1108-2 -20 -10 10 0: 25 25 5 10 5 130 124 - 6(2) 1.
(Longitudinal).
I-
. ee Plate R1108-2 : 0 15 15 15 50 35 55 75 20 78 87 +9 (Transverse)
Weld Metal - 70 -70 0 -40 -40 0 -35 -35 0 125 126 +I III HAZ Metal -110 -110 0 -75 -75 0 -76 -73 3 119 122
- 3(2)
(1) " AVERAGE" is defined as the value read from the curve fitted through the data points of the Charpy tests (Figures 5-1 through 5-4).
l (2) These values reflect scatter in the data and not sessurable changes. Thus, the values util be reported as 130 ft-lb and 78 ft-ib for the plate R1108-2 (longitudtr.a1 and transverse orientations, respectively),125 ft-Ib for the meld setal and 119 ft-ib for the HAZ eetal.
l-i *
- Y a f
f-f
- l 6
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5-13
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i1ill1ill n
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. tr) cA%( 672 460 904 u 666 666 676 dn ei R
n o D i
lt aa) 024 869 238 h g%
on( 675 549 571 To l
222 221 222 E.
t E N
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E V rt V
R e
M oa) f g7 050 500 056 -
U in( 232 320 239 S 0 no 111 111 11 ,.
Ul L
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E E rt V ( u g) 002 130 934 tni R2 ces 373 146 696 Om ark 556 655 666 T c rt(
C / FS A n E
R8 e 1 rn 1 0 uel 394 833 435 1 tel T cre 516 990 374 I X atk 576 654 11 128 N 111 221 7 U 0 Fr S f_
- 7 5 K A 3 e E E r )
L P O ud n 000 075 900 B T tai 681 067 242 .
A E cok T H F aL( 223 322 333 C
- r h 0 F A 5 M 5 a O eh C T tt A a g) 515 059 907 E mni H D iee 394 601 285 T E t rk 888 888 00G 11 T lt(
R AI US O
F D A d S R l E R eh I I it T T gl 226 210 724 R S E L P A
%rk ni ee 378 416 458 665 665 997 O I 2. t (
R R 0S P E I T
E A .
L M t p) 000 550 000 I smF 545 775 445 S ee* 1 5 15 5 N TT( -
E T
er le pb 123 123 1 e3 mm LLL TTT W nW au TTT TTT T AT SN
)
2l 2
-a -)
8n 6e g 0i 0s 1d 1r l 1u 1e a Rt Rv i i s ,
r eg en e tn ta d t ao ar l a lL lT e M P( P( W Y%
ilIll k
r (*D
~
-150 -100 -50 0 50 100 150 200 I I i i j I i I
~
100 o,,or-eLN 2e 91
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f*
Et 60 W
,,t
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l %V**4 ! ! ! t 1 100 2.5
- n n
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\ t e g 60 -
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1.5 7
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- %*-- Wr 10 0 E
20 -
/ <' -
0.5 0
I tM4*1
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o us:mt:ATED e immAtto ern rtute 27 x ':%r' 160 20")
140 R-120 o ' "*- -
160 C
m f 100 5
d 120 e
" Bu -
G 0 ft 60
--- +*-
l,b. -rr 80 5 '
40 -
f
- /,= Iff -
40 l* 20 -
$?
I P"*'e l
- l l l l 0 0 ty, -200 -100 0 100 200 300 400 TEM?ERATURE (*D J. Figure 5-1. Charpy V-Notch Impact Properties for Ccmanche Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 I x-(Longitudinal Orientation) l l
5-15 l
'.j.;su "nr = {
( C)-
-150 -100 -50 0 50 100 150 200 .
j i i I 21 2i I g ; %* g 100 '
8# -
/ ,./
$g/>
s; 50 -
W .
G 40 -
p/ m 20 vY
- *~e ; -
i i i i i 0
100 - - - 2,5 2.0 Cf 80 o ~g
[60 - -
1.5 7,
U ' 40 -
,r Sh[J*g*aft -
1.0 3
- ; '3' i i 1
0 0
o usRAMAirD e im;:ATED SfD fWDCE M x 10 UCPI +
120 1s0 100 -
g u 120 2 80-
"T, o[ k._ 3
~
gg -
80 g __
i- pr.--- 2rr
$ 40 -
y's m CF --
40 ,
2> -
o
- l l ! I I I ! -
3
(
- ws - 2 00 -100 0 100 200 300 400 .
TEMPERATURE ( F1 Figure 5-2. Charpy Y-Hotch Impact Properties for Comanche Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 -
-(Transverse Orientation) 5-16
W
(* C)
-!50 -100 -50 0 50 100 150 200
~
I l l l { l 100
- -- 4 p ,
- -2 9 80 ,oo
= -
. 3
% 60 -
m' W e3b
- v. 40
, ../
20 0
I "^ "2I I I I '
100 2.5
"' 2 h80-g -3 , -
2.0
- 4 g 60 -
d 1.5 m 40 --
_0 0 -
1.0
& . ", e - 'T
}20 0.5
, 0- ! "A ! I I ! !
0 o tomattt
- IRRADMTED CS*f), fdiENC[ 17 m IQI0 n/tn!
^
16J 200 140
- o (*
U 120 v , q 160
- s 100 -- d O l e _
00 D 80 - 5 Ei s 60 - o ol --
gg
- = D. ~ Q.
g.(
40 -
'70
~*4- 9= fr -
49 20 -
,4 l 8 I l l- l l l 0 0 vtts --200 -100 0 100 200 300 400 TEMPERATURE ( F)
F'Oure 5-3. Charpy V-Notch Impact Properties for Comanche Peak Unit 1 Reactor Vessel Surveillance Weld Metal 5-17
r L-s!
(*C)
-150 -100 -50 0 50 100 150 200 l l l l l l l I 100
- . -,e - - e-A
~
80 -
o *
$ 60~ -
e -
w .
G 40 -
of f,'e ,
20
- P .
I ~I ' ! ' '
O 100 2.5 g43:p:h_e 2.0
-)80 1.5 7
a 60 o, x& 40 - _, ,,
30
_.#EN
'a E" 20 0.5 0
l 8E! I I I i 1 0
0 LtJRA3lATEC
. IRRADIA1ED 050*T). ILDEC 17 m id8n/tM 160 200 ,
140 , ,
120 f
0 '{ ,
160 y100 - o
/.
4 0 -
120 -
D :
Sb U a : 80 r
@ 60 a
, j*
37
- 80
- c.
- 40 -
.g g. .
- : -: rr -
40 e(
20 O- 0 '
W ' 200
'100 0 100 200 300 400 TEMPERATURE ( D Figure 5-4. Charpy V-Notch Imoact Properties for Comanche Peak Unit 1 Reactor Vessel Weld Heat-Affected-Zone Metal .
5-18
- . ... -. . . . _ - . - . . - . _ . . . _ . . - . - . . - . - - - . . ~ . - . . - - . . - . - _ .
4
- - r . .
, ' a r W% :yg, , ;]{ i;pp p,g;,73 ,
,YL.' '
hhf ?$~k!?*
^ ~
+ _ ,y.; {$$
l
- TL15 TL14 TL11 TL9 TL10 l
M % " -- w sa.
3
~
)- TL2 TL3 TLS TL1 TL13
) f i-i- TL4 TL12 TL7 TL8 TL5 i
^
Figure 5-5. Charpy Impact Specimen Fracture Surfaces of the Comanche Peak j ,
Unit 1 Reactor Vessel Lower Shell Plate R1108-2 i, (Longitudinal Orientation) i i
k 4
5-19
l l
l n9 n4 n5 n2 n7 l
- -- T;,:.. ._ -
~TT8 TT13 TT11 TT12 TTI -
1.
! ke =
TT10 TT14 TTS TT15 TT3 -
l Figure 5-6. Charpy Impact Specimen Fracture Surfaces of the Comanche Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 (Transverse Orientation) .
5-20 l
l
~
I-i .. ._._,m., m-m. ,~ - -- -
.k. s 4
- TW3 TW12 TW2 TWO TW4 l....
.'3' v ~
^ '
' ~
q.7 _, ,
(
i- TW9 TW13 Th7 TW11 TW1
~. _ . .
-7 :- . ;)
- j. .
i
> +
.J v 1
?
9 I
- lW ;
y;d[
TW15 TW10 TW5 TW8 TW14 l
Figure 5-7. Charpy Impact Specimen Fracture Surfaces of the Comanche Peak
( Unit 1 Reactor Vessel Surveillance Weld Metal i-i 5-21
. .. - ~. . . - .. - . . . - . . - - . . _- . - . - - . - _ - . . - . . - .
.kf \f M ^WY# -
m; e g4 e, n.,=fp.
N-
.' , l} .
4-- < .
'e>
h L, ;, . g .
l
! g ,
I * , ! 4.F ,
~ '
. ~ . . _
i i
TBS TH7 TH2 THIS TH1 lj[ v .
.g
' ;- 4;
~y
- j.4 3
- :.
1
]-
~
TH10 TH3 TH5 TH8 TH14 ,
l
~
sw/ '
i s
E.;gy , .
4-
.}-
[ s, TH0 TH12 TH4 TH11 THIS ,
i- .
l Figure 5-8. Charpy Impact Specimen Fracture Surfaces of the Comanche Peak
- Unit 1 Reactor Vessel Weld Heat-Affected-Zone Metal .
l 5-22 l
i 2_______________. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ , _ . . . __
n _
('C) 0- .50 100 153 200 250 300 f l ~~l l l l l- 800 110 700 100
~ A LLUMATE TENSitE STR[NGin -
600 a so g/ %-
x ,.
& g
, ~ ~
500 G 70
60 -
- O% N y -,e -
403 50
- CR W 54EEN l l l l l 303 40-A O LKEFA:1ATED A # IRRADIATED AT 550'r, tw[NC[ 17 x10!In/ce2 80 j
70 'E O n RE34U3 IN "E6 60 -
8 50 -
3 '40 2
$ 30 ,b-- WAt EttNCATCN a ,
20 0-
' 4'
-s 10 tecRx EttmanCN I !
0 0 100 200 3:0 400 500 600 gg7pg TEMPERATURE ( F)
LDNG Figure 5-9. Tensile Properties for Comanche Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 (Longitudinal Orientation) 5-23
r-
- s ,..
('C) 0 50 100 150 200 250 300 120 j j 800 g _; {
l-110 700 100 .
,- 600 G 90 F \ ULTlWA4 T[N5' 1 [ $iR[N@H h80 - AD E b ,
500 G 70 - .
60 -
- 1- 400 e :
50
- an a su
~
l
' l l l l 40 A O m:RRAD:AitD 18 vc,2 17 x10 A S 1RRAtiAr[3 AT 550'r, FLUENCE 80 70
- NIS IN EA D' O
- U O- v -
in maim u
@ 50 -
3 40 _
~
s s , ,/
20- - 1 L - @ "
9 10 l- '
(AlrCR4 ELONCATICh 0
0 -100 200 300 400 500 600 TEMPERATURE (*F)
Tems Figure 5-10. Tensile Properties for Comanche Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 (Transverse Orientation) 5-24 F .I
n, ,
t.>
('C) 0 50 100 150 200 250 120 -
B00 j j j. l l
}
110 t'
100 N,N' 2
ut11mit itstu STRENGTH M,-]
700 c 90
@N 600
.s
. ; 80 -
- , E Q c27. Yttte sTREN0tH -
500
,_., G 70 -
60 -
400 50 40 a o tNRRAWO A 4 {nADiATO Al SWF, FLU [NCE 17 x13NWtm2 80 P
g Rtoutics IN GEA ,,2
" U 70 "-
j 60 -
@ 50 -
3 40 -
-" 30 -
ICTAL ELcNGATCN 2 6 / i 20 10 e c.- n- 'a gg g,p;s l l l l l l 0 -
0 100 200 300 400 500
,,m9x TEMPERATURE (*f)
... GD Figure 5-11. Tensile Properties for Comanche Peak Unit 1 Reactor Vessel
~
Surveillance Weld Metal 5-25
5% 979 : -
5 3 i
? y[91g/ww.p;T3:4.5 t
p .. s. d, . - ...v . _ r
- )$lw$$%o vi s ,.
e pp. W .tguq9ty *- ' ,
- c. . , .
-' $3 . ' *\,h-he.;
r .;; r +;_.+m
' ] * .e g ,
"'m .
g,j,4
- -2 y.;j ,
- _ #w w .
%y i ?.--
. M fW--
.ay %._; - '.L&gy.l _ . '. a < .=>. ,--yi a--- 1 n .
4 .
1_. _ ,
Specimen TL1 150*F 4
l
,m ,.
v~
-, . . c , ~
~ gt gm . ..w: c.]. r-
. 3
- . s ~
- n' ?
- r. _ Y ,l. i ^ $;j jlJQ$R$i0h l#.k i i
M);l ,. : Q Q W MM. yl . cA.:M m ed
%;W W Xf.
' (in5jg S qagdJ? . mow.
M,l^ij, "
I i
[_,'
--[ [^24 *
- y NIN , 33$
s 3 ,',[ , i M4-/ , [ .. , . , , (.[ gf' ,
,if.l'd., ansi i &diiEkha Al652:5.s "D;. bn] l'I?.$ffN:1 7
l Specimen TL2 40*F l _.
f,(..
g-l :. .wn
... . a: ,
e 4
j ,r b3 j L.
Je
'? F j 1'. .e,
+ -
.N . ,%
m 4
- t I
l l
Y .; :n., : u+ . - . . -._.a e Ma , ,; . :
- ku .-
Specimen TL3 550*F ,
-Figure 5-12. Fractured Tensile Specimens from Comanche Peak Unit 1 Reactor Vessel ~ Lower Shell Plate R1108-2 (Longitudinal Orientation) ,
1
'.- l l
l l
l l
5-26 4
i
( 1 i
1 1
1 * ~ ,. %, v .e .
)
- % Q,,7;%.l m4
% w . si
, .4 '
%=.,,.yh._4;f.,;%n$w;y58 18 N ily&,&m o.fQ ._ ,n . V i
- s. ,. t
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, s.
- .3
. i 5%.g$%
1, . " ..gs %g s. ,
.<,-( , 1. ,
r c 4 r ,5 ,
, 'iN k.. .m, , r-
- :3: :. .k i
~
r
...a,. .-
~
- y. h ..%w- a a g v n x , w n %go:*,s. u-mtv .s vav w.m . 5
.g;; _
f .
! Specimen TT1 75'F i .
l .t.g. yl"g{$'%il%g 4w e . . , [ ,
- p ". , . o
- - p 7v ,. mg.m9,,m... a: g.wtgg .c -7 e, t s
e . .>.
h 4 ...
.N S f.3-l.! I,o, 4,'~
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/
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i a A,
- a
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- c. -
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+ > w. t .
Specimen TT3 550'F Figure 5-13. Fractured Tensile Specimens from Comanche Peak Unit 1 Reactor Vessel Lower Shell Plate R1108-2 (Transverse Orientation) i i 5-27 1
-- - --e- . . . , , . - . . - , . - . - , , . - . . . ,-_ -- ,
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Figure 5-14. Fractured Tensile Specimens from Comanche Peak Unit 1 Reactor Vessel Surveillance Weld Metal -
5-28
s' 100.00 90.00-4 ' co.oo-
___ 70.00-
$2 00.00-
, hcc 50.00-
$ 40.00-30.00-20.00- gj 10.00- 150 F o.00 i . . .
o.2o o.30 o.oo 0.1o STRAIN, IN/IN 4 100.00 90.00-80.00-
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h cc 50.00-
& 40.00-30.00-20.00- g
- 10.00- 40 F o.oc . . . . .
0.20 0.30 0.00 0.1 o STRAIN, IN/IN Figure 5-15. Engineering Stress-Strain Curves for Lower Shell Plate R1108-2 Tensile Specimens TL1 and TL2 (Longitudinal Orientation) 5-29
.r p
.s.- .
L s
N 90.00 80.00- - -
- 70.00-6::
-. ^ M1 L 60.00-y
_ 50.00-
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- O.0c . . . . . -
- O.00 --
. -O.10 O.20 0.30 STRAIN,- IN/IN 0-g b:, s
'=_
Figure 5-16, . ; Engineering Stress-Strain' Curve -for Lower Shell P1 ate R1108-2 J
(Tensile Specimen TL3 (Longitudinal Orientation) 30
e
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+ 3 I h ,', /$, .If h , . .
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1100.0c s
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- s. o.00 . . .. .
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- STRAIN, IN/IN
- +
Figure'5-17. Engineering Stress-Strain Ctrves_ for Lower Shell Plate R1108-2
. : Tensile Specimens TTI and TTE (Transverse Orientation) 5-31
g ..g 4.. .r 1
s'. '
m ym; i -
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- 90.00- -
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. STRAIN, IN/IN-1 1
i
. l t
l Figure ~5-18.- Engineering . Stress-Strain Curve for Lower Shell Pf ite R1108-2
- Tensile Specimen =TT3 (Ttansverse Orientation) k 4
5-32 c
y' _ $ i..
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- : STRAIN, IN/IN :
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.;, ~
- 0.00 .
0.30 ree 1 0.00. 0.10- 0.2o STRAIN,1IN/IN 7
- q,
- g m n Figure
- -5-.19J Engineering' Stress-Strain- Curve for Weld-Metal Tensile' Specimens e
1,4 _
LTW11and _;TW2 .
~
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- = , , ._ . . . , . ~, ,, _.
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b i
< Figure,5-20. . Engineering Stress-Strain Curve for Weld Metal Tensile Specimen-
~
- . . TW3 6
s
/
5-34 I'
f s
+ a i ';'
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~ ~ . ~ - - - ' - .,J.,. , . , _ . - - ___ , _ ,
SECTION 6,0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY
- 6,1 Introduqtiqa Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR
. reactor pressure vessel surveillance programs for two reasons, First, in order to interpret the neutron radiation-induced material property changes observed in the-test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed rnust be known.
Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis.
. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertaintier associated with damage trend curves as well as to a more
", accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water Reactor e
6-1
Surveillance Re;ults," recommends reporting displacements per iron atom (dpa)
~
along with fluence (E > 1.0 MeV) to provide a data base for future refsrence.
The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements oer Atom." The application of the dpa .
Lparameter to the assessment of_ embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the -
Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials."
This section provides the results of the neutron dosimetry evaluations
-performed in conjunction with the analysis of test specimens contained in ~
surveillance Capsule U. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history.
c The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided.
6.2 Discrete Ordinates Analysis
~
l
- A plan view of the reactor geometry at the core midplane is shown ia Figure 4-1. Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillance program. The
-capsules are located at azimuthal angles of 58.5, 61.0*, 121.5*, 238.5',
241.0*, and 301.5* relative to the core cardinal axes as shown in Figure 4-1.
l A plan view of a dual surveillance capsule holder attached to the neutron pad ,
L is shown in Figure 6-1. . The stainless steel specimen containers are 1.182 by 1-inch and approximately 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.
i 6-2
_m W
- From a neutron transport standpoint, the surveillance capsule structures are
- significant. They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron
- pad and the reactor vessel. In order to properly determine the neutron environment at the test specimen locations, the capsules themselves must be
- - included in the analytical model.
. In performing the fast neutron exoosure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations wer :
carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {p(E > 1.0 Hev), d(E > 0.1 Hev), and dpa} through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e.,
dpa/p(E > 1.0 MeV), within the pressure vessel geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e.,
the 1/4T, 1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal loc- 'ns on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest
', for each cycle of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-in of plutonium as the burnup of individual fuel assemblies increased.
6-3
.s e
m
?The? absolute cycle specific data froQ thel adjoint evaluations together with relative neutron energy: spectra and_ radial distribution information from the
' Lforward calculation providedithe means to:
- 1. . Evaluate neutron dosimetry obtained from survelliance capsule-locations.- ._!
l
- 2. Extr apolate dosimetry results to key locations at the inner radius , ,
and thr_ough the. thickness of the pressure vessel wall._ .
3.. Enable'a direct comparison of analytical prediction with measurement.
- 4. Establish a. mechanism for projection of pressure vessel exposure as the _ design _of each new fuel cycle evolves.
1 Th'e- forward! transport calculation for:the reactor model sumarized in Figures-.
L4-1 and 6-1 'was carried out- in R. O geometry using' the: Westinghouse version
- of the DOT: 3.5 two-dimensional discrete ordinates code [12] and the SAILOR, Revision 0-. cross-section libraryII33 ~ The SAILOR library is a 47 group
'l
-ENDFB-IV based _ data. set produced specifically for light water reactor-
.. applications. In these analyses anisotropic scattering was treated with a P3 ,
I expansion of the cross-sections and the angular discretization was modeled with
.an-S8 order of angular quadrature.
The reference core power distribution utilized in the forward analysis was
' derived from statistical; studies of-long-_ term operation.of Westinghouse- 4-loop
) plants. Inherent in the development.of this reference core power distribution-
- is' the~ use of an-out-in fuel. management' strategy; i.e., fresh fuel 'on the core periphery
- Furthermore, Lfor the peripheral. fuel ' assemblies, a 20 -
{ uncertainty derived from-the: statistical evaluation of plant to plant and cycle to cycle variations:in peripheral _- power was used. Since it is unlikely that a
Lsinglef reactor. would: have a power distribution at the nominal +2a level for. 'f
- ailarge-number of: fuel cycles, the~ use of this reference distribution is
_ expected to yield conservative results ranging from approximately 10 percent up to a: factor of two depending on the actual fuel management strategy used at the reactor. ,
1 6-4 D
F L All adjoint analyses were also carried'out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library.
- Adjoint source locations were chosen at several azimuthal locations along the w pressure vessel inner radius as well as the geometric center of each y l" surveillance capsule. Again, these calculations were run in R, 0 geometry e to provide neutron source distribution importance functions for t;ie exposure 1 c ptrameter of interest; in this case, p (E > 1.0 MeV). Having the
- ' .importance functions and appropriate core source distributions, the response of
-interest could be calculated as:
R-(r,0)-frf0[E 1(r, 0, E) S (r, 0, E) r dr d0 dE where: R (r, 0) - p (E > 1.0 MeV) at radius r and azimuthal angle 0 I (r, 9 E) - Adjoint importance function at radius, r, azimuthal angle 0, and neutron source energy E.
S'(r, 0, E) - Neutron source strength at core location r, 0 and energy E.
Although the adjoint importance functions used in the analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field,' changes in the relative neutron energy spectrum are of second order [32]. Thus, for a given location the ratio of dpa/p (E > 1.0 MeV) is insensitive to changing core source distributions. In the application of these-adjoint importance functions to the Commanche Peak Unit I reactor,
.therefore, the iron displacement rates (dpa) and the neutron flux- (E > 0.1 MeV) c, were' computed on a cycle specific basis by using dpa/p (E > 1.0 MeV) and
( (E > 0.1 MeV)/p (E > 1.0 MeV) ratios from the forward analysis in conjunction with.the cycla specific p (E > 1.0 MeV) solutions from the individual adjoint evaluations.
e 6-5
p The reactor _ core power distribution used in the plant specific adjoint calculations was based on the cycle averagad fuel hssembly burnup data taken from the fuel cycle design report for the first operating cycle of Comanche Peak Unit llI43 .
Selected results from the neutron transport analyses are provided in Tables 6-1 .
through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and ,
provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall. .
In Table 6-1, the calculated exposure parameters [p (E > 1.0 MeV),
p(E > 0.1 HeV), and dpa) are given at the geometric center of the two surveillance capsule positions for both the previously described design basis core power distribution and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared.
Similar data is given in Table 6-2 for the pressure vessel inner radius.
Again, the three pertinent exposure parameters are listed for both the design ,
basis and the cycle 1 plant specific power distribution. It is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itseif. Axial variations in the neutron exposure were developed directly from the relative axial power distributions associated with the plant specific and design basis core power distributions.
Radial gradient information for neutron flux (E > 1.0 MeV),
neutron flux (E > 0.1 HeV), and iron atom displacement rate is given in Tables ,'
6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron ,
transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure pcrameter distributions within the wall may be obtained by normalizing the calculated or projected '
exposure at the vessel inner radius to the gradient data given in Tables 6-3 through 6-5.
6-6
q t-4 For example, the neutr:n flux (E > 1.0 MeV).at the 1/4T p sition cn the 45' azimuth is given by:
41/4T(45') = ((220.27, 45') F (225.75, 45')
x where:
41/4T(45')
- Projected neutron flux at the 1/4T position on the 45' azimuth
.- p (220.27,45') - Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth.
F (225.75, 45') - Relative radial distribution function from Table 6-3.
Similar expressions apply for exposure parameters in terms of ( (E > 0.1 MeV) and-dpa/sec.
The DOT calculations were carried out for a typical octant of the reactor.
-However, for the neutron pad arrangement in Commanche Peak Unit 1, the pad extent for all octants is not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the ne w n pad extending from 32.5 - 45.0
, degrees which produces the maximum flux. Other octants have neutron pads spanning larger azimuthal sectors which provide more shielding. For the octant with 12.5 degree pad, the maximum flux to the vessel occurs near 25 degrees and the values in the' tables for the 25 degree angle are vessel maximum values.
Exposure values for 0,15, and 45 degrees can be used for all octants; values in-the tables for 25 and 35 degrees are maximum values and only apply to octants with a 12.5 degree neutron pt.d.
'6.3' Neutron Dosimetry
- , .The passive neutron sensors included in the Commanche Peak Unit I surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determiaation of the various exposure parameters of interest [p (E > 1.0
. Hev),.p (E > 0.1 MeV),'dpa].
6-7
The relative locations of the neutron sensors within the capsules are shawn in Figure 4-2. The iron, nickel, copper, and cobalt- ?uminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors were -
accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest. .
Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on the target material -
~
over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.
In particular, the following variables are of interes':
o The specific activity of each monitor, o The operating history of the reactor. .
o The energy response of the monitor, o The neutron energy spectrum at the monitor location.
o The physical characteristics of the monitor.
The specific activity of each of the neutron monitors was determined using established ASTM procedures (15 through 28]. Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation histcry of the Commanche Peak Unit I reactor during cycle 1 was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" im Se applicable period. ,
The irradiation history applicable to Capsule U is given in Table 6-7. ,'
Heasured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code [29). The ,
6-8
, ~
f f
. FERRET:-approach used.the measured reaction rate data and the calculated neutron energylspectrum at the the center of the surveil _ lance; capsule as input and
-' proceeded to adjust a' priori (calculated) group fluxes to produce a best fit .
- - (in a least squares sense) toithe reaction rate data. . The exposure-parameters
~
- along with associated uncertainties where' then obtained from the adjusted spectra.
=
In the FERRET evaluations,- a log-normal least-squares ~ algorithm weights both
. the a: priori values and the measured data in accordance with the assigned a
uncertainties and correlations. In general, the measured values f are linearly related to the flux'd by some response matrix A:
-( s ,a) - (s) (a) f =12 A 4
- g. .ig g where 'i indexes the measured values belonging to a single data set s, g
- designates the energy group and a delineates spectra that may be simultaneously- adjusted. For example, R =2 .o (
i~ .g -ig g relates a set of measured reaction rates Rj to a single spectrum pg by the multigroup' cross section ajg. (In this case, FERRET ~ also adjusts the cross-sections.) The log-normal approach automatically accounts 'or the
. physical: constraint of positive'. fluxes, even with the large assigned
- uncertainties.-
'= In the FERRET-analysis of the dosimetry data, the continuous quantities (i.e.,
"= fluxes and cross-sections) were approximated in 53 groups. The calculated
-fluxes from the-discrete ordinates analysis were expar. dad into the FERRET group structure using the SAND-II code p0). -This procedure was carried out by
-first expanding the a priori spectrum into the-SAND-II 679 group structure using a SPLINE: interpolation procedure for interpolati' 1 regions where group
> boundaries'do'not coincide.. The 620-point spectrum we i casily collapsed 6-9
10 to the grcup scheme used in FERRET.
The cross-section were also collapsed into the 53 energy-group structure using
' SAND !! with calculated spectra (as expanded to 620 groups) as weighting
. functions. The cross sections were taken from the ENDf/B-V dosimetry file. .
Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data -
and code limitations, but are expected to be unimportant.
For each set of data or_ a priori values, the inverse of the corresponding
- q relative covariance matrix M is used as a statistical weight. In some cases, .
l as for the cross sections, a multigroup covariance matrix is used. More often, l l
a simple' parameterized forin is used:
Mgg , = R[ + R g R,P g gg, where RN specifies an overall fractional normalization uncertainty (i.e.,
complete correlation) for the correspond' 1 set of values. The fractional uncertainties Rg specify additional random uncertainties for group g that are ,
correlated.with a correlation matrix:
Pgg,.= (1 - 0) 69g, + 0 exp_( )
The first' term specifies purely random uncertainties while the second term describes short-range correlations over a range 7 (0 specifies the j strength of the latter tenn). .
1 For.the a priori calculated fluxes, a short-range correlation of y - 6 . i groups was used. This choice implies that neighboring groups are strongly ,
correlated when.0 is close to 1 Strong long-range correlations (or anticorrelations) were justified cased on information presented by R.E.
Maerkerl31). Meerker's results are closely duplicated when y = 6. For
- the integral reaction rate covariances,' simple norr.alization and random ;
-uncertainties were combined as deduced from experimental uncertainties. ,
6-10
, : , .u ; - _ _ _ _ . _ _ __ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
Results of the FERRET eaaluation of the Capsule U dosimetry are givon in Table 6-9. The data sumarized in Table 6-9 indicated that the capsule received an integrated exposure of 3.70 x 10 18 n/cm2 (E > 1.0 HeV) with an associated
.' la uncertainty of i 8%. Also reported are capsule exposures in tt us of fluence (E > 0.1 HeV) and iron stom displacements (dpa). Summaries of the fit of the adjusted spectrum are provided in Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrue to the individual
- . experimental reaction rates. The adjusted spectrum itself is tabulated in 1able 6-11 for the FEP"di 53 energy group structure.
A sumary of the measured and calculated neutron exposure of Capsule V is presented in Table 6-12. The agreement between calculation and measurement falls within i 18% for all f ast neutron exposure parameters listed. The thermal neutron exposure calculated for the exposure period undepredicted the measured value by approximately a factor of two. ,
Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (0.91 EFPY) exposure derived from the Capsule U measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EFPY).
In the evaluation of the future exposure of the reactor pressure vessel the design basis exposure rates from Table 6-2 were employed. Since the Comanche Peak Unit I reactor has operated for only one fuel cycle and equilibrium fuel management has not been fully established, the use of these design basis values is still appropriate. The use of the design basis values should result in conservative predictions of future vessel exposure that can be refined ac additional doshetry becomes available.
', in the calculation of exposure gradients for use in the development of heatup and cooldown curves for the' Comanche Peak Unit I reactor coolant system, exposure projections to 16 EFpY and 32 EFPY were also employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wal are provided in Table 6-14. In order to access RTNDT vs. fluence 6-11
.g trend curves, dpa equivalent fast n:utron fluence lovels for the 1/4T and 3/4T 1 positions were defined by the relations d'L(1/4T) -i (Surface) {d a (S r e ,
, -('(3/4T) -
p (Surface) {d ((u e}} l Using this approach results in the dpa equivalent fluence values listed in ,
-Table 6-14; ,
p In Table 6-15 updated lead factors are listed for each of the Comanche Peak H Unit 1 surveillance capsules.- These data may be used as a guide in I
- establishing future. withdrawal schedules for the remaining capsules.
ke ;
+
t
.. 1 p
i
.k 1
e 4-
; 3.:- g
[) , a
+
6-12 N
.-c . . . _ _ . . . _ ~ 4. _ ....2. . . . . . _ _ , -
_._..__,.--,____..__.._._,,..._..._._...ff
i i U
= 58.58 = G1.08
\
r - $1.825 IN.
; e - ; b E Y Y fP k nM S %
~
%\ i I
,e ,
Figure 6-1. Plan View of a Dtial Reactor Vessel Surveillance Capsule 6-13
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER ((E > 1.0MeV) ((E > 0.1Mev) Iron Displacement Rate In/cm2 -seci In/cm2-sec1 - Idoa/seci 29.0" 31.5* 29.0* 31.5* 29.0* 31.5* DESIGN BASIS 1.13 X 10 Il 1.21 X 10 Il 5.07 X 10Il S.44 X 10 1I 2.21 X 10-10 2.37 X 10-10 CYCLE 1 10 1.06 X 10 ll 4.37 X 10 ll 4.77 X 10 Il 1.91 X 10-10 2.24 X 10-10
.. 9.75 X 10 ?
2 4 g g . O . g . .
TABLE 6-2 1
~
.Y ; CALCL' LATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AT i THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE
-+
d.[E > 1.0MeV) In/cm2 1gpl '
. .0.0' lid' . Zid' 35.0* 4 5. 0 *.
-DESIGN' BASIS' l.78 X 1010 2.66 X 1010 3.01 X 1010 2.45 X 1010 2.81 X 1010 CYCLE 11- 1.49 X 1010 2.26 X 10 10 2.62 X 1010 2.24 X 1010 2.63 X 1010
-(IE > 0.1MeV) In/cm2 _strl
,_Q.A' lid' 25. 0'. 35.0' .41d*
DESIGN BASIS 3.70 X 1010 5.60 X-1010 8.22 X 1010 6.96 X 1010 7.04 X 1010
-3.10--X 10 10.4.76 X 1010 7.15 X 1010 6.36 X 1010 6.59 X 1010 CYCLE: 1-
_lron Atom Disolacement Rate idoa/sec1
.0.0* 15. 0'. Zia' .
3 5 . 0 '. 45.0' l DESIGN BASIS: - 2.77 X 10-11 ' 4'.12 X 10-II ' 5.04 X 10'lI' 4.15 ' X 10-11 4.48 X 10-11
- y CYCLE' l' 2.32LX 10-11 3.50.X 10-11 4.39 X 10-11 3.79 X 10-II'4.19 X-10-Il
-9: , .4 9-7 N _:
b
..) # '
6-15
N[il , , TABLE 6-3 M RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX-(E > 1.0 MeV) ; E WITHIN THE PRESSURE VESSEL WALL r Radius H igg} _ O' 15* 25' 35' 45' , 220.27(I) 1.00 1.00 1.00 1.00 1.00 0.980 0.977 0.979 .' 0.976 0.979 : 220.64
- 221.66 0.888 0.891 0.893 0.891 0.889 ,
b 222.99 0.768 0.770 0.772 0.770 0.766 - 224.31 0.653 0.653 0.657 0.655 0.648 I - 225.63= 0.551 0.550 0.554 0.552 0.543
- 226.95- ~0.462 0.460 0.465 0.A63 0.452- t
- 228.28 -0.386 0.384 0.388 0.386 0.375 .
229.60 0.321 0.319 0.324 0.321 0.311 230.92 0.267 0.263 0.275 0.267 0.257 232.25- 0.221 0.219 0.225 0.221 0.211- , p
= 233'.57 0.183- 0.181 0.185 0.183 0.174 234.89- 0.151 0.149 0.153 0.151 0.142
.- 236.22 0.124_. 0.122 0.126 0.124 0.116
. 237.54 0.102 0.100 0.104 0.102 -0.0945 i 238.86 0.0828: 0.0817. 0.0846 0,0835 0.0762 240.19- 0.0671 0.0660- 0.0689 0.0679- - 0.0608
- .. .241.51L 0.0538 0.0522 0.0550 0.0545' O.0471
. 242.17(2)' O.0506 0.0488 0.0518 0.0521 0.0438 NOTES: 1) Base Metal. Inner Radius l2): Base 6.atal Outer Radius "t.
f 6-16 4 e r --e c-N-'- e,ee- s- s e- --E
~t TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 HeV)
WITHIN THE PRESSURE VESSEL WALL I Radius
. Ism)_ O' _11' _, 25' 35' . -- 45' 220.27(I) 1.00 1.00 1.00 1.00 1.00
. 220.64 1.00 1.00 1.00 1.00 1.00
. 221.66 1.00 1.00 1.00 0.999 0.995 222.99 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901 225.63 0.874 0.865 0.874 0.850 0.842 226.95 0.818 0.808 0.818 0.792 0.782 228.28 0.761 0.750 0.716 0.734 0.721 229.60 0.705 0.693 0.704 0.677 0.662
, 230.92 0.649 0.637 0.649 0.621 0.605 232.25 0.594 0.582 0.594 0.567 0.549 < 233.57 0.540 0.529 0.542 0.515 0.495 234.89 0.487 0.478 0.490 0.465 0.443 236.22 0.436 0.428 0.440 0.413 0.392 237.54- 0.386 0.380 0.392 J.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201 242.17(2) 0.233 0.226 0.237 0.223 0.188
'~
NOTES: 1) Base Metal Inner Radius
- 2) Base Metal- Outer Radius O
6-17 i
TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL P.adius igg}_ O' 15' 25' 35' 45' , 220.27(l) 1.00 1.00 1.00 1.00 1.00 220.64 0.984 0.981 0.984 0.983 0.984 .' 221.66 0.912 0.909 0.917 0.921 0.915 , 222.99 0.815 0.812 0.826 0.833 0.821 - 224.31 0.722 0.719 0.737 0.747 0.730 225.63 0.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572 228.28 0.497 0.493 0.519 0.533 0.506 229.60 0.439 0.435 0.462 0.475 0.447 230.92 0.387 0.383 0.410 0.423 0.394 232.25 0.341 0.338 0.364 0.376 0.347 233.57 0.300- 0.297 0.322 0.334 0.305 234.89 0.263 0.261 0.285 0.295 0.266 236.22 0.230 0.228 0.250 0.260 0.231 237.54 0.199 0.198 0.218 0.227 0.199 238.86 0.171 0.170 0.189 0.196 0.169 240.19. 0.145 0.144 0.161 0.167 0.140 241.51 0.121 0.119 0.135 0.139 0.113 242.17(2) 0.116 0.113 0.128 0.134 0.106 NOTES: 1) Base Metal Inner Radius
- 2)-Base Metal Outer Radius er 6-18
. . e . . , ,
TABLE 5-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Target Fission Reaction Weight Response Product Yleid Monitor of Fraction Rance Half-Life (M Material Interest l Cu63(n,a)Co60 0.6917 E > 4.7 MeV 5.272 yrs Copper l Fe54(n,p)Mn54 0.0582 E > 1.0 MeV 312.2 days Iron N158(n.p)CoS8 0.6830 E > 1.0 MeV 70.90 days Nickel
- a 30.12 yrs 5.99 U238(n,f)Cs137 1.0 E > 0.4 MeV h Uranium-238*
1.0 E > 0.08 MeV 30.12 yrs 6.50 Neptuntum-237* Np237(n,f)Cs137 CoS9(n,7)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobalt-Aluminum
- CoS9(n,1)Co60 0.0015 E > 0.015 MeV 5.272 yrs Cobalt-Aluminum
- Denotes that monitor is cadmium shielded.
4 l J TABLE 6-7 l o~ MONTHLY THERMAL GENERATION DURING THE FIRST FUEL CYCLE .
' 0F THE COMANCHE PEAK UNIT 1 REACTOR ,
. l THERMAL , ,
GENERATION- ;
, MONTH (MW-hr) .
l 4/90 127135 .! 5/90 748810
;6/90 518035 7/90: 1688199'
-8/90_ -1318010 i 9/90- 1480101 ,
10/90 1911524-11/90 1073237 [ 12/90 2374056 1/91 2259446
-2/91 2054786 ,
^
3/91 1375315 4/91 0 5/91 261965 6/91 2439547 +
.7/91 2398615 8/91 ;2521411 9/91 2431361
- 10/91-196474-
' I e +
I e. 6 . p -
- - . -w+, ,,;.., .. , . , , . . , - , . . - , - ,, , , . . . . . ,
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES . Measured Saturated Reaction Monitor and Activity Activity Rate Axial Lottthil (dis /sec-am) (dis / set-ami If25/_tiUCl[US).
. Cu-63 (n a) Co-60 a
Top 4.75 x 104 4.64 x 105 Middle 4.43 x 104 4.32 x 105 Bottom 4.31 x 104 4.21 x 10 5 Average 4.50 x 104 4.39 x 105 6.69 x 10'I7 Fe-34(n.p) Mn-54 Top 1.24 x 10 6 4.15 x 10 6 Middle 1.30 x 10 6 4.35 x 10 6
, Bottom 1.21 x 106 4.05 x 10 6 Average 1.25 x 10 6 4.18 x 10 6 6.66 x 10-15 Ni-58 (n,p) C0-58 Top 7.80 x 10 6 6.84 x 10 7 Middle 7.31 x 10 6 6.41 x 107 Bottom 7.30 x 10 6 6.40 x 107 Average 7.47 x 10 6 6.55 x 10 7 9.35 x 10-15 U-238 (n,f) Cs-137 (Cd)
Middle 1.31 x 10 5 6.42 x 106 4.25 x 10-14 6-21
l i TABLE 6-8 ; s
.m. . MEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd :
Measured Saturated Reaction .
; Monitor'and' Activity Activity Rat. f Axial location- (dis /sec-am) (dis /sec-cm) (RPS/NUCLEU$) . !
. - Np-237(n,f)-Cs-137(Cd)_ .-
' I Middle 1.28 x 106 6.30 x 107 3.81 x 10*I3 f L
Co-59--(n,1) Co-60 r 7
. Top 9.47 x.106 9.24 x 10 Top 8.18 x 106 7.98 x 107 Middle '9 40'x 106 9,37 x 3o7 !
= Middle ~ 7.97 x 106 7.78 x 107 '
E -Bottom 9.20 x.306 8.98 x 10 7-
' Average 8.84 x 106 8.63 x 107 5.63 x 10-12 , ;
.00-59 (n,1)'Co-60-('Cd) i f
. Top' . 4.85 x 106 4.73 x 107 4.81 x 10 7 Middle.- 4.93 x 106
' Average 2
- 4.89 x 106 - 4.77 x 107 3.11 x 10*I2
- Y a
hf 6-22 , 3 L e I
^
4 e.! y.n -e c.+ -r.., . ...p- ,rr- y ei.w .ne w - %n ,..p
,,m.,y,9- vc.mn.g,v,,,39,--,u.,,. v.,%..y_,..%.,, 9
-j.
u ,
, TABLE 6-9 !
t , { p w 1' 1 $UMARY-Of NEUTRON DOSIMETRY _ RESULTS 1 TIME AVERAGED EXPOSURE RATES : i 2 f (E > 1.0 MeV) (n/cm -sec). 1.29 x 10ll i 8% {
)
((E>0.1-MeV)(n/cm.g,c). 2 5.58 x 10ll i 15% i dpa/sec 2.44 x-10-10 3 33g i 2 [ f (E- < 0.414 eV) (n/cm -sec) 1.05 x 101I' i 22% ; l' j INTEGRATED CAPSULE EXPOSURE 3.70 x 10 18 2 ! t (E > 1.0 MeV)- {n/cm ) i 8% r 2 [::, :4.(E->-0.1 MeV) (n/cm ) 1.60 x 1019 - i 15%
..- ,dpa' 7.00 x 10-3 1-11%
2- 3.02 x 1018 - i 22% l t (EL<.0.414 eV)L(n/cm )
! i l- .
U NOTE:, ~ Total -Irradiation Tige - 0.91 EFPY ,e _,4-
- e '
'4 9
23-e,-.s ywr,- . y v 7,em.,-'-,. - - - y y-vm-, awe - revw -,-r74 e = 'r- -
'v- y w
TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER . Adjusted i- Reaction Measured Calculation C/M , Cu-63-(n,a) Co-60 6.69x10-17 6.80x10-II 1.02 -
^
! Fe-54 (n,p) Mn-54 6.66x10-15 6.69x10-15 1.00 Ni-58 (n.p).Co-58 9.35x10-15 9.29x10-15 o,gg U-238 (n,f) Cs-137 (Cd) 4.25x10-14 3.95x10~I4 0.93 Np-237 (n,f) Cs-137 (Cd) 3.81x10-I3 3.95x10~I3 1.04 Co-59 (n,1)-Co-60 (Cd) 5.63x10-12 5.59x10-12 o,gg Co-59 (n,1) Co-60 3.llx10-12 3.13x10-12 3,og I 4 v . 1 s e 6-24 z t f
TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SVRVEILLANCE CAPSVLE CENTER Energy AdjusgedFlux Energy AdjusjedFlux Group (Hav) (n/cm -sec) Group (Mev) (n/cm sec)
. I 1.73x10I 9.53x10 6 28 9.12x10-3 2.46x1010 2 1.49x101 2.15x107 29 5.53x10-3 3.18x1010
- 3 1.35x101 8.30x107 30 3.36x10-3 9.93x109 4 1.16x101 1.86x108 31 2.84x10-3 9.47x109 5 1.00x101 4.10x108 32 2.40x10-3 9.10x109 A .6' 8.61x10 0 7.03x108 33 2.04x10"3 2.56x1010 7 7.41x100 1.62x109 34 1.23x10-3 2.35x10 10 8 6.07x10 0 -2.33x10 9 35 7.49x10-4 2.17x1010 9 4.97x10 0 4.97x109 36 4.54x10-4 2.05x1010 10 3.68x10 0 6.65x109 37 2.75x10-4 2.20x10 10 11 2.87x10 0 1.41x10 10 38 1.67x10-4 2.32x1010 12 2.23x10 0 3,97xio10 39 1.01x10-4 2.37x10 10 13 1.74x10 0 2.78x10l0 40 6.14x10-5 2.36x10 10 14- 1.35x10 0 3,ogxto10 41 3.73x10-5 2.31/J010 15 1.llx10 0 5.65x1010 42 2.26x10-5 2.25x1010 16 8.21x10-1 -6.42x1010 43 1.37x10-5 2.20x1010 17 6.39x10-1 6.63x1010- 44 8.32x10-6 2.10x1010 18 4.98x10-I 4.78x1010 45 5.04x10-6 1.93x1010 19 3.88x10-1 6.67x1010 46 3.06x10-6. 1.80x1010 20 3.02x10-1 6.82x1010 47 1.86x10-6 1.66x1010 21' 1.83x10-l 6.71x1010 48 1.13x10-6 1.23x1010 10 6.83x10-7 1.53x1010 22 1.11x10-1 5.34x10 49
', 23 6.74x10-2 3.70x1010 50 4.14x10-7 1.99x1010 24 4.09x10-2 2.09x1010 51 2.51x10-7 1.92x1010 25- 2.55x10-2 2.74x1010 -52 1.52x10-7 1.78x1010 26 1.99x10-2 1.35x1010 - 53 9.24x10-8 4.86x1010 27 1.50x10-2 1.71x1010 NOTE: Tabulated energy levels represent the upper energy of each group.
6-25
.,c ,,. - - - _
TABLE 6-12 COMPARIS0N OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE U . e Calculated Measured CM 2 3,g3 x 3g 18 3.70 x 1018 0.82 - 4(E > 1.0 MeV) (9/cm ) 2 1.60 x 1019 0.85 f(E > 0.1 MeV) (n/cm ) 1.36 x 1019 dpa 5.94 x 10'3 7.00 x 10-3 0.8S 2 3.02 x 1018 0.48 f(E < 0.414 eV) (n/cm ) 1.44 x 1018 9 v 6-26
*N W-" " - - - - , _ , , , . - _ , ,
an, TABLE 6-13 NEUTRON EX NiURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE _0.91 EFPY 0* 15* 25' 35* 44*
* (E > 1.0 Mev) 5.23 X 10 17 7.91 X 10 I7 9.17 X 10 17 7.83 X 10 I7 9.2iX1017
-In/cm2) t (E
- 3.1 MeV) 1.04 X 10 18 1.61 X 10 18 2.41 X 10 18 2.14 X 10 18 2.22 X 10 18 bi/cm2]
1ron Atom Displacements 7.84 X 10-4 1.18 X 10-3 1.48 X 10-3 1.28 X 10-3 1.41 X 10-3 [dpa) 16.0 ETPY 0* 15* 25* 35* 45*
* (E > 1.0 Mev) 9.00 X 10 1C 1.35 X 10 I9 1.52 X 10 l9 1.24 X 10 19 1.43 X 10 19 '[ [n/cm2]
t (E > 0.1 MeV) 1.87710 13 2.83 X 10l9 '4.15 X 10I9 3.53'X 10I9 3.57 X 10 I9 , [n/cm2) Iron Atom Displacements 1.40 X 10-2 2.08 X 10-2 2.55 X 10-2 2.10 X 10-2 2.27 X 10-2 [dpal 32.0 EFPY 0* 15* 25' 35' 45*
* (E > 1.0 Mev) 1.80 X 10 I9 2.69 I 10 I9 3.04 X 10 19 2.48 X 10 I9 2.85 X 10 I9
[n/cm2]
+ (E > 0.1 MeV) 3.73 X 10I9 5.65 X 10 19 8.30 X 10 l9 7.04 X 10 19 7.13 X 10 I9
[n/cm2) 1ron Atom Displacements 2.80 X 10-2 4.16 X 10-2 5.09 X 10-2 4.20 X 10-2 4.54 x 10-2 [dpa] 1
~ - - -
, a; , _
~ '
- y. ,
TABLE 6-13 (continued) NEUTRON EXPOSURE MIOJECTIONS AT' KEY LOCATIONS-
~ON-THE-PRESSURE VESSEL CLAD /8ASE METAL' INTERFACE; T
48.0 EFFY 0*' 15 * : 25* 35*- 45* J
'(E.> 1.0 Mev) 2.70 X 10 18 4.04 X 10 18 4.56 I 10 19 3.72 X 10 18 ' 4.28 X lol '
.[n/cm2]
4 -(E > 0.1 MeV) 5.60'X 1019 - 8.48 X'10I9 1.25'X 1020 1.06 X 10 20 .1.07 X 1020 [n/cm2] l- , LIron Aton Displacements: 4.20 X.10-2 6.24 X 10-2 7.64 X 10-2 6.30 X 10-2.- "6.81 X 10-2 [ -[dpa) . 4
- e, -
F e 1 p l J-s 1 I 11 , .. . ., . ., .;
TABLE 6-14 NEUTRON EXPOSURE VALUES AT 1/4T AND 3/4T LOCATIONS'FOR 16 AND 32 EFPY 16 EFPY
' NEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 ;(equivalerit n/cs8)
-(n/cm )
Surface 1/4 T 3/4 T Surface- 1/4 T' 3/4 T 0* 9.00 x 10 18- '4.88 x 1018 :1.04 x 10I8 9.00 x 10 I0 ~ 5.68 x ID I8 1.97 x 10 I8 4 15* ~1.35 x 10l9- 7.29'x 10I8 1.53 x'1018 1.35 x 10 I9 8.43 x 10 18 2.92 x 10 18 25 (a) 1.52 X 10 I9 '8.32 X 1018 1.80 X 10 18 1.52 X 10 19 9.89 X 10 18 3.63 X 10 18 35* 1.24'x 10 19 6.77 x 1018 .1.44 x 10 18 1.24 x 10 l9 8.24 x 10 l8 3.09 x 1018
'45* 1.43 x-10.I9- 7.65 x 1018 1.54 x 10 18 1.43 x 10 l9 9.15 x 10 I8 3.13 x 1018
^ m h' 32 EFPY ffEUTRON FLUENCE (E > 1.0 MeV) SLOPE doa SLOPE 2 (n/cm ) (equivaient n/cm2 ) 4 Surface 'l/4 T 3/4 T Surface I!4 T 3/4 T 9.77 x 10 18 2.09 x 10 18 1.13 x 10 I9 0* 1.80'x 1019 1.80 x 10 19 3.94 x 10 18 15* 2.69 x 10 I9' 1.46 x 10l9 3.06 x 10 18 2.69 x 10I9 1.69 x 10 l9 5.83 x 10 18 25 (a) 3.04 X 10l9 1.66 X 10 l9 3.59 X 10 18 3.04 X 10 19 1.98 X 10 19 7.25 X 10 18 35* 2.47 x 1019 1.35 x 10 19 2.87 x 1018 2.48 x 10 19 1.64 x 10 19 6.15 x 10 18 45* 2.84 x 10 I9 1.52 x 1019 3.07 x 10 18 2.85 x 10 I9 1.82 x lb l9 6.24 x 10 18 (a) Maximum point on the pressure vessel i
TABLE 6-15 UPDATED LEAD FACTORS FOR COMMANCHE PEAK UNIT 1 SURVEILLANCE CAPSULES Caosole Lead Factor . U 4.03(a) , Y 3.75(b) . V 3.75(b) , W 4.02(b) X 4.02(b) Z 4.02(b) (a) Plant specific evaluation based on end of cycle 1 calculated fluence. (b) Projection based rn design basis flux.
=
m 9 9 4 6-30 m - - . _ _ - - _ _ _
SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets AS1H E185 82[6] and is recommended for a- future capsules to be removed from the Comanche Peak Unit I reactor vessel: t-F Capsule Estimated [ V Location Lead fluence 2 Capsulo (deg.) Factor Removal Time (b) (n/cm) U 58.5 4.03 0.91 (Removed)(a) 3.70 x 1018 (Actual) Y 241 3.75 8.50 3.03 X 1019 (c) V 61- 3.75- 12.5 4.45 X 10 19 X 238.5 4.02 Standby --- W 121.5 4.02 Standby --- Z -301.5 4.02 -Standby --- I (a) Plant Specific Evaluation (b) Effective full Power Years (EFPY) from plant startup.
'(c) Approximate 32 EfPY fluence at the reactor vessel inner wall location. , , s.
_g.
?
7-1 J
' ~ ' - -
rt y,.- ] y .g. x r F 4 5 e- ' v'~ b' b I;.. 3 N. e' l hi ' f-(v~ -
\
E t i ( , O .- i:
-h i
s I e
- fi h
4 I 6 I b- ' j:- i I
- h. ' '
f I
.)
L 5 e .- j i 5
.6 l.- e t.
[. I ' e i f l >
-4'T f.. e - 6 a
- :I e
o L ! a 4-
.l
(':1 Ia 3 e.. z, h' r-- u t. . _ . . . . .- w_,--._
.m. _,,;l,g.,, e. . . ..m.'* v ~c-' ~
"**'**~~ '^ ' ' " ~ ~ ' ' ~ ^ '~
l SECTION
8.0 REFERENCES
i '
- 1. W.T. Kaiser et al., " Texas Utilities Comanche Peak Unit No. 1 Reactor c.. Vessel Radiation Surveillance Program," WCAP-9475, April 1979.
, 2. Code of federal Regulations,10CFR50, Appendix G. " Fracture Toughness j Requirements", and Appendix H, " Reactor Vessel Material Surveillance
. Program Requirements," U.S. Nuclear Regulatory Commission, Washington.
0.C.
- 3. . Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May 1988.
- 4. Section 111 of the ASME Boiler and Pressure Vessel Code, Appendix G.
" Protection Against Honductile Failure."
- 5. AS1H E208, " Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of ferritic Steels."
- 6. AS1M E185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)."
- 7. ASTM E23-88, " Standard Test Methods for Notched Bar Impact Testing of Metallic Materials."
- 8. ASTM A370-89, " Standard Test Methods and Definitions 'for Mechanical Testing of' Steel Products."
!e
- 9. ASTM E8-89b, " Standard Test Methods of . Tension Testing of Metallic Materials."
- 10. ASTM E21-79 (1988), " Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."
8-1
l
. 'll. ASTM E83-85, " Standard Practice for Verification and Classification of l
-Extensometers." l
- 12. R. G. Soltesz, R. K, Disney, J. Jedruch, and S. L. Ziegler, " Nuclear j 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. ,
- 13. "0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 .
# Neutron,:20 Gamma-Ray, P3, Cross Section Library for Light Water l Reactors".
;14. B. W. Schmidt, et al., "The Nuclear Design. and Core Physics
- Characteristics of the Commanche Peak Unit 1 Nuclear Power Plant - Cycle 1", WCAP-980' 'tev. 3, April 1990. (Proprietary)
- 15. ASTM Designation E482-89, " Standard Guide for Application of-Neutron lansport Methods.for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials,- Philadelphia, PA, 1994.- .
~
16~. ASTM Designation E560-84, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section
-12,' American Society for . Testing and Materials, Philadelphia, PA,1991.
- 17. . ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in~Ferritic Steels in Terms:of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia,-PAL 1991.- ,
-18=. -- ASTM Designation .E706-87, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, American : Society for Testing, and Materials, Philadelphia, PA,1991.
7 I g t 8-2 c
- s. + .-r< L < %,, w. r - .. ,. m Ew , y w q -.e.v -
9 a p 9 -1,p,
_ _ . . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ ~ _ .
- 19. AS1H Designation E853-87, " Standard Practice for Analysis and ,
i Interpretation of Light-Water Reactor Surveillance Results', in ASTM
. Standards Section 12, American Society for Testing and Haterials, Philadelphia, PA, 1991.
- 20. ASTM Designation E261-90, " Standard .iethod for Determining Heutron Flux, .
,_ Fluence, and Spectra by 'tadioactivation Techniques", in ASTM Standards,
-' Section 12, American Ss ety for Testing and Materials, Philadelphia, PA,
. 1991.
21, ASTM Designation E262-86, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in AS1H Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
- 22. ASTM Designation [263-88, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991.
~
- 23. ASTM Designation E264-87, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA,1991.
- 24. ASTM Designation E481-86, " Standard Method for Measuring Neutron-Flux ,
Density by Radioactivation of Cc'a alt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1991.
*~
- 25. ASTM Designation E523-87, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for festing and Materials, Philadelphia, PA,1991.
- 26. ' ASTM Designation E704-90, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American
. Soc.ieti for Testing and Materials, Philadelphia, PA,1991.
8-3
- 27. ASTM Designation E705-90 " Standard Method for Measuring f ast-Neutron flux Density by Radioactivation of Ne,tunium-237", in ASTM Standards, Section 12, American Society for Ter. ling and Materials, Philadelphia, PA,1991. ,
- 23. ASTM Designation Elc;S-84, " Standard Method for Application and Analysis ,
of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1991. .'
- 29. f. A. Schmittroth, ELEREI_ Data Analysis Core HEC fME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
- 30. W. N. McF.1roy, S. Berg and T. Crocket, a_(qmouter-Automated Iterativc Met _ hod of Neytron Flux SpectriLJh.termined by foil Activa 1120, AFL'L-TR-7-41, Vol . I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 31. EPRI-NP-2188, " Development and Demonstration of an Advanced Methu, logy
~
for LWR Dosimetry Applications", R. E. Maerker, et al., 1981.
- 32. R. E. Maerker, et al., " Accounting for Chan91ng Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis", Nucitar Science and Engineering, Volume 94, Pages 291-308, 1986.
+
8-4
4
~
r t t 4 t _ 9
- i APPENDIX A !
Load-Time Records for Charpy Specimen Tests ,
., a L
f j 'h 4
- 4
. h .
s 9 g.
. g s
'f. e
k i . .
., . , ...,.,n,--. . , . , - , , . , , . . , ., - ~ , . . , ~.. ~ , 4,~,
a.
r
;+ =& -
m 4g: l
?
Ag 2. n. S I O D A O A L O L E . R T q U S E T R C R . A T R A . F = p
= 4 w , P P
O. A O L
/ I ; i I l 1
=
M d U r o M c I e X r A [ e M. m
- M i I
t y T - P d a sg l 1 I il t i I I I 1 I I 3 l o , l
= -
" 'ec i
l z , a e d g I t 1 A e r n u W a i g t m F L
- A R
E D , A ENOL .
'G
= D _
y EL l l I l 1 i l l al y
=.. g
.h, P Y t
=
=
o4Oa ,
>s"
\ n wu >w ..
l' l l e e i a i 4 TL15 0.- g4 7 a 3 e2, - w ga o , i e 5*. P 6.6 7.4 8.3 9.I 10.O TIMC ( M3EC ) atone. a i
? i i i
TL14
- I _
,- j s !
.. i D w w
o g- - e o- _. e i M i A. ~ _^
.n i
i
* .V 2.0 4.0 6. 0 0.0 10.0
- TIME ( tGEC ) m10ma-1 I'igure A-2. Load-time records for Specimens TL15 and TL14 1^
A-2 ).' n,
- - g y I l l
- i i i i p . .
TL11 "h 4e.
- 2 a
g g- - n l [' s. ~ e 94 . S.
. j p'
9- . 1 A >
^
A-m } a- , e u n 1
.9 2.0 4.0 6.0 0.0 10.0 j T! tit ( ttSEC ) m10**-1 '
"-- i i i i
TL9 9 ._ _ S4 7.' a . 39" n M' - v. v . d'
.q -
-o - e a n n
.D 2.0 4.0 6.0 8.0 10.0 TIME < ttSEC > m10am-1
- Figure A-3. Load-time records for Specimens TL11 and TL9 A-3
m _ ._ _ _ _ . P a i
~}. i i TL10 l.
0 9 - L ,- m'
- t. ) 39- n v.
.u.
- 4_
, -.- a- - i-. ?: o , u s i (E
? .D 2.0 4.0 6.0 0.0 10.0 TIME ( PCEC ) sicam-1 4 I i (
TL2 "k Jm- _- I e
^ ?a.
g .. y - n v
.N-N
.m N --
o i i i
.9 3.0 6.0 9.0 12.0 15.0 5 .e TitC ( MSCC i mW=-1 4
Figure A-4. Lond-time records for Specimens TL10 and TL2 A-4 Q:-
F# .-. j[b e -
, i i i " ' - ,f .*
TL3
.a._ -
k w -- m-n- 3**a w d *
- w. -
,y 3.,- ...
ie.6 w. o TINC ( M!CC ) =10=s-1 liy e p TLS NA 1 Figure A-5. . Load-time records for Specimens TL3 and TL6 A-6
yt yr:- _){. I i i i TL1 L-* ' g- * - ei
. 3 g- -
n w
. d
%* ~ l l
i o- i i i 1 [ ,t 3.0 6.0 9.0 12.0 15.O TIME < MEC > m10me-1
*, e i i i
e TL13
?- -
& v*
it w
$ *e-i d
q- - V t- s
.?- o i a e i og 3.0 6.0 9.0 12.0 13.0 TIME ( MSEC ) m10am-1 ~
Figure A-6. Load-time records for Specimens TL1 and TL13 A-6
7-I'I s e i 8 1 4 TL4 7 s. '**
&4 4 y
a 3 *." ~ v
- v. - .
9 t0 . 3 N* - a
~
o- , 55' i
,g .5 1.o E.0 5*$
gg7g < EC ) . o i ' 4 4 TL12 2*- - . 5 s. 1 ~
$ *n.-
-v i
I N
~
".~
%w g 1.0
,,3' a.9' *5 TI < ttsCC 3 .
Figure A-7. Load-time records for Specimens TL4 and TL12 A-7 t' . 1
?W' i-
" : I a r J-
.>. TL7 7e _
+:'.. &4 7
e
- . 3 g_
m v
\ _
. A l -
- 9. -
o a i i i
.D .7 1.4 2.1 E.8 3.5 TIMC ( M:;CC )
i i i i e TL8 g e- - a- 4 7-
%- n 3 e- n v
w-, 3._ o i e i i
.g .7 1.4 2.1 2.8 3.5
- TIME < MSEC )
4- Figure A-8. Load-time records for Specimens TL7 and TL8 A-8
3 -;. _.
*~ , i i i 4
TL5 m._ _ h4 . I I i' a-3 *n-v T= l e 1" -
~
9* o-- i . i i 3.5
- j. .t . 7. . 1.4 2.1 2.8 TINE- ( NSEC >
- i i i e
TT9 _ g ., ' _ a' , S ,*- w q_ -
--- ~ .
o a e s s
.D. 2.0 - 4. 0 . 6.0 8.0 10.0
- TIME . '( MSEC > m!0mm-1 .
~
Figure A-9. . Load-time records for Specimens TL5 and TT9 A-9
g s - n s TT4-
? .- -
E &J 7 a
. 4
. 3-n-w
-4 v. -
d
,e
.. N,, -
c~ / _ AA w- -s _/
. e i
.D 2.0 4.0 6.0 0.0 10.0 TINC ( MSEC ) 510am-1 i i i i j
TT5 9 a- - 54 7 4 3 'ni- , v
- v. _
d . n- - o i i i xh i
.D 2.0 4.0 6.0 8.0 10.0
*' TINC ( NSEC ) a10eu-t 1
. Figure A-10. Load-time records for Specimens TT4 and TT5 A-10 l
1
--_ _-_---__-_L___.__.____
.o
' I i J. i TT2 x.
ke 4
- E n
3: M1- -
- v d .
- 9 n i a i a
.V . 2.0 4.0 6.0 8. 0 10.0 TIMC ( MSEC ) m10am-1 a
y 4 I I I TT7
? a- -
a4
-E
$ '*M ~ -
w-
. N u- -
o i i i i
.D 2.0 ' 4. 0 6. 0 8.0 10.0
- TIME ( PCCC ) alose-1 Figure A-11. Load-time records for Specimens TT2 and TT7 A-11
, ; =-;
- u. .
)
I
-. 1 I E 4 l TT8
[ $- e e 9 -, M v 9
. -N-v -
i.
.{
c' i i i i
-..D 2. 0 - 4.0 6.0 0.0 10.O TIME < PCCC ) ut0=e-1 o
g .i i a i TT13
? ._ _
a4 m n - S *. n - v
. . w
.g .
y- _
.D 2. 0 4.0 6. 0 - 8.0 10.0
- TIME C MSEC ) m10am-1 L'
)
-Figure A-12. Load-time records for Specimens TT8 and TT13 A-12
Di f-6
* .. _i i i 4-TT11
' t e-
&4 .
2
-A 3 9,-
s , N .
=
N-
- A o . i i e
.V .5 _ 1. 0 1.5 2.0 2. 5 TIPC ( H3tc )
o i i
- y. 3 4 TT12
? e- -
- - S'i
! 7 l. 3 .9. - e p< '.- ._ :w N-J . o , , , ,
.D .5 1.0 1.5 2.0 2.5 T!rE < MSEC ) .
Figure A-13. Load-time records for Specimens TT11 and TT12 A-13
s . L o a ,
- i. s 4
TT1 m, 3- . w a-3 *n," 8 w-
.. e _
4
.. y_ -
*f .5 1. 0' ggg 1.5' 2.0' 2.5
< tCtc >
.c i '
4 9 1 TT10 rt
&.. w a
g ,e_ _
.y e- -
4
.M *"
4 '. - ** *
- g .5 1.0 1.5' 2. 0 *
- ,, : TIK I 'CEO }
~9 .9-Figure A-14. Load-time records for Specimens TT1 and TT10 A-14
pg, -
- m. - -
4 s e
'* i s
. i e e
TT14 p r. _ _
=
. 54 o-7 39- M v_
v- - .-
"- ~
[- e
' 2.5 i e i i
.t .5 1.0 1. 5 - 2. 0
' TIME ( MSEC )
I] i. i e i TT6 9 o v._ m n . g g_ M v-
.h
>-. -B_
'u_ -
-J o i i e i
.t .5 1. 0 1.5 2. 0 2.5 TIME < MSEC > .
. Figure'A-15. Load-time-records for Specimens TT14 and TT6 A-15
+
. i
;g .. i
,. TTIS
.p..- .
,W 39- n w
a p- q- . a- i i i i
.D .5 1.0 1.5 2.0 2.5 ~
TIME C tt:CC )
., 'lT3 NA Figure A-16. Load-time records for Specimens TT15 and TT3
.- A-16
. .- . . . . 2.____.___.________________
,i t .-
.o.
4 3
- 4~ l-TW3 f
a _
.. Ly$ .
7 a o ~
$ m- 4 .
v o - J-
~ .
- o. '
d s o i e i
.D. 2.0 4.0 6.0 8.0 10.O TINF. ( MSEC ) sloss-1
) i i i TW12 y
e- - S4 7
^ ,
39-. n. v
. - c}
N-
+%t m - , ^ - .-
'.D 8.0 4.0 6.0 8. 0 10.0 TIME- ( MSEC ) stoma-1 Figure A-17. Load-time records for Specimens TW3 and TW12 A-17
yem
";ry; a , , ' , !Y: - ,
g TW2 j .. , , 9-._ - s' ' L.< -
-. p .
a o-
- 9 9_ n v
.; a w_ _
' ^ ^
S e,e' 4,,' > eo io. o um e .etc > ,,,';',
' . I 4
TW6 p.,_ - 04 f 99- n t-w-
+
.N. -
S e.o' - 4,,' ,,,. s. o
,o,,
- . . . Tim - ( mte > , m..,
y.
-~,.
Figure A-18. Load-time records for Specimens TW2 and TW6 A-18 E
1 3 I I TW4 9.. _ a4 . T n 3*- n v 9" - o- i i i i
.9 2. 0 4.0 6.0 8.0 10.0 TINC ( M;EC ) a10am-1 o
I i i y 4 TW9
- ? . _
4 3.** n w m t i , i
.D 2.0 4.0 6.0 8.0 10.0 TIME ( M3EC ) m10am-1 .
. Figure A-10. Load-time records for Specimens TW4 and TW9 A-19
g:; .- T 4 5 5 8 TW13 7 ._ _
-Si 5 4
.if s '. s.
i 3 e$_ _ w
>d:
+-
- , N~ ~
o i . i i
.9 4.0 0. 0 12.0 16.0 20.0
- T!t1C ( M5CC ) st one-t
} . . . .
TW7
.m
, .~ ^
33M _ v
- y. _.
o -- . . . .
.D 4.0 8.0 12.0 16.0 20.0 e TIMC < ttSEC ) m10am-1 i
~
Figure A-20. Load-time records for Specimens TW13 and TW7 A-20
-g ,. -
q t r -i 5 4 4 3 j TW11 [**~ m , 9 *n-c ! v p . na N- -
~
o- s a e ,
.V 4.0 0.0 12.0 16.0 20.0 TIMC t PtSEC ) alcan.1
} i i T TW1 U .- - .
54 17
^ .
S n*- q-
~
o i . i i i
.V- .7 1.4 2.1 2.0 3.5 TINC ( ttSCC > .
Figure A-21. Load-time records for Specimens TW11 and TW1 A-21
'A
- i i i
TW15
, .- ? 9- -
3
.m l
.- 5*- n v
_j. cJ - j y
. N~ ~
.t .7 1.4 E.1 2.0 3.5 TIME ( NSEC )
. i i i i TW10
, ; =_ -
a *? a-3 *.n - - v v I - 1 l q- -
--
- o- . . . i
,0- .7 1.4 2.1 2.8 3.5 e ilhC ( MSEC ) .?
' ~ Figure A-22. Load-time records for Specimens TW15 and TW10 A-22 9,
. ~ = -
*l -- . i e i TW5
?.- -
8 ' 4. 7 a 3 g- - n v t s - d . q- - N [g ~ , 7'- _
.1.4' 2.1 ' 2. 8 ' 3.5 TIME < MSEC )
- i i i i TW8 i
? a- -
54 7 a ., s.9-n v I
~o e i a i
.D .7 1.4 2.1 2.8 3.5 TIME < MSEC >
Figure A-23. Load-time records for Specimens TW5 and TW8 A-23
7;- 4 } ',
.* , i a i TW14 1.-
L.
. m-a 3 3 Mg_
w e- w .,
~
ig -,. oa
.g i: -+ %
N
- 1. . ~- -
- l-
~
, g'- ,7 ' g.4' 2.t' 2.8' 3.5 TlME ( MSEC.)
}
- i i i i THS Qm ..
a4
. m a.
33_ n-v. T _ N - 4
- 9'
,.- O s a i s
.9 .0 1.6 2.4 3.2 4.0
! .g TIME < MSEC ) a1Oum-1
- Figure A-24. Load-time records for Specimens TW14 and TIls a
A-24 i' ..
- , 7 7 c
1 t o-y I I I I
, 4 ;-
TH7
; $ as -
34
- e 9 n*- -
v w _ 4 . N-A
# > e i i
.1 1.6 3.2 4.8 6.4 0. O TINC ( PCEC 3 aloss-1 i
E t. L
*t - . . i i TH2 -
- c. -
1 N
'7m -
54 . 7' a 3.g_ v v -
. , 4 9
'l , , .
A. - - .
.V 3.0 6.0 _ 9.0 12.0 15.0 ,
TIME ( PtstC ) s10er-1
-Figure.A Load-time records for Specimens TH7 and TH2
. t A-25
{if
(; .- ,
??
i r!.. e-y 1 3 5 i THIS 9 5v n- - S
$ n~
..- v
- l
1 ;e- -
. P, N~ -
7 .. -
= i , , P,- --
e ..D. 3.0 6.0 9.0 12.0 15.0 l -? TIMC ( MEC > mlons.1 (.
} s a i i TH1
? a- -
- -.. g4 E
a 39- n w e- -
~
.- ^
o a i. . .
*' .D 3.0 6. 0 9.0 12.0 15.0 TIME ( MCC ) m10ee-1 Figure A-28. Load-time. records for Specimens TH15 and TH1 A-20 4
b
o I i f 3 4 l TH10
?-
a n. S *- n
-v ,
N p . N_ h o ,
*D 3.0 6.0 9. [ 12.h 15.O TIME. ( K*EC ) m10am-1 o .
I d 3 4 y TH3
?
a v9- - a r% 99- m v
~
w 0 - N_ o
.D 3.0' 6. 0' 9.0' 12.0 15*0
- TIME ( MSEC ) *10ma-1 Figure A-27. Load-time records for Specimens TH10 and TH3 A-27
, e o
- g. I i e i
-TH5 0=- _
_L , a
- h. 4 ^
$ e-
- n. i e4 v
L
+ .
'd
. ce
'.p ~
N .
,' o -- i 6 i i i i .9 3.0' 6.0 9.0 12. 0 = 15.0 TIME ( MSEC ) m10 n-1 };;
4 o 4 3 l 1
- . TH8
~
O~ sr-a
. g ae-i -
w
- v. -
A m_ -
-w i '4 e i_
i i i
.D 4.0 0,0 12.0 16.0 20.0
'* TIN ( N3CC ) m10em-1
.+
, .T o
[' Figure A-28. Lond-tiwe records for Specimens TBS and TH8 A-38 i
___,_--s,_ - - - - - - - -
" i i i 3 4
TH14 "g *e - e F A 3 M.- d 3
.. 4.0 8.0 12.0
- 19. 0 20.0 2
firt ( MSCC ) escas.1 i
}- i e i TH9 .
1 e-F -
$*- e 1
J - v f-W, - N w- - o- i i i i
.0 .6 1.2 1.8 E.4 3.0
- T!PL ( MSEC )
Figure A-29. Lond-time records for Specimens TH14 and TH9 A-29
} i i i i TH12 a -
1_ r e
- n 37 0; *
~
_ [ A~ q_ - 4 M [g .4 ' l.t' l.0 ' 2.4' 3. 0
~
TINr. - ( M CC )
"g- i i e i TH4 se .
. ,, E a
e 3 j-v
- v. .
M N
'*I e . ,
i i
, .D .6 1.2 1.9 2.4 3.O TIME ( MSEC )
4-Figure A-30. - Load-time records for Specimetis TH12 and TH4 A-30 Ri
} i i i i TH11
~
$ 7~ v . a $*-n e . y [t 1.0' 2.4' 3.0
.6 1.E' 11HE ( NSCC )
l y I I ~i i TH13 -
~ $,U~ , ,
e 3 _ j "" , v 1 - (h N-a , , i ,
.D .6 1.2 1.0 2.4 3. 0 .
TINC ( ttstC ) Figure A-31. Lond-tico records for Specimens Tilli and TIll3 A-31
.g (,
g 1 APPENDIX 0 lleatup an4 tooldown Limit Curves for Normal Operation for Comancho Peak Unit I for 16 and 32 EfPY
- h' d
-4
.e.
-n.
.e I
r
~. -,-.m - y.. ._.. c. . -y,,,., -
TABLE OF CONTENTS i
~
12.C110.0 11.112 h92 , List of Tables 11 *
- List of Figures 11 B1 Introduction B-1 -
B2 Fracture Toughness Properties B-2 ,' B3 Criteria for Allowable Pressure-Temperature Relationships B-2 B4 Heatup and Cooldown Pressure-Temperature Limit Curves B-6 B5 Calculation of the Adjusted Reference Temperature F-8 B6 References B-21 Attachment B1 : Data Points for Heatup and Cooldown B1-1 Curves (With Margins of 10'F and 110psig for instrumentation Errors) . Attachment B2 : Data Points for Heatup and Cooldown B2-1 Curves (With No Margins for Instrumentation Errors) , 9 0 4 m l'
~ - -
e i e LIST OF TABLES 4 Ithin Title ELqe Bl Comanche Peak Unit 1 Reactor Vessel Beltline Region B-7 Material Properties (Unirradiated) ,
, B2 Summary of Comanche Peak Unit 1 Beltline Region B-10 i Materials Adjusted Refe~ n
- Tempmeatures at 1/4T and 3/4T Locations for 4 rao JY B3 Calculation of Adjusted Reter.w s temperatures at B-ll 1/4T and 3/4T Locations for the Comancho Peak Unit 1 Reactor Vessel t.imiting Materia: at 16 EFPY -
Plate R1108-1 B4 Calculation of Adjusted Reference Temperatures at B-12 1/4T and 3/4T Locations for the Comanche Peak Uni' 1 Reactor Vessel Limiting Material at 32 EFPY - Plate R1108-1 LIST Of flGURES fi9Et lltle EL92 B1 Reactor Coolant System lleatup Limitations B-13 Applicable for the first 16 EFPY (With Margins of-10'F and 110 psig for Instrumentation Errors) w B2 Reactor Coolant System Cooldown Limitations B-14
- Applicable for the First 16 EFPY (With Margins of 10*F and 110 psig for Instrumentation Errors) 11
i LIST OF FIGURES (CON'T) Fiaure litig P_agg 83 Reactor Coolant System Heatup Limitations B-15 , Applicable for the first 32 EFPY (With Margins of 10'F and 110 psig for
- Instrumentation Errors) ,
B4 Reactor Coolant System Cooldown Limitatons 8-16 -
' Applicable for the First 32 EFPY (With Margins of 10*F and 110 psig for Instrumentation Errors)
B5 Reactor Coolant System Heatup Limitations B-17 Applicable for the First 16 EFPY (With No Margins for Instrumentation Errors) B6 Reactor Coolant System Cooldown Limitations B-18 Applicable for the First 16 EFPY ' (With No Margins for Instrumentation Errors) B7 Reactor Coolant System Heatup Limitations B-19 Applicable for the First 32 EFPY (With No Margins for In.trumentation Errors) B8 Reactor Coolant System Cooldown Limitations B-20 Applicable for the First 32 EFPY - (With No Margins for Instrumentation Errors) e iii y r - mm
Bl. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of Adjusted Reference Temperature (ART) for the reactor vessel. The most
. limiting ART of the material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness proporties and estimating the radiation-induced ARTNDT. The unirradiated RTNDT is the higher of either the drop weight nil-ductility transition . temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mils lateral expansion (normal to the major working direction), minus 60*F.
The ART of a material increases as it is exposed to high-energy neutron radiation. Therefore, to determine the most limiting ART at any time period in , the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT, along with margin to account for uncertainties in the calculation. The extent of the shift in RTNDT is enhanced.by certain chemical elements (especially copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has pub 1hned a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. ? (Radiation Embrittlement of Reactor Vessel Materials)[B1), Regulatory Guido 1.99, Revision 2 is used for the calculation of RTNDT values at 1/4T and 3/4T locations (T is the thickness of the reactor vessel at the beltline region, measured from the vessel inner surface). Based on the recent dosirstry results of surveillance capsule U, design basis neutron exposure projections beyond 0.91 EFPY have bean calculated for Comanche Peak' Unit 1 at key locations in the reactor vessel. A new heatup and cooldown ', evaluation using this data has been performed to determine the most limiting material in the beltline region' corresponding to the highest ART values at 1/4T
'~
and 3/4T locations for Comanche Peak Unit 1. This appendix summarizes these results and presents the new heatup and cooldown curves for Comanche Peak Unit I with and without margins for possible instrumentation errors for the service periods up to 16 and 32 EFPY, B-1
in addition, a review of the latest heatup and cooldown evaluation for Comanche . Peak Unit 2 was performed to determine the most limiting material bounding both plants. The results of this review indicated Comanche Peak Unit I heatup and cooldown curves can be applied to Comanche Peak Unit 2. This appendix presents the heatup and cooldown curves bounding Comanche Peak Units 1 and 2 with margins for possible instrumentation errors for a service period up to 16 EFPY. B2 ' FRAGURE TOUGHNESS PROPERTIES ,' The fracture-toughness properties of the ferritic material in the reactor .' coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan [82). The pre-irradiation fracture-toughness properties for the Comanche Peak Unit I vessel materials are presented in Table Bl. B3, CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg, for - the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than.the reference stress intensity factor, KIR' for the metal temperature at that time. Kig is obtained from the reference fracture toughness curve, defined in Appendix G of the ASME Code (83). The KIR curve is given by the following equation: KIR = 26.78 + 1.223 exp (0.0145 (T-RTNDT + 160)) (1) where - KIR - reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature - RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined .. in Appendix G of- the ASME Code (83) as follows-B-2
1 l CcKjg 4 KIT 6KIR (2) i where i Kgg - stress intensity factor caused by membrane (pressure) stress l KIT = stress intensity factor caused by the thermal gradients
',. Kgg function of temperature relative to thi RTNDT of the material
*- C - 2.0 for Level A and Level B service limits C - 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses
< resulting from the temperature gradients through the vessel wall are calculated, and the corresponding (thermal) stress intensity factors, KIT' for the reference flaw are computed. From equation 2, the pressure atress intensity factors are obtained and from these the allowable pressures are calculated, for the calculation of the allowable pressure versus coolant temperature during cooldown, the reference-flaw of Appendix G of the ASME Code is assumed to exist at the inside surface of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside surface because the thermal gradients produce tensile stresses there, which increase with increasing
- cooldowy rates. - Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rates. From these relations, composite
-limit curves are constructed-for each cooldown rate.
.The use of the composite curve in the cooldown analysis is necessary because
.. control of the cooldown procedure relies on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw, B-3
~ - -_. -
h r During cooldown, the 1/4 T vessel location is at a higher temperature than the coolant in contact with the vessel. This condition, however, is not true for steady-state. It follows that, at any given reactor coolant temperature, the
- AT developed during cooldown results in a higher value of KIR at the 1/4 T locstion for finite cooldown rates than for steady-state operation. ,
Furthermore, if conditions exist so that the increase in KIR exceeds KIT' the calculated allowable pressure during cooldown will be greater than the steady-state value. . The above procedures are needed because there is no direct control on .' temperature at the 1/4 T location and, therefore, allowable pressures may P unknowingly be violsted if the rate of cooling is increased at various intervals during cooldown. The use of the composite curve eliminates this problem and ensures safe operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as - finite heatup rate conditions assuming the presence of an axial flaw which extends to the 1/4 T location. At the 1/4 T location, the tensile stresses
; produced by internal pressure are somewhat alleviated by the compressive stresses resulting from thermal gradient. The metal temperature-at the crack tip lags the coolant temperature; therefore, the KIR for the 1/4 T flaw during heatup is lower than the KIR for the 1/4 T flaw during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive f thermal stresses and lower KIR's do not offset each other, and the pressure-
- temperature curve. based on steady-state conditions no longer represents a lower
- bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. - Therefore, both cases must be analyzed to ensure that at any
-coolant temperature the lower value of the allowable pressure calculated for steady-state and~ finite heatup rates is obtained.
e B-4
The second portion of the heatup analysis is the calculation of the pr%sure-temperature limitations for an assumed axial outer surface flaw which
. extends inward to the 3/4 T location. Unlike the situation near the vessel inner surface, the thermal gradients during heatup produce tensile stresses at . the 3/4 T location, which add to the pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the coolant temperature.
Since the thermal stresses near the outside surface are tensile and increase
- with increasing heatup rates, each heatup rate must be analyzed individually.
Following the gener6 tion of the steady-state and finite heatup rate prersure-temperature curves, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the minimum of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over' the course of the heatup, the controlling location of the reference flaw switches from 1/4 T to 3/4 T, and the pressure limit must at all times be based on analysis of the most critical location. Finally, the 1983 Amendment to 10CFR50IB43 has a rule which addresses the
. metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange region must exceed the material RTHDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Comanche Peak Unit 1).
The minimum allowable temperature is based upon the limiting initial RTNDT I the vessel'and closure flange for Comanche Peak Unit 1. Table B1 indicates that tho'unirradiated RTNDT of 40'F occurs in the ciosure head flange of the Comanche-Peak Unit I reactor vessel, so the minimum allowable temperature of this region is 160'F at pressures greater than 621 psig. These limits are shown in Figures B1 through B4 and are more restrictive than the limits shown in Figures B5 through B8, which do not include margins for possible instrumentation errors. 4 B-5 __m______._.____..____.m._
B4. HEATUP AND COOLDOWN PR[SSURE-TEMPERATURE LlHli CURVES Pressure-temperature limit curves for normal heatup and cooldown of the Reactor , Coolant System have been calculated using the methods discussed in Section B3. Figures B1 through B8 contain the heatup and cooldown curves using heatup rates , of 20, 60 and 100'f/hr and cooldown rates up to 100*f/hr for service periods up to 15 and 32 EfPY. These curves were generated with and without margins for possible instrumentation errors for 16 and 32 EfPY. Note, figures .' B1 and B2 show the heatup and couldown curves bounding Comanche Peak Units 1 , and 2 for a service period up to 16 EfPY. A review of Comanche Peak Units 1 and 2 heatup and cooldown evaluations indicate that Comanche Peak Unit 1 ART values of 84*f at 1/4T and 69'F at 3/4T locations are higher than Comanche Peak Unit 2 ART values of 81*f at 1/41 and 62*f at 3/41 locations for a service period up to 16 EIPY. Thus, the latest heatup and cooldown curves generated for Comanche Peak Unit I are mor9 conservative and therefore can be applied to Comanche Peak Unit 2. Allowable combinations of temperature and pressure for specific temperature
- change rates are below and to the right of the limit lines shown in figures B1 through B8. This is in addition to other criteria which must be met before ,
critical reactor operation. All data points for the heatup and cooldown curves with and without margins for possible instiamentation errors are shown in Attachments B1 and 02, respectively. 9 B-6
l 1 TABLE B1 COMANCHE PEAK UNIT 1 REACTOR VESSEL BELTLINE REGION l MATERIAL PROPERTIES (UNIRRADIATED) ; Unirradiated i Cu(a) Hi(a) RTNDT !
.. _aterial M Description (Weloht %) - (Weicht %) _(*F)
.. Intermediate Shell-Core Region ,
t Plate R1107-1 0.06 0.65 10 1 Plate R1107-2-. 0.06 0.64 .-10
. Plate R1107 0.05- 0.688 10
. Longitudinal Weld- 101-124A- 0.04 0.195 -70
- Longitudinal Weld 101-124B 0.04 0.195 -70 Longitudinal _ Weld 101-1240 - 0.04 0.195 q Circumferential-Weld 101-171- -0.04 0.195 -70 j Lower Shell Region: ;
- Plate R1108-1 0.08 0.64 0
- Plate R1108-2" 0.05 0.59- 20 x
. Plate R1108-3 0.07 0.64 0 ;g - Longitudinal _ Weld 101-142A 0.04 0.195 -
70
- Longitudinal Weld.101-142B 0.04 0.195 - 70 Longitudinal ~ Wald 101-142C- . 0.04- 0.195- -70 1 1
- ClosureHesdFlange(C) - 0.77 40 t Vassel-Flange (c) - 0.72 10 (a) All_. %Cu and %Ni . values are obtained from Combustion Engineering material certifications, except for the weld %Ni_ content, which is taken to be-the
^
average of the CE material certification data and the surveillance capsule-matehial analyses (BS). Note. the surveillance capsule weld metal is representative 1of- all welds in the beltline region of the reactor vessel.
- (b)_TheunirradiatedRTNDT values'are measured values.
; (c) Used whe.n-considering flange requirements for heatup/cooldown curvesl84}.
L<
~
B-7 1 (r c-* s- -~ h - J p- , y- g e 'w-.v,-o-
- ..rvn,-wn ,mw- ,c.,w.u, y--+-nw-..e-,-. ,,w.<w.,vw sc ..----.,-e--y4 s-
The leak licit curve shown in Figures B1, B3, 85 and B7 represents the minitu] temperature requirements at the leak test pressure specified by applicable
-codes [B2,B3), The leak test limit curve was determined by using the methods of References B2 and 84. .
The criticality limit curve shown in Figures B1, B3, B5 and B7, specifies . pressure-temperature limits for critical core operation to provide additional margin during actual power production as specified in Reference B4. The , pressure-temperature limits for core operation (except for low power physics . tests) require that the reactor vessel be at a temperature equal to or higher . than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the calculated minimum pressure-temperature curve for heatup and cooldown, as described in Section B3. The minimum temperature for the inservice hydrostatic test for the Comanche Peak Unit I reactor vessel is 230'F and 240*F for a service period up to 16 and 32 EFPY, respectively, using curves with margins for possible instrumentation errors. A vertical line at 230'F and 240'F on the pressure-temperature curve with u.argins for 16 EFPY and 32 EFPY, respectively, intersecting a curve 40'F e- higher than the pressure-temperature limit curve, constitutes the limit for ' core operation. Similarly, the minimum temperature for the inservice hydrostatic test- for the Comanche Peak Unit I reactor vessel is 216*F and 226'F for a' service period up to 16 and 32 EFPY, respectively, using curves without margins for possible instrumentation errors, A vertical . line at 7.16*F and 226'F on the pressure-temperature curve without margins for 16 EFPY and 32 EFPY, respectively, intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation.
-Figures B1 through B8 define limits for ensuring prevention of nonductile
- failure for the Comanche Peak Unit I reactor vessel. ,
B5, CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev 2[BI), the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
- ART- = Initial ' RTNOT + ARTNDT + Margin (3) i B-8
Initial RTNDT is the reference temperature for the unirradiated material as ; defined in paragrapn NB-2331 of Section 111 of the ASME Boiler and Pressure j Vessel Code, ;
,. if measured values of the unirradiated R1NDT for a material are not available, generic mean values for that class of material may be used if !
there are sufficient test results to establish a mean and standard deviation for the class. j i
. ARTNDT is the mean value of the adjustment in reference temperature !
, _ caused by irradiation and is calculated as follows: l ARTNDT = [CF)f(0.28-0.10 log f) (4) ,
To calculate ARTNDT at.4ny depth into the vessel wall (e.g., at 1/4T or 3/4T), the subsurface fluence must first be determined: f(depth X) " Isurface(e .2h) (5) where x-(in inches) is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then used in equation (4) to calculate ARTNDT at the specific depth. The chemistry factor, Cf ('F), is a function of copper and nickel content, and is obtained from Reference B1, in Table I for welds and in Table 2 for bas metals (plates.and forgings). Linear interpolation is permitted. L Applying Regulatory Guide 1.99 Revision 2 procedures to all the beltline region materials it was found tnat the plate R1108-1 was the limiting meterial. The ART values for plate R1108-: dere evaluated at 1/4T and 3/4T locations using
- _ chemistry factor values determined from Table 2 in Regulatory Guide 1.99
' Revision 2 [81),
The ART values at 1/4T and 3/4T-for all materials in the beltline region of the Comanche' Peak Unit I reactor vessel are summarized in Table B2. These values
.:_ were obtained by using the Regulatory Guide 1.99 Revision 2 tables. Sample calculations of ART are shown.in Tables B3 and B4.
t B-9 i..
~-
TABLE B2
SUMMARY
Of COMANCHE PEAK UNIT 1 BELTLINE REGION MATERIALS ADJUSTED REFERENCE TEMPERATURES AT 1/4T and 3/4T LOCATIONS FOR 10 AND 32 EfPY , t ART at 16 EfPY ART at 32_EfPY Component 1/4T ('f) 3/4T (*f) 1/4T ('f) 3/4T (*f) , Intermediate Shell Core Region: . , Plate R1107-1 80 61 87 75 Plate R1107-2 60 41 67 55 Plate R1107-3 70 53 80 64 Longitudinal Weld 101-124A 0 -23 17 -8 Longitudinal Weld 101-124B 10 -14 27 2 Longitudinal Weld 101-124C 10 -14 27 2 Circumferential Weld 101-171 13 -11 29 5 Lower Shell Region: ' Plate R1108-1 84 69 93 79 Plate R1108-2 80 63 90 74 , Plate _R1108-3 77 61 85 73 Longitudinal Weld 101-142A 0 -23 17 -8 Longitudinal Weld 101-1428 10 -14 27 2 Longitudinal Weld 101-142C- 10 -14 27 2 9 9 e 4 6 B-10
TABLE B3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES AT 1/4T AND 3/4T LOCATIONS _FOR THE COMANCHE PEAK UNIT.1 REACTOR VESSEL LIMITING MATERIAL AT 16 IFPY !
. PLATE R1108-1 j Reaulatory Guide 1.99 - Revision 2 j 16 EFPY l
. Ear.Ameter .1/4. T. . 3/4 1 !
+- Chemistry Factor, CF (*F)- 51.00 $1.00 Fluence, ' f.' (10 19 n/cm)(a) 2 0.906 0.322 j Fluence Factor, if 0.972 0.689 ,
1
**********************************************v ******************************* ,
v L 50 35 i ARTNDT =: CF x .ff ('F). _ t , -Initial RTNDT, I ('F) . 0 0 I
- Margin. M ('F) (b) 34 34 i
- ,, r Revision.2 to Regulatory Guide 1.99 ;
1
-Adjusted Reference Temperature, 84 69 ART. = Initial. RTNDT + ARTNDT + Margin r
l(a) Fluence, 'f, is basJd upon' sf urf (10 l9 n/cm2 , E>1-Nev) - 1.52 at 16 1 L EFPY. - The Comanche- Peak Unit I reactor vessel wall thickness is 8.625 l k[ + inches at the beltline region, ) .(b)f Margin is calculated as, M = 2 (og2 + oA 35 0 The ;
; standard ' deviation for.the initial RTNDT. margin. term (aj) is ;
zero*F since the initial RTNDT is a measured value. The standard
- deviation for ART NDT i -(84 ) is 17'F for plate- R1108-1,- s 7 except that oA need not exceed 0.50 times the mean value of ARTNDT*
L B-11 l rE /
,. ._1
_,. , , . _ . , . - . - - , _ . , _ , , , . . , _ . . _ - . , _ ~ , _ . ,, ,,.
7 } TABLE'84
- CALCULATIDN OF ADJUSTfD REFERENCE TEMPERATURES FOR AT 1/4T AND 3/4T l
" f LOCATIONS FOR'THE COMANCHE PEAK UNIT 1 REACTOR VESSEL LIMITING MATERIAL AT 32 ffPY ,f PLATC R1108-1 .!
Reaulatory Guide 1.99 - Revision 2 : i 32 EFPY .
' ParametE 1/4 T 3/4 T .
+
i Chemistry- factor, Cf ('f), 51.00 51.00 O : Fluence,f(10 19 n/cm)(a) 2 1.812' O.644
. Fluence Factor, ff-. 1.163- 0.877 l
-**************************************************v.****************************
e L 59 45 i ARTNDT = CF x ff_('F).
--Initial RTNDT,11-('F) 0 0 l
[ -Margin, M (*f) (0) 34 34 r I
.********************************(.At********************************************
Revision 2:to Regulatory Guide.1.99
' Adjusted Reference Temperature.: 93 79 ART ; Initial RTNDT + ARTHDT_+ Margin-e*******************************************************************************
{ 19 2 (a) Fluence, f, is! based:upon fsurf(10 n/cm.E>1-Mev)=3.04at32 EFPY. The Comanche Peak Unit-1 reactor vessel wall thickness is 8.625 ,'
-inches at' the beltline region. ,
;(b)':;iMargin is calculated as, M -l2 [ag 27, , 2 0.5
~
3 . The ,
- standard deviction for the-initial' RTNDT margin term (oj) is zero'FfsincethelinitialRTNDT_is.a measured value. The standard
< ; deviation foriARTNDTri.(8 4 ) is 17'F: for plate R1108-1, .,
, exceptithat:oA need not exceed 0.50 times the mean value of LARTNDTK-B-12 ,
. J
MATERIAL PROPERTY BASIS i CONTROLLING MATERIAL: LOWER SHELL PLATE R1108-1 (UNIT 1) INTERMEDIATE SHELL PLATE R3807-2 (UNIT 2)
., INITIAL RTNDT: 0'F (UNIT 1),10*F (UNIT 2)
ART AT 16 EFPY: .1/4T : 84'F (UNIT 1), 81*F (UNIT 2) ; 3/4T : 69'F (UNIT 1), 62'F-(UNIT 2) A CURVES BOUNDING COMANCHE PEAK UNITS 1 AND 2. APPLICABLE FOR HEATUP RATES UP TO
$ 100'F/HR FOR THE SERVICE PERIOD UP TO 16 EFFY. CONTAINS MALGIkS OF 10'F AND 110 PSIG FOR POSSIBLE INSTRUMENTATION ERRORS.
I 2500 ym 331,,,;;;;3 ;j , , )j , LEAK TEST LIMIT ~I --H.-4ly I l n$o i
.iii iii _, i i
' CRITICALITY LIMIT i/I.
~ -- -
d -
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OPERATION - -
' f . 4f 2000 ,
i l----/- j I' _ I. L LI P1 .
. i r'-N'Y-bCRITICALITY LIMIT d
+
'1750 4
, HEATUP RATESNd UP 10 ; ,ji/' /
j; i ,/- = CORRESPONDING TO <-- 9 j 20'F/HR .4J g_;ff ;L_f 60'F/HR HEATUP M1E '{C , 3 1500 i , , / -/-lf-/ /i 1 _y f,
- i !
l l_ jjjj 14 s
~~~~
TES r! i i f ~ ~I b CRITICALITY LIMIT i w 1250 hA[. i
%/4/-((-/--[-4 CORRESPONDING TO -
i
! 60'F/HR i i
' <!/ i7 / /!i// i 100*F/HR HEA10P RATE 1
C . f-i' E 1000 l i , { // ! 8 :HEATUP RATES ~ x L r
/
/
/i _,. ACCEPTABLE UP 10 /
%v . 750- 100'F/HR '
N/ / i r OPERATION
-9 ,
/ -l -.* i i ,
i 500 ~
, CRITICALITY LIMIT
'*- i BASED ON INSERVICE .
, HYDROSTATIC TEST '
250 i
';. t TEMPERATURf (230*F)
.%' 1 FOR THE SERVICE -
- j. _
l i i PERIOD UP 10 16 EFPY*
,i,,,,1 , ,,1 i i 7'O . 50 iOO 150 200 250 300 350 400 450 500 C INDIC ATC0 TEWPER ATURE (DEC.F)
- Figure Bl. Reactor C0olant System Heatup Limitations Applicable for the First 16 EFPY B-13
MATFRIAL PROPERTY MSIS CONTROLLING MATERIAL: LOWER SHE ATE R1108-1 (UNIT 1) , INTERMEDIATE SHELL PLATE R3807-2 (UNIT 2) INITIAL RTNDT: 0*F (UNIT 1), 10*F (UNIT 2) , ART AT _16 EFPY: 1/4T : 84'F (UNIT 1), 8.'F (UNIT 2) 3/4T : 69'F (UNIT 1), 62*F (UNIT 2) . CURVES BOUNDING COMANCHE PEAK UNITS 1 AND 2. APPLICABLE FOR C00LDOWN RATES UP TO 100*F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY, CONTAINS MARGINS OF 10'F AND - 110 PSIG FOR POSSIBLE INSTRUMENTATION ERRORS. 2500 7g372 ; 7 ,;._4 ) i [_ ;- ! a
' l d-y
{ 2250 i 4 ,
-]
2000 / ,
-a - - UNACCEPTABLE / ,
i 9 OPERATION 1 1750 jj j {
~
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~ '
G 1500 /j-' i i 4 ! E
/t i g 1250 [ l ,
p,
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.7
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2 750 ~_ COOLDOWN RATES b~ i
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-v F/ R ,
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.'_d ,
{,- . 1 4 j
-A- ! i :' It - !'
U o 50 '00 150 200 250 soo 350 40$ 450 500 INDIC ATEo TEMPER ATURE (CEG,F) Figure B2. Reactor Coolant System Cooldown Liinitations Applicable for the First 16 EFPY B-14
MATERIAt PROPERTY BASIS CONTROLLING MATERIAL: LOWER SilELL PLATE R1108-1 (UNIT 1) INITIAL RTNOT: 0*F 1
, ART AT 32 EFPY: 1/4T : 93'F 3/4T : 79'F
. HEATUP CURVES FOR COMANCHE PEAK UNIT 1. APPLICABLE FOR HFATUP RATES UP TO
, 100'F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. CONTAINS MARGINS OF 10*F AND
.110 PSIG FOR POSSIBLE INSTRUMENTATION ERRORS.
2500 Uw;mi ii iiii iii j ii LEAK TEST LIMIT ~~T #I. _ _ -. . J . 2250 i I
! . i li - - 1 IIT ; RITICALITY LIMIT !
, CORRESPONDING TO .
UNACCEPTABLE ' '[ /I / *--r- 20'F/HR HEATUP RATE [.4__q~
+
OPERATION i >L 2000 j.2_J. ; M$0 i{- Il$f 1 J i
- f r: 1
- 4 CRITICALITY LIMIT ~"
, 4 HEATUPRATESs " i i-11.Jg-- _'_ CORRESPONDING 10 P~bh If-fm - 60'F/HR HEATUP RATE 3 i
UP TO ,
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~
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- HEATUP RATES -d i
f L D-
/ l; f LJf--.-
f iACCEPTABLE * . l
- UP TO / /i 0 7150 : r OPERATION
- E g -100'F/HR 4 i i
x --_h/. / l ; i i e , 1 1 / ! ! i 500 ~--- ,'
;- 4 4 i , 4 CRITICALITY LIMIT- -
.. . ! BASED ON INSERVICE :
HYDROSTATIC TEST - l i l
-050 -
TEMPERATURE (240'F) ', FOR THE SERVICE .
, PERIOD UP 10 32 EFPY -
~
0 SD- 100 150 20 250 300 350 400 450 500
- - . , .. INDICATED TEWPERATURE (DEC.F)
Figure'B3. Reactor Coolant System Heatup Limitations Applicable for the First 32 EFPY B-15 x E- .
MATERIAL PROPERTY BASIS - l t I CONTROLLING MATERIAL:- LOWER SHELL PLATE R1103-1 - l INITIAL RTNDT: 0'F ! ART AT 32 EFPY:- 1/4T : 93'F- , I 3/4T : 79'F ! COOLDOWN CURVES FOR COMANCHE PEAK UNIT 1. APPLICABLE FOR C00LDOWN RATES UP TO . _ 100'F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. CONTAINS MARGINS OF 10'F AND , E 110 PSIG FOR.POSSIBLE INSTRUMENTATION ERRORS. 2500 m .,7ai i i i [
! T I
2250 1 [ i ,
/! !
2000 , 3 ; _, UNACCEPTABLE / OPERATION; i ! 1760 l ,
. i f
- i i f I i ,
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6 >
/ '
i w 1250 l , i [ l , 5 '
/ -ACCEPTABLE ,
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- d. l l l l .
- 750 : COOLDOWN RATES ,-- ,
-f ~._; 1*f/ R g ,
' ! ' i L 500 l 25 - i
- 40My a .
1
*=
i / i >
-a 60 'f i
l250 - - 100 . . - I 4 l g 0 50 _- 100 -150' 200 250 300 350- 400 450- 500
. INDICATED TEWPERATURE (DEG.F) '
i
~ '
i Figure B4.= Reactor Coolant System Cooldown Limitations Applicable fer the First 32 EFPY i B-16
- , . . . . - . ,. -. . . . , a : -, - - -~a,-_,, . -..-;-
MATERIAL PROPERTY BASIS l CONTROLLING MATERIAL: LOWER SHELL PLATE R1108-1 (UNIT 1) INITIAL RTNDT: 0'r
. ART AT 16 EFPY: 1/4T : 84'F 3/4T : 69'F
. HEAS)P CURVES FOR COMANCHE PEAK UNIT 1. APPLICABLE FOR HEATUP RATES UP TO l3 100'F/HR FOR THE SERVICE PERIOD UP TO 16 EfPY. WITHOUT MARGINS FOR POSSIBLE !
!* INSTRUMENTATION ERRORS.
- 2500 na nnc 3 ,,,,,,,,,,,,,,
r ! j [ , l CRITICALITY LIMIT - LEAK TEST LIMIT 2 ' ' ' - CORRE$PONDING TO I
%l l l aL '
2250 l 2,0 'F/HR HEATUP RATE. r r r i r i i . if I I
' ' ' ' d
~
i
-----UNACCEPTABLE-2000 ----
OPERATION
/ / j / '
) CRITICALITY Lit;IT--'---
iiI i a r i r i
-CORRESPONDING TO 1750 --
UP TO
~
x r r m Ii r i T ' ir ; r 20*F/HR
~
I [ 'i [ '"" l CRITICALITY LIMIT-i- -- G 1500 # ' #
/ J -
CORRESPONDING TO :--- j j 100'F/HR H M E --- k
'HEATUP RATES s
- UP 10 x r i r '
i r
/
w 1250 ---- 60'F/HR' i f // / 1 l { k1000 '
~
j "HEATUP RO ES x ; i~ / 0- ;UP TO x r > > ACCEPTABLE ~
' MM
- 750 100'F/HR s=
[ ' L i EM i+ Soo CRITICAll1Y LIMITS BASED ON INSERVICE
.~ HYDROSTATIC TEST 250 TEMPERATURE,(216'F)
- 1. FOR THE SERVICE PERIOD UP 1016 EFPY C iiiiiiIIIIIiII 0
O 50 100 150 -200 250 300 350 400 450 500 INDICATED TEWERATURC (DEC.F) figure B5. Reactor Coolant System Heatup Limitations (Without Margins for' Instrumentation Errors) Applicable for the First 16 EFPY B-1/ E
i 1 MATERIAL PROPERTY BASIS l CONTROLLING 11ATERI AL: LOWER SHELL PLA1E R1108-1 (UNIT 1) . INITIAL RTNOT: 0'F ART AT 16 EFPY: 1/4T : 04'F , 3/4T : 69'F COOLDOWN CURVES FOR CDMANCHE PEAK UNIT 1. APPLICABLE FOR C00LDOWN RATES UP TO .' 100*F/HR FOR THE SERV' .. PERIOD UP 10 16 SFPY. WITHOUT MARGINS FOR POSSIBLE , INSTRUMENTATION ERRORS. IU 1 blic i i i i
! l 8 J i i l ,
# i ! '
2250 !
~' -
' UNACCEPTABLE ! /
_+ _ SERATION l [ , 2000 i l 1750 / ( . y 1500 [ j g w , j W 1250 3 / N r ACCEPTABLE-OPERATION g a i RATES i 5 :c00 p 2 750
~
- 0
! :~:: 20
_ _. 40~ -J / 00 -
;;;;} ,
~
250 y -
.:p -
9
, jg i J -
4 Q:.~:: 0 50 100 150 200 250 3(i0 350 ' 40C, 450 500 INDICAftD TEWPERATURE (.DCC.rt , Figure B6. Reactor Coolant System Cooldown Limilithns (Without Margins for Instrumentation Errers) Applicable for the First 16 EFPY B-18 N
MATERIAL PROPERTY 2 ASIS \I
\ j; . CONTROLLING MATERIAL: LOWER SHELL PLATE R1108-1 (UNIT 1) t e i 0'F INITIAL RTNOT:
j ART AT 32 EFPY: 1/4T : 93'T 3/4T : 79'F APPLICABLE FOR HEATUP RATES UP TO
. HEATUP CURVES FOR COMANCHE PEAK UNIT 1.
,! 100*F/HRFORTHESERVICEPERIODUPTO32EFPY, WITHOUT MARGINS FOR POSSIBLE i
INSTRUMENTATION ERRORS. 2500 7nn m , , , ,, l 1 I i r . r , i r . LEAK TEST LIMIT v ij j j jf ; r CRiflCAlll Y LlPIIl~~~~ a 1 i i; i ; _. CORRESPONDING TO l / l ,'l f jc'. 20 'F/HR HEA1P RATE: 2000 UNACCEPTABLE
- 0PERAil0N .
/ i
/
i j. i, jQ l i , 1 I il' f / C'RITILAL(1Y Llil.T --- H[hiUP RATES ' - - CORRESPONDING TO 1750 UP TO , J / /, / j - 60'F/HR HEATUP RATC::
, 20*FjHR l ll l , ,
^2 1500 1 I ' I i- CRITICALITY LIMIT ----
HEATUP RA1ES
' 2 ' ' ' ' ' CORRESPONDING TO E
~~
- UP TO \ / -
yjj [ / 10,0*FfHR HEATUP RATE::::
~~~~ '
1250 ' ' g I F / .' 1
$ i i f i f l l 'l l o
1000 HF41UPRaTEh . _.
.r
/ ;
/ ,
/ ACCEPTABLE OPERATION w _ U. 10 % e / .r 100'F/HR \ ,
g y
"- 500 CRl11 Call 1Y LIMlls - ---
~ BASED ON INSERVICE 250 JHYDR0 STATIC TEST --::
TEHPERATURE (226*F) ---- FOR THE SERVICE ZI;E q00,U,P ,T,0,3,2, EF,PK:-"" O I I I IIil l'I i I i 0 50 100 150- 200 250 3'JO 350 400 450 500 INDICATED TCWPCRATURE (DCG.r) Figure 87. Reactor Coolant System Heatup Limitations
.(Without Margins for Instrumentation Errors)
Applicable for the first 32 EFPY B-19 q-ng -----ve w .g-
~
MATERIAL PROPEkTY BASIS CONTROLLING MATERIAL: LOWER SHELL PLATE R1108-1 (UNIT 1) . INITIAL RTHDT: O'F ART AT 32 EFPY: 1/4T : 93'F . 3/4T : 79'F COOLDOWN CURVES FOR COMANCHE PEAK UNIT 1. APPLICABLE FOR COOLDOWN RATES UP TO , 100'F/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. WITHOUT MARGINS FOR POSSIBLE , INSTRUMENTATION ERRORS. 2500; y i l . 2250
-UNACCEP1ABLE / -l ;
OPERATION f , I e , 2000 .f i l l . 1750 ) 1 j 1500 n > .1 . b / ACCEPTABLE W 1250 i OPERATION - E / E / w ' s E 1000 8
~
h 750 - g 2
- U$$'
-. 0- '-
~~~
~'- go ,, ,
500 : 40-7 s '
----- 60' ,
----_gne- ,
250 i U O 50 100 150 200 250 30$ 350 400 450 500 INDICATED TEWPERATURE (DEG.7)
. Figure 88. Reactor Coolant System Cooldown Limitations (Without Margins for Instrumentation Frrors)
Acplicable for the First 32 EFPY B-20
B6. REFERENCES
. B1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May,1988.
I 82 " Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of 9,fety
- Analysis Reoorts for Nuclear Power Plants, LWR Edition, NUREG-0800, l
. 1981.
83 ASME Boiler and Pressure Vessel Code, Section Ill, Division 1 - i Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G. Protection Against Honductile failure", pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986. B4 Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No.104, May 27,1983.
=B5 WCAP-9475, " Texas Utilities Comanche Peak Unit No. 1 Reactor Vessel Radiation Surveillance Program," W. T. Kaiset, et al., April 1979. ,
(Westinghouse Proprietary Class 111) 4 e r 9 e 1 B-21
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. *t T8X HEATUP CURfES REG.' GUIDE 1<99.REV.2 WITH MARGIN 09/30/92 COMPOSITE CURVE Pt.0TTED FOR HEATuf'.' PROFILE 4 HEATUP QATE(S) (CEG.F/HR) = 100 O' IRRADIATION PERIDD = 16.000 EFP YEARS FLAW DEP,TH * ( 1- A0 WIN)T INDICATED' INDICATED INDICATED INDICATED.- IMDiCATED INDICATED TEMPERATURE PRESSURE TEMPERATURE- PRESSURE TEMPERATURE PRESSURE-(DEG.F ) (PSI) (DEG.E) (PSI)
(DEG.F) (PSI) 1 85.000 3.r94-58 16 160.000' See-1919 ,g 30 230.000 - 1074.56
-2 90.000 33tMlrt 17' 165.000' Set-96) 3 - 31 235.000 '1136.92 3 95.000 -) 3 a ' 18 170.000' 606.04- 32 240.000 '1203.86 4 100,000 SAG-96 d33g .19 175.000 629 41 33 245.000 1275.'76 .
5 105.000 %G 'M g, 20 ' 180.000 G55.99 34 250.000 1352.67 6 110.000 497--94 *- 21: 185.000 683.23 35 255.000 1435.58. 7 115.000 499-94 22- .190.000 714.21 . 36 260.000 1524.18 8 120.000 . 493.06 23 ~ 195.000 747.78 37 265.000 1G19.09 9 125.000 494.64 .24 -200.000 '784.22 38 270.000 1720.58 10 130.000 498,43 25 205.000 823.76: 39 275.000 .1829.02 11 135.000 504.47 26 ~ 210.000 866.75 40 280.000 1945.09 12 140.000 M 27 215.000. 913.02 41 285.000 20G8.sr 13 145.000 28 220,000 962.92 42 290.000 2201.01 14 150.000 W 599-+9 ggg Ij 29. '225.000 1016.5F 43 295.000. 2341.67 15 155,000 _. U3 4 t E Ir
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TBX'COOLDOWN CURVES REG. GUIDE 1.99.REV.2 WITH MARGIN 09/30/92 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOwN PROFILE 2 z ('20 DEG-F /-HR COOLDOWN ) IRRADIATION PERIDO = 16.000 EFP VEARS FLAW DEPTH = AOWIN'T INDICATED INDICATED INDICATED - INDICATED INDICATED INDICATED TEMPERATURE ' PRESSURE TEMPERATURE PRESSURE ' TEMPERATURE FRESSURE-(DEG.F) '(PSI) (DEG.F) (PSI) (DEG.F) (PSI) 1 85.000 ' 499.71 - 7~ 115.000' 1H)f-twr] ~13 145.000 7taf-26' 8 120.000 ese-s4 14 150.000 EH+-4* 2 90.000 644-G4 ' . 3 95.000 53t7 87 9 125.000 655-09 , 15 155.000 - e49-97$611{6t-4 100.000 - ?47-83 50 ps;. 10 't30.000 m 5" # 16 1s0.000 e 5 105.000 666 11 135.000 J l A gu 17 165.000- 005. P 6 .110.000 See-ee, l12 140.000~ M. 18 170.000 982.Gb m s -a I G m k i
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4 l TBX COOT.DOWN CURVES TEG. CUIDE 1.99.CEV 2 WITH CARGIN - 09/30/92
~
THE FOLLOW!NG DATA WERE PLOTTED FOR COOLDOWN' PROFILE 3' ( 40 DEG-F,/ HR COOLDOWN )
' IRRADIATION PERIOD =' 16.000 EFP YEARS
;. FLAW DEP7H ==A0 WIN T
' INDICATED INDICATED- INDICATED ' INDICATED INDICATED ' INDICATED'-
~ TEMPERATURE PRESSURE' ' TEMPERATURE. PRESSURE TEMPERATU F ' PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) 1 85.000 464.75 7- .115.000 50G,@4' 13 145.000 2 90.000 460.59 '8 120.000- en .14 '. -150.000 M 3 95.000 497,58 9 125.000 est-iH g 15 155.000 M 6II N+
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'T8*. COOLDowN CURVES REG.' GUIDE /1.99.REV.2 WITH MARGIN 09/30/92
- THE'.FOLLOWING DATA WEkE PLOTTED FOR COOLDOWN PROFILE 4 ( 60 DEG-F / HR COOLDOWN l' IRRADIATION PERIOD = 16.000 EFP YEARS FLAW OEPTH = ADWIN T INDICATED' INDICATED' INDICATED INDICATED' INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) ~ (PSI) (DEG.F). (PSI) 1 85.000 429.70. 7 '115.000 Stt-ef 13- 145.000 ?*t-*4 2 90,000- 446.49 '8 120.000 M 14 . 150.000 766-99 ,
3 95.000 464.74 9 -125.000 697-91 15 155.000 09+-* I 61if3.- 4 100.000. 484.27 10 130.000 600-56 0 .' 16 - 160.000 ._s
-5 105.000 505.50 - 11 135.000 &*+-76 17 ~ 165.000 02?.M 6' 110.000' N 5tig 12 L140.000 W, 18. 170.000 986.16 m
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' THE FOLLOW!NG DATA WERE CAtCULATEDFOR THE INSERVICE HYOROSTATIC' LEAK TEST, _
MINIMUM INSERVICE' LEAK TEST *EMPERATURE-( 32.OOO,EFPV) PRESSURE (' PSI)._ TEMPERATURE.(DEG.F)
- 2000 220 2485 240.
' PRES 5URE PRES 5URE STRESS . . 1.5 Ktu (PSI) (PSI) (PSI SO.RT.IN.)
CD 2000 22304 94153 w . . C3 2485 27431 116854 d I
P' 09/30/92 TBX HEATUP CURVES REG.' GUIDE 1.99,REV.2 tdITH MARGIN HEATUP RATE (S) (DEG,F/HR) = 20.0 COMPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2 1RRADIAT!DN PERIOD = 32.000 EFP YEARS FL AW DEPTH = (1- ADWIN)T INDICATED INDICATED INDICATED ) ItJOIC A T ED INDICATEC. INDICATED TEMPERATURE PRESSURE ' TEMPERATURE PRESSURE TEMPERATURE PRESSURE (PSI) (DEG.F) (PSI) (DEG.F) (CEG.F ) . (PSI)
"3' 0' 26 210.000 1316.47 ;
85.000 510.03 14 150.000 . 1334.51 i 15 155.000 6 9-49 27 215.000 ' 2 90.000 510.86 28 220.000 1457.34 3 95.000 444-99' 16 160.000 994 34lS W[' 29 225.000 7535.51 4 100,000 CIC 73 17 165.000 4 4@-04J 1619.07 18 170.000 879.54 30 230.000 5 105.000 404-49 31 235.000 1708.78 M 9-69 19 175.000 921.70 1P04.67 6 110.000 180.000 966.95 32 '240 000 G4+-ee 20 , 7 115.000 1015.61' 33 245.000 1907.35 8- 120.000 +04-6+ >6tf 21' 185.000 1067.83 34 250.000 2016.90 125.000 440 22 190.000 9 195.000 f123.97 35 255 000 2134.40 10 130.000 521 "' , 23 36 260.000 2259.67 24 200.000 1984.12 11 .135.000 'C" C 1248;62 37 265.000 2392.96 12 140.000 fre+-e+ 25 205.000 13- 145.000 '"" CO, C3 w W k k. g 6 $ 8 O g
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l. 4 ATTACHMENT B2 DATA POINTS FOR HEATUP AND C00LDOWN CURVES (With No Margins for Instrumentation Errors) m G-le 4 l
07/06/92-TBX HEATUP CURVES REG. GUIDE.1.99.REV.2 41THOUT MARGIN THE FOLLOWING DATA WERE CALCULATEOFOR THE INSERVICE HYOROSTATIC LEAK TEST. I i MINIstUM INSERVICE LEAK TEST TEMPERATURE ( 16.000 EFPV) ! PRESSURE (PSI) TEMPERATURE (DEG.F) 2000 195 2485 216
.. I PRESSURE PRESSURE STRESS 1.5 KtM (PSI) (PSI) (PSI SQ.RT.IN.)
2000 21142 89085 m 2485 26268 111649
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-T8X' HEATUP CURVES ' REG. : GUIDE ~ t .99,REV.2 'WITtiOUT MARGIN L :07/06/92 -
COMPOSITE CURVE PLOTTED FOR HEATUP-PROFILE 2 HE ATUP - R AT E( 5 ) - (DEG . F/HR ) = " 20.O '
..-IRRAOIATION PERIOD'=.'16.000 EFP YEARS <
1 c l' LAW DEPTH = = ( t-AOWIN)T. . INDICATED' : INDICATED INDICATEb INDICATED' INDICATED- -INDICATED-TEMPERATURE ' PRESSURE - 'TE84PERATURE1 PRESSURE ^TEasPERATURE . PRESSURE-
'(DEG.F)" ' (PSI); .
.(DEG.F)-
~(PSI): -(DEG.F) (PSI)'-
12 - 140.000 $99 23 .'195.000 - 1490.78' t .'86.000' 464,44
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'205.000' 1641.28-100.000- tie-04 15 155.000- +99+-96 26' 210.000- 1724.55
- 5 105.000'--
6 ./16 c160.000 1096.31 ... 27 - 215.000 ~ 1813.98~ ' fG, "110.0004 'Petree >@l F" 17 L ' h185.000 , 1143.98'~ '28l '220.000) 1909.54 ,
't194.68z ; 29 -- 225.000 , , 2011'.90.'
7' L115.000: ??9-16 > 181 ?170.000- . 188 '120.000 6: -907;-13 L191 C-175.000; 11245.38 ; 30 '; J230.OOO_~ i.' 2121.24 J . 49+-46 20. 1180.000 1300.83 L31 -235.000' 2238.21
- 9. '125.000 2362.84-10- :130.000 - 464 44 '21 185.000 1359.84. '32 .240.000' 11- 135.000 494-96 22 190.000., '1422.88 -
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'07/06/92L 3 s
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. , y (DEG.F), .
~(PSI) - (D. EG . F J -(PS.I) 1- 85.000 ' 990-99 13 14 *a . OOO . 990-94 .
205.000- ~1300.08
- 2: .90.000H 99t-95 ' 14 . '150.000- 99&rG4 WBt %. 26 25 210.000 1375.25 3 . 95.000- 990--te 15 155.000. 99+4+ -
27.' 215.000 1449.54 4: 100.000. 499,46- 16 - 160.000 864.57 '28 220.000 1529.11 5' 105.000- 1>99 -- 17 ' 165.000 ^900.16 '225.000 1614.39 06' 1110.000. 440,49 1170 % ' . 939.63'- - - 30 , .230.000- -1705.63-
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-115.000.1 4HWie9 P'h c20;k' ' 18 : l.17! W & !980.22? 31 ~235.000-' * -1803.44
- 8- (120.000i ' S?1 ?? - i180.000." 1025125i e 321 S.240.000. 1907.98 9- ~125.000 -9ee-te- .21 165.000 1073.60- 33' 245.000 2019.67
-10 130.000: W .2^ "22 .190.000: 1125.42 34 :250.000 2139.05 11- 135.000- 9f9-5e -23 -195.000 1181.42 35' 255,000 2266.03.
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[1 35.000 997-9e ' i1S ? I155.OOOC -404,44 W k..' L29_ 226.000l E1246.92- , o- 160.000~ 718.04._ 30 - - 230.000- 1313296 '1 ' -
~2- 90.000 922:08 4 3 95.000 ett-91 ~'&da.M.x 17 16 ; 1s5.000 :73s.4i as- '235.00o iss5.7s 4- 100.000- F 18 170.000' 765.09 -32 240.000 1442.67
'5. 't05,000- 400-04 ' 19 . 175.000' 793.23- 33 .245.000 .-1545.58-L6^ "110.000i '603.06u -
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- 9. 125.000- -614-47
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^ t 1" 135.000 699-94 .'25.. 205.000 1023.02 39 .275.000 -2178.96 '
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.THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILEL1- .( STEADV-STATE COOLDOWN T)
IRRADIATION PERIOD =:'16,000 EFP YEARS' ~ s iPLAW OEPTH = AOWIN Ti' , - IND N TED- INDICATED-- . 'INOICATED - INDICATED.'- INDICATEC' INDICATED-TEMPD..iTURE R PRESSURE. ' TEMPERATURE. ' PRESSURE -TEMPERATLcE- PRESSURE ,.
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