ML20138R174
| ML20138R174 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/30/1985 |
| From: | Boggs R, Fero A, Kaiser W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20138R167 | List: |
| References | |
| WCAP-10814, NUDOCS 8511180477 | |
| Download: ML20138R174 (96) | |
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{{#Wiki_filter:_ s-y WCAP-10814 WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION ANALYSIS OF CAPSULE U FROM THE SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM R. S. Boggs A. H. Fero W.T. Kaiser , June 1985 Work performed under Shop Order No. UCGJ-27302 APPROVED: N #jk T. A. Meyer, Manager Structural Materials and Reliability Technology Prepared by Westinghouse for the South Carolina Electric and Gas Company Although.information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval. WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 l Pittsburgh, Pennsylvania 15230 l 8511180477 851100 yDR ADOCK 05000395 PDR I 8455B:1b-061985 4
r PREFACE This report has been technically reviewed and verified. Reviewer
. Sections 1 through 5 and 7 S. E. Yanichko '/ha L[Lc Section 6 S. L. Anderson d.[. b oom m 6/A118s
. Appendix A W. K. Ma n v7 7 ///f ,) 5 '?ie l
8455B:1b-061985 iii
LIST OF ILLUSTRATIONS (Cont) . Figure Title Page 5-10 Tensile Properties for V. C. Suirmer Unit 1 Reactor 5-25 Vessel Intermediate Shell Plate A9154-1 (Longitudinal)' 5-11 Tensile Properties for V. C. Summer Unit 1 Reactor 5-26 Vessel Intermediate Shell Plate A9154-1 (Tangential) 5-12 Tensile Properties for V. C. Summer Unit 1 Reactor 5-27 Vessel Weld Metal 5-13 Fractured Tensile Specimens of the V. C. Summer 5-28 Unit 1 Reactor Vessel Intermediate Shell Plate A9154-1 (Longitudinal Orientation) 5-14 Fractured Tensile Specimens of the V. C. Summer 5-29 Unit 1 Reactor Vessel Intermediate Shell Plate A9154-1 (Transverse Orientation) 5-15 Fractured Tensile Specimens of the V. C. Summer 5-30 Unit 1 Reactor Vessel Weld Metal 5-16 Typical Stress-Strain curve for Tension Scecimens 5-31 6-1 V. C. Summer Reactor Geometry 6-25 6-2 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-26 6-3 Calculated Azimuthal Distribution of Maximum 6-27 Fast (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel Surveillance Capsule Geometry 6-4 Calculated Radial Distribution of Maximum Fast 6-28 (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel 6-5 Relative Axial Variation of Fast-Neutron Flux 6-29 (E > 1.0 MeV) Within the Reactor Vessel I A-1 Effect of Fluence and Cepper on Shift of RT A-8 NDT for Reactor Vessel Steels Exposed to Irradiation at 550*F i A-2 Fast Neutron Fluence (E > 1.0 MeV) as a Function A-9 l of Full Power Service Life (EFPY) l A-3 V. C. Summer Unit 1 Reactor Coolant System A-10 Heatup Limitations Applicable up to 8 EFPY A-4 V. C. Summer Unit 1 Reactor Coolant System A-11 Cooldown Limitations Applicable up to 8 EFPY 8455B:1b-061985 viii
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LIST OF TABLES (Cent) , Table Title Page 6-3 Calculated Fast Neutron Exposure Parameters for the 6-13 Peak Location of the V. C. Summer Reactor Vessel 6-4 Calculated Fast Neutron Exposure Parameters and Lead 6-14 Factors for the V. C. Summer Surveillance Capsules 6-5 Calculated Fast Neutron Exposure Parameters for 6-15 V. C. Summer Surveillance Capsule U Metallurgical Specimens 6-6 Calculated Neutron Energy Spectrum at the Center of 6-16 V. C. Summer Surveillance Capsule U 6-7 Spectrum-Averaged Reaction Cross Sections at the 6-17 Center of V. C. Summer Surveillance Capsule U 6-8 Irradiation History of V. C. Summer Surveillance 6-18 Capsule U 6-9 Comparison of Measured and Calculated Radiometric 6-19 Monitor Saturated Activities for V. C. Summer Surveillance Capsule U 6-10 Results of Fast Neutron Dosimetry for V. C. Summer 6-22 Surveillance Capsule U 6-11 Product Nuclide Burnout Assessment for V. C. Summer 6-23 Surveillance Capsule U 6-12 Summary of V. C. Summer Fast Neutron Fluence Results 6-24 Based Upen Surveillance Capsule U A-1 Reactor Vessel icughness Data (Unirradiated) A-7 e 8455E:1b-071685 x
TABLE OF CONTENTS Section Title Page 1
SUMMARY
OF RESULTS 1-1 2 INTRODUCTION 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PP0 GRAM 4-1 5- TESTING OF SPECIMENS FROM CAPSULE U 5-1 5-1. Overview 5-1 5-2. Charpy V-Notch Impact Test Results 5-3 5-3. Tension Test Results 5-4 5-4. Wedge Opening Loading Tests 5-5 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 6-1. Introduction 6-1
. 6-2. Discrete Ordinates Analysis 6-1 6-3. Radiometric Monitors 6-4 6-4. Neutron Transport Analysis Results 6-8 6-5. . Dosimetry Results 6-9
- 7. SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8 REFERENCES 8-1 84558:1b-061985 v I
-=. 3
. TABLE OF CONTENTS (Cont)
Section Title Page - Appendix HEATUP AND C00LDOWN LIMIT CURVES FOR A-1 A NORMAL OPERATION A-1. Introduction A-1 A-2. Fracture Toughness Properties A-2 A-3. Criteria For Allowable Pressure-Temperature A-2 Relationships A-4. Heatup and Cooldown Limit Curves , A-5 84558:1b-061385 vi
LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the 4-5 V. C. Summer Unit 1 Reactor Vessel (Updated Lead Factors for Capsules Shown in Parentheses) 4-2 Capsule U Diagram Showing Location of Specimens, 4-6 Thermal Monitors, and Dosimeters 5-1 Irradiated Charpy V-Notch Impact Properties for 5-16 V. C. Summer Unit 1 Reactor Vessel Intermediate Shell Plate A9154-1 (Longitudinal Orientation) 5-2 Irradiated Charpy V-Notch Impact Properties for 5-17 V. C. Summer Unit 1 Reactor Pressure Vessel' Intermediate Shell Plate A9154-1 (Transverse Orientation) 5-3 Irradiated Charpy V-Notch Impact Properties for 5-18 V. C. Summer Unit 1 Reactor Pressure Vessel Weld Metal 5-4 Irradiated Charpy V-Notch Impact Properties for 5-19 V. C. Summer Unit 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for V. C. Summer 5-20 Unit 1 Reactor Pressure Vessel Intermediate Shell Plate A9154-1 (Longitudinal Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for V. C. Summer 5-21 Unit 1 Reactor Pressure Vessel Intermediate Shell Plate A9154-1 (Transverse Orientation) 5 Charpy Impact Specimen Fracture Surfaces for 5-22 V. C. Summer Uni.t 1 Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for 5-23 V. C. Summer Unit 1 Weld Heat Affected Zone Metal 5-9 Comparison of Actual versus Predicted 30 f t Ib 5-24 l Transition Temperature Increases for the V. C. Summer Unit 1 Reactor Vessel Material based on the Prediction Methods of Regulatory Guide 1.99 Revision 1 8455B:1b-061985 vii
l
. )'
l l LIST OF ILLUSTRATIONS (Cont) I 'I Figure Title Page 5-10 Tensile Properties for V. C. Summer Unit 1 Reactor 5-25 Vessel Intermediate Shell Plate A9154-1 (Longitudinal) L 5-11 Tensile Properties for V. C. Summer Unit 1 Reactor 5-26 Vessel Intermediate Shell Plate A9154-1 (Tangential) 5-12 Tensile Properties for V. C. Summer Unit 1 Reactor 5-27 j Vessel Wald Metal L 5-13 Fractured Tensile Specimens of the V. C. Summer 5-28 Unit 1 Reactor Vessel Intermediate Shell Plate A9154-1 (Longitudinal Orientation) 5-14 Fractured Tensile Specimens of the V. C. Summer 5-29 Unit 1 Reactor Vessel Intermediate Shell Plate A9154-1 (Transverse Orientation) ! 5-15 Fractured Tensile Specimens of the V. C. Summer 5-30 l l Unit 1 Reactor Vessel Weld Metal 5-16 Typical Stress-Strain Curve for Tension Specimens 5-31 ! 6-1 V. C. Summer Reactor Geometry 6-25 6-2 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-26 6-3 Calculated Azimuthal Distribution of Maximum 6-27 Fast (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel Surveillance Capsule Geometry 6-4 Calculated Radial Distribution of Maximum Fast 6-28 (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel 6-5 Relative Axial Variation of Fast-Neutron Flux 6-29 , (E > 1.0 MeV) Within the Reactor Vessel A-1 Effect of Fluence and Copper on Shift of RT NDT A-8 for Reactor Vessel Steels Exposed to Irradiation at 550*F A-2 Fast Neutron Fluence (E > 1.0 MeV) as a Function A-9 of Full Power Service Life (EFPY) A-3 V. C. Summer Unit 1 Reactor Coolant System A-10 Heatup Limitations Applicable up to 8 EFPY A-4 V. C. Summer Unit 1 Reactor Coolant System A-11 Cooldown Limitations Applicable up to 8 EFPY 84558:1b-061985 viii
p LIST OF TABLES Table Title Page 4-1 Chemical Composition of the V. C. Summer Unit 1 4-3 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the V. C. Summer Unit 1 4-4 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 5-6 Intermediate Shell Plate A9154-1 Irradiated (Longitudinal) at 550*F, Fluence 6.39 x 10' n/cm 2(E > 1 MeV) 5-2 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 5-7 Intermediate Shell Plate A9154-1 (Tangential) Irradiated at 550*F, Fluence 6.39 x 10' n/cm 2(E > 1 MeV) 5-3 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 5-8 Pressure Vessel Weld Metal Irradiated at 550*F, Fluence 6.39 x 105' n/cm2 (E > 1 MeV) , 5-4 Charpy V-Notch Impact Data for the V. C. Summer Unit 1 5-9 Pressure Vessel Weld Heat Affected Zone Metal Irradiated at 550*F, Fluence 6.39 x 10" n/cm 2 (E > 1 MeV) 5-5 Instrumented Charpy Impact Test Results for V. C. Summer 5-10 Unit 1 Intermediate Shell Plate A9154-1 (Longitudinal Orientation) 5-6 Instrumented Charpy Impact Test Results for V. C. Summer 5-11 Unit 1 Intermediate Shell Plate A9154-1 (Transverse Orientation) 5-7 Instrumented Charpy Impact Test Results for 5-12 V. C. Summer Unit 1 Weld Metal 5-8 Instrumented Charpy Impact Test Results for 5-13 V. C. Summer Unit 1 Weld Heat Affected Zone Metal 5-9 The Effect of 550"F Irradiation at 6.39 x 10" 5-14 (E > 1 MeV) on the Notch Toughness Properties of The V. C. Summer Unit 1 Reactor Vessel Materials 5-10 Tensile Properties for V. C. Summer Unit 1 Reactor Vessel 5-15 Material Irradiated to 6.39 x 10' n/cm2 6-1 SAILOR 47 Neutron Energy Group Structure 6-11 6-2 Nuclear Constants for Radiometric Monitors Contained 6-12 in the V. C. Summer Surveillance Capsules 8455B:1b-061985 ix
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LIST OF TABLES (Cont)
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D Table Title Page 6-3 Calculated Fast Neutron Exposure Parameters for the 6-13 Peak Location of the V. C. Summer Reactor Vessel 6-4 Calculated Fast Neutron Exposure Parameters and Lead 6-14 Factors for the V. C. Summer Surveillance Capsules 6-5 Calculated Fast Neutron Exposure Parameters for 6-15
. V. C. Summer Surveillance Capsule U Metallurgical Specimens 6-6 Calculated Neutron Energy Spectrum at the Center of 6-16 V. C. Summer Surveillance Capsule U 6-7 Spectrum-Averaged Reaction Cross Sections at the 6-17 Center of V. C. Summer Surveillance Capsule U 6-8 Irradiation History of V. C. Summer Surveillance 6-18 Capsule U 6-9 Comparison of Measured and Calculated Radiometric 6-19 Monitor Saturated Activities for V. C. Summer Surveillance Capsule U 6-10 Results of Fast Neutron Dosimetry for V. C. Summer 6-22 Surveillance Capsule U 6-11 Product Nuclide Burnout Assessment for V. C. Summer 6-23 Surveillance Capsule U 6-12' Summary of V. C. Summer Fast Neutron Fluence Results 6-24 Based Upon Surveillance Capsule U A-1 Reactor Vessel Toughness Data (Unirradiated) A-7 l
8455B:1b-071685 x
SECTION 1 SUM 4ARY OF RESULTS The analysis of the reactor vessel material contained in Capsule V, the first surveillance capsule to be removed from the South Carolina Electric and Gas Company V. C. Summer Unit I reactor pressure vessel, led to the following conclusions: o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 6.39 x 10 18 n/cm2 , o Irradiation of the reactor vessel intermediate shell plate A9154-1, to 6.39 x 10 18 n/cm, resulted in 30 and 50 f t-lb transition temperature increases of 40*F and 45*F, respectively for specimens oriented parallel to the major working direction (longitudinal orientation) and increases of 30*F and 35'F, respectively for specimens oriented normal to the major working direction (trans' verse orientation). o Weld metal irradiated to 6.39 x 10 18 n/cm 2 resulted in both a 30 and 50 ft-lb transition temperature increase of 30*F. o The average upper shelf energy of the plate A9154-1 decreased from 130 to 113 ft-lbs and the limiting weld metal decreased from 104 to 75 ft-lbs. Both materials exhibit a more than adequate shelf level for continued safe plant operation, o Comparison of the 30 ft-lb transition temperature increases for the V. C. Summer Unit 1 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1, shows that the plate material and weid metal transition temperature increase were less than predicted or that the embrittlement was less than predicted. 8455B:1b-070385 1-1
SECTION 2 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 South Carolina Electric and Gas Company V. C. Summer Unit i reactor pressure vessel materials under actual operating conditions. The surveillance program for the South Carolina Electric and Gas Company V. C. Summer Unit I reactor pressure ves'sel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and.the preirradiation mechanical properties of the reactor vessel materials are presented by Davidson and Yanichko.III The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-73, " Recommended Practice for Surveillance Tests for Nuclear Reactors".[2] Westinghouse Nuclear Energy Systems personnel were contracted to aid in the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and , Development Laboratory, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed. This report summarizes testing and the postirradiation data obtained from surveillance Capsule U removed from the South Carolina Electric-and Gas Company V. C. Summer Unit i reactor vessel and discusses the analysis of the data. 84558:1b-091085 2.1
SECTION 3 , BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects cf fast neutron irradiation on the mechanical propertie: of low alloy ferritic pressure vessel steels such as SA533 Grade B Class 1 plate (base material of the V. C. Summer Unit I reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation. A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)* s RT NDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 ft 1b (and 35 mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. i The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The K curve is a lower bound of IR dynamic, crack arrest, and static fracture toughness results obtained from ' several heats of p'ressure vessel steel. When a given material is indexed to 84558:1b-012985 3-1
F' {
- the KIR curve, allowable stress intensity factors can be cbtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects'of radiation on the reactor vess ' material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the V. C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program,Ill in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT f r radiation embrittlement. This adjusted RT NDT (RT NDT initial + ARTNDT) is used to index the material to the KIR curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials. d 8455B:1b-012985 3-2
i SECTION 4 DESCRIPTION OF PROGRAM s Six surveillance capsules for monitoring the effects of neutron exposure on tho'V. C. Summer Unit I reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the neutron shielding 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. r Capsule'U was removed after 1.12 effective full power years of plant
. op'eration. This capsule contained Charpy V-notch impact, tensile, and CT
, , specimens (figure 4,-2) from the intermediate shell plate A9154-1 and submerged-
^
arc weld metal _ representative of the beltline weld seams of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) material.- All heat-affected zone specimens were obtained from within the HAZ of plate A9154-1 of th's representative weld. The chemistry and heat treatment of the surveillance material are presented in table 4-1 and table 4-2, respectively. The chemical analyses reported in table 4-1 were obtained from unirradiated material used.in the surveillance program. In addition, a chemical analysis was performed on an irradiated Charpy specimen from the weld metal and is reported in table 4-1. All test specimens were machined from the 1/4 thickness location of the" ~ plate. Test specimens represent material taken at least one plate thickness from the quenched.end of the plate. Base metal Charpy V-notch impact specimens were oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation). Charpy V-notch and tensile specimens from the weld metal were oriented with the 84558:1b-012985 4-1
,. y t' -
J e, j' . 4
, longitudinal axis of the specimens transverse to the welding direction. The CT specimens in Capsule U were machined such that the simulated crack in the
! [ specimen would propagate normal and parallel to the major working direction
; for the plate specimen and parallel to the weld direction.
Capsule U contained dosimeter wires of pure iron, copper, nickel, and
/ aluminum-cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were contained in the capsule.
I i Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule and were located as shown in Figure 4-2. .i The two eutectic alloys and their melting points are:
- 2.5% Ag, 97.5% Pb Melting Point 579'F (304*C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590*F (310*C)
The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule U are shown in Figure 4-2. i 1 ] 84558:1b-012985 4-2
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TABLE 4-1 r CHEMICAL COMPOSITION OF
/s THE V. C. SU M ER UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Plate A9154-1 Weld Metal [c]
Lukens Steel Co. Lukens Steel Co. Element Analysis Analysis C 0.22 0.085 a S 0.015 0.012/0.007(b] N u.0076 0.015 C$ 0.01t, 0.016/0.01[b] Cu 0.10 t Si 0.24 0.05/0.04[b] Mo 0.49 0.48/0.42(b] Ni 0.51 0.49/0.46(b] , Mn 1.30 0.91/0.95[b] 1.32/1.50 b] Cr 0.14/0.12[b] l- V 0.08 0.001 [a] 0.005 l P 0.009 0.013/0.009(b] Sn 0.007 0.0047 x A1 0.024 0.007/0.03(b)
'B 0.0004 0.0005 Ti '
0.0002 0.001 Pb < 0.005 0.0206 Zr 0.001 0.001 As 0.006 0.006 W < 0.01 0.01 [a] Westinghouse Analysis [b] Analysis performed on irradiated weld specimen CW14. [c] Surveillance weld was made of the same RACO INMM wire Heat #4P4784 as the beltline welds of the reactor vessel, i
.E i
8455B:1b-0129B5 4-3
TABLE 4-2 HEAT TREATMENT OF THE V. C. SIAS4ER UNIT 1 REACTOR VESSEL SURVEILLANCE MATERIALS Temperature Material (*F) Time (hr) Coolant Lower 1550*/1650* 1/2 hr/in., min Water quenched Shell Plate A9154-1 1225' + 25' 1/2 hr/in., min Air cooled 1150' 7 25' 43 Furnace cooled to 600*F Weldment 1150' + 25' 12 Furnace cooled 8455B:1b-012985 4-4
12216.lA (3.II) U O' (2.69) Z REACTOR VESSEL CORE BARREL lii I (2.69) Y I I NEUTRON PAD (3.li) x I 20' ) 17' It 4 __. 270* - -- - __ tt s, - -
> l 1 -V (3,1 )
w (2.69) l I l 180' Figure 41 Arrangement of Surveillance Capsules in the V.C. Summer Unit 1 Reactor Vessel (Updated Lead Factors for the Capsules Shown in Parentheses.) 4-5
'* Im*t IEUTRON SHIELD PAD SPECIMEN NUMBERING CODE: [ U l *I CT - PLATE A9154-1 (TRANSVERSE ORIENTATION) '
ir' + CL - PLATE A9 8 54- 1 (LONGITUDINAL ORIENTATION) CW - CORE REGION WELD METAL S; ' CH - FEAT-AFFECTED-ZONE METAL ' CAPSULE U y
/
CORE BANEL VESSEL WALL casas u g23v u2= arse ame g_r _E_N _E_N Tusn c>wev owev g-- -- _E_N _EN-ower ossev ower "atim" vusna cuerv owev oewev owev cnwev u g c.s gg-g g g g c,,, , , , g g g g g g _E_N _E_N .insats c,7 c,, c., c.2 c.i E.i sE.. sE.o E.o E., os a. as as as E E ! n. E E E E E E E cv. cvs cvs y E.. E.. av E. E. cir cri c.. c c> c m c m, c o c o a E.s ev ais crio m io en cv a cv s ai ci. & o ,, w gy aw w WW c-J gg es L-,,_.c. , c, % gg te'L - . , _ , a e,
*TW c-J gg as vv T L_,,_,,e, q'
T* H !! l
'* - bL !! ! F* -
LJLJLJ r* - r*H !: l [,JLJLJr]py_,,_,,,c,,c., [ ][}~ ,, _ ,,, c, s c, , LJLJLJ pir]py ,,_,,,,,,c,, uE.f;6 - 4 se i
! l. l. M- - . .27r6 - 4! to i gg M - Mib- 4 si a
.. = .
! l.l. M -.
aan .m. crassa ; io vtF & WESW.L M EEEM Figure 42.
% ' Capsule U Deayam
_ Thermal Morutors, Showing Location of and Doesmeters 48
l i E SECTION 5 TESTING OF SPECIMENS FROM CAPSULE U 5-1. OVERVIEW - l The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems i personnel. . Testing was performed in accordance with 10CFR50,-Appendices G and H, ASTM Specification E185-82 and Westinghouse Procedure MHL 8402, Revision 0 as modified by RMF Procedures 8102 and 8103. The mechanical test data was document'ed in an R&D report by Lott and Shogan.[3] Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9234.[1] No. discrepancies were found. Examination of the two-low-melting 304'C (579'F) and 310*C (590'F) eutectic alloys indicated'no melting of either type of thermal monitor. Based on this
= examination, the maximum temperature'to which the test specimens were exposeci was less than 304*C (579'F).
The Charpy impact tests were performed per ASTM Specification E23-82 and RMF i Procedure 8103 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instrumentation system. With this system, load-time and energy-time signals
.can be recorded in addition to the standard measurement of Charpy energy (ED ). From the load-time curve, the load of general yielding (PGY),the time to general yielding (tgy), the maximum load (Pg ), and the time to' maximum load (t y ) can be determined. Under some test conditions, a sharp 84558:1b-012985' 5-1
drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (P ),p and the load at which fast fracture terminated is identified as the arrest load (PA)* The energy at maximum load (Eg ) was determined by comparing the energy-time record and the load-time record. The energy at maxi.num 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) is calculated from the three point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also using the three point bend formula. Percentage shear was determined-from postfracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-77. The lateral e..pansion was measured using a dial gage rig similar to that shown in the same specification. Tension tests were performed on a 20,000 pound Instron, split-console test-machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test. Deflection measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67. ! l l Elevated test temperatures were obtained with a three-zone electric resistance l split-tube furnace with a 9-inch hot zone. All tests were conducted in air. l 84558:1b-012985 5-2 3
Q l 1 Because of the difficulty in remotely attaching a thermocouple directly to the
~ specimen, the following procedure was used to monitor specimen temperature.
Chromel-alumel thermocouples were inserted in shallow holes in the center and each and of the gage section of a dummy specimen and in each grip. In test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550*F (288'C). The upper grip was used to control the furnace temperature. During the actual testing the
. grip temperatures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to plus or minus 2*F. The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the
, original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5.2. CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials contained in Capsule U irradiated at 6.39 x 10 18 n/cm2 are presented in Tables 5-1 through 5-8 and Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the Capsule U material are summarized in Te.ble 5-9. Irradiation of vessel intermediate shell plate A9154-1 material (longitudinal orientation) specimens to 6.39 x 10 18 n/cm2 (Figure 5-1) resulted in both 30 and 50 f t-lb transition temperature increases of 40*F and 45'F, respectively, and an upper shelf energy decrease of 2 f t-lb. Irradiation of vessel intermedia 1ie shell plate A9154-1 material (transverse orientation) specimens to 6.39 x 10 10 m/cm2 (Figure 5-2) resulted in both 30 and 50 ft-lb transition temperature increases of 30*F and 35'F respectively. The irradiated upper shelf energy experienced no decrease as compared to the unirradiated data. 8455B:1b-012985 5-3
o 18 Weld metal irradiated to 6.39 x 10 n/cm2 (Figure 5-3) resulted in both 30 and 50 ft-lb transition temperature increases of 30*F and an upper shelf energy decrease of 4 ft-lb. 10
' Weld HAZ metal irradiated to 6.39 x 10 n/cm2 (Figure 5-4) resulted in both 30 and 50 ft-lb transition temperature increases of 30*F and 35'F, respectively, and an upper shelf energy decrease of 15 ft-lb.
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. Figure 5-9 shows a comparison of the 30 ft-lb transition temperature increases for the various V. C. Summer Unit 1 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 1.I43 This comparison shows that the transition temperature increase resulting from irra$iationto,6.39x10 18 n/cm2 is less than predicted by the Guide for plate A9154-1 (longitudinal and transverse orientation). The weld metal transition temperature increase resulting from 6.39 x 10 18 n/cm 2 is also less than the Guide prediction. 5-3. TENSION TEST RESULTS The results of tension tests perform (d on plate A9154-1 (longitudinal and 18 2 transverse orientation) and weld metal irradiated to 6.39 x 10 n/cm are shown in Table 5-10 and Figures 5-10, 5-11 and 5-12, respectively. These results shown that irradiation produced no increase in 0.2 percent yield strength for plate A9154-1 and approximately a 2 ksi increase for the weld metal. Fractured tension specimens for each of the materials are shown in Figures 5-13, 5-14 and 5-15. A typical stress-strain curve for the tension specimens is shown in Figure 5-16.
- 84558
- 1b-012985 5-4
5-4. COMPACT TENSION TEST Per the Surveillance Capsule Testing Contract with South Carolina Electric and Gas, 1/2T compact tension (CT) specimen will not be tested. CT specimen will be stored at the Hot Cell at the Westinghouse R&D Center. 4 i 84558:1b-082185 5-5
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE V. C. STAS 4ER UNIT 1 INTERMEDIATE SHELL PLATE A9154-1 (LONGITUDINAL) IRRADIATED AT 550*F, FLUENCE 6.39 x 1018 n/cm2 (E > 1 NeV) Temperature Impact Energy lateral Expansion Sample No. *F (*C) ft-lbs (Joules) mils (mm) % Shear CL11 -25 (-32) 7.0 ( 9.5) 3.5 (0.09) 11 CL1 0 (-18) 16.0 ( 21.5) 11.0 (0.28) 16 CL5 25 ( -4) 43.0 ( 58.5) 32.0 (0.81) 14 CL2 25 ( -4) 21.0 ( 28.5) 19.0 (0.48) 12 CL15 25 ( -4) 38.0 ( 51.5) 28.0 (0.71) 18 CL9 50 ( 10) 42.0 ( 57.0) 31.5 (0.80) 20 CL7 50 ( 10) 70.0 ( 95.0) 51.0 (1.30) 32 CL8 75 ( 24) 84.0 (114.0) 56.5 (1.44) 38 CL13 100 ( 38) 60.0 ( 81.5) 42.0 (1.07) 37 CL4 100 ( 38) 77.0 (104.5) 55.0 (1.40) 44 CL14 150 ( 66) 109.0 (148.0) 77.0 (1.96) 62 CL3 200 ( 93) 118.0 (160.0) 80.0 (2.03) 97 CL12 250 (121) 126.0(171.0) '90.0 (2.29) 100 CLIO 300 (149) 135.0 (183.0) 86.0 (2.18) 100 CL6 350 (177) 132.0 (179.0) 84.0 (2.13) 100 84558:1b-012985 5-6
TABLE 5-2 CHARPY V-NOTCH INPACT DATA FOR THE V.C. SUW ER UNIT 1 INTERNEDIATE SHELL PLATE A9154-1 (TRANSVERSE) 18 IRRADIATED AT 550*F, FLUENCE 6.39 x 10 n/cm2 (E > 1 NeV) Temperature Impact Energy Lateral Expansion l Sample No. 'F ('C) ft-lbs.(Joules) mils (mm) % Shear CT6 -50 (-46) 5.0 ( 7.0) 2.5 (0.06) 0 CT10 0 (-18) 23.0 ( 31.0) 15.5 (0.39) 2 CT7 0 (-18) 15.0 ( 20.5) 12.5 (0.32) 1 CT4 25 ( -4) 24.0 ( 32.5) 19.0 (0.48) 5 CT12 25 ( -4) 23.0 ( 31.0) 20.5 (0.52) 4 CT3 50 ( 10) 32.0 ( 43.5) 24.5 (0.62) 10 CT1 50 ( 10) 17.0 ( 23.0) 19.0 (0.48) 8 CT14 75 ( 24) 69.0 ( 93.5) 52.5 (1.33) 86 CT2 76 ( 24) 36.0 ( 49.0) 31.5 (0.80) 86 CT15 100 ( 38) 49.0 ( 66.5) 42.5 (1.08) 95 CT13 150 ( 66) 61.0 ( 82.5) 52.0 (1.32) 97 CT9 200 ( 93) 70.0 ( 95.0) 62.5 (1.59) 100 CT11 300 (149) 76.0 (103.0) 65.0 (1.65) 100 CT8 350 (177) 80.0 (108.5) 73.0 (1.85) 100 8455B:1b-012985 5-7
m
. TABLE 5-3 CHARPY V-NOTCH IMPACT DATA FOR THE V.C. SUMMER UNIT 1 PRESSURE VESSEL WELD METAL IRRA0IATED AT 550*F, 18 FLUENCE 6.39 x 10 n/cm2 (E > 1 MeV)
Temperature Impact Energy lateral Expansion Sample No. 'F (*C) ft-lbs (Joules) mils (mm) % Shear CW9 -100 (-73) 7.0 ( 9.5) 7.0 (0.18) 2 CW2 -60 (-51) 24.0 ( 32.5) 18.0 (0.46) 16 CW7 -40 (-40) 26.0 ( 35.5) 20.5 (0.52) 25 CW13 -25 (-32) 22.0 ( 30.0) 23.0 (0.58) 37 CW3 -25 (-32) 38.0 ( 51.5) 26.0 (0.66) 26 CW14 -10 (-23) 43.0 ( 58.5) 33.0 (0.84) 37 CW12 0 (-18) 44.0 ( 59.5) 35.0 (0.89) 42 CW6 0 (-18) 49.0 ( 66.5) 37.0 (0.94) 54 CW4 25 ( -4) 54.0 ( 73.0) 44.0 (1.12) 58 CW15 50 ( 10) 58.0 ( 78.5) 51.5 (1.31) 65 CW11 75 ( 24) 84.0 (114.0) 72.0 (1.83) 100 CW10 100 ( 38) 86.0.(116.5) 70.0 (1.78) 100 CWS 150 ( 66) 86.0 (116.5) 71.0 (1.80) 100 CW1 200 ( 93) 86.0 (116.5) 57.0 (1.45) 100 CW8 300 (149) 92.0 (124.5) 60.5 (1.54) 100 84558:1b-061385 5-8
c TABLE 5-4 CHARPY V-NOTCH IMPACT DATA FOR THE V.C. SIM4ER UNIT 1 PRESSURE VESSEL WELD HEAT AFFECTED ZONE METAL 18 ofc ,2 (E > 1 MeV) IRRADIATED AT 550*F, FLUENCE 6.39 x 10 Temperature Impact Energy Lateral Expansion Sample No. 'F (*C) ft-lbs (Joules) mils (mm) % Shear CH2 -100(-73) 18.0 ( 24.5) 11.0 (0.28) 11 CH13 -60 (-51) 32.0 ( 43.5) 26.5 (0.67) 18 CH14 -40 (-40) 37.0 ( 50.0) 22.5 (0.57) 20 CH3 -40 (-40) 38.0 ( 51.5) 27.0 (0.69) 26 CH7 -25 (-32) 53.0 ( 72.0) 40.0 (1.02) 30 CHil 0 (-18) 79.0 (107.0) 49.5 (1.26) 60 CHIS 25 ( -4) 71.0 ( 96.5) 56.5 (1.44) 57 CH8 25 ( -4) 92.0.(124.5) 60.5 (1.54) 79 CH9 50 ( 10) 95.0 (129.0) 70.0 (1.78) 100 CH6 50 ( 10) 118.0 (160.0) 69.5 (1.77) 100 CH4 75 ( 24) 124.0 (168.0) 80.5 (2.04) 100 CH12 100 ( 38) 99.0 (134.0) 66.5 (1.69) 100 CH1 150 ( 66) 125.0 (169.5) 81.5 (2.07) 100 CH10 200 ( 93) 148.0 (200.5) 84.0 (2.13) 100 CHS 300(149) 138.0 (187.0) 77.0 (1.96) 100 i l 84558:1b-012985 5-9
1 1 ! TABLE 5-5 INSTRIAENTED CHARPY IMPACT TEST RESULTS FOR V. C. Stater UNIT 1 INTERMEDIATE SHELL PLATE A9154-1 (LONGITUDINAL ORIENTATION) Normalized Energies Test Charpy Cnarpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2 2 2 No. (*C) (Joules) (kJ/m ) (kJ/m ) (kJ/m ) (N) (USec) (N) (pSec) (N) (N) (MPa) (MPa) CL11 -32 9.5 119 58 60 15100 90 15500 100 15200 300 779 788 CL1 -18 21.5 271 221 50 15000 105 17400 270 17400 0 770 843 CL2 -4 28.5 356 223 133 15100 95 17200 270 17200 800 776 831 CL15 -4 51.5 644 546 98 15000 95 19100 270 19100 100 769 881 I
~
CL5 -4 58.5 729 634 94 14700 100 19200 570 19200 200 759 874 o CL9 10 57.0 712 559 153 13100 90 17900 665 17900 0 675 798 CL7 10 95.0 1186 643 543 14100 90 19200 625 16700 2200 723 857 CL8 24 114.0 1424 639 785 13100 85 18300 675 16200 6700 675 808 CL13 38 81.5 1017 685 331 12900 95 17800 695 17300 5500 663 789 CL4 38 104.5 1305 636 669 13200 100 18500 775 15800 5200 680 817 CL14 66 148.0 1847 700 1148 13000 95 18000 695 670 799 CL3 93 160.0 2000 584 1416 12800 95 17700 775 660 784 CL12 121 171.0 2135 619 1516 10300 85 16600 765 528 690 CLIO 149 183.0 2288 581 1707 9100 130 16100 780 469 649 CL6 177 179.0 2237 604 1633 11200 90 16500 725 577 713 8455B:1b-061385 . e. a
l TABLE 5-6 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR V. C. SItedER UNIT 1 INTERMEDIATE SHELL PLATE A9154-1 (TRANSVERSE ORIENTATION) Normalized EnergteS l Test Charpy Charpy Maximum Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress l No. (*C) (Joules) (kJ/m ) (kJ/m ) (kJ/m ) (N) (pSec) (N) (pSec) (N) (N) (MPa) (MPa) CT6 -46 7.0 85 38 47 13000 80 13700 85 13400 0 668 686 CT7 -18 20.5 254 147 108 129CO 85 15000 220 14700 0 664 717 - CTIO -18 31.0 390 308 82 14800 100 18200 355 17900 0 763 848 m CT12 -4 31.0 390 280 109 14500 100 17400 335 17400 700 748 822 L
~
CT4 -4 32.5 407 307 100 14300 90 17700 355 17000 500 737 824 CT1 to 23.0 288 111 177 13800 85 15200 160 15200 2600 711 746 l CT3 10 43.5 542 361 181 13200 90 17000 435 17000 0 677 775 CT14 24 93.5 1169 462 708 13000 90 17500 530 16300 8900 668 785 ( CT2 24 49.0 610 278 332 13500 85 17100 335 16800 6700 692 787 l CT15 38 66.5 830 507 323 11900 90 16400 620 15600 6700 613 728 CT13 66 82.5 1034 429 605 12700 100 17000 520 652 763 Cf9 93 95.0 1186 384 802 12600 95 16300 470 647 743 CT11 149 103.0 1288 434 854 10000 50 16300 530 512 676 CT8 177 108.5 1356 407 949 8400 65 14900 560 432 f,99 l l l 8455B:lb-012985
TABLE 5-7 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR V. C. SANFR UNIT 1 WELD ETAL Normallzed Energies Test Charpy Charpy Itax imum Prop Yleid Time Itaximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2 2 No. (*C) (Joutes) (kJ/m ) (kJ/m ) (kJ/m ) (N) (pSec) (N) (pSec) (N) (N) (IAPa) (illPa ) CW9 -73 9.5 119 83 36 16300 100 17300 125 17000 400 841 865 CW2 -51 32.5 407 328 79 15800 95 19000 355 18600 0 815 897 Cw7 -40 35.5 441 335 106 15400 90 19100 360 18700 1500 793 887 Cw13 -32 30.0 373 235 138 15500 105 17800 280 17900 2300 800 858 CW3 -32 51.5 644 387 257 15100 110 18800 435 18800 300 778 874 Cw14 -23 58.5 729 393 336 12900 70 16900 475 16700 2600 664 766 [ Cw12 -18 59.5 746 464 281 14100 85 18700 500 17800 3600 726 843 N Cw6 -18 66.5 830 482 348 15100 95 19100- 510 18500 5900 779 880 Cw4 -4 73.0 915 443 472 13100 85 17700 510 16800 8200 675 792 Cw15 to 78.5 983 510 473 13600 90 18000 570 17400 11100 698 812 Cw11 24 114.0 1424 513 911 12900 80 18200 560 664 800 CwlO 38 116.5 1458 543 915 13700 90 18300 595 705 824 CWS 66 116.5 1458 533 925 13300 00 17800 595 687 803 CW1 93 116.5 1458 563 894 12900 95 17400 650 662 778 CW8 149 124.5 1559 561 998 12600 130 16800 695 648 755 84558:1b-012985 .. c.
( vc
.- s-TABLE 5-8 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR V. C. maaFR UNIT 1 WELD HEAT AFFECTED ZONE METAL Norma 112ed Energies Test Charpy Charpy Maximum Prop Yleid Time Maximum Time to Fracture Arrest Yield Flow Sample Temp. Energy Ed/A Em/A Ep/A Load to Yleid Load Maximum Load Load Stress Stress No. (*C) (Joules) (kJ/m ) (kJ/m2 ) (kJ/m2 ) (N) (pSec) (N) (pSec) (N) (N) (MPa) (MPa)
CH2 -73 24.5 305 267 38 17600 110 19800 295 19800 200 906 '964 CH13 -51 43.5 542 401 141 16900 100 20100 410 20100 1400 872 954 CH14 -40 50.0 627 490 137 16800 100 20200 495 20100 1200 862 951 CH3 -40 51.5 644 493 151 15700 95 19800 505 18900 200 808 914 CH7 -32 72.0 898 579 320 13800 90 18500 650 16700 5200 711 833 CH11 -18 107.0 1339 604 735 16200 90 19800 605 18200 10900 833 925
-4 96.5 1203 530 674 12800 90 18000 600 15400 5900 660 794
[ W CHIS 95 19000 610 776 878 CH8 -4 124.5 1559 584 976 15100 CH9 10 129.0 1610 605 1005 13900 95 18800 650 715 842 CH6 10 160.0 2000 598 1402 13300 85 18200 665 687 811 CH4 24 168.0 2102 721 1381 13100 95 18100 805 674 803 CH12 38 134.0 1678 598 1079 13300 90 18000 675 683 804 CH1 66 169.5 2118 712 1406 13300 90 18700 775 685 822 CH10 93 200.5 2508 606 1903 13100 105 17900 695 673 798 CHS 143 187.0 2339 625 1713 11800 95 17200 740 605 746 84558:1b-012985
TABLE 5-9 EFFECT OF 550*F IRRADIATION AT 6.39 x 1018 ,jc ,2 (E > 1 MeV) ON THE NOTCH TOUGHNESS PROPERTIES OF THE V. C. SubmER UNIT 1 REACTOR VESSEL MATERIALS Average Average 35 mit Average Average Energy Absorption 30 f t - I b Temp (
- F ) Lateral Expanston Temp (*F) 50 f t -I b Temp (
- F ) at Full Shear (ft-Ib)
Material Untrradiated Irradiated AT untrradiated Irradiated AT untrradiated Irradtated AT untrradtated Irradtated A(ft-Ib) Plate -20 20 40 0 40 40 0 45 45 133 131 2 A9154-1 (Longitudinal) Plate 25 55 30 55 75 20 70 105 35 75 75 -O y 09154-1 g (Transverse) weld -55 -25 30 -30 0 30 -15 15 30 91 87 4 Metal HAZ Metal -85 -55 30 -55 -25 30 -65 -30 35 136 121 15 8455B:1b-061385 ,, ,,
e .- 1 TABLE 5-10 TENSILE PROPERTIES FOR V. C. SUledER UNIT 1 REACTOR VESSEL MATERIAL IRRADIATED TO 6.39 x 1018 ,7c ,2
'ast 2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample emp. Strength Strength Load Stress Strength Elongation Elongation in Area No. Material (*F ) (ksi) (kst) (ktp) (kst) (ksi) (%) (%) (%)
CL2 LONG (A9154-1) 100 64.2 88.4 2.80 179.3 57.0 11.0 26.6 68 CII LONG (A9154-1) 250 63.2 86.8 2.80 165.0 57.0 10.5 23.4 65 CL3 LONG (A9154-1) 550 62.1 89.6 3.10 160.1 63.2 10.5 22.7 61 Y
$ CT1 TRANS (A9154-1) 74 64.2 91.7 3.40 146.3 69.3 11.7 23.3 53 CTI TRANS (A9154-1) 200 59.6 84.3 3.00 159.0 61.1 10.7 22.8 62 CT2 TRANS (A9154-1) 550 56.0 87.6 3.30 168.3 67.2 10.5 20.3 60 CW2 WELD Metal O 72.6 97.8 3.30 205.5 67.2 12.3 24.8 67 CW3 WELD Metal 74 72.3 92.5 3 10 185.2 63.2 12.3 24.0 66 CW1 WELD Metal 550 66.1 88.4 3.34 128.4 68.0 9.3 15.5 47*
- Specimen CW1 broke at the extensometer knife edge. auctility values may be in error. (Figure 5-12) t l
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Figure 5 2 Irradiated Charpy V Notch Impact Properties for V.C. Summer Unit 1 Reactor Pressure VesselIntermediate Shell Plate A91541 (Transverse Orientation) 5 17
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Figure 5 4 Irradiated Charpy V Notch Impact Properties for V.C. Summer Unit 1 Reactor Pressure Vessel Weld Heat Affected Zone Metal 5 19
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1 13472A-3 l TEMPERATURE (OC) 0 50 100 150 200 250 300 350
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s SECTION 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY l 6-1. INTRODUCTION Knowledge of the neutron environment within the reactor pressure vessel / surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that exparienced 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. This section describes a discrete ordinates S transport analysis performed n for the V. C. Summer reactor to determine the fast (E > 1.0 Mev) neutron flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules. The analytical data were then used to develop lead factors for use in relating neutron exposure of the reactor vessel to that of the surveillance capsules. Based on the use of spectrum-averaged reaction cross sections derived from this calculation and the V. C. Summer power
' history, the analysis of the neutron dosimetry contained in Capsule U is presented.
6-2. DISCRETE ORDINATES ANALYSIS A plan view of the V. C. Summer reactor geometry at the core midplane is shown in figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a zero- to 45-degree sector is depicted. Six irradiation capsules attached to 1
)
3038e:1d/052185 6-1
~
the neutron pad are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at 16.94 degrees (U,V,X) and 19.72 degrees (W,Y,Z) from the cardinal axes as shown in figure 6-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in figure 6-2. The stainless steel specimen containers are approximately 1-inch square 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. 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 ca rgy 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. This requires at least a two-dimensional calculation. In the analysis of the neutron environment within the V. C. Summer reactor geometry, predictions of neutron flux distributions and energy spectra were -made with the DOT [5] two-dimensional discrete ordinates transport code. The radial and azimuthal distributions were obtained from an R,0 calculation wherein the geometry shown in figures 6-1 and 6-2 was represented in the analytical model. In addition to the R,0 calculation, a second calculation in R,Z geometry was also carried out to obtain relative axial variations of neutron flux throughout the geometry of interest. In the R,Z analysis, the reactor core was treated as an equivalent volume cylinder. The surveillance capsules were not included in the R,Z model. Both the R,0 and R,Z analyses employed 47 neutron energy groups and a P 3 expansion of the scattering cross sections. The cross sections used in the analyses were obtained from the SAILOR cross section library E03 which was developed specifically for light water reactor applications. The neutron energy group structure used in the analysis is listed in table 6-1. 3038e:ld/052185 6-2 l
s A key input parameter in the analysis of the integrated neutron exposure of the reactor vessel is the core power distribution. For this analysis, core power distributions representative of time-averaged conditions derived f rom statistical studies of long-term operation of Westinghouse 3-loop plants were employed. These input distributions include rod-by-rod spatial variations for all peripheral fuel assemblies. This generic, design basis, core power distribution is intended to provide a vehicle for the long-term (end-of-life) projection of reactor vessel exposure. Since plant-specific core power distributions reflect only past operation, tneir use for projection into the future may not be justified. The use of generic data which reflects long-term operation of similar reactor cores may provide a more suitable approach. Benchmark testing of these generic core power distributions and the SAILOR cross sections against surveillance capsule data obtained from two , three ,
' and four-loop Westinghouse plants indicate that this analytical approach yields conservative re >ults, with calculations exceeding measurements f rom 10 to 25 percent.
One further point of interest regarding these analyses is that the design basis assumes an out-in fuel loading pattern (fresh fuel on the periphery). Future commitment to low-leakage core loading patterns could significantly reduce the calculated neutron flux levels presented in section 6-4. In addition, surveillance capsule lead factors could be changed, thereby influencing the withdrawal schedule of the remaining surveillance capsules. Having the results of the R,0 and R,Z calcul.ations, three-dimensional variations of neutron flux may be approximated by assuming that the following relation holds for the applicable regions of the reactor.
$(R,Z,0,E ) = 4(R,0,E ) x F(Z,E ) ( 6-1 )
g g 3038e:ld/052185 6-3 l
d where
$(R,Z,0,Eg ) = neutron flux at point R,Z,0 within energy group g
$(R,0,Eg) = neutron flux at point R,0 within energy group g obtained from the R,0 calculation F(Z,Eg ) = relative axial distribution of neutron flux within energy group g obtained from the R,Z calculation This analysis is consistent with established ASTM standards.[8,9,10,H,12]
6-3. RADIOMETRIC MONITORS The passive radiometric monitors included in the V. C. Summer surveillance program are listed in table 6-2. The first five reactions in table 6-2 are used as fast neutron monitors to relate fast (E > 1.0 MeV) neutron fluence to measured material property changes. In order to address the potential- for burnout of the product nuclides generated by fast neutron reactions, it is
-necessary to also determine the magnitude of the thermal and resonance region neutron fluxes at the monitor location. Therefore, bare and cadmium-shielded cobalt-aluminum monitors are also included.
The relative locations of the various radiometric monitors within the surveillance capsule are shown in figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers
.at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the axial center of the capsule. All monitors are located radially at the center of the capsule and azimuthally within 1 0.23 degrees of the capsule center.
The use of passive monitors such as those listed in table 6-2 does not yield a direct measure of the energy-dependent neutron 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 3038e:ld/052185 6-4
t 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 important. 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 analysis of the passive monitors and the subsequent derivation of the average neutron flux requires two operations. First, the disintegration rate of product nuclide per unit mass of monitor must be determined. Second, in order to define a suitable spectrum-averaged reaction cross section, the neutron energy spectrum at the monitor location must be calculated. The specific activity of each of the monitors is determined using established ASTM p ro c e d u re s . [13,14,15,16,17 ,18,19 ,20,21 ] Following sample preparation, the activity of each monitor is determined by means of a lithium-drif ted germanium, Ge(Li), gamma ray spectrometer. The overall standard deviation of the measured data is a function of the precision of sample weighing, the uncertainty in counting, and the acceptable error in detector calibration. For the samples removed from V. C. Summer, the overall 2a deviation in the measured data is determined to be plus or minus 10 percent. The neutron energy spectrum at the monitor location is determined analytically using the method described in paragraph 6-2. Having the measured activity of the monitors and the neutron energy spectrum at the monitor locations of interest, the calculation of the neutron flux proceeds as follows. The reaction product activity in the monitor is expressed as l I P A=N g FY a(E) $(E) dE p (1 -e-Xt)) ,-Atd (6-2) JE pj max l l 3038e:1d/052185 6-5
d where A = induced product activity (dps per gram) N g
= number of target element atoms per gram F = weight fraction of the target nuclide in the target material Y = number of product atoms produced per reaction a(E) = energy dependent reaction cross section
$(E) = energy dependent neutron flux at the monitor location with the reactor at full (reference) power
= average core power level during irradiation period j P)
P = maximum or reference core power level A ,
= decay constant of the product nuclide
= length of irradiation period j t) t = decay time following irradiation period j d
n = total number of irradiation periods Because the neutron flux distributions are calculated using multigroup transport methods and, further, because the main interest is in the fast (E > 1.0 MeV) neutron flux, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-2) is replaced by the following . elation. a(E) $(E) dE = a 9p JE where 47 a(E) $(E) dE p L "g *g 8" 1o " a=1
. 18 J l MeV +9 g=1 18
$(E) dE = $g
$p =d1 MeV g,)
g = group number f rom Table 6-1 3038e:ld/052485 6-6
Thus, equation (6-2) is rewritten n I A=N FYi$ f (1 -e -Atj ) ,-Atd max j =1 or, solving for the fast (E > 1.0 Mev) neutron flux,
$f= n No FYE (1-e~ j) e -At d (~}
p*** j =1 The total fast (E > 1.0 MeV) neutron fluence is then given by n
*f * #f p t j (6-4) j =1 max where P
p' t) = total ef fective full power seconds of reactor operation up to the time of capsule removal max p) An assessment of the potential for product nuclide burnout may be made using the bare and cadmium shielded cobalt measured activities and published data for the 2200 m/s absorption cross-section and the resonance integral This is done by rewriting equation (6-2) in terms of a monitor 2200 m/s neutron flux and a monitor resonance flux as follows: n A =N bare o F Y (o2200 +2200 + RI $res} P ax j =1 n Acq = No F Y (RI $res) () _, t j) e- td , (6-6) j=1 3038e:1d/052185 6-7
a A = bare induced product activity (dps per gram) bare A cd
= cadmium shielded induced product activity (dps per gram) a = published 2200 m/s absorption cross-section for nuclide 2200 of interest RI = published epicadmium dilute resonance integral for nuclide of interest 2200
= m nitor 2200 m/s neutron flux to be determined from measured activities
$res = monitor resonance neutron flux to be determined from measured activities Equations (6-5) and (6-6) are solved for $ 2200 and $res using the average measured bare and cadmium shielded cobalt activities at the monitor location.
The total loss rate of a product nuclide may then be expressed as the sum of its radioactive decay rate and the neutron absorption rate in that nuclide while the reactor is at power. The product nuclide neutron absorption rate may be estimated from the published data for a2200 and RI and the monitor fluxes determined above. If the neutron absorption rate is small when compared to the decay rate then there is no concern regarding burnout. 6-4. NEUTRON TRANSPORT ANALYSIS RESULTS Results of the discrete ordinates transport calculations for the V. C. Summer reactor are summarized in this section. In figure 6-3, the calculated maximum fast (E > 1.0 MeV) neutron flux levels at the radius of the surveillance capsule center, the reactor vessel inner radius, the reactor vessel 1/4 thickness location, and the reactor vessel 3/4 thickness location are i 3038e:1d/052485 6-8
.~ __ _ _ _ _ _. __ _ _ _ _ , _ . _ . _ , . . _ . . , . _ _ . . . . .
I i presented as a. function of azimuthal angle. The local influence of the j surveillance capsules on the fast neutron flux distribution is clearly evident. In figure 6-4, the radial distribution of maximum fast (E > 1.0 MeV) neutron flux through the thickness of the reactor vessel is shown. The relative axial variation of fast neutron flux within the reactor vessel is given in figure 6-5. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in figure 6-3 or 6-4 by the appropriate values from figure 6-5. Table 6-3 provides the calculated f ast neutron exposure parameters for the V. C. Summer reactor vessel. Table 6-4 provides the calculated fast neutron exposure parameters and updated lead factors for all of the V. C. Summer surveillance capsules. The lead factor is defined as the ratio of the fast (E > 1.0 MeV) neutron flux at the dosimeter block location (capsule center) to the maximum fast neutron flux at the reactor vessel inner radius. Table 6-5 provides the calculated fast neutron exposure parameters for the various metallurgical specimens within V. C. Summer surveillance capsule U. In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. The calculated neutron energy spectrum at the center of the V. C. Summer surveillance capsule U is listed in table 6-6. The calculated spectrum-averaged cross sections for each of the fast neutron reactions are given in table 6-7.
- 5. 00SIMETRY RESULTS The irradiation history of the V. C. Summer reactor up to the time of removal of Capsule U is listed in table 6-8. Comparisons of measured and calculated saturated activity of the radiometric monitors contained in Capsule U based on the irradiation history shown in table 6-8 are listed in table 6-9.
The fast (E > 1.0 Mev) neutron flux and fluence levels derived for Capsule U using the spectrum averaged cross-sections listed in table 6-7 are presented 3038e:1d/052185 6-9 l I
i a in table 6-10. Table 6-11 summarizes the key nuclear data and results of the product nuclide burnout assessment that was performed. Due to the relatively low thermal and resonance neutron fluxes at the surveillance capsule location, the neutron absorption rate is negligably small when compared to the radioactive decay rate. Therefore, no correction has been made for product nuclide burnout. An examination of table 6-10 shows that the average fast (E > 1.0 McV) neutron flux derived from the five threshold reactions is 1.80 x 10 ll n/cm2 -sec with a la standard deviation of 17.5 percent. The calculated flux value of 2.09 x 10 ll n/cm2 -sec exceeds all of the measured values, with calculation to experimental ratios ranging f rom 1.06 to 1.25. A summary of measured and calculated current fast neutron exposures for Capsule U and for key reactor vessel locations is presented in table 6-12. The measured value is given based'on the average of all five threshold reactions listed in table 6-10. End-of-life (EOL) reactor vessel fast neutron fluence projections are also included in table 6-12. Based on the data given in table 6-10, the best estimate fast neutron exposure of Capsule U is 4 = 6.39 x 10 18 n/cm2 (E > 1 MeV) at 1.12 EFPY. 3038e:ld/052185 6-10
~ TABLE 6-1 SAILOR 47 NEUTRON ENERGY GROUP STRUCTURE Group Group Energy Lower Energy Energy Lower Energy Group (MeV) Group (MeV) 1 14.19(a) 25 0.183 2 12.21 26 0.111 3 10.00 27 0.0674 4 8.61 28 0.0409 5 7.41 29 0.0318 6 6.07 30 0.0261 7 4.97 31 0.0242 8 3.68 32 0.0219 9 3.01 33 0.0150
-3 10 2.73 34 7.10x10
-3 11 2.47 35 3.36x10
-3 12 ' 2.37 36 1.59x10
-4 13 2.35 37 4.54x10
~4 14 2.23 38 2.14x10
-4 15 1.92 39 1.01x10
-5 16 1.65 40 3.73x10
-5 17 1.35 41 1.07x10
-6 18 1.00 42 5.04x10
-6 19 0.821 43 1.86x10
-I 20 0.743 44 8.76x10
-I 21 0.608 45 4.14x10
~I 22 0.498 46 1.00x10 23 0.369 47 0.00 24 0.298 a) The upper energy of group 1 is 17.33 MeV.
3038e:1d/052185 6-11
TA8LE 6-2 NUCLEAR CONSTANTS FOR RADIOMETRIC MONITORS CONTAINED IN THE V. C. SUMMER SURVEILLANCE CAPSULES Reaction Target Fission of Weight Product Yield Monitor Material Interest Fraction Ha l f -li f e (%) Iron wire Fe (n p) Mn
- 0.058 312.2 dy Nickel wire Ni (n,p) Co 0.6827 70.91 dy Copper wire Cu (n,a) Co 0.6917 5.272 yr Uranium-23858) in U38 0 0 8 (n,f) Cs 1.0 30.17 yr 6.0 Neptunium-237(# in Np0 Np (n,f) Cs 1.0 30.17 yr 6.5 2
Cobalt-aluminum (a) wire CoS9 (n,y) Co60 0.0015 5.272 yr Cobalt-aluminum wire Co (n,y) Co 0.0015 5.272 yr a) Denotes that the monitor is cadmium-shielded i 1 I 3038e:ld/052185 6-12
TABLE 6-3 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS FOR THE PEAK LOCATION OF THE V. C. SUMMER REACTOR VESSEL Iron Radial Location (a) Fast Neutron Flux Displacement Within the (n/cm -sec) Rate Reactor Vessel (E > 1.0 MeV) (E > 0.1 Mev) , (doa/sec) Inner Surface 6.73 x 10 10 1.63 x 10 II 1.07 x 10 -10 (R = 78.500 inches) 10 ll -II 1/4 Thickness 3.99 x 10 1.44 x 10 7.27 x 10 (R = 80.469 inches) 10 Il -I I 1/2 Thickness 1.98 x 10 1.05 x 10 4.49 x 10 (R = 82.438 inches) 3/4 Thickness 9.26 x 10 9 6.83 x 10 10 2.64 x 10'II (R = 84.406 inches)
~II Outer Surface 3.81 x 10 9 3.46 x 10 10 1.28 x 10 (R =.86.375 inches) i i
a) The peak is located at zero degrees azimuthally and on the core midplane. 3038e:1d/052185 6-13
d TABLE 6-4 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AND LEAD FACTORS FOR THE V. C. SUMMER SURVEILLANCE CAPSULES Iron Azimuthal Fast Neutron Flux Displacement Capsule Location (a) 2 (n/cm -sec) Rate Lead
-I.D. (Dearees) (E > 1.0 Mev) (E > 0.1 MeV) (dpa/sec) Factor (D) y 106.94 2.09 x 10 Il 1.05 x 10 12 4.47 x 10 -10 3.11 W 109.72 1.81 x 10 II 8.80 x 10 ll 3.80 x 10 -10 2.69 X 286.94 2.09 x 10 Il 1.05 x 10 12 4.47 x 10 -10 3.11 Y 289.72 1.81 x 10 II 8.80 x 10 ll 3.80 x 10 -10 2.69 Z 340.28 1.81 x 10 ll 8.80 x 10 II 3.80 x 10 -10 2.69 U 343.06 2.09 x 10 Il 1.05 x 10 12 4.47 x 10 -10 3.11 a) The radius of the surveillance center is 73.310 inches.
b) The lead factor is the ratio of the f ast (E > 1.0 Mev) neutron flux at the center of the surveillance capsule to that at the peak location on the reactor vessel inner surface. l 3038e:1d/052485 6-14
)
r TABLE 6-5 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS FOR V. C. SUMMER SURVEILLANCE CAPSULE U METALLURGICAL SPECIMENS Fast Neutron Flux Iron 2 (n/cm -sec) Displacement Specimen Tvoe and location (E > 1.0 MeV) LE > 0.1 MeV) (doa/sec)
~ CHARPY SPECIMENS ll 12 -10 R = 73.007 in. 0 = 342.75 deg. 2.36 x 10 1.16 x 10 4.98 x 10 Il 12 -10 0 = 343.06 deg. 2.40 x 10 1.18 x 10 5.06 x 10 II 12 -10 0 = 343.37 deg. 2.43 x 10 1.18 x 10 5.10 x 10 ll Il -10 R = 73.613 in. 0 = 342.75 deg. 1.79 x 10 9.17 x 10 3.89 x 10 II Il -10 0 = 343.06 deg. 1.82 x 10 9.27 x 10 3.93 x 10 II II 0 0 = 343.37 deg. 1.84 x 10 9.25 x 10 3.95 x 10 TENSILE SPECIMENS R = 73.507 in. 0 = 342.75 deg. 1.88 x 10 II 9.57 x 10 ll 4.06 x 10 -10 II II -10 0 = 343.06 deg. 1.91 x 10 9.68 x 10 4.11 x 10 II II -10 0 = 343.37 deg. 1.93 x 10 9.66 x 10 4.13 x 10 PRE-CRACKED 8END 8AR
-10 R = 73.310 in. 0 = 343.06 deg. 2.09 x 10 II 1.05 x 10 12 4.47 x 10 COMPACT TENSION SPECIMEN II 12 -10 R = 73.060 in. 0 = 343.06 deg. 2.35 x 10 1.15 x 10 4.96 x 10 II II -10 R = 73.560 in. 1.87 x 10 9.47 x 10 4.02 x 10 3038e:1d/052185 6-15
d TABLE 6-6 CALCULATED NEUTRON ENERGY SPECTRUM AT THE CENTER OF V. C. SUMMER SURVEILLANCE CAPSULE U Energy Neutron Flux Energy Neutron Flux group _ (n/cm2 -sec) Group (n/cm2 -sec) 7 1 3.00 x 10 25 1.29 x 10 Il 9 Il 2 1.11 x 10 26 1.42 x 10 8 Il 3 3.89 x 10 27 1.18 x 10 8 10
.4 7.17 x 10 28 7.38 x 10 5 1.21 x 10 9 29 1.96 x 10 10 9 10 6 2.71 x 10 30 1.04 x 10 10 7 3.80 x 10 9 31 3.16 x.10 9 10 8 7.95 x 10 32 2.25 x 10 10 9 7.46 x 10 9 33 2.96 x 10 9 10 10 6.27 x 10 34 3.63 x 10 .
9 0 11 7.61 x 10 35 6.65 x 10 12 3.81 x 10 9 36 6.74 x 10 10 10 13 1.16 x 10' 37 9.04 x 10 9 10 14 5.90 x 10 38 4.54 x 10 10 10 15 1.64 x 10 33 5.04 x 10 10 10 16 2.25 x 10 40 6.95 x 10 17 '3.60 x 10 10 41 8.04 x 10 10 10 10 18 8.62 x 10 42 4.26 x 10 19 6.58 x 10 10 43 4.37 x 10 10 10 10 20 3.06 x 10 44 2.41 x 10 21 1.21 x 10 ll 45 1.61 x 10 10 10 10 22 9.51 x 10 46 1.68 x 10 23 1.28 x 10 II 47 1.68 x 10 10 ll 24 1.29 x 10 3038e:1d/052185 6-16
n- ,
-- 4 TABLE 6-7 SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF V. C. SUMMER SURVEILLANCE CAPSULE U Spectrum-Averaged Cross Section(a)
Reaction of Interest (barns) Fe54 (n,p) Mn54 0.0515 Ni (n,p) Co$ 0.0726 Cu63 (n.a) Co60 0.000428 U238 (n,f) Cs137 0.301 i Np (n,f) Cs 3.43 f*a a> a.9.(E)$(E)dE a(E) dE J 1 MeV 3038e:1d/052485 6-17 J
TA8LE 6-8 IRRADIATION HISTORY OF V. C. SUMMER SURVEILLANCE CAPSULE U P 'j ' max Irradiation Time Decay Time P) max
' Month Year (MWt) (MWt) (Days) (Davs) 11 1982 426 2775 0.154 14 751 12 1982 872 2775 0.314 31 720 1 1983 1306 2775 0.471 31 689 2 1983 1252 2775 0.451 28 661 3 1983 759 2775 0.274 31 630 4 1983 0 2775 0.000 30 600 5 1983 355 2775 0.128 31 569 6 1983 2316 2775 0.835 30 539 7 1983 2496 2775 0.899 31 508 8 1983 2355 2775 0.848 31 477 9 1983 2421 2775 0.872 30 447 10 1983 2424 2775 0.874 31 416 11 1983 2101 2775 0.757 30 386 12 1983 1072 2775 0.386 31 355 1 1984 2674 2775 0.964 31 324 2 1984 2386 2775 0.860 29 295 3 1984 1955 2775 0.704 31 264 4 1984 165 2775 0.059 30 234 5 1984 2282 2775 0.822 31 203 6 1984 2557 2775 0.921 30 173 7 1984 1116 2775 0.402 31 142 8 1984 2471 2775 0.890 31 111 9 1984 2083 2775 0.751 28 83 CAPSULE U REMOVED Note: 1) Decay time is referenced to 12/20/84
- 2) Total irradiation time is 3.55 x 107 ef fective full power seconds (EFPS) or 1.12 ef fective full power years (EFPY)
- 3) Pj is the average core power level during the irradiation period 3038e:1d/052185 6-18
)
. TA8LE 6-9 COMPARISON OF MEASURED AND CALCULATED RADIOMETRIC MONITOR SATURATED ACTIVITIES FOR V. C. SUMMER SURVEILLANCE CAPSULE U Radiometric Monitor Monitor Saturated Activity and (Disintegrations /second-Gram) l Axial location (a) Measured Basis Calculated C/E Fe54 (n.P) Mn54 hm of W e)
D Top 5.53 x 10 Middle 5.48 x 10 6 Bottom 5.95 x 10 6 Average IDI 6 5.65 x 10 [ 4.6%), 6.98 x 10 6 1.24 Ni' (n.P) Co (gm of wire) Top 9.21 x 10 7 Middle 8.70 x 10 7 Bottom 9.40 x 10 Average ( 9.10 x 10 [14.0%) 1.08 x 10 1.19 l Cu63 (n.a) Co60 (gm of wire) 5 Top 5.59 x 10 Middle 5.51 x 10 5 Bottom 5.75 x 10 5 Average IDI 5 5.62 x 10 [ 2.2%) 5.93 x 10 1.06 3038e:1d/052485 6-19
n
. TABLE 6-9 (continued)
COMPARISON OF MEASURED AND CALCULATED RADIOMETRIC MONITORS $TURATEDACTIVITIESFOR V. C. SUMMER SURVEILLANCE CAPSULE U Radiometric Monitor Monitor Saturated Activity and (Disintegrations /Second-Gram) Axial location (a) Measured Basis Calculated C/E 238 137 U (n.f) Cs (c) (gm of U ) Middle 1.01 x 10 Corrected (d) 8.14 x 10 9.57 x 10 1.18 237 137 ND (n.f) Cs (c) (gm of Np237) Middle 9.52 x 10 7 1.19 x 10 8 1.25 Co (n.Y) Co (C) (gm of Wire) Top 6.83 x 10 Middle 6.95 x 10 I
-Bottom 7.05 x 10 Average ( 6.94 x 10 [11.6%) 7.30 x 10 1.05 CoS9 (n.y) Co 60 (gm of wire) 8 Top 1.24'x 10 Top 1.04,x '10 0 Middle- 1.27 x~10 0 Middle 1.05 x 10 8 8
Bottom 1.27 x 10 8 Bottom 1.12 x 10 Average ( 1.16 x 10 (19.3%] 8.56 x 10 0.74 3038e:1d/052485 6-20
6.
. 1 1
TA8LE 6-9 (continued) COMPARISON OF MEASURED AND CALCULATED RADIOMETRIC l MONITOR SATURATED ACTIVITIES FOR V. C. SUMMER SURVEILLANCE CAPSULE U _ a) Refer to Figure 4-2 for the locations of the various radiometric monitors. b) .The standard deviation (la) of the mean saturated activity is expressed as a percentage of the mean. c) This radiometric monitor was cadmium shielded, d) The measured value has been multiplied by 0.806 to correct for the effect 23E 239 of 323 ppm U and the build-in of Pu , 4 t 4-3038e:1d/052185 6-21
w, ., b lkk
- p] TABLE 6-10 i
RESULTS OF FAST NEUTRON 00SIMETRY FOR V. C. SUMMER SURVEILLANCE CAPSULE U Current ( ) Radiometric Monitor Fast (E > 1.0 MeV) Fast (E > 1.0 Mev) Satured ActivityI ") Neutron Flux Neutron Fluence Reaction (dos /am) _ (n/cm -sec) (n/cm ) of j Interest * .s Measured Calculated Measured Calculated Measured Calculated v lm 6 6 II 18 Fe (n,p) Mn / 5.65x10 6.98x10 1.70x10 6.03x10 8 # Ni (n.p) Co 9.10x10 1.08x10 1.77x10 6. 29 x10 ' d 8 Cu (n,a) Co 5.62x10 5.93x10 1.99x10 7.06x10 II 18 U (n,f) Cs 8.14x10 9.57x10 1.79x10 6.36x10
]h Np (n,f) Cs 9.52x10 1.19x10 8
1.68x10 II 5.97x10 18 Il II 8 g Average 1.80x10 2.09x10 6.39x10 7.43x10
, [17.54]
a) Refer to Table 6-9. b) Total irradiation time for surveillance capsule U is 3.55x10 7 effective full power seconds (EFPS). ,
;g _3038e:1d/052185 6-22 3
~l\ .
a TABLE 6-11 PRODUCT NUCLIOE BURNOUT ASSESSMENT FOR V. C. SUMMER SURVEILLANCE CAPSULE U 3 NUCLEAR DATA E RI "2200 Nuclide Half-life (barns) (barns) 4 Mn 312.2 dy 10.0 - Co$ 70.91 dy 1880 6890 S9 Co Stable 37.2 75.5 60 Co 5.272 yr 2.0 4.3 137 0.50 Cs 30.17 yr 0.11 60 SURVEILLANCE CAPSULE U AVERAGE SATURATED Co ACTIVITY 8 Bare Co-Al wire: A = 1.16 x 10 dps/gm 7 Cd Shielded Co-Al wire: A = 6.94 x 10 dps/gm 60 MONITOR FLUXES DERIVED FROM Co _ SATURATED ACTIVITY N 2
+2200 = 8.22 x 10 10 n/cm2 -sec
$res = 6.00 x 10 n/cm -sec PRODUCT NUCLIDE LOSS RATE COMPARISON Absorption Rate i Decay Constant (a2200 '2200 + RI $res}
l Nuclide (1/sec) (1/sec) 54 Mn 2.57 x 10-8 8.22 x 10-13 Co S8 1.13 x 10- 5.68 x 10 -10 60 4.22 x 10-13 Co 4.17 x 10-9 Cs 137 7.28 x 10 -10 3.90 x 10 -I 4 I 3038e:1d/052485 6-23
TA8LE 6-12
SUMMARY
OF V. C. SUMMER FAST NEUTRON FLUENCE RESULTS 8ASED UPON SURVEILLANCE CAPSULE U End of Life Current Fast (E > 1.0 MeV) Fast (E > 1.0 Mev) Neutron FluenceI *) Neutron Fluence (b) 2 (n/cm ) (n/cm ) location Measured (c) Calculated Measured (c) Calculated 18 18 Capsule U 6.39x10 7.43x10 18 18 I9 I9 Vessel IR 2.05x10 2.39x10 5.84x10 6.80x10 18 18 I9 I9 Vessel 1/4T 1.22x10 1.42x10 3.47x10 4.03x10 II II 18 18 Vessel 3/4T 2.83x10 3.29x10 8.04x10 9.35x10 a) Current fluences are based on operation at 2775 MWt for 1.12 EFPY b) E0L fluences are based on operation at 2775 MWt for 32 EFPY. c) The measured results of surveillance Capsule U were extrapolated to the reactor vessel locations using the following calculated lead factors: Inner Radius - 3.11 1/4 Thickness - 5.24 3/4 Thickness - 22.6 3038e:1d/052185 6-24 C
J 48/55/13546 1 16.94 DEG. (CAPSULES U, V, X) 00 19.72 DEG. (CAPSULES W, Y, Z) l l e REACTOR VESSEL 450 iNNNxx / 11 , /
"' /
i
/
I s f I : j ~
/
/
l :, / l V I / I / l l
/
I / I
/
l l l / l n / l / I / b E___________________________ Figure 61. V.C. Summer Reactor Geometry 6-25
48/55/13546-2 l
- 16.94 DEG. - 19.72 DEG.
Y Y
- 73 .31IN.
I L l Figure 6-2. Plan View of a Dual Reactor Vessel Surveillance Capsule 6-26
i
! l 48/55/13546*J 1012 -
7 - 5 - 2 - 1011 -
~
o _ N
! 7 SURVEILLANCE CAPSULE CENTER x
3 u. Z O 2 - REACTOR VESSEL IR N S Z 1/4 T LOCATION 1010 _ 7 - 5 - _ 3/4 T LOCATION 2 - 109 I I I ! ! ! O 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (degrees) Figure 6-3. Calculated Azimuthal Distribution of Maximum Fast (E >1.0 MeV) Neutron Flux Within the Reactor Vessel- Surveillance Capsule Geometry 6-27
48/55/13546 4 1011 199.39 7 - l l 204.39 5 - lR 1/4T 209.39 2 - E
? l E 1/2T 214.39 !
N 1010 z - @ 7 - 3/4T 219.39 5 w 2 5 - OR 2 - 109 ! I 195 200 205 210 215 220 R ADIUS (cm) Figure 6-4. Calculated Radial Distribution of Maximum Fast (E >1.0 MeV) Neutron Flux Within the Reactor Vessel 6-28
' 48/55/5396A-20 10 0 _ 8 - 6 - 4 - 2 - y 10-1 -- d 8 -- 2 6 - E
$ 2 -
P 5 E 10-2 __ 8 . 6 - 4 - CORE MIDPLANE 2 - TO VESSEL CLOSURE HEAD 10-3
-300 -200 -100 0 100 200 300 400 DISTANCE FROM CORE MIDPLANE (cm)
Figure 6-5. Relative Axial Variation of Fast (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel 6-29
d SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the V. C. Summer Unit I reactor vessel: Vessel Estimated Location Lead Removal Fluence 2 Capsule (deg) Factor Time (a] (n/cm ) 18 U 343* 3.11 1.12 (removed) 6.39 x 10 V 107* 3.11 3 1.66 x 10 19 X 287* 3.11 6 3.32 x 10 19(b] W 110* 2.69 12 5.75 x 10 19(c] Y 290* 2.69 20 9.58 x 10 19 Z 340* 2.69 standby --
- a. Effective full power years from plant startup
- b. Approximate fluence at 1/4 thickness vessel wall at end of life
- c. Approximate fluence at vessel inner wall at end of life 8455B:1b-061385 7-1
)
o SECTION 8 REFERENCES
- 1. Davidson, J. A. and Yanichko, S. E., " South Carolina Electric and Gas Co.
-Virgil C. Summer Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-9234, January 1978.
- 2. ASTM E185-73 "Practive for Surveillance Tests for Nuclear Reactor" in ASTM Standards, Part 10 (1973), American Society for Testing and Materials, Philadelphia, Pa., 1973.
- 3. Lott, R. G. and Shogan, R. P., "V. C. Summer Unit No. 1 Nuclear Pressure vessel Surveillance capsule U Test Program," Westinghouse R&D Memo 85-502-CGEU-M1, 1985.
- 4. Regulatory Guide 1.99, Revision 1 "Ef fects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, April 1977.
- 5. Soltesz, R. G., Disney, R. K., Jedruch, J., and Ziegler, S. L., " Nuclear Rocket Shielding Methods Modification, Updating and Input Data Preparation. Vol. 5 - Two-Dimensional Discrete Ordinates Transport Technique," WANL-PR(LL)-034, Vol 5, August 1970.
- 6. "0RNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".
- 7. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology (to be published).
- 8. ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,
1984. 3038e:ld/052185 8-1
- 9. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 10. 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, Pa., 1984.
- 11. ASTM Designation E706-81a, " Standard Master Matr.ix for Light-Water Reactor Pressure Vessel Surveillance Standards", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 12. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 13. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,
1984.
- 14. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 15. ASTM Designation E263-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 16. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards, Section 12, America Society for Testing and Materials, Philadelphia, Pa., 1984.
3038e:ld/052185 8-2
- 17. ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,
1984.
- 18. ASTM Designation E523-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 19. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 20. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984,
- 21. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 22. Mughabghab, S. F. and Garber, D. I., " Neutron Cross-Sections," BNL-325, 3rd Edition, Volume 1 June 1973.
x 3038e:ld/052185 8-3 s
i 1
" APPENDIX A HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION A-1. . INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNOT (reference nil-ductility temperature). The most limiting RT NOT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART NDT' *
- 9" ** 9 '
NDT the drop weight nil-ductility transition temperature (NOTT) ~ or the temperature at which the material exhibits at least 50 ft Ib of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F. RT nc nases as ma e a s exp se as neu n adadon. h, NOT to find the most limiting RT a any p M od in W nac W s W e, NOT ART due to the radiation exposure associated with that time period must NDT be added to the original unirradiated RT NDT. The extent of the shift in RT is enhanced by certain chemical elements (such as copper) present in NOT reactor vessel steels. Design curves which show the effect of fluence and copper content on ART r mac W vessel steels a m shown in N gu m A-1. NDT Given the copper content of the most limiting material, the radiation-induced ART can e es ma m gu A. as ne n n uence M > 1 Me0 NOT at the vessel inner surface, the 1/4 T (wall thickness), and 3/4 T (wall thickness) ve,ssel locations are given as a function of full-power service life in Figure A-2. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RT NOT' 1236E:10/060785 A-1 s
A-2. FRACTURE TOUGHNESS PROPERTIES ' The preirradiation f racture-toughness properties of the V. C. Summer Unit I reactor vessel materials are presented in Table A-1. The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan.EIl The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the V. C. Summer Unit i Vessel Material Surveillance Program. A-3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various . 1 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, l Kgg, for the metal temperature at that time. K IR is btained from the reference fracture toughness curve, defined in Appendix G of the ASME Code.E I The K IR curve is given by the equation: K IR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)] ( A-1 ) where K IR is the reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code [2] as follows: C K;g + Kit I KIR (A-2) where IM is the stress intensity factor caused by membrane (pressure) stress K g is'the stress intensity factor caused by the thermal gradients 1236E:lD/060785 A-2
K gg is a function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits 1 C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical l At any time during the heatup or cooldown transient, Kg , 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 temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kg , for the reference flaw are computed. From Equation (A-2), th'e pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. For the calculation of the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature,'whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K gg at the 1/4 T location for finite cooldown rates than for steady-state 123R:10/051085 A-3
operation. Furthermore, if conditions exist such that the increase in Kyp ,
)
exceeds K It, the calculated allowable pressure during cooldown will be greater than the steady-state value, i The above procedures are needed because there is no direct control on ) temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K IR f r the 1/4 T crack during heatup is lower than the K IR f r the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K IR '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 whe'n the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperatt're the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 T deep outside surface fl,aw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure
~
stresses present. These thermal stresses are dependent on both the rate of 1236E:10/051085 A-4
G heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. 4 Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a 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 lesser of the three values taken f rom the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.- i Finally, the new 10CFR50[3] rule which addresses the metal temperature of the closure head flange and vessel flange regions is considered. This 10CFR50 rule states that the metal temperature of the closure flange regions must exceed the material RT * * " " * * #' " " NOT pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for V. C. Summer Unit 1). Table A-1 indicates that the limiting RT NDT of 10'F occurs in the head flange of V. C. Summer Unit 1, and the minimum allowable temperature of this region is 130*F at pressures greater than 621 psig. A-4. HEATUP AND C00LDOWN LIMIT CURVES
- Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in Section A-3. The derivation of the limit curves is presented in the NRC Regulatory Standard 4
Review Plan.I43 i
~
i 1236E:10/061085 A-5 J
Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program. Charpy test specimens from Capsule U indicate that both the surveillance weld metal and core region intermediate shell plate code no. A9154-1 exhibited shifts in RT NDT 0*F at a fluence of 6.39 x
- 18 2 10 n/cm . This shift is well within the appropriate design curve (Figure A-1) prediction. Therefore, the heatup and cooldown curves in Figures A-3 and A-4 are based on the trend curve in Figure A-1, and these curves are applicable up to 8 effective full power years (EFPY). The heatup curve in Figure A-3 is not impacted by the new 10CFR50 rule. However, the cooldown curve in Figure A-4 is impacted by this 10CFR50 rule.
I If the Regulatory Guide 1.99 Revision 1 trend curve were to be used, the heatup and cooldown curves would then be applicable up to 11 EPPY. However, the trend curve in Figure A-1 is more realistic. As a result, the 8 EFPY curves in Figures A-3 and A-4 are recommended for normal operation. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure A-3. This is in addition to other criteria which must be met before the reactor is made critical. The leak test limit curve shown in Figure A-3 represents minimum temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of References 2 and 4. t Figures A-3 and A-4 define limits for insuring prevention of nonductile failure. P 1236E:10/061085 A-6 )
'*~
TABLE A-1 REACTOR VESSEL TOUGHNESS DATA (L"lRRADIATED) . 50 Ft-Lb miD(*} NMWD(b) 35 Mil RT Shelf Shel f. Cu P Ni NDT Temp NDT Energy- Energy Component Heat No. Grade 7. % *F *F *F Ft-lbs Ft-lbs. Closure Dome A9231-1 A5338 CL. 1 -
.009 .46 -20 '40 -20 - '106.0 Head Flange 1 5297-V1 A508 CL. 2 -
.009 .70 10 <60 10 -
129.0 V ssel Flange 5301-V1 " "
.007 .70 0 <60 0 -
172.0 Inlet Nozzle 4368-1
.005 .76 -20 <40 -20 -
130.0 436B-2
.005 .8) 0 <60 0 -
114.5 436B-3
.005 .81 -20 <40 -20 - 135.0 Outlet Nozzle 437B-1 " " -
.007 .81 -10 <50 -10 -
146.0 4378-2 " " -
.006 .80 -10 <50 -10 - 165.0 437B-3
.006 .78 0 <50 0 - 150.0 Nozzle Shell C9955-2 A533B CL. 1 .13 .010 .57 -20 78 18 - 100.5 C0123-2
.12 .009 .58 -30 86 26 -
91.0 A9154-1
.10 .009 .51 -20 90 30 136 80.5
{ Inter. Shell
.09 .006 .45 -20 40 -20 141 106.5 A9153-2 tower Shell C9923-2
.08 .005 .41 -10 70 10 161 91.5 C9923-1
.08 .005 .41 -30 70 10 148 106.0 Bsttom Hd. A9249-1
.010 .53 -40 23 -37 - 107.0 Ring Bottom Dome A9231-2
.010 .45 -10 42 -10 - 134.0
.06 .013 .89 -50 16 -44 - 84.0 Inter. to Lower Shell Girth Weld Inter. & Lower Shell Long. Welds .06 .013 .89 -50 16 -44 - 84.0
-70 -37 -70 - 130.0 Weld HAZ (a) Major Working Direction (b) Normal to Major Working Direction
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1017 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 SERVICE LIFE (EFPY) , FIGURE A-2. FAST NEUTRoft FLUEllCE (E>J itV) AS A FUfJCTIOft OF FULL PDWER SERVICE LIFE (EFPY) L . - _ - -- - -- _ _ . . _ _ _ _ _
P 3000 < I f__. 1 l .. _ ! INSERVICE LEAK' l gi ' ', TEST MINIMUM , ,
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TEMPEP W W J / J I MATERIAL BASIS: REACTOR VESSEL INTER. SHELL A9154-1
\ / f / [ [ l Cu = 0.10 wt% g/ f I
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FIGURE A-3 Virgil C. Summer Unit 1 Reactor Coolant System Heatup Limitations Applicable up to 8 EFPY A-10
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FIGURE A-4 Virgil C. Summer Unit 1 Reactor Coolant System Cooldown Limitatiors Applicable up to 8 EFPY A-ll _a
w APPENDIX A I l REFERENCES i
- 1. " Fracture Toughness Requirements," Branch Technical Position MTE8 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants. LWR Edition, NUREG-0800,1981.
i 2. ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Appendices " Rules for Construction of Nuclear Vessels," Appendix G,
" Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York, 1983.
- 3. Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements " U.S. Nuclear Regul'atory' Commission, Washington, D.C.,
Amended May 17, 1983 (48 Fede'ral Register 24010).
- 4. " Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition *:'JREG-0800,1981.
l l l i l , i 1 l A-12 l 1236E: 10/051085 L}}