ML20206E718
| ML20206E718 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/30/1988 |
| From: | Shaun Anderson, Meyer T, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17083C141 | List: |
| References | |
| WCAP-11851, NUDOCS 8811180171 | |
| Download: ML20206E718 (91) | |
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r e e WESTINGHOUSE CLASS 3 CUST!MER DESI'iNATED DISTRCUTION WCAP-11851-o l L ANALYSIS OF CAPSULE U FRCH THE l PACIFIC GAS AND ELECTRIC COMPANY DI ABLO CANYOR 'JNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM f S. E. Yanichko - S. L. Anderson L. Albertin May, 1988 Approved by: . N i bM h [ T. A. Meyer, Manager Structural Materials and Reliability Technology f Work Performed Under Shop Order LOLJ-106 Prepared by Westinghouse Electric Corporation for the Pacific Gas and Electric 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 Power Systems Division P.O. Box 2728
. Pittsburgh, Pennsylvania 15230 2728 xv. man i.
e )
A PREFACE This report has been technica1.ly reviewed and verified. Reviewer Sections 1 through 5 and 7, 8 0. J. Colburn . OM.w, Section 6 E. P. Lippincott ' [4u,//o /n- l
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TABLE OF CONTENTS Section Title Page 1.0
SUMMARY
OF RESULTS 1-1 l
2.0 INTRODUCTION
21 : l 3.0 BAC!, GROUND 3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 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 Compact Tension Tests 5-5 6.0 RADIATION ANALYSIS AND NEUTRON 00SINETRY 6-1 6.1 Introduction 6-1 - l 6.2 Discrete Ordinates Analysis 6-1 1 6.3 Neutron Dosimetry 6-4 6.4 Neutron Transport Analysis Results 6-8 , 6.5 Dosimetry Results 6-9 t
- 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1
8.0 REFERENCES
8-1
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LIST OF ILLUSTRATIONS Figures Title Page 4-1 Location of the Surveillance Test Capsules in the 4-5 - Diublo Canyon Unit 2 Reactor Vessel , 4-2 Arrangement of Sps>cimens, Thermal Monitors, and 4-6 i Dosimeters in Capsule U 5-1 Charpy V-Notch Impact Properties for Diablo Canyon 5-13 t Unit 2 Reactor Yessel Shell Plate B5454-1 (Longitudir tl Orientation) ' 5-2 Charpy V-Notch Impact Properties for Diablo Canyon 5-14 Unit 2 Reactor Vessel Shell Plate B5454-1 : (Transverse Orientation) l I 5-3 Charpy V-Notch Impact Properties for Diablo Canyon 5-15 i
~
Unit 2 Reactor Vessel Wald Metal 5-4 Charpy V-Notch Impact Properties for Diablo Canyon 5-16 Unit 2 Reactor Vessel Weld Heat Affected Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for the Diablo 5-17 Canyon Unit 2 Reactor Vessel Shell Plate B5454-1 l (LongitudinalOrientation) , 5-6 Charpy impact Specimen Fracture Surfaces for the Diablo 5-18 ; Canyon Unit 2 Reactor Vessel Shell Plate B5454-1 ! (Transverse Orientation) 5-7 Charpy Impact Specimen Fracture Surfaces for the Diablo 5-19 !
. Canyon Unit 2 Reactor Vessel Wald Metal !
5-8 Charpy Impact Specimen Fracture Surfaces for tiie Diablo 5-20 l Canyon Unit 2 Reactor Youel Weld HAZ Metal l mi.wm ie y f
LIST OF ILLUSTRATIONS (continued) Figures Title Page 5-9 Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel 5-21 Shell Plate B5454-1 (Longitudinal Orientation) 5-10 Tensile Properties for Diable Canyon Unit 2 Reactor Vessel 5-22 Shell Plate B5454-1 (Transverse Orientation) 5-11 Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel 5-23 Weld Metal i 5-12 Fractured Tensile Specimens from the Diablo Canyon Unit 2 5-24 Reactor Vessel Shell Plate B5454-1 (Longitudinal Orientation) 5-13 Fractured Tensile Specimens from the Diablo Canyon Unit 2 5-25 Reactor Vessel Shell Plate B5454-1 (Transverse Orientation) - 5-14 Fractured Tensile Specimens frem the Diablo Canyon Unit 2 5-26 Reactor Vessel Weld Metal 5-15 Stress-Striin Curve for Shell Plate B5454-1 Tension 5-27 Specimen PL1 . 5-16 Stress-Strain Curve for Shell Plate B5454-1 Tension 5-28 , Specimen PL2 ( 5-17 Stress-Strain Curve for Shell Plate B5454-1 Tension 5-29 Specimen PL3 ! 5-18 Stress-Strain Curve for ShcIl Plate B5454-1 Tansion 5-30 Specimen PT2 , NS$rN!hhh Yh
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[ LIST OF ILLUSTRATIONS (continued) Figures Title Page 5-19 Stress-Strain Curve for Shell Plate B5454-1 Tension 5-31 ; Specimen PT1 5-20 Stress-Strain Curve for Shell Plate B5454-1 Tension 5-32 Specimen PT3 5-21 Stress-Strain Curve for Weld Tension Specimen PW2 5-33 f 5-22 Stress-Strain Curve for Weld Tension Specimen PW1 5-34 5-23 Stress-Strain Curve for Weld Tension Specimen PW3 > 5-35 6-1 Diablo Canyon Unit 2 Reactor Geometry 6-22 6-2 Internal Surveillance Capsule Geometry 6 23 ( r 6-3 Calculated Azimuthal Distribution of Design Basis 6-24 Maximum Fast (E > 1.0 MeV) Neutron Flux Within the ! Reactor Vessel - Surveillance Capsule Geometry
.6-4 Relative Radial Variation of Fast Neutron Exposure 6-25 Within the Reactor Vessel 6-5 Relative Axial Variation of Fast (E > 1.0 MeV) Neutron 6-26 Flux and dpa Within the Reactor Vessel . !
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l LIST OF TABLES I Table Title Page a 3-1 Reactor Vessel Toughness Data (Unirradiated) 3-3 4-1 Chemical Composition of the Diablo Canyon Unit 2 4-3 ) Ceactor Vessel Surveillance Materials g 4-2 Heat Treatment of the Diablo Canyon Unit 2 Reactor 4-4 Vessel Surveillance Materials > 5-1 Charpy V-Notch Impact Data for the Diablo Canyon Unit 2 5-6 Reactor Vessel Shai Plate B5454-1, Irradiated at 550'F, 18 Fluence 3.51 x 10 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Diablo Canyon Unit 1 5-7 Reactor Vessel Weld Metal and HAZ Metal, Irradiated
- at 550'F, Fluence 3.51 x 1018 (E > 1.0 MeV) 5-3 Instrumented Charpy Impact Test Results for the Diablo 5-8 Canyon "nit 2 Reactor Vessel Shell Plate B5454-1, Irradiated at 3.51 x 10 18 n/cm2 (E > 1.0 MeV) 5-4 Instrumented Charpy Impact Test Results for Diablo 5-9 l Canyon Unit 2 Weld Metal and HAZ Metal, Irradiated at 18 3.51 x 10 n/cm2 (E > 1.0 MeV)
Effect of 550'F Irradiation at 3.51 x 10 10 2 5-5 n/cm 5-10 (E > 1.0 MeV) on the Notch Toughness Properties of the Diablo Canyon Unit 2 Reactor Vessel Materials 5-6 Comparison of Diable Canyon Unit 2 Reactor Vessel 5-11 Surveillance Capsule Charpy Impact Test Results With , Regulatory Guide 1.99 Proposed Revision 2 Predictions mi .4mu ie yjjj 9
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, LIST OF TABLES (continued) 4 Table Title Page 5-7 Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel 5-12 <
18 2 Material Irradiated at 550*F to 3.51 x 10 n/cm ; I (E > 1.0 MeV) . 6-1 SAILOR 47 Neutron Energy Group Structures 6-11 Nuclear Parameters for Neutron Flux M(snitors 6-12 6-2 ; r 6-3 Calculated Fast Neutron Exposure Parameters for the 6-1.3 Peak Location of thc Reactor Vessel . 6-4 Calculated Fast Neutron Exposure Parameter and Lead 6-14 Factors for the Surveillance Capsules r
! 6-5 Irradiation History of Neutron Sensors Contained in 6-15
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Capsule U
! 6-6 Measured Nonitor Activities and Reaction Rates for 6-16 l Capsule U t
! i 6-7 Calculated Neutron Energy Spectrum at the Center of 6-17
- Diablo Canyon Unit 2 Surveillanen Capsule U !
a i I, 6-8 Spectrum Averaged Reaction Cross-Sections for Use 6-18 l l in Fast Neutron Dosimetry Evaluations i t 6-9 Re=ults of Fast Neutron Oosimetry for Capsule U 6-19 &
- i Results of Thermal Neutron Dosimetry for Capsule U
, 6-10 6-20 j 6-11 Sussary of Diablo Canyon Unit 2 Fast Neutron Exposure 6-21 ( Evaluation Based on Capsule U 'esults mi.mim is j, ,
1 SECTION 1
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule U, the first capsule to be removed from the Pacific Gas and Electric Company Diablo Canyon Unit 2 reactor pressure vessel, led to the following conclusions: o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 3.51 x 10 18 n/cm2 , o Irradiation of specimens made from the reactor vessel intermediate shell plate B5954-1 to 3.51 x 10 18 n/cm2 resulted in 30 and 50 f t-lb transition temperature shif ts of 65'F and 84'F respectively, for specimens oriented parallel to the major working direction
. (longitudinal orientation) and 73*F for specimens oriented normal to
,the major working dircetion (transverse orientation). <
Specimens made from weld metal irradiated to 3.51 x 10 18 n/cm
- 2 o
resulted in 30 and 50 ft-lb transition temperature increases of
~
174*F and 199'F respectively, o Irradiation to 3.51 x 10 18 n/cm2 resulted in a 36 ft-lb decrease in the upper shelf energy of the weld metal specimens and no decrease in the upper shelf of the transverse oriented shell plate 85454-1 specimens. Bo'th materials exhibit a more than adequate upper shelf level for continued safe plant o w '. tion, o Comparison of the 30 ft-lb transition temperature increases for the Diablo Canyon Unit 2 surveillance material with predicted increases using the methods of NRC Regulatory Guide 199 proposed Revision 2 shows the plate material to be in good agreement with the Guide, whereas the weld metal transition temperature increase was 25'F higher than predicted by the Guide. 11
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- . _ . - - _ - _ _ = . - -
o Capsule U contained specimens from the same heat of weld wire (Heat i 12008 and 21935) as tha limiting reactor vessel weld seam. The , surveillance program base plate C5454-1 is representative of the j limiting reactor vessel material which is considered to be plate . B5454-2 because of its initial higher RTNDT. H wever since both j plates have very similar copper and nickel contents, changes in a properties due to irradiation are expected to be similar. 4 i i A l i 6 m i .. " " 1-2
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 Pacific Gas and Electric Company Diablo Canyon Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Diablo Canyon Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reac+ar vessel materials are presented by Davidson, Etal. (1). The surveillance program was planned to cover the 50 year design life of the reactor pressure vessel and was based on ASTM E-185-70, "Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Power Systems Division personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghoase Research and Development Center, where the
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postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed. 2 This report summarizes the testing of and the postirradiation data obtained ! from surveillance Capsule U removed from the Diablo Canyon Unit 2 reactor vessel and discusses the analyses of these data. i 1 i n 9 5 e k e v . u n u i. 21
SECTION 3 . BACKGROUND - The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 535 Grade B Class 1 (base material of the Diablo Canyon Unit 2 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 periorsing analyses to g r.ed against fast fracture in reactor pressure vessels has been presented in "Protection Against Nonductile 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 (RTNOT)* RT NOT 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-lb (and 35-mil lateral expansion) temperature as determined from Charpy Specimens oriented normal (transverse) to the major working direction of the material. The RTNOT of a given material is used to index that I material to a reference stress intensity facter curve (Kgg curve) which appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Kgg curve, allowable stress Intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing th1se allowable stress intensity factors. I w.w.e. 31 l
RT.dDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel , material properties. The changes in mechanical properties of a given resctor prossure vessel steel due to neutron radiation can be monitored by a reactor . surveillance program such as the Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Pregram (1), in which a surveillance capsule is periodically removed from the operm'ing nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 f t-lb temperature (49TNDT) due to irradiation is added to the original RTNDT to adjust the i RT NDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial + . ARTNDT) is used to index the material to the KIR curve and, in turn, to set opera'.ing limits for the nuclear power plant which take into account the effects o" irradiation on the reactor vessel materials. The fracture toughness of the Diablo Canyon Unit 2 reactor vessel material is identified in table 3-1. 'I d i w .
- m i. 3-2
TABLE 3-1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED) gg)(b) UPPER SHELF HEAT OR WATERIAL Cu Ni P T NDT RT NDT ENERGY MATERIAL CODE NO. SPEC. NO. M M M (*F) (*F) (FT LB) Closure head dome B5457 A5338 C1. 1 0.14 0.55 0.010 -20 - 5' 88' Closure head segment 85456-1 A5338 C1. 1 0.14 0.56 0.012 -50 -17 87 Closure head segment 85456-2 A5338 C1. 1 0.14 0.58 0.012 -20 22 88 Closure head se w nt 85456-3 A5338 C1. 1 0.13 0.56 0.010 -20 -20 81 C!osure head flange B5452 A508 C1. 2 - 0.69 0.009 20 20 98 Vessel flange B5451 A508 C1. 2 0.13 0.65 0.009 -10 -10 103 Inlet nozzle B5461-1 A508 C1. 2 0.09 0.70 0.012 -20 -17 75 Inlet c.zzle B5461-2 A508 Ci. 2 0.09 0.70 0.012 -20 -20 )(a) 77 Ka) Inlet nozzle B5461-3 A508 C1. 2 0.10 0.82 0.013 -40 -40 83 Inlet nozzle 85461-4 A508 C1. 2 0.10 0.81 0.013 -40 -40 84 Outlet nozzle B5462-1 A508 C1., 2 0.11 0.67 0.010 -50 -44 94 Outlet nozzle B5462-2 A508 C1. 2 0.11 0.67 0.009 -40 -26 88 Outlet nozzle 85462-3 A508 C1. 2 0.11 0.67 0.009 -50 -23 85 Outlet nozzle B5462-4 A508 C1. 2 0.11 0.67 0.009 -40 -40 89 asi.mii .
TABLE 3-1 (cont'd.) REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED) NWWD(b) UPPER SHELF HEAT OR MATERIAL Cu Ni P I NDT RT NDT ENERGi MATERIAL CODE NO. SPEC. NO. ]%1 1%1 (%) (*F) (*F) (FT LB) Upper shell B5453-1 A5338 C1. 1 0.11 0.60 0.014 0 28 82 Upper shell 85453-3 A5338 C1. 1 0.11 0.60 0.021 -10 5 89' (a) (a) Upper shell B5011-1 A5338 C1. 1 0.11 0.65 0.015 10 0 >72 Intermediate shell 85454-1 A5338 Cl. 1 0.15 0.62 0.010 -40 52 91 Intermediate shell B5458.-2 A5338 C1. 1 0.14 0.59 0.012 0 67 99 Intermediate shell d5454-3 A5338 Cl. 1 0.15 0.62 0.013 -40 33 90 Lower shell B5455-1 A5338 C1. 1 0.14 0.56 0.010 -20 -15 112 Lower shell 85455-2 A5338 Cl. 1 0.14 0.56 0.011 0 0 122 Lower shell 85455-3 A5338 C1. 1 0.10 0.62 0.010 0 15 100
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Bottom head segment B5009-1 A5338 C1. 1 0.13 0.58 0.010 0 48 >62 Bottom head segment B5009-2 A5338 Cl. 1 0.13 0.57 0.011 -10 70 >55 f(a) Bottom head segment 85009-3 A5338 Cl. 1 0.13 0.60 0.009 -20 -20 84)(a) Bo". tom head dome B5010 A5338 C1. 1 0.14 0.63 0.011 -30 -20, 74 mi m.
l./ TABLE 3-1 (cont'd.) REACTOR VESSEL IOUGHNESS DATA (UNIRRADIATED) NMWD(b) UPPER SHELF HEAT OR MATERIAL Cu Ni P T NDT RT NDT ENERGY MATERIAL CODE NO. SPEC. NO. (%) (%) 1%J (*F) (*F) (FT LB) Inter. shell longi-tudinal weld seams 21935 and 12008 (Tandem) 0.22 0.85 0.018 NA(c) -50 124 Linde 1092 flux Inter. to lower shell girth seams 10120 & Linde 0091 flux 0.04 0.03 0.011 NA(c) -56(d) NA(c) Lower shell longi-tudinal weld seams 33A277 & Linde 124 flux 0.26 0.19 0.015 NA(c) -56(d) NA(c) i NOTES: (a) Estimated per NRC standard review plan, NUREG-0800, Section MTEB 5-2 (b) Normal to major working direction (c) Not available (d) Estimated per methods in 10CFR Part 50, "Analyses of Potential Pressurized Thermal Shock Events," Vol. 50, No.141, July 23,1985. l i r - -
SECTION 4 ) DESCRIPTION OF PROGRAM ; Six surveillance capsules for monitoring the effects of neutron exposure on the Diablo Canyon Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron shield pads and the vessel wall as shown in figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. The capsules contain weld metal specirens made with the same heats of weld wire (Heat No. 12008 and 21935) and Linde 1092 flux as the limiting reactor vessel intermediate shell longitudinal weld seams 2-201A, B and C. Capsule U was removed af ter 0.99 effective full power years of plant operation. This capsule contained Charpy V> notch, tensile, and 1/2 T Compact Tension specimens (Figure 4-2) from the intermediate shell plate B5454-1 and submerged are weld metal representative of the intermediate shell longitudinal 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 B5454-1 of the representative weld. The chemistry and heat treatment of the surveillance material is presented in table 4-1 and 4-2, respectively. The chemical analyses reported in table 4-1 for unirradiated material were obtained from reference (1) used in the surveillance program. Chemical analysis of irradiated material was obtained from Charpy specimens from the weld metal and from base metal plate B5454-1. All base metal test specimens were machined from the 1/4 thickness location of the plate whereas the weld metal specimens were machined at various locations through the weld thickness. Test specimens represent material taken at least one plate thickness from the quenched end of the plate. Base metal Charpy V-notch impact and tensile specimers were oriented with the longitudinal axis w.wus ie 41
of the specimen parallel to the major working direction of the plate (longitudinal crientation) and also with the longitudinal axis of the specimen normal to the major working direction of the plate (transverse orientation). - Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the welding direction. - The 1/2 T Compact Tension 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-0.15% ccbalt (cadmium-shielded and unshielded). In addition, cadmium , $ shielded dosimeters of Neptunium (NP237) and Uranium (U238) were contained in the capsule. Thermal monitors made from the two low-melting eutectic alloys and sealed in l Pyrex tubes were included in the capsule. 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 Melting point: 590*F (310*C) The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule V, are shown in figure 4-2. l i l I I w,wese se 4-2 i 4
TABLE 4-1 CHEMICAL COMPOSITION OF THE DIABLO CANYON UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS Chemical Composition (wt%) ~~ Shell Plate B5454-1 Weld Material (d) Element Unirradiatedla) Irradiated (b) Unirradiated(a) FW9(c) FW12(c) PW15(c) C 0.23 0.243 O s '.3 0.135 0.137 0.137 Si 0.22 0.138 0.22 0.155 0.133 0.128 No 0.43 0.431 0.47 0.495 0.494 0.456 Cu 0.15 0.134 0.22 0.219 0.212 0.213 Ni 0.67 0.653 0.03 0.86 0.88 0.90 Hn 1.28 1.26 1.32 1.43 1.46 1.43 Cr 0.071 0.078 0.031 0.050 0.051 0.053 V 0.001 <0.003 0.001 <0.003 <0.003 <0.003 Co 0.002 0.010 0.012 0.019 0.018 0.016 Sn 0.010 - 0.010 - - - S 0.010 0.015 0.010 0.010 0.010 0.010 P 0.012 0.008 0.017 0.0125 0.0159 0.0146 A1 0.031 - 0.009 - - - B 0.008 - 0.008 - - -
. (a) Fro; Rsference (1).
(b) Analysis performed on irradiated Charpy plate specimen PL4 (c) Analysis performed on irradiated Charpy weld specimens i l (d) Survr,111ance wold was made with a tandem wire submerged are weld process i using the same weld wires Heat 12008 and 21935 and Li.ide 1092 Flux Lot as the aaltline region reactor vessel intermediate shell longitudinal weld seams, Linde 1092 flux lot 3869 was used to fabricate the surveillance weld at.d the intermediate shell lengitudinal weld seams. NOTE: Xray fluorosence spectrometry was used to perform analyses for metallic elenants on unirradiated specimens and inductively coupled pla*- .pe-*.rometry was used on irradiated specimens. e i c wwwwe 43 l l l
l TABLE 4-2 - HEAT TREA1 MENT OF THE DIABLO CANYON UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS Temperature Material (*F) Time (hr) Coolant Shell Plate 85454-1 1600* 1 50' 4 Water quenched 1225' 1 25' 4 Air cooled . 1150' 1 25' 40 furnace cooled ; Weldsent 1150' 1 25' 40 Furnace cooled t t E i l t i i I i t r II7 l i m i.
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Figure 4-1. Location of Surveillance Test Capsules in the Diable Canyon Unit 2 Reactor Vessel , uv.wsee ,e 45
.- .a LEGEND: PL - INTERMEDIATE $ HELL PLATE B5454-1 (LONGITUDINAL)
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1 Figure 4-2. Arrangement of Specimens, Thermal Monitors, and Dosimeters in Capsuls U Sw/Rdt7/ - 0) n :. w w i. 46 S' APERTURE CARD Also Anilable On Aperture Card
SECTION 5 TESTING OF SPECIMENS FROM CAPSULE U > t 5.1 Overview The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Power Systems Division personnel. Testing was, performed in accordance with 10CFR50, Appendices G and [ i H (2] ASTN Specification E185-82, and Westing..,use Procedure RMF-8402, [ Revision 0 as modified by RMF Procedures 8102 and 8103. a f 4 Upon receipt of the capsule at the laboratory, the specimens and spacer blocks ! were carefully removed, inspected for identification nuinber, and checked against the master list in WCAP-8783.(1) No discrepancies were found.
- Examination of the two low-melting point 304'C (579'F) and 310'C (590*F) l eutectic alloys indicated no melting of either type of thermal monitor. Based l on this exa'nination, the maximum temperature to which the test specimens were exposed was less than 304*C (579'F). l l
) The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Prucedure 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 te general yielding (tgy), the maximum load (Pg ), and the time to ! raximum load (tg) can be determined. The load at which fast fracture was initiated is identified as the fast fracture load (P ), pand the load at 4 [ which fast fracture terminated is identified as the arrest load (Pg ). [ F
, The energy at maximum load (E g ) was determined by cwparing the enargy-time [
record and the load-time record. The unsrgy at maximum load is roughly !- equivalent to the energy required to initiate a crack in the specimen. i t i i 3051,.W6M 10 5-1 t 1
Therefore, the propagation energy for the crack (E p
) is the difference between the total energy to fracture (E )D and the energy at maximum load.
The yield stress (,y, is calculated fre'n the three point bend formula. The flow stress is calculated from the average of the yicld and maximum loads, - also using the three point bend formula. Percent shear was determined from postfracture photographs using the ratio-of-areas metnods in compliance with ASTM Specification A370-77. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification. Tension tests were perfor d on a 20,000 pound Instron, split-console test machine (Model 1115) per k . Specifications E8-83 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were made of Inconal 718 hardened to Re 45. The upper pull rod was connected through a universal joint , to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test. Deflection measurements were made with a linear variable displacement , transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 l
! per ASTM E83-67.
i Elevated tect 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 l specimen, the following procedure was used to monitor specimen temperature. Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage saction of a dumy specimen and in each grip. In the test configuration, with a slight lead 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 i i xi " ** "" " 5-2 ,
was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain dtsired specimen temperatures. Experiments indicated that this method is accurate to + 2*F. The yield lead, ultimate lead, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yiel 1.0 MeV) SAMPLE NO. TEMPERATURE IMPACT ENERGY LATERAL EXPANSION SHEAR (*F) ('C) (ft-lbs) (Joules) (mils) (mm) (%) Longitudinal Orientation PL9 0 -18 5.0 7.0 6.5 0.17 5 PLIS 50 10 20.0 27.0 20.0 0.51 15 PL10 60 16 18.0 24.5 19.0 0.48 15 PL3 75 24 40.0 54.0 36.0 0.91 20 PL4 75 24 36.0 49.0 29.0 0.74 20 PL11 76 24 48.0 65.0 38.0 0.97 25 PLS 100 38 41.0 55.5 38.0 0.97 25
?L14 100 38 27.0 36.5 27.0 0.69 20 Pil 125 52 74.0 100.5 58.0 1.47 60 PL12 125 52 39.0 53.0 38.5 0.98 25 PL2 150 66 63.0 85.5 54.0 1.37 55 '
PL8 200 93 90.0 122.0 70.0 1.78 75 PL13 275 135 113.0 153.0 84.0 2.13 100 PL7 350 177 117.0 158.5 62.0 2.08 100 . PL5 425 218 125.0 169.5 90.0 2.29 100 Transverse Orientation PT9 0 -18 12.0 16.5 13.0 0.33 5 PT5 50 10 22.0 30.0 20.5 0.52 10 PT13 75 24 23.0 31.0 22.5 0.57 20 PT6 100 38 27.0 36.5 28.0 0.71 20 I PT12 100 38 29.0 39.5 24.5 0.62 20 PT2 110 43 39.0 53.0 37.5 0.95 30 i PT3 125 52 47.0 63.5 43.0 1.09 35 PT4 125 52 40.0 54.0 38.0 0.97 35 PT8 150 66 37.0 50.0 39.0 0.99 35 , PT7 160 71 54.0 73.0 49.0 1.24 45 PT14 200 93 55.0 74.5 52.5 1.33 50 PT1 225 107 106.0 143.5 75.0 1.91 100 PT15 275 135 92.0 124.5 74.0 1.88 100 < PT10 350 177 90.0 122.0 78.0 1.98 100 PT11 425 218 86.0 116.5 71.5 1.82 100 .
, i nu.ev >u ie - 5-6
TA8LE 5-2 CHARPY V-NOTCH IMPACT DATA FOR DIAPLO CANYON UNIT 2 REACTOR VES*"'. atLO METAL AND HEAT-A IRRADIATEDAT550'F, FLUENCE 3.51x10{[ECTE0 n/cm2 (E > 1.0ZONE MeV) (HAZ) KETAL, SAMPLE NO. TEMPERATURE IMPACT ENERGY LATERAL EXPANSION SHEAR ('F) ('C) (ft-lbs) (Joules) (mils) (mm) {%1 Weld Metal PW13 25 -4 3.0 4.0 4.0 0.10 2 PW4 75 S4 11.0 15.0 12.0 0.30 10 PW7 125 52 24.0 32.5 29.0 0.74 20 PW8 150 66 26.0 35.5 29.0 0.74 25 PW12 150 66 28.0 38.0 30.5 0.77 30 PW1 175 79 32.0 43.5 33.0 0.84 50 PW6 175 79 40.0 54.0 38.0 0.97 65 PW2 200 93 35.0 47.5 36.0 0.91 60 , PW3 200 93 41.0 55.5 39.0 0.99 70 1 PW9 225 107 54.0 73.0 51.0 1,30 75 PW14 225 107 60.0 81.5 58.0 1.47 80 : PW11 275 135 72.0 97.5 65.5 1.66 90 ; PW10 350 177 78.0 103.0 71.0 1.80 100 PW5 425 218 90.0 122.0 75.0 1.91 100 PW15 475 246 78.0 106.0 73.0 1.85 100 HAZ Metal
. PH9 -100 -73 2.0 2.5 2.0 0.05 2 i PH3 -50 -46 26.0 35.5 20.0 0.51 15 PHS -25 -32 MACHINE MALFUNCTION t
PH10 -25 -32 15.0 20.5 13.5 0.34 30 PH1 0 -18 44.0 59.5 33.5 0.98 50 PH6 25 -4 33.0 44.5 27.0 0.69 55 PH4 50 10 59.0 80.0 44.0 1.12 65 ; PH13 75 24 47.0 63.5 36.0 0.91 60 PH14 76 24 35.0 47.5 32.5 0.83 50 t PH15 125 52 36.0 49.0 40.0 1.02 70 ' PH7 125 52 48.0 65.0 47.0 1.19 70 PH12 200 93 62.0 84.0 58.5 1.49 95 l PH2 250 121 98.0 133.0 72.0 1.83 100 : PHS 350 177 84.0 114.0 70.0 1.78 100 PHil 425 218 77.0 104.5 73.0 1.85 100 ; L t i i
*l l
l mswwe 5.y i
i a TA8tE 5-3 INSTRUP.NTED CHARPY IMPACT TEST RESULTS FOR THE DI A8LO CANVON UNIT 2 REACTOR VEs5EL SHELL ? LATE 36454-1 IRRADIATED AT 3.51 x 10'8 n/cm 2 gg . 1.0 mov) i 4 ! NG:MBALI7f D EW2GIES ! TEST OtAAPT CH'a'llEV ERKIEUM resur VIELD TIIIE IIAXIMUM TIME TO FRACTURE AmmEST VIE
- D FLOM 4
SAMPLE TEMP ENEfcGY Est/A Em/A Ep/A LOAD TO VIELD LOAD mRxIssues LOAD LOAD ST8ESS STRESS ] NUW R (* F ) (ft Ibs) (ft-Ibs/1p2) (ktps) (tJ5ec) (ktps) (uSec) (ktps) ikips) (kst) (kst) LONGITUDINAL OR*ENTATION
- 120 3.95 108 820 W Pt 9 0 5.0 40 35 6 3.30 90 3.95 -
l PL15 50 20.O 161 139 22 3.40 85 4.10 325 4.10 O.15 112 124 li PL10 00 18.0 145 102 13 3.15 95 3.90 270 3.05 0.90 103 116 J P11 75 40.0 322 279 43 3.35 85 4.60 500 4.60 0.40 111 132 PL4 75 36.0 290 233 50 3.35 85 4.55 510 4.55 0.55 111 131 l PL11 76 48.0 387 317 49 3.60 80 4.95 650 4.85 0.40 119 141 l PL6 100 41.0 3JO 2f3 44 3.60 85 4.80 575 4.80 0.70 118 138 l PL14 100 27.0 267 172 45 3.30 is 4. 2($ 395 3.75 0.70 110 125 d PL1 125 74.0 596 304 292 3.15 80 4.45 650 4.05 1.90 104 126 PL12 125 39.0 314 22- 91 3.15 85 4.30 505 4.20 1.05 104 123 PL2 150 63.0 507 ;'02 205 3.15 90 4.40 665 4.20 1.55 104 125 1 un PLS 200 90.0 725 191 434 2.95 80 4.25 6*5 3.40 2.05 98 119 i e PL13 275 113.0 910 283 627 2.85 100 4.15 665 - - 94 116 I
- PL7 350 117.0 942 281 666 2.35 63 4.05 665 - -
78 106 1 l PL5 425 125.0 1007 275 732 2.35 55 3.85 670 - - 78 103 TPA,45VEPSE ORIENTATION PT9 0 12.0 97 107 -11 3.65 30 4.10 255 3.8S - 120 .M PTS 50 22.0 177 142 35 3.45 85 4.15 330 4.10 0.30 114 12t, PT13 75 23.0 185 112 73 3.40 90 4.05 275 4.05 0.80 113 164 P16 100 27.0 217 164 53 3. 4'i 85 4.45 360 4.45 1.15 114 131 i PT12 100 29.0 234 186 47 2.90 60 4.50 405 4.40 0.95 96 122 l PT2 110 39.0 314 214 100 2.95 95 4.20 505 4.20 1.35 98 119
- PT3 125 4?.O 378 225 153 3.to 80 4.30 505 4.20 1.30 103 123 PT4 125 40.0 322 221 101 2.95 75 4.25 500 4.15 1.15 97 119 I PT8 150 37.0 298 189 109 2.95 85 4.10 450 4.05 1.20 98 117 i PT7 160 54.0 435 254 131 3.00 85 4.15 E30 4.10 1.50 100 11F PT14 200 55.0 443 212 231 2.95 90 4.10 505 4.00 2.90 98 til Pil 225 106.0 854 295 559 3.00 65 4.30 655 - -
99 129 ! PTIS 275 92.0 741 252 489 2.75 95 4.05 605 - - 90 112 PTIO 350 90.0 725 254 470 2.55 60 4.05 610 - - 84 108 PT11 425 86.0 692 232 460 2.45 60 3.80 580 - - 81 104 1 4 I i 1 1 3051s/051ft88 :10 I S O $ O g
TABLE 5-4 , INSTRutIENTED CHARPY IldPACT TEST RESULTS FOR THE DI ABLO CANYON UNIT 2 18 WELD IEETAL AND HEAT-AFFECTED ZONE OtAZ) W TAL. IRRADIATED AT 3.51 x 10 n/cm2 (E > 1.0 IseV) NDR1 PALI 7E3 ENERGIE$ TEST CHARPY CHARPV RRAXI RAdG PROP VIELD TIME alAXIasuas TIasE TO FRACTURE ARPEST VIELD FLOW TO VIELD LOA 1 alAXIIsuas LOAD LOAD STRESS STRESS SAasPLE 1 E RAP ENERGY Ed/A Em/A [p/A LA)AD NUSANE R (*F)- (ft 1bs) (ft-ids /In ) (ktps) (uSec) (hips) (uSec) (ktpt) (k105) (kS1) (kS1) Weld teetal 25 3.0 24 23 2.75 $5 3.55 85 3.55 0.15 90 103 Pw13 1 PW4 75 11.0 89 65 24 3.70 90 4.00 170 0.45 - 12? 127 Pw7 125 24.0 193 166 28 3.20 80 4.25 375 4.25 0. M 106 123 Pw8 150 26.0 209 142 67 3.05 75 4.10 335 4.10 2.60 100 118 Pw12 150 28.0 225 154 72 3.40 *5 4.20 360 4.15 1.30 112 126 PWI 175 32.0 258 174 84 3.40 90 4.20 400 4.20 1.60 112 126 Pw6 175 40.0 322 191 131 3.35 95 4.20 435 4.20 2.T3 lit 125 Pw2 200 35.0 282 189 93 3.35 SO 4.20 430 4.20 2.90 111 125 PW3 200 41.0 330 197 134 3.35 95 4.20 450 4.15 2.20 111 125 Pw9 225 54.0 435 231 204 3.05 100 4.40 520 4.35 2.70 tot 123 Pw14 225 60.0 483 212 271 2.95 100 4.15 505 4.10 3.60 97 117 Pw11 275 72.0 580 218 362 3.05 95 4.15 510 - - 101 119 PwlO 350 78.0 628 213 415 2.90 85 4.05 505 - - 95 114 Pw5 425 90.0 725 213 512 2.90 85 4.00 505 - - 96 114 600 E', 108 Y m Pw15 475 78.0 628 238 390 2.65 110 3.85 - - HAZ sset a t PH3 -100 2.0 16 10 6 . 85 45 2.55 55 2.55 0.30 61 73 PH3 -! ' 26.0 209 185 24 4.10 90 4.80 370 4.75 0.25 136 148 Pt6 -25 alACHINE MAaJUNCTION PH10 -25 15.0 121 97 24 3.90 360 4.15 435 4.15 0.G5 129 133 PH1 C 44.0 354 249 105 3.75 90 4.80 505 4.70 1.70 125 141 PH6 25 33.0 266 193 73 3.65 85 4.fo 405 4.60 1.45 121 137 PH4 50 59.0 475 286 189 3.50 105 4.t0 605 4.35 2.25 116 134 PH13 75 47.0 378 238 140 3.50 85 4.60 500 4.45 2.60 116 134 PH14 76 35.0 282 213 $9 3.70 100 4.85 435 4.80 2.25 122 141 PH7 125 48.0 387 155 232 3.30 85 4.10 360 4.10 2.35 110 123 PH1r 125 36.0 290 142 147 3.20 80 4.10 335 4.10 2.6G 106 121 PH12 200 62.0 499 215 284 3.05 85 4.10 500 - - 101 188 PH2 250 98.0 789 297 492 2.70 60 4.35 655 - - 89 117 PH8 350 84.0 676 259 4:a 2.60 55 4.05 610 - - 85 109 - PH11 425 77.0 620 202 418 2.25 55 3 85 510 - - 74 101 i 3051s/05 sas:iO
. ..14 TABLE 5-5 18 n/cm2 (E > 1.0 Iney)
EFFECT OF 550*F IRRADIATION AT 3.51 m 10 ON NOICH TOUGesE55 PROPERTIES OF DIABLO CANYON UNIT 2 REACTOR VESSEt. uATERIALS Average 35 mit Average 50 f1-lb Lateral Espansion Average 30 ft-tb average ope.4r Temperat ure ( *F 3 Temperature (*F) Temperat ure (*F ) Shelf Energy (ft-Ib) stat er tal Untrrad!ated Irradlated AT untrradiated Irradsated at unteradtated Irradiate 3 AT Untreadiated Irradiated A(ft-Ib) Plate 85454-1 33 117 84 18 93 75 5 70 65 144 124 20 (L W itudinal) Plate 85454-1 74 147 73 54 116 62 26 99 73 95 94 1 (Transverse) weld metal 15 214 193 -1 17(* 171 -13 160 174 128 85 36 HAZ uetal - 103 - - 59 - - 7 - 148 88 60 8 trl e l>* O O 3051s/051888:10 8 6 e ,
, ,-_.-.--.n_ _ , ,
TABLE 5-6
. COMP:RISON OF DIABLO CANYON UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS WITH REGULATORY GUIDE 1.99 PROPOSED REVISION 2 PREDICTIONS Fluence ARTNOT(*F) USEDecrease(%)
18 2 Material (10 n/cm ) Mess. Pred. Mets. Pred. Plate B5454-1 (Long.) 3.51 65 74 15 18 Plate B5454-1 (Trans) 3.51 73 74 1 18 Weld Metal 3.51 174 149 20 26 P 5-11
UMI .M M TABLE 5 7 TENSILE PROPERTIES FOR DI Act O CANYON UNIT 2 RE ACTGR '.TSSEL MATERI AL 18 3 5tR ADI AILD AT 550*F TO 3.51 a 10 n/cm2 (E > 1.0 Mev)
.2% VIELO ULTIMATE TRACTUHt F R ACTLT
- FRACTUR. UNIFORM TOTAL REDUCTION TEST 1E MPf RATURE STNEhCTH SYR!NGTH LOAD STRESS ST RE NGT H ELONGATION ELONGATION Iff AREA SAMPLE % %
*F k%1 kip kS1 kSS %
NUWftf R MAffRIAL 450 PL1 05454-1 ( Lorig. ) 79 72.9 94.7 3.00 173.0 61.1 12.8 26.9 65 PL2 85454-1 (Long.) 225 08.2 87.6 2.95 163.9 00.1 9.8 20.4 63 PL3 854s4-1 (Long.) 550 63.2 88.6 2.90 149.6 59.1 10.5 22.5 61 202.6 66.2 11.3 22.8 67 PT2 85454-1 (Tr.ns) 78 74.4 94.7 3.25 162.5 64.2 9.8 20.4 61 PT1 B5454-1 (Trans) 225 63.3 88.6 3.*5 63.7 89.5 160.0 70.3 11.1 18.9 56 PT3 B5454-1 (Trans) 550 3.45 77 84.5 101.7 3.20 223.5 67.2 12.0 24.6 70 g PW2 Weld 77.9 93.7 3.1C 133.3 63.2 10.8 21.9 67 4 PW1 Weld 300 19.2 53 Pw3 weld 550 73.3 95.7 3 . 's0 164.1 67.2 9.8 I 3051s/051888:10 O G S g
l-I l e a 180 - 194 - ( oo - sa . I v= 8 o
$ i
(: SS - O U N O "4 g A so . .
; A SO - e O ,
1
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- -see -soo soo .
Q untrrestatte irr ast a tes 100
- O I j ,
a*n - j 136 - 180 - , j nso - a
~
l 800 - J
- a- i i Se = l
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I l H- a
- += -
., y.
O i e , , m m , , , , <
-See -1SG &De 300 SGS P t
l To sus per e t ta r e (F) I a u nier.4 6.i.4 . ice.4i.... , I f Figure 5-1. Charpy V-Notch Impact Properties for Diablo Canyon Unit 2 I < . I j Reactor Yessel Shell Plate B5454-1 (Longitudinal Orientation) ! I i mawee 5-13 ! 1 1 4 I i d 6 { ' l _ _ - - - , , _- = _ - - - . _ . - ~ , , _ _ _ . _ _ _ , , - _ _ _ _ _ _ - - , _ - _ . _ . ._ _
r. o
- l l
1 . u. l l O l
- 8 . _.
} _
9 ~V A f v. - l A U D l H= ' A A l a- g A
. (
i *
*s-s/ - ,
3 g & go. T.mp.r.tur. (F) l- A U ns e r .di .t.d e IFF#488%'d I an. A 4-O M 'A B ^ j w-g ,. - , l w- a a A
- i A ,
'b I ** ~
u-y R
. -a . . .
a unier.. .... * * " ' ' ' " " ' * 'I' irr.. .... Figure 5 2. Charpy V-Notch Impact Properties for Diablo Canyon Unit 2 Reactor Yessel Shell Plate B5454-1 (Transverse Orientation) t noiswsee to 5-14 7
- e j t
a l e ggg e 194 - g
=- 8 g l O ,
90 - , k A
~
79 - SG - A U DG y
..q .
SG [g', A
.. - a/
ae - " -
" A e , , -t i i > i
- age &96 ### 088
** 1 DO Temperature (F)
D Unterndtened a trradiased a
- 33. - r
" H . =. - ~
ise - a ;
- soo -
A, OO " A b 96 - , re - ! M , A f
.. - l /
4 i S?6 *
- y, ,A b a o
- r, s . a m
e--- , . ; , , , .
-see -see so. ,
a voire.. .... **" " "'* 'I) ire.. .... I i I i l Figure 5-3. Charpy V-Notch Impact Properties for Diablo Canyon Unit 2 Reactor Vessel Weld Metal f l mi.ww. 5-15 i I
, ,
a u.arr.4i. ..
- ice.4i.i.4
( l 4 L i i l !
. 1
} Figure 5-4. Charpy V-Notch Impact Pro;:erties for Diablo Canyon Unit 2 ) Reactor Vessel Weld Heat Affected Zone Metal - p I l l w.w . 5 33 I 5 i i
l i l
- l l
l l l
; w l r, *- [ .$k M 'i -
p g ? , o. ,
'a 's . . ., l' v gr '..
i .
. s .
1 l l
. ,, , ~o c , . , , fg-i ),,
,i j).,Y .Qid.% -
Q, .j. ,
- i .
, , *',- 1 *: g.s..-.j ,
o , _n - I PL9 PL15 PL10 PL3 PL4 J , l 1 i l (
' .'. i 7, -4 ) '*
, ' A-
, I, .s] !,ac i ,
. ,, a
.l l 1 L, ,' ] > a ,2.ip. ,. O.
,.,,m. -
3e*< (,L .4 -a l .l l , D-EA, l 5-- - k 29 , 1 i 3 ) s A- ,
'O PL11 Pt 6 PL14 PL1 PL12 i
F O Q,
, {- .
y{} ',
$2. r g@. ,g.E .
y ,/ ' I' l __- . w . .- - PL2 PLS PL13 PL7 PL5 i 1 5 i Figure 5-5. Charpy Impact Speciman Fracture Surfaces of the Diablo Cany:n Unit 2 Reactor Vessel Shell Plate B5454-1 (Longitudinal
) ! Criantation) l M116 tt'tas t 3 '
5-17
,, . us ;
1 i l \
e i l t 1
- t * .
uw QJ L.f=1 .__ ' L e ~~ l'. V.l [' : ;
! PT9 PT5 PT13 PT6 PT12 i
l
@lh ,7; 1yW.:
'( 1 ,
}
7, W l "l c'..'3 i- ?,y :- t
' , l nm ;tt 3: ,
* :- s u i
}m .w: p s _i s
~.
I um
. s h,'
1,- I 1.k..(Css i , ,
, : . u ,- j-- -
-3 ,
u .- -,- w, , i 1 l PT2 PT3 PT4 PT8 PT7 l t l l ,
},-~
, . ,. < C -l f .
7 ,
.., . v_
h p 1i{ i
._he
' - - ' ' a
_, M y
- 6 ;
' n w - ,, , _.,,
.i *s .
t l ts - PT14 PT1 PT15 PT10 PT11 Figure 5-6 Charpy Impact Specimen Fracture Surf aces of the Diablo Canyon ! Unit 2 Reactor Vessel Plata Shell S5454-1 (Transverse Orientation) xv.wm . 5-18
,, . s s . n L
)
9
- i. ,
';g kU,k'.]? c .p j
- ,~
.. h: ,
': ;x, p.. ~
. ) 0*-:.
\
~'
i
=-
K' ' Q , p i' 4 , l ' *f :;.. ft6 i
~ -
w -- -
.#13 PW4 Pn7 Pn8 P#12 l
t
~
-)
t , 'l r
.~~ ,
* ,% (
's -
-(-; .j c; Q;, .1 .
l
. ~ s m '. '
1 . . w m e- ! PW1 Pn6 PW2 PW3 Pw9 ! l l I ! i J l ) w i l ~ ' h_ . %= ' '
\ l
; _' l;5 5 . . .=.
l vc= .
'-q . ,
, 7
- q. ,_
l { ^ sh - 4' to. - : ,_ f7 , j i P414 Pell PA10 P45 P415 Figure 5-7. Charpy Inpset Specieen Fracture Surfaces of the Diablo Canyen ; )- Unit 2 Reseter Vessel Weld Metal l rs' m m is 5-19 . I i < ey . , < > \ , i i
- - _ . . - - - - . - - - __ - _-_-. -_ - - -_ - - - _ __ _ - - _ - . _a
" )*., ~ . , ,
..c llo,
'
- Q~. zi I, i e
, \V 'e x.
). -.
.[ L ,- .'! .,
PH9 PH3 PH10 PH1
"""*""** - , g % ~ ' %
,4
,.; I
\fi } s t,-\ !
a '. . .. '.~ < \1
> Nie tn -gM. ..,) -
L. - W - bg ..- y W '*i' - . b (
~ ,
s j _. fs
~. -_ . .
PH6 PH4 PH13 PH14 PH15 l y 5
- n. '
M {' O j g ' 'r. sg , , t 4 s'f ti!*O o m -, . t) QW'
' ~-
.7 e
l ~.s .. ;'... ,'
. s' f>~
- - M "2 %
PH7 PH12 PH2 PH3 PH11 Figure 5 8. Charpy Impact Specimen Fracture Surfaces of the Diablo Canyen Unit 2 Reactor Vessel Weld HAZ Petal
>:n w ... i s 5-20 i
,,._,,,o
7 e e curve 755521-A
'C
- 50 0 50 100 150 200 250 300 120 , , , , , , , , _ g
. 110 -
4 100 - 700
$90 -
- k. T f ensile Strength ,
W
~
~ V. A 5 80 -
a E g v -- g70 - r0. 2 % Yleid Strength 500 - M - S N3 '[ _ a 50 40 i i i i i i i- 300 Code : Open Points - Unirradiated I8 2 Closed Points - Irradiated to 3,51 x 10 n/cm M i i i i i i i i 70 - o -
. g -
8 ; 'M . - 7 g - Reduction in Area _
$ 40 Total Elongation
]30 -
2
-t- t 20 - -
If, - ; C ? -
' I i i Unifo,rm Elongation ,
0
~100 0 100 200 300 400 500 600 Temperature ( *F)
Figure 5-9. Tensile Properties for Diablo Canyon Unit 2 Reactor Vessel Shell Plate B5454-1 (Longitudinal Orientatien) 5-21
Curvo 755522-A
*C
- 50 0 50 100 150 200 250 300 ,
120 i i i i i i , i yg 110 - 100 - 700
$90 - k . >
~
80 - Tensile Strength-{ [ g{ ( 1: N - 20 - m 70 t 60 - Qeld Strengthy ' - 400 ! y - 4t) i i i i i i _ I _ . 300 i Code : ! Open Points - Unirradiated ! 18 2 l Closed Points - Irradiated to 3. 51 x 10 n/cm r f 80 i i i i i i i i
'O 60 g_
o
- ~a Redur'lon in Area
~
'[
-gg i v
g I -2> o r, E 40 - - j 30 - b . Total Elongation A i m - W
= -
z 10 - e 4 - - 0 , , , , UnMorp Elongon , j
-100 0 100 200 300 400 500 600 l Temperature ( *F) !
l i figure 5-10. Tensile P certies for Diablo Canyon Unit 2 Reactor Vessel , ; Shell Plate B5454-1 (Transverse Orientation) l
. 1 l
mw = io r.22 i
Curvo 755520-A
'C
- 50 0 50 100 150 200 250 300 ..
120 i i i i i i i i _ g 110 - _ 100 ( = 700 r 3% 600g
- Tensile Strength g _
- E g A e ? -
500 - b m 70
- 0. 2 % Yleld Strength i g _
a 50 i i i i i .L i- 300 40 - Code : Open Politts - Unirradiated 2 18 Closed Points - Irradiated to 3. 51 x 10 n/cm 80 i i i i i i i i 70 - 9 Qon in Area
~
60 - 9 -
. vt 50 x -
E 40 - Total Elongation 1 d 30
-- a_
v i . t M - ; I - 10 - Y 9 8 - ! i i i i Unifoqm Elongition i O
~100 0 100 200 300 400 300 600 Temperature (*F)
. Figure 5-11. Tensile Preperties for Diablo Canyon Unit 2 Reactor Vessel l i
Weld Metal j . m.wme " 5-23
e Soecleen PL1 78'F Specimen PL2 225'F Specimen PL3 550*F Figure 5-12. Fracturo Tensile Specimens from the Diablo Canyon Unit 2 ' Reactor Vessel Pla*e B5454-1 (Lengitudinal Orientation) w . * *a ' ' 5-24 LM- 15 d e
]
4 1 . . l I ! l l 1 . I \ 's !I,ElI{-i l (
~
) - '
?. .
s .4,7 I u 7 = .- y , l, ' q .,e
.ca i 1
i' i
- c. >
l j I "pecimen PT2 78'F : I ! l
, l
! . . , .._. ~ i
. , ;e .. ,l1
..'nl,- !
l j =~= - , ,
-. I
,r.- _u r I i .
s -
-, l l
Spee.imen PT1 225'F ; 1 i i
, - < .++ }
Mijf}
- tnw. - .
., ., l c.. l
, ? > I 4 I Specimen PT3 550'F 1~ 1 i Figure 5-13. Fracture Tensile Scecimens from the Diablo Canyon Unit 2 I Reactor Vessel Plate B5454-1 (Transverse Orientation) xmw n. i. 5-25 ! m.1, o , I , i 1 l
i i 1 . . ! i 1
~
,, , .;. ', . 's {T i T. "
{
3 ~$ -
j 4 ,9 7' e >
.s- '
\ Y
, .1 I
l 1 l' . 1 l 3
.. O- , , 'n i j , ,
[ , d3 l
.j . . : . . r.
i Specimen Pw2 77'F .1 L i I 1
, , l
'g .
j , . l e
;: 4 . 4 q
1 I ! <,,,a - , - -. ; j
. = >w41 l
l Specimen PU 300'F ' I l ! o-1 i I 6 i Specison P63 550'F : ! l 1 l Figure 5-14. Fracture Tensile Specians from the Diablo Canyon Unit 2 Reactor Vessel Weld Petal \ l l ) l .:w w u. 5-26 , 1LM-! ! t 3 4 l l
120 , , , IM - 80 - 2 60 - a . E mM - 20 - 0 ' i -- > > 0 0,05 0,10 0,15 0.20 0,25 Strain, in/in Figure 5-15, 5 tress Strain Curve 'or Shell Plate 85454-1 Tension specimen PL1 w ** *
- 5-27
120 . , , , IM - 80 - -
=
a m . 10 5 40 ~ - 20 - 0 ' ' ' ' 0 0.05 0.10 0.15 0.20 0.25 Strain, in/in t i Figure 5-16. Stress-Strain Curve for Shell Plate 85454-1 Tension - Specimen PL2 f m i. w m io 5-28
O 120 , , , 1% - 80 - m b;,, a 60 - N : E . mM - , . M - 0 ' ' i i 0 0.05 0.10 0.15 0.20 0.25-Strain, in/in t i t i
- Figu)e 5-17. Stress-Strain Curve for Shell Plate B5454-1 Tension !
Specimen PL3 ;
. l i
M10MIM410 "[3 *k h l l i L
l H 120 , , , 1% - 80 - E 60 - Oi 2 .
, m 40 , . _
I N - 0 ' ' ' ' O 0.05 0.10 0.15 0.20 0.25 l Strain, in/in f ( l l : l l Figura 5-18. Stress-Strain Curve for Shell Plate B5454-1 Tension - ! Specimen PT2 m.mim i. 5-30 l l L
{ 120 , , , , IM - - l 80 - l D m - Oi 2
. m@ -
0 ' ' > i 0 0,05 0,10 0,15 0,20 0,25 Strain,in/in l l l l l Figure 5-19. Stress-Strain Curve for Shell Plate B5454-1 Tension Specimen PT1 xueue ie 5-31
e 120 , , , , IM - 80 - G 60 .- a E 5@ - M _ 0 ' ' ' i 0 0.05 0.10 0.15 0.20 0.25 Strain, in/in ' I L 1 l i L t l
~
Figure 5-20. Stress-Strain Curve for Shell Plate B5454-1 Tension Specimen PT3 mi.mim in 5-32 ' l
120 i i i i 1M - 80 - D 60 .- E i
. m M -
~
M _ 0 ' ' ' ' i l 0 0.05 0,10 0.15 0.20 0.25 Strain, in/in i l i Figure 5-21 Stress-Stra?' Curve for Weld Tension Specimen PW2 ! i i u.* m i. 5-33 : 1 1 4
---=,e---- - - - - - , - - - - --n,, ,----,
e
-l l 120 , , , ,
IM - - r 80 - - D so .. - N E M 40 - - - m _ _ . , 0 ' ' ' ' 0 0,05 0,10 0,15 0. 20 0. 25 Strain, in/in . i Figure 5-22 Stress-Strain Curve for Wald Tension Specimen PW1 w.misa is 5-34 l
t 120 i , , , IM - 80 - 1 E 60 -
. - i vt i E
, M @ -
l t N -
- l 4
I i i 1 0 ' ' ' i 0 0,05 0,10 0,15 0,20 0,25 . Strain,in/in l 1 t l
- l i !
i ! a ) { Figure 5-23 Stress-Strain Curve for Weld Tension Specimen PW3 I t l
- I L
w.w a. ie 5-35 ' .' 1 I t i
SECTION 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6.1 Introduction Knowledge of the neutron environment within the pressure vessel-surveillance capsule geometry is required as an integral part of the LWR pressure vessel surveillance programs for two reasons. First, in the interpretation of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relating the chenges observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the environment at various positions within the reactor vessel and that experienced by the test specimens must be mtablished. The former requirement is normally met by employing a combination of rigorous analytic &l techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information, on the other hand, is derived solely from analysis. This sect'.on describes a discrete ordinates Sn transport analysis performed for the Diablo Canyon Unit 2 reactor to determine the fast neutron (E > 1.0 MeV) flux and fluence as well as the neutron energy spectra within the reactor vessel and surveillance capsules; and, in turn, to develop data for use in relating neutron exposure of the pressure vossel to that of the surveillance capsules. Based on neutron spectrum data derived from these computations, the analysis of the. neutron dosimetry contained in Capsule V is discussed; and a correlation with neutron dosimetry from the reactor cavity (5) is provided to aid in the assessment of the fast neutron exposure of the reactor vessel itself. 6.2 Discrete Ordinates Analyris . A plan view uf the Diablo Canyon Unit 2 reactor geometry at the core midplane is shown in figure 6-1. Since the reactor exhibits 1/8th core sycrnetry, only a 0*-45' sector is deoicted. Six irradiation capsules attached to the neutron pad are included in the design to constitute the reactor vessel surveillance w.w ui i. 6-1
i program. Four capsules are located symmetrically at 56', 58.5*, 124', 236', 238.5' and 304' relative to the core cardinal axes of 0'. A plan view of a single surveillance capsule attached to the neutron pad is shown in figure 6-2. Th'e stainless steel specimen container is 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. From a neutronic standpoint, the surveillance capsule structures are i significant. In fact, they have a marked impact on the distributions of j neutron flux and energy spectra in the water annulus between the thermal shield and the reactor vessel. Thus, in order to properly ascertain the neutron environment at the test specimen locations, the capsules themselves l ! must be included in the analytical model. Use of at least a two-dimensional computation is, therefore, mandatory. t In the analysis of the neutron environment within the Diablo Canyon Unit 2 reactor geometry, two sets of transport calculdtions wirre carried out. The first, a single computation in the conventional forward mode, was utilized . primarily to obtain spectrum-averaged reaction cross sections and gradient corrections for dosimetry reactions. The second set of calculations consisted of a series of adjoint analyses relating the fast neutron (E > 1.0 MeV) flux at the surveillance capsule center, the dosimetry positions in the reactor l cavity, and selected locations on the pressure vessel inner diameter to the power distributions in the reactor core. These adjoint importance functions, when combined with cycle-specific core power distributions, yield the l I plant-specific fast neutron exposure at the surveillance capsule and pressura vessel locations for each operating fuel cycle. The forward transport calculation was carried out in R, e geometry using the
- 00T two dimensional discrete ordinates code (6) and the SAILOR cross-section library (7). The SAILOR library is 47 group, ENDF-BIV based data set produced .
1 specifically for light water reactor application. Anisotropic scattering is ;
; treated with a P3 expansion of the cross-sections. The energy group .
structure used in the analysis is listed in table 6-1. j mwwm ee gg
-. ,,,,.,.,.-,n , , , - . - - . - - - - - - - - - - , - . . - - - -- - , . - - - . . - . ~ . , - - , ~ . - - - - - - - - , ,
4 1 The design basis 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 design basis core power i distribution is.the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a 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 single reactor would have a power distribution at the nominal
+2a level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results. -
The adjoint analyses were also carried out using the P3 cr ss section , approximation from the SAILOR library. Adjcint source locations were' chosen at the center of the surveillance capsules, at dosimetry positions in the reactor cavity, and at locations on the pressure vessel inner diameter. Again, these calculations were run in R,0 geometry to provide power distribution importance functions for the exposure parameters of interest. Having the adjoint importance functions and appropriate core power distributions, the *esponse of interest is calculated as l Rg,g = /R#8 I (R,0) F (R,0) R dR de where: RR,0 = Responso of interest e.g., # (E > 1.0 MeV) at radius R and azimuthal angle 8. i I (R,0) = Adjoint importance function at radius R and azimuthal angle 0 , I F (R,0) = Full power fission density at radius R and azimutnal angle 0 ' It should be noted that as written in the above equation, the importance function I (R,0) represents an integral over the fission distribution so ' i that the response of interest can be related directly to the spatial ! distribution of fission density within the reactor core, i ! w.wme 6-3 l 1
)
The core power distributions used in the plant specific fast neutron exposure eva', ation for Diablo Canyon Unit 2 were taken from the cycle 1 and cycle 2 fuel cycle design reports (8,9). The data extracted from these fuel cycle , detign reports represented cycle averaged relative assembly powers. Therafore, the adjoint results provided data in terms of fuel cycle averaged , neutron flux which, when multiplied by the fuel cycle length yielded the incremental fast neutron fluence. The exposure calculations for cycle 1 established a direct comparison with dosimetry data from capsule U; whereas the cycle 2 calculation provided the means for exposure projection and capsule lead Sctor evaluation based on low leakage fuel management. The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against neutron dosimetry data obtained at the ORNL PCA facility (10). Extensive comparisons of analytical predictions with measurements from power reactor surveillance capsules and reactor cavity dosimetry programs have also been made. The benchmarking studies indicate that the use of SAILOR cross-secticns and generic design basis power distributions produces flux levels that tend to be conservative by from 7-22%. When plant specific power distributions are used with the adjoint - importance functions, the benchmarking studies show that fluence predictions
~
are within + 15% of measured values at surveillance capsule locations. ' Calculations applicable to reactor cavity locations tend to be biased low by l approximately 20% depending on the thickness of the pressure vessel wall. 6.3 Neutron Dosimetry Tha passive neutron flux monitors included in the Diablo Canyon Unit 2 surveillance program are listed in table 6-2. The first five reactions in table 6-2 are used as fast neutron monitors to relate neutron fluence (E > 1.0 MeV) to measured materials properties changes. To properly account for burnout cf the product isotope generated by fast neutron reactions, it is necessary to also determine the magnitude of the thermal neutron flux at the monitor location. Therefore, bare and cadmium-covered cobalt-aluminum , monitors were also included. Mlls/M144410 g.4
d l The relative locations of the various monitors within tne surveillance capsules are shown in figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within the dosimeter block located near the center of the capsule. l The use of passive monitors such as those listed in table 6-2 does not yield a direct measure of the energy deperdent 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 i from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: i
- o The operating histcry 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 subsequent derivation of the average neutron flux requires completion of two procedures. First, the disintegration rate of product isotope per unit mass of monitor must be determined. Second, 2 in crder 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 l ASTM procedures (11 thru 24). Following sample preparation, the activity of ( each monitor is determined by means of a lithium-drifted germanium, Go(Li), j
- l. gamma spectrometer. The overall standard deviation of the measured data is a j function of the precision of sample weighing, the uncertainty in counting, and :
. the acceptable error in detector calibration. For the samples removed from i Diablo Canyon Unit 2, the overall 2e deviation in the measured data is ! . determined to be + 10 percent. The neutron energy spectra are determined analytically using the method described in section 6.2.
xmww .e 6-5 L i i
- - - - - - - - - - - - - , - - ~ - -. - . - . .
y - - - o Having the measured activity of the monitors and the neutron energy spectra at the locations of interest, the calculation of the neutron flux proceeds as follows. . The reaction product activity in the monitor is expressed as: . N N P At -At d R=ifg Y a(E)#(E) [ h-C3x(1 e d) e (6-1) , E 31 1 where: l R = induced product activity N,
= Avagadro's number l l A = atomic weight of the target isotope fg = weight fraction of the target isotope in the target ,
material 4 Y = number of product atoms produced per reaction o(E) = energy-dependent reaction cross section , ) ((E) = time-averaged energy-dependent neutron flux at the monitor ; location with the reactor at full power , l j P = average core power level during irradiation perio.1 j 3 ! P,,, = maximum or reference core power level i A = decay constant of the produci isotope l 4
= length of irradiation period j i l t) i td = decay time following irradiation period j ;
j = #(E>1.0 MeV) during irradiation period j divided by l C) i the average e(E>1.0 MeV) over the total irradiation l , period. C) is calculated with the adjoint neutron ( ) transport rethod and accounts for the change in neutron j j' monitor response caused by core power distribution l variations from cycle to cycle. Pj /P,,,, which accounts ] for the month-by-month variation of power level within a l cycle, is applied to the full power-based flux ratio, C3 . . l i i~ f i i I
\
i
- w.wme 6-6 i L
1 v------ -e r .- + - "--++-
Since neutron flux distributions are calcul'ated using multigroup transport methods and, further, since the prime interest is in the ' fast neutron flux above 1.0 MeV, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-1) is replaced by the following relation. I _
..)Eu(E) *(E)dE = c *(E > 1.0 MeV) where N
p. o(E) e (E)dE oe gg
, g, N
I. e (E)dE 1 'g J 1.0 MeV G=G 1.0 MeV Thus, equation (6-1) is rewritten N N -it -it d R=jfg Y3e(E>1.0MeV)[ [P max C)(1-e d) e j=1 or, solving for the neutron flux, R N e (E > 1.0 MeV) = N -it -At d [fg Y3 [Pf jul max C) (1-e d) e (6-2) The total fluence above 1.0 MeV is then given by N 6 (E > 1.0 MeV) = # (E > 1.0 MeV) [g 3 max tj (6-3) jal l l where: N P J , total effective full power seconds of reactor P,,, j operation up to the time of capsule removal An assessment of the thermal neutron flux levels within the surveillance
. capsules is obtained from the bare and cadmium covered Co59(n,r)Co60 data by means of cadmium ratios and the use of a 37-barn 2200 m/see cross section. Thus, w ,* * '
6-7
( R D-1 bare D N (0~4) * '
'Th
- N P 3
-it -it d
[ f Yo j [ p (1-e d)e j=1 mar. , where: R bare D is defined as 3 Cd covered 0.4 Neutron Transport Analysis Results Results of the discrete ordinates transport calculations for the Diablo Canyon i Unit 2 reactor are summarized in this section. In figure 6-3, the calculated ; design basis 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 presented as a function of azimuthal angle. In figure 6-4, the radial distribution of design basis maximum fast (E > 1.0 MeV) neutron flux and , ~ displacements per atom (dpa) through the thickness of the reactor vessel is
, shown. The relative axial variation of fast neutron flux and dpa within the .
reactor vessel is given in figure 6-5. Absolute axial variations of fast
> neutron flux may be obtainor' by multiplying the levels given in figure 6-3 or 6-4 by the appropriate values from figure 6-5.
In table 6-3, the calculated fast neutron exoosure parameters for non-low i
- leakage fuel management (design basis) and low leakage fuel management (cycle 2 specific) are provided. The design basis data listed in table 6-3 represent the maximum exposure conditions for the vessel wall. Table 6-4 provides
- '
updated lead factors for all of the Diablo Canyon Unit 2 surveillance capsules tased on both the design basis and cycle 2 specific exposure calculations.
- The lead factor is defined as the ratio of the fast (E > 1.0 MeV) neutron
- flux at the geometric center of the surveillance capsule to the maximum fast i neutron flux at the reactor vessel inner radius.
I w wswo 6-8 F b __._--3..,-.,-__....---r - r--- --------+v--w+,---++--w-----nw w---r7-m--
l I 6.5 Dosimetry Results, l l - The irradiation history of the Diablo Canyon Unit 2 reactor is given in table 6-5. The data were obtained from NUREG-0020, "Licens3d Operating Reactors Status Summary Report," December 1984 through September 1986. Measured and l l saturated rea. tion product specific activities as well as measured full power j reaction rates are listed in table 6-6. These valcas were derived using the < ! irradiation history provided in table 6-5 along with the physical I characteristics of each sensor given in table 6-2. J In order to derive neutron flux and fluence levels from the measured reaction rates, suitable spectrum averaged reaction cross ,ections are required. The neutron energy spectrum calculated at the radial and azimuthal center of Capsule U is tabulated in table 6-7, whereas the resultant spectrum averaged cross-sections for each reaction of interest are listed in table 6-8. The fast neutron (E > 1.0 MeV) flux and fluence levels derived for Capsule V are prea nted in tab 1w 6-9. The thermal neutron flux obtained from the cobalt-aluminum monitors is summarized in table 6-10. Due to the relatively l low thermal neutron flux at the capsule location, no burnup correction was made to any of the measured activities. The maximum error introduced by the assumption is estimated to be less than 1 percent for the Ni58 (n.p) Co58 reaction and evea less significant for all of the other fast neutron reactions. l An exadnation of table 6-9 indicates that the measured fast neutron exposure of Capsule U was 3.51 A 10 18 n/cm 2
. This fluence level is within 15% of the design basis prediction of 4.04 x 10 18 n/cm2 and exceeds the cycle 1 plant specific value of 3.08 x 10 18 n/cm2 by 14%. In either case agreement between measurement and calculation is considered to be excellent.
A summary of measured and calculated fast neutron (E > 1.0 MeV) exposure of the Diablo Canyon Unit 2 pressure vossel is presented in table 6-11. Also
. provided are exposure projections to 8 EFPY as well as to end of vessel design life (40 EFPY). The calculated flux levels based on the cycle 2 plant speejfic analyses (low leakage fuel management) were used to ptrform these projections.
w mien ie 6-9
In regard to the data provided in tablo 6-11, it should be noted that, in addition to the neutron dosimetry contained ia Capsule U, extensive dosimetry was al.o deployed in the Diablo Canyon Unit 2 reactor cavity during the cycle - 1 irradiation. A detailed description of the cavity dosimetry program and of the evaluation of sensors irradiated in cycle 1 is provided in re'erence (5). - In developing the currant "measured" fast neutron exposure of the pressure vessel wall, projections from both the surveillance dosimetry and the cavity dosimetry were employed. The quoted current exposure was taken to be the average of the two sets of projected data. This projection and averaging exercise may be summarized as follows: 2 CYCLE 1 FLUENCE (n/cm ) AVERAGE (b) VESSEL CAPSULE CAVITY VESSEL : LOCATION PROJECTION PROJECTICN EXPOSURE, Capsule V - 3.84 x 10 18 3.51 x 1018 (actual)(a) 6.64 x 10 17 17 6.95 x 10 17 Vessel IR 7.25 x 10 ' Vessel 1/4T 3.54 x 10 17 3.87 x 10 17 3,71 x 10 17 Vessel 3/4T 7.18 x 10 16 7.84 x 10 16 7.51 x 10 16 45' Cavity 2.35 x 10 16 - 2.57 x 10 16 a) Measured capsule exposure b) Based on measureme..t results An examination of this data at indicates excellent agreement between capsule and cavity projections for the vessel wall and supports the methodology used for assessment of exposure gradients through the pressure vessel wall. xv. min i. 6-10 P c - - - - - , , , - - . , . - - - - - , - - - - , -
TABLE 6-1 SAILOR 47 NEUTRON ENERGY GROUP STRUCTURES Lower Energy Lower Energy Genup (MeV) Group (MeV) 1 14.19* 25 0.183 2 12.21 26 0.111 3 10.00 27 0.0674 4 8.61 28 0.0409 l 5 7.41 29 0.0318 l 6 6.07 30 0.0261 7 4.97 31 0.0242 l 1 8 3.68 32 0.0219 9 3.01 33 0.0150 10 2.73 34 7.10 x 10'3 11 2.47 35 3.36 x 10 -3 l' 12 13 2.37 2.35 36 37 1.59 x 10'3 4.54 x 10 ~4 i 14 2.23 38 2.14 x 10 ~4 15 1.92 39 1.01 x 10 ~4 i l 16 1.65 40 3.73 x 10 -5 l 17 1.35 41 1.07 x 10 -5 l 18 1.00 42 5.04 x 10 -6 19 0.821 43 1.86 x 10 -6 20 0.743 44 8.76 x 10 ~I 21 0.608 45 4.14 x 10'I 22 0.498 46 1.00 x 10 ~I 23 0.369 47 0.00 24 0.298
-
- The upper energy of group 1 is 17.33 HeV.
w.eu i. 6-11
TABLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Target Fission Monitor Reaction Weight Response Product Yield
- Material of Interest Fraction Range Half-Life (%)
Copper Cu03(n )Co60 0.6917 E>4.7 MwV 5.27? years Iron Fe54(n.p)Mn54 0.058 E>1.0 MeV 312.2 days ! Nichel NiS8(n.p)CoS8 0.6827 E>1.0 MeV 70.91 days Uranium-238* U238(nf)Cs137 1.0 E>0.4 MeV 30.17 years 6.0 ; Neptunium-237* Np237(n,f)Cs137 1.0 E>0.08 MeV 30.17 years 6.5 ! Cobalt-Aluminum
- CoS9(n,r)Co60 0.0015 0.4eV<E<0.015 MeV 5.272 years Cebalt-Aluminum Co59(n,r)Co60 0.0015 E<0.015 MeV 5.272 years
- Denotes that monitor is cadmium shielded. !
t i i I l
. l
\
t xv.uum ie . 6-12 l
TABLE 6-3 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS FOR THE PEAK LOCATION OF THE REACTOR VESSEL DESIGN BASIS LOW-LEAKAGE RADIAL LOCATION FASTNEgTRONFLUX DISPLACEMENT DISPLACEMENT FASTNE(TRONFLUX WITH N THE (n/cm -sec) RATE (n/cm sec) RATE REACTOR VESSEL (E > 1.0 MeV) (E > 0.1 HeV) (dpa/sec) (E > 1.0 MeV) (! > 0.1 MeV) (dpa/sec) Inner Surface 2.44 x 10 10 6.11 x 10 10 3.88 x 10 -11 1.72 x 10 10 4.31 x 10 10 2.74 x 10-I1 R = 86.72 in. 1/4 Thickness 1.30 x 10 10 5.11 x 10 10 2.48 x 10 -11 9.16 x 109 3.60 x 10 10 1.75 x 10 -11 cn f- R = 88.88 in.
-11 -11 1/2 Thickness 6.02 x 109 3.62 x 10 10 1.49 x 10 4.24 x 109 2.55 x 10 10 1.05 x 10 R = 91.03 in.
3/4 Thickness 2.64 x 109 2.29 x 10 10 8.51 x 10 -12 1.86 x 109 1.62 x 10 10 6.01 x 10-12 R = 93.19 in. Outer Surface 1.07 x 10 9 1.15 x 10 10 4.11 x 10 -12 7.54 x 10 8 8.11 x 109 2.90 x 10-12 R = 95.34 in. NOTE: Reactor vessel inner surface is taken to be at the clad / base metal interface. CLAD IR = 86.50 inches msn.osime os
1ABLE 6-4 CALCULATED FAST NEUTRON EXPOSURE PARANETERS AND LEAD FACTORS FOR THE SURVEILLANCL CAPSULES DESIGN BASIS LOW-LEAKAGE F4ST NEUTRON FLUX DISPLA'.EMENT LEAD FAST NEUTRON FLUX DISPLACEMENT LEAD CAPSULE AZIMUTHAL 2 P'.TE FACTOR ID LOCATION (n/cm -sec) RATE FACTOR (n/cm -sec) (DEG.) (E > 2.0 MeV) (E > 0.1 MeV) (dpa/sec) (E > 1.0 MeV) (E > 0.1 MeV) (dpa/sec)
-10 1I 2.62 x 10 -10 11 U 56 1.29 x 10 6.00 x 10 11 5.29 8.68 x 10 10 4.03 x 10 1.76 x 10 5.05 W 124 1.29 x 10 II 6.00 x 10 11 2.62 x 10-10 5.29 8.68 x 10 10 4.03 x 10 II 1.76 x 10-10 5.05 X 236 1.29 x 10 11 6.00 x 10 II 2.62 x 10 -10 5.29 8.68 x 10 10 4.03 x 10 11 1.76 x 10 -10 5.05 Z 304 1.29 x 10 Il 6.00 x 10 11 2.62 x 10-10 5.29 8.68 x 10 10 4.03 x 10 11 1.76 x 10 -10 5.05 V 58.5 1.08 x 10 ll 5.01 x 10 10 2.19 x 10
-11 4.41 7.23 x 10 10 3.36 x 10 10 1.47 x 10-11 4.21 Y 238.5 1.08 x 10 II 5.01 x 10 10 2.19 x 10
-11 4.41 7.23 x 10 10 3.36 x 10 10 1.47 x 10-11 4.21 a) The radius of the surveillance capsule center is 81.62 in.
b) The lead factor is the ratio of the fast (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. 3051s/SS1888 le 9 . O 9 O .
,n -- - - - - - - - . .------.-,v ---- -- , , , - - . , - - , - , , , , - - , , - , ,-- - , . , .,-
TA8LE 6-5 l IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U IRRADIATION P J b IRRADIATION DECAY ; PERIOD (MWt) Pg ,f TIME (days) TIME (days) l l (
. 10/85 734 0.215 13 731 ,
11/85 1272 0.373 30 701 l 12/85 1305 0.383 31 670 t 1/86 750 0.220 31 639 ! 2/86 341 0.100 28 611 3/86 1845 0.541 31 580 4/86 2859 0.838 30 550 ! 5/86 3088 0.905 31 519 i 6/86 2789 0.818 30 489 7/86 2396 0.702 31 458 8/86 3245 0.951 31 427 l . 9/86 2779 0.815 30 397 l 10/86 3206 0.940 31 366 c
. 11/86 3279 0.961 30 336
< 12/86 3128 0.917 31 305 ! 1/87 3103 0.910 31 274 f i 2/87 2647 0.776 28 246 i 3/87 2029 0.595 31 215 t 4/87 2540 0.745 3 212 ; NOTE: Pg ,f = 3411 MWt [ , Irradiation Time = 0.99 EFPY : I f i { 1 -
! l L ,
r ie
; m i. m i 6 15 i
- _ _ - _ _ _ _ _ _ _ - _ _ _ . . _ . . _ _ _ . - _ _ _ _ . _ .__ _ _ _ _ a
TABLE 6-6 - MEASURED MONITOR ACTIVITIES AND REACTION RATES FOR CAPSULE U Measured Saturated Reaction Monitor and Activity Activity Rate . Axial L :ation (dis /see-am) (dis /see am) (reactions /see-atom) Cu-63Q.2)Co-60 Top 3.50x10f Middle 3.91x10l 4.13 x 10 3.70 x 10 Botton 4.04 x 10 4 3.62 x 10 55 Average 3.61 x 10 5.50 x l'fl7 Fe-54(n,p)Mn-54 Top 3.55x10l Middle 1.13x10f 1.21 x 10 3.80 x 10 0 6 Bottom 1.18 x 10 3.70 x 10 6 5.87 x 10 -15 u Average 3.68 x 10 Ni-58(n,p)Co-58 Top Middle 5.43x10f 5.62 x 10 5.52x10f 5.72 x 10 1 7 6 - a Bottom 5.79 x 10 5.89 x 10 7 -15 l Average, 5.71 x 10 8.15 x 10 U-238/n.f)Cs-137 Midtile 1.37 x 10 55 Corrected
- 1.16 x 10 5.23 x 10 6 3.48 x 10'I4 1
l Np-237(n,t)Cs-137 Middle 1.37 x 106 6.17 x 10 7 3.74 x 10 -13 )
- Corrected by 0.85 to account for U-235 fissions and Pu build-in.
e 5 mi.w= i. 6-16
. 1.,
TABLE 6-7 CALCULATED (8) NEUTRON ENERGY SPECTRUM AT THE CENTER OF DIABLO CANYON UNIT 2 SURVEILLANCE CAPSULE U Energy Neutron Flux Energy NeutrgnFlux Group (n/cm -sec) Group (n/cm*-sec) 1 1.05 x 10 7 25 7.51 x 10 10 2 4.70 x 10 7 26 7.87 x 10 10 3 1.96 x 10 8 27 6.25 x 10 10 4 4.00 x 10 8 28 4.32 x 10 10 5 7.19 x 10 8 29 1.26 x 10 10 6 1.70 x 109 30 6.82 x 10 9 I 7 2.44 x 109 31 1.97 x 10 10 1 8 5.16 x 10 9 32 1.31 x 10 10 9 4.84 x 109 33 2.19 x 10 10 10 4.08 x 10 9 34 3.08 x 10 10 11 , 4.91 x 10 9 35 5.33 x 10 10 I
. 12 2.46 x 10 9 36 4.78 x 10 10
, 13 7.64 x 10 8 37 6.56 x 10 10 1 - 14 3.84 x 10 9 38 3.56 x 10 10 i 15 1.03 x 10 10 39 3.93 x 10 10 16 1.42 x 10 10 40 5.32 x 10 10 l 17
- 21 x 10 10 41 6.25 x 10 10 10 18 5.10 x 10 10 42 3.45 x 10 19 3.98 x 10 10 43 3.89 x 10 10 1 20 1.85 x 10 10 44 2.39 x 10 10 1 21 6.95 x 10 10 45 1.83 x 10 10 22 5.44 x 10 10 46 2.56 x 10 10 10 I 23 6.75 x 10 10 47 3.12 x 10 I h 6.70 x 10 10 i
I i ! f a) Design basis power distribution . l ! { ! - - ., .. 3 17 l
c1
. o TABLE 6-8 SPECTRUN AVERAGED REACTION CROSS-SECTIONS FOR USE IN ,
FAST NEUTRON DOSINETRY EVALUATIONS . ! i Reaction 5(barns) . i Cu63(n )Co60 0.000470 Fe54(n.p)Mn54 0.0529 l NibO(n.p)Co58 0.0746 f U238(n.f)CsIII 0.307 , f Np237(n,f)Cs137 3.36 4 0 4 l , i 1 1 9 e mi.mim i. 6-18
._ _ . _ .m - -
c ._
. . . . . e. 1; f%
. .j TAaLE 61 RESULTS OF FAST NEUTRON 0051esETRY FOR CAPSULE U -
2 2 ( (E > 1.0 mov) (n/cm -sec) 4 (E > 1.0 now) (n/cm ) measureo neaction aate calculatso calculatao Calculates neaction (reacttons/ measured oesign cycle 1 m-ed - Cycle i sec-atom) Sasis Spa ific Scocific 9 40 -17 11 18 CuN(n.e)Co 5.50 m 10 1.17 x 10 3.86 m 10 Fe (n.p)emS4 5.87 a 10
-15 1.10 a 10' 3.48 x 10'8 mi"(n.p)Co" 8.15 m 10-15 ,,og , ,o M 3.42 m 10 38
- 38 U (n.f)Cs* I 3.48 a 10 " 1.13 x 10" 3.55 x 10 .
np237(n. f )Cs'3# ~I3 1.11 m 10 ' 3.40 m 10 3.74 m 10 I e 10 3.51 a 10'8 3.08 x 10'8 5 Average - 1.12 a 10" 1.29 a 10 " 9.34 x 10 l
.i 6
it
- i j
j 1 1 l } 5 ] i k 4 1 3051s/0518SS:10 i 4 ', - - . . . _ - - - - - - - - , - - - - , - - - - - . - - - - - , - - - ,,, - ,- . - - , - - - - - - - - - - - - - - - - - - - - - ..-,--.-..n - - - - - ,v.,--- -- - , -
g .r,,,7 --
., :o ,j , - TABLE 6-10 RESULTS OF THERNAL NEUTRON DOSINETRY FOR CAPSULE U ' (.;
Saturated Activity. (dos /am) 'Th - 2 Location Bare cd-Covered (n/cm -rsc) Top 9.31 x 10 7 5.11 x 10 7 7.41 x 10 10 ; 7 10 Middle 8.59 x 10 7 4.82 x 10 6.65 x 10 7 7 10 ; Botton 8.89 x 10 5.12 x 10 6.65 x 10 Average - - 6.90 x 10 10 f 1 b i B t i i j t F I I i i 4 1 1 4-l wimm ie - 6 20 t
. TABLE 6-11 SUW4ARY OF DIABLO CANYON UNIT 2 FAST NEUTRON EXPOSURE EVALUATION BASED ON CAPSULE U RESULTS
' ' 2 {
FAST NEUTRON FLUENCE (E > 1.0 MeV) (n/en ) LOCATION CURRENT 8 EFPY 40 EFPY ,
' MEASURED MEASURED CALCULATED CALCULATED CALCULATED CAPSULE U 3.51 x 1018(a) 3,08 x 1018(e) . .
VESSEL IR 6.95 x 1017(b) 6.10 x 1017 (c) 4.41 x 1018(d) 2.18 x 1019(d) . VESSEL 1/.T 3.71 x 1017(b) 3.26 x 1017(c) 2.35 x 1018(d) 1.16 x 1019(d) i i VESSEL-3/4T 7.51 x 1016(b) 6.59 x 1016(c) 4.77 x 1017(d) 2.36 x 1010(d) [ l, a) Based on dosimetry from Capsule U [ b) Based on surveillance capsule / cavity dosimetry correlations described in l
. reference 5. '
c) Based on cycle 1 plant specific analysis d) Projected using cycle 2 low ledags core power distr ibution ; i I P [ h
'i
[ l I
- I 4
l i v . m inu ie 6-21
. . _ _ _ ___. __ _ _ _ _ _ _ _ - . _ . _ _ _ _ _ __ _ _ _ .. . . _ _ ~
I eso* , 1 g CORE BARREL
; #m25'
,,34 PEUTRON >
PAD
#345' I /
/i '
//
//
/
I j/ 7
/ /
- ll i // '
ll L
// >
1 h I l l
./
1 i i / i ,,! l Figure 6-1. Diablo Canyon Unit 2 Reactor Geometry . i i mienn ie 6-22
4 i i
?
CHARPY SPECIMEN
-,1 ,1 , -
i s ; 's 1, 1.1.. 1. ,- 1.
///// //////
sssssssssssss\ssssss PEUTRON PAD \
\ :
ks\\\\\\\\\\\\\\\\\\ i I l , I
, [
Figure 6-2. Internal Surveillance Capsule Geometry i I mi.wm io 6-23 l , l i f
\ ,
l
n Il.. . . _, .
. . .mm j-9.... _ .a e... _ f- .+: - :. :
._ ,. --..; - u, t g :. - :- -
7; '
=+-q==di=-i~ ::=i=i+d ==-i-in = :*=i++=i=? =i-i --t I H := ~
~
," 3 =;i
. . . . =_ _ a -. ._ : ;__ -
t -___u =_ + =_=__= :_
- -= = = .
__...__ _- =n _a _L. _'_. .- d- - MiiH++ TEM M -[h.hWt=.=-i!=W Ti i-#9k YINI m=Iu= m:1 .= . .1 2 + ~= M =_2a .. :f+ + = 5 ++m 3~~ d - i-li'? ;i;;lI f 5: bri--I = + ~ --i : I 3- : 22 '2+^ 1_ ic;- P E i ;-~ i-=9 '" Ellili ' ~'
- 2
- 2 2 ~-
Tr__;33;;;i_. . ._r. :;-- -
]c-izi1=:*:
a tx'::iii';ELIMI - ri'_ii-jif ~- Q . , t==._-_=:==______-.-.-a=i=--._--- ._ . _ _ . - . . _
~i t -
.I 1 K-i-
[_ ' A- SU,RVEILLANCE CAPSUL (S
' ~
- .7 I
mai:
. _ - i : 2 i. -
n ,
. r.y. - ,=;== -
g . _ - . =- r :i: --T- - ;__p =2;;===,_.= :4:
- = := --- -
.? ,'Au==
, == ti iz!mru=miiw=;=-tan ==e- e-mi E
o a.
~=wuH%= _w+- =d l i d =--4 - '
p M . ra. . ' h--- _---Y -i E:-
- .2'- - = _ -- ---- - : l !r-- ' ---3 w . _ _ _ : =_ -; 4 _ :. =: .1 .-- - - -
-.. =y: === --
)( 8 . . _ _ - . . . . - . . . . _- -._._..:-- . J.-
-g a : _ _. ._1 --
n - _ :-- =r-- g: = PRES 3URE VESSEL lR ;=.:
.J :- -
N .. _i_ _.-- ,.-.~.2:5._----. +
~+~_d___-*~~; ~ - - - ' ---~-+ - -k g -
e l ,
=
,g! ,r y . _ . e...
a f.
,1 ;n= =. g p_+ .--.4.- m_w =_ ;+ --!_ :- H
, h i:+M==i s iM=4-EMMi=Q-jQi4Ei-- a=j d E+= += E : - =q i
_ , _ _ . _2p ;;aw ;a _e. 4_1 . . _w -'t- t __ _, ,i 4. f . '*R l}; ~'?- . -.__1_ _'! l
=6! H - x Fj _4 --
=- H ~ $ s ----l . ~ ~ 4++:-duL~: + :-d
- i- 2:i::: 5"---- - -
- n 3.
ititi:.li- izi:'
= =; 421h:=nE
=== : ===t =-+--?~~+~-
1."i+:. a ==r = iMinidinsil 3/4 T LOCADON N E5h ==== 2.
~ - ~ . . --! -- = :T!= =_=_ __ =_. . = .It _.-.===4: :
_ . + .. w . C1
--.n_.,
-_ _s.-
i
.....,--_mt.--.----t--- .
_ _- n. _:--- 4 z . _ ;. . _ p __. _ .+; -p.--j 1- -. . . 1_ ;1 I_ 1-1
- i .-1 4. -
g ! I- i ! 0 to 20 30 40 60 AZlMUTHAL ANULE (DEG.) Figure 6-3. Calculated Azimuthal Distribution of Design Basis Maximum Fast (E a 1.0 kev) Neutron Flux Within the Reactor Vessel Surveilhnce Capsule Geometry saw sii. 6-24
1 curve 7554S8. A Base Metal IR l 1/4T
- 1. 0 -
l 1/27 -
~
3/4T
~
~
Base Metal OR E E l . g - l a dpa
. E W
z 0,1 - a: _ l _
~
0 (E d l.0 MeV) 1
- n C % R R -
N N IN N N O.01 220 230 24 0
. Radius ( cm)
Figure 6-4 Relative Radial Variation of Fast Neutron Exposure
- within the Reactor Vessel w.wm i. 6-25
I 10 0 8 - 6 - 4 - t 2 - M 3 10*1 -
,8 -
6 - b g 4 - W I
~
2 -
\
5 1 W E ' 10 2 8 - 6 - CORE M10 PLANE l a l
- TO VESSEL
~ CLOSURE HEAD
- .r ! ! ! !
10 3 - 300 200 100 0 100 200 300 DISTANCE FROM CORE MIDPLANE (om) , l Figure 6-5, Relative Axial Variation of Fast (E > 1.0 MeV) l l Neutron Flux and dpa within the Reactor Vessel , { l mswmo 6-26 i 9 l
SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule based on methods of ASTM E185-82 is recommended for future capsules to be removed from the Diablo Canyon Unit 2 reactor vessel: Estimated Capsule Fluence Removal Location Lead 2 l . Capsula (deg) Factor Time (a) (n/ca )
^
5.05 18(d)
- U 56 0.99 (Removed) 3.51 x 10 X 236 5.05 4 1.10 x 10 19(b)
V 58.5 4.21 10 2.29 x 10 19(c) Y 238.5 4.21 12 2.75 x 10 19 W 124 5.05 Standby - Z 304 5.05 Standby -
- a. Ef fective full power years from plant startup
- b. Aporeximate fluence at 1/4 thickness reactor vessel wall at end of vessel design life
- c. Approximato fluence at reactor vessel inner wall at end of. vessel design life
- d. Actual neutron fluence i
)
l 1 w s w ess ,e l y.$
i SECTION 8 i i~ REFERENCES !
- 1. Davidson, J. A. at al., "Pacific Gas and Electric Co. Diablo Canyon Unit l
- No. 2 Reactor Vessel Radiation Surveillance Program,' WCAP-8783 Dec.1976. !
i i
- 2. Code of Federal Regulations, 10CFR50, Appendix G, "Fracture Toughness !
Requirements: and Appendix H. "Reactor Vessel Material Surveillance ! P.ogram Requi: ements," U.S. Nuclear Regulatory Commission Wathington, D.C. ; i 3
- 3. Regulatory Guide 1.99, Proposed Revision 2. "Radiation Damage to Reactor Yessel Materials," U.S. Nuclear Regulatory Connission, February, 1986. i t
- 4. Nuclear Plant Irradiated Steel Handbook, EPRI NP-4797, Topical Report, I September 1986. ,
t
- 5. Anderson, S. L., "Reactor Cavity Neutron Dosimetry Program for Diablo
- Canyon Unit 2 - Cycle 1 Evaluations. WCAP-11690, December 1987.
. f
] 6. Seitesz, R. G. Di:ney, R. K., Jedruch, J., and Ziegler, S. L., "Nuclear l i Rocket Shielding Methods, Modification, Upciating and Input Data ( i Preparation, Vol. 5 - Two-dimensional Discrete Ordinates Transport l { Techniqua," WANL-PR(LL)-034. Vol. 5, August 1970. l 1 i a i j 7. 'ORNL RSIC Data Library Collection DLC-76. SAILOR Coupled Self-Shleided, j 47 Neutron, 20 Gama-Ray, P3, Cross Section Library for Light Water { ] ] Reactor", i ;
- 8. Pritchett, J. E. et, al., "Nuclear Design and Core Physics [
' Characteristics of the Diablo Canyon Unit 1 Nuclear Power Plant Cycle 1," , WCAP-8408 Rev. 1. January 1983 (Westinghcuse Proprietary). i
- i. !
)
- i mawm ie 31 j f
- 9. Facteau, M. W., Piplics, A. N., Radcliff, R. E. and Zimmermann, M. W. , ;
"Nuclear Nsign and Core Physics Characteristics of the Diablo Canyon :
Power Plant Unit 2 Cycle 2' WCAP-11450, Maf 1987. (Westinghouse , Proprietary).
- 10. Anderson S. L. and Tran, K. C., "8enchmark Testing of Westinghouse l
Neutron Transport Analysis Methodology - PCA Evaluations " (to be ; published). j r
- 11. ASTM Designation E48b 82, "Standard Guide for Application of Neutron !
Transport Methods for Reactor Vessel Surveillance," in ASTM Standards, f Section 12 American Society for Testing and Materials, Philadelphia, Pa., 1984.
- 12. ASTM Designation E560-77, 'Stan:'ard Recommended Practice for
- Extrapolating Reactor Vessel Surveillance Dosimetry Results," in ASTM j 1 Standards Saction 12, American Society for Testing and Waterials, ;
j Philadelphia, Pa., 1984. ' i j 13. ASTM Designation E693-79, ' Standard Practice for Characterizing Neutrcn j Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa) " in ASTM Standards, Section 12, American Society for Testing and f f Materials, Philadelphia, PA., 1984. L i I i : i 14. ASTM Designation E706-81a, "Standard Ht. ster Matrix for Light-Water i, Reactor Pressure Yessel Surveillance Standards ' in ASTM Standards, : j Section 12, American Society for Testing and Materials, Philadelphia, [ j Pa., 1984. i 1 : i 1 15. ASTM Designation E853-84, "Standard Practice for Anelysis and l Interpretation of Light-Water deactor Surveillance Results " in ASTM [ Standards, Section 12, American Society for Testing and Materials, [
. Philadelphia, Pa., 1984. -
l 4 i
" ' " * ' ' 8-2 f I !
- 16. 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.
- 17. 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,
- 18. 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.
- 19. 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, i
- 20. 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.
! 21. ASTM Designation E523-82, "Standard Method for Determining Fast-Neutron Flux Density for Radioactivation of Copper," in ASTM Standards, Section
- 12. American Society for Testing and katerials, Philadelphia, Pa.,1984.
t
- 22. ASTM Designation E704-84, "Otandard Method for Measuring Reaction Rates by Radioactivation of Uranium-238," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1984.
- 23. ASTM Designation E705-79, "Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237," in ASTM Standards, 4 . Section 12, American Society for Testing ana Materials, Philadelphia, Pa., 1984.
f mmme 8-3
e <- .
- 24. ASTM Designation E1005-84, "Standard Method for Application and Analysis I
of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM , 4 Standards, Section 12, American Society for Testing and Materials, , j Philadelphia, Pa., 1984. ( i P h I l i l i i
)
t t h i h I l l i i i y i f l ! L i I
- m ,. . . 84 ;
i (
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