ML20206D142

Analysis of Capsule U from Northeast Utils Svc Co Millstone Unit 3 Reactor Vessel Radiation Surveillance Program

ML20206D142
Person / Time
Site:	Millstone
Issue date:	06/30/1988
From:	Albertin L, Shaun Anderson, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
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ML20206D145	List: ML20206D142
References
WCAP-11878, NUDOCS 8811160562
Download: ML20206D142 (97)

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{{#Wiki_filter:- ______ WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRituTION \\ ) WCAP-11878 s t l ANALYSIS OF CAPSULE U FROM THE NORTHEAST UTILITIES SERVICE COMPANY MILLSTONE UNIT 3 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM S. E. Yanichko S. L. Anderson L. Albertin N. K. Ray June 1988 Work performed under Shop Order No. NXXJ-106 APPROVED:

7. [A h * / h T. A. Meyer, hanager Structural Materials and Reliability Technology Prepared by Westingheuse for the Northeast Utilities Service Company Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse orjits licensees withput the customer's approva1. OK To fTHt457 ELE.4SE PEA'

//floff8 TEwou wffs;gt gy; of NO VnLt7/E3 WESTINGHOUSE ELECTRIC CORPORATION Power Systems Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230 o n.,e w s es1115036k 081107 gos r.ttock0500g3

rf f l f PREFACE i e This report has been technically reviewed and verified. l I Reviewer l [d.[ Sections 1 through 5, 7 and 8 D. J. Colburn Section 6 E. P. LippincottY Appendix A J. C Schmertz 8dM k / / I I { 1 i e b 1 8 [ t l I e 4 I e 1 F D i m ai m.= ggg 4

TABLE OF CONTENTS t Section Title Page 1

SUMMARY

OF RESULTS 1-1 2 INTRODUCTION 2-1 3 BACKGROUND 3-1 2 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE U 5-1 5-1. Overview 5-1 i 5-2. Charpy V-Notch Impact Test Results 5-3 [ 5-3. Tension Test Results 5-4 t 5-4. Compact Tension Tests 5-4 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY 6-1 6-1. Introduction 6-1 6-2. Discrete Ordinates Analysis 6-2 6-3. Neutron Dosimetry 6-8 i 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8 REFERENLES 8-1 r Appendix A Heatup and Cooldown Limit Curves for Normal Operation j i i l

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LIST OF ILLUSTRATIONS Figure Title Page 4-1 Arrangement of Surveillance Capsules in the 4-4 Reactor Vessel l t 4-2 Capsule U Diagram Showing Location of Specimens, 4-5 Thermal Monitors, and Dosimeters [ 5-1 Charpy V-Notch Impact Properties for Millstone w.nt 3 5 12 Reactor Vessel Shell Plate B9805-1 (Longitudinal Orientation) 5-2 Charpy V-Notch Impact Properties for Millstone Unit 3 5-13 Reactor Vessel Shell Plate B9805-1 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for Millstone Unit 3 5-14 Reactor Vessc1 Weld Metal 5-4 Charpy V-Notch Impact Properties for Millstone Unit 3 5 15 1 Reactor Vessel Weld HAZ Metal i 5-5 Charpy Impact Specimen Fracture Surfaces for Millstone 5-16 Ur.it 3 Reactor Vessel Shell Plate B9805-1 (Longitudinal c j Orientation) i a 5-6 Charpy impact Specimen Fracture Surfaces for Millstone 5-17 l Unit 3 Reacto Vessel Shell Plate B9805-1 [ (Transverse Orientation) 1 1 5-7 Charpy Impact Specimen Fracture Surfaces for 5-18 M111stor.e Unit 3 Reactor Vessel Weld Metal i i l 5-8 Charpy impact Specimer Frac +.ure Surfaces for 5-19 Millstone Unit 3 Reactor Vessel Wald HAZ Metal l mm Sm. ygg

l LIST OF ILLUSTRATIONSs(Cont) 1 Figure Title Page 5-9 Tensile Properties for Millstone Unit 3 Reactor 5-20 Vessel Shell Plate B9905-1 (Longitudinal Orientation) 5-10 Tensile Properties for Millstone Unit 3 Reactor 5-21 Vessel Shell Plate B9805-1 (Transverse Orientation) 5-11 Tensile Properties for Millstone Unit 3 Peactor 5-22 i Vessel Weld Metal t 5-12 Fractured Tensile Specimens from the Mil' stone Unit 3 5-23 l Reactor Vessel Shell Plate B9805-1 (Longitudinal Orientation) L 5-13 Fractured Tensile Specimens from the Millstone Unit 3 5-24 Reactor Vessel Shell Plate B9805-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens from the Millstone Unit 3 5-25 Reactor Vessel Weld Metal l 5-15 Typical Stress-Strain Curva for Tension Specimens 5-26 i i \\ 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 6-2 Core Power Distributions Used in Transport Calculations i for Millstone Unit 3 6-14 ( l I 6-3 Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux and dpa Within the Reactor Vessel Wall 6-15 -l -l m,,. m. yisi 4

J u r LIST OF TABLES-Table Title Page 4-1 Chemical Composition and Heat Treatment of the 4-3 Millstone Unit 3 Reactor Ver ? Surveillance Materials 5-1 Charpy Y-Notch Impact Data for the Millstone Unit 3 5-5 Reactor Vessel Shell Plate B9805-1 Irradiated 18 at 550'F, Fluence 4.32 x 10 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Milletone Unit 3 5-6 i Recctor Vessel Weld Metal and HAZ Metal Irradiated at 550'F, Fluence 4.32 x 10 n/cm2 (E > 1.0 MeV) 18 5-3 Instrumented Charpy Impact Test Results for Millstone !)-7 Unit 3 Reactor Vessel Shell Plate B9805-1 5-4 Instrumented Charpy Impact Test Results for S-E I Millstone Unit 3 Reactor Vessel Weld Metal and HAZ Metal 18 2 5-5 The Effect cf 550'F Irradiation at 4.32 x 10 n/cm 5-9 (E > 1.0 MeV) on the Notch Toughness Pro;.wrties of The Millstone Unit 3 Reactor Vess0 thterials t i 5-6 Comparison of Millstone Unit 3 Reactor u m. i 'scillance " 10 I f Capsule Charpy Impact Test Results with Jw U y Guide 1.99 Predictions e 5-7 Tensile Properties for Willstone Unit 3 Reactor Vessel 5-11 18 i Waterial Irradiated to 4.32 x 10 n/cm2(E>1.0MeV) I _ -, -, - - __ _ ~

y [ LIST OF TABLES (Cont) Table Title Page 6-1 Calculated Fast Neutron Exposure Parameters at the 6-16 Surveillance Capsule Center 6-2 Calcelated Fast Neutron Exposure Parameters at the 6-17 Prossure Vessel Clad / Base Metal Interface 6-3 Welative Radial Distributions of Neutron Flux 6-18 (E>1.0 MeV) Within the Pressure Vessel Wall 6-4 Relattve Radial Distributions of Neutron Flux 6-19 (:>0.1 MeV) Within the Pressure Yessel Wall 6-5 Relative Radial Distribution of Iron Displacement 6-20 Rate (dps) hithin the Pressure Vessel Wall f6 Nuclear Parameters for Neutron Flur Mor.itars 6-21 6-7 Irradiation History of Neutron Sensors Contained 6-22 in Capsule U 6-8 Measured Sensor Activities and Reaction Rates 6-23 6-9 Summary of Neutron Dosimetry Recults 6-25 l 6-10 Comparts'.n of Meksured and FERRET Calculated 6-26 1 Ru ns at the Survoiliance Capsula Center pectrum at the Surveillance 6-27 6-11 Ad.. Ct. /. ~ i v n..s e <, y 4 ~~-e -e, +

e LIST OF TA8l.ES (Cont) f Table' Title Page 6-12 Comparison of Calculated and Measured Exposure 6-28 Levels for Capsule U l 6-13 Neutron Expbsure Projections at Key Locations on the 6-29 Pressure Vessel Cled/ Base Metal Interface 6-14 Neutron Exposure Values for Use in the Generation 6-30 of Heatup/Cooldown Curves 6-15 Opdated Lead Factors for Millstone Unit 3 Surveillance 6-31 Capsules t ( O ' e I h L I l i l. 1 1 i f l mu,m, 1

l-SECTION 1 SUW4ARY OF RESULTS The analysis of the reactor vessel material contsined in Capsule U, the first surveillance capsule to be removed from the Northeast Utilities Service Company Millstone Unit 3 reactor pressure vesul, led to the following conclusions: l o The capsule received an average fast neutron fluence (E > 1.0 MeV) i 18 2 of 4.32 x 10 n/cm, o Irradiation of Charpy V-notch impact specimens from the reactor 18 2 vessel intermediate shell Plate 9805-1, to 4.32 x 10 n/cm, resulted in 30 and 50 ft-lb transition temperature increases of 30'F, for specimens oriented parallel to the mtj~~ working direction l (longitudinal crientation) and increases of 35'F for specimens oriented normal to the major working direction (transverse I orientation). t 10 2 o Wald metal impact specimens irradiated to 4.32 x 10 n/cm [ i resulted in 30 and 50 ft-lb transition temperature increases of 40*F l and 35'F respectively. 18 2 o Irradiation to 4.32 x 10 n/cm resulted in a 3 f t-lb decrease in the average upper shelf energy of Plate B9605-1 (transverse l orientation) while the limiting weld metal decreased by 8 ft-lb from i 143 to 135 ft-lbs. Both materials exhibit a more than adequate shelf level for continued sate plant operation. o Comparison of the 30 ft-lb transition temperature increases for the Millstone Unit 3 surveillance material with predicted increases j using the methods of NRC Regulatory Guide 1.99, Proposed Revision 2. shows that the plate material and weld metal transition temperature f herease were 11' and 15' respectively higher than predicted. This is bounded by the 2 sigma allowance for shift prediction of 34*F for base metal and 56'F for weld metal. f i 11

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The surveillance capsule test results do not indicate any significant chtnges in RT values projected for the reactor vessel, and therefore, a low risk NDT of vessel f aili*e from pressurized thermal shock (PTS) events is postulated. Plant heatup and cooldown limit curves are presented in Apppendix A of this report. These curves were developed per the methods of the proposed Revision 2 to the Regulatory Guide 1.99, which identified an exponetial attentuation factor which differs from that adopted in the final revision of Revision 2. However, although the attenuation factors differ, a review of the ARTNOT determined at the vessel 1/4 and 3/4 thickness indicates that the values used to develop the limit curves are censervative since they are higher than those i which would be obtained using the final revisien of Revision 2. 1 i l i i I i i u n.unu ie 1-2

SECTION 2 e INTRCOUCTION I This report presents the results of the examination of Capsule U, the first f capsule to be removed from the reactor in the continuing surveillance program ~ which monitors the effects of neutron irradiation on the Millstone Unit 3 9 reactor pressure vessel materials under actual operating conditions. The surveillance program for the Millstone Unit 3 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the prairradiation mechanical properties of the rea'ctor vessel materials are presented by Singer.Ill The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel and was brsed on ASTM E-185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactors Vessels". Westinghouse Power Systems personnel were contracted to aid in the preparation of procedurer, for removing the capsula from the reactor and its shipment to the Westinghouse Research and Development l. Laboratory, whera the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed. This recort summarizes testing and the postirradiation data ebtained from surveillance Capsule U removed from the Millstone Unit 3 reactor vessel and discusses the analysis of the data, t 1 l l l l. i 1 l l l r I P v wie""" 2-1 I i

SECTION 3 s BACXGROUND 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 moi,t critical region of the vassal because it is subjected to significant fast neutron exposure. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA 533 Grade B Class 1 (base material of the Hillstone Unit 3 reactor pressure vessel beltline) are well documented in the lite.ature. 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 rantor pressure vessels has been presented in "Protection Against Non-ductile Failure," Appendix G to Section III of the ASME biler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the ?. reference nil-ductility temperature (RTNDT)* RT is defined as the greater of either the drop weight nil-ductility NDT transition temperature (NDTT per ASTM E-208) or the temperature 60'F less than the 50 ft Ib (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. The RT f a given material is used to index that NOT material to e reference stress intensity factor curve (KIRcurve)which appears in Appendix G of the ASHE Code. The KIR curve is a lower bound of dynamic, crack arrest, sna static fracture toughness results obtained from several heats of oressure vessel steel. When a given material is indexed to I

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the K g curve, allowable stress intensity factors can be obtained for this g material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors. RTNDT and, in turn, the cperating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlecent or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Millstone Unit 3 Reacto'r Vessel Radiation Surveillance Program,III 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 Ib temperature (ARTNDT) due to irradiation is added to the original RTNDT for radiation embrittlen. ant. This adjusted RTNDT to adjust the RTNOT initial + ARTNDT) is used to index the material to the KIR (RTNDT 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. i I L vn nurse 3.g

4 SECTION 4 DESCRIPTION Ci PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Millstone Unit 3 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules we.e positioned in the reactor vessel between the neutron shield pads and the vessel wall at locations shown in figure 4-1. The vertical center of the capsules is oppesite the vertical center of the core. Capsule U (Figure 4-2) was removed af ter 1.3 effective full power years of plant operation. This capsule contained Charpy V-notch impact, tensile, and 1/2T - Compact Tension fracture mechanics specimens from the reactor vessel intermediate shell ring Plate B9805-1, submerged are weld metal identical to the beltline region girth and 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 B9805-1 of the representative weld. The chemistry and heat treatment of the surveillance material are presented in table 4-1. The chemical analyses reported in table 4-1 were obtained from unirradiated material used in the surveillance program. 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 quenened end of the plate. All base metal Charpy Y-notch impact and tensile specimens were oriented with the longitudinal axis of the specimen both normal to (transverse orientation) and parallel to (longitudinal orientation) the principal acrking direction of the plate. Charpy V notch specimens from the weld metal were oriented with the longitudinal axis of the specimens un. iomma 41

transverse to the weld direction. Tensile specimens were oriented with the lorgitudinal axis of the specimens normal to the welding direction. The 1/2T Compact Tension (CT) test 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 specimens and parallel to the weld direction for weld specimens. All specimens were f atigue procracked per ASTM E399-70T. i Cepsule U contained dosieeter wires of pure iron, copper, nickel, and unshielded aluminum-cobalt. In addition, cadmium-shielded dos meters of Neptunium (Np237) and Uranium (U230) were contained in the capsule. \\ Thermal monitors made from twc low-me'. ting eutoctic alloys and sealed in Pyrex tubes were included in the capsule ar.d were located as shown in Figure 4-2. 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 (510'C) The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule U are shown in Figure 4-2.

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TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE MILLSTONE UNIT 3 REACTOR VESSEL SURVEILLANCE MATERIALS ChemicalComposition(wt%) Element Inter. Shell Plate B9805-1 Weld Waterial C 0.23 0.22 0.24 0.14 0.14 Mn 1.32 1.29 1.33 1.24 1.25 P 0.010 0.018 0.018 0.010 0.011 S 0.010 0.014 0.012 0.009 0.009 Si 0.21 0.23 0.13 0.15 0.12 Ni 0.61 0.62 0.06 0.15 0.03 No 0.57 0.58 0.52 0.54 0.48 f Cr 0.03 0.10 0.067 0.01 0.03 Cu 0.05 0.045 0.038 0.02 0.07 l A1 0.024 0.024 0.003 0.004 l Co 0.012 0.017 0.006 0.006 1 0.001 0.001 Pb W <0.01 <0.01 <0.01 0.01 Ti <0.01 0.006 0.006 0.01 Zr <0.001 <0.002 <0.002 0.001 V 0.006 0.004 0.003 0.005 0.004 l Sn 0.003 0.005 0.005 0.004 As 0.005 0.007 0.006 0.006 Cb <0.01 0.002 <0.002 0.01 N 0.007 0.006 0.010 0.012 2 B <0.001 <0.001 <0.001 0.001 Heat Treatment History [ Material Temperature ('F) Time (Hr) Coolant j Intermediate Shell Austenitizing 1575-16!5 4 Water quanched (Plate B9805-1) Tempered 1200-1250 4 Air cooled Stress Relief 1100-1200 22 Furnace cooled Wald Metal 'I Stress Relief 1100-1200 8 Furnace cooled [ I r i N inis weldment was fabricsted by Combustion Engineering, Inc., using 3/16 l inch Mil B-4 weld filler wire,. heat number 4P6052 and Linde 0091 flux, lot t number 0145 and is identical to that used in the actual fabrication cf the reactor vessel intermediate to lower shell girth weld and associated longitudinal weld seams. J l m.awsu 43

O' REACTOR VESSEL CORE BARREL NEUTRON PAD (309 5') Z CAPSULE U (58 5') V (61 ') M"58. 5 ' 58.5' 61' ,f" 90' 270' ] lp (2 41 ') Y EM (238 5') x s 180' PLAN VIEW 9 Figure 4-1. Arrangement of surveillance capsules in the reactor vessel o n. m.ev u 4-4

4* / LEGEND: EL . INTERMEDIATE SHELL PLATE B9805-1 (LONGITUDINAL) ET INTERMEDIATE SHELL PLATE B98051 (TRANSVERSE) EV/. WELD METAL EH . HEAT AFFECTED ZONE MATERIAL cow 84Cf C 0 wM CT C0w*eCf CowpaCt WW t in tat t itasca Ital @4 C M&A *f C8eaa PT Cnaape t e ntcas tg espoes CMAA89 Ci*8 IW1 (US ( *' t fail twil tag gg got 14 ("I fut two twt twt (#1 twS4 into tws t t e11 gag tg gte IJ ttj ggi rm qq tu? l tai rWie ses twie

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M r.... o .s im -- ,.,,,....n I i I { Figure 4-2. Capsule U Diagram Showing Location of lipcimens, i Thermal Nonitors, and Dosisators c ni 4-5 i i i, ~2'E R \\(oO 5/p2 -6 i I ,t a w r'k

l l l SECTION 5 i TESTING OF SPECIMENS FROM CAPSULE U l 1.- 1' 5-1. OVERVIEW l i The postirradiation mechanical testing of the Charpy V-notch and tentile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and H(2), ASTM Specification E185-82 and Westinghouse Procedure NHL 8402, Revision 0 as modified by RMF Procedures 8102 and 8103. Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked i against the master list in WCAP-10732II) No discrepancies vere found. Examination of the two low-melting 304'C (579'F) and 310'C (590'F) eutectic alloys indicated no melting of either ty:>e of ther. mal monitor. Based on this l examination, the maximum temperature to which the test specimens were exposed was less than 304'C (579'F). f The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74, 358J r4 chine. The tup (striker) of [ the Charpy machine is instrumented with an Effee a Technology model 500 f instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard wasurement of Charpy energy ( (E ). From the load-time curve, thw lead of general yielding (Pgy),the l D time to general yielding (tgy), the maximum load (P ), and the time to [ g maximus, load (t ) can be determined. Under some test conditior.s. a sharp g i l i i line 19442?N 5-1

drop in lead indicative of fast fracture was observed. The load at which fasr fracture was initiated is identified as the fast fracture load (P ), and the p load at which fast fracture terminated is identif'ad as the arrest load (P )* A The energy at maximum load (E ) was determined by comparing the energy-time g record and the lead-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E ) is the difference p betneen the total energy to fracture (E ) and the energy at maximum load, D r The yield stress (c ) is calculated from the three point bend formula. y The f'.ow stress is calculatsd from the average of the yield and maximum leads, 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 expansion was measured using a dial gage rig similar to that shown in l the sarre specification. e Tension tests were performed on a 20,000 pound Instron, split censole test machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF l 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 axialit/ of leading. The tests were conducted at a constant l crosshead speed of 0.05 inch per minute throughout the test. f r Deflection r4asurements were made witn a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-f loaded to the specimen cnd operated through speciren failure. The extenso-meter gage length is 1.00 inch. The extensometer is rated as Class B-2 per l ASTM E83-67. 1 Elevated test temperatures were obtained with a three-zone elactric resistance split-tube furnace with a 9-inch hot zene. All tests were conducted in air. r aw tatwu 52

I l i ] Because of the difficulty in remotely attaching a thermocouple directly to the i specimen, the following procedure was used to monitor specimen temperature. l Chromel-alumel thermocouples were inserted in shallow holes in the center and i ll each end of the jage section of a dummy specimen and in each grip. In the j. test configuration. with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was f i-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. { 2 Experiments indicated that this method is accurate to plus or minus 2'F. 1 i i } The yield lead, ultimate lead, fracture load, total elongation, and uniform l l elongation were determined directly from the lead-extension curve. The yielu ) strength, ultimate strength, and fracture strength were calculated using the j criginal cross-sectional araa. The final diameter and final gage length were determined from postfracture photographs. The fracture area u;.ed to calculate l the fracture stress (true stress at fracture) and percent reduction in area {. was computed using the final diameter measurement. i I'. 5.2. CHARPY V-NOTCH IMPACT TEST RESULTS l )- The results of Charpy V-notch impact tests performed on the various materials contained in Capsule U irradiated to approximately 550'F at 4.32 x 1018 l 2 J n/cm are presented in Tables 5-1 through 5-4 and Figures 5-1 through 5-4. The transition temperature increases and upper shelf energy decreases for the j Capsule U materiai are shewn in Table 5-5. { j l Irradiation of the vessel intermediate shell Plate B9805-1 material l (longitudinal orientatien) specimens to 4.3I x 10 n/cm2 (Figure 5-1) l 18 ) resulted in a 30 and 50 ft-lb transition temperature increase of 30*F, and no 3 upper shelf energy decrease d en compared to the unirradiated data, f j t l. Irradiation of the vessel intermediate shell Plate 9805-1 material (transverse orientation) specimens to 4.32 x 1018,je,2 (Figure 5-2) resulted in both f l. 30 or 50 ft-lb transition temperature increases of 35'F and an upper shelf f i energy decrease of 3 f t-lb when ccepared to the unirr&diated data. I I vt M east'es $,3 i m -,-,-,.-..,--.----..-,-..,g--, ,__m _,,r--_,,,._--, ,,-..~._,.,,-,.. --l

Weld metal irradiated to 4.32 x 10 n/cm2 (Figure 5-3) resulted in a 30 18 and 50 f t-lb transition tsmperature increase of 40*F and 35'T resper.tively, and an upper shelf energy decrease of 8 ft-lb. Weld HAZ metal irradiated to 4.32 x 10 n/cm2 (Figure 5-4) resulted in no 18 30 and 50 ft-lb transition temperature increases of 30 and 25'F mspectively and an upper shelf energy decrease of 2 ft-lb. The fracture appearance of each irradiated Charpy specitaen from the various materials is shown in Figu.es 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasing test temperature. Table 5-6 shows a comparison of the 30 ft-lb transition temperature (t.RTNDT) increases for the various Millstone Unit 3 surveillance mdterials with predicted increases using the methods of proposed NRC Regulatory Guide 1.99, Revision 2.I33 This comparison shows that the transition temperature 18 2 increase resulting from irradiation to 4.32 x 10 n/cm is 6' to 11' higher than predictad by the Guide for Plate B9805-1. The weld metal 18 2 transition temperature increase resulting trom 4.32 x 10 n/cm g, 33 higher than the Guide prediction. 5-3. TENSION TEST RESULTS The results of tension tests performed on Plate B9805-1 (longitudinal and 18 2 transverse orientation) and weld metal irradiated to 4.32 x 10 n/cm are shown in Table 5-7 and Figures 5-9, 5-10 and 5-11, respectively. These results show that irradiatien produced less than a 5 Ksi increase in 0.2 percent yield strength for pit.te B9805-1 and the meld metal, fractured tension specimens for each of the materials are shown in Figures 5-12, 5-13 and 9 14. A typical stress strain curve for the tension specimens is shown in Fipre 5-15. 5-4. CCNPACT TENSION TESTS Per the surveillance capsule tes'.ing contract with Northeast Utilities Service Company,1/2T - Cogact Tension f racture Mechatiics specimens will not be tested and will be stored at the Hot Cell at the Westinghouse R&D Center. u n. * *

• 5-4

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE MILLSTONE UNIT 3 REACTOR VESSEL SHELL PLATE B9805-1 18 IRRADIATED AT 550*F, FLUENCE 4.32 x 10 n/cm2 (E > 1.0 MeV) Temperature Impset Energy Lateral Expansion Shear 5saole No. O',), (*C) (ft-lb) 21 1p,113), (an) ,, %) ( LonnitudinaS Orientation EL4 - 25 -32 8.0 15.0 9.0 0.23 5 EL15 10 -12 10.0 13.5 10.0 0.25 10 EL9 20 -7 23.0 31.0 21.5 0.55 15 EL8 25 -4 29.0 39.5 27.0 0.69 15 EL3 25 -4 30.0 40.5 27.5 0.70 20 EL13 40 4 32.0 43.5 28.0 0.71 20 EL12 50 10 39.0 53.0 36.0 0.91 25 EL14 50 10 59.0 80.0 58.0 1.47 30 EL11 75 24 72.0 97.5 54.5 1.38 55 EL7 78 26 58.0 78.5 46.0 1.17 40 EL2 100 38 70.0 95.0 52.0 1.32 55 EL1 150 66 97.0 131.5 68.5 1,74 70 EL6 200 93 132.0 179.0 85.0 2.16 100 EL5 300 149 143.0 *194.0 84.0 2.13 100 EL10 400 204 '29.0 175.0 83.0 2.11 100 Transverse Orientation ET1 - 25 -32 3.0 4.1 3.0 0.08 5 ET9 10 'J 12.0 16.5 13.0 0.33 10 ET10 25 -4 16.0 21.5 16.5 0.42 15 ET2 40 4 33.0 44.5 28.0 0.71 20 ET4 50 10 32.0 43.5 30.0 0.76 20 ET3 50 10 36.0 4J.0 30.0 0.77 20 ET15 60 16 33.0 44.5 30.0 0.76 20 ET5 75 24 55.0 74.0 34.5 0.88 45 ET13 78 26 45.0 61.0 39.0 0.99 35 ET6 100 38 62.0 84.0 48.0 1.22 45 ET8 150 66 68.0 92.0 55.0 1.40 55 ET7 200 93 107.0 145.0 74.0 1.88 100 ET14 300 149 MACRINE MALFUNCTION ET11 300 149 107.0 145.0 75.0 1,91 100 ET12 400 204 110.0 149.0 73.0 1.85 100 ou.., -, 33

TABLE 5-2 CHARPY Y-NOTCH IMPACT DATA FOR THE MILLSTONE UNIT 3 REACTOR VESSEL WELD AND HAZ METAL IRRADIATED AT 550*F FLUENCE 4.32 x 10 n/cm2 (E > 1.0 MeV) 18 Temperature Impact Energy Lateral Expansion Shear Ssaole No. J'Z), f'C), fft-lb) H}, (sils) f_ag), (0 Weld Wetal EW10 -100 -/3 1.0 1.5 1.0 0.33 0 EW15 -50 -46 6.0 8.0 7.0 0.18 5 EW3 -25 -32 10.0 13.5 11.0 0.28 20 EY8 -10 -23 10.0 13.5 10.0 0.25 20 EW9 0 -18 27.0 36.5 22.5 0.57 40 EW6 0 -18 82.0 111.0 55.0 1.40 80 i EW1 10 -12 35.0 47.5 33.0 0.84 45 EW13 10 -12 30.0 40.5 27.0 0.69 50 EW4 20 -7 87.0 118.0 60.0 1.52 80 EW11 25 -4 29.0 39.5 27.5 0.70 65 EW2 25 -4 73.0 99.0 54.5 1.38 70 EWS 78 26 110.0 149.0 78.0 1.98 05 EW15 150 66 134.0 181.5 89.0 2.26 100 EW14 250 121 144.0 195.0 88.0 2.24 100 EW7 350 177 127.0 172.0 87.5 2.22 100 RAZ Wetal EH11 -200 -129 2.0 2.5 2.0 0.05 1 EH13 -150 -101 2.0 2.5 2.0 0.05 1 EH7 -125 - 87 10.0 13.5 8.0 0.20 5 l EH6 -100 - 73 57.0 77.5 31.0 0.79 40 EH1 -100 - 73 63.0 88.5 35.5 0.90 40 EH2 -100 - 73 25.0 38.0 18.0 0.46 20 EH8 59 64.0 87.0 34.0 0.86 40 EH10 46 MACHINE MALFUNCTION r EH4 59 45.0 61.0 26.5 0.67 40 EH15 46 64.0 87.0 40.0 1.02 45 EH14 0 - 18 121.0 164.0 62.5 1,59 75 EH3 78 26 110.0 149.0 73.0 1.85 90 1 EH9 150 66 145,0 196.5 88.0 2.24 100 E312 250 121 138.0 187.0 87.0 2.21 100 EH5 350 188 138.0 187.0 85.5 2.17 100 [ i l on.io""" 5-6

l TABLE 5-3 INSTRtK MTED CP.ARPY INPACT TEST RESULTS FOR MILLSTONE UNIT 3 i REACTOR VESSEL SHa l PLATE 89805-1 Normalised Energies Test C:arpy Charpy Maximus Prop Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Ee/A Ep/A Load to Yield Load Maximus Load Load Stres: Stress Number f*FJ (itab) (ft-lb/in ) (kips) (psec) (kips) (psee) (kips}_, (kipsl (ksi) (ksi) longitudinal Orientation EL4 - 25 8.0 64 69 - 4 3.80 90 4.15 175 4.10 126 131 ELIS 10 10.0 81 66 15 3.65 90 4.05 175 4.00 0.25 120 127 EL9 20 23.0 185 168 17 3.55 90 4.35 375 4.30 117 130 EL8 25 29.0 234 201 32 3.55 85 4.50 435 4.45 117 133 EL3 25 30.0 242 242 - 0 3.50 85 4.70 505 4.70 115 136 EL13 40 32.0 258 233 25 3.50 90 4.60 500 4.60 0.20 115 134 EL12 50 39.0 314 289 25 3.45 85 4.65 605 4.60 0.40 113 133 EL14 50 59.0 475 319 156 3.65 320 4.95 765 4.8 0.35 120 142 Y EL11 75 72.0 580 348 232 3.30 85 4.60 730 4.35 1.40 109 130 EL7 78 58.0 467 315 151 3.30 105 4.65 680 4.45 1.15 109 131 EL2 100 70.0 564 391 172 3.25 80 4.30 820 4.45 1.25 107 129 EL1 150 97.0 781 389 392 3.15 80 4.45 825 4.10 2.30 104 127 EL6 200 132.0 1063 352 711 2.80 70 4.45 770 93 120 91 116 ELS 300 143.0 1151 329 823 2.75 110 4.25 765 ELIO 400 129.0 1039 260 779 2.65 50 4.10 615 87 111 Transversa Orientation ETI - 25 3.0 24 ET9 10 12.0 97 66 30 3.65 85 4.05 170 3.75 0.30 120 127 ET10 25 16.0 129 112 17 3.65 85 4.15 260 4.15 0.15 121 130 ET2 40 33.0 266 218 48 3.40 90 4.50 475 4.50 0.55 113 130 ET4 50 32.0 258 228 30 3.55 90 4.55 485 4.55 0.45 118 134 j ET3 50 36.0 290 237 53 3.45 85 4.55 505 4.50 0.65 114 133 1 ET15 60 33.0 266 232 34 3.40 85 4.50 500 4.50 C.70 113 131 i ETS 75 55.0 443 313 129 3.10 105 4.60 680 4.55 1.40 102 127 l ET13 78 45.0 362 316 46 3.40 90 4.55 665 4.50 1.00 112 131 l ET6 100 62.0 499 307 193 3.20 85 4.45 665 4.40 1.70 106 126 l ET8 150 68.0 548 306 241 3.15 85 4.45 665 4.30 2.10 103 125 l ET7 200 107.0 862 342 520 2.80 80 4.35 770 93 118 ET14 300 MACHINE MALFUNCTION ET11 300 107.0 862 323 539 2.60 55 4.20 735 85 112 ET12 400 110.0 896 276 609 2.4 65 3.95 665 79 105

TABLE 5-4 INSTRUMENTED Cr3RPY IMPACT TEST RESULTS FOR MILLSTONE UNIT 3 REACTOR VESSEL WELD METAL AND HAZ METAL Normalized Enerr.ics Test Charpy Charpy Mr.ximus 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 Number G (ft-lb) (ft-lb/in ) (kips) (psec) (kips) (psec) (kips) (kips) (ksi) (ksi) Weld Metal EW10 -100 1.0 8 9 - 1 0.90 15 2.15 55 2.05 29 50 EW12 - 50 6.0 48 42 6 3.80 100 4.25 130 4.24 126 133 EW3 - 25 10.0 81 48 l't 4.25 95 4.45 130 4.45 0.30 140 144 EW8 - 10 10.0 81 32 49 2.85 60 4.20 100 4.15 0.75 94 117 EW9 0 27.0 217 142 76 4.3 95 4.65 300 4.60 1.15 133 143 EWS O 82.0 660 260 400 4.05 95 4.95 505 4.50 1.70 135 150 EW1 10 35.0 282 182 100 3.85 85 4.65 370 4.60 1.65 128 141 T EW13 10 30.0 242 127 115 3.95 90 4.55 275 4.50 1.60 132 141 EW4 20 87.0 701 343 358 3.85 85 4.90 655 4.50 2.40 127 145 EW11 25 29.0 234 84 149 3.90 90 4.20 200 4.20 1.80 129 134 EW2 25 73.0 588 274 314 4.35 275 5.40 575 5.15 2.35 144 161 EW5 78 110.0 886 334 552 3.60 85 4.85 655 3.75 2.45 120 140 EWIS 150 134.0 1079 382 751 3.40 80 4.65 670 113 133 EW14 250 144.0 1160 347 813 3.2 90 4.50 735 105 127 EW7 350 127.0 1023 301 722 3.00 95 4.35 6S5 100 122 HAZ Metal EH11 -200 2.0 16 18 - 1 2.65 60 3.30 75 3.30 0.30 88 99 EH13 -150 2.0 16 24 - 8 3.00 55 3.95 80 3.95 99 115 EH7 -125 10.0 81 93 - 13 4.65 105 5.25 200 5.25 154 164 Ell 2 -100 28.0 225 10' 34 3.80 95 5.10 375 5.00 0.40 125 147 EH6 -100 57.0 459 JS 164 4.50 90 5.60 505 5.40 149 It,d EH1 -100 63.0 M7 282 225 43 90 5.40 505 5.20 0.65 142 161 EH4 - 75 45.0 262 292 70 4.40 95 5.40 520 5.30 1.15 145 162 EH8 - 75 64.0 515 277 238 4.10 85 5.25 505 4.85 0.40 136 155 EH10 - 50 MACHINE MALFUNCTIGN EHIS - 50 64.0 515 376 140 4.40 105 5.40 675 4.90 145 162 EH14 0 121.0 97 368 606 3.80 165 5.45 695 4.10 2.30 126 153 EH3 78 110.0 .s86 301 585 3.60 95 4.90 595 4.05 3.35 120 141 EH9 150 145.0 1168 365 802 3.60 275 4.95 805 120 - 142 EH12 250 138.0 1111 352 759 3.15 100 4.60 745 104 128 EHS 350 138.0 till 305 807 3.10 95 4.45 665 103 125

e TABLE 5-5 THE EFFECT OF 550*F IRRADIATION AT 4.32 x 1018,fc,2 (E > 1.0 MeV) ON THE NOTCH TOUGMESS PROPERTIES OF THE i MILLSTONE UNIT 3 REACTOR VESSEL MATERIALS Average Average 35 atI Average Average Energy Aesorptten 30 f t-lb Temp (*F) Lateral Espansson Temp (*F) 50 f t-lb Tome (*F) at Fult Shear (ft-lb) Meterial Untrradiated trradtated AT untrradiated Irradtated AT Untrradtated Irradiated AT untrradiated Irradtated A(ft-te); Plate 99005-1 5 35 30 30 45 15 35 65 30 133 135 +2(a) (longitudinal) Plate.99905-1 5 40 35 35 65 30 60 95 35 til 808 -3 4 (transverse) neeld m2tal -35 5 40 -20 15 35 -15 20 35 143 135 -8 HAZ Isetal -130 -100 30 -105 -65 40 G -70 25 142 140 -2 (a) Increase in se.etf energy i LJ 3929e 992988 90 _ -,. _ _ _ _ _.. - ~ - - - _ _ _ _ _ _,,. _,, _ _

IABLE 5-6 C(BFARISON OF MILLSTONE IsilT 3 REACIOR VESSEL SURVEILLANCE CAPSULE CWJtPY lerACT TEST are T; WITH REGULATORY GUIDE 1.99 PREDICTIONS ARTg y (30 ft-lb Increase) & USE Fluence R.G. 1.99 Rev.2 Measured R.G. 1.99 Rev. 2 Aeasured I9 2 Material Capsule 10 n/cm g.7) g.7) (g) (g) Plate B9805-1 0 0.432 24 30 15.5 0 7 (longitudinal) ?$ Plate 89835-1 0 0.432 24 35 15.5 2.5 l (transverse) Weld Metal U 0.432 25 40 15.5 5.5 30 1.5 HAZ Metal U 0.432 mu n.

TAPLE 5-7 TEIIS~LE PROPERTIES FOR IIILLST0hE tal!T 3 REACTOR VESSEL IIATERIAL IRRADIATED TO 4.32 x 10 n/cm2 (E > 1.0 IIeV) I8 Tert 0.25 Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Trop. Strength Strength Load Stress Strength Bloegation Elongation in Area Material Nueber M (ksil (ksi) (kip) (ksi) (ksi) (5) (5) (5) Plate BbMS-1 Long. ML1 71 67.2 90.2 2.80 217.3 57.0 12.0 26.6 74 Long. EL2 200 63.7 87.6 2.63 177.1 53.6 9.3 23.8 70 Long. EL3 550 58.6 86.6 2.70 175.4 55.0 11.3 23.6 60 Plate T B0806-1 0 Trans. BT1 71 64.9 90.7 2.95 166.9 80.1 12.0 27.0 64 Trans. ET2 200 63.4 83.5 2.70 161.3 55.0 11.3 23.7 66 Trans. ET3 550 57.6 85.6 3.10 142.0 63.2 10.5 20.9 56 Weld EW1 25 83.0 99.8 3.05 225.3 62.1 1G.F 23.9 72 Weld EW2 150 77.4 93.7 2.90 217.3 57.0 9.0 22.8 74 Weld EW3 550 73.3 95.7 3.15 164.2 64.2 10.5 20.8 61 s.n. ees,ee.e

( C) -150 -100 -50 0 50 100 150 200 250 [ 3 ',2 _',2 100 f 80 $M 2 n 2 m O 2 20 2 0 iN I i i i i 100

2. 5 i

i i i i i i i j80 > =. - g

2. 0 1.5'E

~M @m

1. 0 E 9

0 15 F %N 2 0.5 0 i i i 1 0 i i i i i 6 i l 180 20 I 160 2

• /

200 ,_,10 "o e Unftradiated o g 1N o gg ~ 100 l @ 80 120 - Irradiated at.550*F = 18 /cm2 O 4.32 x 10 n o 80 g e g 30 F l l 40 30 F 40 0 i i i 0 - 200 -100 0 100 200 300 400 500 Temperature ( F)

• l figure 5-1 Charpy V-notch Impact Properties for Hillstone Unit 3 Reactor Vessei Shell Plate B9805-1 (longitudinal orientation) l imsma ie 5-12

( C) -150 -100 - 50 0 50 100 150 200 250 i i i i i i i i i 100 3 2 _,2 $ 80 2 >m E 2 m 40 2 2' o 20 2 s2 0 M 2.5 i i i i i i i i i 2 2.0 = 30 7_ e 2

1. 5 ?!

- 60 N40 1.0 30*F $ 20 2 0.5 0 t i i i I 0 e M i i i i i i i i i 180 240 160 200 _ 140 o g120 Unirradiated 9 160 m 100 @ 80 120 C Irradiated at 550 F = oe,N 5 18 2 1 80 60 2 4.32 x 10 n/cm 5*F M O 35*F M 0 i i i i i i 0 - 200 -100 0 100 200 300 400 500 Temperature ( F) Figure 5-2 Charpy V-notch Impact Properties for Hillstone Unit 3 Reacter Vossel Shell Plate B9805-1 (transverse orientation) vn.-em u is 5-13

( C) -150 -100 - 50 0 50 100 150 200 250 i i i i i i 3 '2 100 $ 80 2 %w ~ .E m m m 0 2.5 100 i i i i i i 1 5 80 75 2.0 e e 1.5'E 60 o IO - -35 F

1. 0 0.5 f20 0

i i i i i 0 200 i i ,?e 180 160 2 200 7e. im 5 Unirrad!.w.1 o o lO lD d I ~ 100 o i!R 120 C b 80 Irradiated at 550 F 0 4.32 x 1018 n/cm2 80 g -35 F

1. ~

l 40 F 40 j=u M t 0 i i i i i i 0 - 200 -100 0 100 200 300 00 500 Yemperature ( F) l Figure 5-3 Charpy V-notch Irpact Properties for Hillstone Unit 3 Reacter Vossel Weld Metal 1 v n.-ew a,e 5 14

( o C) -150 -100 - 50 0 50 100 150 200 250 T I I ig Ig i i i i /g 100 o 3 80 60 2 o m m 2 v2 N 2 0 e

'I I

I I i i I 100 2.5 i 1 I b i. I I 1 I .-E 80 o 79 2, 0 o 3 E T o \\8 8 s2 7 60

1. 5 'E 8 40 8

1.0 40 F o, o ( 20 ' 0.5 "0 I i i i i i I 0 i i i i i i i i i 180 240 o l 160 Unirradiated 9. N0 ig Or w w g 120 160 o 8 Irradiated at 550 F ~ 100 18 2 WC 80 8 4.32 x 10 n/cm o 5 Bo o 80 M o 25 F //*30F 40 40 i i i i i i 0 - 200 -100 0 100 200 300 400 500 Temperature ( F) i Figure 5-4 Charpy V-notch lepact Properties for Millstone Unit 3 Reacter Vessel Weld HAZ Hetal 3121s C421641' 5-15 l

i i l k9 I

• g.

['}Q

-;b, _.;.(

'.. *5.% ,3.q p i .,1 o; ! !...' TC p m.,, aj k stw l M +.e.' 7

..R I j

l f'.T.: Q:.I.

j,.c 1

s 4 e o ,N w

e. -

1 l t e j l E4 E15 E9 E8 E3 l l + l ( l Ag,~xn -I h,., l .~

gtg-g i

e..e l .x J i v '. - 1 y.a'AJ l 'i i s I [ m . = -? y, _ A T& (,.,- - + . :g. i~U ) J _-.-. ,,.., w j .~. _ E13 E12 E14 E11 EL7 - ? 1 l iV:: '.s o v-c CV t - 1W ) 1Q i ' h.** i js'TQ ~ W hh I ge o% I,L... [ ! f,3 :.. 'M Q, '; _ _c d m EL2 El ELS E5 EL10 l Figure 5-5 Charpy Impact Specimen Fracture Surfaces for Hillstone Unit 3 Reacter Vessel Shell Plate B9305-1 (lengitudinal crientation) l l I j u n..c.m. 4 5-16 101-16018 i I 1 ~.- - -- -

p

q..

.J:i 7.s

st

.

*~,

$ '$';h ,f ' \\. 3 ]* L w-- w w, w. l ET1 ET9 ET10 ET2 ET4 1 'an(5kk[

\\h/

i ((![' ty.ht ' A ,N , ~ w..y l g r -] L --+ -m .j ammmmm ,m 1 1 1 w u p1 .m q ,h., i, I,, ,{ 'L -. c-g _' i ET3 ET15 ET5 ET13 ETS i i 5,4: c,=g m. E:c.. .g: 1 ~ ,,M G w ET8 ET7 ET11 ET12 l i 1 1 i Figure 5-6 Charpy lepact Specimen fracture Surfaces for Millstone t.' nit 3 l I Reactor Vessel Shell Plate B9805-1 (transverse orientatien' l r n. c w, 5-D M -It019 I j

I r l,,\\ 1 ,p; i n.. .R 'sif;,, \\ ,. _s.. 'b I = j l EW10 EY12 EY3 EW8 EW9 1 ,W \\ 1 IE '[' \\ , ~,.. s -n s

7.. - q j

qu t .h f}'} p )* = w-EWS EW1 EW13 EW4 EW11 . j I I I ju

t. a w

s n I m ammuus )E .1 ~.D' \\$ j l ), ) 5 { '.f, -" h L. 6 w-i l EW2 EW5 EV15 EW14 EW7 j i I i ~* Figure 5-7 Charpy Impact S;:ecimen Fracture Surfaces for Millstone Unit 3 j Reactor Vessel Weld Metal I v n.-ur ss,e s.1g i i e-leo:o 1 1 i .-.-..----l

l h L I s y-.,- g;

7. -- -

,,e i ) p)' .i' (p - (y,j\\ ' h1 $ yj' f f,. C I i

".3. i.

i ~ w-e- i l E311 EH13 EH7 EH6 EH1 l 1 i i 1 i .\\ I i N., { w-i I l

• '^*'*'l

_>J .m. t j l,

~ >

Nl NE~E wn ~. m-y i EH2 EHB EH4 EH15 l .l :' i h..N N' I*.'. i m e., 5 hi L 4, ss s. 4 m w-w w EH14 EH3 EH9 EH12 EH5 l t i j g Figure 5-8 Charpy Impact Specimen Fracture Surfaces of the Hillstone Unit 3 Reactor Vessel HA2 Metal III3 s -C4214410 5-19 ) 4 l 1 u-leo:1 I t I n,,- - - -, - - ,,-,,-...--,--a, ...,,-.,,n ,~,.e

1 'C - 50 0 50 100 150 200 250 300 800 120 i i i i i i I 110 - 700 f _ 100 3 90 600 w r ! 80 2A ^ N2 E s 500 - h 70 Tensile Strength m w 27 400 60 = 27 0.2 % Yield Strength 50 m I I s a s e u- % I Code: Open Points-lInfrradiated Closed Points-Irradiated 4.32 x 1018 n/cm2 h2 ' Reduc \\ ion in' Area ..L = 70 e M is i - 50 l x E 40 Total Elongation l }30 l 0 I I I I i f I I -100 0 100 200 300 400 500 600 l Temperature ( *F) ,l o i l Figure 5-9 Tensile Properties for liillstene Unit 3 Reactor Vessel Shell Plate 89805-1(longitudinalorientatien) l l l vn.-swu se 5.gg

oC - 50 0 50 100 150 200 250 300 120 i i i i i 20 110 700 100 D 7 E 500 - b 70 2 Tensile Strength m ^ 60 400 j 0.2 % Yleid Strength 50 e i i i i i i i- %0 Code: Open Points-Unirradiated Closed Points-Irrad!ated 4.32 x 1018 n/cm2 Reduction in drea - 70 s2 t y _ 60 $m x 1 i E 40 E } Total Elongation g% ,2 i m 20 Un horm Elongation ~ 10 0 I I I I I I i j -100 0 100 200 300 400 500 600 Temperature ( *F) Figure 5-10 Tensile Properties for Millstone Unit 3 Reactor Vessel shell Plate B9805-1 (transverse orientation) 3123s C421M 14 5-21

• C

- 50 0 50 100 150 200 250 300 120 1 i i i i i i 20 110 700 _ 100 m 3 90 2) T2 600 2 ^ a n. h 500 i 70 G 0.2 % Yield Strength 60 400 50 40 8 i i-0 Code: Ooen Points-Unirradiated 18 2 CIosed Points-Irradiated 4.32 x 10 n/cm 80 i i i i i i Reducilon in' Area = 2

• g x

E 40 'E 30 Total Elongation o k = 20 2 ~ Uniform Elongation 10 ^

• 2 0

I I -100 0 100 200 300 400 500 600 Temperature ( F) .L Figure 5-11 Tensile Preperties for Millstone Unit 3 Reactor Vessel Wald Hetal un. e.vis ie 5-22 j i

1 i l i i I 4 1 '!!!.!T*(fff.f.((iaM;t' 1 t wm.' . ;a.: 3- 'h tapr4 5 l '.. 4 j. 4 . 's n.% ._y&_gl 6-i%p_,.b l- ..... J E u.G'd,,. h W k'*L 1..;,l. [. - D j Specimen EL1 71*y l l i 1 l' d I l l i ! ', '. ! a l i H.i q T?,i s e,.,1 i v.- .r I L.ad h i ' l i j .C l 4 ^ i j l p-j i ,:;,;m c,y. ,..t l a j l 3,3, 2 Specimen EL2 200.) \\ \\ ~ -}, ,M' l l 1>d. 2 t 3-j m 1 1 1 t i ) 12 f Specimen EL3 550*F l t 1 ' l Figure 5-12 Fractured Tensile Specimens from the Millstone Unit 3 Reactor Vessel Shell Plate B9805-1 (longitudinal orientation)

l

} l ,, r..m... 3 23 l KM-16022 I l I 9 i

1 l 1 I l M O. ['t 04 Yf.3 i r 'l1LL { 1 y{1 .' l i ~. j 55!'.?_ ._ s l I ..,....,s i l Specimen ET1 71*F l I >,s~ I l j i l i Specimen ET2 200*F l l ' l t l. l l

• b i

t Specimen ET3 500'P Figure 5-13 Fractured '. ensile Specimens from the Millstone Unit 3 Shell Plate B9805-1 (transverse orientation) i v u..esv se,e 5-24 I F5 16023 l

i I i 1 i 'l ie.e .e g t-1-2'3*44'gi4

l

,.i.c.ris 4 e ., > : e ., e 1: i-i ' [Q:... 3, 4,Q) r 1 y h/ h 1 db ,, - I .f.' 3 t a:4 _. z _ ; 4 _'~..,i~j: j Specimen EV1 25'F l I ! i 's 'i l' &> l 4. i l

• g w.4 i

cr.o j

3 4~3e i j 1: l l f '. 'gM.3;m.w< t. g,.w f,.., i 4- ... g,4.a A , tu. A* f s L 4? n ~

. s -

- >.on%

Specimen EY2 150'F l l t J l 1 i I 4 1 Specimen EY3 550'F }- Figure 5-14 Fractured Tensile Specimens from the Hillstone Unit 3 Reactor 4 Vessel Weld Metal 3'23 -c404410 5-25 ut-16024

h l t i I i 120 i 4 IW g l 3 g l j g '3 G e i c, i 20 t i i l 0 i i i 0 0,05 0,10 0,15 0, 20 0, 25 l 1 Strain, in/in i 1 4 t i ? I I ) 1 j Figure 5-15. Typical Stress-Strain Curve for Tension Specirens I t i l l i v n.-ew u n 5-26 lt t t I

SECTION 6 RADIATION ANALYSIS /Jdo NEUTRON DOS! METRY

6.1 INTRODUCTION

Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsula geometry is required as an integral part of LWR reactor pressure vessel surveillance prograns 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 pt.sitions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is derived solely from analysis. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the ieplementation of trend curve data to assess vessel condition. In recent years, however. it has been suggested that an exposure model that accounts for differences in f.eutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associat6d with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel rall. Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data corrwlation. ASTM Standard Practice E853 "Analysis and Interpretation of Light Water Reacter Surveillance Results,' l m-av " " 6-1

recorrmonds reporting displacements per iren atom (dca) along nich fluence (E > 1.0 MeV) to provide a data base for future reference. The energy cependent dpa function to be used for this evaluatien is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." Thw application of the dpa parameter to .L the assessment of embrittlement gradients through tFe thickness of the pressure vessel wall has already been promulgated in Revisien 2 to the Regulatory Guide 1.99, '"Radiation Damage to Reactor Yessel Materials.' f L This section provides the results of the neutron dosimetry evaluations L performed in conjunction with the analysis of test specimens contained in l surveillance capsule U. Fast neutron espesure para eters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Nev), and iron atom displacements (dpa) are established for tre capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the sur veillance capsele and with the projected espesure of the pressure vessel are provided. } i '. l j 6.2 DISCRETE ORDINATES ANALYSIS A pl:n view of the reacter geenetry at the core micolane is sheen in Figure j 4-1. Six irradiation capsules attached to the neutron pads are included in [ the reactor design to constitute the reacter vessel surveillance program. The capsules are located at azimuthal angles of 58.5', 61.0*, 121.5', 238.5', 241.0', and 301.5' relative to the core cardinal area as shc n in Figure 4-1. i i I A plan view of a dual surveillance capsule holder attached to the neutron vad i is shown in Figure 6-1. The stain 19ss steel speciran containers are 1.182 by 1-inen and approximately 56 inches in heignt. The centainers are positioned l axially such that the specimens are centered on the core micplane, thus I spanning the central 5 feet of the 12-fcot high reacter core. -( l \\ f i ,i vnem'a " 6-2 1

i From a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the dist-ibution of neutron flux and the neutron energy spectrum in the water annulus between the neutron I e pad and the reactor vessel. In order to properly determine the neutron l environment at the test specimen locations, the capsules themselves must be included in the analytical model. I f In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions [ throughout the reactor geometry is well as to establish relative radial distributions of exposure parameters (d(E > 1.0 Nov ) v(E 0.1 Nev), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dot *mstry withdrawn from the i surveillance capsule as well as #cr the determination of exposure parameter ratios; i.e., dpah(E > 1.0 MeV), within the pressure vessel geometry. The relatsve radial gradient information was required to permit the projection j of measured exposure parameters to locations interior to the pressure vessel [ wall; i.e, the 1/4T,1/T2, and 3/4T locations. i l The second set of calculations consisted of a series of adjoint analysas l relating the fast neutron flus (E > 1.0 HeV) at surveillance capsule l l positions, and several azimuthal locations on the pressure vessel inner radius j l to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all I absolute exposure projections and comparison with measurement. These [ irportante functions, when corbined with cycle specific neutron source f distributiers, yielded absolute predictions of neutron exposure at the i locations of interest for the cycle 1 irradiation; and established the means i to perform similar predictions and desiretry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron scurce distributions utilized in these analyses included not only spatial variations f of fission rates within the reactor core; but, also accounted for the effects I of varying neutren yield per fiss'on and fission spectrue introduced by the 1 build-up of plutonium as the burnup of individual fuel asserblies increased, f ( i. v n, *v " " 6-3 l

The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the forward calculation provided the means to: l 1. Evaluate neutron dosimetry obtained from surveillance capsule locations. -) 2. Extrapolatu dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall. 3. Enable a direct comparison of analytical prediction with measurement, 4 Establish a mechanism for projection of pressure vessel exposure as i the design of each new fuel cycle evolves. The forward transport calculation for the reactor model summarized in Figures 4-1 and 0-1 was carried out in R. e geometry using the DOT two dimensional discrete ordinates code (4) and the SAILOR cross-section library (5). The SAILOR library is a 47 group ENDFB-IV based data set produced specifically ter light water reactor applications. In these analyses anisotepic scattering was treated with a P3 expansien of the cros3-sections and the angular I discretizatien was modelod with an Sg order of angular quadraturo. The reference core poaer distribution utilized in the forward an61ysis was derived fres statistical studies of long-term cperation of Westinghouse 4-leep a plants. Inherent in the development of this reference core peaer distribution l is the use of an out-in fuel management strategy; i.e., fresh fuel on the cere periphery. Furthermore, for the peripheral fuel assettlies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral poaer was used. Since it '.4 unlikuly that a single reactor would have a power distribution at the ncminal +2e i level for a large nurter of fuel cycles, the use of this reference distribution is expected to yield screwhat c:nservative results. i J l i o n..:wn ie 6-a I

All adjoint analyses were also carried out using an $3 order of angular quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source locations tare chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, e geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, o (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of it.terest could be calculated as: 1(r, 0, E) $ (r, 0, E) r de do dE R (r 0) = /r #e #E o (E > 1.0 MeV) at radius r and azimuthal angle o where: R(r 0) = i Adjoint importance function at radius, r, azimuthal j ! (r, e, E) = angle 0, and neutron source energy E. Neutren source strength at core location r, 0 and S (r, e, E) = energy E. Although the adjoint importance functions used in the Millstone Unit 3 analysis mere based on a response function defined by the threshold neutren j flux (E > 1.0 MeV), prior calculations have shcan that, while the implementation of low leakage leading patterns significantly ir;act the t magnitude and the spatial distribution of the neutron field, changes in the j relative neutron energy spectrum are of second order. Thus, for a given [ location the ratio of dpa/o (E > 1.0 WeV) is insensitive to changing core source distributiens, in the application of these adjoint important functions to the Millstone reactor, therefere, calculation of the iron displacecent l ratos (dpa) and the neutroi flux (E > 0.1 WeV) were computed on a cycle f specific basis by using epa /, (E > 1.0 WeV) and e (E > 0.1 NeV)/e (E > 1.0 WeV) ratios from the fo. aard analysis in conjunction with the cycle specific e (E > 1.0 WeV) soluties from the individual adjoint evaluations. f i i m... r u i. 6-5 1 l

The reactor cera power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design report for the first 7 operating cycle of Millstone Unit 3 (6). The relative power levels in fuel [ assemblies that are significant contributers to the neutron exposure of the pressure vessel and surveillance capsules are sumarized in Figure 6-2. For comparison purne,,es, the core pe=er distribution (design basis) used in t's reference forward calculation is also illustrated in Figure 6 2. SelecNd results from the neutron transport analyses performed for the Millstone Unit 3 reactor are provided in Tables 6-1 through 6 5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiatien period and provide the means to correlate dosimetry results with the corre'pending neutron exposure of the pressure vessel wall. The transport methodclogy, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oak Ridge National Labortory (CRNL) Poolside Critical Assembly (PCA) facility as mell as against the Westinghouse pcwer reactor surveillance ca:sule dats base (7). The bench-marking studies indicate that the use of $4!I.OR cross sections are generic design basis power distributions produces flux levels that tend to be conser-vative by 7 22%. When plant specific power distributions are used with tne adjoint importance functions, the benchmarking studies show that fluence predictions are within + 15% of ressured values at surveillan:e capsule locations. In Table 6-1, the calculated es;csure parareters (e (E > 1.0 WeV), e (E > 0.1 MeV), and dpa) are given at the geometric center of the two surveillance c.apsule positions f:r both the design tasis and the plant l specific core peaer destributiens. The plant specific data, based on the adjoint transpert analysis, are reant to establish tre absolute comparison of ( eessurerent with analysis. The cesign basis date derived frem the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, tFe three ;ertinent exposure vn.-ew ss,e 6-6

r parameters are listed for both the cesign basis and the cycle 1 plant specific i powerdistributions. It is important to note that the data for the vessel j ~ inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself. i = Radial gradient information for neutron flux (E > 1.0 MeV), neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, l and 6-5, respectively. The data, obtained from the forward neutron transport l calculation, are presented on a relative basis for each exposure parameter at ( several azimuthal locations. Exposuae parameter distributions within the wall may be obtained by normalizing the calculated or project >d exposure at the j vessel inner radius to the gradient data given in Tables 6-3 through 6-5. l [ For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' l animuth is given by' e(220.27, 45') F (225.75, 45') egj47(45') = Projected neutron flux at the 1/4T position where egj47(45') = on the 45' azimuth I Projected er calculated neutron flux at the { e (220.27, 45') = vessel inner radius on the 45' azimuth. [ ~ Relativi radial distribution function from Table F (225.75, 45') = 6-3. t Similar espressions apply for exposure parameters in terms of e(E > 0.1 MeV) j and dea /sec. l f The DOT calculations were carried out for a typical octant of the reactor. Hoaever, for the neutron pad arrangement in Millstone Unit 3, the pad extent I t,' for all 7ctants is not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron pad extending frem 32.5* to 45' (.12.5') which produces the maximum vessel flux. Other octants have neutron l I ~ i j r n.

• v u ie 6-7 f

l pads extending 22.5' or 20' which provide more shielding. For the octant with ' the 12.5' pad, the maximum flux to the vessel occurs near 25' and the values in the tables for the 25' angle are vessel maximum values. Exposure values for O',15', and 45' can be used for all octants; values in the tables for 25' and 30' are maximum values and only apply to octants with a 12.5' neutron pad extent. 6.3 NEUTRCN DOSINETRY The passive neutron sensors included in the Hillstone Unit 3 sarveillance pecgram are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear ccnstants t, hat a re used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest (e (E > 1.0 vev), e (E > 0.1 MeV), dpa). The relative locations of the n1utron sensors within the capsules are shown in Figure 4-2. The iron, nickel, ceppe, and cebalt-aluminum monitors, in wire form, were placed in holes drilled in space, s at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors mere acccmedated within the desireter bleck located near the center of the capsule. The use of passive e:nitors such as those listed in Table 5-6 d:es not yield a direct reasure of thi energy dependent flus level at the point of interest. Rather, the activation or fission precess is a measure of the integrated effect that the time-and energy-dependent neutren flut has on the target material over the course of the irradiation period. An accurate assessrent cf j a l the average neutren flux level incident on the various moniters may be derived ( l frem the Petivatien measurements only if the irradiation parat4ters are well j kncan. In particular, the follening variables are of interest: o The specific activity of each m: niter, j o The cperating history of the reacter. o The energy response of the moniter. i i im.-aru s 6-8

o The neutron energy spectrum at the monitor location, o The physical characteristics of the moniter. The specific activity of each of the neutron monitcrs was determined using established ASTM procedures (8 threugh 21). Follcaing sample preparation and weighing, the acti,ity of each monitor was determined by means of a litnuim-drif ted germanium, Ge(Li), gamma spectrometer. The irradiation history of the Millstone Unit 3 reactor during cycle 1 was cbtained from NUREG-0020, "Licensed Operating Reactors Status Sun ary Repert" for the applicable period. The irradiatien history applicable to capsule U is givan in Table 6-7. Measured and saturated reaction product srecific activities as well as l ressured full pc.er reaction rates are listed in Table 6-8. Reacticn rate values were derhea using the pertinent data frem Tables E 6 and 6-7. Values of key fast neutren exposure parameters were derived frem the reasured l reacticn rates using the FERRET least squares adjusteent code (22). The FERRET a; preach used the reasured reaction rate data and the calculated l neutren energy spectrum at the the center cf the surveil' lance capsule as input and proceeded to adjust the priori (calculated) grcup fluxes te produce a best i. fit (in a least squares sense) to the reaction rate data. The exposure paramaters alerg with associated uncertairties where then cetained from the adjusted s;ectre. In the FE;;ET e<sluations, a log normal least-squares algcritta neights both tre prieri vehes and the reasures data in accordance with the assigned uncer W nties and correlations. In general, the ressured values f are li lated to ve flux : by sc e response ratria A: (s.o), g g ,g( e ) (s) 19 g wFere i indenes the ressured values belonging to a single data set s, g designates the energy group and a delineates spectra that may te siNitanerusly adjuste.L. Fer exawple, y n.mse,e s.9

w R I o g=9 ogg g by relates r. set of measured reaction rates R) to a single spectrum og the multigroup nross section ogg. (In this case, FERRET also adjusts the cross-sections.) The legnormal approach automatically accounts for the physical constraint of positive fluses, even with the large assigned uncertainties. In the FERRET analysis of the dosimetry data, the continuous quantities (i.e., fluxes and crocs-sections) were approximated in 53 groups. The calculated flutes frem the discrets ordinates analysis were expanded into the FERRET group structure using the SAND-!! code (23). This procedure was carried out by first expanding the a priori spectrum into the SAND-l! 620 group structure using a SPLINE interpolation precedure for,nterpolation in regions where group boundaries do not coincide. The 620 point spectrum was then easily collapsed to the group scheme used in FERRET. The cross-sections were also collapsed into the 53 energy group strur'.ure using SAND !! with calculated spectra (as expanded to 620 groups) (.s weighting functions. The cross sections were taken from the ENOF/B-V dosimetry file. Uncertainty estimatas and 53 x 53 covariance matrices =ere constructed fer each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant. For each set of data or a priori values, the inverse of the correspeading relative covariance matria M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance estria is used. More often, a simple paramttar '. red form is used: 2+R R.P g, a R g g gg. M where R specifies an overall fractional normalizatien uncertainty (i.e., y ccolete correlation) for the corresponding set of values. The fractional v n* ** ** ' ' 6-10

uncertainties R specify additional random uncertainties for group g that g are correlated with correlation matrix: ' 6 exp [- (g- ')2] Pgg, = (1 - 0) ogg. 23 The 'irst term specifies purely random uncertainties while the second term describes short-range correlations over a r6nge r (8 specifies the strength of the latter term.) For the a priori calculated fluxe. a short-range correlation of r = 6 groups was used. This choice implies that neighbo-ing groups are strengly correlated when e is close to 1. Strong long-range correlations (or anticorrelations) were justified based en information :: resented by R. E. Maerker(24). Paerker's results are closely duplicated when i = 6. For the integral reaction rate covariances, simpla normalization and random uncertainties were ecmbined as deduced frem experimental uncertainties. l Results of the FERRET evaluation of the capsule U desirat"y are given in Table 6-9. The data su carized in Table 6-9 indicated that this capsule received an in'.egrated exposure of 4.32 x 10 n/cm2 (E : 1.0 MeV)..ith an asse-18 l ciated uncertainty of + 8%. Also reported are capsule exposures in terrs of l fluence (E > 0.1 PeV) and iren atem displacements (dpa). Suma-ies of tre fit of the adju,ted spectrum are proviced in Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrun to tre l individual experimental reaction rates. The adjusteC spectrum itself is I tabulated in Table 6-11 fer the FERRET 53 energy group structure. 1 1 A sumsry of the ressured and calculated neutron exposure of capsule V is p a sented in Table 6-12. The agree ent between calculation and measurement falls within + 102 for all exposure parameters listed. The calculated fast l neutron exposure (t (E > 1.0 MeV), 4 (E > 0.1 PeV), dpa) values were low by trem 2-9 percent relative to measurement, whereas, the thermal reutren exposure calculated for cycle 1 exceeded the measured value by 9 percent. l ~ l r n. - c" " 6-11

Neutron exposure projections at key locations on the pressure vessel inner .[ radius are given in Table 6-13. Along with the current (1.3 EFPY) exposure derived from th4 capsule U ressurements, projections are also provided for an escosur-l, period of 10 EFPY and to end of vessel design life (32 EFPY). The esNulated design basis exposure rates given in Table 6-2 were used to perform projections beyond the end of cycle 1. '} In the calculation of espesure gradients for use in the development of heatup and cooldcan curves for the Millstone Unit 3 reactor coolant system, esposure projections to 10 EFPY and 32 EFPY were employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slepe through the vessel wall are provided in Table 6-14 In cruer to access RTNOT i fluence trend curves, dpa equivalent f ast neutron fluence levels for the 1/47 and 3/4T positions were defired by the relations t' 1/4T = 4 (Surfare) ( dpa (1/4T) ) 1 ~ dot (5urface) ) dea (3'4T) I' 3/4T = 4 (Surface) { cpa(SuriTieI)

• I Using this approach results in the dpa equivalent fluence values listed in t

Table 6-14 In Table 6 15 updated lead factors are listed for each of the Millstone Unit 3 surveillance capsules. These cata esy be used as a guide in establishing l future withdrawal schecules for the remaining capsules, j I In Figure 6 3 the relative axial variatien of neutron flux and dpa within the reactor vessel is depicted. j l r n. ww e 6-12 ,,r,,

\\ (TYPICAL) 1 se,se - e1.08 A - 81.425 IN. 7 \\ I NEUTRON PAD \\ i Figure 6-1. Plan View cf a Dual Reacter Vessel Surveillance Capsule o n,-e * ' " ' 6-13

,. i 0.78 0.79 0.76 0.57 Cycle 1 1.01 1.04 0.96 0.7/ Design Basis 1.02 1.05 0.99 0.93 0.81 0.54 1.02 1.10 1.00 1.05 1.10 0.71 1.12 1.10 1.07 1.06 0.98 0.97 1.05 0.87 0.87 1.07 1.00 1.05 1.14 1.12 1.13 1.11 1.18 1.09 1.06 0.68 1.10 1.04 1.13 1.14 1.13 1.14 i 0.90 1.04 1.12 0.92 l i Figure 6-2. Core Pc or Distributions Used in Transport Calculations for Hillstone Unit 3 v n.mr se,e g.14

f l 0 10 l = 6 4 2 M D l 10*1 6 y 4 l ~ 5 \\ l E 30 2 6 f-COFE VIDPLANE / 2 - TO VESSEL ~ CLOSURE HE AD I I I I 30 3 200 200 100 0 100 200 300 DISTANCE FMOM CORE MIDPLANE (om) Figure 6-3. Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux an9 dpa Within the Reactor Vessel Wall on. unn ie 6-15

-TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER DESIGN BASIS CYCLE 1 29.0* 31.5' 29.0* 31.5' 11 11 10 10 9 (E> 1.0 MeV)(a) 1.13 a 10 1.21 x 10 8.81 x 10 9.43 x 10 l 11 11 11 11 + (E> 0.1 MeV)(a) 5.07 x 10 5.43 x 10 3.96 x 10 4.23 x 10 -10 -10 -10 -10' dpa/see 2.21 x 10 2.36 x 10 1.72 x 10 1.84 x 10 2 (a)n/cm-see I i l J I l-e i 4 u n..ew u ie 6-16

TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE DESIGN BASIS O' 15' 25' 30' 45' 10 10 10 10 10 ((E>1.0Mov)3) 1.68 x 10 2.57 x 10 3.02 x 10 2.90 x 10 2.80 x 10 10 10 10 10 10

1. (E>0.1Mev)(a) 3.49 x 10 5.41 x 10 8.25 x 10 7.92 x 10 7.01 x 10

-11 -I1 -11 -11 -11 dpa/see 2.61 x 10 3.98 x 10 5.06 x 10 4.07 x 10 4.46 x 10 CYCLE 1 SPECIFIC O' 15' 25' 30' 45' 10 10 10 10 e(E>1.0Mev)(a) 1.35 x 10O 2.05 x 10 2.37 x 10 2.29 x 10 2.22 x 10 10 10 10 10 10 f* e(E> 0.1Mev)(a) 2.81 x 10 4.31 x 10 6.47 x 10 6.24 x 10 5.56 x 10 ~11 -11 -11 -11 -11 dpa/see 2.10 x 10 3.18 x 10 3.97 x 10 3.82 x 10 3.53 x 10 2 (a) n/cm.3,e G e m"'"*"" 6-17

TABLE 6-3 RELATIVE RADIAL DICTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) - ~ WITHIN THE PRESSURE VESSEL WALL l Radius (cm) 0* 15' 25' 30' 45' 220.27(1) 1.00 1.00 1.00 1.00 1.00 220.64 0.976 0.979 0.980 0.981 0.979 221,66 0.888 0.891 0.893 0.894 0.889 222.99 0.768 0.770 0.772 0.774 0.766 224.31 0.653 0.653 0.657 0.660 0.648 225.63 0.551 0.550 0.554 0.557 0.543 226.95 0.462 0.460 0.465 0.469 0.452 228.28 0.386 0.384 0.388 0.392 0.375 229.60 0.321 0.319 0.324 0.328 0.311 230.92 0.267 0.265 0.271 0.273 0.257 232.25 0.221 0.219 0.223 0.227 0.211 233.57 0.183 0.181 0.185 0.189 0.174 234.89 0.151 0.149 0.153 0.156 0.142 236.22 0.124 0.122 0.126 0.129 0.116 237.54 0.102 0.100 0.104 0.107 0.0945 238.86 0.0828 0.0817 0.0846 0.0875 0.0762 240.19 0.0671 0.0660 0.0689 0.0715 0.0609 241.51 0.0538 0.0522 0.0550 0.0578 0.0471 242.17(2) 0.0506 0.0488 0.0518 0.0548 0.0438 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius t

u n. c.n u io s.18

\\ TABLE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 ) lev) WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0* 15' 25' 30' 45' 220.27(1) 1.00 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 1.00 221.66 1.00 1.00 1.00 1.00 0.995 222.99 0.974 0.969 0.974 0.978 0.956 224.31 0.927 0.920 0.927 0.934 0.901 225.63 0.874 0.865 0.874 0.882 0.842 226.95 0.818 0.808 0.818 0.827 0.782 228.28 0.761 0.750 0.716 0.772 0.721 i 229.60 0.705 0.693 0.704 0.716 0.662 230.9E 0.649 0.637 0.649 0.661 0.605 232.25 0.594 0.582 0.594 0.607 0.549 233.57 0.540 0.529 0.542 0.554 0.495 l 234.89 0.487 0.478 0.490 0.502 0.443 236.22 0.436 0.428 0.440 0.452 0.392 237.54 0.386 0.380 0.392 0.403 0.343 238.86 0.337 0.333 0.344 0.355 0.295 240.19 0.289 0.287 0.298 0.308 0.248 241.51 0.244 0.238 0.249 0.259 0.201 242.17(2) 0.233 0.226 0.237 0.247 0.188 l NOTES:

1) Base Metal Inner Radius l

2) Base Metal Outer Radius

) l l l vn. non to 6-19 l l

1 TABLE 6-5 i RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRLSSURE VE3SEL WALL I Radius ,1 1 (cm) O' 15* 25' 30 45' 220.27(I) 1.00 1.00 1.00 1.00 1.00 220.64 0.984 0.981 0.984 0.986 0.984 221.66 0.912 0.909 0.917 0.925 0.915 222.99 0.815 0.812 0.826 0.840 0.821 ( 224.31 0.722 0.719 0.737 0.755 0.730 225.63 0.638 0.634 0.656 0.678 0.647 226.95 0.563 0.559 0.584 0.608 0.572 228.28 0.497 0.493 0.519 0.545 0.506 ^ 229.60 0.439 0.435 0.462 0.488 0.447 l 230.92 0.387 0.383 0.410 0.436 0.394 232.25 0.341 0.338 0.364 0.389 0.347 233.57 0.300 0.297 0.322 0.347 0.305 ? 234.89 0.263 0.261 0.285 0.308 0.266 236.22 0.230 0.228 0.250 0.272 0.231 i 237.54 0.199 0.198 0.218 0.238 0.199 [ h 238,86 0.171 0.170 0.189 0.207 0.169 ( 240.19 0.145 0.144 0.161 0.178 0.140 241.51 0.121 0.119 0.135 0.150 0.113 l 242.17(2) 0.116 - 0.113 0.128 0.143 0.106 i NOTES:

1) Bale Metal Inner Radius I

2) Base Metal Outer Radius j

( I l ~I i i l + I i u n. esc u se 6-20 i

TABLE 6-6 i NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS Reaction Target Fission Monitor of Neight

Response

Product Yield Material Interest Fraction Range Half-Life (%) Copper Cu63(n,a)Co60 0.6917 E> 4.7 MeV 5.272 yrs Iron Fe54(n.p)Mn54 0.058 E> 1.0 MeV 312.2 days Nickel Ni58(n.p)CoSE 0.6827 E> 1.0 MeV 70.91 days Uranium-238* U238(n f)Cs137 1.0 E> 0.4 MeV 30.17 yrs 6.0 Neptunium-237* Np237(n,f)CsI37 1.0 E> 0.08 MeV 30.17 yrs 6.5 Cebalt-Aluminum

• CoS9(n,r)Co60 0.0015 0.4ev<E< 0.015 MeV 5.272 yrs Cebalt-Aluminum CoS9(n r)Co60 0.0015 E< 0.015 MeV 5.272 yrs o
• Denotes that nonitor is cadmium shielded.

G 9 313)o 19487H 6-21

TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U l P P Irradiation 3 3 Irradiation Decay Period (WW ) P Time (days) Time (days) t Ref. 2/86 267 .0784 28 611 3/86 1237 .363 31 580 4/86 740 .217 30 550 5/86 2858 .838 31' 519 6/86 3399 .996 30 489 7/86 2512 .766 31 458 8/86 1303 .382 31 427 9/86 3043 .892 30 397 10/86 3408 .999 31 366 11/86 3404 .998 30 336 12/86 3404 .998 31 305 1/87 3166 .928 31 274 2/87 3409 1.00 28 246 3/87 1194 .350 31 215 4/87 1943 .570 30 185 5/87 2698 .791 31 154 6/87 2566 .752 30 124 7/87 3410 1.00 31 93 8/87 3405 .998 31 62 i 9/87 3215 .943 30 32 10/87 3180 .932 31 1 g NOTE: Reference Power = 3411 MWt l m m '*"" 6-22

TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec gm) (dis /see-gm) /RPS/ NUCLEUS) Cu-63 (n,a) Co-60 4 5 Top 6.43 x 10 4.11 x 10 4 5 Middle 6.01 x 10 3.84 x 10 4 5 Bottom 5.81 x 10 3.71 x 10 ,, h 4 5 ~17 Average 6.08 x 10 3.89 x 10 5.9.~s x 10 l Fe-54(n.p) Mn-54 6 6 Top 2.35 x 10 3.86 x 10 6 6 l Middle 2.71 x 10 4.45 x 10 6 6 Bottom 2.16 x 10 3.55 x 10 6 6 -15 Average 2.41 x 10 3.95 x 10 6.30 x 10 Ni-58 (n,c) Co-58 7 7 Middle 4.82 x 10 5.46 x 10 ~ I 7 Bottom 4.87 x 10 5.52 x 10 7 7 -15 Average 4.85 x 10 5.49 x 10 7.83 x 10 U-238 (n,f) Cs-137 (Cd) 5 6 -14 Middle 1.50 x 10 5.04 x 10 3.31 x 10 1 l vn, nvu,o s.g3

TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES (Cont'd.) Measured Saturated Reaction Monitor and Activity Activity Rate Axial location (dis /sec-gm) (dis /see-am)_ (RPS/ NUCLEUS) Np-237(n,f) Cs-137 (Cd) Middle 1.47 x 10 4.94 x 10 2.98 x 10'13 0 7 L Co-59 (n,r) Co-60 7 7 Top 1.26 x 10 8.05 x 10 7 7 Middle 1.24 x 10 7.92 x 10 7 7 Bottom 1.24 x 10 7.92 x 10 7 7 -12 Average 1.25 x 10 7.96 x 10 5.20 x 10 i Co-59 (n,r) Co-60 (Cd) 6 7 Middle 6.45 x 10 4.12 x 10 6 7 Bottem 6.42 x 10 4.10 x 10 6 7 -12 Average 6.44 x 10 4.11 x 10 2.68 x 10 i l i L w t i i l l l vn us,u se 6-24

/( of TABLE 6-9 l

SUMMARY

OF NEUTRON 00SIMETRY RESULTS TIME AVERAGED EXPOSURE RATES 2 11

• (E>1.0MeV)(n/cm-sec) 1.04 x 10 r 8%

2 11 + (E> 0.1 MeV) (n/cm -sec) 4.34 x 10 15% -10 dpa/see 1.93 x'10 ggg 2 10 9 (E< 0.414 eV) (n/cm -sec) 4.14 x 10 34g INTEGRATED CAPSULE EXPOSURE 2 18 i (E> 1.0 MeV) (n/cm ) 4.32 x 10 8% 2 19 6 (F.> 0.1 MeV) (n/cm ) 1.80 x 10 115% -3 dpa 8.02 x 10 gig 2 18 I 6 (E< 0.414 eV) (n/cm ) 1.72 x 10 34, 1 i NOTE: Total Irradiation Time = 1.3 EFPY 4 O v i a # v u is 6 25

TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER t Adjusted Reaction -Measured Calculation C/M t Cu63 (n,e) Co60 5.93E-17 6.21E-17 1.05 fe54 (n.p) Mn54 6.30E-15 5.94E-15 0.94 Ni58-(n.p)CoS8 7.83E-15 7.93E-15 1.01 U238 (n f) Cs137 (Cd) 3.31E-14 3.24E-14 0.98 Np237 (n,f) Cs137 (Cd) 2.98E-13 3.10E-13 1.04 f CoS9(n,r)Co60(Cd) 2.68E-12 2.67E-12 1.00 Co59(n,r)Cc60 5.20E-12 5.21E-12 1.00 t i e a ( i i 1 1 i I j I I I i r i I i i j i t 4 j me.mme is 6-26 i 1 ,.-,--,..n.-n.-, .n---.n .----,---.n_.,

TABLE 6-11 ADJUS1ED NEUTRON ENERG'. SPECTRUN AT THE SURVEILLANCE CAPSULE CENTER Energy Adjusted Flux Energy Adjusted Flux 2 2 Group (MeV) (n/cm -see) Group (MeV) (n/cm.3,c) I 1.733E+01 8.616E+06 28 9.119E-03 2.061E+10 2 1.492E+01 1.963E+07 29 5.531E-03 2.680E+10 3 1.350E+01 7.638E+07 30 3.355E-03 8.387E+09 4 1.162E+01

1. 715E+08 31 2.839E-03 8.023E+09 i

5 1.000E+01 3.789E+08 32 2.404E-03 7.734E+09 6 8.607E+00 6.478E+08 33 2.035E-03 2.177E+10 f 7 7.408E+00 1.483E+09 34 1.234E-03 2.003Es 0 l 8 6.065E+00 2.107E+09 35 7.485E-04 1.854E+10 9 4.966E+00 4.409E+09 36 4.540E-04 1.763E+1,0 10 3.679E+00 5.740E+09 37 2.754E-04 1.893E+10 11 2.865E+00 1.179E+10 38 1.670E-04 2.024E+10 12 2.231E+00 1.587E+10 39 1.013E-04 2.032E+10 13 1.738E+00 2.187E+10 40 6.144E-05 2.009E+10 14 1.353E+00 2.407E+10 41 3.727E-05 1.953E+10 15 1.103E+00 4.335E+10 42 2.260E-05 1.885E+10 16 8.208E-01 4.897E+10 43 1.371E-05 1.821E+10 17 6.393E-01 5.046E+10 44 8.315E-06 1.721E+10 18 4.979E-01 3.648E+10 45 5.043E-06 1.567E+10 19 3.877E-01 5.129E+10 46 3.059E-06 1.448E+10 20 3.020E-01 5.294E+10 47 1.855E-06 1.320E+10 21 1.832E-01 5.275E+10 48 1.125E-06 9.661E+09 22 1.111E 01 4.250E+10 49 6.826E-07 9.513E+09 23 6.738E-02 2.977E+10 50 4.140E-07 9.890E+09 24 4.087E 02 1.702E+10 51 2.511E-07 8.055E+09 25 2.554E-02 2.250E+10 52 1,523E-07 6.738E+09 26 1.989E-02 1.116E+10 53 9.237E-08 1.672E410 l 27 1.503E-02 1.423E+10 NOTE: Tabulated energy levels represent the upper energy of each group. m.mvu ie 6-27 l l

TABLE 6-12 i COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE U Calculated Measured C/M 2 18 10 f(E> 1.0 MeV) (n/cm ) 3.92 x 10 '4.32 x 10 0.91 2 19 19 4(E> 0.1 MeV) (n/cm ) 1.76 x 10 1.80 x 10 0.98 dpa 7.65 x 10 8.02 x 10'3 0.95 -3 t 2 18 18 f(E< 0.414 eV) (n/cm ) 1.87 x 10 1.72 x 10 1.09 4 i. F [ l i r i d i i r i. i i a l l j l sin. own ie 6-28 i V

L, TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE AZIMUTHAL AN3LE 0* 15' 25.(b) 30' 45' 1.3 EFPY 17 17 18 18 18 4(E> 1.0 MeV)(a) 6.16 x 10 9.36 x 10 1.08 x 10 1.04 x 10 1.01.t 10 18 18 18 18 18 6(E> 0.1 MeV)(a) 1.28 x 10 1.97 x 10 2.95 x 10 2.84 x 10 2.53 x 10 dpa 9.56 x 10'd 1.45 x 10'3 1.81 x 10 1.74 x 10 1.61 x 10 ~3 ~3 -3 10.0 EFPY 18 18 19 18 18 4(E> 1.0 MeV)(a) 5.22 x 10 7.98 x 10 9.35 x 10 8.98 x 10 8.68 x 10 19 19 19 19 19 6(E> 0.1 MeV)(a) 1.09 x 10 1,68 x 10 2.55 x 10 ?.45 x 10 2.17 x 10 ~3 1.24 x 10-2 1.56 x 10 1.50 x 10 1.38 x 10 -2 -2 -2 dpa 8.10 x 10 32.0 EFPY 19 19 19 19 18 f(E> 1.0 Me'l)(a) 1.69 x 10 2.58 x 10 3.03 x 10 2.91 x 10 2.81 x 10 19 19 19 19 19 ~ f(E> 0.1 MeV)(a) 3.52 x 10 5.43 x 10 8.27 x 10 7.95 x 10 7.04 x 10 -2 -2 -2 -2 -2 dpa 2.62 x 10 3.99 x 10 5.07 x 10 4.87 x 10 4.47 x 10 2 (a) n/cm (b) Maximum point en the pressure vessel I e o n.*vu to 6-29

TABLE 6-14 NEUIRON EXPOSURE VALUES FOR USE IN Tile GENERATION OF liEATUP/C00LDOWN CURVES 10 EfPY NEUIRON FLUENCE (E> 1.0 MeV) SLOPE dpa SLOPE (n/cm ) ,jc,2 2 Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 18 18 I7 18 18 18 0* 5.22 x 10 2.83 x 10 6.06 x 10 5.22 x 10 3.29 x 10 1.14 x 10 I9 18 17 18 18 18 I5* 7.98 x 10 4.33 x 10 9.10 x 10 7.98 x 10 5.00 x 10 1.73 x 10 18 18 18 18 18 18 25*(a) 9.35 x 10 5.13 x 10 1.13 x 10 9.35 x 10 6.28 x 10 2.43 x 10 18 18 I9 18 18 I9 m 30* 8.98 x 10 4.93 x 10 1.09 x 10 8.98.x 01 6.03 x 10 2.33 x 10 b 45' 8.68 x 10 4.64 x 10 9.37 x 10 8.68 x 10 5.56 x 10 1.90 x 10 18 18 I7 18 18 I9 32 EFPY NEUTRON FLUENCE (E> 1.0 MeV) SLOTE dpa SLOPE 2 2 (n/cm ) n/cm Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 19 18 18 I9 I9 18 0* 1.69 x 10 9.16 x 10 1.96 x 10 1.69 x 10 1.07 x 10 3 69 x 10 19 19 I9 19 I9 18 IS' 2.58 x 10 1.40 x 10 2.94 x 10 2.58 x 10 1.62 x 10 5.59 x 10 19 19 18 I9 I9 18 25*(a) 3.03 x 10 1.66 x 10 3.66 x 10 3.03 x 10 2.04 x 10 7.87 x 10 I9 19 18 I9 19 18 30* 2.91 x 10 1.60 x 10 3.53 x 10 2.91 x 01 1.95 x 10 7.55 x 10 I9 I9 18 I9 19 18 45* 2.81 x 10 1.50 x 10 3.03 x 10 2.81 x 10 1.80 x 10 6.15 x 10 (a) Maximum point on the pressure vessel mmme se

TABLE 6-15 UPDATED LEAD FACTORS FOR MILLSTONE UNIT 3 SURVEILLANCE CAPSULES Capsule lead Factor U 3.98(a) X 4.01 W 4.01 2 4.01 V 3.74 Y 3.74 (a) Plant specific evaluation l. O e e h e e ma*me se 6-31

SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTN E185-82 and is recommended for future capsules to be removed from the Millston Unit 3 reactor vessel: Estimated Capsule t Vessel Removal Fluence Location Lead 2 Capsule (deg) Factor Time (a) (n/cm ) 18(b) U 58.5 3.98 1.3 4.32 x 10 19(c) Y 241 3.74 9 3.18 x 10 19 V 61 3.74 16 5.67 x 10 W 121.5 4.01 Standby X 238.5 4.01 Standby Z 301.5 4.01 Standby r a) Effective full power years from plant startup b) Actual fluence c) Approximate fluence at vessel inner wall at end of life i 1 [ t n.o.muu 71 y ._._.-y y. y+.--.--

SECTION 8 s REFERENCES

1. Singer, L. R., "Northeast Utilities Service Company Millstone Unit No. 3 Reactor Vessel Radiation Surveillance Program, WCAP-10732, June 1985.

2. Code of Federal Regulations, 10CFR50, Appendix G. "Fracture Toughnwss Requirements" and Appendix H "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission Washington, DC.

3. Regulatory Guide 1.99, Proposed Revision 2, "Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission. February, 1986.

,4. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, "Nuclear l Rocket Shielding Methods, Modification, Updating and Input Data i Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique", W/.NL-PR(LL)-034 Vol. 5, August 1970.

5. "0RNL RSCI Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, C oss betic' Library for Light Water Reactors".

6. J. V. Alexander, et. al., "The Nuclear Design and Core Physics Characteristics of the Millstone Generating Station Unit 3 Cycle 1",

WCAP-10791, Noveder 1985. (Proprietary)

7. S. H. Anderson and K. C. Tran, "Benche. ark Testing of Westinghouse Neutron Transport Analysis Methodology, PCA Evaluations" to be published.

8. ASTM Designation E482-82, "Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Sectier 12, American Society for Testing and Materials, Philadelphia, PA, a

1984. on.*""" 8-1

9. A3TM Designation E560-77, "Standard Recommanded Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Svetion 12, American Society for Testing and Materials, Philadelphia, PA,1984.

t

10. ASTM Designation E693-79, "Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacemnts per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

11. ASTM Designation E706-81a, "Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards, Section 12, Amrican Society for Testing and Materials, Philadelphia, PA,1984.

12. ASTM Designation E853-84, "Standard Pra:tice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, Anerican 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 Stendards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

14. ASTM Designation E262-77, "Standard Method for Measuring Tnermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984,

15. ASTM Designation E263 d2, "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 5264-82, "Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.

vn. w ou g.g

r

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 fer Testing and Materials, Philadelphia, PA,1984.

19. ASTM Designation E704 84, "Standard Method for Measuring Reaction Rates by Radioactivation of Uraniur 238", in ASTM Standards, Section 12, American Society for Testing and 'iaterials, 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 I

12, American Society for Testing and Materials, Philadelphia, PA,1984. r

21. ASTM Designation E1005-84, "Standard Method for Application and Analysis of Radiemetric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, I

Philadelphia, PA,1984, [ 2 r i

22. F. A. Schmittroth, FERRET _0ata_ Analysis _ Core, HEOL-TME 79-40, Hanford i

Engineering Development Laboratory, Richland, WA, September 1979. 4 l

23. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative l

Method of Neutron Flux Spectra Determined by Foil Activation. AFWL-TR 67-41, Vol. ! IV, Air Force Weapons Laboratory, K'.rkland AFB, NM, i j July 1967. L

24. EPRI-NP-2188, "Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al.,1981.

<e i vn,sevos 83

APPENDIX A s E HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION D

1.0 INTRODUCTION

Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vess91 is determined by using the preservice reactor vessel material fracture tough-RT IS ness properties and estimating the radiation-indu.ed ARTNDT, NDT designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) er the temperature at which the material exhibits at least 50 f t-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F. RT 1.1 creases as the material is exposed to fast-reutron radiation. NDT Therefore, to find the most limiting RTNDT at any time period in the reactor's life, 6RT due to the radiation exposure associated with that NDT The e.. tent of time period must be added to the original unirradiated RTNDT. the shift in RT is enhanced by cartain chemical elements (such as copper, a NDT nickel and phosphorus) present in reactor v6ssel steels. Westinghouse, other NSSS vendors, the U.S. Nuclear Regulatory Commission and others have developed trend curves for predicting adjustment of RTNDT as a function of fluence and copper, nickel and/or phosphorus content. The Nuclear Regulatory Commission (NRC) trend curve is published in Regulatory Guide 1.99 (Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials)(AQ Regulatory Guide 1.99 was originally published in July 1975 with a Revision 1 being issued in April 1977. Currently, a Revision 2(A-2) to Regulatory Guide 1.99 is under consideration within the NRC. The chemistry factor, "CF" ('F), a function of copper and nickel content identified in Regulatory Guide 1.99, Revision 2 is given in table A-! for welds and table A-!! for base metals (plates and forgings). Interpolation is permitted. The value, "f". vn, masse n g.1 l

[ given in figure A-1 is the calculated value of the neutron fluence at the location of interest (inner surface, 1/4T, or 3/4T) in the vessel at the location of the postulated defect, n/cm2 (E > 1 MeV) divided by 1019 The fluence factor is determined from figure A-1. ,[ 2.0 FRACTURE TOUGHNESS PROPERTIES 's The preirradiation fracture-toughness properties of the Millstone Unit 3 [ reactor vessel materials are presented in table A-!!!. The fracture-toughness { properties of the ferritic material in the reactor coolaat pressure boundary I are determined in accordance with the NRC Regulatory Standard Review P l ar.[ A-3) i I 3.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS i l The ASME approach for calculating the allowable limit curves for various f f heatup and cooldown rates specifies that the total stress intensity factor, K,. for the ecmbined thermal and pressure stresses at any time during heatup .j i or cooldown cannot be greater than the reference stress intensity factor, i Kgg, for the metal temperature at that time. K is cbtained from the gg I i reference fracture toughness curve, defined in Appendix G to the ASME Code (A-4), l The KIR curve is given by the following equation: t Kgg = 26.78 + 1.223 exp (0.0145 (T-RTNOT + 160)) (1) i i t a where I K a reference stress intensity factor as a function of the retal gg te perature T and the metal reference nil-ductility temperature RTNOT I i Therefore, the governing equation for the heatup-cooldown analysis is defined j IA'43 as follows: l in appendix G of the ASHE Code j .5 { (2) CKgg + Kli

• E!R r

i t i I l ] A-2 i

where e Kg = stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RTNOT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not.r';ical At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed. From equation 2, the pressure t 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 reference flaw of Appendix G to :he ASME Code is assumed to exist at the inside of the vessel vall. During cooloawn, 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 co,esite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measuremen't of reactor coolant temperatura, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. sm.w is A.3 L

Du.ing cooldown, the 1/4 T vestal 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 ecoldown results in a higher va'ue of K;g at the 1/4 T location for finite cooldown rates than for steady-tate } cperation. Furthermore, if conditions exist so that the increase it X;g exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value. Tt;e 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 ensures conservative cperation of the system for the entire cooldown period. Three separate calculations are required to determine the limit c.'ves for finite heatup rates. As is done in the cooldown analvsis, 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 tne wall that alleviate the tensile stresses produced by internal p essure. The metal temperature at the crack tip lags the coolant for the 1/4 T crack during heatup is loaer terperature; therefore, the KIR than the K for the 1/4 T crack during steady-state conditions at the same IR titre coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of ct.mpressive thermal stresses and lower Xgg's do not offset each other, and the 3ressure-temperature curve bued on steady-state conditions no longer represents a lcaer boand of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be ahelyzed in order to ensure that at any coolant temperature the lower value of the ;%able pressure calculated for steady-state and finite heatup rates is obtained. The second portier, of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, m. m.w ie 44

I f the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the [ rate of 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 analynd on an individual basis. t Following the ger. oration of pressure-temperature curves fer both the steady-I state and finite heatup rate situat ons, the final limit curves are produced by constructing a composite cerve based on a point-by' point comparison of the steady-state and finite heatup rate data. At any given temperature, the i allowable pressure is taken to be the lesser of the three values taken from t 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 heacup 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.

Then, i

composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on figures A-2 and A-3. i e l Finally, the 1983 A:rendment to 10CFR50(A-5) has a rule which addresses the I metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material RT by at least 120'F for normal operation when the NDT pressure exceeds 20 percent of the preservice hydrostatic test pressure. l Table A-3 indicates that the limiting RTNDT of 30'F occurs in the head flange of Millstone Unit 3. so the minimum alloaable temperaturc of this i region is 150'F at pressures greater than 621 psig. These limiss are less restrictive than the limits shown on figure A-2. 4.0 HEATUP AND C00LO(AN LIMIT CURVES i limit curves for normal heatup and cooldown of the primary Reactor Coolant ( System have been calculated using the methods discussed in t.ectica 3. on.m.u ie 45 l -,..-----e.,, - -. - _ - - - - - - -,,,-,,-. -, - -_ -.,,,-, --,-.. -.

Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been cbtained directly from the reactor pressure vessel surveillance prcgram. (f Allowable combinations of temperature and pressure for specific temperature 6 change rates are below and to the right of the limit lines shown in figure A-2. This is in addition to other criterb which must be met before the .[ reactor is made critical. l The leak limit curve shown in figure A-2 represents minimum temperature IA*3' A require ents at the leak test pressure specified by applicable codes The leak test limit curve was determined by methods of references A-3 and A-5. j i Figres A 2 and A-3 define limits for ensuring prevention of nonductile f failure. t I 5.0 ADJUSTED REFERENCE TEMPERATURE i 4 + l From Regulatory Guide 1.99 Rev. 2 the adjusted reference temperature (ART) for i j aach material in the beltline is given by the following expressicn: ART = Initial RTNOT * 'SINDT + Margin (3) I Initial RT is the reference temperature for the unieradiated mat 6 rial as i l NOT defined in paragraph NB-2331 of Section !!! of the ASME Boller and Pressure i for the material in [ Vessel Cede. If reasured values of initial RTNOT l question are not available, generic mean values for that class of material may i L i be used if there are sufficient test results to establish a mean and standard j l deviatica for the class. t l ART is the mean value of the adjustment in reference temperature caused j NOT l i by irradiation and should be calculated as follows: l t 1 (4) g7 sur face = (CF]f(0.2G 0.10 log f) ARi l l l i vn.swwe g.g j

To calculate ARTNDT at any depth (e.g., t.t 1/4T or 3/4T), the following l attenuation formula was used: N ARTNOT = (ARTNDT surface]e where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface. CF (*F) is the chemistry factor, a function of copper and nickel content. CF 1s given in table A-1 for welds and in table A-!! for base untale (plates and forgings). 1.inear interpolation is permitted. In table A-! and A-!! "weight-percent copper" and "weight percent nickel" are the best estimate values for f the material, which will norm',ily be the mean of the reasured values for a j p.' ate or forging or for weld samples made with the weld wire heat numbar that i matches the critical vessel weld. At the vessel inside radius, the calculated neutron fluences for 10 effective I 18 2 full power years (EFPY) is 9.35 x 10 n/cm at the 25' azimuth (where ) *, there is no vertical weld seams). Fluence factor at this locatiran i* O.9812. [ l Applying Regulatory Guide 1.99 Revision 2 procedures to all the beltline f,', region materials, and using tables A-1 and A-!!, it was found that the intermediate shell plate B9805-1 is the limiting vessel material.

J.

For the limiting vessel material, the chemistry factor is 31, based on table i [ A-!!. From equation (4), the ARTNDT at the inner surface is equal to j 30.42'F. Regulatory Guide 1.99 revision 2 provides a formula and rules for esta511shing margin: i Margin =2/o 2 ,2 (6) g i i Since the initial RTNOT value fu. the vessel beltline plate is m asured, the temperature o is taken as O'F. The standard deviation o, is taken !f as 17'F(A 2) g As a result, the margin is 2/0 + 173 = 34' *. Substituting the obtained values for LRTKOT surface and margin, into a t I { si >. w *e4it A.7

- v. g, }>' l Equation (3) gives the adjustad reference temperature (ART) for the inside I-surface: l ART = 60 + 30.42 + 34 a 124.42*F (at 10 EFPY) Using the vessel thickness of 8.625 inches at the beltline. Equation (5) is used to calculate the ART at the 1/4 and 3/4 thicknees locations. These are b 120.32'F ano 113.72*F respectively. The above anal,ysis was used to develop the Millstone Unit 3 heatup and cooldown curves shown in figures A-2.ed A-3, respectively. 1 o l m a w.ie A8

a e Y- = ~- y d e o e A M ce W C3 2: W M .t 4 e e< g a LO o e -e S e l ua e e as A W 1 E o 5 ~ 5 c ss e O N M t D 4 L ^ o Lo "2 a e$ Uc 6 e o h u c e o3 e w o Lo2 .e" to i i l e ? I I ,9

• t 9 N M 81 et

[I ce e g g Seg gg gg gl'MN#$ OW 2 0e A-9

NATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERME01 ATE SHELL PLATE 89805-1 COPPER CONTENT: .05 WT % NICKEL CONTENT: .61 WT % 60'F INITIAL RTNOT: s RT AFTER 10 EFPY: 1/4T, 120'F NOT 3/4T, 114'F 2500 iiiii! i i i i i i i !l i I I r t.eak Test Limit ,1

1

,I 2250 i is r F l } } I, I I I I I 2000 l4 r r 1 1 1, I l-j .I I ) 1750 Unacceptable / / Operation i i Acceptable l l Operation A I I i G 1500 i 1 I / b r i I I u 1250 r i h -Heat Up Rates / [ / f C Up to 60 F/Hr /; [ l E 1000 i 1 t/ / i 0 ( l / y-i i u 750 t Criticality Limit Based 9 / on Inservice Hydro-statfcTestTemperature /' 500 ~/ (265 F) for the Service Period Up to 10 EFPY 250 i 0 50 100 150 200 250 300 350 400 -450 500 0 INDICATED TEWPERATURE (OCG r) Figure A-2. Millstone Unit 3 Reactor Coolant System Heatup Limitations Applicable for the First 10 EFPY vn.sen i. A-10

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE B9805-1 COPPER CONTENT: 0.05 WT % i NICKEL CONTENT: 0 61 WT % INITIAL RTNDT: 60 4 RT AFTER 10 EFPY: 1/4T, 120'F NDT 3/4T, 114*F ~ ~ ~ ~ 2500 i i i 1 2250 1 I T I I 2000 f 1 Unacceptable - f Ooeration f i'i II e. E 1500 '[ I G s l g 1250 / D I f I. Acceptable i e 'l / Operation l { 10'JO c f I a [' h ~~~ Coo 1Do.n mates j " 750 O ~,.,~ *IIHR p ff ~ ,,a bvs,/n G tv gy1 500 su -s v 60--!7i ~~~ 1WHFi i 250 i; 1 1 0 0 50 100 150 200 250 300 350 400 450 500 INDICATCO TEWPERATURE (CCC.F) Figure A-3. Millstone Unit 3 Reactor Coolant System Cooldown Limitations Applicable for the First 10 EFPY l I l mawma 4 3g l L---___----------

Table A-1 CHEMISTRY FACTOR FOR WELOS, 'F

Cepper, Nickel, Wt-%

-{ Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20 O.01 20 20 20 20 20 20 20 0.02 21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 14 43 54 54 54 54 54 4 0.05 26 49 67 68 68 68 68 0.06 29 52 77 82 82 82 82 0.07 32 55 85 95 95 95 95 i 0.08 36 58 90 106 108 108 103 0.09 40 61 94 115 122 122 122 e 0.10 44 65 97 12~2 133 135 135 0.11 49 68 101 130 144 148 148 0.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 0.15 70 88 115 149 178 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 O.20 88 104 129 160 194 223 245 0.21 92 108 133 164 197 229 252 5 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 vnauws g.12

Table A-I (Cont'd.) CHEMISTRY FACTOR FOR WELOS, 'F

Copper, Nickel, Wt-%

Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20 0.25 110 126 148 176 209 243 272 0.26 113 130 151 180 212 246 276 0.27 119 134 155 184 216 249 280 0.28 122 138 160 187 218 251 284 0.29 128 142 164 191 222 254 287 0.30 131 146 167 194 225 257 290 0.31 136 151 172 198 228 260 293 0.32 140 155 175 202 ?31 263 296 0.33 144 160 180 205 234 266 299 0.34 149 164 184 209 238 269 302 0 35 153 168 187 212 241 272 305 0.36 158 172 191 216 245 275 308 0 0.37 162 177 196 220 248 278 311 0.38 166 182 200 223 250 281 314 C.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 on.mme A-13

Table A-!! CHEMISTRY FACTOR FOR BASE METAL, 'F

Copper, Nickel, Wt-%

Wt-% 0 0.20 0.40 0.50 0.80 1.00 1.20 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 O 02 20 20 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31 31 31 'f1 31 31 0.06 28 37 37 37 3i' 37 37 0.07 31 43 44 44 44 44 44 0.08 34 48 51 51 51 51 51 0.09 37 53 58 58 58 58 58 0.10 41 58 65 65 67 67 67 O.11 45 62 72 74 77 77 77 0.12 49 67 79 83 86 86 86 0.13 53 71 85 91 96 96 96 i 0.14 57 75 91 100 105 106 106 l 0.15 61 80 99 110 115 117 117 0.16 65 84 104 118 123 125 125 ) 0.17 69 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 0.19 78 97 120 142 150 154 154 0.20 82 102 125 149 159 164 165 0.21 86 107 129 155 167 172 174 0.22 91 112 134 161 176 181 184 0.23

5 117 138 167 184 190 194 O.24 100 121 143 172 191 199 204 o n. tow.

A.14

l ~ Table A-!! (Cont'd.) l 1 CHEMISTRY FACTOR FOR BASE METAL, 'F l e i

Copper, Nickel. Wt-%

Wt-% 0 0.20 0.40 0.50 0.80 1.00 1.20 0.25 104 126 148 176 199 208 214 0.26 109 130 151 180 205 216 221 l 0.27 114 134 155 184 211 225 230 l l 0.28 119 138 160 187 216 233 239 l 0.29 124 142 164 191 221 241 24P f 0.30 129 146 167 194 225 249 257 0.31 134 151 172 198 22d 255 266 l 0.32 139 155 175 202 231 260 274 0.33 144 160 180 205 234 264 282 0.34 149 164 184 209 238 268 290 l o l 0'.35 153 168 187 212 241 272 298 [ l-0.36 158 173 191 216 245 275 303 0.37 162 177 196 220 248 278 308 f 0.38 166 182 200 223 250 281 313 l 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 l 7 9 4 l i { o n. g.15

VMI JM 1 TABLE A-3 REACIOR VESSEL FRACTU?E 100GitNESS PROPERTIES l Avg. Upper Shelf Cu P Ni T RI WD NDI NDT COMPONENT CODE NO. GRADE M M M (*f) (*f) (ft-lb) {ft-lb) Closure Head Dome B9812-1 A5338, CL.1 0.05 0.010 0.67 -40 0 96.0 Closure llead Torus B9813-1 A5338, Ct.1 0.11 0.010 0.62 -40 10 107.5 Closure IIcad flange B9803-1 A508 Ct.2 0.010 0.72 30 30 121.0 Vessel Flange B9801-1 A508. CL.2 0.15 0.007 0.83 -40 -40 116.5 Inlet Nozzle B9806-3 A508 CL.2 0.09 0.009 0.83 10 10 162.0 ,y inlet Nozzle B9806-4 A508. CL.2 0.09 0.009 0.82 0 0 158.0 Inlet Nozzle RS-3 A508 CL.2 0.07 0.008 0.80 -10 -10 130.0 Inlet Nozzle RS-4 A508, CL.2 0.08 0.009 0.81 0 0 136.0 0.012 0.71 -40 -40 128.0 Outlet Nozzle R6-1 ASCS, CL.2 0.ilet Nozzle R6-2 A508. CL.2 0.006 0.70 -30 -30 127.0 Outeat Nozzle 'AS07-1 A508. CL.2 0.006 0.66 -30 -30 121.0 Outlet Nozzle 898C7-2 ASC8, CL.2 0.005 0.64 -30 -30 126.0 Hozzle Shell B9Sa4-1 A5338, CL.1 0.05 0.012 c.62 -40 40 85.5 l Nozzle Shell B9804-2 A5338, CL.1 0.08 0.010 0.64 -40 20 104.5 Nozzve Shell B9804-3 A5333, CL.1 0.05 0.009 0.65 -50 0 103.5 Inter. Shell 89805-1 A5338 CL.1 0.05 0.010 0.61 -40 60 92.5 113.5 Inter. Shell B9805-2 A5338, CL.1 0.05 0.014 0.62 -60 10 90.0 129.0 Inter. Shell B9805-3 A5338 CL.1 0.04 0.009 0.62 -40 0 106.5 136.5 g > %.u m... 5., t m =-

~.

• o.

0-a 4 TABLE A-3 (Cont.) j REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES 4 Avg. Upper Shelf Ce P Ni i Ri seNO 88G ng gg COMPONENI CODE NO. GRADE 1%1 1%1 11 (*F1 (*F) (ft-lb) (ft-Ib) i ? Lower h il 89820-1 A5338, Ct.1 0.07 0.006 0.62 -50 10 76.5 124.5 Lower W 11 89820-2 A5338. Ct.1 0.06 0.006 0.60 -30 40 75.5 114.5 Lower Shell B M20-3 AS338. CL.1 0.05 0.001 0.58 -30 20 79.5 124.0 Bottom Head Torus 89816-1 A5338. CL.1 0.13 0.012 0.65 -50 -40 91.5 J Botton Head Dome 89817-1 A5338, CL.1 0.15 0.012 0.66 -30 -30 161.0 b Inter & Lower Shell G1.59 SAW 0.07 0.011 0.03 -50 -50 200.0 Long & Girth weld Seams l Notes: NMWD = normal to major working direction 1810 = major working direction SAW = weld wire heat 4P6052 and Linde 0091 flun. 1st 0145 i I j l l

REFERENCES 'e A-1. Regulatory Guide 1.99, Revision 1, "Effects of Residual Elements on Predi:ted Radiation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Comission, April 1977. A 2. "Preposed Revision 2 to Regulatory Guido 1.99, Radiation Damage to Reacter Vessel Materials," U.S. Nuclear Regulatory Comission, February, 1986. A-3. "Fracture Toughness Requirements " Branch Technical Position uTEB u-2, Chapter 5.3.2 in Standard Review Plan fcr the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1931. A-4. ASVE Boiler and Pressure Vessel Code, Section 111, Division 1 - Appendixes, "Rules for Construction of Nuclear Vessels, Appendix G, Protection Against Ncnt.ctile Failure," pp. 559-564, 1983 Editien, A erican Society of Nehanical Engineers, New York,1983. j,. A 5. Ccde of Federal Ragulations,10CFR50, Appendix G. "Fracture Teughness Require?ents," U.S. Nuclear Regulatory Comission Washington, D.C., Pended May 17, 1983 (48 Federal Register 24010). s v n,muis i. A 18}}