ML20043H033

Proposed Tech Specs Re RCS Heatup & Cooldown Limitations Applicable to First 10 EFPYs

ML20043H033
Person / Time
Site:	Beaver Valley
Issue date:	06/11/1990
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20043H030	List: ML20043H029 ML20043H032 ML20043H033
References
NUDOCS 9006210475
Download: ML20043H033 (10)

Text

{{#Wiki_filter:' ATTACHMENT A Revise the Beaver Valley Unit No. 2 Technical Specifications as follows: Remove Paaes Insert Paaes 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 l l i f kffCU05000A3;g 0 90061g FR p FDC l -.. 1

? MATERIAL PROPERTY BASIS 1 CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE 89004 1 INITIAL RTNDT: 60'F RT AFTER 10 EFPY: 1/4T, 140'F NDT 3/4T. 129'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. i '. 41 I ri L A I 8 ! t I. I i l i e + e ! il-1 I I' I I 1 2250 ~"f" Leek Test i i i ri 11 i 1 Limit !J ! It Ii If Ji i !f I! I Ii Ii !t i, i i r dOOO ,I t f i. i I 'i I J _g i 1 im i i !! '.1; I r r. J J 3 6 I I i 1 I I ' i t ' Heatu, nates f / G 150e .Qt 10 r ( SU'F/Hr I 1 i w 1250 / / l i :

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i! i 1 ., i c. 1000 7 4 i o w J i i i 3 !af 750 [ i g i e ,,i i l ..j / Criticality Limit i $00 Hydrostatic Test Based on Inservice i - Temp.(260*F)for Acceptable the Service Peri-250 Operation up to 10 EFPY I i i i 4 I I I f i1 1 1 1 0 50 100 150 200 2 50 300 350 400 450 500 INDICATCO TCWPERATURC (DCC.r) FIGURE 3.4 2 BEAVER VALLEY UNIT 2 REACTOR COOLANT SYSTEM HEATUP l LIMITATIONS APPLICABLE FOR THE FIRST 10 EFPY BEAVER VALLEY - UNIT 2 3/4 4 31 hk0f02D

i WATERIAL PROPERTY 8A$l$ 1 CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE B9004 1 INITIAL RTNOT: 60'F RT AFTER 10 EFPY: 1/4T, 140'F NDT 3/4T, 129'F CURVES APPLICABLE FOR C00LOOWN RATES UP T0100 *F/NR FOR THE SERVICE PERIOD UP TO 10 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POS$1BLE INSTRUMENT 1 ERRORS. 2500 j unn. i i i i i 2250 l l i 1 i i j i i r i i, L j l ll i 2000-m i r I i J 4 i i R 1750 l f l i A l F i i 1 1 J i i i 1 i 1250 Unceeptable r Operation / / Acceptable 1000 j' Operation i 0 7so . cos.im i _. _ + n' '.., i t F/Nr m s i aP d mA F J l 0 . -s .o 500-to ,-_m r g m. 80 l 250 l l i ,0 50 100 140 200 2h0 300 3h0 400 450 500 IN0lCAf tD f tMPERatutt (DEG.F') FIGURE 3.4 3 BEAVER VALLEY UNIT NO. 2 REACTOR COOLANT SYSTEM C00LDOWN LINITATIONS APPLICABLE FOR THE FIR $7 10 EFPY l L BEAVER VALLEY - UNIT 2 3/4 4-32 (NEXT PAE !$ 3/4 4 34) PleMsEb I i_.._,..____...,-.--

I ATTACHMENT B Beaver " alley Power Station, Unit No. 2 i Proposed Te:hnical Specification Change No. 40 REVISION C F TECHNICAL SPECIFICATION 3.4.9.1 HEATUP AND COOLDOWN CURVES AND BASES A. DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would modify the heatup and cooldown limit

curves, Figures 3.4-2 and 3.4-3, to incorporate the curves applicable to 10 Effective Full Power Years (EFPY) provided in WCAP-12406,

" Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit 2 Rehetor Vessel Radiation Surveillance Program." In our response to Generic Letter 88-11 it was concluded that to inplement the requirements of Regulatory. Guide 1.99, Revision 2 " Radiation Embrittlement of Reactor Vessel Materials", the applicability date for the heatup and cooldown curves would be changed to 5 EFPY.

However, the next surveillance capsule is due for removal at 6 EFPY and WCAP-12406 using the methods of Regulatory Guide 1.99 Revision 2 provides heatup and cooldown curves applicable to 10 EFPY, therefore, it is prudent to incorporate the 10 EFPY curves.

B. BACKGROUND The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The Beaver Valley Unit 2 reactor pressure vessel beltline material consists of a low alloy ferritic steel which is affected by an increase in hardness and tensile properties and a decrease in ductibility and toughness during fast neutron irradiation. Appendix G to Section III of t'ae ASME Boiler and Pressure Vessel Code provides a method for performing analyses to guard against fast fracture in pressure vessels. The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT). RTHDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT) or the temperature 60F less than the 50 ft-lb and 35-mil lateral expansion temperature as determined from charpy specimens oriented normal to the major working direction of the material. The RTNDT is used to index the material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME code. The KIR curve is a lower bound of

dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel.

When the material is indexed to the KIR

curve, allowable stress intensity factors can be obtained for this material as a function of temperature.

Allowable operating l limits can then be determined using these allowable stress intensity factors. RTNDT and the operating limits can then be adjusted to account for the effects of radiation on the reactor vessel material properties. Surveillance capsule U was removed from the Beaver Valley Unit 2 reactor vessel after 1.24 EFPY and the encapsulated specimens were tested. The increase in the average charpy V-notch 30 ft-lb temperature due to irradiation is L added to the original RTNDT to adjust the RTNDT for radiation embrittlement.

i PrrpOccd Tcchnical Sp;cificcticn ChOng3 No. 40 Page 2 l This adjusted RTNDT is used to index the material to the KIR J curve and to set operating limits which reflect the effects of irradiation on the reactor vessel materials. C. JUSTIFICATION The reactor vessel material surveillance program complies with 10 CFR 50 Appendix G and Appendix H to ensure the reactor vessel i has an adequate margin of safety with respect to material toughness throughout the service life of the plant. I Specifically, the program develops operating limits (RCS heatup and cooldown limit curves) to prevent non-ductile failure. The heatup and cooldown operating curves have been adjusted in' accordance with the NRC approved methodology of Regulatory Guide 1.99, Revision 2 to account for the cumulative effects of radiation on the reactor vessel material properties and to maintain an adequate margin of safety. l D. SAFETY ANALYSIS l Adjusted reference temperatures have been calculated using the material property and neutron fluence data to determine the most limiting reactor vessel materials. Plate B9004-1 was found to be the most limiting material in the reactor vessel based on these calculations relative to the generation of heatup and cooldown cu rves. The reference flaw of Appendix G to the ASME code is assumed to exist at the inside of the vessel wall for calculating the allowable pressure versus coolant temperature during cooldown. The controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From there relations, composite limit curves are constructed for each cooldown rate of interest. As was done in the cooldown

analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T (vessel wall thickness) defect at the inside of the wall that alleviates the tensile stresses produced by internal pressure.

The heatup analysis also concerns the calculation of pressure-temperature limitations assuming a 1/4 T deep outside surface flaw. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rates, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. The composite curve is necessary to provide conservative heatup limitations because it is possible for cond!.tions to exist wherein, over the course of the heatup

ramp, the controlling condition switches trom the inside to the outside, and the pressure limit must at all times be based on the most limiting criteria.

i Prcpo d Tcchnic21 Sp cific0tien Ch ngo No. 40 1 Page 3 i Based on the above considerations, these changes reflect the application of methodologies recognized by the NRC and Industry ) as providing a sufficient margin of safety. The fracture toughness requirements of 10 CFR 50 Appendix G are satisfied and conservative operating restrictions are applied in the proposed heatup and cooldown

curves, therefore, these changes are considered to be safe and will not reduce the safety of the J

plant. l E. NO SIGNIFICANT RAZARDS EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating 31 cense for a facility licensed under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would nott (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. The following evaluation is provided 'for the no significant hazards consideration standards. 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The heatup and cooldown limit

curves, Figures 3.4-2 and 3.4-3, have been modified to reflect those curves applicable for 10 Effective Full Power Years (EFPY) provided in WCAP-12406,

" Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program." Our response to Generic Letter 88-11 concluded that implementation of Regulatory Guide 1.99 Revision 2, " Radiation Embrittlement of Reactor Vessel Materials" would be provided by changing the applicability date for the h'atup and cooldown curves to 5 EFPY.

However, the next surveillance capsule is due for removal at 6 EFPY and WCAP-12406 using the methods of Regulatory Guide 1.99 Revision 2 provides heatup and cooldown curves applicable to 10 EFPY, therefore, it is prudent to incorporate the 10 EFPY curves.

The reactor vessel material surveillance program complies with 10 CFR 50, Appendix G and H to ensure the reactor vessel has an adequate margin of safety with regard to material toughness throughout the service life of the

Pr:po20d Tcchnical Specification Chnngo No. 40 Page 4 plant. The program develops operating limits to prevent non-ductile failure and the operating limits are adjusted to account for the cumulative effects of radiation on the o reactor vessel material properties. The operating limits provided by these new curves were determined in accordance with the methodology set forth in the regulations to provide an adequate margin of safety to ensure the reactor vessel will withstand the effects of normal cyclic loads due to temperature and pressure changes as well as the loads associated with postulated faulted conditions. Therefore, j the proposed changes do not involve a significant increase in the probability or consequences of an accident previously ) evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Reactor vessel rupture is not part of the Beaver Valley design basis and is not included in the accident analysis. The new heatup and cooldown curves have been determined in accordance with Regulatory Guide 1.99, Revision 2 and WCAP-12406 and contain sufficient margin to ensure that the probability of a reactor vessel rupture is low enough that it is able to be excluded from the accident analysis. Changing the heatup and cooldown curves does not reduce the reliability of the reactor vessel or the procedures involved in plant heatup and cooldown. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the change involve a significant reduction in a margin of safety? The revised heatup and cooldown curves were established in accordance with current regulations and the latest regulatory guidance on RTNDT determinations. Because operation will be within these limits, the reactor vessel materials will behave in a nonbrittle

manner, thus, maintaining the original safety design basis.

i Therefore, the proposed changes do not involve a significant reduction in a margin of safety. F. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above safety

analysis, it is concluded that the activities associated with this license amendment request satisfies the no significant hazards consideration standards of 10 CFR 50.92(c)

and, accordingly, a

no significant hazards consideration finding is justified.

Prcpos d T;chnic31 Sp;cificcticn Chung3 No. 40 Page 5 G. ENVIRONMENTAL EVALUATION The proposed changes have been eyeluated and it has been determined that the changes do not involve (1) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released

offsite, or (iii) a significant increase in individual 1

or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22 (c) (9). Therefore, pursuant to 10 CFR 51.22 (b), an environmental assessment of the proposed changes is not required. l l l l

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==.M w .--a A A m. .e-.* _a ATTACHMENT C UFSAR Changes Beaver Valley Power Station, Unit No.2 Proposed Technical Specification Change No. 40 l \\ l I

BVPS-2 UFSAR TABLE 5.3-1 IRACIURE 100GHN[SS PPOPERilES Of IHE REACIOR VLSSEL So Ft-tb 35 Mi8 Mat'l Cu Ni P NDI Temp. NOI USE Componeng Code peo, __ Spec. 800 J_ 1 _1_

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Clostare head dome 89006-1 A5338 CL. 1 0.13 0.51 0.013 -20 50 -10 137 Closure head flar:9e 89002-1 A508 CL. 2 6.7/ O.012 -10 40 -10 836 l Vessel riange 89001-1 A508 CL. 2 4r94.73 0.010 0 10 0 132.5 I inlet nozzle 89011-1 A508 CL. 2 0.88 0.006 0 to 0 tots Inlet nozzle 89011-2 A508 CL. 2 0.88 0.010 in to to 115 iniet riozzte 89011-3 A508 CL. 2 0.8% 0.009 20 40 20 122 Outlet nozzle 89012-1 A508 CL. 2 0.71 0.00T -10 0 -10 137 Outlet esozzle 89012-2 A508 CL. 2 Destlet nozzle 89012-3 A508 CL. 2 .n e,79 0.006 -10 0 -10 121 l 0.68 0.008 -10 0 -10 112 peozzle simil 89003-1 A5338 CL. 1 0.13 0.61 0.008 -10 110 50 dae 9/ seozzle shett 89003-2 A5338 CL. 1 0.12 0.58 0.009 0 120 60 79.5 peozzle shell 89003-3 A5338 CL. 1 0.13 0.61 0.008 -10 110 50 97.5 Inter, stee l l 89004 A533B CL. 1 0.07 0.53 0.010 0 120 60 83 Inter. shett 89004-2 A5338 CL. 1 0.07 0.59 0.007 -10 100 40 75.5 t ower stie t i 89005-1 A5338 CL. 1 0.08 0.59 0.009 -50 88 28 82 t ower she t t B9005-2 A5338 CL. 1 0.07 0.58 C.009 -40 93 33 71.5 Bottom head torus 89010-1 A5338 CL. 1 0.15 0.49 0.007 -30 56 -4 91 Bottom head dome 89009-1 A5338 CL. 1 0.14 0.53 0.00T -30 35 -25 116 Weld ( inter, & fower sholl long seams & cirth scam)* 0.08 9f.07 .008 -30 30 -30 14g,5 HAZ ( Plate 89004-2) -80 40 -20 16 fe0f t :

• 3ame heat of wire and tot of flux used in all seams including surveillance weldsent.

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i ' WESTINGHOUSE CLASS 3 ,e l WCAP-12406 ANALYS!$ OF CAPSULE U FROM THE r DUQUESNE LIGHT COMPANY BEAVER VALLEY UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM r S. E. Yanichko S. L. Anderson L. Albertin N. K. Ray f September 1989 Work performed under Shop Order No. DF0P-106 . )W St APPROVED: T.A.Meyer,Enager Structural Materials and Reliability Technology Prepared by Westinghouse for the Duquesne Light Company WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230 w u. n u io 8 # c.

C PREFACE This report has been technically reviewed and verified. Reviewer .s Sections 1 through 5 and 7. P E. Terek 2, Iy.,2-i Section 6 E. P. Lippincott Y M ff t t i i Metuosasse to jj

TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 1-1 2 INTRODUCTION 2-1 3 BACKGROUND 3-1 4 DESCRIPTION OF PROGRAM 4-1 5 TESTING OF SPECIMENS FROM CAPSULE V 5-1 5-1. Overview 5-1 r 5-2. Charpy V-Notch Impact Test Results 5-3 I 5-3. Tension Test Results 5-4 5-4. Compact Tension Tests 5-4 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6-1. Introduction 6-1 6-2. Discrete Ordinates Analysis 6-2 6-3. Neutron Dosimetry 6-7 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1 8 REFERENCE 5 8-1 4 Appendix A - Heatup and Cooldown Limit Curves for Normal Operations 1 I . an. ave u $g3

. s - Q % 1 LIST OF ILLUSTRATIONS ( Figure Title Page l [ 4-1 Arrangement of Surveillance Capsules in the 4-5 Beaver Valley Unit 2 Reactor Uessel 4-2 Capsule U Diagram Showing Location of Specimens, 4-6 Thermal Monitors, and Dosimeters 5-1 ~ Charpy V-Notch Impact Data for Beaver Valley Unit 2 5-12 Reactor Vessel Shell Plate B9004-2 (Transverse Orientation) L 5-2 Charpy V-Notch Impact Data for Beaver Valley Unit 2 5-13 Reactor Vetsel Fr. ell Plate B9004-2 (Longitudinal Orientation) t 5-3 Charpy V-Notch Impact Data for Beaver Valley Unit 2 5-14 l Reactor Ve:sel Weld Metal 5-4 Charpy V-Notch Impact Data for Beaver Valley Unit 2 5-15 i Reactor Vessel Weld Heat Affected Zone Metal 5-5 Charpy impact Specimen Fracture Surfaces for Beaver 5-16 Valley Unit 2 Reactor Vessel Shell Plate B9004-2 (Longitudinal Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for Beaver 5-17 r-Valley Unit 2 Reactor Vessel Shell Plate 89004-2 (Transverse Orientation) m. ~ 5-7 Charpy Impact Specimen Fracture Surfaces for 5-18 Beaver Valley Unit 2 Reactor Vessel Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for 5-19 Beaver Valley Unit 2 Reactor Vessel HAZ Metal M 3994t/002740:10

i o e Q LISTOFILLUSTRATIONS(Cont) Figure Title Page 5-9 Tensile Properties for Beaver Valley Unit 2 Reactor 5-20 Vessel Shell Plate B9004-2 (Longitudinal Orientation) I 5-10 Tensile Properties for Beaver Valley Unit 2 Reactor 5-21 Vessel Shell Plate B9004-2 (Transverse Orientation) 5-11 Tensile Properties for Beaver Valley Unit 2 Reactor 5-22 Vessel Weld Metal 5-12 Fractured Tensile Specimens for Beaver Valley Unit 2 5-23 Reactor _ Vessel Shell Plate B9004-2 (Longitudinal Orientation) I l 5-13 Fractured Tensile Specimens for Beaver Valley Unit 2 5-24 Reactor Vessel Shell Plate B9004-2 (Transverse Orientation) 5-14 Fractured Tensile Specimens for Beaver Valley Unit 2 5-25 Reactor Vessel Weld Metal 5-15 Typical Stress-Strain Curve for Tension Specimens 5 i .6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 6-2 Core Power Distributions Use.d in Transport Calculations 6-14 For Beaver Valley Unit 2 3954a/09276910 y b

i o o, LIST OF TABLES Table Title Page 4-1 Chemical Composition of the Beaver Valley Unit 2 4-3 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the Beaver Valley Unit 2 4-4 Reactor Vessel Surveillance Materials 5-1 Charpy V-Notch Impact Data for the Beaver Valley Unit 2 5-5 Reactor Vessel Shell Plate 89004-2 Irradiated 18 at 550'F, Fluence 5.99 x 10 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Beaver Valley Unit 2 5-6 7 Reactor Vessel Weld Metal and HAZ Metal Irradiated I 18 at 550*F, Fluence 5.99 x 10 n/cm2 (E > 1.0 MeV) 5-3 Instrumented Charpy Impact Test Results for Beaver 5-7 Valley Unit 2 Reactor Vessel Shell Plate 89004-2 5-4 Instrumented Charpy impact Test Results for 5-8 Beaver Valley Unit 2 Reactor Vessel Weld Metal and HAZ Metal i 5-5 The Effect of 550*F Irradiation'at 5.99 x 1018 2 n/cm 5-9 (E > 1.0 MeV) on the Notch Toughness Properties of the Beaver Valley Unit 2 Reactor Vessel Materials l 5-6 Comparison of Beaver Valley Unit 2 R& actor Vessel Surveillance 5-10 l Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions l 5-7 Tensile Properties for Beaver Valley Unit 2 Reactor Vessel 5-11 18 Material Irradiated to 5.99 x 10 n/cm2 (E > 1.0 MeV) i 19346492749 10 yj

LIST OF TABLES (Cont) i .l l L Table Title Page 1' L. 6-1 Calculated fast Neutron Exposure Parameters at the 6-15 Surveillance Capsule Center L i l' 6-2 Calculated Fast Neutron Exposure Parameters at the 6-16 Pressure Vessel Clad / Base Metal Interface l I 6-3 Relative Radial Distributions of Neutron Flux 6-17 j. (E>1.0 MeV) Within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-18 (E>0.1 MeV) Within the Pressure Vessel Wall 7. 6-5 Relative Radial Distribution of Iron Displacement 6-19 Rate (dpa) Within the Pressure Vessel Wall 6-6 Nuclear Parameters for Neutron Flux Monitors 6-20 l 6-7 Irradiation History of Neutron Sensors Contained 6-21 in Capsule U 6-8= Measured Sensor Activities and Reaction Rates 6 6-9 Summary of Neutron Dosimetry Results 6-24 L 6-10 Comparison of Measured anc FERRET Cd eulated 6-25 i Reaction Rates at the Surveillance Capsule Center 1 L 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-26 l l Capsule Center L l l m.. m.,o y$$

i: 1 .s LISTOFTABLES(Cont) Table Title Page-6-12 Comparison of Calculated and Measured Exposure 6-27 Levels for Capsule U 6-13 Neutron Exposure Projections at Key Locations on the 6-28 Pressure Vessel Clad / Base Metal Interface 6-14 Neutron Exposure Values for Use in the Generation 6-29 of Heatup/Cooldown Curves 6-15 Updated Lead Factors for Beaver Valley Unit 2 Surveillance 6-30 Capsules t i I nuvomn io y$$$

.+ SECTION 1

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in Capsule U, the first surveillance capsule to be removed from the Duquesne Light Company Beaver Valley Unit 2 reactor pressure vessel, resulted in the following conclusions: The capsule received an average fast neut on fluence (E > 1.0 WeV) o 18 2 of 5.99 x 10 n/cm, o Irradiation of the reactor vessel intermediate shell Plate 89004-2, to 18 5.99 x 10 n/cm, resulted in a 30 and 50 ft-lb transition temperature increase of 30 and 35'F respectively for specimens oriented normal to the major working direction (transverse orienta-tion) and a 15 and 35'F increase respectively for specimens oriented parallel to the major working direction (longitudinal orientation), i i 18 2 o Weld metal irradiated to 5.99 x 10 n/cm experienced a.25'F increase in the 30 and 50 ft-lb transition temperature, 18 2 o Irradiation to 5.99 x 10 n/cm resulted in no decrease in the average upper shelf energy of Plate B9004-2 (transverse orientation) and a 5 ft-lb decrease in the upper shelf energy of the weld metal. Both materials exhibit a more than adequate upper shelf level for continued safe plant operation, o Comparison of the 30 ft-lb transition temperature increases for the Beaver Valley Unit 2 surveillance material with-predicted increases-using the methods of NRC Regulatory Guide 1.99, Revision 2, demon-strated that the Plate B9004-2 material and weld metal transition temperature increases were in relatively good agreement with the predicted increases, o Surveillance capsule test results for reactor vessel intermediate shell Plate B9004-2 and vessel core region weld metal indicate that these materials are not highly sensitive to irradiation at neutron 18 2 l fluences up to 5.99 x 10 n/cm, l . m w ee m,io 11 L s

=,, SECTION 2 INTRODUCTION This report presents the results of the examination of Capsule V, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Duquesne Light Company Beaver PO 4 y Unit 2 reactor pressure vessel materials under actual operating conditions. The surveillance program for the Duquesne Light Company Beaver Valley Unit 2 ~ reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechan'ical_ properties of the reactor vessel materials are presented by Davidson and Yanichko.Ill The surveillance program was planned-to cover the 40 year design life of the reactor pressure vessel and was based on ASTM E-185-73, " Standard Recommended Practice for Surveillance i Tests for Nuclear Reactor Vessels". Westinghouse Electric Corporation personnel performed the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens. This report summarizes testing and the postirradiation data obtained from surveillance Capsule U removed from the Beaver Valley Unit 2 reactor vessel and discusses the analysis of the data. l L 1 animeme io 2-1 1

SECTION 3 BACKGROUND The ability of the large steel pressure vessel, which contains the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is-the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels such as SA533 Grade B Class 1 (base material of the Beaver Valley Unit 2 reactor pressure vessel beltline) are well documented in industry literature. Generally, low alloy ferritic materials demonstrate an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation, t A method for performing analyses to guard against fast fracture in reactor i pressure vessels has been presented in " Protection Against Non-ductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)* 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 NDT material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G of the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable-stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors, i m e. m m io 31

RTNDT and the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program.II) 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 to adjust the RT f r radiation embrittlement. This adjusted RT NDT NDT (RT initial + ARTNDT) is used to index the material to the KIR NDT curve and to set operating limits for the nuclear power plant which reflect i, L the effects of irradiation on the reactor vessel materials. l l I i au.ame io 3-2

SECTION 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Beaver Valley Unit 2 reactor pressure vessti core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were 4 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 opposite the vertical center of the core. Capsule.V (Figure 4-2) was removed after 1.24 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 Plate B9004-2,' submerged arc weld metal identical to that l used for the beltline region girth and longitudinal weld seams of the reactor vessel and Charpy V-notch specimens from weld heat-affected zone (HAZ) y material. All heat-affected zone specimens were obtained from within the HAZ bf Plate B9004-2 of the representative weld. The chemistry and heat treatment of the surveillance material are presented in Table 4-1 and Table 4-2, respectively. The chemical analyses reported in ' Table 4-1 were obtained from unirradiated material used in the surveillance program. 1 All test specimens were machined from the il4 thickness location of the plate, j Test specimens represent material taken at least one plate thickness from the quenched end of the plate. All base metal Charpy V-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 working direction of the plate. Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction. Tensile specimens were oriented with the longitudinal 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 39%4s/092789 to pg

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 fatigue precracked per ASTM E399-70T. Ccpsule U contained dosimeter wires of pure iron, copper, nickel, and unshielded aluminum-cobalt. -In addition, cadmium shielded dosimeters of Neptunium (Np237)andUranium(U238) were contained in the capsule. 1 Thermal monitors made from two low-melting eutec:ve alloys and sealed in Pyrex tubes were included in the capsule and 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 (310*C) The arrangement of the various mechanical test specimens, dosimeters and thermat monitors contained in Capsu~1e U are shown in Figure 4-2. l i au,mme io 4-2

7 %u WESTINGHOUSE CLASS 3 4 TABLE 4-1 l CHEMICAL COMPOSITION OF j THE BEAVER VALLEY UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS Plate B9004-2 Weld Metal (a) l Element (Wt. %) (Wt. %) i C 0.24 0.10 j S-0.016 0.011 l N 0.009 0.028 2 Co-0.009 0.007 l Cu 0.05 0.08 i Si 0.24 0.14 l Mo 0.59 0.49 Ni 0.56 0.07 f j Mn 1,32 1.17 l Cr 0.08 0.07 V 0.003 0.002 P 0.010 0.008 Sn 0.008 0.005 3 Al' O.047 0.001 ) Ti <0.010 < 0. 01. W 0.010 <0.010 Zr 0.002 <0.001 4 As 0.010 0.005 Cb <0.010 <0.010 B 0.0003 <0.001 (a) Surveillance weld specimens were made of the same wire and flux as the Intermediate to lower shell vertical and girth-weld seams (Wire Heat 83642 and Linde 0091 Flux Lot 3536) 3954s/092749 10 4-3

TABLE 4-2 HEAT TREATMENT OF THE BEAVER VALLEY UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS Material Temperature (*F) Time (hr) Coolant Intermediate Shell 1575/1625 4 Water quenched Plate B9004-2 1200/1250 4 Air cooled 1115/1165 30 . Furnace cooled Weld Metal 1125/1175 13.5 Furnace cooled \\ I i 9 Mt/ 7M 10 4,4 d

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=-.. 1 SECTION 5 TESTING 0F SPECIMENS FROM CAPSULE U 5-1. OVERVIEW The postirradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Research and Development Laboratory with consultation by Westinghouse Nuclear Energy Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and g2), ASTM Specification E185-82 and Westinghouse Procedure NHL 8402, Revision 1 as modified by Westinghouse RMF Procedures 8102, Revision 1 and 8103, Revision 1. Upon receipt of the capsule at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked 7 against the master list in WCAP-9615.I13 No discrepancies were found, i Examination of the two low-melting 304*C (579*F) and 310*C (590*F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test, specimens were exposed was less than 304*C (579'F). The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 3580 machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology model 500 instrumentation system. With this system, lead-timi and energy-time signals can be recorded in addition to the standard measuremeat of Charpy energy (E ). From the load-time curve, the load of general yielding D (PGY), the time to general yielding (tgy), the maximum load (P ), and M the time to maximum load (t ) can be determined. Under some test M conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (P ), and the load at which fast fracture terminated is p identified as the arrest load (P )* A i 3954s/092789 10 5-1 l

The energy at maximum load (E ) was determinod by comparing the energy-time M ov erd and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E ) is the difference p between the total energy to fracture (E ) and the energy at maximum load. D The yield stress (oy) is calculated from the three point bend formula. The flow stress is calculated from the average of the yield and maximum loads, also using the three point bend formula. Percentage shear was determined from postfracture photographs using.the ratio-of-areas methods in compliance with ASTM Specification A370-77. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification. Tension tests were performed on a 20,000 pound Instron, split-console test i I machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inch per minute throughout the test. Deflection measurements were made with a linear variable displacement transducer (LVOT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-67. Elevated. test temperatures were obtained with a three-zone ele:tric resistance split-tube furnace with a 9-inch hot zone. All tests were cenducted in air. s Because of the difficulty in remotely attaching a thermNouple directly to the specimen, the following procedure was used to monitor specimen temperature. Chromel-alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In test an.mna in 5-2

i configuration, with a slight load on the specimen, a plot of specimen j temperature versus upper and lower grip and controller temperatures was 6 developed over the range room temperature to 550*F (288*C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures. Experiments indicated that this method is accurate to plus or minus 2*F. u The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were _ calculated using.the original cross-sectional area. The final diameter and final gage length were determined from postfracture photographs. The fracture area used to calculate i the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement. 5.2. CHARPY V-NOTCH IMPACT TEST RESULTS 7 i The results of Charpy V-notch impact tests performed on the various materials contained in Capsule U irradiated to approximately 550*F at 5.99 x 1018 2 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 Capsule U material are shown in Table 5-5. Irradiation of the vessel intermediate shell Plate B9004-2 material (transverse orientation) specimens to 5.99 x 1018 2 n/cm -(Figure 5-1) resulted in a 30 and 50 ft-lb transition temperature increase of 30 and 35'F respectively, and an upper shelf energy increase of 8 ft-lb when compared to the unirradiated data.Ill Irradiation of the vessel intermediate shell Plate 89004-2 material (longitudinal orientation) specimens to 5.99 x 1018 m/cm2 (Figure 5-2) resulted in a 30 and 50 ft-lb transition temperature increase of 15 and 35'F respectively, and an upper shelf energy increase of 10 ft-lb when compared to the unirradiated data. 3954s1092789 to 5-3

t ': 18 Weld metal irradiated to 5.99. 10 n/cm2 (Figure 5-3) resulted in a 30 and 50 ft-lb transition temperature increase of 25'F, and an upper shelf energy decrease of 5 ft-lb. Weld HA2 metal irradiated to 5.99 x 10 n/cm2 (Figure 5-4) resulted in no 18 30 and 50 ft-lb transition temperature increase and an upper shelf energy increase of 18 ft-lb. The fracture appearance of esch irradiated Charpy specimen from the various materials is shown-in Figures 5-5 through 5-8 and show an increasing ductile or tougher appearance with increasing test temperature. Table 5-6 shows a comparison of the 30 ft-lb transition temperature (ARTNOT) increases for the various Beaver Valley Unit 2 surveillance materials with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2 I3) This comparison shows that the transition temperature 18 2 increase resulting from irradiation to 5.99 x 10 n/cm is in relative l agreement with the Guide prediction for Plate B9004-2 and the weld metal. 5-3. TENSION TEST RESULTS The results of tension tests performed on Plate 89004-2 (transverse and i longitudinal or'ientation) and weld metal arradiated to 5.99 x 1018 2 n/cm are shown in Table 5-7 and Figures 5-9, 5-10 and 5-11, respectively. These results show that irradiation produced a 0.2 percent yield strength increase no greater than 7 ksi for Plate B9004-2 and 5 ksi for the weld metal. Fractured tension specimens for each of the materials are shown in Figures 5-12, 5-13 and 5-14. A typical stress-strain curve for the tsnsion specimens is shown in Figure 5-15. 5-4. COMPACT TENSION TESTS l l It was decided that the 1/2T - Compact Tension fracture mechanics specimens l will not be tested and will be stored at the Hot Cell at the Westinghouse R&D Center, since the U.S. Nuclear Regulatory Commission has recommended that surveillance capsule fracture mechanics specimens not be tested at this time, nuvowie io 5-4 0

TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE 89004-2 18 IRRADIATED AT 550'F, FLUENCE 5.99 x 10 n/cm2 (E > 1.0 MeV) Temperature Impact Energy Lateral Expansion Shear Sample No. (*F) -(*C) (ft-lb) ,(,f), (mils) M (%) Longitudinal Orientation E11 -60 -5 4.0 5.5) 5.0 (0.13)- 5 WL3 -30 -3 18.0 24.5 10.0 (0.25) 5 h WL6 25 33.0 44.5 22.0 (0.56) 15 WL1 25 -4) 25.0 34.0 19.0 (0.48) 15 WL12 50-( 10) 29.0 39.5 20.0 (0.51) 20 WL10 50 ( 10 24.0 32.5 -20.0 (0.51 20-WL7 74 ( 23 35.0 ( 47.5 24.0 (0.61 30 WL13 75 ( 24 33.0 ( 45.0)- 24.0 (0.61 30 WL5 100 ( 38) 48.0 ( 65.0) 37.0 (0.94) 45 t WL4 125 ( 52) 53.0 (72.0) 39.0 (0.99) 45 i WL14 125 ( 52) 37.0 (50.0) 33.0 (0.84) 40 WL8 150 ( 66) 55.0 (74.5) 45.0 (1,14) 50 WL2 200 ( 93) 94.0 (127.5) 66.0 (1.68) 100 WL15 300 (149) 109.0 (148.0) 72.0 (1.83) 100 WL9 400 (204) 111.0 (150.5) 77.0 (1.96) 100 Transverse Orientation WT12 -30 (-34) 13.0 ( 17.5) 8.0 (0.40) 5 WT10 25 ( -4) 19.0 ( 26.0) 14.0 (0.36) 10 WT8 40 ( 4) 23.0 (31.0) 23.0 (0.58) '15 WT15 60 ( 16) 37.0 ( 50.0) 27.0 (0.69) 20 WT5 60 (16) 40.0 ( 54.0) 33.0 (0.84) 35 l WT7 74 ( 23) 34.0 ( 46.0) 23.0 (0.58) 30 WT2 75 ( 24) 33.0 ( 44.5) 26.0 (0.66) 30-WT14 100 (38) 39.0 ( 53.0) 34.0 (0.86) 40 WT9 100 ( 38) 43.0 ( 58.5) 37.0 (0.94) 40 WT1 125 ( 52) 45.0 ( 61.0) 38.0 (0.97) 45 WT11 150 ( 66) 58.0 ( 78.5) 46.0 (1.17) 50 i' WT6 200 ( 93) 67.0 ( 91.0) 49.0 (1.24) 95 l: WT3 250 (121) 89.0 (120.5) 65.0 (1.65) 100 WT4 350 (177) 87.0 (118.0) 66.0 (1.68) 100 WT13 450 (232) 85.0 (115.0) 59.0 (1.50) 100' s animme io 5-5 1

j' l TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE BEAVER VALLEY UNIT 2 REACTOR VESSEL WELD METAL AND HAZ METAL IRRADIATED AT 550*F 18 FLUENCE 5.99 x 10 n/cm2(E>1.0MeV) 1 Temperature Impact Energy Lateral Expansion Shear Sample No. ('F) ('01 (f t-lb) 2), (mils) ,(. ngl (%) Weld Metal WW6 -50 (-46)- 11.0 15.0) 12.0 (0.30) 10 WW13 -25 -32) 17.0 23.0) 12.0 (0.30) 10 W3 -20 (-29 28.0 -38.0) 22.0 .(0.56) 25 WW10 -10 (-23 115.0 156.0) 47.0 (l'.19) 100 WW5 0 -18 23.0 31.0) 19.0 (0.48) 20 W12 0 -1 89.0 (120.5) 62.0 (1.57) 85 .W4 10 -1 56.0 (76.0) 45.0 (1.14). 60 WW15. 25 63.0 ( 85.5) 44.0 (1.12) 60 WW8 25 108.0 (146.5) 74.0 1.88)- 100 WW7 50 10) 88.0 (119.5) 62.0 1.57) 85 l W1 75 24) 134.0 (181.5) 83.0 2.11) 100 W2 125 52) 121.0. (164.0)- 83.0 (2.11) 100 W14 200 93) 136.0 (184.5) 76.0 (1.93) 100 l W9 300 (149) 145.0 (196.5) 79.0 (2.01) 100 l W11 400 (204) 164.0 (222.5) 92.0 (2.34) 100 HAZ Metal WH14 -150 -101) 16.0 ( 21.5) 8.0 ' (0.20) 5 WH12 -100 - 73) 25.0 ( 34.0) 19.0 (0.48) 15 WH5 -100 - 73) 27.0 (36.5) 14.0 (0.36) 20 WH8 - 80 (- 62)- 22.0-( 30.0) 17.0 (0.43) 15 1 WH13- - 75 (- 59) 39.0 ( 53.0) 21.0 (0.53) 20 i' WH4 - 75 '-(- 59) 48.0 (65.0) 24.0 (0.61) 35 WH2 - 50 (- 46) 47.0 (63.5) 30.0 (0.76) 40 WH15 - 25 (- 32) 44.0 (59.5)' 29.0 (0.74) 40 WH1 - 25 (-~ 32) 77.0 (104.5) 45.0 (1,14) 55 WH6 0 (-_18) 59.0 ( 80.0) 42.0 (1.07) 55 WH7 25 (- 4) 92.0 (124.5) 51.0 (1.30) 85 WH11 75 ( 24) 89.0 (120.5) 48.0 (1.22) 100 WH10-125 ( 52) 120.0 (162.5) 73.0 (1.85) 100 WH3 200 ( 93) 113.0 (153.0) 63.0 (1.60) 100 WH9 300 ( 149) 113.0 (153.0) 61.0 (1.55) 100 30Sas/09'70910 5-6

e TABLE S-3 INSTRUMENTED CHARPY IMPACT TEST RESULTS FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE 89004-2 Normalised Eneraies Test Charpy Charpy Marinum Prop Yield Time Maximus Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Em/A Ep/A 1,oad to Yield I,oad Maximum I,oad 1,oad Stress Stress Number f*F1 (ft-lb) (ft-lb/in ) (kips) (psec) (kips) (psec) (kips) (hips) (ksi) (ksi) 1,ongitudinal orientation E11 -60 4.0 32 6 26 2.00 25 2.80 35 2.80 66 80 E3 -30 18.0 145 88 57 2.50 200 4.10 330 4.10 83 109 El 25 25.0 201 130 71 2.35 65 4.20 325 4.10 0.15 78 109 E6 25 33.0 266 205 61 2.65 160 4.60 510 4.55 0.20 87 119 E10 50 24.0 193 135 58 2.85 45 4.20 300 4.20 0.45 94 116 E12 50 29.0 234 153 80 3.20 155 4.35 430 4.35 0.30 105 125 E7 74 35.0 282 186 96 2.75 115 4.40 450 4.40 1.05 91 118 E13 75 33.0 266 T E5 100 48.0 387 294 93 3.35 55 4.50 620 4.45 1.40 111 130 E14 125 37.0 298 156 142 2.85 100 4.15 380 4.15 1.55 94 115 E4 125 53.0 427 304 123 2.55 60 4.50 650 4.50 2.15 85 117 E8 150 55.0 443 199 244 '3.05 60 4.20 460 4.05 2.80 101 120 E2 200 94.0 757 281 476 3.05 240 4.20 795 101 120 EIS 300 109.0 878 270 607 2.65 95 3.90 665 88 109 E9 400 111.0 894 282 611 2.45 120 3.95 720 81 106 Transverse Orientation WT12 -30 13.0 105 71 34' 2.75 90 4.20 205 4.20 91 115 WTIO 25 19.0 153 110 43 2.45 55 4.00 270 4.00 0.30 81 107 WT8 40 23.0 185 106 79 3.70 55 4.25 235 4.15 0.30 122 131 MIS 60 37.0 298 205 93 3.00 80 4.35 475 4.35 0.45 98 122 M5 60 40.0 322 217 105 3.10 70 4.45 475 4.45 1.05 102 125 M7 74 34.0 274 170-103 3.20 75 4.35 395 4.30 0.95 107 125 WT2 75 33.0 266 183 103 2.70 80 4.25 400 4.20 1.10 89 115 WT14 100 39.0 314 221 93 2.70 95 4.30 525 4.30 1.25 88 116 M9 100 43.0 346 231 115 3.35 165 4.15 575 4.40 1.30 110 124 WT1 125 45.0 362 196 166 2.30 110-4.15 510 4.15 1.80 76 107 M11 150 58.0 467 252 215 3.00 165 4.30 645 4.20 2.30 100 121 M6 200 67.0 540 217 323 3.15 180 4.10 595 4.00 3.75 104 120 WT3 250 89.0 717 276 441 2.95 166 "* 4.20 705 98 118 M4 350 87.0 701 253 448 2.65 170 3.90 690 88 108 M13 450 85.0 684 209 475 2.50 100 3.55 560 83 100 ~. .. = - - - - - = - - ^ ' ' ' ~

l-TABLE 5-4 INSTRIE NTED CHARPY IMPACT TEST RESULTS FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL WELD ETAL Af5) HAZ ETAL Normalised Enersties Test Charpy Charpy Maximus Prop Yield Time Maxiome Time to Fracture Arrest Yield Flow Sample Temp Energy Ed/A Ee/A Ep/A Load to Yield Load Maximum Load Load Stress Stress 2 Number M (ft-lb) (ft-lb/in ) (kips) (msec) (kips) (ssec) (kips) (kips) (ksi) (ksi) l i Weld Metal j M6 -50 11.0 89 40 49 3.20 140 3.80 170 106 116 W3 -25 17.0 137 71 65 3.60 185 4.00 270 4.00 0.25 119 126 WW3 -20 28.0 225 119 107 2.65 60 4.25 275 4.20 0.65 87 114 WW10 -10 115.0 926 292 634 3.20 80 4.50 605 106 128-I W5 0 23.0 185 96 89 2.65 65 3.90 245 3.90 0.75 87 108 I W12 0 89.0 717 301 416 3.15 190 4.60 725 4.05 1.65 104 128 WW4 10 56.0 451 226 225 3.65 180 4.45 570 4.35 1.80 120 133 WIS 25 63.0 507 '287 220 3.00 230 4.30 755 4.30 1.30 99 121 i g' WW8 25 108.0 870 309 561 3.25 105 4.40 675 108 127 m WW7 50 88.0 709 217 492 3.05 180 4.30 575 4.10 2.50 101 121 WW1 75 134.0 1079 352 727 3.00 145 4.30 840 98 120 i W2 125 121.0 974 336 639 3.10 175 4.10 850 103 119 i WW14 200 136.0 1095 319 776 2.75 80 3.85 775 91 109 t WW9 300 145.0 1168 282 886 2.60 90 3.90 735 85 107 WW11 400 164.0 1321 274 1046 2.05 60 3.80 715 67 97 i RAZ Metal WH14 -150 16.0 129 86 43 3.60 130 5.10 225 5.00 120 144 WH12 -100 25.0 201 148 53 3.85 135 4.70 335 4.70 0.30 127 142 [ WH5 -100 27.0 217 135 82 3.35 85 4.85 285 4.85 0.60 '111 136 WH8 -80 22.0 177 59 118 3.60 65 4.60 145 4.55 0.80 119 136' WH13 -75 39.0 314 194 120 3.90 70 4.85 380 4.85 1.20 128 144 l WH4 -75 48.0 387 230 156 3.50 95 4.75 470 4.60 1.95 115 136 [ WH2 -50 47.0 378 165 214-3.55 245 4.60 500 4.50 2.25 118 135 WH15 -25 44.0 354 137 217 2.95 75 4.55 300 4.55 3.60 98 124 i WB1 -25 77.0 620 252 368 3.65 105 4.65 530 3.85 1.50 121 137 WH6 0 59.0 475 277 198 3.50 130 4.55 610 4.45 2.50 115 132 WH7 25 92.0 741 302 439 3.80 60 4.60 605 3.25 2.05 125 138 WH11 75 89.0 717 195 522 3.15 115 4.40 450 104 125 i WR10 125 120.0 966 282 684 2.90 100 4.35 630 95 120 WH3 200 113.0-910. 280 630 2.85 85-. 4.05 670 94 114 't WH9 300 113.0 910 257 653 2.75 75 4.05 610 91 112 MSUU9H89 30 m

~ 3 6 TABLE 5-5 THE EFFECT OF 550*F IRRADIATION AT 5.99 x'1018,fc,2 (E > 1.0 MeV) ON THE NOTCH TOUGHNESS PROPERTIES OF THE BEAVER VALLEY UNIT 2 REACTOR VESSEL MATERIALS Average Average 35 mii Average Average Energy Absorptton 30 ft-lb Temp (*F) Lateral Expansion Temp (*F ) 50 f t - 1 b Temp (

• F )

at Full Shear (ft-lb) Material Untrradiated Irradiated At Unirradiated Irradiated AT Unirradiated Irradiated AT Untrradtated Irradtated A(ft-Ib) Plate 89004-2 40 70 30 95 '95 0 95 130 35 79 87 +8 (Iransverse) Plate 89004-2 40 55 15 85 120 35 95 130 35 95 105 +10 (Longitudinal) g E Weld Metal -35 -10 25 -25 -5 20 -15 10 25 139 134 -5 HAZ Metal -75 -75 0 .-20 -20 0 -35 -35 0 91 109 +18 1 l l l l i l l 3950s/092789: 10 i

TABLE 5-6 i COMPARISON OF BEAVER VALLEY UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY IMPACT TEST RESULTS' WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS ARTNDT ( F) VSE DECREASE (%) Fluence 19 2 Material Capsule 10 n/cm Meas. Pred; Meas. Pred. I Plate B9004-2 U 0.599 30 27 0 17 (Transverse) Plate 89004-2 U 0.599 15 27 0 17 -(Longitudinal) .i I Weld Metal U 0.599 25 37 4 19 i (: L^ l l l i i 3est-w?nt'e 5-10 l-1

i .) TABLE 5-7 TENSILE PROPERTIES FOR BEAVER VALLEY UNIT 2 REACIOR VESSEL MATERIAL IRRADIATED TO S.99 x 1018,7c,2 (E > 1.0 MeV) l Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction { Sample Temp. Strength Strength Load Stress Strength Blongation Elongation in Area Material Number f*fl (ksi) (ksi) (kip) (ksi) (ksii (5) (%) (5) Plate B9004 WL1 76 76.9 98.8 3.25 132.2 66.2 11.3 24.6 60 (Long. WL2 250 71.3 91.7 2.95 164.1 60.1 9.8 22.2 68 4 Orient.) WL3 550 67.2. 94.7 3.25 163.7 66.2 - 9.8 21.0 60 ~ Plate 9004-2 WT1 75 76.0 98.8 3.50 271.1 71.3 10.5 21.6 64 (Transv. Tf2 250 71.8 92.7 3.20 212.4 65.2 9.0 19.7 64 Drient.) Tf3 550 69.8 94.7 3.55 153.1 72.3 9.0 18.5 53 l Weld. WW1 25 83.5 95.7 2.85 204.2 58.1 9.8 23.7 72 i WW2 200 72.8 87.6 2.50 186.1 50.9 8.3 21.5 73 WW3 550 70.8 87.6 2.80 186.7 57.0 8.6 20.4 69 i ? L t 5 i 3950s/092789 to k f _.. ~. l

\\ "Gl -150 -100 - 50 0 50 100 150 200 250 I I I I I 3l l l 100 - kl _t% B2 i o 5@ 9' 2 E g. 3 0 1 I 1 I I i I l# E5 i i i i i i i i i _5E 10 E-2 L5 e 8@ LO ' "E 45 3 2 0 1 I i i i i i 0 2% 1M 2d 160 200 _ le a p120 gg ~ 100 Unirradiat 2s y - g-120 3 h80 5 / Irradiated at 550'F - 80 60 19 2

5. 99 x 10 n/cm 30'F 20 o

0 i I i i i i 0 - 200 -100 0 100 200 300 @0 500 Temperature (*F) FIGURE 5-1 CHARPY V-NOTCH IMPACT DATA FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE B9004-2 (TRANSVERSE ORIENTATION) as..mne, io 5-12

__

1 tu -150 -100 - 50 0 50 100 150 200 250 1 I I I l l 3 i-100 ~f 80 / j Bm . /. o c cm M 2 g 'N-22 / ,2 i 0 i i i i l l I I I I I I l-5 80 2.0 s 2 - + - - - L5 s f 8# /

1. 0 '

35'F x 9% i i e i i i i 2% i i i i i i i i i 180 2@' 160 200 _ l@ a g120 160 g Unirradiateg 2 { 2-_- 100 ^ }g, .= 80 8 \\ Irradiated at 550*F g 80 3FF 18 2 5.99 x 10 n/cm e ISF O o 0 i i i i i i i i 0 - 200 -100 0 100 200 300 40 500 Temperature (*F) FIGURE 5-2 CHARPY V-NOTCH IMPACT DATA FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE B9004-2 (LONGITUDINAL ORIENTATION) mwome io 5-13

( C) -150 -100 - 50 0 50 100 150 200 250 I I I I I I I I i 3 100 ee e: 4-+--e Ea e l 3e 4 ) 5e l 20 2 ~ 0 l 100

2. 5 i

i i i i i i, i 5 80 T' - 2.0 5 d6' 60

1. 5 3 N4 b

1.0$ f. 20'F g 20 0.5 0 0 200 i i i i 180 2C 12 _ 14 Unirradiated 200 /* J _ _',__ a g120 160 ~ 100 / Irradiated at 550*F 120b ~ I II 2 5 o 5,99 x 10 n/cm h 'F R 29 O 25'F 20 e 0 i i i i i i i i 0 - 200 -100 0 100 200 300 40 500 Temperature ('F) ~ FIGURE 5-3 CHARPY V-NOTCH IMPACT DATA FOR BEAVER VALLEY UNIT 2 RE WELD METAL nu.=nes io 5-14

( S C) -150 -100 - 50 0 50 100 150 200 250 I I I I gi I I I l '3 100 " ^ ^ 2 _f 80 3 M e e am M ,e M 2 3 N 0 i I I I I IM 15 i I i i i i i i i _jM 10 i - 60 S i 1,51 e 8M 1,0 $ 1 ~ i i i i i i i 200 i 4 I I I I I l l l l 180 2e Irradiated at 550*F ig 18 2 le

5. 99 x 10 n/cm 200 8

3 E120 ig y o,s.3-- 9 @M WI" 2 120 ' s e w 60 80 8 Unirradiated g o 3 2 0 i I i 0 - 200 -100 0 100 200 300 00 500 Temperature (*F) FIGURE 5-4 CHARPY V-NOTCH IMPACT DATA FOR BEAVER VALLEY VNIT 2 REACTO WELD HEAT AFFECTED ZONE METAL m e. m niiio 5-15 i e ---,-,---y+- -w,w-,---.,y-y .w-,-,-, .y,, e,w4e,

E11 E3 E6 El E12 k ~ (t';l, i l o /%3 f.D .l. t q m. n ) E10 E7 E13 E5 E4 7m X em, Q:=, E14 E8 E2 E15 E9 i 4 FIGURE 5-5 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE B9004-2 (LONGITUDINAL ORIENTATION) 3.i.. e.n.. i. 5-16 LM-20874

1 i l i Il*lrY h. 'i.},$1 ![ \\f'l j l t, ? 'l .., 5,;A N: l WT12 Mio WT8 WT15 WT5 l I r; . [g' 1 l ),Nd ,I.e2 l i li {W,:! f.*,,. 4 4 . w,- 1 I WT7 WT2 WT14 WT9 WT1 1 i if l \\: .$ft t i l . E $',, f ~,, j,, j I il r w emm-mu A WT11 MS WT3 M4 WT13 I FIGURE 5-6 CHARPY IMPACT SPECIMEN FRACTURE SURFACiS FOR BEAVER VALLEY UNIT 2 l REACTOR VESSEL SHELL PLATE 89004-2 (TRANSVERSE ORIENTATION) l 3essi/osatte to 5-U i RM-20B75 _ -... _. _ _, ~. _ _ _ _ _ _. _ _ _.. _ _ _.. _.. -

l \\ j N6 W13 W3 W10 WS i l i }- ' d s. Q-W12 W4 W15 W8 M7 i dd },g. I p' l N. W1 W2 W14 W9 W11 FIGURE 5-7 CHARPY IMPACT SPECIMEN FRACTURE SVRFACES FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL WELO METAL meiav ' 5-18 RM-20876

l 6 l i!EB;_EE i H14 4 12 US M18 M113 r p

l. _ h'l

• (%-

4 -~ i %+, l g, .A0h(. kd' i ,,. e M4 M2 E15 n1 we i 1 0 L k Mi7 Mill glo mi3 gg FIGURE 5-8 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL WELD HAZ METAL i nsweavis io 5-19 RM-20877 p

r c a n psys,s., l 'C - 50 0 50 100 150 200 250 300 120 i i i i i i i i g 110 - 100 700 Tensile strength $90 7, 28 4 g ~ 30 2 m 70 I e ,2% Yield Strength 500 0 60 8 g 50 40 i i i i i i i-300 Code: Open Points -Unirradiated. IE 2 Closed Points -Irradiated at 5.99 x 10 n/cm 80 i i i i i i i i 60 E g $M ]30 Total Elongation 'd' 20 1 10 0 ,uniforpiongation g -100 0 100 200 300 400 500 600 Temperature ( *F) FIGURE 5-9 TENSILE PROPERTIES FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE 89004-2 (LONGITUDINAL ORIENTATION) mwoenee so 5.gn l .... ~.. - - . _. -.. ~., _ _ _ _ -,-we e a+-e.-w--- -w-

l 1 l

• C

- 50 0 50 100 150 200 250 300 4 120 I 800 110 i 700 Tensile strengtn 100 90 e d 600g 80 a E

• N.-

500 ti, lo 29 e M g 0,2 % Yield Strength 50 l \\ \\ \\ \\ s l I M Code : Open Points -Unirradiated 18 2 Closed Points - Irradiated at 5.99 x 10 n/cm E I I I I I I I l l 70 Reduction in Area j g _fz $M ]30 Total Elongation ^ 20 4- ! 10 I g [ 0 1 I I I Unifor[n Elongption -100 0 100 200 300 400 500 600 Temperature ( *F) FIGURE 5-10 TENSILE PROPERTIES FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE B9004-2 (TRANSVERSE ORIENTATION) nu.avie is 5-21 ,-.y--- ._,_m_ - - _ _, -., -.. _ _,.. - - -..., _.. -

's curve 737516-A 'C - 50 0 50 100 150 200 250 300 120 i i i i i i i i 110 g 100 700 ^ .g g Tensile Strength 4 d. i. i 2 8 g 70 % g_ g 60 0,2 % Yield Strength 400 50 g i i i i i i i-E Code: Open Points - Unirradlated 18 2 Closed Points -Irradiated at 5,99 x 10 n/cm E e l l I I I I I I I 70

• --8 " """""8NS Reduction in Area g

Qi-g $40 ]30 Total Elongation '~ 20 '"""""'-"-------~~~~I Uniform Elongation 10 e- -o g 0 I I I I -100 0 100 200 300 400 500 600 Temperature ( *F) FIGURE 5-11 TENSILE PROPERTIES FOR BEAVER VALLEY UNIT 2 METAL me.une ie 5-22

F- . jf(Qigsagi: J. 1 v at?[J s h ,2 p t.,7. 8 9 Yal? e - 9 l ^ r h4 'dahh!xtei 5pecimen El 25'F ~hi>322W T s .!)i t lone. 1001 % 2 ) o r n. .} by Ws-Specimen WL2 250*F i a% y ??,l;.p; w g .I Specimen WL3 550'F FIGURE 5-12 FRACTURED TENSILE SPECIMENS FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE B9004-2 (LONGITUDINAL ORIENTATION, t 39St e '092789 10 $-23 RM-20878

h.l&y'". . y.,, - na ' .,.x u. [l> "-l'[l} wp { :; -.s 1. -e o 3 8A:: pap" I ' ..,e n- .,g m.: 3 i .*a . itehlit.ir,,, ,W A% Specimen WT1 75'F sqr )r:c l N% Aj mw-y %M ~> .t-T l ( !. I 2 ;, j t .~. I 1o t ><, I toon<. a, 8 2 O t u b s h'ahlub Specimen WT2 250*F z 4. ,f.9 . nygt. [NWh si udd@M9I".'.

• ,... :.. geer - f *e

,'k. k. ,e c. ..w .s.3.

• +

f. -mor I l 1 _ o 8 1 -, P lors I,1.

• p;;.

,loor>6 2 3 4 J p i 11 > > J i. _, Ja da!,5 - Specimen Fr3 550*F FIGURE 5-13 FRACTURED TENSILE SPECIMENS FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL SHELL PLATE B9004-2 (TRANSVERSE ORIENTATION) me. swee o 5 24 m-:osn l l

O %. I gC* E{ v[Os.k.fi I c.'-i 'g a Pl ! ' ^ b J.T..; c Wa 9 TFes r 3,< 0,l ,, A, r Specimen W1 25'F ,q y ;... .3 s i r I a; b, 3 0?? l r teorm .w >1 Specimen W2 200*F l %: g}k;. l .; e ;,.,, "Mi%$ekhmas.- i we 3,% !A( tw 'y \\ Sptcimen WY3 550*F 1 l FIGURE 5-14 FRACTURED ' ENSILE SPECIMENS FOR BEAVER VALLEY UNIT 2 REACTOR VESSEL WEL) METAL me. cin' ' 5-25 RM-2 0880

120 100 w 80

• a M

i g 60 ~ E .hv) 40 t ) 20 i - SPEC.WL3 550F 0 0.04 0.b8 0,i2 O.k6 0.20 0.24 Strain, in/In FIGURE 5-15 TYPICAL STRESS-STRAIN CURVE FOR TENSION SPECIMENS seu.rowss,o 5.gg

SECTION 6 RADIATION ANALYSIS AND NEUTRON 00SIMETRY

6.1 INTRODUCTION

Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment.at various positions within the reactor vessel and that 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 contaired in each of F the surveillance capsules. The latter information is derived solely from analysis. The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested thit an exposure model that accounts for differences in ntutron energy spectra between surveillance capsule locations and positions uithin the vessel wall could load to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall. Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence m w oo m i io 6-1

(E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials." This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule U. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the j projected exposure of the pressure vessel are provided. 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the reactor geometry at the core midplane is shown in Fig 0re 4-1. Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 107', 110', 287', 290', 340', and 343' relative to the core cardinal area as shown in Figure 4-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1. The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core. nu,<omie io 6-2 i

from a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pad and the reactor vessel. In order to properly determine the neutron j environment at the test specimen locations, the capsules themselves must be included in the analytical model. 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 as well as to establish relative r4 dial distributions of exposure parameters (e(E > 1.0 Mev,) e(E > 0.1 Mes), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrewn fre.n the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/e(E > 1.0 MeV), within the pressure vessel geonietry, j The relative radial gradient information was required to permit the projection' of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations. The second set of calculations consisted of a seH es of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for the Cycle 1. irradiation; and established the means-to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects an,mmen ie 6-3 l

o of varying neutron yield per fission and fission spectrum introduced by the build up of plutonium as the burnup of individual fuel assemblies increased. 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: 1. Evaluate neutron dosimetry obtained from surveillance capsule locations. l 2. Extrapolate 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 the design of each new fuel cycle evolves, j The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 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 for light water reactor applications. In these analyses anisotopic scattering was treated with a P3 expansion of the cross-sections and the angular i discretization was modeled with an $ 0" der of angular quadrature. 8 The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loep plants. Inherent in the development of this reference core power distribution is the use of an out-in. fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2e uncertainty derived from 's statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2a me.wme io l 6-4 l

O level for a large number

1. fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.

All adjoint analyses were also carried out using an Sg order of angular quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, O geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, e (E > 1.0 MeV). Having the importance functions and appropriato core source distributions, the response of interest could be calculated as: R (r, 0) = /r #0 E !(r, 0, E) S (r, O E) r dr do dE where: R(r,0) e (E > 1.0 MeV) at radius r and azimuthal angle 0 = t I ! (r, 0, E) Adjoint importance function at radius, r, azimuthal' = angle 0, and neutron source energy E. S (r, O E) N6utron source strength at core location r, e and = energy E. Although the adjoint importance functions used in the Beaver Valley Unit 2 analysis were based on a response function defined by the threshold neutron flux (E > 1,0 MeV), prior calculations have shown that, while the implementation of low leakage leading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/, (E > 1.0 MeV) is insenu tive to changing core source distributions, In the application of these adjoint important functions to the Beaver Valley Unit 2 reactor, therefore, calculation of the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/, (E > 1,0 MeV) and e (E > 0,1 MeV)/e (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific e (E > 1.0 MeV) solutions from the individual adjoint evaluations, me,iosmo,o 6-5

• O The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design report for the first operating cycle of Beaver Valley Unit 2 (6).

The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of the pressure vessel and surveillance capsules are summarized in Figure 6-2. For comparison purposes, the core power distribution (design basis) used in the reference forward calculation is also illustrated in Figure 6-2. Selected results from the neutron transport analyses performed for the Beaver Valley Unit 2 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 irradiation period and provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall. In Table 6-1, the calculated exposure parameters (v (E > 1.0 MeV), e (E > 0.1 MeV), and dpa) are given at the geometric center of the two [ surveillance capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The design basis data derived from the forward calculation are provided as a point of reference against'which plant specific fluence evaluations can be compared. Simi hr data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the Cycle 1 plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself. 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, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are preaanted on a relative basis for each exposure parameter at several azimuthal locatienc. Exposure parameter distributions within the wall may be obtained by normaliziig the calculated or projected exposure at the vessel inner radius to the gtadient data given in Tables 6-3 through 6-5. mei.$.m. " 6-6 U

WESTfN8 HOUSE CLASS 3 For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by: el/4T(45')

• (199.95, 45') F (204.95, 45')

= 9 j47(45') Projected neutron flux at the 1/4T position where = 3 on the 45' azimuth e (199.95, 45') Projected or calculated neutron flux at the = vessel inner radius on the 45' azimuth. F (204.95, 45') Relative radial distribution function from Table = 6-3. Similar expressions apply for exposure parameters in terms of 9(E>0.1MeV)anddpa/sec. I The 00T calculations were carried out for a typical octant of the reactor. However, for the neutron pad arrangement in Beaver Valley Unit 2, the pad extent for all octants is not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron, pad extending from 30' to 45' (15') which produces the maximum vessel flux. Other octants have neutron pads extending 19' to 45' (26') which provide more shielding. 6.3 NEUTRON DOS! METRY The passive neutron sensors included in the Beaver Valley Unit 2 surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest {e (E > 1.0 Mev), e (E > 0.1 HeV), dpa). The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire m e a m io 6-7 r

1 form, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule. The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent flux level at the point of interest. Ra.ner, the activation or fission process is a measure of the integrated effect that the time-and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on 'the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: o The specific activity of each monitor, o The operating history of the reactor, i o The energy response of the monitor. 1 o The neutron energy spectrum at the monitor locatior.. o The physical characteristics of the monitor. The specific activity of each of the neutron monitors was determined using established ASTM procedures (7 through 20). Followiag sample preparation and weighing, the activity of each monitor was determinod by means of a lithium-driftedgermanium,Ge(Li),gammaspectrometer. The irradiation history of the Beaver Valley Unit 2 reactor during Cycle 1 was obtained from NUREG-0020 " Licensed Operating Reactors Status Summary Report" for the applicable period. l The irradiation history applicable to capsule U is given in Table 6-7. Measured and saturated reaction product specifi,c activities as well as measured full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6 6 and 6-7. Values of key fast neutron exposure parameters were derivod from the measured reaction rates using the FERRET least squares adjustment rode (21). The m e.,oi m ii. 6-8

FERRET approach used the measured reaction rate data and the calculated neutron energy spectruns at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra. In the FERRET evaluations, a log normal least-squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations, in general, the measured values f are linearly related to the flux e by some respon:e matrix A: f (s,a) = I A (8) Ie ") i i9 g g where i indexes the measured values belonging to a single data set s, g designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, { R I o g=9 egg g relates a set of measured reaction rates R to a single spectrum e by g g the multigroup cross section og. (In this case, FERRET also adjusts the g cross-sections.) The legnormal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties. In the FERRET analysis of the dosimetry data, the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code (22). This procedure was carried out by first expanding the a priori spectrum into the SAND-Il 620 group structure using a spline interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620 point spectrum was then easily collapsed to the group scheme used in FERRET. me,mme io 6-9

1 The cross-sections were also collapsed into the 53 energy group structure using SAND 11 with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/B-V dosimetry file. Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant, since these are not a major component of the cross section uncertainty treatment. For each set of data or a priori values, the inverse of the corresponding ) relative covariance matrix M is used as a statistical weight. In some cases, i as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is used: Mgg, = R2+R R,P g g gg, where RN specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional t uncertainties R specify additional random uncertainties for group g that g are correlated with a correlation matrix: Pgg, = (1 - 0) 699, + 0 exp [- (g o')2) 2r The first term specifies purely random uncertainties while the second term describes short-range correlations over a range r (0 specifies the strength c,f the latter term.) For the a priori calculated fluxes, a short-range correlation of r = 6 groups was used. This choice implies that neighboring groups are strongly correlated when e is close to 1. Strong long-range correlations (or anticorrelations) were justified based on infoAnation presented by R. E. Maerker (23). Maerker's results are closely duplicated when r = 6. For the integral reaction rate covariances, simple normalization and random l uncertainties were combined as deduced from experimental uncertainties. l p me.am io 6-10

O Results of the FERRET evaluation of the capsule V dosimetry are given in Table 6-9 The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 5.99 x 10 n/cm2 (E > 1.0 MeV) with an 18 associated uncertainty of 1 8%. Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the adjusted spectrum are provided in Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure. A summary of the measured and calculated neutron exposure of capsule U is presented in Table 6-12. The agreement between calculation and measurement falls within 1 12% for all exposure parameters listed. The calculated fast i neutron exposure (f (E > 1.0 MeV), t (E > 0.1 MeV), dpa) values agreed with the measurements to within 1-3% whereas, the thermal neutron exposure calculated for Cycle 1 exceeded the measured value by 12 percent. ? I Nedron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with the current (1.14 EFPY) exposure l derived from the capsule U measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EFPY). The calculated design basis exposure rates given in Table 6-2 were used to perform projections beyond the end of Cycle 1. In the calculation of exposure gradients for use in the development of heatup l and cooldown curves for the Beaver Valley Unit 2 reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14. In order to access RT VS' NDT fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations l 6' (1/4T) = f (Surface) { dpa (1/4T) } apa (Surface) i l nuve maio 6-11

I I' (3/4T) = i (Surface) { dpa (3/4T) ) dpa (Surface) Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Table 6-15 updated lead facters are listed for each of the Beaver Valley Unit 2 surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules, i I i i l l l l L m e.>oiv u io 6-12 i

1 j - 18.94 DEG. -19,72 DEG. ? i Y T r r f - 73.31 IN. 9 k h! \\h ~ % % % I l Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule nu.ame io 6-13 l

i i I i 9.93 9.77 DE5!6N BAl!$ 9.72 S.57 CYCLE 1 0.95 1.07 1.12 c.te 0.95 0.98 0.85 9.59 1.11 4 97 1.92 1.94 9.95 1.14 1.07 1.95 0.98 0.55 l.11 1.11 1.13 1.93 0,92 1.13 1.17 1.19 1.37 9.33 9.97 1.15 1.00 1.11 1.18 1,14 1.18 1.14 Figure 6-2. Core Power Distributions Used in Transport Calculations for Beaver Valley Unit 2 wu,ame io 6 14

o' ? i TABLE 6 1 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE SURVE!LLANCE CAPSULE CENTER DESIGN BASIS CYCLE 1 17' 20' 17' 20' 11 11 11 11 e(E>1.0MeV) 2.00 x 10 1.73 x 10 1.45 x 10 1.26 x 10 2 (n/cm.,,e) j 11 11 11 11 e (E> 0.1 MeV) 1.06 x 10 8.95 x 10 7.71 x 10 6.52.x 10 2 .(n/cm-sec) i -10 -10 -10 l dpa/sec 4.32 x 10-10 3.68 x 10 3.13 x 10 2.68 x 10 l 1 4 6 4

r l

nuvunn ie 6-15 L

TABLE 6-2 CALCULATED FAST NEUTP.ON EXPOSURE PARAMETERS AT -THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE DESIGN BASIS O' 12' 25' 30' 45' 10 10 10 10 10 9(E>1.0Mev) 6.31 x 10 4.18 x 10 3.49 x 10 2.73 x 10 1.89 x 10 2 (n/cm-sec) 10 10 10 10 10 9(E> 0.1Mev) 1.59 x 10 1.03 x 10 7,39 x 19 5.69 x 10 3.90 x 10 2 (n/cm-sec) -11 6.69 x 10'11 5.42 x 10 4.23 x 10'11-2.95 x 10~1$ -11 dpa/sec 1.01 x 10 CYCLE 1 SPECIFIC O' 12' 25' 30' 45' 10 10 10 10 10 e(E> 1.0Mev) 4.93 x 10 3.26 x 10 2.75 x 10 2.16 x 10 1,54 x 10 2 (n/cm.3,c) 9(E>OaMev) 1.25 x 1010 8.03 x 1010 10 10 10. 5.83 x 10 4.50 x 10 3.17 x 10 2 -(n/cm-sec) ~11 5.21 x 10'11 4.27 x 10'11 3.35 x 10'11 2.41 x 10-11 dpa/sec 7.90 x 10 3tlet1927H 10 g.}g 3

1 4 TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX-(E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL i Radius (cm) 0' 12' 20.5 ,30' _45' 199.95(1) 1.00 1.00 1.00 1.00 1.00 200.91 0.926 0.924 0.926 0.927 0.927 202.30 0.799 0.796 0.801 0.801 0.803 203.74 0.667 0.670 0.672 0.673 0.675 205.13 0.554 0.560 0.559 0.562 0.564 206.52 0.456 0.466 0.463 0.465 0.468 207.91 0.374-0.385 0.381 0.384 0.387- [ 209.30 0.306 0.317 0.312 0.315 0.318 210.69 0.249 0.260 0.255 0.258 0.261 212.07 0.202 '0.212 0.209 0.211 0.214 213.46 0.164 0.173 0.170 0.172 0.175 214.85 0.132 0.140 0.138 0.140 0.143 -216.24 0.105 0.113 0.111 0.113 0.116 217.63 0.0828 0.0904 0.0883 0.0900 0.0934 218.86 0.0650' O.0728 0.0708 -0.0726 0.0766 219.95(2) 0.0510 0.0586~ 0.0568 0.0586 0.0628 l l s i NOTES:

1) Bt.se Metal Inner Radius

2) Base Metal Outer Radius

? o w u en n u io 6-17

'o L' TABLE 6-4 RELATIVE. RADIAL OISTRIBUTIONS OF NEUTRON FLVX (E > 0.1 MeV) WITHIN THE PRESSURE VESSEL WALL = Radius (cm) 0' 12' 20.5 30' 45' 199.95(1) 1.00 1.00 1.00 1.00 1.00 200.91 1.00 1.00 1.00 1.00 1.00 202.30 0.965 0.970 0.978 0.978 0.984 203.74 0.905-0.917 0.926 0.925 0.935 205.13-0.839 0.858 0.867 0.866 0.878 206.52 0.771 0.79'S 0.803 0.803 0.817 207.91 0.704 0.732 0.740 0.739 0.754 209.30 0.639 0.668 0.676 0.676 0.693 [ 210.69 0.575 0.606 0.614 0.614 0.632 212.07-0.513 0.546 0.554 0.554 0.573 213.46 0.454 0.486 0.495 0.495 0.515 214.85 0.396 0.429 0.438 0.438 0.458 216.24 0.339 0.372 0.382 0.382 0.403 217.63 0.283 0.316 0.327 0.327 0.350 218.86 0.230 0.265 0.277 0.279 0.303 219.95(21 0.187 0.222 0.235 0.238 0.262 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius

'/ mwowie io 6-18

'O TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0' 12' 20.5_ 30' 45' I1) 199.95 1.00 1.00 1.00 1.00 1.00 200,91 0.939 0.939 0.938 0.938 0.938 202.30 0.841 0.841 0.837 0.835 0.836 z iM.74 0.739 0.744 0.732 0.731 0.732. 2 % 13 0.648 0.657 0.640 0.639 0.641 206.52 0.567 0.579 0.558 0.557 0.560 207.91 0.495 0.509 0.487 0.486 0.490 209.30 0.432 0.447 0.425 0.423 0.428 ? 210.69 0.375 0.391 0.370 0.369 0.374 212.07 0.325 0.341 0.321 0.320 0.326 213.46 0.280 0.296 0.278 0.277 0.283 214.85 0.239 0.255 0.240 - 0.239 0.245 216'.24 0.202 0.217 0.204 0.204 0.211 217.63 0.166 0.182 0.172-0.171 0.180 218.86 0.134 0.152 0.145 0.145 0.155 219.95(2) 0.108 0.127 0.122-0.123 ~0.133 NOTES:

1) Base Metal Inner Radius

2) Base Metal Outer Radius 1

I L auvoame io 6-19 0

WESTINGHOUSE CLASS 3 TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONITORS l l Reaction Target Fission Monitor of Weight Respon:;e Product Yield Material Interest Fraction Range Half-Life (%) Copper Cub (n.a)Co60 0.6917 E> 4.7 MeV 5.272 yrs 1 Iron Fe54(np)Mn54 0.0582 E> 1.0 MeV 312.2 days Nickel NiS8(n.p)CoS8 0.6830 E> 1.0 MeV 70.90 days Uranium-238* U238(n,t)C:137 1.0 E> 0.4 MeV 30.12 yrs 5,99 i i Neptunium-237* Np237(n,f)Cs137 1.0 E> 0.08 MeV 30.12 yrs 6.50 Cobalt-Aluminum

• CoS9(n r)Co60 0.0015 0.4ev<E< 0.015 MeV 5.272 yrs o

Cobalt-Aluminum CoS9(n,r)Co60 0.0015 E<'O.0i5MeV 5.272 yrs

• Denotes that monitor is cadmium shielded.

mu.mme io 6-20

TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE U Irradiation P-P Irradiation Decay 3 3 Period (MW ) P,f, Time (days) Time (days) p t 8/87 524 .198 15 625 9/87 430 .162 30 595 10/87 1530 .577 31 564 11/87 719 .271 30 534 12/87 2511 .947 31 503 1/88 2214 .835 31 472 2/88 1363 .514 29 443 3/88 2637 .994 31 412 i 4/88 2523 .951 30 382 5/88 2639 .995 31 351 -6/88 2493 .940 30 321 7/88 2528 .953 31 290 8/88 2397 .904 31 -259 9/89 2504 .944 30 229 10/88 2530 .954 31 198 11/88 2640 .996 30 168 12/88 2639 .995 31 137 1-/89 2504 .944 31 106 2/89 1516 .572 28 78 3/89 1502 .566 17 61 NOTE: Reference Power = 2652 MW t mwows. io 6-21

O TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated' Reacticn Monitor and -Activity Activity Rate Axial Location (dis /sec am) (dis /see-am) (RPS/ NUCLEUS) Cu-63(n,a)Co-60 4 5 Top 7.54 x 10 5.18 x 10 4 5 Middle 7.17 x 10 4.92 x 10 4 5 Bottom 6.70 x 10 4.60 x 10 4 5 Average 7.14 x 10 4.90 x 10 7.48 x 10'17 Fe-54(n.p) Mn-54 [ 6 6 Top 2.77 x 10 5.33 x 10 6 6 Middle 2.53 x 10 4.86 x 10 6 6 Bottom 2.44 x 10 4.69 x 10 6 6 -15 Average 2.58 x 10 4.96 x 10 7.90 x 10 Ni-58 (n,p) Co-58 i 7 7 Middle 3.49 x 10 7.80 x 10 7 7 Bottom 3.37 x 10 7.53 x 10-7 7 Average 3.43 x 10 7.67 x 10 1.09 x 10'14 U-238 (n,f) Cs-137 (Cd)- 5 6 Middle 2.10 x 10 7.50 x 10 4.95 x 10'14 mu.mma io 6-22

l c m. TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES cont'd Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /see om) (dis /see gm) (RPS/ NUCLEUS) l Np-237(n,f)Cs-137(Cd) 6 7 13 Middle 1.99 x 10 7.11 x 10 4.30 x 10 Co-59 (n,r) Co-60 7 8 Top 1.50 x 10 1.03 x 10 Middle 1.58'x 10 1.09 x 10 f 7 0 7 7 Bottom 1.43 x 10 9.82 x 10 7 8 Average 1.50 x 10 - 1.03 x 10 6.73 x 10 Co-59 (n,r) Co-60 (Cd) 6 7 Top 8.57 x 10 5.88 x 10 6 7 Middle 8.69 x 10 5.97 x 10 6 7 Bottom 9.17 x 10 6.30 x 10 6 7 -12 Average 8.81 x 10 6.05 x 10 3.95 x 10 'l 1 2n.,,em.. io 6:23 9

i q. TABLE 6-9

SUMMARY

OF NEUTRON 00SIMETRY RESULTS TIME AVERAGED EXPOSURE RATES l 2 10. 9 (E> 1.0 MeV) (n/cm -sec) 1.53 x 10 gg 2 11 e (E> 0.1 MeV) (n/cm -sec) 7.21 x 10 1 15% -10 dpa/sec ~ 3.05 x 10 gig 2 10 9(E> 0.414 eV) (n/cm -sec) 3.87 x 10 29% INTEGRATED CAPSL'LE EXPOSURE ? 2 18 4 (E> 1.0 MeV) (n/cm ) 5.99 x 10 8% 2 19 6 (E> 0.1 MeV) (n/cm ) 2.82 x 10 1 15% l i -2 -dpa 1.19 x 10 11% 2 18 + (E> 0.414 ev)'{n/cm ) 1.52'x 10 + 29% 1 NOTE: Total Irradiation Time = 1.24 EFPY animn. io 6-24

l o TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER \\ Adjusted Reaction Measured Calculation C/M Cu-63(na)Co-60 7.48x10'17 7.62x10'17 1.02 -15 -15 Fe-54(n,n)Mn-54 7.90x10 7.84x10 0.99 Ni-58 (n.p) C1-58 1.09x10'14 1.08x10'14 0.99 ~14 ~14 U-238(n.f)Cs-137(Cd) 4.95x10 4.62x10 0.93 ~13 -13 Np-237 (n,f) C -137 (Cd) 4.30x10 4.67x10 1.09 Co-59 (n r) Co 60 (Cd) '6.73x10-12 -12 6.70x10 1.00 o -12 -12 Co-59 (n,r) Co 60 3.95x10 3.95x10 1.00 i i' 4 4 - me,mme io 6-25

s TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE SURVEILLANCE CAPSULE CENTER Energy Adjusged Flux Energy AdjusgedFlux Group (Nev) (n/cm -sec) Group (Mev) (n/cm -sec) 1 7 -3 10 1 -1.73x10 1.04x10. 28 9.12x10 2.71x10 1 7 10 2 1.49x10 2.33x10 29 5.53x10'3 3.20x10 1 7 10 3 1.35x10 8.87x10 30 3.36x10'3 1.09x10 1 8 10 4 1.16x10 2.03x10 31 2.84x10'3 1.10x10 1 8 10 5 1.00x10 4.58x10 32 2.40x10'3 1.12x10 0 8 10 6 8.61x10 7.98x10 33 2.Q4x10'3 3.32x10 0 9 10 7 7.41x10 1.87x10 34 1.23x10'3 2.85x10 0 9 ~4 10 8 6.07x10 2.72x10 35 7.49x10 2.53x10 0 9 10 9 4.97x10 5.87x10 36 4.54x10'4 2.44x10 0 9 -4 10-10 3.68x10 7.88x10 37 2.75x10 2.60x10 7 0 10 10 11 2.87x10 1.64x10 38 1.67x10'4 3.11x10 12 '2.23x100 10 2.30x10 39 1.01x10'4 2.89x1010 0 10 -5 10 13 1.74x10 3.26x10 40 6.14x10 2.81x10 0 10 -5 10 14 1.35x10 3.73x10 41 3.73x10 2.69x10 0 10 -5 10 15-1.11x10 7.00x10 42-2.26x10 2.53x10 16 8.21x10'1 8.20x10 43 1.37x10 2.37x10 10 -5 10 17 6.39x10'1 8.58x10 44 8.32x10 2.15x10 10 -6 10 18 4.98x10'1 6.61x10 45 5.04x10 1.84x10 10 -6 10 19 3.88x10'1 9.55x10 46 3.06x10 1.58x10 10 -6 10 ~1 10 -6 10 20 3.02x10 8.63x10 47 1.86x10 1.35x10 -1 10 -6 9 21 1.83x10 8.99x10 48 1.13x10 9.82x10 -1 10 -7 10 22 1.11x10 7.50x10 49 6.83x10 1.04x10 -2 10 -7 10 23 6.74x10 4.75x10 50 4.14x10 1.19x10 -2 10 9 24 4.09x10 2.45x10 51 2.51x10'7 8.96x10 -2 10 -7 9 25 2.55x10 3.77x10 52 1.52x10 6.26x10 -2 10 -8 10 26 1.99x10 1.41x10 53 9.24x10 1.16x10 -2 10 27 1.50x10 1.45x10 NOTE: Tabulated energy levais represent the upper energy of each group, i a u. m u n io 6-26

1 TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE V Calculated Measured CJ 2 18 18 - t(E> 1.0 MeV) (n/cm ) 5.6ti x 10 5.99 x 10 0.95 2 19 19 f(E>.0.1 MeV) (n/cm ) 3.02 x 10 2.82 x 10 1.07 dpa 1.23 x 10'3 1.19 x 10 1.03 ~3 2 18 18 f(E> 0.414 eV) (n/cm ) 1.05 x 10 1.52 x 10 0.69 ? F n u voe n e io 6-27

is f Ja., TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE AZIMUTHAL ANGLE 0' 15* 25.(a) _-35* 1.14 EFPY 45' 18 18 18 17' 17 +(E> 1.0 MeV)- 2.04 x 10 1.35 x 10 1.13 x 10 8.93 x 10 6.35 x 10 2 (n/cm) 18 18 18 18 18 4(E> 0.1 MeV) 4.56 x 10 2.94 x 10 2.13 x 10 1.65 x 10 1.16 x 10 2 (n/cm) -3 -3 -3 ~3 -3 dpa 4.11 x 10 2.72 x 10 2.20 x 10 1,73 x 10 1.21 x 10 16.0 EFPY i 19 19 19 19 18 +(E> 1.0 MeV) 3.14 x 10 2.08 x 10 1,74 x 10 1.36 x 10 9.44 x 10 2 (n/cm) 19 19 19 +(E> 0.1 MeV) 7.86 x 10 5.09 x 10 3.66 x 10 2.82 x 1019 1.93 x 1019 2 (n/cm) -2 -2 -2 -2 -2 dpa 5.12 x 10 3.39 x 10 2.74 x 10 2.14 x 10 1.50 x 10 32.0 EFPY 19 19 19 19 19 - +(E> 1.0 MeV) - 6.33 x 10 4.19 x 10 3.50 x 10 2.74 x 10 1.90 x 10 2 (n/cm) 20 20 19 19 19 +(E> 0.1 MeV) 1.59 x 10 1.03 x 10 7.39 x 10 5.69 x 10 3.90 x 10 2 (n/cm ) -1 -2 -2 -2 -2 dpa 1.02 x 10 6.77 x 10 5.48 x 10 4.28 x 10 2.99 x 10 (a) Maximum point on the pressure vessel - nuvoune io 6-28 1

l. 4 TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE'IN THE GENERATION OF flEATUP/C00LDOWN CURVES l 1 l 16 EFPY l NEUTRON FLUENCE (E> 1.0 MeV) SLOPE dpa SLOPE 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 19 19 18 19 I9 I9' 0* 3.14 x 10 1.79 x 10 4.08 x 10 3.14 x 10 2.07 x 10 7.41 x 10 19 I9 18 19 I9 18 12* 2.08 x 10 1.19 x 10 2.87 x 10 2.08 x 10 1.39 x 10 5.24 x 10 I9 18 18 19 I9 1d 20.5* 1.74 x 10 9.99 x 10 2.37 x 10 1.74 x 10 1.13 x 10 4.12 x 10 I9 18 18 19 18-18 30* 1.36 x 10 7.83 x 10 1.88 x-10 1.36 x 01 8.85 x 10 3.21 x 10 18 18 18 18 18 I9 45* 9.44 x 10 5.46 x 10 1.33 x 10 9.44 x 10 6.16 x 10 2.29 x ?.0 32 EFPY NEUTRON FLUENCE (E> 1.0 MeV) SLOPE dpa SLOPE 2 2 (n/cm ) (equivalent n/cm ) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 19 I9 18 19 19 19 O' 6.33 x 10 3.60 x 10 8.23 x 10 6.33 x 10 4.18 x 10 1.49 x 10 19 I9 18 19 I9 19 12* 4.19 x 10 2.40 x 10 5.78 x 10 4.19 x 10 2.80 x 10 1.06 x 10 I9 I9 18 I9 19 18 20.5* 3.50 x 10 2.01 x 10 4.76 x 10 3.50 x 10 2.28 x 10 8.30 x 10 19 I9 18 19 I9 18 30* 2.74 x 10 1.58 x 10 3.78 x 10 2.74 x 01 1.78 x 10 6.47 x 10 I9 I9 18 19 19 18 45* 1.90 x 10 1.10 x 10 2.68 x 10 1.90 x 10 1.24 x 10 4.62 x 10 (a) Maximum point on the pressure vessel 3958s/092789 to .c-

_ _.. _ ~ TABLE 6-;5 UPDATED LEAD FACTORS FOR 'EAVER VALLEY UNIT 2 SURVEILLANCE CAPSUL'ES Capsule Lead Factor V 2.94(a) X 3.17 V 3.17 l Z 2.74 W 2.74 Y 2.74 (a) Plant specific evaluation n u. m :2 io 6-30

,1 - SECTION 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recommended for future capsules to be removed from the Beaver Valley Unit 2 reactor vessel: Estimated Vessel Capsule Removal Fluence-Location Lead i Capsule (dog) Factor Time (a) (n/cm ) 2 U -34.3 2.94 1.24 5.99 x 1018(b) V 107-3.17 6 3.76 x 1019(C) W 110 ~2.74 12 6.50 x 1019(d) 19 X 287 3.17 18 11.30 x 10 ? Y 290 2.74 Standby Z 340 2.74 Standby a) Effective full power years from plant startup b) Actual fluence c) Approximate fluence at vessel 1/4 thickness at end of life (32 EFPY) d) Approximate fluence at vessel inner wall at end of life (32 EFPY) i 20W/002789 to -\\ L

SECTION 8 REFERENCES

1. Davidson, J. A., Yanichko, S. E. "Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-9615, Ncvember 1974.

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

3. Regulatory Guide 1.99, Revision 2

" Radiation Embrittlement of Reactor Ve:sel Materials," U.S. Nuclear Regulatory Commission, May, 1988.

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

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

6. S. T. Lesho, et al., "The Nuclear Design and Core Physics Characteristics of the Alvin W. Beaver Valley Unit 2 Nuclear Power Plant - Cycle 1,"

WCAP-11338, November 1986. (Proprietary)

7. ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surve,illance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

n,mme so g.1 ,_------%1.

8. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section l

12, Amer _ican Society for Testing and Materials, Philadelphia, PA, 1984.

9. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron L

Exposures in Ferritic Steeis in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, l; Philadelphia, PA, 1984. L

10. ASTM Designation E706-81a, "' standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillar.ce Standard", in ASTM Standards, Section 12, 3

American Society for Tescing and Materials, Philadelphia, PA, 1984,

11. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984, f

i

12. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

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

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

15. ASTM Designation E264-82, " 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.

t 3954s/092786 10 8-2

16. 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.

17. ASTM Designation E523-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Cooper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

18. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.

19. ASTM Designation E705-79, " Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237", in ASTM Standards, Section 12, American Society for Testing and Meterials, Philadelphia, PA,1984.

20. ASTM Designatwr. E1005-84, " Standard Method for Application and Analysis !

{ of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, i Philadelphia, PA, 1984.

21. F. A. Se%ittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

22. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative I

Method of Neutron Flux Soectra Determined by Foil Activation, AFWL-TR-67-41, Vol. 1-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.

23. EPRI-NP-2188, " Development and Demonstration of an Advaner '.aethodology for LWR Desimetry Applications", R. E. Maerker, et al.,1981.

\\

• **oS m o 8-3

APPENDIX A HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION A1. 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 vessel is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced ART RTNDT.is NDT. designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or 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 increases as the material is exposed to fast-neutron radiation, NDT i I Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ART due to the radiation exposure associated with that NDT time period must be added to the original unirradiated RT The extent of NDT. the shift in RT is enhanced by certain chemical elements (such as copper NDT and nickel) present in reactor vessel steels. The Nucl. ear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)lA-1) The value, "f", given in Figure A-1 is the calculated value of the neutron fluence at the location of interest (inner surface,1/47, or 3/4T) in the vessel at the location of the postulated defect, n/cm (E 19 > 1.0 MeV) divided by 10 The fluence factor is determined from Figure A-1. A2. FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan [A-2) The pre-irradiation fracture-toughness properties of the Beaver Valley Unit 2 reactor vessel are presented in Table A-2. nu. mms to A-1

o A3.. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K;, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, f r the metal temperature at that time. K is obtained from the IR reference fracture toughness curve, defined in-Appendix G to the ASME CodeIA~33 The KIR curve is given by the following equation: KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)) (1) where KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT t i Therefore, the governing equation for the heatup-cooldown analysis is defined ' in Appendix G of the ASME Code (A-3) as follows: .CKyg + KIT

• KIR (2) where KIM = stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RT f the material NDT C

= 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical m e.< cones io A-2

4.- .s A,any time during the heatup or cooldown transient, K is determined by_ IR 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 stress intensity factors are obtained and, from these, the allowable pressures are calculated, i For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From i these relations, composite limit curves are constructed for each cooldown ratei of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K ip exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperat'ure at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various nu,mme io A-3

.s intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K;g for the 1/4 T crack during heatup is lower

than the K f r the 1/4 T crack during steady state conditions at the same IR time coolant temperature.

During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR s do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw f is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. .The'second portion of the heatup analysis concerns the calculation of 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, 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 analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by point comparison of the steady-state and finite heatup rate data. At any given temperature, the mamme " A-4

WESTINGHOUSE CLASS 3 o allowable pressure is taken to be the lesser of the three values taken from-the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup 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. Finally, the 1983 Amendment to 10CFR50(A4) has a rule which addresses the metal temperature of the closure head flange and vessel flange regions.. This rule states that the metal temperature of the closure flange 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-2 indicates that the limiting RT f 0*F occurs in the vessel NDT flange of Beaver Valley Unit 2, so the minimum allowable temperature of this region is 110'F. These limits are less restrictive than the limits shown on Figures A-2 through A-5. A4. HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown af the primary Reactor Coolant -System have been calculated using the methods discussed in Section A3. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures ' A-2 through A-5. This is in addition to other criteria which must be met before the reactor is made critical. The letk limit curve shown in Figures A-2 and A-3 represents minimum temperature requirements at the leak test pressure specified by applicable codes [A-2,A-3) The leak test limit curve was determined by methods of References A-2 and A-4. Figures A-2 through A-5 define limits for ensuring prevention of nonductile failure for the Beaver Valley Unit 2 primary reactor coolant system. 3Hes/09274910

i A5. ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2IA'1I the adjusted reference temperature (ART) for each material in the beltline is given by the following expression: ART = Initial RTNDT + ARTNDT + Margin (3) Initial RT is the reference temperature for the unirradiated material as NOT def_ined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code._ If measured values of initial RT f r the material in HDT question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class. ART is the mean value of the adjustment in reference t a erature caused NOT by irradiation and should be calculated as follows: NOT_= (CF)f(0.28-0.10 log f) (4) ART ~ To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth. I(depth X)

• Isurface (5) where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface. The resultant fluence is then put into Equation (4) to calculate ARTNOT at the specific depth.

CF (*F.) is the chemistry factor, obtained from Reference A-1. Beltline region materials of Beaver Valley Unit 2 are considered for the limiting material. Limiting material is'found to be intermediate Shell plate 89004-1. The calculation of ART for the limiting material is shown in Table A-1. This calculation was used to develop the Beaver Valley Unit 2 heatup and cooldown curves that are shown in Figures A-2 through A-5. I m w o m esio A-6

e..

o TABLE A-1 s

CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING BEAVER VALLEY UNIT 2 REACTOR VESSEL MATERIAL - INTERMEDIATE SHELL PLATE B9004-1 Regulatory Guide 1.99 - Revision 2 [ 5 EFPY 10 EFPY Parameter 1/4 T 3/4 T 1/4 T 3/4 T ChemistryFactor,CF-('F) 44 44 44 44 19 n/cm )(a) 0.61 0.23 1.22

0.47

( 2 Fluenco, f (10 Fluence Factor,-- ff 0.86 0.61 1.06 0.79 i i ,j ARTNDT = CF x ff ('F) 37.9 26.7 46-.4 34.6 NDT, I ( F) (b) Initial RT 60.0 60.0 60.0 60.0 ' Margin,M(*F)(c) 34.0 26.7 34 34, J s. Revision 2 to Regulatory Guide 1.99 AdjustedReference-Tempercture, 132 113 140 129 ART = Initial RTNDT + ARTNDT + Margin 19 2 (a) Fluence, f, is based upon fsurf (10 n/cm, E>1.0 Mev) = 6.33 at 32 EFPY, The Beaver Valley Unit 2 reactor vessel wall thickness is 8.03 inches at the beltline region. (b) The initial RTNOT (I) values for the plates-and welds are measured values. (c) Margin is calculated as, M = 2 2,,2 The standard c 1 deviation for the initial RTNDT margin term (og) is assumed to be zero since the . initial RTNDT values were obtainetl from conservative (i.e., " upper bound") test results. However, the standard dsviation for ART NDT' a, is 28'F for welds and 17'F for base metal, except that a need not exceed 0.50 times the mean value of 3 l ARTNDT* nu.e..m. io A-7~ L l

l _. ~ ~, I l l .[ l' l E "bl00e,l ll all U!Uh f,m{mm -.. lIlllh, yh,,lll,h i - l s ,n, inou llMl?' "y ll HHIDa**m e ii n. m o"" tuu m. a n n"n & g-1 og"l# - n min i un u l. nnllt Iu, ll C. !!!f:S l - lituin t l lll11 l ' ~ flHit b ' liffifHF i - l 1.4 UhlY I Ngtlfl[:c L: lilHHil 7 " p-Un.i!:L. }lb u l l lllllh " l IllHA ; j. !Hm. Iftll'ih. h I ll ;i *. - ililllHillFF!.au lilill! y, lilll!!il ; k l b I lR l14 IINL'l.. HlHjllllHlf"" llINI ':- IlllflH! o2 i l ijf lh.- l I!ll - lil!Zildl!. L lil!! L lill!U :. l E F IE l M llillit a 14 IlIllii Mii+ E Y $1 llb IHill iill h. !W IN-lllll W ti ll0 ll!E Inr NL E llllO $l llLU ll S ' ~, IlH J Hii! ~ t; P I! 4 ihM:i j ( ! ![ - ll{ l.%. l1114: Il!Hi I !!: 1 IM 1 1 2 l 14 h kl % _ l Un.. H!!# r Ikh 4 (i14m: I i Il h !! :. 13 I l lk' - IL. T !!h II!N IllllllW II L L L iW. I li j l lN . :1:i i. ';'. I i :: - Hb: . IR : i llll", C": j!hi "- l:I L l: 0 lN " N' l [. l-[l l} f h.. h lb- -- = ' - - ~ = l HiI! i Y EL lib f g. 1 u - on: [9 n + 1 u lt ; ' j lh, Inn l 1 } H.i,.. i ' 10.., ' uf0 I p in..' a l 1. ' h? j 4, N'm i H. I' t:E. l !i.. 2 3 4 5 6 789 2 3 4 5-6 7891 3 4 5 6 7891 IS" 19# iga F l Ruonce, n/ con (E > 1 ReeWI 'l s I Figure A-1. FluenceFactorforUseintheExhe~ssionforARTNOT 3%8s/03f789 to l ? .~. .,_1.,

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: '!NTERMEDIATE SHELL PLATE B9004-1 INITIAL RTNDT: 60*F RT AFTER 5 EFPY: 1/4T, 132*F NDT 3/4T, 113*F CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 5 EFPY, CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 15 M iwo i i i.,, i, ti I '!l'

ll Leak Test

'7 Limit l ll 1 i e 3 i ( ,iI ,Ii i ..ec20 i f . I ii iif [A .! J e i ine + ft 1 ,i i ie i if 1 .Ii af! i .I i ri Ji f; c0'0 , 1 _ f, fi, i, i j i i ! t I 'i

• 7 i.

i ! . i , i 4 ,. i i,, f. i;, i,., i Heatup Rates i, i i ; i i gg Up To 1 I i i i i l! l 60'F/Hr, [; /; l ,i i ii 1 i e \\ e e } } } r%o 1500 i i i s i, ,i f, i n r.i r, ii i i i / L i e J i .I e w i i i f p 1250 l l l l l lf j,' ll,

p 2

i 4 f 1 i O 0 Unacceptable i fi , i '] -j s. I i e i i i Operation I E 1000 !! t i f. t i r i/ o i 6 y ? I i. i. Ii i . 4 i ,i1 . i a Q i e / i i 6 l l,, i i 2 <50 l,

d. ~!

!/ l lCriticalityLimit iBased on Inservice ; j l i O /i i 3 i lHydrostatic Test i /' ' ', i i l MI l l Temp.(271'F)for 500 ihe Service Period 6 ,i i i i i ; Acceptable --up to S EFPY ii i ,,i, ii Operation

~

i-250 l' l, l ji I i , i i i i i i 4 i 6 i 4 i i i i' i ' i ' 0 O 50 100 150 20 :) 15) 30C 350 400 450 $) INDif t0 TEMPERATURE (CEG.F) Figure A-2. Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 5 EFPY me,mn in A-9 l l

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 INITIAL RTND7: 60'F RT AFTER 10 UPY: 1/4T, 140'F NOT 3/4T, 129'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY. CONTAINS MARGIN OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. - +- LalIE E R 1 A i 1. I I 'l 1 1 , i 1 I i I i, i I . I i leak Itst lI !I ..g,i I i ES ' i I ! i (jmjg J i f: ' I1 i e i i ! i li I I: i i i e i i!! I ' e Il 'It I; II iI J ! i + i a i I i1

1 I

3102 - i, i I! i I!, i i i li I! i ri Ii i i i 1 1! 4 !. e., n 11 4 fi, Ii i i a i . /~ I,, iI I i i 8 !I t I i + i 4 i i ! Ii ! I! i , + i 7 1500 l Heatup Rates (( Il! 'i l i U io -( 60gF/Hr r if i i i i ,f i v !1 /! i u 125) l i l [ / l! : l 4 i i i i iii i; N l-Unacceptable '/ / Il,'! ', ,l',,, W Operation f. iii i i c. 1( 0 -) , i i i i f, i . i i 1 t i , ii _r f. i,, y i i ii ij . i i i i a f i i, 0 .5) i / 'i i f ,. t i o .f i i i i 2 l i ..f i I / CriticalityLimiti, e 50,> i Based on Inservice'" i ! i - i Hydrostatic Test Temp. (200*F) for! i Acceptable the Service Perioe '5) Operation _i ! i vp to 10 EFPY i i-i 3 i , i i ,j i i i l i 1 ) iii i 4 i 4 i i i i it t 'O 50 10( lit 2.0 250 560 150 400 4 51 500 INDICATCo TCWPCRATURC (DCC.F) Figure A-3. Beaver 1.tiiey Unit 2 Rector Coolant System Heatup Limitations Appl 1 Cable for the First 10 EFPY m e,e m iein A-10

NATERIAL PROPERTY BASIS CONTROLLING MATEP!st.,: INTERMEDIATE SHELL PLATE B9004-1 INITIAL RTNDT: 60'F RT AFTER 5 EFPY: 1/4T, 132'F NDT 3/4T, 113'F CURVES APPLICABLE FOR C00LDOWN RATES VP T0100 *F/HR FOR THE SERVICE PERIOD UP TO 5 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 15:0 _ur,. i I i i i I i i, I e 'I cr. t. 0 1 11 t I I I 1 i F. 1 J'0" j i i ii i I j a. i i I i i i j 6 y I. e 1750 l /! l=- t i g i + I i i i e i i i f d i t i e t l i ii i If i i ! t i i + i i i i fi ) I t pgo 15(0 i j j ,,i i i 1 ie i ! !I r i i g i i i a i + + ! T i r gi,h, w l i i ! i iI i i + w ! !i O -- <- *>e -'- l l ,,/ l l l f 1 h It' F .i ./ i i i,, Q i i j I i l i ! j y ii 1. i 2 E I i r 10(0 ,! f i I I I I i i i e Unacceptable s! t ' e F,.iA. w Operation 'ii '+ s.4 ' kr.ceptable l _+ -4[._ v 7g o H -t- ,,j; Operation g m- - ---t- -y = ,j ,,i !. b i !YT i _._ ~ '~ ~ 7 -~~~ N i G Cooldown ism e i i s gates

, j,A,

,i ty i i i SC O - *F/Hr g, ._3 s f i i i ( 1 rh.,T-T ; -; 1 P r f i 0 yr i r i ? 20 i i 250.p 40 ) Z, y' 60 t,.m ,ir q s " $N ~, iki t 'I J. 1 i i ,.i 3 3 i I ' ! I i + + 1. t i 1_ i ii i I'l 0 :' 5). 100 150 200 250 40 ?,50 400 450 5 0 :- i INolCATCo TCWPERA TURE (OCG.F) i Figure A-4. Beaver Valley Unit 2 Reactor Coolant System Cooldown l Limitations Applicable for the First S EFPY l i nuvo.nu a g.33

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 i INITIAL RTNDT: 60'F l RT AFTER 10 EFPY: 1/4T, 140'F NOT 3/4T, 129'F CURVES APPLICABLE FOR COOLDOWN RATES UP T0100 *F/NR FOR THE SERVICE PER TO 10 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. I'0 u+ : r m i i t i j i I i ' r > _1 215:' l'4 !,! l,9, siii ~ i i ! I ! l J i, ,h i a i >c. i 1-i - ' ),;, ' i i i .i i i i i, ,1 ,i i i 4 i i i i r-i! i - i i t i ii I e i is i i 3't, i i 1 e i i i i i i e i i i i r i i i 1 1 I l ,,i j / i g tc00 l ,' i ,ii ii, 7 l l, ,[ l l 'll E i ~ ri i i i i i ; ;, w 1250 Unacceptable r i t l Operation .;i r i 4 i 1 o e J i j i i i i i r Acceptable g. i c. 1000 '..r E Operation ! i i i i i,i o r i i j !F i ! i I i i i i r i i, u 750 o o g l l sm ~ Cooldown' $7 ;j ~ r m l, con Rates .ece e 'F/Hr 'vr #

ar' 2 r 0

3 i 20 .i 250 40 + i' g '. 100 l l l l 't t t i i - - i ! ! - i 9 i O 50 100 150 200 250 300 350 400 450 5)0 INolCATED TEW#ERATURC (OCC,r) Figure A-5. Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 10 EFPY me.wn.so A-12 i \\

l WEST 8NGHOUSE CLASS 3 0 TABLE Ae2 BEAVER VALLEY UNIT 2 REACTOR VESSEL MATERIAL PROPERTIES (UNIRRADIATED) Shelf Energy Haterial Cu Ni NDTT RT NMWD (b) Component Code No. (%) (%) ('F) ('klT (ft-lb) Closure Head Dome B9008-1 0.13 0.51 -20 -10 137 Closure Head Flange B9002-1 0.74 -10 -10 136 Vessel Flange 89001-1 0.73 0 0 132.5 Inlet Nozzle B9011-1 0.88 0 0 104 Inlet Nozzle B9011-2 0.88 10 10 115 Inlet Nozzle B9011-3 0.84 20 20 122 Outlet Nozzle B9012-1 0.71 -10 -10 137 Outlet Nozzle B9012 2 0.74 -10 -10 121 Outlet Nozzle B9012-3 0.68 -10 -10 112 Nozzle Shell B9003-1 0.13 0.61 -10 50 91. Nozzle Shell B9003-2 0.12 0.58 0 60 79.5 Nozzle Shell 89003-3 0.13 0.61 -10 50 97.5 Inter. Shell B9004-1 0.07 0.53 0 60 83 Inter. Shell B9004-2 0.07 0.59 -10 40 75.5 Lower Shell B9005-1 0.08 0.59 -50 28 82 Lower Shell B9005-2 0.07 0.58 -40 33 77.5 Bottom Head Torus B9010-1 0.15 0.49 -30 -4 97 Bottom Head Dome B9009-1 0.14 0.53 -30 -25 116 Inter & Lower Shell G1.42(a) 0.08 0.07 -30 -30 144.5 Vert. Weld Seams and Girth Seam (a). Weld wire heat 83642 and Linde 0091 Flux lot 3536 (b) Normal to major working direction i l me.,eme io A-13

A6. REFERENCES A-1 Regulatory Guide 1.99, Revision 2 " Radiation Embrittlerent of Reactor Vessel. Materia's," U.S. Nuclear Regulatory Commission, May, 1988. A-2. " Fracture Toughness Requirements," Branch T6chnical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981. A-3. ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G. Protection Against Nonductile failure," pp. 558-563, 1986 l Edition, American Society of Mechanical Engineers, New York, 1986. l A-4. Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughness l l Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No. 104, May 27, 1983. l l l l l wavamiin A 14 l

j'\\} -f.-. l. t r 4 N ^ _3 g -1.i i t.' h i it .'. 'g,'; $ 4 3 4 jum

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43

m-Dmw60 f /HR HE Af tsp CtJRVE REG-GUIDE 1.99.REV.2 09/20/89 THE FOLLOtd!NG DAT A tfERE C ALCUL ATEDf 04 THE INSERVICE HYDROSTATIC LE AK TEST. MINIMUI4 INSERVICE LEAM TEST T EasPER A TURE ( 5.000 EFPV) PRESSURE (PSIB T EtePE R A TURE ( DEG. F ) 2000 250 2485 271 PRESSURE PRESSURE STRESS 1.5 M om (PSI) (PSI) ( P">I SQ.RT.IN.B 3:= 2000 21216 86274 e C** 2485 26211 '107488 I => 1

O*eW60 F/HR HEATOP CURVE C.EG. GUIDE 1.99 REv.2 09/20/89-CONPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2 HEATUP RATE (S) ( DEG. F /HR ) 60.0 1RRADIATION PERIOD = 5.000 EFP VEARS FLAW DEPTH * (t-ADWIN)T INDICATED INOICATED INDICATED INDICATED INDICATED INDICATED T E p4PE R A TURE. PRESSURE TEMPERATURE PRESSURE T E GEPE R A TURE PRESSupE ( DEG. F ) (PSI) (DEG.F ) (PSI) (DEG.F) (PSI) 1 85.000 M 17 165.000 612.09 32 240.000 1836.19 2 90.000 5+9-46 18 170.000 631.73 33 245.000 1195.25 3 95.000 568-16 tilfib9 175.000 652.90 34 250.000 1258.92 4 100.000 50+-99 20 180.000 675.96 35 255.000 1327.03 5 105.000 48e-S7 21 185.000 700.70 36 260.000 t399.95 6 110.000 497.95 22 190.000 727.59 37 265.000 1478.19 7 115.000 499.77 23 195.000 756.37 38 270.000 1562.01 8 120.000 503.69 24 200.000 787.55 39 275.000 1651.79 9 125.000 509.60 25 205.000 820.96 40 280.000 1747.81 10 130.000 517.12 26 210.000 856.83 41 285.000 9850.45 11 135.000 526.32 27 215.000 895.61 42 290.000 1960.38 92 140.000 536.96 28 220.000 937.12 43 295.000 2077.62 13 f45.000 549.00 29 225.000 981.74 44 300.000 2203.07 1.8 150.000 562.57 230.000 1029.63 45 305.000 2336.68 A 35 155.000 577.64 31 235.000 1081.04 46 310.000 2479.27 16 160.000 594.14 3> 9 W N I e W ...r. -1

~ e O 9 8 / 02 /90 D EE 3426720222363893 TR) 8643032336782747 AUI C5S 3870808105763701 ISP 72851 96543346826 DE( 0112334567890134 0 NR 1t11111111 11 2222 IP 06 E DR EU) 0000000000000000 TTF. 0000000000000000 AA 0000000000000000 = CRG 0505050505050505 IEE ) DPD 5566778899008122 R NM( 2222222222333333 H IE / T F. G E 4567890123456789 D 3333334444444444 ( ) S ( E T A D R EE 8442'l43410195129 TR) 3354887412406907 P AUI U C5S 1866705212519052. T ISP 7802479258158372 A DE( 5566666777888990 E NR 1 H IP E DR EU) 0000000000000000 2 TTF. 0000000000000000 2 AA 0000000000000000 v CRG E E. IEE 0505050505050505 R. DPD 7788990011223344 t I NM( 1111 1 12222222222 9 F IE 9 O S T R R t P A E E P V 8901234567890123 D U 1122222222223333 l T P u A F 7 G E E g H 0 ry G R 0 E O 0 E F D i, 0' EE 8a689901120907129 E D 1T TR) 8 a. 9 9 9 8 3 5 5 9 8 8 2 7 6 6 8 ro V E ) AUI ~ - +e9M9677i88901+b9 4902618666b15 R T N CSS U T l ISP 245 e C O w DE( S4444444'444455555 L OO NR P P DA IP U I - T E Rt A V E( E E R P DR H U = EU) 00000000000000000 C N R OH T T F. 00000000000000000 AA 00000000000000000 H E IT CRG / T TP IEE 50505050505050505 F I AE DPD 899001 12233445566 S ID NM( 1 1 t1111 11 1 11 1 1 0 O D IE 6 P AW T w M RA e O RL D C IF 23456789o1234567 t 1 1 1 1 1 1 1 a

• 3*pO li

o DeeW COOLDOWN CURVES REG. GUIDE 1.99.REV.2 09/20/89 THE FOLLOWING DATA tfERE PLOTTED FOR COOLDOWN PROFILE i ( STEADV-STATE COOLDOtS8 ) IRRADIATIO8W PERIOD * -5.000 EFP VEARS FLAW DEPTH

• AOWIN T INDICATED INDICATED INDICATED INDICATED INDICATED 18dDICATED T E sePE RA TURE PRESSURE TEMPERATURE PRESSURE T E sePE RA TURE PRESSURE (DEG.F )

(PSI) (DEG.F ) (PSI) (DEG. F ) (PSI) 1 85.000 525.44 16 160.000 719.20 30 230.000 8229.72 2 90.000 532.88 17 165.000 741.19 31 235.000 1289.58 3 95.000 540.88 18 170.000 764.69 32 240.000 9353.58 4 100.000 549.35 19 '75.000 790.10 33 245.000 8422.42 5 105.000 558.58 20 e80.000 887.27 34 250.000 1486.16 6 810.000 568.52 21 185.000 846.58 35 255.000 1575.34 7 915.000 579.29 22 190.000 878.00 36 260.000 1660.23 i I 8 120.000 590.70 23 195.000 981.71 37 265.000 1751.12 9 125.000 602.88 24 200.000 947.99 38 270.000 1848.50 to 130.000 616.17 25 205.000 987.07 39 275.000 1952.68 81 135.000 630.44 26 210.000 1028.93 40 280.000 2064.20 12 140.000 645.75 27 295.000 1073.88 41 285.000 2183.56 13 145.000 662.10 28 220.000 1122.17 42 290.000 2310.84 14 950.000 679.84 29 225.000 1174.07 43 295.000 2446.86 15 155.000 698.85 33 e C=* 9 8 " = - - - ^ - '

OseW COOLDOWN CURVES CEG. GUIDE

1. 99.CE V. 2 0g/20/39 THE FOLLOWItes DATA tfE9E PLOTTED FOR COOLDOWN PROFILE 2 (20 DEG-F / Pet COOLOOWN )

IRRADIATION PERIGO

• 5.000 EFP YEARS FLAW DEPTH = AOWIN T IDeICATED 19e!CATED 10dDICATED 19eD1CATED 19elCATED IselCATED T EsePE R A TURE PRESSURE TEsePERATURE PRES 5URE T EsePE R A TURE PRF55URE (DEG.F)

(PSI) (DEG.F) (PSI) (DES.F) IPSI) 1 85.000 480.00 tt 135.000 598.73 20 180.000 796.59 2 90.000 496.41 12 140.000 614.77 21 185.000 a27.70 3 96.000 504.63-13 145.000 832.19 22 190.000 361.03 4 100.000 513.59 14 150.000 650.74 23 195.000 896.89 5 106.000 523.26 15 155.000 670.92 24 200.000 935.66 6 110.000 533.65 16 160.000 692.56 25 205.000 977.22 7 118.000 644.86 17 185.000 715.75 26 290.000 1021.80 8 120.000 556.77 13 170.000 740.80 27 295.000 106e.75 ) 9 126.000 568.75 19 175.000 767.64 28 220.000 S G e.28 10 930.000 543.71 a l 1 i p e N 1 i i l l 3 i I i i 't l l s O e-* ,n,

DeeW COOLDOWN CURVES REG. GUIDE t.99,REv.2 09/20/89 \\ THE FOLLOtfites DATA tfERE PLUTTED FOR COOLDOWN PftOFILE 3 (40 DEG-F / ## COOLDOtse ) l IRRAOIATION PERIOD = 5.000 EFP VEARS FLAW DEPTH = AOWIN T ISOICATED 18elCATEO 1961CATED 1961CATED IseICA TED 18eICATED TEIEPERATURE PRESSURE TEGEPERATURE PRESSURE Y E8ePE R A TURE PRESSURE l (DEG.F ) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) t 85.000 451.38 10 130.000 551.01 to 175.000 745.17 l 2 90.000 459.31 11 135.000 566.90 20 980.000 776.90 3 95.000 468.01' 12 140.000 583.98 21 195.000 809.93 4 100.000 477.39 13 145.000 602.26 22 190.000 845.61 5 105.000 487.52 14 150.000 572.08 23 195.000 883.92 I 6 110.000 494.42 15 155.000 643.44 24 200.000 925.00 7 115.000 510.00 16 160.000 666.27 25 205.000 969.3e 8 120.000 522.75 17 165.000 691.04 26 290.000 1016.99 9 125.000 535.43 ft 170.000-717.51 ai 215.000 1068.23 I I a N t f h 1 l 0 f 0 i a i ._._.-.._-._-.m

Deew CDOLDOWN CURVES REG. GUIDE 1.99.REV.2 09/20/89 THE FOLLOWING DATA UERE PLOTTED FOR COOLDOUN PROFILE 4 (40 DEG-F / Det COOLDOWN ) .6 IRRAOIATION PERIGO = 5.000 EFP YEARS FLAW DEPTH = AOWIN T INDICATED 18eICATED IteICA TED IteICATED IteICATED INDICATED TEMPERATURE PRES 5URE ' TEMPERATURE PRESSURE T E GIPE R A TURE PRE 55uRE (DEG.F) (PSI) (DEG.F) (PSI) (DES.F) (PSI) 1 85.000 413.37 to 130.000 518.37 se 175.000 725.62 2 90.000 429.70 ft 135.000 535.18 20 180.000 758.31 3 95.000 430.99 12 140.000 553.14 21 186.000 793.72 4 100.000 440.72 13 145.000 572.68 22 190.000 839.69 5 105.000 451.36 to 190.000 593.69 23 195.000 872.58 6 110.000 462.73 15 155.000-616.25 24 200.000 916.53 7 115.000 475.13 16 160.000 640.67 26 205.000 963.87 8 120.000 488.49 17 165.000 666.87 26 210.000 1014.78 9 125.000 502.82 14 170.000 695.19 27 215.000 1069.56. b ~ b O 4 O _______m.__,

m.. 4 DesW COOLDOWN CURVES CEG.. GUIDE 1.99.REV.2 09/20/89 ~

• THE FOLLOWING DATA WERE PLDTTED FOR COOLDOWN PROFILE 5

( 100 DEG-F/M COOLDOWN ) IRRADIATION PERIOD

• 5.000 EFP YEARS l

FLAW OEPTH = AOWIN Y INDICATED INDICATED 18eICATED INDICATED INDICATED 18eICATED T E 80PE R ATURE PRESSURE-TEMPERATURE PRESSURE T E80PE R A TURE PRESSURE i (DEG.F) (PSI) (DEG.F) (PSI) (DEG.F) (PSI) i 1 e5.000 335.71 10 130.000 452.s6 to 175.000 6as.26 2 90.000 344.97 18 135.000 471.81 ?O 190.000 725.66 3 95.000 356.08 12 140.000 492.23 21 186.000 766.00 4 100.000 365.89 13 145.000 514.22 22 190.000 809.44 5 105.000 377.72 14 190.000 538.04 23 105.000 ses.2P 6 110.000 390.50 15 155.000 563.66 24 200.000 906.72 7 895.000 404.29 14 160.000 991.42 25 205.000 961.10 8 120.000 419.25 17 165.000 621.25 26 210.000 1019.63 9 125.000 439.47 18 170.000 653.39 t t h N i ~ I i i' .i . gsg. g o y $r y, -y m.v -g . - = g.-a ' = -. w-ear w y-

DMW COOLDOWN CURVES REG. GUIDE 1.99,REV.2 09/20/89 THE FOLLOWING DATA WERE PLOTTED FOR COOLOOWN PROFILE 1 ( STEADV-STATE COOLDOWN ) IRRADIATION PERIOD

• 10.000 EFP YE A8tS FLAW DEPTH = AOWIN T INDICATED INDICATED INDICATED ICATED ANDICATED INDIC A T E D TEMPERATURE PRESSURE TEMPERATURE

'SURE TEMPERATURE PRESSURE (DEG.F ) (PSI) (DEG.F) r51) (DEG. F ) (PSI) s 1 85.000 513.93 16 160.000 685.30 31 235.000 t189.98 2 90.000 520.59 17 165.000 704.57 32 240.000 1246.69 3 95.000 527.57 18 170.000 725.51 33 245.000 1307.87 4 100.000 535.17 19 175.000 747.95 34 250.000 1373.38 5 105.000 543.34 20 180.000 771.98 35 255.000 1443.44 6 1t0.000 551.99 21 185.000 797.89 36 260.000 1518.91 7 195.000 561.43 22 190.000 825.69 37 265.000 1599.66 8 120.000 571.58 23 195.000 855.46 38 270.000 1686.12 9 125.000 582.50 24 200.000 887.68 39 275.000 1778.77 to 130 000 594.22 25 205.000 922.18 40 280.000 1878.15 ft 135.000 606.69 26 210.000 959.15 49 285.000 1984.76 92 140.000 620.25 27 215.000 998.89 42 290.000 2098.43 13 145.000 634.83 28 220.000 1041.80 43 295.000 2219.86 14 150.000 650.33 29 225 000 1087.75 44 300.000 2349.90 85 155.000 667.18 30 230.000 t137.03 33 e N A e e l E f l } ~e O e O f r

Dmw COOLDowN CURVES REG. GUIDE 1.99,REV.2 4 M/20/89 TifE FOLLOWING DATA trERE PLOTTED FOR COOLDOWN PROFILE 2 (20 DEG-F / HR COOLDOwN ) IRRAOIATION PERIOD

• 10.000 E F P t?ARS

' FL AW OEPTH = AOWIN'T INDfCATED INDICATED INGTCATED INDICATED INDICATED INDICATED T E MPE RA TURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSURE (DEG. F ) (PSI) (DEti. F ) (PSI) (DEG.F ) (PSI) t 85.000 476.47 11 135.000 573.49 29 185.000 775.77 2 90.000 483.27 12 140.000 587.76 22 190.000 805.20 3 95.000 490.61 13 145.000 602.98 23 195.000 837.It 4 100.000 498.51 14 150.000 6t9.51 24 200.000 871.22 5 105.000 506.92 15 155.000 637.32 25 205.000 907.88 6 110.000 596.08 16 960.000 656.28 26 210.000 947.26 7 115.000 525.96 17' 165.000 676.90 27 215.000 999.87 8 120.000 536.58 18 170.000 699.00 28 220.000 1035.42 9 125.000 548.03 19 175.000 722.73 29 225.000 1084.42 10 130.000 560.22 20 180.000 74S.31 30 230.000 1137.03 3= 1 e N UU e L s . ~,

c Deew COOT.DOWN CURVES REr. GUIDE t.99,REV.2 09/20/89 THE FOLLOWING DATA ifERE PLOIIED FOR COOLDOWN PROFILE 3 (40 OEG-F / HR COOLDOWN I IRRADIATION PERIOD = 10.000 EF P VEAR5 FLAW DEPTH = AOWIN T INDICATED INDICATED INDICATED INDICATED INDICATED 19dDI CA S ED 1EnePE RA 1URE PRE 55URE 1 E nePE R A1URE PRES 5URE 1E nePE R A1URE PRES $URE 5 (DEG.F ) (PSI) (DEG.F) (PSI) (DEG. F ) (PSID 1 85.000 438.37 11 135.000 540.f2 21 185.000 754.49 2 90.000 445.44 12 140 000 555.04 22 190.000 786.05 3 95.000 453.00 13 145.000 579.29 23 195.000 819.86 4 100.000 461.23 to 150.000 588.76 24 200.000 856.17 5 105.000 470.14 15 155.000 607.46 25 205.000 895.46 6 110.000 479.72 16 160.000 627.73 26 210.000 937.54 7 185.000 490.08 17 165.000 649.42 27 285.000 982.82 8 120.000 501.13 18 170.000 672.93 28 220.000 1039.49 9 125.000 513.17 19 175.000 698.22 29 225.000 1083.83 to 130.000 526.13 20 180.000 725.34 e N Ch 4

e DMW COOLDOWN CURVES DEG GUIDE t.tts,Rfw.2 9 09/20/89 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 4 (60 DEG-F / HR COOLDOWN I IRRADIATION PERIOD = 10.000 EFP VEARS FLAW OEPTH = AOwlN T INDICATED INDICATED INDICATED INDICATED INDICATED INDICATED T E tePE RA TURE PRESSURE T E asPE RA TURE PRESSURE TEMPERATURE PRESSURE ( DEG. F ) (PSI) ( DEG. F ) -(PSI) (DEG.F ) (PSI) 1 85.000 399.56 11 135.000 506.50 21 185.000 734.49 2 90.000 406.92 12 140.000 522.41 22 190.000 767.96 3 95.000 414.91 13 145.000 539.61 23 195.000 804.01 4 100.000 423.52 14 150.000 558.00 24 200.000 842.93 5 105.000 432.86 15 155.000 577.99 25 205.000 884.82 6 110.000 442.92 16 160.000 599.36 26 290.000 929.79 7 115.000 453.72 17 165.000 622.59 27 285.000 978.22 8 120.000 465.45 18 170.000 647.56 28 220.000 1030.30 + 9 125.000 478.15 99 175.000 674.38 29 225.000 1086.30 10 130.000 491.82 20 180.000 703.21 9 N N e W

088W CDOLDOWN CURVE 3l REG.' GUIDE 1.99.REv.2 09/20/89 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 5 ( 100 DEG-F/HR COOLDOWN ) IRRAOIATION PERIOD =

10. 000 E F P VEARS FLAW DEPTH = AOWIN T IWICATED INDICATED IN01CATED INDECATED I WICATED INDICATED T E 8ePE R A TURE PRESSURE T E sePE R A TURE PRESSURE T EsePE R A TURE PRES 5URE.

( DE G. F ) (PSI) ( DEG. F ) (PSI) (DEG. F ) (PSI) 1 85.000' 320.02 11 135.000 439.18 20 180.000 662.49 2 90.000 328.09 12 140.000 457.01 21 185.000 696.17 3 95.000 336.88 13 145.000 476.43 22 190.000 736.51 4 100.000 346.39 14 150.000 497.36 23 195.000 777.85 5 105.000 356.73 15 155.000 519.89 24 200.000 822.35 6 t10.000 367.84 16 160.000 544.30 25 205.000 ~ 922.00 870 33 7 915.000 379.9s 17 165.000 570.55 26 210.000 8 120.000 393.08 18 170.000 598.96 27 215.000 977.62 9 125.000 407.22 19 175.000 629.56 28 220.000 1037.4s 10 130.000 422.55 e N03 i

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