ML20059E749

Proposed Tech Specs,Revising Reactor Vessel pressure-temp Limits

ML20059E749
Person / Time
Site:	Mcguire, McGuire
Issue date:	08/30/1990
From:	DUKE POWER CO.
To:
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ML20059E748	List: ML20059E743 ML20059E749
References
NUDOCS 9009100303
Download: ML20059E749 (115)

Text

ll.S. Nuclear Regulatory Consnission ATTN: Document Control Desk August 30, 1990' I

i Attachment No. 2 Duke Power Company McGuire Nuclear Station Proposed Changes to Technical Specifications i

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REACTOR COOLANT SYSTEM fr{d !y- Ar 3/4.4.9 PRESSURE / TEMPERATURE LIMITS th_ 1 -

res A. ton M[Y -

, , /

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION _

3.4.9.1 The Reactor Coolant System (except-the pressurizer) terotrature and i

pressure 3.4-2, 3.4s11 4 be Ilmited in accordance with the limit lines shown on Figures inservice, leak and hydrostatic testing with:3.4 4, and 3.4-5 during haatup, l coold a.

Maximum heatup rates as specified in Figures 3.4-2 and 3.4-3 I l

b.

"'simum cooldown rates as specified in Figures 3.4-4 and 3.4-5 l

c.

wximum temperature change 'of less than or equal .o 10*F in any 1-hour period during inservice hydrostatic and-leak testing operations above the heatup and cooldown limit curves. R APPLICA811.!TY: At all times.

ACTION: '

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes;-perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant -

System remains acceptable for continued operation or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the andRCS pressure T,ygto less-than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes'during system  :

heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined,-to-determine changes in' material properties, as 4.4-5. by 10 CFR.50, Appendix H in accordance with the schedule in Table-required The results 3.4-2, 3.4-3, 3.4-4, of andthese examinations shall be used to update Figures 3.4-5.'

l i

McGUIRE - UNITS 1 and 2 3/4 4-30 Amendment No. 82 (Unit 2)

Amendment No.100-(Unit 1)

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h PREIBURE-TEMfERATURE LIMIT %

750 f

ACCEPTABLg .,

OPERATH IN 500 ,

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CRITH ALITY Li dlT BASEI) 250 ON INI mRvice ilvomosTs , Tic

! TEST ' IMPER A' 'URE 131: 'F) l FOR T48 ssRvi :E PERIOD '

UP T010 EPPY 0

0 50 1 ISO 200- 250 300 380- 400 450 500 INDICATED TEMPERATURE (OF)  !

ouswe son MsAtw RATES W /DIR pon Des anavsniAL saass: 4 esavies see w to se am coeffhoLLlos5 esafsRIAL-WsLo het?AL coserA copean comem-eJemen .

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s FIGURE 3.4-2 M8GUIRE UNIT If REACTOR COOLANT SYSTEM, HEATUP LIMITATIONS NRC RQ 1.99 REV 2 McGUIRE - UNITS 1 and 2 3/4 4-31 MnNnNoN Nt9)

Amendner.v. No. -8&- Uni t 2)

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Acceptable 0 aeration Hydrostatic Test' Temperature (311'F)

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' Period Up To.10 EFPY i

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0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG F) i M ATE RIA L B ASIS CURVE APPLICAeLE FOR NE ATUP R ATES

- CO NT R O L LIN G M AT E R I A L-LO N G IT UD IN A L UP TO 88'P /N R.P O R T N E SE R VICE PE R IO D C O PP,E R C O N T E N T : 0.21 w t%

UP TO 1e EPPV. CONT AINS M AROIN8 OP ' WELD RT it*P A ND ee P810 P O R PO SSIB L E ny N0 IN ITI A L: -80'F -

q0T-A P T E R 3 2 E P PY t 1/47,18 5.6* P IN ST R UM E NT E R R O R 8.

3 /4 T ,113' P FIGURE 3.422 MCGUIRE UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS NRC RG 1.99 REV 2 APPLICABLE FOR THE FIRST-10 EFPY

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UP Tt PPY t 0 '

0 50 100 100 200. 290 300 380 400' 450  !

500

- IN0iCATED TEMPRRATURE (0P) assusaska Poa l#ATUP mATSRIALSA85 RA 09 Tg agapnem POR TMS penece up to e 88PY cD8f?AOLLNs3 asAT8AIAL-48A470a se ooortasesesaasineor - vessel imaansecears snei.L as e9 Assoserese soaposerts coseta coastsat e w instmussent som ny,, ,,,yia6 ,

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• Approve by NRC for first 5 EFPY or completion of- iret.asw -

the re eling outage at the end of fuel cycle 6, 8'*T '

base on Generic t.etter 88-11.

PlGURE 3.4 3 MegugRE UNIT 2 REACTOR COOLANT SYSTEM, HEATU$ LIMITATIONS APPLICABLE FOR THE FIRST 8 EPPY McGUIRE - Units 1 and 2 3/4 4 32 Amendment No. 400 (U. 1

- Amendment' No. -99 (Uni t -2

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CU RVE S APPLIC AB LE FO R H E ATUP M AT E RI A L B Atle R A T E S UP.TO 60'P /H R F O R T H E SE R VIC E CO NTR O LLING M ATE RIA L: LOW E R SHE L PE R IOD UP TO 10 E P PY. CO N T AIN S M A R 0 lN S O F 10'P A N D 00 P8le F O R Po ttlR L E COPPER CONTENT: 0.15st%

RT IN ITI A L: . 80'P IN ST R UM E NT E R R O R. .RT y A P T E R 10 E P PY: 1/47.90'F 3 /4 7, 61 ' F d Y j l FIGURE 3.4 3- McGUIR RE LANT 4

' SYSTEM LtMI A NRC RG 1.98 A V 2 APPLICABLE FOR THE FIRST 10 EFPY ,

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0 O 50 150 200 .250 300 350 .400 450 500 4 INDICATED TEMPERATURE (OF)

CUWWW POR Co0LoouuN MATSAIAL SA80s .

RAT 68 70688PAGA PORTag- CoNTROLLite0 MATSAfAL-uv8LD MET AL seny PsRios up To 10 EPPV COPPG A CONTENT .4.3 Bus l Co Awas asAnoiN POn Posseeks . Pwospwonus coNTsNT .e.otsms l l RuleGNT E AROAS, RTsegylN4TIAL-c p. o 1/47,1GS.5'P l RTpegyAPTER 10 GPPV 3/47,11389 .

FIGURE 3,4-4 McGulRE UNIT 1, REACTOR COOLANT SYSTGM, C00LDOWN LIMITATIONS NRC RG 1.90 REV 2' APPLICABLE FOR THE FIRST 10 EFPY McGUIRE - UNITS 1 and 2 3/4 4-331 Amendnent No. be6. (Unit 1 Amendnent No. #- (Unit 2 :

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CURVE APPLICABLE POR COOLDOWN R ATES . M A TE R IA L B ASIS UP TO 198'P/NR FOR THE SE R VICE PE RIOD CONTRO L LINS IA ATER IA L. LO N0lTUDIN A L UP TO 10 EPPV. CONT AINS M ARGINS OF C O PP E R C O N T E N T : 0,.31 s tib it'P AND 80 PSl4 PO R POSSIB L E RT NOTINITIA L -80 P WELD INSTR U AR ENT E R ROR , RT NDT A TER 32 EPPY: 1/47,188 3 /4 T ,113,8, P P FIGURE 3.4-4 ' M8GUIRE UNIT 1, RE ACTOR COOLANT SYSTEM, COOLDOWN LIMITATIONS NRC RG 1.99 REV 2 APPLICABLE FOR THE FIRST 10 EFPY

b fY"M d; g WW Awe 2500 .

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• Aoproved y flRC for first 5 EFPY or completion of the- fueling outage at the end of fuel cycle , based on Generic Letter 88-11. '

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i FIOURE 3.4 5 McGutRE UNIT 2 REACTOR COOLANT "

SYSTEM, COOLDDWN LIMITATIONS j

APPLICABLE FOR THE FIRST 8 EFPY i ficGUIRE - UNITS I and 2 3/4 4-34 Amendment flo.6 (Uni: !!

Amendment No. % (Unit 2 '

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0 I ' ' i i 0 50 100 150 200 250 300 350 400 450 500' i INDICATED TEMPERATURE (DEG F)

CURVES AP9LICA8LE POR COOLDOWN A ATES M A T E R B A L 8 A S18 up To 10e*P/wR PoA TNa sERvlCE PER100 CoNTRo L LING M AT BRI A L - LOW E R SH E LL UP TO to 8PY AND CONTAINS M ARGINS OF C O PP E R CO N T E N T : 0.,8 w m 10*P AND 98 PS48 POR POSSIBLE " NDT N A -30P INSTRUMENT ERRORS. NDT ,,PV: 1/4T,90*P 3/47.61*P FIGURE 3.4-5 McGUIRE UNIT 2, REACTOR COOLANT SYSTEM, COOLDOWN LIMITATIONS NRC RG 1.99 REV 2 APPLICABLE FOR THE FIRST 10 EFPY

IABLE 4.4-5

? .

8 REACIOR VESSEL MATERIAL SURVEILLANCE PAGGRAi1 - WITHDRAWAL SCHEDetE

=

CAPSULE VESSEL- LEAD NIWSER . LOCATION UIIISRAMAL IINE-(EFPY)*

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3. W 124* 4.76 66 Standby Seeney sc'

4. X 236' 4.76 6 .28 t

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-- "approaching Withdrawal time the withdrawal may be modified to coincide with those refueling outages or plant shutdowns most closely schedule.

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. _ _ . . ,, _ ,~ , , ,

U.S. Nuclear Regulatory Commission

= ATTN: Document Control Desk August 30, 1990 Attachment No. 3 Duke Power Company McCuire Nuclear Station- '

Analysis Of Capsule X Fror:. The-Unit 2 Reactor Vessel Radiation Surveillance Program-4

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i WCAP-12556 l

WESTINGHOUSE CLASS 3 1

ANALYSIS 0F CAPSULE X FROM THE-DUKE POWER COMPANY MCGUIRE UNIT 2 REACTOR' VESSEL RADIATION SURVEILLANCE PROGRAN .

1 1

E. Terek S. L. Anderson L. Albertin N. K. Ray i April 1990 Work performed under Shop Order No. DSMJ-106 APPROVED: h./D M T. A. Meyer, S nager Structural Mathrials and Reliability Technology Prepared by Westinghouse for the-Duke Power Company WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Division P.O. Box 2728 Pittsburgh, Pennsylvania, 15230 e 1990 Westinghouse Electric Corp.

'I

~._________

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PitEFACE  :

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i  !

This report has been technically reviewed and verified.

l Reviewer ,

1 Sections 1 through $_and 7, 8 J. M. Chicots M7/ NN<M) C -t

.Section 6. E. P. Lippincott V C P -

ff ,

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r r

w l

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t TABLE OF CONTENTS Section Title Page 1

SUMMARY

OF RESULTS 1-1 2 INTRODUCTION 2 3 BACKGROUND 3-1

.t 4 DESCRIPTION OF PROGRAN 4-1; 5 TESTING OF SPECINENS FRON CAPSULE X 5-1 J 5-1. Overview 5-1 5-2. Charpy V-Notch Impact Test Results 5-3 5-3. Tension Test Results 5-4 5'- 4 . Compact Tension Tests 5-5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 i

6-1. Introduction 6-1 6-2. Discrete Ordinates-Analysis 2:

3.

Neutron Dosimetry 6-7 7

, SURVEILLANCE CAPSULE REMOVAL' SCHEDULE 7-l' 8 REFERENCES 8-1 '

t-Appendix A - Heatup and Cooldown Limit Curves for Normal Operations

. I kWfsM*MM:10 -

gjj n . . .. ... .. _______._- ---- --_

l :

f LIST OF ILLUSTRATIONS Figure Title Page t 4-1 Arrangement of Surveillance Capsules in the .

4-6 McGuire Unit 2 Reactor Vessel l

4-2 Capsulo X Diagram Showing Location of Specimens, 4-7. .

Therr.al Monitors, and Dosimeters 5-1 Charpy V-Notch Impact Data for McGuire Unit 2 5-13 Reactor Vessel Intermediate Shel1~ Forging 05 HT. 526840 (Axial Orientation) 5-2 Charpy V-Notch Impact Data for McGuire Unit 2 5-14 Reactor Vessel Intermediate Shell Forging 05 HT. 526840 (Tangential Orientation) 5-3 Charpy V-Notch Impact Data for McGuire Unit 2 5-15 Reactor Vessel Wald Metal 5-4 Charpy V-Notch Impact Data for McGuire Unit.2 5-16 Reactor Vessel Wald Heat Affected Zone Metal 5-5 Charpy Impact Specimen Fracture Surfaces for McGuire 5-17' I Unit 2 Reactor Vessel Intermediate Shell Forging 05 HT. 526840 (Axial Orientation) ,

t o 5-6 Charpy Impact Specimen Fracture Surfaces for McGuire 5-18

, Unit 2 Reactor Vessel Intermediate Shell Forging 05 l HT. 526840 (Tangential Orientation) l 5-7 Cherpy Imoact Specimen Fracture Surfaces for 5-19 McGuire Unit 2 Reactor Vessel Wald Metal 5-8 Charpy Impact Specimen Fracture. Surfaces for 5-20 McGuire Unit 2 Reactor Vessel HAZ Metal mer.maenao jy

LIST OF ILLUSTRATIONS (Cont)

Figure Title Page 5-9 Tensile Properties for McGuire Unit 2 Reactor 5-21:

Vessel Intermediate Shell Forging 05 HT. 526840 (Axial Orientation) 5-10 Tensile Properties for McGuire Unit 2 Reactor 5-22 Vessel Intermediate Shell Forging 05 HT. 526840 l -

(Tangential Orientation) 5-11 Tensile Properties for McGuire Unit 2 Reactor 5-23.

Vessel Weld Metal '

5-12 Fractured Tensile Specimens for McGuire Unit 2 5-24 Reactor Vessel Intermediate Shell Forging 05 HT. 526840 (Axial Orientation) 5-13 Fractured Tensile Specimens for McGuire Unit 2 5-25 Reactor Vessel Intermediate Shell Forging'05 HT. 526840 (Tangential Orientation) I 5-14 Fractured Tensilo Specimens for McGuire Unit 2 5-26 4 Reactor Vessel Wald Metal '

5-15 Typical Stress-Strain Curve for Tension Specimens- 5-27 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-12 6-2 Core Power Distributions Used in Transport Calculations 6 , For McGuire Unit 2 me.mmeano y

1 t

LIST OF TABLES Y

Table- Title Page-4-1 Chemical Composition of'the McGuire Unit 2 4-4.

Reactor Vessel Surveillance Materials s

4-2 Heat' Treatment of the McGuire Unit 2 4-51 Reactor Vessel Surveillance Natorials 5-1 Charpy V-Notch Impact Data for the McGuire Unit 2 5-6 Reactor Vessel' Intermediate Shell Forging 05 (HT. 526840). Irradiated at 550*F, Fluence 1.45 x 10 18 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the McGuire Unit 2 5-7 i Reactor Vessel Wald Metal and HAZ Metal Irradiated- '

at 550'F, Fluence 1.45 x 10 19 n/cm2 (E > 1.0 MeV) >

5-3 Instrume.ted Charpy Impact Test Results for McGuire 5-8

~

Unit 2 Reactor Vessel Intermediate Shell Forging 05.

L (HT'. 526840) 5-4 Instrumented Charpy Impact Test Results for 5-9 McGuire Unit 2 Reactor Vessel Wald Metal and HAZ Metal 5-5 The Effect of Irradiation ~to-'1.45 x 1019 n/cm2 5-10 (E > 1.0 MeV) at 550'F on the Notch Toughness-Properties of The McGuire Unit 2 Surveillance Capsule Materials-5-6 Comparison of the McGuire Unit 2 Reactor Vessel 5-11' Surveillance Capsule Charpy Impact-Test Results with Regulatory Guide 1.99 Revision 2 Predictions vi

i LIST OF TABLES (Cont) l Table Title  ? age <

5-7 Tensile Properties for the McGuire Unit 2 Reactor Vessel 5-12 Surveillanci Capsule Materials Irradiated to 1.45 x 10 1E e n/cm2 (E > 1.0 MeV) at 550'F s

6-1 Calculated Fast Neutron Exposure Rates-at the: 6-14  ;

Surveillance Capsule _ Center 6-2 Calculated Fast Neutron Exposure Parameters at the 6-15 1 Pressure Vessel Clad / Base Metal Interface >

6-3 Relative Radial Distributions of Neutron Flux 6-16 (E>1.0 MeV) Within the Pressure Vessel Wall i

6-4 Relative Radial Distributions of Neutron Flux 6-17 (E>0.1 Mey) Within the Pressure Vessel Wall- i 6-5 Relative Radial Distribution ofLIron Displacement '6-18 Rate (dpa) Within the Pressure Vessel Wall i

6-6 Nuclear Parameters fo, ;'eutron Flux Monitors 6-19 6-7 Irradiation History of Neutron Sensors Contained 6-20 in Capsule X 6-8 Measured Sensor Activities and Reaction Rates 6-23 i 6-9 Summary of Neutron Oosimetry Results 6-25 6-10 Comparison of Measured and FERRET Calculated 6-26 Reaction Rates at the Surveillance Capsule Center mer.mmeano vii

i LIST OF TABLES (Cont).

9 Table Title Page 6-11 Adjusted Neutron Energy Spectrum at the. Surveillance 6-27 Capsule Center 6-12 Comparison of Calculated and Measured. Exposure 6-28 Levels for Capsule'X 6-13 Neutron Exposure Projections at Key Locations'on the 6-29 Pressure Vessel Clad / Base Metal Interface ,

6-14 Neutron Exposure Values for Use.in the Generation 6-30 of Heatup/Cooldown Curves 6-15 Updated Lead Factors for McGuire Unit 2' Surveillance 6-31 Capsules I

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1 SECTION 1 )

SUleMRY OF RESULTS The analysis of the reactor vessel material contained in Capsule X, the second 4 surveillance capsule to be removed from the Duke Power Company McGuire Unit'2 -

reactor pressuro vs:sel,, led to the following conclusions:

o The capsule received an average fast neutron fluence (E > 1.0.MeV) of 1.45 x 10 19 n/cm2 ,

o Irradiation of the reactor vessel intermediate shell Forging 05 4 19 2 (HT. 526840) to 1.45 x 10 n/cm , resulted in 30'and 50 ft-lb l transition temperature increases of 105 and 120'F respectively, for i specimens oriented normal'to the major working direction of Forging 05 I

(axial orientation) and a 30 and 50 ft-lb transition ~ temperature increase of 100 and 120*F for specimens oriented parallel to the major working direction of Forging 05'(tangential' orientation).

o Wald metal irradiated to 1.45.x 1019 n/cm2 resulted in.a 30.and 50 ft-lb transition temperature increase of 35 and 40*F, respectively. '

o Irradiation to 1.45 x 10 19 n/cm2 resulted in~a decrease in the '

average upper shelf energy of Forging'05 (axial orientation) of 17 ft-lb and no upper shelf energy decrease for the weld" metal. Both materials exhibit a more than adequate shelf level for continued safe plant operation.

o Comparison of the 30 ft-lb transition temperature increases for the.

McGuire Unit 2 surveillance material with predicted increases using; the methods of NRC Regulatorv Guide 1.99, Revision-2, shows that'the l l

mesmm.o 11 1

l I

D n

Forging 05 material transition temperature increase was 21*F less than the prediction. The weld metal showed a transition temperature increase that was 7'F less than the prediction.

Impact Of Test Results On Plant Life Extension.

I o The measured ART OT values are I wer than those values predicted at i 18 1.45 x 10 -n/cm (-22 EFPY) for forging 05. 'This may provide for less restrictive.ASNE,Section III, Appendix'G heatup and cooldown curves for future plant life. The future surveillance. capsule's. test data will be required to determine what potential benefit, if any, may be utilized for heatup and cooldown curves developed for an extended vessel life,-i.e. ,

Plant Life Extension. However. mr Appendix A, when surveillance capsule "

data is used for the generation of heatup and cooldown curves-the limiting material becomes the lower shell forging and not the' intermediate shell forging.

o PTS' margin should exist for some amount of life extension beyond the current license life of the McGuire Unit 2 based on the predicted values 4 l of RT py3 The data reported here can imply. additional PTS margin since the measured RTNDT values for the axial weld material are significantly less than the predicted values using Regulatory Guide 1.99, Revision 2 prediction methods. However, since surveillance capsule data cannot be used in the RTPTS calculation, this benefit cannot be readily-obtained-because the PTS rule requires the use of only predicted RTNOT(i

l RTPTS) values.

l L

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i 1-2 i

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SECTION 2 =

INTRODUCTION-This report presents the results of the examination of Capsulo X, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron . irradiation on the Duke Power Company McGuire Unit 2 reactor pressure vessel materials under actual operating j conditions.

The surveillance program for the-Duke Power Company McGuire Unit 2. reactor pressure' vessel materials was designed and recommended by the Westinghouse 1 Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor. vessel materials are -

presented by Koyama and Davidson.III The surveillance program was planned to cover the 40 year design life of the reactor pressure vessel:and was based on ASTM E-185-73, " Standard Recommended Practica for SurveillanceiTests for '

Nuclear Reactor Vessels". Westinghouse Energy Systems personnel were contracted to aid in the preparation of procedures for removing the capsula from the reacto'r and its shipment'to the Westinghouse'. Science and Technology -

Center Laboratory, where the postirradiation mechanical testing of the Charpy  :

V-notch impact and tensile surveillance specimens was performed.

I This report summarizes testing and the postirradiation data obtained from surveillance Capsule X removed from the Duke Power Company.McGuire Unit 2 reactor vessel and discusses the analysis of the data. The' data are also-compared to Capsule V(2) which was removed from the reactor..in 1986.

g

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l SECTION 3 8ACKGROUS t 1 l

The ability of the large steel pressure vessel containing the reactor core'and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline' region of the reactorf pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of' fast neutron irradiation on the mechanical-properties of low alloy, ferritic pressure vessel steels such as SA 508 Class 2 (base material of the McGuire Unit 2 reactor pressure vessel beltline) are well documented-in the literature. Generally; low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for performing analyses to guard against fast fracture in reactor .;

pressure vessels has been presented in " Protection Against Non-ductile .,

Failure," Appendix G to Section'III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the i

reference nil-ductility temperature (RTNDT)*

RT NDT is defined as.the greater of either the drop weight nil-ductility transition temperature (NDTT per_ ASTM E-208)'or-the temperature 60*F less than the-50 ft-lb (and 35 mil lateral expansion) temperature as determined from Charpy specimens with the longitudinal axis oriented normal (axial orientation) to the major working direction of the forging.- ~ The-RTNOT of a sur.mmno 31 f

4 i

given material is used to index that material to a referer. .. cross intensity

, factor curve (Kgg curve) which appears in Appendix G of the ASME Code. The Kgg curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Kgg curve, allowable stress intensity factors can be obtained for this material as a function of temperature.

Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

RTNOT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel ,

material properties. The 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 McGuire Unit 2 Reactor Vessel Radiation Surveillance Program,UI in which a surveillance capsule is '

periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft Ib l

temperature (ARTNDT) due to irradiation is added to the original RT NOT.

to adjust the RT NOT for radiation embrittlement.

This adjusted RT NDT (RT NOT initial + ARTNOT) is used to index the material to the K gg curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.

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w.4mse ie 32

_ . _ _ _ _ _ _ . _ _ _ . . _ _ . . _ . . . _.__..._s

- _ . - . - - . - . _ - _ - - _ . . - - . . - - _ . _ - - . - . - _ . - - . _ _ - _ ~ _ _ _ _ _ _ _ _ _ -

4 i

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, SECTION 4  !

! DESCRIPT!aN OF PROGRAN Six surveillance capsules for monitoring the effects of neutron exposure on the McGuire Unit 2 reactor 'ressure vessel core region material were inserted in the -l reactor vessel ,.fior to initit.1 plant startup. The capsules were positioned in 4

the reactor vessel between the neutron shield pads and the vessel wall at i locations shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

Capsule X (Figure 4-2) was removed after 4.16 effective full power years of plant operation. This capsule contained Charpy V-notch, tensile, and compact '

tension (CT) specimens from subs.orged are weld metal fabricated with the same WM wire and flux as used in the reactor vessel core region girth weld, and Cnspy V-notch, tensile, CT, and bend bar specimens frort the intermediate shall forging 05. The capsule also contained Charpy V-notch specimens from weld Heat .

AffectedZone(HAZ) metal. All heat affected zone specimens were obtained from the weld HAZ of forging 05.

The chemistry and heat trsatment of the wrveillance mater tal are presented in Table 4 1 and Table 4-2, respectively. The chemical anal)ses reported in table 4-1 were obtained from anirradiated material used in the surveillance pr.ogram.

Test specimens obtained from the intermediate shell forging (after the mal' heat treatment and forming of the forging) were taker at least one forging thickness i

from the quenched ends of the forging. Test specimens were machined ? rom the

, 1/4 thickness location of the forging af ter nerforming a sintated postweld ,

mr.ameo io 41

- - - . . . _ , - , , , , _ . . , _ , . . . , . . - + . - _._.m, _ _ , , , , . , _ , . , . . . . ~ . , . . . _ , . . , _ . . . . ~ _ . . , . . . . , . _ - - . _ . - - . . ., -, .. .-r-.---

1' 1

l 4

i

, stress-relieving treatment on the meterial and also from weld and heat-affected-zone metal of a stress relieved weldment joining the lower and intermediate shell forgings. All heat affected-zone specimens were obtained from the weld heat-affected-zone of the intermediate shell forging, i

Charpy V-notch specimens from the intermediate shell forging were machined in  !

both ti.ngential (longitudinal axis of the specimens parallel to the major working direction) and axial (longitudinal axis normal to the major working direction) orientations. The Charpy V-notch specimens from the weld and weld '

. heat-affected-zone metal were machined perpendicular to the weld direction with i

the notch oriented in the direction of the weld.

Tensile specimens from the intermediate shell forging were machined with the longitudinal axis of the specimen both parallel and normal to the major working ,

direction. Weld specimens were oriented normal to the weld direction.

Bend bar specimens were ma:hined from the intermediate shell forging with the l

longitudinal axis of the spean . orisated parallel to the wo rking direction of the forging such that the simulated crat.k woul'd propagato nn mal to the working direction of the forging. All bend bar specimens were fati ue 3 procracked according to ASTM E-399.

Compact tension test specimens from the intermediate shell forging were machined in both axial and tangential orientations. This was done to obtain toughness data both normal and parallel to the major working direction of the forging.

I Compact tension test specimens from the weld were machined normal to the weld direction with the notch oriented in the direction of the weld. All specimens were fatigue procracked according to ASTM E-399.

Capsule X contained dosimeter wires of pure' iron, copper, nickel, and aluminum-cobalt (cadmium-shieldedandunshielded). In addition, cadmium-shielded dosimeters of Neptunium (Np237) and Uranium (U238) were contained in the capsule.

me.amem 4-2

Thermal monitors made from two low-melting outectic 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 molting points are:

2.5% Ag, 97.5% Pb hitingPoint579'F(304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb hitingPoint590'F(310'C)

The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule X are shown in Figure 4-2.

4 mer.mmee ie 4,3

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i

, TABLE 4-1 CHEMICAL COMPOSITION OF THE MCGUIRE UN.'.T 2 REACTOR VESSEL -

SURVEILUNCE MATERIALS f Forging 05 (HT.526840) Weld Metal M ,

Elemtg (Wt. %)- (Wt. %)

C 0.18 0.055 f

S 0.017 0.015 l N

2 0.004 0.011 Co 0.019 0.007 '

Cu 0.16 0.031 )

l Si 0.23 0.29 Wo 0.58 0.55

! Ni 0.79 0.73 Mn 0.69 1.81 .

Cr 0.43 0.030 V <0.002 <0.002 '

P 0.012 0.016 t Sn 0.008 0.002 Ti <0.001 0.004 Pb 0.001 0.003 ,

W <0.002 <0.002 Zr <0.002 <0.002 As 0.018 0.015 A1 0.016 0.015 B <0.003 <0.003 i Sb <0.002 <0.002 (a) Surveillance weld specimens were made of the same weld wire and flux as i the girth seam weldment between forging: 04 and 05 (Weld Wire Heat No. 6;t075 and Grau L.0, Flux Lot No. P46) ,

s I

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• 4-4

I TA8LE 4-2  !

] HEAT TREATMENT OF THE NCGUIRE UN!i 2  ;

l REACTOR VESSEL SURVEILLANCE MATERIALS l Material Temperature ('fl Time O J Coolant Intermediate Shell 1688-1697 3.5 Water quenched Forging 05 1229/1238 7.5 Air cooled '

(HT. 526840) 1140_+25 22.0 Furnace cooled Weld Metal 1140_+25 15.0 Furnace cooled 4

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r b

B

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1 1

I i REACTOR VESSEL O.

CORE BARREL

! i PEUTMON PAD Z (4.76) i U (4.76) .

fgE V (4.06)

SS' , , 58*-

l d '99. *&

270'- -- -

\ "

90' I

Y (4.06) I I I

X (4.76) 1 W (4.76)  :

100* (

)

Figure 4-1. Arrangement of Surveillance Capsules in the McGuire Unit 2 Reactor Vessel mer.mmen io d 46 '

,<+-,.e .,w-- -e yc--- e . , -%, .-e,--,-. .,-..~,..c-+-,+.-..--~.m.. ...<es.~----,-__m.,. ~ ---I

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4 2

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an. A

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• j _ _ _ _ _ _1 __ - = _

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Figure 4-2 Capsule X Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters  !

,mm. ..

- . , - .. . _ . - _ _ _ _ _ _ _ - - . _ _ __ .. ____ - _ _ _ -__ - _ _ - _ _ _ _ . - _ _ _ .- - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ . -_-____1 - _ . _ . _ - _ _ _ - . - _ - _ _ _ . - __ :

J 4  !

SECTION 5  !

TESTING OF SPECIMENS FROM CAPSULE X

{

5-1. OVERVIEW r The postirradiation mechanical testing of tl's Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center  !

Laboratory with consultation by Westinghoura Energy Systems personnel. t Testing was performed in accordance with 10CFR50, Appendices G and HI33, ,

j ASTM Specification E185-82 and Westinghouse Procedure NHL 8402, Revision 1 as modified by Westinghouse RMF Procedures 8102, Revision 1 and 8103, Revision 1.

c Upon receipt of the capsJ1e at the laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked -

against the master list in WCAP-9488.Ill No discrepancies were found.

Examination of the two low-melting 304'C (579'F) and 310'C '(590*F) eutectic alloys indicated no molting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed waslessthan304*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, 358J machine. The tup (striker) of the Charpy machine is instrumented.with an Effects Technology model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition .to.the standard measurement of Charpy energy (E).D From the load-time curve, the load of general yielding (Pgy),thetimetogeneralyielding(tgy), the maximum load (Pg), and the time to maximum load (tg) can be determined. Under some test  !

me.mwo 5-1 I

i i

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 (Pp ), and the lead at which fast fracture terminated is identified as the arrest load (Pg ).

The energy at maximum load (E g ) was determined by comparing the energy-time record and the load time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen.

Therefore, the propagation energy for the crack (E p

) is the difference -

between the total energy to fracture (E )Dand the energy at maximum load.

The yield strest (oy) a 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 A370-88. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tension tests were perfermed on a 20,000 pound Instron, split-console test machine (Model 1115) per ASTM Specifications E8-83 and E21-79, and RMF Procedurw 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. The upper pull rod was connected trrough a universal joint to improve axiality of loading. The tests were conducted at a l constant crosshead speed of 0.05 inch per minute throughout the test.

) Deflection measurements were made with a linear variable displacement I transducer (LVDT)oxtensometer. 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 electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air.

5-2

Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.

Chromel-alumel thermocouples were inserted in shallow holes in the center and f each end of the gage section of a dummy specimen and in each grip. In test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550'F (288'C). The upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen temperatures.

Experiments indicated that this method is accurate to plus or minus 2*F.

l The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the

. original cross-sectional area. The final diameter and final gage length were  !

determined from postfracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

i 5.2. CHARPY V-NOTCH IMPACT TEST RESULTS The results of Charpy V-notch impact tests performed on the various materials contained in Capsule X irradiated to approximately 1.45 x 1019 n/cm 2 at 550'F 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 Capsula Y ::derial are shown in Table 5-5.

Irradiation of the vessel intermediate shell Forging 05 material (axial orientation) specimens to 1.45 x 10 19 n/cm2 (Figure 5-1) resulted in a 30 and 50 ft-lb transition temperature increase of 105 and 120'F, respectively, and an upper shelf energy increase of 17 ft-lb when compared to the unirradiated data.

Irradiation of the vessel intermediate shell Forging 05 material (tangential orientation) specimens to 1.45 x 10 19 n/cm2 (Figure 5-2) resulted in a 30 and 50 f t-lb transition temperature increase of 100'F and 120'F, respectively, mwoussoae 5-3

and an upper shelf energy decrease of 20 f t-1b when compared to the unirradiateddata.III Weld metal irradiated to 1.45 x 1018 n/cm2 (Figure 5-3) resulted in a 30 and 50 f t-lb transition temerature increase of 35 and 40'F respectively and -

no upper shelf energy decrease.

Weld HAZ metal irradiated to 1.45 x 1019 n/cm2 (Figure 5-4) resulted in a 30 and 50 f t-lb transition temperature increase of 75'F and no upper shelf energy decrease.

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and ebow an increasing ductile or tougher appearance with increasing test temperature.

l Table 5-6 shows a comparison of the 30 ft-lb transition temperature (ARTNOT) increases for the various McGot.e Unit 2 surveillance materials with predicted increases using the Athods of NRC Regulatory Guide 1.99, Revision 2.I43 This comparison rhows that the transition temperature increase resulting from irradiation to 1.45 x 1018 n/cm 2 is less than predicted by the Guide for Forging 05 (HT. 526840) and weld metal.

5-3. TENSION TEST RESULTS The results of tension tests performed on Forging 05 (axial and tangential orientation) and weld metal irradiated to 1.45 x 1018 n/cm2 are shown in '

Table 5-7 and Figures 5-9, 5-10 and 5-11, respectively. These results show that irradiation produced a 8 to 14 Ksi increase.in 0.2 percent yield strength for Forging 05 and 5 to 11 Ksi increase for the weld metal. Fractured tension specimens for each of the materials are shown in Figures 5-12, b13 and 5-14.

A typical stress-strain curve for the tension specimens is shown in Figure 5-15.

sec . m eso "

5-4

5-4. COMPACT TENSION TESTS Per the surveillance capsule testing contract with the Duke Power Company, 1/2T - Compact Tension Fracture Mechanics specimens will not be tested and will be stored at the Hot Cell at the Westinghouse Science and Technology Center.

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i TABLE 5-1 i l

CHARPY V-NOTCH IMPACT DATA FOR THE NCGUIRE UNIT 2 i REACTOR VESSEL INTERMEDIATE SHELL FORGING 05 (HT. 526840) I IRRADIATED AT 550*F, FLUENCE 1.45 x 10 18 n/ce2(E>1.0MeV)

Temperature Impact Energy Lateral Expansion Shear Stanle No. G G & fft-lb) M (mils) (5)

Axial Orientation DT58 -46 (-50 9.5 7.0) 0.08 3.0 2 DT46 -18 0 24.5 18.0) 0.30 12.0 10

'1 DT49 -4 25 23.0 17.0 l 0.20 8.0 10 - '

DT52 10 50 27.0 20.0 0.36 14.0 15 DT56 30 86 40.5 30.0 0.71 28.0 20 DT53 38 58.5 43.0 0.89 35.0 35 DT51 38 63.5 47.0 0.71- 28.0 30 DT47 52 59.5 44.0 0.91- 36.0 35 DT54 66 (150 58.5 43.0 0.79 (31.0 40 '

DT57 66 80.0 (150 59.0 1.17 47.0 55 DT55 79 175) 70.5 52.0 0.94 37.0 50-DT59 93 78.5 58.0 1.22 48.0 60 DT50 107 114.0

84.0) 1.52 60.0 100 DT48 121 99.0 73.0) 1.50 59.0 100 DT60 135 97.5 72.0) 1.52 60.0 100

• Tannential Orientation DL47 -59 -75) 11.0 8.0 0.03 '1.0 2 DL57 -46 -50) 13.5 10.0 0.15 6.0 5

• DL52 -48 -50 11.0 8.0 0.13 5.0 2  ;

DL55 -18 38.0 28.0 0.48 19.0 20 DL48 -4 40.5 30.0 DL60 0.41- 16.0 20 10 78.5 58.0 0.53 21.0 3

35 DL54 30 87.0 DL59 64.5) 1.04 41.0 45 ,

31 142.5 105.0 1.70 DL49 67.0 85 38 1 68.0 50.0 0.91 DL58 36.0 45 38 100) 104.5 77.0 1.37 54.0 DL56 65 52 125) 124.5 92.0 1.47 54.0 DL50 75 66 150 101.5 119.0 2.08 82.0 DL46 95 93 (200 171.5 126.0 1.96 DL51 121 138, (77.0 100 (250 187.0 2.21 (87.0 100

} DL53 135 181.5 i (275 134. 2.13 (84.0 100 l j muueusso is 5-6

TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE WCGUIRE UNIT 2 REACTOR VESSEL WELO METAL A A HAZ WETAL IRRADIATED AT 550*F 18 FLUENCE 1.c.5 x 10 n/cm2 (E > 1.0 WoV)

Temperature Impact Energy Lateral Expansion Shear Samnle No. ,(* g, ,[*I), ,Q), (ft-lb) ,(gQ, fails) (5) ,

Wald M tal DW54' - 32 - 25 24.5 18.0 0.33 15 DW47 - 25 (13.0)

- 32 51.5 38.0 0.74 (20.0) 25 DW58 - 18 0 38.0 28.0 0.81 DW50 - 18 0

.(32.0 25 77.5 57.0 1.07 42.0 50 DW60 -

4 70.5

25) 52.0 0.91 36.0 50 DW57 10 50) 96.5 71.0 1.30 51.0 70 DW51 10 50 104.0 77.0 1.42 DW53 29 84 144,0 (56.0 75-(100.0 1.85 (73.0 90 DW46 38 100 139.5 DW55 (103.0 1.83- 72.0 90 38 100 141.0 (104.0 1.65 DW49 65.0 85 66 150 165.5 2.24 DW48 (122.0 88.0 100 79 175 154.5 114.0 1.98 DW56 *9 175 (78.0 100 188.5 130.0 2.26 (89.0 100 DW52 DW59 93 }200 192.5 10.0 2.13 (84.0 100 135 ( 275 199.5 147.0 2.21 (87.0 100 RAz Metal DH49 - 59 - 75 23.0 17.0) 0.43 1.

- 46 10 DH51 - 50 30.0 22.0) 0.23 10 DB58 - 46 - 50 32.5 DH48 - 3 . - 25 24.0) 0.36 1. 10 "

17.5- 13.0 0.20 8.0 10 DH52 - 18 0 40.5 30.5 0.53 21.0 DB57 - 18 20 0 46.0 34.0 0.50 22.0 DH54 4 25 25 85.5 63.0 0.91 36.0 DE47 10 45 50 99.0 73.0 0.94 DB63 30 37.0 65 86 124.5 92.0 1.37 54.0 DR59 38 80 100 78.0 58.0 1.24 49.0 DR56 60 38 100 123.5 91.0 1.35 DB60 66 53.0 45 150 134.0 99.0 1.65 65.0 DB45 93 100 i 200 143.5 106.0 1.75 69.0 DE55 100 '

121 250 126.0 93.0 1.37 54.0 DR50 100  ;

121 250 152.0 112.0 1.80 71.0 100 w.mmen ie 5-7 A

TABLE S-3 181STHtBEBITED 00ARPY 180'ACT TEST aFSA15 FW IEENIEE OR0li 2 BEACTOR VESSER. IIITEIREBIATE 58 ELL F005BII5 95 (III. 526900)

IIormalised Emeraies Test Obarpy Oberpy Maximum Prop Yield Time Maximus Time to Fractere Arrest Yield Flow Sample . Temp Meergy. Ep/A Load to Yield Load Maximus Lead Load Stress Streso Bd/A Be/A 2 Number f !1 (ft-lb) ___

(ft-lb/in ) (kisel (asec) (hips) (seec) (hiss) (kips) (ksil (ksi)

Asimi Orientation Df58 -50 7.0 56 17 30 2.65 80 2.85 SE 2.85 -

SS 91 9746 0 18.4 145 126 19 3.50 90 4.50 285 4.50 -

116 132 l

Df49 25 17.0 137 55 82 2.7 85 3.15 ISO 3.10 -

89 97 ..

D752 50 20.0 161 11 50 3.15 95 3.95 285 3.90 -

105 118 D756 86 30.0 242 200 42 3.30 95 4.55 435 4.50 -

IM 130 DT53 100 43.0 346 220 126 3.25 90 4.05 505 3.85 0.1 108 121 DTS1 100 47.0 378 183 195 2.55 80 3.45 500 3.45 -

83 98 DT47 125 44.0 354 231 123 3.40 115 4.50 520 4.3 0.35 112 130 DT54 150 43.0 346 199 147 3.20 85 4.35 445 4.3 1.25 106 125 f,

D757 150 50.0 475 248 227 3.15 120 4.05 800 3.90 1.10 105 130 9755 175 52.0 419 147 271 . 30 85 4.10 355 4.0 1.55 96 118 DT50 200 58.0 467 203 264 2.9 85 3.80 500 3.50 1.85 .95 110 DT50 225 84.0 676 171 506 2.15 90 3.25 510 - -

72 90 Df48 250 73.0 588 198 300 2.7 115 3.75 525 - -

88 106 DNO 275 72.0 580 165 415 2.7 125 3.75 455 - -

89 107 Tameential Grientation BL47 .a 8.0 64 1c 48 2.70 85 2.70 95 2.70 0.1 90 90 l BL52 -50 8.0 64 43 22 2.95 65 4.15 *20

, 4.20 -

SS 118 l 9057 -50 10.0 81 24 57 2.80 80 3.20 100 3.2 -

86 96 BLES 0 28.0 225 188 37 3.00 105 4.40 420' 4.40 -

120 133 L

BL48 2F 30.0 242 180 81 3.55 120 4.00 370 4.45 0.1 117 135 DL80 E 58.0 467 303 164 3.15 85 4.35 885 4.10 0.1 104 124 DL54 86 64.0 515 324 192 3.25 80 4.75 GEE 4.55 0.25 107 132 BL5S 88 105.0 845 253 502 2.55 90 3.80 670 2.70 0.25 84 102 l BL49 100 50.0 403 200 113 3.10 95 4.15 870 4.05 0.10 102 120 DL56 125 92.0 741 329 411 3.35 80 4.70 885 4.20 2.10 til 133 DL50 150 119.0 958 281 677 2.85 85 4.05 865 1.50 0.70 94 114 9046 200 126.0 1051 287 728 2.80 85 4.00 680 - -

93 113 DL51 250 138.0 1111 295 816 2.70 90 3.85 720 - -

90 100 l

DL53 275 134.0 1079 228 851 .20 80 3.25 675 - -

66 87 l

l DLS8 Computer malfunction prevented gathering of instrumented Charpy data for this specimes.

3887s4H9BS 80

___._.__-_-__.=.-.:-...=-;-._..

I

. . TAKE 5-4 IIISTMEEIITED CHAAPY IIFACT TEST aFM TS (W IEGIIIK IEIIT 2 j KACTM VESSEL KLB IETAI. M IIAZ IETE braalised Eneraies Test Charpy Charpy llaximum - Prop Yield Time huimum Time to Sample Temp Energy Ed/A Em/A Ep/A Load Fracture Arrest Yield Flow to Yield Load IInximum Load 4

Number f F1 (ft-lb) {ft-lb/in 2) (kine) {asec) (kin.) (seec) (kipe)

Load

{ kins)

Stress Stress (ksi) (keO hid h tal l' DW54 - 25 18.0 145 52 93 2.70 85 3.10 175 3.05 0.55 89 96 DW47 - 25 38.0 306 200 106 3.05 80 4,60 0 430 4.55 1.10 100 126

. DW58 28.0 225 89 137 2.7 85 3.20 j DW50 0 57.0 459 322 270 3.2 1.50 '9 98 137 3.50 90 4.55 670 4.45 1.15

". DW60 25 52.0 419 291 115 133

! 128 3.40 150 4.20 700 4.0 1.55 112 DW57' 50 71.0 572 311 261 3.25 125 85 4.40 865 4.25 2.80 DW51 50 77.0 620 243 377 2.90 108 127 90 3.45 865 3.10 1.65 j DW53 84 100.0 878 292 86 100 c,, 585 2.7 70 4.15 670 2.95 1.90 i

4 DW46 -100 103.0 829 301 89 113 529 3.05 75 4.35 655 3.80 2.4

! DW55 100 104.0 837 277 100 122 580 2.40 85 3.35 765 2.40 1.10 l DW49 150 122.0 982 301 80 95 '

681 2.95 85 3.85 725 i DW48 175 114.0 918 227 97 113 i 691 2 35 105 3.25 675 2.65 2.35 I

DW56 175 139.0 1119 365 754 3.15 78 93

! 95 4.20 825 105 DW52 200 142.0 1143 307 837 2.70 90 122 3.75 770 90 DW59 275 147.0 1184 293 891 2.65 130 3.70 107 780 - -

88 105 EAZ h tal DE49 - 75 17.0 137 130 7 3.30 85 4.45 DE51 22.0 295 4.30 -

100 128

- 50 177 69 100 2.00 .75 3.45 DE58 - 50 24.0 200 3.30 -

85 99 193 130 63 -3.55 80 4.65 DE48 - 25 13.0 105 275 4.65 -

118 136 43 62 3.35 80 4.15 .

DE52 0 30.0 125 4.10 0.20 III 124 242 115 127 3.10 95 4.10 DE57 0 34.0 285 4.10 0.80 102 119 274 131 143 2.7 85 3.80 DE54 25- 63.0 353 3.65 1.10 89 104 507 322 185 3.30 40 i DE47 50 73.0 4.90 610 4.55 1.95 110 136 588 263 325 3.25 85. 4.30 i DE53 86- 92.0 575- 4.10 2.85 108 126 741 242 499 3.20 85 D859 100 58.0 4.70 505 - -

106 131 467 232 235 3.30 100 4.45 DE56 100 91.0 733 510 - -

109 129 244 489 2.50 85 3.50 DE60 150 99.0 797 287 650 3.30 2.55 83 100 i 510 3.15 140 4.10 705 DH45 200 106.0 854 242 104 120 i 612 2.85 95 3.85 595 -

95 DH55 250 93.0 749 171 578 2.20 85 III

.[

DH50 3.25 510 - -

73 91 250 112.0 902 315 587 2.65 95 i

4.00 770 - -

87 110 ,

L

-- . - . . _ . . _ _ _~ .. _ _ _ _____________m _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ . _. _ . _ _ _ _ _ . _ . _ .

. ~ . - .- . . . . . . . . . .. - _- . _ . - ~ _ _ _ . . . - .. ._- _-.-. . _ . . . __ . . . _ . . _ . _ _.

Tan n s-s II .

TE EFFECT OF IARADIATEGII TO 1.4s a 19 n/cm2 (E > 1.9 IIeV)

AT 550*F GI TK INTCII Tm PREFERTIES E TE i

IICEIIIRE ORIIT 2 SIRVEILLAIEE CAPSIEE IETERIALS I

Average Average 35 mtI Average Average Energr ADoorpt son f 30 f t-ID Temp (*F) t_ateral Espansson Tesqp (*F) l 50 f t-lb Temp (*F) at Full Shear (f t-it) l Material Untrradtated Irradiated AT Untrradtated Irradlated AT Untrradtated Irradiated at unteradtated Irrastated A(ft-ID) i i

Forg1og 05 -25 80 105 15 115 100 25 145 120 94 77 -17 (Amtal)

I i forging 05 -75 25 10 0 -70 65 135- -60 60 120 156 136 -20 -

j , .(Tangent 1al) l e c -

teeld eletal -50 35 -45 -5 40 -25 15 40 133 138

• +5

StAZ teetal -90 -15 75 -70 20 90 -55 20 75 104 106 +2 1~

a 3887s/032990: 10 .

. . _ _ . . . . _ . . _ . ~ . _ . . _ . _ _ . _ . . . _ . . _ _ . _ . _ _ _ ._ . __ _.

L  !

u  :

I L

1 TABLE $-6 i

COMPARISf* OF WCGUIRE UNIT 2 I

REACTOR VESSEL SURVEILLANCE CAPSULE CHARPY. IMPACT TEST RESULTS WITH REGULATORY Gul0E 1.99 REVISION 2 PREDICTIONS ARTNDT('F) USE DECREASE (%)

h Fluence R.G 1.99 R.G 1.99 Capsule 19 2 Material 10 n/cm Meas. Prod. Meas. Prod.

I l

Forging 05 V 0.306 70 109 10 18 (Axial) X 1.45 105 126 18 27  ;

i Forging 05 V 0.306 65 109- 14 18 .

(Tangential) X 1.45 100 126 12 27 Wald Metal V 0.306 45 37 0 14 X 1.45 35 42 0 21 HAZ Metal V 0.306 55 -

6 -

X 1.45 75 -

0 -

t t

l l

)

l I

seer. m menie 5-11 l

l TAKE 5-7 TDISILE PROPERIIES FER IIDENIE WIT 2 IIEACTOR VF%SFB SI5lWEILUWICE CAPSEE BIRTERIAL IRRASIATED TO 1.45 x 10 0

m/m2 (E > 1.0 80mW) at $58"F Test 0.25 Yield Ultimate Fractere Freetere Fractere Esifera- Total Beda: tion Searle Temp. Strength Strength Lead Stress Strength Blongation Blongation im Area Material Ilumber M (kei)- (ksi) (his) iksil (keil (W) (5) (5)

Forging 9710 80 78.4 98.8 3.45 171.8 70.3 12.0 24.8 58 (Axial M11 300 87.7 91.7 3.45 135.8 70.3 10.5 19.5 48 -

y Orientatioe) 9712 560 87.7 94.7 3.00 203.7 73.3 9.8 18.8 42 Forging BLIO 78 77.4 97.8 3.00 188.7 81.1 12.8 25.9 82 (7aagential BL11 300 70.8 9(r.7 2.95 231.1 80.1 10.5 21.5 82 Orientation) BLIS 560 87.7 90.7 3.20 321.9 85.2 9.8 20.8 54 Weld 9W10 82 79.5 97.8 3.40 213.2 80.2 11.3 25.4 88 BW11 300 71.3 83.5 2.45 171.2 48.9 9.8 22.7 88 BW12 550 71.8 88.8 2.75 253.8 58.0 10.5 22.2 84 4?

3907:032350 se

. . ~ . . . . . . . . , . . . . . . . . . . - -. . . . . - - - - . . . . -

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- . - - _ _ - . _ - - - - - - . - _ - - - - - - _ _ - - _ _ _ _ _ - _ _ ~ . ..

4 4

(

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' -150 - 100 - 50 0 50 ' 100 150 200 250 I I I I i i

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$ 80 -

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a 10 - -

200 l g120 -

i Ig

~ 100 -

Unirradiated f 80 -

. - 120 :; ,

2 -

120'F 80 g

20

' 7 Irradiated ( 550'F) 19 2 -

l.45 x 10 n/cm

- 200 -100 0 100 200 300 40 500 Temperature (*F) 4

, FIGURE 5-1 CHARPY V-NOTCH IMPACT DATA FOR MCGUIRE UNIT 2 REACTOR VESSEL INTERMEDIATE SHELL FORGING 05 HT, 526840 (AXIAL ORIENTATION) l w.masso ie

' 5-13

. _ _ , . . _ _ . _ _ . _ . ...- ... .. _.____ _ _ _.__ __ _ _ _..___. _ .__ _ __ _I

I ll

( ' C) '

-150 -100 -50 0 50 100 150 200 250  :

j i I I i ' ' '

100 -

'2

= :;

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• i

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5e - 8 "

20 -

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1

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=

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ig

~ 100 -

Irradlated ( 550*F )

b e 80 -

8

• 19 2 -

120 C

1. 45 x 10 n/cm 5

• 60 -

120'F -

80 100*F  ! -

e 0 ' i i '

i i i 0

- 200 -100 0 100 200 300 40 500 Temperdure ('F)

FIGURE 5-2 CHARPY V-NOTCH IMPACT DATA FOR MCGUIRE UNIT 2 REACTOR VESSEL .

INTERMEDIATE SHELL FORGING 05 HT 526840 (TANGENTIAL ORIENTATION) i mer.mmeeio 5-14

('C)

-150 - 100 -50 0 50 100 150 200 250 i i i ' ' '

100 -

=

2'

' 2' '

$ 80 -

3' -

g@ -

a .

mm -

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240 160 -

I <

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Unirradiated

.120 -

160 100 -

2 80 - Irradiated ( 550*F) -

120 0 5 19 2

1. 45 x 10 n/cm

. . . -40* F -

80 C -

r- 35' F 20 -

0 ,2 i i i i i i i 0

- 200 -100 0 100 200 300 00 500  !

Temperature ('F)

FIGURE 5-3 CHARPY V-NOTCH IMPACT DATA FOR WCGUIRE UNIT 2 REACTOR VESSE METAL w.mmee io 5-15

1

, ( ' C) l -150 -100 - 50 0 50 100 150 200 250 100 i ' ' '

i'L2

=

j2,3' l f 80 -

j* -

l mm -

• N -

2s .

0 ' '. ' ' i e i i

1. i i i i i i 2.5 i i i

.j 80 -

e . A 6 2.0 *

~

1.0 s" ': d9'0' M 2 i

,' I i i i i i i i 180 2@

160 l

_ 10 a

200 y120 -

100 - Unirradiated

• 8* -

160 g 80 -

120 0

1. ~

80 75'F -

4 -

, _fa g Irradiated ( 550'F) 2 -

O 20 -

1. 45 x 10 n/cm

- 200 -100 0 100 200 300 00 500 Temperature ('F)

FIGURE 5-4 CHARPY V-NOTCH IMPACT DATA FOR MCGUIRE UNIT 2 REACTOR VESSEL WELD HEAT-AFFECTED ZONE METAL ,

mewousmae 5-16

l

{" ..

( .

M58 M48 Met Mt3 M58 l

l I .

N=

i

-[

E.. M53 Ik E\?<,W' DT51

• M47. M54 DT57 l

' l T -

4j:gl ,

M56 DT50 DT50 Dfes Droo l

l

! FIGURE 5-5 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 2 REACTOR VESSEL INTERMEDIATE SHELL FORGING 05 HT 526840

(AX1ALORIENTATION) wr.mmeo is 5-17 m.ma

S Q &'

, , _ D'f

. $$d)$ "

. $e. a u@sQ ~.

- p- I

. 1 FIGURE 5-6 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 2 REACTOR VESSEL INTERMEDIATE SHELL FORGING 05 HT. 526840 (TANGENTIAL ORIENTATION)

=c. m- . .

5-18 l

._--_-_-__---_______-__--______---------------?"* -

4a- w- ------wm4-4%-- , _ e._-w---m- ---

am- .--mm--mamm-----_mam.-_ .-mmwe_-

_ w-m- --m_w _s wm-es.m-m. mew _4-sw_..,

v+

~

g l

l I

DW57 DW51 D953 998S'y ' ', DW55 DW49 DW48 DW58 D953 DW59 1

FIGURE 5-7 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 2 REACTOR VESSEL WELD METAL w .,o n o io 5-19 l

RW-22621

. _ _ ~ . - _ _ _ . . _ _ . . - . _ _ _ _ . . . _ _ _ _ .-- - - - -'

l

l l

l

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g. l

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.M M

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~ . ,

l. c ,$M 4'.*.iNEl$l,',,'y*ftf,b*b.l(,

\ .... .

sagyu

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, i ti -

FIGURE 5-8 CHARPY IMPACT SPECIMEN FRACTURE SURFACES FOR MCGUIRE UNIT 2 REACTOR VESSEL WELD HA2 METAL 3Mfs/032990 3 5 ^0 M*22022

m -

t l

l

. l

'b

'C

- 50 0 50 100 150 200 250 '300 120

' ' ' ' ~

110 - l #0-100 -

700 n 5, - -

1

' ~

g _ Qnate Tensile Strength ,

@j E

g 70 - N ~<, 0 500<~

si M -

_ ,0. 2 % Yield Strength; _ _g.

2 - *

o. ,

~

40 i i i- i l l- 1 i300 1.egend :

Open Points - Unirradiated -

2 Closed Points - Irradiated at 550*F ( l.45:x l'019 n/cm 3 80 i i , ,_ , , , , ,

70 - . i Reduction in' Area -

E

-g -

.o i

N

~

~

} 40 -  :

] 30 -

Total Elongation - _

20 -

a a n - -

10 - Y p-  % _

0 i i i i Uniform Elongption ,

-100 0 100 200 300 400 500. @.

Temperature ( *F) 3 ' FIGURE 5-9 TENSILE PROPERTIES FOR MCGUIRE UNIT 2 REACTOR VESSEL INTERMEDIATE SHELL FORGING 05 HT. 526840 (AXIAL ORIENTATION)

F 21

I

.C '

- 50 0 50 100- 150 200 250 '300

-120 , , , , , , g i _

110 -

Ultimate Tensile Strength 7

n - 90 N

~

2 %_

600g

^7 A:

0 500 60 -

-G-- o 50 -

3 _

400  ;

, O. 2 % Yield Strenath 40 i i i i i i i- 300 l.egend :

Open Points - Unirradiated Closed Points - Irradiated at 550*F ( 1,45 x- 10 19 n/cm2)

N I I I: I -l- l l l 70 - 8 ^

r

=

_g

, s0

- 50 .

x y@ -

] 30 - Total Elongation 20 - ' - -

9 -

10 - 'e 9 y ._

0 iUnifortg Elonga, tion , , , ,

-100 0 100 20; 300 400 500 600 Temperature ('F) i FIGURE 5-10 TENSILEr'ROPERTIES FOP. MCGUIRE UNIT 2 REACTOR' VESSEL INTERMEDIATE SHELL-FORGING 05 HT. 526840 (TANGENTIAL ORIENTATION) i se n . m m o 5-22 ,

i i L f i

1 1

.C

- 50 0 50 100 150 200 250 300 120 i i ' ' ' ' ' -

110 - 800

_.100 Ultimate Tensile Strength 700 m 90 h 80 1- i 1 I h70 m

5 g

500 No g _

..400 50 - 0. 2 % Yield Strength-40 I i 1 i i i i' -

t 300 l

Legend :

j Open Points - Unirradiated 2

Closed Points - Irradiated at 550*F ( l.45 x 1019 n/cn 80 i i i i. , , , i 70 - C -0 g _

-e w -e liii

~x 50 -

E 40 -

]30 -

Total Elongation -

20 -

~ %--br o_ a _

10 - e- -o o _

O I I i J Unifor,m Elongation ,

-100 .\

0 100 200' 300 400 500 600 i Temperature ( *F)  !

1

.1

'l 1

• l FIGURE 5-11 TENSILE PROPERTIES FOR MCGUIRE UNIT 2 REACTOR VESSEL WELO METAL  !

l

',. i mwo-lo 5-23

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• g nxtm v>g FIGURE 5-12 FRACTURED TENSILE SPECIMENS FOR MCGUIRE UNIT 2 REACTOR VESSEL INTERMEDIATE SHELL FORGING 05 HT. 526840 (AXIAL ORIENTATION)

=>.<oim o 5-24 151-22623

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Specinea IE12- gy FIGURE 5-13 FRACTURED TENSILE SPECIMENS FOR MCGUIRE UNIT 2 REACTO INTERMEDIATE SHELL FORGING 05 HT. 526840 (TANGENTIAL ORI av.,um o 5-25 RW-22624

msg - - - -- - , _ , - - , - - - , - ,- - - - - - - ------ -.

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• Q . , <g dr, e  %%ws e' .

b I' ., d

. g.- ,egi t-044&g -

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Specimen DW12 geg*y l

l l

FIGURE 5-14 FRACTURED TENSILE SPECIMENS FOR NCGUIRE UNIT 2 REACTOR VESSl 1

WELD METAL l

m r.eu m io 5-26 RW-22625 l

i-i i

t i

100 90 - --  %

80 -

70 -

g 80 -

w_

• g. .

30 -

20 -

1 0-SPEC.DL.12 SSOF .

0 8 8 8 i i i i i , , , ,

0 0.04 0.08 0.12- 0.16- U.2 0.24' O.28 STRAIN, IN./IN.

FIGURE 5-15 TYFICAL STRESS-STRAIN CURVE FOR TENSION SPECIMENS sw.mm 5. . g ,

I l

l SECTION 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

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 contained in each of the surveillance capsules. The latter information is derived solely from analysis.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally.been accepted for development of i

damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, hcwever, it has been suggested that an exposure model that accounts for differances in neutron energy: spectra between surveillance capsule ' locations and positions within the vessel wall could lead 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 fluenc6 toward an energy 11ependent damage ifunction for data correlation, ASTM Standard Practice E853,

' Analysis and Interpretation of Light Water Reactor Surveillance Results,"'

recomends reporting displacements per iron atom (dpa) along with fluence 1 mr.mmo 61

(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 L in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlemer.+ gradients through the thickness of the i 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 X. Fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV), fast neutron fluence (E > 0.1 Nev), and iron atom displacements.(dpa) are established for 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 itselE Also uncertainties associated with the derived exposure parameters at th0 surveillance capsule and with the projected exposure of the pressure vessel are provided.

4 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the reactor geometry at the cort midplane is shown in Figure 4-1. Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reacter vessel surveillance program. The capsules are located at azimuthal angles of 56*, 58.5',-124 ,'236', 238.5',

and 304' relative to the core cardinal area as shown in Figure 4-1.

A plan view of a dual surveillance c'apsul'e holder attached to the neutron pad is shown in Figure 6-1. The stainicss 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.

F,om a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the distrioutiot, of neutron movanneo to 6-2

~

11 flux and the neutron energy spectrum in the water annulus between the neutron pad and the reactor vessel. In order to properly determine the neutron environment at the test specimen locations, the capsules themselves must be  !

included in the analytical model.

In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets'of transport calculations were carrie'd out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters (p(E > 1.0 Wev,) e(E > 0.1 Nev),

and dpa) through the vessel wall. The-neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e., dpa/p(E > 1.0 MeV), within the pressure vessel geometry.

The relative radial gradient information was required to permit the projection of measured exoosure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T- and 3/4T locations.

The second. set of calculaLions consisted of a series 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 functier.s generated from these adjoint analyses providea 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 first 5 cycles of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted #or the offects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased, w.maseo" 6-3

H l

The absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the forward calculation provided the means to:

1. Evaluate neutron dosimetry obtained from surveillance capsule locations,

2. Extrapolate do 5etry 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.

The forward transport calculation for the reachr model sunnarized in Figures 4-1 and 6-1 was carried out in R, e geometry using the DOT two-dimensional-discrete ordinates code (5) and the SAILOR cross-se: tion library (6).- 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

!' discretization was modeled with an S 8 rder of angular quadrature.

The reference core power distribution utilized in the forward analysis was derived from statistical studies.of long-term operation of Westinghouse 4-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel manegement strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to-plant and )

i- cycle to cycle variations in peripheral power was-used. Since it is unlikely

that a single reactor would have a power distribution at the nominal +2a level for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.

w uo n secao 6-4

All adjoint analy;es were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library.

Adjoint source locations were chosen at several a:imuthal locations along the j pressure vessel inner radius as well as the geometric center of each '

surveillance capsule. Again, these calculations were run in R, e geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, e (E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as:

R (r, 0) = /r #e #E I(r, 6, E) S (r, e, E) r dr de dE where: R(r,0) =

$ (E > 1.0 MeV) at radius r and azimuthal angle e I (r, e, E) =

Adjoint importance function at radius, r, azimuthal

, angle e, and neutron source energy E. 1 S (r, e, E)

~

=

Neutron source strength at core location r, e and energy E.

Although the adjoint importance functions used in the McGuire 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 loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio-of dpa/4 (E > 1.0 MeV) is insensitive to changing core source distributions.

In the application of these adjoint important functions to the McGuire Unit 2 reactor, therefore, calculation of the iron displacement rates (dpa) and the neutron flux (E > 0.1 WeV) were computed on a cycle specific basis by using dpa/$ (E > 1.0 MeV) and # (E > 0.1 MeV)/e (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific # (E > 1.0 MeV) solutions from the individual adjoint evaluations.

w.mmio

) 6-5

=

The reactor core power distribution used in the plant specific adjoint calculations was taken from the fuel cycle design report for the first five operating cycle of McGuire Unit 2 (7 thru 11). The relative power levels in fuel assemblies that are significant contributors to the neutron exposure of a 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 iilustrated in Figure 6-2.

3 elected results from the neutron transport analyses performed for the McGuire n Unit 2 reactor are provided in Tables 6-1 through 6-5. The data listed in g these tables establish the means for absolute comparisons of analysis and reeasurement for the capsule irradiation period and provide the means to aorrelate dosimetry results with the corresponding neutron exposure of the pressure vessel wall.

In Table 6-1, the calculated exposure parameters (e (E > 1.0 MeV),

e (E > 0.1 WeV), 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. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure parameters are listed for both the design basis and the cycle 1 through 5 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 presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the

, vessel inner radius to the gradient data given in Tables 6-3 through 6-5.

e

- w oonasoio 6-6

1 1

l For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' l aximuth is given by: -l

'1/4T(45')

=

1. (220.27, 45') F (225.75, 45')

where =

Projected neutron flux at the 1/4T_ position

'1/4T(45')

on the-45' azimuth e(220.27,45') = Projected or calculated neutron flux at the i vessel inner' radius on the 45' azimuth.

\

F(225.75,45') =

Relative radial distribution function from' Table 6-3. , j

.Similar expressions apply for exposure parameters in terms of j l *(E > 0.1 MeV) and dpa/sec.  !

n l

6.3 NEUTRON DOSIMETRf l'

l The passive neutron sensors included in the McGuire. Unit 2 surveillance program are listed in Table 6-6. Also given in Table 6-6'are the primary l

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

, (v (E > 1.0 Mov), e (E > 0.1 MeV), 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 I form, were placed in' holes drilled in spacers at several axial levels within the capsules. The cadhium-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.

w ou moao 6-7

l Rather, the activation or fission process is a measure of the integrated {

effect that the time and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux. level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following. variables are of interest:

l o The specific activity of each monitor.

l o The operating history of the reactor.

o The energy response of the monitor.

o The neutron energy spectrum at the monitor location, o The physical characteristics of the monitor.

The specific activity of each of the neutron monitors was determined using established ASTM procedures (12 through 25]. Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the McGuire Unit i reactor during cycles 1 through 5 was obtained l from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable period.

The irradiation history applicable to capsule X is given in Table 6-7.

Measured and saturated reaction product specific activities as well as measured full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7.

Values of key fast neu k on exposure parameters were derived free the measured reaction rates using the FERRET least squares adjustment code (26]. The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum 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 i adjusted spectra.  !

1 I

l l

wr.mmenao 6-8 I I

In the FERRET evaluations,La log normsl least-squares algorithm weights both the a priori values and the measured cata in accordance with the assigned uncertainties and correlations. In go.1eral, the measured _ values f are linearly related to the flux e by some response matrix A:

f i (s,a) = I A (s)

,g( a) g ig 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 e, g o g=9 ' g relates a set of measured reaction rates R g to.a single spectrum # by 9

the multigroup cross section ogg. (In this case, FERRET also adjusts the cross-sections.) The log normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties.

l 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 (27). This procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where-l group boundaries do not coincide. The 620 point spectrum was then easily collapsed to the group scheme used in FERRET.

The cross-sections were also collapsed into the 53 energy group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting j functions. The cross sections were taken from the ENDF/B-V dosimetry file.

( Uncertainty estimates and 53 x 53 covariance matrices wore constructed for l each cross section. Correlations between cross sections were neglected due to data and code limitations, but are expected to be unimportant.

f w .m moio 6-9 i

^

1 L

4 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 esses, as for the cross sections, a multigrcup covariance matrix is used. More of ten, a simple parameterized form is used:

M gg,=Rf+R 9 R,P g gg, where Rg specifies an overall fractional normalization uncertainty (i.e.,

complete orrelation) for the corresponding set of values. The fractional uncertair. ties Rg specify additional random uncertainties for group g that are correlated with a correlation matrix:

Pgg, = (1 - 0) ogg, + 0 exp (- (a-c')2) 2r The first term specifies purely random uncertainties whil's the second term describes short-range correlations over a range r (8 specifies the strength of 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 6 is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R. E.

Maerker(28). Maerker's results are closely duplicated when T = 6. For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.

Results of the FERRET evaluation of the capsulo X dosimetry are given in Table 6-9 The data summarized in Table 6-9 indicated that the capsule received an integrated exposure of 1.45 x 10 19 n/cm2 (E. > 1.0 MeV) with an associated uncertainty of + 8%. Also reported are capsule exposures in terms of fluence (E>0.1MeV)andironatomdisplacements(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.

wuone cao 6-10

u i i

A summary of the measured and calculated neutron o'xposure of capsule X is presented in Table 6-12. The agreement between calculation and measurement  ;

falls within + 11-17% for all exposure parameters listed.

Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13. Along with-the current (4.16 EFPY) exposure derived from the capsule X measurements, projections are also provided for an

~

exposure period of 16 EFPY and to end of vessel design life (32-EFPY). The time averaged exposure rates for the first 4.16 EFPY of operation were used to perform projections beyond the end of cycle 1 through 5 exposure period.

in the calculation of exposure gradients for use in the development of heatup and cooldown curves for the McGuire Unit 2 reactor coolant systa'n, 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 RTNDT vs.

fluence trend curves, dpa equivalent fast neutron fluence levels for the 1/4T L and 3/4T positions were defined by the relations.

+' 1/4T = 6 (Surface) ( dpa (1/4T) )

dpa (Surface) 4' 3/4T = 4 (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 factors are listed for each of the McGuire Unit 2 surveillance capsules. These data may be used as a guide in establishing future withdrawal schedules for the remaining capsules.

mwomo 6-11

i 4

*T F' u ' ' - MCHARPY SPECIMEN '

.j i g

d

'ljs  !

%% T T N-N l / 7 X N .N g 33 vg 3 NEUYMON PA0 :

\ \ Aw gggN \ :g' y A s - g g g g g .

m 9

Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule w.4mio 6-12

l l

g j g' g.34- .a.98 A.77 ' DESIGN >AARIS I 8.74 R.77 -8.89 8.57 C)CLE 1 1.88 1.82 9'. g5 p.78 ' CYCLE.2 9.7) I9.59 0.79 8.54 CYCLE 3 9.92- 8.g9- e,77 9.49 CYCLE'4 9.89. 9.87 6.78 9.48 CYCLE 5

)

1.92 1.18 1.09 1.05 1.18 0.71  !

8.99 1.05, 't.96 8.98 9.92 8.52

'9.86 1.28 9.93: 1.21- 1.83 8.79 9.92 1.19 9.52 1.12 9.88 0.54 l 8.95 1.20 1.04 1.08 0.71 8.43 ,

e,93 1.17 .- l . 83 1.10 9.71 0.38  ;

1.95 0.87 0.97 1.07 1.00 1.95 1.13 1.10 1.11 1.05 0. 98 -- 8.97 a.92 9.80 8.91' 8.99 I.88 8.90-1.24 1.el 1.28 1.01 1.09- 1.68 1.86 1.08 1.31 1.14- 1.22 8.88 1.64 1.00 1.27 1.13 1.28 '8.83 y

1,99 1.06 0.88 1.10 1.04-l.14 1.17 1.13 1.11 1.17 3 6.85 -9.91 1.31 1.38 1.19' '

l.g4 1,29 9.95 1.,36 - 0.99' 1.11 1.32 1.02 1.13 1.19 8.97 1.29 0.97 1.17 1.33 e.99 1.04 1. I'2 9.92 1.18 lalE 1.17 1.13 8.96 - 1.14 1.83 'I.14 9.91 ' 6. 96 1.38 1.82 1.30 9.88 l ~. 99 1.12 lose 1.98 1.39 9.98

. Figure 6-2. Core Power Distributions Used in Transport Calculations for McGuire Unit 2 w .masseno 6-13

TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES  !

AT THE SURVEILLANCE CAPSULE CENTER

.' IRRADIATION _e (E > 1.0 Mev) p(E > 0.1 Mev) dpa/sec TIME 2 (n/cm _,,c) g,fc ,2_,,c3

. CYCLE (EFPS) 31.5* 34.0* 31.5* 34.0* 31.5* 34.0* '

DESIGN BASIS 1.11 x 10 11 1.29 x 10 II 4.88 x 10 1I 5.93 x 10 Il 2.21 x 10 -10' 2.62 x 10 -10 7

CYCLE 1 3.24 x 10 8.33 x 10 10 9.50 x.10 10- 3.66 x 10 11 4.37 x 10 II 1.66 x 10 -10 1.93 x 10 -10 .

. CYCLE 2 2.16 x 10 7 9.69 x 10 10 -

1.11 x 10 1I 4.26 x 10 ll 5.10 x.lo ll 1.93 x 10 -10 2.25 x 10 -10 CYCLE'3 2.31 x 10 7 8.40 x 10 10 9.73 x-10 10 -3.69 x 10 II 4.47 x 10 II 1.67 x 10.10

~

1.97 x 10-10 ,

2.66 x'10 7 7.38 x 10 10 8.34 x 10 10 3.24 x 10 II 1.69's 10-10 CYCLE 4 '3.83 x 1'0 Il' 1.47 x 10-I0 CYCLE 5 2.77 x 10 7 7.09 x 10 10 7.89 x 10 10 3.12 x.10 11 3.63 x 10 11 1.41 x 10 -10 1.60 x 10 -10  ;

t i

y 1

w .m mene

- _ _ _ _ _ _ _ _ - ._______ ~- -

+- n: . -~ ~en - - - . -., . , . . - .

+

TABLE 6-2 CALCULATED FAST NEUTRON EXPOSURE PARAMETERS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE-2

1. (E > 1.0 Mov) { n/cm ,,e)

- O' 15' 30' 45' I

1.45'x 1010 2.21 x 10 10 1.69 x 10 10: 2.44 x 10 10 DESIGN BASIS Cycle 1 l'.08 x 10 10- 1.62 x~10 10 1.27 x'1010 1.80 x 10 10 1.27 x 10 10: .1.92 x 10 10- 1.'48.x 10 10 l Cycle 2

~

2.11 x'10 10.

1.03 x 10 10 1.58 x 10 10 1.28 x 10 10 Cycle 3 1.86 x 10 10 i

Cycle 4 1.06~x 10 10 - 1.b7 x'10 10 1.14 x 10 10 1.57 x 10 10 '

Cycle 5 1.02 x 10 10 -

1.54 x 10 10 1.10 x 10 10 1.47 x 10 10 l.

2

- # (E > 0.1'Mov) ( n/cm .,,c) _

O' '15' 30' 45' -

i DESIGN BASIS 3.02 x 10 10 4.66 x 10 10 4.25 x 10 10 6.11 x 10 10

Cycle 1 2.25 x 10 10 3.42 x 10 10 3.19 x 10 10 4.51 x 10 10

) Cycle 2 2.65 x 10 10 4.05 x 10 10 ;3.72 x.10 10 '5.28 x 10 10 Cycle 3 2.15 x.10 10 . 3.33 x 10 10 3.22 x 10 10 J4.66 x 10 10.

2.21 x 10 10 3.31 x 10 10' Cycle 4 10 3.93 x 10 10 2.87 x~.10 Cycle 5 2.12 x 10 10 3.25 x 10 10 2.77 x 10 10 3.68 x 10 10 4

dpa/sec '

0* 15' 30'- 45' DESIGN BASIS 2.25 x 10 ~11 3.41 x 10 -11 2.73 x 10 -11 3.88 x 10 ~11' Cycle 1 1.68 x 10 -11 2.50 x 10-11 2.05 x 10'11 2.86 x'10 ~11 l Cycle 2 1.97 x 10 -11 2.96 x 10 ~11 2.39 x 10 -11 3.35 x 10 -11' Cycle 3 1.60 x 10 -11 2.44 x 10 ~11 2.07 x 10 ~11 2.96 x 10 -11 4

Cycle 4 1.65 x 10 ~11 2.42 x 10 ~11 1.84 x 10~11 . 2.50 x'10 ~11 i Cycle 6 '1.58 x 10 ~11 2.38 x 10 -11 1.78 x 10 ~11 2.34 x 10 -11 a

wenmeno 6-15

-~ w - , , w ....s+- , - - , -. . - , - -+

( a

'r

'i l

TABLE 6-3 -

RELATIVE RADIAL DISTRIBUTIONS OF_ NEUTRON FLUX _(E.> 1.0 MeV)

WITHIN THE PRESSURE VESSEL WALL Radius t

-(em)- O' 15' 30' 45' - '

t 220.27(1) 1.00 1.00 1.00 1.00 220.64 0.979- 0.979. 0.980- 0.979 231.66- L891 0.891 0.893- 0.889- I

. 222.99 0.771 0.769- 0.773 0.766  !

224.31 0.655 0.652 .0.658 0.648 225.63 0.552 0.549 0.555 0.543 .

226.95 0.463 0.459 0.467 0.452 228.28 0.387 0.383 0.390 0 376 L 229.60 0.322 0.318 0.326 0.311*

230.92 0.268 0.263 0.271 0.257 232.25 0.222 0.218 0.225 0.211 i 233.57 0.183~ 0.180 0.187 0.174 234.89 0.151 0.148 0.155- 0.142- t 236.22 . 0.125 0.121 0.128. 0.116 237.54 0.102 0.0992 0.105 0.0945 238.86 0.0831 0.0807 0.0862 0.0762 240.19 0.0673 0.0650 0.0703 0.0606-241.51 0.0539 0.0512 0.0567 0.0472 l 242.17(2) 0.0508 0.0477 0.0536 0.0438 -

50TES: -1) Base Metal Inner Radius

2) Base Metal Outer Radius w r. - ia 6-16  !

% r i.

TAB'LE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV)

WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0' 15' 30' 45' 220.27(1) 1.00 1.00 1.00- 1.00.

220.64- 1.00. 1.001 1.00 1.00 221.66 1.00 1.00 1.00 0.995. i 222.99 0.974 0.966 0.982 0.956- l 224.31 0.928 0.915! 0.938 0.902 >

225.63 0.875 0.859: 0.886- 0.843 226.95 0.819 0.802 0.832 0.782 228.28 0.762 0.743 0.777- 0.722 1 229.60 0.705 0.686 0.721 0.663 230.02 0.649 0.629 0.665 0.605 232.25 0.594 0.575 . 0.611' ~0.549 233.57 0.540 0,522 0.558 0.495.

234.89 0.488 0.470 0.506 0.443 236.22 0.436 0.421 0.455 0.392 237.54 0.386 0.373 0.406 0.343-238.86 0.337 0.326 0.358 0.296 240.19 0.290 0.280 0.310 0.248 241.51 0.244 0.232 0.261 0.201 242.17(2) 0.233 0.219 0.249 0.188 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius mwomo io 6-17

i I

TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa)  !

WITHIN THE PRESSURE VESSEL WALL-

]

1 Radius (cm) O' 15' -30' 45' 220.27(1) 1.00 1.00 1.00 1.00 220.64 0.982 0.982 0.986- 0.984-221.66 0.911 0.910 0.923 0.915 222.99 0.813 0.812 0.837 0.821 l-224.31 0.721 0.718 0.751 0.730

. 225.63 0.637 0.633 0.673 0.646 226.95 0.562 0.558 0.602 0.572  !

228.28 0.496 0.491 0.539 0.505 229.60 0.438 0.433 0.481- 0.447 ,

230.92 0.387 0.381 0.430 -0.394 I 232.25 0.341 0.335 0.383 0.347 a 233.57 0.300 0.295 0.341 -0.305 234.89 0.263 0.258 0.302 0.266 236.22 0.230 0.225 .0.267 0.231 237.54 0.199 0.195 0.234 0.199 l 238.86 0.171 0.168 0.203 C. 69 240.19 0.145 0.142 0.174 0.140 241.51 0.121 0.117 0.146 0.113 '

242.17(2) 0.116 0.110 - 0.140 0.106 NOTES: 1) Base Metal Inner Radius i

2) Base Metal Outer Radius i

e 1

l w .m assoao 6-18 W

TABLE-6-C-i NUCLEAR PARAMETERS FOR NEUTRON FLUX WONITORS l

Reaction. Target Fission Monitor of Weight Response Product Yield 1 Material Interest Fraction Range Half-Life (%)

Copper Cu63(n.a)Co60 0.6917 E> 4.7 MeV 5.272 yrs Iron FeI4(n.p)Mn54 0.0582 E> 1.0 MeV- 312.2 days-i Nickel NiS8(n.p)CoS8 -0.6830 E> 1.0 MeV 70.90 days Uranium-238* U238(n.f)Cs137 1.0 E> 0.4 MeV 30.12 yrs. 5.99 Neptunium-237* Np237(nf)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  !

Cobalt-Aluminum * ' CoS9(n,r)Co60 0.0015 E< 0.015 MeV 5.272 yrs - -

I

• Denotes that monitor is cadmium shielded. 1 n u.in meeio 6-19

-I TABLE 6-7 IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE X

-IRRADIATION DECAY I PJ- TIME TIME MONTH YEAR (MW)- PJ/ MAX (DAY) (DAY)- ,

5 1983 49.7 0.0146 25 2418~

h 6- 1983 390.8 0.1146 30 2388 L

7 1983 0.7 0.0002 31 2357-8- 1983 959.7 0.2813. 31 2326-l 9 1983 1783.1 0.5227 30= 2296 3 l

10 1983 1891.2 0.5544 31 2265 l 11. 1983 2330.2 0.6831 - 30 2235 12 1983 2200.7 0.6452' 31 2204 r 1 1984 578.8 '0.1697' 31 2173

. 2 1984 2742.2 0.8039 29 2144 l 3 1984 3100.5- 0.9090 31. 2113  ;

4 1984 3197.1 0.9373 30 2083 5 1984 2865.0 0.8399 31 2052- -i 6 1984 3309.0 0.9701 30 2022; 7- 1984 2095.3 0.6143 31- 1991-8 1984 825.8 - 0.2421 31 1960 9 1984 3156.3- 0.9253 30 1930 10 1984 3038.2 0.0907 - 31' 1899 11 1984 2531.6 0.7422 30 1869 l 12 1984 2311.2- 0.6776 ~31 i

1838' 1 1985 2618.6 0.7677 31 1807: ~;

2 1985 0.0 0.0000

.28 1779 3 1985 0.0 0.0000' 31 1748 i 4 1985 0.0 0.0000 30- 1718 5 1985 2169.2 0.6360 31 1687 i

I l

!l l ,

i'

w.ma 6-20

. _ - . _ _ _U

TABLE 6-7 (Cont'd) q IRRADIATION HISTORY OF NEUTRON SENSORS ~ I CONTAINED IN CAPSULE X  :

IRRADIATION DECAY l

PJ TIME TIME MONTH YEAR (MW) fJ/ MAX .(DAY) (DAY) 6 1985 2889.6 0.8471' 30 1657 7 1985 1276.i 0.3743 31 1626 s 8 198$ 2282.6 0.6692 31 1595 1 9 1985 3405.5- 0.9984 30 '1565i i 10 ~1985 3148.1 -0.9229 .-31 1534 11 1985 3286,8 0.9636 30 1504:

12 1985 1785.3 0.5234 31 1473 1 1986 3070.2 0.9001' 31 1442 4 2 1986 3367.0 0.9871 28 1414  :'

3 1986 1439.1 0.4219 31- 1383 4 1986 0.0 0.0000 30 1353 5 1986 0.0 0.0000 31 1322 6 1986 137.3 0.0403 30 .1292 t 7 1986 3252.5 0.9535 31 1261 8 1986 3085.1 0.9045 31 1230 9 1986 3410.9 1.0000 30 1200 10 1986 3049.4 0.8940 31 1169 11 1986 1159.3- 0.3399 30- 1139 12 1986 3413.9 1.0008 31 1108 1 1987 2610.5 0.7653 31 1077 1987 3117.5 0.9139' 28 1049 J 1987 3410.1 0.9997 31 1018 4 1987 * "1. 2 0.9737 30 ~ 988 i 5 1987 28.0 0.0082 31 -957 S 1987 0.0 0.0000 30 927 7 1987 2646.5 0.7759 31. 896 8 1987 3097.8 0.9082 31 865

a 9 1987 3069.0 0.8997 30 835 l L 10 1987 3406.5 0.9987 31 804 11 1987 2904.4 0.8515 30 774 12 1987 3291.9 0.9651. 31 743-1 l

l l

l 4

4 e =1.mme io 6-21

. . _ - _ - . _ - - . - .-- - - . -. ..-.-..x-...- . . - . . . . -

- . . ~. - -- . . _ - . . -. . -. .-

l ,

I i TAB;E 6-7 (Cont'd) ,

IRRADIATION HISTORY OF NEUTRON SENSORS CONTAINED IN CAPSULE X i

IRRADIATION DECAY PJ TIME TIME MONTH YEAR (MW) PJ/ MAX (DAY) (DAY) 1 1988 3193.5 0.9362 11 712 2 1988 3408.3 0.9992 29 683 3 1988 3398.3 0.9963 31 652 4 1988 3397.7 0.9961 30 622 5 1988 2641.2 0.7743 31 591 6 1988 0.0 .O.0000 30 561 1 7 1988. 245.7 0.0720 31 530

8 1988 3108.3 0.9113 31 499 l

' 9 1988 3399.1 0.9965 30 469

• 10 1988 3384.7 0.9923 31 438 11 1988 3327.1 0.9754 30 408 12 1988 3399.0 0.9965 31 377 I

1 1989 3326.7 0.9753 31 346 l 2 1989 3393.6 0.9949 28 318  ;

3022.4 3 1989 0.8861 31 287 l 4 1989 2679.8 0.7856 30 257 5 1989 3036.4 0.8902 31 226 6 1989 3285.2 0.9631 30 166 7 1989 2018.8 0.5918 6 190 L

l Note: Reference Power = 3411 OWT b

= won ie 6-22

, TABLE 6-8 '

MEASURED SENSOR ACTIVITIES AND REACTION RATES  ;

Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dis /sec-om) Nis/see-om) (RPS/ NUCLEUS) +

Cu-63 (n.e) Co-60 Top 1.23 x'10 5 3.43 x 10 5 Middla 1.32 x 10 5 3.68 x 10 5 ,

Botton 1.28 x 10 5 3.57 x 10 5 4

Average 1.28 x 10 5 3.56 x 10 5 5.43 x 10'17 Fe-54(n.p)Mn-54 1.58 x 10 6 Top 3.31 x 10 6

) - Middle 1.71 x 10 6 3.59 x 10 6 ,

Bottom 1.67 x 10 6 3.50 x 10 0 Average 1.65 x 10 6 -

3.47 x 10 6 5.53 x 10 -15 1

Ni-58(n.p)Co-58 Top 6.20 x 10 6 5.24 x 10 7 Middle 6.55 x 10 6 5.53 x 10 7 Bottom 6.44 x 10 6 5.44 x 10 7 Ave ige 6. 'O x 10 6 5.40 x 10 7 7.71 x 10 -15 U-238(n.f)Cs-137(Cd) l .

Middle 5.14 x 10 5 5.81 x 10 0 3.83 x 10'14 1

0

=>.musma e 6-23 i

.s. ., , . , . . .. .- .. . . , - . . . . - , - . - -

. IA?'F.6-8 MEASURED SENSOR ACTIVITIES AND REAC'10N RATES - cont'd -

l l

Measured Saturated Reaction Monitor and Activity Activity Rate Axial Location (dia/sec-om) (dis /sec-am) (RPS/ NUCLEUS)

Np-237(n f) Cs-137 (Cd)

Middle 4.60 x 10 6 5.20 x 10 7 3.15 x 10'13 j i

Co-59(n,r)Co-60 Top 3.33 x 10 7 8.57 x 107-Middle 3.19 x 10 7 9.14 x 10 7 s Bottom 3.20 x 10 7 8.19 x 10 7 Average 3.24 x 10 7 8.63 x 10 7 5.90 x 10 -12 1 Co-59 (n,r) Co-60 (Cd)

Top 1.78 x 10 7 4.97 x 10 7 Middle 1.67 x 10 7 4.66 x 10 7 I 7

Bottom 1.69 x 10 7 4.71 x 10 l

Averaga 1,71 x 10 7 4.78 x 10 7 3.12 x 10-12 l

1 neweemee.ie 6-24

-_. . - .- . . - - _ . _ ~ - -. -. ..

l ,

TABLE 6-9

SUMMARY

OF NEUTRON 00SIMETRY RLSULTS ,

s TIME AVERAGED EXPOSURE RATES l 2

e(E>1.0MeV)(n/cm-sec) 1.10 x 10 11 1 8%

4.88 x 10 1I 2

e (E> 0.1 WeV) (n/(m -sec)  ! 15%

dpa/se: 2.11 x 10-10 11%

2

• (E< 0.414 eV) (n/cm -3ec) 4.33 x 10 10 30%

INTEGRATED CAPSULE EXPOSURE 2

t(E>1.0MeV)(n/cm) 1.45 x 10 18 t 8% 5 2

4 (E> 0.1 WeV) (n/cm ) 6.41 x 10 18 1 15%

dpa 2.77 x 10-2 gig 2

6 (Ec 0.414 eV) (n/cm ) 5.69 x 10 18 30%

NOTE: Total Irradiation Time = 4.16 EFPY

)

i t

mwnsasso ie 6-25

i TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSUi.E CENTER Adjusted )

Reaction Measured Calculation CJ Cu-63 (n,s) Co-60 5.43x10'17 '5.43x10'17 1.00 Fe-54 (n.p) Mn-54 -15 5.53x10 5.62x10-15 1.02 i Ni-58(n.p)Cc-58 -15 7.71x10 7.74x10-15 1.00 0-238(nf)Cs-137(Cd) 3.83x10'14 3.39x10'14 0.88 ,

Np-237(n.f)Cs-137(Cd) 3.15x10'13 3.34x10'13 1.06 Co-59(n,r)Co-60(Cd) -12 3.12x10 3.11x10-12 1.00 Co-59(n,r)Co-60 5.90x10 -12 5.89x10-12 1.00 i

.1 w u."'"

• 6-26

a TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUN AT.

THE SURVEILLANCE CAPSULE CENTER Energy Adjusged Flux Energy Group Group AdjusgedFlux l (Mov) (n/cm -sec) '(Mov) (n/cm -sec) >

1 6 ~3 10 I 1 1.73x10 4.44x10 28 9.12x10 2.21x10 1 7 2 1.49x10 1.09x10 29 5.53x10~3 2.88x10 10 3 1,35x101- 5.04x10 7

30 3.36x10 ~3 9.09x10 8

i 1 8 ~3 8 4 1.16x10 1.29x10 31 2.84x10 8.77x10 '

1 8 8 5 1.00x10 3.13x10 32 2.40x10~3 8.52x10 0 8 ~3 6 8.61x10 5.68x10 33 2.04x10 2.41x10 10 I 0 8 7 7.41x10 1.37x10 34 1.23x10~3 2.22x10 10 ,

0 8 ~4 10 8 6.07x10 2.00x10 35 7.49x10 2.07x10 0 8 10 9 4.97x10 4.27x10 36- 4.54x10~4 1.98x10 0 8 10 3.68x10 5.72x10 37 2.75x10~4 2.13x10 10 0 10 10 11 2.87x10 1.21x10 38 1.67x10'4 2.39x10 0 10 ~4 12 2.23x10 1.68x10 39 1.01x10 2.30x10 10 0 10 13 1.74x10 2.37x10 40 6.14x10 -5 2.25x10 10 0 10 -5 l 14 1.35x10 2.64x10 41 3.73x10 2.17x10 10 0 10 15 1.11x10 4.86x10 42 2.26x10 -5 2.08x10 10  ;

16 8.21x10'1 b.57x10 10 43 1.37x10

-5 10

~

1.99x10 17 6.39x10'I 5.81x10 10 44 8.32x10-6 1.86x10 10

-1 10 18 4.98x10 4.22x10 45 5.04x10-6 1.68x10 10 10 19 3.88x10'1 6.04x10 40 3.06x10 -6 1.53x10 10

~1 10 20 3.02x10 6.00x10 47 1.86x10 -6 1.38x10 10 21 1.83x10'1 10 10 5.99x10 48 1.13x10-6 1.03x10 10 10" 22 1.11x10'I 4.79x10 49 6.83x10 ~7 1.04x10 7 23 6.74x10-2 10 10 3.30x10 50 4.14x10'7 1~.12x10

-2 10 24 4.09x10 1.86x10 51 2.51x10 ~7 9.25x10 8 25 2.55x10

-2 10 ~7 8 2.54x10 52 1.52v10 7.55x10

-2 10 '

26 1.99x10 1.20x10 53 9.24x10 -8 1.53x10 10 27 1.50x10

-2 10 1.51x10 l

NOTE: Tabulated energy levels represent the upper energy of each group.

w. man ie 6-27

TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED

• EXPOSURE LEVELS FOR CAPSULE X Calculated Measured C/M 2

6(E> 1.0 MeV) (n/cm ) 1.21 x 10 18 1.45 x 10 18 0.83 2

4(E> 0.1 MeV) (n/cm ) 5.56 x 10 18 6.42 x 10 18 0.87

. dpa 2.46 x 10 ~2 2.78 x 10-2 0.89 2

(E< 0.414 eV) (n/cm ) 4,74 y 3 18 5.69 x 10 18 0.83 4

4 I

seet.mme ie 6-28

TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS /T KEY LOCATIONS ON THE PRESSURE VESSEL CLAD /8ASE METAL INTERFACE AZIMUTHAL ANGLE 0' 15' 30' 45' 4.16 EFPY

.t 6(E> 1.0 MeV) 1,71 x 10 18 2.58 x 10 18 1.95 x 10 18 2.75 x 10 18  !

2 (n/cm )

6(E> 0.1 NeV) 3.44 x 10 18 5.23 x 10 18 4.73 x 10 18 6.62 x 10 18 2

(n/cm) dpa 2,51 x 10'3 3.74 x 10'3 2.97 x 10'3 4.11 x 10'3 16.0 EFPY 6(E> 1.0 MeV) 6.58 x 10 18 9.92 x 10 18 7.50 x 10 18 1.06 x 10 18 2

(n/cm )

f(E> 0.1 NeV) 1.32 x 10 19 2.01 x 10 18 1.82 x 10 19 2.55 x-10 18 2

(n/cm )

dpa 9.65 x 10'3 1.44.x 10 -2 1.14 x 10-2 1.58 x 10 -2 32.0 EFPY 6(E>1.0MeV) 1.32 x 10 18 1.98 x 10 19 1.50 x 10 18 2.12 x 10 19 2 l (n/cm) 1 6(E> 0.1 NeV) 2.64 x 10 18 4.02 x 10 19 3.64 x 10 18 5.10 x 10 18 2

(n/cm) dpa 1.93 x 10-2 2.88 x 10 -2 2.28 x 10 -2 3.16 x 10 -2 sen.msese ie 6 --

1. . .

TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN Tit: INERATICII 0F HEATUP/C00LO0tel CURVES 16 EFPF BIEUTRON FLUEllCE (E > 1.0 leeV) SLOPE 4a SLOPE (n/cm )

2 g,q,y,,j,,g ,fc ,2)

Surface 1/4 T 3/4.T Serface 1/4 T 3/4 T O' G.58 x 10 18 3.58 x 10 18 7.67 x 10 II 6.58 x 10 18 4.14 x 10 18 1.44 x 10 18 15' 9.92 x 10 18 5.37 x 10 18 1.12 x 10 18 9.92 x 10 18 6.21 x 10 18 2.12 x 10 18 30* 7.50 x 10 18 4.10 x ' 0 18 8.97 x 10 II 7.50 x 01 18 5.00 x 10 18 1.91 x 10 18 45* 1.06 x 10 18 5.67 x 10 18 1.15 x 10 18 1.06 x 10 I8 6.77 x 10 18 2.33 x 10 18 b

32 EFPY NEUTRON FLUElICE (E > 1.0 IdeV) SLOPE 4a SLOPE (n/cm )

2 (equivalent n/ce2)

Surfaca_ 1/4 T 3/4 T Surface 1/4 T 3/4 T 0* 1.32 x 10 I8 7.18 x 10 18 1.53 x 10 18 1.32 x 10 I8 8.29 x 10 18 2.89 x 10 18 15* 1.98 x 10 I8 1.08 x 10 I8 2.23 x 10 18 1.98'x 10 I8 1.24 x 10 I8 4.84 x 10 18 30* 1.50 x 10 I8 8.23 x 10 18 1.79 x 10 18 1.50 x 01 I8 9.99 x 10 18 3.83 x 10 18 45* 2.12 x 10 I8 1.14 x 10 I8 2.29 x 10 18 2.12 x 10 II 1.36 x 10 II 4.65 x 10 18 m,.mm. ..

TABLE 6-15 i j

UPDATED LEAD FACTCRS FOR McGUIRE  ;

UNIT 2 SURVEILLANCE CAPSULES Cac? ult Lead Factor (s)

U 5.28 l X 5.28 W 5.28 2 5.28 ,

V 4.62 Y 4.67 i (a)Plantspecificevaluation l 1

I j

1 mer.mmeo io  ;

6-31

i l

SECTION 7 I SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is recomended for future capsules to be removed from the McGuire Unit 2 reactor vessel: l Estimated Capsule 1 Loa $on Lead Removal Fluence Capsula (dog) 2 Factor Time (*) (n/cm )

V 58.5 4.62 1.03 (removed) 3.06 x 1018(b)

X 236 5.28 4.16(removed) 1.45 x 1019(b)

, V 56 5.28 6 2.11 x 1019(c)

W 124 5.28 10 3.51 x 10 18 2 30.1 5.28 Standby --

j Y 238.5 4.67 Standby --

a) Effective full power years from plant startup b) Actual fluence c) Approximate fluence at vessel inner wall at end of life (32 EFPY)

}

=, ..

71

~

SECTION 8 REFERENCES

1. Koyama, K., and Davidson, J. A., " Duke Power Company William B. McGuire Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9489, May 1979.

2. Yanichko, S.E., et al., " Analysis of Capsule V from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program", WCAP-11029, January 1986.

3. Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughness Requirements" and Appendix H, " Reactor Vessel Material Surveillance Program Requirements" U. S. Nuclear Regulatory Connission, Washington,  !

D.C.

9

4. Regulatory Guide 1.99, Revision 2, " Radiation. Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988.

5. 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.

6. "0RNL RSCI Data Library Collection DLC-76, SAILOR Coepled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors". '

7. V. O. Perone, et al., "The Nuclear Design and Core Physics Characteristics of the W. B. McGuire Unit 2 Nuclear Power Plant - Cycle 1. "WCAP-10182, Ssatomber 1982. (Proprietary) .

8. C. R. Savage, et al., "The Nuclear Design of the McGuire Unit 2 Nuclear Power Plant - Cycle 2," WCAP-10747, March 1985. (Proprietary) semame

• 8-1

, . .. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 i

i 3. P. D. Banning, et al., "The Nuclear Design o' the McGuire Unit 2 Nuclear Power Plant - Cycle 3," WCAP-11048, March 1986. (Proprietary) l

10. P. D. Banning, et al., "The Nuclear Design of the McGuire Unit 2 Nuclear Power Plant - Cycle 4," WCAP-11530, June 1987. (Proprietary)

12. J. R. Lesko, et al., "The Nuclear Design cf the McGuire Unit 2 Nuclear i Power Plant - Cycle 5," WCAP-11891, July 1988. (Proprietary)

12. ASTM Designation E482 52, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12 Americar. Society for Testing and Materials, Philadelphia, PA, 1984, i s

13. ASTM Designation E560-77, " Standard Reconnended Practics for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM -

4 Standards, Section 12, American Society for To ting and Waterials, Philadelphia, PA, 1984. #

, 14. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", '

in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984. -

15. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards.

Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984,

16. ASTM Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, An.orican Society for-Testing and Materials.

Philadelphia, PA, 1984.

l l

sov nsseee se g.g

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

i

18. ASTM Designation E262-77, " Standard Met'~d for Measuring Thermal Neutron Flux by Radioactivation Techniques", in M TM Standards, Section 12,  !

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

l 1

19. ASTM Designation E263 82, " Standard Method for Determining Fast-Neutron Flux Dentity by Radioactivation of Iron", in ASTM Standards, Section 12,  ;

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

20. 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

21. 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. 1

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

23. 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,

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

B-3

f

25. ASTM Designation E1005-84', " Standard Method for Application and Analytic of Radiometric Monitors for Reactor Vessel Surveillance", in ASTM Standards, Sectio- 12 American Society for Testing and Materials, Philadelphia, PA, 1984.

26. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Enginearing Development Laboratory, Richland WA, September 1979,

27. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative  !

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

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

t b

b s ur,4uma n g.4

. . J

APPENDIX A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessei. 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 RT i8 NDT. NDT designated as the higher of either the drop _ weight nil-ductility transition temperature (NDTT) or the tamperature at which the material exhibits at least ,

50 ft-lb of impact energy and 35-mil literal expansion (normal to the inajor i workingdirection)minus60'F.

RT NDT increases as the material is exposed to f ast-neutron radiation.

Therefore, to find the most limiting RTNDT at any time period in '5e reactor's 1.ife, ART due to the radiation exposure associated with that NDT j time period mu'st be added to the original ur radiated RT The extent of NDT.

the shift in RT NDT Is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulator) Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)IA,U. Regulatory Guide 1.99, Revision 2 is used for the calculation of RTNDT values at 1/4T and 3/4T locations (T is the thickness of the vessel at the beltline region).

A-2. FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure bour.dar are determined in accordance with the NRC Regulatory Standard Review Plan (A-2 The pre-irradiation fracture-toughness properties of McGuire Unit 2 of the reactor vessels are presented in table A-1.

j l

1 l

wo.,onen io 41

i i

j A-3, CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS l

The ASME approach for calculating the allowable limit curves for various ,

'tatup and cooldown rates specifies that the total stress intensity factor,

{

i(g, for the combined thermal and pressure stresses t.t any time during heatup

!- or cooldown cannot be greater than the reference stress intensity factor, Kgg, for the metal temperature at that time. K gg is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME CodeIA'33 The KIR curve is given by the following equation:

y 26.78 + 1.223 exp (0.0145 (T RTNDT + 160)) (1) where Kgg = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT Therefore, the governing equatico for the heatup-coolc > .alysis is defined in Appendix G of the ASME CodeIA'33 as follows:

CK.+p KIT

• E!R (2) where KIN = stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1.5 fu. nydrostatic and leak test conditions during which the reactor core is not critic 61 w.musse in '

A-2

l At any time during the heatup or cooldown transient, K gg is determined by

]

l the metal temperature at the tip of the postulated flaw, the appropriate value l

for RTNOT, and the reference fracture toughness curve. The thermal stresses l resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, l

l KIT, for the reference flaw are computed. From equation 2 the pressure l

stress intensity factors are obtained and, from these, the allowable pressures are calculated, l

)

I For the calculation of the allowable pressure versus coolant temperature durino 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 th m =1  !

gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate l

of interest.

The use of the composite curve in the coold'own 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 temperatur& than the fluid adjacent to the vessel 10. This condition, of course, is not true for l the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of j Kgg 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 IR exceeds KIT, the calculated allowable pressure during couldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of coeling is decreased at various w.mmo io A-3

intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for ths entire cooldown period. '

1 Three separate calculations are required to determine the limit curves for l finite heatup rates. As is done in the cooldown analysis, alloweble pressure-i temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 i defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal tempern ure at the crack tip lags the coolant temperature; therefore, the K IR for the 1/4 T crack during heatup is lower than the K gg for the 1/4 T crack during steady-state conditions at the same

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 K;g'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 i flaw is considered. Therefore, bcth cases have to be analyzed in order to ensura l

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, tJnlike the situation at the vessel inside surface, the thermal gradients estaolished 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 i 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 finits heatup rate data. At any given temperature, the w.4u'" "

A-4

I allowable pressure is taken to be the lesser of the three values taken from  :

l 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 10CFR50IA 3 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 NOT by at least 120'F for normal operation when the pressura exceed: 20 percent of the preservice hydrostatic test pressure. -

Table A-1 indicates that the limiting RTNDT of l'F occurs in the closure -

hecd flange of McGuire Unit 2, so the minimum allowable temperature of, this region is 121*F. These limits are shown in figures A-1 and A-2 whenever applicable.

A-4. HEATUP AND C00LDOM LIMIT CURVES 1

Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using th6 w thods discussed in section 3.0.

Figure A-1 is the heatup curve for 4W/hr and applicable for'the first 10 EFPY with margins for possible instrumentation errors. Figure A-2 is the '

cooldown curve up to 100'F/hr and applicable for the first 10 EFPY with margins for possible instrumentation errors.

l Allowable combinations of temperature and pressure for specific-temperature change rates are below and to the right of the limit iines shown in figures A-1 and A-2. This is in addition to other criteria which must be ms.t before the reactor is made critical. '

The leak limit curve shown in figure A-1 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.

i mer.m== is A.5

Figures A-1 and A-2 define limits for ensuring prevention of nonductile failure for the McGuire Unit 2 Primary Reactor Coolant System.

l A-5. ADJUSTED REFERENCE. TEMPERATURE 1

From Regulatory Guide 1.99 Rev. 2 (A-1) the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:

i (3)

ART = Initial RTNOT + ARTNOT + M*F9 " -

Initial RT NDT is the reference temperature'for the unirradiated material as defined in paragraph N8-2331 of Section !!! of the ASME Boiler and Pressure Ves:e1 Code. If measured values of initial RT for the material in NDT 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 NOT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

ART NOT=(CF)f(0.28-0.10logf) (4)

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.

24x '.

f(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 ARTNDT at the specific depth.

CF ('F) is the chemistry factor, obtained from reference A-1. All materials in the beltline region of McGuire Unit 2 were considered for the limiting material. RTNDT at 1/4T and 3/4T are summarized in table A-2. From table A-2, it can be seen that the limiting material is lower shell for heatup and cooldown curves applicable up to 10 EFPY. A sample calculation for RT is-NOT shown in Table A-3.

me.museo "

A-6

TABLE A-l' MCGUIRE UNIT 2 REACTOR YESSEL TOUGUNESS TABLE (Unirradiated)(A-5)

Heat Component Number Cu (%) Ni (%) NDT('F)

Closureheadflange(b) 526916 - -

1(a)

Vessel flange (b) 218572 - -

- 4 (a)

Intermediate Shell 526840 .16 .85 - 4 (c)

Lower Shell 411337-11 .15 .88 -30(c) ,

G Intermediate to Lower .

Shell Weld (d) .05 .70 -68(c)

a. Estimated per U.S. NRC Standard Review Plan (A-2) o

b. To be used for considering flange requirements for heatup/cooldown curves (A-4)  !

c. Based on actual data

d. Submergedareweld(weldwireheat 895075 and Grau Lo Flux Lot. No. Peo) t ammuseo i.

A-7 I

l 1

l l

1 TABLE A-2

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURE (ART) AT 1/4T and 3/4T LOCATION 10 EFPY RT AT NOT .

Component 1/4T {'F) 3/4T ('F)  ;

Intermediate Shell 122(84) 91(60)

Lower Shell 90* 61*

1 Intermediate to Lower Shell Wald 33(-9) -1(-19)

Numbers within ( ) are using chemistry factor based on surveillance capsule data.

• These RTNOT numbers used to generated heatup and cooldown curves applicable up to ,

10 EFPY l

l i

l I

1 mer.ammae A-8

1. -

i i

i TABLE A-3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING MCGUIRE UNIT 2 REACTOR VESSEL MATERIAL - LOWER SHELL I  ;

)

Reculatory Guide-1.99 - Revision 2 10 EfPT l Parameter 1/4 T 3/4 T  !

ChemistryFactor,CF(*F) 116 116 Fluence,f(10 18 n/cm)(a)2

,39g ,344 i Fluence Factor, ff .745 -.494 1

ARTNOT = CF x ff (*F) , 86.4 57.3 Initial RT NOT, I ('F) M -30 -30 Margin, M ('F) IC) 34 34

• • • • • • • • • • • • • • • • • .e...********************,.**************************,,****,,

Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 90 61 ART = Initial RTNDT + ARTNDT + Margin l

19 2 (a) Fluence, f, is based upon fsurf(10 n/cm , E>l Mov) = 0.662 at 10 EFPY, The McGuire Unit 2 reactor vessel wall thickness is 8.465 inches at the beltline. '

region.

(b) The initial RTNOT (I) value for the lower shell is based on ceteal data, (c) Margin is calculated as, M = 2 (el * "A ) *I. The standard deviation i for the initial RTNDT margin term (og) is assumed to to O'F since the initial RTNOT is a measured value. The standard deviation for ARTNDT, (ca) is 17'F for the plate.

4 wer.mmeo in A-9 4

i i

i l

WATER!AL PROPERTY BA _ SIS l

CONTROLLING MATERIAL: LOWER SHELL j INITIAL RTNDT: -30'F

' l RT NOT AFTER 10 EFPY: 1/4T, 90'F  !

3/4T, 61'F j

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lNoiCATED TEWPERAtuRC (OCC.r) l Figure A-1. McGuire Unit 2 Reactor Coolant System Heatup Limitations (Heat up rate up to 60'F/hr) Applicable for the First 10 EFPY (With Margins 10'F and 60 psig For In'atrumentation Errors) nmense se A-10

1 r

t 64ATERIAL PROPERTY BASIS CONTROLLING MATERIAL: LOWER SHELL INIT!AL RTNOT: -304  ;

RT NOT AFTER 10 EFPY: 1/4T,90'F 3/47, 61*f i

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0 50 100 150 200 250 300 350 400 450 500 lN01CAfto TEWPCRATURC (DEC r) >

Figure A-2. Mcguire Unit 2 Reactor Coolant System Cooldown (Cooldown rates upto100'F/hr)LimitationsApplicablefortheFirst10EFPY j (With Margins 10'F and 60 psig For Instrumentation Errors) ,

see w n meeie A.11 r - - -, , , , --

A 7. REFERENCES A-1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials " U.S. Nuclear Regulatory Commission, May, 1988.

A-2 " Fracture Teugi; ness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in

• tan.aard 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 !!!, Division 1 -

Appendixes, " Rules for Construction of Nuclear Power Plant Components.

Appendix G, Protection Against Nonductile Failure " pp. 558-563.-10"#

Edition Araerican Society of Mechanical Engineers, New York,1986.

, A-4 Code of Federal Regulations, 10CFR50, Appendix G " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,

Federal Register, Vol. 48 No.104, May 27 1983.

A-5 Yanichko, S. E., Congedo, T. V., Kaiser, W. T., " Analysis of Capsule V from the Duke Power Company McGuire Unit 2 Reactor Vessel Radia+. ion Surveillance Program," WCAP-11029, January 1986.

4 mm.ameo in A-12

4 ATTACHWENT A DATA POINTS FOR HEATUP AND C00LDOWN CURVF.S (With Margins 10'F and 60 psig for Instrumentation Errors)

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S E R S l O P E S I R M P E r G D E M R R Y T E T A H T S) 7 7 M T I 4 6 -

E S ES 7 8 H C E RP t 6

T I T U( 2 V 2 I

S W R K ) S E A I E 2 S E S R N L P P V I (

E E 0 5 R, E C E 0 8 H I R 0 4 p 9 T V U 2 2 9 R S R E S E 1

O S E R)

F PI R UI 0 5 E D I P SS 0 8 D E SP 0 4 I T M E( 2 2 U A U R G L P U M.

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- .. . .. . ~.. - -

g b 08P60 f/HR HEAT 1JP CURVE REG GUIDE 1.99,REV.2 WITH MARGIN 73/22/90

, COMPOSITE CURVE PLOTTED FOR HEATUF PROFILE 2 HEATUP RATE (S) (DEG.F/HR) = 60.0 i'

IRAAO!ATION PERIOD = 10.000 EFP YEARS FLAW DEPTH.= (t-AOWINIT INDICATED ~ ~IWICATED INDICATED INDICATED IWICATED' IWICATED TEMPERATURE - PRESSURE TEMPERATURE PRESSukE TEMPERATURE PRESSURE' (DEG.F) (P:27 (DEG.F ) (PSI) (DE G . F ) (PSI) 1 85.000 *se.04. 14 1'A 000 736.56 27 215.000 1362.95.

2 90.000 530.80 15 15E u00 765.65 28 220.000 1440.48 i 3 95.000 ESA 41 4 16 160.000 796.83 29 225.000 1523.46 4 100.000 $91ie . 65 17 165.000 b .2183 30 230.000 1612.35 5 105.000 599.63 18 170.000 867 *. 31 235.000' 1707.58 6 110.000 603.05 19 '175.000 908.01 32 240.000 1809.38 .

7 115.000 610.15 20 180.000 951.32 33 .245.000 to16.23 3 120.000 620.19 21 185.000- 996.06' 34 250.000 2015,49 9 125.000 633.03 22 190.000 SG43.40 35 255.000 '2121.35 to 130.000 648.67 23 195.000 t102.60. 36 260.000 2234.57-ti t35.000 667.02 24 200.000 116G.90 31 265.000 23SS.22 12 140.000 687.67 25 205.000 1223.52 38 .270.000 2484.10-13 145.000 711.16 26 210.000 1290.80 i

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. + , . ,

M .. m_c 08P COOLDOWN CURVES REG. SUIDE t.99 REV.2 WITH MARGIN 03/22/90 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILE 1 ( STEADV-STATE COOLDOWN )

IRRADIATION PERIOC

• 10.000 EFP YEARS FLAW DEP'H = ACWIN T INDICATED INDICATED It!OICATED INDICATED INDICATED INDICATED T E 94PE R a T URE PRESSURE TEMPERATURE PRESSURE TEMPERAIURE -PRESSURE (DEG. F ) (PSI) 80'i.F) .(PSI) (DEG.F ) (PSI) 95.000 568.04 13 '45.CDO 802.60 25 205.000 1357.16 1

90.000 590.80- 14 850.000 832.77 26 210.000 s428.43 2 27 215.000 1504.63 3 P6.000 594.41 15 155.000 865.41 609.1 f, 13 460.000 900.37 28 220.000 1586.52 4 .100.000 5 105.000 5e&-990 17 165.000 937.63 29 225.000 1674.10 eet-901,8 18 170.000 978 12 30 230.000 $767.96 6 810.000 ~.31 235.000 136c.4e 7 115.000- e69-e8', 19 175.006 1026.40 .

'20 180.000 1067.90 32 240.000 1975.85 8 ?20.000 6e9=f4 2081.07 125.000 70 9 d e7 21 185.000 1917.87 ' 33 S45.000-

-? 34 :50.000 22t3.79 130.000 723 ' 22 190.000 1871.57

?O 35 255.000 2345.17 11 e35.000 748. , 23 195.000 1229.21 12 140.000 774. 5 24 20C 000 1291.00

~

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w

Df 5 COOLDOWN CURVEQ REG. GUIDE 1.99 REV.2 WITH MARGI'd 03/22/90 THE FOLLOWINQ DATA WERE PLOTTED FOR COOLDOWN PROFILE 2 (20 DEG-F / HR COOLDOWN 1 IRRADIATION PERIOO

• 10.000 EFP VEARS FLAW OEPTH = AOWIN T T

INDICATED INDICATED INDICATED INDICATED INDICA!EO IfeICATED TEMPERATURE PRESSURE TEMPERATURE PRESSURE TEMPERATURE PRESSUPE

= ( DEG. F ) (PSI) (DEG.Fl (PSI) ( DEG. F ) (FSI) t 85.000 538.91 8 120.000 666e48 M 15 155.000 849.19 2 90.000 545.31 9 125.000 673.40 16 160.000 886.60 3 95.000 559.33 30 130.000 697.57 87 165.000 926.84 4 100.000 575.54 11 135. 0^A) 723.73 18 170.000 970.09 5 105.000 592.27 12 140.000 751.72 19 175.000 1016.83 6 110.000 610.49 13 145.000 781.79 20 180.000 1066.67

'7 115.000 666,e3 $ 14 150.000 894.33 i

1. tott OSLG, FLANGE REQ 0ipEMEMr i

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DSP COOLDOWN CURVES REG. GU'DE 1.99,REV.2 WITH MARCIN 03/22/90

~

'THE FOttOWING DATA WERE PLOTTED FOR COOLDOtWN PROFILE 3 (40 DEG-F / HR COOLDOWN )

IRRADIATION PERIOD = 10.000 EFP YEARS

, 'Ft4W DEPTH

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-t INDICATED INDICATED INDICATED INDICA'ED IseICATED . INDICATED TEMPERATURE P.JESSURE TEMPEwATURE PRESSUwE TEMPERATURE PRESSURE (DEG.F) (PSI) (DEG.F) (PSI) ( DEG. F ) (PSI) i Geeste 5 1 85.000 495.45 8- 120,000 14 150.000 797.43 2 90.000 509.80 9 125.000 646.32 15 155.000 834.83 3 95.000 525.28 to 130.000 672.34 16 160.000 875.05 4 100.000 541.81 11 135.000 700.22 17 165.000 918.55

5 105.000 559.70 12 140.000 730.18 18 170.000 965.13 ,

6 110.000 579.12 13 145.000 762.67 19 175.000 1085.23  :

I 7 115.000 599.85

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? Law DEPTH = AOWIN T INDICti~f' INDICATED ~ 1.'ADI CA T E D INDICATED IfeICATED INDICATED TEMPERA'ustF 09 ESSURE TEMPERATURE PRESSURE TEMPERATURE PRE SStiaE (Df s. F ) JPSI) ( DE G. F ) (PSI) (DE G. F ) (PSI) 1 85.400 .MS.17 7 115.000 583.39 13 145.000 716.2%

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