Text

WCAP-16835-NP, Rev. 0, Palo Verde Nuclear Generating Station Units 1, 2, and 3; Basis for RCS Pressure and Temperature Limits Report.
	ML090641016
Person / Time
Site:	Palo Verde
Issue date:	06/30/2008
From:	Byrne S, Ferraraccio F, Ganta B, Paggen V; Westinghouse
To:	; Office of Nuclear Reactor Regulation
References
	102-05960-DCM/GAM WCAP-16835-NP, Rev 0
	Download: ML090641016 (85)
	v • d • e

Enclosure 1 Evaluation of the Proposed TS Change Relocate P-T Limits to PTLR ENCLOSURE 1, ATTACHMENT 5 WCAP-16835-NP, Palo Verde Nuclear Generating Station Units 1, 2, and 3; Basis for RCS Pressure and Temperature Limits Report, June 2008

__ _V_

Westinghouse Non-Proprietary Class 3 WCAP-16835-NP June 200 8 Revision 0 Palo Verde Nuclear Generating Station Units 1, 2 and 3; Basis for RCS Pressure and Temperature Limits Report Westingh0use,

Westinghouse Non-Proprietary Class 3 WCAP-16835, Revision 0 Palo Veirde Nuclear Generating Station" Units-1-12 and 3; Basis for RCS Pressure and Temperature

¶ Limit Repor June 2008

.. . ,, r.

S. T. Byrne

-F P..Ferraraccio.-:

-1;* ".B..R.,Ganta.

V..A. Paggen:

ElecnronicallyApproved.Records ire auihenticatedin the Westinghouse Ele ,tonic.Document Management System

- -- -Pittsburgh,.PA 15230-0355.

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WCAP-16835-NP, Rev 0 Page ii

.June 2008.,

TABLE OF CONTENTS LIST O F TA B L E S ...................................................................................................................................................... v LIST O F F IGU R E S ................................................................................................. .................................................. vi "ABSTRACT. .... ......... .......................... .. :

1.0 VESSEL RNEU.TRON.FLUENcE................................ " .. ....... ............................. 1-1 1.2 DISCRETE ORDINATES ANALYSIS ...............-.... ........... ..................................... 1-11....

1.3 VESSEL FLUENCE ANALYSIS,. ...... J......... ... ........................ ; .............. .......... 1-3 1.4 N EU TRON DO SIM ETRY ........... .... '.. ... ......... . .. ....... I........ '. .................. :............... 1-3 1.5....... A L UIO JN C E-R F-6iýI TIE ..ý...,:"..'... ............... ........................................................... 1-5

. CALCULATIONAL .. CERT N....

.1.5 I....

... 1-5 2.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM 2-1 2.1 TEST MATERIAL SELECTION .......... ........................:.......!-..........1............ 2-1 2.1.1 Plate M aterial Selection .... ...........................................................

............. *......*..........*.....2...2*.*......

.. .......... 2..... 2-1 2A.2.1.2 W eld

. 22 *"W eld M aiteria-l*Sele tiori.

Mate~a*.s~ee '. *.... ....

............................ 2-1 2.2 i 212 E ST SP E C IM EN S.....

TTEST-SPECIENS ..... ............ ?............................. .......... : ..... ................... .................... .2 .2.2.1 T ype a d Qu antity ....................... 0 ................... .. ....... .. ....... . ......... ........ 2-2

. 2.2- " UTrradiated 22*2.2.3 Unirfadiated-Specirm s..:: .*.:.'`.{ `.!. .i. ;. .................................)........'.....:2............

Specime ens.:-.--.....:* .... ...:.-. :......... 2-2

.2.2.3. .... ..... .... . ......................... ............ ';2......... 2-3 3 . SURVEILLANCESPECIMEN-IRRADIATION

.. . . 2-3 2.3.! Specim en Encapsulation ..................................................................................................... 2-3 2.3.2 Flux and Tem perature M easurem ent .................................................................................. 2-4

.2:3.3....Irradiation Specim en Location ......................... . ....................................... ............... 2-5 2.3.4 Surveillance Capsule W ithdrawal Schedule ....................................................................... 2-5 3.0 LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM LIMITS .................................... 3-1 3..... LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM ....:.................3-1 3 .1.1 Introduction ........................................................................................................................ 3-1 3.1.2 LTOP Technical Specifications ......................................................................................... 3-1 3.2 BA SIS FOR LTOP SY STEM LIM ITS ........................................................................................... 3-1 3.2.1 Peak Transient Pressures .................................................................................................... 3-2 3.2.2 Applicable Pressure-Temperature Lim its ........................................................................... 3-4 4.0 ADJUSTED REFERENCE TEMPERATURE ......................................................................................... 4-1 4.1 B A C K GRO U N D ............................................................................................................................. 4-1 4 .2 RE SU L T S ........................................................................................................................................ 4 -1 4.2.1 Fluence C alculation ............................................................................................................ 4-1 4.2.2 Chemistry Factor Calculation ............................................................................................. 4-2 4.2.3 Calculation of ART for the Limiting Plates at 1/4T and 3/4T ............................................ 4-2 4.3 AN A L Y SIS DETA IL S ......................................................... *.......................................................... 4-2 4.3.1 Selection of Representative and Limiting Cases ................................................................. 4-2 4.3.2 Calculation of Fluence at 1/4T and 3/4T ............................................................................ 4-3 4.4 LIMITING ADJUSTED REFERENCE TEMPERATURES .......................................................... 4-3

" WCAP-16835-NP, Rev 0 " - Page iii June 2008

TABLE OF CONTENTS, CONTINUED

'5.0 RCS PRESSURE-TEMPERATURE LIMITSS ...... .......... .................. 5-1 5.1 STRESS INTEN SITY FA CTO R .................................................................................................... 5-1 5.1.1". "'G eneral... ... ...... - ..................................................... .............. -:5-1 ...................

.5.1.2., Pressure-Tem perature Lim its Calculation ............................... ... ..... ..........-............

I 5....

5-1 5.2 FRACTURE TOUGHNESS. CRITERIA ........... ........... ......... .................. 5-2 5.3 TRANSIENT PRESSURE-TEMPERATURE.LIMITS .:......:... ..... :..........v..?.L . ...... . ......... 5-3 55.4 CRACK TIP PRESSUREJEMPERAT .IRELIMITS L. ............. .............. .................. ..................... 5-3

.5.5.., HYDROSTATIC AND LEAK.TEST PRESS"URE-TEMPERATURE LIMITS .... :'. ...... z............ 5-4 5.6 M INIM UM BOLTUP TEM PERATURE ........................................................................................ 5-4

'5.7 MINIMUM PRE . SSURE............ RE................

I M N .-.... 2....:.............."...

. ....... ..................... .5-4 5.8- MLOWEST .SERVICES . TEMPERATURE

................... ... ........... :.. .... . .............................. 5-4

...5.9* "FLANGE* LIM ITS ... ...... ........... ................... 5-5

, ...-}... .. ......

.. . . ... .... . . ..... . . .. .......... . . .............. . .. 'l i : ,[ .. , . * . : . " . . * , _. .

5.10 MINIMUM TEMPERATURE REQUIREMENTS ........................ :................................................ 5-5 S5:11 TEMPERATUREREQUIREMENTS FOR NORMAL OPERATION ........ 5-6 5.12 2. -LTOP . . .. . . .....

ENABLE TEMPERATURE LIMITS

. ....... .............. . . ." !2 '/ . !

D ' . . .

.......... 5-6

. .... .... . . .. . . .. .. . . .. . . . . .. .. .. ...... . ....... .... *:'; " .*J...... .; : , ,-* , * . ......-.. .. .... . . .

... 5.13. PRESSURE AND.TEMPERATURE.CORRECTON FACTORS,~... ......... ,.,.................. 5-7 5.14

SUMMARY

OF RCS PRESSURE-TEI4PERATIJRE .LIMITS. ,- .... 5-8

~~ ~~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . .. ... .. *.72*.,.. .;:::*...;... ;......

6.0 MINIMUM TEMPERATURE REQUIREMENTS ...... ....... ..................... 6-1 7.0 APPLICATION OF SURVEILLANCE CAPSULE DATA ..................................................................... 7-1

8.0 REFERENCES

..........................-. ............. 8-1

. ....... .. . . .. . . ... . .. : " , , i ,L , ..

. . . . ... . . . .. . ... . ... ... ~~~ ~ ~~~~-... . . . . . . . . . . . .i .. ,!. ,_. .-.

S.- .: : ) I.:*

WCAP-416835-'NP, RKV 0 .. aP~ge iv June 2008

LIST OF TABLES 1-1 Calculated Neutron Fluence at Core Midplane through 32 EFPY ..................... 1-8 1-2 Calculated Azimuthal Variation of NeutronFluence for Unit I .-......................... .... 1-8 1-3 - Calculated Azimuthal Variation of Neutron Fluence for Unit 2.......v ....... ......... 1-9.

1-4 Calculated Azimuthal Variation of Neutron Fluence for Unit 3 ...... .................. 1-10 1-55 . Comparison of Measured, Calculated and Best Estimate Reaction Rates for.Unit 2................. 1-10 1-6"'. Comparison of Calcul'ated and Best Estimate Neutron Flux for Unit 2 ...... 1-11 1-7 C6nip'arison'ofMMesured t.o Gacuated'ecinkt t~frUi 1 1-8 Comparison of Best Estimated to' C'alcu'tlate'd Ndutroftifiii* Ratios for Unit 2 ......... 111. -

2-1 - Base Metal.Materials Selected for Su...eillace gam :... . :........ 2 2-2 .. Weld Metal Materials Selectedfor ........ ............ ...... 2-6 2-3 Type and Quantity of Specimens for Irradiation Exposure ....................................... 2-6 2-4 Surveillance Capsule Assembly Removal Schedule through 32 EFPY .................................. 2-6 3-1 RCS Heatup and Cooldown Rate Limits through 32 EFPY ........................... 3-4 4-1 Summary of Limiting ART and RTPTS through 32 EFPY ....................................................... 4-4 4-2 ART Input Values for Unit 1 Beltline Materials .............................................................. 4-4 4-3 ART Input Values for Unit 2 Beltline Materials ................................. 4-4 4-4 ART Input Values for Unit 3 Beltline M aterials ....................................................................... 4-5 4-5 Predicted ART Values for Unit 1 Belfline Materials through 32 EFPY .................................. 4-5 4-6 Predicted ART Values for Unit 2 Beltline Materials through 32 EFPY .................................... 4-5 4-7 Predicted ART Values for Unit 3 Beltline Materials through 32 EFPY .................................... 4-6 4-8 Fluence and Fluence Factors at 1/4T and 3/4T through 32 EFPY ......................................... 4-6 5-1 RCS Pressure and Temperature Heatup Limits through 32 EFPY ............................................ 5-8 5-2 RCS Pressure and Temperature Cooldown Limits through 32 EFPY ...................................... 5-9 5-3 LTOP Enable Temperature Limits through 32 EFPY ............................................................ 5-9 5-4 Allowable In-service Hydrostatic Test Pressure ................................................................. 5-10 5-5 Core Critical Lim its for Heatup at 75°F/hr ......................................................................... 5-10 5-6 Core Critical Limits for Cooldown at 100°F/hr ............................................................. 5-10 6-1 Minimum Indicated RCS Pressure through 32 EFPY .................................................................. 6-1 6-2 Minimum Indicated RCS Temperature through 32 EFPY .................................................... 6-1 7-1 Chemistry Factors for Unit 1 Surveillance Plates and Weld Materials ........................................ 7-2 7-2 Chemistry Factors for Unit 2 Surveillance Plates and Weld Materials ........................................ 7-2 7-3 Chemistry Factors for Unit 3 Surveillance Plates and Weld Materials ........................................ 7-3 7-4 Credibility of Surveillance Measurements for Unit 1............................................................ 7-3 7-5 Credibility of Surveillance Measurements for Unit 2 ........................................................... 7-4 7-6 Credibility of Surveillance Measurements for Unit 3............................................................ 7-4

-WCAP-16835-NP, Rev 0 " - Page v June 2008

LIST OF FIGURES 1-1 Reactor Vessel Geometry at Core M idplane . ... :.............. ....................... ........ ............. 1-12 5-1 RCS'Heatup Pressure-Temperatur Limits through 32 EFPY.......................5-11' 5-2 RCS Cooldowrf Pressure-Temperatufe Limits tlhrouigh 32 EFPY............. ........ 5-12 5-3 RCS Con'iiosite Piesstire-Teniperaffir6 Hedtup Limits thfrough 32 EFPY ....... '.................... 5-13 5-4 RCS Comiposite Presstre-Temperati4e Cooldiown Limits through 32 EFPY '. .5-14 5 -4.... ... ..... .. . ...........................

5-5 RCS Through-wall Thermal Gradients at CrackTlps for.Heatup of 75 0F/hr................5-15.

5-6 RCS Heatup Thermal Stress Intensity Factors at 3/4T Crack Tip .............................................. 5-16 5-7 RCS Through-wall Thermal GradientgattGrickTips$for Coododbwn.of 100°FIhr. .......... ...... 5-17 5-8 -. RCS Cooldown Thermal Stress.Intensiy.Fadtors. at t/4T Crack Tip.....' '........................ 5-18

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7 7 A A ' , 7 ' 7 A II 7 A , -

'7 7 " A

.. . . .. . . -" 77, 7 *

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'A WCAP-16835-NP, Rv 0. .'Page vi June 2008

ABSTRACT Methodology used to establish the pressure-temperature (P-T) curves and low temperature overpressure protection (LTOP) system limits for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3 is documented in topical report CE NPSD-683-A, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," Reference 1. Report CE NPSD-683-A has been reviewed and approved by the NRC staff for compliance with the specific requirements of Appendices G and H to 10 CFR Part 50.

The basis for the PVNGS Units 1, 2 and 3 reactor coolant' system pressure-temperature limit curves, heatup and cooldown rates, low temperature overpressure protection setpoints, vessel adjusted reference temperatures and the projected reactoryesseljl(utpn. f-*ence are deyeloped in this report. These heatup and cooldown limits and LTOP controls, effective through 32 effective full-power years of operation, are designed to prevent potential brittle fracture of the reactor pressure vessel during the most restrictive low temperature overpressure event.

The organization of this report follows that presented in Generic Letter 96-03,

'.'Relocation of Pressure-Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," Reference 2. Low temperature overpressure protection setpoints applicable to PVNGS Units 1, 2 and 3 are developed in Section 3.0. RCS heatup and cooldown pressure-temperature limits are developed in Section 5.0.

This report supports the relocation of specific PVNGS Units 1, 2 and 3 reactor coolant system pressure-temperature curves, heatup and cooldown rate limits and low temperature overpressure protection setpoints from the PVNGS Technical Specifications into a separate Pressure-Temperature Limits Report (PTLR) that is controlled by Arizona Public Service (APS).

WCAP-16835-NP, Rev 0 " Page vii June 2008

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WCAP-16835-NP', R6v' 0' : *. - - " agevlui June 2008

1.0 VESSEL NEUTRON FLUENCE 1.1 OVERVIEW The neutron fluence for the Palo Verde Nuclear Generation Station -(PVNGS)Units 1,-2 and: 3 reactor vessel beltline locationshas-been calculated in accordance with Regulatory Guide.1.190,- Reference 3. The following discussions provide the results of the fluence calculation and the details of the calculational analysis for

.A three-dimensional discrete ordinates transport analysigwsi.pe~fohned.for~thePVNGS Units 1 2,and 3

.:reactors to determine the neutron.radiation environrnent*ithiithe h. ctor pressure vesselandsurveillance capsules:. In'this analysis, fast neuttont'fluence with energydlevels-g~eater thafi .1.0MeV and-iron atom-displacements (dpa) are established on, a:plant-!'indfrel.cy'0ler.spedificf basis.. 7hese calculation's form the basis for projecting.,the reactor pressure vessel .ieutronoexposUret to bperating; periods through 32 effective.full-power years,(EFPY). v .r . .... , -

Current neutron exposure data for the PVNGS reactor vessels, drawn from the sensor sets for each withdrawn capsule and":analyzed.using doSimetryievaluati'on methodology, ar.e summarized :in Section 1.4. -The comfiparison of these dosimeti' evaluation resultB to the analyticalpredictiofislis used to validate the plant-specific neutron transportcalculatiofis.: ;:, .:1.ý' f I .. c -

All PVNGSj dosimetry eyaluations are based onthe latest available. nuclear cross-section data derived from -

ENDF*B-VI;.Reference"4. These neutron transport and dosimietry-evaluation methodologies follow the;:".

guidance issued by the-staff in Regulatory'Guide 1.1 90f - . ..

The peak.reactorxvessel ineutron'.fltunce~val.ues ,at 32 EFPY for,PVNGS Units 1,,2 and 3are given in Table 1-1.

These fluence values,"alciculated;at the. coreniidplane.inmen'radiu's,,.1/4T and 3/4T locations,.are based on those reported in the post-irradiation analyses of PVNGS surveillance specimen capsules, References 5, 6 and 7.

Neutron' fluence values at the,1/4T and 3/.4T :locations rare.,determ:ined using-the attenuation formula from -

Regl'atoryiGuid& L99, Reference 8, .. . . . .f K- 3/4 *, , ..

1.2 :-.DISCRETEYORDINATES.ANALYSIS

:.. .7. S , .. .'

plan view of the PVNGS reactor geometry at the.core midplane is shown in Figure 1-1. Six irradiation capsules .attached to the reactor pressure vessel cladding: are included in the reactor vessel surveillance program.

'These capsules are located-at azimuthal angles 380-and -f 1420 (38? from the core cardinal axes), 230' and 310' (400.from the core cardinal axes), and,43. and 137, (430 from the core-cardinal axes). Froma neutronic standpoint,. thesurveillance capsules and associated support structures are significant because the presence of these materials affects both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the core barrel and the reactor vessel. Therefore, the capsules and capsule holders must be included in the analytical model to accurately establish the neutron environment at the test specimen location.

A series of fuel cycle-specific forward .transport.calculations were performed for the PVNGS reactor vessels and surveillance capsules using the following three-dimensional flux synthesis technique when performing the fast neutron exposure evaluations -. , -. .

"..:(i0,z)"--

'." (r0)* ( z)-,"(Eqn. 1-1).

WCAP-16835-NP, Rev 0 '-Page 1-1 June 2008

In Equation 1-1:

4(r,0,z) = synthesized three-dimensional neutron flux distribution;

ý(r,0) = transport solution in r,0 geometry, 4N(r,z) = two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and

. (r)- one',dimensional solution for a cylindrical reactor model using the same

.. soarc per. unit height as-that used inthe (r,O) two 7dimensional calculation;:_

This synthesis procedure is performed for each operating cycle. .

Two sets of transp6it calciilations are1;used bvhon peiformrirg the fast neutronrexposure evaluations for PVNGS.

The first set of calculations ,is based.on t(e *(r,,mtidelshown in'Figure 1-1, thatincludessurveillance capsules at 380, 40', and 43 0 .- The'.second set of calvulatiohis:based, on the (r,O) model having no surveillance .capsules present. The first set of calculations is' ugeld 't.pej.-fo-l siirveillance, capsule dosimetry evaluations for comparison with, calculatedresults:, jTbe,:stcond set.ofalculationsis used to determine the maximum neutron exposure levels at the pressure vessel wall. UFSAR Figures 5.3-1 through 5.3-4 provide additional, details regarding the arrangement and location of surveillance specimens at PVNGS.

Nominal design dimensions are.employed for thevariolisstructurall componefits .when devel0ping these analytical models, with two.exceptioijs:;. The fi7st zxceptionis.ithat the radius to the.cefiter_ ofthesurveillance, capsule holder and the pressure vessel inner radius are taken from the as-built drawings, Refefence .9, for each PVNGS reactor. The second exception is that the coolant temperature (i.e., coolant density) is treated on a fuel cycle-specific basis, with water temperaturesifti the;reactor ¢esseLcore and by.passrregions..based.i- fullpdWer operating conditions. ,The.reactor, core is treated as a homogeneous mixture .of.fdl, bladding14wateri and-,*_ý,_ :

miscellaneous core structures (such as fuel assembly grids ard.guiide 2tiibes)....The6 geometric.,mesh description of the (r,0) reactor models typically consist of 151 radial by 78 azimuthal intervals. Mesh sizes are chosen to ensure that properconvergerice of the iniierf iterations if achieved .n:aLpoint-wise basis.;:.The ,point-.wise inner iteration flux convergence criterionusedin the (rO). calculationsisset at a value, of 0:001.

The (r,z) model used for the PVNGS calculations extends both radially.fromi :the centerline :of.the ,reactor: core out to a location interior to the primary biological shield and axially from. n~elevation, two feet below the active fuel to two feet above the active fuel. As in the case of the (r,0) models, nominal design dimensions (except for the pressure vessel inner radius as-built dimension) and full pqwer cgooant.densities gre.emplQyed in.the calculations. In this case, the homogenous core region is treated as an equivalent cylinder with a volume equal to that of the active core zone. The'stainl'ess st1eel core'shrbud assenbly gifth rings located'between*ithe cbore' shroud and core barfel regions are exqplicitly ihcluded"inli the modeli The"(r;z)5 ge0ometri!c niesh descripfin"on-f-these reactor models typically';consists of 141" radial by 79 axial intervalg. Consistent withi tfie 6as6-'oth&"(r,O) calculatioins, mesh sizes are chosen io ensure4that proper convergenc6o f the inner it6iati'ons is'achieved on a point-wise'basis. The pdiiti-wise inner iterati0f flux cdinverlg.ce ,criteiion used inri the (rz)'ýalcillati6his is set at a valu~e 'of0.001."' " ".. :: " ': ' * : ;:"* ;.: !. . ... .:: :;-.::::

One-dimensional radial models used -in the' synthe'sig procedure consist *aief the 141 radial mesh intervals included in the (r,z) models. Radial synthesis factors can be determined on a mesh-wise basis throughout the entire. geometry. Further drtails'of the analy'sis are' provided iniRefdrences 5, 6 aiid 7.

Data used in the transport analyses represents cycle-dependent fuel assembly enrichmen ts, bumups, axial power distributions, and radial pin powers. This information is used-to..evelop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport WCAP-16835-NPRe70 .. ...... . '... Page 1-2 June 2008

calculations provide data: in terms of fuel cycle averaged neutron flux. This, w-hen multiplied byth'elappropriate fuel cycle length, generates the incremental fast neutron exposure for each; fuel cycle. Inc'o*nstructing these core source distributions, the energy distribution of the source was based on the initial enrichments and burnup histories ofi individual f6ie asseibl-ies. C6mposite Values of energy release per fis§i8n, neutruon' yield per fission, and fission a.

s .

spectrumn are determinied from!i. the assembly-dependent fission splits.. .. . "- ) ." . * , ":,

Allftranspbrt calculations supporting this analysis are-pe'formd using'the'DORT di'ciet; 'ordinate§ code "'

Version 3.2, Reference 10, and the BUGLE-96 cross-section library Reference 11. Tie'BUGLE-961library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor applic~itions. Inth-ese anaiyses anisoitropic' sca.teri.."i§ t.reated'vWith aP 5 Le0&ende expansion band angular discretization iSmodeled with an' S16 0rder of g4gar iiadi&ate. Enetgy- and spaece-ieperndent core powcr distributions, as well as system operating trnperatii;es; are'treated a fuel cycepecic basis.

an 1.3 VESSEL FLUENCE ANALYSIS Fast neutron fluence results from References 5, 6 and 7, projected at a core power level of 3990 MWt for future cycles,-are provided in Tables: .1-2"througli 4A:forPVNGS Unit.:1; 2 hnd:3; respectively. This&

calculated fast fluence with energy ,levels greatvrlmri11.0 MeV is. given foli the :reactor vessel inner radius at four azimuthal locations. 'Datagive *,n'i*abtes 2-.thrdugh*1*4.are detennined for-the clad/base metal:"

interface; therefore,.the' Tata:epresent-the;maximum calcu'ated fluenceon the iespective vessels,::

Data tabulations include both:pl.nt*-and, fuel cycle-specific caLculatedt:fluence'at the end cf the.most recent operating fuel cycle and future projections through 54 EFPY. These projections are based on a core thermal power of 3990 MWt and the assumption that the core power distributions and associated plant operating characteristics from the most Fcehnt bpei-afing"fuei cycle are re*resenitdtive of futureiplant operation.

1.4 NEUTRON DOSIMETRY",'" '

The validity of the calculated neutron flueiice reported in Section.1.3 is demonstrated by a direct comparison against the measured sensor] reaction rates and by ., least ,squares evaluation performed for each of the capsule dosimetry sets. For completes&'hss,ýaan dsessm eit-df all me-asured 0dosimetfy'rerhoved through 15 EFPY is documented in the oost-irradiatign surveillance capsule, evaluation reports for each PVNGS unit. Comparison ofmeasured dosimetry.results to both the calculated and, least squares adjusted values is summarized herein.

The sensor sets from the capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide'1.190. ; Sensor set results demonstrate that the overall measurements agree. with the calculated and least squares adjusted values to within +20%,as specified by Regulatory Guide 1.190.: This agreement within +/-20% validatesthe calculaecd fluence reported in Section 1.3.

A least squares evaluafioii of the PVNGS surveillance capsule dosimetry Was performed using the FERRET code, Reference 12. FERRET was employed to conbin&the results of ihe plant-specific neutron transport calculations and the sensor. set reaction-rate measurements. to determine best-estimate values of exposure parameters and' associated uncertainties for the surveillance capsules withdrawn through 15 EFPY. The application of the'least squaresmnethodology requires the following inptit:

1. The.calculated neutron energy spectrum and associated uncertainties at the measurement location;

2. The measured reaction rates and associated uncertainty -for each sensor contained: in the multiple,,

foil set; and . . : . .... . . . , , ,. :

WCAP-16835-NP, Rev 0 -'Page 1-3 June 2008

3. The energy-dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set. *, -

The calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations with sensor reaction rates derived from the measured.specific activities. Dosimetryvreaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library, Reference 13. The SNLRML library is an evaluated. dosimetry reaction, crosý-section compilation recommended for use in LWR-evaluations by ASTM Standard E, 1018-01, Reference 14., ¶ -

The uncertainties'associated with the measured reaction rates, dosimetr,' cross-sections, and calculated neutron spectrums are input to the least squares proced,'W inthe form of variances and covariances..Input. uncertainties osign guidance proided in AýTM Standard E 944-02, Reference 15. The flowing discussion of the reaction rate uncertainties is associated with the least squares evaluation of the PVNGS Unit 2 surveillance capsule sensor sets, Reference 6. (See References 5 and 7 for the corresponding results for Units 1 and 3.)

The overall uncertainty associated with.'he raeasizred~rea~zlio'a rates includes components- due to the basic ::,

measurement process, irradiation hisory. orrections, Ltd :.-,t:et'.tc, ns for cmpeting reactions.A high level of accuracy in the reaction rate determinations is;ensired '.y -sirg iaI?0.' tory' procedures that confbrmto .the.ASTM National Con sensus. Standardsfo" reaction-rate sJeter-minationsifor, eacih sen.Gor-type.. I:After:combining all. of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures are assigned the following. net uncertainties tfor input to the¢leas'. squares ewluation::K> ,;

ReaaionUnit: nh cert'anty"1 '-

6

. 46

.,Cu('a) Co '

Ti(n,p) 4 6Sc 5%

54 Fe(n,p)5 gMn 5%.

II '0Ni(n,p)"Co 5~/

233 7.3.

c . ' . .I : *-. 1,0% ... .. " ' "

.-:~o "£ ' Uhcertainties aregivenat o,".-.. *the asrecgitn ..one-s'versu

. ,,"3.

theone-.ýigni&(id)l gm a,* .... c ev....:

3......... i. . . d..

. _n*leve . 3.." -:.:"'

The following summary provides a comparisn ot me'aired zcalcul'atd versus (MiC) f st neutronthreshold ."

reaction rates for'the sensors fro-m Capsule'W230'witidrawrfrorfi PVNGSS Uiiit 2 at'the enofthee 'tieifth ftel cycle..

Unit 2 Reaction Rates.'

Reaction' (rps/atom)". n C

. .. . Ratiot /, 3 . 3 Measured Calculated 46

'Tijn P S 5.23E-16 . .28E-16 , 0.99:?.

54 4

' Fe(n,p) Mn . 2.88E-15 , 2.93E-15 . ,0,98.,

8Ni(n~P) 58Co'(Cd)". 3.59E:-15I' 3.82EL-15" '0.94'

. Average: . 0.97,

,Standard Deviation: "2.8 The reaction rate cross-sections Used inthe least squares evaluations ,a-:e taken from theSNLRML library.

This data libraryprovides reaction cross-sections and associated uncertainties,-including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup WCAP-1683 5 NP, Rev0 ..... . ' Page 1-4 June 2008

structure for use in least squares adjustment applications. These cross-sections are compiled from the most recent cross-sectio'n evaluations and have been Jested with reslect to their accuracy-and consistency for least squares evaluations. Furthermore, the library has been empirically tested for use in fission spectra determination,.as well as in the fluence and energy characterizations of 14 MeV.neutron. sources.

For sensors included indthe PVNGS Unit 2 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package,.:

... Reactioi ...' Uiicertiii:yy. , I i 63 Cti(I a) 60 Co -. . "408'-~.416/ ..

.. Ti(n,p) 46 Sc . . - .44.K(I--

... :,:4.7 .A.

4 1 Fe(n P)1 4Mn 30 .1 58 5 58 Ni(n;pP C I, --.. .0 -:5.16 3M Unjf)'~., -Cs954 .6%.

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in*LXVRirradiatins.: ':/i-,,

,,m. *'-,'i -,4.

.. ., , I. . .,

The measured-to-calcuhated rea tion rate ratios fQr the-Capsule W230. threshold-reactionsrange from 0.94 tot, 0.99; the,averag, M/C:ratio is 0.Q7_+2 8V at t:ihe one-sigma.a ve. Jhis direct comparison falls well within the

+/-20% criterionspeci~fied in Regulatory Guide !,190. .:.v. , .. -

1.5 ',CýA.LCULýATIONAL'UNGCERTAINTIES., - ,*' ' ,' '.

The uncertainty associated with the calculated neutron exposure 6of the PVNGS surveillance capsule and reactor pressure vessel is based on the recommendedapproach provided inRegulatory Guide 1.190. In particular, the qualificationpof the methodology was carried out in the following four stages:

w, c*..rr, 1.n* e-, ,, ,v.

1. Comparison of calculations with benchmark measurements from the Pool Critical Assembiy (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2. Comparisons of calculations t6 surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor bgenchmark experiiment.... '.

3.,.,. .An analytical sensitivity study addressing, the uncertainty components resulting from important input

, parameters applicable to,the plant-specific transport calculations used in the neutron exposure assessments.

4. Comparisons of the plant-specific calculations'to :all available dosimetry -results from the PVNGS .

,survpeillance program ::. . .- .- , , , .,I .

The first phase of the methods qualification, PCA comparisons, addressed the adequacy of basic transport calculation, as ,well as dosimetry evaluation.techniques and associated.cross-sections. However, this phase.,,.

,did not test the,,accuracy of commercial..core neutron source calculations, nor did it address uncertainties in operational or.geometric variables.that impact power reactor calculations. The second phase of the qualification- H. B. Robinson ccomparisons, addressed uncertainties in those additional areas that are primarily methods-related, and ould ,tend to apply generically to all fast neutron exposure evaluations. The third, phase of the qualification, analytical sensitiyity sudies, identified the potential uncertainties..introduced into the overall evaluation due to calculationaj. methods approximations, as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to .he PVNGS analysis is established from results of these three phases of the methods qualification.

WCAP-16835-NP, Rev 0 4 .~' <,.' .Page'1 -5 June 2008

The uncertainties developed from the; first three phases of the methodologyqualification are summarized as follows:

Basis Capsule Vessel IR~l)

PCA.Compa'sons .3% :3%

H. B. Robinson.Comparisons- . . . . 3%.... %.

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors-not Explicitly Evaluated " 5% 5%

Net Calculational Uncertainty - .. . ..... .. 12% .0 13%

) Uncertaintyat IR applies to the V. ssel inside radius.

The net calculational uncertainty was detern'tified by combining the iifdiViduual components in quadrature.

Therefore, the resultant uncertaintýy was fte'ated.asrand-6 m.and.nbosystematic bias was applied to the analytical results.

The fourth phase of the uncertainty assessment, comparisons to PVNGS measu'rermenis, is used:,to.

demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical resutsý Tle'cbrudonly *sa c-h'nt-toi*,i6di-ftthecalcutlated sufrveillance;,

capsule"and pressure vessel nieution. expb'isreS'Rs"esRdts ofth~e-e'ast quat-es"valiatiois 6f the dosimetry from the PVNGS Unit 2 surveillance capsules withdrawn to date ai6 provided 6,Tt7bles i.52and1:6.In Table1.L5.

measured, calculated, and best-estimate values for sensor reaction rates are given for each Unit 2 capsule.

Also provided in this tabulation are ratios of the measqcd reacA:iontrt toboth.the-ealculated*! iarleast squares adjusted reaction rates. These ratios of measured-to-calculated (M/C) and measured-to-best-estimate (M/BE) illustrate the consistency. of the calculaed energy spebiffa ift'tIe 'm*asud re5action riaies'-

rot ~i*fu both before and after adjustment. In Table 1-6, comparison of the calculatedand best estimfateValiues of fast neutron flux (E > 1.0 MeV) and ircn at*m disp iacementý rate are tabulatedi aor'g a the best-estimate -to-ithwh calculated (BE/C).ratios observed for each ofthe capsules., . *

The data comparisons provided in Tables 1-5 and 1-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned tic'erttiahtils foi thecalculated spectra, merasured, snsor reaction rates, and dosimetry reaction cross-sections. Furthermore, these results indicate that the use of the least squares evaluationi results in a reduiction ihi the Uiicerta'intie's associated vith the ,epdsure bf the surveillhinidedcapsules.

For the neutrondxposures giv6riiin Seetionl.3, theutincertaintyassociated Witl the'unadjusted calcnlation.of fast neutron fluence and iron atom displacements at the surveillance capsule locations is specified as"1 O7 at the one-sigma level. The corresponding uncertinties .associated with.the' least squares adjusted exposure parameters, Table 1-6, are 7% for fast neutron flux and 6% for iron atom displacement'rat&.! Agafiri; these uncertainties from the least squares evaluation are at the one-sigma level.

Further companisonsof the measurement results to'calculations are p"6vided in Tlhbi6'*1' anid 1-8" These" omparisons are provided on two levels1 :ITable 1l-7,"c'alcu!laons of midividlU thieshold'sensor reaction rates are compared directly to the &orresponding mueasureienes: The'e ikereshid6d:re'e"tion'i rait*comparisons provide a good evaluati'n of the accuracylof the fast neutron'poition of the calculated energy sp4`dra' hiTable 1-8, calculations of fast neutrbn ýexpo6sure rates iiniterms of nreutron flu"adie irdn atom. displaceinent rate (dpa/s) are compared to the best estimate results obtained from'the lea8tsý§qurfe'ý evaliuation of the capsule dosimetry results. These two levels of Comparison yield consistent and similat 1 sultsl: All mneasurementto-c;alculationI comparisons fail :well within the 20% limits specified as the ;cceptauic& criteria in Regulatory Guide 1.190.

WCAP-1683'5-NP, Rev 0.... .. . ... Pfige 1-6 June 2008

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.94 to 1.16 for the eight samples included in the data set. The overall average M/C ratio for the entire set of PVNGS Unit 2 data is 1.03, with an associated standard deviation of 7.6%.

In the comparisons of best estimate and calculated fast neutron exposure parameters for Unit.2, the corresponding BE/C comparisons for the capsule data sets range ftomý0.96 to 1.01iforofast neutron flux, and fr6mon:96 to 1.00for ironi atom displacemehfirate The 6 a-egBE/C iVitiosfbrastineutron- flux and i ieerarl iron atom'displacement rate are 0.98 with a standard deviation, of3. %;-and 0.99 with a standard deviation of 3.4%, respectively. Results-from-the other two units-arc.compa,-alei-For Uit 1; the correspondig BE/C comparisons for the capsule data sets range from 0.94,to 1.03 for fast neutroi flux,'and'fr ,m 0.94 to 1.03 for iron atom displacement rate. The overall average BE/C ratios fqr.;fqptunue.ytron fliux and iron atom displaceefiiet rate'are 0.98 with-a standardde-idtibifbi* .7%,-4Jd&0:98'Withh -fstandard deviaiirn of4.5%i, respectively. For Unit 3, the corresponding BE/C comparisons for the capsule data sets range from 0.95 to 1.01 for fast neutron flux, and from 0.96 to 1.01 for iron atem dispiacement rate. The overall average BE/C ratios for fast neutron flux and iron atom displacerentrate are 0.98 with a standard deviation of 4.7%, and 0.98 with a standard deviation of 4.4%;,, Tespective!y.<,:v > .:. . . , .

Based on these comparisons.lt..'iclued nt th-ecalated fast..neutro exposures provided in Section 1.3 of this report a~e validated for:,use in tlhe assessment of the condition of thematerials comprising the beltline region of the P"NGS UJnit 1, 2 and 3 reactor pressure vessels. ..... p d t I -

II I - -

Pagej1-7 WCAP-16835-NP, Rev 0 June 2008

Table 1-1 Calculated Neutron Fluence at Core Midplane through 32 EFPY S"'Unii),.:

Uni3) .PNG PVNGS Unit 13 V) .:, PVNGS Unit - PVNGS Unit~nUnit 3(3r Vesel InneiRadids(' 1 1.51E+19 nr/m 2 l.66E+19n/cm 2 -1.74E÷194n/cm";

Vessel 1/4Thic rikss( 2

8.77E+18n/h*n'9 .63E1+18n/cm 22 A, Ol 9n/cm2 .

-- kes 2 esel3/_T_ 3/4Thickness~

Vessel . , .9E 1___ic__. 3.25E+-18 n/cm "3.40E+i8 3.g 0__ n/cm2 ..

() Clad-to-base metal interface. .

(2) £.

CalculatdI asf

fsffae, e -024x)Y 0.2 hee x '4 býiA/4 thicldess dimension 'fo Vesel lower Shell....

(3) Projected at' a coie power of3990M'*Vt. . .

STble1-2 Calculated Azimuthal Variation ot Neutro'on Flufence fol'Unit 1 ,

Cumulative [ , .Neutron Fluence?,) EŽ 1.0:MeV 2

2n/cj Cycle f" irradiation -..

Time- 156 00 45'

__ __ (F )~~m -0 ..........- '.

1 1.22 5.74E+17 8.18E+17 8.32E+/-17 9.62E+17 2 2.00 9.14E+17 1.19E+18 1.24E+ 18 1.40E+ 18 3 3.37 1.57E+18 1.93E+18 1.89E+18 2.08E+18 4 4.57 2.03E+18 2.57E+18 2.49E+18 2.72E+18 5 5.79 2.31E+18 2.92E+18 2.90E+18 3.28E+18 6 6.99 2.57E+18 3.26E+18 3.31E+18. 3.84E+18 7 8.32 2.83E+18 3.61E+18 3.69E+18 4.28E+18 8 9.76 3.14E+18 3.97E+18 4.05E+18 4.71E+18 9 11.13 3.48E+18 4.38E+18 4.54E+18 5.37E+18 10 12.49 3.81E+18 4.80E+18 4.97E+18 5.85E+18 11 13.83 4.16E+18 5.23E+18 5.44E+18 6.47E+18 Future(l) 16.00 4.75E+18 5.94E+18 6.22E+18 7.50E+18 Future(1 ) 32.00 9.06E+18 1.12E+19 1.20E+19 1.51E+19 Future(1 ) 48.00 1.34E+19 1.64E+19 1.78E+19 2.28E+19 Future(l) 54.00 1.50E+19 1.84E+19 1.99E+19 2.56E+19 (1) Future projections for Unit 1 are based on power level of 3990 MWt.

(2) Calculated at reactor vessel clad-to-metal interface.

WCAP-16835-NIP, Re~ 0 Page '1-8 WCAP26835-NP; PeV,0 Jn June 2008

Table 1-3 Calculated Azimuthal Variation of Neutron Fluence for Unit 2 Cumulative Neutro.i Fl~ence 2), E > 1.0 MeV (n/cm 2)

Irradiation7 -

yle Time 100 50 (EFPY)300 450 1 :.15 5.4§9E+17 7*ý77 7 7.96E+17 9.20E+17.

2 2.23 9.10E+17 A1.28E118 1'.34E+18 1.51E+18 3 3'.41f 134E+1i8 3:1.86E+÷18 1L92E+18 2.18E+18:

.4 4.54 i 1.77E+8 4E+18 2.50E+18 2.79E+18" 5 5.52 1A.980Et#18 2.13t+I8 2.97E+18 3.32E+18 6 6.39- 2.21E+18 3.03Et+18 3.29E+18 3.70E+18-7 7.48E+/-18 '343E+18 3.84E+18 4.32E+18-:

8 9.12 , 2.7 E+18 ' .384E+'18 423Et+18; 4.72E+18-*

!t

-910.50 3.iOE-+18 4-2ýE+18 4.67t+18 - 5.32E9+18i 10 -1173 '338E+18 4:64E+ 8 5.14E+18 5.92E+18,;

i;"

- A 1-3:.14..-E.18- 9 -. ,.'2 * -1 ....... 5.64E+18 -6.50E+.18 12 35 '4. ý-:30E-18:` `5;66E+/-8-2 '6.12E+18[ '7.14E+18

,Future(,. .1600 04.83E 18, .6.30E+..l 6.79E18j I 8.03E+18 1

FutureO ) 32.00 9.96E+18 .1.24E+19 1.33E+19 1.66E+19 Future(') 48.00 1.51E+19 1.85E+19 1.97E+19 2.51E+19 1

FutureO ) 54.00 1.70E+19 2.08E+19 2.21E+19 2.83E+19

() Future projections for Unit 2 are based on power level of 3990 MWt.

(2) Calculated at reactor vessel clad-to-metal ifterface.

4 4 I ý -".ý a-4 WCAP-16835-NP, Rev 0 "Page 1-9 June 2008

Table 1-4 Calculated Azimuthal Variation of Neutron Fluence for Unit 3 Cumulative Neutron Fluence(2 ), E > 1.0 MeV (n/cm 2 )

Irradiation Cyc0Time 0 150 300 450

- _ __ _ (EFPY); ., _" .. - . ..

11.07 5':31E+1'7 7.71 E+ 17 8.94E+17,

-2.14. 8&9EA-i.1.6.

. .1.30E+18 1.47E+18ý 3 3.31 1.32E+18 1.76E+18 1.81E+18, 2:09E+18 4 4.44 1.62E+18 2.17E+18 2.28E+18 2.68E+18 75 * -5.67 ...... 797 18- ,59E18' 2.77E+18 3.26E+18; 6.89 222E+18 i 2.94E÷18 3.20E+18: 3.78E+18

8.33 2 53E+18 33 SE+18 3.61E+18 4.22E+18 8- . .9.75 2.86E+18 3.84E+18 4.12E+18, 4.86E+1g 9 .1.1.07 3 14+ 8 4..23E+8 4.63E+18: 5.49E+189 12.44 346E+18 ,468E+,18 5.24E+18, 6.20E+18i

11, . 13.75 ? 77E+l,8 5.15E+18 5.84E+18 6,95E+18, Future(') 16.00 4.32E+18 5.96E+-18 6.87E+18 8.23E+18 Future(i) , 32.00 .8.25E+18 1.17+19 .1.42E+19 1174E+19

~~19 Future() " 48.00 :,1 22E+19 .1.75E+19 i 2.45E+19' 2.65E+19,;

.Future(') 54.00 1.37E+-19 - 1.97E+.19 2.43E+19 2.99E+ 19

-(1) Future projections for Unit 3 are. based onpow.r -levelof 399,0 .MWt.

'(2) Calculated at reactor iiessel clad-to-metal interface. ' -

Table 1-5

-Coiiip:ris6fnofMeaisured, Calculated anid Best EstiiiimiteReactionhR: ates-for'UniF2 Capsule W137 Reaction Rate(t) Capsule W137 Ratios Reaction Measured Calculated Best Estimate M/C MIBE 63 Cu(n,CC) 60Co (Cd) 4.84E-17 4.18E-17 4.72E-17 1.16 1.03 46 Ti(n,p) 46Sc 7.24E-16 6.44E- 16 7.1OE-16 1.12 1.02 54 Fe(n,p) 54Mn 3.75E-15 3.60E-15 3.80E-15 1.04 0.99 "8Ni(n,p) 58Co (Cd) 4.85E-15 4.69E- 15 4.92E- 15 1.03 0.99 238 U(n,f)137Cs (Cd) l.16E-14 1.2 1E-14 1.24E-14 0.96 0.94 Capsule W230 Reaction Ratet l) Capsule W230 Ratios Reaction Measured Calculated Best Estimate M/C MIBE 46 Ti(n,p) 46Sc 5.23E-16 5.28E-16 5.17E-16 0.99 1.01 54 Fe(n,p) 54Mn 2.88E-15 2.93E-15 2.85E-15 0.98 1.01 58 Ni(n,p) 58Co (Cd) 3.59E-15 3.82E-15 3.67E-15 0.94 0.98

(') Reaction rate in reactions per second per atom.

' WCA.P-16835-NP, R 0. . -.: -Page 1-10 June 2008

Table 1-6 Comparison of Calculated and-Best Estimate Neutron Flux for Unit 2 2

Capsule ID(2) NeutronFlux, E > 1.0 MeV (n/cm -s)

SCalculated') Best Estimate *Uncertainty (1).. BE/C W137 2.70E+10 2.72E+10 6% 1.007 W230 2.19E+ 10 2.1l1E÷101 7% 0.965

______ _ I-ro'n Atom Displacement Rate (dpa/s)

Capsulem2 Calculated() Best Estimi'ae'.] Uincertainty (lo) BE/C W137 3.92E-I1 3.93E-1. .....-. 6% 1.002 W230 3.19E-11 3.04E-1II M.-.6%.... 0.955 Calculated results are based on the synthesized transpori calulatirns*taken at the core midplane following the completion of each;i especfiv6.capsules irr'a*iaticn period.

Capsule identification is based on'azimuYi&al position.

Table 1-7 Comparison of Measured/to Calculated Reaction Rate'Ratios for Unit 2 1

Reaction. *' Measured/Calculated Ratio( )

Reaction; .easure "

Capsule W137 .Capsule W230.

63 Cu(n,a)06CO (Cd) 1.16 -...... . --

Ti(n,P) 4 6 sC : 1.12 0.99 54 Fe(n,P) 54Mn 1.04 0.98.

5 Ni(np)5 8Co (Cd) 1.03. '0:94 238 37 U(nf)1 Cs (Cd) 0.96 --

Average 1.06 0.97 Standard Deviation .7.4% 2.8%

(t The overall average M/C ratio for the set of eight sensor measurements is 1.03 with an associated standard deviation of 7.6%.

Table 1-8 Comparison of Best Estimate to Calculated Neutron Flux Ratios for Unit 2 Capsule ID Best Estimate/Calculated Ratio Neutron Flux, E > 1.0 MeV dpa/s W137 1.01 1.00 W230 0.96 0.96 Average 0.99 0.98 Standard Deviation 3.0% 3.4%

WCAP-16835-NP, Rev 0 ". Page 1-11 June 2008

I Figure 1-1 Reactor Vessel Geometry at Core Midplane 100 156 200 250 " 300 ' 350 R Axis (cm) y

. WCAP-16835-NP, R 0 ........- .............. . . . ... Page 1-12 June 2008

2.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM.

The surveillance program for PVNGS Units 1, 2 and 3 is based on ASTM E185-79, Reference .16,.Which presents criteria for monitoring changes in the fracture toughness properties of reactor vessel beltli'ne materials.

This program and the surveillance-capsule withdrawal schedule for PVNGS ard des~ribedini this' section:-

Reports describing the pre-irradiation evaluations of the surveillance materials are 6onitain.ed in References' 17,

18 and 19. Reteren6ýs 5, 6 and-7 are the most' recent reports that describe the postirradliatibn evaluations of the PVNGS esur*cillance miate ials. . ' ; -...

The:PVNGS surveillance piogram adheresito all ASTM E185- 9giidelin'es' and to 10 CFR 50, Appendix H, Referencfe 20. All react 6 r vessel' surveillance specimen capsule, holders are attached.to ttheinside vessel wall, cladding in the beltline region at PVNGS. This! 'pgiuleý'Oldei; ti'chinient methOdnrheets the design and*

inspection requirements of the ASME Code, Sec"tion 'lI a"hd1XI7-:

2.1 TEST MATERIAL SELECTION Three metallurgically different mateiasreiparesertative, ofthe.reactor~vessel are. used 'ortest sp1e1..ens-in

,.accordance withlthe general guidejlines of ASTM'E:185-J,9. These. matri.s include base metal,i~wed metal, and heat affected;,zone.(.HAZ,) materials. .. . .. '. .. , . :. ,.

2.1.1 Plate Material Selection The lower shell and a portion of the intermediate shell plate materials are neare;st'to the eactor core; hence they will-s:.ustain the:great'est neutronp.exposure: E l-ach,ofthe, six!plates which make uap the intermediate and l.lower? shll copurses'were evaluated (Referencps g7 L, 21, 2,nd723)interms of initial RTNDT, residual copper and phosphorus content, and. the cumulative effect of. irradiation .on RTNDT shift. The baseq materials selected for..

the PVNGS surveillance programs, including two plates selected for Unit 1, are listed in Table 2-1.

Base] metal test materials, for PVNGS-Units 1,,2 and.3 ai-e manufactured.from sections of the shell -lates listed in Table 2-1,:. ,,The...section ofthe shell plate tused wa...s.adjacent to the. test material used for ASME Code Section III tests and was, at a distance-of at,least one. plate thickness from any .wpter-quenched.edge. This material was heat-treated to a metallurgical condition representative of the final metallurgical condition of the base metal in the completed reactor vessel.

"*."'..-i

,. 'i":*'

"* < ".:," :......................,...............-......2 "*!

.I2.1.2 Weld Material Selection . ' -. . . ..... , ,

The weld materials f6r .the PVNGS Units 1, 2 and 3.surveillance programs are selected to duplicate the materials in the lower shell axial weld seams. Table 2-2 lists the weld materials selected for the PVNGS surveillance programs.

Weld metal an'd HAZ specimeis. for PV NGS ,are: produced by weldin g together. sections from the beltline plates (i.e., the lower she'llplate. listed in Table 2-1).. The sections of the shell plate used for weld metal and HAZ test material are adjacentto ihe test material used for ASME Code Section III tests, and are at a distance of at.least one plate thickness from any water-quenched edge. These specimens are heat-treated to a condition representative of the final metallurgical condition of the weld metal in the completed reactor vessel.

asummary of the materials, and .typg_f specimens included in the six PVNGS surveillance capsules is presented in Table 2-3. Table 2-3 also defines speinmein quantities for each surveillance capsule .assembly by type (i.e., pre-cracked Charpy or comp&act tension), as described below.

WCAP-16835-NP, Rev 0 " . ; 'Page2-T June 2008

2.2 TEST SPECIMENS 2.2.1 Type and Quantity The magnitude of the neutron-induced property changes, of the,reactor vessel materials is determined by comparing the results; pf -ests using irradiated specimens-to the.-results of similar tests using. unirradiated.,

..specimens. For, example, .changes in toughness. of the vessel materials are determined using the amount of the temperature shift in the Charpy impact test curves between the unirradiated material and the irradiated material, measured at 30 ft-lbs impact energy. Drop weight, Charpy impact, pre-cracked Charpy, compact tension, and tensile test specimens are provided for unirradiated, tests:, Drop weight tests were conducted in accordance with ASTM E208.- Charpyimpact tests were coqqn4ucpd in accgrdance with ASTM E23. Tensile tests were conducted in accordance ESand.2,....I*se fdrop weight and Charpy impact tests to establish pith.ASTM initial reference temperature was done in accor-4ance with-NB-23099:of the ASME ,Code,Section III. Charpy.,

impact, pre-cracked Charpy, compact tension, and tensile test specimens are provided for post-irradiation tests.

The total quantity of specimens furnished for carrying out the overall requirements of this program is presented in

References:

21, 22':and 23. Anr'amount of base imfetalj 'weld-metal arid HAZ'test materiai sufficient to provide two additional sets of'test ispecimrens has'-beenreained with, fi1t 'o'umentatio ,and identlncation for. future use.

Each of the test materials was chemically analyzed for approximately 21 el-ements,* incliuding'al!those listed in Paragraph 5.7 of ASTM E185-79.

2.2.2 Unirradiated Specimens,_............-.,...,.

The type and quantity of test 'pecimens prbvided fort'establishirng tlie'propertieý;of the-unir idiated i &actor.-

vessel materialsaare presented in Referendes 21',22:. anid-23. Th* &dtafrot-iitests' ofthese _9pecianens pr6vide'the 6pronIerty b1sis for detenrmining the neutronincduced cha~g~ s doTthe .neaeto vesselg.materials. ":. . .'

Drop Weight Test Specimens: Twelve drop weight test specimens each of the base metal (transverse

orientatidn),'weld-nietal, and HAZ'materidal are'ptovided'fo. establishing the'nil ductilitytransiti n tempertu.re (NDTT) Of the unirradiated survefilahnce materisr. -'Thesedata'.fOn. th&basis' for the reference Yernperatute' RTNDDT, from whih subsequent neutron-induo'ed'chianges'for thePVNGS pldet and' weld 'materials!* are*:.:.

determined. ' . . , ' . '. 3/4 . D 9 , ~ '

Charpy Impact Test Specimens: Eighteen to twenty-four test specimens each of base metal (longitudinal and transverse orientation), weld metal, and HAZ material are provided for impact testing:, .This quantity. exceeds the minimum number of test specimens recommended by ASTM E 185-79 for developing a .Charpy impact eneigy transition temperature curve for these materials' Thesed h ith the op weight NDTT are used 'to establish the ieference temperature, RTNDT 'for each material.'

Uniaxial Tension Test Specimens: Nine to twelve tensile test specimens each of base metal (longitudinal and transverse orientation)'and weld metal miaterials are tprd~id&e. for tension tesfting"'. Thi' quantity*iso exceeds the minimum number of test specimens recommended by A'STM El 85 to accurately establish the-tenisil6 properties for these materials at a 'minimum of three test temperatures (e.g., amnbient, operating, andidesigri)..

Pre-cracked Charpy Impact Test Specimens: TWelve 1re-cradked Chaipy test 'specimens each of base metal (longitudinal and transverse orientation) and weld metal materials are provided for fracture toughness testing. These test specimens are'provided f6r supplemen'tal to ge P".ope'4ydieteri'iinati6in, and a:re:in addition to the drop weight, standaid Charpy, and uniakial tension sjpecimens required by ASTM E185-79.

WCAP-1683'5-NP, Re'"O' . . ,' Page 2-2 June 2008

Compact Tension Test Specimens: Four 1/2T and eight IT compact tension test specimens eadchof base metal (transverse) and weld metal material are provided for fracture toughness testing. Similar to the pre-cracked Charpy specimens, the'se test specimefis are provided for supplemental toughness-proPeiy determination, and are ini addition tothose required byASTME185-79..

2.2.3 Irradiated Specimnens Charpy impact, pre-cracked Charpy, compact tension, and uniaxial tension test specimensz are used to determine changes in the static and dynamic properties of the PVNGS reactor vessel materials due to irradiation. The type and quantity of test specimens provided for establishing the properties of tlhe irradiated materials over the lifetime of the vessel, References 21, 22, and 23, are presented in Table 2-3.

I,' -- .. *.

2.3 SURVEILLANCE SPECIMEN IRRADT., TION ,'. -

2.3.1 Specimen Encapsulation -**,, ,, .: , , . . , -

The test specimens-are housed w~ittiicrirosionsii-istanltcaps'ule asseilbllesý in an inert eniviro'nmeni to:

' spe imens by th pmary coolant during irradiation;;

" Prevent corrosion of' the.- carbo steel test

" Physically locate the test specimens in selected locations within the reacdor; and

" Facilitate the removal of a desired quantity of test specimens from the reactor when a specified neutron fluence has been attained..... . .

A typical PVNGS capsule assembly (References 21, 22, and 23) consists of a series of three specimen compartments connected by wedge couplings and a lock assembly. Each compartment enclosure of the capsule assembly is internally supported by the surveillance specimens,'arid i'sexternally pressdire~teste'd during final

'fabrication.. The wedgescouplingý, also :ser:e.as. end caps for,the. specimen compartments, and position the:;

compartraents within the capsule holders:that areattachcdto the reactor vessel. cladding.,: The lockassemblies

!fixthe-locdtions (f the cap'ules*Aiithin the-holdersand p.eventxrelative'rmotion. The lock assemblies,also serve as a point of attachment-forthe toolng used to remove the-capsulesfroem the reactor.,..

Each surveillance specimen compartment consists of two sections connected by a spacer. Capsule compartments are assigned a unique identification so that a complete record of test specimen location within eachiconpadtinentVcan be maintained. Each PVNGS reactor vesselfcontaiins six suirVeillance capsule~s, including both p *re-cr,,cked',Chayla'serbhesiand comp*6t'tensidn assembli&s. PV'NGS Unit 1 has-two pre-cracked Charpy a5 semblies' and '6oie-compact tension' assemb'y for loweir shAl-plate M-43 r I-1 specimens, and a second set for intermediate shell plate M-6701-2 specimens. The two sets of assemblies are necessary to accommodate th'twvo :diffeientiUiit 1 base' miital surveillance materials. PVNGS Units 2 and 3 have -three each of the pre--

cracked Charpy a'ssmblies'and dompa6t tension-assemblies:. Th'Seiissemblies are further described below:

2.3.1.i Pro-cracked Charpy Assemblies The pre-cracked Charpy capsu..'e assemblies consist of'three separate compartments.' Two compartments,..

designated as Charpyand. flux. compartments; contain Charpy impact (standard and pre-cracked) specimens and neutron flux spectrum monitors. Charpy specimens are provided to establish an impact energy transition curve for a given irradiated material. These specimens are arranged vertically in one-by-three arrays, and are oriented with: the notch, toward the core. The temperature differential between the specimens andthe reactor coo l ant is

- "Page 2-3

-WCAP-16835-NP, Rev 0 June 2008

minimized by using spacers between the specimens and the compartment, and by sealing the entire assembly in an atmosphere of helium.

One compartment (designated the temperature, flux, tension, and Charpy compartment) contains the uniaxial tension test specimens, neutron flux spectrum monitors, temperature monitors, and standard.reference material (SRM) Charpy impact specimens. Tensile specimens are placed in a housing machined to fit inside the compartment. Split spacers are placed around the gage length of the tensile specimens to minimize the temperature differential between the specimen gage length and the coolant. The'entire compartifrent is seaied within an atmosphere-ofheliumi. ... . .

2.3.1.2 : lise s!" .. "."

"Compact'1enislion":'... Assembhe The compact tension capsule assemblies consist of three separate compartments. Two compartments (designated as Charpy, flux, and compact tension compartments) contain Charpy impact (standard and pre-cracked) specimens and neutron flux spectrurnmcr-iltr"', -Sti*JAar'd Chaipy speciinensiare p:ovided'to establish an impact energy transition curve for a given irradiated material. The Charpy specimens are arranged vertically in one-by-three arrays, and are oriented with the notch toward the core. T Ie 112T'ci ipact tension specimens are oriented with the. opening of the~ crack.,starer n.tc1 faciagthe top of the compartment.(1i,e., to obtain a.-

uniform fluence gradient across the crack front). The temperature differential between the specimens and the reactor coolantis inlnlmlzeci by using gpaC6ers' betWeen the specimffns anitdttie conmpartment, as weli asby sealing the entire assembly in an atmnosphl ýf iehuim. ..... . .6... .. . .. ,'

One compartment is designated as a temperature, flux, tension, and Cfiarpý'6 mnpaf e Tristl '61jpaitment is the same as described for the pre-cracked Charpy assemblies.

2.3.2,. Flux and Temperature Measurement . " .

The chaanges .in,the. reactor vesselý materials toughness are defived.fiomspecimens irradiated to"'. arious .fluence levels andin different neutron energy spectra. -.,iConrmdlte.informati.3nUon the neutron flux; neutron' energy.:,.

spectra', and the irradiation iemperatureof ;the su-veillance;speciimens: is cbtained, from the flux; and temperatLre monitors in order to permit accurate interpreiatioriof-the srrllveilla?-ce material test results. .. ,

2.3.2.1 Flux Measurements ........

Neutron flux measurements are obtained from detectors located.ineach of the six irradiation capsules., Such detectors are particularly suited for th, proposed application because their effective threshold energies lie, in the low.MeV range. .(References 21, 22 and 23, provide a list, of neutron detectors installed at.PVNGS.)

Neutron threshold detectors can..be-used to monitor the thermaland:fast neutron' spectra incident;on the test, specimens.. These detectors possess. reasonably long1.aff-lives -and activation cross.sections covering the desired neutron energy range. One set of neutron flux spectrum monitors is included irn each tensile-monitor compartment. Each detector is placed inside a sheath that identifies the material and facilitates handling. Cadmium covers are used for those materials (e.g., uranium, niickel, copper,a ind cobalt)that have competing neutron capture activities: The flux-monitors are pladed:in holesdrilied in stainless steel housings at three'axial locations-in each capsule assembly.to provide anaxial.fluenice profile~f0reach set-of test specimens. ,..

In addition tb these detectors, the PVNGS surveillance programfalso3 includes standard referenice material.

(SRM) specimens. These are Charpy impact test specimens made from a reference heat of ASTM A533 "WCAP-16835-NP, RevO ... Page 2-4 June 2008

B 1. Those specimens are being irradiated along with the specimens made from the PVNGS reactor vessel materials to serve as a correlation monitor material. The changes in impact properties of the reference material provide a cross-check on the neutron irradiation environment in anyt given surveillance -program.

These changes also piiovide data for correlating the result from this surveillance program'-with the results from experimental irradiations and otherreactor. surveillance programs using the samexeference material.

2.3.2.2 Temperature Estimates -

Changes in mfechainiidal and impnat propelties of urkadiated 6pe-5ifimefis arle highly' depefndent'bn the irradiation temperature. Therefore, it is necessary to have knowledge of the temperatures of both the specimens and the pressure vessel. During irradiation, instrumented' capsales are not practical for a surveillance program extending over the design lifetime of a power reactor. The maximur~i t&mperatui6 off.theirradiated specimens can be estimated With reasonable accuracy'by in.cluding small piiepes of low,melting point:ailoys or pure metals in the capsule assemblies. The compositions oft ionitcr matie*na- atni-miitg points in the 6per.ting range of power reactors are listed in Referenices 21, 22 aid-23. Theibfiit-fe d to bracket-the- pe-ating temperature of the reactor vessel., V.,.

The temperature monitors consist of a helix of low melting point alloy wire inside a sealed quartz tube. A stainless steel weight supported by the wire ensures that monitor wires having reached their melting point are clearly indicated. The composition* aný.* theýefcre,-.thee.elting.termperatures ef the temperature monitors are differentia'te'd b" the "h"'ca lefit- 0f thee iu be-s, thi.tcontini- th-6 114

-ires.. .

A set-of teperatuire rnotors-Isictuded-in each temperature, flux,-tensio'r;-and Charpycompartment. The temperature monitors are placed-in holes drilledin stainless-steel-housings to provide the mr*iimfii temperature to which-the specimensare exposed.------

. ' -t F 2.3.3 Irradiation.Specimen-Locatien ...

The ýt'sspeimens are enclos~d within-ea-uaule ,ass-e-mblies. -They areir'radiated- at six radial- psiiion around the active-c6e iirfd i bui-ietl9 bih--idplaneof tie c0i-i_ Test sp-eciiiieiis 6ntained in the capsule assemblies arerused -to monitor the: iiradiation-inducedrproperty. changes ofthe reactor vessel materials.

Therefore, these capsules are positioned near the inside wall of the reactor vessel so that the irradiation conditions (fluence, flux spectrum, temperature) of the test specimens resemble, as closely as possible, the irradiation condition of the reactor vessel. The neutrdr'i fluence of the test specimens is within approximately 15%.of that.seen. by. th adjacent.vessel wall. Thref6ioe, the RTNL'T. changes resulting from the irradiation of these..specimens.-wil~i.closely appr6ximate the RTN-changes in_ýthe. materials 6'f the reactorsvessel.

2.3.4 -- Surveilance Capsule-Withdrawal Schedule"- . . .- -

Surveillance:capsule assemblies are withdrawn during antappropriate refueIing outage when the test specimens have *7-tiained the d-i- p for he capsule assemblies and the haeatandte sired fluence. 'Table 2ý4. presents the azimiuthl loaI~ foIh act-ial tfie of capsurle fmva[ ih terms dsffecfiVi-e"fll power years. .

The capsule a§semiblies0ldtMd in the 137-d.i0gee and 230-degree position have been withdrawn frin each of the PVNGS units: Additi6nally, the38-degree c*psuie has been WithdraWfffrom Unit 1. The actual time of capsule removal may b6 *modified to coiicdide with the refueling outage. or plant shutdown most closely approaching the scheduled time for-withdrawal-. ...... .

- WCAP-16835-NP, Rev 0 - :, '.Page 2-5 June 2008

Table 2-1

'Base Metal Materials Selected for SuryfillanceProgram Proram PVNGS Unit Plate ID Number P ate Location Unit 1 . M-4311-1 Lower Shell .

Unit 1 M-6701-2 Intermediate Shell Unit 2 F-773-1 LowerShell Unit 3 F-6411-2 Lower Shell

.*=*: ... ...

i: .:i,!..-"::"

"/,-. .*;,*!: ",................................:-...............

Weld Metal Mztelais Selected for Surveillance Program.

, PVNGS! WireffH.t uirair* 'r. Flux Type :":Flufx Lot Number,

]Unit 1I '. :-. .007i ' h 'L-,iideo0091 . . 054 .

Unit2 " " P7317"; ' t Linde 124 ': 0662 Unit 3 4P7869 Linde 124 0281 " L

.'lype'and Quautityý6f.Spe ln6en -for Irradiatio ni"E io sut'e

_______....____'_lPrciakc e Cv CmcT"ension.Assembly CApsseiily-Material Std. Cv Tension PCv Std, Cv Tension 1/2T CT

__, __ 9= - -

Base.Metal (Long.) 9

- . i Base Metal (Trans.) 15 3 9 15 . . 3 10 Weld Metal 15 3 9 15 3 10 HAZ 12 - - 12 SRM(') 9 9 -

T6tal in Alsembly:' '60' - :6 27 5'1.: 20

,,1]

Standard Reference Material (SRM):charatenized by heavy sectibh 'steel technology ........... '"' "'

program. Sp'ecimens are provided ioply forcon'relati6n'with-.diaracterizatiofi tests'. i" :..

Table 2-4 ..

Surveillance Capsule Assembly Removal. Scledule through 32 EFPy Azimuthal, Unit 1.: ,- . Unit'2 .. ' , . Uni Location Capsule Removal Time Capsule Removal Time Capsule Removal Time (degrees) Number (EFPY)(') Number (EFPy)) . Number. (EFPY)(1 !

38 1 9.76 1 Standby 1 Standby 43- ' 2 Standby -!2,, 'Standb 2- Stnby"'

'137. '7 "3 :' ': " 45 4: - '1431 . 44 142 4 Standby ' '4'4 ";; :Stan~dby-. -33>-1>* !.X3/4,fatdby' 230 5 13.83 5 14.35 5 13.75 310 ' 6 ': :18 :24 "> ' * '&': '; :"' !18'--24 i "'::,

,6::, ' : ... 1'8 -24 :*

8I -* 2' `8' 24 Removal time maybe adjusted to coincide with the refuelihng outage.or scheduled shutdown most closely app'roximating the withdrawal schedule.: ' "' '.

WCAP'-16835:NP, *Ve0.. . S. .* Page'2-6 June 2008

3.0 LOW TEMPERATURE OVERPRESSURE PROTECTION:'

SYSTEM LIMITS 3.1 LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM 3.1.1 Intrduction.

The low temperature overpressure protection (LTOP) system protects the PVNGS reactor coolant system (RCS) pressure boundary integrity by ensuring that the pressure rermains below the.applicable. P-T limits of 10 CFR 50, Appendix G, at low temperature, cQditions when theRCS is water-solid.

Technical Specifications require two shutdown cooling system (SCS) relief valves ("LTOP valves") to be operable or the RCS vented with a vent area ,greater 14an or eg a9tpoj, square inches when any2RCS cold leg temperature is less than the specified LTOP enable temperature. Once enabled, the LTOP system automatically maintains the RCS Oressiir&"e1eothebrittie friPire thr'e'shold withOut requiring action or intervention by the plant operator. * , ' , " - ' I This Section describes the process for developing the LTOP systerai.imitsand demonstrating adequate,..

LTOP performance at PVNGS Units 1, 2 and 3. :.

3.1.2 LTOP Technical Specifications .

Protection of the PVNGS reactor coolant pressure boundary against brittle fracture isoprovided during.

Mode 4 heatup or cooldown, when all cold leg temperatures are above the LTOP enable temperature, by at least one operable pressurizer safety valve with a lift pressure setting as specified in Technical, Specification LCO 3.4.11.

Similarly, RCS brittle fracture protectimn ls provided-when any RCS 'cold'I leg temiper'ature is less than or equalto the spe~ifi~d LTOP. eriable teml'erature by' t'oýoperabe 'SCS relif Valves f6r, if the RCS is depressurized, by providing:an'RCS vent area *qual to 'or gteatertthai' sixteen square inches:, This limit on plant operations applies in Modes 4, 5 and 6 as specified in TS LCO 3.4.13. TS LCO 3.4.13 also prohibits startup of a reactor coolant puijp inriMode 4: if the' seoiidaiy- sidic"W-ater ieniperature in' ýithei stean' generator is more than !00°F above:any.RCS cold.leg temperature. .' ,

3.2 BASISFOR LTOP SYSTEM LIMITS , . .. - , ' ,

PVNGS LTOP system and controls develop using the ethool6gy approved by the staff in CE NPSD-683-A, Reference 1, and the plant-specific methodologoypresented in this report:'

Technical Specifications require! two SCS. suction ,!in('JTTOP',) relief valves to be operable, although event mitigation only credits one valve. along with controls on the heatup, and cooldown rateý whenever any RCS cold leg temperature is below an enable setpoint. The SCS relief valve se'tpoint and capacity, and the LTOP heatup and cooldown enable setpoint temperatures support analyses that ensure the peak RCS pressure resulting from postulated overpressure events does not.exceed.the allowable, RCS P-T limits.

Calculated limiting temperatures for LTOP hedtup and cooldown rate protection are given in Table 3-1.

Between the minimum boltup temperature determined in accordance with Section 5.6 and the LTOP enable temperature determined in accordance with Section 5.12, the peak RCS pressures resulting from postulated

,WCAP-16835-NP,Rev 0 ' Page 3-1 June 2008

overpressure events does not-exceed either the minimum pressure value determined in accordance with Section 5.7 or the corrected allowable pressures determined in Section 5.14.

3.2.1 Peak Trafisient Pressures Analyses of the limiting case mass addition (HPSI start) and energy addition (RCP start) events are performed using conservative LTOP assumptions to establish the peak transient RCS pressures at PVNGS.

Calculatedresults ccfifirmh that the shutdowin"cooling system relief-valves' (SI-179, SI-189; also-termed-ý "LTOP valves"):installed at PVNGS' Units 1, 2.`and 3 Ihave Viccess capacity relatiVe' to relieving requirements. Therefore, the limitingmas's~a'nrd e'nergy~additicn oVerprIessure events are qui*kly ierrininted upon valve actuation.

These design basis&p6ik pr~sgur anaiy~esihc5ipA~tet: folli0iMing'LmTOP assumptions:

The. pressurizer is initially water- solid with no, steam space and no credit for the presence of a cover gas (e.g., nitrogen), . ,

The RCS pressure boundary is rigid, i.e., no expansion due to pressure or thermal effects,

No heat is transferredto or:front the: RCS,:",: '." . :" . ..

The RCS letdown flow is isolated, " ': ' -" ,

" All pumps attain rated speed instantaneously,

Only one SCS relief valve actuates to mitigate the transient,< "...

. No 'operator'action-is requir d, and " - , - ; . ;, .

Conservative energy addition sources are used to ev4auate both energy. addmon a mass-addition transients, including: . ....... ' '

o Full heat output from all pressurizer heaters (1800 kW) is assumed for the duration of the transient in prder to maximize. the energyinput into the,RCS,; and ,

0 Decay. heat;, with .two-sigma.uuncertairltie~s ,isassumed~constant th~rughoithet transient ,I :."

at a value consistent with the earliest time after shutdown thatithe-,transient'can occur._,

The following additional assumptions are made to ensure a conservative ,analysis:

" The SCS is assumed isolated at the starof the transientmin order to minumzeethe total ,

volume absorbing the heat/mass addition and to isolate any heat removal from the RCS,

" The SCS relief valve opening profile is consistent withth* A'SM'Erodel'described'inv' -.

-Section 312.1.1; this model results in a ,delayed resppnse.to the relief valve lift anda .. , ..... , .

resulting delay in providing the relief 6apacity .

No RCP seal leakage or controlled bleed-off is assumed, The&RCS is isothermal and is-not cooled orheated b.y an m.s.s "additionand;

The initial conditions are chosen to maximize... the pressuire transients I- in-ordeirto-develop

. .'.,* : i ',-;" :

the greatest rate of pressure rise. " . .... .

The transient analysis methodology used to establish resuits ftr tlre design basis mwio temperature overpressure events at PVNGS Units 1, 2 and 3 is discussed in the following sections.

WCAP-16835-NP, Rev 0................"". . . . Page.3-2 June 2008

3.2.1.1 Relief Valve Overpressure Protection Low temperature overpressure protection at PVNGS Units 1, 2 and 3 is provided by spring actuated relief valves

("LTQP" valves) installed in the SCS suction lines. These relief valves are placed in service during RCS cooldown at or above the LTOP enable temperature, and at or below the eniable temperaftur during heatup, to ensure the reactor vessel is protected from brittle fracture in the e.'ent of 'a low temperature overpressure event.

The SCS relief valve opening and discharge characteristics are c6nsisteient anid colnseriviV6 'elative to the ASME Code requirements for spring loaded safety valves. These :.lalyes follow theASME,Code model with an initial SCS.relief valve opening of 30% at 3% accumulation and fulikopening atI 0% accumulation at which point each valve has a relieving capacity of 5635 gpm. This relievi;ng 'capaqctyis substan*ti2!ly~greate*r tha that needed to mitigate the design basis RCS mass addition and energy. additiou transients.

The setpoint for SCS relief valve actuationa.lift'pressure of,467 'psig as definediinTS LCO 3.4.13,i£ established by the limiting pressure to which any SCS component (e.g., LPSI pump seals) nidy be exposed. This setpoint is sufficiently below the limiting reactor coolant system pressure established by 10 CFR 50 Appendix G. Therefore,lovi 'temperature overpressure prot.cfion-o)f thereactor vessel is assured since the calculkited RCS pressure rise is .terminated befoie the ,reliefvalve.reac~s ffall-open' i' both thedifiihmtng .energy-addition and mass addition LTOP.transients.,: . "', 'B K ;'-'.': :", -, , . ..... . ,. : . "..

3.2.1.2 Mass Addition Overpressure Event The limiting design basis mass addition transient is an inadvertent actuation of two HPSI puimpsýwhile all thr~e charging pumps: ahe' oiei'ating at their-desigin'flo-Vrate.. Massaddition~fromn safety injection. tanks is not coniside-ed, i.,this-t-ansient; since.Technical .S4ecific.atiobn 3.5.:2 .allowsý the t5nks to be isolated during Mode 4.

,:.operation ,viththe .pressurizerpressute;below 430 psia., . . :' .' ' . ..

This event is analyzed using the methodology of CE NPSD-683-A. That method determines inputs for HPSI massaddition; charginig pump mass additibi4 and the eluival&nt mass addition'that results from energy addition.

The iimagriitude*if the pressurization. is deteimined:by superposition of.the mass addition-curve onto the relief vdI, e:discharge.,curve, -both.of wihich desdribe, mass' flow rate as'a functior of piessuri~er pressure. Thei-eqtilibrium pressure is taken as the intersection of the two pressure c&Vii-~es' ass umifig liquid input and discharge, at which the mass addition rate matches the relief valve discharge flow rate.

The HPSI mass addition is obtained from the maxikfuiii volumetric delivery curves developed for emergency core cooling system calculatins using the pressure difference between the reactor coolant system and the refueling water tank:. A conser.vative (low) temperature is assumed. for the, supply.water to establish the greatest rate of mass addition. ' ' -".

A rapid RCS pressure.rise.occjgrs upon initjation".f a design basismass aiddifich transient at PVNGS. The transient is quickly mitigated"5'with.a peak R.CS,plressure.at the.pressurize of less than 499 psia, due to the large capacity discharge throughý.theSCS.suction lineirelief valves. ...

3.2.1.3 Energy Addition Overpressure Event The limiting energy addition LTOP transient at PVNGS is modeled by simulating the pressure increase in a water-solid, idle RCS due to reverse heat transfer from a hot steam generator when a' single reactor coolant pump is started. This model, described in Reference 24 and applied to LTOP analyses for PVNGS, simulates the discharge from a relief valve and determines the RCS pressure during the relieving action.

'WCAP-16835-NP, Rev 0 ".. Page 3-3 June 2008

The following conservative assumptions are included in the analysis of the design basis energy addition

ýtransienit:

'The steam generator secondary temperature is assumed to b' 100°F hotter than the primarycoolafit temperaiture throughout the transient, 0 One RCP is assumed to start and instantaneously reach rated speed to initiate the transient. The model assumes a constant heat, input for the duration of the analysis,,.

The RCS bouridary-,i§sassumed iiid; i.e.,4 there-is 'noincrease in RCS volume -with an increase in..

-,."RC-S pressure or 'temperafilre. Alro,-R.CPPseal leakage is assumed and the letdown flow paths are isolated, vdth masswrleaseonly thrroighxveliof valves, :

The RCS pressure at the initiation 6f the transientt is selected to bie'435 psia', and'

The raethod for the determiriing; the ,decy 'heat,.conrtr~biition is cZonsistent with-that described in CE

" .. :" i . -. '... ' '-

With the reatively large capacity SCS relief.valves installed at! VNGS, the, energy. addition pressure

.transientis quickly'mitigatedlupori, the .valve-opening:a'3%arc.um'ulation. Analysis Con-frms that.. the peak pressure at the pressurizer during the energy addition transient remains below 499 psiaand complies with,"A' NRC requirements in a conservative manner.

3.2.2 - Applicable.Pressure-Temperature Limits . ' .. . ... ..

Applicab!e P-T heatup and cobldown limits :iised-.to *fipport LTOP.,:.ontrols *are established based on the:,-, :;:.,

methodology described in CE NPSD-683-- .Using this methodlohgyAthe limiting ratesi.are:-sel.ected* based'on a comparative evaluation of the family of heatup and cooldoawri.rate.basedtpreessure?.temperature limitsi(described in Section 5.0) to the peak transient pressures described in Section 3.2.1.

The applicable P-T heatup and cooldown rate: limits; ,developed in Sections 5.6 and,5. 2 lrespectively`.are:

dontrolling.over the range of R.CS, temperatures from thetminimumi hblt up:temperature (80TF).to the LTOP'-F enable temperature (221 'F).- RCS heatup: and cooldown rates ,less'than oi .equal to. those .shown.in ,Table33-1

'apply to Units 1,-2 and 3.through 32.EFPY:.: ". i ""*.: : .i . " 1., ",, ',.

,,RCSHeatup and Cooldown Rate, Limits through 32 EFPY Indicated RCS Cold Leg, Heatup Rate ' Cooldown Rate' Temperature ('F) ( 0F/hr) ( 0F/hr) 80 0 to 920 < 75 < 30

>92°to.100' ' ... . <75. -5

> 100to°<2210 j .2 .<75:, ,00 ,

>2210 2, 75 j, !00

, 2...

, ? ' ." .: . : : , . : : * : .2' : .") " " . )" . . i !., "

WCAP-16835-NP, Rev'0 -... 'Page 3-4 June 2008

4.0 ADJUSTED REFERENCE TEMPERATURE Adjusted reference temperatures (ART) for the reactor vessel beltline region are deterni*ned using NRC-accepted methodologies, as described below. Limiting ART values for the PVNGS Units 1, 2, and 3 reactor vessel'beitline regions corresponding ty.32 effective full poxwer years (EFPY) f6rthe-114T. a'nd 3/4T locations are:

Limiting Material Location ART Unit 1 -"intermediate Shell Plate M-6701 - 4T" : " . 60F Unit 1 - Intermediate Shell Plate M-6701-2 3/4T 1030F RTPTS values for the PVNGS Units 1, 2, and 3 reactor vessels are calculated in acco~rdance .with 10 CFR Part 50.61 using the neutron fluence at the clad-to-base metal interface through 32 EFPY. The highest predicted RTPTS value is 123'F, and corresponds to the intermediate shell plate M-6701-2 of Unit 1.

Section 4:4 describes the determination of RTPTS . -

4.1 BACKGROUND

Determiniati6n of the&ARTf aduef-rTtiheePýVNOS Uniits P1, 2,; and3"feadtoirVessel beltline materials throiugh 32 EFPY is discussed in this Section. ART values are coiistf'AiiVe Wiitfi tespect to the measuremenit 'obtained' from the irradiated surveillance capsules, as described in Section 7.

4.2 RE .

The pred:icteo adjusted reference temperature and RTPTs resil~s through 32 EFPY for-the..limiting material in each PVNGS unit are summarized in Table 4-1. All ART values are conservatively predicted relative to the measured properties, i.e., to the Charpy transition terniperaituisfhifts mneasured as parit of the suiveillainc'"

capsule evaluations as discussed in Section 7.

4.2.1 fluence Calculation'~._:,-,.. K .

Thde'peaak desigin fast flue'nce a t the vessel clad-to.base rnetaf interface through 32 EFPY is 3.29E+ 19 *cm For conservatism, if was assuiied thai thisý pea' fiuence i§ applied*oeach ofthe PVNGS reactor vessel beltline plates and welds; i.e., no reduction factor is applied to account for axial or azimuthal variations from the peak value. From Table 1-1, it should be noted that the peak calkulatedrneutron fluence thfough;32.EFPY based on evaluation of irradiated surveillance capsules is substantially less than the design yalue of 3.29E+.19 n/cm 2 and reflects plant-specific fuel management.

The following equation from Regulatory Guide 1.99 is used to determine thee attenuation of neutron fluence with distance into the'plate: . ., ... ., - .

2 4 f furf (e x) (Eqn. 4-1)

In Equation 4-'l: .K.. ',

f . . = neutronfluence at-the. desired location,,

S*fsurf=neutron, fluence at the inside' vessel.surface.,

x .. distance from. the inside~vessel surface, inches......... . .

WCAP-16835-NP, Rev 0 Page 4-1 June 2008

For PVNGS, the given neutron fluence is assumed to apply at the clad-to-base metal interface. The distance 'x' is measured from the clad'to-base metal surface into the plate. -- .

4.2.2-. Chemistry Factor Calculation .

The chemisti fadior'(CF) is determined using Tables 1 ahd 2 of Reference 8 for the welds and bbase metal (plates), respectively. Chemistry factor values determined fof each"of these PVNGS matridas and ihe' respective initial RTNDT values are reported in Tables 4-2 through 4-4....

Surveillance data.when available, can'be used'd to deternine a chemistry,factor accordingI to the following equations froin-Re&ference 8:......................... ..... ....... ......

A RTNDT (CF)

ff (Eqn. 4-2) where:, .. . . 2 -

7- , ...... (*Eqn.4-3)

One test of the validity of the estimated chemistry factor consists of calculating ARTNDT for a given fluence and comparing it with the measured. ARTNDT for that fluence. The measured ARTTrýT rmist fall within plus or minus one sigma. (C-A), of the calculated ,-RTq-_, where l7,F.for base metal-and 5,6, 2-87y .fgr,.welds I,

(Reference 8).,. That assessment is described. in, Section..7._. . 'li.:,- . '- ; : -

4.2.3 Calculation of ART for the Limiting Plates at 1/4T and 3/4T Adjusted reference temperatures are calculated using the following equation from Refiehe-8 ? '-`

ZART =Initial RThDT4 ARTNDT +/Iargin: ' " - . '- , -" *(n

'4'5)d" All temperatures in Equation 4-5. are in degrees Fahrenheit. *-

,,,,.,i .?*i?' . ;-, . "".;:*5

.".. .'*. 'i'... . .:.,. .:,',-.' '..

4.3 ANALYSIS DETAILS Ferritic materials in the PVNGS reactor vessels that may have accumulated ki-e~atron'fluence' i....cess of.

1.0E+1 7 n/cm 2 are assessed according to CE NPSM-683-A, Reference I.. The materials considered cornpris-the lower, .intermediate and.upper shell course-plates, bottom head plates and included welds..

., , . . . . .. ;. * : . .. . , . . . .. . . . . . C,:.

. . - , .. ... ":. ' K, U * ',-

4.3.11. Selection of Representative'and Limiting. Cases: ',c: , 1,, .:c . .

For each of the PVNGS uhits,-the tentlc piat'es and weld iiimterial§lo66'ted within thie region-lmmedltely surrounding the active core are evaluated to identify the limiting material Af the'li/4T and3/4Tfidatio is".That includes materials fromjthe intermediate shell course for which only a small section is within the effective height of the"tore buit receives sufficient ntitron li-radiatioifýo*econsidered.i.'thie-idenitifidation 'ofthe most limiting beltline material. The adjusted RTNDT Values for the limiting beltline materials are used to establish the heat-up and cool-down limits. .

The neutron fluence may exceed 1.OE+17 n/cm2 for the ferritic plate and weld materials located above and.

below the region immediately surrounding the active! core. However,-the:.fluencein .that surrounding region is too low to make it more limiting relative to the sectionvwithin the effective height of the-core. The initial RTNDT values of those ferritic materials are comparable to the inital RTi-m$ valties f'orthe materials -fromthe lower and intermediate shell courses; therefore, the adjustment to RTNDT (i.e., the predicted shift) will be smaller than for WCAP-16835-NP; Rev-0 "Paft'4-2 June 2008

the materials surrbunding the effective height of the core. Hence, the ferriticplate'anid Wld miter'alsi*oate'dt above and below the region immediately surroundling the effective height of the core in the.PVNGS'are not limiting 'with 'respect to estdblishing'the heat-up'and c6ol-downlimits. These additiofial materials comprise the upper shell course and bottom head plates and welds that are located immediately above the intermediate shell course and below the lower shell course.

Tables 4-5 through 4-7 present the projected ART values at l/4T and 3/4T locations in the reactor vessel wall for each of the PVNGS units. ART values are determine'd using the full vaiue of nIargin, i.e., no credit is taken for the case where the predicted shift was less than two standard ideviations (.,UA; see Section 4.4)."As shown in Table 4-1,tlfe mfiateridl With the highestpredicted ART at 32-EFPY forPVNGS -is the Unit 1 intermediate shell plate M-6701-2. -Note in'Table 4-5 -that comparable resuilts- are -als obtained for-plate-M-6701-3........

4.3.2. Calculation of Fluence at 114T and 314T ...

As stated in Section 4.2.1, the design value of peak fast fluence through 32 EFPY for PVNGS Units 1, 2, and 3 is 3.29E+19 n/cm2 (E > 1.0 MeV) and applies to the vessel clad-to-base metal interface. Table 4-8 provides the values of fluence and fluence factors through 32 EFPY'determined at the 1/4T and 3/4T locations in the vessel wall. These fluence values are calculated "sing 4-, and ae used in the calculation of ART in Tables Equation 4-5 through 17-3. 4,;7.- Fluence factors listed injTable 4-8 .ae-a,!culated using Equation The following values of depth (x)areused for iteediiteadlowe ihiels:. hieie Location at 1/4T; ".'." at 3/4T

"-- ......... Intermedihe shell s xx =2.80

= .0xm . in. * /x 8.39,in. ;' ,

LowerShell x = 2.27 in.: x = 6.80:nn.'

4.4 LIMITING ADJUSTED REFERENCE TEMPERATURES The. adjusted-reference temperatures for PVNGS at 32EFPY. are calculated usingEquation 4-5. Chemistry factor and initial RTNDT values for each of the PVNGS beltline materials are given in Tables 4-2 through 4-4.

They are calculated using the values of fluence (f) and fluence factor (ff) given in Table 4-8. Margin values are derived using Equation 4-6 (from Reference 8) as-'.,

((.V2 2'* 1/2. . . .. . . . ... . ., . . " .. . .. . ..( n .4 6 M argin .=2 + 3A..)

+... -... ... ... ... (Eqn. 4-,6)

The stanidard-deviatlon for the inifial!.RTN'r,-di, is taken-as 'zero, givehnthe 2use of rfmeasured values for each of the PYVNGSjplates and welds.- The.standard deviation for the predicted.shift, Ga, is 28F.fdr w'elds arid .17F for plates... Theiefore, with c7 being zero, ihe margin becomes twic e the value of (. Per Refe'rence' 8, CA need not exceed one-half of the predicted shift;.:The projected ART values for each of these materialsgiven in Tables 4-5 through 4-Tare determined using the full value of rhargin, i.e,,.no. credit is taken in-this case when one standard deviation for shift exceeded one-half the predicted shift.*.

The'high6ýtcalchlatiid'Valiieg 6f ARTo-fi -each PVNGS Unit appierih Table 4-1.- The highest adjusted refetenie temperature is used tw define the-P:T limits for PVNGS. Units-1; 2 -and 3 through 32 EFPY as described in.Section 6.- - - - - - - - - - - - - - "

Similar calculations are performed for the limiting plate and weld at 32 EFPY following the fracture toughness requirements of 10 CFR Part 50.61. RTpTs is calculated for each of the beltline plates and welds using the values of chemistry factor and initial RTNDT shown in Tables 4-2 through 4-4, and margin as WCAP-16835-NP, Rev 0 .. Page 4-3 June 2008

described, above.- Calculations are based on a peak fluence at the vessel clad-to-base metal interface through 32 EFPY of 3.29E+i19 n/cm 2 in.both the lower shell and in the intermediate shell. The highest predicted RTPTS is 123°F and corresponds to-intermediate shell plate M-6701-2,in Unit 1. Projected RTPTS values through 32 EFPY for the limiting material at each PVNGS Unit are shown in Table 4-1.

.. . Table 4-1

,Surnmaryof Limiting ART and RTprs Values throuigh.32 EFPY J,

,;"i Unhit oCation, Mite rii` i1i4T ART (OF) 3/4TART (F) RTPTS ('F) 1" Inter. Shell Plate M-67012 116 103 123 2 Inter. Shell Plate F-765-6 74 64 78 3 Lower Shell Plate F-6411-2... .,65 ,; =__57 .68

- ,Table 4-2 ,

ART Input Values forUnit .I Beltline Materials Beltlirie Material Plate or Weld Numbrf *hemstiyi'Fattcr (0F)' --Initial RTNDT (CF)

Lower Shell Plate M-4311-1 ', ':26.10.

Lower Shell Plate M-4311-2 20 -40 Lower Shell Plate M-4311,3

20 - 20 Intermediate Shell Plate, :' M-6701: II .- 44 / , +30 Intermediate Shell Plate,' M-6701:2.. 37; , +40 Intermediate Shell Plate M-6701-3 37 + 40 Inter. Shell Axial Weld 101-124 35.45 - 50 Low. Shell Axial Weld " ,101-142'

ý : 27,8w " ______...'_-80 _

Inter/Low. Girth Weld 101-171 3.4.05 -- 70 Table 4-1 ART Input Values for Unit 2 Beltline Materials Beltline Material Plate or Weld Number. Chemistry Factor.(F) ,Initial RTNDT(F)

Lower Shell,Plate .. F-773-1 .' 20, .1 . , .,

Lower Shell Plate F-773-2 ,-26. 0":,;

Lower Shell Plate FF-773-3

- 3. , 31.t , -6.0 Intermediate Shell Plate ... F-765-4 ' 20 , -20.-

Intermediate Shell Plate F-765-5 .0 , 0. . . +:10-Intermediate Shell Plate F-765-6 26 + 10 Inter. Shell Axial Weld 101"2f-124 33.6 ' -60 Low. Shell Axial Weld 101-142 4'4.2 - -80.

Inter/Low. Girth Weld 101-171 26.55 -30

WCAP- 16835'-NP, R&0VO Page:4-4 June 2008

Table 4-4 ART Input Values for Unit 3 Beltline Materials Beltline Material Plate or Weld Number Chemistry Factor (OF) Initial RTNDT(OF)

Lower Shell Plate . F-641 1-1 26 - 40 Lower ShellPlate . F-6411-2 26 . 0 Lower Shell Plate F-6411-3 26 - 60 W

Intermediate Shell Plate . F-6467-4 26-, -- 3.*-3'0 Intermediate Shell Plate F-6407-5------- -- - -1 ' .. - 20 -"

Intermediate Shell Plate F-6407-6 26 - 20 259 Inter. Sheli Axial Weld _101-124 . . . -- -50" Low. Shell Axial Weld 1i01-142 . . 30.65 . - -50.

inter/Low. Girth Weld '101-4 1- , 10 Tab'.e 4-5 Predicted ART Values for Unitit Biltline Materials thirough 32 EFPY Plate or Weld Predicted 1/4T 1/4T ART Predicted 3/4T 3/4T ART Number Shift ")9 (OF) Shift (°F) (OF)

Lower Shell Plate ,*r .M;43 -. : 3 j" '-..

33..: 55 .. - 23 47 Lower Shell-Plate. ....

> 31913 F 2 .24 .. 8 . .8 12 Lower Shell-Plate. 7- M431 13 -24

-. . 38 18. - 32 Intermediate-Shell.Plate:.- M-6701[2 .. l114.. " 23 -1 983 Intermediate Shell Plate - M-6701-2 ..2 .. 116 29 103 Intermediate Shell Plate M-6701-3 42 n4- ~ 2 103 Inter. Shell Axiai'Weld . 101-124 .. .... 4-1 47-... ........ 27 33 Lower Shell Axial Weld 101-142 33 9 24 0 Inter./Low. Girth Weld 101-171 40 26 30 16 Table 4-6 Predicted ART Values for Unit 2 Beltline Materials through 32 EFPY Plate or Weld Predicted 1/4T 1/4T ART Predicted 3/4T 3/4T ART Number Shift (OF) (OF) Shift (OF) (OF)

Lower Shell Plate F-773-1 24 68 18 62 Lower Shell Plate F-773-2 31 65 23 57 Lower Shell Plate F-773-3 36 10 27 1 Intermediate Shell Plate F-765-4 23 37 15 29 Intermediate Shell Plate F-765-5 23 67 15 59 Intermediate Shell Plate F-765-6 30 74 20 64 Inter. Shell Axial Weld 101-124 38 34 26 22 Lower Shell Axial Weld 101-142 52 28 39 15 Inter./Low. Girth Weld 101-171 31 57 23 49 WCAP-16835-NP, Rev 0 Page'4-5 June 2008

... .... .. T able 4-7 ...

--Predicted ART Values for -Unit 3 Beltline Materials through 32 EFPY

. Plateor-Weld Predicted 1/4T 1/4T ART Predicted 3/4T --3/4T ART Beltline Material I i M Number .- Shift (9F) .. ).. Shift (°F) . . )

Lower Shell Plate1. F-641-1 ......... 31 25 23Y . '.17 Lower, Shell Plate .. . F-641-2 ....... _31 65 23:, 57 Lower Shell Plate, F,641'.-3 . 31_.. 5 .- I 323 IntermediAte Shell Plate- . F-6407 .- 3 . 34 20.. 24 Intermediate Shell Plate ... F-640_7-5 .. ____.._. .____ 50 24 38 Intermediate Shell Plate .. F-640,-'6 ... .. .3Q. - 44... . : 20 34.

Inter. Shell Axial Weld 101-124 30 36 20 26 Lower Shell Axial Weld 101-142 36 42 27 33 Inter./Low. Girth Weld 101-171 40 26 30 16

,2.

dtable414 aE Fluence and Fiunece Factors at:1/4T and 3/4Tthrough 32 EFPY L-ocation-- . ..,1/4T -f

.(n/em2)() - 11/4T_.f* --.- 3/ f(ne 2)(1)i ..... 3/ Ti~f*-*'('

Inter. Shell.... -1.681E+19 1431 4 .. 8 . ...-0....390E+

Lower Shell 1.910E+19' 1.1770 . :.. 6.4ý38E*18 i .866

() nutrofluence uton er-.

(2) ff fluence factor per Equation-4-3. - =- '- "-

JuneWCAP-2180*NP08evb .". Pag 4-6 June 2008

5.0 RCS PRESSURE-TEMPERATURE LIMITS RCS pressure-temperature limits developed for PVNGS employ the analytical methodology approved by the NRC in CE NPSD-683-A, Reference 1., and use conservative reactor vessel beltline re sildualfracture toughness values based on those obtained from post-irradiation examination of surveillance specimens.

The pressure-termperature limits for thebeltline region are combined vwith non-beithne regions, as appropriate, to develop the set of composite curves for the'.opjerational modes at PVNGS. The lower bound of these composite curves'defines the pressure-tempeirtuire limit for PVNGS Units 1, 2 and 3 for eclh opera'tiofial mode.

Pressure-temperature limits establishled for non-beltlihie locations do not change significantly a'scompared with the most limiting 1/4T and 3/4T beltine locations during the-vessel lifetimie due to the. lower exposure to neutron flux experienced at nonr-beltline locations. Non-beltmef n aitons are considered when updating RCS pressure-temperature limits thrfighout plant 1f6 s'ince these'locations are embodied in the plant design basis.

5.1- STRESSUINTENSITY FACTOR, .... , ,,,"

5.1.1 General The analytical procedure for developing the PVNG S r act- "e` e tkime regionP-Tlim.ts use'stle" t _.

methods of linear elastic fracture mechafics and the guidance found in A SME.Section X1. Appendix G, Reference 25., For thes- ianaIyses,sthe ,.Mode. I (C'crackopening". mode) stress initensity*factors are used for the

,solution basis couoledwith a superposition technique and the influence coefficient methodology described in Pressure-temperature limit curves for the reactor coolant system are established based on the beltline limits for a family ofheatup and cooldownrates. These PjT curyes aef then corrected to account for pressure. .

elev*ation effects and-instrument measurement unceritaintv.: The ad ustment addresses RCS hydraulic pressure drop due to flow and corrects fqr the.change in elevation btween the.vessel beltline region and the .

pressurizer, as well as for temperature and pressure instrument uncertainties.

The final P-T limits also identify the minimum boltup' temperhature, loiest service iemtperattife and'the flange limit:' Both the minimum boltup temperature and the'lowest service temperature are determined usingthe available materi'a information. 'Minimum -tempeiafure requirements' for PT, limits are determin ed using the

criteriaiestablishedby 10CFR 50 Appeadi' G. LTOP enable teImperatures are'deteriniried using the heat transfer results and-applying theASME Code criteiia of Reference 25.. j".

5.1.2 Pressure-Temperature Limits Calculation Combustion Engineering developed afinite element methodology to calculate the .allowable pressures due to membrane stress intensity factors in reactor pressure vessels. This methodology,.approved by the staff in report CE NPSD-683-.A.is used. to.calculate. the.PVNGS reactor vessel beltline pressure-temperature and LTOP. limits during RCS. heatup,, cooldown,. isothermal, and test conditions.. The application of this methodology is.briefly described in the following ýections., . .. ... .

Heat Transfer Analysis . . .,

A one-dimensional, multi-node, finite eleinent heat. transfer-model is usedto compute the thermal gradients in the reactor vessel beltline region. The model is solved numerically to establish the temperature distribution through the vessel wall as a function of radius, time, and heatup or cooldown rates. A convective boundary condition is applied to the inside wall of the vessel and an insulated boundary on the outside vessel wall.

WCAP-16835-NP, Rev0 * " Page 5-1 June 2008

Variation in material properties are modeled using average properties for each material over the temperature range of interest assumed in the analysis.

Crack Tip Stregs Intensity Factors for Membrane Tensile Pres"ure' Stresses.

Two-dimensional fiiite 6lement models, one with a' postulated crack at onelquarler of the vessel wall ihi~klness and a second with a crack located at three-quarter of the vessel wall thickness are employed to analyze the crack tip stress initensity. Theiinmhodol6gy for 'alculating the stiess intensit* factor correspondini to mnembrahne tension resulting from pressure loading of the reactor vessel, KIM, involves these finite element miodels.

These finite element crack models are loaded with internal pressure on the cylindrical inside surface. For the, case with an inside crack, pressure was 4lso' applied to the~crack face. Membrane, stressintensity factor influence coefficients, KIM, are then computd for a unit internal pressure of 1,000 psi. Finally, the KiM coedfficientis are extracted6frf - rals assulng a crack'aspect ration of1:6 and corrected using ASME Code ratios to account for three-dimensional versus two-dimensional flaws. Corrected values are used to calculate the stress intensity factors at any applied internal pressure. The membrane stress intensity factor coefficients, KIM values, due to unit (1,000 psi) internal pressure'-loadmings for PVNGS reactor vessel geometry are then used in place of the simplified pressure stress and the KIM values available in the ASME Code.

Crack Tip Stress Intensit Factors for Thern Stressvs';.

l '.,.' ' ,.,, . -..

A temperatureprfile-ba'sed supe'rpositionte*lfhi4d&ii 'sed t6e'sftblish-thie crack 'iesttestip intf'eniy' factor dfie to 'thermal stress; with a third order pdlymndiiiil fit ie'ld t6-i.*dde' the'l r'ni &ridistribitti6f0s' in the Wall: Unit stiess intensity' factors, Kl,; are"calculated 'foreach' i*t-m of thAepolyiimnl Usin'ga tW6d*imensiona1 finite eleinent modeling method. These unit values are then summed to determine the total K, value for the a*pp'lied loa ds, under any general temperature profile in the wall that occurs during the thermal transient.

Temperature-base§d influ'ence Ioeffii6nts' are used .to6 edeteifhiine` 'the -theth!ai ...in gity ft 6fator:

§'sres ian location. -'Using'miethods from* CE NPSDl83di tese;udent pdtusfig a"-tibo:dimhnio'af, reactor vessel model'with a crack adjiisted to a'c*dutit f6i? tli~ediiir nal effecit'. '* ' --

5.2 FRACTURE TOUGHNES$ CRITERIA The reiereace'pressure stress intensityKlR;.is obtained from'a reference fracture'toughness curve for~feactor:

vessel low alloy steels and iskdefined in.AppendicesA andGof Section XI of the ASME :Code...,Thisarefeaence pressure' stress intensity isideterminedby two properties, KiA'and Kic.that represent critical values.of thestress intensity factor. In this report, KIR is defined as Kri; with KIc defined as .the lower' bound.of static 'initiation.::,

critical K, values measured as a function of temperature.

ASME Code Case N-640 permits application of the lower bound static crack initiation critical stress intensity factor~equation (i.e-, K16 equation)`as the bas'is for establishiiig the P-T curves-sin hieu-0f using the Io'. w"ebound crack 'arrest.critical stress intensity fatc 6r equation (i.-e4 Ki2'whch' is'based on. c6nditi . ns eetd"" toarrest a' propagating crack,' and whichiiS the method invo'ked by AIppendix G1to Sectioin-X[ of the ASME Code. -Use of the Kiequatioiuitotdeiermine the lower bou'ind fractu'retoughnies whe'ncomputing P-T-tuives is. more .

technically correct than the use of the KIA equation since the rate of loading 'during a heitup' or codld6wn is slow, and. since crack initiation, which is more representative of a static condition than a dynamic condition, is principally at issue. The Kic equation appropriately implements the use of the static initiation fracdure toughness be'havior to evaluate the controlled heatup.and cooldown process. of aireactor vessel., ' .

WCAP-16835-NP," ev 0 .... ' Pag"e 5-2 June 2008

5.3 TRANSIENT PRESSURE-TEMPERATURE LIMITS A thermal stress intensity factor, KIT, is calculated at the 1/4Tand 3/4T crack tip locations for anyinstant during the plant heatup or cooldown. Typically, thermal stress intensity factors are calculatedusirig the temperature profile through the vessel wall as a function of time. These stress intensity factors are subtracted from the available.Klc value to find the allowable pressure stress intensity factor and, consequz.nti.y,;-the-allowable ,

pressure. For PVNGS, the allowable pressure is defined as:

'[..:, ~~~~~~~~

~ ',* ~~..".

J .... ,*....... . "... .. . : . ....

KKITcIT -.

P.=11 c2k for heatup and cooldown transienis;.-and, . ., ,(Eqn. 5.1) 2kiM

, ,. 2k1

.:*"C. o steady-state (isotheimail) cdt iditri dt , ? '--i "::, !(Eqna..5-2) :

where:

kIM = membrane load stress intensity .factor due to -unitinternal pressure, ksi'/inpsia, KIT . =thermal..load stress intensitypfact r,:ks in, .,* * ... , ..

K fracture toughness, ksnn. -, .. .. ",:

Pall = allowable RCS pressure, psia.

Isothermal and transient conditions are analyzed at the selected ART values. Cooldown transients are analyzed at rates of 10 F/nr, 'u oF/hr,.30oFrtirI 40cF/hr, '5'0'/l, 7.5 F/lir and 1006F/Fr, h regim ng at a bulk coolanht

..temperature of550tan te g at-70'F. Heatup transients are analyzed at rAtes of 10.F7fi, 207F/hr, r F .r bulk temperature of 70F an terminating at 550F....

Hydrostatic limits are obtained only'for isothermal, conditions,-

.. .- ........... ¢.--. , ":. " ,: ': .', ,.;

5.4 CRACK TIP PRESSURE-TEMPERATURE LIMITS Crack tip pressure-temperature limits are developed bgy first conducting-a'heat trarisfer analysis of' limiting' heatup.4 qnd c0ooldo~wn transients.-.Results are ex.tracted from; the heat transfer analyses, includi ng~through-wa'l therrnal; gradierit.profifes and metal temperatures at the .1A4T and '3/4T crack tip..locations". :Thermalstress intensity factors, KiT, for the heatup.dind th* cooldown transiehts dre thbn computed from the through-wall thermal gradients. JiNext, the allowable fficture toughmess is comp'ufed based on the crack tip, metal.' "

temperatures. Finally, the limiting allowable pressures for transient and steady-state conditions are computed iinrgl ttlie '6b6ye equations.' 'The mnost limiting allowable presure 'en eoped'for eachstartup and cooldown transient is evaluated considering that during heatup, the thermal bending stress lis compressive at the reactor vessel inside wall and tensile at the reactor vessel outside wall. Internal pressure creates a tensile stress at the inside wall and outside wall locations with the outside wall location having*the larger total stress. However, neutron embrittlement, shift in material RTNDT, and redufction in fracture. toughness are greater at the inside location than the outside. Therefore, results from both the inside and 6utsideI flaw locations must be compared to ensure that the most iimiting condition is recognized.

During cooldown, membrane and thermal bending stresses act together in tension at the reactor vessel inside wall. This results in the pressure stress intensity factor, KIM, and the thermal stress intensity factor, KIT, acting in unison to amplify the stress intensity. Tensile pressure stress andcompressive t.hermal stress act in opposition at the reactor vessel outside wall, resulting in a lower total stress at the outside as compared to the inside wall location. 'Also, the shift in RT-., the reda*ztion in: fracture toughness, and neutron embrittlement are less'severe ai the'outside wall than at tIhe inside wdallidoation. Consequently, the inside.flaw location is more limiting for the c6oldown event. .- . . : . ..

'WCAP-16835-NP, Rev 0 ". .. -Page"5-3 June 2008

Although allowable pressures for some transients are limited by conditions, calculating the crack tip P-T limits by the above method, i.e., enveloping the isothermal and all events at 1/4T 'arid 3/4T locations and includingthe heatup or'cooldown rates below the value under consideration, is more conservative thaii the:;

ASME Appendix G method. .

5.5 HYDROSTATIC AND LEAK TEST. PRESSURE-TEMPERATURE LIMITS The purpose of the hydrostatic test limit is to establish the miinimum femperature required at the correspoinding hydrostatic test pressure. The in-service hydrostatic test for CE NSSS designs is based on a test pressure corresponding to 1.1 times the operating pressur.e,,with.the reactor core nct~critical.

Pressure-temperature limits for the hydrostatic and leak tests are determined using KIM values due to the applied pressure loading at isothermal conditions per ASME .Code procedure. A safety, factor of-1.5. is applied to KIM along with the condition that 1.5*KIM must be less than or equal to Kic when establishing the maximum allowable P-T limits for these tests.

A gradual reactor coolant systemitemipeeattire 'chatig6 of t0`Fin d .nY '1-_hoiir ýeriddis assumed to induce negligible thermal stresses. Therefore, a change in RCStelYperature'of +/-10°F in any 1-hourperiod is the maximum permitted during inservice hydrostatic and leak tes-;,ing. . ,: -

5.6 MINIMUM BOLTUP TEMPERATURE The minimum reactor vessel flange bojtu.p Xemperatuaxre is defined asthe initial,,RT ,.tetqnperature for the limiting material.in the stressed,(iange) region, plus 4nyefts Qofirdiation. Since there'is.no meaningfuil irradiation effect in the flnge e n, the mimmum .oltup temperature fr P.)NGS is set as the inital RihDr of 60'F plus the instrumient uncertainty of 13 .2' ,or 73..F. For prac-tiality and 6'nserv'atsma value of 80 Fis set as the minimum boltup temperature for PVNGS.' ThiS hinimtiiiV bol tup tempeeate istappic'abie for pressures less than 20% of the pre-service hydrostatic pressure.

5.7 MINIMUM PRESSURE REQU.!IR NT,.

F*, ,, .. t.,.

I.

Appendix G Of 1' CFR 50 Speci-fieS fracture {ioughiress f.0-quireeients*;frfefritid inatetials8us&1 iai:r~actoriCý01ant system pregsurem-retaining components ill orddr:toaiisare th-e; pressure boundary integrity over:its iservice-lifetime.

The minimum pressure as defined.in 1.0 CER 5&Appendix. G is 20%t 6ftho prerservic&hydrostatic..test pressure.

For PVNGS, this minimum uncorrected design.pre~ssur .is20% x (2500 :psia x t%.25) or .625 psia. ... ;

When corrected for pressure elevation. effects and,.tempeýature meas rýemei uncertainty* the minimum pressure requirement at PVNQS becomes: .. ' , * :. . - ,.

For T < 2000F, P < 750 psia, . ..

20% Pprehydr= (625. 111) =514 psia . , ... , .

For T > 200 0F, P < 750 psia, .- .... . .

TRCS 200 + 13.2 =213.20 F

'-;20% Ppre'.hydr6 (625 --128).= 497 psia ,. .,.. ....

5.8 LOWEST SERVICETEMPERATuRE :"

The lowest service temperatures for piping, pumps, and valves with-material thickness greater than .2.5 inches is specified in ASME Section II! Article NB-321 1 as equal to the.highest RTND.pluss100F, Reference 26. For PVNGS, the highest RTNDT for piping, pumps, and'valves' is 40'F,.Reference 27. With instrument uncertainty included, the lowest service temperature becomes (40'F + 100°F + 13.2°F =) 153.2°F.

WCAP-l6835-NPRev"O............ 5 June 2008

5.9 FLANGE LIMITS Minimum required temperatures for the allowable P-T limits depend on'the h1ighest RTDT ý6f the closuie flange region in the highly stressed regions. For PVNGS Units 1, 2 and 3, the highest flange region RTNDT value is 60 0 F. This value is considered in generating the minimum temperature limits fodrthe hydrostatic test, and for normal operation with and without the core critical.

5.10 MINIMUM TEMPERATURE REQUIR"MENTS Pressure-temperature limits for the reactor pressure .vessel fIange for in-service hydrostatic and leak tests, heatup and cooldown transients with the core not critical are directly e'valuated using 10 CFR Part 50 Appendix G Criteria la or lb (hydro and leak tests) and Criteria 2a or 2b (normal operation including heatup and cooldown). These criteria establish that when the 'RCSipressure 'siess-than'or equal to 20%' of the pre-seivice hydrostatic test pressure; the minimum reactor Vessel rAq.e.must:be at least as high as the RTNDT for the yae.

limitingmaterial in. the closure flange region stressed by bolt preload. ,W.Ven tteRCS pressure is .greater,than 20% of the:pre-servi.ce hydrostatictte.st pres sure.j--1eIminimum reactgr;yessel temperature must be at least as.

high as the RT1I.- for the limiting materjal in thle ec sure flangeregip iqstresses byvbolt preload.plus..90'F for

- hydro or leak. tes~tiing, or plus 1600 for-norma!, operati'n, .nc~uding heatup.and cooldown.....

Hydrostatic and Leak Tests (.ARepnix G., Crteria 'a-Atd_I ,

Minimum temperature ii. te vesser and -the core not" critical are given by:

Tmin ? RTNDT-flange (P <20% Ppr-e-hydro), and , " , ........ '

Ž RTNDT-flange + 90 0F. ,.(P> 20o% Pre-hyd.),.

Tmin where:

Tmin is the minimum temperature limit required for the flange, RTNDT-flange is the highest reference lemp'eratu rleocffihe meaterial inrthelohsure flange'iegion', and, Ppre-hydro is the pre-service hydrostftic test pressure: . . .

Applying these minimum temperature limits to PVNGS Units 1, 2 and 3 results in:

-"Crii 6ial'1.a':Pressure:< 20%.Pre-Service Hydrostatic Test Limit: " .

Tmin-Hydrd.- '='RTNb-i'-_jange ,+ AT.uncitainty ,..

= 60°F + 13.2 F = 73.2 F 0 0 (A boltup temperature of 80'F is used.)

Criteria lb: Pressure > 20% Pre-Service Hydrostatic Test Limit Trin-Hydro RTNDT-flange0 + 0 F + ATuncertainty

,::...*-`600P 90 F,+13.2OF 163.2'F' Normal-Op'eration -*CorieNot'Critical (Appeadi.xtG,-Criteria 2a and 2b)

For normal operation, including heatipi and cooldown, the minimum temperature requirements are evaluated using the criteria in 10 CFR 50 Appendix G. For the core not critical condition, the minimum temperature requirements are given by:

- Tmin RTNDT fang*'-' : (P-<20%0P*r*-hydro), and . ... -

TM *in> RTNDT-flange +I 20EF.' (P".' '2 0%.'Ppre-hydr6r)."

WCAP-16835-NP, Rev 0 Page'5-5 June 2008 . I;

Applying these limits to PVNGS Units 1, 2 and 3 produces:

Criteria 2a: Pressure,<20% Pre-Service Hydrostatic Test Limit rTnirinNoP -- RTNDT-flýnge + ATuncertainty

=.60'F:+ 13.2°F = 73.2°F (A b oltup temperature of 80'F is used.)

Criteria 2b: Pressure > 20% Pre-Service Hydrostatic Test Limit Tmin-NoP = RTNDT-flange + 120 0F +..AT*¢celainty

= 60OF + 120°F + 13'.2 F = 193'.20F These minimum temlperatures are incoip'orated iinto the P-T limits for the hydrostatic test.

J.J.,

5.11 TEMPERATTJURE RQUIREMENTS FOR NORMAL OPERATION Minimumirinliperaeire criteria foi. cri'ecriiia1i edIticn eSblished by 0 CFR 50 Appendix G Criteria 2ciand 2d, ýpe:ifyý-th6 following P--T -linits. in th'. catVh thcn th .I&RCS pressure is; less thanf or equal to 20% of the pre-service hydrostatic-test pressure, the nuiinilmnmieatori vessel tempeiatuie must'be :at least'as high as the RTlDv for.thehlimiting material in the closureflange recin stes~.d b pblfelobad plis 4W0F, 6r the minimumf' permissible temperatuie for ith6 in-servicbe hydrostati :pirssu ie test, Whichev6er.is larger. When theRCS, pressure is greater than 20% of the pre-service hydrostatic test pressure, the minimum reactor vessel temperature must be at least as high as the RTNDT for the limiting-.rnatrial-iii y 0 a1ius pi(,suflange-r:giinSS 0

160 F, or the minimumpermiissible-tfrnperaVture for,the: in-service-hydrostatic, pe~ssre testiwhichever is:larger.

When the core is critical, minimum temperatures required for PVNGS Units 1, 2 and 3 aie given by:

Criteria 2c: Pressure < 20% Pre-Service Hydrostatic Test'Limiit.

Tnif-NoP = Larger of ((WHydro) or (RTNDi-IIgII + 40OF)) + T - -

= Larger of ((168.2°F) or (60 0F.+ 40'F)) + 13.2'F = 181.4 0 F Criteria,2d: Pressure,> 20% Pre-Se1yijce Hydrostatic Test Limit

- .÷ .... 1.°.

.- AT.. .......

Tmin-NoP = Larger of ((THydro) or (RT+T,-nnh ...... AT

= Larger of ((168.2°F) or (60'F + 160'F)) + 13.2 0F 233.2 0 F The minimum in-service hydrostatic test temperature (THydro, uncorrected) is 168.2 0 F and corresponds to a pressure of 2,475 psi (uncorrected) from Table:5-1. These minimam temperatures are-incorpcrated-into theP-T limits for the heatup transients in Figure 5-1, and for cooldown transients in Figure 5-2. -,

5.12 LTOP ENABLE TEMPERATURE LIMITS LTOP enable temperatures are implemented at PVNGS to protect against brittle fracture during. reactor start-up and shutdown operations due to low temperature oveipressure events-for Se. vice Level A 'or B conditions.

Computed LTOP heatup and cooldown enable temperatures are shown in Table 5-3. Note that the shutdown cooling system suction line ("LTOP"') relief valves canxemiain.,in service above.221?F during heat-upor-.

cooldown until the RCS pressure reaches the maximum SCS operating pressure (cf., TS LCO 3.4.13).

Heatup Transients . "

ASME Code criteria requires that LTOP systems must be effective at coolant temperatures less than 2000 F, or at reactor coolant inlet temperatures corresponding to a reactor vessel metal temperature, at the 1/4T crack tip location of less than (RTNDT+5 0 °F), whichever is greater. For heatup transients, the LTOP enable temperature is based on the l/4T crack tip metal temperature limit of 166°F, (RTNDT = 1 16°F) which corresponds to a fluid temperature, obtained from heat transfer analysis, of 207.3°F. The temperature lag at the 1/4T crack tip is a

, WdCAP-6835IP, Re2008 . ...: ' ... '4Pag&5-6

",:5 June 2008

function of the heatup rate,"iessel dimensions, and the heat transfer propertiesu inthi.. Te resulting LTOP heatup enable~temperature including uncertainty is 220.59F (= 207.3'F : I 3.2°F).'. Per ASME Code criteria, theLTOP enable temperature for heatup events is set at the maximum of:(200'F; 220.5°F),or,:..

220.5'F, which for operational purposes is rounded to 22 IF.

Cooldown Transients. ,. .:,. . . ..

The 'PVNGS LTOP cooldovwn enable' temperature is conserVativeiy set equal to t6'rikb: fihe isothermal case, and is expressed as the maximum of 200'F Or RTNDT'+'500-;F.' The vessel metal tenp 'ratiiie'is faken:at.a*`

distance one-fourth of the vessel thickness from the inside surface in the vessel beltline region. RTNDT is the highest adjusted reference teinperature for the weld oi bas rrietl'lii.tlhei bltlirie region at ,14T;.measured from the vessel inner surface. This, is determined usin'g thfe-'redu!res-of Regulatory Guide 1.99, Revision 2.

Instrumentation correction needs to be considered to.aMve at the final LTOP enable temperature. For an

.isothermal condition at PVNGS Units. f.2 and 3_,includm g iunkTi iiiipmenftuncertainty, the LTQP enable Stemoerature becomes RNOT -I 50F. +;urfeetainty',-= ( l ,F*R-- 5°F + 13.2'F) 179.2°F.-As 'this calculated.,

temperature is below the ASMECode requiremenit of2000 F,, the LUTOF enable temperature for cooldown .

-transients, including-uncertainties, becomcs. (-200 9-F +-j13:2-,'F) =2 1-3:2 9ýF: -.- ..........

M3-Combined Ehabled Tempeirature

, _Fo~pratialiY~a i -- ',, Limit. *,** . ,-mp rat r.rvau ,of,22l.F "

-For pacticaliWand conservatism; asirigl6 terriperitue~vaie f*:2...F is-selected as-the hea'tup and c6oldown:

LTOP efinblktedipefrtufeýfor PVNGSUfiitsT1` 2-a s"shown in Table 5-3.-- -

5.13:- PRESSIURE AND .TEMPERAT-URE CORRE*CTON FACTORS..

All. RS pressur-te-mperature limits calcuilated fborl h*ep c6old6owr and test conditioi& are; correced. for 0

hyd;:ulici conditions-and instrument uncertainy. A .constant temperature instrument uncertainty of 4+13.2 F is Ned alli66rpu td RCS fermperafrures.. Differe.tvgress5ure 60fectionslthifadjust for differences ii elevation apiid.t6 the-pressurizer pressure.-instruienti.ocation and the-raator vessel,.reactor coolant.system flowrate,, and

- .... surizer pressure i ntuncertaiinty are applied in. .iffent regimes of pressure and temperature' 'These nstrum pressuire correction-factors-vary from-1-1I psi' to 169 psi, Reference -27, and apply as shown below.-

, " i-RCS ... i .. :RCSRC PrPressure Pressure* "Pes*e ... ) mp-erature .. .

... Temperature- Coreion .crrecti'on'

...... i. ..

....... .. ..! ' :7... < 750 Oigia .. .:. ~ s :.. .. + 13.2°F ..

.00

{ 750 psia

<7 50 p si

-15&3psi.

.l 2 8 p si . .. . .

+13.2 0F I 3 .2 °F .

+1

' ._>750-psa----- -169 psi . . +1-3.2 0F

The aeppicabletemperature rafng-e is consist"ent!with+the illowabl& RCP usage specified by'Technidal

- Specifica:ions T.hýat is, withiii-the-LTNP*.b1rdige, ipeiiion of two RCPS is permitted below 200'F 'and

-three RC*s at 200'F-and above. Similarly, the applicable pressure range is limited by the range of the pressurizer narrow range instrunmentation .

Indicated'RCS pressure-temperature -limits are generated by-subtracting the pressure correction from the calculated all6wable pressure- then adding the temperature correction to the calculated RCS temperature. This correctionprocess is followedto determine the allowable P-T limits for all heatup and cooldown transients;-and for RCSo&test conditions...'....... . ... ..... . ... ...

WCAP-16835-NP, Rev 0 Page.5-7 June 2008

5.14 _, i

SUMMARY

OF RCS PRESSURE-TEMPERATURE LIMITS Table'5:-i* shownfi plotted in'Figure 5-1, provides PVNGS heattip P-T limits corrected for instrument uncertainty through 32 EFPY':YTable 5-2, shown plotted in Figure 5-2, illustrates tcomparable PT co01down limits... LTOP enable temperatures through 32 EFPY, corrected for uncertainties, are given in Table15-3.

Figures 5-3 and 5-4 show the composite P-T limits, including core critical values, corrected for uncertaintfies.'

Tables 5-4, 5.-5 nd 5- Jit tthe core critical P-T limits, including corrections for instrument uncertaint, for I hydrostatic testhig and for. lreatup (75?F/hr) and cooldown (1900 F /hr) events.

Figures 5-5and, 5-7 shqw. .the thermal, grajieats.-at. I/,4T, and 3/4T for heatup and cooldown events through 32 EFPY.. Corresponding hieatup andcloldo~vnt.lea/ str s intensityfactors are shown in Figures 5-6.and 5-8.

essur-anT empsature eaiLimits through 32 EFPY Tempe~ature Pressure . i -RCS lkressdre(pfia)z@,Heatup Rate , ... . Hydrostatic Isothermal,* ............ -*4O'Fih. i @5' 0 Flhr @75 0F/hr Test (psia)

(psia) r /

310_Fhr"_ _r2_ .. . -*

80 680.6 680.6 680.6 671.1 650.2 622.2 602.2 9544 83.2 690.2 690.2 690.2 676.2 650.2 622.2 602.2 967.2 93.2 727.2 727.2 705.2 676.2 6502:!..:'62-2.2,;, 602.22r`-I. : i016.2 103.2 772.2 . 772.2 7j0.,2 676.2 650.2. 622.2 .602.2. 1075.2 113.2 8262 .. '86.2. 735.2' 68i2. :6502' 622 "602.2. 1148.2 123.2 893.2 893:'2- 718.2 :: 700.2' 5 :-" ý6222:2ý " 662,2" " i237:2 133.2 974.2 974.2 839.2 738.2 672.2 627.2 602.2 1346.2 143.2 1074.2 . 1074.2 9t21%2 Q790.2,7 ' 705.2 :6,15.2.. "(602.2- 1478.2 153.2 1195.2 1195.2 1018.2 862.2 754.2 676.2 604.2 1640.2 163.,2 134:42:

f- 1335.2 1142.2 :- :95"4.2: 192'- -i; '1i72P.2 T J617.2- '1838.2 171.5 .i1494:.8; ,'1467,5: A2§9.5- ;;`"l1049`0 . 889.9 -772.8 "2:638.0 :: ,.0399A

!,- 172.1 1I507.01 . 1478.3 1279%9 L-.. F15:7*0 ,,:-7773'896.7i -. ,598.0W0- 205,3.6 173.2 .- .1525.2;. , 149.4.2 12952. 1068. , 904,2: 783.2 . 600.;2, 2080.2

,183.2 _ 17,7.2 . 1689.2, 1484.2,:; 1213..2 . .,10 ,2 865.2 . ..... .637.2, - _., 2375.2 186.7 1841.7 , 1772.5 .1565.4 . 1275.5 1062.2 . 902.0 655.4 2500.0 193.2 2017.2 927.26.2 391.2 970.2 689'.2 203.2 2347.2- 221-7.2- A1998.2 . 1610.2 11320.2 1-101.2- -757.2 207.0 '500.0'

'ý 2351.5 2'29.0 1713Z:- 1399.2 1162.4 790.6 211.2 ....... - 2500.0 - 2274:2- - -:1827.0-- 1486-.6- -..12300 -827.6 213.2 .. 2343,2 11881.2. ._528.2 1262.2, 845.2 213.2 2327.2 1865.2-.; .15-12.2 1246.2 829.2 217.3 2500.0 1998.9 . 1616.3 1327.8 91874.7 223.2 .... 2191.2 . 1766.2- 1445.t2- 940.2 230.8 2500.0 2008.6 1634.4 -. 1045.8 233.2 2085.2 1694.2 1079.2 243.2 ' -, 2474.2 " -@ Od 22 : ý'1250.2

'243.7- .

L6 2500.6'i :-77201,8.K8' -4',260.8:' ,

253.2*, . f.........__-_,___ .--372: 2 *! "1461.2. f -.

256.0 '. 2500A-0. -1!533.4-1 -

263.2 1719.2 273.2 _.,. , , :, : - " 20342 .

283.2 2418.2 284.9 ' .2500.0 (1) Corrected for instrumeni uncertainty and for RCS pressure and elevation 'effect. j.

'. W CAP-16835-N P, Rev -0'..... .. .. .. .. .. ... .. .. . . . .. ............. ... , , . - . Pague 5-8 June 2008

Table 5-2 RCS Pressure and Temperature Cooldown Limits through 32 EFPY Temperature RCS Pressure (psia) (& Cooldown Rate (I)

(OF) () Isothermal @10F/hr @20°F/hr @30'F/hr @40 °F/hr @5 0 °F/hr @75°F/hr @100°F/hr 80 680.6 612.3 589.0 527.1 469.5 416.6 329.2 237.6 83.2 690.2 '623.2 ,.601.2

. 541.2' 485.2 -433.2 329.2 272.2 90.9 718.6 655:4 638.0 583.4 533.5 492.2 402.8 372.6 91.3 720.1 - 657:2 598.0 .585:7.. - . 536.1 495.4 406.8 378.1 93.2 727.2 665-.2.. 607.2 596.2 - -548.2 -- 510.2 425.2 403.2 99.6 756.1 .. 698.0 644.5 ... .. 638.0-. 507-1 ... 559.7 501.1 493.2 99.9 757.5 .... 699.6. 646.3 .-. 598.0 -599{.4 562.1 504.7 497.5 103.2 772.2 716.2 665.2 619.2 624.2 587.2 543.2 543.2 104.7 780.4 725.6 676. 1.. ,631 "36 38:9- 604.8 565.0 565.0 104.9 781.6 727.0 .677.7 .. 633.'1. . 598U0 607.3 568.2 568.2 107.6 795.8 743.4 696.7 654.2 622.1 638.0 606.3 606.3 107.8 796.8 744.4 698.0 655.6 623.6 598.0 608.7 608.7 109.8 807.8 757.0 712.6 - :671.9 642.1 621.6 638.0 638.0 109.9 808.5 757'9 . 713 .... '6"73:0 6434-A 623.2 598.0 598.0 113.2 826.2.i 778.2 _ 713.737.2 699 .2 673.21_. 661.2 645.2 645.2 123.2 893.2 .. 854-2-2 -823-2 .- 7988.1- - -. - 781.2 -776.2 776.2 776.2 133.2 974.2;;.... 947.2-- .929:2 918.2--9-8:2 .. -918.2

. 918.2 918.2 143.2 1074.2.- --.1060.2 A.057.2 . L-.105.7.2 -... 10'57.2.. 105'7.2 1057.2 1057.2 153.2 1195.2!--........ . .. ,195.2. 1195.2 ._..:.T1-95.2 -1195.2 1195.2 1195.2 163.2 1344.2 1344.2. 134.4.2 . .-..1344.2 '1344.2 1344.2 1344.2 1344.2 173.2 1525.2, 1525.2 525.2 1525.2 1525.2 1525.2 1525.2 1525.2 183.2 1747.2i 1747.2 , 1747.2 1747.2 ' 1747.2 1747.2 1747.2 1747.2 193.2 2017.2 2017.2- 2017.2 2017.2 2017.2 2017.2 2017.2 2017.2 203.2 2347.21 234-32 . -.- 2347.2 . 2347'.2 . -2347.2"2-. 2347.2 2347.2 2347.2 207.1 2500.0. .2500.0 . 250010 .... 2500.0 - 2500.0 - 2500.0 2500.0 2500.0 (1) Corrected for instrfiniient uincirain"tfy ifd for RCS Pres-gue and. e6ldation-effects.'

-Table 5-3.

LTOP Enable Temperature Limits through 32 EFPY gRTNDT . Crack 1Uncorrected Instrument LTOP Enable Case 1/4T M S, Tcoo.ant , Uncertainty Enable Temperature (0' F..... I * °F ," Te.... m ~ ~(OF) a .ure . ., . ... . .. (OF) (OF) (1) (OF) (1)

Heatup @ 75 0 F/hr .116 - 166-. . L' 207.3 ...... 13.2 220.5 Heatup, Code Minimum t t- --

200 132.

13.2 213.2 220.5 Cooldown/Isothermal ... 116. 50 -' 166 . 166. 1312 179.2 Cooldown, Code 132 232 213.2

. .. . ...... ...... ..... 23.2 213.20 Mini m um...

Combined Heatup and eo1dow Lim .rit, includin g uncertainiy (rounded). 221.0 (1) Corrected for instrument uncetairt*...

WCAP-16835-NP, Rev 0 Page 5-9 June 2008

Table 5-4 Allowable In-service Hydrostatic Test P'essfire' RCS Temperature (oF)l) P-Allowable (psia).(.1) 80 0.

?80 514 1.63.2 ". . . 514 163.2 __, 1838.2,'

.2 ,,173.2 ,2.-.,:.. 2080.2 181.4 '2306 (1) Corre~tedftemperatuie and:res'sures.

!.f.ble5-2 . , 25-

"7

- Core,Crtical Limits or Heatup at 751F/h:

RCS Temperature ( 0F) . .P-Allowble (psia) (1)

=J8 ... 4i.......*"::,- . ......:2:! ) ..... 0 !:"5:1

..., 18.4. ............ .'

-0 __

,, , -,j..~ 1.8,1.4

~~~~~~~~

.* '.'" . ... . .... = .. ... .. -,z , :" --..

".... ~~~~ 5'14.0 i .*..... _.-, ___ =

2 -233.2 - 514.0 2 23.2 : 2 689.2.; f 42.0 :; 748.7

- 245.7; 779.2 25+2.1 - . .. 835.7, ,

2660.:. - 7 979. 2 2

'i~:=.'....i , 2;78.9 . ,...._ .. .'"::'i.; 11 . 7613 :::* = ...L - :

291.4 J14237 .

320.3 2306.0 (1) Corrected temperatures and pressures.

Table 5-6

" Core Critical Limits for Cooldown at IOOoF/hi "RCSiTemperaiure (oF)(6) 1 P'Allowable (psi.a),(')

181.4 -. - 2'

.. .. . 181.4 " 2.... 514,0 -

233.2 . - . .514.0 "233.2 2017.2

.. ..235.0 -.-. . 2076.6 246.0 -, " .". 224 ..

242.2 123066.0 (I) Corrected temperatures and pressures.

- .WCAP" 16835-NP;ReV'0 Page 5-10 Juhe 2008

Figure 5-1 RCS Heatup Pressure-Temperature Limits through 32 EFPY 2500 ____ ________

....~.. .... .........

i... ... . ...... ....

. ... ...i ..... ...'..... !.............. ..... .....

..... .... .;1 ..... .....

2000

t~ . 't......

~~~~~

.10 i'° ..........

~1500 U)

.i .... ........... . .:

~20'

[ *~~~~

N____ ........ * ! ........ ...

~~ .... . ; .. ... ?"

40-, - ...........'. . ......[.....

U) -..i..............:.:.

.....i.,.:.. . ) .. : :.....:: * ,-i....... ..... ... .. . ...........

a.

1000/

,/ F/h"r' i50

...i.i...

, :.... - - "" " " ... 750 F/hr : - 7.... ... .. ........ .,,:

........ ..:.... . .i :..... . . .. .

5o 0 .

0 0 100 150 200 250 300 350 400 Indicated RCS Temperature (*F)

WCAP-16835-NP, Rev 0 Pageý5-11 June 2008

Figure 5-2 RCS Cooldown Pressure-Temperature Limits through 32 EFPY 2500 2000 S 1500

.N

'1000 0

500 0

50 1.00 .150 .200 250 300 350 400 Indicated RCS Temperature (*F)

W CAP.1 6835-NP, Rev 0 Page 5-12 June 2008

Figure 5-3 RCS Composite Pressure-Temperature Heatup Limits through 32 EFPY 2500 I I I I II I - I-- -

I i ,

Lowest LI Service -

Flange K __

Lim it 2000-- i i r r l 7

-Temp

- 193.2-F I

i - i .i- T T T CL

-~ I CL 1500 I ,I I rr-r I- r I I Hydrostatic Test III ,." I

- -r U)

,, /,,t T'T--T CL.

-r--7--I- I- --- -i M~ ' I 41 ' J *,J .

I I

~j1/2 I I 1000 1111 I 14 I i i, 1 I L I I l Isothiemal i I, " ' T

-L 1-- L L~

- 1 I I" L4

I I I I 1 i

/ I LTOP':f Enable -

i. i o *

Te m p 2210 500 II I

I I

I I . . . L . . . . I , m , , , I . . . . . ;

I I I I I I I. I I I I I I I I ,

- - I~ - IT - F - -~ - ~I - - -

Minimum

- - r - r - F - Boltup - - II 7 3-,p 'L I

I I.

I I

I

'Temp 80°F - LI - ~- - L -

I I I I I I I I I I I I I I I I - I I I I I I I' 0

-Q .50 100 t.150 200 250 300

I*ndicated RCS Temperature Tc (OF)

'WCAP-16835-NP, Rev 0 Page 5-13 June 2008 'i

Figure 5-4 RCS Composite Pressure-Temperature Cooldown Limits through 32 EFPY 2500 T-p I II r I T I I 7 T -

2000 I'"StH F II Il I Flange I-.-

Limit 7

I

'Ser-ite I I

I I -I I

Temp I

I I Iir Temp ' 1193,.201 S I 1I L L L L , - 1 L 153.2'F C,

-1 J L L'__ _1- A_ L :

S/i

(. 77 I I - I i *I i I ,

= 1500 I I L

I I "

I-- -1=- 4-J- 4 I

2-I "1 i I 1"i I I I

4 -4 7 L 0. FlydrostatidTe Zs N : i I t I I I I I I I

-i- 4-I---

CD I- -i - I - I - ý1,- I- 1 7 LTOP 1000 II Enable II Temp a) 221 OF 001, I I II i

I TI T

r F

7! Is othermal T F ~'1'

... 500 - I I Ih 34O 0Fhr

.~A I i I I

--- LT T ..

11, 50OF/h r -- -

Minimum 514i pia i? i i T F 1.,Boltup ,: I ,

Te m p: L 1 _-1 0

80 F 'I. . . ."

! I_ I_1_

I I80F i:- i "

0 0 50 00o 200 250 300 Indicated RCS.Temperature Tc (OF)

"WCAP-16835-NP, Revi 0.. ,Pag&5-'14 June 2008

Figure 5-5 RCS Through-wall Thermal Gradients at Crack Tips for Heatup of 751F/hr

.100 90 80 70 a,.

60 I--

50 o

ig 40 40

, Il S 30 20 10 0

0 50 100 150 200 250 300 350 400 450 500 550 RCS Temperature (°F)

WCAP-16835-NP, Rev 0 -Page5,45 June 2008

Figure 5-6 RCS Heatup Thermal Stress Intensity Factors at 3/4T Crack Tip 25 20 C

.15 o

5 0

0 50 100 150 200 250 300 350 400 450 500 550

. .. . . . .RCSTemperature (°F)

WCAPl 683 5-NP" Re6v' Pag*) 5-16 June 2008

Figure 5-7 RCS Through-wall Thermial Gradients at Crack Tips for Cooldown of 100°F/hr 130 120 110 100 90 80 70 0

0 60 c- 50

<0

-40 30 20 10 0

550 500 450 400 350 300 250 200 150 100 50 0 RCS Temperature (*F)

WCAP-16835-NP, Rev 0 Page -17 June 2008

Figure 5-8 RCS Cooldown Thermnal Stress Intensity Factors at 1/4T Crack Tip 45 40 35 30 25 20 C

to 15 10 5

0 1.i-550 500 450 400 350 300 250 200 150 100 `*50 0 RCS Temperature (°F)

WCAP-f6835-NP, R6v 0 *

Page 5-18 June 2008

6.0 MINIMUM TEMPERATURE REQUIREMENTS The minimum temperature requirements specified in Appendix G of 10 CFR 50 are applied to the pressure-temperature curves using the NRC-approved methodologies as described in Section 6.0 of CE NPSD-683-A, Reference 1.

The lowest service temperature is established for PVNGS based on the limiting RTNDT for the reactor coolant pumps. Also, pressure-temperature limits developed for PVNGS use the more conservative of either the lowest service temperature or other minimum temperature requirement for the reactor vessel when the RCS is pressurized to greater than 20% of the pre-service hydrostatic test pressure.

The "minimum pressure criteria" specified in 10 CFR 50 Appendix G serves as a regulatory breakpoint in the development of pressure-temperature limits and is defined as 20% of the pre-service hydrostatic test pressure.

For PVNGS, the pre-service hydrostatic test pressureis defined as 1.25 times the design pressure. When developing these pressure-temperature limits, the minimum pressure establishes the point of transition between the various temperature-only based pressure-temperature limits (for example: minimum boltup and the lowest service temperature or flange limits).

For PVNGS Units 1, 2 and 3, the minimum (uncorrected) pressure of 625 psia is calculated in Section 5.0.

The limiting minimum pressure for PVNGS Units 1, 2 and 3 through 32 EFPY, corrected for instrument uncertainty, elevation and flow, is shown in Table 6-1. Table 6-2 lists the minimum indicated temperature values applied to the pressure-temperature curves of PVNGS Units 1, 2 and 3 through 32 EFPY.

Table 6-1 Minimum Indicated RCS Pressure through 32 EFPY 1 1 TemperatureM ) Minimum Pressure( )

TRcs < 200OF PRCS = 514 psia TRCS ? 200°F PRCS = 497 psia (1) Corrected temperatures and pressures.

Table 6-2 Minimum Indicated RCS Temperature through 32 EFPY Requirement Minimum Temperature(0)

Minimum Boltup Temperature 80°F Minimum Hydrostatic Test Temperature 181.4 0F Lowest Service Temperature 153.2 0F Minimum Flange Limit (Normal Operation) 193.2 0F Minimum Flange Limit (Hydrostatic Test) 163.2 0F (1) Corrected temperatures.

" WCAP-1683 5-NP, Rev 0 ' Page 6-1 June 2008

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This page is intentionally blank..

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-r ;

,'-i '-A IFS.I.:' 3 31 F F F F

...WCAP-16835-NP, Rev 0.... - . ..... P,* : -

June 2008

7.0 APPLICATION OF SURVEILLANCE CAPSULE DATA Post-irradiation surveillance capsule test results for PVNGS units are reported in References 5, 6 and 7. Test results were evaluated with respect to the credibility criteria of Regulatory Guide 1.99, Reference 8. Results of the credibility assessment are:

The surveillance program plates and welds are those-judged to be most likely: ontrolling with

.

regard-to radiation-induced embrittlerent, ... - .. .

.. Charpy data scatter does not, cause ambiguity '.in the deternination of the 304"ftl: shift, Measured RTN'DT shifts are consistent with the, predicted: shifts, ..

o Capsule.irradiation temperature matches. that .ftevesel .aflanU..

- - *..-Correlation monitor data falls within the scatter band. fortijt't.material-and therefore meets the

. .. .est.

cr dibiity

.....: ... . .. .. .................................................................. , , ?

credibility test. . . ......

Two or-nore credible data sets are-available for e*hbof the6PVNGS',njt*'.' Those surveillance data' sets are used to assess the chemiiistry factor arid the-marn iirinerni iii-4 c6-dafic ..... fh-tli' fiethodo0ogy prescribed in Position

2. . of Regulatory Guide 1.99. Calculated ch*-histry factor values.for the surveillance plate andweld mateiials

.are shoxvn in Tables.7-1,..7-2 and 7-3 for PVNGS Units.l,2 and. 3,respectively. Thecorresponding credibility test for the surveillance

.I capsul ee_asiirems ntseand the n ea ur'd shift for the correlation monitor material are s hown in Tables 7-4, 7-5 and 7-6. . . . . .

The derived chemistry factor for the Unit 1 surveillance plate (M-6701-2) antd surveillance weld'(Heat 90071)

Sare 27.5SF and 4;9.F*,.respectively*:aszshown 4n Table 7-1-. These chemistry, factor values compare conservatively with their respective values detern-ined following Position 1.1 of Regulatory Guide 1.99, chemistry factors of 37°F and 27.8°F. Table 7-4 demonstrates that the surveillance plate and weld measurements are predictable given that the difference between the measured and predicted shift is less than one standard deviation for the predicted shift (17°F for plates and 2,8.F for welds). Similar conservative results for Unit 2 plate (F-773-1) and surveillance weld (Hqat '3P73,17) are given inTables.:7-2 and. 7.-5,'and for Unit 3 plate (F-6411-2) and: surveillance-weld (Heat.4P7869)-in Tables,7-3 and:7-6.. Therefore, surveillance results-are shown

.tobepredictable and'credible for each of.the PVNGS umits. .

Thecorrelation mrinitor materials fromeach of.theePVNQN unitg meet the,,dredibility. test to be'within the scatte~iband of the. database.for that"maiterial...This is demonstrdtbd ii{ Tables 7.4,.7-5 and 7-6 in which the shift

-meas.remerts available toý date are co'mpared to pi*dictions based on :hicherhistry factor determined following Positiorr.1- of Regulatory Guide. 1.99. 'The-difference-between the measured and predicted shift is less than 19?F-for'all seven' measurements, andl-the averageyvalue of these diflfiences is 7.7,F........

6Th-e calculafion of adjii'iedrieference teieaimeiýature' ARTfdr use in determiining pressure-temperature liimits is

described in Section. 4. ,The most limiting (highest), valueof ART.from the three PVNGS units is applied to all three units. Even though the information presented in this section demonstrates that the post-irradiation" surveillance capsule test results for PVNGS units are credible, the calculation of ART takes no credit for those

.credible results. .Tis is conservative given that:; ..

The most limiting (highest) value of ART from the three PVNGS units is applied to all three units, and

The derived chemistry factors from the credible surveillance data are all conservative relative to the chemistry factors determined following Position 1.1 of Regulatory Guide 1.99.

Supplemental surveillance data are not used in the calculations of the chemistry factor. Therefore, the issue of whether or not the copper and nickel content of the surveillance weld differs from that of the vessel weld, and how such differences are applied to adjusted values of ART, is not applicable.

WCAP-16835-NP, Rev 0 ; ' "'Page 7-1 June 2008

Table 7-1

Ch-emistryFactors for Unit-'1 Surveillance Plates and-Weld Materials, (3)D (ff) 3 Material-,-: ...Capsule. CCapsule f,' ARTNDT ff*ART.T Plate M-6701-2 137 3.65E-+8,tj. 9.7216, 34.2°F 24.678 . 0.5207 Longitudinal 230 _-',0.9629 A86E+/-18 - ... 15.3 0F 14.734 - 0.9271

- , 1... 3.65E-.8:.l8 ..... A0.7216., 13°F 9.386 0.5207

'Plate M-6701-2 230 8.76E+18 0.9629 31.9 0F 30.720, . 0.9271 Transverse Sum:. 79.518 2.8956 CFM_6 7,ol 2 -2 R (ff)2 = (79.518 + 2.8956) = 27.5°F 137 3'.651 +18 ";0.7216l: 0OF('4 ) " '0 0.5207' SWeld08697. 6.7`F' 5:8299 0.7563 (Heat 9007 1)

. 230 - ;76E+18. *,0.9629' 5F;* - 4913 0.9271 2 Sum: 10.742 2.2041 CF RTNr.

N. -:.2(f) E f=:(ff*-(10742-2041) 4.9°F,

() f fluence (n/cm 2, E >281.0 MeV).

(2) ff = fluence factor = ( - 0.1 og f)

(3) ARTNDT values are the measured 30 ft-ibshift yvaues .. -

Actual value for the_weld;='.- 2.87F.,.

- " : Chemistry Fctors f'orUxiit2 Surveillance Pliates ad-WdiMateials :: .

Material Capsule Capsule el) "ff<hW .'. A'RTJ2NDT.: ;A" .... . t)....

0 Plate F-773-1 137 3.87E+18 0.7372 13.3 F 9.804 0.5434 Longitudinal .230 9.92E+18, 0;9978, 17.7F . 17.660 .... "0.9955r<-¢ Plate , 137, 3.87E+18 0.7372 . 5°F 7.003, 0.5434, F-7731 , -230 -9.92E+18 0.9978 19.3°F 19.257 0.9955 Transverse . -. _.. Sum:-' 53.724. 0 . 3.078"

.__ran__s_*_ .... CFG_77311 = Z(ff* RTN'DT) 53 0778) = 17.5 F 137 3.87E+18 : 0.7372 0oF " 0 ' '0.5434 Weld (Heat 3P7317), 230 9.92E+18 + 0.9978 2.5 OF 2.494 0.9955 a 7)... . . .: , .. . .Sum! 2.4941 , ;- 15389.-

CFwed_.*RTi_,_- - (2.494- 1.5389) 1.60 F, -

Sfý=fluen6 (n/cm 2; E > 21.0 8 MeV)....*'... .

(2) ff=fluence factor- f(0" -0Olo*ygf) . . , . ." ,i ..- . "

(3) ARTNDT values are the measured 30 ft-lb shift values.

.. . . . *

i *.. .

WCAP-1683 5-NP,;Rev 0. Piige' 7-2 June 2008

Table 7-3 Chemistry Factors for Unit 3 Sur1eillance Plates-and Weld Materials Material Capsule Capsule f(l). ff( 2) ARTNDT3). ff*ARTNDT (ff)2 PlaceF-641l-2 "

---L230 9.07E+18 0.9726 . 6.3.F -. ..128. 0.9460ý

Longitudinal____________j_______

0 14230 3.48E+18 0.7090 .13.1 7F . ..

8. 0.5026 Plate 230 9.07E+18 -.. 0.9726-",. _9.2F1'. . 8.948.--, 0.9460

F-6411-2 ;, Sum:. -'24.363 ' -: 23947

"... Transverse.FF41- = ff .- 2 - , ;

Transverse...- CFi 1-2 RT64 --- (24363-+2.3947)' 10.2 0F "

-- 142(4) 3148E+.18 --.. .. 0,7090- ..'... ----27.5 F ,: .....:19.497 0.5026 .

.... Weld (Heat 4P7869) 230 9.07E+18 0.9726 24. 1 '23.440 - 0.9460

.Sum: 2.93 1.4487

- . - . CFwe~d -='(fff*RTNbt)W-- Z(ff 2 =-(4?.93÷-1-l.4487)= 29.6°F (0 f=fluence (n/cm 2; E> 1.0 MeV). ... .. .. ... ...

(2) (2ff = fluence~factor = & ý.(:8 o.

11ýog1

- 0. *ogf 1.. - - -..L..T..........

(3) ARTNDT values are the measured 30 ft-lb shift values.

(4) Irradiated at 137 degree position in Unit 3 vessel.

o S -Y 4rui

....... ......- Crdibilit or Siiiyeiifa.nce Measuirements for Unit 1------------

2 Predicted Shift. '...-Difference.

"Maeial.i. .Caisule CF*,~" * :ff2 Measured_ -..

137 C .. Shift (-F) , (OF- (OF),j

.. 3P - .4.2 0.7216< -I.34.2

___5 - 19.8 -

-Lorigitidinal 230 '%27.5 0.9629 15f3'_"' 26.5 d.1.:

'- M.7E, - I.'27.5 .0.96290 , 31.. .. 26.5 +.5.4 Tralls~erse 23 137 4.9 0.72,16, 283.5 63' (Heat 90071) ,4.9 ... 0.8697,;. 67 6.38... , . , 43. . +23.

0.9629 5.1 4.7 + 0.4

...C6freiati68 .... j3'230 .... 4.90 i31.7 ..... 0.7/2'16 .3-----------------------........

10....f1" -¢'.

9.P 5.0 +6 6.3 Monitor071610.C~d._

Monitor 38 131.7(') 0.8697 114.1 114.5 Material -0.4 230 131.7(') 0.9629 129.2 126.8 +2.4

(') Chemistry factor based on 0.174 Cu and 0.665 Ni using Table 2 of Regulatory Guide 1.99, R02.

28 (2) f = fluence factor = f(0. - 0.1 *og f)

WCAP-16835-NP, Rev 0 S-' Page.7-3 June 2008

Table 7-5 Credibility of Surveillance Measurements for Unit 2 Materl Capsule ,CF. fMeasured* *Shift.,(OF) <Predicted

-*(CF;*ff), (Shift ).,(OF)......Difference Plate F-773-1 137 17.5 0.7372 13.3 12.9.. 4 0.4 Longitudinal - 230 15 609978 17.5 +0.2 Plate F-773-1 ..13.7 17.5. 0.7372 - 9.5 , . 12.9ý -3.4 TVansverse . -230 7.5 . 09978" 1903 17.5 + 1.8 Weld 137 1- ; 7-372 _ O 0 " 1.2 -1.2 (Heat 3P7317) '1 230 . 6.... ýV-.*98, . 5 1

!.6 ' +0.9 Correlation " :137 13 1. 772--. 16.0 97.1 + 18.9

, -Monitor- .. -. -_:. . -_- _ -

131.4

-Material ;- ... 13!!7}1 ?..99978,.. 14 - + 1.0

(') Chemistry Factor based on 0:174-Cu anid G.6655N using ubie2 of Regulatory Guide 1.99; R02. "... . .

(2) if= fluence factor = f(028 -0.1 *log f)

Table 7-6 Credibility of Surveillance.Measurements for Unit 3 CF 'MeA.si irId Shift Ptedicted Shift Material -. apsule CF- .O.......f DiffeFen Plate F-54.11-2.

....- L6nitdina- - 230....

-. 10.2. 0.9726: -6.3 .. ..... 9.9-i 6...

- Plate F-6411-2 142- - -10. 2 . 0.7090- 3.1 - .. 7---

7 2' .+9/

Transyerse 230 : --. 10.2 0.9726.,. : 9.2f, . ." 9.9 0.7.

.-Weld 142 -29.6 0.7090------- 27-.5------.. - 22.1.0- - 6.5- -

(Heat 4P7869).. 230 29.6 0.9726' . 241 - .. .. .28.8 - '4.7T.

Correlation I 142 -1 131.7() 0.70901:. 82.5 93.4 0.9....9

M o nito r . . . . . . .. . . . . . . . . . .. . . .. .

Material [ 23.0- , '..0.9726 131.7. .14.' '128 - + 17

.. (2) Chemistry-Factor based.on 0.17.4 Cu.and 0.665 Ni using Table 2 o f egulator Guide 1.99; R02., -

(2) ff= fluence factor = *O.28 -.0.*log f) "

". 'WCAP-1 6g35-iNP',Rev, 9 4 Pa'e.

June 2008

8.0 REFERENCES

1. CE NPSD-683-A, Rev 06, "Development of a RCS Pressure and Temperature Limrits Repiort for the Removal of P-T Limits and LTOP Requirements from the Technfical.Specifications,'."April 2001:'

2.: NRC Generi*c Letter 96-03, "Relocation of Pressure-Temperature Limit Cuives anid Low Temperature Overpressure Protection System Limits,"' January 31, 1996.

3. U.S. Nuclear Regulatory Commission Regulatory Guide 1..10, "Caculationaand Do1.simetry Methods for Determining Pressure Vessel Neutron Fluence,"' Mah2ch61.di- .. ' "

4. H.D. Lemmel, P.K. McLaughlin, V.G. Pronyaev, "ENDF/B-VI Release 7, The U.S. Evaluated Nuclear

Data Library for Neutron Reaction Data by theU.S National'Niuclear Data Center - 1990' including revisions up to April 2000," International Atoimiiienergy' Kgenc'y'report IAEAi:-NDS' 00,; ke. 10, June 2000.'" '; "

5. WCAP-16374-NP, "An ayysis of CapsuIe 2ýOc fr"6r'Arizoina Pu"'ic Service Company Paio Vefde Unit 1 Reactor Vessel Radiation S eif1`fice Pitgrdi'aFtebruary 2005. *

6. WCAP-16524-NP, "Analysis of Capsule 230' from Arizona Public Service Company Palo Verde Unit 2 Reactor Vessel Radiation'Survielanice Pr6gram, Feoruary 2006.ý........ "

7. WCAP-16449-NP, "Analysis of Capsule 2300 from Arizona Public Service Company Palo Verde Unit 3 Reactor Vessel Radiation Surveillance Program," August 2005.

8. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear

.Reguilatory-Comnrission, May 1998) .. ....

9. Combustion Engineering, Reactor Vessel as-Built Drawings E-78173-16 1-00 1, R03 (Unit 1),

E91 i001 02 U 'nit2) d E-65*J34f26 (nit 3).: ...

10. RSICC Computer Code Collection CCC-650, DOORS 3.2, "One, Two- and Three-Dimensional Discrete

" Odinlates'Neutrýonihoton TransportCodeSystem, 1998.

'April --

11. RSIC Data Library Collection DLC- 185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

12.. A. Schmittroth, FERRET Data Analysis Code, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

13. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium," July 1994.

14. ASTM E 1018-01, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB),"

American Society for Testing and Materials.

15. ASTM Standard E 944-02, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)," American Society for Testing and Materials.

16. ASTM E 185-79, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials.

17. TR-V-MCM-012, "Arizona Public Service Company, Palo Verde Unit 1, Evaluation of Baseline Specimens, Reactor Vessel Materials Irradiation Surveillance Program," Combustion Engineering Report dated January 31, 1987.

WCAP-16835-NP, Rev 0 ' Page.8-1 June 2008

18. TR-V-MCM-013, "Arizona Public Service Company, Palo Verde Unit 2, Evaluation of Baseline.

Specimens, Reactor Vessel Materials Irradiation Surveillance Program," Combustion Engineeriing Report dated November 4, 1992.

19. TR-V-MCM-014ý,"Arizona Public Service Company, Palo' Verde Unit 3, Evaluation of Baseline;

,Specimens, Reactqor Vessel-.Materials Irradiation Surveillance Prfgram,".Combustion Engineering Report dated November 6, 1992.

20. Code ofFedeijal , pegulations,.10 CFR 50, Appendix G, "Fracture Toughness Requirements," and Appendix H, "Reactor Vessel MaterialSurveillaince Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.

21. TR-V7MCM-02., "Sumn.ary R epordonManufacture of Test Specimens and Assembly of.Capsules for

Irradiation Surveillance PofPlo VexPde Unit 1 Reactor Vessel Materials," Combustion Engineering Report dated July 14, 1982.

22. TR-V-MCM-004 ."Summary, Repor on Manuf4,ctre, of Test Specimens and Assembly of Capsules for, Irradiation Surveillance of Palo Verde Unit 2 Reactor Vessel Materials." Combustion Engineering Report dated June 30, 1983.

23. TR-V-MCM-010, "Summary Report onallufactureofTest Specimens and Assembly of Capsules for Irradiation Surveillance of Palo Verde Unit 3 Reactor Vessel Materials," Combustion Engineering Report dated November 5'1992 '..i,- , ,, :..: * ,! < : :. . 1,

24. WCAP-15688, Rev 00, "CE-NSSS LTOP Energy Addition Transient Analysis Methodology," May 2001.

25.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix. G, ',Fracture TouglhnessCriteria for Protection against Failure," 2001 Edition with 2003 Addenda.

26. ASME Boiler and Pressure Vessel Code Section, III, Article NB-321 1,and Article'NB2332, ?OOI Edition with 2003 Addenda.

27. Letter, A. Meeden (APS) to J. Olsze*vski (Westinghouse), ."APS Palo VererUnits 1.2 and 3 PTLR Plant Data Request," Letter No. 448-00708 dated October 17, 2007.

3:"' -. " ' ' ;

, , . *i " - . . ' .. ;.I",*'

. ['*: " /'i " *: , :..3 ,*' i" . '

WCAP--16835-NP, ReO . '- ,Pag&8-2 June 2008

WCAP-16835-NP, Revision 0 Westinghouse Non-Proprietary Class 3

-' 3/4 S*

westinghouse Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

ENCLOSURE 1, ATTACHMENT'6' APS Responses to the NRC Request for Additional Information Related to the San Onofre Nuclear Generating Station (SONGS) PTLR Amendment Request The precedent cited in Section 4.2 of the Evaluation of the Proposed TS Change was the PTLR amendment request for the San Onofre Nuclear Generating Station.

(SONGS). Southern California Edison (SCE) submitted a license amendment request to the NRC by letter dated January 28, 2005 (ADAMS Accession No. ML050320286),

and supplemented this request by letter dated January 12, 2006 (ADAMS Accession No. ML060190101), for SONGS Units 2 and 3 Operating License amendments to relocate the RCS P/T limits and LTOP limits from the TSs to a licensee-controlled PTLR. The NRC approved the SONGS Operating License amendments in a letter dated July 13, 2006 (ADAMS Accession No. ML062170006).

In their January 12, 2006, submittal, SCE identified, and provided responses to, nine NRC requests for additional information (RAIs) related to the PTLR amendment request. Provided below are APS responses to the nine RAIs.

NRC RAI No. 1 for SONGS In the staff's safety evaluation (SE) on topical report CE-NPSD-683, Revision 6, dated March 16, 2001, the staff included 26 action items that would need to be addressedin a pressure-temperature(P-T)limits report (PTLR) license amendment request that invoked the methods of the topical report. Your PTLR submittal of January28, 2005, does not specifically identify how the proposed San Onofre Nuclear GeneratingStation, Units 2 and 3 (SONGS 2 and 3) PTLRs resolve the action items in the SE of March 16, 2001.

The staff requests that you supplement your application with your responses to these 26 action items. If your PTLR submittal already includes information that satisfies any of these action items, please specify which information in the PTLR satisfies resolution of a particularaction item. If the PTLR does not include information which satisfies a particularaction item, please provide supplemental information which satisfies resolution of the particularaction item of concern.

The staff recognizes that several of these action items have become obsolete due to updates in the allowable editions and addenda of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section X1, Appendix G, which have been incorporatedby reference in Title 10 of the Code of FederalRegulations, Part 50 (10 CFR Part 50). If such an action item falls under this categoryplease designate it as such.

1

Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment APS Response No. 1 Responses to the 26 action items of the staffs safety evaluation (SE) on topical report CE-NPSD-683, Revision 6, dated March 16, 2001, are provided in Section 3.3 of the Evaluation of the Proposed TS Change.

NRC RAI No. 2 for SONGS The ASME Code, Section Xl, Appendix G provides a methodology for calculating stress intensity factors correspondingto membrane tension (KIM) and thermal stress (KIT) for the postulated axial defect. Calculationsof KIT are based on stress influence coefficients from finite element modeling (FEM) analyses for inside (1/4T) and outside (3/4T) surface flaws. Calculationsof the maximum allowable KIM are based on a closed-form solution to an equation such as 2KIM +KIT < KIC, where KIT has been determined from solutions based on stress influence coefficients, and Kic was determined using the equation representingthe analyticalapproximationto the lower bound fracture toughness curve, Kic (in ksi*sqrt(in.)) = 33.2 + 20.734exp[O.02(T -RTNDT)], where RTNDT is the materialnil-ductility transitionreference temperature and T is the actual temperature of the material.

The Combustion Engineering (CE) nuclearsteam supply system (NSSS) methodology differs from the ASME Code, Section X1, Appendix G methodology in several respects. The CE NSSS methodology for calculating KIT is based on thermal influence coefficients from FEM analyses, as opposed to stress influence coefficients. Furthermore,the CE NSSS methodology for calculating KIM does not involve a closed-form solution based on calculationsof KIT and Kc factors, and instead applies FEM methods for estimating the KIM factors.

Please supplement Section 5.0 of the SONGS 2 and 3 PTLRs with a discussion of the specific methodologies that will be applied in the PTLRs for SONGS 2 and 3 for calculatingstress intensity factors at the 1/4T and 3/4T crack depth locations:

a. Discuss the methodology for calculating the thermal stress intensity factor, KIT.

b. Discuss the methodology for calculating the stress intensity factor correspondingto membrane tension resulting from pressureloading of the reactorvessel, KIM. Please specify whether KIM is determined by obtaining a closed-form solution (as prescribedby the ASME Code, Section Xl, Appendix G) or determined by applying FEM methods (as prescribedby the CE NSSS methodology).

Peryour response to action item 21 in RAI 1, if your methodology applies the CE NSSS method for calculating KIM stress intensity values, then your 2

Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment application will need to include a request for an exemption from the requirements of 10 CFR Part 50, Appendix G for P-T limits. The need for an exemption for calculatingP-T limits using the CE NSSS method is specified in the fourth paragraph(pages 20-21) of Section 2.5.4 and in action item 21 (page

27) of Section 5.0 of the SE on topical report CE-NPSD-683, Revision 6, dated March 16, 2001. The CE Owners Group (CEOG) agreed to this requirement in their final version of topical report CE-NPSD-683, Revision 6. The requirement for the exemption is specified in the "Note" on page 5-15 of the topical report.

APS Response No. 2 Westinghouse methodology for calculating the stress intensity factors corresponding to membrane tension resulting from pressure loading of a CE NSSS reactor vessel, KIM, is based on a two-dimensional finite element model with a unit internal pressure loading.

This model considers a 1/4-thickness crack originating from the inside surface as well as a second model with a 1/4-thickness crack originating from the outside vessel surface. These models were loaded with internal pressure on the cylindrical inside surface. For the case with an inside crack, pressure was also applied to the crack face.

Membrane stress intensity factor influence coefficients (KIM) were then computed for unit internal pressure of 1000 psi. These KIM coefficients are then corrected to represent three-dimensional surface cracks with an aspect ratio of 1-to-6 using the ASME Code ratios for 3D versus 2D flaws. Finally, these corrected values are used to calculate the actual stress intensity factors at any applied internal pressure.

In order to comply with the conditions listed in the safety evaluation for CE NSPD-683-A, a request for exemption from the requirements of 10CFR Part 50, Appendix G is provided as Enclosure 2 of this submittal.

NRC RAI No. 3 for SONGS In support of the NRC staffis review of the P-T limit curves contained in the PTLR submittal, please supplement your application with data for the through-wall thermal gradients (AT) and thermal stress intensities (KIT) for the 1/4T and 3/4T crack depth locations. These data are necessary for the NRC staff to perform independent calculations of P-T limits to verify that the P-T limit curves are at least as conservative as those that would be obtained as a result of applying the methods of 10 CFR Part50, Appendix G, or as modified using the CE NSSS methodology. In addition,if you are requesting to use the CE NSSS methodology for KIM determinations,please submit the plant-specific KIM data to support the staffs review of these calculations.

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Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment APS Response No. 3 PVNGS Units 1, 2, and 3 reactor vessel stress intensity factors (KIT) and through-wall thermal gradients as a function of heatup or cooldown rate are discussed in Section 5.1 and presented in Figures 5-5 through 5-8 of WCAP-16835 (Attachment 5).

The membrane stress intensity factor coefficient (KiM) due to unit (1000 psi) internal pressure loading for the PVNGS reactor vessel geometry with a base metal thickness of 11.2 inches is 25.0 (ksi'/in) for a 1/4T crack depth location and 23.3 (ksi4in) at the 3/4T crack depth location.

NRC RAI No. 4 for SONGS In all cases P-T limit curves must be determined using the most limiting conditions in the reactorvessel. Forheatup and cooldown transientsthe applicationof the PTLR methodology and calculationsof P-T limits must always take into considerationthe different conditions at the 1/4T and 3/4T locations during the thermal transient,and the resulting P-T limit curves must always representthe most limiting of these conditions.

Please supplement Section 5.0 of the SONGS 2 and 3 PTLRs with a discussion of how the P-T limit curves account for the most limiting conditions in the reactorvessel. The discussion should address the following points:

a. Please discuss how the calculation of P-T limit curves for SONGS 2 and 3 addressesheatup and cooldown transients,specifically taking into consideration the different conditions at the 1/4T and 3/4T crack depth locations.

b. Please discuss how the calculationof P-T limit curves for SONGS 2 and 3 addressesthe assessment of the 1/4T location for steady state conditions in addition to the 1/4T and 3/4T locations under heatup and cooldown transientconditions. Please supplement the P-T limit curves for SONGS 2 and 3 with a P-T limit curve representingthe 1/4T location under steady state conditions.

APS Response No. 4 The PTLR methodology of CE NSPD-683-A, Revision 06, first conducts a transient heat transfer analysis for all heatup and cooldown transients. Results from the heat transfer analyses, through-wall thermal gradient profiles, as well as the metal temperatures at the inside (1/4T) and outside (3/4T) crack tip locations are extracted. From the through-wall thermal gradients, the thermal stress intensity factors, KIT, for all heatup and cooldown transients are then computed for all transients. Allowable fracture toughness is then computed using the crack tip metal temperatures. The membrane stress intensity factor, KIM is then computed for a unit pressure loading case. Finally, limiting 4

Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment allowable pressures are computed for all transients as well as steady-state conditions.

This approach is conservative when compared to the ASME Appendix G method as it includes the steady-state condition for all heatup and cooldown transients. Allowable pressures for some of these transients are limited by the steady-state condition at high pressures.

NRC RAI No. 5 for SONGS Table I of 10 CFR Part 50, Appendix G, specifies six different minimum temperature requirements that must be met when generating the pressure-temperature (P-T) limits for U.S. operating pressurized water reactors (PWRs):

a. Those for pressure test conditions with the ReactorCoolant System (RCS) pressure less than or equal to 20% of the reactor'spreservice hydrostatic test pressure.

b. Those for pressure test conditions with the RCS pressure greaterthan 20% of the reactor'spreservice hydrostatictest pressure.

c. Those for normal operating conditions (includingheatups and cooldowns of the reactorand transientoperating conditions) with the RCS pressure less than or equal to 20% of the reactor'spreservice hydrostatic test pressure, at times the reactoris not in the criticaloperating mode.

d. Those for normal operating conditions (including heatups and cooldowns of the reactorand transientoperating conditions) with the RCS pressure greaterthan 20% of the reactor'spreservice hydrostatic test pressure at times the reactoris not in the criticaloperating mode.

e. Those for normal operatingconditions (including heatups and cooldowns of the reactorand transient operatingconditions) with the RCS pressure less than or equal to 20% of the reactor'spreservice hydrostatictest pressure at times the reactoris in the criticaloperating mode.

f. Those for normal operating conditions (including heatups and cooldowns of the reactorand transient operatingconditions) with the RCS pressure greaterthan 20% of the reactor'spreservice hydrostatic test pressure at times the reactoris in the criticaloperatingmode.

Criterion 6 in Attachment I to GL 96-03 states that the above minimum temperature requirementsof 10 CFR Part50, Appendix G shall be incorporated into the P-T limit curves, and PTLRs shall identify minimum temperatureson the P-T limit curves such as the minimum boltup temperature and the hydrotest temperature.

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Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment Section 6.0 of the SONGS 2 and 3 PTLRs, provides a listing and brief discussion of the minimum temperature requirementsthat have been incorporatedinto the P-T limit curves for SONGS 2 and 3. However, the discussion does not adequatelydemonstrate how the P-T limit curves for pressure testing conditions and normal operationswith the core criticaland core not criticalwill be in compliance with the appropriateminimum temperature requirements as given in Table I to Appendix G to 10 CFR Part 50. This information is needed to satisfy action item 23 from staff's safety evaluation (SE) on topical report CE-NPSD-683, Revision 6.

Peryour response to action item 23 in RA/ 1, update Section 6.0 of the PTLRs for SONGS 2 and 3 to provide a discussion on how the P-T limit curves will meet all of the minimum temperature requirementsmandated by Table 1 of 10 CFR Part 50, Appendix G. Include in this discussion the value for the highest reference temperature of the materialin the closure flange region that is highly stressed by the bolt preload and how this value is applied along with minimum permissible hydrostatic test temperature to determine minimum temperature requirements that will be applied to the P-T limit curves for SONGS 2 and 3. This information is necessary to ensure that the SONGS 2 and 3 P-T limit curves will continue to comply with the minimum temperature requirements.of Table I of 10 CFR Part 50, Appendix G, and that the PTLR will conform to the provisions of Criterion 6 in Attachment I to Generic Letter (GL) 96-03.

APS Response No. 5 The pressure-temperature curves for pressure testing conditions and normal operations with the core critical, and core not critical are in compliance with the appropriate minimum temperature requirements as given in Table 1 of Appendix G to 10 CFR Part

50. This is illustrated below along with each of the minimum temperature requirements of (a) through (f).

PVNGS Units 1, 2, and 3 Design pressure = 2500 psia Normal operating pressure = 2250 psia Preservice hydrostatic pressure = 3125 psia Minimum bolt-up temperature = 80°F Flange region RTNDT = 60'F Initial piping, pumps and valves RTNDT = 40°F Adjusted RTNDTat 1/4T after 32 EFPY = 116 0 F Adjusted RTNDT at 3/4T after 32 EFPY = 103 0 F 20% Preservice hydrostatic pressure = 625 psia Preservice hydrostatic pressure with correction for instrument uncertainty

= 20% x preservice hydro pressure + RCS instrument uncertainty

= 625 psia - 111 psi = 514 psia for TRcs < 200'F 6

Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment

= 625 psia - 128 psi = 497 psia for TRCS > 200'F Inservice hydrostatic pressure

= 1.1 x (Operating Pressure) + (pressurizer instrument uncertainty)

= 1.1 x (2250 psia) + 70 psi = 2545 psia Minimum Temperature requirements for Inservice Hydrostatic/Leak Tests a) Minimum temperature for pressures at or below 625 psia per 10 CFR 50 App G, Table 1, item l.a

= highest flange RTNDT + instrument uncertainty

= 60 + 13.2 = 73.2°F (80'F used) b) Minimum temperature for pressures above 625 psia, per 10 CFR 50 App G, Table 1, item 1.b

= highest flange RTNDT + 90'F + instrument uncertainty

= 60 + 90 + 13.2 = 163.2 0 F Minimum Temperature requirements for Normal Operation (core not critical) c) Minimum temperature for pressure at or below 625 psia, per 10 CFR 50 App G, Table 1, item 2.a

= highest flange RTNDT + instrument uncertainty

= 60 + 13.2 = 73.2°F (80'F used) d) Minimum temperature for pressure above 625 psia, per 10 CFR 50 App G, Table 1, item 2.b

= highest flange RTNDT + 120'F + instrument uncertainty

= 60 + 120 + 13.2 = 193.2 0 F Minimum Temperature requirements for Normal Operation (core critical) e) Minimum temperature for pressure at or below 625 psia, per 10 CFR 50 App G, Table 1, item 2.c.

Larger of [(THydro) or (RTNDT-flange + 40 0 F)] + instrument uncertainty

- [(168.2) or (60 + 40)] + 13.2 = 181.4 0F The minimum inservice hydrostatic test temperature (THydro, uncorrected) is 168.2°F and corresponds to an uncorrected pressure of 2475 psi. When corrected for instrument uncertainty, these values correspond to a temperature of 181.4 0F and a pressure of 2322.1 psia per Table 5-1 of WCAP-1 6835 (Attachment 5).

Lowest service temperature per ASME Section +11,Division 1, Article NB-2332

- Initial piping, pumps and valves RTNDT + 100 0 F + instrument uncertainty

-40 + 100 + 13.2 = 153.2 0 F 7

a

Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment Minimum temperature required for normal operation for pressures below 20%

preservice hydro

- 181.4°F as given in Section 5.11 of WCAP-16835.

f) Minimum temperature for pressure above 625 psia, per 10 CFR 50 App G, Table 1, item 2.d:

= Larger of [(THydro) or (RTNDT-flange + 160 0 F)] + instrument uncertainty

= [(168.2) or (60 + 160)] + 13.2 = 233.2 0 F Minimum temperature required normal operation for pressures at or above 20%

preservice hydro

= 233.2°F as given in Section 5.11 of WCAP-1 6835.

These minimum pressure and temperature requirements for hydrostatic test, and heatup and cooldown transients for PVNGS Units 1, 2, and 3 with the core not critical or critical, corrected for instrument uncertainty and RCS pressure and elevation effects are tabulated in Tables 5-1 and 5-2 of WCAP-16835 (Attachment 5).

NRC RAI No. 6 for SONGS Section 5.0 of the PTLRs for SONGS 2 and 3 provides a footnote indicating that pressure and temperaturelimit values are adjusted for instrument uncertainty, and for RCS pressure and elevation effects. Please supplement Section 5.0 of the SONGS 2 and 3 PTLRs with a detailed discussion of how instrument uncertaintiesare treated in the development of the PTLR P-T limit curves for SONGS 2 and 3. Include in this discussion numerical values for the instrument uncertaintiesas well as numerical values for factors that compensate for RCS pressure and elevation effects. Please discuss how these factors are applied in the calculation of the P-T limit curves.

APS Response No. 6 The calculated reactor vessel pressure and temperature limit values shown in Tables 5-1 through 5-6 of WCAP-1 6835 (Attachment 5) are adjusted for instrument uncertainty and for RCS pressure and elevation effects. These instrumentation corrections ensure that the calculated beltline pressure-temperature limits are conservatively interpreted by pressurizer pressure and RCS temperature instrumentation. Section 5.13 of WCAP-16835 describes the application of instrumentation corrections at PVNGS.

The pressure correction factors applied to Tables 5-1 through 5-6 of WCAP-16835 of (Attachment 5) consist of three components:

1. A pressure differential corresponding to the static water head between the pressurizer water level and the reference point. in the reactor vessel (APELEV);

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Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment

2. The flow-induced pressure drop between the reactor vessel downcomer and the surge nozzle in the hot leg (APFLOW; a value that depends on the number of operating RCPs); and

3. The pressurizer pressure instrumentation loop uncertainty (APINSTR).

These components are individually established using conservative assumptions, then summed into the pressure correction factor. Pressure correction factors are subtracted from the analytical values to conservatively reduce the allowable pressure limit. The explicit pressure correction factor values applied depend on the number of operating RCPs and the pressure instrument in service. For Palo Verde, the pressure correction, including instrument uncertainty ranges from -111 psid to -169 psid as shown in Section 5.13 of WCAP-1 6835 (Attachment 5).

The heatup and cooldown data in Tables 5-1 and 5-2 of WCAP-16835 (Attachment 5) are also adjusted for temperature instrumentation uncertainty. For PVNGS, an uncertainty of +13.2 OF is added to all computed temperatures as shown in Section 5.13 of WCAP-1 6835.

NRC RAI No. 7 for SONGS The proposed P-T limit curves included in Section 5.0 of the PTLRs for SONGS 2 and 3 are proposed to be effective through 32 effective full power years of operation (EFPY). The existing P-T limit curves contained in the Technical Specifications (TS) are stated to be effective through 20 EFPY. Confirm whether the changes to the P-T limit curves included in Section 5.0 of the PTLRs for SONGS 2 and 3 reflect only the increasein the EFPY for which the curves will be applied. If there are other factors, such as different parametersor methods, which contribute to the changes to the curves, provide a detailed discussion of these factors and how they affect the PTLR P-T limit curves.

APS Response No. 7 The proposed pressure-temperature limit curves for PVNGS are confirmed to be effective through 32 EFPY. These curves are based on the RTNDT shifts for the projected beltline fluence through 32 EFPY for the most restrictive of Units 1, 2, or 3.

NRC RAI No. 8 for SONGS Criterion 7 of the Table in Attachment I to GL 96-03 specifies that an analysis of the credibility of the surveillance data must be provided in the PTLR.

Regulatory Position 2.1 of Regulatory Guide (RG) 1.99, Revision 2 specifies that when two or more credible surveillance data sets become available from the reactorin question, they may be used to determine the Adjusted Reference Temperature (ART) values. If the procedure of Regulatory Position 2.1 for determining the ART values based on the surveillance data 9

Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment results in a higher value for the ART than that given by using the procedures of Regulatory Position 1.1 of the RG, RG 1.99, Revision 2 specifies that the surveillance data should be used for the ART and chemistry factor determination. If the procedure of Regulatory Position 2.1 results in a lower value for the ART, either may be used.

Please confirm that the credibility analysis of the SONGS 2 surveillance data from Section 7.0 of the SONGS 2 P.TLR demonstrated that the surveillance data sets for SONGS 2 are credible.

Section 7.0 of the SONGS 2 PTLR states that the surveillance data were not used to generate a chemistry factor in accordance with the methodology prescribedin Regulatory Position 2.1 of RG 1.99, Revision 2. Please confirm whether the ART values for the limiting materials were calculated using the procedure of Regulatory Position 1.1 of RG 1.99, Revision 2.

If the procedure of Regulatory Position 1.1 of RG 1.99, Revision 2 was used to calculate the ART values for the limiting materials,please indicate why this is an acceptableprocedure, given the credibility of the surveillancedata.

Please supplement Section 7.0 of the PTLR for SONGS 2 with the following information:

a. Table 7-1 of the SONGS 2 PTLR provides chemistry factors for the two surveillance materialsplate C-6404-2 and weld 9-203. Please indicate how these chemistry factors were derived.

b. There is no explicit calculationin the SONGS 2 PTLR demonstrating that chemistry factor values for the limiting materialsderived from the tables in RG 1.99, Revision 2 would result in limiting ART values that are more conservative than those determined using chemistry factors derived from surveillance data. Peryour response to action item 24 in RAI I please supplement Section 7.0 of the PTLR for SONGS 2 with detailed calculations of the chemistry factors for each of the surveillancematerials based on the calculation methods specified in Regulatory Position 2.1 of RG 1.99, Revision 2.

The calculationsof the chemistry factors for the surveillance materialsfor SONGS 3, provided in Table 7-1 of the SONGS 3 PTLR represent an acceptable format for presenting surveillance material chemistry factor calculations.

APS Response No. 8 An analysis performed in accordance with Position 2.1 of RG 1.99 found that credibility criteria are met for surveillance plate and weld data from capsules withdrawn from 10

Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment PVNGS. Tables 7-4 through 7-6 in WCAP-16835 (Attachment 5) demonstrate the credibility of surveillance measurements for PVNGS Units 1, 2, and 3, respectively. It is also confirmed that ART values for PVNGS reactor vessel beltline materials were calculated using Regulatory Position 1.1 of RG 1.99, Revision 2. This is conservative because analysis based on Regulatory Position 2.1 produce credible results that would yield lower values for predicted adjusted reference teMperatures.

Chemistry factors for the limiting PVNGS surveillance plate and weld materials are shown in Tables 4-2, 4-3 and 4-4 of WCAP-16835 and were derived using Tables 1 and 2 of RG 1.99, Revision 2. These chemistry factors are compared to those derived using Regulatory Position 2.1 of RG 1.99, Revision 2 in the following table:

Position 1.1 Position 2.1 PVNGS Material ID Chemistry Chemistry Factor Factor Unit 1 Plate M-6701-2 37 0 F 27.5 0 F Unit 1 Weld Heat 90071 34.1OF 4.9 0 F Unit 2 Plate F-773-1 20°F 17.5 0 F Unit 2 Weld Heat 3P7317 26.6 0 F 1.6 0F Unit 3 Plate F-6411-2 26 0 F 10.2 0 F Unit 3 Weld Heat 4P7869 34.1OF 29.6 0 F The preceding demonstrates that the chemistry factors derived based on the PVNGS surveillance plate and weld data are less than those derived using regulatory position 1.1 of RG 1.99, Revision 2. Therefore, the calculation of adjusted reference temperature using chemistry factors based on Regulatory Position 1.1 will result in more conservative values than would be obtained using plant-specific chemistry factors based on Regulatory Position 2.1. Predicted ART values at 1/4T and 3/4T for PVNGS are given in tables 4-5 through 4-7 of WCAP-1 6835 (Attachment 5).

NRC RAI No. 9 for SONGS Regulatory Position 2.1 of RG 1.99, Revision 2 states that if there is clear evidence that the copper or nickel content of the surveillance weld differs from that of the vessel weld, the measured values of ART should be adjusted by multiplying them by the ratio of the chemistry factor for the vessel weld to that for the surveillance weld.

Please indicate in the SONGS 2 and 3 PTLRs whether the copper and nickel content of the surveillance weld differs from that of the vessel weld. If so, please supplement Section 7.0 of the PTLRs for SONGS 2 and 3 with detailed calculations for determining the adjustments to the measured values for DRT for the surveillance weld, and indicate whether these adjusted values of DRT were used in the determinationof the chemistry factor for the surveillance weld.

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Attachment 6, APS Responses to NRC RAI for SONGS PTLR Amendment APS Response No. 9 Chemistry and fluence factors for surveillance weld materials and the measured ARTNDT values, obtained per Position 2.1, for the PVNGS surveillance capsule weld materials are shown in-Tables 7-1 through 7-3 of WCAP-1.6835 (Attachment 5). Adjusted values of ARTNDT are not used in the determination of the chemistry factor for the surveillance weld because the copper and nickel content of the surveillance welds do not differ from that of the vessel welds. The bases for the chemical content of the vessel and surveillance plates and welds are provided in the APS response to Generic Letter 92-01.

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ENCLOSURE 2 Application for Exemption from Certain 10 CFR Part 50, Appendix G Requirements when computing Pressure-Temperature Limits for Palo Verde. Nuclear Generating Station (PVNGS) Units 1, 2, and 3 1.0 Introduction 2.0 Exemption Request 3.0 Discussion 3.1 Exemption is Authorized by Law 3.2 Granting this Exemption Will Not Present an Undue Risk to the Public Health and Safety 3.3 Granting this Exemption is Consistent with the Common Defense and Security 3.4 Special Circumstances Support the Issuance of an Exemption 4.0 Precedent 5.0 Conclusion 6.0 References 1

Enclosure 2 Application for Exemption from 10 CFR Part 50, Appendix G 1.0 Introduction Appendix G to 10 CFR 50 establishes fracture toughness requirements to be applied to ferritic reactor coolant pressure boundary materials of light water nuclear power reactors. The purpose of such requirements is to ensure adequate margins of safety exist during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.

The American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME)

Code forms the basis for the requirements promulgated in Appendix G to 10 CFR Part

50. The rules of ASME Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components" are used when developing pressure and temperature limits for the beltline region of the PVNGS reactor vessels. The sections, editions and addenda of the ASME Boiler and Pressure Vessel Code, and any limitations and modifications thereof, which are approved by the staff for use in developing pressure and temperature limits, are specified in 10 CFR 50.55a.

The methodology developed by Combustion Engineering to calculate RCS pressure-temperature curves, heatup and cooldown limits and LTOP requirements is documented in topical report CE NPSD-683-A (Ref. 1). The staff noted in its March 16, 2001 safety evaluation for this report that "[t]he CE NSSS [nuclear steam supply system]

methodology does not invoke the methods in the 1995 edition of Appendix G to the Code for calculating KIM factors, and instead applies FEM [finite element modeling]

methods for estimating the Kim factors for the RPV shell ... the staff has determined that the KIM calculation methods apply FEM modeling that is similar to that used for the determination of the KIT factors [as codified in the ASME Code,Section XI, Appendix G].

The staff has also determined that there is only a slight non-conservative difference between the P/T limits generated from the 1989 edition of Appendix G to the Code and those generated from CE NSSS methodology as documented in Evaluation No. 063-PENG-ER-096, Revision 00. The staff considers that this difference is reasonable and that it will be consistent with the expected improvements in P/T generation methods that have been incorporated into the 1995 edition of Appendix G to the Code."

The staff has advised licensees to specify whether membrane stress intensity factors due to pressure loading, KIM, are determined by obtaining a closed-form solution (per the ASME Code,Section XI, Appendix G) or determined by applying finite element modeling methods (per CE NPSD-683-A, Revision 6). Stress intensity values, KIM, for PVNGS are computed using the CE NSSS finite element modeling methods, therefore, APS requests an exemption from the requirements of Appendix G to 10 CFR Part 50 to apply this model when calculating applicable Units 1, 2, and 3 pressure-temperature curves and LTOP limits.

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Enclosure 2 Application for Exemption from 10 CFR Part 50, Appendix G 2.0 Exemption Request Reactor coolant system pressure-temperature curves and LTOP limits for PVNGS Units 1, 2, and 3 are based on the specific methodology developed by Combustion Engineering and approved by the NRC in CE NPSD-683-A. This methodology employs a finite element analysis model and a crack stress intensity factor, Kic, of ASME Code Case N-640 (Ref. 2). Results produced. by this method are slightly less conservative than the use of KIA stress intensity factors and the linear elastic fracture mechanics methodology promulgated in Appendix G to 10 CFR Part 50. Specifically,Section IV.A.2 of Appendix G to 10 CFR Part 50 establishes the following criteria for generating plant-specific pressure-temperature limits:

The pressure-temperature limits for an operating plant must be at least as conservative as those that would be generated if the methods of analysis and the margins of safety from Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code were applied.

Pursuant to 10 CFR 50.12, APS hereby applies for an exemption from the requirements of the above 10 CFR 50, Appendix G, criterion. This exemption is requested since the specific RCS pressure-temperature limits developed for PVNGS employ a finite element modeling methodology developed by Combustion Engineering and applied to CE NSSS plants for calculating KIM stress intensity values.

APS addresses and satisfies the criteria of 10 CFR 50.12 in this exemption request. As required by 10 CFR 50.12(a)(1), and as more fully discussed below, this exemption is authorized by law, does not present an undue risk to the public health and safety, and is consistent with the common defense and security. Further, in accordance with 10 CFR 50.12(a)(2), the request demonstrates that special circumstances support issuance of the exemption.

3.0 Discussion The reference pressure stress intensity, KIR, used for calculation of RCS pressure and temperature limits at PVNGS is obtained from a reference fracture toughness curve for reactor vessel low alloy ferritic steels and is defined in Appendices A and G of Section XI of the ASME Code. This reference pressure stress intensity is determined by two properties, KIA and KIc that represent critical values of the stress intensity factor. For PVNGS, KIR is defined as Kic, with KIc defined as the lower bound of static initiation critical K, values measured as a function of temperature.

Title 10 CFR 50 Appendix G criteria require that pressure-temperature limit curves be generated from the most conservative combinations of the limiting P/IT data points and the minimum temperature requirements listed in Appendix G to 10 CFR Part 50. The NRC Staff endorsed Appendix G to the ASME Code through the 1995 Edition at the time that the PTLR analysis methodology described in CE NSPD-683-A was approved.

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Enclosure 2 Application for Exemption from 10 CFR Part 50, Appendix G ASME Code Case N-640 permits application of the lower bound static crack initiation critical stress intensity factor equation (i.e., Kic equation) as the basis for establishing the P/T curves in lieu of using the lower bound crack arrest critical stress intensity factor equation (i.e., KIA) which is based on conditions needed to arrest a propagating crack, and which is the method invoked by Appendix G to Section XI of the ASME Code. Use of the Kic equation to determine the lower bound fracture toughness when computing P/T curves is more technically correct than the use of the KIA equation since the rate of loading during a heatup or cooldown is slow, and since crack initiation, which is more representative of a static condition than a dynamic condition, is principally at issue. The Kic equation appropriately implements the use of the static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of a reactor vessel.

Appendix G to 10 CFR 50 required the use of the conservative KIA equation since 1974, when the equation was codified. Use of the conservative KIA equation was considered necessary due to a limited knowledge of reactor pressure vessel material properties at the time. A significant amount of additional materials property data have been collected about RPV fabrication materials since 1974 and have provided the staff with a better understanding of how the RPV materials behave in service. For this reason, the staff has concluded that this additional information is sufficient to permit a lower bound static crack initiation critical stress intensity factor (Kic equation) coupled with a finite element analysis methodology to be used when calculating P/T limits, as described in Section 1.4.2 of the NRC Safety Evaluation related to Topical Report CE NPSD-683-A, Revision 6 (Ref. 1). In addition, P/T curves based on the K1c equation will enhance overall plant safety by opening the P/T operating window with the greatest safety benefit in the region of low temperature operations. Thus, pursuant to 10 CFR 50.12(a)(2)(ii), the underlying purpose of the regulation will continue to be served.

Exemptions from the requirements of 10 CFR 50 Appendix G may be granted by the Commission in accordance with 10 CFR 50.12. For the reasons discussed below, the exemption criteria in Section 50.12 are satisfied by this application.

3.1 Exemption is Authorized by Law Title 10 CFR 50.12(a)(1) requires a demonstration that an exemption from NRC regulations is authorized by law. This demonstration is found in 10 CFR 50.60 which defines acceptance criteria for fracture prevention measures for normal operation of light water nuclear power reactors.

Paragraph (a) of 10 CFR 50.60 requires that PVNGS Units 1, 2, and 3 meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in 10 CFR 50 Appendices G and H. Paragraph (b) of 10 CFR 50.60 advises that proposed alternatives to the described requirements in 10 CFR 50 Appendices G and H may be used when an exemption is granted by the 4

Enclosure 2 Application for Exemption from 10 CFR Part 50, Appendix G Commission pursuant to 10 CFR 50.12. Accordingly, this exemption request is authorized by law, as required by Section 50.12(a)(1).

3.2 Granting this Exemption Will Not Present an Undue Risk to the Public Health and Safety Title 10 CFR 50.12(a)(1) requires a demonstration that the granting of an exemption from the requirement in question will not present an undue risk to the public health and safety. As demonstrated below, this exemption request fully satisfies this criterion.

Requirements to monitor and control the pressure and temperature imposed on the PVNGS Units 1, 2, and 3 reactor coolant system pressure boundaries during heatup, cooldown, testing and normal operation remain unchanged as a result of this exemption request. Further, any conceivable risk would be equivalent to that inherent in any other license application where an exemption request to apply the KIc crack stress intensity factors was permitted by the staff.

Any risk to the public health and safety created by granting this exemption request will be mitigated by a number of factors. First, existing licensee programs and activities which serve to ensure safe plant operation (e.g., operational, maintenance, engineering, and corrective action programs and processes) will remain in effect during operation of PVNGS Units 1, 2, and 3. Second, the full array of NRC inspection and oversight activities will remain in effect, including the agency's authority to shut down any or all units at PVNGS. Third, these inspection and oversight activities will be further and fully informed by NRC Staff review of this PVNGS license amendment request, which will have been completed by the time the exemption request is granted. Fourth, no changes are made to the methods used to develop the PTLR nor to the application of such P/T results to operation of the Palo Verde Nuclear Generating Station, therefore granting the requested exemption will not present an undue risk to the health and safety of the public.

3.3 Granting this Exemption is Consistent with the Common Defense and Security NRC requirements relating to maintaining the integrity of the reactor coolant system pressure boundary are fully met by this exemption request. The exemption requested in no way affects the security or safeguards features or programs at PVNGS. Such features and programs will remain in full effect during the term of each unit's operating license. Accordingly, granting the requested exemption is consistent with the common defense and security.

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Enclosure 2 Application for Exemption from 10 CFR Part 50, Appendix G 3.4 Special Circumstances Support the Issuance of an Exemption Title 10 CFR 50.12(a)(2) requires a showing of at least one of six "special circumstances" to support issuance of the requested exemption. One of the special circumstances identified in Section 50.12(a)(2) appliesto this request, that is, the application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of the regulations in 10 CFR Part 50, Appendix G, is to provide an acceptable margin of safety against brittle failure of the RCS during any condition of normal operation to which the pressure boundary may be subjected over its service lifetime. Special circumstances, pursuant to 10 CFR 50.12(a)(2)(ii), apply to this exemption request in that continued operation of PVNGS Units 1, 2, and 3 with P/T limit curves developed in accordance with the ASME Code, Section Xl, Appendix G, without the authorization to utilize the alternative KIM calculational methodology of CE NPSD-683-A, Revision 6, is not necessary to achieve the underlying purpose of 10 CFR Part 50, Appendix G. Application of the calculational methodology documented in CE NPSD-683-A, Revision 6, in lieu of the calculational methodology specified in the ASME Code, Section X1, Appendix G, provides an acceptable alternative evaluation procedure that will continue to meet the underlying purpose of 10 CFR Part 50, Appendix G.

Therefore, APS requests an exemption based on the special circumstances of 10 CFR 50.12(a)(2)(ii), "Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule."

4.0 Precedent The analysis methodology of CE NPSD-683-A employs an alternate finite element analysis method for calculating stress intensity factors for the reactor pressure vessel shell. Upon review of an application by Southern California Edison (SCE) for the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 (Ref. 3), the NRC determined that sufficient information was presented to assess the method for calculating KIM factors. Except for loading inputs, the staff found that the KIM calculation methods, as applied to the P/T limits developed for SONGS-2 and -3, utilize finite element analysis modeling in a manner that is similar to that endorsed by Appendix G of the ASME Code. The staff also determined that only a slight non-conservative difference existed between the P/T limits generated from the 1989 edition of Appendix G to the ASME Code as compared to those generated using the methodology of CE NPSD-683-A. The staff found this difference to be reasonable and consistent with the expected improvements in P/T generation methods that have been incorporated into later editions of Appendix G to the ASME Code. Therefore, the staff concluded that the methodology of CE NPSD-683-A for generating P/T limits is equivalent to the current methodology in the 1995 edition of Appendix G to the Code and is acceptable for P/T limit applications.

6

Enclosure 2 Application for Exemption from 10 CFR Part 50, Appendix G Consistent with the conditions imposed by the staff's safety evaluation concerning topical report CE NPSD-683-A, Revision 06, SCE's January 12, 2006, submittal (Ref. 4) included a request for an exemption from certain requirements of 10 CFR Part 50, Appendix G when calculating P/T limits. The specific exemption requested involved the application of CE Nuclear Steam Supply System finite element analysis methodology for calculating KIM stress intensity values due to internal pressure loading rather than the linear elastic fracture mechanics methods promulgated in Appendix G to 10 CFR 50 and described in Appendix G to the ASME Code Section X1. The NRC staff authorized the Southern California Edison exemption request in a letter dated June 5, 2006 (Ref. 5).

5.0 Conclusion Title 10 CFR 50.60(b) permits licensees to use alternatives to the requirements of Appendix G to Part 50 if an exemption is granted by the Commission pursuant to the provisions and exemption acceptance criteria of 10 CFR 50.12. The staff has previously granted permission to Southern California Edison through the exemption request process to apply CE finite element methods and ASME Code Case N-640 to the calculation of plant-specific P/T limits (Refs. 5 and 6).

Analytical procedures employed by Westinghouse to develop the PVNGS reactor vessel P/T limits use the finite element analysis methods developed by Combustion Engineering and the guidance found in Appendix G of ASME Section Xl. Use of Kic crack stress criteria to calculate the allowable fracture toughness when establishing P/T limits for PVNGS is consistent with Section XI of the 2001 Edition of the ASME Code.

The justification presented in this application provides sufficient grounds for issuance of the requested exemption to APS.

As required by Section 50.12 of the NRC regulations, the exemption sought is authorized by law, presents no undue risk to public health and safety, is consistent with the common defense and security, and is supported by special circumstances. A thorough safety and technical review of the PVNGS PTLR and proposed changes to the Technical Specifications by the staff are sufficient to provide reasonable assurance of continued safe operation of PVNGS Units 1, 2, and 3 without increased risk to the health and safety of the public and without any potential environmental impact.

Accordingly, APS respectfully requests that the NRC grant the exemption from the requirements of Section IV.A.2 of Appendix G to 10 CFR Part 50 as applied to the development of RCS pressure-temperature limits for PVNGS Units 1, 2, and 3.

6.0 References

1. Combustion Engineering Owners Group Topical Report CE NPSD-683-A, Revision 6, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," April 2001 (ADAMS Accession No. ML011350387).

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Enclosure 2 Application for Exemption from 10 CFR Part 50, Appendix G

2. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for ASME Section XI, Division 1."

3. Letter from Southern California Edison Company (SCE) to the NRC, "San Onofre Nuclear Generating Station Units 2 and 3, Docket Nos. 50-361 and 50-362, Proposed Change Number NPF-10/15-551, License Amendment Request,

'Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)'," January 28, 2005 (ADAMS Accession No. ML050320286).

4. Letter from Southern California Edison Company (SCE) to the NRC, "San Onofre Nuclear Generating Station, Units 2 and 3, Docket Nos. 50-361 and 50-362, Proposed Change Number NPF-1 0/15-551, License Amendment Request,

'Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)'," January 12, 2006 (ADAMS Accession No. ML0601901011).

5. Letter, N. Kalyanam (NRC) to R. M. Rosenblum (SCE), "San Onofre Nuclear Generating Station, Units 2 and 3 - Exemption from the Requirements of Appendix G to 10 CFR Part 50, (TAC NOS. MC5773 AND MC5774)" dated June 5, 2006 (ADAMS Accession No. ML0611730433).

6. Letter from the NRC to Southern California Edison Company (SCE), "San Onofre Nuclear Generating Station, Units 2 And 3 - Issuance of Amendments Re:

Reactor Coolant System (RCS) Pressure And Temperature Limits Report (PTLR)

(TAC NOS. MC5773 AND MC5774)," July 13, 2006 (ADAMS Accession No. ML062170006).

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ML090641016

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