ML20248D363
| ML20248D363 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/30/1989 |
| From: | Shaun Anderson, Meyer T, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML093440949 | List: |
| References | |
| WCAP-11045, WCAP-11045-R01, WCAP-11045-R1, NUDOCS 8910040223 | |
| Download: ML20248D363 (54) | |
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7-- O ATTACHMI T I TO IPN-89-060 WCAP-11045 REVISION 1
- ' INDIAN POINT UNIT 3 REACTOR VESSEL FLUENCE AND RTi rts EVALUATIONS" NEW YORK POWER AUTHORITY INDIAN POINT 3 NUCLEAR POWER PLANT DOCKET NO. 50-286
y ... WESTINGHOUSE CLASS 3-l . *; . WCAP-11045 Revision 1. I b INDIAN POINT' UNIT 3 REACTOR VESSEL FLUENCE AND RT EVALUATIONS PTS S. E. Yanichko S. L. Anderson N. K. Ray
. Work Performed for New York Power Authority June 1989 Approved by:
DN' T. A. Meyer, Mana'ger ' Structural Materials and Reliability Technology Work Performed Under Shop Order IVGP-108 Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval. 1
^- . WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 3706s/061000.10 m______..______. _ . _ . . . _
^l'J:;
@m TABLE OF CONTENTS:
'Section Title Pager TABLE.0F CONTENTS i LIST OF TABLES ii
' LIST 0F FIGURES iii'
.I.- INTRODUCTION 1 I.1 The Pressurized Thermal Shock Rule 1 1.2 The Calculation of RTPTS 3 II. ~ NEUTRON EXPOSURE EVALUATION 5 11.1 Method of Analysis 6-11.2 Fast Neutron Fluence Results .10
'III. MATERIAL PROPERTIES 27 III.1 Identification and Location of Beltline Region Materials 27 III.2 Definition and Source of Material Properties for All 27 Vessel Locations III.3 Summary of Plant-Specific Material Properties 28 IV. ~ DETERMINATION OF RT PTS VALUES FOR ALL BELTLINE 31 REGION MATERIALS IV.1 Status of Reactor Vessel Integrity in Terms of RTPTS 31 versus Fluence.Results
-IV.2 Discussion of Results 32 1
V. CONCLUSIONS AND RECOMMENDATIONS 35 VI. REFERENCES 37 l-. VII. APPENDICES A-1 A. Core Power Distributions B. Weld Chemistry B-1 C. RT PTS Values of Indian Point Unit 3 Reactor Vessel C-1 Beltline Region Materials m m m eio $ I: L .- - - - - - - - - - _ - _
+
LIST OF TABLES 1 Page o 11.2-1 . Indian Point Unit 3 Fast Hautron (E>1.0 MeV) Exposure at the 14 Reactor Vessel ~ Inner Radius - 0* Azimuthal Angle II.2-2 Indian Point Unit 3 Fast Neutron (E>1.0 MeV) Exposure at the - 15 Reactor Vessel Inner Radius - 15' Azimuthal Angle . . II.2-3 Indian Point Unit 3 Fast Neutron (E>1.0 MeV) Exposure at the 16
% actor Vessel Inner Radius - 30* Azimuthal. Angle .
'II.2-4 Indian Point Unit 3 Fast' Neutron (E>1.0 MeV) Exposure at the 17 Reactor Vessel Inner Radius - 45' Azimuthal Angle 11.2-5 Indian' Point Unit 3 Fast Neutron'(E>1.0 MeV) Exposure at the 18 4* Surveillance Capsule Center 71.2-6 Indian Point Unit 3 Fast Neutron (E>1.0 MeV) Exposure at the 19 40' Surveillance Capsule Center II.2-7 Relative Radial Distributions of Neutron Flux (E>1.0 MeV) 20 Within the Pressure Vessel Wall II.2-8 .dpa/ Fluence (E>1.0 MeV) Ratios at the Pressure Vessel Inner 21 Radius and at the Center of Surveillance Capsules 11.2-9. Relative Radial Distributions of Iron Atom Displacement Rate 22 (dpa) Within the Pressure Vessel Wall III.3-1 Indian Point Unit 3 Reactor Vessel' Beltline Region Material 29 Properties IV.1-1 RT PTS Values for Indian Point Unit 3 33 A-1 Indian Point Unit 3 Core Power Distributions Used in A-3 the' Fluence Analysis B-1 Indian Point Unit 3 Intermediate to Lower Shell Circum- B-2 ferential Weld Chemistry From WOG Materials Data Base -
Wire Haat Number 13253 C.1 RT Values for Indian Point Unit 3 Reactor Vessel Beltline C-2 RehI5nMaterialsatVariousFluences C.1-2 RT Values for Indian Point Unit 3 Reactor Vessel Beltline C-4
- l. RehI5n Materials 9 Current (6.73 EFPY) Fluence Values C.1-3 RT Values for Indian Point Unit 3 Reactor Vessel Beltline C-5
- RehIhnMaterials@LicenseExpiration(21.88EFPY) h
_ .. ii
-4 LIST OF FIGURES ~
PAGE II.1-1 Indian Point Unit 3, R Theta Reactor Geometry 23 11.1-2 Surveillance Capsule Geometry 24 11.2-1 Indian Point Unit 3 Fast Neutron (E>1.0 MeV) Fluence at 25 the Vessel Beltline as a Function of Operating Time II.2-2 Azimuthal Variation of Fast Neutron (E>1.0 MeV) Fluence 26 at the Vessel as a Function of Operating Time III.1-1 Identification and Location of Beltline Region Material 30 for the Indian Point Unit 3 Reactor Vessel IV.1-1 Indian Point Unit 3 - RT Curves per PTS Rule Method [1] 34 DocketedBaseMateriala$d3WOG Data Base Mean Material Properties A-1 Indian Point Unit 3 Core Description for Power Distribution A-4 Nap 9 ie 6 me. min io $$$
I o 1 p SECTION I INTRODUCTION s The purpose of this report is to submit .he reference temperature for I pressurized thermal shock (RTPTS) values for the Indian Point Unit 3 reactor vessel to address the Pressurized Thermal Shock (PTS) Rule. Section I discusses the Rule and provides the methodology for calculating RTPTS* Section 11 presents the results of the neutron exposure evaluation assessing the effects that past and present core management strategies have had on neutron fluence levels in the reactor vessel. Section III provides the reactor vessel beltline region material properties. Section IV provides the RTPTS calculations from present through the projected end-of-license fluence values. 1.1 THE PRESSURIZED THERMAL SHOCK RULE The Pressurized Thermal Shock (PTS) Rule [1] was approved by the U.S. Nuclear
- Regulatory Commissioners on June 20, 1985, and appeared in the Federal Register on July 23, 1985. The date that the Rule was published in the Federal Register is the date that the Rule became a regulatory requirement.
The Rule outlines regulations to-address the potential for PTS evente on pressurized water reactor (PWR) vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC). PTS events have been shown from operating experience to be transients that result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.
~
O me.m io io i
The Rule establishes the following requirements for all domestic, operating PWRs:
- Establishes the RT (measure of fracture resistance) Screening CriterionfortheNhetorvesselbeltlineregion '
270*F for plates, forgings, axial welds 300*F for circumferential weld materials
- 6 Months From Date of Rule: All plants must submit their present ,
RT vaNhsattheexpirationdateoftheoperatinglicense. values Thed$ (pertheprescribedmeth this submittal must be received by the NRC for plants with operating licenses is January 23, 1986. l
- 9 Months From Date of Rule: Plants projected to exceed the PTS Screening Criterion shall submit an analysis and a schedule for implementation of such flux reduction programs as are reasonably practicable to avoid reaching the Screening Criterion. The data for this submittal must be received by the NRC for plants with operating licenses by April 23, 1986.
- Requires plant-specific PTS Safety Analyses before a plant is within 3 years of reaching the Screening Criterion, including analyses of alternatives to minimize the PTS concern. .
- Requires NRC approval for operation beyond the Screening Criterion. .
In the Rule, the NRC provides guidance regarding the calculation of the toughness state of the reactor vessel materials - the " reference temperature for nil ductility transition" (RTNDT). For purposes of the Rule, RTNDT is now defined as "the reference temperature for pressurized thermal shock" (RTPTS) and calculated as prescribed by 10 CFR 50.61(b) of the Rule. Each - USNRC licensed PWR must submit a projection of RT PTS values from the time of the submittal to the license expiration date. This assessment must be submitted within 6 months after the effective date of the Rule, on January 23, 1986, with updates whenever changes in core loadings, surveillance measurements, or other information indicate a significant change in projected values. The calculation must be made for each weld and plate, or forging, in the reactor vessel beltline. The purpose of this report is to provide the uen u RT PTS values for Indian Point Unit 3. , me.a= in 2
u e p .. ; g I.2 THE CALCULATION 0F RT PTS i I In-the PTS Rule, the NRC Staff has selected a conservative and uniform method for, determining plant-specificsvalues of RT PTS at:a given time. The prescribed equations in the PTS rule for calculating RT PTS are actually
.one of.several ways to calculate RT NDT. For the purpose;of comparison with the Screening' Criterion, the value of RTPTS f r the reactor vessel must be calculated for each weld and plate, or forging in the beltline region as given below. For each material, RT PTS is the lower of the results given by Equations 1 and 2.
Equation 1: 0 RT PTS = I + M + [-10 + 470(Cu) + 350(Cu)(Ni)] f .270 Equation 2: 0 RT PTS = I + M + 283 f .194 where I = the initial reference transition temperature of the unirradiated material measured as defined in the ASME Code, NB-2331. If a measured value is not available, the following-generic mean values must be used: 0*F for welds made with Linde 80. flux, and -56*F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes. M = the margin to be added to cover uncertainties in the values of initial RTNDT, copper and nickel content, fluence, and calculation procedures. In Equation 1, M=48"F if a measured value of I was used, and M=59'F if the generic mean value of I was used. In Equation 2, M=0*F if a measured value of
. I was used, and M=34'F if the generic mean value of I was used.
. me.mnom to 3
i
.Cu and Ni = the best estimate weight percent of, copper and nickel in the material.
f = the maximum neutron fluence, in units of 1019n/cm2 (E greater'than or ,
- equal to 1 MeV), at.the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question.-
Note, since the chemistry values given in equations 1 and 2 are best estimate mean values, and the margin, M, causes the RTPTS values to be upper bound predictions, the mean material chemistry values are to be used, when available, so as not to compound conservatism. The basis for the Cu and Ni values used in the RTPTS calculations for Indian Point Unit 3 are discussed in Section III.2. 6 m e me.<=mun 4
SECTION II. NEUTRON EXPOSURE EVALUATION
-This section describes a discrete ordinates Sn transport analysis performed for the Indian Point Unit 3 reactor to determine the neutron radiation environment within the reactor vessel and surveillance capsules. Fast neutron
> . exposure parameters in terms of fast neutron fluence (E>1.0 MeV) and iron atom displacements (dpa) are established on a plant and fuel cycle specific basis for the first seven reactor operating fuel cycles; and, based on these evaluations, projections of vessel exposure for future operating periods using the most recent cycle design are estimated. Neutron dosimetry results from the first three surveillance capsules withdrawn from the Indian Point Unit 3 reactor are integrated with the analytically derived exposure values to provide an overall "best estimate" of the current and future exposure of the pressure vessel. The use of fast neutron fluence (E>1.0 MeV) to correlate measured material
- properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutr'on energy spectra among surveillance capsule locations and positions within the pressure vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853,
" Analysis and Interpretation of Light Water Reactor Surveillance Results",
recommends reporting iron atom displacements (dpa) along with fluence (E>1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of
- Displacements per Atom". The Application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure 3786s/08f 589 to 5
.vesselwallhasalreadybeenpromulgated'inRevision2SoRegulatoryGuide
- 1.99,- " Radiation Embrittlement of Reactor Vessel Materials". Therefore, in keeping with the philosophy espoused in the' current-ASTM standards governing
' pressure vessel exposure evaluations, dpa data'is'also' included in.this~ ,
section. 11.1- METHOD OF ANALYSIS In performing the fast neutron exposure evaluations for the Indian Point Unit
'3 reactor pressure vessel, two distinct sets of transport calculations were carried out.- The first, a single computation in the conventional-forward mode, was used nrimarily to_obtain relative neutron energy distributions throughout the reactor geometry as well as'to establish relative radial distributions of. exposure parameters [ fluence (E>1.0 MeV) and dpa) through
'the vessel wall. The neutron spectral information is required for the
. interpretation of neutron dosimetry withdrawn from surveillance capsules as
^
well as for.the determination of exposure parameter ratios; i.e., dpa/ fluence (E>1.0 MeV), within the pressure vessel geometry. The relative radial. . gradient information is required to permit the projection of measured exposure parameters to locations ~ interior to the pressure vessel. wall; i.e., the 1/4T, - 1/2T, and 3/4T locations. The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E>1.0 MeV) at surveillance capsule positions and several azimuthal locations on the pressure vessel. inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provide the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded' absolute predictions of neutron exposure at the locations of interest for.each of the first seven operating fuel cycles; and established the means to perform similar predictions and dosimetry evaluations for subsequent fuel i cycles. It is important to note that the cycle spec Uic neutron source . distributions utilized in tf ese analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects -
* * " ' 6
. , ex.
of. varying neutron yield per fission and fission spectrum introduced by the build-in of plutonium as the burnup of individual. fuel assemblies increased. The absolute cycle specific data from the adjoint evaluations together with- l
~'
relative energy spectra and radial distribution information from the forward calculation provide the means to: ;
}
- 1. Evaluate neutron dosimetry obtained from the surveillance capsule ,
program. 1
- 2. Extrapolate dosimetry results te key locations at the inner radius and through the thickness of the pressure vessel wall. ;
I
- 3. Enable a direct comparison of analytical prediction with measurement.
- 4. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
A plan view of the Indian Po nt Unit 3 reactor geometry _at_the core midplane Js_showninfigure11.1-1., ince the reactor exhibits 1/8 core symmetry oni Q 0-45~aegree sector-is_idepictedNaddition to the core, reactor iiiternals, pressurevessel,an'$primarybiologicalshield,themodelalsoincluded explicit' representations of the surveillance capsules, the pressure vessel cladding, and the insulation located external to the pressure' vessel. From a neutronic standpoint, the inclusion of the surveillance capsules in the analytical model is significant. Since the presence of the capsules and associated support structure has a marked impact on both the flux magnitude and energy spectra at dosimetry locations, a meaningful comparison of measurement and calculation can be made only if these perturbation effects are
^
properly accounted for in the analysis. These analytical / experimental comparisons are important since, on a plant specific basis, they establish the
. overall accuracy of the analytical technique. A plan view of a single surveillance capsule attached to the thermal shield is shown in figure 11.1-2.
mummua 7
-The forward transport calculation for the reactor model summarized in figures 11.1-1 and II.1-2 was carried out in R, Theta geometry using the DOT
~
two-dimensional discrete ordinates code [2] and the SAILOR cross-section library [3]. The SAILOR library is a 47 group ENDFB-IV based data set , produced specifically for light water reactor applications. In this analysis anisotropic scattering was treated with a P3 expansion of the scattering cross-sections and the angular discretization was modeled with an 58 order of angular quadrature. The reference forward calculation was normalized to a core midplane rower density characteristic of operation at a thermal power level of 3025 MWt. The reference core power distribution utilized in the forward analysis was derived from statistical studies of long term operation of Westinghouse 4-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel mananment strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2 sigma uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was added to the nominal - assembly power level for all fuel assemblies adjacent to the core baffle plates. An axial peaking factor of 1.20 was also employed to scale the , axially averaged power distribution to the midplane value. Since it is unlikely that a single reactor would have a power distribution at the nominal
+ 2 sigma level and would maintain an axial peaking factor of 1.20 for a large number of cycles, the use of this reference case is expected to yield somewhat conservative results. This is especially true in cases where low leakage fuel I
management has been employed. The reference core power distribution data used in the analysis is provided in Appendix A to this report. The data listed in Appendix A represents cycle averaged assembly powers. All adjoint analyses were also carried out using a P3 cross-section approximation from the SAILOR library and an S8 order of angular quadrature. i Adjoint source locations were chosen at several locations at the pressure vessel inner radius and at the geometric center of surveillance capsules , positioned at 4 degrees and 40 degrees relative to the core cardinal axes. f Again, these calculations were run in R, Theta geometry to provide neutron -I source distribution importance functions for the exposure parameter of i 3786s/061060.10 g
y ,
. interest;.in this" case flux (E > 1.0 MeV). Having the importanca functions. l
,- and appropriate core source distributions, the response. of interest could be calculated as-
#(Ro',eo) = / # #R8E I(R,0,E) S(R,0,E) RdRdedE
- where:- ((Ro,so) = Neutron flux (E > 1.0 MeV) at the adjoint source location of radius Ro'and an Azimuthal angle so I(.R,0,E) = . Adjoint importance function at radius R, azimuthal angle e, and neutron source energy E
'S(R,0,'E) = Neutron source strength at core location R,0 and energy E
. Although the adjoint importance functions used in the Indian Point Unit 3 analyses were based on a response function defined by the threshold neutron r
flur.-(E > 1.0 MeV), prior calculations have shown that, while the-
,- implementation of. low leakage loading patterns significantly impact the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location.the ratio of dpa/ fluence (E > 1.0 MeV) is insensitive to changing
. core source-distributions. In the application of these importance functions to the Indian Point Unit 3 reactor, therefore, iron atom displacements (dpa) were computed on a cycle specific basis by using dpa/ fluence-(E.> 1.0 MeV) ratios.from the forward analysis in conjunction with the cycle specific fluence (E > 1.0 MeV) solutions from the individual adjoint evaluations.
The core power distributions used in the plant specific adjoint evaluations were taken from the appropriate fuel cycle design reports for the Indian Point Unit 3 reactor [4 through 10]. The data extracted from the design reports -
-represented. cycle averaged relative assembly powers, burnups, and axial peaking factors. Therefore, the adjoint results provided data in terms of
' fuel cycle averaged neutron flux which, when multiplied by the appropriate fuel cycle length, in turn yielded the incremental fast neutron exposure for the fuel cycle. In constructing the cycle specific energy dependent source me.mmman g
,/ . ,. ~ distributions account was taken of the burnup dependent inventory of' fissioning isotopes,-including U-235, U-238, Pu-239, Pu-240, Pu-241, and 4 Pur242. For comparison purposes with the reference c. ore power distribution-used in the forward analysis, the cycle specific relatfve assembly power ,
~ distributions used in the adjoint analyses are also provided in Appendix A to this report.
The transport methodology, both forward and adjoint, using the SAILOR cross section library has been benchmarked against' neutron dosimetry data obtained at the ORNL PCA facility [11). Extensive comparisons of analytical predictions with measurements from power reactor surveillance capsules and reactor cavity dosimetry programs have also been made.. The benchmarking studies. indicate that the use of SAILOR cross-sections and the generic reference core power distribution produces flux levels that tend to be conservative by from 7-22%. When plant specific' power distributions are used with the adjoint importance functions, the benchmarking studies show a tendency.to underpredict fluence levels at surveillance capsule positions by from 5-10%; while calculations applicable to reactor cavity locations tend to . be biased low by approximately 10-20% depending on the thickness of the pressure vessel.'In performing the exposure evaluation for the Indian Point - Unit'3 reactor the comparisons of predictions with' measurements obtained from
~ previously withdrawn surveillance capsules (T,Y, and Z) were factored into the overa11' analysis to provide best estimate exposure levels with a minimum uncertainty.- =
II.2' FAST NEUTRON FLUENCE RESULTS Best estimate fast neutron (E > 1.0 MeV) exposure results for the Indian Point Unit 3 reactor are presented in Tables II.2-1 through II.2-7 and in Figures 11.2-1 and 11.2-2. Data is presented at several azimuthal locations on-the inner radius of the pressure vessel as well as at the center of each surveillance capsule. Factors to convert the data provided in Tables 11.2-1 through 11.2-7 and Figures 11.2-1 and 11.2-2 to exposure in terms of iron atom . displacements rather than fluence (E > 1.0 MeV) are given in Tables II.2-8 and 11.2-9. -i l 10 l 3
- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - i
In: Tables 11.2-1 through'II.2-4 cycle specific maximum neutron flux and fluence levels at 0,15, 30, and 45 degrees on the reactor vessel inner radius
~ '
of.' Indian Point Unit'3 are listed for the period of operation up to May
,. 31,1989 and projected fluence levels are provided for irradiation periods L encompassing Lthe expiration date of the current operating license
!- (8/13/2009). LAlso presented are the reference fluence levels predicted using the generic 4-loop core power distribution at the nominal + 2 sigma level. Similar data applicable to the center of surveillance capsules located at 4 degrees'and 40 degrees are given in Tables 11.2-5 and 11.2-6, respectively. In regard to the cycle specific average flux and fluence levels listed in Tables 11.2-1 through 11.2-6,.it should be noted that the calculated neutron radiation levels have been multiplied by a factor of 1.086 to account for biases observed between cycle specific calculations and the results of neutron dosimetry for'the first three surveillance capsules [12,13,14] removed from the Indian Point Unit 3 reactor. The factor of 1.086 was derived by taking the average of the measurement to calculation. ratios (M/C) as follows. 4 CALCULATED MEASURED FLUENCE FLUENCE M/C 2 2 (n/cm ) (n/cm ) CAPSULE T 2.66E+18 2.99E+18 1.124 CAPSULE Y 7.03E+18 7.58E+18 1.078 CAPSULE Z 1.04E+19 1.10E+19 1.058
-AVERAGE 1.086 These three surveillance capsules were withdrawn from the 40 degree location at the end of cycles 1, 3, and 5, respectively. Application of this bias factor to the analytical results assures that the vessel exposure predictions represent conservative values with a minimum uncertainty. -- Graphical representations of the plant specific fast neutron fluence at key locations on the reactor vessel are shown in Figure 11.2-1. Reactor vessel a exposure data is presented for the 45 degree location on the circumferential l
me.amm 11
4 weld and shell plates as well as for all of the longitudinal welds located in the beltline' region (see Section III.1). In regard to Figure II.2-1, the
~
solid portinns of the fluence curves are based directly on the cycle specific q core loadings for the first six fuel cycles. The dashed portions of these , curves, however, involve a projection into the future. The neutron flux calculated for cycle 7 was used to project these future fluence levels. It should be noted that implementation of a more severe low leakage pattern than that represented by the cycle 7 loading pattern would reduce the projections of fluence at key locations. On the other hand, relaxation of the cycle 7 loading pattern or a return to out-in fuel management would increase those projections. The RTpts assessment must be updated per 10CFR50.61(b) [1] whenever, among other things, changes in core loadings significantly impact the fluence and RTpts projections. In Figure II.2-2, the azimuthal variation of maximum fast neutron (E > 1.0 MeV) fluence at the inner radius of the reactor vessel is presented. Data are depicted for both current (E0C 6) and projected conditions. In Table 11.2-7, . the relative radial variation of fast neutron flux and fluence within the reactor vessel wall is listed for the azimuthal locations of each of the key - vessel materials. Similar data showing the cycle specific relative axial variation of fast neutron flux and fluence over the beltline region of the reactor are included with the remainder of the cycle specific core power distribution information in Appendix A to this report. A two-dimensional description of the maximum fast neutron exposure of the reactor vessel wall can be constructed using the data given in Tables II.2-1 through II.2-7 along with the relation f(R,0) = f(e) F(R) , I where: f(R,0) = Fast neutron fluence at location R,0 within the reactor vessel wall 6(0) = Fast neutron fluence at azimuthal location r on the reactor vessel inner radius from Tables II.2-1 - j through II.2-4 me.mme.io 12 _ ___.___ .______ _______ _ _ m-_.-
^ +
F(R) = Relative fast neutron fluence at radius R into tne. vessel wall from Table-II.2-7 i Analysis has shown that the radial variations within the vessel wall are relatively insensitive toi the implementation of low leakage-fuel management schemes. Thus, the above relationship provides~a vehicle for a reasonable evaluation of fluence gradients within the vessel' wall. If, in addition to the R, Theta variation of maximum neutron expc.sure, an axial variation ove'r the beltline region of the vessel.is desired, these distributions-can be constructed by using the cycle specific axial power distributions. listed in Appendix A and reintegrating over the appropriate irradiation periods. All of the best estimate fast neutron (E > 1.0 MeV) exposure data can be converted to iron atom displacements (dpa) by making use of the data listed in Tables 11.2-8 and II.2-9. The exposure parameter ratios given in Table 11.2-8 can be employed in conjunction with the fluence information from Tables II.2-1 through 11.2-4 to calculate the dpa values at the pressure vessel inner
- radius. Distributional information within the vessel wall may then be calculated by normalizing the inner radius values to the relative radial gradient data given in Table II.2-9. ,6 4
I me.meinee in 13 m , _____ -_--- ---_ - --- -
J I
& TABLE II.2-1 k
- INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE '
REACTOR-VESSEL INNER RADIUS - 0 DEGREE AZIMUTHAL. ANGLE (a) l BELTLINE REGION ELAPSED CUMULATIVE FLUENCE (n/cm2) IRRADIATION IRRADIATION AVG. FLUX (b) PLANT INTERVAL TIME (EFPY) (n/cm2-sec) SPECIFIC (b) REFERENCE (c) CY-1 1.37 6.47E+09 2.80E+17 3.23E+17 CY-2' 2.23 7.70E+09 4.88E+17 5.26E+17 CY-3 3.29 7.95E+09 7.53E+17 7.76E+I7 CY-4 4.41 6.97E+09 1.00E+18 1.04E+18 , CY-5 5.55 6.41E+09 1.23E+18 1.31E+18 CY-6(d) 6.73 5.91E+09 1.45E+18 1.59E+18 - CY7-EOL(e) 21.88 5.53E+09 4.10E+18 5.16E+18 (a) Applicable to longitudinal weld 2-042C in the intermediate shell (b) Includes an analytical bias factor of 1.086 (c) Reference fast neutron flux = 7.47E+09 n/cm2-sec at 3025 MWt (d) Current neutron fluences are defined as of the end of cycle 6 (e) Fuel cycle projections are based'on the average neutron flux for cycle 7 and an assumed capacity factor of 0.75 e
" * " " ' 14
TABLE 11.2-2 INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE
. REACTOR VESSEL _ INNER RADIUS - 15 DEGREE AZIMUTHAL ANGLE (a)
BELTLINE REGION ELAPSED CUMULATIVE FLUENCE (n/cm2)
. IRRADIATION IRRADIATION AVG. FLUX (b) PLANT INTERVAL TIME (EFPY) (n/cm2-sec) SPECIFIC (b) REFERENCE (c)
CY-1 1.37 1.03E+10 4.46E+17- 5.23E+17 CY-2 2.23 1.22E+10 7.75E+17 8.52E+17 CY-3 3.29 1.25E+10 1.19E+18 1.26E+18 CY-4 4.41 1.05E+10 1.56E+18 1.68E+18 CY-5 5.55 9.34E+09 1.90E+18 2.12E+18
. CY-6(d) 6.73 8.84E+09 2.23E+18 2.57E+18 CY7-EOL(e)~ 21.88 8.95E+09 6.51E+18 8.35E+18 (a) Applicable to longitudinal welds 3-042A and 3-042B in the lower shell (b) Includes an analytical bias factor of 1.086 (c) Reference fast neutron flux = 1.21E+10 n/cm2-sec at 3025 MWt (d) Current neutron fluences are defined as of the end of cycle 6 (e) Fuel cycle projections are based on the average neutron flux for cycle 7 and an assumed capacity factor of 0.75 ,
I m e. w ou in 15 ;
- __ _____o
TABLE.II.2-3 INDIAN POINT UNIT 3' FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THEL , REACTOR VESSEL INNER RADIUS - 30 DEGREE AZIMUTHAL ANGLE'(a) BELTLINE REGION ELAPSED CUMULATIVE FLUENCE (n/cm2) IRRADIATION IRRADIATION AVG. FLUX (b) . PLANT INTERVAL TIME (EFPY) (n/cm2-sec)' SPECIFIC (b) REFERENCE (c) CY-1 1.37 1.27E+10 5.50E+17 6.49E+17 CY-2 2.23 1.59E+10 9.79E+17 1.06E+18 CY-3 3.29 1.40E+10 1.45E+18 1.56E+18 CY-4 4.41- 1.13E+10 1.85E+18 2.09E+18 CV-5 5.55 9.34E+09 2.18E+18 2.63E+18
- CY-6(d) 6.73 8.98E+09' 2.52E+18 3.19E+18 -
CY7-EOL(e) 21.88 7.98E+09 6.33E+18 1.04E+19 (a) Applicable to longitudinal welds 2-042A and 2-042B in the intermediate shell (b) Includes an analytical bias factor of 1.086 (c) Reference fast neutron flux = 1.50E+10 n/cm2-sec at 3025 MWt (d) Current neutron fluences are defined as of the end of cycle 6 (e) Fuel cycle projections are based on the average neutron flux for cycle 7 and an assumed capacity factor of 0.75 me.msiano 16
TABLE II.2 - INDIAN POINT UNIT 3 FAST. NEUTRON'(E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - 45 DEGREE AZIMUTHAL ANGLE (a) BELTLINE REGION ELAPSED CUMULATIVE FLUENCE (n/cm2) IRRADIATION IRRADIATION AVG. FLUX (b) PLANT INTERVAL TIME (EFPY) (n/cm2-sec) SPECIFIC (b) REFERENCE (c) CY-1 1.37 1.94E+10 8.40E+17- 1.00E+18 CY-2 2.23- 2.53E+10 1.52E+18 1.63E+18 CY-3 3.29 2.09E+10 2.22E+18 2.41E+18 CY-4 4.41 1.67E+10 2.81E+18 3.23E+18 CY-5 5.55 1.38E+10 3.31E+18 4.06E+18
, CY-6(d) 6.73 1.30E+10 3.79E+18 4.93E+18 CY7-EOL(e) 21.88 1.05E+10 8.81E+18 1.60E+19 (a) Applicable to longitudinal weld 3-042C in the lower shell, the intermediate to lower shell circumferential weld 9-042, and all shell plates (b) Includes an analytical bias factor of 1.086 (c) Reference fast neutron flux = 2.32E+10 n/cm2-sec at 3025 MWt (d) Current neutron fluences are defined as of the end of cycle 6 (e) Fuel cycle projections are based on the average neutron flux for cycle 7 and an assumed capacity factor of 0.75 me.mme in 17
_ _ _ _ _ _ - _~_
. TABLE II.2-5 INDIAN POINT UNIT 3 FAST NEUTRON-(E > 1.0 MeV) EXPOSURE AT.THE 4 DEGREE SURVEILLANCE CAPSULE CENTER BELTLINE REGION ELAPSED CUMULA.TIVE FLUENCE (n/cm2)
IRRADIATION IRRADIATION AVG. FLUX (a) PLANT INTERVAL TINE (EFPY) (n/cm2-sec)' SPECIFIC (a) REFERENCE (b) CY-1 1.37 2.16E+10 9.37E+17 1.08E+18 CY-2 2.23 2.58E+10 1.63E+18 1.76E+18 CY-3 3.29 2.66E+10 2.52E+18 2.60E+18 CY-4 4.41 2.33E+10 3.35E+18 3.48E+18 CY-5 5.55 2.14E+10 4.12E+18 4.38E+18 CY-6(c) 6.73 1.98E+10 4.85E+18 5.31E+18 . CY7-EOL(d) 21.88 1.85E+10 1.37E+19 1.73E+19 (a) Includes an analytical bias factor of 1.086 (b) Reference fast neutron flux = 2.50E+10 n/cm2-see at 3025 MWt-(c) Current neutron fluences are defined as of the end of cycle 6 (d) Fuel cycle projections are based on the average neutron flux for cycle 7 and an assumed capacity factor of 0.75 l I me.mm io 18 i
4
.'h-
. TABLE 11.2-6 i INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE-40 DEGREE-SURVEILLANCE CAPSULE CENTER BELTLINE REGION ELAPSED CUMULATIVE FLUENCE (n/cm2)
IRRADIATION IRRADIATION AVG. FLUX (a) PLANT
-INTERVAL. TIME-(EFPY) (n/cm2-sec) SPECIFIC (a) REFERENCE (b)
CY-1 1.37 6.67E+10 2.89E+18 3.47E+18 CY-2 2.23 8.79E+10 5.26E+18 5.64E+18 ' CY-3 3.29 7.14E+10 7.64E+18 8.33E+18 CY-4 4.41 5.67E+10 9.65E+18 1.12E+19 CY-5 5.55 4.64E+10 1.14E+19 1.40E+19
, CY-6(c) 6.73 4.34E+30 1.30E+19 1.70E+19 CY7-EOL.(d) 21.88 3.40E+10 2.92E+19 5.54E+19 (a)' Includes an analytical bias factor of 1.086 (b) Reference fast neutron flux = 8.02E+10 n/cm2-sec at 3025 MWt (c) Current neutron fluences are defined as of the end of cycle 6 (d) Fuel cycle projections are based on the average neutron flux for cycle 7 and an assumed capacity factor of 0.75
'S m w os m e.ie 19
TABLE 11.2-7 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL i Radius (cm) 0 Degrees 15 Degrees 30 Degrees - 45 Degrees t 220.27(a) 1.000 1.000 1.000 1.000 220.64 0.977 0.978 - 0.979 0.977 221.66 0.884 0.887 0.889 0.885 222.99 0.758 0.762 0.765 0.756 224.31 0.641 0.644 0.648 0.637 225.63 0.537 0.540 0.545 0.534 226.95 0.448 0.451 0.455 0.443 228.28 0.372 0.375 0.379 0.367 , 229.60 0.309 0.310 0.315 0.303 230.92 0.255 0.257 0.261 0.250 . 232.25 0.211 0.212 0.216 0.206 233.57 0.174 0.175 0.178 0.169 234.89 0.143 0.144 0.147 0.138 236.22 0.117 0.118 0.121 0.113 237.54 0.0961 0.0963 0.0989 0.0912 238.86 0.0783 0.0783 0.0807 0.0736 240.19 0.0635 0.0632 0.0656 0.0584 241.51 0.0511 0.0501 0.0519 0.0454 242.17(b) 0.0483 0.0469 0.0487 0.0422 (a) Indicates base metal inner radius (b) Indicates base metal outer radius , j mwmo ,o 20 ________.____m -_____._L.__ _ . . . _ ._.___..___.___._e _ . _ . _ _ . . _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _________.__.__._____.___.__-..____._____________-_______________.____.______.___m
'I TABLE 11.2-8 i
dpa/ fluence (E > 1.0 MeV) RATIOS AT THE PRESSURE VESSEL. INNER RADIUS AND AT THE CENTER OF SURVEILLANCE CAPSULES l [dpa]/ fluence (E'> 1.0 MeV) 0 deg. Vessel IR- 1.63E-21 l 15 deg. Vessel IR 1.61E-21 30 deg. Vessel IR 1.61E-21 45 deg. Vessel IR 1.63E-21
- . 4 deg.. Capsule 1.62E-21 40 deg. Capsule 1.71E-21 3780eM100010 21
i
. k TABLE II.2-9 RELATIVE RADIAL DISTRIBUTIONS OF IRON ATOM DISPLACEMENT RATE (dpa)
WITHIN THE PRESSURE VESSEL WALL 1 Radius (cm) 0 Degrees 15 Degrees 30 Degrees 45 Degrees 220.27(a) 1.000 1.000 1.000 1.000 220.64 0.983 0.983 0.984 0.983 221.66 0.913 0.914 0.918 0.915 222.99 0.818 0.819 0.827 0.820 224.31 0.728 0.728 0.739 0.730 225.63 0.647 0.646 0.659 0.647 226.95 0.574 0.573 0.587 0.573 228.28 0.510 0.507 0.523 0.507 . 229.60 0.453 0.450 0.466 0.449 230.92 0.402 0.399 0.414 0.397 - 232.25 0.356 0.353 0.368 0.349 233.57 0.315 0.312 0.327 0.307 234.89 0.277 0.275 0.289 0.269 236.22 0.243 0.241 0.254 0.233 237.54 0.212 0.210 0.222 0.201 238.86 0.182 0.181 0.192 0.170 240.19 0.155 0.154 0.164 0.141 241.51 0.131 0.128 0.137 0.113 242.17(b) 0.125 0.122 0.130 0.106 (a) Indicates base metal inner radius (b) Indicates base metal outer radius . 3786s/061089 10 22
. i l 'i, ,
(- e a.. la l I q - Oq Vg # . El ' 1 ae*
/ as'
% ,' %ei, ,e /
i y
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//
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Figure II.1-1. Indian Point Unit 3 R, Theta Reactor Geometry 3783e/060100 to 23
l
~
3.49 . 2.63_ . . 2.48 . [ 1.60 l 1.77 [ Radius (cm) , 213.85
<s\\\\\\\\\\\
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C - r / i ll / is C / 213.01 212,sg FLUX WIRES - \ h h 2d.18 a r P' 4P FLUX WIRES - , \ V V
/\ 211.68 CAPSULE CENTER - N C h 4e C h /\ 211.41
/\
FLUX WIRES - \/ a r a r 211.18 ,
! 210.68
\ f SPACER
/ /// / / / / / / / ////
\ _h,*gf
\\DDDDD\ 205.97 THERMAL SHIELD
'l l Figure II.1-2. Surveillance Capsule Geometry 3783s/060189:10 g
- c. = 3
.e c ' * :: __
E: : ... . . .. . . . . . .. .
- : use!ne!NAL WEm 3-OA C.
"ll ,l : ,l
. l l l , . , . 1
- CIRCLAFEFENTIAL WELD S-062 llllll l - lll l l ' l , l L 4 ALL W -- -
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..g . . 6 4 s . . e s . s a s .# . . ~
, ee I _l LONSITLSINAL WLDS 3-062A.S AT 15*
siiiii ii e i i i i ti I iois R: ~ W g t AT 30-
. . _ . .=. 5:3E E126--6 A
- LONSITLCINAL WELD 24lM2C AT O*.
g, - (, - _-- . c. w . "
. 2 v
~
- d. I : _- _--
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--- l N
j ,- [j - -==m.-.. ~ ggg
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- p n ir i i .
( E Jf s I t b I 1 ff f
- 1 . e ! . , e , , . , . . , , .
I I II I t4 i l e6 . 64 .., e e i . , . 6 . . 6e . .i I 'I ll .' i1 8 ' *I8 * *i 8 . *t
- 8 4 t*i I e i8 i eI t + * . ee 4 *ii.
~
10881 lli / > * 'i* ' ' ' *** * - ' ' ' ti>i 6 *
.mI .. -
. EEFf - -
a
. liN a . .
g =.__ a
. E 0 10 20 _ __ 30 40 50 60 70 OPEMTING TIME (EPPY)
' Figure 11.2-1. Indian Point Unit 3 Fast Neutron (E > 1.0 Mev) Fluence at the Vessel Beltline as a Function of Operating Time smmu in 25
- .t. 9 4
i c;20 - 9... = :- J g-m Ae ; ; :- # j;- 9 - q.- g;=sv a g;gs=. s rsz==s = 2== u $ = m a=r e s- gl 8 . .. - 7... 6... : _ _ _ 5... . . . . .
. _ . . . . . . - r . - = -.:q : . . : :: m= 2 - T . -- - - _:
;--.__.:____z ::: . __ _ a. = = = = - : _ x as a=l-s=
4... 3 ~~ ~ . N E u= N ' i ---4. C 2... i' ; , , : ; , ! , i i l l i l l l l l l l l Ld .: i i
,' i i i -- 60 CY i g i i i i ,i _e i i i i . i . . .i , i l i l i il I i l _ " I i f I I I l 'I e i I '
l _ ' I *l *I s l
- 4 1 l 8 l
~
g lO I9 i i t- ' ' ' , ,, ', i -, ", ~*"i i . 40 CY i i
, , , !=._ ._f 3 e=id R_ _ . - m .-- m ___=-. Z s.. M '- # - ==E2 LICENSE EXPIRATION.ii g ,~~
_- ,,.= .
;_~~
6.. : ._-- - - "
- s .. y - :
y ,; gx -- - s2=i [ i_ - __. __ __. w __ . _ . . .T 3 . .l __ l=--
^. _ _- z-9. r - 2_K- -M_==iE .._ _ _ __-~ _. __ . . : =E=: _ z . ~.~ :-R ]
y _g - ..- 4 l_. _ f . _ _ _ , .. ... t = -- - . := 2.. I # ~ ~ ~ '
-I e' ACTUAL d.
; y ;
i [ j .----- PROJECTED i i j
-1038 ! .
-{---
0- 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEGREE) Figure II.2-2. Azimuthal Variation of Fast Neutron (E > 1.0 Mev) Fluence at the Vessel Beltline as a Function of Operating Time
. .. _ , u
1 . . SECTION III MATERIAL PROPERTIES For the RTPTS calculation, the best estimato copper and nickel chemical ' composition of the reactor vessel beltline material is necessary. - The material properties for the' Indian Point Unit 3 beltline region will. be presented in this section. III.1' IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS The beltline region is defined by the Rule [1] to be "the region of the reactor vessel (she11' material including welds, heat affected zones, and plates or-forgings) that directly surrounds the. effective height of the active core and adjacent regions of the reactor vessel.that are predicted to experience sufficient neutron irradiation damage to be considered in the selection of the most limiting' material with regard to radiation damage."
' Figure III.1-1 identifies the location of all beltline region materials for the reactor vessel.
III.2 DEFINITION AND SOURCE OF MATERIAL PROPERTIES FOR ALL VESSEL LOCATIONS Material property values for the shell plates, which have been docketed with the NRC in Reference 12, were derived from vessel fabrication test certificates. The property data for the welds were derived from weld qualification test records and have also been reported in Reference 15. The tests were performed by the reactor vessel vendor, Combustion Engineering (CE), at the time of fabrication. Fast neutron irradiation-induced changes in the tensile, fracture, and impact properties of reactor vessel materials are largely dependent on chemical i composition, particularly in the copper concentration. The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration within the weldments. 3786s/061000 10
. 27
For'each weld in the Indian Point Unit 3 beltline region. a material data search was performed using the WDG Reactor Yessel Beltline Region Weld Materials Data Base. The WOG data base, which was developed in 1984 and is - continually being updated, contains information from weld qualification
~ -records (including those for Indian Point Unit 3), surveillance capsule reports, the B&W report BAW-1799, and the Materials Properties Council (MPC)-
and the NRC Mender MATSURV data bases. Searches were performed for materials having the identical weld wire heat number as those in the Indian Point Unit 3 vessel, but any combination of wire and flux was allowed. For all of the data found for a particular wire, the copper, nickel, phosphorous and silicon values were averaged and the standard deviations were calculated. Although phosphorous and silicon are not needed for the PTS Rule, they are provided for the sake of completeness. The information obtained from the data base searches is found in Appendix B. III.3
SUMMARY
OF PLANT-SPECIFIC MATERIAL PROPERTIES A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the Indian Point Unit 3 reactor vessel are . given in Table III.3-1. All of the initial RT NDT values (I) are given in Table III.3-1. The longitudinal weld initial RTNDT is the generic mean value as defined by the PTS rule [1]. The data in Table III.3-1 is used to evaluate the RTPTS values for the Indian Point Unit 3 reactor vessel. l l 3786s/061080 10 2g
;L ,
' TABLE:III.3-1 L' .
-. INDIAN POINT UNIT 3 REACTOR VESSEL' BELTLINE REGION MATERIAL PROPERTIES Cu Ni P 1(a)
(Wt.%) (Wt.%) (Wt.%) -(*F) Intermediate Shell Plate B2802-1: .20 .50 .010 5 L Intermediate Shell Plate B2802-2: .22 .53- .015 -4 Intermediate Shell Plate B2802-3: .20 .49 .011 17 Lower Shell-Plate B2803-1: .19 .47 .012 49
. Lower Shell Plate B2803-2: .22 .52 .011 '5 l.
Lower Shell Plate B2803-3: .24 .52 .012 74 Longitudinal Welds 042 A,B,C and 3-042 A,B,C, Wire Heat 34B009, Flux Linde 1092 WOG Data Base .19 1.00 .012 -56 Circum. Weld 9-042'- Intermed. to Lower Shell, Wire Heat 13253, Flux Linde 1092 Weld Qualification Value .27 .74 .023 -70
.WOG Data Base Mean .25 .72 .020 -54 (a) All Plate and circumfersatial Weld RTNDT (I) values are actual values.
The remaining value for the Longitudinal welds is a generic mean value as defined by the PTS rule [1]. 3786s/061089 10 29 l-' . i
[. CIRCUMFERENTIA.L SEAMS VERTICAL SEAMS 2-0428 90*
- B2802-2 g C -
8-042 21" f 30
\L
=
r 2-042C h J CORE 4-r180* O' -- -
= 30-N
< B2802-1 B2002-3 144" 5
$ ^
$ 2-042A 270* '
Q, . _ z g -=--- 0 4 2 15.9" g
~
90* tr f J B2003-B2OO3-2
= '
51"
/ CORE v / o. .. -
180* [I- 5' 3-042A B2003-1 3-042C 270* Figure 111.1-1 Identification and Location of Beltline Region Material for the Indian Point Unit 3 Reactor Vessel m w o.oise io 30
- g. . ' 'h' SECTION IV:
DETERMINATION OF RT"TS VALUES FOR ALL BELTLINE REGION MATERIALS
+
, )
~
Using the methodology prescribed-in Section I.2, the results of the fast. neutron exposure provided in Section II, and the material properties discussed. in;Section III, the RT PTS values for Indian Point Unit 3 can now be determined. IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF.RTPTS VERSUS FLUENCE RESULTS
- Using the prescribed PTS Rule methodology, RTPTS values were generated for
- all beltline region materials of the Indian Point Unit 3 reactor vessel as a
- function of several fluence values and pertinent vessel lifetimes. The
- tabulated results from the total evaluation are presented in Appendix C for
- all beltline region materials for both units.
Figure IV.1-1 presents the RT PTS values for the limiting longitudinal weld, _.. circumferential weld and limiting basemetal of the Indian Point Unit 3 va:sel
~
in terms'of RTPTS versus fluence
- curves. The curves in these figures can
- be used:
o to provide guidelines to evaluate fuel reload' options in relation to the NRC RT PTS Screening Criterion for PTS (i.e., RT PTS values can be readily projected for any options under consideration, provided fluence is known),and o - to show the current-(6.73 EFPY), and end-of-license (21.88 EFPY) RT PTS values _using actual and projected fluence. i
*The EFPY can be determined using Figure 11.2-1.
L.
- m mnonio 31
_-.E _________m_- m-_ _ . _ _ _ _ _ . _ _ _ _ . _ _ _. _ _ _ _ _ _ _ _ , . . . . _ _ _ _ . - _ _ _ _ .
Table IV.1-1 provides a sumary of the RTPTS values for all beltline region materials for the lifetime of interest. IV.2 DISCUSSION OF RESULTS
- As shown in Figure IV.1-1 and Table IV.1-1, lower shell plate B2803-3 is the limiting location relative to PTS.. At license expiration, plate B2803-3 is seen to have an RT PTS value of 264*F. All of the RT PTS values, including that for plate B2803-3, are below the NRC' screening criteria through license expiration.
e 9 a m e. m me,o 32
c.:
' TABLE IV.1-1 1 RT VALUES FOR INDIAN POINT UNIT 3 PTS n,
.RTPTS' Values-(*F)-
Present- End-of-License Screening Location Vessel Material (6.73 EFPY) (21.88 EFPY) Criteria i 1 Intermediate shell plate B2802-1 145 168 270 2 Intermediate shell plate B2802 147 174 270 3 -Intermediate'shell. plate B2802-3 156 179 270 4 Lower shell plate B2803-1 182 204 270 5 Llower shell plate B2803-2 146- 172 270 6 Lower shell plate B2803-3 235 264 270 7 Limiting Longitudinal Weld 3-042C 115 144 270 8 intermediate'to lower shell 125 159 300 circumferential weld 9-042 Longitudinal weld 2-042C 90 118 270 Longitudinal weld 3-042A,8 100 133 270 Longitudinal weld 2-042A,B 104 132 270 newoems:o 33 [ _. _ _ - ._ ___ _ __ _ _
1 L 1 400 -
. LIMITING BASEMETAL 350 --
NRC RTRTS SCREENING VALUE (3OO'F) - CIRCUhFERENTIAL WELD 3 - - -- - - - ------ TEC sTRTS 55EENING vAtt.i-(270 r r-
.C'_^IE jND goNGI{ WING
_wELOS g 250 - C L g 200 0
'E LONGITUDINAL WELDS INTERMEDIATE TO LOWER SHELL i50 -
CIRCUhFERENTIAL WELD I00 - 50 - 0 I I I I I III I I I I I I II 1010 1019 3o20 NEUTRON FLUENCE (n/cm2) LEGEND: A = Current Life (6.73 EFPY) and a = End of License 21.88 EFPY) RT Values Using Plant Specific and Projected Plant Specific Fluence PTS Values (Limiting Fluences Used for all Locations) Figure IV.1-1 Indian Point Unit 3 RT PTS Curves Per PTS Rule Method [1] Docketed Base Material and WOG Data Base Mean Material . Properties mwoeices io 34 __-____ _ __-__-_ D
l, f [ SECTION V CONCLUSIONS AND RECOMMENDATIONS Calculations have been completed in order to submit RT PTS values for the Indian Point Unit 3 reactor vessel in meeting the requirements of the NRC Rule for Pressurized Thermal Shock [1]. This work entailed a neutron exposure evaluation and a reactor vessel material study in order to determine the. RT PTS values. t Detailed fast neutron exposure evaluations using plant specific cycle by cycle core power dist,ributions and state of-the-art neutron transport methodology have been completed for the Indian Point Unit 3 reactor vessel. Explicit calculations were' performed for the first fuel cycles. Projection of the fast neutron exposure beyond Cycle 6 was based on continued implementation of low leakage fuel management similar to that employed during cycle 7.
' Plant specific evaluations have demonstrated that during fuel cycles using out-in-in fuel management, the maximum fast neatron (E > 1.0 MeV) flux incident on the reactor vessel was, on the average, 12 percent less than predictions based on the design basis core power distributions. With regard to the low leakage fuel management strategy in place at Indian Point Unit 3, the plant specific evaluations have shown that the average fast neutron (E > 1.0 MeV) flux at the 45* azimuthal position (peak location) was reduced by 26 percent relative to that prior to the implementation of low leakage.
It should be noted that significant deviations from the low leakage scheme already in place will affect the exposure projections beyond the current operating cycle. A move toward a more severe form of low leakage (lower relative power on the periphery) would reduce the projection. On the other hand, a relaxation of the loading pattern toward higher relative power on the core periphery would increase the projections beyond those reported. As each future fuel cycle evolves, the loading patterns should be evaluated to determine their potential impact on projections made in this report. 3786s/061089 10 35
o [ j; The fast neutron-fluence values from the plant specific calculations have been ]! compared directly with measured fluence levels derived from neutron dosimetry contained in surveillance capsules withdrawn from Indian Point Unit 3. The
- ratio of measured to calculated fluence values ranges from 1.058 to 1.124 for
~
the three capsule data points. This reasonably good agreement between calculation and measurement supports the use of this analytical approach to perform a plant specific evaluation for the Indian Point-Unit 3 reactor. Material property values for the Indian Point Unit 3 reactor vessel beltline region components were determined. The pertinent chemical and mechanical properties for the shell plates remain the same as those that were originally reported in the vessel fabrication test certificates. The weld material
. properties are obtained from the WOG Material Data Base.
Using the prescribed PTS Rule methodology, RTPTS values were generated for all beltline region materials of the Indian Point Unit 3 reactor vessel as a function cf several fluence values and pertinent vessel lifetimes. All of the RT values remain below the NRC screening values for PTS using the . PTS projected fluence exposure through the expiration date of the operating license. The most limiting value at end of-license is 264*F for the lower - shell plate B2803-3. A review of the material properties of shell plate B2803-3 indicates that a high initial value of RTNDT is the primary facto ~r causing the high RT PTS value. The results in this report are provided to enable New York Power Authority to comply with the requirements of the PTS rule. mwosme.in 36
SECTION VI REFERENCES
- 1. Nuclear Regulatory Commission 10CFR Part 50, " Analysis of Potential Pressurized Thermel Shock Events," Federal Register, Vol. 50, No.141, July 23, 1985.
- 2. Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation Vol. 5 - Two Dimensional, Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.
- 3. " SAILOR RSIC Data Library Collection DLC-76." Coupled. Self-Shielded, 47 Neutron, 20 Gamma-Ray, P , Cross-Section Library for :iskt Water 3
Reactors.
- 4. WCAP-8360, " Core Physics Characteristics of the Indian Point Nuclear Plant Unit III, Cycle 1," P. J. Sipush, et al., October 1974.
(WestinghouseProprietary)
- 5. WCAP-9244, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3, Cycle 2," D. M. Lucoff, et al., January 1978.
(Westinghouse Proprietary)
- 6. WCAP-9599, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3 Cycle 3," J. A. Penkrot, et al., September 1979.
(Westinghouse Proprietary)
- 7. WCAP-10051, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3, Cycle 4," M. A. Kotun and M. F. Muenks, March 1982.
(WestinghouseProprietary)
- 8. WCAP-10839, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3, Cycle 5," M. A. Kotun and R. H. Pitulski, June 1985.
(WestinghouseProprietary) 3786s/061089 10 37
- 9. WCAP-11537 "Nuclet Parameters and Operations Package for Indian Point Unit 3 Nuclear Power Plant Cycle 6," F. J. Silva, et al., July 1987.
(Westinghouse Proprietary) 10.- WCAP-12169 " Nuclear Parameters and Operations Package for Indian Point Unit 3 Nuciaar Powcr Plant Cycle 7," T. Downs, J. Cole and F. Silva, April 1989. (Westinghouse Proprietary)
- 11. NUREG/CR-1861, " LWR Pressure Vessel Surveillance Dosimetry Improvement Program - PCA Experiment and Blend Tests," W. N. McElroy, July 1981.
- 12. WCAP-9491, " Analysis of Capsule T from the Indian Point Unit No. 3
- Reactor Vessel Radiation Surveillance Program," J. A. Davidson, S. L.
Anderson and W. T. Kaiser, April 1979.
- 13. WCAP-10300, " Analysis of Capsule Y from the Power Authority of the State of New York, Indian Point Unit 3 nactor Vessel Radiation Surveillance Program," S. E. Yanichko, S. L. Anderson and W. T. Kaiser, March 1983. .
- 14. WCAP-11815, " Analyses of Capsule Z from the New York Power Authority -
Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko, S. L. Anderson, and L. Albertin, March 1988.
- 15. Letter from W. J. Cahill, Jr., of Consolidated Edison to the Director of Nuclear Reactor Regulation, Mr. R. W. Reid, Docket 50-286, March 8,1978.
N me.ame in 38
APPENDIX A i - CORE POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the Indian Point Unit 3 reactor vessel were derived from the following fuel cycle nuclear design reports: Fuel Cycle Nuclear Design Report 1 WCAP-8360 2 WCAP-9244 3 WCAP-9599 4 WCAP-10051 5 WCAP-10839 6 WCAP-11537 7 WCAP-12169 A schematic diagram of the core configuration applicable to Indian Point Unit 3 is shown in Figure A-1. Cycle averaged relative assembly powers for each fuel cycle are listed in Table A-1 along with the design basis core power distribution. On Figure A-1 and in Table A-1 an identification number is assigned to each fuel assembly location. Three regions consisting of subsets of fuel assemblies are defined. In performing the adjoint evaluations, the relative power in the fuel assemblies comprising Region 3 has been adjusted to account for known biases in the prediction of power in the peripheral fuel assemblies while the relative power in the fuel assemblies comprising Region 2 has been maintained at the cycle average value. Due to the extreme self-shielding of the reactor core neutrons born in the fuel assemblies comprising Region 1 do not contribute significantly to the neutron exposure of either the surveillance capsules or the reactor vessel. Therefore, core power distribution data for fuel assemblies in Region 1 are not listed in Table A-1. mwostoano A-1
(<.
. In each of.the' adjoint evaluations, within assembly spatici gradients.have
- been superimposed on the average assemblyL power levels. For the peripheral
' assemblies (Region' 3), these spatial' gradients also include adjustments to account for analytical deficiencies that tend to. occur near the boundaries of .
the core region. m b a W ) 37 foeio .so A2 L _ ._____.___.._______________________________O.
jf, ;
. i L
.* ?
, TABLE A-11
-INDIAN POINT UNIT 3 CORE. POWER DISTRIBUTIONS USED'IN THE FLUENCE ANALYSIS Design Cycle Averaged Relative Assembly Power
' Basis Fuel Cycle
' Fuel Relative Assembly Power 1 2 '3 4 5 6' 7
.1- 1.06 0.78 0.93 1.04 0.95 0.75 0.66 0.60 2 1.09- 0.84 0.94 1.02 0.94 0.90 0.83 0.62 3 1.01 0.74 0.88 0.95 0.89 0.79 0.65 0.81 4 0.81 0.63 0.70 0.76. 0.52 0.40 0.45 0.46 5 :1.15 0.88 0.99 0.98 0.90. 0.68 0.63 0.61 o
6 0.75 0.54 0.72 0.61 0.44 0.34' O.31 0.21
- 7. 1.02 .1.01 0.99 1.08 1.20 1.10 1.10 0.87
. 8 1.10 1.00 1.23 1.20 1.15 1.22 1.26 1.24 9 1.00 0.98 0.96 0.95 1.21 1.05 1.02 1.05 10 1.05- 0.98 1.16 1.23 1.12 1.05 1.08 1.14 11 1.07 1.06 1.01 0.93 1.21 1.09 1.08 1.18 l
12 1.00 0.94 1.07 0.97 1.09 1.13 1.15 1.18 13 ~1.05 0.96 1.16 0.70 0.87 0.79 0.74 0.60 l L l Note: Refer to Figure A-1 for fuel assen61y location. 1 A-3
1 i I L
~
O' l I-l 2 3' 4 t
- 7- 8 9 10 5 6 45'
/ t 14- 15 16 II 12 13 l
1
'l7- 18 19- 20 21 fl l
.22 23 24 25 26- 27 28 1
-29 30 REGION ASSEMBLIES I 14-31 31 2 7-13 ,
3 1-6 3 Figure A-1 Indian Point Unit 3 Core Description for Power Distribution Map me.was in A-4
APPENDIX B b - WELD CHEMISTRY Table B.1-1 provides the weld data output from the WOG Material Data Base. The pertinent material chemical compositions are given, along with j the wire / flux identification. The mean chemistry values and the. . population standard deviation are then calculated. The mean values of copper and nickel are used in the RTPTS analysis. Weld Chemistry Data Source and Plant: AEP - Donald C. Cook Unit 1 Cu Weight % of. Copper INT - Indian Point Unit 3 KEP - Mihama Unit 1 Ni - Weight % of Nickel P Weight % of Phosphorous PGE - Diablo Canyon Unit 1 PNJ - Salem Unit 2
. PSE -
Salem Unit 1 SC - Surveillance Capsule Si - Weight % of Silicon WQ Weld Qualification S 3766s/061089 10 g.}
. h . e h e S d h S h l S S r e r . t eW r e e t r e n e w n Iw N I c o o O n L L I o a o T t l o t o P l t t I e i e R l e r l r C z v e z ze t .
S z r t - E o u nI o n D N S ' N I T N - A E P T P E L G E N E S 0 2 P P A I. K P 0 8 0 7 5 7
~
3 7 0 0 2 0 L 9 8 L 2 1 0 0 E E i H S S 0 0 0 3 S A 0 4 B 0 2 R 0 4 E A 7 3 2 0 W T 1 2 0 0 O A 0 0 L D 0 0 P 0 0 O S 3 0 3 T L 5 0 1 _ A 2 - 0 2 E I 3 0 0 5 1 T R 1 1 4 2 2 A E 7 7 7 0 I T R i 1 D A E N 0 0 0 0
- E M B B M M 0 4 R G U 0 8
_. E E O N 0 0 0 2 L T W 3 7 0 8 B N : T 2 2 5 2 A I M A u 2 0 T O E C 0 0 3 R H 0 0 F T- - E N I Y R E N R I H E C C U T W S C D A R C S, 5, T I L T U J P N M E A O N E I E W D S P A O H P C X 4 1 N D U T 7 9 A L L O 7 7 I E F L 3 3 DW N I 2 2 9 9 0 0 1 1 E E X E D D U P N N L Y I I F T L L D 0 O 0 M N E E R P 4 4 I Y - - W T B B 3 3 3 3 3 E T 5 5 5 5 5 . R A 2 2 2 2 2 v I E 3 3 3 3 3 e WH 1 1 1 1 1 d 0 4 4 4 4 n . 9 9 9 9 9 a d D 2 2 2 2 2 e t I 0 0 0 0 0 m s [
' L
APPENDIX C RT VALUES OF INDIAN POINT UNIT 3 PTS REACTOR VESSEL BELTLINE REGION MATERIALS Tables C.1-1 through C.1-3 provide the RTPTS values, as a function of both constant fluence and constant EFPY.(assuming the projected fluences values), for all beltline region materials of the Indian Point Unit 3 reactor vessel. The RT values are calculated in accordance with the PTS rule, which is PTS Reference [1] in the main body of this report. The vessel location numbers in-the following tables correspond to the vessel materials identified.below and in Table III.3-1 of the main report. Location Vessel Material 1 Intermediate shell plate B2802-1 2 Intermediate shell plate B2802-2 g 3 Intermediate shell plate B2802-3 4 Lower shell plate B2803-1
- 5 Lower shell plate B2803-2 6 Lower shell plate B2803-3 7 Limiting Longitudinal Weld 3-042C 8 Intermediate to lower shell Circumferential Weld 9-042 m.,== io C-1
0 2
+
E 0 6 6 8 0 4 9 9 0 E 2 9 0 0 3 0 9 7 0 C 1 2 2 2 2 2 1 2 N ' E U 0 L 2 F + E T 0 2 8 3 8 6 8 9 5 A 1 7 7 8 0 7 6 4 6 1 1 1 2 1 2 1 1 E S U E L 9 e R C A 1 u O N V + l a T E E C U S 0 2 5 3 9 4 3 4 5 v A L T 5 5 5 6 8 5 4 2 3 E F P 1 1 1 1 1 2 1 1 T R T D S R N 3 U 9 T O 1 R T I + R E l I N A 0 7 6 9 6 5 1 1 6 a U V 1 1 1 2 5 1 0 8 8 ) i t
-. 1 1 1 1 1 2 T T 1 C i N A - I n I C R i O S E . . . . . . . . E ~
1 P L P M. M. M. M. M. M. WW eN f g E o
- A Y . .
1 N I T B B B B B B L C a G A R . p r p e C. D L I I e o y E N A C e t L I M L L L L L L I L s l B E A A A A A A R A ( a s A E N U U U U U U U E U u e t t F T H O L T T T T T T N T l T I A C C C C C C E C a c o g G V A A A A A A G A i a n R E r e e O R e , d d F t T E 5 4 7 9 5 4 6 4 a D " n S N I - 1 4 - 7 5 5 mN C d i T
- I l E I -
U L l R R e e L T e E d W r A L 0 5 1 2 1 2 2 0 s f N l a V E 1 1 1 1 1 1 1 2 s o E e l B P 0 0 0 0 0 0 0 0 e G W a s S v e " - i e r T L 0 0 0 0 0 0 0 0 u l t P E f l r a n ut T S o a o l n e R S 0 3 9 7 2 2 0 2 v a i r a E 5 5 4 4 5 5 0 7 n t d e r V I o l L" e u f m p e N 0 0 0 0 0 0 1 0 i a A M t t i U i u m a t T e g c e 0 2 0 9 2 4 9 5 c i C s n r t 2 2 2 1 2 2 1 2 o n A a o i U . . L I " B L C e C 0 0 0 0 0 0 0 0 c
= = = = = = n e
e r T u . . . e N l W. W f A T T T T T T T T D a M. . e L N N N N N N N N I I V B L C R P I I I I I I I I s e t D o I 1 2 3 4 5 6 7 8 N [ [
;j; l I g g
= t~ c -- p 0 E 2 C + N E E 0 4 1 5 4 9 0 0 2 U 7 5 7 6 8 6 7 5 8 L 2 2 2 2 2 3 2 2 F S T 0 E A 2 R C E
+
E O N T E U 0 6 2 7 6 9 0 0 1 C U. L 6 4 6 5 7 5 6 4 7 A L A 2 2 2 2 2 3 2 2 E F V R S S 0 3 U T 2 O P + T I T E I R R 0 6 9 7 3 7 5 5 2 N A 4 2 3 3 5 3 3 1 4 U V 2 2 2 2 2 3 2 2
) T T T N A
, N I E . . . . . . . .
O O S C P L P Y M. M. M. M. M. M. WW . . ( - A T B B B B B B L C N I
, 1 A R
- I E 1 D T C N A L L L L L L I L C I M E A A A A A A R A U U U U U U U E U E E N L T T T T T T N T L H O I A C C C C C C E C B T V A A A A A A G A A G T R E O R F 5 4 7 9 5 4 6 4 E I - 1 4 - 7 5 5 S N - -
E I U L L T 0 5 1 2 1 2 2 0 A L 1 1 1 1 1 1 1 2 V E 0 0 0 0 0 0 0 0 B P S 0 0 0 0 0 0 0 0 T l P r T S 0 3 9 7 2 2 0 2 R S 5 5 4 4 5 5 0 7 F I V N 0 0 0 0 0 0 1 0 0 2 0 9 2 4 9 5 2 2 2 1 2 2 1 2 U C 0 0 0 0 0 0 0 0 T N A T T T T T T T T
, L N N N N N N N N 0 P I I I I I I I I t 1 9
9 10 6 0/ s 0 D 7 8 I 1 2 3 4 5 6 7 8 3 [ [ lfl
l-l' TABLE C.1-2 RTPTS VALUES FOR THE INDIAN POINT UNIT 3 REACTOR VESSEL BELTLINE REGION MATERIALS AT CURRENT LIFE (6.73 EFPY) , FLUENCE VALUES 1 -
.ID PLANT CU NI P I VALUE TYPE FLUENCE RTPTS L 1- INT 0.20 0.50 0.010 5 ACTUAL B.M. 0.38E+19 145-2 INT 0.22 .0.53 0.015 -4 ACTUAL B.M. 0.38E+19 147 I
3 INT 0.20 0.49 0.011 17 Ar,TUAL B.M. 0.38E+19 156 4 INT 0.19 0.47 0.012 49 ACTUAL B.M. 0.38E+19 182 5- INT 0.22- 0.52 0.011 -5 ACTUAL B.M. 0.38E+19 146 6 INT 0.24 0.52- 0.012 74 ACTUAL B.M. 0.38E+19 235 7 INT 0.19 1.00 0.012 -56 GENERIC L.W. 0.38E+19 115 8 INT 0.25 -0.72 0.020 -54 ACTUAL C.W. 0.38E+19 125 9
+
W 4 m foeiano C-4
;4 i
t'- TABLE C.1-3 g RTPTS VALUES 70R-THE INDIAN POINT UNIT 3 REACTOR VESSEL I* BELTLINE REGION MATERIALS AT END OF LIFE (21.88 EFPY) FLUENCE VALUES i: .. I D PLANT- CU NI P I VALUE TYPE FLUENCE RTPTS 1 INT 0.20 0.50 0.010 5 ACTUAL B.M. 0.88E+19 168 2 INT. 0.22 0.53 .0.015 -4 ACTUAL B.M. 0.88E+19 174 3 INT 0.20 0.49 0.011 17 ACTUAL B.M. . 0.88E+19 179
.4 INT 0.19 0.47 0.012 49 ACTUAL B.M. 0.88E+19 204 5 INT 0.22 0.52 0.011 -5 ACTUAL B.M. 0.88E+19 172 6 INT 0.24 0.52 0.012 74 ACTUAL B.M. 0.88E+19 264 7' INT 0.19 1.00 0.012 -56 GENERIC L.W. 0.88E+19 144-8 INT 0.25 0.72 0.020 -54 ACTUAL ~ C.W. 0.88E+19 159 4
4 9 a ms.mme ,o c.5 _ _ _ _ .____ ____ _ __ ___ _ ___ -}}