ML20137K818
| ML20137K818 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/31/1986 |
| From: | Hirst C, Lau F, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20137K810 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR 4070E:1D-010986, 4070E:1D-10986, TAC-59953, TAC-59954, WCAP-11047, NUDOCS 8601240234 | |
| Download: ML20137K818 (75) | |
Text
.-_ _ _ _ _ _ _. _ _ _
k WCAP-11047 WESTINGHOUSE PROPRIETARY CLASS 3 o
4 CUSTOMER DESIGNATED DISTRIBUTION i
J. M. FARLEY UNITS 1 AND 2 REACTOR VESSEL FLUENCE AND RT AWATIONS PTS T. Congedo T. Rens R. Turner M. Weaver S. Yanichko Work Performed for Alabama Power Company January 1986 i
APPROVED:
M APPROVED: I Am>.
T. A. Meyer, Manager F. L. Lau, Manager Stru:tural Materials Radiation and Systems and Reliability Technology Analysis APPROVED: f fV W C. W. Hirst, Manager Reactor Coolant System Components Licensing i
Although the information contained in this report is nonproprietary, no distribution shall be made ~outside Westinghouse or its Licensees without the l,
customer's approval.
l l
WESTINGHOUSE ELECTRIC C')RPORATIDN NUCLEAR ENERGY SYSTEMS P. O. 80X 355 PITTS8URGH, PENNSYLVANIA 15230 8601240234 e60120 PDR ADOCK 05000348
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4070e:1d/010986
TABLE OF CONTENTS PAGE l
i TABLE OF CONTENTS i
LIST OF TABLES 11 LIST OF FIGURES v
I.
INTRODUCTION 1
I.1 The Pressurized Thermal Shock Rule 1
1.2 The Calculation of RTPTS 3
II.
NEUTRON EXPOSURE EVALUATION 5
II.1 Method of Analysis 5
11.2 Fast Neutron Fluence Results 8
III.
MATERIAL PROPERTIES 35 111.1 Identification and Location of Beltline Region Materials 35 111.2 Definition and Source of Material Properties for All 35 Vessel Locations III.3 Summary of Plant-Specific Material Properties 36 IV.
DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE 42 REGION MATERIALS IV.1 Status of Reactor Vessel Integrity in Terms of RTPTS 42 versus Fluence Results IV.2 Discussion of Results 43 V.
CONCLUSIONS AND RECOMMENDATIONS 48 VI.
REFERENCES 50 VII.
APPENDICES A.
Power Distribution A-1 B.
Weld Chemistry B -1 C.
RTpys Values of Farley Units 1 and 2 Reactor Vessel C -1 BeTtline Region Materials 4070e:ld/010986 i
LIST OF TABLES Page 11.2-1 Farley Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 12 Pressure Vessel Inner Radius - 0* Azimuthal Angle 11.2-2 Farley Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 13 Pressure Vessel Inner Radius - 12' Azimuthal Angle 11.2-3 Farley Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 14 Pressure Vessel Inner Radius - 21' Azimuthal Angle 11.2-4 Farley Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 15 Pressure Vessel Inner Radius - 30' Azimuthal Angle 11.2-5 Farley Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 16 Pressure Vessel Inner Radius - 45' Azimuthal Angle II.2-6 Farley Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 17 17* Surveillance Capsule Center 11.2-7 Farley Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 18 20' Surveillance Capsule Center 11.2-8 Farley Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 19 Pressure Vessel Inner Radius - 0' Azimuthal Angle 11.2-9 Farley Unit 2 Fast Neutron (E>l.0 MeV) Exposure at the 20 Pressure Vessel Inner Radius - 12* Azimuthal Angle 11.2-10 Farley Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 21 Pressure Vessel Inner Radius 21' Azimuthal Angle II. 2-11 Farley Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 22 Pressure Vessel Inner Radius 30' Azimuthal Angle 11.2-12 Farley Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 23 Pressure Vessel Inner Radius 45* Azimuthal Angle 3
11.2-13 Farley Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 24 17* Surveillance Capsule Center 11.2-14 Farley Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 25 20' Surveillance Capsule Center 111.3-1 J. M. Farley Unit 1 Reactor Vessel Beltline Region 38 Material Properties 111.3-2 J. M. Farley Unit 2 Reactor Vessel Beltline Region 39 Material Properties 4070e:1d/010986 11
LIST OF TABLES (Continued)
Page IV.1 -1 RTPTS Values for J. M. Farley Unit 1 44 IV.1 -2 RTPTS Values for J. M. Farley Unit 2 45 A-1 Farley Unit 1 Core Power Distributions Used in the Fluence A-4 Analysis A-2 Farley Unit 2 Core Power Distributions Used in the Fluence A-5 Analysis B.1 -1 J. M. Farley Intermediate Shell Longitudinal Welds Chemistry B-2 From WOG Materials Data Base - Wire Heat 33A277 C.1 -1 RTpis Values for the J. M. Farley Unit 1 Reactor Vessel C-3 Beltline Region Materials at Various Fluences C.1-2 RTpis Values for the J. M. Farley Unit 1 Reactor Vessel C -5 Beltline Region Materials at Current Life (4.79 EFPY)
C.1-3 RTPTS Values for the J. M. Farley Unit 1 Reactor Vessel C-6 Beltline Region Materials at End of License (26.49 EFPY)
C.1 -4 RTpys Values for the J. M. Farley Unit 1 Reactor Vessel C -7 Beltline Region Materials at 32 EFPY C.2-1 RTPTS Values for the J. M. Farley Unit 2 Reactor Vessel C-8 Eeltline Region Materials at Various Fluences C.2-2 RTPTS Values for the J. M. Farley Unit 2 Reactor Vessel C -10 Beltline Region Materials at Current Life (3.20 EFPY)
C.2-3 RTPTS Values for the J. M. Farley Unit 2 Reactor Vessel C-I l Beltline Region Materials at End of License (24.90 EFPY)
C.2-4 RTPTS Values for the J. M. Farley Unit 2 Reactor Vessel C-12 Beltline Region Materials at 32 EFPY l
4070e:ld/010986 iii
l LIST OF FIGURES
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PAGE I I.1 -1 J. M. Farley Reactor Geometry 26 11.1-2 J. M. Farley Reactor Geometry - 15' Neutron Pad 27 II.1-3 J. M. Farley Reactor Geometry - 26' Neutron Pad 2B 11.2-1 Farley Unit 1 Maximum Fast Neutron (E>l.0 MeV) Fluence 29 at the Beltline Weld Locations as a Function of Full Power Operating Time II.2-2 Farley Unit 2 Maximum Fast Neutron (E>l.0 MeV) Fluence 30 i
at the Beltline Weld Locations as a Function of Full Power Operating Time 11.2-3 Farley Unit 1 Maximum Fast Neutron (E>1.0 Mev) Fluence 31 at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 11.2-4 Farley Unit 2 Maximum Fast Neutron (E>1.0 MeV) Fluence 32 at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 11.2-5 Farley Units 1 and 2 Relative Radial Distribution 33 of Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall II.2-6 Farley Units 1 and 2 Relative Axial Distribution of 34 Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 111.1-1 Identification and Location of Beltline Region Material 40 for the J. M. Farley Unit 1 Reactor Vessel III.1 -2 Identification and Location of Beltline Region Material 41 for the J. M. Farley Unit 2 Reactor Vessel IV.1 -1 J. M. Farley Unit 1 - RTpys Curves per PTS Pule Method [1]
46 IV.1 -2 J. M. Farley Unit 2 - RTpTS Curves per PTS Rule Method [1]
47 A-1 J. M. Farley Units 1 and 2 Core Description for Power A-3 Distribution Map
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4070e:11/010986 iv
i i
SECTION I INTRODUCTION i
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The purpose of this report is to submit the reference temperature for pressurized thermal shock (RTPTS) values for the J. M. Farley Units 1 and 2 reactor vessels to address the Pressurized Thermal Shock (PTS) Rule.Section I discusses the Rule and provides the methodology for calculating RT PTS
- 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 for both units.Section IV provides the RT calculations from present through.the projected PTS end-of-license fluence values.
.l 1.1 THE PRESSURIZE 0 THERMAL SH0CK RULE The Pressurized Thermal Shock (PTS) Rule [1] was approved by the U.S. Nuclear Regulatory Comissioners 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.
1 The Rule outlines regulations to address the potential for PTS events on pressurized water reactor (PWR) vessels in nuclear power plants that are operated with a license f rom the United States Nuclear Regulatory Commission (USNRC). PTS events have been shown f rom 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 j
irradiation. Such an event may produce the propagation of flaws postulated to i
exist near the inner wall surface, thereby potentially af fecting the integrity of the vessel.
l 4070e:1d/010886 1
The Rule establishes the following requirements for all domestic, operating PWRs:
Establishes the RTPTS (measure of fracture resistance) Screening Criterion for the reactor vessel beltline region 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 RTpis values (per the prescribed methodology) and projected RTPTS values at the expiration date of the operating license. The date that this submittal must be received by the NRC for plants with operating licenses is January 23, 1986.
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 i
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 j
alternatives to minimize the PTS concern.
Requires NRC approval for operation beyond the Screening Criterion.
For applicants of operating licenses, values of the projected RT are to PTS be provided in the Final Safety Analysis Report. This requirement is added as part of 10CFR Part 50.34.
In the Rule, the NRC provides guidance regarding the calculation of the toughness state of the reactor vessel materials - the " reference temperature j
for nil ductility transition" (RTNOT). For purposes of the Rule, RT 5
NOT 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 RTPTS **
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 measure-ments, or other information indicate a significant change in projected 4070e:ld/010986 2
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 RT values for J. M. Farley Units 1 and 2.
PTS I.2 THE CALCULATION OF RTPTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RT at a given time.
p73 The prescribed equations in the PTS rule for calculating RT are actually PTS one of several ways to calculate RT purpose of compadson wM NDT.
the Screening Criterion, the value of RT f r the reactor vessel must be PTS calculated for each weld and plate, or forging in the beltline region as given below. For each material, RT is the lower of the results given by PTS Equations 1 and 2.
Equation 1:
PTS = I + M + [-10 + 470(Cu) + 350(Cu)(Ni)] f.270 0
RT Equation 2:
= 1 + M + 283 f PTS 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 RTf40T, copper and nickel content, fluence, and calculation procedures.
In 4070e:1d/010886 3
.---w
t 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.
1 Cu and Ni = the best estimate weight percent of copper and nickel in the material.
i t
I f = the maximum neutron fluence, in units of 10 n/cm2 (I greater than or I9 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 j
fluence for the period of service in question.
h I
Note, since the chemistry values given in equations 1 and 2 are best estimate mean values, and the margin, M, causes the RT values to be upper bound PTS 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 t
values used in the RT calculations for J. M. Farley Units 1 and 2 are j
PTS discussed in Section 111.2.
l l
l r
f i
i i
I f
i i
t t
l l
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4070e:Id/010886 4
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i i
i SECTION II 1
j NEUTRON EXPOSURE EVALUATION This section presents the results of the application of Westinghouse derived
+
j adjoint importance functions to the calculation of the J. M. Farley Units'l and 2 reactor vessel fluence for Alabama Power Company. The use of adjoint i
importance functions provides a cost effective tool to assess the effects that past and present core management strategies have had on neutron fluence levels l
in the reactor vessel.
I i
j 11.1 METHOD OF ANALYSIS j
A plan view of the J. M. Farley reactor geometry at the core midplane is shown j
in Figure 11.1-1.
In general, the reactor may be described with octant j
symmetry. Three octants, however, have elongated neutron pads with dual surveillance capsule holders attached. Figure 11.1-2 shows a zero-to,
]
degree sector with a 15-degree neutron pad segment included. This is the I
geometry for which the reactor vessel neutron flux adjoint functions, to be i,
j discussed shortly, have been developed. Figure 11.1-3 shows a zero-to j l
degree sector with a 26-degree neutron pad segment and a dual surveillance l
capsule holder. This is the geometry for which surveillance capsule evaluations are performed.
i 4
Two sets of neutron transport calculations were carried out in performing the fast neutron exposure evaluations for the reactor geometries shown in Figures j
11.1-2 and 11.1-3.
The first, a single computation in the conventional
)
forward mode, was utilized to provide baseline data derived from a design basis core power distribution against which cycle by cycle plant specific calculations can be compared. The second set of calculations consisted of a l
series of adjoint analyses relating the response of interest (neutron flux >
l 1.0 MeV) at several selected locations within the reactor geometry to the power distributions in the reactor core. These adjoint importance functions when combined with cycle specific core power distributions yield the plant l*
specific exposure data for each operating fuel cycle.
l i
I 4070e:ld/010986 5
,,-.~%---
s+--w
-w,wv.wr, y--w---
,,--www.,--yy,r*
y,-~
-m,,+-.-ww.--v
---m----,-g--+--v
,w.--wm,..-
--w-m -
,-v~---,
y.-.--r--,
4 h
1 4
The forward transport calculation was carried out in R,e geometry using the DOT discrete ordinates code [2] and the SAILOR cross-section library [3]. The
~
.i SAILOR library is a 47 group, ENDF/B-IV based data set produced specifically
]
for light water reactor applications. Anisotropic scattering is treated with I
aP expansion of the cross-sections. An S angular quadrature was used.
3 6
The design basis core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 3-loop plants. Inherent in the development of this design basis core power j
distribution is the use of an out-in fuel management strategy; i.e., f resh
]
fuel on the core periphery.
Furthermore, for the peripheral fuel assemblies, l
a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal
+2a level for a large number of fuel cycles, the use of this design basis i
j distribution is expected to yield somewhat conservative results. This is j
especially true in cases where low leakage fuel management has been employed.
The design basis core power distribution data used in the analysis is provided in Appendix A of this report. The data listed in Appendix A represent cycle averaged relative assembly powers.
The adjoint analyses were also carried out using the P cross-section 3
approximation from the SAILOR library. Adjoint source locations were chosen i
at the center of each of the surveillance capsules as well as at positions i
along the inner radius of the pressure vessel. Again, these calculations were l
run in R,9 geometry to provide power distribution importance functions for the exposure parameter of interest (neutron flux > 1.0 MeV). Having the adjoint importance functions and appropriate core power distributions, the response of interest is calculated as:
R
- I I I I (R,0,E) F (R,0,E) dE R dR de R,9 R e E 4070e:ld/010886 6
i
. -. _ -. ~. - -
i.
i where:
j.-
i l,
R, Response of interest (+ (E > 1.0 MeV)) at radius R and
=
azimuthal angle e.
Adjoint importance function at radius R and azimuthal I (R,e,E)
=
angle e for neutron energy group E.
Full power fission density at radius R and azimuthal angle l
F (R,e,E)
=
j e for neutron energy group E.
i j
The fission density distributions used reflect the burnup-dependent inventory j
of fissioning actinides, including U-235, U-238, Pu-239, and Pu-241.
i Core power distributions for use in the plant specific fluence evaluations for Farley Units 1 and 2 were derived from core design documentation as described l
in Appendix A of this report. The specific power distribution data used in l
the analysis is provided in Appendix A of this report. The data listed in Appendix A represents cycle averaged relative assembly powers. Therefore, the adjoint results were in terms of fuel cycle averaged neutron flux wnich when f
multiplied by the fuel cycle length yields the incremental fast neutron
{
fluence.
J The projection of reactor vessel fast neutron fluence into the future to the expiration date of the operating license requires that a few key assumptions 1
{
be made. Current neutron fluences, based on past core loadings, are defined i
l as of June 30, 1985. The operating license for the Farley Units expires on j
August 16, 2012 (forty years after the construction permit was issued). This l
report includes fluence projections from June 30, 1985 to August 16, 2012 using the cycle-averaged core power distribution of the current operating cycle (Cycle 7 for Unit 1 and Cycle 4 for Unit 2) and an assumed future f*
capacity factor of 80%. All fluence projections into the future reflect the j
low leakage fuel management strategy exemplified by these core loadings.
)*
Finally, it has been assumed that the Farley cores will continue to operate at 2652 MWth*
1 3
i e
4070e:ld/010886 7
i 4.
.m
The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oakridge National Laboratory (ORNL) Poolside Critical Assembly (PCA) facility as well as against the Westinghouse power reactor surveillance capsule data base [4). The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be conservative by 7-22%. When plant specific power distributions are used with the adjoint importance functions, the benchmarking studies show that fluence predictions are within i 15% of measured values at surveillance capsule locations.
11.2 FAST hEUTRON FLUENCE RESULTS Calculated fast neutron (E >1.0 MeV) exposure results for Farley Units 1 and 2 are presented in Tables II.2-1 through 11.2-14 and in Figures 11.2-1 through II.2-6.
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.
In Tables 11.2-1 through 11.2-5 cycle-specific maximum neutron flux and fluence levels at 0*,
12*, 21*, 30', and 45' on the pressure vessel inner radius of Farley Unit I are listed for the period of operation up to June 30, 1985, and projected to the expiration date of the operating license, as well as to 32.0 EFPY, representing 40 years of operation at a capacity factor of l
80%. Also presented are the design basis fluence levels predicted using the generic 3-loop core power distribution at the nominal + 2a level. Similar data for the center of surveillance capsules located at 17' and 20' are given in Tables 11.2-6 and 11.2-7, respectively.
l In addition to the calculated data given for the surveillance capsule j
locations, measured fluence data f rom previously withdrawn surveillance l
capsules are also presented for comparison with analytical results.
In the case of Unit 1, capsules were removed from the 17' location at the end of cycles 1 and 4.
To date, none have been removed from the 20* location.
4070e:ld/010986 8
4 4
Cycle-specific and design basis fast neutron flux and fluence data at the
~
inner radius of the pressure vessel of Farley Unit 2 are given in Tables 11.2-8 through II.2-12 for the period of operation up to June 30, 1985, and projected to the expiration date of the operating license, and 32.0 EFPY, as explained above. As in the case of Unit 1, data are presented for the 0*,
12*, 21*, 30*, and 45' azimuthal angles. Evaluations of plant specific and design basis fluence levels at the two surveillance capsule locations are given in Tables 11.2-13 and 11.2-14.
For Unit 2, a surveillance capsule was removed from the 17* position following cycle 1.
A dosimetry evaluation from this capsule witidrawal is listed in Table 11.2-13.
Several observations regarding the data presented in Tables 11.2-1 through 11.2-14 are worthy of note. These observations may be summarized as follows:
1.
For both Farley units, calculated plant specific fast neutron (E > 1.0 MeV) fluence levels at the surveillance capsule center are in excellent agreement with measured data. The maximum difference between the plant specific calculations and the measurements is approximately 11%.
Differences of this magnitude are well within the uncertainty of the experimental results.
2.
For both Farley units, the f ast neutron (E > 1.0 MeV) flux incident on the pressure vessel during Cycle 1 was, on the average, 8 to 9% less than predictions based on the design basis core power distributions. This result is consistent with the statement that the design basis power distributions produce flux levels that tend to be conservative by 7-22%.
3.
The low leakage fuel management employed during cycle 7 of Farley Unit 1, i
which is used for projection into the future, has reduced the peak fast neutron flux (0* azimuthal position) on the pressure vessel by a factor of 1.31 relative to the design basis flux.
(In subsequent discussions,
~
factors of fast neutron flux reduction, defined as the ratio of the design 4070e:1d/010986 9
1
basis flux to the cycle-specific flux, will be quoted.) The cycle 7 core loading produced flux reduction factors ranging from 1.32 to 1.35 at the other azimuthal locations.
4.
In Farley Unit 2, the low leakage core loading used for projection into the future (cycle 4) yielded a flux reduction factor of 1.32 at the peak flux location and factors ranging from 1.32 to 1.37 at the remaining azimuths.
Graphical presentations of the plant specific fast neutron fluence at key locations on the pressure vessel are shown in Figures 11.2-1 and 11.2-2 as a function of full power operating time for Farley Units 1 and 2, respectively.
For both Units 1 and 2, pressure vessel data is presented for the O' location on the circumferential weld as well as for the 45' longitudinal welds (see Section III.1).
In regard to Figure 11.2-1 and 11.2-2, the solid portions of the fluence curves are based directly on the cycle-specific core loadings as of. lune 30, 1985. The dashed portions of these curves, however, involve a projection into the future. As mentioned in Section 11.1, the neutron flux average over cycle 7 of Farley Unit I and Cycle 4 of Farley Unit 2, respectively, were used to project future fluence levels.
It should be noted that implementation of a more severe low leakage pattern than that of Cycle 7 for Unit 1 or Cycle 4 for Unit 2 would act to reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those projections. The RT assessment must be updated per PTS 10CFR50.61(b)(1) whenever, among other things, changes in core loadings significantly impact the fluence and RT projections.
~
In Figures 11.2-3 and 11.2-4, the azimuthal variation of maximum fast neutron (E > 1.0 MeV) fluence at the inner radius of the pressure vessel is presented
(
as a function of azimuthal angle for Units 1 and 2, respectively. Data are presented for both current and projected expiration-of-operating-license 1
4070e:Id/010886 10 l
conditions.
In Figure 11.2-5, the relative radial variation of fast neutron
~
flux and fluence within the pressure vessel wall is presented. Similar data showing the relative axial variation of fast neutron flux and fluence over the beltline region of the pressure vessel is shown in Figure 11.2-6.
A i
three-dimensional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figure II.2-3 through 11.2-6 along with the relation
$(R, e,Z) = +(e) F(R) G(Z)
Fast neutron fluence at location R, e, Z within where: + (R,e,Z)
=
the pressure vessel wall Fast neutron fluence at azimuthal location e on
+ (e)
=
the pressure vessel inner radius f rom Figure 11.2-3 or 11.2-4 Relative fast neutron flux at depth R into the F (R)
=
pressure vessel from Figure 11.2-5 Relative fast neutron flux at axial position Z from G (Z)
=
Figure 11.2-6 4
Analysis has shown that the radial and axial variations within the vessel wall are relatively insensitive to 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.
i 4070e:Id/010986 11 I
TABLE 11.2-1 FARLEY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSU_RE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE 8eltline Region 4
2 Elapsed Cumulative Fluence (n/cm )
Desigg Irradiation Irradiation Avg.2 Flux Plant Basis,}
Interval Time (EFPY)
(n/cm -sec)
SDecific 8
i CY-1 1.05 5.87 X 10 1.94 x 10 2.12 x 10 CY-2 1.81 6.62 x 10 3.53 x 10 3.66 x 10 10 18 18 CV-3 2,19 5.79 x 10 4.23 x 10 4.44 x 10 I
CY-4 2.98 6.06 x 10 5.74 x 10 6.04 x 10 10 18 18 CY-5 3.81 5.00 x 10 7.04 x 10 7.71 x 10 0
18 18 CY-6 4.72 4.35 x 10 8.30 x 10 9.56 x 10 CY-7 (to 6/30/85)(b) 4.79 4.89 x 10 8.41 x 10 9.71 x 10 10 18 18 6/30/85 4 End of License "I 26.49 4.89 x 10 4.19 x 10 5.37 x 10 I
10 I9 End of LicenseICI + 32.0 EFPY 32.0 4.89 x 10 5.04 x 10" 6.48 x 10 10 (a) Design basis fast neutron flux = 6.42 x 1010 n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which the current neutron fluxes are defined.
(c) License expires on 8/16/2012.
e l
4070e:1d/010886 12 1
l
TABLE II.2-2 FARLEY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 12' AZIMUTHAL ANGLE 8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant DesiggI Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CV-1 1.05 3.93 X 10 1.30 x 10 1.45 x 10 CY-2 1.81 4.47 x 10 2.37 x 10 2.49 x 10 10 18 18 CY-3 2.19 3.97 x 10 2.85 x 10 3.02 x 10 4
18 CY-4 2.98 4.20 x 10 3.90 x 10 4.11 x 10 10 18 18 CY-5 3.81 3.28 x 10 4.75 x 10 5.25 x 10 I
8 8
CY-6 4.72 2.98 x 10 5.61 x 10 6.51 x 10 10 18 18 CY-7 (to 6/30/85)IDI 4.79 3.24 x 10 5.68 x 10 6.61 x 10 6/30/85 4 End of License #
26.49 3.JA x 10 2.79 x 10 3.65 x 10" 10 End of License (c) + 32.0 EFPY 32.0 3.24 x 10 3.35 x 10 "
4.41 x 10" 10 (a) Design basis fast neutron flux = 4.37 x 1010 n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
4070e:1d/010886 13
TABLE 11.2-3 FARLEY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 21* AZIMUTHAL ANGLE 8eltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.05 3.29 X 10 1.09 x 10 1.23 x 10 8
18 CY-2 1.81 3.78 x 10 1.99 x 10 2.12 x 10 10 18 18 CY-3 2.19 3.42 x 10 2.41 x 10 2.57 x 10 10 I
CY-4 2.98 3.62 x 10 3.31 x 10 3.50 x 10 10 18 18 CY-5 3.81 2.85 x 10 4.05 x 10 4.47 x 10 CY-6 4.72 2.60 x 10 4.80 x 10 5.54 x 10 10 18 18 CY-7 (to 6/30/85)(b) 4.79 2.82 x 10 4.86 x 10 5.62 x 10 6/30/85 + End of License (c) 26.49 2.82 x 10 2.42 x 10" 3.11 x 10 End of License (c) + 32.0 EFPY 32.0 2.82 x 10 2.91 x 10 "
3.76 x 10 "
10 (a) Design basis fast neutron flux = 3.72 x 1010 n/cm2-sec at 2652 MWth.
(b) 6/30/35 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
i 1
i
\\
4070e:Id/010806 14 l
l l
- L
^
TABLE 11.2-4 FARLEY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE 8eltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg. Flux Plant Desigga)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 II I
CY-1 1.05 2'.55 X 10 8.43 x 10 9.63 x 10 10 18 CY-2 1.81 2.98 x 10 1.55 x 10 1.66 x 10 10 18 18 CY-3 2.19 2.69 x 10 1.88 x 10 2.01 x 10 0
18 18 CY-4 2.98 2.81 x 10 2.58 x 10 2.74 x 10 10 18 18 CY-5 3.81 2.09 x 10 3.12 x 10 3.49 x 10 10 18 18
~
CY-6 4.72 1.97 x 10 3.69 x 10 4.33 x 10 CY-7 (to 6/30/85)I I 4.79 2.15 x 10 3.74 x 10 4.40 x 10 10 18 18 I
10 I
I9 6/30/85 4 End of License '
25.49 2.15 x 10 1.85 x 10 2.43 x 10 End of LicenseIC) + 32.0 EFPY 32.0 2.15 x 10 2.22 x 10 2.94 x 10 10 I9 I9 e
(a) Design basis fast neutron flux = 2.91 x 1010 n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
i f
e
]
4070e:1d/010886 15
TA8LE II.2-5 FARLEY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45* AZIMUTHAL ANGLE 8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desig Interval Time (EFPY)
(n/cm -sec)
Specific Basis a) 10 II II CY-1 1.05 1.74 x 10 5.76 x 10 6.55 x 10 10 18 18 CY-2 1.81 2.06 x 10 1.07 x 10 1.13 x 10 10 18 18 CY-3 2.19 1.87 x 10 1.30 x 10 1.37 x 10 0
18 18 CY-4 2.98 1.87 x 10 1.76 x 10 1.86 x 10 10 18 18 CY-5 3.81 1.33 x 10 2.11 x 10 2.38 x 10 10 8
CY-6 4.72 1.38 x 10 2.51 x 10 2.95 x 10 10 18 18 CY-7 (to 6/30/85)I I 4.79 1.49 x 10 2.54 x 10 2.99 x 10 6/30/85 + End of License (#
26.49 1.49 x 10 1.27 x 10 1.66 x 10 0
I9 I9 End of License (c) + 32.0 EFPY 32.0 1.49 x 10 1.53 x 10 2.00 x 10 (a) Design basis fast neutron flux = 1.98 x 1010 n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
4070e:1d/010886 16
TABLE 11.2-6 FARLEY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 17' SURVEILLANCE CAPSULE CENTER 8eltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Cesigg)
Capsg Interval Time (EFPY)
(n/cm -sec)
Specific Basis Data I
18 18 CY-1 1.05 1.80 x 10' 5.96 x 10 6.65 x 10 5.83 x 10 CY-2 1.81 2.06 x 10" 1.09 x 10 1.15 x 10I I
CY-3 2.19 1.85 x 10 1.31 x 10 '
1.39 x 10 '
I9 I9 I9 CY-4 2.98 1.96 x 10" 1.80 x 10 1.89 x 10 1.65 x 10 I
I CY-5 3.81 1.55 x 10" 2.21 x 10 2.41 x 10 I9 I9 CY-6 4.72 1.41 x 10' 2.61 x 10 2.99 x 10 I9 l
4.79 1.53 x 10' 2.64 x 10 3.04 x 10 '
CY-7 (to 6/30/85)(b)
ICI 20 20 6/30/85 + End of License 26.49 1.53 x 10" 1.31 x 10 1.68 x 10 20 20 End of LicenseICI + 32,0 EFPY 32.0 1.53 x 10" 1.58 x 10 2.03 x 10 ll 2
(a) Design basis fast neutron flux = 2.01 x 10 n/cm -sec at 2652 MWth.
(b) 6/30/85 is the date at which the current neutron fluxes are defined.
(c) License expires on 8/16/2012.
(d) Data f rom p. 7-1 of Reference 5.
4070e:1d/010886 17
l TABLE 11.2-7 FARLEY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 20' SURVEILLANCE CAPSULE CENTER
- 4..
Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigga)
Capsg Interval Time (EFPY)
(n/cm -sec)
Specific Basis Data CY-1 1.05 1.54 x 10' 5.10 x 10 5.76 x 10 18 18 CY-2 1.81 1.77 x 10' 9.34 x 10 9.92 x 10 18 18 CY-3 2.19 1.60 x 10' 1.13 x 10 1.20 x 10 '
II l
CY-4 2.98 1.70 x 10" 1.55 x 10" 1.64 x 10 CY-5 3.81 1.34 x 10' 1.90 x 10 2.09 x 10 CY-6 4.72 1.22 x 10' 2.25 x 10 2.59 x 10" CY-7 (to 6/30/85)I )
4.79 1.33 x 10" 2.28 x 10 2.63 x 10" 6/30/85 -+ End of License '
26.49 1.33 x 10" 1.14 x 10 1.45 x 10 I
0 20 IC 20 20 End of License
-+ 32.0 EFPY 32.0 1.33 x 10" 1.37 x 10 1.76 x 10 (a) Design basis fast neutron flux = 1.74 x 1011 n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which the cure nt neutron fluxes are defined.
(c) License expires on 8/16/2012.
(d) To date, no capsules have been withdrawn from this position.
4070e:1d/010886 18
TABLE 11.2-8 FARLEY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXFOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg,)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.09 5.87 x 10 2.03 x 10 2.22 x 10 18 CY-2 1.86 6.57 x 10 3.63 x 10 3.78 x 10 10 18 18 CV-3 2.95 5.54 x 10 5.53 x 10 5.98 x 10 CY-4 (to 6/30/85)(
0 I
18 3.20 4.85 x 10 5.90 x 10 6.47 x 10 ICI 10 I9 I9 6/30/85 + End of License 24.90 4.85 x 10 3.91 x 10 5.04 x 10 0
End of License #
4 32.0 EFPY 32.0 4.85 x 10 5.00 x 10 6.48 x 10 (a) Design basis fast neutron flux = 6.42 x 1010 n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
4070e:1d/010886 19
TABLE 11.2-9 FARLEY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 12' AZIMUTHAL ANGLE 1
Beltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg. Flux Plant Desig a)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.09 3.93 x 10 1.36 x 10 1.51 x 10 10 18 18 CY-2 1.86 4.46 x 10 2.44 x 10 2.57 x 10 10 18 18 CY-3 2.95 3.63 x 10 3.69 x 10 4.07 x 10 10 18 18 CY-4 (to 6/30/85)ID) 3.20 3.26 x 10 3.94 x 10 4.41 x 10 10 I9 I9 6/30/85 + End of License (c) 24.90 3.26 x 10 2.62 x 10 3.43 x 10 10 I9 I9 End of License (c) + 32.0 EFPY 32.0 3.26 x 10 3.35 x 10 4.41 x 10 (a) Design basis fast neutron flux = 4.37 x 1010 2
n/cm -sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
4070e:1d/010886 20
TABLE II.2-10 FARLEY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 21* AZIMUTHAL ANGLE Beltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg,)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis 10 18 18 CY-1 1.09 3-30 x 10 1.14 x 10 1.28 x 10 I
CY-2 1.86 3.78 x 10 2.06 x 10 2.19 x 10 10 18 18 CY-3 2.95 2.98 x 10 3.09 x 10 3.47 x 10 CY-4 (to 6/30/85)(
3.20 2.81 x 10 3.30 x 10 3.75 x 10 6 00/85 + End of License (c) 24.90 2.81 x 10 2.25 x 10 2.92 x 10 10 E:
af License ('
+ 32.0 EFPY 32.0 2.81 x 10 2.88 x 10 3.76 x 10 I
(a) Design basis fast neutron flux = 3.72 x 1010 2
n/cm -sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
4070e:1d/010886 21
TABLE II.2-11 FARLEY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIIIS - 30' AZIMUTHAL ANGLE 8eltline Region 2
Elapsed Cumulative Fluence (n/cm )
l Irradiation Irradiation Avg.2 Flux Plant Desigga)
Interval Time (EFPY)
(n/cm -sec)
Specific Basis I
18 CY-1 1.09 2.57 x 10 8.85 x 10 1.00 x 10 10 18 I
CY-2 1.86 2.98 x 10 1.61 x 10
- 1.71 x 10 10 18 18 CV-3 2.95 2.09 x 10 2.33 x 10 2.71 x 10 10 18 18 CY-4 (to 6/30/85)I I 3.20 2.12 x 10 2.49 x 10 2.93 x 10 IC) 10 I9 I9 6/30/85 + End of License 24.90 2.12 x 10 1.70 x 10 2.29 x 10 10 I9 I
End of License (c) + 32.0 EFPY 32.0 2.12 x 10 2.17 x 10 2.94 x 10 10 2
(a) Design basis fast neutron flux = 2.91 x 10 n/cm -sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
s l
l l
l l
l 4070e:1d/010986 22
TABLE 11.2-12 FARLEY UNIT 2 o
FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigg,}
Interval Time (EFPY)
(n/cm -sec)
SDccific Basis 10 II I7 CY-1 1.09 1.76 x 10 6.08 x 10 6.83 x 10 10 18 18 CY-2 1.86 2.06 x 10 1.11 x 10 1.17 x 10 10 18 18 CY-3 2.95 1.27 x 10 1.55 x 10 1.85 x 10 0
18 CY-4 (to 6/30/85)( }
3.20 1.48 x 10 1.66 x 10 2.00 x 10 I) 10 I9 I9 6/30/85 4 End of License 24.90 1.48 x 10 1.18 x 10 1.56 x 10 End of LicenseI ) + 32.0 EFPY 32.0 1.48 x 10 1.51 x 10 2.00 x 10 (a) Design basis fast neutron flux = 1.98 x 1010 n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
4070e:1d/010886 23
TA8LE 11.2-13 FARLEY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 17* SURVEILLANCE CAPSULE CENTER 8eltline Region Elapsed Cumulative Fluence (n/cm )
Irradiation I, radiation Avg.2 Flux Plant Desig Capsg
)
Interval T t_-w ( E F PY)
(n/cm -sec)
Specific Basis Data l
CY-1 1.09 1.80 x 10" 6.23 x 10 6.94 x 10 5.6 x 10 18 18 CY-2 1.86 2.06 x 10" 1.12 x 10 1.18 x 10 CY-3 2.95 1.67 x 10" 1.70 x 10 1.87 x 10 CY-4 (to 6/30/85)I )
3.20 1.54 x 10" 1.82 x 10 2.03 x 10 6/30/85 -+ End of License 24.90 1.54 x 10" 1.24 x 10 1.58 x 10 II 20 20 End of LicenseIC) -+ 32.0 EFPY 32.0 1.54 x 10" 1.58 x 10 2.03 x 10 20 20 (a) Design basis fast neutron flux = 2.01 x 10ll n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
(d) Data from p. 7-1 of Reference 6.
4070e:1d/010886 24
TABLE 11.2-14 FARLEY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 20" SURVEILLANCE CAPSULE CENTER Beltline Region 2
Elapsed Cumulative Fluence (n/cm )
Irradiation Irradiation Avg.2 Flux Plant Desigga)
Capsg Interval Time (EFPY)
(n/cm -sec)
Specific Basis Data II 18 18 CY-1 1.09 1.55 x 10 5.34 x 10 6.00 x 10 II 18 IS CY-2 1.86 1.77 x 10 9.66 x 10 1.02 x 10 II I9 I9 CY-3 2.95 1.41 x 10 1.45 x 10 1.62 x 10 CY-4 (to 6/30/85)(b) 3.20 1.R x 10' 1.55 x 10 1.75 x 10 I9 l9 6/30/85 4 End of License (C) 24.90 1.32 x 10" 1.06 x 10 1.37 x 10 0
0 20 20 End of LicenseIC) + 32.0 EFPY 32.0 1.32 x 10" 1.35 x 10 1.76 x 10 (a) Design basis fast neutron flux = 1.74 x 10ll n/cm2-sec at 2652 MWth.
(b) 6/30/85 is the date at which current neutron fluxes are defined.
(c) License expires on 8/16/2012.
(d) To date, no capsules have been withdrawn f rom this position.
4070e:1d/010886 25
15873 7 RE ACTOR VESSEL 7
NEUTRGN PAD 7
i-a J
l -,
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+
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s I
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l Figure 11.1 -1.
J.M. Farley Reactor Geometry 26
16149 1 O'
NEUTRON PAD I
l 15' REACTOR VESSEL I
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J.M. Farley Reactor Geometry - 150 Neutron Pad 1-27
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16149 2 l
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NEUTRON PAD 16.94 DEG. (CAPSULES U,X,Y IN FARLEY 1 and U,X,V IN FARLEY 2) 0 0
19.72 DEG. (CAPSULES W,V,Z IN FARLEY 1 and g
W,Y,Z IN FARLEY 2) g REACTOR VESSEL k\\
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0 Figure 11.1 -3.
J.M. Farley Reactor Geometry - 26 Neutron Pad 28
16248 1 20 10 0
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LICENSE EXPIRATION
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O 10 20 30 40 50 60 4.79 26.49 32.00 JUNE 30,1985 EFPY Figure 11.2-1. Farley Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Beltline Weld Locations as a Function of Full Power Operating Time 29
16248-2 20 10 D*
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Farley Unit 2 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Beltline Wefd Locations as a Function of Full Power Operating Time l
30
16248 3 20 10 N
E 19 LICENSE EXPIRATION o
10
)
(26.49 EFPY)
Lu U
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Ey JUNE 30.1985 (4.79 EFPY)
W H
18 9
h 37 10 I
l l
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Figure 11.2 3.
Farley Unit 1 Maximum Fast Neutron (E >1.0 MeV) Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 31
16248-4 20 10 w.
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39 10 LICENSE EXPIRATION (24.90 EFPY) w Zw 3
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a WZ JUNE 30,1985 F
18 (3.20 EFPY) lO N
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17 10 I
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Figure 11.2-4. Farley Unit 2 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Pressure Vesselinner Radius as a Function of Azimuthal Angle l
32
O 16248-5 199.39 l
1.0
=-
204.39 p
IR I
I/4T 214.39 i
1 O*I 219.39 3/4T OR IIII IIII IIII IIII II'I
.01 200 205 210 215 220 225 Figure ll.2-5. Farley Units 1 and 2 Relative Radial Distribution of Fast Neutron (E> 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 33
.I i1 l
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SECTION III MATERIAL PROPERTIES For the RT calculation, the best estimate copper and nickel chemical PTS composition of the reactor vessel beltline material is necessary. The matec;al properties for the J. M. Farley Units 1 and 2 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 (shell material including welds, heat af fected 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."
Figures III.1-1 and 111.1-2 identify and indicate the location of all beltline region materials for the J. M. Farley Units 1 and 2 reactor vessels.
III.2 DEFINITION AND SOURCE OF MATERIAL PROPERTIES FOR ALL VESSEL LOCATIONS Material property values for the shell plates were derived f rom vessel fabrication test certificates.
Fast neutron irradiation-induced changes in the tension, f racture, and impact properties of reactor vessel materials are largely dependent on chemical 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.
To address the variation in weld materials the weld material prc;erties are obtained from the WOG Reactor Vessel Beltline Region Weld Metal Data Base.
The WOG data base, which was developed in 1984 and is continually being 4070e:ld/010986 35
updated, contains information from weld qualifications records, surveillance 1
~
capsule reports, the Materials Properties Council (MPC) and NRC Mender MATSURV data bases.
For each of the welds in the J. M. Farley Units 1 and 2 beltline region, a material data search was performed using the WOG data base. Searches were i
performed for materials having the identical weld wire heat number as those in j
the Farley vessels, 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.
q Although phosphorous and siliren are not needed for the PTS Rule, they are provided for the sake of completeness. Only one or two data points are available for the weld materials found in the Farley Vessels with exception of i
the Unit 1 intermediate shell longitudinal weld, wire heat 33A277. The information obtained for this weld is presented in Appendix B.
The material properties for plates are identical to those reported recently in an evaluation of heatup and cooldown curves (References 9 and 10). Weld properties are similar to values found in References 9 and 10, the dif ferences are due to additional data points used to calculate average material i
properties.
III.3
SUMMARY
OF PLANT-SPECIFIC MATERIAL PROPERTIES A sunnary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the J. M. Farley Units 1 and 2 reactor f
vessels are respectively given in Tables 111.3-1 and 111.3-2.
Although 2
phosphorus is no longer used in the calculation of RT with respect to the NOT PTS rule [1], it is given for reference since it is currently used in the Regulatory Guide 1.99,. Revision 1 trend curve [7].
]
When available actual initial values of RT were used. These values were l
NOT l
estimated according to Branch Position MTEB 5-2 [8]. Actual values were available for all the base metal and some of the weld materials. When actual l
initial RT values were not available for weld materials the generic mean MDT i
4070e:1d/010986 36' 1
I
value as defined in the PTS rule [1] was used. The data in Tables III.3-1 and 111.3-2 are used to evaluate the RT values for the Farley Unit 1 and 2 PTS reactor vessels.
T 4
~
b s
4 P
i J
e l
4070e:1d/010986 37
TABLE 111.3-1 J. M. FARLEY UNIT 1 REACTOR VESSEL BELTLINE REGION
~
MATERIAL PROPERTIES Cu Ni P
I(a)
(Wt.%)
(Wt.%)
(Wt.%)
(*F)
Intermediate Shell Plate B6903-2 0.13 0.60 0.011 0
Intermediate Shell Plate B6903-3 0.12 0.56 0.014 10 Lower Shell Plate B6919-1 0.14 0.55 0.015 15 Lower Shell Plate B6919-2 0.14 0.56 0.015 5
Intermediate Shell Longitudinal Weld Wire Heat - 33A277 0.251 0.21 0.017
-56(b)
Lower Shell Longitudinal Weld II Wire Heat 90099 0.17 0.20 0.022
-56 Intennediate to Lower Shell Circumferential Weld Wire I)
Heat - 6329637 0.225 0.20 0.011
-56 Notes:
(a) The initial RTNDT values for the plates are estimated according to l
Branch Technical Position MTEB 5-2 [B].
I (b) The initial RTyor values for the welds are the generic mean value defined by the PTS rule [1].
4070e:ld/010986 38 l
TABLE III.3-2 J. M. FARLEY UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P
I(a)
(Wt.%)
(Wt.%)
(Wt.%)
(*F)
Intermediate Shell Plate B7203-1 0.14 0.60 0.010 15 Intermediate Shell Plate 87212-1 0.20 0.60 0.018
-10 Lower Shell Plate B7210-1 0.13 0.56 0.010 18 Lower Shell Plate 87210-2 0.14 0.57 0.015 0
Intermediate Shell Longitudinal Weld Wire Heat - H0DA and BOLA 0.02 0.96 0.0095
-56(b)
Lower Shell Longitudinal Welds Wire Heat - 83640 0.05 0.20 0.006
-70 Intermediate to Lower Shell Circumferential Weld Wire Heat - SP5622 0.13 0.20 0.016
-40 i
j Notes:
(a) The initial RTNDT values for the plates and welds (with the exception of the intermediate shell longitudinal weld) are estimated according to Branch Technical Position MTEB 5-2 [8], see note (b).
(b) The initial RTNDT values for this weld is the generic mean value defined by the PTS rule [1].
4070e:1d/010986 39
FIGURE III.1-1
!OENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL I
FOR THE J.M. FARLEY UNIT 1 REACTOR VESSEL CIRCUMFERENTIAL SEAMS VERTICAL SEAMS 19-8948 86903-3
,e g_ 10-894 r-1 8.4" 45 '
CORE n
CORE 5
1h 1
L 144.0" 3
B6903-2 19-894A I
c
~
C f_
20.1" q
11-894 3
86919-2 20-8948 U
'45
.E CORE e
y u
l d
I 48.75" t-- a r
20-894A B6919-1 l
l 40 l
a FIGURE III.1-2 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE J.M. FARLEY UNIT 2 REACTOR VESSEL i
CIRCUMFERENTIAL SEAMS VERTICAL SEAMS l
l 19-923B 87212-1 q _ 10-923 r 7 45
- 8. "
CORE h
CORE 144.0" B7203-1 19-923A l
i
~
~
C f -
o 20-923B 87210-2 y
F-7
'45 di CORE p
b l
i 48.75" e
m 20-923A B7210-1 41
SECTION IV DETERMINATION OF RT VALES FOR AR BELUNE REGION MAMIALS PTS 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 values for J. M. Famy Unhs 1 and 2 can now be PTS determined.
IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF RT VERSUS FLUENCE PTS RESULTS Using the prescribed PTS Rule methodology, RT values were generated for PTS all beltline region materiais of the J. M. Farley Units 1 and 2 reactor vessels as a function of several fluence values and pertinent vessel lifetimes. The location of the maximum fluence on the Unit 1 and 2 reactor vessels is at the 0* azimuthal angle. The circumferential welds and baseplate materials use this fluence to determine RT values. The longitudinal pg welds are located at the 45* azimuthal angle. Therefore, RT values are PTS calculated using the corresponding fluence at these locations. The tabulated results f rom the total evaluation are presented in Appendix C for all beltline region materials for both units.
Figures IV.1-1 and IV.1-2 present the RT values for the limiting PTS longitudinal weld, circumferential weld and shell plate of the J. M. Farley Units 1 and 2 vessels in terms of RT PTS in these figures can be used:
0 to provide guidelines to evaluate fuel reload options in relation to the NRC RT Screening Criterion for PTS (i.e., RT va kes can be PTS PTS
- The EFPY can be determined using Figure 11.2-1 for Unit I and Figure 11.2-2 for Unit 2.
4070e:1d/010986 42
1
=
readily projected for any options under consideration, provided fluence is known), and o
to show the current (4.79 EFPY for Unit 1 and 3.20 EFPY for Unit 2), and end-of-license (26.49 EFPY for Unit 1 and 24.90 EFPY for Unit 2) RTPTS values using actual and projected fluence.
Table IV.1-1 and IV.1-2 provide a summary of the RT values for all PTS beltline region materials for the lifetime of interest.
IV.2 DISCUSSION OF RESULTS i
As shown in Figures IV.1-1 and IV.1-2, the shell plates are the governing locations for both reactor vessels relative to PTS. All the RT values PTS remain below the NRC screening values for PTS using the projected fluence values through the license expiration. The most limiting RT value at PTS license expiration is 185*F for the lower shell plate B6919-1 of Unit 1 and 220*F for the intermedfate shell plate 87212-1 of Unit 2.
I r
l I
ls 4070e:ld/010986 43 v.
TABLE IV.1-1 RT VALUES FOR J. M. FARLEY UNIT 1 PTS RT Values (*F)
PTS Present End-of-License Screening Location Vessel Material (4.79 EFPY)
(26.49 EFPY) 32 EFPY Criteria 1
Intermediate shell plate B6903-2 123 163 169 270 2
Intermediate shell plate B6903-3 125 161 166 270 3
Lower shell plate B6919-1 142 185 191 270 4
Lower shell plate B6919-2 132 176 182 270 5
Intermediate shell longitu-90 138 145 270 dinal welds 19-894A, 19-8948 i
6 Lower shell longitudinal welds 20-894A, 20-8948 60 90 95 270 7
Intennediate to lower shell 109 167 176 300 circumferential weld 11-894
)
4070e:1d/010986 44
TABLE IV.1-2 RT VALUES FOR J. M. FARLEY UNIT 2 PTS i
l RT Values ( F)
PTS Present End-of-License Screening Location Vessel Material (3.2 EFPY)
(24.9 EFPY) 32 EFPY Criteria 1
i l
1 Intermediate shell plate B7203-1 137 186 195 270 2
Intermediate shell plate B7212-1 147 220 223 270 3
Lower shell plate B7210-1 132 177 184 270 4
Lower shell plate B7210-2 121 169 177 270 l
5 Intermediate shell longitu-7 9
10 270 l
Lower shell longitudinal welds20-923
-12
-4
-3 270 7
Intermediate to lower shell 60 95 101 300 l
circumferential weld 11-923 l
l l
l 4070e:1d/010986 45 l..
o l
FIGURE IV.1-1 J.N.FARLEY UNIT 1 - RTPTS CURVES PER PTS RULE NETHOD [1]
3 00 - ]- - - - - - - - - -- - - - - - - - - - - - - - - - - - - ~ - - -
e NRC RT Screening Value - Circumferential Welds py3 280 l
7 _ _NRC RT 260 PTS Screening value - Plates and Longitudinal Weld
~
INTERNEDIATE SHELL LON6ITUDINAL WELD 200 gas m180 8-167 0
~
INTERNEDIATE TO LOWER A
M LOWER SHELL PLATE.
142 CIRCUNFERENTIAL WELD g140 -86919-1 iss l
9"-" 120 90 109 80 t
60 40 20 0
1018 1019 102C FLUENCE, NEUTRONS / CW2 y
e= CURRENT LIFE (4.79 EFPY) AND A =END OF LICENSE (26.49 EFPY)
RTPTS VALUES USING PLANT SPECIFIC AND PROJECTED PLANT SPECIFIC FLUENCE VALUES i
FIGURE IV.1-2 J.M.FARLEY UNIT 2 - RTPTS CURVES PER PTB RULE METHOD (13 300----_-________________.____________________._
Screening Value - Circumferential Welds PTS 280 Screening Value Plates and Longitudinal Weld 260 PTS 240 220 220 I
INTERMEDIATE SHELL PLATE 87212-1 200 m180 ta.
Ov160 147 N
M g140 INTERMEDIATE TO LOWER i
e CIRCUMFERENTIAL WELD E 120 100 80 60 60 40 INTERMEDIATE SHELL LON61TUDINAL NELD 20 7
i i
e i
n i e i 1 9,
i e
i i i i I 1018 1019 102C 2
FLUENCE, NEUTRONS / CW l
e= CURRENT LIFE (3.2 EFPY) AND A=END OF LICENSE (24.99 EFPY)
RTPTS VALUES USING PLANT SPECIFIC AND PROJECTED PLANT SPECIFIC FLUENCE VALUES
_._,.,r...._
SECTION V
~
CONCLUSIONS AND RECOMMENDATIONS Calculations have been completed in order to submit RT values for the PTS J. M. Farley Units 1 and 2 reactor vessels 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 values.
PTS Detailed fast neutron exposure evaluations using plant specific cycle by cycle core power distributions and state-of-the-art neutron transport methodology have been completed for the Farley Units 1 and 2 reactor vessels. Explicit calculations were performed for all operating cycles of both units through June 30, 1985. For both units, projection of the fast neutron exposure beyond June 30, 1985 was based on continued implementation of low leakage fuel management similar to that employed at that time in each unit.
In regard to the low leakage fuel management already in place at the Farley Units, the plant specific evaluations have demonstrated that for the cycle core loadings as of June 30, 1985 the peak fast neutron flux at the 0*
azimuthal position has been reduced by a factor of 1.31 in Unit 1 and a factor of 1.32 in Unit 2 relative to the flux based on the design basis core power distribution.
This location represents the maximum fast neutron flux incident on the reactor pressure vessel. At other locations on the vessel, as well as at the surveillance capsules, the impact of low leakage will differ from the data presented above.
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 tend to reduce the projection. On the other hand, a relaxation of the loading pattern toward higher relative power 4070e:1d/010986 48 a
i t
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 projc<tions made in this report.
i t
The f ast neutron fluence values f rom the plant specific calculations have been j
compared directly with measured fluence levels derived from neutron dosimetry contained in surveillance capsules withdrawn from each of the Farley Units.
3 For Unit 1 the ratio of calculated to measured fluence values ranges from 1.02 to 1.09 for the two capsule data points. The corresponding ratio for Unit 2 is 1.11 for the capsule removed from that reactor. This excellent 1
agreement between calculation and measurement supports the use of this l
analytical approach to perform plant specific evaluations for the Farley reactors.
Material property values for the Farley Units 1 i.nd 2 reactor vessel beltline region components were determined. The pertinent chemical and mechanical I
properties for the shell plates remain the same as those that have been reported in the original vessel fabrication test certificates. The weld i
material properties are obtained from the WOG Material Data Base.
Using the prescribed PTS Rule methodology RT values were generated for PTS l
all beltline region materials of the J. M. Farley Units 1 and 2 reactor vessels as a function of several fluence values and pertinent vessel lifetimes. For both reactor vessels, all the RT values remain below the PTS l
NRC screening values for PTS using the projected fluence exposure through the l
expiration date of the operating license. The most limiting values at end-of-license (26.49 EFPY for Farley Unit 1 and 24.90 EFPY for Farley Unit 2) are 185'F and 220*F for the lower shell plate B6919-1 of Unit 1 and the l
intermediate shell plate 87212-1 of Unit 2, respectively.
The results in this report are provided to enable Alabama Power Company to l
comply with the initial 6 months submittal requirements of the USNRC PTS Rule.
4070e:1d/010986 49
~
SECTION VI REFERENCES 1.
Nuclear Regulatory Commission,10CFR Part 50, " Analysis of Potential Pressurized Thermal 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, 31screte Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5 August 1970.
3.
" SAILOR RSIC Data Library Collection DLC-7b." Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P, Cross-Section Library for Light Water 3
Reactors.
4.
Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -
to be published.
5.
" Analysis of Capsule U f rom the Alabama Power Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program", WCAP-10474, R. S.
Boggs et al, Westinghouse Electric Corporation, February 1984.
6.
" Analysis of Capsule U f rom the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program", WCAP-10425, M. K. Kunka et al, Westinghouse Electric Corporation, October 1983.
7.
" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Regulatory Guide 1.99 - Revision 1. U.S. Nuclear Regulatory Commission, Washington, April 1977.
4 8.
NUREG-0800 - U.S. NRC Standard Review Plan, Branch Technical Position 5-2, Revision 1, July 1981.
4070e:ld/010986 50
9.
"Heatup and Cooldown Limit Curves for the Alabama Power Company Joseph M.
Farley Unit 1 Reactor Vessel", WCAP-10934, W. T. Kaiser et al, September 1985.
- 10. "Heatup and Cooldown Limit Curves for the Alabama Power Company Joseph M.
Farley Unit 2 Reactor Vessel", WCAP-10910, W. T. Kaiser et al, September 1985.
- 11. Radclif fe, R. and Holmes, R., " Revised Nuclear Design Data for Cycle 2 of Farley Unit 1", Westinghouse Nuclear Fuels Division, NDS-79-096, April 25, 1979.
i I
r 1
J I
l l
4070e:1d/011086 51 l
l
APPENDIX A POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the J. M. Farley Unit 1 and 2 reactor vessels were derived f rom core design data as explained below. A schematic diagram of the core configuration applicable to the J. M. Farley plants is shown in Figure A-1.
Cycle averaged relative assembly powers for each operating fuel cycle of J. M. Farley Unit 1 are listed in Table A-1; similar data for J. M. Farley Unit 2 are listed in Table A-2.
On Figure A-1 and in Tables A-1 and A-2 an identification number is assigned to each fuel assembly location; and three regions consisting of subsets of fuel assemblies are defined.
In performing the adjoint evaluations, the relative power in assemblies comprising Region 3 has been adjusted to account for known biases in the analytical or design prediction of power in the peripheral assemblies while the relative power in 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 fuel assemblies comprising Region 1 do not contribute significantly to the neutron exposure either at the surveillance capsules or at the pressure vessel. Therefore, power distribution data for assemblies in Region 1 are not listed in Table A-1.
Cycle averaged core powr ' distributions for Farley Unit I were developed using c
the following Westinghouse fuel cycle design reports for each operating cycle to date:
Fuel Cycle Report 1
WCAP-8515 and Ref. 11
~
2 Ref.11 and WCAP-9761 3
WCAP-9761 and WCAP-10036 4A WCAP-10036 and WCAP-10308 5
WCAD-10308 and WCAP-10525 6
WCAP-10525 and WCAP-10795 7
WCAP-10795 and WCAP-10795 Rev.1 4070e:1d/011086 A -1
Of these, Cycles 1 through 4A utilized out-in fuel loading patterns, and Cycles 5, 6 and 7 implemented low leakage fuel loading patterns.
Cycle averaged core power distributions for Farley Unit 2 were developed using the following Westinghouse fuel cycle design reports for each operating cycle to date:
Fuel Cycle Report 1
WCAP-10674 Of these, Cycles 1 and 2 utilized out-in fuel loading patterns, and Cycles 3 and 4 implemented low leakage fuel loading patterns.
The power distributions employed represent cycle averaged relative assembly powers. Therefore, the adjoint results are in tenms of fuel cycle averaged neutron flux, which when multiplied by the fuel cycle length yields the incremental fast neutron fluence.
In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly 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.
4 l
l l
l 1
I 4070e:ld/011086 A-2 i
16003-26 O*
(MAJOR AXIS)
BAFFLE
//
7 CORE BARREL xNNN I
2
\\
45o xNNNN.'
6 7
3 4
\\
i 8
9 10 5
/
II I2 Figure A 1.
Farley Units 1 & 2 Core Description for Powe-Distribution Map A3
TABLE A-1 FARLEY UNIT 1 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant SDecific Cycle Averaged Relative Assembiv Power Design Fuel Cycle Basis Assembly Relative Power 1
2 3
4 5
6 7
t 1
0.93 0.810 0.891 0.767 0.773 0.742 0.598 0.718 2
0.77 0.642 0.731 0.639 0.685 0.438 0.425 0.447 3
1.12 0.924 1.025 0.953 1.009 0.958 0.844 0.921 4
0.80 0.640 0.747 0.664 0.716 0.453 0.401 0.462 5
0.85 0.684 0.809 0.721 0.721 0.443 0.471 0.517 6
0.95 1.023 0.970 0.945 1.140 1.150 0.886 0.965 7
1.07 1.021 1.195 1.145 1.095 1.135 1.051 1.119 8
0.97 1.103 0.970 1.205 1.095 1.140 1.070 1.071 9
1.02 1.038 0.980 0.980 1.120 1.225 1.234 1.191 10 1.04 0.971 1.185 1.130 1.080 1.075 1.084 1.042 11 1.03 1.038 1.000 1.225 1.155 1.215 1.239 1.201 12 0.92 0.885 1.040 0.920 0.825 0.765 1.019 1.090 r
~
i 4070e:1d/010886 A -4
TABLE A-2 FARLEY UNIT 2 CORE POWER DISTRIBUTION USED IN THE FLUENCE ANALYSIS Plant SDecific Cycle Averaged Relative Assembiv Power Design Fuel Cycle Basis Assembiv Relative Power 1
2 3
4 1
0.93 0.809 0.876 0.793 0.685 2
0.77 0.644 0.731 0.536 0.461 3
1.12 0.927 1.025 0.979 0.934 4
0.80 0.644 0.752
.0.464 0.431 5
0.85 0.695 0.803 0.412 0.512 6
0.95 1.015 0.970 1.090 0.952 7
1.07 1.015 1.195 1.140 1.091 8
0.97 1.100 0.975 1.160 1.111 9
1.02 1.035 0.980 1.215 1.230 10 1.04 0.975 1.180 1.070 1.085 11 1.03 1.040 0.995 1.180 1.207 12 0.92 0.895 1.040 0.740 1.101 l
l l
1 i
4070e:1d/010986 A-5 l
1
APPENDIX B WELD CHEMISTRY Table B.1-1 provides the weld data output f rom the WOG Material Data Base for wire heat 33A277. The pertinent material chemical compositions are given, along with 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 RT analysis.
PTS Weld Chemistry Data Source and Plant:
ALA J. M. Farley Unit 1 j
CCl Calvert Cliffs 1 CE Combustion Engineering Cu Weight % of Copper MPC DB Materials Properties Council Data Base Ni Weight % of Nickel P
Weight % of Phosphorous SC Surveillance Capsule Si Weight % of Silicon WQ Weld Qualification i
i 4070e:ld/010886 8 -1
TABLE B.1-1 J.M.FARLEY UNIT 1 INTERMEDIATE SHELL LONGITUDINAL WELDS l
CHEMISTRY FROM WOG MATERIALS DATA BASE - WIRE HEAT 33A277 SELECT REPORT un n-..u u-u uu no..u unun u n u un.n u s on.u... un n = = u.u.= ===.nu.n uu====.u= = ==========
ID WIRE WIRE FLUI FLUI WELDCHER Cu ut P
Si PLMT DESCRIPi!ON HEAT TYPE TYPE LOT DATA SOURCE
~
0002 33A277 8-4 LINDE 124 3878 CE,WQ 0.320 0.016 0.330 PES L0dR SHELL LONS 0078 33A277 B-4 LINDE 0091 3922 CE,hD 0.300 0.013 0.180 ALA SURVEILLANCE WELD CCI INTER TO LOWER SHELL CCI SURVEILLANCE WELD CC2 N0ZILE 10 INTER SHELL CE2 LOWER SHELL LONG ML2 N0llLE TO INTER SHELL RY INTER 10 LDER SHELL 0000 33A277 S-4 LINDE 0091 3977 CE,WG 0.230 0.017 0.180 0193 33A277 B-4 LINDE 1092 3669 CE,WQ 0.320 4.015 CE 0.320 9.015 CE,hQ 0194 33A277 8-4 LINCE 60 B651 W
0.270 0.015 0.140 ALA INTER SHELL LON6 0197 33A277 B-4 LINDE 1092 3889 ALA,Q 5C 0.140 0.190 0.016 0.270 ALA SURVEILLANCE ELD 0292 33A277 B-4 LlhDE 0091 3922 CCI INTER TO L0hER SHELL ED CCl SURVEILL MCE WELD E
CC2 N0ZILE 10 INTER SHELL CC2 LOWER SHELL LON6 ML2 C ILE 10 INTER SHELL MY INTER TO LOWER SHELL 0317 33A277 E-4 LINDE 0091 3922 CE,WD 0.230 0.013 0.210 ALA SURVEILLANCE WELD CCl INTER 10 LOWER SHELL CCI SURVEILLANCE WELD CC2 N0llLE TO INTER SHELL CC2 LDER SHELL LONG RL2 N0llLE T0 INTER SHELL MY INTER 10 LOWER SHELL 0666 33A277 6-4 LINDE 0091 3922 RPC,DE CCI,SC 0.290 0.180 0.014 0.200 ALA SURVEILLMCE ELD CCl INTER 10 LOWR SHELL CCI SURVEILLANCE WELD CC2 N0ZZLE 10 INTER SHELL CC2 LGER SHELL LONG ML2 N0llLE 10 INTER SHELL MV INTER 10 LOWER SHELL 0732 33A277 B-4 LINDE 0091 3922 CC 0.140 0.270 0.040 0.320 ALA.
SURVEILLMCE WELO CCl INTER TO L0hER SHELL CCI SURVEILLMCE HELD CC2 N0lILE 10 INTER SHELL CC2 LOWER SHELL LON6 ML2 N0ZILE 10 INTER SHELL MV INTER TO LOWER SHELL 0.251000 0.213333 0.017400 0.228750 sean std.dev.
0.068872 0.049329 0.000044 0.069703
- - - u = = =.u n.
u = = u...n u n = =. = u = = u.. n.u..u.. u.. n o n n o n. u = = = = n.u. =. n j
I
APPENDIX C RT VALUES OF J. M. FARLEY UNITS 1 AND 2 PTS REACTOR VESSEL BELTLINE REGION MATEkIALS C.1 FARLEY UNIT 1 Tables C.1-1 through C.1-5 provide the RT values, as a function of both PTS constant fluence and constant EFPY (assuming the projected fluences values),
for all beltline region materials of the J. M. Farley Unit I 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 111.3-1 of the main report.
Location Vessel Material 1
Intermediate shell plate B6903-2 2
Intermediate shell plate B6903-3 3
Lower shell plate B6919-1 4
Lower shell plate B6919-2 5
Intermediate shells longitudinal welds 19-894A, 19-8948 6
Lower shells longitudinal welds 20-894A, 20-894B 7
Intermediate to lower shell circumferential weld 11-894 C.2 FARLEY UNIT 2 Tables C.2-1 through C.2-5 provide the RT values, as a function of both PTS constant fluence and constant EFPY (assuming the projected fluence values),
for all beltline region materials of the J. M. Farley Unit 2 reactor vessel.
The RI values are calculated in accordance with the PTS rule, which is PTS j
Reference [1] in the main body of this report. The vessel location numbers i
4070e:ld/010886 C-1
~-
Ps
. - =
in the following tables correspond to the vessel materials identified below and in Table 111.3-2 of the main report.
Location Vessel Material 1
Intermediate shell plate B7203-1 2
Intermediate shell plate B7212-1 3
Lower shell plate 87210-1 4
Lower shell plate 87210-2 5
Intermediate shell longitudinal welds 4
19-923 6
Lower shell longitudinal welds20-923 7
Intermediate to lower shell circumferential weld 11-923 1
4 e
- l l
l 4070e:ld/010886 C-2
.-- - _ _ ~
o TABLE C.1-1 RTPTS VALUES FOR THE J.M.FARLEY UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS AT VARIOUS FLUENCES
!!D ! PLANT! CU ! NI
! P !
! VALUE
! TYPE !
RTPTS VALUE AT FLUENCE
.10E+19
.50E+19
.10E+20
.20E+20 1
1 ALA 0.130 0.600 0.011 0
ACTUAL B.M.
90 113 126 143 2
ALA 0.120 0.560 0.014 10 ACTUAL B.M.
96 116 128 142 3
ALA 0.140 0.550 0.015 15 ACTUAL B.M.
107 132 146 163 4
ALA 0.140 0.560 0.015 5
ACTUAL B. !1.
98 122 136 153 5
ALA 0.251 0.210 0.017
-56 GENERIC L.W.
71 108 129 155 6
ALA 0.170 0.200 0.022
-56 GENERIC L.W.
47 71 85 102 7
ALA 0.225 0.200 0.011
-56 GENERIC C.W.
63 95 115 137 Notes:
ID
= Location of vessel maternal (tsee page c-4)
I
= Initial value of RTNDT, ac tuesl or tant i mat ed Value=" ACTUAL" denotes that t.h e 11sted values of RTNDT are actual rather than " GENERIC" values.
B.M.
= Base Metal L.W.
= Longitudinal Weld C.W.
=Circumferential Weld Reference temperatures are in dey F C-3
TABLE C.1-1'(CONT)
RTPTS VALUES FOR THE J.M.FARLEY UNIT 1 REACTOR VESSEL BELTLINE RE6103 MATERIALS AT VARIOUS FLUENCES i
l i
i i
!!D ! PLANT! CU ! NI
! P !
! VALUE
! TVPE !
RTPTS VALUE AT FLUENCE
.40E+20
.60E+20
.70E+20 1
APR 0.140 0.600 0.010 15 ACTUAL B.M.
187 201 207 g
2 APR 0.200 0.600 0.01B
-10 ACTUAL B.M.
221 242 251 3
APR 0.130 0.560 0.010 18 ACTUAL B.M.
177 190 196 4
APR 0.140 0.570 0.015 0
ACTUAL B.M.
170 184 190 5
APR 0.'020 0.960 0.010
-56 GENERIC L.W.
12 13 13 6
APR 0.050 0.200 0.006
-70 ACTUAL L.W.
3 6
7 7
APR 0.130 0.200 0.016
-40 ACTUAL C.W.
96 106 110 2
h j
C-4 O
O O
0 0
TABLE C.1-2 RTPTS VALUES FOR THE J.M.FARLEY UNIT 1 REACTOR VESSEL BELTLINE P"910N MATERIALS AT CURRENT LIFE (4.29 EFPY)
!!D ! PLANT! CU ! NI
! P !
! VALUE
! TYPE ! FLUENCE ! RTPTS !
1 ALA 0.130 0.600 0.011 0
ACTUAL B.M.
0.84E+19 123 2
ALA 0.120 0.560 0.014 10 ACTUAL B.M.
0.94E+19 125 3
ALA 0.140 0.550 0.015 15 ACTUAL B.M.
0.84E+19 142 4
ALA 0.140 0.560 0.015 5
ACTUAL B.M.
0.84E+19 132 5
ALA 0.251 0.210 0.017
-56 GENERIC L.W.
0.25E+19 90 6
ALA 0.170 0.200 0.022
-56 GENERIC L.W.
0.25E+19 60 7
ALA 0.225 0.200 0.011
-56 GENERIC C.W.
0.84E+19 109 t
C-5
TABLE C.1-3 RTPTS VALUES FOR THE J.M.FARLEY LNiT 1 REACTOR VESSEL SELTLINC REGION MATERIALS AT END OF LICENSE (26.49 EFPYi 4
4 t
!!D ! PLANT! CU ! MI
! P !
! VALUE
! TYPE ! FLUENCE ! RTPTS !
1 ALA 0.130 0.600 0.011 0
ACTUAL B.M.
0.42E+20 163 2
ALA 0.120 0.560 J.014 10 ACTUAL B.M.
0.42E+20 161 3
ALA 0.140 0.550 0.015 15 ACTUAL B.M.
0.42E+20 185 4
ALA 0.140 0.560 0.015 5
ACTUAL B.M.
0.42E+20 176 5
,ALA 0.251 0.210 0.017
-56 GENERIC L.W.
0.13E+20 138 6
ALA 0.170 0.200 0.022
-56 GENERIC-L.W.
0.13E+20 90 7
ALA 0.225 0.200 0.0!!
-56 GENERIC C.W.
0.42E+20 167 I
l 1
I 7
C-6 l
TABLE C.1-4 RTPTS VALUES FOR THE J.M.FARLEY UNIT 1 REACTOR I
VESSEL BELTLINE REGION MATERIALS AT 32 EFPY t
!!D ! PLANT! CU ! NI
! P !
! VALUE
! TYPE ! FLUENCE ! RTPTS !
1 ALA 0.130 0.600 0.011 0
ACTUAL B.M.
0.50E+20 169 2
ALA 0.120 0.560 0.014 10 ACTUAL B.M.
0.50E+20 166 3
ALA 0.140 0.550 0.015 15 ACTUAL 8.M.
0.50E+20.
191 4
ALA 0.140 0.560 0.015 5
ACTUAL B.M.
0.50E+20 182
,5 ALA 0.251 0.210 0.017
-56 GENERIC L.W.
0.15E+20 145 6
ALA 0.170 0.200 0.022
-56 GENERIC L.W.
0.15E+20 95 7
ALA 0.225 0.200 0.011
-56 GENERIC C.W.
0.50E+20 176 C-7
table C.2-1 RTPTS VALUES FOR THE J.M.FARLEY UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS AT VARIOUS FLUENCES
!!D ! PLANT! CU ! NI
! P !
! VALUE
! TYPE !
RTPTS VALUE AT FLUENCE l
.10E+19.50E+19
.10E+20.20E+20 1
APR 0.140 0.600 0.010 15 ACTUAL B.M.
109 134 148 166 2
APR 0.200 0.600 0.018
-10 ACTUAL B.M.
106 142 164 190 3
APR 0.130 0.560 0.010 18 ACTUAL B.M.
107 130 143 158 4
APR 0.14,0 0.570 0.015 0
ACTUAL B.M.
93 117 132 149 5
APR 0.020 0.960 0.010
-56 GENERIC L.W.
6 8
9 10 6
APR 0.050 0.200 0.006
-70 ACTUAL L.N.
-13
-8
-5
-2 7
APR 0.130 0.200 0.016
-40 ACTUAL C.W.
40 58 68 81 C-8
_ - _ _. ~. ~ _. _. -. -.. _ _. -. -.
-. - - ~ _ -
TABLE C.2-1 (CONT)
RTPTS VALUES FOR THE J.M.FARLEY UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS AT VARIOUS FLUENCES i
i
!!D ! PLANT! CU !
NI
! P !
! VALUE
! TYPE !
RTPTS VALUE AT FLUENCE
.40E+20
.60E*20
.70E+20 1
ALA 0.130 0.600 0.011 0
ACTUAL B.M.
162 175 181 2
ALA.0.120 0.560 0.014 10 ACTUAL B.M.
16 r.
171 176 3
ALA 0.140 0.550 0.015 15 ACTUAL B.M.
103 197 203 4
ALA 0.140 0.560 0.015 5
ACTUAL B.M.
174 188 194 5
ALA 0.251 0.210 0.017
-56 GENERIC L.W.
187 208 217 6
ALA 0.170 0.200 0.022
-56 GENERIC L.W.
122 136 141 7
ALA 0.225 0.200 0.011
-56 6ENERIC C.W.
165 184 192 4
b 4
4 l
C-9
TABLE C.2-2 RTPTS VALUES FOR THE J.M.FARLEY U%IT 2 REACTOR VESSEL BELTLINE REGION MATERIALS AT CURRENT LIFE (3.2 EFPY)
!!D ! PLANT! CU ! NI
! P !
I
! VALUE
! TYPE ! FLUENCE ! RTPTS !
1 APR 0.140 0.600 0.010 15 ACTUAL B.M.
0.59E+19 137 2
APR 0.200 0.600 0.018
-10 ACTUAL B.M.
0.59E+19 147 3
APR 0.130 0.560 0.010 18 ACTUAL B.M.
0.59E+19 132 4
APR 0.140 0.570 0.015 0
ACTUAL B.M.
0.59E+19 121 5
APR 0.020 0.960 0.010
-56 GENERIC L.W.
0.17E+19 7
6 APR 0.050 0.'200 0.006
-70 ACTUAL L.,W.
0.17E+19
-12 7
AFR 0.130 0.200 0.016
-40 ACTUAL C.W.
0.59E+19 60 C-10
=
TABL E C.2-3 RTPTS VALUES FOR THE J.M.FARLEY UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS AT END OF LICENSE (24.90 EFPYi
!!D ! PLANT! CU ! NI
! P !
! VALUE
! TYPE ! FLUENCE ! RTPTS !
1 APR 0.140 0.600 0.010 15 ACTUAL B.M.
0.39E+20 186 2
APR 0.200 0.600 0.018
-10 ACTUAL B.M.
0.39E+20 220 3
APR 0.130 0.560 0.010 18 ACTUAL B.M.
0.39E+20 177 4
APR 0.140 0.570 0.015 0
ACTUAL B.M.
0.39E+20 169 5
APR 0.020 0.960 0.010
-56 6ENERIC L.W.
0.12E+20 9
6 APR 0.050 0.200 0.006
-70 ACTUAL L.W.
0.12E+20
-4 7
APR 0.130 0.200 0.016
-40 ACTUAL C.W.
0.39E+20 95 J
f C-11
. 'I.
I TABLE C.2-4 RTPTS VALUES FOR THE J.M.FARLEY UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS AT 32 EFPY
!!D ! PLANT! CU ! NI
! P !
! VALUE
! TYPE ! FLUENCE ! RTPTS !
1 APR 0.140 0.600 0.010 15 ACTUAL B.M.
0.50E+20 195 2
APR 0.200 0.600 0.018
-10 ACTUAL B.M.
0.50E+20 233 3
APR 0.130 0.560 0.010 18 ACTUAL B.M.
0.50E+20 184 4
APR 0.140 0.570 0.015 0
ACTUAL B.M.
0.50E+20 177
,5 APR 0.020 0.960 0.010
-56 GENERIC L.W.
0.15E+20 10 6
APR 0.050 0.200 0.006
-70 ACTUAL L.W.
0.15E+20
-3 7
APR 0.130 0.200 0.016
-40 ACTUAL C.W.
0.50E+20 101 C-12 2
I
'