ML20140A682
| ML20140A682 | |
| Person / Time | |
|---|---|
| Site: | Zion, 05000000 |
| Issue date: | 12/31/1985 |
| From: | Shaun Anderson, Balkey K, Furchi E, Perone V, Weaver M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20140A658 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR 1297:1D-122085, 1297E:1D-122085, WCAP-10962, NUDOCS 8601230308 | |
| Download: ML20140A682 (95) | |
Text
.
WESTINGHOUSE PROPRIETARY CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION WCAP-10962 ZION UNITS 1 AND 2 REACTOR VESSEL FLUENCE AND RT EVALUATIONS PTS E. L. Furchi V. A. Perone S. L. Anderson M. A. Weaver a
K. R. Balkey Work Performed f or Commonwealth Edison Company December 1985 APPROVED:
M. b A4 APPROVED:
A4A-T. A. Meyer, Wanager F. L. Lau, Manager Structural Materials Radiation and Systems and Reliability Technology Analysis APPROVED: [k C. W. Hirsf, Manager Reactor Coolant System B601230308 860117g5 -
Components Licensing DR ADOCK O Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval.
WESTINGHOUSE ELECTRIC CORPORATION NbCLEAR ENERGY SYSTEMS P. O. 80X 355 PITTSBURGH, PENNSYLVANIA 15230
TABLE OF CONTENTS PAGE TABLE OF CONTENTS i
LIST OF TABLES 11 LIST OF FIGURES v
I.
INTRODUCTION 1
1.
The Pressurized Thermal Shock Rule 1
3 2.
The Calculation of RTPTS II.
NEUTRON EXPOSURE EVALUATION 5
1.
Method of Analysis 5
2.
Fast Neutron Fluence Results 7
III.
MATERIAL PROPERTIES 31 1.
Identification and Location of Beltline Region Materials 31 2.
Definition and Source of Meterial Properties for All 31 Vessel Location 3.
Material Chemistry Study for Weld Wire Heat 72105 33 3.1 Statistical Evaluation 33 3.1.1 Statistical Analysis Results 33 3.1.2 Filler Wire Examination 35 3.1.3 Impact of Flux Lot for Other Filler Wires 35 3.2 Calculation of Adjusted RTNOT 36 3.3 Summary of Material Chemistry Study for Weld 37 Heat Number 72105 4.
Summary of Plant-Specific Material Properties 37 IV.
DETERMINATION OF RTpis VALUES FOR ALL BELTLINE 45 REGION MATERIALS 1.
Status of Reactor Vessel Integrity in Terms of RTPTS 45 versus Fluence Results 2.
Discussion of Results 46 V.
CONCLUSIONS AND RECOMMENDATIONS 51 VI.
REFERENCES 54 VII.
APPENDICES A.
Power Distribution A-1 B.
Weld Chemistry B-1 C.
RTpyS Values of Zion Units 1 and 2 Reactor Vessel C-1 Beltline Region Materials i
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LIST OF TABLES P,a28 11.2-1 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 12 Inner Radius - 0* Azimuthal Angle - Zion Unit 1 11.2-2 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 13 Inner Radius - 15* Azimuthal Angle - Zion Unit 1 11.2-3 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 14 Inner Radius - 30' Azimuthal Angle - Zion Unit 1 11.2-4 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 15 Inner Radius - 45' Azimuthal Angle - Zion Unit 1 l
11.2-5 Fast Neutron (E>1.0 MeV) Exposure at the 4* Surveillance 16 Capsule Center - Zion Unit 1 11.2-6 Fast Neutron (E>1.0 MeV) Exposure at the 40' surveillance 17 Capsule Center - Zion Unit 1 11.2-7 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 18 Inner Radius 0* Azimuthal Angle - Zion Unit 2 11.2-8 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 19 Inner Radius 15' Azimuthal Angle - Zion Unit 2 11.2-9 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 20 Inner Radius 30' Azimuthal Angle - Zion Unit 2 11.2-10 Fast Neutron (E>1.0 MeV) Exposure at the Pressure Vessel 21 Inner Radius 45' Azimuthal Angle - Zion Unit 2 11.2-11 Fast Neutron (E>1.0 MeV) Exposure at the 4* Surveillance 22 Capsule Center - Zion Unit 2 11.2-12 Fast Neutron (E>1.0 MeV) Exposure at the 40' Surveillance 23 Capsule Center - Zion Unit 2 III.3-1 Summary of Results from Westinghouse Statistical Materials 34 Study for Heat 72105 III.3-2 Summary of Archive Data f rom BAW-1799 Appendix B 38 III.3-3 Chemistry Factor (CF) for Welds per Proposed Reg. Guide 1.99 39 111.4-1 Zion Unit 1 Reactor Vessel Beltline Region Material Properties 40 III.4-2 Zion Unit 2 Reactor Vessel Beltline Region Material Properties 41 ii 1297E:10/122085
- _ _. ~. - _.
LIST OF TABLES (Continuid) fA21 IV. 3-1 Zion Unit 1 RTPTS Values 47 IV.3-2 Zion Unit 2 RTpys Values 48 A.1 -1 Core Power Distributions Used in the Plant Specific A-4 Fluence Analysis - Zion Unit 1 A.1 -4 Core Power Distributions Used in the Plant Specific A-5 Fluence Analysis - Zion Unit 2 i
8.1 -1 Zion Unit 1 Intermediate and Lower Shell Longitudinal B-2 Welds Chemistry From WOG Materials Data Base -
l Wire Heat Number 8T1762 1
8.1 -2 Zion Unit 2 Intermediate Shell Longitudinal Weld Chemistry From 8-3 i
WOG Material Data Base - Wire Heat Number 72102 8.1-3 Zion Unit 2 Beltline Circumferential Weld Chemistry From 8-4 WOG Materials Data Base - Wire Heat Number 71249 l
8.1 -4 Zion Unit 1 Beltline Circumferential Weld and Zion Unit 2 B-7 Loar Shell Longitudinal Weld Chemistry From WOG Materials Data Base - Wire Heat Number 72105 i
C.1 -1 RTpys Values for Zion Unit 1 Reactor Vessel Beltline C-2 Region Materials 9 Fluence = 1.0 x 1018 n/cm2 C-3 C.1 -2 RTPTS Values for Zion Unit 1 Reactor Vessel Beltline Reg 1on Materials 9 Fluence = 5.0 x 1018 n/caz l
C.1-3 RTPTS Values for Zion Unit 1 Reactor Vessel feltline C-4 Region Materials 9 Fluence = 1.0 x 10l9 n/cmd 1
j C.1 -4 RTpis Values for Zion Unit 1 Reactor Vessel Beltline C-5 j
Region Materials 9 End of License (25.8 EFPY) -
Projected Fluence Value C.1 -5 RTPTS Values for Zion Unit 1 Reactor Vessel Beltline C-6 Reg 1on Materials 9 32 EFPY - Projected Fluence Values C. 2-1 RTPTS Values for Zion Unit 2 Reactor Vessel Beltline C -7 l;
Region Materials 9 Fluence = 1.0 x 1018 n/cm2 C.2-2 RTpys Values for Zion Unit 2 Reactor Vessel Beltline C-8 Region Materials 9 Fluence = 5.0 x 1018 n/cm2 C.2-3 RTpys Values for Zion Unit 2 Reactor Vessel Beltline C-9 I
Region Materials 9 Fluence = 1.0 x 1019 n/cm2 i
l' iii 1297E:lD/122085
LIST OF TA8LES (Continued)
East C.2-4 RTPTS Values for Zion Unit.2 Reactor Vessel Beltline C-10 Region Materials 9 End of License (25.3 EFPY) -
Projected Fluence Value C.2-5 RTPTS Values for Zion Unit 2 Reactor Vessel Beltline C-11 Region Materials 9 32 EFPY - Projected Fluence Values 1
i l
i i
l I
1 i
l 4
iv 1297E:10/121985
~
LIST OF FIGURES PAGE 11.1-1 Zion Reactor Geometry 24 11.2-1 Maximum Fast Neutron (E>l.0 MeV) Fluence at the Beltline 25 Weld locations as a Function of Full Power Operating Time - Zion Unit 1 11.2-2 Maximum Fast Neutron (E>1.0 MeV) Fluence at the Beltline 26 Weld Locations as a Function of Full Power Operating Time - Zion Unit 2 11.2-3 Maximum Current and Projected EOL Fast Neutron (E>l.0 MeV) 27 Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle - Zion Unit i 11.2-4 Maximum Current and Projected EOL Fast Neutron (E>l.0 MeV) 28 Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle - Zion Unit 2 11.2-5 Relative Radial Distribution of Fast Neutron (E>1.0 MeV) Flux 29 and Fluence Within the Pressure Vessel Wall Zion Units 1 and 2 11.2-6 Relative Axial Distribution of Fast Neutron (E>l.0 MeV) Flux 30 and Fluence Within the Pressure Vessel Wall Zion Units 1 and 2 111.1-1 Identifica'. ion and Location of Beltline Region Material 42 for the Zian Unit No. 1 Reactor Vessel l
III.1 -2 Identification and Location of Beltline Region Material 43 for the Zion Unit No. 2 Reactor Vessel 111.3-1 B&W Archive Data Flux 8773 and 8669 Comparison 44 1
IV. 4-1 Zion Unit 1 - RTPTS Curves per PTS Rule Methodology 49 IV.4-2 Zion Unit 2 - RTpys Curves per PTS Rule Methodology 50 A.1 -1 Zion Units 1 and 2 Core Description for Power A-3 Distribution Maps l
v 1297E:10/122085
SECTION I INTRODUCTION The purpose of this report is to determine the reference temperature for pressurized thermal shock (RTPTS) values for the Zion 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 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 for both units.Section IV provides the RT calculations for the present, projected end-of-license, PTS and end-of-life fluence values.
I.1 THE PRESSURIZED THERMAL SHOCK RULE The Pressurized Thermal Shock (PTS) Rule [1] was approved by the U.S. Nuclear l
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 pressurized thermal shock (PTS) of 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 af fecting the integrity of the vessel.
The Rule establishes the following requirements for all domestic, operating PWRs:
1297E:l D/121985 1
~
Establishes the RTPTS (measure of f racture resistance) Screening Criterion for the reactor vessel beltline region J
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 PTS valurs (per the prescribed methodology) and projected RT values at PTS 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: Flants 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.
i Requires NRC approval for operation beyond the Screening Criterion.
For applicants of operating licenses, values of the projected RT am 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 for nil ductility transition" (RTNDT). 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 RT values from the time of PTS the submittal to the license expiration date. This assessment must be submitted within 6 months af ter the ef fective date of the Rule, on January 23, 1297E:lD/121985 2
1986, with updates shenever changes occur af fecting 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 RT values PTS for Zion Units 1 and 2.
I.2 THE CALCULATION OF RT PTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method f
for determining plant-specific values of RT at a given time.
PTS l
The prescribed equations in the PTS rule for calculating RT are actually PTS one of several ways to calculate RT For the purpose of comparison with 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 8
"5
- 9 PTS Equations 1 and 2.
Equation 1:
RTPTS = 1 +
+ [-10 + 470(Cu) + 350(Cu)(Ni)] f Equation 2:
=I+
+ 83 f PTS where i
j I = the initial reference transition temperature of the unirradiated material measured as defined in the ASME Code, N8-2331.
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 8-5 weld fluxes.
M = the margin to be added to cover uncertainties in the values of initial RTNOT, c pper 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 l
1297E:10/121985 3
generic mean value of I was used.
In Equation 2, M-0*F if a measured value of I was used, and M-24*F if the generic mean value of I was used.
Cu and Ni = the best estimate weight percent of copper and nickel in the
- material, i
f = the maximum neutron fluence, in units of 10 'n/cm2 (E greater than or I
i equal to 1 Mev), at the clad-base-metal interface on the inside surf ace of the t
vessel at the location where the material in question receives the highest j
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 RT values to be upper bound PTS i
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 RT calculations for Zion Units 1 and 2 are discussed PTS in Section III.
1 1
i j
1, 1
'1 f
1297E:10/121985 4
SECTION II NEUTRON EXPOSURE EVALUATION This section presents the results from the application of the Westinghouse derived adjoint flux program to the Zion Units 1 and 2 reactor vessels for Commonwealth Edison Company. The use of adjoint importance functions provides a cost effective tool to assess the effects that past and present core management strategies have had an neutron fluence levels in the reactor vessel. The use of adjoint importance functions provides a cost ef fective tool to assess the effects that past and present core nunagement strategies have had on neutron fluence levels in the reactor vessel. Both of the Zion plants have recently operated using low leakage core nunagement schemes.
i II.1 METHOD OF ANALYSIS A plan view of the Zion Units 1 and 2 reactor geometry at the core midplane is shown in Figure II.1-1.
Since the reactor exhibits 1/8th core symmetry only a 0*-45' sector is depicted. Eight irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program. Two capsules are located symmetrically in each quadrant at azimuthal positions of 4* and 40* f rom the reactor core cardinal axes as shown in Figure 11.1-1.
In performing the fast neutron exposure evaluations for the reactor geometry shown in Figure 11.1-1, two sets of transport calculations were carried out.
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 cor. pared. The second set of calculations consisted of series of adjoint analyses relating the response of interest (neutron flux > 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 specific exposure data for each operating fuel cycle, i
1297E:10/122085 5
l
The fcrward transport calculatien was carried cut in R,0 g:ometry using the 00T discrete ordinates code [2] and the SAILOR cross-section library (3). The SAILOR library is a 47 group, ENDF-BIV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with aP expansi n f the cross-sections.
3 The design basis core power distribution utilized in the forward analysis was derived from statistical studies of long-tens operation of Westinghouse 4-loop plants. Inherent in the development of this design basis core power 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, 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 distribution is expected to yield somewhat conservative results. This is especially true in cases where low leakage fuel management has been employed.
The adjoint analyses were also carried out using the P C' 55-5
3 approximation from the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positions i
along the inner diameter of the pressure vessel. Again, these calculations were run in R,e geometry to provide power distribution importance functions for the exposure parameters 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,,= f, f,I (R,e) F (R,9) R @ &
where:
Response of interest (+ (E > 1.0 MeV)) at radius R and R
=
g,,
azimuthal angle 9.
l l
l 1297E:10/122085 6
l
~w
I (R,e)
Adjoint importance function at radius R and azimuthal
=
angle e.
1 F (R,0)
Full power fission density at radius R and azimuthal angle
=
e.
It should be noted that as written in the above equation, the importance function I (R 9) represents an integral over the fission distribution so l
that the response of interest can be related directly to the spatial j
distribution of fission density within the reactor core.
Core power distributions for use in the plant specific fluence evaluations for Zion Units 1 and 2 were taken from the design of each operating cycle for the two reactors. The specific power distribution data used in 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 which when multiplied by the fuel cycle length yields the incremental fast neutron I
fluence.
1 The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oak Ridge National Laboratory (ORNL) Poolside Critical Assembly (PCA) f acility as well as against i
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.
II.2 FAST NFUTRON FLUENCE RESULTS Calculated fast neutron (E >l.0 MeV) exposure results for Zion Units I and 2 are presented in Tables II.2-1 through II.2-12 and in Figures 11.2-1 through i
11.2-6.
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.
I
{
1297E:10/121985 7
In Tables 11.2-1 through II.2-4 plant sp;cific maximum n:utron flux and i
fluence levels at 0*,15*, 30', and 45* on the pressure vessel inner radius 3
are listed for the first 7 completed fuel cycles of Zion Unit 1.
Also presented are the design basis fluence levels predicted using the generic 4-loop core power distribution at the nominal + 2a level. Similar data for the center of surveillance capsules located at 4' and 40' are given in Tables 11.2-5 and 11.2-6, respectively.
l In addition to the calculated data given for the surveillance capsule locations, measured fluence data from previously withdrawn surveillance capsules are also presented for comparison with analytical results.
In the case of Unit 1, capsules were removed from the 40' location at the end of cycles 1, 4 and 6.
i Cycle-specific and design basis fast neutron flux and fluence data at the l
inner radius of the pressure vessel are given in Tables 11.2-7 through 11.2-10 for the first 7 completed fuel cycles of Zion Unit 2.
As in the case of Unit 1, data are presented for the 0*,15*, 30', and 45' azimuthal angles.
Evaluations of cycle-specific and design basis fluence levels at the two surveillance capsule locations are given in Tables II.2-11 and 11.2-12.
For Unit 2, surveillance capsules were removed from the 40* position following
]
cycles I and 4.
Dosimetry evaluations from these two capsule withdrawals are I
also listed in Table 11.2-12.
1 Several observations regarding the data presented in Tables 11.2-1 through j
11.2-12 are worthy of note. These observations may be sumarized as follows:
)
1.
For both units, calculated cycle-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 cycle-specific calculations and the measurements is less than 7%.
1 Differences of this magnitude are well within the uncertainty of the experimental results.
I t
1297E:10/122085 8
~
2.
For Unit 1, low lcakag) fuel management introduced following cycle 6 has
, reduced the peak flux on the pressure vessel by about 35%.
3.
For Unit 2, low leakage fuel management introduced following cycle 5 has reduced the peak flux on the pressure vessel by about 25%. This reduction has been maintained over the last 2 operating cycles.
4.
For both of the Zion reactors the maximum neutron flux incident on the pressure vessel (45' azimuthal position) during the fuel cycles using out-in fuel management (cycles 1 through 6 for Unit 1 and 1 through 5 f or J
Unit 2) was, on the average, approximately 15% less than predictions based on the design basis core power distributions.
j 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 l
function of full power operating time for Zion Units 1 and 2, respectively.
l For both Units 1 and 2, pressure vessel data is presented for the 45' location on the circumferential weld as well as for the 0* longitudinal welds (see
{
Section III.1).
i i
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 evaluations presented in this report. The dashed portions of these curves, however, involve a projection into the future. Since both Zion Units are committed to a consistent form of j
low leakage fuel management, the average neutron flux at the key locations over the low leakage fuel cycles was used for all temporal projections.
In j
particular, the neutron flux average over cycle 7 was used to project future fluence levels for Unit 1, while the neutron flux average over cycles 6 and 7 was employed for Unit 2.
a The fluence projections in Figures 11.2-1 and 11.2-2 have been carried out to
]
32 effective full power years. However, since RT data corresponding to PTS
~
the license expiration date must be supplied to the NRC in response to the Pressurized Thermal Shock Rule (10CFR50.61(b)(1)), the fluences corresponding 4
1297E:lD/122085 9
I A
to the license expiraticn date aro indicated in Figurcs 11.2-1 and 11.2-2.
An 80% capacity factor was assumed for operation beyond Cycle 7 (including the j
seventh refueling outage) for each Zion Unit.
1 It should be noted that implementation of a more severe low leakage pattern j
j would act to reduce the projections of fluence at key locations. On the other i
hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those projections.
In any event the RTPTS assessment must be updated per 10CFR50.61(b)(1) whenever, among other things, f
changes in core loadings significantly impact the fluence and RTPTS projections.
j 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 i
as a function of azimuthal angle for Units 1 and 2, respectively. Data are I
presented for both current and projected end-of-life conditions.
In Figure i
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 i
axial variation of fast neutron flux and fluence over the beltline region of l
the pressure vessel is shown in Figure 11.2-6.
A three-dimensional l
{
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
>f the relation i
1 4
i
$(R, e,1) = +(e) F(R) G(Z) where: $ (R,e,1)
Fast neutron fluence at location R, e, Z within
=
j l
the pressure vessel wall
+ (e)
Fast neutron fluence at azimuthal location e on
=
the pressure vessel inner radius from Figure II.2-3 or 11.2-4 L
i 4
1 i
i 1297E:10/122085 10 i
t s -. _ _
F (R)
.=
Relative' fast neutron flux at depth R into the pressure vessel from Figure 11.2-5 Relative fast neutron flux at axial position Z from G (Z)
=
Figure II.2-6 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 i
L 1297E:lD/122085 11
TA8LE 11.2-1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE (a)
ZION UNIT 1 Elapsed Beltline Region 2
Irradiation Cycle Avg.
Cumulative Fluence (n/cm )
Cycle Time Flux Cycle Design No.
(EFPY)
(n/cm2-sec)
Specific Basis (b) 9 Il II 1
1.2 7.46 x 10 2.73 x 10 3.37 x 10 9
II II 2
2.1 8.46 x 10 5.13 x 10 5.99 x 10 9
I II 3
2.7 8.63 x 10 6.99 x 10 7.98 x 10 9
I 18 4
3.5 8.10 x 10 8.98 x 10 1.02 x 10 9
18 18 5
4.3 8.82 x 10 1.11 x 10 1.24 x 10 9
18 18 6
5.0 8.52 x 10 1.31 r 10 1.46 x 10 9
18 18 7
5.9 8.66 x 10 1.54 x 10 1.71 x 10 IC) 9 18 18 Cy 8 + EOL 25.8 8.66 x 10 6.98 x 10 7.50 x 10 9
I9 18 EOL 32.0 EFPY 32.0 8.66 x 10 8.68 x 10 9.31 x 10 a) Applicable to the longitudinal welds at 0*,
90*, 180*, 270* in the peak axial flux.
9 2
b) Design basis fast neutron flux = 9.22 v. 10 n/cm -sec at 3391 m th' c) Current neutron fluences are defined as of the beginning of Cycle 8 (February 9,1984).
The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26, 2008) using the Cycle 7 average flux and an 80% capacity factor.
1297E:10/122085 12
TABLE 11.2-2 FAST NEUTRON (E > 1.0 MeV1 EXPOSURE AT THE PRtSSURE VESSEL INNER RADIUS - 15* AZIMUTHAL ANGLE ZION UNIT 1 l
Elapsed 8eltline Region Irradiation Cycle Avg.
Cumulative Fluence (n/cm2)
Cycle Time Flux Cycle Design No.
(EFPY)
(n/cm2-sec)
Specific Basis (a) 10 lI I
1 1.2 1.16 x 10 4.82 x 10 5.27 x 10 10 I7 I
2 2.1 1.32 x 10 8.57 x 10 9.36 x 10 10 18 18 3
2.7 1.35 x 10 1.15 x 10 1.25 x 10 10 18 18 4
3.5 1.29 x 10 1.46 x 10 1.60 x 10 10 18 I9 5
4.3 1.45 x 10 1.81 x 10 1.94 x 10 10 18 18 6
5.0 1.41 x 10 2.15 x 10 2.29 x 10 10 18 18 7
5.9 1.19 x 10 2.47 x 10 2.67 x 10 Cy 8 + EOL(b) 25.8 1.19 x 10 9.95 x 10 1.17 x 10 10 I9 I9 10 I9 I9 EOL 32.0 EFPY 32.0 1.19 x 10 1.23 x 10 1.45 x 10 10 2
a) Design basis fast neutron flux = 1.44 x 10 n/cm -sec at 3391 MWth" b) Current neutron fluences are defined as of the beginning of Cycle 8 (February 9, 1984)
The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26, 2008) using the Cycle 7 average flux and an 80% capacity factor.
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TA8LE 11.2-3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30* AZIMUTHAL ANGLE ZION UNIT 1 Elapsed 8eltline Region Irradiation Cycle Avg.
Cumulative Fluence (n/cm2)
Cycle Time Flux Cycle Design No.
(EFPY)
(n/cm2-sec)
Specific Basis (a) 10 II I7 1
1.2 1.46 x 10 5.34 x 10 6.55 x 10 10 Il 18 2
2.1 1.60 x 10 9.89 x 10 1.16 x 10 10 18 18 3
2.7 1.69 x 10 1.35 x 10 1.55 x 10 10 18 18 4
3.5 1.68 x 10 1.77 x 10 1.99 x 10 10 18 I9 5
4.3 1.81 x 10 2.19 x 10 2.41 x 10 10 18 18 6
5.0 1.79 x 10 2.62 x 10 2.84 x 10 10 18 18 7
5.9 1.14 x 10 2.93 x 10 3.32 x 10 10 I9 I9 Cy 8 + EOL(b) 25.8 1.14 x 10 1.01 x 10 1.46 x 10 10 I9 I9 EOL 32.0 EFPY 32.0 1.14 x 10 1.23 x 10 1.81 x 10 f.
10 2
a) Design basis fast neutron flux = 1.79 x 10 n/cm -sec at 3391 MWth" b) Current neutron fluences are defined as of the beginning of Cycle 8 (February 9,1984).
The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26,2008) using the Cycle 7 average flux and an 80% capacity factor.
1297E:10/122085 14
TABLE 11.2-4 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45* AZIMUTHAL ANGLE ("I l
ZION UNIT 1 Elapsed Beltline Region Irradiation Cycle Avg.
Cumulative Fluence (n/cm2)
Cycle Time Flux Cycle Design No.
(EFPY)
(n/cm2-sec)
Specific 8 asis (b) 10 I
18 1
1.2 2.18 x 10 7.98 x 10 1.01 x 10 0
10 18 2
2.1 2.35 x 10 1.47 x 10 1.80 x 10 10 18 18 3
2.7 2.64 x 10 2.03 x 10 2.40 x 10 0
18 18 4
3.5 2.63 x 10 2.68 x 10 3.08 x 10 10 18 18 5
4.3 2.68 x 10 3.32 x 10 3.73 x 10 10 18 18 6
5.0 2.81 x 10 3.99 x 10 4.40 x 10 10 18 18 7
5.9 1.63 x 10 4.43 x 10 5.14 x 10 Cy 8 - E0L(c) 25.8 1.63 x 10 1.47 x 10 "
2.25 x 10 10 0
E0L -, 32.0 EFPY 32.0 1.63 x 10 1.79 x 10 2.80 x 10 a) Maximum fast neutron flux incident upon the intermediate and lower shell plates and the intermediate to lower shell circumferential weld, 10 2
b) Design basis fast neutroc flux = 2.77 x 10 n/cm -sec at 3391 MWth*
c) Current neutron fluences are defined as of the beginning of Cycle 8 (February 9,1984).
The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26, 2008) using the Cycle 7 average flux and an 80% capacity factor.
1297E:10/121985 15
TABLE 11.2-5 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 4' SURVEILLANCE CAPSULE CENTER - ZION UNIT 1 Elapsed 8eltline Region 2
Irradiation Cycle Avg.
Cumulative Fluence (n/cm )
Cycle Time Flux Cycle Design 2
(n/cm -sec)
Specific Basis No.
(EFPY) 10 I
18 1
1.2 2.28 x 10 8.35 x 10 1.03 x 10 10 18 18 2
2.1 2.59 x 10 1.57 x 10 1.83 x 10 10 18 18 3
2.7 2.64 x 10 2.14 x 10 2.43 x 10 10 18 18 4
3.5 2.48 x 10 2.75 x 10 3.13 x 10 10 18 18 5
4.3 2.70 x 10 3.40 x 10 3.80 x 10 10 18 18 6
5.0 2.61 x 10 4.01 x 10 4.48 x 10 10 18 18 7
5.9 2.65 x 10 4.71 x 10 5.23 x 10 10 2
Note: Design Basis 4 = 2.82 x 10 n/cm -sec at 3391 MWth*
1297E:10/121985 16
TA8LE 11.2-6 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 40*
SURVEILLANCE CAPSULE CENTER - ZION UNIT 1 Elapsed 8eltline Region 2
Irradiation Cycle Avg.
Cumulative Fluence (n/cm )
Cycle Time Flux Cycle Design Capsule l
No.
(EFPY)
(n/cm2-sec)
Specific Basis 9ata 10 18 18 18 1
1.2 6.90 x 10 2.53 x 10 3.21 x 10 2.73 x 10 10 18 18 2
2.1 7.44 x 10 4.65 x 10 5.70 x 10 10 8
18 3
2.7 8.36 x 10 6.43 x 10 7.59 x 10 10 18 18 18 4
3.5 8.33 x 10 8.49 x 10 9.74 x 10 8.64 x 10 0
5 4.3 8.49 x 10 1.05 x 10 1.18 x 10 '
10 6
5.0 8.90 x 10 1.26 x 10 1.39 x 10 '
1.24 x 10 0
7 5.9 5.16 x 10 1.40 x 10 1.63 x 10 10 2
Note: Design Basis + = 8.77 x 10 n/cm -sec at 3391 MWth*
l l
l 1297E:10/121985 17
TA8LE 11.2-7 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE (a)
ZION UNIT 2 Elapsed 8eltline Region 2
Irradiation Cycle Avg.
Cumulative Fluence (n/cm )
Cycle Time Flux Cycle Design (n/cm -sec)
Specific Basis (b) 2 No.
(EFPY)
I lI 1
1.2 7.33 x 10 2.84 x 10 3.58 x 10 9
II I
2 2.0 8.99 x 10 5.00 x 10 5.79 x 10 9
I II 3
2.8 7.88 x 10 6.94 x 10 8.06 x 10 I
I 4
3.4 6.65 x 10 8.28 x 10 9.92 x 10 18 18 5
4.3 8.18 x 10 1.05 x 10 1.25 x 10 18 18 6
5.0 8.19 x 10' 1.24 x 10 1.46 x 10 9
18 18 7
5.7 7.49 x 10 1.41 x 10 1.67 x 10 Cy 8 -+ EOL('}
8 25.3 7.85 x 10 6.26 x 10 7.36 x 10 E0L-+ 32.0 EFPf 32.0 7.85 x 10' 7.92 x 10 9.31 x 10 18 18 a) Applicable to the longitudinal welds at 0*,
90*, 180*, 270* in the peak axial flux.
2 b) Design basis fast neutron flux = 9.22 x 10 n/cm -sec at 3391 MWth*
c) Current neutron fluences defined as of the beginning of Cycle 8 (July 9, 1984).
The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26, 2008) using the average of the Cycle 6 and 7 fluxes and an 80% capacity factor.
1297E:10/121985 18
TABLE II.2-8 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15' AZIMUTHAL AMGLE ZION UNIT 2 Elapsed 8eltline Region Irradiation Cycle Avg.
Cumulative Fluence (n/cm2)
Cycle Time Flux Cycle Design No.
(EFPY)
(n/cm2-sec)
Specific Basis (a) 10 lI II 1
1.2 1.15 x 10 4.46 x 10 5.59 x 10 10 I
lI 2
2.0 1.39 x 10 7.80 x 10 9.04 x 10 10 18 18 3
2.8 1.23 x 10 1.08 x 10 1.26 x 10 10 18 18 4
3.4 1.19 x 10 1.32 x 10 1.55 x 10 10 18 18 5
4.3 1.27 x 10 1.67 x 10 1.95 x 10 10 18 18 6
5.0 1.17 x 10 1.95 x 10 2.28 x 10 10 18 18 7
5.7 1.11 x 10 2.19 x 10 2.60 x 10 II 10 18 I9 Cy 8 + EOL 25.3 1.14 x 10 9.23 x 10 1.15 x 10 10 I9 I9 EOL 32.0 EFPY 32.0 1.14 x 10 1.16 x 10 1.45 x 10 10 2
a) Design basis fast neutron flux = 1.44 x 10 n/cm -sec at 3391 MWth*
b) Current neutron fluences defined as of the beginning of Cycle 8 (July 9, 1984).
l The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26,2008) using the average of the Cycle 6 and 7 fluxes and an 80% capacity factor.
i i
1297E:10/122005
,19 n
TA8LE 11.2-9 FAST NEUTRON (E > 1.0 MeV1 EXPOSURE AT THE PRESSt!RE VESSEL INNER RADIUS - 30' A7IMUTHAL ANGLE 710N UNIT 2 Elapsed 8eltline Region Irradiation Cycle Avg.
Cumulative Fluence (n/cm2)
Cycle Time Flux Cycle Design No.
(EFPY)
(n/cm2-sec)
Specific Basis (a) 10 II II
~
1 1.2 1.42 x 10 5.51 x 10 6.95 x 10 10 Il 18 2
2.0 1.63 x 10 9.42 x 10 1.12 x 10 10 18 18 3
2.8 1.58 x 10 1.33 x 10 1,56 x 10 10 18 18 4
3.4 1.64 x 10 1.66 x 10 1.93 x 10 10 18 18 5
4.3 1.61 x 10 2.10 x 10 2.42 x 10 10 18 18 6
5.0 1.24 x 10 2.40 x 10 2.84 x 10 10 l9 18 7
5.7 1.26 x 10 2.67 x 10 3.23 x 10 10 I9 I9 Cy 8 -+ EOL(b) 25.3 1.25 x 10 1.04 x 10 1.43 x 10 10 I9 I9 EOL 32.0 EFPY 32.0 1.25 x 10 1.30 x 10 1.81 x 10 10 2
a) Design basis fast neutron flux = 1.79 x 10 n/cm -sec at 3391 MWth*
b) Current neutron fluences defined as of the beginning of Cycle 8 (July 9, 1984).
The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26, 2008) using the average of the Cycle 6 and 7 fluxes and an 80% capacity factor.
1297E:10/122085 20
TABLE 11.2-10 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT Tile PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE (a)
IION UNIT 2 l
Elapsed Beltline Region Irradiation Cycle Avg.
Cumulative Fluence (n/cm2)
Cycle Time Flux Cycle Design No.
(EFPY)
(n/cm2-sec)
Specific Basis (b) 10 II 18 1
1.2 2.09 x 10 0.11 x 10 1.07 x 10 10 18 18 2
2.0 2.37 x 10 1.38 x 10 1.74 x 10 10 IO 18 3
2.8 2.50 x 10 1.99 x 10 2.42 x 10 10 18 18 4
3.4 2.70 x 10 2.54 x 10 2.98 x 10 10 18 18 5
4.3 2.64 x 10 3.27 x 10 3.74 x 10 10 18 18 6
5.0 1.72 x 10 3.67 x 10 4.39 x 10 10 18 18 7
5.7 1.82 x 10 4.07 x 10 5.00 x 10 Cy 8 + EOL(c) 25.3 1.76 x 10 1.49 x 10 2.21 x 10 10 I9 I9 10 I9 I9 EOL 32.0 EFPY 32.0 1.76 x 10 1.87 x 10 2.80 x 10 a) Maximum fast neutron flux incident upon the intermediate and lower shell plates and the intermediate to lower shell circumferential weld.
10 2
b) Design basis fast neutron flux = 2.77 x 10 n/cm -sec at 3391 MWth*
l l
c) Current neutron fluences defined as of the beginning of Cycle 8 (July 9,1984).
The data pertaining to Cy 8 and beyond represent a projection to the license expiration date (December 26, 2008) using the average of the Cycle 6 and 7 fluxes and an 80% capacity factor.
1297E:10/122085 21
TABLE II.2-11 FAST NEUTRON (E > 1.0 MeV) EXPO AT THE 4*
SURVEILLANCE CAPSULE CENTER - ZION UNIT 2 Elapsed 8eltline Region 2
Irradiation Cycle Avg.
Cumulative Fluence (n/cm )
Cycle Time Flux Cycle Design 2
(n/cm -sec)
Specific Basis No.
(EFPY) 10 I
IO 1
1.2 2.24 x 10 8.69 x 10 1.09 x 10 10 18 18 2
2.0 2.75 x 10 1.53 x 10 1.77 x 10 3
2.8 2.41 x 10 2.12 x 10 2.46 x 10 g
0 8
18 10 18 18 4
3.4 2.03 x 10 2.53 x 10 3.03 x 10 10 18 18 5
4.3 2.50 x 10 3.21 x 10 3.81 x 10 10 18 18 6
5.0 2.50 x 10 3.79 x 10 4.47 x 10 10 8
18 7
5.7 2.29 x 10 4.31 x 10 5.09 x 10 10 2
Note: Design Basis + = 2.82 x 10 n/cm -sec at 3391 MWth*
1297E:10/121985 22
i TABLE 11.2-12 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 40' SURVEILLANCE CAPSULE CENTER - ZION UNIT 2 Elapsed 8eltline Region Irradiation Cycle Avg.
Cumulative Fluence (n/cm2)
Cycle Time Flux Cycle Design Capsule 2
l No.
(EFPY)
(n/cm -sec)
Specific Basis Data 10 18 18 18 1
1.2 6.62 x 10 2.57 x 10 3.40 x 10 2.70 x 10 10 18 18 2
2.0 7.50 x 10 4.37 x 10 5.51 x 10 18 18 3
2.8 7.92 x 10 6.30 x 10 7.66 x 10 10 18 18 18 4
3.4 8.55 x 10 8.04 x 10 9.44 x 10 8.49 x 10 10 I9 I9 5
4.3 8.36 x 10 1.04 x 10 1.18 x 10 10 I9 I
6 5.0 5.45 x 10 1.16 x 10 1.39 x 10 '
10 I
I 7
5.7 5.76 x 10 1.29 x 10 '
1.58 x 10 10 2
Note: Design Basis + = 8.77 x 10 n/cm -sec at 3391 MWth' i
I l
i 1297E:10/121985 23
Figure II. 1-1 ZION REACTOR GEOMETRY i
I 0'
CAPSEES) 4o
- s. v. w Z J 8#A Cto9 kE CAPSULES M
N(
f 0
T.U.X,Y 40
/////
/
<s*
I
'"E
/
/
I
%9 i s8,,'O v //s I
i
/
l
,_-1 11 ://11 11 111
/
j
/
(
b'
/
l
'I
/
/ '
I
/
/
I
/
' /
\\
//
1
/'
/
I 24
_ -._ _ ___ _._.._, _.-- ~._ _.
~ -.
f j-mwmy m
.m.
ri ore 11. 2-1 m -
1 e emm m
-h==
"2
~
5-MAXIMUM FAST NEUTRON (E>1.0 MeV) FLUENCE r.__
-=
~~
6~"
A FUNCTION OF FULL POWER OPERATING TIME
=__ =..._t,
- =
r. _- __- ___
ZION UNIT 1
_p. 4 a.__
.; ;, i - :2
'"- :m =_ x 1s t gg 4,_
sii:*-===r-f-E Ei Y:P : -
-t - f i i! P
-2. _
Ir~ E$ =C#5_
- 55 I 30
_2.4
- _ _ 4 2: _3
-a -
=~;
3...
+
'~~
2._
'$C
,e
,s*
a i.
1 10!
' = =X=- k_M-W-=?= 4=i i-5 ~=i 9tE
- ' M== =
==-- ?-#
~-
~ =, =
=i- -
==
=-
= = ' -
g_ =EEW M. _.+.__ :--_. =. _:-- r '+4_t =_ = :_ _ _
__ _ W',
'+::i ? - -- E.,.-- -e :
=
a-- P' 5:
J _.-". _. _ = _ = - -=_==_=_=v 2
. _ s.-- - :- -
- - - - = _ = _. _- - - -
y
.. _. - _., = - _ -. -_ _ _._ _ _ - _ _.
g --
g 6..
-_.2_,_.
y-
- , " -t==E --li-i==i F#A
- t=+: rf+4 #i I-E '=+33I S 5 I_ i=tt tEi= ?
M4 3-':-i:4#+ ++,14 ivN U s' *^~ t E-t ~=2i f==& '?- =; :M ":. E== =-;5%- 4 =--# #;h-?== =t-
+-
==~!.=i= += E4 #
jg y.
_ _ _ -g-3x gg
_g3.. __
====.===;.
. = = =
x
3, p~
~
=
La-3..
5
/'
t l
m w 2 --
1 z
f E
A j
I I,
a
/
/
1 I
J T
I
/
i 18 r
/ "
10
=+-
- -==f.1 E4 =+,, _ - _
=m z __
y i. =
y_
- 8. 54MN E-
~ M=
~O
==C'Mt 45' '4~-MM M -Er E i~ 'E*:-Ni =2 n_==.
. - _ - -..... =. =.= _ _. _ = = _
.. -. =.. =.
- =_._u..._..
= : a
=.= = -
:
.-_.__--I_._g 7,
.. lf ~CL 7_ _ _ _ _ __===3 _3 qu
'~~
~
~ ~ ^ ' -
-____p,=-
t
$, -7._
. ] : 5 :. -
--i z = n=
? i.-== = = _. - = -
= +, + -. =i-g:-. L_r y
.; 5 y-3 - - L.
._--1=.. ; _::
li-f $ f i- ?-i. ' ~~ * : ~: 'i. f -:=V.=--@.hk' =
. = - ~ = =
= ;-~515: ?- =5^t-. : =35 4:-3_ E ~.:== kl.Ei
- i'? - :
~
4_
j 7 f - p- =_:
=1.iE__"_L 1" i=- 2i-_=.. s:--__.a _i._W==.J3- ~ :=.-12_
r-~,
- --a t5:==.p_-g-g NN5-55 N
I~
~ ~ ~
3.
- t.. : -.. -
.=.:
-:=
- = -~ -
. = - -
- ::J ::- :-
- ==.:1. _:.;==_- -
+-- ::
Actual 2.
-- = g _... 5
-5=~
Projected ~"
=__.
4 _ _.. L. __.
~~t' ~~
License Expiration i
e i
10E O
5 10 15 20 23 30 3)
- OPERATING TIME (EFPV) 25
i g-2 mu iu+1 i w.
g,4 w s'"
MAXIMUM FAST NEUTRON (E>1.0 MeV) FLUENCE r-
.- -p :
F1gure I!.2-2 EEM - Fi^-
-.+i4 ?
-e.
- 1_ _:.
- _.. s.. _ :-
7...
AT THE BELTLINE WELD LOCATIONS AS
- :=, -I== :
-== =
=
A FUNCTION OF FULL POWER OPERATING TIME
- = = = _ * = _ = _ = _ =
=-t 3~
- =.
ZION UNIT 2 n
us
+=t. #ss i==V===e 1 't=.= M =-
a*"
r=* -.t+=:- =
_.. __ _ _ = _ _
_--~ _ y- _
_~ __ ;
_._r..._;
3...
_v_
g = --
g_ _.
<w A
1 i.
dd' 1
e i
~ 10 9 r
4 g... -e + = =
z-
==cr
&+=
= rr+ += u
r+ m
i ess rm t 4 i ^tw -
+ c+ = u LM
=- _M:4 # * =.+4 4C"i =_=id= *~M, 2#--
M
- +=
+
s.. = * =. =.
=_=._r - =. _1-
' W. =
._'.- - '=.
=
'? :- *g ~5
_ - - _ _ _ _ -, - _ - - - ^ ', _ _ _ _ '.
.___4-_~'___-
'- ~ - ' '
^-
'- r.f-'
y,,
s --
_.=_
j-u_ =_
g 6 5 --
m.'_____-,K_-:=-Er.
.vr. ai.. w_..._a. _ _.... = = _ = _...=.+.+_1=-
==~==?= = Lei s
unsi++1mi5.-
==-a +-
. _ _c-fz-. _ _ _ _. _ -.i==.. _..' i.. =.5 tw 4;3
__.:.- = = = =
. ry -
=__,__q _,-- -
_q_2.-
- =.
- . 7..
4--
7 o,--
_~~~
g h
I 2
y
/
E 2" B
,r w
E s
I.
7 f-
~'~'
/
18
/
10 fi.
'iel = : =M -
- =e- #iW++-n
,==-
-- r
.tt++=
==-
=
- 8. M#W-W t ~~'*= '" MH=
W-4-M^= #=*^ M
=M
=F=*
M = M ~2 -
. __. } 3. : =. = -.. _. =_ -
=. =.. =. =.
- - := -..
- :=.
- _ : :.2-
= _
- _ _r =_y
==;=.
y,
-*j== -_
~~
=
6.
,r-__ = _ _ _
_ _ _ = _ -
+ - -
5,
+HI Mi w=' a-t-i-t = L= -E- =+i =-
= + + =w i=+
n= 4 +.
=
1-+ -
Mi -
+
"-^-
~_T i.i 4-}~ r i
.i f il_:
i p s i.
.ia;:--
_fli-si d lifil :.il:~r~ if.' - ~-] 2 =-i -' ' l ~y:.; 4 n=I-f. - 4'
== -i=M Mt &=_ L-~ ~TL -::==+ir=: t ElwiHV%=t- =*4L s=====_ 3.y = =_ = ;=. E= =M7 :===___:-=+ ~ ~ - - =--r- - - -== 3_E_ EE=- 7 =v. :T= _. = =: === : :.- _.. 3._:_ =. _= -=__ Actual-im _ --r
- ~
Projected.--- j License Expiration f 1 I II I 10 O 5 10 15 20 25 30 D OPERATING TIME (EFPV), 26
I l ifit' ~~ '~ MAXIMUM CURRENT AND PROJECTED EOL FAST ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~ g-.y 3 + w_t #.# w l =:+4 r:+:,&, = s$ gg Figure II.2-3 si 41 1 IIl i W=g s
== E 5 ~~ -M Z 7... ^ ~ ~ ~ ~ ~ ~ ~~- 6"- NEUTRON (E>1.0 MeV) FLUENCE AT THE ~ PRESSURE VESSEL INNER RADIUS AS A FUNCTION OF AZIMUTHAL ANGLE L.... E a i_m... --: w=+- .re-=--=----. =ss-
- - s r+
_.=g =. +.1; ;. m-ZION UNIT 1 g Ee
== =t =. Jag m -+ 3... 2... _ _ _ _rn, = -- - - ry
- trr, e
F. - magg F fg4 -- P-r I I f 8 I 8 I t / ? l I I i 2 1_ 1,0 ' -:=== =-g> = y' ~ cn
===g.,.ua g gm. tw gum. y >=mv=-np n =. - zus. e .w;;;- z_ . ;g-. , _ _.. _4 _ : :, m .=,;- = _ = = _ = _ _. =. m. ,C_- .-e_-- .._m - - - - '- + -. ew '.56..
- r n==i=Ew+= w=75f-!-- srt? atn 5.a EF: i*itPtu += =r : =2ri==i+
. - - + = ~=ETiE2 M ki-4=h n tii= 4 ~~ =^M ^".#9*3 MW413# fi ?? i EL z .m Ws -.=_ _^ ~_:_ f,.- = y I:'3 " p7 f.. B g3.. 1 1 1 A 1 10[8
-- * + +
w s+M- ++++ ; =_ - i- +* W44
- =P w t*
- i E.=4WMM ia*=- =
P'i~i".tS _ * ~M44 - * *=.
~ =
- s. C=n C _:_I'. 4h=_1 ~.M W si Mi + - iM *
==
- . m m.
,_ mm y. m.==. __._ . _~_..._.. _ _ 6. ~~ + 5 _= _..u. A t.g +3a ;; 4.g r 3_3 g-a 5.; y4m 43#s:324 7=+ 4_.3_. . ag_ 2-M=m:ss:_ 2.p. =m
==++ tas3 zi*t tum
- ==5 55
=.4.it :+n 421-gs 1 Ei33% 5===. " = fi =" i. 'l : E t?M =i + ! ~14-iniHt-it 4#s-4== E . _ =.. -.... -
- - - - ~.- Current-2-
I I Projected i kf f. ! I 6 6 8 1 l i 10"7 1
- 1 O
10 20 M 40 bo i AZIMUTHAL ANGLE (deg.) 27
s u u. uu - gg=m.g n=E 8... Z:2.T -~.- -~~:: T. _- Figure II.2-4 MH 1" U t : 2-s= 2.5w= se . ;;.2 : =~__:- 2 7... ~ MAXIMUM CURRENT AND PROJECTED EOL FAST
- - := =
.0% MEMM 6-" PRESSURE VESSEL INNER RADIUS AS A -- ~ FUNCTION OF AZIMUTHAL ANGLE ...._=+ -.+++ = + =- 5... n +=.. = = =uaw-wi =Rii=W. s-ZION UNIT 2 h=h 4
- i i-i + i-i = i
. _. 3, 3... 2... y ' 41.E.FEU ; > renw r c... n_ mne. > nntrattgg} .i. ,,ii 3., I_ 10 ' Hu" -;;;'i -;=+T22 wimaa=ti. =v== M=1 = M n-g.. itwir=_== t" M+n w+'=s=1 ~~ iM 4--- W -T ~ y -- ._1-_
. _
I~M -* ~M=_= -I_ _2 :J- ".. _1 N= 1.- . s.. _. = _ _.
=
- - = - . _ _ _ _~.____ ev f 6.. '- c 5.. 4+ -+g3==g t+3y5m: . _ _. ;a=++. w =a=-=g 4 m m;' n+a= Ligz=L=.=5ua.
=
w + n
.a.z :: 3 :_1.
= = :.:-
- - : =.
=_:_ =r-o f _;2
- _ =-== =-' _ _ _ =. =__.
U _ ar -
- =::2: :
-- 2 5 * -- -n a f_ F l 3.,_@ ~f w x: 2.. J'1 s s w" 1 10.8 . =.. 7 ;; g
- g. - _ _.-. _; = =i=_ *MT2=:M.: 4 =-+ ti:Lb 4&EE =.1 L
-i = =,.:=DLV i==
==. ='g
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SECTION III l MATERIAL PROPERTIES For the RT calculation, the best estimate copper and nickel chemical PTS composition of the reactor vessel beltline material is necessary. The material properties for the Zion Units 1 and 2 beltline region will be presented in this section. L i 111.1 IDENTIFICATION ANO LOCATION OF BELTLINE REGION MATERIALS t ( The beltline region is defined by the Rule [1] to be "the region of the l reactor vessel (shell material including welds, heat af fected zones, and h plates or forgings) that directly surrounds the offective height of the active core and adjacent regions of the reactor vessel that are predicted to experience suf ficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage." Figures III.1-1 and III.1-2 identify and indicate the location of all beltline region materials for the Zion Units 1 and 2 reactor vessels. j III.2 OEFINITION AND SOURCE OF MATERIAL PROPERTIES FOR ALL VESSEL LOCATIONS l Material property values for the shell plates, which have been docketed with the NRC in Reference 5, were derived from vessel fabrication test certificate i results. The property data for the welds have also been docketed with the NRC in Reference 5, however, the weld properties cannot be used in the same direct i manner as the properties for the plates. Fast neutron irradiation-induced changes in the tension, fracture, 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. j To address the variation in chemistry, Babcock & Wilcox (B&W) performed a reactor vessel beltline weld chemistry study of eight B&W vessels, including ) Zion Units 1 and 2, and reported the results in BAW-1799 [6] for the Westinghouse Owners Group (WOG). The scope of the work included collecting [ 1297E:10/121985 31 ,- ~
existing sources of chemistry data, performing extensive chemical analysis on 4 the available archive reactor vessel weldments, and developing predictive methods with the aid of statistical analyses to establish the chemistry of the reactor vessel beltline weldments in question. In addition to the 8W report 8AW-1799, the WOG Reactor Vessel Beltline Region Weld Metal Data Base was used. The WOG data base, which was developed in 1984 and is continually being updated, contains information from weld qualification i records, surveillance capsule reports, the BW report BAW-1799, and the Materials Properties Council (MPC) and the NRC Mender MATSURV data bases. For each of the welds in the Zion Units 1 and 2 beltline region, a material data search was performed using the WOG data base. Searches were performed for materials having the identical weld wire heat number as those in the Zion 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. 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 8. i When the results of the material data base searches were evaluated, it was found that a large scatter existed in the measured as-deposited copper values for the data obtained for weld wire heat 72105, which is associated with the WF-70 weld seams, and weld wire 71249, which is associated with the Unit 2 girth weld. Preliminary RT calculations showed that the submerged arc PTS welds fabricated with filler wire heat number 72105 were limiting in regards to PTS for both Zion Units 1 and 2 reactor vessels. Since the chemical 1 composition of these welds significantly impacts the PTS concern for the Zion vessels, a statistical materials program, which is discussed in the next section, was developed to address the large scatter in copper content. The chemical composition values for heat 71249 have already been addressed via evaluations of reactor vessel materials data for Turkey Point Units 3 and 4 (see Reference (7]). i l 1297E:10/121985 32
III.3 MATERIAL CHEMISTRY STULY FOR WELD WIRE HEAT 72105 i l Filler wire heat number 72105 was used with Linde 80 flux lot number 8669 in making the beltline welds of both reactor vessels and with Linde 80 flux lot number 8773 in making the Zion reactor vessel surveillance capsules. The following is a susnery of the results f rom the statistical materials program undertaken to address the large scatter found in the measured as-deposited copper values for filler wire heat number 72105. 111.3.1 STATISTICAL EVALUATION III.3.1.1 Statistical Analysis Results As part of the Westinghouse Statistical Materials Program, all available data on the chemical composition of heat 72105 was gathered. B&W previously recommended in report BAW-1799 that mean chemistry values of 0.35 wt% and 0.595 for copper and nickel content be respectively defined for weld wire heat number 72105. These values were also used to define RT values for the NOT respective Zion 1 and 2 reactor vessel welds in the NRC Policy Issue on PTS ] SECY-82-465 (11). However, 30 more data points than those used to I substantiate the B&W recommended values were obtained via the WOG data base development. These additional chemical measurements, which were obtained from several sources, change the previous recommendations given in BAW-1799. From a total population of 87 data points, which include 51 data pointe from B&W report BAW-1799 [6), the mean copper value obtained was 0.32% with a coefficient of variation (standard deviation divided by the mean) of 21%. (A high coef ficient of variation indicates that there is a significant amount of scatter in the data). The mean copper value obtained f rom only the B&W archive data in BAW-1799 is 0.35% with a coef ficient of variation of 17%. The total population of 87 data points contains measurements for heat 72105 in combination with three different flux lot numbers. The data follows for the three different lots of flux. 1297E:10/122005 33
TA8LE III.3-1 j
SUMMARY
OF RESULTS FROM WESTINGHOUSE STATISTICAL MATERIALS STUDY FOR HEAT 72105* i j FLUX LOT NUM8ER OF DATA MEAN Cu COEFFICIENT NUM8ER POINTS (WTX) 0F VARIATION 8688 2 0.2555 f 8669 17 0.4055 16% l 8773 68 0.2955 14.5% i j All data is provided in Appendix 8 of this report. I 1 j The 68 data points for flux lot 8773 were obtained using four different chemical analyses techniques, four dif ferent laboratories, and both irradiated I; and unirradiated samples. Except for two data points, the data for flux lot l 8669 wre obtained by using emission spectrometry analyses on a weldsent from a nozzle bore dropout taken from some vessel other than Zion Units 1 or 2. r The other two points were obtained from a weld metal qualification test and a retest of this specimen. It appears f rom the data found in 8AW-1799 Appendix 8 (6) ( see Figure l 111.3-1), that two very distinct populations of data exist for the same filler } wire even though all of the measurements were done by 84W using emission 4 i spectrographic analysis of unirradiated specimens. Such a large discrepancy i l should not exist between dif ferent flux lots of welds made f rom the same heat ( j of filler wire, since the flux does not contain an appreciable amount of f copper. The principal source of copper is from the filler wire coating, and } the flux neither extracts copper from the weld deposit nor significantly j contributes copper to the weld deposit. In the next two sections, the impact of the flux is evaluated to determine whether the flux populations are truly different or if a bias exists in one of the populations. I i + r 1297E:10/122005 34 i ...-..-.__,..___.-_.-,.------..----.---m.
111.3.1.2 Filler Wire Examination To examine the potential for a biased set of data, the copper content in the filler wire, which is the principal source of copper in the weldment, was evaluated. In Appendix A of the B&W report, an examination of the filler wire heat number 72105 was performed. In this analysis, the filler wire was stripped of its copper coating. The copper quantity in both the coating and the bare metal was measured to determine the principle source of copper in the as-deposited weld metal, which is usually the surf ace coating. The quantity of copper content present on each wire sample was measured in compliance with ASTM 0-168-77, Method 0-Atomic Absorption Spectrophotometry. B&W looked at five different heats of wire, one of which was 72105. They used samples that were 8-10 inches long obtained from two or more spools of filler wire. As shown in Tables A-3, A-4, and A-5 in BAW-1799, 86W obtained a copper concentration of 0.230 wt% for the coating and a copper concentration of 0.075 wt% for the bare filler wire for heat 72105. Thus, the total copper concentration found for wire heat number 72105 in this analysis was 0.30 wt%. This number is in agreement with the mean copper value of 0.295% determined for all available measurements with flux lot number 8773 f rom the Westinghouse statistical materials study (see Table !!!.3-1). The 0.30 wt% value is also in agreement with chemical measurements of weld metal qualification material made with flux lots 8688 and 8669 (average measured values equal 0.255 wt% for flux 8688 and 0.305 wt% for flux 8669 - see Appendix 8 of this report), f 111.3.1.3 Impact of Flux Lot for Other Filler Wires in order to determine whether or not other welds made f rom the same filler wire with more than one flux result in distinct populations, the archive data in Appendix 8 of BAW-1799 were examined. Table !!!.3-2 summarizes the result of the calculated means, standard deviations, ard coef ficient of variations of the populations examined. The same phenomenon of two distinctly dif ferent means with small coefficients of variation, as found for the 12105 data, only exists for one other heat, number 71249. However, the NRC has already reviewed the data for heat 71249 and has accepted the mean value from all the data for the copper weight percentage (7). 1297E:lD/121985 35
From other data in the 84W report, flux 8669 did not result in the same high i copper values as found for the nozzle bore dropout weldsent when used with j other filler wires (i.e., heat 72442), and discrepancies in the various flux populations do not exist (see Table !!!.3-2). From the above observations, there are two passible explanations for the high copper reported in the nozzle dropout. Either the chemical analysis is in l error or the particular spool (s) used for the nozzle dropout contained much higher copper than all the other 72105 spools. !!!.3.2 CALCULATION OF A0 JUSTED RT NOT Even if the data from the nozzle bore dropout with flux lot 8669 is biased, it l may be difficult to defend eliminating this data from the total population because this lot of flux was used in making welds in the two Zion vessels. The methodology for calculating the RTPTS ' 41ues prescribed in the PTS Rule is just one of many ways for calculating RT "C' IV *** NOT* l issued for calculating RTPTS **I"
- U 0" l.99 Revision 2, is under development (8]. This trend curve represents the latest methodology for calculating RT Therefore, in order to put the NOT.
difference in copper values for filler wire 72105 into perspective, the difforence in terms of the adjusted RTNOT, as prescribed in the proposed Regulatory Guide 1.99 Rev. 2 (see Table !!!.3-3) was calculated. I For the limiting welds of the Zion reactor vessels, using 0.56% Ni and copper mean values of 0.32% and 0.40% (the maximum copper value provided in Table !!!.3-3), and predicted fluence values representative of 32 effective full power years, the dif ference in the mean shif t of RT S NOT is well within the margin added to the mean shift of RTNOT, which is 56*F per the provision of the proposed Regulatory Guide 1.99 Revision 2.
- Thus, averaging all svallable data proves to be a viable approach for weld heat number 72105 for the Zion Units I and 2 reactor vessels.
I I 1297E:10/121985 36
=_. 111.3.3
SUMMARY
OF MATERIAL CHEMISTRY STUDY FOR WELO HEAT NUM8ER 72105 Based on the above logic, the copper content for weld heat number 72105 is 0.32 wt%, a number that is the average of all data with a standard deviation of 0.067 wt%. The mean nickel content is 0.56 wt% with a standard deviation of 0.06 wt%. These numbers are used to calculate the RT values f und in PTS Section IV. 111.4
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 Zion Units 1 and 2 reactor vessels are respectively given in Tables 111.4-1 and 111.4-2 along with the references for this information. Although phosphorus is no longer used in the calculation of RT wt respect to the PTS rule [1], it is given for reference since it NOT is currently used in the Regulatory Guide 1.99, Revision I trend curve (10). The initial RT value f 0*F, which is shown for all of the Zion Units 1 NOT and 2 reactor vessel beltline weldments,. is the generic mean value defined in i the PTS rule [1] for welds made with Linde 80 flux. The data in Tables !!!.4-1 and !!!.4-2 are used to evaluate the RT values FTS for the Zicn Unit 1 and 2 reactor vessels. j 1 i N 1297E:10/121985 37
4 TA8tE III.3-2
SUMMARY
OF ARCHIVE DATA FROM 8AW 1799 APPENDIX B Filler Wire Flux Lot # Number of Mean Cu Standard Coefficient of Data v Deviation Variation o/u Points (wt%) o (wt%) 72105* 8669 15 0.42 0.048 11.4% 8773 36 0.32 0.028 9.0% J 299L44 8650 48 0.352 .024 6.9% 8596 11 0.375 .010 2.8% 71249 8445 9 0.181 .027 14.7% 8738 26 0.286 .020 7.1% i 61782 8436 12 0.204 .045 21.9% 8457 29 0.270 .043 16.0% ~ 406L44 8688 21 0.318 .012 3.7% 8773 8 0.275 .012 4.3% 72442* 8669 13 0.222 .066 29.6% 8579 18 0.262 .028 10.7%
- 0ne weld contained these two filler wires with flux lot number 8669.
4 I 1297E:10/121985 38
T TABLE 111.3-3 CHEMISTRY FACTOR (CF) FOR WELOS PER PROPOSED REG. GUIDE 1.99, Rev. 2 MEAN SHIFT RT NOT " l Copper. Nickel, Wt. % Wt. % 0 0.20 0.40 0.60 0.80 1.00 1.20 l 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 ~ 0.04 24 43 54 54 54 54 54 0.05 26 49 67. 68 68 68 68 0.06 29 52 77 82 82 82 82 0.07 32-55 85 95 95 95 95 0.08 36 58 90 106 108 108 108 0.09 40 61 94 115 122 122 122 s 0.10 44 65 97 122 133 135 135 0.11 49 68 101 130 144 148 148 O.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 0.16 70 88 115 149 178 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122' 154 187 214 230 0.19 83 100 126 157 191 220 238 0.20 88 104 129 160 194 223 245 0.21 92 108 133 164 197 229 252 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 0.25 110 126 148 176 209 243 272 0.26 113 130 151 180 212 246 276 0.27 119 134 155 184 216 249 280 0.28 122 138 16t 187 218 251 284 0.29 128' 142 164 191 222 254 287 0.30 131 146 167 194 225 257 290 0.31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263 296 0.33 1K 160 180 205 234 266 299 0.34 149 164 '184 209 238 269 302 0.35 153 168 187 212 241 272 305 0.36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 311 0.38 166 182 203 227 254 285 317 200 223 250 281 314 0.39 171 185 0.40 175 109 207 231 257 288 320 1297E:10/121985 39
TA8LE 111.4-1 ZION UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P I (Wt.%) (Wt.%) (Wt.%)('F) Source Intermediate Shell Plate 8-144-2: 0.12 0.49 0.010 10(a)Ref. [5] Intermediate Shell Plate 8-144-1: 0.12 0.49 0.010 5 Ref. [5] Lower Shell Plate 9-144-1: 0.13 0.48 0.013 -4 Ref. [5] Lower Shell Plate 9-144-2: 0.15 0.50 0.010 20 Ref. [5] Circumferential Weld - Intermed. to Lower Shell WF-70, Heat 72105, Linde 80 Flux 8669: 0.32 0.56 0.017 0(b) WOG Material Data 8ase, Statistical Materials Study Sec-tion 111.3 Longitudinal Nelds - Intermed. & Lower Shell WF-4/WF-8, Heat No. 8T1762, Linde 80 Flux 8597/8632: 0.29 0.55 0.0130(b) Ref. [6] Notes: (a) The initial RTNDT value for this plate is estimated according to Branch Position MTEB 5-2 [9] (b) The initial RTNDT values for the welds are the generic mean value defined by the PTS rule [1] for Linde 80 welds. 1297E:10/122085 40 l i
TA8LE 111.4-2 ZION UNIT 2 REACTOR VESSEL 8ELTLINE REGION MATERIAL PROPERTIES Cu Ni P I l (Wt.%) (Wt.%) (Wt.%) (*F) Source Intermediate Shell Plate 8-152-1: 0.12 0.51 0.010 22(a) Ref. [5] Intermediate Shell Plate 8-152-2: 0.12 0.53 0.010 22 Ref. [5] Lower Shell Plate 9-152-1: 0.12 0.54 0.010 10(*'
- ,4. [5]
Lower Shell Plate 9-152-2: 0.14 0.52 0.008 2(a) Ref. [5] Circumferential Weld - Intermed. to Lower Shell SA-1769, Heat 71249, Linde 80 Flux 8738: 0.26 0.60 0.019 0(b) was Material Data Base, Ref. [6] and Ref. [7)(C) Longitudinal Welds - Intermed. Shell WF-29, Heat 72102, Linde 80 Flux 8650: 0.23 0.63 .019 O(b) Ref. [6] Longitudinal Welds - Lower shell WF-70, Heat 72105, Linde 80 Flux 8669: 0.32 0.56 0.017 O(b) was Material Data Base, Statistical Materials Study - Sec-tion III.3 Notes: (a) The initial RTNDT value for these plates are estimated according to Branch Position MTE8 5-2 [9] (b) The initial RTNOT values for the welds are the generic mean value defined by the PTS rule [1] for Linde 80 welds. (c) Agreement exists between References [6] and [7] and the WOG Material Data Base for heat 71249. 1297E:10/122085 41
FIGURE III.1-1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE ZION UNIT NO.1 REACTOR VESSEL CIRCtNFERENTIAL SEAMS VERTICAL SEAMS 270 WF-4 & WF-8 g I W -WF-154 45 0 ad b 5 180* 0* ~ CORE g WF-70 WF-4 & WF-8 9O 270* l [ d 45 0 WF-8 180* 2 0' i 90* i l 42
FIGURE III.1-2 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE ZION UNIT NO. 2 REACTOR VESSEL t CIRCUMFERENTIAL SEAMS VERT 1 CAL SEh45 l 270* d - WF-200 45 0 $G -- 0* 180* g W = 5 v L COP.E 4 - SA-1769 \\ 90* WF-29 270* a x m f5 45 0 ~ WF-70 / a 180* Z t 0* WF-70 4 90* 43
FIGURE III.3-1 B&W ARCHIVE DATA
- FLUX 8773 AND 8669 COMPARISON Copper Histogram 17 18 -
8773 [/ I ~ = 0.32 14 - r = 0.028 13 - /. 9g [ 12 - 8669 11 - [j = 0.42 to - g. / r = 0.048 / = 11.4% g. g 7- / 53 ~ ~ 0.285 0.205 0.325 0.355 0.385 0.415 0.445 0.475 r#2 8773 8669
- Data is found in BAW 1799 [6] Appendix B l
44
SECTION IV OETERMINATION OF RT O_R A R B E NE R W ON MAT W ALS PTS Using the methodology prescribed in Section 1.2, the results of the fast neutron exposure provided in Section II, and the material properties discussed in Section III, the RT values f r Zion Units 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 materials of the Zion Units 1 and 2 reactor vessels 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. Figures IV.1-1 and IV.1-2 present the RTPTS. values for the limiting longitudinal weld, circumferential weld and shell plate of the Zion Units 1 and 2 vessels in terms of RT versus fluence
- curves. The curves in these PTS figures can be used:
o to provide guidance to evaluate fuel reload options in relation to the NRC RT creening CMedon for MS O.e., RT values can be madh PTS PTS projected for any options under consideration, provided fluence is known), and o to show the current (5.9 EFPY for Zion 1 and 5.7 EFPY for Zion 2), end-of-license (25.8 EFPY for Zion 1 and 25.3 EFPY for Zion 2) and end-of-life (32 EFPY for both Zion Units) RT values using actual and PTS projected fluence.
- The EFPY can be determined using Figure II.2-1 for Unit 1 and Figure II.2-2 for Unit 2.
1297E:10/121985 45
Table IV.3-1 and IV.3-2 provide a summary of the RT values for all PTS beltline region materials for the lifetime of interest. IV.2 DISCUSSION OF RESULTS As shown in Figures IV.1-1 and IV.1-2, the welds are the governing locations for both reactor vessels relative to PTS. All the RT values remain below PTS the NRC screening values for PTS using the projected fluence values through j end-of-life (32 EFPY). 4 4 l l l 1297E:lD/121985 46 I
6 ' -~
- TABLE IV.3-1 RT VALUES FOR ZION UNIT 1 PTS RT Values ( F)
PTS Present End-of-License End-of-Life location Vessel Material (5.9 EFPY) (25.8 EFPY) (32. EFPY) 1 Interinediate shell plate 8-144-2 112 132 136 2 Intermediate shell plate 8-144-1 107 127 131 3 Lower shell plate 9-144-1 103 125 129 4 Lower shell plate 9-144-2 138 164 169 5 Interinediate to lower shell circumferential weld WF-70 222 284 297 6 Interinediate shell longitudinal welds WF-4/WF-8 169 224 235 1297E:10/122085 47
TA8LE IV.3-2 RT VA UES FOR ZION UNIT 2 PTS RT Values (*F) p73 Present End-of-License End-of-LifC Location vessel Material (5.7 EFPY) (25.3 EFPY) (32 EFPY) 1 Intermediate shell plate 8-152-1 123 146 150 2 Intermediate shell plate 8-152-2 124 146 151 3 Lower shell plate 9-152-1 112 135 140 4 Lower shell plate 9-152-2 114 141 146 5 Intermediate to lower shell circumferential weld SA-1769 190 245 256 6 Intermediate shell longitudinal welds WF-29 147 190 199 7 Lower shell longitudinal welds WF-70 179 238 250 1297E:10/122085 48 l
S25 NRC RT Screening Value - Circumferential Weld PTS ,300 NRC RT Screening Value - Plates and longi inal Welds PTS 2$0 Circumferential Weld 225 200 C Longitudinal Welds S175 em KEY 7150 g A Current Life (5.9 EFPY) Limiting Plate E End-of-License (25.8) 125 using actual and projected fluence values 100 G End-of-Life (32 EFPY) using actual and projected 75 fluence values 50 25 ' ' 'ni O 1018 1018 1020 FLUENCE. NEUTRONS / CW2 Figure IV.l.1 Zion Unit 1 - RT Curves per PTS Rule Methodology [1] PTS
I l-NRC RT Screening Value - Circumferential Welds PTS g,_____________________.____________-_ NRC RT Screening Value - Plates and Longitudinal Weld 280 PTS 260 240 Lower Shell L noitudinal Welds 220 200 N Circumferential Weld mjgg ta.
- 160 v
e4 KEY
- 140 m
E A Current Life (5.7 EFPY) w" 120 g End-of-License (25.3 EFPY) Limitina Plate using actual and projected 100 fluence values 80 O End-of-Life (32 EFPY) using actual and projected fluence values 60 40 20 I I 0 1018 1018 1020 FLUENCE. MEUTR0NS / CW2 Figure IV. 2.1 Zion Unit 2 - RT Curve per PTS Rule Pfethodology [1] PTS
S'ECTION V CONCLUSIONS AND RECOMMENDATIONS Calculations have been completed in order to determine RT values for the PTS Zion Units 1 and 2 reactor vessels to meet the requirements of the NRC Rule for Pressurized Thermal Shock [1]. This work entailed a neutron exposure f evaluation and a reactor vessel material study. l Detailed fast neutron exposure evaluations using plant specific cycle by cycle l core power distributions and state-of-the-art neutron transport methodology have been completed for the Zion Units 1 and 2 pressure vessels. Explicit calculations were performed for the first seven operating cycles of both units. For both units, projection of the fast neutron exposure beyond the current operating cycle was based on continued implementation of low leakage fuel management similar to that employed during cycle 7 for Unit I and cycles 6 and 7 for Unit 2. In regard to the low leakage fuel management already in place at the Zion Units, the plant specific evaluations have demonstrated that for the low leakage case the average fast neutron flux at th'e 45' azimuthal position has I been reduced by about 35% at Unit 1 and 25% at Unit 2 relative to that existing prior to implementation of low leakage. In particular, the following data applies at the 45' location. 2 + (n/cm -sec) Unit 1 Unit 2 j 10 Out-In Pattern 2.51 x 10 2.42 x 10 10 10 Low Leakage Pattern 1.63 x 10 1.76 x 10 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 lov leakage will differ from the data presented above. 1297E:lD/121985 51
It should b3 natsd that significant deviaticns from the low leakage scheme already in place will affect the exposure projections beyond the current l 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 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. i l The fast neutron fluence values from the plant specific calculations have been compared directly with measured fluence levels derived from neutron dosimetry contained in the three surveillance capsules withdrawn from Zion Unit I and l the two surveillance capusles withdrawn from Zion Unit 2. For Unit 1, the I ratio of calculated to measured flueiicE values ranges f rom 0.93 to 1.02 for i the three capsule data points. The corresponding ratio for Unit 2 is 0.95 for both capsules removed from that reactor. This excellent agreement between l calculation and measurement supports the use of this analytical approach to j perform a plant specific evaluations for the Zion reactors. Material property values for the Zion Units 1 and 2 reactor vessel beltline J' region components were determined. TnD pertinent chemical and mechanical properties for the shell plates remain the same as those that have been j docketed with the NRC in Reference 5. The weld material properties are consistent with those reconenended by 88W in reptet BAW-1799 [6] with the j exception of those weldments made with weld wird heat number 72105. For this l weld material, a material chemistry study was completed by Westinghouse that { included more data than that reported in BAW-1799 in order to define mean chemistry values of copper and nickel (0.325 and 0.56% respectively) for these limiting weld locations. l l Using the prescribed PTS Rule methodology, RT values were generated for PTS all beltline region materials of the Zion 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 NRC screening values PTS for PTS using the projected fluence exposure through 32 EFPY. The m st i 1297E:lD/122085 5,2 1 I
limiting values at end-of-license (25.8 EFPY for Zion 1 and 25.3 EFPY for Zion
- 2) are 284*F for the circumferential weld for Unit 1 and 238'F for the longitudinal welds in the lower shell of Unit 2.
This report is provided to enable Connonwealth Edison Company to comply with the initial 6 months submittal requirements of the USNRC PTS Rule. l I r ( 1297E:10/121985 53
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, Discrete Ordinates Transport Technique," WANL-PR(LL)034 Vol. 5, August 1970. 3. " SAILOR RSIC Data Library Callection OLC-76." Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P, Cross-Sectigi Library for Light Water 3 Reactors. 4. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology - to be published. 5. Commonwealth Edison letter from D. E. O'Brien to Mr. A. Schwenur of the NRC, " Zion Station Units 1 and 2 NRC Docket Nos. 50-295 and 50-304," September 7,1977. 6. B&W Owners Group Report, BAW-1799, "B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study", July 1983. 7. NRC Letter Docket Nos. 50-250 and 50-251, " Evaluation of Reactor Vessel Materials Data for Turkey Point Plant Units 3 and 4 Reactor Vessels", f rom S. A. Varga to J. W. Williams, Jr., of Florida Power and Light Company, April 26,1984. 8. Advisory Committee for Reactor Safeguards (ACRS) Metal Components Subcommittee Meeting, "Oraft Regulatory Guide 1.99, Revision 2, Radiation Damage to Reactor Vessel Materials", Washington, D. C., September 4-5, 1985. i 1297E:10/121985 54
9. NUREG-0800 - U.S. NRC Standard Review Plan, Branch Technical Position 5-2, Revision 1, July 1981.
- 10. " 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.
- 11. NRC Policy Issue
" Pressurized Thermal Shock," SECY-82-465, November 23, 1982.
- 12. Commonwealth Edison Letter from H. E. Bliss to S. Anderson (Westinghouse),
"IS0 TOTE and 20 Calculated Edits of the Assembly Average Burnup", June 22,1984. l l l i i i 1297E:10/121985 55
APPENDIX A POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the Zion Units 1 and 2 pressure vessels were derived from the following fuel cycle design reports and verified by comparison with burnup i l data supplied by Commonwealth Edison [12]. l Fuel Cycle Unit 1 Unit 2 1 WCAP-7675-R1 WCAP-7675 2 WCAP-8616 WCAP-8881 3 WCAP-9114 WCAP-9246 4 WCAP-9356 WCAP-9458 5 WCAP-9568 WCAP-9687 6 WCAP-9859 WCAP-9959 7 WCAP-10047 WCAP-10282 A schematic diagram of the core configuration applicable to Zion Units 1 and 2 is shown in Figure A.1-1. Cycle averaged relative assembly powers for each operating fuel cycle of Zion Units 1 and 2 are listed in Tables A.1-1 and A.1-2, respectively. On Figure A.1-1 and in Tables A.1-1 and A.1-2 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-1297E:10/121985 A-1
survaillenco capsules er at tha prsssure v:ssol. Therefere, power distribution data for assemblies in Region 1 are not listed in Tables A.1-1 and A.1-2. 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. I l 1, 1 l l 1297E:1D/102185 A-2 l- --
Figure A.1-1 Zion' Units 1 and 2 Core Description for Power Distribution Maps 1 2 3 4 7 8 9 10 5 6 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 l 26 27 28 29 30 Region 1 Assemblies 17 - 31 Region 2 Assemblies 7 - 16 l Region 3 Assemblies 1-6 31 t A-3
TABLE A.1-1 CORE POWER DISTRIBUTIONS USED IN THE PLANT SPECIFIC FLUENCE ANALYSIS ZION UNIT 1 Fuel Cycle j Assembly 1 2 3 4 5_ 6 7 1 0.81 0.98 0.98 0.91 1.02 0.96 1.00 2 0.89 0.98 1.01 0.95 1.01 0.99 1.03 3 0.76 0.92 0.95 0.89 1.01 1.01 0.95 4 0.66 0.73 0.75 0.73 0.86 0.80 0.47 5 0.93 0.99 1.03 1.06 1.10 1.10 0.64 6 0.56 0.63 0.71 0.71 0.74 0.77 0.39 7 1.04 0.97 0.87 0.83 0.92 0.88 0.79 8 1.05 1.16 1.23 1.17 1.16 0.91 1.13 9 1.01 0.96 0.99 0.96 1.01 1.19 1.13 10 1.02 1.18 1.14 1.16 1.19 1.11 1.05 11 1.17 0.94 0.82 0.86 0.82 1.08 0.89 12 1.12 0.95 0.87 'O.89 0.92 1.15 1.01 13 1.15 0.95 0.93 0.90 1.29 1.13 1.23 14 1.08 1.11 1.19 1.01 1.02 1.02 1.08 15 0.97 0.99 0.99 1.19 1.08 1.16 1.13 16 1.04 0.90 1.12 1.01 0.84 0.99 1.04 1 i i 1297E:10/102185 A-4 ---,--.m.- ,-,__,,c__
TABLE A.1-2 CORE POWER DISTRIBUTIONS USED IN THE PLANT SPECIFIC FLUENCE ANALYSIS ZION UNIT 2 Fuel Cycle Assemb1v 1 2 3 4 5 6 7 l 1 0.80 1.04 0.89 0.59 0.95 0.88 0.72 O.87 1.05 0.92 0.81 0.95 0.99 0.92 3 0.75 0.98 0.85 0.83 0.90 0.91 0.85 4 0.65 0.77 0.68 0.70 0.70 0.47 0.46 5 0.92 0.99 0.98 1.01 0.98 0.91 0.96 6 0'.53 0.64 0.67 0.73 0.72 0.41 0.43 7 1.02 0.96 0.87 0.77 0.88 0.89 1.19 8 1.03 1.20 1.18 1.16 1.11 1.12 1.17 9 1.00 0.98 0.94 0.92 0.89 1.16 0.90 10 1.00 1.20 1.14 1.14 .1.05 1.09 1.12 11 - 1.17 0.87 0.93 0.98 1.05 1.09 1.21 12 1.12 1.03 1.00 0.98 0.97 0.99 1.07 13 1.15 0.94 1.10 1.01 1.19 1.18 1.24 14 1.08 0.96 1.12 0.97 0.92 1.15 1.01 15 0.97 0.88 1.01 1.03 0.88 0.91 1.04 16 1.02 0.87 1.06 1.16 1.11 0.97 1.05 1297E:1D/102285 A-5
APPENDIX B WELD CHEMISTRY Tables B.1-1 through B.1-4 provide the weld data output f rom the WOG Material Data Base. Given are the searches of all available data for the wire heat in the Zion Units 1 and 2 reactor vessels beltline region. 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 g in the RT analysis. PTS Weld Chemistry Data Source and Plant: AN1 Arkansas Nuclear 1 BAW-1799 - Babcock & Wilcox Report Number B&W Babcock & Wilcox COM Zion 2 CR3 Crystal River 3 Cu Weight % of Copper CW'E Zion 1 ESA Emission Spectrographic Analysis FLA Turkey Point 4 FPL Turkey Point 3 MATSU2V NRC Mender MATSURV Data Base MPC Materials Properties Council Data Base Ni Weight % of Nickel DC1 Oconee 1 P Weight % of Phosphorous RGE Robert Emmett Ginna RS1 Rancho Seco 1 SC Surveillance capsule Si Weight % of Silicon TMI Three Mile Island 1 VIR Surry 1 WEP Point Beach 1 WMQR Weld Metal Qualification Retest WQ Weld Qualification s 1297E:10/102185 B-1
TABLE B.1-1 ZION UNIT 1 INTERMEDIATE AND LOWER SHELL LONGITUDINAL WELDS CHEMISTRY FRUM WOG MATERIALS D WIRE HEAT NUMBER 8T1762 ID WIRE WIRh FLHX FLUX WELDCHtM Cu Ns. P St PLANT DESCRIPTION HEAT TYPE TYPE LDT DATA SutK4CE ~~ i 0286 GTl762 MN-MO-NI LINDE WO 8632 BW,WQ O.200 0.650 0.009 0.530 CR3 INTER SHELL LONG CIT INTEM SHELL LONG Cd LOWER SHELL LONG TM1 INTER SHELL LONG VIR LOWER SHELL LONG sure Hil762 MM-NO-NI lINDE 80 BSS3 BW.WQ O.800 0.650 0.087 0.430 OCl LOWER SHELL LONG WE P NOllL E TO INTER GHELL 027tl GTl162 MN-NO-NI LINDE 80 8650 BW.WQ O.105 0.4SO O.004 0.390 ANS INTER SHELL LONil ANS LOWE R liHI LL LONG CR3 INTEN SHELL LONG 0335 eTl762 MN-MO-Ni LINDE 90 8397 BW WO O.570 0.530 0.017 0.500 CWE INTER SHELL LON(i VIR LOWER 66f LL LDNQ 0336 0T5762 MN-MO-NI LINDE 80 8553 BAW-l?99 WQ O.560 0.600 0.087 0.430 OCl LOWER SHELL. LONG WEP NOlFLE TO INTER SHELL l m N 0337 ST1762 MN-MO-NI LINDE 80 8596 BAW-l?99,WQ O.220 0.600 0.015 O.430 CR3 LOWER SHELL LONG g o772 8T8762 MN-MO-Ni LINDE 80 8578 BAW-1799,WO O.220 0.430 0.057 0.460 O.379286 0.547143 0.083754 0.454296 0.040252 0.078467 0.005587 0"049686 l mean std.dev.
- Since these values are limited to the weld metal qualification test reports, report BAW-1799 recommends that a mean value of copper equal to 0.29 wt% with a standard deviation equal to 0.07 wt% be applied because a bias exists in the retest of similar type welds.
- BAW.1799 recommends that a mean value of nickel equal to 0.55 wt%
be applied.
TABLE B.1-2 ZION UNIT 2 LOWER SHELL LONGITUDINAL WELD CHEMISTRY FROM WOG MATERIAL DATA BASE - WIRE HEAT NUMBER 72102 ID WIRE WIRE Flux FL ux WEL DCHEM Cu N4 P 56 PLANT DESCRIPTION HEAT T YM TYPE LOT DATA tH alRCE 0217 72302 MM-MO-N! LINDE 90 6650 SW,WQ 0.860 0.270 0.087 0.420 CDM INTER SHELL LONG RSI INTER SHELL LONS RSI LOWER SELL LONG 0245 723a2 f1N M -NI LINDE 90 0479 BW, W 0.250 0.530 0.022 0.470 MN-MO-N! LINDE SG 8650 isAW-l?9?.WQR 0.250 0.630 0.085 0.420 COM INTER ENLL LONG u?64 72102 ~ RSS INTER ShELL LONG RSS LOWER SHELL LONS O.I93333 0.476&&7 0.OI8000 0.436447 mean 0.029868 0.895834 0.003406 0.029868 etd.dev. e,n w
- BAW-1799 reconnends that a mean value of copper equal to 0.23 wt% with a standard deviation equal to 0.07 wt% be applied based upon retest of weld metal qualification test samples for similar welds.
- BAW.17'9 reconnends that a mean value of nickel equal to 0.63 wt% be applied.
L
TABLE B.1-3 ZION UNIT 2 BELTLINE CIRCUMFEP.ENTIAL WELD Cl!EMISTRY FROM WOG MATERIALS DATA BASE - WIRE HEAT NUMBER 71249 - - - - - - - - - -.. - ~ - ~ - - - - - - - - - - - - - - - - - - - - - -. - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ID WIRE WIEE F t. tax FLua WLtDQEM Cu Na P St PLANT DESCRIPTION etat TVPE TYPE LOT DATA e StastCE -w 0219 78249 MN-MO-NI LINDE 80 3733 SW.WQ 0.190 0.660 0.021 0.450 CID9 INTER TO LOWER SHELt. ER3 MOlitE TO INTER SHELt. 0223 78249 fee-MO-N I LINDE 80 8445 ins.WG 0.200 0.570 0.028 0.520 FLA INT 6R TO LOWER 9 ELL FPL INTEA TO LouER SetLL FPL SURVEILLANCE tee 10 REsE NO22LE TO INTER SHEL.L teep INTER TO LOWER SHELL 0243 73249 res-MO-NI LINDE SO 8669 bed, WQ. 0.230 0.550 0.082 0.430 0273 78249 MN4tO-MI LINDE 80 9457 8es,WG 0.230 0.550 0.020 0.580 FLA
- SUHVEILLAfCE WELO v296 78249 MM-MO-MI LINDE SG S445 FPL,9C 0.330 0.570 0.033 0.660 FLA INTE R TO LOuER 3 HELL FPL INTER TO LOWER SenLL FPL SURVEILLANCS WELD FeSE NO22LE TO INTER SHLLL WEP IseTER TO LOWER SHELL 0297 78249 MN-f40-N B LIpeDE 80 9457 FLA,6C 0.300 0.600 0.084 0.500 FLA SUNVEILL ANCE WELD 0454 78249 MM-f4D-N s LIseDE 80 8445 SAme-8799,ESA 0.160 0.550 0.089 0.540 FLA IteTER TO LOWER SonLL FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD kGE NO22LE TO INTER SHELL WE'P INTER TO LOWR SHELL 0455 78249 Pee-MO-NI LIpeDE SO S445 SAW-t?99 ESA 0.850 0.540 0.080 0.550 FLA IseTER TO LOWER SDELL G3 FPL INTER TO 8 MR SPELL h
FPL SURVEILLANCE WELD FeSE NDZZLE TO INTER SefLL edEP INTER TO LOWER SHELL 0456 73249 fee-fG-MI LipsSE SO S445 SAW-8799,ESA 0.880 0.550 0.019 0.540 FLA INTER TO LOWER SHELL FPL INTE.R TO LOdR SHELL FPL St#1VEILLANCE WELD RSE NO22LE TO INTER SHELt. esEP INTER TO LouER SHELL 0457 74249 fee-f40-MI LINDE 80 8445 84W-8799,ESA 0.890 0.540 0.019 0.610 FLA INTER TO LOWER SHELL FPL INTER TO LOWER SHELL FPL SURVEILLANCE WELD RGE NO22LE TO INTER SHELL WEP INTER TO t ObsER SELL 04*J 74 249 Pe6-MD-N I LINDE SO S445 SAW-8799,ESA 0.850 0.550 0.020 0.600 FLA Ips1Lk TO LuuER 6*ft L F H. INTER TO LOWER SHELL FPL SURVEILLANCE WELD RSE NO22LE TO INTER G6 ELL tdEP INTER TO LOWER SonLL 0459 78249 fee-fq0-N! LIDoDE 80 8445 SAW-8799,ESA 0.170 0.540 0.019 0.620 FLA INTER TO LDutR tiHELL FPL INTER TO LOWER SHELL FPL SuWEILLANCE ndELD stGE NO22L E TO INTER SHELL edEP INTER TO LowFFt SHELL 0460 78249 res-MO-NI LiteDE 80 9445 84W-8799,ESA 0.200 0.540 0.020 0.630' FLA IN1ER TO LOWER tiHELL FFt INTER TO LONEk SHELL FFt SUNVEILLANCE 6dkLD RGE' NOZZtf TO INIER St(LL WE P IN18.h TO LOMER SHEtL u461 73249 fee-MO-NI LIseDE 80 8445 EeAW-8799,ESA 0.200 u.540 0.019 ~ u.63o Fs A INIE R 10 Line k SHE L4 FH INit h fit i OWE N GHELL Fet fdlNVE IL LWe.1 edE 4.D edd Nu22LE 50 INILM LHL L t.
TABLE B.1-3 (c:nt'd) WE P B Nit R TO LOWER SHELL 0462 78249 NN-f10-NI LINDei 90 9445 IsAW-I F99,ESA 0.230 0.520 0.087 0.62H fl A INTER TO LOWER SHI:LL FFt INTER TO t OWEh SHELL FFt SitKWE ILS ANCI WELD
- EE IJOllLE TO INTER SHELL WE P INTER TO LOWER SHELL 0443 78249 MN-MO-NI t.INDE 80 is7'8 DAti-l F99,E SA 0.380 0.640 0.021 0.550 LOM INith TO LOWER SHELL ER3 NOllLE TO INTER 66 ELL 0464 75249 f1N-MO-N I LINDE 80 87 3ag Babe-I F99,E SA 0.270 0.640 0.028 0.550 COM INTER TO LOWER SHELL CR3 NOllLE TO INTEft SHELL 0465 7t249 MN-MO-NI LINDE SO 9738 BAW-8799,ESA 0.270 0.640 0.020 0.560 COM It.TER TO LOWER SHELL CR3 NOllLE TO INTER SHELL 0466 71249 MN-MO-NI LINDE 80 8738 Bate-8 799,ESA O.280 0.640 0.028 0.560 COM TNTER TO LOWER SHLLL CR3 NOllLE TO INTER SHELL 0467 78249 MN-MO-N!
LINDE 80 9734 DAte
- 4 799,EinA 0.000 0.640 0.028 0.550 COM INTER TO LOWER SHELL CR3 NOllLE TO INTER 96 ELL 0464 73249 NN-MO-NI LINDE 80 9738 DAte-1799,ESA 0.380 0.640 0.020 0.560 COM INTER TO LOWER SHELL CR3 NOllLE TO INTER SHELL 0469 73249 MN-MO-NI LINDE 80 9734 DAW-1799,ESA 0.280 0.640 0.020 0.550 COM INTER TO LOWER SHELL CR3 NOllLE TO INTER SHELL 0470 78249 MN-MO-NI LINDE 80 8734 DAu-1799,ESA 0.700 0.630 0.028 0.550 COM INTER TO LOWER SHELL CR3 NOl2LE TO teeter SHELL 0478 71249 MN-NO-NI LINDE SO 9738 8 TAW-l ?99,E SA O.270 0.630 0.020 0.550 COM INTER TO LOWER SHELL CR3 NOllLE TO INTER SHELL 0472 73249 NN-NO-MI LIreDE 80 8738 BAW-4799,ESA 0.260 0.620 0.089 0.550 COM INTER TO LOWER SHELL CR3 NO2TLE TO INTER SHELL 0473 78249 MN-MO-NI LINDE 80 9734 DAte-8799,ESA 0.240 0.620 0.020 0.560 COM INTE R TO lot 4 R SHE LL CR3 NOi!LE TO INTER 9 HELL 0474 78249 MN-MO-NB LINDE 80 0738 DAnf-t F99,ESA 0.280 0.620 0.020 0.570 COM INTER TO LOWER G& ELL g
CR3 NOZZLF TO INTER SHELL L73 0475 78249 MM-MO-NI LINDE 30 9738 Band-1799,ESA O.290 0.620 0.020 0.570 COM INTER.O LOWER S> ELL e CR3 NOZ2LE TO INTER SHELL 0476 78249 f#e-MO-M I LINDE 80 5738 Band-8799,ESA C.270 0.620 0.089 0.540 COM INTER TO LOWER SHELL CR3 NO22LE TO INTER SHELL 0477 73249 MM-MO-MI LINDE 80 8734 Dame-1799,ESA 0.290 0.630 0.020 0.570 COM INTER TO 8 M R SHELL CR3 NOl2LE TO INTER SHELL 0478 73249 MN-MO-MI LINDE 30 0738 BAW-1799,ESA 0.300 0.630 0.021 0.500 COM INTER TO LOWER SHELL CR3 NOZZLE TQ INTER 66 ELL 0479 73249 MN-MO-MI LINDE SO G738 BAW-1799,ESA 0.340 0.630 0.020 0.340 COM INTER TO LOWER SHELL CR3 NOl2LE TO INTER S6 ELL 0400 78249 MM-MO-NI LINDE 90 8738 DAnd-8799,ESA 0.200 0.620 0.020 0.570 COM INTER TO LOWER SHELL CR3 NOZILE TO INTER StELL 0401 78249 MN-MO-MI LINDE 80 8738 DAnd-1799,ESA 0.290 0.620 0.Cl9 0.540 COM INTER TO LOWER SHELL CR3 NOZILE TO INTER SHELL 0402 78249 MN-MO-NI LINDE 80 3738 Bald-1799,ESA 0.380 0.620 0.020 0.560 COM INTER TO LOteER GHELL CR3 NO2TLE TO INTER SHELL 0453 71249 f#e-MO-NI LINDE 80 0734 emed-1799,ESA 0.280 0.620 0.020 0.570 COM INTER TO LOWER SHELL CR3 NO22LE TO INTER 96Et t 04e4 74249 MM-MO-MI LINDE.40 0738 DAnd-1799,ESA 0.700 0.634 0.019 0.500 COM INTER TO LOWER SHELL CH3 NO2FLE TO INTER 66 ELL 0405 78249 MM-NO-NI LINDE 80 9738 DAW-t?99,ESA 0.270 0.630 0.080 Q.550 COM INTER TO LOWER IpfLL CR3 NOllLE TO INTER SHELL 0486 75249 MN-MO-NI LINDE 80 9738 bAhe-1799,ESA O.380 0.630 0.088 0.560 CDM INIEk TO LOWER S& ELL CR3 NO27tE TO INTER SHELL 0407 78249 MN-MO-NB LINDE 80 0714 Dame-8799,ESA 0.350 0.630 0.010 0.560 COM INTER TO LOWEN SHL LL CR3 NO2TLE TO INTER SHE LL 0498 18249 MN-MO-NI t.INDE 80 9738 BAW-8799,ESA 0.290
- 4. es 30 0.nle 0.560 LOM INTER TO L OWER tiHELL Ch3 NO27LE 90 INTER is>EL L 0773 78249 MN-f10-NI LINDE Sca e492 DAW-179y,teo (s.2 se o.5 7(o 0.023
- 0. 4 %e 0775 73749 MN-MO-NI t INDE See D445 f f t.,SC
%. " Tee & I #4 INff h 70 tot 4R SHFSt. &St IN18R T O L OWLR GHE L L tii f A 4M 14 4 AP4 f WI I D i
TABLE B.1-3 (cont'd) 3 6 04 Nestit C 30 INit es GiLLL 6d I-INikk 10 t MFr D E.Lt. 0774 18249 Pue-HO-N g LINDE t00 44443 FPL,bc
- o. 34*'t et A BHit.H TQIOWLH SWL1 FFt INTt N 10 E M b (dittL FPL id SWEILL #deCE. tde t D
~ Rf*. Ntsilt k. TO JNTER DetL NEP INTEM TO LOMth tiHtti 0777 78249 PW4-MO-N3 LIse0E 80 8445 FPL,6C. - u.320 FLA INTER TO LOWik sieELL Fet. INTEh TO Landkk SHEll FPL SiW4 ILt.AMCE TEE.D RG6 NDilLE TQ INTEN EHL&L DELP INTER To a wat SDELL ascern O.260444 0.602388 0.Ot9843 0.5569u5 mtd.dev. 0.053338 0.04u133 0.u02893 0.0474B05 e e Q% 4 e e
l l ZION UNIT 1 BELTLINE CIRCUMFERENTIAL WELD AND ZION UNIT 2 LOWER SHELL LONGIT TABLE B.1-4 CHEMISTRY FROM WOG MATERIALS DATA BASE - WIRE HEAT NUMBER 72105 l l
==============================**=*=============*======**========**===========
,==,... =
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===.==u==. ee============= .ID WIFEE WIRE FLUX FLUM WELDCHEM Cu NI P 98 PLANT DESCRIPTION HE AT TvFE TYPE LOT DATA SOBJHCE -~= 0289 72305 f#4-MO-N I LINDE 80 8669 lud.WQ 0.270 0.460 0.084 0.400 COM LDWER SHELL LONS CR3 INTER TO LOWER SHELL CWE INTER TO LOWER SHELL FLA N0!ZLE TO INTER SHELL OC3 INTER TO LOWER SHELL RSS LOWER SHELL LONG TMS NO22LE TO INTER SELL 0274 72105 MN-MO-NI LINDE 80 9773 BW.WO 0.300 0.490 0.020 0.560 COM SURVEILLANCE ELD CR3 CNE SLAtVEILLADCE ELD DC2 SURVEILLANCE WELO OC3 SURVEILLANCE WELD 0294 72305 MN-MO-NI LINDE 90 9773 CWE,SC 0.350 0.570 0.020 0.690 COM SURVEILLANCE WELD CR3 Ckt SURVEILLAsct ELD OC2 SURVEILLANCE WELD OCJ SURVEILLANCE WELD 0295 72805 MN-MO-NI LINOf 80 9773 COM.SC O.290 +N 0.017 0.470 COM SURVEILLANCE WE'LD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD DC3 SURVEILLANCE WELD CD 0403.72105 MN-NO-NI LINDE 80 9669 BAW-l?99,ESA 0.430 0.590 0.021 0.650 LOM LOWER SHELL LONG. Ch3 INTER TO LOWER SHELL I CWE INTER TO LOWER SHELL N FLA N022LE TO INTER SHELL OC3 INTER TO LOWER SHELL RSS LOWER SELL LONS TMI N0llLE TO INTER SELL 0404 72105 MM-tW1-N! LINDE 80 9669 PAW-1799,ES4 0.420 0.590 0.020 0.600 COM LOWER SE LL LONG CR3 INTER To LOWER SHELL CWE INTER TO LOWER SELL FLA N0ZZLE.70 INTER SHELL OC3 INTER TO LOWER SHELL RSS LOWER SELL LCDSB TML NOI2LE TO INTER SELL 0405 72105 MN-MO-NI LINDE 80 8669 BAW-1799,ESA 0.400 0.590 0.020 0.600 CfM LOWER SNELL LONO CR3 INTER TO LCtdER SELL CWE INTER TO LOWER SHELL FLA NO22LE TO INTER SHELL OC3 INTER TO LOWER SHELL RSl LOWER SHELL LONG TMt N022LE TO INTER SHELL 0406 72805 NN-MO-N! LINDE 80 8669 BAW-1799.ESA 0.390 0.590 0.019 0.560 COM LOWER SHELL LONO CR3 INTER To LOWER SHFLL CWE INTER TO LOWER SE LL FLA N0llLE TO INTER SHEll DC3 INTER TO LOLKR SHELL Riit LOWFR fiHELL I ONO TMI NullL E 10 INTER SHCLL 0407 72805 MN-MO-N! 8.INDE 90 0669 BAW-1799,ESA 0.350
- 0. =M0 0.019 0.530 COM LHWFR GHELL 10NG CR3 INTER TO LOWER fiHELL CWE INTER TO LOWER SHELL FI A NOllt E TO INTER SHELL
TABLE B.1-4 (cont'd) OC3 INTER 70 TOE R SHEtt RSS LOWER SHELL LONG TMS N0lltE TO INTER UHELL 0400 72805 MN-MO-N! 8.INDE 80 9669 8%W-4 799.ESA 0.350 0.580 0.089 0.540 CON LOWER SHELL LONO CR3 INTER TO LOWFR SHELL CWE INTER TO LOWER SHELL FLA NOllLE TO INTER SHELL
- DC3 INTER TO LOWER SHELL RBI LOWER SE LL LONG TMl N0Z2LE TO INTER BELL 0409 72805 MN-MO-N!
LINDE 80 8669 9tM-1799.ESA O.390 0.5e0 0.089 0.550 COM LOWER SHELL LONG CR3 INTER TO LOWER SHELL CNE INTER TO LOWER SELL FLA N02!LE TO INTER SHELL OC3 INTER TO LOWER SELL RSI LOWER SELL LONS TMl N0ZZLE TO INTER SHELL 0410 72805 MN-MO-NI LINCE 90 8669 BAW-8799.ESA 0.370 0.590 0.085 0.540 COM LOWER SELL LONG CR3 INTER TO LOWER SELL CWE INTER TO LOWER $4 LL FLA NO!!LE TO INTER BHELL OC3 INTER TO LOWER SHELL RSS LOWER S ELL LONG TM1 NDZILE TO INTER SHELL 0411 72805 MN-MO-NI LINDE 80 8669 DAW-1799.ESA 0.360 0.590 0.089 0.530 CON LOWER SELL LONS CR3 INTER TO LOWER SHELL CNE INTER TO LOER SELL FLA NOTZLE TO INTER SHELL OC3 INTER TO LOWER SELL RBI LOWER SELL LEMB TMS NO22LE TO INTER SELL m e 0482 72105 MN-MD-NI LINDE 80 8669 DAW-1799.ESA 0.490 0.590 0.018 0.530 COM LOWER SELL LONG C3 CR3 INTER TO LOWER SELL CE INTER 10 LOER SELL FLA ND2FLE TO INTER SHELL OC3 INTER TO LDER SHELL RSS LOWER SHELL LO W TMt NO22LE TO INTER SHELL 0483 721,05 MN-MO-NI LINCE 80 8469 BAW-8799.ESA 0.470 0.600 0.089 0.520 COM LOER SELL LONG CR3 INTER TO LOWER SELL s CWE INTER TO LOE R SELL FLA NO22LE TO INTER SHELL OC3 INTER TO LOWER SELL RSI LOWER SHELL LONG TMS NO22LE TO INTER SELL 0414 72105 MN-MO-MI LINDE 80 8469 DAW-1799.ESA 0.470 0.650 0.087 0.490 CON LOWER SELL LONG CR3 INTER TO LDMER 9 HELL CME INTER TO LOWER SHE*d FLA NO27LE TO INTER SHELL OC3 INTER TO LOWER SHELL RSI LOE R SHELL LONG TMI NOFFLE TD INTER SELL 0415 72505 MN-MO-NI LINDE 90 8669 BAW-1799 ESA 0.490 0.650 0.089 0.490 COM LOWER SHELL LONG CR3 INTER TO LOWER SELL CWE INTER TO LOER 94LL F1 A NDilLE 70 INTER SHELL OC3 INTFR TO LOWER SHELL RS$ LOWER GHELL LONG TMt Nntit E TO INTER SELL f*416 72805 MM-MO-NI 4.INDE Rn BAA9 BAW-1799.E SA 0.470 0.680 0.017 0.490 COM LOWFR SHELL LONG CR3 INTE$r 10 LOWFR flNELL CWE INTER TO LOWER SHELL
TABLE B.1-4 (c:nt'd) FLA NOffLE TO INTFR SHFLL OC3 INTER TO LOWER GHELL RSS LOWER SHELL LONO TMt NOllLE TO INTER SHELL 0417 72105 MN-NO-NI LINDE 80 9669 BAW-1799,ESA 0.44n O.600 0.089 0.490 COM LOWFR SHELL LONG CR3 INTER TO LOWER SHELL CWE INTER TO LOWER SHELL FLA NO22LE TO INTER SHEL.L OC3 INTER TO LOWER SHELL RSI LOWE R SNLL LONG TMl NOZILE TO INTER SHELL 041R 72805 MN-MO-NI LINDF BG 8773 BAW-8799,ESA 0.350 0.590 0.023 0.690 COM SURVEILLANCE WELD CRS CWE SURVEILLANCE WELD OC2 SURWILLANCE WELD OC3 SURWILL ANCE KLD 0449 72105 MN-MO-N, LINDE 80 9773 DA& l?99,ESA 0.360 0.590 0.022 0.640 COM SURVEILLANCE WELD CR3 CNE SURVEILLANCE WELD OC2 9URVEILLANCE WELD OC3 SURWILLANCE WELD 04?O 72805 NN-MO-NI LINDE 80 8773 BAW-1799,ESA 0.350 0.580 0.028 0.630 COM SURVEILLANCE WEL' CR3 CWE SURVEILLAN(l' WELD OC2 SURWILLANCE FLD OC3 SURVEILLANCE 16 %9 0428 72105 MN-MO-HI LINDE 90 8773 BAW-8799,ESA 0.360 0.590 0.023 0.690 COM SURVEILLANCE WILD CR3 CWE SURVEILLANCE dELD* f OC2 SURVEILLANr$ tdELD e OC3 SLRVEILLANCE 4dELD 0422 72805 MN-MO-NI LINDE to 9773 BAW-8799,ESA 0.360 0.580 0.021 0.600 COM SURVEILLANCE WELD CR3 CWE SURVEILLAfCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE ndELD 0423 72805 MN-NO-NI LINDE 9e 8773 BAW-1799,ESA 0.360 0.570 0.022 0.640 COM SURVEILLANCE WELD CR3 CWE SLAIVEILLANCE DELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0424 72105 MN-MO-NI LINDE 80 8773 BAW-8799,ESA 0.370 0.590 0.089 0.540 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE %dELD OC2 SURVEILLANCE edELD OC3 SURVEILLANCE WELD 0425 72105 NN-MO-NI LINDE 80 8773 SAW-l?99,ESA 0.350 0.680 0.087 0.530 COM SimVEILLANCE WELD CR3 CWE SURVEILLANCE tELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0426 72105 MN-MO-MI LINDE 90 8773 BAW-1799,ESA 0.370 0.600 0.019 0.560 COM SURVEILL ANCC WELD CR3 CWE SURWILL ANCE WELD OC2 SURVEILLANCE WELD OC3 StRVF ILL ANCE bdE LD 0427 72805 MN-MO-NI LINDF 90 9773 RAW-l?99,F9A o.330 0.620 0.019 0.540 COM fauRWILLANCE WFLD CR3 CWE SURVEILL ANCF WEL D OC2 fuRVEILL ANCE tFLD (10 3 SLWiVEILL ANCE WFLD
2 e s 2 NYYW Wi$$ eses es-e e 2 e WWWW WWWW W e 223 ssse s es $$$$ WNNW IWIW e esss eIWY WWWWWWWW WYWW W WWWW W g 111111l11l,11,11111111111l1ll111111111111111 sta=
- mas _
_e
- a a
e i !!Il !!il lill lill lill 1161 lill lill lill 111! !!!! !!il ! i sagisagisnaisugasussanassassangssnaisusssagismin a s s a b e e e 2 e e e e e e e e e e e d d d d d d d d d d d d d i ~ b O O O O O O C C d b O d d i d d d d d 0 0 C O O O O O O O. O C ^ }30 TR A 2 2 8 2 2 2 2 8 2 8 A o m e e n m 3$ 5 5 5 d d d d d d d d d is '5 5 8 3 5 3 3 3 5 5 8 5 h h_ h_ h_ h_ N N m-w. N E E E E u u E E E E u E E E 3 E E E E 2 3 S 3 2 3 8 8 8 8 8 8 8 8 8 8 8 8 2 I I I - I I I I I I I -sI 3 3 3 s a a a a a s E 1 5 5 5 5 5 5 5 5 5 1 ~ h h h h h h h h h E E E E E E E E E E E E E ese u u u u u u u u u u u u u l u 7, 12 4 E A ? o 3 3 3 3 3 3 3 3 3 3 3 3 l B-10 l - ~
l TABLE B.1-4 (c:nt'd) OC2 91RVFILLANCE KLD G.3 BLRVE BL L ANCE WF L D 0441 72 8 m'5 MN -MO-- N 1 1.INDF Ho 0 773 BAW-1799.E SA 0.200 0.590 0.016 0.550 COM SIRWILLANCE WELD CR3 CWE SLRVF ILL ANCE WELD OC2 SURVF ILLANCE WLD OC3 SURVE IL LANCF WELD 0442 12805 NN-MO-NI LINDE On 9773 BAW-1799.E H4 n.3Oo 0.500 0.036 0.560 COM 9ffWffLLANCL WELD 0943 CWE St9WILLANCE WL D OC2 9JRVLILLANCE KLD OC3 SURVLILLANCE WELD 0443 728u*5 NN-MO-NI LINDE 80 8773 BAW-8799.ESA 0.300 n.590.0.016 0.560 COM SURWILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD UC3 SURVEILLANCE WELD 04a4 y23o5 MN-MO-NI LINDE 80 G773 BM-1799.ESA 0.270 0.500 0.024 0.770 COM RURVEILLANCE WELD CR3 CWE SURVEILt ANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0445 72805 NN-MO-N! LINDE 80 8773 BN-1799 ESA 0.300 0.500 0.023 0.690 COM SUR'KILLANCE WELD CR3 CWE SURVEILLANCE WELD. OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0446 72301 MN-MO-NI LINDE 80 G773 DN-1799.FSA 0.380 0.500 0.022 0.660 COM SURVEILLANCE WFLD CR3 CWE SLRVEILL.ANCE WELD OC2 SURVEILLANCE WELD f. OC3 SURVEILLANCE WELD W 0447 72105 NN-MO-N! LINDE 90 8773 BAW-1799.ESA 0.320 0.500 0.022 0.660 COM SURVEILLANCE WELD CR3 g CWE SURVEILLANCE WE1D OC2 EUPVEILLANCE WELD OC3 SURVEILLANCE WELD n440 72105 NN-MO-MI LINDE 80 e773 BAW-1799.ESA O.290 0.570 0.08G 0.690 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD Q449 72805 NN-MO-NI LINIE 80 8773 BAW-1799.E SA 0.300 0.S90 0.017 0.590 CON SURVEILLANCE WELD CR3 CWE SURUEILLANCE WELD OC2 GURtKILLANCE WELD OC3 SURVEILL ANCE WELD 0450 72105 MN-NO-N! LINDE 80 8773 D N -8799 ESA 0.290 0.570 0.086 0.570 COM SLRVEILLANCE MLD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE K LD OC3 SURVEILLANCE WLD 0451 72105 NN-NO-NI LINDE 80 8773 DAW-1799.ESA 0.290 0.500 0.087 0.600 COM SURVEILLANCE WELD CR3 CK SURVEILL ANCE WEL D OC2 S ERVE It.L ANCE WE D OC3 SLRVE ILL ANCE WFL D n4S? 72tn3 MN-MO-N! LINDF B0 R173 BAW-1799.F *iA 0.300 0.500 0.006 0.600 (~f 'M fM@VFILLANCE WELD Ch' OF St RYF lL t ANCE WE L D inC7 m ovr IL i ANCF Wrt D Ut3 9RVE ILL ANs F WFLD
TABLE B.1-4 (cont'd) 04S3 72105 MN-MO-Ni LINDE 80 U/73 l* Ate-8 79V.EfiA U. 2FM 0.50) 0.017 0.aOO COM fiURVEILL ANCE WELD CH3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SUNVEILLANCE WELD l A6*,8 72805 MN-MO-NI LINDE 80 9773 MPC,DB,0C2,SC 0.360 0.580 0.022 0.650 COM SURVEILLANCE WELD CR3 CWE SURVEIL. LANCE WLD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD Gasy 72305 NN-MO-MI LINDE to e773 MPC,DS,0C3,5C 0.300 0.500 0.017 0.610 COM SURVEILLANCE WELD CR3 CNE SURVEILLANCE WELD OC2 StmVEILLANCE WELD OC3 SURVEILLANCE WELD 0664 72805 MN-MO-MI LINDE 90 0773 MPC.D8,CR3,9C 0.390 0.100 0.021 1.000 COM SURVEILLANCE WELD CR3 CME SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0678 72805 MN-MO-MI LINDE 80 8773 CWE,SC 0.2th 0.530 0.087 0.619 COM SURVEILLANCE WELD Ch3 CNE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0479 72105 MN-MO-NI 'LINDE 80 9773 CWE,9C 0.270 0.570 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE ELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD f 0680 72105 MM-MO-HI LINDE 80 8773 CWE.SC 0.239 0.545 0.018 0.445 COM ' SURVEILLANCE WELD w Ck3 N CNE SURVEILLANCE ELD OC2 SURVEILLANCE WELD OC3 StRVEILLANCE WELD 06e: 72805 MN-NO*NI LINDE 90 8773 CWE.SC 0.250 0.490 COM SURVEILLANCE WELD CR3 CNE SURVEILLANCE WELD OC2 BLRVEILLANCE WELD OC3 54AVEILLANCE ELD 06G2 72805 MN-MO-N! LINDE 80 B773 CWE.SC 0.260 0.540 COM SalRVEILLANCE WELD CH3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0481 72105 MN-MO-NI LINDE 90 3773 CWE SC 0.260 0.540 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0604 72405 MN-MO-NI LINDE 80 3773 CWE,SC 0.240 0.550 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE ELD OC2 SURVEILLANCE WELD OC3 StDVEILLANCE WELD 0685 72805 MN-MO-MI LINDE SO 8773 CWh,SC 0.260 0.530 COM SURVEILLANCE WELD CR3 CNE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WFLD n486 72805 194-MO-NI LINDE 00 8773 CWE.SC
- 0. 2tM 0.560 CtM SURVE ILLANCE WELD CR3
i l i TABLE B.1-4 (c:nt'd) CHE SURVEILLANCE WELD 002 SURVEILLANCE WEL3 OC3 t>URVE ILL ANCE WELD COM SURVEILLANCE WELD 0687 72I05 NN-MO-N! LINDE 80 8773 CWE.GC 0.250 0.540 CR3 CWE SURWILLANCE WFLD UC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0600 7210$ MN-MO-NI LINDE 80 8773 COM.SC 0.190 0.520 COM SimVEILLANCE WELD CN3 CWE GURWILLANCE WELD OC2 SURVLILLANCE WELD OC3 SURVE ILLANCE WELD 0689 72I05 NN-MO-NI LINDE 80 8773 COM.SC O.230 0.520 COM SURVEILLANCE WELD CR3 CME SURVEILLANCE WELD OC2 StmVEILLANCE WELD OC3 SURVEILLANCE WELD 0690 72105 MN-MO-NI LINDE 80 8773 COM,6C 0.260 0.530 0.024 0.520 COM IREVEILLANCE WELD CR3 CWE SURVEILLADCE WELD OC2 SURVCILLANCE WELD OC3 SURVEILLANCE WELD 0691 72105 MN-MO-HI LINDE 80 8773 COM.SC 0.230 0.540 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0692 72105 MM-MO-NI LINDE SO 9773 COM.SC 0.250 0.S30 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD 03 OC3 SURVEILLANCE WELD 8 W 0693 72105 MM-MO-N! LINDE 80 0773 COM SC 0.310 0.520 0.024 0.270 COM SURVEILLANCE WELD CR3 CNE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0694 72105 MN-MO-N! LINDE 80 8773 COM.SC 0.210 0.400 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0695 7210S MM-MO-N! LINDE S0 8773 COM.SC 0.290 0.550 0.026 0.490 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0696 72105 NN-MD-N! LINDE 80 8773 COM.SC O.230 0.470 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 l SURVEILLANCE WELD OC3 SURVEILLANCE WELD 0697 7210S MM-NO-MI LINDE 90 8773 CON SC 0.220 0.520 COM SURVEILLANCE WELD CR3 CWE SURVEILLANCE WELD OC2 SURVL ILLANCE WELD DC3 tIURVL IL L ANCF WE L D 0698 7210S NN-NO-NI LINDE 80 G773 COM,5C 0.251 0.560 COM EdlRVE ILLANCE WELD CR3 O*. GamVEILL ANCE WFI D OC2 SUNVF ILLANCE WFLD
l l TABLEB.1-4(cont'd) OC3 IMIRVEILL ANCE WELD 0784 72105 MN-MG-NI L int 1 So 9773 COM,SC 0.270 0.S30 COM IRNM.lLt ANCE 6dELD CH3 CWE SURVEILLANCE WEL D OC2 SURVEILLANCE WELD UC3 SURVEILLANfK WELD 0785 72805 MM-MO-MI LINDE 80 9773 COM,SC 0.260 0.540 COM SURVEILLANC9: WELD CR3 CE SL9tVEILLANCE WELD OC2 SURVEILLANCE WELD OC3 St#tvEILLANCE WELD 0744 72805 MM-MO.NI LINDE SO 3773 MPC,DS,0C2 0.340 0.600 0.083 COM StAtVEILLANCE ELD CR3 CE SURVEILLANCE edELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE E LD 0752 72805 NN-MD-NI LINDE SO 9669 DAW-l?99, WOR 0.340 0.580 0.089 0.580 COM LOWER SHELL LONG CR3 INTER TO LOWER SHELL CWE INTER TO LOWER SHELL FLA NG22LE TO INTER LL OC3 INTER TO LOWER LL RSS LOWER SNELL LONS TMS MOZ2LE TO INTER SHELL 0753 72805 Pee-MO-MI LINDE SO 8688 SAW-l?99, WOR 0.300 0.680 0.087 0.580 0754 72805 MM-MO-NI LINDE SO 8773 Sate-1799,6dQR 0.400 0.590 0.028 0.570 COM SINIVEILLANCE ELD m CR3 e CWE SURVEILLANCE WELD M* OC2 SURVEILLANCE tdELD DC3 St#tVEILLRNCE ELD 0778 72505 fee-MO-MI LINDE 80 8688 SAN-8799,WO 0.280 0.590 0.088 0.590 0774 72105 fee-MO-MI LINDE SO 8773 MPC,DS,0C3 0.290 0.590 0.087 CIM SuREiLLADCE ELD CR3 CNE StAtVEILLANCE ELD OC2 SURVEILLANCE WELD OC3 SURVEILLANCE ndELD / 0.385563 0.563506 0.089464 0.584627 mean std.dev. 0.067005 0.060055 0.002747 0.080673 4
APPENDIX C RT A S OF ZION UNITS 1 AND 2 PTS REACTOR VESSEL BELTLINE REGION MATERIALS C.1 ZION UNIT 1 values, as a function of both Tables C.1-1 through C.1-6 provide the RTPTS constant fluence and constant EFPY (assuming the projected fluences values), for all beltline region materials of the Zion Unit 1 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 i in Table 111.4-1 of the main report. Location Vessel Material 1 Intermediate shell plate 8-144-2 2 Intermediate shell plate 8-144-1 3 Lower shell plate 9-144-1 4 Lower shell plate 9-144-2 5 Intermediate to lower shell circumferential weld WF-70 6 Intermediate and lower shells longitudinal welds WF-4/WF-8 1 1297E:lD/102285 C-1
C.2 ZION UNIT 2 Tables C.2-1 through C.2-6 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 Zion Unit 2 reactor vessel. The RT values are calculated in accordance with the proposed PTS rule, which PTS is 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 111.4-2 of the main report. Location Vessel Material 1 Intermediate shell plate 8-152-1 2 Intermediate shell plate 8-152-1 3 Lower shell plate 8-152-1 4 Lower shell plate 9-152-2 5 Intermediate to lower shell circumferential weld SA-1769 6 Intermediate shell longitudinal weld WF-29 7 Lower shell longitudinal weld WF-70 l i i i I 1297E:10/102285 C-2
TABLE C.1-1 RT VALUES FOR ZION UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS PTS 18 2 9 FLUENCE = 1.0 x 10 n/cm LOC netants cu m p eersette vatw tres actuessets strets 1 cw .t:: 4es.osos so. i actuat i s.m. e.soe+sei m.: 2 cw .s s.4es.osos s. actuat i s.a. i.soe+ses oe.: 3 , e., ,.is 4es.oss 4. actuat a s.m. s.sor+eer es.: nb 4
- CWE
.tSt.if0s.olot 2o. i actuat i S.W. s. foe +fe 119. 5 cw i.3::.ses.oirs
- o. e erwenic : c.w. s.see+ e ses.:
6 ew .ois, o. eeware i t.. i,,og.,o ,st.: I I
TABLE C.1-2 RT VALUES FOR ZION UNIT 1 REACTOR' VESSEL BELTLINE REGION MATERIALS PTS 18 2 0 FLUENCE = 5.0 x 10 n/cm __'C. LO IPLAmT cu : N i e infuoras watw a svea IFLUENCEl RTPTS t cut t. ins.491.0t?? to. I ACTUAL i S.W. l.9oE*191 114.3 2 i Cw a.tal.491.olot
- 5. I ACTUAL t 3.u. t.got+ toe too,g 3
cwe .ts 4es.ots: -4. ACTUAL i s.m. i sor+1st tod.: n e b 4 I cut t.15f. Sos.otol So. I ACTUAL I s.N. g.gog+teg 140,3 5 e cw a.33:. sos.otti
- o. i erweic : c.w. e.noa+tes 327.
5 cw .se.ssi.otsi
- o. i sensene ~s L.w. i.noa+1ei s to. :
i j i 1 i l l
TABLE C.l.3 4 VALUES FOR ZION UNIT 1 REACTOR VESSEL BELTLINC MATERIALS RTPTS 19 2 W FLUENCE = 1.0 x 10 n/cm 4 P 18788T38 WALUE I TYPE OFLUE88CEI STPts I g IPLA8fft CU l NI I I CWE I.lal.495.0001
- 10. I ACTUAL I 5.W. 1.10E+201 129.1
] 2 i CW I.tal 488.0001 S. I ACTUAL i S.M. 1.50E*201 120.1 3 i CW I.52: 4e1.0521 -4. I ACTUAL I a.M. 1.10E+20I 157.1 n 4 I CWE I.158.901.0101
- 20. I ACTUAL t 5.N. 1. TOE +20f 158.8 5
I Cwc I.22:. sal.0til O. I GENERIC l C.W. 1. TOE *201 2s2.1 I 4
( TABLE C.1-4 RT VALUES FOR ZION UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS PTS 9 CURRENT (5.9 EFPY) FLUENCE e invuorie vatue i Tvet irtuences artis LOC.irtauri cu i wi i............................................................ 1 i cws i.i:: 4es.osos to. i actuat i s m. s.44e+ist ein.: 2 i cwe .ini.4ei.oio
- s. i actuai. i s.m. i.44r+ tes so7.
3 cwe .is ... oisi
- 4. i actuat i s.m. i.44e+isi sos.:
4 i cwt .its.Soi.oto: 2o. i aCTuat i S.m. i.44E+i9: ist.: n s 5 i cwe .32:.ssi.o m
- o. i newenic : c.w. i.44E+ts: 222.i ca cwe i.2s:.sst.oisi
- o. i onweenc L.w.
i.ist+ssi ses.: 6
.~ ~ J TABLE C.1-5 RT VALUES FOR ZION UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS PTS SEND OF LICENSE (25.8 EFPY) - PROJECTED FLUENCE VALUE P leifeTII VALUE f vvPE IFLUfMCE f RTPT5 I LOC.lPLANil CU f NI f I 1 i cwe i.i2: 4si.osoi so. i ACTUAL i s.m. i.ise 2os is2.s 2 I cwE I.i21 4st.oeal
- s. I ACTUAL t s.N. I.ist*2of i27.I 3
cwe t.is .dat.oist -4.
- ACTUAL s.w. i.isE*2on 32s.:
g, SN 4 I cut t.nst.sof.osol 2o. I ACTUAL l s.N. I.esE*2of is4.I 5 cwe i.32:.ses.onvi
- o. t ceNeesc : c.w. i.ese*2o 2e4.s 6 i cwt i.2s
.ssi.ossi o. osNearc i L.w. i.7oc+ s 224.s 9 T
TABLE C.1 6 VALUES FOR ZION UNIT 1 REACTOR VLSSEL BELTLINE REGION MAIERIALS RTPTS 9 32 EFPY - PROJECTED FLUENCE VALUES IPLANTI CU i N i P leTNOTIl VALUE I TVPE IFLUENCEI RMS 8 LOC............................................................ ] i Cut i.42: 4es.otol to. : ACTUAL I S.M. 1.teg.303 gas.: 2 e cw .s::.4es.otos
- s. : ACTUAL e s.m. i.iee+ oi ess.:
3 e cvs .sai.Aes.ois 4. ACTUAL e.m. e.nes+ son : e.: p 4 I cut .99:.Soi.otos to. 8 ACTUAL i S.M. .98E.301 tee.l L3 l; 5 cw .33:.ssi.oive
- o. i esmenic c.w. i.ses+ o met.:
6
- cw
.2e:.ssi.ossi o. essente L.w. i. eve +ses ass.: ) ll s [
l TABLE C.2-1 RT VALUES FOR ZION UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS PTS 18 2 0 FLUENCE = 1.0 x 10 n/cm netanti cu i m e satserie vatur a tver artinmer:RTris s LOC............................................................ ] I C000 t.529.Sil.olot
- 22. I ACTUAL I S.m.
- 1. toe +198 tos.:
2
- Com I.121.S2I olol
- 22. I ACTUAL i S.m.
- 1. toe +191 1o7.9 3
com :.12 .54 otos to. : Actunt i s.m.
- i. toe +tes os.:
4 I COel :.441.528.0008
- 2. I Actual i S.m.
i. tok'+ t9 8 94.1 n e s.3 5 e cons I.2si. sos.otes
- o. : armnic c.w.
- i. tot *ses ide.
~ 6 8 com I.22:.828.otes
- o. i se w are : L.w.
- i. toc +ssi s2s.:
7 i com .32:.ssi.o Ts
- o. i samanic : L.w.
- i. tor +,os ise.:
TABLE C.2-2 RT VALUES FOR ZION UNIT 2 REACTOR VESSEL BELTLINE REGION PTS 9 FLUENCE = 5.0 x 1018 2 n/cm 1.0C ertanti cu w .................i e inveetts vatu ..................t i tres st ............'..tuence s niets ] 3 coal I.t2l.598.0908
- 22. I Attual I S.N 3.30g+s33 g3g.g 2
I Ceu I.121.538.0008
- 22. t Actual i S.R l.90E+t90 127.8 3
8 coal I.121.548.000f
- 90. 8 Actual i S.M.
3.30geget agg,g 4 9 CON l.348.829.0001
- 2. 6 ACiual i B.N 0.90E*191 997.3
.....................................s-...... n 5 i com .:.e0s.oies
- o. i ermaic c.v. i.30e*,e,,,
a.> 6 8 **" ' " ' **' * * ' "'"* ' ' * ' ...............................................**.'.'.'.'.'.'.88' 7 i c0,.32,.s.v.0ive
- o. i new= c : t.w v.soe+1st 227.8
TABLE C.2-3 RT VALUES FOR ZION UNIT 2 REACTOR VESSEL BELTLINE MATERIALS PTS 19 2 9 FLUENCE = 1.0 x 10 n/cm ............. ortusacsintris i Tver LOC .n e.t.Am..r.t.c.u...I.w..a..i..e..s a vee..v.i.l...vatue............... 1 i Com e.i2:.stl.oson
- 32. I Actual i s.m. i.10s+20s ise.
i Case I.12:.531.010
- 22. 1 ACTUAL I 5.N.
1.10E+201 139.1 g 3 I Coal I.i2:.541.0101
- 10. I ACTUAL I B.M.
i.10E+201 127.1 n ( I COIS I.941.528.0001
- 2. I ACTUAL i S.M.
1.10E+201 131.1 4 * ..a 5
- Com i.as.e05.0tes
- 0. I sensa:C i C.w. 1.10t+201 22s.:
I TABLE C.2-4 RT VALUES FOR ZION UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS PTS @ CURRENT (5.7 EFPY) FLUENCE LOC.iPLANTI cu e at : e seisastr e vatus i tvPr artuencri arris i ] Cool .32: .5 .osol 22. ACTUAL 5.m. i.dlE+its s23.8 j 2 e coes :. ins. Sat.osos an. i Acrust i s.m. i.4 .isi ia4.s 3 com .ine.s4:.osos so. Actuat a.m. s.4:s+ si sia.: O1 4
- Coat i.id
.52s.000: 2. ACTUAL i S.m.
- l. die +its si4.8 PJ 5
a com :. ass. son.ossi
- o. i semnic : c.w.
i.4 s.ies too.: 6 e coes :.33:.szt.oisi
- o. I semenic i L.w.
- i. net +tsi i47.1 7
e com .32:.sse.os7:
- o. i erw eic L.w.
i.ier+ies 37s.: I e
o TABLE C.2-5 RT VALUES FM ZIM MIT 2 REACTM VESSEL BEWE EIM MRIRS PTS 9END OF LICENSE (25.3 EFPY) - PROJECTED FLUENCE VALUE LOC netaura cu : w e termots: wa .................................tue Tree ertuences stars 1 e com . ::.ssi.oso
- 32. i actuat i s.m. i.ise.no, is..
2 e com :. ::.sse.oso
- 22. i actuat i s.se. v.ise.ror. 34s.:
3
- ca
.,2:.e4,o,o, ,o. i actuat i s.m. i.ise.no 33s., n .............................................g.............., i 4 com .34:.52.o00,
- 2. e actual : 3 m. a.g5e*2o: 343.
s.s 5 com :.as:. sos.oisi o. osunic : c.w. i. se.2os 24s.: 6 e com .za.ess.oisi o e eenente i t... e,.3e.,,, 7 ' C"" :.32:.ser.oire o. eenemic i t... i.s3s.se 23e.:
TABLE C.2 6 RT VALUES FOR ZION UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS PTS 932 EFPY - PROJECTED FLUENCE VALUES setauri cu at :. e seter va LOC _............................ne.....tus e tres artusaces arris j i ces t.tSt. Sit.otol St. I Actual i S.N. B. tM*301 100.8 2 i ceu i.sas.nse.osos ss. : Actuat i s.m. s. M+ son ses.: t 3 I coat t g2:.548.09of lo. I Actual i B.N. t.(M*3ot too. 4 3 c088 I.tel.538.0008
- 3. I ACTUAL I a.08. g.gm eses gag,g r1 e
5 e com .m . son.oses o. eeneerc : c.w. i.,w.soi 3., 6 com .nse.ess.oses
- o. I eelmesc : L.w.
e.aos.nes ese. 7 8 co" 8.32:.sse.oire o. orwenic L.w.. sog.,e, aso. I [}}