ML20212D255
| ML20212D255 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/30/1996 |
| From: | Brassart G WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20212D250 | List: |
| References | |
| WCAP-14687, NUDOCS 9710310050 | |
| Download: ML20212D255 (68) | |
Text
.
WESTINCHOUSE NON PROPRIETARY CLASS 3 WCAP 14687 1
JOSEPH M. FARLEY UNITS 1 AND 2 RADIATION ANALYSIS AND NEUTRON DOSIMETRY EVALUATION 1
1 R. L. Bencini June 1996 Work Performed Under Shop Order AWUP-1109A Prepared tiy Westinghouse Electric Corporation for the Southern Nuclear Company 1
Approved:
L G; 'A.'Brassah, Manager Radiation Engineering and Analysis WESTINGHOUSE ELECIRIC CORPORATION P.O. Box 355 Piusburgh, Pennsylvania 15230-0355 C 1996 Westinghouse Electric Corporation All Rights Reserved h
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4 TABLE OF CONTENTS i -
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1.0 INTRODUCTION
11 2.0 DISCRETE ORDINATES ANALYSIS 2-1 3.0 NEUTRON DOSIMETRY 3-1 1
4.0
- PROJECTIONS OF PRESSURE VESSEL EXPOSURE 41 1
1
5.0 REFERENCES
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LIST OF TABLES 4
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_ILtl.t P_ mar 2 Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center 25 2-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates at the Pressure Vessel Clad / Base Metal Interface 28 2-3 Relative Radial Distribution of $(E > 1.0 MeV) within the Pressure Vessel Wall 2 11 2-4 Relative Radial Distribution of $(E > 0.1 MeV) within the Pressure Vessel Wall 2-12 4
2-5 Relative Radial Distribution of dpa/see within the Pressure Vessel Wall 2-13 l
3-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors 3-6 3-2 Monthly 'Ihermal Generation 3-7 3-3 Measured and Saturated Sensor Activities and Reaction Rates 3-12 3-4 Summary of Neutron Dosimetry Results for Surveillance Capsules 3-19 4
3-5 Comparison of Measured and Ferret Calculated Reaction Rates at the Surveillance Capsule Center-3-21 3-6 Adjusted Neutron Energy Spectrum at the Capsule Center 3-25 3-7 Comparison of Calculated and Measured Neutron Exposure Levels 3-32 4-1 Neutron Exposure Projations at Key Locations on the Pressure Vessel Clad / Base Metal Interface 4-4 4-2 Neutron Exposure Values Within the Reactor Vessel 4-6 4 Updated Lead Factors for Farley Units 1 and 2 Surveillance Capsules 4-10 m:Uo6ow.wpf:lt@0196 ii
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LIST OF FIGURES N
Fimare Tigt P.us.
J 21 Reactor Geometry Showing a 45' r,0 Sector.
2 14 4
2-2 Plan View of a Dual Reactor Vessel Surveillance Capsule 2-15 d,
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- 1.0; INTRODUCTION Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule
' geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two j
reasons. First, in order to interyet the neutron radiation induced material property changes observed in the test specimens; the neutror environment (energy spectrum, flux, fluence) to which the test
{
. specimens were exposed mur, be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established l
.between the neutmn environment at various positions withm the pressure vessel and that experienced by the test specimens. - The former requirenent is normally met by employing a combination of -
rigorous analytical techniques and measurements obtained with passive neutron N aonitors
[
contained in each of the surveillance capsules. De latter information is generally aerived solely from l
- analysis.
ne use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the maserial has traditionally been accepted for development of damage trend j
curves as well as for the implementation of trend curve data to assess vessel condition. In recent
{
years, however, it has been suggested that an exporure model that accounts for differences in neutron
{-
energy spectra between surveillance capsule locationr and positions within the vessel wall could lead l
to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall, i
Because of this potential shift away from a threshold fluence'toward an energy dependent damage function for data correlation ASD( Standard Practice E853, " Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements pr.r iron atom (dpa) along -
[
- with fluence (E > 1.0 MeV) to provide a dan base for future reference. The energy dependent dpa -
function to be used for this evaluation is specified in ASTM Standart' Practice E693, "Charactenzing -
i-Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the
. dpa p.r.J c r to the assessment of embrittlement gradients through the :Sickne::. of the guessure vem! -
wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Embrittlement i
of Reactor Vessel Materials,"
i-g his section provides an update of the dosirnetry evaluation for Capsules Y, U, X, and W for Farley
[
Unit I withdrawn at the end of cycles 1,4,7, and 12, respectively, ne update includes Capsules U,-
l W, and X for Farley Unit 2 withdrawn at the end of cycles 1,4, and 6, respectively. This update is
[
based on current state-of-the-art methodology and nuclear data including recently released neutron'
. transport and dosimetry cross-section libraries derived from the ENDF/B-VI data base. This report provides a consistent up-to-date neutron exposure data base for use in evaluating the material properties of the Farley Units 1 and 2 reactor vessels.
-[
In each of the capsule dosimetry evaluations, fast neutron exposure parameters in terms of neutron j
fluence (E > 1.0 MeV), neutron fluence (E > 0.1 MeV), and iron atom displacements (dpa) are
[
established for the capsule irradiation history. Tne analytical formalism relating the measured capsule m:uonow.wpt: b-os2996 11 m,_
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exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel wall. Also, uncertainties associated with the derived exposure parameters at the surveillance capsules and with the projected exposure of the pressure vessel are provided, 1
1 0
4 m:UO60w.wpf;1b o62996 1-2
2.0 DISCRETE ORDINATES ANALYSIS A plan view of the reactor geometry at the core midplane is shown in Figure 21. Since the reactor exhibits 1/8 core symmetry only a 0-45 degree sector is depicted. Six irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program. He capsules are located at azimuthal angles of 107',110', 287*, 290*, 340*, and 343' relative to the core cardinal axis as shown in Figure 2-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 2-2. The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in height. The containers ne positioned axially such that the test specimens a e centered on the core midplane, thus spanning the central 5 feet of the 12 foot high reactor core.
From a neutronic stridpoint, the surveillance capsules and associated support structures are significant.
De presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pad and the reactor vesse' order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, l
two distinct sets of transport calculations were carried out. De first, a single computation in the j
conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the teactor geometry as well as to establish relative radial distributions of exposure parameters ($(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec} through the vessel wall. He neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios; i.e.,
[dpa/sec]/[$(E > 1.0 MeV)], within the pressure vessel geometry. He relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T,1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux,
$(E > 1.0 MeV), at surveillance capsule positions and at several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The source importance functions generated from these adjoint analyses provided the basis for all absolute exposure calculations and comparison with measurement. Dese importance functions, when combined with fuel cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the i
locations of interest for each cycle of irradiation. They also established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles, it is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core but also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the bumup of individual fuel assemb'ies increased.
m-uo60...pt:ib.o62996 2-1
The absolute cycle specific data from the adjoint evaluations together with the relative neutron energy spectra and radisl distribution information from the reference forward calculation provided the means to:
- 1.
Evaluate neutron dosimetry obtained from surveillance capsules.
2.
Relate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.
3.
Enable a direct comparison of analyticM prediction with measurement.
i 4.
Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
The forward transport calculation for the reactor model summanzed in Figures 2-1 and 2-2 was carried out in R,0 geometry using the DORT two<iimensional discrete ordinates code Version 2.7.3"3 and the
~
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BUGLE-93 cross-section library. The BUGLE-93 library is a 47 energy group ENDF/B-VI based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was treated with a P expansion of the scattermg cross-sections and the angular discretization 3
was modeled with an Se order of angular quadrature.
'Ihe core power distribution utihzed in the reference forward transport calculadon was derived from -
statistical studies of long-terrr operation of Westinghouse 3-loop plants. Inherent in the development of this referew core power distribution is the use of an out in fuel reanagement strategy; i.e., fresh fuel on the core periphery. F=ho. ore, for the periphers,1 fuel assemblies, the neutron source was increased by a 20 margm derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power. Since it is unlikely that any single teactor would exhibit power levels on the core p-riphery at the nominal + 20 value for a large number of fuel cycles, tie use of this
- reference distribution is expected to yield somewhat conservative results.
All adjoint calculations were also carned out using an S order of angular quadrature and the P cross-3 section approximation from the BUGLE-93 library. Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as at the geometric center of each
- surveillance capsule. Again, these calculations were run in R,0 geometry to provide neutron source
- distribution importance functions for the exposure parameter of interest, in this case $(E > 1.0 MeV).
Having the importance functions and appropriate core source distributions, the response of interest
- could be calenlmed as:
~~~
R(r,0) =
I(r,0,E) S(r,0,E) r dr de dE
", L 's m oosow.=pt: b m oi,6 2-2
where: _ R(r,0) -_ = L $(E > 1.0 MeV) at radius r and azimuthal angle 8.
LI(r,0,E) = ; Adjoint source importance function at radius r, azimuthal angle 9, and neutron source energy S.
S(r,0,E) = Neutron source strength at core location r,0 and energy E.
- Although the adjoint importance functions used in this analysis were based on a response function defined by the threshold neutron flux $(E > 1.0 MeV), prior calculations"3 have shown that, while the implementation of low leakage loading pattems significantly impacts both the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Dus, for a given location the ratio of [dpa/sec]/[$(E > 1.0 MeV)] is insensitive to changing core source distributions. In the application of these adjoint importance functions to the Parley Unit I and 2 reactors, therefore, the iron atom displacement rates (dpa/sec) and the neutron flux -
$(E > 0.1 MeV) were computed on a cycle specific basis by using [dpa/sec)/[$(E > 1.0 MeV)] and
($(E > 0.1 MeV))/[$(E > 1.0 MeV)] ratios from the forward analysis in conjunction with the cycle
_ specific $(E > 1.0 MeV) solutions from the individual adjoint evaluations.
- The reactor core power distributions used in the plant specific adjoint c@~ ions were taken from the -
fuel cycle design reports for the first fourteen operating cycles of Farley Unit I and the first eleven operating cycles of Farley Unit 2'W3 E
Selected results from the neutron E pen analyses are provided in Tables 2-1 through 2-5. De data listed in these tables establish the means for absolute comparisons of analysis and measurement for the-upsule irrediation periods and provide the means to correlate dosimetry resuhs with the corresponding'
. exposure of the pressure vessel wall.
In Table 21, the calculated exposure parameters'[$(E >'l.0 MeV), $(E > 0.1 MeV), and dpa/sec) are given at the geometric center of the two azimuthally symmetric surveillance capsule positions (17* and 20*) for both the reference and the plant specific cose power distributions. De plant specific data, 1
based on the adjoint E. spen analysis, are meant to establish the absolute comparison of measurement with analysisc De reference data derived from the forward calculation are provided as a conservative -
exposure evaluation against which plant specific fluence calculations can be compared. Similar data
- are given in Table 2-2 for the pressure vessel inner radius. : Again, the inree yd.w.t exposure 1 parameters are listed for the plant specific power distributions for each operatirig cycle.
~
It is important to note that the data for the vessel inner radius were taken at the clad / base metal -
interface; and, thus, represent the maximum predicted exposure levels 'of the vessel plates and welds.
Also, in regard to the pressure vessel, two sets of values are included for the 15' and 30* azimuthal.
locations. His is necessary due to the non-symmetry of the neutron pads in the vicinity of surveillance capsule attachments. With no capsule present the pad span ranges from 0* to 15' in the respective octant. However, pad spans of 0* to 26' exist in the three octants containing double surveillance capsule holders.~
- m:uoso..wpt:i m 2996 3
Radial gradient infonnation applicable to $(E > 1.0 MeV),- $(E > 0.1 MeV), and dpa/sec is given ir.
- Tables 2 3,2 4, and 2-5, respectively. 'Ihe data, obtained from the reference forward neutron transpon 2
calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Tables 2 3 through 2 5.
For example, the neutron flux $(E > 1.0 MeV) at the 1/4T depth in the pressure vessel wall along the 0* azimuth is given by:
i s-Q 40') = $(199.95, O*) F(204.95, 0*)
4 where: $3d 0*)
= Projected neutron flux at the 1/4T position on the O' azimuth.
$(199.95,0*) = Projected or calculated neutron flux at the vessel inner radius on the O'
[
azimuth.
F(204.95,0*) = Ratio of the neutron flux at the 1/4T position to the flux at the vessel inner I
radius for the 0* azimuth. This data is obtained from Table 2-3.
Similar expressions apply for exposure parameters expressed in terms of $(E > 0.1 MeV) and dpa/see where the attenuation function F is obtamed from Tables 2-4 and 2-5 respectively.
1 1
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TABLE 21 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER FARLEY UNIT 1
$ (E > 1.0 MeV) [n/cm' sec)
Location 17' 20' Reference 2.30E+11 1.99E+11 Cycle 1 1.77E+11 1.54E+11 Cycle 2 1.79E+11 1.54E+11 Cycle 3 1,81E+11 1.56E+11 Cycle 4 2.10E+11 1.74E+11 Cycle 5 1.56E+11 1.36E+11 Cycle 6 1.41E+11 1.22E+11 Cycle 7 1.52E+11 1.33E+11 Cycle 8 1.28E+11 1.llE+11 Cycle 9 1,41E+11 1.27E+11 Cycle 10 1.55E+11 1.45E+11 Cycle 11 1.24E+11 1.10E+11 Cycle 12 1.21E+11 1.03E+11 8
$ (E > 0.1 MeV) [n/cm -sec]
Location 17' 20' Reference 1.12E+12 9.27E+11 Cycle 1 8.55E+11 7.16E+11 Cycle 2 8.68E+11 7.20E+11 Cycle 3 8.75E+11
- 7.26E+11 Cycle 4 1.02E+12 8.09E+11 Cycle 5 7.55E+11 6.32E+11 Cycle 6 6.81E411 5.69E+11 Cycle 7 7.37E+11 6.19E+11 Cycle 8 6.18E+11 5.19Etl1 Cycle 9 6.83E+11 5.90E+11 Cycle 10 7.49E+11 6.78E+11 Cycle 11 6.03E+11 5.llE+11 Cycle 12 5.84E+11 4.80E+11 muoso. pcita2996 2-5 m
TABLE 21 (continued)
CALCULATED FAST NEUTRON EXPOSURE
{
RATES AT THE SURVEILLANCE CAPSULE CENTER Iron Displacement Rate [dpa/sec)
.Lesaggg 17' 20' i
Reference 4.63E-10 3.92E-10 Cycle 1 3.55E-10 3.03E 10 Cycle 2 3.61E-10 3.04E-10 Cycle 3 3.64E-10 3.07E-10 Cycle 4 4.22E 10 3.42E-10 Cycle 5 3.14E 10 2.67E-10 Cycle 6 2.83E-10 2.40E-10 Cycle 7 3.06E 10 2.62E-10 Cycle 8 2.57E-10 2.19E-10 Cycle 9 2.84E-10 2.49E-10 Cycle 10 3.llE-10 2.87E-10 Cycle 11 2.51E 10 2.16E-10 Cycle 12 2.43E-10 2.03E-10 FARLEY UNIT 2 8
$ (E > 1.0 MeV) [n/cm sec]
Location 17*
20' Reference 1.12E+12 9.27E+11 Cycle 1 8.55E+11 7.16E+11 Cycle 2 8.68E+11 7.20E+11 Cycle 3 8.75E+11 7.26E+11 Cycle 4 1.02E+12 8.09E+11 Cycle 5 7.55E+11 6.32E+11 Cycle 6 6.81E+11 5.69E+11
$ (E > 0.1 MeV) [n/cm' sec)
Location 17' 20" Reference 1.12E+12 9.27E+11 Cycle 1 8.82E+11 7.16E+11 Cycle 2 9.31E+11 7.20E+11 Cycle 3 8.42E+11 7.265+11 Cycle 4 6.97E+11 8.09E+11 Cycle 5 6.77E+11 6.32E+11 Cycle 6 6.00E+11 5.69E+11 m:uo60w.wpr:id-062m 2-6
TABLE 21 (ccntinued)
CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER Iron Displacement Rate [dpa/sec) l
)
Location 17' 20' Reference 4.63E-10 3.92E 10 Cycle 1 3.67E-10 3.03E-10 Cycle 2 3.87E-10 3.043-10 Cycle 3 3.50E 10 3.07E-10 Cycle 4 2.90E-10 3.42E-10 Cycle 5 2.81E 10 2.67E-10 Cycle 6 2.50E-10 2.40E 10 l
maceow.wprdw2m 2-7
TABLE 2 2
' CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE e
FARLEY UNIT 1 8
$(E > 1.0 MeV) [m/cm -sec]
E E
.11181 E
20,,*131 E
Reference 6.92E+10 3.82E+10 3.05E+10 2.85E+10 2.75E+10 1.90E+10 Cycle 1 5.30E+10 2.%E+10 236E+10 2.22E+10 2.14E+10 1.54E+10 Cycle 2 5.45E+10 3.01E+10 2.41E+10 2.21E+10 2.13E+10 1.48E+10 Cycle 3 533E+10 3.02E+10 2.41E+10 2.22E+10 2.14E+10 1.53E+10 Cycle 4 6.20E+10 3.52E+10 2.82E+10 2.34E+10 2.26E+10 1.54E+10 Cycle 5 4.65E+10 2.61E+10 2.09E+10 1.84E+10 1.78E+10 1.19E+10 Cycle 6 3.96E+10 2.35E+10 1.88E+10 1.72E+10 1.65E+10 1.22E+10 Cycle 7 4.49E+10 2.55E+10 2.04E+10 1.87E+10 1.81E+10 133E+10 Cycle 8 335E+10 2.13E+10 1.71E+10 1.61E+10 1.55E+10 1.15E+10
- Cycle 9 3.49E+10 233E+10 1.86E+10 1.77E+10 1,70E+10 1.15E+10 Cycle 10 3.77E+10 2.56E+10 2.05E+10 2.24E+10 2.16E+10 1.19E+10 Cycle 11 337E+10 2.09E+09 1.67E+09 1.62E+10 1.57E+10 1.20E+10 Cycle 12 3.41E+10 2.03E+09 1.62E+09 1.43E+10 138E+10 1.07E+10 Cycle 13 3.40E+10 2.00E+10 1.60E+10 1.28E+10 1.24E+10 9.47E+09 8
$(E > 0.1 McV) [n/cm -sec]
E E
.l.f.381 E
29*181 E
Reference -
1.81E+11 9.23E+10 738E+10 6.13E+10 5.91E+10 4.01E+10 Cycle l 138E+11 7.15E+10 5.72E+10 4.77E+10 4.60E+10 3.27E+10 Cycle 2 1.42E+11 7.28E+10 5.83E+10 4.75E+10 4.58E+10 3.13E+10 Cycle 3 139E+11 7.29E+10 5.83E+10 4.76E+10 4.60E+10 3.25E+10 Cyc!: 4 1.61E+11 8.51E+10 6.81E+10 5.02E+10 4.84E+10 3.26E+10 Cycle 5 1.21E+11 632E+10 5.06E+10 3.96E+10 3.82E+10 2.52E+10 Cycle 6 1.03E+11 5.68E+10 4.54E+10 3.68E+10 3.55E+10 2.58E+10 Cycle 7 1.17E+11 6.16E+10 4.93E+10 4.02E+10 3.88E+10 2.81E+10 Cycle 8 8.7?S+ 2 5.16E+10 4.13E+10 3.46E+10 3.34E+10 2.43E+10 Cycle 9 9.10E+10 5.63E+10 4.51E+10 3.79E+10 3.66E+10 2.44E+10 Cycle 10 9.83E+10 6.19E+10 4.95E+10 4.80E+10 4.63E+10 2.51E+10 Cycle 11 8.78E+10.
5.04E+09 4.03E+09 3.48E+10 336E+10 2.54E+10 Cycle 12 8.87E+10 4.91E+09 3.93E+09 3.0CE+10 2.95E+10 2.27E+10 Cycle 13 8.84E+10 4.83E+10 3.86E+10 2.75E+10 2.65E+10 2.01E+10 (a) Indicates location in octants with a 26' neutron pad span.
w\\3060w.wpf;1b 070196 2-8
TABLE 2 2 (continuerl)
CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE Iron Aton Displacement Rate [dpa/sec)
E
' jE JJ3nl E
30'fe)
E Reference 1.10E 10 6.00E Il 4.80E 11 4.37E Il 4.21E-11 2.92E Il Cycle 1 8.44E.ll 4.65E Il 3.72E 11 3.40E Il 3.28E Il 2.38E Il Cycle 2 8.67E Il 4.74E ll 3//9E 11 3.39E 11 3.27E 11 2.28E Il Cycle 3 8.49E-Il 4.74E Il 3.79E 11 3.40E 11 3.28E Il 2.37E Il Cycle 4 9.86E Il 5.54E-Il 4.43E 11 3.58E 11 3.46E Il 2.38E Il Cycle 5 7.40E Il 4.!!E-Il 3.295 11 2.82E 11 2.72E-Il 1.83E Il Cycle 6 6.30E 11 3.69E-11 2.95E 11 2.63E 1I 2.54E 11 1.88E.I1 Cycle 7 7.14E Il 4.00E Il 3.20E-Il 2.87E Il 2.77E Il 2.05E 11 Cycle 8 5.33E-Il 3.36E 11 2.68E Il 2.47E 11 2.38E Il 1.77E Il Cycle 9 5.56E Il 3.66E Il 2.93E Il 2.70E-11 2.61E Il 1.78E Il Cycle 10 6.00E-11 4.03E-!!
3.22E Il 3.43E Il 3.31E Il 1.83E Il I
Cycle 11 5.37E Il 3.28E 12 2.62E 12 2.49E 11 2.40E Il 1.85E Il Cycle 12 5.42E-Il 3.19E 12 2.55E 12 2.18E 11 2.11E 11 1.65E Il Cycle 13 5.40E Il 3.14E Il 2.51E 11 1.96E Il 1.89E Il 1.46E 11 FARLEY UNIT 2 8
$(E > 1.0 MeV) [n/cm sec)
E 3E 3E(Al E
3.3 51 E
Reference 6.92E+10 3.82E+10 3.05E+10 2.85E+10 2.75E+10 1.90E+10 Cycle 1 5.48E+10 3.05E+10 2.44E+10 2.25E+10 2.17E+10 1.56E+10 Cycle 2 5.79E+10 3.22E+10 2.57E+10 2.36E+10 2.28E+10 1.63E+10 Cycle 3 5.2SE+10 2.92E+10 2.34E+10 1.95E+10 1.88E+10 1.21E+10 Cycle 4 4.33E+10 2.41E+10 1.93E+10 1.72E+10 1.66E+10 1.24E+10 Cycle 5 3.93E+10 2.34E+10 1.87E+10 1.72E+10 1.66E+10 1.26E+10 Cycle 6 3.44E+10 2.08E+10 1.66E+10 1.55E+10 1.50E+10 1.13E+10 Cycle 7 3.74E+10 2.25E+10 1.80E+10 1.77E+10 1.70E+10 1.21E+10 Cycle 8 3.42E+10 2.llE+10 1.69E+10 1.62E+10 1.56E+10 1.17E+10 Cycle 9 3.20E+10 1.98E+10 1.58E+10 1.56E+10 1.50E+10 1.14E+10 Cycle 10 3.98E+10 1.23E+01 9.81E+00 1.69E+10 1.63E+10 1.19E+10 a) Indicates location in octants with a 26' neutron pad span.
muo60w.wpf:steoi96 2-9 l
I
TABLE 2 2 (continued)
CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE
$(E > 0.1 MeV) [n/cm'sec)
E E
.lE(El E
M.*1ll E
Reference 1.81E+11 9.23E+10 738E+10 6.13E+10 5.9tE 10 4.01E+10 Cycle 1 1.43E+11 7.38E+10 5.90E+10 4.83E+10 4.66E+10 331E+10 Cycle 2.
1.51E+11 7.78E+10 6.23E+10 5.07E+10 4.89E+10 3,46E+10 Cycle 3 138E+11 7.07E+10 5.65E+10 4.18E+10 4.04E+10 2.56E+10 Cycle 4 1.13E+11 5.83E+10 4.66E+10 3.68E+10 3.56E+10 2.OE+10 l
Cycle 5 1.02E+11 5.66E+10 4.53E+10 3.69E+10 3.56E+10 2.66E+10 Cycle 6 8.95E+10 5.03E+10 4.02E+10 335F'10 3.22h+10 2.40E+10 l
Cycle 7 9.74E+10 5.44E+10 4.35E+10 3.79E+10 3.66E+10 2.55E+10 Cycle 8 8.90E+10 5.llE+10 4.09E+10 3.47E+10 335E+10 2.48E+10 l
Cycle 9 833E+10 4.78E+10 3.82E+10 334E+10 3.22E+10 2,42E+10 l
Cycle it, 1.04E+11 2.97E+01 237E+01 3.62E+10 3.49E+10 2.53E+10 Iron Atoen Displacessent Rate [dpa/sec)
E E
.lf181 E
aT.Isl E
Reference 1.10E-10 6.00E-11 4.80E Il 437EIl 4.21E-!!
2.92E-11 Cycle 1 8.73E-Il 4.80E Il 3.84E Il 3.45E Il 333E-11 2.41E-Il Cycle 2 9.21E Il 5.06E 11 4.05E Il 3.62E-Il 3.49E Il 2.52E-Il Cycle 3 8.40E-Il 4.59E Il 3.68E-Il 2.98E-Il 2.88E-Il 1.86E-Il Cycle 4 6.89E 11 3.79E 11 3.03E-11 2.63E-11 2.54E-11 1.91E-11 Cycle 5 6.25E-Il 3.68E 11 2.95E-Il 2.63E Il 2.54E-11 1.94E Il Cycle 6 5.47E Il 3.27E Il 2.62E-Il 238E Il 2.30E-Il 1.75E 11 Cycle 7 5.95E-11 3.54E Il 2.83E Il 2.70E Il 2.61E-Il 1.86E-Il Cycle 8 5.43E-1I 333E 11 2.66E-11 2.48E 11 2.39E-11 1.81E 11 Cycle 9 5.09E Il 3.!!E Il 2.49E-Il 2.38E Il 230E-Il 1.76E-Il Cycle 10 633E-11 1.93E 20 1.54E-20 2.58E-11 2.49E 11 1.84E I' (a) Indicates location in octants with a 26' neutron pad span.
s m uo60w.wpr;iba2996 2-10 t
c..
TABLE 2 3 RELATIVE RADIAL DISTRIBUTION OF $(E > 1.0 MeV)
WITIIIN TIIE PRESSURE VESSEL WALL Radius ign),
O. 0*
15.0*
30.0*
45.0' 199.95"'
l.00 1.00 1.00 1.00
~
200.54 0.%1 0.963 0.%3 0.963 201.72 0.860 0.866 0.866 0.867 202.89 0.752 0.763 0.762 0.764 204.07 0.650 0.664 0.662 0.665 204.95*
0.581 0.597 0.595 0.598 20:i.25 0.559 0.575 0.572 0.576 206.42 0.478 0.496 0.492 0.496 207.60 0.407 0.425 0.421 0.426 j
208.78 0.347 0.364 0.360 0.365 209.95 0.294 0.311 0.307 0.312 211.13 0.249 0.265 0.262 0.266 212.30 0.211 0.226 0.222 0.226 213.48 0.177 0.192 0.188 0.192 214.66 0.149 0.162 0.159 0.163 214.95*
0.143 0.156 0.153 0.157 215.83 0.125 0.137 0.134 0.138 217.01 0.104 0.115 0.113 0.117 218.19 0.086 0.096 0.094 0.098 219.36 0.069 0.080 0.078 0.083 219.95")
0.066 0.077 0.075 0.080 NOW.S:
- 1) Base Metal Inner Radius
- 2) Base Metal 1/4T
- 3) Base Metal 3/4T
- 4) Base Metal Outer Radius m uo60w.wyr:1b 062996 2-11
TABLE 2-4 REL ATIVE RADIAL DISTRIBUTION OF $(E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius (cm)
O. O' 15.0*
30.0*
45.0' 199.95*
1.00 1.00 1.00 1.00 200.54 1.01 1.01 1.01 1.01 201.72 0.986 1.00 0.998 1.00 202.89 0.944 0.972 0.%3 0.972 204.07 0.894 0.931 0.919 0.930 204.95*
0.854 0.897 0.883 0.896 20535 0.841 0.886 0.871 0.885 206.42 0.787 0.837 0.820 0.836 207.60 0.733 0.787 0.769 0.787 l
208.78 0.679 0.737 0.718 0.737 209.95 0.o27 0.687 0.667 0.688 211.13 0.576 0.637 0.618 0.639 21230 0.526 0.587 0.569 0.591 213.48 0.477 0.539 0.521 0.544 214.66 0.430 0.491 0.474 0.498 214.93*
0.418 0.479 0.463 0.487 215.83 0384 0.444 0.429 0.453 217.01 0338 0.397 0.384 0.409 218.19 0.292 0351 0340 0366 21936 0.243 0303 0.295 0323 219.95*
0.233 0.294 0.286 0.315 NOTES:
- 1) Base MetalInner Radius
- 2) Base Metal 1/4T l
- 3) Dase Metal 3/4T
- 4) Base Metal Outer Radius
- m. wow.wpt:tum2996 2-12 v
TABLE 2 5 RELATIVE RADIAL DISTRIBUTION OF DPA/SEC MTI'lIIN THE PRESSURE VESSEL WALL Radius IcIP.l.
O. 0*
,)M' 30.0' 45.0*
199.95*
1.00 1.00 1.00 1.00 200.54 0.%8 0.971 0.%8 0.%9 201.72 0.888 0.896 0.889 0.890 202.89 0.804 0.816 0.804 0.805 204.07 0.723 0.739 0.722 0.725 204.955 0.668 0.685 0.666 0.669 i
205.25 0.650 0.667 0.647 0.651 206.42 0.582 0.602 0.580 0.584 207.60 0.521 0.542 0.518 0.523 208.78 0.466 0.488 0.463 0.468 209.95 0.416 0.439 0.414 0.420 211.13 0.371 0.394 0.370 0.376 212.30 0.331 0.353 0.330 0.336 213.48 0.293 0.316 0.293 0.300 214.66 0.b9 0.281 0.260 0.267 214.95m 0.251 0.573 0.253 0.260 215.83 0.227 0.250 0.230 0.238 217.01 0.197 0.229 0.202 0.210 218.19 0.169 0.192 0.176 0.186 219.':6 0.141 0.166 0.152 0.163 219.95*
0.135 0 !61 0.148 0.159 s
NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal 1/4T
.)
- 3) Base Metal 3/4T
- 4) Base Metal Outer Radius i
i
FIGURE 21 REACTOR GEOMETRY SHOWING A 45' R, O SECTOR 18.M CAPSULES U. X, V 0*
I/
19.72 CAPSULES W. Y, Z
\\
4s-N REACTOR VESSEL I
NEUTRON PAD
/
REACTOR d
1 f
CORE BARREL CORE l
I/
4 m11060w.wpt:1ba62996 2-14
FIGURE 2 2 l
A,mNNl fW////
xNy s
N NNN %
A' 4
m:UO60w.wpf:lbM2996 2-15
3.0 NEUTRON DOSIMETRY
'Ihe passive neutron sensors included in the Parley Units 1. vid 2 surveillance pmgram are listed in Table 3-1. Also given in Table 3-1 are the primary nuck:ar mactions and associated nuclear constats that were used in the evaluation of the neutron energy speceum within the surveillance capsules and e
in the subsequent determinction of the various exposure parameters of interest [$(E > 1.0 MeV),
$(E > 0.1 MeV), dpa/sec). The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules. 'lhe cadmium shielded uranium and neptunium fission monitors were accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in Table 3-1 does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average peuaon flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. L. particular, the fohwing variables are of interest:
4 The measured specific activity of each morutor, The physical characteristics of each monitor.
2 The operating history of the reactor.
The energy response of each monitor.
The neutron energy spectmm at the monitor location.
The specific activity of each of the neutron monitors was determined using established ASTM procedures **"""'3. Following sample preparation and weighing, the activity of each monitor was determmed by means of a lithium-drifted germanium, Ge(LI), gamma spectrometer. The irradiation histories of the Farley Unit I and 2 reactors was obtamed from NUREG-0020, " Licensed Operating Reactors Status Summary Report". The irradiation history applicable to the exposure of Farley Unit 1 Capsules Y, U, X, and W and Earley Unit 2 capsules U, W, and X is given in Table 3-2.
Having the measured specific activitics, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full power operation were determined from the following equation:
A R=
P N, F Y E C, [1-e *] [e *]
nr I
where: R
= Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,(rpdnucleus) o A
= Measured speciric activity (dps/gm) a m UO60w.wpf:Ib 070196 3-1
N. = Number of target clement atoms per gram of sensor F
= Weight fraction of the target isotope in the sen.;or material
- Y
=; Number of product atoms produced per reaction P
3
= Average core power level during irradiation penod j (MW)
P,,, _ = Maxinunn or reference power level of the reactor (MW) s C
= Calculated ratio of $(E > 1.0 MeV) during irradiation period j to the time weighted 3
average $(E > 1.0 MeV) over the entire irradiation period
[
A
= Decay constant of the product isotope (1/sec) tf = Length of irradiation period j (sec)
Decay time following irradiation period j (sec) t, z
and the summation is carried out over the total number of monthly intervals comprising the irradiation Period.
l In the equation describing the reaction rate calculation, the ratio _(Py[P,,,1 accounts for month by -
month variation of reactor com power level within any given fuel cycle as well as over multiple fuel -
~
cycles. The ratio C, which can be calculated for each fuel cycle using the adjoint transport technology 3
discussed in Section 1.2 accounts for the change if sensor reaction rates caused by vanations in flux -
level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. - For a single cycle irradiation C is normally taken to be 1.0. However, for nmitiple cycle irradiations, particularly 3
those employing low leakage fuel management. the additional C term should be employed. The.
3
.-. impact of changing flux levels for constant power operation can be quite significant for sensor sets that-have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in su veillance capsules that have been moved from one capsule location to another.
For the irradiation history of Capsules in Farley Units 1 and 2, the flux level term in the reaction rate
- calculations was developed from the plant specific analysis provided in Table 2-1. Measured and
> saturated reaction product specific activities as well as the derived full power reaction rates are listed in Table 3-3. The specific activities and reaction rates of the U-238 sensors provided in Table 3-3 include corrections for U-235 impurities, plutonium build-in, and gamma ray induced fissions.
Corrections for gemma ray induced fissions were also included in the specific activities and reaction
- rates for the Np 237 sensors as well.
mooso...pt:1w2,96 3-2
4 Values of key fast neutron exposure parameters were desived from the measured reaction rates using.
' the FERRET least squares adjustment code'*3. De FERRET approach used the measured reaction rate l
data, sensor reaction cross-sections, and 6 calculated trial spectrum as input and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured J
= reaction rate data. De " measured" exposure parameters along with the associated uncertainties were then obtained from the adjust-4 spectrum.
In the FERRET evaluations, a log normal least squares algorithm weights both the a priori ve*ws and the measured data in accordance with the assigned uncertainties and correlations. In genera'. x measured values f are linearly related to the flux $ by some response matrix A:
f r - t A y,~
s where i indexes the measured values belonging to a single data set s, g designates the energy group, and et delineates spectra that may be simultaneously adjusted. For example, R,- E o,,+,
a relates a set of measured reaction rates R, to a single spectrum $, by the multigroup reaction cross-section o,,
The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
'In the least squares adjustment, the continuous ouantities (i.e., neutron spectra and cross-sections) were approximated in a multi group format consisting of 53 energy groups. De trial input spectrum was converted to the FERRET 53 group structure using the SAND II code'53. This proceb was carried -
l out by first expanding the 47 group calculated woum into the SAND-II 620 group structuse using a SPLINE interpolanon procedure in regions where group boundaries do not coincide. De 620 point -
spectrum was then re-collapsed into the group structure used in FERRET.
He sensor set reaction cross-sections, @iaad from the ENDF/B VI dosimetry file!"3, were also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting funcdon in the cross section collapsing procedure. Reaction cross-section uncertamties in the fomi of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/B-VI data files.' nese matrices included energy group to anergy group uncertainty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were not included. The omission of thi: additional uncertainty information does not significantly impact the results of the adjustment.-
Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual.pectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation muoso...pr:1ws2996 3-3
was taken from the center of the surveillance capsule modeln in the reference forward transport calculation. While the 53 x 53 group covariance matrices upplicable to the sensor reaction cross-sections ~were' developed from the ENDF/B VI data files, the covariance matrix for the input trial spectrum was constructed from the following relation:
M,, = R,8 + R, R,, P,,
where R, specifies an overall fracdonal normalization uncertainty (i.e., complete couelation) for the set of values. The fractional uncertainties R, specify additional random uncertainties for group g that are correlated with a correlation matrix given by:
l P,, = [1 -0] 6,, + 0 e
- i where:
H = I8'I'}
2 2y The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describe: short range correlations over a group range y (8 specifies the strength of the latter term). The value of 8 is I when g = 3' and 0 otherwise. For the trial spectrum used in the l
cuneut evaluations, a short range conelation of y = 6 groups was used. ' Itis choice implies that ncighboring groups are strongly correlated when 9 is close to 1. Strong long range conelations /.ar anti-c..letions) were justified based on information presented by R. E. MaerkeF. Maerker's senits are closely duplicated when y = 6.
The uncertainties associated with the measured reaction rates included both staustical (coenting) and systematic components. 'Ihe systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, irradiation history conections, and conections for competmg reactions-in the individual sensors.
Resuhs of the FERRET evr.luations of the Farley Units 1 and 2 capsule dosimetry are riven in Table 3-4. The data summarized in this table include fast neutron exposure evaluations in terms of 4(E > 1.0 MeV), @(E > 0.1 MeV), and dpa. In general, excellent results were achieved in the fits of the adjusted spectra to the indivutual measured reaction rates. 'Ih: measured and FERRET adjusted reaction rates for each reaction are given in Table 3-5. An examination of Table 3-5 shows that, in all cases, reaction rates eM*d with the adjusted spectra match the measured reaction rates to better than 14%. 'Ibe ' adjusted spectra from the least squares evaluation is given in Table 3-6 in the FERRET 53 energy group structure.
=:uosow.wpr: bam296 3-4
i i
a l
In Table 3-7, absolute comparisons of the measured and calculated fluence at the center of each d
capsule are presented. De results for the Parley Unit 1 Capsules Y, U, X dosimetry evaluations (M/C ratios of 0.90 - 0.93 for $(E > 1.0 MeV)) and Farley Unit 2 capsules U, W, and W dosimetry evaluations ( M/C ratios of 0.83 - 0.91 for $(E > 1.0 MeV))are consistent with results obtained (rom i
similar evaluations of dosimetry from other reactors using methodologies based on ENDF/B-VI cross.
sections. De evaluation of Farley Unit 1 Capsule W was, however, not consistent with the results for -
q the other capsules and was therefore not used as a factor in determining best estimate exposures.
l l
n w sow = prob m a29s 3-5
TABLE 31 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS Tsget Heston Reaction Weight Product Yield Monitor Material of laterset Fraction Ressoase Ramme Half Life M)
Copper Cu"(n.a)Co*
0.6917 E > 4.7 MeV 5.271 yrs Iron Fe"(n.p)Mn" 0.0580 E > 1.0 MeV 312.5 days Nickel Ni"(n.p)Co" 0.6827 E > l.0 MeV 70.78 days U.mnium 238' US(n,f)Cs'"
0.9996 E > 0.4 MeV 30.17 yrs 6.00 Neptunium 237' Np'"(n.f)Cs'"
1.0 E > 0.08 MeV 30.17 yrs 6.27 Cobalt Aluminum *
< (n,y)Co" 0.0015
.4ev>E>0.015 MeV 5.271 yrs Cobalt Alaminum Co"(n,y)Co" 0.0015 E > 0.015 MeV 5.271 yrs
- Denotes that monitor is ca knium shielded.
4 I
mwmo..wyt;te 3-6
TABLE 3 2 MONTHLY THERMAL GENERATION
IRST THIRTEEN FUEL CYCLES OF THE FARLEY UNIT 1 REACTOR Thermal Thermal Thermal Month la Generstaan Month la Generation Monta la Generation Cycle 1 (MWier)
Cycle 2 (MWiar)
Cycle 3 (MWier) 8/77 67,513 4
0 12 0
9 864,971 5
0 1/81 0
10 355,140 6
0 2
0 11 1.131,938 7
0 3
0 12 1,304,003 8
0 4
1,134,435 108 1,354,896 9
0 1,639,348 2
1,426,824 10 0
6 1,836,579 3
1,884,641 11 457,481 7
1,889,655 4
1,662,759 12 1,760,937 8
1,902,225 5
1,650,988 1/80 1,721,418 9
552,776 6
1,784,321 2
639,186 7
1,830,416 3
1,800,910 8
1,797.336 4
1,870,2X, 9
941,142 5
1,927,148 10 1,501,768 6
822,411 11 1,8 %,861 7
1,201,139 12 1,787,293 8
1,694,62R 109 1,558,383 9
1,398,265 2
1,616,052 10 1,933,931 3
381,913 11 382J70 TOTAL 26,799,158 TOTAL 17,610,294 TOTAL 8,955,018 Therusal Thermal Thermal Mosthla Generation Month im Generation Month in Generation Cycle 4 (MWt4r)
Cycle 5 (MWt br)
Cycle 6 (MWier) 10 0
2 0
3 0
11 0
3 33,094 4
124,058 12 0
4 1,489,953 5
1,807,741 1/82 0
5 1,956,770 6
1,908,735 2
0 6
1,810,449 7
1,973,062 3
1,254,0.18 7
1,973,088 8
1,973,088 4
1,523,773 8
1,971,686 9
1,904,651 5
1,955.244 9
1,909,440 10 1,949,788 6
1,875,9Ii 10 1,789,981 11 1,898,355 7
1, % 8,060 11 1,909,228 12 1,898,787 8
1,679,390-12 1,973,088 1/85 1,973,088 9
1,812,488 1/84 1,729,356 2
1,782,144 10 1,710,147 2
631,428 3
1,708,471 11 1,909,130 4
329,500 12 1,856,957 1/83 794,698 TOTAL 18,339,836 TOTAL 19,177,563 TOTAL 21.231,468 nuom.wpubo62996 37
TABLE 3 2 (continued)
MONTHLY THERMAL GENERATION FIRST THIRTEEN FUEL CYCLES OF THE FARLEY UNIT 1 REACTOR Therinal Thernaal Thennal Moeth la Generation Month la Generation Month la Generstion Cycle 7 (MWt br)
Cycle 8 (MWi br)
Cyde 9 (MWi br) 5 85,286 11 0
4 0
6 1,655,389 12 1,716,656 5
388,449 7
1,844,333 1/87 1,751,070 6
1,773,947 8
1,958,242 2
1,768,852 7
1.973,088 9
1,871,312 3
1,955,044 8
1,970,025 10 1,954,081 4
1,216.358 9
1,4 9,440 i
l1 1,865,690 5
1,906,563 10 1,884,883 12 1,943,431 6
1,909,366 11 1,908,164 1/86-1,877,709 7
1,973,035 12 1,951,416 2
1,741,224 8
1,748,771 1/89 1,964,305 3
1,812.234 9
1,909,437 2
1,757,472 4
1,872,495 10 1,812,021 3
1,932.709 5
1,853,814 11 1,909,437 4
1,892,334 6
1,888,407 12 1,235,697 5
1,973,088 7
1.846,629 1/88 1,973,088 6
1,906,099 8
1,775,600 2
1,845,755 7
1,967,625 9
1,890,035 3
1,581.648 8
1,969,301 10 181,323 9
1,383,665 TOTAL 29,917,234 TOTAL 28,212,798 70TAL 30,506,010 l
Therusal Theriaal Thenmal Month la Generation Month in Generation Month in Generation Cycle 10 (MWt br)
Cycle 11 (MWt-br)
Cycle 12 (MWi br) 10 0
4 0
10 0
11 855,994 5
467,378 11 257 12 1,968,102 6
1,777,930 12 1,560,307 1/90 1,970,192 7
1,919,659 1/93 1,928,56'4 2
1,652.567 8
1,725,229 2
1,775fo4 3
1,% 2,103 9
1,890,966 3
1,293,224 4
1,906,730 10 1,879,448 4
1,905,594 5
1,846,696 11 1,909,440 5
1.971,444 6
1,758,695 12 1,961,350 6
1,908,088 7
1,690,942 1/92 1,971,348 7
1,972,237 8
1.971,720 2
1,844,169 8
1,972,213 9
1,867,822 3
1,972,001 9
1,907,885
'O 1,974,496 4
1,860,541 10 1,975,210 11 1,906,403 5
1.971,974 11 1,909,034 12 1,971,698 6
1,907,928 12 1,972,229 1/91 1,970,245 7
1,971,762 1/94 1,969, % 7 2
1,780,012 8
1,971,823 2
1,763,942 3
488,986 9
1,564,342 3
227,593 TOTAL 29,543,403 TOTAL 30,567,289 TOTAL 28,013.396
~
m:uo60w.wpt; bo62996 3-8
TABLE 3 2 (continued)
MONTHLY THERMAL GENERATION FIRST THIRTEEN FUEL CYCLES OF THE FARLEY UNIT l' REACTOR Thermal Month la Generation Cycle 13 (MWl4r) 4 275,479 5
1,840,449 6
1,887,519 7
1,946,860 8
1,972,995 I
9 1,909,440 10 1,966,909 11 1,909,440 12 1,972,720 165 1,668,639 2
1,772,517 3
1,971,635 4
1,904,131 5
1,969,707 6
1,432,290 7
1,972,982 8
1,926,386
~
9 791,505 TOTAL 31,091,603 ma6o..wyr;ibo62996 39
TABLE 3 2 (continued)
MONTHLY THERMAL GENERATION FIRST TEN FUEL CYCLES OF THE FARLEY UNIT 2 REACTOR Thermal N rmal Nrmal Month b Generstlos Month la Generation Month in Generation Cycle 1 (MWi-kr)
Cycle 2 (MWi kr)
Cycle 3 (MWt.hr) 5/81 49,210 11 0
10 163,398 1
6 369,436 12 1,572,848 11 1,479,567 1
7 118.595 1/83 1,967,195 12 1,917,263 8
1,830,227 2
1,778,710 1/84 1.711,893 9
1,837,261 3
1, % 9,863 2
1,845,593 l
10 1,948,774 4
1,899,797 3
1,895,910 11 1,818,795 5
1,970,049 4
1,845,190 12 1,912,506 6
1,905,327 5
1,950,480 l
1/82 1,707,819 7
1,960,067 6
1,909,440 2
10,186 8
1,966,925 7
1, % 8,023 3
1,541,170 9
937,869 8
1,951,599 4
1,855,620 9
944,645 5
1,913,932 to 1,813,687 6
1,836,202 11 1,906,767 7
1,925.451 12 1,771,677 8
1,963,933 1/85 239,473 9
1,814,278 10 1,387,065 TOTAL 24,453,395 TOTAL 17,928,650 TOTAL 25,314,605 Nrmal thenmal Nranal Month in Generatiou Month in Generation Month la Generation Cycle 4 (MWi kr)
Cycle 5 (MWi-br)
Cycle 6 (MWi br) 2 0
5 396,193 11 0
3 368,968 6
1,744,613 12 240,584 4
1,861,582 7
1,625,000 1/88 1,971,889 5
1,925,137 8
1,865,663 2
1,806,542 6
1,865,725 9
1,871, % 7 3
1,973,062 7
1,769,634 10 1,966,262 4
1,906,960 8
1,739,953 11 1,905,245 5
1,973,062 9
1,867,238 12 1.970,404 6
1,904,396 10 1,943,516 1/87 789,042 7
1,973,019 Ii 1,874,503 2
1,756,640 8
1,973,083 12 1, % 2,073 3
1,839,164 9
1,767,980 1/86 1,798,238 4
1,873,088 10 1,975,740 2
1,755,415 5
1,972,677 11 1,690.586 3
1,945,120 6
1,771,494 12 1,938,045 4
239,320 7
1,970,667 1/89 1, % 7,158 8
1.782.971 2
1,687.171 9
1,909,196 3
1.516,400 10 119,759 TOTAL 22,916,422 TOTAL 29,130,045 TOTAL 28,265,677 nouo60wwpf;1t462996 3-10
.~
i TABLE 3 2 (continued)
MONTHLY THCtMAL GENERATION FIRST TEN FUEL CYCLES OF THE FARLEY UNIT 2 REACTOR Thermal Thermal Theriaal Month in Generation Month in Generation Month la Generation Cycle 7 (MWi br)
Cycle 8 (MWt br)
Cycle 9 (MWi br) 4 0
11 0
4 0
5 172.773 12 0
5 513,304 6
1,786,822 ISI 1,459,221 6
1,783,981 7
1,864,633 2
1,744,i14 7
1,877,44I 8
1,971,608 3
1,971,550 8
1,%5,981 9
1,456,223 4
1,237,745 9
1,779,781 10 1,848,890 5
1,965,554 10 1,742,799 11 1,815,074 6
1,905,656 11 1,890,054 12 1,971,879 7
1,966,074 12 1,952,058 ISO I,970,306 8
1,874,009 163 1,879,525 2
1,631,313 9
1,901.630 2
1,114,049 3
1.971,205 10 1,973,717 3
1,936.119 4
1,694,856 11 1,784,506 4
1,885,890 5
1,168,621 12 1,971,598 5
1,902,839 6
1,907,902 162 1,853,267 6
1,889,227 7
1,970,982 2
1,845,792 7
1,906,713 8
1,970,095 3
367,169 8
1,952,573 9
1,907.101 9
1,433,273 10 755,300 TOTAL 29,835,583
'IDTAL 25,821,600 TOTAL 29,405,606
'Ibersaal Month in Generation Cycle 10 (MWt br) 10 0
l1 0
12 1,104,1%
164 1,953,209 2
1,772,701 3
1, % 2,548 4
1,893,222 5
1,% 3,755 6
1,903,898 7
1,973,017 8
1,881,8 %
9 1,909,440 10 1,975,737 11 1,905,343 12 1,709,587 ISS 1,728,025 2
1,742,523 3
555,111 TOTAL 27.934,208 mM060w.wpf;1b462996 3-1I J
TABLE 3 3 MEASURED AND SATURATED SENSOR ACTIVITIES AND REACTION RATES FARLEY ! SURVEILLANCE CAPSULE Y Measural Satursted Reaction AetMey Activity Rate Reacties (don / man)
(don /nm)
(rsa/ atom)
Cu-63 (n.a) Co-60 Top 6.59E+04 5.10E+05 7.77E 17 Mid 6.28E+04 4.86E+05 7.41E.17 Bottom 6.81E+04 5.27E+05 8.04E 17
" AVERAGES "
6.56E+04 5.08E+05 7.74E 17 Fe-54 (n.p) Mn 54 T'9 1.95E+06 5.33E+06 8.52E 15 Mid 1.90E+06 5.19E+06 8.30E 15 Bottom 1.99E+06 5.44E+06 8.69E 15
- AVERAGES "
1.947+06 5.32E+06 8.50E 15 Ni 58 (n.p) Co 58 Top I.01E+07 8.05E+07 1.1SE.I4 Middle 9.63E+06 7.67E+07 1.10E 14 Bottom 1.04E+07 8.29E+07 1.18E 14
" AVERAGES "
1.00E+07 8.00E+07 1.14E 14 i
Co 59 (n.1) Co-60 Top 1.44E d7 1.11E+08 7.27E 12 Middle 1.50E+07 1.16E+08 7.57E 12 Bottom 1.46E+07 1.13E 48 7.37E 12
" AVERAGES "
1.467407 1.14E+08 7.40E 12 Co 59 (n.1) Co-60 (Cd)
Top 1.44E+07 1.11E48 7.27E 12 Middle 1.50E+07 1.16E+08 7.57E 12 Bottom 1.46E+07 1.13E+08 7.37E-12
" AVERAGES "
1.47E+07 1.14E+07 7.40E 12 U 238 (n.0 Cs 137 (Cd)-
Middle 2.59E+05 1.01E+07 6.63E 14 Np 237 (n.0 Cs 137 (Cd)
Middle 1.82E+06 7.07E+07 4.44E 13 I
i i
4 i
l 4
m9060w.wpt:1w2996 3-12
TABLE 3 3 (continued)
MEASURED AND SATURATED SENSOR ACTIVITIES AND REACTION RATES FARLEY 1 SURVEILLANCE CAPSULE U Reettien Measured Satursted Reaction ActMty ActMty Rate (don / ann (don /am) frus/ atom)
Co-63 (n.a) Co 60 Top l.35E+05 5.06E+05 7.72E 17 Mid 1.34E+05 5.024+05 767IE 17 Bottom 1.42E+05 5.32E+05 8.12E 17
" AVERAGES "
1.37E+05 5.14E+05 7.84E 17 Fe 54 (n.p) Mn 54 Top 1.88E+06 5.22E+06 8.36E 15 Mid 1.86E+06 5.16E+06 8.25E 15 Bottom 1.94E+06 5.40E+06 8.64E 15
" AVERAGES "
1.89E+06 5.26E+06 8.42E 15 Ni 58 (n.p) Co 58 Top 4.16E+06 7.90E+07 1.13E 14 Middle 4.03E+06 7.66E+07 1.09E 14 Bottom 4.33E+06 8.21E+07 1.17E 14
" AVERAGES "
4.17E+06 7.93E+07 1.13E 14 Co 59 (n,y) Co-60 Top 3.56E+07 1.33E+08 8.71E 12 Middle 3.57E+07 1.34E+08 8.74E 12 Bottom 3.60E+07 1.35E+08 3.79E 12
" AVERAGES **
3.587+07 1.34E+08 8.75E 12 Co 59 (n,Y) Co-60 (Cd)
Top No Data No Data No Data Middle No Data No Data No Data Bottom 1.92E+07 1.35E+08 4.68E 12
" AVERAGES "
No Data No Data No Data n
U 238 (n,0 Cs 137 (Cd)
Middle 7.00E+05 1.07E+07 7.03E 14 Np-237 (n.O Cs 137 (Cd)
Middle 6.05E+06 9.23E+07 5.79E-13 awow*pt: b06:996 3-13 i
TABLE 3 3 (continued)
MEASURED AND SATURATED SENSOR ACTIVITIES AND REACTION RATES FARLEY 1 SURVEILLANCE CAPSULE X Measurid Saturated Reaction Activity Activity Rate Reection (don /ne)
/
(datsud.
(rse/stosa)
)
Cu-63 (n.a) Co-60 Top 2.13E405 4.72E+05 7.20E 17 Mid 2.12E+05 4.70E+05 7.17E 17 Bottom 2.22E+05 4.92E+05 7.51E 17
" AVERAGES "
2.16E+05 4.78E+05 7.295 17 l
Fe 54 (n.p) Mn-54 Top 2.27E+06 4.59E+06 7.33E 15 Mid 2.24E+06 4.53E+06 7.24E 15 Bottom 2.40E+06 4.85E+06 7.75E 15
" AVERAGES "
2.307+06 4.65E+06 7.75E 15 Ni 58 (n.p) Co-58 Top 7.83E+06 7.33E+07 1.05E 14 Middle 7.55E+06 7.07E+07 1.01E 14 Bottom 8.05E+06 7.54E+07 1.08E 14
" AVERAGES "
7.81E+07 7.31E+07 1.04E 14 Co-59 (n,1) Co.60 t
Top 5.11E+07 1.13E+08 7.39E 12 Middle 5.24E407 1.16E+08 7.58E 12 Bottom 4.76E+07 1.06E+08 6.89E-12
" AVERAGES "
5.04E+07 1.12E+08 7.28E 12 Co-59 (n,y) Co-60 (Cd)
Top 2.84E+07 6.30E+07 4.llE 12 Middle No Data No Data No Data Bottom 2.64E+07 5.85E+07 3.82E 12
" AVERAGES "
2.74E+07 6.08E+07 3.82E 12 U 238 (n.f) Cs 137 (Cd)
Middle 1.28E+06 1.02E+07 6.71E 14 Np-237 (n.f) Cs 137 (Cd)
Middle 9.88E+06 7.86E+07 4.93E 13 m:uo60..wpu b-062996 3-14
TABLE 3 3 (continued)
MEASURED AND SATURATED SENSOR ACTIVITIES AND REACTION RATES FARLEY 1 SURVEILLANCE CAPSULE W M e red Satursted Reaction Activity Activity Rate Reaction (de->mm)
(des /nm)
(rseletosal Cu 63 (n.a) Co-60 Top 2.36E+05 3.68E+05 5.61E 17 Mid 2.35E+05 3.66E+05 5.59E 17 Bottom 2.49E+05 3.88E+05 5.59E 17
" AVERAGES "
2.40E+05 3.74E+05 5.93E 17 Fe 54 (n.p) Mn 54 q
Top 1.73E+06 3.38E+06 5.41E 15 Mid 1.70E+06 3.33E+06 5.32E 15 Bottom 1.84S+06 3.60E+06 5.76E 15
" AVERAGES "
1.76E+06 3.43E+06 5.76E 15 Ni 58 (n.p) Co-58 Top 8.46E+06 5.60E+07 8.00E 15 Middle 8.08E+06 5.35E+07 7.64E 15 Bottom 8.79E+06 5.82E+07 8.31E 14
" AVERAGES "
8.44E+06 5.59E+07 7.98E 15 Co-59 (n,y) Co-60 Top No Data No Data No Data Middle No Data No Data No Data Bottom No Data No Data No Data
" AVERAGES "
No Data No Data No Data Co-59 (n,1) Co-60 (Cd)
Top 4.15E+07 6.47E+07 4.22E 12 Middle 4.12E+07 6.43E+07 4.19E 12 Bottom 4.15E+07 6.47E+07 4.22E 12
" AVERAGES "
4.14E+07 6.46E+07 4.21E 12 U 238 (n.0 Cs 137 (Cd)
Middle 1.54E+06 6.52E+06 4.30E 14 Np-237 (n.0 Cs 137 (Cd)
Middle 1.15E+07 4.87E+07 3.06E 13 mD060w.wpf.1b462996 3-1$
T,tBLE 3 3 (continued)
MEASURED AND SATURATED SENSOR ACTIVITIES AND REACTION RATES FARLEY 2 SURVE!LLANCE CAPSULE U Measusid Saturated React 6on Activity Activity Rate Reaction (dednes)
(dos /m)
(rse/atosa)
Cu-63 (n,a) Co 60 Top 6.42E+04 5.256+05 8.02E 17 Mid 6.47E+04 5.29E+05 8.08E 17 Bo tom 6.79E+04 5.56E+05 8.48E 17
" AVERAGES "
6.56E+04 5.37E+05 8.19E 17 Fe 54 (n.p) Mn 54 Top 1.681+06 5.43E+06 8.68E-15 Mid 1.62E+06 5.23E+06 8.36E 15 Bottom 1,71E+06 5.54E+06 8.86E 15
" AVERAGES "
1.67E+06 5.40E+06 8.63E 15 Ni.58 (n,p) Co 58 Top 5.17E+06 8.29E+07 1.18E 14 Middle 4.84E+06 7.76E +07 1.1lE 15 Bottom 5.29E+06 8.48E+07 1.2tE 14
" AVERAGES "
5.10E+06 8.17E+07 1.17E 15 Co-59 (n,$ Co 60 Top 1.36E 47 1.12E+08 7.28E 12 Middle 1.42E+07 1.16E+08 7.58E 12 Bottom 1.40E+07 1.14E+08 7.45E 12
" AVERAGES "
1.39E+07 1.14E+08 7.44E 12 Co 59 (n,$ Co-60 (Cd)
Top 7.77E+06 6.36E+07 4.14E-12 Middle 8.09E+06 6.62E+07 4.32E 12 Bottom 7.82E+06 6.40E+07 4.1BE 12
" AVERAGES "
7.89E+06 6.46E407 4.22E 12 U 238 (n.0 Cs 137 (Cd)
Middle 2.29E+05 9.26E+06 6.10E 14 Np-237 (n.O Cs 137 (Cd)
Middle 2.15E+06 8.69E+07 5.46E 13 muosow.wpr;tb-062996 3-16
TABLE 3-3 (continued)
MEASURED AND SATURATED SENSOR ACTIVITIES AND REACTION RATES FARLEY 2 SURVEILLANCE CAPSULE W MeasurW Saturated Reaction Activity Activity Rate Rees, ting (don /nm)
(dos / min)
(rna/atoen)
Cu.63 (n.a) Co 60 Top 1.53E+05 4.311+05 6.58E 17 Mid 1.55E+05 4.37E+05 6.66E 17 Bottom 1.69E+05 4.76E+05 7.27E 17
" AVERAGES "
1.59E+05 4.48E+05 6.84E 17 Fe 54 (n.p) Mn 54 Top 2.11E+06 4.37E+06 6.98E 15 Mid 2.05E+06 4.24E+06 6.78E 15 Bottom 2.19E+06 4.53E+06 7.24E 15
" AVERAGES "
2.12E+06 4.38E+06 7.00E 15 Ni-58 (n.p) Co-58 Top 7.28E+06 6.88E+07 9.82E 15 Middle 6.91E+06 6.53E+07 9.32E 15 Bottom 8.04E406 7.60E+07 1.09E 14
" AVERAGES "
7.41E+06 7.00E+07 1.00E 14 Co 59 (n,Y) Co-60 Top 2.77E+07 7.81 E+0','
5.09E 12 Middle
' 2.92E+07 8.23E+07 5.37E 12 Bottom 2.78E+07 7.83E+07 5.1IE 12
" AVERAGES "
2.82E+07 7.%E+07 5.19'3 12 Co.59 (n,y) Co-60 (Cd)
Top 1.56E+07 4.40E+07 2.87E 12 Middle 1.64E+07 4.62E+07 3.02E 12 Bottom 1.62E+07 4.57E+07 2.98E 12
" AVERAGES "
1.61E+07 4.53E+07 2.95E 12 U 238 (n.f) Cs 137 (Cd)
Middle 5.70E+05 6.72E+06 4.43E 14 Np.237 (n,f) Cs 137 (Cd)
Middle 5.54E,06 6.53E+07 4.10E 13 muom.wyt.ib.o62996 3-17
TABLE 3 3 (continued)
MEASURED AND SATURATED SENSOR ACT!Y! TIES AND REACTION RATES FARLEY 2 SURVEILLANCE CAPSULE X Measural Saturated Reaction Activity Activity Rate Reaction (dos /am)
(dea /am)
(rsvatosa)
Cu-63 (n.a) Co-60 Top 2.27E+05 4.521+05 6.90E 17 l
Mid 2 21E405 4.40E+05 6.72E 17 i
Bottom 2.33E+05 4.64E+05 7.08E 17
- AVERAGES "
2.27E+05 4.52E+05 6.90E 17 Fe-54 (n.p) Mn 54 Top 2.85E+06 4.39E+06 7.CIE 15 Mid 2.80E+06 4.31E+06 6.89E 15 Bottom 2.95E+06 4.54E+06 7.26E 15
" AVERAGES "
2.87E406 4.41E+06 7.0$E 15 N158 (n.p) Co-58 Top 3.24E+07 7.20E+07 1.03E 14 Middle 3.07E+07 6.82E407 9.738 15 Bottom 3.34E+07 7.42E+07 1.06E.14
" AVERAGES "
3.2 t E+07 7.14E+07 1.02E 14 Co-59 (n.1) Co-60 Top 4.90E+07 9.76E+07 6.37E 12 Middle 5.09E+07 1.01E+08 6.62E 12 Bottom 5.06E+07 1.01E+08 6.58E 12
" AVERAGES "
5.02E+07 9.99E+07 6.52E 12 Co 59 (n1) Co-60 (Cd)
Top 2.90E+07 5.78E+07 3.77E 12 Middle 2.76E+07 5.50E+07 3.59E 12 Bottom 2.75E407 5.48E+07 3.58E 12 Bottom 2.67E+07 5.32E+07 3.58E 12
", AVERAGES "
2.77E407 5.52E+07 3.47E 12 U 238 (n,0 Cs 137 (Cd)
Middle No Data No Data No Data Np 237 (n,0 Cs 137 (Cd)
Middle No Data No Data No Data mwmow.wpf.lM42996 3 18
TABLE 3-4
SUMMARY
OF NEUTRON DOSIMETRY RESULTS FOR SURVEILLANCE CAPSULES Measured Flux and Duence for Fadey Unit 1 Capsule Y Hus Huence Ouantity (m/cm sec)
(m/ce')
Uncertainty 8
$ (E > l.0 MeV) 1.60E+11
$.80E+18 m 8. %
$ (E > 0.1 MeV) 7.58E+11 2.76E+ 19 216.%
dpa/sec 3.19E 10 1.16E 02 211. %
l
$ (E < 0.414 ev) 1.30E+11 4.74E+18 221. %
l Measured Mur and Muence for Farley Unit 1 Capsule U l
Hux Duence Omantity in/cm sec)
(m/ce')
Uncertalmtv 8
$ (E > l.0 MeV) 1.74E+ 11 1.69E+19 2 8. %
$ (E > 0.1 MeV) 9.30E+ 11 9.05E+19 216.%
dpa/sec 3.73E 10 3.63E-02 212.%
$ (E < 0.414 ev) 1.60E+11 1.56E+19 220. %
Measured Mux and noence for Fadey Unit 1 Capsule X Mux Huence Ousstity (m/cmi'-see)
(m/cm')
Uncertalmtv
$ (E > 1.0 McV) 1.53E+11 2.95E+ 19
- 8%
$ (E > 0.1 MeV) 8.08E+11 1.56E+20 216. %
dpa/sec 3.26E 10 6.28E-02 212.%
$ (E < 0.414 ev) 1.32E+11 2.54E+19 121 %
Measured Flux and Ruence for Farley Unit 1 Capsule W Mux Muence Ovantity (n/cm sec)
(a/ca')
Uncertainty 8
$ (E > 1.0 MeV) 9.73E+10 3.82E+19 2 8. %
$ (E > 0.1 MeV) 4.90E+11 1.92E+19 216.%
dpa/see 2.03E 10 7.95E-02 212 %
$ (E < 0.414 ev) 3.37E+10 1.32E+19 279. %
atuo60 wpub.062996 3-19
i TABLE 3-4 (continued)
SUMMARY
OF NEUTRON DOSIMETRY RESULTS FOR SURVEILLANCE CAPSULES Measured Hux and Huence for Farley Unit 2 Capsule U Hus Huence Ouestity in/ce'see)
(a/ca')
Uncertainty
$ (E > 1.0 HeV) 1.66E+11 5.79E+18 2 8. %
$ (E > 0.1 MeV) 8.63E+11 3.03E+19 216.%
dpa/sec 3.51E 10 1.23E-02 212.%
$ (E < 0.414 ev) 1.29E+11 4.52E+18 224.%
l i
Measured Hux and Huence for Farley Unit 2 Capsule W Hux Huence Ouantity (m/ce'-sec)
(a/ce')
UnceHalaty
$ (E > 1.0 MeV) 1.24E+11 1.54E+19 2 8. %
$ (E > 0.1 MeV) 6.25E+11 7.80E+19 s16. %
dpa/sec 2.58E 11 3.22E-02
- 11. %
$ (E < 0.414 ev) 8.93E+10 1.12E+19 224. %
Measured Hux and Huence for Farley Unit 2 Capsule X Hux Huence Ouantity (a/ce' sec)
(m/ce')
Uncertainty
$ (E > 1.0 MeV) 1.30E+11 2.64E+19 28 %
$ (E > 0.1 MeV) 6.67E+11 1.35E+20 s16 %
dpa/sec 2.74E 10 5.55E-02 211. %
$ (E < 0.414 ev) 1.16E+11 2.35E+19 124.%
i l
m:uo60w.wpt.iba62996 3 20
TABLE 3 5 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER FARLEY UNIT 1 SURVEILLANCE CAPSULE Y REACTION RATE RATIO (RPS/ NUCLEUS)
MEAS / CALC Reaction Hgg Adl Cale Adl Cale Cu63(n.a)Co60 7.74E 17 7.63E 17 1.01 Fe 54(n,OMn 54 8.50E 15 7.63E 15 0.99 Ni 58(n.p)Co-58 1,14E 14 1.18E 14 0.97 U 238(n,0Cs 137 (Cd) 5.58E 14 4.88E 14 1.15 Np 237(n,0Cs 137(Cd) 4.41E 13 4.68E 13 0.94 Co39(n.Y)Co60 7.40E 12 7.34E 12 1.01 CoS9(n,Y)Co60 (Cd) 4.12E 12 4.14E 12 0.99 f
FARLEY UNIT 1 SURVEILLANCE CAPSULE U REACTION RATE RATIO (RPS/ NUCLEUS)
MEAS / CALC Reaction Man Adi Cale Adi Cale Cu63(n,a)Co60 7.86E 17 7.71E 17 1.02 Fe 54(n,OMn 54 8.42E 15 8.55E 15 0.98 Ni 58(n.p)Co 58 1.13E 14 1.17E 14 0.%
U 238(n,0Cs 137 (Cd) 5.61E 14 5.07E 14 1.11 Np.237(n,0Cs 137(Cd) 5.76E 13 5.72E 13 1.01 CoS9(n,1)Co60 8.75E 12 8.67E 12 1.01 CoS9(n,Y)Co60 (Cd) 4.68E 12 4.71E 12 0.99 m9060..wpr;iba2996 3 21
TABLE 3 5 (condnued)
COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVELLLANCE CAPSULE CENTER FARLEY UNIT 1 SURVEILLANCE CAPSULE X REACTION RATF.
RATIO (RPS/ NUCLEUS)
MEAS / CALC Reaction Msg MLCalt Adi Cale Cu63(n,u)Co60 7.29E 17 7,12E 17 1.02 Fe 54(n,OMn 54 7.44E.15 7.66E 15 0 07 Ni 58(n.p)Co 58 1.04E 14 1.07E 14 0.97 U 238(n,0Cs 137 (Cd) 5.04E 14 4.51E 14 1.I2 Np-237(n,0Cs 137(Cd) 4.90ii 13 4.93E 13 0.99 CoS9(n,7)Co60 7.28E-12 7.22E 12 1.01 CoS9(n,1)Co60 (Cd) 3.% E 12 3.99E 12 0.99 FARLEY UNIT 1 SURVEILLANCE CAPSULE W REACTION RATE RA710 (RPS/ NUCLEUS)
MEAS / CALC Reection Msg A41 Cale AdJ Cale Cu63(n.a)Co60 5.71E 17 5.61E 17 1.02 Fe-54(n OMn 54 5.49D15 5.63E-15 0.97 Ni 58(n.p)Co-58 7.98E 14 7.98E 14 1.00 U 238(n.OCs 137 (Cd) 2.99E 14 2.96E 14 1.01 Np.137(n.OCs 137(Cd) 3.04E 13 3.06E 13 0.99 CoS9(n,y)Co60 No Data No Data No Data CoS9(n,1)Co60 (Cd) 4.68E 12 4.71E 12 0 99 m9060.*pt;nb.os2996 3 22
TABLE 3 5 (continued)
COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CEN'ITR FARLEY UNIT 2 SURVEILLANCE CAPSULE U REACTION RATE RATIO (RPS/ NUCLEUS)
MEAS / CALC Reection Mang AdlCale Adl Calc Cu63(n.a)Co60 8.19E.17 8.04E 17 1.02 Fe 54(n,OMn 54 8.63E.15 8.72E 15 0.99 Ni 5B(n.p)Co 58 1.17E 14 1.20E.14 0.97 U 238(n.f)Cs 137 (Cd) 5.13E.14 4.89E 14 1.05 I
Np-237(n.OCs 137(Cd) 5.42E 13 5.375 13 1.01 CoS9(n,y)Co60 7.44E 12 5 73E.12 1.01 CoS9(n,y)Co60 (Cd) 4.22E.12 4.24E.12 0.99 FARLEY UNIT 2 SURVELLLANCE CAPSUi.E W REACTION RATE RATIO (RPS/ NUCLEUS)
MEAbMC Reaction Mang Adl Cale Adi Calc Cu63(n.a)Co60 6.83E-17 6.75E 17 1.01 Fe-54(n,OMn 54 7.00E 15 7.07E-15 0.99 Ni 58(n.p)Co 58 1.00E 14 9.99E.15 1.00 U 238(n.f)Cs 137 (Cd) 3.53E 14 3,70E.14 0.95 Np.237(n,0Cs 137(Cd) 4.08E.13 3.98E 13 1.02 CoS9(n,Y)Co60 5.195 12 5.15E 12 1,N CoS9(n,y)Co60 (Cd) 2.95E-12 2.97E 12 0.99 nr\\3060w.wpf:1b462996 3-23
~ -. -
TABLE 3 5 (continued)
CGMPAR1 SON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER
^
FARLEY UNIT 2 SURVEILLANCE CAPSULE X REACTION RATE RATIO (RPS/ NUCLEUS)
MEAS / CALC Reaction
- Mgy, Adl Cale AdlCale Cu63(n.a)Co60 6.908 17 6.79E 17 1.02 Fe 54(n,f)Mn 54 7.05E 15 7.19E 15 0.98 Ni 58(n.p)Co $8 1.02E 14 1.02E 14 1.00 U 23P(n.f)Cs 137 (Cd)
No Data No Data No Data Np 237(n f)Cs 131(Cd)
No Data No Data No Data CoS9(n,1)Co60 6.52E 12 6.47E 12 1.01 CoS9(n,y)Co60 (Cd) 3.62E 12 3.62E 12 0.59 a
t m9060w.wytm2996 3-24
TABLE 3 6 i
ADJUt/MD NEUTRON ENERGY SPECTRUM AT THE CAPSULE CENTER FARREY UNIT 1 SURVEILLANCE CAPSULE Y Mus nux pang,1 Enerny (MeV)
(Mg Groum # Enerry (MeV) in/cm nec) 8 1
1.73E+0!
9.0 % 06 28 9.12E43 2.92E+10 2
1.4%.M!
!.73E+0'l 29 5.53E 03 3.62E+10 3
1.35E*ol
%?5E+07 30 3.36E 03 1.23E+10 4
1.165+0!
2 0$E+08 31 2.5v 03 1.24E+10 5
1.00E+01 4.72E 4 32 2.40E-03 1.25E+10 6
8.6t E+00 8.42E+08 33 2.04E-03 3.67E+10 7
7.41E+00 2.0 TE+09 34 1.23E-03 3.42E+10 8
6.07E+00 3.23E+09 35 7.49E-04 3.07E+10 9
4.97E+00 6.92E409 36 4.54E 04 2.74E+10 10 3.68E400 8.39E49 37 2.75E 04 2.94E+10 11 2.87E+00 1.68E+10 38 1.67E-04 3.11E+10 12 2.23E+00 2.398+10 39 1.01E-04 3.09E+10 i
13 1.74E+00 0.34E+10 40 6.14E-05 3.00E+10 14 1.35Et00 3.98E+10 41 3.73E-05 2.91E+10 15 1,llE400 6.86E+10 42 2.26E-05 2.79E+10 I6 8.21 E 01 8.28E+10 43 1.37E-05 2.55E+10 17 6.39E-01 9.09E+10 44 8.32E 06 2.45E+10 18 4.98E il 6.56E+10 45 5.04E-06 2.21E+10 19 3.88E-01 1.02E+11 46 3.06E 06 2.08E+10
?
20 3.02E-01 9.48E+10 47 1.66E-06 1.95E+10 21 1.83E-01 1.02E+11 48 1.13E-06 1.25E+10 22 1.1IE 01 7.33E+10 49 -
6.83E-07 1.37E+10 23 6.74E-02 5.27E+10 50 4.14E-07 2.01E+10 24 4.09E-02 2.64E+10 51 2.51E=07 2.09E+10 25 2.55E-02 3.60E+10 52 1.52E-07 2.08E+10 26 1.99E 12 1.33E+10 53 9.24E-08 6.86E+10 27 1.50E-02 2.08E+10 Note: Tabulated energy levels r nresent the upper energy in each group.
muo60..wpt;iba2996 3-25
TABLE 3 6 (continued)
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CAPSULE CENTER
- FARLEY UNIT 1 SURVEILLANCE C.tPSULE U Max Mux
,G.ggg,,,f, f aerry (MeV)
(a/cm -sec)
Groun #
Emerry (MeV) in/cm eect 8
8 i
1.73E+01 9.72E+06 28 9.12E-03 3.41E+10 2
1.49E+01 2.07E+07 29 5.53E-03 4.18E+ 10 3
1.35E+01 7.68E+07 30 3.36E-03 1.42E+10 4
1.16E+01 2.10E408 31 2.84E-03 1.42E+10 5
1.00E+01 4.77E+08 32 2.40E-03 1.42E+10 6
8.61E+00 8.38E+08 33 2.04E-03 4.16E+10 7
7.41E40 2.04E+09 34 1.23E-03 3.89E+10 8
6.07E+00 3.16E+09 35 7.49E 04 3.48E+10 9
4.97E+00 6.77E49 36 4.54E-04 3.llE+10 10 3.68E+00 8.34E+09 37 2.75E-04 3.35E+10 i
11 2.87E+00 1.70E+10 38 1.67E-04 3.54E+10 12 2.23E+00 2.50E+10 39 1.01E-04 3.52E+10 13 1.74E+00 3.64E+10 40 6.14E-05 3.44E+10 14 1.35E+00 4.54E+10 41 3.73E-05 3.32E+10 15 1.llE+00 8.15E+ 10 42 2.26E-05 3.17E+10 16 8.2tE 01 1.02E+11 43 1.37E-05 3.00E+10 17 6.39E-01 1.14E+11 44 8.32E 06 2.78E+10 18 4.98E-01 8.35E+10 45 5.04E-06 2.50E+10 19 3.88E-01 1.31E+11 46 3.06E-06 2.35E+10 20 3.02E-01 1.22E+11 47 1.86E-06 2.20E+10 21 1.83E-01 1.30E+11 48 1.13E-06 1.41E+10 I
22 1.llE-01 9.32E+10 49 6.83E-07 1.57E+10 23 6.74E-02 6.61E+10 50 4.14E-07 2.32E+10 24 4.09E-02 3.27E+10 51 2.51E-07 2.46E+10 25 2.55E-02 4.37E+10 52 1.52E-07 2.50E+10 26 1.99E-02 1.59E+10 53 9.24E OR 8.72E+10 27 1.50E-02 2.46E+10 Note: Tabulated energy levels represent the upper energy in each group.
mm@w.wpf:lW2996 3 26
\\
TABLE 3 6 (continued)
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CAPSULE CENTER FARLEY UNIT 1 SURVEILLANCE CAPSULE X Hux Mux Groun #
Emerry (MeV)
W em sec)
Groun #
Emerry (MeV)
(m/ce6eec) 8 1
1.73E+01 9.00E+06 28 9.12E-03 3.08E+10 2
1.49E+01 1.92E+07 29 5.53E-03 3.79E+10 l
3 1.35E+01 7.12E+07 30 3.36E-03 1.29E+10 4
1.16E+01 1.95E+08 31 2.84E-03 1.30E+10 5
1.00E+0!
4.41E+08 32 2.40E-03 1.32E+10 6
8.61E+00 7.72E+08 33 2.04E 03 3.92E+10 7
7.41E+00 1.87E+09 34 1.23E-03 3.72E+10 8
6.07E+00 2.87E+09 35 7.49E-04 3.39E+10 9
4.97E+00 6.10E+09 36 4.54E-04 3.08E+10 l
10 3.68E+00 7.46E+09 37 2.75E-04 3.35E+10 11 2.87E+00 1.52E+10 38 1.67E 04 3.83E+10 12 2.23E+00 2.22E+10 39 1.01E-04 3.54E+10 13 1.74E+00 3.20E+10 40 6.14E-05 3.41E+10 14 1.35E+00 3.%E+10 41 3.73E-05 3.23E+10 15 1.llE+00 7.07E+10 42 2.26E-05 3.04E+10 16 8.21E-01 8.77E+10 43 1.37E-05 2.84E+10 17 6.39E-01 9.85E+10 44 8.?2E-06 2.59E+10 1R 4.98E-01 7.21E+10 45 5.04E-06 2.29E+10 19 3.88E-01 1.13E+11 46 3.06E-06 2.UE+10 20 3.02E-01 1.06E+11 47 1.86E-06 1.98E+10 21 1.83E-01 1.13E411 48 1.13E-06 1.26E+10 22 1.llE-01 8.14E+10 49 6.83E-07 1.32E+10 23 6.74E 02 5.80E+10 50 4.1dE-07 1.83E+10 24 4.09E-02 2.88E+10 51 2.51E-07 1.79E+10 25 2.55E-02 3.88E+10 52 1.52E-07 1.69E+10 26 1.99E-02 1.42E+10 53 9.24E-08 4.73E+10 27 1.50E-02 2.20E+10 Note: Tabulated energy levels represent the upper energy in each group.
- m. wow.wpr%.062996 3-27
TABLE 3-5 (continued)
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CAPSULE CENTER FARLEY UNIT 1 SURVEILLANCE CAPSULE W Hux nux Groue #
Enerry (MeV) fiveni' sec)
Grous #
Enerry (MeV)
(n/ca' sed 1
1.73E+01' 7.38E+06 28 9.12E-03 2.07E+10 2
1.49E+01 1.58E407 29 5.53E-03 2.54E+10 3
1.35E+0!
5.85E+07 30 3.36E-03 8.72E+09 4
1.16E+01 1.60E+08 31 2.84E-03 8.90E+09 l
5 1.00E+01 3.58E+08 32 2.40E 03 9.15E+09 i
6 8.61E+00 6.17E+08 33 2.04E-03 2.77E+10 7
7.41E+00 1.47E+09 34 1.23E-03 2.69E+10 8
6 07E+00 2.21E+09 35 7.49E-04 2.52E+10 9
4.97E+00 4.52E+09 36 4.54E 04 2.35E+10 10 3.68E+00 5.25E+09 37 175E-04 2.60E+10 11 2.87E+00 1.02E+10 38 1.67E-04 3.36E+10 12 2.23E400 1.42E+10 39 1.01E-04 2.75E+10 13 1.74E+00 1.97E+10 40 6.14E-05 2.58E+10 14 1.35E+00 2.4 tE+10 41 3.73E-05 2.38E+10 15 1.llE+00 4.21E+10 42 2.26E 05 2.16E+10 16 8.21E 01 5.17E+10 43 1.37E-05 1.%E+10 17 6.39E-01 5.79E+10 44 8 'E-06 1.74E+10 18 4.98E-01 4.28E+10 45 5.04E 06 1.51E+10 19 3.88E-01 6.80E+10 46 3.06E-06 1.37E+10 20 3.02E-01 6.43E+10 47 1.86E-06 1.25E+10 21 1.83E-01 6.98E+10 48 1.13E-06 7.81E+09 22 1.11E-01 5.08E+10 49 6.83E47 6.86E+09 23 6.74E-02 3.67E+10 50 4.14E-07 8.17E+09 24 4.09E-02 1.85E+10 51 2.51E 07 6.97E+06 25 2.55E-02 2.50E+10 52 1.52E-07 5.95E+09 26 1.99E-02 9 32E+09 53 9.24E 08 1.26E+10 27 1.50E-02 1.46E+10 Note: Tabulated energy levels represent the upper energy in each group, i
m:uo60w.wyt.iba62996 3-28
TABLE 3 6 (continued)
ADJUSTED NEUTRON ENERGY SPECTRUM AT TIIE CAPSULE CEhTER FARLEY UNIT 2 SURVEILLANCE CAPSULE U Flux Dux Group #
Enem (MeV) in/cm -sejc Grnuo #
Enem (MeV)
(n/cm'-sec) 8 1
1.73E+01 9.98E+06 28 9.12E-03 3.26E+10 2
1.49E+01 2.15E+07 29 5.53E.03 4.00E+10 3
1.35E+0!
8.02E+07
'K) 3.36E-03 1.35E+10 4
1.16E+01 2.21E+08 31 2.84E-03 1.35E+10 5
1.00E+01 5.02E48 32 2.40E-03 1.35E+10 l
6 8.61E+00 8.81E+08 33 2.NE-03 3.93E+10 7
7.41E100 2.13E+09 34 1.23E-03 3.64E+10 8
6.07E40 3.28E+09 35 7.49E-04 3.24E+10 9
4.97E+00 6.95E+09 36 4.54E-04 2.88E+10 10 3.68E+00 8.39E+09 37 2.75E-04 3.08E+10 11 2.87E+00 1.67E+10 38 1.67E-N 3.17E+10 12 2.23E400 2.40E+10 39 1.01E-04 3.24E+10 13 1.74E+00 3.42E+10 40 6.14E-05 3.17E+10 l
14 1.35E+00 4.25E+10 41 3.73E-05 3.07E+10 15 1.11E+00 7.54E+10 42 2.26E.05 2.95E+10 16 8.21E-01 9.34E+10 43 1.37E-05 2.81E+10 17 6.39E-01 1.0$E+11 44 8.32E-06 2.61E+10 18 4.98E-01 7.66E+10 45 5.04E-06 2.35E+10 19 3.88E-01 1.21E+11 46 3.06E-06 2.21E+10 20 3.0?E-01 1.13E+11 47 1.86E-06 2.07E+10 21 1.83E-01 1.21E+11 48 1.13E-06 1.33E+10 22 1.11E-01 8.72E+10 49 6.83E-07 1.44E+10 23 6.74E-02 6.23E+10 50 4.14E-07 2.08E+10 24 4.09E-02 3.09E+10 51 2.51E-07 2.13E+10 25 2.55E-02 4.16E+10 52 1.52E-07 2.09E+10 26 1.99E-02 1.52E+10 53 9.24E-08 6.57E+10 27 1.50E-02 2.35E+10 Note: Tabulated energy levels represent the upper energy in each group, m:uo60w.wyt:im 3-29
TADLE 3-6 (continued)
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CAPSULE CENTER FARLEY UNIT 2 SURVEILLANCE CAPSULE W Flux flus Group #
Enerny (MeV)
(eveni'sec)
Grous #
Eneruv (MeV)
(n/can sec) 8 1
1.73E+0!
8.65E406 28 9.12E-03 2.39E+ 10 2
1.49E+01 1.86E+07 29 5.53E-03 2.89E+10 3
1.35E+0!
6.92E407 30 3.36E-03 9.72E+09 4
1.16E+01 1.90E+08 31 2.84E-03 9.68E+09 5
1.00E+01 4.30E+08 32 2.40E-03 9.65E+09 6
8.61E+00 7.49E+08 33 2.04E-03 2.81E+10 l
l 7
7.41E400 1.81E+09 34 1.23E.03 2.59E+10 8
6.07E+00 2.74E+09 35 7.49E-04 2.30E+10 9
4.97E+00 5.69E+09 36 4.54E-04 2.04E+10 10 3.68E+00 6.63E+09 37 2.75E-04 2.17E+10 11 2.87E+00 1.29E+10 38 1.67E-04 2.21E+10 12 2.23E+00 1.79E+10 39 1.01E-04 2.28E+10 13 1.74d+00 2.50E+10 40 6.14E-05 2.23E+10 j
14 1.35E400 3.09E+10 41 3.73E-05 2.16E+10 15 1.llE+00 5.42E+10 42 2.26E 05 2.08E+10 j
16 8.21E-01 6.67E+10 43 1.37E-05 1.98E+10 17 6.39E-01 7.45E+10 44 8.32E-06 1.84E+10 18 4.98E-01 5.49E+10 45 5.04E-06 1.65E+10 19 3.88E-01 8.66E+10 46 3.06E-06 1.55E+10 20 3.02E-01 8.15H+10 47 1.86E'4)6 1.45E+10 21 1.83E-01 8.78E+10 48 1.13E-06 9.34E+09 22 1.llE-01 6.34E+10 49 6.83E 07 1.01E+10 23 6.74E-02 4 53E+10 50 4.14E-07 1.45E+10 24 4.09E-02 2 26E+10 51 2.51E-07 1.48E+10 25 2.55E-02 3.02E+10 52 1.52E-07 1.45E+10 26 1.99E-02 1llE+10 53 9.24E-08 4.55E+10 27 1.50E-02 1.71E+10 Note: Tabulated energy levels represent the upper energy in each group.
muo60w.wpt:iw2996 3-30
TABLE 3 6 (continued)
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CAPSULE CENTER FARLEY UNIT 2 SUI'.VEILLANCE CAPSULE X Flus flux Groun #
Enerry (MeL (a/cmi'.wc)
Groun #
Enerry (MeV) in/cin'.sec)
I 1.73E 41 8.47E+06 28 9.12E-03 2.68E+10 2
1.49E+01 1.83E+07 29 5.53E-03 3.29E+10 3
1.35E+01 6.85E+07 30 3.36E-03 1.12E+10 4
1.16E+01 1.89E+0A 31 2.84E-03 1.12E+10 5
1.00E+01 4.28E+08 32 2.40E 03 1.12E+10 l
6 8.61E+00 7.49E+08 33 2.04E-03 3.27E+10 l
7 7.41E+00 1.80E+09 34 1.23E-03 3.05E+10 8
6.07E+00 2.76E+09 35 7.49E-04 2.72E+10 9
4.97E+00 5.79E+09 36 4.54E 04 2.42E+10 10 3.68E+00 6.87E+09 37 2.75E-04 2.60E+10 11 2.87E+00 1.35E+10 38 1.67E-04 2.7t E+10 12 2.23E+00 1.89E+10 39 1.0lE-04 2.74E+10 13 1.74E+00 2.65E+10 40 6.14E-05 2.68E+10 14 1.35E+00 3.28E+10 41 3.73E-05 2.59E+10 15 1.llE+00 5.77E+10 42 2.26E-05 2.48E+10 16 8.21E-01 7.12E+10 43 1.37E-05 2.36E+10 17 6.39E-01 7.98E+10 44 8.32E-06 2.19E+10 18 4.98E-01 5.85E+10 45 5.04E-06 1.97E+10 19 3.88E-01 9.26E+10 46 3.06E-06 1.85E+10 20 3.02E 01 8.75E+10 47 1.86E-06 1.74E+10 21 1.83E-01 9.47E+10 48
- 1.13E-06 1.11E+10 -
22 1.11E-01 647E+10 49 6.83E-07 1.23E+10 23 6.74E-02 4.95E+10 50 4.14E-07 1.79E*10 24 4.09E-02 2.48E+10 51 2.51E-07 1.86E+10 25 2.55E-02 3.36E+10 52 1.52E-07 1.85E+10 26 1.99E-02 1.24E+10 53 9.24E 08 6.10E+10 27 1.50E-02 1.92E+10 Note: Tabulated energy levels represent the upper energy in each group.
i amosow.wpt: be996 3-31
__________u
._ - _ =.
TABLE 3 7 COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVt.'LS FARLEY UNIT 1 SURVEILLANCE CAPSULES CAPSULE Y Calculated higarind E
Fluence (E > l.0 Mev) [n/cm2-sec) 6.42E+18 5.80E+18 0.90 Fluence (E > 0.l Mev) [n/cm2ec) 3.11E+1) 2.75E+19 0.89 dpa 1.29E-02 1.16E-02 0.90 CAPSULE U Calculated Measured E
Fluence (E > 1.0 Mev) [n/cm2 sec) 1.8IE+19 i.69E+19 0.93 Fluence (E > 0.1 Mev) [n/cm2 sec) 8.79E+19 a05E+19 1.03 dpa 6.65E-02 6.27E-02 1.00 CAPSULE X Calculated hiragand E
Fluence (E > 1.0 Mev) [n/cm2 sec) 3.24E+19 2.95E+19 0.91 Fluence (E > 0.1 Mev) [n/cm2 sec]
1.57E+20 1.55E+20 0.99 dpa 6.53E-02 6.27E-02 0.%
CAPSULE W I
Calculated Measund E
Fluence (E > l.0 Mev) [n/cm2 sec) 5.17E+19 3.82E+19 0.73 l
Fluence (E > 0.1 Mev) [n/cm2 sec) 2.41E+20 1.92E+19 0.80 j
dpa 1.02E41 7.95E 02 0.78 1
3 l
neoo60w.wpf;1b@0296 3 32 l
TABLE 3 7 (continued)
COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FARLEY UNIT 2 SURVEILLANCE CAPSL1ES CAPSULE U l
Calculated Measured E
Fluence (E > 1.0 Mev) [n/cm2 sec) 6.39E+18 5.79E+18 0.91 Fluence (E > 0.1 Mev) [n/cm2 sec) 3.10E+19 3.03E+19 0.98 dpa 1.29E.02 1.23E-02 0.%
CAPSULE W Calculated Measured E
Fluence (E > 1.0 Mev) [n/cm2 sec) 1.85E+19 1.54E+19 0.84 Fluence (E > 0.1 Mev) [n/cm2 sec) 8.61E+19 7.79E+19 0.91 dpa 3.64E-02 3.22E.02 0.88 CAPSULE X
_CMSidalgd Measured E
Fluence (E > 1.0 Mev) [n/cm2-sec) 3.18E+19 2.64E+19 0.83 Fluence, (E > 0.1 Mev) [n/cm2-sec]
1.54E+20 1.35E+20 0.88 dp 6.40E-02 5.55E-02 0.87 i
m oos>..wpt: w s2996 3-33
4.0 PROJECTIONS OF PRESSURE VESSEL EXPOSURE l
he best estimate exposure of the Farles Units I and 2 reactor pressure vessels was developed using a combination of absolute plant specific transport calculations and all available plant specific measurement data, in the case of Parley Unit 1, the measurement data base consists of the four surveillance capsules discussed in this report. For Farley Unit 2, the measurement data base consists of the three surveillance capsules discussed in tV.s report.
Combining this measurement data ca,e with the plant specific calculations, the best estimate vessel exposure is obtained from the following relationship:
= K @c, where: $%= ne best estimate fast neutron exposure at the location of interest.
K ne plant specific measurement / calculation (M/C) bias factor derived from the
=
surveillance capsule dosimetry data.
$ce The absolute calculated fast neutron exposure at the location of interest.
=
De approach defined in the above equation is based on the premise that the measurement data represent the most accurate plant specific information available at the locations of the dosimetry; and, further that the use of the measurement data on a plant specific basis essentially removes biases present in the analytical approach and mitigates the uncertainties that would result from the use of analysis alone.
nat is, at the measurement points the uncertainty in the best estimate exposure is dominated by the uncertainties in the measurement process. At locations within the pressure vessel wall, additional uncertainty is incurred due to the analytically determined relative ratios among the various measurement points and locations within the pressure vessel wall.
For Farley Umt 1, the derived plant specific bias factors were 0.92,0.97, and 0.95 for $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa, respectively, excluding the data for Capsule W. For Farley Unit 2, the dutived plant specific bias factors were 0.86,0.92, and 0.90 for $(E > 1.0 MeV), $(E > 0.1 MeV),
and dpa, respectively. Bias factors of this maFnitude are fully consistent with experience using the BUGLE-93 cross-section libmry.
The use of the bias factors derived from the measurement cata base acts to remove plant specific biases associated with the definition of the core source, actual vs. assumed reactor dimensions, and operational variations in water density within the reactor. As a r:sult, the overall uncertainty in the best estimate exposure projections within the vessel wall depends on the individual uncertainties in the m:uo60...pt:nco296 4-1
- ~ - - -.. - - -. - - - - -. -...
i^
measurement proces, the uncertainty in the dosimetry location, and, in the uncertainty in the calculated ratio of the neutron exposure at the point of interest to that at the measurement location.
j --
De uncertainty in the derived neutron flux for an imiividual measurement is obtained directly from -
the results of a least squares evaluation of dosimetry data. The least squares approach combines individual uncertainty in the calculated neutron energy spectrum, the uncertainties in dosimetry cross-sections, and the uncertainties in measured foil specific activities to produce a net uncertainty in the l
derived neutron flux at the measurement point. The associated uncertainty in the plant specific bias i
factor, K, derived from the M/C data base, in turn, depends on the total number of available measurements as well as on the uncertainty of each measurement.
In developing the overall uncertainty associated whh the pressure vessel exposure, the positioning i[
uncertainties for dosimetry are taken from parametric studies of sensor position performed as part a series of analytical sensitivity studies included in the qualification of the methodology The j 'i uncertainties in the exposure ratios relating dosimetry results to positions within the vessel wall are.
again based on the analytical sensitivity studies of the vessel thickness tolerance, downcomer water density vanations and vessel inner radius tolerance. Dus, this portion of the overall uncertainty is
[
controlled entirely by dimensional tolerances associated with the reactor design and by the operational i
characteristics of the reactor.
l
_ The net uncertainty in the bias factor, K, is combined with the uncertainty from the analytical sensitivity study to define the overall 'luence uncertainty at the pressure vessel wall. In the case of
. Farley Unit 1, the derived uncertainties in the bias factor, K, and the additional uncertainty from the analytical sensitivity studies combine to yield a net uncertainty of 26%. 'In the cas of Farley Unit 2, j
the derived uncertainties in the bias factor, K, and the additional uncertainty from the analytical l
- sensitivity studies combine to yield a net uncertainty of 28%.
i j
' Based on this best estimate approach, neutron exposure projections at key locations on the pressure
- vessel inner radius are given in Table 4-1. Along with the current (13.82 EFPY for Farley Unit 1 and 11.30 EFPY for Farley Unit 2) exposure, projections are also provided for exposure periods of 16,32, l
36, and 54 EFPY. Projections for future operation were based on the assumption that the average exposure rates averaged over the cycles 9 through 13 irradiation period for Farley Unit I and cycles 7 through 10 irradiation period for Farley Unit 2 would continue to be applicable up to the first cycle of uprated power for each unit. Projections assume that Farley Unit I uprates to 2775 MWt in Cycle 16
-at 16.5 EFPY Projections assume that Farley Unit 2 uprates to 2775 MWt in Cycle 13 at 13.8 EFPY, i
i j-In the calculation of exposure gradients within the pressure vessel wall for the Farley Units 1 and 2 p
reactor vessels, exposure projections to 16,32,36, and 54 EFPY were also employed. Uprated power levels were projected as described above. Data based on both a @(E > 1.0 MeV) slope and a plant j
specific dpa slope :hrough the vessel wall are provided in Table 4-2.
a
'i
~
i m:u060w.wpf:lbM2996 4-2
In order to access RT,er vs nuence curves, dps equivalent fast neutron Duence levels for the 1/4T and 3/4T positions were defined by the relations:
$(1/4T) = $(OT) dpa(1/4T) dpa(UT) and
$(3/4T) = $(OT) dpa(3/4T) dpa(OT)
Using this approach results in the dpa equivalent fluence values listed in Table 4 2. In Table 4-3 updated lead factors are listed for each of the Farley Units I and 2 surveillance capsules. Lead factor data based on the accumulated fluence through ;ycle 13 for Farley Unit I and cycle 10 for Farley Unit 2 are provided for each remaining capsule.
l msosow.wpt:Ibmo196 43
TABLE 41 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE FARLEY UNIT 1 l
BEST ESTIMATE EXPOSURE (13.8 EFPY) AT THE PRESSURE VESSEL INNER RADIUS 9'
E
.11*(11 E
19'd!1 E
E > 1.0 1.65E+19 9.83E+18
~ 7.86E+18 7.26E+18 7.01E+18 4.93E+18 l
E > 0.1 4.53E+19 2.51E+19 2.01E+19 1.65E+19 1.59E+19 1.10E+19 dpa 2.72E-02 1.60E-02 1.28E-01 1.15E-02 1.llE-02 7,89E-03 BEST ESTIMATE EXPOSURE (16 EFPt's AT THE PRESSURE VESSEL INNER RADIUS 9*.
E' 11*D!)
E 1 9*181 15' E > 1.0 1.87E+19 1.12E+19 8.97E+18 8.31E+18 8.02E+18 5.63E+18 E > 0.1 5.13E+19 2.86E+19 2.29E+19 1.88E+19 1.82E+19 1.26E+19 dpa 3.08E-02 1.83E-02 1.46E-02 1.32E-02 1.27E-02 9.01E-03 BEST ESTDdATE EXPOSURE (32 EFPY) AT THE PRESSURE VESSEL INNER RADIUS 9.*
E
.11*L81 E
19.*(11 31' E > 1.0 3.55E+19 2.18E+19 1.75E+19 1.63E+19 1.58E+19 1.10E+19-E > 0.1 9.75E+19
- 5.57E+19 4.45E+19 3.70E+19 3.57E+19 2.45E+19 dpa.
5.86E-02 3.56E 02 2.85E 02 2.59E-02 2.50E-02 1.76E-02 BEST ESTIMATE EXPOSURE (36 EFPY) AT THE PRESSURE VESSEL INNER RADIUS p*
E jf,'(gl E
30*(n) 41' E > 1.0 3.97E+19 2.45E+19 1.%E+19 1.83E+19 1.77E+19 1.23E+19 E > 0.1 1.0/6+20 6.25E+19 5.00E+19 4.15E+19 4.01E+19 2.75E+19 dpa 6.55E-02 3.99E-02 3.19E-02 2.91E-02 2.81E.02 1.97E-02 BEST ESTIMA~IE EXPOSURE (54 EFPY) AT THE PRESSURE VESSEL INNER RADIUS p'
Ji*
15'(a)
E 30'(a) 4 5',
E > 1.0 5.87E+19 3.64E+19 2.91E+19 2.74E+19 2.64E+19 1.83E+19 E > 0.1 1.61E+20 9.30E+19 7.44E+19 6.20E+19 5.98E+19 4.10E+19 dpa 9.68E-02 5.94E-02 4.75E-02 4.35E-02 4.20E-02 2.94E-02 (a) Indicates Iccadons in octats with a 26' neutron pad span.
m43060w.wpf:lb 062996 4-4
TABLE 41 (continued)
NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE FARLEY UNIT 2 BEST ESTIMATE EXPOSURE (11.3 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E
E 15'(a) 20' 30*(a)
E E > 1.0 1.27E+19 7.38E+18 5.90E+18 5.46E+18 5.27E+18 3.83E+18 E > 0.1 3.55E+19 1.92E+19 1.53E+19 1.26E+19 1.21E+19 8.72E+18 dpa 2.12E-02 1.22E-02 9.77E-03 8.80E-03 8.50E-03 6.23E-03 BEST ESTIMATE EXPOSURE (16 EFPY) AT TIIE PRESSURE VESSEL INNER RADIUS E
E
.!1'fal E
20*L81
.il' E > 1.0 1.73E+19 1.02E+19 8.14E+18 7.61E+18 7.35E+18 3.WE+18 E > 0.1 4.85E+19 2.64E+19 2.llE+19 1.75E+19 1.69E+19 102E+19 dpa 2.90E-02 1.68E-02 1.35E-02 1.23E-02 1.18E-02 8.71E-03 BEST ESTIMATE EXPOSURE (32 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E
E
.!1'Ial E
29*L81 4.1*
E > 1.0 3.35E+19 1.99E+19 1.59E+19 1.51E+19 1.46E+19 1.07E+19 E > 0.1 9.37E+19 5.16E+19 4.13E+19 3.48E+19 3.36E+19 2.43E+19 i
dpa 5.61E-02 3.29E-02 2.63E-02 2.43E-02 2.35E-02 1.73E-02 BEST ESTIMATE EXPOSURE (36 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E
E J1'(g) 20' 30*(a) 45' 5
E > 1.0 3.75E+19 2.23E+19 1.78E+19 1.70E+19 1.64E+19 1.20E+19 E > 0.1 1.05E+20 5.79E+19 4.63E+19 3.91E+19 3.77E+19 2.73E+19 dpa 6.29E-02 3.69E-02 2.95E-02 2.73E-02 2.64E-02 1.95E 02 BEST ESTIMATE EXPOSURE (54 EFPY) AT THE PRESSURE VESSEL INNER RADIUS E
J.I' 15'(a)
E 30*(a) 45' E > 1.0 5.57E+19 3.32E+19 2.66E+19 2.54E+19 2.45E+19 1.80E+19 E > 0.1 1.56E+20 8.63E+19 6.90E+19 5.85E+19 5.64E+19 4.09E+19 dpa 9.33E-02 5.50E-02 4.40E-02 4.09E-02 3.95E-02 2.92E-02 (a) Indicates locations in octants with a 26' neutron pad span.
m:uo60w.wyr:iw62996 4-5
TABLE 4 2 NEUTRON EXPOSURE VALUES WITHIN THE REACTOR VESSEL FARLEY UNIT 1 FLUENCE BASED ON E > 1.0 MeV SLOPE 0*
15' 15'(a) 30' 30*(a) 45' 16 EFPY FLUENCE SURFACE 1.87E+19 1.12E+19 8.97E+18 831E+18 8.02E+18 5.63E+18 1/4T 1.09E+19 6.69E+18 5.35E+18 4.95E+18 4.77E+18 3.37E+18 3/4T 2.67E+18 1.75E+18 1.40E+'. 3 1.27E+18 1.23E+18 8.84E+17 32 EFPY FLUENCE SURFACE 3.55E+19 2.18E+19 1.75E+19 1.63E+19 1.58E+19 1.10E+19 1/4T 2.06E+19 1.30E+19 1.04E+19 9.72E+18
?38E+18 6.57E+18 3/4T 5.08E+18 3.40E+18 2.72E+18 2.50E+18 2.41E+18 1.72E+18 36 EFPY FLUENCE SURFACE 3.97E+19 2.45E+19 1.96E+19 1.83E+19 1.77E+19 1.23E+19 1/4T 2.31E+19 1.46E+19 1.17E+19 1.09E+19 1.05E+19 7.37E+18 3/4T 5.68E+18 3.82E+18 3.05E+18 2.81E+18 2.71E+18 1.93E+18 54 EFPY FLUENCE SURFACE 5.87E+19 3.64E+19 -
2.91E+19 2.74E+19 2.64E+19 1.83E+19 1/4T 3.41E+19 2.17E+19 1.74E+19 1.63E+19 1.57E+19 1.10E+19 3/4T 8.39E+18 5.68E+18 4.55E+18 4.19E+18 4.04E+18 2.88E+18 (a) Indicates locations in octants with a 26* neutron pad span auC60wspMb.062996 4-6
TABLE 4 2 (continued)
NEUTRON EXPOSURE VALUES WTTHIN THE REACTOR VESSEL.
FARLEY UNIT 1-FLUENCE BASED ON dpa SLOPE 0'
Lt 15'(a) 30' 39*(u) 45*
16 EFPY FLUENCE SURFACE 1.87E+19 1.12E+19 8.97E+18 8.31E+18 8.02E+18 5.63E+18 1/4T 1.25E+19 7.68E+18 6.14E+18 5.54E+18 5.34E+18 3.77E+18 3/4T 4.69E+18 3.06E+18 2.45E+18 2.10E+18 2.03E+18 1.46E+18 32 EFPY FLUENCE SURFACE 3.55E+19 2.18E+19 1.75E+19 1.63E+19 1.58E+19 1.10E+19 1/4T 2.37E+19 1.49E+19 1.20E+19 1.09E+19 1.05E+19 7.34E+18 3/4T 8.91E+18 5.%E+18 4.77E+18 4.13E+18 3.99E+18 2.85E+18 36 EFPY FLUENCE SURFACE 3.97E+19 2.45E+19 1.%E+19 1.83E+19 1.77E+19 1.23E+19 1/4T 2.65E+19 1.68E+19 1.34E+19 1.22E+19 1.18E+19 8.24E+18 3/4T 9.97E+18 6.68E+18 5.35E+18 4.64E+18 4.46E+18 3.20E+18 54 EFPY FLUENCE SURFACE 5.87E+19 3.64E+19 2.91E+19 2.74E+19 2.64E+19 1.83E+19 1/4T 3.92E+19 2.50E+19 2.00E+19 1.82E+19 1.76E+19 1.23E+19 3/4T 1.47E+19 9.94E+18 7.%E+18 6.93E+18 6.68E+18 4.77E+18 (a) Indicates locations in octants with a 26' neutron pad span moooow.wpt:ib.o62996 4-7
TABLE 4 2 (continued)
NEUTRON EXPOSURE VALUES WITHIN THE REACTOR VESSEL FARLEY UNIT 2 FLUENCE BASED ON E > 1.0 MeV SLOPE
'O' 15' 15'(al, 36'
.20,,'{a).
45' 16 EFPY FLUENCE SURFACE 1.73E+19 1.02E+19 8.14E+18 7.61E+18 735E+18 536E+18 1/4T 1.01E+19 6.07E+18 4.86E+18 4.53E+18 437E+18 3.20E+18 3/4T 2.48E+18 1.59E+18 1.27E+18 1.16E+18 1.12E+18.
8.41E+17 4
32 EFPY FLUENCE SURFACE 335E+19 1.99E+19 1.59E+19 1.51E+19 1.46E+19 1.07E+19 1/4T 1.95E+19 1.19E+19 9.50E+18 8.98E+18 8.67E+18 638E+18 3/4T 4.79E+18 3.10E+18 2.48E+18 2 31E+18 2.23E+18 1.67E+18 i
36 EFPY FLUENCE SURFACE 3.75E+19 2.23E+19 1.78E+19 1.70E+19 1.64E+19 1.20E+19 1/4T 2.18E+19 133E+19 1.07E+19 1.01E+19 9.74E+18 7.17E+18 3/4T 537E+18 3.48E+18 2.78E+18 2.60E+18 2.50E+18 1.88E+18
+
54 EFPY FLUENCE SURFACE 5.57E+19 332E+19 2.66E+19 2.54E+19 2.45E+19 1.80E+19 1/4T 3.24E+19 1.98E+19 1.59E+19 1.51E+19 1.46E+19 1.07E+19 3/4T '
7.97E+18 5.18E+18 4.15E+18 3.88E+18 3.75E+18 2.82E+18 (a) Indicates locations in octants with a 26' neutron pad span.
f m:UO60w.wpf;lt4)62996 4-8
TABLE 4 2 (continued)
NEUTRON EXPOSURE VALL'ES WITHIN THE REACTOR VESSEL FARLEY UNIT 2 FLUENCE BASED ON dpa SLOPE c'
15' 15'(a) 30*
30*(s) 45' 16 EFPY FLUENCE SURFACE 1.73E+19 1.02E+19 8.14E+18 7.61E+18 7.35E+18 -
5.36E+18 1/4T 1.16E+19 6.97E+18 5.57E+18 5.07E+'i 8 4.89E+18 3.59E+18 3/4T 4.35E+18 2.78E+18 2.22E+18 1.93E+18 1.86E+18 1.39E+18 32 EFPY FLUENCE SURFACE 3.35E+19 1.99E+19 1.59E+19 1.51E+19 1.46E+19 1.07E+19 1/4T 2.24E+19 1.36E+19 1.09E+19 1.01E+19 9.70E+18 7.14E+18 3/4T 8.41E+18 5.43E+18 4.34E+18 3.82E+18 3.68E+18 2.77E+18 36 EFPY FLUENCE SURFACE 3.75E+19 2.23E+19 1.78E+19 1.70E+19 1.64E+19 1.20E+19 1/4T 2.51E+19 1.53E+19 1.22E+19 1.13E+19 1.09E+19 8.02E+18 3/4T 9.42E+18 6.09E+18 4.87E+18 4 29E+18 4.14E+18 3.12E+18 54 EFPY FLUENCE SURFACE 5.57E+19 332E+19 2.66E+19 2.54E+19 2.45E+19 1.80E+19 1/4T 3.72E+19 2.28E+19 1.82E+19 1.69E+19 1.63E+19 1.20E+19 3/4T 1.40E+19 9.07E+18 7.26E+18 6.42E+18 6.20E+18 4.67E+18 (a) Indicates locations in octants with a 26' neutron pad span.
m oo60w.wpfdM62996 4-9
TABLE 4 3 UPDATED LEAD FACTORS FOR FARLEY UNITS 1 AND 2 SURVEILLANCE CAPSULES Fadey Unit I l
Cansules Lead Factor Yid 3.33 U*l 3.34 Xk!
3.38 W'8 3.13 i
V'l 3.11 Z"3 3.11
[a] - Withdrawn at the.nd of Cycle 1
[b] - Withdrawn at the end of Cycle 4
[c] - Withdrawn at N r -i Cycle 7 L
[d] - Withdrawn at M
.if Cycle 12
[e] - Capsules remaining in the reactor Fadey Umst 2 Canseles Lead Factor _
U'*3 3.32 WS' 2.86 X"3 3.40 VI9 3.09 Yl*
2.67 ZiS 2.67
[a] - Withdrawn at ths end of Cycle 1
[b] - Withdrawn at the end of Cycle 4
[c] - Withdrawn at the end of Cycle 6
[d] - Capsules remaining in the reactor m:uo60w.wyt.Ibc70296 4-10
/
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m:uosow.wpf:lW2996 5-4
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