ML20137B351

Radiation Analysis & Neutron Dosimetry Evaluation

ML20137B351
Person / Time
Site:	Beaver Valley
Issue date:	06/30/1996
From:	Brassart G WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20137B329	List: ML20137B323 ML20137B351
References
WCAP-14554, NUDOCS 9703210285
Download: ML20137B351 (74)

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I Westinghouse Non-Proprietary Gass 3

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Beaver Valley Unit 1 Radiatiori Analysis and Neutron Dosim~etry Evaluation

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O Westinghouse Energy Sys tems -

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14554 i

3 BEAVER VALLEY UNIT 1 RADIATION ANALYSIS AND NEUTRON DOSIMETRY EVALUATION ii S. L. Anderson June 19%

1 Work Performed Under Shop Order DROP-450 Prepared by Westinghouse Electric Corporation for the Duquesne Light Company Approved by: _ a# ^?t - -

r G. A. Brassart, Manager Radiation Engineering and Analysis ,

1 4

i l

^

WESTINGHOUSE ELECTRIC CORPORATION l Systems and Major Projects Division i P.O. Box 355 l Pittsburgh, Pennsylvania 15230-0355 l C 1996 Westinghouse Electric Corporation All Rights Reserved

TABLE OF CONTENTS pace TABLE OF CONTENTS i LIST OF FIGURES ii ,

LIST OF TABLES iii

1.0 INTRODUCTION

1-1 2.0 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION 2-1 i METHODOLOGIES 2.1 Neutmn Transport Analysis Methods 2-1 2.2 Neutmn Dosimetry Evaluation Methodology 2-7 i 2.3 Determination of Best Estimate Pirssure Vessel Exposure 2-12 3.0 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 3-1 3.1 Reference Forward Calculation 3-1 ,

3.2 Fuel Cycle Specific Adjoint Calculations 3-2 4.0 EVALUATIONS OF SURVEILLANCE CAPSULE DOSIMETRY 4-1 4.1 Measured Reaction Rates 4-1 l 4.2 Results of the Least Squares Adjustment Pmcedure 4-2 l

5.0 COMPARISON OF CALCULATIONS WITH MEASUREMENTS 5-1  ;

5.1 Comparison of Least Squares Adjustment Results with Calculation 5-1 j 5.2 Comparisons of Measured and Calculated Sensor Reaction Rates 5-2 6.0 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE 6-1 VESSEL MATERIALS 6.1 Exposure Distributions Within the Beltline Region 6-1 6.2 Uncenainties in Exposure Projections 6-6 6.3 Updated Lead Factors for Surveillance Capsules 6-8

7.0 REFERENCES

7-1 i

l

9 4

LIST OF FIGURES Ficure Title Pace 1.0-1 Description of Pressure Vessel Beltline Materials 1-3 2.1-1 Reactor Geometry Showing a 45' r,0 Sector 2-5 2.1-2 Intemal Surveillance Capsule Geometry 2-6 j

t A

ii

l LIST OF TABLES Table Title _ Pace 3.1-1 Calculated Reference Neutron Energy Spectra at Surveillance A Capsule Center 3.1-2 Reference Neutron Sensor Reaction Rates and Exposure Parameters 3-6 at the Center of Surveillance Capsules 3.1-3 Radial Gradient Corrections for Sensors Contained in Internal 3-7 Surveillance Capsules 3.1-4 Summary of Exposure Rates at the Pressure Vessel Clad / Base 3-8 Metal Interface 3.1-5 Relative Radial Distribution of Neutron Flux (E > 1.0 MeV) 3-9 Within the Pressure Vessel Wall 3.1-6 Relative Radial Distribution of Neutron Flux (E > 0.1 MeV) 3-10 l Within the Pressure Vessel Wall l

i 3.1-7 Relative Radial Distribution of Iron Displacement Rate (dpa) 3-11 Within the Pressure Vessel Wall 3.2-1 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Center of 3-12 Reactor Vessel Surveillance Capsules 3.2-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of 3-13 Reactor Vessel Surveillance Capsules 3.2-3 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Pressure Vessel 3-14 Clad / Base Metal Interface 3.2-4 Calculated Fast Neutron Fluence (E > l.0 MeV) at the Pressure 3-15 Vessel Clad / Base Metal Interface 3.2-5 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Center of 3-16 Reactor Vessel Surveillance Capsules iki i

l i

i

' LIST OF TABLES l 4 i l

}- Table Title Page

) +

3.2 Calculated Fast Neutror Fluence (E > 0.1 MeV) at the Center of 3-17 Reactor Vessel Surveillance Capsules 3.2-7 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Pressure Vessel 3-18

]

Clad / Base Metal Interface i

3.2-8 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Pressure 3-19 I Vessel Clad / Base Metal Interface 4

. 3.2-9 Calculated Imn Displacement Rate at the Center of Reactor Vessel 3-20

Surveillance Capsules i.

3.2-10 Calculated Imn Displacements at the Center of Reactor Vessei 3-21 j Surveillance Capsules

! 3.2-11 Calculated Iron Displacement Rate at the Pressure Vessel Clad / Base 3-22

{ Metal Interface a

< 3.2-12 Calculated Imn Displacements at the Pressum Vessel Clad / Base 3-23

Metal Interface J

! 4.1-1 Measured Sensor Specific Activities for Capsule V 4-3 4.1-2 Measured Sensor Specific Activities for Capsule U 4-4 4.1-3 Measured Sensor Specific Activities for Capsule W 4-5 3 i j 4.1-4 Irradiation History of Surveillance Capsules V, U, and W .

4-6

4.1-5 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets 4-7 i Sets Withdrawn fmm Intemal Surveillance Capsules l

4.2 1 Derived Exposure Rates from Surveillance Cap ule V Withdrawn at 4-8 ,

the End of Fuel Cycle 1 I i 4.2-2 Derived Exposure Rates from Surveillance Capsule U Withdrawm at 4-9

. the End of Fuel Cycle 4

, iv 3

e y , ,. _ - . ,

LIST OF TABLES Table Title Pace 4.2 3 Derived Exposure Rates fmm Surveillance Capsule W Withdrawn at 4-10 the End of Fuel Cycle 6 5.1-1 Comparison of Measured and Calculated Exposure Rates from 5-3 Surveillance Capsule Irradiations 5.2-1 Comparison of Measured and Calculated Neutron Sensor Reaction 5-4 Rates from Surveillance Capsule Irradiations 6.1-1 Best Estimate Fast Neutron Fluence (E > 1.0 MeV) for Materials 6-3 Comprising the Beltline Region of the Reactor Vessel 6.1-2 Best Estimate Fast Neutron Fluence (E > 0.1 MeV) for Materials 6-4 Comprising the Beltline Region of the Reactor Vessel 6.1-3 Best Estimate Iron Atom Displacements for Materials Comprising 6-5 Comprising the Beltline Region of the Reactor Vessel 1

l i

V

i

^L i I

l SECTION 1.0 j INTRODUCTION d

In the assessment of the state of embrittlement of light water reactor pressure vessels, an accurate j evaluation of the neutmn exposure of the materials comprising the beltline region of the vessel is i required. This exposure evaluation must, in general, include assessments not only at locations of l- maximum exposure at the inner diameter of the vessel, but, also, as a function of axial, azimuthal, and I radial location throughout the vessel wall.

A schematic of the beltline region of the Beaver Valley Unit I reactor pressure vessel is provided in  !

, Figure 1.01. In this case, the beltline region is constructed of four (4) shell plates, four (4) longitudinal welds, and a circumferential weld Joining the intermediate and lower shells. Each of these j nine materials must be considered in the overall embrittlement assessment of the pressure vessel.  ;

i j in order to satisfy the requirements of 10CFR50 Appendix G for the calculation of pressure / temperature limit curves for normal heatup and cooldown of the reactor coolant system, fast j neutron exposure levels must be dermed at depths within the vessel wall equal to 25 and 75 percent of the wall thickness for each of the matedals comprising the beltline region. These locations are n commonly referred to as the 1/4T and 3/4T positions in the vessel wall. The 1/4T neutron exposures I are also used in the detemiination of upper shelf fracture toughness as specified in 10CFR50 Appendix 4

G.

i in the determination of values of RT,,rs for compariscn with applicable pressurized thennal shock screening criteria for plates and welds, maximum neutmn exposure levels experienced by each of the beltline materials are required. These maximum levels will, of course, occur at the vessel inner radius.

l

) In the event that a probabalistic fracture mechanics evaluation of the pressure vessel is perfonned, or if l an evaluation of thermal armealing and subsequent material re-embrittlement is undenaken, a complete embrittlement profile is required for the entire volume of the pressure vessel beltline. .The j determination or this embrittlement profile would, in tum, necessitate the evaluation of neutmn y exposure gradients throughout the entire beltline.  ;

i' The purpose of this report is to describe the approach used to determine the best estimate fast neutron exposum experienced by the Beaver Valley Unit I reactor pressure vessel; and to establish the  !

uncenainties associated with those projections. The overall methodology derives from the guidance provided in ASTM Standard E853, " Analysis and interpretation of Light Water Reactor Surveillance

' Results" and Draft regulatory Guide DG-1025, " Calculational and Dosimetry Methods for Detemiining

. Pressure Vessel Neutron Fluence. The methodology is dependent on a blend of plant specific neutron transport calculations and availaNe measured data to produce an accurate assessment of the pressure

vessel exposure while minimizing the uncertainty associated with that assessment.

1-1 4

4

The methodology used to provide the best estimate neutron exposure evaluations for the Beaver Valley  ;

l Unit I pressure vessel is based on the underlying philosophy that,in order to minimize the uncertainties associated with vessel exposure projections, plant specific neutron transport calculations must be supported by benclunarking of the analytical approach, comparison with industry wide power reactor data bases of surveillance capsule and reactor cavity dosimetry, and, ultimately, by validation with the plant specific measurement data base.

That is, t progression is made from the use of a purely analytical approach tied to experimental benchmarks to an approach that makes use of industry and plant specific power reactor measurements to remove potential biases in the analytical method, knowledge regarding the neutron environment applicable to a specific reactor vessel is increased and the uncertainty associated with vessel exposure projections is minimized.

In subsequent sections of this report, the neutron transport and dosimetry evaluation methodologies used to perform calculations of the neutron environment within the Beaver Valley Unit I reactor geometn are described; and, the procedures used to combine measurements with calculations to produce the final best estimate exposure of the reactor pressure vessel are discussed. Results of exposure evaluations from surveillance capsule dosimetry withdrawn at the end of Fuel Cycles 1,4, and 6 are combined with the results of plant specific neutron transpon calculations to provide the integrated exposure of the pressure vessel from plant startup through the end of Cycle 11. Also, uncenainties associated with the derived exposure parameters at the measurement locations and with the projected exposure of the pressure vessel are provided. In addition to the evaluation of the current exposure of the reactor vessel beltline materials, pmjections of the future exposure of the vessel are also provided. Current evaluations and future projections are provided for each of the beltline weldments as well as for the plates comprising both the intermediate and lower shells.

All of the calculations and dosimetry evaluations presented in this repon are intended to meet the requirements of Draft Regulatory Guide DG-1025," Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"; and, have been based on the latest available nuclear cross-section data derived from ENDF/B-VI.

l-2

i .

j i

l j FIGURE I.0-1

! 1 4 1 DESCRFilON OF PRESSURE VESSEL BELTLINE MATERIALS '

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j SECTION 2.0 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION METHODOLOGIES As noted in Section 1.0 of this report, the best estimate exposure of the reactor pressure vessel was developed using a combination of absolute plant specific neutmn transpon calculations and plant specific measurements from the three intemal surveillance capsules withdrawn to date. In this section, the neutron transport and dosimetry evaluation methodologies are discussed in some detail; and the j approach used to combine the calculations and measurements to produce the best estimate vessel exposure is presented.

2.1 - Neutron Transpon Analysis Methods Fast neutron exposure calculations for the Beaver Valley Unit I reacter geometry were carried out using both forward and adjoint discnte ordinates transport techniques. A single forward calculation provided the relative energy distribution of neutrons for use as input ta neutron dosimetry evaluations l

as well as for use in relating measurement results to the actual exposure at key locations in the pressure vessel wall. A series of adjoint calculations, on the odier hand, established the means to compute absolute exposure rate values using fuel cycle specific core power distributions; thus, pmviding a direct comparison with all dosimetry results obtained over the operating history of the reactor. i l

In combination, the absolute cycle specific data from the adjoint evaluations together with relative neutmn energy spectra distributions from the forward calculation pmvided the means to:

1- Evaluate neutmn dosimetry from the intemal surveillance capsule locations.

2- Enable a direct comparison of analytical prediction with measurement. l l

l 3- Determine plant specific bias factors to be used in the evaluation of the best estimate exposure of the reactor pressure vessel.

I Establish a mechanism for pmjection of pressure vessel exposure as the design of each new fuel cycle evolves.  !

l 2.1.1 - Reference Forward Calculation A plan view of the reactor geometry at the core midplane elevation is shown in Figum 2.1-1. Since the reactor exhibits 1/8 core symmetry only a 0-45 degree sector is depicted. In addition to the core, 2-1 i

y. -

grp--

reactor internals, pressure vessel, and the primary biological shield, the model also included explicit f representations of the surveillance capsules, the pressure vessei cladding, and the mirror insulation j located external to the vessel. l I

The model depicted in Figure 2.1-1 was developed using nominal design dimensions for all components. Specified tolerances in the design dimensions are reflected in the overall uncenainty assessments associated with projected neutron exposures. This modeling approach is consistent with )

the guidelines of DG-1025. j i

A description of a single surveillance capsule attached to the thennal shield is shown in Figure 2.1-2. i From a neutronic standpoint, the inclusion of the surveillance capsules and associated support structures in the analytical model is significant. Since the presence of the capsules and stmeture has a marked impact on the magnitude of the neutron flux as well as on the relative neutron energy spectra at dosimetry locations within the capsules, a meaningful comparison of measurement and calculation can be made only if these perturbation effects are properly accounted for in the analysis.

The forward transport calculation for the reactor model depicted in Figures 2.1-1 and 2.1-2 was carried out in r,0 geometry using the DORT two-dineional discrete onlinates codeN and the BUGLE-93 cross-section librarym. The BUGLE-93 library is a 67 group (47 Neutron,20 Gamma-Ray), ENDFB-VI based, data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a P3 expansion of the scattering cross-sections and the angular discretization was modeled with an S, order of angular quadraturr. The reference forward calculation was nomialized to a core midplane power density characteristic of operation at a thermal power level of 2652 MWt.

The spatial core power distribution utilized in the reference forward calculation was derived from statistical studies of long-term operation of Westinghouse 3-loop plants. Inherent in the development of this reference core power distribution was the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 20 uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used.

Due to the use of this bounding spatial power distribution, the results from the reference forward calculation establish conservative exposure projections for reactors of this design operating at a rated power level of 2652 MWt. Since it is unlikely that actual reactor operation would result in the implementation of a power distribution at the nominal +20 level for a large number of fuel cycles and, further, because of the widespread implementation of low leakage fuel management strategies, the fuel cycle specific calculations for this reactor result in exposure rates well below these conservative predictions. This difference between the conservative forward calculation and the fuel cycle specific computations is illustrated by a comparison of the analytical results given in Section 3.0 of this report.

2-2

2.1.2 - Cycle Specific Adjoint Calculations All adjoint analyses were also carried out using an S, order of angular quadrature and the P3 cross-section approximation from the BUGLE-93 library, Adjoint source locations were chosen at several key azimuths on the pressure vessel inner radius as well as at the geometric center of surveillance capsules located at 15,25",35 and 45 degrees relative to the core cardinal axes. 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).

The imponance functions generated from these individual adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when I combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each of the fuel cycles to date; and, established the means to perfonn similar predictions and dosimetry evaluations for all subsequent fuel cycles.

Having the imponance functions and appropriate com source distributions, the response of interest can be calculated as:

$(Ro,00 ) = f, J ja 1(r 0,E) S( ,0,E) r dr d0 dE where: $(Ro,00) = Neutron flux (E > 1.0 MeV) at radius Ro and azimuthal angle O.o l(r,0,E) = Adjoint importance function at radius r, azimuthal angle 0, and neutron source energy E.

S(r 0,E) = Neutron source strength at core location r,0 and energy E.

It is important to note that the cycle specific neutron source distributions, S(r,0,E), utilized with the adjoint importanca functions,1(r,0,E), pennitted the use not only of fuel cycle specific spatial variations of fission rates within the reactor core; but, also allowed for the inclusion of the effects of the differing neutron yield per fission and the variation in fission spectmm introduced by the build-in of plutonium isotopes as the burnup of individual fuel assemblies increased.

Although the adjoint imponance functions used in these analyses were based on a response function defined by the threshold neutmn flux (E > 1.0 MeV), prior calculations"3 have shown that, while the implementation of low leakage loading pattems significantly impact the magnituJe and the spatial distribution of the neutron field, changes in the mlative neutron energy spectrum are of second order.

Thus, for a given location the exposure parameter ratios such as [dpa/sec]/[$(E > 1.0 MeV)] are 2-3

insensitive to changing core source distributions. In the application of these adjoint importance functions to the current evaluations, therefore, calculation of the iron displacement rates (dpa/sec) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using the appropriate

[dpa/sec]/[$(E > 1.0 MeV)] and [$(E > 0.1 MeV)]/[$(E > 1.0 MeV)] ratios from the reference forward analysis in conjunction with the cycle specific $(E > 1.0 MeV) solutions from the individual adjoint evaluations.

In particular, after defining the following exposure rate ratios,

[dpa/sec]

R=3

$(E> 1.0 MeV)

, $(E > 0.1 MeV)

$(E > 1.0 MeV) the corresponding fuel cycle specific exposure rates at the adjoint source locations were computed from the following relations:

dpa/sec = ($(E > 1.0 MeV)] R 3

$(E > 0.1 MeV) = ($(E > 1.0 MeV)] R2 All fuel cycle specific calulations were also normalized to the current rated power level for Beaver Valley Unit 1,2652 MWt.

i 2-4  ;

4 I

I l

\

i j '

FIGURE 2.1-1 l REACTOR GEOMETRY SHOWING A 45 r,0 SECTOR

! {

! i

! l

l l l l

. 1 I

I 4

I i

O' (MAJOR AXIS) '

1,5' (CAPSULES V,X).

) #

25' (CAPSULES Y,W,U)

/y /

l l

35' (CAPSULES Z.T)

> f i j \

/

i 45' (CAPSULE 5)

I

/

x\\\\

s A

/ / PRESSURE VESSEL

" Nxxxxx\3

• /  % /

8 / \ /

I / g

/ / , N THERMAL SHIELD

~ l I

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/

/

/

I / /

/ / / CORE BARRE

~ l ,I ,/ t I

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REACTOR CORE

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2-5 l

l l

s+-a,s4a .a a.s,wJ.- e.J -.6a a 4-w,.rasaa  ; -_g ..,s.,a__a,,,_ , , _m.4..w . ,__y_ , , ..a.._,d,__ _,_J ..aJ_._,.4 m . A A._.. 3,.__-4-..M4e. m 4 MAF._,J-,ea4 d e s- h_A wh_mu-16-.4 .e A a hr+b *_^_ww -4 0

0 FIGURE 2.1-2 INTERNAL SURVEILLANCE CAPSULE GEOMETRY li3. U 2.

311.1 %

t 3

m.24 vi 3

_. s ei.so g s ee.m u t aa. tit 3

) al.%o i

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l 2.2 - Neutron Dosimetry Evaluation Methodology l l

The use of passive neutron sensors such as those included in the intemal surveillance capsule i dosimetry sets does not yield a direct measure of the energy dependent neutron flux at the i measurement location. Rather, the activation or fission process is a measun: of the integrated effect i 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 flux level and, herice, time integrated exposure (fluence) experienced by the sensors may be developed from the measurements only if the sensor characteristics and the parameters of the irradiation are well known. In particular, the following variables are of interest:

1 - The measured specific activity of each sensor.

2 - The physical characteristics of each sensor.

3 - The operating history of the reactor.

4 - The energy response of each sensor.

5 - The neutron energy spectmm at the sensor location.

In this section the procedures used to determine sensor specific activities, to develop reaction rates for individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described. These pmcedures apply to all of the evaluations provided in this report.

2.2.1 - Detennination of Sensor Reaction Rates Following irradiation, the multiple foil sensor sets from the surveillance capsule irradiations were recovered and transponed to Pittsburgh for evaluation. Analysis of all radiometric foils sensors was performed at the Westinghouse Analytical Services Laboratory.

The specific activity of each of the radiometric sensors was determined using established ASTM procedures * "d' "1 Following sample preparation and weighing, the specific activity of each sensor was detennined by means of a lithium drifted germanium, Ge(Li), gamma spectrometer. These analyses were perfomied by direct counting of each of the individual wires; or, in the case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution and chemical separation of cesium fmm the sensor.

The irradiation history of the reactor over its operating lifetime was obtained from NUREG-0020,

" Licensed Operating Reactors Status Summary Report". In particular, operating data were extracted from that repon on a monthly basis from reactor startup to the end of the current evaluation period.

For the sensor sets utilized in suiveillance capsule irradiations, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate i

l 2-7 l

l 1 representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. ,

l 1

Itaving the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation at 2652 MWt were '

determined from the following equation:

A R=

N F Y[J o C, [1 - e *l] e *'

rtf I

where: A = measured specific activity (dps/gm).

R = reaction rate averaged over the irradiation period and referenced to operation at a core power level of P.,(rps/ nucleus).

No = number of target element atoms per gram of sensor.

F = weight fraction of the target isotope in the sensor material.

Y = number of product atoms produced per reaction.

P, = average core power level during irradiation period j (MW).

P, = maximum or reference core power level of the reactor (MW). l

= calculated ratio of $(E > 1.0 MeV) during irradiation period j to the I C,

time weighted average $(E > 1.0 MeV) over the entire irradiation period.

A = decay constant of the product isotope (sec"). j t, = length of irradiation period j (sec). l t, = decay time following irradiation period j (sec).

and the summation is carried out over the total number of monthly intervals comprising the total irradiation period.

In the above equation, the ratio P/P , accounts for month by month variation of power level within a given fuel cycle. The ratio C, is calculated for each fuel cycle using the adjoint transport methodology and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel cycle. For a single cycle irradiation C, =

1.0. Ilowever, for multiple cycle irradiations, panicularly those employing low leakage fuel management the additional C, correction must be utilized.

2-8

2.2.1.3 - Corrections to Reaction Rate Data Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 2.2.2 of this repon, corrections were made to the U-238 foil measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. In addition to the corrections made for the presence of U-235 in the U-238 fission sensors, corrections were also made to both the U-238 and Np-237 sensor reaction rates to account for gamma ray induced fission reactions occuring over the course of die irradiation.

In performing the dosimetry evaluations for the intemal surveillance capsules, the sensor reaction rates measured at the locations shown in Figure 2.1-2 were indexed to the geometric center of the capsules prior to use in the spectrum adjustment procedure. This indexing procedure required correcting the measured reaction rates by the application of analytically detennined spatial gradients. For the Beaver Valley Unit I surveillance capsules, the gradient correction factors for each sensor reaction were obtained from the reference fonvard transpon calculation and were used in a multiplicative fashion to relate individual measured reaction rates to the corresponding value at the geometric center of the surveillance capsule.

2.2.2 - Least Squares Adjustment Procedurr.

Values of key fast neutmn exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code"U. The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as inpat and proceeded to adjust the group fluxes fmm the trial spectrum to produce a best fit (in a least squares sense) to Ole measured reaction rate data. The " measured" exposure parameters along with the associated uncenainties were then obtained fmm the adjusted spectrum.

In the FERRET evaluations, a log-nonnal least squares algorithm weights both the trial values and the measured data in accordance with the assigned uncenainties and correlations. In general, the measured values f are linearly related to the flux $ by some response matrix A:

fl'* = s[ Al &f where i indexes the measured values belonging to a single data set s, g designates the energy gmup, and a delineates spectra that may be simultaneously adjusted. For example, 2-9 1

1 l

l l

R, = [ og 4, s

relates a set of measured reaction rates R, to a single spectrum Q, 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 quantities (i.e., neutron spectra and cross-sections) were approximated in a multi-group fonnat consisting of 53 energy groups. The trial input spectrmn was converted to the FERRET 53 group structure using the SAND-Il code"0 This procedure was carried out by first expanding the 47 gmup calculated spectrum into the SAND-Il 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET.

The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry fileU73, 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 cs a weighting function in the cross-section collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed fnn the information contained on the ENDF/B-VI data files. These matrices included energy group to energy group uncertainty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were not included. The omission of this additional uncertainty information does not significantiy 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 spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation was obtained from the plant specific calculation for each dosimetry location. While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the cmss-section data files, the covariance matrix for the input trial spectmm was constructed from the following relation:

M ,1 = R$ + R, R,1 P,1 where R,, specifies an overall fractional nonnalization uncertainty (i.e., complete correlation) for the set of values. The fractional uncertainties R, specify additional random uncertainties for group g that are corTelated with a correlation matrix given by:

2-10

e l b l

i 3

P,, = [1 -0] 5,, + 0 e -"

1, where:

1_

1. =W 2y 2 4

1 j The first temi in the correlation matrix equation specifies purely random uncertainties, while the i i

second tenn describes shon range correlations over a gmup range 7 (0 specifies the strength of the *

] latter tenn). The value of 6 is I when g = g' and 0 otherwise. For the trial spectnun used in the current evaluations, a shon range correlation of y = 6 groups was used. This choice implies that neighboring gmups are strongly correlated when 0 is close to 1. Strong long range correlations (or l

~

4 anti-correlations) were justified based on information presented by R. E. MaerkerD H. Maerter's results are closely duplicated when y = 6. For the integral reaction rate covariances, simple normalization  !

and nmdom uncenainties were combined as deduced from experimental uncenainties.

t 4'

In performing the least squares adjusunent with the FERRET code, the input spectra fmm the reference forward transpon calculations were normalized to the absolute calculations fmm the cycle ,

specific adjoint analyses. The specific normalization factors for individual evaluations depended on the location of the sensor set as well as on the neutron flux level at that location.

i The specific assignment of uncenainties in the measured reaction rates and the input (trial) spectra  ;

i used in the FERRET evaluations was as follows:

REACrlON RATE UNCERTAINTY 5% l FLUX NORM ALIZATION UNCERTAIN TY 30%

FLUX GROUP UNCERTAINTIES (E > 0.0055 MeV) 30 %

(0.68 ev < E < 0.0055 MeV) 58% -

(E < 0.68 ev) 104 %  !

i

. SHORT RANGE CORRELATION .l

! (E > 0.0055 MeV) 0.9  ;

(0.68 ev < E < 0.0055 MeV) 0.5 l

(E < 0.68 ev) 0.5 i 2-11 E

Y FLUX GROUP CORRELATION RANGE (E > 0.0055 MeV) 6 (0.68 ev < E < 0.0055 MeV) 3 (E < 0.68 ev) 2 It should be noted that the uncenainties listed for the upper energy ranges extend down to the lower range. Thus, the 58% group uncertainty in the second range is made up of a 30% uncertainty with a 0.9 shon range correlation and a range of 6, and a second pan of magnitude 50% with a 0.5 correlation and a range of 3.

These input uncertainty assignments were based on prior experience in using the FERRET least squares adjustment approach in the analysis of neutron dosimetry from surveillance capsule, mactor cavity, and benctunark irradiations. The values are liberal enough to permit adjustment of the input spectrum to fit the measured data for all practical applications.

2.3 - Detennination of Best Estimate Pressure Vessel Exposure -

As noted earlier in this report, the best estimate exposure of the reactor pressure vessel was developed using a combination of absolute plant specific transport calculations based on the methodology discussed in Section 2.1 and plant specific measurement data detemiined using the measurement evaluation techniques described in Section 2.2. In particular, the best estimate vessel exposure is obtained from the following relationship:

• saa ra. " K 6g where: @ ,..s, = The best estimate fast neutron exposure at the location of interest.

K = The plant specific measurement / calculation (M/C) bias factor derived from all available surveillance capsule dosimetry data.

$cm. = The absolute calculated fast neutron exposure at the location of interest.

The approach defined in the above equation is based on the premise that the measurement data represent the most accurate plant specific infonnation 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 2-12 I

analysis alone. That 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.

The implementation of this approach acts to remove plant specific biases associated with the definition of the core source, actual vs. assmned reactor dimensions, and operational variations in water density within the reactor. As a result, the overall uncenainty in the best estimate exposure pmjections within the vessel wall depend on the individual uncertainties in the measurement process, 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.

For the Beaver Valley Unit i neutron exposure evaluation, the uncertainties in the measured exposure parameters were derived directly from the results of the least squares evaluation of dosimetry data.

The positioning uncertainties were taken fmm parametric studies of sensor position perfomied as part of an analytical sensitivity evaluation of the Beaver Valley Unit I reactor geometry. The uncertainties I in the exposure ratios relating dosimetry results to positions within the vessel wall were based on ,

analytical sensitivity studies of the downcomer water density variations and vessel inner radius  !

tolerance for the surveillance capsule measurements.  !

2-13

l l

SECrlON 3.0 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 3.1 Reference Forward Calculation )

1 1

l As noted in Section 2. 0 of this report, data from the reference forward transport calculation were used in evaluating dosimetry from surveillance capsule irradiations as well as in relating the results of these evaluations to the neutron exposure of the pressure vessel wall. In this section, the key data extracted I from the reference forward calculation is presented and its relevance to the dosimetry evaluations and vessel exposure projections is discussed. The reader should recall that the results of the reference forward transport calculation were intended for use on a relative basis and, therefore, should not be used for absolute comparison with measurement. All absolute comparisons were based on the results of the fuel cycle specific adjoint calculations discussed in Section 3.2.

3.1.1 - Surveillance Capsule Locations Data from the reference fonvard calculation pertinent to surveillance capsule evaluations are provided in Tables 3.1-1 through 3.1-3.

In Table 3.1-1, the calculated neutron energy spectra at the geometric center of surveillance capsules located at 15,25,35, and 45 degrees relative to the core cardinal axes are listed. These data represent the trial spectra used as the starting guess in the FERRET least squares adjustment evaluations of the surveillance capsule sensor sets. On a relative basis, these calculated energy distributions establish a baseline against which adjusted spectra may be compared; and, when coupled with the adjoint results of Section 3.2, provide an analytical prediction of absolute neutron spectra at the sensor set locations for each irradiation period.

In Table 3.1-2, the calculated neutron sensor rextion rates and exposure rate ratios associated with the spectra from Table 3.1-1 are provided along with the calculated exposure rates in terms of Q(E > 1.0 MeV), $(E > 0.1 MeV) and dpa/sec. Again, these data are applicable to the geometric center of each surveillance capsule. Also listed are the associated exposure rate ratios calculated for each of the sensor set locations. These reference reaction rates, exposure rates, and exposure rate ratios were used in conjunction with fuel cy;le specific calculations from Section 3.2 to provide calculated sensor set reaction rates am - n *

• wasor set exposures in terms of $(E > 0.1 MeV),

and dpa/sec for each irradiation period.

As noted earlier in this report, surveillance capsule dosimetry evaluations also require spatial gradient corrections to be applied to measured reaction rates in sensors dispersed throughout the capsule. In the case of the Beaver valley Unit I surveillance capsules, neutron sensors were positioned within the 3-1

i I

l l

I specimen array as shown in Figure 2.1-2. In Table 3.1-3, gradient correction factors applicable to the various dosimetry locations are provided for each sensor reaction.

3.1.2 - Pressure Vessel Wall  !

Data from the reference forward calculation pertinent to the pressure vessel wall are provided in Tables 3.1-4 through 3.1-7.

1 In Table 3.14, die calculated azimuthal distribution of exposure rates in terms of Q(E > 1.0 MeV),

$(E > 0.1 MeV), and dpa/sec are listed at approximately 5 degree intervals over the reactor geometry.

These data are applicable to the clad / base metal interface. Also given in Table 3.1-4 are die exposum rate ratios [$(E > 0.1 MeV)]/[$(E > 1.0 MeV)] and [dpa/secl/[$(E > 1.0 MeV)] that provide an indication of the variation in neutron spectmm as a function of azimuutal angle at the pressure vessel inner radius.

Radial gradient information for Q(E > 1.0 MeV), Q(E > 0.1 Mev), and dpa/sec is given in Tables 3.1-5,3.1-6, and 3.1-7, respectively. These data are presented on a relative basis for each exposure parameter at the 0",15". 30", and 45* azimuthal locations. Exposure rate distributions within the vessel wall are obtained by normalizing the calculated or best estimate exposure at the vessel inner radius to the gradient data given in Tables 3.1-5 through 3.1-7.

3.2 - Fuel Cycle Specific Adjoint Calculations Results of the fuel cycle specific adjoint transport calculations for the first 1i cycles of operation at Beaver Valley Unit I are summarized in Tables 3.2-1 through 3.2-12. The data listed in these tables establish the means for absolute comparison of analysis and measurement for the three sets of surveillance capsule dosimetry withdrawn to date. These results also provide the fuel cycle specific relationship among the surveillance capsule measurement locations and key positions at the inner radius of the pressure vessel wall.

l The core power distributions used in the cycle specific fast neutron exposure calculations for Fuel Cycles I thmugh 11 were taken from the fuel cycle design reports applicable to Beaver Valley Unit 1"""**. The data extracted from the fuel cycle design reports represented cycle averaged relative fuel assembly powers and burnups as well as cycle averaged relative axial distributions. Therefore, the results of the adjoint evaluation provided data in tenns of fuel cycle averaged neutron flux which, ,

when mdtiplied by the appropriate fuel cycle length, produced the incremental fast neutron exposure j for th- (ml cycle.

The calculated fast neutron flux (E > 1.0 MeV) and cumulative fast neutron lluence at the center of surveillance capsules located at 15,25,35, and 45 degrees am provided for each of the i1 operating 3-2

i i

fuel cycles in Tables 3.2-1 and 3.2-2, respectively. The data as tabulated are applicable to the axial j core midplane. Similar data applicable to the pressure vessel irmer radius are given in Tables 3.2-3 1 and 3.2-4.

Exposure parameter ratios necessary to convert the cycle specific data listed in Tables 3.2-1 through 3.2-4 to other key fast neutron exposure units are given in Section 3.1 of this report. Application of these ratios to the data from Tables 3.2-1 through 3.2-4 yielded corresponding exposure data in tenns i l

of flux / fluence (E > 0.1 MeV) (Tables 3.2.5 through 3.2.8) and iron atom displacements (Tables 3.2.9 l through 3.2.12).

l l

l 3-3

. _ . . . , - = _ . __. _

.m.m.. __ _ _ . - _ - . _ m i _. _.

• _ _ _ . _ _ . . . .

J TABLE 3.1-1 CA.LCULATED REFERENCE NEUTRON ENERGY SPECTRA 2

SURVEILLANCE CAPSULE CENTER [n/cm -sec] ,

2652 MWt; F, = 1.2 I

LOWER '

ENERGY AZIMUTilAL ANGLE  !

(MeV) 15.0" 25.0" 35.0 45.0 I 1.42E+01 1.65E+07 1.32E+07 1.04E+07 9.60E+06 i 1.22E+01 5.37E+07 4.22E+07 3.29E+07 2.99E+07 l.00E+01 2.39E+08 1.85E+08 1.41E+08 1.26E+08 8.61E+00 4.73E+08 3.62E+08 2.73E+08 2.41E+08 7.41E+00 8.46E+08 6.38E+08 4.73E+08 4.10E+08 6.07E+00 2.1 IE+09 1.58E+09 1.I6E+09 9.93E+08  !

4.97E+00 3.31E+09 2.42E+09 1.74E+09 1.46E+09 .

3.68E+00 6.96E+09 4.88E+09 3.40E+09 2.75E+09 l 3.01E+00 5.77E+09 3.90E+09 2.65E+09 2.10E+09 2.73E+00 4.52E+09 3.05E+09 2.06E+09 1.62E+09 2.47E+00 5.38E+09 3.59E+09 2.41E+09 1.89E+09 2.37E+00 2.69E+09 1.80E+09 1.20E+09 9.38E+08 2.35E+00 7.51E+08 4.99E+08 3.34E+08 2.61E+08 2.23E+00 3.79E+09 2.51E+09 1.68E+09 1.31E+09 1.92E+00 1.05E+10 6.89E+09 4.59E+09 3.55E+09 1.65E+00 1.26E+10 8.08E+09 5.35E+09 4.l lE+09 1.35E+00 1.92E+10 1.22E+ 10 8.06E+09 6.16E+09 1.00E+00 3.63E+ 10 2.25E+10 1.48E+10 1.12E+10 8.21E-01 2.50E+ 10 1.53E+10 9.96E+09 7.51E+09  :

7.43E-01 1.26E+10 7.69E+09 5.01E+09 3.78E+09 )

6.08E-01 3.72E+10 2.21E+10 1.43E+10 1.07E+10 4.98E-01 3.02E+ 10 1.77E+10 1.14E+10 8.52E+09 3.69E-01 3.28E+10 1.94E+10 1.25E+10 9.33E+09 NOTE: The upper energy of gmup 1 is 17.33 Mev.

3-4

i

. 1 1

TABLE 3.1-1 (continued) i CALCULATED REFERENCE NEUTRON ENERGY SPECTRA j 2

SURVEILLANCE CAPSULE CENTER [rvem -sec]

2652 MWt; F, = 1.2 i l

LOWER  :

ENERGY AZIMUTHAL ANGLE

]

(MeV1 15.(f 25.ff 35.0 45.(f '

2.97E-01 3.09E+ 10 1.81E+ 10 1.I6E+10 8.63E+09 1.83E-01 4.07E+10 2.38E+ 10 1.53E+10 1.14E+10 1.1IE-01 3.87E+ 10 2.25E+10 1.44E+10 1.07E+ 10 6.74E-02 2.89E+10 1.68E+10 1.07E+ 10 7.97E+09 4.09E-02 2.36E+10 1.37E+10 8.75E+09 6.49E+09 )

3.18E-02 8.31E+09 4.81E+09 3.08E+09 2.29E+09 4 2.61E-02 3.44E+09 2.00E+09 1.28E+09 9.58E+08 2.42E-02 8.39E+09 4.83E+09 3.09E+09 2.28E+09 l 2.19E-02 5.26E+09 3.01E+09 1.93E+09 1.42E+09 j 1.50E-02 1.I6E+10 6.67E+09 4.27E+09 3.I 8E+09 j 7.1OE-03 2.22E+ 10 1.28E+10 8.22E+09 6. I 1E+09 j 3.36E-03 2.72E+ 10 1.57E+ 10 1.00E+ 10 7.45E+09 l 1.59E-03 2.53E+ 10 1.46E+10 9.28E+09 6.89E+09 4.54E-04 4.13E+10 2.37E+ 10 1.51E+ 10 1.12E+10 2.14E-04 2.25E+ 10 1.29E+10 8.20E+09 6.07E+09 1.01E-(M 2.52E+ 10 1.44E+ 10 9.I 8E+09 6.79E+09 l 3.73E-05 3.32E+ 10 1.89E+10 1.20E+ 10 8.89E+09 l 1.07E-05 4.(ME+ 10 2.30E+ 10 1.46E+10 1.08E+10 5.(ME-06 2.29E+ 10 1.30E+ 10 8.24E+09 6.08E+09 1.86E-06 2.98E+ 10 1.69E+10 1.07E+10 7.88E+09 l 8.76E-07 2.17E+ 10 1.23E+ 10 7.77E+09 5.73E+09 4.14E-07 1.74E+10 9.82E+09 6.24E+09 4.61E+09 1.00E-07 3.47E+ 10 1.96E+10 1.25E+10 9.21E+09 0.00 7.84E+10 4.39E+10 2.84E+10 2.10E+10 NOTE: The upper energy of group 1 is 17.33 Mev.

I i

i 3-5

TABLE 3.1-2 REFERENCE NEUTRON SENSOR REACTION RATES AND EXPOSURE PARAMETERS AT THE SURVEILLANCE CAPSULE CENTER 2652 MWt; F, = 1.20 i l AZIMUTHAL ANGLE l 15.(P 25.(F 35.0" 45.0"  !

Reaction Rate (rps/ nucleus) j Cu-63(n.tx) (Cd) 7.53E- 17 5.67E-17 4.20E-17 3.64E-17 Fe-54(n.p) - (Cd) 8.37E-15 5.91E-15 4.16E-15 3.41E-15 Ni-58(n.p) (Cd) 1.15E- 14 8.05E-15 5.63E-15 4.59E- 15 f

U 238(n,0 (Cd) 3.97E- 14 2.66E-14 1.80E-14 1.43E-14 i i Np-237(n,0 (Cd) 2.98E- 13 1.88E-13 1.24E-13 9.55E-14 4 Co-59(n.7) 4.91E-12 2.78E-12 1.78E-12 1.32E-12 Co-59(n,y) (Cd) 2.34E-12 1.34E- 12 8.50E-13 6.29E-13 U-238(y,0 1,78E-15 1.09E-15 7.79E-16 6.22E- 16 Np-237(y,0 4.95E-15 3.03E-15 2.17E-15 1.74E-15 Neutron Flux (n/cm 2-sec)

{

$(E > 1.0 MeV) 1,16E+11 - 7.52E+10 5.04E+ 10 3.91E+ 10 ,

$(E > 0.1 MeV) 3.70E+ 11 2.25E+ 11 1.47E+11 1. I IE+11 dna/sec Displacement Rate 1.89E-10 1.20E-10 7.97E-11 6.16E- 11 i

$(E > 0.1)/$(E > 1.0) 3.20 2.99 2.92 2.85

[dpa/sec]/$(E > 1.0) 1.63E-21 1.59E-21 1.58E 21 1.57E-21 U238(y,0/U238(n,0 0.045 0.041 0.043 0.043 i Np237(y,0/Np237(n.0 0.017 0.016 0.018 0.018 l

3-6 i

e i i

TABLE 3.1-3 RADIAL GRADIENT CORRECTIONS FOR SENSORS CONTAINED IN l INTERNAL SURVEILLANCE CAPSULES i

15' CAPSULE f Radial Location (cm) l 190.93 191.16 191.41 191.93 l Cu-63(n.a) 0.956 l Fe-54(n.p) 1.050

-l Ni-58(n.p) 1.158 U-238(n,0 Cd 1.000 Np-237(n,0 Cd 1.0(0  ;

Co-59(nty) 0.957  :

Co-59(n,y) Cd 1.157 J

b

)

l 25' CAPSULE Radial Location (cm) 190.93 191.16 191.43 191.93 Cu-63(n,a) 0.956 Fe-54(n.p) 1.051 Ni-58(n.p) 1.163 U-238(n,0 Cd 1.0C0 Np-237(n,0 Cd 1.000 Co-59(n,y) 0.956 Co-59(n,y) Cd 1.155 l

l 1

I 3-7

(

I

TABLE 3.1-4

SUMMARY

OF EXPOSURE RATES AT THE PRESSURE VESSEL l CLAD / BASE METAL INTERFACE l

1 FLUX (n/cm2 sec)

THETA IE > 0.11 dna/sec j f.d, egl (E > 1.0) (E > 0.1) doa/sec f E > l .01 IE > l .01 l 0.79 7.02E+ 10 1.88E+11 1.13E-10 2.68 1.61E-21 5.66 6.5 iE+ 10 1.74E+11 1.05E-10 2.68 1.61E-21 10.44 5.03E+10 1.36E+ 11 8.I3E-11 2.71 1.62E-21 15.00 3.44E+10 9.20E+10 5.56E-11 2.68 1.62E-21 20.58 2.47E+ 10 6.42E+10 3.98E-I l 2.60 1.61E-21 25.00 2.19E+10 5.64E+10 3.51E-I l 2.57 1.60E-21 29.90 1.87E+10 4.78E+10 2.99E-11 2.56 1.60E-21 35.00 1.53E+10 3.87E+ 10 2.45E-I l 2.53 1.60E-21 39.90 1.28E+10 3.22E+ 10 2.05E-11 2.52 1.60E-21 44.91 1.23E+10 3.07E+ 10 1.97E-11 2.50 1.60E-21 i J

l l

i 3-8

TABLE 3.1-5 RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 1.0 MeV)

WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) (P 15" 3ff 45*

199.95 1.000 1.000 1.000 1.000 200.54 0.959 0.964 0.961 0.960 201.72- 0.858 0.862 0.860 0.857 202.89 0.749 0.755 0.753 0.750 204.07 0.646 0.655 'O.651 0.651 204.95 0.577 0.589 0.583 0.5 84 205.25 0.554 0.566 0.561 0.562 206.42 0.473 0.487 0.481 0.483 207.60 0.403 0.417 0.411 0.414 208.78 0.342 0.357 0.350 0.354 209.95 0.290 0.305 0.298 0.302 211.13 0.246 0.259 0.253 0.257 212.30 0.208 0.220 0.215 0.218 213.48 0.175 0.187 0.182 0.185 214.66 0.147 0.I58 0.154- 0.157 214.95 0.141 0.152 0.148 0.151 215.83 0.123 0.I34 0.130 0.133 217.01 0.102 0.I12 0,109 0.112 218.19 0.084 0.094 0.091 0.095 219.36 0.067 0.077 0.076 0.080 219.95 0.064 0.074 0.073 0.077 Note: Base Metal Inner Radius = 199.95 cm.

Base Metal 1/4T = 204.95 cm.

Base Metal 1/2T n 209.95 cm.

Base Metal 3/4T = 214.95 cm.

Base Metal Outer Radius = 219.95 cm.

3-9

. 1 TABLE 3.1-6 RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 0.1 MeV)

WITillN TIIE PRESSURE VESSEL WALL RADIUS AZIMUTilAL ANGLE (cm) (f 15* 3ff 45" 199.95 1.(X10 1.(XX) 1.000 1.000 200.54 1.005 1.013 1.009 1.01I 201.72 0.984 0.998 0.992 0.993 202.89 0.942 0.963 0.955 0.958 204.07 0.891 0.920 0.909 0.915 204.95 0.850 0.885 0.872 0.880 205.25 0.837 0.873 0.860 0.869 206.42 0.782 0.824 0.809 0.820 207.60 0.727 0.774 0.757 0.771 208.78 0.674 0.723 0.707 0.722 209.95 0.621 0.673 0.657 0.674 211.13 0.570 0.623 0.608 0.626 212.30 0.520 0.575 0.560 0.579 213.48 0.472 0.526 0.513 0.533 214.66 0.425 0.479 0.468 0.488 214.95 0.413 0.467 0.457 0.477 215.83 0.379 0.433 0.423 0.444 217.01 0.333 0.387 0.380 0.402 218.19 0.287 0.342 0.338 0.360 219.36 0.239 0.294 0.296 0.320 219.95 0.228 0.285 0.287 0.313 i

Note: Base Metal Inner Radius = 199.95 cm. l Base Metal 1/4T = 2N.95 cm. I Base Metal 1/2T = 209.95 cm.

Base Metal 3/4T = 214.95 cm.

Base Metal Outer Radius = 219.95 cm.

3-10

. _ - _= ._ - - , .--

TABLE 3.1-7 RELATIVE RADIAL DISTRIBUTION OF IRON DISPLACEMENT RATE (dpa) l- WITHIN TiiE PRESSURE VESSEL WALL l

, RADIUS AZIMUTilAL ANGLE (cm) 0" 15* 30" 45*

199.95 1.(XX) 1.(XX) 1.000 1.(XX) 200.54 0.967 0.973 0.969 0.968 201.72 0.888 0.896 0.890 0.888 202.89 0.804 0.816 0.808 0.805 2(M.07 0.723 0.740 0.728 0.727 204.95 0.668 0.687 0.674 0.674 205.25 0.649 0.670 0.655 0.656 206.42 0.582 0.605 0.589 0.591 207.60 0.520 0.546 0.530 0.532 208.78 0.466 0.493 0.476 0.479 209.95 0.416 0.444 0.427 0.432 211.13 0.371 0.400 0.384 0.388 212.30 0.330 0.359 0.344 0.349 213.48 0.292 0.321 0.307 0.313 i 214.66 0.258 0.287 0.274 0.280 j 214.95 0.250 0.278 - 0.266 0.273  ;

215.83 0.226 0.254 0.244 0.250  !

217.01 0.1% 0.224 0.215 0.223 218.19 0.168 0.1% 0.189 0.198 219.36 0.139 0.169 0.166 0.175 219.95 0.133 0.164 0.161 0.172 i

i Note: Base Metal Inner Radius = 199.95 cm.

Base Metal 1/4T = 2(M.95 cm.

Base Metal I/2T = 209.95 cm.

Base Metal 3/4T = 214.95 cm. .

Ba : Metal Outer Radius = 219.95 cm.

l l 3-11

d i

l TABLE 3.2-1 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES NEUTRON FLUX  :

2 (n/cm -sec)

CYCLE No. IS' 25' 35" 45' I 9.30E+ 10 6.01E+10 4.07E+10 3.21E+10 2 1.09E+11 7.12E+10 4.88E+ 10 3.85E+10  !

3 1.00E+11 6.36E+10 4.26E+ 10 3.29E+10 4 8.00E+ 10 5.15E+10 . 3.41E+ 10 2.66E+ 10 5 7.74E+ 10 4.95E+ 10 3.34E+10 2.66E+10 6 6.86E+ 10 5.04E+10 3.39E+ 10 2.64E+ 10 7 7.47E+ 10 4.82E+ 10 3.22E+10 2.57E+10 7

8 7.77E+10 5.1IE+10 3.30E+ 10 2.57E+10 9 7.07E+10 4.89E+10 - 3.43E+10 2.81E+ 10 10 6.18E+10 4.53E+ 10 3.43E+ 10 2.90E+10 11 6.40E+ 10 4.65E+10 3.43E+10 2.79E+10 t

l l

3-12

.D ,

i TABLE 3.2-2 i

CALCULATED FAST NEUTRON FLUENCE (E > 1.0 MeV) AT THE ,

CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES IRRADIATION CUMULATIVE FLUENCE e

END OF TIME 2 (n/cm )

CYCLE (EFPS) 15* 25" 35" 45" 1- 3.66E+07 3.40E+ 18 2.20E+ 18 1.49E+17 1.I8E+17 2 5.91E+07 5.85E+18 3.80E+ 18 2.59E+ 18 2.(ME+ 18 3 8.40E+07 8.34E+18 5.39E+18 3.65E+18 2.86E+ 18 i 4 1.13E+08 1.07E+ 19 6.88E+18 4.64E+18 3.63E+18 5 1.51E+08 1.36E+19 8.74E+18 5.90E+ 18 4.64E+ 18 ,

6 1.87E+08 1.61E+19 1.06E+ 19 7.13E+ 18 5.59E+18 7 2.26E+08 1.90E+19 1.25E+19 8.40E+ 18 6.61E+ 18 8 2.61E+08 2.17E+19 1.43E+19 9.55E+ 18 7.50E+ 18 9 3.05E+08 2.48E+ 19 1.64E+19 1.10E+19 8.72E+ 18 10 3.42E+08 2.71E+19 1.81E+19 1.23E+19 9.81E+ 18 ,

I i

)

I 3-13 ,

I

.=. . - _ . . .- -. . . .

TABLE 3.2-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE NEUTRON FLUX

) 2 (n/cm sec)

CYCLE No. (f 15" 30' 45*

1 5.65E+ 10 2.76E+ 10 1.50E+ 10 1.01E+10 <

2 6.55E+ 10 3.23E+ 10 1.78E+10 1.2]E+10 3 3 6. l lE+10 2.96E+ 10 1.58E+10 1.04E+ 10 4 4.77E+ 10 2.38E+10 1.27E+10 8.44E+09 l 5 4.60E+10 2.30E+10 1.23E+10 8.44E+09 6 3.59E+ 10 2.05E+10 1.26E+10 - 8.38E+09 ,

7 4.22E+10 2.22E+ 10 1.19E+10 8.16E+09 8.18E+09 8 4.31E+ 10 2.31E+ 10 1.25E+ 10

• 9 3.82E+ 10 2.10E+10 1.24E+ 10 8.86E+09 l

10 3.57E+ 10 1.86E+10 1.19E+ 10 9.12E+09 l 11 3.45E+ 10 1.91E+ 10 1.22E+10 8.82E+09 4

i l

3-14

___ - __ ~. . -

.- /

TABLE 3.2-4 CALCULATED FAST NEUTRON FLUENCE (E > 1.0 MeV) AT TiiE PRESSURE VESSEL CLAD / BASE METAL INTERFACE IRRADIATION CUMULATIVE FLUENCE l 2 END OF TIME (n/cm )

CYCLE (EFPS) O' 15' 3ff 45' i 1 3.66E+07 2.07E+18 1.01E+18 5.48E+17 3.70E+17

, 2 5.91E+07 3.54E+ 18 1.73E+18 9.50E+17 6.42E+ 17 3 8.40E+07 5.06E+18 2.47E+ 18 1.34E+18 9.00E+ 17 4 1.13E+08 6.45E+18 3.16E+ 18 1.71E+18 1.15E+18  !

5 1.51E+08 8.18E+ 18 4.03E+ 18 2.18E+ 18 1.46E+18 6 1.87E+08 9.48E+ 18 4.77E+ 18 2.63E+18 1.77E+18 7 2.26E+08 1.12E+19 5.65E+18 3.10E+18 2.09E+ 18 8 2.61E+08 1.27E+19 6.45E+18 3.54E+ 18 2.37E+ 18 9 3.05E+08 1.43E+19 7.36E+18 4.08E+18 2.76E+18 10 3.42E+08 1.57E+19 8.06E+ 18 4.52E+ 18 3.10E+ 18 l

l 1

3-15 l

l TABLE 3.2-5 l

CALCULATED FAST NEUTRON FLUX (E > 0.1 MeV) AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES ,

NEUTRON FLUX ,

2 (n/cm -sec) ,

CYCLE No. 15* 25* 35" 45" 1 2.98E+ 11 1.80E+11 1.19E+11 9.15E+ 10 2 3.49E+ i t 2.13E+11 1.43E+11 1.10E+11 ,

3 3.20E+ 11 1.90E+11 1.24E+11 9.38E+ 10 4 2.56E+11 1.54E+11 9.95E+10 7.58E+10 5 2.48E+ 11 1.48E+11 9.75E+10 7.59E+10 6 2.20E+ 11 1.51E+11 9.91E+ 10 7.52E+10 ,

7 2.39E+ 11 1.44E+11 9.40E+ 10 7.33E+10 8 2.49E+11 1.53E+11 9.64E+10 7.32E+10 9 2.26E+ 11 1.46E+11 1.00E+11 8.01E+10 10 1.98E+11 1.35E+11 1.00E+ 11 8.27E+10 11 2.05E+ 11 1.39E+ 11 1.00E+11 7.94E+ 10 3-16

- . .- . . - . - ..- . = . . .- . - . .

I =

I TABLE 3.2-6 CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT TiiE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES IRRADIATION CUMULATIVE FLUENCE 2

END OF TIME (n/cm )

CYCLE (EFPS) 15* 25* 35* 45*

1 3.66E+07 1.09E+19 6.58E+18 4.35E+ 18 3.35E+18 l 2 5.91E+07 1.87E+19 1.I4E+19 7.56E+18 5.82E+18 3 8.40E+07 2.67E+ 19 1.61E+19 1.07E+19 8.1SE+ 18 4 1.13E+08 3.41E+19 2.06E+ 19 1.36E+19 1.04E+ 19 5 1.51E+08 4.35E+ 19 2.6 iE+ 19 1.72E+19 1.32E+ 19 6 1.87E+08 5.14E+19 3.16E+ 19 2.08E+19 1.59E+19 7 2.26E+08 6.08E+ 19 3.73E+ 19 2.45E+19 1.88E+19 8 2.61E408 6.95E+19 4.26E+19 2.79E+19 2.14E+ 19 9 3.05E+08 7.93E+19 4.90E+19 3.22E+19 2.49E+19 10 3.42E+08 8.68E+19 5.41E+19 3.60E+19 2.80E+ 19 l

l

\

l r 3-17

e TABLE 3.2-7 i

CALCULATED FAST NEUTRON FLUX (E > 0.1 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE NEUTRON FLUX 2

(n/cm -sec)

CYCLE No. (f 15* 3(f 45' i 1.51E+11 7.39E+10 3.84E+ 10 2.53E+ 10 2 1.76E+11 8.64E+10 4.57E+10 3.03E+ 10 3 1.64E+11 7.93E+ 10 4.04E+10 2.59E+10 4 1.28E+11 6.38E+ 10 3.26E+10 2.l lE+10 -

5 1.23E+ 11 6.17E+ 10 3.15E+10 2.11E+ 10 6 9.63E+ 10 5.48E+ 10 3.22E+ 10 2.10E+ 10 7 1.13E+11 5.95E+10 3.05E+10 2.04E+ 10 8 1.15E+11 6.19E+10 3.19E+10 2.05E+ 10 9 1.02E+11 5.63E+10 3.17E+10 2.21E+ 10 l 10 9.56E+ 10 4.97E+10 3.06E+10 2.28E+ 10 11 9.25E+10 5.13E+10 3.12E+ 10 2.21E+10 3-18 l

TABLE 3.2-8 CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT Tile PRESSURE VESSEL CLAD / BASE METAL INTERFACE ,

1 IRRADIATION CUMULATIVE FLUENCE 2

END OF TIME (n/cm )

CYCLE (EFPS) 0" 15' 30' 45*

I 3.66E+07 5.54E+18 2.70E+18 1.40E+18 9.24E+ 17 2 5.91E+07 9.49E+18 4.65E+18 2.43E+ 18 1.6iE+18 3 8.40E+07 1.36E+19- 6.62E+ 18 3.44E+18 2.25E+ 18 4 1.13E+08 1.73E+19 8.48E+ 18 4.38E+ 18 2.86E+18 5 1.51E+08 2.19E+19 1.08E+19 5.57E+ 18 3.66E+ 18 6 1.87E+08 2.54E+19 1.28E+19 6.74E+ 18 4.42E+ 18 7 2.26E+08 2.99E+ 19 1.5iE+19 7.94E+ 18 5.22E+18 8 2.61E+08 3.39E+19 1.73E+19 9.05E+ 18 5.93E+18 9 3.05E+08 3.84E+ 19 1.97E+19 1.04E+19 6.90E+18 10 3.42E+08 4.19E+19 2.16E+ 19 1.16E+19 7.75E+18 I

3-19

i 4

l TABLE 3.2-9 CALCULATED IRON ATOM DISPLACEMENT RATE AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES DISPLACEMENT RATE (dpa/sec)

CYCLE No. IS* 25" 35' 45" i 1.52E-10 9.55E-11 6.44E-11 5.NE-1 I 2 1.78E-10 1.13E-10 7.71E-I l 6.05E-I l 3 1.63E-10 1.01E-10 6.73E-I l 5.17E-I l 4 1.30E-10 8.19E- 11 5.38E-11 4.17E-11 5 1.26E-10 7.87E-I l 5.28E-Il 4.18E-I l 6 1.12E-10 8. ole-Il 5.36E-Il 4.14E-I l 7 1.22E-10 7.66E-I l 5.09E-I l 4.04E-Il 8 1.27E-10 8.13E-11 5.22E-11 4.03E-11 j 9 1.15E-10 7.78E-I l 5.42E-I l 4.41E-I l 10 1.01E-10 7.20E-I l 5.42E-Il 4.55E-I l ii 1.04E-10 7.40E-11 5.41E-11 4.38E-1 I I

3-20

[

l TABLE 3.2-10 CALCULATED IRON ATOM DISPLACEMENTS AT THE  ;

CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES

, IRRADIATION CUMULATIVE DISPLACEMENTS END OF TIME (dpa) .

CYCLE (EFPS) 15* 25* JS*_ 45*

1 3.66E+07 5.55E-03 3.50E-03 2.36E-03 1.84E-03 2 5.91E+07 9.54E-03 6.04E-03 4.09E-03 3.20E-03 3 8.40E+07 1.36E-02 8.56E-03 5.77E-03 4.49E-03 4 1.13E+08 1.74E-02 1.09E-02 7.33E-03 5.71E-03 ;

5 1.51E+08 2.21E-02 1.39E-02 9.32E-03 7.28E-03 6 1.87E+08 2.62E-02 1.68E-02 1.13E-02 8.78E-03 j 7 2.26E+08 3.10E-02 1.98E-02 1.33E-02 1.04E-02 8 2.61E+08 3.54E-02 2.27E-02 1.51E-02 1.I8E-02 9 3.05E+08 4.04E-02 2.61E-02 1.74E-02 1.37E-02 10 3.42E+08 4.42E-02 2.88E-02 1.95E-02 1.54E-02 3-21

TABLE 3.2-11 CALCULATED IRON ATOM DISPLACEMENT RATE AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE DISPLACEMENT RATE (dpa/sec)

CYCLE No. 0" 15" 30" 45' I 9.09E- 11 4.47E-11 2.40E-11 1.62E-11 2 1.06E-10 5.22E-1I 2.86E-11 1.94E-11 3 9.84E-I l 4.79E-I l 2.53E-I l I.66E-I l 4 7.68E- 1 I 3.86E-1 I 2.(ME-11 1.35E-11 5 7.41E- 11 3.73E- 11 1.97E-11 1.35E-11 6 5.78E-I l 3.31E-I l 2.01E-I l 1.34E-I l 7 6.80E-I l 3.60E-I l 1.91E-I l 1.31E-11 8 6.93E-11 3.74E-11 2.00E- 11 1.31E-11 9 6.14E-11 3.40E-11 1.98E-11 1.42E- 11 10 5.74E-11 3.01E-11 1.91E-11 1.46E-11 11 5.56E-11 3.10E-11 1.95E-11 1.41E-11

.1 1

3-22 i

1 l

i t

TABLE 3.2-12  ;

f CALCULATED IRON ATOM DISPLACEMENTS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE j IRRADIATION CUMULATIVE FLUENCE END OF TIME 2 (n/cm )

CYCLE (EFPS) ff 15 30 45* I i 3.66E+07 3.33E-03 1.63E-03 8.77E-(M 5.92E-(M

. 2 5.91E+07 5.70E-03 2.81E-03 1.52E-03 1.03E-03 3 8.40E+07 8.15E-03 4.00E-03 2.15E-03 1.44E-03 4 1.13E+08 1.04E-02 5.13E-03 2.74E-03 1.83E-03 i 5 1.51E+08 1.'32E-02 6.53E-03 3.48E-03 2.34E-03 6 1.87E+08 1.53E-02 7.73E-03 4.21E-03 2.83E-03

l 7 2.26E+08 1.80E-02 9.15E-03 4.96E-03 3.34E-03 i 8 2.61E+08 2.04E-02 1.05E-02 5.66E-03 3.80E-03 9 3.0$E+08 - 2.30E-02 1.19E-02 6.52E-03 4.42E-03 10 3.42E+08 2.52E-02 1.31E-02 7.24E-03 4.96E-03

, t 1

a

'I I

r 1

b 1

i l

3-23 l

i I

l l

\ .

l SECTION 4.0 EVALUATIONS OF SURVEILLANCE CAPSULE DOSIMETRY In diis section, the results of the evaluations of the three neutron sensor sets wididrawn as a part of the Beaver Valley Unit 1 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

AZIMUTHAL WITHDRAWAL IRRADIATION CAPSULE ID LOCATION TIME TIME (EFPS)

V 15* END OF CYCLE 1 3.66E+07 U 25" END OF CYCLE 4 1.13E+08 W 25 END OF CYCLE 6 1.87E+08 4.1 - Measured Reaction Rates Radiometric counting of each of these capsule dosimetry data sets was accomplished by Westinghouse using the pmcedures discussed in Section 2.0 of this report. The measured specific activities are provided in Tables 4.1-1 thmugh 4.1-3 for Capsules V, U, and W, respectively. The irradiation history of the Beaver Valley Unit I reactor during the finit 6 fuel cycles is listed in Table 4,1-4. The irradiation history was obtained from NUREG-0020," Licensed Operating Reactors Status Summary Report for the applicable operating periods.

Since Capsules U and W experienced irradiation over multiple fuel cycles, flux level adjustment factors for each cycle were also required as input to the reaction rate calculatiins. Appropriate adjustment factors for the Capsule U and W analyses were determined from the fuel cycle specific adjoint calculations described in Section 3.2 of this report.

Based on the irradiation history, the individual sensor characteristics, capsule gradient corrections, flux level adjustment factors, and the measured specific activities, reaction rates averaged over the appropriate irradiation periods and referenced to a core power level of 2652 MWt were computed for the sencor sets removed from Capsules V, U, and W. The computed reaction rates for the multiple foil sensor sets from each of the three intemal surveillance capsules are provided in Table 4.1-5.

In regard to tne data listed in Table 4.1-5, the fission rate measurements for the U-238 sensors include corrections for U-235 impurities, the build-in of Plutonium isotopes during the long irradiations, and j for the effects of 7.f reactions. Likewise, the fission rate measurements for the Np-237 sensors include adjustments for 7,f reactions occuring over the course of the respective irradiation periods.

l 4-1

4.2 - Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the three sets of surveillance capsule dosimetry are provided in Tables 4.2-1 through 4.2-3. In these tables, the derived exposure experienced by the capsule along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations are the 10 uncertainties associated with each of the derived exposure rates.

. In regard to the comparisons listed in Tables 4.2-1 through 4.2-3, it should be noted that the columns labeled " trial calc" were obtained by nomializing the neutron spectral data from Table 3.1-1 to the absolute calculated neutron flux (E > 1.0 MeV) averaged over the applicable irradiation periods (Cycle i for Capsule V, Cycles I through 4 for Capsule U, and Cycles I through 6 for Capsule W). Thus, the comparisons illustrated in Tables 4.2-1 through 4.2-3 indicate the degree to which the calculated neutron energy specta matched the measured sensor data before and after adjustment. Absolute comparisons are discussed further in Section 5.0 of this report.

Based on the d (simetry evaluations described in this report, Capsule V was irradiated for a period of 1.16 effective full power years and experienced the following integrated fast neutron exposun :

2

$ (E > 1.0 Mev) = 3.16e+18 1 8% n/cm 2

& (E > 0.1 Mev) = 1.07e+19 1 15% n/cm dpa = 5.31e-03 1 10% dpa Capsule U was irradiated for a period of 3.58 effective full power years and experienced the following integrated fast neutmn exposure:

2

$ (E > 1.0 Mev) = 6.91e+18 1 8% n/cm 2

& (E > 0.1 Mev) = 2.26e+19 1 15% n/cm dpa = 1.15e-02 1 10% dpa Capsule W was irradiated for a period of 5.92 effective full power years and experienced the following integrated fast neutron exposure:

2

$ (E > l.0 Mev) = 9.15c+18 1 9% n/cm 2

& (E > 0.1 Mev) = 2.93c+19 1 15% n/cm dpa = 1.51e-02 i 10% dpa 4

4 f

4-2 i

l TABLE 4.1-1 MEASURED SENSOR SPECIFIC ACTIVITIES FOR CAPSULE V ClW.WCAL ANALY115' REPORT"I'y ,WESTINADVANGD REACTOR BlVill0N - - -

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thT F1Ln 4ba Co-60 M n -s+ L,o -"

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i. T - 12 e. 5* Al.-Ce (c1) Lee 9 x m3 96 - l 70 s- R- 5.94

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- . 9.57 5 fo' 26 -17 ese Fe -

5.3F x 6o ' 8

'27 -l709 Cu. 4.14 *10' -

• M*h. 12. -97Io Ni

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.R t. % - 17 & Fa 5.57 x 10' anuna ,

cwvLes CS-Ib7 dfs />g ' (9-/6-Vo )

Np -237 70-8719 .NP~E31 9.70

• IC#'

l) .qSY 90 -17:;o {} - 23V I. %

• 10**

Dooorsne Llwr Dosonermes - DLQ-4-3

TABLE 4.1-2 MEASURED SENSOR SPECIFIC ACTIVITIES FOR CAPSULE U I

onmicamym espoer mustzusMouns Anymoun maner synnues' AseM.YTICE. LAacRATURrss amas.asov soouest se inaa.Tz uru. srTE ( 19 g,

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aWit'ars eiasantTsas BEAVE R V A U.EY UN4T

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'U Q APsow ths m cT N A Lr e K Mn 7 m at.seev DosMcTRY AC.T tV 4 T T M ATERe AL ISOTOPE APRtt 3.1985 & f 2. foe 85-900 U enes u t)n o . Cs - <37 188.6 dxs /mr of t]-33e

- not He e.sma Wo o, cs - 137 213 R dys ! $ eS Ne437

- 401 "r'?$ en A l es - La lt74O d.as ll<., e f wsu

- 903 l Co Al en co Lo 4271

-904 Fe. M. 5V f 2. I 'l

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top Meb Mi to SS 4%10

- 9 0 (, e u. Co-LO 44.42 907

FC Mw-54 iI33

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Fe. Mw-sv 1222.

9 st Bo7 Nih Me co 68 492)

- 01,2. eu f.o 6 0 9.9.01

=Q3 L T Fe. Ma S4 i I (ao j ~ 4 r4 Ber (% Af ta 60 I1890

- 9 16 l O_n Al ch to - (oo 4178

- 9 f t, Y Fe- Mu-54 II43 4' 4-4 I

1 TABLE 4.1-3

]

MEASURED SENSOR SPECIFIC ACTIVITIES FOR CAPSULE W Westinghouse Elsetric Corporation Advanced Energy Systeens - Analytical Laboratory REPORT Waltz Mill Site Reoueste 13468 Originator S.L. Anderson (WINTD, Energy Center (478) 5.E.Yantenko (WISTSD, R&D Center 781 Blog. (298) Received: 7/27/88 Reported: 8/11/88

! IRESULTS OF ANALYSIS 3 J

W Dosametrys Seaver Valley Unit #1, Caosule 89*

Originator Lat. Dosteeter (AUG.10,1988)

ID Samole # Material Nuclide dos /ag

• 2 mQua FISSON MONITORS U-238 LCST U-238 Cs-137 (See remarks)

No-237 48-1175 No-237 Cs-137 2.57E+e3 +/- 2.4E+81 TOP WIRES i AICo 88-1176 A1Co Co-64 1.36E+44 +/- 1.4E+02 A1Co(Cd) 88-1177 A1Co Co-68 4.93E+83 +/- 9.5E+41 Je 88-1179 Fe Mn-54 1.90E+43 +/- 9.8E+00 4

TOP MID

, Cu 88-1188 Cu Co-68 1.14E+42 +/- 1.4E+00 1 N1 88-1181 Ni Co-58 2.17E+03 +/~. 1.3E+61 l Fe 88-1182 Fe Mn-54 9.29E+42 +/- 1.9E+41 j Fe 88-1183 Fe Mn-54 4.88E+42 +/- 1.4E+41 MID' Cu 88-1184 Cu Co-68 1.20E+42 +/- 1.3E+ee 1

Fe 88-1185 Fe Mn-54 9.44E+e2 +/- 1.6E+41

) 30TTOM MID Cu 88-1186 Cu Co-64 1.18E+42 +/- 1.3E+98 i Ni 68-1187 Ni Co-58 2.20E+e3 +/- 1.3E+01

Fe 88-1388 Fe Mn-54 4.85E+42 +/- 1.3E+01 l Fe 88-1189 Fe Mn-54 8.92E+et +/- 1.6E+el BOTTOM AICo 88-1191 AICo Co-64 1.44E+e4 +/- 1.7E+02 j AICo(Cd) 88-119e Al~a Co-68 4.99E+43 +/- 7.5E+01 Fe 88-1192 Fe Mn-54 9.19E+et +/- 1.2E+el Remarks

• Results are an units of dos /(mg of Dosameter Material).

The uranium caosule esas received cut open. No uranium present.

AL File: 13468 References Lab.Booke 29 page 248 Proceduress A-512 A-513 Analyst WRM, WTF, CAB Acorovede dd. /44*6- I'll*E

. . . . _ . __ - - -. . . . - ~ _._ __ . _ . . . _ ._. - _. _

E t

TABLE 4.1-4  :

l a

IRRADIATION HISTORY OF SURVEILLANCE CAPSULES V, U, AND W '

4 1 1

- Thornet normal Thermal . h rmal D D D D sont h 181N-Br) - Skrath (IIN-Er) 8 toot h (IAf-B r ) Hont h teep-Br) 4 B/76 258 4/79 9 3/82 0 2/05 1495792 4/74 141574 t/79 9 4/82 6 3/45 , 1867714 1

7/74 199944 4/19 9 S/82 9 4/85 1519174 8/16 112417 - 7/79 9 6/02 6 $/85 1568343

'9/14 431683 8/79 650167 1/a2 975422 6/85 1848524 1./7. ... 7. 9/79 1 19442 ./.2 1 9,9 3 ,/.. 1 15 1 11/74 199449 19/79 784544 9/82 994746 8/85 1799641 4 12/76 - 410132 11/79 692354 10/s2 1633919 9/45 1827823 J

1/77 189115 12/79 0 11/s2 1864403 19/05 149154S 2/17 0 1/89 0 12/02 1814831 11/85 1864543 3

$ /77 704513 2/se 9 1/43 1734339 12/85 1951648

4/77 96S179 3/st i 2/83 1$98797 1/a4 1949567

, S/77 1864148 4/80 0 Stel 1939771 3/04 1543257

{ 4/77 1949724 $/89 0 4/63 1845670 3/46 1965574 i 1/77 1649449 6/06 5 S/43 1732947 4/46 1774390 1'

4/77 1116291 7/80 0 6/43 545214 5/04 825172 a 9/77 0 4/40 0 7/83 9 6/04 6 10/77 46194 9/80 0 8/83 0 7/86 0 l i- . .

um meno tem e 9m wm em an.6. i

um m.n t m .. not.9 1. m m 93.. 9,.4 16 9ut t i?. 157. 2. 12/.. 91665. il/.3 1895. 3 .1./.4 1 4 41.

2/18 1872227 1/41 1118511 12/83 . 1893541 ' .'11/86 1790031 Sm 19.nu 2m . .um 1m usant - um u5sn?. .;

4/7s tussel mi e 2/a4 1s114ae 1/e7 19ntee i S/78 9 4/81 916478 3/84 1263131 2/07 1485904 4/75 141140 S/81 1999194 4/04 1819222 3/87 1952434 7/78 1821224 6/81 1874754 5/64 1812753 4/47 1506172 6/78 6 7/81 .1988094 4/94 1533814 6/87 20911 9/78 0 4/81 1850578 7/04 1731014 6/87 1661777 10/14 0 9/81 1745793 8/44 1949946 1/07 1984814 11/78 6 19/01 1651329 9/84 1674344 S/87 1841509 12/18 .594540 11/04 1702396 10/44 658494 t/a? 1752315 1/19 707964 12/61 1148932 11/64 0 10/87 1945242 2/79 1433640 1/82 0 12/84 e 11/87 1762194 9

3/79 334146 2/81 0 1/85 1230684 12/87 469765 4-6

o TABLE 4.1-5

SUMMARY

OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS WITHDRAWN FROM INTERNAL SURVEILLANCE CAPSULES REACI' ION RATE (rns/ nucleus)

CAPSULE CAPSULE CAPSULE REACTION V U W Cu-63 (n ot) Co-60 5.72E-17 4.71E-17 4.18E-17 Fe-54 (n p) Mn-54 5.94E-15 4.49E-15 3.82E-15 Ni-58 (n.p) Co-58 8.12E-15 6.06E-15 5.08E-15 U-238 (n,f) Cs-137 Cd 2.97E-14 2.08E-14 Np-237 (n,f) Cs-137 Cd 2.38E-13 1.75E-14 1.33E-13 Co-59 (n,y) Co-60 4.1 lE-12 2.44E-12 2.10E-12 Co-59 (n,y) Co-60 Cd 1.76E-12 1.06E-12 9.13E-13 4-7

/

l

.. .-.e- - . , . ~ - . _-_a. 1 -- -s- - - _.--a a - - .- - - - - - a a.

TABLE 4.2-I f DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE V DOSIMETRY WITIIDRAWN AT Tile END OF FUEL CYCLE 1 -

TRIAL ADJUSTED 10 VALUE VALUE UNCERTAINTY

$(E > 1.0 MeV) 9.30E+ 10 8.63E+ 10 8%

$(E > 0.1 MeV) 2.97E+ 11 2.93E+11 15 %

$(E < 0.414 eV) 9.00E+10 1.01E+11 20%

dpa/sec 1.52E-10 1.45E-10 10 %

I 1

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES j SURVEILLANCE CAPSULE V REACTION RATE (rps/ nucleus)

TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu-63 (n,a) Co-60 5.72E-17 6.05E-17 5.66E-17 0.95 1.01 Fe-54 (n.p) Mn-54 5.94E-15 6.73E-15 6.00E-15 0.88 0.99  !

Ni-58 (n.p) Ni-58 8.12E-15 9.24E-15 8.20E-15 0.88 0.99 U-238 (n,f) Cs-137 Cd 2.97E-14 3.19E-14 2.92E-14 0.93 1.02 i Np-237 (n.f) Cs-137 Cd 2.38E-13 2.44E-13 2.33E-13 0.99 1.02 Co-59 (n,y) Co-60 4.1IE-12 3.95E- 12 4.10E-12 1.04 1.00 Co-59 (n,y) Co-60 Cd 1.76E-12 1.8EE-12 1.76E-12 0.94 1.00 4-8

TABLE 4.2-2 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE U DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 4 TRIAL ADJUSTED 1o VALUE VALUE UNCERTAINTY

$(E > 1.0 MeV) 6.09E+ 10 6.1 IE+ 10 8%

$(E > 0.1 MeV) 1.82E+11 1.99E+11 15 %

$(E < 0.414 eV) 5.09E+10 5.97E+10 20%

dpa/sec 9.69E-11 1.01E-10 10 %

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE U REACTION RATE (rps/ nucleus)

TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu-63 (n,a) Co-60 4.71E-17 4.59E-17 4.64E-17 1.03 1.02 Fe-54 (n.p) Mn-54 4.49E-15 4.78E-15 4.54E-15 0.94 0.99 Ni-58 (n.p) Co-58 6.06E-15 6.51E-15 6.14E-15 0.93- 0.99 U-238 (n.O Cs-137 Cd 2.08E-14 2.15E-14 2.10E-14 0.97 0.99 Np-237 (n.0 Cs-137 Cd 1.75E-13 1.52E-13 1.66E-13 1.15 1.05 Co-59 (n,y) Co-60 2.44E-12 2.25E-12 2.44E-12 1.08 1.00 Co-59 (n.7) Co-60 Cd 1.06E-12 1.08E-12 1.06E-12 0.98 1.(X) 4-9 I . .

e TABLE 4.2-3 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE W DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 6 TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY

$(E > 1.0 MeV) 5.66E+ 10 4.90E+ 10 9%

$(E > 0.1 MeV) 1.69E+11 1.57E+11 15 %

$(E < 0.414 eV) 4.73E+10 5.14E+10 20%

dpa/sec 9.00E-11 8.10E-11 10 % l l

I COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE W ,

REACTION RATE (rps/ nucleus) ,

TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu-63 (n a) Co-60 4.I 8E-17 4.26E-17 4.1IE-17 0.98 1.02 Fe 54 (n.p) Mn-54 3.82E-15 4.44E-15 3.85E-15 0.86 0.99 Ni-58 (n.p) Co-58 5.08E-15 6.05E-15 5.15E-15 0.84 0.99 Np-237 (n,f) Cs-137 Cd 1.33E-13 1.41E-13 1.29E-13 0.94 1.03 Co-59 (n,y) Co-60 2.1OE-12 2.09E-12 2.1OE-12 1.01 1.00 Co-59 (n,y) Co-60 Cd 9.13E- 13 1.00E-12 9.14E-13 0.91 1.00 4-10

SECTION 5.0 COMPARISON OF CALCULATIONS WITH MEASUREMENTS As described in Section 2.3, the best estimate neutron exposure projections for the Beaver Valley Unit I pressure vessel were based on a combination of plant specific neutmn transport calculations and plant specific measurements. Direct comparisons of the transport calculations with the Beaver Valley Unit I measurement data base were used to quantify the biases that may exist due to the transport methodology, reactor modeling, and/or reactor operating characteristics over the respective irradiation periods.

In this section, comparisons of the measurement results from surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels. In the first instance, predictions of fast neutmn exposure rates in terms of $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec are compared with the results of the FERRET least squares adjustment procedure; while, in the second case, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratory. It is shown that these two levels of comparison yield consistent and similar results, indicating that the least squares adjustment methodology is producing accurate exposure results and that the measurement / calculation (M/C) comparisons yield an accurate plant specific bias factor that can be applied to neutron transport calculations perfonned for the Beaver Valley Unit I reactor to produce "best estimate" exposure projections for the pressure vessel wall.

5.1 Comparison of Least Squares Adjustment Results with Calculation In Table 5.1-1, comparisons of measured and calculated exposure rates for the three surveillance capsule dosimetry sets withdrawn to date are given. In all cases, the calculated values were based on the fuel cycle specific exposure calculations averaged over the appropriate irradiation period.

An examination of Table 5.1-1 indicates that, considering all of the available core midplane data, the measured exposure rates were less than calculated values by factors of 0.932,1.003, and 0.965 for

$(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec, respectively, The standard deviations associated with each of the 3 sample data sets were 0.056 (6.0%),0.068 (6.8%), and 0.059 (6.1%), respectively.

5,2 Comparisons of Measured and Calculated Sensor Reaction Rates in Table 5.2-1, measurement / calculation (M/C) ratios for each fast neutron sensor reaction rate from the surveillance capsule irradiations are listed. This tabulation, provides a direct comparison, on an absolute basis, of calculation and measurement prior to the application of the least squares adjustment procedure as represented in the FERRET evaluations.

5-1

An examination of Table 5.2-1 shows consistent behavior for all reactions and all measurement points.

The overall average M/C ratio for the entire data set has an associated to standard deviation of 0.078 (8.2%). Funhennore, the average M/C bias of 0.947 observed in the reaction rate comparisons is in excellent agreement with the values of 0.932,1.003, and 0.965 observed in the exposure rate  !

comparisons shown in Table 5.1 1.

5-2 l

TABLE 5.1 1 COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE IRRADIATIONS 2

$(E > 1.0 MeV) [n/cm -sec]

INTERNAL CAPSULES CALCULATED MEASURED M/C V (15*) 9.30E+10 8.63E+ 10 0.928 U (25") 6.09E+10 6.I 1E+10 1.003 W (25") 5.66E+ 10 4.90E+10 0.866 AVERAGE M/C BIAS FACTOR (K) 0.932 STANDARD DEVIATION (10) 0.056 2

$(E > 0.1 MeV) [n/cm -sec]

INTERNAL CAPSULES CALCULATED MEASURED M/C V (15*) 2.97E+11 2.93E+11 0.987 U (25*) 1.82E+ 11 1.99E+11 1.093 W (25*) 1.69E+11 1.57E+ 11 0.929 AVERAGE M/C BIAS FACTOR (K) 1.003 STANDARD DEVIATION (10) 0.068 Iron Displacements [dpa-sec]

INTERNAL CAPSULES CALCULATED MEASURED pjlC V (15 ) 1.52E-10 1.45E-10 0.9 ~4 U (25*) 9.69E-11 1.01E-10 1.042 W (25*) 9.00E-11 8.IOE-11 0.900 AVERAGE M/C BIAS FACTOR (K) 0.965 STANDARD DEVIATION (10) 0.059 5-3

TABLE 5.2-1 COMPARISON OF MEASURED AND CALCULATED NEUTRON SENSOR REACTION RATES FROM SURVEILLANCE CAPSULE IRRADIATIONS M/C RATIO INTERNAL CAPSULES Cu63(n.cx) Fe54(n 0) NiS8(n.p) U238(n 0 NO237(n.0 V (15') 0.945 0.883 0.879 0.931 0.975 U (25") 1.026 0.939 0.931 0.967 1.151 W (25*) 0.981 0.860 0.840 0.943 AVERAGE 0.984 0.894 0.883 0.949 1.023 STD DEV (10) 0.033 0.033 0.037 0.018 0.091 OVERALL AVERAGE 0.947 STD DEV (10) 0.078 l

5-4

SECTION 6.0 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE VESSEL MATERIALS In this section the measurement msults provided in Section 4.0 are combined with the results of the neutron transport calculations described in Section 3.0 to establish the best estimate neutron exposure of the materials comprising the beltline region of the Beaver Valley Unit I reactor pressure vessel through the completion of Cycle 11. Based on the continued use of the Cycle 1I fuel loading pattem incorporating part length hafnium absorbers in several peripheral fuel assemblies, projections of future vessel exposure to 32 and 48 effective full power years of operation are also pmvided.

6.1 Exposure Distributions Within the Beltline Region As described in Section 2.3 of this report, the best estimate vessel exposum was determined fmm the following relationship:

• satz.n. = K @w, where: $wir . = The best estimate fast neutron exposure at the location of interest.

K = The plant specific measurement / calculation (M/C) bias factor derived fmm all available surveillance capsule and reactor cavity dosimetry data.

= The absolute calculated fast neutmn exposure at the location of interest.

$c.5.

From the data provided in Table 5.1-1, the plant specific bias factors (K) to be applied to the calculated exposure values given in Section 3.2 were as follows:

$(E > 1.0 MeV) 0.932 0.056 (6.0 %)

$(E > 0.1 MeV) 1.003 0.068 (6.8 %)

dpa 0.965 0.059 (6.1 % )

These bias factors were based on the results of the comparison of calculations with measured data from three intemal surveillance capsules irradiated through the first 10.8 effective full power years of operation.

6-1

t The micertainties listed with the individual bias factors are at the la level and are given on an absolute and percentage basis. Additional uncertainties associated with the evaluation of the best estimate vessel exposure are discussed in Section 6.2.

The tv ~t estimate results applicable to the vessel iruler surface are incorporated into Table 6.1-1 throup 6.1-3 for exposure parameters expressed in terms of $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa, respectively. Exposure distributions thmugh the vessel wall, can be developed using these surface exposures and radial distribution functions from Section 3.0.

At the end of Cycle 10, the Beaver Valley Unit I reactor had accrued 10.8 effective full powc years (EFPY) of operation. In order to establish a framework for the assessment of future vessel condition, exposure projections are also provided as a function of operating time through 27.1 EFPY in Tables 6.1-1 through 6.1-3. The end of life value of 27.1 EFPY was based on an assumed capacity factor of 0.80 between the end of Cycle i1 and the expiration of the plant operating license on 01/29/2016.

The extrapolations into the future were based on the assumption that the best estimate exposure levels characteristic of Cycle 11 would remain applicable throughout plant life. Examination of these projected exposure levels establishes the long term effectiveness of the low leakage fuel management incorporated in Cycle 11 and can be used as a guide in assessing strategies for future vessel exposure management.

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l

9 TABLE 6.1-1 BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) FOR MATERIALS COMPRISING THE BELTLINE REGION OF THE REACTOR VESSEL Operating Lower Lower Time Shell Shell Int. Shell Int. Shell (EFPY) Plates Long. Weld Cire. Weld Plates Long. Weld 10.8 1.38e+ 19 2.73e+18 1.38e+19 1.38e+19 2.73e+18 12 1.48e+19 3.0le+18 1.48e+19 1.48e+19 3. Ole + 18 14 1.66e+19 3.50e+ 18 1.66e+19 1.67e+19 3.50e+18 16 1.83e+ 19 3.99e+ 18 1.83e+19 1.86e+19 3.99e+18 18 2. Ole +19 4.48e+18 2.0le+19 2.05e+ 19 4.48e+18 20 2.19e+ 19 4.97e+18 2.19e+19 2.24e+19 4.97e+ 18 22 2.37e+19 5.46e+18 2.37e+19 2.43e+19 5.46e+18 24 2.54e+19 5.95e+ 18 2.54e+19 2.62e+19 5.95e+ 18 26 2.72e+ 19 6.45e+18 2.72e+19 2.81e+19 6.45e+18 27.1 2.82e+ 19 6.72e+ 18 2.82e+ 19 2.92e+19 6.72e+18 6-3

TABLE 6.1-2 BEST ESTIMATE FAST NEUTRON FLUENCE (E > 0.1 MeV) FOR MATERIALS COMPRISING THE BELTLINE REGION OF THE REACTOR VESSEL Operating Lower Lower Time Shell Shell Int. Shell Int. Shell (EFPY) Plates Long. Weld Cire. Weld Plates Long. Weld 10.8 4.21e+19 8.34c+18 4.21e+19 4.21e+19 8.34e+18 12 4.52e+ 19 9.20e+ 18 4.52e+ 19 4.54c+19 9.20e+18 14 5.06e+ 19 1.07e+19 5.06e+19 5.13e+19 1.07e+19 16 5.60e+19 1.22e+19 5.60e+19 5.71e+19 1.22e+19 18 6.14e+19 1.37e+19 6.14e+19 6.30e+19 1.37e+19 20 6.68e+19 1.52e+19 6.68e+19 6.89e+ 19 1.52e+19 22 7.22e+ 19 1.67e+19 7.22e+ 19 7.47e+19 1.67e+19 24 7.76e+19 1.82e+ 19 7.76e+19 8.06e+19 1.82e+19 26 8.30e+19 1.97e+19 8.30e+ 19 8.64e+19 1.97e+19 27.1 8.59e+ 19 2.05e+ 19 8.59e+19 8.97e+19 2.05e+ 19 6-4

TABLE 6.1-3 BEST ESTIMATE IRON ATOM DISPLACEMENTS FOR MATERIALS COMPRISING THE BELTLINE REGION OF THE REACTOR VESSEL

. Operating Lower Lower Time - Shell Shell Int. Shell Int. Shell (EFPY) Plates Lonc. Weld Cire. Weld Plates Lonc. Weld 10.8 2.43e-02 4.82e-03 2.43e-02 2.43e-02 4.82e-03 12 2.61e-02 5.32e-03 2.61e-02 2.63e-02 5.32e-03 14 2.92e-02 6.19e-03 2.92e-02 2.96e-02 6.I 8e-03 16 3.23e-02 7.05e-03 3.23e-02 3.30e-02 7.05e43 18 3.55e-02 7.92e-03 3.55e-02 3.64e-02 7.91e-03 20 3.86e-02 8.79e-03 3.86e-02 3.98e-02 8.78e-03 22 4.17e-02 9.66e-03 4.17e-02 4.32e-02 9.64e-03 24 4.48e-02 1.05e-02 4.48e-02 4.66e-02 1.05e-02 26 4.79e-02 1.14c-02 4.79e-02 5.00e-02 1.14c-02 27.1 4.97e-02 1.19e-02 4.97e-02 5.18e-02 1.18e-02 1

E 6-5

6.2 Uncenainties in Exposure Pmjections The overall uncertainty in the best estimate exposure projections within the pressure vessel wall stem primarily from two sources;

1) The uncertainty in the bias factor (K) derived from the plant specific measurement data base; and

2) The analytical uncertainty associated with relating the results at the measurement locations to the desired results within the pressure vessel wall.

Uncertainty in the bias factor derives directly from the individual uncertainties in the measurement process, in the least squares adjustment procedure, and in the location of the surveillance capsule sensor sets. The analytical uncertainty in the relationship between the exposure of the pressure vessel and the exposure at the measurement locations are based on capsule positioning uncenainties, downcomer water density variations and vessel inner radius tolerance relative to the surveillance capsule data.

The la uncertainties associated with the bias factors applicable to @(E > 1.0 MeV), @(E > 0.1 MeV),

and dpa are given in Section 6.1 of this report. The additional infonnation pertinent to the required analytical uncertainty for vessel locations has been obtained from benchmarking studies using the Westinghouse neutron transport methodology and from several comparisons of power reactor intemal surveillance capsule dosimetry and reactor cavity dosimetry for which the irradiation history of all sensors was the same.

Based on these benchmarking evaluations the additional uncertainty associated with the tolerances in dosimetry positioning, vessel inner radius and downcomer temperature was estimated to be approximately 6% for all exposure parameters. These uncertainty components were then combined as follows:

la UNCERTAINTY

@(E > 1.0 MeV) @(E > 0.1 MeV) h Bias Factor 6.0% 6.8% 6.1 %

Analytical 6.07c 6.0% 6.0%

Combined 12.0 % 12.8 % 12.17c 6-6 l

Thus, the total uncenainty associated with the neutron exposure projections at the pressure vessel clad / base metal interface for Beaver Valley Unit I was estimated to be:

la Uncenainty

$(E > 1.0 MeV) 12 %

$(E > 0.1 MeV) 13 %

dpa 12 %

These uncertainty values are well within the 20% 10 uncertainty in vessel fluence projections required

- by the FTS rule.

6-7 i-

4 6.3 Updated Lead Factors for Surveillance Capsules in Table 6.3-1 updated lead factors are listed for each of the Beaver Valley Unit I surveillance capsules. In Table 6.3-1 the individual capsule lead factors are provided as a function of operating time over the remaining licensed lifetime of the reactor. Lead factor data are based on the accumulated fluence experienced by the individual capsules and reactor vessel and reflect the relocation of Capsule T from a 35" location to a 25' location and Capsule Z from a 35' location to a 15* location at the conclusion of Cycle 10.

TABLE 6.31 UPDATED LEAD FACTORS FOR BEAVER VALLEY UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULES CUMULATIVE LEAD FACTOR Operating Time Capsule Capsule Capsule Capsule Capsule (EFPY) X Y T Z S 10.8 1.73 1.16 0.79 0.79 0.63 12 1.71 1.16 0.82 0.84 0.64 14 1,69 1.17 0.87 0.91 0.65 16 1.66 1.17 0.90 0.97 0.66 18 1.64 1.18 0.93 1.01 0.67 20 1.63 1.18 0.96 1.05 0.67 22 1.62 1.19 0.98 1.08 0.68 24 1.61 1.19 1.00 1.11 0.68 26 1.60 1.19 1.01 1.13 0.69 27.1 1.59 1.19 1.02 ' l.15 0.69 6-8

SECTION

7.0 REFERENCES

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9. ASTM Designation E264-92, " Standard Method for Determining Fast Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1993.

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10. ASTM Designation E48186 (Reapproved 1991)," Standard Method for Measuring Neutron Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1993.

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22. Hoskins, K. C., et. al., "The Nuclear Design and Core Management of the Beaver Valley Unit i Power Plant - Cycle 4," WCAP-10330, May 1983. [ Proprietary Class 2]

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