ML20199L747

Fast Neutron Fluence Evaluations for Fort Calhoun Unit 1 Reactor Pressure Vessel

ML20199L747
Person / Time
Site:	Fort Calhoun
Issue date:	11/30/1995
From:	Shaun Anderson WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20199L711	List: LIC-98-0009, Forwards Application for Amend to License DPR-40,requesting Deletion of Section 3.E,license Term,Which Requires Monitoring of Long Term Load Factor & Reevaluation of End of License Fluence w/ENDF/B-VI Cross Sections,From License ML20199L722 ML20199L729 ML20199L747
References
SE-REA-95-003, SE-REA-95-3, NUDOCS 9802100021
Download: ML20199L747 (95)

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_ - - - - _ - _ _ _ _ - _ _ _

l-

-FAST NEUTRON FLUENCE EVALUATIONS FOR THE FORT CALHOUN UNIT i REACTOR PRESSURE _ VESSEL SE-REA-95-003, November 1995

, . b nDQD%

S. L. Anderson Radiation Engineering and Analysis Prepared by Westinghouse for the Omaha Public Power District Purchase-Order No.-S082884 Work performed under Shop Order No. MKUP-450 9902100021 990130 PDR ADOCK 05000285-P PDR,

C EXECUTIVE

SUMMARY

This report describes an evaluation of.the best estimate fast neutron exposure'of the Fort:Calhoun Unit 1 reactor pressure-vessel.- The overall exposure evaluation methodologyRis based on guidance'provided in Draft Regulatory Guide DG-1025,

" Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" and makes use of the latest ENDF/B-VI neutron transport and dosimetry cross-sections included in the BUGLE-93 library.

In' addition to the general description of the methodology, the qualification of the overall approach and the uncertainties associated with the use of the methodology-are also.provided based on:

1- Comparison of absolute calculations with measurements obtained at the PCA assembly at the Oak Ridge National Laboratory.

2- Comparison of absolute calculations and measurements for two pressurized water reactors similar in' design to i Fort Calhoun; including five sets of internal surveillance capsule measurements and eight cycles of ex-vessel reactor cavity measurements (four at each' reactor)..

3- Comparison of absolute calculations with-three sets of internal surveillance capsule-measurements-from the Fort Calhoun' reactor.

4- An analytical sensitivity; study of important: input parameters _ applicable _to the Fort Calhoun transport calculations.

The results of these fluence evaluations demonstrate that the best estimate fas neutron exposure of the pressure vessel can be determined with a lo uncertainty of 113% for 4(E > 1.0 MeV)=,.1 9%- 1 for @(E > 0.1 MeV), and 114% for dpa. These uncertainties are well within the 120% guideline specified in DG-1025.

Application of the exposure methodology to the Fort Calhoun reactor pressure vessel indicates that'at the conclusion of Cycle 4

Ag __

14 (13.6 Effective Full Power Years) the critical weld material 3-410 had accrued a maximum fast neutron fluence (E > 1.0 MeV) of 9.24e+18 n/cm 2 and had reached a corresponding RT,73 value of 233.9 "F based on the correlations provided in Regulatory Guide 1.99, Rev. 2.

Based on the use of low leakage fuel management as embodied in j the design of Fuel Cycles 15 and 16, projections of future fast neutron exposure indicate-that at the license expiration date

(~30.5 Effective Full Power Years), the critical weld material will have accrued a maximum fast neutron fluence (E > 1.0 Mev) of 1.53e+19 n/cm 2 and will have reached an RT,13 value of 265.8 *F.

P

TABLE OF CONTENTS

.E.ilL92 TABLE OF-CONTENTS i LIST OF TABLES ii

-LIST OF FIGURES v

1.0 INTRODUCTION

1-1

~. 0 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION 2-1 METHODOLOGIES ,

2.1 Neutron Transport Analysis Methods 2-1 2.2 Peutron Dosimetry Evaluation Methodology 2-8' 2.3 Determination of the Best Estimate Pressure 2-14 Vessel Exposure 3.0 METHODS QUALIFICATION AND UNCERTAINTY EVALUATIONS 3-1 3.1 Comparisons with the PCA Pressure Vessel 3-1 Simulator Benchmark-3.2 Comparisons with Power-Reactor Measurements 3-12 3.3 Analytical Sensitivity Studies 3-23 4.0- RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4-1 4.1 Reference Forward Transport Calculation 4-1 4.2 Fuel Cycle Specific Adjoint Calculations 4-9 5.0 . EVALUATIONS OF SUFVEILLANCE CAPSULE DOSIMETRY 5-1 5.1 Measured Reaction' Rates 5-1 5.2~Results of the Least Squares Adjustment 5-2 Procedure 6.0 BEST ESTIMATE NEUTRON EXPOSURE AND RTng PT>JECTIONS 6-1 FOR' FORT.CALHOUN PRESSURE. VESSEL MATERIALS.

L6.1 Comparison of Calculations with Measurements 6-1 s 6.2 Exposure Distributions'Within the Beltline 6-6 Region 6.3 Projected RTn, of Limiting Beltline Material 6-13 6.4 Uncertainties in Exposure Projections 6  !

7.0 REFMRENCES 7-1 i

d tv -

O LIST OF TABLES Table Tit 1e Egqg

- 3 .1 -l ' Summary of Measurement Locations within the 3-5 PCA 12/13 Configuration 3.1-2. Measured Sensor Reaction Rates in the PCA 12/13 3-9 Configuration

-3.1-3 Calculated Sensor Reaction Rates in the PCA 3-10

-12/13 Configuration 3.1-4 Ratio of' Measurement to Calculation (M/C) in 3-11 PCA 12/13 Configuration -

3.2-1 . Comparison of Measured and Calculated Exposure 3-15 Rates from Surveillance Capsule and Cavity Dosimetry Irradiations-Plant 1 3.2-2 . Comparison of Measured-and Calculated Neutron 3-18 Sensor Reaction l Rates from Surveillance Capsule t and Cavity Dosimetry Irradiations-Plant 1 3.2-1 Comparison of Measured and Calculated Exposure. 3-19 Rates from Surveillance Capsule and Cavity-Dosimetry Irradiations-Plant 2' 3.2-2 Comparison of Measured and Calculated Neutron 3-22 Sensor Reaction Rates from Surveillance Capsule and Cavity, Dosimetry Irradiations-Plant 2 4.1-1 Calculated Reference Neutron Energy Spectra at 4-3 Surveillance Capsule Locations 4.1-2 Reference Neutron Sensor Reaction Rates and 4-4 Exposure Parameters at the Center of Surveillance Capsules 4.1-3 Summary of Exposure Rates at the Pressure 4-5 Vessel Clad / Base Metal Interface ii-u

____j

LIST OF TABLES (Continued)

Table Title Pace 4.1-4 Relative Radial Distribution of $(E 2 1.0 MeV) 4-6 within the Pressure Vessel Wall 4.1-5 Relative Radial Distribution of $(E 2 0.1 MeV) 4-7 within the Pressure Vessel Wall 4.1-6 Relative Radial Distribution of dpa/sec within 4-8 the Pressure Vessel Wall 4.2-1 Calculated Fast Neutron Flux (E 2 1.0 MeV) at 4-10 the Surveillance Capsule Center ,

4.2-2 Calculated Fast Neutron Flux (E 2 1.0 MeV) at 4-11 Pressure Vessel Clad / Base Metal Interface 4.2-3 Calculated Fast Neutron Flux (E 2 0.1 MeV) at 4-12 the Surveillance Capsule Center 4.2-4 Calculated Fast Neutron Flux (E 2 0.1 MeV) at 4-13 Pressure Vessel Clad / Base Metal Interface 4.2-5 Calculated Iron Atom Displacement Rate at 4-14 the Surveillance Capsule Center 4.2-6 Calculated Iron Atom Displacement Rate at 4-15 Pressure Vessel Clad / Base Metal Interface 5.1-1 Summary of Reaction Rates Derived from Multiple 5-3 Foil Sensor Sets Withdrawn from Internal Surveillance Capsules 5.2-1 Derived Exposure Rates from Surveillance 5-4 Capsule W225 - Withdrawn at the End of Fuel Cycle 1 iii

LIST OF; TABLES (Continued)

-Table- Title 2A22 5.2-2 .. Derived Exposure Rates from Surveillance 5-5 Capsule W265 Withdrawn at-the End-of Fuel Cycle 7 5.2-3L Derived-Exposure-Rates from Surveillance 5-6

' Capsule W275 - Withdrawn at the End of-Fuel Cycle--14 6.1-1 Comparison of Measured and Calculated Exposure 6-4 Rates-from Surveillance Capsule Irradiations 6 .1 - 2 . Comparison of Measured and Calculated Neutron- 6-5 l Sensor Reaction Rates from Surveillance Capsule Irradiations I

6 '. 2 - 1 Neutron Exposure Projections at Key Locations 6-9 on the Pressure Vessel Clad / Base Metal Interface 6.3 Projected RTn3 Values for Weld Material-3-410- ~ 6-14 r Based on Best Estimate Fluence Projections I

iv l

l

LIST OF FIGURES Fioure Title Pace

-1.0-1 Schematic of the Fort Calhoun Reactor Vessel 1-4 Beltline 2,1-1 Fort Calhoun r,0 Reactor Geometry 2-6 2.1-2 Surveillance Capsule Geometry 2-7

-3,1-1 PCA 12/13 Configuration - X,Y Geometry 3-3 3.1-2 PCA 12/13 Configuration - Y,Z Gecmetry 3-4 ,

6,2-1 Neutron Exposure Projections at the Pressure 6-;2 Vessel Clad / Base Metal Interf?:e ,

i I

v

SECTION 1.0 INTRODUCTION a

In the assessment of-the state of embrittlement of light water reactor pressure vessels, an accurate evaluation of the neutron exposure of:the materials compricing the beltline region of the '

vessel dr.' required. This exposure evaluation must, in general, include' assessments not only at locations of_ maximum exposure at  !

the inner-diameter of the vessel, but, also, as a function of axial, azimuthal, and radial location throughout the vessel wall.

A schematic of the~ beltline region of the Fort Calhoun reactor pressure vessel is provided in Figure 1.0-1. In this case,-the

, beltline region is constructed of six (6) shell plates, six-(6) longi:udinal w9 ids, and one (1) circumferential weld. Each of these thirteen naterials must- be considered in the overall i embrittlement assessments of the beltline region.

I In order to saticfy the requirements of 10CFREO Appendix G-for the calculation of pressure /temperatute-limit curves for normal

, heatup and cooldawn of the retctor coolant system, fast neutron

} exposure levels must be defined at depths 1within the; vessel wall L equal to 25 and 75 percent of the wall thickness for each of the materials. comprising the beltline region. These locations are commonly referred to as'the 1/4T and 3/4T positions in the-vessel wall. The 1/4T exposure levels are also used in the determination of upper shelf fracture toughness as specified'in--

10CFR50 Appendix G. 'In the determination of values of RTns for comparison with the applicable pressurized thermal shock screening criterion, maximum neutron exposure levels experienced by each of the beltlina materials sie required. These maximum levels will, of course, occur at the vessel inner radius.

Furthermore,' in the event that a probabalistic fracture mechanics evaluation of the pressure vessel is performed or if an evaluation of thermal annealing and subsequent material re-embrittlement is undertaken, a complete embrittlement profile is required for the entire volume of the pressure vessel beltlane. i The determination of this embrittlement profile, in turn, 1-1 i

l

11 necessitates the. evaluation of neutron exposure gradients throughout the entire beltline.

The purpose of this report is to describe the approach used to

' determine the best estimate fast neutron exposure to the Fort Calhoun reactor pressure vessel; and to establish the uncertainties associated with those projections. The overall methocology derives from the guidance provided in ASTM Standard E853 " Analysis and Interpretation of Light Water Reactor Surveillance Results"I" and Draft Regulatory Guide DG-1025

" Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.rai; and, is dependent on a blend of plant ,

specific neutron transport calculations and available measured data to produce an accurate assessment of the pressure vessel exposure while minimizing the uncertainty associated with that assessment.

The ovcrall exposure evaluation methodology is based on the underlying philosophy that, in order to minimize the uncertainties in vessel exposure projections, plant specific neutron transport calculations must be suppcrted by;

1) benchmarking of the analytical approach, 2) comparison with power reactor surveillance capsule and reactor cavity industry wide data bases, and, ultinately, 3) by comparison with plant specific measurements. That is, as the 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 enalytical 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 methodologies used to perform calculations of the neutron environment within the Fort Calhoun reactor geometry are desct!. bed. The methods utilized in the evaluation of neutron dosimetry from either surveillance capsule or reactor cavity irradiations are also described; and, the procedures used to combine measurements with calculations to prcduce the final best estimate exposure of the reactor vassel are; discussed.

1-2

l 1

l l In addition to the general description of the methodology, the qualification of tho overall approach and the uncertainties associated with the use of the methodology are also prr.ided.

The methods qualification and uncertainty assessments are based ons 1- Comparison of absolute calculations with measurements obtained at the PCA assembly at the Oak Ridge National Laboratory.

2- Comparison of absolute calculations and measurements for two pressurized water power reacto s similar in design to Port Calhoun; including two sets of internal surveillance capsule measurements and four. cycles of reactor cavity measurements for Plant 1 and three sets of internal surveillance capsule measurements an' four cycles of reactor cavity measurements for Plant .

3- Comparison of absolute calculations with three sets of i internal surveillance capsule measurements fron. the

' ort Calhoun reactor.

4- ..n analytical sensitivity study of important input parameters applicable to the Fort Calhoun transport calculations.

Finally, this report provides an eva3uation of the current best estimate neutron exposure of the Fort Calhoun pressure vessel in terms of fast neutron fluence, $(E 2 1.0), fact neutron fluence,

$(E 2 0.1 MeV), and iron atom displacemencs, dpa, in addition,

[

based on the continued use of the Cycles 15 and 16 fuel loading patterns, projections of the future exposure of the vessel are provided. Also, uncertainties associated with tha current and projected exposure of che pressure vessel are discussed.

t 1-3 a -

i FIGURE 10-1 1

SCHEMATIC OF THE FORT CALHOUN REACTOR VESSEL BELTLINE

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______ -____________,/___ --_____-__

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see ]O O ap 23e 360 A!!rtui RIISII 9

single computation in the conventional forward mode, was used l primarily to obtain rslative neutron e.tergy distributions i throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {$(E 2 1.0 MeV),

i

$(E 2 0.1 MeV), and dpa/sec) through the vessel wall. The I neutron spectral information was required for the interpretation '

of neutron dosimetry withdrawn from the surveillance capsules as l well as for the determination of exposure parameter ratios; i.e.,

l [dpa/sec)/[$(E A 1.0 MeV)), within the pressure vessel geometry.

The relative radial gradient information was required to permit i the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1 '4T,

, ;T, and 3/4T locations.

The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, $(E 2 1.0 MeV), at surveillence 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. These importance functions, when combined with operating cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation; and, established the means to perform similar predictions and dosimetry eveluations 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 burnup of individual fuel assemblies increased.

The absolute cycle specific data from the aujoint evaluations together with the relative neutron energy spectra and radial distribution information from the reference forward calculation provided the means to:

1- Evaluate neutron dosimetry from the surveillance capsule locations.

2-2 -

I SECTION 2.0 l NEUTRON ThANSPCAT AND DOSIMETRY EVALUATION METHODOLOGT.ES ,

i As noted in-Section 1.0 of this report, the best estimate exposure of the reactor pressure vessel was developed using a  :

I combination of absolute plant specific neutron transport calculations and plant specific measurements obtained from the reactor vessel materials surveillance program. In this section, the neutron transport and dosimetry evaluation methodologies are discussed in some detail; and the approach used to combine the  ;

calculations and measurements to produce the best estimate vessel  ;

exposure is presented.

' 2.1:- Neutron Transport Analysis Methods i A plan view of the Fort Calhoun Station Unit No. 1 reactor .

geometry at the core midplane is shown in Figure 2.1-1. Six surveillance capsules. attached to the pressure vessel wall which are removed on an individual basis at frequencies defined'in the Fort Calhoun Station Updated Safety Analysis Report Section 4.5

- are included in the reactor design to constitute the reactor vessel ~ surveillance program. The capsules are located at azimuthal angles of- 45', _85', 95', 225', 265', and 275' relative- to the core cardinal axis as shown in Figure 2.1-1. A plan view of a surveillance capsule-holder attached to the pressure vessel wall is shown in Figure 2.1-2.

From_a neutronic standpoint, the surveillance capsule. structures are significant. - The 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 thermal shield and the reactor vessel. In' order to determine the neutron I

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 Fort ,

. Calhoun surveillance capsules and reactor vessel, two distinct sets of-transport calculations were carried out. The first, a I

2e1

.5---e2- -M ., --,w,, -etty- [- ,.r v [ .- m c.-_-,-- -_----m,m won,w,,,,-- --.,,-,--p.m.,w~.r.--.-r.v- ....--,..ym-y,.-- +-.,wm, --.v52---,,,,-.-,,-e--------,w*--.--- ..e.-

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

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

4- Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.

l The forward transport calculation for the reactor model

! summarized in Figures 2.1-1 and 2.1-2 was carried out in r,0 geometry using the DORT two-dimensional discrete ordinates code H and the BUGLE-93 cross-section library"l. The BUGLE-93 library is a 47 energy group ERDF/B-V1 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 quadrature. The core power distribution utilized in the reference forward transport calculation was representative of the burnup weighted average over the first 14 cycles of operation.

All adjoint calculations 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 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 2 1.0 MeV).

Having the adjoint importance functions and appropriate core source distributions, the response of interest could be calculated as:

2-3

r r

R(r,0) = [ P

( I(r,0, E) S(r,0,5) r dr do dE J w t 0E Where: R(r,0) = $(E > 1.0 Mt/) at radius r and azimuthal angle 9. -

I(r,0,E)= Adjoint source importance function at radius r, azimuthal angle 9, and neutron source energy E.

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

Although the adjoint importance functions used in this analysis were based on a response function defined by the threshold neutron flux $(E 2 1.0 MeV), prior calculations 80 have shown that, while the implementation of low leakage loading patterns significantly impacts both the magnitude and. spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of (dpa/sec]/[$(E 2 1.0 MeV)) is insensitive to changing core source distributions. In the application of these adjoint importance functions to the Fort Calhoun-rescter. therefore, the iron atom displacement rates (dpa/sec) and the neutroc. flux

$(E 2 0.1 MeV) were computed on a cycle specific basis by using

[dpa/sec)/($(E 2 1.0 MeV)) and ($(E 1 0.1 MeV))/($(E 1 1.0 MeV))

ratios from the forward analysis in conjunction with the cycle specific $(E 1 1.0 MeV) solutions from the individual adjoint evaluations.

l In particular, after defining the following exposure rate ratios, (dpa/sec) g1 ,

$ (E 2.1. 0 MeV) 2-4

l l

l

$(E l 0.1 MeM 3"

$ (E 1 1. 0 MeM the corresponding fuel cycle specific exposure rates at the

, adjoint source locations were computed from the following relations:

dpa/sec = [$ (E 1 1. 0 MeV) ] R 3 l

1

$ (E l 0.1 NeV) = ($ (E l 1. 0 MeV)] R A

The reactor core power distributions used in t: 9 plant specific adjoint calculations were supplied by OPPD for the first 14 operating cycles of Fort Calhoun; and, for the predicted design distributions for Cycles 15 and 16.

2-5

FIGURE 2.1 .

FORT CALHOuti r,0 REACTOR GEOMETRY

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i 2.2 - Neutron Dosimetry Evaluation Methodology The use of passive neutron sensors such as those included in the Fort Calhoun surveillance program does not yield a direct measure

• of the energy dependent neutron flux level at the measurement location. Rather, the activation or fission process'is a measure of the integrated effect that the time- and energy-dependent neutrcn flux has on the target naterial over the course of the irradiation period. An accurate assessment of the average flux level and, hence, 'ime integrated exposure (fluence) experienced ty 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 spectrum at the sensor location In this section the procedures used to determine eensor specific '

activities, to oevelop 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.

L 2.2.1 - Determination of Sensor Reaction Rates

-The measured specific activity of each of the radiometric sensors contained in the three surveillance capsules withdrawn to date from the Fort Calhoun reactor was reported in Reference 6.

The irradiation history of the reactor over its operating lifetime was obtained from NUREG-0020, " Licensed operating Reactors Status Summary Report"DI. In particular, operating di a were extracted from that report on a monthly bases from reactor  ;

l startup to.the end of the current evaluation period. For the sensor sets utilized in surveillance capsule irradiations, the half-lives of the product isotopes'are long enough that a monthly 2-8

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

I I

histogram describing reactor operation has proven to be an ahequate representation for use in radioactive decay corrections -l for the reactionsLof interest in the exposure evaluations"1 llaving the measured specific activities, the operating history of

-l the reactor, and the physical characteristics of the sensors, j reaction rates referenced to full power operation were determined from the following equations i

i t

R. '

NFy A a C, \ 1 - e ** *1) e '"'

where:

• A = measured specific activity (dp.,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 naterial.

Y = number.of product-atoms produced per reaction.

P3 = average core' power 'evel.during irradiation period j (MW).

P,,, e- maximum.or_ reference core power level _of the reactor (MW).

C3 = calculated ratio of $(E 2 1 1.0 MeV).during frradiation period j to the time weighted average

$(E A 1.0 MeV) over the entire irradiation' period.

L = decay constant of the product isotope (sec) . l

.t-3

= length of irradiation period j (sec). -

t, = decay time following irradiation period-j (sec).

i and-the summation is carried out over the_ total number of monthly--

intervals comprising the total irradiation period.

1 In the above equation, the ratio P /P,,, 3 accounts for month by l month variation of_ power level within a given fuel cycle. The.

ratio C 3 _ is calculated for each fuel cycle using the adjoint l- 2-9 ,

P

.,,-.-.,,.--,-...--.a,.-,- - - . , . , . _ . . , .r-..-  :-...-~ . _ , , . , . . . - , - _ . - . ,- - , - . , - .

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

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 fuol cycle to fuel cycle. For a single cycle irradiation C = 1.0. However, for 3

multiple cycle irradiations, particularly those employing low leakage fuel management the additional C c;orrection must be 3

utilized.

2.2.2 - Corrections to Reaction Rate Datti Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 3.2.2 of this report, additional corrections were made to the U-238 measur,ements 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 additien to the corrections made for the presence of U-235 in the U-238 fission sensors, corrections were also made to the U-238 sensor reaction rates to account for ganna ray induced fission reactions occuring over the course of the irradiation. These photo-fission corrections were, likewise, location dependent and were based on the reference transport calculations described in Section 2.1.

2.2.3 - Least Squares Adjustment P ocedure Values of key fast. neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment codel ". The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust the group fluxes from the trial spec.t*;um to produce a best fit (in a least squares sense) to the mea'sured reaction rate data. The

" measured' exposure parameters along with the associ ated uncertainties were then obtaintsd f rom the adjusted spectrum.

In the FERRET evaluations, a log-nornal least squares algorithm  ;

weights both the trial values and the measured data in accordance l with the assigned uncertaintiss and correlations. In general,  !

1 2-10 l

l

i 9 the' measured values f are linearly related to the flux $ by some response matrix As fs"'"})Ala'$a

a where i indexes the measured _ values belonging to a single data l -set s, g designates the energy group, and a delineates spectra that may be simultaneously adjusted. For example, ni 1: a i, +,

relates a set of measured reaction rates Ri to a single spectrum

$, by the multigroup reaction cross-section ci,. 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 format consisting of 53 energy groups. The-trial input L

spectrum was converted to the FERRET 53 group structure using the SAND-II code"83 -This procedure-was carried out by first expanding the 47' group calculated spectrum into the SAND-II 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 ENDFIB-VI dosimetry file'"8, 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 function in the cross-section collapsing procedure.

-Reaction cross-section uncertainties in the form cf a 53 x 53 covariance matrix for each-sensor reaction were also constructed from the information contained on the ENDF/B-VI data files.

These natricese 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 inforr:.ation does not significantly impact the results of the adjustment.

l Due to the importance of providing a trial spectrum that exhibits i 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 cross-section data files, the >

covariance matrix _for the input trial spect rum was constructed from the following relation:

M,,o = Rn' + R, R,o P,,o  ;

where R, specifies an overall fractional normalization uncertainty (i.e. , - complete correlation) for-the set of values.

The f ractional uncertainties R, specify additional random +

uncertainties for group g that are correlated with a correlation matrix given by:

P,,, = (1 -0] 6,,o + 0 e-"

where:

1. = I9~7U 2y a The first term in the correlation matrix equation specifies purely-random uncertainties, while the second term describes short range correlations over a group range y (6 specifies the

-strength of the latter term). The value of 6 is 1 when g = g'

'and 0 otherwise. For the trial spectrum used in the current evaluations, a-short range correlation of y = 6 groups was used.-  !

g_33

l This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerker"U. Maerker's results are closely duplicated when y = 6. For the integral reaction rate l covariances, simple nornalization and random uncertainties were combined as deduced from experimental uncertainties.

In performing the least squr.res adjustment with the FERRET code, the input spectra from the reference forward transport calculations were normalized to the absolute calculations from the cycle specific adjoint analyses. The specific normalization factors for individual evaluations depended on the location of i the sensor set as well as on the neutron flux level.at that location.

The specific assignment of uncertainties in the measured reaction rates and the input (a priori) spectra used in the FERRET evaluations was as follows:

REACTION RATE UNCERTAINTY 5%

FLUX NORMALIZATION UNCERTAINTY 30%

FLUX GROUP UNCERTAINTIES (E > 0.0055 MeV) 30%

(0.68 ev s E s 0.0055 MeV) 58%

(E < 0.68 ev) 104%

SHORT RANGE CORRELATION-(E > 0.0055 MeV) 0.9 (0.68 ev 1 E s 0.0055 MeV) 0.5 (E < 0.68 ev) 0.5 FLUX GROUP CORRELATION RANGE (E > 0.0055 MeV) 6 (0.68 ev 1 E s 0.0055 MeV) 3 (E < 0.68 ev) 2 2-13

It should bc noted that the uncertainties 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 short range correlation and a range of 6, j

- and-a second part 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 i in the analysis of neutron dosimetry from surveillance capsule,  !

reactor cavity, and benchmark irradiations. The values are l liberal enough to permit adjustment of the input spectrum to fit ,

the measured data for all-practical applications.

2.3 - Determination of Best Estimate Pressure Vessel Exposure l 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 4

! methodology. discussed in Section 2.1 and plant specific l measurement data determined using the measurement-evaluation .

techniques described in Section 2.2. In particular, the best estimate vessel exposure is obtained from the following;  ;

relationship O seet set.

K

• calc.

where @g.,,, ,,,, = 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 and reactor cavity  !

dosimetry data.

$c. w , . = The absolute calculated fast neutron exposure at-the location of interest.

2-14 ,

o wr 7 w f* Wr

• w w w v- m - -w - wr 'r=,wyye.,.w n w-- -e-vs' -vr4 g- 1y, "y---- y **w +"

l

• l The approach defined in the above equation is based on the premise that the measurement data replasent the most accurate plant specific information available at the locations of the dosimetry; and, further that the use of the measurement data on a l plant specific basis essentially removes biases present in the l analytical approach and mitigates the uncertainties that would l result from the use of analysis alone. That is, at the l 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. assumed reactor dimensions, and operational variations in water density within the reactor. As a result, the overall uncertainty in the best estimate exposure projections within the vessel wall depend on the individual uncertainties in the i 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.

The uncertainties in the measured flux were derived directly from the results of the least squares evaluation of dosimetry data.

The positioning uncertainties were taken from the analytical sensitivity studies of sensor position performed as part of the overall benchmarking evaluation of the Best Estimate Fluence.

Methodology described in Section 3.0.

2-15

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

, o

! SECTION 3.0 j ME'1 HODS QUALIFICATION AND UNCERTAINTY EVALUATIONS 4

I

As noted in Section'1.0, the qualification of the transport methodology used in the analysis of the fast neutron exposure of i

! the Fort Calhoun pressure. vessel consisted of the following three parts:

1- Comparisons with benchmark measurements from the PCA  !

simulator at ORNL.

l 2- Comparisons with a series of power reactor, measurements ,

that-include data both from internal surveillance capsule dosimetry and reactor cavity dosimetry.  ;

3- An analytic sensitivity study investigating the 3

dominant sources of uncertainty in the transport model. -

t The results of-these studies, when combined with the Fort-Calhoun measurement data base discussed in Section 5.0'of this-report, ,

define the biases and uncertainties required to provide projections of the best estinate neutron exposure of the Fort Calhoun pressure vessel.  ;

9

^ 3.1 Comparisons with the PCA Pressure Vessel Simulator Benchmark The pressure vessel simulator benchmark comparisons used in the qualification of the neutron transport methodologyjare based-on (

- the analysis of the PCA 12/13 experimental configuration (see i Re:erences 13, 14, and 15). A schematic description of this  ;

. configuration is provided in Figures 3.1-1 and 3.1-2. A plan

- view of the PCA reactor and pressure vessel simulator showing materials characteristic of the core axial midplane is shown in Figure 3.1-1; whereas, a sectivn view through the center of the mockup is shown in Figure 3.1-2. The description of the 12/13 l

- configuration was-derived from information provided in References-  !

-through115 and reflects the-latest available geometric data

. forfthe-simulator.

31

+

err v w-w +

e--r+ww mm7,--c I---+w n **--w--r e-n-r <~- ,-e--ws- wrw *-er m- we+=----wa-s w-w- w - - - * + v -- i-+ww.m,e-wu-n--w--w----yw erg.---r wwr,

l 4 The 12/13 configuration was chosen for the methods evaluation due  !

to the similarity of this particular mockup to the thermal shield 4

- downcomer - pressure vessel designs that are typical of most '

pressurized water reactors. Of particular note in regard to the areas of similarity are the 12 cm. water gap on the core side of i the thermal shield, the 13 cm, water gap between the thermal shield and the pressure vessel simulator, the 6 cm. thick thermal +

shield, the 22.5 cm thick low alloy steel pressure vessel, and the simulated reactor cavity (void box) positioned behind the  ;

pressure vessel mockup. I Frcm the viewpoint of fast neutron attenuation, the 12/13 experimental configuration results in a reduction factor for  !

$(E z 1.0 MeV) of approximately 10' between the reactor core and I the inner surface of the pressure vessels and a corresponding reduction factor of about 30 from the inner surface to the outer surface of the pressure vessel wall, These similarities in the geometry and attenuation properties of the PCA mockup and LWR  ;

plant configurations provide additions) confidence that  :

judgements made regarding measurement / calculation comparisons in the simulator environment can be related to the subsequent analyses performed for operating Light Water Reactors. ,

During the PCA experiments, measurements were obtained at several l locations within the mockup to-provide traverse data extending i from the reactor core outward through the pressure vessel simulator and on into the void box. The specific measu*:ement  !

locations are illustrated-on' Figure 3,1-1 and listed in Table 3.1-1. From Figure 3.1-1 it is noted that all of the p measurements were obtained on the lateral centerline of the mockup. _ Furthermore, all of-the measurement points were also positioned on the axial midplane of the simulator.  ;

t 3-2

. . . ,,r.- -,.v.-- t',%..,.e,- 4,,.v- p.,,-- .r-e--, --:m e .,,--------ew,me . v -

.,,-w,-w- w+,w---r--n- . - , .,.y,.., % m- e, ,.rmq.y,---w3--wr+-.1-e,- w- , ...e-,r.,,,wrree-t ir y rv etw+ +w-vn- w w-- -w- d

FIGURE 3.1-1 PCA 12/13 CONFIGURATION - X,Y GEOMETRY canz pacz amautAina m age eht.94 WIMM

. u .

m msu m - m-) , A . c. . g .

y g s -

fl.s N. ,N '

~

~

f f y 00000 n

V O@o @ o ~

g4 ,

. y <, >,W .

s a u,,

s como .

Ir - --

p  % O@O0 0 s

x\ s e

j ,#

g

,3 g

",C OOOO i

a da n-

• M

~

48 3 '

-  : 4,60 M

- 0*e  ; O z . '

_ 99 9 1

~1 as.t T

_~ i so.s T j m.S '~

1 S .4. , ,,

30 5 ;1; 28.5  ;

It.6 et  ; 18.5 ~g 3-3 4

FIGURE 3.1.-2 PCA 12/13 CONFIGURATION - Y,Z GEOMETRY l

l l

A4 A8,81 A7 A6A5 A3M A2 A1 * * * " "

pr

.-i-,

-, -, .a 1

p

.). - .

,i --

g

,! u i

a , . .

I I d

I -

1 , 1

%'i g v N

2 l ' '

g 4

' les A ]:h 1%

I b / Y, e <

4 4 'i '

i ,

t! d / 9

- s / x

! l A

a w

t s w.menou a mo  ;

m , c

,7_ i g8 L i

r - .

w,,

9 , ,

' , , ( .

2

/

f. I '

,e '; l s

,, , B \ ' ,, e '-

x$J,

1. 2 --

n 4

7< g s

j i s -

s o .

. .., ;4, .. o_ . ., g .

, A s

4 4., , s o e a v y .

3 L- ,b . b m. ic o in N

m u __ y.. m e - w --1 ^ra_M AEE VCHQ.. 801 l

l l

i i

l 3-4 i

l TABLE 3.1-1

SUMMARY

OF MEASUREMENT LOCATIONS WITHIN THE PCA 12/13 CONFIGURATION LOCATION _lQ_ Y(em)

CORE CENTER- A0- -20.75 THERMAL SHIELD FRONT Al 11.98 THERMAL SHIELD BACK A2 22.80

! PRESSURE VESSEL FRONT A3 29.71 l PRESSURE VESSEL 1/4T A4 39.51 PRESSURE VESSEL 1/2T A5 44.67.

PRESSURE VESSEL 3/4T A6 50.13 L -VOID BOX A7 , 59.13 l

L Note: Y dimensions are referenced to the core side face of the aluminum window (see Figure 3.1-1).

The mcasurement location specified in Table 3.1-1 provide. data sufficient to generate measurement / calculation comparisons

-throughout the entire 12/13 configuration. Data-from locations A4, AS, and A6 establish the means for verification of calculated exposure gradients within the pressure vessel wall itself. Since measurements at operating power reactors can, at best, provide data in.the downcomer region internal to the vessel wall or in the cavity external to the vessel wall, these PCA data points located interior to the thick walled vessel establish a key set of comparisons to aid in the accurate determination of exposure gradients within the pressureLvessel wall.

3.1.1 Method of Analysis-The neutron transport analysis of the PCA 12/13 configuration'was carried-out using two DORT two-dimensional discrete ordinates transport calculations, one in X,Y geometry and one in Y,Z geometry, as well as a single one-dimensional DORT calculation in planar l(Y) geometry to synthesize a three-dimensional solution throughout the PCA~ simulator. The synthsis was carried out using the folicwing relationship:

325

l 1

1 l

4,(y, :)

4,(x,y,z) = 4,(x,y)

• l where: $,lx,y,z) = The group-g neutron flux at position x,y,z within the problem geometry.

$, (x , y ) = The group-g neutron flux solution from the x,y DORT computation.

$, br, z ) = The group-g neutron flux solution from the y,z DORT computation.

$, br) = The group-g neutron flux solution from the y DORT computation.

In this synthesis approach the ratio ($,(y, z)) / ($,br)) represents an energy dependent axial shape factor that accounts for the finite height of the PCA core as well as for axial leakage effects introduced by the simulator geometry.

l In the calculation of the PCA 12/13 configuration, all of the DORT computations were carried out in 67 energy groups (47 neutron, 20 gamma-ray) using a P3 cross-section expansion from l

the BUGLE 93 library and an S order of angular quadrature. The geometric models used in the calculations consisted of 71 X 131, 131 X 71, and 131 mesh cell arrays for the x,y, y,z, and y problems, respectively. Material descriptions for each of the regions comprising the simulator geometry were taken as specified in References 13 through 15. Likewise, the spatial distribution of the neutron source within the PCA core was obtained directly from References 13 through 15. In generating the energy dependent source for use in the transport calculations, the specified spatial distribution was coupled with the ENDF/B-VI U-235 fission spectrum supplied with the BUGLE 93 library.

Dosimeter reaction rates for comparison with PCA measurements were derived from the synthesized three-dimensional neutron flux distribution using the ENDF/B-VI reaction cross-sections also supplied with the BUGLE 93 library.

3-6

3.1.2 Comparison of'PCA Calculations with Measurements Measured data from the PCA experiments using the 12/13 simulator configuration have been documented and discussed extensively in References 13, 14, and 15. In these documents, individual sensor measurements were provided in terms of either equivalent fission flux per source neutron or absolute reaction rates per source neutron for a variety of reactions =with responses spanning the fast neutron energy range. For-the comparisons presented in this report all equivalent fission fluxes were converted to absolute i reaction rates using' fission spectrum averaged-reaction cross-sections that.were also reported in the PCA documentation. In particular, the following reaction cross-sections were. employed to perform the required conversions:

REACTION o, (barns)

Al-27(n,a) 0.000705 Ni-58(n,p) 0.1085 In-115(n,n') 0.189 U-238(n,f) 0.308 Np-237(n,f) -1,334 L

'The' appropriate measured reaction etes used for comparison with analytical prediction are sunnari : l in. Table 3.1-2.

In regard to the reaction rates listed in Table 3.1-2 it is-important to note that, based on discussions contained in Reference 15, the U-238 and Np-237 data for locations within the pressure vessel wall (positions A4, AS, and A6) differ somewhat

-from the reaction rates given in References 13 and 14. In the earlier reports a 10% bias was noted between fission. chamber measurements and solid state track recorder (SSTR) data. As a result, recommended reaction rates were taken to be the average of the two data sets. Since publication of those earlier documents, the observed bias wre determined to be caused by perturbations in the neutron field caused by the presence of the fission chamber structure. Therefore, the SSTR measurements prov:ied a more-accurate representation of the U-238(n,f) and Np-23?(n,f) reaction rates within the pressure vessel wall. The data = listed in Table 3.1-2 incorporate only the SSTR results for

. 1 I

Fission rate data for all other positions A4, AS, and A6.

locations within the 12/13 configuration remain as reported in References 12 and 14.

In addition to the measured reaction rates for_each of the-individual neutron sensors, the documentation of the PCA experiments also provides recommended values for important energy

-dependent exposure parameters at each of the measurement y locations. LThese derived exposure parameters resulting from the (

. application of-least squares-adjustment proceduresuto fit an  :

appropriate trial neutron energy spectrum to each-set of measured  !

reaction rate data include $(E 2; 1.0 MeV), $ (E 2; 0.1 MeV) , and the iron atom displacement rate (dpa/sec). The recommended l values of exposure parameters applicable to the.12/1)

-configuration-are also listed in Table 3.1 2. The derived  ;

exposure, parameters for locations A4, A5, and A6 raflect the use of U-238.and Np-237 fission rates measured by means of the SSTR

-. technique. Thus, the influence of the-previously mentioned bias associated with fission chamber perturbations has also been removed.from these integral results.

The cG1culated reaction rates and exposure parameters applicable

-to the PCA 12/13~ configuration are listed in Table 3.1-3.

Comparisons of these analytical predictions with the measurements

- are'provided-in. Table 3.1-4.

E3 -8

TABLE 3.1-2 hEASURED SENSOR REACT 10N RATES IN THE PCA 12/13 CONFIGURATION REACI1ON RATE (rpsAusciens-sousce neumon)

Al-27(n.a) Ni-58(m.o) In-il5(n.n4 U-2380Lf) No-237(m.f) '

AI 5A8e-33 631c-31 1.05e-?J A2 7.16e-34 6.72e-32 1.14e-31 730e-31 A3 - 3.13e-34 2.50e-32 3.68e-32 5.91e-32 3.05e-31 A4 7.15e-35 5.69e-33 IIIe-32 1.79e-32 1.20e-31 A5 2.92e-35 2.25e-33 5.20e-33 7.88e-33 6.56e-32 A6- I.12e-35 7.99e-34 2.23e-33 3.26e-33 3A7e-32 A7 4.29i.-36 6.43e-34 8.65e-34 9.60e-33

- w' ME > I.0 MeV) ME > 0.I MeV) dialsec 4 Al 5.85e-28 "A2 4.0le-07 7A7e-07 A3 A4- '4.50e-08 135e-07 7 Ale-29 A5 2.21c-06 9.0le-08 4.20e-29 A6 9.73e-09 537e-08 2.22e-29 A7 Nose: Neutron flux values ase in uniss of st/ca'-sec-source meusson.

Iron " , ' - = rases ase in units of dpe/sec-sossce neumon.

6 TABLE 3.1-3 CALCULATED SENSOR REACTION RATES IN 11tE PCA 12/13 CONHGURATION w.

REACTION RATE (rps/ nucleus-souste acuami) l Al-27(n.a) Ni-58(n.o) In-115(n.n') U-238(n.f) No-237(n.O I At 5.30e-33 6.03e-31 9.77e-31 1.68e-30 8.19e-30 1 A2 6.86e-34 ' 6.47e-32 1.06e-31 I.80e-31 9.27e-31 t

' A3 3.10e-34 2.46e-32 3.60e-32 6.28e-32 2.99e-3I r

}. A4 6.86e-35 5.49e-33 1.07e-32 1.72e-32 1.16e-3I [

l A5 ' 2.75e-35 2.14e-33 4.82e-33 739e-33 631e-32  ;

A6 1.04e-35 . 7.89e-34 2.03e-33 2.95e-33 3.13e-32

A7 3.12e-36 1.96e-34 4.82c-34 6.95e-34 738c-33  !

i. y ME > I.0 MeV) ME > 0.1 MeV) thefsec j i.

w AI 3.76e-06 6.76e-06 5.66e-27  :

A2 4.18e-07 837e-07 637c-28 [

A3 138e-07 2.45e-07 2.12e-28 [

A4- - 4.54e-08 1.40e-07 7.48e-29  !

. A. 2.13e-08 9.13e-08 4.11e-29 I A6 9.15e-09 5.17c-08 2.07c-29 i l

A7 2.16e-09 1.15e-05 4.69e-30 l

1 I [

l Note: Neutsosi flux values ase mi usnes of afcus*-sec-source neutson.  !

j Ison e rases are in units of dpshec-sousce neutson. [

^

f I

l

, +

l f i

e

,a , ,..,m. .J.- r -m. ., L - - -

I

'I .

TABLE 3.! d RATIO OF MEASUREMENT TO CALCULATION (M/C) IN THE PCA 12/13 COST 1GURATION Al-27(n.u) Ni-58(n.p) In-115(n.n') U-238(rJ) No-237(nJ)

Ai 1.03 1.04 1.08 A2 1.04 1.04 1.08 0.79 A3 1.01 1.02 1.02 034 1.02 A4 1.04 1.04 1.04 1.04 1.04 A5- 1.06 1.05 ~ 1.08 1.07 1.04 A6 1.08 1.01 1.10 1.11 1.11 A7 1.38 1.33 1.24 1.30 I

ME > 1.0 MeV) ME > 0.1 MeV) Msec y A1

- A2 0.96 0.89 0.92 A3 A4 0.99 - 0.96 0.99 A5 1.04 0.99 1.02 A6 1.06 1.04 1.07 A7 2

3.2 - Comparisons with Power Reactor Measurements In this uection, comparisons of the measurement results from internal surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented for two Westinghouse Light Water Reactors (similar

- in design to Fort Calhoun) that implemented an ex-vessel dosimetry system prior to initial plant startup. As such, the internhl and external dosimetry sets have experienced the same reactor operating conditions for extended irradiation periods.

The current data base applicable to the first reactor (Plant 1) consists of two sets of' internal surveillance capsule dosimetry and four sets of ex-vessel reactor cavity dosimetry. Each set of reactor cavity measurements includes data from four azimuthal angles on the core axial midplane, thus, providing a total of 16 cavity data points for comparison. The data base for the second reactor (Plant 2) consists of three sets of internal surveillance capsule dosimetry and four sets of ex-vessel reactor cavity dosimetry. As was the case with Plant 1, the Plant 2 data base also provides 16 csvity data points on the reactor core midplane.

These comparisons of calculation and measurement are provided on t two levels. In the first instance, predictions of fast neutron exposure rates in terms of $(E 2,1.0 MeV), $(E 1 0.1 MeV), and s dpa/sec are compared with the results of the FERRET least squares adjustment procedure; while, in the second case, calculationsLof individual sensor. reaction rates are compared directly with the measured data from the counting laboratories. 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 indication of the b!.as that exists between measurement and calculation. ,

3.2.1 - Comparison of Least Squares Adjustment Resultr .iith Calculation i

.: 3bles 3.2-1 and 3.2-3, comparisons of measured and calculated' expos.re rates for the internal surveillance capsule dosimetry s

3-12

(

l

V

. i sets as well as for fode cycles of reactor cavity midplane dosimetry sets are giveh for Plants 1 and 2, respectively. In l all cases, the calcula(ed. values were based en the methodology c _ described in Section 2,,0 using fuel cycle spacific exposure

>> calculations averaged over the app.opriate irradiation period.

i An c2 amination of Table 3.2-1 and 3.2-3 indicates that, considering all of the available core midplsne data, the measured exposure rates for Plant 1 were less than calculated values by factors of 0.93, 0,99, and 0.94 for $(E 1 1.0 MeV), $(E 2 0.1 MeV), and dpa/seci .respectively. The standard deviations associated with each of the 18 sample data sets were 8.4%, 10.2%,

and 9.4%, respectively. In the case of-Plant 2, the measured exposure rates were also less than the measured values by factors of 0.90, 0.93, and 0.92 for $(E 2 1.0 MeV), $(E 1 0.1 MeV), and dpa/sec, respectively. The standard deviations associated with eachfof the 19 sample data sets were 8.1%, 10.4%, and 9.5%,

respectively.

i s

\ s N' <

3.2.2 - Comparisons of Mea.sured and Calculated Sensor Reaction s Raten' Tn Table 3.2<9 an'd 3.2-4, measurement / calculation (M/C) ratios j f or 'each fast neutron sensor reaction rate f rom the internal l ,j surveillance ca sule and external reactor cavity irradiations are fisted. These' tabulations, provides a direct comparison, on an i, ' absolute liais, of calculation and m<asurement prior to the app,licaticu-)f the least squares adjustment procedure as represented in the' FERRET evaluations.

An' examination of ' Tables 3.2-2 and 3.2-4 shows consistent behavior for 'all reactions and all measurement points. For Plant 1 thefstandard deviations observed for the six fast neutron 3 reactions range from 6.4% to 11.7% on an individual reaction .;

3 basis;' whereas, the overall average M/C ratio for the entire data ,

set Sas an ar.ociated 10 standard deviation of 11.3%.

Furthbrmore, the average M/C bias of 0.90 observed in the reaction rate comparisons is in excellent agreement with the values of 0.93, 0.99, and 0.94 observed in toe exposure rate compariQons shown in Table 3.2-1. '

3-13 i ,

x 3

l_ ..

u i ,

For Plant 2 the standard deviations observed for the six fast neutron reactions range from 4.3% to 18.4% on an individual reaction basis; whereas, the overall average M/C ratio for the entire data set has an associated 10 standard deviation of 10.2%.

Furthermore, the average M/C bias of 0.92 observed-in the reaction rate comparisons is in excellent agreement with the values of 0.90, 0.93, and 0.92 observed in the exposure rate comparisons shown in Table 3.2-1.

The-data comparisons given in Tables 3.2-1 through 3.2-4 indicate that, in tha power reactor case, the current transport methodolo c using the BUGLE-93 cross-section library tends to overpredict measurement at both in-vessel and ex-vessel measurement locations. Thus, providing a bias factor less than unity r,o be applied to calculated value; in order to produce best estimate results.

L 1

\

3-14

.. i

'N , 4 TABLE 3.2 ! COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE-AND CAVITY DOSIMETRY IRRADIATIONS-PLANT 1

$(E .>_ 1.'0 Mev) .[n/cm)2 CALCULATED MEASURED ht]Q, . ,

INTERNAL CAPSULE j Capsule-'l- -

7.37e+10 7'.49e+10 ' 12

' Capsule:2 6.11e+10 5.08e+10 0 83 0 DEGREE CAVITY Cycle A 3.14e+08 3.-16e+08 ' 1,-01 Cycle B- 2.79e+08 2.89e+08, 1.34 Cycle C- 2.50e+08 2.'27e+08 0.91 i

, Cycle D 2.44e+08 2.58e+08 1.06-11 DEGREE CAVITY Cycle A 4.61e+08 3.86e+08 0.84 Cycle B- 3.58e+08 -3.34e+08 0 93-Cycle C 3.34e+08 3.06e+08 0.92 -'

~ Cycle D 3.20e+08- 3.16e+08 '0.99 35 DEGREE CAVITY

, Cycle-A 7.17e+08 6.00e+08- 0.84-Cycle B 6.55e+08 6.64e+08 1.01 4- _

Cycle C 6.28e+08 5.36e+08 :0.85 ,

4.99e+08 4.49e+08 Cycle D 0.90_

45 DEGREE' CAVITY p Cycle A 8.07e+08 7.83e+08 0.97-

. Cycle B- 7.13e+08 6.69e+08 0.94

• Cycle C 6.89e+08 5.71e+08 0.83 L Cycle-D 5.25e+08 4.48e+0.8 -0.85 AVERAGE M/C BIAS FACTOR (K) 0.93 PERCENT STANDARD DEVIATION (10) 8.4%

3-15 l

B b 4 TABLE 3.2-1 (continued)

-COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM t

- SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS-PLANT 1

$(E 2,0.1 MeV) [n/cm) 2 CALCULATED MEASURED M/C INTERNAL CAPSULE-

-Capsule 1 2.42e+11 2.72e+11 1.12 Capsule 2 2.01e+11 1.79e+11 0.89 0 DEGREE CAVITY Cycle A 3.59e+09 4.17e+09 1.16 Cycle.B 3.19e+09 3.33e+09, 1.04 Cycle C 2.86e+09 2.69e+09 0.94 Cycle D 2.79e+09 3.05e+09 1.09 11 DEGREE CAVITY Cycle A- 5.34e+09 4.96e+09 0.93 Cycle B 4.17e+09 3.87e+09 0.93 Cycle C 3.89e+09 3.90e+09 1.00 Cycle-D- 3.73e+09 4.20e+09 1.13 35 DEGREE CAVITY Cycle 1. 8.64e+09 7.32e+09' O.85 I Cycle B 7.76e*09 8.63e+09 1.11 Cycle C- 7.44e+09 6.62e+09 0.89

, Cycle D 5.91e+09 5.58e+09 0.94 15_QE,qREE CAVITY Cycle A 9.08e+09 8.68e+09 0.96

' Cycle B -8.35e+09 8.41e+09 1.01 Cycle C 8.06e+09 6.87e+09 0.89 Cycle D 6.14e+09 5.59e+09 0.94 AVERAGE M/C BIAS FACTOR (K) 0.99 PERCENT STANDARD DEVIATION (10) 10.2%

i 3-16 l

'4 TABLE 3.2-1 (continued)

't COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM-SURVEILLANCE CAPSULE AND CAV1TY DOSIMETRY-IRRADIATIONS-PLANT 1 2 '

Iron Displacements- (dpa) i CALCULATED MEASURED M/C INTERNAL-CAPSULE Capsule 1 1.24e 1.'30e-10 1.05 Capsule 2 1.03e-10 8.71e-11 0.85-O DEGREE CAVITY Cycle A- 1.24e-12 1.35e-12, 1.09

^

Cycle B 1.10e-12 1.11e-12 1.01 Cycle C 9.88e-13 8.95e-13 0.91-Cycle D 9.64e-13 1.01e-12 1.05 l'1 DEGREE-CAVITY '

Cycle-A' 1.84e-12 1.62e-12 0.88 ,

Cycle B 1.44e-12 1.29e-12 0.90 Cycle-C 1.34e-12 1.28e-12 0.96-Cycle D ' 28e-12

. 1.36e 12 l '. 0 6_

35 DEGREE CAVITY

Cycle A 2.93e-12 2.41e-12 0.82

~

Cycle:B- 2.67e-12 2.79e-12 1.04

' Cycle _C 2.56e-12 2.18e-12 0.85i

_ Cycle D 2.04e 1.83e-12 0.90 45 DEGREE CAVITY Cycle A 3.12e-12 2.88e-12 - 0.92 I Cycle B 2.83e-12 2.73e-12 0.96

. Cycle C 2.73e-12 2.25e-12' O.82 Cycle D. 2.08e-12 1.82e-12 0.88

}- AVERAGE-M/C BIAS FACTOR (K) 0.94 PERCENT STANDARD DEVIATION (10) 9.4%

l 3-17

, u o f

TABLE.3.2-2

COMPARISON OF MEASUPED AND CALCULATED NEUTRON SENSOR REACTION RATES-

'FROM SURVE1LLANCE; CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS-PLANT 1 I

Cu63(n.a) Ti46(n.ol Fe54(n.0) NiS8(n.0) U238 (n. fi No237 (n f)

INTERNAL CAPSULE Capsule 1 1.09 0.88: 1.01 1.03 1.14 Capsule 2 0 .97 0.78 0.84 0.89-0 DEGREE CAVITY Cycle A- 0.92 0.89- 0.84 0.85 0.97 1.19l Cycle B' O.98 0.93 1.05 1.04 L

' . Cycle C 0.89: 0.93 0.85 0.83 0.92 0.93 Cycle D 0.94 1.02 0.88 0.93 1.14 1.09 11 DEGREE CAVITY Cycle A 0.78 0.82- 0.71 0.72 0.86 0.92 Cycle B 0.87 0.81 0.99 .0.89 Cycle C 0.85 'O.91 0.79 0.88 1.01 w 0.80 0.83 0.98 1.15

• Cycle D 'O.87 0.94 5 35 DEGREE CAVITY 0.86 0.78 0.77 0.87. 0.83 Cycle A 0 . 18 0 -

Cycle B 0.90 0.83 '1.02 1.10 Cycle-C 0.88' O.87. 0.77 0.90- 0.85-Cycle D O.91 -0.96 0.81 '0.86 0.94 0.91 45 DEGREE CAVITY Cyc1O AL 0.88 0.88" 0.81 0.88 1.06 0.92 Cycle B. 0.88- -0.76 0.96 0.99 Cycle C 0.78 0.83s 0.74 0.75 0.87 Cycle D 0.83 0.88 0.71 0.79- 0.90 0.88 AVERAGE 0.89 0.90 0.81. 0.84 0.96 0.99

% ST. DEV. (16) , 8.5 6 .' 4 7. :2 -9.5 8.2 11.7 0.90 OVERALL AVERAGE M/C RATIO PERCENT STANDARD DEVIATION (16) 11.3%

~^

TABLE 3.2-3 COMPARISON OF MEASURED-AND CALCULATED EXPOSURE RATES FROM

' SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS-PLANT 2

$(E > 1.0 MeV) (n/cm) 2 CALCULATED MEASURED HLQ SURVEILLANCE CAPSULES Capsule 1 0.49e+10 1.14e+11 1.20 Capsule-2 -.s5e+10 8.83e+10 1.03 Capsule 3 6.94a+10 5.93e+10 0.86 0 DEGREE CAVITY ,

Cycle A 5.33e+08 4.67e+08 0.88

-Cycle B 4.15e+08 3.82e+08 0.92 Cycle C 4.42e+08 3.57e+08 0.81 Cycle D 3.90e+08 3.37e+08 0.86 !

10 DEGREE CAVITY Cycle A 7.39e+08 6.00e+08 0.81 Cycle B 5.24e+08 4.18e+08 O.80 Cycle C- 5.62e,08 4.25e+08 0.76 Cycle D 4'.91e+08 3.91e+08 0.80

[

35 DEGREE CAVITY Cycle A' 7.15e+08 6.28e+08 0.88 Cycle B '6.44e+08 6.16e+08 0.96 Cycle'C 5.73e+08 4.75e+08 0.83 Cycle D 5.40e+08 5.17e+08 0.96 45 DEGREE CAVITY Cycle A 8.10e+08 7.94e+08 0.98 Cycle B 7.11e+08 6.55e+08 0.92 Cycle C 6.13e+08. 5.69e+08 0.93

' Cycle D 5.92e+08 5.56e+08 0.94 AVERAGE M/C BIAS FACTOR (K) 0.90 PERCENT STANDARD DEVIATION (10) 8.1%

3-19

s TABLE 3.2-3 (continued)

' COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS-PLANT 2

$(E > 0.1 MeV) (n/cm]

2 CALCULATED MEASURED - liLQ, SURVEILLANCE CAPSULES Capsule 1 4.25e+11 5.51e+11 1.30 Capsule 2 3.83e+11 4.32e+11 1.13 Capsule 3 3.05e+11 2.66e+11 0.87 0 DEGREE CAVITY Cycle A '4.85e+09 4.14e+09 0.86 Cycle B- 3.77e+09 3.47e+09 v.92 Cycle C 4.02e+09 3.20e+09 0.80 Cycle D 3.55e+09 3.14e+09 0.89-10 DEGREE CAVITY Cycle A 6.90e+09 5.35e+09 0.77 Cycle B 4.92e+09 3.88e+09- 0.79

, Cycle C 5.28e+09 3.88e+09 0.74 Cycle D 4.61e+09 3.76e+09 0.82 >

35 DEGREE CAVITY Cycle A -8.13e+09- 7.32e+09 0.90

- Cycle-B- 7.35e+09 7.56e+09 1.03 Cycle C 6.54e+09 5.66e+09 0.87 Cycle D 6.16e+09 -6.35e+09 1.03 45 DEGREE CAVITY Cycle A S.32e+09 8.11e+09 0.97 Cycle B 7.30e+09 6.82e+09 -0.93 Cycle C 6.30e+09 '6.44e+09 1.02 Cycle D 6.08e+09 5.80e+09 0.95 AVERAGE M/C BIAS FACTOR (K) 0.93 PERCENT STANDARD DEVIATION (10) 10.4%

3-20

TABLE 3.2-3 (continued)

COMPARISON OF MEASURED AND. CALCULATED EXPOSURE RATES FROM

-SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS-PLANT 2 Iron Displacements-[dpa)

CALCULATED: MEASURED jil.Q, ,

SURVEILLANCE CAPSULES

-. Capsule 1 1.84e 2.30e-10 1.26 Capsule-2 1.65e 1.80e-10 1.09.

Capsule 3 1.33e-10 1.16e-10 0.87 0 DEGREE CAVITY ,

Cycle A 1.73e-12 1.48e-12 0.86

! Cycle B 1.34e-12 1.23e-12 0.92-Cycle C 1.43e-12 1.15e-12 0.80 i Cycle D 1.26e-12 1.11e-12 0.88 10 DEGREE CAVITY Cycle A- 2.44e-12 1.91e-12 0.78 j Cycle B- 1.74e-12 1.38e-12 0.79.

j_ Cycle C 1.86e-12 1.39e-12 0.74 Cycle _D _1.63e-12 _1.33e-12. 'O.81 35 DEGREE CAVITY Cycle.A -2.74e-12 2.44e-12 0.89 Cycle B 2.47e-12 2.49e-12 1.01' Cycle C' 2.19e-12 1.88e 0.86 Cycle-D 2.07e-12 _2.09e-12 1.01 ,

l C EGREE CAVITY Cycle A 2,84e-12 2.75e-12 0.97

- Cycle B 2.49e 2.30e-12 0.93 Cycle C 2.15e-12 2.13e-12 0.99 Cycle D 2.07e-12 1.96e-12 0.95 AVERAGE M/C BIAS FACTOR (K) 0.92-PERCENT STANDARD DEVIATION (10) 9 .' 5 %

3-21

TABLE 3.2-4 COMPARISON OF MEASURED AND CALCULATED NEUTRON SENSOR REACTION RATES i

- FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS-PLANT 2 Cu63(n.a) Ti46(n.0) Fe54(n,0) NiS8(n.o) U238 (n, f) NO237 (n, f)

SURVEILLANCE CAPSULES Capsule 1 1.18 1.09 1.08 1.19 1.33 Capsule 2 1.10 0.94 0.93 1.09 1.12 Capsule 3 1.08 0.89 0.89 0.90 0.84 i

'O DEGREE CAVITY Cycle A 0.94 0.95 0.93 0.89 0.89 0.85 l

Cycle B 0.93 0.98 0.94 0.89 0.94 0.92 l

Cycle C 0.87 0.94 0.84 0.86 0.83 0.78 Cycle D 0.93 1.00 0.85 0.89 0.91 10 DEGREE CAVITY Cycle A 0.86 0.89 0.91 0.81 0.83 0.76 i Cycle B 0.89 0.94 0.88 0.84 0.79 0.78 Y Cycle C 0.89 0.94 0.83 0.84 0.78 0.71 y Cycle D 0.94 0.99 0.83 0.87 0.81 35 DEGREE CAVITY Cycle A 0.83 0.91 0.88 0.85 r 91 Cycle B 0.91 0.97 0.92 0.85 s 97 1.05 Cycle C 0.87 0.92 0.81 0.83 0.86 0.85 Cycle D -0.95 C.98 0.86 '0.90 1.01 1.02 45 DEGREE CAVITY ,

Cycle A 0.94 0.95 0.96 0.93 1.01 Cycle B 0.92 0.93 0.92 0.86 0.95 0.94 Cycle C 0.96 0.86 0.90 0.90 0.88 1.05 Cycle D 1.02 0.99 0.90 0.94 1.00 AVERAGE 0.95 0.95 0.90 0.89 0.92 0.93

% ST. DEV. (10) 9.4 4.3 7.1 6.5- 11.4 18.4 OVERALL AVERAGE M/C RATIO 0.92 PERCENT STANDARD DEVIATION (10) 10.2%

....v_.,z.._._u,.

- . . .:_..c.... w.....

.r

3.3 - Anclytical Sensitivity Studies The1overall uncertainty associated with calculated exp6sure-rates and integrated exposuresLcan be conveniently subdivided

~ 3 into'two broad categories.. The:first category involves biases i or errors that.may be present due to inadequacies in the method itself or in the -basic nuclear: data input to the calc.lation. These potential. biases.are addressed via 5

validation of the analytical techniqu'efthrough comparison with .

measurements from controlled benchmarkLexperiments, from power-reactor surveillance capsule and reactor cavity measurement data bases,-and,-ultimately, from plant specific surveillance capsule and cavity 1 irradiations. The use of these .

! analytical / measurement comparisons to effectively re. move i potential biases from analytical predictions results in best estimate projections of vessel exposure;1and is discussed'in Sections 2.3 and in Sections 6.2 and 6.4 of this report.

The second- category of uncertainty in the analysis of vessel Lexposure' involves variations that may exist-in reactor dimensions, coolant temperature, neutron-source strength and source' distribution,'as well as in other parameters that may-vary from; reactor to. reactor or fuel cycle to fuel' cycle.

This. category of'uncertaintytis most easily addressed via-sensitivityJstudies performed for each of the variables importe.nt to the overall evaluation.

For the: methodology used in the Fort Calhoun" neutron exposure

. evaluations, several sensitivity 3studits' were carried out to test theJeffect of variations in reactor. geometry and-neutron

. source-definition on the calculated ~ vessel exposure based on the'analytica'l approach outlined in Section 2.0. These studies'are not all inclusive, but do encompass the major contributorsLto uncertainties in the analytical approach.

.Important input-parameters addressed in these studies include theffollowing:

Geometry and Material Density

- stainless steel reactor internals water annuli

- reactor pressure vessel 3-23

-- - , - -. , - - - . , . , . n- .. = . - , - -

- core periphery modeling dosimetry positioning-(capsule / cavity)

Core Neutron: Source

- peripheral assembly source nagnitude

- peripheral-assembly burnup

-- >xial power distribution

- relative spatial distribution of the source As noted earlier in this section, the effects of transport cross-section errors and-uncertainties as well as biases

' introduced ly methods approximations were obtained by direct' ,

comparisons with-measured data rather than via a series of analytical studies. .

3.3.1 - Geomatric Modeling and Material Density With'the exception of the location of the surveillarce capsule

geometric center and the pressure vessel inner radius, the calculations performed for the Fort Calhoun reactor made use of nominal design dimensions for all internals components to establish the reactor geometry used in the transport model.

In'the case of the surveillance capsule and vessel = inner radius, as-built data were available and were incorporated into the model.- Likewise, nominal average full power coolant temperatures were.used to determine-water density in the. core and downcomer regions. ' Sensitivity of the calculated fast neutron exposure of the pressure vessel to each of these-variables was addressed via a seriestof parametric studies.-

To determini the potential impact of the reactor internals manuf acturing and asserably tolerances on the analytical' prediction of the fast neutron exposure of the pressere vessel, calculationa were performed for cases representing minimum shielding between the reactor core and the pressure vessel.(i.e., all components at minimum thickness) and for-

.maximumLshielding between the core and the pressure vessel (i.e., all components at maximum thickness). These extreme conditions were then compared to the nominal calculation to 3-24 1

establish an upper bound uncertainty in the use of nominal vs as-built internals dimensions. The resultant uncertainty in the calculated exposure of the pressure vessel is 13%.

The censitivity of the calculated vessel exposure to fluctuations in water temperature was likewise determined via a parametric study in which water temperature and, hence, coolant density was varied over a range of several degrees F relative to nominal conditions. The results of this stady indicate that a bounding uncertainty of 14% results from a temperature variation of 1 10 degrees F. A 110 degree fluctuation in water temperature would exceed variations expected during normal operation of the plant over a given fuel cycle. . Thus, the projected 4% uncertainty is considered to represent a conservative upper bound estimate.

The modeling of the rectilinear core baffle in r,0 geometry represents another potential source of uncertainty in the geometric modeling of the reactor. The sensitivity of the solution to the modeling approach was determined by a direct comparison of the results of an r,0 computation with those of an X,Y calculation in which the baffle region and core periphery were modeled explicitly. The comparisons of interest- were taken at various locations external to the core baffle. Results of these' calculations, in general agreed within the pointwise flux convergence criterion specified for the transport analyses, thus, demonstrating the adequacy of the modeling approach. Therefore, the bounding analytical uncertainty-associated witn this modeling approximation is taken to be less than 11%.

It should be noted that the X,Y vs r,0 comparisons described in the preceding paragraph, address not only the ade quacy of the geometric modeling of the core periphery, but, aAso demonstrate the adequacy of the transformation of the core neutron source from pin powers to the r,0 DOT model.

The inner radius of the reactor vessel itself and the position of surveillance capsule dosimetry are extremely important in the determination of the exposure of the pressure vessel wall both-from an analytical standpoint and from the viewpoint of 3-25

l surveillance capsule dosimetry interpretation. Therefore, sensitivity studies based on the as-built dimensions for both the vessel inner radius and capsule position were also performed.

Parametric evaluations of vessel inner radius indicate that s variations in vessel inner radius result in a change in calculated vessel fast neutron exposure of 15%. Uncertainties associated with the positioning of capsule dosimetry are extremely important in the evaluation of comparisons of calculation with measurement and the subsequent determination of plant specific bias factors. Parameter studies using the as-built position variations result in positioning uncertainties of 14% for curveillance capsules. ,

In developing the above uncertainties, the parametric studies assumed that, in the case of the surveillance dosimetry, displacement of the sensors either introduced or removed water from the area between the reactor core and the sensors.

3.3.2 - Core Neutron Source In addition to the sensitivity of the transport calculation to tolerances in the geometric model, several studies were also carried out to establish the sensitivity to the strength and spatial distribution of the neutron source within the reactor core. In particular, investigations were carried out to determine the sensitivity of calculated results to the absolute source strength in fuel-assemblies on the core periphery, the pin by pin spatial distribution of neutron source en the core periphery, the burnup of peripheral fuel assemblies, and the axial power distribution used in the flux synthesis procedure. It should be noted that the impacts of changing fission spectra, energy release per fission, and neutron yield per fission were encompassed in the parametric variation of fuel assembly burnup.

In regard to the absolute power level of peripheral fuel assemblies, the self-attenuation afforded by the core materials results in the neutron environment external to the core being dominated by these edge assemblies. An examination 3-26 l

l

of the adjoint transport evaluations performed for the Fort Calhoun reactor demonstrates that 90-95% of the external s environment results from neutrons born in these locations.

Therefore, the fluence uncertainty associated with the absolute core power level is directly dependent on the uncertainties in the power production of those peripheral assemblies. Based on comparisons of calculated vs measured (derived from in-core flux maps) peripheral power distributions for pressurized water reactors a bounding uncertainty for peripheral power magnitude has been determined to be 15%.

In a fashion similar to the peripheral assembly power, the uncertainty in the axial power distribution averaged.over the irradiation period, translater, directly to an uncertainty in the calculated neutron environment external to the core. Over the course of a 9.ven fuel cycle, the variation in the axial l peaking factor at maximum flux locations is typically 10%.

That is, the maximum axia.1 peaking factor may -

7e from a value of approximately 1.15 at beginning of c. to 1.05 at end of cycle, yielding a cycle average peaking factor of 1.10.

This observation was drawn from an examination of numerous axial distributions from a wide variety of pressurized water reactors employing both low leakage and non-low leakage fuel mansgement. In order to bound the uncertainty associated with this cycle average value, a variation of 15% is taken to be applicable. This uncertainty value is liberal-enough to encompass the entire change-in axial shape over the course of the fuel cycle.

Sensitivity studies involving source parameters such as fission spectrum, neutron yield-per fission and energy release per fission were performed via an evaluation of the sensitivity of the calculated fast neutron flux at the pressure vessel inner radius to the burnup of assemblies on the periphery of the reactor core. These burnup studies encompass significant perturbations in these source parameters due to the build-in of plutonium isotopes as the assembly burn"p increases.

For the studies in question, burnup was varied from an 3-27

\

assembly average of-3,000 MWD /MTU to 45,000 MWD /MTU. -The results of this' evaluation' indicated that the net change in vessel flux is approximately- 0.4%/1,000 MWD /MTV in:the burnup-range.of 3,000-15,000 MWD /MTU and 0.2%/1,.J0 MWD /MTU in the

-burnup' range ofil5.000-45,000 MWD /MTU. The total increase in calculated : flux at a burnup of 45- 000 ' MWD /MTU relative to that ,

-based'on a_burnup of 3,000 M.iD/MTU is about 10%.

~

-The values. quoted'in the preceding paragraph are typical of c light water reactors. Actual values will vary slightly depending on: reactor core configuration, core loadings,cand point of interest on-the vessel wall. However, these smallerL changes are of second order and, therefore, the_ data discussed

above provide an adequate evaluation of the sensitivity of the.

neutron flux at the pressure ~ vessel =and at dosimetry locations to'these particulcr parameters.

In.the assignment of an overall sensitivity to fuel assembly burqup aLliberal approach was utilized. It was first assumed l- that-the_ sensitivity to burnup effects was 0.4% per 1,000 MWD /MTU; i.e.,:the largest value obtained from the sensitivity

~

i study. It-was then further assumed that from the plant _

--specificicore designninformation,--a 5,000 MWD /MTU: uncertainty exists in the calculated fuel assembly burnup. This is-clearly a conservative evaluation, particularly at;1ow to intermediate levels of burnup. Combining these.two values-yields a bounding-sensitivity >to fission spectrum, neutron yield per fission,'and energyErelease per_ fission =of 12%.

Core managementcstudies on Westinghouse designed fuel cycles

' indicate that" uncertainties in the relative pin powers in -

peripheral 1 fuel assemblies-can be'on-the order of 8-10%. Due

-to-the use-of similar design methodologies, this uncertainty should-apply as:well to fuel designs of'other manufacturers.

Translating this coreidesign uncertainty to vessel exposure-results in'a 1 4%1 uncertainty in vessel exposure.

3-28

3.3;3 --Summary of Analytical Sensitivity Studies The results of analytically based sensitivity studies of geometric and source distribution input parameters may.be summarized as-follows:

VESSEL IR CAPSULE r,0 Modeling 1% 1%

Internals Dimensions '3 % 3%

Vessel Inner Radius 5%

Vesr'l Thickness Water Temperature 4% 4%

Peripheral Assembly Source Strength 5% 5%

Exial Power Dittribution 5% 5%

Peripheral Assembly Burnup 2%- .2 %

Spatial Distribution-of the Source 4% 4%

Capsule Dosimetry Positioning 4%

TOTAL 11% 10%

l When combined these individual sensitivities result in a net impact on the calculated flux levels in-the vicinity of the pressure vessel of 111%. The uncertainty evaluated at the dosimeter locations within internal surveillance capsules is 110%.

-These uncertainties due to potential variations in design and operating parameters'for individual reactors must, of course

.txa combined with-biases resulting from methods and cross-section errors to determine the total' uncertainty in the calculated results. This evaluation of the total uncertainty for the Fort Calhoun fluence evaluations is discussed in Section 6.0.

i 3-29 i.

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

o S'.CTION 4.0 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4.1 - Reference Forward Transport Calculation l As noted in Section 2.0 of.this-report,-a reference forward transport calculation based on a core' power ~ distribution representative of the burnup weighted average over the first 14 cycles of operation provided data for use in evaluating-neutron dosimetry from surveillance capsule evaluations as wellias in relating the results of these evaluations to the neutron exposure.

at locations interior to the pressure. vessel wall. ,In this section the key data extracted from this 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 the. measurements discussed in Section 5.0. -All absolute comparisons were based on the results of the fuel cycle specific adjoint calculations discussed in Section 4.2.

4.1.1 - Surveillance Capsule Locations Data'from the reference forward calculation pertinent to surveillance capsule evaluations are provided in Tables 4.1-1 and 4.1-2.

In Table 4.1-1, the calculated neutron energy spectra at the-geometric center of surveillance capsules located at the 225 oand 265'/275 azimuthal locations are listed. In Table 4.1-2, the calculated neutron sensor reaction rates and exposure rate ratios associated with the spectra from Table 4.1-1 are provided along with the calculated exposure rates in terms of $(E 2; 1.0 MeV), ~

$(E 2r0.1 MeV), and dpa/sec. Again, these data are applicsble to the geometric center of each surveillance capsule.

4 4-1

Thess reference reaction rates, expsure rates, and exposure rate ratios were used in conjunction with the results of fuel cycle specific adjoint transport calculatiens from Section 4.2 to provide calculated sensor reaction ra*:o and to project sensor set exposures in terms of $(E 2 0.1 MeV) and dpa/sec for each capsule irradiation period.

4.1.2 -

Pressure Vessel Wall Data from the reference forward calculation pertinent to the pressure vessel wall are provided in Tables 4.1-3 through 4.1-6.

In Table 4.1-3, the calculated azimuthal distribution of exposure rates in terms of $(E 2 1.0 MeV), $(E 2 0.1 MeV), and dpa/sec are listed at approximately 5 degree intervals over the reactor geometry. These data are applicable to the clad / base metal interface. Also given in Table 4.1-3 are the exposure rate ratios ($(E 2 0.1 MeV))/[$(E 2 1.0 MeV)) as well as (dpa/sec)/($(E 2 1.0 MeV)] that provide an indication of the variation in neutron energy spectrum as a function of azimuthal angle at the pressure vessel inner radius.

Radial gradier. information for $(E 2 1.0 MeV), $(E 2 0.1 MeV),

and dpa/sec is given in Tables 4.1-4, 4.1-5, and 4.1-6, respectively. These data are presented on a relative basis for each exposure parareter at the 0 , 15 , 30, and 45 azimuthal locations. Exposure rate distributions within the vessel wall were obtained by normalizing the calculated ($cuc.) or best estimate ( Q,,,, m ) exposure at the vessel inner radius to the gradient data given in Tables 4.1-4 through 4.1-6.

4-2

e TABLE--4.1-1 3 CALCULATED REFERENCE NEUTRON ENERGY SPECTRA AT ,

SURVEILLANCE CAPSULE LOCATIONS

• LOWER LOWER ENERGY ENERGY (MeV)- 225'- 265*/275 (MeV) 225" 265'/275'-

1.42E+01 1.10e+07 9.73e+66 2.97E-01 1 59e+10 -1.02e+10 1.22E+01 -3.55e+07 3.08e+07 1.83E 1.44e+10 9.29e+09 1.00E+01 1.56e+08 1.31e+08 1.11E-01 1.17e+10 -7=.55e+09

-8.61E+00- 3.05e+08 2.54e+08 6.74E-02 1.00e+10 .6.43e+09-7.41E+00 5.44e+08 4.42e+08 4.09E-02 '4.27e+09 2.73e+09 6.07E+00 1.38e+09 1.11e+09 3.18E-02 3.00e+09, 1.90e+09 4.97E+00 2.13e+09 1.66e+09 2.61E-02 2.32e+09 1.51e+09 ,

3.66E+00 -4.29e+09 3.19e+09 2.42E-02 1.88e+09 1.21e+09

.3.01E+00 3.35e+09 2.40e+09 2.19E-02 5.55e+09: 3.45e+09 2.73E*00 2.55e+09 1.81e+09 1.50E-02 1.20e+10 7.65e'09 +

2.47E+00 2.95e+09 2.07e*09 7.10E 1.34e+10 8.58e+09 2.37E+00 1.47e+09' 1.03e+09- 3.36E 1.29e+10- 8.21e+09 2.35E+00 3.99e+08 2,77e+08 1.59E-03 .2 . 0 6 e+10 - 1.31e+10 ,

2.23E+00 1.98e+09 1.38e+09 4.54E-04 1.25e+10 7.93e+09 1.92E+00 5.25e+09 3.52e+09 2.14E 1.28e+10 8.09e+09 1.65E+00 5.81e+09 3.96e+09 1.01E-04 1.72e+10 1.08e+10 1.35E+00 8.55e+09 5.77e+09 3.73E-05 2.15e+10 1.35e+10 1.00E+00 1.39e,10 9.29e+09 1.07E-05 1.28e+10 8.04e+09 8 '. 21E- 01 8.95e+09 5.92e+09: .5.04E-06 1.74e+10 1.09e+10 7.43E-01 5.06e+09 3.34e+09 1.86E-06 1.34e+10 8.34e+09 6.08E-01 1.20e+10 7.81e+09 8.76E-07 '1.34e+10 8.30e+09 4.98E-01 1.03e+10 6.67e+09 4.14E-07 3.19e+10 1.96e+10 3.69E-01 1.15e+10 7.48e+09 1.00E 1.28e+11 7.60e+10 2.97E-01 9.54e+09 6.19e+09 0.00 NOTE: The upper energy of group 1 is 17.33 Mev.

4-3

TABLE 4.1-2 REFERENCE NEUTRON SENSOR REACTION RATES AND EXPOSURE PARAMETERS AT THE CENTER OF SURVEILLANCE CAPSULES 225*- 265'/275*

Reaction Rate (ros/ nucleus)

' Cu-63 (n, ot) 4.80e-17 3.91e-17' Ti-46(n,p) 7.92e-16 6.31e-15

.Fe-54 (n, p). 5.09e-15 3.84e-15 4 Ni-58(n,p) 6.68e-15 5.01e-15

-U-238(n,f)- (Cd) 2.10e 1.50e-14 U-238(y,f) 1.60e-15 1.02e-15 Neutron Flux (n/cm _gge) a

$(E > 1.0 MeV) 5.52e+10 3.85e+10

$(E > 0.1 MeV) 1.45e+11 9.70e+10 doa/see Displacement Rate 8.74e-11 6.03e-11

$(E >-0.1)/$(E > 1.0). 2.63 2.52 (dpa/sec)/$(E > 1.0) 1.58E-21 .1. 57 E U238(y,f)/U238(n,f) 0.076 0.068

~

4-4

~ ...

e 4

TABLE-4.1 ~

SUMMARY

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

-FLUX'(n/cm .3,c) 2 THETA IE >'0.11 doa/sec (deo) (E > 1.0) ( E s - 0.1 )' -doa/sec IE > 1.01'fE > 1.01--

0 . 0 0'- -- 2 . 51 e + 10 6.58e+10 -4.05e-11 2.62 1.61E-21 5.00- 2.42e+10- 6'.81e+10 3.97e-11 2.81 1.64E-21 9'75 2.50e+10 6'56e+10 4.04e-11~ 2.62- 1.62E-21 15.25 '2-23e+10 -5.92e+10'.3.61e-11

. 2.65 1.' 6 2 E-21 19.78 -1.98e+10- 5.07e+10 3.22e-11 2.56- 1.63E-21

-24.75 -l.93e+10 4.86e+10 3.14e 2.52 1.63E-21.

30.00- '2.35e+10 6.21e+10 3.81e-11 2.64 1.62E-21 35.25 3.02e+10 8.66e+10 4.91e-11' 2.87 1.63E-21 39.71 3 . 4 0e+10, --1. 01e+11 - 5.53e 2.97 1.63E-21 45.00 3.44e+10 . 00e+11' 5.68e-11 2.91'. 1.65E-21 i

t 4-5 l

4

TABLE 4.1-4 RELATIVE RADIAL DISTRIBUTION OF $(E 2 1.0 MeV)

WITHIN THE PRESSURE VESSEL WALL RADIUS (cm) 90.0* 75.0* 60.0* M 0.0*

179.85'" 1.000 1.000 1.000 1.000 1.000 180.36 0.973- 0;959 0.966 0.966 0.973 182.37 0.785 0.788 0.796 0.787 0.785 184.09 0.637 0.641 0.642 0.640 0.637-184.37'" 0.617 0.621 0,621 0.620- 0.617 186.39- 0.470 0.473 0.473 0.472 0.470 188.40 0.357 0.359 0.362 0.359 .0.357 190.41 0.270 0.272 0.274 0.271- 0.270 192.42 0.202 0.204 0.206- 0.202 0.202 193.14 0.183: 0.185 0.187 0.183 0.183 193.42'"- 0.176 0.178 0.180 0.176 0.176 194.43 0.150 0.151 0.153 0.149 0.150 196.44 0.108 0.110 0.111- 0.106 0.108 197.94"' O 089 0.092 0.093 0.088 0.089 NOTES: (1) Indicates Base Metal Inner Radius

= (2) Indicates Base Me:tal 1/4T (3) Indicates Base Metal-3/4T-(4)- Indicates Base Metal Outer Radius 4-6 9

N

. . . _ . . . . . . . _ . - - . _ . _ _ _ _ . - _ , _ . . . _ _ _ _ _ - _ _ _ _ _ _ . . . _ _ . _ - . . ~ .

i TABLE 4.1-5 ,

. RELATIVE RADIAL DISTRIBUTIONLOF $(E 2 0.1 MeV) f WITHIN THE PRESSURE VESSEL PALL--

-RADIUS ,

, ( cm) - 90.0* 75.0" $243* 152S1 0 , 0* -

" 179.85"'- 1.0001 1.000- 1.000 1.000 1.000-l 180.36 1.000J -1.000: :1.000 1.000 1.000-1182.37 0.960 0.957 0.964- 0.941 0.960 184.09 0.887- -0.884 0.894- -0.863 0.887

'184.37 3' 0.874 0.871 0.881 0.849 0.874 186.39 0.780 0.777. 0.790 0.751 0.780-188.40 0.685 0.682 0.698 0.654 . ~ 0 ~.' 6 8 5 -

190.41 0.592- 0.591: 0.604- 0.559 0.592~

192.42 0.500- 0.503 0.515- 0.468 0.500

193.14 0.469 0.472 0.483 -0.437 0.469 i: 193,42'" 0'.457- 0.460 0.471 0.425 -0.457 194'.43 0.413 0.417 0.426 0.381 .0.413

. -196.44 0.326 0,332- -0.340 0.293: 0.326

. 197.94"' O.277 0.285 0.292 0.243 0.277 NOTES: (1) Indicates Base Metal-Inner Radius-3 (2) Indicates Base Metal 1/4T (3) Indicates = Base Metal 3/4T 7 (4)- Indicates Base Metal 10 uter Radius ~

t.

4-7

TABLE 4.1-6 RELATIVE RADIAL DISTRIBUTION OF dpa/sec WITHIN THE PRESSURE VESSEL WALL RADIUS (cm) 90.0* 7 5 . 0". 60.0 45.0 0. 0*

179.85 1.000 1.000 1.000 1.000 1.000 180.36 0.974 0.973 0.974 0.974 0.974 182.37 0.832 0.831 0,836 0.834 0.832 184.09 0.714 0.712 0.719 0.717 0.714 184.37'" 0.697 0.695 0.702 0.700 0.697 186.39 0.575 0.573 0.583 0.576 0.575 188.40 0.474 0.472 0.481 0.473 ,

0.474 190.41 0.388 0.387 0.397 0.387 0.388 192.42 0.315 0.318 0.322 0.311 0.315 193.14 0.292 0.295 0.300 0.287 0.292 193.42'" 0.283 0.286 0.291 0.278 0.283 194.43 0.250 0.254 0.260 0.245 0.250 196.44 0.194 0.199 0.203 0.186 0.194 197.94"' O 165 0.171 0.175 0.154 0.165 NOTES: (1) Indicates Base Metal Inner Radius (2) Indicates Base Metal 1/4T (3) Indicates Base Metal 3/4T (4) Indicates Base Metal Outer Radius 4-8 l

4.2 - Fuel Cycle Specific Adjoint Calculations Results of the' fuel cycle specific adjoint transport calculations for the first 16 cycles of operation at Fort Calhoun are summarized in tables 4.2-1 through 4.2-6. The data listed in these tables establish the means for absolute comparison of analysis _ and n easurement 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.

The calculated fast neutron flux (E 2,1.0 MeJ) at the center of surveillance capsules located at azimuthal positions.of 225 and 265"/275* are provided for each of the sixteen operating cycles in Table 4.2-1. The data as tabulated represent the maximum flux location in the axial distribution. Similar data applicable to the pressure vessel inner radius are given in Table 4.2-2.

l L Exposure parameter ratios necessary to convert the cycle specific data listed in Tables 4.2-1 and 4.2-2 to other key fast neutron {

exposure units'are given in Section 4.1 of this report. h Application of these ratios to the data from Tables.4.2-1 and

[

4.2-2 yielded corresponding exposure in terms of neutron flux (E 2 0.1 MeV) (Tables 4.2-3 and 4.2-4) and iron atom displacement rates (Tables 4.2-5 and 4.2-6).

l.

4-9 i

i

TABLE 4.2-1 CALCULATED FAST NEUTRON FLUX (E 2 1.0 MeV) AT THE SURVEILLANCE CAPSULE CENTER CAPSULE LOCATION 225* 265"/275" Cycle 1 6.85e+10 4.98e+10 Cycle 2 6.72e+10 4.79e+10 Cycle 3 8.03e+10 5.08e+10 Cycle 4 6.91e+10 4.83e+10 g

Cycle 5 7.14e+10 4.60e+10 I Cycle 6 7.36e+10 5.27e+10-Cycle 7 7.40e+10 5.49e+10 Cycle 8 5.15e+10 3.75e+10 Cycle'9 4.90e+10 2.75e+10 Cycle ?.0 3.35e+10 5.02e+10 Cycle 11 5.80e+10 3.09e+10 Cycle 12 4.87e+10 3.04e+10 Cycle 13 5.10e+10 2.72e+10 Cycle 14 3.53e+10 2.94e+10 0'

Cycle 15 2.73e+10 2.94e+10 Cycle 16 3.04e+10 2.53e+10

)

4-10 i

l

TABLE 4.2-2 CALCULATED FAST NEUTRON FLUX (E 2 1.0 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE AZIMUTHAL ANGLE '

90.0* 75.0* 60.0" 45.0' O.0" , ,

Cycle 1 3.27e+10 2.80e+10 2.80e+10 4.26e+10 3.2'le+10 Cycle 2 3.17e+10 2.6be+10 2.76e+10 4.18e+10 3.17e+10 Cycle 3 3.35e+10 2.84e+10 3.12e+10 4.99e+10 3.35e+10 i Cycle 4 3.18e+10 2.72e+10 2.84e+10 4.31e+10 3.18ei10

! Cycle 5 3.04e+30 2.58e+10 2.86e+10 4.44e+10 3.04e+10 Cycle 6 3.48e+10 2.93e+10 3.02e+10 4.58e+10 3.04e+10 .

Cycle 7 3.62e+10 3.04e+10 3.10e+10 4.61e+10 3.62e+10 Cycle 8 2,41e+10 2.40e+10 2.45e+10 3.22e+10 2.41e+10 Cycle 9 1.79e+10 1.68e+10 2.11e+10 3.06e+10 1.74e+ic .i ; 4 Cycle 10 3.36e+10 2,57e+10 1.670+10 2.1La+10 1.05e+10 l Cycle 11 1.97e+10 2.14e+10 2.59e+10 3.63e+10 1.80e+16 '

Cycle 12 1.95e+10 2.04e+10 2.34e+10 3.05e+10 2.11e+10 I

Cycle 13- 1.80e+10- 1.70e+10 2.28e+10 3.19e+10 1.66e+10 * -

Cycic 14 1.95e+10 1.63e+10 1.45e+10 2.20e+10 1.08e+10 Cycle 15 1.96e+10 1.61e+10 1.29e+10 1.71e+10 1.32e+10 Cyc)n 16 1.660+10 1.50e+10 1.34e+10 1.90e+10 1.12e+10 I

\1 i<

-(

a l

TABLE 4.2-3 CALCULATED FAST !JEUTROli FLUX (E 2,0.1 MeV) AT THE SURVEILLA!!CL CAPSULL CE!JTER l

, CAPSULE LOCATIOli 225* 265"/275*

l Cycle 1 1.80e+11 1.25e+11

Cycle 2 1.77e+11 1.21e+11 Cycle 3 2.13e+11 1.28e+11

, Cycle 4 1.82e+11 1.'le411 Cycle 5 1.88e+11 1..,ie+11 h-Cycle 6 1.93e+11 1.33e+11 Cycle 7 1.94e+11 1.38e+11 Cycle 8 1.35e+11 9.44e+10 Cycle 9 1.29e+11 6.91e+10 Cycle 10 8.82e+10 1.26e+11

) Cycic 11 1.5.?e+11 7.79e+10 [

b Cycle 12 1.28e+11 7.64e+10 t Cycle 13 1.34e+11 6.86e+10 Cycle 14 9.27e+10 7.39e+10 Cycle 15- 7.16e+10 7.40e+10 Cycle 16 7.99e+10 6.37e+10 1

d

\'

d e t 4

4-12 4 '

\

TABLE 4.2-4 CALCULATED FAST NEUTRON FLUX (E 2 0.1 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE AZ1MUTHAL ANGLE 90.0' 75.0* 60.0' 45.0* 0. 0*

l Cycle 1 8.59e+10 7.39e+10 7.43e+10 1.24e+11 8.59e+10 L Cycle 2 8.32e+10 7.01e+10 7.32e+10 1.22e+11 8.32e+10 l Cycle 3 8.80e+10 7.51e+10 8.28e+10 1.45e+11 8.80e+1' Cycle 4 8.35e+10 7.19e+10 7.53e+10 1.25e+11 8.35e Cycle 5 7.97e+10 6.81e+10 7.58e+10 1.29e+11 7.97e+1s Cycle 6 9.13e+10 7.74e+10 8.00e+10 1.33e+11 9.13e+10 Cycle 7 9.51e+10 8.03e+10 8.21e+10 1.34e+11 9.51e+10 Cycle 8 6.33e+10 6.34e+10 6.51e+10 9.37e+10 6.33e+10 Cycle 9 4.71e+10 4.42e+10 5.59e+10 8.91e+10 4.71e+10 Cycle 10 8.82e+10 6.80e+10 4.43e+10 6.11e+10 2.76e+10 Cycle 11 5.17e+10 5.66e+10 6.88e+10 1.05e+11 4.73e+10 Cycle 12 5.12e+10 5.39e+10 6.19e+10 8.87e+10 5.54e+10 Cycle 13 4.72e+10 4.49e+10 6.06e+10 9.28e+10 4.36e+10 Cycle 14 5.12e+10 4.30e+10 3.84e+10- 6.40e+10 2.84e+10 Cycle 15 5.16e+10 4.24e+10 3.42e+10 4.96e+10 3.47e+10 Cycle 16 e. 37e+10 3.96e+10 3.54e+10 5.52e+10 2.94e+10 4-13

_ _. _ j

TABLE 4.2-5 CALCULATED IRON ATOM DISPLACEMENT RATE AT THE SURVEILLANCE CAPSULE CENTER CAPSULE LOCATION 225' 265*/275" Cycle 1 1.09e-10 7.80e-11 Cycle 2 1.06e-10 7.50e-11 Cycle 3 1.27e-10 7.96e-11 Cycle 4 1.09e-10 7.56e-11 Cycle 5 1.13e-10 7.20e-11 Cycle 6 1.17e-10 8.26e-11 Cycle 7 1.17e-10 8.59e-11 Cycle 8 8.15e-11 5.87e-11 Cycle 9 7.75e-11 4.30e-11

, Cycle 10 5.31e-11 7.87e-11 l

Cycle 11 9.19e-11 4.85e-11 Cyclo 12 7.71e-11 4.76e-11 Cycle 13 8.08e-11 4.27e-11 Cycle 14 5.59e-11 4.60e-11 Cycle 15 4.32e-11 4.60e-11 Cycle 16 4.81e-11 3.96e-11 4-14

I TABLE 4.2-6 CALCULATED IRON ATOM DISPLACEMENT RATE AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE i.

AZIMUTHAL ANGLE 90.0* 75.0* 60.0' 45.0" 0.0' Cycle 1 5.29e-11 4.53e-11 4.55e-11 7.04e-11 5.29e-11 Cycle-2 5.12e-11 4.30e-11 4.48e-11 6.91e-11 5.12e-11 Cycle 3 5.41e-11 4.61e-11 5.07e-11 8.24e-11 5.41e-11 Cycle 4 5.14e-11 4.41e-11 4.61e-11 7.11e-11 5.14e-11 Cycle 5 4.91e-11 4.18e-11 4.64e-11 7.34e-11 4.91e-11 Cycle 6 5.62e-11 4.75e-11 4.89e-11 7.56e-11 5.62e-11 Cycle ? 5.85e-11 4.93e-11 5.03e-11 7.61e-11 5.85e-11 Cycle B 3.90e-11 3.89e-11 3.98e-11 5.32e-11 3.90e-11 Cycle 9 2.90e-11 2.71e-11 3.42e-11 5.06e-11 2 90e-11 Cycle 10 5.43e-11 4.17e-11 2.71e-11 3.47e-11 1.70e-11 Cycle 11 3.18e-11 3.47e-11 4.21e-11 5.99e-11 2.91e-11 Cycle 12 3.15e-11 3.31e-11 3.79e-11 5.04e-11 3.41e-11 Cycle 13 2.90e-11 2.75e-11 3.71e-11 5.27e-11 2.68e-11 Cycle 14 3.15e-11 2.64e-11 2.35e-11 3.63e-11 1.75e-11 Cycle 15 3.17e-11 2.60e-11 2.10e-11 2.82e-11 2.23e-11 Cycle 16 2.69e-11 2.43e-11 2.17e-11 3.13e-11 1.81e-11 4-15

SECTION 5.0 EVALUATIONS.0F SURVEILLANCE CAPSULE DOSIMETRY In this section, the results of the evaluations of the three neutron sensor sets-withdrawn as a part of the Fc.c Calhoun

, 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:

L AZIMUTHAL WITdDPAWAL IRRADIATION CAPSULE ID. LOCATION TIME __ TIME (EFPS)

W225 225' END OF CYCLE 3 7.72e+07 W265 265* END OF CYCLE 7 1.87e+08 W275 275* END OF CYCLE 14 4.28e+08 5.1 - Measured Reaction Rates The radiometric covnting of dosimetry from these three surveillance capsules was carried out by Ccmbustion Engineering Inc. for Capsules W225 and W265 and by B&W Nuclear Technologies I Inc. for Capsule W275. The measured specific activities for each of the sensors contained in these dosimetry-sets are.provided.in

-Reference 6.

.The irradiation history of the Fort CalhounLreactor during Cycles 1 through 14: Was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report *D' for the applicable operating periods. -The-data in NUREG-0020 is based on. core follow information provided on a monthly basis-by OPPD. The fine detail, i.e. monthly intervals, is necessary in performing radioactive decay corrections for-each of the neutron sensors.

In addition to the reactor power 1.istory, for.the multiple cycle-irradiations of Capsules W225, W265,-and W275, the flux level adjustment factors for each cycle were determined from the fuel cycle specific adjoint calculations described in Section 4.2 of

-this report.

5-1

4 I

Based on the irradiation history, the individual sensor characceristics, and the measured specific activities, reaction ,

rates averaged over the appropriate irradiation periods were  :

computed for the sensor sets removed from Capsules W225, W265, dnd W275.. The computed reaction rates for the multiple foil ,

sensor sets are provided in Table 5.1-1.

j In regard to the data listed in Table 5.1-1, the fission rate l measurements from the U-238 sensors include corrections-for U-235  !

impurities, the_ build-in of plutonium isotopes during the long i irradiations, and for the effects of Y,f reactions.

5.2 - 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 5.2-1 through 5.2-3. In these tables, the ,

-derived exposure experienced by each capsule along with data l illustrating the fit of both the trial end adjusted spectra to t the measurements are given. Also included in the tabulations are.

~

the-10 uncertainties associated with each-of the derived exposure l

~

rates.

f In regard to the comparisons listed in Tables 5.2-1 through 5.2-3, it should be noted that the~ columns labeled " trial cale"'

were obtained by normalizing the-neutron spectral data from Table J 4.1-1 to'the absolute calculated $(E 2 1.0 MeV) averaged over the l applicable irradiation periods (Cycles 1-3 for Capsule W225, Cycles.1-7_for Capsule W265, and Cycles _1-14 for Capsule W275) as i discussed .h. Section 2.0. Thus, the comparisons illustrated in Tables 5.2ei through 5.2-3 indicate the degree to which the calculated-neutron energy spectra matched the. measured sensor data bciore and after adjustment. Absolute comparisons are discussed further in Section 6.0 of this report.

5-2

__ _ _ - , __.a._ _

TABLE 5.1-1

SUMMARY

OF REACTION RATES DERIVED FROM MULTIPL3 FOIL SENSOR SETS WITHDRAWN FROM INTERNAL SURVEILLANCE CAPSULES CAEgCLE W225 REACTION RATE (rps/ nucleus)

M MIDDLE BOTTOM-Cu-63(n,a) Cd 6.45e-17 6.57e-17 6.15e Ti-46(n,p) _9.77e-16 9.55e-16 9.71e-16 Fe-54(n,p) 6.00e-15 6.39e-15 5.80e-15 Ni-58(n,p) Cd 8.36e-15 8.53e-15 7.74e-15 U-238(n,f) Cd 3.09e-14 3.03e-14 2.55e-14 CAPSULE W265 REACTION RATE (rps/ nucleus)

M MIDDLE BOTTOM Cu-63(n a) Cd 5.94e-17 6.40e-17 5.29e-17 Ti-46(n,p) 7.69e-16 7.23e-16 7.06e-16 Fe-54(n,p) .4.79e-15 4.22e-15 4.32e-15 Ni-58(n.p) Cd 6.05e-15 5.80e-15 5.21e-15 U-238(n,f) Cd 1.71e-14: 1.60e-14 1.53e-14 CAPSULE W275 REACTION RATE (rps/ nucleus)

M MIDDLE BOTTOM Cu-63(n,a) Cd 6.45e-17 6.57e-17 6.15e-17 Ti-46(n,p) 4.97e-16 6.31e-16 5.78e-16 Fe-54(n,p) 3.67e-15 3.38e-15 3.02e-15 Ni-58(n,p) Cd 4.16e-15 3.90e-15 3.47e-15 U-238(n,f) Cd -1.47e-14 1.36e-14 1.20e-14

. = - _

5-3 4_

TABLE-5.2-1 DERIVED EXPOSURE RATES FROM' SURVEILLANCE CAPSULE W225 DOSIMETRY WITHDRAWN AT THE EN1. OF FUEL CYCLE 3 TRIAL- ADJUSTED 10 M M UNCERTAINTY

$(E y. 1.0 MeV) 7.13e+10 7.07e+10 10%

$(E 2,0.1 MeV)- 1.80e+11 1.94e+11 19%

dpa/sec 1.12e-10 1.12e-10 12%

l

\

COMPARISON ~OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE W225 REACTION RATE (rps/ nucleus)

-TRIAL ADJUSTED M/C M/C MEASURED CALC. Em . TRIAL ADJUSTED Cu-63(n,u) Cd 6.39e-17 6.18e-17 6.28e-17 1.03 1.02 Ti-46(n p) 9.68e-16 1.02e-15 9.74e-16 0.95 0.99 Fe-54(n,p) 6.06e-15 6.55e 6.22e 0.93 0.97 Ni-58(n,p) Cd 8.21e-15 8.60e-15 8.25e-15 0.95 l'.00 U-238(n,f) Cd 2.89e-14 2.71e-14 2.68e-14 1.07- 1.08-4 e

e..:

TABLE ~5.2-2 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE W265 DOSIMETRY  ;

WITHDRAWN AT THE Elm OF FUEL CYCLE 7 l

TRIAL- ADJUSTED 10 VALUE VALUE ,Q{, CERTAINTY

_$(E 2 1.0 MeV) 5.00e+10 4.07e+10 9%

$(E _2 0.1 MeV) 1.31e+11 1.05e+11- -19%

dpa/see 1.06e-10 6.52e-11 12%

4 COMPARISON-OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE W265 REACTION RATE (rpr/ nucleus)

TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. M ADJUSTED Cu-63(n,u). Cd 5.88e-17 5.05e-17 5.64e 17 1.16 1.04 Ti-46(n,p) 7.33e-16 8.15e-16 7.52e-1(, 0.90 0.97 Fe-54(n.p) 4.44e-15 4.96e-15 4.46e-18 0.90 1.00 Ni-58(n,p)- _Cd 5.68e-15 6.47e-15 5.73e-15 0.88 0.99 U-238(n,f) Cd 1.61e-14 -1.94e-14 1.63e-14 0.83 0.99 5-5 l

TABLE 5.2-3 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE W275 DOSIMETRY WITHDRAW 11 AT THE END OF FUEL CYCLE 14 TRIAL ADJUSTED 10 VALUE VALUE UNCERTAINTY

$(E 2 1.0 MeV) 4.03e+10 3.29e+10 10%

$(E 2 0.1 MeV) 1.06e+11 8.60e+10 19%

dpa/sec 6.38e-11 5.13e-11 12%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE W275 REACTION RATE (rps/ nucleus)

TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Ti-46(n,p) 5.69e-16 6.57e-16 5.62e-16 0.87 1.01 Fe-54(n,p) 3.36E-15 4.00e-15 3.30E-15 0.84 1.02 Ni-58(n,p) Cd 3.84E-15 5.22e-15 4.u2E-15 0.74 0.96 U-238(n,f) Cd 1.34E-14 1.56e-14 1.28E-14 0.86 1.05 5-6 l

[

SECTION 6.0 .

BEST ESTIMATE NEUTRON EXPOSURE AND RTm PROJECTIONS FOR FORT CALHOUN PRESSURE VESSEL MATERIALS i

In this section the measurement results provided in Section 5.0 are combined with the results of the neutron transport 4

calculation's described in Section 4.0 to establish a mapping of the best (stimate neutron exposure of the beltline region of the i Fort Calhoun reactor pressure vessel through the completion of 1 Cycle 14. Based on the continued use of the low leakage core power distributions characteristic of the design of Cycles.15 and ,

'16, projections of future vessel exposure are also provided. In  :

addition to the spatial mapping of the fast neutron exposure over t the beltline region, RTn., projections pertinent to the limit beltline material oro highlighted.

6.1 Comparison of Calculations with Measurements As described in Section 2.3, the best estimate neutren exposure [

projectionb for the Fort Calhoun pressure vessel were based on a combination of plant specific neutron transport calculations and plant specific meacurements. Direct comparisons of the transport calcu)ations with the Fort Calhoun 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. i

. .In-this section, comparisons of the meararement results from ,

surveillance capsules W225, W265, and W275-with corresponding analytical predictions at the measurement locations are presented. These comparisc.1s are provided on two levels. In the first' instance, predictions of fast-neutron exposure rates in '

terms of $(E 2; 1.0 MeV), 9(E 2 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 n*:e-compared directly-with the measured data-from the counting laboratories. It is shoen that these two levels of-comparison yield-consistent and similar resu.its, 6-1

l l

)

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 performed for the Fort Calhoun reactor to produce "best estimate"-exposure projections for the pressure vessel  !

wall.  !

i 6.1.1 Comparison of Least Squares Adjustment Results with l Calculation In Table 6.1-1, comparisons of measured and calculated exposure ces 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 Yable-6.1-1 indicates that, considering all of the available core midplane data, the measured exposure rates were less than calculated values by factors of.0.874, 0.897, and 0.880 for $(E 2 1.0 MeV), $(E 2 0.1 MeV), and dpa/sec, respectively. The 10 standard deviations associated with each of the 3 sample data sets were 11.7%, 17.7%, and 12.1%,

respectively.

6.1.2 Comparisons of Measured and Calculated Sensor Reaction Rates In Table 6.1-1, measurement / calculation (M/C) ratios for each fast neutron sensor reaction rate from the three 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. i An examination of Table 6.1-2 shows concistent behavior for all 1

-reactions and all measurement points. The standard deviations .

observed for the six fast neutron reactions range from 6.7% to 13.9% on an individual reaction basist whereas, the overall 6-2 s

-_ -.- . . . . _ _ _ _ _ _ ___ - . _ , . - . . _ . _ _ _ _ . , . . ~ . - _ - - . . - _ . _ , , - , .

average M/C ratio for the entire data set has an associated 30 standard deviation of 12.8%. Perthermore, the average M/C bias of 0.918 observed in the reaction rate comparisons is in excellent agreement with tese values of 0.874, 0.897, and 0.880

. observed in the exposure rate comparisons shown in Table 6.1-1.

C-3

___.._.._._ __- ______ . . _ _ _ _ _ _ _ .-~. _ .- _ . _ _ . _ _ . . . _

I TABLE 6.1-1  ;

COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM ,

SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS

{

6(E

• 1.0-MeV) In /cm;l _

CALCULATED MEASUPED ji/,C, j Capsule WR25 1.13e+10 7.07e+10 0.992 =!

Capsule W265; (5.00e+10 4.07e+10 0.814 {

'apsule W275 4.03e+10 3.29e+10 0.316 AVERAGE M/C BIAS FACTOR (K) 0.874 PERCENT STANDARD DEVIATION (10) , 11.7% j 4

6(E > 0.1 MeV) f n /cm 81 CALCULATED MEASURED ji/,,,Q, Capsule W225 1.80e+11 1.94e+11 1.080  ;

Capsule W265 .

1.31e+11 1.05e+11 0.799  ;

Capsule W275 1.06e+11 8.60e+10 0.812 -i i

AVERAGE M/C BIAS FACTOR (K) 0.897 PERCENT STANDARD DEVIATION-(10)' 17.7%

t Iron Disolacement Rate'fdoa/seci ,

CALCULATED MEASURED jff.,C, Capsule W225 1.12e-10 1.12e-10 1.003 Capsule W265 7.92e-11 6.52e-11 0.823 Capsule W275 6.38e-11 5.19e-11 0.813 J

AVERAGE M/C BIAS FACTOR (K) 0.880 PERCENT STANDARD DEVIATION (10) 12 '%

6-4  ;

s -. en w-- + w-- w i =- w av o -em. -g- ,+re-- ww- - + - - - -,m-me.m-w u--v=-a*,v+m e-- , +. ew en w o m, +'v-e,- w-*r e-m-* --r--v=t--w w -

w ~ p w, r wrrv- vcue g-w o p v

l 1

l TABLE 6.1-2 j i

COMPARISON OF ME,'.SURED AND CALCULATED NEUTRON SENSOR REACTION RATES FROM SURVERLANCE CAPSULE AND CAVITY DOSIMETRY BRRADIATIONS Cu6VM T46(nm) Fe54(nm) NiS8(nm) 53238(n.O CAPSULE W225 TOP I.044 0.958 0.916 0 S72 1.140 MIDDLE I.063 0336 0376 0.992 1.118 BOTTOM 0.995 0.952 0.885 0.900 0.94I CAPSULE W265 TOP I.169 0 S 38 0.960 OS29 0.877 MIDDLE I.260 0.882 0.846 0.891 0.821 BOTTOM 1.041 0.861 0.866 0.743 0.785 cn J, CAPSULE W275 TOP 0.753 0913 0.792 0336 MIDDLE 0.956 0.841 0.743 0.866 BOTTOM 0.876 0.751 0.661 0.764 AVERAGE 1.095 0.901 0.884 0.853 0.916 8.2 6.7 63 12.4 13.9

% STD DEV (lo)

OVERALL AVERAGE ME RATIO 0918 PERCENT STANDARD DEVIATION (1o) ,

12.8%

4_ . _ .

6.2 Exposure Distributions Within the Beltline Region As' described in Section 2.3~of this report, the best estimate vesselfexposure was determined from the following relationships

- es = K e m ,

wheret

~

4,,,, en , = The best estimate fast neutron exposure at the location of interest.

-K = The plant specifi'c measurement / calculation (M/C)  ;

bias factor derived fium all available. surveillance capsule and reactor cavity dosimetry. j data.  !

4..ic, =- The absolute calculated fast j neutron exposure at the l location of interest. ,

From:the data provided in Table 6.1-1, the plant specific bias factors (K).to be applied to the calculated exposure-values given r in Section 4.2 were as follows:  ;

4(E > 1.0 MeV)~ 0.874 1 11.7%

$(E >-0.1 MeV) 0.897 1 17.7% i

- dpa 0.880 2 12.1%

e i

These bias factors were based on the results of ,. se' continuous 1

. monitoring: program at Fort Calhoun that has pr;>vided measured  :

data from three internal surveillance capsules through the first 13.6 effective full power years of operation.

F

~ 6-6

< A

- , , - - r,'j~ N g~. .p... e. . , , , , , p-w. -

n + e,- -

-e ---m- -,-o,-vw -- wr-- ~~w---* nwc ~, ----v - ~ , ~w---- w--,e---- --+ se , A~-~sr- -~l - ~ =~w c

C The uncertainties listed with the individual bias factors are at the lo level. Additional uncertainties associated with the evaluation of the best estimate vessel exposure are discussed in Section 6.4.

6.2.1 Exposure Accrued During Cycles 1 through 14 To assess the incremental exposure resulting from the Cycles 1 through 14 irradiations, the bias factors listed in Section 6.1 were applied-directly to the calculated values from section 4.2 for the vessel clad / base metal interface to produce best estimate fluence levels characteristic of the materials comprising the beltline region of the recctor pressure vessel. '1e best estimate results applicable to the vessel inner surface are incorporated into Table 6.2-1 to establish the exposure accrued by ts.* reactor vessel through the end of Cycle 14. Exposure distribu' ions through the vessel wall, can be developed using these surfa7e exposures and radial distribution functions from Section 4.0. .

6.2.2 Projection of Future Vessel Exposure At the end of Cycle 14, the Fort Calhoun reactor had accrued 13.6 effective full power years (EFPY) of operation. .In order to establish a framework for the assessment of future vessel condition, e::posure projections through- 32 EFPY are also included in Table 6.2-1 in addition to the plant _ specific exposure assessments through the end of Cycle 14. These exposure projections for the Fort Calhoun pre',sure vessel 'are illustrated graphically in Figure 6.2-1.

These-temporal extrapolations into the future were based on the assumption _that the best estimate neutron; exposure rates averaged over Cycles 15 and 16 were representative of all future fuel cycles. That is, that future fuel designs would incorporate the

' low leakage fuel management concept. employed during Cycles 15 and

16. Examination of these projected exposure levels establishes the long-term effectiveness of the low leakage-fuel management incorporated to date and can be used as a guide in assessing 6-7

strategies for future vessel exposure management. The validity of these projections for future operation will be confirmed via r the next scheduled surveillance capsule withdrawal.

9 C

e 9

1 6-8

--.v . - iy -

---

.-r --- - - -

..m--y n, m ~ y, sm-. ,, ,

o 1

1 5 TABLE 6.2-1 i

] NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ,

ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 4(E 2,1.0 MeV) [n/cm')

Iffl 90' 75' _ 60' 45' 0*  ;

- EOC 1 7.17e+17 6.14e+17 6.14e+17 9.35e+17 7.17e+17 L

EOC 2 1.56e+18 - 1.32e+18 1.35e+18 2.05e+18 1.56e+18 EOC 3 2.19e+18 1.86e+18 1.94e+18 2.99e+18 2.19e+18 EOC- 4 2.83e+18 2.40e+18 2.51e+18 3.86e+18 2.83e+18 EOC 5 3.61e+18 3.07e+18 3.24e+18 5.00e+18 3.61e+18 EOC 6 4.53e+18 3.85e+18 4.04e+18 - 6.21e+18 4.53e+18 '

4 EOC 7 5.38e+18 4.56e+18 4.77e+18 7.29e+18 5.38e+18 EOC 8 5.95e+18 5.13e+18 5.35e+18 8.06e+18 5.95e+18 EOC 9 6.51e+18 5.65e+18 6.01e+18 9.01e+18 6.51e+18 L EOC 10 7.47e+18 6.38e+18 6.48e+18 9.61e+18 6.81e+18

, EOC'11 8.11e+18 7.08e+18 7.33e+18 1.08e+19 7.40e+18 '

EOC 12 8.62e+18 7.61e+18 7.93e+18 1.16e+19 7.95e+18 EOC 13 9.27e+18 8.23e+18 8 . 7 6 e +1,8 1.27e+19 8.55e+18

. EOC 14 9.91e+18 8.77e+18 '9.24e+18 1.35e+19 -8.76e+18 16.0 1.11e+19 9.78e+18 1.01e+19 1.50e+19 9.55e+18

18.0 1.21e+19 1.06e+19 1.08e+19 1.62e+19 1.02e+19 .

20.0 1.30e+19 1.15e+19 1.15e+19 1.75e+19 1.09e+19 22.0 1.40e+19 1.23e+19 1.22e+19 1.87e+19 1.15e+19 '

24.0 1.50e+19 1.31e+19 1.30e+19 2.00e+19 1.22e+19 26.0 1.60e+19 1.40e+19 1.37e+19 2.13e+19 1.28e+19 28.0 1.70e+19 1.48e+19 1.44e+19 2.25e+19 1.35e+19

-30.0- 1.79e+19 1.57e+19 1.51e+19 2.38e+19 1.42e+19 32.0 1.89e+19 1.65e+19 1.58e+19 2.50e+19 - 1.48e+19 .

j t

i I.

4 Y

t j

e%-, ,.,y. ,+,,..,-eve, e -.,--.--,-e-. ~y- . . . - . . , wv .---ww,--,-rv -y,w.--, . , - . . . - ~ - . - ..,.-r- , - - . -

l

)

i TABLE 6.2-1 (Continued)  !

NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS )

i ON THE PRESSURE VESSEL CLAD /BASF 'ETAL INTERFACE

$(E 2' 0.1 MeV) In/cm) 8 ygfj( 90' 75' 60' 45' ,

O' EOC 1 1.93e+18 1.66e+18 1.67e+18 2.79e+18 1.93e+18 EOC 2 4.20e+18 3.57e+18- 3.67e+18 6.11e+18 4.20e+18 EOC 3 5.91e+18 5.03e+18 5.28e+18 8.93e+18 5.91e+18 EOC 4 7.63e+18 6.51e+18 6.83e+18 1.15e+19 7.63e+18

- EOC 5 9.73e+18 8.31e+18 8.83e+18 1.49e+19 9.73e+18 ,

- EOC 6 1.22e+19 1.04e+19 1.10e+19 1.86e+19 1.22e+19 l EOC 7 - 1.45e+19 1.24e+19 1.30e+19 2.18e+19- 1.45e+19 EOC 8 - 1,60e+19 1.39e+19 1.46e+19 2.41e+19 1.60e+19 EOC 9 1.75e+19 1.53e+19 1.64e+19 2.69e+19 1.75e+19 EOC 10 2.01e+19 1.73e+19 1.77e+19 2.87e+19 1.84e+19 i

EOC 11 2.19e+19 1.92e+19 1.99e+19 3.22e+19 1.99e+19 EOC 12 2.32e+19 2.06e+19 2.16e+19 3.46e+19 2.14e+19 EOC 13 2.50e+19 2.23e+19 2.39e+19 3.80e+19 2.30e+19

- EOC 14 2.60e+19 2.31e+19 2.45e+19 3.92e+19 2.30e+19 16.0 2.91e+19 - 2.58e+19 2.68e+19 4.3Ce+19 2.51e+19 18.0- 3.17e+19 2.80e+19 2.87e+19' 4.72e+19 2.68e+19 20.0 3.43e+19 3.02e+19 3.06e+19 5.09e+19 2.85e+19 22.0 3.68e+19 3.25e+19 3.25e+19 5.45e+19 3.03e+19 24.0 3.94e+19 3.47e+19 3.44e+19 5.82e+19 3.20e+19 26.0 4.20e+19 3.69e+19 3.63e+19 6.19e+19 3.37e+19 28.0 4.45e+19 3.91e+19 3,82e+19 6.55e+19 3.55e+19 f

30.0- 4.71e+19- 4.14e+19 4.01e+19 6.92e+19 3.72e+19 32.0 4.97e+19 4.36e+19 4.20e+19 7.29e+19 3.89e+19 6-10

. - - ,_--r-w-,-,%.e- -.y .c-,v4,- w y -.

.m...

..m,,w- ,,y- T-- , , . , , , m,,,y,, 3v--.i , - . , - - - , ---,-r,,~ . -.

~

TABLE 6.2-1 (Continued)

-NEUTRON EXPOSURE PROJECTIONS AT E -LOCA? IONS ON THE PRESSURE VESSEL CLAD / BASE METJ L IN' ERFACE IRON DISPLACEMENTS (dpa)

EEE1 90* 75" 60" 45" Oa EOC- 1 1.17e-03 1.00e-03 1.00e-03 1.55e-03 1.17e-03:

EOC 2 2.54e-03 2.15e 2.20e-03 -3.40e-03 2.54e-03 EOC 3 3.57e-03 3.036-03 3.17e-03 4.98e-03 3.57e-03 EOC 4 4.61e-03 3.92e-03' 4.10e-03 6.41e-03 4.61e-03 EOC 5 5.88e-03 5.00e-03 5~.30e-03 8.31e-03 5.88e-03 EOC 6= 7.38e-03 6.27e-03 6.61e-03 1.03e-02 '7.38e-03 EOC 7- 8.76e-03 7.43e-03 7.80e-03 1.21e-02 8.76e-03 EOC 8 9.69e-03 8.36e-03 8.75e-03 1.34e-02 9.69e-03 EOC 9 1.06e-02 9.21e-03 9.82e-03 1.50e-02 1.06e-02 EOC 10 1.22e-02 1.04e-02 1.06e-02 1.60e-02 1.11e-02 EOC-11 1.32e-02 1.15e-02 '1.20e-02 1.79e-02 1.20e-02 EOC 12 1.40e-02 1,24e-02 1.30e-02 1.93e-02 1.29e-02 EOC 13 1.51e-02 1.34e-02 1.43e-02 2.12e-02 1.39e-12 EOC 14~ 1.60e-02 1.42e-02 1.50e-02 2.22e-02 -1.42e 16.0 1.79e-02 1.58e-02 1.64e 2.47e-02 1.54e-02 18.0 1.95e-02 1.72e-02 1.76e-02 2.68e-02 1.65e-02 20.0 2.11e-02 1.86e-02 1.87e-02 2.89e-02 1.76e-02 22.0 2.27e-02 1'-99e-02 1.99e-02 3.10e-02 1.86e-02:

24.0 2.43e-02 2.13e-02 2.10e-02 3.30e-02 1.97e-02' 26.0 2.-58e-02 '2.26e-02 2.22e-02 3.51e-02 2.08e-02 28.0- 2.74e-02 2.40e-02 2.34e-02 3.72e-02 2.18e-02 30.0 2.90e-02 2.54e-02 2.45e-02 3.93e-02 2.29e-02 32.0- 3.06e-02 2.67e-02 2.57e-02 4.14e 2.40e-02 6-11 i

4 FIGURE 6.2-1 NEUTRON EXPOSURE PROJECTIONS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACM 1.0E+20 e

~

o o "

_. o 1.0E+19 a "

tP"' - w 1

1.0E+) 8 0 10 20 30 40 50 60 70 80 90 Azimuthal nngle (Degrees)

~

• 13.6 EFPY

• 32.0 EFPY i

6-12 li l

l - - - - - ,-

q W 6.3 Projected RTror of Limiting Beltline Material As shown in Figure 1.0-1, the beltline region 'f the Fort Calhoun reactor pressure vessel is. comprised of a series of six shell plates (3 intermediate shell and three lower shell), six longitudinal welds (3 intermediate and 3 lower), and a circumferential weld joining the two shells. The circumferential weld is centered below the axial midplane of the active core in the vicinity of the maximum vessel exposure; while the intermediate shell forging extends upward to an' elevation above the active fuel and the lower shell forging extends-downward to an elevation below the bottom of the active fuel. The maximum neutron exposure expurienced by each of these beltline naterials can be extracted from the data provided in Table 6.1-1 The limiting naterial in the vessel beltline from the standpoint of pressurized thermal shock is longitudinal weld.3-410. This weld raterial occurs at azimuthal locations of O' and 60* in the vessel lower shell with the-maximum neutron exposure point occuring along the 60' azimuth. The Regulatory Guide =1.99 Rev. 2 chemistry factor associated with this material is 229.0 'F, the initial RTun is taken as the generic value of -56 'F, and the Lappropriate margin term for RTn, application is 66 'F. Based on these naterial-propertiet and the maximum best. estimate fluence experienced by the weld, values of RTn, were generated as a function of neutron exposure and are listed in Table 6.3-1. From Table 6.3-1, _ it may be noted that the RTn,- screening criterion of (

270 'F is not exceeded through 32 EFPY of operation.

The data included in' Table 6.3-1 extend from the end of Cycle 14 e (13.6LEFPY) to 32.0 EFPY, Based on an assumed reactor capacity factor of 0.85-in the operating period between the end of Cycle 14 (09/24/93)^and the-license expiration date for the Fort Calhoun Station 1(08/09/2013), approximately 30.5 EFPY will have been accumulated at;the license expiration date. Therefore, weld 3-410 is not projected to exceed- the RTn, screening criterion

- throughout the licensed lifetime of the plant.

6-13

._ . _ - - _ _ . m_._ _ _ _ _ _ _ _ . . - . _ _ . _ . _ _ _ - . _ _ _ . _ _ . - . . _ . _ - - _ .... _ ._ _ _ .._

C +

> TABLE 6.3-1 ,

PROJECTED RTm VALUES FOR WELD MATERIAL 3-410 BASED i ON BEST ESTIMATE FLUENCE PROJECTIONS OPERATING  !

TIME $(E y.,1.0 MeV) RTm -  ;

(EFPY) (n/cm ) 2

(F ) l 13.6 9.24e+18 -233.9 l 16.0 1.01e+19 239.6-18.0 1.08e+19 244.0 20.0 1.15e+19 248.1

• 251.9 22.0 1.22e+19

< 24.0 1.30e+19 255.5  ;

~

26.0 1.37e+19 258.9 i 28.0 1.44e+19 262.1 30.0 1.51e+19 265.1-32.0 1.586+19 268.0- a

,a..

. Note: Based on a capacity' factor of 0.85 for the interval .

- between 09/30/93 (EOC 14) and-08/09/2013 (license  ;

expiration),-the Fort Calhouncreactor will reach

- approximately 30.5 EFPY at-license expiration. i T

+

L Y

l 6 '

. 4 h h

- .-, . :. _. _,~. ,

---.,;;_,,,. , , , , . .,~ .....,,,..-_......,,_.-.,-n.,...,..- , . , . . ,

1 4

6.4 Uncertainties in Exposure Projections The_overall uncertainty in the best estimate exposure projections l within the pressure vessel wall stem primarily from two sources; a) the uncertainty in the bias-factor (K) derived from the plant specific r..<asurement data base and b) the analytical uncertainty associated with relating the results at the measurement locations to the desired results within the pressure vessel wall.

I - 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 e.nd cavity dosimetry sensor sets. The i analytical uncertainty in the relationship between the exposure of the-pressure vessel and the exposure at the measurement

locations are based on the variations in the as-built dimensions 3 1 of the surveillance capsules and the pressure vessel inner diameter, as well as on downcomer water density variations.

The lo. uncertainties associated with the bias factors applicable.

to $(E 2 1.0 MeV),'4(E 2 0.1 MeV), and dpa are given in Section  ;

6.2 of this report. The additional information pertinent to the required analytical uncertainty for vessel locations has been obtained from the analytical _ uncertainty studies described in Section 3.0 of this report.

l Based on the analytical sensitivity studies the additional ,

i' uncertainty associated with the tolerances in dosimetry positioning, vessel inner radius, and downcomer temperature was estimated-to be-6% for-all exposure parameters. These uncertainty. components were then combined as follows:

la UNCERTAINTY 4(E > 1.0 MeV) 4(E > 0.1 MeV) dpg, Bias Factor 11.7% 17.7% 12.1%

Analytical 6.0% 6.0% 6.0%

Combined 13.1% 18.7% 13.5%

4 6-15 j

%.., ._-.-.,m,- , , , , . . , _ , . . . , , - ,_ . . . _ , . . . . - . . . . - . _ , . _ , . - _ _ , , - - ,

e Thus, the total uncertainty associated with the neutron exposure projections at the pressure vessel clad / base metal interface for Fort Calhoun was estimated to be:

la Uncertainty

$(E > 1.0 MeV) 13g

@(E > 0.1 MeV) 19%

dpa 14%

These uncertainty values are well within the 20% 10 uncertainty in vessel fluence projections required by the PTS rule.

i 6-16 l

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

.e*

SECTION 7.0 i

REFERENCES e.

1- ASTM Designation E853-87, ' Standard Practice for Analysis and Interpretation of Light Water Reactor Surveillance i -Results," in ASTM Standards, Section 12, American Society

) for Testing and Materials, Philadelphia, Pa. 1993.

2- Draft RegulatoryiGuide DG+1025, ' Calculational and Dosimetry Methods!for Determining Pressure Vessel Neutron Fluence,

U.S. Nuclear Regulatory Commission, Office of Nuclear

. Regulatory Research, September.1993, t

}

3- RSIC Computer Code Collection CCC-543, ' TORT-DORT Two- and l Three-Di.nensional Discrete- Ordinates Transport, Version

2.8.14,

• January 1994. i 4- RSIC Data Library Collection DLC-175, " BUGLE-93, Production

• and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 .

j Broad Group Neutron / Photon Cross-Section Libraries Derived  :

from ENDF/B-VI' Nuclear Data," April 1994, i  !

. 5- Maerker, R. E., et. al., ' Accounting 1for Changing Source  :

Distributions in Light Water Reactor Surveillance Dosiretry Analysis," Nuclear Science and Engineering, '!olume 94, pages ,

-291-308, 1986.-

,, 4 6- DeVan, M. J., et. al., " Evaluation of Irradiated Capsule, ,

W-275: 0MAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit' Number 1 Reactor Vessel Materials Irradiation Surveillance  :

Prf ram," DAW-2126, November 1994.

i  ?- NV1 -0020, ' Licensed Operating Roactors Status: Summary ^

< Report, Nuclear Regulatory Commission Monthly Publication, $

September-1973 through October 1993.

7-1 h

r

._ -- , - , .-_ . . . .  ; - ..a . . . - - --- .. - - _ -- - . - . - , . . . . -

~e, e

i

( .

8- Norris, E. B., "

Ef f ect of Thermal Pcwer Averaging Method on ,

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