ML20249C001
| ML20249C001 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 11/30/1997 |
| From: | Ford C, Schleicher T VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML18152B752 | List: |
| References | |
| VEP-NAF-3, NUDOCS 9806250158 | |
| Download: ML20249C001 (94) | |
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Attachment 3 REACTOR VESSEL FLUENCE ANALYSIS METHODOLOGY TOPICAL REPORT (VEP-NAF-3)
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- ; Reactor Vessel Fluence Analysis Methodology
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1: Nuclear Analysis and Fuel
. Nuclear Engineering & Services November,1997
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Reactor Vessel Fluence Analysis Methodology Nuclear Analysis and Fuel
- Nuclear Engineering & Services November,1997
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VIRGINEA POWER
l VEP-NAF-3 Reactor Vessel Fluence Analysis Methodology By C. A. Ford and T. W. Schleicher Nuclear Analysis and Fuel Department Nuclear Engineering Services Virginia Electric and Power Company Richmond, Virginia November,1997 Recommended for Approval:
~PM D. Dziadosf f Supervisor, Nuclear Core Design Approved:
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! K. L. Iiasehore I Manager, Nuclear Analysis and Fuel s
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l CLASSIFICATION /DISCLAMER
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The data, information, analytical techniques, and conclusions in this report have i
l been prepared solely for use by the Virginia Electric and Power Company (the Company), and they may not be appropriate for use in situations other than those for which they were specifically prepared. The Company therefore makes no claim or warranty whatsoever, express or implied, as to their accuracy, usefulness, or applicability.
In particular, THE COMPANY MAKES NO WARRANTY OF MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USEAGE OF TRADE, with respect to this report or any of the data, information, analytical techniques, or conclusions in it. By making this report available, the Company does not authorize its use by others, and any such use is expressly forbidden except with the prior written approval of the Company. Any such written approval shall itself be deemed to incorporate the disclaimers ofliability and disclaimers of warranties provided herein. In no event shall '
the Company be liable, under any legal theory whatsoever (whether contract, tort, warranty, or strict or absolute liability), for any property damage, mental or physical injury or death, loss of use of property, or other damage resulting from or arising out of the use, authorized or unauthorized, of this report or the data, information, and analytical techniques, or conclusions in it.
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ABSTRACT A methodology has been developed for calculating the neutron fluence through a reactor pressure vessel. This methodology includes both a discrete ordinates model and a Monte Carlo model. The discrete ordinates model uses a flux-synthesis technique for calculating a three-dimensional flux distribution, while the Monte Carlo model directly solves the transport equations in three dimensions. Both models use cross sections derived from ENDF/B-VI and both models use neutron source distributions derived from three-dimensional PDQV2 calculated power distributions.
The models were validated by comparisons with the Pool Critical Assembly Benchmark, Configuration 12/13; in-vessel surveillance capsule measurements, and ex-vessel cavity dosimetry measurements. In all cases the calculations and measurements agreed within 20% (lc).
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TABLE OF CONTENTS l
Page CLAS SIFICATION/DI SCLAIMER ........................................................ .............. ..... i ABSTRACT..............................................................................................................iii TA B L E OF CONTENTS .. .. .... .. .. ... . .... ......... .... .. .. .. .. .... . ..... ... .... ... ....... ......... . . .... ... ... . .. .v LI ST O F TAB L ES . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . .. . . . .. . . . . . ... . . . . . . . . . . . . vii LI ST O F FI G U RES . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .. . . . .. .. . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .... . . . . . . . i x S ECTI ON 1 - INTRODUCTI ON . . . ..... ... ... ... ....... ....... ..... . .... ....... . .. .. .. ......... ... .. .... .. ..... 1 SECTION 2 -MODEL DESCRIPTION 2.1 I n trod uc ti o n . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . .. . . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 5 2.2 Cal culation S eq uences .... .. ... .. .. ..... . ... . . . .. . .... .. .. ... .. . .. .. . .... . . . . .. .. .... .. .. .. .. . . ... 5 2.3 M odel G eometry. . ... . . . . . .. ... . . . . .... . .. . ... .. . . .. . .. ..... .. .. .. . .. .. .. . . . .. . . . . . ...... . .. .... . ... 6 2.4 Material Compositions ............................................. ... .............. ............ 16 2.5 Cross Section Generation ....................................................................... 16 2.6 Neutron S ourc es ... .... .... ... . . .... . . .. ..... . .. .. ........... . . .. . ... .. ... . ... . ..... . .. . ..... .... .... I 7 2.7 Other Modeling Parameters............................................................. ....... I 8 SECTION 3 - BENCHMARKING 3 .1 I ntrod u c t i o n . . . .. .. . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . 2 1 3 .2 PC A Experiment .. .. ..... .. .. . ... . .. . .. .......... .. . .... .. . . . . . . . ... . ... .... .... . ... ... ..... . . . .... 21 3.3 Surveillance Capsules.................................................................. .......... 2 8 3.4 Ex-vessel cavity Dosimetry............................. .................................. ... 3 9 SECTION 4 - UNCERTAINTY ESTIMATES 4.1 Analytical Uncertainty Analysis ............................................................ 47 4.2 Comparisons with Benchmark Results..... ............................................ 49 l SECTION 5 -
SUMMARY
AND CONCLUSIONS ..... ........................................... 53 S ECTI ON 6 - REFERENC ES . ..... ....... .... ..... .. ... .... ....... .. .. ... . .... ..... .... ......... . ..... . .. .. 5 5 APPENDIX 1 -
SUMMARY
OF COMPlJANCE WITH DRAFT REGULATORY GUIDE DG-10$3 i
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LIST OF TABLES Table Title Page 1 Measured PCA Reaction Rates . ........ . ....... ........ .. .. ....... ... ... . .... ..... .. 25 2 PCA Reaction Rates Calculated with the DORT Code...... . . . .. ........ . . .. 25 3 PCA C/M Ratios Using the DORT Code.. . ...................................25 4 PCA Reaction Rates Calculated with the MCNP Code ... ... . ....... .. ... .. .... 26 5 PCA C/M Ratios Using the MCNP Code ....... ...... ... ....... .. . ..................... 26 6 Comparison of Calculated and Measured Dosimeter Reaction Rates... ..... . 34 7 Normalized Calculated and Measured Dosimeter Reaction Rates....... .. . 35 8 Comparison of Calculated and Measured Dosimeter Reaction Rates after A dj ustments...... .. ... ... . .... .......... ... .. ...... . .... .. .. ..... .. . ... . ... . 3 6 9 Calculated Reaction Rates for Dosimeter SI-T,15' Location, and Irradiated in Cycle S 1 C 1 .... .... .... .. .... .... .... . . ... . .. ... ........ .. ........ ...... .. .. 3 7 10 Calculated Reaction Rates for Dosimeter SI-W,35 Location, and Irradiated in Cycles S 1 C 1 -S 1 C4 ... ......... . . ... .......... .......... ... . .... ... .... 3 7 11 Calculated Reaction Rates for Dosimeter SI-V,15' Location, and Irradiated in Cycles S 1 C 1 -S 1 C8 ................ ............... . . . ................... .. .... 3 8 12 Cavity Dosimetry Results for the Multi-Element Capsules..... ............. ... . 42 54 13 Comparison of Fe Results at the O' Location........ ...... .. . . ..... ......... . . 43
' 5s 14 Comparison of Ni Results at the 0 Location ...... ........ . ................. .. ...... 44 5d 15 Comparison of Fe Results at the 45* Location. ... . ... . .......... . .... ........ 45 t
16 Comparison of5sNi Results at the 45 Location ...... ......... ...... ... ........... .46 17 Sources of Uncertainty in the Calculated Flux at the Vessel Inner S urface . . .. ..... .. . . . ..... . .. ..... ... . .... . .. ... . . . . . . . .. .... . ..... 5 0 18 Reactor Vessel Flux Aggregate Uncertainties for Various Models and Vessel locations ........... ...... . ......................51 l
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LIST OF FIGURES Figure Title Page l
1 Calculational Sequence for the DORT Model ............................................... 9 2 Calculational Sequence for the MCNP Model.............................................10 3 Reactor Geometry Showing a 45' R,0 Sector............................. .............. .11 4 Axial Reactor Geometry ..... ... .. ... .... .... .. ..... ... ...... .. . . .. . . . . . ..... . . . .. . ... .. .. ... . .... .. 12 5 Internal Surveillance Capsule Geometry ......... .. .........................................I 3 6 DORT R0 Model Region Boundaries......................................... ................. I 4 7 Typical DORT R0 Model of a Surveillance Capsule ................. ................15 8 Pool Critical Assembly Benchmark, Configuration 12/13, Plan View ................................................ ................. 2 7 9 0* R0 Cross-Section of the Ex-Vessel Dosimetry and Neutron Shield Tank Detector Well.................... .. . ........ ............. ............. 41 i
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SECTION 1 -Introduction The purpose of this report is to document the reactor vessel fluence analysis methodology developed by Virginia Electric and Power Company (Virginia Power).
This methodology can be used to calculate the neutron fluence to which the North Anna and Surry reactor vessels have been exposed.
The reactor vessel neutron fluence is determined in two steps. First, the neutron flux (or fluence rate) at the reactor vessci is calculated for each cycle assuming nominal full power operating conditions. The cycle-specific neutron fluxes are then multiplied by the effective full power days (EFPD) for the associated cycle and the results are summed to determine the total neutron fluence. When projecting the neutron fluence for End-of-License (EOL) operation, an appropriate capacity factor is used to reflect expected operating strategies.
This methodology uses two models for neutron flux calculations; one using discrete ordinates techniques and one using Monte Carlo techniques. The discrete 6 2 ordinates model uses the industry-standard computer codes DORT , GIP", DOTSOR ,
and CELL 2" in addition to Virginia Power's FEDIT, ZEDIT, GEOM, GPLOT, and 7
SYNT11 computer codes. The Monte Carlo model uses the Los Alamos MCNP l
computer code in addition to Virginia Power's FEDIT and MCNPSRC computer codes. j i
Both models derive the spatial neutron source distribution from calculations performed with the PDQ Two Zone model". The vessel fluence models use cross section libraries derived from ENDF/B-VI data.
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The discrete ordinates model uses a flux-synthesis technique to combine one- and two-dimensional calculations into three-dimensional flux distribution. This technique provides detailed information regarding the spatial and energy distribution of the neutron flux through the model geometry. The applicability of the flux-synthesis technique is )
l limited to source distributions where the assembly axial power distributions are reasonably consistent around the core periphery.
l The design of Surry Unit 1, Cycles 13,14, and 15 included part-length hafnium flux suppression inserts to protect the reactor vessel welds from neutron embrittlement.
The variation in peripheral assembly axial flux shapes caused by the flux suppression I
inserts complicates the use of the flux synthesis technique and the discrete ordinates model. In these instances, the three-dimensional Monte Carlo model can be used.
Although the Monte Carlo model does not have the synthesis limitations inherent in the discrete ordinates model, the Monte Carlo model only provides flux information at specific, pre-determined locations within the model geometry.
Appendix A compares the models described in this report to the provisions contained in Draft Regulatory Guide DG-10531. It should be noted that the sections of Reference I that concern dosimetry methods are not applicable here because this report discusses only calculational methods. Additionally, the draft regulatory guide includes provisions for validating fluence calculation methods to a set of proposed calculational i benchmark problems. liowever, the calculational benchmark problems had not yet been published when Virginia Power's fluence methodology was being benchmarked. It is
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Virginia Power's fluence methodology folbwing final publication of the regulatory guide.
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l SECTION 2 - Model Description 1 2.1 Introduction 1
The Virginia Power fluence methodology incorporates two distinct models for neutron flux calculations. The first model is a multi-group two-dimensional discrete-ordinates transport model that synthesizes three-dimensional fluxes from calculations of lower dimensions. This model is used to determine the integral fluxes for core loading patterns that do not have strong axial or azimuthal power distribution variations.
Additionally, the discTie-ordinates model is used to determine the flux spectrum as well as the flux profile as a function of depth into the reactor vessel.
The second model is a continuous-energy three-dimensional Monte Carlo transport model. This model is used to determine the integral fluxes when strong axial or azimuthal power distribution variations exist (i e., an axial zoned hafnium absorber inserted in several peripheral assemblies). This model does not require synthesis approximations or multi-group approximations, but requires significantly more
( computational time.
2.2 Calculation Sequences The calculation sequence for the DORT model is sl.own in Figure 1. The GIP computer code is used to mix the isotopic cross sections for use by the DORT code. The GEOM computer code is used to generate the variable-mesh geometry input for the f
DORT model. The GPLOT computer code is used to verify the adequacy of the geometry input. The power distribution generated by the PDQ Two Zone Model at c . nominal operating conditions is transformed into source distributions for use in the 5
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1 DORT by using the FEDIT, ZEDIT, and DOTSOR computer codes. Finally, the SYNTH ,
l computer code is used to combine (r,0), (r,z), and (r) DORT flux files into a three-dimensional flux distribution.
The calculation sequence for the MCNP model, as shown in Figure 2, is significantly simpler. The power distribution generated by the PDQ Two Zone Model at j nominal operating conditions is transformed into a source distribution for use with the MCNP code using the FEDIT and MCNPSRC computer codes. Generic importance maps used to bias the MCNP calculations are initially generated using an adjoint DORT calculation, and then are adjusted to optimize the MCNP calculation.
The calculated neutron fluxes for each cycle are multiplied by the EFPD for the associated cycle to determine the accumulated fluence for that cycle. The cycle fluence are summed to determine the accumulated vessel fluence at various axial, radial, and l
azimuthal locations. When projecting the neutron fluence for EOL operation, an appropriate neutron flux and capacity factor are used to reflect expected operating strategies. The neutron flux used for projecting EOL operation is taken from a recent l equilibrium cycle. A 90% capacity factor is used when projecting future operation for l
nominal 18 month cycles. Significant changes in the cycle length or capacity factor will require a re-evaluation of the EOL neutron fluences.
2.3 ModelGeometry The azimuthal synmetry of the reactor core, and vessel intemals, and reactor vessel permit the geometry to be modeled as a single octant. Figure 3 shows a typical 45 section that is representative of the North Anna and Surry reactor vessels. Note that all of 6
the surveillance capsules and capsule supports are not located in a single octant, but have been represented in Figure 3 by rotating the locations into a single octant for the purposes of the model. Figure 4 shows the relative placement of the reactor core, the surveillance capsules, and the reactor vessel. Figure 5 includes a more detailed representation of the surveillance capsule geometry.
Discrete Ordinates Model The discrete ordinates model synthesizes (r,0), (r,z), and (r) discrete ordinates calculations into a three-dimensional flux distribution. The discrete ordinates calculations are performed with the DORT computer code. Details of the (r,0) model are shown in Figures 6 and 7. The (r,0) model uses the DORT variable mesh option and includes 127 azimuthal intervals, with up to 134 radial intervals in each azimuthal interval. The large number of azimuthal intervals allows for accurately modeling the fuel assemblies and the in-vessel dosimeters. The largest azimuthal interval is approximately 1, with intervals decreasing in size as necessary to model the geometry. The model regions representing the fuel assemblies reproduce the true physical f> el assembly area to within 0.01%. The total area of the peripheral assemblies is also modeled to within 0.01%. This level of detail, coupled with the maximum radial mesh width of 1.27 cm in the fuel region, is expected to adequately reproduce the pin-wise gradients in the power distribution.
The radial mesh structure varies with each azimuthal mesh, but limits on the 1
radial mesh widths are specified to ensure adequate modeling of the flux gradients. The maximum radial mesh widths are 2 intervals per inch in peripheral assemblies,3 intervals 7
per inch in water, and approximately 1.5 intervals per inch in steel. A coarser mesh of 3 ,
1 inches per radial mesh interval is used in the center of the core.
l The (r,z) model uses a fixed axial mesh structure. The axial mesh spacing varies with axial location with a maximum of approximately 2 inches per mesh interval. Finer axial mesh spacings are used in the reflector regions and at the top and bottom of the fuel where the flux gradients are larger. The mesh spacings are adjusted as necessary to accurately model the physical locations of the upper and lower reflector regions as well as the former plates located between the core barrel and the baffle plates.
The three-dimensional flux synthesis is performed for each energy group using the SYNTH code. The SYNT11 code performs the synthesis using the equation:
i' p(r,0,:, E) = p(r,0, E) p(r,:, E) p(r, E)
This method of synthesis is equivalent to the synthesis method used in the DOTSYN computer code, part of the LEPRICON package'.
Monte Carlo Model The Monte Carlo model is a three-dimensional model covering one octant of the reactor core, reactor internals, axial reflector regions, and reactor vessel. The model is constructed using the combinatorial geometry features of the MCNP computer code.
I Each region boundary is modeled by the union and intersection of various surfaces. This approach permits detailed modeling of such features as assembly / baffle interfaces and in-vessel dosimetry configurations.
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Figure 3: Reactor Geometry Showing a 45' R,0 Sector 7 [ [ N 11
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I l l 2.4 MaterialCompositions l The material compositions for water in the downcomer, baffle area, and fuel I I regions of the DORT model are based on cycle-specific calculations reflecting actual operating conditions. Axial variations in water densities are included in the model. The water densities are also adjusted to include the effects oflow-power fuel assemblies ; loaded in the periphery of the core. I The compositions for stainless steel and pressure vessel steel are based on . 1 nominal design values. The isotopic compositions and densities for the fuel are based on as-built isotopics and subsequent depletion calculations. The fuel is modeled as having either 0 MWD /MTU or 45,000 MWD /MTU burnup. This approximation in determining macroscopic cross sections was shown to have only minor effects on the resulting radiation transport calculations. The error introduced by this approximation is included in the analytic uncertainty analysis. Typical isotopic compositions and densities used in the MCNP model are derived from the DORT compositions to ensure consistency in the models. 1 l 2.5 Cross Section Generation BUGLE-93 cross sections (Reference 3) are used in the DORT discrete-ordinates l model. The BUGLE-93 library, distributed by the Radiation Shielding Information l l Center, was developed from the ENDF/B-VI based VITAMIN-B6 library. The l BUGLE-93 library includes 47 neutron energy groups, with 26 energy groups above i approximately 0.1 MeV. The DORT model used a P-3 Legendre expansion of the i scattering cross sections. The BUGLE-93 cross sections have been benchmarked against 16
i the 199 energy group VITAMIN-B6 library using the PCA Blind Test configuration 4/12. l The results at a simulated surveillance capsule and several locations within the pressure vessel agreed to within 4%. 1 l The MCNP model uses cross sections from the ENDF/B-VI based MCNPDAT6 library distributed by the Radiation Shielding Information Center (Reference 8). This 1 library features a continuous-energy treatment that avoids the approximations inherent in multi-group processing. 2.6 Neutron Sources The spatial source distributions used in the fluence models are derived from three-dimensional power distributions calculated using the PDQ Two Zone Model (Reference 17). Power distributions from at least three burnup steps for a given cycle are averaged to produce a spatially-distributed source which is representative of the source integrated over the entire length of the cycle. This source distribution is then converted to source distributions for R0, RZ, and R DORT calculations using the DOTSOR computer program (Reference 2). The R0 spatial source distribution is representative of the PDQV2 power distribution integrated ( over the axial height of the fuel. The RZ and R spatial source distributions are j representative of the RZ plane at the azimuthal location of the peak vessel fluence. The cycle-average power distribution is converted to a three-dimensional source distribution for use in the MCNP calculations using the MCNPSRC computer code. The resulting MCNP source distribution includes pin-wise three-dimensional source distributions for the peripheral assemblies and quarter-assembly averaged three-l 17
I dimensional source distribution's for the remainder of the core. l ( The magnitude of the total source strength, in neutrons /MW/sec, is determined l { from data generated by the CELL 2 lattice code (References 4 and 5). The total source strength is calculated for each cycle and is calculated based on a weighted average burnup and initiaa enrichment of the peripheral fuel assemblies. 1 l l The source spectrum is chosen as a weighted average of the fission spectra for 235 24i U,2nU, 239Pu,and Pu. Each of the fission spectra weighting factors are based on data generated by the CELL 2 lattice code. The weighting factors are cycle-specific and ; are chos:n based on a weighted average burnup and initial enrichment of the peripheral fuel assemblies. 2.7 OtherModeling Parameters Other miscellaneous modeling parameters are required by the DORT and MCNP models. The DORT model uses an Sa fully symmetric angular quadrature and a P3 Legendre expansion of the scattering cross sections. A theta-weighted difference flux extrapolation model is used with a point flux convergence of 0.001. The P3 Legendre expansion of the scattering cross sections and the theta-weighted difference flux extrapolation model have been found to be acceptable for reactor vessel fluence calculations (Reference 11). The Si angular quadrature is also typical for fluence calculations, but may be too coarse when streaming is an issue. The suitability of using Si angular quadrature in Virginia Power's DORT model was verified by performing a sensitivity study which used an Sa fully symmetric angular quadrature and simultaneously halving the spatial mesh widths in each direction. Similarly, the 18
suitability of using a point flux convergence of 1.0E-3 was verified by rerunning a test case with the convergence reduced to 1.0E-4. Issues such as quadrature, Legendre expansion, and point flux convergence are not applicable to the MCNP Monte Carlo model. Ilowever, the finite size of the MCNP flux tallies can introduce a bias in the calculated flux. A surface flux tally is used at the wetted vessel surface to calculate the peak neutron flux at the vessel inner surface and the peak flux at the beltline welds. This tally is 5 wide, extends the axial height of the model, and is segmented axially into approximately 40 cm segments. The axial segments are positioned to allow the axial variation in the flux to be monitored without introducing any significant bias into the calculated fluxes. The 5' tally width was chosen to improve the tally efficiency while introducing only a minimum bias in the results. Sensitivity studies show that the azimuthal flux variation within the tally region introduces a 2% bias when a localized flux peak occurs within the tally region. This bias is accounted for by increasing the calculated fluxes at 0* by 2%. A surface flux tally is also used at the wetted vessel surface to calculate the j neutron flux at the Surry longitudinal welds. This tally is also 5 wide, extends the axial height of the model, and is segmented axially into approximately 40 cm segments, j l Ilowever, the azimuthal variation of the flux at all axial locations has a minimum at the i longitudinal welds so the calculated tally flux is slightly larger than the flux at the weld location. This bias is conservative and no correction is applied to the calculated results. ] MCNP integral flux tally results must pass several tests prior to being considered statistically valid. First, the tally results must pass the ten statistical checks performed by ; l 19 l
I l 1 1 the MCNP computer code. Passing these ten statistical checks provides additional l confidence in both the estimated mean and the estimated standard deviation. 1 Additionally, the probability density function for each tally is evaluated to determine if a l tally, which may meet the first ten checks, may be biased because an important region of l 1 the problem was not well sampled. Further information regarding these statistical checks l
~
can be found in Reference 7. The statistical uncertainty (one standard deviation) must be less than 10% for each integral flux tally to pass the statistical checks mentioned earlier. Furthermore, the statistical uncertainty of calculated fluxes used for fluence calculations should be less than 5%. A bounding standard deviation of 5% has been included in the statistical i uncertainty analysis performed for this fluence methodology. Integral flux tally standard i deviations in excess of 5% will be added as a bias to the calculated fluences. l l 20
Section 3 - Benchmarking 3.1 Introduction The vessel fluence methodology was benchmarked using a combination of plant-specific surveillance capsules, pressure vessel simulator measurements, and plant-specific l ex-vessel cavity dosimetry measurements. The pressure vessel simulator measurements I provide experimental results with well-known and documented uncertainties, while the l 1 plant-specific surveillance capsule benchmarking includes actual reactor materials, geometry, and operating conditions. The inclusion of the Surry Unit 1, Cycle 13 ex-vessel cavity dosimetry measurements provides additional validation for use of the MCNP model in the presence of the part-length hafnium inserts. l l 3.2 PCA Experiment The Pool Critical Assembly (PCA) Benchmark, Configuration 12/13, is a well-documented pressure vessel simulator experiment (Reference 11). A plan view, published in Reference i1, is shown in Figure 8. The PCA benchmark was modeled using both DORT and MCNP. In each case the cross-sections, modeling parameters, and synthesis techniques used in the vessel fluence models were applied to the PCA benchmark. A one-quarter model was used for both the DORT and MCNP calculations to take advantage of the symmetry found in Figure 8. The DORT calculation used BUGLE-93 cross sections, a 2n U fission spectrum, an S8 angular quadrature, a P3 Legendre expansion, a point convergence criterion of 0.001, and an R0/RZ/R synthesis technique to generate a three-dimensional flux distribution. The RZ model included the full height of the fuel and reflectors to ensure l 21 l
'j
that the slightly asymmetric axial power shape was adequately modeled. The physical dimensions, material compositions, and source distribution information were taken from Reference 11. The DOTSOR code was used to convert the rectangular-mesh source distribution information into R0, RZ, and R source distributions. The SYNTIl code was l used to combine the resulting one- and two-dimensional flux distributioris into a tinee- , 1 dimensional flux distribution. The measured reaction rates are presented in Table 1. These values were taken from Tables 8.2.1,8.3.1, and 7.1.7 of Reference 11. Note that in some instances Reference 11 provided the measured data in terms of equivalent fission fluxes rather than ,1 l reaction rates. Those equivalent fission fluxes were then divided by the fission spectrum cross sections provided in Table 7.1.1 of Reference 11 to derive the reaction rates shown l in Table 1. The fluxes and reaction rates from the DORT calculations are presented in Table 1
- 2. The data is presented in the form of calculation-to-measurement (C/M) ratios in Table
- 3. All locations are relative to the centerline of the core. Note that location Al is between the core and the thermal shield and location A3 is between the thermal shield l and the pressure vessel. Locations A4, A5, and A6 are 1/4,1/2, and 3/4 thickness of the pressure vessel, respectively.
All of the calculated reaction rates are within il2% of the experimental data. The 1 i calculated flux with energies above one MeV shows a consistent positive bias, ranging from +4% to +10%, with respect to experimental data. This is inconsistent with the dosimeters with high-energy reaction thresholds that show a small but definite under-l l 22
1 prediction within the pressure vessel. It is possible that DORT is over-predicting the flux in an energy range above 1 MeV that is not adequately monitored by the dosimetry. Another possible reason for the discrepancy lies in the definition of a " measured" (or experimental) flux above one MeV. This flux was determined by using a least squares fit to fluxes calculated from the various dosimeter reaction rates using some specified dosimeter cross section. This process may have introduced a bias into the experimental flux. Additionally, Reference 11 indicates that the la uncertainty associated with the experimental flux above 1 MeV varies from 7% to 12% with increasing distance from the core. The agreement between calculated and experimental fluxes is generally within the 1 experimental 2a limits. 2n The MCNP calculation used ENDF-B/VI based cross sections, a U fission spectrum, and surface flux tallies located at each of the detector locations. A quadrant-symmetric model was used which included the full axial height of the core and reflectors. Biasing techniques included the use of spatial importance factors, source energy biasing, source location biasing, and an energy cutoff of approximately 0.1 MeV. The results of the MCNP calculations are shown in Table 4. The calculated fluxes and reaction rates include corrections ranging from 1% to 3% to account for spatial flux l variations within the tally regions. Each entry in the table includes a calculated flux or reaction rate and the la statistical uncertainty associated with the calculation. The PCA experiment detectors difTer from Virginia Power's in-vessel dosimetry 2n because the PCA fission detectors were not cadmium shielded. This will cause the Np reaction rates calculated with a 0.1 MeV cutoff to be biased low if a large thermal flux is 23
I l present. This effect is expected to be most pronounced at the Al detector location, between the core and the thermal shield. To verify this, an additional calculation was 237 performed for detector location A1 without the 0.1 MeV cutoff. The Np reaction rate increased by 9.7%. This effect is not present in the analysis of power reactor in-vessel 237 l dosimetry because the Np dosimeters are shielded with a cadmium cover. Table 5 shows the ratios of calculated and experimental results after including the bias adjustment for non-zero tally sizes. Note that, as expected, the calculated full-spectrum 237 Np reaction rate at detector location Al agrees much better with the l experimental results. Additionally, all of the calculated reaction rates are within i10% of the experimental data. This is equivalent to or better than earlier Oak Ridge National Laboratory (ORNL) studies, which have shown differences among measurements and calculations between approximately 10% and 20% (Reference 18). The calculated flux with energies above one MeV shows a consistent positive bias, ranging from +7% to +14%, with respect to experimental data. This is generally I consistent with the DORT results and is within the experimental 2o limits. I I I I I 24 I 'I
I Table 1: Measured PCA Reaction Rates Detector Flux >1.0 MeV Flux >0.1 MeV "Ni *'Al '"in **U N p A1 3.71 E-06 6.94E-06 6.35E-07 5.59E-09 1.06E-06 8.71E-06 A3 1.33E-07 2.49E-07 2.52E-08 3.18E-10 3.76E-08 2.98E-07 A4 4.30E-08 1.39E-07 5.78E-09 7.24E-11 1.11E-08 d. 86E-08 1.22E-07 A5 2.07E-08 9.35E-08 2.28E-09 2.91E-11 5.22E-09 8.36E-09 6.80E-08 A6 9.11 E-09 5.57E-08 8.10E-10 1.09E-11 2.21E-09 3.42E-09 3.54E-08 Table 2: PCA Reaction Rates Calculated with the DORT Code Detector Flux >1.0 MeV Flux >0.1 MeV "Ni 2'Al '"in '"U 28'Np A1 3.93E-06 7.05E-06 6.35E-07 5.65E-09 1.02E-06 1.74E-06 8.49E-06 A3 1.43E-07 2.57E-07 2.54E-08 3.20E-10 3.74E-08 6.52E-08 3.14E-07 A4 4.73E-08 1.48E-07 5.65E-09 7.00E-11 1.11E-08 1.79E-08 1.21E-07 AS 2.21E-08 9.75E-08 2.19E-09 2.75E-11 4.99E-09 7.62E-09 6.63E-08 A6 9.44E-09 5.55E-08 7.93E-10 1.03E-11 2.09E-09 3.00E-09 3.29E-08 Table 3: PCA C/M Ratios Using the DORT Code Detector Location Flux >1.0 Flux >0.1 "Ni 27 Al '"in 8"U 28'Np (cm) MeV MeV A1 32.57 1.06 1.02 1.00 1.01 0.96 0.97 A3 50.27 1.08 1.03 1.01 1.01 0.99 1.05 A4 60.07 1.10 1.06 0.98 0.97 1.00 0.96 0.99 l A5 65.27 1.07 1.04 0.96 0.95 0.96 0.91 0.98 l A6 70.67 1.04 1.00 0.98 0.94 0.95 0.88 0.93 j l
\
)
l l 1 i l 25 1
Table 4: PCA Reaction Rates Calculated with the MCNP Code
**7 Detector Flux >1.0 MeV Flux >0.1 MeV "Ni *'Al "'In 8'* U Np A1 3.973E 6 7.131E-6 6.337E-7 5.630E-9 1.030E-6 1.767E-6 7.913E-6 (0.9%) (0.8%) (1.1%) (1.6%) (0.9%) (0.9%) (0.8%)
A1 (see 8.678E-6 Note 1) (1.3%) : A3 1.464E-7 2.623E-7 2.583E-8 3.245E-10 3.829E-8 6.664E-8 2.929E-7 i (0.9%) (0.8%) (1.1%) (1.5%) (1.0%) (1.0%) (0.9%) A4 4.895E-8 1.381E-7 5.743E-9 7.058E-11 1.137E-8 1.829E-8 1.206E-7 (1.1%) (0.8%) (0.9%) (1.2%) (0.7%) (1.2%) (0.6%) A5 2.285E-8 8.865E-8 2.229E-9 2.784E-11 5.125E-9 7.787E-9 6.585E-8 (1.1%) (0.8%) (1.3%) (1.7%) (1.1%) (1.2%) (0.9%) A6 9.895E-9 5.164E-8 8.123E-101.067E-11 2.185E-9 3.142E-9 3.342E-8 (1.0%) (0.6%) (1.7%) (1.4%) (0.9%) (1.2%) (0.7%) 237 Note 1: The Np reaction rate was recalculated with an energy cutoff ofless than 0.01 eV. Table 5: PCA C/M Ratios Using the MCNP Code Detector Location Flux >1.0 Flux >0.1 "Ni Al "'In 22sU
7 Np (cm) MeV MeV A1 32.57 1.07 1.03 1.00 1.01 0.97 0.91 A1 (see 32.57 1.00 Note 1)
A3 50.27 1.10 1.05 1.03 1.02 1.02 0,98 A4 60.08 1.14 0.99 0.99 0.97 1.02 0.98 0.99 A5 65.28 1.10 0.95 0.98 0.96 0.98 0.93 0.97 A6 70.67 1.09 0.93 1.00 0.98 0.99 'O.92 0.94 237 Note 1: The Np reaction rate was recalculated with an energy cutoff ofless than 0.01 eV. 26
Figure 8: Pool Critical Assembly Benchmark, Configuration 12/13, Plan View 3 h
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3.3 Surveillance Capsules Virginia Power's vessel fluence methodology was also benchmarked against a set ofin-vessel surveillance capsule measurements made at the Surry and North Anna reactors. Although each capsule is documented in a separate evaluation report, the reaction rates used here were taken from a summary report, Reference 10. Calculations for the surveillance capsule reaction rates consisted of performing i individual DORT and MCNP calculations for each cycle, using cycle-specific power , distributions and material properties in accordance with the methodology described in j Section 2. 1 l It should be noted that the measured reaction rates have been determined relative l l to a predetermined reference power level for each plant. The reference power level used by Westinghouse for North Anna was 2893 MWth, while the reference power level for l Surry was 2441 MWth. All of the Surry surveillance capsules included in this analysis were removed prior to uprating the Surry units to 2546 MWth. However, all of the North I Anna Unit 2 calculations and the first five cycles of North Anna I were performed herein using an eighth core power consistent with 2775 MWth. Therefore, the results from those cycles were multiplied by 2893/2775 for an accurate comparison with the measured data. DORT Calculations Table 6 includes the measured reaction rates from Reference 11, the DORT l calculated reaction rates, and a ratio of the calculated and measu ed reaction rates (C/M). Analysis of the C/M values shows a mean C/M of 1.02 with a standard deviation of 12%. However, several of the individual C/M values exceed 20%. The most noticeable 28
deviations involve all of the Cu dosimeters removed from Surry Unit 1. The C/M values for all of these dosimeters are approximately 0.76, including the Cu dosimeter removed after SIC 1. This dosimeter is significant because the S1C1 core was very similar to the S2Cl core, and the copper dosimeter from S2-X (removed after SICl) does 6 not show a C/M similar to the Surry I data. This indicates that the Surry 1 'Cu measured data may be suspect. Further investigation entailed normalizing the dosimeter values to the "Fe response, both for the measured values as well as the calculated values. Normalizing in this fashion emphasizes inconsistencies due to spectrum behavior as well as measurement error. The normalization removes most of the errors (such as power distribution errors) which would equally affect all of the dosimetry material in a capsule. Table 7 shows the normalized results. The normalized results should be reasonably consistent for each dosimeter material at a particular azimuthal location, independent ofirradiation history or operating unit. 63 Note that while most of the Cu normalized calculated reaction rates are reasonably well grouped, the Surry Unit I measured data are consistently higher. Reference 11 indicates that an average normalized value for three loop thermal-shield Westinghouse plants is 9.93E-3 i8.4% at 15 degrees and 1.061E-2 1.7% at 25%. No i l average data is available at 35 degrees because the Suny Unit I capsule was the only I ) capsule removed from a 35 degree location at the time of the report. I 1 l The normalized calculated6'Cu reaction rates are generally within I standard l ) deviation of the average Westinghouse values for all of the capsules from the Surry and 29
North Anna units. Ilowever, all of the normalized measured Surry 1 Cu reactions rates are greater than I standard deviation away from the Westinghouse averages. It appears that the measured63Cu reaction rates for Surry Unit 1 may be biased high by approximately 10 to 15 percent. Note that this was first identified in the capsule report . for the SI-T surveillance capsule (Reference 14). However, because the magnitude of the 6 bias is only approximately known, the Surry Unit 1 'Cu data will be disregarded. i Another dosimeter showing a large deviation between measured and calculated 237 reaction rates is the Np dosimeter in the Surry Unit 2 W capsule, removed after four cycles. The C/M value of 1.28 is the largest of the ratios. Note, however, that the 237 normalized measured reaction rate for this Np dosimeter is markedly different from the other capsule data. Reference 11 indicates that the average normalized measured reaction rate at 25 degrees in a Westinghouse three-loop thermal-shield plant is 36 i4.4%. This dosimeter had a normalized measured reaction rate of 23.2, or 3 standard deviations from the average. In addition, the remaining dosimeters in capsule W are much more consistent with the respective Westinghouse averages. Therefore, the Suny Unit 2, 237 Capsule W Np dosimeter will be disregarded. 23: Finally the U dosimeters from many of the capsules have values of C/M greater i than one, with two of the dosimeters having C/M values greater than 1.20. It should be 23s l noted that the U measured data have been subjected to larger adjustment factors than have the other dosimeters. Most of the dosimeters are adjusted in Reference 11 only to account for radial flux variations, while the 23s U dosimeters have additional reductions of 239 l approximately 4.8% for photofission and between 12% and 19% for 235 U and Pu 1 30
237 j contributions. (Other corrections included 1.85% for photofission in Np and 2.3% for i 23s flux variations in nickel). Note that the sum of these enrrections for U, between 16.8% and 23.8%, represents an adjustment significantly larger than the reported standard deviations of the normalized 23sU reaction rate for Westinghouse plants (approximately ) 6.2% in three-loop thermal shield plants). Additionally, Reference 12 indicates that 23s 239 because the U dosimeters were cadmium shielded, the contribution from Pu is l expected to be insignificant (<2%). Ilowever, Reference 13 indicates that a correction for 235 235 U contributions is necessary when the U concentration is greater than 40 ppm. 235 23s Reference 11 specifies the U concentration in the U dosimeters as 300 ppm. 23: 239 Therefore, the 11 reaction rates were adjusted to remove the Pu fission correction while leaving the other corrections intact. Table 8 shows the comparison of calculated and measured reaction rates after excluding the questionable measurements and using the 1 modified 23sU measured data. Note that the 23sU data are now in better agreement. Averaging the C/M values shown in Table 8, the overall C/M for the calculational model is 1.01, indicating a very low bias. In addition, the standard deviation of the C/M values is 9.05%, significantly less than the recommended maximum of 20%. However, it should be noted that the modifications to the measured reaction rates did not have a large I effect on these statistics - the average C/M using the original measured data was 1.02 with a standard deviation of 12%. In addition, the mean C/M values for each of the } ) surveillance capsules ranged from 0.88 to 1.12. i I In summary, the DORT model produces agreement between measured and calculated dosimeter reaction rates that are within the expected range of approximately 31 l
t i20%. The average C/M for each surveillance capsule was within il2%. These C/M values are larger than were seen in the PCA evaluation, reflecting more uncertainty in the source distribution and the dosimetry materials. However, the agreement between measured and calculated dosimeter reaction rates is viithin the desired range of 20% (Reference 1). f i MCNP Calculations MCNP Calculations were performed only for a limited number of surveillance l capsules, primarily to show that the MCNP model closely matches the results produced ; by the DORT model. This verifies that the uncertainties in the calculations are not caused 4 by the differences between the Monte Carlo and discrete ordinate methods, but rather i d from the model input parameters such as power distributions, material properties, and ENDF/B-VI cross section data. Surry Unit I surveillance capsules were also chosen for the MCNP benchmarking because:
. the Surry Unit I capsules experienced irradiation periods ranging from one to eight i
cycles and included high leakage as well as low leakage loading patterns, e the MCNP model will be used to calculate vessel fluxes for Surry Unit I when part-length hafnium flux suppression inserts are used, and i e evaluating Surry Unit I capsules would provide additional analysis of the possible bias in the Cu dosimeters. , Tables 9 through 11 show the MCNP calculated reaction rates for the three Surry Unit I surveillance capsules. Also shown are comparisons with the earlier DORT calculated reaction rates and the measured reaction rates. Note that for most of the < 32
dosimeters the MCNP results agreed with the DORT results to within 2 standard deviations. The MCNP results for the 35 dosimeter may indicate a small negative bias (~5%) relative to the DORT results. Ilowever, the average percent difference for the capsule (excluding the Cu dosimeter as discussed earlier) is -16.8%, within the desired range of120%. l l l 33
Table 6: Comparison of Calculated and Measured Dosimeter Reaction Rates Capsule Azimuthal Operating "Cu "Fe "Ni "U "Np Angle Cycles f (Degrees) N1-V(calc.) 15 1 6.071E-17 6.601E-15 8.998E-15 3.084E-14 2.205E-13 , (meas.) 5.670E-17 5.830E-15 7.980E-15 2.580E-14 1.910E-13 ( C/M 1.07 1.13 1.13 1.20 1.15 N 1-U(calc.) 25 1-6 4.135E-17 4.116E-15 5.558E-15 1.807E-14 1.2133E-13 l (meas.) 3.790E-17 3.580E-15 5.330E-15 1.580E-14 1.270E-13 C/M 1.09 1.15 1.04 1.14 0.96 N2-V(calc.) 15 1 6.001E-17 6.530E-15 8.902E-15 3.052E-14 2.183E-13 (meas.) 5.800E-17 5.860E-15 8.120E-15 2.450E-14 1.990E-13 , C/M 1.03 1.11 1.10 1.25 1.10 N2-U(calc.) 25 1-6 4.194E-17 4.191E-15 5.661E-15 1.844E-14 1.2397E-13 (meas.) 4.040E-17 3.860E-15 5.410E-15 1.410E-14 1.350E-13 C/M 1.04 1.09 1.05 1.31 0.92 SI-T(calc.) 15, 1 5.190E-17 5.596E-15 7.617E-15 2.590E-14 1.832E-13 i (meas.) 6.940E-17 6.180E-15 8.050E-15 2.560E-14 1.880E-13 C/M 0.75 0.91 0.95 1.01 0.97 SI W(calc.) 35 1-4 3.026E-17 2.894E-15 3.887E-15 1.225E-14 7.958E-14 (meas.) 3.970E-17 3.130E-15 4.610E-15 N/A N/A C/M 0.76 0.92 0.84 N/A N/A SI-V(calc.) 15 1-8 4.956E-17 5.278E-15 7.178E-15 2.428E-14 1.711E-13 (meas.) 6.420E-17 5.830E-15 7f)0E-15 2.380E-14 1.900E-13 C/M 0.77 0.91 0.97 1.02 0.90 S2-X(calc.) 15 1 5.264E-17 5.678E-15 7.728E-15 2.628E-14 1.8599E-13 (meas.) 5.810E-17 6.380E-15 7.890E-15 2.680E-14 1.830E-13 C/M 0.91 0.89 0.98 0.98 1.02 S2-W(calc.) 25 1-4 4.179E-17 4.239E-15 5.726E-15 1.869E-14 1.2511E-13 (meas.) 4.480E-17 4.210E-15 5.800E-15 1670E-14 9.750E-14 C/M 0.93 1.01 0.99 1.12 1.28 S2-V(cale.) 15 1-8 4.969E-17 5.293E-15 7.198E-15 2.434E-14 1.7152E-13 (meas.) 4.650E-17 5.260E-15 6.600E-15 2.400E-14 1.740E-13 C/M 1.07 1.01 1.09 1.01 0.99 i 34
i Table 7: Normalized Calculated and Measured Dosimeter Reaction Rates Capsule Azimuthal Operating "Cu "Fe "Ni "U '"Np r Angle Cycles (Degrees) N1-V(calc.) 15 1 9.20E-03 1.00 1.36 4.67 33.4 i (meas.) 9.73E-03 1.00 1.37 4.43 32.8 N1-U(calc.) 25 1-6 1.00E-02 1.00 1.35 4.39 29.5 (meas.) 1.06E-02 1.00 1.49 4.41 35.5 N2-V(calc.) 15 1 9.19E-03 1.00 1.36 4.67 33.4 (meas.) 9.90E-03 1.00 1.39 4.18 34.0 N2-U(cale.) 25 1-6 1.00E-02 1.00 1.35 4.40 29.6 (meas.) 1.05E-02 1.00 1.40 3.65 35.0 SI-T(calc.) 15 1 9.27E-03 1.00 1.36 4.63 32.7 (meas.) 1.12E-02 1.00 1.30 4.14 30.4 SI-W(calc.) 35 1-4 1.05E-02 1.00 1.34 4.23 27.5 (meas.) 1.27E-02 1.00 1.47 N/A N/A SI-V(calc.) 15 1-8 9.39E-03 1.00 1.36 4.60 32.4 (meas.) 1.10E-02 1.00 1.27 4.08 32.6 S2-X(calc.) 15 1 9.27E-03 1.00 1.36 4.63 32.8 (meas.) 9.11E-03 1.00 1.24 4.20 28.7 S2-W(calc.) 25 1-4 9.86E-03 1.00 1.35 4.41 29.5 (meas.) 1.06E-02 1.00 1.38 3.97 23.2 S2-V(calc.) 15 1-8 9.39E-03 1.00 1.36 4.60 32.4 ! (meas.) 8.84E-03 1.00 1.25 4.56 33.1 1 ) 35 1
I Table 8: Comparison of Calculated and Measured Dosimeter Reaction Rates after Adjustments Capsule Azimuthal Operating "Cu "Fe "Ni "'U "'Np g Angle Cycles 3 (Degrees) N1-V(cr.lc.) 15 1 6.071E-17 6.601E-15 8.998E-15 3.084E-14 2.205E-13 g (meas.) 5.670E-17 5.830E-15 7.980E-15 2.772E-14 1.910E-13 5 C/M 1.07 1.13 1.13 1.11 1.15 N1-U(calc.) 25 1-6 4.135E-17 4.116E-15 5.558E-15 1.807E-14 1.213E-13 (meas.) 3.790E-17 3.580E-15 5.330E-15 1.748E-14 1.270E-13 C/M 1.09 1.15 1.04 1.03 0.96 N2-V(calc.) 15 6.001E-17 6.530E-15 8.902E-15 3.052E-14 2.183E-13 1 5.800E-17 5.860E-15 8.120E-15 2.630E-14 1.990E-13 l (meas.) C/M 1.03 1.11 1.10 1.16 1.10 N2-U(calc.) 25 1-6 4.194E-17 4.191E-15 5.661E-15 1.844E-14 1.240E-13 (meas.) 4.040E-17 3.860E-15 5.410E-15 1.57]E-14 1.350E-13 C/M 1.04 1.09 1.05 1.17 0.92 g SI-T(calc.) 15 1 5.596E-15 7.617E-15 2.590E-14 1.832E-13 5 (meas.) 6.180E-15 8.050E-15 2.748E-14 1.880E-13 C/M 0.91 0.95 0.94 0.97 S I-W(calc.) 35 1-4 2.894E-15 3.887E-15 1.225E-14 7.958E-14 (meas.) 3.130E-15 4.610E-15 N/A N/A C/M 0.92 0.84 N/A N/A SI-V(calc.) 15 1-8 5.278E-15 7.178E-15 2.428E-14 1.711E-13 (meas.) 5.830E-15 7.400E-15 2.753E-14 1.900E-13 C/M 0.91 0.97 0.88 0.90 S2-X(cale.) 15 1 5.264E-17 5.678E-15 7.728E-15 2.628E-14 1.860E-13 (meas.) 5.810E-17 6.380E-15 7.890E-15 2.876E-14 1.830E-13 g C/M 0.91 0.89 0.98 0.91 1.02 3 S2-W(cale.) 25 1-4 4.179E-17 4.239E-15 5.726E-15 1.869E-14 (meas.) 4.480E-17 4.210E-15 5.800E-15 1.822E-14 C/M 0.93 1.01 0.99 1.03 S2-V(calc.) 15 1-8 4.969E-17 5.293E-15 7.198E-15 2.434E-14 1.715E-13 (meas.) C/M 4.650E-17 5.260E-15 6.600E-15 2.749E-14 1.N0E-13 1.07 1.01 1.09 0.89 0.99 ll l l I; 1 Il l 36 L_._._________ _ I
Table 9: Calculated her.ction Rates for Dosimeter SI-T,15 Location, and Irradiated in Cycle SICI
""U Cu "*Ni "'Np "Fe l MCNP 2.568E-14 5.140E-17 7.595E-15 1.788E-13 5.582E-15 3Mt to 2.48 % 3.01% 2.81 % 1.82 % 2.99 %
DORT 2.59E-14 5.19E-17 7.62E-15 1.83E-13 5.60E-15
% Diff -0.85% -0.96% -0.33% -2.40% -3.21 %
(MCNP vs. DORT) Measured 2.75E-14 6.94E-17 8.08E-15 1.88E-13 6.18E-15
% Diff -6.62% -25.94 % -6.00% -5.00% -9.68%
l (MCNP vs. I Measured) l l Table 10: Calculated Reaction Rates for Dosimeter SI-W,35" Location, and l Irradiated in Cycles SICl-sic 4 l ""U "Cu "*Ni "'Np "Fe MCNP 1.17E-14 2.88E-17 3.66E-15 7.61 E-14 2.72E-15 l MCNPle 1.68 % 1.98 % 2.39 % 1.20 % 2.56 % l DORT 1.22E-14 3.03E-17 3.89E-15 7.96E-14 2.89E-15
% Diff -4.10% -4.95% -5.91% -4.40% -5.88%
(MCNP vs. DORT) Measured N/A 3.97E-17 4.61 E-15 N/A 3.13 E-15
% Diff N/A -27.46 % -20.61 % N/A -13.01 %
(MCNP vs. Measured) 37
Table 11: Calculated Reaction Rates for Dosimeter SI-V,15' Location, and ' l Irradiated in Cycles SICl-S1C8 "U "Cu "Ni "'Np "Fe ., MCNP 2.47E-14 4.95E-17 7.33E-15 1.69E-13 5.40E-15 f MCNPlo 1.26% 1.70 % 1.80% 0.87 % 1.94 % DORT 2.43E-14 4.96E-17 7.18E-15 1.71E-13 5.28E-15 f
% Diff 1.65 % -0.20% 2.09 % -1.17% 2.27 %
(MCNP vs. DORT) Measured 2.75E-14 6.42E-17 7.40E-15 1.90E-13 5.28E-15 I i l
% DifT -10.18 % -22.90 % -0.95% -11.05 % +2.27%
(MCNP vs. l Measured) I 4 1 l t I I ( 1 38
r 3.4 Ex-vessel cavity Dosimetry Ex-vessel cavity dosimetry was installed external to the Surry Unit I reactor prior to the start of Cycle 13. The dosimetry was irradiated at two ex-vessel positions,0* and 45*, and was removed for evaluation after Cycle 13. Multi-element dosimetry samples were contained in cadmium or aluminum capsules and included wires of 237 23s cobalt / aluminum, iron, copper, niobium, nickel, and spheres of Np and U. Lengths i of stainless steel wire were used to provide reaction rates along the axial length of the dosimetry assembly. The irradiation of the ex-vessel cavity dosimetry was modeled using the MCNP computer code. The MCNP code was chosen because the Cycle 13 power distribution > was axially asymmetric due to the use of part length hafnium flux suppression inserts in the peripheral assemblies at 0 and 45 . The MCNP cavity model is an extension of the in-vessel model developed for vessel fluence calculations. The cavity model extends the in-vessel model by including the vessel lagging, the water-filled neutron shield tank surrounding the vessel, and the ex-vessel cavity dosimetry. A large number of features of the neutron shield tank were incorporated into the model. These included the outer steel walls, the internal support plates, the neutron ) shield tank vent shrouds, and the ex-vessel plant nuclear instrumentation assemblies. The modeling features used in the cavity analysis were identical to those used for j in-vessel fluence analysis. These included the fission spectrum, material properties, l physical dimensions, and source distribution, and the variance reduction techniques. The I y axial dimension of the model encompassed the active fuel height and sufficient reflector ) 39 )
on either end of the core to adequately model ex-vessel fluxes in axial planes i corresponding to the fuel. Additionally, the ex-vessel cavity dosimetry model included material and geometry modeling for the reactor vessel insulation, neutron shield tank, and i water within the neutron shield tank. The densities used for the steel components were j l based on design specifications. The density of the water in the neutron shield tank was . l specified to reflect normal operating conditions. The tally regions used in this evaluation corresponded to the volume of the I I dosimetry string and the dosimetry guide tubes at O' and 45 . Figure 9 shows the geometry of the 0* neutron shield tank detector well at the core midplane, including the . ex-core neutron instrumentation and the cavity dosimetry. Table 12 shows a comparison of the measured and calculated reaction rates for the multi-clement capsules at O' and 45". The calculations are within 20% of the meamrements, with the iron and nickel dosimeters showing the largest average l difference. Tables 13 through 16 show comparisons of the measured and calculated reaction rates for "Fe and 5sNi in the dosimeter wires. These values also agree within 20% and continue to show a generally positive bias in the calculated values. 1 l l i i 40
Figure 9: 0' R0 Cross-Section of the Ex-Vessel Dosimetry and Neutron Shield Tank Detector Well e . J \
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l Table 12: Cavity Dosimetry Results for the Multi-Element Capsules ; Target Azimuthal Axial Measured Predicted Difference Material Angle Position Reaction Reaction (Relative to ; (degrees) (cm) Rate Rate Measured)
"Fe 0 -91.4 2.108E-17 2.439E-17 16% "Fe 0 0 2.201E-17 2.632E-17 20 % i "Fe 0 91.4 2.341E-17 2.584E-17 10 %
Avg: 15 %
"Ni 0 -91.4 3.204E-17 3.617E-17 13 % "Ni 0 0 3.335E-17 3.893E-17 17 % "Ni 0 91.4 3.630E-17 3.863E-17 6% ,
Avg: 12 %
Cu 0 -91.4 3.102E-19 2.779E-19 -10% Cu 0 0 3.066E-19 3.190E-19 4% Cu 0 91.4 3.658E-19 3.119E-19 -15%
Avg: -7% 1
* U 0 -91.4 1.546E-16 1.607E-16 4% '" U 0 0 1.657E-16 1.691E-16 2% '" U 0 91.4 1.728E-16 1.743E-16 1% i Avg: 2% '"Np 0 -91.4 3.030E-15 2.971 E-15 -2% ' N p 0 0 3.738E-15 3.256E-15 -13% ' N p 0 91.4 3.880E-15 3.381 E-15 -13%
Avg: -9.3%
"Fe 45 -123.4 4.859E-18 5.259E-18 8% i "Ni 45 -123.4 6.963E-18 7.550E-18 8% Cu 45 -123.4 8.690E-20 7.889E-20 -10% i '" U 45 -123.4 2.988E-17 2.910E-17 -3% '"Np 45 -123.4 5.653E-16 5.350E-16 -6%
1 1 i 42 l
Table 13: Comparison of"Fe Results at the O' Location Axial Measured Predicted Difference Position Reaction Reaction (Relative to (cm) Rate Rate Measured)
-182.4 9.453E-18 1.183E-17 20%
i -161.4 1.553E-17 1.715E-17 9%
-140.4 1.889E-17 2.070E-17 9%
-119.4 2.104E-17 2.339E-17 10 %
-98.7 2.142E-17 2.452E-17 13 %
-86.6 2.021 E-17 2.437E-17 17 %
-65.85 2.091 E-17 2.426E-17 14 %
-45.1 1.807E-17 2.527E-17 See Note 1
-24.35 2.226E-17 2.541E-17 12 %
-4.65 2.192E-17 2.610E-17 16 %
4 2.207E-17 2.622E-17 16 % 12.2 2.262E-17 2.600E-17 13 % 30.95 2.377E-17 2.637E-17 10 % 49.7 2.484E-17 2.720E-17 9% 68.45 2.429E-17 2.757E-17 12 % 86.65 1.994E-17 2.618E-17 See Note 1 97.2 2.378E-17 2.605E-17 9% 116.7 2.410E-17 2.615E-17 8% 136.2 2.180E-17 2.396E-17 9% 155.7 1.780E-17 1.915E-17 7% 175.2 1.288E-17 1.334E-17 3% Note 1: These measured values appear to be in error and no valid comparisons can be made. l 1 43 1
f ss Table 14: Comparison of Ni Results at the O' Location Axial Measured Predicted Difference Position Reaction Reaction (Relative to f
- (cm) Rate Rate Measured)
-182.4 1.498E-17 1.751E-17 14 % !
-161.4 2.462E-17 2.534E-17 3% (
-140.4 2.954E-17 3.062E-17 4%
-119.4 3.295E-17 3.463E-17 5% i
-98.7 3.240E-17 3.628E-17 11 %
-86.6 3.143E-17 3.612E-17 13 %
-65.85 3.083E-17 3.586E-17 14 % j
-45.1 3.176E-17 3.734E-17 15 %
-24.35 3.369E-17 3.772E-17 11 %
-4.65 3.313E-17 3.865E-17 14 % i 4 3.354E-17 3.887E-17 14 %
12.2 3.526E-17 3.875E-17 9% l i 30.95 3.708E-17 3.936E-17 6% 49.7 3.852E-17 4.044E-17 5% 68.45 3.814E-17 4.092E-17 7% q 86.65 3.177E-17 3.909E-17 See Note 1 97.2 3.712E-17 3.882E-17 4% l 116.7 3.871E-17 3.861 E-17 0% { 136.2 3.613E-17 3.524E-17 -3% 155.7 2.987E-17 2.840E-17 -5% j 175.2 2.107E-17 1.997E-17 -6% Note 1: This measured value appears to be in error and no valid comparison can be made. i i i 44
l Table 15: Comparison of"Fe Results at the 45' Location 1 Axial Measured Predicted Difference Position Reaction Reaction (Relative to (cm) Rate Rate Measured)
-183.9 2.151 E-18 2.415E-18 11 %
! -165.2 3.170E-18 3.554E-18 11 %
-146.5 4.267E-18 4.755E-18 10%
-128.1 4.724E-18 5.241E-18 10 %
-117.6 5.024E-18 5.281 E-18 5%
-96.6 5.081E-18 5.412E-18 6%
-75.6 5.298E-18 5.777E-18 8%
-54.6 5.418E-18 5.934E-18 9%
-33.6 5.561E-18 6.141E-18 9%
-12.6 5.970E-18 6.742E-18 11 %
8.4 6.078E-18 6.657E-18 9% 29.4 6.265E-18 6.863E-18 9% 50.4 6.418E-18 7.034E-18 9% 71.4 6.053E-18 6.745E-18 10 % t 92.1 5.903E-18 6.867E-18 14 % 97.2 5.876E-18 6.790E-18 13 % 116.7 5.756E-18 6.441 E-18 11 % i 136.2 5.406E-18 5.725E-18 6% 155.7 4.342E-18 4.485E-18 3%
- 175.2 3.540E-18 3.325E-18 -6%
l ) l ) 1 45 f
\
Table 16: Comparison of ssNi Results at the 45* Location i t Axial Measured Predicted Difference Position Reaction Reaction (Relative to < (cm) Rate Rate Measured)
-183.9 3.229E-18 3.454E-18 7%
-165.2 4.870E-18 5.066E-18 4% i
-146.5 6.124E-18 6.765E-18 9%
-128.1 6.867E-18 7.505E-18 8% !
-117.6 7.081E-18 7.605E-18 7%
-96.6 7.374E-18 7.782E-18 5%
-75.6 7.498E-18 8.274E-18 9% )
-54.6 7.674E-18 8.545E-18 10 %
-33.6 7.697E-18 8.861 E-18 13 %
-12.6 8.449E-18 9.640E-18 12 %
8.4 8.487E-18 9.577E-18 11 % 29.4 9.025E-18 9.862E-18 8% 50.4 9.564E-18 1.013E-17 6% 71.4 8.866E-18 9.759E-18 9% 92.1 9.012E-18 9.864E-18 9% ! 97.2 8.673E-18 9.747E-18 11 % 116.7 8.975E-l8 9.228E-18 3% 136.2 8.073E-18 8.203E-18 2% i 155.7 6.774E-18 6.462E-18 -5% 175.2 5.515E-18 4.794E-18 -15% 1 i 46
l Section 4 - Uncertainty Estimates 4.1 Analytical Uncertainty Analysis An analytical uncertainty analysis was performed to demonstrate the overall accuracy of the vessel fluence methodology. Contributions to the calculated fluence uncertainty stem from the calculated fluxes as well as the operating history used to determined the integrated fluence. In determining the uncertainty of the calculated fluxes, several important sources of uncertainty were identified. Sensitivity calculations were performed to determine the effects that these uncertainties have on the calculated reactor vessel flux. Table 17 includes a list of uncertainty components that were investigated and the resultant uncertainties in the flux. The analytical uncertainties calculated by statistically combining the component uncertainties are given in Table 18. The most significant uncertainties in the flux were introduced by the iron inelastic scattering cross section (9.5%), vessel out-of-roundness (8.5%), the relative source distribution (7.4%), and the fission spectrum (5.6%). Calculated fluxes at weld locations above the active fuel region also had a 5.4% uncertainty caused by uncertainty in the axial location of the weld. The uncertainty due to the relative source distribution is a i bounding value that is governed primarily by the uncertainty in assembly-wise powers for l 1 1 assemblies containing part-length hafnium flux suppression inserts. The vessel out-of- > roundness contribution was calculated based on manufacturing tolerances for eccentricity I and using a uniform distribution to calculate a standard deviation for the eccentricity. ) -
- y. !
F I 47 1
A small bias associated with tally dimensions used in the Monte Carlo model has been identified and a correction factor was added to the model. A larger, conservative i bias associated with the relative power distributions has also been identified. This bias, estimated to be approximately 10%, is caused by the coarse axial mesh used in the PDQV2 calculations to model the part-length hafnium flux suppression insens. Studies i i indicate that the calculated assembly axial power shapes may be over-predicting the power around the hafnium inserts, resulting is an over-prediction of the beltline weld I fluences for Surry Unit 1. Ilowever, this bias has not been removed from the model and the Surry Unit I fluence calculations. Cycle lengths in EFPD are combined with the calculated neutron flux at the vessel to determine the total neutron fluence. The cycle lengths for completed cycles contribute ! only a negligible amount to the calculated fluence. The capacity factors assumed for ; I future operation can make a contribution to the uncertainty of the calculated EOL fluence. i A capacity factor of 90% is used in this methodology for modeling future operations. It is unlikely that future operations using 18 month cycles will result in frequent cycle capacity factors of greater than 95%. Statistically combing a conservatively large i uncertainty of 5% (la) in the future capacity factor, the resulting analytical uncertainty in the reactor vessel fluence remains less than 20% for all cases listed in Table 18. Note that significant changes in both cycle length and loading pattems require a re-evaluation of 1 the vessel fluence. l 48 (
4.2 Comparisons with Benchmark Results The analytical uncertainty estimates for calculated fluxes range from 16.4% to 18.1% (10). These values bound the differences identified by comparing measurements and predictions for the benchmark cases. The PCA experiments showed agreements between measurements and predictions ranging up to 12% for reaction rates and 14% for fluxes. The in-vessel dosimetry comparisons showed a standard deviation of approximately 12%. Note that is both cases the differences between measurements and calculations includes some measurement uncertainty. The discrepancy between the analytical uncenainty estimate and the benchmarking results are likely due to the conservative estimates used for some of the analytical uncertainties. The ex-vessel cavity calculations showed a larger difference when compared to measurements; however this was expected because of the uncertainty in the iron cross sections used to model the pressure vessel. 49
Table 17: Sources of Uncertainty in the Calculated Flux at the Vessel Inner Surface Source of Uncertainty Flux Uncertainty ; Fission Spectrum 5.6% ( Neutron Production Rate 2.2% Fuel Composition 1% Vessel Location 8.5% j Downcomer Temperature 1.2% Core Coolant Temperature 1.6% l t Carbon Steel Density 0% Stainless Steel Density 2.6% f ron inelastic Scattering Cross Section 9.5% RO Quadrature (Discrete Ordinates Only) 1.5% RZ Quadrature (Discrete Ordinates Only) 1% (Active Fuel Height) 2.5% (Above the Active Fuel) I l Synthesis Approximation (Discrete Ordinates Only) 1.1% (Active Fuel Height) 1 5.0% (Above the Active Fuel) i Tally Uncertainty (Monte Carlo Only) 5.0% ; Neutron Source Distribution 7.4% Axial Location of Welds Relative to the Fuel 5.4% I Axial Location of Welds Relative to the Flux 0.2% Suppression Inserts f 1 l ( 50
] Table 18: Reactor Vessel Flux Aggregate Uncertainties for Various Models and Vessel Locations Model and Location Total Uncertainty (la) Discrete Ordinates Model, 16.4 % f Peak Vessel Fluence Locations ! Monte Carlo Model, 17.0 % Peak Vessel Fluence Locations I Discrete Ordinates Model, 18.1 % Upper Circumferential Weld Monte Carlo Model, 17.9 % Upper Circumferential Weld Discrete Ordinates Model, 16.4 % > Welds Shadowed By liafnium Inserts Monte Carlo Model, 17.0 % Welds Shadowed By llafnium inserts 51
)
( l I l i 4 k I 1 I i l I I 1 l k I 52 l
) Section 5 - Summary and Conclusions A methodology has been developed for calculating the neutron fluence through a reactor pressure vessel. This methodology includes both a discrete ordinates model and a l Monte Carlo model. The discrete ordinates model uses a flux-synthesis technique for calculating a three-dimensional flux distribution, while the Monte Carlo model directly solves the transport equations in three dimensions. Both models use cross sections I derived from ENDF/B-VI and both models use neutron source distributions derived from three-dimensional PDQV2 calculated power distributions. The models were validated by comparisons with the Pool Critical Assembly i Benchmark, Configuration 12/13; in-vessel surveillance capsule measurements, and ex- ) vessel cavity dosimetry measurements. In all cases the calculations and measurements agreed within 20% (1c). An analytical uncertainty analysis was performed to evaluate the uncertainty in the fluence calculated at the reactor vessel inner surface. The resulting analytical uncertainty was dependent on the model used and the location on the vessel surface. Ilowever, in all instances the analytical uncertainty was less than 20%. ) I 53
{ i 4 I i i l i
)
i I f f l i 1 I i ( I 54
l SECTION 6 - REFERENCES
- 1. " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron I
Fluence," Draft Regulatory Guide DG-1053, U.S. Nuclear Regulatory Commission, June 1996. 1 l
- 2. M. L. Williams,"DOTSOR: A Module in the LEPRICON Computer System for Representing the Neutron Source Distribution in LWR Cores," EPRI Interim Report, December 1985.
- 3. " BUGLE-93, Production and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron / Photon Cross Section Libraries Derived From ENDF/B-VI Nuclear Data," RSIC Data Library Collection, DLC-175, Oak Ridge National Laboratory, February 1994.
- 4. G. R. Poetschat, et al., "EPRI-PRESS, Volume 2: User's Manual," Part II, Chapter 5, ARMP-02, EPRI NP-4574, August 1986.
- 5. R. E. MacFarlane et al.," Description of the CELL-2 Fast and Thermal Cross Section Libraries," Part II, Chapter 2 ARMP-02, EPRI NP-4574, October 1987.
- 6. W. A. Rhoades and R. L. Childs,"An Updated Version of the DOT 4 One-and-Two Dimensional Neutron / Photon Transport Code," ORNL-5851, Oak Ridge National Laboratory, July 1982.
i
- 7. J. F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Panicle Transport Code, Version 4A." LA-12625-M, Los Alamos National Laboratory, November 1993.
- 8. J. S. Hendrics, et al., "ENDF/B-VI Data for MCNP," LA-12891, Los Alamos National Laboratory, December 1994.
- 9. M. L. Williams, P. Chowdhury, and B. L. Broadhead,"DOTSYN: A Module for Synthesizing Three-Dimensional Fluxes in the LEPRICON Computer Code System," EPRI Interim Report, December 1985.
- 10. E. P. Lippincott," Westinghouse Surveillance Capsule Neutron Fluence
! Reevaluation," WCAP-14044, Westinghouse Electric Corporation, April 1994. I 1. W. N. McElroy, Editer," LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments and Blind Test," NUREG/CR-1861 ) (Hanford Engineering Development Laboratory,IIEDL-TME 80-87), July 1981.
- 12. ASTM Standard E704-90," Standard Method for Measuring Reaction Rates by
! Radioactiviation of Uranium-238"
- 13. ASTM Standard E844-86," Standard Guide for Sensor Set Design and Irradiation for
) Reactor Surveillance". 55
- 14. J. S. Perrin, et al.,"Surry Unit No.1 Pressure Vessel Irradiation Capsule Program: j Examination and Analysis of Capsule T," BATTELLE Columbus Laboratories. June i 1975.
1
- 15. C. R. Mitchell and M. F. Murphy, " Pressure Vessel Dosimetry Measurements on d Virginia Power Surry Unit 1, May 1994 to September 1995," AEA Technology plc, April 1996. ,
l
- 16. W. A. Rhoades and M. B. Emmett," DOS: The Discrete Ordinates System,"
ORNLffM-8362, Oak Ridge National Laboratory, September 1982. ,
- 17. R. A. Hall "The PDQ Two Zone Model," VEP-NAF-1, Virginia Power, July 1990.
I 8. W. N. McElroy, R. Gold, E. D. McGarry, Editors, " LWR Pressure Vessel { Surveillance Dosimetry Improvement Program," NUREG/CR-3320 (WHC-EP-0204), Volume 2, July 1992. i l l i j
)
i i l 56
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