NUREG/CR-6887BNL-NUREG-75098-2005

DORT/TORT Analyses ofthe Hatch Unit-i Jet Pump RiserBrace Pad Neutron DosimetryMeasurements with Comparisonsto Predictions Made with RAMA

Brookhaven National Laboratory

U.S. Nuclear Regulatory CommissionOffice of Nuclear Regulatory ResearchWashington, DC 20555-0001

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NUREG/CR-6887BNL-NUREG-75098-2005

DORT/TORT Analyses of the HatchUnit-1 Jet Pump Riser Brace PadNeutron Dosimetry Measurementswith Comparisons to PredictionsMade with RAMA

Manuscript Completed: April 2004Date Published: November 2005

Prepared byJ.F. Carew, K. Hu, A. Aronson,A.N. Mallen and M. Todosow

Brookhaven National LaboratoryUpton, NY 11973-5000

With contributions byDJB. Jones and K.E. Watkins

TransWare Enterprises, Inc.

W.E. Norris, NRC Project Manager

Prepared forDivision of Engineering TechnologyU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001NRC Job Code Y6883

ABSTRACT

Regulatory Guide 1.190 for determining pressure vessel fast neutron fluence requires that the vessel fluence calculationalmethodology be evaluated using dosimetry measurement benchmarks. To insure the viability of the underwater weldingthat has been proposed for repairing Boiling Water Reactor (BWR) highly irradiated stainless and high nickel alloyvessel internals, an accurate and well benchmarked calculational method is required for determining the BWR thermalfluence. The recent Electric Power Research Institute (EPRI)/ Nuclear Regulatory Commission (NRC) measurementprogram performed at Hatch-I provides both a fast and thermal neutron dosimetry data-base for benchmarking BWRfluence calculation methodologies.

Under the NRC "BWR Fluence" Program (JCN-Y-6391), the neutron dosimetry measurements performed at Hatch-1are being used as a data base for assessing the accuracy of calculational methodologies used to predict neutron fluenceaccumulated by BWR internal components and the vessel. Results from both the well established DORT/TORTcomputer codes and the new state-of-the-art RAMA Fluence Methodology were to be compared in order to reduce theuncertainty in fluence estimates, and in the future, to permit a better evaluation of the feasibility of underwater weldingtechniques of highly irradiated components.

Detailed calculations of theHatch-1 jet pump riser brace pad thermal and fast neutron dosimetry measurements have beenperformed byBrookhavenNational Laboratory (BNL) using theDORT/TORTdiscrete ordinates transport methodologyand by Transware Enterprises Inc. (TWE) using the RAMA three-dimensional fluence methodology. The calculationsfor both code models were performed using nuclear data primarily based on the BUGLE-96 nuclear data library. Thecalculations were performed using a detailed description of the Hatch-I core/internals/vessel material and geometricalconfiguration. The core neutron source includes the effects of the pin-wise power distribution on the core periphery andthe effects of plutonium buildup on the magnitude and energy dependence of the neutron source.

Comparisons of the fluences calculated by DORT/TORT (performed by BNL), and fluences and activations calculatedby RAMA (performed by TWE), and the Hatch-i measurements have been performed to assess the accuracy of themethodologies for predicting the fast and thermal neutron fluence of BWR internal components and the vessel.Measurement-to-Calculation (M/C) fluence comparisons were also performed. The DORT/TORT methodology wasfound to predict the fast and thermal fluence measurements to within - 5% and -15%, respectively. TheDORT/TORTfluence measurement predictions by BNL are considered to be within the combined accuracy of the calculations andmeasurements. The DORT/TORT thermal fluence calculations showed an average measurement to calculation ratio of1.17. The DORT/TORT calculation under-predicted the thermal fluence measurement by -15%.

The RAMA methodology by TWE was found to predict the fast fluence measurements to within - 7%. The thermalfluence measurements were predicted by RAMA with a measurement to calculation ratio of -0.62 with a standarddeviation of 0.2%. The RAMA methodology performs a direct three-dimensional solution of the transport equation andcalculates fluence predictions. The fast predictions are in good agreement with measurements. However, the RAMAcode over-predicted the measured thermal fluence by -60%. The agreement between calculated and measured fast andthermal activations was similar to that of the fluence in both direction and magnitude. The large difference between theRAMA calculated thermal fluence and the measured thermal fluence at the pressure vessel wall warrants furtherinvestigation.

TheDORT/TORT-to-RAMA differences forthe fast fluence measurements are .4%, and are consistent with the accuracyof the two calculational methods. The large over-prediction (-60%) in the RAMA thermal fluence, compared to themeasured fluence, prevents any meaningful comparison with the DORT/TORT synthesis approach at BNL. TheDORT/TORT approach under-predicted the thermal fluence measurements by an average of -15%.

iii

CONTENTS

Page

Abstract .i....................:; .. HiiList of Figures .......... . ; ; viList of Tables ........ :. viiForeword ............ . . .... ixAcknowledgments ................. ;; .. xi

1; . Introduction ................................... .. .-..............

2 . Hatch-I Riser Brace Pad Dosimetry Measurements .... . .............. 2-1

3 . Hatch-i Plant Data ........................ 3-i3.1 Introduction .... . : 3-13.2 Core, Internals and Vessel Geometry and Materials Data ................. 3-13.3 Jet Pump Riser Brace Pad Geometry and Materials Data .3-23.4 Core Neutron Source Data .; 3-2

4 Calculational Methods ............................ -;r..-.v...................... 4-14.1 Introduction .......... ....................... 4-14.2 Neutron Cross Sections and Fission Spectra .. .......... '.4-1

4.2.1 . DORT/TORT Fission Spectra . . .I: '.. 4-14.2.2 RAMA Fission Spectra .. 4-1

4.3 Core Neutron Source................................. 4-14.3.1 DORT/TORT Source Calculation .. 4-14.3.2 RAMA Source Calculation .. 4-2

4.4 DORT/TORT Neutron Transport Calculations . . .4-24.4.1 Neutron Transport Method .. 4-24.4.2 Flux Synthesis Method .. 4-24.4.3 Dosimetry Cross Sections .. 4-34.4.4 Thermal Fluence Calculation .. 4-34.4.5 TORT Calculations of the Effect of the Local Brace Pad Geometry . .4-3

4.5 RAMA Neutron Transport Calculations .. 4-44.5.1 Neutron Transport Method.. 4-44.5.2 The RAMA Fluence Model .. 4-44.5.3 Dosimetry Cross Sections .. 4-54.5.4 Thermal Fluence Calculation .. 4-5

5 Comparisons of Calculations and Measurements . . .5-15.1 Introduction .. 5-15.2 Brace Pad Dosimetry and Fluence Comparisons .. 5-1

5.2.1 Comparison of Fluence Calculations with Measurement .5-15.2.2 Comparison of DORT/TORT and RAMA Fluence Results .5-15.2.3 Comparison of RAMA Activity Calculations With Measurements .5-2

6 Summary and Conclusion . . .6-1

7 References . . .7-1

v

FIGURES

Page

2.1 Hatch- I Jet Pump and Riser Assembly . .................................................. 2-32.2 Hatch-1 Jet Pump Riser and Brace ....................................................... 2-42.3 Azimuthal Locations of Hatch- I Riser Brace Pad Dosimetry Measurements ...... ................ 2-52.4 Location of Hatch-I Brace Pad Dosimetry Samples Viewed from Inside the Vessel ..... ........... 2-63.2.1 Hatch-1 Fuel Assembly Lattice Geometry . 3-83.2.2 Hatch-1 Core Configuration . 3-93.2.3 Radial Fuel Assembly Material Groups . 3-103.3.1 Hatch-I Jet Pump Riser and Brace Configuration . 3-113.3.2 Azimuthal Location of Jet Pump Riser Brace Pad on Pressure Vessel . 3-123.3.3 Horizontal View of Jet Pump Riser Brace Pad . 3-133.3.4 Horizontal and Vertical View of Jet Pump Riser Brace . 3-143.3.5 Vertical View of Jet Pump Riser Brace . 3-154.1 Hatch-1 Planar Geometry. ............................................ 4-74.2 Hatch-i Axial Geometry ............................................ 4-84.3 Top View of the Hatch-I TORT Riser and Brace Model ....................................... 4-94.4 Side View of the Hatch-i TORT Riser and Brace Model .................. .................. 4-104.5 Horizontal View of the Hatch-i TORT Riser and Brace Model ............ ................... 4-114.6 Top View of the Hatch-I RAMA Model Near the Jet Pump Riser Brace Assembly Elevation .... ... 4-124.7 Axial View of the Hatch-I RAMA Jet Pump Riser Brace Assembly Model ...................... 4-134.8 Front View of the Hatch-i RAMA Jet Pump Riser Brace Pad ................................ 4-14

vi

TABLES

Page

2.1 Hatch-i Brace Pad Fast and Thermal Fluence Measured Reaction Rates ....... .................. 2-23.2.1 Hatch-1 Core/Internals/Vessel Geometry and Materials Data ......... ......................... 3-33.2.2 Cycle 1-4 Core Material Compositions .................................................... 3-43.2.3 Cycle 5-12 Core Material Compositions ................................................... 3-53.2.4 Cycle 13-19 Core Material Compositions ............. .................................... 3-63.2.5 Vessel and Internals Material Compositions . .............................................. 3-74.2.1 Fraction of Fissions by Isotope as a Function of Fuel Exposure ........ ........................ 4-64.2.2 TORT Calculations of the Effect of the Riser Brace and Brace Pad on the Dosimetry Measurements ... 4-65.2.1 Comparison of the Jet Pump-3 and 13 Measured and DORT/TORT Calculated Fluences .5-35.2.2 Comparison of the Jet Pump-5 and 15 Measured and DORT/TORT Calculated Fluences .5-35.2.3 Comparison of the Jet Pump-3 and 13 Measured and RAMA Calculated Fluences .5-45.2.4 Comparison of the Jet Pump-5 and 15 Measured and RAMA Calculated Fluences .5-45.2.5 Comparison of the DORT/TORT and RAMA Predictions of the Jet Pump-3 and 13 Measurements 5-55.2.6 Comparison of the DORT/TORT and RAMA Predictions of the Jet Pump-S and 15 Measurements 5-55.2.7 Comparison of Jet Pump-3 and 13 Activation Measurements (gCi/mg) to RAMA Calculated Activations 5-65.2.8 Comparison of Jet Pump-5 and 15 Activation Measurements (g±Ci/mg) to RAMA Calculated Activations 5-75.2.9 Comparison of the Activation Measurements to RAMA Calculated Activities ...... ............... 5-7

vii

FOREWORD ., .

Industry operating experience has demo-nstrated the occurrence of cracking in stainless steel and highNickel alloy core internal components of boiling water reactors (BWRs), and the incidence of cracking isexpected to increase as U.S.'nuclear pow'er plants continue to age. To address this issue, the ElectricPower Research Institute (EPRID is studying a variety of repair and mitigation strategies. Mechanicalrepairs are not alw ys practical because of obstructions. As a result, for many components, weldedrepairs may be the only viable strategy. For irradiated stainless steel's,how'ever, welded repairs may notbe viable due to the potential for cacking.

The issue at hand is that Helium,' which is produced in stainless steel &omponents as a result ofirradiation, accumulates during welding anid forms bubbles that gro-w'apidly, and these bubbles can leadto cracking during the welding process. Helium results from the transmutation of Boron and Nickel;however, the Boron content of U.S. BWR components is usually un mw-i because Boron is an impurityin stainlcss stcels . .'

EPRI began gathering information in the late 1990s concerning the feasibility of welding stainless steelswith varying Helium conterit. 'However, EPRI soon determined that additional information would beneeded in order to render a sound technical judgment concerning the cceptability of such welding forBWRs in the United States. *

The U.S. Nuclear Regulatory Commission (NRC) and EPRI signed a cooperative agreement on January11, 2000, to address the cracking of stainless steel and high Nickel alloy in-vessel components of U.S.BWRs. Research conducted under that agreement has identified Helium content thresholds to ensureacceptable weldability of highly irradiated stainless steels.

Helium content can be determined through measurement or estimated by analytical methods. The mostreliable method is measurement where very small samples are removed from the components of interestto measure their respective Heliumn content. However, sampling and measurement are costly and time-consuming. Analytical methods (computer codes) can be used to calculate Helium content provided thatthe fluence of the given material (a measure of the number of neutrons strilng the material) is known.Prior to this study, the fluen'ce ad not yet been assessed for many of the components of interest. Smallsamples were taken by Framnatoine ANP from typical BWR jet pump riser brace pads (JPRBPs). Thesamples were analyzed by Pacific Northwest National Laboratory (PNNL) for Helium content, initialBoron content, and accumulated fluence.

The NRC staff has relied for many years on the well established DORT/TORT computer code forcalculating pressure vessel fluence. However, modeling the complex three-dimensional geometry ofin-vessel components is difficult using DORT/TORT. In addition, the code had been benchmarked onlyfor the estimation of fast neutron (high-energy) fluence. Epithermal and resonance thermal neutroncapture is also required to accurately calculate the concentration of Helium. Independently, EPRI hadpreviously sponsored TransWare Enterprises Inc. (TWE) to develop a new, state-of-the-art three-dimensionai computer code, Known as RAMA to estimate fast and thermal neutron fluence. Thisinformation could then be used for comparisons with estimates from analytical methods.

During the course of their cooperative effort, the NRC and EPRI discussed the need to comparethe laboratory-measured fluence (from the work at PNNL) to those calculated by the DORT/TORT(BNL) and RAMA (TWE) codes.

ix

The study originally intended to evaluate how well the DORT/TORT and RAMA transport codesestimated the fast and thermal neutron fluences of the JPRBPs compared to the laboratory-measuredvalues of the actual samples. The fast fluence evaluations were completed, but shortly after the thermalfluence comparison efforts were initiated, several high priority emerging issues compelled a re-evaluation of research activities. The thermal fluence evaluations were initiated but not fully completed.The code comparisons indicated that, for the specific comparisons made, the fast neutron fluences (E >I.O MeV) ralculated usinma both the DORTITORT and RAMA codes agree well with the laboratory-measured values of the 'actual samples. Specifically, DORT/TORT predicted fast fluence to withinapproximately 5 percent of the laboratory-measured values of the actual samples, and RAMLA predictedfast fluence to within approximately 7 percent, which is within the uncertainties of the two calculational.methods. The thermal fluence calculations, however, did not show the same consistency. Rather, theDORT/TORT code underpredicted thermal fluence relative to the laboratory-measured values of theactual samples by approximately 15 percent on average, while the RAMA code overpredicted thermalfluence by approximately 60 percent. This was a first attempt at benchmarking thermal fluence, and thecauses of the discrepancies are not utndeirstood at this time. Therefore, the industry has initiatedadditional investigation into the source of the discrepancies in the thermal fluence calculations.

The cooperative research described in this report describes preliminary efforts to estimate the Heliumncontent and fluences of irradiated stainless steels through the use of computer codes. Although work wasterminated before completion, the preliminary results provide a starting point for further analysis inefforts to use computer codes in lieu of sampling in-vessel components to make weldabilitydeterminations.

Carl J. Pdperidlo, DirectorOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory Commission

x

ACKNOWLEDGMENTS

The work documented in this report was performed under the auspices of the United States Nuclear RegulatoryCommission (USNRC). It was funded by the Materials Engineering Branch of the Division of Engineering Technologyin the Office of NuclearRegulatoryResearch, underJob CodeJCNY-6883. Theprogram was monitored at theUSNRCby W. E. Norris whose support of this work is greatly appreciated. The authors are also grateful to Nilesh Chokshi andMichael Mayfield of the Office of Nuclear Regulatory Research for many valuable discussions and contributions to thiswork.

The authors would like to thank the Southern Nuclear Operating Company staff for their cooperation and support inproviding the detailed Hatch- 1 plant data required for this evaluation, especially Nancy Folk, Kenneth Folk, Robin Dyleand Marty Sims. The cooperation of Dean Jones and Kenneth Watkins of TransWare Enterprises Inc., and KennethWolfe and Robert Carter of the Electric Power Research Institute is also greatly appreciated.

TransWare Enterprises Inc. (TWE) performed all the RAMA calculations and provided the RAMA results used in thisreport under a subcontract to Brookhaven National Laboratory (BNL).

xi

1 INTRODUCTION-;

The fracture toughness of light water reactor pressure'vessel materials is measured in terms of the ReferenceTemperature for nil-ductility transition (RTNw.) TheRTNm is defined as a function of the material chemistry(concentration of Cu and Ni) and the fast neutron ( >1MeV) vessel fluence. Because of the limited marginbetween the predicted and limiting value of RTNmr atcertain plants and in order to provide the necessaryconfidence in the fracture toughness and integrity of thereactor pressure vessel, an accurate calculation of thevessel >1 MeV fluence is required.

In addition, underwater welding has been proposed as ameans of repairing BWR stainless and high nickel alloyvessel internals that have experienced environmentallyassisted cracking (Reference-l). However, the feasibilityof the underwater welding of highly irradiatedcomponents is dependent on the concentration of Hepresent in the steel due to thermal neutron capture intrace amounts of B-10 and Ni-58 present in the originalmaterial.' Consequently, in order to determine the Heconcentration and insure the viability of the proposedwelds, an accurate estimate of the neutron fluenceaccumulated by the vessel internals is required.

Because of the several decades of attenuation in theneutron flux between the core and the pressure vessel,the vessel fluence calculation is extremely sensitive tothe material and geometrical representation, the nuclearcross section data, and the numerical schemes used in itsdetermination. These factors combine to make anaccurate calculation of the internal components andvessel fluence difficult. To provide the necessary levelof confidence in the vessel fluence predictions,Regulatory Guide 1.190 for determining pressure vesselneutron fluence (Reference-2) requires that the vesselfluence calculational methodology be evaluated usingdosimetry measurement benchmarks. I

As part of the work being performed under anAddendum to an NRC/EPRI Memorandum ofUnderstanding (MOU), material samples from severaloperating BWRs have been removed and characterized.The characterization of these samples includes: (1)

'The He producing reactions are:(1) B'0 + n - LF + a and (2) Ni5l + n - Ni59 +y

Ni59 + n - Fe56 + ai.

dosimetry activations'and the determination of thermaland fast nutiiron fluence (Refererice-3a and Reference-3b)an'd (2) heii5 and boron concentrations of scrapingsfrom the]je'tpump riser brace pads'located on the vesselinner wall'(Reference-3a). Under the NRC "BoilingWater Reactor (BWR) Fluence Prdgram (VCN Y-6391),"these measurements are being used to provide a dosimetrydata-base for benchmarking calculation methodologiesused to predict the fluence accumulated by the BWRinternal compoinents and vessel.

This report fdocuments the DORT/TORT' FluenceMeth6dology ' calculations performed by BrookhavenNational Laboratory (BNL) and the RAMA' FluenceMethodology calculations 'performed by TransWareEnterprises' Inc. ' (TWE) of the Hatch-I dosimetrymeasurements. The DORT/TORT discrete ordinatestransport code system used 'at BNL represents a wellestablished "methodology used for 'radiation transportcomputation' The DORT/TORT code 'system wasdeveloped by the OakRidgeNationalLaboratory (ORNL)(Reference4). The RAMA Fluence Methodology(RAMA) that was used by TWE represents a new state-of-the-art methodology for fluence determination (Reference-5). RAMA was developed by TransWare Enterprises Inc.(TWE) under sponsorship of the Electric Power ResearchInstitute (EPRI) and the Boiling Water Reactor Vesseland Internals Project (BWRVIP).

The purpose of the comparisons of the DORT/TORT andthe RAMA fluence calculations was to compare theuncertainties in fluence estimates, allow the evaluation ofa more modem computer code, and improve and advancethe state of knowledge in the area of radiation transportcomputation. The calculations were performed using adetailed description of the Hatch-b core/internals/vesselmaterial and geometrical configuration provided by theLicensee, Southern Nuclear Operating Company. In orderto insure an accurate calculation, a substantial effort wasmade to determine the core neutron source through Cycle-19 when the brace pad scrapings were taken. The neutronsource is based on the Hatch-1 three-dimensional powerand exposure distributions, and pin-wise powerdistributions for the fuel bundles close to the coreboundary. The neutron source was based on plant ProcessComputer data for cycles 1-12, and on plant nodalcalculations for cycles 13-19.

I1-1

The present analysis differs from traditional vesselfluence evaluations in two respects. First, the presentanalysis requires the determination of the neutronthermal reaction rates and fluences whereas typicalvessel analyses focus primarily on the > 1 MeV and >0.1MeV fluences. Second, because the relatively complexlocal geometry involving the riser, brace yoke and leavessurrounding the brace pads has a substantial three-dimensional character, special consideration wasrequired to insure accurate estimates of the fluence at thebrace pads.

The Hatch-i jet pump riser brace pad dosimetrymeasurements are described in Section-2. The Hatch-Iplant data including: (1) the core/internals/vesselmaterial and geometrical configuration and (2) the coreneutron source are given in Section-3. TheDORT/TORT and RAMA calculational methods used topredict the brace pad measurements are described inSection-4. The measurement-to-calculation ratios for theDORT/TORT and RAMA methods, and theDORT/TORT-to-RAMA fluence comparisons arepresented in Section-5. A summary of the results, and theconclusions, are presented in Section-6.

1-2

2 HATCH-1 RISER BRACE PAD DOSIMETRY MEASUREMENTS

The Hatch-I jet pump riser brace pad (JPRBP)measurements were made, in part, to provide a fast andthermal neutron dosimetry data base for an operatingBWR plant. The samples were taken by FramatomeTechnologies after the Cycle-19 shutdown on September30, 2000. The samples consisted of small "divots"drilled from the inside surface of the Inconel-182 bracepads to a depth of 10.1524 cm. The divots were drilledfrom a location in the brace pads located immediately tothe right (viewed from inside the vessel) of Jet Pumps 3,5, 13 and 15. (Note:. The Jet Pumps described here arealso referred to as mixer tubes in this report.) Thespecific location of the samples is shown in Figures 2.1 -2.4.. As shown in Figure-2.4, the samples were takenfrom the left-hand side (viewed from inside the vessel) ofthe brace pad. It is noteworthy that Jet Pump-3 and JetPump-13 are at azimuthally-symmetric locations and,except for differences between the design and as-builtplant configuration, should yield identical results. JetPump-5 and Jet Pump-15 are also at azimuthally-symmetric locations.

considered unnecessary because of the relatively lowfluence at the vessel. The thermal activities werereduced to account for the contribution from epithermalabsorption.;

The fluence estimates inferred from these activities,include adjustments for isotope decay, interfering

reactions and epithermal absorption. -When the fluenceestimates for the individual reaction rates are combined,the one-sigma-. fluence uncertainty - due -to -themeasurement is -10%. However, in view of therelatively-large . (-30% in the case -of -the thermalmeasurements) asymmetries - observed between

- symmetric locations, the total measurement uncertainty,for both fast and thermal fluence, is believed to besomewhat larger (especially for the thermal fluence)-10-15%.

The >1 MeV, >0.1 MeV and thermal fluence for thebrace pad dosimetry samples at each of the four jet pumplocations were determined from the measured activitiesfor the reaction rates given in Table-2.1 (taken fromReference-3a). The >1 MeV and >0.1 MeV fast neutronfluences are based on measurements of the Fe-54 (n, p)Mn-54, Ni-58 (n, p) Co-58 and Nb-93 (n, n') Nb-93mresidual gamma activities. The thermal neutron fluenceestimates are based on measurements of the gammaactivities associated with the Fe-54 (n, y) Fe-55, Ni-62 (n,y) Ni-63, Fe-58 (n, y) Fe-59, Cr-50 (n, y) Cr-51 and Co-59 (n, y) Co-60 reactions. The samples were gammacounted and then adjusted to the Hatch-i initial fullpower (2436 MW1) using the plant operating history.2

The activities were adjusted to account for reactionproduct decay during plant operation and up to the timeof counting. Corrections were made for gamma self-absorption and interfering reactions. However, becausethe samples were taken from the surface of the brace pad,the fluence did not require an adjustment for the effectsof neutron self-absorption. Adjustments for burnup andtransmutation of target and product isotopes were also

2The gamma counting and processing were performedby L. R. Greenwood and are described in detail inReference-3a.

.2-1

Table 2.1 - Hatch-I Brace Pad Fast and Thermal Fluence Measured Reaction Rates.

Thermal | Fast > 0.1-MeV I Fast > -MeV |

54Fe(n, y)55Fe 9 3Nb(n, n') 93mNb 93npbl(n, ) 9 3mm

62Ni(n, y)63Ni 5Fe(n, p)54Mn' -Fe(n p)5Mn'

58Fe(n, y)59Fe 58Ni(n, p)58Co 58Ni(n, p)58Co

Cr(n, y) 5 Cr '

5 9 Co(n, y)60Co ;

2-2

Ccre Shroud

* ~-- Jet PurrpBeam Bcits

RlserBraces

1.-Mer

D ffuser

'-Senshn LineBrackets.

~- -- Baffl Plate

Figure 2.1. Hatch-i IJet Pump and Riser Assembly

2-3

Figure 2.2. Hatch-i Jet Pump Riser and Brace

2-4

* : Measurement- 3Measurement-15

Measurement-5

Measurement-1 3

* Figure 2.3. Azimuthal Locations of Hatch-l Riser. Brace Pad Dosimetry Measurements

2-5

NDosimetry

Sample

I

I ~I

z_ . . _ w

z 1

4.78

*1 I I! .. _I_z I

0.356 X _-RISER

(Dimensions in cm)

Figure 2.4. Location of Hatch-i Brace Pad DosimetrySamples Viewed from Inside the Vessel

2-6

- -3 HATCH-I PLANT DATA

3.1 Introduction -

Because of the strong attenuation of the neutron fluencebetween the core and the vessel (- three to four decadesdepending on location) and the resulting sensitivity of thefluence transport calculation, an accurate and detaileddescription of the core/internals/vessel configuration isrequired. The data required for the calculation of theHatch-i riser brace pad measurements was requestedfromSouthernNuclear Operating Companyin Reference-7. The data was provided in a series of transmittalsincluding detailed drawings of the plant configuration andelectronic files describing the core neutron source.Additional measurement related data was taken fromReference-3a and Reference-3b and core design data wastaken from References 8-9. BNL reviewed this data andprepared the input required for the transport calculations.The data and input preparation were consistent with therequirements of Regulatory Guide 1.190.

3.2 Core, Internals and Vessel Ge-ometry and Materials Data

TheHatch-i core geometry and material composition datawere taken from References 8-9 and the material providedby Southern Nuclear Operating Company. The corepower level, inlet temperature and operating pressure aregiven in Table 3.2.1. Fuel assembly and core geometryconfigurations are given in Figures 3.2.1 and 3.2.2,respectively. Figure 3.2.1 illustrates a fuel assemblydesign with a 7x7 array of fuel pin cells. This design isrepresentative of the early operating cycles of Hatch-1.Later cycles of Hatch-i use 8x8 designs. Figure 3.2.1illustrates that the Hatch-i fuel assembly designs havenon-uniform water gaps with alternate narrow and widewater gaps separating the neighboring fuel assemblies.

For this analysis, the fuel assembly material compositionswere homogenized over the fuel assembly pitch, includingthe water gap. The brace pad measurements and vesselfluence are most sensitive to the fuel compositions of theassemblies in the outer-most rows of fuel assemblies. Toaccount for the radial dependence of the core void distri-bution, region-specific average fuel assembly composi-tions were determined for the four regions defined inFigure 3.2.3: (1) inner core region, (2) outer core region(3) core boundary region and (4) the low void region.The axial void distributions were included. The use of the

group-average fuel compositions, rather than assembly-specific compositions, results in less than a -1% error inthermeasurementpredictions. Plant Process Computer(Reference 10) data was'used to determine the coolantdensities for-Cycles 1-12 and core nodal calculationswere used for Cycles 13-19. For convenience Cycles 1-4were combined (Process Computer data for the 144 inchcore), Cycles 5-12 were combined (Process Computerdata for the 150, inch core) and Cycles 13-19 werecombined (nodal calculation data with the 150 inch core).In combining this data, the cycle-specific data wasweighted by the cycle exposure. The region-wise fuelcompositions for Cycles 1-4, Cycles 5-12 and Cycles 13-19 are given in Tables 3.2.2 - 3.2.4, respectively, at theelevation of the dosimetry measurements.

The fuel number densities provided in Tables 3.2.2 -3.2.4 for the 2'U, U 8U, and Of,.,. isotopes are appropriatefor the DORT/TORT discrete ordinate methods, but notfor RAMA. In addition to these isotopes, the RAMAmethod requires number densities for the 239Pu, 240Pu,241Pu, and 242Pu isotopes., Furthermore, the ratio of U/Punumber densities must be consistent with fuel exposure.To generate the fuel material data needed for the RAMA'flueri i xiodel,an 8x8 fuel assembly design typical ofthe latei'-Hatch-l cor& loadings was depleted. Theuraniuim mrd pilut6nium number densities from thedepletions were tabulated as functions of exposure andvoid fractio. Using the Hatch-i fuel exposure and voiddata for each operating cycle, fuel number densities forthe RAMA fluence model were determined. The coolantnumber densities in Tables 3.2.2 through 3.2.4 were usedas provided.

The geometry and material'data for the Hatch-l vesseland internals were taken from drawings provided bySouthermf Nuclear Operating Company.' The componentdimensions for the core shroud, jet pump riser and mixertubes, pressure vessel, mirror insulation, and biologicalshield are 'given in Table 3.2.1. The indicated geometryand materials apply at elevations in the neighborhood ofthe jet pump riser brace pads (viz., it axial elevationsaround 742.95 7cm that is the mid-elevation of the jetpump brace assembly).

c..i ~.-. .:.. t: . :. :. .-

- S.' :'1 _r'..........!'..'i.

3-1

3.3 Jet Pump Riser Brace Pad Ge-ometry and Materials Data

The local geometry surrounding the brace pad dosimetrylocations is relatively complex due to the proximity of thejet pump mixer pipe, riser pipe, riser brace, core shroudand pressure vessel. The geometrical arrangementsurrounding the brace pad dosimetry locations is given inFigures 2.2 - 2.4 and Figures 3.3.1 - 3.3.5. As indicatedin Figures 2.2 and 2.4, the dosimetry samples were takenfrom the left-hand side of the left-hand brace pad (asviewed from the inside of the vessel). The dimensions ofthe brace pad and the relative location of the brace leaveson the pad are shown in Figure 3.3.2. As shown in Figure3.3.3, the brace pads are located at an' elevation of 742.95cm (relative to the inside of the bottomn of the vessel). Asa reference, the core mid-plane is at an elevation of712.622 cm. The dimensions of the riser brace are givenin Figures 3.3.4 and 3.3.5.

gradient in the remaining inner fuel bundles is negligible(<1%) to the ei-core fluence problem and was notincluded in the analysis. The fuel bundle pin-wise powerdistribution data for Hatch-1 Cycles 1-19 was not avail-able at the time of these calculations. However, since thesource gradient effect has been determined to have an-4% effect on the vessel' fluence based on extensiveearlier analyses, a set of typical pin powers was used inthis analysis.; The effect of this approximation is esti-mated to have a negligible effect (<1%) on the fluencepredictions. The pin powers used in this analysis weretaken from Reference-13.

The jet pump mixer pipe, riser pipe and brace are SS-304,and the brace pad is Inconel-182. The specific materialcompositions are given in Table 3.2.5.

3.4 Core Neutron Source' Data

The core neutron source is determined by the three-dimensional power and exposure distributions. Thepower distribution is used to determine the spatial distri-bution of the fission rate and source density.. The expo-sure distribution is used to account for the increasednumber and harder (more penetrating) spectrum ofneutrons produced in plutonium fission.

The Southern Nuclear Operating Company provided thecore power and fuel exposure distribution data for Cycles1-19. The core neutron source was determined using thecycle-specific beginning-of-cycle (BOC) and end-of-cycle(EOC) three-dimensional burnup distributions. The three-dimensional exposure'distributions for Cycles 1-12 weredetermined using Proceiss Co mputer data based on plantTraversing In-core Probe (TIP) and Local Power RangeMonitor (LPRM) neutron flux measurements. The sourcedistributions for Cycles 13-19 were determined usingcalculated cycle-specific BOC and EOC three-dimen-sional burnup distributions.

Fuel bundle pin-wise power distributions were used todetermine the radial source gradient for the three outer-most rows of fuel bundles (i.e., the fuel bundles closest tothe core boundary). The effect of the pin-wise source

3-2

Table 3.2.1 - Hatch.I Core/Internals/Vessel Geometry and Materials Data

Reactor Parameter | Parameter Value Material

Thermal Power Cycles 1-16 - 2436 MWtCycles 17-18 - 2558 MW,Cycle-19 - 2763 MW, l

Core Inlet Temperature , 531 OF ,

Core Operating Pressure 1062 psia ,_,_._._.

Core Saturated Temperature 551.7 0F

Inner Radius of Shroud 221.6150 cm

Shroud Thick'ness 3.81 cm , SS-304

Inner Radius of RPV Liner 279.5588 cm

Thickness of RPV Liner 0.7938 cm SS-304

Inner Radius of RPV 280.3526 cm

RPV Wall Thickness 13.6525 cm A-533 B

Inner Ridius - Insulation Liner" 307.880 c'm-n-, . .

Thickness'of Insulation Liner+ 0.159 cm SS-304

Thickness of Insulation! 8.571 cm'' Al

Inner Radius:- Concrete Shield' 392.970 cm

Thickness of Concrete Shield' 40.62 cm Concrete

Center-Line Radius of the Riser 246.8118 cm

Number of Jet Pumps 20

Location of Jet Pumps , + 8.1771° of Riser

Location of Risers - At 30° intervals starting at 30°'

Jet Pump Inside Diameter 16.4592 cm - SS-304

Jet Pump Outside Diameter 18.9992 cm .

Riser Inside Diameter 25.7454 cm SS-304

Riser Outside Diameter 27.3050 cm '

* The material, thickness and separation from the pressure vessel of the insulation and concrete shield have been taken to be the same asthe BNVR benchmark problem of NUREG-6115.

7,The core flats are 00, 900, 1800 and 270°. There are no risers at the two recirculation suction nozzles at 0° and 180°.

3-3

Table 3.2.2 Cycle 1-4 Core Material Compositions

Material Component' Dersities (atoms/barncm)

Inner Core H 2.0627-02:o 1.0313-02

o (Fuel) 1.0677-02-U-235 5.4093-05U-238 5.2025-03-

Zr .5.7167-03lOuter Core H 2.1678-02:

O -1.0839-02O (FueDl 1.0677-02-U-235 5.4093-05:U-238 5.2025-03

Zr 5.7167-03 -

CoreBondary H 2.1586-02--O 1.0793-02 -

O (Fuel) 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

Low Void Region H 2.4165-02O 1.2082-02

O (FueD 1.0677-02U-235. 5.4093-05U-238 5.2025-03 :

zr 5.7167-03

3-4

Table 3.2.3 Cycle 5-12 Core Material Com" positions

Matexial . Com nent |Densities (atoim/barncm)

Inner Core *H 2.0021-02. I1.0011-02

- (FueD ; 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

OuterCore H 2.1448-02o 1.0724-02

0 (FueD 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

Core Boundary H 2.2210-02o 1.1105-02

0 (FueD 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

; owVoidRegion H -2.4480-021o 1.2240-02

-0 (FueD) 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

.3-5

Table 3.2.4 Cycle 13-19 Core Material C6mpositi6ns

Material | Component | Densities (atons/ban-cm)

Inner Core H 1.7898-02O 8.9489-03

o (Fuel) 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

Outer Core H 2.0461-02O 1.0231-02

O (Fuel) 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

Core Boundary H 2.3173-02O 1.1587-02

O (FueD 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

Low Void Region H 2.5168-02O 1.2584-02

O (Fuel) 1.0677-02U-235 5.4093-05U-238 5.2025-03

Zr 5.7167-03

3-6

Table 3.2.5 Vessel and Irternals Material Compositions. , .. . .

Materia Component ' -; Densities (atonE/barnan)

.- : ' Cr -.- 1.74000E-02Mn .- q r 1.52000E-03

SS-304 Ni 7 - - 855000E-03- Fe ; -- - . - 5.83000E-02

; Ii--C -- ~ >-237000E-04- -- Si ' ' 2893000E04

Fe : >. .'-- 8.265800E-02A-533 B - Ni 4.418000E-04

- ----- -- -:---- :--1.115800E-03C 9.814000E-04

Insulation ., . Al 6.060300E-03

Brace Pad/Inconel-182 Cr 1.4365E-02Mn 6.5263E.03Fe 6.5280E-03Ni 6.0129E-02Nb 8.5000E04

= H 1.51367E-02C -.-- C -- - 224032E3040 8.53268E-02Na 2.04551E-03

Coicrete- Mg --' 2.88319E-04. -- - - -l --4.65596E-03

, Si-: 3.07780E-02: . K 1350031303-. ,- Ca- - -- 4.46115E-03- .. Fe - 6.09755E-04

- -- ,.' -. , ; I I.-, I - . - _ - -. .' I. . g-j .- ;I - .

3-7

C -1- H- R-0- N

00.00 op'° °° B 00°

O 000 0 00

00000 00O0 0 00

000000I 0 0 0.000,

11I

I

II

rp 00000

000 ooooC

000000,0

jop-0000-00-00

I 00000

N( bhT-

:,< .,

_-

_ . . oo)e,

1.4 - -- Ki. - mfnl'(DG9 FE

G) (D ( ) . a. .

1D @''s @,4 (a 0 e

D e-

0

0

aC

0.000000

IlII

'1

y

SI-DRIINSIDE RAO)

t--

I

DIM.IDENTIFICATION A - | C O E |F I G H [t I .1 IDIM. INCHES 12.0 5.278 0375 | 0.C0 0.175 0.14350.7'8

DIM. IDENTIFICATION tC | L M N O aC R | SDIM. INCHES 0.188 I 0.380

I - -

Figure 3.2.1. Hatch-i Fuel Assembly Lattice Geometry

3-8

* '7 ME' M MOk -

45 *oqooCooo nE OQ .

211 ,odr 000 DOO

,, *IT ~poOQ5 ~lloC°PD°S:

04 12 1J

N0UM0 BO O UL A010 °11° 1T 00 °GCI0,0 0r C 001°10 DlO °1 °

13+1

*IC UOO Du7C OPSOF LL C COINE °FUEL BUNDLES W4 14 ORIFICE DIAMET -D7

+°B1°~~~ °+-1 r1^1 0 0 [:L °1 °DOIDO a1 °1 Ot ~lOOoooooo~olol000~z10

To~ o lo lo12'oo oo o 2...

,, , ,05,, ,,,,E.OF ... ' * ENT°1SE°1 UE°1 3 °

'K COMO POITO .F12R*20 LL'tIPMA -I

)n FUEOMAVNDESWITION FO In.ORIF1C O IAMHINETES

*FUELOUNDLESVATM .4g.In.ORIFICEDIAMETER-76

. . . , : w,,. .1 . - ., t . I

. -; , . .. . ;I . . I* 'V , I *. ; .- *

Figure 3.2.2. Hatch-i Core Configuration

3-9

W Inner Core Assemblies

OOuter Core Assemblies

W .Core Boundary Assemblies

.. *\7,Avr~.7V_.:

_ _ _9

.,. , n�

Iz-I:7, Aetna',,1- -9m - .-

68 71...., -Low Void Assemblies

_ _ _

-63 64 g g ..:.....56 57 58 :59 1

47 1481 49 50 51 52 1ffl5.

37 38 39 401 41 42 43111�\

,.......- a-i - I�i�I�t� I. - *I�9 -9--

26 27 28 29 30 31 32 33 .- l- , . - - - --- -114 15 16117118119 20 21 22 1...............

1 2 3 4 5 6 7j8 1910M

Figure 3.2.3. Radial Fuel Assembly Material Groups

3-10

! s I '. .1-

sLT 's ' ." .. I = - .

v. r-cNBRACE TORISERRS-8 NTO RS-9

YORI

SUPPORT PAD

PAD

I

- r _ I

-: Figure 3.3.1. Hatch-l Jet Pump Riser 'iid Brace Configuration

i3-11

SUPPORT PADINCONEL 182

NZZLEN2

RB-ic11.4

RB-id

(Dimensions in cm)

Figure 3.3.2. Azimuthal Location of Jet Pump Riser Brace Pad on Pressure Vessel

3-12

0.953. - .

- -- - c , . . . _

742;95,1 RPV .ELEV.I i

, ?

SD- 0.794 SS

0.476(Dimensions in cm)

: .. F. .,. .. . . . _.. X :. . ' U - - -,

Figure 3.3.3. Horizontal View of Jet Pump Riser Brace Pad

3-13

I,RB-2b (UPPER)

/ RB-2d (LOWER)

8.89

-V304 SS,

I i0%*0

I1-v.

\ RB-2a (UPPER)RB-2c (LOWER)

T .991

(4 PLACES)

(Dimensions in cm)

Figure 3.3.4. Horizontal and Vertical View of Jet Pump Riser Brace

3-14

(Dimensions in cm)

Figure 3.3.5. Vertical View of Jet Pump Riser Brace

3-15

-' - 4 CALCULATIONAL METHODSI~~ .. I i. :f- i-- ! ~

I -;1 . , ;. * * .- . - . .

4.1 Introduction

The DORT/TORT (used at BNLQ and RAMA (used atTWE) nethodologies used to perform the Hatch-l jetpump riser brace pad measurements generally follow themethods'and approach described'in Regulatory Guide1.190 for pressure vessel fluence calculation and measure-ment methods.;However, the present calculations have asomewhat broader catio in'that Regulatory Guide1.190 is primarily'concerned wvith the fast fluence whilethe present analysis also requires the determination of thethermal neutron fluence. In addition, the calculation ofthe bra' pad measurements requires the modeling of therelatively complex three-dimensional riser/brace/vesselgeometry surrounding the brace pad sample locations (seeFigures 2.2 and 3.3.1), 'which is typically not required inthe determination of the vessel fluence.

The' nuclear cross sections, determination of the coreneutron'so'urce,' transport calculations, DORT/TORTsynthesis approach, and the dosimetry cross sections are'described in the Sections 4.2-4.2.1, 4.3-4.3.1,'4.4-4.4.3,respectively.' The method used to determine the thermalneutron fluence is described-in'Section 4.4.4 and themethods used to model the local riser/brace/vessel geome-try are described in Section 4.4.5

The RAMA' fission spectra and the RAMA source aredescribed in-sections 4.2.2 and 4.3.2 respectively. 'The

'neutron transport calculations, fluence'model, dosimetrycross-sections,' 'and thermal fluence 'calculation 'aredescribed in sections 4.5-4.5.4, respectively.

4.2 Neutro'n.Cross'Sections-andFission Spectra

The DORTWTORT a'nd RAMA ne'utron' transport calcula-"tions were performed using data from the BUGLE-96 datalibrary (Reference-i 1). The BUGLE-96 library providesa 47 neutron/20 gamma-ray broad-grou-penergy represen-tation for determining neutron and photon space-energydistributions in light water reactor shielding and pressurevessel fluence applications. The BUGLE-96 library wasdetermined by collapsing the ENDFIB-VI VITAMIN-B16-fine-group cross section set using spatially 'dependent-spectra typical of -light ;water'reactor configurations.-Special tabulations of- cross-sections are provided inBUGLE-96 for core materials typical ofaBWR operating

environment. The' BUGLE-96 library" includes theENDF/B-VI updates of the iron, hydrogen and oxygencross sectiois, which are known to have a' significant'effect on pressure vessel fluence predictions.

'4.2.1' DORT/TORT Fission Spectra

The fission spectra for U-235, U-238, and Pu-239 thatwere used for the DORT/TORT transport calculationswere taken from BUGLE-96. The fission spectrum forPu-241,was determined with NJOY using ENDF/B-VIdata. These_ spectra were -used :in -the MESH(Reference:12) calculations that prepare the coreneutron source for the DORT/TORT calculations.

4.2.2 RAMA Fission Spectra

The fission spectra for U-235, U-238, Pu-239, Pu-240,Pu-241, and Pu-242 that are used in the RAMA trans-port calculations were taken directly from the latestrelease of the BUGLE-96 data library. RAMA calcu-lates a weighted fission spectrum based on the relativecontributions of the fuel isotopes that is used in the

,transport calculation. -, . ,

4.3 Core NeutroniSource

The Hatch-1 'core neutron source for both theDORT/TORT and RAMA transport calculations wasdeterminied using the cycle-specific BOC and EOCthree-dimensional burnup distributions. Thesource

-distributiois accounted for the radial power gradient inthelfuel as'semblies ioaded near the core boundary bymodeling ihe pin-wise source distributions in the outerthiee rows of fuel assemblies. I Th'e magnitude of theneutroni s6urce accouits for the contributi6n of pluto-niuum fissions as alfunction of fuel assembly exposure.

4.3.1 DORT/TORT SourceCalculation ,- ' i-

The DORT/TORT source distribution was specifiedcycle-wise for each assembly and for each fifteen cmaxial node. The MESH code was used to allocate thepin-wise power to the individual DORT (r, 0) meshblocks. This allocation was performed by a numericalintegration of the power distribution, defined on the

4-1

(x, y) pin-wise mesh, over the (r, 0) mesh block; Thisnumerical integration typically employed 2100 integrationmesh points per fuel pin and was shown to be accurate towithin S 1% for each (r, 0) mesh block.

The magnitude of the neutron source increases with fuelburmiup due to the higher number of neutrons released perMeV of energy produced by Pu fission. This was takeninto account by calculating the number of neutrons perMeV, V/K [neutrons/MeV], using the fuel burnup depend-ent isotopic fission fractions provided by SouthernNuclear Operating Company and given in Table 4.2.1. Inaddition, the fission spectrum was considered to bedependent on the fuel burnup in order to account for theharder more penetrating neutron spectrum characteristicof the Pu fissions in the high burnup fuel. This exposuredependence was also determined using the Table 4.2.1data and was included in the DORT source distribution.

4.3.2 RAMA Source Calculation

The neutron source for the RAMA transport calculation iscalculated in RAMA using the input power density factorsfor the different fuel regions and data from the RAMAnuclear data library. The' fission spectrum used in theRAMA source calculation is described in Section 4.2.2.The RAMA nuclear data library provides the cross sectiondata (including fission cross sections), neutron releasefactors (v), and energy release factors (K) needed toconvert the input powers to source terms.

The RAMA core model is constructed to replicate thenodal geometry used in core simulator codes and processcomputers, including appropriate representations for thepin-wise power and source term distributions in the outerrows of fuel assemblies. In the current analysis, BNLprovided three-dimensional relative power distributionsfor the core region and pin- wise power distributions forthe fuel assemblies for each Haich- "operating cycle.Appropriate uranium and plutonium number densities forthe fission spectrum calculation were not available forRAMA, therefore fuel assembly depletion calculationswere performed using a fuel assembly design typical ofthe later Hatch-i operating cycles. Using' the cycle-dependent exposure data provided for Hatch-I, uraniumand plutonium number densities for each fuel region werecalculated from the fuel assembly depletion data and usedin the source calculation.'

4.4 ' DORT/TORT Neutron Trans-port Calculations (BNL)

4.4.1 Neutron Transport Method

The DORT/TORT transport calculations were per-formed in a fixed source mode for a radial (r, 0) plane,an axial'(r, z) plane, and in'a 'one-dimensional (r)geometry. The calculational model represented a one-eighth azimuthal sector of the core and vessel geometry.The Hatch-i planar and axial geometries used in theDORT analysis are given in Figures 4.1 and 4.2, respec-tively.

The transport calculations were performed using 'an Ssquadrature and 'a P-3 arigular decomposition of thescattering cross-sections. The (r, 0) mesh included 72angular mesh intervals, and 226 radial mesh intervals.The'angular (0) and radial (r) mesh densities wereincreased at material interfaces where the geometry waschanging rapidly and at the measurement locations. Theradial mesh used in. the axial (r, z) calculations wasessentially identical to that used in the (r, P) calcula-tions. The axial model included 179 mesh intervals forCycles 1-4 where the fuel length was 144 inch and 183mesh intervals for Cycles 5-19 where 150 inch fuel wasused.

Vacuum boundary conditions were'used on the outerradial and axial boundaries of the problems and reflect-ing boundary conditions were used on the internal 0 =O0 and 0 '= 45° azimuthal boundaries. A pointwise fluxconvergence of 103 was used together with an inte-grated flux convergence criteria of i3.

4.4.2 Flux Synthesis Method

The vessel fluxes were determined using the fluxsynthesis method of NUREG/CR-6115, BNL-NUREG-52395 (Reference-13) to combine the DORT (r, 0) aridDORT (r, z) calculated fluxes. The flux at the (r. 0, z)location was determined' by'the relation

4)/r, 0, z) = [4)(r, 6I4,W(r)I.(r, z), (1)

where fg(r, 6), 4(r, z) and 4(r) are the group-gfluxes calculated in the indicated geometry. ,Theeffective core radius Reff used in the cylindrical geome-try (r, z) and (r) calculations was determined so that theflux above 1-MeV, ¢,X, at the vessel inner-wall satisfies

4-2

the condition - z. _R I . . . b .. . _ .

0 X (r; Rff ) = 2 - | 0 I,(r, ZdOS. (2)

The radial source distribution in these cylindrical calcula-tions was determined as an azimuthal average of the (r, 0)source distribution.

4.4.3 Dosimetry Cross Sections

The reaction rates for the riser brace pad samples weredeterminedbycounting and analysis stPNNL(Reference-3a and Refererice-3b).'BNL calculated a 47 group spec-trum for the Hatch-i plant'at the'radial location of thebrace pad using DORT and the BUGLE-96 library. Fastreaction rates were determined for the dosimeter measure-ments listed in Table 2.1 using ENDF/B-VI cross sectiondata and the BNL supplied Hatchli spectrum.' PNNLcalculated the'fast (>0.1 MeV and >1.0'MeV) crosssections (Reference-3b) based on theENDF/B-VI dosime-try file collapsed in 100 groups using the BNL Hatch-lspectrum. The cross secti6ns'fdi the'thermal'dosimetryreactions of Table 2.1 were calculated at BNL by averag-ing the MATXS-12 library (Reference-14) reaction crosssections over the thermal portion of the BNL Hatch-I 69group neutron spectrum at the location of ihe' brace padsamples.

4.4.4 Thermal Fluence Calculation

The ratio of the thermal fluence to the fast >1 MeVfluence at the measurement location was determined usinga detailed one-dimensional model of thedowncomer/brace-pad interface. The calculations wereperformed with BUGLE-96 cross sections and includedup-scattering in the thermal groups. The neutron spec-trum was edited at the brace-pad depth of the measureddosimetry samples, and the ratio of the E < 0.414 eVthermal fluence to the E > 1 MeV! fast fluence wasdetermined. This ratio was used as an ad-hoc multiplier,applied to the TORT calculated fast fluence, to determinethe thermal fluence to be compared with measurement.

..-. . .... ...

4.4.5 -;TORT Calculations of the Ef-fect of the Local Brace PadGeometry

The geometrical arrangement of the riser, brace andvessel introduces a three-dimensional spatial depend-ence into the transport calculation of the brace pad

-measurements.' In order to evaluate the effect of thelocal geometry on the measurements, three-dimensionalTORT calculations were performed with and -withoutthe jet punip brace present. The TORT model repre-sented an (r, 0, z) segment of the geometry including:(1) the radial geometry from the inside of the shroud tothe outside of the vessel (2) an azimuthal section from00(at the core flats) to 8.20 (passing through the centerof the neighboring jet pump) and (3) a '30.5 cm axialsection centered on the jet pump brace.-

The TORT calculation was performed using an (x, y, z)model in which the cylindrical riser and jet pump 'cross-sectional areas were preserved. The shroud and vesselwere considered to be flat, since the amount of curva--ture introduced over this small section ( £ 1 cm radialdeviation from linearity over the entire brace pad) hasa negligible effect on the calculated fluence ratios.'TheTORT geometry is shown in Figures 4.3-4.5. Source

*boundary conditions, determined using a full three-dimensional DORT synthesis model, were specified onall six external surfaces. ' '

The results of the TORT calculations by BNL with andwithout the riser brace and brace pads included arepresented in Table 4.2;2 . The calculated adjustmentfactor (BIA) is the ratio of: (1) the calculated fluence atthe location of the samples on the riser brace pad withthe full riser brace and pad geometry included and (2)the calculated sample fluence without the riser braceand pad present; '

To account for the local three-dimensional geometryeffects of the riser bracelpad geometry on the DORTmeasurement calculations, the B/A ratio of the TORTpredictions with and without the riser brace and padspresent wkas applied as'a multiplicative adjustment to theDORT calculations. - .

., . .. >'. ..- , - . . -

4-3

4.5 -RAMA Neutron Transport Cal-culations (TVE)

4.5.1 Neutron Transport Method

The RAMA Fluence Methodology (TWE) supports a fullthree-dimensional transport solver that provides directsolutions of the neutron flux, fluence, and activation. TheRAMA geometry. modeling capability is based uponcombinatorial geometry techniques. This allows models tobe built with accurate representations of componentshapes, dimensions, and position in the reactor, includingthe mixing of rectangular and cylindrical body surfaces.

The transport calculation is based upon a three-dimensional deterministic volume integral technique withtreatments for fixed-source and anisotropy. The user mayselect the angular quadrature for the transport calculation.An Ss quadrature was used for this analysis. Anisotropy istreated with high order angular decompositions of thescattering cross-sections. In this analysis, the heavy nu-clides are treated with a P-S and lighter nuclides with a P-7 Legendre expansion of the scattering cross-sections. Apointwise flux convergence criterion of 10.2 was used inthe transport calculation.

The RAMA transport methodology includes integratedcapabilities for calculating the fission spectrum andneutron source for the transport calculation from commonuser input parameters and information provided in thenuclear data library.

The RAMA methodology calculates fluence and activa-tion using the neutron flux distributions calculated by thetransport calculation, isotopic activation and decay chains,and reactor daily operating history.

The RAMA methodology includes a nuclear data librarythat contains cross-section data and nuclear constants forseveral reactor material nuclides. The. RAMA cross-section data is derived from the BUGLE-96 data libraryand are represented in the same 47-neutronl20-gammaenergy groups. The RAMA data library includes addi-tional cross-section data for the ?39Pu, .`Pu, 214 Pu, and242Pu fuel isotopes and other nuclear data constantsincluding energy release per fission (ic) factors that areneeded for the RAMA source and transport calculations.

4.5.2 The RAMA Fluence Model-

The RAMA fluence model for Hatch-I is illustrated inFigure4.6. TheRAMAnmodel assumes octant symmetryin which the north-northeast octant is represented in theazimuthal dimension. Assuming mirror reflection alongthe diagonal boundary, the RAMA model solves thesame problem as the east-northeast DORT/TORTmodel illustrated in Figure 4.1 (not to scale).

Vacuum boundary conditions are used on the outerradial and axial boundary surfaces of the problem.Reflective boundary conditions are used on the internalsurfaces for the 0 and 45 degree azimuths. Figure 4.6shows the basic reactor geometry near the axial eleva-tion corresponding to thejet pump brace assembly. TheRAMA model uses square rectangular bodies to repre-sent the fuel assembly geometries and cylindrical bodiesto describe the ex-core regions and components.Appropriate combinations of bodies are used in the corereflector region between the core region and shroud todescribe the transition from planar to cylindrical geome-tries. Figure 4.6 also shows that the jet pump riser pipe,mixer pipes, and yoke assembly are appropriatelyrepresented in the downcomer region of the model.Assuming symmetry conditions, each of the jet pumps3, 5, 13, and 15 are appropriately accounted for in theRAMA model.

Figure, 4.2 (not to scale) provides an axial representa-tion of the Hatch-i reactor, including noting the axialelevations around the jet pump brace assembly. TheRAMA model used the same axial elevation iriformna-tion. The RAMA model agreed with that shown inFigure 4.2 with the following exceptions: (1) the jetpump assemblies were explicitly modeled in thedowncomer region between radial dimensions 225.425and 279.559 cm and (2) the axial extent of the RAMAmodel spanned the elevations 651.6624 to 834.5424cm, relative to reactor vessel zero. It was determined byparametric'study that this axial elevation range modeledin the RAMA transport code produced an asymptoticsolution at the brace pad elevation of 742.95 cm (shownin Figure 3.3.3).

Figures 4.7 and 4.8 illustrate the detail of the Hatch-ibrace pad model used in the RAMA calculations.Figure 4.7 provides an axial view (side view) of the jetpump riser brace assembly. The brace assembly wasmodeled in the correct geometrical form with thefollowing exceptions: (1) the upper leaf was homogen-ized with the water above the leaf to the top of the brace

4-4

pad to avoid small axial heights in the flux model; (2) thelower leaf was homogenized with the water below the leafto the bottom of the brace pad to avoid small axial heightsin the flux model; and (3) the yoke was expanded axiallyto coincide with the axial extent of the brace assemblyafter the adjustments in (1) and (2). Note that the bracepad is recessed into the reactor, pressure vessel (RPV)cladding to interface with the RPV wall. Figure 4.8provides a front view (looking out from the core region)of the jet pump riser brace pad.This figure shows that thefull dimensional size of the brace pad is modeled and theradial and axial location where the sample was extracted.

The three-dimensional production model for the RAMAanalysis contains 30,108 mesh regions. This model wasused in the RAMA transport &lcuilatidris for each of the -

:19 cycles of operating history data.

4.5.3 Dosimetry Cross Sections

The RAMA reaction rates for the fast activation reactionswere determined by TWE for the dosimeter measurementsgiven in Table 2.1 using the flat spectrum weightedresponse functions provided in the RAMA nuclear datalibrary. The RAMA response functions were derived fromthe BUGLE-96 data library. The thermal reaction crosssections were determined by collapsing the 199 groupVITAMIN-B6 (Reference-15) reaction cross sections tothe 47 group BUGLE-96 structure using a BWRdowncomer spectrum. 9.

-1-1 ----

I

'I , i, -I., - �." - � I .", 11, f . I

4.5.4 Thermal Fluence Calculation

The thermal fluence at the measurement location wasdetermined using a detailed model of the downcomer, jetpump assembly, brace assembly, and brace pad interface.The thermal calculations included up-scattering in thethermal groups. The thermal fluence was edited forenergy <0.414 eV at the brace pad depth of 0.1524 cm ofthe measured dosimetry samples. The thermal fluencecalculated directly by the RAMA Fluence Methodologywas compared with the measurements.

. i I .

I I

... ....- ---. - . - -

4-5

Table 4.2.1 - Fraction of Fissions by Isotope as a Function of Fuel Exposure.

I .. .. . .

,FUeI Exposure U. 35 . U Pu- 39 , P24 ,~(GW D/t).. ----____--

..0-- 0.9247 0.0753 - - 0.0000 0.0000

5.0 0.6823 0.0770 '. 0.2243 0.0164

10.0 0.5460 0.0818 0.3381 0.0341

15.0 0.4432 0.0836 0.4143 . - 0.0589

20.0 0.3590 0.0886 0.4637 0.0887

25.0 - 0.2843 0.0887 0.5056 0.1214

-';30.0 0.2169 0.0924 0.5398 0.1510

35.0 0.1472 0.0942 . 0.5663 0.1923

Table 4.2.2 - TORT Calculations of the Effect of the Riser B raceand Brace Pad on the Dosimetry Measurements.

'.--:.A -'Respon at: B-- Response at-- - Ef ectof Brace andRESPONSE-. -Sampb Location w/o: Sample Location with-. i;.: Pad on Response-,F;

.t,, , ,,;_, . ,,. RiserBrace and Pad RiserBrace and-Pad' -B IA-- Adjustment '.

>; 1.0 MeV (nkc 2-sec) 1.1323 +9 1.1660 +9 1.0298

>0.1MeV (n/cm2 -sec) 2.0857 +9 -2.2155 +9 1.0622

4-6

.. . .. .- , . .-} ,r"-+ ', , . . I . t

. , . - . . .

.. . . .

... .. . .

. . . .

.

--' Hatch-1 Planar Geometry; . . . . .......... ........ .

.

X- Left -(

0

0..

NOTE:

.7

- Ii1 11 131 .X1 fflgsrv : I

L-.. ..L.z. .4...~• _ 0., c, -g -cm CD ge _ O 0

-'* LO C. C- U O 0 : '

-: CD . -ZD ' CN M:' C~lCj J N .J : 'lC) O C) C

', ' cm CM 'I ;,3 .'; .' .

Ic.

. 4.

-. C

Figure 41. Hatch-i Planar Geometry

:4-7

Hatch-1 Axial Geometry(Core flats at 0 = 00)

... , ... ::: ... , . : . .: .: . hi: : . .. : i ,..,..., ...., .,....,.-,.. ,, .' ,''' '.' '' ',

. . - ...... ., ....... .- . . ... ... .. ..: ;: . . . ; ... : ,: ; : ,:. : .,, : ... : ., :., ,: : 8.,

'', . .'.., STEAM SEPARATOR AND DRYER REGION ,' '.-"'.;. ' , . '; *.;. ', .. , * .>, . . .' ' .: , . r: '-. W '; ' , " X ,, ' . � '," .' -; . . .. ". X 'r * , ." *, . .

*. e. .;.: .¢,*.'. ,- :^. .' *,.-,,-.''. .'. '. :.-..'.,'.* -; - -*-* La -*.- - * ' . . ' r,' ' . .' ' '.' ' , .': ' '. .- ' ' - .' r. -. * .- '

t ;w i- ; ;

In ;s Am- sir .r.

U

E-0

C) C "' .- e'J U1, 4=c-

LO co a) j c- Je'JCD Cn 7

RADIUS (cm)

Figure 4.2. Hatch-i Axial Geometry

4-8

66.06 cm -I

i

Shroud 35.22 cm

II.,

. 1

Figure 4.3. Top View of the Hatch-i TORT Riser and Brace Model

4-9

Clad

BraceYoke X

Riser

Pressure Vessel

Figure 4.4. Side View of the Hatch-i TORT Riser and Brace Model

4-10

. : 4 4 -, ,. ri I -J

- 66.06 cm;, , ; ..; I

. I

I

I-- Clad---

z..

Shroud

Brace Yoke

--- 30.48 cm

I I

Brace Leaf Brace Pad

Figure 4.5. Horizon'tal View' of the Hatch-i' TORT Riser and Brace Model

4-11

North 00

Measurements 5 & 15

(Q) Measurements <

Downcomer

Core Shroud

Core ReflectorCore Region

Drawing Is not to scale.

Figure 4.6. Top View of the Hatch-i RAMA Model Near theJet Pump Riser Brace Assembly Elevation

3& 13

4-12

Core Shrou

/ /_ , . . .

_P I :,.g.',.-; . .

:....-.:

.,

.''.'' ,1':''s.

-.':

'..-:.. � ,.# *'''

".

;

.....

I

..7

.

"'I

d Jet Pump RiserDowncomer /

Yoke Leaf + Water;.. o --e

w/ J F

RPV Clad

: ,; 1- .I

. _ .I.

' i 0.9398

I �

0 9906i-,S,# ? - ---------- zz/YzXzY< zzo/1a

--- ElevationsRelative to

Bottom of RPV

.,- 748.665

_- -742.95

4- 737.235

11.43

.I.

Sample /Location

Brace Pad

_. n..... I 1.wa-

I 1 Di 225A250

4i l

-� I �

. . 246.8118-.1 I. I

+! 1

278.6063279.5588280.3526-

Note:AII Dimensions In cm.

Figure 4.7. Axial View of the Hatch-i RAMA Jet PumpRiser Brace Assembly Model

.4-13

s * E. -

ElevationsSample Relative toLocation Bottom of RPVLeaf + Water

_ , | _ _ _ - - _ _/- 748.665

7.5692 742.95

11.43L

737.235

16.51

Note: All Dimensions in cm.

Figure 4.8. Front View of the Hatch-i RAMA Jet Pump Riser Brace Pad

4-14

5 COMPARISONS OF CALCULATIONS AND MEASUREMENTSz - ': : ~~~~~~~~. ............... ... Ot"J-.2..

5.1. Introduction

The ,Hatch-I measurement, program has provided athermal and fast neutron dosimetry data-base forbenchmarking BWR fluence calculation methodologies.'Detailed calculations of the fast and thermal fluence ofthe brace pad scrapings have been performed using theDORT/TORT discrete ordinates transport methodologyand the RAMA Fluence Methodology.

Comparisons of the DORT/TORT (performed by BNL)and RAMA (performed by TWE) predictions with thebrace pad measurements were made to provide a directassessment of the accuracy of the methodologies.. Inaddition, comparisons of the DORT/TORT predictionswith the RAMA methodology were made, to provide acomparison of the two calculational methods. TheRAMA methodology also calculates activities forirradiated specimens that are compared to measurements.

5.2 Brace Pad Dosimetry andFluence Comparisons

The DORT/TORT and RAMA fluence predictions arecompared with the measurements in Section 5.2.1 andthe DORT/TORT to RAMA inter-code comparisons arepresented in Section 5.2.2. Additional comparisons ofRAMA predictions to activation measurements arepresented in Section 5.2.3.

5.2.1 Comparison of Fluence Calcu-lations with Measurement

Before making the comparisons, it is first noted that themeasurements corresponding to Jet Pump-3 and 13 arelocated at diametrically opposite locations on the vessel.Similarly, the locations of the Jet Pump 5 and 15 mea-surements are also diametrically opposed. Except forslight asymmetries (s2%) resulting from core loadingstrategies, these symmetric measurements should beidentical. However, a review of the Hatch-i measure-ment data indicates that, while the symmetric locationsgenerally indicate an asymmetry of s 9 ± 2 %, thesymmetrically located thermal measurements for JetPumps 3 and 13 differ by 30%. In addition, while thethermal-to-fast fluenceratios for the JetPump-13,5 and15 measurements are all in close agreement (to within

3%),' the thermal-to-fast ratio 'for the Jet Pump-3measurements differs by.209;o. 'Consequently, the'JetPump-3 measurement is considered unreliable and hasbeen eliminated from the MIC com'parisons. TheFe-58(n, y) Fe-59 thermal data for the symmetric Jet Pu'mp 5and 15 measurements indicates a 30% asymmetry andhas also been eliminated from the MWC comparisons.'

The comparisons of the DORT/TORT calculations withthe brace pad 'measurements are presented in Tables5.2.1'and 5.2.2.- The DORT/TORT calculations of thefast fluence (i:e., >1 MeV and >0.1 MeV) are seen toagree with' the eight measurements with an averagemeasurement-to-calculation ratio of MIC = 0.97, +

.083. 'The comparison of the DORT/TORT thermal'fluence'calcu'lation and the three measurements' givesan average MIC of 1.17 + 0.07. The measured thermal

' values were reduced by 0.7% to make the 'adjustedmeasured value correspond to the group boundarytusedin ihe calculation, namely 0.414 eV. The 0.7% reduc-tion factor was the calculated flux contribution (from0.414 eV to 0.5 eV) compared to the calculated total'thermal flux (from 0 to 0.5 eV). These M/C differences'are within the estimated combined -1 5% DORT/TORT-calculationr uncertainty 'and -10-15% measurementuncertainty.

TheRAMA calculations of. the brace pad neutrondosimetry measurements are presented in Tables 5.2.3and 5.2.4. The RAMA calculations of the fast fluences(i.e., energies >1 MeV and >0.1 MeV) agree with the

' mesasuremeants with an average measurement-to-calcula-tion atio 'of MWC = 0.93 _ 0.04. Comparison of theRAMA thermal fluence calculation with the brace 'pad

'miisureme nt indicates an average WC of 0.62 '0.002.' The WC differences for the fast fluences areconsidered to be consistent with the estimated cal6ula-tion and measurement uncertainties. The M/C differ-ences for the thermal fluence is discussed 'in'-Section

" 5.2.3.: "'* - : :

'.5.2.2 ;..Comparison of DORT/TORT- and RAMA Fluence Results,.

.. 'u .. ..A .-

3A1l + differences are understood to be one-sigmavalues.

5-1

The comparisons of the DORT/TORT (BNL) andRAMA (TWE) fluence predictions of the Jet Pump-3/13and Jet Pump-5/15 measurements are presented in Tables5.2.5 and 5.2.6, respectively. The fast neutron fluencecomparisons indicate an average code-to code ratio ofDORT/RAMA = 0.97 ± 0.04.. The thermal neutronfluence comparisons indicate an average code-to-coderati6'of DORT/RAMA - 0.52 t 0.06. The fast fluencedifferences between the DORT/TORT and RAMA arewithin the estimated uncertainty of the two calculations.

The calculated thermal fluence comparisons for bothmethodologies do not show the consistency of the fastfluence predictions. Additional results edited from theRAMA methodology are presented in Section 5.2.3.

The. DORT/TORT calculations showed smaller differ-ences between the thermal fluence predictions and thethermal fluence measurements. The RAMA thermalfluence calculations showed larger differences in corn-parison to measurements. The DORT/TORT thermalfluence results were an under prediction of the fluence,while the RAMA results were an over prediction.Relative to the measurements, the DORT/TORT thermalfluences presented in Tables 5.2.1'and 5.2.2 were -15%lower on average, and theTWE RAMA results presentedin Tables 5.2.3 and 5.2.4 were -60% higher on average.

thermal activations is 0.59 E 0.05; Table 5.2.9 showsthe averages for all measurement comparisons. Theaverage M/C ratios for the fast activations is 1.02 ±0.13 and for the thermal activations is 0.59 ± 0.05.Measurement results for certain nuclides have beenomitted from the comparisons (identified with an "n/a"in the tables) in accordance with discussions presentedin Section 5.2.1. The niobium-93' meta-stable dataprovided in the BUGLE-96 data library has beendetermined to be unreliable and has also been omittedfrom the tables.

TWE believes the cause of the low M/C ratios forthermal activations could be attributable to theBUGLE-96 data library upon which the RAMA nucleardata library is based. The BUGLE-96 data has beenderived for fast fluence predictions using spectraweighting and a group structure that may not be appro-priate for the RAMA code's thermal predictions at thelocations where the brace pad measurements weretaken.

5.2.3 Comparison of RAMA ActivityCalculations with Measure-ments

The RAMA Fluence Methodology performs a threedimensional calculation that produces direct solutions(i.e., no multiplicative factors are used) for activationsand neutron fluence. The comparisons of the RAMAcalculated to measured activations are presented inTables 5.2.7 - 5.2.9.

Table 5.2.7 shows comparisons of the symmetric JetPump-3 and 13 activation measurements to the RAMAcalculated measurements. The average measurement tocalculated (MIC) ratios for the fast activations is 1.03±0.16 and for the thermal activations is 0.57 ±' 0.05.Similarly, Table 5.2.8 shows comparisons of the sym-metric Jet Pump-5 and 15 activation measurements tothe RAMA calculated measurements. The average M/Cratios for the fast activations is 1.02 t 0.13 and for the

. .I .

5-2

Table 5.2.1 - Comparison of the Jet Pump-3 and 13 Measured and DORT/TORTCalculated Fluences (x 10-'7 n/cm2).

| Fuence - JetPup-3 -| -Jet ump-13 |` Average 'Calculated e Measurement/.-Energy -Me mnt me1 Measuremenit -Measiremient n'Fluence Calculation"

-E 3i7MeV-62 8.02 1.00

|.E>01MeV i.8 14.3 1.1 ' ' 14.1 1.07

E<041ieV NA 13.3 133 10.5 .1.27

I - . 4

Table 5.2.2 - Comparison of the Jet Pump-5 and 15 Measured and DORT/TORTCalculated Fluences (x 10-'7 n/cm2).

rFuence ':.-Jet Pump-S :: Jet Pump-15 -- Average, -Calculated; Measurement/:Energy,. Measurement' Measurement M asurement Fluence -. Calculon '

., * ,, ',' ",' _ _, _ _ _..' -. _ .1 _ ,. :._._ _ _ _

.E>1-MeV§- 4.88 4.53 4.71 5.38 0.88

E > 0.1'MeV-- --9.10 *8.47 8.79 - 9.48 -0.93

E <O.414 eV 7.99 - 7.88 7.93;> . 7.06 1.12

I -

� I 4

* 4

.5-3

Table 5.2.3 - Comparison of the Jet Pump-3 and 13 Measured and RAMACalculated Fluences (x i0-'7 n/cm2 ).

Fluence ,. Jet mp-3 : Jet p -13. | A'verage, |Calculated-' . Mes iirel' e'ftEnr IF-~~I- Calculatiofi- Flunrgyence.=- Energy- I ~~~~~Measuremnent' .Measuireme'nt.-I MeasureimentF ~-Fune '.acuain>

E>, l-MeV 8.38 7.62 8.00 | 8.50 | 0.94

-E >'0.1 MeV-- 15.8 14.3 15.1 15.3 0.98

E <0.414 eV: NA 13.3 13.3 21.6 0.61

Table 5.2.4 - Comparison of the Jet Pump-5 and 15 Measured and RAMACalculated Fluences (x 10-7 ncm2 ). -

Fluence : Jet Pmp-5 Jet Pump-15-. 'Average';-| Calculated,- -MeasurementU',Ener. y. .Measurementr Measurement Measurement' Fiuence' .- Calcultion

.,,,,,, -.t cui Io

_E> 1-MeV 4.88 4.53 4.71 5.29 J 0.89

E> 0.1 MeV 9.10 8.47 8.79 9.63 0.91

E < 0.414 eV 7.99 7.88 7.93 12.9 T 0.62

5-4

Table 5.2.5 - Comparison of the DORT/TORT and RAMA Predictions of theJePup3ad1I ~~~~Measurements xI 10.1 jn/cj e Pm- ad1

-Fuence J DORT/TrORT-.J RAMA9 } Ratio DORT/TrORT-'-_____ ____ ____ ___ _ _ _ _ _ _ _ _ _3_ _An d3 13W nergy Pmp 3 an~ .PMP3nd } -

802 8.50 0 .94'..- -

.',,E >0.1`MeV I 411 - -- 15.3 - .. 0.92

.EO44V..-.10.5 21.6 ->0.49

- I . , , , i iI . I .

-Table 5.2.6-Comparison of the DORT/TORT and RAMA Predictions of the Jet Pump-5 and15-Measurements (x I0'- n/cm2)

_ :lbe.2 0MA f'Rati DORTfTORT-Tluen& -.DORT/TOR up; n 5-~ oRMEnryPump-5S and 15 Pun'5aid1

E>1MV5.38 5.29 1.02

E> 0.1 ,MeV ,9.48 9.63 0.98

E 044V7.612.9 10.55

5-5

Table 5.2.7 - Comparison of Jet Pump-3 and 13 Activation Measurements (tLCi/mg) to RAMA CalculatedActivations

/Product. Jet ump-3 '-Jet P'rmp-13 : Avera'ge Me - Calculated ' Mearement/.Nuclide ; MeasMeasure mea's rement surement -Ac ivation - 'Calculation

,--Thermal . , ., . ,,

.,,- Fe-55 n/a 7.34E-03 7.34E-03 1.15E-02 0.64

Ni;63 - n/a 1.58E-02 1.58E-02 2.82E-02 0.56

|-, Fe-59j n/a 1.14E-04 1.14E-04' 1.82E-04 0.62

' Cr-51 ' n/a 4.84E-02 4.84E-02, 9.28E-02 0.52

n/a 5.18E-03 5.18E-03 9.79E-03 0.53

-_ _ _ _ _ _ :. _ , _ _ Average 0.57

| _ Std. Dev. 0.05

Fast..:

. n/a n/a n/a n/a n/a

, n-54 5.30E-04 5.59E-04 5.44E-04 4.69E-04 1.16

Co- 8' 4.71E-02 4.32E-02 4.52E-02 5.03E-02 0.90

Average 1.03

--- . , Std.Dev. 0.16

5-6

Table 5.2.8 - Comparison of Jet Pump-5 and 15 Activation Measurements ([LCi/mg) to RAMA Calcu-lated Activations

i

~PoUct .7 4 e upS~IJtPin-5 I vr e;[2Calulated MesremeneViNul6 ~ sie~tj-esieint msureiiet .f-A:Cilv~atiOn "Xa1cu'lation.

|Th..ermal

=Fe-55- 4.48E-03 4.52E-03 4.50E-03 7.02E-03 0.64

N --63 9.1 IE-03 8.99E-03 9.05E-03 1.72E-02 0.52

-Fe-59.i, n/a n/a n/a n/a n/a

Cr-51 2.90E-02 2.70E-02 2.80E-02 4.99E-02 0.54

_'rCo- ' 3.66E-03 3.54E-03 3.60E-03 5.74E-03 0.62

._.___________ A verage 0.59

-____.___._.Std. Dev. 0.05

,- F a s t>-_ _ _ _ __.

;Nb-93m n/a n/a n/a n/a n/a

Mn-54 , 3.1 1E-04 3.25E-04 3.18E-04 2.85E-04 1.12

Co-58 2.77E-02 2.68E-02 2.72E-02 2.98E-02 0.91

_ _ _ ; _- Average 1.02

_ _ _ Std. Dev. 0.13

Table 5.2.9 - Comparison of the Activation Measurements to RAMA Calculated Activities

ie m se tlMeisurements(MC)- ' )T- a I t Ii~ ..(MC

iF - N i-6 F & 9 C - 1oC- O A e j " t ~ M S5 8 ." td.e50-3 1- X&59'. co i .% - o- Averagec

'-4 - i' -' |iD . -,- iD ev.

.64 .54 .62 .54 .58 .59 10.5 E1.14 0.90 1.02 0.13

5-7

6 SUMMARY AND CONCLUSION

In order to provide a benchmarking and assessment ofBWR fluence calculation methods, detailed calculationsof the Hatch-1 jet pump riser brace pad fast and thermalneutron dosimetry measurements have been performedusing the DORT/TORT discrete ordinates transportmethodology and the RAMA Fluence Methodology.The DORT/TORT and RAMA calculations have beencompared with the measurements and with the predic-tions calculated with each methodology. Bothmeasurement-to-code and code-to-code comparisons ofthe calculations have been performed to assess theaccuracy of these methodologies for predicting the fastand thermal neutron fluence of BWR internal compo-nents and the vessel.

The calculations were performed using accepted(DORT/TORT) and state-of-the-art (RAMA)calculational methods together with the BUGLE-96nuclear data library. The DORT/TORT and RAMAcalculational models include a detailed description ofthe Hatch-I corelintemals/vessel material and geometri-cal configuration. The models include a cycle-by-cycledescription of the Hatch-1 Cycle 1-19 operating history.The core neutron source includes the effects of the pin-wise power distribution on the core periphery and theeffects of Pu buildup on the magnitude and energydependence of the neutron source.

The DORT/TORT methodology (BNL) was found topredict the fast fluence measurements to within -5%and the thermal fluence measurements to within -15%.The observed agreement between the DORT/TORTcalculation methodology and fluence measurements isgenerally considered to be consistent with the uncertain-ties in the calculations and measurements.

The observed agreement for the fast fluence betweenthe RAMA calculation methodology and measurementsis generally considered to be consistent with the uncer-tainties in the calculations and measurements.

The DORT/TORT-to-RAMA differences for the fastfluence measurements are -4%. The observed agree-ment between the fast fluence calculations and measure-ments and between the code predictions is generallyconsidered to be consistent with the uncertainties in thecalculations and measurements.

This report has described work performed under anAddendum to the NRC/EPRI Memorandum of Under-standing (signed January 11, 2000) to remove andcharacterize samples from BWR in-vessel components.Under this cooperative agreement, each party wasresponsible for funding and administering certainactivities. The NRC contracted BNL to perform workwhich included the following objectives: calculate fastand thermal fluences at the belt-line region of the core;compare the calculated values with fluences based onmeasurements; compare calculated BWR fluence resultsfrom a well-established computer code (DORT/TORT)with those from a newer, state-of-the-art, code(RAMA), and document the results. While the twocodes were in good agreement for the prediction of fastfluence, there was a significant discrepancy in thethermal fluence predictions. With the completion ofthis report, the objectives of the BNL efforts have beenaccomplished. However, EPRI intends to initiate anactivity later in 2004 to identify the source of the over-prediction of thermal fluence by RAMA.

The RAMA fluence methodology was found to predictthe fast fluence measurements to within -7%. TheRAMA thermal fluence predictions exceeded measuredvalues by approximately 60%. Similar trends wereobserved in comparing fast and thermal activationpredictions to measurements. That is, predicted fastactivations were in much better agreement with mea-surements than were predicted thermal activations.While TWE currently believes that the differences in thethermal spectrum relative to the fast may be attributableto the derived BUGLE-96 thermal cross-section dataused in the analysis, the reason for the differencesrequires further investigation.

6-1

7 REFERENCES

1. V. N. Shah and P. E. McDonald, eds., "Agingand Life Extension of Major Light WaterReactor Components:' Elsevier SciencePublishers B. V., Amsterdam, Netherlands,1993.

2. Office of Nuclear Regulatory Research,"Calculational and Dosimetry Methods forDeterminingPressure Vessel Neutron Fluence,"Regulatory Guide 1.190, U.S. NuclearRegulatory Commission, March 2001.

3. (a) L. R. Greenwood, "Retrospective NeutronDosimetry for the Plant 1 Reactor," Attachment-A and "Retrospective Neutron Dosimetry forthe Plant 2 Reactor," Attachment B to "Helium,Boron, and Radiometric Analysis of ReactorSteel-Plant-i, Project No. 40645 (JCN Y6249)," Letter, B. M. Oliver (PNNL) to W.Norris (NRC), dated August 16, 2001.(b) "Calculations:' Letterreferring to Hatch-i,L. R. Greenwood (PNNL) to John Carew(BNL) dated October 15, 2002.

4. W. A. Rhoades et. al.,"TORT-DORT Two- andThree-Dimensional Discrete OrdinatesTransport, Version 2.8.14," CCC-543,Radiation Safety Information ComputationalCenter, Oak Ridge National Laboratory, 1994.

5. D. B. Jones, K. E. Watkins and S. P. Baker,RAMA, Radiation Analysis ModelingApplication, Proceedings PHYSOR 2000, ANSInternational Topical Meeting, Pittsburgh, PA,May 2000.

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Electric Power Research Institute Report EPRINP-562, January 1979.

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7-1

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER(9-2004) (Assigned by NRC. Add Vol., Supp., Rev.NRCMD 3.7 and Addendum Numbers, If any.)

BIBLIOGRAPHIC DATA SHEET(See istrucons on the revrse) NUREG/CR-6887

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

DORT/TORT Analyses of the Hatch Unit-1 Jet Pump Riser Brace Pad Neutron Dosimetry MONTH YEAR

Measurements with Comparisons to Predictions Made with RAMA November 2005

4. FIN OR GRANT NUMBER

Y68835. AUTHOR(S) 6. TYPE OF REPORT

J.F. Carew, K. Hu, A. Aronson, A.N. Mallen, and M. TodosowTechnical

7. PERIOD COVERED (Inlustve Dates)

Sep 2003 - Jun 20048. PERFORMING ORGANIZATION - NAME AND ADDRESS (tfNRC. provide Dmsbn. Offxce or Regtion. .S. NudarRegulatotyCommission, and maiing address: It contractor,

provide name and matng address.)

Division of EngineeringOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001

9. SPONSORING ORGANIZATION -NAME AND ADDRESS (ItNRC, te Same as abe;ff contratorm provideNRCDsin, Office orRegion, U.S. NudCear Regulatoty Commission,and maiing addrss )

same as above

10. SUPPLEMENTARY NOTES

W.E. Norris, NRC Project Manaaer11. ABSTRACT (2X0 words oress)

Detailed calculations of the Hatch-1 jet pump riser brace pad thermal and fast neutron dosimetry measurements have beenperformed by Brookhaven National Laboratory using the DORT/TORT discrete ordinates transport methodology and byTransware Enterprises, Inc., using the RAMA three-dimensional fluence methodology. The calculations were performed usinga detailed description of the Hatch-1 core/intemals/vessel material and geometrical configuration. Comparisons of the fluencescalculated by DORT/TORT and fluences and activations calculated by RAMA and the Hatch-1 measurements were performedto assess the accuracy of the methodologies for predicting the fast and thermal neutron fluence of the BWR Internalcomponents and the vessel. Measurement-to-Calculation fluence comparisons were also performed. The core neutron sourceincludes the effects of the pin-wise power distribution on the core periphery and the effects of plutonium buildup on themagnitude and energy dependence of the neutron source.

12. KEY WORDS/DESCRIPTORS (List words orphrases that wil assistresearchersi loating te report) 13. AVAILABILITY STATEMENT

NUREGICR-6887 unlimitedfast neutron 14. SECURITY CLASSIFICATION

thermal neutron (This Page)neutron dosimetry measurements . unclassifiedBWR pressure vessel fluence , A-7hi5 Report)fluence methodology uncasDORT/TORT unclassifiedRAMA 15. NUMBER OF PAGES

6016. PRICE

NRC FORM 335 (9 200-4) PRINTED ON RECYCLED PAPER

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NUREG/CR-6887 DORT/TORT ANALYSES OF THE HATCH UNIT 1- JET PUMP RISER BRACE PADNEUTRON DOSIMETRY MEASUREMENTS WITH

COMPARISONS TO PREDICTIONS MADE WITH RAMA

NOVEMBER 2005

UNITED STATESNUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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