Monte Carlo Calculation of Dancoff Factors in Irregular Geometries

Fehér, S.; Hoogenboom, J. Eduard; Leege, P.F.A. de; Valko, J.

doi:10.13182/nse94-a21500

Cited by 25 publications

(5 citation statements)

References 4 publications

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“…There are several well-known analytical and numerical methods for the Dancoff factor determination implemented in the CSAS control module [12] and widely used in the calculations of the regular lattice nuclear reactors. In all these procedures, as well as in the recent ones based on the Monte Carlo techniques [13], determination of the Dancoff factor is based on its definition derived from the theory of the first neutron collision probability. For the most widely used arrangements, such as regular square or hexagonal lattices, the SCALE-4.4a code system provides Dancoff factors.…”

Section: Calculation Methodologymentioning

confidence: 99%

Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

Ljubenov

Milošević

2004

Serb J Electr Eng

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The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114

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Section: Calculation Methodologymentioning

confidence: 99%

Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

Ljubenov

Milošević

2004

Serb J Electr Eng

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“…This was accomplished by creating two HTTR models, one using the provided DOUBLEHET unit cell treatment available in SCALE 5.1, while the other bypasses the feature. Instead the D ancoff cor rection factor i s independently de termined by t he cod e s ystem DANCOFF-MC [40] and manually entered into the model as an external parameter. Table 4.11 provides a comparison of the results for the two different treatments of the heterogeneity effects.…”

Section: Benchmark Analysismentioning

confidence: 99%

“…The lag time variable for the AP1000 i ncludes 4 br eakpoints a t: [ 5,10,15,20] years. T he p acking f raction for T RISO particles in the VHTR represents the third input variable and its corresponding breakpoint set is: [20,30,40 ] percent. T he final input variable is the lag time associated with the VHTR, which has the breakpoint set of: [5,10,15,20] years.…”

Section: Integrated System Model For the Nuclear Energy Systemmentioning

confidence: 99%

High fidelity nuclear energy system optimization towards an environmentally benign, sustainable, and secure energy source.

Tsvetkov

Rodríguez²,

Ames

et al. 2009

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Approved for public release; further dissemination unlimited. 2Issued by S andia N ational Laboratories, operated for the U nited States D epartment o f E nergy b y Sandia Corporation. NOTICE:This report was prepared as an account of work sponsored by an agency of the United States Government. N either t he U nited S tates Government, n or an y a gency t hereof, n or an y o f their e mployees, nor a ny of t heir c ontractors, s ubcontractors, or t heir e mployees, m ake an y warranty, e xpress or i mplied, or a ssume a ny l egal l iability or responsibility for t he a ccuracy, completeness, or us efulness of a ny i nformation, a pparatus, pr oduct, or process di sclosed, or represent t hat its us e would not infringe pr ivately owned r ights. R eference h erein t o an y s pecific commercial p roduct, p rocess, o r s ervice b y t rade n ame, t rademark, m anufacturer, o r o therwise, does not ne cessarily c onstitute or i mply i ts e ndorsement, r ecommendation, or f avoring b y t he United States Government, any agency thereof, or any of their contractors or subcontractors. T he views and opinions expressed herein do not necessarily state or reflect those of the United States Government, any agency thereof, or any of their contractors.Printed i n the U nited S tates o f America. This r eport has b een r eproduced d irectly from t he b est available copy. AbstractA new hi gh-fidelity i ntegrated s ystem m ethod a nd a nalysis approach w as de veloped a nd implemented f or c onsistent a nd c omprehensive evaluations of a dvanced f uel c ycles l eading t o minimized T ransuranic ( TRU) inventories. T he method ha s be en i mplemented i n a de veloped code system integrating capabilities of Monte Carlo N -Particle Extended (MCNPX) for highfidelity fuel cycle component simulations.In this report, a Nuclear Energy System (NES) configuration was developed to take advantage of used fuel r ecycling and t ransmutation capabilities i n waste m anagement s cenarios l eading t o minimized TRU waste inventories, long-term activities, and radiotoxicities. The reactor systems and f uel cycle components t hat m ake up the NES w ere s elected f or t heir abi lity t o perform i n tandem to produce clean, safe, and dependable energy in an environmentally conscious manner. The di versity i n performance and spectral ch aracteristics w ere us ed to enhance T RU w aste elimination while e fficiently ut ilizing ur anium resources a nd p roviding a n a bundant energy source.A computational modeling approach was developed for integrating the individual models of the NES. A general approach was ut ilized allowing for t he Integrated S ystem M odel ( ISM) t o be modified in order to provide s imulation for ot her s ystems w ith similar a ttributes. B y ut ilizing this approach, the ISM is capable of performing system evaluations under many different design parameter opt ions. A dditionally, t he p redictive capabilities of the ISM a nd its c omputational time e fficiency a llow f or s ystem s ensitivity/uncertainty a nalysis a nd the impl...

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“…Several code packages are in use to perform this calculation [19]. We have also developed a Monte Carlo code for cylindrical geometries that closely reproduces the results of other codes [20] but with computational savings [21]. A plot of the Dancoff factor, for cylindrical pins in a square lattice, as a function of Rr and V 1 /V 0 , applicable to the test case used in this study is shown in Fig.…”

Section: Fuel and Moderator/coolant Escape And Transmission Probabilimentioning

confidence: 99%

“…Near the end of fuel residence time, it is possible for Eq. (20) to imply a negative control absorber concentration. Physically this corresponds to a local multiplication constant that is less than one: the burned-out fuel pin being modeled 'borrows' neutrons from adjacent fresher fuel for the chain reaction to be maintained.…”

Section: Determining the Lethargy Dependent Fluxmentioning

confidence: 99%

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

Schneider

Deinert

Cady

2006

Journal of Nuclear Materials

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Monte Carlo Calculation of Dancoff Factors in Irregular Geometries

Cited by 25 publications

References 4 publications

Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

High fidelity nuclear energy system optimization towards an environmentally benign, sustainable, and secure energy source.

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

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