• Title/Summary/Keyword: integral reactor

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Experimental Determination of Differential Fast Neutron Spectra in a Reactor using Threshold Detectors

  • Kim, Dong-Hoon
    • Nuclear Engineering and Technology
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    • v.4 no.4
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    • pp.280-293
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    • 1972

  • The differential fast neutron spectra above 0.5 Mev at particular spatial positions in tile reactor(TRIGA MARK-II) core has been determined experimentally using several threshold activation detectors. The series expansion technique utilizing the concept of least squares optimization was used to obtain an approximate solution to the set of integral equations which are defined by the experimentally determined activation data. The influence of use of different weighting functions in the solution was analyzed in each measurement. To carry out the necessary mathematical calculations, a computer code for the UNIVAC 1106 digital computer has been prepared. Good agreement was achieved between the differential fast neutron spectra determined in this work and the computed flux determined independently using space-independent multigroup transport theory.

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ASUSD nuclear data sensitivity and uncertainty program package: Validation on fusion and fission benchmark experiments

  • Kos, Bor;Cufar, Aljaz;Kodeli, Ivan A.
    • Nuclear Engineering and Technology
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    • v.53 no.7
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    • pp.2151-2161
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    • 2021

  • Nuclear data (ND) sensitivity and uncertainty (S/U) quantification in shielding applications is performed using deterministic and probabilistic approaches. In this paper the validation of the newly developed deterministic program package ASUSD (ADVANTG + SUSD3D) is presented. ASUSD was developed with the aim of automating the process of ND S/U while retaining the computational efficiency of the deterministic approach to ND S/U analysis. The paper includes a detailed description of each of the programs contained within ASUSD, the computational workflow and validation results. ASUSD was validated on two shielding benchmark experiments from the Shielding Integral Benchmark Archive and Database (SINBAD) - the fission relevant ASPIS Iron 88 experiment and the fusion relevant Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) mock-up experiment. The validation process was performed in two stages. Firstly, the Denovo discrete ordinates transport solver was validated as a standalone solver. Secondly, the ASUSD program package as a whole was validated as a ND S/U analysis tool. Both stages of the validation process yielded excellent results, with a maximum difference of 17% in final uncertainties due to ND between ASUSD and the stochastic ND S/U approach. Based on these results, ASUSD has proven to be a user friendly and computationally efficient tool for deterministic ND S/U analysis of shielding geometries.

  • https://doi.org/10.1016/j.net.2021.01.034 인용 PDF KSCI

Analysis of VVER-1000 mock-up criticality experiments with nuclear data library ENDF/B-VIII.0 and Monte Carlo code MCS

  • Setiawan, Fathurrahman;Lemaire, Matthieu;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • v.53 no.1
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    • pp.1-18
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    • 2021

  • The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article.

  • https://doi.org/10.1016/j.net.2020.06.015 인용 PDF KSCI

Optimal control of tubular reactors described by partial differential equations

  • Choe, Young-Soon;Lee, In-Beum;Soo, Chang-Kun
    • 제어로봇시스템학회:학술대회논문집
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    • 1992.10b
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    • pp.436-439
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    • 1992

  • A tubular reactor model represented by partial differential equations was studied as one of nonlinear distributed parameter optimal control problems. An optimal control theory in the form of maximum principles based on nonlinear integral equations was used to develop an algorithm to solve the problem.

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