• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.128.2-128.2
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• 2003

• Proceedings of the Korean Vacuum Society Conference
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• 2008.08a
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• pp.187-187
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• 2008

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.281-290
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• 2014

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.598-603
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• 1996

When locked rotor event. occurs, instantaneously affected loop and core flow were quickly reduced, which resulted in an increase in coolant temperature and system pressure. Analysis method of this event was that constant core inlet temperature and system pressure as well as change in core flow calculated from COAST code were statically used as an input variable to HERMITE code, because of no tools to simulate NSSS behavior and 1-D core neutronics transient coincidently. With employing TASS code revised with 1-D neutronics model, this event was analyzed in point of DNBR. By doing so, analysis procedure could be simplified and unreasonable conservatism might be removed in DNBR calculation by consideration of pressure increase.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.520-545
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• 2010

This paper presents the neutronics modeling capabilities of the fast reactor simulation system SHARP, which ANL is developing as part of the U.S. DOE's NEAMS program. We discuss the three transport solvers (PN2ND, SN2ND, and MOCFE) implemented in the UNIC code along with the multigroup cross section generation code $MC^2$-3. We describe the solution methods and modeling capabilities, and discuss the improvement needs for each solver, focusing on massively parallel computation. We present the performance test results against various benchmark problems and ZPR-6 and ZPPR critical experiments. We also discuss weak and strong scalability results for the SN2ND solver on the ZPR-6 critical assembly benchmarks.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1219-1225
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• 2017

In this work, a scalable algorithm for model calibration in nuclear engineering applications is presented and tested. The algorithm relies on the construction of surrogate models to replace the original model within the region of interest. These surrogate models can be constructed efficiently via reduced order modeling and subspace analysis. Once constructed, these surrogate models can be used to perform computationally expensive mathematical analyses. This work proposes a surrogate based model calibration algorithm. The proposed algorithm is used to calibrate various neutronics and thermal-hydraulics parameters. The virtual environment for reactor applications-core simulator (VERA-CS) is used to simulate a three-dimensional core depletion problem. The proposed algorithm is then used to construct a reduced order model (a surrogate) which is then used in a Bayesian approach to calibrate the neutronics and thermal-hydraulics parameters. The algorithm is tested and the benefits of data assimilation and calibration are highlighted in an uncertainty quantification study and requantification after the calibration process. Results showed that the proposed algorithm could help to reduce the uncertainty in key reactor attributes based on experimental and operational data.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.244-250
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• 1996

Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.05a
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• pp.128-129
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• 2007

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.89.1-89
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.67-67
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• 2002