• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.65-70
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• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.

• 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.

• Journal of Sensor Science and Technology
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• v.25 no.4
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• pp.264-270
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• 2016

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1036-1044
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• 2023

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3591-3598
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• 2023

Photoneutron background spectroscopy observations at linac are essential for directing accelerator shielding and subtracting background signals. Therefore, we constructed a Bonner Sphere Spectrometer (BSS) system based on an array of BF₃ gas proportional counter tubes. Initially, the response of the BSS system was simulated using the MCNP5 code. Next, the response of the system was calibrated by using neutrons with energies of 2.86 MeV and 14.84 MeV. Then, the system was employed to measure the spectrum of the ²⁴¹Am-Be neutron source, and the results were unfolded by using the Gravel and EM algorithms. Using the validated system, the undesirable neutron spectrum of the 120 MeV electron linac was finally measured and acquired. In addition, it is demonstrated that the equivalent undesirable neutron dose at a distance of 3.2 m from the linac is 19.7 mSv/h. The results measured by the above methods could provide guidance for linac-related research.

• Journal of Radiation Industry
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• v.4 no.1
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• pp.33-38
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• 2010

The MCNP5 neutron, electron, photon Monte Carlo transport program was installed on the KISTI's SUN Tachyon computer using the parallel programming. Electron beam accelerators were modeled and shielding calculations were performed in order to investigate the reduction of computation time in the supercomputer environment. It was observed that a speedup of 40 to 80 of computation time can be obtained using 64 CPUs compared to an IBM PC.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1183-1184
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• 2004

• 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.

• Journal of the Korean Society of Radiology
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• v.2 no.1
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• pp.31-34
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• 2008

The neutron capture spectrum for the light nuclide was very useful to study the nuclear structure. In the present study, the capture gamma-ray from the 27-keV resonance of $^{19}F(n,g)^{20}F$ reaction were measured with an anti-Compton NaI(Tl) spectrometer and the 3-MV Pelletron accelerator of the Research Laboratory for Nuclear Reactors at the Tokyo institute of technology. A neutron Time-of-Flight method was adopted with a 1.5 ns pulsed neutron source by the $^7Li(p,n)^7Be$ reaction. In the present experiment, a Teflon(($CF_2$)n) sample was used The sample was disk with a diameter of 90mm. The thickness of sample was determined so that reasonable counting rates could be obtained and the correction was not so large for the self-shielding and multiple scattering of neutrons in the sample, and was 5mm. The primary gamma-ray transitions were compared with previous measurement of Kenny.