• Journal of Radiation Protection and Research
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• v.47 no.2
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• pp.99-106
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• 2022

Background: The radionuclide inventory calculation codes such as ORIGEN and FISPACT collapse neutron reaction libraries with energy spectra and generate an effective one-group cross-section. Since the nuclear cross-section data, energy group (g) structure, and other input details used by the two codes are different, there may be differences in each code's activation inventory calculation results. In this study, the calculation results of neutron-induced activation inventory using ORIGEN and FISPACT were compared and analyzed regarding radioactive waste classification and worker exposure during nuclear decommissioning. Materials and Methods: Two neutron spectra were used to obtain the comparison results: Watt fission spectrum and thermalized energy spectrum. The effective one-group cross-sections were generated for each type of energy group structure provided in ORIGEN and FISPACT. Then, the effective one-group cross-sections were analyzed by focusing on ⁵⁹Ni, ⁶³Ni, ⁹⁴Nb, ⁶⁰Co, ¹⁵²Eu, and ¹⁵⁴Eu, which are the main radionuclides of stainless steel, carbon steel, zircalloy, and concrete for decommissioning nuclear power plant (NPP). Results and Discussion: As a result of the analysis, ¹⁵⁴Eu and ⁵⁹Ni may be overestimated or underestimated depending on the code selection by up to 30%, because the cross-section library used for each code is different. When ORIGEN-44g, -49g, and -238g structures are selected, the differences of the calculation results of effective one-group cross-section according to group structure selection were less than 1% for the six nuclides applied in this study, and when FISPACT-69g, -172g, and -315g were applied, the difference was less than 1%, too. Conclusion: ORIGEN and FISPACT codes can be applied to activation calculations with their own built-in energy group structures for decommissioning NPP. Since the differences in calculation results may occur depending on the selection of codes and energy group structures, it is appropriate to properly select the energy group structure according to the accuracy required in the calculation and the characteristics of the problem.

• Journal of the Korean Society of Radiology
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• v.13 no.1
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• pp.45-53
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• 2019

This study is a neutron activation for concrete that shields medical linear accelerator facilities. Comparison of general concrete and low activation concrete. The simulation method was simulated using MCNPX (Ver. 2.5.0) and FISPACT-2010, and the shielding ability for photon and neutron beams was calculated and neutron activation evaluation was carried out. As a result, the shielding capacity was 20 ~ 50 cm efficient in general concrete, and activate evaluation in low activation concrete was calculated to be low in radioactivity concrete, but all were estimated to not exceed their own allowable concentration in self - disposal. As a result of the comprehensive analysis, it is considered effective to use ordinary concrete.

• 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.49 no.6
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• pp.1346-1353
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• 2017

The FISPACT-II inventory simulation platform is a modern computational tool with advanced and unique capabilities. It is sufficiently flexible and efficient to make it an ideal basis around which to perform extensive simulation studies to scope a variety of responses of many materials (elements) to several different neutron irradiation scenarios. This paper briefly presents the typical outputs from these scoping studies, which have been used to compile a suite of nuclear physics materials handbooks, providing a useful and vital resource for material selection and design studies. Several different global responses are extracted from these reports, allowing for comparisons between materials and between different irradiation conditions. A new graphical output format has been developed for the FISPACT-II platform to display these "global summaries"; results for different elements are shown in a periodic table layout, allowing side-by-side comparisons. Several examples of such plots are presented and discussed.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2683-2689
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• 2024

This work investigated the most suitable type of degrader (Cu, Al or Nb) and its thickness, taking into consideration the salient aspects of concrete activation for high-purity ⁸⁹Zr production by ⁸⁹Y(p,n)⁸⁹Zr nuclear reaction. The MCNP and FISPACT codes were used to determine the optimal degrader thickness and the radioactivity of shielding concrete by neutron activation, respectively. The results showed that the optimal thickness of the beam degraders was 1.16, 3.19, and 1.33 mm for Cu, Al, and Nb, respectively. The neutron production rate per proton and the energy and angular distributions of neutrons varied depending on the type of degrader. Considering the radioactivity of the target-room concrete and the amount of radioactive waste expected to be generated, the use of a 1.33-mm-thick Nb degrader for ⁸⁹Zr production was determined to be the best choice.

• Journal of the Korean Society of Radiology
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• v.15 no.4
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• pp.415-427
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• 2021

Korea Institute of Radiological and Medical Sciences has provided various beam irradiation services to researchers using a 50 MeV cyclotron beam line. In particular, since the neutron beam service uses the nuclear reaction between protons and beryllium, the possibility of activation of the irradiated sample increases by using a high current. In this study, MCNP 6.2 and FISPACT-II 4.0 were used to evaluate the possible activation during the 35 MeV 20 ㎂ neutron beam service, which is preferred by the researchers. As a result of the calculation, if the iron, copper, and tungsten samples were irradiated for more than 1 hour, long-lived radioisotopes were produced and their radioactivity exceeded the standard level for self-disposal. Under the conditions of 2 hours of daily irradiation, no activation occurred in the building materials, and the internal exposure of workers due to air activation inside the irradiation room was very insignificant. And when this air was discharged to environment, the radioactivity including this air was also satisfied the emission standard.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.378-379
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• 2011

• Journal of Radiation Industry
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• v.17 no.4
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• pp.443-449
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• 2023

KAERI(Korea Atomic Energy Research Institute) is conducting research and development of large-scale radiation generators and the latest radiation measuring instruments. In particular, research and development of security screening equipment using an electron beam accelerator and a neutron generator is in progress recently. Globally, 20 ft containers are used to transport imports and exports, and electron beam accelerators are radiation sources to measure the shape of the material inside the container during customs inspections in each country. KAERI is developing a device that can use an electron beam accelerator and a neutron generator sequentially to grasp the shape of various materials as well as the location of the internal target material. In this study, when using the neutron generator, the radiation dose and the degree of activation by neutron for the facility and surrounding environment, facility equipment were simulated using MCNP and FISPACT code. As a result, the shielding structures inside and outside the radiation control area were satisfactory to the reference level established conservatively based on the Korean Nuclear Act.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.63-69
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• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.