• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.675-680
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• 2006

The neutron transmutation doping of silicon (NTD), as a method to produce a high quality semiconductor, utilizes the transmutation of a silicon element into phosphorus by neutron absorption in a silicon single crystal. In this paper, we present the design of a neutron screen for a 6' Si ingot irradiation in the NTD2 hole of HANARO. The goal of the design is to achieve an even flat axial distribution of the resistivity, or $Si^{30}(n,{\gamma})Si^{31}$ reaction rate, in the irradiated Si ingot. We used the MCNP4C code to simulate the neutron screen and to calculate the reaction rate distribution in the Si ingot. The fluctuations in the axial distribution were estimated to be within ${\pm}2.0%$ from the average for the final neutron screen design; thus, they satisfy the customers' requirement for uniform irradiation. On the other hand, we determined the optimal insertion depths of the Si ingots by varying the critical control rod position, which greatly affects the axial flux distribution.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4137-4141
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• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.

• Journal of Radiation Protection and Research
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• v.21 no.4
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• pp.263-271
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• 1996

The ANSI N13.32 recommends that a study of the angular response of a dosimeter be carried out once, although no pass/fail criterion is given for angular response. Gamma dose equivalent conversion and angular dependence factors were calculated by using MCNP code for the case of ANSI N13.32 extremity phantoms(finger and arm) at the depth of $7mg/cm^2$. Those extremity dosimeters were assumed to be irradiated from both monoenergitic photons and ISO X-ray narrow beams. These calculated gamma dose equivalent conversion and angular dependence factors were compared to B. Grosswendt's result calculated by using X-ray beams. The result showed that the dose equivalent conversion factors of this study agreed well with that of B. Grosswendt for all energies within 2% except 7% in the case of the low energies. In the case of angular dependence factors comparison, they agreed within 3%. It was shown that angular dependence factors of the finger phantom decreased as the horizontal angle of the phantom increased for the ISO X-ray beams less than 60keV. For the higher energy X-ray beams range they decreased slightly around 40 degree, but then increased from this energy to 90 degree.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.93-98
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• 2006

The radial uniformity of neutron irradiation in silicon ingots for neutron transmutation doping (NTD) at HANARO is examined by both calculations and measurements. HANARO has two NTD holes named NTD1 and NTD2. We have been using the NTD2 hole for 5 in. NTD commercial service, and we intend to use two holes for 6 in. NTD. The objective of this study is to predict the radial uniformity of 6 in. NTD at the two holes. The radial neutron flux distributions inside single crystal and noncrystal silicon loaded at the NTD2 hole are calculated by the VENTURE code. For NTD1, the radial distributions of the reaction rate for a 6 in. NTD with a neutron screen are calculated by MCNP, and measured by gold wire activation. The results of the measurements are compared with those of the calculations. From the VENTURE calculation, it is confirmed that the neutron flux distribution in the single crystal silicon is much flatter than that in the non-crystal silicon. The non-uniformities of the measurements for radial neutron irradiation are slightly larger than those of the calculations. However, excluding local dips in the measurements, the overall trends of the distributions are similar. The radial resistivity gradient (RRG) for a 5 in. silicon ingot is estimated to be about $1.5\%$. For a 6 in. ingot, the RRG of a silicon ingot irradiated at HANARO is predicted to be about $2.1\%$. Also, from the experimental results, we expect that the RRG would not be larger than $4.4\%$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Progress in Medical Physics
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• v.33 no.4
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• pp.129-135
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• 2022

Purpose: A full-energy-peak (FEP) efficiency correction is required through a Monte Carlo simulation for accurate radioactivity measurement, considering the geometrical characteristics of the detector and the sample. However, a relative deviation (RD) occurs between the measurement and calculation efficiencies when modeling using the data provided by the manufacturers due to the randomly generated dead layer. This study aims to optimize the structure of the detector by determining the dead layer thickness based on Monte Carlo simulation. Methods: The high-purity germanium (HPGe) detector used in this study was a coaxial p-type GC2518 model, and a certified reference material (CRM) was used to measure the FEP efficiency. Using the MC N-Particle Transport Code (MCNP) code, the FEP efficiency was calculated by increasing the thickness of the outer and inner dead layer in proportion to the thickness of the electrode. Results: As the thickness of the outer and inner dead layer increased by 0.1 mm and 0.1 ㎛, the efficiency difference decreased by 2.43% on average up to 1.0 mm and 1.0 ㎛ and increased by 1.86% thereafter. Therefore, the structure of the detector was optimized by determining 1.0 mm and 1.0 ㎛ as thickness of the dead layer. Conclusions: The effect of the dead layer on the FEP efficiency was evaluated, and an excellent agreement between the measured and calculated efficiencies was confirmed with RDs of less than 4%. It suggests that the optimized HPGe detector can be used to measure the accurate radioactivity using in dismantling and disposing medical linear accelerators.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• The Journal of Korean Society for Radiation Therapy
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• v.25 no.2
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• pp.145-151
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• 2013

Purpose: In this study, we analyzed how the dose change by field size effects on atomic number of shielding materials while using 6 MeV election beam. Materials and Methods: The parallel plate chamber is mounted in $25{\times}25cm^2$ the phantom such that the entrance window of the detector is flush with the phantom surface. phantom was covered laterally with aluminum, copper and lead which thickness have 5% of allowable transmission and then the doses were measured in field size $6{\times}6$, $10{\times}10$ and $20{\times}20cm^2$ respectively. 100 cGy was irradiated using 6 MeV electron beam and SSD (Source Surface Distance) was 100 cm with $10{\times}10cm^2$ field size. To calculate the photon flux, electron flux and Energy deposition produced after pass materals respectively, MCNPX code was used. Results: The results according to the various shielding materials which have 5% of allowable transmission are as in the following. Thickness change rate with field size of $6{\times}6cm^2$ and $20{\times}20cm^2$ that compared to the field size of $10{\times}10cm^2$ found to be +0.06% and -0.06% with aluminum, +0.13% and -0.1% with copper, -1.53% and +1.92% with lead respectively. Compare to the field size $10{\times}10cm^2$, energy deposition for $6{\times}6cm^2$ and $20{\times}20cm^2$ had -4.3% and +4.85% respectively without shielding material. With aluminum it had -0.87% and +6.93% respectively and with lead it had -4.16% and +5.57% respectively. When it comes to photon flux with $6{\times}6cm^2$ and $20{\times}20cm^2$ of field sizes the chance -8.95% and +15.92% without shielding material respectively, with aluminum the number -15.56% and +16.06% respectively and with copper the chance -12.27% and +15.53% respectively, with lead the number +12.36% and -19.81% respectively. In case of electron flux in the same condition, the number -3.92% and +4.55% respectively without shielding material respectively, with aluminum the number +0.59% and +6.87% respectively, with copper the number -1.59% and +3.86% respectively, with lead the chance -5.15% and +4.00% respectively. Conclusion: In this study, we found that the required thickness of the shielding materials got thinner with low atomic number substance as the irradiation field is increasing. On the other hand, with high atomic number substance the required thickness had increased. In addition, bremsstrahlung radiation have an influence on low atomic number materials and high atomic number materials are effected by scattered electrons.

• Journal of the Korean Society of Radiology
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• v.17 no.4
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• pp.489-495
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• 2023

Radiation shielding facilities are constructed in locations where diagnostic radiation generators are installed, with the aim of preventing exposure for patients and radiation workers. The purpose of this study is seek to compare and validate the trend of attenuation thickness of lead, the primary material in these radiation shielding facilities, at different maximum tube voltages by Monte Carlo simulations and measurement. We employed the Monte Carlo N-Particle 6 simulation code. Within this simulation, we set a lead shielding arrangement, where the distance between the source and the lead sheet was set at 100 cm and the field of view was set at 10 × 10 cm². Additionally, we varied the tube voltages to encompass 80, 100, 120, and 140 kVp. We calculated energy spectra for each respective tube voltage and applied them in the simulations. Lead thicknesses corresponding to attenuation rates of 50, 70, 90, and 95% were determined for tube voltages of 80, 100, 120, and 140 kVp. For 80 kVp, the calculated thicknesses for these attenuation rates were 0.03, 0.08, 0.21, and 0.33 mm, respectively. For 100 kVp, the values were 0.05, 0.12, 0.30, and 0.50 mm. Similarly, for 120 kVp, they were 0.06, 0.14, 0.38, and 0.56 mm. Lastly, at 140 kVp, the corresponding thicknesses were 0.08, 0.16, 0.42, and 0.61 mm. Measurements were conducted to validate the calculated lead thicknesses. The radiation generator employed was the GE Healthcare Discovery XR 656, and the dosimeter used was the IBA MagicMax. The experimental results showed that at 80 kVp, the attenuation rates for different thicknesses were 43.56, 70.33, 89.85, and 93.05%, respectively. Similarly, at 100 kVp, the rates were 52.49, 72.26, 86.31, and 92.17%. For 120 kVp, the attenuation rates were 48.26, 71.18, 87.30, and 91.56%. Lastly, at 140 kVp, they were measured 50.45, 68.75, 89.95, and 91.65%. Upon comparing the simulation and experimental results, it was confirmed that the differences between the two values were within an average of approximately 3%. These research findings serve to validate the reliability of Monte Carlo simulations and could be employed as fundamental data for future radiation shielding facility construction.