• Advances in Energy Research
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• v.5 no.2
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• v.6 no.4
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• pp.239-247
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• 1974

A human phantom of polyethylene has been designed and sculptured for studying the effective radiation safety control. The phantom has the approximate size of the Korean adult and was sliced into thirty-five transverse slabs, 2.5 cm thick, The relative dose at the specified position was determined from the exposure that a TLD badge worn on the surface of the phantom body received from external ${\gamma}$-ray. The variation of the exposure as a function of depth in the phantom was measured for uncollimated ${\gamma}$-ray using TLD rods, and also isodose curves were obtained for the anatomical cross-section of the critical organs of the body. To simulate radiation exposure condition in the nuclear facility, measurements were made for given angles of incident ${\gamma}$-ray. The front to back attenuation factor for human phantom of thickness 20 cm was 0.439 for Cs$^{137}$ ${\gamma}$-ray which is in reasonable agreement with the published data.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.344-351
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• 2020

AgInCd alloy is widely used as neutron absorber in nuclear reactors. However, the AgInCd control rods may fail during service due to the irradiation swelling. In the present study, a calculational method is proposed to calculate the composition change of the AgInCd absorber. Calculated results show that neutron fluence has significant impact on the chemical compositions. Ag and In contents gradually decrease while Cd and Sn conversely increases from the center to the rim of AgInCd absorber due to the depression of neutron flux. The composition change at the surface is higher almost two times than that at the center. Based on the calculated compositions, six simulated AgInCdSn alloys were prepared and examined. With the increase of Cd and Sn, the simulated AgInCdSn alloys transform from a single fcc phase into the mixed fcc and hcp phases, and finally into the single hcp phase. The atomic volume of the hcp phase is obviously larger than the fcc phase. The fcc-hcp transformation results in considerable volume swelling of the AgInCd absorber. Moreover, the lattice parameters of the fcc and hcp phases gradually increase with Cd and Sn contents, which also can induce small volume swelling.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.8-19
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• 1993

The natural convection model of the consolidated system has been developed to make sure the removal of decay heat generated in the spent fuel for the loss of forced cooling accident. The numerical technique employed was based on the ADI scheme. The calculation of heat generation rate in the spent fuel was peformed by the ANS-79 decay heat model, and the nonuniform surface heat flux is assumed with a chopped sine curve for the conservative decay heat generation input. The sensitivity study was performed to examine the possibility of the pool bulk boiling by varying the various parameters, i.e. inter-fuel spacing ratio, heat generation power, and radius of the fuel rod. The application results of this model show that the natural circulation flow through compacted spent fuel bundles enables the pool temperature to control in a safe and effective manner, after the required cooling time. The corresponding acceptance criteria of the cooling time for rearranging the spent fuel rods were also found.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.941-950
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• 2013

The Korean Nuclear-Hydrogen Technology Development (NHTD) Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about $1030^{\circ}C$ in BOC (Beginning of Cycle) during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about $1.7{\times}10^{21}\;n/cm^2$, PIE (Post-Irradiation Examination) of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility). This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.1-5
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• 2017

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor ($K_{eff}$) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of $D_2O$, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

• Journal of the Korean Society of Industry Convergence
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• v.21 no.6
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• pp.337-344
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• 2018

Food irradiation is important not only in ensuring safety but also improving antioxidant activity of peaches. Our objective was to establish the best irradiation treatment for peaches by calculating dose distribution using Monte Carlo simulation. 3-D geometry and component densities of peaches, extracted from CT scan, were entered into MCNP to obtain simulated dose distribution. Radiation energies for electron beam were 1.35 MeV (low energy) and 10 MeV (high energy). Co (1.25 MeV) and the Husman irradiator, containing three sealed Cs source rods in an annular array, were used for gamma irradiation. At 1.35 MeV electron beam simulation, electrons penetrated well beyond the peach skin, enough for surface treatment for microorganisms and allergens. At 10 MeV electron beam simulation, for top-beam only treatment, doses at the core were the highest and for double beam treatment, the electron energy was absorbed by the entire sample. At Co source, the radiation doses were presented on the whole area. At Cs source, the dose uniformity ratios were 2.78 for one source and 1.48 for three ones at 120 degrees interval. Proper control of irradiation treatment is critical to establish confidence in the irradiation process.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• International conference on construction engineering and project management
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• 2015.10a
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• pp.401-404
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• 2015

Demand on free-form buildings is gradually increasing, yet owing to the difficulty of production-installation work, several problems occur in the construction phase upon construction of a building, including the increased cost and construction duration, and reduced constructibility. To solve these problems, a techonology to produce FCP using a CNC(Computerized Numeric Control) machine is developed. The technology is that the information of designed free-form buildings to the CNC machine is transferred, and the transferred information is used for RTM(Rod-Type Mold, the mold shaped by back-up rods) and PCM(Phase Change Material) shaping, and the shaped RTM and PCM have the role of molds to produce FCP. Construction duration and project cost are limited in building sites, so the efficiency of processes like production-installation of FCP for application of the technology is significant. Since it is almost impossible to change the production-installation process at the construction phase when they are established, process should be deliberately decided. Therefore, the purpose of the study is to propose a production-installation simulation model of free-form concrete panels, in aspect of PCM. This paper is establishing the process for production-installation of FCP, estimating time required by each construction type and proposing a time simulation model that changes according to various constraints based on the analyses. With the time simulation model, it will be possible to build a cost model and to review the optimal construction duration and project cost.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.