• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.155-159
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• 2016

Background: Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. Materials and Methods: To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% $^{235}U$ contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. Results and Discussion: We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. $^{241}Pu$) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate Conclusion: Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

• Journal of Environmental Health Sciences
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• v.39 no.2
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• pp.187-195
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• 2013

Objectives: For the practical application of the dielectric barrier discharge plasma reactor, a plasma reactor able to manage large volumes of water is needed. This study investigated the possibility of the practical application of a multi-plasma reactor which is a scaled-up version of a single plasma reactor. Methods: The multi-plasma reactor consists of several high-voltage transformers and plasma modules (discharge, ground electrodes and quartz dielectric tubes). The effects of water characteristics such as voltage (30-120 V), air flow rate (1-5 l/min), number of high-voltage transformers and plasma modules, and water quality on Escherichia coli (E. coli) disinfection and decrease of COD and $UV_{254}$ absorbance were investigated. Results: The experimental results showed that at a voltage of over 80 V, most of the E. coli were disinfected within 90 seconds. E. coli inactivation was not affected by the air flow rate. E. coli disinfection in the multiplasma process showed the traditional log-linear form of the disinfection curve. E. coli inactivation performance by transformer 3-Reactor 5 and transformer 3-Reactor 3 were similar. The disinfection performance of the UV process was affected by artificial sewage water. However, the plasma process was less affected by the artificial sewage within the standards for effluent water quality. Conclusions: Disinfection performance with several low voltages and plasma modules of three to five in number applied to the plasma process was higher than that concentrating a small amount of high voltage through a single plasma reactor. Removal of COD, $UV_{254}$ absorbance, and E. coli disinfection with the plasma process were better than with the UV process.

• Journal of Korean Society of Industrial and Systems Engineering
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• v.39 no.2
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• pp.113-118
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• 2016

Project management is a tool for smooth operation during a full cycle from the design to normal operation including the schedule, document, and budget management, and document management is an important work for big projects such as the JRTR (Jordan Research and Training Reactor). To manage the various large documents for a research reactor, a project management system was resolved, a project procedure manual was prepared, and a document control system was established. The ANSIM (Advanced Nuclear Safety Information Management) system consists of a document management folder, document container folder, project management folder, organization management folder, and EPC (Engineering, Procurement and Construction) document folder. First, the system composition is a computerized version of the Inter-office Correspondence (IOC), the Document Distribution for Agreement (DDA), Design Documents, and Project Manager Memorandum (PM Memo) works prepared for the research reactor design. Second, it reviews, distributes, and approves design documents in the system and approves those documents to register and supply them to the research reactor user. Third, it integrates the information of the document system-using organization and its members, as well as users' rights regarding the ANSIM document system. Throughout these functions, the ANSIM system has been contributing to the vitalization of united research. Not only did the ANSIM system realize a design document input, data load, and search system and manage KAERI's long-period experience and knowledge information properties using a management strategy, but in doing so, it also contributed to research activation and will actively help in the construction of other nuclear facilities and exports abroad.

• Applied Chemistry for Engineering
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• v.10 no.8
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• pp.1200-1209
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• 1999

Prediction model was composed by optimal parameter estimation from best fitting on reactant temperature profile, inlet and outlet temperature of coolant and yield of dual fixed-bed catalytic reactor(FBCR) which was measured in the industrial field. In order to design the FBCR which could obtain maximum conversion and yield, we investigated the effect of catalyst bed length and reactor radius changes. An uniform activity FBCR showed the best performance at z = 2.8 m of total catalysst bed length in case of reactor radius r = 0.01241 m and z =2.80 m(upper layer: 1.88 m, lower layer: 0.92 m) under reactor radius r = 0.01254 m for a dual activities FCBR. In case of reactor radius changes, the axial temperature profile and maximum radial temperature was rapidly risen for radius increase. The reactor radius decrease showed the opposite result.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.573-580
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• 2013

The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The $CANDU^{(R)}$ reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.354-358
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• 2016

Background: To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition Materials and Methods: MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. Results and Discussion: It was found that the leak rate is $0.1%{\cdot}day^{-1}$ of air, $0.004%{\cdot}day^{-1}$ of noble gas and $3.7{\times}10^{-5}%{\cdot}day^{-1}$ of aerosol during typhoon passing. The air leak rate of $0.1%{\cdot}day^{-1}$ can be converted into $1.36m^3{\cdot}hr^{-1}$, but the design leak rate in HANARO safety analysis report was considered as $600m^3{\cdot}hr^{-1}$ under the condition of $20m{\cdot}sec^{-1}$ wind speed outside of the building by typhoon. Conclusion: Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

• Journal of Environmental Health Sciences
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• v.34 no.3
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• pp.233-239
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• 2008

In order to improve reactor performance of existing sewage treatment plants, the feasibility of enhancing reactor performance by bioaugmentation using EM as bioaugmentation agent and the effects of anoxic: oxic time ratio on reactor performance were investigated. Continuous and intermittent aeration modes were compared under the 6 hr of HRT. Three different types of intermittent aeration modes, that is, 15 min, of anoxic:45 min of oxic, 30 min of anoxic: 30 min of oxic, and 45 min of anoxic: 15 min oxic respectively were chosen as test modes to study the effects of anoxic : oxic time ratios on reactor performance. The optimum anoxic: oxic time ratio was 30 min:30 min when considering simultaneous removal of organic, nitrogen and phosphorus. When applying EM into a continuously aerated reactor under the varying dosing rates of 50-200 ppm, reactor performance in terms of organic and nitrogen removal efficiencies was not improved at all. Nitrogen removal efficiency was increase when the EM dosing rate was increased. However the degree of improvement was slight when the EM was injected above 100 ppm. However optimum phosphorus removal was found at the EM dosing of 200 ppm. Thus it was found that optimum injection concentration of EM is 200 ppm. It is apparent that putting EM into a sewage treatment plant significantly affects the T-N removal efficiency of the reactor by enhancing denitrification efficiency especially in operational conditions of relatively long anoxic periods. To achieve reciprocal condition in a reactor with intermittent aeration it is necessary to enhance the reactor performance by EM injection. In the case of modifying existing continuously aerated reactors into intermittent aerated reactors, it is obvious that operating costs of aeration would be reduced by reducing aeration time when compared with existing conventional sewage treatment plants.

• 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 Korean Society of Industrial and Systems Engineering
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• v.34 no.4
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• pp.49-56
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• 2011

The Project of Jordan Research and Training Reactor (JRTR) officially launched in Aug. 2010. JRTR is the first made-in-Korea nuclear system to be built abroad by year 2015, and Korea Atomic Energy Research Institute (KAERI) is responsible for the design of major systems including the reactor core. While the PDCS (Project Document Control System) being operated by EPC company controls all the documents of the whole Project, KAERI is supposed to have its own system for KAERI documents. Meeting such a need; KAERI has implemented a document control for the JRTR Project into already existing ANSIM (KAERI Advanced Nuclear Safety Information Management) system. The documents of JRTR project to be controlled are defined in the PPM (Project Procedures Manual), QAP (Quality Assurance Procedure) and PEP (Project Execution Program). The ANSIM consists of the document management holder, document container holder and organization management holder. The document management holder, which is the most important part of ANSIM-JRTR, consists of the DDA (Document Distribution for Agreement), IOC (Inter-office Correspondence), PM Memo. (Project Manager Memorandum) and cover sheets of design documents. Other materials such as meeting minutes, sub-department materials and design information materials are stored in an independent COP (Community of Practice). This established computerized document control system, ANSIM, could lessen a burden for project management team and enhance the productivity as well.