• Journal of Energy Engineering
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• v.16 no.4
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• pp.155-163
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• 2007

To insure nuclear reactor safety, the reactivity of control rods should be calculated by measuring the criticality of reactor core and it is regularly performed during the annual physics test period. Also, the core criticality should be monitored during the start-up operation to avoid reactivity induced accidents. Many research works on control rod reactivity measurement and subcriticality measurement have been accomplished throughout the world for decades and recently a new method named "Modified Neutron Source Multiplication Method (MNSM)" was proposed in Japan which is known to be improved overcoming limitations of traditional Neutron Source Multiplication Method (NSM). In this study, MNSM was tested in calculation of subcriticalities and in evaluation of application validity using the educational reactor in Kyung Hee University, AGN-201. For this study, a revised nuclear data library and a neutron transport code system TRANSX - PARTISN were established. Correction factors for various control rod positions were produced using the k-effective values and the corresponding flux distributions and adjoint flux distributions. Experimental values of the core criticality were obtained using the neutron count rates of the BF3 proportional counters. The results showed that the expected reactivity worth of control rods by MNSM agreed well with the theoretical values and the correction factors contributed much for this purpose.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.63-72
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• 2014

Pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems in Korea. The Korea Atomic Energy Research Institute has been studying pyroprocess technology, and the conceptual design of an engineering-scale pyroprocess facility, called the Advanced Fuel Cycle (AFC) facility, has been performed on the basis of a 10tHM throughput per year. In this paper, the concept of the AFC facility was introduced, and its safety evaluations were performed. For the safety evaluations, anticipated accident events were selected, and environmental safety analyses were conducted for the safety of the public and workers. In addition, basic radiation shielding safety analyses and criticality safety analyses were conducted. These preliminary safety studies will be used to specify the concept of safety systems for pyroprocess facilities, and to establish safety design policies and advance more definite safety designs.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.413-418
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• 1996

The main purpose of this paper is to develop the modified LTC code for accurate analysis of the boron concentration behavior of all components in the Nuclear Steam Supply System (NSSS). This is achieved by adapting a multi-cell mad to the existing Long Term Cooling (LTC) code. To verify the modified LTC, the simulated results were compared with the actual test results measured during YGN 4 initial criticality test. It was shown that the simulated results of this modified LTC were in good agreement with the actual test results. Also, the boron concentration behavior analysis were performed using the modified LTC code for both direct and indirect dilution/boration nude using YGN 3,4 design data. This modified LTC code can provide a valuable information in predicting boron concentration behavior during power maneuvering such as startup operation, shutdown operation and load follow operation. It is expected that the modified LTC can be applied to both on-line and off-line mode using Plant Computer System(PCS).

• Journal of the Korean Society for Industrial and Applied Mathematics
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• v.13 no.2
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• pp.73-86
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• 2009

The problem of thermal stability of an exothermic reactive viscous fluid between two parallel walls in the plane Poiseuille and Couette flow configurations is investigated for different thermal boundary conditions. Neglecting reactant consumption, the closed-form solutions obtained from the momentum equation was inserted into the energy equation due to dissipative effect of viscosity. The resulting energy equation was analyzed for criticality using the variational method technique. The problem is characterized by two parameters: the Nusselt number(N) and the dynamic parameter($\Lambda$). We observed that the thermal and dynamical boundary conditions of the wall have led to a significant departure from known results. The influence of the variable pre-exponential factor, due to the numerical exponent m, also give further insight into the behavior of the system and the results expressed graphically and in tabular forms.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1226-1233
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• 2017

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors and representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. The database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.41-50
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• 2006

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by $\sim20$ tons.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.153-162
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• 2005

In this Paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by ${\~}$20 tons.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.245-258
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• 1989

A 69-group cross section library consisting of more than 130 materials was generated for thermal reactor applications using the NJOY nuclear data processing system and the recent version of evaluated nuclear data files available from IAEA Nuclear Data Section. The multigroup library was validated through the analysis of various criticality experiments and depletion results of PWR. When used with the WIMS-KAERI code, the average $K_{eff}$ obtained for 47 uranium-oxide and 41 uranium metal fueled critical configurations is 0.9997 with a standard deviation of 0.69 percent. The calculated burnup dependent isotopic inventories of uranium and plutonium generally show good agreement with measured values obtained from depleted PWR pins.s.

• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.56-60
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• 1982

A calculation of the effective multiplication factor has been made for GODIVA and JEZEBEL critical assemblies by using a computer code, ANISN, with having the Watt's, Cranberg's and Maxwellian formulae for the prompt fission neutron spectrum as a fission source. Then the calculated values have been compared with experimental data obtained by others. The Maxwellian formula seems to be the best one for representing the prompt fission neutron spectrum since the effective multiplication factor based on it shows a better agreement with the experimental value compared to the rest formulae.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4759-4775
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• 2022

The RVACS is a versatile and robust safety system driven by two natural circulations: in-vessel coolant and ex-vessel air. To observe interaction between the two natural circulations, SINCRO-IT facility was designed with two different similarity laws simultaneously. Bo' based similarity law was employed for the in-vessel, while Ishii's similarity law for the ex-vessel excluding the radiation. Compared to the prototype, the sodium and air system, SINCRO-IT was designed with Wood's metal and air, having 1:4 of the length reduction, and 1.68:1 of the time scale ratio. For the steady state, RV temperature limit was violated at 0.8% of the decay heat, while the sodium boiling was predicted at 1.3%. It showed good accordance with the system code, TRACE. For an arbitrary re-criticality scenario with RVACS solitary operation, sodium boiling was predicted at 25,100 s after power increase from 1.0 to 2.0%, while the system code showed 30,300. Maximum temperature discrepancy between the experiments and system code was 4.2%. The design and methodology were validated by the system code TRACE in terms of the convection, and simultaneously, the system code was validated against the simulating experiments SINCRO-IT. The validated RVACS model could be imported to further accident analysis.