• Proceedings of the Korean Society of Medical Physics Conference
• /
• 1999.11a
• /
• pp.54-57
• /
• 1999

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.34 no.5
• /
• pp.553-561
• /
• 2010

A medium-scale helium loop that can simulate a VHTR (very-high-temperature reactor) is now under construction at the Korea Atomic Energy Research Institute. The heaters of the test helium loop electrically heat helium fluid up to $950^{\circ}C$ at pressures of 1 to 9 MPa. To optimize the design specifications of the experimental helium loop, the conjugate heat transfer in the high-temperature helium heater was analyzed by performing a CFD simulation. The analysis results indicate that the maximum temperature does not exceed the allowable limit. It is confirmed that the thermal characteristics of the loop with the given geometry satisfy the design requirements.

• Nuclear Engineering and Technology
• /
• v.56 no.4
• /
• pp.1165-1203
• /
• 2024

In the paper we validate theoretical models of the pulse against experimental data from the Jozef Stefan Institute TRIGA Mark II research reactor. Data from all pulse experiments since 1991 have been collected, analysed and are publicly available. This paper summarizes the validation study, which is focused on the comparison between experimental values, theoretical predictions (Fuchs-Hansen and Nordheim-Fuchs models) and calculation using computational program Improved Pulse Model. The results show that the theoretical models predicts higher maximum power but lower total released energy, full width at half maximum and the time when the maximum power is reached is shorter, compared to Improved Pulse Model. We evaluate the uncertainties in pulse physical parameters (maximum power, total released energy and full width at half maximum) due to uncertainties in reactor physical parameters (inserted reactivity, delayed neutron fraction, prompt neutron lifetime and effective temperature reactivity coefficient of fuel). It is found that taking into account overestimated correlation of reactor physical parameters does not significantly affect the estimated uncertainties of pulse physical parameters. The relative uncertainties of pulse physical parameters decrease with increasing inserted reactivity. If all reactor physical parameters feature an uncorrelated uncertainty of 10 % the estimated total uncertainty in peak pulse power at 3 $ inserted reactivity is 59 %, where significant contributions come from uncertainties in prompt neutron lifetime and effective temperature reactivity coefficient of fuel. In addition we analyse contribution of two physical mechanisms (Doppler broadening of resonances and neutron spectrum shift) that contribute to the temperature reactivity coefficient of fuel. The Doppler effect contributes around 30 %-15 % while the rest is due to the thermal spectrum hardening for a temperature range between 300 K and 800 K.

• Nuclear Engineering and Technology
• /
• v.32 no.6
• /
• pp.566-594
• /
• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1658-1668
• /
• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• Proceedings of the Korean Society of Medical Physics Conference
• /
• 1999.11a
• /
• pp.58-61
• /
• 1999

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.22 no.1
• /
• pp.40-48
• /
• 2010

The purpose of the research is to develop the high performance solar chemical reactor for producing hydrogen using steam reforming reaction of methane. A specific shape chemical reactor is suggested : spiral type reactor. The reactor is installed on the dish-type solar thermal system of Inha University. The temperatures, $CH_4$ conversion rates, and Hz proportion are measured. At specific condition, $CH_4$ conversion rates of the spiral type reactor are about 91%, and Hz proportion are about 66%. The spiral type reactor gives reasonably good performance without any problems caused by highly concentrated solar radiation.

• 한국태양에너지학회:학술대회논문집
• /
• 2009.04a
• /
• pp.55-60
• /
• 2009

The purpose of the research is to develop the high performance solar chemical reactor for producing hydrogen using steam reforming reaction of methane. A specific shape chemical reactor is suggested: spiral type reactor. The reactor is installed on the dish-type solar thermal system of Inha University. The temperatures, $CH_4$ conversion rates are measured. At specific condition, $CH_4$ conversion rates of the spiral type reactor are about 92%. The spiral type reactor gives reasonably good performance without any problems caused by highly concentrated solar radiation.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.05a
• /
• pp.1001-1007
• /
• 1995

A Comprehensive Vibration Assessment Program (CVAP) has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibrations prior to commercial operation. The theoretical evidence for the structural integrity of the reactor internals and the basis for measurement and inspection are provided by the analysis. Flow induced hydraulic loads and reactor internals vibration response data were measured during pre-core hot functional testing in YGN 4 site. Also, the critical areas in the reactor internals were inspected visually to check any existence of structural abnormality before and after the pre-core hot functional testing. Then, the measured data have been analyzed and compared with the predicted data by analysis. The measured stresses are less than the predicted values and the allowable limits. It is concluded that the vibration response of the reactor internals due to the flow induced vibration under normal operation is acceptable for long term operation.

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2975-2981
• /
• 2020

Radiation damage is one of the aging important causes in nuclear reactors. Radiation damage causes changes in material properties. In this study, this effect has been evaluated and analyzed on the clad of the Tehran research reactor (TRR). A grade 6061 aluminum is used as a clad in the TRR. The MCNPX code is used to designate the most sensitive location of the reactor and calculate neutron flux distribution. Then, a software using FORTRAN language programming is developed to process the particle track (PTRAC) output file of the MCNPX code. The SRIM code is used here to calculate the rate of displacement per atom. Moreover, the SPECOMP and SPECTER codes are also applied to estimate the displacement rate and compared with the results attained using the SRIM code. The rate of displacement per atom by the SPECTER and SRIM codes have been obtained 2.54 × 10^-7 dpa/s and 2.44 × 10^-7 dpa/s (QD method), respectively. Also, the mechanical properties have been evaluated using the RCC-MRx code and have been compared with experimental results. Finally, the change in the matter specification has been analyzed as a function of time.