• Nuclear Engineering and Technology
• /
• v.30 no.6
• /
• pp.541-554
• /
• 1998

A computer code COSMOS has been developed based on the CARO-D5 for the thermal analysis of LWR UO$_2$ and MOX fuel rod under steady-state and transient operating conditions. The main purpose of the COSMOS, which considers high turnup characteristics such as thermal conductivity degradation with turnup and rim formation at the outer part of fuel pellet, is to calculate temperature profile across fuel pellet and fission gas release up to high burnup. A new mechanistic fission gas release model developed based on physical processes has been incorporated into the code. In addition, the features of MOX fuel such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been also taken into account so that it can be applied to MOX fuel. Another important feature of the COSMOS is that it can analyze fuel segment refabricated from base irradiated fuel rods in commercial reactors. This feature makes it possible to analyze database obtained from international projects such as the MALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of the COSMOS has been tested with some number of experimental results obtained from the HALDEN, RISO and FIGARO programs. Comparison with the measured data indicates that, although the COSMOS gives reasonable agreement, the current models need to be improved. This work is being performed using database available from the OECD/NEA.

• Proceedings of the KSR Conference
• /
• 2004.06a
• /
• pp.702-709
• /
• 2004

Modified Gurson model (Gurson-Tvergaard-Needleman model) was used to analyze crack growth in M(T) and C(T) specimens. A commercial finite element code ABAQUS/Explicit is used to account for total failure of material point by cavity coalescence, and crack growth was simulated by finite element extinction. Crack growth resistance curve was obtained by calculating J-integral. Crack growth under residual stress was investigated.

• Steel and Composite Structures
• /
• v.43 no.2
• /
• pp.271-278
• /
• 2022

Buckling is one of the most critical phoneme in the design of steel structures. Lateral torsional buckling (LTB) is particularly significant for slender beams generally subjected to loading in plane. The web distortion effects on LTB are not addressed explicitly in standards for flexural design of steel I-section members. Hence, the present study is focused to predict the influence of the web distortion on the elastic (L_r) limiting lengths given in American Institute of Steel Construction (AISC) code for the lateral torsional buckling (LTB) behavior of steel beams due to no provision in the code for consideration of web distortion. For this aim, the W44x335 beam is adopted in the buckling analysis carried out by the ABAQUS finite element (FE) program since it is one of the most critical sections in terms of lateral torsional buckling (LTB). The strain results at mid-height of the web at mid-span of the beam are taken into account as the monitoring parameters. The web strain results are found to be relatively greater than the yield strain value when L/L_r is equal to 1.0. In other words, the ratio of L/L_r is estimated from the numerical analysis to be about 1.5 when the beam reaches its first yielding at mid-span of the beam at mid-height of the section. Due to the effect of web distortion, the elastic limiting length (L_r) from the numerical analysis is obtained to be considered as greater than the calculated length from the code formulation. It is suggested that the formulations of the limiting length proposed in the code can be corrected considering the influence of the web distortion. This correction can be a modification factor or a shape factor that reduces sectional slenderness for the LTB formulation in the code.

• Nuclear Engineering and Technology
• /
• v.38 no.2
• /
• pp.147-154
• /
• 2006

Thermal fatigue of the coolant circuits of PWR plants is a major issue for nuclear safety. The problem is especially accute in mixing zones, like T-junctions, where large differences in water temperature between the two inlets and high levels of turbulence can lead to large temperature fluctuations at the wall. Until recently, studies on the matter had been tackled at EDF using steady methods: the fluid flow was solved with a CFD code using an averaged turbulence model, which led to the knowledge of the mean temperature and temperature variance at each point of the wall. But, being based on averaged quantities, this method could not reproduce the unsteady and 3D effects of the problem, like phase lag in temperature oscillations between two points, which can generate important stresses. Benefiting from advances in computer power and turbulence modelling, a new methodology is now applied, that allows to take these effects into account. The CFD tool Code_Saturne, developped at EDF, is used to solve the fluid flow using an unsteady L.E.S. approach. It is coupled with the thermal code Syrthes, which propagates the temperature fluctuations into the wall thickness. The instantaneous temperature field inside the wall can then be extracted and used for structure mechanics computations (mainly with EDF thermomechanics tool Code_Aster). The purpose of this paper is to present the application of this methodology to the simulation of a straight T-junction mock-up, similar to the Residual Heat Remover (RHR) junction found in N4 type PWR nuclear plants, and designed to study thermal striping and cracks propagation. The results are generally in good agreement with the measurements; yet, in certain areas of the flow, progress is still needed in L.E.S. modelling and in the treatment of instantaneous heat transfer at the wall.

• Nuclear Engineering and Technology
• /
• v.56 no.4
• /
• pp.1204-1212
• /
• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

• Journal of Energy Engineering
• /
• v.17 no.1
• /
• pp.23-30
• /
• 2008

The neutron kinetic analysis methods for the molten-salt reactors are quite different from those for conventional solid-fuel reactors, which do not take into account the flowing-fuel-induced neutronics effects. Therefore, for dynamics and safety analyses of the molten-salt reactor systems, the conventional kinetics codes would not be appropriate to accurately predict its transient behaviors. A point-kinetics with flowing- fuel model has been used to assess the fluid-fuel reactor system safety, but recognized as not to be sufficient to simulate spatial distributions of delayed-neutron precursors and neutron populations during transients for given detail reactor models. In order to meet this requirement, AMBIKIND, a 2-group, 2-dimensional neutron kinetics code suitable for the molten-salt reactor systems was developed. This paper explains the code's theoretical and numerical descriptions and, as a part of its verification, includes some simulation results of MSRE stability experiments. Even though the present reactor model does not include the recirculation effect of the fuel-salt through the reactor system, the AMBIKIN2D code should be able to predict the power and phase shift at various power levels and reactivity insertions with better accuracy.

• Nuclear Engineering and Technology
• /
• v.41 no.5
• /
• pp.729-738
• /
• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Journal of Radiation Protection and Research
• /
• v.48 no.2
• /
• pp.90-99
• /
• 2023

Background: Quantitative risk assessments should be accompanied by uncertainty analyses of the risk models employed in the calculations. In this study, we aim to develop a computational code named SUMRAY for use in cancer risk projections from radiation exposure taking into account uncertainties. We also aim to make SUMRAY publicly available as a resource for further improvement of risk projection. Materials and Methods: SUMRAY has two versions of code written in R and Python. The risk models used in SUMRAY for all-solid-cancer mortality and incidence were those published in the Life Span Study of a cohort of the atomic bomb survivors in Hiroshima and Nagasaki. The confidence intervals associated with the evaluated risks were derived by propagating the statistical uncertainties in the risk model parameter estimates by the Monte Carlo method. Results and Discussion: SUMRAY was used to calculate the lifetime or time-integrated attributable risks of cancer under an exposure scenario (baseline rates, dose[s], age[s] at exposure, age at the end of follow-up, sex) specified by the user. The results were compared with those calculated using another well-known web-based tool, Radiation Risk Assessment Tool (RadRAT; National Institutes of Health), and showed a reasonable agreement within the estimated confidential interval. Compared with RadRAT, SUMRAY can be used for a wide range of applications, as it allows the risk projection with arbitrarily specified risk models and/or population reference data. Conclusion: The reliabilities of SUMRAY with the present risk-model parameters and their variance-covariance matrices were verified by comparing them with those of the other codes. The SUMRAY code is distributed to the public as an open-source code under the Massachusetts Institute of Technology license.

• Progress in Medical Physics
• /
• v.14 no.1
• /
• pp.1-7
• /
• 2003

In this work we developed a EGS4 control code to calculate the dose distributions for high energy electron beams in water phantom applied longitudinal magnetic field. We reviewed the electron's motion in magnetic field and delivered equations for direction changs of the electron by the external magnetic field. The mathematical results are inserted into the EGS4 code system to account for the presence of external magnetic fields in phantom. The electron pencil beam paths of 6 MeV in water phantom are calculated for magnetic fields of 1-3 T and the dose distributions for a field of 1.0 cm in diameter are calculated for magnetic fields of 0.6-1 T using the code. From the results of path calculations we found that the lateral ranges of the electrons are reduced in the magnetic field of 3 T. For a field of 1 cm diameter and a magnetic field of 1 T, the small dose enhancement near the range of the electrons on the depth dose and the penumbra reduction of 0.15 cm on the beam profile are observed. We discussed and evaluated the results from the theoretical concepts.

• Applied Chemistry for Engineering
• /
• v.32 no.1
• /
• pp.102-109
• /
• 2021

Recently, an interest in risk calculation methods has been increasing in Korea due to the establishment of classification code for explosive hazardous area on gas facility (KGS CODE GC101), which is based on the international standard of classification of areas - explosive gas atmospheres (IEC 60079-10-1). However, experiments to check for leaks of combustible or toxic gases are very difficult. These experiments can lead to fire, explosion, and toxic poisoning. Therefore, even if someone tries to provide a laboratory for this experiment, it is difficult to install a gas leakage equipment. In this study we find out differences among actual experiments, CFD by using FLACS and calculation based on classification code for explosive hazardous area on gas facility (KGS CODE GC101) by comparing to each other. We develpoed KGS HAC (hazardous area classification) program which based on KGS GC101 for convenience and popularization. As a result, actual gas leak, CFD and KGS HAC are showing slightly different results. The results of dispersion of 1.8 to 2.7 m were shown in the actual experiment, and the CFD and KGS HAC showed a linear increase of about 0.4 to 1 m depending on the increase in a flow rate. In the actual experiment, the application of 3/8" tubes and orifice to take into account the momentum drop resulted in an increase in the hazardous distance of about 1.95 m. Comparing three methods was able to identify similarities between real and CFD, and also similarities and limitations of CFD and KGS HAC. We hope these results will provide a good basis for future experiments and risk calculations.