• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2953-2959
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• 2021

A new practical modeling of the Norton's power law creep is proposed and implemented to analyze the high temperature behaviors of Alloy 690 SG tube material. In the model, both the stress exponent n and the rate constant B are simply treated as the temperature dependent parameters. Based on the two-step optimization procedure, the temperature function of the rate constant B(T) was determined for the data set of each B value after fixing the stress exponent n value by using the prior optimized function at each temperature. This procedure could significantly reduce the numerical errors when using the power law creep equations. Based on the better description of the steady-state creep rates, the experimental rupture times could also be well predicted by using the Monkman-Grant relationship. Furthermore, the difference in tensile strengths at high temperatures could be very well estimated by assuming the imaginary creep stress related to the given strain rate after correcting the temperature effects on the elastic modulus.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.455-468
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• 2022

Colloid Formation and Migration (CFM) project is being carried out within the Grimsel Test Site (GTS) Phase Ⅵ. Since 2008, the Korea Atomic Energy Research Institute (KAERI) has joined CFM to investigate the behavior of colloid-facilitated radionuclide transport in a generic Underground Research Laboratory (URL). The CFM project includes a long-term in-situ test (LIT) and an in-rock bentonite erosion test (i-BET) to assess the in-situ colloid-facilitated radionuclide transport through the bentonite erosion in the natural flow field. In the LIT experiment, radionuclide-containing compacted bentonite was equipped with a triple-packer system and then positioned at the borehole in the shear zone. It was observed that colloid transport was limited owing to the low swelling pressure and low hydraulic conductivity. Therefore, a postmortem analysis is being conducted to estimate the partial migration and diffusion of radionuclides. The i-BET experiment, that focuses more on bentonite erosion, was newly designed to assess colloid formation in another flow field. The i-BET experiment started with the placement of compacted bentonite rings in the double-packer system, and the hydraulic parameters and bentonite erosion have been monitored since December 2018.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Journal of Radiation Protection and Research
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• v.24 no.1
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• pp.31-38
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• 1999

A real-time dose assessment system(FADAS : Following Accident Dose Assessment System) has been developed for the real-time accident consequence assessment against a nuclear accident. Field tracer experiment near Younggwang nuclear power plant was performed to improve the accuracy of developed system and to parameterize the site-specific parameters into the FADAS. The mean values and turbulent components of wind profile obtained through field experiment have been reflected to FADAS with site-specific conditions. The simulated results of diffusion model agreed well with the measured data through tracer experiment. The developed system is being used as a basic module of emergency preparedness system in Korea. The diffusion model which can be reflected site-specific parameters will be improved through field experiments continuously.

• Journal of Radiation Protection and Research
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• v.25 no.1
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• pp.11-19
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• 2000

The sensitivity analysis of input parameters for a dynamic food chain model DYNACON was conducted as a function of deposition date for the long-lived radionuclides $(^{137}Cs,\;^{90}Sr)$. Also, the influence of input parameters for the short and long-terms contamination of selected foodstuffs (cereals, leafy vegetables, milk) was investigated. The input parameters were sampled using the LHS technique, and their sensitivity indices represented as PRCC. The sensitivity index was strongly dependent on contamination period as well as deposition date. In case of deposition during the growing stages of plants, the input parameters associated with contamination by foliar absorption were relatively important in long-term contamination as well as short-term contamination. They were also important in short-term contamination in case of deposition during the non-growing stages. In long-term contamination, the influence of input parameters associated with foliar absorption decreased, while the influence of input parameters associated with root uptake increased. These phenomena were more remarkable in case of the deposition of non-growing stages than growing stages, and in case of $^{90}Sr$ deposition than $^{137}Cs$ deposition. In case of deposition during growing stages of pasture, the input parameters associated with the characteristics of cattle such as feed-milk transfer factor and daily intake rate of cattle were relatively important in contamination of milk.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.283-296
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• 2008

The numerical simulations for groundwater flow were carried out to support the input parameters for safety assessment in LILW repository site. As the input parameters for safety assessment, the groundwater flux into the underground facilities during construction, flow rate through the disposal silo after closure of disposal silo and flow pathway from the disposal silo to discharge area were analyzed using the 10 cases groundwater flow simulations. From the total 10 numerical simulation results, the statistics of estimated output were similar to among 10 cases. In some cases, the analyzed input parameters were strongly governed by locally existed high permeable fracture zone at radioactive waste disposed depth. Indeed, numerical simulation for well scenario as a human intrusion scenario was carried out using the hydraulically severe case model. Using the results of well scenario, the input parameters for safety assessment were also obtained through the numerical simulation.

• KSTLE International Journal
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• v.4 no.2
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• pp.47-51
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• 2003

Sliding wear tests have been carried out in room temperature air and water in order to compare the wear resistance of Zr-xNb-xSn alloys of various alloying elements (Nb and Sn). The main focus was to quantitatively compare the wear properties of the recently developed Zr-xNb-xSn alloys with the commercial ones using the evaluation parameters of the wear resistance with the consideration of the worn area. As a result, the recently developed alloys had a similar wear resistance compared with the commercial ones. The dominant factor governing the wear resistance was the protruded volume of the wear debris that was formed on the worn area in the air condition, but the accommodation of the plastic deformation on the contact area in water. In addition, the worn area size appeared to be very different depending on the tested alloys. To evaluate the wear resistance of each test specimen, the ratio of the wear volume or the protruded volume to the worn area ($D_e$ or $D_p$) is investigated and proposed as the evaluation parameters of the wear resistance.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.461-468
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• 2020

The improvements of the DeCART/MUSAD code system for uncertainty analysis of HTGR neutronic parameters are presented in this paper. The function for quantifying an uncertainty of critical-spectrumweighted few group cross section was implemented using the generalized adjoint B₁ equation solver. Though the changes between the infinite and critical spectra cause a considerable difference in the contribution by the graphite scattering cross section, it does not significantly affect the total uncertainty. To reduce the number of iterations of the generalized adjoint transport equation solver, the generalized adjoint B₁ solution was used as the initial value for it and the number of iterations decreased to 50%. To reflect the implicit uncertainty, the correction factor was derived with the resonance integral. Moreover, an additional correction factor for the double heterogeneity was derived with the effective cross section of the DH region and it reduces the difference from the complete uncertainty. The code system was examined with the MHTGR-350 Ex.II-2 3D core benchmark. The k_eff uncertainty for Ex.II-2a with only the fresh fuel block was similar to that of the block and the uncertainty for Ex.II-2b with the fresh fuel and the burnt fuel blocks was smaller than that of the fresh fuel block.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.361-371
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• 2000

The 123Te(p,n)123I, 124Te(p,n)124I and 124Te(p,2n)123I reactions, among the many reaction channels opened, are the major reactions under consideration from a diagnostic purpose because reaction residuals as the gamma emitters are used for most radiophamaceutical applications involving radioiodine. Based on the available experimental data, the absorption cross sections and elastic scattering angular distributions of the proton-induced nuclear reaction on Te isotopes below 60 MeV are calculated using the optical model code APMNK. The transmission coefficients of neutron, proton, deuteron, trition and alpha particles are calculated by CUNF code and are fed into the GNASH code. By adjusting level density parameters and the pair correction values of some reaction channels, as well as the composite nucleus state density constants of the pre-equilibrium model, the production cross sections and energy-angle correlated spectra of the secondary light particles, as well as production cross sections and energy distributions of heavy recoils and gamma rays are calculated by the statistical plus pre-equilibrium model code GNASH. The calculated results are analysed and compared with the experimental data taken from the EXFOR. The optimized global optical model parameters give overall agreement with the experimental data over both the entire energy range and all tellurium isotopes.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.3
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• pp.307-312
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• 2001

Ultrasonic signals from Charpy impact test specimen have been analyzed in order to evaluate the integrity of reactor pressure vessel. Base and weld metal that were extracted from reactor vessel doting plant outages according to the schedule of the surveillance test required by the related regulations have been used and the ultrasonic test parameters including velocity, attenuation, etc. showed a close correlations with the amount of neutron irradiation for base metal, relatively homogeneous materials. This result showed certain possibility where a nondestructive method could be used to predict the fluence of the Irradiation due to neutron in nuclear reactor vessel materials.