• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Nuclear Engineering and Technology
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• v.13 no.3
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• pp.153-161
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• 1981

The dynamic behavior of the Ko-ri Unit 1 nuclear reactor following some credible and postulated accidents has been analyzed to a certain extent by means of neutronics and temperature equations formulated in terms of point reactor model. In general, the result of numerical calculation is harnessed to be incorporated in more elaborate models so as to predict transient behavior in a reliable mode as a part of accident analysis. It is shown in the case of power response upon an uncontrolled withdrawal of rod cluster control assembly at hot full power that the point reactor kinetics model proves to be good enough to reproduce the generic features described in the final safety analysis report of the Ko-ri Unit 1.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.36-46
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• 2004

The ISAAC fission product release calculation is based on detailed FPRAT models developed by Jaycor. For volatile fission product release calculations, either the Cubicciotti steam oxidation correlation or the NUREG-0772 correlation is used. In this study, evaluation is carried out for these volatile fission product release models. As a result, in the case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, the NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option is evaluated to show mitigated conservative results. In addition, a sensitivity study on detailed core nodalization is performed. In the study, 380 horizontal fuel channels in the Wolsong plant are nodalized into 12 (6 channels per loop, $3{\times}3$ Core Pass) representative channels and detailed by 16/20/24 channels. For reference accidents, LOAH and large LOCA are selected as representing high and low pressure sequences, respectively. According to the results, the original 12 channel approach with $3{\times}3$ core passes is evaluated to be sufficient as an optimal scheme.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.355.2-355
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• 1999

The spnng back phenomena occurring in the coiling process of a steam generator tube induces the dimensional inaccuracy and makes the coiling procedure difficult. In this research, an analytical model was developed to evaluate the amount of the spring back for SMART steam generator tubes. The model was developed on the basis of beam theory and elastic-perfectly plastic material property. This model was extended to consider the effect of plastic hardening and the effect of the tensile force on the spring back phenomena. Parametric studies were performed for various design variables of steam generator tubes in order to minimize the spring back in the design stage. A sensitivity analysis has shown that the low yield strength, the high elastic modulus, the small helix diameter, and the large tube diameter result in a small amount of the spring back. The amount of the spring back can be controlled by the selection of adequate design values in the basic design stage and reduced to an allowable limit by the application of the tensile force to the tube during the coiling process.rocess.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.745-754
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• 2012

The correlations between Large Early Release Frequency (LERF) and Early Fatality need to be investigated for risk-informed application and regulation. In Regulatory Guide (RG) -1.174, while there are decision-making criteria using the measures of Core Damage Frequency (CDF) and LERF, there are no specific criteria on LERF. Since there are both huge uncertainties and large costs needed in off-site consequence calculation, a LERF assessment methodology needs to be developed, and its correlation factor needs to be identified, for risk-informed decision-making. A new method for estimating off-site consequence has been presented and performed for assessing health effects caused by radioisotopes released from severe accidents of nuclear power plants in this study. The MACCS2 code is used for validating the source term quantitatively regarding health effects, depending on the release characteristics of radioisotopes during severe accidents. This study developed a method for identifying correlations between LERF and Early Fatality and validates the results of the model using the MACCS2 code. The results of this study may contribute to defining LERF and finding a measure for risk-informed regulations and risk-informed decision-making.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.600-607
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• 2010

Rhenium-186, having a half-life of 90.64 h, is an important radionuclide, used in metabolic radiotherapy and radio immunotherapy. $^{186}Re$ hydroxyethylidene diphosphonate (HEDP) is a new compound used for the palliation of painful skeletal metastases. Its production is achieved via charged-particle-induced reactions; the data are available in EXFOR library. For the work discussed in this paper, production of $^{186}Re$ was done via $^{nat}W(p,n)^{186}Re$ nuclear reaction. Pellets of $^{nat}W$ were used as targets and were irradiated with 15, 17.5, 20, 22.5, 25 MeV proton beams at 5 ${\mu}A$ current. The radiochemical separation was performed by the ion exchange chromatography method. The production yield achieved at 25 MeV was 1.91 $MBq{\cdot}{\mu}A^{-1}{\cdot}h^{-1}$. Excitation functions for the $^{186}Re$ radionuclide, via $^{186}W(p,n)^{186}Re$ and $^{186}W(d,2n)^{186}Re$ reactions were calculated by ALICE-ASH and TALYS-1.0 codes to validate and fit the experimental data and to obtain a recommended set of data for $^{186}W(p,n)^{186}Re$ reaction. Required thickness of the targets was obtained by SRIM code for each reaction.

• Nuclear Engineering and Technology
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• v.37 no.3
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• pp.265-272
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• 2005

Severe accidents in nuclear power plants can cause hydrogen-generating chemical reactions, which create the danger of hydrogen combustion and thus threaten containment integrity. For containment analyses, a three-dimensional mechanistic code, GOTHIC-3D has been applied near source compartments to predict whether or not highly reactive gas mixtures can form during an accident with the hydrogen mitigation system working. To assess the code applicability to hydrogen combustion analysis, this paper presents the numerical calculation results of GOTHIC-3D for various hydrogen combustion experiments, including FLAME, LSVCTF, and SNU-2D. In this study, a technical base for the modeling oflarge- and small-scale facilities was introduced through sensitivity studies on cell size and bum modeling parameters. Use of a turbulent bum option of the eddy dissipation concept enabled scale-free applications. Lowering the bum parameter values for the flame thickness and the bum temperature limit resulted in a larger flame velocity. When applied to hydrogen combustion analysis, this study revealed that the GOTHIC-3D code is generally able to predict the combustion phenomena with its default bum modeling parameters for large-scale facilities. However, the code needs further modifications of its bum modeling parameters to be applied to either small-scale facilities or extremely fast transients.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.