• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1099-1105
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• 2018

How to choose defaults in risk-informed regulations depends on the conservatism implicated in regulatory defaults. Without a universal agreement on the approaches dealing with the conservatism of defaults, however, the desirability of conservatism in regulatory risk analyses has long been controversial. The opponent views it as needlessly costly and irrational, and the proponent as a form of protection against possible omissions or underestimation of risks. Moreover, the inherent ambiguity of risk makes it difficult to set suitable defaults in terms of risk. This paper, the extension of the previous work [1], focuses on the effects of different levels of conservatism implicated in regulatory defaults on the estimates of risk. According to the postulated behaviors of regulated parties and the diversity of interests of regulators, in particular, various measures for evaluating the effect of conservatism in defaults are developed and their properties are explored. In addition, a simple decision model for setting regulatory defaults is formulated, based on the understanding of the effect of conservatism implicated in them. It can help decision makers evaluate the levels of safety likely to result from their regulatory policies.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2367-2375
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• 2022

In this work, we propose a new mechanistic model for the treatment of helium behaviour at the grain boundaries in oxide nuclear fuel. The model provides a rate-theory description of helium inter-granular behaviour, considering diffusion towards grain edges, trapping in lenticular bubbles, and thermal resolution. It is paired with a rate-theory description of helium intra-granular behaviour that includes diffusion towards grain boundaries, trapping in spherical bubbles, and thermal re-solution. The proposed model has been implemented in the meso-scale software designed for coupling with fuel performance codes SCIANTIX. It is validated against thermal desorption experiments performed on doped UO₂ samples annealed at different temperatures. The overall agreement of the new model with the experimental data is improved, both in terms of integral helium release and of the helium release rate. By considering the contribution of helium at the grain boundaries in the new model, it is possible to represent the kinetics of helium release rate at high temperature. Given the uncertainties involved in the initial conditions for the inter-granular part of the model and the uncertainties associated to some model parameters for which limited lower-length scale information is available, such as the helium diffusivity at the grain boundaries, the results are complemented by a dedicated uncertainty analysis. This assessment demonstrates that the initial conditions, chosen in a reasonable range, have limited impact on the results, and confirms that it is possible to achieve satisfying results using sound values for the uncertain physical parameters.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1280-1286
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• 2023

MgF₂ as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT). Regarded as important for applications, the thermal neutron scattering data of MgF₂ were calculated, based on the phonon expansion model. The structural properties of MgF₂ were researched by the VASP code based on the ab-initio methods. The PHONOPY code was employed to calculate the phonon density of states. Furthermore, the NJOY code was used to calculate the thermal neutron scattering data of MgF₂. The calculated inelastic cross sections plus absorption cross sections are in agreement with the available experimental data. The neutron transport in the BSA has been simulated by using a hybrid Monte-Carlo-Deterministic code NECP-MCX. The results indicated that compared with the calculation of the free gas model, the thermal neutron flux and epithermal neutron flux at the BSA exit port calculated by using the thermal neutron scattering data of MgF₂ were reduced by 27.7% and 8.2%, respectively.

• Journal of Mechanical Science and Technology
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• 제14권8호
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• pp.816-822
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• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.99-108
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• 2006

SNU-RCCS is a water pool type RCCS (Reactor Cavity Cooling System) developed for VHTR (Very High Temperature Reactor) application by SNU (Seoul National University). Since radiation heat transfer is the major process of passive heat removal in a RCCS, it is important to determine the precise emissivity of the reactor vessel. Review studies have used a constant emissivity in the passive heat removal analysis, even though the emissivity depends on many factors such as temperature, surface roughness, oxidation level, wavelength, direction, atmosphere conditions, etc. Therefore, information on the emissivity of a given material in a real RCCS is essential in order to properly analyze the radiation heat transfer in a VHTR. The objectives of this study are to develop a method for compensation of the factors affecting the emissivity measurement using an infrared thermometer and to estimate the true emissivity from the measured emissivity via the developed method, especially in the SNU-RCCS environment. From this viewpoint, we investigated factors such as the attenuation effect of the window, filling gas, and the effect of background radiation on the emissivity measurements. The emissivity of the vessel surface of the SNU-RCCS facility was then measured using a sight tube. The background radiation was subsequently removed from the measured emissivity by solving a simultaneous equation. Finally, the calculated emissivity was compared with the measured emissivity in a separate emissivity measurement device, yielding good agreement with the emissivity increase with vessel temperature in a range of 0.82 to 0.88.

• Nuclear Engineering and Technology
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• 제10권3호
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• pp.137-144
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• 1978

계측장치에서 관측한 계수율에 대한 비직선성 교정을 결정하는 한 방법을 제시하였다. 불감시간은 계수율이 의존성을 가지고 있다는 가정하에서 발전시킨 한 표현식은 실험결과와 잘 일치하였다. 계측장치의 펄스전압의 계수율에 따른 ,변화량을 파고분석장치를 이용하여 측정 하였다. 이 방법은 광범한 계수율 범위에 걸치어 불감시간에 대한 정착찬 값을 주었다. 개요한 이 측정기술로 불감손실로 인한 비직선성의 정확한 교정할 수 있었다. 불감시간은 계수율이 증가함에 따라 감소하는 것이 관측되었다.

• Nuclear Engineering and Technology
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• 제40권2호
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• pp.133-138
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• 2008

The use of Supercritical Fluids(SCF) has been proposed for numerous power cycle designs as part of the Generation IV advanced reactor designs, and can provide for higher thermal efficiency. One particular area of interest involves the behavior of SCF during a blowdown or depressurization process. Currently, no data are available in the open literature at supercritical conditions to characterize this phenomenon. A preliminary computational analysis, using a homogeneous equilibrium model when a second phase appears in the process, has shown the complexity of behavior that can occur. Depending on the initial thermodynamic state of the SCF, critical flow phenomena can be characterized in three different ways; the flow can remain in single phase(high temperature), a second phase can appear through vaporization(high pressure low temperature) or condensation(high pressure, intermediate temperature). An experimental facility has been built at the University of Wisconsin to study SCF depressurization through several diameter breaks. The preliminary results obtained show that the experimental data can be predicted with good agreement by the model for all the different initial conditions.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.415-420
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• 2013

A determination method for $^{237}Np$ in spent nuclear fuel samples was developed using an isotope dilution method with $^{239}Np$ as a spike. In this method, inductively coupled plasma mass spectrometry (ICP-MS) was taken for the $^{237}Np$ instead of the previously used alpha spectrometry. $^{237}Np$ and $^{239}Np$ were measured by ICP-MS and gamma spectrometry, respectively. The recovery yield of $^{237}Np$ in synthetic samples was $95.9{\pm}9.7$% (1S, n=4). The $^{237}Np$ contents in the spent fuel samples were 0.15, 0.25, and $1.06{\mu}g/mgU$ and these values were compared with those from ORIGEN-2 code. A fairly good agreement between the measurements (m) and calculations (c) was obtained, giving ratios (m/c) of 0.93, 1.12 and 1.25 for the three PWR spent fuel samples with burnups of 16.7, 19.0, and 55.9 GWd/MtU, respectively.

• 대한의용생체공학회:의공학회지
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• 제26권1호
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• pp.17-22
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• 2005

Small animal PET using a dual layer phoswich detector has been developed to obtain high and uniform spatial resolution. In this study, a simple analytic model to optimize the lengths of a dual layer phoswich detector was derived and validated by Monte Carlo simulation. For a small animal PET scanner with a 10㎝ ring diameter, the optimal length of the phoswich detector consisting of various crystal materials, such as LSO and LuYAP, were calculated analytically and validated using GATE. The detector module consisted of 8×8 arrays of crystals, with each phoswich detector element having a 2㎜×2㎜ sensitive area. The total crystal length was fixed to 20㎜. The optimal lengths of the phoswich detector layers, as functions of the crystal materials and order, conveniently derived by the analytic equation, showed good agreement with those estimated by the time consuming simulation. The simple analytical model can be used for the fast and accurate design of an optimal phoswich detector for small animal PET to achieve high spatial resolution and uniformity.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2788-2802
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• 2021

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MW_th metal-fueled SFR (MET-1000) and a 3600 MW_th oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core k_eff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core k_eff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.