• Nuclear Engineering and Technology
• /
• 제46권3호
• /
• pp.313-342
• /
• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• Nuclear Engineering and Technology
• /
• 제51권2호
• /
• pp.369-376
• /
• 2019

A complete solution for a soluble-boron-free (SBF) small modular reactor (SMR) is pursued with a new burnable absorber concept, namely centrally-shielded burnable absorber (CSBA). Neutronic flexibility of the CSBA design has been discussed with fuel assembly (FA) analyses. Major design parameters and goals of the SBF SMR are discussed in view of the reactor core design and three CSBA designs are introduced to achieve both a very low burnup reactivity swing (BRS) and minimal residual reactivity of the CSBA. It is demonstrated that the core achieves a long cycle length (~37 months) and high burnup (~30 GWd/tU), while the BRS is only about 1100 pcm and the radial power distribution is rather flat. This research also introduces a supplementary reactivity control mechanism using stainless steel as mechanical shim (MS) rod to obtain the criticality during normal operation. A further analysis is performed to investigate the local power peaking of the CSBA-loaded FA at MS-rodded condition. Moreover, a simple $B_4C$-based control rod arrangement is proposed to assure a sufficient shutdown margin even at the cold-zero-power condition. All calculations in this neutronic-thermal hydraulic coupled investigation of the 3D SBF SMR core are completed by a two-step Monte Carlo-diffusion hybrid methodology.

• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3888-3896
• /
• 2022

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm k_eff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of ⁵²Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3897-3908
• /
• 2022

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on "Neutronics Benchmark of CEFR Start-Up Tests" offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

• 한국전산유체공학회:학술대회논문집
• /
• 한국전산유체공학회 2004년도 춘계 학술대회논문집
• /
• pp.214-219
• /
• 2004

The HANARO, a multi-purpose research reactor of 30 MWth, open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. The HANARO is composed of inlet plenum, grid plate, core channel with flow tubes and chimney. The reactor core channel is located at about twelve m (12 m) depth of the reactor pool and cold by the upward flow that the coolant enters the lower inlet of the plenum, rises up through the grid plate and the core channel and exit through the outlet of chimney. A guide tube is extended from the reactor core to the top of the reactor chimney for easily un/loading a target under the reactor normal operation. But active coolant through the core can be Quickly raised up to the top of the chimney through the guide tube by jet flow. This paper is described an analytical analysis to study the flow behavior through the guide tube under reactor normal operation and unloading the target. As results, it was conformed through the analysis results that the flow rate, about fourteen kilogram per second (14 kg/s) suppressed the guide tube jet and met the design cooling flow rate in a circular flow tube, and that the fission moly target cooling flow rate met the minimum flow rate to cool the target.

• Nuclear Engineering and Technology
• /
• 제54권3호
• /
• pp.790-802
• /
• 2022

Large-scale reactor simulation often requires the use of Monte Carlo calculation techniques to estimate important reactor parameters. One drawback of these Monte Carlo calculation techniques is they inevitably result in some uncertainty in calculated quantities. The present study includes parametric uncertainty quantification (UQ) and sensitivity analysis (SA) on the Advanced Test Reactor Critical (ATRC) facility housed at Idaho National Laboratory (INL) and addresses some complications due to Monte Carlo uncertainty when performing these analyses. This approach for UQ/SA includes consideration of Monte Carlo code uncertainty in computed sensitivities, consideration of uncertainty from directly measured parameters and a comparison of results obtained from brute-force Monte Carlo UQ versus UQ obtained from a surrogate model. These methodologies are applied to the uncertainty and sensitivity of k_eff for two sets of uncertain parameters involving fuel plate geometry and fuel plate composition. Results indicate that the less computationally-expensive method for uncertainty quantification involving a linear surrogate model provides accurate estimations for k_eff uncertainty and the Monte Carlo uncertainty in calculated k_eff values can have a large effect on computed linear model parameters for parameters with low influence on k_eff.

• 해양환경안전학회지
• /
• 제29권1호
• /
• pp.76-85
• /
• 2023

신재생에너지로서 수소에 대한 수요가 증가하고 있으나 기존의 화석 연료와 달리 수소는 연료 공급을 위한 전용 충전소가 필요하며, 이러한 인프라 확보를 위해서 수소충전소의 위험성 평가가 선행되어야 한다. 따라서 본 연구에서는 정성적 위험성평가와 정량적 위험성 평가로 구분하여 수소충전소에 대한 위험성평가를 수행하였다. 정성적 평가는 Hazard and Operability Analysis(HAZOP) 기법을 사용하여 Dispenser Module을 두 개의 Node로 평가하였으며, Criticality Estimation Matrix에 따라 Filter의 막힘으로 인한 사고와 고압 사고의 위험도가 High Level로 평가되었다. 정량적 위험성 평가는 Hydrogen Korea Risk Assessment Module(Hy-KoRAM)을 사용하여 화재의 형상과 피해영향범위를 나타냈고, 개인적 위험도와 사회적 위험도에 대한 평가를 수행하였다. 개인적 위험도는 수소충전소로부터 약 100m 떨어진 공공시설 부근까지 추가적인 안전조치가 고려되는 As Low As Reasonably Practicable(ALARP) 구간의 위험도를 보였고, 사회적 위험도 역시 약 10명의 사망자가 발생할 사고빈도가 1E-04/year로 도출되며 ALARP 구간 내에 나타났다. 정성적·정량적 위험성 평가 결과, 공정 단계의 추가적인 안전 조치와 수소충전소 부근의 제한구역 설정을 통하여 안전성 향상 방안을 제시하였다.

• 한국항공우주학회지
• /
• 제45권1호
• /
• pp.63-70
• /
• 2017

본 논문에서는 가항력돛을 이용한 궤도이탈 장치를 설계 및 제작하고 전개 특성을 연구하였다. 형상기억합금을 이용한 분리장치를 개발하고, 형상기억합금의 구동에 따라 테잎 스프링의 복원력을 이용해 구동되는 새로운 궤도이탈 장치를 설계하고 실험하였다. 가항력돛의 효율적인 수납 및 전개를 위해 origami flasher 방식 중 original ISO flasher 방식을 선정하였으며, 반복적인 실험을 하기 위해서 다른 재료들에 비해 저렴한 우주재료인 mylar film을 가항력돛의 재료로 사용하였다. 또한, 일회성 장치 신뢰성 평가 방법 중 하나인 FTA(fault tree analysis) 방법을 통해 장치의 신뢰도(0.997572)를 평가하고 가장 치명도가 높은 부분이 Roller failure임을 확인하였다. 최종적으로 가항력돛의 전개장치의 제작 및 실험을 통하여 향후 궤도이탈 장치의 개발 가능성을 확인하였다.

• 디지털융복합연구
• /
• 제13권1호
• /
• pp.257-262
• /
• 2015

해킹과 같은 사이버 공격의 약 75%가 애플리케이션의 보안 취약점을 악용하기 때문에 안전행정부에서는 코딩 단계에서부터 사이버 공격을 막을 수 있고 보안취약점을 제거할 수 있는 시큐어 코딩 가이드를 제공하고 있다. 본 논문에서는 안행부가 제시한 시큐어 코딩 가이드 7개의 항목들에 대해 AHP기법을 사용하여 우선순위를 찾고 중요도 분석을 하였다. 그 결과 '에러 처리'가 가장 중요한 항목으로 결정되었다. 현재 소프트웨어 감리에 시큐어 코딩에 관한 항목이 없는데, 이 분석 결과는 소프트웨어 개발 과정 중 감리 기준으로 유용하게 사용될 것이다.

• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.788-793
• /
• 2018

High-density spent fuel (SF) storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others.