• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2007년도 학술논문요약집
• /
• pp.77-78
• /
• 2007

• Nuclear Engineering and Technology
• /
• 제55권8호
• /
• pp.2823-2834
• /
• 2023

Austenitic stainless steel welds (ASSWs) of nuclear components undergo aging-related degradations caused by high temperature and neutron radiation. Since irradiation leads to the change of material characteristics, relevant quantification is important for long-term operation, but limitations exist. Although ion irradiation is utilized to emulate neutron irradiation, its penetration depth is too shallow to measure bulk properties. In this study, a systematic approach was suggested to estimate mechanical properties of ion irradiated 308 ASSW. First of all, weld specimens were irradiated by 2 MeV proton to 1 and 10 dpa. Microstructure evolutions due to irradiation in δ-ferrite and austenite phases were characterized and micropillar compression tests were performed. In succession, dislocation density based stress-strain (S-S) relationships and quantification models of irradiation defects were adopted to define phases in finite element analyses. Resultant microscopic S-S curves were compared to verify material parameters. Finally, macroscopic behaviors were calculated by multiscale simulations using real microstructure based representative volume element (RVE). Validity of the approach was verified for the unirradiated specimens such that the estimated S-S curves and 0.2% offset yield strengths (YSs) which was 363.14 MPa were in 10% agreement with test. For irradiated specimens, the estimated YS were 917.41 MPa in 9% agreement.

• Nuclear Engineering and Technology
• /
• 제54권5호
• /
• pp.1789-1803
• /
• 2022

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in k_eff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.1177-1178
• /
• 2005

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.335-341
• /
• 1997

The generic analysis method for core flow instability is suggested to confirm that the core flow instability would not occur on PWR conditions. For the confirmation, the stability criteria of each fuel type are provided. Instability investigations in various accident conditions prove that the locked rotor accident is the most limiting case to instability. Parametric Effects are surveyed and in good agreement with available studies. The effects of heat flux distribution become negligible as the subcooling number is decreased. The power margin to instability is calculated quantitatively in various accident conditions.

• 핵의학기술
• /
• 제15권1호
• /
• pp.51-57
• /
• 2011

광대역 재구성(wide beam reconstruction, WBR) 기법인 Xpress.cardiac$^{TM}$ 프로그램을 적용하여 기존 OSEM (ordered subsets expectation maximization) 기법과 심근 내 분절별 관류와 국소벽 운동에서의 일치율을 확인하여 WBR 기법의 임상적 유용성을 알아보고자 하였다. 관상동맥질환의 병력이 없고 핵의학 전문의에 의한 판독상 이상소견이 없는 총 20명(남7명, 여자13명: 정상군)과 관상동맥질환을 진단받은 총 10명(남6명, 여자4명: 비정상군)을 대상으로 휴식기 $^{201}Tl$/부하기 $^{99m}Tc$-MIBI 심근관류 SPECT를 실시하였다. 영상 획득과 재구성은 휴식기 시 투사영상당 30초, 곧바로 15초씩 영상을 얻고 부하기 시 투사영상 당 25초, 곧바로 13초씩 영상을 얻어 OSEM과 WBR 기법을 적용하였고 심근 내 분절별 관류과 국소벽 운동은 AutoQuant 프로그램의 QPS/QGS 알고리즘의 20분절 모델을 적용하였다. 관류상태는 5등급(0=정상, 1=경도, 2=중등도, 3=심한 결손, 4=섭취 없음), 국소벽 운동은 5등급(0=정상, 1=경도, 2=중등도, 3=심한운동저하, 4=무운동)으로 분류한 반정량값을 이용해 기존 OSEM 기법과 WBR 기법에서의 일치율을 평가하였다. 정상군에서 기존 OSEM 기법과 WBR 기법에서의 일치율은 휴식기 시 분절별 관류에서 99% (396/400, k=0.662, p<0.0001), 국소벽 운동에서 83.8% (335/400, k=0.283), 부하기 시 분절별 관류에서 95.8% (383/400, k=0.656), 국소벽 운동에서 87.3% (349/400, k=0.390)의 일치율을 보였다. 비정상군에서 휴식기 시 분절별 관류에서 83% (166/200, k=0.605), 국소벽 운동에서 55.5% (111/200, k=0.385), 부하기 시 분절별 관류에서 79.5% (159/200, k=0.682), 국소벽 운동에서 63.5% (127/200, k=0.486)의 일치율을 보였다. 관상동맥 질환의 진단 및 예후 예측에 있어 중요한 의미를 갖는 심근 내 분절별 관류와 국소벽 운동 기능 평가의 지표들을 이용한 WBR 기법은 기존 OSEM 기법과 비교하여 정상 비정상군 모두에서 심근 내 분절별 관류의 일치율은 높았지만 국소벽 운동에서는 의미 있게 낮은 일치율을 보였다. WBR 기법은 높은 해상도와 대조도를 제공할 수 있다고 하나 심근관류 SPECT에서의 적용은 유용성이 떨어진다고 사료된다.

• Nuclear Engineering and Technology
• /
• 제54권7호
• /
• pp.2635-2649
• /
• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• Nuclear Engineering and Technology
• /
• 제55권11호
• /
• pp.4048-4056
• /
• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

• 방사선산업학회지
• /
• 제9권1호
• /
• pp.37-46
• /
• 2015

This study examines mutual perceptions between general public and nuclear plant employees on understanding nuclear policy communication applying the co-orientation model. The total of 414 responses were analyzed including 211 of the general public and 203 of plant employees. Results indicate that agreement between general public and plant employees is relatively high, in that general public tends to have negative evaluation to nuclear policy communication, but plant employees tends to have positive one. In terms of congruence, general public perceive that plant employees might have more positive evaluation than themselves, and nuclear plant employees perceive that general public might have more negative evaluation than themselves. Finally, in terms of accuracy, general public accurately estimate how nuclear plant employees perceive on policy communication, whereas nuclear plant employees unaccurately estimate how general public perceive on policy communication.

• Nuclear Engineering and Technology
• /
• 제53권2호
• /
• pp.562-571
• /
• 2021

In this work, we propose a new mechanistic model for the treatment of helium behaviour which includes the description of helium solubility in oxide fuel. The proposed model has been implemented in SCIANTIX and validated against annealing helium release experiments performed on small doped fuel samples. The overall agreement of the new model with the experimental data is satisfactory, and given the mechanistic formulation of the proposed model, it can be continuously and easily improved by directly including additional phenomena as related experimental data become available.