지원사업
학술연구/단체지원/교육 등 연구자 활동을 지속하도록 DBpia가 지원하고 있어요.
커뮤니티
연구자들이 자신의 연구와 전문성을 널리 알리고, 새로운 협력의 기회를 만들 수 있는 네트워킹 공간이에요.
논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2016.6
- 수록면
- 642 - 649 (8page)
이용수
초록· 키워드
오류제보하기
Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5?1 pcm.
목차
등록된 정보가 없습니다.
참고문헌 (13)
참고문헌 신청
C.H. Miller / 1974 / Some basic physics aspects of the Canadian Nuclear Power Program / Joint CAP-CAS Congress '74, Memorial University of Newfoundland
G. Roh / 2011 / Improvement of power coefficient by using burnable poison in the CANDU reactor / Nucl. Eng. Des 241:1565~1578
T.E. Schaubel / CNSC staff review of Pickering NGS-B integrated safety review: Safety analysis safety factors report, Letter to D.P. McNeil
M. Ouisloumen / 1991 / A model for neutron scattering off heavy isotopes that accounts for thermal agitation effects / Nucl. Sci. Eng 107:189~200
W. Rosenstein / 1996 / Neutron scattering kernels in pronounced resonances for stochastic Doppler effect calculations / Ann. Nucl. Energy 23:441~458
G. Roh / 2011 / Improvement of power coefficient by using burnable poison in the CANDU reactor / Nucl. Eng. Des 241:1565~1578
T.E. Schaubel / CNSC staff review of Pickering NGS-B integrated safety review: Safety analysis safety factors report, Letter to D.P. McNeil
M. Ouisloumen / 1991 / A model for neutron scattering off heavy isotopes that accounts for thermal agitation effects / Nucl. Sci. Eng 107:189~200
W. Rosenstein / 1996 / Neutron scattering kernels in pronounced resonances for stochastic Doppler effect calculations / Ann. Nucl. Energy 23:441~458
함께 읽어보면 좋을 논문
논문 유사도에 따라 DBpia 가 추천하는 논문입니다. 함께 보면 좋을 연관 논문을 확인해보세요!
-
NEUTRONICS INVESTIGATION OF CANADA DEUTERIUM URANIUM 6 REACTOR FUELED (TRANSURANICeTH) O2 USING A COMPUTATIONAL METHOD
Nuclear Engineering and Technology
2015 .01
-
Effect of DUPIC Cycle on CANDU Reactor Safety Parameter
Nuclear Engineering and Technology
2016 .01
-
Analysis of the CREOLE experiment on the reactivity temperature coefficient of the UO2 light water moderated lattices using Monte Carlo transport calculations and ENDF/B-VII.1 nuclear data library
Nuclear Engineering and Technology
2020 .01
-
Spent fuel characterization analysis using various nuclear data libraries
Nuclear Engineering and Technology
2022 .09
-
Strain‑Dependent Polar Optical Phonon Scattering and Drive Current Optimization in Nanoscale Monolayer MoS2 FETs
Electronic Materials Letters
2020 .01
-
Impact of fuel temperature on nuclear core design calculations
Nuclear Engineering and Technology
2024 .09
-
SARAPAN-A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses
Nuclear Engineering and Technology
2017 .01
-
Improving the Neutronic Characteristics of a Boiling Water Reactor by Using Uranium Zirconium Hydride Fuel Instead of Uranium Dioxide Fuel
Nuclear Engineering and Technology
2016 .06
-
Simulation of low-enriched uranium burnup in Russian VVER-1000 reactors with the Serpent MonteCarlo code
Nuclear Engineering and Technology
2021 .09
-
Monte-Carlo 시뮬레이션 기반 Power Curve의 전력변환 오차 개선에 관한 연구
대한전기학회 학술대회 논문집
2019 .07
-
Evaluate of the Spent Nuclear Fuel Rod Model Using the Monte Carlo Simulation
한국방사성폐기물학회 학술대회
2017 .01
-
Modelling of the Spent Nuclear Fuel Assembly Using the Monte Carlo Method
한국방사성폐기물학회 학술대회
2017 .01
-
Measurement of deuterium concentration in heavy water utilizing prompt gamma neutron activation analysis (PGNAA) in comparison with MCNPX simulation results
Nuclear Engineering and Technology
2022 .11
-
Experimental study of the influence of borehole parameters on prompt fission neutron uranium logging and its corrections
Nuclear Engineering and Technology
2024 .08
-
Neutronic Study of Utilization of Discrete Thorium-Uranium Fuel Pins in CANDU-6 Reactor
Nuclear Engineering and Technology
2019 .01
-
표준태풍 생성을 위한 Monte-Carlo 시뮬레이션
대한토목학회 학술대회
2015 .10
-
Physics Study of Canada Deuterium Uranium Lattice with Coolant Void Reactivity Analysis
Nuclear Engineering and Technology
2017 .01
-
Domain Decomposition Strategy for Pin-Wise Full-Core Monte Carlo Depletion Calculation with the Reactor Monte Carlo Code
Nuclear Engineering and Technology
2016 .06
-
Systems Analyses of Alternative Technologies for the Recovery of Seawater Uranium
방사성폐기물학회지
2018 .01
-
Option values of quantity flexibility in nuclear fuel long-term contracts: an application of the least squares Monte Carlo
Geosystem Engineering
2020 .01
이 논문의 저자 정보
최근 본 자료
댓글(0)
0