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
- 2022.10
- 수록면
- 3,888 - 3,896 (9page)
- DOI
- 10.1016/j.net.2022.05.020
이용수
초록· 키워드
오류제보하기
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 keff 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 1s 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%
목차
등록된 정보가 없습니다.
참고문헌 (0)
참고문헌 신청함께 읽어보면 좋을 논문
논문 유사도에 따라 DBpia 가 추천하는 논문입니다. 함께 보면 좋을 연관 논문을 확인해보세요!
-
Verification of OpenMC for fast reactor physics analysis with China experimental fast reactor start-up tests
Nuclear Engineering and Technology
2022 .10
-
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
-
Neutronics analysis of TRIGA Mark II research reactor
Nuclear Engineering and Technology
2018 .01
-
Analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/BVIII.0 nuclear data library
Nuclear Engineering and Technology
2020 .12
-
Criticality benchmarking of ENDF/B-Ⅷ.0 and JEFF-3.3 neutron data libraries with RMC code
Nuclear Engineering and Technology
2020 .09
-
Copper neutron transport libraries validation by means of a 252 Cf standard neutron source
Nuclear Engineering and Technology
2021 .10
-
Current Status of ACE Format Libraries for MCNP at Nuclear Data Center of KAERI
방사선방어학회지
2016 .01
-
Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016
Nuclear Engineering and Technology
2017 .01
-
Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor
Nuclear Engineering and Technology
2018 .01
-
Inverse method to obtain reactivity in nuclear reactors with P1 point reactor kinetics model using matrix formulation
Nuclear Engineering and Technology
2021 .02
-
Spent fuel characterization analysis using various nuclear data libraries
Nuclear Engineering and Technology
2022 .09
-
Supercritical CO2-cooled fast reactor and cold shutdown system for ship propulsion
Nuclear Engineering and Technology
2024 .03
-
Evaluation of neutronics parameters during RSG-GAS commissioning by using Monte Carlo code
Nuclear Engineering and Technology
2023 .05
-
Sensitivity and Uncertainty quantification of neutronic integral data in the TRIGA Mark II Research Reactor
Nuclear Engineering and Technology
2022 .02
-
An inter-comparison between ENDF/B-VIII.0-NECP-Atlas and ENDF/B-VIII.0-NJOY results for criticality safety benchmarks and benchmarks on the reactivity temperature coefficient
Nuclear Engineering and Technology
2021 .08
-
Use of similarity indexes to identify spatial correlations of sodium void reactivity coefficients
Nuclear Engineering and Technology
2020 .11
-
The reactivity of SO₂ and NO₂ on different mineral oxides
한국에너지학회 학술발표회
2018 .04
-
JSI TRIGA fuel rod reactivity worth experiments for validation of Serpent-2 and RAPID fuel burnup calculations
Nuclear Engineering and Technology
2024 .08
-
Study on Nuclear Test Scheme and Test Results for Safety Neutron Measurement System in Research Reactor
한국방사성폐기물학회 학술대회
2017 .01
-
A critical study on best methodology to perform UQ for RIA transients and application to SPERT-III experiments
Nuclear Engineering and Technology
2022 .05
이 논문의 저자 정보
최근 본 자료
댓글(0)
0