Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved]
Background This paper discusses the development and benchmarking of IDEA (Indigenous Diffusion Equation Analyzer), an in-house transient, multi-group, 3-dimensional neutron diffusion equation (NDE) code. IDEA can perform both steady-state criticality calculations and reactivity transient analysis fo...
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F1000 Research Ltd
2025-01-01
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| Series: | Nuclear Science and Technology Open Research |
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| Online Access: | https://nstopenresearch.org/articles/3-4/v1 |
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| author | Ubaid ur Rehman Zahid Rasool M. Jawad Hussain Ali Mansoor Farhana Kausar Haseeb ur Rehman Aman ur Rehman |
| author_facet | Ubaid ur Rehman Zahid Rasool M. Jawad Hussain Ali Mansoor Farhana Kausar Haseeb ur Rehman Aman ur Rehman |
| author_sort | Ubaid ur Rehman |
| collection | DOAJ |
| description | Background This paper discusses the development and benchmarking of IDEA (Indigenous Diffusion Equation Analyzer), an in-house transient, multi-group, 3-dimensional neutron diffusion equation (NDE) code. IDEA can perform both steady-state criticality calculations and reactivity transient analysis for 3-dimensional rectangular geometries, with potential extension to cylindrical geometries. Methods The IDEA code employs the finite difference method (FDM) for spatial and temporal discretization, enhanced by source extrapolation-based acceleration. Its performance was evaluated by comparing its results with those of established NDE-based codes, PRIDE and CITATION. Various 2D and 3D steady-state and transient reactivity benchmark problems were used for this comparison. Results The results obtained from IDEA were found to be in close agreement with reference values from publicly available literature. In addition, the IDEA code demonstrated superior accuracy and consistency compared to PRIDE and CITATION in both steady-state and transient reactivity analyses. A neutronic analysis of a typical small PWR core under clean and cold conditions also showed validation against reference results. Conclusions The IDEA code has shown promising figures of merit to be used for practical neutronic calculations. Moreover, a multitude of users can benefit from its further development due to ease of access and lack of proprietary restrictions. |
| format | Article |
| id | doaj-art-62a3ce07ef8e491c833d7ec83565dade |
| institution | DOAJ |
| issn | 2755-967X |
| language | English |
| publishDate | 2025-01-01 |
| publisher | F1000 Research Ltd |
| record_format | Article |
| series | Nuclear Science and Technology Open Research |
| spelling | doaj-art-62a3ce07ef8e491c833d7ec83565dade2025-08-20T02:43:20ZengF1000 Research LtdNuclear Science and Technology Open Research2755-967X2025-01-01310.12688/nuclscitechnolopenres.17629.118912Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved]Ubaid ur Rehman0Zahid Rasool1M. Jawad Hussain2Ali Mansoor3https://orcid.org/0000-0002-2013-3391Farhana Kausar4Haseeb ur Rehman5Aman ur Rehman6Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Islamabad, Islamabad Capital Territory, 45650, PakistanDepartment of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Islamabad, Islamabad Capital Territory, 45650, PakistanDepartment of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Islamabad, Islamabad Capital Territory, 45650, PakistanDepartment of Aerospace and Mechanical Engineering, The Ohio State University Graduate School, Columbus, Ohio, 43210, USADepartment of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Islamabad, Islamabad Capital Territory, 45650, PakistanDepartment of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Islamabad, Islamabad Capital Territory, 45650, PakistanDepartment of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Islamabad, Islamabad Capital Territory, 45650, PakistanBackground This paper discusses the development and benchmarking of IDEA (Indigenous Diffusion Equation Analyzer), an in-house transient, multi-group, 3-dimensional neutron diffusion equation (NDE) code. IDEA can perform both steady-state criticality calculations and reactivity transient analysis for 3-dimensional rectangular geometries, with potential extension to cylindrical geometries. Methods The IDEA code employs the finite difference method (FDM) for spatial and temporal discretization, enhanced by source extrapolation-based acceleration. Its performance was evaluated by comparing its results with those of established NDE-based codes, PRIDE and CITATION. Various 2D and 3D steady-state and transient reactivity benchmark problems were used for this comparison. Results The results obtained from IDEA were found to be in close agreement with reference values from publicly available literature. In addition, the IDEA code demonstrated superior accuracy and consistency compared to PRIDE and CITATION in both steady-state and transient reactivity analyses. A neutronic analysis of a typical small PWR core under clean and cold conditions also showed validation against reference results. Conclusions The IDEA code has shown promising figures of merit to be used for practical neutronic calculations. Moreover, a multitude of users can benefit from its further development due to ease of access and lack of proprietary restrictions.https://nstopenresearch.org/articles/3-4/v1neutron diffusion equation finite difference method PRIDE CITATION steady state and transient reactivity benchmarks.eng |
| spellingShingle | Ubaid ur Rehman Zahid Rasool M. Jawad Hussain Ali Mansoor Farhana Kausar Haseeb ur Rehman Aman ur Rehman Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved] Nuclear Science and Technology Open Research neutron diffusion equation finite difference method PRIDE CITATION steady state and transient reactivity benchmarks. eng |
| title | Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved] |
| title_full | Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved] |
| title_fullStr | Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved] |
| title_full_unstemmed | Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved] |
| title_short | Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved] |
| title_sort | neutronics analysis of pressurized water reactor cores using idea code version 1 peer review 2 approved |
| topic | neutron diffusion equation finite difference method PRIDE CITATION steady state and transient reactivity benchmarks. eng |
| url | https://nstopenresearch.org/articles/3-4/v1 |
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