Monte Carlo with fuel burnup method for the ENHS benchmark calculations
Estimates of the uncertainties arising from approximations in the methods used in different nuclear data processing and neutron transport codes are usually obtained by inter-comparing calculations made using different code systems. This paper gives details of an investigation of differences between...
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| Format: | Article |
| Language: | English |
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VINCA Institute of Nuclear Sciences
2003-01-01
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| Series: | Nuclear Technology and Radiation Protection |
| Subjects: | |
| Online Access: | http://www.doiserbia.nb.rs/img/doi/1451-3994/2003/1451-39940302003M.pdf |
| _version_ | 1819140166210551808 |
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| author | Milošević Miodrag 1 Greenspan Ehud Vujić Jasmina Lj. |
| author_facet | Milošević Miodrag 1 Greenspan Ehud Vujić Jasmina Lj. |
| author_sort | Milošević Miodrag 1 |
| collection | DOAJ |
| description | Estimates of the uncertainties arising from approximations in the methods used in different nuclear data processing and neutron transport codes are usually obtained by inter-comparing calculations made using different code systems. This paper gives details of an investigation of differences between results obtained by using different codes for a single zone model of the Encapsulated Nuclear Heat Source (ENHS) benchmark core fuelled with metallic alloy of Pu, U, and Zr. The ENHS is a new lead-bismuth or lead cooled novel reactor concept for 20 effective full power years without refuelling and with very small reactivity swing. The computational tools benchmarked include MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on ENDF/B-VI evaluation; KENO-V.a/ORIGEN2.1 code system, recently developed by authors of this paper, with the ENDFB-V based 238 group library; the design-oriented procedure based on the simplified one-dimensional (1D) geometry model and SAS2H control module; and the well-established fast reactor neutronics design tools in use at Argonne National Laboratory. Calculations made for the ENHS benchmark have shown that the differences between the results obtained when using different code schemes are quite significant and should be taken into account in assessing the quality of the nuclear data library. |
| first_indexed | 2024-12-22T11:34:14Z |
| format | Article |
| id | doaj.art-c8bab03b9ecd4f80b7a35776d4713138 |
| institution | Directory Open Access Journal |
| issn | 1451-3994 |
| language | English |
| last_indexed | 2024-12-22T11:34:14Z |
| publishDate | 2003-01-01 |
| publisher | VINCA Institute of Nuclear Sciences |
| record_format | Article |
| series | Nuclear Technology and Radiation Protection |
| spelling | doaj.art-c8bab03b9ecd4f80b7a35776d47131382022-12-21T18:27:29ZengVINCA Institute of Nuclear SciencesNuclear Technology and Radiation Protection1451-39942003-01-0118231110.2298/NTRP0302003MMonte Carlo with fuel burnup method for the ENHS benchmark calculationsMilošević Miodrag 1Greenspan EhudVujić Jasmina Lj.Estimates of the uncertainties arising from approximations in the methods used in different nuclear data processing and neutron transport codes are usually obtained by inter-comparing calculations made using different code systems. This paper gives details of an investigation of differences between results obtained by using different codes for a single zone model of the Encapsulated Nuclear Heat Source (ENHS) benchmark core fuelled with metallic alloy of Pu, U, and Zr. The ENHS is a new lead-bismuth or lead cooled novel reactor concept for 20 effective full power years without refuelling and with very small reactivity swing. The computational tools benchmarked include MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on ENDF/B-VI evaluation; KENO-V.a/ORIGEN2.1 code system, recently developed by authors of this paper, with the ENDFB-V based 238 group library; the design-oriented procedure based on the simplified one-dimensional (1D) geometry model and SAS2H control module; and the well-established fast reactor neutronics design tools in use at Argonne National Laboratory. Calculations made for the ENHS benchmark have shown that the differences between the results obtained when using different code schemes are quite significant and should be taken into account in assessing the quality of the nuclear data library.http://www.doiserbia.nb.rs/img/doi/1451-3994/2003/1451-39940302003M.pdfMonte Carlofuel burnupORIGEN2.1ENHS benchmark |
| spellingShingle | Milošević Miodrag 1 Greenspan Ehud Vujić Jasmina Lj. Monte Carlo with fuel burnup method for the ENHS benchmark calculations Nuclear Technology and Radiation Protection Monte Carlo fuel burnup ORIGEN2.1 ENHS benchmark |
| title | Monte Carlo with fuel burnup method for the ENHS benchmark calculations |
| title_full | Monte Carlo with fuel burnup method for the ENHS benchmark calculations |
| title_fullStr | Monte Carlo with fuel burnup method for the ENHS benchmark calculations |
| title_full_unstemmed | Monte Carlo with fuel burnup method for the ENHS benchmark calculations |
| title_short | Monte Carlo with fuel burnup method for the ENHS benchmark calculations |
| title_sort | monte carlo with fuel burnup method for the enhs benchmark calculations |
| topic | Monte Carlo fuel burnup ORIGEN2.1 ENHS benchmark |
| url | http://www.doiserbia.nb.rs/img/doi/1451-3994/2003/1451-39940302003M.pdf |
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