Fuel burnup analysis for Thai research reactor by using MCNPX computer code

Fuel burnup analysis for Thai research reactor by using MCNPX computer code

This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiati...

Ausführliche Beschreibung

Gespeichert in:
Bibliographische Detailangaben
Veröffentlicht in:Journal of physics. Conference series 2017-06, Vol.860 (1), p.12033
Hauptverfasser: Sangkaew, S, Angwongtrakool, T, Srimok, B
Format: Artikel
Sprache:eng
Schlagworte:
Electric power distribution
Fission products
FORTRAN
Fuel cycles
Fuels
Neutron counters
Neutron flux
Neutrons
Nuclear fuels
Physics
Radiation transport
Tubes
Online-Zugang:Volltext
Tags: Tag hinzufügen
Keine Tags, Fügen Sie den ersten Tag hinzu!
container_end_page
container_issue 1
container_start_page 12033
container_title Journal of physics. Conference series
container_volume 860
creator Sangkaew, S
Angwongtrakool, T
Srimok, B
description This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.
doi_str_mv 10.1088/1742-6596/860/1/012033
format Article
fullrecord <record><control><sourceid>proquest_cross</sourceid><recordid>TN_cdi_proquest_journals_2574660506</recordid><sourceformat>XML</sourceformat><sourcesystem>PC</sourcesystem><sourcerecordid>2574660506</sourcerecordid><originalsourceid>FETCH-LOGICAL-c461t-40f746a4b1a409780c9bfd369f8722a754c53235a49af523a76d0a0af050f0803</originalsourceid><addsrcrecordid>eNqFkN1LwzAUxYMoOKf_ggR88qH25qNp-ijF-TV14ATfQtomrmNba9I-7L83ozIRBJ_u4eacc8kPoXMCVwSkjEnKaSSSTMRSQExiIBQYO0Cj_cPhXkt5jE68X0JwMJaO0OOkNytc9G7Tt1hv9Grra49t4_B8oWvsjDfalYsgdNmFbbHFva83H_gpf56947JZt31nXBCVOUVHVq-8OfueY_Q2uZnnd9H05fY-v55GJRekizjYlAvNC6I5ZKmEMitsxURmZUqpThNeJoyyRPNM24QynYoKNGgLCViQwMboYuhtXfPZG9-pZRM-EE4qmoRqEYwiuMTgKl3jvTNWta5ea7dVBNQOnNoxUTs-KoBTRA3gQpAOwbppf5r_DV3-EXqY5a-_fKqtLPsCZqh7EQ</addsrcrecordid><sourcetype>Aggregation Database</sourcetype><iscdi>true</iscdi><recordtype>article</recordtype><pqid>2574660506</pqid></control><display><type>article</type><title>Fuel burnup analysis for Thai research reactor by using MCNPX computer code</title><source>Institute of Physics IOPscience extra</source><source>IOP_英国物理学会OA刊</source><source>Alma/SFX Local Collection</source><source>Free Full-Text Journals in Chemistry</source><source>EZB Electronic Journals Library</source><creator>Sangkaew, S ; Angwongtrakool, T ; Srimok, B</creator><creatorcontrib>Sangkaew, S ; Angwongtrakool, T ; Srimok, B</creatorcontrib><description>This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.</description><identifier>ISSN: 1742-6588</identifier><identifier>EISSN: 1742-6596</identifier><identifier>DOI: 10.1088/1742-6596/860/1/012033</identifier><language>eng</language><publisher>Bristol: IOP Publishing</publisher><subject>Electric power distribution ; Fission products ; FORTRAN ; Fuel cycles ; Fuels ; Neutron counters ; Neutron flux ; Neutrons ; Nuclear fuels ; Physics ; Radiation transport ; Tubes</subject><ispartof>Journal of physics. Conference series, 2017-06, Vol.860 (1), p.12033</ispartof><rights>Published under licence by IOP Publishing Ltd</rights><rights>2017. This work is published under http://creativecommons.org/licenses/by/3.0/ (the “License”). Notwithstanding the ProQuest Terms and Conditions, you may use this content in accordance with the terms of the License.</rights><lds50>peer_reviewed</lds50><oa>free_for_read</oa><woscitedreferencessubscribed>false</woscitedreferencessubscribed><citedby>FETCH-LOGICAL-c461t-40f746a4b1a409780c9bfd369f8722a754c53235a49af523a76d0a0af050f0803</citedby><cites>FETCH-LOGICAL-c461t-40f746a4b1a409780c9bfd369f8722a754c53235a49af523a76d0a0af050f0803</cites></display><links><openurl>$$Topenurl_article</openurl><openurlfulltext>$$Topenurlfull_article</openurlfulltext><thumbnail>$$Tsyndetics_thumb_exl</thumbnail><linktopdf>$$Uhttps://iopscience.iop.org/article/10.1088/1742-6596/860/1/012033/pdf$$EPDF$$P50$$Giop$$Hfree_for_read</linktopdf><link.rule.ids>314,780,784,27924,27925,38868,38890,53840,53867</link.rule.ids></links><search><creatorcontrib>Sangkaew, S</creatorcontrib><creatorcontrib>Angwongtrakool, T</creatorcontrib><creatorcontrib>Srimok, B</creatorcontrib><title>Fuel burnup analysis for Thai research reactor by using MCNPX computer code</title><title>Journal of physics. Conference series</title><addtitle>J. Phys.: Conf. Ser</addtitle><description>This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.</description><subject>Electric power distribution</subject><subject>Fission products</subject><subject>FORTRAN</subject><subject>Fuel cycles</subject><subject>Fuels</subject><subject>Neutron counters</subject><subject>Neutron flux</subject><subject>Neutrons</subject><subject>Nuclear fuels</subject><subject>Physics</subject><subject>Radiation transport</subject><subject>Tubes</subject><issn>1742-6588</issn><issn>1742-6596</issn><fulltext>true</fulltext><rsrctype>article</rsrctype><creationdate>2017</creationdate><recordtype>article</recordtype><sourceid>O3W</sourceid><sourceid>ABUWG</sourceid><sourceid>AFKRA</sourceid><sourceid>AZQEC</sourceid><sourceid>BENPR</sourceid><sourceid>CCPQU</sourceid><sourceid>DWQXO</sourceid><recordid>eNqFkN1LwzAUxYMoOKf_ggR88qH25qNp-ijF-TV14ATfQtomrmNba9I-7L83ozIRBJ_u4eacc8kPoXMCVwSkjEnKaSSSTMRSQExiIBQYO0Cj_cPhXkt5jE68X0JwMJaO0OOkNytc9G7Tt1hv9Grra49t4_B8oWvsjDfalYsgdNmFbbHFva83H_gpf56947JZt31nXBCVOUVHVq-8OfueY_Q2uZnnd9H05fY-v55GJRekizjYlAvNC6I5ZKmEMitsxURmZUqpThNeJoyyRPNM24QynYoKNGgLCViQwMboYuhtXfPZG9-pZRM-EE4qmoRqEYwiuMTgKl3jvTNWta5ea7dVBNQOnNoxUTs-KoBTRA3gQpAOwbppf5r_DV3-EXqY5a-_fKqtLPsCZqh7EQ</recordid><startdate>20170601</startdate><enddate>20170601</enddate><creator>Sangkaew, S</creator><creator>Angwongtrakool, T</creator><creator>Srimok, B</creator><general>IOP Publishing</general><scope>O3W</scope><scope>TSCCA</scope><scope>AAYXX</scope><scope>CITATION</scope><scope>8FD</scope><scope>8FE</scope><scope>8FG</scope><scope>ABUWG</scope><scope>AFKRA</scope><scope>ARAPS</scope><scope>AZQEC</scope><scope>BENPR</scope><scope>BGLVJ</scope><scope>CCPQU</scope><scope>DWQXO</scope><scope>H8D</scope><scope>HCIFZ</scope><scope>L7M</scope><scope>P5Z</scope><scope>P62</scope><scope>PIMPY</scope><scope>PQEST</scope><scope>PQQKQ</scope><scope>PQUKI</scope><scope>PRINS</scope></search><sort><creationdate>20170601</creationdate><title>Fuel burnup analysis for Thai research reactor by using MCNPX computer code</title><author>Sangkaew, S ; Angwongtrakool, T ; Srimok, B</author></sort><facets><frbrtype>5</frbrtype><frbrgroupid>cdi_FETCH-LOGICAL-c461t-40f746a4b1a409780c9bfd369f8722a754c53235a49af523a76d0a0af050f0803</frbrgroupid><rsrctype>articles</rsrctype><prefilter>articles</prefilter><language>eng</language><creationdate>2017</creationdate><topic>Electric power distribution</topic><topic>Fission products</topic><topic>FORTRAN</topic><topic>Fuel cycles</topic><topic>Fuels</topic><topic>Neutron counters</topic><topic>Neutron flux</topic><topic>Neutrons</topic><topic>Nuclear fuels</topic><topic>Physics</topic><topic>Radiation transport</topic><topic>Tubes</topic><toplevel>peer_reviewed</toplevel><toplevel>online_resources</toplevel><creatorcontrib>Sangkaew, S</creatorcontrib><creatorcontrib>Angwongtrakool, T</creatorcontrib><creatorcontrib>Srimok, B</creatorcontrib><collection>IOP_英国物理学会OA刊</collection><collection>IOPscience (Open Access)</collection><collection>CrossRef</collection><collection>Technology Research Database</collection><collection>ProQuest SciTech Collection</collection><collection>ProQuest Technology Collection</collection><collection>ProQuest Central (Alumni)</collection><collection>ProQuest Central UK/Ireland</collection><collection>Advanced Technologies &amp; Aerospace Collection</collection><collection>ProQuest Central Essentials</collection><collection>ProQuest Central</collection><collection>Technology Collection</collection><collection>ProQuest One Community College</collection><collection>ProQuest Central Korea</collection><collection>Aerospace Database</collection><collection>SciTech Premium Collection</collection><collection>Advanced Technologies Database with Aerospace</collection><collection>ProQuest advanced technologies &amp; aerospace journals</collection><collection>ProQuest Advanced Technologies &amp; Aerospace Collection</collection><collection>Publicly Available Content Database</collection><collection>ProQuest One Academic Eastern Edition (DO NOT USE)</collection><collection>ProQuest One Academic</collection><collection>ProQuest One Academic UKI Edition</collection><collection>ProQuest Central China</collection><jtitle>Journal of physics. Conference series</jtitle></facets><delivery><delcategory>Remote Search Resource</delcategory><fulltext>fulltext</fulltext></delivery><addata><au>Sangkaew, S</au><au>Angwongtrakool, T</au><au>Srimok, B</au><format>journal</format><genre>article</genre><ristype>JOUR</ristype><atitle>Fuel burnup analysis for Thai research reactor by using MCNPX computer code</atitle><jtitle>Journal of physics. Conference series</jtitle><addtitle>J. Phys.: Conf. Ser</addtitle><date>2017-06-01</date><risdate>2017</risdate><volume>860</volume><issue>1</issue><spage>12033</spage><pages>12033-</pages><issn>1742-6588</issn><eissn>1742-6596</eissn><abstract>This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.</abstract><cop>Bristol</cop><pub>IOP Publishing</pub><doi>10.1088/1742-6596/860/1/012033</doi><tpages>8</tpages><oa>free_for_read</oa></addata></record>
fulltext fulltext
identifier ISSN: 1742-6588
ispartof Journal of physics. Conference series, 2017-06, Vol.860 (1), p.12033
issn 1742-6588
1742-6596
language eng
recordid cdi_proquest_journals_2574660506
source Institute of Physics IOPscience extra; IOP_英国物理学会OA刊; Alma/SFX Local Collection; Free Full-Text Journals in Chemistry; EZB Electronic Journals Library
subjects Electric power distribution
Fission products
FORTRAN
Fuel cycles
Fuels
Neutron counters
Neutron flux
Neutrons
Nuclear fuels
Physics
Radiation transport
Tubes
title Fuel burnup analysis for Thai research reactor by using MCNPX computer code
url https://sfx.bib-bvb.de/sfx_tum?ctx_ver=Z39.88-2004&ctx_enc=info:ofi/enc:UTF-8&ctx_tim=2025-01-01T10%3A02%3A01IST&url_ver=Z39.88-2004&url_ctx_fmt=infofi/fmt:kev:mtx:ctx&rfr_id=info:sid/primo.exlibrisgroup.com:primo3-Article-proquest_cross&rft_val_fmt=info:ofi/fmt:kev:mtx:journal&rft.genre=article&rft.atitle=Fuel%20burnup%20analysis%20for%20Thai%20research%20reactor%20by%20using%20MCNPX%20computer%20code&rft.jtitle=Journal%20of%20physics.%20Conference%20series&rft.au=Sangkaew,%20S&rft.date=2017-06-01&rft.volume=860&rft.issue=1&rft.spage=12033&rft.pages=12033-&rft.issn=1742-6588&rft.eissn=1742-6596&rft_id=info:doi/10.1088/1742-6596/860/1/012033&rft_dat=%3Cproquest_cross%3E2574660506%3C/proquest_cross%3E%3Curl%3E%3C/url%3E&disable_directlink=true&sfx.directlink=off&sfx.report_link=0&rft_id=info:oai/&rft_pqid=2574660506&rft_id=info:pmid/&rfr_iscdi=true