Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor
•Thermal hydraulics coupling with the one-point neutron kinetics of a molten salt fast reactor.•User defined function of FLUENT is used to implement the neutronics.•One-point kinetic parameter of RELAP5-3D is modified to consider outflow and inflow of fuel salt.•ULOF event is calculated using both c...
Gespeichert in:
| Veröffentlicht in: | Nuclear engineering and design 2020-11, Vol.368 (C), p.110793, Article 110793 |
|---|---|
| 1. Verfasser: | |
| Format: | Artikel |
| Sprache: | eng |
| Schlagworte: | |
| Online-Zugang: | Volltext |
| Tags: |
Tag hinzufügen
Keine Tags, Fügen Sie den ersten Tag hinzu!
|
| container_end_page | |
|---|---|
| container_issue | C |
| container_start_page | 110793 |
| container_title | Nuclear engineering and design |
| container_volume | 368 |
| creator | Mochizuki, Hiroyasu |
| description | •Thermal hydraulics coupling with the one-point neutron kinetics of a molten salt fast reactor.•User defined function of FLUENT is used to implement the neutronics.•One-point kinetic parameter of RELAP5-3D is modified to consider outflow and inflow of fuel salt.•ULOF event is calculated using both codes and results show almost the same trends.•Loss of single pump trip event is calculated with RELAP5-3D.
The present study proposes a method for analyzing the thermal hydraulics coupling with the neutron point kinetics of a molten salt fast reactor. First, it has been analyzed by the FLUENT code that the flow behaviors in the reactor under the steady state conditions are greatly influenced by the installation condition of the inlet piping. Among them, the case where the inlet pipes are not offset is selected as a representative case. Next, the method of coupling the neutronics and thermal-hydraulics using UDF of the FLUENT code is explained when a transient analysis is conducted. In the case of using the system code RELAP5-3D, it is explained that thermophysical property binary files of the molten salts are generated, and the kinetic parameters for the implemented point kinetics model are modified in consideration of the inflow and outflow of delayed neutron precursors. A ULOF event has been simulated by the FLUENT code and the RELAP5-3D code, and it has been confirmed that the short-term evolutions in the comparison of reactor power and outlet temperature by the two codes show similar trends. It has been confirmed that when all the pumps are tripped, the core average temperature increases and the negative temperature reactivity coefficient feedback reduces the reactor power to the decay heat level. An event of single pump trip is also calculated with two codes, and it has been confirmed that the molten salt fast reactor has self-control characteristics on the basis of the negative temperature reactivity coefficient. |
| doi_str_mv | 10.1016/j.nucengdes.2020.110793 |
| format | Article |
| fullrecord | <record><control><sourceid>proquest_osti_</sourceid><recordid>TN_cdi_osti_scitechconnect_2279664</recordid><sourceformat>XML</sourceformat><sourcesystem>PC</sourcesystem><els_id>S0029549320302879</els_id><sourcerecordid>2462183753</sourcerecordid><originalsourceid>FETCH-LOGICAL-c529t-5b9b6a08423143d972b2d24edd4cd0fc4925bd7149036dead0b1055f0ef597a03</originalsourceid><addsrcrecordid>eNqFkU9vEzEQxVcIJELhM2DBeVP_XcfHqKSAFBWEWomb5bVnG0cbO9jeSjnx1fFqKVd8GfnN741m9JrmPcFrgkl3fVyHyUJ4dJDXFNOqEiwVe9GsyEbSVgr182WzwpiqVnDFXjdvcj7i-Sm6an7fwVRSDN5mZIJD5QDpZMb2cHHJTOMs2zidRx8ea9-Ml-wzmvL8rSi63T_s7u4r4uDZjn7s9tvvomWfFnmICRl0imOBgLIZCxpMLiiBsSWmt82rwYwZ3v2tV83D7e7-5ku7__b5681231pBVWlFr_rO4A2njHDmlKQ9dZSDc9w6PFiuqOidJFxh1jkwDvcECzFgGISSBrOr5sMyN-bidba-gD3YGALYoimVqut4hT4u0DnFXxPkoo9xSvXqrCnvKNkwKVil5ELZFHNOMOhz8ieTLppgPUeij_pfJHqORC-RVOd2cUK99MlDmheBYMH5NO_hov_vjD-QCZiB</addsrcrecordid><sourcetype>Open Access Repository</sourcetype><iscdi>true</iscdi><recordtype>article</recordtype><pqid>2462183753</pqid></control><display><type>article</type><title>Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor</title><source>Access via ScienceDirect (Elsevier)</source><creator>Mochizuki, Hiroyasu</creator><creatorcontrib>Mochizuki, Hiroyasu</creatorcontrib><description>•Thermal hydraulics coupling with the one-point neutron kinetics of a molten salt fast reactor.•User defined function of FLUENT is used to implement the neutronics.•One-point kinetic parameter of RELAP5-3D is modified to consider outflow and inflow of fuel salt.•ULOF event is calculated using both codes and results show almost the same trends.•Loss of single pump trip event is calculated with RELAP5-3D.
The present study proposes a method for analyzing the thermal hydraulics coupling with the neutron point kinetics of a molten salt fast reactor. First, it has been analyzed by the FLUENT code that the flow behaviors in the reactor under the steady state conditions are greatly influenced by the installation condition of the inlet piping. Among them, the case where the inlet pipes are not offset is selected as a representative case. Next, the method of coupling the neutronics and thermal-hydraulics using UDF of the FLUENT code is explained when a transient analysis is conducted. In the case of using the system code RELAP5-3D, it is explained that thermophysical property binary files of the molten salts are generated, and the kinetic parameters for the implemented point kinetics model are modified in consideration of the inflow and outflow of delayed neutron precursors. A ULOF event has been simulated by the FLUENT code and the RELAP5-3D code, and it has been confirmed that the short-term evolutions in the comparison of reactor power and outlet temperature by the two codes show similar trends. It has been confirmed that when all the pumps are tripped, the core average temperature increases and the negative temperature reactivity coefficient feedback reduces the reactor power to the decay heat level. An event of single pump trip is also calculated with two codes, and it has been confirmed that the molten salt fast reactor has self-control characteristics on the basis of the negative temperature reactivity coefficient.</description><identifier>ISSN: 0029-5493</identifier><identifier>EISSN: 1872-759X</identifier><identifier>DOI: 10.1016/j.nucengdes.2020.110793</identifier><language>eng</language><publisher>Amsterdam: Elsevier B.V</publisher><subject>CFD ; Computational fluid dynamics ; Coupling ; Coupling analysis ; Fluid flow ; Hydraulics ; Inlet pipes ; Kinetics ; Molten salts ; MSR ; Parameter modification ; Reactors ; Salts ; Single pump trip ; System code ; Thermophysical properties ; Transient analysis ; ULOF</subject><ispartof>Nuclear engineering and design, 2020-11, Vol.368 (C), p.110793, Article 110793</ispartof><rights>2020 Elsevier B.V.</rights><rights>Copyright Elsevier BV Nov 2020</rights><lds50>peer_reviewed</lds50><oa>free_for_read</oa><woscitedreferencessubscribed>false</woscitedreferencessubscribed><citedby>FETCH-LOGICAL-c529t-5b9b6a08423143d972b2d24edd4cd0fc4925bd7149036dead0b1055f0ef597a03</citedby><cites>FETCH-LOGICAL-c529t-5b9b6a08423143d972b2d24edd4cd0fc4925bd7149036dead0b1055f0ef597a03</cites><orcidid>0000-0002-6030-4781 ; 0000000260304781</orcidid></display><links><openurl>$$Topenurl_article</openurl><openurlfulltext>$$Topenurlfull_article</openurlfulltext><thumbnail>$$Tsyndetics_thumb_exl</thumbnail><linktohtml>$$Uhttps://dx.doi.org/10.1016/j.nucengdes.2020.110793$$EHTML$$P50$$Gelsevier$$H</linktohtml><link.rule.ids>230,314,780,784,885,3550,27924,27925,45995</link.rule.ids><backlink>$$Uhttps://www.osti.gov/biblio/2279664$$D View this record in Osti.gov$$Hfree_for_read</backlink></links><search><creatorcontrib>Mochizuki, Hiroyasu</creatorcontrib><title>Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor</title><title>Nuclear engineering and design</title><description>•Thermal hydraulics coupling with the one-point neutron kinetics of a molten salt fast reactor.•User defined function of FLUENT is used to implement the neutronics.•One-point kinetic parameter of RELAP5-3D is modified to consider outflow and inflow of fuel salt.•ULOF event is calculated using both codes and results show almost the same trends.•Loss of single pump trip event is calculated with RELAP5-3D.
The present study proposes a method for analyzing the thermal hydraulics coupling with the neutron point kinetics of a molten salt fast reactor. First, it has been analyzed by the FLUENT code that the flow behaviors in the reactor under the steady state conditions are greatly influenced by the installation condition of the inlet piping. Among them, the case where the inlet pipes are not offset is selected as a representative case. Next, the method of coupling the neutronics and thermal-hydraulics using UDF of the FLUENT code is explained when a transient analysis is conducted. In the case of using the system code RELAP5-3D, it is explained that thermophysical property binary files of the molten salts are generated, and the kinetic parameters for the implemented point kinetics model are modified in consideration of the inflow and outflow of delayed neutron precursors. A ULOF event has been simulated by the FLUENT code and the RELAP5-3D code, and it has been confirmed that the short-term evolutions in the comparison of reactor power and outlet temperature by the two codes show similar trends. It has been confirmed that when all the pumps are tripped, the core average temperature increases and the negative temperature reactivity coefficient feedback reduces the reactor power to the decay heat level. An event of single pump trip is also calculated with two codes, and it has been confirmed that the molten salt fast reactor has self-control characteristics on the basis of the negative temperature reactivity coefficient.</description><subject>CFD</subject><subject>Computational fluid dynamics</subject><subject>Coupling</subject><subject>Coupling analysis</subject><subject>Fluid flow</subject><subject>Hydraulics</subject><subject>Inlet pipes</subject><subject>Kinetics</subject><subject>Molten salts</subject><subject>MSR</subject><subject>Parameter modification</subject><subject>Reactors</subject><subject>Salts</subject><subject>Single pump trip</subject><subject>System code</subject><subject>Thermophysical properties</subject><subject>Transient analysis</subject><subject>ULOF</subject><issn>0029-5493</issn><issn>1872-759X</issn><fulltext>true</fulltext><rsrctype>article</rsrctype><creationdate>2020</creationdate><recordtype>article</recordtype><recordid>eNqFkU9vEzEQxVcIJELhM2DBeVP_XcfHqKSAFBWEWomb5bVnG0cbO9jeSjnx1fFqKVd8GfnN741m9JrmPcFrgkl3fVyHyUJ4dJDXFNOqEiwVe9GsyEbSVgr182WzwpiqVnDFXjdvcj7i-Sm6an7fwVRSDN5mZIJD5QDpZMb2cHHJTOMs2zidRx8ea9-Ml-wzmvL8rSi63T_s7u4r4uDZjn7s9tvvomWfFnmICRl0imOBgLIZCxpMLiiBsSWmt82rwYwZ3v2tV83D7e7-5ku7__b5681231pBVWlFr_rO4A2njHDmlKQ9dZSDc9w6PFiuqOidJFxh1jkwDvcECzFgGISSBrOr5sMyN-bidba-gD3YGALYoimVqut4hT4u0DnFXxPkoo9xSvXqrCnvKNkwKVil5ELZFHNOMOhz8ieTLppgPUeij_pfJHqORC-RVOd2cUK99MlDmheBYMH5NO_hov_vjD-QCZiB</recordid><startdate>20201101</startdate><enddate>20201101</enddate><creator>Mochizuki, Hiroyasu</creator><general>Elsevier B.V</general><general>Elsevier BV</general><general>Elsevier</general><scope>AAYXX</scope><scope>CITATION</scope><scope>7SP</scope><scope>7ST</scope><scope>7TB</scope><scope>8FD</scope><scope>C1K</scope><scope>FR3</scope><scope>KR7</scope><scope>L7M</scope><scope>SOI</scope><scope>OTOTI</scope><orcidid>https://orcid.org/0000-0002-6030-4781</orcidid><orcidid>https://orcid.org/0000000260304781</orcidid></search><sort><creationdate>20201101</creationdate><title>Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor</title><author>Mochizuki, Hiroyasu</author></sort><facets><frbrtype>5</frbrtype><frbrgroupid>cdi_FETCH-LOGICAL-c529t-5b9b6a08423143d972b2d24edd4cd0fc4925bd7149036dead0b1055f0ef597a03</frbrgroupid><rsrctype>articles</rsrctype><prefilter>articles</prefilter><language>eng</language><creationdate>2020</creationdate><topic>CFD</topic><topic>Computational fluid dynamics</topic><topic>Coupling</topic><topic>Coupling analysis</topic><topic>Fluid flow</topic><topic>Hydraulics</topic><topic>Inlet pipes</topic><topic>Kinetics</topic><topic>Molten salts</topic><topic>MSR</topic><topic>Parameter modification</topic><topic>Reactors</topic><topic>Salts</topic><topic>Single pump trip</topic><topic>System code</topic><topic>Thermophysical properties</topic><topic>Transient analysis</topic><topic>ULOF</topic><toplevel>peer_reviewed</toplevel><toplevel>online_resources</toplevel><creatorcontrib>Mochizuki, Hiroyasu</creatorcontrib><collection>CrossRef</collection><collection>Electronics & Communications Abstracts</collection><collection>Environment Abstracts</collection><collection>Mechanical & Transportation Engineering Abstracts</collection><collection>Technology Research Database</collection><collection>Environmental Sciences and Pollution Management</collection><collection>Engineering Research Database</collection><collection>Civil Engineering Abstracts</collection><collection>Advanced Technologies Database with Aerospace</collection><collection>Environment Abstracts</collection><collection>OSTI.GOV</collection><jtitle>Nuclear engineering and design</jtitle></facets><delivery><delcategory>Remote Search Resource</delcategory><fulltext>fulltext</fulltext></delivery><addata><au>Mochizuki, Hiroyasu</au><format>journal</format><genre>article</genre><ristype>JOUR</ristype><atitle>Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor</atitle><jtitle>Nuclear engineering and design</jtitle><date>2020-11-01</date><risdate>2020</risdate><volume>368</volume><issue>C</issue><spage>110793</spage><pages>110793-</pages><artnum>110793</artnum><issn>0029-5493</issn><eissn>1872-759X</eissn><abstract>•Thermal hydraulics coupling with the one-point neutron kinetics of a molten salt fast reactor.•User defined function of FLUENT is used to implement the neutronics.•One-point kinetic parameter of RELAP5-3D is modified to consider outflow and inflow of fuel salt.•ULOF event is calculated using both codes and results show almost the same trends.•Loss of single pump trip event is calculated with RELAP5-3D.
The present study proposes a method for analyzing the thermal hydraulics coupling with the neutron point kinetics of a molten salt fast reactor. First, it has been analyzed by the FLUENT code that the flow behaviors in the reactor under the steady state conditions are greatly influenced by the installation condition of the inlet piping. Among them, the case where the inlet pipes are not offset is selected as a representative case. Next, the method of coupling the neutronics and thermal-hydraulics using UDF of the FLUENT code is explained when a transient analysis is conducted. In the case of using the system code RELAP5-3D, it is explained that thermophysical property binary files of the molten salts are generated, and the kinetic parameters for the implemented point kinetics model are modified in consideration of the inflow and outflow of delayed neutron precursors. A ULOF event has been simulated by the FLUENT code and the RELAP5-3D code, and it has been confirmed that the short-term evolutions in the comparison of reactor power and outlet temperature by the two codes show similar trends. It has been confirmed that when all the pumps are tripped, the core average temperature increases and the negative temperature reactivity coefficient feedback reduces the reactor power to the decay heat level. An event of single pump trip is also calculated with two codes, and it has been confirmed that the molten salt fast reactor has self-control characteristics on the basis of the negative temperature reactivity coefficient.</abstract><cop>Amsterdam</cop><pub>Elsevier B.V</pub><doi>10.1016/j.nucengdes.2020.110793</doi><orcidid>https://orcid.org/0000-0002-6030-4781</orcidid><orcidid>https://orcid.org/0000000260304781</orcidid><oa>free_for_read</oa></addata></record> |
| fulltext | fulltext |
| identifier | ISSN: 0029-5493 |
| ispartof | Nuclear engineering and design, 2020-11, Vol.368 (C), p.110793, Article 110793 |
| issn | 0029-5493 1872-759X |
| language | eng |
| recordid | cdi_osti_scitechconnect_2279664 |
| source | Access via ScienceDirect (Elsevier) |
| subjects | CFD Computational fluid dynamics Coupling Coupling analysis Fluid flow Hydraulics Inlet pipes Kinetics Molten salts MSR Parameter modification Reactors Salts Single pump trip System code Thermophysical properties Transient analysis ULOF |
| title | Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor |
| url | https://sfx.bib-bvb.de/sfx_tum?ctx_ver=Z39.88-2004&ctx_enc=info:ofi/enc:UTF-8&ctx_tim=2024-12-29T05%3A21%3A02IST&url_ver=Z39.88-2004&url_ctx_fmt=infofi/fmt:kev:mtx:ctx&rfr_id=info:sid/primo.exlibrisgroup.com:primo3-Article-proquest_osti_&rft_val_fmt=info:ofi/fmt:kev:mtx:journal&rft.genre=article&rft.atitle=Neutronics%20and%20thermal-hydraulics%20coupling%20analysis%20using%20the%20FLUENT%20code%20and%20the%20RELAP5-3D%20code%20for%20a%20molten%20salt%20fast%20reactor&rft.jtitle=Nuclear%20engineering%20and%20design&rft.au=Mochizuki,%20Hiroyasu&rft.date=2020-11-01&rft.volume=368&rft.issue=C&rft.spage=110793&rft.pages=110793-&rft.artnum=110793&rft.issn=0029-5493&rft.eissn=1872-759X&rft_id=info:doi/10.1016/j.nucengdes.2020.110793&rft_dat=%3Cproquest_osti_%3E2462183753%3C/proquest_osti_%3E%3Curl%3E%3C/url%3E&disable_directlink=true&sfx.directlink=off&sfx.report_link=0&rft_id=info:oai/&rft_pqid=2462183753&rft_id=info:pmid/&rft_els_id=S0029549320302879&rfr_iscdi=true |