Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket
Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-life nuclear waste, and having inherent safety. Considering the fast development of nuclear power industry and the abundant thorium resources in China, the concept of a thorium-based breeding blanket fo...
Gespeichert in:
| Veröffentlicht in: | Fusion engineering and design 2010-12, Vol.85 (10), p.2227-2231 |
|---|---|
| Hauptverfasser: | , , , , , , |
| Format: | Artikel |
| Sprache: | eng |
| Schlagworte: | |
| Online-Zugang: | Volltext |
| Tags: |
Tag hinzufügen
Keine Tags, Fügen Sie den ersten Tag hinzu!
|
| container_end_page | 2231 |
|---|---|
| container_issue | 10 |
| container_start_page | 2227 |
| container_title | Fusion engineering and design |
| container_volume | 85 |
| creator | Ma, X.B. Chen, Y.X. Wang, Y. Zhang, P.Z. Cao, B. Lu, D.G. Cheng, H.P. |
| description | Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-life nuclear waste, and having inherent safety. Considering the fast development of nuclear power industry and the abundant thorium resources in China, the concept of a thorium-based breeding blanket for fuel production is proposed. The blanket using ThN or ThO
2 dispersed in graphite or BeO is investigated under a first neutron wall loading of 0.57
MW/cm
2 as ITER. The design helium cooled pebble bed design is employed according to the experience in the fusion reactor field. Preliminary neutronic calculations are performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. The behavior of the neutronic potential is observed for 960 days. The cumulative fissile fuel enrichment values varied between 6.88% and 8.56% depending on the fuel types. The tritium breeding ratio is greater than 1.05 for all investigated fuel types and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those modes during the operation period. The blanket energy multiplication factor
M, varies between 13.66 and 15.85 depending on the fuel types at the end of the operation period. In addition, the effect of
233Pa on the
233U production and
k
eff
are also discussed. |
| doi_str_mv | 10.1016/j.fusengdes.2010.08.044 |
| format | Article |
| fullrecord | <record><control><sourceid>proquest_cross</sourceid><recordid>TN_cdi_proquest_miscellaneous_869796894</recordid><sourceformat>XML</sourceformat><sourcesystem>PC</sourcesystem><els_id>S0920379610004035</els_id><sourcerecordid>869796894</sourcerecordid><originalsourceid>FETCH-LOGICAL-c377t-84ce078a9c3d8c31f8caf84e79ae51a1f4cc1eddd0d63f76b77983437769d1e23</originalsourceid><addsrcrecordid>eNqFkM9KAzEQxoMoWKvP4F7E09Zks26SYxH_QVEEPYd0MrGp201NdoXefAff0CcxpeJVGMiQfF--mR8hp4xOGGXNxXLihoTdq8U0qWi-pXJC63qPjJgUvBRMNftkRFVFSy5Uc0iOUlpSykSuEXl6wKGPofNQgGlhaE3vQ5eK4ApT9IsQ_bAq5yahLXJMfvr-_HI-bbtisZlHb4uIBvoQi3lrujfsj8mBM23Ck99zTF5urp-v7srZ4-391XRWAheiL2UNSIU0CriVwJmTYJysUSiDl8wwVwMwtNZS23AnmrkQSvI6extlGVZ8TM53_65jeB8w9XrlE2Cbp8AwJC0bldeVqs5KsVNCDClFdHod_crEjWZUbxnqpf5jqLcMNZU6M8zOs98MkzIeF00HPv3ZKy7yRBXLuulOh3nhD49RJ_DYAVofEXptg_836wfCto6q</addsrcrecordid><sourcetype>Aggregation Database</sourcetype><iscdi>true</iscdi><recordtype>article</recordtype><pqid>869796894</pqid></control><display><type>article</type><title>Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket</title><source>Elsevier ScienceDirect Journals</source><creator>Ma, X.B. ; Chen, Y.X. ; Wang, Y. ; Zhang, P.Z. ; Cao, B. ; Lu, D.G. ; Cheng, H.P.</creator><creatorcontrib>Ma, X.B. ; Chen, Y.X. ; Wang, Y. ; Zhang, P.Z. ; Cao, B. ; Lu, D.G. ; Cheng, H.P.</creatorcontrib><description>Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-life nuclear waste, and having inherent safety. Considering the fast development of nuclear power industry and the abundant thorium resources in China, the concept of a thorium-based breeding blanket for fuel production is proposed. The blanket using ThN or ThO
2 dispersed in graphite or BeO is investigated under a first neutron wall loading of 0.57
MW/cm
2 as ITER. The design helium cooled pebble bed design is employed according to the experience in the fusion reactor field. Preliminary neutronic calculations are performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. The behavior of the neutronic potential is observed for 960 days. The cumulative fissile fuel enrichment values varied between 6.88% and 8.56% depending on the fuel types. The tritium breeding ratio is greater than 1.05 for all investigated fuel types and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those modes during the operation period. The blanket energy multiplication factor
M, varies between 13.66 and 15.85 depending on the fuel types at the end of the operation period. In addition, the effect of
233Pa on the
233U production and
k
eff
are also discussed.</description><identifier>ISSN: 0920-3796</identifier><identifier>EISSN: 1873-7196</identifier><identifier>DOI: 10.1016/j.fusengdes.2010.08.044</identifier><identifier>CODEN: FEDEEE</identifier><language>eng</language><publisher>Amsterdam: Elsevier B.V</publisher><subject>233Pa effect ; Applied sciences ; Blanket ; Blanketing ; Controled nuclear fusion plants ; Energy ; Energy. Thermal use of fuels ; Exact sciences and technology ; Fission nuclear power plants ; Fuels ; Fusion-fission hybrid reactors ; Hybrid reactor ; Installations for energy generation and conversion: thermal and electrical energy ; Mathematical analysis ; Monte Carlo methods ; Nuclear engineering ; Nuclear fuels ; Nuclear reactor components ; Preparation and processing of nuclear fuels ; Thorium ; Transport ; Tritium</subject><ispartof>Fusion engineering and design, 2010-12, Vol.85 (10), p.2227-2231</ispartof><rights>2010</rights><rights>2015 INIST-CNRS</rights><lds50>peer_reviewed</lds50><woscitedreferencessubscribed>false</woscitedreferencessubscribed><citedby>FETCH-LOGICAL-c377t-84ce078a9c3d8c31f8caf84e79ae51a1f4cc1eddd0d63f76b77983437769d1e23</citedby><cites>FETCH-LOGICAL-c377t-84ce078a9c3d8c31f8caf84e79ae51a1f4cc1eddd0d63f76b77983437769d1e23</cites></display><links><openurl>$$Topenurl_article</openurl><openurlfulltext>$$Topenurlfull_article</openurlfulltext><thumbnail>$$Tsyndetics_thumb_exl</thumbnail><linktohtml>$$Uhttps://www.sciencedirect.com/science/article/pii/S0920379610004035$$EHTML$$P50$$Gelsevier$$H</linktohtml><link.rule.ids>309,310,314,776,780,785,786,3537,23911,23912,25120,27903,27904,65309</link.rule.ids><backlink>$$Uhttp://pascal-francis.inist.fr/vibad/index.php?action=getRecordDetail&idt=23734321$$DView record in Pascal Francis$$Hfree_for_read</backlink></links><search><creatorcontrib>Ma, X.B.</creatorcontrib><creatorcontrib>Chen, Y.X.</creatorcontrib><creatorcontrib>Wang, Y.</creatorcontrib><creatorcontrib>Zhang, P.Z.</creatorcontrib><creatorcontrib>Cao, B.</creatorcontrib><creatorcontrib>Lu, D.G.</creatorcontrib><creatorcontrib>Cheng, H.P.</creatorcontrib><title>Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket</title><title>Fusion engineering and design</title><description>Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-life nuclear waste, and having inherent safety. Considering the fast development of nuclear power industry and the abundant thorium resources in China, the concept of a thorium-based breeding blanket for fuel production is proposed. The blanket using ThN or ThO
2 dispersed in graphite or BeO is investigated under a first neutron wall loading of 0.57
MW/cm
2 as ITER. The design helium cooled pebble bed design is employed according to the experience in the fusion reactor field. Preliminary neutronic calculations are performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. The behavior of the neutronic potential is observed for 960 days. The cumulative fissile fuel enrichment values varied between 6.88% and 8.56% depending on the fuel types. The tritium breeding ratio is greater than 1.05 for all investigated fuel types and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those modes during the operation period. The blanket energy multiplication factor
M, varies between 13.66 and 15.85 depending on the fuel types at the end of the operation period. In addition, the effect of
233Pa on the
233U production and
k
eff
are also discussed.</description><subject>233Pa effect</subject><subject>Applied sciences</subject><subject>Blanket</subject><subject>Blanketing</subject><subject>Controled nuclear fusion plants</subject><subject>Energy</subject><subject>Energy. Thermal use of fuels</subject><subject>Exact sciences and technology</subject><subject>Fission nuclear power plants</subject><subject>Fuels</subject><subject>Fusion-fission hybrid reactors</subject><subject>Hybrid reactor</subject><subject>Installations for energy generation and conversion: thermal and electrical energy</subject><subject>Mathematical analysis</subject><subject>Monte Carlo methods</subject><subject>Nuclear engineering</subject><subject>Nuclear fuels</subject><subject>Nuclear reactor components</subject><subject>Preparation and processing of nuclear fuels</subject><subject>Thorium</subject><subject>Transport</subject><subject>Tritium</subject><issn>0920-3796</issn><issn>1873-7196</issn><fulltext>true</fulltext><rsrctype>article</rsrctype><creationdate>2010</creationdate><recordtype>article</recordtype><recordid>eNqFkM9KAzEQxoMoWKvP4F7E09Zks26SYxH_QVEEPYd0MrGp201NdoXefAff0CcxpeJVGMiQfF--mR8hp4xOGGXNxXLihoTdq8U0qWi-pXJC63qPjJgUvBRMNftkRFVFSy5Uc0iOUlpSykSuEXl6wKGPofNQgGlhaE3vQ5eK4ApT9IsQ_bAq5yahLXJMfvr-_HI-bbtisZlHb4uIBvoQi3lrujfsj8mBM23Ck99zTF5urp-v7srZ4-391XRWAheiL2UNSIU0CriVwJmTYJysUSiDl8wwVwMwtNZS23AnmrkQSvI6extlGVZ8TM53_65jeB8w9XrlE2Cbp8AwJC0bldeVqs5KsVNCDClFdHod_crEjWZUbxnqpf5jqLcMNZU6M8zOs98MkzIeF00HPv3ZKy7yRBXLuulOh3nhD49RJ_DYAVofEXptg_836wfCto6q</recordid><startdate>20101201</startdate><enddate>20101201</enddate><creator>Ma, X.B.</creator><creator>Chen, Y.X.</creator><creator>Wang, Y.</creator><creator>Zhang, P.Z.</creator><creator>Cao, B.</creator><creator>Lu, D.G.</creator><creator>Cheng, H.P.</creator><general>Elsevier B.V</general><general>Elsevier</general><scope>IQODW</scope><scope>AAYXX</scope><scope>CITATION</scope><scope>7SP</scope><scope>7TB</scope><scope>7U5</scope><scope>8FD</scope><scope>FR3</scope><scope>KR7</scope><scope>L7M</scope></search><sort><creationdate>20101201</creationdate><title>Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket</title><author>Ma, X.B. ; Chen, Y.X. ; Wang, Y. ; Zhang, P.Z. ; Cao, B. ; Lu, D.G. ; Cheng, H.P.</author></sort><facets><frbrtype>5</frbrtype><frbrgroupid>cdi_FETCH-LOGICAL-c377t-84ce078a9c3d8c31f8caf84e79ae51a1f4cc1eddd0d63f76b77983437769d1e23</frbrgroupid><rsrctype>articles</rsrctype><prefilter>articles</prefilter><language>eng</language><creationdate>2010</creationdate><topic>233Pa effect</topic><topic>Applied sciences</topic><topic>Blanket</topic><topic>Blanketing</topic><topic>Controled nuclear fusion plants</topic><topic>Energy</topic><topic>Energy. Thermal use of fuels</topic><topic>Exact sciences and technology</topic><topic>Fission nuclear power plants</topic><topic>Fuels</topic><topic>Fusion-fission hybrid reactors</topic><topic>Hybrid reactor</topic><topic>Installations for energy generation and conversion: thermal and electrical energy</topic><topic>Mathematical analysis</topic><topic>Monte Carlo methods</topic><topic>Nuclear engineering</topic><topic>Nuclear fuels</topic><topic>Nuclear reactor components</topic><topic>Preparation and processing of nuclear fuels</topic><topic>Thorium</topic><topic>Transport</topic><topic>Tritium</topic><toplevel>peer_reviewed</toplevel><toplevel>online_resources</toplevel><creatorcontrib>Ma, X.B.</creatorcontrib><creatorcontrib>Chen, Y.X.</creatorcontrib><creatorcontrib>Wang, Y.</creatorcontrib><creatorcontrib>Zhang, P.Z.</creatorcontrib><creatorcontrib>Cao, B.</creatorcontrib><creatorcontrib>Lu, D.G.</creatorcontrib><creatorcontrib>Cheng, H.P.</creatorcontrib><collection>Pascal-Francis</collection><collection>CrossRef</collection><collection>Electronics & Communications Abstracts</collection><collection>Mechanical & Transportation Engineering Abstracts</collection><collection>Solid State and Superconductivity Abstracts</collection><collection>Technology Research Database</collection><collection>Engineering Research Database</collection><collection>Civil Engineering Abstracts</collection><collection>Advanced Technologies Database with Aerospace</collection><jtitle>Fusion engineering and design</jtitle></facets><delivery><delcategory>Remote Search Resource</delcategory><fulltext>fulltext</fulltext></delivery><addata><au>Ma, X.B.</au><au>Chen, Y.X.</au><au>Wang, Y.</au><au>Zhang, P.Z.</au><au>Cao, B.</au><au>Lu, D.G.</au><au>Cheng, H.P.</au><format>journal</format><genre>article</genre><ristype>JOUR</ristype><atitle>Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket</atitle><jtitle>Fusion engineering and design</jtitle><date>2010-12-01</date><risdate>2010</risdate><volume>85</volume><issue>10</issue><spage>2227</spage><epage>2231</epage><pages>2227-2231</pages><issn>0920-3796</issn><eissn>1873-7196</eissn><coden>FEDEEE</coden><abstract>Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-life nuclear waste, and having inherent safety. Considering the fast development of nuclear power industry and the abundant thorium resources in China, the concept of a thorium-based breeding blanket for fuel production is proposed. The blanket using ThN or ThO
2 dispersed in graphite or BeO is investigated under a first neutron wall loading of 0.57
MW/cm
2 as ITER. The design helium cooled pebble bed design is employed according to the experience in the fusion reactor field. Preliminary neutronic calculations are performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. The behavior of the neutronic potential is observed for 960 days. The cumulative fissile fuel enrichment values varied between 6.88% and 8.56% depending on the fuel types. The tritium breeding ratio is greater than 1.05 for all investigated fuel types and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those modes during the operation period. The blanket energy multiplication factor
M, varies between 13.66 and 15.85 depending on the fuel types at the end of the operation period. In addition, the effect of
233Pa on the
233U production and
k
eff
are also discussed.</abstract><cop>Amsterdam</cop><pub>Elsevier B.V</pub><doi>10.1016/j.fusengdes.2010.08.044</doi><tpages>5</tpages></addata></record> |
| fulltext | fulltext |
| identifier | ISSN: 0920-3796 |
| ispartof | Fusion engineering and design, 2010-12, Vol.85 (10), p.2227-2231 |
| issn | 0920-3796 1873-7196 |
| language | eng |
| recordid | cdi_proquest_miscellaneous_869796894 |
| source | Elsevier ScienceDirect Journals |
| subjects | 233Pa effect Applied sciences Blanket Blanketing Controled nuclear fusion plants Energy Energy. Thermal use of fuels Exact sciences and technology Fission nuclear power plants Fuels Fusion-fission hybrid reactors Hybrid reactor Installations for energy generation and conversion: thermal and electrical energy Mathematical analysis Monte Carlo methods Nuclear engineering Nuclear fuels Nuclear reactor components Preparation and processing of nuclear fuels Thorium Transport Tritium |
| title | Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket |
| url | https://sfx.bib-bvb.de/sfx_tum?ctx_ver=Z39.88-2004&ctx_enc=info:ofi/enc:UTF-8&ctx_tim=2025-01-28T00%3A07%3A05IST&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=Neutronic%20calculations%20of%20a%20thorium-based%20fusion%E2%80%93fission%20hybrid%20reactor%20blanket&rft.jtitle=Fusion%20engineering%20and%20design&rft.au=Ma,%20X.B.&rft.date=2010-12-01&rft.volume=85&rft.issue=10&rft.spage=2227&rft.epage=2231&rft.pages=2227-2231&rft.issn=0920-3796&rft.eissn=1873-7196&rft.coden=FEDEEE&rft_id=info:doi/10.1016/j.fusengdes.2010.08.044&rft_dat=%3Cproquest_cross%3E869796894%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=869796894&rft_id=info:pmid/&rft_els_id=S0920379610004035&rfr_iscdi=true |