Loading…
Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel
U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate sample...
Saved in:
| Published in: | Journal of nuclear materials 2014-09, Vol.452 (1-3), p.533-547 |
|---|---|
| Main Authors: | , , , , , , , , |
| Format: | Article |
| Language: | English |
| Subjects: | |
| Citations: | Items that this one cites Items that cite this one |
| Online Access: | Get full text |
| Tags: |
Add Tag
No Tags, Be the first to tag this record!
|
| cited_by | cdi_FETCH-LOGICAL-c423t-74e435a17ff3859495d9c6e021e7cefe769fab955b12810dc2f439ad013c21343 |
|---|---|
| cites | cdi_FETCH-LOGICAL-c423t-74e435a17ff3859495d9c6e021e7cefe769fab955b12810dc2f439ad013c21343 |
| container_end_page | 547 |
| container_issue | 1-3 |
| container_start_page | 533 |
| container_title | Journal of nuclear materials |
| container_volume | 452 |
| creator | Zweifel, T. Valot, Ch Pontillon, Y. Lamontagne, J. Vermersch, A. Barrallier, L. Blay, T. Petry, W. Palancher, H. |
| description | U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500°C and 670°C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500°C FGs are released from IDL/matrix interfaces. The second peak at 670°C representing the main release of FGs originates from the interaction between U–Mo and matrix in the vicinity of the cladding. |
| doi_str_mv | 10.1016/j.jnucmat.2014.05.052 |
| format | article |
| fullrecord | <record><control><sourceid>proquest_hal_p</sourceid><recordid>TN_cdi_hal_primary_oai_HAL_hal_01021884v1</recordid><sourceformat>XML</sourceformat><sourcesystem>PC</sourcesystem><els_id>S0022311514003298</els_id><sourcerecordid>1611622057</sourcerecordid><originalsourceid>FETCH-LOGICAL-c423t-74e435a17ff3859495d9c6e021e7cefe769fab955b12810dc2f439ad013c21343</originalsourceid><addsrcrecordid>eNqFUMtqVEEQbUTBSfQThLvUxZ1U9eM-VmEIagIjLjTgrul0V2sPPbcn3XdC3PkP_qFfkp5MyFYoOFTVqUOdw9g7hCUCdmeb5Wba262ZlxxQLkHV4i_YAodetHLg8JItADhvBaJ6zU5K2QCAGkEt2I_VNJGJYfrZzFTm0iTfhKndhUhNyNm4YGZyTboPjhqbHpvrf3_-fklnq1jxW2hcKDvKpS7qF5FMbvye4hv2yptY6O0TnrLrTx-_X1y266-fry5W69ZKLua2lySFMth7LwY1ylG50XYEHKm35KnvRm9uRqVukA8IznIvxWgcoLAchRSn7MNR95eJepfD1uTfOpmgL1drfZgBVrFhkHdYue-P3F1Ot_tqV29DsRSjmSjti8YOseMcVF-p6ki1OZWSyT9rI-hD6nqjn1LXh9Q1qFq83p0f76h6vguUdbGBJksuZLKzdin8R-EB80CORg</addsrcrecordid><sourcetype>Open Access Repository</sourcetype><iscdi>true</iscdi><recordtype>article</recordtype><pqid>1611622057</pqid></control><display><type>article</type><title>Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel</title><source>ScienceDirect Journals</source><creator>Zweifel, T. ; Valot, Ch ; Pontillon, Y. ; Lamontagne, J. ; Vermersch, A. ; Barrallier, L. ; Blay, T. ; Petry, W. ; Palancher, H.</creator><creatorcontrib>Zweifel, T. ; Valot, Ch ; Pontillon, Y. ; Lamontagne, J. ; Vermersch, A. ; Barrallier, L. ; Blay, T. ; Petry, W. ; Palancher, H.</creatorcontrib><description>U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500°C and 670°C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500°C FGs are released from IDL/matrix interfaces. The second peak at 670°C representing the main release of FGs originates from the interaction between U–Mo and matrix in the vicinity of the cladding.</description><identifier>ISSN: 0022-3115</identifier><identifier>EISSN: 1873-4820</identifier><identifier>DOI: 10.1016/j.jnucmat.2014.05.052</identifier><language>eng</language><publisher>Elsevier B.V</publisher><subject>Engineering Sciences ; Materials</subject><ispartof>Journal of nuclear materials, 2014-09, Vol.452 (1-3), p.533-547</ispartof><rights>2014 Elsevier B.V.</rights><rights>Distributed under a Creative Commons Attribution 4.0 International License</rights><lds50>peer_reviewed</lds50><oa>free_for_read</oa><woscitedreferencessubscribed>false</woscitedreferencessubscribed><citedby>FETCH-LOGICAL-c423t-74e435a17ff3859495d9c6e021e7cefe769fab955b12810dc2f439ad013c21343</citedby><cites>FETCH-LOGICAL-c423t-74e435a17ff3859495d9c6e021e7cefe769fab955b12810dc2f439ad013c21343</cites><orcidid>0000-0001-7684-6625 ; 0000-0001-8603-8162</orcidid></display><links><openurl>$$Topenurl_article</openurl><openurlfulltext>$$Topenurlfull_article</openurlfulltext><thumbnail>$$Tsyndetics_thumb_exl</thumbnail><link.rule.ids>230,314,780,784,885,27924,27925</link.rule.ids><backlink>$$Uhttps://hal.science/hal-01021884$$DView record in HAL$$Hfree_for_read</backlink></links><search><creatorcontrib>Zweifel, T.</creatorcontrib><creatorcontrib>Valot, Ch</creatorcontrib><creatorcontrib>Pontillon, Y.</creatorcontrib><creatorcontrib>Lamontagne, J.</creatorcontrib><creatorcontrib>Vermersch, A.</creatorcontrib><creatorcontrib>Barrallier, L.</creatorcontrib><creatorcontrib>Blay, T.</creatorcontrib><creatorcontrib>Petry, W.</creatorcontrib><creatorcontrib>Palancher, H.</creatorcontrib><title>Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel</title><title>Journal of nuclear materials</title><description>U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500°C and 670°C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500°C FGs are released from IDL/matrix interfaces. The second peak at 670°C representing the main release of FGs originates from the interaction between U–Mo and matrix in the vicinity of the cladding.</description><subject>Engineering Sciences</subject><subject>Materials</subject><issn>0022-3115</issn><issn>1873-4820</issn><fulltext>true</fulltext><rsrctype>article</rsrctype><creationdate>2014</creationdate><recordtype>article</recordtype><recordid>eNqFUMtqVEEQbUTBSfQThLvUxZ1U9eM-VmEIagIjLjTgrul0V2sPPbcn3XdC3PkP_qFfkp5MyFYoOFTVqUOdw9g7hCUCdmeb5Wba262ZlxxQLkHV4i_YAodetHLg8JItADhvBaJ6zU5K2QCAGkEt2I_VNJGJYfrZzFTm0iTfhKndhUhNyNm4YGZyTboPjhqbHpvrf3_-fklnq1jxW2hcKDvKpS7qF5FMbvye4hv2yptY6O0TnrLrTx-_X1y266-fry5W69ZKLua2lySFMth7LwY1ylG50XYEHKm35KnvRm9uRqVukA8IznIvxWgcoLAchRSn7MNR95eJepfD1uTfOpmgL1drfZgBVrFhkHdYue-P3F1Ot_tqV29DsRSjmSjti8YOseMcVF-p6ki1OZWSyT9rI-hD6nqjn1LXh9Q1qFq83p0f76h6vguUdbGBJksuZLKzdin8R-EB80CORg</recordid><startdate>20140901</startdate><enddate>20140901</enddate><creator>Zweifel, T.</creator><creator>Valot, Ch</creator><creator>Pontillon, Y.</creator><creator>Lamontagne, J.</creator><creator>Vermersch, A.</creator><creator>Barrallier, L.</creator><creator>Blay, T.</creator><creator>Petry, W.</creator><creator>Palancher, H.</creator><general>Elsevier B.V</general><general>Elsevier</general><scope>AAYXX</scope><scope>CITATION</scope><scope>7ST</scope><scope>C1K</scope><scope>SOI</scope><scope>1XC</scope><scope>VOOES</scope><orcidid>https://orcid.org/0000-0001-7684-6625</orcidid><orcidid>https://orcid.org/0000-0001-8603-8162</orcidid></search><sort><creationdate>20140901</creationdate><title>Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel</title><author>Zweifel, T. ; Valot, Ch ; Pontillon, Y. ; Lamontagne, J. ; Vermersch, A. ; Barrallier, L. ; Blay, T. ; Petry, W. ; Palancher, H.</author></sort><facets><frbrtype>5</frbrtype><frbrgroupid>cdi_FETCH-LOGICAL-c423t-74e435a17ff3859495d9c6e021e7cefe769fab955b12810dc2f439ad013c21343</frbrgroupid><rsrctype>articles</rsrctype><prefilter>articles</prefilter><language>eng</language><creationdate>2014</creationdate><topic>Engineering Sciences</topic><topic>Materials</topic><toplevel>peer_reviewed</toplevel><toplevel>online_resources</toplevel><creatorcontrib>Zweifel, T.</creatorcontrib><creatorcontrib>Valot, Ch</creatorcontrib><creatorcontrib>Pontillon, Y.</creatorcontrib><creatorcontrib>Lamontagne, J.</creatorcontrib><creatorcontrib>Vermersch, A.</creatorcontrib><creatorcontrib>Barrallier, L.</creatorcontrib><creatorcontrib>Blay, T.</creatorcontrib><creatorcontrib>Petry, W.</creatorcontrib><creatorcontrib>Palancher, H.</creatorcontrib><collection>CrossRef</collection><collection>Environment Abstracts</collection><collection>Environmental Sciences and Pollution Management</collection><collection>Environment Abstracts</collection><collection>Hyper Article en Ligne (HAL)</collection><collection>Hyper Article en Ligne (HAL) (Open Access)</collection><jtitle>Journal of nuclear materials</jtitle></facets><delivery><delcategory>Remote Search Resource</delcategory><fulltext>fulltext</fulltext></delivery><addata><au>Zweifel, T.</au><au>Valot, Ch</au><au>Pontillon, Y.</au><au>Lamontagne, J.</au><au>Vermersch, A.</au><au>Barrallier, L.</au><au>Blay, T.</au><au>Petry, W.</au><au>Palancher, H.</au><format>journal</format><genre>article</genre><ristype>JOUR</ristype><atitle>Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel</atitle><jtitle>Journal of nuclear materials</jtitle><date>2014-09-01</date><risdate>2014</risdate><volume>452</volume><issue>1-3</issue><spage>533</spage><epage>547</epage><pages>533-547</pages><issn>0022-3115</issn><eissn>1873-4820</eissn><abstract>U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500°C and 670°C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500°C FGs are released from IDL/matrix interfaces. The second peak at 670°C representing the main release of FGs originates from the interaction between U–Mo and matrix in the vicinity of the cladding.</abstract><pub>Elsevier B.V</pub><doi>10.1016/j.jnucmat.2014.05.052</doi><tpages>15</tpages><orcidid>https://orcid.org/0000-0001-7684-6625</orcidid><orcidid>https://orcid.org/0000-0001-8603-8162</orcidid><oa>free_for_read</oa></addata></record> |
| fulltext | fulltext |
| identifier | ISSN: 0022-3115 |
| ispartof | Journal of nuclear materials, 2014-09, Vol.452 (1-3), p.533-547 |
| issn | 0022-3115 1873-4820 |
| language | eng |
| recordid | cdi_hal_primary_oai_HAL_hal_01021884v1 |
| source | ScienceDirect Journals |
| subjects | Engineering Sciences Materials |
| title | Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel |
| url | http://sfxeu10.hosted.exlibrisgroup.com/loughborough?ctx_ver=Z39.88-2004&ctx_enc=info:ofi/enc:UTF-8&ctx_tim=2024-12-26T13%3A41%3A15IST&url_ver=Z39.88-2004&url_ctx_fmt=infofi/fmt:kev:mtx:ctx&rfr_id=info:sid/primo.exlibrisgroup.com:primo3-Article-proquest_hal_p&rft_val_fmt=info:ofi/fmt:kev:mtx:journal&rft.genre=article&rft.atitle=Annealing%20tests%20of%20in-pile%20irradiated%20oxide%20coated%20U%E2%80%93Mo/Al%E2%80%93Si%20dispersed%20nuclear%20fuel&rft.jtitle=Journal%20of%20nuclear%20materials&rft.au=Zweifel,%20T.&rft.date=2014-09-01&rft.volume=452&rft.issue=1-3&rft.spage=533&rft.epage=547&rft.pages=533-547&rft.issn=0022-3115&rft.eissn=1873-4820&rft_id=info:doi/10.1016/j.jnucmat.2014.05.052&rft_dat=%3Cproquest_hal_p%3E1611622057%3C/proquest_hal_p%3E%3Cgrp_id%3Ecdi_FETCH-LOGICAL-c423t-74e435a17ff3859495d9c6e021e7cefe769fab955b12810dc2f439ad013c21343%3C/grp_id%3E%3Coa%3E%3C/oa%3E%3Curl%3E%3C/url%3E&rft_id=info:oai/&rft_pqid=1611622057&rft_id=info:pmid/&rfr_iscdi=true |