diff --git a/_data/pub.json b/_data/pub.json index 0609bf0f..859de8c3 100644 --- a/_data/pub.json +++ b/_data/pub.json @@ -1,7 +1,7 @@ [ { "key": "4R7ATE42", - "version": 30399, + "version": 30807, "library": { "type": "group", "id": 10058, @@ -41,14 +41,25 @@ } } }, + "lastModifiedByUser": { + "id": 112658, + "username": "gonuke", + "name": "", + "links": { + "alternate": { + "href": "https://www.zotero.org/gonuke", + "type": "text/html" + } + } + }, "creatorSummary": "Mummah and Wilson", "parsedDate": "2024-06-19", "numChildren": 1 }, - "bibtex": "\n@inproceedings{mummah_cyclus_2024,\n\taddress = {Las Vegas, NV},\n\ttitle = {Cyclus {Toolkit} {Enhancements} to {Simulate} {Nuclear} {Material} {Buying} {Patterns}},\n\tvolume = {130},\n\tbooktitle = {Transactions of the {American} {Nuclear} {Society}, {Annual} {Meeting} 2024},\n\tauthor = {Mummah, Kathryn A. and Wilson, Paul P.H.},\n\tmonth = jun,\n\tyear = {2024},\n\tnote = {(accepted)},\n}\n", + "bibtex": "\n@inproceedings{mummah_cyclus_2024,\n\taddress = {Las Vegas, NV},\n\ttitle = {Cyclus {Toolkit} {Enhancements} to {Simulate} {Nuclear} {Material} {Buying} {Patterns}},\n\tvolume = {130},\n\turl = {https://www.ans.org/pubs/transactions/article-55965/},\n\tbooktitle = {Transactions of the {American} {Nuclear} {Society}, {Annual} {Meeting} 2024},\n\tauthor = {Mummah, Kathryn A. and Wilson, Paul P.H.},\n\tmonth = jun,\n\tyear = {2024},\n\tnote = {(accepted)},\n}\n", "data": { "key": "4R7ATE42", - "version": 30399, + "version": 30807, "itemType": "conferencePaper", "title": "Cyclus Toolkit Enhancements to Simulate Nuclear Material Buying Patterns", "creators": [ @@ -76,7 +87,7 @@ "DOI": "", "ISBN": "", "shortTitle": "", - "url": "", + "url": "https://www.ans.org/pubs/transactions/article-55965/", "accessDate": "", "archive": "", "archiveLocation": "", @@ -92,7 +103,7 @@ "owl:sameAs": "http://zotero.org/groups/10058/items/9PR5P2RR" }, "dateAdded": "2024-03-20T19:47:42Z", - "dateModified": "2024-07-23T14:49:03Z" + "dateModified": "2024-09-09T14:48:20Z" } }, { @@ -349,7 +360,7 @@ }, { "key": "WTKDTRL7", - "version": 29705, + "version": 30812, "library": { "type": "group", "id": 10058, @@ -369,6 +380,12 @@ "alternate": { "href": "https://www.zotero.org/groups/10058/items/WTKDTRL7", "type": "text/html" + }, + "attachment": { + "href": "https://api.zotero.org/groups/10058/items/DLUFAQQI", + "type": "application/json", + "attachmentType": "application/pdf", + "attachmentSize": 441719 } }, "meta": { @@ -383,14 +400,25 @@ } } }, + "lastModifiedByUser": { + "id": 112658, + "username": "gonuke", + "name": "", + "links": { + "alternate": { + "href": "https://www.zotero.org/gonuke", + "type": "text/html" + } + } + }, "creatorSummary": "Mummah", "parsedDate": "2023-11", - "numChildren": 0 + "numChildren": 1 }, - "bibtex": "\n@inproceedings{mummah_bridging_2023,\n\taddress = {Washington D.C.},\n\ttitle = {Bridging the {Fidelity} {Gap} in {System}-{Scale} {Nuclear} {Fuel} {Cycle} {Simulations} for {Realistic} {State}-{Level} {Nuclear} {Material} {Accounting}},\n\tauthor = {Mummah, Kathryn A.},\n\tmonth = nov,\n\tyear = {2023},\n}\n", + "bibtex": "\n@inproceedings{mummah_bridging_2023,\n\taddress = {Washington D.C.},\n\ttitle = {Bridging the {Fidelity} {Gap} in {System}-{Scale} {Nuclear} {Fuel} {Cycle} {Simulations} for {Realistic} {State}-{Level} {Nuclear} {Material} {Accounting}},\n\turl = {https://www.ans.org/pubs/proceedings/article-55005/},\n\tbooktitle = {Proceedings of {Advances} in {Nonproliferation} {Technology} and {Policy} {Conference} 2023},\n\tauthor = {Mummah, Kathryn A.},\n\tmonth = nov,\n\tyear = {2023},\n}\n", "data": { "key": "WTKDTRL7", - "version": 29705, + "version": 30812, "itemType": "conferencePaper", "title": "Bridging the Fidelity Gap in System-Scale Nuclear Fuel Cycle Simulations for Realistic State-Level Nuclear Material Accounting", "creators": [ @@ -402,7 +430,7 @@ ], "abstractNote": "", "date": "November 2023", - "proceedingsTitle": "", + "proceedingsTitle": "Proceedings of Advances in Nonproliferation Technology and Policy Conference 2023", "conferenceName": "Advances in Nonproliferation Technology and Policy Conference 2023", "place": "Washington D.C.", "publisher": "", @@ -413,7 +441,7 @@ "DOI": "", "ISBN": "", "shortTitle": "", - "url": "", + "url": "https://www.ans.org/pubs/proceedings/article-55005/", "accessDate": "", "archive": "", "archiveLocation": "", @@ -427,7 +455,7 @@ ], "relations": {}, "dateAdded": "2024-03-20T19:47:42Z", - "dateModified": "2024-03-20T19:47:42Z" + "dateModified": "2024-09-09T14:49:15Z" } }, { @@ -933,7 +961,7 @@ }, { "key": "UBSWG585", - "version": 27218, + "version": 30815, "library": { "type": "group", "id": 10058, @@ -976,10 +1004,10 @@ "creatorSummary": "D'Angelo and WILSON", "numChildren": 1 }, - "bibtex": "\n@inproceedings{dangelo_sdr_2022,\n\taddress = {Seattle, WA, USA},\n\ttitle = {{SDR} {Calculations} {Involving} {Geometry} {Movement} {After} {Shutdown}},\n\tauthor = {D'Angelo, Chelsea and WILSON, Paul P. H.},\n\tmonth = sep,\n\tyear = {2022},\n}\n", + "bibtex": "\n@inproceedings{dangelo_sdr_2022,\n\taddress = {Seattle, WA, USA},\n\ttitle = {{SDR} {Calculations} {Involving} {Geometry} {Movement} {After} {Shutdown}},\n\turl = {https://www.ans.org/pubs/proceedings/article-52048/},\n\tbooktitle = {Porceedings of the 14th {International} {Conference} on {Radiation} {Shielding} and 21st {Topical} {Meeting} of the {Radiation} {Protection} and {Shielding} {Division}},\n\tauthor = {D'Angelo, Chelsea and WILSON, Paul P. H.},\n\tmonth = sep,\n\tyear = {2022},\n}\n", "data": { "key": "UBSWG585", - "version": 27218, + "version": 30815, "itemType": "conferencePaper", "title": "SDR Calculations Involving Geometry Movement After Shutdown", "creators": [ @@ -996,7 +1024,7 @@ ], "abstractNote": "", "date": "2022-09-25 9/25/22-9/29/22", - "proceedingsTitle": "", + "proceedingsTitle": "Porceedings of the 14th International Conference on Radiation Shielding and 21st Topical Meeting of the Radiation Protection and Shielding Division", "conferenceName": "14th International Conference on Radiation Shielding and 21st Topical Meeting of the Radiation Protection and Shielding Division", "place": "Seattle, WA, USA", "publisher": "", @@ -1007,7 +1035,7 @@ "DOI": "", "ISBN": "", "shortTitle": "", - "url": "", + "url": "https://www.ans.org/pubs/proceedings/article-52048/", "accessDate": "", "archive": "", "archiveLocation": "", @@ -1021,7 +1049,7 @@ ], "relations": {}, "dateAdded": "2022-11-02T11:00:29Z", - "dateModified": "2022-11-02T11:02:28Z" + "dateModified": "2024-09-09T14:50:11Z" } }, { @@ -1697,7 +1725,7 @@ }, { "key": "QW38HRSJ", - "version": 25069, + "version": 30816, "library": { "type": "group", "id": 10058, @@ -1737,14 +1765,25 @@ } } }, + "lastModifiedByUser": { + "id": 112658, + "username": "gonuke", + "name": "", + "links": { + "alternate": { + "href": "https://www.zotero.org/gonuke", + "type": "text/html" + } + } + }, "creatorSummary": "Mummah and WIlson", "parsedDate": "2020-07", - "numChildren": 1 + "numChildren": 2 }, - "bibtex": "\n@inproceedings{mummah_integrating_2020,\n\ttitle = {Integrating {Acquisition} {Pathway} {Analysis} {Into} {The} {Cyclus} {Fuel} {Cycle} {Simulator}},\n\tabstract = {The IAEA considers a State’s entire fuel cycle capability when evaluating and implementing safeguards, a process known as the State-Level Approach. Conducting Acquisition Path Analysis (APA) is one aspect of ensuring efficient use of safeguards resources and an objective evaluation of member States. APA is designed to identify, characterize, and rank technically-feasible pathways through a fuel cycle to produce weapons-usable material. This paper covers the integration of APA techniques into the Cyclus fuel cycle simulator. Material flowing through a nuclear fuel cycle can be represented by a directed graph (digraph) with vertices V(D) repre-\nsenting facilities and edges E(D) representing trade or material transport. In a Cyclus input file, a user defines a set of facility prototypes and the commodities that can be traded between them.\nFrom this user-specified fuel cycle, a digraph is generated representing all possible commodity trades between facilities. Graph traversal techniques are used to enumerate all pathways for\nmaterial to flow through the given fuel cycle. Pathways that produce weapons-usable material are filtered and further analyzed. Due to the flexibility of the Cyclus fuel cycle simulator, this\nmethod works for any fuel cycle, including ones that use closed facility models that are not part of the open source Cyclus and Cycamore facility libraries.},\n\tbooktitle = {Proceedings of the 61st {INMM} {Meeting}},\n\tauthor = {Mummah, Kathryn and WIlson, P. P.H},\n\tmonth = jul,\n\tyear = {2020},\n}\n", + "bibtex": "\n@inproceedings{mummah_integrating_2020,\n\ttitle = {Integrating {Acquisition} {Pathway} {Analysis} {Into} {The} {Cyclus} {Fuel} {Cycle} {Simulator}},\n\turl = {https://resources.inmm.org/annual-meeting-proceedings/integrating-acquisition-pathway-analysis-cyclus-fuel-cycle-simulator},\n\tabstract = {The IAEA considers a State’s entire fuel cycle capability when evaluating and implementing safeguards, a process known as the State-Level Approach. Conducting Acquisition Path Analysis (APA) is one aspect of ensuring efficient use of safeguards resources and an objective evaluation of member States. APA is designed to identify, characterize, and rank technically-feasible pathways through a fuel cycle to produce weapons-usable material. This paper covers the integration of APA techniques into the Cyclus fuel cycle simulator. Material flowing through a nuclear fuel cycle can be represented by a directed graph (digraph) with vertices V(D) repre-\nsenting facilities and edges E(D) representing trade or material transport. In a Cyclus input file, a user defines a set of facility prototypes and the commodities that can be traded between them.\nFrom this user-specified fuel cycle, a digraph is generated representing all possible commodity trades between facilities. Graph traversal techniques are used to enumerate all pathways for\nmaterial to flow through the given fuel cycle. Pathways that produce weapons-usable material are filtered and further analyzed. Due to the flexibility of the Cyclus fuel cycle simulator, this\nmethod works for any fuel cycle, including ones that use closed facility models that are not part of the open source Cyclus and Cycamore facility libraries.},\n\tbooktitle = {Proceedings of the 61st {INMM} {Meeting}},\n\tauthor = {Mummah, Kathryn and WIlson, P. P.H},\n\tmonth = jul,\n\tyear = {2020},\n}\n", "data": { "key": "QW38HRSJ", - "version": 25069, + "version": 30816, "itemType": "conferencePaper", "title": "Integrating Acquisition Pathway Analysis Into The Cyclus Fuel Cycle Simulator", "creators": [ @@ -1772,7 +1811,7 @@ "DOI": "", "ISBN": "", "shortTitle": "", - "url": "", + "url": "https://resources.inmm.org/annual-meeting-proceedings/integrating-acquisition-pathway-analysis-cyclus-fuel-cycle-simulator", "accessDate": "", "archive": "", "archiveLocation": "", @@ -1786,12 +1825,12 @@ ], "relations": {}, "dateAdded": "2020-11-12T15:04:00Z", - "dateModified": "2021-01-21T05:47:23Z" + "dateModified": "2024-09-09T14:51:38Z" } }, { "key": "H4VPVVT5", - "version": 25203, + "version": 30817, "library": { "type": "group", "id": 10058, @@ -1813,10 +1852,10 @@ "type": "text/html" }, "attachment": { - "href": "https://api.zotero.org/groups/10058/items/HCUMB62F", + "href": "https://api.zotero.org/groups/10058/items/WFMNLVJP", "type": "application/json", "attachmentType": "application/pdf", - "attachmentSize": 290296 + "attachmentSize": 409270 } }, "meta": { @@ -1831,14 +1870,25 @@ } } }, + "lastModifiedByUser": { + "id": 112658, + "username": "gonuke", + "name": "", + "links": { + "alternate": { + "href": "https://www.zotero.org/gonuke", + "type": "text/html" + } + } + }, "creatorSummary": "Park et al.", "parsedDate": "2020-04-29", - "numChildren": 1 + "numChildren": 2 }, - "bibtex": "\n@inproceedings{park_evaluation_2020,\n\taddress = {Cambridge, United Kingdom},\n\ttitle = {Evaluation of {Critical} {Experiments} in the {University} of {Wisconsin} {Nuclear} {Reactor} ({UWNR}) with {Uncertainty} {Quantification}},\n\tisbn = {978-1-5272-6447-2},\n\tabstract = {An improved computational model for the University of Wisconsin Nuclear Reactor (UWNR) has been developed to facilitate automated input generation, data provenance, and modularity for alternate representations. This development was initiated as part of efforts to evaluate recent data acquired during an experimental campaign conducted at UWNR to generate benchmark data for validation. Specifically, this evaluation effort aims to contribute a number of fresh and depleted critical (CRIT) configurations of UWNR as well as steady-state and transient reaction-rate (RRATE) measurements. Previous efforts led to a scripted UWNR model that supports automated generation of inputs for MCNP and Serpent. Recently, this capability was extended to SCALE/KENO, which required significant changes to the underlying geometry and material representations. All three Monte Carlo tools (MCNP, Serpent, and KENO) are being used to evaluate a variety of zero-power, fresh-critical configuration and will be used to model burnup for evaluation of depleted-critical configurations. The inclusion of SCALE/KENO input generation makes possible a variety of sensitivity and uncertainty analyses using the TSUNAMI and SAMPLER modules of SCALE. In addition, an automated mesh-generation option was added based on the UW-developed, MCNP-to-CAD plugin. As a result, a meshed geometry for use with deterministic tools (e.g., MAMMOTH/Rattlesnake) can be produced that is fully consistent with the Monte Carlo models. Work is ongoing to develop a full core model in MAMMOTH/Rattlesnake, which is a deterministic code based on the MOOSE framework. This model will be used for the evaluation of several transient experiments conducted at UWNR. Preliminary results of fresh-critical configurations show a good agreement among the four codes and experimental data. Also, preliminary results of depleted-critical configurations indicate that the depleted core model successfully tracks core reactivity over time as long as an initial (but relatively small) reactivity bias is eliminated. Formal uncertainty quantification will be carried out using SCALE to study the impact of model uncertainties on the effective multiplication factor and other observables. In conclusion, the evaluation of UWNR benchmark data provides increased confidence in various states/configurations of the UWNR computational model and will provide a unique model for use by other analysts.},\n\tbooktitle = {Proceedings of the {PHYSOR} 2020},\n\tauthor = {Park, YoungHui and Cheng, Ye and Elzohery, Rabab and Wilson, Paul P.H. and Roberts, Jeremy A. and DeHart, Mark D.},\n\tmonth = apr,\n\tyear = {2020},\n\tpages = {10},\n}\n", + "bibtex": "\n@inproceedings{park_evaluation_2020,\n\taddress = {Cambridge, United Kingdom},\n\ttitle = {Evaluation of {Critical} {Experiments} in the {University} of {Wisconsin} {Nuclear} {Reactor} ({UWNR}) with {Uncertainty} {Quantification}},\n\tisbn = {978-1-5272-6447-2},\n\turl = {https://doi.org/10.1051/epjconf/202124710032},\n\tabstract = {An improved computational model for the University of Wisconsin Nuclear Reactor (UWNR) has been developed to facilitate automated input generation, data provenance, and modularity for alternate representations. This development was initiated as part of efforts to evaluate recent data acquired during an experimental campaign conducted at UWNR to generate benchmark data for validation. Specifically, this evaluation effort aims to contribute a number of fresh and depleted critical (CRIT) configurations of UWNR as well as steady-state and transient reaction-rate (RRATE) measurements. Previous efforts led to a scripted UWNR model that supports automated generation of inputs for MCNP and Serpent. Recently, this capability was extended to SCALE/KENO, which required significant changes to the underlying geometry and material representations. All three Monte Carlo tools (MCNP, Serpent, and KENO) are being used to evaluate a variety of zero-power, fresh-critical configuration and will be used to model burnup for evaluation of depleted-critical configurations. The inclusion of SCALE/KENO input generation makes possible a variety of sensitivity and uncertainty analyses using the TSUNAMI and SAMPLER modules of SCALE. In addition, an automated mesh-generation option was added based on the UW-developed, MCNP-to-CAD plugin. As a result, a meshed geometry for use with deterministic tools (e.g., MAMMOTH/Rattlesnake) can be produced that is fully consistent with the Monte Carlo models. Work is ongoing to develop a full core model in MAMMOTH/Rattlesnake, which is a deterministic code based on the MOOSE framework. This model will be used for the evaluation of several transient experiments conducted at UWNR. Preliminary results of fresh-critical configurations show a good agreement among the four codes and experimental data. Also, preliminary results of depleted-critical configurations indicate that the depleted core model successfully tracks core reactivity over time as long as an initial (but relatively small) reactivity bias is eliminated. Formal uncertainty quantification will be carried out using SCALE to study the impact of model uncertainties on the effective multiplication factor and other observables. In conclusion, the evaluation of UWNR benchmark data provides increased confidence in various states/configurations of the UWNR computational model and will provide a unique model for use by other analysts.},\n\tbooktitle = {Proceedings of the {PHYSOR} 2020},\n\tauthor = {Park, YoungHui and Cheng, Ye and Elzohery, Rabab and Wilson, Paul P.H. and Roberts, Jeremy A. and DeHart, Mark D.},\n\tmonth = apr,\n\tyear = {2020},\n\tpages = {10},\n}\n", "data": { "key": "H4VPVVT5", - "version": 25203, + "version": 30817, "itemType": "conferencePaper", "title": "Evaluation of Critical Experiments in the University of Wisconsin Nuclear Reactor (UWNR) with Uncertainty Quantification", "creators": [ @@ -1886,7 +1936,7 @@ "DOI": "", "ISBN": "978-1-5272-6447-2", "shortTitle": "", - "url": "", + "url": "https://doi.org/10.1051/epjconf/202124710032", "accessDate": "", "archive": "", "archiveLocation": "", @@ -1900,7 +1950,7 @@ ], "relations": {}, "dateAdded": "2021-02-07T00:18:47Z", - "dateModified": "2021-02-07T00:43:37Z" + "dateModified": "2024-09-09T14:52:31Z" } }, { @@ -10423,7 +10473,7 @@ }, { "key": "HU8FELQI", - "version": 21903, + "version": 30749, "library": { "type": "group", "id": 10058, @@ -10467,10 +10517,10 @@ "parsedDate": "2017-09-01", "numChildren": 2 }, - "bibtex": "\n@article{shriwise_particle_2017,\n\tseries = {Special {Issue} on {International} {Conference} on {Mathematics} and {Computational} {Methods} {Applied} to {Nuclear} {Science} and {Engineering} 2017 ({M}\\&{C} 2017)},\n\ttitle = {Particle tracking acceleration via signed distance fields in direct-accelerated geometry {Monte} {Carlo}},\n\tvolume = {49},\n\tissn = {1738-5733},\n\turl = {http://www.sciencedirect.com/science/article/pii/S1738573317303145},\n\tdoi = {10.1016/j.net.2017.08.008},\n\tabstract = {Computer-aided design (CAD)-based Monte Carlo radiation transport is of value to the nuclear engineering community for its ability to conduct transport on high-fidelity models of nuclear systems, but it is more computationally expensive than native geometry representations. This work describes the adaptation of a rendering data structure, the signed distance field, as a geometric query tool for accelerating CAD-based transport in the direct-accelerated geometry Monte Carlo toolkit. Demonstrations of its effectiveness are shown for several problems. The beginnings of a predictive model for the data structure's utilization based on various problem parameters is also introduced.},\n\tnumber = {6},\n\turldate = {2017-11-24},\n\tjournal = {Nuclear Engineering and Technology},\n\tauthor = {Shriwise, Patrick C. and Davis, Andrew and Jacobson, Lucas J. and Wilson, Paul P. H.},\n\tmonth = sep,\n\tyear = {2017},\n\tkeywords = {CAD, DAGMC, Monte Carlo, Radiation Transport},\n\tpages = {1189--1198},\n}\n", + "bibtex": "\n@article{shriwise_particle_2017,\n\tseries = {Special {Issue} on {International} {Conference} on {Mathematics} and {Computational} {Methods} {Applied} to {Nuclear} {Science} and {Engineering} 2017 ({M}\\&{C} 2017)},\n\ttitle = {Particle tracking acceleration via signed distance fields in direct-accelerated geometry {Monte} {Carlo}},\n\tvolume = {49},\n\tissn = {1738-5733},\n\turl = {http://www.sciencedirect.com/science/article/pii/S1738573317303145},\n\tdoi = {10.1016/j.net.2017.08.008},\n\tabstract = {Computer-aided design (CAD)-based Monte Carlo radiation transport is of value to the nuclear engineering community for its ability to conduct transport on high-fidelity models of nuclear systems, but it is more computationally expensive than native geometry representations. This work describes the adaptation of a rendering data structure, the signed distance field, as a geometric query tool for accelerating CAD-based transport in the direct-accelerated geometry Monte Carlo toolkit. Demonstrations of its effectiveness are shown for several problems. The beginnings of a predictive model for the data structure's utilization based on various problem parameters is also introduced.},\n\tnumber = {6},\n\turldate = {2017-11-24},\n\tjournal = {Nuclear Engineering and Technology},\n\tauthor = {Shriwise, Patrick C. and Davis, Andrew and Jacobson, Lucas J. and Wilson, Paul P. H.},\n\tmonth = sep,\n\tyear = {2017},\n\tkeywords = {CAD, DAGMC, Monte Carlo, Radiation Transport, product},\n\tpages = {1189--1198},\n}\n", "data": { "key": "HU8FELQI", - "version": 21903, + "version": 30749, "itemType": "journalArticle", "title": "Particle tracking acceleration via signed distance fields in direct-accelerated geometry Monte Carlo", "creators": [ @@ -10533,21 +10583,23 @@ { "tag": "Radiation Transport", "type": 1 + }, + { + "tag": "product" } ], "collections": [ - "UKXV4KID", - "H442QZRN", - "CMA2SK5V" + "APMMJXES", + "UKXV4KID" ], "relations": {}, "dateAdded": "2017-11-24T19:24:01Z", - "dateModified": "2017-11-24T19:24:01Z" + "dateModified": "2024-09-08T17:29:03Z" } }, { "key": "PG5YTTAG", - "version": 21903, + "version": 30748, "library": { "type": "group", "id": 10058, @@ -10591,10 +10643,10 @@ "parsedDate": "2017-07-04", "numChildren": 2 }, - "bibtex": "\n@article{el-guebaly_design_2017,\n\ttitle = {Design and {Evaluation} of {Nuclear} {System} for {ARIES}-{ACT2} {Power} {Plant} with {DCLL} {Blanket}},\n\tvolume = {72},\n\tissn = {1536-1055},\n\turl = {https://doi.org/10.1080/15361055.2016.1273669},\n\tdoi = {10.1080/15361055.2016.1273669},\n\tabstract = {The ARIES team has examined a multitude of fusion concepts over a period of 25 years. In recent years, the team wrapped up the Advanced Research, Innovation, and Evaluation Study (ARIES) series by completing the detailed design of the ARIES–Advanced and Conservative Tokamak (ARIES-ACT2) power plant—a plant with conservative physics and technology, representing a tokamak with reduced-activation ferritic/martensitic (RAFM) structure and dual-coolant lead-lithium blanket. The integration of nuclear assessments (neutronics, shielding, and activation) is an essential element to ARIES-ACT2 success. This paper highlights the design philosophy of in-vessel components and characterizes several nuclear-related issues that have been addressed during the course of the study to improve the ARIES-ACT2 design: sufficient breeding of tritium to fuel the plasma, well-optimized in-vessel components that satisfy all design requirements and guarantee the shielding functionality of its radial/vertical builds, survivability of low-activation/radiation-resistant structural materials in 14-MeV neutron environment, activation concerns for RAFM and corrosion-resistant oxide-dispersion-strengthened alloys, and an integral approach to handle the mildly radioactive materials during operation and after decommissioning.},\n\tnumber = {1},\n\turldate = {2018-04-05},\n\tjournal = {Fusion Science and Technology},\n\tauthor = {El-Guebaly, L. and Mynsberge, L. and Davis, A. and D’Angelo, C. and Rowcliffe, A. and Pint, B. and Team, ARIES-ACT},\n\tmonth = jul,\n\tyear = {2017},\n\tkeywords = {Activation analysis, DCLL blanket, neutronics},\n\tpages = {17--40},\n}\n", + "bibtex": "\n@article{el-guebaly_design_2017,\n\ttitle = {Design and {Evaluation} of {Nuclear} {System} for {ARIES}-{ACT2} {Power} {Plant} with {DCLL} {Blanket}},\n\tvolume = {72},\n\tissn = {1536-1055},\n\turl = {https://doi.org/10.1080/15361055.2016.1273669},\n\tdoi = {10.1080/15361055.2016.1273669},\n\tabstract = {The ARIES team has examined a multitude of fusion concepts over a period of 25 years. In recent years, the team wrapped up the Advanced Research, Innovation, and Evaluation Study (ARIES) series by completing the detailed design of the ARIES–Advanced and Conservative Tokamak (ARIES-ACT2) power plant—a plant with conservative physics and technology, representing a tokamak with reduced-activation ferritic/martensitic (RAFM) structure and dual-coolant lead-lithium blanket. The integration of nuclear assessments (neutronics, shielding, and activation) is an essential element to ARIES-ACT2 success. This paper highlights the design philosophy of in-vessel components and characterizes several nuclear-related issues that have been addressed during the course of the study to improve the ARIES-ACT2 design: sufficient breeding of tritium to fuel the plasma, well-optimized in-vessel components that satisfy all design requirements and guarantee the shielding functionality of its radial/vertical builds, survivability of low-activation/radiation-resistant structural materials in 14-MeV neutron environment, activation concerns for RAFM and corrosion-resistant oxide-dispersion-strengthened alloys, and an integral approach to handle the mildly radioactive materials during operation and after decommissioning.},\n\tnumber = {1},\n\turldate = {2018-04-05},\n\tjournal = {Fusion Science and Technology},\n\tauthor = {El-Guebaly, L. and Mynsberge, L. and Davis, A. and D’Angelo, C. and Rowcliffe, A. and Pint, B. and Team, ARIES-ACT},\n\tmonth = jul,\n\tyear = {2017},\n\tkeywords = {Activation analysis, DCLL blanket, neutronics, product},\n\tpages = {17--40},\n}\n", "data": { "key": "PG5YTTAG", - "version": 21903, + "version": 30748, "itemType": "journalArticle", "title": "Design and Evaluation of Nuclear System for ARIES-ACT2 Power Plant with DCLL Blanket", "creators": [ @@ -10668,21 +10720,23 @@ { "tag": "neutronics", "type": 1 + }, + { + "tag": "product" } ], "collections": [ - "UKXV4KID", - "H442QZRN", - "CMA2SK5V" + "APMMJXES", + "UKXV4KID" ], "relations": {}, "dateAdded": "2018-04-05T12:34:36Z", - "dateModified": "2018-04-05T12:34:36Z" + "dateModified": "2024-09-08T17:29:03Z" } }, { "key": "PP39E2UP", - "version": 21903, + "version": 30781, "library": { "type": "group", "id": 10058, @@ -10737,10 +10791,10 @@ "parsedDate": "2017-07", "numChildren": 1 }, - "bibtex": "\n@article{biondo_transmutation_2017,\n\ttitle = {Transmutation {Approximations} for the {Application} of {Hybrid} {Monte} {Carlo}/{Deterministic} {Neutron} {Transport} to {Shutdown} {Dose} {Rate} {Analysis}},\n\tvolume = {187},\n\turl = {http://www.tandfonline.com/doi/abs/10.1080/00295639.2016.1275848},\n\tabstract = {In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate\n(SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic\nneutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using\ndeterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed\nhere, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR\nproblem, GT-CADIS provides speedups of 200 ± 100 relative to global variance reduction with the Forward Weighted (FW)-CADIS method and 9 ± 5 · 10{\\textasciicircum}4 relative to analog. This work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR\nanalysis.},\n\tnumber = {1},\n\tjournal = {Nuclear Science and Engineering},\n\tauthor = {Biondo, Elliott D. and Wilson, Paul P.H.},\n\tmonth = jul,\n\tyear = {2017},\n\tpages = {27--48},\n}\n", + "bibtex": "\n@article{biondo_transmutation_2017,\n\ttitle = {Transmutation {Approximations} for the {Application} of {Hybrid} {Monte} {Carlo}/{Deterministic} {Neutron} {Transport} to {Shutdown} {Dose} {Rate} {Analysis}},\n\tvolume = {187},\n\turl = {http://www.tandfonline.com/doi/abs/10.1080/00295639.2016.1275848},\n\tabstract = {In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate\n(SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic\nneutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using\ndeterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed\nhere, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR\nproblem, GT-CADIS provides speedups of 200 ± 100 relative to global variance reduction with the Forward Weighted (FW)-CADIS method and 9 ± 5 · 10{\\textasciicircum}4 relative to analog. This work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR\nanalysis.},\n\tnumber = {1},\n\tjournal = {Nuclear Science and Engineering},\n\tauthor = {Biondo, Elliott D. and Wilson, Paul P.H.},\n\tmonth = jul,\n\tyear = {2017},\n\tkeywords = {CNERG:HK20 Final Report, product},\n\tpages = {27--48},\n}\n", "data": { "key": "PP39E2UP", - "version": 21903, + "version": 30781, "itemType": "journalArticle", "title": "Transmutation Approximations for the Application of Hybrid Monte Carlo/Deterministic Neutron Transport to Shutdown Dose Rate Analysis", "creators": [ @@ -10777,16 +10831,21 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + }, + { + "tag": "product" + } + ], "collections": [ - "UKXV4KID", - "4MDZ29N8", - "H442QZRN", - "CMA2SK5V" + "APMMJXES", + "UKXV4KID" ], "relations": {}, "dateAdded": "2017-06-23T02:45:37Z", - "dateModified": "2017-06-23T02:45:37Z" + "dateModified": "2024-09-08T17:38:42Z" } }, { @@ -10906,7 +10965,7 @@ }, { "key": "H4P52FWV", - "version": 21903, + "version": 30749, "library": { "type": "group", "id": 10058, @@ -10950,10 +11009,10 @@ "parsedDate": "2017", "numChildren": 1 }, - "bibtex": "\n@article{harb_effect_2017,\n\ttitle = {The {Effect} of {Constructed} {Mesh}-{Based} {Fluxes} on {Shutdown} {Dose} {Rate} {Calculations} in {Fusion} {Energy} {Systems}},\n\tvolume = {117},\n\tnumber = {1},\n\tjournal = {Transactions of the American Nuclear Society},\n\tauthor = {Harb, Moataz and Wilson, Paul P.H. and Davis, Andrew},\n\tmonth = nov,\n\tyear = {2017},\n\tpages = {1216--1219},\n}\n", + "bibtex": "\n@article{harb_effect_2017,\n\ttitle = {The {Effect} of {Constructed} {Mesh}-{Based} {Fluxes} on {Shutdown} {Dose} {Rate} {Calculations} in {Fusion} {Energy} {Systems}},\n\tvolume = {117},\n\tnumber = {1},\n\tjournal = {Transactions of the American Nuclear Society},\n\tauthor = {Harb, Moataz and Wilson, Paul P.H. and Davis, Andrew},\n\tmonth = nov,\n\tyear = {2017},\n\tkeywords = {product},\n\tpages = {1216--1219},\n}\n", "data": { "key": "H4P52FWV", - "version": 21903, + "version": 30749, "itemType": "journalArticle", "title": "The Effect of Constructed Mesh-Based Fluxes on Shutdown Dose Rate Calculations in Fusion Energy Systems", "creators": [ @@ -10995,15 +11054,18 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "product" + } + ], "collections": [ - "UKXV4KID", - "H442QZRN", - "CMA2SK5V" + "APMMJXES", + "UKXV4KID" ], "relations": {}, "dateAdded": "2017-11-24T19:31:43Z", - "dateModified": "2017-11-24T19:32:39Z" + "dateModified": "2024-09-08T17:29:03Z" } }, { @@ -11940,7 +12002,7 @@ }, { "key": "USAJZ3FD", - "version": 21903, + "version": 30781, "library": { "type": "group", "id": 10058, @@ -11995,10 +12057,10 @@ "parsedDate": "2016", "numChildren": 1 }, - "bibtex": "\n@article{biondo_shutdown_2016,\n\ttitle = {Shutdown {Dose} {Rate} {Analysis} with {CAD} {Geometry}, {Cartesian}/{Tetrahedral} {Mesh}, and {Advanced} {Variance} {Reduction}},\n\tvolume = {106},\n\tissn = {0920-3796},\n\turl = {http://www.sciencedirect.com/science/article/pii/S0920379616302009},\n\tdoi = {http://dx.doi.org/10.1016/j.fusengdes.2016.03.004},\n\tabstract = {In fusion energy systems (FES) high-energy neutrons born from burning plasma activate system components to form radionuclides. The biological dose rate that results from photons emitted by these radionuclides after shutdown—the shutdown dose rate (SDR)—must be quantified for maintenance planning. This can be done using the Rigorous Two-Step (R2S) method, which involves separate neutron and photon transport calculations, coupled by a nuclear inventory analysis code. The geometric complexity and highly attenuating configuration of FES motivates the use of CAD geometry and advanced variance\nreduction for this analysis.\n\nAn R2S workflow has been created with the new capability of performing SDR analysis directly from CAD geometry with Cartesian or tetrahedral meshes and with biased photon source sampling, enabling\nthe use of the Consistent Adjoint Driven Importance Sampling (CADIS) variance reduction technique. This workflow has been validated with the Frascati Neutron Generator (FNG)-ITER SDR benchmark using both Cartesian and tetrahedral meshes and both unbiased and biased photon source sampling. All results are\nwithin 20.4\\% of experimental values, which constitutes satisfactory agreement. Photon transport using\nCADIS is demonstrated to yield speedups as high as 8.5·10{\\textasciicircum}5 for problems using the FNG geometry.},\n\tjournal = {Fusion Engineering and Design},\n\tauthor = {Biondo, Elliott D. and Davis, Andrew and Wilson, Paul P.H.},\n\tmonth = may,\n\tyear = {2016},\n\tpages = {77--84},\n}\n", + "bibtex": "\n@article{biondo_shutdown_2016,\n\ttitle = {Shutdown {Dose} {Rate} {Analysis} with {CAD} {Geometry}, {Cartesian}/{Tetrahedral} {Mesh}, and {Advanced} {Variance} {Reduction}},\n\tvolume = {106},\n\tissn = {0920-3796},\n\turl = {http://www.sciencedirect.com/science/article/pii/S0920379616302009},\n\tdoi = {http://dx.doi.org/10.1016/j.fusengdes.2016.03.004},\n\tabstract = {In fusion energy systems (FES) high-energy neutrons born from burning plasma activate system components to form radionuclides. The biological dose rate that results from photons emitted by these radionuclides after shutdown—the shutdown dose rate (SDR)—must be quantified for maintenance planning. This can be done using the Rigorous Two-Step (R2S) method, which involves separate neutron and photon transport calculations, coupled by a nuclear inventory analysis code. The geometric complexity and highly attenuating configuration of FES motivates the use of CAD geometry and advanced variance\nreduction for this analysis.\n\nAn R2S workflow has been created with the new capability of performing SDR analysis directly from CAD geometry with Cartesian or tetrahedral meshes and with biased photon source sampling, enabling\nthe use of the Consistent Adjoint Driven Importance Sampling (CADIS) variance reduction technique. This workflow has been validated with the Frascati Neutron Generator (FNG)-ITER SDR benchmark using both Cartesian and tetrahedral meshes and both unbiased and biased photon source sampling. All results are\nwithin 20.4\\% of experimental values, which constitutes satisfactory agreement. Photon transport using\nCADIS is demonstrated to yield speedups as high as 8.5·10{\\textasciicircum}5 for problems using the FNG geometry.},\n\tjournal = {Fusion Engineering and Design},\n\tauthor = {Biondo, Elliott D. and Davis, Andrew and Wilson, Paul P.H.},\n\tmonth = may,\n\tyear = {2016},\n\tkeywords = {CNERG:HK20 Final Report, product},\n\tpages = {77--84},\n}\n", "data": { "key": "USAJZ3FD", - "version": 21903, + "version": 30781, "itemType": "journalArticle", "title": "Shutdown Dose Rate Analysis with CAD Geometry, Cartesian/Tetrahedral Mesh, and Advanced Variance Reduction", "creators": [ @@ -12040,16 +12102,21 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + }, + { + "tag": "product" + } + ], "collections": [ - "UKXV4KID", - "4MDZ29N8", - "H442QZRN", - "CMA2SK5V" + "APMMJXES", + "UKXV4KID" ], "relations": {}, "dateAdded": "2016-04-04T19:55:27Z", - "dateModified": "2017-01-11T03:36:09Z" + "dateModified": "2024-09-08T17:38:42Z" } }, { @@ -12754,8 +12821,7 @@ "extra": "", "tags": [], "collections": [ - "UKXV4KID", - "RI2DQ3B2" + "UKXV4KID" ], "relations": {}, "dateAdded": "2016-10-26T19:23:01Z", @@ -12875,7 +12941,7 @@ }, { "key": "E5STAZZD", - "version": 21761, + "version": 30777, "library": { "type": "group", "id": 10058, @@ -12930,10 +12996,10 @@ "parsedDate": "2015-04-19", "numChildren": 2 }, - "bibtex": "\n@inproceedings{biondo_accelerating_2015,\n\taddress = {Nashville, Tennessee},\n\ttitle = {Accelerating {Fusion} {Reactor} {Neutronics} {Modeling} by {Automatic} {Coupling} of {Hybrid} {Monte} {Carlo}/{Deterministic} {Transport} on {CAD} {Geometry}},\n\tisbn = {978-0-89448-720-0},\n\tabstract = {Detailed radiation transport calculations are necessary for many aspects of the design of fusion\nenergy systems (FES) such as ensuring occupational safety, assessing the activation of system components\nfor waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid\nMonte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily\nshielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of\nFES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required\nthe translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion\nGenerator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified\nto support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and\neliminating the need for this translation step. This was done by adding a new ray tracing routine\nto ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC)\nsoftware library. This new capability is demonstrated with a prompt dose rate calculation for an ITER\ncomputational benchmark problem using both the Consistent Adjoint Driven Importance Sampling\n(CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters\nproduced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry.\nSignificant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as\na factor of 59.6).},\n\tbooktitle = {Mathematics \\& {Computations} ({M}\\&{C}+{SNA}+{MC} 2015)},\n\tauthor = {Biondo, Elliott and Ibrahim, Ahmad M. and Mosher, Scott W. and Grove, Robert E.},\n\tmonth = apr,\n\tyear = {2015},\n}\n", + "bibtex": "\n@inproceedings{biondo_accelerating_2015,\n\taddress = {Nashville, Tennessee},\n\ttitle = {Accelerating {Fusion} {Reactor} {Neutronics} {Modeling} by {Automatic} {Coupling} of {Hybrid} {Monte} {Carlo}/{Deterministic} {Transport} on {CAD} {Geometry}},\n\tisbn = {978-0-89448-720-0},\n\tabstract = {Detailed radiation transport calculations are necessary for many aspects of the design of fusion\nenergy systems (FES) such as ensuring occupational safety, assessing the activation of system components\nfor waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid\nMonte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily\nshielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of\nFES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required\nthe translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion\nGenerator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified\nto support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and\neliminating the need for this translation step. This was done by adding a new ray tracing routine\nto ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC)\nsoftware library. This new capability is demonstrated with a prompt dose rate calculation for an ITER\ncomputational benchmark problem using both the Consistent Adjoint Driven Importance Sampling\n(CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters\nproduced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry.\nSignificant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as\na factor of 59.6).},\n\tbooktitle = {Mathematics \\& {Computations} ({M}\\&{C}+{SNA}+{MC} 2015)},\n\tauthor = {Biondo, Elliott and Ibrahim, Ahmad M. and Mosher, Scott W. and Grove, Robert E.},\n\tmonth = apr,\n\tyear = {2015},\n\tkeywords = {CNERG:HK20 Final Report},\n}\n", "data": { "key": "E5STAZZD", - "version": 21761, + "version": 30777, "itemType": "conferencePaper", "title": "Accelerating Fusion Reactor Neutronics Modeling by Automatic Coupling of Hybrid Monte Carlo/Deterministic Transport on CAD Geometry", "creators": [ @@ -12979,14 +13045,18 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + } + ], "collections": [ "UKXV4KID", - "4MDZ29N8" + "QNJGHVT4" ], "relations": {}, "dateAdded": "2015-07-09T20:06:20Z", - "dateModified": "2017-09-04T20:57:46Z" + "dateModified": "2024-09-08T17:38:36Z" } }, { @@ -13332,7 +13402,7 @@ }, { "key": "MSA966E9", - "version": 21903, + "version": 30781, "library": { "type": "group", "id": 10058, @@ -13387,10 +13457,10 @@ "parsedDate": "2015", "numChildren": 2 }, - "bibtex": "\n@article{biondo_rigorous_2015,\n\tseries = {Best of {Radiation} {Protection} and {Shielding} {Division} 2014---{II}},\n\ttitle = {Rigorous {Two}-{Step} {Activation} for {Fusion} {Systems} with {PyNE}},\n\tvolume = {112},\n\tjournal = {Transactions of the American Nuclear Society},\n\tauthor = {Biondo, Elliott and Davis, Andrew and Scopatz, Anthony and Wilson, Paul P.H.},\n\tyear = {2015},\n\tpages = {617--620},\n}\n", + "bibtex": "\n@article{biondo_rigorous_2015,\n\tseries = {Best of {Radiation} {Protection} and {Shielding} {Division} 2014---{II}},\n\ttitle = {Rigorous {Two}-{Step} {Activation} for {Fusion} {Systems} with {PyNE}},\n\tvolume = {112},\n\tjournal = {Transactions of the American Nuclear Society},\n\tauthor = {Biondo, Elliott and Davis, Andrew and Scopatz, Anthony and Wilson, Paul P.H.},\n\tyear = {2015},\n\tkeywords = {CNERG:HK20 Final Report, product},\n\tpages = {617--620},\n}\n", "data": { "key": "MSA966E9", - "version": 21903, + "version": 30781, "itemType": "journalArticle", "title": "Rigorous Two-Step Activation for Fusion Systems with PyNE", "creators": [ @@ -13437,16 +13507,21 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + }, + { + "tag": "product" + } + ], "collections": [ - "UKXV4KID", - "4MDZ29N8", - "H442QZRN", - "CMA2SK5V" + "APMMJXES", + "UKXV4KID" ], "relations": {}, "dateAdded": "2015-07-09T20:27:36Z", - "dateModified": "2017-01-11T04:03:41Z" + "dateModified": "2024-09-08T17:38:42Z" } }, { @@ -15259,8 +15334,8 @@ } }, { - "key": "UM2DP2WJ", - "version": 21760, + "key": "WKPH3AP6", + "version": 30781, "library": { "type": "group", "id": 10058, @@ -15274,28 +15349,28 @@ }, "links": { "self": { - "href": "https://api.zotero.org/groups/10058/items/UM2DP2WJ", + "href": "https://api.zotero.org/groups/10058/items/WKPH3AP6", "type": "application/json" }, "alternate": { - "href": "https://www.zotero.org/groups/10058/items/UM2DP2WJ", + "href": "https://www.zotero.org/groups/10058/items/WKPH3AP6", "type": "text/html" }, "attachment": { - "href": "https://api.zotero.org/groups/10058/items/RWQ34QDW", + "href": "https://api.zotero.org/groups/10058/items/62VDJJT6", "type": "application/json", "attachmentType": "application/pdf", - "attachmentSize": 733265 + "attachmentSize": 663460 } }, "meta": { "createdByUser": { - "id": 162605, - "username": "kldunn", - "name": "", + "id": 708524, + "username": "erelson", + "name": "Eric Relson", "links": { "alternate": { - "href": "https://www.zotero.org/kldunn", + "href": "https://www.zotero.org/erelson", "type": "text/html" } } @@ -15311,29 +15386,34 @@ } } }, - "creatorSummary": "Dunn and Wilson", + "creatorSummary": "Relson et al.", "parsedDate": "2013-05-05", - "numChildren": 1 + "numChildren": 2 }, - "bibtex": "\n@inproceedings{dunn_monte_2013,\n\taddress = {Sun Valley, ID},\n\ttitle = {Monte {Carlo} {Mesh} {Tallies} based on a {Kernel} {Density} {Estimator} {Approach} using {Integrated} {Particle} {Tracks}},\n\tbooktitle = {M\\&{C} 2013},\n\tpublisher = {American Nuclear Society},\n\tauthor = {Dunn, K. L. and Wilson, P. H.},\n\tmonth = may,\n\tyear = {2013},\n}\n", + "bibtex": "\n@inproceedings{relson_improved_2013,\n\taddress = {Sun Valley, ID},\n\ttitle = {Improved {Mesh} {Based} {Photon} {Sampling} {Techniques} {For} {Neutron} {Activation} {Analysis}},\n\tabstract = {The design of fusion power systems requires analysis of neutron activation of large, complex volumes, and the resulting particles emitted from these volumes. Structured mesh-based discretization of these problems allows for improved modeling in these activation analysis problems. Finer discretization of these problems results in large computational costs, which drives the investigation of more efficient methods. Within an ad hoc subroutine of the Monte Carlo transport code MCNP, we implement sampling of voxels and photon energies for volumetric sources using the alias method. The alias method enables efficient sampling of a discrete probability distribution, and operates in O(1) time, whereas the simpler direct discrete method requires O(log(n)) time. By using the alias method, voxel sampling becomes a viable alternative to sampling space with the O(1) approach of uniformly sampling the problem volume. Additionally, with voxel sampling it is straightforward to introduce biasing of volumetric sources,\nand we implement this biasing of voxels as an additional variance reduction technique that can be applied. We verify our implementation and compare the alias method, with and without biasing, to direct discrete sampling of voxels, and to uniform sampling. We study the behavior of source biasing in a second set of tests and find trends between improvements and source shape, material, and material density. Overall, however, the magnitude of improvements from source biasing appears to be limited. Future work will benefit from the implementation of efficient voxel sampling – particularly with conformal unstructured meshes where the uniform sampling approach cannot be applied.},\n\tbooktitle = {M\\&{C} 2013},\n\tpublisher = {American Nuclear Society},\n\tauthor = {Relson, E. and Wilson, P.P.H. and Biondo, Elliott},\n\tmonth = may,\n\tyear = {2013},\n\tkeywords = {ALARA, CNERG:HK20 Final Report, MCNP, Photons, R2S-ACT, Sampling},\n}\n", "data": { - "key": "UM2DP2WJ", - "version": 21760, + "key": "WKPH3AP6", + "version": 30781, "itemType": "conferencePaper", - "title": "Monte Carlo Mesh Tallies based on a Kernel Density Estimator Approach using Integrated Particle Tracks", + "title": "Improved Mesh Based Photon Sampling Techniques For Neutron Activation Analysis", "creators": [ { "creatorType": "author", - "firstName": "K. L.", - "lastName": "Dunn" + "firstName": "E.", + "lastName": "Relson" }, { "creatorType": "author", - "firstName": "P. H.", + "firstName": "P.P.H.", "lastName": "Wilson" + }, + { + "creatorType": "author", + "firstName": "Elliott", + "lastName": "Biondo" } ], - "abstractNote": "", + "abstractNote": "The design of fusion power systems requires analysis of neutron activation of large, complex volumes, and the resulting particles emitted from these volumes. Structured mesh-based discretization of these problems allows for improved modeling in these activation analysis problems. Finer discretization of these problems results in large computational costs, which drives the investigation of more efficient methods. Within an ad hoc subroutine of the Monte Carlo transport code MCNP, we implement sampling of voxels and photon energies for volumetric sources using the alias method. The alias method enables efficient sampling of a discrete probability distribution, and operates in O(1) time, whereas the simpler direct discrete method requires O(log(n)) time. By using the alias method, voxel sampling becomes a viable alternative to sampling space with the O(1) approach of uniformly sampling the problem volume. Additionally, with voxel sampling it is straightforward to introduce biasing of volumetric sources,\nand we implement this biasing of voxels as an additional variance reduction technique that can be applied. We verify our implementation and compare the alias method, with and without biasing, to direct discrete sampling of voxels, and to uniform sampling. We study the behavior of source biasing in a second set of tests and find trends between improvements and source shape, material, and material density. Overall, however, the magnitude of improvements from source biasing appears to be limited. Future work will benefit from the implementation of efficient voxel sampling – particularly with conformal unstructured meshes where the uniform sampling approach cannot be applied.", "date": "May 5-9, 2013", "proceedingsTitle": "M&C 2013", "conferenceName": "International Conference on Mathematics and Computational Methods Applied to Nuclear Science & Engineering (M&C 2013)", @@ -15354,17 +15434,37 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "ALARA" + }, + { + "tag": "CNERG:HK20 Final Report" + }, + { + "tag": "MCNP" + }, + { + "tag": "Photons" + }, + { + "tag": "R2S-ACT" + }, + { + "tag": "Sampling" + } + ], "collections": [ - "UKXV4KID" + "UKXV4KID", + "QNJGHVT4" ], "relations": {}, - "dateAdded": "2013-07-01T15:56:13Z", - "dateModified": "2014-01-15T21:21:58Z" + "dateAdded": "2013-04-04T22:35:53Z", + "dateModified": "2024-09-08T17:38:42Z" } }, { - "key": "WKPH3AP6", + "key": "UM2DP2WJ", "version": 21760, "library": { "type": "group", @@ -15379,28 +15479,28 @@ }, "links": { "self": { - "href": "https://api.zotero.org/groups/10058/items/WKPH3AP6", + "href": "https://api.zotero.org/groups/10058/items/UM2DP2WJ", "type": "application/json" }, "alternate": { - "href": "https://www.zotero.org/groups/10058/items/WKPH3AP6", + "href": "https://www.zotero.org/groups/10058/items/UM2DP2WJ", "type": "text/html" }, "attachment": { - "href": "https://api.zotero.org/groups/10058/items/62VDJJT6", + "href": "https://api.zotero.org/groups/10058/items/RWQ34QDW", "type": "application/json", "attachmentType": "application/pdf", - "attachmentSize": 663460 + "attachmentSize": 733265 } }, "meta": { "createdByUser": { - "id": 708524, - "username": "erelson", - "name": "Eric Relson", + "id": 162605, + "username": "kldunn", + "name": "", "links": { "alternate": { - "href": "https://www.zotero.org/erelson", + "href": "https://www.zotero.org/kldunn", "type": "text/html" } } @@ -15416,34 +15516,29 @@ } } }, - "creatorSummary": "Relson et al.", + "creatorSummary": "Dunn and Wilson", "parsedDate": "2013-05-05", - "numChildren": 2 + "numChildren": 1 }, - "bibtex": "\n@inproceedings{relson_improved_2013,\n\taddress = {Sun Valley, ID},\n\ttitle = {Improved {Mesh} {Based} {Photon} {Sampling} {Techniques} {For} {Neutron} {Activation} {Analysis}},\n\tabstract = {The design of fusion power systems requires analysis of neutron activation of large, complex volumes, and the resulting particles emitted from these volumes. Structured mesh-based discretization of these problems allows for improved modeling in these activation analysis problems. Finer discretization of these problems results in large computational costs, which drives the investigation of more efficient methods. Within an ad hoc subroutine of the Monte Carlo transport code MCNP, we implement sampling of voxels and photon energies for volumetric sources using the alias method. The alias method enables efficient sampling of a discrete probability distribution, and operates in O(1) time, whereas the simpler direct discrete method requires O(log(n)) time. By using the alias method, voxel sampling becomes a viable alternative to sampling space with the O(1) approach of uniformly sampling the problem volume. Additionally, with voxel sampling it is straightforward to introduce biasing of volumetric sources,\nand we implement this biasing of voxels as an additional variance reduction technique that can be applied. We verify our implementation and compare the alias method, with and without biasing, to direct discrete sampling of voxels, and to uniform sampling. We study the behavior of source biasing in a second set of tests and find trends between improvements and source shape, material, and material density. Overall, however, the magnitude of improvements from source biasing appears to be limited. Future work will benefit from the implementation of efficient voxel sampling – particularly with conformal unstructured meshes where the uniform sampling approach cannot be applied.},\n\tbooktitle = {M\\&{C} 2013},\n\tpublisher = {American Nuclear Society},\n\tauthor = {Relson, E. and Wilson, P.P.H. and Biondo, Elliott},\n\tmonth = may,\n\tyear = {2013},\n\tkeywords = {ALARA, MCNP, Photons, R2S-ACT, Sampling},\n}\n", + "bibtex": "\n@inproceedings{dunn_monte_2013,\n\taddress = {Sun Valley, ID},\n\ttitle = {Monte {Carlo} {Mesh} {Tallies} based on a {Kernel} {Density} {Estimator} {Approach} using {Integrated} {Particle} {Tracks}},\n\tbooktitle = {M\\&{C} 2013},\n\tpublisher = {American Nuclear Society},\n\tauthor = {Dunn, K. L. and Wilson, P. H.},\n\tmonth = may,\n\tyear = {2013},\n}\n", "data": { - "key": "WKPH3AP6", + "key": "UM2DP2WJ", "version": 21760, "itemType": "conferencePaper", - "title": "Improved Mesh Based Photon Sampling Techniques For Neutron Activation Analysis", + "title": "Monte Carlo Mesh Tallies based on a Kernel Density Estimator Approach using Integrated Particle Tracks", "creators": [ { "creatorType": "author", - "firstName": "E.", - "lastName": "Relson" + "firstName": "K. L.", + "lastName": "Dunn" }, { "creatorType": "author", - "firstName": "P.P.H.", + "firstName": "P. H.", "lastName": "Wilson" - }, - { - "creatorType": "author", - "firstName": "Elliott", - "lastName": "Biondo" } ], - "abstractNote": "The design of fusion power systems requires analysis of neutron activation of large, complex volumes, and the resulting particles emitted from these volumes. Structured mesh-based discretization of these problems allows for improved modeling in these activation analysis problems. Finer discretization of these problems results in large computational costs, which drives the investigation of more efficient methods. Within an ad hoc subroutine of the Monte Carlo transport code MCNP, we implement sampling of voxels and photon energies for volumetric sources using the alias method. The alias method enables efficient sampling of a discrete probability distribution, and operates in O(1) time, whereas the simpler direct discrete method requires O(log(n)) time. By using the alias method, voxel sampling becomes a viable alternative to sampling space with the O(1) approach of uniformly sampling the problem volume. Additionally, with voxel sampling it is straightforward to introduce biasing of volumetric sources,\nand we implement this biasing of voxels as an additional variance reduction technique that can be applied. We verify our implementation and compare the alias method, with and without biasing, to direct discrete sampling of voxels, and to uniform sampling. We study the behavior of source biasing in a second set of tests and find trends between improvements and source shape, material, and material density. Overall, however, the magnitude of improvements from source biasing appears to be limited. Future work will benefit from the implementation of efficient voxel sampling – particularly with conformal unstructured meshes where the uniform sampling approach cannot be applied.", + "abstractNote": "", "date": "May 5-9, 2013", "proceedingsTitle": "M&C 2013", "conferenceName": "International Conference on Mathematics and Computational Methods Applied to Nuclear Science & Engineering (M&C 2013)", @@ -15464,30 +15559,13 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [ - { - "tag": "ALARA" - }, - { - "tag": "MCNP" - }, - { - "tag": "Photons" - }, - { - "tag": "R2S-ACT" - }, - { - "tag": "Sampling" - } - ], + "tags": [], "collections": [ - "UKXV4KID", - "4MDZ29N8" + "UKXV4KID" ], "relations": {}, - "dateAdded": "2013-04-04T22:35:53Z", - "dateModified": "2014-01-15T21:21:53Z" + "dateAdded": "2013-07-01T15:56:13Z", + "dateModified": "2014-01-15T21:21:58Z" } }, { diff --git a/_data/theses.json b/_data/theses.json index 267f660b..d4524115 100644 --- a/_data/theses.json +++ b/_data/theses.json @@ -1050,7 +1050,7 @@ }, { "key": "6VTPU3C5", - "version": 30664, + "version": 30784, "library": { "type": "group", "id": 10058, @@ -1105,10 +1105,10 @@ "parsedDate": "2016-07-21", "numChildren": 2 }, - "bibtex": "\n@phdthesis{biondo_hybrid_2016,\n\taddress = {Madison, WI, United States},\n\ttype = {{PhD} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Hybrid {Monte} {Carlo}/{Deterministic} {Neutron} {Transport} for {Shutdown} {Dose} {Rate} {Analysis}},\n\turl = {https://digital.library.wisc.edu/1711.dl/5DMMVSMDW5LUU9A},\n\tabstract = {In fusion energy systems (FES) neutrons are born from a burning plasma and subsequently activate surrounding system components. The photon dose rate after shutdown from the resultant radionuclides must be quantified for maintenance planning. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for this purpose. This requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work one such formulation is introduced which is valid when a specific set of transmutation criteria are met, referred to as the Single Neutron Interaction and Low Burnup (SNILB) criteria. These criteria are quantitatively evaluated for typical FES scenarios and are shown to be met within a reasonable margin. Groupwise Transmutation (GT)-CADIS, proposed here, is an implementation of MS-CADIS that calculates this adjoint neutron source using a series of irradiation calculations. For a simple SDR problem, GT-CADIS provides speedups of 200 ± 100 relative to global variance reduction with the Forward Weighted (FW)-CADIS method and 90,000 ± 50,000 relative to analog. When the SNILB criteria are egregiously violated, GT-CADIS modifications are proposed and are shown to provide significant performance improvements. Finally, GT-CADIS is applied to a production-level problem involving a Spherical Tokamak Fusion Nuclear Science Facility (ST-FNSF) device. This work shows that GT-CADIS is broadly applicable to FES scenarios and will significantly reduce the computational resources necessary for SDR analysis.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Biondo, Elliott Dean},\n\tmonth = jul,\n\tyear = {2016},\n}\n", + "bibtex": "\n@phdthesis{biondo_hybrid_2016,\n\taddress = {Madison, WI, United States},\n\ttype = {{PhD} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Hybrid {Monte} {Carlo}/{Deterministic} {Neutron} {Transport} for {Shutdown} {Dose} {Rate} {Analysis}},\n\turl = {https://digital.library.wisc.edu/1711.dl/5DMMVSMDW5LUU9A},\n\tabstract = {In fusion energy systems (FES) neutrons are born from a burning plasma and subsequently activate surrounding system components. The photon dose rate after shutdown from the resultant radionuclides must be quantified for maintenance planning. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for this purpose. This requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work one such formulation is introduced which is valid when a specific set of transmutation criteria are met, referred to as the Single Neutron Interaction and Low Burnup (SNILB) criteria. These criteria are quantitatively evaluated for typical FES scenarios and are shown to be met within a reasonable margin. Groupwise Transmutation (GT)-CADIS, proposed here, is an implementation of MS-CADIS that calculates this adjoint neutron source using a series of irradiation calculations. For a simple SDR problem, GT-CADIS provides speedups of 200 ± 100 relative to global variance reduction with the Forward Weighted (FW)-CADIS method and 90,000 ± 50,000 relative to analog. When the SNILB criteria are egregiously violated, GT-CADIS modifications are proposed and are shown to provide significant performance improvements. Finally, GT-CADIS is applied to a production-level problem involving a Spherical Tokamak Fusion Nuclear Science Facility (ST-FNSF) device. This work shows that GT-CADIS is broadly applicable to FES scenarios and will significantly reduce the computational resources necessary for SDR analysis.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Biondo, Elliott Dean},\n\tmonth = jul,\n\tyear = {2016},\n\tkeywords = {CNERG:HK20 Final Report, product},\n}\n", "data": { "key": "6VTPU3C5", - "version": 30664, + "version": 30784, "itemType": "thesis", "title": "Hybrid Monte Carlo/Deterministic Neutron Transport for Shutdown Dose Rate Analysis", "creators": [ @@ -1134,17 +1134,21 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + }, + { + "tag": "product" + } + ], "collections": [ "6259B6TV", - "4MDZ29N8", - "H442QZRN", - "CMA2SK5V", "34I86HPD" ], "relations": {}, "dateAdded": "2016-08-20T14:20:24Z", - "dateModified": "2024-09-08T16:49:48Z" + "dateModified": "2024-09-08T17:38:42Z" } }, { @@ -1596,7 +1600,7 @@ }, { "key": "GN6PK84Z", - "version": 26427, + "version": 30781, "library": { "type": "group", "id": 10058, @@ -1640,10 +1644,10 @@ "parsedDate": "2013-07", "numChildren": 2 }, - "bibtex": "\n@phdthesis{relson_improved_2013,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Improved {Methods} {For} {Sampling} {Mesh}-{Based} {Volumetric} {Sources} {In} {Monte} {Carlo} {Transport}},\n\tabstract = {This research focuses on developing mesh-based techniques for sampling distributed,\n volumetric sources in Monte Carlo particle transport codes, such as MCNP. This work\n culminated in several source sampling techniques being implemented within a 3-D neutron\n activation workflow.\n\nThe most significant development is the implementation of an efficient voxel sampling\n technique. Voxel sampling can be applied to source meshes with any number of mesh\n elements thanks to efficient sampling via the alias method, and meshing of non-source volumes\n can be avoided. Voxel sampling in turn enables straight-forward implementation of source\n biasing for variance reduction, and also the use of unstructured source meshes using\n tetrahedral mesh elements. The uniform sampling technique used in past work is effectively a\n biasing scheme, and thus can be implemented more efficiently with biased voxel sampling.\n\nFor this work, the source meshes are inherited from neutron mesh tallies. Cartesian\n structured meshes, which provide straight-forward compatibility with legacy tools can be\n sampled with either the voxel or uniform sampling methods. Alternately, using an unstructured\n mesh (via the unstructured mesh tally capabilities in DAG-MCNP) allows for better conforming\n meshes – particularly with geometries that do not align well with a structured mesh, or where\n the source region is spread out through a region of non-source materials, such as systems of\n pipes.\n\nThe set of source sampling techniques is useful as a toolkit for obtaining quality answers\n from a variety of scenarios. This thesis supplements methods development and\n implementation with experiments to identify and understand which sampling techniques\n should be used in different scenarios. The new sampling methods and workflows are shown to\n be in good agreement with results from older methods. While there remain several aspects of\n the new methods’ behavior to characterize, voxel sampling and its derivatives have fully\n replaced older sampling methods in neutron activation analysis work at UW-Madison.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Relson, Eric},\n\tmonth = jul,\n\tyear = {2013},\n}\n", + "bibtex": "\n@phdthesis{relson_improved_2013,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Improved {Methods} {For} {Sampling} {Mesh}-{Based} {Volumetric} {Sources} {In} {Monte} {Carlo} {Transport}},\n\tabstract = {This research focuses on developing mesh-based techniques for sampling distributed,\n volumetric sources in Monte Carlo particle transport codes, such as MCNP. This work\n culminated in several source sampling techniques being implemented within a 3-D neutron\n activation workflow.\n\nThe most significant development is the implementation of an efficient voxel sampling\n technique. Voxel sampling can be applied to source meshes with any number of mesh\n elements thanks to efficient sampling via the alias method, and meshing of non-source volumes\n can be avoided. Voxel sampling in turn enables straight-forward implementation of source\n biasing for variance reduction, and also the use of unstructured source meshes using\n tetrahedral mesh elements. The uniform sampling technique used in past work is effectively a\n biasing scheme, and thus can be implemented more efficiently with biased voxel sampling.\n\nFor this work, the source meshes are inherited from neutron mesh tallies. Cartesian\n structured meshes, which provide straight-forward compatibility with legacy tools can be\n sampled with either the voxel or uniform sampling methods. Alternately, using an unstructured\n mesh (via the unstructured mesh tally capabilities in DAG-MCNP) allows for better conforming\n meshes – particularly with geometries that do not align well with a structured mesh, or where\n the source region is spread out through a region of non-source materials, such as systems of\n pipes.\n\nThe set of source sampling techniques is useful as a toolkit for obtaining quality answers\n from a variety of scenarios. This thesis supplements methods development and\n implementation with experiments to identify and understand which sampling techniques\n should be used in different scenarios. The new sampling methods and workflows are shown to\n be in good agreement with results from older methods. While there remain several aspects of\n the new methods’ behavior to characterize, voxel sampling and its derivatives have fully\n replaced older sampling methods in neutron activation analysis work at UW-Madison.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Relson, Eric},\n\tmonth = jul,\n\tyear = {2013},\n\tkeywords = {CNERG:HK20 Final Report},\n}\n", "data": { "key": "GN6PK84Z", - "version": 26427, + "version": 30781, "itemType": "thesis", "title": "Improved Methods For Sampling Mesh-Based Volumetric Sources In Monte Carlo Transport", "creators": [ @@ -1669,20 +1673,23 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + } + ], "collections": [ "6259B6TV", - "4MDZ29N8", "Y4UI9B4X" ], "relations": {}, "dateAdded": "2013-11-10T16:12:41Z", - "dateModified": "2013-11-10T16:15:41Z" + "dateModified": "2024-09-08T17:38:42Z" } }, { "key": "4BD4NW6X", - "version": 30671, + "version": 30782, "library": { "type": "group", "id": 10058, @@ -1726,10 +1733,10 @@ "parsedDate": "2012-06", "numChildren": 2 }, - "bibtex": "\n@phdthesis{ibrahim_automatic_2012,\n\taddress = {Madison, WI, United States},\n\ttype = {{PhD} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Automatic {Mesh} {Adaptivity} for {Hybrid} {Monte} {Carlo}/{Deterministic} {Neutronics} {Modeling} of {Difficult} {Shielding} {Problems}},\n\turl = {https://digital.library.wisc.edu/1711.dl/7KMFF4JZJEU4S87},\n\tabstract = {Over the last decade, the role of neutronics modeling has been shifting from analysis of\neach component separately to high fidelity, full-scale analysis of the nuclear systems entire\ndomains. The high accuracy, associated with minimizing modeling approximations and including\nmore physical and geometric details, is now feasible because of advancements in computing\nhardware and development of efficient modeling methods. The hybrid Monte Carlo/deterministic\ntechniques, CADIS and FW-CADIS dramatically increase the efficiency of neutronics modeling,\nbut their use in the design of large and geometrically complex nuclear systems is restricted by the\navailability of computing resources for their preliminarily deterministic calculations and the\nlarge computer memory requirements of their final Monte Carlo calculations.\nTo reduce the computational time and memory requirements of the hybrid Monte\nCarlo/deterministic techniques while maintaining their efficiency improvements, three automatic\nmesh adaptivity algorithms were developed and added to the Oak Ridge National Laboratory\nAutomateD VAriaNce reducTion Generator (ADVANTG) code. First, a mixed-material\napproach, which we refer to as the macromaterial approach, enhances the fidelity of the\ndeterministic models without having to refine the mesh of the deterministic calculations. Second,\na deterministic mesh refinement algorithm improves the accuracy of structured mesh\ndeterministic calculations by capturing as much geometric detail as possible without exceeding\nthe total number of mesh elements that is usually determined by the availability of computing\nresources. Finally, a weight window coarsening algorithm decouples the weight window mesh\nfrom the mesh of the deterministic calculations to remove the memory constraint of the weight\nwindow map from the deterministic mesh resolution.\nii\nTo analyze the combined effect of the three algorithms developed in this thesis, they were\nused to calculate the prompt dose rate throughout the entire ITER experimental facility. This\ncalculation represents a very challenging shielding problem because of the immense size and\ncomplexity of the ITER structure and the presence of a two meter thick biological shield.\nCompared to a FW-CADIS calculation with the same storage size of the variance reduction\nparameters, the use of the three algorithms resulted in a 23.3\\% increase in the regions where the\ndose rate results are achieved in a 10 day Monte Carlo calculation and increased the efficiency of\nthe Monte Carlo simulation by a factor of 3.4. Because of this significant increase in the Monte\nCarlo efficiency which was not accompanied by an increase in the memory requirements, the use\nof the three algorithms in FW-CADIS simulations enabled the simulation of this difficult\nshielding problem on a regular computer cluster using parallel processing of Monte Carlo\ncalculations. The results of the parallel Monte Carlo calculation agreed at four points with a very\nfine mesh deterministic calculation that was performed on the super-computer, Jaguar.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Ibrahim, Ahmad},\n\tmonth = jun,\n\tyear = {2012},\n}\n", + "bibtex": "\n@phdthesis{ibrahim_automatic_2012,\n\taddress = {Madison, WI, United States},\n\ttype = {{PhD} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Automatic {Mesh} {Adaptivity} for {Hybrid} {Monte} {Carlo}/{Deterministic} {Neutronics} {Modeling} of {Difficult} {Shielding} {Problems}},\n\turl = {https://digital.library.wisc.edu/1711.dl/7KMFF4JZJEU4S87},\n\tabstract = {Over the last decade, the role of neutronics modeling has been shifting from analysis of\neach component separately to high fidelity, full-scale analysis of the nuclear systems entire\ndomains. The high accuracy, associated with minimizing modeling approximations and including\nmore physical and geometric details, is now feasible because of advancements in computing\nhardware and development of efficient modeling methods. The hybrid Monte Carlo/deterministic\ntechniques, CADIS and FW-CADIS dramatically increase the efficiency of neutronics modeling,\nbut their use in the design of large and geometrically complex nuclear systems is restricted by the\navailability of computing resources for their preliminarily deterministic calculations and the\nlarge computer memory requirements of their final Monte Carlo calculations.\nTo reduce the computational time and memory requirements of the hybrid Monte\nCarlo/deterministic techniques while maintaining their efficiency improvements, three automatic\nmesh adaptivity algorithms were developed and added to the Oak Ridge National Laboratory\nAutomateD VAriaNce reducTion Generator (ADVANTG) code. First, a mixed-material\napproach, which we refer to as the macromaterial approach, enhances the fidelity of the\ndeterministic models without having to refine the mesh of the deterministic calculations. Second,\na deterministic mesh refinement algorithm improves the accuracy of structured mesh\ndeterministic calculations by capturing as much geometric detail as possible without exceeding\nthe total number of mesh elements that is usually determined by the availability of computing\nresources. Finally, a weight window coarsening algorithm decouples the weight window mesh\nfrom the mesh of the deterministic calculations to remove the memory constraint of the weight\nwindow map from the deterministic mesh resolution.\nii\nTo analyze the combined effect of the three algorithms developed in this thesis, they were\nused to calculate the prompt dose rate throughout the entire ITER experimental facility. This\ncalculation represents a very challenging shielding problem because of the immense size and\ncomplexity of the ITER structure and the presence of a two meter thick biological shield.\nCompared to a FW-CADIS calculation with the same storage size of the variance reduction\nparameters, the use of the three algorithms resulted in a 23.3\\% increase in the regions where the\ndose rate results are achieved in a 10 day Monte Carlo calculation and increased the efficiency of\nthe Monte Carlo simulation by a factor of 3.4. Because of this significant increase in the Monte\nCarlo efficiency which was not accompanied by an increase in the memory requirements, the use\nof the three algorithms in FW-CADIS simulations enabled the simulation of this difficult\nshielding problem on a regular computer cluster using parallel processing of Monte Carlo\ncalculations. The results of the parallel Monte Carlo calculation agreed at four points with a very\nfine mesh deterministic calculation that was performed on the super-computer, Jaguar.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Ibrahim, Ahmad},\n\tmonth = jun,\n\tyear = {2012},\n\tkeywords = {CNERG:HK20 Final Report},\n}\n", "data": { "key": "4BD4NW6X", - "version": 30671, + "version": 30782, "itemType": "thesis", "title": "Automatic Mesh Adaptivity for Hybrid Monte Carlo/Deterministic Neutronics Modeling of Difficult Shielding Problems", "creators": [ @@ -1755,16 +1762,18 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + } + ], "collections": [ "6259B6TV", - "4MDZ29N8", - "RI2DQ3B2", "34I86HPD" ], "relations": {}, "dateAdded": "2013-01-21T21:43:04Z", - "dateModified": "2024-09-08T16:51:31Z" + "dateModified": "2024-09-08T17:38:42Z" } }, { @@ -1861,7 +1870,7 @@ }, { "key": "7CCN265G", - "version": 26427, + "version": 30781, "library": { "type": "group", "id": 10058, @@ -1905,10 +1914,10 @@ "parsedDate": "2011-08", "numChildren": 3 }, - "bibtex": "\n@phdthesis{snouffer_validation_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Validation and {Verification} of {Direct} {Accelerated} {Geometry} {Monte} {Carlo}},\n\tabstract = {As both Monte Carlo radiation transport codes and 3D CAD modeling become more widely used, there has been an increasing number of efforts to combine these tools. One such effort is the Direct Accelerated Geometry Monte Carlo (DAGMC) software package being developed at the University of Wisconsin-Madison . DAGMC performs the particle tracking needed for\nMonte Carlo radiation transport code directly on CAD geometries. DAGMC has been in development for a number of years and is in need of validation and verification in order to build user confidence in DAGMC's reliability and accuracy.\n\nThis work performs extensive testing of DAGMC implemented with the radiation transport code Monte Carlo N-Particle 5 (DAG-MCNP5). Four tests suites have been compiled for DAG-MCNP5 to ensure the accuracy of the code now and for future developers. These test suites are based largely on the test suites for MCNP5 and include: a suite of 80 regression tests, a suite of 75 verification tests, a suite of 30 validation criticality tests, and a suite of 19 validation shielding tests. These tests encompass a wide range of geometries, materials, and physics to test almost all of the features of DAG-MCNP5. The results of these test were compared to both analytical and experimental results, where appropriate, and MCNP5 results. A faceting tolerance study\nwas also performed for many of these test. It was found that a faceting tolerance of not large than 10-4 cm produces statistically similar results to MCNP5 on a consistent basis for all problem types. It is concluded that DAG-MCNP5 performs as accurately as MCNP5 for these test problems, and that DAG-MCNP5 can be considered a reliable neutronics code.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Snouffer, Patrick},\n\tmonth = aug,\n\tyear = {2011},\n}\n", + "bibtex": "\n@phdthesis{snouffer_validation_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Validation and {Verification} of {Direct} {Accelerated} {Geometry} {Monte} {Carlo}},\n\tabstract = {As both Monte Carlo radiation transport codes and 3D CAD modeling become more widely used, there has been an increasing number of efforts to combine these tools. One such effort is the Direct Accelerated Geometry Monte Carlo (DAGMC) software package being developed at the University of Wisconsin-Madison . DAGMC performs the particle tracking needed for\nMonte Carlo radiation transport code directly on CAD geometries. DAGMC has been in development for a number of years and is in need of validation and verification in order to build user confidence in DAGMC's reliability and accuracy.\n\nThis work performs extensive testing of DAGMC implemented with the radiation transport code Monte Carlo N-Particle 5 (DAG-MCNP5). Four tests suites have been compiled for DAG-MCNP5 to ensure the accuracy of the code now and for future developers. These test suites are based largely on the test suites for MCNP5 and include: a suite of 80 regression tests, a suite of 75 verification tests, a suite of 30 validation criticality tests, and a suite of 19 validation shielding tests. These tests encompass a wide range of geometries, materials, and physics to test almost all of the features of DAG-MCNP5. The results of these test were compared to both analytical and experimental results, where appropriate, and MCNP5 results. A faceting tolerance study\nwas also performed for many of these test. It was found that a faceting tolerance of not large than 10-4 cm produces statistically similar results to MCNP5 on a consistent basis for all problem types. It is concluded that DAG-MCNP5 performs as accurately as MCNP5 for these test problems, and that DAG-MCNP5 can be considered a reliable neutronics code.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Snouffer, Patrick},\n\tmonth = aug,\n\tyear = {2011},\n\tkeywords = {CNERG:HK20 Final Report},\n}\n", "data": { "key": "7CCN265G", - "version": 26427, + "version": 30781, "itemType": "thesis", "title": "Validation and Verification of Direct Accelerated Geometry Monte Carlo", "creators": [ @@ -1934,15 +1943,18 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + } + ], "collections": [ "6259B6TV", - "4MDZ29N8", "Y4UI9B4X" ], "relations": {}, "dateAdded": "2012-12-07T14:29:06Z", - "dateModified": "2012-12-07T14:31:43Z" + "dateModified": "2024-09-08T17:38:42Z" } }, { @@ -2031,8 +2043,8 @@ } }, { - "key": "VDD6RCJQ", - "version": 30683, + "key": "TW3VBHD8", + "version": 30781, "library": { "type": "group", "id": 10058, @@ -2046,22 +2058,33 @@ }, "links": { "self": { - "href": "https://api.zotero.org/groups/10058/items/VDD6RCJQ", + "href": "https://api.zotero.org/groups/10058/items/TW3VBHD8", "type": "application/json" }, "alternate": { - "href": "https://www.zotero.org/groups/10058/items/VDD6RCJQ", + "href": "https://www.zotero.org/groups/10058/items/TW3VBHD8", "type": "text/html" }, "attachment": { - "href": "https://api.zotero.org/groups/10058/items/GUXS9QRA", + "href": "https://api.zotero.org/groups/10058/items/MUUA6GAS", "type": "application/json", "attachmentType": "application/pdf", - "attachmentSize": 8600060 + "attachmentSize": 1351946 } }, "meta": { "createdByUser": { + "id": 708524, + "username": "erelson", + "name": "Eric Relson", + "links": { + "alternate": { + "href": "https://www.zotero.org/erelson", + "type": "text/html" + } + } + }, + "lastModifiedByUser": { "id": 112658, "username": "gonuke", "name": "", @@ -2072,32 +2095,32 @@ } } }, - "creatorSummary": "Smith", + "creatorSummary": "Moule", "parsedDate": "2011", "numChildren": 2 }, - "bibtex": "\n@phdthesis{smith_robust_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{PhD} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Robust {Tracking} and {Advanced} {Geometry} for {Monte} {Carlo} {Radiation} {Transport}},\n\turl = {https://www.proquest.com/pqdtglobal/docview/885040837/abstract/49781E1E9AC841E4PQ/1?sourcetype=Dissertations%20&%20Theses},\n\tabstract = {A set of improved geometric capabilities are developed for the Direct Accelerated Geometry for Monte Carlo (DAGMC) library to increase its ease of use and accuracy. The improvements are watertight faceting, robust particle tracking, automatic creation of nonsolid space, and overlap tolerance. Before being sealed, adjacent faceted surfaces do not have the same discretization along shared curves. Sealing together surfaces to create a watertight faceting prevents leakage of particles between surfaces. The tracking algorithm is made robust by ensuring numerical consistency and avoiding geometric tolerances. Monte Carlo simulation requires all space to be defined, whether it be vacuum, air, coolant, or a solid material. The implicit creation of nonsolid space reduces human effort otherwise required to explicitly create nonsolid space in a CAD program. CAD models often contain small gaps and overlaps between adjacent volumes due to imprecise modeling, file translation, or intentional deformation. Although gaps are filled by the implicit creation of nonsolid space, overlaps cause geometric queries to become unreliable. The particle tracking algorithm and point inclusion test are modified to tolerate small overlaps of adjacent volumes. Overlap-tolerant particle tracking eliminates manual repair of CAD models and enables analysis of meshed finite element models undergoing structural deformation. These improvements are implemented in a coupling of DAGMC with the Monte Carlo N-Particle (MCNP) code, known as DAG-MCNP. The elimination of both manual CAD repair and lost particles are demonstrated with CAD models used in production calculations.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Smith, Brandon M.},\n\tyear = {2011},\n}\n", + "bibtex": "\n@phdthesis{moule_sampling_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Sampling {Material} {Composition} of {CAD} {Geometries}},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Moule, Damien},\n\tyear = {2011},\n\tkeywords = {CNERG:HK20 Final Report},\n}\n", "data": { - "key": "VDD6RCJQ", - "version": 30683, + "key": "TW3VBHD8", + "version": 30781, "itemType": "thesis", - "title": "Robust Tracking and Advanced Geometry for Monte Carlo Radiation Transport", + "title": "Sampling Material Composition of CAD Geometries", "creators": [ { "creatorType": "author", - "firstName": "Brandon M.", - "lastName": "Smith" + "firstName": "Damien", + "lastName": "Moule" } ], - "abstractNote": "A set of improved geometric capabilities are developed for the Direct Accelerated Geometry for Monte Carlo (DAGMC) library to increase its ease of use and accuracy. The improvements are watertight faceting, robust particle tracking, automatic creation of nonsolid space, and overlap tolerance. Before being sealed, adjacent faceted surfaces do not have the same discretization along shared curves. Sealing together surfaces to create a watertight faceting prevents leakage of particles between surfaces. The tracking algorithm is made robust by ensuring numerical consistency and avoiding geometric tolerances. Monte Carlo simulation requires all space to be defined, whether it be vacuum, air, coolant, or a solid material. The implicit creation of nonsolid space reduces human effort otherwise required to explicitly create nonsolid space in a CAD program. CAD models often contain small gaps and overlaps between adjacent volumes due to imprecise modeling, file translation, or intentional deformation. Although gaps are filled by the implicit creation of nonsolid space, overlaps cause geometric queries to become unreliable. The particle tracking algorithm and point inclusion test are modified to tolerate small overlaps of adjacent volumes. Overlap-tolerant particle tracking eliminates manual repair of CAD models and enables analysis of meshed finite element models undergoing structural deformation. These improvements are implemented in a coupling of DAGMC with the Monte Carlo N-Particle (MCNP) code, known as DAG-MCNP. The elimination of both manual CAD repair and lost particles are demonstrated with CAD models used in production calculations.", - "thesisType": "PhD Nuclear Engineering and Engineering Physics", + "abstractNote": "", + "thesisType": "MS Nuclear Engineering and Engineering Physics", "university": "University of Wisconsin-Madison", "place": "Madison, WI, United States", "date": "2011", - "numPages": "145", + "numPages": "71", "language": "English", "shortTitle": "", - "url": "https://www.proquest.com/pqdtglobal/docview/885040837/abstract/49781E1E9AC841E4PQ/1?sourcetype=Dissertations%20&%20Theses", + "url": "", "accessDate": "", "archive": "", "archiveLocation": "", @@ -2105,20 +2128,23 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + } + ], "collections": [ "6259B6TV", - "4MDZ29N8", - "34I86HPD" + "Y4UI9B4X" ], "relations": {}, - "dateAdded": "2012-12-07T14:36:11Z", - "dateModified": "2024-09-08T16:57:56Z" + "dateAdded": "2013-02-07T21:51:45Z", + "dateModified": "2024-09-08T17:38:42Z" } }, { - "key": "5WKCU3AF", - "version": 26428, + "key": "VDD6RCJQ", + "version": 30781, "library": { "type": "group", "id": 10058, @@ -2132,12 +2158,18 @@ }, "links": { "self": { - "href": "https://api.zotero.org/groups/10058/items/5WKCU3AF", + "href": "https://api.zotero.org/groups/10058/items/VDD6RCJQ", "type": "application/json" }, "alternate": { - "href": "https://www.zotero.org/groups/10058/items/5WKCU3AF", + "href": "https://www.zotero.org/groups/10058/items/VDD6RCJQ", "type": "text/html" + }, + "attachment": { + "href": "https://api.zotero.org/groups/10058/items/GUXS9QRA", + "type": "application/json", + "attachmentType": "application/pdf", + "attachmentSize": 8600060 } }, "meta": { @@ -2152,32 +2184,32 @@ } } }, - "creatorSummary": "Nygaard", + "creatorSummary": "Smith", "parsedDate": "2011", - "numChildren": 0 + "numChildren": 2 }, - "bibtex": "\n@phdthesis{nygaard_notitle_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Nygaard, Erik},\n\tyear = {2011},\n}\n", + "bibtex": "\n@phdthesis{smith_robust_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{PhD} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Robust {Tracking} and {Advanced} {Geometry} for {Monte} {Carlo} {Radiation} {Transport}},\n\turl = {https://www.proquest.com/pqdtglobal/docview/885040837/abstract/49781E1E9AC841E4PQ/1?sourcetype=Dissertations%20&%20Theses},\n\tabstract = {A set of improved geometric capabilities are developed for the Direct Accelerated Geometry for Monte Carlo (DAGMC) library to increase its ease of use and accuracy. The improvements are watertight faceting, robust particle tracking, automatic creation of nonsolid space, and overlap tolerance. Before being sealed, adjacent faceted surfaces do not have the same discretization along shared curves. Sealing together surfaces to create a watertight faceting prevents leakage of particles between surfaces. The tracking algorithm is made robust by ensuring numerical consistency and avoiding geometric tolerances. Monte Carlo simulation requires all space to be defined, whether it be vacuum, air, coolant, or a solid material. The implicit creation of nonsolid space reduces human effort otherwise required to explicitly create nonsolid space in a CAD program. CAD models often contain small gaps and overlaps between adjacent volumes due to imprecise modeling, file translation, or intentional deformation. Although gaps are filled by the implicit creation of nonsolid space, overlaps cause geometric queries to become unreliable. The particle tracking algorithm and point inclusion test are modified to tolerate small overlaps of adjacent volumes. Overlap-tolerant particle tracking eliminates manual repair of CAD models and enables analysis of meshed finite element models undergoing structural deformation. These improvements are implemented in a coupling of DAGMC with the Monte Carlo N-Particle (MCNP) code, known as DAG-MCNP. The elimination of both manual CAD repair and lost particles are demonstrated with CAD models used in production calculations.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Smith, Brandon M.},\n\tyear = {2011},\n\tkeywords = {CNERG:HK20 Final Report},\n}\n", "data": { - "key": "5WKCU3AF", - "version": 26428, + "key": "VDD6RCJQ", + "version": 30781, "itemType": "thesis", - "title": "", + "title": "Robust Tracking and Advanced Geometry for Monte Carlo Radiation Transport", "creators": [ { "creatorType": "author", - "firstName": "Erik", - "lastName": "Nygaard" + "firstName": "Brandon M.", + "lastName": "Smith" } ], - "abstractNote": "", - "thesisType": "MS Nuclear Engineering and Engineering Physics", + "abstractNote": "A set of improved geometric capabilities are developed for the Direct Accelerated Geometry for Monte Carlo (DAGMC) library to increase its ease of use and accuracy. The improvements are watertight faceting, robust particle tracking, automatic creation of nonsolid space, and overlap tolerance. Before being sealed, adjacent faceted surfaces do not have the same discretization along shared curves. Sealing together surfaces to create a watertight faceting prevents leakage of particles between surfaces. The tracking algorithm is made robust by ensuring numerical consistency and avoiding geometric tolerances. Monte Carlo simulation requires all space to be defined, whether it be vacuum, air, coolant, or a solid material. The implicit creation of nonsolid space reduces human effort otherwise required to explicitly create nonsolid space in a CAD program. CAD models often contain small gaps and overlaps between adjacent volumes due to imprecise modeling, file translation, or intentional deformation. Although gaps are filled by the implicit creation of nonsolid space, overlaps cause geometric queries to become unreliable. The particle tracking algorithm and point inclusion test are modified to tolerate small overlaps of adjacent volumes. Overlap-tolerant particle tracking eliminates manual repair of CAD models and enables analysis of meshed finite element models undergoing structural deformation. These improvements are implemented in a coupling of DAGMC with the Monte Carlo N-Particle (MCNP) code, known as DAG-MCNP. The elimination of both manual CAD repair and lost particles are demonstrated with CAD models used in production calculations.", + "thesisType": "PhD Nuclear Engineering and Engineering Physics", "university": "University of Wisconsin-Madison", "place": "Madison, WI, United States", "date": "2011", - "numPages": "", - "language": "", + "numPages": "145", + "language": "English", "shortTitle": "", - "url": "", + "url": "https://www.proquest.com/pqdtglobal/docview/885040837/abstract/49781E1E9AC841E4PQ/1?sourcetype=Dissertations%20&%20Theses", "accessDate": "", "archive": "", "archiveLocation": "", @@ -2185,19 +2217,23 @@ "callNumber": "", "rights": "", "extra": "", - "tags": [], + "tags": [ + { + "tag": "CNERG:HK20 Final Report" + } + ], "collections": [ "6259B6TV", - "Y4UI9B4X" + "34I86HPD" ], "relations": {}, - "dateAdded": "2016-11-27T21:57:05Z", - "dateModified": "2020-12-30T15:03:12Z" + "dateAdded": "2012-12-07T14:36:11Z", + "dateModified": "2024-09-08T17:38:42Z" } }, { - "key": "TW3VBHD8", - "version": 26427, + "key": "5WKCU3AF", + "version": 26428, "library": { "type": "group", "id": 10058, @@ -2211,33 +2247,16 @@ }, "links": { "self": { - "href": "https://api.zotero.org/groups/10058/items/TW3VBHD8", + "href": "https://api.zotero.org/groups/10058/items/5WKCU3AF", "type": "application/json" }, "alternate": { - "href": "https://www.zotero.org/groups/10058/items/TW3VBHD8", + "href": "https://www.zotero.org/groups/10058/items/5WKCU3AF", "type": "text/html" - }, - "attachment": { - "href": "https://api.zotero.org/groups/10058/items/MUUA6GAS", - "type": "application/json", - "attachmentType": "application/pdf", - "attachmentSize": 1351946 } }, "meta": { "createdByUser": { - "id": 708524, - "username": "erelson", - "name": "Eric Relson", - "links": { - "alternate": { - "href": "https://www.zotero.org/erelson", - "type": "text/html" - } - } - }, - "lastModifiedByUser": { "id": 112658, "username": "gonuke", "name": "", @@ -2248,21 +2267,21 @@ } } }, - "creatorSummary": "Moule", + "creatorSummary": "Nygaard", "parsedDate": "2011", - "numChildren": 2 + "numChildren": 0 }, - "bibtex": "\n@phdthesis{moule_sampling_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Sampling {Material} {Composition} of {CAD} {Geometries}},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Moule, Damien},\n\tyear = {2011},\n}\n", + "bibtex": "\n@phdthesis{nygaard_notitle_2011,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Nygaard, Erik},\n\tyear = {2011},\n}\n", "data": { - "key": "TW3VBHD8", - "version": 26427, + "key": "5WKCU3AF", + "version": 26428, "itemType": "thesis", - "title": "Sampling Material Composition of CAD Geometries", + "title": "", "creators": [ { "creatorType": "author", - "firstName": "Damien", - "lastName": "Moule" + "firstName": "Erik", + "lastName": "Nygaard" } ], "abstractNote": "", @@ -2270,8 +2289,8 @@ "university": "University of Wisconsin-Madison", "place": "Madison, WI, United States", "date": "2011", - "numPages": "71", - "language": "English", + "numPages": "", + "language": "", "shortTitle": "", "url": "", "accessDate": "", @@ -2284,13 +2303,11 @@ "tags": [], "collections": [ "6259B6TV", - "4MDZ29N8", - "RI2DQ3B2", "Y4UI9B4X" ], "relations": {}, - "dateAdded": "2013-02-07T21:51:45Z", - "dateModified": "2013-11-10T16:19:44Z" + "dateAdded": "2016-11-27T21:57:05Z", + "dateModified": "2020-12-30T15:03:12Z" } }, { @@ -2379,8 +2396,8 @@ } }, { - "key": "23P9KT43", - "version": 30686, + "key": "NGJ8UC47", + "version": 30758, "library": { "type": "group", "id": 10058, @@ -2394,22 +2411,33 @@ }, "links": { "self": { - "href": "https://api.zotero.org/groups/10058/items/23P9KT43", + "href": "https://api.zotero.org/groups/10058/items/NGJ8UC47", "type": "application/json" }, "alternate": { - "href": "https://www.zotero.org/groups/10058/items/23P9KT43", + "href": "https://www.zotero.org/groups/10058/items/NGJ8UC47", "type": "text/html" }, "attachment": { - "href": "https://api.zotero.org/groups/10058/items/IEPH8GF8", + "href": "https://api.zotero.org/groups/10058/items/GED9HPK8", "type": "application/json", "attachmentType": "application/pdf", - "attachmentSize": 303098 + "attachmentSize": 2141571 } }, "meta": { "createdByUser": { + "id": 144819, + "username": "gidden", + "name": "", + "links": { + "alternate": { + "href": "https://www.zotero.org/gidden", + "type": "text/html" + } + } + }, + "lastModifiedByUser": { "id": 112658, "username": "gonuke", "name": "", @@ -2420,33 +2448,33 @@ } } }, - "creatorSummary": "Kiedrowski", + "creatorSummary": "Oliver", "parsedDate": "2009", - "numChildren": 4 + "numChildren": 6 }, - "bibtex": "\n@phdthesis{kiedrowski_adjoint_2009,\n\taddress = {Madison, WI, United States},\n\ttype = {{PhD} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {Adjoint {Weighting} for {Continuous}-{Energy} {Monte} {Carlo} {Radiation} {Transport}},\n\turl = {https://www.proquest.com/pqdtglobal/docview/305033185/abstract/A62FE598B9FC46AEPQ/1?sourcetype=Dissertations%20&%20Theses},\n\tabstract = {Methods are developed for importance or adjoint weighting of individual tally scores within a continuous-energy k-eigenvalue Monte Carlo calculation. These adjoint-weighted tallies allow for the calculation of certain quantities important to understanding the physics of a nuclear reactor.\n\nThe methods, unlike traditional approaches to computing adjoint-weighted quantities, do not attempt to invert the random walk. Rather, they are based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. This can be calculated in a strictly forward calculation, and this factor can be applied to previously computed tally scores.\n\nThese methods are implemented in a production Monte Carlo code and are used to calculate parameters requiring adjoint weighting, the point reactor kinetics parameters and reactivity changes based upon first-order perturbation theory. The results of these calculations are compared against experimental measurements, equivalent discrete ordinates calculations, or other Monte Carlo based techniques.},\n\tlanguage = {English},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Kiedrowski, Brian C.},\n\tyear = {2009},\n}\n", + "bibtex": "\n@phdthesis{oliver_geniusv2_2009,\n\taddress = {Madison, WI, United States},\n\ttype = {{MS} {Nuclear} {Engineering} and {Engineering} {Physics}},\n\ttitle = {{GENIUSv2}: {Software} {Design} and {Mathematical} {Formulations} for {Multi}-{Region} {Discrete} {Nuclear} {Fuel} {Cycle} {Simulation} and {Analysis}},\n\turl = {http://kyleoliver.net/work/thesis_KMO.pdf},\n\turldate = {2012-11-25},\n\tschool = {University of Wisconsin-Madison},\n\tauthor = {Oliver, Kyle M.},\n\tyear = {2009},\n}\n", "data": { - "key": "23P9KT43", - "version": 30686, + "key": "NGJ8UC47", + "version": 30758, "itemType": "thesis", - "title": "Adjoint Weighting for Continuous-Energy Monte Carlo Radiation Transport", + "title": "GENIUSv2: Software Design and Mathematical Formulations for Multi-Region Discrete Nuclear Fuel Cycle Simulation and Analysis", "creators": [ { "creatorType": "author", - "firstName": "Brian C.", - "lastName": "Kiedrowski" + "firstName": "Kyle M.", + "lastName": "Oliver" } ], - "abstractNote": "Methods are developed for importance or adjoint weighting of individual tally scores within a continuous-energy k-eigenvalue Monte Carlo calculation. These adjoint-weighted tallies allow for the calculation of certain quantities important to understanding the physics of a nuclear reactor.\n\nThe methods, unlike traditional approaches to computing adjoint-weighted quantities, do not attempt to invert the random walk. Rather, they are based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. This can be calculated in a strictly forward calculation, and this factor can be applied to previously computed tally scores.\n\nThese methods are implemented in a production Monte Carlo code and are used to calculate parameters requiring adjoint weighting, the point reactor kinetics parameters and reactivity changes based upon first-order perturbation theory. 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