Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant’s structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.
{"title":"Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor","authors":"D. Guillen","doi":"10.2172/911720","DOIUrl":"https://doi.org/10.2172/911720","url":null,"abstract":"Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant’s structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"15 1","pages":"827-828"},"PeriodicalIF":0.0,"publicationDate":"2006-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86214406","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
H. M. McIlroy, D. McEligot, R. Schultz, Daniel P. Christensen, R. Pink, R. Johnson
The Very-High-Temperature Reactor (VHTR) is one of six reactor technologies chosen for further development by the Generation IV International Forum. In addition this system is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. In preparation for the thermal-hydraulics and safety analyses that will be required to confirm the performance of the NGNP, work has begun on readying the computational tools that will be needed to predict the thermal-hydraulics conditions and safety margins of the reactor design. Meaningful feasibility studies for VHTR designs will require accurate, reliable predictions of material temperatures which depend upon the thermal convection in the coolant channels of the core and other components. Unfortunately, one-dimensional system codes for gas-cooled reactors typically underpredict these temperatures, particularly for reduced power operations and hypothesized accident scenarios. Likewise, most turbulence models in general-purpose CFD codes also underpredict these temperatures. Matched-Index-of-Refraction (MIR) fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical VHTR plenum geometries in the limiting case of negligiblemore » buoyancy and constant fluid properties.« less
{"title":"Experimental measurement of flow phenomena in a VHTR lower plenum model","authors":"H. M. McIlroy, D. McEligot, R. Schultz, Daniel P. Christensen, R. Pink, R. Johnson","doi":"10.2172/948590","DOIUrl":"https://doi.org/10.2172/948590","url":null,"abstract":"The Very-High-Temperature Reactor (VHTR) is one of six reactor technologies chosen for further development by the Generation IV International Forum. In addition this system is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. In preparation for the thermal-hydraulics and safety analyses that will be required to confirm the performance of the NGNP, work has begun on readying the computational tools that will be needed to predict the thermal-hydraulics conditions and safety margins of the reactor design. Meaningful feasibility studies for VHTR designs will require accurate, reliable predictions of material temperatures which depend upon the thermal convection in the coolant channels of the core and other components. Unfortunately, one-dimensional system codes for gas-cooled reactors typically underpredict these temperatures, particularly for reduced power operations and hypothesized accident scenarios. Likewise, most turbulence models in general-purpose CFD codes also underpredict these temperatures. Matched-Index-of-Refraction (MIR) fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical VHTR plenum geometries in the limiting case of negligiblemore » buoyancy and constant fluid properties.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"1 1","pages":"367-368"},"PeriodicalIF":0.0,"publicationDate":"2006-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79662597","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A secure and affordable energy supply is essential for achieving U.S. national security, in continuing U.S. prosperity and in laying the foundation to enable future economic growth. The next generation energy workforce in the U.S. is a critical element in meeting both national and global energy needs. The Center for Advanced Energy Studies (CAES) was established in 2005 in response to U.S. Department of Energy (DOE) requirements. CAES, located at the new Idaho National Laboratory (INL), will address critical energy education, research, policy study and training needs. CAES is a unique joint partnership between the Battelle Energy Alliance (BEA), the State of Idaho, an Idaho University Consortium (IUC), and a National University Consortium (NUC). CAES will be based in a new facility that will foster collaborative academic and research efforts among participating institutions.
{"title":"The new center for advanced energy studies (CAES)","authors":"L. Bond, R. A. Whadon, A. Kadak","doi":"10.18260/1-2--1328","DOIUrl":"https://doi.org/10.18260/1-2--1328","url":null,"abstract":"A secure and affordable energy supply is essential for achieving U.S. national security, in continuing U.S. prosperity and in laying the foundation to enable future economic growth. The next generation energy workforce in the U.S. is a critical element in meeting both national and global energy needs. The Center for Advanced Energy Studies (CAES) was established in 2005 in response to U.S. Department of Energy (DOE) requirements. CAES, located at the new Idaho National Laboratory (INL), will address critical energy education, research, policy study and training needs. CAES is a unique joint partnership between the Battelle Energy Alliance (BEA), the State of Idaho, an Idaho University Consortium (IUC), and a National University Consortium (NUC). CAES will be based in a new facility that will foster collaborative academic and research efforts among participating institutions.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"74 1","pages":"95-96"},"PeriodicalIF":0.0,"publicationDate":"2006-06-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73694617","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper discusses the technologies applied at Hanford's K-Basins to mitigate risk and reduce the source term in preparing the basins for deactivation and demolition. These project technologies/strategies (in various stages of implementation) are sequential in nature and are the basis for preparing to dispose of the K Basins--two highly contaminated concrete basins at the Hanford Site in southeastern Washington State. A large collection of spent nuclear fuel stored for many years underwater at the K Basins has been removed to stable, dry, safe storage. Remediation activities are underway to prepare the basin structures for de-inventory, decontamination, and disposal.
{"title":"SOURCE TERM REMEDIATION & DEMOLITION STRATEGY FOR THE HANFORD K-AREA SPENT FUEL BASINS","authors":"G. B. Chronister","doi":"10.2172/881670","DOIUrl":"https://doi.org/10.2172/881670","url":null,"abstract":"This paper discusses the technologies applied at Hanford's K-Basins to mitigate risk and reduce the source term in preparing the basins for deactivation and demolition. These project technologies/strategies (in various stages of implementation) are sequential in nature and are the basis for preparing to dispose of the K Basins--two highly contaminated concrete basins at the Hanford Site in southeastern Washington State. A large collection of spent nuclear fuel stored for many years underwater at the K Basins has been removed to stable, dry, safe storage. Remediation activities are underway to prepare the basin structures for de-inventory, decontamination, and disposal.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"18 1","pages":"312-313"},"PeriodicalIF":0.0,"publicationDate":"2006-03-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91162224","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Critical experiments were conducted at the Oak Ridge Critical Experiment Facility (ORCEF) to determine the critical concentration for an unreflected 69.2-cm-diameter sphere of UO{sub 2}F{sub 2}, at an enrichment of {approx}37 percent U{sup 235}, by weight. These experiments were a continuation of previous efforts to determine critical dimensions for fissile materials in simple geometry. Some of the earlier experiments in this vessel have been published as part of the OECD handbook. The reports concerning these experiments have only recently become available. Until August 2005, Refs. 2 and 3 were still classified. These documents, along with experimental logbooks and unclassified papers available on the experimental campaign and facility are being used to generate a computer model for this critical experiment.
{"title":"Calculation of Critical Experiments involving U(37)O2F2 Solution","authors":"K. L. Goluoglu","doi":"10.2172/885548","DOIUrl":"https://doi.org/10.2172/885548","url":null,"abstract":"Critical experiments were conducted at the Oak Ridge Critical Experiment Facility (ORCEF) to determine the critical concentration for an unreflected 69.2-cm-diameter sphere of UO{sub 2}F{sub 2}, at an enrichment of {approx}37 percent U{sup 235}, by weight. These experiments were a continuation of previous efforts to determine critical dimensions for fissile materials in simple geometry. Some of the earlier experiments in this vessel have been published as part of the OECD handbook. The reports concerning these experiments have only recently become available. Until August 2005, Refs. 2 and 3 were still classified. These documents, along with experimental logbooks and unclassified papers available on the experimental campaign and facility are being used to generate a computer model for this critical experiment.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"28 1","pages":"150-151"},"PeriodicalIF":0.0,"publicationDate":"2006-03-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77393590","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k{sub eff}) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present amore » method for selecting conservative and bounding configurations for a wide range of end conditions.« less
{"title":"Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion","authors":"D. Kimball, C. Sanders","doi":"10.2172/893931","DOIUrl":"https://doi.org/10.2172/893931","url":null,"abstract":"The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k{sub eff}) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present amore » method for selecting conservative and bounding configurations for a wide range of end conditions.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"12 1","pages":"145-146"},"PeriodicalIF":0.0,"publicationDate":"2006-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89821844","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The report presents the results of extensive experimental work on the oxidation of Zircaloy-4 in air at high temperatures. The experimental program was aimed at mechanistic phenomenology of the reaction between Zircaloy and air and investigation of air attack under prototypical conditions for air ingress under the conditions of an hypothetical severe nuclear reactor accident, i.e. at temperatures 800-1500 °C and consideration of mixed air(nitrogen)-steam atmospheres and pre-oxidation. The oxidation in air as well as in air and nitrogen-containing atmospheres leads to strong degradation of the cladding material. The main mechanism for this process is the formation of zirconium nitride and its re-oxidation. From safety point of view, the barrier effect of the fuel cladding is lost much earlier than during accident transients with only a steam atmosphere. Pre-oxidation in steam prevents air attack as long as the oxide scale is intact, i.e. at temperatures above 1050 °C (beyond breakaway regime) and as long as oxidising gases are available (no steam starvation conditions). Under steam/oxygen starvation conditions the oxide scale is reduced and significant external nitride formation takes place. Stronger degradation of cladding tubes was also observed in air-steam and nitrogen-steam mixtures over a wide range of compositions. Regarding modelling of air ingress in severe accident computer codes one conclusion is that parabolic correlations for oxidation in air may be applied only for high temperatures (>1400 °C) and for pre-oxidized cladding (> 1100 °C). For all other conditions faster, more linear reaction kinetics should be applied. The results presented in this report are mainly of phenomenological nature. Therefore, the program will be extended by selected experiments oriented on the determination of kinetic correlations.
{"title":"Prototypical experiments on air oxidation of zircaloy cladding at high temperatures","authors":"M. Steinbrück, U. Stegmaier, T. Ziegler","doi":"10.5445/IR/270067262","DOIUrl":"https://doi.org/10.5445/IR/270067262","url":null,"abstract":"The report presents the results of extensive experimental work on the oxidation of Zircaloy-4 in air at high temperatures. The experimental program was aimed at mechanistic phenomenology of the reaction between Zircaloy and air and investigation of air attack under prototypical conditions for air ingress under the conditions of an hypothetical severe nuclear reactor accident, i.e. at temperatures 800-1500 °C and consideration of mixed air(nitrogen)-steam atmospheres and pre-oxidation. The oxidation in air as well as in air and nitrogen-containing atmospheres leads to strong degradation of the cladding material. The main mechanism for this process is the formation of zirconium nitride and its re-oxidation. From safety point of view, the barrier effect of the fuel cladding is lost much earlier than during accident transients with only a steam atmosphere. Pre-oxidation in steam prevents air attack as long as the oxide scale is intact, i.e. at temperatures above 1050 °C (beyond breakaway regime) and as long as oxidising gases are available (no steam starvation conditions). Under steam/oxygen starvation conditions the oxide scale is reduced and significant external nitride formation takes place. Stronger degradation of cladding tubes was also observed in air-steam and nitrogen-steam mixtures over a wide range of compositions. Regarding modelling of air ingress in severe accident computer codes one conclusion is that parabolic correlations for oxidation in air may be applied only for high temperatures (>1400 °C) and for pre-oxidized cladding (> 1100 °C). For all other conditions faster, more linear reaction kinetics should be applied. The results presented in this report are mainly of phenomenological nature. Therefore, the program will be extended by selected experiments oriented on the determination of kinetic correlations.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"12 1","pages":"136-137"},"PeriodicalIF":0.0,"publicationDate":"2006-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78256970","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.
{"title":"Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores","authors":"A. Krass","doi":"10.2172/885963","DOIUrl":"https://doi.org/10.2172/885963","url":null,"abstract":"This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"335 1","pages":"582-583"},"PeriodicalIF":0.0,"publicationDate":"2005-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76904896","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The U.S. Department of Energy (DOE) Complex includes many sites and laboratories that store quantities of low-level, solid nuclear waste in drums and other types of shipping containers. The drums may be stored for long periods of time prior to being transported and final dispositioning. Based on the radioactivity (e.g., Pu{sup 239} equivalent), chemical nature (e.g. volatile organic compounds) and other characteristics of the stored waste, flammable gases may evolve. Documented safety analyses (DSAs) for storage of these drums must address storage and safety management issues to protect workers, the general public, and the environment. This paper discusses an improved analytical method for determining the explosion effects flammable gas-air mixtures as well as the subsequent accident phenomenology.
{"title":"An Improved Analytical Approach to Determine the Explosive Effects of Flammable Gas-Air Mixtures","authors":"Joong M. Yang","doi":"10.2172/888583","DOIUrl":"https://doi.org/10.2172/888583","url":null,"abstract":"The U.S. Department of Energy (DOE) Complex includes many sites and laboratories that store quantities of low-level, solid nuclear waste in drums and other types of shipping containers. The drums may be stored for long periods of time prior to being transported and final dispositioning. Based on the radioactivity (e.g., Pu{sup 239} equivalent), chemical nature (e.g. volatile organic compounds) and other characteristics of the stored waste, flammable gases may evolve. Documented safety analyses (DSAs) for storage of these drums must address storage and safety management issues to protect workers, the general public, and the environment. This paper discusses an improved analytical method for determining the explosion effects flammable gas-air mixtures as well as the subsequent accident phenomenology.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"13 2 1","pages":"177-177"},"PeriodicalIF":0.0,"publicationDate":"2005-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85659813","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The Yucca Mountain repository requires the use of a waste package (WP) transporter to transport a WP from a process facility on the surface to the subsurface for underground emplacement. The transporter is a part of the waste emplacement transport systems, which includes a primary locomotive at the front end and a secondary locomotive at the rear end. The overall system with a WP on board weights over 350 metric tons (MT). With the shielding mass constituting approximately one-third of the total system weight, shielding optimization for minimal weight will benefit the overall transport system with reduced axle requirements and improved maneuverability. With a high contact dose rate on the WP external surface and minimal personnel shielding afforded by the WP, the transporter provides radiation shielding to workers during waste emplacement and retrieval operations. This paper presents the design approach and optimization method used in achieving a shielding configuration with minimal weight.
{"title":"Repository waste package transporter shielding weight optimization","authors":"C. Sanders, S. Su","doi":"10.2172/840150","DOIUrl":"https://doi.org/10.2172/840150","url":null,"abstract":"The Yucca Mountain repository requires the use of a waste package (WP) transporter to transport a WP from a process facility on the surface to the subsurface for underground emplacement. The transporter is a part of the waste emplacement transport systems, which includes a primary locomotive at the front end and a secondary locomotive at the rear end. The overall system with a WP on board weights over 350 metric tons (MT). With the shielding mass constituting approximately one-third of the total system weight, shielding optimization for minimal weight will benefit the overall transport system with reduced axle requirements and improved maneuverability. With a high contact dose rate on the WP external surface and minimal personnel shielding afforded by the WP, the transporter provides radiation shielding to workers during waste emplacement and retrieval operations. This paper presents the design approach and optimization method used in achieving a shielding configuration with minimal weight.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"6 1","pages":"31-32"},"PeriodicalIF":0.0,"publicationDate":"2005-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84673283","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：