G. Cota-Sanchez, K. Leeder, D. Bureau, M. Watson, Chris MacCready, D. Turgeon, D. Woods
The Recycle Fuel Fabrication Laboratory (RFFL) is a licensed nuclear facility designed to produce experimental quantities of mixed oxide fuel for reactor physics tests and demonstration irradiations. Since its refurbishment in 1996, several fuel fabrication campaigns and research projects have been completed. The RFFL is the only facility in Canada capable of handling significant quantities of plutonium and other actinides. In this context, effort has been put forth in recent years to demonstrate capabilities in new fields, such as actinide chemistry. One of the first experiments in actinide chemistry conducted in the Facility was devoted to the separation and purification of 241Am from aged PuO2. The experimental work presented here, was conducted to establish actinide separation capability at the RFFL and demonstrate the staff expertise to perform these types of separations. The separation experiments were carried out using ion exchange columns packed with basic anion exchange AG1-X4 resin. More than 92% of the 241Am contained in the starting PuO2 solution was recovered in the experiments; however, some plutonium was also found in the washing effluent fraction. The final americium concentration of the washing effluent fraction was estimated to be about 50% of the total heavy elements present. More than 93% of the purified plutonium retained in the column was eluted. The experimental results are discussed in terms of the speciation behaviour of the Am–Pu–N–H2O systems using E-pH thermodynamic equilibrium diagrams and the maximum ionic exchange capacity of the resin.
{"title":"SEPARATION OF AMERICIUM FROM AGED PLUTONIUM DIOXIDE","authors":"G. Cota-Sanchez, K. Leeder, D. Bureau, M. Watson, Chris MacCready, D. Turgeon, D. Woods","doi":"10.12943/CNR.2017.00015","DOIUrl":"https://doi.org/10.12943/CNR.2017.00015","url":null,"abstract":"The Recycle Fuel Fabrication Laboratory (RFFL) is a licensed nuclear facility designed to produce experimental quantities of mixed oxide fuel for reactor physics tests and demonstration irradiations. Since its refurbishment in 1996, several fuel fabrication campaigns and research projects have been completed. The RFFL is the only facility in Canada capable of handling significant quantities of plutonium and other actinides. In this context, effort has been put forth in recent years to demonstrate capabilities in new fields, such as actinide chemistry. One of the first experiments in actinide chemistry conducted in the Facility was devoted to the separation and purification of 241Am from aged PuO2. The experimental work presented here, was conducted to establish actinide separation capability at the RFFL and demonstrate the staff expertise to perform these types of separations. The separation experiments were carried out using ion exchange columns packed with basic anion exchange AG1-X4 resin. More than 92% of the 241Am contained in the starting PuO2 solution was recovered in the experiments; however, some plutonium was also found in the washing effluent fraction. The final americium concentration of the washing effluent fraction was estimated to be about 50% of the total heavy elements present. More than 93% of the purified plutonium retained in the column was eluted. The experimental results are discussed in terms of the speciation behaviour of the Am–Pu–N–H2O systems using E-pH thermodynamic equilibrium diagrams and the maximum ionic exchange capacity of the resin.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2018-08-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49056311","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}
Thoria–plutonia (ThO2–PuO2) pellets with a nominal composition of 9.0 wt% PuO2 were prepared using a fabrication route similar to an industrial process for production of urania–plutonia (UO2–PuO2) mixed oxide fuel. The green fuel pellets were separated into 2 batches and the sintering of each batch was carried out under a reducing atmosphere at 1820 °C or 1750 °C. The distribution of plutonium (Pu) in the sintered pellets was investigated by electron probe microanalysis using X-ray mapping and quantitative point analyses. The results show that the pellet samples consist of Pu-rich agglomerates with Pu content close to that of the mastermix blend and a thorium (Th)-rich matrix. The matrix and the Pu-rich agglomerates are separated by a transition zone with Pu content varying from practically nil to the Pu content of the Pu-rich agglomerates. X-ray maps taken from random regions of the centre of the pellets show different sizes of Pu-rich agglomerates irregularly dispersed in the Th-rich matrix. Image analysis of the Pu X-ray maps indicate that the average diameter of the Pu-rich agglomerates of the material sintered at 1820 °C and 1750 °C were 68 μm and 161 μm, respectively.
{"title":"CHARACTERIZATION OF PLUTONIUM DISTRIBUTION IN THO2–PUO2 MIXED OXIDES BY ELECTRON PROBE MICROANALYSIS","authors":"D. Woods, M. Saoudi, C. Mayhew, R. Ham-su","doi":"10.12943/CNR.2017.00014","DOIUrl":"https://doi.org/10.12943/CNR.2017.00014","url":null,"abstract":"Thoria–plutonia (ThO2–PuO2) pellets with a nominal composition of 9.0 wt% PuO2 were prepared using a fabrication route similar to an industrial process for production of urania–plutonia (UO2–PuO2) mixed oxide fuel. The green fuel pellets were separated into 2 batches and the sintering of each batch was carried out under a reducing atmosphere at 1820 °C or 1750 °C. The distribution of plutonium (Pu) in the sintered pellets was investigated by electron probe microanalysis using X-ray mapping and quantitative point analyses. The results show that the pellet samples consist of Pu-rich agglomerates with Pu content close to that of the mastermix blend and a thorium (Th)-rich matrix. The matrix and the Pu-rich agglomerates are separated by a transition zone with Pu content varying from practically nil to the Pu content of the Pu-rich agglomerates. X-ray maps taken from random regions of the centre of the pellets show different sizes of Pu-rich agglomerates irregularly dispersed in the Th-rich matrix. Image analysis of the Pu X-ray maps indicate that the average diameter of the Pu-rich agglomerates of the material sintered at 1820 °C and 1750 °C were 68 μm and 161 μm, respectively.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2018-08-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44815796","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}
Concrete containment buildings (CCBs) are important safety structures in nuclear power plants; however, degradation may occur in CCBs as they age. For post-tensioned CCBs, prestressing losses could occur and may affect the CCBs’ performance under accident conditions. CANDU CCBs contain cement-grouted post-tensioning (P-T) cables. The grouting of P-T cables prevents direct monitoring of prestressing losses by traditional lift-off testing. Instrumented monitoring has been recommended as an indirect approach by some guidelines for integrity evaluation of CCBs with grouted prestressing systems. As part of the investigation on the relationship between instrumentation data and the integrity of CCBs, sensitivity analyses have been performed using finite element models to develop an understanding of the sensitivity of strain changes to degradation factors that contribute to prestressing losses, such as creep and shrinkage of the concrete, stress relaxation, and deterioration of prestressing systems. Strain measurements from a CANDU CCB were analysed to assess the measurement noise, which was compared with the predicted strain changes due to degradation to evaluate whether the degradation of concrete and prestressing systems can be captured by strain instrumentation. The analysis reveals that the strain changes due to degradation, except the creep and shrinkage during the early years of CCBs, were comparable with the level of noise observed in the measured strain data. Degradation mechanisms related to prestressing losses have conflicting effects on strain changes and are difficult to assess individually. Therefore, it could be difficult to detect the prestressing losses and the effect of individual degradation issues using strain instrumentation.
{"title":"THE IMPACT OF CONCRETE AND POST-TENSIONING CABLE DEGRADATION ON STRAIN MEASUREMENTS OF CONCRETE CONTAINMENT BUILDINGS","authors":"Yuqing Ding, S. Jaffer","doi":"10.12943/CNR.2017.00012","DOIUrl":"https://doi.org/10.12943/CNR.2017.00012","url":null,"abstract":"Concrete containment buildings (CCBs) are important safety structures in nuclear power plants; however, degradation may occur in CCBs as they age. For post-tensioned CCBs, prestressing losses could occur and may affect the CCBs’ performance under accident conditions. CANDU CCBs contain cement-grouted post-tensioning (P-T) cables. The grouting of P-T cables prevents direct monitoring of prestressing losses by traditional lift-off testing. Instrumented monitoring has been recommended as an indirect approach by some guidelines for integrity evaluation of CCBs with grouted prestressing systems. As part of the investigation on the relationship between instrumentation data and the integrity of CCBs, sensitivity analyses have been performed using finite element models to develop an understanding of the sensitivity of strain changes to degradation factors that contribute to prestressing losses, such as creep and shrinkage of the concrete, stress relaxation, and deterioration of prestressing systems. Strain measurements from a CANDU CCB were analysed to assess the measurement noise, which was compared with the predicted strain changes due to degradation to evaluate whether the degradation of concrete and prestressing systems can be captured by strain instrumentation. The analysis reveals that the strain changes due to degradation, except the creep and shrinkage during the early years of CCBs, were comparable with the level of noise observed in the measured strain data. Degradation mechanisms related to prestressing losses have conflicting effects on strain changes and are difficult to assess individually. Therefore, it could be difficult to detect the prestressing losses and the effect of individual degradation issues using strain instrumentation.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2017-12-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45688732","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}
Experimental and operational data are valuable assets for the field of nuclear science and technology. It is very important to develop software tools to assist scientists to manage the data effectively and to conveniently access and share the data. This paper presents 5 case studies on software development pertaining to topical areas such as nuclear fuel performance, waste management, biological research, reactor physics, and chemistry analysis at Canadian Nuclear Laboratories (CNL). Each case study illustrates the design and development of the database and user interface for its target research area and end users. While datasets based on flat files are still used in reactor physics studies, full-scale relational databases are developed for most of the other studies. The user interfaces developed for these areas include web applications, desktop applications, and mobile applications. These software tools have become essential parts of the research activities at CNL.
{"title":"CASE STUDIES OF NUCLEAR RESEARCH SOFTWARE DEVELOPMENT","authors":"Huiping Yan, S. Yatabe","doi":"10.12943/CNR.2017.00013","DOIUrl":"https://doi.org/10.12943/CNR.2017.00013","url":null,"abstract":"Experimental and operational data are valuable assets for the field of nuclear science and technology. It is very important to develop software tools to assist scientists to manage the data effectively and to conveniently access and share the data. This paper presents 5 case studies on software development pertaining to topical areas such as nuclear fuel performance, waste management, biological research, reactor physics, and chemistry analysis at Canadian Nuclear Laboratories (CNL). Each case study illustrates the design and development of the database and user interface for its target research area and end users. While datasets based on flat files are still used in reactor physics studies, full-scale relational databases are developed for most of the other studies. The user interfaces developed for these areas include web applications, desktop applications, and mobile applications. These software tools have become essential parts of the research activities at CNL.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2017-12-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48679262","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}
Shielding analysis and design are important tools for ensuring that humans and the environment are protected from the detrimental effects of high levels of radiation. The fundamental principles and methodologies for shielding analysis and design, especially for reactor applications, have been developed and refined since the 1940s and the beginning of nuclear power research programs in Canada and internationally. Other applications are gaining importance due to both increased need and technological advances. In this work, a high-level survey of emerging areas in shielding research and development is provided. Areas of topical interest include remote reactor monitoring, source reconstruction and inverse shielding methods, waste management and decommissioning applications, accelerator, cyclotron, and other advanced medical shielding applications, space exploration, and new materials development. Each of these areas of interest is evaluated based on current capacity of the research community. They are also evaluated in terms of the benefits for the scientific community and industry arising from performing research including development of new technologies and techniques.
{"title":"EMERGING AREAS OF SHIELDING RESEARCH—A BIRD’S EYE VIEW","authors":"R. Dranga, F. Adams","doi":"10.12943/CNR.2017.00008","DOIUrl":"https://doi.org/10.12943/CNR.2017.00008","url":null,"abstract":"Shielding analysis and design are important tools for ensuring that humans and the environment are protected from the detrimental effects of high levels of radiation. The fundamental principles and methodologies for shielding analysis and design, especially for reactor applications, have been developed and refined since the 1940s and the beginning of nuclear power research programs in Canada and internationally. Other applications are gaining importance due to both increased need and technological advances. In this work, a high-level survey of emerging areas in shielding research and development is provided. Areas of topical interest include remote reactor monitoring, source reconstruction and inverse shielding methods, waste management and decommissioning applications, accelerator, cyclotron, and other advanced medical shielding applications, space exploration, and new materials development. Each of these areas of interest is evaluated based on current capacity of the research community. They are also evaluated in terms of the benefits for the scientific community and industry arising from performing research including development of new technologies and techniques.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2017-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42665114","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}
M. McDonald, M. Moore, D. Wojtaszek, Nicholas Chornoboy, G. Edwards
An incremental approach to introducing thorium to the conventional pressure-tube heavy-water reactor natural uranium fuel cycle is investigated. The approach involves the replacement of the centre fuel element of the bundle with an element of thorium dioxide. Increasing the operating margin of a key safety parameter, the coolant void reactivity, is a prime motivating factor. The analyses showed that the simple use of a single pin of thorium is unlikely to be economically advantageous due to a large burnup penalty and increased fuel costs. However, a slight reduction in the void reactivity is observed, and this approach does allow the exploitation of the energy potential available in thorium as an alternative nuclear fuel resource through the development of a U-233 resource. This bundle concept may also be advantageous from a fuel disposal point of view, as the fuel requires less time in storage before emplacement in a deep geological repository.
{"title":"ON THE USE OF A CENTRAL THORIUM FUEL ELEMENT IN PRESSURE-TUBE HEAVY-WATER REACTOR FUEL BUNDLES","authors":"M. McDonald, M. Moore, D. Wojtaszek, Nicholas Chornoboy, G. Edwards","doi":"10.12943/CNR.2017.00009","DOIUrl":"https://doi.org/10.12943/CNR.2017.00009","url":null,"abstract":"An incremental approach to introducing thorium to the conventional pressure-tube heavy-water reactor natural uranium fuel cycle is investigated. The approach involves the replacement of the centre fuel element of the bundle with an element of thorium dioxide. Increasing the operating margin of a key safety parameter, the coolant void reactivity, is a prime motivating factor. The analyses showed that the simple use of a single pin of thorium is unlikely to be economically advantageous due to a large burnup penalty and increased fuel costs. However, a slight reduction in the void reactivity is observed, and this approach does allow the exploitation of the energy potential available in thorium as an alternative nuclear fuel resource through the development of a U-233 resource. This bundle concept may also be advantageous from a fuel disposal point of view, as the fuel requires less time in storage before emplacement in a deep geological repository.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2017-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47667635","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 “dirty bomb” is a type of hypothetical radiological dispersal device (RDD) that has been the subject of significant safety and security concerns given the disruption that would result in a postulated terrorist attack. Reliable and accurate predictions of dispersion of radiological material from an RDD are absolutely necessary for first responders and emergency decision makers to plan effective response strategies. Development of high-fidelity, mechanistic models of a dirty bomb are complicated because dispersion over areas with the greatest risk of contamination is highly sensitive to the source of contaminant particles, and this source term is governed by processes over much smaller temporal and spatial length scales than the dispersion. New work on accelerating high-fidelity models of RDDs has been initiated that looks to incorporate the multiscale aspects of the problem and enhance predictive capabilities that may assist in anti-terrorism activities.
{"title":"CHALLENGES FOR PHYSICS-BASED MODELS OF A RADIONUCLIDE DISPERSAL DEVICE","authors":"D. Hummel, L. Ivan","doi":"10.12943/CNR.2017.00005","DOIUrl":"https://doi.org/10.12943/CNR.2017.00005","url":null,"abstract":"A “dirty bomb” is a type of hypothetical radiological dispersal device (RDD) that has been the subject of significant safety and security concerns given the disruption that would result in a postulated terrorist attack. Reliable and accurate predictions of dispersion of radiological material from an RDD are absolutely necessary for first responders and emergency decision makers to plan effective response strategies. Development of high-fidelity, mechanistic models of a dirty bomb are complicated because dispersion over areas with the greatest risk of contamination is highly sensitive to the source of contaminant particles, and this source term is governed by processes over much smaller temporal and spatial length scales than the dispersion. New work on accelerating high-fidelity models of RDDs has been initiated that looks to incorporate the multiscale aspects of the problem and enhance predictive capabilities that may assist in anti-terrorism activities.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2017-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44849271","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}
Accidental hydrogen production in nuclear reactors has been a significant focus of nuclear reactor safety for decades. However, since the accident at Fukushima Daiichi nuclear generating station, hydrogen safety in nuclear reactors is a more relevant topic. As new reactor concepts, such as the supercritical water-cooled reactor (SCWR), are designed and developed the risk of unintentional hydrogen generation is not eliminated; however, it can be mitigated in the design. A systematic assessment of the hydrogen risk from both normal and accident conditions in the Canadian SCWR design was performed, in which various techniques to mitigate the hydrogen combustion potential were considered. While the rate of hydrogen generation under normal operating conditions was found to be low when held at supercritical water conditions, conservative estimates suggest that a significant quantity of hydrogen may be produced and released to the containment building in a severe accident. As a result, a hydrogen–oxygen management concept has been proposed to mitigate the hydrogen produced in a severe accident that includes a nitrogen-inerted containment building to reduce the combustion potential of hydrogen and the installation of passive autocatalytic recombiners for oxygen management. This hydrogen–oxygen management concept results in significant design changes and likely significant economic and operational impacts on the Canadian SCWR design.
{"title":"DEVELOPMENT OF A HYDROGEN MANAGEMENT CONCEPT FOR THE CANADIAN SUPERCRITICAL WATER-COOLED REACTOR","authors":"L. Gardner, D. Ryland, S. Suppiah","doi":"10.12943/CNR.2017.00004","DOIUrl":"https://doi.org/10.12943/CNR.2017.00004","url":null,"abstract":"Accidental hydrogen production in nuclear reactors has been a significant focus of nuclear reactor safety for decades. However, since the accident at Fukushima Daiichi nuclear generating station, hydrogen safety in nuclear reactors is a more relevant topic. As new reactor concepts, such as the supercritical water-cooled reactor (SCWR), are designed and developed the risk of unintentional hydrogen generation is not eliminated; however, it can be mitigated in the design. A systematic assessment of the hydrogen risk from both normal and accident conditions in the Canadian SCWR design was performed, in which various techniques to mitigate the hydrogen combustion potential were considered. While the rate of hydrogen generation under normal operating conditions was found to be low when held at supercritical water conditions, conservative estimates suggest that a significant quantity of hydrogen may be produced and released to the containment building in a severe accident. As a result, a hydrogen–oxygen management concept has been proposed to mitigate the hydrogen produced in a severe accident that includes a nitrogen-inerted containment building to reduce the combustion potential of hydrogen and the installation of passive autocatalytic recombiners for oxygen management. This hydrogen–oxygen management concept results in significant design changes and likely significant economic and operational impacts on the Canadian SCWR design.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2017-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44529461","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}
M. Piro, D. Sunderland, W. Revie, S. Livingstone, I. Dimayuga, A. Douchant, Michael Wright
Potential mitigation strategies for preventing stress corrosion cracking (SCC) failures in CANDU fuel cladding that are based on lessons learned on both domestic and international fronts are discussed in this paper. Although SCC failures have not been a major concern in CANDU reactors in recent decades, they may resurface at higher burnup for conventional fuels or with nonconventional fuels that are currently being investigated, such as MOX or thoria-based fuels. The motivation of this work is to provide the foundation for considering possible remedies for SCC failures. Three candidate remedies are discussed, namely improved fabrication methods for fuel appendages, barrier-liner cladding, and fuel doping. In support of this effort, recent advances in experimental characterization methods are described—methods that have been successfully used in non-nuclear materials that can be used to further elucidate SCC behaviour in CANDU fuel. The overall objective is to outline a path forward for characterizing material behaviour as an essential part of investigating remedies to SCC failure. This will allow increased fuel discharge burnup, maximum linear power, and plant manoeuvrability, while maintaining a high degree of reliability.
{"title":"POTENTIAL MITIGATION STRATEGIES FOR PREVENTING STRESS CORROSION CRACKING FAILURES IN HIGH-BURNUP CANDU FUEL","authors":"M. Piro, D. Sunderland, W. Revie, S. Livingstone, I. Dimayuga, A. Douchant, Michael Wright","doi":"10.12943/CNR.2016.00011","DOIUrl":"https://doi.org/10.12943/CNR.2016.00011","url":null,"abstract":"Potential mitigation strategies for preventing stress corrosion cracking (SCC) failures in CANDU fuel cladding that are based on lessons learned on both domestic and international fronts are discussed in this paper. Although SCC failures have not been a major concern in CANDU reactors in recent decades, they may resurface at higher burnup for conventional fuels or with nonconventional fuels that are currently being investigated, such as MOX or thoria-based fuels. The motivation of this work is to provide the foundation for considering possible remedies for SCC failures. Three candidate remedies are discussed, namely improved fabrication methods for fuel appendages, barrier-liner cladding, and fuel doping. In support of this effort, recent advances in experimental characterization methods are described—methods that have been successfully used in non-nuclear materials that can be used to further elucidate SCC behaviour in CANDU fuel. The overall objective is to outline a path forward for characterizing material behaviour as an essential part of investigating remedies to SCC failure. This will allow increased fuel discharge burnup, maximum linear power, and plant manoeuvrability, while maintaining a high degree of reliability.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2016-09-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"66365477","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}
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