CANDU fuel bundles experience plastic deformations over time, and the horizontal configuration of the bundle in a crept pressure tube (PT) causes coolant to bypass the sagged lower half of the bundle. Bundle segments where the flow is limited may become more susceptible to dryout due to reactor aging. A finite element model of a 37-element fuel bundle was constructed using the commercial finite element software ANSYS to study the mechanical deformation behaviour of the bundle to maintain a coolable geometry. The main focus was on the contact between the fuel elements and between the fuel elements and PT. The complexity of the model due to all the contact pairs necessitated the use of high-powered computing hardware. Contact was demonstrated between the appendages, and sensitivity of the deformation to different boundary conditions (BC) was investigated. In particular, the radial position where the elements were welded to the endplate significantly impacted the magnitude of the element bowing. Expanding the PT up to 8% diametral creep demonstrated the proper functioning of the spacer pads (SP) and bearing pads in preventing sheath-to-sheath contact at the midplane and sheath-to-PT contact. However, the quarter plane was deemed to be the critical region due to the lack of SPs preventing excessive element bowing. This work has successfully illustrated the deformation of a CANDU fuel bundle, with contact, and its similarity with the bow profiles when compared with post-irradiation examination results and bundle heat-up tests.
{"title":"FINITE ELEMENT ANALYSIS OF AN EMPTY 37-ELEMENT CANDU® FUEL BUNDLE TO STUDY THE EFFECTS OF PRESSURE TUBE CREEP","authors":"Kyuhwan Lee, D. Wowk, P. Chan","doi":"10.12943/cnr.2020.00003","DOIUrl":"https://doi.org/10.12943/cnr.2020.00003","url":null,"abstract":"CANDU fuel bundles experience plastic deformations over time, and the horizontal configuration of the bundle in a crept pressure tube (PT) causes coolant to bypass the sagged lower half of the bundle. Bundle segments where the flow is limited may become more susceptible to dryout due to reactor aging. A finite element model of a 37-element fuel bundle was constructed using the commercial finite element software ANSYS to study the mechanical deformation behaviour of the bundle to maintain a coolable geometry. The main focus was on the contact between the fuel elements and between the fuel elements and PT. The complexity of the model due to all the contact pairs necessitated the use of high-powered computing hardware. Contact was demonstrated between the appendages, and sensitivity of the deformation to different boundary conditions (BC) was investigated. In particular, the radial position where the elements were welded to the endplate significantly impacted the magnitude of the element bowing. Expanding the PT up to 8% diametral creep demonstrated the proper functioning of the spacer pads (SP) and bearing pads in preventing sheath-to-sheath contact at the midplane and sheath-to-PT contact. However, the quarter plane was deemed to be the critical region due to the lack of SPs preventing excessive element bowing. This work has successfully illustrated the deformation of a CANDU fuel bundle, with contact, and its similarity with the bow profiles when compared with post-irradiation examination results and bundle heat-up tests.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49191750","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}
Discharged CANDU fuel is stored under water in irradiated fuel bays (IFBs) to remove their decay heat. If the fuel is exposed to air, a self-sustaining reaction could result when the Zircaloy-4 sheathing reaches temperatures sufficient for a breakaway oxidation. To predict when the transition occurs, a 2-D fuel bundle cross-section model in air was developed using the COMSOL Multiphysics® platform. Breakaway was predicted to occur at its earliest within 2.6 hours for a range of recently discharged bundle powers. It was concluded due to the time required for heat up and cracking of the oxide layer, sufficient margin exists for operators to intervene before a passively cooled, isolated bundle undergoes breakaway. To examine the effect of multiple bundles, a 3-D model based on a quarter of a stand-alone spent fuel rack was developed to calculate the steady-state temperature and mass fluxes of air. The model provided a lower bound for the ambient temperatures because the flow resistance of the bundle was not considered. The correct incorporation of flow resistance is a necessary step before conclusions could be made about the safety of IFBs. However, the analysis using a Computational Fluid Dynamics model for a 0.5 MW fuel rack, indicated that the maximum temperature of the air within the rack was 642 K and located at the centre of the outlet. This result is encouraging to support the safety of IFBs, as the temperature is well below the 873 K, which is approximately the minimum required for a breakaway reaction.
{"title":"HEAT TRANSFER OF CANDU FUEL BUNDLES AFTER A LOSS OF COOLANT ACCIDENT IN AN IRRADIATED FUEL BAY","authors":"Derek Logtenberg, P. Chan, E. Corcoran","doi":"10.12943/cnr.2019.00013","DOIUrl":"https://doi.org/10.12943/cnr.2019.00013","url":null,"abstract":"Discharged CANDU fuel is stored under water in irradiated fuel bays (IFBs) to remove their decay heat. If the fuel is exposed to air, a self-sustaining reaction could result when the Zircaloy-4 sheathing reaches temperatures sufficient for a breakaway oxidation. To predict when the transition occurs, a 2-D fuel bundle cross-section model in air was developed using the COMSOL Multiphysics® platform. Breakaway was predicted to occur at its earliest within 2.6 hours for a range of recently discharged bundle powers. It was concluded due to the time required for heat up and cracking of the oxide layer, sufficient margin exists for operators to intervene before a passively cooled, isolated bundle undergoes breakaway. To examine the effect of multiple bundles, a 3-D model based on a quarter of a stand-alone spent fuel rack was developed to calculate the steady-state temperature and mass fluxes of air. The model provided a lower bound for the ambient temperatures because the flow resistance of the bundle was not considered. The correct incorporation of flow resistance is a necessary step before conclusions could be made about the safety of IFBs. However, the analysis using a Computational Fluid Dynamics model for a 0.5 MW fuel rack, indicated that the maximum temperature of the air within the rack was 642 K and located at the centre of the outlet. This result is encouraging to support the safety of IFBs, as the temperature is well below the 873 K, which is approximately the minimum required for a breakaway reaction.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43139775","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}
Nuclear power plants could potentially be deployed in a type of nuclear hybrid energy system (NHES) in which their power is used primarily to drive an industrial process but can be diverted to meet demands for electricity when needed. The purpose of this study is to analyze the effects of deploying NHESs as reserve power for the transmission grid in Ontario on the overall Canadian fuel cycle. In this scenario, the fuel cycle demands of 2 high-temperature gas-cooled reactor (HTGR) concepts are analyzed with respect to costs, resource consumption, and enrichment requirements. One HTGR concept is a 30 MW-thermal (MWth) reactor that is based on the UBattery concept, and the other is the Xe-100, which is a 200 MWth reactor. Calculations indicate that such a deployment of HTGRs would have a substantial effect on the fuel cycle in Canada. In particular, NU and enrichment demands would be greatly affected. Beginning this HTGR deployment in the year 2030 would more than double the annual NU demands in Canada, and deplete the uranium resources with extraction costs of <$80/kgU by the year 2142. The uranium enrichment demands of this fleet would be >35% of the US capacity for uranium enrichment.
{"title":"FUEL CYCLE IMPLICATIONS OF DEPLOYING HTGRS IN HYBRID ENERGY SYSTEMS AS RESERVE POWER GENERATION IN ONTARIO","authors":"D. Wojtaszek, S. Golesorkhi","doi":"10.12943/cnr.2020.00002","DOIUrl":"https://doi.org/10.12943/cnr.2020.00002","url":null,"abstract":"Nuclear power plants could potentially be deployed in a type of nuclear hybrid energy system (NHES) in which their power is used primarily to drive an industrial process but can be diverted to meet demands for electricity when needed. The purpose of this study is to analyze the effects of deploying NHESs as reserve power for the transmission grid in Ontario on the overall Canadian fuel cycle. In this scenario, the fuel cycle demands of 2 high-temperature gas-cooled reactor (HTGR) concepts are analyzed with respect to costs, resource consumption, and enrichment requirements. One HTGR concept is a 30 MW-thermal (MWth) reactor that is based on the UBattery concept, and the other is the Xe-100, which is a 200 MWth reactor. Calculations indicate that such a deployment of HTGRs would have a substantial effect on the fuel cycle in Canada. In particular, NU and enrichment demands would be greatly affected. Beginning this HTGR deployment in the year 2030 would more than double the annual NU demands in Canada, and deplete the uranium resources with extraction costs of <$80/kgU by the year 2142. The uranium enrichment demands of this fleet would be >35% of the US capacity for uranium enrichment.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44631028","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}
Dipankar Mukherjee, G. Choudhuri, R. Pal, Sanjoy Das
Fuel failures are always a cause of concern in any nuclear reactors as it increases the manrem consumption of radiation workers. Although performance of the fuels in pressurized heavy water reactors is good, but still fuel failures occur occasionally. Post irradiation examination (PIE) of the failed fuel elements indicates internal hydriding, not deuteriding, as a major cause for the failures, although secondary deuteriding occurs and, in a few cases, failures are associated with defects in the end plug weld. The sources of hydrogen are either fuel pellets or the clad or the graphite coating. Restriction has been imposed on maximum content of total hydrogen in the fuel element to 1 mg to prevent hydriding of the Zircaloy clad tube. Accidental pick up of hydrogen occurs, which could lead to failure of the fuel bundles. Experimental investigations have been conducted to understand the individual effect of iodine and accidental pick up of moisture on the microstructure of Zircaloy-4 end cap welded samples with graphite coating. Results indicate that severe hydriding in Zircaloy-4 samples due to the existence of internal moisture in presence of graphite under service condition may result in fuel failure and justifies the findings of PIE.
{"title":"EFFECT OF IODINE AND MOISTURE ON THE MICROSTRUCTURE OF ZIRCALOY-4 UNDER SERVICE CONDITION IN PHWR","authors":"Dipankar Mukherjee, G. Choudhuri, R. Pal, Sanjoy Das","doi":"10.12943/cnr.2019.00015","DOIUrl":"https://doi.org/10.12943/cnr.2019.00015","url":null,"abstract":"Fuel failures are always a cause of concern in any nuclear reactors as it increases the manrem consumption of radiation workers. Although performance of the fuels in pressurized heavy water reactors is good, but still fuel failures occur occasionally. Post irradiation examination (PIE) of the failed fuel elements indicates internal hydriding, not deuteriding, as a major cause for the failures, although secondary deuteriding occurs and, in a few cases, failures are associated with defects in the end plug weld. The sources of hydrogen are either fuel pellets or the clad or the graphite coating. Restriction has been imposed on maximum content of total hydrogen in the fuel element to 1 mg to prevent hydriding of the Zircaloy clad tube. Accidental pick up of hydrogen occurs, which could lead to failure of the fuel bundles. Experimental investigations have been conducted to understand the individual effect of iodine and accidental pick up of moisture on the microstructure of Zircaloy-4 end cap welded samples with graphite coating. Results indicate that severe hydriding in Zircaloy-4 samples due to the existence of internal moisture in presence of graphite under service condition may result in fuel failure and justifies the findings of PIE.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49010635","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}
Incoloy 800H is one of several candidates for a fuel cladding material in super-critical water nuclear reactor concepts. The objective of this work is to determine the effect of the gas tungsten arc welding (GTAW) process on the microstructure and resulting tensile properties of Incoloy 800H tubes. In this work, GTAW was used to join Incoloy 800H. During welding, the weld thermal cycle produces differently featured heat-affected zone and fusion zone (FZ) microstructures. Microstructural examination revealed that weld-characteristic columnar and equiaxed dendritic structures were formed in the FZ. In comparison with the optimum heat input, both increase and decrease of heat input led to the formation of more columnar dendritic structures in the FZ. The chemical element distribution analysis using scanning electron microscopy/energy dispersive X-ray spectroscopy showed the segregation of Ti in the form of Ti-rich carbides and nitrides; other elements did not display any obvious segregation. Tensile test results revealed that Incoloy 800H alloy welds exhibit an excellent combination of strength and ductility almost equal to the base metal (BM) at the optimum and higher than optimum heat input conditions with full penetration. The welding process has no obvious effect on the microhardness across the whole welding zone. The refinement of grain size and morphology in the FZ can contribute to the improvement in the mechanical properties. As a result, the Incoloy 800H weldment shows the comparable mechanical properties to the BM.
{"title":"MICROSTRUCTURAL CHARACTERIZATION AND TENSILE PROPERTIES ASSESSMENT OF GTAW WELDED INCOLOY 800H ALLOYS FUEL CLADDING FOR SCWR","authors":"Lin-fa Xiao, G. Cota-Sanchez","doi":"10.12943/cnr.2020.00001","DOIUrl":"https://doi.org/10.12943/cnr.2020.00001","url":null,"abstract":"Incoloy 800H is one of several candidates for a fuel cladding material in super-critical water nuclear reactor concepts. The objective of this work is to determine the effect of the gas tungsten arc welding (GTAW) process on the microstructure and resulting tensile properties of Incoloy 800H tubes. In this work, GTAW was used to join Incoloy 800H. During welding, the weld thermal cycle produces differently featured heat-affected zone and fusion zone (FZ) microstructures. Microstructural examination revealed that weld-characteristic columnar and equiaxed dendritic structures were formed in the FZ. In comparison with the optimum heat input, both increase and decrease of heat input led to the formation of more columnar dendritic structures in the FZ. The chemical element distribution analysis using scanning electron microscopy/energy dispersive X-ray spectroscopy showed the segregation of Ti in the form of Ti-rich carbides and nitrides; other elements did not display any obvious segregation. Tensile test results revealed that Incoloy 800H alloy welds exhibit an excellent combination of strength and ductility almost equal to the base metal (BM) at the optimum and higher than optimum heat input conditions with full penetration. The welding process has no obvious effect on the microhardness across the whole welding zone. The refinement of grain size and morphology in the FZ can contribute to the improvement in the mechanical properties. As a result, the Incoloy 800H weldment shows the comparable mechanical properties to the BM.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46569567","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}
Determination of the effective delayed neutron fraction (βeff) and neutron generation time (Λ), on account of their important role in the reactivity transients analysis, safety, and control of nuclear reactors, is of the great importance in the reactor physics calculations. In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs is calculated. Software was developed to automate the procedure of kinetic parameters calculations. We used both a deterministic and a probabilistic method for calculation of the delayed neutron parameters. The results performed well in comparison to the reference.
{"title":"KINETIC PARAMETERS CALCULATION DURING FIRST CYCLE OF THE WWER-1000 REACTOR CORE","authors":"M. Akbari, Samira Rezaei, F. Khoshahval","doi":"10.12943/CNR.2017.00017","DOIUrl":"https://doi.org/10.12943/CNR.2017.00017","url":null,"abstract":"Determination of the effective delayed neutron fraction (βeff) and neutron generation time (Λ), on account of their important role in the reactivity transients analysis, safety, and control of nuclear reactors, is of the great importance in the reactor physics calculations. In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs is calculated. Software was developed to automate the procedure of kinetic parameters calculations. We used both a deterministic and a probabilistic method for calculation of the delayed neutron parameters. The results performed well in comparison to the reference.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2019-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43373668","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}
Nuclear waste management facilities at Chalk River Laboratories (CRL) use below-ground “tile-holes” to store solid waste from various activities such as medical isotope production. After long periods of isotopic decay, some of the waste has decayed down to low activities and can be transferred to low-level waste storage facilities. This paper presents a method to verify the radiation level of the waste inside tile-holes by performing gamma radiation scans along the depth of waste storage tile-holes. Such measurements allow for noninvasive verification of tile-hole contents and provide input to the assessment of radiological risk associated with removal of the waste. Using the radiation profile system, the radiation level of the radioactive waste may be identified based on the radiation profile. This information will support planning for possible transfer of this waste to a licensed waste storage facility designed for low-level waste, thus freeing storage space for possible tile-hole re-use for more highly radioactive waste. CRL-developed small diode-based gamma radiation sensors have been used in these radiation scans. The diode sensors were deployed into verification tubes adjacent to the tile-holes to measure the radiation profile. Over 10 tile-holes have been scanned using this technique since 2009.
{"title":"VERIFYING NUCLEAR WASTE TILE-HOLES USING GAMMA RADIATION SCANNING","authors":"J. Johnston, Shuwei Yue, J. Stewart","doi":"10.12943/CNR.2017.00016","DOIUrl":"https://doi.org/10.12943/CNR.2017.00016","url":null,"abstract":"Nuclear waste management facilities at Chalk River Laboratories (CRL) use below-ground “tile-holes” to store solid waste from various activities such as medical isotope production. After long periods of isotopic decay, some of the waste has decayed down to low activities and can be transferred to low-level waste storage facilities. This paper presents a method to verify the radiation level of the waste inside tile-holes by performing gamma radiation scans along the depth of waste storage tile-holes. Such measurements allow for noninvasive verification of tile-hole contents and provide input to the assessment of radiological risk associated with removal of the waste. Using the radiation profile system, the radiation level of the radioactive waste may be identified based on the radiation profile. This information will support planning for possible transfer of this waste to a licensed waste storage facility designed for low-level waste, thus freeing storage space for possible tile-hole re-use for more highly radioactive waste. CRL-developed small diode-based gamma radiation sensors have been used in these radiation scans. The diode sensors were deployed into verification tubes adjacent to the tile-holes to measure the radiation profile. Over 10 tile-holes have been scanned using this technique since 2009.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2019-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44780250","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}
Fe–Cr–Ni alloys have experienced localized degradation, such as stress-corrosion cracking (SCC), when used for steam generator tubes in nuclear power plants. The tube surface can be covered by a porous deposit layer resulting primarily from fouling. This porous layer acts as a barrier to the mass transfer for the chemical species in the main fluid to the tube surface. Thus, it influences the interfacial chemistry at the metal surface and the susceptibility of Fe–Cr–Ni alloys to SCC. While the chemistry of the main fluid can be controlled and monitored, this interfacial chemistry must be determined indirectly. Numerical models can be used to predict the interfacial chemistry and provide insight to SCC initiation and propagation. In the present work, a numerical model has been developed to calculate the mass-transfer rate of a chemical species, such as dissolved oxygen (DO), from main fluid to tube surface through an unreactive porous layer under single-phase liquid flow conditions. Major features of the model were validated against available literature data at room temperature (25 °C). The numerical results for high pressure (5 MPa) and high temperature (250 °C) conditions show that the effect of advection on the mass-transfer rate of DO through an unreactive porous layer dominates over that of diffusion.
{"title":"CONVECTIVE MASS TRANSFER THROUGH AN UNREACTIVE POROUS DEPOSIT LAYER UNDER HIGH TEMPERATURE CONDITIONS","authors":"Lan Sun, Qi Chen, S. Laroche","doi":"10.12943/CNR.2017.00018","DOIUrl":"https://doi.org/10.12943/CNR.2017.00018","url":null,"abstract":"Fe–Cr–Ni alloys have experienced localized degradation, such as stress-corrosion cracking (SCC), when used for steam generator tubes in nuclear power plants. The tube surface can be covered by a porous deposit layer resulting primarily from fouling. This porous layer acts as a barrier to the mass transfer for the chemical species in the main fluid to the tube surface. Thus, it influences the interfacial chemistry at the metal surface and the susceptibility of Fe–Cr–Ni alloys to SCC. While the chemistry of the main fluid can be controlled and monitored, this interfacial chemistry must be determined indirectly. Numerical models can be used to predict the interfacial chemistry and provide insight to SCC initiation and propagation. In the present work, a numerical model has been developed to calculate the mass-transfer rate of a chemical species, such as dissolved oxygen (DO), from main fluid to tube surface through an unreactive porous layer under single-phase liquid flow conditions. Major features of the model were validated against available literature data at room temperature (25 °C). The numerical results for high pressure (5 MPa) and high temperature (250 °C) conditions show that the effect of advection on the mass-transfer rate of DO through an unreactive porous layer dominates over that of diffusion.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2019-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46194239","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}
Preliminary, conceptual studies have been performed previously using deterministic lattice physics (WIMS-AECL) and core physics codes (RFSP) to estimate performance and safety characteristics of various thorium-based fuels and uranium-based fuels augmented by small amounts of thorium for use in pressure tube heavy-water reactors (PT-HWRs). To confirm the validity of the results, the WIMS-AECL/RFSP results are compared against predictions made with the stochastic neutron transport code MCNP. This paper describes the development of a method for setting up an MCNP core model of at PT-HWR for comparison with WIMS-AECL/RFSP results, using a core with 37-element natural uranium fuel bundles as a test case for sensitivity studies. These studies included evaluating the sensitivity of the bias of the effective neutron multiplication factor (keff), a source convergence study, uncertainties correction with multiple independent simulations, the impact of irradiation map binning methods, and the impact of reflector models. A Python-based software scripting tool was developed to automate the creation, execution, and post-processing of reactor physics data from the MCNP models. The software tool and algorithm for creating an MCNP core model using data from the WIMS-AECL and RFSP models are described in this paper. Based on the preliminary evaluations of the simulation parameters with the base model, reactor physics analyses were performed for PT-HWR cores with thorium-based fuels in a 35-element bundle type. Code-to-code results demonstrate good agreement between MCNP and RFSP, giving confidence in the method developed and its applicability to other fuels and core types.
{"title":"MODELING STUDIES AND CODE-TO-CODE COMPARISONS FOR PRESSURE TUBE HEAVY WATER REACTOR CORES","authors":"Huiping Yan, B. Bromley, C. Dugal, A. V. Colton","doi":"10.12943/CNR.2018.00003","DOIUrl":"https://doi.org/10.12943/CNR.2018.00003","url":null,"abstract":"Preliminary, conceptual studies have been performed previously using deterministic lattice physics (WIMS-AECL) and core physics codes (RFSP) to estimate performance and safety characteristics of various thorium-based fuels and uranium-based fuels augmented by small amounts of thorium for use in pressure tube heavy-water reactors (PT-HWRs). To confirm the validity of the results, the WIMS-AECL/RFSP results are compared against predictions made with the stochastic neutron transport code MCNP. This paper describes the development of a method for setting up an MCNP core model of at PT-HWR for comparison with WIMS-AECL/RFSP results, using a core with 37-element natural uranium fuel bundles as a test case for sensitivity studies. These studies included evaluating the sensitivity of the bias of the effective neutron multiplication factor (keff), a source convergence study, uncertainties correction with multiple independent simulations, the impact of irradiation map binning methods, and the impact of reflector models. A Python-based software scripting tool was developed to automate the creation, execution, and post-processing of reactor physics data from the MCNP models. The software tool and algorithm for creating an MCNP core model using data from the WIMS-AECL and RFSP models are described in this paper. Based on the preliminary evaluations of the simulation parameters with the base model, reactor physics analyses were performed for PT-HWR cores with thorium-based fuels in a 35-element bundle type. Code-to-code results demonstrate good agreement between MCNP and RFSP, giving confidence in the method developed and its applicability to other fuels and core types.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2018-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44997634","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}
To ensure long-term nuclear energy security, it is advantageous to consider the use of externally driven sub-critical systems for producing fissile fuel to supply fleets of thermal-spectrum reactors as an alternative to using fast-spectrum or thermal-spectrum breeder reactors. Computational/analytical neutronics and heat transfer studies have been carried out for gas-cooled fuel bundle lattices with mixtures of fertile thorium and depleted uranium (DU) that could be used in the blanket region of a sub-critical fast reactor driven either by a fusion reactor in a hybrid fusion-fission reactor (HFFR) system, or an accelerator-based spallation neutron source in an accelerator driven system (ADS). The HFFR or ADS concept envisioned is one with a simple cylindrical geometry. The annular-cylindrical blanket is approximately 10 m long, can be made 2–5 m thick (1.0 m ≤ Rblanket ≤ 3.0 m to 6.0 m), and is filled with a repeating square lattice of pressure tubes filled with 0.5 m long fuel bundles that are made of (DU,Th)O2, with various mixtures of Th and DU, and refuelled periodically online. Although using blankets made of pure DUO2 or ThO2 are viable options to analyze, mixing DUO2 with ThO2 can help alleviate any potential proliferation concerns, since any 233U produced from breeding will be denatured by the presence of 238U in (DU, Th)O2. Lattice calculations demonstrate that the total fissile content in the fuel after an extended period of burnup (50 MWd/kg) will be approximately the same, regardless of the mixture of DU and thorium used, and that the content of americium and 232U in the irradiated fuel will be <0.01 wt%/initial heavy metal.
{"title":"ASSESSMENT OF FAST-SPECTRUM BLANKET LATTICES FOR BREEDING FISSILE FUEL FROM THORIUM AND DEPLETED URANIUM IN AN EXTERNALLY DRIVEN SUB-CRITICAL GAS-COOLED PRESSURE TUBE REACTOR","authors":"B. Bromley, J. Alexander","doi":"10.12943/CNR.2018.00010","DOIUrl":"https://doi.org/10.12943/CNR.2018.00010","url":null,"abstract":"To ensure long-term nuclear energy security, it is advantageous to consider the use of externally driven sub-critical systems for producing fissile fuel to supply fleets of thermal-spectrum reactors as an alternative to using fast-spectrum or thermal-spectrum breeder reactors. Computational/analytical neutronics and heat transfer studies have been carried out for gas-cooled fuel bundle lattices with mixtures of fertile thorium and depleted uranium (DU) that could be used in the blanket region of a sub-critical fast reactor driven either by a fusion reactor in a hybrid fusion-fission reactor (HFFR) system, or an accelerator-based spallation neutron source in an accelerator driven system (ADS). The HFFR or ADS concept envisioned is one with a simple cylindrical geometry. The annular-cylindrical blanket is approximately 10 m long, can be made 2–5 m thick (1.0 m ≤ Rblanket ≤ 3.0 m to 6.0 m), and is filled with a repeating square lattice of pressure tubes filled with 0.5 m long fuel bundles that are made of (DU,Th)O2, with various mixtures of Th and DU, and refuelled periodically online. Although using blankets made of pure DUO2 or ThO2 are viable options to analyze, mixing DUO2 with ThO2 can help alleviate any potential proliferation concerns, since any 233U produced from breeding will be denatured by the presence of 238U in (DU, Th)O2. Lattice calculations demonstrate that the total fissile content in the fuel after an extended period of burnup (50 MWd/kg) will be approximately the same, regardless of the mixture of DU and thorium used, and that the content of americium and 232U in the irradiated fuel will be <0.01 wt%/initial heavy metal.","PeriodicalId":42750,"journal":{"name":"CNL Nuclear Review","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"2018-11-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49435769","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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