This paper presents the results of the numerical analysis performed to asses the structural integrity of spent nuclear fuel (SNF) stainless steel canisters when subjected to impact loads associated with free gravity drops from heights not exceeding 20 ft. The SNF canisters are to be used for the Shipment of radioactive material from the Oak Ridge National Laboratory (ORNL) Site to the Idaho National Engineering and Environmental Laboratory (INEEL) for storage. The Idaho chemical Processing Plant Fuel Receipt Criteria Questionnaire requires that the vertical drop accidents from two heights be analyze. These heights are those that are considered to be critical at the time of unloading the canisters from the shipping cask. The configurations a...
This study examined the safety of nuclear spent-fuel (NSF) transport casks against accidental punctu...
In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to ...
In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to ...
1999PDFResearch PaperSnow, Spencer D.Morton, D. KeithWare, A. G.Rahl, Tom E.Smith, NancyIdaho Nation...
The Department of Energy (DOE) has developed a design concept for a set of standard canisters for th...
The National Spent Nuclear Fuel Program (NSNFP) has pursued a number of structural testing projects ...
A series of analyses were performed to assess the structural response of spent nuclear fuel dry cask...
A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped ...
This paper comparatively evaluates stress-based and strain-based acceptance criteria that are sugges...
During fiscal year 1999, a total of nine 18-inch diameter test canisters were fabricated at the Idah...
This study investigates the effects of potential drops of a typical shipping cask, waste container, ...
The inventory of spent nuclear fuel (SNF) generated in nuclear power plants is continuously increasi...
The stress-based criteria specified in ASME Section III (American Society of Mechanical Engineers, 2...
This work concerns the evaluation of spent nuclear fuel disposal canister behaviour and integrity in...
This work concerns the evaluation of spent nuclear fuel disposal canister behaviour and integrity in...
This study examined the safety of nuclear spent-fuel (NSF) transport casks against accidental punctu...
In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to ...
In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to ...
1999PDFResearch PaperSnow, Spencer D.Morton, D. KeithWare, A. G.Rahl, Tom E.Smith, NancyIdaho Nation...
The Department of Energy (DOE) has developed a design concept for a set of standard canisters for th...
The National Spent Nuclear Fuel Program (NSNFP) has pursued a number of structural testing projects ...
A series of analyses were performed to assess the structural response of spent nuclear fuel dry cask...
A crush pad has been designed and analyzed to absorb the kinetic energy of a hypothetically dropped ...
This paper comparatively evaluates stress-based and strain-based acceptance criteria that are sugges...
During fiscal year 1999, a total of nine 18-inch diameter test canisters were fabricated at the Idah...
This study investigates the effects of potential drops of a typical shipping cask, waste container, ...
The inventory of spent nuclear fuel (SNF) generated in nuclear power plants is continuously increasi...
The stress-based criteria specified in ASME Section III (American Society of Mechanical Engineers, 2...
This work concerns the evaluation of spent nuclear fuel disposal canister behaviour and integrity in...
This work concerns the evaluation of spent nuclear fuel disposal canister behaviour and integrity in...
This study examined the safety of nuclear spent-fuel (NSF) transport casks against accidental punctu...
In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to ...
In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to ...