scholarly journals Structural integrity of KJRR-F fresh nuclear fuel under vehicle-induced vibration for normal transport condition

2021 ◽  
Author(s):  
Gil-Eon Jeong ◽  
Yun-Young Yang ◽  
Kyoung-Sik Bang
Keyword(s):  
Nuclear Fuel ◽  
Structural Integrity ◽  
Transport Condition ◽  
Normal Transport

Structural Integrity of Spent Nuclear Fuel Storage Casks Subjected to Drop

1996 ◽  
Vol 114 (1) ◽  
pp. 111-121
Author(s):  
John R. Stokley ◽  
David H. Williamson
Keyword(s):  
Nuclear Fuel ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Fuel Storage

scholarly journals Toward Robotic Inspection of Dry Storage Casks for Spent Nuclear Fuel

2017 ◽  
Vol 139 (3) ◽  
Author(s):  
C. J. Lissenden ◽  
S. Choi ◽  
H. Cho ◽  
A. Motta ◽  
K. Hartig ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Crack Detection ◽  
Guided Waves ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Guided Wave ◽  
Radiation Environment ◽  
Breakdown Spectroscopy ◽  
Dry Storage ◽  
Horizontal Wave

Extended dry storage of spent nuclear fuel makes it desirable to assess the structural integrity of the storage canisters. Stress corrosion cracking of the stainless steel canister is a potential degradation mode especially in marine environments. Sensing technologies are being developed with the aim of detecting the presence of chloride-bearing salts on the surface of the canister as well as whether cracks exist. Laser-induced breakdown spectroscopy (LIBS) methods for the detection of Chlorine are presented. In addition, ultrasonic-guided wave detection of crack-like notches oriented either parallel or perpendicular to the shear horizontal wave vector is demonstrated using the pulse-echo mode, which greatly simplifies the robotic delivery of the noncontact electromagnetic acoustic transducers (EMATs). Robotic delivery of both EMATs and the LIBS system is necessary due to the high temperature and radiation environment inside the cask where the measurements need to be made. Furthermore, the space to make the measurements is very constrained and maneuverability is confined by the geometry of the storage cask. In fact, a large portion of the canister surface is inaccessible due to the presence of guide channels on the inside of the cask's overpack, which is strong motivation for using guided waves for crack detection. Among the design requirements for the robotic system are to localize and track where sensor measurements are made to enable return to those locations, to avoid wedging or jamming of the robot, and to tolerate high temperatures and radiation levels.

Robotic Inspection of Dry Storage Casks for Spent Nuclear Fuel

2016 ◽  
Author(s):  
C. J. Lissenden ◽  
S. Choi ◽  
H. Cho ◽  
A. Motta ◽  
K. Hartig ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Crack Detection ◽  
Guided Waves ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Radiation Environment ◽  
Breakdown Spectroscopy ◽  
Dry Storage ◽  
Laser Induced Breakdown ◽  
Horizontal Wave

Extended dry storage of spent nuclear fuel makes it desirable to assess the structural integrity of the storage canisters. Stress corrosion cracking of the stainless steel canister is a potential degradation mode especially in marine environments. Sensing technologies are being developed with the aim of detecting the presence of chloride-bearing salts on the surface of the canister as well as whether cracks exist. Laser induced breakdown spectroscopy (LIBS) methods for the detection of Chlorine are presented. Detection of a notch oriented either parallel or perpendicular to the shear horizontal wave vector is demonstrated using the pulse-echo mode, which greatly simplifies the robotic delivery of the noncontact electromagnetic acoustic transducers (EMATs). Robotic delivery of both EMATs and the LIBS system is necessary due to the high temperature and radiation environment inside the cask where the measurements need to be made. Furthermore, the space to make the measurement is very constrained and maneuverability is confined by the geometry of the storage cask. In fact, a large portion of the canister surface is inaccessible due to the presence of guide channels on the inside of the cask’s overpack, which is strong motivation for using guided waves for crack detection. Among the design requirements for the robotic system are: to localize and track where sensor measurements are made to enable return to those locations, to avoid wedging or jamming of the robot, and to tolerate high temperatures and radiation levels.

Japanese Codes and Standards for Dual Purpose Metal Casks for Spent Nuclear Fuel

2014 ◽  
Author(s):  
Toshiari Saegusa ◽  
Makoto Hirose ◽  
Norikazu Irie ◽  
Masashi Shimizu
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Spent Fuel ◽  
Dual Purpose ◽  
Service Conditions ◽  
Interim Storage ◽  
And Storage

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied. The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME. JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added. The Code has differences from that for NPP components with considerations to DPC characteristics; - The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1). - Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function. - Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages. - Stress limits for earthquakes during storage are specified. - Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels. DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.

scholarly journals Structural integrity investigations of spent nuclear fuel with finite element modeling

2021 ◽  
Vol 1 ◽  
pp. 13-14
Author(s):  
Efstathios Vlassopoulos ◽  
Susanne Pudollek ◽  
Olympios Alifieris ◽  
Dimitrios Papaioannou ◽  
Ramil Nasyrow ◽  
...  
Keyword(s):  
Finite Element ◽  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Structural Integrity ◽  
Mechanical Response ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Final Disposal ◽  
National Cooperative

Abstract. Radioactive waste in Switzerland will be disposed of in a deep geological repository (DGR). Responsible for the planning and preparation of realization of this task is National Cooperative for the Disposal of Radioactive Waste (Nagra). Spent fuel assemblies (SFA) constitute the main high-level waste (HLW) stream that will be disposed in the DGR. Prior to final disposal they will be transferred or transported to an encapsulation plant, where they will be loaded into final disposal canisters. To ensure that the structural integrity of SFAs is not compromised during handling and transportation, it is desirable to characterize the expected mechanical parameters of SFAs after long-term interim storage. Experimental research activities performed at the JRC Karlsruhe include safety aspects of radioactive waste management, encompassing also spent fuel storage and spent fuel/HLW disposal activities. Nagra and JRC have established a collaboration to jointly study relevant properties and behaviours of spent fuel rods, with the support of the Gösgen nuclear power plant and of Framatome, and in collaboration with other partners in Europe and internationally. As part of this collaboration, 3-point bending and impact tests were performed at the hot-cell facilities of JRC Karlsruhe, to determine the mechanical response of spent fuel rodlets under quasi-static and dynamic loads. The structural integrity of fuel rods was also evaluated under different handling scenarios using finite element (FE) analysis. Starting with the construction of a static 3D FE model of a Pressurized Water Reactor (PWR) nuclear fuel rodlet in ANSYS Mechanical, Nagra has developed a series of FE models over the years. Mechanical properties of the original rodlet model were derived through an extensive validation process, using experimental data from the 3-point bending tests. To evaluate the mechanical response of an SFA in different loading scenarios, this model was expanded using 1D beam modeling approach. The development of the simplified 1D models is shown in this presentation. In particular, the effect of the contact formulation between the spacer grid and the rods is discussed. Finally, preliminary results of the bending response of a 15×15 PWR SFA sub-model are presented.

Thermal Analysis of Spent Fuel Transportation Cask

2006 ◽  
Author(s):  
Seik Mansoor Ali ◽  
P. Goyal ◽  
Vishnu Verma ◽  
A. K. Ghosh ◽  
H. S. Kushwaha
Keyword(s):  
Finite Element Method ◽  
Thermal Analysis ◽  
Fire Test ◽  
Spent Fuel ◽  
The Finite Element Method ◽  
Thermal Test ◽  
Test Conditions ◽  
Transport Condition ◽  
Normal Transport ◽  
Transient Thermal

Spent fuel transportation casks are required to meet among others, the regulatory thermal test conditions in order to demonstrate their ability to withstand specified accidental fire conditions during transport. This paper describes the transient thermal analysis performed with the above intention for a transportation cask that has not undergone drop test and is not damaged. The original dimensions are used for computations. The analysis was carried out using a heat conduction code employing the Finite Element Method (FEM). At the outset, a benchmark exercise was carried out with the present code to ascertain its capability to carry out calculations involving phase change. The benchmarking was done against the solution given by commercial CFD code STAR CD. After assessing the suitability of the FEM code, it was employed for thermal analysis of the transportation cask. The computation covers normal transport condition, half an hour fire test at 800°C and also the post fire cool down period. The objective of the analysis was to determine the maximum outer surface temperature for normal transport conditions and to assess the extent of melting of lead during fire and post fire period.

Safety Analysis of Irradiated Nuclear Fuel Transportation Container

2007 ◽  
Author(s):  
E. Uspuras ◽  
S. Rimkevicius
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Analysis ◽  
Normal Operation ◽  
Spent Fuel ◽  
Safety Level ◽  
Fuel Assemblies ◽  
Accident Conditions ◽  
Transportation Container

Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit 1 on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level.

Structural Intergrity Evaluation Approach for PWR Spent Nuclear Fuel

2010 ◽  
Author(s):  
Y. S. Nam ◽  
Y. H. Kim ◽  
K. L. Jeon ◽  
S. K. Lee ◽  
K. S. Choi ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Mechanical Characteristics ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Technical Information ◽  
Design Parameters ◽  
Evaluation Approach ◽  
Analysis Methodology ◽  
Hypothetical Accident

PWR fuel assembly (FA) experiences many changes from the time it is manufactured, loaded in the reactor and removed from the reactor for reprocessing or stroage etc. Any of these alterations which impact spent nuclear fuel (SNF) integrity should be considered to design a cask/canister. Regarding the cask/canister design, there could be a freedom to design a system that mitigates the forces transmitted to SNF and fuel rods. If the cask/canister design prevents or mitigates forces transmitted to its contents such that structural integrity is not significantly compromised, the detailed SNF properties are necessary to make a decision of the elaborated design parameters. An approach to those work formations is to analyze mechanical characteristics of structural components. Those informations are also used to evaluate hypothetical accident, to select limiting FA for cask/canister to accommodate various kinds of SNFs and to design transportation/storage system for SNFs. Especially, FA structural properties are a sort of essential data. Thus, in this paper, some approaches to evaluate SNF mechanical characteristics are suggested through the existing technical information review, some test data and the analysis methodology, and also closely study the mechanical characteristics of a representative SNF for its general comprehension.

scholarly journals A Framework to Develop Flaw Acceptance Criteria for Structural Integrity Assessment of Multipurpose Canisters for Extended Storage of Used Nuclear Fuel

2014 ◽  
Author(s):  
Poh-Sang Lam ◽  
Robert L. Sindelar ◽  
Andrew J. Duncan ◽  
Thad M. Adams
Keyword(s):  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Storage Period ◽  
Corrosion Cracking ◽  
External Loading ◽  
Acceptance Criteria ◽  
Integrity Assessment ◽  
Used Nuclear Fuel

A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

Sign in / Sign up

Export Citation Format

Share Document