scholarly journals Finite Element Modeling and Analysis of Nonlinear Impact and Frictional Motion Responses Including Fluid—Structure Coupling Effects

1997 ◽  
Vol 4 (5-6) ◽  
pp. 311-325 ◽  
Author(s):  
Yong Zhao
Keyword(s):  
Finite Element ◽  
Nuclear Power ◽  
Time History ◽  
Spent Fuel ◽  
Free Standing ◽  
Surrounding Water ◽  
Fluid Structure ◽  
Modeling And Analysis ◽  
Motion Responses ◽  
Fuel Assemblies

A nonlinear three dimensional (3D) single rack model and a nonlinear 3D whole pool multi-rack model are developed for the spent fuel storage racks of a nuclear power plant (NPP) to determine impacts and frictional motion responses when subjected to 3D excitations from the supporting building floor. The submerged free standing rack system and surrounding water are coupled due to hydrodynamic fluid-structure interaction (FSI) using potential theory. The models developed have features that allow consideration of geometric and material nonlinearities including (1) the impacts of fuel assemblies to rack cells, a rack to adjacent racks or pool walls, and rack support legs to the pool floor; (2) the hydrodynamic coupling of fuel assemblies with their storing racks, and of a rack with adjacent racks, pool walls, and the pool floor; and (3) the dynamic motion behavior of rocking, twisting, and frictional sliding of rack modules. Using these models 3D nonlinear time history dynamic analyses are performed per the U.S. Nuclear Regulatory Commission (USNRC) criteria. Since few such modeling, analyses, and results using both the 3D single and whole pool multiple rack models are available in the literature, this paper emphasizes description of modeling and analysis techniques using the SOLVIA general purpose nonlinear finite element code. Typical response results with different Coulomb friction coefficients are presented and discussed.

The holistic risk analysis of spent fuel pool considering the fuel route risk

2021 ◽  
pp. 1748006X2199851
Author(s):  
Zhixin Xu ◽  
Ming Wang ◽  
Binyan Song ◽  
WenYu Hou ◽  
Chao Wang
Keyword(s):  
Nuclear Power ◽  
Comprehensive Evaluation ◽  
Point Of View ◽  
Spent Fuel ◽  
Nuclear Disaster ◽  
Evaluation Approach ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Large Water ◽  
Fukushima Nuclear Disaster

The Fukushima nuclear disaster has raised the importance on the reliability and risk research of the spent fuel pool (SFP), including the risk of internal events, fire, external hazards and so on. From a safety point of view, the low decay heat of the spent fuel assemblies and large water inventory in the SFP has made the accident progress goes very slow, but a large number of fuel assemblies are stored inside the spent fuel pool and without containment above the SFP building, it still has an unignored risk to the safety of the nuclear power plant. In this paper, a standardized approach for performing a holistic and comprehensive evaluation approach of the SFP risk based on the probabilistic safety analysis (PSA) method has been developed, including the Level 1 SFP PSA and Level 2 SFP PSA and external hazard PSA. The research scope of SFP PSA covers internal events, internal flooding, internal fires, external hazards and new risk source-fuel route risk is also included. The research will provide the risk insight of Spent Fuel Pool operation, and can help to make recommendation for the prevention and mitigation of SFP accidents which will be applicable for the SFP configuration risk management.

scholarly journals Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

2021 ◽  
Vol 7 (1) ◽  
pp. 9-13
Author(s):  
David A. Hakobyan ◽  
Victor I. Slobodchuk
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Cooling Mode ◽  
Temperature Conditions ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

Fluid-structure Interactions

1998 ◽  
Vol 120 (04) ◽  
pp. 66-68 ◽  
Author(s):  
Klaus-Ju¨rgen Bathe
Keyword(s):  
Finite Element ◽  
Fluid Structure Interactions ◽  
Physical Processes ◽  
Fluid Structure ◽  
Modeling And Analysis ◽  
Structure Interaction ◽  
Design And Manufacturing ◽  
Physical Phenomena ◽  
Acoustic Fluid ◽  
Made In

This article reviews finite element methods that are widely used in the analysis of solids and structures, and they provide great benefits in product design. In fact, with today’s highly competitive design and manufacturing markets, it is nearly impossible to ignore the advances that have been made in the computer analysis of structures without losing an edge in innovation and productivity. Various commercial finite-element programs are widely used and have proven to be indispensable in designing safer, more economical products. Applications of acoustic-fluid/structure interactions are found whenever the fluid can be modeled to be inviscid and to undergo only relatively small particle motions. The interplay between finite-element modeling and analysis with the recognition and understanding of new physical phenomena will advance the understanding of physical processes. This will lead to increasingly better simulations. Based on current technology and realistic expectations of further hardware and software developments, a tremendous future for fluid–structure interaction applications lies ahead.

Model of a Dry Storage Facility for a Medium Nuclear Power Plant

2001 ◽  
Author(s):  
Mile Bace ◽  
Kresimir Trontl ◽  
Dubravko Pevec
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Radiation Dose ◽  
Dose Rate ◽  
Nuclear Power ◽  
Storage Facility ◽  
Spent Fuel ◽  
Radiation Dose Rate ◽  
Dry Storage ◽  
Fuel Assemblies

Abstract The intention was to model a dry storage facility that could satisfy the needs of a medium nuclear power plant similar to the NPP Krsko. The attention has been focused on radiation dose rate analyses and criticality calculations. Using the SCALE 4.4 code package and modified QAD-CGGP code, we modeled a facility that satisfies the basic criteria for public radiation protection. The capacity of the storage is 1,400 spent fuel assemblies which is adequate for a forty years medium NPP lifetime.

Seismic Design of Free Standing Racks in Japanese Nuclear Power Plants

2017 ◽  
Author(s):  
Yu Takaki ◽  
Katsuhiko Taniguchi ◽  
Junichi Kishimoto ◽  
Akihisa Iwasaki ◽  
Yoshitsugu Nekomoto ◽  
...  
Keyword(s):  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Design Method ◽  
Seismic Resistance ◽  
Spent Fuel ◽  
Analysis Model ◽  
Free Standing ◽  
Scale Free ◽  
Seismic Design Method

The free standing racks are spent fuel storage racks with self-sustained structure without fixation to the pit floor or pit walls. If a free standing rack receives a force to move it due to an earthquake, the force acting on each member of the rack is reduced in compared to the floor-anchored racks owing to sliding of the free standing rack. Now it is planned to exchange the existing floor-anchored racks with the free standing racks to secure higher seismic resistance. In previous studies, efforts were made to establish a behavior analysis model that allows for evaluation of sliding and rocking behaviors of free standing racks and to make out a seismic design method based on an evaluation technique to evaluate, in a conservative manner, vibration test results of full-scale free standing racks. The free standing racks which consist of connected eight racks are designed with this seismic design method. It was confirmed that the free standing racks have enough seismic resistance by performing evaluation using the basic seismic motion and making an analysis on beyond the design event.

Recommendations for Conducting Probabilistic Risk Assessment of Spent Fuel Storage Facilities

2017 ◽  
Author(s):  
Sai Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong ◽  
Zhixin Xu
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Spent Fuel ◽  
Complementary Method ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Storage Units ◽  
Storage Facilities ◽  
Probabilistic Risk

Currently, the probabilistic risk assessments (PRA) for the nuclear power plant (NPP) sites are primarily focused on the reactor counterpart. However, evoked by the 2011 Fukushima Daiichi accident, it has been widely recognized that a complete site risk profile should not be confined to the reactor units, but should cover all the radiological sources in a site, e.g. spent fuel storage facilities. During the operation of the reactor units, the used fuel assemblies will be unloaded from the reactor core to the storage facilities in a continuous or periodical manner. Accident scenarios involving such facilities can occur with non-negligible frequencies and significant consequences, posing threat to public safety. Hence, the risk contributions from such scenarios should be carefully estimated and integrated into the safety goal evaluations. The spent fuel storage facilities can be categorized as two types: pool storage units and dry cask storage facilities. In the former type, spent fuel assemblies are stored in large pools inside or outside the reactor building, with the residual heat removed by natural or forced water circulation. The latter type, where air or inert gas circulation plays an important role, appear mostly as a complementary method, along with the pool storage units, to expand the plant’s storage capacity. For instance, at the Daiichi plant, there are several fuel pool units holding some fresh fuel and some used fuel, the latter awaiting for its transfer to the dry cask storage facilities on site. Note that, as well as in a joint manner, both storage facilities can be designed to serve the NPPs independently. As a fully developed method to identify potential risk in a logical and quantitative way, the framework of PRA can be generally applied to the spent fuel storage facilities with some special considerations. This paper is aimed at giving recommendations for the spent fuel storage facility PRAs, including (1) clarifying the analysis scope of risk from spent fuel storage facilities; (2) illustrating four key issues that determines such risk; (3) presenting three essential considerations when conducting PRAs to evaluate such risk. Also, this paper integrates the insights obtained from two representative case studies involving two NPP sites with different types of both fuel elements and storage facilities.

scholarly journals Approaches to Safety Justification for Loading of VSC-VVER Containers in ZNPP DSFSF

2019 ◽  
pp. 82-87
Author(s):  
Ya. Kostiushko ◽  
O. Dudka ◽  
Yu. Kovbasenko ◽  
A. Shepitchak
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Spent Fuel ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
The Government

The introduction of new fuel for nuclear power plants in Ukraine is related to obtaining a relevant license from the regulatory authority for nuclear and radiation safety of Ukraine. The same approach is used for spent nuclear fuel (SNF) management system. The dry spent fuel storage facility (DSFSF) is the first nuclear facility created for intermediate dry storage of SNF in Ukraine. According to the design based on dry ventilated container storage technology by Sierra Nuclear Corporation and Duke Engineering and Services, ventilated storage containers (VSC-VVER) filled with SNF of VVER-1000 are used, which are located on a special open concrete site. Containers VSC-VVER are modernized VSC-24 containers customized for hexagonal VVER-1000 spent fuel assemblies. The storage safety assessment methodology was created and improved directly during the licensing process. In addition, in accordance with the Energy Strategy of Ukraine up to 2035, one of the key task is the further diversification of nuclear fuel suppliers. Within the framework of the Executive Agreement between the Government of Ukraine and the U.S. Government, activities have been underway since 2000 on the introduction of Westinghouse fuel. The purpose of this project is to develop, supply and qualify alternative nuclear fuel compatible with fuel produced in Russia for Ukrainian NPPs. In addition, a supplementary approach to safety analysis report is being developed to justify feasibility of loading new fuel into the DSFSF containers. The stated results should demonstrate the fulfillment of design criteria under normal operating conditions, abnormal conditions and design-basis accidents of DSFSF components.  Thus, the paper highlights both the main problems of DSFSF licensing and obtaining permission for placing new fuel types in DSFSF.

scholarly journals Minimizing the Standard Deviation of the Thermal Load in the Spent Nuclear Fuel Cask Loading Problem

Energies ◽  
2020 ◽  
Vol 13 (18) ◽  
pp. 4869
Author(s):  
Joaquín Bautista-Valhondo ◽  
Lluís Batet ◽  
Manuel Mateo
Keyword(s):  
Standard Deviation ◽  
Nuclear Fuel ◽  
Thermal Load ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Search Algorithm ◽  
Deterministic Algorithm ◽  
Mixed Integer ◽  
Spent Fuel ◽  
Fuel Assemblies

The paper assumes that, at the end of the operational period of a Spanish nuclear power plant, an Independent Spent Fuel Storage Installation will be used for long-term storage. Spent fuel assemblies are selected and transferred to casks for dry storage, with a series of imposed restrictions (e.g., limiting the thermal load). In this context, we present a variant of the problem of spent nuclear fuel cask loading in one stage (i.e., the fuel is completely transferred from the spent fuel pool to the casks at once), offering a multi-start metaheuristic of three phases. (1) A mixed integer linear programming (MILP-1) model is used to minimize the cost of the casks required. (2) A deterministic algorithm (A1) assigns the spent fuel assemblies to a specific region of a specific cask based on an MILP-1 solution. (3) Starting from the A1 solutions, a local search algorithm (A2) minimizes the standard deviation of the thermal load among casks. Instances with 1200 fuel assemblies (and six intervals for the decay heat) are optimally solved by MILP-1 plus A1 in less than one second. Additionally, A2 gets a Pearson’s coefficient of variation lower than 0.75% in less than 260s CPU (1000 iterations).

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

2021 ◽  
Author(s):  
Davor Grgic ◽  
Mario Matijevic ◽  
Paulina Duckic ◽  
Radomir Jecmenica
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Variance Reduction ◽  
Spent Fuel ◽  
Dose Rates ◽  
Dry Storage ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
The Impact

Abstract In this paper shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW loaded with Spent Fuel Assemblies (SFA) from Nuclear Power Plant (NPP) Krsko Spent Fuel Dry Storage (SFDS) Campaign one. The HI-TRAC VW is a multi-layered cylindrical vessel designed to accept a Multi Purpose Canister (MPC) during loading, unloading and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using ORIGEN-S module of the SCALE code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron induced gammas (n-g reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, FW-CADIS variance reduction scheme, as implemented in ADVANTG code, was adopted for accelerating final MCNP6 calculations. The dose rates around HI-TRAC VW cask were calculated using MCNP6 code for all 16 casks loading belonging to Campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization (Holtec International).

The Sliding and Overturning Analysis of a Free-Standing Cask under Earthquake

2007 ◽  
Vol 120 ◽  
pp. 207-212 ◽  
Author(s):  
R.F. Shie ◽  
Kung Ting ◽  
J.S. Yu ◽  
C.T. Liu
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Computer Code ◽  
Storage Facility ◽  
Spent Fuel ◽  
Power Station ◽  
Free Standing ◽  
Safety Assurance ◽  
Maximum Earthquake ◽  
Interim Storage

A freestanding spent fuel interim storage facility is possible to be selected to house the spent fuel for ChinShan nuclear power station in a period of 5 to 50 years. The design specifications of the storage facility should provide adequate safety assurance under the event of a major earthquake. ANSYS/LS-DYNA computer code is employed in this study to simulate the dynamic behavior of the cask under extremely seismic loading, the criterion which is used to judge whether freestanding cask will slide or overturn is thus established. A cylindrical cask is used as an object which subjected to safety shutdown earthquake of support pad of ChinShan nuclear power station. The cylindrical cask will slide but will not overturn is obtained under maximum earthquake loading.

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