Drop Tests Assessment of Internal Shock Absorbers for Packages Loaded With Encapsulations for Damaged Spent Nuclear Fuel

2020 ◽  
Author(s):  
Lars Müller ◽  
R. Scheidemann ◽  
T. Schönfelder ◽  
S. Komann ◽  
F. Wille
Keyword(s):  
Nuclear Fuel ◽  
Shock Absorber ◽  
Spent Nuclear Fuel ◽  
Radioactive Material ◽  
Test Facility ◽  
Package Design ◽  
Shock Absorbers ◽  
Internal Shock ◽  
Fuel Assemblies ◽  
Drop Tests

Abstract Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts. BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber. The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs.

DROP TESTS ASSESSMENT OF INTERNAL SHOCK ABSORBERS FOR PACKAGES LOADED WITH ENCAPSULATIONS FOR DAMAGED SPENT NUCLEAR FUEL

2021 ◽  
Author(s):  
Lars Mueller ◽  
Robert Scheidemann ◽  
Thorsten Sch\xf6nfelder ◽  
Steffen Komann ◽  
Frank Wille
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Shock Absorbers ◽  
Internal Shock ◽  
Drop Tests

Mechanical and Thermal Assessment by BAM of a New Package Design for the Transport of SNF From a German Research Reactor

2020 ◽  
Author(s):  
Steffen Komann ◽  
Viktor Ballheimer ◽  
Thomas Quercetti ◽  
Robert Scheidemann ◽  
Frank Wille
Keyword(s):  
Research Reactor ◽  
Drop Test ◽  
Radioactive Material ◽  
Test Facility ◽  
Test Sequence ◽  
Thermal Safety ◽  
Package Design ◽  
Drop Tests ◽  
Approval Procedure ◽  
The Impact

Abstract For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged. This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package. For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence. The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design. The FE models include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters. Additional calculations by using the verified numerical model were done to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure. The thermal safety assessment was based on analytical energy balance calculations and FE analyses. As an additional point of evaluation in frame of approval procedure, the effect of possible impact limiter burning under accident conditions of transport was considered by the applicant and assessed by BAM. This paper describes the package design assessment from the point of view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods.

scholarly journals Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal

2021 ◽  
Vol 11 (14) ◽  
pp. 6499
Author(s):  
Matthias Frankl ◽  
Mathieu Hursin ◽  
Dimitri Rochman ◽  
Alexander Vasiliev ◽  
Hakim Ferroukhi
Keyword(s):  
Uncertainty Quantification ◽  
Nuclear Fuel ◽  
Evaluation Method ◽  
Spent Nuclear Fuel ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Evaluation Methodology ◽  
Geological Disposal ◽  
Fuel Assemblies ◽  
Criticality Safety

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

Dynamic analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

2013 ◽  
Vol 260 ◽  
pp. 155-164 ◽  
Author(s):  
Gintautas Dundulis ◽  
Albertas Grybenas ◽  
Renatas Karalevicius ◽  
Vidas Makarevicius ◽  
Sigitas Rimkevicius ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Experimental Research ◽  
Shock Absorber ◽  
Fuel Assemblies

Static analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

2009 ◽  
Vol 239 (1) ◽  
pp. 1-8 ◽  
Author(s):  
Gintautas Dundulis ◽  
Albertas Grybenas ◽  
Renatas Karalevicius ◽  
Vidas Makarevicius ◽  
Sigitas Rimkevicius ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Experimental Research ◽  
Shock Absorber ◽  
Fuel Assemblies

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 Conceptual Design of a MEDE Treatment System for Sodium Bonded Fuel

2008 ◽  
Author(s):  
Carl E. Baily ◽  
Karen A. Moore ◽  
Collin J. Knight ◽  
Peter B. Wells ◽  
Paul J. Petersen ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Department Of Energy ◽  
Test Facility ◽  
Evaporation Chamber ◽  
Oak Ridge ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Metal Fuel ◽  
Enriched Uranium

Unirradiated sodium bonded metal fuel and casting scrap material containing highly enriched uranium (HEU) is stored at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL). This material, which includes intact fuel assemblies and elements from the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor-II (EBR-II) reactors, as well as scrap material from the casting of these fuels, has no current use under the terminated reactor programs for both facilities. The Department of Energy (DOE), under the Sodium-Bonded Spent Nuclear Fuel Treatment Record of Decision (ROD), has determined that this material could be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for commercial nuclear reactors. A plan is being developed to prepare, package, and transfer this material to the DOE HEU Disposition Program Office (HDPO), located at the Y-12 National Security Complex in Oak Ridge, Tennessee. Disposition of the sodium bonded material will require separating the elemental sodium from the metallic uranium fuel. A sodium distillation process known as MEDE (Melt-Drain-Evaporate), will be used for the separation process. The casting scrap material needs to be sorted to remove any foreign material or fines that are not acceptable to the HDPO program. Once all elements have been cut and loaded into baskets, they are then loaded into an evaporation chamber as the first step in the MEDE process. The chamber will be sealed and the pressure reduced to approximately 200 mtorr. The chamber will then be heated as high as 650 °C, causing the sodium to melt and then vaporize. The vapor phase sodium will be driven into an outlet line where it is condensed and drained into a receiver vessel. Once the evaporation operation is complete, the system is de-energized and returned to atmospheric pressure. This paper describes the MEDE process as well as a general overview of the furnace systems, as necessary, to complete the MEDE process.

New ASME Rules for Spent Nuclear Fuel and High Level Radioactive Material and Waste Storage Containments

2004 ◽  
Author(s):  
Ralph S. Hill ◽  
Gerald M. Foster
Keyword(s):  
Nuclear Fuel ◽  
Pressure Vessel ◽  
Spent Nuclear Fuel ◽  
The Body ◽  
Radioactive Material ◽  
Comprehensive Understanding ◽  
Waste Storage ◽  
American Society ◽  
High Level

In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) will be published. The new Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. Some time after publication, CC N-717 will be incorporated into the body of the Code. This paper provides an informative insight to the Code Case so that Owners, regulators, designers, and fabricators have a more comprehensive understanding.

Research for Crossing Strait Transportation Accident of Nuclear Fuel Assemblies

2013 ◽  
Author(s):  
Xiaoxiao Xu ◽  
Xuexin Wang ◽  
Jiangang Zhang ◽  
Chaoduan Li ◽  
Gao Fan ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Hainan Island ◽  
Radioactive Material ◽  
Sea Transport ◽  
Fuel Assemblies ◽  
Qiongzhou Strait ◽  
Transport Accidents

Hainan nuclear power plant (HNPP) is the first nuclear power plant built on China’s Hainan Island. Therefore the nuclear fuel assemblies must transport through the Qiongzhou Strait. There are two transportation plans to be used in crossing strait transportation of the fuel assemblies. One plan is railway ferry stretching across sea; the other is road vehicular crossing strait on roll-on/roll-off (Ro/Ro) ships. According to crossing strait transportation scenario and statistical analysis of sea transport accidents in Qiongzhou Strait, three ferrying transportation accidents are considered in this paper. Through research of ship-to-ship collision, fire and sunk, the following conclusions: Collision, fire or foundered are not caused by the leakage of radioactive material, the environmental impact is very small. The accident hazards of crossing strait transportation does not lie in the radiological consequences, but in the effects of public psychology and international repercussions.

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