Safety Aspects of Long-term Dry Interim Storage Of Type-b Spent Fuel and Hlw Transport Casks

2004 ◽  
Vol 15 (3-4) ◽  
pp. 207-214 ◽  
Author(s):  
D. Wolff ◽  
U. Probst ◽  
H. Völzke ◽  
B. Droste ◽  
R. Rödel
Keyword(s):  
Type B ◽  
Spent Fuel ◽  
Safety Aspects ◽  
Interim Storage

Safety Aspects of a Dry Interim Storage Facility for Spent Fuel Cernavoda NPP

1997 ◽  
pp. 171-175
Author(s):  
Maria Radu ◽  
Marcela Stanciu ◽  
Adrian Panait ◽  
Traian Barbu ◽  
Silvia Mateescu ◽  
...  
Keyword(s):  
Storage Facility ◽  
Spent Fuel ◽  
Safety Aspects ◽  
Interim Storage

Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine

2003 ◽  
Vol 807 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Dissolution Rate ◽  
Spent Fuel ◽  
Salt Solutions ◽  
Fuel Surface ◽  
Fuel Sample ◽  
Salt Brine ◽  
High Burnup ◽  
Interim Storage ◽  
The Impact

ABSTRACTDuring long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.

CONSTOR® Transport and Storage Cask for Spent Fuel and for High Active Waste

2001 ◽  
Author(s):  
Rudolf Diersch ◽  
Robert Gartz ◽  
Konrad Gluschke
Keyword(s):  
Nuclear Power ◽  
Type B ◽  
Spent Fuel ◽  
The Czech Republic ◽  
Active Waste ◽  
Intermediate Storage ◽  
Near Future ◽  
And Storage ◽  
Rbmk Fuel

Abstract The CONSTOR® was developed with special consideration to an economical and effective way of manufacturing by using conventional mechanical engineering technologies and common materials. The main objective of this development was to fabricate these casks in countries not having highly specialized industries for casting or forging of thick-walled casks. Nevertheless, the CONSTOR® concept fulfills both the internationally valid IAEA criteria for transportation and the German criteria for long-term intermediate storage. The basic cask concept has been designed for adaptation to different spent fuel specifications as well as handling conditions in the NPPs. Adaptations have been made for spent fuel from RBMK and VVER reactors, and also for BWR spent fuel and high active waste. So far, 30 CONSTOR® RBMK-1500 original casks have been manufactured and delivered to Ignalina Nuclear Power Plant in Lithuania. Two of these have been succesfully loaded during hot trial tests and placed in storage. The CONSTOR® cask for RBMK fuel has obtained the type B(U)F verification certificate from the Russian authority GAN and the release for manufacturing as a storage cask by the Lithuanian Authority VATESI. Following the successful hot trials, the final storage license from VATESI is expected in the near future. The type B(U)F licensing in the Czech Republic will be finished in 2001.

CASTOR® BR3: Design of a Transport and Storage Cask for Spent Fuel From a Belgian Nuclear Power Plant

2001 ◽  
Author(s):  
Ulrich Knopp
Keyword(s):  
Nuclear Power ◽  
Storage Period ◽  
Nuclear Research ◽  
Ductile Cast Iron ◽  
Radioactive Material ◽  
Research Centre ◽  
Spent Fuel ◽  
Interim Storage ◽  
And Storage

Abstract The CASTOR® BR3 cask has been designed and manufactured to accomodate irradiated fuel (U and MOX) from the BR3 test reactor at the nuclear research centre SCK/CEN in Dessel near Mol, Belgium, which is currently being dismantled. The CASTOR® BR3 is designed as a Type B(U)F package for transport and will be licensed in Belgium. In addition, the CASTOR® BR3 needs a license as a storage cask to be operated in an interim cask storage facility. To obtain these licenses, the cask design has to observe the international regulations for the safe transport of radioactive material as well as the special requirements for the cask storage. The CASTOR® BR3 is a member of the CASTOR® family of spent fuel casks, delivered by the German company GNB. In this way, the cask has such typical features as the following: • monolithic cask body made of ductile cast iron; • double-lid system consisting of primary and secondary lid for long-term interim storage of the fuel. This family of casks has been used for over 20 years for transport and storage of spent fuel. In this paper, the IAEA regulatory requirements for transport casks are summarized and it is shown by selected examples how these requirements have been converted into the cask design and the analyses performed for the cask. Finally, the cask features for an interim storage period of up to 50 years will be spotlighted. Main topics are the evaluation of the long term behaviour of selected cask components and the cask monitoring system for the surveillance of the leak tightness of the cask during the storage period.

Extending Periodic Leakage Rate Testing of 9977 Packages With Elastomeric O-Rings

2014 ◽  
Author(s):  
Zenghu Han ◽  
Vikram N. Shah ◽  
Yung Y. Liu
Keyword(s):  
Leakage Rate ◽  
Radioactive Material ◽  
Ring Test ◽  
Type B ◽  
Testing Interval ◽  
Rate Testing ◽  
Arrhenius Function ◽  
Material Transportation ◽  
Interim Storage

According to ANSI N14.5, the periodic leakage rate testing of Type B radioactive material transportation packages is performed within 12 months prior to each shipment. The purpose of performing periodic leakage rate testing is to confirm that packages built to an approved design can perform their containment function as required after a period of use. However, certain transportation packages, e.g., Model 9975 and 9977 Type B packages, have been used for interim storage for a period > 12 months, and it is desirable to extend the periodic leakage rate testing interval to reduce personnel radiation exposure and cost. Long-term leak performance tests on O-ring test fixtures have been conducted at 200°F (366K) and higher temperatures since 2004 for the purpose of interim storage of 9975 packages. The test data are adopted and evaluated in this paper by using the Arrhenius function and the Weibull statistics to establish the basis for extending the periodic leakage rate testing interval. The results show that the testing interval can be extended to 5 and 2 years for Model 9977 packages with Viton® GLT and GLT-S elastomeric O-rings (Parker Seals V0835-75 and VM835-75), respectively, if the O-ring service temperature is kept below 200°F (366K) and verified with continuous temperature monitoring.

scholarly journals Impact of Radiation-Induced Microstructures on the Integrity of Spent Nuclear Fuel (SNF) Elements in Long-Term Storage

2021 ◽  
Vol 1 ◽  
pp. 17-18
Author(s):  
Neslihan Yanikömer ◽  
Rahim Nabbi ◽  
Klaus Fischer-Appelt
Keyword(s):  
Mechanical Properties ◽  
Monte Carlo ◽  
Radiation Damage ◽  
Spent Fuel ◽  
Radiation Induced ◽  
Primary Damage ◽  
Interim Storage ◽  
Fuel Elements ◽  
Cladding Material

Abstract. The current safety concept provides for a period in the range of 40 years for interim storage of spent fuel elements. Since the requirement for proof of safety for to up to 100 years arises, the integrity of the spent fuel elements in prolonged interim storage and long-term repositories is becoming a critical issue. In response to this safety matter, this study aims to assess the impact of radiation-induced microstructures on the mechanical properties of spent fuel elements, in order to provide reliable structural performance limits and safety margins. The physical processes involved in radiation damage and the effect of radiation damage on mechanical properties are inherently multiscalar and hierarchical. Damage evolution under irradiation begins at the atomic scale, with primary knock-on atoms (PKAs) resulting in displacement cascades (primary damage), followed by the defect clusters leading to microstructural deformations. In this context, we have developed and applied a multiscale simulation methodology consistent with the multistage damage mechanisms and the corresponding effects on the mechanical properties of spent fuel cladding and its integrity. Within the improved hierarchical modelling sequence, the effect of the radiation field on the fuel element cladding material (Zircalloy-4) is assessed using Monte Carlo methods. A molecular dynamics method is employed to model damage formation by PKAs and primary damage defect configurations. The formation of clusters and evolution of microstructures are simulated by extending the simulation sequence to a longer time scale with the kinetic Monte Carlo (KMC) method. Transferring the calculated radiation-induced microstructures into macroscopic quantities is ultimately decisive for the structural/mechanical behaviour and stability of the cladding material, and thus for long-term integrity of the spent fuel elements. Results of the multiscale modelling and simulations as well as a comparison with experimental results will be presented at the conference session.

scholarly journals Long-Term Radiotoxicity Evaluation of PWR Spent Uranium and MOX Fuel and Highly Active Waste

2020 ◽  
Vol 207 ◽  
pp. 01024
Author(s):  
Petar Paunov ◽  
Ivaylo Naydenov
Keyword(s):  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Potential Danger ◽  
Spent Fuel ◽  
Highly Active ◽  
Active Waste ◽  
Interim Storage ◽  
The Given ◽  
Nuclear Power Production

One of the main concerns related to nuclear power production is the generation and accumulation of spent nuclear fuel. Currently most of the spent fuel is stored in interim storage facilities awaiting final disposal or reprocessing. The spent fuel is stored in isolation from the environment in protected facilities or specially designed containers. Nevertheless, spent fuel and highly active waste might get in the environment in case the protective barriers are compromised. In such a case, spent fuel may pose risk to the environment and human health. Those risks depend on the concentration of the given radionuclide and are measured by the value of potential danger. The potential danger is called also ’radiotoxicity’. The paper examines spent uranium and MOX fuels from a reference PWR, as well as the highly radioactive wastes of their reprocessing. The radiotoxicity of the four materials is examined and evaluated for a cooling time of 1000 years. The contribution of different radionuclides is assessed and the cases of reprocessing and no reprocessing of spent fuel have been compared.

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