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.

VALIDATION OF A MONTE CARLO MODEL OF THE FORK DETECTOR WITH A CALIBRATED NEUTRON SOURCE

2014 ◽  
Vol 27 ◽  
pp. 1460151
Author(s):  
ALESSANDRO BORELLA ◽  
LIVIU-CRISTIAN MIHAILESCU
Keyword(s):  
Nuclear Power ◽  
Monte Carlo Model ◽  
Measurement Data ◽  
Nuclear Research ◽  
Research Centre ◽  
Spent Fuel ◽  
The Past ◽  
Research Areas ◽  
Fuel Elements ◽  
Comparison Exercise

The investigation of experimental methods for safeguarding spent fuel elements is one of the research areas at the Belgian Nuclear Research Centre SCK•CEN. A version of the so-called Fork Detector has been designed at SCK•CEN and is in use at the Belgian Nuclear Power Plant of Doel for burnup determination purposes. The Fork Detector relies on passive neutron and gamma measurements for the assessment of the burnup and safeguards verification activities. In order to better evaluate and understand the method and in view to extend its capabilities, an effort to model the Fork detector was made with the code MCNPX. A validation of the model was done in the past using spent fuel measurement data. This paper reports about the measurements carried out at the Laboratory for Nuclear Calibrations (LNK) of SCK•CEN with a 252Cf source calibrated according to ISO 8529 standards. The experimental data are presented and compared with simulations. In the simulations, not only was the detector modeled but also the measurement room was taken into account based on the available design information. The results of this comparison exercise are also presented in this paper.

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.

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.

Materials Unconditional Release Practices

2001 ◽  
Author(s):  
Lucien Teunckens
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Waste Treatment ◽  
Nuclear Research ◽  
Electricity Production ◽  
Research Centre ◽  
Nuclear Facilities ◽  
Treatment Facilities ◽  
Interim Storage ◽  
Reprocessing Plant

Abstract Belgium started its nuclear programme quite early. The first installations were constructed in the fifties, and presently, more than 55% of the Belgian electricity production is provided by nuclear power plants. After 30 years of nuclear experience, Belgium started the decommissioning of nuclear facilities in the eighties with two main projects: the BR3-PWR plant and the Eurochemic reprocessing plant. The BR3-decommissioning project is carried out at the Belgian Nuclear Research Centre, while the decommissioning of the former Eurochemic reprocessing plant is managed and operated by Belgoprocess n.v., which is also operating the centralised waste treatment facilities and the interim storage for Belgian radioactive waste.

Danube Field Excursion 1988: Tritium Content of River Water; Radioactivity of Danube Sediments

1990 ◽  
Vol 22 (5) ◽  
pp. 203-210 ◽  
Author(s):  
D. Rank ◽  
F. J. Maringer ◽  
W. Papesch ◽  
V. Rajner
Keyword(s):  
River Water ◽  
Bottom Sediments ◽  
Nuclear Power ◽  
Radioactive Material ◽  
Reactor Accident ◽  
Tritium Content ◽  
Fine Sediments ◽  
Fish Samples ◽  
Field Excursion

Water, sediment, and fish samples were collected during the Danube excursion 1988, within a coordinated sampling program of the Radiology Working Group of the “Internationale Arbeitsgemeinschaft Donauforschung ” (K.Hübel, Munich; I. Kurcz, Budapest; D.Rank, Vienna). The H-3 content of the river water and the radioactivity of the bottom sediments were measured at the BVFA Arsenal, Vienna. The determined H-3 content of the Danube water corresponds with the long-term trend in the H-3 content of the hydrosphere; the values lie in the range of 3 Bq/kg downstream from Belgrade, upstream from Belgrade they are about 4 Bq/kg. It was only in the waste water plume of the nuclear power station of Kozloduj that a slightly elevated H-3 value - 6 Bq/kg - was determined. The content of the sediments of artificial radionuclides was found, at the time of the Danube field excursion, to be almost exclusively due to the radioactive material released following the reactor accident at Chernobyl in April 1986 (mainly Cs-137 and Cs-134). As a consequence of the air currents and precipitation conditions prevailing at the time of the accident, the bottom sediments in the lower course of the Danube were less contaminated than those in the upper course. The fine sediments were found to contain over 3000 Bq/kg of Cs-137 in the upper course of the Danube.

Last Developments About Spent Fuel Management in Belgium

2001 ◽  
Author(s):  
V. Wittebolle
Keyword(s):  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Research Centre ◽  
Total Output ◽  
Storage Facility ◽  
Spent Fuel ◽  
Fuel Management ◽  
Dry Storage ◽  
Wet Storage ◽  
Interim Storage

Abstract In Belgium 57% of the electricity is presently generated by 7 nuclear units of the PWR type located in Doel and Tihange. Their total output amounts to 5632 MWe. Part of the spent fuel unloaded from the first three units has been sent till 2000 for reprocessing in the Cogema facility at La Hague. As the reprocessing of the spent fuel produced by the last four units is not covered by the contracts concluded with Cogema, Synatom, the Belgian utilities’ subsidiary in charge of the front- and back-end of the nuclear fuel cycle for all PWR reactors in Belgium, decided to study the possible solutions for a temporary storage of this spent fuel. End of 1993, the Belgian government decided that reprocessing (closed cycle) and direct disposal (open cycle) of spent fuel had to be considered as equal options in the back-end policy for nuclear fuel in Belgium. The resolution further allowed continued execution of a running reprocessing contract (from 1978) and use of the corresponding Pu for MOX in Belgian NPP’s, but requested a reprocessing contract concluded in 1990 (for reprocessing services after 2000) not to be executed during a five-year period. During this period priority was to be given to studies on the once-through cycle as an option for spent fuel management. Figure 1 is a chart showing the two alternatives for the spent fuel cycle in Belgium. In this context, Synatom entrusted Belgatom1 to develop a dedicated flask (called “bottle”) for direct disposal of spent fuel, to perform a design study of an appropriate encapsulation process and to prepare a preliminary feasibility study of a complete spent fuel conditioning plant. Meanwhile preparation works were made for the construction of an interim storage facility on both NPP sites of Doel and Tihange in order to meet increasing storage capacity needs. For selecting the type of interim storage facility, Belgatom performed a technical-economical analysis. Considerations of design and safety criteria as well as flexibility, reversibility, technical constraints, global economical aspects and construction time led to adopt dry storage with dual purpose casks (in operation since end 1995) for the Doel site and wet storage in a modular pool for the Tihange site (in operation since 1997). In parallel, ONRAF/NIRAS, the Belgian Agency for the management of radioactive waste and enriched fissile materials and the Belgian nuclear research centre, SCK•CEN, conduct underground investigations in view of geological disposal. The paper describes the methodology that Belgatom has developed to provide the utilities with appropriate solutions (reracking, dry storage in casks, wet storage in ponds, etc.) and how Belgatom demonstrated also the feasibility of spent fuel conditioning with a view to direct disposal in clay layers. The spent fuel storage facilities in operation in Belgium and designed and built by Belgatom are then briefly presented.

Special Waste Generated by the Belgian Nuclear Research Centre

2001 ◽  
Author(s):  
Veerle Van Alsenoy ◽  
Yves Demeulemeester ◽  
Luc Noynaert ◽  
Michel Klein ◽  
Nicolas Lardot
Keyword(s):  
Waste Treatment ◽  
Nuclear Research ◽  
Research Centre ◽  
Current Position ◽  
Joint Effort ◽  
Waste Forms ◽  
Need To Evaluate ◽  
Special Studies ◽  
And Storage ◽  
Research Centres

Abstract At research centres, test programs generate very particular waste forms, which cannot be treated using the known regular scenario’s and procedures, and therefore special studies need to evaluate safe handling of these special waste forms. This paper describes how a joint effort between the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, the Belgian Waste Treatment and Storage Facility Belgoprocess, and the Belgian Nuclear Research Centre worked out solutions for the 1st beryllium moderator of the BR2 and the highly activated internal parts of the BR3. The current position regarding activated graphite and aluminium is also summarised.

Unit Dose Factors for Transportation of Radioactive Materials

2015 ◽  
Author(s):  
Kevin J. Connolly ◽  
Elena Kalinina
Keyword(s):  
Dose Rate ◽  
Nuclear Power ◽  
Large Scale ◽  
Spent Nuclear Fuel ◽  
Radioactive Material ◽  
Unit Distance ◽  
The Public ◽  
Environmental Impact Statements ◽  
Unit Dose ◽  
Interim Storage

It will be necessary in the future to transport spent nuclear fuel on a large-scale basis from nuclear power plant sites to interim storage and/or a repository. Shipments of radioactive material are required to comply with regulations limiting the dose rate to no more than 0.1 mSv (10 mrem) per hour at 2 meters from the sides of the vehicle transporting the package. Determining the resulting dose to the public will be necessary for a number of reasons (e.g., stakeholder concerns, environmental impact statements). In order to understand the dose consequence of such a transportation system, this paper describes a method for determining unit dose factors. These are defined as the dose to the public per unit distance traveled along a road, rail, or waterway from one shipment assuming unit values for the other route specific parameters. The actual dose to the public is calculated using unit dose factors, the dose rate due to the radiation field emanating from the package, and characteristics of the route itself. Route specific parameters include the speed of the conveyance, the population density, and characteristics of the environment surrounding the route; these are provided by a routing tool. Using these unit dose factors, in conjunction with a routing tool, it will be possible to quantify the collective dose to the public and understand the ramifications of choosing specific routes.

Developing a Standard Code for Spent Fuel and High Level Waste Containment Baskets

2008 ◽  
Author(s):  
Donald Wayne Lewis
Keyword(s):  
Radioactive Material ◽  
High Level Waste ◽  
Plant Design ◽  
Support Structures ◽  
Spent Fuel ◽  
Waste Containment ◽  
Internal Support ◽  
Division 1 ◽  
High Level ◽  
And Storage

ASME Section III, Division 3, “Containments for Transportation and Storage of Spent Nuclear Fuel and High Level Radioactive Material and Waste” currently addresses the design of transportation and storage containment shells but it has yet to address the containment internal support structure that holds the spent fuel or high level waste in place. However, the code for internal support structures, hereafter referred to by its common name “basket”, has been under development by ASME for the past 2 years. Development of the new code, to be known as Subsection WD, “Internal Support Structures” was deemed necessary because current containment system basket construction is a piecemeal approach using ASME Section III, Division 1, Subsection NF, “Supports” and/or ASME Section III, Division 1, Subsection NG, “Core Support Structures” or some other engineering method. Approvals for the various combinations are granted from the regulatory authority. The piecemeal approach tries to capture the critical elements important for a containment basket. However, Subsections NF and NG are based on nuclear power plant design which has different design goals than for a spent fuel or high level waste containment. The issuance of Subsection WD will ensure standardization of future containment baskets, assist the regulatory agency in the review and approval of the baskets, and ensure that the essential criteria in the basket related to spent fuel and high level waste storage transportation and disposal is adequately addressed. The purpose of the basket is primarily to ensure that the radioactive components in the containment are supported in a way as not to create a criticality event. Current acceptance is typically based on a no yield design that the containment manufactures all say is too conservative and based on unreasonable criteria. What should the basket design be based on, how should Subsection WD address them, etc.? The purpose of this paper is to inform interested parties of the progress that has been made in development of Subsection WD, what construction provisions it will initially include and what is planned for it, and when is it scheduled to be issued.

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