In-House Developed Methodologies and Tools for Decommissioning Projects

2007 ◽  
Author(s):  
Michel Detilleux ◽  
Baudouin Centner
Keyword(s):  
Nuclear Power ◽  
Assessment Tool ◽  
Nuclear Research ◽  
Research Centre ◽  
Primary System ◽  
Time Schedule ◽  
Operational Parameters ◽  
Dose Rates ◽  
User Friendly

The paper describes different methodologies and tools developed in-house by Tractebel Engineering to facilitate the engineering works to be carried out especially in the frame of decommissioning projects. Three examples of tools with their corresponding results are presented: - The LLWAA-DECOM code, a software developed for the radiological characterization of contaminated systems and equipment. The code constitutes a specific module of more general software that was originally developed to characterize radioactive waste streams in order to be able to declare the radiological inventory of critical nuclides, in particular difficult-to-measure radionuclides, to the Authorities. In the case of LLWAA-DECOM, deposited activities inside contaminated equipment (piping, tanks, heat exchangers ...) and scaling factors between nuclides, at any given time of the decommissioning time schedule, are calculated on the basis of physical characteristics of the systems and of operational parameters of the nuclear power plant. This methodology was applied to assess decommissioning costs of Belgian NPPs, to characterize the primary system of Trino NPP in Italy, to characterize the equipment of miscellaneous circuits of Ignalina NPP and of Kozloduy unit 1 and, to calculate remaining dose rates around equipment in the frame of the preparation of decommissioning activities; - The VISIMODELLER tool, a user friendly CAD interface developed to ease the introduction of lay-out areas in a software named VISIPLAN. VISIPLAN is a 3D dose rate assessment tool for ALARA work planning, developed by the Belgian Nuclear Research Centre SCK·CEN. Both softwares were used for projects such as the steam generators replacements in Belgian NPPs or the preparation of the decommissioning of units 1&2 of Kozloduy NPP; - The DBS software, a software developed to manage the different kinds of activities that are part of the general time schedule of a decommissioning project. For each activity, when relevant, algorithms allow to estimate, on the basis of local inputs, radiological exposures of the operators (collective and individual doses), production of primary, secondary and tertiary waste and their characterization, production of conditioned waste, release of effluents, ... and enable the calculation and the presentation (histograms) of the global results for all activities together. An example of application in the frame of the Ignalina decommissioning project is given.

scholarly journals RADIOLOGICAL ASSESSMENT OF IONIZING RADIATION IMPACT ON THE TERRESTRIAL NON-HUMAN BIOTA IN LITHUANIA

2015 ◽  
Vol 23 (4) ◽  
pp. 295-301 ◽  
Author(s):  
Marina KONSTANTINOVA ◽  
Nina PROKOPČIUK ◽  
Arūnas GUDELIS ◽  
Donatas BUTKUS
Keyword(s):  
Ionizing Radiation ◽  
Nuclear Power ◽  
Activity Concentration ◽  
Assessment Tool ◽  
Natural Radionuclides ◽  
Soil Layer ◽  
Software Application ◽  
Dose Rates ◽  
Top Soil

The quantitative assessment of radionuclides transfer to non-human biota using their activity concentration ratios is required for models of predictive doses of ionizing radiation. Based on long-term data regarding activity concentration of radionuclides in the top soil layer of the entire territory of Lithuania, and with the help of ERICA Assessment Tool – a software application that calculates dose rates to selected biota, we estimated the radiological impact on the terrestrial non-human biota with special emphasis on the protected areas located in the vicinity of Ignalina Nuclear Power Plant (INPP). Estimated total dose rates of artificial radionuclides – after-Chernobyl 137Cs and 90Sr as well as discharged by INPP – and natural radionuclides, such as 238U and 232Th, were found to be less than ERICA screening value of 10 μGy h–1.

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.

Radiological Characterization of the Nuclear Waste Streams of the Belgian Nuclear Research Centre SCK•CEN

2007 ◽  
Author(s):  
Patrick Maris ◽  
Rene´ Cornelissen ◽  
Michel Bruggeman
Keyword(s):  
Nuclear Research ◽  
Physical Models ◽  
Research Centre ◽  
Waste Streams ◽  
Activity Ratios ◽  
Disposal Option ◽  
The Many ◽  
Non Destructive ◽  
Radiological Characterization

The radiological characterization of nuclear wastes of a research centre is difficult seen the many different processes that generate waste. Since these wastes may contain radionuclides relevant for the disposal option, the nuclide content and activity have to be known. Considering the fact that some wastes are generated only in minor quantities, complex approaches, involving sampling and successive analysis are not justified. Basic physical models can generally be applied to estimate activity ratios, from which the radionuclide inventory can be determined by non-destructive assay on waste-packages. This article discusses waste streams at the Belgian Nuclear Research Centre SCK•CEN and explains how nuclide inventories and activity are determined. The physical models, used to derive activity ratios, and other simple approaches are discussed.

Spanish Approach to Large Decommissioning Projects

2011 ◽  
Author(s):  
Juan Luis Santiago ◽  
Alejandro Rodri´guez
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Research ◽  
Lessons Learned ◽  
Research Centre ◽  
Nuclear Facilities ◽  
Material Management ◽  
Site Remediation ◽  
Key Aspects

The Spanish experience related to the decommissioning of nuclear facilities includes the decommissioning of the Vandello´s I Nuclear Power Plant, the decommissioning of the CIEMAT Nuclear Research Centre and the decommissioning of the Jose´ Cabrera Nuclear Power Plant. This paper reviews the key aspects of these projects and describes the lessons learned related to preparatory activities, auxiliary facilities, decommissioning technologies, material management and site remediation and release.

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.

The Involvement of SCK•CEN in Decommissioning Projects

2013 ◽  
Author(s):  
Luc Noynaert ◽  
Jérôme Dadoumont ◽  
René Cornelissen ◽  
Kurt Van den Dungen
Keyword(s):  
Nuclear Power ◽  
Nuclear Research ◽  
Nuclear Material ◽  
Research Centre ◽  
Pressurized Water Reactor ◽  
Rate Reduction ◽  
Pressurized Water ◽  
Know How ◽  
Reactor Technology ◽  
Nuclear Power Production

SCK•CEN is the Belgian Nuclear Research Centre. Founded in the mid-fifties, it has accumulated experience and know-how in all fields of the nuclear power production: in the neutronics calculation, radiation protection, waste management, fuel performance and analysis, nuclear measurements, radiochemistry, reactor technology, etc. Since 1989, SCK•CEN has launched Decommissioning activities to deal with the Technical Liabilities created by 40 years of operation. The main projects started were: • the dismantling of the BR3 PWR reactor; • the dismantling of active laboratories and the decontamination of buildings for unrestricted reuse; • the management of the waste arising from the refurbishment activities of the BR2, especially the management of the high active beryllium matrix. In 1989, the BR3 reactor, a Pressurized Water Reactor, was selected by the European Commission as one of the four pilot dismantling projects in the framework of the EC five year RTD program on dismantling nuclear installations. Through this project, SCK•CEN has built up a broad know-how in dismantling and decommissioning operations. This know how concerns the decontamination for dose rate reduction and/or free release of materials, teleoperated techniques for cutting highly activated components of a reactor, concrete decontamination techniques, characterization techniques of radioactive waste or for free release of components and development of decommissioning management and recordkeeping of material streams and of nuclear material accountancy. SCK•CEN is now actively involved in other decommissioning projects in Belgium and in expertise abroad. After giving an overview of the main achievements and the perspectives of the decommissioning of the BR3 reactor, the paper intends to present the involvement of SCK•CEN in the other decommissioning projects and to give an overview of our activities and capacities.

scholarly journals Characterization of gamma field in the JSI TRIGA reactor

2018 ◽  
Vol 170 ◽  
pp. 04001 ◽  
Author(s):  
Klemen Ambrožič ◽  
Vladimir Radulović ◽  
Luka Snoj ◽  
Adrien Gruel ◽  
Mael Le Guillou ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Electronic Devices ◽  
Field Measurements ◽  
Operational Parameters ◽  
Triga Reactor ◽  
Ongoing Work ◽  
Recent Developments ◽  
Stefan Institute ◽  
Gamma Field

Research reactors such as the “Jožzef Stefan” Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR’s structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

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.

IDEA: An Integrated Set of Tools for Sustainable Nuclear Decommissioning Projects

2009 ◽  
Author(s):  
Pierre Cnapelinckx ◽  
Fanny Castillo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Radioactive Waste ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
Research Centre ◽  
Time Schedule ◽  
Waste Streams ◽  
Computer Tools ◽  
Difficult To Measure Radionuclides

Projects Decommissioning of nuclear installations constitutes an important challenge and shall prove to the public that the whole nuclear life cycle is fully mastered by the nuclear industry. When ceasing operation, nuclear installations owners and operators are looking for solutions in order to assess and keep decommissioning costs at a reasonable level, to fully characterize waste streams (in particular radiological inventories of difficult-to-measure radionuclides) and to reduce personnel exposure during the decommissioning activities taking into account several project, site and country specific constraints. In response to this need, Tractebel Engineering has developed IDEA (Integrated DEcommissioning Application), an integrated set of computer tools, to support the engineering activities to be carried out in the frame of a decommissioning project. IDEA provides optimized solutions from an economical, environmental, social and safety perspective. IDEA is based on the integration of the following computer tools: LLWAA-DECOM, VISIMODELLER/VISIPLAN and DBS. The LLWAA-DECOM module has been developed for the radiological characterization of contaminated systems and equipment. The module constitutes a specific part of more general software that was originally developed to characterize NPP radioactive waste streams in order to assist the Operators when declaring the radiological inventory of critical nuclides, in particular difficult-to-measure radionuclides, to the Authorities. In the case of LLWAA-DECOM, deposited activities inside contaminated equipment (piping, tanks, heat exchangers ...) and scaling factors between nuclides, at any given time of the decommissioning time schedule, are calculated on the basis of physical characteristics of the systems and of operational parameters of the nuclear power plant. The VISIMODELLER tool, a user friendly CAD interface developed to ease the introduction of lay-out areas in a software named VISIPLAN. VISIPLAN is a 3D dose rate assessment tool for ALARA work planning, developed by the Belgian Nuclear Research Centre SCK•CEN. The DBS computer tool has been developed to manage the different kinds of activities that are part of the general time schedule of a decommissioning project. For each activity, when relevant, algorithms allow to estimate, on the basis of local inputs, radiological exposures of the operators (collective and individual doses), production of primary, secondary and tertiary waste and their characterization, production of conditioned waste, release of effluents, ... and enable the calculation and the presentation (histograms) of the global results for all activities together. Based on design and operating data from the Nuclear Power Plant to be dismantled and on the specificities of the country regarding radioactive waste management and disposal routes, IDEA will enable to prepare and manage a decommissioning project, in a sustainable way, leading to a greenfield or a reuse of the nuclear site after decommissioning of the plant. Moreover thanks to the characterization and definition of the optimal waste treatment and conditioning techniques, IDEA contributes to the long term safe management of the radioactive waste.

“Real-World” Microscopy: Understanding Environment-Sensitive Behavior of Ni-Base Welds

1998 ◽  
Vol 4 (S2) ◽  
pp. 528-529
Author(s):  
M. G. Burke ◽  
R. J. Wehrer ◽  
C.M. Brown
Keyword(s):  
Nuclear Power ◽  
Base Alloy ◽  
Analytical Electron Microscopy ◽  
Furnace Cool ◽  
Environmental Embrittlement ◽  
Reported Study ◽  
Compositional Changes ◽  
Insight Into ◽  
Sensitive Behavior

Ni-base alloy welds such as EN82H weld metal are frequently employed in nuclear power applications where resistance to corrosion is required. Results of a recently reported study of the mechanical properties of EN82H welds show that this alloy is susceptible to low-temperature (∼100°C) environmental embrittlement (LTEE) in hydrogenated water. LTEE is a manifestation of hydrogen embrittlement in these alloys.1 Recent LTEE tests have demonstrated a beneficial effect of a high-temperature (∼1100°C) anneal and furnace-cool in alleviating the material's susceptibility to LTEE. Understanding the reason for the reduction in LTEE susceptibility requires detailed characterization of the microstructure so that the specific structural and compositional changes that have been induced by the solution-anneal can be identified. This study reports the results of light optical and analytical electron microscopy (AEM) characterization of the microstructures of as-fabricated and as-solution-annealed EN82H welds with the objective of providing insight into the observed LTEE behavior.

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