scholarly journals MARIA Reactor Irradiation Technology Capabilities towards Advanced Applications

Energies ◽  
2021 ◽  
Vol 14 (23) ◽  
pp. 8153
Author(s):  
Marek Migdal ◽  
Emilia Balcer ◽  
Łukasz Bartosik ◽  
Łukasz Bąk ◽  
Agnieszka Celińska ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Scientific Community ◽  
Research Reactor ◽  
Nuclear Research ◽  
Material Testing ◽  
Reactor Irradiation ◽  
Fuel Conversion ◽  
Nuclear Installation ◽  
Power Studies ◽  
Irradiation Technology

The MARIA research reactor is designed and operated as a multipurpose nuclear installation, combining material testing, neutron beam experiments, and medical and industrial radionuclide production, including molybdenum-99 (99Mo). Recently, after fuel conversion to LEU and rejuvenation of the staff while maintaining their experience, MARIA has been used to respond to the increased interest of the scientific community in advanced nuclear power studies, both fission and fusion. In this work, we would like to introduce MARIA’ s capabilities in the irradiation technology field and how it can serve future nuclear research worldwide.

Using Probabilistic Criteria in an Assessment of the Potential Radiological Consequences of the Decommissioning of a Nuclear Research Reactor

2013 ◽  
Author(s):  
Christian Wallner ◽  
Anna-Maria Rall ◽  
Severin Thummerer
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Research Reactor ◽  
Nuclear Research ◽  
Exposure Dose ◽  
Occurrence Frequency ◽  
Normal Operation ◽  
Nuclear Facilities ◽  
Nuclear Research Reactor

In order to assess the risk of radiological consequences of incidents and accidents in nuclear facilities it is important to contemplate their frequency of occurrence. It has to be shown that incidents and accidents occur sufficiently seldom according to their radiological consequences i. e. the occurrence frequency of radiological doses has to be limited. This is even demanded by the German radiation protection ordinance (StrlSchV), which says that in nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, the occurrence frequency of incidents and accidents shall be contemplated in order to prove the design of safety measures and safety installations. Based on the ideas of the ICRP64, we developed a risk based assessment concept for nuclear facilities, which fulfils the requirements of the German regulations concerning dose limits in normal operation and design basis accidents. The general use of the concept is dedicated to nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, where the regulation of risk assessment is less sophisticated. The concept specifies occurrence frequency limits for radiation exposure dose ranges, i. e. the occurrence frequency of incidents and accidents has to be limited according to their radiological effects. To apply this concept, scenarios of incidents and accidents are grouped in exposition classes according to their resulting potential effective dose to members of the general public. The occurrence frequencies of the incidents and accidents are summarized in each exposition class whereas the sum must not exceed the frequency limits mentioned above. In the following we introduce the application of this concept in the assessment of the potential radiological consequences of the decommissioning of a nuclear research reactor. We carried out this assessment for the licensing process of the decommissioning on behalf of German authorities.

Estimation of Emergency Planning Zones (EPZs) for Nuclear Research Reactor Using Plume Dispersion Code

2012 ◽  
Author(s):  
Mazzammal Hussain ◽  
Salah Ud-Din Khan ◽  
Waqar A. Adil Syed
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Research Reactor ◽  
Nuclear Research ◽  
Emergency Planning ◽  
Dispersion Modeling ◽  
Plume Dispersion ◽  
Nuclear Research Reactor ◽  
Accident Scenarios

After Fukushima nuclear power plant accident, a demand for re-evaluation of emergency planning zones for nuclear facilities has emerged to ensure that in case of nuclear accident, the population, the environment and the property should lie in the safe zone. Emergency planning zones (EPZs) around a facility are the designated areas where protective measures are adopted according to predefined emergency plan for different emergency situations. Deterministic and probabilistic approaches have been adopted to assess and design emergency planning zones for design base accidents (DBAs), however, after Fukushima nuclear power plant accident, multi-layer failure of the plant systems could not be ruled out. To estimate the dispersion of radioactive material and respective radiation doses received by the affected people, different techniques are used including on-field measurements through survey meters, on-line monitoring network and state-of the art codes for plume dispersion modeling. In this study plume dispersion modeling and estimation of emergency planning zones has been carried out using plume dispersion code for hypothetical accident scenarios at a 10MW nuclear research reactor. Different accident scenarios were considered with different release characteristic and environmental conditions to study the affect of the parameters including release height, heat content, release time, atmospheric stability. The simulation results have been analyzed to assess the existing emergency planning zones of nuclear research reactor.

scholarly journals AN APPLICATION OF IRIDM IN THE DECISION MAKING PROCESS ON FUEL CONVERSION OF THE MARIA REACTOR

2013 ◽  
Vol 3 (3) ◽  
pp. 3-6 ◽  
Author(s):  
Mieczysław Józef Borysiewicz ◽  
Karol Kowal ◽  
Piotr Andrzej Prusiński ◽  
Marcin Dąbrowski
Keyword(s):  
Decision Making ◽  
Research Reactor ◽  
Nuclear Research ◽  
Decision Making Process ◽  
Fuel Conversion ◽  
National Centre ◽  
Uranium Fuel ◽  
Integrated Risk ◽  
Enriched Uranium ◽  
Global Threat

Poland, when acceded to GTRI (Global Threat Reduction Initiative) in 2004, has committed to convert the nuclear fuel of the Research Reactor MARIA, operated by the National Centre for Nuclear Research (NCBJ) in Świerk. The conversion means giving up of high enriched uranium fuel containing 36% of U-235, which was used so far, and replacing it with the low enriched uranium fuel (19.7% U-235). This article describes the potential usability of the Integrated Risk Informed Decision Making (IRIDM) methodology in optimization of the fuel conversion procedure.

scholarly journals MCNP model of L-54M nuclear research reactor: validation by preliminary graphite radiological characterization

2019 ◽  
Vol 322 (3) ◽  
pp. 1341-1350
Author(s):  
Eros Mossini ◽  
Luca Codispoti ◽  
Giorgio Parma ◽  
Filippo Maria Rossi ◽  
Elena Macerata ◽  
...  
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Nuclear Research Reactor ◽  
Radiological Characterization

Site Decommissioning of AECL Whiteshell Laboratories

2004 ◽  
Vol 824 ◽  
Author(s):  
Grant W. Koroll
Keyword(s):  
Waste Management ◽  
Environmental Assessment ◽  
Materials Science ◽  
Nuclear Research ◽  
Management Program ◽  
Nuclear Facilities ◽  
Safety Research ◽  
Nuclear Installation ◽  
Underground Research Laboratory ◽  
Nuclear Fuel Waste

AbstractAECL Whiteshell Laboratories (WL), near Winnipeg, Canada has been in operation since the early 1960s. R&D programs carried out at WL include a 60 MW organic-cooled research reactor, which operated from 1965 to 1985, reactor safety research, small reactor development, materials science, post irradiation examinations, chemistry, biophysics and radiation applications. The Canadian Nuclear Fuel Waste Management Program was conducted and continues to operate at WL and also at the nearby Underground Research Laboratory.In the late-1990s, AECL began to consolidate research and development activities at its Chalk River Laboratories (CRL) and began preparations for decommissioning WL. Preparations for decommissioning included a staged shutdown of operations, planning documentation and licensing for decommissioning. As a prerequisite to AECL's application for a decommissioning licence, an environmental assessment (EA) was carried out according to Canadian environmental assessment legislation. The EA concluded in 2002 April when the Federal Environment Minister published his decision that WL decommissioning was not likely to cause significant adverse environmental effects and that no further assessment by a review panel or mediation would be requiredIn 2002 December, the Canadian Nuclear Safety Commission issued a decommissioning licence for WL, valid until December 31, 2008. The licence authorized the first planned phase of site decommissioning as well as the continuation of selected research programs. The six-year licence for Whiteshell Laboratories was the first overall decommissioning license issued for a Canadian Nuclear Research and Test Establishment and was the longest licence term ever granted for a nuclear installation of this complexity in Canada.The first phase of decommissioning is now underway and focuses on decontamination and modifications to nuclear facilities, such as the shielded facilities, the main R&D laboratories and the associated service systems, to achieve a safe state of storage-with-surveillance. Later phases have planned waste management improvements for selected wastes already in storage, eventually followed by final decommissioning of facilities and infrastructure and removal of most wastes from the site.This paper provides an overview of the planning, environmental assessment, licensing, and organizational processes for decommissioning and selected descriptions of decommissioning activities currently underway at AECL Whiteshell Laboratories.

Study on Supervision Mode of Floating Nuclear Power Plant With Small Modular Reactor

2018 ◽  
Author(s):  
Lei Wan ◽  
Guiyong Li ◽  
Min Rui ◽  
Yongkang Liu ◽  
Jue Yang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Government Regulation ◽  
Mutual Recognition ◽  
Nuclear Safety ◽  
Small Modular Reactor ◽  
License Application ◽  
Mutual Recognition Agreement ◽  
Nuclear Installation

A floating nuclear power plant (FNPP) with small modular reactor (SMR) is a combination of a civilian nuclear infrastructure and an offshore installation, which is defined as a floating nuclear facility. The article draws the lessons from studying of the engineer combination like Floating Production Storage and Offloading (FPSO) under the regulation of several government departments. It puts forward recommendations for license application and government regulation as follows in consideration with current license application for nuclear power plant and ship survey. A FNPP shall follow the requirements of construction, fueling and operation for civil nuclear installation combined with ship survey. Application is submitted to nuclear safety regulator for construction permit, while the design drawings shall be submitted to department of ship survey which checks the drawings whether meet the requirements of ship survey, considering some nuclear safety needs. The result of ship survey shall be represented in the safety analysis reports. The construction and important devices manufacturing shall be under the supervision of nuclear installation regulators and ship survey departments. In conclusion, National Nuclear Safety Administration (NNSA) and Maritime Safety Administration of the People’s Republic of China (MSA) shall establish united supervisory system for SMR on sea in China. It is suggested that NNSA is in charge of the overall safety of a FNPP, while MSA is responsible of the ship survey. The operator shall undertake obligation of a FNPP and evaluate the ship cooperating with experienced agency. It is suggested that government departments build the mutual recognition agreement of safety review. It is better to solve the vague questions by coordination.

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