Investigation of mineralogy, porosity and pore structure of Olkiluoto bedrock

2014 ◽  
Vol 1665 ◽  
pp. 31-37 ◽  
Author(s):  
Juuso Sammaljärvi ◽  
Antero Lindberg ◽  
Jussi Ikonen ◽  
Mikko Voutilainen ◽  
Marja Siitari-Kauppi ◽  
...  
Keyword(s):  
Pore Structure ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Mean Value ◽  
High Porosity ◽  
Rock Matrix ◽  
Crystalline Bedrock ◽  
Rock Heterogeneity ◽  
Polarized Microscopy

ABSTRACTSpent nuclear fuel from TVO's (Teollisuuden Voima Oy) and Fortum's nuclear power plants will be deposited deep in the crystalline bedrock in Olkiluoto, Western Finland. The bedrock needs to be well characterized to assess the risks inherent to the waste disposal at the site. If radionuclides (RN) are transported, it happens via water conducting fractures. Retardation may occur either by diffusion into stagnant pore water or by immobilization on mineral surfaces of the rock matrix.RN’s retardation from flowing water is linked to parameters defining porosity and microscopic rock pore structure, such as pore size distribution, connectivity, tortuosity and constrictivity, and by the mineralogy and chemical nature of the minerals and charge of the pore surfaces.In this work, centimeter scale rock cores from Olkiluoto were investigated. The work is part of the in situ project REPRO (Experiments to investigate Rock Matrix Retention Properties) where the diffusion and sorption of RN are studied experimentally. Porosity and pore structures were characterized with the PMMA autoradiography method and polarized microscopy, which was used also to ascertain the mineralogy of the samples.The results show that the rock from the REPRO site has low porosity with a mean value of 0.5% and a range of 0.1-1.5%. Rock heterogeneity explains the variation of porosity values. Correlation between the porosity and the mineralogy was found. Areas of high porosity correspond to areas of altered minerals, such as cordierite, biotite and plagioclase, which cover spatially between 10 and 20% of the rock volume

Through diffusion study on Olkiluoto veined gneiss and pegmatitic granite from a structural perspective

MRS Advances ◽  
2016 ◽  
Vol 1 (61) ◽  
pp. 4041-4046 ◽  
Author(s):  
Mikko Voutilainen ◽  
Jussi Ikonen ◽  
Juuso Sammaljärvi ◽  
Jukka Kuva ◽  
Antero Lindberg ◽  
...  
Keyword(s):  
Pore Structure ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Knudsen Diffusion ◽  
Effective Diffusion ◽  
Crystalline Bedrock ◽  
Through Diffusion ◽  
The Difference ◽  
Micrometer Scale ◽  
Pegmatitic Granite

ABSTRACTSpent nuclear fuel from Finnish power plants is planned to be deposited deep in the crystalline bedrock in Olkiluoto, Finland. The bedrock needs to be well characterized to assess the risks inherent to the long term safety of the site. In the bedrock the possibly released radionuclides are mainly transported by water conducting fractures and their transport is retarded by diffusion and sorption. In porous materials these properties are typically linked to microscopic pore structure (pore size distribution, tortuosity and constrictivity) and chemical nature of the minerals and groundwater.In this work transport properties of veined gneiss (VGN) and pegmatitic granite (PGR) samples from Olkiluoto were studied using various through diffusion experiments and the C-14-PMMA autoradiography. Through diffusion experiments were performed on rock cores using HTO and 36Cl in water phase and He in gas phase as tracers. The effective diffusion coefficients (De) determined for the VGN were found to be dependent on the tracer molecule (De(HTO) < De(He) < De(Cl)) whereas for the PGR such a dependence was not found. The porosity distributions determined by the C-14-PMMA autoradiography revealed the difference in the pore structure between the samples. The porosity of VGN consists mostly of grain boundary pores and pores between biotite lamellae. Due to a high content of nanometer scale pores anion exclusion affected the results of 36Cl and Knudsen diffusion the ones of He for VGN. Furthermore, in the PGR micrometer scale intra- and intergranular fissures form a connected network for diffusive transport and thus all tracers diffuse at the same rate.

scholarly journals Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

2021 ◽  
Vol 7 (1) ◽  
pp. 9-13
Author(s):  
David A. Hakobyan ◽  
Victor I. Slobodchuk
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Cooling Mode ◽  
Temperature Conditions ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

Operational Experience of TN 24 Dual Purpose Spent Fuel Storage Casks at the DOEL Nuclear Power Plants in Belgium

2001 ◽  
Author(s):  
C. Baroux ◽  
M. Detrilleaux ◽  
G. Demazy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
High Capacity ◽  
Radiation Shielding ◽  
Ring Closure ◽  
Operational Experience ◽  
Spent Fuel ◽  
Dual Purpose ◽  
New Members

Abstract Spent nuclear fuel has been stored at the DOEL power station in Belgium in dual-purpose metal casks since 1995. The casks were procured from TRANSNUCLEAIRE by SYNATOM to meet the operational demands for on-site dry storage solutions for fuel arising from the four PWR reactors at DOEL. The TN 24 type of cask was chosen and a range of different cask types were developed. The initial requirement was for dual purpose cask to contain fuel from the DOEL units 3 and 4, these having similar fuel types but different lengths, and thus two new members of the TN 24 family were developed; the TN 24 D and TN 24 XL with capacities of 28 and 24 SFA’s. These casks were licensed as B(U) fissile packagings with approval certificates granted by the French and validated by the Belgium competent authorities for the transport configurations. Both cask designs were also analyzed by TRANSNUCLEAIRE in their storage configurations to ensure that the criteria for safe interim storage could be met. Since 1995, a total of 18 TN 24 D and TN 24 XL casks have been loaded with spent fuel assemblies with an average burn-up of 40,000 MWd/tU. SYNATOM subsequently decided to purchase further casks for DOEL 3 and 4 fuels with higher enrichments, higher burn-ups and shorter cooling times. TRANSNUCLEAIRE developed the TN 24 DH and TN 24 XLH casks within the similar envelope size and weight limits. The increase in performance was achieved by an in-depth optimization of each design in terms of radiation shielding, heat transfer and criticality safety. This paper shows how this optimization process was undertaken for the TN 24 DH and TN 24 XLH casks, 16 of which have been ordered by SYNATOM. DOEL 1 and 2 units use much shorter PWR fuel and it was decided to ship the fuel to unit 3 with an internal transfer cask because the handling limitations in the DOEL 1 and 2 pool prohibited the loading of a high capacity dual purpose transport/storage cask. The TN 24 SH cask was subsequently designed for DOEL 1 and 2 PWR fuel with a capacity of 37 assemblies and nine of there casks have been ordered by SYNATOM. The casks are fitted with monitoring devices to detect any change in the performance of the double metal O ring closure system and none of the casks has shown any deterioration in leaktightness. This paper examines the operation experience of loading and storing more than 30 TN 24 dual purpose casks and compares the performance with design expectations.

Centralization of Canada’s Spent Nuclear Fuel

2013 ◽  
Author(s):  
Krista Nicholson ◽  
John McDonald ◽  
Shona Draper ◽  
Brian M. Ikeda ◽  
Igor Pioro
Keyword(s):  
Nuclear Fuel ◽  
Site Selection ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Selection Process ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Candu Reactors ◽  
High Level

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel. The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method. The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.

Ways of technetium and neptunium localization in extraction reprocessing of spent nuclear fuel from nuclear power plants

2014 ◽  
Vol 56 (5) ◽  
pp. 501-514 ◽  
Author(s):  
N. D. Goletskii ◽  
B. Ya. Zilberman ◽  
Yu. S. Fedorov ◽  
A. S. Kudinov ◽  
A. A. Timoshuk ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel

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 The Work Results of the Institute for Safety Problems of Nuclear Power Plants of the NAS of Ukraine in 2020

2021 ◽  
Vol 20 ◽  
pp. 96-105
Author(s):  
V. S. Havrylenko ◽  
◽  
I. V. Kutsyna ◽  
D. I. Кhvalin ◽  
◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Scientific Work ◽  
Development Effort ◽  
Nuclear Facilities ◽  
Energy Agency ◽  
Public Authorities ◽  
Organizational Activities

The year 2020 has become a real challenge for almost all aspects of life all over the world. Under these conditions, Ukrainian science, which has been in a state of crisis for more than a year, has been forced to overcome additional difficulties. However, due to the efforts of scientists, scientific work not only did not stop, but also received a large development effort in new directions and formats. In 2020, the Institute for Safety Problems of Nuclear Power Plants was one of the scientific institutions that under the new conditions made every effort to carry out the planned scientific researches and perform works in accordance with its activities. The main results of scientific and scientific-organizational activities of the Institute for Safety Problems of Nuclear Power Plants in 2020 are presented in the article. Despite the difficult economic situation due to the epidemic, the staff of the Institute obtained important results in studies of nuclear and radiation safety of the Shelter object, and in works aimed at improving the reliability and safety of existing Ukrainian and foreign nuclear power plants. The results of works in the field of the Shelter object transformation into an ecologically safe system, safe operation of nuclear facilities, decommissioning of nuclear facilities, spent nuclear fuel and radioactive waste management are presented in the article. The results of the work performed under the International Atomic Energy Agency (IAEA) grant and the Ukrainian-Japanese Science and Technology Research Partnership for Sustainable Development (SATREPS) project are described. The implementation of the results of these studies in practice is indicated. Scientific and expert activities in the interests and at the request of public authorities, cooperation with national and foreign scientific organizations as well as scientific and organizational activities in 2020 are described. Information on internal certification of employees, participation in international and national events, publishing and etc. is presented.

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