Browsing by Author "Şahin, Sümer"
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Article Citation Count: 6Assessment of criticality and burn up behavior of candu reactors with nuclear waste trans uranium fuel(Pergamon-elsevier Science Ltd, 2012) Şahin, Sümer; Ahmed, Rizwan; Khan, Mohammad Javed; Department of Mechanical EngineeringLarge quantities of nuclear waste plutonium and minor actinides (MAs) have been accumulated in the civilian light water reactors (LWRs) and CANDU reactors. These trans uranium (TRU) elements are all fissionable, and thus can be considered as fissile fuel materials in form of mixed fuel with thorium or naturanium in the latter. CANDU fuel compacts made of tristructural-isotropic (TRISO) type pellets would withstand very high burn ups without fuel change. As carbide fuels allow higher fissile material density than oxide fuels, following fuel compositions have been selected for investigations: (1) 90% nat-UC + 10% TRUC, (2) 70% nat-UC + 30% TRUC and (3) 50% nat-UC + 50% TRUC. Higher TRUC charge leads to longer power plant operation periods without fuel change. The behavior of the criticality k(infinity) and the burn up values of the reactor have been pursued by full power operation for > similar to 12 years. For these selected fuel compositions, the reactor criticality starts by k(infinity) = 1.4443, 1.4872 and 1.5238, where corresponding reactor operation times and burn up values have been calculated as 2.8 years, 8 years and 12.5 years, and 62, 430 MW.D/MT, 176,000 and 280,000 MW.D/MT, with fuel consumption rates of similar to 16, 5.68 and 3.57 g/MW.D respectively. These high burn ups would reduce the nuclear waste mass per unit energy output drastically. The study has show clearly that TRU in form of TRISO fuel pellets will provide sufficient criticality as well as reasonable burn up for CANDU reactors in order to justify their consideration as alternative fuel. (c) 2012 Elsevier Ltd. All rights reserved.Article Citation Count: 5Candu reactors with reactor grade plutonium/thorium carbide fuel(Carl Hanser Verlag, 2011) Şahin, Sümer; Khan,M.J.; Ahmed,R.; Department of Mechanical EngineeringReactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC+70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k∞,0 = 1-4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are ~ 2.7, 8.4, and 15 years and with burn ups of ∼ 72000, 222000 and 366000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. © Carl Hanser Verlag, München.Article Citation Count: 10Commercial utilization of weapon grade plutonium as TRISO fuel in conventional CANDU reactors(Pergamon-elsevier Science Ltd, 2012) Şahin, Sümer; Sahin, Haci Mehmet; Acir, Adem; Department of Mechanical EngineeringLarge quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Plutonium and heavy water moderator can give a good combination with respect to neutron economy. TRISO type fuel can withstand very high fuel burn up levels. The paper investigates the prospects of utilization of TRISO fuel made of WG-plutonium in CANDU reactors. Three different fuel compositions have been investigated: (1): 90% ThC + 10% PuC, (2): 70% ThC + 30% PuC and (3): 50% ThC + 50% PuC. The temporal variation of the criticality k(infinity) and the burn-up values of the reactor have been calculated by full power operation up to 17 years. Calculated startup criticalities for these fuel modes are k(infinity.0)= 1.6403, 1.7228 and 1.7662, respectively. Attainable burn up values and reactor operation times without new fuel charge will be 94700, 265000 and 425000 MW.D/MT and along with continuous operation periods of similar to 3.5, 10 and 17 years, respectively, for the corresponding modes. These high burn ups would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically. (C) 2012 Elsevier Ltd. All rights reserved.Conference Object Citation Count: 9COMPARISONS OF THE CALCULATIONS USING DIFFERENT CODES IMPLEMENTED IN MCNPX MONTE CARLO TRANSPORT CODE FOR ACCELERATOR DRIVEN SYSTEM TARGET(Amer Nuclear Soc, 2012) Şahin, Sümer; Sahin, Sumer; Gunay, Mehtap; Celik, Yurdunaz; Department of Mechanical EngineeringThe MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation-fission models and CEM2k cascade-exciton model. The study analyzes the main quantities determining ADS performance such as neutron yield, neutron leakage spectra, and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models, cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. Target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. Target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 5.3 cm and an inner radius 5.0 cm. The maximum of the neutron flux in the target is observed on the axis similar to 10 cm below the beam window, where the maximum difference between 7 different models is similar to 15 %. The total neutron leakage out of the of the target calculated with the Bertini/ABLA is 1.83x10(17) n/s, and is about 14 % higher than the value calculated by the INCL4/Dresner (1.60x10(17) n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20 %, 2.3 %, 77.6 % of the total leakage, respectively, whereas, they become 18.6 %, 2.3 %, 79.4 % with INCL4/Dresner combination.Conference Object Citation Count: 33Criticality and burn up evolutions of the Fixed Bed Nuclear Reactor with alternative fuels(Pergamon-elsevier Science Ltd, 2010) Şahin, Sümer; Sahin, Haci Mehmet; Acir, Adem; Department of Mechanical EngineeringTime evolution of criticality and burn-up grades of the Fixed Bed Nuclear Reactor (FBNR) are investigated for alternative fuels. These are: (1) low enriched uranium, (2) weapon grade plutonium, (3) reactor grade plutonium, and (4) minor actinides in the spent fuel of light water reactors (LWRs). The criticality calculations are conducted with SCALE 5.1 using S(8)-P(3) approximation in 238 neutron energy groups with 90 groups in thermal energy region. The main results of the study can be summarized as follows: (1) Low enriched uranium (UO(2)): FBNR with an enrichment grade of 9% and 19% will start with k(eff) = 1.2744 and k(eff) = 1.36 and can operate similar to 8 and >15 years with the same fuel charge, where criticality drops to k(eff) = 1.06 and a burn-up grade of 54 000 and >110000 MW.D/t can be attained. (2) Weapon grade plutonium: Such a high quality nuclear fuel suggests to be mixed with thorium. Second series of criticality calculations are conducted with fuel compositions made of thoria (ThO(2)) and weapon grade PuO(2), where PuO(2) component has been varied from 1% to 100%. Criticality with k(eff) > 1.0 is achieved by similar to 2.5% PuO(2). At 4% PuO(2), the reactor criticality will become satisfactory (k(eff) = 1.1121), rapidly increasing with more PuO(2). A reasonable mixture will by around 20% PuO(2) and 80% ThO(2) with a k(eff) = 1.2864. This mixed fuel would allow full power reactor operation for >20 years and burn-up grade can reach 136 000 MW.D/t. (3) Reactor grade plutonium: Third series of criticality calculations are conducted with fuel compositions made of thoria and reactor grade PuO(2), where PuO(2) is varied from 1% to 100%. Reactor becomes critical by 8% PuO(2) content. One can achieve k(eff) = 1.2670 by 35% PuO(2) and would allow full power reactor operation also for >20 years and burn-up grade can reach 123 000 MW.D/t. (4) Minor actinides in the spent fuel of LWRs: Fourth series of criticality calculations are conducted with fuel compositions made of thoria and MAO(2), where MAO(2) is varied from 1% to 100%. Reactor becomes critical by similar to 17% MAO(2) content. Reasonably high reactor criticality (k(eff) = 1.2673) is achieved by 50% MAO(2) for a reactor operation time of 15 years with a burn up of 86 000 MW.D/t without fuel change. On that way, the hazardous nuclear waste product can be transmuted as well as utilized as fuel. (C) 2010 Elsevier Ltd. All rights reserved.Editorial Citation Count: 1Editor's notes on ICENES'2013, 16th International Conference on Emerging Nuclear Energy Systems(Pergamon-elsevier Science Ltd, 2015) Şahin, Sümer; Department of Mechanical Engineering[No Abstract Available]Article Editor’s notes on ICENES’2013, 16th International Conference on Emerging Nuclear Energy Systems(Progress in Nuclear Energy, 2015) Şahin, Sümer; Department of Mechanical EngineeringThe 16th ICENES 2013 was held in Madrid (Spain) on May 26e 30, 2013, by Universidad Politécnica of Madrid (UPM), according to a scientific tradition of 35 years, started in 1978 as an autono mous, self-organized event, when a group of independent nuclear scientists met in Graz, Austria in order to find new routes for improving the use of the immense energy of the atomic nucleus. Throughout the years, ICENES became a very important series of well established conferences and acted as an open forum for inno vation and challenges in the field as long as they were based on sound deliberations and were addressing the key issue of providing energy for humankind.Article Editor’s Report on NURER2012, The III. International Conference on Nuclear and Renewable Energy Resources, _ Istanbul, Türkiye (20–23rd May 2012)(Energy Conversion and Management, 2012) Şahin, Sümer; Department of Mechanical EngineeringThe Journal of Energy Conversion and Management covers a wide range of topics related to energy such as the energy efficiency and management; heat pipes; thermo-siphons and capillary pumped loops; thermal management of spacecraft; space and terrestrial power systems; hydrogen production and storage; renewable energy; nuclear power; conventional power; single and combined cycles; miniaturized energy conversion and power systems; fuel cells and advanced batteries; biomass, and water management and desalinationEditorial Citation Count: 0Editor's Report on NURER2012, The III. International Conference on Nuclear and Renewable Energy Resources, Istanbul, Turkiye (20-23rd May 2012)(Pergamon-elsevier Science Ltd, 2013) Şahin, Sümer; Department of Mechanical Engineering[No Abstract Available]Article ‘‘EDITOR’S REPORT’’, IREC 2011, The International Renewable Energy Congress, Hammamet, Tunisia(Energy Conversion and Management, 2011) Şahin, Sümer; Department of Mechanical EngineeringThe Journal of Energy Conversion and Management covers a wide range of topics related to energy such as the energy efficiency and management; heat pipes; thermo-siphons and capillary pumped loops; thermal management of spacecraft; space and ter restrial power systems; hydrogen production and storage; renew able energy; nuclear power; single and combined cycles; miniaturized energy conversion and power systems; fuel cells and advanced batteries; and water management and desalination.Editorial Citation Count: 1EDITOR'S REPORT, IREC 2011, The International Renewable Energy Congress, Hammamet, Tunisia (December 20-22, 2011)(Pergamon-elsevier Science Ltd, 2012) Şahin, Sümer; Department of Mechanical Engineering[No Abstract Available]Editorial Citation Count: 0Editorial notes on the 2012 International Youth Nuclear Congress (IYNC), Charlotte, North Carolina, USA (5-11 August 2012)(Pergamon-elsevier Science Ltd, 2013) Şahin, Sümer; Department of Mechanical Engineering[No Abstract Available]Article Editorial notes on the 2012 International Youth Nuclear Congress (IYNC), Charlotte, North Carolina, USA (5–11 August 2012)(Energy Conversion and Management, 2012) Şahin, Sümer; Department of Mechanical EngineeringPrimary purpose of the International Youth Nuclear Congress (IYNC) series is to transfer knowledge from the current generation of leading scientists and engineers to the next generation. Scien tific, political, public and corporate views regarding the develop ment of different nuclear issues are presented to provide comprehensive discussions on all sides of the subject. With this aim, The 2012 International Youth Nuclear Congress has been held in Charlotte, North Carolina. IYNC2012 was focused on the use of nuclear energy more than ever after the accident at Fukushima power plant and also in the framework of the nuclear renaissance in many developing countries and world powers as well. IYNC2012 offered the opportunity to share knowledge, experience, best prac tices and information about nuclear energy between generations and also between peers in the ongoing mission to promote the peaceful use of nuclear power.Editorial Citation Count: 0Editorials, "IREC 2012, The International Renewable Energy Congress," Sousse, Tunisia (December 19-22, 2012)(Pergamon-elsevier Science Ltd, 2014) Şahin, Sümer; Department of Mechanical Engineering[No Abstract Available]Article Citation Count: 9Emergency Planning Zones Estimation for Karachi-2 and Karachi-3 Nuclear Power Plants using Gaussian Puff Model(Hindawi Ltd, 2016) Şahin, Sümer; Ali, Muhammad; Department of Mechanical EngineeringEmergency planning zones (PAZ and UPZ) around the Karachi-2 and Karachi-3 nuclear power plants (K-2/K-3 NPPs) have been realistically determined by employing Gaussian puff model and Gaussian plume model together for atmospheric transport, diffusion, and deposition of radioactive material using onsite and regional data related to meteorology, topography, and land-use along with latest IAEA Post-Fukushima Guidelines. The analysis work has been carried out using U.S. NRC computer code RASCAL 4.2. The assumed environmental radioactive releases provide the sound theoretical and practical bases for the estimation of emergency planning zones covering most expected scenario of severe accident and most recent multiunit Fukushima Accident. Sheltering could be used as protective action for longer period of about 04 days. The area about 3 km of K-2/K-3 NPPs site should be evacuated and an iodine thyroid blocking agent should be taken before release up to about 14 km to prevent severe deterministic effects. Stochastic effects may be avoided or minimized by evacuating the area within about 8 km of the K-2/K-3 NPPs site. Protective actions may become more effective and cost beneficial by using current methodology as Gaussian puff model realistically represents atmospheric transport, dispersion, and disposition processes in contrast to straight-line Gaussian plume model explicitly in study area. The estimated PAZ and UPZ were found 3 km and 8 km, respectively, around K-2/K-3 NPPs which are in well agreement with IAEA Post-Fukushima Study. Therefore, current study results could be used in the establishment of emergency planning zones around K-2/K-3 NPPs.Article Citation Count: 11Energy multiplication and fissile fuel breeding limits of accelerator-driven systems with uranium and thorium targets(Pergamon-elsevier Science Ltd, 2015) Şahin, Sümer; Sarer, Basar; Celik, Yurdunaz; Department of Mechanical EngineeringThe study analyses the integral U-233 and Pu-239 breeding rates, neutron multiplication ratio through (n,xn)- and fission-reactions, heat release, energy multiplication and consequently the energy gain factor in infinite size thorium and uranium as breeder material in an accelerator driven systems (ADS), irradiated by a 1-GeV proton source. Energy gain factor has been calculated as M-energy = 1.67, 4.03 and 5.45 for thorium, depleted uranium (100% U-238) and natural uranium, respectively, where the infinite criticality values are k(infinity) = 0.40, 0.752 and 0.816. Fissile fuel material production is calculated as 53 Th-232(n,gamma)U-233, 80.24 and 90.65 U-238(n,gamma)Pu-239 atoms per incident proton, respectively. The neutron spectrum maximum is by similar to 1 MeV. Lower energy neutrons E < 1 MeV have major contribution on fissile fuel material breeding (>97.5%), whereas their share on energy multiplication is negligible (0.2%) for thorium, depleted uranium. Major fission events occur in the energy interval 1MeV < E < 50 MeV. Copyright (C) 2015, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved.Article Citation Count: 7Evaluation of integral quantities in an accelerator driven system using different nuclear models implemented in the MCNPX Monte Carlo transport code(Pergamon-elsevier Science Ltd, 2013) Şahin, Sümer; Sahin, Sumer; Celik, Yurdunaz; Gunay, Mehtap; Department of Mechanical EngineeringThe MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation-fission models and CEM2k cascade-exciton model. This study analyzes the main quantities determining ADS performance, such as neutron yield, neutron leakage spectra, heating and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. The target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. The target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 53 cm and an inner radius of 5.0 cm. The maximum value of the neutron flux in the target is observed on the axis similar to 10 cm below the beam window, where the maximum difference between 7 different models is similar to 15%. The total neutron leakage of the target calculated with the Bertini/ABLA is 1.83 x 10(17) n/s, and is about 14% higher than the value calculated by the INCL4/Dresner (1.60 x 10(17) n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20%, 2.3%, 77.6% of the total leakage, respectively, whereas, the calculated fractions are 18.6%, 2.3%, 79.4%, respectively, with INCL4/Dresner combination. The largest heat deposition density by considering all particles in the beam window calculated with CEM2k model is 104 W/cm(3)/mA, which is 9.0% greater than the lowest value predicted with INCL4/Dresner model (95.4 W/cm(3)/mA). The maximum average heat deposition density for all particles in the target is calculated as 6.87 W/cm(3)/mA with INCL4/ABLA. (C) 2013 Elsevier Ltd. All rights reserved.Article Citation Count: 3Experimental evaluation of surveillance capsule assemblies for life assessment of CHASNUPP Unit-1 reactor pressure vessel(Pergamon-elsevier Science Ltd, 2016) Şahin, Sümer; Saeed, Asim; Department of Mechanical EngineeringNeutron flux and energy spectrum were determined at the surface of three in-vessel Surveillance Capsule Assemblies (SCAs) removed from CHASNUPP Unit-1 after 2nd, 4th, and 9th fuel cycles for the life assessment of reactor pressure vessel belt line region. Dosimetry data were measured from radiometric sensors irradiated in base material section of SCAs. Fast neutron flux (E > 1.0 MeV) was best estimated at the surface of three SCAs corresponding to the center of C-1 core using the least square method by employing LSL-M2 package. These results were compared with fast neutron flux calculated using DOT3.5 code and both results are within good agreement of +/- 20% acceptance criteria as described in Regulatory Guide 1.190. Therefore, calculational model was validated by dosimetry evaluation and these results can be used in the life assessment of CHASNUPP Unit-1 pressure vessel belt line region. (C) 2015 Elsevier Ltd. All rights reserved.Article Citation Count: 25Fissile fuel breeding and minor actinide transmutation in the life engine(Elsevier Science Sa, 2011) Şahin, Sümer; Khan, Mohammad Javed; Ahmed, Rizwan; Department of Mechanical EngineeringProgress on The National Ignition Facility (NIF) brings fusion a viable energy source in foreseeable future. Energy multiplication in a fusion-fission (hybrid) reactor could lead earlier market penetration of fusion energy for commercial utilization. Originally, scientists at the Lawrence Livermore National Laboratory (LLNL) have worked out a hybrid reactor design concept; the so-called Laser Inertial Confinement Fusion-Fission Energy (LIFE) engine, which has consisted of a spherical fusion chamber of similar to 5 m diameter, surrounded by a multi-layered blanket with a beryllium multiplier zone after the first wall. However, earlier work had indicated extreme power peaks at immediate vicinity of the first wall of a hybrid assembly, if a beryllium multiplier is used. Hence, in the current work, the beryllium multiplier zone has been removed in order to mitigate fission power peaks at the vicinity of the first wall as a result of neutron moderation on beryllium. Furthermore, minor actinides (MA) will cause significant neutron multiplication under fusion neutron irradiation so that an extra beryllium multiplier will not be needed. Present work has made following modifications on the LLNL design of the original (LIFE) engine: Omission of beryllium multiplier. TRISO fuel has been suspended as micro-size particles in Flibe coolant in lieu of being dissolved in uranium salt or imbedded carbon matrix in macro-size pebbles. Carbide fuel is used. Fissionable fuel charge is kept lower than in the LLNL (LIFE) engine. The modified (LIFE) engine is kept similar to the LLNL design to a great degree in order to allow mutual feedback between two geographically separated teams towards a more advanced and improved design under consideration of totally independent views. The first wall is made of ODS (2 cm) and followed by a Li17Pbg3 zone (2 cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2 cm) from the Flibe molten salt zone (50 cm), containing MA as fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a fusion driver power of 500 MWth in S-8-P-3 approximation using 238-neutron groups. Minor actinides (MA) out of the nuclear waste of LWRs are used as fissile carbide fuel in TRISO particles with volume fractions of 0,2,3,4 and 5% have been dispersed homogenously in the Flibe coolant. For these cases, tritium breeding at startup is calculated as TBR= 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. In the course of plant operation, TBR and fissile neutron multiplication factor decrease gradually. For a self-sustained reactor, TBR > 1.05 can be kept for all cases over 8 years. Higher fissionable fuel content in the molten salt leads also to higher blanket energy multiplication, namely M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5% TRISO volume fraction at start up, respectively. For all investigated cases, fissile burn up exceeds 400000 MW D/MT. Major damage mechanisms have been calculated as DPA = 50 and He = 176 appm per year. This implies a replacement of the first wall every 3 years. (C) 2011 Elsevier B.V. All rights reserved.Conference Object Citation Count: 3Fixed bed nuclear reactor for electricity and desalination needs of Middle-East countries(2010) Şahin, Sümer; Şahin,H.M.; Sefidvash,F.; Al-Kusayer,T.A.; Department of Mechanical EngineeringA new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are being developed by the INPRO program of the International Atomic Energy Agency (IAEA). It is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The Fixed Bed Nuclear Reactor (FBNR) is a small (70 MWel) nuclear reactor that meets all the requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve in a dual purpose plant to produce simultaneously both electricity and desalinated water, thus making it especially suitable to the needs of the Middle-East Countries. FBNR is being developed with the support of the International Atomic Energy (IAEA) under its program of Small Reactors Without On-Site Refueling (SRWOSR). The reactor uses the pressurized water reactor (PWR) technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact.
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