Release Of Radioactive Material Research Articles

Design Change and Operational Consideration of the HVAC System during Nuclear Power Plant Decommissioning

The heating, ventilation, and air conditioning (HVAC) system plays a crucial role in ensuring the safety of workers and preventing the release of gaseous radioactive materials into the environment during the decommissioning of a nuclear power plant (NPP). To establish an HVAC operation strategy, decommissioning phases were divided into four stages, and the HVAC systems were reclassified. In addition, assumptions have been made regarding design modifications and maintenance for the reactor containment building (RCB) HVAC, fuel handling building (FHB) HVAC, and main control room (MCR) HVAC. Based on these, for RCB HVAC operation, natural ventilation and RCB purge operation during the transition period are proposed. In the decommissioning stage, recirculation operation, entire ventilation operation consisting of continuous operation and purge operation, and finally partial ventilation operation to purify local space were proposed. Moreover, during the transition period, the FHB HVAC was proposed to operate as normal NPP, and the MCR HVAC was suggested to operate with safety-related equipment removed.

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

Efficient steam condensation is crucial for safe nuclear power plant (NPP) operations, preventing pressure buildup, overheating, and the release of radioactive materials. However, the presence of noncondensable gases (NCGs), such as air, nitrogen, hydrogen, or helium, can hinder the condensation process by creating a thermal resistance layer that impedes steam diffusion and condensation on the system’s surface. Maximizing the efficiency of steam condensation requires a thorough grasp of the fundamental processes, theories, advancements, and technical hurdles. Therefore, this work thoroughly addresses these needs, with a particular emphasis on addressing the challenges posed by NCGs by dividing the work into four thematic areas. The first theme relates to a comprehensive examination of pure steam condensation phenomena, which includes an exploration of familiar condensation scenarios and various film condensation types. The second theme examines condensation in the presence of NCGs, their mixture properties, and related theories and modelling of heat and mass transfer. The third theme investigates condensation in NPP by exploring passive cooling systems and condensation phenomena under both natural and forced convection conditions during nuclear accidents, the origin of NCGs in NPP and their transportation aspects. This is followed by experimental work related to condensation scenarios and scale. Finally, the last theme looks upon the recent advancements in computational fluid dynamics (CFD) modelling of wall condensation, system analysis codes coupling with CFD, and the implementation of machine learning (ML) for predicting the condensation HTC. By bridging the gap between fundamental knowledge and practical applications, the four thematic areas presented in this work are aimed at providing a comprehensive foundation for researchers and experts in the field of steam condensation when NCGs are involved. The ultimate objective is to bolster the safety and efficacy of NPP operations by understanding the heat and mass transfer mechanisms while mitigating the risk of catastrophic events.

Safety evaluation of Multiple Steam Generator Tube rupture accident using the best estimate plus uncertainty approach

Following the disaster in Fukushima Nuclear Power Plant (NPP) in 2011, the awareness of securing the safety of NPP under extreme events has been raised. For this purpose, the concept of Design Extension Conditions (DECs) was introduced, to enhance the plant’s capability to withstand events that are more severe than Design Basis Accidents (DBA), such as Multiple Steam Generator Tube Rupture (MSGTR) accident, given its rapid progression and significant potential for large release of radioactive materials to the environment. It is worth noting that MSGTR has never occurred during the commercial operation of nuclear power plant, therefore the knowledge of consequences of the event and potential mitigation strategy is limited. In this study, a postulated MSGTR (simultaneous rupture of 5 U-tubes in a single SG) accident is analysed using the best-estimate thermal hydraulics code RELAP5/SCDAPSIM/MOD3.4. The main focus of this work is evaluation of appropriate operator mitigation actions under various inherent uncertainties. To this end, the best estimate plus uncertainty (BEPU) approach is adopted, whereby RELAP5/SCDAPSIM/MOD3.4 was coupled to the statistical analysis software, DAKOTA, to conduct the uncertainty quantification (UQ). Using BEPU analysis, an ensemble of system responses under various epistemic and aleatory uncertainties is obtained and used to validate the efficacy of operator actions in bringing the plant to a safe shutdown condition.

Comprehensive integrity assessment of spent nuclear fuel cladding during normal and postulated conditions of transportation

As a part of Spent Nuclear Fuel (SNF) management, it is important to maintain the cladding integrity to prevent the potential release of hazardous radioactive materials. Accordingly, assessing the integrity of SNF under the generated shock and vibration loads during normal conditions is essential for safe transportation. In this study, failure criteria for two zirconium alloys were derived by utilizing data from tensile property tests, mechanical model calculations, and machine learning estimations. Integrity assessment under normal and postulated transportation conditions was also performed via finite element analyses. Loading conditions were generated by real time-acceleration data collected during transportation tests in the Republic of Korea. Shock and vibration loads were applied to a simplified light-water reactor assembly. Furthermore, postulated transportation conditions and pre-cracked cladding were assumed for more challenging conditions. Consequently, a comprehensive integrity assessment of SNF cladding under shock and vibration loads for both free-defect and defective (pre-cracked) cladding was conducted. In each case, the resultant maximum principal strains, stress intensity amplitudes, and stress intensity factors were significantly smaller than the strain-based failure criteria, S-N curve, and fracture toughness, respectively. Therefore, the integrity of SNF cladding under transportation conditions remained intact even when assuming further challenging postulated transportation load conditions.

Application of uncertainty and sensitivity analysis in dose assessment during a postulated LBLOCA for VVER-1000 nuclear reactor

A key element in the process of accident analysis for the event of a severe nuclear emergency such as the large break loss of coolant accident (LBLOCA) is to properly quantify safety margins using reliable tools. Assessing the consequences of radioactive material release and the equivalant dose after the LBLOCA can provide clarity about the accident. Reliable safety analysis methods for nuclear power plants (NPPs) can be used to obtain relatively realistic outcomes considering the complex uncertainties of NPP events. In this study, an appropriate technique for sensitivity and uncertainty analysis is employed to develop a methodology for performing dose assessment within a postulated LBLOCA including uncertainties. Considering the approved reliability over the years and safety aspects, VVER reactor is among the most widely used nuclear reactors, which is selected as the case study in this article. The main problem that the study aims to solve is to illustrate whether the inordinate deviation of the input parameters from their reference values lead to violation of the allowed TEDE value by conducting uncertainty analysis. For performing calculations of total effective dose equivalent (TEDE), the environmental contamination with radioactive materials has been simulated using HotSpot code and the obtained results were compared with the solutions of JRODOS program. To obtain the most influential parameters on the obtained TEDE values in different distances from the NPP, an appropriate sensitivity analysis has been carried out, following which the Wilks’ statistical method is used to generate 59 data samples with uncertainty propagation and 95 % confidence. The resulting uncertainty bands demonstrate that the maximum TEDE values of calculated data lie within about 7 to 15 mSv in the postulated LBLOCA scenario, which are below the threshold limit of 50 mSv recommended by International Atomic Energy Agency.

ORCA: A Tool for Radiological Consequences for Accidental Releases

Evaluating atmospheric dispersion and radiological doses in the vicinity of buildings is required for small modular reactors (SMRs) because of the reduced size of their exclusion area boundary. The current Canadian nuclear industry tool for these calculations implements the methodology defined in CSA Standard N288.2-M91, which was written to support large Canada Deuterium Uranium (CANDU) nuclear reactors as opposed to SMRs. The ORCA (On/offsite Radiological Consequences of Accidents) code has been developed to address this technical concern in addition to evaluating atmospheric dispersion and doses in the far field. The code calculates worker and public doses following an airborne release of radioactive material into the atmosphere under postulated accident conditions at a nuclear facility. The current paper presents the key assumptions and methods utilized in ORCA and discusses qualification of the software to the requirements of CSA Standard N286.7-16. The new model is applicable to SMRs and existing reactor designs and reduces conservatisms in the near field (i.e., <1 km from the source) relative to the methods in CSA N288.2-M91.

Environmental impacts of underground nuclear weapons testing

ABSTRACT Each nuclear weapon test contributes to a global burden of released radioactivity. Between 1945 and 1996, more than 2,000 nuclear tests were conducted, three-quarters of which were underground. Underground nuclear tests involved placing the nuclear device in a cavity drilled or excavated beneath the surface. The goal was to contain the explosion and its effects to the immediate vicinity of the detonation point, ultimately minimizing the release of radioactive materials into the atmosphere. While these tests successfully curtailed the atmospheric release and radioactive fallout, they created dynamic responses within crustal formations caused by local shock waves from the explosion. This paper discusses the legacy of underground nuclear testing, addressing issues from containment failure to the phenomenological effects after underground detonations and the pathways of subsequent dispersal of radionuclides to the environment.

Realistic estimation framework of radioactive release distributions into the environment during nuclear power plant accidents

Since the level 2 PSA of OPR-1000 was the requirement for regulatory purposes, Cs-137 release estimation was contained as the Nuclear Safety Act of ROK in which the Cs-137 release frequency exceeding 100 TBq was determined to happen less than 1.0E-6 per year after the Fukushima Daiichi Accident. However, Cs-137 release estimation from the conventional level 2 PSA of OPR-1000 provided uncertainty due to dominant accident sequence consideration. Thus, this study aimed to develop systematic methods through the overall framework to quantify realistic uncertainty concerns of radioactive material release using sensitivity and uncertainty analysis methods and apply them to OPR-1000. This framework helped to quantify confidential value for the Cs-137 release under the BEPU approach using both parametric and non-parametric methods to cover both realistic and conservative points. Uncertainty propagation analysis showed the unexpected uncertainty increase of Cs-137 release exceeding 100 TBq. The non-parametric uncertainty analysis provided higher conservative concerns for safety than the realistic concerns in terms of economics when compared with the parametric uncertainty analysis. Wilks’ uncertainty analysis showed the importance to consider conservative Cs-137 release in order to reach the higher safety need. Sensitivity analysis showed reasonable relationships between engineering safety parameters with the Cs-137 release.

Concrete ablation depth analysis of OPR1000 NPP during top flooding ex-vessel cooling

This study assesses the integrity of OPR1000 nuclear power plant (NPP) containment during top flooding ex-vessel cooling in the event of a postulated severe accident. Maintaining containment integrity during ex-vessel cooling is critical for preventing substantial release of radioactive materials to the environment. To analyze the feasibility of top flooding ex-vessel cooling of the OPR1000 NPP, the spread of molten corium in the dry reactor cavity is first calculated, followed by calculating the concrete ablation depth according to the reactor cavity flooding delay time. It is judged whether the leak-tightness integrity of the containment during ex-vessel cooling is maintained based on whether the containment liner plate (CLP) is damaged or not. In the first step analysis, the spread of molten corium in the dry reactor cavity is calculated using computational fluid dynamics (CFD) methodology for various cases, and results indicate a uniform deposition of approximately 143 tonnes of molten corium, with a thickness of about 33 cm, meaning that the thermal load is distributed uniformly in the reactor cavity. In the second step analysis, concrete ablation depths are calculated base on the delay time for flooding the reactor cavity. For the postulated severe accident, injecting coolant into the reactor cavity one hour after breach of the reactor vessel results in a concrete ablation depth of about 0.74 m, indicating no damage to the CLP and ensuring containment integrity. The maximum allowable delay time for flooding the reactor cavity for the postulated severe accident is evaluated to be 2.24 hours maintaining the leak-tightness integrity of the OPR1000 NPP containment. The research methodology and findings presented in this study are expected to aid in assessing the feasibility of top flooding ex-vessel cooling and establishing appropriate accident management strategies for nuclear power plants like the OPR1000 NPP.

Coupled AC2-CFD simulations for a high-pressure core melt accident scenario

Abstract Even though very unlikely to occur, severe accident scenarios in nuclear power plants have to be analyzed. During high-pressure core meltdown scenarios in a pressurized water reactor the primary circuit should fail first. Previous analyses found that a free convection flow within the vertical steam generator (SG) tubes with a simultaneous stratified gas counterflow in the hot legs could arise. This phenomenon leads to higher thermal loads on individual SG tubes which might then fail leading to a containment bypass and the release of radioactive material into the environment. Lumped parameter system codes used for safety analyses do not provide the models necessary to simulate phenomena like mixing in three-dimensional flows and could not consider local turbulence effects. Computational fluid dynamic (CFD) codes provide such capabilities but are much more computationally expensive. Coupling of a system code with a CFD code can therefore be used to simulate such phenomena. The advantages of both approaches can be maximized by splitting up the simulation domain between the codes, depending on the expected flow conditions. The system code AC2 coupled with the CFD code OpenFOAM was used to simulate part of the severe accident transient. Free convection in the hot leg and the U-tubes of the vertical SG was observed in case of high-pressure severe accident sequences. The thermal load of individual SG tubes has been estimated from the results. These loads can be used as inputs for structural-mechanical analyses to estimate which part of the primary circuit would fail first.

3D analysis of spray effect on long-term depressurization of VVER-1000 containment during LB-LOCA

Containment as the last barrier of defence in depth strategy plays the most important role in protecting the environment and public from the release of radioactive material. As a result, analysing its performance in nuclear accidents and the reliability and effectiveness of its engineering safety features (ESFs) in the depressurization of containment is a key step in safety assessment. This study demonstrates the effect of the spray system as one of the ESFs in long-term depressurization of VVER-1000 containment during an LB-LOCA accident by using GOTHIC code. In the first step toward this goal, the as-built 3D structure of VVER-1000/V446 containment has been prepared in AutoCAD and used as GOTHIC geometry input by applying the input preparation process. Long-term 3D profiles of temperature and pressure were simulated in GOTHIC without spray actuation. The next step deals with considering the effects of spray and sensitivity analysis of spray temperature and droplet diameter on long-term depressurization. 3D profiles of thermal-hydraulic parameters during long-term depressurization of containment have been presented as research outcomes and validated against FSAR data and finally, the effects of spray on profiles were concluded. Results can obviously prove the effect of spray on the long-term depressurization of containment. The required time to achieve half of the peak pressure of the containment was calculated as 1380 s which is completely in agreement with NRC safety criteria. Research outcomes confirm the success of the spray system to support the integrity of containment which could be jeopardized without spray system intervention.

Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

2023 Fukushima Medical University International Symposium on the Fukushima Health Management Survey, "Thinking Together about Health, Life, and our Future in Fukushima".

In response to the release of radioactive materials and evacuation of residents after the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Plant accident, caused by the Great East Japan Earthquake, the Fukushima Health Management Survey (FHMS) began in June of 2011. This survey aims to provide long-term follow-up of the physical and mental health of Fukushima residents and to maintain and improve their health for the future. Every year since 2019, Fukushima Medical University (FMU) has organized the FHMS International Symposium to share survey results with people in Fukushima Prefecture and beyond. The fifth annual symposium convened at FMU's Ekimae Campus on Saturday, 4 March 2023, with the theme, "Thinking Together about Health, Life and our Future in Fukushima."

Direct numerical simulation of internally heated natural convection in a hemispherical geometry

Internally heated (IH) natural convection can be found in nature, industrial processes, or during a severe accident in a light water reactor. In this accident scenario, the nuclear reactor core and some internal structures can melt down and relocate to the lower head of the reactor pressure vessel (RPV) and interact with the remaining coolant. Subsequent re-heating and re-melting under decay and oxidation heat creates a transition from a debris bed to a molten pool. The molten pool, which can involve more than hundred tons of dangerously superheated oxidic and metallic liquids, imposes thermo-mechanical loads on the vessel wall that can lead to a thermal and/or structural failure of the vessel and subsequent release of radioactive materials to the reactor pit, and can possibly make its way to the environment. This study uses Direct Numerical Simulation (DNS) to investigate homogeneous IH molten pool convection in a hemispherical domain using Nek5000, an open-source spectral element code. With a Rayleigh number of 1.6 × 1011, the highest reached through DNS in this confined hemispherical geometry, and a Prandtl number of 0.5, which corresponds to a prototypic corium, the study provides detailed information on the thermo-fluid behavior. The results show a turbulent flow with three distinct regions, consistent with the general flow observations from the BALI experiments. The study also presents detailed information on turbulence, such as turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance. Additionally, the study provides 3D heat flux distributions along the boundaries. The heat fluxes along the top boundary fluctuate due to the turbulent eddies in the vicinity, while along the curved boundary the heat fluxes increase nonlinearly from the bottom to the top.

Metallographic examinations and hydrogen measurements of high-burnup spent nuclear fuel cladding

In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). To investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.

Introducing DEPDOSE, a Tool to Calculate Dose Coefficients to Members of the Public for Radioactive Aerosols.

This paper presents DEPDOSE, an open-source computer application that combines the KDEP respiratory tract deposition fractions for inhaled aerosols with DC_PAK committed equivalent dose coefficients for a unit deposition in each region of the respiratory tract. DEPDOSE allows the user to rapidly produce tables of dose coefficients for workers and members of the public inhaling precisely defined, user-specified aerosols using the ICRP Publication 60 methodology. Combined with a plume dispersion modeling system, such as the Quick Urban & Industrial Complex (QUIC) Dispersion Modeling System, this makes it possible to predict radiation doses downstream from an accidental or intentional release of radioactive materials. For this work, a radioactive plume was calculated to members of the public downstream from a dirty bomb in Chicago. DEPDOSE is published under an open source license, and can be downloaded at https://github.com/lanl/DEPDOSE .

Soil surface roughness impacts the risk arising from a hypothetical urban radiological dispersive device activation.

This study considers a deliberate hypothetical release of radioactive material over an inhabited urban zone. The event is initiated by the activation of a radiological dispersion device. The main threat is the deposition of radioactive material onto the soil's surface. The radiation represents the threat-defining risks, which depend on the main variables, i.e. soil surface roughness, sex, age of the exposed individuals and the moment of the release (day or nighttime). This study aims to evaluate the effect of soil surface roughness on the radiological risk. The simulation was performed by an analytical method using the HotSpot Health Physics code within the first 100h. The results found relevant elements that allow for differentiating consequences as a function of the time of release (whether daytime or nighttime), thus allowing decision-makers to be supported with a little more detail about the situation, although in a critical initial phase.

Analysis of High-Pressure extended SBO accident in VVER-1000: A sensitivity study on lower plenum structures modeling and core region nodalization in MELCOR

The lower plenum structures of VVER-1000 reactors exhibit a unique shape. This paper aims to investigate the sensitivity of various important parameters related to safety using different models in the MELCOR 1.8.6 code's COR package. The study focuses on modeling the lower plenum structures and the core region using different CVH package nodalizations during a High Pressure Extended Station Black-Out accident. At first, MELCOR calculated parameters are investigated during the course of a severe accident for a basic model of the lower plenum structures and core region nodalization. Different important parameters from a safety point of view are also examined. These include the total amount of in-vessel hydrogen production, the lower head breach time and elevation, the synchronicity between lower head breach and debris ejection, the total amount of hydrogen produced during molten corium concrete interaction, the total amount of ejected debris to the cavity, and the total amount of radioactive material release. Then, by considering sixteen different simulations corresponding to four lower plenum structures modeling and four core region nodalizations, the sensitivity of the mentioned parameters is investigated. The obtained results show a sensitivity of about 12.7% for the timing of the accident (equivalent to), 103% for the hydrogen produced inside the vessel (equivalent to 417.9 to 847.8 kg), 143% for the hydrogen produced outside the vessel (equivalent to 445.7 to 1081.4 kg), 47% for the total hydrogen produced (equivalent to 1207.5 to 1780.7 kg), and 76% for total radioactive material release (equivalent to 625.311 to 1101.892 kg) up to 24 h after the accident.

Evaluation of the probability of overflow from the Abadia de Goiás repository by the Fokker-Planck equation using the Trotter’s formula

A model was constructed with the objective of developing the Fokker-Planck equation for the calculation of the reliability attributes required for risk-informed decision-making involved in the analysis of water infiltration inside the repository located in the Abadia de Goiás municipality, in the metropolitan region of the city of Goiânia. The solution of the Fokker-Planck equation was analytically obtained by the Trotter’s method for a near surface repository. This solution provides the probability density and the probability functions for the height of the liquid (water + radionuclides) column, generated by the infiltrated water in the concrete crypt of the repository. The modeling was unidimensional, and the calculations were performed for the geometric center of the repository crypt, which is the place of greater sensitivity of occurrence of accidental scenarios. The concept of the present model can be extended to other places in the crypt without loss of generality. The developed model is appropriate to analyze the repository overflow, a kind of event that occurs when the height of the liquid column inside the repository reaches the value of the inner height of the concrete crypt of the facility. This accident scenario produces the greatest release of radioactive material to the region outside the repository and, consequently, it is responsible for the main radiological impact for both members of the public and the environment. Calculations performed for the Abadia de Goiás repository indicated that the overflow probability for this facility is strongly dependent on the rainfall rate and also on the institutional time period, varying from about 3% with no rainfall penetration up 99% with the maximum rainfall rate for an institutional time period control of 300 yr. Such results can help making decisions on the problem of risk scenarios related to the disposal of radioactive waste.

A Chinese–French Study on Nuclear Energy and the Environment

This article focuses on the environmental impact of nuclear energy and addresses the following major environmental issues associated with nuclear power generation: ① controlling the radioactive discharge from nuclear installations under normal operation and evaluating their non-radioactive environmental impact (water withdrawals and non-radioactive discharges); ② long-term management of spent fuel and radioactive waste (radwaste), notably that disposed off in geological repositories; ③ prevention and mitigation of severe nuclear accidents and their radioactive releases and; ④ improving nuclear safety to restrict its environmental impact and to contribute toward the public acceptance of nuclear energy. Nuclear energy, with its very low emissions of green-house gases, has a unique capacity to generate massive and on-demand dispatchable amounts of electricity. The annual effective radiation dose delivered to the public surrounding nuclear power plants under normal operation is negligible. Considerable efforts have been made to define sustainable management of high-level long-lived radwaste that is disposed in geological formations. The return of experience from severe nuclear accidents in the past has informed and propelled major improvements in several aspects of nuclear energy production—including reactor design and operational management as well as in the development of accident-management guidelines—and has proved to be highly valuable. The environmental risks in the event of a severe accident have been substantially reduced and protocols have been established to minimize the release of radioactive materials and avoid the large-scale evacuation of people in the event of a severe nuclear accident. Efforts must be continued to improve reactor safety and enhance the transparency of the industry and the authorities that support and control nuclear power to further reduce the environmental impact.