Fuel Storage Facility Research Articles

Monitoring the technical condition of the fuel storage facility at the Ukrainian Antarctic Akademik Vernadsky station

Monitoring the technical condition of the fuel storage facility at the Ukrainian Antarctic Akademik Vernadsky station

МЕДИКО-САНИТАРНЫЕ ПОСЛЕДСТВИЯ РАДИАЦИОННЫХ АВАРИЙ ПРИ ОБРАЩЕНИИ С ОТРАБОТАННЫМ ЯДЕРНЫМ ТОПЛИВОМ

Purpose: To consider the features of potential health effects for personnel and for the population in the event of radiation accidents at facilities for post-reactor spent nuclear fuel management (including spent nuclear fuel pools and general plant storage facilities, transportation of nuclear materials and their subsequent processing at radiochemical plants). Results: Based on the analysis of publications, the potential radiation consequences for personnel and for the population in the event of various types of radiation accidents at the stages of spent nuclear fuel management were systematized. Conclusion: Advanced technologies underlying the closed nuclear fuel cycle provide a high level of radiation safety for personnel and for the population. At the same time, a large amount of accumulated radioactive and nuclear materials makes it necessary to maintain the health care system in preparedness at all stages of spent nuclear fuel management. Health effects for personnel may be associated with exposure because of criticality accident and internal intake of fission products of uranium and actinides (by inhalation and through wound surfaces). In case of accidents at radiochemical production, combined radiation-thermal and radiation-chemical injuries are also possible. The main potential hazard for the population in the event of radiation accident at the spent nuclear fuel storage facility is the contamination of the environment and exposure to long-lived uranium fission products and actinides. This requires clarification and development of appropriate criteria and derived intervention levels for making decisions on protective measures. Hypothetically, the health effects in case of radiation accident when managing “fresh” spent nuclear fuel can be comparable to a large-scale reactor accident and require urgent protective measures, including evacuation and iodine prophylaxis. An important factor that should also be taken into account when planning health care measures is the need for decontamination of victims.

Hydrotalcite Colloidal Stability and Interactions with Uranium(VI) at Neutral to Alkaline pH.

In the United Kingdom, decommissioning of legacy spent fuel storage facilities involves the retrieval of radioactive sludges that have formed as a result of corrosion of Magnox nuclear fuel. Retrieval of sludges may re-suspend a colloidal fraction of the sludge, thereby potentially enhancing the mobility of radionuclides including uranium. The colloidal properties of the layered double hydroxide (LDH) phase hydrotalcite, a key product of Magnox fuel corrosion, and its interactions with U(VI) are of interest. This is because colloidal hydrotalcite is a potential transport vector for U(VI) under the neutral-to-alkaline conditions characteristic of the legacy storage facilities and other nuclear decommissioning scenarios. Here, a multi-technique approach was used to investigate the colloidal stability of hydrotalcite and the U(VI) sorption mechanism(s) across pH 7–11.5 and with variable U(VI) surface loadings (0.01–1 wt %). Overall, hydrotalcite was found to form stable colloidal suspensions between pH 7 and 11.5, with some evidence for Mg2+ leaching from hydrotalcite colloids at pH ≤ 9. For systems with U present, >98% of U(VI) was removed from the solution in the presence of hydrotalcite, regardless of pH and U loading, although the sorption mode was affected by both pH and U concentrations. Under alkaline conditions, U(VI) surface precipitates formed on the colloidal hydrotalcite nanoparticle surface. Under more circumneutral conditions, Mg2+ leaching from hydrotalcite and more facile exchange of interlayer carbonate with the surrounding solution led to the formation of uranyl carbonate species (e.g., Mg(UO2(CO3)3)2–(aq)). Both X-ray absorption spectroscopy (XAS) and luminescence analysis confirmed that these negatively charged species sorbed as both outer- and inner-sphere tertiary complexes on the hydrotalcite surface. These results demonstrate that hydrotalcite can form pseudo-colloids with U(VI) under a wide range of pH conditions and have clear implications for understanding the uranium behavior in environments where hydrotalcite and other LDHs may be present.

To the question of assessment reliability of the thermal regime of an underground spent nuclear fuel storage facility

To the question of assessment reliability of the thermal regime of an underground spent nuclear fuel storage facility

Analysis of Loss of Water Inventory at the Philippine Research Reactor-1 Fuel Storage Facility

The Philippine Research Reactor-1 (PRR-1) fuel storage facility is a wet storage for irradiated nuclear fuel. The behavior of the facility in case of accidental loss of water is currently unknown. Safety analysis of a repository of fissile material involves criticality safety assessment and radiation dose estimation for normal and accident scenarios. To determine the radiological consequences of loss of water inventory in the fuel storage facility, we calculated the effective multiplication factor (keff) and gamma dose rate distribution in the system using Monte Carlo N-particle (MCNP) radiation transport code. Results show that keff will decrease from a maximum of 0.61424 ± 0.00013 when the water level is decreased from its fully moderated condition. In contrast, keff will increase to a maximum of 0.7468 ± 0.0002 when voids are introduced in the water. These results indicate that the system will remain subcritical (keff<1) and a runaway supercritical fission chain reaction cannot result from any moderating conditions. Furthermore, the dose assessment for the case of complete loss of water revealed a maximum dose rate of 5.52 ± 0.045 mSv/h and 288 ± 4.65 𝛍Sv/h at the tank surface and tank platform, respectively. These values are significantly above background radiation levels; however, the calculated spatial distribution is highly asymmetric and the dose rate falls off rapidly with increasing distance. These results suggest that a facility-level emergency response is sufficient to address the accident. Our findings can augment the existing facility emergency procedures, thereby contributing to the improvement of radiological safety.

Methodology for Determining the Thermal and Thermal-Stress States of a Concrete Storage Container for Spent Nuclear Fuel for Assessment of Its Service Life

The work is devoted to the development of methodologies for determining the thermal and thermal-stress states of the main equipment in dry container storage facilities for spent nuclear fuel. Storage facilities of this type are most common for spent fuel of nuclear power reactors. The safety of storage equipment in terms of assessing its service life is not covered widely enough in the world scientific literature. In particular, there are no effective methods for calculating the thermal and thermal-stress states of the equipment that would take into account the influence of many external factors throughout the life of a storage facility. To assess the thermal state of the containers, forward conjugate heat transfer problems, accounting for the mutual heat transfer in both a solid body and in the fluid environment (air), are proposed to be solved. Based on the solution of the conjugate heat transfer problems, the boundary conditions are to be determined to further assess the thermal-stress state of storage containers using inverse heat transfer problems. The proposed approach to determining the thermal and thermal-stress states of a concrete spent fuel container will promote more effective methods for assessing the service life of dry spent fuel storage facilities, which is, in turn, necessary in the development of ageing management programs for storage equipment and long-term safe operation.

Radiation survey and environmental impact assessment at the site of temporary storage at Andreeva Bay (16 years of studies)

The site of temporary storage for spent fuel and radioactive waste at Andreeva Bay is a largest nuclear legacy site in the Russian Northwest. The radiation situation within the industrial site and in the surrounding health protection zone (HPZ) is characterised by significant local contamination of the topsoil, which is a source of environmental contamination and potential spread of radioactivity within and outside the HPZ, including the adjacent marine environment. The highest levels of radioactive contamination of soil due to manmade radionuclides have been registered around the spent nuclear fuel storage facilities, where the specific activity of Cs-137 reaches 5 × 107 Bq kg−1, and that of Sr-90, 5.7 × 106 Bq kg−1. Determination of the mobility of these main dose-forming radionuclides has demonstrated that about 90% of the Sr-90 and 30% of the Cs-137 in the soil are in a mobile form. The corresponding figures for bottom sediments are 75% and 60%. The assessment of the impact of contamination in the different areas of the Andreeva Bay site, and the assessment of the adequacy of the previously developed remediation criteria for environmental protection (Sneve et al 2015 J. Radiol. Prot. 35 571) demonstrated that when carrying out remedial work according to any scenario provided for in the Guidelines ‘Criteria and norms for remediation of sites and facilities contaminated with manmade radionuclides’, of the Federal State Unitary Enterprise ‘The Northern Federal Enterprise for Radioactive Waste Management’ for the supervision area, recommendations for environmental protection are followed. For scenarios such as conversion and conservation, reference levels of biota exposure for some species of representative organisms could be exceeded, based on a preliminary conservative assessment.

Experimental investigation on pressure drop of a PWR fuel assembly under low Re conditions

The accurate calculation of the flow resistance of the fuel assembly is crucial for the design and development of the advanced reactors, spent fuel pools, and spent fuel storage facilities. A low-Re experiment was carried out to measure the flow resistance of a full-size AFA3G 17 × 17 fuel assembly. The present experimental data shows that the prediction error of the Cheng-Todreas correlation is large in the laminar and the transition regime. Hence, a modified Cheng-Todreas correlation is proposed based on the present experimental data to improve its prediction accuracy. In addition, the modified Cheng-Todreas correlation is compared with the experimental results in the literature, and the results proved that the modified Cheng-Todreas correlation does reduce the prediction error compared to the original model.

Separation of Flux/Dose Contributions from Multiple Transport Pathways in Monte Carlo Shielding Analyses of a Consolidated Interim Spent Nuclear Fuel Storage Facility

The characteristics of the radiation field around a consolidated interim spent nuclear fuel storage facility were investigated comprehensively through Monte Carlo simulations. Neutron and gamma-ray flux/dose contributions from multiple transport pathways, including direct, streaming, skyshine, groundshine, and multishine, were isolated using a modified version of the method that was originally developed by Oh et al. [J. Korean Phys. Soc., Vol. 69, 1057 (2016)] for the evaluation of neutron skyshine from a high-energy electron accelerator. The application of the methodology was demonstrated in this paper, and the flux/dose contributions of individual pathways were examined and compared. The results provided additional insight into how the radiation propagated from the source to off-site locations. The modified method for separating five transport pathways can provide valuable information for shielding optimization during the design phase and is generally applicable to Monte Carlo shielding analyses of other nuclear facilities.

Time-dependent variations in the radiological health impact of an interim spent fuel storage facility

Radiation shielding calculations on spent fuel casks are an inherent part of the nuclear analyses performed for the site safety report (SSR) when siting an interim spent fuel storage facility (ISFSF). The basis for this is a legislative requirement that the licensee must demonstrate to the nuclear regulator that no member of the public will be exposed to a radiation health risk as result of the operation of the ISFSF. Because of the enormous number of isotopes that are generated as result of the operation of a nuclear reactor, shielding calculations have to perform the following analyses which are the objectives of this research paper: (a) identify a group/set of nuclides that are important in radiation shielding with a view to ensuring that the shielding material/strategy adopted envelops the radiation emissions of all nuclides so as to prevent a deficiency in the radiation protection measures adopted (i.e. optimisation of protective measures); (b) determine the mathematical expression that links the behaviour (i.e. growth or decay) as a function of time within a stipulated time frame (i.e. the boundary conditions); (c) develop a mechanism for ranking nuclide importance in shielding; (d) determine whether there is a relationship between taking credit for burnup or not (Leotlela et al., 2015; Leotlela et al., 2020) and the resulting effective dose; and (e) determine whether there is a relationship between the effective dose and the duration of cask storage in the ISFSF.

Numerical simulation of the thermal regime of an underground spent fuel storage facility (built-in structure variant)

The results of a numerical simulation of the thermal regime of an underground facility for long-term storage of spent nuclear fuel in a built-in reinforced concrete structure are presented. Two computer models were constructed in a three-dimensional formulation in the COMSOL programme. The first model is based on the incompressible fluid approximation, while the second model is based on the "incompressible ideal gas" approximation. The mathematical basis of models: the continuity equation, Navier - Stokes equations averaged by Reynolds, the standard (k - ε) turbulence model, and the general heat transfer equation. Consideration of mixed convection conditions is implemented in the "incompressible ideal gas" approximation, where the air density is a function of temperature only. The most thermally stressful arrangement of spent fuel placement is investigated: U-Zr - defective - U-Be. The air rate is varied in the range from 21 to 0.656 m3/s. Numerical experiments were performed for up to 5 years of fuel storage. The principal difference between the non-stationary structure of the velocity fields predicted in the "incompressible ideal gas" model and the "frozen" picture of the aerodynamic parameters in the incompressible fluid model is emphasized. It is shown that the requirements for exceeding the temperature limit values are met when the object operates under conservative ventilation conditions (rate 0.656 m3/s) with a minimum of costs for organizing ventilation. The dynamics of heat flows directed into the rock mass through the base and from the surface of the built-in structure of the U-Zr fuel compartment to the air environment are analyzed. The predominance of the heat flow from the surface of the structure and the different times when the curves of the heat flow dynamics reach their maximum values are noted. The heat flow to the array reaches its maximum significantly faster than to the air.

A numerical study of thermal behavior of CASTOR RBMK-1500 cask under fire conditions

Casks are widely used for interim storage of spent nuclear fuel after its pre-storage in water pools. Interim storage of spent nuclear fuel is a very important part in the whole cycle of nuclear energy generation, so the casks must be designed according to relevant safety standards. Like criticality and doses, thermal safety is an important condition for spent nuclear fuel storage in a dry spent nuclear fuel storage facility and should be analyzed thoroughly. During the entire period of interim storage of spent nuclear fuel, normal operation and emergency situations should be investigated. Analysis of various accident scenarios is one of the means to prevent damage of storage casks and spent nuclear fuel. There are many publications on thermal cask behavior during normal operation conditions, yet publications on cask behavior during accidents are rather limited. There are a few publications presenting fire impact analysis of thermal behavior of casks during transportation. Therefore, this paper presents a numerical study that has been performed to evaluate thermal behavior of a cask under fire condition loaded with spent nuclear fuel stored in an open type dry storage facility pre-stored in a water pool for five years. Thermal performance of a 2D vertical CASTOR RBMK-1500 cask under such conditions (steady-state, fire, post-fire) has been numerically studied by using the ANSYS Fluent code. Also, a local sensitivity analysis of parameters that mostly influence the cask’s temperature was performed.

Skyshine Calculations for a Large Spent Nuclear Fuel Storage Facility with SCALE 6.2.3

The SCALE code system developed at Oak Ridge National Laboratory includes state-of-the-art capabilities for radiation source term and radiation transport simulations that can be used in numerous applications, including dose rate analyses of complex consolidated interim storage facilities (CISFs). A licensed CISF could be used to store tens of thousands of tonnes of spent nuclear fuel discharged from commercial power reactors using various cask and storage pad designs. A CISF design must comply with the regulatory requirements provided in 10 CFR Part 72, including requirements related to annual dose limits applicable to real individuals located beyond the area controlled by the licensee. Therefore, calculating a dose to the public is a necessary part of the licensing process for the construction of a CISF. These calculations are very challenging because of the complexity of the CISF design and the low magnitude of dose rate at large distances from the facility. This paper describes detailed far-field dose rate calculations performed for a proposed CISF using MAVRIC, the Monte Carlo radiation shielding sequence in SCALE 6.2.3, with automated variance reduction based on discrete ordinates calculations. The method presented in this paper uses a detailed Monte Carlo radiation transport simulation in one step from source to dose rate. A series of independent simulations was made using the complete site geometry (all casks present), but with only one cask containing radiation sources to obtain the dose rate maps produced by each storage cask. The CISF dose rate map was obtained by adding the dose rate maps produced by the independent individual cask simulations. Ample volumes of air and soil extending beyond the location of interest for dose rate calculation were included in the calculation model to properly simulate important radiation attenuation and scattering events that affect far-field dose rates. A comprehensive sensitivity study is included in this paper to illustrate the importance of selecting appropriate air volume, mass density, and composition for CISF skyshine dose rate calculations. Dry soil and soil containing water were analyzed to determine their effects on groundshine radiation.

Methods and Technologies for Creating a New Generation of Fuel Storage and Transportation Facilities

The main scientific an practical results in the field of creating petroleum products supply facilities of a new generation based on the modified polymeric materials are presented. The technical requirements, evaluation methods, techniques and technologies for the modification of polymeric materials to create elastic tanks and pressure lay-flat pipelines for fuel are presented. The compliance of the operational properties of the developed fuel storage and transportation facilities with the modern requirements has been established.

Моделювання та візуалізація радіаційного стану на майданчику зберігання контейнерів ЦСВЯП за допомогою програмного комплексу ChNPP VRDose Planner

The results of three-dimensional modeling of working conditions of personnel in radiation-hazardous conditions of the Centralized Spent Nuclear Fuel Storage Facility (CSNFSF) of WWER type reactors of domestic NPPs, are presented using the software complex ChNPP VRDose Planner. Today, the software package ChNPP VRDose Planner, developed by the Institute of Energy Technology (Norway), is an effective tool for estimating the does load on personnel. It allows interpreting radiological information based on three-dimensional models of radiationhazardous objects where human activities are planned. The VRDose software complex allows creating virtual models of each worker and reproducing detailed technological process of performance of works with participation of the person. This creates a visual representation of the behavior of staff at all stages of work. The software complex is created for planning and implementation of works in difficult radiation conditions, allows carring out optimization of radiological protection and choosing the safest scenarios of performance of operations. Using this software package: 1) the predicted values of the dose rate on the surface of the container model and the model of the storage site were obtained; 2) the dose for the dosimeter was simulated and evaluated during the operation no. 32 “Control of gamma dose rate and neutron radiation of the loaded storage container HI-STORM at the storage site” for the time of operation on arrival and departure with the arrival and departure from the venue; 3) the contribution of the main emitting nuclides to the dose rate when measured on each side of the container is determined; 4) the possibility of its use to control the radiation impact on the environment and its minimization on the person during the construction and commissioning of the CSNFSF was assessed. By verifying the results obtained during the work on scientific and technical support of construction and commissioning of CSNFSF with the data obtained using the software package ChNPP VRDose Planner, it can be argued that they are sufficiently correlated. Given the above, the software package ChNPP VRDose Planner can be recommended for use during the work on scientific and technical support of the operation of CSNFSF and control of radiation exposure to the environment and its minimization for personnel for other design work to be performed, as part of work on the transformation of the Shelter object into an ecologically safe system, as well as in other projects of construction, operation and decommissioning of facilities using radiation and nuclear technologies.

Тривимірна нейтронно-­фізична модель системи зберігання відпрацьованого ядерного палива

The management of spent nuclear fuel is one of the most pressing problems of Ukraine’s nuclear energy. To solve this problem, as well as to increase Ukraine’s energy independence, the construction of a centralized spent nuclear fuel storage facility is being completed in the Chornobyl exclusion zone, where the spent fuel of Khmelnytsky, Rivne and South Ukrainian nuclear power plants will be stored for the next 100 years. The technology of centralized storage of spent nuclear fuel is based on the storage of fuel assemblies in ventilated HI-STORM concrete containers manufactured by Holtec International. Long-term operation of a spent nuclear fuel storage facility requires a clear understanding of all processes (thermohydraulic, neutron-physical, aging processes, etc.) occurring in HI-STORM containers. And this cannot be achieved without modeling these processes using modern specialized programs. Modeling of neutron and photon transfer makes it possible to analyze the level of protective properties of the container against radiation, optimize the loading of MPC assemblies of different manufacturers and different levels of combustion and evaluate biological protection against neutron and gamma radiation. In the future it will allow to estimate the change in the isotopic composition of the materials of the container, which will be used for the management of aging processes at the centralized storage of spent nuclear fuel. The article is devoted to the development of the three-dimensional model of the HI-STORM storage system. The model was developed using the modern Monte Carlo code Serpent. The presented model consists of models of 31 spent fuel assemblies 438MT manufactured by TVEL company, model MPC-31 and model HISTORM 190. The model allows to perform a wide range of scientific tasks required in the operation of centralized storage of spent nuclear fuel.

Оптимізація часу і статистичної похибки розрахунку біологічного захисту контейнерів зберігання відпрацьованого ядерного палива

The radiation protection is an important issue in the operation of nuclear power plants and artificial radioactive sources, which include spent nuclear fuel storage facilities. The Monte-Carlo codes are the effective instruments for calculation of radiation shielding properties and radiation field characteristics for complex geometries. However, achieving a satisfactory statistical error of the results in modeling the passage of neutrons and photons through biological protection may require excessively long calculation time. To solve this problem, Monte-Carlo codes use methods to reduce the variance to direct particles to regions with detectors to improve statistical accuracy. Our paper presents the application of the variance reduction function based on weight windows in the Monte-Carlo Serpent code, the function is investigated on the example of a simplified 2D model of the spent nuclear fuel storage container HI-STORM 190. The simple approach option of variance reduction with fixed cartesian and cylindrical meshes was investigated for different mesh nodes and for different dimensions of nodes. Also, global variance reduction option with fixed cartesian and cylindrical meshes was analyzed for case of achieving satisfactory results for the entire simulated volume. For a qualitative assessment of the variance reduction function, the indicator — figure of merit (FOM) used in our paper which proposed by the developers of the Serpent Monte-Carlo code. It is shown that the use of the variance reduction function leads to a significant decrease of statistical error and decrease of the calculation time, and therefore can be useful for biological protection calculations. As conclusions we can note that: the cylindrical mesh is not as effective in terms of FOM compared to Cartesian mesh; for both cylindrical and Cartesian meshes it is possible to find the recommended grain (node) size; the use of azimuthal partition of the cylindrical mesh together with radial partition leads to an increase in FOM; the application of global variance reduction is useful in the case of asymmetric biological protection geometries, while the FOM decreases.

Modeling and validation of RELAP5 for natural circulation flow in a single PWR fuel assembly

The natural circulation loops have been widely used in the design of next generation nuclear reactors and spent fuel storage facilities. However, the friction correlations for fuel bundle and grid spacers in RELAP5 code are not applicable for natural circulation flow. This paper summarizes the modeling methods in previous researches, and proposed a flow resistance model for rod bundle and grid spacers based on experimental data. A RELAP modeling method of radiation heat transfer in fuel assembly is proposed for natural circulation flow of air in the fuel assembly. The new model is verified using experimental data from the Spent Fuel Project by NRC. The results of the new model are in good agreement with the experimental data. The current model will be useful for analyzing the flow and heat transfer characteristics of natural circulation in the reactor core of next generation nuclear reactors and spent fuel storage facilities.

The use of long-span coating plates in case of reconstruction of “wet” spent nuclear fuel storage facilities

Spent nuclear fuel is one of the most dangerous threats on Earth. Its storage is carried out in various facilities, which include stand-alone “wet” storage facilities for spent nuclear fuel (SNFS). Such storage facilities were designed and built decades ago, when there were no industry requirements for accounting for extreme external impacts. Given the design features of the “wet” SNFS, in the event of their reconstruction it is advisable to consider the possibility of replacing the frame part of the structures with monolithic reinforced concrete elements that can withstand earthquakes and other effects of the required intensity.

Study of neutron energy and directional distribution at the Beloyarsk NPP selected workplaces

Energy and directional distribution of neutrons at the Beloyarsk NPP workplaces is a subject of this study. Measurements of H∗(10) rate and neutron energy distribution were taken at 8 workplaces, which can be divided into three categories: work with spent or fresh nuclear fuel, work with radionuclide neutron sources, work at the rooms adjusted to reactors. The Hp(10) measurements were performed only at 6 out of 8 locations, due to the fact that long term placing of an effective neutron moderator in fresh nuclear fuel storage facility is forbidden. As a result of the research energy and direction distribution of the neutron fields at 8 locations of the Beloyarsk NPP workplaces was obtained. To estimate the accuracy of the H∗(10) rate and Hp (10) measurements the reference values of dose equivalents were calculated using energy and directional distribution. To take into account the difference between the reference values and the measured results site-specific correction factors were calculated.