Spent Fuel Storage Research Articles

Operational Experience in Degradation (Ageing Effects) and Implementation of Ageing Management Programmes for Dry Storage Systems in Ukraine

The Coordinated Research Projects (CRPs) of International Atomic Energy Agency (IAEA) is a power tool to engage simultaneously a lot of different countries that operate Spent Fuel Storage Facilities for solving important safety issues at a deep level of competence and expertise. One of the topics that needs thorough research is behavior and state assessment of spent fuel assemblies and fuel rods during long/very long term storage in spent fuel storage facilities. It is naturally that over a long period, special attention is drawn to degradation mechanisms, ageing effects and ageing management programs that should be implemented in order to avoid and mitigate the impact of degradation mechanisms on spent fuel elements. This article describes the first step of implementing new IAEA CRP project “T13019. Performance Assessment of Storage Systems for Extended Durations (PASSED)” with contribution from the State Scientific and Technical Center for Nuclear and Radiation Safety (SSTC NRS), which conducts research regarding the “Operational experience on degradation (ageing effects) of dry storage systems” during the first stage” at the first stage. The overall objective of the research proposed by the SSTC NRS is to strengthen technical knowledge, experience and understanding of the long-term behavior of the dry spent fuel storage systems, and of the currently applied inspection and monitoring technologies. At the same time, a specific objective of involving the SSTC NRS in this project was not only to present national experience and practices, but also to study the existing experience of the participating countries, which also joined the CRP proposed by the IAEA. The article considers the existing storage systems for spent nuclear fuel at the Chornobyl and Zaporizhzhya NPP sites, provides examples of the aging management system for components and structures, as well as examples of the issues that require development, implementation and maintenance of aging management programs.

Application of an Independent Temporary Spent Fuel Storage Pool Cooling System for Decommissioning of Pressurized Water Reactor Kori Units 3 and 4

The operating license of Kori units 3 and 4 are to be expired in 2024 and 2025. If the plants are decided to be decommissioned, the spent nuclear fuels in the spent fuel storage pool (SFP) have to be removed from the site. However, no proper storage facility is currently available in Korea. To overcome the difficulty, this study proposes application of an independent temporary SFP cooling system. It is expected to safely isolate the SFP from the plant. The case study conducted showed this concept is achievable and technically mature. Moreover, the installation cost per unit is USD 9.51 million obtained from an analogous estimating. This is much lower cost than the dry cask options. Then, system requirements and applicable design are developed. Maximum heat generation of unit 4 is estimated as 1.67 MW, through fuel characteristics categorizations and ORIGEN calculations. To remove the decay heat, installation of an indirect air-cooled method is selected by evaluating three cooling methods. The modification includes new heat exchanger, secondary cooling pumps, and chillers. Furthermore, two single failures are considered; cooling pump and normal power. This can be covered by two redundant trains and a back-up diesel generator provided.

乏燃料贮存格架结构改进设计研究

乏燃料贮存格架结构改进设计研究

Evaluation of functional loss to radiation detector in tomographic device for spent-fuel inspection by high-energy photons and neutrons: A preliminary study

The aim of this study was to investigate functional loss of radiation detectors used for spent fuel inspections due to radiations from spent fuel. Dose to a radiation detector was evaluated by Geant4 simulation. Subsequently, the radiation detector measured counts, according to the irradiation time, for the test source with presence of external 60Co and 252Cf sources in almost identical condition to those of the simulation, where gamma and neutron emission rates of the sources were similar to the actual spent fuels. The counts during the 60Co irradiation were changed with discrepancy of 1.17% while neutron beam decreased more than 25% of total counts after its irradiation. Additionally, abnormalities in the detector’s measured background counts were observed following the neutron irradiation. The results can be applied to efforts to define the tolerance dose for radiation detectors used in neutron-abundant environments such as nuclear power plants or spent-fuel storage facilities.

A Study on the Decontamination of Cs-137 and Sr-90 Contained in the Liquid Radioactive Waste Discharged from the Spent Fuel Storage Tank Using Microalgae

A Study on the Decontamination of Cs-137 and Sr-90 Contained in the Liquid Radioactive Waste Discharged from the Spent Fuel Storage Tank Using Microalgae

Tolerance assessment and crack growth of chloride-induced stress corrosion cracking for Chinshan dry storage system

Evaluating the susceptibility of canister used in Chinshan Independent Spent Fuel Storage Installation (CSISFSI) to chloride-induced stress corrosion cracking (CISCC) has been an aging management plan in an initial license period of 20 years. The presence of concentrated chlorides, temperature and relative humidity, and sufficient weld residual stress are the three key environmental conditions for atmospheric CISCC. A single 2-D model was used to simulate natural exposure of austenitic stainless steel (SS) of SS304 and SS304L dry storage system (DSS) for transportable storage canister (TSC). This model will calculate the temperature on the TSC surface based on CFD method. This study estimated RH and DRH on the TSC surface by the realistic environmental conditions (e.g., ambient RH, ambient temperature, and the temperature on the TSC surface) after 1st, 10th, and 20th years, respectively. The flaw tolerance to part-depth (PD) and through-wall (TW) critical flaw were evaluated, and the two-phase CGR induced by CaCl2 or MgCl2 of the TSC surface under 0.18 m height were predicted. The results show that the fastest CISCC CGR was on the bottom induced by CaCl2 in the 20th year, and the slowest CGR was at the 0.18 m of TSC induced by MgCl2 in the 10th year. The two-phase CGR trends were very close if the deliquescence threshold had been considered under different environmental conditions.

Study on a High-Boron-Content Stainless Steel Composite for Nuclear Radiation.

In this research, a high-boron-content composite material with both neutron and γ rays shielding properties was developed by an optimized design and manufacture. It consists of 304 stainless steel as the matrix and spherical boron carbide (B4C) particles as the functional particles. The content of B4C is 24.68 wt%, and the particles’ radius is 1.53 mm. The density of the newly designed material is 5.17 g·cm−3, about 68.02% of that of traditional borated stainless steel containing 1.7 wt% boron, while its neutrons shielding performance is much better. Firstly, focusing on shielding properties and material density, the content and the size of B4C were optimized by the Genetic Algorithm (GA) program combined with the MCNP program. Then, some samples of the material were manufactured by the infiltration casting technique according to the optimized results. The actual density of the samples was 5.21 g cm−3. In addition, the neutron and γ rays shielding performance of the samples and borated stainless steel containing 1.7 wt% boron was tested by using an 241Am–Be neutron source and 60Co and 137Cs γ rays sources, respectively, and the results were compared. It can be concluded that the new designed material could be used as a material for nuclear power plants or spent-fuel storage and transportation containers with high requirements for mobility.

Utilizing of Spent Fuel Storage Pool in Gamma Irradiation Experiments

Abstract The objective of this study is suggesting a new method for gamma irradiating experiment using spent fuel elements (SFEs). An irradiation device was developed to contain the sample during the gamma irradiation process under water in spent fuel storage pool. The maximum burned fuel elements were used as a source of gamma rays with intensity and energy spectrum depend on the decay time from the reactor core withdrawal. A SFE configuration was introduced to verify an ideal condition of irradiating the sample with uniform gamma flux. The gamma dose rate inside the sample chamber was calculated using SCALE/MAVRIC code for the SFEs configuration source at decay times ranging between 0 and 5 years. A relationship was introduced to obtain the required accumulated dose depending on choosing the proper SFE configuration with the proper irradiation time.

Study on Heat Transfer Characteristics and Optimization of Spent Fuel Storage and Transportation Containers Based on Heat-Fluid-Solid Coupled Model for Radiation, Convection and Conduction

Based on the heat-fluid-solid coupled computational fluid dynamics (CFD) model for radiation,convection and conduction, this paper compares the changes in maximum temperature, natural convection flow rate, outer surface radiation and convection power of various parts under different storage status and ambient temperatures, with different packing media in the containers, and with or without annular fin. According to the study results, horizontal arrangement improves the convection heat transfer on the annular fin; each time the ambient temperature increases by 10 ℃, the fuel cladding temperature and outer wall surface temperature increase by 6.5 ℃ and 8.3℃, respectively; when the medium filling the fuel basket is changed from neutron absorption plate or aluminum block to helium, the overall heat transfer capacity of the container deteriorates dramatically, with the temperature inside the container increasing and the outer wall surface temperature remaining almost unchanged; with the solar radiation ignored, a annular fin will help improve convection heat transfer and thus decrease the overall temperature of the container, otherwise, the annular fin absorbs more solar energy, causing the overall temperature of the container to exceed that of a container with smooth wall; and the CFD radiation power is basically same as the calculation results gained in the formula method, as shown in the verification of the CFD radiation model based on the algebraic analysis method and diffuse gray surface model.

Methodology to incorporate damaged BWR spent fuel as contents in the NAC-UMS® storage system

Following Taiwan’s nuclear waste management policy, INER launched Taiwan’s first Independent Spent Fuel Storage Installation (ISFSI) project in 2005. The technology of dry storage system adopted is NAC-UMS system. INER made some design modifications of NAC-UMS system in order to accommodate the constraints at Chin-Shan Nuclear Power Plant (CSNPP). The modified UMS was named as INER High Performance System (INER-HPS). The corresponding Safety Analysis Report (SAR) of ISFSI was approved by AEC on November of 2008. Through years of dedication, the first dry run test has been accomplished on November 2012. INER-HPS is capable to cover most of the spent fuels dry storage needs in Taiwan with current NAC-UMS License. The current CoC for NAC-UMS includes undamaged BWR fuel with a burn-up of not greater than 45 GWD/MTU. For the purpose of NAC-UMS to incorporate damaged BWR fuel as approved contents, INER and NAC further enter into cooperation to developing and licensing NAC-UMS to include up to four BWR damaged fuel cans and a new BWR damaged fuel basket. This paper summarized the project progress, results of trial fabrication of damaged fuel can, and the methodology to incorporate damaged BWR spent fuel as contents in the NAC-UMS storage system.

Development of Underwater Lifting Tool for Spent Fuel Storage Rack in Nuclear Power Plants

In response to the requirement for the underwater lifting of spent fuel storage racks in a second-generation nuclear power plant, according to the current situation of the narrow underwater layout of spent fuel storage racks and the analysis of the safety and convenience of underwater lifting interfaces, a grab-and-release rack was developed. Bottom-bottom lattice lifting tool with underwater self-locking lattice is realized by running water holes on the bottom of the rack, and the structure, operation and working principle of this underwater lifting tool are introduced in detail. Finally, the lattice is used in a second-generation nuclear power plant for a lifting operation of 20 rack frames in the factory. The field application shows that the lifting tool is easy to operate and reliable in structure, which meets the requirements for the underwater lifting of the lattice frame.

Comparison of Gamma and Neutron Dose Rate Variations with Time for Cast Iron and Metal–Concrete Casks Used for RBMK-1500 Spent Fuel Storage

The present SF management concept in Lithuania envisages that spent RBMK-1500 fuel will be stored in dry storage containers for 50 years, before being disposed of in a deep geological repository. However, the risk that a deep geological repository will not be constructed at the planned time should be taken into account, and the extension of SF storage over 50 years should be considered. This paper presents a comparison of gamma and neutron dose rate distributions and variations with planned and extended storage times for cast iron and metal–concrete containers loaded with RBMK-1500 SF. All calculations were performed using the SCALE computer codes system. The modeling results show that the overall shielding properties of the CONSTOR® RBMK-1500 container containing the same neutron and gamma sources are better than those of the CASTOR® RBMK-1500 container. During an extended storage period (from 50 to 300 years), the total dose rate would decrease considerably and the dose rate due to neutrons would become dominant for both containers.

Температурный режим в бассейне выдержки РБМК при нарушении режима его охлаждения

Температурный режим в бассейне выдержки РБМК при нарушении режима его охлаждения

The simulate of the stratification of an explosive mixture of hydrogen and air in the RBMK-1000 spent-fuel storage pool

Currently, in Russia, codes with concentrated parameters are used as software tools to justify the hydrogen explosion safety of nuclear power units, which use the concept of “average concentration of hydrogen” in the calculated volume and do not take into account the stratification of hydrogen, due to which local concentrations of hydrogen can significantly exceed the “average”. This paper presents the hydrogen explosion safety analysis results during a beyond-design accident in the RBMK-1000 spent-fuel storage pool taking into account local concentrations of gas components using the gas dynamic code ANSYS FLUENT. The vapor-gas mixture propagation problem in the spent-fuel storage pool was solved in a two-dimensional approximation. The following sources of hydrogen were considered radiolysis hydrogen generation, hydrogen generation by reducing its solubility in water and the vapor-zirconium reaction. As a result of modeling, the possible accumulation areas of an explosive mixture and their sizes were determined. It is shown that under certain combinations of boundary conditions, due to the presence of convective flows, an explosive concentration is not achieved even if the mixture components stratification is taken into account.

Minimizing the Standard Deviation of the Thermal Load in the Spent Nuclear Fuel Cask Loading Problem

The paper assumes that, at the end of the operational period of a Spanish nuclear power plant, an Independent Spent Fuel Storage Installation will be used for long-term storage. Spent fuel assemblies are selected and transferred to casks for dry storage, with a series of imposed restrictions (e.g., limiting the thermal load). In this context, we present a variant of the problem of spent nuclear fuel cask loading in one stage (i.e., the fuel is completely transferred from the spent fuel pool to the casks at once), offering a multi-start metaheuristic of three phases. (1) A mixed integer linear programming (MILP-1) model is used to minimize the cost of the casks required. (2) A deterministic algorithm (A1) assigns the spent fuel assemblies to a specific region of a specific cask based on an MILP-1 solution. (3) Starting from the A1 solutions, a local search algorithm (A2) minimizes the standard deviation of the thermal load among casks. Instances with 1200 fuel assemblies (and six intervals for the decay heat) are optimally solved by MILP-1 plus A1 in less than one second. Additionally, A2 gets a Pearson’s coefficient of variation lower than 0.75% in less than 260s CPU (1000 iterations).

Prediction of dose rates around the interim spent fuel storage facility

This paper presents the results of radiation shielding analyses performed for siting of the Interim Spent Fuel Storage Facility (ISFSF) with a view of determining what the radiation levels will be around the ISFSF when it is made up of a certain number of casks of a specific design (Radulescu et al., 2018) (Radiation Shielding Analy, 2004). It will also determine the equivalent neutron energy where each cell damage scenarios occurs as a result ionisation.

Calculation of decay heat rate in VVER-1000 spent fuel pool including uncertainties

This paper presents a decay heat rate analysis of the VVER-1000 (unit 6 in “Kozloduy NPP” Plc.) spent fuel pool (SFP) using the method outlined by the United States (US) Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54 (Revision 2, 2018), titled “Spent Fuel Heat Generation in an Independent Spent Fuel Storage Installation.” The usage of RG 3.54 for SFP calculations is extremely attractive because it enables fast determination of the decay status of a large number of stored fuel assemblies (FAs) in an SFP and further evaluation of their uncertainties.Two batches of fuel FAs in an SFP were analyzed in the study: The first batch consisted of 402 FAs that were stored for 321 days of reactor power operation from 2014 to 2015. The second batch consisted of 355 FAs that were stored for 326 days of reactor power operation from 2018 to 2019. The total decay heat rate and the corresponding decay heat rates of the FA that was placed in each of three VVER-1000 SFP sections were evaluated. Additionally, the propagation of the input uncertainties introduced by the initial fuel mass, burnup, and operation time on full power effective days was estimated.The applied method was derived from the uncertainty method, which was previously introduced by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), for free from deliberate pessimism safety analyses (named best estimate). The variation in the selected input parameters produced decay heat generation uncertainty as follows: ±73.2 kW (2014)/±59.1 kW (2018) at the beginning of each batch storage, and correspondingly, ±12.0 kW (2015)/±10.7 kW (2019) at the end of each batch storage. The results indicate that the method is sufficiently sensitive to the initial mass variation and irradiation uncertainties.The degree of conservatism (via a parameter referred to as a ‘safety factor’), introduced in RG 3.54 to address the numerical and nuclear data uncertainties in pressurized water reactor (PWR) decay heat calculations, was calculated for the stored VVER-1000 FA. An original approach for calculation of the scaling coefficient kcorr is suggested in the paper for estimation of the level of conservatism using the safety factor in RG 3.54 and other sources. The obtained results are reported.The recommendations by RG 3.54 were implemented in an input script for the MAXIMA Computer Algebra System (CAS). The MAXIMA CAS calculations of the VVER-1000 SFP decay heat rate were verified with the corresponding Oak Ridge Isotope Generation (ORIGEN) (SCALE 6.1) data at the beginning and end of the analyzed time periods. The authors determined that the obtained values of kcorr = 0.91–0.93 are useful in the re-evaluation of the RG 3.54 degree of conservatism. Furthermore, the scaled safety factor decay heat generation output was replaced with the calculated upper uncertainty bound. The upper boundary values in the decay heat rate uncertainty range showed flexibility and reduced conservatism in comparison with the upper 10% of the conservative bound of the nominal values.

Ball-milling properties and sintering behavior of Al-based Gd2O3–W shielding materials used in spent-fuel storage

To overcome the disadvantages of commercial neutron absorbers used for spent-fuel storage, novel Al-based Gd2O3–W composites were designed by calculating the neutron absorption and γ-ray transmission and synthesized via ball milling, cold isostatic pressing, and sintering. During the ball milling, the cubic Gd2O3 was partially transformed into monoclinic Gd2O3, and new Al3Gd and Gd3Al5O12 phases were formed. The sintering induced chemical reactions that formed Gd3Al5O12, Al3Gd, and Al12W phases. Nanoparticles with sizes of 100–300 nm were uniformly distributed in the sintered bulk materials. Moreover, with increasing sintering time, sintering temperature, and W content, the microhardness of the Al-based Gd2O3–W metal matrix composites (MMCs) increased. The compressive strength of the Al-based Gd2O3–W MMCs increased with W content, in contrast to the thermal expansion coefficient. Overall, the properties of the synthesized composites were better than those of the reference Al–B4C.

On residual stress analysis and microstructural evolution for stainless steel type 304 spent nuclear fuel canisters weld joint: Numerical and experimental studies

Chloride-induced stress corrosion cracking (CISCC) in the weldments of austenitic stainless steel canisters is one of the primary safety concerns during the dry storage of used nuclear fuel at Independent Spent Fuel Storage Installations in coastal areas. In order to evaluate the CISCC behavior in the canister, a 3D sequentially coupled thermo-mechanical finite element model was built to simulate the residual stresses induced by two intersecting longitudinal and circumferential multi-pass welds in austenitic stainless steel mockup canister. Weld-induced residual stresses from simulation were in good agreement with the experimental measurement results by deep-hole drilling and contour methods. Through-wall high tensile axial and hoop stresses were observed in both the fusion zone and heat affected zone (HAZ) in the longitudinal and circumferential weld, respectively. On the other hand, the microstructure within the weldments of 304 stainless steel mockup was characterized using optical microscopy, scanning electron microscopy, electron back-scattered diffraction and energy dispersive X-ray spectroscopy methods. A few amount of MnS inclusions in both the base metal and HAZ, and a lot of (Mn,Si)xOy inclusions in the fusion zone were found. Meanwhile, there is some α′-martensite in the base metal, while the martensite was reversed in the HAZ. In the fusion zone, there is a lot of δ-ferrite. The martensite reversion phenomenon was also verified by the combination of Feritscope test and thermal simulation, which could predict a martensite reversion temperature as 400C°. The welding residual stresses and microstructure analysis results could offer the instructive information for the prediction of CISCC behavior of the spent nuclear fuel canisters.

Деякі проблемні питання в ядерній енергетичній галузі України

Institute for Safety Problems of Nuclear Power Plants (NPP) of the National Academy of Sciences of Ukraine at a meeting of the National Commission for Radiation Protection of Ukraine on June 1, 2020. It is described the crisis in the nuclear industry due to approval of the new energy balance for 2020 by the Ministry of Energy and Environmental Protection of Ukraine. According to this balance, the volume of electricity production by NPP decreased and at the same time the share of thermal power plants and “green energy”, which produce much more expensive electricity, increased in the energy market. The situation is analyzed and recommendations are given for the implementation of measures to improve it. It is shown that the main task of the Ukrainian energy sector will be to continue the course of market relations and to abandon excessive market regulation. Market relations will lead to the fact that generating companies with cheaper energy will be able to sell it not at the cost level price, but at tariffs that have formed in the market. The price of nuclear energy will be brought up to the level of the price of energy produced by thermal power plants. Consumption costs will increase, but government generation will receive more funds, and the state will decide where to direct them: to subsidies for the population or to the development of nuclear or renewable energy. It is proposed to make the necessary changes to the legislation on “green energy”; establish a reasonable tariff for energy produced at NPPs; to introduce technical and organizational measures to ensure the operation of NPP power units in shunting modes. It is concluded that the optimal solution for the future from the standpoint of the both environmental and economic policy of Ukraine is the gradual replacement of NPP power units, which have “exhausted” project resource, with modern new ones with a higher level of safety. This will support the share of NPPs in the energy balance of Ukraine at 40−50%, as well as contribute to the fulfillment of its commitments to increase the use of low-carbon technologies in energy. As for solving energy problems of Ukraine on the basis of promising technologies — small modular reactors (SMR), this strategy does not yet have a proper feasibility study and is not confirmed by the necessary proven practice, as in the world such reactors are still being developed. Observations on the construction of the Central Spent Fuel Storage Facility (CSFSF) for long-term storage of spent fuel from the power units of Rivne, Khmelnytskyi and South-Ukraine NPPs were made. The construction of the CSFSF is under threat because already at the final stage of construction it was decided to liquidate the division of NNEGC “Energoatom”, which was engaged in the construction. Such an incomprehensible decision will not lead to anything positive, but will only delay the commissioning period of the storage facility by 1−2 years. It is shown that the construction of a New Safe Confinement (NSC) over the Shelter object of the Chornobyl NPP is just an intermediate stage in the transformation of the Shelter object into an ecologically safe system. In the near future, it is necessary to perform dismantling of building structures of the Shelter object, the term of operation of which ends in 2023. This is prompted by cases of destruction and collapse of building structures of this object, which occurred last year, and therefore this issue requires immediate solution. The critical state of nuclear science in Ukraine is reflected and it is warned that if the financial support does not change properly, in the coming years the country will be left without a unique scientific institution that conducts research on the safety of nuclear facilities, including the most dangerous in the world Shelter object. Emphasis is placed on the complete absence of a safety culture at the highest level of the country’s leadership, the lack of which is manifested in violations of national norms and international standards, the absence of some leaders in nuclear industry and dangerous enterprises, as well as the attitude of government agencies to the nuclear industry.