scholarly journals DEVELOPMENT OF A MODELING APPROACH TO ESTIMATE RADIATION FROM A SPENT FUEL ROD QUIVER

2021 ◽  
Vol 247 ◽  
pp. 16006
Author(s):  
Zs. Elter ◽  
V. Mishra ◽  
S. Grape ◽  
E. Branger ◽  
P. Jansson ◽  
...  
Keyword(s):  
Gamma Radiation ◽  
Spent Nuclear Fuel ◽  
Radiation Transport ◽  
Neutron Radiation ◽  
Nuclear Safeguards ◽  
Monte Carlo Code ◽  
Spent Fuel ◽  
Fuel Rods ◽  
History Of ◽  
New Type

Before encapsulation of spent nuclear fuel in a geological repository, the fuels need to be verified for safeguards purposes. This requirement applies to all spent fuel assemblies, including those with properties or designs that are especially challenging to verify. One such example are quivers, a new type of containers used to hold damaged spent fuel rods. After placing damaged rods inside the quivers, they are sealed with a thick lid and the water is removed. The lid is thick enough to significantly reduce the amount of the gamma radiation penetrating through it, which can make safeguards verification from the top using gamma techniques difficult. Considering that the number of quivers at storage facilities is foreseen to increase in near future, studying the feasibility of verification is timely. In this paper we make a feasibility study related to safeguards verification of quivers, aimed at investigating the gamma and neutron radiation field around a quiver designed by Westinghouse AB and filled with PWR fuel rods irradiated at the Swedish Ringhals site. A simplified geometry of the quiver and the detailed operational history of each rod are provided by Westinghouse and the reactor operator, respectively. The nuclide inventory of the rods placed in the quiver and the emission source terms are calculated with ORIGEN-ARP. The radiation transport is modeled with the Serpent2 Monte Carlo code. The first objective is to assess the capability of the spent fuel attribute tester (SFAT) to verify the content for nuclear safeguards purposes. The results show that the thick quiver lid attenuates the gamma radiation, thereby making gamma radiation based verification from above the quiver difficult.

scholarly journals IDENTIFICATION OF THE FUEL ROD CLADDING DESTRUCTION FROM THE CHANGE OF THE SNF STORAGE CASKS RADIATION

2020 ◽  
pp. 111-119
Author(s):  
V.G. Rudychev ◽  
N.A. Azarenkov ◽  
I.O. Girka ◽  
Y.V. Rudychev
Keyword(s):  
Numerical Modeling ◽  
Gamma Radiation ◽  
Dose Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Side Surface ◽  
Neutron Radiation ◽  
External Radiation ◽  
Fuel Rods ◽  
Fuel Rod

Two options for changing the distribution of spent nuclear fuel due to the possible destruction of the cladding of fuel rods, which causes a change in radiation outside the cask, are considered for VSC-24 casks used for storage of spent nuclear fuel by the dry method. The effect of height reduction due to the destruction of the fuel rods of all 24 SFAs and 10 central SFAs on external radiation is studied analytically and by numerical modeling in the MCNP package. The destruction of 24 SFA is shown to lead to a significant decrease in the dose rate of neutrons and gamma-radiation from 60Co on the weather lid of the cask, and of gamma-radiation from SNF isotopes at the mid-height of the side surface of the cask. The destruction of the ten central SFAs can be determined only from a change in the neutron radiation in the air inlets of the cask.

As mentioned in the previous chapter, many experiments on food irradiation in the 1950s were carried out with spent-fuel rods from nuclear reactors. Such fuel rods contain a mixture of many fission products, with greatly differing half-lives, emitting different types of radiation with different energies. The composition of fuel rods changes all the time because the radionuclides with short half-lives disappear quickly, whereas those with longer half-lives remain. Although fuel rods are primarily a source of gamma radiation (the less penetrating alpha and beta radiation are absorbed by the steel hull of the rods) they do give off some neutrons. Since the latter can produce radioactivity when they interact with matter such as food, fuel rods have not been used for irraditation of foods since the early 1960s. Because of their constantly varying composition, fuel rods also make dosimetry difficult, and this was another reason for abandoning their use. Individual constituents of spent fuel rods can be separated in reprocessing plants by chemical methods. One of the radionuclides obtainable in this way is Cs. With a half-life of 30 years and emission of gamma radiation (0.66 MeV) and beta radiation (0.51 MeV and 1.18 MeV), '^C s decays to stable '^B a (barium). After the ,37Cs is separated from the other constituents of the fission waste in the form of CsCl it is triply encapsulated in stainless steel containers because CsCl is soluble in water. If it leaked out it could cause contamination of the environment. As provided by the Waste Encapsulation and Storage Facility (WESF) at Hanford, Washington, the 137Cs capsule is 400 mm in active length (500 mm in total length) and 67 mm in diameter. There are only a few reprocessing plants in the world and the capacity for extracting ,37Cs from spent fuel rods is very limited. Plans for building several commercial reprocessing facilities in the United States were canceled by Presi­ dent Carter’s 1977 decision that the United States would not engage in commer­ cial reprocessing of spent nuclear fuel. As a consequence, not much ,37Cs is available and there are not many gamma radiation facilities which use ,Cs. No

1995 ◽  
pp. 31-31
Keyword(s):  
United States ◽  
Gamma Radiation ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
The United States ◽  
Food Irradiation ◽  
Beta Radiation ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Active Length

scholarly journals Quantitative imaging and automated fuel pin identification for passive gamma emission tomography

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Ming Fang ◽  
Yoann Altmann ◽  
Daniele Della Latta ◽  
Massimiliano Salvatori ◽  
Angela Di Fulvio
Keyword(s):  
Inverse Problem ◽  
Image Reconstruction ◽  
Image Resolution ◽  
Emission Tomography ◽  
Nuclear Safeguards ◽  
Activity Levels ◽  
Spent Fuel ◽  
Gamma Emission ◽  
Cross Sectional ◽  
Fuel Rods

AbstractCompliance of member States to the Treaty on the Non-Proliferation of Nuclear Weapons is monitored through nuclear safeguards. The Passive Gamma Emission Tomography (PGET) system is a novel instrument developed within the framework of the International Atomic Energy Agency (IAEA) project JNT 1510, which included the European Commission, Finland, Hungary and Sweden. The PGET is used for the verification of spent nuclear fuel stored in water pools. Advanced image reconstruction techniques are crucial for obtaining high-quality cross-sectional images of the spent-fuel bundle to allow inspectors of the IAEA to monitor nuclear material and promptly identify its diversion. In this work, we have developed a software suite to accurately reconstruct the spent-fuel cross sectional image, automatically identify present fuel rods, and estimate their activity. Unique image reconstruction challenges are posed by the measurement of spent fuel, due to its high activity and the self-attenuation. While the former is mitigated by detector physical collimation, we implemented a linear forward model to model the detector responses to the fuel rods inside the PGET, to account for the latter. The image reconstruction is performed by solving a regularized linear inverse problem using the fast-iterative shrinkage-thresholding algorithm. We have also implemented the traditional filtered back projection (FBP) method based on the inverse Radon transform for comparison and applied both methods to reconstruct images of simulated mockup fuel assemblies. Higher image resolution and fewer reconstruction artifacts were obtained with the inverse-problem approach, with the mean-square-error reduced by 50%, and the structural-similarity improved by 200%. We then used a convolutional neural network (CNN) to automatically identify the bundle type and extract the pin locations from the images; the estimated activity levels finally being compared with the ground truth. The proposed computational methods accurately estimated the activity levels of the present pins, with an associated uncertainty of approximately 5%.

scholarly journals Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder

2020 ◽  
Vol 2020 ◽  
pp. 1-12
Author(s):  
Young-Hwan Kim ◽  
Yung-Zun Cho ◽  
Jin-Mok Hur
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Energy Research ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Compressive Stresses ◽  
Simulated Fuel ◽  
Ceramic Fuel ◽  
Experimental Approaches

We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

Study on Periodic Shielding Performance Measurements of Spent Fuel Transportation Package

2017 ◽  
Author(s):  
Hongchao Sun ◽  
Guoqiang Li ◽  
Xuexin Wang ◽  
Dajie Zhuang ◽  
Renze Wang ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Measurement Instrument ◽  
Neutron Radiation ◽  
Radiation Shielding ◽  
Radioactive Material ◽  
Transportation Safety ◽  
Spent Fuel ◽  
Performance Measurements ◽  
Radiation Level ◽  
Neutron Measurement

The radioactive activity of spent nuclear fuel is high, and the transportation safety is concerned by public and specialist. The periodic radiation shielding performance measurements of spent fuels package is important content to ensure transportation safety of spent fuels. The radiation shielding performance of package must meet the requirements of “Regulations for the safe transport of radioactive material” (GB11806-2004). However, some of the problems and difficulties reflected in practice need to be solved, such as the measurements results of neutron radiation level of spent fuels package outer are not always reliable. In this paper, the periodic shielding performance measurements of one type of spent fuel transportation package are presented. The monitoring results of using both the neutron multi-sphere spectrometer and portable neutron measurement instrument are compared, and the Monte Carlo simulation is done to verify the measurements results. Some factors are discussed, and an optimized scheme is recommended.

Nuclear Waste

2014 ◽  
Author(s):  
Michael H. Fox
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Spent Nuclear Fuel ◽  
Nuclear Fission ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Fuel Pellets ◽  
Water Filling ◽  
Nuclear Fuel Waste ◽  
Ranch House

I gazed over the railing into the crystal clear cooling pool glowing with blue Cherenkov light caused by particulate radiation traveling faster than the speed of light in water. I can see a matrix of square objects through the water, filling more than half of the pool. It looks like you could take a quick dip into the water, like an indoor swimming pool, but that would not be a good idea! It is amazing to think that this pool, about the size of a ranch house, is holding all of the spent fuel from powering the Wolf Creek nuclear reactor in Burlington, Kansas, for 27 years. The reactor was just refueled about a month before my visit, so 80 of the used fuel rod assemblies were removed from the reactor and replaced with new ones. The used fuel rods were moved underwater into the cooling pool, joining the approximately 1,500 already there. There is sufficient space for the next 15 years of reactor operation. There is no danger from standing at the edge of this pool looking in, though the levels of radon tend to be somewhat elevated and may electrostatically attach to my hard hat, as indeed some did. What I am gazing at is what has stirred much of the controversy over nuclear power and is what must ultimately be dealt with if nuclear power is to grow in the future—the spent nuclear fuel waste associated with nuclear power. What is the hidden danger that I am staring at? Am I looking at the unleashed power of Hephaestus, the mythical Greek god of fi re and metallurgy? Or is this a more benign product of energy production that can be managed safely? What exactly is in this waste? And is it really waste, or is it a resource? To answer that question, we have to understand the fuel that reactors burn. The fuel rods that provide the heat from nuclear fission in a nuclear reactor contain fuel pellets of uranium, an element that has an atomic number of 92 (the number of protons and also the number of electrons).

scholarly journals Problems of radiation safety calculations related to spent nuclear fuel transport casks

2020 ◽  
Vol 6 (1) ◽  
pp. 43-47
Author(s):  
Artem V. Sobolev ◽  
Pavel A. Danilov
Keyword(s):  
Gamma Radiation ◽  
Effective Dose ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Radiation Safety ◽  
Spent Fuel ◽  
Preliminary Assessment ◽  
Radiation Sources ◽  
Fuel Assemblies ◽  
Long Time

The paper discusses the stages of calculating the radiation safety of spent nuclear fuel (SNF) transport packages, in particular, transport casks and some related problems. The problem of describing the source of neutrons and gamma radiation of spent nuclear fuel is shown. For individual designs of fuel assemblies, data are given on isotopes that make the main contribution to the neutron source as well as on gamma rays in nuclear fuel material and structural materials. The authors emphasize the necessity of analyzing the influence of the initial spent fuel parameters on the formation of the radiation spectrum and, therefore, on the radiation situation around the transport casks. Consideration is given to the problem of assessing the attenuation of gamma radiation in calculating protection analytically and using software. Due to the ambiguity of the position of the zone with the highest effective dose value on the SNF transport cask surface, it is indicated that preliminary estimates are required to take into account all radiation sources and their nonuniformities. All the problems presented in the paper are currently being solved by means of rather complex and voluminous calculations that take a long time. In order to be able to conduct a preliminary assessment of the radiation situation around the transport casks, the authors propose to create a methodology that will determine the type of interrelations between the maximum effective dose and input parameters, such as fuel burnup, decay, fuel composition, protection material in the SNF transport cask, etc. This methodology will make it possible to improve the efficiency of the process of designing the SNF transport casks, avoid possible design errors and, in particular, when used as intended, resolve the issue of the SNF cask loading configuration.

scholarly journals The Three-­Dimensional Neutron-­Physical Model of Spent Nuclear Fuel Storage System

2021 ◽  
Vol 20 ◽  
pp. 51-59
Author(s):  
О. R. Trofymenko ◽  
◽  
І. M. Romanenko ◽  
М. І. Holiuk ◽  
C. V. Hrytsiuk ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Storage Facility ◽  
Spent Fuel ◽  
Wide Range ◽  
Fuel Assemblies ◽  
Fuel Storage

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.

Estimation of the long term helium production in high burn-up spent fuel due to alpha decay and consequences for the canister

2014 ◽  
Vol 1665 ◽  
pp. 297-302
Author(s):  
Alba Valls ◽  
Mireia Grivé ◽  
Olga Riba ◽  
Maita Morales ◽  
Kastriot Spahiu
Keyword(s):  
Spent Nuclear Fuel ◽  
Swelling Pressure ◽  
Alpha Decay ◽  
Nuclear Fuels ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Α Decay ◽  
Alpha Emitters ◽  
Maximal Pressure

ABSTRACTIn the KBS-3 repository concept and safety analysis, the copper container with a cast iron insert plays a central role in assuring isolation of the waste from the surrounding during long periods of time. All processes that affect its stability are thoroughly analysed, including potential detrimental processes inside the canister. For this reason, an estimation of the helium produced during the long term decay of alpha emitters in the spent fuel is necessary to evaluate if the pressures generated inside can have consequences for the canister.The spent nuclear fuel to be disposed of in Sweden is mainly LWR fuel. The maximum burn-up expected is 60 MWd/kg U for BWR and PWR. A small quantity of BWR MOX is expected to be stored with a maximum burn-up of 50 MWd/kg U.This work has focused on carrying out calculations of the amounts of He generated during more than 1 million years in Swedish spent nuclear fuels with a benchmarking exercise by using both codes AMBER and Origen-ARP. The performance and agreement of the codes in the He generation from α-decay have been checked and validated against data reported in literature [1].In the calculation of the maximal pressure inside the canister, the quantity of helium used to pre-pressurise the fuel rods has been accounted for. The pressure inside the canister due to He generation is at all times much lower than the hydrostatic pressure and/or the bentonite swelling pressure outside the canister.

scholarly journals Operational and Technical Results from the Spent Fuel Test-Climax

1986 ◽  
Vol 29 (1) ◽  
pp. 51-54
Author(s):  
Wesley Patrick
Keyword(s):  
Spent Nuclear Fuel ◽  
Radiation Transport ◽  
Storage Period ◽  
Test Site ◽  
Department Of Energy ◽  
Mechanical And Thermal Properties ◽  
Spent Fuel ◽  
Technical Data ◽  
Geologic Storage ◽  
Fuel Assemblies

The technical feasibility of short-term storage and retrieval of spent nuclear fuel assemblies has recently been demonstrated in a test of deep geologic storage at the U.S. Department of Energy Nevada Test Site (NTS). Handling systems and procedures developed and deployed on this test functioned safely and reliably to emplace eleven intact spent-fuel assemblies and retrieve them three years later. Three exchanges of spent fuel were conducted at regular intervals during the storage period to maintain the proficiency of personnel and the readiness of the handling system. Technical data was collected using nearly 1,000 instruments. These data show that the mechanical and thermal properties of granites are compatible with nuclear waste isolation objectives. Measured and calculated temperatures are in excellent agreement, confirming the adequacy of available heat transfer codes. Radiation transport calculations were of high quality, exceeding the accuracy of available long-term dosimetry techniques which were used on the test. We also found good agreement between measured and calculated displacements within the rock mass.

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