Seismic isolation of pool-type tanks for the storage of nuclear spent fuel assemblies

Seismic isolation of pool-type nuclear spent fuel storage tanks requires careful investigation of dynamic behavior of the fluid–structure–isolator interaction system to satisfy the requirements of safety functions and the prevention of nuclear criticality. This paper presents the investigation, results and discussions on the seismic design considerations of isolated pool-type tanks for the storage of nuclear spent fuel assemblies. A three-dimensional boundary element-finite element method is presented for the analysis of the fluid–structure–isolator systems in time domain. Scaled model tests were performed to verify the numerical method and to study the dynamic behavior of isolated pool-type storage tanks. Important factors affecting the dynamic behavior of tanks with a fixed base are further investigated as is the case for isolated tanks using base isolators with different mechanical properties. The base isolators are the high damping rubber-bearing type and are modeled using a bilinear analysis model. Based on the numerical analysis and experimental results, some conclusions and discussions on the design considerations for isolated storage tanks are presented. In general, it is shown that careful selection of mechanical properties of the isolators with a certain lower limit on the effective frequency can guarantee the reduction of the dynamic responses of the storage tanks and the enhancement of the stability of stored spent fuel assemblies against earthquake excitations.     

Thermal criteria for terminal storage of spent nuclear fuel assemblies in bedded salt

Design considerations, particularly thermal design considerations, for terminal storage of spent nuclear fuel in a bedded salt repository are discussed in a set of four papers which address: thermal criteria, uncoupling thermal problems, near-field temperatures, and optimization of mine arrangement. This paper outlines the thermal criteria used in the conceptual design of a bedded salt repository and discusses the historical rationale behind developments of each criterion.     

Uncoupling thermal problems of terminal storage of spent nuclear fuel assemblies in bedded salt

Spent nuclear fuel assemblies stored in bedded salt can be modeled with a large array of dimensioned decay heat sources (spent fuel assemblies) in an extended thermal conducting media. Although a finite-difference or finite-element representation of the total storage facility could be established, regions of the repository should be analyzed separately since a model of the total repository would require formidable digital computing capacity. This paper explains the basis for thermally analyzing the total storage facility with separate models for the stored fuel assembly package and the salt medium. In addition, the effect of fuel assembly packaging on the maximum fuel temperature, the related problems of fuel handling prior to storage, and uncoupling of the effects of mine ventilation and conduction in the salt medium are discussed.     

Recycling of uranium from spent RBMK fuel in the nuclear fuel cycle

It is shown that there is promise in using the uranium product obtained by reprocessing spent nuclear fuel from RBMK reactors as a non-initial fuel source for thermal reactors. A technical path for spent nuclear fuel from RBMK reactors is proposed: radiochemical reprocessing and obtaining oxides of recycled uranium. Oxides of the category RBMK-poor are packed and then stored in a near-surface storage facility; oxides of the category RBMK-rich are fluoridated, and UF₆ is fed into separation production for additional enrichment to the required content of ²³⁵U. Additional advantages of recycled RBMK uranium as a source of non-initial ²³⁵U are the low content of ²³²U and the relatively low activity of spent fuel, which simplifies its reprocessing.     

Decontamination of Zircaloy cladding hulls from spent nuclear fuel

The feasibility of decontaminating spent fuel cladding hulls using hydrofluoric acid (HF) was investigated as part of the Global Energy Nuclear Partnership (GNEP) Separations Campaign. The concentrations of the fission product and transuranic (TRU) isotopes in the decontaminated hulls were compared to the limits for determining the low level waste (LLW) classification in the United States (US). The ⁹⁰Sr and ¹³⁷Cs concentrations met the disposal criteria for a Class C LLW; although, in a number of experiments the criteria for disposal as a Class B LLW were met. The TRU concentration in the hulls generally exceeded the Class C LLW limit by at least an order of magnitude. The concentration decreased sharply as the initial 30-40 μm of the cladding hull surface were removed. At depths beyond this point, the TRU activity remained relatively constant, well above the Class C limit.     

Characterization of alloy particles extracted from spent nuclear fuel

We characterized, for the first time, submicro- and nanosized fission product-alloy particles that were extracted nondestructively from spent nuclear fuel, in terms of noble metal (Mo–Ru–Tc–Rh–Pd–Te) composition, atomic level homogeneity and lattice parameters. The evidences obtained in this work contribute to an improved understanding of the redox chemistry of radionuclides in nuclear waste repository environments and, in particular, of the catalytic properties of these unique metal alloy particles.     

Dynamic analysis of storage racks for spent fuel assemblies

The dynamic response of storage racks for spent fuel assemblies subjected to base excitation is calculated. While classical methods of linear structural dynamics may be adequate at low levels of excitation, nonlinear effects due to uplift, for example, can no longer be neglected at high excitation levels. Several nonlinear dynamic analyses have been performed for different types of storage racks with uplift capability. With the help of the numerical results, the rocking behavior of storage racks and their structural integrity has been examined.     

Validation of nuclear and radiation safety of a container for spent amb reactor fuel assemblies at the Beloyarskaya nuclear power plant

An approach is proposed for validating the nuclear and radiation safety of a container for spent fuel assemblies from AMB-100 and-200 reactors at the Beloyarskaya nuclear power plant. To validate the radiation safety, the characteristics of fuel assemblies and their classification according to the average fuel burnup in the casing, and the intensities of n and γ radiation in the casing are analyzed. Nuclear safety is validated on the basis of the concept of a “model” casing. This model makes it possible to obtain an upper estimate of the effective coefficient of neutron multiplication for all real casings with fuel assemblies. Calculations are used to determine the minimum necessary thickness of the vessel, bottom, and cover for 17-and 35-place casings. It is shown that no special neutron protection is needed. The container design to be developed meets the IAEA and OPBZ-83 safety standards. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 423–428, June, 2006.     

Tokamak based hybrid systems for fuel production and recovery from spent nuclear fuel

The effect of the internal elements of a tokamak with different coolants on the transmutation rate of long-lived actinides is studied. The Monte Carlo method is used to calculate nuclear reactions in a homogeneous model of a blanket and in specific designs of a blanket. The neutron-physical calculations of a homogeneous model of a blanket showed that the absorption of neutrons by the central column of the tokamak and their moderation by the beryllium coating slow the transmutation rate to 35% of the initial value with the most efficient utilization of lead coolant. For water coolant, this effect is negligible. In a heterogeneous model of a blanket where water coolant is used, plutonium must be added to the actinides (50%/50%). The use of lead as the coolant will increase the transmutation rate of the actinides without using plutonium. In this case, it will be possible to reprocess spent nuclear fuel from more than 10 VVER-1000 in the JUST-T hybrid reactor.     

Method for identifying 17-place casings with spent fuel assemblies of AMB reactors at the Beloyarskaya nuclear power plant

A method is proposed for identifying a cylindrical case with a jacket containing 17 spent fuel assemblies from the AMB reactors at the Beloyarskaya nuclear power plant on the basis of measurements of the radiation characteristics in lateral surface of the jacket opposite the fuel assemblies in the outer and inner rows. Computational validation using the PRIZMA and MCNP computer codes is given for the method. It is shown that the collection of signals from detectors is specific to each jacket, and this makes it possible to use it as an identifying indicator. __________ Translated from Atomnaya énergiya, Vol. 102, No. 6, pp. 363–367, June, 2007.     

Modelling the dynamical behaviour of nuclear reactor fuel assemblies

The computation of the motion of fuel rods in a nuclear reactor core is generally difficult because the number of elements is very large. This paper is devoted to an homogenization process likening the fluid-tube-spacer system to an equivalent homogeneous medium for which the constitutive laws are derived. It is shown how non-linearity (non-linear spacers, shocks) can be taken into account. Some numerical computation schemes are evoked.