RT-1 Plant Research Articles

Adaptation of RT-1 Plant Technology to Processing of New Types of Spent Nuclear Fuel

Adaptation of RT-1 Plant Technology to Processing of New Types of Spent Nuclear Fuel

Analysis of the Possibility of Using Carbamide for Separation of U and Pu in Extractive Processing of SNF

Behavior of carbamide in HNO3 solutions on Pt and SnO2 electrodes was studied. It was found that the oxidation rate of Co(NH2)2 on the SnO2 electrode is approximately two orders of magnitude lower than that on the Pt electrode. The electrochemical reduction of U(VI) in a cell having no diaphragm on a Ti (cathode)–SbO2 (anode) pair of electrodes was examined at its concentration of 10 to 100 g/L in HNO3 solutions (0.5–2.0 M) containing carbamide (5–30 g/L). It was found that the concentration of these components affects the completeness and rate of U(IV) formation and the current efficiency. It was shown that the reduction efficiency decreases in the presence of technetium ions, and, after the current is switched off, the already formed U(IV) is oxidized to U(VI). The rate of this process grows with increasing concentration of Te and HNO3. The electrochemical behavior of carbamide on an insulated Ti cathode was examined in aqueous nitric acid solutions. It was found that, in this case, the carbamide solutions acquire on being subjected to electrolysis reducing properties toward Pu(IV) and Np(VI). A laboratory installation of mixers-settlers, assembled in accordance with the technological scheme of the first cycle of SNF processing at RT-1 plant was used to perform experiments with the use of electrochemically processed carbamide as a re-extractive agent in the procedure of Pu and Np separation from uranium. It was found that a high rate of U purification to remove Np is reached in this case, but a satisfactory mutual separation of U and Pu is not provided.

Development and Pilot Testing of New Flow Diagrams for Plutonium Purification at the RT-1 Plant

To modernize the extraction facility of the RT-1 plant, two flow diagrams for plutonium purification have been developed. In the first flow diagram, the Pu(IV)–Np(IV) valence pair is stabilized and both Pu and Np are stripped and, in the second flow diagram, the Pu(IV)–Np(V) valence pair is stabilized and only Pu is stripped. The problem of stabilizing the pair of the desired components in the required valence state is solved using reagents that had already been used at the RT-1 plant: hydrogen peroxide, hydrazine nitrate, and a catalyst (Fe). Both flow diagrams are adapted to the extraction facility of the plant with minimum changes in equipment and successfully pass extended process tests. As the result of this work, the volume and salt content of the raffinate are reduced.

Determination of the Most Likely Wear Zones of Pipe Bends in the Pulsed Pneumatic Conveyor at the RT-1 Plant

The results of an experimental investigation and numerical modeling of the most likely high-wear zones in 90° pipeline bends in the pulsed pneumatic conveyor system at the RT-1 plant of the Mayak Industrial Association are presented. The experimental studies were conducted on a model of a pulsed pneumatic conveyor system; the numerical modeling of the gas-dynamic flow in the pipeline channel was performed in a model of a two-dimensional non-stationary motion of a compressible medium. It was found on the basis of the experimental results obtained using the model and numerical modeling in the homogeneous approximation that local regions with high wear of walls in the 90° pipe bends in the pulsed pneumatic conveyor do indeed exist and their experimentally found positions agree with the calculations.

The influence of deviations in process parameters on the purification of uranium from different radionuclides

In the present study, the effect of deviations in the process has been determined during the process of the extraction processing of spent nuclear fuel (SNF) VVER-440 on RT-1 plant on the purification of uranium from 99Tc, 125Sb, 106Ru, 137Cs, 228Th, etc. The uranium decontamination factors during purification from these nuclides for both cycles of the Purex scheme of plant are calculated. It has been shown that, upon deviations in the process (the short-term shutdown of process, fluctuations of hydrodynamic mode, deposition of elements in the extractors, etc.), an increase in the uranium decontamination factors during purification from most unwanted nuclides is observed, except for technetium. However, it has been found that, at an increased inflow of Tc to the second extraction cycle, the purification of the final uranium product from this element may be insufficient.

Gamma-Spectrometric Determination of the Plutonium Mass in Containers

A method is proposed for determining the mass and isotopic composition of the plutonium warehoused in containers at the RT-1 plant of the Industrial Association Mayak. A single measurement system with a detector based on ultrapure germanium is used for the measurements. The mass is determined according to the pulse count rate in 239Pu peaks. The correction for self-absorption of radiation in a container is determined by measuring the transmission of radiation emitted from an external source and by Monte Carlo calculations. Estimates show that the total error in determining the plutonium mass by the new method is 11% (P = 0.95). The method can be used to perform confirmatory measurements in the accounting and control system for nuclear materials.

Used Mixed Oxide Fuel Reprocessing at RT-1 Plant

Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX UNF in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing.

The Development and Testing of the New Flowsheets for the Plutonium Purification Cycle of the Purex Process

In order to improve the extraction flowsheet of RT-1 Plant two versions of plutonium purification unit flowsheet were developed: a flowsheet with stabilization of Pu(IV)-Np(IV) valence pair and Pu, Np co-recovery, and a flowsheet with stabilization of Pu(IV)-Np(V) valence pair and Pu recovery. The task related to stabilization of the valence pair of the target components in the required state was solved with the use of reactants already applied at RT-1 Plant, namely, hydrogen peroxide, hydrazine nitrate and catalyst (Fe). Both flowsheets were adapted for the plant purification facility with minimum modifications of the equipment, and passed the full scale industrial testing. As a result of this work, reduction in volume and salt content of the raffinate was achieved.

Preliminary studies of dynamic regimes in the extraction-scrubbing groups of units of the experimental and demonstration center (EDC) using mathematical modeling

The plutonium accumulation in the extraction-scrubbing cascade in the course of reprocessing of spent nuclear fuel from water-cooled water-moderated energy reactors (WWER) via the combined flowsheet of the RT-1 plant of Mayak production association and via the flowsheet of the Experimental and Demonstration Center (EDC) of the Mining and Chemical Combine, with transformation of the scrubbing section into a zirconium stripping unit, was compared using a dynamic model of extraction processes. The limiting level by which the productive capacity can be increased without emergencies is ∼15% for RT-1 and 22% for EDC. In the latter case, the accumulation with the potential discharge of a part of Pu into the Zr strip occurs in the third stage of the Zr stripping unit, which is confirmed by rig trials and can be detected in practice by neutron control method because of the absence of Am and Cm. Complete recovery of Zr without its accumulation in the extraction section can be ensured by its coextraction with the Cr(VI) anion fed as a separate small stream. Stopping this stream leads to a reversible decrease in the Zr recovery and to a sharp increase in the entrainment of aqueous microemulsion containing 137Cs. This fact confirms the assumption that the Zr extraction in the cascade in the presence of U and at increased HNO3 concentration has micellar character [without Cr(VI)].

Comparison of correlation relations for determining the relative 242Pu content in plutonium produced in the RT-1 plant at the Industrial Association Mayak

The correlation relations for determining the 242Pu mass fraction which are used in the MGA and FRAM computer codes are compared with the relations proposed in the present article for plutonium stored in the RT-1 plant of the Industrial Association Mayak. The need to determine the 242Pu mass fraction using the correlations arises in nondestructive analysis of plutonium. The MGA and FRAM codes use correlations which are based on data on the isotopic composition of the plutonium that has accumulated in American reactors so that the possibility of using these codes for Russian plutonium must be checked. This check was based on an analysis of data obtained on more than 6000 isotopic compositions during certification of the RT-1 plutonium. A characteristic of the relation proposed in this article is that it takes account of the change in the isotopic composition of plutonium over time.

Gamma spectrometric determination of the isotopic decomposition of plutonium in containers using MGA and FRAM software

Plutonium dioxide is stored in special containers, holding about 3 kg plutonium each, in a dump at the RT-1 plant of the Industrial Association Mayak. The certificates on the containers with plutonium produced in 1990–2000 do not contain any data on the isotopic composition. Without removing the plutonium from a container, the isotopic composition can be determined only by γ-spectrometry. This paper compares the MGA and FRAM computer codes used to determine the isotopic composition of the plutonium when performing measurements on samples with a large mass.

Nitric acid purification from admixtures of volatile acids by fractionation during waste evaporation at a spent nuclear fuel reprocessing company

We present the results of the computer simulation of a method for better purification of nitric acid from HF and CH3COOH admixtures by creating a stripping section in the fractionation column with HNO3 scrubbing and by the feeding of a parasitic water flow to the bottom section. The production column in the evaporator section of the RT-1 plant at the PA “Mayak” was upgraded using the calculation results and test trials on real process high-level waste were performed, which made it possible to improve the purification of regenerated 12 mol/L HNO3 from HF and simultaneously from HCl by a factor of 3–5. The specification figures for three flowsheet of nitric acid regeneration that were used at different times since 1979 for commercial purification from halogens and radioactive contaminants were compared.

Possibility of Various Types SNF Reprocessing at the PA Mayak exampled with AMB SNF

For the purpose of reprocessing of irradiated nuclear fuel from the water-cooled graphite-moderated pressure-tube reactor named AMB from decomissioned Russian “Atom Peaceful Big”, modernization of the process flow-sheet of the RT-1 plant is being carried out at PA Mayak with participation of FSUE KRI and VNIINM. A particular AMB SNF feature is extremely broad range of fuel compounds with the main ones being the uranium-molybdenum metal, uranium oxide and uranium carbide compositions usually dispersed in magnesium or calcium. Wide range of fuel compositions required to amend SNF dissolution, extraction processing, evaporation of high-level radioactive wastes and vitrification of high-level radioactive wastes. The above set of laboratory research was completed with dynamic tests using samples of AMB from the water-cooled graphite-moderated pressure-tube reactor. Tests have shown the possibility of processing the entire range of AMB SNF at the radiochemical plant RT-1 plant of the PA Mayak. Thus, the ability of the RT-1 plant to process different fuel compositions, including the long-term research reactor fuel have been proved experimentally.

Handling spent nuclear fuel at a temporary storage site in Guba Andreev

Spent nuclear fuel has been stored in dry-storage units at a shore base of the naval fleet for 35–45 year. The total activity of the spent nuclear fuel is 170 PBq. This article presents data which characterize the state of the fuel (from normal to defective), the radiation conditions, and information on the individual and collective irradiation dose to workers. The results of an inventory check of the cells and jackets which contain fuel assemblies are presented. The corrosion processes are described and ideas for handling the spent fuel at the RT-1 plant of the Mayak Industrial Association, including handling fuel assemblies and jackets in cases, are described.

Technetium Distribution throughout Purex Process Cycles at RT-1 Plant

Data are presented on the Tc content in the major products of the RT-1 plant and also on its distribution throughout the process flowsheet for reprocessing spent fuel from VVER-440 (440-MW watercooled water-moderated power reactor). At RT-1 plant the process flowsheet allows more than 99% recovery of Tc in the joint extraction cycle and then about 98.5% removal of Tc to raffinate of the refining cycle. The decontamination factor of U from Tc in the first extraction cycle is about 500, and in the second (refining) U extraction cycle, 5-10. The decontamination factors of Pu and Np from Tc are 104 and 106, respectively. In reprocessing BN-350 (350-MW fast neutron reactor) fuel on an extraction line with a low TBP concentration, about 2/3 of technetium goes to high-level raffinate and 1/3 to Pu strip.