Heavy Water Moderator Research Articles

External Moderation of Reactor Core Neutrons for Optimized Production of Ultra-Cold Neutrons

The ultra-cold neutron (UCN) source being commissioned at North Carolina State University’s PULSTAR reactor is uniquely optimized for UCN production in the former graphite-filled thermal column outside of the reactor pool. The source utilizes a remote moderation design, which is particularly well suited to the PULSTAR reactor because of its high thermal and epithermal neutron leakage from the core face. This large non-equilibrium flux from the core is efficiently transported to the UCN source through the specially designed beam port in order to optimize UCN production at any given reactor power. The increased distance to the source from the core also greatly limits the heat load on the cryogenic system. A MCNP (Monte Carlo N-Particle) model of this system was developed and is in good agreement with gold foil activation measurements using a test configuration as well as with the real UCN source’s heavy water moderator. These results established a firm baseline for estimates of the cold neutron flux available for UCN production and prove that remote moderation in a thermal column port is a valuable option for future designs of cryogenic UCN sources.

Superhydrophilic metal-organic frameworks film modified surface for tritium removal from tritiated heavy water

Tritium extraction from heavy water moderators with catalytic exchange reactions had hydrogen explosion, fire risk, and complicated system design, which restricted the industrial scale usage. Herein, we reported a safer and simpler tritiated heavy water distillation process with metal-organic frameworks (MOFs) modified surface. MOFs with microstructure, microporous pores, and hydrophilic nature showed superhydrophilic properties, which was a benefit to better interface contact properties between water drop, vapor, and packing. High separation efficiency was achieved with height equivalent to a theoretical plate (HETP) of 1.53 cm under optimized operation conditions. Pilot-scale two-column tritiated heavy water distillation system was simulated and optimized based on experimental results with a high separation factor and recovery rate of tritium. This work demonstrates the first attempt at MOFs engineered surface for tritiated heavy water treatment, advancing a new concept in the design of tritiated heavy water distillation with intrinsic safety for industrial applications.

Design of a direct-cycle supercritical CO2 nuclear reactor with heavy water moderation

A new reactor concept is described that directly couples a supercritical CO2 (sCO2) power cycle with a CO2-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO2 concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (∼20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

Tailoring neutron beam properties by target-moderator-reflector optimisation

Compact accelerator-driven neutron sources allow to operate multiple optimised target-moderator-reflector (TMR) units adapted to the requirements of the respective instruments. The compact design of the TMR units allows an efficient coupling of neutron production, neutron moderation and extraction, but requires a novel way of optimisation. The neutronic performance of different TMR units based on polyethylene, heavy water and a mixture of heavy and light water moderators together with Pb and Be reflectors and a borated polyethylene absorber is discussed. Extraction channels for thermal and cold neutrons are investigated regarding the energy and time spectra.

Photoneutron production in heavy water reactor fuel lattice from accelerator-driven bremsstrahlung

The coupling of low-energy electron linear accelerators (eLINACs) to a large heavy water reactor is proposed to create an accelerator-driven photoneutron source (ADS). Photoneutron yields of 1012 pn/s per kW of beam power can be achieved in the ADS-CANDU concept where the wide fuel channel spacing of heavy water reactors represents a near-optimal geometry for conversion of accelerator-driven bremsstrahlung into photoneutrons in the heavy water moderator with minimal parasitic attenuation of photons in the fuel channels. Twenty MeV electron beam energy is most efficient at producing photoneutrons despite having the largest fuel channel shielding effect (28% attenuation compared to an infinite heavy water medium). The majority of photoneutrons are produced during source-photon first flights, so the spatial distribution and emission spectra of the ADS in the secondary (γ,n) converter are correlated to the doubly differential angle and energy distribution of the bremsstrahlung emitted from the (e-,γ) converter. Compton-scattered photons and tertiary bremsstrahlung originating from the electron–positron pair and recoil electron secondary particles are important contributors to the photoneutron yield of higher energy eLINACs systems. Photonuclear data for the deuterium photoneutron reaction cross section and secondary electron transport and tertiary bremsstrahlung production physics implemented in Monte-Carlo radiation transport codes can dominate ADS simulation results and are the same magnitude as the physical phenomena such as the fuel channel shielding effect.

On the Amount of Moderation in the CANDU Lattice

Throughout the years, various reports and training manuals on CANada Deuterium Uranium (CANDU) reactors have mentioned that the CANDU lattice is overmoderated. Overmoderation is not always defined in such documents but often appears associated with the positive void reactivity of the CANDU lattice. Some documents refer, logically, to overmoderation as meaning that the lattice pitch is larger than the pitch that maximizes the infinite-lattice multiplication constant but do not demonstrate this is the case for CANDU. We demonstrate that in fact, the CANDU lattice is undermoderated; that is, the current 28.575-cm lattice pitch is smaller than the pitch for which the infinite-lattice multiplication constant reaches its maximum. We hypothesize that the misconception of CANDU overmoderation may have originated from attributing the CANDU positive void reactivity to too much moderator by incorrectly equating the effect of losing heavy water coolant with the effect of losing heavy water moderator.

Preliminary analysis of fuel cycle performance for a small modular heavy water-moderated thorium molten salt reactor

Heavy water-moderated molten salt reactors (HWMSRs) are novel molten salt reactors that adopt heavy water rather than graphite as the moderator while employing liquid fuel. Owing to the high moderating ratio of the heavy water moderator and the utilization of liquid fuel, HWMSRs can achieve a high neutron economy. In this study, a large-scale small modular HWMSR with a thermal power of 500 MWth was proposed and studied. The criticality of the core was evaluated using an in-house critical search calculation code (CSCC), which was developed based on Standardized Computer Analyses for Licensing Evaluation, version 6.1. The preliminary fuel cycle performances (initial conversion ratio (CR), initial fissile fuel loading mass, and temperature coefficient) were investigated by varying the lattice pitch (P) and the molten salt volume fraction (VF). The results demonstrate that the temperature coefficient can be negative over the range of investigated Ps and VFs for both 233U-Th and LEU-Th fuels. A core with a P of 20 cm and a VF of 20% is recommended for 233U-Th and LEU-Th fuels to achieve a high performance of initial CR and fuel loading. Regarding TRU-Th fuel, a core with a smaller P (~ 5 cm) and larger VF (~ 24%) is recommended to obtain a negative temperature coefficient.

Trace determination of gadolinium by conductivity-based approach

A conductivity-based technique is developed for the determination of Gd3+ in the heavy water moderators of pressurized heavy water reactors (PHWRs). The method involves monitoring extremely small shifts in conductivity, in the order of few nS/cm, due to the continuous addition of a suitable complexing agent to Gd3+ in aqueous medium. The resulting plot gives two distinct regions with vastly differing slopes. Two multidentate ligands, ethylenediaminetetraacetic acid (EDTA) and diethylenetriaminepentaaceticacid (DTPA), as complexing agents are compared. A high performing conductivity detector based on a new class of sensors called pulsating sensors that works entirely in the digital domain is deployed to monitor the conductivity shifts. Titration plots are studied in both H2O and D2O, and the observed difference between the plots in the two matrices is discussed in detail. Boron did not interfere in the analysis. The method was validated using the UV–vis spectrophotometric technique. The method is sensitive and rapid, as each analysis takes 3 min. The limit of detection in H2O and D2O are 1.27×10−7 mol/L and 5.1×10−7 mol/L, respectively. The precision in analysis lies between 1.9% and 5.3%. This method has important application in the nuclear industry for the routine analysis of gadolinium.

A novel concept for a molten salt reactor moderated by heavy water

The Heavy Water moderated Molten Salt Reactor (HW-MSR) is a novel concept of a thermal heterogeneous nuclear reactor and pursues for Thorium-Uranium (Th-U) breeding. It adopts heavy water rather than graphite as moderator while employs the same liquid fuel of molten salt reactor (MSR). Thus this new concept inherits the positive properties of MSR (feasible for online reprocessing) and heavy water reactor (high neutron moderating ratio). Furthermore it can address the problems of depleted graphite management and positive temperature coefficient due to neutron spectral shift propitious to 233U fission existed in Molten Salt Breeding Reactor (MSBR). HW-MSR consists of the primary loop system, intermediate cooling loop, energy conversion system, external cooling system and online reprocessing system. Since there is a huge temperature difference between molten salt (higher than 600 °C) and heavy water (lower than 100 °C to keep high density) in the core, effectively preventing the heat transfer between them is one main challenge. To address this problem, a thermal insulator made of Yttria Stabilized Zirconia (YSZ) is applied and analyzed. The calculation results show that both the fuel salt and moderator outlet temperature across the core can satisfy the design requirements. Similar to a traditional MSR, the HW-MSR also implements an online reprocessing system which can online extract and recycle useful fuel during operation, providing a feasible approach for Th-U fuel cycle. In addition, because of the outstanding neutron performance of heavy water moderator, a small initial 233U loading mass is required for HW-MSR, and its breeding ratio can achieve 1.073, corresponding to a doubling time of 12 years at the reprocessing cycle of 10 days. Compared with traditional light water reactor (LWR) whose transuranium (TRU) production is 69 kg/GWy, HW-MSR produces only 0.51 kg/GWy. This much lower radiotoxicity of nuclear waste in the HW-MSR would significantly enhance the sustainability of nuclear energy.

Large-scale energy production with thorium in a typical heavy-water reactor using high-grade plutonium as driver fuel

Thorium can be introduced into the energy vector in combination with high-grade plutonium (HG-Pu). Excellent neutron economy of heavy-water moderator allows use of mixed ThO2/HG-PuO2 fuel in heavy-water reactors with high efficiency, leading to the exploitation of large world thorium reserves and extending the availability of the nuclear energy by two orders of magnitude. In the present work, the criticality calculations have been performed with the code MCNPX 3-D geometrical modeling of a typical heavy-water reactor, where the structure of all fuel rods and bundles is represented individually. In the course of time calculations, nuclear transformation and radioactive decay of all actinide elements as well as fission products are considered. Five different fuel compositions have been selected for investigations: (1) 97% thoria (ThO2) + 3% PuO2; (2) 96% ThO2 + 4% PuO2; (3) 95% ThO2 + 5% PuO2; (4) 94% ThO2 + 6% PuO2; and (5) 92% ThO2 + 8% PuO2. The behavior of the criticality k∞ and the burn-up values of the reactor have been pursued by full power operation at 640 MWel (2180 MWth) for approximately 7 years. Time calculations have been conducted with MCNPX and CINDER codes under consideration of all nuclear transformation and radioactive decay processes on the actinide isotopes and fission fragments. As the reactor allows fuel recharging at on-power operation mode, the reactor criticality has been followed down to keff,end = ~1.05. The corresponding burn-up values and operation periods for the investigated modes are (1) 18 GWd/MT and 780 days; (2) 27 GWd/MT and 1200 days; (3) 35 GWd/MT and 1560 days; (4) 44 GWd/MT and 1940 days; and (5) 60 GWd/MT and 2640 days. Among the investigated four modes, 94% ThO2 + 6% PuO2 seems a reasonable choice under consideration of the high price of the HG-Pu as driver fuel. The mixed fuel has the potential of an extensive exploitation of thorium resources. Reactor will run with the same fuel charge for approximately 5 years and allow a fuel burn-up approximately 44 GWd/MT, comparable with conventional light-water reactors (LWRs). Plutonium component of the mixed fuel will become nonprolific after few months of plant operation through the accumulation of even isotopes. Addition of few percent natural uranium to the initial mixed fuel charge will keep the 233U component below 22% and hence at nonprolific level over the entire plant operation period. Replacement of 4% ThO2 with 4% nat-UO2 will practically not change main technical parameters of the reactor.

Physics characteristics of internally cooled annular fuel for potential application in pressure tube heavy water reactors

This paper summarizes the results of exploratory lattice physics studies of alternative, advanced fuel bundle concepts that could potentially be implemented in pressure tube heavy water reactors (PT-HWRs). The lattice physics code WIMS-AECL was used to analyze the physics performance and operational characteristics of an 18-element and a 12-element internally cooled annular fuel (ICAF) fuel bundle, made with (LEU,Th)O2 fuel, with both low-burnup and high-burnup options. Such fuel bundles with annular fuel elements may be able to operate at higher bundle power levels and with higher linear element (LER) ratings than fuel bundles with conventional solid cylindrical fuel elements. In addition, the use of thorium mixed with LEU can help extend uranium resources, exploit the energy potential in thorium, and also reduce the production of plutonium and minor actinides, due to the smaller fraction of 238U in the fuel. The influence of improvements in the neutron economy of these lattices was also investigated, by incorporating higher-purity heavy water moderator and coolant (99.90 at.% D2O) in the models, along with enriched zirconium (95 wt% 90Zr/Zr) for the zirconium alloys used in the structural components. The results were compared with those for a more conventional 37-element PT-HWR fuel bundle using natural uranium (NU) fuel. Results show that annular fuels could be very attractive, being able to achieve higher burnup, comparable or better fissile utilization, reduced coolant void reactivity, and comparable or more negative fuel temperature coefficients.

Optimization of the Canadian SCWR Core Using Coupled Three-Dimensional Reactor Physics and Thermal-Hydraulics Calculations

The Canadian supercritical water-cooled reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water above the thermodynamic critical point as a coolant and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling, a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons, and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures, and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined so as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM/MOD4.0 model to accurately predict the MCST as a function of time during a fueling cycle.

A new approach on modeling of the b-viii, the ultimate achievement of the second “Uranverain”

German Nazi state conducted researches in nuclear technologies as an attempt to achieve various military goals. As the result of these researches, German scientists developed different, advanced nuclear technologies in years before and during World War II. In an attempt to develop the ?Uranmaschinen?, in which controlled release of high energy in fission process can be achieved, various approaches were examined, theoretically and experimentally. These studies were conducted under support of the German Nazi state and were known as the First and Second ?Uranverain? (Uranium Society/Club). Versions of the ?Uranmaschinen? were based, mainly, on natural uranium fuel and moderators of heavy water, regular water or paraffin. The latest known fission device was the subcritical nuclear fission reactor B-VIII, re-built in village Haigerloch, Bavaria, Southern Germany, in first months of 1945. It was a tank type device with natural uranium metal fuel and heavy water moderator, reflected by graphite. Radiation shielding of the device was achieved, primarily, by surrounding the reactor tank by regular water. The whole device construction was assembled inside a concrete hole in the floor of an underground cave, ex beer cellar. A recent neutronics study of this reactor was done, assuming fuel rods with lumped parameters approximation, by Italian Bologna University LIN (Laboratorio Ingegneria Nucleare) research group in 2009. This paper is a new approach to the neutronics study of the B-VIII reactor with an attempt to model real fuel-moderator geometry. This study points out many approximations and simplifications, made during the B-VIII material composition and geometry modeling, due to missing data. The paper investigates the influence to criticality of numerous uncertainties in the material compositions, mass densities and geometry of the facility. The Monte Carlo MCNP6.1 code with the latest ACE type neutron nuclear cross section data is used for that purpose. Additionally, an attempt of estimation of the uncertainty of the experimental result of the neutron multiplication was given. Differences in the calculated values of the neutron multiplication and the experimental one are investigated and tried to explain. These analyses show that the B-VIII was a subcritical device, as it was shown by the experimental results of the German scientists achieved in March-April 1945 in Haigerloch. <br><br><font color="red"><b> This article has been corrected. Link to the correction <u><a href="http://dx.doi.org/10.2298/NTRP1802230E">10.2298/NTRP1802230E</a><u></b></font>

Fuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor

The Canadian supercritical water-cooled nuclear reactor (SCWR) is a 2540 MWth channel-type SCWR concept that employs 336 fuel channels in the reactor core. Each fuel channel includes a pressure tube that is submerged in a heavy water moderator and contains a removable fuel assembly. The fuel assembly is designed so that all in-core components exposed to high radiation fields (other than the pressure tube) are part of the fuel assembly, which is removed from the reactor core as part of the assembly after three operating cycles. This design feature significantly reduces the likelihood of component failures due to radiation damage. To achieve high (>45%) power conversion efficiency, the Canadian SCWR operates at a supercritical water pressure (25 MPa) and high temperatures (350 °C at the inlet, 625 °C at the outlet). These conditions lead to fuel cladding temperatures close to 800 °C. Because of the reduced material strength at this temperature and higher fission gas production of the fuel, collapsible fuel cladding is selected over internally pressurized cladding. To increase heat transfer and to reduce cladding temperatures, turbulence-inducing wire-wraps are employed on fuel elements. Numerical models have been developed to analyze the thermal-structural behavior of Canadian SCWR fuel at normal and accident conditions. It was found that axial ridging, a possible failure mechanism with collapsed fuel cladding, can be avoided if the cladding thickness is larger than 0.4 mm. Detailed numerical analysis showed that the maximum fuel cladding temperature for the worst-case accident scenario is below the melting point by a small margin. This result was obtained with conservative assumptions, suggesting that the actual margin is greater. Hence, one of the design goals, the exclusion of the possibility of melting of the fuel, which is called the “no-core-melt” concept, seems attainable. However, this needs to be demonstrated more rigorously by removing the conservative assumptions in the analysis and performing supporting experimental work. This paper presents a description of the Canadian SCWR fuel assembly concept, its unique features, the rationale used in the concept development and the results of various numerical analyses demonstrating the performance and characteristics of the Canadian SCWR fuel channel.

Investigation of Fuel Burnup Impacts on Nuclear Reactor Safety Parameters in the Canadian Pressure Tube Supercritical Water-Cool Reactor

The Canadian pressure-tube super critical water-cooled reactor (PT-SCWR) is an advanced generation IV reactor concept which is considered as an evolution of the conventional Canada Deuterium Uranium (CANDU) reactor that includes both pressure tubes and a low temperature and pressure heavy water moderator. The Canadian PT-SCWR fuel assembly utilizes a plutonium and thorium fuel mixture with supercritical light water coolant flowing through the high-efficiency re-entrance channel (HERC). In this work, the impact of fuel depletion on the evolution of lattice physics phenomena was investigated starting from fresh fuel to burnup conditions (25 MW d kg−1 [HM]) through sensitivity and uncertainty analyses using the lattice physics modules in standardized computer analysis for licensing evaluation (SCALE). Given the evolution of key phenomena such as void reactivity in traditional CANDU reactors with burnup, this study focuses on the impact of fission products, 233U breeding, and minor actinides on fuel performance. The work shows that the most significant change in fuel properties with burnup is the depletion of fission isotopes of Pu and the buildup of high-neutron cross section fission products, resulting in a decrease in cell k∞ with burnup as expected. Other impacts such as the presence of protactinium and uranium-233 are also discussed. When the feedback coefficients are assessed in terms of reactivity, there is considerable variation as a function of fuel depletion; however, when assessed as Δk (without normalization to the reference reactivity which changes with burnup), the net changes are almost invariant with depletion.

OPTIMIZATION OF COLLIMATOR NEUTRON DESIGN FOR BORON NEUTRON-CAPTURE CANCER THERAPY (BNCT) BASED CYCLOTRON 30 MeV

<span>This research in BNCT has a goal to design a collimator that can be used for cancer therapy. Simulations were carried out by MCNPX software. A collimator is designed by cyclotron 30 MeV as a neutron generator. Independent variables varied were material and thickness of each collimator’s component to get five of IAEA’s standard of the neutron beam. The result is two collimator designs that can pass all IAEA’s standard. Those designs are cyclotron collimator I and cyclotron collimator II. Collimator designs obtained are tube collimator consisting of a cylindrical target </span><sup>7</sup><span>Be length of 1.4 cm and radius 1 cm, a lead wall with thickness 23 cm, cylindrical heavy water moderator (D</span><sub>2</sub><span>O) with radius 3 cm. Filter Cd-nat for cyclotron collimator I with a thickness of 1 mm and a radius 3 cm. Cyclotron collimator II uses </span><sup>60</sup><span>Ni with a thickness of 5 cm as a filter. The radius aperture is 3 cm. These two collimator designs can be used for cancer treatment with BNCT. Dosimetry calculation and manufacture of prototypes are needed to test the application of this design.</span>

Investigations of natural convection and circulation in Passive Moderator Cooling System of an advanced reactor in a scaled test facility

Passive systems are being examined for the future advanced nuclear reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator passively in case of an extended Station Black Out (SBO) condition. The heavy-water moderator gets heated due to heat transferred from the main heat transport system and the core structures; and rises upward due to buoyancy. This is cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The heat exchanger, in turn, is cooled by water from gravity driven water pool (secondary loop). The natural circulation loops provide sufficient cooling to prevent the increase of moderator temperature and pressure inside Calandria vessel beyond safe limits. In addition to this, decay heat is also dumped into GDWP tank through submerged isolation condensers.A scaled test facility was setup to simulate the thermal hydraulic characteristics of the PMCS and to analyze the performance behavior of the PMCS. Prolonged SBO transient experiment was performed in the test facility and analyzed using system code RELAP5/Mod 3.2. Besides, CFD simulations were also carried out for analyzing the temperature distribution inside calandria vessel of the test facility. Experimental results show efficient removal of the heat from the moderator and isolation condensers. The simulation results show that RELAP5 is able to predict the natural circulation phenomena very well and three dimensional temperature fields inside the vessel can be predicted by CFD. Temperature field as predicted by CFD is found to be in good agreement with the experiment.

Water distillation as a method of detritiation of heavy-water moderator

A D2O distillation column cascade for detritiating the heavy-water moderator of a nuclear reactor has been optimized for maintaining a steady-state tritium concentration of 5 Ci/kg therein. Use of a structured rolled ribbon–screw packing in columns with a large diameter and a random small-sized spiral prismatic packing at the final stage makes it possible to attain the necessary degree of separation in a five-column cascade at capital and operating costs typical of deprotiation.

Prediction of the moderator temperature field in a heavy water reactor based on a cellular neural network

Reactors with heavy water coolants and moderators have been used extensively in today's power industry. Monitoring of the moderator condition plays an important role in ensuring normal operation of a power plant. A cellular neural network, the architecture of which has been adapted for hardware implementation, is proposed for use in a system for prediction of the heavy water moderator temperature. A reactor model composed in accordance with the CANDU Darlington heavy water reactor design was used to form the training sample collection and to control correct operation of the neural network structure. The sample components for the adjustment and configuration of the network topology include key parameters that characterize the energy generation process in the core. The paper considers the feasibility of the temperature prediction only for the calandria's central cross-section. To solve this problem, the cellular neural network architecture has been designed, and major parts of the digital computational element and methods for their implementation based on an FPLD have also been developed. The method is described for organizing an optical coupling between individual neural modules within the network, which enables not only the restructuring of the topology in the training process, but also the assignment of priorities for the propagation of the information signals of neurons depending on the activity in a situation analysis at the neural network structure inlet. Asynchronous activation of cells was used based on an oscillating fractal network, the basis for which was a modified ring oscillator. The efficiency of training the proposed architecture using stochastic diffusion search algorithms is evaluated. A comparative analysis of the model behavior and the results of the neural network operation have shown that the use of the neural network approach is effective in safety systems of power plants.

Simulation and optimization of the deprotiation cascade of a heavy-water moderator

A simulation procedure and results of simulation are reported for the operation of a cascade of packed distillation columns for heavy water deprotiation. The mathematical model is based on the results of laboratory-scale tests of a rolled ribbon–screw structured packing developed at the Mendeleev University of Chemical Technology of Russia an a random spiral prismatic packing made from stainless steel. Use of the rolled ribbon–screw packing in the columns provides means to diminish the volume of the separation equipment without changing the cascade throughput capacity and separation factor.