Ship Reactor Research Articles

Research on Collapse Testing of Nuclear Icebreaker Reactor Hull Structure Based on Distortion Similarity Theory

In this study, the finite element method combined with the model test method are used to investigate the ultimate strength of a target ship reactor hull structure under a pure bending load. Based on the distortion similarity theory and nonlinear similarity method, a scale model of the actual ship reactor hull structure is designed and the model collapse test is conducted. The ultimate bending moment obtained by the model test is transformed to the actual ship through the similarity transformation relationship and compared with the nonlinear finite element analysis result of the actual structure. The results are consistent with each other, which indicates that the collapse characteristics of the actual ship reactor hull structure can be better forecasted using the model test results when the test model is designed based on the nonlinear similarity method and distortion similarity theory.

Study on the seawater cooled PRHRS of a nuclear propulsion ship with the upper heat exchanger for driving force and the seawater heat exchanger as an ultimate heat sink

As international maritime trade among countries increases, great efforts are being made to reduce the amount of greenhouse gases emitted from ships using conventional fossil fuels. A variety of environment-friendly power sources for ships are being considered, and nuclear energy is one of them. The research described in this paper focuses on elemental technologies related to nuclear-powered ships rather than the nuclear reactor technology itself. Among those technologies, the Passive Residual Heat Removal System (PRHRS) of a nuclear-powered ship is studied to remove residual heat by using seawater as an alternative heat sink. New concept of PRHRS is proposed and it is named as SWC-PRHRS. SWC-PRHRS has two heat exchangers: one is for driving force and another is for dumping the most of residual heat to the sea. Numerical analysis and experimental studies are conducted to improve the performance of SWC-PRHRS. The numerical results show that most of the residual heat comes from a reactor core is eliminated by sea water cooling proving the concept of SWC-PRHRS is working properly without any operator intervention and additional water supply. Experimental studies are conducted based on the numerical analysis results. From the experimental results, it is found that the mass flow rate of SWC-PRHRS is very sensitive to the filling ratio of SWC-PRHRS and there are three distinctive filling ratio ranges. Through the experiment, it is confirmed that SWC-PRHRS is working efficiently with the lower mass flow rate and the smaller upper heat exchanger without any operator intervention. SWC-PRHRS could remove a large amount of residual decay heat from the ship's reactor and release it to the sea. Considering the size, weight, and exceptional heat removal capability of SWC-PRHRS, it is clear that SWC-PRHRS is well suited for marine application usage.

Effects of rolling motion on helical coil once-through steam generator thermal-hydraulic characteristics

Steam generators are essential for the safe, economical, and reliable operation of nuclear reactors. Helical Coil Once-through Steam Generators (HCOTSG) offer unique advantages over other types of steam generators for ship reactors or offshore platforms. To investigate the effect of rolling motion due to ocean conditions on the safe and economic operation of HCOTSG, a new model and a new code are developed in this paper. The program is developed and allows numerical simulation of HCOTSG under ocean conditions. The validity of the code was verified by comparing the simulation results with the design parameters of the Marine Reactor X (MRX) steam generator and the simulation results of other dedicated programs. The code was further used to perform the International Reactor Innovative and Secure (IRIS) transient operating conditions under typical rolling motions. Then, the influences of different swing directions, angles, periods, and positions of swing axes from the origin of the system operating parameters are analyzed.In the cases discussed in this paper, the following conclusions are obtained: (a) The direction of the swing significantly affects the system. The most dangerous situation is around the x-axis, and the case around the z-axis is the safest, in which the rolling situation has little effect on the system because the centripetal force is perpendicular to the tube wall, and the gravitational pressure drop is constant. (b) when the swing angle increases, except the fluctuation speed is faster, the fluctuation range of the parameters also increases. (c) when the swing period changes, the parameters’ fluctuation range also change (d) the orientation of the swing axis from the origin affects the magnitudes of the parameters’ changes (e) the distance of the swing axis from the origin affects the magnitude of the parameter variation and the farther the swing axis is from the origin, the greater the parameter fluctuation.

Reactivity on Coolant Dynamics in Modern Ship Reactor Plants

Reactivity on Coolant Dynamics in Modern Ship Reactor Plants

Health risk analysis as per radiation and chemical factors in a zone influenced by a nuclear legacy object in the north-western part of Russia

Rehabilitation programs for nuclear legacy objects in north-western regions in Russia involve decommissioning; a former coast technical support base belonging to the RF Navy and located on the Gremikha peninsula in Murmansk region is now being removed from service. At present spent nuclear fuel is being extracted from ship reactors with liquid-metal coolant and nuclear wastes that have been accumulated during active operations are being treated. The current activities at the site result in complex influence on the environment determined by both radiation and non-radiation contaminants. The present work focuses on examining a hygienic situation existing in residential areas located close to this nuclear legacy object. In general, radiation situation in residential areas is determined by technogenic contaminants, basically 137Cs and 90Sr radionuclides. Our research has revealed that at the moment their contents in the environmental objects correspond to background levels that are typical for the examined region. Chemical contamination detected in residential areas is characterized with elevated heavy metals contents (Cd, As, Pb, V, Zn, Cu etc.) in soils in urban settlements that are higher than background levels. We detected lead and zinc concentrations that were higher than MPC. Total soil contamination with heavy metals (Zc) is still «permissible» in most residential areas. Carcinogenic health risk for people living in the examined residential areas does not exceed 10-6 as per radiation and chemical factors. Research results allow assessing the current hygienic situation as normal and concluding that decommissioning of SZC «SevRAO» objects (Gremikha section) does not exert negative impacts on the environment and health of people living in residential areas located nearby. The obtained data give grounds for controlling activities when environmental monitoring is accomplished during works aimed at recovering contaminated areas where nuclear legacy objects are located.

Анализ риска по радиационным и химическим факторам для населения в районе расположения объекта ядерного наследия в северо-западном регионе России

Rehabilitation programs for nuclear legacy objects in north-western regions in Russia involve decommissioning; a former coast technical support base belonging to the RF Navy and located on the Gremikha peninsula in Murmansk region is now being removed from service. At present spent nuclear fuel is being extracted from ship reactors with liquid-metal coolant and nuclear wastes that have been accumulated during active operations are being treated. The current activities at the site result in complex influence on the environment determined by both radiation and non-radiation contaminants. The present work focuses on examining a hygienic situation existing in residential areas located close to this nuclear legacy object. In general, radiation situation in residential areas is determined by technogenic contaminants, basically 137Cs and 90Sr radionuclides. Our research has revealed that at the moment their contents in the environmental objects correspond to background levels that are typical for the examined region. Chemical contamination detected in residential areas is characterized with elevated heavy metals contents (Cd, As, Pb, V, Zn, Cu etc.) in soils in urban settlements that are higher than background levels. We detected lead and zinc concentrations that were higher than MPC. Total soil contamination with heavy metals (Zc) is still «permissible» in most residential areas. Carcinogenic health risk for people living in the examined residential areas does not exceed 10-6 as per radiation and chemical factors. Research results allow assessing the current hygienic situation as normal and concluding that decommissioning of SZC «SevRAO» objects (Gremikha section) does not exert negative impacts on the environment and health of people living in residential areas located nearby. The obtained data give grounds for controlling activities when environmental monitoring is accomplished during works aimed at recovering contaminated areas where nuclear legacy objects are located.

Refueling Complexes and Equipment for Civilian and Naval Ship Reactor Installations: Creation and Development

The history of the creation and development of reloading systems and reloading equipment for refueling reactor installations on civilian and naval ships, commencing in the late 1950s, is described. The PU-1 equipment was created for unloading fuel from the reactor of our country’s first nuclear submarine. Subsequently, reloading equipment kits were created in different years for refueling and unloading nuclear fuel from naval ship reactors – PU-2, OK-300PB, OK-300PBM/650-3, KN-3PB, OK-300PBU, and AUPO. The next phase of the development of the reloading equipment became the creation of refueling complexes for civilian ships: the floating nuclear thermal power plant Akademik Lomonosov and the general-purpose nuclear icebreaker (design 22220). For over 70 years, the listed equipment has been used on naval and civilian ships for refueling and unloading of nuclear fuel.

OKBM Afrikantov Research and Testing Complex: From Creation to the Present Day

Testing and experimental substantiation are a significant and integral part at all phases of the life cycle of reactor equipment. The Research and Testing Complex (RTC) operates at OKBM Afrikantov to ensure that these tasks are successfully fulfilled. The RTC provides the opportunity of performing a wide range of research and testing during the creation of reactor installations as well as the development of the main and auxiliary equipment for the nuclear energy industry. The most significant work performed on the experimental base at OKBM in the last 15–20 years reduces to neutronics and thermotechnical studies of cores and FA, testing equipment for naval and civilian ship reactor installations and NPP, experimental substantiation of systems and study of processes occurring in them and the elements of the equipment in nuclear installations, and the study of the corrosion and strength properties of materials. Work on modernization of currently operating and creation of new test stands is ongoing at the RTC.

Nuclear Ship Reactor Installations: From Gen 1 to 5

The use of nuclear energy on submarines has made autonomous and independent navigation possible for long periods of time. OKBM Afrikantov – one of the leading enterprises of the Rosatom State Corporation – has been supplying Russia’s naval ships with nuclear reactor installations for more than 65 years commencing with the development of equipment for the installations on first-generation submarines. Currently, all active ships – second, third, and fourth generation submarines and nuclear surface ships – are powered by reactor installations developed by OKBM. Commencing with the complete nuclear reactor installation set for the Yasen class nuclear submarine, OKBM has functioned as the complete supplier of equipment for all installations developed and being developed for ships and vessels which are under construction or being modernized. Currently, OKBM is designing ship nuclear reactor installations for promising naval ships.

МОДЕЛИРОВАНИЕ ТЕПЛОПЕРЕДАЧИ В КАНАЛАХ КРУГЛОЙ ФОРМЫ МАШИН НЕПРЕРЫВНОГО ЛИТЬЯ ЗАГОТОВОК С ПЕРСПЕКТИВНЫМ СВИНЦОВО-ВИСМУТОВЫМ КРИСТАЛЛИЗАТОРОМ

As part of the search for new areas of application of liquid metal coolants, their use for cooling crystallizers of steel in continuous casting machines instead of water has been proposed. This solution will reduce energy costs in the production of steel due to the implementation of the steam-power cycle with the generation of electricity. This article describes the results of the justification of the optimal liquid metal coolant and the modeling of cooling of a round crystallizer channel by it. It is shown that the eutectic alloy of lead with bismuth, which was successfully used as a coolant for ship reactors in the Soviet Union, is best suited to solve this problem. A comparative simulation of cooling of hardened steel in a round channel with a lead-bismuth eutectic with temperatures of 180 °C and 450 °C and water with a temperature of 20 °C revealed that replacing water with a liquid metal coolant practically does not reduce the efficiency of steel production. The greatest heat flux is removed from the steel crystallizer at a temperature of lead-bismuth eutectic at the inlet of the steel crystallizer equal to 180 °C. The temperature of the lead-bismuth eutectic at the outlet of the steel crystallizer reaches 400 °C, which is sufficient to generate electricity in the steam-power cycle. The choice in favor of modeling in comparison with the experiment was made in accordance with current trends in cost reduction due to the use of computer-based calculation tools. It is advisable to continue research to select the wall material and determine the optimal shape of the cooling channels of the crystallizer for maximum intensification of heat transfer between hardening steel and lead-bismuth coolant.

Fuel options for nuclear ship reactors featuring reactivity swing below one dollar

Environmental protection against air pollution and climate change has drawn great and increasing public attention in recent decades. Nuclear power involving no pollutant and CO2 emissions provides a more sustainable way for commercial shipping activities. Nine fuel options are simulated to find the combinations of key parameters that make the core to have a lifetime reactivity swing below one dollar. Such a low reactivity swing, when combined with a negative feedback reactivity coefficient, may facilitate passive load follow and passive safety, which means the reactivity can be controlled through temperature or thermal expansion of the reactor core. These fuel options are compared regarding reactor physics and thermal hydraulics, under certain constraints associated with ship reactors. WIMS code is used for burnup, perturbation and parameter calculating. MONK and SERPENT codes are used to verify the WIMS model. It is found that ship-based lead-cooled fast reactor can be designed to have superior core safety and good economics. Different fuel options show diverse characteristics in terms of power output limits, refuelling intervals, natural circulation cooling capabilities and other safety related parameters.

Enhanced Life Assurance for Ship Reactors

The reasons for the appearance of alternating deviations of the parameters in transient regimes, which negatively impact the resource characteristics of ship reactors, are investigated. A definitive physical mechanism and certain key phenomena responsible for the appearance and evolutionary character of the deviations of the defining parameters are revealed on the basis of qualitative analysis confirmed by computational studies. This assures the development of the most effective technical measures for reducing the dynamic overshoots in operating and in-design plants.

Radiation Safety in Handling High-Level Wastes in Gremikha Village

As part of the work on eliminating the nuclear legacy temporary storage facilities for high-level solid wastes were opened, radiometric inspection of the wastes was performed, wastes were sorted, control rods of ship reactors were identified in the wastes and removed, and spectrometric determination of the induced activity in each rod was made. Forty-four rods with activity totaling 1.35·1013 Bq were packaged and shipped to a long-term storage site. The activity distribution was determined from samples and will be taken into account in planning subsequent work on the handling of such wastes and picking the most appropriate type of packaging. The work was performed using remote-control machines in compliance with the standards and regulations for the radiation safety of personnel.

Nuclear Power Plants for the Icebreaker Fleet and Power Generation in the Arctic Region: Development Experience and Future Prospects

The technical characteristics of a power series of reactor installations developed at Afrikantov OKBM are presented. The technical qualities making facilities equipped with modern ship reactor installations commercially attractive are shown to have undergone evolutionary improvement. First and foremost these are facilities for universal nuclear icebreakers under construction and an icebreaker in the design stage. Designs of optimized floating and ground-based power-generating units with improved user qualities are being developed on the basis of the experience gained in the development of a floating nuclear heat-and-power plant based on the Akademik Lomonosov floating power-generating unit. Promising reactors of the type RITM are used in these designs. They differ fundamentally from the block-type units by the integral steam-generating unit, heightened energy resource, and long intervals between refuelings, which are possible owing to the innovative core with metal-ceramic fuel and new structural materials. The possibility of significantly increasing the resource of the new reactor installations is confirmed by its stepped increase on operating nuclear ships and by scientific-technical studies.

Project Development of an Underground Nuclear Power Plant on the Basis of the Integrated Ship Reactor, KN-3

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Primary design and optimization of shielding for nuclear medical ship reactor

Primary design and optimization of shielding for nuclear medical ship reactor

March of the jellies

Desperate attempts to deal with jellyfish swarms may be wrong-headed, says Tamar Stelling

Non-dimensional analysis of the transient flow rate and the coolant temperature in a reactor coolant system under different pump failures

ABSTRACTThe mathematical models are developed to solve the non-dimensional transient flow rates in two loops and a reactor core under different power failures of reactor coolant pumps. Comparison of the experimental results of the Qinshan Nuclear Power Plant and the test results of the nuclear ship reactor in Japan under one pump power failure shows an excellent agreement. The non-dimensional flow rates are determined by the established non-dimensional parameters λ, ϵ, and α. Under the sequential power failure of two pumps, the non-dimensional flow rates are determined by the established λ, ϵ, α, and ΔT parameters. λ, ϵ, α, and ΔT are four important non-dimensional parameters in the prediction of flow transients. λ indicates the resistance coefficient ratio of the single loop to the reactor core, ϵ indicates the fluid inertia ratio of the reactor core to the single loop, α indicates the ratio of the initial kinetic energy of the single loop coolant fluid to the effective initial kinetic energy of the reactor coolant pump, and ΔT means the non-dimensional time interval of the sequential power failure of two pumps. The effects of λ, ϵ, α, and ΔT on the non-dimensional flow rates and the temperature change are investigated.

Curvature weight method of solving the point reactor neutron kinetic equations

The point kinetic equations are the system of a couple stiff ordinary differential equations. Many studies have focused on the development of more advanced and efficient methods of solving the equations, such as the high order Taylor polynomials method, the Haar wavelet operational method, the fractional point-neutron kinetic model method, the basis function method, the homotopy analysis method, and other methods. Most of these methods are successful in some specific problems, but still have, more or less, disadvantages. For example, the accuracy of the Haar wavelet operational method is limited by the collocation points, and it needs more computing time for a high precision. Aiming at the requirements that some numerical calculation results must have the higher precision and only the positive error in the nuclear reactor safety engineering and ship reactor for the maneuverability, in this paper we try to look for a new numerical method to satisfy that the calculation value is slightly higher than the real value when the actual curve is upward convex or downward concave, and the error is not greater than that by the Euler and improved Euler method. The new method is so-called the curvature weight (CW) method, which is based on the curvature circle method and considers the contributions of two curvatures at the interval step point to the average curvature inside the interval step. Using the decoupling method to remove the stiffness of equations and the instantaneous jump approximation to derive the neutron differential equations, the first and second derivative of neutron density are obtained. Then the CW method is used to solve the point reactor neutron kinetic equations, and thus obtaining the numerical solution. Compared with the results by the Euler and improved Euler method, the numerical calculation results by the CW method are always higher than the real value, and the calculation accuracy and speed are improved significantly. When this new method is used to solve the point reactor neutron differential equations with the step and linear reactivity inserted into the subcritical reactor, the numerical results which satisfy the requirements of positive calculation error and high precision can be obtained quickly. After improving the calculation step length, the precision reduction by the CW method is significantly lower than that by the Euler and improved Euler method. So the CW method can greatly shorten the total computing time, and it is also effective for most of differential equation systems.

Accident process and consequence research for LOCA combining with blackout accident of ship reactor

Accident process and consequence research for LOCA combining with blackout accident of ship reactor