Use of weapons plutonium in VVÉR-1000 reactors

One scenario for using excess Russian weapons plutonium is to load it into VVéR-1000 reactors. It is proposed that up to 40% of the fuel assemblies with uranium fuel be replaced with structurally similar fuel assemblies with mixed uranium-plutonium fuel. The stationary regime for burning fuel has the following characteristics: the run time is about 300 or 450 eff. days, the yearly plutonium consumption reaches 450 kg, the neutron-physical characteristics are close to the corresponding regimes with uranium fuel. The nuclear safety criteria and the irradiation dose for workers handling fresh and spent mixed fuel remain within the limits of the normative values. The use of mixed fuel makes it necessary to upgrade certain systems at nuclear power plants. A substantial quantity of weapons plutonium can be loaded every year into VVéR-1000 reactors, effectively using the energy potential of this plutonium. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 215–222, October, 2007.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

Spent fuel burnup estimation by Cerenkov glow intensity measurement

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to the declared burnups for the tested spent fuel assemblies for cooling times ranging from 900-2.000 d     

Analysis of decay heat removal from RBMK-1500 reactor in decommissioning phase by natural circulation of water and air

The main problem in nuclear energy is providing of safety at all stages of lifetime of nuclear installations in conditions of normal operation, accidents and at shutdown. Ignalina NPP, located in Lithuania, is one of the latest with RBMK reactors at highest capacity. Ignalina NPP has two units, both are closed for decommissioning now (in 2004 and 2009). Both units are equipped with RBMK-1500 reactors, the thermal power output is 4200 MW, the electrical power capacity is 1500 MW for each. In RBMK-1500 reactor the fuel assemblies remain for long time inside reactor core after the final shutdown. The paper discusses possibility of heat removal from the RBMK-1500 core at shutdown condition by natural circulation of water (1) and air (2) inside the fuel channels. In first case the decay heat from fuel assemblies is removed due to natural circulation of water and the piping above reactor core should be cooled by means of ventilation in the drum separator compartments. To warrant free access of air in to fuel channels (in the second case) the reactor cooling system should be completely dry out and the pressure headers and the steam discharge valves in steam lines should be opened. If mentioned conditions will be fulfilled, the reactor core will be cooled by natural circulation of water or air and fuel rods remain intact.     

Experience in using uranium-erbium fuel in power-generating units with RBMK-1000 reactors

The main reasons for and the results of switching to uranium-erbium fuel in the units of the Lengingrad, Kursk, and Smolensk nuclear power plants are presented. It is shown that uranium-erbium fuel made it possible to regulate the steam coefficient of reactivity, upgrade the control rods, lower the power density in the core, increase the reliability of the fuel assemblies, increase burnup, decrease the volume of spent fuel, and improve the commercial indicators. The prospects for improving the characteristics of uranium-erbium fuel for RBMK-1000 reactors are also presented. __________ Translated from Atomnaya énergiya, Vol. 100, No. 3, pp. 165–170, March, 2006.     

Analysis of accidents leading to local flow stagnation in parallel fuel channels of RBMK-type reactors

The RBMK-type nuclear power reactors, still operating in Russia, are graphite-moderated with vertical fuel channels, using low-enriched nuclear fuel. The main challenge, which leads to the overheating of the fuel assemblies, fuel channels and other core components in channel type nuclear reactors, is a misbalance between heat generation in core structures and heat sink, which can appear due to the loss of coolant accident. In this accidental case, the emergency core cooling system ensures the core cooling. In RBMK-type reactors this system consists of hydro-accumulators and a number of pumps, taking water from large water reservoirs. This equipment injects water into the fuel channels through the group distribution headers at high pressure. However, the direct supply of cold water from emergency core cooling system into fuel channels is possible only if check valves on group distribution headers are closed properly. If these check valves failed, the part of water would be lost through the break, the flow stagnation in channels could occur, which might lead to overheating of fuel assemblies in the fuel channels. This paper presents the results of deterministic safety analysis, performed using system thermal hydraulic code RELAP5. Using this code the reactor cooling system of RBMK-1500 was modelled and analyses of loss of coolant accidents with failure of few check valves in group distribution headers were performed. The results of the calculations are used for the development of symptom-based emergency operating procedures for RBMK-1500. The basic principles that describe the complex distribution of water flows in vertical forced circulation circuit with parallel fuel channels can be adjusted for the RBMK-1000 reactors, still operating in Russia.     

Thermohydraulic Aspects of Deep Burnup of Nuclear Fuel in Fast Reactors

The results of thermomechanical and thermohydraulic studies showing the relative effect of the deformation of fuel-element claddings and lattices in fast-reactor fuel assemblies on their temperature regimes are presented. It is shown that the temperature nonuniformities in fuel assemblies largely determine the deformation of fuel assemblies and, in turn, the operating efficiency and, correspondingly, the degree of burnup of nuclear fuel in fast reactors. The increase in the efficiency of the fuel assemblies is largely due to temperature smoothing, including smoothing of local temperature nonuniformities. Various solutions to technical and structural problems can accomplish this.     

Analysis of criticality in shipment and storage of fuel at a nuclear power plant with a VVÉR reactor

The substantiation of nuclear safety during shipment and storage of fresh and spent fuel at nuclear power plants with VVéR reactors is examined in the light of the more stringent nuclear safety rules. Possible technical measures for satisfying the safety criterion are examined, for example, the concept of subcritical fresh fuel. An example of the estimation of the probability of the formation of a critical mass as result of fuel assemblies falling randomly out of a container is presented. Certain characteristic features of the calculation of the neutron-physical characteristics of fuel in a cooling pond are presented, for example, the nonconservative nature of a separate analysis in the infinite approximation. 4 figures, 5 references. OKB “Gidropress”. Translated from Atomnaya éneriya, Vol. 87, No. 1, pp. 11–16, July, 1999.     

Transmutation of Americium in a Heavy-Water Reactor

It is shown that 22.5 metric tons of americium from the spent fuel of 30 VVÉR reactors which operated for 30 yr can be transmuted in a 1 GW(t) heavy-water system in 103 yr using as fuel the plutonium from the same spent VVÉR fuel. This means that 7.5 VVÉR reactors (CUF = 0.85) must be maintained simultaneously for fuel storage time 30 yr (for a 3-yr fuel storage period, the number of VVÉR reactors maintained increases to 25). In the entire period of operation of the system, a substantial quantity of plutonium from the spent fuel is used – about 150 metric tons (with total plutonium production in VVÉR reactors of about 200 metric tons) and about 38 metric tons of fissioning isotopes are burned. Therefore, with up to 98% burnup of americium in the target material the conversion coefficient defined as the ratio of the mass of the americium annihilated to the mass of the spent fissioning material is about 0.57.     

Conceptual designing of reduced-moderation water reactor with heavy water coolant

The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06–1.11.     

Work on a program for lowering fuel enrichment in research reactors to 19.7% at the Russian Science Center Kurchatov Institute

The work performed at the Russian Science Center Kurchatov Institute on the program for lowering the enrichment of fuel in research reactors is briefly described. The developed fuel assemblies and fuel elements with 19.7% ²³⁵U enrichment fuel, their testing in reactors, and post-reactor studies are described.     

Possibility of long-term nuclear power development based on today’s technology

The possibility of long-term nuclear power development with a uranium fuel cycle based on ²³⁸U burnup and todays industrial technology is investigated. It is shown that such development is possible with fast reactors, including with sodium coolant. In this case, incomplete fuel reprocessing is admissable in a closed fuel cycle employing a pyroelectrochemical technology, which allows some fission products and actinides to be present in the fresh fuel prepared for reloading after reprocessing. These fission products and actinides can be burned in a reactor, thereby decreasing the quantity of radioactive wastes compared with the complete reprocessing with chemical separation of the fuel elements and decreasing the radiation load on the environment.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 252–260, October, 2004.     

Density effect of reactivity in a heavy-water power reactor operating in the <Superscript>233</Superscript>U self-fueling regime

The void coefficients of the reactivity of different channel-type power reactors are compared. It is shown that a heavy-water channel reactor operating in a self-fueling regime within a uranium–thorium fuel cycle is just as nuclear-safe as CANDU type reactors. When composite fuel assemblies containing fuel elements with fuel and a ThO₂ target are used, such a reactor possesses negative void and therefore power coefficient of reactivity. Consequently, its nuclear safety is substantially higher than that of channel power reactors cooled by heavy or light water. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 249–254, November, 2008.     

Work on the program for lowering the fuel enrichment in research reactors to 36% at the Russian Science Center Kurchatov Institute

The work performed at the Russian Science Center Kurchatov Institute on the program for lowering the fuel enrichment in research and experimental reactors is briefly described. The designs of fuel assemblies and fuel elements containing fuel with 36% enrichment with ²³⁵U, their testing in reactors, and post-reactor studies are described.     

Special Features of the Statistical Processing of Seal Monitoring Data on Fuel-Cell Cladding in the ÉGP-6 Reactors at the Bilibino Nuclear Power Plant

The results of a statistical analysis of the indications of the seal-monitoring system for the outer cladding of the fuel elements in the fuel assemblies of the ÉGP-6 reactors at the Bilibino nuclear power plant are presented. It is shown that the distribution of the indications deviates substantially from the normal law, approaching the log-normal law. It is suggested that this phenomenon is due to seepage of reactor gas into the gas channels of a fuel assembly through the gaps between the graphite sleeves of the assemblies.     

Analysis of the processes in spent fuel pools of Ignalina NPP in case of loss of heat removal

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shutdown for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. According to the design, the spent fuel should be returned for reprocessing to Russia. However actually any fuel assembly has not been taken out from territory of the Ignalina NPP and all assemblies of spent fuel are stored in the spent fuel pools and dry on-site storage facility. Thus, the safety justification of facilities for intermediate spent fuel assemblies’ storage in Ignalina NPP is very important. This paper presents the results of loss of heat removal accidents (the most probable beyond design basis accident) in spent fuel pools of Ignalina NPP. The analysis was performed by employing best-estimate system thermal hydraulic code RELAP5 and codes for severe accidents ATHLET-CD and ASTEC. The best-estimate analysis, performed using RELAP5, allowed to investigate in the details the water evaporation, uncovering and fuel assemblies heat-up processes, when heat removal from the structures of buildings and pools are evaluated. The processes of spent fuel assemblies’ degradation due to loss of long-term heat removal were analyzed using ATHLET-CD and ASTEC codes. The results of calculations showed that the increase in water temperature in the pools from 50 °C up to 100 °C takes approximately 80-110 h, the evaporation of water volume down to uncovering of fuel assemblies takes approximately 220-260 additional hours. Later, after 200-300 h, the temperature of fuel claddings exceeds 800-1000 °C and the failures of fuel claddings occur due to cladding ballooning. The total amount of hydrogen generated up to time of complete water evaporation from spent fuel pools is about 7500-16,000 kg. These results of performed analysis were used for development of accident management guidelines for spent fuel pools of RBMK-1500.     

Determination of the Fuel-Assembly Burnup in a Research Reactor by Repeated Short-Time Irradiation Followed by γ Spectrometric Measurements

Experiments performed to determine the absolute fuel burnup in spent fuel assemblies in the IRT research reactor at the Moscow Engineering Physics Institute are described. The method is based on measuring the residual amount of ²³⁵U in the spent fuel asemblies with respect to the activity of the fission product ¹⁴⁰La accumulated in fresh and spent fuel assemblies after they were irradiated for a short time in the reactor core. A fresh fuel assembly with known uranium mass was used as a standard. The neutron flux was monitored using Al + Cu and Al + Co wires placed at the center of the fuel assembly. Small corrections for the difference in the spectrum amd the flux density of the neutrons in fuel assemblies with different uranium content were obtained from the calculations. The burnup of the three fuel assemblies studied was determined to within less than 2%.     

Handling spent nuclear fuel and radioactive wastes on the shore servicing base in Gremikha

The situation which has developed at the shore base in Gremikha involving fuel assemblies in the removed cores of water-moderated water-cooled reactors (first-generation submarines), which are located on an open site in containers and receiving chambers, and involving solid and liquid radioactive wastes present at the base is examined. Data are presented on the number of fuel assemblies and their technical state and on the state and amount of solid and liquid radioactive wastes. Suggestions on what should be done with the fuel assemblies and wastes are discussed. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 61–65, July, 2006.     

Handling spent nuclear fuel from the research reactors of the Russian Science Center Kurchatov Institute

Shipping out the spent fuel of the research reactors at the Institute for reprocessing is examined. The spent fuel is characterized by a great diversity of structural characteristics of the fuel assemblies and fuel elements, fuel compositions, and the enrichment, burnup, and cool-down times of the fuel as well as the state of the components of the assemblies and the structural materials. A classification and quantitative indicators of the accumulated spent fuel from the standpoint of the modern state of its reprocessing technology and the requirements for delivery to the Mayak Industrial Association are presented. The structural features of the TKU-19 and -128 shipment containers are presented, and the loading of spent fuel assemblies into them for shipment to reprocessing is described. The plans and goals of further work on the removal of spent fuel from the Institute’s territory are presented. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 99–105, February, 2009.