Preparation and removal for reprocessing of spent VVR-2 and or nuclear fuel from the Russian Science Center Kurchatov Institute

The details of the preparation and removal of spent nuclear fuel from the Institute’s VVR-2 and OR research reactors for chemical reprocessing are presented. The spent fuel is represented by fuel assemblies which have different shapes and contain EK-10 fuel elements with similar construction and UO₂–Mg 10% enrichment kernels or S-36 fuel elements with U–Al alloy kernels with 36% enrichment. The storage conditions for the spent fuel are described. The details of the procedures developed to identify fuel assemblies by type of fuel elements are presented. The choice of the TUK-19 shipment container for loading and transporting spent fuel for reprocessing is validated. The details of the loading of spent fuel assemblies into TUK-19 are described; these operations are performed by workers under a protective layer of water in a handling room specially designed for such purposes. Translated from Atomnaya énergiya, Vol. 106, No. 4, pp. 201–209, April, 2009.     

Residual energy release and actinide and fission product radiotoxicities during long-term storage of high burnup spent VVÉR fuel

The residual energy release and radiotoxicity of spent high burnup VVéR-1000 fuel during long-term storage is investigated as a function of time. The contributions of α, β, γradiation and radiotoxicity-the maximum admissible activity of nuclides in air and water-are taken into account in the calculations of the energy release. The data presented can be used to develop methods for handling spent nuclear fuel from prospective power reactors. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 292–296, May, 2007.     

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. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 56–61, July, 2006.     

Safety of the storage of spent submarine reactor compartments with no nuclear fuel on or under water

This article is a continuation of the radiation safety analysis of storage of spent submarine compartments with no nuclear fuel in an inlet. The first results were, published in Atomnaya énergiya,85, No. 3, 233–238 (1998). In contrast to the previous analysis, the danger of leakage of radioactive substances not only from shielding materials but also from a PWR type reactor with no fuel is examined. Marine storage of spent compartments in a sealed container is shown to be safe under conditions such that water flows past the container at a rate of 10⁷ m³/yr. 1 figure, 4 tables, 7 references. Russian Science Center “Kurchatov Institute” Translated from Atomnaya énergiya, Vol. 86, No. 3, 225–232, March, 1999.     

Salvaging the Lepse tender

The basic problems of salvaging the Lepse tender of the Murmansk Shipping Company are examined. The main sources of nuclear and radiation danger are determined, the characteristics of their structural features which prevent the appearance of a spontaneous reaction are given, the consequences of long-time storage of spent fuel assemblies are described, and different variants of their removal from storage are proposed. A computational validation of the required degree of subcriticality of the spent nculear fuel and an assessment of possible accidents, including accidents accompanying sinking of the vessel at its permanent base, are given. The salvaging steps are described, the bottlenecks are determined, schemes for handling the vessel are validated, and the radiation load on the workers is estimated. __________ Translated from Atomnaya énergiya, Vol. 101, No. 1, pp. 65–69, July, 2006.     

Materials-technology investigations of fuel elements, irradiated in a reactor at the obninsk nuclear power plant, after standing for 38 years in a depository

The results of materials-technology investigations of a spent fuel assembly from a reactor at the Obninsk nuclear power plant, the first nuclear power plant in the world, before the rated burnup and after prolonged dry storage (for about 40 years) were presented. It was established that the fuel elements from the fuel assembly studied are in satisfactory condition. No appreciable damage due to the prolonged storage was found: the outer diameter remains within the technological tolerance limits and the strength and the plasticity of the jackets are high. Only surface corrosion damage to 10 μm depth was found on the fuel-element jackets. The fuel composition remained whole. 6 figures, 1 table, 3 references. State Science Center of the Russian Federation—A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 3, pp. 183–188, March, 2000.     

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.     

Microbial biofilm growth on irradiated, spent nuclear fuel cladding

A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 10³ Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.     

Some aspects of nuclear and radiological safety of operating a spent nuclear fuel repository

A model of an irregular situation in a spent nuclear fuel repository with the introduction of excess reactivity into the system, consisting of containers with spent fuel assemblies and water, is examined. The neutron kinetics of a critical system is calculated taking account of the thermohydraulics of the system. The character of the flow of a short-time self-sustained chain reaction — “neutron burst” — is described. It is found that an excursion of the system in the range of reactivity introduction rates examined will result in heating of the system and self-quenching of the chain reaction by negative reactivity effects with respect to fuel temperature. Intense fluxes of fission neutrons and prompt gamma rays, accompanying a self-sustained chain reaction, are formed in the excursion process. A mixed neutron and gamma ray field near the system considered is investigated. __________ Translated from Atomnaya énergiya,Vol. 104, No. 3, pp. 141–147, March, 2008.     

Innovative development of nuclear power in Russia

The basic principles for performing analysis and the systems requirements for large-scale nuclear power in our country are formulated. The problems of modern nuclear power are examined and ways for modern nuclear power to transition to innovative development while satisfying these systems requirements for fuel use, handling spent fuel and wastes, and nonproliferation are indicated. The basic scenario of innovative development in the near term (up to 2030) is based on using predominantly ²³⁵U as fuel and water-moderated water-cooled reactors, which have been well mastered, for increasing nuclear capacities with limited introduction of fast reactors for solving the problem of spent fuel from thermal reactors. In the long term (2030–2050), a transition to ²³⁸U as the primary raw material with fast reactors predominating and complete closure of the nuclear power fuel cycle will be made. The journal variant of a report “New-Generation Nuclear Energy Technologies” presented at a meeting of the Scientific and Technical Council of Rosatom, Moscow, September 27, 2006. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 147–155, September, 2007.     

Challenges in the renewal of the Swedish sea transport system

Abstract

Since 1985, SKB has successfully operated a sea transport system for transport of spent nuclear fuel and radioactive waste to the intermediate storage facility, Clab and the final repository, SFR, in Sweden. The main components in the system are the ship M/S Sigyn, transport casks for spent fuel and core components, IP2 containers and terminal vehicles.     

Experimental investigation of the physical properties of salt melts containing sodium and lithium fluorides and beryllium difluoride

Experimental measurements of the basic physical properties of the melt NaF-LiF-BeF₂ are presented as validation of the concept of a molten-salt reactor for burning actinides from spent fuel from light-water reactors. Compositions which are characterized by the minimal molar fraction LiF 15–17% and BeF₂ 25–27% and meet the special requirements for a fuel salt for the concept under study are found. The melts of the fluorides of three metals have an acceptable melting temperature (<500°C), permit dissolution of actinide and lanthanide trifluorides to molar fraction 2% and higher at 600°C, possess good neutron-physical (even without enrichment with respect to ⁷Li) and thermophysical properties, are compatible with nickel-molybdenum alloys to temperature 750°C, are inexpensive and are not strongly activated by neutrons so that they do not present a long-term disposal risk. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 364–372, November, 2006.     

Optimization of the thorium-uranium regime in CANDU heavy-water reactors

Two operating regimes of a heavy-water power reactor operating in a thorium-uranium fuel cycle are examined: accumulation of the required amount of ²³³U and self-fueling with ²³³U fuel. The parameters of 10 variants of the core-lattice cells of a heavy-water power reactor are calculated. The lattice spacing is kept constant and the variants differ by the number of channels, containing fuel elements or targets, and the ²³³U content in ThO₂. Combined channels containing fuel elements and a target at the same time are examined. Preference is given to a cell variant where a cell contains three channels with fuel elements and one channel with targets. For this variant, the fuel burnup is ∼8 MW-days/kg. A large increase of the burnup, i.e., decrease of the amount of reprocessed targets, can be achieved by decreasing the minimum reactivity excess and also by changeing and increasing the complexity of the technology. For example, a large effect is expected from using combined fuel assemblies. In so doing, provisions must be made for performing complicated operation of disassembling highly active fuel assemblies consisting of fuel elements and targets. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 327–336, November, 2006.     

Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation

The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO₂; (2) ThO₂ denaturated with 10% natural-UO₂ and (3) ThO₂ denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li₂BeF₄’) are selected for the nuclear heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m² and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO₂ produces weapon grade ²³³U. The denaturation of the ²³³U fuel is realized with a homogenous mixture of 90% ThO₂ with 10% natural-UO₂ as well as with 10% LWR spent nuclear fuel. The homogenous mixture of 90% ThO₂ with 10% natural-UO₂ can successfully denaturate ²³³U with ²³⁸U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO₂ with 10% LWR spent nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.     

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.     

Computerised Modelling of Radiolysis and Corrosion Effects in Transport Containers for Fuel Spent Nuclear

Abstract

The radiolysis of water and/or gases within transport containers for spent nuclear fuel may result in the generation of hydrogen and oxygen gases and also the enhanced corrosion of the materials in contact with the water. These effects are important, particularly when the fuel container is also used for storage post-transport prior to reprocessing or disposal. The behaviour of a range of radiolytic systems has been studied. Plant behaviour has been simulated in numerous laboratory experiments: plant and experimental results have been linked by a computerised model describing the radiolysis mechanism and predicting the quantities and production rate of gaseous and corrosive species. This allows prediction of plant performance over a long time scale. The model is based on a well-accepted radiolysis mechanism supplemented with specific measurements made at the Harwell laboratory. Model capabilities include inert atmospheres, materials corrosion, variations in water and gas volumes or aqueous chemistry. The model has been applied to design stage radiolysis assessments of transport containers; information from operating plant has been interpreted to advise on design improvement, e.g. diminution of gas production using easily corroded scavengers to remove oxygen. Radiolysis in gas filled dry storage containers for spent nuclear fuel has been studied; corrosive product production (e.g. nitric acid), which is important for fuel cladding integrity has been assessed. The development and use of this computerised model is described with a current summary.     

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.     

Computational estimate of the corrosion resistance of fuel microelements with silicon carbide cladding in a water medium at supercritical pressure

A computational estimate of the corrosion resistance of a silicon carbide coating on spherical fuel microelements in a supercritical-pressure water medium at 350–700°C is made for the nominal operating regime of the core of a light-water nuclear reactor. A model of the thermal dissociation of water and dissolution of a silicon oxide film is used to estimate the corrosion resistance of silicon carbide. An expression describing the dependence of the corrosion depth of a silicon carbide layer on the temperature and pressure of the medium (concentration of the products of dissociation of water) and the operating time of the fuel microelements is obtained on the basis of the model developed. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 168–173, March, 2007. An erratum to this article is available at .     

Computational-experimental determination of the temperature of the components of a shipping container with spent fuel

The results of a determination of the expected temperatures of the structural components of the container with long-term dry storage and shipment of spent nuclear fuel from the AMB reactors at the Beloyarakaya nuclear power plant, under normal and accident conditions, are presented. The requirements for its thermalprotection characteristics are formulated on the basis of normative-technical documentation, and the basic assumptions and initial prerequisites, used to perform the calculations, are presented. The computational-experimental method is used together with the results of tests performed on a prototype model to confirm the possibility of ensuring the required temperatures of the basic components of the container and the fuel assembly. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 428–431, June, 2006.     

Possibilities of BREST reactors and transmutation fuel cycle under conditions implementing modern plans for the development of nuclear power

The possible dynamics of the development of BREST-1200 fast reactor capacities after 2030 on the basis of plutonium and other actinides accumulated in the spent fuel of thermal reactors is examined. It is shown that by 2100 the power BREST reactors could be 114–176 GW, and subsequently they will develop as a result of their own breeding of plutonium. Calculations have shown that the rate at which BREST reactors are put into operation can be doubled by using enriched uranium obtained from natural uranium and regenerated spent fuel from thermal reactors. It is shown that the development of fast reactors with a closed fuel cycle solves the problem of transmutation of long-lived high-level actinides and makes it possible to implement a transmutation fuel cycle in nuclear power. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 21–28, July, 2007.