Prospects for Using Mixed Oxide Fuel in Fast Sodium-Cooled Reactors

When designing new fast reactors, it is desirable to increase as much as possible the breeding occurring in the core in order to ensure the minimum excess reactivity for burnup on the one hand and a closed fuel cycle without replenishment with external plutonium and without separating plutonium from uranium during chemical reprocessing of irradiated fuel on the other. The latter requirement greatly decreases the risk of plutonium proliferation in such a fuel cycle. This requires a core breeding ratio 1.05–1.08. Such values can be achieved by using technologically perfected and tested oxide fuel with its volume fraction in the core increased to 55–60%. The results of computational-theoretical studies on the selection and optimization of cores with high fuel fractions for BN-1600 and BN-800 reactors are presented in this article. It is shown that such cores can be built in principle.     

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.     

Study of the radiotoxicity of actinides recycling in boiling water reactors fuel

In this paper the production and destruction, as well as the radiotoxicity of plutonium and minor actinides (MA) obtained from the multi-recycling of boiling water reactors (BWR) fuel are analyzed. A BWR MOX fuel assembly, with uranium (from enrichment tails), plutonium and minor actinides is designed and studied using the HELIOS code. The actinides mass and the radiotoxicity of the spent fuel are compared with those of the once-through or direct cycle. Other type of fuel assembly is also analyzed: an assembly with enriched uranium and minor actinides; without plutonium. For this study, the fuel remains in the reactor for four cycles, where each cycle is 18 months length, with a discharge burnup of 48 MWd/kg. After this time, the fuel is placed in the spent fuel pool to be cooled during 5 years. Afterwards, the fuel is recycled for the next fuel cycle; 2 years are considered for recycle and fuel fabrication. Two recycles are taken into account in this study. Regarding radiotoxicity, results show that in the period from the spent fuel discharge until 1000 years, the highest reduction in the radiotoxicity related to the direct cycle is obtained with a fuel composed of MA and enriched uranium. However, in the period after few thousands of years, the lowest radiotoxicity is obtained using the fuel with plutonium and MA. The reduction in the radiotoxicity of the spent fuel after one or two recycling in a BWR is however very small for the studied MOX assemblies, reaching a maximum reduction factor of 2.     

Pyrochemical conversion of weapons plutonium into plutonium mononitride,monocarbide and mixed uranium-plutonium fuel for fast reactors

The main directions and results of research on pyrochemical reprocessing of weapons plutonium in fuel for fast reactors are presented. It is shown that this technology is economical and ecologically validated, compact, fire and explosion safe, especially for reprocessing in carbide-nitride as well as oxide fuel for fast reactors. It satisfies the principle of nonproliferation. For reprocessing weapons plutonium in oxide fuel with deep removal of ²⁴¹Am and Ga, a combined process which combines pyrochemical conversion of plutonium into oxide or nitride powder, and dissolution in acids and extraction of impurities. It is shown that the fuel kernels made from nitride, carbide, and oxide powers both from individual PuN, PuC_0.86, and PuO₂ powders as well as mixed plutonium compounds with uranium are fabricated by means of the conventional regime and provide the required density and content of gallium of <0.001 wt. %.     

Advanced mixed oxide fuel assemblies with higher plutonium contents for pressurized water reactors

Optimizing fuel cycle costs by increasing the final burnup leads to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in mixed oxide (MOX) fuel assemblies (FAs) reduces further the amount of plutonium which has to be disposed of in final storage. Increasing the final burnup requires higher initial enrichments of uranium fuel to be matched by an advanced design of MOX FAs with higher plutonium contents. The neutronic design of these MOX FAs has to consider the licensing status of nuclear power plants concerning the use of MOX fuel. The Siemens Nuclear Fuel Cycle Division, with more than 20 years' experience in the production of MOX fuel, has designed several advanced MOX FAs of different types (14 × 14 to 16 × 16) with fissile plutonium contents up to 4.60 w/o.     

Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10×10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10×10 lattice; in the second approach, an 11×11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11×11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10×10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly.     

Purex co-processing of spent LWR fuels: flow sheet

Purex co-processing of spent LWR fuel is investigated. In purex co-processing, uranium and plutonium in spent fuel are processed and recovered together as a single stream, while in standard purex reprocessing uranium and plutonium are obtained as separate streams. A two-step (co-decontamination and co-stripping) flow sheet for purex co-processing is devised; concentrations, recoveries and decontamination factors are calculated; and methods to co-convert uranium–plutonium nitrate to mixed oxide are reviewed. A closed nuclear fuel cycle in which at no point uranium and plutonium are separated from each other is reached.     

Increasing fuel burnup in VVER

The results of investigations of fuel burnup increase in VVER are presented. The influence of the costs of different technological stages, changes in the number of refuelings and run time, fuel enrichment and waste, and the consumption of natural uranium on increasing burnup is examined. An analysis taking account of the uncertainty of future prices is performed. The price for natural uranium up to 2020 is estimated using a model. The results presented in this article show that the cost reduction in the fuel component with an increase of VVER fuel burnup in an open fuel cycle can be 6%. __________ Translated from Atomnaya énergiya, Vol. 104, No. 3, pp. 137–141, March, 2008.     

Plutonium fuel and fuel elements for power reactors

Conclusions The use of plutonium in the fuel cycle during complex utilization of thermal and fast reactors in nuclear energetics permits solving the problem of ensuring nuclear fuel for a long period. Oxide uranium-plutonium fuel facilitates the development of technology of fast reactors and so far it is considered as the basic type of fuel. At the same time, oxide fuel cannot ensure the required rate of plutonium accumulation, in view of which the investigations of more efficient fuel and constructional materials become a pressing problem. The use of uranium-plutonium oxide fuel in thermal reactors requires improvements in the construction of fuel elements and organization of large-scale completely automatic production.Translated from Atomnaya Énergiya, Vol. 43, No. 5, pp. 412–417, November, 1977. Editors' Remarks. For the completeness of the discussion of the problem it is, of course, necessary to consider the possibility of using plutonium in fast and thermal reactors as done by the authors. However, it should be kept in mind that by its nuclear-physical parameters plutonium as a nuclear fuel is more suitable for use in fast reactors than in thermal reactors. The use of plutonium in thermal reactors can reduce the demands of natural uranium for the development of nuclear power in all by 10–15%, whereas its use in fast reactors reduces the demand for uranium by a factor of 10.All this indicates the feasibility of using plutonium only in fast reactors even if its accumulation is required over a certain period.     

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.     

Expanded 233U Breeding in the Heavy-Water Blanket of an Electronuclear System

Comparative investigations are performed of the neutron-physical characteristics and their variations during burnup for two types of fuel Th–U and U–Pu with heavy- and light-water lattices. Analysis of the results gives the basic parameters of the electronuclear system operating in a thorium–uranium fuel cycle. It is shown that an external neutron source (accelerator + target) makes it possible to switch to the thorium–uranium fuel cycle with expanded ²³³U breeding without using fissioning uranium or plutonium isotopes. There are grounds for believing that with a higher accelerator current it is possible to achieve a regime where the heavy-water blanket can be replenished only with thorium with no ²³³U.     

Fuel utilization in VVÉR-1000: Status and prospects

The status and prospects for increasing the fuel utilization efficiency of VVéR-1000 reactors are reviewed. It is shown that the main trends in the development of water moderated and cooled reactors are reflected in an improved design with a four-year fuel run, different variants of which are now being implemented in nuclear power plants operating in Russia, Ukraine, and Bulgaria: weakly neutron-absorbing materials are used in the fuel assemblies, part of the excess reactivity of the core is compensated by an absorber (gadolinium) which is integrated with the fuel, the fuel load is designed with a reduced radial neutron leakage, and the number of refuelings is increased. Promising directions for improving fuel utilization are noted: increasing the energy content of the fuel load, operating the reactor with reduced values of the parameters at the end of a run after the reactivity excess has been exhausted on burnup, and reusing (recycling) the uranium and plutonium contained in the spent fuel. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 139–146, March, 2007.     

Conversion of high enriched uranium in thorium-232-based oxide fuel for light and heavy water reactors: MOX-T fuel

This paper discusses the possibility of using military high enriched uranium and plutonium in thorium oxide fuel for light and heavy water reactors (LWRs and HWRs). It is shown that such a fuel has several important advantages: (i) ²³⁹Pu and other long-living actinides are generated in quantities which are at least 100 times less than in conventional fuel; (ii) neutron emission is lower by a factor of more than 100; (iii) ²³³U is generated and burnt (the conversion factor for LWRs is 0.64–0.68 and for HWRs about 0.88); (iv) thorium is utilized and the total available amount of nuclear fuel is increased. The problem of non-proliferation of fissile material is also discussed and it is shown that the supervision of such fuel does not differ essentially from the supervision of low enriched uranium fuel with plutonium generation.     

Use of Recovered Uranium and Plutonium in Thermal Reactors

A possible version of the VVER-1000 fuel cycle without separation of uranium and plutonium during reprocessing of spent fuel is examined. In this fuel cycle, the uranium-plutonium regenerate obtained, from which other actinides and fission products have been removed, is used after enriched natural uranium is added for preparing VVER fuel. The results of a calculation of the content of uranium and plutonium isotopes in the spent uranium-plutonium fuel after one and two recycles in VVER-1000 are presented. The main advantages of the fuel cycle are discussed: lower risk of plutonium proliferation, savings of natural uranium, and less spent fuel as compared with an open uranium fuel cycle. __________ Translated from Atomnaya Energiya, Vol. 99, No. 2, pp. 136–141, August 2005.     

Investigation of 85Kr Emission and Its Effect on the Radiation Conditions during the Preparation of BOR-60 Fuel for Recovery

The main stages of the BOR-60 closed fuel cycle, implemented on the experimental base at the Scientific-Research Institute of Nuclear Reactors, are examined. The ⁸⁵Kr emission at the stages of preparation of the spent BOR-60 fuel assemblies for recovery is determined experimentally. It is shown that the maximum ⁸⁵Kr emission as a result of destruction of fuel element cladding with oxide uranium fuel is 68%; its contribution to the irradiation dose to the public as a result of mechanical disassembly of the fuel elements in a single BOR-60 fuel assembly with 10% burnup and a 10-yr holding time does not exceed 1·10^–4% of the dose limit (1 mSv/yr).     

Possibility of operating GT-MHR with different fuels

The results of a computational investigation of the neutron-physical characteristics and parameters of GT-MHR for different fuels are presented: weapons-grade plutonium, reactor plutonium, low-enrichment uranium, mixed fuel based on uranium and weapons-grade plutonium, fuel based on high-enrichment uranium and thorium, and fuel with added americium and curium. Analysis has shown that GT-MHR can operate on different fuels; this is determined by the structural implementation incorporated in the design and the neutron-physical characteristics of the reactor. This feature makes it possible to use different fuels in this reactor without changing the basic structural components: fuel assemblies, size of the core and reflectors, and the number of reactivity compensation rods. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 68–72, January, 2007.     

Fast Reactor Operating on Uranium–Plutonium Oxide Fuel in a Self-Regulatable Regime

Mathematical simulation is used to show that it is possible to develop a fast reactor operating on uranium–plutonium oxide fuel (UO₂)_1–x (PuO₂) _x , the same for all fuel elements in the core, and with uranium carbide in breeding elements with heavy coolant (PbBi eutectic). A self-regulatable regime is obtained in the reactor. This enhances safety while minimizing control. Tailings uranium with 0.1% ²³⁵U and a mixture of plutonium isotopes, which is obtained from spent fuel, making it possible to conduct operation in an actinide-closed fuel cycle, is used in the fuel and uranium carbide. ²³⁸U is actually consumed in the reactor, but most fission products are produced from ²³⁹Pu.     

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once - through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.     

VVER nuclear fuel burnup with different absorbers

A computational study is performed of the fuel burnup in VVER-1000 using different absorbers in open and closed fuel cycles. It is shown that mixtures of plutonium isotopes (energy and others) can give the same effect as gadolinium, which is currently used. Fuel burnup increases. When neptunium, americium, and curium isotopes are used as a consumable absorber in a closed fuel cycle, the accompanying effect is elimination of long-lived α-emitting radionuclides which have accumulated in long-term repositories.     

Radiation Characteristics of Fuel and Wastes in the Uranium–Plutonium and Thorium–Uranium Fuel Cycle

The radiation characteristics of fuel cycles of various reactors – replacement candidates in the future nuclear power – are compared. Proceeding from the basic requirements (safety, fuel supply, and nonproliferation of fissioning materials), inherently safe fast reactors of the BREST type can be used as the basis for large-scale nuclear power. Thermal reactors, which can burn enriched uranium, thorium–uranium fuel, or mixed uranium–plutonium fuel with makeup with fissioning materials from fast reactors, will operate for a long time simultaneously with fast reactors in the future nuclear power. VVÉR-1000 and CANDU reactors are examined as representatives of thermal reactors; for each of these reactors the operation in various variants of the fuel cycle is simulated. It is shown that with respect to radiation characteristics of the fuel and wastes the thorium–uranium fuel cycle has no great advantages over the uranium–plutonium cycle.