Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.     

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium–plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.     

Plutonium and minor actinides management in the nuclear fuel cycle: assessing and controlling the inventory

The mastering of the plutonium and minor actinides inventory in the French Nuclear Cycle is based on a progressive approach from the present status, dealing with the partial reprocessing of spent fuels and the recycling of Pu in the MOX assemblies loaded in the 20 licensed PWRs. This strategy keeps the door open long-term, for example, for the eventual multirecycling of excess Pu in dedicated new assemblies, such as APA or CORAIL in order to stabilise the Pu inventory in the fuel cycle or allow its utilization in new types of fast reactors. Presently, in the framework of 1991 law, scenario studies relying on present and/or innovative technologies are carried out in order to transmute both Pu and minor actinides, thus minimising the quantities to be for disposal. To cite this article: H. Mouney, C. R. Physique 3 (2002) 773–782.     

关于压水堆产武器级钚的模拟计算

防止核扩散是国际社会共同努力的目标,其中武器级核材料的防扩散是重中之重.钚是反应堆的副产品,如果不计较经济效益,利用铀为核燃料的反应堆都可以生产武器级钚.本文基于日本Takahama-3压水堆建立了五个模型,并进行中子和燃耗计算,得到两种燃料棒产武器级钚的条件、燃料棒轴向的燃耗分布、组件内燃料棒燃耗的变化区间和全堆芯燃料棒径向燃耗分布.基于上述模型和计算数据给出压水堆堆芯内含有武器级钚的准确位置和UO_2燃料棒中武器级钚的产量.这种低燃耗的乏燃料给国际核不扩散带来了巨大风险,国际社会应该加强对此类乏燃料的监管.     

Redox state of plutonium in irradiated mixed oxide fuels

Nowadays, MOX fuels are used in about 20 nuclear power plants around the world. After irradiation, plutonium co-exists with uranium oxide. Due to the redox sensitive nature of UO₂ other plutonium oxides than PuO₂ potentially present in the fuel may interact with the matrix. The aim of this study is to determine which plutonium species are present in heterogeneous and homogeneous MOX. The results provided by X-ray Absorption Near Edge Spectroscopy (XANES) for non-irradiated as well as irradiated (center and periphery) homogeneous MOX fuel were published earlier and are completed by Extended X-ray Fine Structure (EXAFS) analysis in this work. The EXAFS signals have been extracted using the Athena code and the analyses were carried using Excure98 as performed earlier for an analogous element. EXAFS shows that plutonium redox state remains tetravalent in the solid solution and that the minor fraction of trivalent Pu must be below 10%. Independently, the study of homogeneous MOX was also approached by thermodynamics of solid solution of (U,Pu)O₂. Such solid solutions were modeled using the Gibbs Energy Minimisation (GEM)-Selektor code (developed at LES, NES, PSI) supported by the literature data on such solid solutions. A comparative study was performed showing which plutonium oxides in their respective mole fractions are more likely to occur in (U,Pu)O₂. In the modeling, these oxides were set as ideal and non-ideal solid solutions, as well as separate pure phases. Pu exists mainly as PuO₂ in the case of separate phases, but can exist under its reduced forms, PuO_1.61 and PuO_1.5 in minor fraction i.e. ~15% in ideal solid solution (unlikely) and ~10% in non-ideal solid solution (likely) and at temperature around 1300 K. This combined thermodynamic and EXAFS studies confirm independently the results obtained so far by Pu XANES for the same MOX samples.     

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.     

On the equilibrium isotopic composition of the thorium–uranium–plutonium fuel cycle

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium–uranium–plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.     

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be ∼13 kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.      

Closed Thorium-Uranium-Plutonium Cycle Concept in Nuclear Power Engineering

The use of 232Th instead of 238U as a fertile isotope, 233U instead of 239Pu as the main fissile isotope, heavy water instead of light water as a coolant, and its dilution with light water in the VVER reactor campaign make possible self-enrichment of fuel with fissile isotopes, including the time upon achieving the balanced isotopic abundance ratio of actinides, and also provide conditions for closing the Th-U-Pu fuel cycle. This allows increasing the fuel lifetime by around two orders of magnitude, making it much easier to handle radioactive waste, reducing the nuclear hazard of PWE reactors, and providing a technological barrier to prevent the distribution of fissile materials and nuclear technologies.     

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium–Uranium–Plutonium Fuel Cycle

It is shown for a closed thorium–uranium–plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain ~54 kg of fission products, ~0.8 kg of thorium, ~0.10 kg of uranium isotopes, ~0.005 kg of plutonium isotopes, ~0.002 kg of neptunium, and “trace” amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.     

不同燃耗计算模型对商用压水堆乏燃料组件核素成分的影响分析

燃耗计算精度对提高乏燃料贮存效率有着重要影响,在应用燃耗信用制时,燃耗计算得到的核素成分偏差决定了乏燃料贮存的临界安全裕量。不同燃耗计算模型所得到的核素成分偏差各不相同,为提高燃耗计算精度,提出了一种装载不同燃料富集度的多组件燃耗计算模型,并使用不同燃耗计算模型分别对TMI-1反应堆NJ07OG组件中的6个样本进行了计算、对比和分析。结果表明,相比其他模型,考虑不同燃料富集度的多组件模型得到的²³⁵U、²³⁸U和²³⁹Pu等核素平均相对偏差更接近于零且6个样本的相对偏差分布更为平均。     

HPR1000堆芯装载50%MOX组件的燃料管理方案北大核心CSCD

针对HPR1000压水堆堆芯,开展了应用MOX(混合氧化物燃料)组件的燃料管理方案初步研究。对MOX燃料组件进行设计,研究了MOX燃料成分及燃料棒在组件内的布置。在此基础上,开展了1/4堆芯年换料、18个月长周期换料,并装载50%MOX组件这两种燃料管理方案研究。通过与UO2堆芯的对比,分析了装载50%MOX组件堆芯的核特性。分析结果表明,两种50%堆芯装载MOX组件的燃料管理方案,其堆芯主要物理参数均满足核设计准则要求。     

Determination of Separated Plutonium,Americium and Curium in Environmental Samples by Sequential Extraction Techniques

A simple, reliable and practical radiochemical method for sequential isolation and determination of plutonium, americium and curium in a wide variety of environmental samples including soils, river sediments and water was developed. The isotopes determined are: ²³⁸Pu, ^239,240Pu, ²⁴¹Am, ²⁴²Cm and ²⁴⁴Cm. The methods involve leaching of soil or sedimental samples with concentrated nitric acid using a pressure digestion technique. Subsequent concentration and separation of nuclides of interest from major matrix elements and other interfering alpha-emitters are carried out by coprecipitation with ferric hydroxide and solvent extractions. Sources suitable for alpha-spectrometry are prepared by electrodeposition from acidic ammonium chloride solution. ²⁴²Pu and ²⁴³Am are used as tracer isotopes of plutonium and americium-curium elements, respectively. Some results of analysis of soils, sediments and water are given. The alpha peaks from ²³⁸Pu, ^239,240Pu and ²⁴²Pu as well as ²⁴¹Am, ²⁴⁴Cm, ²⁴²Cm and ²⁴³Am are well resolved. The entire analytical procedures for plutonium, americium and curium are completed in less than sixteen hours.     

Variants of closing the nuclear fuel cycle

Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.     

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing ²³³U from ²³²Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.     

Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the ²³⁶U isotope is compensated by re-enrichment in the ²³⁵U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the ²³²U isotope are obtained.     

Plutonium isotopes concentration in the ground level air and rain samples from Kraków

The aim of the present work was to look at the background levels of plutonium in air and rain samples. Two sets of Petryanov type filters through which ca. 0.5 Mm³ each of air had passed and two large samples of collected rain (170 L and 182 L) were analysed for the content of plutonium alpha-emitters. In the article the radiochemical procedure applied to these samples is described. Obtained Pu activities in air were 0.5±0.1 nBqm^?3 and 0.7±0.1 nBqm^?3 for²³⁸Pu and 3.1±0.2 nBqm^?3 and 8.3±0.6 nBqm^?3 for^239+240Pu. Activities of rain samples were 7.5±0.7 mBqm^?3 and 14.3±0.8 mBqm^?3 for^239±240Pu and 0.4±0.1 mBqm^?3 and 2.1±0.2 mBqm^?3 for²³⁸Pu, respectively, for the two samples. The most interesting finding is the largely different²³⁸Pu to^239+240Pu activity ratios (from 0.03±0.01 to 0.18±0.01) measured in the samples. This result needs conformation, but as now it suggests different origins of plutonium isotopes in incoming to Kraków air.     

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction ^232+233+234U and ²³¹Pa are formulated. (1) The fuel cycle would shift from fissile ²³⁵U to ²³³U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of ²³¹Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.     

Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the ²³¹Pa–²³²U–²³³U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of ²³²U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.     

On feasibility of a closed nuclear power fuel cycle with minimum radioactivity

Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem.