Spent fuel reprocessing: A vital link in Indian nuclear power program

The success of the three stage Indian nuclear energy program is inter-linked with the establishment of an efficient closed fuel cycle approach with recycling of both fissile and fertile components of the spent fuel to appropriate reactor systems. The Indian reprocessing journey was started way back in 1964 with the commissioning of a plant based on PUREX technology to reprocess aluminum clad natural uranium spent fuel from the research reactor CIRUS. After achieving the basic skills, a power reactor reprocessing facility was built to reprocess spent fuel from power reactors. Adequate design and operating experience was gained from these two plants for mastering the reprocessing technology. The first plant, being the maiden venture, based on indigenous technology had to undergo many modifications during its operation and finally needed refurbishment for continued operation. Decommissioning and decontamination of this plant was carried out meticulously to allow unrestricted access to the cells for fresh installation. A third plant was built for power reactor spent fuel reprocessing to serve as a design standard for future plants with the involvement of industry. Over the years, spent fuel reprocessing based on PUREX technology has reached a matured status and can be safely deployed to meet the additional reprocessing requirements to cater to the expanding nuclear energy program. Side by side with the developments in the spent natural uranium fuel reprocessing, irradiated thoria reprocessing is also perused to develop THOREX into a robust process. The additional challenges in this domain are being addressed to evolve appropriate technological solutions. Advancements in the field of science and technology are being absorbed to meet the challenges of higher recovery combined with reduced exposure and environmental discharges.     

Combined vitrification of chloride and phosphate wastes by pyroelectrochemical reprocessing of nuclear fuel

The results of an investigation of the combined vitrification of chloride and phosphate wastes, which were obtained as a result of experimental reprocessing of irradiated nuclear fuel by the pyroelectrochemical method, are presented. The optimal composition of the glass matrix is determined, the characteristics of the model and real samples obtained are presented, and the hydrolytic stability and the thermal stability of the samples are studied. The proposed method of combined vitrification makes it possible to salvage more than 1.5 times more wastes per unit mass of the glass block than separate vitrification. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 438–440, December, 2006.     

后处理厂1A柱有机相着火事故源项估算

我国对于后处理工业的需求随着核电事业的迅猛发展变得愈发强烈,为了满足后处理工业安全发展必不可少的核应急需求,为核应急工况下后处理厂的核应急响应与决策支持提供依据。针对后处理厂1A柱有机相着火事故这一基准事故,结合实际工艺流程及监测手段,选取了核应急工况下的可获得参数(有机溶剂泄漏质量等)作为输入,在有机相燃烧速率经验公式基础上,结合后处理的工艺特点,引入少量修正建立了后处理厂1A柱有机相着火事故源项估算模型,并使用FORTRAN编程语言开发了相应软件。数值验证结果表明,该估算模型可以满足后处理厂1A柱有机相着火事故的核应急需求。     

On nuclear power intrinsically protected against long-lived actinide wastes,weapons proliferation and heavy accidents

The “Mobile fuel reactor (MFR) concept” [Slessarev, I., Stukalov,V., Subbotin, S., 1984. Problems of development of fast reactors self-fuel-provision without fuel reprocessing. Atomkernenergie, 45, N1; Slessarev, I., et al., 2004. WISE: a new fuel cycle concept based on a mobile fuel reactor. Nuclear Technol., 146 (N3), 230–243] and re-organization of the fuel cycle may allow to build the “actinide waste free” nuclear power (NP) and to improve radically its proliferation resistance. These goals can be achieved on the base of an “intrinsic approach” respecting long-lived wastes, safety and non-proliferation aspects simultaneously. It implies elimination of feed fuel enrichment, irradiated fuel reprocessing as well as the need for a long-lived, radiotoxic actinide repository. The key concept is: the application of “mobile” (liquid or gaseous) reactor fuels allowing for a simple mixing of fresh and already irradiated fuel components without “forced” FP separation.     

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.     

Recycling of uranium from spent RBMK fuel in the nuclear fuel cycle

It is shown that there is promise in using the uranium product obtained by reprocessing spent nuclear fuel from RBMK reactors as a non-initial fuel source for thermal reactors. A technical path for spent nuclear fuel from RBMK reactors is proposed: radiochemical reprocessing and obtaining oxides of recycled uranium. Oxides of the category RBMK-poor are packed and then stored in a near-surface storage facility; oxides of the category RBMK-rich are fluoridated, and UF₆ is fed into separation production for additional enrichment to the required content of ²³⁵U. Additional advantages of recycled RBMK uranium as a source of non-initial ²³⁵U are the low content of ²³²U and the relatively low activity of spent fuel, which simplifies its reprocessing.     

Ecological problems in the development of nuclear power

Conclusions The initial and middle stages of the nuclear fuel cycle, i.e., mining and reprocessing of ore, uranium enrichment, production of fuel elements, and the normal operation of a nuclear power plant, do not cause any serious danger to the environment. Comparisons show that the negative-effect coal-fired HEP is much greater.The probability for accidents involving the emission of a large quantity of radionuclides in modern nuclear power plants equipped with tested safety systems is significantly lower than the accident probability in other areas of industry. This conclusion is valid, however, if safety requirements, starting with the nuclear power plant, are satisfiedunscrupulously, if the strictest technological discipline, making sure that all the components have sufficient reliability, is followed, and if constant efforts are made to train personnel.It is as yet impossible to evaluate quantitatively the environmental effects of reprocessing plants.Czechoslovakian Technical University, Prague. Translated from Atomnaya Énergiya, Vol. 49, No. 6, pp. 352–357, December, 1980.     

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.     

Development of a real-time detection strategy for process monitoring during nuclear fuel reprocessing using the UREX+3a method

Reprocessing nuclear fuel is becoming more viable in the United States due to the anticipated increase in construction of nuclear power plants, the growing stockpile of existing used nuclear fuel, and a public desire to reduce the amount of this fuel. A new reprocessing facility will likely have state of the art controls and monitoring methods to safeguard special nuclear materials, as well as to provide real-time monitoring for process control. The focus of this research was to create a proof of concept to enable the development of a detection strategy that uses well established gamma and neutron measurement methods to characterize samples from the Uranium Extraction Plus 3a (UREX+3a) reprocessing method using a variety of detector types and measurement times.A facility that implemented real-time gamma detection equipment could improve product quality control and provide additional benefits, such as waste volume reduction. In addition to the spectral analyses, it was determined by Monte Carlo N Particle (MCNP) simulations that there is no noticeable self-shielding for internal pipe diameters less than 5.08 cm, indicating that no self-shielding correction factors are needed. Further, it was determined that High Purity Germanium (HPGe) N-type detectors have the high gamma ray energy resolution and neutron damage resistance that would be required in a reprocessing facility. Finally, the gamma ray spectra for the measured samples were simulated using MCNP and then the model was extended to predict the responses from an actual reprocessing scenario from UREX+3a applied to fuel that had a decay time of 3 years. The 3-year decayed fuel was more representative of commercially reprocessed fuel than the acquired UREX+3a samples. It was determined that the 3-year decayed fuel is easier to apply real-time process monitoring due to an increased number of short lived detectable isotopes. This research found that real-time gamma ray detection for process monitoring would be beneficial to a reprocessing facility and that commercially available detectors may be adequate for the neutron environment.     

Decladding of Irradiated Fuel Elements

The results of investigations of the leading operation of a nonaqueous technology for reprocessing fuel elements from nuclear reactors — dissolution of fuel claddings in a zinc-based melt — are presented. Data obtained in experiments on simulators and samples of irradiated fuel elements in standard BOR-60 and SM-2 packages with different burnup and holding time are presented. In the experiments, the metallic melt was separated from the fuel by filtering through a mesh and regenerated by vacuum distillation for reuse. The uranium and plutonium extraction was 99.99%. The behavior of individual radionuclides is described. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 273–276, October, 2005.     

Handling Irradiated RBMK-1000 and VVÉR-1000 Fuel during the Growth of Nuclear Power

On-site storage facilities, consisting of ponds with water, for irradiated RBMK-1000 fuel are now close to being filled. To continue operating nuclear power plants with RBMK reactors, it is necessary to select one possible method for handling irradiated fuel.A variant of long-term storage followed by reprocessing is examined and considerations are presented for future use of reprocessed irradiated RBMK and VVÉR fuel as fuel for an initial load for naturally-safe fast reactor. Important points in handling irradiated RBMK-1000 fuel include economic assessments and requirements for a strategy for development of nuclear power in Russia based on closure of the nuclear fuel cycle with radiation-equivalent burial of wastes and utilization of accumulated plutonium for fast reactors. 3 figures, 2 tables, 8 references.     

Recovery of Uranium and Plutonium from Spent Fuel Elements of Nuclear Reactors

The results of investigations of the preliminary removal of the products of radioactive decomposition from irradiated nuclear fuel to obtain uranium and plutonium which are suitable for reuse in fuel fabrication are presented. Nitrate-alkali melts are used for the operation. The experiments are performed on simulators and irradiated samples of BOR-60 fuel in remote-controlled hot boxes. The coefficients of removal of fission products are presented. A technological scheme, which will shorten the fuel cycle, for purifying hot nuclear fuel is recommended. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 387–392, November, 2005.     

Investigation of the incorporation of spent alkali chloride melt in ceramic

Investigations to validate a method of immobilizing high-level wastes from pyroelectrochemical reprocessing of spent nuclear fuel in mineral-like crystal matrices have been conducted. Methods for obtaining and the microstructure, elemental and phase compositions, and chemical stability of a ceramic matrix based on the mineral kosnarite, intended to be used for immobilizing wastes of alkali-metal chloride melts, are studied using inactive model materials, x-ray phase analysis, scanning electron microscopy, x-ray microspectrometry, and leach tests. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 178–182, March, 2007.     

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.     

Distribution of Fission Products in Irradiated Vibrationally Compacted Oxide Fuel

To substantiate the safety of temporary storage and shipment, development of a technology and optimization of conditions for reprocessing irradiated nuclear fuel, it necessary to follow carefully, during materials science investigations, the distribution of diverse fission products in the fuel, the local concentration and chemical state of the fission products, and the influence of the fission products on the change in the state of the fuel components and claddings. In the present paper, the results of an investigation by electron microprobe analysis and scanning electron microscopy of the distribution of fission products in vibrationally compacted fuel after irradiation in fast reactors are presented. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 380–387, November, 2005.     

Nuclide separation from spent nuclear fuels by using tertiary pyridine resin

The newly nuclide separation system from spent nuclear fuels is proposed. The proposed separation system consists of recovery of nuclear fuel elements, separation of trivalent minor actinide from lanthanide, and separation of some fission products such as strontium. This separation system is based on the chromatographic technique using the tertiary pyridine resin. Separation experiments using mixed oxide fuel highly irradiated in fast reactor “Joyo” were carried out. The recovery of plutonium, the separation of minor actinide from fission products including lanthanides, and the separation of americium and curium were achieved. The recovery or removal of platinum group elements and technetium was also investigated, and the removal of these elements prior to the main reprocessing process has been proposed.     

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.     

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.     

Reduction recrystallization of uranium oxides in molybdate melts

The presence of traces of water dissolved in a sodium molybdate melt and the interaction of this water with a metal surface (NP-2 nickel) in contact with the melt are confirmed. It is shown experimentally that U₃O₈ can be reduced to UO₂ in sodium molybdate melt by the hydrogen that is released during such an interaction. A technological process of reprocessing spent oxide nuclear fuel using the phenomenon studied is described. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 290–292, May, 2007.     

Reprocessing free nuclear fuel production via fusion fission hybrids

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th²³²–U²³³ conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO₂ matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U²³⁵ fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.