Long-Lived-Radionuclide Transmutation Scenarios

The results of multigroup calculations of continuous irradiation of Np, Am, and Cm in VVÉR-1000, PHWR-880, Superphoenix-1200, BREST-1000, and ÉLYaU-800 reactors are used to compare transmutation efficiency. The sources of continuous replenishment for the transmuters were Np, Am, and Cm extracted after a 3-yr holding period from the VVÉR and Superphoenix spent fuel. It is shown that the most effective transmuter is a subcritical liquid-fuel ÉLYaU system with an average thermal-neutron flux in the blanket 2·10¹⁵ sec^–1·cm^–2. For solid-fuel reactors, the continuous-irradiation model makes it possible to describe approximately the multiple transmutation regime. In the foreseeable future, one-time transmutation of Np, Am, and Cm in a solid-fuel reactor followed by storage in a long-term storage facility is feasible. The results of different computational variants for such regimes show that for transmutation in 10 yr in PHWR the radiotoxicity of Np, Am, and Cm accumulated in long-term storage reaches an equilibrium in no longer than 100 yr.     

Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

In the frame of Partitioning and Transmutation (P&T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a parametric sensitivity study of both the fuel cycle and the electricity generating costs.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

Research reactor RBT-6

Conclusions Reactor RBT-6 is simple in construction and is easily accessible for conducting experiments. The values of neutron flux in it are high for small thermal power; this together with the large duration of continuous operation ensures the possibility of conducting a wide range of experimental investigationa. Such a reactor may be recommended as a research reactor for irradiation of samples of materials up to moderate flux values (10¹⁹–10²¹ neutrons/cm²) and for conducting experiments for studying the change in the properties of materials during irradiation, which are becoming increasingly more important. Estimates show that the number of used fuel assemblies of the SM-2 reactor are sufficient for the operation of several such reactors.If necessary, the number of experimental channels in the active zone of such a reactor can be increased by increasing the number of fuel assemblies and the thermal power. Beryllium can be used as the lateral reflector. This results in a decrease of the volume of the active zone and the thermal power of the reactor, but increases its cost.Translated from Atomnaya Énergiya, Vol. 43, No. 1, pp. 3–7, July, 1977.     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Development of a fuel element with low harmful neutron absorption for the SM high-flux research reactor

Information about the SM research reactor and its characteristic features and advantages over other research reactors is presented. The reasons for updating the reactor and the optimal method of solving the problem are indicated. The upgrade program preserves the essential structural features of the rector and allows for the insertion of additional irradiation channels in the fuel part of the core by removing some fuel elements. The reactivity loss arising in so doing is compensated by increasing the uranium content in the remaining fuel elements. A new type of fuel element based on materials with reduced harmful absorption of neutrons is being developed to improve further the technical and economic performance of the reactor. The design and the technology of the fuel element have been developed for three implementations, and experimental fuel elements for reactor tests have been fabricated. The fuel elements have been checked for adherence to the requirements. It has been shown that normal operation of the fuel elements is possible with heat flux density at the surface 9–12 MW/m², which meets the initial requirements.     

In-core fuel management strategy for the basket-fuel-assembly molten salt reactor

Molten salt reactor, with good economics and inherent reliability, is one of the six types of Generation IV candidate reactors. The Basket-Fuel-Assembly Molten Salt Reactor(BFAMSR) is a new concept design based on fuel assemblies composed of fuel pebbles made of TRISOcoated particles. Four refueling patterns, similar to the fuel management strategy for water reactors, are designed and analyzed for BFAMSR in terms of economy and security.The MCNPX is employed to calculate the parameters, such as the total duration time, cycle length, discharge burnup,total discharge quantity of ~(235)U, total discharge quantity of ~(239)Pu, neutron flux distribution and power distribution. The in–out loading pattern has the highest burnup and duration time, the worst neutron flux and power distribution and the lowest neutron leakage. The out–in pattern possesses the most uniform neutron flux distribution, the lowest burnup and total duration time, and the highest neutron leakage.The out–in partition alternate pattern has slightly higher burnup, longer total duration time and smaller neutron leakage than that of the out–in loading pattern at the cost of sacrificing some neutron flux distribution and power distribution. However, its alternative distribution of fuelelements cut down the refueling time. The low-leakage pattern is the second highest in burnup, and total duration time, and its neutron flux and power distributions are the second most uniform.     

Possibility of developing low-power reactors with a pebble-bed core

The possibilities of a two-section pebble-bed core with coolant delivered along the radial direction are examined for a water-cooled water-moderated 100–400 MW(t) reactor with 4 mm in diameter fuel pebbles. The construction of the two-section core includes two ring-shaped fuel layers (inner and outer) separated by a ring-shaped channel for coolant delivery. Water filters through the inner fuel layer from the periphery toward the center and through the outer layer from the center toward the periphery. The basic characteristics of such a reactor are determined. It is shown that the power density of such a core with 4 mm in diameter fuel pellets is two times higher than that in a VVÉR-1000 reactor, and there is a substantial margin with respect to the average specific heat flux. Thermohydraulic and neutron-physical calculations showed that two-section pebble-bed cores in low-power nuclear power reactors are promising.Translated from Atomnaya Énergiya, Vol. 97, No. 3, pp. 168–172, September, 2004.     

Fusion-fission hybrid design with analysis of direct enrichment and non-proliferation features (the SOLASE-H study)

The role of a fusion-fission hybrid in the context of a nuclear economy with and without reprocessing is examined. An inertial confinement fusion driver is assumed and a consistent set of reactor parameters are developed. The form of the driver is not critical, however, to the general concepts. The use of the hybrid as a fuel factory within a secured fuel production and reprocessing center is considered. Either the hybrid or a low power fission reactor can be used to mildly irradiate fuel prior to shipment to offsite reactors thereby rendering the fuel resistant to diversion. A simplified economic analysis indicates a hybrid providing fuel to 10 fission reactors of equal thermal power is insensitive to the recirculating power fraction provided reprocessing is permitted. If reprocessing is not allowed, the hybrid can be used to directly enrich light water reactor fuel bundles fabricated initially from fertile fuel (either ThO₂ or ²³⁸UO₂). A detailed neutronic analysis indicates such direct enrichment is feasible but the support ratio for ²³³U or ²³⁹Pu production is only 2, making such an approach highly sensitive to the hybrid cost. The hybrid would have to produce considerable net power for economic feasibility in this case. Inertial confinement fusion performance requirements for hybrid application are also examined and an integrated design, SOLASE-H, is described based upon the direct enrichment concept.     

Applicability of neutron noise technique for diagnostics of in-core components for heavy water reactors

Neutron flux signal is composed of a steady or mean component resulting from the flux produced by power operation of the reactor and a very small fluctuating component called ‘noise’ component. Analysis of neutron noise from suitably located sensors is a proven technique to monitor the in-core components of light water reactors (LWRs). However, the use of neutron noise has been rare, if any, for heavy water reactors (HWRs) as it was generally felt that the unfavourable transfer function characteristics of the reactors would limit its applicability. To assess the applicability of technique in pressurised heavy water reactors (PHWRs), experiments were carried out using in-core and out-of-core neutron sensors in a research reactor. This paper discusses the measurement details and results of the experiment. This paper concludes that the neutron noise technique can be effectively utilised for diagnostics/characterisation of the in-core components of heavy water reactors.     

Pu-breeding feasibility in irradiation channels of research reactors

The breeding potential in the irradiation channels of research reactors is of safeguards concern, because of lacking continuous supervision on the type of experiments in all the irradiation channels. Moreover, the irradiation time can be optimized in order to breed high quality weapon grade plutonium. With regard to the safeguards measures currently adopted, IAEA concentrates its efforts on those reactors whose thermal power is greater than 25 MWth, because it was calculated that a 25 MWth LEU-fuelled reactor produces not more than one Significant Quantity of Pu (8 kg)/year in its spent fuel and a HEU-fuelled reactor of this power would require an annual reload of not more than one Significant Quantity of U₂₃₅ (25 kg). In order to investigate whether it would be possible to determine an analogous power level threshold to estimate the clandestine plutonium production capability of different research reactors, the Monte Carlo method was used to determine the neutron flux in the irradiation channels and to calculate the plutonium breeding potential for three different reactor types: (1) a Triga Mark II with 250 kWth, representative for a small size research reactor; (2) a Material Test Reactor (MTR) with 5 MWth, representative for a medium size research reactor; (3) a High Flux Reactor (HFR) with 45 MWth, representative for a large size research reactor. It was observed that the most important factors for plutonium breeding are the neutron flux (to which reaction rates are proportional) and the available space to place irradiation samples. The breeding capability scales fairly well with the reactor power level and from about 10 MWth onwards the proliferation concern raises with increasing power level and available sample space.     

Studies on physics parameters of metal (U–Pu–Zr) fuelled FBR cores

For faster growth of nuclear power in India, it is essential to shift to the use of metal-fuels in fast breeder reactors (FBR), which gives a higher breeding ratio (BR) and lower doubling time (DT). Also, future commercialization of the FBR fuel cycle necessitates the use of metallic fuel along with the pyro-process recycling, which can be less costly than oxide fuel reprocessing. Two-dimensional diffusion calculations have been performed to investigate the various physics parameters of metal (U–Pu–Zr) fuelled FBR cores as a function of reactor parameters like reactor power, smear density, zirconium content in the fuel and the number of rows in radial blankets. A 1000 MWe fast reactor with U–Pu fuel (i.e. metal-fuel with no zirconium – which is a theoretical possibility now, due to the lack of irradiation experience) can attain a breeding ratio of 1.61 and a reactor fuel doubling time of 6.6 yrs. Two methods to reduce the sodium void reactivity, which is high and positive in metal-fuelled FBR cores, are suggested.     

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.     

BN-800 as a New Stage in the Development of Fast Sodium-Cooled Reactors

The role of fast reactors in a strategy for developing nuclear power in Russia because of the inevitable exhaustion of natural uranium deposits in the foreseeable future is discussed. The BN-800 reactor, which is under construction and incorporates unique solutions – greatly enhancing the safety of the reactor – to technical and constructional problems, is examined. Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely reactors in the world – water-moderated water-cooled reactors.Closing the BN-800 nuclear fuel cycle will make it possible to solve the problem of utilizing plutonium and actinides. This makes fast reactors safer for the environment.     

Assessment of criticality and burn up behavior of candu reactors with nuclear waste trans uranium fuel

Large quantities of nuclear waste plutonium and minor actinides (MAs) have been accumulated in the civilian light water reactors (LWRs) and CANDU reactors. These trans uranium (TRU) elements are all fissionable, and thus can be considered as fissile fuel materials in form of mixed fuel with thorium or nat-uranium in the latter. CANDU fuel compacts made of tristructural-isotropic (TRISO) type pellets would withstand very high burn ups without fuel change.As carbide fuels allow higher fissile material density than oxide fuels, following fuel compositions have been selected for investigations: ① 90% nat-UC + 10% TRUC, ② 70% nat-UC + 30% TRUC and ③ 50% nat-UC + 50% TRUC. Higher TRUC charge leads to longer power plant operation periods without fuel change. The behavior of the criticality k_∞ and the burn up values of the reactor have been pursued by full power operation for > ∼12 years. For these selected fuel compositions, the reactor criticality starts by k_∞ = 1.4443, 1.4872 and 1.5238, where corresponding reactor operation times and burn up values have been calculated as 2.8 years, 8 years and 12.5 years, and 62, 430 MW.D/MT, 176,000 and 280,000 MW.D/MT, with fuel consumption rates of ∼16, 5.68 and 3.57 g/MW.D respectively. These high burn ups would reduce the nuclear waste mass per unit energy output drastically. The study has show clearly that TRU in form of TRISO fuel pellets will provide sufficient criticality as well as reasonable burn up for CANDU reactors in order to justify their consideration as alternative fuel.     

Nuclear power development in market conditions with use of multi-purpose modular fast reactors SVBR-75/100

This article presents an innovative nuclear power technology, based on the use of modular type fast-neutron reactors SVBR-75/100 having heavy liquid-metal coolant, i.e. eutectic lead–bismuth alloy, which was mastered in Russia for the nuclear submarines’ reactors. Reactor SVBR-75/100 possesses inherent self-protection and passive safety properties that allow excluding of many safety systems necessary for traditional type reactors. Use of this nuclear power technology makes it possible to eliminate conflicting requirements among safety needs and economic factors, which is particularly found in traditional reactors, to increase considerably the investment attractiveness of nuclear power based on the use of fast-neutron reactors for the near future, when the cost of natural uranium is low and to assure development of nuclear power in market conditions. On the basis of the factory-fabricated “standard” reactor modules, it is possible to construct the nuclear power plants with different power and purposes. Without changing the design, it is possible for reactor SVBR-75/100 to use different kinds of fuel and operate in different fuel cycles with meeting the safety requirements.     

Application of candle burnup strategy for future nuclear energy utilization

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. If this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means about 40% of natural or depleted uranium can be utilized without either enrichment or reprocessing.

In the ideal nuclear energy utilization system, the radioactive toxicity in the environment should remain or decrease after the utilization. This requirement is very severe and difficult to be satisfied. It may take too much time for its realization. The CANDLE burnup may substitute this period. Though it is a once-through fuel cycle, the discharged fuel burnup is about ten times of the present value for light water reactors. The space necessary for final disposal can be drastically reduced. However, in order to realize such a high burnup of discharged fuels some innovative technologies should be developed. Either new material standing still for such a high burnup or intermediate recladding will be required. Especially new fuel development will take a lot of time. For the time being a small reactor with CANDLE burnup may be a good option for nuclear power generation. Even this kind of reactor requires some innovative technologies and a long period for their developments. For the first stage of CANDLE burnup the prismatic fuel high-temperature gas cooled reactor is preferable. Since the design of this reactor fits to the CANDLE burnup very well, only a little time is required for its research and development.     

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR／CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexi-bility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nu-clear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (-22.5%), increase the energy output (-41%), decrease the quantity of spent fuels to be disposed (-2/3) and lower the cost of nuclear poower, Because of the inherent flexibility of nuclearfuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modifica-tion of the reactor core structure and operation mode.It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.     

Application of Gas-Cooled Reactor Technology to the Fast Breeder Concept

Presently discernible limitations to the extent of low-cost uranium resources provide an obvious incentive for seeking better utilization of fissile material, which may well become dominating by the end of the century. The high degree of world interest in breeder reactors, which could produce more fissile material than they consume, is thus very understandable, for the substantial conversion of fertile into fissile material thereby offered would greatly increase the effective energy yield of the mined fuel. However, the large-scale use of fast breeders which would be necessary for substantial impact on resource conservation would also require that these reactors be entirely acceptable also in other ways. Particularly they would have to compete with other routes to nuclear power in areas of capital cost, maintainability, and siting flexibility. Such considerations have stimulated the evolution of the gas-cooled fast breeder and the contemplation of combinations of fast breeders and advanced high temperature converter reactors aimed at taking best advantage of the special merits of each type. Predominant influences here are the enhanced importance of breeding ratio to the effectiveness of such a combination, the special worth of U233 as a thermal reactor fuel and the value of the high temperature capability, both directly and as a factor broadening the options available to power plant design.