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30 Years of experience in operating the BN600 sodium-cooled fast reactor
Abstract
Experience in operating the BN-600 sodium-cooled fast reactor during its nominal service life as well as its service life
extension period, an additional 15 years, is described. Information is presented on the performance indicators which were
achieved and deviations from the normal operating regime which occurred when the reactor was first started up. The degree
to which they affect the safety and technical-economic performance of the facility is evaluated. It is concluded on the basis
of an analysis of the BN-600 operating experience that sodium-cooled fast reactors have now been mastered commercially and
that their prospects for further development are good.
... Proven technologies and experience acquired in the design, commissioning and operation of BN-600 and BN-800 reactors should have broad use in the BN-1200 design to the greatest extent possible [18,19]; Testing and validation of improvements in the reactor safety, economic competitiveness and effectiveness of fuel management incorporating innovative technologies through the appropriate RD&D activities using existing and newly developed research facilities; Optimization of the infrastructure requirements can be achieved through the selection of the appropriate value of the BN-1200 electrical power rating 18 and may involve unification of requirements for the NPP siting and unification of the electric generators and electric components used for the plant connection to the grid; Transportation of the NPP components to the construction site by railroad. ...
... Proven technologies and experience acquired in the design, commissioning and operation of BN-600 and BN-800 reactors should have broad use in the BN-1200 design to the greatest extent possible [18,19]; Testing and validation of improvements in the reactor safety, economic competitiveness and effectiveness of fuel management incorporating innovative technologies through the appropriate RD&D activities using existing and newly developed research facilities; Optimization of the infrastructure requirements can be achieved through the selection of the appropriate value of the BN-1200 electrical power rating 18 and may involve unification of requirements for the NPP siting and unification of the electric generators and electric components used for the plant connection to the grid; Transportation of the NPP components to the construction site by railroad. ...
... [20]. 18 Installed electrical power rating of BN-1200 was defined as equal to the electrical power rating of the advanced Russian water cooled reactors (AES-2006 or VVER-TOI). Based on the optimization studies performed in the Russian Federation for both sodium cooled fast reactors and water cooled reactors, the concept of 1200 MW(e) monobloc was adopted for the BN-1200 power plant. ...
An example of a limited scope INPRO sustainability assessment of an innovative nuclear energy system used the BN-1200 fast reactor as a case study. The INPRO assessment was performed at the full-depth criteria level and helped to identify actions for sustainable long term deployment of sodium cooled fast reactors. The publication describes the application of the INPRO sustainability assessment method to the innovative nuclear energy system based on BN-1200 fast reactors in the areas of economics and safety of nuclear reactors. The case study is intended to verify readiness of the updated INPRO Methodology for assessment of the sodium cooled fast reactors and to develop recommendations for further improvements to the INPRO assessment method.
... The goal of the ASTRID reactor was to demonstrate safety and superior operability properties of Gen-IV SFR reactors [9]. Experience gained from the operation of experimental prototype and commercial [10][11][12][13] size SFRs were used to support the design of the ASTRID reactor. The examined core includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. ...
This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRID-like design. The core is characterised by a highly heterogeneous configuration and was likely to challenge the accuracy of the Serpent-DYN3D sequence. It includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. Consequently, the influence of generation and correction methods of various homogenised, few-group cross-sections (XS) on the accuracy of the full-core nodal diffusion DYN3D calculations is presented. An attempt to compare the approximate time effort spent on models preparation against the accuracy of the result is made. Results are compared to reference full-core Serpent MC (Monte Carlo) solutions. Initially, XS data was generated in Serpent using traditional methods (2D single assemblies and 2D super-cells). Full core calculations and MC simulations offered a moderate agreement. Therefore, XS generation with 2D fuel-reflector models and 3D single assembly models was verified. Super-homogenisation (SPH) factors for XS correction were applied. In conclusion, the performed work suggests that Serpent-DYN3D sequence could be used for the analysis of highly heterogeneous SFR designs similar to the studied ASTRID-like, with an only small penalty on the accuracy of the core reactivity and radial power distribution prediction. However, the XS generation route would need to include the correction with SPH factors and generation of XS with various MC models, for different core regions. At a certain point, there are diminishing returns to using more complex XS generation methods, as the accuracy of full-core deterministic calculations improves only slightly, while the time effort required increases significantly.
... В расчетных ячейках БР встречаются ячейки с наличием внешнего стального чехла и центральной ячейки с гомогенно представленными в ней теплоносителем и какии миилибо конструкциями (концевиками твэлов в реакторах типа БН [10,11] или техх нологическими трубами, например, как на рис. 2). ...
The importance of calculation of radiation fields inside in-reactor cavities is associated with the necessity to simulate the emergency modes in fast breeder reactors (FBR), as well as reactor states with different coolant levels in special dedicated channels of passive feedback devices in lead-cooled fast reactors (LFR) of BREST type or in sodium cavities in sodium-cooled fast reactors (SFR).
The Last Flight (LF) method (Bell and Glesston 1974, Davison 1960, DOORS3.2 1988, Mynatt et. al. 1969, Rhoades and Childs 1988, Rhoades and Sipmson 1997, SCALE 2009, Voloschenko et. al. 2012), or the method of the unscattered component is widely known and is commonly used in computer codes based on the method of spherical harmonics for obtaining solution in a gas medium at a certain distance from the calculated volume domain (DORT (Rhoades and Childs 1988), TORT (Rhoades and Sipmson 1997) and others (SCALE 2009)). The practice of its application (DOORS3.2 1988) demonstrated that acceptable results are achievable at considerable distances from the surface separating dense and gas media (more than two meters). Obtaining high-quality solution is not guaranteed for cavities within the calculation area.
In addition, it is desirable to implement the cavities calculation methodology within the framework of the approximations used in reactor calculations introducing certain specific features. In particular isotropy of the neutron flux density and the necessity of forced introduction of a “conditional” calculation cell on the boundary surface of the void cavity are assumed in the diffusion approximation. If the LF method is oriented on the connection of the source point with the detection point, then it is necessary to determine in the calculation of neutron field in the cavities the neutrons escaping the surface area of the source and neutrons reaching a certain surface area of the cavity. In order to solve the problem, the authors suggested using the approximate solution presented in the paper.
Thus, an algorithm for calculation of in-reactor cavities using the diffusion approximation was developed and implemented by the authors.
... Nonetheless, as shown in Table III, this plant has operated well for a long time duration. 21,22 III.C. Other Fission Power Plants ...
As fusion researchers look toward the future, there have been discussions on what the plant operation goals should be for a demonstration fusion power plant (DEMO). The U.S. Research Needs Workshop (ReNeW) in 2009 stated that power producers (the companies owning power plants) could not expect an ultimate fusion power plant availability of 80% or more if a DEMO reactor cannot demonstrate a 50% or higher availability. The ReNeW panel also stated that achieving 50% availability with a DEMO plant would be a huge accomplishment. Other recent DEMO design studies have given goals for plant availability as well. This technical note presents historical plant availability values of new technology fission power plants to compare fission achievements with the suggested goals for fusion DEMO plant designs. Demonstration fission plants that met or exceeded 40% average annual availability were generally considered to be successful. These data help to support the goal values that have been put forward in various studies.
... Being the main component of nitride nuclear fuel, uranium mononitride (UN) exhibits a higher metal atom density (e.g., the densities of U 0.8 Pu 0.2 O 2 and U 0.8 Pu 0.2 N at room temperature are 9.75 g cm 3 and 13.5 g cm 3 , respectively [5]) and higher thermal conductivity (20.6 W m À1 K À1 for UN [6] versus 3.5 W m À1 K À1 for UO 2 e at 1000 K [7]) compared to the conventional oxide fuel. The high linear heat generation rate (LHGR) in the fuel elements of fast reactors (up to 42 kW/m in the lead-cooled fast reactor [8] and up to 54 kW/m in sodium-cooled fast reactors [9]) results in an increased heat load on nuclear fuel as compared to thermal reactors (corresponding average LHGR 15e20 kW/m in PWR [10]). Hence, good thermal conductivity (meaning higher achievable LHGR) and larger density (meaning a smaller size of the reactor core and a larger breeding ratio) are points in favour of using the nitride fuel. ...
It is demonstrated that using the mathematical formalism of the interaction parameters for multicomponent metallic solution, it is possible to predict mass transfer in a system consisting of two dissimilar metals separated by low-melting one containing non-metallic impurities. The calculation of the interaction parameters in three-and four-component systems was carried out using the equations of the coordination-cluster model. Comparison between theory and data reported in the literature for corrosion in sodium loop led to the conclusion about the most probable mechanism of the influence of the oxygen impurity content on the corrosion rate of iron under non-isothermal conditions.
Dense uranium mononitride (UN) pellets with controlled microstructures and tailored grain size from large-grained to a few microns are synthesized by spark plasma sintering (SPS) combined with high energy ball milling. The impacts of the sintering conditions on fuel microstructure, grain size, physical density, and phase behavior are systematically investigated, and the thermal-mechanical properties and oxidation behavior of the SPS densified UN pellets are characterized. Higher sintering temperatures and longer ball milling durations and thus finer starting UN powders promote sintering and densification, and dense UN pellets above 95% theoretical density can be achieved by SPS at 1873 K for 10 mins. UN phase purity is maintained in the SPS-densified pellets sintered at a lower temperature and short duration. A phase heterogeneity with secondary UO2 or uranium sesquinitride (U2N3) occurs for the UN pellets sintered at higher temperatures using finer UN powders. The hardness and fracture toughness of the SPS-densified UN pellets increase with smaller grain sizes and higher densities to 7.9 GPa and 3.5 MPa m1/2, respectively. Both small (1-2 μm) and large grain-sized (30-50 µm) UN pellets exhibit good thermal conductivity. Dynamic oxidation testing by a thermogravimetric analyzer in air shows that the onset temperature for oxidation varies with microstructure and phase heterogeneity of the SPS densified UN pellets. Particularly, the smaller-grained (micron-sized) UN pellets containing uranium oxides and U2N3 display lower weight gain and significantly-reduced oxidation kinetics, and full oxidation completes at a temperature above 1173 K when tested with a ramp rate of 10 K/min.
This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRIDlike design. The core is characterised by a highly heterogeneous configuration and was likely to challenge the accuracy of the Serpent-DYN3D sequence. It includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. Consequently, the influence of generation and correction methods of various homogenised, few-group crosssections (XS) on the accuracy of the full-core nodal diffusion DYN3D calculations is presented. An attempt to compare the approximate time effort spent on models preparation against the accuracy of the result is made. Results are compared to reference full-core Serpent MC (Monte Carlo) solutions. Initially, XS data was generated in Serpent using traditional methods (2D single assemblies and 2D super-cells). Full core calculations and MC simulations offered a moderate agreement. Therefore, XS generation with 2D fuel-reflector models and 3D single assembly models was verified. Super-homogenisation (SPH) factors for XS correction were applied. In conclusion, the performed work suggests that Serpent-DYN3D sequence could be used for the analysis of highly heterogeneous SFR designs similar to the studied ASTRID-like, with an only small penalty on the accuracy of the core reactivity and radial power distribution prediction. However, the XS generation route would need to include the correction with SPH factors and generation of XS with various MC models, for different core regions. At a certain point, there are diminishing returns to using more complex XS generation methods, as the accuracy of full-core deterministic calculations improves only slightly, while the time effort required increases significantly.
The design of the BN-1200 sodium-cooled fast reactor is analyzed from the standpoint of its compliance with the criteria fitting the promising Gen-IV reactor facilities. The basic technical solutions adopted in the BN-1200 design making it possible to meet the requirements of consistent development of non-proliferation and safety, cost-effectiveness, and physical protection are described. The BN-1200 concept is evaluated on the basis of 26 criteria developed as part of the Generation IV international forum for the Gen-IV sodium-cooled fast reactor. The compliance of the safety requirements used in the development of the BN-1200 design with the design criteria for the Gen-IV sodium-cooled fast reactor developed as part of the Generation IV international forum is analyzed.
Achieving nuclear disarmament, stopping nuclear proliferation, and preventing nuclear terrorism are among the most critical challenges facing the world today. Unmaking the Bomb proposes a new approach to reaching these long-held goals. Rather than considering them as separate issues, the authors – physicists and experts on nuclear security – argue that all three of these goals can be understood and realized together if we focus on the production, stockpiling, and disposal of plutonium and highly enriched uranium – the fissile materials that are the key ingredients used to make nuclear weapons. The authors describe the history, production, national stockpiles, and current military and civilian uses of plutonium and highly enriched uranium, and propose policies aimed at reducing and eventually eliminating these fissile materials worldwide. These include an end to the production of highly enriched uranium and plutonium for weapons, an end to their use as reactor fuels, and the verified elimination of all national stockpiles. © 2014 Massachusetts Institute of Technology. All rights reserved.
After description of the uranium / plutonium fuel cycle results are reported for various isotopic compositions of reactor-grade plutonium for usability in hypothetical nuclear explosive devices. It is shown that it is mainly thermal analyses, besides neutron physics analyses, which indicate the limits of concentration of the Pu-238 plutonium isotope. Above these limits, such hypothetical nuclear explosive devices are not feasible technically. In the light of this finding, future proliferation-proof fuel cycles are proposed which make use of recent methods of actinide transmutation. The author is honorary professor at the Karlsruhe Institute of Technology (KIT) and emeritus director of the former Institute for Neutron Physics and Reactor Engineering of the former Karlsruhe Research Center. For many years he was a member and, for some time, chairman of the German Advisory Committee on Reactor Safeguards (RSK). He is a member of the European Nuclear Society (ENS) and a member and fellow of the American Nuclear Society (ANS).
In order to meet the steadily increasing worldwide energy demand, nuclear power is expected to continue playing a key role in electricity production. Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years' nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it – after the initial loading – to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor-years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback, viz. there is generally a positive reactivity effect when sodium coolant is removed from the core. Furthermore, this so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed improvements address both neutronics and thermal-hydraulics aspects. Furthermore, emphasis has been placed on not only the beginning-of-life (BOL) state of the core, but also on the beginning of closed equilibrium fuel cycle (BEC) state. An important context for the current thesis is the 7th European Framework Program's Collaborative Project for a European Sodium Fast Reactor (CP-ESFR), the reference 3600 MWth ESFR core being the starting point for the conducted research. The principally employed computational tools belong to the so-called FAST code system, viz. the fast-reactor neutronics code ERANOS, the fuel cycle simulating procedure EQL3D, the spatial kinetics code PARCS and the system thermal-hydraulics code TRACE. The research has been carried out in essentially three successive phases. The first phase has involved achieving a clearer understanding of the principal phenomena contributing to the SFR void effect. Decomposition and analysis of sodium void reactivity have been carried out, while considering different fuel cycle states for the core. Furthermore, the spatial distribution of void reactivity importance, in both axial and radial directions, is investigated. For the reactivity decomposition, two methods – based on neutron balance considerations and on perturbation theory, respectively – have been applied. The sodium void reactivity of the reference ESFR core has been, accordingly, decomposed reaction-wise, cross-section-wise, isotope-wise and energy-group-wise. Effectively, the neutron balance based method allows an in-depth understanding of the "consequences" of sodium voidage, while the perturbation theory based method provides a complementary understanding of the "causes". The second phase of the research has addressed optimization of the reference ESFR core design from the neutronics viewpoint. Four options oriented towards either the leakage component or the spectral effect have been considered in detail, viz. introducing an upper sodium plenum and boron layer, varying the core height-to-diameter (H/D) ratio, introducing moderator pins into the fuel assemblies, and modifying the initially loaded plutonium content. The sensitivity of the principal safety and performance parameters, viz. void reactivity, Doppler constant, nominal reactivity and breeding gain, has been evaluated with respect to each of the options. Finally, two synthesis core concepts – representing different selected combinations of the optimization options – have been proposed. The third and last phase has been to test the proposed optimized designs in terms of the dynamic core response to a representative ULOF accident, in which sodium boiling occurs. Such a hypothetical transient is of prime importance for SFR safety demonstration, since it may lead to a severe accident situation, where both cladding and fuel melt. As compared to the response of the reference ESFR core, certain improvements are indeed achieved with the static neutronics optimization carried out. However, these are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting. Thermal-hydraulic optimization has thus been considered necessary to: 1) prevent sodium flow blockage in the fuel channels and 2) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. Following implementation of appropriate thermal-hydraulics related design changes, one arrives at a final configuration of the SFR core in which – for the selected accident scenario – a new "steady state" involving stable sodium boiling is shown to be achievable, with melting of neither cladding nor fuel. Such satisfactory behavior has been confirmed not only for the beginning-of-life core state, but also for the equilibrium closed fuel cycle.
The article examines the possibility of using different fuels to achieve acceptable plutonium breeding performance in fast reactors with a fuel load meeting the requirements. The moderate and high growth scenarios will require efficient breeder reactors with high core power-density and the possibility of producing plutonium in the core and screens with breeding ratio 1.4-1.5. A high breeding ratio will make it possible to decrease the fraction of fast reactors with thermal and intermediate neutron spectra. The amount of plutonium in the core was 3.4 tons in the variants examined. Because of the different fuel density, the content of the initial material in all cases is different. In the case of metal, nitride, and oxide fuel, 67.6, 63.1, and 38.9 tons of fuel are present in the reactor, respectively. The plutonium production with nitride and oxide fuel is lower than for metal fuel, which, in turn, results in a higher rate of reactivity loss with burn up.
The Generation IV International Forum (GIF) has among its main goals the excellence in safety and reliability of the proposed innovative nuclear systems. The development of computational tools that are able to assist the design and safety analysis of these innovative reactor concepts is crucial. In this line, the JRC–IET is developing a static and dynamic integrated safety analysis platform with the objective to perform an integrated core and safety analysis of nuclear reactor systems. The first application of this platform has been made in the framework of the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) that is also part of EURATOM contribution to GIF. Two core designs have been currently proposed for the 3600 MWth sodium-cooled reactor concept based on oxide and carbide fuel respectively.Using the integrated safety analysis platform, static calculation on neutronics (incl. reactivity coefficients) and thermal–hydraulic characteristics for both oxide and carbide reference cores have been conducted. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the tools and the models applied in the study together with the relevant results for the oxide and carbide core.
New aspects of the development of a future nuclear power system based on the advanced technologies of a closed nuclear fuel
cycle with fast-neutron reactors are discussed on the basis of an analysis of systems problems pertaining to present-day nuclear
power. The systems requirements ensuring adequate fuel for nuclear power with any installed capacity and maximum use of natural
uranium and thorium brought into the fuel cycle are formulated. Sodium-cooled fast reactors, which thus far possess the highest
level of technological readiness for commercialization, are given a special role in the formation of a new technological platform
for large-scale nuclear power of the 21st century.
The data are presented on the operation experience of the BN-600 reactor with the sodium coolant in the primary and secondary circuits. High reliability of the equipment and sodium circuit systems, operating during 23 years, is shown. Violations of the reactor normal operation were effectively fixed and localized by the NPP safety systems. The unique experience of service and maintenance of the sodium equipment is obtained.
In the paper are presented the main results of the BN-600 power unit operation during the last 16 years. Operation of the reactor core, main equipment, fuel charging and recharging systems, steam generators, sodium and steam-water circuits and electric equipment is described. Disturbances of normal operation and introduced improvements and modernization are described. It is shown that the BN-600 power unit by its technical and economical performances is in the number of 50% of the best NPP in the world.
The various types of reactor systems which OKBM designed for different purposes over the 60 years of its existence are reviewed:
commercial uranium-graphite and heavy-water reactors, reactors and steam-generating systems for naval and for civilian (icebreakers)
ships, fast reactors, low-and medium-capacity reactor systems for regional power generation, and high-temperature gas-cooled
reactors. The results of work performed by OKBM on the development of the cores of propulsion reactors and alternative fuel
for VVR-1000 reactors are also presented.
Data on the experience in operating the BN-600 reactor with sodium coolant in the first and second loops are presented. It is shown that the equipment and the systems in the sodium loops, which have operated for more than 23 years, are highly realiable. The average installed capacity utilization factor for this period reached about 74%. The losses due to rupture of the heat-transfer pipes in the steam-generator modules and the sodium leaks from the loops were about 0.3%. Disruptions of normal operation are detected reliably and contained by the safety systems present in the unit. Unique experience in performing maintenance and repair work on the sodium equipment has been gained on the BN-600 reactor.
Assessment of the operating efficiency of a power-generating unit with a BN-600 fast-neutron reactor at the Beloyarskaya nuclear power plant over 25 years of operation
- N N Oshkanov
- O A Potapov
- P P Govorov
- NN Oshkanov
Operating experience and direction of development of BN-600 reactor core,” in: Nuclear Power Technologies with Fast Reactors
- V B A A Vasiliev
- O V Rogov
- M R Mishin
- Farakshin
Twenty eight years of operation. Power-generating unit with a BN-600 fast reactor: main technical-economic indicators
- N N Oshkanov
- P P Govorov
- A V Kuznetsov
Fast reactor operating experience gained in Russia: analysis of anomalies and abnormal operation cases
- Y M Ashurko
- R P Baklushin
- Y I Zagorelko
The main results of Beloyarskaya NPP unit 3 operation
- N N Oshkanov
- M V Bakanov
- O A Potapov
Problems of building BN-800 and the possibility of developing advanced fast reactors
- V I Kostin
- B A Vasiliev
- VI Kostin
Experience in operating power-generating unit with a BN-600 reactor
- O M Saraev
- N N Oshkanov
- V V Vylomov
- OM Saraev
Operating experience and direction of development of BN-600 reactor core
- B A Vasiliev
- V A Rogov
- O V Mishin
- M R Farakshin