Improving SFR economics through innovations from thermal design and analysis aspects

Achieving economic competitiveness as compared to LWRs and other Generation IV (Gen-IV) reactors is one of the major requirements to attract large-scale investment in commercial sodium cooled fast reactor (SFR) power plants. Advances in R&D for advanced SFR fuel and structural materials provide key long-term opportunities to improve SFR economics. In addition, other new opportunities are emerging to further improve SFR economics. This paper provides an overview on potential ideas from the perspective of thermal hydraulics to improve SFR economics. These include: (1) a new hybrid loop-pool reactor design to further optimize economics, safety, and reliability of SFRs with more flexibility, (2) a multiple-reheat and intercooling helium Brayton cycle to improve plant thermal efficiency and to reduce safety related overnight and operation costs, and (3) modern multi-physics thermal analysis methods to reduce analysis uncertainties and associated requirements for over-conservatism in reactor design. This paper reviews advances in all three areas and their potential beneficial impacts on SFR economics.     

小型氟盐冷却高温堆耦合布雷顿循环系统分析与研究

为满足小型氟盐冷却高温堆（FHR）能量转换需求,开发与之匹配的高效、紧凑、无水冷却动力转换系统,本文对比了超临界二氧化碳（SCO₂）、空气、氩气（Ar）、氮气（N₂）、氙气（Xe）5种气体工质在不同布雷顿循环构型中的热电转换效率、?效率、?损失分布。研究发现,SCO₂布雷顿循环相比其它工质循环具有最高的热电转换效率和?效率,且结构更为紧凑,易于小型化和模块化,与小型氟盐冷却高温堆耦合更具优势;进而对SCO₂布雷顿循环进行构型优化,得出匹配小型氟盐冷却高温堆的最佳循环构型方式,构成固有安全模块化小型氟盐冷却高温堆热电转换系统,为西部能源利用提供新研究思路。      

Medium temperature carbon dioxide gas turbine reactor

A carbon dioxide (CO₂) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 °C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 °C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO₂; and consideration of variation in CO₂ specific heat at constant pressure, C_p, with pressure and temperature into cycle configuration. Lowering temperature to 650 °C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 °C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO₂ have been proven during extensive operation in AGRs. In the previous study, the CO₂ cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO₂ gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors.     

Supercritical Carbon Dioxide Power Conversion System Applied to Sodium-cold Fast Reactor

Because of the high efficiency, compactness and avoiding sodium water reaction, the supercritical carbon dioxide (SCO₂) Brayton cycle is an ideal power conversion system for sodium-cooled fast reactors. In this paper, the 1 200 MWe Sodium-cooled Fast Reactor was used as the heat source of the system, and the temperature and heat load of the sodium loop were used as the operating boundary of the circulation system. The system performance and key equipment performance of different supercritical carbon dioxide Brayton cycles were compared. The coupling between the inter-stage cooling and recompression cycle and the characteristics of the heat source of the sodium-cooled reactor is the best, and the cycle efficiency is the highest (40.7%). Furthermore, the influence of different operating parameters on the efficiency of the inter-stage cooling and recompression cycle was studied, and the sensitivity of the efficiency of the circulation system to each of the key influencing factors was given. It is found that the efficiency of the circulation system is the most sensitive to the cold-end parameters, followed by the split ratio and turbine inlet parameters, and the weakest to the main compressor inter-stage pressure ratio.     

应用于钠冷快堆的超临界二氧化碳动力转换系统研究

超临界二氧化碳（SCO₂）布雷顿循环由于高效、紧凑和可避免钠水反应等特性而成为钠冷快堆的理想动力转换系统。本文以1 200 MWe大型池式钠冷快堆为系统热源,钠回路温度及热负荷为循环系统运行边界,对比研究了不同SCO₂布雷顿循环系统性能和关键设备性能的变化规律。研究发现,级间冷却再压缩循环与钠冷快堆热源特性匹配性最佳,且循环效率最高（40.7%）。进而研究了不同运行参数对级间冷却再压缩循环效率的影响规律,给出了循环系统效率对各关键影响因素的敏感度,发现循环系统效率对冷端参数的敏感度最强,其次为分流比和透平入口参数,对主压缩机级间压比的敏感度最弱。     

Investigation of alternative layouts for the supercritical carbon dioxide Brayton cycle for a sodium-cooled fast reactor

Analyses of supercritical carbon dioxide (S-CO₂) Brayton cycle performance have largely settled on the recompression supercritical cycle (or Feher cycle) incorporating a flow split between the main compressor downstream of heat rejection, a recompressing compressor providing direct compression without heat rejection, and high and low temperature recuperators to raise the effectiveness of recuperation and the cycle efficiency. Alternative cycle layouts have been previously examined by Angelino (Politecnico, Milan), by MIT (Dostal, Hejzlar, and Driscoll), and possibly others but not for sodium-cooled fast reactors (SFRs) operating at relatively low core outlet temperature. Thus, the present authors could not be sure that the recompression cycle is an optimal arrangement for application to the SFR. To ensure that an advantageous alternative layout has not been overlooked, several alternative cycle layouts have been investigated for a S-CO₂ Brayton cycle coupled to the Advanced Burner Test Reactor (ABTR) SFR preconceptual design having a 510 °C core outlet temperature and a 470 °C turbine inlet temperature to determine if they provide any benefit in cycle performance (e.g., enhanced cycle efficiency). No such benefits were identified, consistent with the previous examinations, such that attention was devoted to optimizing the recompression supercritical cycle. The effects of optimizing the cycle minimum temperature and pressure are investigated including minimum temperatures and/or pressures below the critical values. It is found that improvements in the cycle efficiency of 1% or greater relative to previous analyses which arbitrarily fixed the minimum temperature and pressure can be realized through an optimal choice of the combination of the minimum cycle temperature and pressure (e.g., for a fixed minimum temperature there is an optimal minimum pressure). However, this leads to a requirement for a larger cooler for heat rejection which may impact the tradeoff between efficiency and capital cost. In addition, for minimum temperatures below the critical temperature, a lower heat sink temperature is required the availability of which is dependent upon the climate at the specific plant site.     

Sodium fast reactor evaluation: Core materials

In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.     

Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO₂ (S-CO₂) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480 °C, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO₂ turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO₂ cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m³ and 7240 m³ for the steam turbine system and S-CO₂ turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO₂ than in H₂O. Therefore, the S-CO₂ turbine system is recommended to the fusion reactor system than the steam turbine system.     

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-temperature reactor, with reactor-coolant exit temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-temperature gas-cooled reactors.     

Potential improvements of supercritical recompression CO2 Brayton cycle by mixing other gases for power conversion system of a SFR

A sodium-cooled fast reactor (SFR) is one of the strongest candidates for the next generation nuclear reactor. However, the conventional design of a SFR concept with an indirect Rankine cycle is subjected to a possible sodium-water reaction. To prevent any hazards from sodium-water reaction, a SFR with the Brayton cycle using Supercritical Carbon dioxide (S-CO₂) as the working fluid can be an alternative approach to improve the current SFR design. However, the S-CO₂ Brayton cycle is more sensitive to the critical point of working fluids than other Brayton cycles. This is because compressor work is significantly decreased slightly above the critical point due to high density of CO₂ near the boundary between the supercritical state and the subcritical state. For this reason, the minimum temperature and pressure of cycle are just above the CO₂ critical point. In other words, the critical point acts as a limitation of the lowest operating condition of the cycle. In general, lowering the rejection temperature of a thermodynamic cycle can increase the efficiency. Therefore, changing the critical point of CO₂ can result in an improvement of the total cycle efficiency with the same cycle layout. A small amount of other gases can be added in order to change the critical point of CO₂. The direction and range of the critical point variation of CO₂ depends on the mixed component and its amount. Several gases that show chemical stability with sodium within the interested range of cycle operating condition were chosen as candidates for the mixture; CO₂ was mixed with N₂, O₂, He, and Ar. To evaluate the effect of shifting the critical point and changes in the properties of the S-CO₂ Brayton cycle, a supercritical Brayton cycle analysis code with a properties program, which has the most accurate mixture models, was developed. The CO₂-He binary mixture shows the highest cycle efficiency increase. Unlike the CO₂-He binary mixture, the cycle efficiencies of CO₂-Ar, CO₂-N₂, and CO₂-O₂ binary mixtures decreased compared to the pure S-CO₂ cycle. It was found that the increment of critical pressure led to a decrease in cycle operating pressure ratio which resulted in a negative effect on total cycle efficiency. In addition, the effects from changed minimum operating condition and property variations of multi-component working fluid changed the recuperated heat in the cycle which was closely related to the cycle performances.     

Potential application of Rankine and He-Brayton cycles to sodium fast reactors

Traditionally all the demos and/or prototypes of the sodium fast reactor (SFR) technology with power output, have used a steam sub-critical Rankine cycle. Sustainability requirement of Gen. IV reactors recommends exploring alternate power cycle configurations capable of reaching high thermal efficiency.By adopting the anticipated working parameters of next SFRs, this paper investigates the potential of some Rankine and He-Brayton layouts to reach thermal efficiencies as high as feasible, so that they could become alternates for SFR reactor balance of plant. The assessment has encompassed from sub-critical to super-critical Rankine cycles and combined cycles based on He-Brayton gas cycles of different complexity coupled to Organic Rankine Cycles. The sub-critical Rankine configuration reached at thermal efficiency higher than 43%, which has been shown to be a superior performance than any of the He-Brayton configurations analyzed. By adopting a super-critical Rankine arrangement, thermal efficiency would increase less than 1.5%. In short, according to the present study a sub-critical layout seems to be the most promising configuration for all those upcoming prototypes to be operated in the short term (10-15 years). The potential of super-critical CO₂-Brayton cycles should be explored for future SFRs to be deployed in a longer run.     

Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor

The high temperature gas-cooled reactor (HTGR) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of HTGR gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. Furthermore they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analyzed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbomachines which are required for the three optimized Brayton cycles are aerodynamically analyzed and compared and their fundamental characteristics are obtained. Helium turbocompressor has lower stage pressure ratio and more stage number than those for nitrogen and air machines, while helium and nitrogen turbocompressors have shorter blade length than that for air machine.     

氦氙混合气体冷却反应堆单通道程序开发

闭式布雷顿循环在大功率空间核电源方面具有广阔的应用前景。采用氦氙混合气体作循环工质可提高布雷顿循环的性能,且由于其良好的换热能力,可直接用作反应堆的冷却剂。在概念设计阶段,为适应频繁改动反应堆设计的需要,开发了氦氙混合气体冷却反应堆单通道分析程序。其中氦氙混合气体的物性参数采用理论方法预测。程序采用FORTRAN 90语言编写,并集成画图功能。程序可用于计算环形流道和圆管流道,通过对同一个算例的单通道计算结果与FLUENT计算结果或实验数据进行对比,初步验证了程序的正确性。     

Metallic fuels for advanced reactors

In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数（反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数）入手,对5个循环方案进行比较分析。综合考虑各种工程因素,上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的,其中下位循环在高温气冷堆范围是亚临界参数循环,在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化,且反应堆入口温度在反应堆压力壳材料允许的范围内,具有足够的安全性。     

Thermal striping limits for components of sodium cooled fast spectrum reactors

With the objective of establishing thermal striping limits for future sodium cooled fast spectrum reactors (SFR), a fracture mechanics-based method employing ‘σ-d approach’ recommended in RCC-MR: Appendix A16 has been followed. Towards this, an idealized geometry, thermal fluctuations in the form of constant power spectral density and pessimistic material data were considered and temperature and thermal stresses are computed taking in to account frequency-dependent thermal attenuation on the structural wall. The effect of attenuation is found to be significant. The limits are derived at various potential locations in SFRs, which are also subjected to creep-fatigue damage due to major cycles caused by startup, shutdown, power failures and pump trips, etc. The maximum range of temperature fluctuations can be as high as 70 K where there is practically no accumulated creep-fatigue damage and 45 K is acceptable where the creep-fatigue is significant (0.9). These limits are found to be consistent with the broad limits extrapolated from the failure experiences of international SFRs and sodium facilities. Pool hydraulic computations carried out to identify and quantify the thermal striping zones confirmed that the proposed limits can be respected with good margins for SFRs.     

Economic benefits of advanced materials in nuclear power systems

A key obstacle to the commercial deployment of advanced fast reactors is the capital cost. There is a perception of higher capital cost for fast reactor systems than advanced light water reactors. However, cost estimates come with a large uncertainty since far fewer fast reactors have been built than light water reactor facilities. Furthermore, the large variability of industrial cost estimates complicates accurate comparisons. Reductions in capital cost can result from design simplifications, new technologies that allow reduced capital costs, and simulation techniques that help optimize system design. It is plausible that improved materials will provide opportunities for both simplified design and reduced capital cost. Advanced materials may also allow improved safety and longer component lifetimes. This work examines the potential impact of advanced materials on the capital investment cost of fast nuclear reactors.     

Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies.In this paper, two fast reactor systems are discussed—the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries.First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW_e) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems.We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating minor actinides. The tight pin lattice SFRs (P/D = 1.2) appears to have a better neutron economy than wide channel LFRs (P/D = 1.6), resulting in larger BOL actinide inventories and lower burn-up swings for LFRs. The reactivity burn-up swing of an LFR self-breeder employing BeO moderator pins could be limited to 1.3$ in 1 year. For a 600 MW_e LFR burner, LWR-to-burner support ratio was about two for (U, TRU)O₂-fuelled system, while it increased to approximately 2.8 when (Th, TRU)O₂ fuel was employed. The corresponding figures for an SFR were somewhat lower. The calculations revealed that LFRs have an advantage over SFRs in coping with the investigated severe accident initiators (ULOF, ULOHS, TLOP). The reason is better natural circulation behavior of LFR systems and the much higher boiling temperature of lead. A ULOF accident in an LFR only leads to a 220 K coolant outlet temperature increase whereas for an SFR the coolant may boil. Regarding the economics, the LFR seems to have an advantage since it does not require an intermediate coolant circuit. However, it was also proposed to avoid an intermediate coolant circuit in an SFR by using a supercritical CO₂ Brayton cycle. But in an LFR, the reduced concern about air and water ingress may decrease its cost further.     

Verification of NETFLOW code using plant data of sodium cooled reactor and facility

The plant simulation code NETFLOW on PC applicable to the liquid-metal cooled reactors has been developed on the basis of the models developed for single-phase and two-phase light water flow systems. The functions of this code have been verified by individual tests for light water flow systems and a sodium flow system. In order to apply this code to a sodium cooled fast reactor, several extra functions were verified using the plant data obtained using 50 MW steam generators and the Monju fast breeder reactor. Finally, the turbine trip transient of the Monju was simulated and the result was compared with the measured plant data. Good agreements were obtained in these verifications. As a result of the present study, the code can be applied as an education tool for students.     

Characterization of a sodium-cooled fast reactor in an MHR–SFR synergy for TRU transmutation

In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50–65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR–SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core.