Sodium cooled small fast long-life reactor “4S”

CRIEPI and Toshiba Corp. have been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. A conceptual design of 4S (Super-Safe, Small and Simple) reactor is proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void reactivity are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core lifetime is more than 10 years. The 4S reactor is a metallic fueled sodium cooled fast reactor. A target of an electrical output is 10–50 MW. A remarkable feature of 4S is that its reactivity is not controlled by neutron absorber rods but by neutron reflectors to cope with a long core lifetime and a negative coolant void reactivity.

This study includes a design consideration of 4S. Design discussions are mainly focused on various core designs to meet above requirements. A tall core active height is adopted to gain long core lifetime. An averaged fuel burn-up is tried to be increased up to 100 GWd/ton from a point of economic view. The reference 4S designs are 10 MWe (30 years core lifetime) and 50 MWe (10 years core lifetime).     

On a fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel

A fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel is given. The present cermet fuel consists of an oxide solid solution of Th and minor actinides and Mo-inert matrix. It has been proposed as a high-performance device that can enhance minor actinide incineration in a fast reactor cycle. It is used in an independent small sub-cycle, whereby dedicated cycle technologies are adopted. Two-step reprocessing process was proposed for the present cermet fuel; it consists of a pre-removal of Mo-inert matrix and an actinide recovery. A preliminary test for the pre-removal of Mo-inert matrix was carried out using a surrogate cermet fuel. Burnup characteristics of a fast reactor core loaded with the cermet fuel were investigated by using neutronic calculation codes. It was revealed that a heterogeneous composition of Mo-inert matrix in inner and outer cores may lead to an effective transmutation of minor actinides and a flattened power density. It was concluded that the present cermet fuel was potentially promising as a high-performance incineration device of minor actinides for fast reactors.     

Effects of fuel type on the safety characteristics of a sodium cooled fast reactor. Part II: Simulation results

The objective of this investigation is to examine the impact of the fuel type on the inherent safety characteristics of Liquid Metal Fast Reactors (LMFRs). To perform this study, the responses to various transient conditions are examined for metallic, oxide and nitride cores of a baseline LMFR. GE-Hitachi’s Super Power Reactor Innovative Small Module (S-PRISM) was chosen as the baseline LMFR. In Part I of this paper, the background on S-PRISM’s metal and oxide cores are described and the redesign of a new nitride fueled S-PRISM core were introduced. Reactivity feedback and power profile data necessary for transient simulations with RELAP5-3D/ATHENA (RELAP5-3D, 2009) code are also presented and discussed. In this Part II of our paper, we present the results of accident simulations and a comparison between the metal, oxide and nitride cores based on their performance during the selected accident scenarios. Loss of Flow, Loss of Heat Sink, Loss of Power and inadvertent control rod withdrawal accidents were simulated for each core at beginning (BOC), middle (MOC) and end of a fuel cycle (EOC). The simulations were stopped at the initiation of melting of fuel or cladding. The results showed that in most of the transients the metal core came closer to its melting temperature while the strong reactivity feedbacks of the oxide and nitride cores limited their fuel temperature increases. Overall, the oxide and nitride cores had similar performance with respect to their inherent safety characteristics.     

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd₂O₃. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.     

Design study of PbBi- and NaK-cooled small deep sea fast reactors

The liquid lead-bismuth eutectic (PbBi) has good compatibility with water, which is different from sodium. It is expected that the PbBi could be used as a coolant of the deep sea fast reactor (DSFR). Physics analysis of the PbBi-cooled small reactor cores with and without inner control rods performed using the computer program of a neutronics code system (SRAC95) shows that PbBi is suitable for the coolant of small reactors as well as NaK.     

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(BB) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/~(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PBB) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.     

Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors

The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under progress in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been also carried out. Crucial development issues have been found out for each FR system concept and their development plans for the key technologies are summarized as the roadmap. As a result, it has been confirmed that the sodium-cooled FR concept is highly suited to the development targets and R&D issues are related enhancing the economy with certain perspectives for realization. A flexible and robust development program for the FR cycle system will be proposed taking account of the characteristics for each FR concept until the end of the Phase II study.     

A CFD simulation process for fast reactor fuel assemblies

A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-? and SST (Menter) k-ω were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.     

Nuclear burning wave in fast reactor with mixed Th-U fuel

The possibility of creating a self-sustained regime of a running nuclear burning wave in the critical fast reactor with the mixed Th-U fuel is demonstrated. The calculations were performed in the deterministic approach based on solving the non-stationary multi-group diffusion equation of neutron transport together with the set of equations of the fuel component burn-up and the nuclear kinetics of precursor nuclei of delayed neutrons. The presence of the constructional material Fe and the coolant (the Pb-Bi eutectic) in the reactor composition is taken into account. The calculation results of the space-time evolution of neutron flux and fuel component concentrations are presented for different values of the Th-U ratio in the fuel. The calculations show the remarkable stability of the nuclear burning wave regime against neutron flux distortions in the reactor, which is a result of the negative feedback on reactivity inherent to this regime. This is one of the most important features of the reactor of this type, which ensures its intrinsic safety.     

Core and fuel cycle for an advanced sodium-cooled fast reactor

The development of BN-1200 is based on the greatest possible use of tested and scientifically validated and developed technical solutions implemented in BN-350, -600, and the BN-800 design as well as new technical solutions that increase facility cost-effectiveness and safety. The BN-1200 design must permit the reactor to operate with different cores, including with denser fuel. The main fuel variant considered is oxide fuel and for the nearest term nitride fuel, for which the production technology involves the same steps as the oxide technology. The main approaches for choosing the parameters of the BN-1200 core as well as the results of computational studies are presented.     

Comparison of MCNPX-C90 and TRIPOLI-4-D for fuel depletion calculations of a Gas-cooled Fast Reactor

The Gas-cooled Fast Reactor is one of the reactor concepts selected by the Generation IV International Forum for the next generation of innovative nuclear energy systems. Several fuel design concepts are being investigated. Burnup depletion of mixed fuel of uranium and plutonium, cooled with gas in a fast neutron energy spectrum must be simulated. Various codes are being developed and/or adapted to improve the quality of the results, and also to reduce the computing time required for the simulations.     

深燃耗球床式气冷快堆研究

气冷快堆是未来发展的第四代先进核能系统候选堆型之一,它可以满足核能的可持续性、安全可靠性和经济性要求.从反应堆物理和热工水力学的角度出发,设计了热功率300 MW的球床式气冷快堆,选择了碳化物燃料作为气冷快堆的燃料.用耦合燃耗计算程序COUPLE2.0模拟得到了深燃耗气冷快堆的铀燃料循环的平衡态.平衡态研究结果表明基于深燃耗的300 MW球床式气冷快堆可以提高铀资源的利用率同时降低乏燃料中的次锕系核素的含量.当燃料球直径为6 cm,燃料区的直径为5.5 cm,燃料占燃料区的体积的70%,燃料形式为UC,其中235U的初始富集度为12%时,燃料球通过堆芯的时间可以达到12 600 d,重金属燃耗深度为164.38 GWd/t,总的铀资源的利用率可以达到为28.03%.     

A preliminary approach to the ALFRED reactor control strategy

In this paper, a preliminary approach to the definition of a suitable control strategy for the Advanced Lead Fast Reactor European Demonstrator (ALFRED), developed within the European 7th Framework Program, has been undertaken. The Generation IV reactors offer new challenges for what concerns the nuclear power plant control since several constraints both on primary and secondary loops have to be faced, differently from the conventional Light Water Reactors. A simulator of the ALFRED plant has been developed in a previous work (Ponciroli et al., 2014) with the main purpose of studying the system free dynamics and stability features in a control-oriented perspective. Based on the outcomes of these investigations, in the present work, the possibility of adopting decentralized control schemes has been investigated. Accordingly, Single Input Single Output control laws have been applied directly to the selected couples of input–output variables, which have been identified first on the basis of the preliminary plant dynamics analyses, and then confirmed by the indications of the Relative Gain Array method. Afterwards, two different control schemes have been studied depending on the number of available inputs, and then implemented and compared in order to evaluate the effect of each control action on the associated potential control strategy effectiveness. As a last step, the ALFRED control system has been finalized. The regulator design has been set up based on a simultaneous feedforward-feedback scheme incorporating four closed feedback loops. A controlled power reduction and a controlled overpower transient have been simulated in order to assess the performance of the two proposed control schemes. Results show that both the adopted control strategies can assure an efficient control of the thermal power while guaranteeing an effective control of lead and steam temperatures as well. In addition, some non-negligible differences between the two schemes have been observed and discussed in the simulation results of control and controlled variables.     

Effects of fuel type on the safety characteristics of a sodium-cooled fast reactor: Part I: Background,modeling tools and pre-transient calculations

The objective of this investigation is to examine the impact of the fuel type on the inherent safety characteristics of Liquid Metal Fast Reactors (LMFRs). To perform this study, the responses to various transient conditions are examined for metallic, oxide and nitride cores of a baseline LMFR. GE-Hitachi’s Super Power Reactor Innovative Small Module (S-PRISM) was chosen as the baseline LMFR. In this paper the background on S-PRISM’s metal and oxide cores are described and the redesign of a new nitride fueled S-PRISM core is introduced. Reactivity feedback and power profile data necessary for transient simulations with RELAP5-3D/ATHENA code are also presented and discussed. Part II of this paper will present the results of accident simulations and a comparison between the metal, oxide and nitride cores based on their performance during the selected accident scenarios.     

小型模块化熔盐快堆燃料管理初步分析

由于燃料随熔盐流动的特性以及可以进行在线添料与处理的特点,液态燃料熔盐堆的燃耗分析与燃料管理和传统固态燃料反应堆有很大不同,需要针对液态燃料熔盐堆的特点重新开发燃耗分析与管理程序。本文针对液态燃料熔盐堆的熔盐流动特性以及在线添料与处理功能,基于MCNP5和ORIGEN2.1燃耗耦合程序,开发了适用于液态燃料熔盐堆的燃料管理程序,并应用于一种小型模块化熔盐快堆的燃料管理和分析,对比分析了5种不同运行方案以及分批在线添料情况下,运行30年期间keff的变化情况及重要核素的演化情况。计算结果表明,采用不断调整添料率的连续在线添料运行方案和固定批量添料的运行方案,都可以让小型模块化熔盐快堆维持运行在一个较小的keff波动范围之内。开发的燃料管理程序适用于液态燃料熔盐堆的研究,同时可以为液态燃料熔盐堆的设计及燃耗管理和分析提供有价值的参考。     

Current perspective on fast reactor containment

The concept of “containment” is to provide a series of physical barriers between the radioactive products of the fission process and the public. All nuclear reactors have several such barriers and LMFBRs have more than most. These barriers are, successively:

1. fuel, which retains fission products;

2. fuel cladding, which encloses the fuel;

3. sodium coolant, which absorbs fission products released through fuel caldding;

4. primary coolant boundary, which has energy absorption and leakage control capabilities;

5. containment building, hereafter referred to as containment, which provides the final engineered barrier for control of radioactive releases;

6. exclusion distance, which provides space for natural attenuation of radioactive releases before reaching the public.

These barriers, along with the design approaches and features which protect their integrity under normal and accident conditions, assure that the public is adequately protected from the potential hazards of radioactivity residing in the core. It is only in the case of hypothesized core disruptive accidents (HCDAs) that these successive barriers can be sufficiently threatened as to pose a significant threat to the public. These HCDAs involve an extremely low probability sequence of successive failures resulting in core cooling imbalances which lead to fuel overheating. Under such conditions, the fuel and cladding barriers can be lost and energy sources can be generated which threaten the primary coolant boundary and containment. This paper addresses current perspectives on containment of HCDAs with emphasis on the approach and programs in the US.     

A study on reactivity insertion controlled LMR cores with metallic fuel

An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity.

A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core.

The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences.     

船用堆燃料棒包壳疲劳寿命分析

船用堆瞬态变工况下燃料棒包壳温度和冷却剂压力波动较大,引起包壳的疲劳损伤,因此包壳疲劳寿命分析至关重要。本文利用ANSYS软件模拟船用堆瞬态变工况下燃料棒的热机械行为,结合锆包壳疲劳寿命设计曲线,考察包壳温度、冷却剂压力、燃料棒内压以及辐照对船用堆燃料棒包壳疲劳寿命的影响。计算结果表明,瞬态变工况使得包壳疲劳寿命有很大降低;包壳温度变化与冷却剂压力变化相比,前者对包壳疲劳寿命的影响小;辐照会降低包壳疲劳寿命。在不影响核动力船舶机动性的前提下,可采取一些必要的措施来降低包壳的疲劳损伤。     

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO₂ spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.