Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.     

Study of thorium advanced fuel cycle utilization in light water reactor VVER-440

The first step in investigation of thorium fuel is evaluation of the results obtained from the spectral code for this type of fuel. The benchmark summarized by IAEA in 2003 was used for partial validation of the code HELIOS 1.9. The benchmark was focused on a comparison of the methods and basic nuclear data. Acceptable results of benchmark comparison allowed examining and comparing different advanced nuclear fuel cycles under light water reactor conditions, especially in VVER-440. Cycles, calculations and results for VVER-440 reactors are presented in the paper. Two of the investigated thorium based fuels include one solely plutonium–thorium based fuel, while the other one is a plutonium–thorium based fuel with a content of reprocessed uranium. The third examined fuel cycle is a cycle with an inert-matrix fuel consisting of reprocessed plutonium and minor actinides (MA) fixed in an yttria-stabilized zirconium matrix. All of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The Pu transmutation rate and cumulating of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The fuel cycle with an inert-matrix fuel was proven to be the best cycle for minimizing the production of Pu in the VVER-440 reactors.     

Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.     

Preliminary study on feasibility of large and small water cooled thorium breeder reactor in equilibrium states

Fuel breeding is one of the essential performances for a self-sustaining reactor system which can maintains the fuel sustainability while the reactor produces energy and consumes the fissile materials during operation. Thorium cycle shows some advantageous on higher breeding characteristics in thermal neutron spectrum region as shown in the Shippingport reactor and molten salt breeder reactor (MSBR) project. In the present study, the feasibility of large and small water cooled thorium breeder reactors is investigated under equilibrium conditions where the reactors are fueled by ²³³U–Th oxide and they adopts light water coolant as moderator. The key properties such as required enrichment, breeding capability, and initial fissile inventory are evaluated. The conversion ratio and fissile inventory ratio (FIR) are used for evaluating breeding performance. The results show the feasibility of breeding for small and large reactors. The breeding performance increases with increasing power output and lower power density. The small reactor may achieve the breeding condition when the fuel pellets' power density of about 22.5 W/cm³ and burnup of about 20 GWd/t.     

Design of a boiling water reactor core based on an integrated blanket–seed thorium–uranium concept

This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket–seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the ²³³U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average ²³⁵U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly.     

A constraint-based approach to intelligent support of nuclear reactor design

Constraint is a powerful representation to formulate and solve problems in design; a constraint-based approach to intelligent support of nuclear reactor design will be proposed in this paper. We will first discuss the features of the approach, and then present the architecture of a nuclear reactor design support system under development. In this design support system, the knowledge base contains constraints useful to structure the design space as object class definitions, and several types of constraint resolvers are provided as design support subsystems. The adopted methods of constraint resolution will be explained next in detail. The usefulness of the approach will be demonstrated using two design problems: design window search and multiobjective optimization in nuclear reactor design.     

Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor

The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a near-term production reactor. Four tokamak reactor concepts (TORFA/FED-R, AFTR/ZEPHYR, Riggatron, and Superconducting Coil) of approximately 500-MW fusion power are compared with regard to their demands on plasma performance, required fusion technology development, and blanket configuration characteristics. Because of its relatively moderate requirements on fusion plasma physics and technology development, as well as its superior configuration of production blankets, the TORFA/FED-R type of reactor operating with a fusion power gain of about 3 is found to be the most suitable tokamak candidate for implementation as a near-term production reactor.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing from its 1982 version (except for Tables II and III and Fig. 1), explaining the fact that some of the material is dated.     

Some characteristics of a fast reactor with thorium shield

The regeneration factor of Pu²³⁹ in U²³³ was determined in the BR-1 experimentaI fast reactor with a Pu²³⁹ core and a Th²³² shield. The breeding characteristics of thorium, the utilization factor of fission neutrons (D) and the neutron multiplication factor (k) were also studied.Translated from Atomnaya Énergiya, Vol. 17, No. 4, pp. 294–299, October, 1964Deceased     

Petri-net based modelling approach for ALFRED reactor operation and control system design

In this work, the Petri-net modelling approach applied to the control system design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is presented, paying particular attention to the startup procedure. The reactor startup is the operational transient in which all the systems of the plant are brought from the cold shutdown condition to the full power mode, close to load-frequency control. In this phase, the several control actions to be taken need to be properly coordinated. To this end, the operational sequence which constitutes the reactor startup procedure has been described by adopting the Petri-nets approach, i.e., a useful formalism for the modelling and the analysis of Discrete Event Systems. Thanks to this quantitative representation, it is possible to easily derive the corresponding control scheme. In addition, the Petri-nets approach has been also exploited for the two-level control system architecture, namely a master system coordinates the operation of the plant by sending suitable signals to the slave system, in which feedback controllers are implemented. As a major outcome of this work, the procedure for the reactor startup and the transition to the full power mode has been simulated in order to assess the control system performance.     

Neutron-physical characteristics and fuel utilization in a sodium-cooled fast reactor with high-density fuel at the initial stage of the transition to a closed cycle

The results of neutron-physical investigations of a fast reactor using high-density fuel (UC) at the initial stage of the transition to a closed fuel cycle are presented. Validation is given for the possibility of making the transition to a closed fuel cycle with self-supply of fissile nuclei starting with the first recycle. Translated from Atomnaya énergiya,Vol. 106, No. 1, pp. 8–15, January, 2009.     

钍资源的核能利用问题探讨

分析了钍/铀燃料循环特点,评估了国际上钍资源利用研究开发现状和发展趋势,并试图按照科学发展观提出了我国钍资源核能利用的战略思考和钍/铀燃料循环前瞻性研究开发课题.     

A method for statistical steady state thermal analysis of reactor cores

This paper presents a method for performing a statistical steady state thermal analysis of a reactor core. The technique is only outlined here since detailed thermal equations are dependent on the core geometry. The method has been applied to a pressurised water reactor core and the results are presented for illustration purposes.Random hypothetical cores are generated using the Monte-Carlo method. The technique shows that by splitting the parameters into two types, denoted core-wise and in-core, the Monte Carlo method may be used inexpensively. The idea of using extremal statistics to characterise the low probability events (i.e. the tails of a distribution) is introduced together with a method of forming the final probability distribution. After establishing an acceptable probability of exceeding a thermal design criterion, the final probability distribution may be used to determine the corresponding thermal response value. If statistical and deterministic (i.e. conservative) thermal response values are compared, information on the degree of pessimism in the deterministic method of analysis may be inferred and the restrictive performance limitations imposed by this method relieved.     

Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation.This paper presents a discussion of the history and some of the results of this effort, with respect to advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The purpose is to describe the nonlinear dynamic analysis techniques employed to analyze the HTGR core. Correlation of the codes is beyond the scope of the paper and will be discussed in subsequent publications.Each fuel column in the HTGR core is composed of stacked elements doweled together to ensure alignment of the coolant channels. Gaps exist between columns, allowing the elements to impact during a seismic disturbance. Analysis of this type of structure by standard structural dynamics techniques is not possible since both nonlinearities and discontinuities exist. One- and two-dimensional models of the three-dimensional core have been developed with explicit time integration methods. Various methods to treat the impact between elements are discussed.Three computer codes were developed. CRUNCH-1D models a one-dimensional horizontal strip through the core; CRUNCH-2D, a two-dimensional horizontal planar section; and MCOCO, a two-dimensional vertical planar section. The dynamic characteristics of these three representations of the full core structure are compared and the methods evaluated in the text. Plans for additional development and work to improve the techniques are also discussed.     

The utilization of thorium for long-life small thermal reactors without on-site refueling

Thorium cycle has many advantages over uranium cycle in thermal and intermediate spectrum nuclear reactors. In addition to large amount of resources in the world which up to now still not utilized optimally, thorium based thermal reactors may have high internal conversion ratio so that they are very potential to be designed as long-life reactors without on-site refueling based on thermal spectrum cores. In this study preliminary study for application of thorium cycle in some of thermal reactors has been performed.

We applied thorium cycle for small long-life high temperature gas reactors without on-site refueling. Calculation results using SRAC code show that 10 years lifetime without on-site refueling can be achieved with excess reactivity of about 10% dk/k.

The next application of thorium cycle has been employed in long-life small and medium PWR cores without on-site refueling. Relatively high fuel volume fraction is also applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 20–300 MWth PWR with maximum excess reactivity of a few %dk/k.

The last application of thorium cycle has been employed in long-life BWR cores without on-site refueling. Relatively high fuel volume fraction is applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 100–600 MWth BWR with maximum excess reactivity of a few %dk/k.     

Integrated design approach of the pebble BeD modular reactor using models

The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts.This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results.This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR.     

Fast reactor decay heat removal: approach to the safety system design in Japan and Europe

In the framework of the cooperation on fast reactor between the European and Japanese electrical utilities, the design companies responsible for the demonstration fast breeder reactor (DFBR) in Japan and the European fast reactor (EFR) have performed a comparative evaluation of the safety qualified decay heat removal systems of the two reactor designs. At the level of overall safety and concept design there is an obvious similarity between the two DHR systems. In both cases heat is removed directly from the reactor vessel primary sodium by systems designed according to a similar deterministic methodology, with a probabilistic assessment performed to demonstrate achievement of the required reliability. Nevertheless, the evaluation revealed a number of differences resulting from different national practices. These include the application of diversity and redundancy philosophy, the extent of passivity taken into account, the consequences of postulated maintenance outage on the design and the decay heat curve.