Breeder Reactor Concept Research Articles

Modeling and simulation of online reprocessing in the thorium-fueled molten salt breeder reactor

In the search for new ways to generate carbon-free, reliable base-load power, interest in advanced nuclear energy technologies, particularly Molten Salt Reactors (MSRs), has resurged with multiple new companies pursuing MSR commercialization. To further develop these MSR concepts, researchers need simulation tools for analyzing liquid-fueled MSR depletion and fuel processing. However, most contemporary nuclear reactor physics software is unable to perform high-fidelity full-core depletion calculations for a reactor design with online reprocessing. This paper introduces a Python package, SaltProc, which couples with the Monte Carlo code, SERPENT2 to simulate MSR online reprocessing by modeling the changing isotopic composition of MSR fuel salt. This work demonstrates SaltProc capabilities for a full-core, high-fidelity model of the commercial Molten Salt Breeder Reactor (MSBR) concept and verifies these results to results in the literature from independent, lower-fidelity analyses.

Core design of PWR-type seed-blanket core breeder reactor with tightly packed fuel assembly

Pressurized water reactor (PWR) is the reactor type with the most abundant operation experience in the world. However, studies on designing PWR-type fast reactors have been limited and there have not been any PWR-type fast breeder reactor design concepts. In this study, the concept of seed-blanket PWR-type breeder reactor with tightly packed fuel assembly (TPFA) has been developed by coupled three dimensional neutronics and thermal-hydraulics core calculations. For the seed-blanket heterogeneous core using mixed oxide (MOX) fuel and depleted uranium, it has been shown that the core height is limited to about 1.0 m or less, in order to satisfy the design criterion of negative void reactivity. Moreover, it has been shown that increasing the core power density is difficult, as it leads to substantial increase in the core pressure drop. Consequently, the core characteristics are featured by low breeding performance with compound system doubling time (CSDT) of about 150 years or more (fissile plutonium surviving ratio of below 1.01) and low thermal power of about 700 MW or less. For the core using 4.95 wt% enriched uranium for the blanket assembly, it is possible to improve the void reactivity characteristics, breeding performance and thermal power by reducing reactivity difference between the seed and the blanket fuel assemblies. By utilizing enriched uranium, the concept of breeding PWR core with CSDT of 60 years and thermal power of 1000 MW has been shown. In addition, PWR-type breeder reactor concept using enriched uranium that the fissile surviving ratio including uranium and plutonium exceeds 1 was shown for the first time.

Analysis of a sustainable gas cooled fast breeder reactor concept

Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5GWdMTHM. TGFBR concept is to address the eventual shortage of 235U and nuclear waste management issues. A mixture of thorium and uranium (232Th+233U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239Pu. Initial feed of 233U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233U a theoretical evaluation has been used to derive the data for the source of 233U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5years assuming the design thermal power output as 445MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233U as an alternate to 235U and an alternate use for light water reactor used fuel as a blanket for the production of plutonium.

The Energy Multiplier Module: Advancing the Nuclear Fuel Cycle through Technology Innovations

To achieve long-term energy security in an environmentally acceptable manner, fission technology needs to make further advances in the areas of lower financial risk, better resource utilization, and reduced volumes of high-level waste. Without such progress, these concerns may be limiting factors in the exploitation of this vital resource. “Convert-and-burn” fast reactors offer the potential for advances in each of these areas without the specter of increased proliferation risk that accompanies breeder reactor concepts. An example is Energy Multiplier Module (EM2), a compact, helium-cooled fast reactor that augments its fissile fuel load with either depleted uranium or used nuclear fuel (UNF). The convert-and-burn in situ operating mode results in a core predicted to last 30 years without the need to add or shuffle fuel. EM2 can endure a station blackout, even one combined with a loss-of-coolant accident, using only passive safety systems to prevent radioactivity release or loss of plant. The end-of-cycle fuel and/or light water reactor UNF can be refabricated in a manner that does not separate out heavy metal, permitting reuse in subsequent generations at reduced proliferation risk. Proliferation resistance is further enhanced by eliminating the need for enrichment beyond that needed for the first-generation fuel load. Waste problems are mitigated by several factors: higher burnup, fuel use in multiple generations, and conversion of existing waste to energy. Economically attractive power costs are anticipated through a combination of high efficiency, simplicity of the direct-cycle gas turbine, and relatively small subsystems that can be shop fabricated and shipped by road to the site. Reactor materials have been carefully chosen to achieve a safe, economically affordable, and proliferation-resistant energy source.

Possible Configurations for the Thorium Molten Salt Reactor and Advantages of the Fast Nonmoderated Version

Molten Salt Reactors based on the thorium cycle were studied in the 1950 to 1960s to lead to the Molten Salt Breeder Reactor concept, which was finally discontinued prior to any industrial development. In the past few years, this concept has once again been studied in order to generalize it and seek configurations ensuring a high intrinsic safety level, an initial inventory compatible with intensive deployment on a worldwide scale, and a not-too-demanding salt chemical reprocessing scheme.The Thorium Molten Salt Reactor (TMSR) thus defined is studied in the Th-233U cycle in various configurations obtained by modulating the amount of graphite in core to obtain a thermal, an epithermal, or a fast spectrum. In particular, configurations of a fast spectrum TMSR have been identified with outstanding safety characteristics and minimal fuel-reprocessing requirements.

Physics principles to achieve comparable fission power from fertile and fissile rods of the conceptual ATBR/FTBR reactors

Loading of seedless fertile rods has been used as the central principle to maximize fertile to fissile conversion in the two thorium breeder reactor concepts, viz. ATBR and FTBR [Jagannathan, V., Pal, Usha, Karthikeyan, R., Ganesan, S., Jain, R.P., Kamat, S.U., 2001. ATBR – a thorium breeder reactor concept for an early induction of thorium in an enriched uranium reactor. Nuclear Technology 133, 1–32; Jagannathan, V., Pal, Usha, Karthikeyan, R., Raj Devesh, Srivastava, Argala, Ahmad Khan, Suhail, 2007. Reactor physics ideas to design novel reactors with faster fissile growth. In: Paper accepted for oral presentation in ‘ICENES 2007 – 13th International Conference on Emerging Nuclear Energy Systems, 3--8 June 2007, Istanbul, Turkey]. At fresh state the seedless thoria rods will produce practically no fission power, or nearly thousand times less fission rate compared to the seed fuel rods. Hence it is conceived that the fuel assembly would be constituted by assembling the fresh seed rods with one fuel cycle irradiated fertile thoria rods. Even in this state there is a wide disparity between the fissile content of these rods. By judicious choice of the rod dimensions and their relative locations, a degree of balance in the fission rate is achieved in the fresh state of seeded rods. Remarkably as the burnup proceeds the initially seedless fertile rods have a continuous growth of fissile content up to an asymptotic value for a given spectrum and the fissile content in seeded rods monotonically decreases. If the discharge burnup is sufficiently large by design, it is seen that the power share of the initially seedless fertile rods can even exceed that of the seed fuel rods. The physics principles of achieving this characteristic are presented in this paper.

Revisiting the thorium-uranium nuclear fuel cycle

europhysicsnews 24 • volume 38 • number 2 amounts of uranium 233 available remained tiny, insufficient to allow the rapid development of a thorium-uranium concept. It appears today that the growth rate of nuclear power worldwide does not require the fast development of breeder reactor concepts. It is then possible, as we show in this paper, to constitute a stockpile of uranium 233 that could allow the development of a fleet of thorium-uranium reactors. We show also that such a concept has many major advantages, in particular concerning the disposal of radioactive waste and the risks of nuclear proliferation.We give a brief description of the types of reactors being considered to implement this fuel cycle and of a strategy that makes use of today’s reactors to create the initial stockpiles of uranium 233.

The thorium molten salt reactor: Moving on from the MSBR

A re-evaluation of the molten salt breeder reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the thorium molten salt reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the molten salt reactor configurations that deserve further evaluation.

Conceptual Design of a Modular Island Core Fast Breeder Reactor “RAPID-M”

A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor performance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified refueling by replacing IFAs in which all the core and blanket fuel elements are comprised. In this paper, the 600 MWe RAPID-M design consists of 7 IFAs is presented. Significant reactor mass savings and the improvement of inherent safety features are discussed. Plant dynamics analyses using the multi-point reactor kinetics equations to accommodate the modular core configuration demonstrated a favorable transient response in case of unprotected transient over power (UTOP).

Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M".

A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor performance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified refueling by replacing IFAs in which all the core and blanket fuel elements are comprised. In this paper, the 600 MWe RAPID-M design consists of 7 IFAs is presented. Significant reactor mass savings and the improvement of inherent safety features are discussed. Plant dynamics analyses using the multi-point reactor kinetics equations to accommodate the modular core configuration demonstrated a favorable transient response in case of unprotected transient over power (UTOP).

ATBR—A Thorium Breeder Reactor Concept for Early Induction of Thorium in an Enriched Uranium Reactor

A new reactor concept has been proposed for induction of thorium in an enriched uranium reactor. The neutronic characteristics of the fissile and fertile materials have been exploited to arrive at optimal fuel assembly and core configurations. Each fuel assembly consists of an enriched uranium seed zone and a thoria blanket zone. They are in the form of ring-type fuel clusters. The fuel is contained in vertical pressure tubes placed in a hexagonal lattice array in a D2O moderator. Boiling H2O coolant is used. The 235U enrichment is ~5.4%. The thoria rods contain the 233U bred in situ by irradiation of one batch load of mere thoria clusters (without the seed zone) for one fuel cycle in the same reactor. There is no need for external feed enrichment in thoria rods. Additionally, some moveable thoria clusters are used for the purpose of xenon override. The fissile production rate from the fertile material and the consumption rate of fissile inventory is judiciously balanced by the choice of U/Th fuel rod diameter and the number and location of thoria rods in the fuel assembly and in the core. During steady-state operation at rated power level, there is no need for any conventional control maneuvers such as change in soluble boron concentration or control rod movement as a function of burnup. Burnable poison rods are also not required. A very small reactivity fluctuation of ±2 mk in 300 effective full-power days of operation is achieved and can be nearly met by coolant inlet enthalpy changes or moveable thoria clusters. Control is required only for cold shutdown of the reactor. The uranium as well as thoria rods achieve a fairly high burnup of 30 to 35 GWd/tonne at the time of discharge. Since the excess reactivity for hot-full-power operation is nearly zero at all times during the fuel cycle and since the coefficients of reactivity due to temperature and density variations of coolant are nearly zero by design, there is hardly any possibility of severe accidents involving large reactivity excursions.

Innovative fast breeder reactor concept ‘RAPID’ for improvement of reactor performance and proliferation resistance

The 60 MWe metal fueled fast breeder reactor concept ‘RAPID’ to improve reactor performance and proliferation resistance has been demonstrated. The reactor can be operated without refueling for up to 5 years. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. RAPID concept enables quick and simplified refueling by replacing an IFA in which all the core and blanket fuel elements are comprised. An on-site storage cask achieves on-site decay heat removal of an IFA. After 3 years of on-site storage, an IFA together with an on-site storage cask can be transported directly to the reprocessing plant without any intermediate steps. Significant improvement of inherent safety features and plant availability has been discussed. Decay heat removal capability, safety consideration on criticality of the IFA and shielding design of the on-site storage cask has been confirmed. The RAPID refueling concept possesses high resistance to state-supported removal of plutonium for nuclear weapons production.

Comprehensive Analysis of Passive Safety Test Phase IIB in the Fast Flux Test Facility

Power Reactor and Nuclear Fuel Development Corporation (PNC), using computer codes developed and/or modified at PNC, has analyzed the Phase IIB passive safety test (PST) proposed for the Fast Flux Test Facility (FFTF). Major interests of PST are understanding core bowing and mitigating extremely low probability accidents. It is confirmed that the PNC code system is applicable to all aspects of the FFTF Phase IIB PST program. Results of the pretest analysis indicate that the proposed Phase IIB PST in FFTF can be devised to provide very useful data for validation of the analytical models that treat reactivity feedback effects due to core bowing. Recommendations to the test program are also made. With those analytical tools, future fast breeder reactor concepts will incorporate to a greater degree than earlier designs passive safety features on neutronic aspects as well as on thermohydraulic aspects.

Safety Design Study of Fast Breeder Reactors in Japan

Two fast breeder reactor (FBR) concepts, the tank type and the loop type, have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle of the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant.

Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as a parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs.

Safety Assessment of the Fusion Breeder

This paper presents the safety features and describes the supporting analyses of the fusion breeder reactor concepts developed by the Fusion Breeder Program (FBP). The reactor and blanket concepts studied include suppressed and fast-fission blankets for both the tandem mirror and tokamak confinement schemes. Helium and lithium cooled blankets are considered. Most of the effort was directed toward a lithium cooled, fission-suppressed, tandem mirror blanket.1,2 Other concepts are evaluated with comparatively minimal analysis. Mobile fuel, which can be gravity dumped to separately cooled tanks, together with other design solutions and safety systems, can result in an acceptably safe fusion breeder.

Instrumentation for Reactor Two Phase Noise Diagnosis

The aim of this work is directed towards developing two particular applications of two phase noise diagnosis in nuclear power plants, cavitation damage prediction in circulating pumps, particularly for sodium cooled fast breeder reactor concepts and the detection of incipient boiling and cavitation in pressurized water or sodium cooled reactors. For the above purpose a high frequency pressure-bar probe and a digital acquisition and processing system has been designed to find the amplitude and statistical characterics of pressure pulses from cavitation in a venturi which simulates conditions in hydraulic machinery and high speed flow channels. Significant results have been obtained about two phase inception phenomenon for the development of instrument systems for the above two applications.