Bruno Merk

Use of Zirconium-Based Moderators to Enhance Feedback Coefficients in a MOX-Fueled Sodium-Cooled Fast Reactor

Abstract This work shows the effect of the use of moderating layers on the sodium void effect in sodium-cooled, mixed oxide-fueled fast breeder reactors. The moderating layers consist of either zirconium boride ZrB2 or zirconium hydride ZrH2. The two investigated ZrH2 layers (0.1 and 0.2 mm thick) cause a strong reduction of the sodium void effect. Additionally, these layers significantly improve the fuel temperature effect and the coolant effect of the system. All changes caused by the insertion of the ZrH2 layers result in a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. The effect in the infinite system can be fully combined with the traditional methods of increasing the neutron leakage.

On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

Transmutation of All German Transuranium under Nuclear Phase Out Conditions - Is This Feasible from Neutronic Point of View?

The German government has decided for the nuclear phase out, but a decision on a strategy for the management of the highly radioactive waste is not defined yet. Partitioning and Transmutation (P&T) could be considered as a technological option for the management of highly radioactive waste, therefore a wide study has been conducted. In the study group objectives for P&T and the boundary conditions of the phase out have been discussed. The fulfillment of the given objectives is analyzed from neutronics point of view using simulations of a molten salt reactor with fast neutron spectrum. It is shown that the efficient transmutation of all existing transuranium isotopes would be possible from neutronic point of view in a time frame of about 60 years. For this task three reactors of a mostly new technology would have to be developed and a twofold life cycle consisting of a transmuter operation and a deep burn phase would be required. A basic insight for the optimization of the time duration of the deep burn phase is given. Further on, a detailed balance of different isotopic inventories is given to allow a deeper understanding of the processes during transmutation in the molten salt fast reactor. The effect of modeling and simulation is investigated based on three different modeling strategies and two different code versions.

An Analytical Solution of the Time-Dependent Diffusion Equation in a Composite Slab

The time-dependent, one-dimensional diffusion equation is solved for a finite slab of two layers. An external source is supplied to one of the layers. The differential equations are subject to the reflecting boundary conditions at the two outer boundary surfaces. The flux and the current density are continuous across the interface between two media. The exact analytical solution is expressed in terms of a Green’s function. The solution is developed by the application of the Laplace transformation.

Demand driven salt clean-up in a molten salt fast reactor – Defining a priority list

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

A disruptive approach to eliminating weapon-grade plutonium – Pu burning in a molten salt fast reactor

The successful implementation of disarmament treaties of the last centuries has led to significant amounts of weapon-grade Plutonium which are currently stored in high security storage facilities. Disposing this Plutonium should be seen as ‘good housekeeping’ avoiding unnecessary costs and the hazards of storing this material indefinitely. In addition, the disarmament is only brought to a successful end when the Plutonium isn’t available for the production of new weapons anymore. We propose a disruptive approach for Plutonium disposition and demonstrate the feasibility in a neutronic study. Burning of weapon-grade Plutonium in a molten salt fast reactor is significantly more efficient than in the studied other reactors, while efficient process design has the potential to reduce the security concerns significantly. The proposed system could turn about 1.25 tons of weapon-grade Plutonium into electric energy worth £ 0.5 to 1 billion/year, depending on the electricity price while avoiding the hassle and eliminating the risk of high security Plutonium storage. In conclusion, burning of the weapon-grade Plutonium resulting from disarmament could be an economically very attractive approach to reduce the nuclear threat.

Separation of 103Ru from a proton irradiated thorium matrix: A potential source of Auger therapy radionuclide 103mRh

Ruthenium-103 is the parent isotope of 103mRh (t1/2 56.1 min), an isotope of interest for Auger electron therapy. During the proton irradiation of thorium targets, large amounts of 103Ru are generated through proton induced fission. The development of a two part chemical separation process to isolate 103Ru in high yield and purity from a proton irradiated thorium matrix on an analytical scale is described herein. The first part employed an anion exchange column to remove cationic actinide/lanthanide impurities along with the majority of the transition metal fission products. Secondly, an extraction chromatographic column utilizing diglycolamide functional groups was used to decontaminate 103Ru from the remaining impurities. This method resulted in a final radiochemical yield of 83 ± 5% of 103Ru with a purity of 99.9%. Additionally, measured nuclear reaction cross sections for the formation of 103Ru and 106Ru via the 232Th(p,f)103,106Ru reactions are reported within.

An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF

A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

SP3 Solution versus Diffusion Solution in Pin-by-Pin Calculations and Conclusions Concerning Advanced Methods

Opportunities for and limitations of the use of the SP3 solution instead of the diffusion solution are given and discussed. Due to existing limitations, the use of a hybrid method consisting of nodal full core calculations coupled with an advanced transport solution based on the current coupling collision probability method with an orthonormal flux expansion is proposed. The method seems to be promising compared to adaptive mesh using refined geometry but without refined detail information, which is deleted by the homogenization process and compared to brute force full core pin-by-pin using advanced transport solvers.