Neutronic analysis of SFR core with HELIOS-2, Serpent, and DYN3D codes

Neutronic benchmark of the molten salt fast reactor in the frame of the EVOL and MARS collaborative projects

This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR). The MSFR operates with the thorium fuel cycle and can be started with 233U-enriched U and/or TRU elements as initial fissile load. All three compositions were covered by the present benchmark. The calculations have confirmed that the MSFR has very favorable characteristics not present in other Gen4 fast reactors, like strong negative temperature and void reactivity coefficients, a low-fissile inventory, a reduced long-lived waste production and its burning capacities of nuclear waste produced in currently operational reactors.

Reactor safety research within the Helmholtz Association

The reactor safety research of the Helmholtz Association is an integral part of the national provident research. It focuses on the safety of nuclear power plants in Germany and abroad as well as on safety aspects of internationally developed innovative reactor concepts. The research in the three involved Helmholtz centers Forschungszentrum Jülich, Helmholtz-Zentrum Dresden-Rossendorf and Karlsruhe Institute of Technology covers important areas of design basis and beyond design basis accidents. A unique combination of code and model development supported by own large-scale experiments ensures the active preservation of the reactor safety competences. The research that is embedded in a strong international co-operation will be continued after the completion of the national phase-out from the use of nuclear energy for electricity production in 2022.

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.

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

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.

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.

The reactor dynamics code DYN3D and its trigonal-geometry nodal diffusion model

The reactor dynamics code DYN3D is a three-dimensional best-estimate tool for simulating steady states and transients of light-water reactors and innovative reactor designs. An overview of the DYN3D features is provided. This paper further focuses on the recently developed trigonal-geometry diffusion model DYN3D-TRIDIF including a description of the underlying nodal approach and the characteristics of trigonal geometries. Via a mesh refinement study by means of a VVER-1000-type core benchmark using a fine-mesh diffusion reference solution, DYN3D-TRIDIF shows spatial convergence. Furthermore, the performance of DYN3D-TRIDIF is verified by means of a single-assembly problem on pin-cell level. Good agreement between DYN3D-TRIDIF and the detailed-geometry transport reference is achieved with an average deviation in power of less than 1%.

Fine Distributed Moderating Material with Improved Thermal Stability Applied to Enhance the Feedback Effects in SFR Cores

The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effect in sodium-cooled fast reactor cores is described. The influence of the moderating material on the fuel assembly geometry, the neutron spectrum, the feedback effects, the power and burnup distribution, and the transmutation performance is given. An overview on possible materials is provided and the relationship between hydrogen content and thermal stability is described. A solution for the problem of the limited thermal stability of primarily proposed hydrogen-bearing moderating material ZrH1.6is developed by the use of yttrium-mono-hydride. The similarity in the effects reached by ZrH and YH is demonstrated by comparison calculations. The topic is closed by an overview on material properties, manufacturing issues, experience in fast reactors, and a comparison of raw material costs.