In the past years, more concerns are focused on the nuclear waste management due to the very long half-lives of various actinides produced in Light Water Reactors (LWRs). Sodium Fast Reactors (SFRs) are thus becoming more attractive since they are known to be very efficient to transmute long-lived radionuclides present in spent fuel. However, the current simulation tools (thermal-hydraulics code with point kinetics) and safety assessment methods are not as mature as for LWR applications and need to be enhanced.
This thesis aims at filling the gap in safety analysis of SFR cores to reach a standard similar to LWR applications by applying multi-physics modelling. In contrast to LWRs, the reactivity in SFRs is affected by three main feedback: the Doppler broadening reactivity effect, the sodium density change reactivity effect and the thermal expansion of several mechanical components of the reactor.
In this thesis, the thermal-hydraulic system code ATHLET is coupled with the three-dimensional neutron-physics code PARCS for transient analysis. Developed at GRS, ATHLET was recently upgraded for sodium coolant properties. The nodal diffusion codes PARCS, developed at the University of Michigan, can solve the multi-group diffusion equation in hexagonal geometry. While both codes already have the main features to simulate SFRs, the development of models dedicated to the thermal expansion effect of reactivity is necessary. The latter has three main origins i.e. the core axial thermal expansion effect (caused by the fuel and the cladding axial thermal expansion), the core radial thermal expansion effect (caused by the diagrid thermal expansion), the control rod displacement due to the thermal expansion of the Control Rod Drive Lines (CRDLs), the strongback and the reactor vessel. Thus, the three main new developments achieved in the scope of this work are:
- Development of a method to generate homogenized multi-energy-group neutron macroscopic cross sections (needed by PARCS) for SFR applications which consider not only the Doppler temperature and sodium density but also the core axial and radial thermal expansion.
- Development of a three-dimensional core radial thermal expansion model and its implementation in PARCS. A core axial thermal expansion model has already been developed for PARCS prior to this work.
- Development of a module in ATHLET for modelling the control rod displacement as a result of the influence of the reactor structures thermal expansion.
The parametrized homogenized multi-energy-group neutron macroscopic cross section libraries for PARCS applications are generated with the Monte Carlo reactor-physics code Serpent. For all materials contained in fuel assemblies, a three-dimensional model is used while the SPH method is applied to materials contained in non-fuel assemblies (e.g. control rods, etc.). The cross section libraries are collapsed into a 12-energy-group structure. Furthermore, a dedicated module was successfully developed and implemented within the core simulator KMACS (developed at GRS).
The core radial thermal expansion effect is implemented in PARCS using a coordinate transformation of the diffusion equation from the expanded state to the nominal geometry. The core radial thermal expansion depends on the diagrid temperature. It is calculated by ATHLET and transferred to PARCS by the extended interface between both codes.
The modelling of the control rod displacement as a result of the reactor structures thermal expansion is performed by a module linked to ATHLET. The strongback, the reactor vessel and the CRDLs are modelled as heated structures in ATHLET, which calculates their respective temperature. The module can compute the thermal expansion of each structure as well as the total control rod banks displacement.
The new techniques are verfied on a selected case study, the ASTRID core design. First, full core criticality simulations are performed with the Monte Carlo reactor-physics code Serpent (considered as reference calculations) and with PARCS. Good agreement between the two codes is achieved in terms of multiplication factors and power distribution. This allows to conclude that the developed method for neutron cross section libraries can be used for SFR applications.
The newly implemented core radial expansion model in PARCS is successfully verified on the ASTRID core with the standalone version of PARCS. Then, various transient simulations are performed in order to separately analyse the different contributions to the reactivity by: the Doppler broadening effects, the sodium density change effect, the core radial and axial thermal expansion effect and the control rod displacement effect. It is demonstrated that the core power responses are plausible which allows the conclusion that all the different thermal expansion models are properly implemented.
Furthermore, the presented simulations show very different core power responses. It appears that the effect of the sodium density change on reactivity is a parameter that is strongly heterogeneous (depending on the core location). This shows the importance of using a three-dimensional neutron kinetics model rather than a point-kinetic model for transient simulations with thermal-hydraulic codes. Moreover, the time-scale of the various effects are ranging from few seconds to several hundred seconds. While the Doppler broadening, the sodium density change, as well as the core axial and radial thermal expansion effects on reactivity are fast, the thermal expansion of the strongback and the vessel only appears after several hundred seconds. This emphasizes the importance of considering all thermal expansion effects in addition to the usual thermal-hydraulic feedback parameters (e.g. fuel temperature, coolant density etc.) to be able to compute the core behavior realistically.:Contents
Abstract II
List of Figures VII
List of Tables X
List of Acronyms XI
Acknowledgments XIII
1 Introduction
1.1 Sodium cooled fast reactors
1.1.1 Fast reactor development
1.1.2 Comparison of sodium fast reactor and pressurized water reactor designs
1.1.2.1 Neutron spectrum
1.1.2.2 Breeding
1.1.2.3 Partitioning and Transmutation
1.1.2.4 Control of the reactivity in the core
1.1.2.5 Coolant properties
1.1.2.6 Reactivity feedback
1.1.2.7 Comparison summary
1.2 Objectives and structure of the thesis
1.2.1 Objectives
1.2.2 Structure of the thesis
2 State of the art of Sodium Fast Reactor safety assessment
2.1 Relevant safety events to consider for Sodium Fast Reactors
2.2 Major reactivity feedback mechanisms
2.3 State of the art of safety analysis methods for Sodium Fast Reactor
3 Methods and codes for safety assessment of sodium cooled fast reactors
3.1 Neutronics core calculations
3.1.1 Core calculations with the diffusion code PARCS
3.1.2 Generation of nodal few-group cross sections with the Monte Carlo code Serpent
3.1.3 Core simulator KMACS
3.2 Thermal-hydraulics simulations with the system code ATHLET
3.3 Coupled three-dimensional thermal-hydraulics / neutronics calculations
4 Development of three-dimensional thermal expansion models
4.1 General calculation approach proposed for safety assessment
4.2 Thermal expansion in solids
4.3 Model for generating nodal few-energy-group cross sections for deterministic core analysis
4.3.1 Energy group structure
4.3.2 Full-scale three-dimensional fuel assembly models in Serpent
4.3.3 Two-dimensional non-fuel assembly models in Serpent
4.3.4 Super homogenization method for non-multiplying media
4.3.5 Automated creation of Serpent models for parametrized cross section generation with KMACS
4.4 Core radial thermal expansion effect
4.4.1 Description of the core radial thermal expansion phenomenon
4.4.2 Coordinate transformation of the diffusion equation
4.4.3 Implementation of the coordinates transformation in PARCS
4.4.4 Adapted cross section parametrization scheme for the core radial expansion model
4.4.5 Diagrid model in ATHLET and temperature transfer
4.5 Core axial thermal expansion effect
4.5.1 Description of the core axial thermal expansion phenomenon
4.5.2 Implementation of a core axial thermal expansion model in PARCS
4.5.3 Appropriate cross section parametrization scheme
4.6 Control rod displacement due to reactor structures thermal expansion effects
4.6.1 Modelling scheme
4.6.2 Strongback model in ATHLET
4.6.3 Vessel model in ATHLET
4.6.4 Control rods drive lines ATHLET model
5 Verification on a case study
5.1 Description of the ASTRID reactor
5.2 Full core models
5.2.1 Full core Serpent reference models of the ASTRID core
5.2.2 Three-dimensional neutron kinetics model of ASTRID core in PARCS
5.2.3 Generation of appropriate few-group cross sections
5.2.4 Thermal-hydraulic model in ATHLET and ATHLET-PARCS feedback mapping
5.3 Verfications of the radial core expansion model
5.4 Assessment of the Doppler and sodium density effects
5.4.1 Assessment of the Doppler effect
5.4.2 Assessment of the sodium density effect
6 Coupled three-dimensional thermal-hydraulics/neutron-physics transient simulations with ATHLET-PARCS
6.1 Description of the models and transient simulations
6.2 Simulation 1: Doppler effect
6.2.1 Description
6.2.2 Results
6.3 Simulation 2: Sodium density effect
6.3.1 Description
6.3.2 Results
6.4 Simulation 3: Doppler and sodium density effects
6.4.1 Description
6.4.2 Results
6.5 Simulation 4: Core radial thermal expansion effect
6.5.1 Description
6.5.2 Results
6.6 Simulation 5: Doppler, Sodium density and core radial thermal expansion effects
6.6.1 Description
6.6.2 Results
6.7 Simulation 6: Core axial thermal expansion effect
6.7.1 Description
6.7.2 Results
6.8 Simulation 7: Doppler, Sodium density and core axial thermal expansion effects
6.8.1 Description
6.8.2 Results
6.9 Simulation 8: Doppler effect, Sodium density effect, core radial thermal expansion effect and core axial thermal expansion effect
6.9.1 Description
6.9.2 Results
6.10 Simulation 9: Doppler effect, Sodium density effect, core radial thermal expansion effect, core axial thermal expansion effect and control rod displacement due to reactor structures thermal expansion effect
6.10.1 Description
6.10.2 Results
6.11 Preliminary conclusions of the test calculations
7 Conclusion and outlook for future developments
7.1 Summary and conclusions
7.2 Suggestions for future work
Appendices
A The Boltzmann equation
B Macro-group structure
Bibliography
| Identifer | oai:union.ndltd.org:DRESDEN/oai:qucosa:de:qucosa:78793 |
| Date | 11 April 2022 |
| Creators | Bousquet, Jeremy |
| Contributors | Weiß, F. - P., Hampel, Uwe, Macian-Juan, R., Technische Universität Dresden |
| Source Sets | Hochschulschriftenserver (HSSS) der SLUB Dresden |
| Language | English |
| Detected Language | English |
| Type | info:eu-repo/semantics/publishedVersion, doc-type:doctoralThesis, info:eu-repo/semantics/doctoralThesis, doc-type:Text |
| Rights | info:eu-repo/semantics/openAccess |
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