Development of Ultra-long Cycle Small Modular Sodium-cooled Fast Reactor Core Concept

The sodium-cooled fast reactor (SFR) is one of the six Gen-IV reactor concepts, and it is a near-term deployable reactor concept with its technological achievement and the experience through that many countries have operated and developed. On the other hand, the small modular reactor (SMR) concept is representative reactor design concept for the near future regarding the advanced plant design trend. There are some SMR concepts developed in some countries and it is expected that the SMR market would be enlarged to fit the global electricity demand with the flexibility of SMRs. In this thesis, a hybrid concept of SFR and SMR is developed with the characteristics of an ultra-long cycle operation and the advanced safety feature arisen from the combination design of the two reactor design concepts. There are some important parameters in the fast reactor physics such as conversion ratio, capture to fission ratio, fertile fission bonus, and they can be very different according to the design option. The coolant void reactivity and the thermal expansion coefficients are the critical parameters in fast reactor core design that decide the safety of the core. Choosing metal fuel and liquid metal coolant also secure the inherent safety of fast reactor with its reactivity parameters so the materials of them dominate the key characteristics of the core. At the same time, to have transportability of the reactor modules, which is one of the important identities of SMR, the diameter of the reactor vessel or the core barrel is limited less than 3 m. Before reactor design, the feasibility study of the ultra-long cycle operation was performed for the target of 30-year operation. For this, existing small fast reactor concepts with long-cycle operation were assessed first, and the fuel, coolant, and structure materials were reviewed. In the reactor design stage, the reactor and core design requirements were set based on the previous SFR and SMR design studies and the feasibility study of long cycle operation of small fast reactors. With the core design requirement, pin, assembly and the core design parameters were determined so as to have the safety feature in any state during the long cycle operation. With the fixed core geometry, five breed-and-burn strategies were tested in the five different cores, and finally an optimized breed-and-burn strategy was developed through the combination of the test cases. The core performance of the final core design was evaluated and the operation feasibility was assessed in the neutronics point of view. It was confirmed that the 30-year operation of the core is feasible even its conversion ratio is less than unity and that the active core breeds in radial direction from the peripheral region and the power peak moves to the core center as the burnup propagates. The Pressurized Water Reactor (PWR) spent fuel was tested by loading for the blanket material and the operation feasibility of it was also confirmed. Core kinetics parameters and reactivity feedback coefficients were assessed, and the sodium void worth and expansion coefficients show negative values. The control rod system was investigated that it is optimized to control the core reactivity to make it critical throughout its operation time and to secure the shutdown margin and another primary control rod material was proposed. With the integral reactivity parameters, the quasi-static reactivity balance analysis performed for the anticipated transient without scram (ATWS) events, the unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), and unprotected transient over power (UTOP) events, and it shows the inherent safety of the core. Furthermore, the transient analysis was also performed and it has been confirmed that this newly developed core is inherently safe and has capability to cope with the beyond design basis accident (BDBA). Both Monte Carlo and deterministic approaches were utilized by using neutronics computer codes for the steady-state analysis in the design and evaluation stage of the core modeling. Moreover, safety analysis computer code system was also utilized to perform the transient analysis for the modeled core.