Reactor local power distribution

As shown in Table 4.2, large break LOCA events involve the most physical phenomena and, therefore, require the most extensive analysis methods and tools. Typically, 3D reactor space-time kinetics physics calculation of the power transient is coupled with a system thermal hydraulics code to predict the response of the heat transport circuit, individual channel thermal-hydraulic behavior, and the transient power distribution in the fuel. Detailed analysis of fuel channel behavior is required to characterize fuel heat-up, thermochemical heat generation and hydrogen production, and possible pressure tube deformation by thermal creep strain mechanisms. Pressure tubes can deform into contact with the calandria tubes, in which case the heat transfer from the outside of the calandria tube is of interest. This analysis requires a calculation of moderator circulation and local temperatures, which are obtained from computational fluid dynamics (CFD) codes. A further level of analysis detail provides estimates of fuel sheath temperatures, fuel failures, and fission product releases. These are inputs to containment, thermal-hydraulic, and related fission product transport calculations to determine how much activity leaks outside containment. Finally, the dispersion and dilution of this material before it reaches the public is evaluated by an atmospheric dispersion/public dose calculation. The public dose is the end point of the calculation. 

Find the fuel distribution and the corresponding flux that give a constant power distribution in an infinite-slab reactor. Assume (1) a bare one-velocity reactor, (2) the extrapolation boundary condition, and (3) that the diffusion coefficient is unaffected by local variations in fuel concentration. 

A detailed treatment of the axial power distribution local heat transfer, two-phase mixture dynamics, and coupling with the rest of the reactor coolant system requires the use of complex computer models. Figure 3.2-1 compares the predictions based on Eq. (3.2-1) with code calculations for a Zion station blackout scenario compounded by failure of turbine-driven auxiliary feedwater (the so-called TMLB scenario). As indicated by the comparison, the exponentially decreasing function defined by Equations 3.2-1 and 3.3-2 is a reasonable approximation for the water level in the core region during this stage of the accident. 

It is also true that the study of possible accidents, even if limited, leads to the provision of abundant water for core submersion and for the shutdown of the chain reaction. The area of possible improvement concerns the systems which diagnose the conditions of possible danger to the eore itself. For this reason the group reeommended, in the first place, the installation, as far as teehnologjcally feasible on eaeh reactor, of instrumentation capable of directly and reliably measuring the water level, and the temperature and power local distribution, in the core. 

For the described limits, one possible solution to combine high-throughput with good thermal management is a multi-injection microchannel reactor, where one reactant is injected along the reactor. This concept is used as one of the approaches for scale-up [34, 35]. Distributed feeding of one reactant in multiple locations reduces the local heat power released depending on the number of injection points. The concept corresponds to the semibatch operation of conventional reactor vessels. 

Answer Since the teflq)erature cycling Is usually the result of a series of local turnarounds and power changes In certain parts of the pile> the only way to prevent cycling is to maintain an even flux distribution over the. various sections of the reactor. Even flux and temperature distribution can be maintained only through the correct interpretation of representative monitoring data, which includes adequate frequency and sampling ... 

---