Coupled neutronic and thermal-hydraulic

Laureau, A., Rubiolo, P., Heuer, D., Merle-Lucotte, E., Brovchenko, M., 2013. Coupled neutronics and thermal-hydraulics numerical simulations of the molten salt fast reactor... 

The block diagram used for coupled neutronic and thermal-hydraulic stability of the Super LWR is shown in Fig. 1.28. The neutronic model is used to find the forward transfer function G(i) and the thermal-hydraulic heat transfer and ex-core models are used to determine the backward transfer function H(s). The fi-equency... 

The time delay of the heat transfer to the coolant and moderator water is an important factor in the mechanism of coupled neutronic and thermal-hydraulic instability. The Super LWR is a reactor system with a positive density coefficient of reactivity and a large time delay constant. If there is no time delay, a decrease in density would cause a decrease in power generation, which suppresses any further decrease in density, stabilizing the system. However, if there is a large time delay, it causes a decrease in the gain of the density reactivity transfer function, and reduces the effect of density reactivity feedback, making the system less stable. The time delay of the heat transfer to the water rods is much larger than that to the coolant. Thus the reactor system becomes less stable when the water rod model is included than the case without it. 

Figure 1.31 shows the decay ratio contour map for coupled neutronic and thermal-hydraulic stability of the Super LWR. The decay ratio contour line of DR =1.0 indicates the stability boundary on the power versus flow rate map of the Super LWR. At the high power low flow rate region, the reactor becomes unstable. At low power operation and during startup, it is necessary to take care to satisfy the stability criteria. At the low power low flow rate region, the unstable conditions should be avoided by carefully adjusting the flow rate. 

Fig. 1.31 Decay ratio map for coupled neutronic and thermal-hydraulic stability of the Super LWR...

The presence of water rods reduces the density reactivity feedback effect due to the large time delay in the heat transfer to the water rods, and this affects the coupled neutronic and thermal-hydraulic stability. 

In summary, at the subcritical pressure operation during the pressurization phase, thermal criteria are more limiting due to dryout. The startup scheme prior to line switching is mainly determined by thermal criteria. The thermal-hydraulic stability criterion is satisfied by applying a sufficient orifice pressure drop coefficient. The coupled neutronic and thermal-hydraulic stability is also satisfied, since the power to flow rate ratio is low during this phase. 

The main differences of the HPLWR design compared to the Super LWR studied by the University of Tokyo and the thermal spectrum SCWR developed by Japanese industries are the three-pass core and the wire wrapped fuel assembly. The first coupled neutronic and thermal-hydraulic analyses of the core were performed for full load, steady-state conditions. They showed that the envisaged power profile and coolant density distribution are feasible. CFD analyses of coolant mixing inside assemblies as well as in the mixing chambers above and below the core predicted an acceptable temperature distribution at the inlet of each heat up step. Stress and deformation analyses of the reactor pressure vessel, the major reactor internals, and of the assembly boxes indicated areas for design optimization which are going to be addressed with the next design iteration. 

J. Zhao, P. Saha and M. Kazimi, Coupled Neutronic and Thermal-Hydraulic Ont-of-Phase Stability of Supercritical Water Cooled Reactors, Proc. ICAPP 06, Reno, NV, Jtme 4-8, 2006, Paper No. 6424 (2006)... 

Chapter 2 covers design and analysis of the core and fuel. It includes core and fuel design, coupled neutronic and thermal hydraulic core calculations, subchannel analysis, statistical thermal design methods, fuel rod design, and fuel rod behavior and integrity during transients. 

Chapters 3-5 treat the plant system and behaviors. They include system components and configuration, plant heat balance, the methods of plant control system design, plant dynamics, plant startup schemes, methods of stability analysis, thermal-hydraulic analyses, and coupled neutronic and thermal-hydraulic stability analyses. 

Coupled neutronics/thermal hydraulics stability analyses of the STAR reactor at these plant equilibrium states at full and partial load will be required. Such analyses have been conducted already for the STAR-LM which shares the neutronics and thermal hydraulics properties of STAR-H2 reactor, - and stability has been demonstrated. 

The high temperatore core without the critical heat flux criterion (i.e. the MDHFR) was designed in 1998 [12]. The two-dimensiraial R-Z model of the core cannot accurately predict bum-up of fuel rods. The three-dimensional coupled neutro-nic-thermal-hydraulic core calculation was developed in 2003 [18]. It is shown in Fig. 1.9. This calculation considered the control rod pattern and fuel loading pattern [19, 20] and was similar to the core calculation for BWRs. But the finite difference code SRAC [21] was used for the three-dimensional neutronic calculation instead of a nodal code. The core design of the Super FR also adopted the three dimensional neutronic and thermal hydraulic coupled core bum-up calculation. 

Fig. 1.9 Three-dimensional neutronic and thermal-hydraulic coupled core calculation...

A. Yamaji, Y. Oka and S. Koshizuka Three-dimensional Core Design of SCLWR-H with Neutronics and Thermal-Hydraulic Coupling, Proc. Global2003, New Orleans, November 16-20, 2003, 1763-1771 (2003)... 

The core calculations consist of neutronic and thermal-hydraulic parts. These parts are coupled to evaluate the core characteristics such as the core power or coolant temperatures. 

The coupling of neutronic and thermal-hydraulic calculations is especially important for designing the Super LWR core. The density change of the coolant (and moderator) is large and sensitive to the enthalpy rise of the coolant as it flows from the core inlet to the outlet. On the other hand, the core neutronic characteristics strongly depend on the coolant and moderator density distributions. 

The COREBN code does not have the coupling function. Hence, the bumup calculations for one cycle of the core operation is divided into a number of bumup steps. Within each bumup step, the neutronic and thermal-hydraulic calculations are coupled by the core power and density distributions (within each bumup step, the coolant density distribution is assumed to be constant). These calculations are repeated until the core power distribution and the density distributions are converged. Once the convergence is obtained, the bumup step proceeds to the next step. For the coupling calculations, the macro-cross section sets of the fuel assemblies are prepared for different coolant and moderator densities and these are interpolated by bumups. 

The core calculations were also introduced, including the neutronic and thermal-hydraulic parts. The core thermal-hydraulic characteristics are unique and strongly coupled with the neutronic characteristics of the core. 

The main control module automatically prepares the inputs required for both neutronic and thermal-hydraulic calculations. It also produces the pin power distribution and bumup distribution, and then prepares the peak and average channels for each fuel assembly. The neutronic and thermal-hydraulic calculations are executed by the main control module internally and coupled to each other by pin power and coolant density distributions. 

The thermal hydraulic design of CAREM reactor core was carried out using an improved version of the 3D two fluid model realized in the THERMIT code. In order to take into account strong coupling of the thermal-hydraulic and neutronic characteristics of the core, THERMIT was linked with the neutronic code CITVAP. This coupled model makes it possible to produce a 3D map of power and thermal-hydraulic parameters at any moment of the bum-up cycle. 

The reactivity in point kinetics equations depends on time-dependent average fuel temperature and time-dependent distributions of coolant and moderator density, and hence this model couples reactor neutronics with thermal-hydraulics. This model receives the fuel temperature distribution from the fuel rod heat transfer model and the coolant density distribution and moderator density distribution from the two thermal-hydraulic models. Then, it generates the power distribution to the fuel rod heat transfer model. 

Preliminary reactor core concepts have been proposed at various institutions. Among them, the concept with mixed neutron spectrum proposed by SJTU [95,96] has achieved the special attention of Chinese researchers. The mixed spectrum SCWR core combines the merits of both thermal and fast spectra as far as possible. The basic idea is to divide the reactor core into two zones with different neutron spectra. In the outer zone, the neutron energy spectrum is similar to that of PWRs. To assess the performance of the reactor core, a coupled neutron-physics and thermal-hydraulics analysis was conducted [96]. 

In the case of a BWR, the operation point of the average heated fuel assembly should correspond to a decay ratio less than 0.5 for a single-channel density wave oscillation, and a decay ratio less than 0.25 should correspond to the coupled thermal-hydraulic/neutronic density wave oscillation. Furthermore, the whole operation range, also including hot fuel assemblies, should be in the linear stable region of the stability map. 

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