Core thermal hydraulics control

The orifices in BWRs are mainly used for improving the core thermal-hydraulic stabilities. Generally, the BWR channel stability improves when the pressure drops and inertia in the single-phase flow region are increased. This is why inlet orifices are used in BWRs. For LMFBRs, inlet orifices are used to control the coolant flow rate to the fuel assemblies to effectively cool the fuel. 

The CAREM reactor under development in Argentina is a 100 MW(th) (about 27 MW(e)) design based on natural circulation. It has an integrated primary circuit comprising the core, steam generators, control rods with their drive mechanisms and the entire primary coolant. Several experiments examining core neutronics and thermal hydraulics have been conducted in test facilities. The construction of a prototype is planned. 

Such interactions can only occur, however, when the volatile fission products and the primary aerosols appear simultaneously in the primary system, in spite of the large differences in their volatilization behavior. As was discussed above, uniform thermal-hydraulic conditions do not prevail within the reactor core during a severe accident (for example, the peripheral fuel rods may fail relatively late in the accident sequence, at a point when a large part of the central rods may already be molten) and it can be assumed that the broad time-envelope of significant release of structural aerosols will encompass the release of the volatile fission products. However, as was mentioned in Section 7.3.1.2., the amount of primary aerosols formed and the timing of their formation depend highly on the specific accident sequence this is particularly true for the control rod materials. 

TWINKLE is a multidimensional spatial neutron kinetics code, whieh is patterned after steady-state codes currently used for reactor core design. The code uses an implicit finite-difference method to solve the two-group transient neutron diffusion equations in one, two, and three dimensions. The code uses six delayed neutron groups and contains a detailed multi-region fuel-clad-coolant heat transfer model for calculating point-wise Doppler and moderator feedback effects. The code handles up to 2000 spatial points and performs its own steady-state initialisation. Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as input basic driving functions, such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. Various edits are provided (for example, channel-wise power, axial offset, enthalpy, volumetric surge, point-wise power, and fuel temperatures). 

Over the next several years, the PRA was detailed to the point it included detailed fault trees of the mechanical, electrical, and instrumentation and control systems and the scope was expanded to include shutdown, fire, flood events, and large release frequency and off-site dose quantifications. Core damage frequency PRA was supported by extensive plant thermal-hydraulic analysis to justify success criteria. Extensive testing and thermal-hydrauUc analysis, to support containment integrity during core melt sequences, underpirmed the large release PRA. 

Related to the Reactor Core Coolant System, modeling and qualification are boosted by the testing performed in a high pressure natural circulation rig (CAPCN), covering thermal hydraulics, reactor control and operating techniques. Several sets of experiments were conducted at nominal and ex-nominal conditions. The CAPCN facility may also test the second shutdown system and some in-vessel instrumentation probes. 

The CPT is responsible for integration and implementation of the core design and safety analyses, which includes responsibility for design specifications, verification that fuel received is within specifications, verification that core conditions lie within core design and safety analysis criteria, operating limits, physics ihd thermal-hydraulic data for operational assessment and operational data for core design and safety analyses. The CPT is also responsible for core performance analysis which includes estimated critical position predictions, shutdown reactivity predictions, control rod worth curves, operational anomaly analysis and observed performance summaries. 

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. 

The plant starts up by heat entering from the primary pump and the system temperature rises to 350°C from the cold shutdown state. Under this condition, all parts of the system, including the recirculation line in the water system, are uniformly heated. Then, a neutron absorber at the center of the core is withdrawn. At temperatures below 350°C, the neutron absorber cannot be withdrawn by the self-connected mechanism using the thermal expansion difference between the stainless steel and Cr-Mo steel (Fig. 14). After withdrawal of the neutron absorber, the reflector is lifted up by the hydraulic system to reach critical condition at 350 C. A ficzy control system is employed for this approach and a fully automatic operation circuit is provided because no malfunction causes severe reactivity insertion as described previously. 

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