Water rod

F Region with homogeneous mesophase with two-dimensional hexagonal structure water rods reversed structure. 

Non-lamellar lipid mesophases (Fig. 4) may also be identified by their characteristic small-angle diffraction pattern. The structure of the inverse hexagonal lipid-water mesophase (denoted as Hu) is based on cylindrical water rods, which are surrounded by lipid monolayers. The rods are packed in a two-dimensional hexagonal lattice with Bragg peaks positioned at... 

Fig. 9.7. Grand Gulf BWR control rod and fuel bundle cross sections (courtesy of General Electric Company and Nuclear Engineering International). O, Fuel rod water rod 0, tie rod.

The association between a surfactant-stabilized reverse structure and a gelatin-water rod network provides a organo-gel structure that has been called a three-dimensional cross-linked colloidal necklace which is dispersed in an oil phase [81]. Such structures can be used to immobilize enzymes as catalysts for organic synthesis such as esterification or hydrolysis [82] and for optical resolution of racemic mixtures [75]. 

Increased moderator/fuel ratio. The moderator/fuel ratio is substantially increased, mainly through the implementation of six water rods (Figure 4.7) and the reduction of the original fuel rod outer diameter to a smaller value. This leads to increased reactivity for this fuel design. 

The moderator temperature in the water rods should be below the pseudo-critical temperature to keep the moderator density high. Thin layer of zirconia (Zr02) is used for thermal insulation on the water rods. The thermal insulation also reduces the stress of stainless steel plates of water rods below allowable stress level. 

Moderator temperature in water rods = 384°C (pseudo critical temperature at 25 MPa)... 

Thermal spectmm core Many/Large water rods... 

Moderator temperature below pseudo-critical Insulation of water rod wall... 

A thin thermal insulation of Zirconia is provided between the water rods and fuel coolant channels. Gadolinia is used for compensating bum-up reactivity and axial power flattening. The control rods are the cluster rod type. The control elements are inserted in the guide tubes that are located in the central water rods. 

The core power of the Super LWR was found not to be sensitive to the feedwater flow rate due to the existence of many water rods. 

Boiling (and dryout) must be prevented in the water rods at subcritical pressures (in sliding pressure startup scheme). 

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. 

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. 

The results of the total loss of reactor coolant flow accident are explained in Fig. 1.40. The heat conduction to the water rods increases and the water rods serve as a heat sink . This heat conduction also thermally expands the water in the water rods and temporarily supplies water to the fuel channels. Thus, water rods serve as a water source also and enable the backup pmnps (AFSs) to have a realistic delay time. The results of loss of turbine load without turbine bypass transient are shown in Fig. 1.41. This is a type of pressurization event and an important one for... 

The SCRELA code was developed for large LOCA analyses for the SCFR, an early version of the Super FR [72, 73]. The SPRAT-DOWN, including the downward flow water rod model for the Super LWR, was extended to the SPRAT-DPWN-DP code for the large LOCA analyses [71]. The critical flow at supercritical pressure is not known. Then, the correlation at the subcritical pressure has also been used in the supercritical pressure for the LOCA analyses since the duration of supercritical pressure is very short. Both codes were verified in comparison with the REFLA-TRAC code. The SPRAT-DOWN code was applied to the small LOCAs of the Super LWR because the system pressure stays in supercritical region at the small LOCAs [71]. 

Top dome and water rods serve as an in-vessel accumulator Loss of flow mitigated by water rods Short period of high cladding temperature at transients Mild behavior at transients, accidents and ATWS Simple safety principle (keeping flow rate) due to once-through cooling cycle... 

One-dimensional bum-up and two-dimensional R-Z core calculation procedures for the fast reactor are found in refs. [108,109]. The LOCA analysis code, SCRELA was described in ref. [110]. The procedure for a simplified PSA is also described there. The single channel thermal-hydrauUc calculation code SPROD, the two-dimensional coupled core calculation scheme of the thermal spectram reactor with water rods and transient and accident analysis code at supercritical-pressure, SPRAT are described in ref. [111]. 

Safety analysis of the Super LWR is described in ref. [121]. The SPRAT-DOWN code for the analysis of downward flowing water rods and the SPRAT-DOWN-DP code for depressurization in an LOCA were prepared. The LOCA analysis of the Super LWR was performed in combination with SPRAT-DOWN-DP and the reflooding module of SCRELA. ATWS analysis is also described in ref. [121]. The momentum equation is included in the SPRAT-DOWN code from the ATWS analysis. The design of the two-pass core of the Super LWR and the safety analysis at subcritical pressure during startup are described in ref. [122]. 

The temperature difference between the moderator in the water rods and the coolant in the fuel chaimels is large, approximately 250°C. Without thermal insulator, thermal stress exceeds the tensile strength of typical stainless steel as shown in Fig. 1.65 [122]. Yittria-stabilized zirconia was developed for the thermal insulator... 

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