Core Design Criteria

The core design criteria are summarized in Table 1.3. The maximum linear heat generation rate (MLHGR) at rated power is 39 kW/m. It is slightly lower than those of PWRs (42.6 kW/m) and BWRs (44 kW/m) due to the high average coolant temperature. The fuel centerline temperature stays nearly the same as that of LWRs. The fission gas release rate from the fuel pellets is similar to that of LWRs. The fuel design of the Super LWR follows that of LWRs. 

All the reactor coolant is purified after condensation in the once-through coolant cycle of the Super LWR and Super FR. This differs from BWRs and PWRs in which reactor coolant is circulated in a closed loop as recirculating coolant and primary loop coolant, respectively. The purity of reactor coolant is therefore different from that of LWRs. 

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. 

Maximum linear heat generation rate (MLHGR) at rated power 39 kW/m 

Maximum cladding surface temperature at rated power 650°C for Stainless Steel cladding 

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Among basic thermal-hydraulic core design criteria, fully satisfied in the safety analyses is the minimum departure from nucleate boiling ratio (DNBR) must be always higher than 3.0, for Level 1 (normal events). Level 2 (unplanned events probability higher than 3.10 ... 

The nuclear design of the KALIMER core is based on the core design criteria discussed below and on the constraints given in Table XX-2, derived from the currently available metal fuel database. Altogether, they define the important performance parameters of the reactor that assure proper performance and safety of fuel and core. 

Therefore, in the core design of the Super LWR, it is possible to eliminate the CHF from the core design criteria. In this case, the occurrence of the heat transfer deterioration may be permitted as long as the fuel cladding temperature is kept below its limit. If the core design of the Super LWR were limited by the CHF to prevent the heat transfer deterioration, the core outlet average coolant temperature... 

The core design criteria of the Super LWR are significantly different from those of PWRs there is no criterion like the minimum departiue from nucleate boiling ratio (DNBR) and the heat transfer deterioration at supercritical pressure is not such a violent phenomenon as DNB at subcritical pressiue. 

However, it is important to recognize that a shorter yielding length results in higher inelastic strains in the core which will result in higher forces in the braces, an important design criterion. 

The responsibility criterion also requires that Reactor Operations establish the necessary interface agreements with the SRL for needed support. The staff found that these interface agreements have been defined in SCS-RGP-89-0020, "Centralization of Reactor Safety Responsibility" (Reference 3). These agreements establish the responsibilities for the flow and exchange of various information that is needed by the various organizations to accomplish the core design and safety analysis. 

The staff concludes that the organizational changes recently implemented by WSRC provide adequate alignment of responsibilities and interface channels to satisfy the restart criterion for core design responsibility. 

Removing the critical heat flux criterion (i.e., the MDHFR) from the core design and taking the MCST criterion makes it possible to raise the outlet coolant temperature of the Super LWR and Super FR to that of the supercritical FPP. The high enthalpy rise and low coolant flow rate are advantages of the once-through coolant cycle. 

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 startup curves based on the results of the analysis are shown in Fig. 5.77. The condenser pressure and dissolved oxygen level in the reactor are plotted by referring to the startup curves of BWRs. In the system transient analysis, the MCST is also calculated in order to consider the fuel rod integrity during the startup. The MCSTs in the three hot channels are shown in Fig. 5.78. They do not exceed 650°C which is the criterion for the three-dimensional core design (see Chap. 7). 

The same plant characteristics as in Chap. 4 are used for the safety analyses. The initial conditions are shown in Table 6.10. The hottest cladding temperature of 650°C is the same as the criterion applied in the three-dimensional core design... 

Table 7.29 [26] shows the fuel rod design results. The fuel rod diameter and P/D are 7.0 mm and 1.16, respectively. The rod arrangements in the seed and blanket assemblies are the same as those of the 1,000 MWe class design as shown in Fig. 7.57 [26]. The fuel loading pattern and flow distribution design are shown in Fig. 7.58 [26]. The distributions of the MCST calculated by single channel analyses and subchannel analyses are shown in Fig. 7.59 [26]. The highest value is kept well below the criterion of 650°C by modifying the subchannel shapes as introduced in Sect. 7.6. The core design results are summarized in Table 7.30 [26]. Based on the reference core design, two important performances are improved in Sects. 7.8.2 and 7.8.3.

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