Fuel Rod Design Criteria

The fuel rod design criteria of the Super FR for each failure mode are summarized in Table 7.5 [1]. Each criteriOTi is explained below. 

As one thermal design criteriOTi to prevent overheating of the fuel pellet, melting of the pellet centerline should be avoided even considering various uncertainties. The fuel centerline temperature of 1,900°C and the MLHGR of 39 kW/m are used as 

Thermal design criteria Fuel centerline temperature 1,900°C 

Hydrodynamic design criteria Flow dynamic design consideration 0.02 MPa 

Thermo-mechanical design criteria Pressure difference 1/3 Buckling collapse 

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The fuel rods are designed to satisfy the fuel rod design criteria for rod bumup levels up to the design discharge bumup using the extended bumup design methods described in the Extended Bumup Evaluation report (Reference 6.9). 

Fuel Rod Design Table 7.4 Fuel rod design criteria for LMFBRs. (Taken from [1]) 455... 

The fuel rod design criteria except those of collapse and creep collapse are investigated with single channel thermal hydraulic analysis and fuel rod thermomechanical analysis. Design criteria are assessed by the FEMAXI-6 code as is also used for the Super LWR (see Chap. 2). Figure 7.12 shows the interaction of the fuel... 

Fuel rod failure modes and associated fuel rod design criteria were established. As an example of the fuel rod design, the parameters were determined with those criteria including thermo-hydrodynamic and thermo-mechanical considerations. 

The four basic design criteria in the fuel rod design are as follows, for both normal and abnormal transients ... 

Among the criteria, the criterion for the peak cladding temperamre had been estimated by simple but conservative calculations. However, it is expected to depend on the material of the cladding and the fuel rod design. Based on the above requirements, the following four principles are adopted to derive rationahzed new criteria for abnormal transients. 

The fuel rod cladding material of the Super LWR is being developed and tested. For fuel rod design of the Super LWR in the cmicept development phase, typical austenitic stainless steels or Ni-alloys are applied as described in Chap. 2. The principle of the safety criteria for fuel rod integrity is shown in Table 6.6. Since heat transfer deterioration is a much milder phenomenon than boiling transition, the minimum deterioration heat flux ratio was eliminated from the transient criterion related to fuel rod heat-up [7]. The types of abnormalities are separated into loss of cooling and overpower . 

Strain controlled limit has been applied in LMFBRs rather than load controlled limit. Inelastic or creep strain have been used as design criteria in LMFBRs. In the Fast Flux Test Facility or CRBRP fuel rod designs, inelastic hoop strain was limited to 0.1% on average and 0.2% as the peak for normal operation conditions, 0.3% at an anticipated transient, and 0.7% at an unlikely event [14]. Inelastic strain limit was an earlier approach used for a failure criterion and is a straightforward concept. 

Severe emergency core cooling system criteria require that the builders of water reactors increase the linear feet of fuel in the reactor core for the same power in order to reduce LOCA (loss-of-cooling accident) fuel temperatures. In the unit described here, an assembly with a 16 x 16 fuel rod array of smaller diameter rods is used in the same assembly envelope that was occupied by a 14 x 14 assembly in earlier designs. This results in a maximum linear heat rate decrease in the assembly of about 25%. 

The fuel rods are to be internally pressurized with helium gas as in BWRs and PWRs. The initial internal pressure of the fuel rods should be optimized to minimize the stresses and especially the pressure difference on the cladding. However, the internal pressure should not exceed the normal operating coolant pressure (25 MPa) to prevent any creep deformations that causes the gap between the pellet and cladding to increase. The four basic design criteria were determined to ensure the fuel integrity at all anticipated transients based on simple, but conservative evaluations [30]. 

However, such conservative criteria severely limited the plant operability during anticipated transients. In order to maximize the economical potential of the Super LWR and Super FR, and minimize the R D efforts, the criteria were rationalized based on detailed fuel analyses. The FEMAXI-6 code [31] for LWR fuel analyses was used for the study. The principle of rationalization of the criteria for anticipated transients of Super LWR was developed [32, 33]. The design and integrity analysis of the Super LWR fuel rods is summarized in ref. [34]. 

The fuel and core design is the central issue for a nuclear power plant (NPP). This chapter describes the design concepts of the Super LWR core including the fuel rod and fuel assembly designs. The core characteristics are explained together with the design method, criteria, and the research and development subjects to comprehensively develop these concepts. 

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. 

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