Fuel centerline temperature

Recently, however, limited use of best estimate plus uncertainty analysis methods has been undertaken. This is consistent with the international trend toward use of such methods. In this approach, more physically realistic models, assumptions, and plant data are used to yield analysis predictions that are more representative of expected behavior. This requires a corresponding detailed analysis of the uncertainties in the analysis and their effect on the calculated consequences. Typically, the probability of meeting a specific numerical safety criterion, such as a fuel centerline temperature limit, is evaluated together with the confidence limit that results from the uncertainty distributions associated with governing analysis parameters. The best estimate plus uncertainties approach addresses many of the problematic issues associated with conservative bounding analysis by... 

Higher fuel temperatures effect a pronounced transport of cesium from the pellet center to the outer regions. This can be seen from Fig. 3.23., representing measurements performed on a fuel pellet which, after a bumup of about 20 MWd/kg U, was subjected to a power ramp with a terminal power of 404 W/cm (corresponding to a fuel centerline temperature of about 1970 K) for about 48 hours subsequently, irradiation was continued at the lower, normal power level to a final bumup of about 22 MWd/kg U. The depletion of radioactivity in the central region is particularly pronounced for Cs, again reflecting the differences in the production mechanisms for both cesium isotopes. 

The thermal conductivity of a fuel is of importance in the computation of the fuel centerline temperature. The thermal conductivities of MOX and UO2 decrease as functions of temperature up to temperatures around 1527 and 1727°C, respectively, and then fliey increase as the temperature increases (see Fig. 18.1). In general, the thermal conductivity of MOX fuel is slightly lower than that of UO2. In other words, the addition of small amounts of PUO2 decreases the thermal conductivity of the mixed fuel. However, the thermal conductivity of MOX does not decrease significantly when... 

In their analysis, the thermal conductivity values have been calculated based on Eq. [18.3], which requires the measured values of the thermal diffusivity, specihc heat, and density of these solid solutions. In the current study, the correlation developed by Jain et al. (2006), which is shown as Eq. [18.4], has been used in order to calculate the thermal conductivity of Th02 fuel for the purpose of the calculation of the fuel centerline temperature. In Eq. [18.4], T is the temperature in K. 

The analyzed reactor core consists of 336 high-efficiency reentrant fuel channels. The inlet temperature of the coolant is 350°C at a pressure of 25 MPa, and the outlet temperature is 625°C. As a conservative approach, the thermal power corresponding to a fuel channel with the maximum thermal power was used in order to calculate the fuel centerline and sheath temperatures with the use of a one-dimensional thermal-hydraulic code. The temperature variation of the fuel hottest element in the radial direction is shown in Fig. 18.23. The maximum fuel centerline temperature of the UO2 fuel reaches 2196°C in the hottest fuel element of a fuel channel with a maximum thermal power of 10.23 MWn,. The temperature profiles of the coolant and the cladding (ie, CLaDding Temperature (CLDT)), as well as the Heat Transfer Coefficient (HTC) are shown in Fig. 18.24. 

For the same fuel bundle design and heat flux profile, a high thermal conductivity fuel such as UC or UO2—sdicon carbide (SiC) shows significantly less fuel centerline temperatures. Figs. 18.26 and 18.27 show the temperature variation of the fuel hottest... 

Abdalla, A., Peiman, W., Pioro, I., Gabriel, K., 2012. Sensitivity analysis of fuel centerline temperature in SCWRs. In Proceedings of the 20th fntemational Conference on Nuclear Engineering (ICONE-20) — ASME 2012 POWER Conference, July 30—August 3, Anaheim, CaUfomia, USA. Paper 54530, 9 pp. 

For. some preliminary calculations, the fuel pellet diameter and active core height are taken as 8 mm and 3.66 m (12 ft), respectively. Design considerations can change both of these dimensions when needed. Some of the considerations for the fuel shape are that it can be easily manufactured, as in an extrusion process, and that it has a reasonable diameter with uniform fuel density. The fuel height and diameter should be made such that the rod is structurally rugged and able to withstand flow induced vibrations. Finally, thermal considerations require that the fuel thickness be compatible with the power density and the available heat removal rate to avoid excessive fuel centerline temperatures. 

With the fuel surface temperature and internal heat generation rate known, the fuel centerline temperature is calculated from ... 

Case 2 evaluated the increased fuel centerline temperature that would result if aluminum cladding is replaced with stainless steel. The results indicate that using stainless steel cladding will only cause the fuel centerline temperature to increase by 10 K above that obtained with aluminum cladding. Thus, at these low fuel power densities it makes little difference from a thermal-hydraulic viewpoint what cladding material is used for the fuel pin. 

The fuel centerline temperature data in LOFT large break experiments LP-02-6 and LP-LB-1 were analyzed to determine the bias at peak cladding temperature (PCT) in the cladding exterior surface-mounted thermocouples and the effect of the thermocouple cable on the thermal behavior of the cladding. A statistically determined bias of 11.4 K 16.2 K was found in the cladding thermocouples (measured less than actual PCT). The fin effect of the thermocouple cable was determined to be small and within the uncertainty of the data in the blowdown phase of the transients in which PCT occurred. The PCT in LOFT experiments LP-02-6 and LP-LB-1 was determined to be 1104.8 K and 1284.0 K respectively. 

Figure 11. Calculated and Measured Fuel Centerline Temperature in LP-02-6...

Given the fuel pin power generation rates and the surface heat transfer coefficients, the CARTS fuel code (discussed in Ref. 3) was employed to predict the fuel centerline temperatures. At a core power of 15 MW (fuel pin LHGR of 3.9 kW/ft), the predicted fuel centerline temperature is 320°C. For core powers of 30 and 45 MW, the centerline temperatures are 663 and 1043°C, respectively. Thus, of the core powers considered here, only a core power of 15 MW yields fuel temperatures less than the 450°C disassociation temperature for the UO3 fuel. As interpolated, the core power could likely approach 21 MW with the given assumptions. 

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. 

The linear heat generation rate affects the fuel centerline temperature strongly. The average linear heat generation rate and the average coolant core outlet temperature are assumed to be 170 W/cm and 500°C, respectively. All other core characteristics are assumed to be the same as the reference core design summarized in Table 2.4 [9]. 

Flow induced vibration Fuel centerline temperature Gap conductance Fission gas release fraction Gas plenum pressure Thermo-mechanical behavior of cladding... 

Overheating of fuel pellets Fuel centerline temperature (°C) ... 

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... 

The Baron model [15], which assumes considerable degradation of thermal conductivity by bumup, is used for the thermal conductivity of MOX fuel. It has been known to be more conservative than MATPRO-11 model [16] in most aspects of the fuel rod thermal behavior such as fuel centerline temperature and fission gas release [17]. The amount of fission gas generation in MOX fuel is assumed to be the same as that in UO2 fuel. Fission gas release is predicted by the White and Tucker-Speight model [18,19]. The Studsvik model [20] is adopted as it is a representative fuel pellet swelling model. Other material properties are taken from the MATPRO-11 model. 

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