Hot channel factors for rod temperature calculations in LMFBR assemblies

Carelli, Mario D.; Friedland, A.J.

doi:10.1016/0029-5493(80)90027-8

Cited by 12 publications

(12 citation statements)

References 7 publications

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“…(9) with the values in Tables 10 and 11. The results of RTDP are shown in Table 12, where ' PF is 18.80 C. This value is 2.6% larger than the value determined by the Monte Carlo procedure.…”

Section: Results and Comparisonmentioning

confidence: 95%

“…Nominal results have to be corrected to account for such effects as calculation approximations, measurement errors, instrumentation ÓAtomic Energy Society of Japan accuracy, manufacturing and fabrication tolerances, correlation uncertainties and so on. 9) Various methods have been developed and utilized successfully to evaluate and combine the engineering uncertainties for most types of nuclear reactors except for Super LWR. These methods treat the uncertainties by using values directly or by using dimensionless factors, and combine all the uncertainties by different methods including the deterministic method, the semi-statistical method, and the statistical method.…”

Section: Introductionmentioning

confidence: 99%

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Development of Statistical Thermal Design Procedure to Evaluate Engineering Uncertainty of Super LWR

Yang

Oka

Liu

et al. 2006

Journal of Nuclear Science and Technology

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The thermal performance of a nuclear reactor core contains various engineering uncertainties. These uncertainties often arise from calculation, measurement, instrumentation, manufacture, fabrication and data processing. Statistical techniques are useful to evaluate and combine these uncertainties in the thermal design of nuclear reactors. In this paper, a statistical method is developed and employed in the thermal design of the supercritical pressure light water reactor (Super LWR) to evaluate the statistical engineering uncertainties. This method is referred as the Monte Carlo Statistical Thermal Design Procedure for Super LWR (MCSTDP). This method uses the maximum cladding surface temperature (MCST) as a crucial criterion and a sub-channel code is utilized to perform the core thermal hydraulic analysis. The engineering uncertainties are considered with strict respect to the 95/95 limit of Super LWR. The main purpose of this paper is to establish the statistical evaluation methodology. The engineering uncertain for the thermal design of Super LWR is evaluated by using this method to get an approximate quantification. The results are compared with those of the Revised Thermal Design Procedure (RTDP) of Super LWR.

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Section: Results and Comparisonmentioning

confidence: 95%

Section: Introductionmentioning

confidence: 99%

Development of Statistical Thermal Design Procedure to Evaluate Engineering Uncertainty of Super LWR

Yang

Oka

Liu

et al. 2006

Journal of Nuclear Science and Technology

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“…A more complete set of hot channel factors for the CRBRP heterogeneous core design is available [29], which takes into consideration the blanket assembly hot channel factors and other aspects not considered in the earlier work. The FFTF hot channel factor analysis [27] was similar to the early CRBRP analysis but we focus here on the early CRBRP approach since it came later in design and applies to a power reactor design.…”

Section: Crbrp and Fftf Hot Channel Factorsmentioning

confidence: 99%

“…Methods used are strongly influenced by precedents used in the licensing process; indeed, LWR precedents have already influenced fast reactor methods. Since we have available the extensive uncertainty analyses for FFTF [27] and CRBRP [28][29][30], these are used here as guides to the methodology for calculating uncertainties. Analyses for later designs will continue to become more elaborate, particularly as more experimental data become available, more advanced numerical models are developed, and as incentives grow to decrease design margins.…”

Section: Hot Channel Factorsmentioning

confidence: 99%

Core Thermal Hydraulics Design

Waltar

Todd²

2011

Fast Spectrum Reactors

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In the previous chapter we explored the methodology for determining the temperature field for a single fuel pin. Since a typical fast reactor core comprises several thousand fuel pins clustered in groups of several hundred pins per assembly, a complete thermal-hydraulic analysis requires knowledge of coolant distributions and pressure losses throughout the core. This chapter will address these determinations.We focus first on the coolant velocity and temperature distributions within an assembly. Then the discussion is broadened to include coolant velocity and pressure distributions throughout the reactor vessel. Having established these temperature and flow fields, the important design question of hot channel factor determination is then addressed. Assembly Velocity and Temperature DistributionIn Chapter 9 we presented a simple relation for the average axial coolant temperature rise, based on an average bundle mass flow rate [Eq. (9.37)]. The problem is actually more complicated, for several reasons. First, the flow areas for the three types of channels in Fig. 9.13 are different, so the velocity and mass flow rates in each are different. Second, there is a radial power distribution, or power skew, across individual fuel assemblies. Third, there is crossflow, i.e., flow transverse to the axial direction, between assembly channels. In the first part of this section a simplified approximate velocity distribution will be derived that accounts for the variation in flow area between channels. In the second part the more complex set of equations accounting for crossflow and turbulent mixing between channels will be presented. Approximate Velocity DistributionAn approximate velocity distribution, or flow split, between flow channels in Fig. 9.13 can be obtained from the physical fact that the axial pressure drop across each channel of a fuel assembly must be the same since the flow begins and ends in common inlet and outlet plenum regions. This technique was used by Novendstern [1] in his method for evaluating pressure drop in a fast reactor fuel assembly and A. Waltar (B) Pacific Northwest

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“…In order to estimate the maximum cladding and coolant temperatures due to uncertainties and power peaking factors, the uncertainty factors listed in Table 3.2 were used [39]. Originally, these factors were computed in detail for CRBR cores [51,52] and have been modified for metal fueled cores in an "overly conservative fashion" [39]. The overall hot spot factor for the channel in consideration is obtained by a combination of the direct factors and statistical factors.…”

Section: Corementioning

confidence: 99%

Thermal hydraulic aspects of an unconventional liquid metal reactor

Sökmen

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Hot channel factors for rod temperature calculations in LMFBR assemblies

Cited by 12 publications

References 7 publications

Development of Statistical Thermal Design Procedure to Evaluate Engineering Uncertainty of Super LWR

Development of Statistical Thermal Design Procedure to Evaluate Engineering Uncertainty of Super LWR

Core Thermal Hydraulics Design

Thermal hydraulic aspects of an unconventional liquid metal reactor

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