Manifestations of nonlinearity in fuel center thermocouple steady-state and transient data: implications for data analysis

Lanning, D.D.; Barnes, B.K.; Williford, R.E.

doi:10.2172/6321877

Cited by 7 publications

(8 citation statements)

References 2 publications

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“…This has also been shown by the analysis of scram data. ( 19 ) Since the thermal expansion is localized within each distinct fragment, there is no driving force to pull the fuel away from the cladding •as the power is reduced. There should be no tension in a medium composed of separated fragments.…”

Section: • •mentioning

confidence: 99%

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

Williford¹,

Mohr²,

Lanning³

et al. 1980

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2a Cladding Elongation Data at Beginning-of-Life 2b Comparison of Cladding Elongation Data to Code Calculations 3 Flow Diagram for Program BM2DNS 4a Crack Pattern with Circumferential Crack. 4b Radial Cracks Only • 5 Crack Pattern of Pellet 39 from Rod 6 of IFA-431 6 Thermal Conductivity Factor Versus Hydrostatic Stress 7 The Pellet-Cladding Gap Under Hydrostatic Conditions 8 Radial Elastic Modulus as a Function of Fuel Available Void 9 Axial Elastic Modulus as a Function of Total Available Void 10 Relationship Between Radial and Axial Elastic Moduli for Cracked Pellets .

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Section: • •mentioning

confidence: 99%

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

Williford¹,

Mohr²,

Lanning³

et al. 1980

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“…( 1 ) In that report we showed, in effect, that centerline temperature versus power data can be explained by constrained combinations of fuel and gap resistance spanning the full range of physical possibility for each resistance separately. Yet predicted transient cladding behavior during the loss-of-coolant accident is significantly affected by the resistance partition that is chosen.…”

mentioning

confidence: 86%

“…The analytical solution to the transient heat transfer equation is exact. The transient equation is simply ( 1) where Tcool is coolant temperature and Q(t) is heat generation (in watts) . First of all, we will consider the effect of a temperature dependence for the lumped parameter conductance, since all the parameters of a real fuel rod are in fact temperature dependent.…”

Section: The Constant Parameter Modelmentioning

confidence: 99%

A Procedure for the Qualitative Interpretation of Fuel Centerline Thermocouple Response to Step-Power Decreases

Author¹

1979

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“…1000 q where: R = thermal resistance, (K-m/W) Tcl = local centerline temperature (K) Tcool = coolant temperature (K) q = local linear heat generation rate (kW/m). The concept of thermal resistance previously has been described in detail (Lanning et al, 1979); thus, only a brief summary will be given here. The total local resistance of the fuel rod is the sum of the thermal resistance across the cladding, the resistance across the fuel-to-cladding gap, and the resistance through the fuel.…”

Section: Fission Gas Releasementioning

confidence: 99%

Evaluation of the in-pile pressure data from instrumented fuel assemblies IFA-431 and IFA-432.

Bradley¹,

Cunningham²,

Lanning³

et al. 1979

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This report includes results of the examination of the in-pile pressure data from instrumented test assemblies IFA-431 and 432. The pressure data have been used to estimate the fission gas release fraction as a function of fuel burnup. Included are comparisons of the estimated release functions and those predicted by three fission gas release models using the experimental temperature histories of the fuel rods. These comparisons show that fuel temperature is the primary factor in determining fission gas release and that burnupenhanced fission gas release is not important in U0 2 fuels irradiated to 1700 GJ/kgU (20,000 MWd/MTM). v • SUMMARY AND CONCLUSIONS This report includes the internal pressure data from fuel rods in two of the Nuclear Regulatory Commission/Pacific Northwest Laboratory's instrumented fuel assemblies. Three rods in each assembly were equipped with pressure transducers for monitoring fission gas release during irradiation. The experimental parameters were consistent with current light water reactor fuel and include fuel temperature, fuel density, fuel stability with respect to densification, and burnup. Instrumented fuel assembly 431 (IFA-431) was discharged after achieving burnups in excess of 350 GJ/kgU (4 GWd/MTM). Assembly 432 is still being irradiated, but current exposures in the IFA-432 rods are in excess of 1700 GJ/kgU (20 GWd/MTM). Meaningful pressure data was obtained from five of the six pressure transducers. Data from all five shows a pressure decrease during the initial stages of irradiation, which is due to the combined effects of fuel densification and helium loss from the fuel rod system. The relative contribution from each of these could not be established and, thus, a range of gas release fractions was calculated from the pressure data. In all cases, the estimated gas release was less than 15%. The estimated release fractions were compared to the predictions from gas release models currently being used in the GAPCON and FRAPCON series of steadystate fuel performance computer codes. Specifically, the GASREL (Beyer, et ale 1975), FGASRL (Reymann, ed., 1978), and ANS54 (Rausch and Panisko, 1979), subroutines for fission gas release were used in conjunction with the measured temperature/power histories of each rod. From these comparisons, the following has been concluded regarding fission gas release from U0 2 fuels. • Fuel temperature is the predominant factor in fission gas release. • Fuel density has only a minor influence on fission gas release.

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Manifestations of nonlinearity in fuel center thermocouple steady-state and transient data: implications for data analysis

Cited by 7 publications

References 2 publications

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

A Procedure for the Qualitative Interpretation of Fuel Centerline Thermocouple Response to Step-Power Decreases

Evaluation of the in-pile pressure data from instrumented fuel assemblies IFA-431 and IFA-432.

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