Comparison of Rupture Data from Irradiated Fuel Rods and Unirradiated Cladding

Hobson, D.O.; Osborne, M.F.; Parker, G.W.

doi:10.13182/nt71-a30845

Cited by 18 publications

(3 citation statements)

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“…Data points from the work of Emery et al 12 for two heating-rate ranges, i.e., [8][9][10][11][12][13][14][15] and 45-70°C/s, are included in the figure.…”

Section: Effect Of Heating Rate On the Deformation Behavior Of Zircalmentioning

confidence: 99%

Pilot-Lot Fabrication of ZPPR Oxide-Rod Elements.

Ayer¹,

Hins²,

McCuaig³

1970

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“…Data points from the work of Emery et al 12 for two heating-rate ranges, i.e., [8][9][10][11][12][13][14][15] and 45-70°C/s, are included in the figure.…”

Section: Effect Of Heating Rate On the Deformation Behavior Of Zircalmentioning

confidence: 99%

Pilot-Lot Fabrication of ZPPR Oxide-Rod Elements.

Ayer¹,

Hins²,

McCuaig³

1970

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“…High temperature creep data is required to analyse the cladding behaviour during postulated LOCA conditions to assess the performance of cladding and to assure the Keywords: PHWR fuel; accident behaviour; high temperature creep; ballooning, loss of coolant accident (LOCA) integrity of the fuel pin. Traditionally ballooning and burst studies on Zircaloy cladding reported in literature have been mostly carried out using internally pressurized unirradiated cladding tubes heated under argon or vacuum [1][2][3][4][5][6][7][8][9]. There are only a few reported works on irradiated cladding [6].…”

Section: Introductionmentioning

confidence: 99%

“…Traditionally ballooning and burst studies on Zircaloy cladding reported in literature have been mostly carried out using internally pressurized unirradiated cladding tubes heated under argon or vacuum [1][2][3][4][5][6][7][8][9]. There are only a few reported works on irradiated cladding [6]. In the present work actual irradiated (non-failed) fuel pins discharged from operating reactors have been directly used for the experiment.…”

Section: Introductionmentioning

confidence: 99%

Study of creep deformation of irradiated zircaloy cladding by isothermal heating of irradiated PHWR fuel pins in temperature range 700°C–900°C

Sah

Viswanathan

Anantharaman

2010

Trans Indian Inst Met

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Fuel pins removed from actual irradiated fuel bundles discharged from Pressurised Heavy Water Reactors (PHWRs) have been used for experimental study of high temperature creep deformation as functions of cladding temperature and internal fission gas pressure. Experiment consisted of localized heating of 100 mm long segment of the fuel pins in a furnace in inert atmosphere at temperatures 700 o C, 800 o C, 850 o C and 900 o C for 10-15 minutes. The internal pressure and the total void volume in the fuel pins were estimated by puncture test on sibling pins from the same fuel bundle. After the heating experiment the diameter of the pin along the length was measured to obtain the diameter increase due to high temperature creep. Analysis of the experimental data for fuel pins with internal pressure 0.55 ± 0.05 MPa, provided the following empirical correlation for creep rate of the cladding as a function of temperature:Creep rate (s -1 ) = 2.23 x 10 10 x exp (-305500/RT), for temperatures in the range of 800 o C -900 o C, where, R is gas constant, 8.314 J/mol K and T is temperature in K. For fuel pins with different internal fission gas pressures, the correlation obtained for the cladding strain as a function of internal pressure (at room temperature) was Cladding strain = 118.22 x 10 -3 exp (0.53P), for temperature = 900 o C where, P is the internal pressure in MPa at room temperature. This paper presents the details of the experiment and the results.

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Verlust des Reaktorkühlmittels

Smidt¹

1979

Reaktor-Sicherheitstechnik

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Comparison of Rupture Data from Irradiated Fuel Rods and Unirradiated Cladding

Cited by 18 publications

References 0 publications

Pilot-Lot Fabrication of ZPPR Oxide-Rod Elements.

Pilot-Lot Fabrication of ZPPR Oxide-Rod Elements.

Study of creep deformation of irradiated zircaloy cladding by isothermal heating of irradiated PHWR fuel pins in temperature range 700°C–900°C

Verlust des Reaktorkühlmittels

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