Postirradiation Examination of the Pm-3a Type 1 Serial 2 Core. Part 1. Postirradiation Examination of Fuel Tubes

Brown, John; Storhok, V.W.; Gates, J.E.

doi:10.21236/ad0655290

Cited by 4 publications

(8 citation statements)

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“…The spent fuel rod examination was requested due to the increasing fission product accumulation in the primary coolant system. The examination determined that the fission product buildup was due to leakage through inter-granular cracks in the outer cladding of the fuel rods (Brown et al, 1967).…”

Section: Refueling and Maintenancementioning

confidence: 99%

“…The Battelle Memorial Laboratory report on the reactor spent fuel rods (Brown et al, 1967) was reviewed to determine if the leaking fuel rods could have created new, different, or unexpected sources of or pathways for radiation exposure for support personnel at McMurdo Station. The leaking fuel rods could have caused the amount of radioactive waste generated at the NPP to be relatively greater than otherwise expected or could have caused increased radiation releases of radioactive materials through the stack.…”

Section: Refueling and Maintenancementioning

confidence: 99%

“…The radioactive waste packages would have contained fission products removed from the demineralizers of the primary coolant loop of the NPP. Thus, any fission products that leaked into the primary coolant through fractures in the cladding of the reactor fuel rods would have been removed by the demineralizer and processed by the radioactive waste control system (Brown et al, 1967). The exposure of support personnel to radiation from packages containing radioactive materials is presumed to have occurred during any DF year.…”

Section: Movement Of Packages Containing Radioactive Materials Outsidmentioning

confidence: 99%

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Upper-Bound Radiation Dose Assessment for Military Personnel at McMurdo Station, Antarctica, between 1962 and 1979, Revision 1

Dunavant¹,

Chehata²,

Morris³

et al. 2016

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Section: Refueling and Maintenancementioning

confidence: 99%

Section: Refueling and Maintenancementioning

confidence: 99%

Section: Movement Of Packages Containing Radioactive Materials Outsidmentioning

confidence: 99%

See 1 more Smart Citation

Upper-Bound Radiation Dose Assessment for Military Personnel at McMurdo Station, Antarctica, between 1962 and 1979, Revision 1

Dunavant¹,

Chehata²,

Morris³

et al. 2016

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“…Subsequent observation made this clearer e.g., by showing a similar effect of the water purity (as measured solution conductivity) on unirradiated and irradiated stainless steel (Figure 4 ). The initial reports of IASCC occurred in the early 1960s (Andresen & Ford, 1995;Andresen, Ford, Murphy, & Perks, 1990a;Armijo, Low, & Wolff, 1965;Brown, Storhok, & Gates, 1967;Bruemmer et al, 1999;Cheng, 1964Cheng, , 1970Cheng, , 1975Cowan & Gordon, 1977;Duncan et al, 1965;Hanninen & Aho-Mantila, 1988;Jacobs & Wozadlo, 1985;Multer, 1975;Pashos et al, 1964;Pasupathi & Klingensmith, 1981;Schaffer, 1962;Scott, 1994;Storrer & Locke, 1970;Was & Andresen, 1992a;2007) and involved intergranular cracking of stainless steel fuel cladding initiating from the water side. Post-irradiation mechanical tests, by comparison, were performed in inert environments at various strain rates and temperatures and exhibited mostly ductile, transgranular cracking.…”

Section: Plant Data On Iasccmentioning

confidence: 99%

“…IASCC is clearly not confi ned to a particular reactor design, material, component or water chemistry. For example, stainless steel fuel cladding failures were reported years ago in commercial PWRs and in PWR test reactors (Andresen et al, 1990;Brown Jr et al, 1967;Cheng, 1964Cheng, , 1970Cheng, , 1975Garzarolli, Rubel, & Steinberg, 1984;Hanninen & AhoMantila, 1988;Multer, 1975;Pasupathi & Klingensmith, 1981;Schaffer, 1962;Storrer & Locke, 1970). At the West Milton PWR test loop, intergranular failure of vacuum annealed type 304 stainless steel fuel cladding was observed (Cheng, 1970) in 316 ° C ammoniated water (pH 10) when the cladding was stressed above yield.…”

Section: Plant Data On Iasccmentioning

confidence: 99%

Irradiation-assisted stress corrosion cracking

Was¹,

Ashida²,

Andresen³

2011

Corrosion Reviews

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Irradiation-assisted stress corrosion cracking (IASCC) refers to the process(es) by which irradiation enhances the inherent susceptibility of alloys to stress corrosion cracking (SCC). This article begins with an introduction to the IASCC problem experienced in light water reactors, followed by a summary of existing plant and laboratory data that reveal its dependence on irradiation, environment and alloy parameters. The specifi c effects of irradiation on IASCC are classifi ed into two categories: the effects on water chemistry and the effects on microstructure. Water chemistry effects include radiolysis and its effects on corrosion potential, and effects of corrosion potential on IASCC. Microstructure effects include radiationinduced segregation (RIS), irradiated microstructure, swelling and creep, and hydrogen and helium generation. Leading mechanisms proposed to explain the role of radiation in the SCC process are: radiolysis and crack tip strain rate, grain boundary chromium depletion, irradiation hardening, localized deformation and RIS of minor elements. As nuclear power plants operate for longer, an increased incidence of SCC can be expected unless active mitigation steps are taken. Drawing on the accumulated understanding over the past four decades of the key processes believed to control IASCC, a set of attributes of an ideal, IASCC-resistant alloy are identifi ed as a fi rst step.

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Radiation-induced segregation in multicomponent alloys: Effect of particle type

Was

Allen

1994

Materials Characterization

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The problem of irradiation-assisted stress-corrosion cracking (IASCC) in reactor cores is currently being addressed using different types of particle irradiation (electrons, protons, and heavy ions) to study the effect of neutron damage. The effect of radiation damage of greatest interest to the IASCC problem is the segregation of impurities or redistribution of alloying elements in the vicinity of the grain boundary. Differences among the types of irradiation include particle type, temperature, dose, and dose rate. The different particle type results in different "effective" displacement rates due to fundamental differences in the radiation damage state. The effect of displacement efficiency is incorporated into existing models for radiation-induced segregation and compared with experimentally determined values. Comparisons between theory and experiment were made using neutron-, proton-, and ion-irradiated stainless steels. Results showed that the measured chromium depletion prohle is generally narrower and shallower than that calculated from theory. Agreement between experiment and model was better in almost all cases when the particle efficiency was taken into account. Agreement was best for proton-irradiated steels, and less so for neutron and heavy ion-irradiated steels. The origins of this difference are probably due to spatial and depth resolution of the scanning transmission electron microscopy and Auger electron spectrometry measurement techniques, respectively, the lack of knowledge of the material parameters in modeling, and the uncertainty in the true displacement rate. Nevertheless, beyond 2 nm from the grain boundary, the shape of the experimental and calculated profiles agree reasonably well.

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Postirradiation Examination of the Pm-3a Type 1 Serial 2 Core. Part 1. Postirradiation Examination of Fuel Tubes

Cited by 4 publications

References 0 publications

Upper-Bound Radiation Dose Assessment for Military Personnel at McMurdo Station, Antarctica, between 1962 and 1979, Revision 1

Upper-Bound Radiation Dose Assessment for Military Personnel at McMurdo Station, Antarctica, between 1962 and 1979, Revision 1

Irradiation-assisted stress corrosion cracking

Radiation-induced segregation in multicomponent alloys: Effect of particle type

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