Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

Ott, L.J.; Robb, Kevin; Wang, Dean

doi:10.1016/j.jnucmat.2013.09.052

Cited by 234 publications

(74 citation statements)

References 5 publications

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“…Second option is by using thermal conductivities and specific heat of fuel and cladding as a temperature functions (the value changed based on temperature reference) that will be calculated by MATPRO sub routine [13]. The sub routine of MATPRO uses several equations for calculating thermal conductivities and specific heat of fuel and cladding as a temperature functions, so that MATPRO has been used in analysis in VVER [12] and LWR [14][15].…”

Section: Cobra-en Codementioning

confidence: 99%

Validation of Simbat-PWR Using Standard Code of Cobra-en on Reactor Transient Condition

Isnaini

Subekti

2016

Tri Dasa Mega

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VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION.The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR) using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which is able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that are calculated by MATPRO sub routine (COBRA-EN+MATPRO) for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher than COBRA-EN and 7.70% lower than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the difference of the gap between heat transfer coefficient and the cladding thermal conductivity.Keywords: transient, thermal-hydraulics, PWR, simulator, COBRA-EN, MATPRO. ABSTRAK VALIDASI SIMBAT-PWR MENGGUNAKAN KODE STANDAR COBRA-EN DALAM KONDISI REAKTOR TRANSIEN. Telah dilakukan validasi Simulator Pembangkit Listrik Tenaga Nuklir tipe Pressurized Water Reactor yang dikembangkan oleh BATAN (SIMBAT-PWR

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Section: Cobra-en Codementioning

confidence: 99%

Validation of Simbat-PWR Using Standard Code of Cobra-en on Reactor Transient Condition

Isnaini

Subekti

2016

Tri Dasa Mega

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“…During the temperature ramp up, two different heating rates were adopted: 5 and 10 K/min. The temperature ramp rates on the progression of a specific accident scenario can vary significantly, from few to more than one hundred Kelvin per minute, depending on the accident severity [29].…”

Section: Experiments Arrangementmentioning

confidence: 99%

Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

Tang

Jianu

Steinbrueck

et al. 2018

Journal of Nuclear Materials

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FeCrAl alloys are proposed and being intensively investigated as alternative accident tolerant fuel (ATF) cladding for nuclear fission application. Herein, the influence of major alloy elements (Cr and Al), reactive element effect and heating schedules on the oxidation behavior of FeCrAl alloys in steam up to 1500°C was examined. In case of transient ramp tests, catastrophic oxidation, i.e. rapid and complete consumption of the alloy, occurred during temperature ramp up to above 1200°C for specific alloys. The maximum compatible temperature of FeCrAl alloys in steam increases with raising Cr and Al content, decreasing heating rates during ramp period and doping of yttrium. Isothermal oxidation resulted in catastrophic oxidation at 1400°C for all examined alloys. However, formation of a protective alumina scale at 1500°C was ascertained despite partial melting. The occurrence of catastrophic oxidation seems to be controlled by dynamic competitive mechanisms between mass transfer of Al from the substrate and transport of oxidizing gas through the scale both toward the metal/oxide scale interface.

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“…Replacing the zircaloy cladding with another alloy such as stainless steel or iron-chromium-aluminum alloy (FeCrAl) [18] can improve oxidation resistance during an accident. However, steels tend to weaken and melt at lower temperatures compared to zirconium-based alloys [15], so molybdenum-based alloys have also been proposed as an alternative that can withstand high temperatures [19].…”

Section: Literature Reviewmentioning

confidence: 99%

Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

Younan

Novog

2018

Science and Technology of Nuclear Installations

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The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO 2 -SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO 2 . Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels. In addition, none of the fuel types have a significant advantage over UO 2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN) enriched in 15 N would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

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Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

Cited by 234 publications

References 5 publications

Validation of Simbat-PWR Using Standard Code of Cobra-en on Reactor Transient Condition

Validation of Simbat-PWR Using Standard Code of Cobra-en on Reactor Transient Condition

Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

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