Fission product behaviour and graphite corrosion under accident conditions in the HTR

Katscher, W.; Moormann, R.; Verfondern, Karl; Decken, C.B. von der; Iniotakis, N.; Hilpert, K.; Christ, A.; Lohnert, G.H.; Wawrzik, U.

doi:10.1016/0029-5493(90)90107-9

Cited by 16 publications

(4 citation statements)

References 4 publications

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“…For there to be any substantive radiological consequence from a severe accident at a HTGR-type SMR, it is thus necessary to postulate a means of failing the TRISO fuel. The SiC layer in TRISO has been demonstrated to be largely resistant to chemical attack and maintains its capability to retain fission products up to 1600°C [21][22][23][24][25]. In analyses of various HTGR accident scenarios, including pressurized loss of forced Ccrculation (equivalent to a loss-of-flow accident in a water-cooled plant), Depressurized loss of forced circulation (D-LOFC, equivalent to a loss-of-coolant accident in a watercooled plant), and reactivity-initiated accidents (including failure to scram), peak fuel temperatures are below this 1600°C limit [18,19,[26][27][28].…”

Section: Hypothetical Accident Progressionmentioning

confidence: 99%

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Results of a Phenomena Identification and Ranking Table (Pirt) Exercise for a Severe Accident in a Small Modular High-Temperature Gas-Cooled Reactor

et al. 2019

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Canada has attracted specific interest from developers of nonwater-cooled small modular reactor (SMR) technologies, including concepts based on high-temperature gas-cooled reactors (HTGRs). It is anticipated that some research and development (R&D) will be necessary to support safety analysis and licensing of these reactors in Canada. The Phenomena Identification and Ranking Table (PIRT) process is a formalized method in which a panel of experts identifies which physical phenomena are most relevant to the reactor safety analysis and how well understood these phenomena are. The PIRT process is thus a tool to assess current knowledge levels and (or) predictive capabilities of models, thus providing direction to a focused R&D program. This paper summarizes the results of a PIRT process performed by a panel of experts at Canadian Nuclear Laboratories for a limiting or “worst-case” accident scenario at a generic HTGR-type SMR. Suggestions are given regarding the highest priority R&D items to support severe accidents analysis of these reactors.

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Section: Hypothetical Accident Progressionmentioning

confidence: 99%

“…The postulated sequence of events considered in this PIRT exercise, based upon information available in literature [18][19][20][21][22][23][24][25][26][27][28][29][30][31][32][33][34][35], is thus as follows:…”

Section: Hypothetical Accident Progressionmentioning

confidence: 99%

Results of a Phenomena Identification and Ranking Table (Pirt) Exercise for a Severe Accident in a Small Modular High-Temperature Gas-Cooled Reactor

et al. 2019

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“…Both oxidation and potential erosion of the graphite block by graphite dust can significantly degrade the surface-structure of the graphite component. The effect of such surface degradation on the strength and fracture toughness [25][26][27] is an important consideration in extrapolating the properties of test bars to component behavior in a reactor. c. The effects of cyclic [28] and static fatigue [29] in reducing the allowable operational stress should be incorporated in a thorough analysis.…”

Section: Limitationsmentioning

confidence: 99%

On estimating the fracture probability of nuclear graphite components

Srinivasan

2008

Journal of Nuclear Materials

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“…The ingress of larger amounts of water or steam into the primary circuit of RDE is seen to be an accident of considerable safety relevance. The two impacts of water to the graphite structures of core and fuel elements are: moderation and graphite corrosion effect [6,7]. The changing of the concentration moderated nuclei the neutron balance is affected and by chemical oxidation of the graphite the structural integrity of the core is attacked and explosive gases may be produced.…”

Section: Introductionmentioning

confidence: 99%

Analysis of helium purification system capability during water ingress accident in RDE

Sriyono¹,

Kusmastuti²,

Bakhri³

et al. 2018

J. Phys.: Conf. Ser.

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Abstract. The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

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Fission product behaviour and graphite corrosion under accident conditions in the HTR

Cited by 16 publications

References 4 publications

Results of a Phenomena Identification and Ranking Table (Pirt) Exercise for a Severe Accident in a Small Modular High-Temperature Gas-Cooled Reactor

Results of a Phenomena Identification and Ranking Table (Pirt) Exercise for a Severe Accident in a Small Modular High-Temperature Gas-Cooled Reactor

On estimating the fracture probability of nuclear graphite components

Analysis of helium purification system capability during water ingress accident in RDE

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