Updating irradiated graphite disposal: Project ‘GRAPA’ and the international decommissioning network

Wickham, A.J.; Steinmetz, Hans Jürgen; O’Sullivan, Patrick; Ojovan, Michael I.

doi:10.1016/j.jenvrad.2017.01.022

Cited by 50 publications

(24 citation statements)

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“…In 2015, an environmental radiological characterization was carried out to assess unrealistic contamination of the surrounding area and obtain the reference radiological blank [8]. In 2016, Politecnico di Milano joined the IAEA collaborative research project on irradiated GRAphite Processing Approaches (GRAPA) [9]. This paper describes the 3D neutronic model of L-54M reactor developed in order to accurately assess the neutron activation of the graphite stack occurred in 20 years of operations [10] and better address the future radiological characterization efforts [11].…”

Section: Politecnico DI Milano Carried Out Several Activities Accordimentioning

confidence: 99%

MCNP model of L-54 M nuclear research reactor: development and preliminary verification

Parma

Rossi

Mossini

et al. 2018

J Radioanal Nucl Chem

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Since shutdown , Politecnico di Milano managed L-54M nuclear research reactor according to deferred dismantling strategy. Recently, decommissioning activities have been started with preliminary radiological characterization, even though a more extended campaign would be required. In this work, a Monte Carlo N-Particle 3D model of the reactor has been proficiently developed to estimate graphite stack activation. To verify the model accuracy, several simulated criticality data have been satisfactorily compared with experimental ones, validating this computational approach as valid support to forthcoming radiological characterization campaign.

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Section: Politecnico DI Milano Carried Out Several Activities Accordimentioning

confidence: 99%

MCNP model of L-54 M nuclear research reactor: development and preliminary verification

Parma

Rossi

Mossini

et al. 2018

J Radioanal Nucl Chem

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“…In some cases, the value of selective yield of 14 C reaches 70% with relatively low loss of mass of graphite (within the limits of 5%). Practical implementation of the process in question can significantly reduce potential danger of graphite radioactive wastes due to the reduction of both the graphite activity for the given isotope, and the rate of its leaching (LaBrier and Dunzik-Gougar 2015, Wickham et al 2017) by the removal in the processing of the weakest bound and heterogeneously distributed 14 С isotope (Fig. 6).…”

Section: Arrangement Of Batches Of Samples In Two Differentmentioning

confidence: 99%

Potential of application of IRT-T research reactor as the solution of the problem of graphite radwaste disposal

Pavliuk¹,

Kotlyarevskiy²,

Bespala³

et al. 2018

NUCET

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Aspects of handling irradiated graphite during decommissioning uranium-graphite reactors (UGR) of different types were investigated. It was demonstrated that handling reactor graphite is complicated by the presence in the composition of graphite of long-lived radionuclides, especially 14C, which may get entrained in biological cycles since carbon constitutes one of the main components of biological chains. Practical implementation of the process of selective separation of 14С can significantly reduce potential danger represented by graphite radioactive wastes due to the reduction of graphite activity as related to the isotope in question, as well as due to the reduction of the leaching rate by separating 14С isotope which is the most weakly bound within the graphite structure. Conclusion was formulated that analytical measurement methodologies and calculation methods allow reliably estimating only the total quantity of 14C accumulated in graphite, the contribution of 14C accumulation channel from 13C(n, γ)14C reaction, as well as the total contribution of 14N(n, p)14C reaction on nitrogen impurities and on nitrogen contained in purge gas. Method was suggested for estimating the values of contributions of different channels of accumulation on nitrogen impurities and nitrogen contained in purge gas using IRT-T research reactor (Tomsk, Tomsk Region). Parallel irradiation of batches of samples of non-irradiated (fresh) reactor-grade graphite contained in different gaseous media constitutes the basis of the study. Algorithm was suggested for calculating contributions of all channels of 14C accumulation according to the results of measurements to be obtained in the proposed studies. Recommendations were formulated on the use of all brands of graphite applied for manufacturing elements of graphite stacks of uranium-graphite reactors designed in Russia for determining selectively separated fraction of 14C for all types of graphite radioactive wastes by the companies in the RF which operated (are operating) the uranium-graphite reactors. Time of exposure of samples of irradiated graphite in the GEK-4 horizontal experimental channel of the IRT-T reactor was calculated and was found to be equal to ~ 10 days. Methodology was suggested for conducting a series of experiments for determining the values of contributions of 14C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated 14C in irradiated graphite elements of practically all uranium-graphite nuclear reactors, including reactors operated abroad Russia, under the condition of maintaining carbon dioxide gas atmosphere in one of the irradiated containers.

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“…In recent years, a number of substantial initiatives have been undertaken to investigate irradiated graphite characterization, processing, immobilization and Figure 1. World inventory of irradiated graphite waste (tonnes), not including China (based on data from [1] [4] and the IAEA Project GRAPA (Irradiated Graphite Processing Approaches) [5]. The radionuclides in irradiated graphite presenting the most significant long-term hazard are Carbon 14 (C-14) and…”

Section: Introductionmentioning

confidence: 99%

The Significance of Carbon 14 in Graphite Reactor Components at End of Generation

Metcalfe¹,

Tzelepi²

2019

JEP

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Updating irradiated graphite disposal: Project ‘GRAPA’ and the international decommissioning network

Cited by 50 publications

References 1 publication

MCNP model of L-54 M nuclear research reactor: development and preliminary verification

MCNP model of L-54 M nuclear research reactor: development and preliminary verification

Potential of application of IRT-T research reactor as the solution of the problem of graphite radwaste disposal

The Significance of Carbon 14 in Graphite Reactor Components at End of Generation

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