Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

Johnson, L. H.; Ferry, Cécile; Poinssot, Christophe; Lovera, Patrick

doi:10.1016/j.jnucmat.2005.04.071

Cited by 72 publications

(77 citation statements)

References 6 publications

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“…During the dissolution process three releases occurred at the same time coming from: matrix, grain boundaries and gap and fractures dissolution [1,6,7]. The quantification of the gap and grain boundaries contribution can be determined by subtracting the matrix dissolution release.…”

Section: Determination Of Instant Release Fraction At Time 10 and 200mentioning

confidence: 99%

“…Radionuclides in gap and grain boundary have been considered as instant release fraction (IRF) [6,7]. This fraction includes: fission gases, volatiles ( 129 I, 137 Cs, 135 Cs, 36 Cl, 79 Se) and segregated metals ( 99 Tc, 107 Pd, 126 Sn) [8].…”

Section: Introductionmentioning

confidence: 99%

“…This layer is observed for BU's higher than 40 MWd·kgU -1 [9][10][11][12][13][14][15][16] and is a function of BU and the irradiation temperature, being the temperature threshold 1100ºC [13,15]. As a conservative approach for those SNF's, the radionuclides located in the HBS were also included in the IRF [6,8].…”

Section: Introductionmentioning

confidence: 99%

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Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

González-Robles

Serrano‐Purroy

Sureda

et al. 2015

Journal of Nuclear Materials

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The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd·kgU

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Section: Determination Of Instant Release Fraction At Time 10 and 200mentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

González-Robles

Serrano‐Purroy

Sureda

et al. 2015

Journal of Nuclear Materials

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“…Some studies have been conducted for investigating radionuclide characteristics of HTR-10 however it is limited only about the inventory activity [6,7]. Radionuclide activity is important in assessing the performance of spent fuel in geological disposal [8]. Thermal power quantification of spent fuel is used as the basis for safety analysis of heat removal in the spent fuel storage tank [9,10].…”

Section: Introductionmentioning

confidence: 99%

Radionuclide Characteristics of Rde Spent Fuels

Husnayani

Udiyani

2018

Tri Dasa Mega

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RADIONUCLIDE CHARACTERISTICS OF RDE SPENT FUELS. Reaktor Daya Eksperimental(RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that is planned to be constructed by National Nuclear Energy Agency of Indonesia (BATAN) in Puspiptek complex, Tangerang Selatan. RDE utilizes low enriched UO 2 fuel coated by TRISO layers and loaded into the core by means of multipass loading scheme. Determination of radionuclide characteristics of RDE spent fuel; such as activity, thermal power, neutron and photon release rates; are very important because those characteristics are crucial to be used as a base for evaluating the safety of spent fuel handling system and storage tank. This study is aimed to investigate the radionuclide characteristics of RDE spent fuel at the end of cycle and during the first 5 years cooling time in spent fuel storage. The method used to investigate the radionuclide characteristics is burnup calculation using ORIGEN2.1 code. In performing the ORIGEN2.1 calculation, one pebble fuel was assumed to be irradiated in the core for 5 cycles and then decayed for 5 years. At the end of the fifth cycle, it is obtained that the total activity, thermal power, neutron production, and photon release rates from all radionuclides inside one spent fuel are approximately 105.68 curies, 0.41 watts, 2.65 x 10 3 neutrons/second, and 1.79 x 10 4 photons/second, respectively. The results for the radionuclides characteristics during the first 5 years cooling time in the spent fuel storage show that the radioactivity characteristics from all radionuclides are rapidly decreasing at the first year and then slowly decreasing at the second until the fifth year of cooling time. The results obtained in this study can provide data for safety evaluation of fuel handling and spent fuel storage, such as the calculation of sourceterm, radiation dose rate, and the determination of radiation shielding.Keywords: RDE, spent fuel, radionuclide activity, thermal power, neutron production, photon releaserates ABSTRAK KARAKTERISTIK RADIONUKLIDA DI DALAM BAHAN BAKAR RDE. Reaktor Daya Eksperimental (RDE) adalah reaktor tipe Reaktor Temperatur Tinggi Berpendingin Gas dengan daya termal 10MW yang akan dibangun oleh BadanTenagaNuklirNasional (BATAN) di kawasanPuspiptek, Tangerang Selatan. RDE menggunakan bahan bakar UO 2 yang dilapisi dengan lapisan TRISO dan dimasukkan ke dalam teras RDE menurut skema multipass (5 siklus

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“…Dissolution studies of spent nuclear fuel (SNF) represent one main contribution to the safety assessment of a nuclear waste repository since the leaching of radionuclides (RN) into the groundwater is associated with the risk of ecological and human exposure [1,2,3,4]. SNF is characterized as being a complex system composed of the fuel matrix UO 2 doped with a variety of components such as oxide precipitates of Rb, Cs, Ba, Zr, Nb, Mo, Tc or metallic precipitates -so called -particles -of Mo, Ru, Pd, Tc and Rh [5].…”

Section: Introductionmentioning

confidence: 99%

Comparative study of the structural and electrochemical properties of noble metal inclusions in a UO₂matrix

Stumpf¹,

Tobias²,

Seibert³

et al. 2010

IOP Conf. Ser.: Mater. Sci. Eng.

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Abstract. The intention of the presented study is to elucidate the influence of noble metal inclusions (fission products) on the structure as well as on the electrochemical properties of spent nuclear fuel (SNF). To this aim, thin UO 2 films doped with metal inclusions such as Pd, Mo and Au are prepared by sputter deposition. The films are characterized by spectroscopic (XPS, EXAFS, XRD) as well as by microscopic (AFM, SEM) methods. In a next step the electrochemical properties of these model systems are comparatively investigated by cyclo voltammetry (CV). The sputter technique in combination with the heating treatment of the films allows the formation of a crystalline UO 2 matrix as it is found in SNF. The co-deposition with Au results in the dispersion of the pure metal in the oxide matrix. Pd as well as Mo are oxidized due to the deposition at RT. Heating the films involves a further oxidation of MoO 2 to MoO 3 . By contrast Pd agglomerates and forms metallic -phases as it is found in SNF. Electrochemical investigations of the UO 2 -Pd samples indicate an inhibiting influence of Pd on the oxidative dissolution of UO 2 . When it comes to the formation of secondary phases under reducing conditions such influence is passivated. The precipitates finally dominate the overall redox behaviour of the model system.

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Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

Cited by 72 publications

References 6 publications

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Radionuclide Characteristics of Rde Spent Fuels

Comparative study of the structural and electrochemical properties of noble metal inclusions in a UO₂matrix

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Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

Cited by 72 publications

References 6 publications

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Radionuclide Characteristics of Rde Spent Fuels

Comparative study of the structural and electrochemical properties of noble metal inclusions in a UO2matrix

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Comparative study of the structural and electrochemical properties of noble metal inclusions in a UO₂matrix