scholarly journals Radioactive Fission Waste from Molybdenum-99 Production and Proliferation Risks

2021 ◽  
Vol 927 (1) ◽  
pp. 012041
Author(s):  
Aisyah ◽  
Pungky Ayu Artiani ◽  
Jaka Rachmadetin
Keyword(s):  
Nuclear Proliferation ◽  
Fission Products ◽  
High Level Waste ◽  
Blending Method ◽  
Activation Products ◽  
Enriched Uranium ◽  
High Level ◽  
Solid Form ◽  
Interim Storage

Abstract Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.

scholarly journals Control Experiences Regarding Clearable Materials from Nuclear Power Plants and Nuclear Installations: Scaling Factors Determination and Measurements’ Acceptance Criteria Definition

Environments ◽  
2019 ◽  
Vol 6 (11) ◽  
pp. 120
Author(s):  
Luca Albertone ◽  
Massimo Altavilla ◽  
Manuela Marga ◽  
Laura Porzio ◽  
Giuseppe Tozzi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Scaling Factors ◽  
Activation Products ◽  
Long Time ◽  
Nuclear Installation ◽  
Enriched Uranium

Arpa Piemonte has been carrying out, for a long time, controls on clearable materials from nuclear power plants to verify compliance with clearance levels set by ISIN (Ispettorato Nazionale per la Sicurezza Nucleare e la Radioprotezione - National Inspectorate for Nuclear Safety and Radiation Protection) in the technical prescriptions attached to the Ministerial Decree decommissioning authorization or into category A source authorization (higher level of associated risk, according to the categorization defined in the Italian Legislative Decree No. 230/95). After the experience undertaken at the “FN” (Fabbricazioni Nucleari) Bosco Marengo nuclear installation, some controls have been conducted at the Trino nuclear power plant “E. Fermi,” “LivaNova” nuclear installation based in Saluggia, and “EUREX” (Enriched Uranium Extraction) nuclear installation, also based in Saluggia, according to modalities that envisage, as a final control, the determination of γ-emitting radionuclides through in situ gamma spectrometry measurements. Clearance levels’ compliance verification should be performed for all radionuclides potentially present, including those that are not easily measurable (DTM, Difficult To Measure). It is therefore necessary to carry out upstream, based on a representative number of samples, those radionuclides’ determination in order to estimate scaling factors (SF), defined through the logarithmic average of the ratios between the i-th DTM radionuclide concentration and the related key nuclide. Specific radiochemistry is used for defining DTMs’ concentrations, such as Fe-55, Ni-59, Ni-63, Sr-90, Pu-238, and Pu-239/Pu-240. As a key nuclide, Co-60 was chosen for the activation products (Fe-55, Ni-59, Ni-63) and Cs-137 for fission products (Sr-90) and plutonium (Pu- 238, Pu-239/Pu-240, and Pu-241). The presence of very low radioactivity concentrations, often below the detection limits, can make it difficult to determine the related scaling factors. In this work, the results obtained and measurements’ acceptability criteria are presented, defined with ISIN, that can be used for confirming or excluding a radionuclide presence in the process of verifying clearance levels’ compliance. They are also exposed to evaluations regarding samples’ representativeness chosen for scaling factors’ assessment.

CAREM X INPRO Case Study

2004 ◽  
Author(s):  
Pablo C. Florido ◽  
Dari´o Delmastro ◽  
Daniel Brasnarof ◽  
Osvaldo E. Azpitarte
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
High Level Waste ◽  
Fuel Cycles ◽  
Surface Intermediate ◽  
Geological Repository ◽  
Enriched Uranium ◽  
Intermediate Storage ◽  
High Level

Argentina is performing CAREM X Nuclear System Case Study based on CAREM nuclear reactor and Once Through Fuel Cycle, using SIGMA for enriched uranium production, and a deep geological repository for final disposal of high level waste after surface intermediate storage in horizontal natural convection silos, to verify INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) methodology. Projections show that developing countries could play a crucial role in the deployment of nuclear energy, in the next fifty years. This case study will be highly useful for checking INPRO methodology for this scenario. In this contribution to ICONE 12, the preliminary findings of the Case Study are presented, including proposals to improve the INPRO methodology.

Interim Storage Technology of Spent Fuel and High-Level Waste in Germany

2007 ◽  
Author(s):  
H. Geiser ◽  
J. Schro¨der
Keyword(s):  
Normal Operation ◽  
High Level Waste ◽  
Spent Fuel ◽  
Horizontal Orientation ◽  
Long Term Storage ◽  
Fuel Storage ◽  
Safety Features ◽  
High Level ◽  
Interim Storage ◽  
Aircraft Crash

The idea of using casks for interim storage of spent fuel arose at GNS after a very controversial political discussion in 1978, when total passive safety features (including aircraft crash conditions) were required for an above ground spent fuel storage facility. In the meantime, GNS has loaded more than 1000 casks at 25 different storage sites in Germany. GNS cask technology is used in 13 countries. Spent fuel assemblies of PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high level waste containers are stored in full metal casks of the CASTOR® type. Also MOX fuel of PWR and BWR has been stored. More than two decades of storage have shown that the basic requirements (safe confinement, criticality safety, sufficient shielding and appropriate heat transfer) have been fulfilled in any case — during normal operation and in case of severe accidents, including aircraft crash. There is no indication of problems arising in the future. Of course, the experience of more than 20 years has resulted in improvements of the cask design. The CASTOR® casks have been thoroughly investigated by many experiments. There have been approx. 50 full and half scale drop tests and a significant number of fire tests, simulations of aircraft crash, investigations with anti tank weapons, and an explosion of a railway tank with liquid gas neighbouring a loaded CASTOR® cask. According to customer and site specific demands, different types of storage facilities are realized in Germany. Firstly, there are facilities for long-term storage, such as large ventilated central storage buildings away from reactor or ventilated storage buildings at the reactor site, ventilated underground tunnels or concrete platforms outside a building. Secondly, there are facilities for temporary storage, where casks have been positioned in horizontal orientation under a ventilated shielding cover outside a building.

Determination of the Corrosion Mechanisms of High-Level Waste Containing Glass

1981 ◽  
Vol 11 ◽  
Author(s):  
Horst Scholze ◽  
Reinhard Conradt ◽  
Heinrich Engelke ◽  
Hans Roggendorf
Keyword(s):  
Rock Salt ◽  
High Level Waste ◽  
Extreme Conditions ◽  
Safety Standard ◽  
Waste Forms ◽  
Corrosion Mechanisms ◽  
High Level

The German concept of high level waste final storage provides the use of certain glasses containing radioelement oxides as glass components. These waste forms are to be stored in rock salt formations in order to isolate the waste from the biosphere. The efficiency of this isolation is a most important question. The aim is to achieve a high safety standard that remains valid under extreme conditions such as the uncontrolled water entrance to the deposit.

scholarly journals Characterization of Composite Ceramic High Level Waste Forms

1997 ◽  
Vol 481 ◽  
Author(s):  
S. M. Frank ◽  
K. J. Bateman ◽  
T. DiSanto ◽  
S. G. Johnson ◽  
T. L. Moschetti ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Composite Ceramic ◽  
Waste Form ◽  
High Level Waste ◽  
Ceramic Waste ◽  
Chloride Form ◽  
Physical Measurements ◽  
High Level

ABSTRACTArgonne National Laboratory has developed a composite ceramic waste form for the disposition of high level radioactive waste produced during electrometallurgical conditioning of spent nuclear fuel. The electrorefiner LiCl/KCl eutectic salt, containing fission products and transuranics in the chloride form, is contacted with a zeolite material which removes the fission products from the salt. After salt contact, the zeolite is mixed with a glass binder. The zeolite/glass mixture is then hot isostatic pressed (HIPed) to produce the composite ceramic waste form. The ceramic waste form provides a durable medium that is well suited to incorporate fission products and transuranics in the chloride form. Presented are preliminary results of the process qualification and characterization studies, which include chemical and physical measurements and product durability testing, of the ceramic waste form.

scholarly journals Development of a ceramic waste form for high-level waste disposal

1999 ◽  
Vol 556 ◽  
Author(s):  
D. W. Esh ◽  
K. M. Goff ◽  
K. T. Hirsche ◽  
T. J. Battisti ◽  
M. F. Simpson ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
High Level Waste ◽  
Durability Test ◽  
Ceramic Waste ◽  
High Level ◽  
The Impact

AbstractA ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.

scholarly journals Spectroscopic and chemical investigations on volatile fission and activation products within the fuel-cladding interface of irradiated pressurised water reactor fuel rod segments

2021 ◽  
Vol 1 ◽  
pp. 5-6
Author(s):  
Tobias König ◽  
Ron Dagan ◽  
Kathy Dardenne ◽  
Michel Herm ◽  
Volker Metz ◽  
...  
Keyword(s):  
Mass Spectrometry ◽  
Nuclear Fuel ◽  
High Level Waste ◽  
Fuel Cladding ◽  
Interaction Layer ◽  
X Ray ◽  
Fuel Rod ◽  
Activation Products ◽  
Interim Storage ◽  
Fission And Activation Products

Abstract. In Germany, the present waste management concept foresees the direct disposal of spent nuclear fuel (SNF) in deep geological repositories for high-level waste available by 2050, at best. Until then, SNF is encapsulated in dual-purpose casks and stored in dry interim storage facilities. Licenses for both casks and facilities will expire after 40 years following loading of the cask and emplacement of the first cask in the storage location. Yet, due to considerable delays in the site selection process and the estimated duration for construction and commissioning of a final repository of at least 2 decades, a prolonged dry interim storage of SNF is inevitable (ESK, 2015). Concerning these considerable timespans, integrity of the cladding is of utmost importance regarding the ultimately conditioning of the fuel assemblies for final disposal. Various processes strain the structural integrity of Zircaloy cladding during reactor operation and beyond such as delayed hydride cracking, fuel-cladding chemical interactions or irradiation damage induced by α-emitters present in the fuel pellet's rim zone (Ewing, 2015). Especially with higher burn-up, the gap between fuel and cladding closes and results in the formation of an interaction layer, in which precipitates of fission and activation products are present, displaying an interface for degradation processes. For chemical analysis and speciation of these agglomerates, Zircaloy-4 and SNF specimens were sampled from fuel rod segments irradiated in commercial pressurised water reactors during the 1980s. Zircaloy-4 specimens were taken from an UOX (50.4 GWdtHM-1) and mixed oxide fuel (MOX) (38.0 GWdtHM-1). In addition, SNF fragments were sampled from the closed gap of both fuel types to examine volatile activation and fission products, which had been segregated from the centre to the pellet periphery during irradiation and thus contribute to the possible chemically assisted cladding degradation effect of the precipitates within the fuel-cladding interface. Spectroscopic analysis of precipitates within the interface layer between fuel and cladding were performed by optical microscopy, X-ray absorption and X-ray photoelectron spectroscopy, as well as by energy-dispersive scanning electron microscopy. Moreover, the radionuclide inventory of the respective Zircaloy-4, fuel and interaction layers was determined using liquid scintillation counting, γ-spectroscopy, gas mass spectrometry, ion chromatography and inductive-coupled plasma mass spectrometry and compared to results received by MCNP/CINDER and webKORIGEN calculations. In this study, we provide results regarding the speciation and chemical composition of previously identified Cs-U-O-Zr-Cl-I bearing compounds found in the interaction layer of irradiated nuclear fuel and inventory analyses of radionuclides present therein, with particular emphasis on Cl-36 and I-129. Furthermore, the agglomerates within the fuel-cladding interface were characterised for the first time utilising synchrotron radiation-based Cl K-edge and I K-edge measurements, resulting in compounds with structural similarities to CsCl and CsI. The outcomes obtained from this study provide further insights into the complex chemistry within the fuel-cladding interface with respect to the aging management and integrity of SNF under the conditions of interim storage. In future studies we will examine whether the different compounds at the fuel-cladding interface have the potential to affect the mechanical properties of Zircaloy cladding.

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