Frontiers in Nuclear Engineering最新文献

{"title":"Toward long-term storage of nuclear materials in MOX fuels fabrication facility","authors":"Shun Hirooka, S. Nakamichi, Taku Matsumoto, Ryota Tsuchimochi, Tatsutoshi Murakami","doi":"10.3389/fnuen.2023.1119567","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1119567","url":null,"abstract":"Storage of plutonium (Pu)-containing materials requires extremely strict attention to physical safety and material accounting. For the emphasized importance of storage management, many reports are available in the public, e.g., experience in PuO2 storage in the UK and safety standards in the storage of Pu-containing materials in the US. Japan also stores plutonium, mainly in the form of uranium-Pu mixed oxide (MOX) mostly in powder form. Robust conditioning and storage arrangements must be adopted that are appropriate to the characteristics of the MOX, such as raw powder obtained by reprocessing of spent Light Water Reactor fuels, MOX samples from research and development on the remains of fuel fabrication, which can contain organic materials, and dry-recycled powder during fuel fabrication. A reduction in MOX fuel fabrications operations and experience in degradation of MOX containers during extended period of storage have led to JAEA to review the MOX storage arrangements in the Plutonium Fuel Development Center in Japan Atomic Energy Agency. The present work discusses the various nuclear materials, storage methods, experience in degradation of containers that occur during storage, and strategies for future long-term storage.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"17 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122743960","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Investigating 99Mo output changes in high-sustainability uranium targets by modifying target volume and geometry","authors":"Robert Raposio, G. Braoudakis, Amir Rosenfeld, Juniper Bedwell-Wilson, Gordon Thorogood","doi":"10.3389/fnuen.2023.978990","DOIUrl":"https://doi.org/10.3389/fnuen.2023.978990","url":null,"abstract":"Introduction: The most common current uranium target type for 99Mo production uses plate geometry. This paper investigates the effects of geometry on 99Mo output and target sustainability. Methods: MCNP6.2 was used to model rectangular, spherical, and cylindrical targets ranging from 12.03 cm3 to 120.34 cm3 to determine the target geometry and volume effects on 99Mo output, sustainability, proliferation concerns, and heating. 4-7 days irradiation was used with a consistent target density of 2 g/cm3 UO2 for all target types. Results and Discussion: The cylindrical target was found to have the best performance due to having the second highest sustainability, the highest 99Mo output and produce the lowest amounts of 239Pu compared to the other target geometries. The heating comparison showed that there were negligible heating concerns for all target volumes and geometries.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"84 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134314717","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Influence of gluconate on the retention of Eu(III), Am(III), Th(IV), Pu(IV), and U(VI) by C-S-H (C/S = 0.8)","authors":"Sophie Dettmann, N. Huittinen, N. Jahn, Jérôme Kretzschmar, M. Kumke, Tamara Kutyma, J. Lohmann, T. Reich, K. Schmeide, Salim Shams Aldin Azzam, Leon Spittler, J. Stietz","doi":"10.3389/fnuen.2023.1124856","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1124856","url":null,"abstract":"The retention of actinides in different oxidation states (An(X), X = III, IV, VI) by a calcium-silicate-hydrate (C-S-H) phase with a Ca/Si (C/S) ratio of 0.8 was investigated in the presence of gluconate (GLU). The actinides considered were Am(III), Th(IV), Pu(IV), and U(VI). Eu(III) was investigated as chemical analogue for Am(III) and Cm(III). In addition to the ternary systems An(X)/GLU/C-S-H, also binary systems An(X)/C-S-H, GLU/C-S-H, and An(X)/GLU were studied. Complementary analytical techniques were applied to address the different specific aspects of the binary and ternary systems. Time-resolved laser-induced luminescence spectroscopy (TRLFS) was applied in combination with parallel factor analysis (PARAFAC) to identify retained species and to monitor species-selective sorption kinetics. 13C and 29Si magic-angle-spinning (MAS) nuclear magnetic resonance (NMR) spectroscopy and X-ray photoelectron spectroscopy (XPS) were applied to determine the bulk structure and the composition of the C-S-H surface, respectively, in the absence and presence of GLU. The interaction of Th(IV) with GLU in different electrolytes was studied by capillary electrophoresis-inductively coupled plasma mass spectrometry (CE-ICP-MS). The influence of GLU on An(X) retention was investigated for a large concentration range up to 10−2 M. The results showed that GLU had little to no effect on the overall An(X) retention by C-S-H with C/S of 0.8, regardless of the oxidation state of the actinides. For Eu(III), the TRLFS investigations additionally implied the formation of a Eu(III)-bearing precipitate with dissolved constituents of the C-S-H phase, which becomes structurally altered by the presence of GLU. For U(VI) sorption on the C-S-H phase, only a small influence of GLU could be established in the luminescence spectroscopic investigations, and no precipitation of U(VI)-containing secondary phases could be identified.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129913803","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Uptake of Ni(II), Eu(III) and Pu(III/IV) by Hardened Cement Paste in the Presence of Proxy Ligands for the Degradation of Polyacrylonitrile","authors":"Péter G. N. Szabó, A. Tasi, X. Gaona, A. Maier, S. Hedström, M. Altmaier, H. Geckeis","doi":"10.3389/fnuen.2023.1117413","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1117413","url":null,"abstract":"The uptake of 63Ni(II), 152Eu(III) and 242Pu(III/IV) by hardened cement paste (HCP, CEM I) in the degradation stage II (pH ≈ 12.5 [Ca] ≈ 0.02 M) was investigated in the absence and presence of α-hydroxyisobutyric, 3-hydroxybutyric and glutaric acids. These organic ligands were previously identified as proxies for the degradation products of UP2W (a polyacrylonitrile-based material used as filter aid in nuclear power plants) under repository conditions. Sorption experiments were conducted with various ligand concentrations (10−4 M ≤ [L]tot ≤ 0.1 M) and solid-to-liquid ratios (0.5 g⋅dm–3 ≤ S:L ≤ 20 g⋅dm–3). Redox conditions in the Pu systems were buffered with either hydroquinone (HQ, pe + pH ≈ 10) or Sn(II) (pe + pH ≈ 2). Strong sorption is observed for 152Eu(III) and 242Pu(III/IV) in the absence of proxy ligands, with distribution coefficients (log Rd ≈ 2.2–4, with Rd in m3⋅kg–1) in line with data reported in the literature. No differences are observed for sorption experiments with Pu in HQ and Sn(II) systems. Lower Rd values are determined for 63Ni(II) (log R d , N i 63 ≈ 0-1), consistently with previous studies. In combination with log R d , N i determined on the basis of the concentration of stable Ni(II) in pristine HCP and in cement porewater, values of the partition coefficient (α) close to 1 are determined. This suggests that the uptake of 63Ni is possibly driven by isotopic exchange with the complete species inventory of stable Ni present in pristine HCP, including Ni in solid phases and associated with surfaces, e.g., of C-S-H phases. The presence of proxy ligands has a negligible effect on the uptake of 152Eu(III) up to [L]tot = 0.1 M. A slight decrease in the distribution ratios for 63Ni(II) and 242Pu(III/IV) is observed at [L]tot > 10−2 M, although the effect is less evident in the case of plutonium due to the dispersion of the data and the increase of the detection limits with increasing ligand concentrations. Compared to strongly complexing ligands like isosaccharinic acid or gluconate, the investigated proxy ligands show a minor capacity for radionuclide mobilization in cementitious systems, even at high concentrations.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-02-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126475908","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Nuclear fuel irradiation testbed for nuclear security applications","authors":"B. Wilson, A. Conant, T. Ulrich, A. Kercher, Luke R. Sadergaski, T. Gerczak, A. Nelson, C. Petrie, J. Harp, A. Shields","doi":"10.3389/fnuen.2023.1123134","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1123134","url":null,"abstract":"The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"26 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126273535","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Oxygen diffusion in the fluorite-type oxides CeO2, ThO2, UO2, PuO2, and (U, Pu)O2","authors":"Masato Kato, Masashi Watanabe, Shun Hirooka, Romain Vauchy","doi":"10.3389/fnuen.2022.1081473","DOIUrl":"https://doi.org/10.3389/fnuen.2022.1081473","url":null,"abstract":"This study evaluates the self-diffusion and chemical diffusion coefficients of oxygen in the fluorite-type oxides CeO2, ThO2, UO2, PuO2, and (U, Pu)O2 using point defect chemistry (oxygen vacancies and interstitials). The self-diffusion coefficient changed in proportion to the 1/n power of oxygen partial pressure, similar to the defect concentration. All parameters used to represent the diffusion coefficients were determined, and the experimental data were accurately stated. The defect formation and migration energies of the oxides were compared, and the change in Frenkel defect concentration was found to affect the high-temperature heat capacities of CeO2 and ThO2. The oxygen chemical diffusion was evaluated in the oxides, excluding the line compound ThO2, and the coefficients increased dramatically around the stoichiometric composition, i.e., the chemical diffusion coefficient was much higher at stoichiometric composition, with the oxygen-to-metal ratio equal to 2.00, than in low oxygen-to-metal oxides. This difference altered the mechanism of the reduction and oxidation processes. In the reduction process, the chemical diffusion control rate was dominant and a new phase with the oxygen-to-metal ratio equal to 2.00 was formed, which then expanded from the surface in the oxidation process from a low oxygen-to-metal ratio to the stoichiometric composition.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"95 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133750465","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A review of void and gas bubble superlattices self-organization under irradiation","authors":"Yongfeng Zhang","doi":"10.3389/fnuen.2023.1110549","DOIUrl":"https://doi.org/10.3389/fnuen.2023.1110549","url":null,"abstract":"Irradiation by high-energy particles has been well known as a destructive force that “damages” crystalline materials by creating lattice defects. One surprising outcome from irradiation is the self-organization of void superlattices and gas bubble superlattices in various materials under irradiation. While these superlattices exhibit crystal structures that mimic atomic lattices, their self-organization takes place in far-from-equilibrium environment. A thermodynamic driving force that entails ordering is either absent or yet to be identified. In the past few decades, extensive research efforts have been made to generate such superlattices and to discern their formation mechanisms. While a consensus is yet to reach, these studies have substantially enriched our understanding on defect evolution and self-organization under irradiation. Appending previous reviews that are mostly done two decades ago, this article presents a comprehensive review of new experimental, theoretical, and simulational studies of void and gas bubble superlattices in the past two decades. An in-depth discussion on the formation mechanisms and their implications on superlattice properties is provided for the purpose of encouraging future studies.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"41 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114483855","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Investigating an alternative sustainable low-enriched uranium target for the manufacture of 99Mo using MCNP6.2 modelling with CINDER90","authors":"Robert Raposio, G. Braoudakis, Amir Rosenfeld, Gordon Thorogood, Juniper Bedwell-Wilson","doi":"10.3389/fnuen.2022.978948","DOIUrl":"https://doi.org/10.3389/fnuen.2022.978948","url":null,"abstract":"Introduction: Current fission-based methods of 99Mo production use targets that are based on high output without taking into consideration the amount of radioactive waste produced. We examine the idea of using a low enriched target (<20%) to reduce the amount of nuclear and chemical wastes produced by the manufacture of 99Mo via fission of uranium-based targets. Methods: MCNP6.2 was used to model targets of 20%, 10%, 7%, 3%, and 1% enrichment for 99Mo output and sustainability. Results and Discussion: The 1% enriched target at the lowest density of 0.2 g/cm3 of UO2 was found to have the highest sustainability score at 6 days irradiation but it had a low 99Mo output. On the other hand, the highest output target was found to be 20% enriched with a density of 8.0 g UO2/cm3 with an irradiation time of 20 days. Target security and safeguards concerns were modelled and found to be of minimal concern.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124499802","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo","authors":"Peter J. Doyle, C. Massey, D. Richardson, I. Greenquist, R. Seibert, G. Helmreich, T. Ulrich, Ryan C. Gallagher, Kara Godsey, R. Fielding, A. Nelson, J. Harp","doi":"10.3389/fnuen.2022.1054288","DOIUrl":"https://doi.org/10.3389/fnuen.2022.1054288","url":null,"abstract":"Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swelling measurement of the specimen sizes studied here required development and assessment of new methods for volume determination before and after irradiation. Laser profilometry and X-ray computation tomography (XCT) were used to provide preirradiation characterization of samples to determine the error and applicability of each to determine swelling following irradiation. These outcomes are contextualized through use of BISON simulations performed to assess the predicted expansion of U-Mo fuels subjected to the irradiation conditions of this work. Use of existing BISON fuel performance models predicted a maximum of 7% swelling under the irradiation conditions of this study. Pre-irradiation characterization revealed the as-cast U-Mo fuel samples were uniformly large-grained fully cubic U crystals with small U-C/N bearing precipitates and pores distributed throughout. Samples were found to contain a bulk porosity between .4 and 3% because of the casting process. Local porosity in areas far from large, interconnected pores was found by Slice-and-View to be under .2%. Nanometer-sized precipitates rich in C and N were identified in all samples, likely because of impurities during the fabrication process. Dendritic bands were also observed throughout the samples. These bands were characterized by variable Mo content that deviated from the overall Mo content by 2–3 wt %. No other microstructural features were correlated to these bands. Mechanical properties were found to be slightly strengthened compared to literature reports of bulk U-Mo fuels due to the nano-scale precipitates throughout the sample.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115550635","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Oxygen potential, oxygen diffusion, and defect equilibria in UO2±x","authors":"Masashi Watanabe, Masato Kato","doi":"10.3389/fnuen.2022.1082324","DOIUrl":"https://doi.org/10.3389/fnuen.2022.1082324","url":null,"abstract":"Since the oxygen potential and the oxygen diffusion coefficient of UO2 have a significant impact on fuel performance, many experimental data have been obtained. However, experimental data of the oxygen potential and the oxygen diffusion coefficient in the high temperature region above 1673 K are very limited. In the present study, we aimed to obtain these data and analyze them by defect chemistry. the oxygen potentials and the oxygen chemical diffusion coefficient of UO2 were measured by the gas equilibrium method in the near stoichiometric region at temperatures ranging from 1673 to 1873 K. A data set of oxygen potentials was made together with literature data and analyzed by defect chemistry. The oxygen potential of UO2 was determined as a function of O/U ratio and temperature, and an equation representing the relationship was derived. The oxygen chemical diffusion coefficient values obtained in this study were reasonably close to the literature values. The oxygen partial pressure dependence of the oxygen chemical diffusion coefficients was predicted from the evaluated results of the oxygen potential data, but no clear dependence was observed.","PeriodicalId":400400,"journal":{"name":"Frontiers in Nuclear Engineering","volume":"113 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117037133","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}