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Sample records for fabricating 124i nuclear

  1. Nuclear Fabrication Consortium

    Energy Technology Data Exchange (ETDEWEB)

    Levesque, Stephen [EWI, Columbus, OH (United States)

    2013-04-05

    This report summarizes the activities undertaken by EWI while under contract from the Department of Energy (DOE) Office of Nuclear Energy (NE) for the management and operation of the Nuclear Fabrication Consortium (NFC). The NFC was established by EWI to independently develop, evaluate, and deploy fabrication approaches and data that support the re-establishment of the U.S. nuclear industry: ensuring that the supply chain will be competitive on a global stage, enabling more cost-effective and reliable nuclear power in a carbon constrained environment. The NFC provided a forum for member original equipment manufactures (OEM), fabricators, manufacturers, and materials suppliers to effectively engage with each other and rebuild the capacity of this supply chain by : Identifying and removing impediments to the implementation of new construction and fabrication techniques and approaches for nuclear equipment, including system components and nuclear plants. Providing and facilitating detailed scientific-based studies on new approaches and technologies that will have positive impacts on the cost of building of nuclear plants. Analyzing and disseminating information about future nuclear fabrication technologies and how they could impact the North American and the International Nuclear Marketplace. Facilitating dialog and initiate alignment among fabricators, owners, trade associations, and government agencies. Supporting industry in helping to create a larger qualified nuclear supplier network. Acting as an unbiased technology resource to evaluate, develop, and demonstrate new manufacturing technologies. Creating welder and inspector training programs to help enable the necessary workforce for the upcoming construction work. Serving as a focal point for technology, policy, and politically interested parties to share ideas and concepts associated with fabrication across the nuclear industry. The report the objectives and summaries of the Nuclear Fabrication Consortium

  2. Fabricating nuclear components

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Activities of the Nuclear Engineering Division of Vickers Ltd., particularly fabrication of long slim tubular components for power reactors and the construction of irradiation loops and rigs, are outlined. The processes include hydraulic forming for fabrication of various types of tubes and outer cases of fuel transfer buckets, various specialised welding operations including some applications of the TIG process, and induction brazing of specialised assemblies. (U.K.)

  3. Nuclear fuel fabrication in India

    Energy Technology Data Exchange (ETDEWEB)

    Kondal Rao, N

    1975-01-01

    The important role of a nuclear power program in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned.

  4. Nuclear fuel fabrication in India

    International Nuclear Information System (INIS)

    Kondal Rao, N.

    1975-01-01

    The important role of a nuclear power programme in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned. (K.B.)

  5. Positron emission intensities in the decay of 64Cu, 76Br and 124I

    International Nuclear Information System (INIS)

    Qaim, S.M.; Bisinger, T.; Hilgers, K.; Nayak, D.; Coenen, H.H.

    2007-01-01

    The relatively long-lived positron emitters 64 Cu (t 1/2 = 12.7 h), 76 Br (t 1/2 = 16.2 h) and 124 I (t 1/2 = 4.18 d) are finding increasing applications in positron emission tomography (PET). For precise determination of their positron emission intensities, each radionuclide was prepared via a charged particle induced reaction in a ''no-carrier-added'' form and with high radionuclidic purity. It was then subjected to γ-ray and X-ray spectroscopy as well as to anticoincidence beta and γγ-coincidence counting. The positron emission intensities measured were: 64 Cu (17.8 ± 0.4)%, 76 Br (58.2 ± 1.9)% and 124 I (22.0 ± 0.5)%. The intensity of the weak 1346 keV γ-ray emitted in the decay of 64 Cu was determined as (0.54 ± 0.03)%. Some implications of the precisely determined nuclear data are discussed. (orig.)

  6. Evaluation of {sup 124}I PET/CT and {sup 124}I PET/MRI in the management of patients with differentiated thyroid cancer

    Energy Technology Data Exchange (ETDEWEB)

    Dercle, Laurent; Deandreis, Desiree; Terroir, Marie; Leboulleux, Sophie; Lumbroso, Jean; Schlumberger, Martin [Institut Gustave Roussy and University Paris Saclay, Department of Nuclear Medicine and Endocrine Oncology, Villejuif Cedex (France)

    2016-06-15

    The work of Binse and colleagues points out that there are probably some research perspectives for the use of {sup 124}I PET/ CT and PET/MRI in patients with DTC. It shows that there is no substantial advantage of {sup 124}I PET/MRI over {sup 124}I PET/CT for the detection of tumour lesions in the neck when using similar PET devices. It confirms the superiority of {sup 124}I PET over CT and MRI for the detection of iodine-positive lesions. It demonstrates that the use of a more sensitive PET device and a longer acquisition time leads to the detection of more lesions. {sup 124}I PET is a promising research tool for pretherapy dosimetry, the evaluation of response to {sup 131}I treatment and the staging of recurrent or residual disease. The recognized advantages of MRI are the evaluation of aerodigestive tract lesions and suprahyoid region lesions. The coregistration of MRI and {sup 124}I PET/CT might thus be more convenient than {sup 124}I PET/ MRI (shorter time of acquisition, better cost-effectiveness and more accurate attenuation correction). The benefits of these procedures in terms of patient outcome, and for the clinician and the healthcare system remain to be determined.

  7. 2'-fluoro-2'-deoxy-1-β-D-arabinofuranosyl-5-[124I]iodouracil ([124I]FIAU)

    International Nuclear Information System (INIS)

    Chae, Min Jeong; Lee, Tae Sup; Kim, June Youp

    2008-01-01

    The HSV1-tk gene has been extensively studied as a type of reporter gene. In hepatocellular carcinoma (HCC), only a small proportion of patients are eligible for surgical resection and there is limitation in palliative options. Therefore, there is a need for the develoopement of new treatment modalities and gene therapy is a leading candidate. In the present study, we investigated the usefulness of substrate, 2'-fluoro-2'-deoxy-1-β -D-arabino-furanosyl-5-[ 124/125 I]iodo- uracil ([ 124/125 I]FIAU) as a non-invasive imaging agent for HSV1-tk gene therapy in hepatoma model using small animal PET. With the Morris hepatoma MCA cell line and MCA-tk cell line which was transduced with the HSV1-tk gene, in vitro uptake and correlation study between [ 125 I]FIAU uptake according to increasing numeric count of percentage of MCA-tk cell were performed. The biodistribution data and small animal PET images with [ 124 I]FIAU were obtained with Balb/c-nude mice bearing both MCA and MCA-tk tumors. Specific accumulation of [ 125 I]FIAU was observed in MCA-tk cells but uptake was low in MCA cells. Uptake in MCA-tk cells was 15 times higher than that of MCA cells at 480 min. [ 125 I]FIAU uptake was linearly correlated (R2=0.964, p=0.01) with increasing percentage of MCA-tk numeric cell count. Biodistribution results showed that [ 125 I]FIAU was mainly excreted via the renal system in the early phase. Ratios of MCA-tk tumor to blood acting were 10, 41, and 641 at 1 h, 4 h, and 24 h post-injection, respectively. The maximum ratio of MCA-tk to MCA tumor was 192.7 at 24 h. Ratios of MCA-tk tumor to liver were 13.8, 66.8, and 588.3 at 1 h, 4 h, and 24 h, respectively. On small aninal PET, [ 124 I]FIAU accumulated in substantial higher levels in MCA-tk tumor and liver than MCA tumor. FIAU shows selective accumulation to HSV1-tk expressing hepatoma cell tumors with minimal uptake in normal liver. Therefore, radiolabelled FIAU is expected to be a useful substrate for non-invasive imaging

  8. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  9. On the study of proton-irradiated Tellurium targets relevant for production of medical radioisotopes 123I and 124I

    International Nuclear Information System (INIS)

    Imam Kambali; Hari Suryanto; Daya Agung Sarwono; Cahyana Amiruddin

    2014-01-01

    The energy loss distribution and range of energetic proton beams in tellurium (Te) target have been simulated using the Stopping and Range of Ion in Matter (SRIM 2013) codes. The calculated data of the proton's range were then used to determine the optimum thickness of Te targets for future production of 123 I and 124 I from 123 Te(p,n) 123 I, 124 Te(p,n) 124 I and 124 Te(p,2n) 123 I nuclear reactions using the BATAN's Cs-30 cyclotron. It was found that for an incidence angle of 0° with respect to the target normal, the optimum thickness of 123 Te and 124 Te targets for 123 I production should be 644 µm and 1.8 mm respectively, whereas a 649 µm thick 124 Te target would be Required for 124 I production. In addition, the thickness should be decreased with increasing incidence angle. The EOB yield could theoretically reach up to 13.62 Ci of 123 I at proton energy of 22 Me V and beam current of 30 µA if the 124 Te is irradiated over a period of 3 hours. The theoretical EOB yield is comparable to the experimental data with accuracy within 10%. (author)

  10. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  11. Nuclear fuel conversion and fabrication chemistry

    International Nuclear Information System (INIS)

    Lerch, R.E.; Norman, R.E.

    1984-01-01

    Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO 2 material or a mixed PuO 2 -UO 2 fuel material is generated. Several methods are available for preparation of the pure PuO 2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO 2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO 2 . The UO 2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO 2 -PuO 2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO 3 -PuO 2 starting materials is accomplished by calcination-reduction to UO 2 -PuO 2 followed by pellet fabrication. (orig.)

  12. [124I]FIAU: Human dosimetry and infection imaging in patients with suspected prosthetic joint infection

    International Nuclear Information System (INIS)

    Zhang, Xiaoyan M.; Zhang, Halle H.; McLeroth, Patrick; Berkowitz, Richard D.; Mont, Michael A.; Stabin, Michael G.; Siegel, Barry A.; Alavi, Abass; Barnett, T. Marc; Gelb, Jeffrey; Petit, Chantal; Spaltro, John; Cho, Steve Y.; Pomper, Martin G.; Conklin, James J.; Bettegowda, Chetan; Saha, Saurabh

    2016-01-01

    Introduction: Fialuridine (FIAU) is a nucleoside analog that is a substrate for bacterial thymidine kinase (TK). Once phosphorylated by TK, [ 124 I]FIAU becomes trapped within bacteria and can be detected with positron emission tomography/computed tomography (PET/CT). [ 124 I]FIAU PET/CT has been shown to detect bacteria in patients with musculoskeletal bacterial infections. Accurate diagnosis of prosthetic joint infections (PJIs) has proven challenging because of the lack of a well-validated reference. In the current study, we assessed biodistribution and dosimetry of [ 124 I]FIAU, and investigated whether [ 124 I]FIAU PET/CT can diagnose PJIs with acceptable accuracy. Methods: To assess biodistribution and dosimetry, six subjects with suspected hip or knee PJI and six healthy subjects underwent serial PET/CT after being dosed with 74 MBq (2 mCi) [ 124 I]FIAU intravenously (IV). Estimated radiation doses were calculated with the OLINDA/EXM software. To determine accuracy of [ 124 I]FIAU, 22 subjects with suspected hip or knee PJI were scanned at 2–6 and 24–30 h post IV injection of 185 MBq (5 mCi) [ 124 I]FIAU. Images were interpreted by a single reader blinded to clinical information. Representative cases were reviewed by 3 additional readers. The utility of [ 124 I]FIAU to detect PJIs was assessed based on the correlation of the patient's infection status with imaging results as determined by an independent adjudication board (IAB). Results: The kidney, liver, spleen, and urinary bladder received the highest radiation doses of [ 124 I]FIAU. The effective dose was 0.16 to 0.20 mSv/MBq and doses to most organs ranged from 0.11 to 0.76 mGy/MBq. PET image quality obtained from PJI patients was confounded by metal artifacts from the prostheses and pronounced FIAU uptake in muscle. Consequently, a correlation with infection status and imaging results could not be established. Conclusions: [ 124 I]FIAU was well-tolerated in healthy volunteers and subjects with

  13. [(124)I]FIAU: Human dosimetry and infection imaging in patients with suspected prosthetic joint infection.

    Science.gov (United States)

    Zhang, Xiaoyan M; Zhang, Halle H; McLeroth, Patrick; Berkowitz, Richard D; Mont, Michael A; Stabin, Michael G; Siegel, Barry A; Alavi, Abass; Barnett, T Marc; Gelb, Jeffrey; Petit, Chantal; Spaltro, John; Cho, Steve Y; Pomper, Martin G; Conklin, James J; Bettegowda, Chetan; Saha, Saurabh

    2016-05-01

    Fialuridine (FIAU) is a nucleoside analog that is a substrate for bacterial thymidine kinase (TK). Once phosphorylated by TK, [(124)I]FIAU becomes trapped within bacteria and can be detected with positron emission tomography/computed tomography (PET/CT). [(124)I]FIAU PET/CT has been shown to detect bacteria in patients with musculoskeletal bacterial infections. Accurate diagnosis of prosthetic joint infections (PJIs) has proven challenging because of the lack of a well-validated reference. In the current study, we assessed biodistribution and dosimetry of [(124)I]FIAU, and investigated whether [(124)I]FIAU PET/CT can diagnose PJIs with acceptable accuracy. To assess biodistribution and dosimetry, six subjects with suspected hip or knee PJI and six healthy subjects underwent serial PET/CT after being dosed with 74MBq (2mCi) [(124)I]FIAU intravenously (IV). Estimated radiation doses were calculated with the OLINDA/EXM software. To determine accuracy of [(124)I]FIAU, 22 subjects with suspected hip or knee PJI were scanned at 2-6 and 24-30h post IV injection of 185MBq (5mCi) [(124)I]FIAU. Images were interpreted by a single reader blinded to clinical information. Representative cases were reviewed by 3 additional readers. The utility of [(124)I]FIAU to detect PJIs was assessed based on the correlation of the patient's infection status with imaging results as determined by an independent adjudication board (IAB). The kidney, liver, spleen, and urinary bladder received the highest radiation doses of [(124)I]FIAU. The effective dose was 0.16 to 0.20mSv/MBq and doses to most organs ranged from 0.11 to 0.76mGy/MBq. PET image quality obtained from PJI patients was confounded by metal artifacts from the prostheses and pronounced FIAU uptake in muscle. Consequently, a correlation with infection status and imaging results could not be established. [(124)I]FIAU was well-tolerated in healthy volunteers and subjects with suspected PJI, and had acceptable dosimetry. However, the

  14. Dual monitoring using 124I-FIAU and bioluminescence for HSV1-tk suicide gene therapy

    International Nuclear Information System (INIS)

    Lee, T. S.; Kim, J. H.; Kwon, H. C.

    2007-01-01

    Herpes simplex virus type I thymidine kinase (HSV-tk) is the most common reporter gene and is used in cancer gene therapy with a prodrug nucleoside analog, ganciclovir (GCV). The aim of this study is to evaluate therapeutic efficacy of suicide gene therapy with 2'-fluoro-2'-deoxy-1-D-arabinofuranosyl-5-[ 124 I] iodouracil ( 124 I - FIAU) and bioluminescence in retrovirally HSV -tk and firefly luciferase transduced hepatoma model. The HSV -tk and firefly luciferase (Luc) was retrovirally transduced and expressed in MCA rat Morris hepatoma cells. Nude mice with subcutaneous tumors, MCA and MCA-TK-Luc, were subjected to GCV treatment (50mg/Kg/d intraperitoneally) for 5 day. PET imaging and biodistribution with ( 124 I-FIAU) were performed at before and after initiation of therapy with GCV. Bioluminescent signal was also measured during GCV treatment. Before GCV treatment, no significant difference in tumor volume was found in tumors between MCA and MCA-TK-Luc. After GCV treatment, tumor volume of MCA-TK-Luc markedly reduced compared to that of MCA. In biodistribution study, 124 I-FIAU uptake after GCV therapy significantly decreased compared with pretreatment levels (34.8 13.67 %ID/g vs 7.6 2.59 %ID/g) and bioluminescent signal was also significantly decreased compared with pretreatment levels. In small animal PET imaging, 124 I-FIAU selectively localized in HSV -tk expressing tumor and the therapeutic efficacy of GCV treatment was evaluated by 124 I-FIAU PET imaging. 124 I-FIAU PET and bioluminescence imaging in HSV-tk suicide gene therapy were effective to evaluate the therapeutic response. 124 I-FIAU may serve as an efficient and selective agent for monitoring of transduced HSV1-tk gene expression in vivo in clinical trials

  15. Hard beta and gamma emissions of 124I. Impact on occupational dose in PET/CT.

    Science.gov (United States)

    Kemerink, G J; Franssen, R; Visser, M G W; Urbach, C J A; Halders, S G E A; Frantzen, M J; Brans, B; Teule, G J J; Mottaghy, F M

    2011-01-01

    The hard beta and gamma radiation of 124I can cause high doses to PET/CT workers. In this study we tried to quantify this occupational exposure and to optimize radioprotection. Thin MCP-Ns thermoluminescent dosimeters suitable for measuring beta and gamma radiation were used for extremity dosimetry, active personal dosimeters for whole-body dosimetry. Extremity doses were determined during dispensing of 124I and oral administration of the activity to the patient, the body dose during all phases of the PET/CT procedure. In addition, dose rates of vials and syringes as used in clinical practice were measured. The procedure for dispensing 124I was optimized using newly developed shielding. Skin dose rates up to 100 mSv/min were measured when in contact with the manufacturer's vial containing 370 MBq of 124I. For an unshielded 5 ml syringe the positron skin dose was about seven times the gamma dose. Before optimization of the preparation of 124I, using an already reasonably safe technique, the highest mean skin dose caused by handling 370 MBq was 1.9 mSv (max. 4.4 mSv). After optimization the skin dose was below 0.2 mSv. The highly energetic positrons emitted by 124I can cause high skin doses if radioprotection is poor. Under optimized conditions occupational doses are acceptable. Education of workers is of paramount importance.

  16. Development of Nuclear Fuel Remote Fabrication Technology

    International Nuclear Information System (INIS)

    Lee, Jung Won; Yang, M. S.; Kim, S. S. and others

    2005-04-01

    The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world

  17. Introduction to Exxon nuclear fuel fabrication plant

    International Nuclear Information System (INIS)

    Schneider, R.A.

    1985-01-01

    The Exxon Nuclear low-enriched uranium fuel fabrication plant in Richland, Washington produces fuel assemblies for both pressurized water and boiling water reactors. The Richland plant was the first US bulk-handling facility selected by the IAEA for inspection under the US-IAEA Safeguards Agreement. The plant was under IAEA inspection from March 1981 through October 1983. This text provides a written description of the plant layout, operation and process. The text also includes a one ton-a-day model (or reference) plant which was adapted from the Exxon Nuclear plant. The Model Plant provides a generic example of a low-enriched uranium (LEU) bulk-handling facility. The Model Plant is used to illustrate in a more quantitative way some of the key safeguards requirements for a bulk-handling facility

  18. Study of {sup 124} I contamination in {sup 123} I used in medical applications

    Energy Technology Data Exchange (ETDEWEB)

    El-Samman, H [Faculty of Science, Menoufia University, Shibin El-Kom (Egypt); Arafa, W [Physics Department, Faculty of Women, Ain Shams University, Cairo (Egypt)

    1997-12-31

    The decay of 0.2 mCi capsules of iodide ({sup 123} I) used for diagnostic purpose and delivered to hamad hospital in Qatar, was studied using HPGe detector of (30% efficiency and 1.8 KeV energy resolution), coupled to a computer based 4096 multichannel analyzer. The acquisition parameters were controlled by computer program. The gamma spectra were analyzed using well developed gamma spectrum analysis program gamanl. Results showed that the isotope used is not pure {sup 123} I but it is a mixture of {sup 123} I and {sup 124} I. The percentage of the unwanted {sup 124} I isotope was estimated to be 15%. The dose taken by the patient due to the unwanted {sup 124} I isotope was estimated. Half-lives time of the {sup 123} I and {sup 123} I isotopes were determined with high accuracy and compared to the published values. 3 tabs.

  19. Autoradiolytic decomposition and reductant-free sodium sup 124 I- and sup 123 I-iodide

    Energy Technology Data Exchange (ETDEWEB)

    Sajjad, M.; Lambrecht, R.M.; Bakr, S.A. (King Faisal Specialist Hospital and Research Centre, Riyadh (Saudi Arabia). Radionuclide and Cyclotron Operations)

    1990-01-01

    The presence of salts and metal cations in {sup 124}I- and {sup 123}I-sodium iodide solutions separated from {sup 124}Te targets promots autoradiolytic decomposition of iodide to several different iodine species dependent upon the chemical environment. The stabilization of the radioiodine as iodide by removal of trace salts and trace metal cations and in the absence of reducing agents is described. The high specific activity {sup 123}I- and {sup 124}I-iodide is suitable for labeling antibodies, proteins and radiopharmaceuticals. (orig.).

  20. Nuclear fuel control in fuel fabrication plants

    International Nuclear Information System (INIS)

    Seki, Yoshitatsu

    1976-01-01

    The basic control problems of measuring uranium and of the environment inside and outside nuclear fuel fabrication plants are reviewed, excluding criticality prevention in case of submergence. The occurrence of loss scraps in fabrication and scrap-recycling, the measuring error, the uranium going cut of the system, the confirmation of the presence of lost uranium and the requirement of the measurement control for safeguard make the measurement control very complicated. The establishment of MBA (material balance area) and ICA (item control area) can make clearer the control of inventories, the control of loss scraps and the control of measuring points. Besides the above basic points, the following points are to be taken into account: 1) the method of confirmation of inventories, 2) the introduction of reliable NDT instruments for the rapid check system for enrichment and amount of uranium, 3) the introduction of real time system, and 4) the clarification of MUF analysis and its application to the reliability check of measurement control system. The environment control includes the controls of the uranium concentration in factory atmosphere, the surface contamination, the space dose rate, the uranium concentration in air and water discharged from factories, and the uranium in liquid wastes. The future problems are the practical restudy of measurement control under NPT, the definite plan of burglary protection and the realization of the disposal of solid wastes. (Iwakiri, K.)

  1. 124I-PET dosimetry in advanced differentiated thyroid cancer: therapeutic impact

    International Nuclear Information System (INIS)

    Freudenberg, L.S.; Jentzen, W.; Goerges, R.; Knust, J.; Bockisch, A.; Marlowe, R.J.

    2007-01-01

    Purpose: This study evaluated the impact of 124 I-positron emission tomography (PET) dosimetry on post-primary surgery therapy in radioiodine-naive patients with advanced differentiated thyroid cancer (DTC). Patients, material, methods: In each of 28 thyroidectomized patients with high-risk DTC (one or more of pT4, pN1 or pM1), we gave 23-50 MBq of 124 I as an oral capsule and performed PET dosimetry to calculate the individualized therapeutic 131 I activity that would, insofar as possible, achieve a radioiodine dose ≥ 100 Gy to all metastases without exceeding 2 Gy to the blood (a surrogate for bone marrow toxicity). We thus determined the absorbed lesion dose per GBq of administered 131 I activity (LDpA) based on serial PET (4, 24, 48, 72 and 96 h after oral 124 I intake) and PET/computed tomography (25 h after 124 I intake) and the critical blood activity (CBA) based on blood and whole-body radiation counting (2, 4, 24, 48, 72, 96 h after 124 I intake). We compared the dosimetry-based interventions with our standard empirical protocol. Results: 25 patients had a total of 126 iodine-positive metastases. 18 (72%) of the 25 had solely iodine-avid metastases, while seven (28%) had both iodine-avid and -non-avid metastases. In two patients (8%), none of the iodine-avid metastases could have been practically treated with a sufficient radiation dose. Relative to the empirical protocol, 124 I-PET dosimetry findings changed management in 7 (25%) patients, e. g. allowing application of activities >11 GBq 131 I. Further changes included implementation of hematological back-up in a patient found to be at risk of life-threatening marrow toxicity, and early multimodal therapy in 9 (32%) patients. Conclusion: 124 I-PET dosimetry is a useful routine procedure in advanced DTC and may allow safer or more effective radioiodine activities and earlier multimodal interventions than do standard empirical protocols. (orig.)

  2. {sup 124}I-PET dosimetry in advanced differentiated thyroid cancer: therapeutic impact

    Energy Technology Data Exchange (ETDEWEB)

    Freudenberg, L.S.; Jentzen, W.; Goerges, R.; Knust, J.; Bockisch, A. [Duisburg-Essen Univ., Essen (Germany). Dept. of Nuclear Medicine; Petrich, T. [Medizinische Hochschule Hannover (Germany); Marlowe, R.J.

    2007-07-01

    Purpose: This study evaluated the impact of {sup 124}I-positron emission tomography (PET) dosimetry on post-primary surgery therapy in radioiodine-naive patients with advanced differentiated thyroid cancer (DTC). Patients, material, methods: In each of 28 thyroidectomized patients with high-risk DTC (one or more of pT4, pN1 or pM1), we gave 23-50 MBq of {sup 124}I as an oral capsule and performed PET dosimetry to calculate the individualized therapeutic {sup 131}I activity that would, insofar as possible, achieve a radioiodine dose {>=} 100 Gy to all metastases without exceeding 2 Gy to the blood (a surrogate for bone marrow toxicity). We thus determined the absorbed lesion dose per GBq of administered {sup 131}I activity (LDpA) based on serial PET (4, 24, 48, 72 and 96 h after oral {sup 124}I intake) and PET/computed tomography (25 h after {sup 124}I intake) and the critical blood activity (CBA) based on blood and whole-body radiation counting (2, 4, 24, 48, 72, 96 h after {sup 124}I intake). We compared the dosimetry-based interventions with our standard empirical protocol. Results: 25 patients had a total of 126 iodine-positive metastases. 18 (72%) of the 25 had solely iodine-avid metastases, while seven (28%) had both iodine-avid and -non-avid metastases. In two patients (8%), none of the iodine-avid metastases could have been practically treated with a sufficient radiation dose. Relative to the empirical protocol, {sup 124}I-PET dosimetry findings changed management in 7 (25%) patients, e. g. allowing application of activities >11 GBq {sup 131}I. Further changes included implementation of hematological back-up in a patient found to be at risk of life-threatening marrow toxicity, and early multimodal therapy in 9 (32%) patients. Conclusion: {sup 124}I-PET dosimetry is a useful routine procedure in advanced DTC and may allow safer or more effective radioiodine activities and earlier multimodal interventions than do standard empirical protocols. (orig.)

  3. Quantitative performance evaluation of 124I PET/MRI lesion dosimetry in differentiated thyroid cancer

    Science.gov (United States)

    Wierts, R.; Jentzen, W.; Quick, H. H.; Wisselink, H. J.; Pooters, I. N. A.; Wildberger, J. E.; Herrmann, K.; Kemerink, G. J.; Backes, W. H.; Mottaghy, F. M.

    2018-01-01

    The aim was to investigate the quantitative performance of 124I PET/MRI for pre-therapy lesion dosimetry in differentiated thyroid cancer (DTC). Phantom measurements were performed on a PET/MRI system (Biograph mMR, Siemens Healthcare) using 124I and 18F. The PET calibration factor and the influence of radiofrequency coil attenuation were determined using a cylindrical phantom homogeneously filled with radioactivity. The calibration factor was 1.00  ±  0.02 for 18F and 0.88  ±  0.02 for 124I. Near the radiofrequency surface coil an underestimation of less than 5% in radioactivity concentration was observed. Soft-tissue sphere recovery coefficients were determined using the NEMA IEC body phantom. Recovery coefficients were systematically higher for 18F than for 124I. In addition, the six spheres of the phantom were segmented using a PET-based iterative segmentation algorithm. For all 124I measurements, the deviations in segmented lesion volume and mean radioactivity concentration relative to the actual values were smaller than 15% and 25%, respectively. The effect of MR-based attenuation correction (three- and four-segment µ-maps) on bone lesion quantification was assessed using radioactive spheres filled with a K2HPO4 solution mimicking bone lesions. The four-segment µ-map resulted in an underestimation of the imaged radioactivity concentration of up to 15%, whereas the three-segment µ-map resulted in an overestimation of up to 10%. For twenty lesions identified in six patients, a comparison of 124I PET/MRI to PET/CT was performed with respect to segmented lesion volume and radioactivity concentration. The interclass correlation coefficients showed excellent agreement in segmented lesion volume and radioactivity concentration (0.999 and 0.95, respectively). In conclusion, it is feasible that accurate quantitative 124I PET/MRI could be used to perform radioiodine pre-therapy lesion dosimetry in DTC.

  4. Development and preclinical evaluation of new 124I-folate conjugates for PET imaging of folate receptor-positive tumors

    International Nuclear Information System (INIS)

    AlJammaz, I.; Al-Otaibi, B.; Al-Rumayan, F.; Al-Yanbawi, S.; Amer, S.; Okarvi, S.M.

    2014-01-01

    In an attempt to develop new folate radiotracers with favorable biochemical properties for detecting folate receptor-positive cancers, we have synthesized [ 124 I]-SIB- and [ 124 I]-SIP-folate conjugates using a straightforward and two-step simple reactions. Radiochemical yields for [ 124 I]-SIB- and [ 124 I]-SIP-folate conjugates were greater than 90 and 60% respectively, with total synthesis time of 30–40 min. Radiochemical purities were always greater than 98% without HPLC purification. These synthetic approaches hold considerable promise as rapid and simple method for 124 I-folate conjugate preparation with high radiochemical yield in short synthesis time. In vitro tests on KB cell line showed that the significant amounts of the radioconjugates were associated with cell fractions. In vivo characterization in normal Balb/c mice revealed rapid blood clearance of these radioconjugates and favorable biodistribution profile for [ 124 I]-SIP-folate conjugate over [ 124 I]-SIB-folate conjugate. Biodistribution studies of [ 124 I]-SIP-folate conjugate in nude mice bearing human KB cell line xenografts, demonstrated significant tumor uptake. The uptake in the tumors was blocked by excess injection of folic acid, suggesting a receptor-mediated process. These results demonstrate that [ 124 I]-SIP-folate conjugate may be useful as a molecular probe for detecting and staging of folate receptor-positive cancers, such as ovarian cancer and their metastasis as well as monitoring tumor response to treatment

  5. 124I-Epidepride: A PET radiotracer for extended imaging of dopamine D2/D3 receptors

    International Nuclear Information System (INIS)

    Pandey, Suresh; Venugopal, Archana; Kant, Ritu; Coleman, Robert; Mukherjee, Jogeshwar

    2014-01-01

    Objectives: A new radiotracer, 124 I-epidepride, has been developed for the imaging of dopamine D2/3 receptors (D2/3Rs). 124 I-Epidepride (half-life of 124 I = 4.2 days) allows imaging over extended periods compared to 18 F-fallypride (half-life of 18 F = 0.076 days) and may maximize visualization of D2/3Rs in the brain and pancreas (allowing clearance from adjacent organs). D2/3Rs are also present in pancreatic islets where they co-localize with insulin to produce granules and may serve as a surrogate marker for imaging diabetes. Methods: 124 I-Epidepride was synthesized using N-[[(2S)-1-ethylpyrrolidin-2-yl]methyl]-5-tributyltin-2, 3-dimethoxybenzamide and 124 I-iodide under no carrier added condition. Rats were used for in vitro and in vivo imaging. Brain slices were incubated with 124 I-epidepride (0.75 μCi/cc) and nonspecific binding measured with 10 μM haloperidol. Autoradiograms were analyzed by OptiQuant. 124 I-Epidepride (0.2 to 0.3 mCi, iv) was administered to rats and brain uptake at 3 hours, 24 hours, and 48 hours post injection was evaluated. Results: 124 I-Epidepride was obtained with 50% radiochemical yield and high radiochemical purity (> 95%). 124 I-Epidepride localized in the striatum with a striatum to cerebellum ratio of 10. Binding was displaced by dopamine and haloperidol. Brain slices demonstrated localization of 124 I-epidepride up until 48 hours in the striatum. However, the extent of binding was reduced significantly. Conclusions: 124 I-Epidepride is a new radiotracer suitable for extended imaging of dopamine D2/3 receptors and may have applications in imaging of receptors in the brain and monitoring pancreatic islet cell grafting

  6. International light water nuclear fuel fabrication supply. Are fabrication services assured?

    International Nuclear Information System (INIS)

    Rothwell, Geoffrey

    2010-01-01

    This paper examines the cost structure of fabricating light water reactor (LWR) fuel with low-enriched uranium (LEU, with less than 5% enrichment). The LWR-LEU fuel industry is decades old, and (except for the high entry cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added at Nth-of-a-Kind cost to the maximum capacity allowed by a site license. The industry appears to be competitive: nuclear fuel fabrication capacity is assured with many competitors and reasonable prices. However, nuclear fuel assurance has become an important issue for nations now to considering new nuclear power plants. To provide this assurance many proposals equate 'nuclear fuel banks' (which would require fuel for specific reactors) with 'LEU banks' (where LEU could be blended into nuclear fuel with the proper enrichment) with local fuel fabrication. The policy issues (which are presented, but not answered in this paper) become (1) whether the construction of new nuclear fuel fabrication facilities in new nuclear power nations could lead to the proliferation of nuclear weapons, and (2) whether nuclear fuel quality can be guaranteed under current industry arrangements, given that fuel failure at one reactor can lead to forced shutdowns at many others. (author)

  7. Cold-crucible fabrication of nuclear glasses

    International Nuclear Information System (INIS)

    Boen, R.

    2010-01-01

    Vitrification has stood the nuclear industry in good stead, for many years now, as a safe long-term conditioning technology for high-level waste. Major advances are nonetheless still being made, with the development of the cold-crucible technology, affording as it does new possibilities, in terms of volume reduction, and of extending the range of waste products amenable to incorporation. Indeed, by allowing higher melting temperatures to be achieved (1200 - 1400 C degrees), this process opens the way to a considerable increase in glass production capacities, and the fabrication of novel matrices, involving higher incorporation rates than current glasses. In the cold-crucible technology, materials put into the crucible are heated directly through induction. The crucible made of metal is cooled by water circulation. Where the glass comes into contact with the cold wall, a thin layer of solidified glass forms, with a thickness of 5-10 mm preventing the metal forming the crucible from coming into contact with the molten glass. A full scale pilot of the cold crucible was constructed at the La Hague vitrification workshop

  8. Nuclear fuel fabrication - developing indigenous capability

    International Nuclear Information System (INIS)

    Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K.

    1997-01-01

    Nuclear Fuel Complex (NFC), established in early 70's for production of fuel for PHWRs and BWRs in India, has made several improvements in different areas of fuel manufacturing. Starting with wire-wrap type of fuel bundles, NFC had switched over to split spacer type fuel bundle production in mid 80's. On the upstream side slurry extraction was introduced to prepare the pure uranyl nitrate solution directly from the MDU cake. Applying a thin layer of graphite to the inside of the tube was another modification. The Complex has developed cost effective and innovative techniques for these processes, especially for resistance welding of appendages on the fuel elements which has been a unique feature of the Indian PHWR fuel assemblies. Initially, the fuel fabrication plants were set-up with imported process equipment for most of the pelletisation and assembly operations. Gradually with design and development of indigenous equipment both for production and quality control, NFC has demonstrated total self reliance in fuel production by getting these special purpose machines manufactured indigenously. With the expertise gained in different areas of process development and equipment manufacturing, today NFC is in a position to offer know-how and process equipment at very attractive prices. The paper discusses some of the new processes that are developed/introduced in this field and describes different features of a few PLC based automatic equipment developed. Salient features of innovative techniques being adopted in the area Of UO 2 powder production are also briefly indicated. (author)

  9. PET imaging with the non-pure positron emitters: 55Co, 86Y and 124I

    DEFF Research Database (Denmark)

    Braad, Poul-Erik; Hansen, S B; Thisgaard, H

    2015-01-01

    PET/CT with non-pure positron emitters is a highly valuable tool in immuno-PET and for pretherapeutic dosimetry. However, imaging is complicated by prompt gamma coincidences (PGCs) that add an undesired background activity to the images. Time-of-flight (TOF) reconstruction improves lesion...... detectability in 18F-PET and can potentially also improve the signal-to-noise ratio in images acquired with non-pure positron emitters. Using the GE Discovery 690 PET/CT system, we evaluated the image quality with 55Co, 86Y and 124I, and the effect of PGC-correction and TOF-reconstruction on image quality...... and quantitation in a series of phantom studies. PET image quality and quantitation for all isotopes were significantly affected by PGCs. The effect was most severe with 86Y, and less, but comparable, with 55Co and 124I. PGC-correction improved the image quality and the quantitation accuracy dramatically for all...

  10. A recommendation for revised dose calibrator measurement procedures for 89Zr and 124I.

    Directory of Open Access Journals (Sweden)

    Bradley J Beattie

    Full Text Available Because of their chemical properties and multiday half lives, iodine-124 and zirconium-89 are being used in a growing number of PET imaging studies. Some aspects of their quantitation, however, still need attention. For (89Zr the PET images should, in principle, be as quantitatively accurate as similarly reconstructed 18F measurements. We found, however, that images of a 20 cm well calibration phantom containing (89Zr underestimated the activity by approximately 10% relative to a dose calibrator measurement (Capintec CRC-15R using a published calibration setting number of 465. PET images of (124I, in contrast, are complicated by the contribution of decays in cascade that add spurious coincident events to the PET data. When these cascade coincidences are properly accounted for, quantitatively accurate images should be possible. We found, however, that even with this correction we still encountered what appeared to be a large variability in the accuracy of the PET images when compared to dose calibrator measurements made using the calibration setting number, 570, recommended by Capintec. We derive new calibration setting numbers for (89Zr and (124I based on their 511 keV photon peaks as measured on an HPGe detector. The peaks were calibrated relative to an 18F standard, the activity level of which was precisely measured in a dose calibrator under well-defined measurement conditions. When measuring (89Zr on a Capintec CRC-15R we propose the use of calibration setting number 517. And for (124I, we recommend the use of a copper filter surrounding the sample and the use of calibration setting number 494. The new dose calibrator measurement procedures we propose will result in more consistent and accurate radioactivity measurements of (89Zr and (124I. These and other positron emitting radionuclides can be accurately calibrated relative to 18F based on measurements of their 511 keV peaks and knowledge of their relative positron abundances.

  11. Remote fabrication of nuclear fuel: a secure automated fabrication overview

    International Nuclear Information System (INIS)

    Nyman, D.H.; Benson, E.M.; Yatabe, J.M.; Nagamoto, T.T.

    1981-01-01

    An automated line for the fabrication of breeder reactor fuel pins is being developed. The line will be installed in the Fuels and Materials Examination Facility (FMEF) presently under construction at the Hanford site near Richland, Washington. The application of automation and remote operations to fuel processing technology is needed to meet program requirements of reduced personnel exposure, enhanced safeguards, improved product quality, and increased productivity. Commercially available robots are being integrated into operations such as handling of radioactive material within a process operation. These and other automated equipment and chemistry analyses systems under development are described

  12. Fabrication of pressure vessels for nuclear power plants

    International Nuclear Information System (INIS)

    Sampaio, M.S.P. de

    1982-01-01

    The status of the technology used in the fabrication of pressure vessel for nuclear power plants and the performance of the Brazilian industry in this area are presented. The followng aspects are discussed: qualification of the industries for the supplying equipment in its requirement categories; the calculation of the components; the choice of the materials; the fabrication process; and, the destructive and nondestructive tests associated to the fabrication. (E.G.) [pt

  13. Prototype fuel fabrication for nuclear reactors of Laguna Verde

    International Nuclear Information System (INIS)

    Nocetti, C.; Torres, J.; Medrano, A.

    1996-01-01

    Four prototype fuel bundles for the Laguna Verde Nuclear Power Plant have been fabricated. the type of nuclear fuel produced is described and the process used is commented. As an example of the fabrication criteria adopted, the production model to determine the density of the U O 2 pellets for the different batches of ceramic powder is described. the results are evaluated using the statistical indexes C p and C pk . (author)

  14. Nuclear target foil fabrication for the Romano Event

    International Nuclear Information System (INIS)

    Weed, J.W.; Romo, J.G. Jr.; Griggs, G.E.

    1984-01-01

    The Vacuum Processes Lab, of LLNL's M.E. Dept. - Material Fabrication Division, was requested to provide 250 coated Parylene target foils for a nuclear physics experiment titled the ROMANO Event. Due to the developmental nature of some of the fabrication procedures, approximately 400 coated foils were produced to satisfy the event's needs. The foils were used in the experiment as subkilovolt x-ray, narrow band pass filters, and wide band ultraviolet filters. This paper is divided into three sections describing: (1) nuclear target foil fabrication, (2) Parylene substrate preparation and production, and (3) foil and substrate inspections

  15. Maintenance and Fabrication of Nuclear Electronic Equipment

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chong Eun; Koo, In Soo; Hong, Seok Boong (and others)

    2006-12-15

    Based on the development of Hanaro RMS software, the RMS software system for another radiation facility(RI and IMEF) at KAERI will be developed next year. Development and its application of a wire-mesh sensor could contribute to the safe operation of nuclear power plants and could be used to develop other precision sensors for nuclear applications. Development of an ERMS could be used not only in nuclear facility but also in other radiation application institution such as acceleration facility etc.

  16. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1985-01-01

    In the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors, the regulations have all been revised on the fabrication business of nuclear fuel materials. The revised regulations are given : application for permission of the fabrication business, application for permission of the alteration, application for approval of the design and the construction methods, application for approval of the alteration, application for the facilities inspection, facilities inspection, recordings, entry limitations etc. for controlled areas, measures concerning exposure radiation doses etc., operation of the fabrication facilities, transport within the site of the business, storage, disposal within the site of the business, security regulations, designation etc. of the licensed engineer of nuclear fuels, collection of reports, etc. (Mori, K.)

  17. Role of ion chromatograph in nuclear fuel fabrication process at Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Balaji Rao, Y.; Prasada Rao, G.; Prahlad, B.; Saibaba, N.

    2012-01-01

    The present paper discusses the different applications of ion chromatography followed in nuclear fuel fabrication process at Nuclear Fuel Complex. Some more applications of IC for characterization of nuclear materials and which are at different stages of method development at Control Laboratory, Nuclear Fuel Complex are also highlighted

  18. Dual monitoring using {sup 124}I-FIAU and bioluminescence for HSV1-tk suicide gene therapy

    Energy Technology Data Exchange (ETDEWEB)

    Lee, T. S.; Kim, J. H.; Kwon, H. C. [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of)] (and others)

    2007-07-01

    Herpes simplex virus type I thymidine kinase (HSV-tk) is the most common reporter gene and is used in cancer gene therapy with a prodrug nucleoside analog, ganciclovir (GCV). The aim of this study is to evaluate therapeutic efficacy of suicide gene therapy with 2'-fluoro-2'-deoxy-1-D-arabinofuranosyl-5-[{sup 124}I] iodouracil ({sup 124}I - FIAU) and bioluminescence in retrovirally HSV -tk and firefly luciferase transduced hepatoma model. The HSV -tk and firefly luciferase (Luc) was retrovirally transduced and expressed in MCA rat Morris hepatoma cells. Nude mice with subcutaneous tumors, MCA and MCA-TK-Luc, were subjected to GCV treatment (50mg/Kg/d intraperitoneally) for 5 day. PET imaging and biodistribution with ({sup 124}I-FIAU) were performed at before and after initiation of therapy with GCV. Bioluminescent signal was also measured during GCV treatment. Before GCV treatment, no significant difference in tumor volume was found in tumors between MCA and MCA-TK-Luc. After GCV treatment, tumor volume of MCA-TK-Luc markedly reduced compared to that of MCA. In biodistribution study, {sup 124}I-FIAU uptake after GCV therapy significantly decreased compared with pretreatment levels (34.8 13.67 %ID/g vs 7.6 2.59 %ID/g) and bioluminescent signal was also significantly decreased compared with pretreatment levels. In small animal PET imaging, {sup 124}I-FIAU selectively localized in HSV -tk expressing tumor and the therapeutic efficacy of GCV treatment was evaluated by {sup 124}I-FIAU PET imaging. {sup 124}I-FIAU PET and bioluminescence imaging in HSV-tk suicide gene therapy were effective to evaluate the therapeutic response. {sup 124}I-FIAU may serve as an efficient and selective agent for monitoring of transduced HSV1-tk gene expression in vivo in clinical trials.

  19. Comparison of Imaging Characteristics of 124I PET for Determination of Optimal Energy Window on the Siemens Inveon PET

    Directory of Open Access Journals (Sweden)

    A Ram Yu

    2016-01-01

    Full Text Available Purpose.124I has a half-life of 4.2 days, which makes it suitable for imaging over several days over its uptake and washout phases. However, it has a low positron branching ratio (23%, because of prompt gamma coincidence due to high-energy γ-photons (602 to 1,691 keV, which are emitted in cascade with positrons. Methods. In this study, we investigated the optimal PET energy window for 124I PET based on image characteristics of reconstructed PET. Image characteristics such as nonuniformities, recovery coefficients (RCs, and the spillover ratios (SORs of 124I were measured as described in NEMA NU 4-2008 standards. Results. The maximum and minimum prompt gamma coincidence fraction (PGF were 33% and 2% in 350~800 and 400~590 keV, respectively. The difference between best and worst uniformity in the various energy windows was less than 1%. The lowest SORs of 124I were obtained at 350~750 keV in nonradioactive water compartment. Conclusion. Optimal energy window should be determined based on image characteristics. Our developed correction method would be useful for the correction of high-energy prompt gamma photon in 124I PET. In terms of the image quality of 124I PET, our findings indicate that an energy window of 350~750 keV would be optimal.

  20. Modeling fabrication of nuclear components: An integrative approach

    Energy Technology Data Exchange (ETDEWEB)

    Hench, K.W.

    1996-08-01

    Reduction of the nuclear weapons stockpile and the general downsizing of the nuclear weapons complex has presented challenges for Los Alamos. One is to design an optimized fabrication facility to manufacture nuclear weapon primary components in an environment of intense regulation and shrinking budgets. This dissertation presents an integrative two-stage approach to modeling the casting operation for fabrication of nuclear weapon primary components. The first stage optimizes personnel radiation exposure for the casting operation layout by modeling the operation as a facility layout problem formulated as a quadratic assignment problem. The solution procedure uses an evolutionary heuristic technique. The best solutions to the layout problem are used as input to the second stage - a simulation model that assesses the impact of competing layouts on operational performance. The focus of the simulation model is to determine the layout that minimizes personnel radiation exposures and nuclear material movement, and maximizes the utilization of capacity for finished units.

  1. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    International Nuclear Information System (INIS)

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-01-01

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  2. Review of training methods employed in nuclear fuel fabrication plants

    International Nuclear Information System (INIS)

    Box, W.D.; Browder, F.N.

    1975-01-01

    A search of the literature through the Nuclear Safety Information Center revealed that 86 percent of the incidents that have occurred in fuel fabrication plants can be traced directly or indirectly to insufficient operator training. In view of these findings, a review was made of the training programs now employed by the nuclear fuel fabrication industry. Most companies give the new employee approximately 20 hours of orientation courses, followed by 60 to 80 hours of on-the-job training. It was concluded that these training programs should be expanded in both scope and depth. A proposed program is outlined to offer guidance in improving the basic methods currently in use

  3. Chemical aspects of nuclear fuel fabrication processes

    Energy Technology Data Exchange (ETDEWEB)

    Naylor, A; Ellis, J F; Watson, R H

    1986-04-01

    Processes used by British Nuclear Fuels plc for the conversion of uranium ore concentrates to uranium metal and uranium hexafluoride, are reviewed. Means of converting the latter compound, after enrichment, to sintered UO/sub 2/ fuel bodies are also described. An overview is given of the associated chemical engineering technology.

  4. Quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Abdelhalim, A.S.; Elsayed, A.A.; Shaaban, H.I.

    1988-01-01

    The department of metallurgy, NRC Inchass is embarking on a programme of on a laboratory scale, fuel pins containing uranium dioxide pellets are going to be produced. The department is making use of the expertise and equipment at present available and is going to utilize the new fuel pin fabrication unit which would be shortly in operation. The fabrication and testing of uranium dioxide pellets then gradually adapt them and develop, a national know how in this field. This would also involve building up of indigenous experience through proper training of qualified personnel. That are applied to ensure quality of U o 2 pellets, the techniques implemented, the equipment used and the specifications of the equipment presently available. The following parameters are subject to quality control tests: density. O/U ration, hydrogen content, microstructure, each property will be discussed, measurements related to U o 2 powders, including flow ability, bulk density, O/U ratio, bet surface area and water content will be critically discussed. Relevant tests to ensure Q C of pellets are reviewed. These include surface integrity, density, dimensions, microstructure.4 fig., 1 tab

  5. Maintenance and fabrication of nuclear electronic equipment

    International Nuclear Information System (INIS)

    Hong, Seok Boong; Chung, Chong Eun; Hwang, In Koo; Koo, In Soo; Park, Bum; Kim, Soo Hee; Lee, Seong Joo; Kim, Min Seok; Choi, Wha Lim

    2011-12-01

    - process equipment at PIEF, Chemical Analysis Team and RWFTF have been calibrated. - The electronic equipment and radiation equipment at RWTF and PIEF have been prepared. - Development and installation of integrated RMS software for Hanaro Cold Neutron Laboratory Building(CNLB) RMS, and development and performance upgrade of a portal monitor for CNLB. - Performance test of the Hardware/Software of digital unit controller has been performed, and functional upgrade of the Hardware/Software of stimulator for SMART MMIS performance test facility has also been performed. - A controller of high voltage power supply for a small electron beam generator and a controller for razer pinning and a remote monitoring apparatus of cathode power supply for a 0.2 Mev. small electron beam generator have been designed and fabricated. - Database construction for effective maintenance for the process equipment and radiation instruments are designed and constructed

  6. Nuclear waste package fabricated from concrete

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kennedy, J.M.

    1987-03-01

    After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400 0 C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs

  7. Artificial vision in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Dorado, P.

    2007-01-01

    The development of artificial vision techniques opens a door to the optimization of industrial processes which the nuclear industry cannot miss out on. Backing these techniques represents a revolution in security and reliability in the manufacturing of a highly technological products as in nuclear fuel. Enusa Industrias Avanzadas S. A. has successfully developed and implemented the first automatic inspection equipment for pellets by artificial vision in the European nuclear industry which is nowadays qualified and is already developing the second generation of this machine. There are many possible applications for the techniques of artificial vision in the fuel manufacturing processes. Among the practices developed by Enusa Industrias Avanzadas are, besides the pellets inspection, the rod sealing drills detection and positioning in the BWR products and the sealing drills inspection in the PWR fuel. The use of artificial vision in the arduous and precise processes of full inspection will allow the absence of human error, the increase of control in the mentioned procedures, the reduction of doses received by the personnel, a higher reliability of the whole of the operations and an improvement in manufacturing costs. (Author)

  8. Property-process relationships in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Tikare, V.

    2015-01-01

    Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)

  9. Fabrication of nanoporous nuclear track membranes

    International Nuclear Information System (INIS)

    Peng Liangqiang; Wang Shicheng; Ju Xin; Masaru Yoshida; Yasunari Maekawa

    2001-01-01

    Polyethylene terephthalate (PET) and polycarbonate (PC) films were irradiated by S, Kr and Xe ions and were illuminated with ultraviolet light. The normalized track etch rate for PET and PC films etched in different conditions were measured by conductometric experiments. It is shown that normalized track etch rate can be over 1000 for PET films, 2000 for PC films under optimized condition. TEM photographs of copper nanowires electroplated into nanoporous nuclear track membranes show that the narrowest wire diameter of copper nanowires is 20 nm and that the pore diameter calculated by conductometric experiments is in agreement with the wire diameter measured by TEM when the pore diameter is over 30 nm

  10. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of nuclear electricity generation in Mexico in 1976 is described: two nuclear reactors were under construction but no definite programme on the type and start-up dates for the next power plants existed. However, the existence of a general plan on nuclear power plants is mentioned, which, according to the latest estimates, will provide 10,000MW installed by 1990. The national intention, as laid down in an appropriate Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reloading of the two BWRs at the first national station in Laguna Verde, required at the end of 1981 and 1982, respectively. Before this can be achieved and to provide the relatively small amounts of fuel elements for the two reactors, Mexico must adopt a strategy of fuel elements fabrication. The two main options are analysed: (1) to delay local fabrication until a national nuclear programme has been defined, meanwhile purchasing abroad the necessary initial cores and refuelling; (2) to start local fabrication of fuel elements as soon as possible in order to provide the first refuelling of the first unit of Laguna Verde, confronting the economic risks of such a decision with the advantages of immediate action. Both options are analysed in detail, comparing them especially from the economic point of view. Current information from potential licensors for design and manufacture are used in the analysis. (author)

  11. Radiological surveillance in the nuclear fuel fabrication in Mexico

    International Nuclear Information System (INIS)

    Garcia A, J.; Reynoso V, R.; Delgado A, G.

    1996-01-01

    The objective of this report is to present the obtained results related to the application of the radiological safety programme established at the Nuclear Fuel Fabrication Pilot Plant (NFFPF) in Mexico, such as: surveillance methods, radiological protection criteria and regulations, radiation control and records and the application of ALARA recommendation. During the starting period from April 1994 to April 1995, at the NFFPF were made two nuclear fuel bundles a Dummy and other to be burned up in a BWR the mainly process activities are: UO 2 powder receiving, powder pressing for the pellets formation, pellets grinding, cleaning and drying, loading into a rod, Quality Control testing, nuclear fuel bundles assembly. The NFFPF is divided into an unsealed source area (pellets manufacturing plant) and into a sealed source area (rods fabrication plant). The control followed have helped to detect failures and to improve the safety programme and operation. (authors). 1 ref., 3 figs

  12. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  13. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  14. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  15. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1977-01-01

    As regards an application for permission of an fabricating business of nuclear fuel materials, it should describe the site of the fabricating facilities and the structure and equipments of buildings (fire-resistant, aseismatic, waterproof, ventilating and air-tight structures), etc. The business plan to be attached to the foregoing application should contain 1) scheduled date when the fabricating business starts, 2) scheduled amounts of products classified by the kinds in each business year within 5 years since the business starts, 3) the amount and the procurement plan of funds necessary for the operation, etc. For the permission of change of a fabricating business, an application must be filed. One who wants to obtain the permission of design and construction of fabricating facilities must file an application. One who wants to undergo inspection of the construction of fabricating facilities must file an application in which various items must be written. After such inspection has been done and it is regarded as passable, a certificate of passing inspection will be given. (Rikitake, Y.)

  16. Review of training methods employed in nuclear fuel fabrication plants

    International Nuclear Information System (INIS)

    Box, W.D.; Browder, F.N.

    A search of the literature through the Nuclear Safety Information Center revealed that approximately 86 percent of the incidents that have occurred in fuel fabrication plants can be traced directly or indirectly to insufficient operator training. In view of these findings, a review was made of the training programs now employed by the nuclear fuel fabrication industry. Most companies give the new employee approximately 20 h of orientation courses, followed by 60 to 80 h of on-the-job training. It was concluded that these training programs should be expanded in both scope and depth. A proposed program is outlined to offer guidance in improving the basic methods currently in use. (U.S.)

  17. Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Panda, Binayak; Shah, Sandeep

    2005-01-01

    Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

  18. Application of plasma deposition technology for nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Jung, I. H.; Moon, J. S.; Park, H. S.; Song, K. C.; Lee, C. Y.; Kang, K. H.; Ryu, H. J.; Kim, H. S.; Yang, M. S.

    2001-01-01

    Yttria-stabilized-zirconia (m.p. 2670.deg. C), was deposited by induction plasma spraying system with a view to develop a new nuclear fuel fabrication technology. To fabricate the dense pellets, the spraying condition was optimized through the process parameters such as, chamber pressure, plasma plate power, powder spraying distance, sheath gas composition, probe position particle size and its morphology. The results with a 5mm thick deposit on rectangular planar graphite substrates showed 97.11% theoretical density, when the sheath gas flow rate was Ar/H 2 120/20 L/min, probe position 8cm, particle size-75 μm and spraying distance 22cm. The microstructure of YSZ deposit by ICP was lamellae and columnar perpendicular to the spraying direction. In the bottom part near the substrate, small equiaxed grains bounded in a layer. In the middle part, relatively regular size of columnar grains with excellent bonding each other were distinctive

  19. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1979-01-01

    The regulations are entirely revised under the law for the regulations of nuclear materials, nuclear fuel materials and reactors and provisions concerning the fabricating business in the order for execution of the law. Basic concepts and terms are defined, such as: exposure dose; accumulative dose; controlled area; inspected surrounding area; employee and radioactive waste. The application for permission of the fabricating business shall include: location of processing facilities; structure of building structure and equipment of chemical processing facilities; molding facilities; structure and equipment of covering and assembling facilities, storage facilities of nuclear fuel materials and disposal facilities of radioactive waste, etc. Records shall be made and kept for particular periods in each works and place of enterprise on inspection of processing facilities, control of dose, operation, maintenance, accident of processing facilities and weather. Specified measures shall be taken in controlled area and inspected surrounding area to restrict entrance. Measures shall be made not to exceed permissible exposure dose for employees defined by the Director General of Science and Technology Agency. Inspection and check up of processing facilities shall be carried on by employees more than once a day. Operation of processing facilities, transportation in the works and enterprise, storage, disposal, safety securing, report and measures in dangerous situations, etc. are in detail prescribed. (Okada, K.)

  20. Process for the fabrication of nuclear fuel oxide pellets

    International Nuclear Information System (INIS)

    Francois, Bernard; Paradis, Yves.

    1977-01-01

    Process for the fabrication of nuclear fuel oxide pellets of the type for which particles charged with an organic binder -selected from the group that includes polyvinyl alcohol, carboxymethyl cellulose, polyvinyl compounds and methyl cellulose- are prepared from a powder of such an oxide, for instance uranium dioxide. These particles are then compressed into pellets which are then sintered. Under this process the binder charged particles are prepared by stirring the powder with a gas, spraying on to the stirred powder a solution or a suspension in a liquid of this organic binder in order to obtain these particles and then drying the particles so obtained with this gas [fr

  1. The fabrication of nozzles for nuclear components by welding

    International Nuclear Information System (INIS)

    Moraes, M.M.; Krausser, P.; Echeverria, J.A.V.

    1986-01-01

    A nozzle with medium outside diameter of 1000 mm and medium thickness of 150 mm composed integrally by deposited metal by submerged-arc (wire S3NiMo1, 0.5mm) was fabricated in NUCLEP. The nondestructive, mechanical, metallographic and chemical testing carried out in a test sample made by the same procedure and welding parameters, showed results according to specifications established for primary components for nuclear power plants, and the tests presented mechanical properties and tenacity better than similar nozzle samples. This nozzle is cheapest concerning to importations, in respecting to its forged similar. (M.C.K.) [pt

  2. Method of fabricating self-powered nuclear radiation detector assemblies

    International Nuclear Information System (INIS)

    Playfoot, K.; Bauer, R.F.; Sekella, Y.M.

    1982-01-01

    In a method of fabricating a self-powered nuclear radiation detector assembly an emitter electrode wire and signal cable center wire are connected and disposed within the collector electrode tubular sheath with compressible insulating means disposed between the wires and the tubular sheath. The above assembly is reduced in diameter while elongating the tubular sheath and the emitter wire and signal cable wire. The emitter wire is reduced to a predetermined desired diameter, and is trimmed to a predetermined length. An end cap is hermetically sealed to the tubular sheath at the extending end of the emitter with insulating means between the emitter end and the end cap. (author)

  3. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1980-11-01

    The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

  4. Beyond 18F-FDG: Characterization of PET/CT and PET/MR Scanners for a Comprehensive Set of Positron Emitters of Growing Application--18F, 11C, 89Zr, 124I, 68Ga, and 90Y.

    Science.gov (United States)

    Soderlund, A Therese; Chaal, Jasper; Tjio, Gabriel; Totman, John J; Conti, Maurizio; Townsend, David W

    2015-08-01

    This study aimed to investigate image quality for a comprehensive set of isotopes ((18)F, (11)C, (89)Zr, (124)I, (68)Ga, and (90)Y) on 2 clinical scanners: a PET/CT scanner and a PET/MR scanner. Image quality and spatial resolution were tested according to NU 2-2007 of the National Electrical Manufacturers Association. An image-quality phantom was used to measure contrast recovery, residual bias in a cold area, and background variability. Reconstruction methods available on the 2 scanners were compared, including point-spread-function correction for both scanners and time of flight for the PET/CT scanner. Spatial resolution was measured using point sources and filtered backprojection reconstruction. With the exception of (90)Y, small differences were seen in the hot-sphere contrast recovery of the different isotopes. Cold-sphere contrast recovery was similar across isotopes for all reconstructions, with an improvement seen with time of flight on the PET/CT scanner. The lower-statistic (90)Y scans yielded substantially lower contrast recovery than the other isotopes. When isotopes were compared, there was no difference in measured spatial resolution except for PET/MR axial spatial resolution, which was significantly higher for (124)I and (68)Ga. Overall, both scanners produced good images with (18)F, (11)C, (89)Zr, (124)I, (68)Ga, and (90)Y. © 2015 by the Society of Nuclear Medicine and Molecular Imaging, Inc.

  5. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  6. Longitudinal monitoring adipose-derived stem cell survival by PET imaging hexadecyl-4-{sup 124}I-iodobenzoate in rat myocardial infarction model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Hwan [Molecular Imaging Research Center, Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); School of Life Sciences and Biotechnology, Korea University, Seoul (Korea, Republic of); Woo, Sang-Keun; Lee, Kyo Chul; An, Gwang Il [Molecular Imaging Research Center, Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Pandya, Darpan [Department of Molecular Medicine, BK21 Plus KNU Biomedical Convergence Program, Kyungpook National University, Daegu (Korea, Republic of); Park, Noh Won; Nahm, Sang-Soep; Eom, Ki Dong [College of Veterinary Medicine, Konkuk University, Seoul (Korea, Republic of); Kim, Kwang Il; Lee, Tae Sup [Molecular Imaging Research Center, Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Kim, Chan Wha [School of Life Sciences and Biotechnology, Korea University, Seoul (Korea, Republic of); Kang, Joo Hyun [Molecular Imaging Research Center, Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Yoo, Jeongsoo, E-mail: yooj@knu.ac.kr [Department of Molecular Medicine, BK21 Plus KNU Biomedical Convergence Program, Kyungpook National University, Daegu (Korea, Republic of); Lee, Yong Jin, E-mail: yjlee@kirams.re.kr [Molecular Imaging Research Center, Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of)

    2015-01-02

    Highlights: • We developed a safe, simple and appropriate stem cell labeling method with {sup 124}I-HIB. • ADSC survival can be monitored with PET in MI model via direct labeling. • Tracking of ADSC labeled with {sup 124}I-HIB was possible for 3 days in MI model using PET. • ADSC viability and differentiation were not affected by {sup 124}I-HIB labeling. • Survival of ADSC in living bodies can be longitudinally tracked with PET imaging. - Abstract: This study aims to monitor how the change of cell survival of transplanted adipose-derived stem cells (ADSCs) responds to myocardial infarction (MI) via the hexadecyl-4-{sup 124}I-iodobenzoate ({sup 124}I-HIB) mediated direct labeling method in vivo. Stem cells have shown the potential to improve cardiac function after MI. However, monitoring of the fate of transplanted stem cells at target sites is still unclear. Rat ADSCs were labeled with {sup 124}I-HIB, and radiolabeled ADSCs were transplanted into the myocardium of normal and MI model. In the group of {sup 124}I-HIB-labeled ADSC transplantation, in vivo imaging was performed using small-animal positron emission tomography (PET)/computed tomography (CT) for 9 days. Twenty-one days post-transplantation, histopathological analysis and apoptosis assay were performed. ADSC viability and differentiation were not affected by {sup 124}I-HIB labeling. In vivo tracking of the {sup 124}I-HIB-labeled ADSCs was possible for 9 and 3 days in normal and MI model, respectively. Apoptosis of transplanted cells increased in the MI model compared than that in normal model. We developed a direct labeling agent, {sup 124}I-HIB, and first tried to longitudinally monitor transplanted stem cell to MI. This approach may provide new insights on the roles of stem cell monitoring in living bodies for stem cell therapy from pre-clinical studies to clinical trials.

  7. Quality control in nuclear fuel fabrication on the inspection basis

    International Nuclear Information System (INIS)

    Fuentes S, A.

    1997-01-01

    Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B 'Quality assurement for nuclear plants' specially in the criteria 'Inspections' that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)

  8. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of the nucleoelectrical generation in Mexico by 1976 is described: two nuclear reactors under construction but no defined program on the type and start-up dates for the next power plants. However the existence of a general plan on nuclear power plants is mentioned, which, according to the last estimates reaches to 10,000 MW installed by 1990. The national intension, definitely expressed in the Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reload for the two BWR's at the first national station in Laguna Verde, which will be required at the end of 1981 and of 1982, respectively. Before such circumstances and the relatively short amounts of fuel elements that should be produced for those two unique reactors, Mexico already has to adopt a strategy to follow in respect to fuel elements fabrication. The two main options are analyzed: 1. To delay the local fabrication until a National Nuclear Program may be defined, meanwhile purchasing abroad the necessary reloads and initial cores; and 2. To start as soon as possible the local fuel elements fabrication in order to supply fuel for the first reload of the first unit of Laguna Verde, confronting the economical risks of such posture with the advantages of an immediate action. Both options are analyzed in detail comparing them specially under the economic point of view, standing out immediately the big effect of some factors which are economically imponderable, as experience and independance that would be gained with the second option. Emphasis is made on the advantages and risks of any case. According to the first option and once a National Program is defined, the work would be heavy but of simple strategy. On the contrary, the second option requires the adoption of a more complicated strategy, as either the project of the factory as its initial operation should be made under transient conditions, in view of the expected future expansion still

  9. {sup 124}I-L19-SIP for immuno-PET imaging of tumour vasculature and guidance of {sup 131}I-L19-SIP radioimmunotherapy

    Energy Technology Data Exchange (ETDEWEB)

    Tijink, Bernard M.; Perk, Lars R.; Budde, Marianne; Stigter-van Walsum, Marijke; Leemans, C.R. [VU University Medical Center, Department of Otolaryngology/Head and Neck Surgery, Amsterdam (Netherlands); Visser, Gerard W.M.; Kloet, Reina W. [VU University Medical Center, Nuclear Medicine and PET Research, Amsterdam (Netherlands); Dinkelborg, Ludger M. [Bayer Schering Pharma AG, Global Drug Discovery, Berlin (Germany); Neri, Dario [Swiss Federal Institute of Technology, Institute of Pharmaceutical Sciences, Zurich (Switzerland); Dongen, Guus A.M.S. van [VU University Medical Center, Department of Otolaryngology/Head and Neck Surgery, Amsterdam (Netherlands); VU University Medical Center, Nuclear Medicine and PET Research, Amsterdam (Netherlands)

    2009-08-15

    The human monoclonal antibody (MAb) fragment L19-SIP is directed against extra domain B (ED-B) of fibronectin, a marker of tumour angiogenesis. A clinical radioimmunotherapy (RIT) trial with {sup 131}I-L19-SIP was recently started. In the present study, after GMP production of {sup 124}I and efficient production of {sup 124}I-L19-SIP, we aimed to demonstrate the suitability of {sup 124}I-L19-SIP immuno-PET for imaging of angiogenesis at early-stage tumour development and as a scouting procedure prior to clinical {sup 131}I-L19-SIP RIT. {sup 124}I was produced in a GMP compliant way via {sup 124}Te(p,n){sup 124}I reaction and using a TERIMO trademark module for radioiodine separation. L19-SIP was radioiodinated by using a modified version of the IODO-GEN method. The biodistribution of coinjected {sup 124}I- and {sup 131}I-L19-SIP was compared in FaDu xenograft-bearing nude mice, while {sup 124}I PET images were obtained from mice with tumours of <50 to {proportional_to}700 mm{sup 3}. {sup 124}I was produced highly pure with an average yield of 15.4 {+-} 0.5 MBq/{mu}Ah, while separation yield was {proportional_to}90% efficient with <0.5% loss of TeO{sub 2}. Overall labelling efficiency, radiochemical purity and immunoreactive fraction were for {sup 124}I-L19-SIP: {proportional_to}80, 99.9 and >90%, respectively. Tumour uptake was 7.3{+-}2.1, 10.8{+-}1.5, 7.8{+-}1.4, 5.3{+-}0.6 and 3.1{+-}0.4%ID/g at 3, 6, 24, 48 and 72 h p.i., resulting in increased tumour to blood ratios ranging from 6.0 at 24 h to 45.9 at 72 h p.i. Fully concordant labelling and biodistribution results were obtained with {sup 124}I- and {sup 131}I-L19-SIP. Immuno-PET with {sup 124}I-L19-SIP using a high-resolution research tomograph PET scanner revealed clear delineation of the tumours as small as 50 mm{sup 3} and no adverse uptake in other organs. {sup 124}I-MAb conjugates for clinical immuno-PET can be efficiently produced. Immuno-PET with {sup 124}I-L19-SIP appeared qualified for sensitive

  10. Novel fabrication of silicon carbide based ceramics for nuclear applications

    Science.gov (United States)

    Singh, Abhishek Kumar

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

  11. Determination and quantification of impurities found in samples of {sup 124}I using gamma spectrometry; Determinacao e quantificacao de impurezas encontradas em amostra de {sup 124}I usando a espectrometria gama

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Ronaldo Lins da; Delgado, Jose Ubiratan; Araujo, Miriam Taina Ferreira de; Laranjeira, Adilson da Silva; Poledna, Roberto; Veras, Eduardo de; Almeida, Maria Candida M. [Instituto de Radioprotecao e Dosimetria, (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Braghirolli, Ana Maria S. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    {sup 124}I, positron emitter is produced in the IEN/CNEN-RJ and used in diagnosis image of tumors. In this radioisotope production process impurities appear and health agency requires that the level of these characteristics is quantified. These radionuclides emit gamma and X-radiation, allowing the identification and quantitation by gamma spectrometry. With the use of HPGE detector, coupled with the efficiency curve, was identified {sup 125}I and {sup 126}I. The impurity levels measured in the sample were in the range of 0.5% to 90%, respectively, indicating the feasibility of the method for controlling the quality of the radiopharmaceutical.

  12. Support of the radioactive waste treatment nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Park, H.H.; Han, K.W.; Lee, B.J.; Shim, G.S.; Chung, M.S.

    1982-01-01

    Technical service of radioactive waste treatment in Daeduck Engineering Center includes; 1) Treatment of radioactive wastes from the nuclear fuel fabrication facility and from laboratories. 2) Establishing a process for intermediate treatment necessary till the time when RWTF is in completion. 3) Technical evaluation of unit processes and equipments concerning RWTF. About 35 drums (8 m 3 ) of solid wastes were treated and stored while more than 130 m 3 of liquid wastes were disposed or stored. A process with evaporators of 10 1/hr in capacity, a four-stage solvent washer, storage tanks and disposal system was designed and some of the equipments were manufactured. Concerning RWTF, its process was reviewed technically and emphasis were made on stability of the bituminization process against explosion, function of PAAC pump, decontamination, and finally on problems to be solved in the comming years. (Author)

  13. Process for the fabrication of a nuclear fuel

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1970-01-01

    Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

  14. Investigation of the thermal performance of {sup nat}Te target for {sup 124}I production in the RARS cyclotron

    Energy Technology Data Exchange (ETDEWEB)

    Azizakram, Hamid; Zolfagharpour, Farhad [Univ. of Mohaghegh Ardabili, Ardabil (Iran, Islamic Republic of). Dept. of Physics; Sadeghi, Mahdi [Iran Univ. of Medical Science, Tehran (Iran, Islamic Republic of). Medical Physics Dept.; Ashtari, Parviz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School (RARS); Nikjou, Amir [Payam Noor Univ., Tehran (Iran, Islamic Republic of). Dept. of Physics

    2017-12-15

    Regarding the low thermal conductivity of {sup nat}Te element, the provision of an effective cooling system is one of the critical issues in cyclotron targetry to prevent melting of target matter during the irradiation to {sup 124}I production via {sup nat}Te(p,xn){sup 124}I reaction. Heat transfer on Te target and efficiency of cooling fluid in the solid target system have been simulated based on a Finite Element Analysis (FEA) code for the thermal behavior of the target during the irradiation and under different beam currents, coolant flow rates, substrate matters and target geometry. The results on the routinely used solid target in Radiation Application Research School (RARS) cyclotron showed that in a 3 m/s coolant flow rate, by using a fined-cooling system and a nickel substrate coated on copper backing plate, the irradiation beam current can be raised up to 180 μA without any risk of melting. The cooling flow rates greater than 3 m/s do not noticeably improve the heat dispersion of target layer. As expected, a linear increase was observed for the temperature and temperature gradient of plates in the beam currents of 100-300 μA.

  15. Interpretation of bioassay data from nuclear fuel fabrication workers

    International Nuclear Information System (INIS)

    Melo, D.; Xavier, M.

    2005-01-01

    Full text: In nuclear fuel fabrication facilities, workers are exposed to different compounds of enriched uranium. Although in this kind of facility the main route of intake is inhalation, ingestion may occur in some situations. The interpretation of the bioassay data is very complex, since it is necessary taking into account all the different parameters, which is a big challenge. Due to the high cost of the individual monitoring programme for internal dose assessment in the routine monitoring programmes, usually only one type of measurement is assigned. In complex situations like the one described in this paper, where several parameters can compromise the accuracy of the bioassay interpretation it is need to have a combination of techniques to evaluate the internal dose. According to ICRP 78 (1997), the general order of preference in terms of accuracy of interpretation is: body activity measurement, excreta analysis and personal air sampling. Results of monitoring of working environment may provide information that assists in interpretation on particle size, chemical form and solubility, time of intake. A group of seventeen workers from controlled area of the fuel fabrication facility was selected to evaluate the internal dose using all different available techniques during a certain period. The workers were monitored for determination of uranium content in the daily urinary and faecal excretion (collected over a period of 3 consecutive days), chest counting and personal air sampling. The results have shown that at least two types of sensitivity techniques must be used, since there are some sources of uncertainties on the bioassay interpretation, like mixture of uranium compounds intake and different routes of intake. The combination of urine and faeces analysis has shown to be the more appropriate methodology for assessing internal dose in this situation. (author)

  16. Applications of ultrasonic phased array technique during fabrication of nuclear tubing and other components for the Indian nuclear power program

    International Nuclear Information System (INIS)

    Kapoor, K.

    2015-01-01

    Ultrasonic phased array technique has been applied in fabrication of nuclear fuel and structural at NFC. The integrity of the nuclear fuel and structural components is most crucial as they are exposed to severe environment during operation leading to rapid degradation of its properties during its lifecycle. Nuclear Fuel Complex has mandate for the fabrication of the nuclear fuel and core structurals for Indian PHWRs/BWR, sub-assemblies for the PFBR and steam generator tubing for PFBR and PHWRs which are the most critical materials for the Indian Nuclear Power program. NDE during fabrication of these materials is thus most crucial as it provides the confidence to the designer for safe operation during its lifetime. Many of these techniques have to be developed in-house to meet unique requirements of high sensitivity, resolution and shape of the components. Some of the advancements in the NDE during the fabrication include use of ultrasonic phased array which is detailed in this paper

  17. Diagnosis and dosimetry in differentiated thyroid carcinoma using 124I PET: comparison of PET/MRI vs PET/CT of the neck

    International Nuclear Information System (INIS)

    Nagarajah, James; Jentzen, Walter; Hartung, Verena; Rosenbaum-Krumme, Sandra; Bockisch, Andreas; Stahl, Alexander; Mikat, Christian; Heusner, Till Alexander; Antoch, Gerald

    2011-01-01

    This study compares intrinsically coregistered 124 I positron emission tomography (PET) and CT (PET/CT) and software coregistered 124 I PET and MRI (PET/MRI) images for the diagnosis and dosimetry of thyroid remnant tissues and lymph node metastases in patients with differentiated thyroid carcinoma (DTC). After thyroidectomy, 33 high-risk DTC patients (stage III or higher) received 124 I PET/CT dosimetry prior to radioiodine therapy to estimate the absorbed dose to lesions and subsequently underwent a contrast-enhanced MRI examination of the neck. Images were evaluated by two experienced nuclear medicine physicians and two radiologists to identify the lesions and to categorize their presumable provenience, i.e. thyroid remnant tissue (TT), lymph node metastasis (LN) and inconclusive tissue. The categorization and dosimetry of lesions was initially performed with PET images alone (PET only). Subsequently lesions were reassessed including the CT and MRI data. The analyses were performed on a patient and on a lesion basis. Patient-based analyses showed that 26 of 33 (79%) patients had at least one lesion categorized as TT on PET only. Of these patients, 11 (42%) and 16 (62%) had a morphological correlate on CT and MRI, respectively, in at least one TT PET lesion. Twelve patients (36%) had at least one lesion classified as LN on PET only. Nine (75%) of these patients had a morphological correlate on both CT and MRI in at least one LN PET lesion. Ten patients (30%) showed at least one lesion on PET only classified as inconclusive. The classification was changed to a clear classification in two patients (two LN) by CT and in four (two TT, two LN) patients by MRI. Lesion-based analyses (n = 105 PET positive lesions) resulted in categorization as TT in 61 cases (58%), 16 (26%) of which had a morphological correlate on CT and 33 (54%) on MRI. A total of 29 lesions (27%) were classified as LN on PET, 18 (62%) of which had a morphological correlate on CT and 24 (83%) on MRI

  18. Induction plasma deposition technology for nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Jung, I. H.; Bae, K. K.; Lee, J. W.; Kim, T. K.; Yang, M. S.

    1998-01-01

    A study on induction plasma deposition with ceramic materials, yttria-stabilized-zirconia ZrO 2 -Y 2 O 3 (m.p. 2640 degree C), was conducted with a view of developing a new method for nuclear fuel fabrication. Before making dense pellets of more than 96%T.D., the spraying condition was optimized through the process parameters, such as chamber pressure, plasma plate power, powder spraying distance, sheath gas composition, probe position, particle size and powders of different morphology. The results with a 5mm thick deposit on rectangular planar graphite substrates showed a 97.11% theoretical density when the sheath gas flow rate was Ar/H 2 120/20 l/min, probe position 8cm, particle size -75 μm and spraying distance 22cm by AMDRY146 powder. The degree of influence of the main effects on density were powder morphology, particle size, sheath gas composition, plate power and spraying distance, in that order. Among the two parameter interactions, the sheath gas composition and chamber pressure affects density greatly. By using the multi-pellets mold of wheel type, the pellet density did not exceed 94%T.D., owing to the spraying angle

  19. Hydrothermal synthesis for fabrication and reprocessing of MOX nuclear fuel

    International Nuclear Information System (INIS)

    Ohta, Suguru; Yamamura, Tomoo; Shirasaki, Kenji; Satoh, Isamu; Shikama, Tatsuo

    2011-01-01

    To improve the nuclear proliferation resistance and to minimize use of chemicals, a new reprocessing and fabrication process of 'mixed oxide' (MOX) fuel was proposed and studied by using simulated spent fuel solutions. The process is consisting of the two steps, i.e. the removal of fission product (FP) from dissolved spent fuel by using carbonate solutions (Step-1), and hydrothermal synthesis of uranium dioxides (Step-2). In Step-1, rare earth (the precipitation ratio: 90%) and alkaline earth (10-50% for Sr) as FP were removed based on their low solubility of hydroxides and carbonate salts, with uranium kept dissolved for the certain carbonate solutions of weak base (Type 2) or mixtures of relatively strong base and weak base (Type 3). In Step-2, the features of uranium dioxides UO 2+x particles, i.e. stoichiometry (x=0.05-0.2), size (0.2-3 μm) and shape (cubic, spherical, rectangular parallelpiped, etc.), were controlled, and the cesium was removed down to 40 ppm by an addition of organic additives. The decontamination factors (DF) for cesium exceeds 10 5 , whereas the total DF of all the simulated FP were as low as the order of 10 which requires future studies for removal of alkaline earth, Re and Tc etc. (author)

  20. Ion-exchange separation of radioiodine and its application to production of {sup 124}I by alpha particle induced reactions on antimony

    Energy Technology Data Exchange (ETDEWEB)

    Shuza Uddin, Md. [Forschungszentrum Juelich (Germany). Inst. fuer Neurowissenschaften und Medizin, INM-5: Nuklearchemie; Atomic Energy Research Establishment, Inst. of Nuclear Science and Technology, Dhaka (Bangladesh); Qaim, Seyed M.; Spahn, Ingo; Spellerberg, Stefan; Scholten, Bernhard; Coenen, Heinz H. [Forschungszentrum Juelich (Germany). Inst. fuer Neurowissenschaften und Medizin, INM-5: Nuklearchemie; Hermanne, Alex [Vrije Univ. Brussel (Belgium). Cyclotron Lab.; Hossain, Syed Mohammod [Atomic Energy Research Establishment, Inst. of Nuclear Science and Technology, Dhaka (Bangladesh)

    2015-07-01

    The basic parameters related to radiochemical separation of iodine from tellurium and antimony by anion-exchange chromatography using the resin Amberlyst A26 were studied. The separation yield of {sup 124}I amounted to 96% and the decontamination factor from {sup 121}Te and {sup 122}Sb was > 10{sup 4}. The method was applied to the production of {sup 124}I via the {sup 123}Sb(α, 3n) reaction. In an irradiation of 110 mg of {sup nat}Sb{sub 2}O{sub 3} (thickness ∝0.08 g/cm{sup 2}) with 38 MeV α-particles at 1.2 μA beam current for 4 h, corresponding to the beam energy range of E{sub α} = 37 → 27 MeV, the batch yield of {sup 124}I obtained was 12.42 MBq and the {sup 125}I and {sup 126}I impurities amounted to 3.8% and 0.7%, respectively. The experimental batch yield of {sup 124}I amounted to 80% of the theoretically calculated value but the level of the radionuclidic impurities were in agreement with the theoretical values. About 96% of the radioiodine was in the form of iodide and the inactive impurities (Te, Sb, Sn) were below the permissible level. Due to the relatively high level of radionuclidic impurity the {sup 124}I produced would possibly be useful only for restricted local consumption or for animal experiments.

  1. Monte Carlo simulations of GeoPET experiments: 3D images of tracer distributions (18F, 124I and 58Co) in Opalinus clay, anhydrite and quartz

    Science.gov (United States)

    Zakhnini, Abdelhamid; Kulenkampff, Johannes; Sauerzapf, Sophie; Pietrzyk, Uwe; Lippmann-Pipke, Johanna

    2013-08-01

    Understanding conservative fluid flow and reactive tracer transport in soils and rock formations requires quantitative transport visualization methods in 3D+t. After a decade of research and development we established the GeoPET as a non-destructive method with unrivalled sensitivity and selectivity, with due spatial and temporal resolution by applying Positron Emission Tomography (PET), a nuclear medicine imaging method, to dense rock material. Requirements for reaching the physical limit of image resolution of nearly 1 mm are (a) a high-resolution PET-camera, like our ClearPET scanner (Raytest), and (b) appropriate correction methods for scatter and attenuation of 511 keV—photons in the dense geological material. The latter are by far more significant in dense geological material than in human and small animal body tissue (water). Here we present data from Monte Carlo simulations (MCS) reflecting selected GeoPET experiments. The MCS consider all involved nuclear physical processes of the measurement with the ClearPET-system and allow us to quantify the sensitivity of the method and the scatter fractions in geological media as function of material (quartz, Opalinus clay and anhydrite compared to water), PET isotope (18F, 58Co and 124I), and geometric system parameters. The synthetic data sets obtained by MCS are the basis for detailed performance assessment studies allowing for image quality improvements. A scatter correction method is applied exemplarily by subtracting projections of simulated scattered coincidences from experimental data sets prior to image reconstruction with an iterative reconstruction process.

  2. Fabrication of the fuel elements cladding for utilization in the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Schaeffer, L.; Sefidvash, F.

    1986-01-01

    A method for the fabrication of cladding of the spherical fuel elements for the utilization in the fluidized bed nuclear reactor is presented. Some prelimminary experiments were performed to adopt a method which adapt itself to mass production with the desired high quality. Still methods for cladding fabrication are under study. (Author) [pt

  3. In vivo cell tracking imaging of hexadecyl-4-[{sup 123,} {sup 124}I]iodobenzoate labeled adipose derived stem cells (ADSCs) in rat heart

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Hwan; Lee, Yong Jin; Lee, Kyo Chul [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of)

    2011-10-15

    Monitoring of transplanted stem cells for cardiac repair is important part in regenerative medicine. Direct cell labeling techniques using [{sup 18}F]FDG, [{sup 64}Cu]PTSM and [{sup 99m}Tc]-HMPAO have been developed for in vivo imaging. Especially, {sup 18}F-labeled derivates have been widely used for direct labeling agent. But the {sup 18}F has short half life (T{sub 1/2}={approx}2 h), thus this imaging agent has limitation of in vivo imaging. We used {sup 123}I or {sup 124}I which has relative long half life, to track the transplanted stem cells for a long-term imaging. This study is aimed to track the transplanted adipose derived stem cells (ADSCs) in rat heart using hexadecyl-4-[{sup 123,} {sup 124}I]iodobenzoate ([{sup 123,} {sup 124}I]HIB) mediated direct labeling method in vivo

  4. The market for nuclear equipment - engineering and fabrication

    International Nuclear Information System (INIS)

    Tait, D.R.

    1977-01-01

    The role of electronic equipment in a CANDU power station is explained. Costs of installations and outages are outlined. The nuclear market for electronic equipment utilizes many components common to non-nuclear applications. (E.C.B.)

  5. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

  6. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

  7. Imaging with 124I in differentiated thyroid carcinoma: is PET/MRI superior to PET/CT?

    International Nuclear Information System (INIS)

    Binse, I.; Poeppel, T.D.; Ruhlmann, M.; Gomez, B.; Bockisch, A.; Rosenbaum-Krumme, S.J.; Umutlu, L.

    2016-01-01

    The aim of this study was to compare integrated PET/CT and PET/MRI for their usefulness in detecting and categorizing cervical iodine-positive lesions in patients with differentiated thyroid cancer using 124 I as tracer. The study group comprised 65 patients at high risk of iodine-positive metastasis who underwent PET/CT (low-dose CT scan, PET acquisition time 2 min; PET/CT 2 ) followed by PET/MRI of the neck 24 h after 124 I administration. PET images from both modalities were analysed for the numbers of tracer-positive lesions. Two different acquisition times were used for the comparisons, one matching the PET/CT 2 acquisition time (2 min, PET/MRI 2 ) and the other covering the whole MRI scan time (30 min, PET/MRI 30 ). Iodine-positive lesions were categorized as metastasis, thyroid remnant or inconclusive according to their location on the PET/CT images. Morphological information provided by MRI was considered for evaluation of lesions on PET/MRI and for volume information. PET/MRI 2 detected significantly more iodine-positive metastases and thyroid remnants than PET/CT 2 (72 vs. 60, p = 0.002, and 100 vs. 80, p = 0.001, respectively), but the numbers of patients with at least one tumour lesion identified were not significantly different (21/65 vs. 17/65 patients). PET/MRI 30 tended to detect more PET-positive metastases than PET/MRI 2 (88 vs. 72), but the difference was not significant (p = 0.07). Of 21 lesions classified as inconclusive on PET/CT, 5 were assigned to metastasis or thyroid remnant when evaluated by PET/MRI. Volume information was available in 34 % of iodine-positive metastases and 2 % of thyroid remnants on PET/MRI. PET/MRI of the neck was found to be superior to PET/CT in detecting iodine-positive lesions. This was attributed to the higher sensitivity of the PET component, Although helpful in some cases, we found no substantial advantage of PET/MRI over PET/CT in categorizing iodine-positive lesions as either metastasis or thyroid remnant

  8. Imaging with {sup 124}I in differentiated thyroid carcinoma: is PET/MRI superior to PET/CT?

    Energy Technology Data Exchange (ETDEWEB)

    Binse, I.; Poeppel, T.D.; Ruhlmann, M.; Gomez, B.; Bockisch, A.; Rosenbaum-Krumme, S.J. [University of Duisburg-Essen, Medical Faculty, Department of Nuclear Medicine, Essen (Germany); Umutlu, L. [University of Duisburg-Essen, Medical Faculty, Department of Radiology, Essen (Germany)

    2016-06-15

    The aim of this study was to compare integrated PET/CT and PET/MRI for their usefulness in detecting and categorizing cervical iodine-positive lesions in patients with differentiated thyroid cancer using {sup 124}I as tracer. The study group comprised 65 patients at high risk of iodine-positive metastasis who underwent PET/CT (low-dose CT scan, PET acquisition time 2 min; PET/CT{sub 2}) followed by PET/MRI of the neck 24 h after {sup 124}I administration. PET images from both modalities were analysed for the numbers of tracer-positive lesions. Two different acquisition times were used for the comparisons, one matching the PET/CT{sub 2} acquisition time (2 min, PET/MRI{sub 2}) and the other covering the whole MRI scan time (30 min, PET/MRI{sub 30}). Iodine-positive lesions were categorized as metastasis, thyroid remnant or inconclusive according to their location on the PET/CT images. Morphological information provided by MRI was considered for evaluation of lesions on PET/MRI and for volume information. PET/MRI{sub 2} detected significantly more iodine-positive metastases and thyroid remnants than PET/CT{sub 2} (72 vs. 60, p = 0.002, and 100 vs. 80, p = 0.001, respectively), but the numbers of patients with at least one tumour lesion identified were not significantly different (21/65 vs. 17/65 patients). PET/MRI{sub 30} tended to detect more PET-positive metastases than PET/MRI{sub 2} (88 vs. 72), but the difference was not significant (p = 0.07). Of 21 lesions classified as inconclusive on PET/CT, 5 were assigned to metastasis or thyroid remnant when evaluated by PET/MRI. Volume information was available in 34 % of iodine-positive metastases and 2 % of thyroid remnants on PET/MRI. PET/MRI of the neck was found to be superior to PET/CT in detecting iodine-positive lesions. This was attributed to the higher sensitivity of the PET component, Although helpful in some cases, we found no substantial advantage of PET/MRI over PET/CT in categorizing iodine

  9. Hybrid pellets: an improved concept for fabrication of nuclear fuel

    International Nuclear Information System (INIS)

    Matthews, R.B.; Hart, P.E.

    1979-09-01

    The feasibility of fabricating fuel pellets using gel-derived microspheres as press feed was evaluated. By using gel-derived microspheres as press feed, the potential exists for eliminating dusty operations like milling, slugging, and granulation, from the pelleting process. The free-flowing character of the spheres should also result in limited dust generation during powder transport and pressing operations. The results of this study clearly demonstrate that fuel pellets can be successfully fabricated on a laboratory scale using UO 2 gel microspheres as press feed. Under moderate sintering conditions, 1,500 0 C for 4 h in Ar-4% H 2 , UO 2 pellets with densities up to 96% TD were fabricated. A range of pellet microstructures and densities were achieved depending on sphere forming and calcining conditions. Based on these results, a set of necessary sphere properties are suggested: O/U less than 2.20, crystallite size less than 500 A, specific surface area greater than 8 m 2 /g, and sphere size 200 and 400 μm. Preliminary attempts to fabricate ThO 2 and ThO 2 -UO 2 pellets using microspheres were unsuccessful because the requisite sphere properties were not achieved. Areas requiring additional development include: demonstration of the process on scaled-up equipment suitable for use in a remote fuel fabrication facility and evaluation of the irradiation performance of pellet fuels from gel-spheres

  10. Environmental aspects based on operation performance of nuclear fuel fabrication facilities

    International Nuclear Information System (INIS)

    2001-07-01

    This publication was prepared within the framework of the IAEA Project entitled Development and Upgrading of Guidelines, Databases and Tools for Integrating Comparative Assessment into Energy System Analysis and Policy Making, which included the collection, review and input of data into a database on health and environmental impacts related to operation of nuclear fuel cycle facilities. The objectives of the report included assembling environmental data on operational performance of nuclear fabrication facilities in each country; compiling and arranging the data in a database, which will be easily available to experts and the public; and presenting data that may be of value for future environmental assessment of nuclear fabrication facilities

  11. New fabrication techniques for the nuclear fuels of tomorrow

    International Nuclear Information System (INIS)

    Babelot, J.F.; Bokelund, H.; Gerontopoulos, P.; Gueugnon, J.F.; Richter, K.

    1995-01-01

    The shift of the emphasis of the work at the Institute for Transuranium Elements (ITU) from the development of fuels based on uranium and plutonium to safety aspects concerning the use of plutonium and other of actinides, necessitates the production of targets containing appreciable amounts of minor actinides for irradiation experiments. The handling of minor actinides requires additional protective measures, combined with improved fuel fabrication techniques. The boundary conditions for a suitable process are flexibility, adaptability to remote control, and minimization of dust formation. A method based on the sol-gel fabrication technique meets these criteria, and was selected for the present developments at ITU. (author)

  12. Basic tendencies of restructured UO2 nuclear fuels fabrication industry for water-moderated reactors

    International Nuclear Information System (INIS)

    Makhova, V.A.; Bokshitskij, V.I.; Blinova, I.V.

    2002-01-01

    Processes of reformation and consolidation of firms and frontier nuclear fuels fabrication industry associated with processes of globalization and deregulation of electric power market are analyzed. Current state of nuclear fuel market and basic factors influenced on the market are presented. The role of nuclear fuel in increasing competition of NPP and fundamental directions of innovation action on the creation of perspective kinds of fuel were considered [ru

  13. Effect of the positron range of 18F, 68Ga and 124I on PET/CT in lung-equivalent materials.

    Science.gov (United States)

    Kemerink, Gerrit J; Visser, Mariëlle G W; Franssen, Renee; Beijer, Emiel; Zamburlini, Mariangela; Halders, Servé G E A; Brans, Boudewijn; Mottaghy, Felix M; Teule, Gerrit J J

    2011-05-01

    The aim of this study was to investigate the effect of positron range on visualization and quantification in (18)F, (68)Ga and (124)I positron emission tomography (PET)/CT of lung-like tissue. Different sources were measured in air, in lung-equivalent foams and in water, using a clinical PET/CT and a microPET system. Intensity profiles and curves with the cumulative number of annihilations were derived and numerically characterized. (68)Ga and (124)I gave similar results. Their intensity profiles in lung-like foam had a peak similar to that for (18)F, and tails of very low intensity, but extending over distances of centimetres and containing a large fraction of all annihilations. For 90% recovery, volumes of interest with diameters up to 50 mm were required, and recovery within the 10% intensity isocontour was as low as 30%. In contrast, tailing was minor for (18)F. Lung lesions containing (18)F, (68)Ga or (124)I will be visualized similarly, and at least as sharp as in soft tissue. Nevertheless, for quantification of (68)Ga and (124)I large volumes of interest are needed for complete activity recovery. For clinical studies containing noise and background, new quantification approaches may have to be developed.

  14. Method for the fabrication of nuclear fuel bodies

    International Nuclear Information System (INIS)

    Davis, D.E.; Leary, D.F.

    1976-01-01

    According to the method, graphite particles are treated with a liquid impregnating agent containing heat-hardenable resin components; the resulting particles are mixed with nuclear fuel particles, and a nuclear fuel body is formed by binding the mixture of particles into a cohesive mass by means of a carbon-contained binder. The claim concerns the details of the process. (UA) [de

  15. Development of fabrication technology for ceramic nuclear fuel

    International Nuclear Information System (INIS)

    Lee, Young Woo; Sohn, D. S.; Na, S. H.

    2003-05-01

    The purpose of the study is to develop the fabrication technology of MOX fuel. The researches carried out during the last stage(1997. 4.∼2003. 3.) mainly consisted of ; study of MOX pellet fabrication technology for application and development of characterization technology for the aim of confirming the development of powder treatment technology and sintering technology and of the optimization of the above technologies and fabrication of Pu-MOX pellet specimens through an international joint collaboration between KAERI and PSI based on the fundamental technologies developed in KAERI. Based on the studies carried out and the results obtained during the last stage, more extensive studies for the process technologies of the unit processes were performed, in this year, for the purpose of development of indigenous overall MOX pellet fabrication process technology, relating process parameters among the unit processes and integrating these unit process technologies. Furthermore, for the preparation of transfer of relevant technologies to the industries, a feasibility study was performed on the commercialization of the technology developed in KAERI with the relevant industry in close collaboration

  16. Negative predictive value of 124I-PET/CT imaging in patients affected by metastatic thyroid cancer and treated with 131I

    International Nuclear Information System (INIS)

    Pettinato, C.; Civollani, S.; Nanni, C.; Celli, M.; Allegri, V.; Zagni, P.; Fanti, S.; Monari, F.; Cima, S.; Mazzarotto, R.; Spezi, E.

    2015-01-01

    Full text of publication follows. Aim: patients affected by metastatic Differentiated Thyroid Cancer (mDTC) are treated with 131 I even in presence of negative diagnostic 131 I whole body (WB) scan. Actually, very often, these patients present positive post therapy 131 I whole body scan, showing iodine avid metastases that were not seen with the diagnostic imaging. The aim of this work was the evaluation of the feasibility to use 124 I PET/CT images to predict patients who will not benefit from the iodine therapy, because of the absence of avidity, avoiding useless treatments. Material and methods: 25 patients affected by mDTC were enrolled in the study approved by the ethical Committee of our Institution, with the aim to evaluate the usefulness of 124 I PET/CT sequential scans to predict absorbed doses to metastatic thyroid cancer patients undergoing 131 I therapy. Patients (pts) were divided into 4 groups, based on their histology: group A, 4 pts with follicular cancer; group B, 13 pts with papillary cancer; group C, 2 pts with papillary tall cells cancer; group D, 6 patients with papillary cancer with follicular variant. Patients showing negative 124 I-PET/CT were treated with a reduced dose of 131 I (3700 MBq) and post treatment WB scans were acquired 96 hours after the therapeutic administration. Results: 12 patients showed at least one metastatic lesion at 124 I PET/CT imaging, and most of the lesions were visible at the 24 hours scan (4 pts group A, 3 pts group B, 5 pts group D). The remaining 13 patients did not show any uptake of all known metastatic lesions at each PET/CT time points (10 pts group B, 2 pts group C, 1 pt group D). Negative PET/CT findings were confirmed by post therapy WB scan. Discussion and Conclusion: 124 I-PET/CT scan is a useful diagnostic tool to discriminate patients with iodine avid metastases. Actually, when they are present, the superiority of PET/CT resolution and sensitivity, compared to standard 131 I planar imaging, allow the

  17. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1978-01-01

    The Regulation is revised on the basis of ''The law for the regulations of nuclear source materials, nuclear fuel materials and reactors'' and the ''Provisions concerning the enterprises processing nuclear fuel materials'' in the Enforcement Ordinance for the Law, to enforce such provisions. This is the complete revision of the regulation of the same name in 1957. Terms are explained, such as exposure radiation dose, cumulative dose, control area, surrounding inspection area, persons engaged in works, radioactive wastes, area for incoming and outgoing of materials, fluctuation of stocks, batch, real stocks, effective value and main measuring points. For the applications for the permission of the enterprises processing nuclear fuel materials, the location of an enterprise, the construction of buildings and the construction of and the equipments for facilities of chemical processing, forming, coating, assembling, storage of nuclear fuel materials, disposal of radioactive wastes and radiation control must be written. Records shall be made and maintained for the periods specified on the inspection of processing facilities, nuclear fuel materials, radiation control, operation, maintainance, accidents of processing facilities and weather. Limit to entrance into the control area, measures for exposure radiation dose, patrol and inspection, operation of processing facilities, transport of materials, disposal of radioactive wastes, safety regulations are provided for. Reports to be filed by the persons engaging in the enterprises processing nuclear fuel materials are prescribed. (Okada, K.)

  18. Production of 68Ge, 64Cu, 86Y, 89Zr, 73Se, 77Br and 124I positron emitting radionuclides through future laser-accelerated proton beams at ELI-Beamlines for innovative PET diagnostics

    Directory of Open Access Journals (Sweden)

    Antonio Italiano

    2016-05-01

    Full Text Available The development of innovative production pathways for high-Z positron emitters is of great interest to enlarge the applicability of PET diagnostics, especially in view of the continuous development of new radiopharmaceuticals. We evaluated the theoretical yields of 64Cu, 86Y, 89Zr, 73Se, 77Br and 124I PET isotopes, plus the 68Ge isotope, parent of the 68Ga positron emitter, in the hypothesis of production through laser-accelerated proton sources expected at the ELI-Beamlines facility. By means of the TALYS software we simulated the nuclear reactions leading to the above radionuclides, hypothesizing three possible scenarios of broad proton spectra, with maximum energies of about 9, 40 and 100 MeV. The production yields of the studied radionuclides, within the expected fluences, appear to be suitable for pre-clinical applications.

  19. Process and device for fabricating nuclear fuel assembly grids

    International Nuclear Information System (INIS)

    Thiebaut, B.; Duthoo, D.; Germanaz, J.J.; Angilbert, B.

    1991-01-01

    The method for fabricating PWR fuel assembly grids consists to place the grid of which the constituent parts are held firmly in place within a frame into a sealed chamber full of inert gas. This chamber can rotate about an axis. The welding on one face at a time is carried out with a laser beam orthogonal to the axis orientation of the device. The laser source is outside of the chamber and the beam penetrates via a transparent view port

  20. Aspects for selection of materials and fabrication processes for nuclear component manufacturing

    International Nuclear Information System (INIS)

    Pernstich, K.

    1980-01-01

    For components of the Nuclear steam supply System of Light Water Reactors an extremely high safety standard is required. These requirements only can be met by adequate selection of materials and fabrication processes and their proper application in combination with strict quality assurance and control measurements. A general overview of the basic aspects to be considered in this connection is presented together with an indication of the present state of art for the main materials and fabrication processes. (author) [pt

  1. The role of a multinational nuclear fuel fabrication supplier

    International Nuclear Information System (INIS)

    Beard, S.J.

    1987-01-01

    The author argues that international markets and multinational suppliers provide large benefits to utilities. It represents a long term commitment to the nuclear business that these companies will be able to supply nuclear technology on the long haul. The technology that is available around the world becomes available to everyone through the international markets and multinational suppliers. The increased experience base is seen as valuable in that errors that have been made or have not been made yet can be avoided through the transfer or experience. The security of supply is discussed as important to any utility that is operating a reactor

  2. Pneumatic conveying of sensitive compounds during nuclear fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Sielck, Franz-Christian; Braehler, Georg [NUKEM Technologies GmbH (Germany)

    2009-07-01

    Any transport of nuclear material is associated with the risk of contamination after release into working areas or environment. stationary installed safe geometry vessels with pneumatic transfer between them offer unique safety features and reduce operating costs. The article describes the case of HTR fuel spheres, where a specially designed conveying system has been developed and the prototype conveyor has been tested.

  3. Pneumatic conveying of sensitive compounds during nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Sielck, Franz-Christian; Braehler, Georg

    2009-01-01

    Any transport of nuclear material is associated with the risk of contamination after release into working areas or environment. stationary installed safe geometry vessels with pneumatic transfer between them offer unique safety features and reduce operating costs. The article describes the case of HTR fuel spheres, where a specially designed conveying system has been developed and the prototype conveyor has been tested.

  4. Evaluation of radiation protection conditions in nuclear gauges fabrication

    International Nuclear Information System (INIS)

    Sekiguchi, Marcelo Ferreira; Borges, Jose Carlos

    1999-01-01

    The objective of this work was to evaluate the radioprotection conditions in the work place, of a industry that produces nuclear gauges. The survey was divided, basically, in two parts; first took place a physical monitoring area, individual and contamination and biological, through the analysis of excretes and cytogenetic dosimetry. (author)

  5. The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

    International Nuclear Information System (INIS)

    Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I

    2009-01-01

    The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed

  6. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    International Nuclear Information System (INIS)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-01-01

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels

  7. Fabrication and Characterization of Surrogate Glasses Aimed to Validate Nuclear Forensic Techniques

    Science.gov (United States)

    2017-12-01

    the glass formed during a nuclear event, trinitite [14]. The SiO2 composition is generally greater than 50% for trinitite and can vary appreciably...CHARACTERIZATION OF SURROGATE GLASSES AIMED TO VALIDATE NUCLEAR FORENSIC TECHNIQUES by Ken G. Foos December 2017 Thesis Advisor: Claudia...December 2017 3. REPORT TYPE AND DATES COVERED Master’s thesis 4. TITLE AND SUBTITLE FABRICATION AND CHARACTERIZATION OF SURROGATE GLASSES AIMED TO

  8. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  9. KHIC's experience in the design and fabrication of nuclear components

    International Nuclear Information System (INIS)

    Suh, S.-C.

    1992-01-01

    Since 1980, Korea Heavy Industries ampersand Construction Company, Ltd. (KHIC) has specialized in the design and equipment supply for nuclear power facilities in Korea. In April 1987, KHIC became the prime contractor for the construction of Yonggwang 3 ampersand 4 (YGN 3 ampersand 4) nuclear power project. Accordingly, KHIC's technological self-reliance capability for the manufacturing processes of the primary system equipment and components has increased from 18% during the initial stage of Yonggwang 1 ampersand 2 (YGN 1 ampersand 2) project to 63% for YGN 3 ampersand 4 project. Self-reliance capability for the secondary system equipment and components has increased from 28% to 84% during the same period of time as well. The ultimate goal is to achieve complete and total assurance that our products are of the finest quality in the nuclear industry in the world market. Henceforth, we will be able to guarantee complete customer satisfaction and reliability of our products with safety assurance and leading edge technology

  10. Material engineering to fabricate rare earth erbium thin films for exploring nuclear energy sources

    Science.gov (United States)

    Banerjee, A.; Abhilash, S. R.; Umapathy, G. R.; Kabiraj, D.; Ojha, S.; Mandal, S.

    2018-04-01

    High vacuum evaporation and cold-rolling techniques to fabricate thin films of the rare earth lanthanide-erbium have been discussed in this communication. Cold rolling has been used for the first time to successfully fabricate films of enriched and highly expensive erbium metal with areal density in the range of 0.5-1.0 mg/cm2. The fabricated films were used as target materials in an advanced nuclear physics experiment. The experiment was designed to investigate isomeric states in the heavy nuclei mass region for exploring physics related to nuclear energy sources. The films fabricated using different techniques varied in thickness as well as purity. Methods to fabricate films with thickness of the order of 0.9 mg/cm2 were different than those of 0.4 mg/cm2 areal density. All the thin films were characterized using multiple advanced techniques to accurately ascertain levels of contamination as well as to determine their exact surface density. Detailed fabrication methods as well as characterization techniques have been discussed.

  11. Summary of current radiation dose estimates to humans from 123I, 124I, 126I, 130I, and 131I as sodium rose bengal

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Estimated absorbed doses to human gall bladder, gastrointestinal tract, liver, ovaries, bone marrow, and testes from 123 I, 124 I, 126 I, 130 I, and 131 I after a single intravenous administration as sodium rose bengal are summarized. The greatest uncertainty in these dose estimates is due to the variability in time for the movement of radioiodine through the biliary tract, gall bladder, and gastrointestinal tract

  12. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1987-01-01

    Regulations specified here cover application for such matters as permission for an undertaking of processing, alteration (of location, structure, arrangements, processing method, etc.), approval of design and construction plan, approval of alteration (of design and construction plan of processing facilities), and inspection of the facilities. The regulations also cover execution of facilities inspection, certificate of facilities inspection, processing facilities subject to welding inspection, application for welding inspection, execution of welding inspection, facilities not subject to welding inspection, approval of welding method, welding inspection for imported equipment, certificate of welding inspection, application for approval of joint management, notice of alteration, etc., cancellation of permission, record keeping, restriction on access to areas under management measures concerning exposure to radioactive rays, patrol and checking in processing facilities, operation of processing equipment, transportation within plant or operation premises, storage, waste disposal within plant or operation premises, safety rules, public notification concerning examination and successful applicants, procedure for application for examination, reissue of certificate for nuclear fuel handling expert, return of certificate for nuclear fuel handling expert, submission of report, measures for emergency, notice of abolition of business, measures concerning cancellation of permission, identification card, etc. (Nogami, K.)

  13. Advanced methods of quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Onoufriev, Vladimir

    2004-01-01

    Under pressure of current economic and electricity market situation utilities implement more demanding fuel utilization schemes including higher burn ups and thermal rates, longer fuel cycles and usage of Mo fuel. Therefore, fuel vendors have recently initiated new R and D programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel. In the beginning of commercial fuel fabrication, emphasis was given to advancements in Quality Control/Quality Assurance related mainly to product itself. During recent years, emphasis was transferred to improvements in process control and to implementation of overall Total Quality Management (TQM) programmes. In the area of fuel quality control, statistical control methods are now widely implemented replacing 100% inspection. This evolution, some practical examples and IAEA activities are described in the paper. The paper presents major findings of the latest IAEA Technical Meetings (TMs) and training courses in the area with emphasis on information received at the TM and training course held in 1999 and other latest publications to provide an overview of new developments in process/quality control, their implementation and results obtained including new approaches to QC

  14. Design, fabrication and erection of steel structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety Standard for Civil Engineering Structures Important to Safety of Nuclear Facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design, fabrication and erection of steel structures important to safety

  15. Study of developing nuclear fabrication facility's integrated emergency response manual

    International Nuclear Information System (INIS)

    Kim, Taeh Yeong; Cho, Nam Chan; Han, Seung Hoon; Moon, Jong Han; Lee, Jin Hang; Min, Guem Young; Han, Ji Ah

    2016-01-01

    Public begin to pay attention to emergency management. Thus, public's consensus on having high level of emergency management system up to advanced country's is reached. In this social atmosphere, manual is considered as key factor to prevent accident or secure business continuity. Therefore, we first define possible crisis at KEPCO Nuclear Fuel (hereinafter KNF) and also make a 'Reaction List' for each crisis situation at the view of information-design. To achieve it, we analyze several country's crisis response manual and then derive component, indicate duties and roles at the information-design point of view. From this, we suggested guideline to make 'Integrated emergency response manual(IERM)'. The manual we used before have following few problems; difficult to applicate at the site, difficult to deliver information. To complement these problems, we searched manual elements from the view of information-design. As a result, we develop administrative manual. Although, this manual could be thought as fragmentary manual because it confined specific several agency/organization and disaster type

  16. Fabrication of high performance components for Indian nuclear reactors

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2011-01-01

    Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

  17. 25 years of NDE in fabrication of zirconium alloy mill products and nuclear fuel in the Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Mistry, R.K.; Laxminarayana, B.; Srivastava, R.K.

    1996-01-01

    Failure of nuclear fuel is highly undesirable from both economic and operational aspects. Hence all the components require rigorous QC and inspection checks. NDT plays a major role in assuring the quality of the products both at final and intermediate stages. This paper gives an overall review of NDT methods employed in achieving the integrity of nuclear products. The NDE procedures followed in NFC are visual inspection, radiography, penetrant testing, eddy current testing, ultrasonic testing and helium leak testing. NFC's quality assurance programme is organised to achieve the desired objectives by carrying out in process and final inspection at all critical steps of fabrication. (author)

  18. Fabrication and closure development of nuclear waste disposal containers for the Yucca Mountain Project: Status report

    International Nuclear Information System (INIS)

    Domian, H.A.; Robitz, E.S.; Conrardy, C.C.; LaCount, D.F.; McAninch, M.D.; Fish, R.L.; Russell, E.W.

    1991-09-01

    In GFY 89, a project was underway to determine and demonstrate a suitable method for fabricating thin-walled monolithic waste containers for service within the potential repository at Yucca Mountain. A concurrent project was underway to determine and demonstrate a suitable closure process for these containers after they have been filled with high-level nuclear waste. Phase 1 for both the fabrication and closure projects was a screening phase in which candidate processes were selected for further laboratory testing in Phase 2. This report describes the final results of the Phase 1 efforts. It also describes the preliminary results of Phase 2 efforts

  19. Scanning tunnelling microscope fabrication of phosphorus array in silicon for a nuclear spin quantum computer

    International Nuclear Information System (INIS)

    O'Brien, J.L.; Schofield, S.R.; Simmons, M.Y.; Clark, R.G.; Dzurak, A.S.; Prawer, S.; Adrienko, I.; Cimino, A.

    2000-01-01

    Full text: In the vigorous worldwide effort to experimentally build a quantum computer, recent intense interest has focussed on solid state approaches for their promise of scalability. Particular attention has been given to silicon-based proposals that can readily be integrated into conventional computing technology. For example the Kane design uses the well isolated nuclear spin of phosphorous donor nuclei (I=1/2) as the qubits embedded in isotopically pure 28 Si (I=0). We demonstrate the ability to fabricate a precise array of P atoms on a clean Si surface with atomic-scale resolution compatible with the fabrication of the Kane quantum computer

  20. Development of a facility for fabricating nuclear waste canisters from radioactively contaminated steel

    International Nuclear Information System (INIS)

    Logan, J.A.; Larsen, M.M.

    1986-01-01

    This paper describes design of a facility and processes capable of using radioactively contaminated waste steel as the principal raw material for fabricating stainless steel canisters to be used for disposal of nuclear high-level waste. By such action, expenditure (i.e., permanent loss to society) of thousands of tons of uncontaminated chromium and nickel to fabricate such canisters can be avoided. Moreover, the cost and risks involved in disposing of large accumulations of radioactively contaminated steel as low-level radioactive waste (LLRW), that would otherwise be necessary, can also be avoided. The canister fabrication processes (involving centrifugal casting) described herein have been tested and proven for this application. The performance characteristics of stainless steel canisters so fabricated have been tested and agreed to by the organizations that have been involved in this development work (Battelle Memorial Institute, DuPont, EGandG and the Savannah River Laboratory) as equivalent to the performance characteristics of canisters fabricated of uncontaminated wrought stainless steel. It is estimated that the production cost for fabricating canisters by the methods described will not differ greatly from the production cost using uncontaminated wrought steel, and the other costs avoided by not having to dispose of the contaminated steel as LLRW could cause this method to produce the lowest ultimate overall costs

  1. Control and balance of nuclear matters used for core fabrication of Super Phenix

    International Nuclear Information System (INIS)

    Beche, M.; Guillet, H.; Heyraud, H.; Levrard, J.; Pajot, J.

    1987-05-01

    The fabrication of the core of the fast breeder reactor set up at Creys Malville ended in March 1984. It started in 1978 and it required, for the fabrication of the 410 assemblies, the utilization of 7438 kg of plutonium. To satisfy national and international regulations, DPFER/SFER has used a methodology to follow and to control the movements of the nuclear materials. These controls are achieved by physical methods, chemical methods and empiric methods. Euratom has conducted a succession of inspections during the 5.5 years of that campaign. The inventory difference, in the fabrication of that core, represents about 0.1% of the total mass of the plutonium handled [fr

  2. Application of vacuum technology during nuclear fuel fabrication, inspection and characterization

    International Nuclear Information System (INIS)

    Majumdar, S.

    2003-01-01

    Full text: Vacuum technology plays very important role during various stages of fabrication, inspection and characterization of U, Pu based nuclear fuels. Controlled vacuum is needed for melting and casting of U, Pu based alloys, picture framing of the fuel meat for plate type fuel fabrication, carbothermic reduction for synthesis of (U-Pu) mixed carbide powder, dewaxing of green ceramic fuel pellets, degassing of sintered pellets and encapsulation of fuel pellets inside clad tube. Application of vacuum technology is also important during inspection and characterization of fuel materials and fuel pins by way of XRF and XRD analysis, Mass spectrometer Helium leak detection etc. A novel method of low temperature sintering of UO 2 developed at BARC using controlled vacuum as sintering atmosphere has undergone successful irradiation testing in Cirus. The paper will describe various fuel fabrication flow sheets highlighting the stages where vacuum applications are needed

  3. Tuning up and fabrication of U3Si2 nuclear material

    International Nuclear Information System (INIS)

    Pasqualini, Enrique E.; Echenique, Patricia N.; Rossi, Gustavo S.; Canil, Eduardo E.; Esteban, Adolfo; Lopez, Marisol; Adelfang, Pablo

    2000-01-01

    This work describes the tuning up and fabrication of uranium-silicide powder for its utilization as nuclear fuel in material testing reactors taking in account NUREG-1313 recommendations, the experience of several suppliers and the one acquired in this work.Several alloy compositions were melted with natural uranium at temperatures of about 1800 degree C for adjusting composition and ingot homogeneity. Alumina, magnesia and zirconia-5% stabilized yttria crucibles were tested to evaluate the degree of contamination introduced by chemical attack of molten uranium and silicon. The fabrication procedure of 20% enriched uranium-silicide powder was established for building up the P-06 fuel element that actually is being irradiated at the RA-3 reactor facility. The selected procedures of fabrication and the critical analysis for the interpretation of several specifications are discussed. Results are shown of the obtained ingots and powder produced with the enriched uranium-silicide. (author)

  4. Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O' Brien

    2013-02-01

    Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

  5. Fabrication and closure development of nuclear waste containers for storage at the Yucca Mountain, Nevada repository

    International Nuclear Information System (INIS)

    Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

    1989-04-01

    US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 1 fig., 7 tabs

  6. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

  7. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P. O. 1236909. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design.

  8. Evaluation of methods for seismic analysis of nuclear fuel reprocessing and fabrication facilities

    International Nuclear Information System (INIS)

    Arthur, D.F.; Dong, R.G.; Murray, R.C.; Nelson, T.A.; Smith, P.D.; Wight, L.H.

    1978-01-01

    Methods of seismic analysis for critical structures and equipment in nuclear fuel reprocessing plants (NFRPs) and mixed oxide fuel fabrication plants (MOFFPs) are evaluated. The purpose of this series of reports is to provide the NRC with a technical basis for assessing seismic analysis methods and for writing regulatory guides in which methods ensuring the safe design of nuclear fuel cycle facilities are recommended. The present report evaluates methods of analyzing buried pipes and wells, sloshing effects in large pools, earth dams, multiply supported equipment, pile foundations, and soil-structure interactions

  9. Establishing QC/QA system in the fabrication of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Suh, K.S.; Choi, S.K.; Park, H.G.; Park, T.G.; Chung, J.S.

    1980-01-01

    Quality control instruction manuals and inspection methods for UO 2 powder and zircaloy materials as the material control, and for UO 2 pellets and nuclear fuel rods as the process control were established. And for the establishment of Q.A programme, the technical specifications of the purchased materials, the control regulation of the measuring and testing equipments, and traceability chart as a part of document control have also been provided and practically applied to the fuel fabrication process

  10. Tumor immunolocalization using 124I-iodine-labeled JAA-F11 antibody to Thomsen-Friedenreich alpha-linked antigen

    International Nuclear Information System (INIS)

    Chaturvedi, Richa; Heimburg, Jamie; Yan, Jun; Koury, Stephen; Sajjad, Munawwar; Abdel-Nabi, Hani H.; Rittenhouse-Olson, Kate

    2008-01-01

    Clinical immunolocalization has been attempted by others with an anti-Thomsen-Friedenreich antigen (TF-Ag) mAb that bound both alpha- and beta-linked TF-Ag. In this report, 124 I-labeled mAb JAA-F11 specific for alpha-linked TF-Ag showed higher tumor specificity in in vivo micro-positron emission tomography (micro-PET) of the mouse mammary adenocarcinoma line, 4T1, showing no preferential uptake by the kidney. Labeled product remained localized in the tumor for at least 20 days. Glycan array analysis showed structural specificity of the antibody

  11. Direct characterization of cotton fabrics treated with di-epoxide by nuclear magnetic resonance.

    Science.gov (United States)

    Xiao, Min; Chéry, Joronia; Keresztes, Ivan; Zax, David B; Frey, Margaret W

    2017-10-15

    A non-acid-based, di-functional epoxide, neopentyl glycol diglycidyl ether (NPGDGE), was used to modify cotton fabrics. Direct characterization of the modified cotton was conducted by Nuclear Magnetic Resonance (NMR) without grinding the fabric into a fine powder. NaOH and MgBr 2 were compared in catalyzing the reaction between the epoxide groups of NPGDGE and the hydroxyl groups of cellulose. Possible reaction routes were discussed. Scanning electron microscopy (SEM) images showed that while the MgBr 2 -catalyzed reaction resulted in self-polymerization of NPGDGE, the NaOH-catalyzed reaction did not. Fourier transform infrared spectroscopy (FTIR) showed that at high NaOH concentration cellulose restructures from allomorph I to II. NMR studies verified the incorporation of NPGDGE into cotton fabrics with a clear NMR signal, and confirmed that at higher NaOH concentration the efficiency of grafting of NPGDGE was increased. This demonstrates that use of solid state NMR directly on woven fabric samples can simultaneously characterize chemical modification and crystalline polymorph of cotton. No loss of tensile strength was observed for cotton fabrics modified with NPGDGE. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Fabrication and characterization of CeO{sub 2} pellets for simulation of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    García-Ostos, C.; Rodríguez-Ortiz, J.A. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Arévalo, C., E-mail: carevalo@us.es [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Cobos, J. [CIEMAT, Avenida Complutense, 40, Madrid (Spain); Gotor, F.J. [Materials Science Institute of Seville (CSIC-US), Av. Américo Vespucio, 49, 41092 Seville (Spain); Torres, Y. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain)

    2016-03-15

    Highlights: • CeO{sub 2} is presented as a surrogate material for UO{sub 2} to study nuclear fuel. • Powder-metallurgy methods are applied to fabricate CeO{sub 2} pellets with controlled porosity. • An optimization of the fabrication parameters is established. • Microstructural and tribo-mechanical characterizations are performed. • Properties are compared to those of the nuclear fuel. - Abstract: Cerium Oxide, CeO{sub 2}, has been shown as a surrogate material to understand irradiated Mixed Oxide (MOX) based matrix fuel for nuclear power plants due to its similar structure, chemical and mechanical properties. In this work, CeO{sub 2} pellets with controlled porosity have been developed through conventional powder-metallurgy process. Influence of the main processing parameters (binder content, compaction pressure, sintering temperature and sintering time) on porosity and volumetric contraction values has been studied. Microstructure and physical properties of sintered compacts have also been characterized through several techniques. Mechanical properties such as dynamic Young's modulus, hardness and fracture toughness have been determined and connected to powder-metallurgy parameters. Simulation of nuclear fuel after reactor utilization with radial gradient porosity is proposed.

  13. Improved fabrication of HgI2 nuclear radiation detectors by machine-cleaving

    International Nuclear Information System (INIS)

    Levi, A.; Burger, A.; Schieber, M.; Vandenberg, L.; Yellon, W.B.; Alkire, R.W.

    1982-01-01

    The perfection of machine-cleaved sections from HgI 2 bulk crystals was examined. The perfection of the machine-cleaved sections as established by gamma diffraction rocking curves was found to be much better than the perfection of hand-cleaved sections or as grown thin platelets, reaching a perfection similar to that of the wire-sawn sections of HgI 2 . A correlation between the perfection and the thickness of the machine-cleaved section was also found, i.e., the thicker the cleaved-section the more perfect it is. The reproducibility of the fabrication was significantly improved by using machine cleaving in the process of fabrication. Large single crystals of HgI 2 weighing 20 to 200 g, can be grown from the vapor phase using the TOM Technique. In order to fabricate nuclear radiation detectors from these single crystals, thin sections of about 0.4 to 0.8 mm thickness have to be prepared. Up till now, the state-of-the-art of fabricating HgI 2 nuclear radiation detectors involved two methods to get thin sections from the large single crystals: (1) hand-cleaving using a razor-blade and (2) solution wire sawing. The chemical wire sawing method involves a loss of about 50% of the crystal volume and is usually followed by a chemical polishing process which involves a significant loss of volume of the original volume. This procedure is complicated and wasteful. The traditional fabrication method, i.e., hand-cleaving followed by rapid nonselective chemical etching, is simpler and less wasteful

  14. Thoria-based nuclear fuels thermophysical and thermodynamic properties, fabrication, reprocessing, and waste management

    CERN Document Server

    Bharadwaj, S R

    2013-01-01

    This book presents the state of the art on thermophysical and thermochemical properties, fabrication methodologies, irradiation behaviours, fuel reprocessing procedures, and aspects of waste management for oxide fuels in general and for thoria-based fuels in particular. The book covers all the essential features involved in the development of and working with nuclear technology. With the help of key databases, many of which were created by the authors, information is presented in the form of tables, figures, schematic diagrams and flow sheets, and photographs. This information will be useful for scientists and engineers working in the nuclear field, particularly for design and simulation, and for establishing the technology. One special feature is the inclusion of the latest information on thoria-based fuels, especially on the use of thorium in power generation, as it has less proliferation potential for nuclear weapons. Given its natural abundance, thorium offers a future alternative to uranium fuels in nuc...

  15. The Combination of In vivo 124I-PET and CT Small Animal Imaging for Evaluation of Thyroid Physiology and Dosimetry

    Directory of Open Access Journals (Sweden)

    Henrik H. El-Ali

    2012-06-01

    Full Text Available Objective: A thyroid rat model combining functional and anatomical information would be of great benefit for better modeling of thyroid physiology and for absorbed dose calculations. Our aim was to show that 124I-PET and CT small animal imaging are useful as a combined model for studying thyroid physiology and dose calculation. Methods: Seven rats were subjects for multiple thyroid 124I-imaging and CT-scans. S-values [mGy/MBqs] for different thyroid sizes were simulated. A phantom with spheres was designed for validation of performances of the small animal PET and CT imaging systems. Results: Small animal image-based measurements of the activity amount and the volumes of the spheres with a priori known volumes showed a good agreement with their corresponding actual volumes. The CT scans of the rats showed thyroid volumes from 34–70 mL. Conclusions: The wide span in volumes of thyroid glands indicates the importance of using an accurate volume-measuring technique such as the small animal CT. The small animal PET system was on the other hand able to accurately estimate the activity concentration in the thyroid volumes. We conclude that the combination of the PET and CT image information is essential for quantitative thyroid imaging and accurate thyroid absorbed dose calculation.

  16. Design of a quality assurance system in the nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Garcia Rojas Palacios, L.

    1992-01-01

    A)For the first time a project on nuclear fuel fabrication is going to be lead in this country. For this reason the work is oriented to establish a quality assurance system for the different stages of fuel fabrication. C) The work of this thesis was developed first by means of an analysis of quality philosophies of Deming, Ishikawa, Juran and Crosby from which several important points were stracted to be used in the designed quality system. Metrology and normalization are so important for quality control that a study of them is made considering definitions, unit systems and type of errors (for Metrology) as well as standards for quality systems, qualification, destructive and non destructive tests, shipment, packing for nuclear power plants. With the standards as a basis, the working strategy for the system was reached, as well as the design of control cards and the design of documents for inspection control, personnel and its documentation and finally the diagrams for each one of the fabrication stages

  17. CDMS Detector Fabrication Improvements and Low Energy Nuclear Recoil Measurements in Germanium

    Energy Technology Data Exchange (ETDEWEB)

    Jastram, Andrew [Texas A & M Univ., College Station, TX (United States)

    2015-12-01

    As the CDMS (Cryogenic Dark Matter Search) experiment is scaled up to tackle new dark matter parameter spaces (lower masses and cross-sections), detector production efficiency and repeatability becomes ever more important. A dedicated facility has been commissioned for SuperCDMS detector fabrication at Texas A&M University (TAMU). The fabrication process has been carefully tuned using this facility and its equipment. Production of successfully tested detectors has been demonstrated. Significant improvements in detector performance have been made using new fabrication methods, equipment, and tuning of process parameters. This work has demonstrated the capability for production of next generation CDMS SNOLAB detectors. Additionally, as the dark matter parameter space is probed further, careful calibrations of detector response to nuclear recoil interactions must be performed in order to extract useful information (in relation to dark matter particle characterzations) from experimental results. A neutron beam of tunable energy is used in conjunction with a commercial radiation detector to characterize ionization energy losses in germanium during nuclear recoil events. Data indicates agreement with values predicted by the Lindhard equation, providing a best-t k-value of 0.146.

  18. Probabilistic safety analysis for nuclear fuel cycle facilities, an exemplary application for a fuel fabrication plant

    International Nuclear Information System (INIS)

    Gmal, B.; Gaenssmantel, G.; Mayer, G.; Moser, E.F.

    2013-01-01

    In order to assess the risk of complex technical systems, the application of the Probabilistic Safety Assessment (PSA) in addition to the Deterministic Safety Analysis becomes of increasing interest. Besides nuclear installations this applies to e. g. chemical plants. A PSA is capable of expanding the basis for the risk assessment and of complementing the conventional deterministic analysis, by which means the existing safety standards of that facility can be improved if necessary. In the available paper, the differences between a PSA for a nuclear power plant and a nuclear fuel cycle facility (NFCF) are discussed in shortness and a basic concept for a PSA for a nuclear fuel cycle facility is described. Furthermore, an exemplary PSA for a partial process in a fuel assembly fabrication facility is described. The underlying data are partially taken from an older German facility, other parts are generic. Moreover, a selected set of reported events corresponding to this partial process is taken as auxiliary data. The investigation of this partial process from the fuel fabrication as an example application shows that PSA methods are in principle applicable to nuclear fuel cycle facilities. Here, the focus is on preventing an initiating event, so that the system analysis is directed to the modeling of fault trees for initiating events. The quantitative results of this exemplary study are given as point values for the average occurrence frequencies. They include large uncertainties because of the limited documentation and data basis available, and thus have only methodological character. While quantitative results are given, further detailed information on process components and process flow is strongly required for robust conclusions with respect to the real process. (authors)

  19. Development of Automatic Quality Check Software in Mailbox Declaration For Nuclear Fuel Fabrication Plants

    International Nuclear Information System (INIS)

    Kim, Minsu; Shim, Hye Won; Jo, Seong Yeon; Lee, Kwang Yeol; Ban, Myoung Jin

    2014-01-01

    Short Notice Random Inspection (SNRI) is a new IAEA safeguards inspection regime for bulk handing facility, which utilities random inspection through a mailbox system. Its main objective is to verify 100% of the flow components of the safeguarded nuclear material at such a facility. To achieve the SNRI objective, it is required to provide daily mailbox declaration, by a facility's operator, to the IAEA with regard to information, such as the receipt and shipment of nuclear materials. Mailbox declarations are then later compared with accounting records so as to examine the accuracy and consistency of the facility operator's declaration at the time of the SNRI. The IAEA has emphasized the importance of accurate mailbox declarations and recommended that the ROK initiate its own independent quality control system in order to improve and maintain its mailbox declarations as a part of the SSAC activities. In an effort to improve the transparency of operational activities at fuel fabrication plants and to satisfy IAEA recommendation, an automatic quality check software application has been developed to improve mailbox declarations at fabrication plants in Korea. The ROK and the IAEA have recognized the importance of providing good quality mailbox declaration for an effective and efficient SNRI at fuel fabrication plants in Korea. The SRA developed an automatic quality check software program in order to provide an independent QC system of mailbox declaration, as well as to improve the quality of mailbox declaration. Once the automatic QC system is implemented, it will improve the quality of an operator's mailbox declaration by examining data before sending it to the IAEA. The QC system will be applied to fuel fabrication plants in the first half of 2014

  20. IAEA safeguards to prevent nuclear matrials diversion for fabrication of nuclear explosives

    International Nuclear Information System (INIS)

    Preuschen von und zu Liebenstein, R.

    1982-01-01

    The IAEA precautionary measures in accordance with the Non-Proliferation Treaty can be characterized as measures creating confidence. They constitute at present the essential basis for peaceful use of atomic energy. Even though there is a lot of criticism concerning the efficiency of the precautionary measures, and all justified calls for the elaboration of further legal instruments against nuclear materials diversion must not be neglected, the IAEA precautionary measures have already in a credible way contributed to contain the proliferation of nuclear weapons. (orig./HSCH) [de

  1. Flexible manufacturing systems and their relevance in nuclear fuel fabrication in India

    International Nuclear Information System (INIS)

    Ramakumar, M.S.

    1989-01-01

    Fabrication of nuclear reactor fuel bundle involves several materials and a number of complicated technologies and the process of manufacture has to conform to stringent standards. The Indian Nuclear Programme relies heavily on indigeneous capability of manufacture of nuclear fuels as well as automation of the related facilities. Automation of the existing nuclear facilities is a challenge in view of the characteristic plant environments and process demands as well as the various mechanical and metallurgical steps involved. This paper discusses their requirements and the measures initiated for achieving a high order of automation in Indian nuclear facilities. As a first step, specific automation steps are being incorporated in the existing plants. Such interface automation will enhance productivity and avoid the need for building new totally automated palnts. Flexible manufacturing system as applied here, has a different connotation vis-a-vis conventional manufacturing industry. Robotic devices, even for stacking jobs, have not been used on a large scale the world over. (author). 6 figs

  2. Maintenance of nuclear chemical and fuel fabrication plants [Invited talk no. IT-3

    International Nuclear Information System (INIS)

    Prasad, A.M.

    1981-01-01

    Though the objective of the maintenance practices followed in nuclear facilities is to optimise production as in other conventional production plants, the radioactivity associated with nuclear materials is a major constraint in all maintenance jobs on equipment of the nuclear facility. Often non-routine maintenance have to be adopted. Maintenance aspect has to be taken into consideration at the design stage of the nuclear facility. The maintenance concept adopted in a nuclear facility depends on the type of plant and varies from full indirect remote maintenance to direct contact maintenance. This is illustrated by discussing maintenance practices followed in a fuel reprocessing plant, a high level radioactive waste management facility, a fuel fabrication plant, and a heavy water plant. Exposure of maintenance staff to radiation has to be kept within limits governed by safety regulations. Along with planning and scheduling of maintenance, training of manpower with mock-up facilities assumes importance and the maintenance jobs must be carried out under strict supervision. (M.G.B.)

  3. Lesion dose in differentiated thyroid carcinoma metastases after rhTSH or thyroid hormone withdrawal: {sup 124}I PET/CT dosimetric comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Freudenberg, Lutz Stefan; Jentzen, Walter; Brandau, Wolfgang; Bockisch, Andreas [University of Duisburg/Essen, Department of Nuclear Medicine, Essen (Germany); Petrich, Thorsten; Knapp, Wolfram H. [Hanover University School of Medicine, Department of Nuclear Medicine, Hanover (Germany); Froemke, Cornelia [Hanover University School of Medicine, Institute of Biometry, Hanover (Germany); Marlowe, Robert J. [Spencer-Fontayne Corporation, Jersey City, NJ (United States); Heusner, Till [University of Duisburg/Essen, Department of Diagnostic and Interventional Radiology and Neuroradiology, Essen (Germany)

    2010-12-15

    Renal radioiodine excretion is {proportional_to}50% faster during euthyroidism versus hypothyroidism. We therefore sought to assess lesion dose/GBq of administered {sup 131}I activity (LDpA) in iodine-avid metastases (IAM) of differentiated thyroid carcinoma (DTC) in athyreotic patients after recombinant human thyroid-stimulating hormone (rhTSH) versus after thyroid hormone withdrawal (THW). We retrospectively compared mean LDpA between groups of consecutive patients (N = 63) receiving {sup 124}I positron emission tomography/computed tomography ({sup 124}I PET/CT) aided by rhTSH (n = 27) or THW (n = 36); we prospectively compared LDpA after these stimulation methods within another individual. Data derived from serial PET scans and one CT scan performed 2-96 h post-{sup 124}I ingestion. A mixed model analysis of covariance (ANCOVA) calculated the treatment groups' mean LDpAs adjusting for statistically significant baseline intergroup differences: non-IAM were more prevalent, median IAM count/patient lower in cervical lymph nodes and higher in distant sites, median stimulated thyroglobulin higher, mean cumulative radioiodine activity greater and prior diagnostic scintigraphy more frequent in the rhTSH patients. Mean LDpAs were: rhTSH group (n = 71 IAM), 30.6 Gy/GBq; THW group (n = 66 IAM), 51.8 Gy/GBq. The difference in group means (rhTSH less THW), -21.2 Gy/GBq, was statistically non-significant (p = 0.1667). However, the 95% confidence interval of that difference (-51.4 to + 9 Gy/GBq) suggested a trend favouring THW. The within-patient comparison found 2.9- to 10-fold higher LDpAs under THW. We found some suggestions, but no statistically significant evidence, that rhTSH administration results in a lower radiation dose to DTC metastases than does THW. A large, well-controlled, prospective within-patient study should resolve this issue. (orig.)

  4. Imaging Expression of Cytosine Deaminase-Herpes Virus Thymidine Kinase Fusion Gene (CD/TK Expression with [124I]FIAU and PET

    Directory of Open Access Journals (Sweden)

    Trevor Hackman

    2002-01-01

    Full Text Available Double prodrug activation gene therapy using the Escherichia coli cytosine deaminase (CDherpes simplex virus type 1 thymidine kinase (HSV1-tk fusion gene (CD/TK with 5-fluorocytosine (5FC, ganciclovir (GCV, and radiotherapy is currently under evaluation for treatment of different tumors. We assessed the efficacy of noninvasive imaging with [124I]FIAU (2′-fluoro-2′-deoxy-1-β-d-arabinofuranosyl-5-iodo-uracil and positron emission tomography (PET for monitoring expression of the CD/TK fusion gene. Walker-256 tumor cells were transduced with a retroviral vector bearing the CD/TK gene (W256CD/TK cells. The activity of HSV1-TK and CD subunits of the CD/TK gene product was assessed in different single cell-derived clones of W256CD/TK cells using the FIAU radiotracer accumulation assay in cells and a CD enzyme assay in cell homogenates, respectively. A linear relationship was observed between the levels of CD and HSV1-tk subunit expression in corresponding clones in vitro over a wide range of CD/TK expression levels. Several clones of W256CD/TK cells with significantly different levels of CD/TK expression were selected and used to produce multiple subcutaneous tumors in rats. PET imaging of HSV1-TK subunit activity with [124I]FIAU was performed on these animals and demonstrated that different levels of CD/TK expression in subcutaneous W256CD/TK tumors can be imaged quantitatively. CD expression in subcutaneous tumor sample homogenates was measured using a CD enzyme assay. A comparison of CD and HSV1-TK subunit enzymatic activity of the CD/TK fusion protein in vivo showed a significant correlation. Knowing this relationship, the parametric images of CD subunit activity were generated. Imaging with [124I]FIAU and PET could provide pre- and posttreatment assessments of CD/TK-based double prodrug activation in clinical gene therapy trials.

  5. Time efficient 124I-PET volumetry in benign thyroid disorders by automatic isocontour procedures: mathematic adjustment using manual contoured measurements in low-dose CT.

    Science.gov (United States)

    Freesmeyer, Martin; Kühnel, Christian; Westphal, Julian G

    2015-01-01

    Benign thyroid diseases are widely common in western societies. However, the volumetry of the thyroid gland, especially when enlarged or abnormally formed, proves to be a challenge in clinical routine. The aim of this study was to develop a simple and rapid threshold-based isocontour extraction method for thyroid volumetry from (124)I-PET/CT data in patients scheduled for radioactive iodine therapy. PET/CT data from 45 patients were analysed 30 h after 1 MBq (124)I administration. Anatomical reference volume was calculated using manually contoured data from low-dose CT images of the neck (MC). In addition, we applied an automatic isocontour extraction method (IC0.2/1.0), with two different threshold values (0.2 and 1.0 kBq/ml), for volumetry of the PET data-set. IC0.2/1.0 shape data that showed significant variation from MC data were excluded. Subsequently, a mathematical correlation using a model of linear regression with multiple variables and step-wise elimination (mIC0.2/1.0), between IC0.2/1.0 and MC, was established. Data from 41 patients (IC0.2), and 32 patients (IC1.0) were analysed. The mathematically calculated volume, mIC, showed a median deviation from the reference (MC), of ±9 % (1-54 %) for mIC0.2 and of ±8.2 % (1-50 %) for mIC1.0 CONCLUSION: Contour extraction with both, mIC1.0 and mIC0.2 gave rapid and reliable results. However, mIC0.2 can be applied to significantly more patients (>90 %) and is, therefore, deemed to be more suitable for clinical routine, keeping in mind the potential advantages of using (124)I-PET/CT for the preparation of patients scheduled for radioactive iodine therapy.

  6. The influence of saliva flow stimulation on the absorbed radiation dose to the salivary glands during radioiodine therapy of thyroid cancer using 124I PET(/CT) imaging

    International Nuclear Information System (INIS)

    Jentzen, Walter; Schmitz, Jochen; Freudenberg, Lutz; Eising, Ernst; Bockisch, Andreas; Stahl, Alexander; Balschuweit, Dorothee; Hilbel, Thomas

    2010-01-01

    A serious side effect of high-activity radioiodine therapy in the treatment of differentiated thyroid cancer is radiogenic salivary gland damage. This damage may be diminished by lemon-juice-induced saliva flow immediately after 131 I administration. The aim of this study was to assess the effect of chewing lemon slices on the absorbed (radiation) doses to the salivary glands. Ten patients received (pretherapy) 124 I PET(/CT) dosimetry before their first radioiodine therapy. The patients underwent a series of six PET scans at 0.5, 1, 2, 4, 48 and ≥96 h and one PET/CT scan at 24 h after administration of 27 MBq 124 I. Blood samples were also collected at about 2, 4, 24, 48, and 96 h. Contrary to the standard radioiodine therapy protocol, the patients were not stimulated with lemon juice. Specifically, the patients chewed no lemon slices during the pretherapy procedure and neither ate food nor drank fluids until after completion of the last PET scan on the first day. Organ absorbed doses per administered 131 I activity (ODpAs) as well as gland and blood uptake curves were determined and compared with published data from a control patient group, i.e. stimulated per the standard radioiodine therapy protocol. The calculations for both groups used the same methodology. A within-group comparison showed that the mean ODpA for the submandibular glands was not significantly different from that for the parotid glands. An intergroup comparison showed that the mean ODpA in the nonstimulation group averaged over both gland types was reduced by 28% compared to the mean ODpA in the stimulation group (p=0.01). Within each gland type, the mean ODpA reductions in the nonstimulation group were statistically significant for the parotid glands (p=0.03) but not for the submandibular glands (p=0.23). The observed ODpAs were higher in the stimulation group because of increased initial gland uptake rather than group differences in blood kinetics. The 124 I PET(/CT) salivary gland

  7. Fabrication and optical characterization of cadmium sulfide needles using nuclear track membrane

    International Nuclear Information System (INIS)

    Peng, L.Q.; Wang, S.C.; Ju, X.; Xiao, H.; Chen, H.; He, Y.J.

    1999-01-01

    Cadmium sulfide needles with a diameter of 0.2 μm have been fabricated in nuclear track polyethylene-terephthalate (PET) membrane by electrochemically depositing from organic solvent dimethylsulfoxide (DMSO) containing CdCl 2 and elemental sulfur at the temperature 110 deg. C. The characterization of the sample of CdS needles was studied by scanning electron microscope, X-ray diffraction, absorption and photoluminescence spectra. The optical experiments show that in the sample of CdS needles there is an absorption peak that could be assigned to the interface states of the CdS needles

  8. Fabrication and optical characterization of cadmium sulfide needles using nuclear track membrane

    Energy Technology Data Exchange (ETDEWEB)

    Peng, L.Q.; Wang, S.C.; Ju, X.; Xiao, H.; Chen, H.; He, Y.J

    1999-06-01

    Cadmium sulfide needles with a diameter of 0.2 {mu}m have been fabricated in nuclear track polyethylene-terephthalate (PET) membrane by electrochemically depositing from organic solvent dimethylsulfoxide (DMSO) containing CdCl{sub 2} and elemental sulfur at the temperature 110 deg. C. The characterization of the sample of CdS needles was studied by scanning electron microscope, X-ray diffraction, absorption and photoluminescence spectra. The optical experiments show that in the sample of CdS needles there is an absorption peak that could be assigned to the interface states of the CdS needles.

  9. Design, fabrication and installation of irradiation facilities -Advanced nuclear material development-

    International Nuclear Information System (INIS)

    Kim, Yong Seong; Lee, Jeong Yeong; Lee, Seong Ho; Ji, Dae Yeong; Kim, Seok Hoon; An, Seong Ho; Kim, Dong Hoon; Seok, Ho Cheon; Kim, Joon Yeon; Yang, Seong Hong

    1994-07-01

    The objective of this study is to design and construct the steady state fuel test loop and non-instrumented capsules to be installed in KMRR. The principle contents of this project are to design, fabricate the steady-state fuel test loop and non-instrumented capsule to be installed in KMRR for nuclear technology development. This project will be completed in 1996, so preparation of design criteria for fuel test loop have been performed in 1993 as the first year of the first phase in implementing this project. Also design and pressure drop test of non-instrumented capsule have been performed in 1993

  10. Application of gas shielded arc welding and submerged arc welding for fabrication of nuclear reactor vessels

    International Nuclear Information System (INIS)

    Gehani, M.L.; Rodrigues, W.D.

    1976-01-01

    The remarkable progress made in the development of knowhow and expertise in the manufacture of equipment for nuclear power plants in India is outlined. Some of the specific advances made in the application of higher efficiency weld processes for fabrication of nuclear reactor vessels and the higher level of quality attained are discussed in detail. Modifications and developments in submerged arc, gas tungsten arc and gas metal arc processes for welding of Calandria which have been a highly challenging and rewarding experience are discussed. Future scope for making the gas metal arc process more economical by using various gas-mixes like Agron + Oxygen, Argon + Carbon Dioxide, Argon + Nitrogen (for Copper Alloys) etc., in various proportions are outlined. Quality and dimensional control exercised in these jobs of high precision are highlighted. (K.B.)

  11. Nuclear materials accountancy in an industrial MOX fuel fabrication plant safeguards versus commercial aspects

    International Nuclear Information System (INIS)

    Canck, H. de; Ingels, R.; Lefevre, R.

    1991-01-01

    In a modern MOX Fuel Fabrication Plant, with a large throughput of nuclear materials, computerized real-time accountancy systems are applied. Following regulations and prescriptions imposed by the Inspectorates EURATOM-IAEA, the State and also by internal plant safety rules, the accountancy is kept in plutonium element, uranium element and 235 U for enriched uranium. In practice, Safeguards Authorities are concerned with quantities of the element (U tot , Pu tot ) and to some extent with its fissile content. Custom Authorities are for historical reasons, interested in fissile quantities (U fiss , Pu fiss ) whereas owners wish to recover the energetic value of their material (Pu equivalent). Balancing the accountancy simultaneously in all these related but not proportional units is a new problem in a MOX-plant where pool accountancy is applied. This paper indicates possible ways to solve the balancing problem created by these different units used for expressing nuclear material quantities

  12. PET-based compartmental modeling of {sup 124}I-A33 antibody: quantitative characterization of patient-specific tumor targeting in colorectal cancer

    Energy Technology Data Exchange (ETDEWEB)

    Zanzonico, Pat; O' Donoghue, Joseph A.; Humm, John L. [Memorial Sloan Kettering Cancer Center, Department of Medical Physics, New York, NY (United States); Carrasquillo, Jorge A.; Pandit-Taskar, Neeta; Ruan, Shutian; Larson, Steven M. [Memorial Sloan Kettering Cancer Center, Department of Radiology, New York, NY (United States); Smith-Jones, Peter [Memorial Sloan Kettering Cancer Center, Department of Radiology, New York, NY (United States); Stony Brook School of Medicine, Departments of Psychiatry and Radiology, Stony Brook, NY (United States); Divgi, Chaitanya [Columbia University Medical Center, New York, NY (United States); Scott, Andrew M. [La Trobe University, Olivia Newton-John Cancer Research Institute, Melbourne (Australia); Kemeny, Nancy E.; Wong, Douglas; Scheinberg, David [Memorial Sloan Kettering Cancer Center, Department of Medicine, New York, NY (United States); Fong, Yuman [Memorial Sloan Kettering Cancer Center, Department of Surgery, New York, NY (United States); City of Hope, Department of Surgery, Duarte, CA (United States); Ritter, Gerd; Jungbluth, Achem; Old, Lloyd J. [Memorial Sloan Kettering Cancer Center, Ludwig Institute for Cancer Research, New York, NY (United States)

    2015-10-15

    The molecular specificity of monoclonal antibodies (mAbs) directed against tumor antigens has proven effective for targeted therapy of human cancers, as shown by a growing list of successful antibody-based drug products. We describe a novel, nonlinear compartmental model using PET-derived data to determine the ''best-fit'' parameters and model-derived quantities for optimizing biodistribution of intravenously injected {sup 124}I-labeled antitumor antibodies. As an example of this paradigm, quantitative image and kinetic analyses of anti-A33 humanized mAb (also known as ''A33'') were performed in 11 colorectal cancer patients. Serial whole-body PET scans of {sup 124}I-labeled A33 and blood samples were acquired and the resulting tissue time-activity data for each patient were fit to a nonlinear compartmental model using the SAAM II computer code. Excellent agreement was observed between fitted and measured parameters of tumor uptake, ''off-target'' uptake in bowel mucosa, blood clearance, tumor antigen levels, and percent antigen occupancy. This approach should be generally applicable to antibody-antigen systems in human tumors for which the masses of antigen-expressing tumor and of normal tissues can be estimated and for which antibody kinetics can be measured with PET. Ultimately, based on each patient's resulting ''best-fit'' nonlinear model, a patient-specific optimum mAb dose (in micromoles, for example) may be derived. (orig.)

  13. Investigation of 3′-debenzoyl-3′-(3-([124I]-iodobenzoyl))paclitaxel analog as a radio-tracer to study multidrug resistance in vivo

    International Nuclear Information System (INIS)

    Sajjad, M.; Riaz, U.; Yao, R.; Bernacki, R.J.; Abouzied, M.; Erb, D.A.; Chaudhary, N.D.; Veith, J.M.; Georg, G.I.; Nabi, H.A.

    2012-01-01

    A study was carried out to identify a suitable radioactive paclitaxel analog and to use it to investigate tumor multidrug resistance in vivo. 3′-Debenzoyl-3′-(3-([ 124 I]-iodobenzoyl))paclitaxel was prepared by aromatic iodination of 3′-debenzoyl-3′-(3-trimethylstannylbenzoyl)paclitaxel. Uptake of the labeled paclitaxel analog in nude mice bearing tumor with the paclitaxel sensitive cancer cell lines MCF7 and MDA-435/LCC6(WT), and multidrug resistant cell lines NCI/ADR-RES and MDA-435/LCC6(MDR), was studied. There was no difference in drug level between the sensitive and resistant MDA-435/LCC6 tumors at 6 h post-injection. However, at 6 h, there was a significant increase in drug level for the MCF7 tumor as compared with the NCI/ADR-RES tumor, presumably due to increased drug retention. At 24 h, drug uptake/retention was significantly higher in both sensitive tumor cell lines as compared to their drug resistant counterparts. Pretreatment of mice with MDR transport modulators, Cyclosporine or tRA 96029, did not increase the level of labeled paclitaxel analog in the drug resistant MDA-435/LCC6(MDR) tumor. On the other hand, at 24 h Cyclosporine apparently increased analog level in the drug sensitive MDA-435/LCC6(WT) tumor, aiding drug imaging studies. - Highlights: ► 3′-Debenzoyl-3′-(3-iodobenzoyl)paclitaxel cytotoxicity was comparable to paclitaxel. ► 3′-Debenzoyl-3′-(3-([ 124 I]-iodobenzoyl)paclitaxel was synthesized. ► Uptake of the drug was higher in sensitive tumor compared to the resistant tumor. ► The Pgp-modulators had a positive effect on drug-sensitive tumor. ► The sensitive tumor was visible in images obtained using micoPET.

  14. Nondestructive assay of special nuclear material for uranium fuel-fabrication facilities

    International Nuclear Information System (INIS)

    Smith, H.A. Jr.; Schillebeeckx, P.

    1997-01-01

    A high-quality materials accounting system and effective international inspections in uranium fuel-fabrication facilities depend heavily upon accurate nondestructive assay measurements of the facility's nuclear materials. While item accounting can monitor a large portion of the facility inventory (fuel rods, assemblies, storage items), the contents of all such items and mass values for all bulk materials must be based on quantitative measurements. Weight measurements, combined with destructive analysis of process samples, can provide highly accurate quantitative information on well-characterized and uniform product materials. However, to cover the full range of process materials and to provide timely accountancy data on hard-to-measure items and rapid verification of previous measurements, radiation-based nondestructive assay (NDA) techniques play an important role. NDA for uranium fuel fabrication facilities relies on passive gamma spectroscopy for enrichment and U isotope mass values of medium-to-low-density samples and holdup deposits; it relies on active neutron techniques for U-235 mass values of high-density and heterogeneous samples. This paper will describe the basic radiation-based nondestructive assay techniques used to perform these measurements. The authors will also discuss the NDA measurement applications for international inspections of European fuel-fabrication facilities

  15. Pulsed TIG welding in the fabrication of nuclear components and structures

    International Nuclear Information System (INIS)

    Lucas, W.; Males, B.O.

    1979-01-01

    TIG welding is an important welding technique in nuclear plant fabrication for the welding of critical components and structures where a high level of weld integrity is demanded. Whilst the process is ideally suited to precision welding, since the arc is a small intense heat source, it has proved to be somewhat intolerant to production variations in 'difficult' applications, such as tube to tube plate welding and orbital tube welding with tube in the fixed position. Whilst the problems directly associated with this intolerance (of the welding process) are less frequently observed when used manually, difficulties are experienced in fully mechanised welding operations particularly when welding to a relatively rigid approved procedure. Pulsing of the welding current was developed as a technique to achieve greater control of the behaviour of the weld pool. Instead of moving the weld pool in a continuous motion around the joint, welding was conducted intermittently in the form of overlapping spots. This technique, which offers significant advantages over continuous current welding has been exploited in nuclear fabrication for welding those components which demand a high level of weld quality. In this paper, the essential features of this technique are described and, in indicating its advantages, examples have been drawn from recent experiences on the welding of two types of joint for the Advanced Gas Cooled Reactor, a tube sheet and a butt joint in the G Position. (author)

  16. Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.

    1997-01-01

    Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

  17. Statistical methods to assess and control processes and products during nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Weidinger, H.

    1999-01-01

    Very good statistical tools and techniques are available today to access the quality and the reliability of fabrication process as the original sources for a good and reliable quality of the fabricated processes. Quality control charts of different types play a key role and the high capability of modern electronic data acquisition technologies proved, at least potentially, a high efficiency in the more or less online application of these methods. These techniques focus mainly on stability and the reliability of the fabrication process. In addition, relatively simple statistical tolls are available to access the capability of fabrication process, assuming they are stable, to fulfill the product specifications. All these techniques can only result in as good a product as the product design is able to describe the product requirements necessary for good performance. Therefore it is essential that product design is strictly and closely performance oriented. However, performance orientation is only successful through an open and effective cooperation with the customer who uses or applies those products. During the last one to two decades in the west, a multi-vendor strategy has been developed by the utility, sometimes leading to three different fuel vendors for one reactor core. This development resulted in better economic conditions for the user but did not necessarily increase an open attitude with the vendor toward the using utility. The responsibility of the utility increased considerably to ensure an adequate quality of the fuel they received. As a matter of fact, sometimes the utilities had to pay a high price because of unexpected performance problems. Thus the utilities are now learning that they need to increase their knowledge and experience in the area of nuclear fuel quality management and technology. This process started some time ago in the west. However, it now also reaches the utilities in the eastern countries. (author)

  18. Characterisation and fabrication of zirconia and thoria based ceramics for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Barrier, D C

    2005-11-01

    The reduction of the long term radiotoxicity of nuclear waste during disposal is the aim of the research called ''Partitioning and Transmutation of Minor actinides (MAs)'', which also requires the development of inert ceramic support materials. Moreover, after separation, if the transmutation is not available, the actinides can be conditioned into stable dedicated solid matrices (Partitioning and Conditioning strategy). Yttrium-stabilized zirconia and thoria are discussed in the international nuclear community as candidates for the fixation of long-lived actinides as target material for transmutation and as stable materials for long-term final disposal. The aims of the following work are twofold: determine the impact of the addition of actinides, simulated by cerium on the properties of the matrices and study the possibility of synthesising homogeneous ceramics using simple fabrication routes. Within this framework, (ZrY)O{sub 2-x}-CeO{sub 2} and ThO{sub 2}-CeO{sub 2} powders with variable ceria contents (from 0 to 100 %) were synthesised by a co-precipitation method of nitrate solution. The influence of ceria concentration on the powder' properties, such as thermal behaviour and the evolution of material crystallisation during annealing, was investigated in detail by thermogravimetry (TG) coupled with differential scanning calorimetry (DSC) and X-ray diffraction (XRD). Both systems crystallise at high temperature in a stable solid solution, fcc, fluorite type structure and follow the Vegard's law for the complete range of ceria. For both systems a critical concentration of 20 mol% has been established. For ceria concentrations lower than 20%, the properties of the system are mainly controlled by the matrix. Pellets with different ceria concentrations were compacted from these powders by using different technological cycles. In order to obtain materials with reliable properties, the technological parameters of each chosen fabrication route, have been optimised. By

  19. Characterisation and fabrication of zirconia and thoria based ceramics for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Barrier, D.C.

    2005-11-01

    The reduction of the long term radiotoxicity of nuclear waste during disposal is the aim of the research called ''Partitioning and Transmutation of Minor actinides (MAs)'', which also requires the development of inert ceramic support materials. Moreover, after separation, if the transmutation is not available, the actinides can be conditioned into stable dedicated solid matrices (Partitioning and Conditioning strategy). Yttrium-stabilized zirconia and thoria are discussed in the international nuclear community as candidates for the fixation of long-lived actinides as target material for transmutation and as stable materials for long-term final disposal. The aims of the following work are twofold: determine the impact of the addition of actinides, simulated by cerium on the properties of the matrices and study the possibility of synthesising homogeneous ceramics using simple fabrication routes. Within this framework, (ZrY)O{sub 2-x}-CeO{sub 2} and ThO{sub 2}-CeO{sub 2} powders with variable ceria contents (from 0 to 100 %) were synthesised by a co-precipitation method of nitrate solution. The influence of ceria concentration on the powder' properties, such as thermal behaviour and the evolution of material crystallisation during annealing, was investigated in detail by thermogravimetry (TG) coupled with differential scanning calorimetry (DSC) and X-ray diffraction (XRD). Both systems crystallise at high temperature in a stable solid solution, fcc, fluorite type structure and follow the Vegard's law for the complete range of ceria. For both systems a critical concentration of 20 mol% has been established. For ceria concentrations lower than 20%, the properties of the system are mainly controlled by the matrix. Pellets with different ceria concentrations were compacted from these powders by using different technological cycles. In order to obtain materials with reliable properties, the technological parameters of each chosen fabrication

  20. Characterisation and fabrication of zirconia and thoria based ceramics for nuclear applications

    International Nuclear Information System (INIS)

    Barrier, D.C.

    2005-11-01

    The reduction of the long term radiotoxicity of nuclear waste during disposal is the aim of the research called ''Partitioning and Transmutation of Minor actinides (MAs)'', which also requires the development of inert ceramic support materials. Moreover, after separation, if the transmutation is not available, the actinides can be conditioned into stable dedicated solid matrices (Partitioning and Conditioning strategy). Yttrium-stabilized zirconia and thoria are discussed in the international nuclear community as candidates for the fixation of long-lived actinides as target material for transmutation and as stable materials for long-term final disposal. The aims of the following work are twofold: determine the impact of the addition of actinides, simulated by cerium on the properties of the matrices and study the possibility of synthesising homogeneous ceramics using simple fabrication routes. Within this framework, (ZrY)O 2-x -CeO 2 and ThO 2 -CeO 2 powders with variable ceria contents (from 0 to 100 %) were synthesised by a co-precipitation method of nitrate solution. The influence of ceria concentration on the powder' properties, such as thermal behaviour and the evolution of material crystallisation during annealing, was investigated in detail by thermogravimetry (TG) coupled with differential scanning calorimetry (DSC) and X-ray diffraction (XRD). Both systems crystallise at high temperature in a stable solid solution, fcc, fluorite type structure and follow the Vegard's law for the complete range of ceria. For both systems a critical concentration of 20 mol% has been established. For ceria concentrations lower than 20%, the properties of the system are mainly controlled by the matrix. Pellets with different ceria concentrations were compacted from these powders by using different technological cycles. In order to obtain materials with reliable properties, the technological parameters of each chosen fabrication route, have been optimised. By employing mild wet

  1. Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

    A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

  2. Fabrication and characterization of joined silicon carbide cylindrical components for nuclear applications

    Science.gov (United States)

    Khalifa, H. E.; Deck, C. P.; Gutierrez, O.; Jacobsen, G. M.; Back, C. A.

    2015-02-01

    The use of silicon carbide (SiC) composites as structural materials in nuclear applications necessitates the development of a viable joining method. One critical application for nuclear-grade joining is the sealing of fuel within a cylindrical cladding. This paper demonstrates cylindrical joint feasibility using a low activation nuclear-grade joint material comprised entirely of β-SiC. While many papers have considered joining material, this paper takes into consideration the joint geometry and component form factor, as well as the material performance. Work focused specifically on characterizing the strength and permeability performance of joints between cylindrical SiC-SiC composites and monolithic SiC endplugs. The effects of environment and neutron irradiation were not evaluated in this study. Joint test specimens of different geometries were evaluated in their as-fabricated state, as well as after being subjected to thermal cycling and partial mechanical loading. A butted scarf geometry supplied the best combination of high strength and low permeability. A leak rate performance of 2 × 10-9 mbar l s-1 was maintained after thermal cycling and partial mechanical loading and sustained applied force of 3.4 kN, or an apparent strength of 77 MPa. This work shows that a cylindrical SiC-SiC composite tube sealed with a butted scarf endplug provides out-of-pile strength and permeability performance that meets light water reactor design requirements.

  3. Recycling of nuclear fuel swarf at the fabrication of UO sub(2)-pellets and its influence on the irradiation behavior

    International Nuclear Information System (INIS)

    Dias, M.S.; Lameiras, F.S.; Santos, A.M.M. dos

    1991-01-01

    From the fabrication of UO sub(2) pellets for light water reactor fuel rods, nuclear fuel scraps results in form of UO sub(2) grinding swarf and UO sub(2) sinter scraps oxidized to U sub(3)O sub(8) powder. Detailed investigations on five types of UO sub(2) pellets fabricated with different portions of this scrap kinds added to the UO sub(2) press powder showed that there is only a small influence of such scrap additions on the irradiation behavior, especially for the fission gas release. This allows to recycle the fabrication scrap in a simple and economic way. (author)

  4. Low-activity 124I-PET/low-dose CT versus 99mTc-pertechnetate planar scintigraphy or 99mTc-pertechnetate single-photon emission computed tomography of the thyroid: a pilot comparison.

    Science.gov (United States)

    Darr, Andreas M; Opfermann, Thomas; Niksch, Tobias; Driesch, Dominik; Marlowe, Robert J; Freesmeyer, Martin

    2013-10-01

    The standard thyroid functional imaging method, 99mTc-pertechnetate (99mTc-PT) planar scintigraphy, has technical drawbacks decreasing its sensitivity in detecting nodules or anatomical pathology. 124I-PET, lacking these disadvantages and allowing simultaneous CT, may have greater sensitivity for these purposes. We performed a blinded pilot comparison of 124I-PET(/CT) versus 99mTc-PT planar scintigraphy or its cross-sectional enhancement, 99mTc-PT single-photon emission CT (SPECT), in characterizing the thyroid gland with benign disease. Twenty-one consecutive adults with goiter underwent low-activity (1 MBq/0.027 mCi) 124I-PET/low-dose (30 mAs) CT, 99mTc-PT planar scintigraphy, and 99mTc-PT-SPECT. Endpoints included the numbers of “hot spots” with/without central photopenia and “cold spots” detected, the proportion of these lesions with morphological correlates, the mean volume and diameter of visualized nodules, and the number of cases of lobus pyramidalis or retrosternal thyroid tissue identified. 124I-PET detected significantly more “hot spots” with/without central photopenia (P < 0.001), significantly more nodules (P < 0.001), and more “cold spots” than did 99mTc-PT planar scintigraphy or 99mTc-PT-SPECT, including all lesions seen on the 99mTc-PT modalities. Ultrasonographic correlates were found for all nodules visualized on all 3 modalities and 92.5% of nodules seen only on 124I-PET. Nodules discernible only on 124I-PET had significantly smaller mean volume or diameter (P < 0.001) than did those visualized on 99mTc-PT planar scintigraphy or 99mTc-PT-SPECT. 124I-PET(/CT) identified significantly more patients with a lobus pyramidalis (P < 0.001) or retrosternal thyroid tissue (P < 0.05). 124I-PET(/CT) may provide superior imaging of benign thyroid disease compared to planar or cross-sectional 99mTc-PT scintigraphy.

  5. Radiation dosimetry and first therapy results with a 124I/131I-labeled small molecule (MIP-1095) targeting PSMA for prostate cancer therapy

    International Nuclear Information System (INIS)

    Zechmann, Christian M.; Afshar-Oromieh, Ali; Mier, Walter; Armor, Tom; Joyal, John; Stubbs, James B.; Hadaschik, Boris; Kopka, Klaus; Debus, Juergen; Babich, John W.; Haberkorn, Uwe

    2014-01-01

    Since the prostate-specific membrane antigen (PSMA) is frequently over-expressed in prostate cancer (PCa) several PSMA-targeting molecules are under development to detect and treat metastatic castration resistant prostate cancer (mCRPC). We investigated the tissue kinetics of a small molecule inhibitor of PSMA ((S)-2-(3-((S)-1-carboxy-5-(3-(4-[ 124 I]iodophenyl)ureido)pentyl)ureido) pentan edioicacid; MIP-1095) using PET/CT to estimate radiation dosimetry for the potential therapeutic use of 131 I-MIP-1095 in men with mCRPC. We also report preliminary safety and efficacy of the first 28 consecutive patients treated under a compassionate-use protocol with a single cycle of 131 I-MIP-1095. Sixteen patients with known prostate cancer underwent PET/CT imaging after i.v. administration of 124 I-MIP-1095 (mean activity: 67.4 MBq). Each patient was scanned using PET/CT up to five times at 1, 4, 24, 48 and 72 h post injection. Volumes of interest were defined for tumor lesions and normal organs at each time point followed by dose calculations using the OLINDA/EXM software. Twenty-eight men with mCRPC were treated with a single cycle of 131 I-MIP-1095 (mean activity: 4.8 GBq, range 2 to 7.2 GBq) and followed for safety and efficacy. Baseline and follow up examinations included a complete blood count, liver and kidney function tests, and measurement of serum PSA. I-124-MIP-1095 PET/CT images showed excellent tumor uptake and moderate uptake in liver, proximal intestine and within a few hours post-injection also in the kidneys. High uptake values were observed only in salivary and lacrimal glands. Dosimetry estimates for I-131-MIP-1095 revealed that the highest absorbed doses were delivered to the salivary glands (3.8 mSv/MBq), liver (1.7 mSv/MBq) and kidneys (1.4 mSv/MBq). The absorbed dose calculated for the red marrow was 0.37 mSv/MBq. PSA values decreased by >50 % in 60.7 % of the men treated. Of men with bone pain, 84.6 % showed complete or moderate reduction in pain

  6. Radiation dosimetry and first therapy results with a {sup 124}I/{sup 131}I-labeled small molecule (MIP-1095) targeting PSMA for prostate cancer therapy

    Energy Technology Data Exchange (ETDEWEB)

    Zechmann, Christian M.; Afshar-Oromieh, Ali; Mier, Walter [University Hospital Heidelberg, Department of Nuclear Medicine, Heidelberg (Germany); Armor, Tom; Joyal, John [Molecular Insight Pharmaceuticals, Boston, MA (United States); Stubbs, James B. [Radiation Dosimetry Systems RDS, Inc., Apharetta, GA (United States); Hadaschik, Boris [University Hospital Heidelberg, Department of Urology, Heidelberg (Germany); Kopka, Klaus [Division Radiopharmaceutical Chemistry, DKFZ, Heidelberg (Germany); Debus, Juergen [University Hospital Heidelberg, Department of Radiation Oncology, Heidelberg (Germany); Babich, John W. [Molecular Insight Pharmaceuticals, Boston, MA (United States); Cornell University, Division of Radiopharmacy, Department of Radiology, New York, NY (United States); Haberkorn, Uwe [University Hospital Heidelberg, Department of Nuclear Medicine, Heidelberg (Germany); Clinical Cooperation Unit Nuclear Medicine, DKFZ, Heidelberg (Germany)

    2014-07-15

    Since the prostate-specific membrane antigen (PSMA) is frequently over-expressed in prostate cancer (PCa) several PSMA-targeting molecules are under development to detect and treat metastatic castration resistant prostate cancer (mCRPC). We investigated the tissue kinetics of a small molecule inhibitor of PSMA ((S)-2-(3-((S)-1-carboxy-5-(3-(4-[{sup 124}I]iodophenyl)ureido)pentyl)ureido) pentan edioicacid; MIP-1095) using PET/CT to estimate radiation dosimetry for the potential therapeutic use of {sup 131}I-MIP-1095 in men with mCRPC. We also report preliminary safety and efficacy of the first 28 consecutive patients treated under a compassionate-use protocol with a single cycle of {sup 131}I-MIP-1095. Sixteen patients with known prostate cancer underwent PET/CT imaging after i.v. administration of {sup 124}I-MIP-1095 (mean activity: 67.4 MBq). Each patient was scanned using PET/CT up to five times at 1, 4, 24, 48 and 72 h post injection. Volumes of interest were defined for tumor lesions and normal organs at each time point followed by dose calculations using the OLINDA/EXM software. Twenty-eight men with mCRPC were treated with a single cycle of {sup 131}I-MIP-1095 (mean activity: 4.8 GBq, range 2 to 7.2 GBq) and followed for safety and efficacy. Baseline and follow up examinations included a complete blood count, liver and kidney function tests, and measurement of serum PSA. I-124-MIP-1095 PET/CT images showed excellent tumor uptake and moderate uptake in liver, proximal intestine and within a few hours post-injection also in the kidneys. High uptake values were observed only in salivary and lacrimal glands. Dosimetry estimates for I-131-MIP-1095 revealed that the highest absorbed doses were delivered to the salivary glands (3.8 mSv/MBq), liver (1.7 mSv/MBq) and kidneys (1.4 mSv/MBq). The absorbed dose calculated for the red marrow was 0.37 mSv/MBq. PSA values decreased by >50 % in 60.7 % of the men treated. Of men with bone pain, 84.6 % showed complete or

  7. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    International Nuclear Information System (INIS)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

    2011-01-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  8. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    Energy Technology Data Exchange (ETDEWEB)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  9. Steel, specially for the fabrication of welded structure working under pressure in nuclear installations

    International Nuclear Information System (INIS)

    Dolbenko, E.T.; Astafiev, A.A.; Kark, G.S.

    1981-01-01

    The present invention is in the field of metallurgy. Steels have found an increasing number of applications in mechanical constructions, and notably in the construction of materials for the production of energy and for the fabrication of welded structures operating under pressure at temperatures as high as 450 0 C. A possible application is the pressurized vessels of nuclear facilities. The steels of interest contain carbon, silicon, manganese, nickel, molybdenum, vanadium, aluminium, nitrogen, phosphorus and iron, but are characterized by the fact that they also contain arsenic, tin and calcium. The sum of the weighted percentages of nickel and manganese and the weighted percentage of phosphorous are related as follows: (Ni + Mn) . P [fr

  10. Subsoil exploration of the estimated building site for nuclear fuel development and fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Song, In Taek [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    The objective of this report, as the result of subsoil exploration, is to provide basic design data of structural plan for nuclear fuel development and fabrication facility that is builded on Duckjin 150, Yusong, Taejeon, Korea, and provide basic data for execution of work. The soft rock level of estimated building site is deep(18.0m:BH-1, 20.5m:BH-2, 25.5m:BH-3) and the hard rock level of it is very deep (33.0m:BH-1, 46.0m:BH-2, 34.5m:BH-3) , for structural design, the hard rock shall be the bottom of foundation. 9 figs., 19 tabs. (Author)

  11. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  12. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N.

    2004-01-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO 2 , Al 2 O 3 , Gd 2 O 3 , etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO 2 ) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al 2 O 3 ) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO 2 for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate casks with variable cermet compositions

  13. Informal presentations by fuel fabricators and others [contributed by A. Nishiyama, Nuclear Fuel Industries, Ltd.

    International Nuclear Information System (INIS)

    Nishiyama, A.

    1993-01-01

    This paper contains a brief summary of activities in the field of research reactor fuel fabrication in Nuclear Fuel Industries Sumitomo and Furukawa Industries. Since 1956 2 million dollars were spent for development of nuclear fuels and plant facilities including complete manufacturing and testing capabilities. Now this company is the only fuel supplier for the research reactors in Japan. The fabrication process starts with the melting, alloying, and casting of U-Al. The uranium billets are prepared by foreign fabricators. The uranium content varies from 13 to 22 wt % according to the purchaser's specifications. In making fuel plates, the picture frame method is applied. In this case, the original procedure is sufficiently effective in avoiding dogboning. The plates are finished by hot and cold roll milling and inspected dimensionally, metallurgically, and mechanically, and at the same time the blister test and X-ray radiographic tests are performed. Fuel elements are assembled by rolling flat or curved plates into side plate grooves and end-fit welding. Finished elements are tested dimensionally and hydraulically. Nominal losses during operation are less than 1% of the uranium metal. Our present capacity licensed by the Japanese Government is approximately 950 fuel elements a year. About 35 employees including engineers are engaged in development and manufacturing of fuels. Owing to the small limited demand of the research reactor fuels in Japan during the past 20 years (mostly in last 10 years), we processed only about 350 kg of highly enriched uranium and supplied approximately 1000 fuel elements to JAERI, Kyoto University, and others, and we have been suffering red-ink balance of budget every year. Some of trials in development are briefly discussed. In case of UO 2 -Al metal fuel plates, the vibratory compacting method was very popular among many researchers about 10 years ago. A lot of time and money was spent to study the economic fabrication process of

  14. The Hanau atomic energy laws. Nuclear fuel fabrication and the administrative law system

    International Nuclear Information System (INIS)

    Becker-Neetz, G.; Uebersohn, G.

    1989-01-01

    The review concentrates on administrative law aspects in the discussion of problems relating to the licences and preliminary notices of approval issued for the Hanau nuclear industry. The authors deal with the licences granted in 1974 (according to sec. 9 Atomic Energy Act), with the extended licensing requirements of sec. 7 Atomic Energy Act as amended by the 3rd amendment (concerning fabrication and handling of nuclear fuels), and the criminal court proceedings examining the conduct of the Alkem management and senior officers of the Hessian Ministry of Economics. Specific aspects investigated in the review include continuation of existing operations in accordance with transitory provisions, replacement of existing by new installations, and preliminary notice of approval. The preliminary notices of approval given up to the date of December 31, 1977 are said to have been illegal and extinct at that date, but the court's decision to abstain from punishment is accepted. The authors outline some possibilities of giving more concrete shape to the judicial control by administrative courts. (RST) [de

  15. Radiological and environmental safety aspects of uranium fuel fabrication plants at Nuclear Fuel Complex, Hyderabad

    International Nuclear Information System (INIS)

    Viswanathan, S.; Surya Rao, B.; Lakshmanan, A.R.; Krishna Rao, T.

    1991-01-01

    Nuclear Fuel Complex, Hyderabad manufactures uranium dioxide fuel assemblies for PHWRs and BWRs operating in India. Starting materials are magnesium diuranate received from UCIL, Jaduguda and imported UF. Both of these are converted to UO 2 pellets by identical chemical processes and mechanical compacting. Since the uranium handled here is free of daughter product activities, external radiation is not a problem. Inhalation of airborne U compounds is one of the main source of exposure. Engineered protective measures like enclosures around U bearing powder handling equipment and local exhausts reduce worker's exposure. Installation of pre-filters, wet rotoclones and electrostatic precipitators in the ventillation system reduces the release of U into the environment. The criticality hazard in handling slightly enriched uranium is very low due to the built-in control based on geometry and inventory. Where airborne uranium is significant, workers are provided with protective respirators. The workers are regularly monitored for external exposure and also for internal exposure. The environmental releases from the NFC facility is well controlled. Soil, water and air from the NFC environment are routinely collected and analysed for all the possible pollutants. The paper describes the Health Physics experience during the last five years on occupational exposures and on environmental surveillance which reveals the high quality of safety observed in our nuclear fuel fabricating installations. (author). 4 refs., 6 tabs

  16. Interpretation of the results from individual monitoring of workers at the Nuclear Fuel Fabrication Facility, Brazil

    International Nuclear Information System (INIS)

    Castro, Marcelo Xavier de

    2005-01-01

    In nuclear fuel fabrication facilities, workers are exposed to different compounds of enriched uranium. Although in this kind of facility the main route of intake is inhalation, ingestion may occur in some situations, and also a mixture of both. The interpretation of the bioassay data is very complex, since it is necessary taking into account all the different parameters, which is a big challenge. Due to the high cost of the individual monitoring programme for internal dose assessment in the routine monitoring programmes, usually only one type of measurement is assigned. In complex situations like the one described in this study, where several parameters can compromise the accuracy of the bioassay interpretation it is need to have a combination of techniques to evaluate the internal dose. According to ICRP 78 (1997), the general order of preference of measurement methodologies in terms of accuracy of interpretation is: body activity measurement, excreta analysis and personal air sampling. Results of monitoring of working environment may provide information that assists in the interpretation on particle size, chemical form, solubility and date of intake. A group of fifteen workers from controlled area of the studied nuclear fuel fabrication facility was selected to evaluate the internal dose using all different available techniques during a certain period. The workers were monitored for determination of uranium content in the daily urinary and faecal excretion (collected over a period of 3 consecutive days), chest counting and personal air sampling. The results have shown that at least two types of sensitivity techniques must be used, since there are some sources of uncertainties on the bioassay interpretation, like mixture of uranium compounds intake and different routes of intake. The combination of urine and faeces analysis has shown to be the more appropriate methodology for assessing internal dose in this situation. The chest counting methodology has not shown

  17. Development of advanced nuclear materials - Fabrication of Zr-Nb alloy used in PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang In; Kim, Won Baek; Lee, Chul Kyung; Choi, Kuk Sun; Kang, Dae Kyu; Seo, Chang Ryul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The following conclusions can be made from the second year research: 1. Easy control for alloying elements can be made for the following adding metals like Nb, V, Sn, Mo, Fe due to low vapor pressure. In case of Cr and Te= known to have high vapor pressure, they are controlled by adding master alloy(Zr-Cr) or quite excess of aimed composition. However, Bi was found to be very difficult to charging the certain amount into the melt. 2. Oxygen content can be adjusted by adding the Zr-10%O master alloy considering the inherent amount of oxygen in sponge zirconium. 3. The charging rod of 38 mm in diameter, 96 mm in length was made by a series of button melting, casting and vacuum welding, from this, Zr-2.5Nb ingot of 50 mm in diameter and 550 mm in length was fabricated by EB drip melting process. 4. The amount of Nb can be successfully adjusted at 2.8% with charging 15% excess. Nb as adding element is easily controlled due to high-melting -point metal and its low vapor pressure. 5. Oxygen content is not varied during remelting, casting, and drip melting, only slight change was observed in button melting stage due to uptake the desorbed gases during the melting operation. Nuclear materials in domestic nuclear power plants depend on import and this amount reaches 100 million dollars per year. The increase in demand for the development of new zirconium based alloys are expecting. All the results involving this research can be applied for the melting of reactive metals, vacuum refining and alloy design. 13 refs., 6 tabs., 10 figs., 10 ills. (author)

  18. Decontamination chamber for the maintenance of DUPIC nuclear fuel fabrication and process equipment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. H.; Park, J. J.; Yang, M. S.; Lee, H. H.; Shin, J. M

    2000-10-01

    This report presents the decontamination chamber of being capable of decontaminating and maintaining DUPIC nuclear fuel fabrication equipment contaminated in use. The decontamination chamber is a closed room in which contaminated equipment can be isolated from a hot-cell, be decontaminated and be reparired. This chamber can prevent contamination from spreading over the hot-cell, and it can also be utilized as a part of the hot-cell after maintenance work. The developed decontamination chamber has mainly five sub-modules - a horizontal module for opening and closing a ceil of the chamber, a vertical module for opening and closing a side of the chamber, a subsidiary door module for enforcing the vertical opening/closing module, a rotary module for rotating contaminated equipment, and a grasping module for holding a decontamination device. Such sub-modules were integrated and installed in the M6 hot-cell of the IMEF at the KAERI. The mechanical design considerations of each modules and the arrangement with hot-cell facility, remote operation and manipulation of the decontamination chamber are also described.

  19. Mortality and cancer incidence experience of employees in a nuclear fuels fabrication plant

    International Nuclear Information System (INIS)

    Hadjimichael, O.C.; Ostfeld, A.M.; D'Atri, D.A.; Brubaker, R.E.

    1983-01-01

    The mortality and cancer incidence experience of 4,106 employees in a nuclear fuels fabrication plant was evaluated in this retrospective cohort study. Standardized mortality (SMR) and incidence ratios were calculated for groups of employees holding different jobs in the company associated with various types of industrial exposures and with low levels of radiation. Connecticut population mortality rates and Connecticut Tumor Registry incidence rates, specific for age-sex, calendar year and cause of death or cancer site, were used for the calculation of expected rates. Results showed the SMR for all male employees to be significantly lower than expected for all causes and what would be expected for all cancer deaths. More deaths were observed than expected from diseases of the central and peripheral nervous system and from obstructive pulmonary disease. The overall cancer incidence experience of the male employees was significantly lower than expected among the industrial employees. There was no risk associated with any particular job exposure group. Log linear models analysis showed no significant effect from industrial and radiation exposures or from their combined influence

  20. Decontamination chamber for the maintenance of DUPIC nuclear fuel fabrication and process equipment

    International Nuclear Information System (INIS)

    Kim, K. H.; Park, J. J.; Yang, M. S.; Lee, H. H.; Shin, J. M.

    2000-10-01

    This report presents the decontamination chamber of being capable of decontaminating and maintaining DUPIC nuclear fuel fabrication equipment contaminated in use. The decontamination chamber is a closed room in which contaminated equipment can be isolated from a hot-cell, be decontaminated and be reparired. This chamber can prevent contamination from spreading over the hot-cell, and it can also be utilized as a part of the hot-cell after maintenance work. The developed decontamination chamber has mainly five sub-modules - a horizontal module for opening and closing a ceil of the chamber, a vertical module for opening and closing a side of the chamber, a subsidiary door module for enforcing the vertical opening/closing module, a rotary module for rotating contaminated equipment, and a grasping module for holding a decontamination device. Such sub-modules were integrated and installed in the M6 hot-cell of the IMEF at the KAERI. The mechanical design considerations of each modules and the arrangement with hot-cell facility, remote operation and manipulation of the decontamination chamber are also described

  1. Fabrication development for high-level nuclear waste containers for the tuff repository

    International Nuclear Information System (INIS)

    Domian, H.A.; Holbrook, R.L.; LaCount, D.F.; Babcock and Wilcox Co., Alliance, OH

    1990-09-01

    This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler ampersand Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs

  2. Test Operation of Oxygen-Enriched Incinerator for Wastes From Nuclear Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Kim, J.-G.; Yang, H.cC.; Park, G.-I.; Kim, I.-T.; Kim, J.-K.

    2002-01-01

    The oxygen-enriched combustion concept, which can minimize off-gas production, has been applied to the incineration of combustible uranium-containing wastes from a nuclear fuel fabrication facility. A simulation for oxygen combustion shows the off-gas production can be reduced by a factor of 6.7 theoretically, compared with conventional air combustion. The laboratory-scale oxygen enriched incineration (OEI) process with a thermal capacity of 350 MJ/h is composed of an oxygen feeding and control system, a combustion chamber, a quencher, a ceramic filter, an induced draft fan, a condenser, a stack, an off-gas recycle path, and a measurement and control system. Test burning with cleaning paper and office paper in this OEI process shows that the thermal capacity is about 320 MJ/h, 90 % of design value and the off-gas reduces by a factor of 3.5, compared with air combustion. The CO concentration for oxygen combustion is lower than that of air combustion, while the O2 concentration in off-gas is kept above 25 vol % for a simple incineration process without any grate. The NOx concentration in an off-gas stream does not reduce significantly due to air incoming by leakage, and the volume and weight reduction factors are not changed significantly, which suggests a need for an improvement in sealing

  3. The second answers and questions on the licence of the fabrication project for the nuclear fuel of research reactors

    International Nuclear Information System (INIS)

    Park, Hee Dae; Kim, C. K.; Kim, K. H.

    2002-07-01

    KINS has examined the application for licensing of research reactor fuel fabrication for seven months, from May to Dec. 2000. The most hot issues during examination, in order to understand whether the design and facilities are fitted to the regulation criteria or not, were the availability of basic ground, design criteria on safety, availability and methodology of design, seismic criteria, availability of nuclear fuel fabrication, safety related criticality, safety related the process, availability of nuclear waste management, validity of organization and procedure for radioactivity management, and the validity of both selection and analysis about predicted accident. Moreover, another issues such as the radioactivity inspection plan for waste treatment, effect on both radioactive material and accidant, method of decrease of damage on environment, and environmental inspection plan of radioactivity, were severely examined

  4. Fabrication and closure development of corrosion resistant containers for Nevada's Yucca Mountain high-level nuclear waste repository

    International Nuclear Information System (INIS)

    Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

    1989-11-01

    US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 2 figs., 4 tabs

  5. On the Generalized Correlation Equation of Welding Current for the Tig Welding Machine Used in Nuclear Fuel Fabrication

    International Nuclear Information System (INIS)

    Umar, Efrizon

    1995-01-01

    In nuclear fuel fabrication, welding plays a very important role to join the end cap to the tube. In order to determine the welding current in TIG welding process for various materials, weld geometries and welding rates, the correlation between the welding current and the other parameters are needed. This paper presents the correlation of those parameters mentioned above. The proposed correlation was tested and produced satisfactory results. (author). 8 refs., 2 tabs., 2 figs

  6. Site Specific Discrete PEGylation of 124I-Labeled mCC49 Fab′ Fragments Improves Tumor MicroPET/CT Imaging in Mice

    Science.gov (United States)

    Ding, Haiming; Carlton, Michelle M.; Povoski, Stephen P.; Milum, Keisha; Kumar, Krishan; Kothandaraman, Shankaran; Hinkle, George H.; Colcher, David; Brody, Rich; Davis, Paul D.; Pokora, Alex; Phelps, Mitchell; Martin, Edward W.; Tweedle, Michael F.

    2014-01-01

    The tumor-associated glycoprotein-72 (TAG-72) antigen is highly overexpressed in various human adenocarcinomas and anti-TAG-72 monoclonal antibodies, and fragments are therefore useful as pharmaceutical targeting vectors. In this study, we investigated the effects of site-specific PEGylation with MW 2–4 kDa discrete, branched PEGylation reagents on mCC49 Fab′ (MW 50 kDa) via in vitro TAG72 binding, and in vivo blood clearance kinetics, biodistribution, and mouse tumor microPET/CT imaging. mCC49Fab′ (Fab′-NEM) was conjugated at a hinge region cysteine with maleimide-dPEG12-(dPEG24COOH)3 acid (Mal-dPEG-A), maleimide-dPEG12-(dPEG12COOH)3 acid (Mal-dPEG-B), or maleimide-dPEG12-(m-dPEG24)3 (Mal-dPEG-C), and then radiolabeled with iodine-124 (124I) in vitro radioligand binding assays and in vivo studies used TAG-72 expressing LS174T human colon carcinoma cells and xenograft mouse tumors. Conjugation of mCC49Fab′ with Mal-dPEG-A (Fab′-A) reduced the binding affinity of the non PEGylated Fab′ by 30%; however, in vivo, Fab′-A significantly lengthened the blood retention vs Fab′-NEM (47.5 vs 28.1%/ID at 1 h, 25.1 vs 8.4%/ID at 5 h, p Fab′-NEM by 70%, blood retention, microPET/CT imaging tumor signal intensity, and residual 72 h tumor concentration by 49% (3.83 ± 1.50 vs 1.97 ± 0.29%ID/g, p < 0.05) and 63% (3.83 ± 1.50 vs 1.42 ± 0.35%ID/g, p < 0.05), respectively. We conclude that remarkably subtle changes in the structure of the PEGylation reagent can create significantly altered biologic behavior. Further study is warranted of conjugates of the triple branched, negatively charged Mal-dPEG-A. PMID:24175669

  7. New trends in design and fabrication of signal and power cables to increase nuclear safety

    International Nuclear Information System (INIS)

    Salmen, Florin; Florescu, Gheorghe; Ionescu, Aurel

    2007-01-01

    Based on NPP operating experiences in the past, it was found that the inadequate management of aging degradation caused shortening of the lifetime of equipment, which in turn, hindered plant life extension. Aging degradation of plant structures and components should be properly managed to ensure the designated safety function of plant systems during design life and extended life. From a safety perspective, aging management means maintaining the aging degradation level in major equipment and structures below the allowable limit and holding the capacity to sustain abnormal operating condition. Cable aging was not considered as a significant factor in relation to the nuclear power plant maintenance due to its long life which is almost the same as the plant design life. Attempts to extend the lifetime of NPP has become one of the major concern in the nuclear industry world wide. Consequently, life evaluation and lifetime management of cables to survive over 40 years has become major topic of discussion. In connection to this, accelerated aging must be studied in detail in order to simulate the natural aging in NPP. Test results for evaluating aging degradation after accelerated aging of polyethylene jacket will be described herein.There are hundred types of cables in NPPs. These cables can be classified as medium/low voltage cable, low power cable, instrument and control cable, panel connect line cable, special cable, security line cable, phone line cable and ground cable. Insulation and jacket material in electrical cables are fabricated of polymer materials combined with a number of additives and filler to provide the required mechanical, electrical and fire retardant proprieties. The most commonly used insulation materials are XLPE/EPR/EPDM and PVC. PVC has been widely used as an insulation material, particularly in old plants, but it is less used in modern plants. Neoprene/CSPE/PVC are commonly used material for nuclear cable jacket. The old types of cables

  8. A Computer Simulation to Assess the Nuclear Material Accountancy System of a MOX Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Portaix, C.G.; Binner, R.; John, H.

    2015-01-01

    SimMOX is a computer programme that simulates container histories as they pass through a MOX facility. It performs two parallel calculations: · the first quantifies the actual movements of material that might be expected to occur, given certain assumptions about, for instance, the accumulation of material and waste, and of their subsequent treatment; · the second quantifies the same movements on the basis of the operator's perception of the quantities involved; that is, they are based on assumptions about quantities contained in the containers. Separate skeletal Excel computer programmes are provided, which can be configured to generate further accountancy results based on these two parallel calculations. SimMOX is flexible in that it makes few assumptions about the order and operational performance of individual activities that might take place at each stage of the process. It is able to do this because its focus is on material flows, and not on the performance of individual processes. Similarly there are no pre-conceptions about the different types of containers that might be involved. At the macroscopic level, the simulation takes steady operation as its base case, i.e., the same quantity of material is deemed to enter and leave the simulated area, over any given period. Transient situations can then be superimposed onto this base scene, by simulating them as operational incidents. A general facility has been incorporated into SimMOX to enable the user to create an ''act of a play'' based on a number of operational incidents that have been built into the programme. By doing this a simulation can be constructed that predicts the way the facility would respond to any number of transient activities. This computer programme can help assess the nuclear material accountancy system of a MOX fuel fabrication facility; for instance the implications of applying NRTA (near real time accountancy). (author)

  9. Radiation doses and cause-specific mortality among workers at a nuclear materials fabrication plant

    International Nuclear Information System (INIS)

    Checkoway, H.; Pearce, N.; Crawford-Brown, D.J.; Cragle, D.L.

    1988-01-01

    A historical cohort mortality study was conducted among 6781 white male employees from a nuclear weapons materials fabrication plant for the years 1947-1979. Exposures of greatest concern are alpha and gamma radiation emanating primarily from insoluble uranium compounds. Among monitored workers, the mean cumulative alpha radiation dose to the lung was 8.21 rem, and the mean cumulative external whole body penetrating dose from gamma radiation was 0.96 rem. Relative to US white males, the cohort experienced mortality deficits from all causes combined, cardiovascular diseases, and from most site-specific cancers. Mortality excesses of lung and brain and central nervous system cancers were seen from comparisons with national and state rates. Dose-response trends were detected for lung cancer mortality with respect to cumulative alpha and gamma radiation, with the most pronounced trend occurring for gamma radiation among workers who received greater than or equal to 5 rem of alpha radiation. These trends diminished in magnitude when a 10-year latency assumption was applied. Under a zero-year latency assumption, the rate ratio for lung cancer mortality associated with joint exposure of greater than or equal to 5 versus less than 1 rem of both types of radiation is 4.60 (95% confidence limits (CL) 0.91, 23.35), while the corresponding result, assuming a 10-year latency, is 3.05 (95% CL 0.37, 24.83). While these rate ratios, which are based on three and one death, respectively, lack statistical precision, the observed dose-response trends indicate potential carcinogenic effects to the lung of relatively low-dose radiation. There are no dose-response trends for mortality from brain and central nervous system cancers

  10. High Resolution Magic Angle Spinning Nuclear Magnetic Resonance (HRMAS NMR) for Studies of Reactive Fabrics

    Science.gov (United States)

    2015-11-01

    spectroscopy (NMR) Self- decontaminating fabric Reactive fabric...reactions of reagents including chemical weapons on materials like concrete, soil , and sand, as well as reactive polymers.3,4,5,6,7 There are...sample. The rotor and cap can be cleaned by rinsing with solvent or decontamination solution and reused. 12.0 DATA ANALYSIS AND CALCULATIONS 12.1

  11. The data acquisition system for the management of nuclear materials involved in the fabrication of MOX fuel at the Cogema plant in Cadarache

    International Nuclear Information System (INIS)

    Crousilles, M.; Beche, M.; Dalverny, G.

    2001-01-01

    This article presents the follow-up system of all the nuclear materials that are involved in the industrial process of MOX fuel fabrication. This system, called Concerto, allows the management of MOX fabrication but also of any nuclear material transfer and of the stockpile of nuclear materials with taking into account their own specificity such as the risk of criticality. Operators that intervene on the different steps of the fabrication process, supply Concerto with information so Concerto can be considered as a near real-time system providing and recording the localization, the composition, the weight, the container,... of any batch of nuclear materials. Concerto complies with the requirements of quality assurance but also of nuclear safety by forbidding any transfer whenever the maximal authorized quantity would be exceeded. (A.C.)

  12. Improvements in the consistency of fabrication and the reliability of nuclear fuels through quality assurance

    International Nuclear Information System (INIS)

    Sifferlen, R.

    1976-01-01

    By establishing correlations between rejection level and fabrication parameters, quality assurance can guide corrective action for improving the consistency of fabrication and the reliability of fuel elements. The author cites examples relating to the quality of the uranium in metallic fuels, the influence of the parent metal on the welding of zirconium alloys, the behaviour of the springs inside the cladding during the welding of plugs and the behaviour of uranium oxide pellets. (author)

  13. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    International Nuclear Information System (INIS)

    Vasudevamurthy, G.; Radecka, A.; Massey, C.

    2015-01-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  14. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

    2015-07-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  15. Development of uranium reduction system for incineration residue generated at LWR nuclear fuel fabrication plants in Japan

    International Nuclear Information System (INIS)

    Sampei, T.; Sato, T.; Suzuki, N.; Kai, H.; Hirata, Y.

    1993-01-01

    The major portion of combustible solid wastes generated at LWR nuclear fuel fabrication plants in Japan is incinerated and stored in a warehouse. The uranium content in the incineration residue is higher compared with other categories of wastes, although only a small amount of incineration residue is generated. Hence, in the future uranium should be removed from incineration residues before they are reduced to a level appropriate for the final disposal. A system for processing the incineration residue for uranium removal has been developed and tested based on the information obtained through laboratory experiments and engineering scale tests

  16. Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Heestand, R. L.; Kizer, D. E.

    1973-01-01

    The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

  17. Development of Hi-Tech ceramics fabrication technologies - Development of advanced nuclear materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Thae Kap; Park, Ji Youn; Kim, Sun Jae; Kim, Kyong Ho; Jung, Choong Hwan; Oh, Seok Jin [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-15

    The objective of the present work is to prepare the foundation of hi-tech ceramics fabrication technologies through developing important processes i.e., tape casting, sol-gel, single crystal growing, compacting and sintering, and grinding and machining processes. Tape casting process is essential to manufacture hard and functional thin plates and structural elements for some composite materials. For the fabrication of spherical mono-sized micropowders of oxides, sol-gel process has widely been used. Piezoelectric elements that are the core parts of the sensors of LPMS (loose part monitoring system) and ALMS (acoustic leakage monitoring system) are used in single crystal forms. Compacting and sintering processes are general methods for fabricating structural parts using powders. Grinding and machining processes are important to achieve the final dimensions and surface properties of the parts. (Author).

  18. Production of 68Ge, 64Cu, 86Y, 89Zr, 73Se, 77Br and 124I positron emitting radionuclides through future laser-accelerated proton beams at ELI-Beamlines for innovative PET diagnostics

    OpenAIRE

    Italiano, Antonio; Amato, Ernesto; Minutoli, Fabio; Margarone, Daniele; Baldari, Sergio

    2016-01-01

    The development of innovative production pathways for high-Z positron emitters is of great interest to enlarge the applicability of PET diagnostics, especially in view of the continuous development of new radiopharmaceuticals. We evaluated the theoretical yields of 64Cu, 86Y, 89Zr, 73Se, 77Br and 124I PET isotopes, plus the 68Ge isotope, parent of the 68Ga positron emitter, in the hypothesis of production through laser-accelerated proton sources expected at the ELI-Beamlines facility. By mean...

  19. Some problems on domestic technology development from a point of fabricator of nuclear power plant. [Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watamori, T [Hitachi Ltd., Tokyo (Japan)

    1976-06-01

    During past 20 years, the nuclear power industry in Japan has introduced foreign technology, digested it in a short period, and continued to research and develop domestic technology. Now, 95% of the machinery and equipments for nuclear power generation with light water reactors can be produced domestically, and some technologies are going to be exported. However, the nuclear power industry is still in a severe environment. The progress of the development of nuclear power plants passed the periods of organizational preparation, the construction of research reactors, the import of foreign technologies and reactors for practical use, and the construction of domestically produced reactors for practical use. The supplying capacity of the nuclear power industry in Japan reached 6 units of 1,000 MW yearly, but in order to meet the long term plan of nuclear power generation, this capacity must be further enhanced. The problems in the promotion of domestic production are the establishment of independent technologies, the promotion of standardization, the strengthening of business basis, the upbringing of relating enterprises, and the acceleration of national projects. Since the energy crisis, the trend of filling up energy demand with nuclear power generation became conspicuous, but for the expansion of export, the problems of safety guarantee, nuclear fuel cycle, and financial measures must be solved with government aid.

  20. Some problems on domestic technology development from a point of fabricator of nuclear power plant

    International Nuclear Information System (INIS)

    Watamori, Tsutomu

    1976-01-01

    During past 20 years, the nuclear power industry in Japan has introduced foreign technology, digested them in short period, and continued to research and develop domestic technology. Now, 95% of the machinery and equipments for nuclear power generation with light water reactors can be produced domestically, and some technologies are going to be exported. However, the nuclear power industry is still in severe environment. The progress of the development of nuclear power plants passed the periods of organizational preparation, the construction of research reactors, the import of foreign technologies and reactors for practical use, and the construction of domestically produced reactors for practical use. The supplying capacity of the nuclear power industry in Japan reached 6 units of 1,000 MW yearly, but in order to meet the long term plan of nuclear power generation, this capacity must be further enhanced. The problems in the promotion of domestic production are the establishment of independent technologies, the promotion of standardization, the strengthening of business basis, the upbringing of relating enterprises, and the acceleration of national projects. Since the energy crisis, the trend of filling up energy demand with nuclear power generation became conspicuous, but for the expansion of export, the problems of safety guarantee, nuclear fuel cycle, and financial measures must be solved with government aid. (Kako, I.)

  1. Quality control in nuclear fuel fabrication on the inspection basis; Control de calidad para fabricacion de combustible nuclear en base a inspecciones

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes S, A. [Instituto Tecnologico de Toluca, Toluca (Mexico)

    1997-12-31

    Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B `Quality assurement for nuclear plants` specially in the criteria `Inspections` that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)

  2. Quality control in nuclear fuel fabrication on the inspection basis; Control de calidad para fabricacion de combustible nuclear en base a inspecciones

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes S, A [Instituto Tecnologico de Toluca, Toluca (Mexico)

    1998-12-31

    Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B `Quality assurement for nuclear plants` specially in the criteria `Inspections` that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)

  3. Feasibility study for the implementation of NRTMA system for an industrial nuclear fuel fabrication plant

    International Nuclear Information System (INIS)

    Aparo, M.; Dionisi, M.; Graziani, M.; Remetti, R.

    1989-01-01

    In the frame of the problems arising from the fissile materials safeguards into the facilities of the nuclear fuel cycle, the International Safeguards devoted, in the recent years, R and D efforts on a new Dynamic Accountability procedures (Near Real Time Material Accountancy) appealing to the needs of timeliness in detecting diversion. This paper deals with a feasibility study of a NRTMA system to be applied to a nuclear fuel fabbrication plant for light water reactor. Such a feasibility study was performed by developing a dynamic model and a computer program, written in FORTRAN 77, in order to simulate all the processes and measurement procedures involved in the nuclear material accountancy

  4. Fabrication and Testing of Nuclear-Thermal Propulsion Ground Test Hardware, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Efficient nuclear-thermal propulsion requires heating a low molecular weight gas, typically hydrogen, to high temperature and expelling it through a nozzle. The...

  5. Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P.

    1999-08-01

    This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs

  6. Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P

    1999-08-01

    This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs.

  7. Romanian nuclear fuel fabrication and in-reactor fuel operational experience

    International Nuclear Information System (INIS)

    Budan, O.

    2003-01-01

    A review of the Romanian nuclear program since mid 60's is made. After 1990, the new Romanian nuclear power authority, RENEL-GEN, elaborated a realistic Nuclear Fuel Program. This program went through the Romanian nuclear fuel plant qualification with the Canadian (AECL and ZPI) support, restarting in January 1995 of the industrial nuclear fuel production, quality evaluation of the fuel produced before 1990 and the recovery of this fuel. This new policy produced good results. FCN is since 1995 the only CANDU fuel supplier from outside Canada recognised by AECL as an authorised CANDU fuel manufacturer. The in-reactor performances and behaviour of the fuel manufactured by FCN after its qualification have been excellent. Very low - more then five times lesser than the design value - fuel defect rate has been recorded up to now and the average discharge of this fuel was with about 9% greater than the design value. Since mid 1998 when SNN took charge of the production of nuclear generated electricity, FCN made significant progresses in development and procurement of new and more efficient equipment and is now very close to double its fuel production capacity. After the completion of the recovery of the fuel produced before June 1990, FCN is already prepared to shift its fuel production to the so-called 'heavy' bundle containing about 19.3 kg of Uranium per bundle

  8. Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

    International Nuclear Information System (INIS)

    1978-01-01

    This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser

  9. Structural Component Fabrication and Characterization of Advanced Radiation Resistant ODS Steel for Next Generation Nuclear Systems

    International Nuclear Information System (INIS)

    Noh, Sang Hoon; Kim, Young Chun; Jin, Hyun Ju; Choi, Byoung Kwon; Kang, Suk Hoon; Kim, Tae Kyu

    2016-01-01

    In a sodium-cooled fast reactor (SFR), the coolant outlet temperature and peak temperature of the fuel cladding tube will be about 545 .deg. C and 700 .deg. C with 250 dpa of a very high neutron dose rate. To realize this system, it is necessary to develop an advanced structural material having high creep and irradiation resistance at high temperatures. Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling severely occurred to 120dpa at high temperatures and this eventually leads to a decrease of the mechanical properties and dimensional stability. Advanced radiation resistant ODS steel (ARROS) has been newly developed for the in-core structural components in SFR, which has very attractive microstructures to achieve both superior creep and radiation resistances at high temperatures [4]. Nevertheless, the use of ARROS as a structural material essentially requires the fabrication technology development for component parts such as sheet, plate and tube. In this study, plates and tubes were tentatively fabricated with a newly developed alloy, ARROS. Microstructures as well as mechanical properties were also investigated to determine the optimized condition of the fabrication processes.

  10. Computerized information system for inventory-taking and verification at a nuclear fuel fabrication plant with closed production lines

    International Nuclear Information System (INIS)

    Bahm, W.; Brueckner, C.; Hartmann, G.

    1976-01-01

    By means of a model the use of electronic data processing is studied for preparing inventory listings and for inventory verification in a fabrication plant for Pu-U mixed-oxide fuel pins. It is postulated that interruptions in operation should be avoided as much as possible. Closed-Line production is assumed so that access to nuclear material calls for special withdrawal via locks. The production line is subdivided into sections with measuring points placed in between to record the nuclear material flow. The measured results are fed to a central data acquisition and reporting system capable of calculating on-line from these results the book inventories present in the individual sections. Inventory-taking and verification are carried out simultaneously in the sections of the production line using the EDP system. The production is not interrupted for this purpose. The production stream is tagged prior to reaching a section to be measured and is subsequently measured when entering the respective section until the tag has reached the end of the section. The measurement can be verified by inspectors. Movements of nuclear materials in and from other plant areas such as the storage area are likewise fed into the central data processing system so that inventory lists can be recalled at any moment. By this means the inventory can be taken quickly and at any time. The inventory is verified in the conventional way. (author)

  11. Some fabrication problems in nuclear power plants heat exchanges, its detectability and implications

    International Nuclear Information System (INIS)

    Condessa, N.C.; Oliveira, R.

    1988-01-01

    On the design and manufacturing follow-up of heat-exchangers of nuclear power plants some care are took into account in order to assure a high degree of confiability allowing the heat-exchanger in operation under severe and aggressive conditions be operating during the useful life of the nuclear power plant. However, despite the care, some problems can ocurr as the ones described on this job; that, if not detected in due time could bring umpleasant problems to the component or to the system in which it is working during operation. (author) [pt

  12. Economic analysis to compare fabrication of nuclear power and fossil fuel power plants at Iran

    International Nuclear Information System (INIS)

    Rasouliye Koohi, Mojtaba

    1997-01-01

    Electric power due to its many advantages over other forms of energies covers most of the world's energy demands.The electric power can be produced by various energy converting systems fed by different energy resources like fossil fuels, nuclear, hydro and renewable energies, each having their own appropriate technologies. The fossil fuel not only consumes the deplete and precious sources of non conventional energies but they add pollution to environment too. The nuclear power plants has its own share of radioactive pollutions which, of course can be controlled by taking precautionary measures. The investing cost of each generated unit (KWh) in the nuclear power plants, comparing with its equivalent production by fossil fuels is investigated. The various issues of economical analysis, technical, political and environmental are the different aspects, which individually can influence the decisions for kind of power plant to be installed. Finally, it is concluded that the fossil and nuclear power generations both has its own advantages and disadvantages. Hence, from a specializing point of view, it may not be proper to prefer one over the others

  13. Development of Advanced Technologies to Reduce Design, Fabrication and Construction Costs for Future Nuclear Power Plants

    International Nuclear Information System (INIS)

    DiNunzio, Camillo A.; Gupta, Abhinav; Golay, Michael; Luk, Vincent; Turk, Rich; Morrow, Charles; Geum-Taek Jin

    2002-01-01

    OAK-B135 This report presents a summation of the third and final year of a three-year investigation into methods and technologies for substantially reducing the capital costs and total schedule for future nuclear plants. In addition, this is the final technical report for the three-year period of studies

  14. Modern requirements to quality assurance and control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Weidinger, H.G.

    1999-01-01

    This lecture have shown a new type of quality assurance management has already successfully introduced in various industries and now starts to be used increasingly in the nuclear fuel industry. Static authority regulations and a tendency to bureaucratic understanding and handling of these regulations lead to a delayed start and a relatively slow progress of these quality strategies in the nuclear fuel technology. However, the economic pressure of strong competition and increasing demands of the utilities as the user of nuclear fuel result in a more determined introduction also to this area. The different use of statistical methods of two different fuel vendors are shown. Vendor A uses old fashioned methods. The focus is on the expensive final product control and few emphasis is on design of experiments and process control. Consequently, this vendor will have high costs, not only for QC and rejection but also for repair and replace actions after delivery. To the contrary, vendor B invests primarily in the design of experiments and process control. This vendor will profit only from lower direct costs but also from being at the front line of technical development and from enjoying a satisfied and happy customer. Many well examined quality management tools are available today which help not only to improve the quality but also decrease the costs. Still, the progress in using these techniques in nuclear fuel technology is limited and not comparable to the progress in other industries like automobile production or the electronic industry. (author)

  15. Development of Advanced Technologies to Reduce Design, Fabrication and Construction Costs for Future Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    DiNunzio, Camillo A. [Framatome ANP DE& S, Marlborough, MA (United States); Gupta, Abhinav [Univ. of North Carolina, Raleigh, NC (United States); Golay, Michael [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Luk, Vincent [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Turk, Rich [Westinghouse Electric Company Nuclear Systems, Windsor, CT (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jin, Geum-Taek [Korea Power Engineering Company Inc., Yongin-si, Gyeonggi-do (Korea, Republic of)

    2002-11-30

    This report presents a summation of the third and final year of a three-year investigation into methods and technologies for substantially reducing the capital costs and total schedule for future nuclear plants. In addition, this is the final technical report for the three-year period of studies.

  16. INMACS - An approach to on-line nuclear materials accounting and control in a fuel fabrication environment

    International Nuclear Information System (INIS)

    Yan, G.; L'Archeveque, J.V.R.; Paul, R.N.

    1977-08-01

    Taking advantage of modern system technologies, the concept of an Integrated Nuclear Materials Accounting and Control System (INMACS) was formulated as an alternative solution to manual inventory procedures. The selected approach offers prospects for tackling the more general fissile materials inventory problem while satisfying the immediate requirements of the Fuel Fabrication Pilot Line at CRNL. A PDP-11/40 minicomputer system was purchased, and a Data Base Management System (DBMS) was designed and implemented to provide a uniform file handling capability. The specific requirements of the Pilot Line were met by a package of application programs. About 16 man-years have been spent on the project. INMACS has been installed in the field and its usefulness as an on-line inventory system will be demonstrated in the Pilot Line. (author)

  17. The future supply of and demand for candidate materials for the fabrication of nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    Grover, L.K.

    1990-01-01

    This report summarizes the findings of a literature survey carried out to assess the future world supply of and demand for titanium, copper and lead. These metals are candidate materials for the fabrication of containers for the immobilization and disposal of Canada's nuclear used-fuel waste for a reference Used-fuel Disposal Centre. Such a facility may begin operation by approximately 2020, and continue for about 40 years. The survey shows that the world has abundant supplies of titanium minerals (mostly in the form of ilmenite), which are expected to last up to at least 2110. However, for copper and lead the balance between supply and demand may warrant increased monitoring beyond the year 2000. A number of factors that can influence future supply and demand are discussed in the report

  18. Fabrication of nuclear ship reactor MRX model and study on inspection and maintenance of components

    International Nuclear Information System (INIS)

    Kasahara, Yoshiyuki; Nakazawa, Toshio; Kusunoki, Tsuyoshi; Takahashi, Hiroki; Yoritsune, Tsutomu.

    1997-10-01

    The MRX (Marine Reactor X) is an integral type small reactor adopting passive safety systems. As for an integral type reactor, primary system components are installed in the reactor vessel. It is therefor important to establish the appropriate procedure for construction, inspection and maintenance, dismauntling, etc., for all components in the reactor vessel as well as in the reactor containment, because inspection space is limited. To study these subjects, a one-fifth model of the MRX was fabricated and operation capabilities were studied. As a result of studies, the following results are obtained. (1) Manufacturing and installing problems of the reactor pressure vessel, the containment vessel and internal components are basically not abserved. (2) Heat transfer tube structures of the steam generator and the heat exchangers of emergency decay heat removal system and containment water cooler were not seen of any problem for fabrication. However, due consideration is required in the detailed design of supports of heat transfer tubes. (3) Further studies should be needed for designs of flange penetrations and leak countermeasures for pipes instrument cables. (4) Arrangements of equipments in the containment should be taken in consideration in detail because the space is narrow. (5) Further discussion is required for installation methods of instruments and cables. (author)

  19. Updates of the fire protection system of the Juzbado Nuclear Fuel Fabrication Plant; Actualizaciones del Sistema de Proteccion Contra Incendios de la Fabrica de Combustible Nuclear de Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    Dorado, P.; Palomo, J. J.; Romano, A.

    2015-07-01

    The Juzbado Nuclear Fuel Fabrication Plant fire protection system is one of the most important safety system of the plant. Every year, a large part of the annual investment is employed to improve this system, to update its technology, in order to improve detection and extinction capability to minimize fire risk. Over the last few years, several improvement projects have been carried out that focused on fire detection technology update and on optimization of local detectors integration with a centralized control system, as well as on an advanced public address system, which used clear and unambiguous messages improving personnel response to a plant evacuation. Planned projects and those, which are currently under development, focus on improving passive fire protection means as well as fire protection of key emergency response equipment s such as emergency diesel generators and fire extinguishing bombs. (Author)

  20. The influence of saliva flow stimulation on the absorbed radiation dose to the salivary glands during radioiodine therapy of thyroid cancer using {sup 124}I PET(/CT) imaging

    Energy Technology Data Exchange (ETDEWEB)

    Jentzen, Walter; Schmitz, Jochen; Freudenberg, Lutz; Eising, Ernst; Bockisch, Andreas; Stahl, Alexander [Universitaet Duisburg-Essen, Klinik fuer Nuklearmedizin, Essen (Germany); Balschuweit, Dorothee; Hilbel, Thomas [Fachhochschule Gelsenkirchen, Fachbereich Physikalische Technik, Gelsenkirchen (Germany)

    2010-12-15

    A serious side effect of high-activity radioiodine therapy in the treatment of differentiated thyroid cancer is radiogenic salivary gland damage. This damage may be diminished by lemon-juice-induced saliva flow immediately after {sup 131}I administration. The aim of this study was to assess the effect of chewing lemon slices on the absorbed (radiation) doses to the salivary glands. Ten patients received (pretherapy) {sup 124}I PET(/CT) dosimetry before their first radioiodine therapy. The patients underwent a series of six PET scans at 0.5, 1, 2, 4, 48 and {>=}96 h and one PET/CT scan at 24 h after administration of 27 MBq {sup 124}I. Blood samples were also collected at about 2, 4, 24, 48, and 96 h. Contrary to the standard radioiodine therapy protocol, the patients were not stimulated with lemon juice. Specifically, the patients chewed no lemon slices during the pretherapy procedure and neither ate food nor drank fluids until after completion of the last PET scan on the first day. Organ absorbed doses per administered {sup 131}I activity (ODpAs) as well as gland and blood uptake curves were determined and compared with published data from a control patient group, i.e. stimulated per the standard radioiodine therapy protocol. The calculations for both groups used the same methodology. A within-group comparison showed that the mean ODpA for the submandibular glands was not significantly different from that for the parotid glands. An intergroup comparison showed that the mean ODpA in the nonstimulation group averaged over both gland types was reduced by 28% compared to the mean ODpA in the stimulation group (p=0.01). Within each gland type, the mean ODpA reductions in the nonstimulation group were statistically significant for the parotid glands (p=0.03) but not for the submandibular glands (p=0.23). The observed ODpAs were higher in the stimulation group because of increased initial gland uptake rather than group differences in blood kinetics. The {sup 124}I PET

  1. Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands

    Energy Technology Data Exchange (ETDEWEB)

    Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)

    2015-06-01

    The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.

  2. On-line item control at a high enriched nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Lewis, T.W.; Lewis, H.M.

    1984-01-01

    The on-line item control system at Nuclear Fuel Services, Inc., is a near-real time method capable of tracking uniquely identified items from creation through disposition. The system provides for improved control, timeliness, accuracy and usability of company information and the necessary data required to support the regulatory program for the protection against diversion of Special Nuclear Materials. The system consists of software applications (approximately 150 programs) with man/machine interface controls which provide facilities for correct data entry and for the protection of data integrity. This system went into stand-alone operation in September, 1983 after a twenty month parallel test run with the previous keybatched (manual forms) item control system

  3. Nuclear

    International Nuclear Information System (INIS)

    2014-01-01

    This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)

  4. Approaching six sigma quality in nuclear fuel fabrication - an Indian perspective

    International Nuclear Information System (INIS)

    Laxminarayana, B.; Kamalesh Kumar, B.; Saratchandran, N.; Ganguly, C.

    1999-01-01

    Nuclear Fuel complex (NFC), Hyderabad, manufactures fuel and structural components for both Boiling Water Reactors (BWR) and Pressurised Heavy water (PHWR). Customer and product quality has always been assigned top priority at NFC. At present, NFC is pursuing the goal of attaining six sigma quality levels, the paper brings out the details of various steps initiated and progress made towards the same, with a special reference to end closure welds. (author)

  5. Fabrication of mechanical components and piping design for Brazilian nuclear reactors

    International Nuclear Information System (INIS)

    Deppe, L.O.

    1987-01-01

    The supply of Brazilian equipment and piping design for Angra 2 (and Angra 3 in some cases) have reached an advanced status in spite of the continuous outside difficulties which affect these nuclear power plants. The achieved quality is similar to the quality achieved in foreign countries and the nationalization program foreseen in 1975 is being largely surpassed. In this paper the actual situation is presented as well as the future perspectives. (Author) [pt

  6. The development of SiC whisker fabrication technology for nuclear applications

    International Nuclear Information System (INIS)

    Kang, Thae Khapp; Kuk, Il Hiun; Kim, Chang Kyu; Lee, Jae Chun; Lee, Ho Jin; Park, Soon Dong; Im, Gyeong Soo

    1991-02-01

    Some important experiments for whisker growth reactions, fabrication processes, and experiments for fabricarion of whisker reinforced composites have been performed. In order to investigate growth reaction of SiC whiskers, a conventional carbothermic reaction was tested. Based on the results of carbothermic process, a new process called silicothermic reaction was planned and some basic experiments were performed. Reaction characteristics of silicon monoxide, core material for SiC whisker growth in both of the reactions were investigated for basic data. Additionally, a hydrofluoric acid leaching process was tested for developing SiC whisker recovery process, and powder metallurgy process and melt sqeeze process were tried to develop aluminum-SiC whisker composites. (Author)

  7. Evaluation of environmental control technologies for commercial uranium nuclear fuel fabrication facilities

    International Nuclear Information System (INIS)

    Perkins, B.L.

    1983-01-01

    At present in the United States, there are seven commercial light-water reactor uranium fuel fabrication facilities. Effluent wastes from these facilities include uranium, nitrogen, fluorine, and organic-containing compounds. These effluents may be either discharged to the ambient environment, treated and recycled internally, stored or disposed of on-site, sent off-site for treatment and/or recovery, or sent off-site for disposal (including disposal in low-level waste burial sites). Quantities of waste generated and treatment techniques vary greatly depending on the facility and circuits used internally at the facility, though in general all the fluorine entering the facility as UF 6 is discharged as waste. Further studies to determine techniques and procedures that might minimize dose (ALARA) and to give data on possible long-term effects of effluent discharge and waste disposal are needed

  8. Special equipment for the fabrication and quality control of nuclear fuel elements

    International Nuclear Information System (INIS)

    Guse, K.; Herbert, W.; Jaeger, K.

    1989-01-01

    For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG) [de

  9. Integrated software package for nuclear material safeguards in a MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Schreiber, H.J.; Piana, M.; Moussalli, G.; Saukkonen, H.

    2000-01-01

    Since computerized data processing was introduced to Safeguards at large bulk handling facilities, a large number of individual software applications have been developed for nuclear material Safeguards implementation. Facility inventory and flow data are provided in computerized format for performing stratification, sample size calculation and selection of samples for destructive and non-destructive assay. Data is collected from nuclear measurement systems running in attended, unattended mode and more recently from remote monitoring systems controlled. Data sets from various sources have to be evaluated for Safeguards purposes, such as raw data, processed data and conclusions drawn from data evaluation results. They are reported in computerized format at the International Atomic Energy Agency headquarters and feedback from the Agency's mainframe computer system is used to prepare and support Safeguards inspection activities. The integration of all such data originating from various sources cannot be ensured without the existence of a common data format and a database system. This paper describes the fundamental relations between data streams, individual data processing tools, data evaluation results and requirements for an integrated software solution to facilitate nuclear material Safeguards at a bulk handling facility. The paper also explains the basis for designing a software package to manage data streams from various data sources and for incorporating diverse data processing tools that until now have been used independently from each other and under different computer operating systems. (author)

  10. Fabrication and characterization of nuclear localization signal-conjugated glycol chitosan micelles for improving the nuclear delivery of doxorubicin

    Directory of Open Access Journals (Sweden)

    Zhao J

    2012-09-01

    Full Text Available Jingmou Yu,1 Xin Xie,1 Meirong Zheng,1 Ling Yu,2 Lei Zhang,1 Jianguo Zhao,1 Dengzhao Jiang,1 Xiangxin Che11Key Laboratory of Systems Biology Medicine of Jiangxi Province, College of Basic Medical Science, Jiujiang University, Jiujiang, 2Division of Nursing, 2nd Affiliated Hospital, Yichun University, Yichun, People's Republic of ChinaBackground: Supramolecular micelles as drug-delivery vehicles are generally unable to enter the nucleus of nondividing cells. In the work reported here, nuclear localization signal (NLS-modified polymeric micelles were studied with the aim of improving nuclear drug delivery.Methods: In this research, cholesterol-modified glycol chitosan (CHGC was synthesized. NLS-conjugated CHGC (NCHGC was synthesized and characterized using proton nuclear magnetic resonance spectroscopy, dynamic light scattering, and fluorescence spectroscopy. Doxorubicin (DOX, an anticancer drug with an intracellular site of action in the nucleus, was chosen as a model drug. DOX-loaded micelles were prepared by an emulsion/solvent evaporation method. The cellular uptake of different DOX formulations was analyzed by flow cytometry and confocal laser scanning microscopy. The cytotoxicity of blank micelles, free DOX, and DOX-loaded micelles in vitro was investigated by 3-(4,5-dimethylthiazol-2-yl-2,5-diphenyltetrazolium bromide (MTT assay in HeLa and HepG2 cells.Results: The degree of substitution was 5.9 cholesterol and 3.8 NLS groups per 100 sugar residues of the NCHGC conjugate. The critical aggregation concentration of the NCHGC micelles in aqueous solution was 0.0209 mg/mL. The DOX-loaded NCHGC (DNCHGC micelles were observed as being almost spherical in shape under transmission electron microscopy, and the size was determined as 248 nm by dynamic light scattering. The DOX-loading content of the DNCHGC micelles was 10.1%. The DOX-loaded micelles showed slow drug-release behavior within 72 hours in vitro. The DNCHGC micelles exhibited greater

  11. Prognostic impact of incomplete surgical clearance of radioiodine sensitive local lymph node metastases diagnosed by post-operative {sup 124}I-NaI-PET/CT in patients with papillary thyroid cancer

    Energy Technology Data Exchange (ETDEWEB)

    Sabet, Amir; Binse, Ina; Grafe, Hong; Goerges, Rainer; Poeppel, Thorsten D.; Bockisch, Andreas; Rosenbaum-Krumme, Sandra J. [University Duisburg-Essen, Department of Nuclear Medicine, Essen (Germany); Ezziddin, Samer [Saarland University, Department of Nuclear Medicine, Homburg (Germany)

    2016-10-15

    Nodal involvement is an independent risk factor of recurrence in papillary thyroid cancer (PTC). Neither the international guidelines nor the recently introduced ongoing risk adaptation concept consider the extent of initial surgical clearance of radioiodine sensitive lymph node metastases in their stratification systems. We investigated the prognostic relevance of incomplete initial surgical clearance in patients with purely lymphogeneous metastatic PTC (pN1 M0) despite successful radioiodine therapy. Accurate assessment of pre-ablative nodal status was attempted using PET/CT studies with both {sup 124}I-NaI and {sup 18}F-FDG along with high-resolution cervical ultrasound. Sixty-five patients with histologically diagnosed lymph node metastases (pN1 M0) were retrospectively analyzed. Patients with iodine-negative lymph node metastases diagnosed by {sup 18}F-FDG PET/CT or distant metastases were excluded from the analysis. The association of disease recurrence with the pre-ablative nodal status, as well as other baseline characteristics, were examined applying nonparametric tests for independent samples and multiple regression analysis. Patients with persistent lymph node metastases in {sup 124}I-NaI PET/CT were further divided according to the additional presence or absence of FDG-uptake in {sup 18}F-FDG PET/CT. Survival analyses were performed using Kaplan-Meier curves and the Cox proportional hazards model for uni- and multivariate analyses to assess the influence of prognostic factors on progression free survival (PFS). Incomplete metastatic lymph node resection captured by {sup 124}I-NaI PET/CT (n = 33) was an independent risk factor for recurrence (61 % vs 25 %, p = 0.006) and shorter PFS (46 months vs not reached, HR 4.0 [95 %-CI, 1.7-9.2], p = 0.001). Ultrasound could detect lymph node metastases only in 19/33 patients (58 %). Among patients with positive nodal status, FDG-avidity of metastatic iodine positive lymph nodes worsened the outcome (16 vs 69

  12. Stainless steel fabrication for high quality requirements in the nuclear industry

    International Nuclear Information System (INIS)

    Wareing, A.J.

    1990-01-01

    In this paper the author explains the welding procedures and practices adopted within the nuclear industry to achieve the high quality and standards of welds required. The changeover to mechanised welding, orbital TIG welding and synergic MIG welding, has resulted in consistent achievement of high quality standards as well as optimising the productivity. However, the use of mechanised welding machines does require the welder operating them to be fully trained and qualified. The formally organised training courses are described and the cost savings and production rates achieved by utilising the mechanised method are discussed. (author)

  13. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

    2018-01-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

  14. Verification of nuclear material balances: General theory and application to a highly enriched uranium fabrication plant

    International Nuclear Information System (INIS)

    Avenhaus, R.; Beedgen, R.; Neu, H.

    1980-08-01

    In the theoretical part it is shown that under the assumption, that in case of diversion the operator falsifies all data by a class specific amount, it is optimal in the sense of the probability of detection to use the difference MUF-D as the test statistics. However, as there are arguments for keeping the two tests separately, and furthermore, as it is not clear that the combined test statistics is optimal for any diversion strategy, the overall guaranteed probability of detection for the bivariate test is determined. A numerical example is given applying the theoretical part. Using the material balance data of a Highly Enriched Uranium fabrication plant the variances of MUF, D (no diversion) and MUF-D are calculated with the help of the standard deviations of operator and inspector measurements. The two inventories of the material balance are stratified. The samples sizes of the strata and the total inspection effort for data verification are determined by game theoretical methods (attribute sampling). On the basis of these results the overall detection probability of the combined system (data verification and material accountancy) is determined both for the MUF-D test and the bivariate (D,MUF) test as a function of the goal quantity. The results of both tests are evaluated for different diversion strategies. (orig./HP) [de

  15. Transference of know-how for the fabrication of heavy components for nuclear power reactors

    International Nuclear Information System (INIS)

    Meier, F.

    1977-01-01

    1) Heavy components for nuclear power reactors. Reactor pressure vessels with total weight of 540 tons; steam generators: heat exchangers with U-type tube bundles, total weight 420 tons. 2) Choice of know-how recipient. Technical criteria, i.e. manufacturing facilities, existing quality assurance system, location of the workshops, possibilities for training, infrastructures. 3. Measures for transferring know-how to a newly established company. Planning and erection of the factory: organisational set up of the company; personnel selection and training; transfer of documentation; transfer of know-how that cannot be transferred in a written form. 4) Contracts for assuring the transfer of know-how. Stipulation of mutual rights and obligations of the know-how owner and receiver in individual contracts: engineering services contract, technical information contract, personnel training contract, license contract. (orig.) [de

  16. Development of maintenance equipment for nuclear material fabrication equipment in a highly active hot cell

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. J.; Yang, M. S.; Kim, K. H. and others

    2000-09-01

    This report presents the development of a maintenance system for a highly contaminated nuclear material handling equipment at a hot-cell. This maintenance system has mainly three subsystems - a gamma-radiation measurement module for detecting a gamma-radiation level and identifying its distribution in-situ, a dry-type decontamination device for cleaning up contaminated particles, and a maintenance chamber for isolating contaminated equipment. The mechanical design considerations, controller, capabilities and remote operation and manipulation of the maintenance system are described. Such subsystems developed were installed and tested in the IMEF (Irradiated Material Examination Facility) M6 hot-cell after mock-up tests and performed their specific tasks successfully.

  17. Development of maintenance equipment for nuclear material fabrication equipment in a highly active hot cell

    International Nuclear Information System (INIS)

    Park, J. J.; Yang, M. S.; Kim, K. H. and others

    2000-09-01

    This report presents the development of a maintenance system for a highly contaminated nuclear material handling equipment at a hot-cell. This maintenance system has mainly three subsystems - a gamma-radiation measurement module for detecting a gamma-radiation level and identifying its distribution in-situ, a dry-type decontamination device for cleaning up contaminated particles, and a maintenance chamber for isolating contaminated equipment. The mechanical design considerations, controller, capabilities and remote operation and manipulation of the maintenance system are described. Such subsystems developed were installed and tested in the IMEF (Irradiated Material Examination Facility) M6 hot-cell after mock-up tests and performed their specific tasks successfully

  18. Fabrication of advanced targets for laser driven nuclear fusion reactions through standard microelectronics technology approaches.

    Czech Academy of Sciences Publication Activity Database

    Picciotto, A.; Crivellari, M.; Bellutti, P.; Barozzi, M.; Kucharik, M.; Krása, Josef; Swidlovsky, A.; Malinowska, A.; Velyhan, Andriy; Ullschmied, Jiří; Margarone, Daniele

    2017-01-01

    Roč. 12, October (2017), č. článku P10001. ISSN 1748-0221 Grant - others:OP VK 2 LaserGen(XE) CZ.1.07/2.3.00/20.0087; LaserZdroj (OP VK 3)(XE) CZ.1.07/2.3.00/20.0279 Institutional support: RVO:61389021 ; RVO:68378271 Keywords : Nuclear instruments and methods for hot plasma diagnostics * Plasma generation (laserproduced, RF, x ray-produced) * Plasma diagnostics - charged-particle spectroscopy Subject RIV: BL - Plasma and Gas Discharge Physics; BL - Plasma and Gas Discharge Physics (FZU-D) OBOR OECD: 2.11 Other engineering and technologies; 2.11 Other engineering and technologies (FZU-D) Impact factor: 1.220, year: 2016 http://iopscience.iop.org/article/10.1088/1748-0221/12/10/P10001/meta

  19. Fabrication of 121Sb isotopic targets for the study of nuclear high spin features

    Science.gov (United States)

    Devi, K. Rojeeta; Kumar, Suresh; Kumar, Neeraj; Abhilash, S. R.; Kabiraj, D.

    2018-06-01

    Isotopic 121Sb targets with 197Au backing have been prepared by Physical Vapor Deposition (PVD) method using the diffusion pump based coating unit at target laboratory, Inter University Accelerator Centre (IUAC), New Delhi, India. The target thickness was measured by stylus profilo-meter and the purity of the targets was investigated by Energy Dispersive X-ray Analysis (EDXA). One of these targets has been used in an experiment which was performed at IUAC for nuclear structure study through fusion evaporation reaction. The excitation function of the 121Sb(12C, yxnγ) reaction has been performed for energies 58 to 70 MeV in steps of 4 MeV. The experimental results were compared with the calculations of statistical models : PACE4 and CASCADE. The methods adopted to achieve best quality foils and good deposition efficiency are reported in this paper.

  20. Application of Self-Propagating High Temperature Synthesis to the Fabrication of Actinide Bearing Nitride and Other Ceramic Nuclear Fuels

    International Nuclear Information System (INIS)

    Moore, John J.; Reigel, Marissa M.; Donohoue, Collin D.

    2009-01-01

    The project uses an exothermic combustion synthesis reaction, termed self-propagating high-temperature synthesis (SHS), to produce high quality, reproducible nitride fuels and other ceramic type nuclear fuels (cercers and cermets, etc.) in conjunction with the fabrication of transmutation fuels. The major research objective of the project is determining the fundamental SHS processing parameters by first using manganese as a surrogate for americium to produce dense Zr-Mn-N ceramic compounds. These fundamental principles will then be transferred to the production of dense Zr-Am-N ceramic materials. A further research objective in the research program is generating fundamental SHS processing data to the synthesis of (i) Pu-Am-Zr-N and (ii) U-Pu-Am-N ceramic fuels. In this case, Ce will be used as the surrogate for Pu, Mn as the surrogate for Am, and depleted uranium as the surrogate for U. Once sufficient fundamental data has been determined for these surrogate systems, the information will be transferred to Idaho National Laboratory (INL) for synthesis of Zr-Am-N, Pu-Am-Zr-N and U-Pu-Am-N ceramic fuels. The high vapor pressures of americium (Am) and americium nitride (AmN) are cause for concern in producing nitride ceramic nuclear fuel that contains Am. Along with the problem of Am retention during the sintering phases of current processing methods, are additional concerns of producing a consistent product of desirable homogeneity, density and porosity. Similar difficulties have been experienced during the laboratory scale process development stage of producing metal alloys containing Am wherein compact powder sintering methods had to be abandoned. Therefore, there is an urgent need to develop a low-temperature or low-heat fuel fabrication process for the synthesis of Am-containing ceramic fuels. Self-propagating high temperature synthesis (SHS), also called combustion synthesis, offers such an alternative process for the synthesis of Am nitride fuels. Although SHS

  1. Process development for fabrication of zircaloy- 4 of dissolver assembly for spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Tonpe, Sunil; Saibaba, N.; Jairaj, R.N.; Ravi Shankar, A.; Kamachi Mudali, U.; Raj, Baldev

    2010-01-01

    Spent fuel reprocessing for fast breeder reactor (FBR) requires a dissolver made of a material which has resistance to corrosion as the process involves Nitric Acid as the process medium. Various materials to achieve minimum corrosion rates have been tried for this operation. Particularly the focus was on the use of advanced materials with high performance (corrosion rate and product life) for high concentrations greater than 8 N and temperatures (boiling and vapour) of Nitric Acid employed in the dissolver unit. The different commercially available materials like SS316L , Pure Titanium, Ti - 5% Ta and Ti - 5% Ta - 1.8% Nb were tried and the corrosion behavior of these materials was studied in detail. As this is continuous process of evolution of new materials, it was decided to try out zircaloy - 4 as the material of construction for construction due to its excellent corrosion resistance properties in Nitric Acid environment. The specifications were stringent and the geometrical configurations of the assembly were very intricate in shape. On accepting the challenge of fabrication of dissolver, NFC has made different fixtures for Electron Beam Welding and TIG Welding. Various trials were carried out for optimization of various operating parameter like beam current, Acceleration voltage, welding speed to get adequate weld penetration. Both EB welding and TIG welding process were standardized and qualified by carrying out a number of trials and testing these welds by various weld qualification procedures like radiography, Liquid dye penetrant testing etc. for different intricate weld geometries. All the welds were simulated with samples to optimize the weld parameters. Tests such as include metallographic (for microstructure and HAZ), mechanical (for weld strength) and chemical (material analysis for gases) were conducted and all the weld samples met the acceptable criteria. Finally the dissolver was made meeting stringent specifications. All the welds were checked

  2. Study of developing nuclear fabrication facility's integrated emergency response manual

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taeh Yeong; Cho, Nam Chan; Han, Seung Hoon; Moon, Jong Han; Lee, Jin Hang [KEPCO, Daejeon (Korea, Republic of); Min, Guem Young; Han, Ji Ah [Dongguk Univ., Daejeon (Korea, Republic of)

    2016-05-15

    Public begin to pay attention to emergency management. Thus, public's consensus on having high level of emergency management system up to advanced country's is reached. In this social atmosphere, manual is considered as key factor to prevent accident or secure business continuity. Therefore, we first define possible crisis at KEPCO Nuclear Fuel (hereinafter KNF) and also make a 'Reaction List' for each crisis situation at the view of information-design. To achieve it, we analyze several country's crisis response manual and then derive component, indicate duties and roles at the information-design point of view. From this, we suggested guideline to make 'Integrated emergency response manual(IERM)'. The manual we used before have following few problems; difficult to applicate at the site, difficult to deliver information. To complement these problems, we searched manual elements from the view of information-design. As a result, we develop administrative manual. Although, this manual could be thought as fragmentary manual because it confined specific several agency/organization and disaster type.

  3. Fabrication development for high-level nuclear waste containers for the tuff repository; Phase 1 final report

    Energy Technology Data Exchange (ETDEWEB)

    Domian, H.A.; Holbrook, R.L.; LaCount, D.F. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

    1990-09-01

    This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs.

  4. Nuclear purity and the production of uranium (1962); La purete nucleaire et la fabrication de l'uranium (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Verte, P [Commissariat a l' Energie Atomique, Centre du Bouchet, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    When the production of 'nuclear grade' uranium is dealt with, it is difficult, the author of this study points out, to separate its chemical, technical, and economical bearings. While recalling the evolution of chemical processes in various countries and describing the technic of uranium manufacture in the plant of the French 'Commissariat a l'Energie Atomique' at Le Bouchet, the author outlines the effect of economical contingencies on the problems the chemists and engineer are faced with. The question of cost price is also considered here with particular attention. (author) [French] Lorsqu'il s'agit de la production d'uranium de 'qualite nucleaire', il est difficile, souligne l'auteur de cette etude, de separer les aspects chimique, technique et economique. Aussi, en retracant l'evolution des procedes chimiques dans divers pays et decrivant les techniques de fabrication de l'uranium a l'usine du Bouchet du Commissariat a l'Energie Atomique, l'auteur ne manque-t-il pas de rappeler les incidences de la conjoncture economique sur les problemes posees au chimiste et a l'ingenieur. La question du prix de revient, egalement, est traitee ici avec une attention particuliere. (auteur)

  5. Control of nuclear material hold-up: The key factors for design and operation of MOX fuel fabrication plants in Europe

    International Nuclear Information System (INIS)

    Beaman, M.; Beckers, J.; Boella, M.

    2001-01-01

    Full text: Some protagonists of the nuclear industry suggest that MOX fuel fabrication plants are awash with nuclear materials which cannot be adequately safeguarded and that materials 'stuck in the plant' could conceal clandestine diversion of plutonium. In Europe the real situation is quite different: nuclear operators have gone to considerable efforts to deploy effective systems for safety, security, quality and nuclear materials control and accountancy which provide detailed information. The safeguards authorities use this information as part of the safeguards measures enabling them to give safeguards assurances for MOX fuel fabrication plants. This paper focuses on the issue of hold-up: definition of the hold-up and of the so-called 'hidden inventory'; measures implemented by the plant operators, from design to day to day operations, for minimising hold-up and 'hidden inventory'; plant operators' actions to manage the hold-up during production activities but also at PIT/PIV time; monitoring and management of the 'hidden inventory'; measures implemented by the safeguards authorities and inspectorate for verification and control of both hold-up and 'hidden inventory'. The examples of the different plant specific experiences related in this paper reveal the extensive experience gained in european MOX fuel fabrication plants by the plant operators and the safeguards authorities for the minimising and the control of both hold-up and 'hidden inventory'. MOX fuel has been fabricated in Europe, with an actual combined capacity of 2501. HM/year subject, without any discrimination, to EURATOM Safeguards, for more than 30 years and the total output is, to date, some 1000 t.HM. (author)

  6. Calibration setting numbers for dose calibrators for the PET isotopes "5"2Mn, "6"4Cu, "7"6Br, "8"6Y, "8"9Zr, "1"2"4I

    International Nuclear Information System (INIS)

    Wooten, A. Lake; Lewis, Benjamin C.; Szatkowski, Daniel J.; Sultan, Deborah H.; Abdin, Kinda I.; Voller, Thomas F.; Liu, Yongjian; Lapi, Suzanne E.

    2016-01-01

    For PET radionuclides, the radioactivity of a sample can be conveniently measured by a dose calibrator. These devices depend on a “calibration setting number”, but many recommended settings from manuals were interpolated based on standard sources of other radionuclide(s). We conducted HPGe gamma-ray spectroscopy, resulting in a reference for determining settings in two types of vessels containing one of several PET radionuclides. Our results reiterate the notion that in-house, experimental calibrations are recommended for different radionuclides and vessels. - Highlights: • Dose calibrators measure radioactivity by ionization of gas from emitted radiation. • Accuracy of dose calibrators depends on “calibration setting numbers” for isotopes. • Many manufacturer settings are interpolated from emissions of other radionuclides. • As a high-precision reference, HPGe gamma-ray spectroscopy was conducted. • New calibrations were found for PET isotopes "5"2Mn, "6"4Cu, "7"6Br, "8"6Y, "8"9Zr, and "1"2"4I.

  7. Risks, costs and benefits analysis for exhumation of buried radioactive materials at a nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Kirk, J.S.; Moore, R.A.; Huston, T.E.

    1996-01-01

    A Risks, Costs and Benefits analysis provides a tool for selecting a cost-effective remedial action alternative. This analysis can help avoid transferring risks to other populations and can objectively measure the benefits of a specific remedial action project. This paper describes the methods and results of a Risks, Costs and Benefits analysis performed at a nuclear fuel fabrication facility. The analysis examined exhuming and transporting radioactive waste to an offsite disposal facility. Risks evaluated for the remedial action project were divided into two categories: risks posed to the worker and risks posed to public health. Risks to workers included exposure to radioactive contaminants during excavation and packaging of waste materials and the use of heavy machinery. Potential public health risks included exposure to radioactive materials during transport from the exhumation site to the disposal facility. Methods included use of site-specific and published data, and existing computer models. Occupational risks were quantified using data from similar onsite remedial action projects. Computer modeling was used to evaluate public health risks from transporting radioactive materials; the consequences or probability of traffic accidents; and radiation exposure to potential inhabitants occupying the site considering various land use scenarios. A costs analysis was based on data obtained from similar onsite remedial action projects. Scenarios used to identify benefits resulting from the remedial action project included (1) an evaluation of reduction in risks to human health; (2) cost reductions associated with the unrestricted release of the property; and (3) benefits identified by evaluating regulatory mandates applicable to decommissioning. This paper will provide an overview of the methods used and a discussion of the results of a Risks, Costs and Benefits analysis for a site-specific remedial action scenario

  8. Construction, fabrication, and installation

    International Nuclear Information System (INIS)

    1992-05-01

    This standard specifies the construction, fabrication, and installation requirements that apply to concrete containment structures of a containment system designated as class containment components, parts and appurtenances for nuclear power plants

  9. I-124 and its applications in nuclear medicine and biology

    International Nuclear Information System (INIS)

    Weinreich, R.; Wyer, L.; Crompton, N.; Nievergelt-Egido, M.C.; Guenther, I.; Roelcke, U.; Leenders, K.L.; Knust, E.J.; Blasberg, R.G.

    1998-01-01

    4.15-d 124 I decays simultaneously by positron emission (25.6 %) and by electron capture (74.4 %). This dualistic decay allows in principle to use 124 I in both diagnostic and therapeutic approaches. In some high-current measurements, 124 I was produced via the nuclear reaction 124 Te(p,n) 124 I using 12.6 MeV protons in yields 25% below those of the mainly used reaction 124 Te(d,2n) 124 I, but with a very much lower contamination by long-lived 125 I and 126 I. The minimum obtained value for the sum of all impurities was 0.14% of the 124 I activity, at 6 days after end of bombardment, using 99.8% enriched 124 TeO 2 as target material. This yield/purity ratio also permits the production of 124 I by low-energy ''baby'' cyclotrons which could considerably increase the general availability of this nuclide. [ 124 ]IUdR was synthesized by direct electrophilic labelling in good yield (45-65 %), high radiochemical purity (>95%) and high specific activity for functional PET imaging of brain tumours. One day after administration to patients and after completion of the ''washout'', the only remaining activity was that in tumour structures. The comparison with the tumour labelling index showed that PET with [ 124 ]IUdR introduces a novel imaging approach: tumour diagnostics by the measurement of cell proliferation. [ 124 ]IodoHoechst 33258 was synthesized by direct electrophilic labelling in yields of 70 % and in a radiochemical purity of 99 %. In cell culture experiments using HTB-40 (human adenocarcinoma line), it was shown to be taken up by the DNA as well as the unlabelled fluorescence dye H 33258. Furthermore, its radiobiological activity was equal to that of the 125 I-labelled H 33258, but markedly stronger than that of the 131 I-labelled derivative. This suggests a mechanism for Auger-electron induced radiobiological activity as a novel therapeutical approach. p-[ 124 ]Iodophenylalanine and [ 124 ]iodo-α-methyltyrosine are two other compounds labelled with 124 I that

  10. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

  11. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  12. Pairwise comparison of 89Zr- and 124I-labeled cG250 based on positron emission tomography imaging and nonlinear immunokinetic modeling: in vivo carbonic anhydrase IX receptor binding and internalization in mouse xenografts of clear-cell renal cell carcinoma

    International Nuclear Information System (INIS)

    Cheal, Sarah M.; Punzalan, Blesida; Doran, Michael G.; Osborne, Joseph R.; Evans, Michael J.; Lewis, Jason S.; Zanzonico, Pat; Larson, Steven M.

    2014-01-01

    The PET tracer, 124 I-cG250, directed against carbonic anhydrase IX (CAIX) shows promise for presurgical diagnosis of clear-cell renal cell carcinoma (ccRCC) (Divgi et al. in Lancet Oncol 8:304-310, 2007; Divgi et al. in J Clin Oncol 31:187-194, 2013). The radiometal 89 Zr, however, may offer advantages as a surrogate PET nuclide over 124 I in terms of greater tumor uptake and retention (Rice et al. in Semin Nucl Med 41:265-282, 2011). We have developed a nonlinear immunokinetic model to facilitate a quantitative comparison of absolute uptake and antibody turnover between 124 I-cG250 and 89 Zr-cG250 using a human ccRCC xenograft tumor model in mice. We believe that this unique model better relates quantitative imaging data to the salient biological features of tumor antibody-antigen binding and turnover. We conducted experiments with 89 Zr-cG250 and 124 I-cG250 using a human ccRCC cell line (SK-RC-38) to characterize the binding affinity and internalization kinetics of the two tracers in vitro. Serial PET imaging was performed in mice bearing subcutaneous ccRCC tumors to simultaneously detect and quantify time-dependent tumor uptake in vivo. Using the known specific activities of the two tracers, the equilibrium rates of antibody internalization and turnover in the tumors were derived from the PET images using nonlinear compartmental modeling. The two tracers demonstrated virtually identical tumor cell binding and internalization but showed markedly different retentions in vitro. Superior PET images were obtained using 89 Zr-cG250, owing to the more prolonged trapping of the radiolabel in the tumor and simultaneous washout from normal tissues. Estimates of cG250/CAIX complex turnover were 1.35 - 5.51 x 10 12 molecules per hour per gram of tumor (20 % of receptors internalized per hour), and the ratio of 124 I/ 89 Zr atoms released per unit time by tumor was 17.5. Pairwise evaluation of 89 Zr-cG250 and 124 I-cG250 provided the basis for a nonlinear immunokinetic

  13. Pairwise comparison of {sup 89}Zr- and {sup 124}I-labeled cG250 based on positron emission tomography imaging and nonlinear immunokinetic modeling: in vivo carbonic anhydrase IX receptor binding and internalization in mouse xenografts of clear-cell renal cell carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Cheal, Sarah M.; Punzalan, Blesida; Doran, Michael G.; Osborne, Joseph R. [Memorial Sloan-Kettering Cancer Center, Department of Radiology, New York, NY (United States); Evans, Michael J. [Memorial Sloan-Kettering Cancer Center, Human Oncology and Pathogenesis Program, New York, NY (United States); Lewis, Jason S. [Memorial Sloan-Kettering Cancer Center, Department of Radiology, New York, NY (United States); Memorial Sloan-Kettering Cancer Center, Program in Molecular Pharmacology and Chemistry, New York, NY (United States); Memorial Sloan-Kettering Cancer Center, Radiochemistry and Imaging Sciences Service, New York, NY (United States); Zanzonico, Pat [Memorial Sloan-Kettering Cancer Center, Department of Radiology, New York, NY (United States); Memorial Sloan-Kettering Cancer Center, Molecular Pharmacology and Therapy Service, New York, NY (United States); Memorial-Sloan Kettering Cancer Center, New York, NY (United States); Larson, Steven M. [Memorial Sloan-Kettering Cancer Center, Department of Radiology, New York, NY (United States); Memorial Sloan-Kettering Cancer Center, Program in Molecular Pharmacology and Chemistry, New York, NY (United States); Memorial Sloan-Kettering Cancer Center, Molecular Pharmacology and Therapy Service, New York, NY (United States)

    2014-05-15

    The PET tracer, {sup 124}I-cG250, directed against carbonic anhydrase IX (CAIX) shows promise for presurgical diagnosis of clear-cell renal cell carcinoma (ccRCC) (Divgi et al. in Lancet Oncol 8:304-310, 2007; Divgi et al. in J Clin Oncol 31:187-194, 2013). The radiometal {sup 89}Zr, however, may offer advantages as a surrogate PET nuclide over {sup 124}I in terms of greater tumor uptake and retention (Rice et al. in Semin Nucl Med 41:265-282, 2011). We have developed a nonlinear immunokinetic model to facilitate a quantitative comparison of absolute uptake and antibody turnover between {sup 124}I-cG250 and {sup 89}Zr-cG250 using a human ccRCC xenograft tumor model in mice. We believe that this unique model better relates quantitative imaging data to the salient biological features of tumor antibody-antigen binding and turnover. We conducted experiments with {sup 89}Zr-cG250 and {sup 124}I-cG250 using a human ccRCC cell line (SK-RC-38) to characterize the binding affinity and internalization kinetics of the two tracers in vitro. Serial PET imaging was performed in mice bearing subcutaneous ccRCC tumors to simultaneously detect and quantify time-dependent tumor uptake in vivo. Using the known specific activities of the two tracers, the equilibrium rates of antibody internalization and turnover in the tumors were derived from the PET images using nonlinear compartmental modeling. The two tracers demonstrated virtually identical tumor cell binding and internalization but showed markedly different retentions in vitro. Superior PET images were obtained using {sup 89}Zr-cG250, owing to the more prolonged trapping of the radiolabel in the tumor and simultaneous washout from normal tissues. Estimates of cG250/CAIX complex turnover were 1.35 - 5.51 x 10{sup 12} molecules per hour per gram of tumor (20 % of receptors internalized per hour), and the ratio of {sup 124}I/{sup 89}Zr atoms released per unit time by tumor was 17.5. Pairwise evaluation of {sup 89}Zr-cG250 and {sup

  14. An evaluation of UO2-CNT composites made by SPS as an accident tolerant nuclear fuel pellet and the feasibility of SPS as an economical fabrication process for the nuclear fuel cycle

    Science.gov (United States)

    Cartas, Andrew R.

    The innovative and advanced purpose of this study is to understand and establish proper sintering procedures for Spark Plasma Sintering process in order to fabricate high density, high thermal conductivity UO2 -CNT pellets. Mixing quality and chemical reactions have been investigated by field emission scanning electron microscopy (FESEM), wavelength dispersive spectroscopy (WDS), and X-ray diffraction (XRD). The effect of various types of CNTs on the mixing and sintering quality of UO2-CNT pellets with SPS processing have been examined. The Archimedes Immersion Method, laser flash method, and FE-SEM will be used to investigate the density, thermal conductivity, grain size, pinning effects, and CNT dispersion of fabricated UO2-CNT pellets. Pre-fabricated CNT's were added to UO 2 powder and dispersed via sonication and/or ball milling and then made into composite nuclear pellets. An investigation of the economic impact of SPS on the nuclear fuel cycle for producing pure and composite UO2 fuels was conducted.

  15. Role of non-destructive examinations in leak testing of glove boxes for industrial scale plutonium handling at nuclear fuel fabrication facility along with case study

    International Nuclear Information System (INIS)

    Aher, Sachin

    2015-01-01

    Non Destructive Examinations has the prominent role at Nuclear Fuel Fabrication Facilities. Specifically NDE has contributed at utmost stratum in Leak Testing of Glove Boxes and qualifying them as a Class-I confinement for safe Plutonium handling at industrial scale. Advanced Fuel Fabrication Facility, BARC, Tarapur is engaged in fabrication of Plutonium based MOX (PuO 2 , DDUO 2 ) fuel with different enrichments for first core of PFBR reactor. Alpha- Leak Tight Glove Boxes along with HEPA Filters and dynamic ventilation form the promising engineering system for safe and reliable handling of plutonium bearing materials considering the radiotoxicity and risk associated with handling of plutonium. Leak Testing of Glove Boxes which involves the leak detection, leak rectification and leak quantifications is major challenging task. To accomplish this challenge, various Non Destructive Testing methods have assisted in promising way to achieve the stringent leak rate criterion for commissioning of Glove Box facilities for plutonium handling. This paper highlights the Role of various NDE techniques like Soap Solution Test, Argon Sniffer Test, Pressure Drop/Rise Test etc. in Glove Box Leak Testing along with procedure and methodology for effective rectification of leakage points. A Flow Chart consisting of Glove Box leak testing procedure starting from preliminary stage up to qualification stage along with a case study and observations are discussed in this paper. (author)

  16. A review of the environmental impact of mining and milling of radioactive ores, upgrading processes, and fabrication of nuclear fuels

    International Nuclear Information System (INIS)

    Costello, J.M.; Davy, D.R.; Cattell, F.C.R.; Cook, J.E.

    1980-01-01

    The subject is discussed under the headings: uranium mining; milling of uranium ores; manufacture of uranium hexafluoride; uranium enrichment; fuel manufacture and fabrication; environmental impact (use of natural resources; effluents from fuel cycle operations; occupational health; public health); alternative fuel cycles; additional waste treatment. (U.K.)

  17. pHluorin-assisted expression, purification, crystallization and X-ray diffraction data analysis of the C-terminal domain of the HsdR subunit of the Escherichia coli type I restriction-modification system EcoR124I

    Czech Academy of Sciences Publication Activity Database

    Grinkevich, Pavel; Iermak, Iuliia; Luedtke, N.A.; Mesters, J. R.; Ettrich, Rüdiger; Ludwig, Jost

    2016-01-01

    Roč. 72, č. 9 (2016), s. 672-676 ISSN 2053-230X R&D Projects: GA ČR GAP207/12/2323; GA MŠk(CZ) LM2015055 Institutional support: RVO:61388971 Keywords : restriction-modification system * EcoR124I * HsdR Subject RIV: EE - Microbiology, Virology Impact factor: 0.799, year: 2016

  18. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  19. Modular nuclear reactor for a land-based power plant and method for the fabrication installation and operation thereof

    International Nuclear Information System (INIS)

    Craig, E. R.; Blumberg, B. Jr.

    1985-01-01

    A self-contained modular nuclear reactor which can be prefabricated at a factory location, nuclear-certified at the factory, transported to a field location for final assembly and connection to a large-scale electric-power generating facility. The modular reactor includes a prefabricated nuclear heat supply module and a plurality of shell segments which can be assembled about the heat supply module and which provide a form for the pouring and curing of a cementatious biological shield about the heat supply module. The modular reactor includes passive shutdown heat removal systems sufficient to render the reactor safe in an emergency. A large-scale power plant arrangement is disclosed which incorporates a plurality of the modular reactors

  20. Operational experience in the non-destructive assay of fissile material in General Electric's nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Stewart, J.P.

    1976-01-01

    Operational experience in the non-destructive assay of fissile material in a variety of forms and containers and incorporation of the assay devices into the accountability measurement system for General Electric's Wilmington Fuel Fabrication Facility measurement control programme is detailed. Description of the purpose and related operational requirements of each non-destructive assay system is also included. In addition, the accountability data acquisition and processing system is described in relation to its interaction with the various non-destructive assay devices and scales used for accountability purposes within the facility. (author)

  1. Cost effectiveness of robotics and remote tooling for occupational risk reduction at a nuclear fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Lochard, Jacques

    1989-08-01

    This case study, related to the design stage of a fuel fabrication facility, presents the evaluation of alternative options to manipulate mixed oxide fuel rods in a quality control shop. It is based on a study performed in the framework of the 'MELOX project' developed by COGEMA in France. The methodology for evaluating robotic actions is resulting from a research work part funded by the IAEA under the co-ordinated research programme on 'Comparison of cost-effectiveness of risk reduction among different energy systems', and by the commission of the European Communities under the research and training programme on radiation protection.

  2. Cost effectiveness of robotics and remote tooling for occupational risk reduction at a nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Lochard, Jacques

    1989-01-01

    This case study, related to the design stage of a fuel fabrication facility, presents the evaluation of alternative options to manipulate mixed oxide fuel rods in a quality control shop. It is based on a study performed in the framework of the 'MELOX project' developed by COGEMA in France. The methodology for evaluating robotic actions is resulting from a research work part funded by the IAEA under the co-ordinated research programme on 'Comparison of cost-effectiveness of risk reduction among different energy systems', and by the commission of the European Communities under the research and training programme on radiation protection

  3. Modular nuclear fuel element, modular capsule for a such element and fabrication process for a modular capsule

    International Nuclear Information System (INIS)

    Chotard, A.

    1988-01-01

    The nuclear fuel rod is made by a tubular casing closed at both ends and containing a series of modular capsules with little play with the casing and made by a jacket closed by porous plugs at both ends and containing a stack of fuel pellets [fr

  4. Results from a Field Trial of the Radio Frequency Based Cylinder Accountability and Tracking System at the Global Nuclear Fuel Americas Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Fitzgerald, Peter; Laughter, Mark D.; Martyn, Rose; Pickett, Chris A.; Rowe, Nathan C.; Younkin, James R.; Shephard, Adam M.

    2010-01-01

    The Cylinder Accountability and Tracking System (CATS) is a tool designed for use by the International Atomic Energy Agency (IAEA) to improve overall inspector efficiency through real-time unattended monitoring of cylinder movements, site specific rules-based event detection, and the capability to integrate many types of monitoring technologies. The system is based on the tracking of cylinder movements using (radio frequency) RF tags, and the collection of data, such as accountability weights, that can be associated with the cylinders. This presentation will cover the installation and evaluation of the CATS at the Global Nuclear Fuels (GNF) fuel fabrication facility in Wilmington, NC. This system was installed to evaluate its safeguards applicability, operational durability under operating conditions, and overall performance. An overview of the system design and elements specific to the GNF deployment will be presented along with lessons learned from the installation process and results from the field trial.

  5. Fabrication of nuclear fuel by powder injection moulding: Study of the binders systems and the de-binding of feedstock containing actinide powder

    International Nuclear Information System (INIS)

    Bricout, J.

    2012-01-01

    Powder Injection Moulding (PIM) is identified as an innovative process for the nuclear fuel fabrication. Technological breakthrough compared to the current process of powder metallurgy, the impact of actinide powder's specificities on the different steps of PIM is performed. Alumina powders simulating actinide powder have been implemented with a reference binders system. Thermal and rheological studies show the injectability and the de-binding of feedstocks with adequate solid loading (≥50 %vol), thanks to the de-agglomeration during the mixing step, which allow to obtain net shape fuel pellet. Specific surface area of powders, acting as a key role in behaviour's feedstocks, has been integrated in analysis models of viscosity prediction according to the shear rate. Also conducted studies on uranium oxide powder show that the selected binders systems, which have a compatible rheological behaviour with PIM process, impact the de-agglomeration of powder and final microstructure of the fuel pellet, consistent with the results obtained on alumina powders. Independent behaviour of binders and uranium oxide powder, showing no adverse chemical reaction against the PIM process, show a residual mass of carbon of about 150 ppm after sintering. Binders system using polystyrene, resistant to radiolysis phenomena and loadable more than 50 %(vol) of actinide powder, shows the promising potential of PIM process for the fuel fabrication. (author) [fr

  6. Review of experience gained in fabricating nuclear grade uranium and thorium compounds and their analytical quality control at the Instituto de Energia Atomica, Sao Paulo, Brazil

    International Nuclear Information System (INIS)

    Abrao, A.; Franca, J.M. Jr.; Ikuta, A.; Pueschel, C.R.; Federgruen, L.; Lordello, A.R.; Tomida, E.K.; Moraes, S.; Brito, J. de; Gomes, R.P.; Araujo, J.A.; Floh, B.; Matsuda, H.T.

    1977-01-01

    This paper summarizes the main activities dealing with the fabrication of nuclear grade uranium and thorium compounds at the Instituto de Energia Atomica, Sao Paulo. Identification of problems and their resolutions, the experience gained in plant operation, the performance characteristics of an ion-exchange facility and a solvent extraction unit (a demonstration plant based on pulsed columns for purification of uranium and production of ammonium diuranate) are described. A moving-bed facility for UF 4 preparation and its operation is discussed. A pilot plant for uranium and thorium oxide microsphere preparation based on internal gelation for HTGR fuel type is also described. A solvent extraction pilot plant for thorium purification based on a compound extraction-scrubbing column and a mixer-settler battery and the involved technology for thorium purification are commented. The main products, namely ammonium diuranate, uranyl amonium tricarbonate, uranium trioxide, uranium tetrafluoride, thorium nitrate and thorium oxalate and their quality are commented. The development of necessary analytical procedures for the quality control of the mentioned nuclear grade products is summarized. A great majority of such procedures was particularly suitable for analyzing traces impurities. Designed for installation are the units for denitration of uranyl nitrate solutions and pilot plants for elemental fluorine and UF 6 . The installation of a laboratory-scale plant designed for reprocessing irradiated uranium and an experimental unit for the recovery of protactinium from irradiated thorium is in progress

  7. Preliminary experiments for the fabrication of clad for a spherical fuel for a research fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Almeida, L.A.A.

    1982-01-01

    A preliminary experiments using 1100 aluminum 0,5mm thick hemispheres welded on 10mm diameter steel and ceramic spheres in order to determine a method to clad spherical fuel for a research fluidized bed nuclear reactor were studied. The processes of hot press, T.I.G. and resistance we use for welding. A qualitative compression and metalographic tests of welded pieces are performed. By the analysis of the results of the tests we conclude that the resistance welding was the best. The experimental methods and the results with their analysis are presented in the paper. (Author) [pt

  8. Fabrication of surfactant-free quercetin-loaded PLGA nanoparticles: evaluation of hepatoprotective efficacy by nuclear scintigraphy

    Energy Technology Data Exchange (ETDEWEB)

    Ganguly, Soumya; Gaonkar, Raghuvir H. [CSIR-Indian Institute of Chemical Biology, Infectious Diseases and Immunology Division (India); Sinha, Samarendu; Gupta, Amit [Thakurpukur Cancer Centre and Welfare Home Campus, Regional Radiation Medicine Centre (India); Chattopadhyay, Dipankar [University of Calcutta, Department of Polymer Science & Technology, University College of Science & Technology (India); Chattopadhyay, Sankha [Variable Energy Cyclotron Centre, Radiopharmaceuticals Laboratory, Board of Radiation and Isotope Technology (India); Sachdeva, Satbir S. [Radiopharmaceuticals Production (India); Ganguly, Shantanu [Thakurpukur Cancer Centre and Welfare Home Campus, Regional Radiation Medicine Centre (India); Debnath, Mita C., E-mail: mitacd@iicb.res.in, E-mail: mita-chdebnath@yahoo.com [CSIR-Indian Institute of Chemical Biology, Infectious Diseases and Immunology Division (India)

    2016-07-15

    The purpose of this study was to develop surfactant-free quercetin-loaded PLGA nanoparticles (Qr-NPs) and investigate the hepatoprotective efficacy of the product non-invasively by nuclear scintigraphy. The nanoparticles were prepared using PLGA by dialysis method and ranged in size between 50 and 250 nm with a narrow range of distribution. They were found to arrive at the fenestra of liver sinusoidal epithelium for accumulation. The sizes of nanoparticles (batch S1) were optimal to reach the target and offer enough protection of the hepatocytes degenerated by CCl{sub 4} intoxication as determined by various biochemical and histopathological tests. In vitro studies exhibited the cytotoxic effect of the formulation against HepG2 cell line. The hepatoprotective efficacy of Qr-NPs evaluated non-invasively by nuclear scintigraphic technique using {sup 99m}Tc-labelled sulphur colloid revealed abnormality in liver at the area of decreased uptake in rats of CCl{sub 4}-treated group, which disappeared in Qr-NP-treated group. In dynamic studies with {sup 99m}Tc-mebrofenin, excretion was severely impaired in CCl{sub 4}-treated group but was moderate in drug-treated group, proving the recovery of animals from damage.Graphical Abstract.

  9. Fabrication Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Fabrication Facilities are a direct result of years of testing support. Through years of experience, the three fabrication facilities (Fort Hood, Fort Lewis, and...

  10. Environmental concerns in regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    Santos, Glaucia R.T.; Durazzo, Michelangelo; Carvalho, Elita F.U.; Riella, Humberto G.

    2008-01-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the main programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel -CCN- is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt % 235 U), to supply its IEA-R1 research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts.This study aims to implant the Sustainability Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  11. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U.; Riella, H. G.

    2008-01-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% 2 35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  12. Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

    International Nuclear Information System (INIS)

    Fleming Rubio, Peter Alex

    2010-01-01

    The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

  13. Effect of prior machining deformation on the development of tensile residual stresses in weld-fabricated nuclear components

    International Nuclear Information System (INIS)

    Prevey, P.S.; Mason, P.W.; Hornbach, D.J.; Molkenthin, J.P.

    1996-01-01

    Austenitic alloy weldments in nuclear systems may be subject to stress-corrosion cracking (SCC) failure if the sum of residual and applied stresses exceeds a critical threshold. Residual stresses developed by prior machining and welding may either accelerate or retard SCC, depending on their magnitude and sign. A combined x-ray diffraction and mechanical procedure was used to determine the axial and hoop residual stress and yield strength distributions into the inside-diameter surface of a simulated Alloy 600 penetration J-welded into a reactor pressure vessel. The degree of cold working and the resulting yield strength increase caused by prior machining and weld shrinkage were calculated from the line-broadening distributions. Tensile residual stresses on the order of +700 MPa were observed in both the axial and the hoop directions at the inside-diameter surface in a narrow region adjacent to the weld heat-affected zone. Stresses exceeding the bulk yield strength were found to develop due to the combined effects of cold working of the surface layers during initial machining and subsequent weld shrinkage. The residual stress and cold work distributions produced by prior machining were found to influence strongly the final residual stress state developed after welding

  14. Fabrication of the shafts of the liquid metal pumps for the Creys-Malville nuclear power station

    International Nuclear Information System (INIS)

    Pasqualini, G.; Lefebvre, B.; Archer, J.; Gravier, M.

    1982-01-01

    This report is a synthesis of the considerations with regard to the project work and the work executes in the field of metallurgy, which have made it possible to manufacture the shafts of primary and secondary pumps intended for the Creys-Malville nuclear power station. In the first part of this report attention is drawn to the most important items of this equipment with regard to the performance specifications. These specifications are the expression of the experiences made in France in the industrial manufacture of pumps for liquid metals for this type of application Rapsodie (1967) and Phenix (1974). In the second part of the report on hand, in particular the technical aspects of the welding operations with regard to the use of the chosen material (austenitic corrosion resisting steel Z 15 CNW 22-12, maual TIG welding, the type of steel of the filler metal being the same as the parent metal) will be discussed. Finally, a testified comment on the most important steps of the manufacture of these shafts in the works at Jeumont will be described. (orig.) [de

  15. Preliminary results for the Co-Rolling process fabrication of plate-type nuclear fuel based in U-10Mo monolithic meat and zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Pedrosa, Tercio A.; Brina, Jose Giovanni M.; Paula, Joao Bosco de; Lameiras, Fernando S.; Ferraz, Wilmar B.

    2013-01-01

    The fabrication process of plate-type nuclear fuel with monolithic meat is under development at CDTN. The U-10Mo alloy was chosen as the meat material due to its high density, corrosion resistance and the higher dimensional stability proportioned by the metastable gamma phase, which presents a lesser extension of the breakaway swelling phenomena occurrence during irradiation tests. The monolithic meat was cut from an U-10Mo ingot, that was induction melted at CDTN. The co-rolling process was adopted due to the higher mechanical properties and melting point of the Zircalloy-4 cladding material, which presents a lesser discrepancy in relation to the meat material properties, when compared to the aluminum 6061 alloy. Preliminary plates were obtained by means of the co-rolling process, performed at 650, 800, 950 deg C with total thickness reduction of 80%, followed by a pickling step and cold co-rolling passes. The plates were characterized through bending tests, optical microscopy and radiography. The co-rolling temperature of 800 deg C presented the best results, with a homogeneous distribution of the total thickness reduction between the cover plates and the meat, and the absence of delamination in the bending test samples. It was observed the occurrence of meat thickening in its ends, according to its longitudinal axle, parallel to the rolling direction, that is known as the d og bone , for the three co-rolling temperatures. (author)

  16. TU-F-12A-01: Quantitative Non-Linear Compartment Modeling of 89Zr- and 124I- Labeled J591 Monoclonal Antibody Kinetics Using Serial Non-Invasive Positron Emission Tomography Imaging in a Pre-Clinical Human Prostate Cancer Mouse Model

    Energy Technology Data Exchange (ETDEWEB)

    Fung, EK; Cheal, SM; Chalasani, S; Fareedy, SB; Punzalan, B; Humm, JL; Osborne, JR; Larson, SM; Zanzonico, PB [Memorial Sloan Kettering Cancer Center, New York, NY (United States); Otto, B; Bander, NH [Weill Cornell Medical College, New York, NY (United States)

    2014-06-15

    Purpose: To examine the binding kinetics of human IgG monoclonal antibody J591 which targets prostate-specific membrane antigen (PSMA) in a pre-clinical mouse cancer model using quantitative PET compartmental analysis of two radiolabeled variants. Methods: PSMA is expressed in normal human prostate, and becomes highly upregulated in prostate cancer, making it a promising therapeutic target. Two forms of J591, radiolabeled with either {sup 89}Zr or {sup 124}I, were prepared. {sup 89}Zr is a radiometal that becomes trapped in the cell upon internalization by the antigen-antibody complex, while radioiodine leaves the cell. Mice with prostate cancer xenografts underwent non-invasive serial imaging on a Focus 120 microPET up to 144 hours post-injection of J591. A non-linear compartmental model describing the binding and internalization of antibody in tumor xenograft was developed and applied to the PET-derived time-activity curves. The antibody-antigen association rate constant (ka), total amount of antigen per gram tumor (Ag-total), internalization rate of antibody-antigen complex, and efflux rate of radioisotope from tumor were fitted using the model. The surface-bound and the internalized activity were also estimated. Results: Values for ka, Ag-total, and internalization rate were found to be similar regardless of radiolabel payload used. The efflux rate, however, was ∼ 9-fold higher for {sup 124}I-J591 than for {sup 89}Zr-J591. Time-dependent surface-bound and internalized radiotracer activity were similar for both radiolabels at early times post-injection, but clearly differed beyond 24 hours. Conclusion: Binding and internalization of J591 to PSMA-expressing tumor xenografts were similar when radiolabeled with either {sup 89}Zr or {sup 124}I payload. The difference in efflux of radioactivity from tumor may be attributable to differential biological fate intracellularly of the radioisotopes. This has great significance for radioimmunotherapy and antibody

  17. Interpretation of the results from individual monitoring of workers at the Nuclear Fuel Fabrication Facility, Brazil; Interpretacao de resultados de monitoracao individual interna da Fabrica de Combustivel Nuclear - FCN

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Marcelo Xavier de

    2005-07-01

    In nuclear fuel fabrication facilities, workers are exposed to different compounds of enriched uranium. Although in this kind of facility the main route of intake is inhalation, ingestion may occur in some situations, and also a mixture of both. The interpretation of the bioassay data is very complex, since it is necessary taking into account all the different parameters, which is a big challenge. Due to the high cost of the individual monitoring programme for internal dose assessment in the routine monitoring programmes, usually only one type of measurement is assigned. In complex situations like the one described in this study, where several parameters can compromise the accuracy of the bioassay interpretation it is need to have a combination of techniques to evaluate the internal dose. According to ICRP 78 (1997), the general order of preference of measurement methodologies in terms of accuracy of interpretation is: body activity measurement, excreta analysis and personal air sampling. Results of monitoring of working environment may provide information that assists in the interpretation on particle size, chemical form, solubility and date of intake. A group of fifteen workers from controlled area of the studied nuclear fuel fabrication facility was selected to evaluate the internal dose using all different available techniques during a certain period. The workers were monitored for determination of uranium content in the daily urinary and faecal excretion (collected over a period of 3 consecutive days), chest counting and personal air sampling. The results have shown that at least two types of sensitivity techniques must be used, since there are some sources of uncertainties on the bioassay interpretation, like mixture of uranium compounds intake and different routes of intake. The combination of urine and faeces analysis has shown to be the more appropriate methodology for assessing internal dose in this situation. The chest counting methodology has not shown

  18. Nuclear power

    International Nuclear Information System (INIS)

    Abd Khalik Wood

    2005-01-01

    This chapter discussed the following topics related to the nuclear power: nuclear reactions, nuclear reactors and its components - reactor fuel, fuel assembly, moderator, control system, coolants. The topics titled nuclear fuel cycle following subtopics are covered: , mining and milling, tailings, enrichment, fuel fabrication, reactor operations, radioactive waste and fuel reprocessing. Special topic on types of nuclear reactor highlighted the reactors for research, training, production, material testing and quite detail on reactors for electricity generation. Other related topics are also discussed: sustainability of nuclear power, renewable nuclear fuel, human capital, environmental friendly, emission free, impacts on global warming and air pollution, conservation and preservation, and future prospect of nuclear power

  19. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  20. Advanced fabrication technology

    International Nuclear Information System (INIS)

    Sheely, W.F.

    1986-01-01

    The Fuel Cycle Plant is a multipurpose nuclear facility located on the Hanford Nuclear Reservation in eastern Washington state. The facility is part of the Hanford Engineering Development Laboratory which is operated by Westinghouse Hanford Company for the Department of Energy. The Fuel Cycle Plant is currently being prepared to support the Liquid Metal Reactors Program with fuel fabrication services for the Fast Flux Test Facility and other LMR programs. This report describes the technical innovations to be utilized in the operation of this plant

  1. Digital fabrication

    CERN Document Server

    2012-01-01

    The Winter 2012 (vol. 14 no. 3) issue of the Nexus Network Journal features seven original papers dedicated to the theme “Digital Fabrication”. Digital fabrication is changing architecture in fundamental ways in every phase, from concept to artifact. Projects growing out of research in digital fabrication are dependent on software that is entirely surface-oriented in its underlying mathematics. Decisions made during design, prototyping, fabrication and assembly rely on codes, scripts, parameters, operating systems and software, creating the need for teams with multidisciplinary expertise and different skills, from IT to architecture, design, material engineering, and mathematics, among others The papers grew out of a Lisbon symposium hosted by the ISCTE-Instituto Universitario de Lisboa entitled “Digital Fabrication – A State of the Art”. The issue is completed with four other research papers which address different mathematical instruments applied to architecture, including geometric tracing system...

  2. OPO fabric decontamination

    International Nuclear Information System (INIS)

    Severa, J.; Bar, J.; Grujbar, V.

    1978-01-01

    Samples of five polypropylene-based man-made fabrics were studied with regard to the degree of contamination and possibilities of decontamination in order to assess their suitability as material for protective clothing in the nuclear industry. The contamination degree of the fabrics in an aqueous solution of a fission product mixture was found to be low. Soaking in a mixture of the Sapon detergent and sodium hexametaphosphate at a concentration of both materials of 1 g/l with subsequent washing in a solution of the Zenit detergent at a concentration of 3 g/l was suggested as the most suitable decontamination procedure. It reduces the initial contamination by almost 99%. (Z.M.)

  3. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hardy, C.J.; Silver, J.M.

    1985-09-01

    The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research

  4. Nuclear

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    The first text deals with a new circular concerning the collect of the medicine radioactive wastes, containing radium. This campaign wants to incite people to let go their radioactive wastes (needles, tubes) in order to suppress any danger. The second text presents a decree of the 31 december 1999, relative to the limitations of noise and external risks resulting from the nuclear facilities exploitation: noise, atmospheric pollution, water pollution, wastes management and fire prevention. (A.L.B.)

  5. Metal finishing and vacuum processes groups, Materials Fabrication Division progress report, March-May 1984

    International Nuclear Information System (INIS)

    Dini, J.W.; Romo, J.G.; Jones, L.M.

    1984-01-01

    Progress is reported in fabrication and coating activities being conducted for the weapons program, nuclear test program, nuclear design program, magnetic fusion program, and miscellaneous applications

  6. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2013-06-28

    The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative (GTRI) Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors.

  7. Stainless steel fabrications: past and present

    International Nuclear Information System (INIS)

    Daniels, R.

    1986-01-01

    The paper deals with stainless steel fabrications of Fairey Engineering Company for the nuclear industry. The manufacture of stainless steel containers for Magnox and Advanced Gas Cooled Reactors, flexible fabrication facility, and welding development, are all briefly described. (U.K.)

  8. Interfacing robotics with plutonium fuel fabrication

    International Nuclear Information System (INIS)

    Bowen, W.W.; Moore, F.W.

    1986-01-01

    Interfacing robotic systems with nuclear fuel fabrication processes resulted in a number of interfacing challenges. The system not only interfaces with the fuel process, but must also interface with nuclear containment, radiation control boundaries, criticality control restrictions, and numerous other safety systems required in a fuel fabrication plant. The robotic system must be designed to allow operator interface during maintenance and recovery from an upset as well as normal operations

  9. Fabricated Elastin.

    Science.gov (United States)

    Yeo, Giselle C; Aghaei-Ghareh-Bolagh, Behnaz; Brackenreg, Edwin P; Hiob, Matti A; Lee, Pearl; Weiss, Anthony S

    2015-11-18

    The mechanical stability, elasticity, inherent bioactivity, and self-assembly properties of elastin make it a highly attractive candidate for the fabrication of versatile biomaterials. The ability to engineer specific peptide sequences derived from elastin allows the precise control of these physicochemical and organizational characteristics, and further broadens the diversity of elastin-based applications. Elastin and elastin-like peptides can also be modified or blended with other natural or synthetic moieties, including peptides, proteins, polysaccharides, and polymers, to augment existing capabilities or confer additional architectural and biofunctional features to compositionally pure materials. Elastin and elastin-based composites have been subjected to diverse fabrication processes, including heating, electrospinning, wet spinning, solvent casting, freeze-drying, and cross-linking, for the manufacture of particles, fibers, gels, tubes, sheets and films. The resulting materials can be tailored to possess specific strength, elasticity, morphology, topography, porosity, wettability, surface charge, and bioactivity. This extraordinary tunability of elastin-based constructs enables their use in a range of biomedical and tissue engineering applications such as targeted drug delivery, cell encapsulation, vascular repair, nerve regeneration, wound healing, and dermal, cartilage, bone, and dental replacement. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  10. Nuclear energy and the environment

    International Nuclear Information System (INIS)

    El-Hinnawi, E.E.

    1980-01-01

    Chapters are presented concerning the environmental impact of mining and milling of radioactive ores, upgrading processes, and fabrication of nuclear fuels; environmental impacts of nuclear power plants; non-radiological environmental implications of nuclear energy; radioactive releases from nuclear power plant accidents; environmental impact of reprocessing; nuclear waste disposal; fuel cycle; and the future of nuclear energy

  11. Center for emergency response at the ENUSA fuel fabrication plant in Juzbado; El centro de gestion de las emergencias de la fabrica de combustible nuclear de ENUSA en Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    Alvaro Perez, C.; Romano, A.

    2016-08-01

    Effective emergency preparedness and management is critical for a safe exploitation of nuclear installations like the Enusa fuel fabrication plant. In 2012, an important project was carried out at the plant which enlarged and remodeled the Emergency Room used until then to give response to the Internal Emergency Plan postulated scenarios. This project was motivated after carefully analyzing the results of audits, inspections and operation experience as well as after studying the conclusions of the Fukushima accident emergency management weaknesses. The new Center for Emergency Response, which hosts the plant control room, devoted to monitoring the plant safety systems on a constant basis, greatly improves both technical means available and operative procedures as well as human interactions during an emergency. This paper describes the most relevant technical features of this Center, the safety systems which support its operation and the emergency management process that takes place in it. (Author)

  12. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    International Nuclear Information System (INIS)

    Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.

    1976-06-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables

  13. Review of qualifications for fuel assembly fabrication

    International Nuclear Information System (INIS)

    Slabu, Dan; Zemek, Martin; Hellwig, Christian

    2013-01-01

    The required quality of nuclear fuel in industrial production can only be assured by applying processes in fabrication and inspection, which are well mastered and have been proven by an appropriate qualification. The present contribution shows the understanding and experiences of Axpo with respect to qualifications in the frame of nuclear fuel manufacturing and reflects some related expectations of the operator. (orig.)

  14. An Ethology of Urban Fabric(s)

    DEFF Research Database (Denmark)

    Fritsch, Jonas; Thomsen, Bodil Marie Stavning

    2014-01-01

    The article explores a non-metaphorical understanding of urban fabric(s), shifting the attention from a bird’s eye perspective to the actual, textural manifestations of a variety of urban fabric(s) to be studied in their real, processual, ecological and ethological complexity within urban life. We...... effectuate this move by bringing into resonance a range of intersecting fields that all deal with urban fabric(s) in complementary ways (interaction design and urban design activism, fashion, cultural theory, philosophy, urban computing)....

  15. Nuclear Safety

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1989-01-01

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  16. Immobilization of high activity nuclear wastes in sintered glass. Fabrication of blocks at semi-industrial scale by hot pressing technique

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi, N.B.; Riquelme, R.; Sterba, M.E.; Audero, M.A.

    1990-01-01

    The sintering process under glass pressure has been studied as an alternative of melting with the aim of obtaining a monolytic material apt to preserve the high activity nuclear wastes. Different properties of the products obtained have been evaluated where the material is selected on the basis of the results attained. The purpose of this work is the equipment development and the process adjusting for the blocks obtainment. (Author) [es

  17. Review of experience gained in fabricating nuclear grade uranium and thorium compounds and their analytical quality control at the Instituto de Energia Atomica, Sao Paulo, Brazil

    International Nuclear Information System (INIS)

    Abrao, A.; Franca Junior, J.M.; Ikuta, A.

    1977-01-01

    The main activities developed at 'Instituto de Energia Atomica' Sao Paulo, Brazil, on the recovery of uranium from ores, the purification of uranium and thorium raw concentrates and their transformation in nuclear grade compounds, are reviewed. The design and assemble of pilot facilities for ammonium diuranate (ADV) uranium tetrafluoride, uranium trioxide, uranium oxide microspheres, uranyl nitrate denitration, uranim hexafluoride and thorium compounds are discussed. The establishment of analytical procedures are emphasized [pt

  18. Study and fabrication of a broadband receiver with frequency change for a spectrometer by Nuclear Magnetic Resonance based on pulse in liquids

    International Nuclear Information System (INIS)

    Avril, Robert

    1974-01-01

    The objective of this work has been, by using the pulsed NMR technique, to study and produce a set for the reception of the nuclear precession signal for a broadband (1-60 MHz) spectrometer. In a first part, the author recalls some fundamental elements of pulsed NMR, and discusses the assessment of the precession signal amplitude which can be expected. Then, he gives a detailed description of the various components of the reception chain (modulator, amplifiers, broadband pre-amplifiers, diode gate, frequency changer, synchronous detection), and, through experimentation and measurements, outlines the ease of implementation and adaptation to measurement conditions

  19. Design, Fabrication, and Testing of a Laboratory-Scale Voloxidation System for Removal of Tritium and Other Volatile Fission Products from Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Spencer, Barry B; DelCul, Guillermo D; Bradley, Eric Craig; Jubin, Robert Thomas; Hylton, Tom D; Collins, Emory D

    2008-01-01

    Advanced nuclear fuel processing methodologies are being demonstrated at the Oak Ridge National Laboratory (ORNL) as part of the Global Nuclear Energy Partnership (GNEP) program. A coupled end-to-end (CETE) research and development (R and D) capability is being installed to provide all primary processing operations, ranging from spent fuel receipt to production of products and waste forms. This R and D capability is designed for small, laboratory-scale throughput and will permit conduct of experiments in the range of 20 kg of spent fuel per year. The head-end processing segment includes single-pin shearing, voloxidation to remove tritium from the fuel before it enters the aqueous based separations systems, cleanup of the cladding hulls for disposition, and transfer of the fuel powder to the dissolution process. This paper describes the voloxidation system design and presents results from the cold checkout of the hardware. Preliminary results of the initial processing campaign with spent fuel is presented as well

  20. Fabrication and Prototyping Lab

    Data.gov (United States)

    Federal Laboratory Consortium — Purpose: The Fabrication and Prototyping Lab for composite structures provides a wide variety of fabrication capabilities critical to enabling hands-on research and...

  1. Fabrication of recyclable superhydrophobic cotton fabrics

    Science.gov (United States)

    Han, Sang Wook; Park, Eun Ji; Jeong, Myung-Geun; Kim, Il Hee; Seo, Hyun Ook; Kim, Ju Hwan; Kim, Kwang-Dae; Kim, Young Dok

    2017-04-01

    Commercial cotton fabric was coated with SiO2 nanoparticles wrapped with a polydimethylsiloxane (PDMS) layer, and the resulting material surface showed a water contact angle greater than 160°. The superhydrophobic fabric showed resistance to water-soluble contaminants and maintained its original superhydrophobic properties with almost no alteration even after many times of absorption-washing cycles of oil. Moreover, superhydrophobic fabric can be used as a filter to separate oil from water. We demonstrated a simple method of fabrication of superhydrophobic fabric with potential interest for use in a variety of applications.

  2. Applications of polymer coatings for the fabrication of copper-based containers for the ultimate disposal of Canada's spent nuclear fuel

    Science.gov (United States)

    Mortley, Aba

    Oxygen-free, phosphorous doped copper containers have been proposed for the storage of the used nuclear fuel bundles as a part of Canada's multi-barrier, adaptive phased management procedure for long term storage of spent nuclear fuel bundles. The spent nuclear fuel disposal system proposed for Canada has been engineered based on the multi-barrier approach intended to minimize the risk that the radioactive materials enter the biosphere. Copper is known to be susceptible to corrosion and it is thought that the simultaneous exposure to aggressive ionizing radiation field and residual heat produced by the spent nuclear fuel and the surrounding groundwater would all challenge the container's integrity. The goal of the present work is to reduce the impact of corrosion in the early stages of emplacement with the addition of a protective coating. Specifically, castor oil based polyurethanes were assessed as coatings and their ability to act as an additional physical barrier in the multi-barrier system mentioned previously. The novelty of this work stems from the use of a naturally derived non-petroleum based material in the form of castor oil as the polyol component. Two types of castor oil polyurethanes were investigated, one based on an aliphatic hexamethylene diisocyanate (HMDI), and the other based on an aromatic 2,4-toluene diisocyanate (TDI). Radiation and saturation tests were conducted using varying conditions. Mixed field ionizing radiation was provided by a SLOWPOKE-2 pool-type nuclear research reactor, up to accumulated doses of 6 MGy at dose rates of 37 kGy h-1 and 55.5 kGy h-1. Weight gain immersion studies, at temperatures of 25° C, 50° C, 70° C, were used to determine the mass uptake of several different solutions. The solutions utilized in the present work included hydrochloric acids of varying pHs, distilled water, and buffered solutions, which simulated chloride and sulphide rich calcium-sodium bicarbonate waters. After being exposed to radiation and

  3. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1976-10-01

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

  4. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U. [IPEN, CNEN-SP, P.O. Box 11049, CEP 05422-970, Sao Paulo (Brazil); Riella, H. G. [Universidade Federal de Santa Catarina, Departamento de Engenharia Quimica, Campus Universitario, Florianopolis, CEP 88040-900 (Brazil)]. e-mail: grsantos@ipen.br

    2008-07-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% {sup 2}35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

  5. Nuclear fuel activities in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Bairiot, H

    1997-12-01

    In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes.

  6. Description of methods for making activation detectors for use in nuclear reactors; Description des procedes de fabrication des detecteurs d'activation utilises dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Barbalat, R; Le Coguie, R; Leger, P; Salon, L; Thierry, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    A brief description of methods currently used for making activation detectors, thin films and various deposits used in nuclear reactors. The thicknesses required vary from about a few tenths of a micron to a few tenths of a millimeter. Different techniques are used for fixing the large variety of elements: rolling, moulding, painting, electrolysis, vacuum deposition, thin films, wires, enamels, protective linings, etc. (authors) [French] Expose succinct des procedes actuellement mis en oeuvre pour la realisation des detecteurs d'activation, feuilles minces et depots divers utilises dans les reacteurs nucleaires. La gamme des epaisseurs necessaires s'etendant approximativement des dixiemes de micrometre aux dixiemes de millimetre. La diversite des elements a fixer justifiant les techniques differentes selon les cas: laminage, moulage, peinture, electrolyse, depot sous vide, couches minces, fils, emaux, revetements protecteurs, etc. (auteurs)

  7. Polymorphous computing fabric

    Science.gov (United States)

    Wolinski, Christophe Czeslaw [Los Alamos, NM; Gokhale, Maya B [Los Alamos, NM; McCabe, Kevin Peter [Los Alamos, NM

    2011-01-18

    Fabric-based computing systems and methods are disclosed. A fabric-based computing system can include a polymorphous computing fabric that can be customized on a per application basis and a host processor in communication with said polymorphous computing fabric. The polymorphous computing fabric includes a cellular architecture that can be highly parameterized to enable a customized synthesis of fabric instances for a variety of enhanced application performances thereof. A global memory concept can also be included that provides the host processor random access to all variables and instructions associated with the polymorphous computing fabric.

  8. The quality challenge for fuel fabrication

    International Nuclear Information System (INIS)

    Lannegrace, J.-P.

    1990-01-01

    Fuel fabrication is a key segment of the nuclear fuel cycle, since safe and economic operation of reactors is highly dependent on the quality of the fuel. Achieving and controlling quality is, therefore, of paramount importance to fuel fabricators dominating nearly every aspect of the business. The quality policy, concepts and assurance system at three French plants are outlined. The need for integrated inspection, process optimization and good employee motivation is stressed. (author)

  9. A column exchange chromatographic procedure for the automated purification of analytical samples in nuclear spent fuel reprocessing and plutonium fuel fabrication

    International Nuclear Information System (INIS)

    Zahradnik, P.; Swietly, H.; Doubek, N.; Bagliano, G.

    1992-11-01

    A Column Exchange Chromatographic procedure using Tri-n-Octyl-Phosphine-Oxide (TOPO) as stationary phase, is in routine use at SAL since 1984 on nuclear spent fuel reprocessing and on Pu product samples, prior to alpha and mass spectrometric analysis. This standard procedure was further on modified in view of its automation in a glove box; the resulting new procedure is described in this paper. Laboratory Robot Compatible (LRC) disposable columns were selected because their dimensions are particularly favorable and reproducible. A less corrosive HNO 3 -HI mixture substituted the former HC1-HI plutonium eluant. The inorganic support of the stationary phase used to test the above mentioned changes was unexpectedly withdrawn from the market so that another support had to be selected and the procedure reoptimized accordingly. The resulting procedure was tested with the robot and validated against the manual procedure taken as reference: the comparison showed that the modified procedure meets the analytical requirements and has the same performance than the original procedure. (author). Refs, figs and tabs

  10. Strategies for CT tissue segmentation for Monte Carlo calculations in nuclear medicine dosimetry

    DEFF Research Database (Denmark)

    Braad, P E N; Andersen, T; Hansen, Søren Baarsgaard

    2016-01-01

    in the ICRP/ICRU male phantom and in a patient PET/CT-scanned with 124I prior to radioiodine therapy. Results: CT number variations body CT examinations at effective CT doses ∼2 mSv. Monte Carlo calculated absorbed doses depended on both the number of media types and accurate......Purpose: CT images are used for patient specific Monte Carlo treatment planning in radionuclide therapy. The authors investigated the impact of tissue classification, CT image segmentation, and CT errors on Monte Carlo calculated absorbed dose estimates in nuclear medicine. Methods: CT errors...

  11. Design, fabrication and installation of irradiation facilities

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Bong Shick; Kim, Y. S.; Lee, C. Y. and others

    1999-03-01

    The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and fabrication of main equipment, licensing and technical review for fuel test loop have been performed during 2 years(1997, 1998) for this project. Following contents are described in the report. - Procurement and fabrication of the equipment, piping for OPS - IPS manufacture - License - Technical review and evaluation of the FTL facility. As besides, as these irradiation facilities will be installed in HANARO, review of safety concern, discussion with KINS for licensing and review ofHANARO interface have been performed respectively. (author)

  12. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle for use in establishing ''as low as practicable'' guides: fabrication of light-water reactor fuels containing plutonium

    International Nuclear Information System (INIS)

    Groenier, W.S.; Blanco, R.E.; Dahlman, R.C.; Finney, B.C.; Kibbey, A.H.; Witherspoon, J.P.

    1975-05-01

    A cost-benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model light-water plutonium recycle reactor fuel fabrication plant, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as practicable'' in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 300 metric tons of LWR plutonium recycle fuel. Additional radwaste treatment equipment is added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitment are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Some of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costsand the radiological doses, detailed calculations, and tabulations are presented in Appendixes A and B. (U.S.)

  13. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle for use in establishing ''as low as practicable'' guides: fabrication of light-water reactor fuel from enriched uranium dioxide

    International Nuclear Information System (INIS)

    Pechin, W.H.; Blanco, R.E.; Dahlman, R.C.; Finney, B.C.; Lindauer, R.B.; Witherspoon, J.P.

    1975-05-01

    A cost-benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model enriched-uranium, light-water reactor (LWR) fuel fabrication plant, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as practicable'' in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment equipment is added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitment are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Some of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. (U.S.)

  14. China's nuclear safety regulatory body: The national nuclear safety administration

    International Nuclear Information System (INIS)

    Zhang Shiguan

    1991-04-01

    The establishment of an independent nuclear safety regulatory body is necessary for ensuring the safety of nuclear installations and nuclear fuel. Therefore the National Nuclear Safety Administration was established by the state. The aim, purpose, organization structure and main tasks of the Administration are presented. At the same time the practical examples, such as nuclear safety regulation on the Qinshan Nuclear Power Plant, safety review and inspections for the Daya Bay Nuclear Power Plant during the construction, and nuclear material accounting and management system in the nuclear fuel fabrication plant in China, are given in order to demonstrate the important roles having been played on nuclear safety by the Administration after its founding

  15. Typical IAEA operations at a fuel fabrication plant

    International Nuclear Information System (INIS)

    Morsy, S.

    1984-01-01

    The IAEA operations performed at a typical Fuel Fabrication Plant are explained. To make the analysis less general the case of Low Enriched Uranium (LEU) Fuel Fabrication Plants is considered. Many of the conclusions drawn from this analysis could be extended to other types of fabrication plants. The safeguards objectives and goals at LEU Fuel Fabrication Plants are defined followed by a brief description of the fabrication process. The basic philosophy behind nuclear material stratification and the concept of Material Balance Areas (MBA's) and Key Measurement Points (KMP's) is explained. The Agency operations and verification methods used during physical inventory verifications are illustrated

  16. Iran's nuclear programme

    International Nuclear Information System (INIS)

    Boureston, J.; Marvin, B.

    2004-01-01

    Iran's nuclear program is discussed, activity of enterprises connected with the nuclear industry of the country is evaluated. IAEA initiated inspection of some industrial sites with the aim of data acquisition about nuclear developments of Iran. Uranium ore mining and reducing to small size, uranium conversion, uranium enrichment, fabrication of nuclear fuel, production of plutonium: plant of heavy water production, spent fuel reprocessing are discussed [ru

  17. Chilean fuel elements fabrication progress report

    International Nuclear Information System (INIS)

    Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

    1993-01-01

    Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

  18. Fuel fabrication and post-irradiation examination

    Energy Technology Data Exchange (ETDEWEB)

    Venter, P J; Aspeling, J C [Atomic Energy Corporation of South Africa Ltd., Pretoria (South Africa)

    1990-06-01

    This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

  19. Fuel fabrication and post-irradiation examination

    International Nuclear Information System (INIS)

    Venter, P.J.; Aspeling, J.C.

    1990-01-01

    This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

  20. Final Report on Design, Fabrication and Test of HANARO Instrumented Capsule (07M-13N) for the Researches of Irradiation Performance of Parts of X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of a X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens of control rod materials of AP1000 reactor requested by Westinghouse Co. were inserted in this capsule as a preliminary irradiation test and Polyimide specimens requested by Hanyang university were also inserted. 463 specimens such as buckling and spring test specimens of cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718, Polyimide, Ag and Ag-In-Cd alloys were placed in the capsule. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of neutron fluence monitors installed in the capsule. A new friction welded tube between STS304 and Al1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during a capsule cutting process in HANARO. The capsule was irradiated for 95.19 days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 230 {approx} 420 .deg. C. The specimens were irradiated up to a maximum fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1.0MeV) and the dpa of the irradiated specimens were evaluated as 1.21 {approx} 1.97. The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell and the obtained results will be very valuable for the related researches of the users.

  1. South Korea's nuclear fuel industry

    International Nuclear Information System (INIS)

    Clark, R.G.

    1990-01-01

    March 1990 marked a major milestone for South Korea's nuclear power program, as the country became self-sufficient in nuclear fuel fabrication. The reconversion line (UF 6 to UO 2 ) came into full operation at the Korea Nuclear Fuel Company's fabrication plant, as the last step in South Korea's program, initiated in the mid-1970s, to localize fuel fabrication. Thus, South Korea now has the capability to produce both CANDU and pressurized water reactor (PWR) fuel assemblies. This article covers the nuclear fuel industry in South Korea-how it is structures, its current capabilities, and its outlook for the future

  2. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Scurr, I.F.; Silver, J.M.

    1990-01-01

    Australian Nuclear Science and Technology Organization maintains an ongoing assessment of the world's nuclear technology developments, as a core activity of its Strategic Plan. This publication reviews the current status of the nuclear power and the nuclear fuel cycle in Australia and around the world. Main issues discussed include: performances and economics of various types of nuclear reactors, uranium resources and requirements, fuel fabrication and technology, radioactive waste management. A brief account of the large international effort to demonstrate the feasibility of fusion power is also given. 11 tabs., ills

  3. Brazing process in nuclear fuel element fabrication

    International Nuclear Information System (INIS)

    Katam, K.; Sudarsono

    1982-01-01

    The purpose of the brazing process is to join the spacers and pads of fuel pins, so that the process is meant as a soldering technique and not only as a hardening or reinforcing process such as in common brazing purposes. There are some preliminary processes before executing the brazing process such as: materials preparation, sand blasting, brazing metal coating tack welding the spacers and pads on the fuel cladding. The metal brazing used is beryllium in strip form which will be evaporated in vacuum condition to coat the spacers and pads. The beryllium vapor and dust is very hazardous to the workers, so all the line process of brazing needs specials safety protection and equipment to protect the workers and the processing area. Coating process temperature is 2470 deg C with a vacuum pressure of 10 -5 mmHg. Brazing process temperature process is 1060 deg C with a vacuum pressure of 10 -6 mmHg. The brazing process with beryllium coating probably will give metallurgical structural change in the fuel cladding metal at the locations of spacers and pads. The quality of brazing is highly influenced by and is depending on the chemical composition of the metal and the brazing metal, materials preparations, temperature, vacuum pressure, time of coating and brazing process. The quality control of brazing could be performed with methods of visuality geometry, radiography and metallography. (author)

  4. Nuclear shelters

    International Nuclear Information System (INIS)

    Prendergast, L.; Prendergast, P.; Prendergast, E.J.; Prendergast, L.J.

    1982-01-01

    An underground nuclear shelter, fabricated from a series of transverse metal hoop frames connected by longitudinal metal members and plates, is described. The shelter chamber has two hatches in the roof each with a shaft fitted with a shield cover. One shaft houses an air inlet and filter, the other is used for access. Two or more shelter units can be linked. (U.K.)

  5. Fabric based supercapacitor

    International Nuclear Information System (INIS)

    Yong, S; Tudor, M J; Beeby, S P; Owen, J R

    2013-01-01

    Flexible supercapacitors with electrodes coated on inexpensive fabrics by the dipping technique. This paper present details of the design, fabrication and characterisation of fabric supercapacitor. The sandwich structured supercapacitors can achieve specific capacitances of 11.1F/g, area capacitance 105 mF.cm −2 and maintain 95% of the initial capacitance after cycling the device for more than 15000 times

  6. The nuclear age

    International Nuclear Information System (INIS)

    Leclercq, J.

    1986-01-01

    This book is divided into 7 chapters bearing on: (1) Nuclear power: an historical perspective (2) Diversity of reactor designs (3) Safety and the environment (4) Architecture and heavy construction (5) Fabrication and installation (6) Nuclear fuel (7) Working for the electric power industry

  7. Design of the MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Johnson, J.V.; Brabazon, E.J.

    2001-01-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  8. Design of the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.V. [MFFF Technical Manager, U.S. dept. of Energy, Washington, DC (United States); Brabazon, E.J. [MFFF Engineering Manager, Duke Cogema Stone and Webster, Charlotte, NC (United States)

    2001-07-01

    A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

  9. Fabrics in Function

    DEFF Research Database (Denmark)

    Bang, Anne Louise

    2007-01-01

    sensing of fabrics in function. It is proposed that tactile and visual sensing of fabrics is a way to investigate and express emotional utility values. The further purpose is to use experiments with repertory grid models as part of the mapping of the entire research project and also as a basis...

  10. Fabricating architectural volume

    DEFF Research Database (Denmark)

    Feringa, Jelle; Søndergaard, Asbjørn

    2015-01-01

    The 2011 edition of Fabricate inspired a number of collaborations, this article seeks to highlight three of these. There is a common thread amongst the projects presented: sharing the ambition to close the rift between design and fabrication while incorporating structural design aspects early on...

  11. Smart Fabrics Technology Development

    Science.gov (United States)

    Simon, Cory; Potter, Elliott; Potter, Elliott; McCabe, Mary; Baggerman, Clint

    2010-01-01

    Advances in Smart Fabrics technology are enabling an exciting array of new applications for NASA exploration missions, the biomedical community, and consumer electronics. This report summarizes the findings of a brief investigation into the state of the art and potential applications of smart fabrics to address challenges in human spaceflight.

  12. Steel structures for nuclear facilities

    International Nuclear Information System (INIS)

    1993-01-01

    In the guide the requirements concerning design and fabrication of steel structures for nuclear facilities and documents to be submitted to the Finnish Centre for Radiation and Nuclear Safety (STUK) are presented. Furthermore, regulations concerning inspection of steel structures during construction of nuclear facilities and during their operation are set forth

  13. Russian nuclear survey

    International Nuclear Information System (INIS)

    2002-07-01

    This document gives a broad overview of the organization of nuclear activities in the Russian federation: Minatom activities, nuclear park and availability (reactors, performances, export activity), perspectives of development (improvement of safety, age of reactors, new realizations); fuel cycle (uranium production, conversion and enrichment, fuel fabrication, spent fuel reprocessing); wastes management (storage and disposal sites); R and D activities (organizations) and nuclear safety authority. (J.S.)

  14. Optics fabrication technical challenges

    International Nuclear Information System (INIS)

    Chabassier, G.; Ferriou, N.; Lavastre, E.; Maunier, C.; Neauport, J.; Taroux, D.; Balla, D.; Fornerod, J.C.

    2004-01-01

    Before the production of all the LMJ (MEGAJOULE laser) optics, the CEA had to proceed with the fabrication of about 300 large optics for the LIL (laser integration line) laser. Thanks to a fruitful collaboration with high-tech optics companies in Europe, this challenge has been successfully hit. In order to achieve the very tight requirements for cleanliness, laser damage threshold and all the other high demanding fabrication specifications, it has been necessary to develop and to set completely new fabrication process going and to build special outsize fabrication equipment. Through a couple of examples, this paper gives an overview of the work which has been done and shows some of the results which have been obtained: continuous laser glass melting, fabrication of the laser slabs, rapid-growth KDP (potassium dihydrogen phosphate) technology, large diffractive transmission gratings engraving and characterization. (authors)

  15. MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

    Directory of Open Access Journals (Sweden)

    Miroslav Cech

    2016-12-01

    Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

  16. Nuclear power and its fuel cycle

    International Nuclear Information System (INIS)

    Wymer, R.G.

    1986-01-01

    A series of viewgraphs describes the nuclear fuel cycle and nuclear power, covering reactor types, sources of uranium, enrichment of uranium, fuel fabrication, transportation, fuel reprocessing, and radioactive wastes

  17. National cyclotron centre at the Institute for Nuclear Research and Nuclear Energy

    Science.gov (United States)

    Tonev, D.; Goutev, N.; Asova, G.; Artinyan, A.; Demerdjiev, A.; Georgiev, L. S.; Yavahchova, M.; Bashev, V.; Genchev, S. G.; Geleva, E.; Mincheva, M.; Nikolov, A.; Dimitrov, D. T.

    2018-05-01

    An accelerator laboratory is presently under construction in Sofia at the Institute for Nuclear Research and Nuclear Energy. The laboratory will use a TR24 type of cyclotron, which provides a possibility to accelerate a proton beam with an energy of 15 to 24 MeV and current of up to 0.4 mA. An accelerator with such parameters allows to produce a large variety of radioisotopes for development of radiopharmaceuticals. The most common radioisotopes that can be produced with such a cyclotron are PET isotopes like: 11C, 13N, 15O, 18F, 124I, 64Cu, 68Ge/68Ga, and SPECT isotopes like: 123I, 111In, 67Ga, 57Co, 99mTc. Our aim is to use the cyclotron facility for research in the fields of radiopharmacy, radiochemistry, radiobiology, nuclear physics, materials sciences, applied research, new materials and for education in all these fields including nuclear energy. Presently we perform investigations in the fields of target design for production of radioisotopes, shielding and radioprotection, new ion sources etc.

  18. New polymorphous computing fabric

    International Nuclear Information System (INIS)

    Wolinski, Christophe; Gokhale, Maya; McCabe, Kevin P.

    2002-01-01

    This paper introduces a new polymorphous computing Fabric well suited to DSP and Image Processing and describes its implementation on a Configurable System on a Chip (CSOC). The architecture is highly parameterized and enables customization of the synthesized Fabric to achieve high performance for a specific class of application. For this reason it can be considered to be a generic model for hardware accelerator synthesis from a high level specification. Another important innovation is the Fabric uses a global memory concept, which gives the host processor random access to all the variables and instructions on the Fabric. The Fabric supports different computing models including MIMD, SPMD and systolic flow and permits dynamic reconfiguration. We present a specific implementation of a bank of FIR filters on a Fabric composed of 52 cells on the Altera Excalibur ARM running at 33 MHz. The theoretical performance of this Fabric is 1.8 GMACh. For the FIR application we obtain 1.6 GMAC/s real performance. Some automatic tools have been developed like the tool to provide a host access utility and assembler.

  19. The Text of the Agreement of 22 July 1977 between Argentina and the Agency for the Application of Safeguards in Connection with a Contract Concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union Gmbh Hanau (Federal Republic of Germany) for Co-Operation in the Field of Fabrication of Fuel Elements for Peaceful Nuclear Activities

    International Nuclear Information System (INIS)

    1977-01-01

    The text of the Agreement of 22 July 1977 between Argentina and the Agency for the application of safeguards in connection with the Contract of 13 August 1976 concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union GmbH (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities is reproduced in this document for the information of all Members. The Agreement entered into force, pursuant to Section 26, on 22 July 1977.

  20. The Text of the Agreement of 22 July 1977 between Argentina and the Agency for the Application of Safeguards in Connection with a Contract Concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union Gmbh Hanau (Federal Republic of Germany) for Co-Operation in the Field of Fabrication of Fuel Elements for Peaceful Nuclear Activities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-11-30

    The text of the Agreement of 22 July 1977 between Argentina and the Agency for the application of safeguards in connection with the Contract of 13 August 1976 concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union GmbH (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities is reproduced in this document for the information of all Members. The Agreement entered into force, pursuant to Section 26, on 22 July 1977.

  1. Fabrication process of expanded cooling jackets

    International Nuclear Information System (INIS)

    Weber, C.M.

    1980-01-01

    The present invention concerns the fabrication process of heat exchangers and in particular, the fabrication and assembly process of cooling jackets of the system driving the control rods used in nuclear reactors. The cooling jackets are assembled for cooling the stator of a tubular motor displacing the control rods. The fabrication and assembling of the cooling jacket is made up by the following operations: - a sleeve has an inner fluid inlet and outlet ways, - an external socket is fitted to the sleeve, - on the external socket a continuous welding is realized, which join the socket to the sleeve, and define a serie of parallel welded turns, - a pressure is established between the sleeve and the socket by a fluid through the outlet or inlet ways of the sleeve. When the other way is sealed up, the socket expands between the welded turns, and the fluid can pass through the jacket [fr

  2. Design, fabrication and installation of irradiation facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Sung; Lee, C. Y.; Kim, J. Y.; Chi, D. Y.; Kim, S. H.; Ahn, S. H.; Kim, S. J.; Kim, J. K.; Yang, S. H.; Yang, S. Y.; Kim, H. R.; Kim, H.; Lee, K. H.; Lee, B. C.; Park, C.; Lee, C. T.; Cho, S. W.; Kwak, K. K.; Suk, H. C. [and others

    1997-07-01

    The principle contents of this project are to design, fabricate and install the steady-state fuel test loop and non-instrumented capsule in HANARO for nuclear technology development. This project will be completed in 1999, the basic and detail design, safety analysis, and procurement of main equipment for fuel test loop have been performed and also the piping in gallery and the support for IPS piping in reactor pool have been installed in 1994. In the area of non-instrumented capsule for material irradiation test, the fabrication of capsule has been completed. Procurement, fabrication and installation of the fuel test loop will be implemented continuously till 1999. As besides, as these irradiation facilities will be installed in HANARO, review of safety concern, discussion with KINS for licensing and safety analysis report has been submitted to KINS to get a license and review of HANARO interface have been performed respectively. (author). 39 refs., 28 tabs., 21 figs.

  3. Design, fabrication and installation of irradiation facilities

    International Nuclear Information System (INIS)

    Kim, Yong Sung; Lee, C. Y.; Kim, J. Y.; Chi, D. Y.; Kim, S. H.; Ahn, S. H.; Kim, S. J.; Kim, J. K.; Yang, S. H.; Yang, S. Y.; Kim, H. R.; Kim, H.; Lee, K. H.; Lee, B. C.; Park, C.; Lee, C. T.; Cho, S. W.; Kwak, K. K.; Suk, H. C.

    1997-07-01

    The principle contents of this project are to design, fabricate and install the steady-state fuel test loop and non-instrumented capsule in HANARO for nuclear technology development. This project will be completed in 1999, the basic and detail design, safety analysis, and procurement of main equipment for fuel test loop have been performed and also the piping in gallery and the support for IPS piping in reactor pool have been installed in 1994. In the area of non-instrumented capsule for material irradiation test, the fabrication of capsule has been completed. Procurement, fabrication and installation of the fuel test loop will be implemented continuously till 1999. As besides, as these irradiation facilities will be installed in HANARO, review of safety concern, discussion with KINS for licensing and safety analysis report has been submitted to KINS to get a license and review of HANARO interface have been performed respectively. (author). 39 refs., 28 tabs., 21 figs

  4. Fabrication and characterisation of fabric supercapacitor

    OpenAIRE

    Yong, Sheng

    2016-01-01

    Fabric supercapacitor is a flexible electrochemical device for energy storage application. It is designed to power up flexible electronic systems used for, for example, information sensing, data computation and communication. The development of a flexible supercapacitor is important for e-textiles since supercapacitor can achieve higher energy density than a standard parallel plate capacitor and a larger power density compared with a battery. This research area is currently facing barriers on...

  5. Robotic fabrication and inspection for power plants

    International Nuclear Information System (INIS)

    Date, Ranjit

    2002-01-01

    The usage of Robotic Automation is now an integral part of the modern manufacturing systems. Applications in nuclear power plants is no exception. As a matter of fact, as a result of the hazards of radiations for the human workers makes automation of the on-site working highly desirable. This presentation will focus on the broad benefits by use of automation in Power plants. Various processes and technologies for robotic applications in fabrication, maintenance and inspection will be highlighted. The specific technology features for use in nuclear environments will be highlighted

  6. The text of the agreement of 22 July 1977 between Argentina and the Agency for the application of safeguards in connection with a contract concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reactor Brennelement Union GmbH Hanau (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities

    International Nuclear Information System (INIS)

    1995-01-01

    The Agreement between the Republic of Argentina, the Federative Republic of Brazil, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials and the International Atomic Energy Agency for the Application of Safeguards came into force on 4 March 1994. As a result of the coming into force of the aforesaid Agreement for Argentina, the application of safeguards under the Agreement of 22 July 1977 between Argentina and the IAEA for the application of safeguards in connection with a contract concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reactor Brennelement Union GmbH Hanau (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities has been suspended

  7. Mechanical components: fabrication of major reactor structures

    International Nuclear Information System (INIS)

    Nicholson, S.

    1985-01-01

    The paper examines the validity of criticisms of quality assurance of mechanical plant and welded products within major reactor structures, taking into account experience gained on the AGR's. Various constructive recommendations are made aimed at furthering the objectives of quality assurance in the nuclear industry and making it more cost-effective. Current levels of quality related costs in the fabrication industry are provided as a basis for discussion. (U.K.)

  8. Fabrication and utilization of semiconductor radiation detectors

    International Nuclear Information System (INIS)

    Lemos Junior, Orlando Ferreira

    1969-01-01

    This paper describes the assembly of the equipment for the fabrication of Ge-Li drifted detectors and the technique used in the preparation of a Planar detector of 7 cm 2 x 0,5 cm for the Laboratory of the Linear Accelerator at the University of Sao Paulo, as well as the utilization of a 22 cm 3 coaxial detector for the analysis of fission product gamma rays at the Instituto de Engenharia Nuclear, Rio de Janeiro, R J, Brazil. (author)

  9. Nuclear power plant piping prefabrication and assembly

    International Nuclear Information System (INIS)

    Schmidt, H.

    1990-01-01

    The piping design for nuclear power plants projects reveals, at the beginning, a modification through the application of new fabrication techniques for prefabrication and assembly. This report presents a fabrication methodology which aims to minimize the fabrication and assembly costs as well as to improve and assure quality. (Author) [es

  10. Junction and circuit fabrication

    International Nuclear Information System (INIS)

    Jackel, L.D.

    1980-01-01

    Great strides have been made in Josephson junction fabrication in the four years since the first IC SQUID meeting. Advances in lithography have allowed the production of devices with planar dimensions as small as a few hundred angstroms. Improved technology has provided ultra-high sensitivity SQUIDS, high-efficiency low-noise mixers, and complex integrated circuits. This review highlights some of the new fabrication procedures. The review consists of three parts. Part 1 is a short summary of the requirements on junctions for various applications. Part 2 reviews intergrated circuit fabrication, including tunnel junction logic circuits made at IBM and Bell Labs, and microbridge radiation sources made at SUNY at Stony Brook. Part 3 describes new junction fabrication techniques, the major emphasis of this review. This part includes a discussion of small oxide-barrier tunnel junctions, semiconductor barrier junctions, and microbridge junctions. Part 3 concludes by considering very fine lithography and limitations to miniaturization. (orig.)

  11. Experimental Fabrication Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Provides aviation fabrication support to special operations aircraft residing at Fort Eustis and other bases in the United States. Support is also provided to AATD...

  12. Alloy Fabrication Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — At NETL’s Alloy Fabrication Facility in Albany, OR, researchers conduct DOE research projects to produce new alloys suited to a variety of applications, from gas...

  13. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  14. Development of PHWR fuel fabrication in Korea

    International Nuclear Information System (INIS)

    Suh, K.S.; Yang, M.S.; Kim, D.H.; Rim, C.S.

    1988-01-01

    Korea Advanced Energy Research Institute (KAERI) started a research project to develop the PHWR (CANDU) nuclear fuel fabrication technology in 1981. Based on the results of the intensive developmental work, several prototype fuel bundles were fabricated and tested in the Hot Test Loop at KAERI continuously in 1983 and 1984. After that, irradiation test and post-irradiation examination were carried out for two KAERI-made fuel bundles at Chalk River Nuclear Laboratories in Canada in 1984. Since the results of in-pile and out-of-pile tests with prototype fuel bundles proved to be satisfactory, 48 additional fuel bundles were loaded in Wolsung reactor (CANDU) in 1984 and 1985, and all of them were discharged without a defect after excellent performance in the power reactor. In 1985, the Korean government decided that KAERI supplies all the fuel necessary for the Wolsung reactor. For the mass production of nuclear fuel bundle, several process equipment, facilities and automation methods have been improved making use of experience accumulated during research. A quality assurance program was also established, and quality inspection technology was reviewed and improved to fit the mass production. This paper deals with the development experience so far obtained with the design and fabrication of the Korean PHWR fuel

  15. Design and fabrication of NDA standards

    International Nuclear Information System (INIS)

    Long, S.M.; Hsue, S.T.

    1996-01-01

    The Plutonium Facility, TA-55, at Los Alamos National Laboratory is currently producing NDA calibration standards used by various laboratories in the DOE complex. These NIST traceable standards have been produced to calibrate NDA instruments for accountability measurements used for resolving shipper/receiver differences, and for accountability in process residues and process waste. Standards are needed to calibrate various NDA (Non-destructive Assay) instruments such as neutron coincidence counters, gamma-ray counters, and calorimeters. These instruments measure various ranges of nuclear material being produced in the DOE nuclear community. Los Alamos National Laboratory has taken a lead role in fabrication of uranium and plutonium standards, along with other actinides such as neptunium and americium. These standards have been fabricated for several laboratories within the complex. This paper will summarize previous publications detailing the careful planning encompassing components such as precise weighing, destructive analysis, and the use of post fabrication NDA measurements to confirm that the standards meet all preliminary expectations before use in instrument calibration. The paper will also describe the specialized containers, diluents, and the various amount of nuclear materials needed to accommodate the calibration ranges of the instruments

  16. Technical study report on fuel fabrication system

    International Nuclear Information System (INIS)

    Kono, Shusaku; Tanaka, Kenya; Ono, Kiyoshi; Iwasa, Katsuyoshi; Hoshino, Yasushi; Shinkai, Yasuo

    2000-07-01

    The feasibility study of FBR and related fuel cycle is performed for developing the FBR recycle system which ensures safety, economic competitiveness, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation under consistency of FBR reactor and fuel cycle systems. In this study, a conceptual design study and system characteristics evaluation are conducted for fuel fabrication systems of pellet process, vibropack process for oxide and nitride fuel and casting process for metal fuel. Technical issues in each process are also extracted. In 1999 fiscal year, a conceptual design study were conducted for the fuel fabrication plants adopting (1) the short pellet process which simplifies the conventional MOX pellet fabrication processes, (2) vibropack processes of aqueous gelation process, improved RIAR process, improved ANL process and fluoride volatility process, (3) casting processes of injection process, centrifuging process. As a result, attainable perspective was obtained for each fuel fabrication system through the evaluation of apparatuses, layout and facility volume, etc. In each fuel fabrication system, technical issues for practical use were made clear. Hereafter, more detailed study will be performed for each system, and research programs for phase II study will be planned. (author)

  17. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  18. Fabrication of HTTR first loading fuel

    International Nuclear Information System (INIS)

    Kato, S.; Yoshimuta, S.; Hasumi, T.; Sato, K.; Sawa, K.; Suzuki, S.; Mogi, H.; Shiozawa, S.; Tanaka, T.

    2001-01-01

    This paper summarizes the fabrication of the first loading fuel for HTTR, High Temperature engineering Test Reactor constructed by JAERI, Japan Atomic Energy Research Institute. The fuel fabrication started at the HTR fuel facility of NFI, Nuclear Fuel Industries, Ltd., June 1995. 4,770 fuel rods were fabricated through the fuel kernel, coated fuel particle and fuel compaction process, then 150 fuel elements were assembled in the reactor building December 1997. Fabrication technology for the fuel was established through a lot of R and D activities and fabrication experience of irradiation examination samples spread over about 30 years. Most of all, very high quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and the automatic compaction process. As for the inspection technology, the development of the automatic measurement equipment for coated layer thickness of a coated fuel particle and uranium content of a fuel compact contributed to the higher reliability and rationalization of the inspection process. The data processing system for the fabrication and quality control, which was originally developed by NFI, made possible not only quick feedback of statistical quality data to the fabrication processes, but also automatic document preparation, such as inspection certificates and accountability control reports. The quality of the first loading fuel fully satisfied the design specifications for the fuel. In particular, average bare uranium fraction and SiC defective fraction of fuel compacts were 2x10 -6 and 8x10 -5 , respectively. According to the preceding irradiation examinations being performed at JMTR, Japan Materials Testing Reactor of JAERI, the specimen sampled from the first loading fuel shows good irradiation performance. (author)

  19. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    1998-05-01

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  20. Nuclear technology international 1987

    International Nuclear Information System (INIS)

    Geary, Neville

    1987-01-01

    A total of 59 articles cover a wide range of subjects within the scope of nuclear power generation. The first 13 are concerned with the design and construction of nuclear reactors - PWRs, AGRs, Magnox reactors, fast reactors. The final article in this section is on reactor decommissioning. The next 33 papers all concern services to the nuclear power industry. These include the supply of uranium, uranium enrichment, fuel fabrication, reprocessing, spent fuel storage, robotics and remote handling and radioactive waste disposal. The 13 articles in the safety and public acceptability section concern fears over the Chernobyl accident, safety aspects of nuclear power including risk assessment, fire protection, quality assurance, earthquake tolerance, non-proliferation of nuclear weapons and finally, general problems of balancing advances in nuclear technology and economic desirability against a lack of public confidence in the industry. All reactor and fuel types are represented. Most of the articles concern nuclear power in Europe or North America. All are indexed separately. (UK)

  1. LLNL/YMP Waste Container Fabrication and Closure Project

    International Nuclear Information System (INIS)

    1990-10-01

    The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) Program is studying Yucca Mountain, Nevada as a suitable site for the first US high-level nuclear waste repository. Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing and developing the waste package for the permanent storage of high-level nuclear waste. This report is a summary of the technical activities for the LLNL/YMP Nuclear Waste Disposal Container Fabrication and Closure Development Project. Candidate welding closure processes were identified in the Phase 1 report. This report discusses Phase 2. Phase 2 of this effort involved laboratory studies to determine the optimum fabrication and closure processes. Because of budget limitations, LLNL narrowed the materials for evaluation in Phase 2 from the original six to four: Alloy 825, CDA 715, CDA 102 (or CDA 122) and CDA 952. Phase 2 studies focused on evaluation of candidate material in conjunction with fabrication and closure processes

  2. Understanding core conductor fabrics

    International Nuclear Information System (INIS)

    Swenson, D E

    2011-01-01

    ESD Association standard test method ANSI/ESD STM2.1 - Garments (STM2.1), provides electrical resistance test procedures that are applicable for materials and garments that have surface conductive or surface dissipative properties. As has been reported in other papers over the past several years 1 fabrics are now used in many industries for electrostatic control purposes that do not have surface conductive properties and therefore cannot be evaluated using the procedures in STM2.1 2 . A study was conducted to compare surface conductive fabrics with samples of core conductor fibre based fabrics in order to determine differences and similarities with regards to various electrostatic properties. This work will be used to establish a new work item proposal within WG-2, Garments, in the ESD Association Standards Committee in the USA.

  3. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  4. Fabrication of multilayer nanowires

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Jasveer, E-mail: kaurjasveer89@gmail.com; Singh, Avtar; Kumar, Davinder [Department of Physics, Punjabi University Patiala, 147002, Punjab (India); Thakur, Anup; Kaur, Raminder, E-mail: raminder-k-saini@yahoo.com [Department of Basic and Applied Sciences, Punjabi University Patiala, 147002, Punjab (India)

    2016-05-06

    Multilayer nanowires were fabricated by potentiostate ectrodeposition template synthesis method into the pores of polycarbonate membrane. In present work layer by layer deposition of two different metals Ni and Cu in polycarbonate membrane having pore size of 600 nm were carried out. It is found that the growth of nanowires is not constant, it varies with deposition time. Scanning electron microscopy (SEM) is used to study the morphology of fabricated multilayer nanowires. An energy dispersive X-ray spectroscopy (EDS) results confirm the composition of multilayer nanowires. The result shows that multilayer nanowires formed is dense.

  5. MOX Fabrication Isolation Considerations

    Energy Technology Data Exchange (ETDEWEB)

    Eric L. Shaber; Bradley J Schrader

    2005-08-01

    This document provides a technical position on the preferred level of isolation to fabricate demonstration quantities of mixed oxide transmutation fuels. The Advanced Fuel Cycle Initiative should design and construct automated glovebox fabrication lines for this purpose. This level of isolation adequately protects the health and safety of workers and the general public for all mixed oxide (and other transmutation fuel) manufacturing efforts while retaining flexibility, allowing parallel development and setup, and minimizing capital expense. The basis regulations, issues, and advantages/disadvantages of five potential forms of isolation are summarized here as justification for selection of the preferred technical position.

  6. Fabrication of multilayer nanowires

    International Nuclear Information System (INIS)

    Kaur, Jasveer; Singh, Avtar; Kumar, Davinder; Thakur, Anup; Kaur, Raminder

    2016-01-01

    Multilayer nanowires were fabricated by potentiostate ectrodeposition template synthesis method into the pores of polycarbonate membrane. In present work layer by layer deposition of two different metals Ni and Cu in polycarbonate membrane having pore size of 600 nm were carried out. It is found that the growth of nanowires is not constant, it varies with deposition time. Scanning electron microscopy (SEM) is used to study the morphology of fabricated multilayer nanowires. An energy dispersive X-ray spectroscopy (EDS) results confirm the composition of multilayer nanowires. The result shows that multilayer nanowires formed is dense.

  7. San Rafael mining and fabrication complex today

    International Nuclear Information System (INIS)

    Navarra, Pablo; Aldebert, Sergio R.

    2005-01-01

    In Mendoza province, 35 km West San Rafael city, is located a CNEA installation for uranium ore extraction and concentration: the San Rafael Mining and Fabrication Complex. By the middle of the nineties, as a consequence of the very low prices of uranium concentrate in the international market and of the high internal production costs, uranium extraction was stopped. To day, the international price of the concentrate had a very important increase and the Government has decided the completion of the Atucha II Nuclear Power Station construction. Moreover, studies have been started for new nuclear power plants. In such circumstances the reactivation of the Complex will make sure the uranium supply for our nuclear power stations, contributing to the improvement of the energy generation mix in our country. (author) [es

  8. Nuclear power generation and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Walske, C.

    1978-01-01

    In the future outlook around year 2000 of nuclear power, thought must be given to fuel reprocessing and plutonium utilization. The adverse utilization of plutonium may be prevented by the means balanced with its economical value. As the method of less cost with lower effect of nonproliferation, combination of fuel reprocessing and fuel fabrication facilities and mixed plutonium/uranium processing are possible. As the method of more cost with higher effect of nonproliferation the maintenance of high radioactivity and inaccessibility of plutonium is conceivable. As for the agreeable methods in 2000, seven principles may be mentioned, such as the dependence upon the agreements among major nations and upon nuclear exporting countries. These are still inadequate, however. What is important is to provide with the sufficient safeguards to countries concerned to negate the need for nuclear weapons. Efforts are then necessary for leading nuclear countries to extend aids to other nuclear-oriented countries. (Mori, K.)

  9. Text-Fabric

    NARCIS (Netherlands)

    Roorda, Dirk

    2016-01-01

    Text-Fabric is a Python3 package for Text plus Annotations. It provides a data model, a text file format, and a binary format for (ancient) text plus (linguistic) annotations. The emphasis of this all is on: data processing; sharing data; and contributing modules. A defining characteristic is that

  10. PIGMI mechanical fabrication

    International Nuclear Information System (INIS)

    Hart, V.E.

    1976-01-01

    A prime goal of the mechanical design effort associated with the PIGMI (Pion Generator for Medical Irradiations) program is to investigate new materials and fabrication techniques in an effort to obtain increased machine efficiency and reliability at a reasonable cost. The following discussion deals with the modeling program that LASL is pursuing for 450-MHz and 1350-MHz PIGMI development. (author)

  11. Micromechanical Structures Fabrication; FINAL

    International Nuclear Information System (INIS)

    Rajic, S

    2001-01-01

    Work in materials other than silicon for MEMS applications has typically been restricted to metals and metal oxides instead of more ''exotic'' semiconductors. However, group III-V and II-VI semiconductors form a very important and versatile collection of material and electronic parameters available to the MEMS and MOEMS designer. With these materials, not only are the traditional mechanical material variables (thermal conductivity, thermal expansion, Young's modulus, etc.) available, but also chemical constituents can be varied in ternary and quaternary materials. This flexibility can be extremely important for both friction and chemical compatibility issues for MEMS. In addition, the ability to continually vary the bandgap energy can be particularly useful for many electronics and infrared detection applications. However, there are two major obstacles associated with alternate semiconductor material MEMS. The first issue is the actual fabrication of non-silicon micro-devices and the second impediment is communicating with these novel devices. We have implemented an essentially material independent fabrication method that is amenable to most group III-V and II-VI semiconductors. This technique uses a combination of non-traditional direct write precision fabrication processes such as diamond turning, ion milling, laser ablation, etc. This type of deterministic fabrication approach lends itself to an almost trivial assembly process. We also implemented a mechanical, electrical, and optical self-aligning hybridization technique for these alternate-material MEMS substrates

  12. Fabrication activity for nanophotonics

    DEFF Research Database (Denmark)

    Malureanu, Radu; Chung, Il-Sug; Carletti, Luca

    We present the fabrication and characterization of new structures and materials to be used in nanophotonics. The first structure presented is a fractal metallic metasurface designed to be used as a high-sensitivity sensor for 810nm wavelength. A second structure is a high index contrast grating...

  13. Selection of engineering materials and fabrication of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Patriarca, P.

    1975-01-01

    Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

  14. Status report, canister fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Emilsson, Goeran [CSM Materialteknik AB, Linkoeping (Sweden)

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  15. Status report, canister fabrication

    International Nuclear Information System (INIS)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika; Emilsson, Goeran

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  16. India's nuclear program

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    India made an early commitment to being as self-sufficient as possible in nuclear energy and has largely achieved that goal. The country operates eight nuclear reactors with a total capacity of 1,304 MWe, and it remains committed to an aggressive growth plan for its nuclear industry, with six reactors currently under construction, and as many as twelve more planned. India also operates several heavy water production facilities, fabrication facilities, reprocessing works, and uranium mines and mills. Due to India's decision not to sign the Treaty on the Non-Proliferation of Nuclear Weapons (NPT), the country has had to develop nearly all of its nuclear industry and infrastructure domestically. Overall, India's nuclear power program is self-contained and well integrated, with plans to expand to provide up to ten percent of the country's electrical generating capacity

  17. Homemade nuclear bomb syndrome

    International Nuclear Information System (INIS)

    Meyer, W.; Loyalka, S.K.; Nelson, W.E.; Williams, R.W.

    1977-01-01

    With the publication of Nuclear Theft: Risks and Safeguards by Willrich and Taylor, significant attention has been focused by the media and the public on the possibility of fissile materials being stolen by a terrorist organization and diverted to the actual building, or the threat of building, of a nuclear explosive device. The implication has been created that one or several relatively inexperienced individuals could obtain the materials necessary and fabricate a low-yield nuclear explosive. This article examines these contentions in some detail. The safeguards and use-denial methods presently used in the nuclear fuel cycle are considered and the difficulties they present in obtaining significant amounts of strategic nuclear materials are examined. The characteristics of reactor-grade plutonium are discussed, and the difficulties associated with the assembly of an efficient nuclear explosive device are outlined

  18. Fabrication experiments for large helix heat exchangers

    International Nuclear Information System (INIS)

    Burgsmueller, P.

    1978-01-01

    The helical tube has gained increasing attention as a heat transfer element for various kinds of heat exchangers over the last decade. Regardless of reactor type and heat transport medium, nuclear steam generators of the helix type are now in operation, installlation, fabrication or in the project phase. As a rule, projects are based on the extrapolation of existing technologies. In the particlular case of steam generators for HTGR power stations, however, existing experience is with steam generators of up to about 2 m diameter whereas several projects involve units more than twice as large. For this reason it was felt that a fabrication experiment was necessary in order to verify the feasibility of modern steam generator designs. A test rig was erected in the SULZER steam generator shops at Mantes, France, and skilled personnel and conventional production tools were employed in conducting experiments relating to the coiling, handling and threading of large helices. (Auth.)

  19. Inspection of NFT-type cask fabrication

    International Nuclear Information System (INIS)

    Takani, M.; Umegaki, O.

    1998-01-01

    NFT-type cask has been developed to transport the high burn-up spent fuel from Japanese nuclear power stations to the reprocessing plant of Japan Nuclear Fuel Limited which is under construction in Rokkasho-mura, Aomori prefecture. NFT placed orders of 53 casks to 5 fabricators in Japan and overseas, and these casks have been fabricated since 1994. There are two types of NFT-type casks for PWR spent fuel and four types of NFT-type cask for BWR spent fuel. These are designed in consideration of the number of spent fuels accommodated into each type of casks and the handling conditions at domestic nuclear power stations. According to Japanese notification, it is required to be confirmed by competent authority that casks are manufactured in accordance with approved designs. Furthermore, additional tests are performed such as through-gauge test for basket and pressure test on the shielding material space to ensure the performance of cask by NFT other than items inspected by the competent authority. In order to enhance maintainability of casks, replacement parts such as bolts and valves are shared as much as possible. (authors)

  20. Les fabricants de superlourds

    CERN Multimedia

    Thivent, Viviane

    2006-01-01

    Who said that russian science was not competitive any more? Today, the nuclear physics laboratory in Dubna asserts the record of the heaviest chimical element ever manufactured: 118 protons. (4 pages)

  1. Nuclear law - Nuclear safety

    International Nuclear Information System (INIS)

    Pontier, Jean-Marie; Roux, Emmanuel; Leger, Marc; Deguergue, Maryse; Vallar, Christian; Pissaloux, Jean-Luc; Bernie-Boissard, Catherine; Thireau, Veronique; Takahashi, Nobuyuki; Spencer, Mary; Zhang, Li; Park, Kyun Sung; Artus, J.C.

    2012-01-01

    This book contains the contributions presented during a one-day seminar. The authors propose a framework for a legal approach to nuclear safety, a discussion of the 2009/71/EURATOM directive which establishes a European framework for nuclear safety in nuclear installations, a comment on nuclear safety and environmental governance, a discussion of the relationship between citizenship and nuclear, some thoughts about the Nuclear Safety Authority, an overview of the situation regarding the safety in nuclear waste burying, a comment on the Nome law with respect to electricity price and nuclear safety, a comment on the legal consequences of the Fukushima accident on nuclear safety in the Japanese law, a presentation of the USA nuclear regulation, an overview of nuclear safety in China, and a discussion of nuclear safety in the medical sector

  2. Automated breeder fuel fabrication

    International Nuclear Information System (INIS)

    Goldmann, L.H.; Frederickson, J.R.

    1983-01-01

    The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures

  3. Study of internal exposure to uranium compounds in fuel fabrication plants in Brazil; Estudo da exposicao interna a compostos de uranio na fabricacao do elemento combustivel nuclear no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Maristela Souza

    2006-07-01

    The International Commission on Radiological Protection (ICRP) Publication 66 and Supporting Guidance 3) strongly recommends that specific information on lung retention parameters should be used in preference to default values wherever appropriate, for the derivation of effective doses and for bioassay interpretation of monitoring data. A group of 81 workers exposed to UO{sub 2} at the fuel fabrication facility in Brazil was selected to evaluate the committed effective dose. The workers were monitored for determination of uranium content in the urinary and faecal excretion. The contribution of intakes by ingestion and inhalation were assessed on the basis of the ratios of urinary to fecal excretion. For the selected workers it was concluded that inhalation dominated intake. According to ICRP 66, uranium oxide is classified as insoluble Type S compound. The ICRP Supporting Guidance 3 and some recent studies have recommended specific lung retention parameters to UO{sub 2}. The solubility parameters of the uranium oxide compound handled by the workers at the fuel fabrication facility in Brazil was evaluated on the basis of the ratios of urinary to fecal excretion. Excretion data were corrected for dietary intakes. This paper will discuss the application of lung retention parameters recommended by the ICRP models to these data and also the dependence of the effective committed dose on the lung retention parameters. It will also discuss the problems in the interpretation of monitoring results, when the worker is exposed to several uranium compounds of different solubilities. (author)

  4. Nuclear power in Pakistan

    International Nuclear Information System (INIS)

    Siddiqui, Z.H.; Qureshi, I.H.

    2005-01-01

    Pakistan started its nuclear power program by installing a 137 M We Canadian Deuterium Reactor (Candu) at Karachi in 1971 which became operational in 1972. The post-contract technical support for the Karachi Nuclear Power Plant (KANUPP) was withdrawn by Canada in 196 as a consequence of Indian nuclear device test in 1974. In spite of various difficulties PAEC resolved to continue to operate KANUPP and started a process for the indigenous fabrication of spare parts and nuclear fuel. The first fuel bundle fabricated in Pakistan was loaded in the core in 1980. Since then KANUPP has been operating on the indigenously fabricated fuel. The plant computer systems and the most critical instrumentation and Control system were also replaced with up-to date technology. In 2002 KANUPP completed its original design life of 30 year. A program for the life extension of the plant had already been started. The second nuclear power plant of 300 M We pressurized water reactor purchased from China was installed in Chashma in 1997, which started commercial operations in 2001. Another unit of 300 M We will be installed at Chashma in near future. These nuclear power plants have been operating under IAEA safeguards agreements. PAEC through the long-term performance of the two power plants has demonstrated its competence to safely and successfully operate and maintain nuclear power plants. Pakistan foresees an increasingly important and significant share of nuclear power in the energy sector. The Government has recently allocated a share of 8000 MWe for nuclear energy in the total energy scenario of Pakistan by the year 2025. (author)

  5. Intentionally fabricated autobiographical memories

    OpenAIRE

    Justice, LV; Morrison, CM; Conway, MA

    2017-01-01

    Participants generated both autobiographical memories (AMs) that they believed to be true and intentionally fabricated autobiographical memories (IFAMs). Memories were constructed while a concurrent memory load (random 8-digit sequence) was held in mind or while there was no concurrent load. Amount and accuracy of recall of the concurrent memory load was reliably poorer following generation of IFAMs than following generation of AMs. There was no reliable effect of load on memory generation ti...

  6. Colored fused filament fabrication

    OpenAIRE

    Song, Haichuan; Lefebvre, Sylvain

    2017-01-01

    Filament fused fabrication is the method of choice for printing 3D models at low cost, and is the de-facto standard for hobbyists, makers and schools. Unfortunately, filament printers cannot truly reproduce colored objects. The best current techniques rely on a form of dithering exploiting occlusion, that was only demonstrated for shades of two base colors and that behaves differently depending on surface slope. We explore a novel approach for 3D printing colored objects, capable of creating ...

  7. Advanced fuel fabrication

    International Nuclear Information System (INIS)

    Bernard, H.

    1989-01-01

    This paper deals with the fabrication of advanced fuels, such as mixed oxides for Pressurized Water Reactors or mixed nitrides for Fast Breeder Reactors. Although an extensive production experience exists for the mixed oxides used in the FBR, important work is still needed to improve the theoretical and technical knowledge of the production route which will be introduced in the future European facility, named Melox, at Marcoule. Recently, the feasibility of nitride fuel fabrication in existing commercial oxide facilities was demonstrated in France. The process, based on carbothermic reduction of oxides with subsequent comminution of the reaction product, cold pressing and sintering provides (U, Pu)N pellets with characteristics suitable for irradiation testing. Two experiments named NIMPHE 1 and 2 fabricated in collaboration with ITU, Karlsruhe, involve 16 nitride and 2 carbide pins, operating at a linear power of 45 and 73 kW/m with a smear density of 75-80% TD and a high burn-up target of 15 at%. These experiments are currently being irradiated in Phenix, at Marcoule. (orig.)

  8. Applications of ion plating in metals fabrication

    International Nuclear Information System (INIS)

    Bell, R.T.; Thompson, J.C.

    1974-01-01

    Use of ion plating at the Oak Ridge Y-12 Plant to solve problems encountered in metals fabrication and processing are discussed. Three typical areas are covered. The first is the use of strike coats on various substrates for subsequent electrodeposition. The second area in which ion plating is shown to contribute to a process is in cold welding or room temperature bonding of metals. The third application involves plating U to promote safe handling, fission-product retention, and corrosion protection in nuclear reactors

  9. Automated fuel fabrication- a vision comes true

    International Nuclear Information System (INIS)

    Hemantha Rao, G.V.S.; Prakash, M.S.; Setty, C.R.P.; Gupta, U.C.

    1997-01-01

    When New Uranium Fuel Assembly Project at Nuclear Fuel Complex (NFC) begins production, its operator will have equipment provided with intramachine handling systems working automatically by pressing a single button. Additionally simple low cost inter machine handling systems will further help in critical areas. All these inter and intra machine handling systems will result in improved reliability, productivity and quality. The fault diagnostics, mimics and real time data acquisition systems make the plant more operator friendly. The paper deals with the experience starting from layout, selection of product carriers, different handling systems, the latest technology and the integration of which made the vision on automation in fuel fabrication come true. (author)

  10. Radiation exposure doses of employees in reactor facilities for test and research and under research and development stages, and in facilities for nuclear fuel refining, fabrication, reprocessing and usage

    International Nuclear Information System (INIS)

    1980-01-01

    (1) Radiation exposure doses in reactor facilities. The owners of reactor facilities are obliged by law to control the radiation exposure doses of the employees below the permissible levels. The data based on the reports made in this connection are given in tables for the fiscal year 1978 (from April 1978 to March 1979). It was revealed that the radiation exposure doses of the employees were far below the permissible levels. The distributions of exposure doses in Japan Atomic Energy Research Institute, Power Reactor and Nuclear Fuel Development Corporation and so on are presented for the whole year and the respective quarters. (2) Radiation exposure doses in facilities for nuclear fuel. The owners are similarly obliged to control radiation exposure. The data in this connection are given, and the doses were far below the permissible levels. The distributions in the private enterprises and so on are presented for the whole year. (J.P.N.)

  11. Standard model for safety analysis report of fuel fabrication plants

    International Nuclear Information System (INIS)

    1980-09-01

    A standard model for a safety analysis report of fuel fabrication plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.) [pt

  12. Present state and problems of uranium fuel fabrication businesses

    International Nuclear Information System (INIS)

    Yuki, Akio

    1981-01-01

    The businesses of uranium fuel fabrication converting uranium hexafluoride to uranium dioxide powder and forming fuel assemblies are the field of most advanced industrialization among nuclear fuel cycle industries in Japan. At present, five plants of four companies engage in this business, and their yearly sales exceeded 20 billion yen. All companies are planning the augmentation of installation capacity to meet the growth of nuclear power generation. The companies of uranium fuel fabrication make the nuclear fuel of the specifications specified by reactor manufacturers as the subcontractors. In addition to initially loaded fuel, the fuel for replacement is required, therefore the demand of uranium fuel is relatively stable. As for the safety of enriched uranium flowing through the farbicating processes, the prevention of inhaling uranium powder by workers and the precaution against criticality are necessary. Also the safeguard measures are imposed so as not to convert enriched uranium to other purposes than peacefull ones. The strict quality control and many times of inspections are carried out to insure the soundness of nuclear fuel. The growth of the business of uranium fuel fabrication and the regulation of the businesses by laws are described. As the problems for the future, the reduction of fabrication cost, the promotion of research and development and others are pointed out. (Kako, I.)

  13. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

    2002-01-01

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  14. Highlights of 50 years of nuclear fuel development

    International Nuclear Information System (INIS)

    Simnad, M.T.

    1989-01-01

    The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel-cycle costs and reliable performance in the fuel elements are discussed in the historical context. 10 refs

  15. Highlights of 50 years of nuclear fuels developments

    International Nuclear Information System (INIS)

    Simnad, M.T.

    1989-01-01

    The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel cycle costs and reliable performance in the fuel elements are discussed in the historical context

  16. Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

    1983-01-01

    The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

  17. Projected developments in the US and European LWR fabrication market

    International Nuclear Information System (INIS)

    Anderson, C.K.; Varley, G.

    1996-01-01

    There are several important factors influencing change in the fuel fabrication supply industry in Western Europe today. The most important is the changing supply relationships resulting from lower costs in the USA. With US fabrication prices substantially lower than in Western Europe, a situation which is sustainable based on production costs, a major force for change prevails which will have significant implications for the commercial development of both markets over the next decade. The extent to which supply from the USA will drive the Western Europe market will depend, to some extent, on three other factors: the new US-Euratom agreement on nuclear trade; electricity market deregulation in the European Union (EU); near term MOX fuel fabrication capacity. The main purpose of this paper is to discuss the current situation in the USA and Europe, and the manner in which this will lead to changing fuel fabrication supply dynamics in both markets. (author)

  18. Fracture toughness of fabrication welds investigated by metallographic methods

    International Nuclear Information System (INIS)

    Canonico, D.A.; Crouse, R.S.

    1978-01-01

    The intermediate scale test vessels (ITV) were fabricated to provide test specimens that have sufficient wall thickness and simulate light water reactor pressure vessels. They were fabricated from grades of steel that are similar to those used for nuclear pressure vessels, having a wall thickness of 150mm and the same welded construction. They are, however, considerably smaller in height and diameter than actual vessels. To date, ten vessels have been fabricated and eight have been tested. In preparation for testing the eighth vessel (ITV-8), an extensive investigation was conducted of the toughness properties of the fabrication weld. It was thoroughly characterized and the fracture specimens used in this metallographic investigation were taken from that weld metal

  19. Impact of high (131)I-activities on quantitative (124)I-PET

    DEFF Research Database (Denmark)

    Braad, P E N; Hansen, Søren B.; Høilund-Carlsen, P F

    2015-01-01

    relevant [Formula: see text]I/[Formula: see text]I-activities were performed on a clinical PET/CT-system. Noise equivalent count rate (NECR) curves and quantitation accuracy were determined from repeated scans performed over several weeks on a decaying NEMA NU-2 1994 cylinder phantom initially filled...... [Formula: see text]I-activities was good and image quantification unaffected except at very high count rates. Quantitation accuracy and contrast recovery were uninfluenced at [Formula: see text]I-activities below 1000 MBq, whereas image noise was slightly increased. The NECR peaked at 550 MBq of [Formula......: see text]I, where it was 2.8 times lower than without [Formula: see text]I in the phantom. Quantitative peri-therapeutic [Formula: see text]I-PET is feasible....

  20. Intraocular lens fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Salazar, Mike A. (Albuquerque, NM); Foreman, Larry R. (Los Alamos, NM)

    1997-01-01

    This invention describes a method for fabricating an intraocular lens made rom clear Teflon.TM., Mylar.TM., or other thermoplastic material having a thickness of about 0.025 millimeters. These plastic materials are thermoformable and biocompatable with the human eye. The two shaped lenses are bonded together with a variety of procedures which may include thermosetting and solvent based adhesives, laser and impulse welding, and ultrasonic bonding. The fill tube, which is used to inject a refractive filling material is formed with the lens so as not to damage the lens shape. A hypodermic tube may be included inside the fill tube.

  1. Intraocular lens fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Salazar, M.A.; Foreman, L.R.

    1997-07-08

    This invention describes a method for fabricating an intraocular lens made from clear Teflon{trademark}, Mylar{trademark}, or other thermoplastic material having a thickness of about 0.025 millimeters. These plastic materials are thermoformable and biocompatable with the human eye. The two shaped lenses are bonded together with a variety of procedures which may include thermosetting and solvent based adhesives, laser and impulse welding, and ultrasonic bonding. The fill tube, which is used to inject a refractive filling material is formed with the lens so as not to damage the lens shape. A hypodermic tube may be included inside the fill tube. 13 figs.

  2. Mask fabrication process

    Science.gov (United States)

    Cardinale, Gregory F.

    2000-01-01

    A method for fabricating masks and reticles useful for projection lithography systems. An absorber layer is conventionally patterned using a pattern and etch process. Following the step of patterning, the entire surface of the remaining top patterning photoresist layer as well as that portion of an underlying protective photoresist layer where absorber material has been etched away is exposed to UV radiation. The UV-exposed regions of the protective photoresist layer and the top patterning photoresist layer are then removed by solution development, thereby eliminating the need for an oxygen plasma etch and strip and chances for damaging the surface of the substrate or coatings.

  3. Nuclear energy

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    The Administrative Court of Braunschweig judges the Ordinance on Advance Funding of Repositories (EndlagervorausleistungsVO) to be void. The Hannover Regional Court passes a basic judgment concerning the Gorleben salt mine (repository) and an action for damages. The Federal Administrative Court dismisses actions against part-permits for the Hanau fuel element fabrication plant. The Koblenz Higher Administrative Court dismisses actions against a part-permit for the Muelheim-Kaerlich reactor. 31st Amendment of the German Criminal Code passed, involving amendments in environmental criminal code, defined in the 2nd amendment to the Act on Unlowful Practices Causing Damage to the Environment (UKG); here: Amendments to the law relating to the criminal code and penal provisions governing unlawful conduct in the operation of nuclear installations. (orig.) [de

  4. Alternative Fabrication of Recycling Fast Reactor Metal Fuel

    International Nuclear Information System (INIS)

    Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

    2015-01-01

    Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

  5. Fabrication of zein nanostructure

    Science.gov (United States)

    Luecha, Jarupat

    The concerns on the increase of polluting plastic wastes as well as the U.S. dependence on imported petrochemical products have driven an attention towards alternative biodegradable polymers from renewable resources. Zein protein, a co-product from ethanol production from corn, is a good candidate. This research project aims to increase zein value by adopting nanotechnology for fabricating advanced zein packaging films and zein microfluidic devices. Two nanotechnology approaches were focused: the polymer nanoclay nanocomposite technique where the nanocomposite structures were created in the zein matrix, and the soft lithography and the microfluidic devices where the micro and nanopatterns were created on the zein film surfaces. The polymer nanoclay nanocomposite technique was adopted in the commonly used zein film fabrication processes which were solvent casting and extrusion blowing methods. The two methods resulted in partially exfoliated nanocomposite structures. The impact of nanoclays on the physical properties of zein films strongly depended on the film preparation techniques. The impact of nanoclay concentration was more pronounced in the films made by extrusion blowing technique than by the solvent casting technique. As the processability limitation for the extrusion blowing technique of the zein sample containing hight nanoclay content, the effect of the nanoclay content on the rheological properties of zein hybrid resins at linear and nonlinear viscoelastic regions were further investigated. A pristine zein resin exhibited soft solid like behavior. On the other hand, the zein hybrid with nanoclay content greater than 5 wt.% showed more liquid like behavior, suggesting that the nanoclays interrupted the entangled zein network. There was good correspondence between the experimental data and the predictions of the Wagner model for the pristine zein resins. However, the model failed to predict the steady shear properties of the zein nanoclay nanocomposite

  6. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  7. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  8. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2014-04-17

    The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors. Therefore, the overriding motivation behind the FFC R&D program described in this plan is to foster closer integration between fuel design and fabrication to reduce programmatic risk. These motivating factors are all interrelated, and progress addressing one will aid understanding of the others. The FFC R&D needs fall into two principal categories, 1) baseline process optimization, to refine the existing fabrication technologies, and 2) manufacturing process alternatives, to evaluate new fabrication technologies that could provide improvements in quality, repeatability, material utilization, or cost. The FFC R&D Plan examines efforts currently under way in regard to coupon, foil, plate, and fuel element manufacturing, and provides recommendations for a number of R&D topics that are of high priority but not currently funded (i.e., knowledge gaps). The plan ties all FFC R&D efforts into a unified vision that supports the overall Convert Program schedule in general, and the fabrication schedule leading up to the MP-1 and FSP-1 irradiation experiments specifically. The fabrication technology decision gates and down-selection logic and schedules are tied to the schedule for fabricating the MP-1 fuel plates, which will provide the necessary data to make a final fuel fabrication process down-selection. Because of the short turnaround between MP-1 and the follow-on FSP-1 and MP-2 experiments, the suite of specimen types that will be available for MP-1 will be the same as those available for FSP-1 and MP-2. Therefore, the only opportunity to explore parameter space and alternative processing

  9. An ISO response to the needs of the nuclear industry

    International Nuclear Information System (INIS)

    Buyers, R.

    1982-01-01

    The history and development of the International Standards Organisation's standard DIS6215, Nuclear Power Plants - Quality Assurance, is presented. The aim is to produce quality assurance standards covering the design, procurement, fabrication, construction, commissioning, operation, maintenance and decommissioning of nuclear power plants. The remit was later extended to include fuel fabrication and reprocessing plants. (U.K.)

  10. Nuclear for curious

    International Nuclear Information System (INIS)

    Fontesse, Max

    2015-01-01

    The author aims at proposing a popularisation of nuclear science. In a first part which addresses general aspects and uses of radioactivity, he presents and describes the matter structure, proposes a historical overview, introduces the various forms of nuclear activity (α, β, and γ radioactivity) and natural radioactivity. He explains why radiation can be called ionizing radiation (action of ionizing radiations on matter, comparison between α, β, X, and γ radiations). He presents the various radiological units, addresses issues related to radiation protection (main damages on human body, impact of a radiation on the whole body), and describes various uses of radioactivity (nuclear medicine, dating, heritage conservation, and others). In the second part, the author addresses uses of nuclear energy. He describes artificial radioactivity (nuclear fission, criticality), the various types of nuclear reactors (natural uranium graphite gas, heavy water, pressurised water, boiling water, fast breeders), and the nuclear fuel cycle (mining extraction, refinery, isotopic enrichment, assembly fabrication, cooling in pool, reprocessing, nuclear wastes, MOX fuel). The third part addresses perspectives: EPR, future reactors, nuclear fusion

  11. Intentionally fabricated autobiographical memories.

    Science.gov (United States)

    Justice, Lucy V; Morrison, Catriona M; Conway, Martin A

    2018-02-01

    Participants generated both autobiographical memories (AMs) that they believed to be true and intentionally fabricated autobiographical memories (IFAMs). Memories were constructed while a concurrent memory load (random 8-digit sequence) was held in mind or while there was no concurrent load. Amount and accuracy of recall of the concurrent memory load was reliably poorer following generation of IFAMs than following generation of AMs. There was no reliable effect of load on memory generation times; however, IFAMs always took longer to construct than AMs. Finally, replicating previous findings, fewer IFAMs had a field perspective than AMs, IFAMs were less vivid than AMs, and IFAMs contained more motion words (indicative of increased cognitive load). Taken together, these findings show a pattern of systematic differences that mark out IFAMs, and they also show that IFAMs can be identified indirectly by lowered performance on concurrent tasks that increase cognitive load.

  12. Fabrication progress of the ITER vacuum vessel sector in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.C., E-mail: bckim@nfri.re.kr [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Y.J.; Hong, K.H.; Sa, J.W.; Kim, H.S.; Park, C.K.; Ahn, H.J.; Bak, J.S.; Jung, K.J. [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Park, K.H.; Roh, B.R.; Kim, T.S.; Lee, J.S.; Jung, Y.H.; Sung, H.J.; Choi, S.Y.; Kim, H.G.; Kwon, I.K.; Kwon, T.H. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of)

    2013-10-15

    Highlights: ► Fabrication of ITER vacuum vessel sector full scale mock-up to develop fabrication procedures. ► The welding and nondestructive examination techniques conform to RCC-MR. ► The preparation of real manufacturing of ITER vacuum vessel sector. -- Abstract: As a participant of ITER project, ITER Korea has to supply two ITER vacuum vessel sectors (Sector no. 6, no. 1) of total nine ITER VV sectors. After the procurement arrangement with ITER Organization, ITER Korea made the contract with Hyundai Heavy Industries (HHI) for fabrication of two sectors. Then the start of the manufacturing design was initiated from January 2010. HHI made three real scale R and D mock-ups to verify the critical fabrication feasibility issues on electron beam welding, 3D forming, welding distortion and achievable tolerances. The documentation according to IO and the French nuclear safety regulation requirement, the qualification of welding and nondestructive examination procedures conform to RCC-MR 2007 were proceed in parallel. The mass production of raw material was done after receiving ANB (agreed notified body) verification of product/parts and shop qualification. The manufacturing drawing, manufacturing and inspection plan of VV sector with supporting fabrication procedures are also verified by ANB, accordingly the first cutting and forming of plates for VV sector fabrication started from February 2012. This paper reports the latest fabrication progress of ITER vacuum vessel Sector no. 6 that will be assembled as the first sector in the ITER pit. The overall fabrication route, R and D mock-up fabrication results with forming and welding distortion analysis, qualification status of welding and nondestructive examination (NDE) are also presented.

  13. Project Plan Remote Target Fabrication Refurbishment Project

    International Nuclear Information System (INIS)

    Bell, Gary L.; Taylor, Robin D.

    2009-01-01

    In early FY2009, the DOE Office of Science - Nuclear Physics Program reinstated a program for continued production of 252 Cf and other transcurium isotopes at the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). The FY2009 major elements of the workscope are as follows: (1) Recovery and processing of seven transuranium element targets undergoing irradiation at the High Flux Isotope Reactor (HFIR) at ORNL; (2) Development of a plan to manufacture new targets for irradiation beginning in early- to mid-FY10 to supply irradiated targets for processing Campaign 75 (TRU75); and (3) Refurbishment of the target manufacturing equipment to allow new target manufacture in early FY10 The 252 Cf product from processing Campaign 74 (recently processed and currently shipping to customers) is expected to supply the domestic demands for a period of approximately two years. Therefore it is essential that new targets be introduced for irradiation by the second quarter of FY10 (HFIR cycle 427) to maintain supply of 252 Cf; the average irradiation period is ∼10 HFIR cycles, requiring about 1.5 calendar years. The strategy for continued production of 252 Cf depends upon repairing and refurbishing the existing pellet and target fabrication equipment for one additional target production campaign. This equipment dates from the mid-1960s to the late 1980s, and during the last target fabrication campaign in 2005- 2006, a number of component failures and operations difficulties were encountered. It is expected that following the target fabrication and acceptance testing of the targets that will supply material for processing Campaign 75 a comprehensive upgrade and replacement of the remote hot-cell equipment will be required prior to subsequent campaigns. Such a major refit could start in early FY 2011 and would take about 2 years to complete. Scope and cost estimates for the repairs described herein were developed, and authorization for the work

  14. NCSX Vacuum Vessel Fabrication

    International Nuclear Information System (INIS)

    Viola ME; Brown T; Heitzenroeder P; Malinowski F; Reiersen W; Sutton L; Goranson P; Nelson B; Cole M; Manuel M; McCorkle D.

    2005-01-01

    The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120 o vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1-inch of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120 o vessel segments are formed by welding two 60 o segments together. Each 60 o segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8-inch (20.3 cm) wide spacer ''spool pieces''. The vessel must have a total leak rate less than 5 X 10 -6 t-l/s, magnetic permeability less than 1.02(micro), and its contours must be within 0.188-inch (4.76 mm). It is scheduled for completion in January 2006

  15. Readiness Review of BWXT for Fabrication of AGR 5/6/7 Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sharp, Michelle Tracy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-02-01

    In support of preparations for fabricating compacts for the Advanced Gas Reactor (AGR) fuel qualification irradiation experiments (AGR-5/6/7), Idaho National Laboratory (INL) conducted a readiness review of the BWX Technology (BWXT) procedures, processes, and equipment associated with compact fabrication activities at the BWXT Nuclear Operations Group (BWXT-NOG) facility outside Lynchburg, VirginiaVA. The readiness review used quality assurance requirements taken from the American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008/1a-2009) as a basis to assess readiness to start compact fabrication.

  16. Nuclear Medicine

    Science.gov (United States)

    ... Parents/Teachers Resource Links for Students Glossary Nuclear Medicine What is nuclear medicine? What are radioactive tracers? ... funded researchers advancing nuclear medicine? What is nuclear medicine? Nuclear medicine is a medical specialty that uses ...

  17. Fabrication of superhydrophobic cotton fabrics using crosslinking polymerization method

    Science.gov (United States)

    Jiang, Bin; Chen, Zhenxing; Sun, Yongli; Yang, Huawei; Zhang, Hongjie; Dou, Haozhen; Zhang, Luhong

    2018-05-01

    With the aim of removing and recycling oil and organic solvent from water, a facile and low-cost crosslinking polymerization method was first applied on surface modification of cotton fabrics for water/oil separation. Micro-nano hierarchical rough structure was constructed by triethylenetetramine (TETA) and trimesoyl chloride (TMC) that formed a polymeric layer on the surface of the fabric and anchored Al2O3 nanoparticles firmly between the fabric surface and the polymer layer. Superhydrophobic property was further obtained through self-assembly grafting of hydrophobic groups on the rough surface. The as-prepared cotton fabric exhibited superoleophilicity in atmosphere and superhydrophobicity both in atmosphere and under oil with the water contact angle of 153° and 152° respectively. Water/oil separation test showed that the as-prepared cotton fabric can handle with various oil-water mixtures with a high separation efficiency over 99%. More importantly, the separation efficiency remained above 98% over 20 cycles of reusing without losing its superhydrophobicity which demonstrated excellent reusability in oil/water separation process. Moreover, the as-prepared cotton fabric possessed good contamination resistance ability and self-cleaning property. Simulation washing process test showed the superhydrophobic cotton fabric maintained high value of water contact angle above 150° after 100 times washing, indicating great stability and durability. In summary, this work provides a brand-new way to surface modification of cotton fabric and makes it a promising candidate material for oil/water separation.

  18. Nuclear safety. Seguranca nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Aveline, A [Rio Grande do Sul Univ., Porto Alegre, RS (Brazil). Inst. de Fisica

    1981-01-01

    What is nuclear safety Is there any technical way to reduce risks Is it possible to put them at reasonable levels Are there competitiveness and economic reliability to employ the nuclear energy by means of safety technics Looking for answers to these questions the author describes the sources of potential risks to nuclear reactors and tries to apply the answers to the Brazilian Nuclear Programme. (author).

  19. Competition still fierce in the US fuel fabrication market

    International Nuclear Information System (INIS)

    Schwartz, M.H.

    1990-01-01

    The US market for nuclear fuel fabrication services is characterized by an annual production capacity significantly in excess of both current and anticipated demand. The trends toward longer operating cycle lengths and higher burnup fuel continue in the United States. This, together with the lack of any prospects for new light water reactors coming on line in the US during the next ten years, is expected to hold the annual demand for fuel fabrication services from US LWRs at around 2000t of uranium into the next century. (author)

  20. MOX fuel fabrication technology in J-MOX

    International Nuclear Information System (INIS)

    Osaka, Shuichi; Yoshida, Ryouichi; Yamazaki, Yukiko; Ikeda, Hiroyuki

    2014-01-01

    Japan Nuclear Fuel Ltd. (JNFL) has constructed JNFL MOX Fuel Fabrication Plant (J-MOX) since 2010. The MIMAS process has been introduced in the powder mixing process from AREVA NC considering a lot of MOX fuel fabrication experiences at MELOX plant in France. The feed material of Pu for J-MOX is MH-MOX powder from Rokkasho Reprocessing Plant (RRP) in Japan. The compatibility of the MH-MOX powder with the MIMAS process was positively evaluated and confirmed in our previous study. This paper describes the influences of the UO2 powder and the recycled scrap powder on the MOX pellet density. (author)

  1. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  2. Design, fabrication and transportation of Si rotating device

    International Nuclear Information System (INIS)

    Kimura, Nobuaki; Imaizumi, Tomomi; Takemoto, Noriyuki; Tanimoto, Masataka; Saito, Takashi; Hori, Naohiko; Tsuchiya, Kunihiko; Romanova, Nataliya; Gizatulin, Shamil; Martyushov, Alexandr; Nakipov, Darkhan; Chakrov, Petr; Tanaka, Futoshi; Nakajima, Takeshi

    2012-06-01

    Si semiconductor production by Neutron Transmutation Doping (NTD) method using the Japan Materials Testing Reactor (JMTR) has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency (JAEA) in order to expand industry use. As a part of investigations, irradiation test of silicon ingot for development of NTD-Si with high quality was planned using WWR-K in Institute of Nuclear Physics (INP), National Nuclear Center of Republic of Kazakhstan (NNC-RK) based on one of specific topics of cooperation (STC), Irradiation Technology for NTD-Si (STC No.II-4), on the implementing arrangement between NNC-RK and the JAEA for 'Nuclear Technology on Testing/Research Reactors' in cooperation in research and development in nuclear energy and technology. As for the irradiation test, Si rotating device was fabricated in JAEA, and the fabricated device was transported with irradiation specimens from JAEA to INP-NNC-RK. This report described the design, the fabrication, the performance test of the Si rotating device and transportation procedures. (author)

  3. Invisible nuclear; converting nuclear

    International Nuclear Information System (INIS)

    Park, Jongmoon

    1993-03-01

    This book consists of 14 chapters which are CNN era and big science, from East and West to North and South, illusory nuclear strategy, UN and nuclear arms reduction, management of armaments, advent of petroleum period, the track of nuclear power generation, view of energy, internationalization of environment, the war over water in the Middle East, influence of radiation and an isotope technology transfer and transfer armament into civilian industry, the end of nuclear period and the nuclear Nonproliferation, national scientific and technological power and political organ and executive organ.

  4. Process for fabrication of cermets

    Science.gov (United States)

    Landingham, Richard L [Livermore, CA

    2011-02-01

    Cermet comprising ceramic and metal components and a molten metal infiltration method and process for fabrication thereof. The light weight cermets having improved porosity, strength, durability, toughness, elasticity fabricated from presintered ceramic powder infiltrated with a molten metal or metal alloy. Alumina titanium cermets biocompatible with the human body suitable for bone and joint replacements.

  5. CW RFQ fabrication and engineering

    International Nuclear Information System (INIS)

    Schrage, D.; Young, L.; Roybal, P.

    1998-01-01

    The design and fabrication of a four-vane RFQ to deliver a 100 mA CW proton beam at 6.7 MeV is described. This linac is an Oxygen-Free Electrolytic (OFE) copper structure 8 m in length and was fabricated using hydrogen furnace brazing as the joining technology

  6. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Patarin, L.

    2002-01-01

    This book treats of the different aspects of the industrial operations linked with the nuclear fuel, before and after its use in nuclear reactors. The basis science of this nuclear fuel cycle is chemistry. Thus a recall of the elementary notions of chemistry is given in order to understand the phenomena involved in the ore processing, in the isotope enrichment, in the fabrication of fuel pellets and rods (front-end of the cycle), in the extraction of recyclable materials (residual uranium and plutonium), and in the processing and conditioning of wastes (back-end of the fuel cycle). Nuclear reactors produce about 80% of the French electric power and the Cogema group makes 40% of its turnover at the export. Thus this book contains also some economic and geopolitical data in order to clearly position the stakes. The last part, devoted to the management of wastes, presents the solutions already operational and also the research studies in progress. (J.S.)

  7. Nuclear plants - military hostages

    International Nuclear Information System (INIS)

    Ramberg, B.

    1986-01-01

    Recent events suggest that nuclear reactors could make tempting military or terrorist targets. Despite the care with which most reactors are built, studies document their vulnerability to willful destruction through disruption of coolant mechanisms both inside and outside the containment building. In addition to reactors, such nuclear support facilities as fuel fabrication, reprocessing, and waste storage installations may be attractive military targets. A nuclear bomb which exploded in the vicinity of a reactor could increase its lethal effects by one-third. The implications of this is vulnerability for Middle East stability as well as to other volatile regions. The author suggests several avenues for controlling the dangers: international law, military and civil defense, facility siting, increasing plant safety, and the international management of nuclear energy. 21 references

  8. DRAPING SIMULATION OF WOVEN FABRICS

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, William [General Motors LLC; Jin, Xiaoshi [ESI Group NA; Zhu, Jiang [Optimal CAE; Wathen, Terrence [General Motors LLC; Doroudian2, Mark [ESI Group NA; Aitharaju, Venkat [General Motors LLC

    2016-09-07

    Woven fabric composites are extensively used in molding complex geometrical shapes due to their high conformability compared to other fabrics. Preforming is an important step in the overall process, where the two-dimensional fabric is draped to become the three-dimensional shape of the part prior to resin injection. During preforming, the orientation of the yarns may change significantly compared to the initial orientations. Accurate prediction of the yarn orientations after molding is important for evaluating the structural performance of the final part. This paper presents a systematic investigation of the angle changes during the preform operation for carbon fiber twill and satin weave fabrics. Preforming experiments were conducted using a truncated pyramid mold geometry designed and fabricated at the General Motors Research Laboratories. Predicted results for the yarn orientations were compared with experimental results and good agreement was observed

  9. Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

    CERN Document Server

    Kessler, Günter

    2012-01-01

    Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

  10. Nuclear power generation and nuclear fuel

    International Nuclear Information System (INIS)

    Okajima, Yasujiro

    1985-01-01

    As of June 30, 1984, in 25 countries, 311 nuclear power plants of about 209 million kW were in operation. In Japan, 27 plants of about 19 million kW were in operation, and Japan ranks fourth in the world. The present state of nuclear power generation and nuclear fuel cycle is explained. The total uranium resources in the free world which can be mined at the cost below $130/kgU are about 3.67 million t, and it was estimated that the demand up to about 2015 would be able to be met. But it is considered also that the demand and supply of uranium in the world may become tight at the end of 1980s. The supply of uranium to Japan is ensured up to about 1995, and the yearly supply of 3000 st U 3 O 8 is expected in the latter half of 1990s. The refining, conversion and enrichment of uranium are described. In Japan, a pilot enrichment plant consisting of 7000 centrifuges has the capacity of about 50 t SWU/year. UO 2 fuel assemblies for LWRs, the working of Zircaloy, the fabrication of fuel assemblies, the quality assurance of nuclear fuel, the behavior of UO 2 fuel, the grading-up of LWRs and nuclear fuel, and the nuclear fuel business in Japan are reported. The reprocessing of spent fuel and plutonium fuel are described. (Kako, I.)

  11. Secure Automated Fabrication: an overview of remote breeder fuel fabrication

    International Nuclear Information System (INIS)

    Nyman, D.H.; Graham, R.A.

    1983-10-01

    The Secure Automated Fabrication (SAF) line is an automated, remotely controlled breeder fuel pin fabrication process which is to be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at Hanford and is scheduled for completion in 1984. The SAF line is scheduled for startup in 1987 and will produce mixed uranium-plutonium fuel pins for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). The fabrication line and support systems are described

  12. LEU fuel fabrication program for the RECH-1 reactor. Status report

    International Nuclear Information System (INIS)

    Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

    2000-01-01

    In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

  13. MOX fuel fabrication at AECL

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Jeffs, A.T.

    1995-01-01

    Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is nearing completion. This project is funded by the CANDU Owners' Group (COG). The initial fabrication campaign will consist of the production of thirty-eight 37-element (U,Pu)O 2 bundles containing 0.2 wt% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of the Rehabilitation Project will be given. (author)

  14. Nanomaterials and nanotechnologies in nuclear energy chemistry

    International Nuclear Information System (INIS)

    Shi, W.Q.; Yuan, L.Y.; Li, Z.J.; Lan, J.H.; Zhao, Y.L.; Chai, Z.F.

    2012-01-01

    With the rapid growth of human demands for nuclear energy and in response to the challenges of nuclear energy development, the world's major nuclear countries have started research and development work on advanced nuclear energy systems in which new materials and new technologies are considered to play important roles. Nanomaterials and nanotechnologies, which have gained extensive attention in recent years, have shown a wide range of application potentials in future nuclear energy system. In this review, the basic research progress in nanomaterials and nanotechnologies for advanced nuclear fuel fabrication, spent nuclear fuel reprocessing, nuclear waste disposal and nuclear environmental remediation is selectively highlighted, with the emphasis on Chinese research achievements. In addition, the challenges and opportunities of nanomaterials and nanotechnologies in future advanced nuclear energy system are also discussed. (orig.)

  15. Fabric circuits and method of manufacturing fabric circuits

    Science.gov (United States)

    Chu, Andrew W. (Inventor); Dobbins, Justin A. (Inventor); Scully, Robert C. (Inventor); Trevino, Robert C. (Inventor); Lin, Greg Y. (Inventor); Fink, Patrick W. (Inventor)

    2011-01-01

    A flexible, fabric-based circuit comprises a non-conductive flexible layer of fabric and a conductive flexible layer of fabric adjacent thereto. A non-conductive thread, an adhesive, and/or other means may be used for attaching the conductive layer to the non-conductive layer. In some embodiments, the layers are attached by a computer-driven embroidery machine at pre-determined portions or locations in accordance with a pre-determined attachment layout before automated cutting. In some other embodiments, an automated milling machine or a computer-driven laser using a pre-designed circuit trace as a template cuts the conductive layer so as to separate an undesired portion of the conductive layer from a desired portion of the conductive layer. Additional layers of conductive fabric may be attached in some embodiments to form a multi-layer construct.

  16. Fabrication of cotton fabric with superhydrophobicity and flame retardancy.

    Science.gov (United States)

    Zhang, Ming; Wang, Chengyu

    2013-07-25

    A simple and facile method for fabricating the cotton fabric with superhydrophobicity and flame retardancy is described in the present work. The cotton fabric with the maximal WCA of 160° has been prepared by the covalent deposition of amino-silica nanospheres and the further graft with (heptadecafluoro-1,1,2,2-tetradecyl) trimethoxysilane. The geometric microstructure of silica spheres was measured by transmission electron microscopy (TEM). The cotton textiles before and after treatment were characterized by using scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS). The wetting behavior of cotton samples was investigated by water contact angle measurement. Moreover, diverse performances of superhydrophobic cotton textiles have been evaluated as well. The results exhibited the outstanding superhydrophobicity, excellent waterproofing durability and flame retardancy of the cotton fabric after treatment, offering a good opportunity to accelerate the large-scale production of superhydrophobic textiles materials for new industrial applications. Copyright © 2013 Elsevier Ltd. All rights reserved.

  17. Fabrication of nanowires and nanostructures

    DEFF Research Database (Denmark)

    Mátéfi-Tempfli, Stefan; Mátéfi-Tempfli, M.; Piraux, L.

    2009-01-01

    We report on different approaches that we have adopted and developed for the fabrication of nanowires and nanostructures. Methods based on template synthesis and on self organization seem to be the most promising for the fabrication of nanomaterials and nanostructures due to their easiness and low...... cost. The development of a supported nanoporous alumina template and the possibility of using this template to combine electrochemical synthesis with lithographic methods open new ways for the fabrication of complex nanostructures. The numerous advantages of the supported template and its compatibility...

  18. Quantum Bridge Fabrication Using Photolithography

    International Nuclear Information System (INIS)

    Quinones, R.

    2001-01-01

    The need for high-speed performance electronics in computers integrated circuits and sensors, require the fabrication of low energy consumption diodes. Nano fabrication methods require new techniques and equipment. We are currently developing a procedure to fabricate a diode based on quantum-effects. The device will act like a traditional diode, but the nanometer scale will allow it to reach high speeds without over heating. This new diode will be on a nano-bridge so it can be attenuated by an electromagnetic wave. The goal is to obtain similar current vs voltage response as in a silicon diode

  19. Fabricating Copper Nanotubes by Electrodeposition

    Science.gov (United States)

    Yang, E. H.; Ramsey, Christopher; Bae, Youngsam; Choi, Daniel

    2009-01-01

    Copper tubes having diameters between about 100 and about 200 nm have been fabricated by electrodeposition of copper into the pores of alumina nanopore membranes. Copper nanotubes are under consideration as alternatives to copper nanorods and nanowires for applications involving thermal and/or electrical contacts, wherein the greater specific areas of nanotubes could afford lower effective thermal and/or electrical resistivities. Heretofore, copper nanorods and nanowires have been fabricated by a combination of electrodeposition and a conventional expensive lithographic process. The present electrodeposition-based process for fabricating copper nanotubes costs less and enables production of copper nanotubes at greater rate.

  20. Industrias Nucleares do Brasil in the context of the Brazilian nuclear program; A INB no contexto do programa nuclear brasileiro

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    The activities carried out by Industrias Nucleares Brasileiras (INB) related to the nuclear fuel cycle are described. These activities comprise presently uranium prospecting and processing and fuel elements assembly.Starting in 1997,INB will also perform the reconversion of enriched uranium hexafluoride and the fabrication of fuel pellets.Furthermore, INB produces as well rare earth oxides 2 figs., 1 tab.

  1. Natural fabric of Hildegardia populifolia composites

    CSIR Research Space (South Africa)

    Guduri, BBR

    2006-12-01

    Full Text Available The influence of Hildegardia populofolia fabric content, fabric orientation, sodium hydroxide (NaOH) and silane coupling agent treatment on the surface properties of the fabric, mechanical and fracture properties of Hildegardia populifolia...

  2. Properties of natural fabric Polyalthia cerasoides

    CSIR Research Space (South Africa)

    Jayaramudu, J

    2009-06-01

    Full Text Available of this fabric were compared with those of two natural fabrics reported in the literature. This uniaxial fabric has sufficient tensile modulus and can be used as reinforcement in the development of green composites....

  3. Advanced accounting techniques in automated fuel fabrication facilities

    International Nuclear Information System (INIS)

    Carlson, R.L.; DeMerschman, A.W.; Engel, D.W.

    1977-01-01

    The accountability system being designed for automated fuel fabrication facilities will provide real-time information on all Special Nuclear Material (SNM) located in the facility. It will utilize a distributed network of microprocessors and minicomputers to monitor material movement and obtain nuclear materials measurements directly from remote, in-line Nondestructive Assay instrumentation. As SNM crosses an accounting boundary, the accountability computer will update the master files and generate audit trail records. Mass balance accounting techniques will be used around each unit process step, while item control will be used to account for encapsulated material, and SNM in transit

  4. Automatic welding and cladding in heavy fabrication

    International Nuclear Information System (INIS)

    Altamer, A. de

    1980-01-01

    A description is given of the automatic welding processes used by an Italian fabricator of pressure vessels for petrochemical and nuclear plant. The automatic submerged arc welding, submerged arc strip cladding, pulsed TIG, hot wire TIG and MIG welding processes have proved satisfactory in terms of process reliability, metal deposition rate, and cost effectiveness for low alloy and carbon steels. An example shows sequences required during automatic butt welding, including heat treatments. Factors which govern satisfactory automatic welding include automatic anti-drift rotator device, electrode guidance and bead programming system, the capability of single and dual head operation, flux recovery and slag removal systems, operator environment and controls, maintaining continuity of welding and automatic reverse side grinding. Automatic welding is used for: joining vessel sections; joining tubes to tubeplate; cladding of vessel rings and tubes, dished ends and extruded nozzles; nozzle to shell and butt welds, including narrow gap welding. (author)

  5. [Nuclear theory

    International Nuclear Information System (INIS)

    Haxton, W.

    1990-01-01

    This report discusses research in nuclear physics. Topics covered in this paper are: symmetry principles; nuclear astrophysics; nuclear structure; quark-gluon plasma; quantum chromodynamics; symmetry breaking; nuclear deformation; and cold fusion

  6. Fabrication of integrated metallic MEMS devices

    DEFF Research Database (Denmark)

    Yalcinkaya, Arda Deniz; Ravnkilde, Jan Tue; Hansen, Ole

    2002-01-01

    A simple and complementary metal oxide semiconductor (CMOS) compatible fabrication technique for microelectromechanical (MEMS) devices is presented. The fabrication technology makes use of electroplated metal layers. Among the fabricated devices, high quality factor microresonators are characteri......A simple and complementary metal oxide semiconductor (CMOS) compatible fabrication technique for microelectromechanical (MEMS) devices is presented. The fabrication technology makes use of electroplated metal layers. Among the fabricated devices, high quality factor microresonators...

  7. Foil fabrication for the ROMANO event. Revision 1

    International Nuclear Information System (INIS)

    Romo, J.G. Jr.; Weed, J.W.; Griggs, G.E.; Brown, T.G.; Tassano, P.L.

    1984-01-01

    The Vacuum Processes Lab (VPL), of LLNL's M.E. Dept. - Material Fabrication Division (MFD), conducted various vacuum related support activities for the ROMANO nuclear physics experiment. This report focuses on the foil fabrication activities carried out between July and November 1983 for the ROMANO event. Other vacuum related activities for ROMANO, such as outgassing tests of materials, are covered in separate documentation. VPL was asked to provide 270 coated Parylene foils for the ROMANO event. However, due to the developmental nature of some of the procedures, approximately 400 coated foils were processed. In addition, VPL interacted with MFD's Plastics Shop to help supply Parylene substrates to other organizations (i.e., LBL and commercial vendors) which had also been asked to provide coated foils for ROMANO. The purposes of this report are (A) to document the processes developed and the techniques used to produce the foils, and (B) to suggest future directions. The report is divided into four sections describing: (1) nuclear target foil fabrication, (2) Parylene substrate preparation and production, (3) calibration foil fabrication, and (4) foil and substrate inspections

  8. 76 FR 24018 - Notice of Availability of the Draft Supplemental Environmental Impact Statement for the Nuclear...

    Science.gov (United States)

    2011-04-29

    ... includes two construction options, the Deep Excavation Option and the Shallow Excavation Option. The two... technological capabilities that support nuclear materials handling, processing and fabrication; stockpile...

  9. International conference on Asian nuclear prospects 2010

    International Nuclear Information System (INIS)

    2010-10-01

    The proceedings of the second international conference on Asian nuclear prospects was held during October 10-13, 2010. The topics covered were atomic minerals exploration, fuels and fuel fabrication, fast reactor engineering and technology, thermal reactors, fuel reprocessing, materials development, nuclear instrumentation and electronics, reactor and environmental safety, waste management, knowledge management and HRD. Papers relevant to INIS are indexed separately

  10. Material development for India's nuclear power programme

    Indian Academy of Sciences (India)

    rials with emphasis on development of fabrication routes of zirconium alloys for .... nuclear power programme, which envisages design and construction of thermal breeder ... Production of Hf-free nuclear grade zirconium ..... Later on for pressure tubes specified limit for hydrogen content in the as manufactured condition.

  11. Method for processing spent nuclear reactor fuel

    International Nuclear Information System (INIS)

    Levenson, M.; Zebroski, E.L.

    1981-01-01

    A method and apparatus are claimed for processing spent nuclear reactor fuel wherein plutonium is continuously contaminated with radioactive fission products and diluted with uranium. Plutonium of sufficient purity to fabricate nuclear weapons cannot be produced by the process or in the disclosed reprocessing plant. Diversion of plutonium is prevented by radiation hazards and ease of detection

  12. Quality assurance monitoring during nuclear fuel production in JSC 'TVEL'

    International Nuclear Information System (INIS)

    Filimonov, G.; Tchirkov, V.

    2000-01-01

    The paper describes Quality Assurance (QA) monitoring during fabrication of nuclear fuel in Russian Federation. Joint Stock Company 'TVEL', natural state monopoly of the type of holding that fabricates and supplies nuclear fuel for the NPPs of Russia, CIS and Europe, incorporates the major enterprises of the nuclear fuel cycle including JSC 'Mashinostroitelny zavod', Electrostal (fabrication of fuel pellets, rods and assemblies for different types of reactors), JSC 'Novosibirsky zavod khimconcentratov', Novosibirsk (fabrication of fuel rods and assemblies for WWER-440 and WWER-1000), JSC 'Tchepetsky mechanitchesky zavod', Tchepetsk (fabrication of Zr tubing). Monitoring of QA is an important element of Quality Management System (QMS) developed and implemented at the above-mentioned enterprises of the JSC 'TVEL' and it is performed on three levels including external and internal audits and author's supervision. Paper also describes short- and long-term policies of the JSC 'TVEL' in nuclear fuel quality field. (author)

  13. Experience in developing countries in monitoring procurement and fabrication

    International Nuclear Information System (INIS)

    Csik, B.J.

    1977-01-01

    Owner's responsibility in monitoring procurement and fabrication. Monitoring ectivity, tasks, knowledge and personnel requirements, scope and organization. Contractual arrangements, commitments, responsibilities, rights and obligations. Domestic and foreign supplies. Staff and consultants. Experience in developing countries. Problem areas: availability of qualified staff, organization, methodology standards, codes, specifications, availability and flow of information, language, technical knowledge, access to suppliers' facilities, delays, nuclear safety related components, modifications and additionals. (orig.) [de

  14. Pilling Resistance of Knitted Fabrics

    Directory of Open Access Journals (Sweden)

    Gita BUSILIENĖ

    2011-09-01

    Full Text Available Knitted fabrics with different quantity of elastane, conspicuous by high viscosity and elasticity, having one of the most important performance properties - resistance to pilling are often used in the production of high quality sportswear. During technological process imitating operating conditions, the behaviour of knitted fabrics may be changed by different industrial softeners from 12 % to 20 % of active substance, for example fatty acid condensate (Tubingal 5051 or silicone micro emulsion (Tubingal SMF. The aim of this investigation is to define the influence of fibrous composition and chemical softeners to the propensity of fuzzing and pilling of plain and plated jersey pattern knitted fabrics. The results of investigations showed that fibrous composition and thickness of materials (up to 6 % and washing as well as softening (from 33 % to 67 % change the resistance of knitted fabrics to pilling.http://dx.doi.org/10.5755/j01.ms.17.3.597

  15. Geoacoustic Physical Model Fabrication Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Fabricates three-dimensional rough surfaces (e.g., fractals, ripples) out of materials such as PVC or wax to simulate the roughness properties associated...

  16. Fabricating plasmonic components for nanophotonics

    DEFF Research Database (Denmark)

    Boltasseva, Alexandra; Nielsen, Rasmus Bundgaard; Jeppesen, Claus

    2009-01-01

    We report on experimental realization of different metal-dielectric structures that are used as surface plasmon polariton waveguides and as plasmonic metamaterials. Fabrication approaches based on different lithographic and deposition techniques are discussed....

  17. Silicone nanocomposite coatings for fabrics

    Science.gov (United States)

    Eberts, Kenneth (Inventor); Lee, Stein S. (Inventor); Singhal, Amit (Inventor); Ou, Runqing (Inventor)

    2011-01-01

    A silicone based coating for fabrics utilizing dual nanocomposite fillers providing enhanced mechanical and thermal properties to the silicone base. The first filler includes nanoclusters of polydimethylsiloxane (PDMS) and a metal oxide and a second filler of exfoliated clay nanoparticles. The coating is particularly suitable for inflatable fabrics used in several space, military, and consumer applications, including airbags, parachutes, rafts, boat sails, and inflatable shelters.

  18. Safeguards through secure automated fabrication

    International Nuclear Information System (INIS)

    DeMerschman, A.W.; Carlson, R.L.

    1982-01-01

    Westinghouse Hanford Company, a prime contractor for the U.S. Department of Energy, is constructing the Secure Automated Fabrication (SAF) line for fabrication of mixed oxide breeder fuel pins. Fuel processing by automation, which provides a separation of personnel from fuel handling, will provide a means whereby advanced safeguards concepts will be introduced. Remote operations and the inter-tie between the process computer and the safeguards computer are discussed

  19. Nuclear topics

    International Nuclear Information System (INIS)

    Lukner, C.

    1982-07-01

    The pamphlet touches on all aspects of nuclear energy, from the world energy demands and consumption, the energy program of the Federal Government, nuclear power plants in the world, nuclear fusion, nuclear liability up to the nuclear fuel cycle and the shutdown of nuclear power plants. (HSCH) [de

  20. Nuclear liability act and nuclear insurance

    International Nuclear Information System (INIS)

    Clarke, Roy G.; Goyette, R.; Mathers, C.W.; Germani, T.R.

    1976-01-01

    The Nuclear Liability Act, enacted in June 1970 and proclaimed effective October 11, 1976, is a federal law governing civil liability for nuclear damage in Canada incorporating many of the basic principles of the international conventions. Exceptions to operator liability for breach of duty imposed by the Act and duty of the operator as well as right of recourse, time limit on bringing actions, special measures for compensation and extent of territory over which the operator is liable are of particular interest. An operator must maintain $75,000,000. of insurance for each nuclear installation for which he is the operator. The Nuclear Insurance Association of Canada (NIAC) administers two ΣPoolsΣ or groups of insurance companies where each member participates for the percentage of the total limit on a net basis, one pool being for Physical Damage Insurance and the other for Liability Insurance. The Atomic Energy Control Board recommends to the Treasury Board the amount of insurance (basic) for each installation. Basic insurance required depends on the exposure and can range from $4 million for a fuel fabricator to $75 million for a power reactor. Coverage under the Operator's Policy provides for bodily injury, property damage and various other claims such as damage from certain transportation incidents as well as nuclear excursions. Workmen's Compensation will continue to be handled by the usual channels. (L.L.)

  1. Nuclear fuel element leak detection system

    International Nuclear Information System (INIS)

    John, C.D. Jr.

    1978-01-01

    Disclosed is a leak detection system integral with a wall of a building used to fabricate nuclear fuel elements for detecting radiation leakage from the nuclear fuel elements as the fuel elements exit the building. The leak detecting system comprises a shielded compartment constructed to withstand environmental hazards extending into a similarly constructed building and having sealed doors on both ends along with leak detecting apparatus connected to the compartment. The leak detecting system provides a system for removing a nuclear fuel element from its fabrication building while testing for radiation leaks in the fuel element

  2. Nuclear power and nuclear weapons

    International Nuclear Information System (INIS)

    Vaughen, V.C.A.

    1983-01-01

    The proliferation of nuclear weapons and the expanded use of nuclear energy for the production of electricity and other peaceful uses are compared. The difference in technologies associated with nuclear weapons and nuclear power plants are described

  3. Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    Bambang Galung Susanto

    2007-01-01

    The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO 2 /year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

  4. Fabrication of control rods for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sease, J.D.

    1998-01-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A

  5. Fabrication of control rods for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  6. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Smith, John [General Atomics, San Diego, CA (United States)

    2016-09-23

    The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

  7. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1977-01-01

    Concerning refining, fabrication and reprocessing operations of such materials as well as the installation and operation of reactors, necessary regulations are carried out. Namely, in case of establishing the business of refining, fabricating and reprocessing nuclear materials as well as installing nuclear reactors, applications for the permission of the Prime Minister and the Minister of International Trade and Industry should be filed. Change of such operations should be permitted after filing applications. These permissions are retractable. As regards the reactors installed aboard foreign ships, it must be reported to enter Japanese waters and the permission by the Prime Minister must be obtained. In case of nuclear fuel fabricators, a chief technician of nuclear fuel materials (qualified) must be appointed per each fabricator. In case of installing nuclear reactors, the design and methods of construction should be permitted by the Prime Minister. The standard for such permission is specified, and a chief engineer for operating reactors (qualified) must be appointed. Successors inherit the positions of ones who have operated nuclear material refining, fabrication and reprocessing businesses or operated nuclear reactors. (Rikitake, Y.)

  8. PHWR fuel fabrication with imported uranium - procedures and processes

    International Nuclear Information System (INIS)

    Rao, R.V.R.L.V.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N.

    2010-01-01

    Following the 123 agreement and subsequent agreements with IAEA & NSG, Government of India has entered into bilateral agreements with different countries for nuclear trade. Department of Atomic Energy (DAE), Government of India, has entered into contract with few countries for supply of uranium material for use in the safeguarded PHWRs. Nuclear Fuel Complex (NFC), an industrial unit of DAE, established in the early seventies, is engaged in the production of Nuclear Fuel and Zircaloy items required for Nuclear Power Reactors operating in the country. NFC has placed one of its fuel fabrication facilities (NFC, Block-A, INE-) under safeguards. DAE has opted to procure uranium material in the form of ore concentrate and fuel pellets. Uranium ore concentrate was procured as per the ASTM specifications. Since no international standards are available for PHWR fuel pellets, Specifications have to be finalized based on the present fabrication and operating experience. The process steps have to be modified and fine tuned for handling the imported uranium material especially for ore concentrate. Different transportation methods are to be employed for transportation of uranium material to the facility. Cost of the uranium material imported and the recoveries at various stages of fuel fabrication have impact on the fuel pricing and in turn the unit energy costs. Similarly the operating procedures have to be modified for safeguards inspections by IAEA. NFC has successfully manufactured and supplied fuel bundles for the three 220 MWe safeguarded PHWRs. The paper describes various issues encountered while manufacturing fuel bundles with different types of nuclear material. (author)

  9. Supply, operation and radioactive waste disposal of nuclear power plants

    International Nuclear Information System (INIS)

    Mohrhauer, H.; Krey, M.; Haag, G.; Wolters, J.; Merz, E.; Sauermann, P.F.

    1981-07-01

    The subject of 'Nuclear Fuel Cycle' is treated in 5 reports: 1. Uranium supply; 2. Fabrication and characteristics of fuel elements; 3. Design, operation and safety of nuclear power plants after Harrisburg; 4. Radioactive waste disposal of nuclear power plants - changed political scenery after 1979; 5. Shutdown and dismantling of LWR-KKW - state of knowledge and feasibility. (HP) [de

  10. Fabrication of thermo-responsive cotton fabrics using poly(vinyl caprolactam-co-hydroxyethyl acrylamide) copolymer.

    Science.gov (United States)

    Xiao, Min; González, Edurne; Monterroza, Alexis Martell; Frey, Margaret

    2017-10-15

    A thermo-responsive polymer with hydrophilic to hydrophobic transition behavior, poly(vinyl caprolactam-co-hydroxyethyl acrylamide) P(VCL-co-HEAA), was prepared by copolymerization of vinyl caprolactam and N-hydroxyethyl acrylamide via free radical solution polymerization. The resulting copolymer was characterized by Fourier transform infrared spectroscopy (FTIR), 1 H nuclear magnetic resonance (NMR), gel permeation chromatography (GPC), differential scanning calorimetry (DSC), and thermogravimetric analysis (TGA). The lower critical solution temperature (LCST) of P(VCL-co-HEAA) was determined at 34.5°C. This thermo-responsive polymer was then grafted onto cotton fabrics using 1,2,3,4-butanetetracarboxylic acid (BTCA) as crosslinker and sodium hypophosphite (SHP) as catalyst. FTIR and energy dispersive X-ray spectroscopy (EDS) studies confirmed the successful grafting reaction. The modified cotton fabric exhibited thermo-responsive behavior as evidenced by water vapor permeability measurement confirming decreased permeability at elevated temperature. This is the first demonstration that a PVCL based copolymer is grafted to cotton fabrics. This study provides a new thermo-responsive polymer for fabrication of smart cotton fabrics with thermally switchable hydrophilicity. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Nuclear fuel banks

    International Nuclear Information System (INIS)

    Anon.

    2010-01-01

    In december 2010 IAEA gave its agreement for the creation of a nuclear fuel bank. This bank will allow IAEA to help member countries that renounce to their own uranium enrichment capacities. This bank located on one or several member countries will belong to IAEA and will be managed by IAEA and its reserve of low enriched uranium will be sufficient to fabricate the fuel for the first load of a 1000 MW PWR. Fund raising has been successful and the running of the bank will have no financial impact on the regular budget of the IAEA. Russia has announced the creation of the first nuclear fuel bank. This bank will be located on the Angarsk site (Siberia) and will be managed by IAEA and will own 120 tonnes of low-enriched uranium fuel (between 2 and 4.95%), this kind of fuel is used in most Russian nuclear power plants. (A.C.)

  12. A facile method to fabricate superhydrophobic cotton fabrics

    Science.gov (United States)

    Zhang, Ming; Wang, Shuliang; Wang, Chengyu; Li, Jian

    2012-11-01

    A facile and novel method for fabricating superhydrophobic cotton fabrics is described in the present work. The superhydrophobic surface has been prepared by utilizing cationic poly (dimethyldiallylammonium chloride) and silica particles together with subsequent modification of (heptadecafluoro-1,1,2,2-tetradecyl) trimethoxysilane. The size distribution of silica particles was measured by Particle Size Analyzer. The cotton textiles before and after treatment were characterized by using scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS). The wetting behavior of cotton samples was investigated by water contact angle measurement. Moreover, the superhydrophobic durability of coated cotton textiles has been evaluated by exposure, immersion and washing tests. The results show that the treated cotton fabrics exhibited excellent chemical stability and outstanding non-wettability with the WCA of 155 ± 2°, which offers an opportunity to accelerate the large-scale production of superhydrophobic textiles materials for new industrial applications.

  13. Nuclear rights - nuclear wrongs

    Energy Technology Data Exchange (ETDEWEB)

    Paul, E.F.; Miller, F.D.; Paul, J.; Ahrens, J.

    1986-01-01

    This book contains 11 selections. The titles are: Three Ways to Kill Innocent Bystanders: Some Conundrums Concerning the Morality of War; The International Defense of Liberty; Two Concepts of Deterrence; Nuclear Deterrence and Arms Control; Ethical Issues for the 1980s; The Moral Status of Nuclear Deterrent Threats; Optimal Deterrence; Morality and Paradoxical Deterrence; Immoral Risks: A Deontological Critique of Nuclear Deterrence; No War Without Dictatorship, No Peace Without Democracy: Foreign Policy as Domestic Politics; Marxism-Leninism and its Strategic Implications for the United States; Tocqueveille War.

  14. Nuclear moments

    CERN Document Server

    Kopferman, H; Massey, H S W

    1958-01-01

    Nuclear Moments focuses on the processes, methodologies, reactions, and transformations of molecules and atoms, including magnetic resonance and nuclear moments. The book first offers information on nuclear moments in free atoms and molecules, including theoretical foundations of hyperfine structure, isotope shift, spectra of diatomic molecules, and vector model of molecules. The manuscript then takes a look at nuclear moments in liquids and crystals. Discussions focus on nuclear paramagnetic and magnetic resonance and nuclear quadrupole resonance. The text discusses nuclear moments and nucl

  15. Fabrication of Nanoimprint stamps for photonic crystals

    International Nuclear Information System (INIS)

    Kouba, J; Kubenz, M; Mai, A; Ropers, G; Eberhardt, W; Loechel, B

    2006-01-01

    We report on fabrication of nanoimprint stamps for fabrication of two dimensional photonic crystals in visible range of spectra. Nanoimprint stamps made of silicon and/or nickel were successfully fabricated using electron beam lithography and advanced dry etching techniques. The quality of the stamps was evaluated using scanning electron microscopy. The fabricated stamps were also evaluated by imprinting them into suitable polymer materials

  16. 14 CFR 23.605 - Fabrication methods.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fabrication methods. 23.605 Section 23.605... Fabrication methods. (a) The methods of fabrication used must produce consistently sound structures. If a... fabrication method must be substantiated by a test program. [Doc. No. 4080, 29 FR 17955, Dec. 18, 1964; 30 FR...

  17. 14 CFR 29.605 - Fabrication methods.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fabrication methods. 29.605 Section 29.605... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Design and Construction General § 29.605 Fabrication methods. (a) The methods of fabrication used must produce consistently sound structures. If a fabrication process...

  18. 14 CFR 27.605 - Fabrication methods.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fabrication methods. 27.605 Section 27.605... STANDARDS: NORMAL CATEGORY ROTORCRAFT Design and Construction General § 27.605 Fabrication methods. (a) The methods of fabrication used must produce consistently sound structures. If a fabrication process (such as...

  19. 14 CFR 25.605 - Fabrication methods.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fabrication methods. 25.605 Section 25.605... STANDARDS: TRANSPORT CATEGORY AIRPLANES Design and Construction General § 25.605 Fabrication methods. (a) The methods of fabrication used must produce a consistently sound structure. If a fabrication process...

  20. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  1. Nuclear power plants - Quality assurance

    International Nuclear Information System (INIS)

    1980-01-01

    This International Standard defines principles for the establishment and implementation of quality assurance programmes during all phases of design, procurement, fabrication, construction, commissioning, operation, maintenance and decommissioning of structures, systems and components of nuclear power plants. These principles apply to activities affecting the quality of items, such as designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, testing, commissioning, operating, inspecting, maintaining, repairing, refuelling and modifying and eventually decommissioning. The manner in which the principles described in this document will be implemented in different organizations involved in a specific nuclear power project will depend on regulatory and contractual requirements, the form of management applied to a nuclear power project, and the nature and scope of the work to be performed by different organizations

  2. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  3. LLNL/YMP Waste Container Fabrication and Closure Project; GFY technical activity summary

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-10-01

    The Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM) Program is studying Yucca Mountain, Nevada as a suitable site for the first US high-level nuclear waste repository. Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing and developing the waste package for the permanent storage of high-level nuclear waste. This report is a summary of the technical activities for the LLNL/YMP Nuclear Waste Disposal Container Fabrication and Closure Development Project. Candidate welding closure processes were identified in the Phase 1 report. This report discusses Phase 2. Phase 2 of this effort involved laboratory studies to determine the optimum fabrication and closure processes. Because of budget limitations, LLNL narrowed the materials for evaluation in Phase 2 from the original six to four: Alloy 825, CDA 715, CDA 102 (or CDA 122) and CDA 952. Phase 2 studies focused on evaluation of candidate material in conjunction with fabrication and closure processes.

  4. Electromagnetic micropores: fabrication and operation.

    Science.gov (United States)

    Basore, Joseph R; Lavrik, Nickolay V; Baker, Lane A

    2010-12-21

    We describe the fabrication and characterization of electromagnetic micropores. These devices consist of a micropore encompassed by a microelectromagnetic trap. Fabrication of the device involves multiple photolithographic steps, combined with deep reactive ion etching and subsequent insulation steps. When immersed in an electrolyte solution, application of a constant potential across the micropore results in an ionic current. Energizing the electromagnetic trap surrounding the micropore produces regions of high magnetic field gradients in the vicinity of the micropore that can direct motion of a ferrofluid onto or off of the micropore. This results in dynamic gating of the ion current through the micropore structure. In this report, we detail fabrication and characterize the electrical and ionic properties of the prepared electromagnetic micropores.

  5. Integrated Fabrication of a Microgripper

    Institute of Scientific and Technical Information of China (English)

    1999-01-01

    Successful implementation of simple mechanism on silicon chip is a prerequisite for monolithic microrobot-ic systems. This paper describes the integrated fabrication of polycrystalline silicon microgripper. Link, fixed andactive joint, and sliding flange structures with dimensions of micrometers have been fabricated on the substrate ofmonocrystalline silicon using silicon microfabrication technology. This microgripper, which may be applied to trans-ducers or sensors, can be batch-fabricated in IC-compatible process. The movable mechanical elements are built onlayers that are later removed, so that they are free for translation and rotation. Under external driving, a microgrip-per cut from substrate would be able to catch tiny filament or small particle with dimension of 10~ 200 micrometers.

  6. Cascade reactor: granule fabrication processes

    International Nuclear Information System (INIS)

    Erlandson, O.D.; Winkler, E.O.; Maya, I.; Pitts, J.H.

    1985-01-01

    A key feature of Cascade is the granular blanket. Of the many blanket material options open to Cascade, fabrication of Li 2 O granules was felt to offer the greatest challenge. The authors explored available methods for initial Li 2 O granule fabrication. They identified three cost-effective processes for fabricating Li 2 O granules: the VSM drop-melt furnace process, which is based on melting and spheroidizing irregularly shaped Li 2 O feed granules; the LiOH process, which spheroidizes liquefied LiOH and uses GA Technologies' sphere-forming procedures; and the Li 2 CO 3 sol-gel process, used for making spherical fuel particles for the high-temperature gas-cooled reactor (HTGR). Each process is described below

  7. SRF Cavity Fabrication and Materials

    CERN Document Server

    Singer, W

    2014-07-17

    The technological and metallurgical requirements of material for highgradient superconducting cavities are described. High-purity niobium, as the preferred metal for the fabrication of superconducting accelerating cavities, should meet exact specifications. The content of interstitial impurities such as oxygen, nitrogen, and carbon must be below 10μg/g. The hydrogen content should be kept below 2μg/g to prevent degradation of the Q-value under certain cool-down conditions. The material should be free of flaws (foreign material inclusions or cracks and laminations) that can initiate a thermal breakdown. Defects may be detected by quality control methods such as eddy current scanning and identified by a number of special methods. Conventional and alternative cavity fabrication methods are reviewed. Conventionally, niobium cavities are fabricated from sheet niobium by the formation of half-cells by deep drawing, followed by trim machining and Electron-Beam Welding (EBW). The welding of half-cells is a delicate...

  8. Stirling Microregenerators Fabricated and Tested

    Science.gov (United States)

    Moran, Matthew E.

    2004-01-01

    A mesoscale Stirling refrigerator patented by the NASA Glenn Research Center is currently under development. This refrigerator has a predicted efficiency of 30 percent of Carnot and potential uses in electronics, sensors, optical and radiofrequency systems, microarrays, and microsystems. The mesoscale Stirling refrigerator is most suited to volume-limited applications that require cooling below the ambient or sink temperature. Primary components of the planar device include two diaphragm actuators that replace the pistons found in traditional-scale Stirling machines and a microregenerator that stores and releases thermal energy to the working gas during the Stirling cycle. Diaphragms are used to eliminate frictional losses and bypass leakage concerns associated with pistons, while permitting reversal of the hot and cold sides of the device during operation to allow precise temperature control. Three candidate microregenerators were fabricated under NASA grants for initial evaluation: two constructed of porous ceramic, which were fabricated by Johns Hopkins Applied Physics Laboratory, and one made of multiple layers of nickel and photoresist, which was fabricated by Polar Thermal Technologies. The candidate regenerators are being tested by Johns Hopkins Applied Physics in a custom piezoelectric-actuated test apparatus designed to produce the Stirling refrigeration cycle. In parallel with the regenerator testing, Johns Hopkins is using deep reactive ion etching to fabricate electrostatically driven, comb-drive diaphragm actuators. These actuators will drive the Stirling cycle in the prototype device. The top photograph shows the porous ceramic microregenerators. Two microregenerators were fabricated with coarse pores and two with fine pores. The bottom photograph shows the test apparatus parts for evaluating the microregenerators, including the layered nickel-and-photoresist regenerator fabricated using LIGA techniques.

  9. Industrias Nucleares do Brasil in the context of the Brazilian nuclear program

    International Nuclear Information System (INIS)

    1996-10-01

    The activities carried out by Industrias Nucleares Brasileiras (INB) related to the nuclear fuel cycle are described. These activities comprise presently uranium prospecting and processing and fuel elements assembly.Starting in 1997,INB will also perform the reconversion of enriched uranium hexafluoride and the fabrication of fuel pellets.Furthermore, INB produces as well rare earth oxides

  10. Advanced fabrication of hyperbolic metamaterials

    DEFF Research Database (Denmark)

    Shkondin, Evgeniy; Sukham, Johneph; Panah, Mohammad Esmail Aryaee

    2017-01-01

    Hyperbolic metamaterials can provide unprecedented properties in accommodation of high-k (high wave vector) waves and enhancement of the optical density of states. To reach such performance the metamaterials have to be fabricated with as small imperfections as possible. Here we report on our...... advances in two approaches in fabrication of optical metamaterials. We deposit ultrathin ultrasmooth gold layers with the assistance of organic material (APTMS) adhesion layer. The technology supports the stacking of such layers in a multiperiod construction with alumina spacers between gold films, which...

  11. MQXFS1 Quadrupole Fabrication Report

    CERN Document Server

    Ambrosio, G; Bossert, R; Cavanna, E; Cheng, D; Chlachidize, G; Cooley, L D; Dietderich, D; Felice, H; Ferracin, P; Ghosh, A; Hafalia, R; Holik, E F; Izquierdo Bermudez, S; Juchno, M; Krave, S; Marchevsky, M; Muratore, J; Nobrega, F; Pan, H; Perez, J C; Pong, I; Prestemon, S; Ravaioli, E; Sabbi, G L; Santini, C; Schmalzle, J; Schmalzle, J; Stoynev, S; Strauss, T; Vallone, G; Wanderer, P; Wang, X; Yu, M

    2017-01-01

    This report presents the fabrication and QC data of MQXFS1, the first short model of the low-beta quadrupoles (MQXF) for the LHC High Luminosity Upgrade. It describes the conductor, the coils, and the structure that make the MQXFS1 magnet. Qualification tests and non-conformities are also presented and discussed. The fabrication of MQXFS1 was started before the finalization of conductor and coil design for MQXF magnets. Two strand design were used (RRP 108/127 and RRP 132/169). Cable and coil cross-sections were “first generation”.

  12. MQXFS1 Quadrupole Fabrication Report

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosio, G. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Anerella, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Bossert, R. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Cavanna, E. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Cheng, D. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Chlachidize, G. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Cooley, L. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Dietderich, D. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Felice, H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Ferracin, P. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Ghosh, A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hafalia, R. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Holik, E. F. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Bermudez, S. Izquierdo [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Juchno, M. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Krave, S. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Marchevsky, M. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Muratore, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Nobrega, F. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Pan, H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Perez, J. C. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Pong, I. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Prestemon, S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Ravaioli, E. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Sabbi, G. L. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Santini, C. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Schmalzle, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Stoynev, S. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Strauss, T. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Vallone, G. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Wanderer, P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Wang, X. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Yu, M. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States)

    2017-07-16

    This report presents the fabrication and QC data of MQXFS1, the first short model of the low-beta quadrupoles (MQXF) for the LHC High Luminosity Upgrade. It describes the conductor, the coils, and the structure that make the MQXFS1 magnet. Qualification tests and non-conformities are also presented and discussed. The fabrication of MQXFS1 was started before the finalization of conductor and coil design for MQXF magnets. Two strand design were used (RRP 108/127 and RRP 132/169). Cable and coil cross-sections were “first generation”.

  13. The Flexible Fabric of Space

    Science.gov (United States)

    VanNorsdall, Erin Leigh

    2015-08-01

    This poster will clearly illustrate my understanding of how the fabric of space behaves. The poster will be on a large trampoline with a heavy bowling ball in the center. The observer will be able to clearly understand the much more complicated property of how an object in space, such as a star, literally bends the fabric of the space around as a result of its density. This will also help to explain, in very simple terms, how space-time is bendable, and therefore, travel in space can be as well.

  14. Purification of HGI2 for nuclear detector fabrication

    International Nuclear Information System (INIS)

    Schieber, M.M.

    1978-01-01

    A process for purification of mercuric iodide (HgI 2 ) to be used as a source material for the growth of detector quality crystals. The high purity HgI 2 raw material is produced by a combination of three stages: synthesis of HgI 2 from Hg and I 2 , repeated sublimation, and zone refining

  15. Nuclear energy research in Indonesia

    International Nuclear Information System (INIS)

    Supadi, S.; Soentono, S.; Djokolelono, M.

    1988-01-01

    Indonesia's National Atomic Energy Authority, BATAN (Badan Tenaga Atom Nasional), was founded to implement, regulate and monitor the development and launching of programs for the peaceful uses of nuclear power. These programs constitute part of the efforts made to change to a more industrialized level the largely agricultural society of Indonesia. BATAN elaborated extensive nuclear research and development programs in a variety of fields, such as medicine, the industrial uses of isotopes and radiation, the nuclear fuel cycle, nuclear technology and power generation, and in fundamental research. The Puspiptek Nuclear Research Center has been equipped with a multi-purpose research reactor and will also have a fuel element fabrication plant, a facility for treating radioactive waste, a radiometallurgical laboratory, and laboratories for working with radioisotopes and for radiopharmaceutical research. (orig.) [de

  16. Concrete structures for nuclear facilities

    International Nuclear Information System (INIS)

    1996-01-01

    The detailed requirements for the design and fabrication of the concrete structures for nuclear facilities and for the documents to be submitted to the Finnish Centre for Radiation and Nuclear Safety (STUK) are given in the guide. It also sets the requirements for the inspection of concrete structures during the construction and operation of facilities. The requirements of the guide primarily apply to new construction. As regards the repair and modification of nuclear facilities built before its publication, the guide is followed to the extent appropriate. The regulatory activities of the Finnish Centre for Radiation and Nuclear Safety during a nuclear facility's licence application review and during the construction and operation of the facility are summarised in the guide YVL 1.1

  17. Application of new technology to the fabrication and installation of stainless steel pipework

    International Nuclear Information System (INIS)

    Halford, P.; Carrick, L.

    1989-01-01

    BNFL has been constructing new reprocessing plant on a continuous basis since the late seventies. Initially the productivity achieved when fabricating and welding stainless steel to nuclear standards was poor. In order to complete projects to programme and cost BNFL developed a Total Fabrication System that is now applied to all of their construction projects and has resulted in overall productivity increases by a factor of 2.4 with major quality and cost benefits. The development of the Total Fabrication System is described

  18. Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

    International Nuclear Information System (INIS)

    Berg, Thomas A.; Disney, Richard K.

    2004-01-01

    Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs

  19. The fabrication of uniform cylindrical nanoshells and their use as spectrally tunable MRI contrast agents

    International Nuclear Information System (INIS)

    Zabow, G; Dodd, S J; Koretsky, A P; Moreland, J

    2009-01-01

    A new form of tunable magnetic resonance imaging agent based on precisely dimensioned cylindrical magnetic nanoshells is introduced. Using top-down prepatterned substrates, the nanoshells are fabricated by exploiting what is usually regarded as a detrimental processing side-effect, namely the redeposition of material back-sputtered during ion-milling. The well-resolved nuclear magnetic resonance peaks of the resulting nanostructures attest to the nanoscale fabrication control and the general feasibility of such sputter redeposition for fabrication of a variety of self-supporting, highly monodisperse nanoscale structures.

  20. Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version 2

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

    2002-10-01

    Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institut, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make specification documents for the fuel pellet, the sphere-pac fuel particles, the vipac fuel fragments, and the fuel segment fabrication. JNC should make the fabrication drawings for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Holland). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Holland). The fabrication drawings have been made under the design assignment with PSI, and consist of the drawings of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications was discussed and agreed among all partners. According to this agreement, the fabrication drawings were revised in January 2002. After the earlier revision, the shape of particle retainer to be made by PSI was modified from its drawing beforehand delivered. In this report, the fabrication drawings revised again will be shown, and the fabrication procedure (welding Qualification Tests) will be modified in accordance with the result of discussion on the 3rd technical meeting held in September 2002. These design works have been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center under the commission of Plutonium Fuel