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Космические корабли и станции, автоматические КА и методы их проектирования, бортовые комплексы управления, системы и средства жизнеобеспечения, особенности технологии производства ракетно-космических систем

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Мониторинг СМИ

Мониторинг СМИ и социальных сетей. Сканирование интернета, новостных сайтов, специализированных контентных площадок на базе мессенджеров. Гибкие настройки фильтров и первоначальных источников.

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Применить Всего найдено 4353. Отображено 100.
21-03-2013 дата публикации

Removal equipment and method for the storage facility for transuranium compounds

Номер: US20130072741A1
Автор: Lee Chung-Shin, Liang Kuo-Long, TSIA CHENG-TO, Wu Jlang-Tzuan
Принадлежит:

A nuclear chemistry laboratory adopts In-site Cutting Technique to remove the large-scale glove box contaminated by transuranium compounds. During removal operation, to prevent further spreading of contamination, it is necessary to build an alpha airtight quarantine tent around the glove box that is ready to be cut. Each section of the quarantine tent maintains a stable airflow and sufficient air exchange to meet negative pressure requirements and effectively prevent leak of α contamination. It is necessary to install a negative pressure ventilation system to assure the operation of the quarantine tent that has a pressure gradient and allows airflow from low contamination area to high contamination area to effectively prevent spreading of α contamination and also increase the safety for the transuranium contaminated glove box that is not in service. 1. A removal operation process for a large glove box contaminated by transuranium compounds comprising the following steps:(a) dismantling the transuranium compounds contaminated glove box in removal operation section of a quarantine tent;packing a dismantled part of the transuranium compounds contaminated large-scale wastes with a transuranium shielding bag as first-layer package, and moving the package to large object transfer room, and repeating the packing and moving to next transfer room in the foregoing steps at least two times for reducing transuranium contamination through negative pressure ventilation, and finally the package is subject to radiation detection, andmoving the package from the quarantine tent into a container when there is no contamination detected;(b) the airtight quarantine tent enclosing the glove box is running with negative pressure ventilation system, allowing the airflow to follow pressure gradient from low negative side to high negative side in the quarantine tent and enabling effective control of airborne α particle flow;(c) after the negative pressure ventilation system starts running, ...

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Номер записи: 1
09-05-2013 дата публикации

PROCESS FOR THE DETRITIATION OF SOFT HOUSEKEEPING WASTE AND PLANT THEREOF

Номер: US20130115156A1
Автор: Borgognoni Fabio, Ghirelli Nicolas, Lefebvre Xavier, Liger Karine, Santucci Alessia, Tosti Silvano, Trabuc Pierre
Принадлежит: ENEA - AGENZIA NAZIONALE PER LE NUOVE TECNOLOGIE, L'ENERGIA E LO SVILUPPO ECONOMIC

A process for removal of tritium from materials that are contaminated thereby envisages the use of a detritiation reactor RT, in which the reaction for the removal of tritium from the waste takes place, the waste being recovered by a flow of moist inert gas in which an extremely low percentage of humidity is used. The heated waste releases a current of tritiated gases, the current of gases being removed from the reactor via the moist inert gas, which conveys it into a membrane reactor RM for decontamination. The membrane reactor, in fact, is able to remove selectively the tritium present in the mixture of gases: there is thus the dual advantage of purifying the mixture of gases and of recovering the tritium contained therein. 1. A process for the detritiation of soft housekeeping waste , i.e. , of radioactive waste produced by laboratories and plants that use tritium , characterized in that it envisages carrying out a thermal desorption by subjecting said waste , appropriately placed in a detritiation reactor (RT) , to a flow of moist gas and subsequently recovering tritium in the form of gas by means of a membrane reactor (RM) in order to valorize it and re-use it; for this purpose there being basically envisaged the following steps:A) shredding and uniformly mixing the waste to be detritiated;B) placing said material to be treated in a detritiation reactor (RT);C) sending inert gas and demineralized water to an evaporation/mixing device;D) feeding said moist gaseous mixture, constituted by inert gas and vapour, to said reactor (RT) so that said mixture traverses all the material to be detritiated, giving rise to the formation of a moist gaseous current containing tritium;E) sending said gaseous current containing tritium to a purposely provided membrane reactor (RM); andF) feeding said membrane reactor (RM) with a swamping gas, thus obtaining exit from the reactor (RM) itself, as end products, of a gaseous current of isotopes containing tritium extracted from the ...

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Номер записи: 2
11-07-2013 дата публикации

APPARATUS AND METHOD FOR THE GRANULATION OF RADIOACTIVE WASTE, AND VITRIFICATION METHOD THEREOF

Номер: US20130178685A1
Автор: CHO Hyun-Je, Choi Young-bu, Kim Cheon-Woo, Kim Deuk-man, Park Jong-kil
Принадлежит:

An apparatus and method for the granulation of radioactive waste in which a preprocessing method for the vitrification of radioactive waste is simplified to conform to onsite conditions of a nuclear power plant, additives are improved, and pellets suitable for vitrification are manufactured. The apparatus for the granulation of radioactive waste includes: a body frame having an inlet and an outlet; a hopper supplying the radioactive waste to be transferred and fed through the inlet; a feeder transferring/supplying the radioactive waste supplied to a specific position and in a certain quantity; a stirrer pulverizing/mixing lumps of the radioactive waste supplied; an additive supply part supplying a lubricant to the radioactive waste fed into the stirrer; and a pellet press pressing the radioactive waste fed through the feeder into a pellet shape and discharging the pellet through the outlet. 1. An apparatus for the granulation of radioactive waste comprising:a body frame having an inlet and an outlet;a hopper supplying the radioactive waste to be transferred and fed through the inlet;a feeder transferring/supplying the radioactive waste supplied through the hopper to a specific position and in a certain quantity;a stirrer pulverizing/mixing lumps of the radioactive waste supplied through the hopper;an additive supply part disposed at a side of the stirrer to supply a lubricant to the radioactive waste fed into the stirrer; anda pellet press pressing the radioactive waste fed through the feeder into a pellet and discharging the pellet of radioactive waste through the outlet.2. The apparatus of claim 1 , further comprising a pollution spread preventing film installed around the body frame to prevent pollution spreading that may occur during manufacturing of the pellet.3. The apparatus of claim 2 , further comprising an exhaust pipe on a top portion of the pollution spread preventing film to discharge dust.4. The apparatus of claim 1 , further comprising a sleeve glove ...

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Номер записи: 3
11-07-2013 дата публикации

METHOD AND DEVICE FOR LIMITING THE DEGASSING OF TRITIATED WASTE ISSUED FROM THE NUCLEAR INDUSTRY

Номер: US20130178686A1
Автор: Lefebvre Xavier, Liger Karine, Troulay Michele
Принадлежит: Commissariat A L'Energie Atomique Et Aux Energies Alternatives

A method and device for limiting the degassing of tritiated waste issued from the nuclear industry are provided. The method reduces an amount of generated tritiated hydrogen (Tor HT) and/or tritiated water (HTO or TO) including at least one piece of tritiated waste from the nuclear industry. The method includes placing the package in contact with a mixture including manganese dioxide (MnO) combined with a component that includes silver; and placing the package in contact with a molecular sieve. 1. A method for reducing an amount of tritiated hydrogen (Tor HT) and/or tritiated water (HTO or TO) generated by at least one package including at least one piece of tritiated waste from the nuclear industry , the method comprising:{'sub': '2', 'placing the package in contact with a mixture including manganese dioxide (MnO) combined with a component comprising silver; and'}placing the package in contact with at least one molecular sieve.2. The method according to claim 1 , wherein the silver is in the mixture appearing as at least one of a silver oxide claim 1 , a silver salt claim 1 , and a silver complex.3. The method according to claim 1 , wherein the silver appears as silver oxide in the mixture claim 1 , in which the mass concentration of manganese dioxide in the mixture ranging from 80% to 99% claim 1 , and in which the mass concentration of silver oxide in the mixture is ranging from 20% to 1%.4. The method according to claim 3 , wherein the mass concentrations of manganese dioxide and silver oxide in the mixture are about 90% and 10% claim 3 , respectively.5. The method according to claim 1 , wherein the mixture includes a platinum compound in which the mass concentration in the mixture is ranging from 0.1% to 1%.6. The method according to claim 5 , wherein the platinum compound is made of platinum black 10% Pt.7. The method according to claim 5 , wherein the silver appears as silver oxide in the mixture claim 5 , in which the mass concentration of manganese dioxide ...

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Номер записи: 4
15-08-2013 дата публикации

METHOD FOR PARTIALLY DECONTAMINATING RADIOACTIVE WASTE

Номер: US20130211172A1
Автор: Baginski Kathrin, Lensa Werner F., Nabbi Rahim, Steinmetz Hans-Juergen, Vulpius Dirk
Принадлежит:

Methods for partially decontaminating radioactive waste wherein the waste is first mixed, or brought in contact, with at least one corrosive medium. Activation energy is then supplied to the corrosive medium, so that at least a portion of the radionuclide present in the waste is converted into at least one gaseous reaction product, or is dissolved, by hydrogen or hydrogen ions, oxygen or oxygen ions, and/or halogen (for example chlorine) or halogen ions from the corrosive medium. The aim is that of decontaminating a C/C-containing porous solid waste, which is contaminated with the C radionuclide. For this purpose, COand/or hydrogen are applied as corrosive media to the waste, so that at least a portion of the waste is reacted to form at least one gaseous reaction product, wherein the process temperature is selected so that the C radionuclide is enriched in the reaction product over C/C. 2. The method according to claim 1 , wherein the hydrogen claim 1 , hydrogen ions claim 1 , oxygen claim 1 , oxygen ions claim 1 , halogen claim 1 , and/or halogen ions react in the nascent state with the radionuclide.35-. (canceled)6. A method according to claim 1 , wherein at least one radionuclide claim 1 , with which the waste is contaminated claim 1 , was formed by a nuclear reaction claim 1 , and more particularly nuclear fission or neutron activation.7. The method according to claim 6 , wherein the nuclear reaction took place in the material of the waste.8. A method according to claim 1 , wherein waste is selected that contains a further isotope of the same element claim 1 , in addition to the radionuclide.9. A method according to claim 1 , wherein porous waste is selected.10. (canceled)11. A method according to claim 1 , wherein waste containing C and/or C is selected.12. A method according to claim 1 , wherein waste that is present in the form of solid pieces claim 1 , granules or powder is selected.1314-. (canceled)15. A method according to claim 1 , wherein a liquid ...

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Номер записи: 5
29-08-2013 дата публикации

METHOD FOR SURFACE DECONTAMINATION

Номер: US20130220366A1
Автор: GASSEN RAINER, ZEILER Bertram
Принадлежит: AREVA NP GMBH

A method for chemical decontamination of an oxide-coated surface of a metal structural part or of a system in a nuclear power plant with several cleaning cycles, involves oxidation steps, in which the oxide layer is treated with an aqueous solution containing an oxidation agent, and a subsequent decontamination step, in which the oxide layer is treated with an aqueous solution of an acid. At least one oxidation step is carried out in an acid solution and at least one oxidation step in an alkaline solution. 1. A process for chemical decontamination of a surface having an oxide layer of a metallic component or of a system of a nuclear power station , which comprises the steps of:performing a plurality of cleaning cycles each containing oxidation steps in which the oxide layer is treated with an aqueous solution having an oxidant and a subsequent decontamination step in which the oxide layer is treated with an aqueous solution of an acid, wherein at least one of the oxidation steps is carried out in an acidic solution and at least one of the oxidation steps is carried out in an alkaline solution; and{'sub': 3', '2', '8, 'sup': '2−', 'selectin the oxidant from the group consisting of O, SO and a cerium(IV) compound for use in the oxidation steps.'}2. The process according to claim 1 , which further comprises setting a pH of the acidic solution to be <6 and a pH of the alkaline solution to be >8.3. The process according to claim 2 , which further comprises setting a pH of the acidic solution to be <4 and a pH of the alkaline solution of >10.4. (canceled)5. The process according to claim 1 , which further comprises selecting the oxidant from the group consisting of HMnO claim 1 , HMnOwith HNO claim 1 , and KMnOwith HNOfor use in an oxidation step in the acidic solution.6. The process according to claim 1 , which further comprises using KMnOtogether with an alkalizing agent for an oxidation step in the alkaline solution.7. The process according to claim 6 , which further ...

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Номер записи: 6
12-09-2013 дата публикации

DECONTAMINATION METHOD AND APPARATUS FOR SOLID-STATE MATERIAL CONTAMINATED BY RADIOCESIUM

Номер: US20130237741A1
Автор: Futami Kenichi, Miama Takahiro, Sakashita Daichi, Sekine Tomokazu, Shimomura Tatsuo
Принадлежит:

A decontamination method of solid-state material contaminated by radiocesium comprising bringing the solid-state material containing radiocesium in contact with a first processing solution and preferably eluting cesium ion from the solid-state material to the liquid phase under the presence of potassium ion or ammonium ion. 1. A decontamination method of solid-state material contaminated by radiocesium comprising:bringing solid-state material contaminated by radiocesium in contact with a first process liquid containing fluorine ion; andeluting cesium ion from the solid-state material to a liquid phase.2. The decontamination method according to claim 1 , comprising:bringing a processed solid obtained by eluting the cesium ion to the liquid phase in further contact with a second process liquid containing an acidic material; andeluting cesium ion remaining in the processed solid.3. The decontamination method according to or claim 1 , wherein the fluorine ion is derived from fluoride selected from hydrogen fluoride gas claim 1 , hydrofluoric acid and its salt claim 1 , and fluorosilicate and its salt.4. The decontamination method according to wherein the fluoride is ammonium hydrogen fluoride claim 3 , potassium hydrogen fluoride claim 3 , potassium fluoride or sodium hydrogen fluoride.5. The decontamination method according to claim 2 , wherein the acidic material is selected from hydrochloric acid claim 2 , nitric acid claim 2 , phosphoric acid claim 2 , sulfamic acid claim 2 , and organic acid.6. The decontamination method according to claim 1 , wherein the solid-state material is a fine-grained fraction or a coarse-grained fraction which is obtained by a wash and classification process.7. A decontamination method of a solid-state material contaminated by radiocesium comprising:bringing a solid-state material containing radiocesium as a contaminant in contact with a first process liquid containing fluorine ion, andeluting cesium ion from the solid-state material to a ...

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Номер записи: 7
31-10-2013 дата публикации

Decontamination method of cladding hull wastes generated from spent nuclear fuel and apparatus thereof

Номер: US20130289329A1
Автор: Chang Hwa Lee, Geun Il Park, Kweon Ho KANG, Min Ku Jeon
Принадлежит: Korea Atomic Energy Research Institute KAERI

The present disclosure relates to a decontamination method and apparatus for cladding hull wastes of spent nuclear fuels, capable of decontaminating a small quantity of spent nuclear fuels remaining on surfaces of the cladding hull wastes and radioactive fission products penetrated into the cladding hulls through an electrochemical dissolution. The method includes inserting the cladding hull waste into an anodic basket, immersing a reference electrode and a cathodic electrode as well as the anodic basket in a molten salt, dissolving a surface of the cladding hull waste by applying a voltage or current to the anodic basket with respect to the cathodic electrode or the reference electrode, removing the anodic basket, and removing a salt remaining on the surface of the cladding hull waste.

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Номер записи: 8
07-11-2013 дата публикации

Melting apparatus for melt decontamination of radioactive metal waste

Номер: US20130294473A1
Автор: Hyun Gyu Kang, Jae Bong RYU, Jeung Gun Seol, Suk Ju CHO, Wook Jin Han, Yong Jae Kim, Young Bae Lee
Принадлежит: Kepco Nuclear Fuel Co Ltd

Disclosed herein is a melting apparatus for melt-decontaminating radioactive metal waste. The melting apparatus includes a melting furnace, a high frequency generator, a ladle, a bogie, a cooling unit and a dust collector. The melting furnace includes a crucible into which the metal waste is input, and an induction coil which is wound around the crucible to melt the metal waste. The induction coil has a hollow hole in which cooling fluid flows. The high frequency generator applies high-frequency current to the induction coil. The ladle supplies molten metal, from which slag has been removed in the crucible, into molds. The bogie is disposed adjacent to the ladle and is provided with the molds, each of which forms an ingot using the molten metal supplied thereinto. The cooling unit cools the cooling fluid and circulates it along the induction coil. The dust collector filters out dust and purifies gas.

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Номер записи: 9
28-11-2013 дата публикации

APPARATUS FOR REDUCING VOLUME OF RESIN CONTAINING RADIOACTIVE MATERIAL, AND METHOD FOR OPERATING THE APPARATUS

Номер: US20130313227A1
Автор: Katagiri Gen-ichi
Принадлежит: FUJI ELECTRIC CO., LTD.

Provided is an ion exchange resin volume reduction apparatus in which the ignition of plasma is facilitated and the plasma is prevented from extinguishing. A volume reduction apparatus according to aspects of the present invention includes a stage carrying thereon a resin to be treated, a CCP power source, and an ICP power source. The volume reduction apparatus according to a certain aspect of the present invention is provided with a supply mechanism, and the CCP power source continues operating when the resin to be treated is supplied in a depressurized state to a vacuum vessel. In the volume reduction apparatus according to a certain aspect of the present invention, the CCP power source continues operating when a gas condition under which gas is supplied into the vacuum vessel is changed. 1. A volume reduction apparatus , comprising:a heatable stage that is disposed inside a vacuum vessel and formed to place thereon a resin to be treated which carries a radioactive substance;a CCP power source that supplies a voltage or an electric field by capacitive coupling to a space above the stage inside the vacuum vessel; andan ICP power source that supplies power by inductive coupling to oxygen plasma induced in the space, whereinthe CCP power source supplies to the space a voltage or an electric field that ignites the oxygen plasma.2. A volume reduction apparatus , comprising:a heatable stage that is disposed inside a vacuum vessel and formed to place thereon a resin to be treated which carries a radioactive substance;a CCP power source that supplies a voltage or an electric field by capacitive coupling to a space above the stage inside the vacuum vessel;an ICP power source that supplies power by inductive coupling to oxygen plasma induced in the space; anda supply mechanism configured to supply the resin to be treated to the stage while maintaining a depressurized state of the vacuum vessel, whereinthe CCP power source supplies the voltage or electric field when the ...

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Номер записи: 10
13-02-2014 дата публикации

MAGNETIC COMPOSITE PARTICLE FOR DECONTAMINATION, METHOD FOR FABRICATING THE SAME, RADIOACTIVE SUBSTANCE FAMILY DECONTAMINATION SYSTEM, AND RADIOACTIVE SUBSTANCE FAMILY DECONTAMINATION METHOD

Номер: US20140042068A1
Автор: NAMIKI Yoshihisa
Принадлежит: The Jikei University

Provided is a radioactive substance collecting system and a radioactive substance collecting method which are capable of collecting radioactive substances with high efficiency. The radioactive substance collecting system according to the present invention removes radioactive substances (radioactive cesium ) contained in a liquid (radioactive substances-contaminated water ) and includes, as means for removing radioactive substances from the liquid, a radioactive substance trapping composite including at least a magnetic particle and a radioactive substance trapping compound that traps radioactive substances, and magnetic accumulation means for accumulating the radioactive substance trapping composite 1. A radioactive substance family decontamination system comprising:a magnetic composite particle for decontamination that traps a radioactive substance family in a liquid; anda magnetic accumulation unit for accumulating the magnetic composite particle for decontamination in the liquid,wherein the magnetic composite particle for decontamination has a multilayer structure including: a magnetic particle formed in a core portion; a trapping compound formed in a surface layer to trap the radioactive substance family in the liquid; and an intermediate layer that directly covers the magnetic particle and is formed substantially between the magnetic particle and the trapping compound, andwherein the magnetic particle has an average grain size of 1 nm to 10 mm.25.-. (canceled)6. The radioactive substance family decontamination system according to claim 1 , wherein the magnetic accumulation unit includes a structure that enables ON-OFF control of a magnetic force claim 1 , accumulates the radioactive substance family contained in the liquid by the magnetic force of the magnetic accumulation unit claim 1 , and separates the radioactive substance family from the magnetic accumulation unit after the accumulation.7. (canceled)8. The radioactive substance family decontamination ...

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Номер записи: 11
27-02-2014 дата публикации

Immobilization of Technetium by Electroless Plating

Номер: US20140058183A1
Автор: Hagerty Kevin J.
Принадлежит: AREVA NP Inc.

A process of incorporating technetium into an electroless deposit, forming an alloy that is extremely resistant to corrosion and reduces the mobility of technetium on a geologic time scale is disclosed and claimed. The process includes providing a liquid containing technetium, such as an aqueous waste stream generated during the used nuclear fuel reprocessing activities. The technetium is collected and concentrated, and provided into an electroless deposition bath. A substrate, such as suitably prepared zero valent iron or stainless steel, is introduced into the bath to initiate autocatalytic electroless deposition of the technetium onto the substrate due to the difference in electrochemical potential between the plating bath and ti metals in solution. This causes a layer of technetium metal to form on the substrate. The electroless deposition is continued until virtually all of the technetium has been removed from the bath, and then continues to build a layer of technetium-free material on the outermost surface of the substrate. One or more additional deposition steps may be performed to armor the plated substrate against leaching/corrosion in a nuclear waste disposal facility. 1. A process , comprising:{'sup': '99', 'providing a liquid containing Tc;'}adding chemicals to form an electroless plating bath;{'sup': '99', 'introducing a suitably prepared plating substrate into said bath to initiate electroless deposition of said Tc onto said substrate; and'}{'sup': '99', 'maintaining said substrate within said bath until substantially all of said Tc is removed from said bath and deposited onto said substrate.'}2. The process of claim 1 , further comprising:filtering said liquid to remove unwanted particulates;{'sup': '99', 'eluting Tc from said liquid; and'}{'sup': '99', 'collecting eluted Tc as a concentrated solution.'}3. The process of claim 2 , wherein said eluting includes performing a pertechnetate specific ion exchange.4. The process of claim 1 , wherein said ...

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Номер записи: 12
07-01-2021 дата публикации

SYSTEMS AND METHODS FOR NUCLEAR WASTE DISPOSAL USING GRIDS

Номер: US20210005338A1
Автор: Crichlow Henry
Принадлежит:

Embodiments of the present invention include systems and methods for long-term disposal of nuclear and/or radioactive waste materials, in liquid, solid, and/or other physical forms, using an array deeply located human-made caverns (caverns), wherein the array of caverns are within a deep geologic rock formation and below a grid pattern on a surface of the Earth. Each cavern is made from a substantially vertical wellbore, by drilling and under reaming operations upon a distal portion of the substantially vertical wellbore. At least some of the caverns may be connected by intersecting substantially lateral wellbores that may facilitate injection of protective materials into the caverns that are so intersected. The nuclear and/or radioactive waste may be preprocessed from original surface storage site(s), transported, temporarily surface stored, and then finally further processed at a selected wellsite before injection into a given of the subterranean deep caverns within the deep geologic rock formation. 1. A method for disposing of radioactive waste into a plurality of human-made caverns that are arranged in a predetermined array pattern within a deep geological formation , wherein the method comprises steps of:(a) forming a predetermined grid pattern on a surface of the Earth that is vertically directly above the deep geological formation, wherein the predetermined grid pattern comprises a plurality of grids, wherein a sub-set of the plurality of grids comprises at least one drill site per grid selected from the sub-set;(b) placing a first walking drill rig at one of the at least one drill sites;(c) drilling a substantially vertical wellbore from the surface directly down to the deep geological formation using the first walking drill rig, wherein the substantially vertical wellbore at least touches the deep geological formation;(d) under-reaming a terminal portion of the substantially vertical wellbore into the deep geological formation using the first walking drill ...

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Номер записи: 13
03-01-2019 дата публикации

TREATMENT METHOD OF RADIOACTIVE WASTE WATER CONTAINING RADIOACTIVE CESIUM AND RADIOACTIVE STRONTIUM

Номер: US20190006055A1
Автор: IZUMI Takeshi, KINOSE Yutaka, Komatsu Makoto, KOZASU Kenta, MIYABE Shinsuke, NOGUCHI Eiji, Sakamoto Takeshi, Sakuma Takashi
Принадлежит:

A treatment method of radioactive waste water containing radioactive cesium and radioactive strontium, comprising passing the radioactive waste water containing radioactive cesium and radioactive strontium through an adsorption column packed with an adsorbent for cesium and strontium, to adsorb the radioactive cesium and radioactive strontium on the adsorbent, wherein the adsorbent for cesium or strontium comprises a crystalline silicotitanate having a crystallite diameter of 60 Å or more and having a half width of 0.9° or less of the diffraction peak in the lattice plane (100), the crystalline silicotitanate represented by the general formula: ATiSiO.nHO. 1. A treatment method of radioactive waste water containing radioactive cesium and radioactive strontium , comprising passing the radioactive waste water containing radioactive cesium and radioactive strontium through an adsorption column packed with an adsorbent for cesium and strontium , to adsorb the radioactive cesium and radioactive strontium on the adsorbent ,{'sub': 4', '4', '3', '16', '2, 'wherein the adsorbent for cesium or strontium comprises a crystalline silicotitanate having a crystallite diameter of 60 Å or more and having a half width of 0.9° or less of the diffraction peak in the lattice plane (100), the crystalline silicotitanate represented by the general formula: ATiSiO.nHO wherein A is Na or K or a combination thereof, and n represents a number of 0 to 8,'}wherein the adsorbent for cesium and strontium is a granular adsorbent having a grain size of 250 μm or more and 1200 μm or less,wherein the absorbent is packed to a height of 10 cm or more and 300 cm or less in the adsorption column, and{'sup': '−1', 'wherein the radioactive waste water is passed through the adsorption column at a linear velocity (LV) of 1 m/h or more and 40 m/h or less and a space velocity (SV) of 200 hor less.'}2. The treatment method according to claim 1 , wherein the radioactive waste water is waste water containing a Na ...

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Номер записи: 14
12-01-2017 дата публикации

METHOD FOR PRODUCING NONATITANATE OF ALKALI METAL

Номер: US20170007983A1
Автор: KINOSE Yutaka, KOZASU Kenta, MIYABE Shinsuke
Принадлежит: Nippon Chemical Industrial Co., Ltd.

A method for producing a nonatitanate of an alkali metal, the method having: a first step for reacting an alkali metal hydroxide with titanium tetrachloride and producing Ti(OH); a second step for mixing the resulting Ti(OH)and an alkali metal hydroxide; and a third step for heating the mixture obtained in the second step, the alkali metal hydroxide being used so that the A/Ti molar ratio (A represents an alkali metal element) falls within a range of 1.0-5.0 in the second step, wherein a nonatitanate of an alkali metal can be produced economically. 1. A method for producing a nonatitanate of an alkali metal , the method comprising:{'sub': '4', 'a first step of allowing an alkali metal hydroxide to act on titanium tetrachloride to produce Ti(OH);'}{'sub': '4', 'a second step of mixing the resulting Ti (OH)and an alkali metal hydroxide; and'}a third step of heating the mixed solution obtained in the second step,wherein, in the second step, the alkali metal hydroxide is used so that an A/Ti molar ratio (A represents an alkali metal element) is in a range of 1.0 to 5.0.2. The method according to claim 1 , wherein the third step is performed in a temperature range of 100° C. or more and 300° C. or less under spontaneous pressure.3. The method according to claim 1 , wherein the third step is performed in a temperature range of less than 105° C. under atmospheric pressure.4. The method according to claim 1 , wherein at least one selected from sodium hydroxide and potassium hydroxide is used as an alkali metal hydroxide.5. The method according to claim 1 , wherein there is obtained at least one nonatitanate selected from the group consisting of NaTiO.nHO claim 1 , KTiO.nHO claim 1 , and (NaK)TiO.nHO where x represents a number of more than 0 and less than 1 and n represents a number of 0 or more.6. An adsorbent for strontium claim 1 , wherein a nonatitanate of an alkali metal obtained by the production method according to is used.7. An adsorbent for strontium claim 1 , ...

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Номер записи: 15
27-01-2022 дата публикации

SELF LOADING WASTE DISPOSAL SYSTEMS AND METHOD

Номер: US20220023925A1
Автор: Crichlow Henry, Schneider Gary
Принадлежит: NuclearSAFE Technology LLC

Self-loading systems and methods for disposal of waste materials in a deep underground formation may include at least one wellbore that runs from the Earth's surface to the deep underground formations, wellbore viscous fluid within that at least one wellbore, and at least one waste capsule, wherein the at least one waste capsules houses some waste and is configured to fall within both the at least one wellbore and the wellbore viscous fluid. The systems and methods may also include at least one human-made cavern located in the deep underground formation and connected to the at least one wellbore, wherein the at least one human-made cavern may be configured to receive the at least one waste capsule. The systems and methods may also include a counter for counting waste capsules and/or a robot for dropping waste capsules into a wellhead leading to the at least one wellbore. 1. A system for disposing of waste , wherein the system comprises:at least one wellbore that runs from the Earth's surface and into a deep geological formation, wherein the deep geological formation is located at least five thousand feet beneath the Earth's surface;a quantity of wellbore viscous fluid that has a predetermined viscosity, wherein at least some portion of the quantity of the wellbore viscous fluid occupies at least some portion of the at least one wellbore before at least one waste capsule is dropped into the at least one wellbore; wherein the wellbore viscous fluid is non-gaseous;the at least one waste capsule, wherein the at least one waste capsule is configured to receive and house at least some quantity of waste;wherein when the at least one waste capsule is dropped into the at least one wellbore, the at least some portion of the quantity of the wellbore viscous fluid controls a velocity of the at least one waste capsule such that the at least one waste capsule comes to a stop within the deep geological formation and without rupturing.2. The system according to claim 1 , wherein ...

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Номер записи: 16
10-01-2019 дата публикации

ADSORBENT FOR RADIOACTIVE ANTIMONY, RADIOACTIVE IODINE AND RADIOACTIVE RUTHENIUM, AND TREATMENT METHOD OF RADIOACTIVE WASTE WATER USING THE ADSORENT

Номер: US20190009245A1
Автор: IZUMI Takeshi, KINOSE Yutaka, Komatsu Makoto, KOZASU Kenta, MIYABE Shinsuke, Sakamoto Takeshi, Sakuma Takashi, Satou Kiyoshi, Sugihara Kaori, Tokutake Mari
Принадлежит:

An adsorbent capable of adsorbing radioactive antimony, radioactive iodine and radioactive ruthenium, the adsorbent containing cerium(IV) hydroxide in a particle or granular form having a particle size of 250 μm or more and 1200 μm or less; and a treatment method of radioactive waste water containing radioactive antimony, radioactive iodine and radioactive ruthenium, the treatment method comprising passing the radioactive waste water containing radioactive antimony, radioactive iodine and radioactive ruthenium through an adsorption column packed with the adsorbent, to adsorb the radioactive antimony, radioactive iodine and radioactive ruthenium on the adsorbent, wherein the absorbent is packed to a height of 10 cm or more and 300 cm or less of the adsorption column, and wherein the radioactive waste water is passed through the adsorption column at a linear velocity (LV) of 1 m/h or more and 40 m/h or less and a space velocity (SV) of 200 hor less. 1. An adsorbent capable of adsorbing radioactive antimony , radioactive iodine and radioactive ruthenium , the adsorbent comprising cerium(IV) hydroxide ,wherein the cerium(IV) hydroxide has the following properties:(1) a granular form having a particle size of 250 μm or more and 1200 μm or less,(2) in a thermogravimetric analysis, a weight reduction ratio is 4.0% or more and 10.0% or less when the temperature is increased from 200° C. to 600° C., and{'sup': −1', '−1', '−1', '−1', '−1', '−1, '(3) in an infrared absorption spectrum analysis, absorption peaks are observed in ranges of 3270 cmor more and 3330 cmor less, 1590 cmor more and 1650 cmor less, and 1410 cmor more and 1480 cmor less.'}2. The adsorbent according to claim 1 , wherein a content of the cerium(IV) hydroxide is 99.0 wt % or more.3. The adsorbent according to claim 1 , wherein a content of the cerium(IV) hydroxide is 90.0 wt % or more and 99.5 wt % or less claim 1 , and the adsorbent further comprises silver phosphate in a content of 0.5 wt % to 10.0 wt %.4 ...

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Номер записи: 17
08-01-2015 дата публикации

Solidification method of radioactive waste

Номер: US20150011816A1
Автор: Akio Sayano, Hirotada Hayashi, Masamichi Obata, Shohei Kawano, Yoshiyuki KAWAHARADA, Yusuke NOHARA
Принадлежит: Toshiba Corp

A solidification method of radioactive waste is provided, including kneading a binder and an inorganic adsorbent to obtain a kneaded object, the in organic adsorbent included radionuclides; extruding the kneaded object to obtain an extruded material object; cutting the extruded material object to obtain at least one extruded material block; and firing the at least one extruded material block to solidify the at least one extruded material block.

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Номер записи: 18
11-01-2018 дата публикации

A Method For Converting Elements, Such As Calcium, Copper, Magnesium, And Cesium, Into More Useful Elements, And A Method For Making Radioactive Substances Harmless By Applying This Element Conversion Method

Номер: US20180012673A1
Автор: OMASA Ryushin
Принадлежит:

The method according to the present invention comprises using a high-frequency vibrating stirrer that is confirmed to include a treatment tank I, a high-frequency vibrating motor fixed to a table positioned above the treatment tank , two vibrating rods extending toward the bottom of the treatment tank and coupled to the table, and multistage vibrating blades mounted to the lower parts of the vibrating rods and surface-plated with palladium or platinum serving as a catalyst in element transmutation, characterized in that the high frequency vibrating motor is controlled by an inverter so as to vibrate the multistage vibrating blades at a frequency of 100-170 Hz in an aqueous solution containing an element to be transmuted in the treatment tank , thereby transmuting the element in the aqueous solution into another element. By adding heavy water to the solution to be treated, the transmutation efficiency can be elevated. By adding tritium water with an appropriate concentration as a substitute for the heavy water, the element transmutation can be completed within a short period of time and, at the same time, the tritium water that is seemingly the main cause of radioactive contamination can be effectively utilized and the radioactivity thereof can be attenuated or detoxified. 2. A method described in claim 1 , wherein heavy water is added to said aqueous solution with a concentration of 0.1 to 5% claim 1 , to shorten the duration of element conversion.3. A method described in claim 1 , wherein tritium water of 0.5 to 5 μSv is added to said aqueous solution with a concentration of 5 to 50% to shorten the duration of element conversion claim 1 , while utilizing the tritium water claim 1 , which is said to cause radioactive contamination claim 1 , and considerably mitigating or eliminating the radioactivity of the tritium.4. A method described in claim 1 , wherein said multistep vibrating vanes are equipped with positive and negative electrodes and nano or micro bubbles ...

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Номер записи: 19
10-01-2019 дата публикации

TREATMENT METHOD OF RADIOACTIVE WASTE WATER CONTAINING RADIOACTIVE CESIUM AND RADIOACTIVE STRONTIUM

Номер: US20190013107A1
Автор: IZUMI Takeshi, KIKUCHI Masahiro, KINOSE Yutaka, Komatsu Makoto, MIYABE Shinsuke, Sakamoto Takeshi, Sakuma Takashi
Принадлежит:

The present invention provides a treatment method of radioactive waste water containing radioactive cesium and radioactive strontium, comprising passing the radioactive waste water containing radioactive cesium and radioactive strontium through an adsorption column packed with an adsorbent for cesium and strontium, to adsorb the radioactive cesium and radioactive strontium on the adsorbent, wherein the adsorbent for cesium and strontium comprises: at least one selected from crystalline silicotitanates represented by the general formulas: NaTiSiO.nHO, (NaK)TiSiO.mHO and KTiSiO.lHO wherein x represents a number of more than 0 and less than 1, and n, m and l each represents a number of 0 to 8; and at least one selected from titanate salts represented by the general formulas: NaTiO.qHO, (NaK)TiO.rHO and KTiO.tHO wherein y represents a number of more than 0 and less than 1, and q, r and t each represents a number of 0 to 10; wherein the adsorbent is a granular adsorbent having a grain size of 250 μm or more and 1200 μm or less, wherein the absorbent is packed to a height of 10 cm or more and 300 cm or less in the adsorption column, and wherein the radioactive waste water is passed through the adsorption column at a linear velocity (LV) of 1 m/h or more and 40 m/h or less and a space velocity (SV) of 200 hor less. 1. A treatment method of radioactive waste water containing radioactive cesium and radioactive strontium , comprising passing the radioactive waste water containing radioactive cesium , radioactive strontium , a Na ion , a Ca ion and a Mg ion through an adsorption column packed with an adsorbent for cesium and strontium , to adsorb the radioactive cesium and radioactive strontium on the adsorbent , [{'sub': 4', '4', '3', '16', '2', 'x', '(1-x)', '4', '4', '3', '16', '2', '4', '4', '3', '16', '2, 'at least one selected from crystalline silicotitanates represented by the general formulas: NaTiSiO.nHO, (NaK)TiSiO.mHO and KTiSiO.lHO wherein x represents a number of ...

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Номер записи: 20
09-01-2020 дата публикации

CHEMICAL DECONTAMINATION METHOD

Номер: US20200013519A1
Автор: ISHIDA Kazushige, IWASA Junji, KAZAMA Masahiko, OUCHI SATOSHI, TSUBOKAWA Naobumi
Принадлежит:

A chemical decontamination method includes a dissolution step in which a radioactive insoluble substance containing a metal oxide, the radioactive insoluble substance being adhered to a decontamination object including carbon steel, is dissolved in a decontamination solution and a metal-ion removal step in which the decontamination solution containing the metal ion, the decontamination solution being produced in the dissolution step, is brought into contact with a cation-exchange resin in order to remove the metal ion, the dissolution step including a reductive dissolution step conducted using a decontamination solution containing formic acid, ascorbic acid and/or erythorbic acid, and a corrosion inhibitor. 1. A chemical decontamination method comprising a dissolution step in which a radioactive insoluble substance containing a metal oxide , the radioactive insoluble substance being adhered to a decontamination object including carbon steel , is dissolved in a decontamination solution and a metal-ion removal step in which the decontamination solution containing the metal ion , the decontamination solution being produced in the dissolution step , is brought into contact with a cation-exchange resin in order to remove the metal ion , the dissolution step including a reductive dissolution step conducted using a decontamination solution containing formic acid , ascorbic acid and/or erythorbic acid (hereinafter , referred to as “ascorbic acid , etc.”) , and a corrosion inhibitor.2. The chemical decontamination method according to claim 1 , wherein the decontamination object includes carbon steel and stainless steel claim 1 , and wherein the dissolution step includes an oxidative dissolution step conducted using a decontamination solution containing permanganic acid and/or a permanganic acid salt (hereinafter claim 1 , referred to as “permanganic acid (salt)”) at a concentration of 100 to 2 claim 1 ,000 mg/L claim 1 , a reductive decomposition step in which a reducing ...

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Номер записи: 21
19-01-2017 дата публикации

HIGH-PRESSURE FLUID DISCHARGE DEVICE

Номер: US20170018322A1
Автор: KITAMURA Masafumi
Принадлежит: IHI CORPORATION

Provided is a high-pressure fluid discharge device in which a pipe, to which a nozzle is connected, is routed and a high pressure fluid transferred through the pipe is discharged from the nozzle, wherein the pipe is formed by alternately connecting first pipes and second pipes having a larger flow passage area than the first pipes. 1. A high-pressure fluid discharge device in which a pipe , to which a nozzle is connected , is routed and a high pressure fluid transferred through the pipe is discharged from the nozzle , wherein the pipe is formed by alternately connecting first pipes and second pipes having a larger flow passage area than the first pipes.2. The high-pressure fluid discharge device according to claim 1 , wherein the first pipes are formed in expandable coil shapes.3. The high-pressure fluid discharge device according to claim 1 , wherein at least one of the second pipes is connected to one of the first pipes formed in the coil shape in a tangential direction thereof.4. The high-pressure fluid discharge device according to claim 1 , wherein the second pipes are longer than the first pipes.5. The high-pressure fluid discharge device according to claim 1 , wherein the first pipes and the second pipes are formed of stainless steel.6. The high-pressure fluid discharge device according to claim 1 , wherein the high pressure fluid is liquid nitrogen. This application is a continuation application based on a PCT Patent Application No.PCT/JP2015/061823, filed on Apr. 17, 2015, whose priority is claimed on Japanese Patent Application No. 2014-85419, filed Apr. 17, 2014, the content of both the PCT Application and the Japanese Application are incorporated herein by reference.Embodiments described herein relates to a high-pressure fluid discharge device.High-pressure fluid discharge devices are used for, for example, decontamination/disassembly processes of nuclear reactor power generation facilities, and so on. The high-pressure fluid discharge devices can spray ...

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Номер записи: 22
22-01-2015 дата публикации

Method of drying high-level radioactive wastes and device thereof

Номер: US20150020405A1
Автор: Chang Kuo-Yuan, Chen Jhong-Heng
Принадлежит:

A method of drying high-level radioactive wastes and device thereof contains a suspension mechanism for hanging a manual elevating mechanism, and a shielding cover. The manual elevating mechanism couples with a basket for accommodating wastes by using a hanging rope, and the shielding cover is fixed below the suspension mechanism and in a moving path of the basket. The basket is moved to a storage tank containing water in which radioactive wastes are stored, and the radioactive wastes are pumped into the basket by means of a pump. The basket is then lifted above a water surface of the storage tank and received in the shielding cover, and then the shielding cover is moved into a water holder so as to drain waters in the basket. The basket is further moved onto a heating seat to be heated and a vacuum equipment is started to dry the radioactive wastes. 1. A method of drying high-level radioactive wastes comprising steps of:placing a basket into a water tank in which radioactive wastes are stored and collecting the radioactive wastes in the basket, wherein a hanging device moves a basket to a storage tanks containing water in which radioactive wastes are stored, and the radioactive wastes B in the storage tanks are moved into the basket;moving the basket out of the water tank and placing the basket in a shielding cover; wherein the basket for collecting the radioactive wastes is lifted above a water surface of the storage tank and is received in the shielding cover so that the basket moves the radioactive wastes without exposing radiation;moving the basket into a water holder so as to drain water drops, wherein the basket is moved into a water holding groove of the water holder so as to drain the waters contained in the radioactive wastes; andhanging the basket on a heating tray and connecting the shielding cover with a vacuuming equipment so as to dry the radioactive wastes, wherein a vacuum equipment is connected with the shielding cover so as to heat and vacuum free ...

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Номер записи: 23
17-01-2019 дата публикации

Treatment Method for Volume Reduction of Spent Uranium Catalyst

Номер: US20190019590A1
Автор: AHN Byung-Gil, CHUNG Dong Yong, Kim Kwang-Wook, Lee Keun-Young, SEO Bum Kyoung
Принадлежит: KOREA ATOMIC ENERGY RESEARCH INSTITUTE

The present invention relates to a volume reduction treatment method of a spent uranium catalyst. According to the volume reduction treatment method of a spent uranium catalyst of the present invention, the disposal cost of the spent uranium catalyst can be reduced and the utilization of the repository can be improved since the method can significantly reduce the volume of the final disposal waste of the spent uranium catalyst. 1. A treatment method of a spent uranium catalyst comprising the following steps:selectively dissolving the support component by immersing the spent uranium catalyst in an alkali solution (step 1);separating the dissolution solution and the undissolved solid materials in the step 1 by solid-liquid separation (step 2);selectively precipitating the silicon ions included in the dissolution solution as silicon dioxide (step 3);separating the silicon dioxide generated in step 3 by solid-liquid separation and its purification (step 4);precipitating uranium ions as uranium phosphate by adding phosphate to the residual solution separated in the step 4 (step 5);separating the uranium phosphate generated in step 5 by solid-liquid separation (step 6);mixing the undissolved solid materials separated in step 2 with the precipitate of uranium phosphate separated in step 6 and then adding a glassification agent thereto, followed by heat-treatment to fix the mixture in the form of a glass-ceramic composite medium (step 7).2. The treatment method of a spent uranium catalyst according to claim 1 , wherein the spent uranium catalyst of step 1 has a form in which USbMO(M is one or more materials selected from the group consisting of Fe claim 1 , Al claim 1 , Mo claim 1 , V claim 1 , and Bi; w claim 1 , x claim 1 , y claim 1 , and z indicate the molar ratio of the elements constituting the oxide) is supported on a silicon dioxide support.3. The treatment method of a spent uranium catalyst according to claim 1 , wherein the alkali solution of step 1 is one or more ...

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Номер записи: 24
22-01-2015 дата публикации

Decontamination Method for Radioactively Contaminated Material

Номер: US20150025294A1
Автор: Grosse Karl-Heinz, Seemann Richard
Принадлежит:

This invention relates to a method for decontaminating radioactively contaminated material, for example construction waste. The material is comminuted by means of voltage pulses and can be divided with a high degree of selectivity into non-contaminated or only weakly contaminated material and contaminated remainder. The majority therefore represents non-contaminated or only weakly contaminated waste material which can be disposed of much more easily than the contaminated remainder. Therefore, the method is particularly suitable for reducing the volume of radioactive waste material which is subject to stringent safety requirements in respect of its storage and disposal and therefore its storage and disposal incur high costs. 1. Method for separating radionuclides from contaminated material , for example construction waste , comprising the steps ofplacing of the contaminated material in a container containing a liquid and comprising at least one first and one second electrode;inducing of at least one voltage pulse between the electrodes so that the contaminated material is comminuted, wherein the radionuclides are accumulated in the liquid;optionally dividing of the contaminated material into at least one contaminated and at least one non-contaminated or only weakly contaminated material phase;separating of solid and liquid constituents; andisolating of the radionuclides from the liquid constituents.2. Method according to claim 1 , wherein the liquid comprises water claim 1 , halogenated hydrocarbons claim 1 , silicone oil or a mixture of these liquids.3. Method according to or claim 1 , wherein the method comprises the division of the contaminated material into at least one contaminated and at least one non-contaminated or only weakly contaminated material phase and wherein a contaminated material phase is cement and a non-contaminated or only weakly contaminated material phase is gravel.4. Method according to claim 3 , wherein the contaminated cement phase and the ...

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Номер записи: 25
23-01-2020 дата публикации

STORING HAZARDOUS MATERIAL IN A SUBTERRANEAN FORMATION

Номер: US20200023416A1
Автор: Muller Elizabeth, Muller Richard A.
Принадлежит:

A hazardous material storage repository includes a drillhole extending into the Earth and including an entry. The drillhole includes a vertical drillhole portion, a transition drillhole portion coupled to the vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion. The hazardous material storage drillhole portion is located below a self-healing geological formation and is vertically isolated, by the self-healing geological formation, from a zone that comprises mobile water. The repository includes a storage canister positioned in the hazardous material storage drillhole portion and sized to fit from the drillhole entry through the vertical drillhole portion, the transition drillhole portion, and into the hazardous material storage drillhole portion. The storage canister includes an inner cavity sized to enclose hazardous material. 163-. (canceled)64. A method for storing hazardous material , comprising:moving a storage canister through an entry of a drillhole that extends into a terranean surface, the entry at least proximate the terranean surface, the storage canister comprising an inner cavity sized enclose hazardous material;moving the storage canister through the drillhole that comprises a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, the hazardous material storage drillhole portion located below a self-healing geological formation, the hazardous material storage drillhole portion vertically isolated, by the self-healing geological formation, from a subterranean zone that comprises mobile water;moving the storage canister into the hazardous material storage drillhole portion; andforming a seal in the drillhole that isolates the storage portion of the drillhole from the entry of the drillhole.65. The method of claim 64 , ...

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Номер записи: 26
28-01-2021 дата публикации

DEEP HUMAN-MADE CAVERN CONSTRUCTION

Номер: US20210025241A1
Автор: Crichlow Henry
Принадлежит:

Systems and/or methods of waste disposal use human-made caverns that are constructed within deep geological formations. A given human-made cavern may be constructed by first drilling out a vertical wellbore to a deep geological formation. Then a bottom portion of the vertical wellbore is jet drilled using an abrasive jetting fluid to form a launch chamber of void volume, that is sized to fit a reaming tool in its deployed open configuration. A reaming tool, in a closed configuration, is then inserted into the vertical wellbore for landing in the launch chamber. The reaming tool is then deployed into its open configuration while in the launch chamber. Reaming operations then occur from the launch chamber directed downwards within the deep geological formation, forming a given human-made cavern. The newly formed human-made cavern may be conditioned and/or configured for receiving amounts of the waste for long-term disposal and/or storage. 1. A method for disposing of waste within at least one human-made cavern , wherein the method comprises steps of:(a) drilling at least one wellbore from a drill site located on a terrestrial surface, wherein the at least one wellbore is substantially vertical and drilled out to at least a predetermined depth;(b) inserting at least one jetting tool within the at least one wellbore to a predetermined location;(c) jet drilling into a geological formation that axially surrounds a portion of the at least one wellbore using the at least one jetting tool to form a launch chamber of a volume of void space within the geological formation;(d) landing at least one reaming tool within the launch chamber;(e) reaming portions of a deep geological formation that are located below the launch chamber to form the at least one human-made cavern; and(f) inserting at least some of the waste into the at least one human-made cavern.2. The method according to claim 1 , wherein the waste is selected from one or more of: nuclear waste claim 1 , radioactive ...

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Номер записи: 27
28-01-2021 дата публикации

MANAGING THE DISPOSAL OF HIGH-LEVEL NUCLEAR WASTE

Номер: US20210027902A1
Автор: Crichlow Henry
Принадлежит:

A method for managing disposal of high-level nuclear waste (HLW) may include: generating electrical power from nuclear fuel; producing HLW as a byproduct from generating the electrical power; encapsulating the HLW within waste-capsules, forming a deep geologic repository for disposing of the encapsulated HLW; and/or loading the HLW into lateral wellbore(s) of the deep geologic repository. The method may also include other steps such as, but not limited to: surface storage and transporting steps of the HLW; licensing steps; receiving payments; closing the deep geologic repository; monitoring, maintaining and/or providing security with respect to the deep geologic repository; and/or using the deep geologic repository for either temporary HLW disposal or permanent HLW disposal. At least some of the steps in the method may be carried by a nuclear power generating company and/or agent(s) thereof; such that the nuclear power generating company takes an active role in the disposal of HLW. 1. A method for managing disposal of high-level nuclear waste , the method comprising steps of:(a) generating electrical power from nuclear fuel;(b) generating the high-level nuclear waste as a byproduct of the step (a);(c) encapsulating predetermined amounts of the high-level nuclear waste within waste-capsules;(d) forming a deep geologic repository, wherein the deep geologic repository comprises at least one lateral wellbore within a deep geologic rock formation, wherein the at least one lateral wellbore is configured to receive the waste-capsules, wherein the deep geologic rock formation is located at least 5,000 feet below a surface of the Earth; and(e) loading the waste-capsules, with the predetermined amounts of the high-level nuclear waste, within the at least one lateral wellbore.2. The method according to claim 1 , wherein the step (a) and the step (b) are carried out by at least one nuclear generating power utility company.3. The method according to claim 2 , wherein the at ...

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Номер записи: 28
31-01-2019 дата публикации

LIGAND-CONTAINING CONJUGATED MICROPOROUS POLYMER AND USE THEREOF

Номер: US20190030513A1
Автор: HAN Xiaoli, HUA Daoben, XU Meiyun
Принадлежит:

The present invention relates to a ligand-containing conjugated microporous polymer, which is obtained by covalent coupling of a conjugated microporous polymer and a uranium complexing ligand. The conjugated microporous polymer comprises an aromatic ring and/or a heterocyclic ring. The uranium complexing ligand is selected from the group consisting of a compound with a group containing phosphorus, a compound with a group containing nitrogen, and a compound with a group containing sulfur. The invention further provides use of the ligand-containing conjugated microporous polymer as a uranium adsorbent. The ligand-containing conjugated microporous polymer the invention is capable of adsorbing the radioactive element uranium in strongly acidic and strong-radiation environments. 1. A ligand-containing conjugated microporous polymer , obtained by covalent coupling of a conjugated microporous polymer and a uranium complexing ligand , wherein the conjugated microporous polymer comprises an aromatic ring and/or a heterocyclic ring , and the uranium complexing ligand is selected from a compound with a group containing phosphorus , a compound with a group containing nitrogen , a compound with a group containing sulfur and any combination thereof.2. The ligand-containing conjugated microporous polymer as claimed in claim 1 , wherein the conjugated microporous polymer is obtained by copolymerization of a first monomer and a second monomer claim 1 , the first monomer and the second monomer being independently selected from the group consisting of benzene claim 1 , a benzene derivative claim 1 , fluorene claim 1 , a fluorene derivative claim 1 , porphyrin claim 1 , a porphyrin derivative claim 1 , pyridine claim 1 , a pyridine derivative claim 1 , thiophene and a thiophene derivative.3. The ligand-containing conjugated microporous polymer as claimed in claim 2 , wherein the group containing phosphorus is selected from a phosphonic acid group claim 2 , a phosphate ester group claim ...

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Номер записи: 29
30-01-2020 дата публикации

RADIONUCLIDE ADSORBENT, METHOD FOR PREPARING THE SAME AND METHOD FOR REMOVING RADIONUCLIDE USING THE SAME

Номер: US20200030771A1
Автор: CHOI Min Kee, HAN Eun Hye, Kim Il Gook, Lee Kune Woo, PARK Chan Woo, YANG Hee-Man, Yoon In Ho
Принадлежит:

Provided is a radionuclide absorbent, a method of preparing the same and a method of removing a radionuclide using the same. The radionuclide absorbent, compared to conventional zeolite, may more selectively remove radioactive cesium (Cs) and/or strontium (Sr) ions even in the presence of various competitive ions (e.g., Na, K, Mg, and Ca) in groundwater or seawater. In addition, the radionuclide absorbent may be prepared by a simple method of thermally treating a mixture of sulfur and zeolite, and thereby, sulfur may be uniformly dispersed in zeolite. 1. A radionuclide absorbent comprising a sulfur-zeolite composite in which sulfur is dispersed in zeolite.2. The radionuclide absorbent of claim 1 , wherein the sulfur-zeolite composite comprises 1 to 25 wt % of sulfur on the basis of the total weight of the composite.3. The radionuclide absorbent of claim 1 , wherein the sulfur-zeolite composite comprises 3 to 20 wt % of sulfur on the basis of the total weight of the composite.4. The radionuclide absorbent of claim 1 , wherein the sulfur-zeolite composite comprises 4 to 12 wt % of sulfur on the basis of the total weight of the composite.5. The radionuclide absorbent of claim 1 , wherein the zeolite is one selected from the group consisting of chabazite (CHA) claim 1 , mordenite (MOR) claim 1 , NaA claim 1 , NaX claim 1 , faujasite (FAU) claim 1 , Linde Type A (LTA) claim 1 , analcime (ANA) claim 1 , Linde Type L (LTL) claim 1 , EMT (EMC-2) claim 1 , MFI (ZSM-5) claim 1 , ferrierite (FER) claim 1 , heulandite (HEU) claim 1 , beta polymorph A (BEA) and MTW (ZSM-12) structures claim 1 , or a combination thereof.6. The radionuclide absorbent of claim 1 , wherein the radionuclide comprises one or more selected from cesium or strontium.7. The radionuclide absorbent of claim 1 , wherein the sulfur-zeolite composite comprises 1 to 25 wt % of sulfur on the basis of the total weight of the composite claim 1 , and the radionuclide is cesium.8. The radionuclide absorbent of claim ...

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Номер записи: 30
01-02-2018 дата публикации

Abyssal Sequestration of Nuclear Waste and Other Types of Hazardous Waste

Номер: US20180033509A1
Автор: Germanovich Leonid, Murdoch Lawrence C., Robinowitz Marvin
Принадлежит:

A system and method of disposing nuclear waste and other hazardous waste includes means for, and the steps of, blending a waste stream, which includes either a radioactive waste or a hazardous waste (or both), with a liquid and, optionally, a solid material to produce a dense fluid and pumping the dense fluid into a tubing string of an injection boring. The dense fluid then exits a perforation in a casing of the injection boring and enters a fracture in a rock strata, where it continues to propagate downward until it reaches an immobilization point. The dense fluid may be a slurry formed by a metal and a cross-linked polymer gel or hydrated clay slurry. The metal can be one that has a melting temperature less than the temperature at the bottom of the injection boring. The solid material could also be other nuclear waste or a radionuclide. 1. A system for abyssal sequestration of nuclear waste and other types of hazardous waste , the system comprising:a gravity fracture filled with a fluid having at least one waste selected from the group consisting of a radioactive waste and a hazardous waste, the fluid being denser than a rock formation into which the fluid is to be disposed so as to cause the rock formation to gravity fracture, the fluid propagating downward in the gravity fracture as the gravity fracture propagates downward.2. A system according to wherein the fluid has a density of at least 3.0 g/cm.3. A system according to wherein the fluid is a slurry.4. A system according to further comprising the slurry including a solid material which is blended with the at least one waste.5. A system according to wherein the solid material is a metal.6. A system according to wherein the metal is selected from the group consisting of bismuth claim 5 , iron claim 5 , lead claim 5 , and copper.7. A system according to wherein the solid material contains one or more radionuclides.8. A system according to wherein a liquid component of the slurry is a metal having a melting ...

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Номер записи: 31
05-02-2015 дата публикации

BERM AND METHOD OF CONSTRUCTION THEREOF

Номер: US20150037100A1
Автор: Dudding Carlton
Принадлежит:

A berm comprises fill material, a covering portion, and a drainage system. The fill material comprises contaminated fill material fully encapsulated by an impermeable membrane. The covering portion comprises structural fill material and at least partially covers the fill material. The drainage system is within the impermeable membrane. 1. A berm comprising:fill material comprising contaminated fill material fully encapsulated by an impermeable membrane;a covering portion comprising structural fill material, the covering portion at least partially covering the fill material; anda drainage system within the impermeable membrane;wherein the contaminated fill material comprises one or more of fossil fuel combustion product, fly ash, bottom ash, boiler slag, flue gas desulphurization material, non-hazardous contaminated soil, contaminated crushed glass, contaminated crushed concrete, contaminated crushed asphalt, sand blast grit, foundry sand, properly de-watered dredge spoils, or combinations thereof; and wherein the contaminated fill material is contaminated with one or more of a metal, an acid, a base, a volatile organic compound, a semi-volatile organic compound, a petroleum product, selenium, mercury, lead, boron, cadmium, thallium, a polycyclic aromatic hydrocarbons compound, or combinations thereof.2. The berm of claim 1 , wherein the impermeable membrane comprises a geomembrane.3. The berm of claim 2 , wherein the geomembrane comprises one of low-density polyethylene (LDPE) claim 2 , high-density polyethylene (HDPE) claim 2 , polyvinyl chloride (PVC) claim 2 , polyurea or polypropylene (PP).4. The berm of claim 1 , wherein the impermeable membrane comprises a plurality of impermeable membrane sections joined with impermeable seams.5. The berm of claim 4 , wherein the plurality of impermeable membrane sections are joined using extrusion welding or fusion welding.6. The berm of claim 1 , wherein the drainage system comprises a perforated pipe surrounded by a ...

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Номер записи: 32
30-01-2020 дата публикации

CONCRETE CASK

Номер: US20200035371A1
Автор: Arai Hiroaki, GOTO Masanori, Okada Keisuke
Принадлежит: HITACHI ZOSEN CORPORATION

A concrete cask enabling suppression of occurrence of stress corrosion cracking (SCC) in a lid welded part of a canister. The concrete cask includes: a metal canister accommodating spent fuel; a concrete container body for accommodating the canister inside the container body; a cooling passage provided between the external peripheral surface of the canister and the internal peripheral surface of the container body, and allowing air for cooling the external peripheral surface of the canister to pass; and a top space provided between the top surface part of the canister, and the inside of a lid of the container body. A baffle plate for suppressing introduction of air rising through the cooling passage to the top space is provided. 1. A concrete cask comprising:a metal canister accommodating spent fuel;a concrete container body for accommodating the canister inside the container body;a cooling passage provided between an external peripheral surface of the canister and an internal peripheral surface of the container body, and allowing air for cooling the external peripheral surface of the canister to pass; anda top space provided between a top surface of the canister, and inside of a lid of the container body, whereina baffle plate for suppressing introduction of air rising through the cooling passage to the top space is provided.2. The concrete cask according to claim 1 , whereinthe baffle plate is mounted on a top external peripheral surface of the canister, and has such a shape that an external periphery expands toward a top.3. The concrete cask according to claim 1 , whereina mounting bracket for mounting the baffle plate on the canister is provided, and the baffle plate is mounted on a top end part of the canister or near the top end part of the canister by the mounting bracket.4. The concrete cask according to claim 1 , whereina coefficient of thermal expansion of the mounting bracket is smaller than a coefficient of thermal expansion of a metal material forming ...

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Номер записи: 33
30-01-2020 дата публикации

METHOD FOR REDUCING AMOUNT OF RADIOACTIVITY OF LIQUID

Номер: US20200035372A1
Автор: Mizuno Minoru
Принадлежит:

According to the present invention, a method which dissolves hydrogen in a liquid that includes a radioactive substance is able to reduce the amount of radioactivity of the liquid. With respect to this method, the radioactive substance may include radioactive cesium, and hydrogen may be dissolved in the liquid by mixing a substance that contains a radioactive substance with a hydrogen water that contains 1.0 ppm or more of hydrogen. 1. A method for reducing an amount of radioactivity of a liquid including a radioactive substance , the method comprising:dissolving hydrogen in the liquid.2. The method according to claim 1 , wherein the hydrogen is dissolved in the liquid by mixing a substance that includes the radioactive substance with a hydrogen water that includes 1.0 ppm or more of hydrogen.3. The method according to claim 1 , wherein the radioactive substance includes radioactive cesium.4. The method according to claim 2 , wherein the radioactive substance includes radioactive cesium. The present invention relates to a method for reducing the amount of radioactivity of liquid.Hydrogen-contained water (also referred to as “hydrogen water” in this specification) has a high reducing power, and therefore has been attempted to be suitable for various uses. For example, Patent Document 1 discloses a radioactivity decontamination device comprising a hydrogen supply unit that supplies a hydrogen gas, a hydrogen water production unit that dissolves the hydrogen gas in a raw material water to produce a hydrogen water, and a spray unit that sprays the produced hydrogen water as a water flow. According to the device, it is possible to efficiently remove (decontaminate) radioactive substances attached to soil or the like.Patent Document 1: Japanese Patent Application Laid-Open No. 2015-78836However, according to the results described in the embodiments of Patent Document 1, by taking soil as a sample, the difference between the case where decontamination is performed using a ...

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Номер записи: 34
12-02-2015 дата публикации

Mechanical press system and method of removing salt using the same

Номер: US20150040727A1
Автор: Zachary William KOSSLOW
Принадлежит: GE HITACHI NUCLEAR ENERGY AMERICAS LLC

A method of removing salt from a dendritic mixture includes loading the dendritic mixture into a mechanical press system. The dendritic mixture includes a metallic dendrite and salt dispersed within the metallic dendrite. The dendritic mixture is heated to liquefy the salt without volatilizing one or more metals of the metallic dendrite. The dendritic mixture is also compressed to obtain a fluidic mixture and an ingot of the metallic dendrite. The fluidic mixture may include molten salt and residual metallic dendrite. The fluidic mixture may be filtered to separate the residual metallic dendrite from the molten salt.

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Номер записи: 35
24-02-2022 дата публикации

SOLIDIFYING-AGENT COMPOSITION CONTAINING ALUMINA CEMENT FOR SOLIDIFYING RADIOACTIVE WASTE AND METHOD FOR SOLIDIFYING RADIOACTIVE WASTE USING SAME

Номер: US20220059249A1
Автор: Cho Nam-Chan, Ju Young-Jong, Ki Kyoung Kuk, Kim Jeong-myeong, Lim Dong Han, Park Jung Hoon
Принадлежит: KEPCO NUCLEAR FUEL CO., LTD.

This invention relates to a solidifying agent for solidifying radioactive waste, and more particularly to a solidifying-agent composition for solidifying radioactive waste, including alumina cement and a gypsum powder. The solidifying-agent composition including alumina cement and a gypsum powder is capable of effectively minimizing an increase in the volume of a solidified radioactive waste product to a level satisfying physical and chemical safety regulations upon the solidification of radioactive waste. 1. A method of solidifying radioactive waste , comprising the steps of:(1) adding a fluidizing agent and water to radioactive waste and performing stirring;(2) adding a solidifying-agent composition comprising alumina cement, and a gypsum powder, and a defoaming agent to the radioactive waste containing the fluidizing agent and the water added in the step (1) and performing stirring; and(3) curing the radioactive waste containing the solidifying-agent composition added in the step (2).2. The method of claim 1 , wherein in the step (2) claim 1 , the solidifying-agent composition is added 33 to 68 parts by weight of 100 parts by weight of the radioactive waste.3. The method of claim 1 , wherein the alumina cement is contained 10 to 70 parts by weight of 100 parts by weight of the composition claim 1 , and the gypsum powder is contained 5 to 50 parts by weight of 100 parts by weight of the composition.4. The method of claim 3 , wherein the composition further comprises claim 3 , of the total of 100 parts by weight thereof claim 3 , 1 to 10 parts by weight of a resin powder claim 3 , 0.01 to 3 parts by weight of a reaction accelerator claim 3 , 0.01 to 5 parts by weight of a retention agent claim 3 , 0.01 to 5 parts by weight of the defoaming agent claim 3 , and 0.1 to 10 parts by weight of a fluidizing agent.5. The method of claim 1 , wherein the curing in the step (3) is performed for 28 days. This patent application is a divisional of U.S. patent application Ser. ...

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Номер записи: 36
01-05-2014 дата публикации

Thermal Treatment of Carbonaceous Waste, Improved by the Choice of Gas Injected

Номер: US20140121440A1
Автор: Laurent Gérard
Принадлежит: ELECTRICITE DE FRANCE

A method is provided for the decontamination of radioactive carbonaceous material, such as graphite, in which an injection of steam is planned into the material, concurrent with a first roasting thermal treatment of the material at a temperature between 1200° C. and 1500° C. Advantageously, the first treatment may be followed by a second treatment at a lower temperature with an injection of carbon oxide for oxidation according to the Boudouard reaction. 1. A method of decontamination of radioactive carbonaceous material , comprising:inducting steam into said material, concurrent with a first roasting thermal treatment of the material at a temperature between 1200° C. and 1500° C.2. The method according to claim 1 , wherein steam is injected into a reactor claim 1 , whose content of water is measured for controlling an amount of steam injected.3. The method according to claim 2 , further comprising:drying the material to control the amount of water present in the reactor prior to the first roasting thermal treatment.4. The method according to claim 1 , further comprising:injecting the steam with a gaseous fluidizing agent comprising hydrogen.5. The method according to claim 1 , wherein a temperature of about 1300° C. is applied during the first thermal treatment.6. The method according to claim 1 , further comprising:providing a second thermal treatment of roasting with injection of carbon oxide gas.7. The method according to claim 6 , wherein the second thermal treatment is performed at a temperature between 900° C. and 1100° C.8. The method according to claim 6 , further comprising:diluting the carbon oxide gas in an inert gas.9. The method according to claim 8 , wherein the dilution of the carbon oxide gas is increased during the second thermal treatment from a proportion of about 75% inert gas to about 90% inert gas at the end of the second thermal treatment.10. The method according to claim 6 , wherein the temperature applied during the second thermal treatment ...

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Номер записи: 37
07-02-2019 дата публикации

METHOD AND DEVICE FOR DISPOSING NUCLEAR WASTE USING DEEP GEOLOGICAL REPOSITORY

Номер: US20190043629A1
Автор: Han Jinglan, SUN Liyao
Принадлежит: HARBIN QINGRUI SCIENCE AND TECHNOLOGY DEVELOPMENT CO., LTD.

A device comprises a raw material conveyor, a raw material mixer, a liquid waste conveying pipeline, an additive tank, a powder waste conveyor, an output pump, a liquid supply pump, a liquid supply manifold, an output manifold, a mixed liquid conveying pipeline, a high-pressure injection pump, a high-pressure pipeline, and a wellhead sealing device. The method includes: 1. A method for disposing nuclear waste using a deep geological repository , comprising the following steps:drilling a well down to the granite stratum;forming a fracture in the granite stratum by injecting liquid for forming fracture into the underground granite stratum; andinjecting a sand-carrying feed liquid containing powder nuclear waste or liquid nuclear waste to the fracture in the underground granite stratum by a high-pressure injection pump, so that the sand-carrying feed liquid stays in the fracture of the granite stratum for coagulating, and after disposal a wellhead is sealed by wellhead cementing concrete, thereby permanently storing the nuclear waste in the fracture of the underground granite stratum, and achieving the purpose of effective disposal to the nuclear waste.2. The method for disposing nuclear waste using a deep geological repository of claim 1 , further comprising the following steps:sampling the granite stratum; and{'sup': '2', 'forming the fracture in the granite stratum, wherein the pressure is set according to the density of the granite stratum, and the pressure can preferably be set at 40 to 140 MPa/cm.'}3. The method for disposing nuclear waste using a deep geological repository of claim 1 , further comprising the following step:adding an active agent into the liquid for forming fracture, wherein the active agent is composed of a surfactant and oxalic acid in a ratio of 4.5 to 5:1 to 1.5%, in which the surfactant is linear alkylbenzene sulfonates, tetrapropylene benzene sulfonate, dioctyl sulfosuccinate, sodium dodecyl benzene sulfonate or sodium stearyl sulfate, and ...

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Номер записи: 38
19-02-2015 дата публикации

SYSTEMS, METHODS, AND FILTERS FOR RADIOACTIVE MATERIAL CAPTURE

Номер: US20150049852A1
Автор: Loewen Eric P.
Принадлежит: GE-HITACHI NUCLEAR ENERGY AMERICAS LLC

A system configured to passively filter radioactive materials from a flow may include one or more particulate removal devices; one or more water removal devices; and/or one or more radionuclide removal devices. At least one of the one or more particulate removal devices may mechanically remove particulates of the radioactive materials from the flow. At least one of the one or more water removal devices mechanically may remove water from the flow. At least one of the one or more radionuclide removal devices may remove radioactive aerosols, reactive radioactive gases, or radioactive aerosols and reactive radioactive gases from the flow using engineered filter media. A filter may include a body, including an inlet and an outlet. The body may be configured to store filter media, to contain pressure from gas explosions, and/or to allow the stored filter media to move toward the outlet when pressure at the inlet increases. 1. A system configured to passively filter radioactive materials from a flow , the system comprising:one or more particulate removal devices;one or more water removal devices; andone or more radionuclide removal devices;wherein at least one of the one or more particulate removal devices mechanically removes particulates of the radioactive materials from the flow,wherein at least one of the one or more water removal devices mechanically removes water from the flow, andwherein at least one of the one or more radionuclide removal devices removes radioactive aerosols, reactive radioactive gases, or radioactive aerosols and reactive radioactive gases from the flow using engineered filter media.2. The system of claim 1 , wherein the radioactive materials comprise one or more of reactive gaseous radioactive materials claim 1 , liquid radioactive materials claim 1 , and particulate radioactive materials.3. The system of claim 1 , wherein in a direction of the flow claim 1 , the one or more particulate removal devices precede the one or more water removal ...

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Номер записи: 39
14-02-2019 дата публикации

Abyssal Sequestration of Nuclear Waste and Other Types of Hazardous Waste

Номер: US20190051423A1
Автор: Germanovich Leonid N., Murdoch Lawrence C., Robinowitz Marvin
Принадлежит:

A system and method of disposing nuclear waste and other hazardous waste includes means for, and the steps of, blending a waste stream, which includes either a radioactive waste or a hazardous waste (or both), with a liquid and, optionally, a solid material to produce a dense fluid and pumping the dense fluid into a tubing string of an injection boring. The dense fluid then exits a perforation in a casing of the injection boring and enters a fracture in a rock strata, where it continues to propagate downward until it reaches an immobilization point. The dense fluid may be a slurry formed by a metal and a cross-linked polymer gel or hydrated clay slurry. The metal can be one that has a melting temperature less than the temperature at the bottom of the injection boring. The solid material could also be other nuclear waste or a radionuclide. 1a fluid denser than a rock formation into which the fluid is to be disposed, wherein when the fluid is disposed in the rock formation, the fluid causes the vertical downward propagating fracture in the rock formation; said fracture continuing to propagate vertically downward as the fluid propagates vertically downward in said fracture until an immobilization point of the fluid is reached.. A system for creating a vertical downward propagating fracture in a rock formation, the system comprising: This application is a continuation of U.S. patent application Ser. No. 15/682,238 filed Aug. 21, 2017, which was a continuation of U.S. patent application Ser. No. 15/344,243, filed Nov. 4, 2016, now U.S. Pat. No. 9,741,460, which was a continuation of U.S. patent application Ser. No. 14/942,643, filed Nov. 16, 2015, now U.S. Pat. No. 9,700,922, which was a continuation of U.S. patent application Ser. No. 14/129,504 filed Dec. 26, 2013, now U.S. Pat. No. 9,190,181, which was a United States National Phase of PCT Patent Application No. US2012/045084 filed on Jun. 29, 2012, which claimed priority to U.S. Provisional Patent Application No. 61/ ...

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Номер записи: 40
25-02-2021 дата публикации

Disassembly and disposal of munition components

Номер: US20210057121A1
Автор: Henry Crichlow
Принадлежит: Individual

Methods for disposing of munition components may include separating propellants from heavy metal penetrators and disposing of those separated components into different types of geological formations. The initially solid form propellants may be converted into a stable liquified propellant form, by a particular disclosed process, that may be injected within salt water (injection) disposal wells, where distal portions of such salt water disposal wells may be located in a geological formation of substantially at least one salt. The separated heavy metal penetrators (with or without their associated projectile jackets) may be disposed of within human-made caverns, where such human-made caverns may be located within a deep geological formation that is often 2,000 feet or more below the Earth's surface. The heavy metal penetrators may include uranium (depleted uranium). Portions of a given munition, to be disposed of, may be radioactive.

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Номер записи: 41
25-02-2021 дата публикации

DISPOSAL OF DEPLETED URANIUM PRODUCTS IN DEEP GEOLOGICAL FORMATIONS

Номер: US20210057122A1
Автор: Crichlow Henry
Принадлежит:

The invention is of systems and methods for long-term disposal/storage of depleted uranium products and materials (such as munitions), in solid, liquid, and other physical forms in in lateral wellbores and/or in human-made caverns derived from a wellbore, located within deep geologic rock formations. Converted and/or modified depleted uranium products, materials, and/or wastes may be processed, made into slurries, chemically treated for long duration disposal, and/or implemented in waste disposal capsules and/or maintained as cementitious material or solids which are then transported and finally disposed of into lateral wellbores or human-made caverns within the deep geologic rock formations. Void space around depleted uranium products, materials, and/or wastes may be filled with a protective medium. 1. A method for disposing of depleted uranium waste into at least one geologically deep repository , wherein the method comprises steps of:(a) collecting at least some of the depleted uranium waste; wherein the depleted uranium waste is of at least one of the following different formats: depleted uranium munitions with depleted uranium penetrators; the depleted uranium penetrators without other components of the depleted uranium munitions; solid depleted uranium material; liquid depleted uranium material; or depleted uranium hexafluoride; wherein when the different format of the at least some of the depleted uranium waste is the depleted uranium munitions with the depleted uranium penetrators, the step (b) comprises separating the depleted uranium penetrators from the other components of the depleted uranium munitions and packaging the depleted uranium penetrators into cylindrical capsules;', 'wherein when the different format of the at least some of the depleted uranium waste is the depleted uranium penetrators without the other components of the depleted uranium munitions, the step (b) comprises packaging the depleted uranium penetrators into the cylindrical capsules ...

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Номер записи: 42
13-02-2020 дата публикации

ZINC DOSING FOR DECONTAMINATING LIGHT-WATER REACTORS

Номер: US20200051706A1
Автор: Jordan David, NIEDER Dietmar
Принадлежит:

The invention relates to a method for decontaminating a radioactively contaminated metal surface, wherein the metal surface is brought in contact with a decontamination solution, which comprises a complexing agent and a transition metal. The invention further relates to such a decontamination solution and to the use thereof to decontaminate a metal surface. 1. A method for decontaminating a radioactively contaminated metal surface , comprising the step of:bringing at least a portion of the metal surface into contact with a decontamination solution comprising a complexing agent and a transition metal.2. The method as per claim 1 , wherein the transition metal is selected from the group consisting of zinc claim 1 , nickel claim 1 , cobalt or mixtures thereof.3. The method as per at least one of the preceding claims claim 1 , wherein the concentration of the transition metal is in the range of from ≥0.5 to ≤15 mg/kg.4. The method as per at least one of the preceding claims claim 1 , wherein the transition metal is zinc and is present in a concentration in the range of from ≥2 to ≤5 mg/kg.5. The method as per at least one of the preceding claims claim 1 , wherein Co and/or Co ions are removed from the metal surface.6. The method as per at least one of the preceding claims claim 1 , wherein the decontamination solution is introduced into the primary circuit of a nuclear reactor.7. The method as per at least one of the preceding claims claim 1 , wherein the decontamination solution is circulated.8. The method as per at least one of the preceding claims claim 1 , wherein claim 1 , as the first method step claim 1 , the method comprises a pre-oxidation step or a reduction step for oxidizing or reducing the radioactively contaminated metal surface.9. The method as per at least one of the preceding claims claim 1 , wherein the method also comprises the step of:removing at least some of the radioactive isotopes present in the decontamination solution.10. The method as per ...

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Номер записи: 43
17-03-2022 дата публикации

STORING HAZARDOUS MATERIAL IN A SUBTERRANEAN FORMATION

Номер: US20220080481A1
Автор: Muller Elizabeth, Muller Richard A.
Принадлежит:

Techniques for storing hazardous material include moving a storage canister sized to enclose radioactive hazardous material through an entry of a drillhole that extends into a terranean surface and is at least proximate the terranean surface; moving the storage canister from the entry through an angled drillhole portion that is coupled to the entry and deviates from true vertical at an angle; moving the storage canister from the angled drillhole portion to a hazardous material storage drillhole portion coupled to the angled drillhole portion; moving the storage canister into the hazardous material storage drillhole portion; and forming a seal in the drillhole that isolates the hazardous material storage portion of the drillhole from the entry of the drillhole. 1. (canceled)2. A method for storing hazardous material , comprising:moving a storage canister through an entry of a drillhole that extends into a terranean surface, the entry at least proximate the terranean surface, the storage canister comprising an inner cavity sized to enclose radioactive hazardous material;moving the storage canister from the entry through an angled drillhole portion that is coupled to the entry and deviates from true vertical at an angle;moving the storage canister from the angled drillhole portion to a hazardous material storage drillhole portion coupled to the angled drillhole portion;moving the storage canister into the hazardous material storage drillhole portion; andforming a seal in the drillhole that isolates the hazardous material storage portion of the drillhole from the entry of the drillhole.3. The method of claim 2 , wherein the angled drillhole portion comprises a proximate end coupled to the entry at a first depth and a distal end coupled to the hazardous material storage drillhole portion opposite the proximate end at a second depth deeper than the first depth.4. The method of claim 2 , wherein the hazardous material storage drillhole portion comprises at least one of a ...

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Номер записи: 44
22-05-2014 дата публикации

Method and apparatus for identification, stabilization and safe removal of radioactive waste and non hazardous waste contained in buried objects

Номер: US20140142365A1
Автор: Halliwell Steve, Soyfer Boris, Stone Keith
Принадлежит: VJ TECHNOLOGIES INC.

A method and apparatus for the stabilization and safe removal of buried waste that is tested and classified as being transuranic or not transuranic waste and disposed accordingly. The buried waste (usually in vertical pipe units) is enclosed in a casing and ground and mixed with the surrounding soil. This process allows for chemical reactions to occur that stabilizes the mixture. The entire process is contained within the casing to avoid contamination. In situ or external testing is done for radio isotopes to classify the waste. If it is classified as transuranic the waste is removed in a controlled way into a retrieval enclosure and disposed off in drums. If the waste is not transuranic then grout is introduced into the mixture, allowed to set and the resulting monolith is removed and buried in trenches. 2. The method of wherein an enclosure base is used for centering the casing over the buried waste.3. The method of wherein the casing is mechanically driven to enclose the buried waste.4. The method of wherein the grinding mechanism is a rotating augering tool that is housed in an augering tool enclosure concentrically fitted over the casing via an interface enclosure that is fitted over the enclosure base.5. The method of wherein the rotating motion of the augering tool is provided by a drilling rig.6. The method of wherein dust is controlled by using dust control chemicals inserted through one or more openings provided in the interface enclosure.7. The method of wherein the augering tool is cleaned prior to removal by using high pressure water through openings provided in the interface enclosure.8. The method of wherein testing the mixture is done by inserting a detector through a hollow stem auger.9. The method of wherein the detector tests for the presence of radio isotopes in the mixture in situ as the hollow stem auger is rotating in the mixture.10. The method of wherein test results are remotely monitored for classification as transuranic or not transuranic ...

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Номер записи: 45
12-03-2015 дата публикации

Method of Chemical Decontamination for Carbon Steel Member of Nuclear Power Plant

Номер: US20150073198A1
Автор: AIZAWA Motohiro, HOSOKAWA Hideyuki, ISHIDA Kazushige
Принадлежит:

A circulation pipe of a chemical decontamination apparatus including a malonic acid injection apparatus and an oxalic acid injection apparatus is connected to a purification system pipe, which is made of carbon steel, of a boiling water nuclear power plant. A malonic acid aqueous solution is injected from the malonic acid injection apparatus into the circulation pipe. An oxalic acid aqueous solution is injected from the oxalic acid injection apparatus into the circulation pipe. A reduction decontaminating solution including a malonic acid of 5200 ppm and an oxalic acid within a range of 50 to 400 ppm is supplied into the purification system pipe through the circulation pipe. Reduction decontamination for an inner surface of the purification system pipe is executed. After the reduction decontamination for the purification system pipe finishes, the malonic acid and oxalic acid included in the solution are decomposed and furthermore, the solution is purified. 1. A method of chemical decontamination for a carbon steel member of a nuclear power plant , comprising steps of:bringing a reduction decontaminating solution including a malonic acid and an oxalic acid within a range from 50 to 400 ppm into contact with a surface of a carbon steel member of a nuclear power plant; andexecuting reduction decontamination for the surface of the carbon steel member by the reduction decontaminating solution.2. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 1 , comprising step of:removing cations eluted from the carbon steel member into the reduction decontaminating solution by the reduction decontamination, from the reduction decontaminating solution.3. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 1 , comprising step of:injecting oxygen gas into the reduction decontaminating solution including the malonic acid and the oxalic acid within the range from 50 to 400 ...

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Номер записи: 46
16-03-2017 дата публикации

Method and System for Removing Radioactive Nuclides from Water

Номер: US20170073248A1
Автор: Gabriel Kay Gunther, Kaschek Martin, Poth Heiko, Shoar Behnaz
Принадлежит: ItN Nanovation AG

The present invention concerns a method for the removal of radionuclides from water, wherein at least one absorption additive that has an absorptive effect on the nuclides and at least one filter device that is impermeable to the adsorption additive and the nuclides absorbed thereon are used. An improved removal rate is achieved with reduced equipment expenditure when an adsorption layer is formed from the adsorption additive on an inflow-side surface of the respective filter device. 1. Method for the removal of radionuclides from water ,wherein at least one absorption additive that has an absorptive effect on the nuclides is used, andwherein at least one filter device that is impermeable to the respective adsorption additive is used,characterized in thatan adsorption layer is produced from the adsorption additive on an inflow-side surface of the respective filter device.2. Method as claimed in claim 1 ,characterized in thatthe adsorption layer is produced from an adsorption additive that is essentially free of nuclides.3. Method as claimed in claim 1 ,characterized in thatthe adsorption layer is produced from an adsorption additive that is at least 50% free of nuclides.4. Method as claimed in claim 1 ,characterized in thatadsorption of the nuclides takes place during a filtration phase of a filtration process largely or essentially in the adsorption layer.5. Method as claimed in claim 1 ,characterized in thatat least 50% of adsorption of the nuclides takes place during a filtration phase of a filtration process in the adsorption layer.6. Method as claimed in claim 1 ,characterized in thatproduction of the adsorption layer takes place during a start phase of a filtration process by means of addition of the adsorption additive to a water flow flowing through the respective filter device.7. Method as claimed in claim 6 ,characterized in thatthe adsorption additive is added to the water flow largely or essentially only during the start phase.8. Method as claimed in ...

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Номер записи: 47
05-06-2014 дата публикации

GREEN RECYCLED MATERIAL COMPONENT WET WELL

Номер: US20140154013A1
Автор: Mehr Nasser Fred
Принадлежит:

A wet well design that utilizes recycled material in sectional components to construct wet wells on site is disclosed. Traditional wet well components are made of pre-cast concrete. Green Recycled Material Component Wet Well components are constructed using recycled plastic, recycled steel and recycled Styrofoam. The method described herein for constructing the components and assembling the wet well on site addresses several logistical problems associated with the pre-cast concrete design including reducing project start to completion time, delivery costs, large crane rental costs and power line relocation costs. This design also enables construction of oblong wet wells in medians and other restricted areas as components can be straight or curved sections. Finally, this design eliminates the shifting of traditional cement well components due to uplift from underground water pressure through the use of a new anchoring system. Shifting can result in groundwater intrusion into the well. 1. Green wastewater pump station circular and oblong wet wells constructed by stacking 2 , 3 , 4 and 5 modular circular cylinder and straight sections for required well depths of 12′ , 18′ , 24′ and 30′ respectively and having a mechanical anchoring system; said modular cylinder and straight sections constructed from recycled plastic , recycled Styrofoam and recycled steel scraps in the form of welded wire mesh forming the walls of the wet well and by wet well top and bottom slabs enclosing the modular cylinder wet well walls; said circular and straight sections comprising identical top and bottom frames made of recycled plastic with an inner diameter equal to the desired wet well diameter , vertical members constructed from recycled PVC materials , Styrofoam space refill that fills the space between the vertical members , connecting angles used to connect said vertical members to said top and bottom frames and module stacking connecting bolts that connect said top module frames to ...

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Номер записи: 48
07-03-2019 дата публикации

METHOD FOR TREATING WASTE WATER FROM THE DECONTAMINATION OF A METAL SURFACE, WASTE-WATER TREATMENT DEVICE AND USE OF THE WASTE-WATER TREATMENT DEVICE

Номер: US20190074099A1
Автор: SEMPERE BELDA Luis, TOPF Christian
Принадлежит: Framatome GmbH

A method for treating waste water from the decontamination of a metal surface in a primary coolant circuit of a nuclear reactor comprises discharging a predetermined amount of an oxidation solution from the primary coolant circuit into a reduction zone connected to the primary coolant circuit and reacting the oxidation solution with a reducing agent to form a reaction solution that is freed of oxidizing agent, and passing the reaction solution over an ion-exchange resin in order to form a desalinated solution, and returning the desalinated solution to the primary coolant and/or disposing of the desalinated solution. A waste water treatment apparatus for carrying out the method is also provided. 1. A method for treating waste water from decontamination of a metal surface in a primary coolant circuit of a nuclear reactor , characterized in that the method comprises steps:a) introducing an oxidizing agent into a primary coolant in the primary coolant circuit to form an oxidation solution, and circulating the oxidation solution in the primary coolant circuit to contact the oxidation solution with the metal surface;b) during or after step a), discharging a predetermined amount of the oxidation solution from the primary coolant circuit into a reduction zone connected to the primary coolant circuit;c) in the reduction zone, reacting the oxidizing agent with a reducing agent to form a reaction solution freed of the oxidizing agent;d) passing the reaction solution over an ion-exchange resin to form a desalinated solution; ande) returning the desalinated solution to the primary coolant and/or temporarily storing and/or disposing of the desalinated solution.2. The method of claim 1 , characterized in that the oxidizing agent is a permanganate.3. The method according to claim 1 , characterized in that the reducing agent is an aliphatic dicarboxylic acid.4. The method according to claim 1 , characterized in that the primary coolant circuit comprises at least one coolant pump ...

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Номер записи: 49
16-03-2017 дата публикации

ABYSSAL SEQUESTRATION OF NUCLEAR WASTE AND OTHER TYPES OF HAZARDOUS WASTE

Номер: US20170076829A1
Автор: Germanovich Leonid, Murdoch Lawrence C., Robinowitz Marvin
Принадлежит:

A system and method of disposing nuclear waste and other hazardous waste includes means for, and the steps of, blending a waste stream, which includes either a radioactive waste or a hazardous waste (or both), with a liquid and, optionally, a solid material to produce a dense fluid and pumping the dense fluid into a tubing string of an injection boring. The dense fluid then exits a perforation in a casing of the injection boring and enters a fracture in a rock strata, where it continues to propagate downward until it reaches an immobilization point. The dense fluid may be a slurry formed by a metal and a cross-linked polymer gel or hydrated clay slurry. The metal can be one that has a melting temperature less than the temperature at the bottom of the injection boring. The solid material could also be other nuclear waste or a radionuclide. 1. A method of disposing waste , the method comprising the steps of:(i) blending a waste material to be disposed of with a liquid to produce a dense fluid denser than a surrounding rock formation;(ii) pumping a portion of the dense fluid into a tubing string of an injection boring; and(iii) gravity fracturing the surrounding rock formation using the portion of the dense fluid;the portion of the dense fluid after step (iii) continuing to propagate downward in a gravity fracture as the gravity fracture continues to propagate downward.2. A method according to wherein a second portion of the dense fluid after being pumped into the tubing string of the injection boring enters the gravity fracture and continues a downward travel as the dense fluid drains from the injection boring.3. A method according to wherein the portion of the dense fluid after entering the gravity fracture continues a downward travel after becoming detached from any dense fluid remaining in the injection boring.4. A method according to wherein the portion of the dense fluid after entering the gravity fracture continues a downward travel and remains connected by a ...

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Номер записи: 50
18-03-2021 дата публикации

GRIT BLASTING

Номер: US20210082593A1
Автор: HAILEMARIAM Leaelaf Mengislu, Majamaa Katariina
Принадлежит: DDP Specialty Electronic Materials US, Inc.

Provided is a process for blast cleaning comprising subjecting a work surface to a blast stream, wherein the work surface is metal that is contaminated with one or more radioactive moiety, wherein the blast stream comprises water and one or more resin particles. 1. A process for blast cleaning comprising subjecting a work surface to a blast stream ,wherein the work surface is metal that is contaminated with one or more radioactive moiety,wherein the blast stream comprises water and one or more resin particles.2. The process of claim 1 , wherein the metal is steel.3. The process of claim 1 , wherein the radioactive moiety comprises one or more radioisotope of cobalt claim 1 , nickel claim 1 , or chromium.4. The process of claim 1 , wherein the radioactive moiety comprises Co.5. The process of claim 1 , wherein the resin particles comprise styrenic resin.6. The process of claim 1 , wherein the resin particles comprise particles of one or more ion exchange resin selected from the group consisting ofion exchange resins having covalently bound sulfonic groups;ion exchange resins having covalently bound carboxylic groups;ion exchange resins having covalently bound secondary amine groups;ion exchange resins having covalently bound tertiary amine groups;ion exchange resins having covalently bound quaternary ammonium groups;and mixtures thereof.7. The process of claim 1 , wherein the pH of the blast stream is 4 or lower.8. The process of claim 1 , wherein the blast stream has a temperature of 15° C. to 35° C. It is often desired to decontaminate metal surfaces that have become contaminated with radioactive isotopes. For example, when a nuclear reactor is partially or fully decommissioned, it becomes desirable to remove radioactivity from metal parts. In particular, steel items that have been exposed to neutron radiation become contaminated with cobalt-60, and much of that contamination is located near the surface of the steel item. Previously known methods of decontamination ...

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Номер записи: 51
31-03-2022 дата публикации

CHEMICAL DECONTAMINATION METHOD

Номер: US20220102019A1
Автор: Hamada Ryo, HOSOKAWA Hideyuki, Oohira Takashi, Oouchi Satoshi, Ota Nobuyuki, Yanagisawa Shintaro
Принадлежит: HITACHI-GE NUCLEAR ENERGY, LTD.

A chemical decontamination method capable of improving the decontamination efficiency of chemical decontamination of a steam dryer in the RPV is provided. In particular, the decontamination method includes feeding a chemical decontamination aqueous solution into a reactor pressure vessel in which a steam dryer is arranged, and after chemical decontamination of the steam dryer, the water level of the chemical decontamination aqueous solution existing in the reactor pressure vessel is lowered to a first water level below the lower end of the steam dryer. 1. A chemical decontamination method comprising:feeding a chemical decontamination aqueous solution into a reactor pressure vessel in which a steam dryer is arranged, andafter chemical decontamination of the steam dryer, a water level of the chemical decontamination aqueous solution existing in the reactor pressure vessel is lowered to a first water level below a lower end of the steam dryer.2. The chemical decontamination method according to claim 1 , whereinthe chemical decontamination of the steam dryer is performed by raising the water level of the chemical decontamination aqueous solution in the reactor pressure vessel to a second water level above an upper end of the steam dryer, andthen the water level of the chemical decontamination aqueous solution is lowered from the second water level to the first water level.3. The chemical decontamination method according to claim 2 , whereinthe first water level is a water level in a range below the lower end of the steam dryer and above a shroud head attached to an upper end of a cylindrical reactor core shroud arranged in the reactor pressure vessel.4. The chemical decontamination method according to claim 2 , whereinthe second water level is located above the upper end of the steam dryer, and is a water level within a range at or below a position of a lower surface of a pressing member that is attached to an upper lid that seals the reactor pressure vessel and presses ...

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Номер записи: 52
31-03-2022 дата публикации

ELECTROLYTE FOR ELECTROCHEMICAL DECONTAMINATION, AND PREPARATION METHOD AND APPLICATION THEREOF

Номер: US20220102020A1
Автор: CUI Yanmin, DI Yutong, Gu Wenjie, XU Liguo, ZHAO Lingjun
Принадлежит: China Nuclear Sichuan Environmental Protection Engineering Co., Ltd.

The present disclosure relates to the technical field of radioactive waste treatment and provides an electrolyte for electrochemical decontamination and a preparation method and application thereof. The electrolyte for electrochemical decontamination provided in the present disclosure is an aqueous solution including the following solutes: phosphoric acid, oxalic acid, citric acid, tartaric acid, hydrogen peroxide and glacial acetic acid. According to the present disclosure, an electrolyte for electrochemical decontamination that has a good decontamination effect and allows for fast decontamination is obtained by reasonably combining different types of solutes and controlling the levels of the solutes, and resulting secondary waste solution and residues are easy to treat. The electrolyte for electrochemical decontamination is suitable for overall or local electrochemical decontamination of radioactively contaminated stainless steel scrap. In addition, in the electrolyte for electrochemical decontamination provided in the present disclosure, the oxalic acid, the citric acid, the tartaric acid and the glacial acetic acid are organic acids and no N and Cl ions are present, which may be highly advantageous for a glass solidification system for spent decontamination solution. Thus, the resulting waste solution can be treated readily by the glass solidification system. 1. An electrolyte for electrochemical decontamination , wherein the electrolyte for electrochemical decontamination is an aqueous solution comprising the following solutes by mass:45% to 80% by mass of phosphoric acid, 5 g/L to 10 g/L of oxalic acid, 1 g/L to 10 g/L of citric acid, 1 g/L to 2 g/L of tartaric acid, 1 g/L to 5 g/L of hydrogen peroxide, and 5 g/L to 10 g/L of glacial acetic acid.2. The electrolyte for electrochemical decontamination according to claim 1 , wherein the electrolyte for electrochemical decontamination comprises 50% to 70% by mass of the phosphoric acid claim 1 , 5.5 g/L to 8 g/L ...

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Номер записи: 53
05-05-2022 дата публикации

Systems and methods for low level waste disposal

Номер: US20220134397A1
Автор: Henry Crichlow
Принадлежит: Individual

Open pit mine (OPM) structures are modified or built new for use in disposing of low-level radioactive/nuclear waste (LLW). A drainage system is added to the OPM to drain water, such as, but not limited to, rain water, out of a volume of the OPM and to a particular geologic zone located far below the OPM that is isolated away from the local water table. Cells are formed within the volume of the OPM that are configured to receive the LLW. Cells are added to the OPM from a bottom towards a top of the OPM. Void spaces around the LLW materials within the cells are filled in with a protective-medium to mitigate against radionuclide migration away from the LLW materials within the cells. The protective-medium may be a blend of carbon nanotubes and a foam cement slurry. The carbon nanotubes may be made from reacting ethylene with vermiculite.

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Номер записи: 54
19-06-2014 дата публикации

METHOD FOR CHEMICALLY STABILIZING URANIUM CARBIDE COMPOUNDS, AND DEVICE IMPLEMENTING THE METHOD

Номер: US20140171724A1
Автор: Brothier Méryl, Fiquet Olivier, Marchand Mickael
Принадлежит: Commissariat A L'Energie Atomique Et Aux Energies Alternatives

A process for chemical stabilization of a uranium carbide compound having formula: UC+yC with x≧1 or 2 and y>0, x and y being true numbers, placed in a stabilization chamber, comprises: a rise in chamber internal temperature for “oxidation” of the compound based on uranium carbide between approximately 380° C. and 550° C., the chamber being fed with a neutral gas; isothermal oxidative treatment at the oxidation temperature, the chamber being placed under Opartial pressure; controlling completion of stabilization of the compound, comprising monitoring the amount of molecular oxygen consumed and/or carbon dioxide or carbon dioxide and carbon monoxide given off, until achievement of an input set-point value for the amount of molecular oxygen, of a minimum threshold value for the amount of carbon dioxide or minimum threshold values for the carbon dioxide and carbon monoxide. A device implements the process. 2. The process for the chemical stabilization of a uranium carbide compound as claimed in claim 1 , wherein the stage of controlling the completion of the stabilization additionally comprises the monitoring of variation in weight of the solid compounds based on carbon and uranium in the chamber claim 1 , an increase in weight being correlated with the oxidation of uranium carbide in progress.3. The process for the chemical stabilization of a uranium carbide compound as claimed in claim 1 , wherein stage of controlling the completion of the stabilization is carried out with the application of a rise in temperature of the internal temperature of said chamber between said oxidation temperature and the temperature of oxidation of the carbon claim 1 , said temperature being excluded from the interval claim 1 , and monitoring the presence of COgiven off.4. The process for the chemical stabilization of a uranium carbide compound as claimed in claim 1 , comprising the introduction of a water vapor partial pressure into said chamber before and/or during and/or after the ...

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Номер записи: 55
30-03-2017 дата публикации

MIGRATION PREVENTION SYSTEM FOR RADIOACTIVE WASTEWATER OF UNDERGROUND NUCLEAR POWER PLANT

Номер: US20170092384A1
Автор: FANG Yanguo, HAN Qianlong, He Jie, LI Hongbin, LIU Haibo, NIU Xinqiang, SHI Huatang, SU Lijun, XU Nianfeng, YANG Qigui, YU Weina, ZHAO XIN, ZHOU Shuda
Принадлежит:

A migration prevention system for radioactive wastewater from an underground nuclear power plant. The underground nuclear power plant includes a nuclear island including an underground cavern group including a reactor cavity and auxiliary cavities. The migration prevention system includes a protective layer coating the reactor cavity and an impermeable layer surrounding the nuclear island. The protective layer includes an inner liner, a drainage layer, and a filling layer of rock fractures in that order. The inner liner is configured to prevent exosmosis of the radioactive wastewater of the reactor cavity. The drainage layer is configured to gather and drain seepage water. The impermeable layer is disposed in the periphery of the underground cavern group including the reactor cavity and the auxiliary cavities, and is configured to isolate the underground cavern group from natural underground water. 2. The system of claim 1 , wherein the filling layer of rock fractures comprises an inner filling layer of rock fractures and an outer filling layer of rock fractures.3. The system of claim 2 , wherein the reactor cavity is surrounded by the inner liner claim 2 , the inner filling layer of rock fractures claim 2 , the drainage layer claim 2 , and the outer filling layer of rock fractures from inside to outside in that order.4. The system of claim 3 , wherein the inner liner is a reinforced concrete structure claim 3 , or the inner liner is a reinforced concrete structure plus waterproof board.5. The system of claim 4 , wherein the reinforced concrete structure is impermeable.6. The system of claim 3 , wherein the inner filling layer of rock fractures and the outer filling layer of rock fractures both comprise grouting materials in the rock fractures and rock mass.7. The system of claim 3 , wherein the drainage layer comprises multiple layers of first drainage tunnels and first drainage holes communicating with the first drainage tunnels.8. The system of claim 1 , wherein ...

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Номер записи: 56
16-04-2015 дата публикации

Radioactive sludge transfer apparatus

Номер: US20150101692A1
Автор: Masahiro Aizawa, Osamu Ogata
Принадлежит: Taihei Dengyo Kaisha Ltd

There is provided a radioactive sludge transfer apparatus for safely and surely transferring radioactive sludge contaminated by radioactive substance stored in a sludge storage tank to a transfer tank for the sake of tank inspection. A radioactive sludge transfer apparatus for transferring sludge stored in a sludge storage tank with supernatant solution to a transfer tank includes: a transfer apparatus body 1 ; a stirring apparatus 2 for blasting the supernatant solution to the sludge to thereby produce sludge solution in the one tank; sludge solution transfer means 8 for transferring the sludge solution to the transfer tank; an attitude control float 15 ; a floating force control ballast tank 16 ; and control means for remotely controlling the stirring apparatus 2 , the sludge solution transfer means 8 and the floating force control ballast tank 16 , wherein the stirring apparatus 2 , the sludge solution transfer means 8 , the attitude control float 15 , and the floating force control ballast tank 16 are mounted respectively to the transfer apparatus body, the stirring apparatus 2 is provided with a supernatant solution suction pump 3 for sucking the supernatant solution and an injection nozzle 4 for jetting the supernatant solution to the sludge in the one tank to thereby form the sludge solution, the injection nozzle having freely controllable nozzle angle, the sludge solution transfer means 8 is provided with a sludge solution suction pump 9 for sucking the sludge solution and transfer the sucked sludge solution to the transfer tank, and the sludge solution suction pump 9 has suction ports 9 A including bottom suction ports 10 and side suction ports 11 which are opened or closed by the open/close means 13 , and when the sludge solution remaining inside the sludge storage tank is sucked out, the side suction ports 11 are closed by the open/close means 13.

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Номер записи: 57
05-04-2018 дата публикации

PLASMA SURFACE DECONTAMINATION: METHOD AND APPARATUS FOR REDUCING RADIOACTIVE NUCLEAR WASTE AND TOXIC WASTE VOLUME

Номер: US20180096745A1
Автор: Park JaeYoung
Принадлежит:

Plasma reactors and methods are provided for reducing the volume of radioactive nuclear wastes or toxic wastes where the surface contamination is the main source of radioactivity or toxicity. The radioactive or toxic wastes are prepared in the form of small particles and fed into a pulsed plasma reactor operating in fluidized bed configuration. The repetitively pulsed radio-frequency (rf) powered plasma reactor generates high power plasma for pulse duration between 10 μ-10 ms. During the pulse, the plasma deliver a short burst of intense heat flux to the surface of waste feed particles. Due to the short pulse duration, the heat flux is concentrated on the surface without propagating much to the core of the particles. The localized heat flux preferentially removes the surface contaminants via vaporization. The removed waste in the vapor phase will be transferred out of the reactor where it may undergo additional treatments or disposed accordingly in a reduced volume. The residual particles which are free from surface contaminants can then be recycled or disposed as non-toxic or non-radioactive waste. By controlling the treatment time inside the plasma reactor along with the pulse power, duration and repetition rate, and the plasma chemical composition, the thickness of removed surface layers can be controlled to provide the efficient surface decontamination. 1. An apparatus , comprising:a pulsed radio frequency source that is driven with a pulse duration and generates radio frequency signal;a decontamination reactor where the pulsed plasma generated from the pulsed radio frequency source removes surface contaminants of radioactive and toxic materials from the waste feed particles and leaves residual particles, the pulsed plasma delivering intense pulsed heat flux to a surface of waste feed particles to vaporize the surface radioactive and toxic materials into volatile gas species;an inductive antenna, coupled to the pulsed radio frequency source, that surrounds a ...

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Номер записи: 58
12-05-2022 дата публикации

METHOD FOR DECONTAMINATING TRITIUM RADIOACTIVE CONTAMINATED WATER

Номер: US20220148750A1
Автор: TAKATSUKA Hikaru
Принадлежит: KABUSHIKIKAISHA GABRIEL

A decontamination method in which tritium radiation is attenuated or eliminated from radioactive contaminated water. In the decontamination method, the following steps are executed: a step for performing an addition treatment to add a prescribed amount of a mineral powder such as a silicon dioxide ore and a nano-level carbon liquid to heated tritium-contaminated water in an addition treatment tank; a step for pumping the addition-treated water from the addition treatment tank to a mineral solid filled tank using a hydraulic pump; a step for causing the addition-treated water to collide with a mineral solid and passing the addition-treated water through the mineral solid filled tank; a step for returning the passed water to the addition treatment tank using the hydraulic pump; and a circulation process step for repeating the aforementioned steps in a prescribed period of time. 1. A tritium radioactive contaminated water decontamination method of reducing or removing tritium radiation from radioactive contaminated water , the method comprising the steps of:a first step of performing an addition treatment in an addition treatment tank, the addition treatment including adding 0.5-6 parts by weight of a mineral powder obtained by pulverizing a mineral comprising one or more selected from silicon dioxide ore, old shellfish fossil and radium ore, and 0.5-6 parts by weight of a nano-level carbon liquid to 100 parts by weight of tritium radioactive contaminated water heated to 30-80° C.;a second step of pumping the addition treated water from the addition treatment tank to a mineral solid filled tank by a hydraulic pump of 1 to 7 atmosphere pressure, the mineral solid filled tank being filled with a mineral solid in which the mineral is crushed to a predetermined size;a third step of passing the pumped addition treated water through the mineral solid filled tank, while the pumped addition treated water is colliding with the mineral solid;a fourth step of returning the ...

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Номер записи: 59
04-04-2019 дата публикации

REPOSITORY FOR STORING HAZARDOUS MATERIAL IN A SUBTERRANEAN FORMATION

Номер: US20190099790A1
Автор: Muller Elizabeth, Muller Richard A.
Принадлежит:

A hazardous material storage repository includes a drillhole extending into the Earth and including an entry at least proximate a terranean surface, the drillhole including a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion, at least one of the transition drillhole portion or the hazardous material storage drillhole portion including an isolation drillhole portion; a storage canister positioned in the hazardous material storage drillhole portion, the storage canister sized to fit from the drillhole entry through the substantially vertical drillhole portion, the transition drillhole portion, and into the hazardous material storage drillhole portion of the drillhole, the storage canister including an inner cavity sized enclose hazardous material; and a seal positioned in the drillhole, the seal isolating the hazardous material storage drillhole portion of the drillhole from the entry of the drillhole. 1. A hazardous material storage repository , comprising:a drillhole extending into the Earth and comprising an entry at least proximate a terranean surface, the drillhole comprising a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, at least one of the transition drillhole portion or the hazardous material storage drillhole portion comprising an isolation drillhole portion that is directed vertically toward the terranean surface and away from an intersection between the substantially vertical drillhole portion and the transition drillhole portion;a storage canister positioned in the hazardous material storage drillhole portion, the storage canister sized to fit from the drillhole entry through the substantially vertical drillhole portion, the transition drillhole ...

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Номер записи: 60
12-04-2018 дата публикации

DECONTAMINATION METHOD REDUCING RADIOACTIVE WASTE REMARKABLY AND A KIT THEREFOR

Номер: US20180102194A1
Автор: Choi Mansoo, CHOI Wangkyu, JUNG Chong-Hun, JUNG Jun Young, Kim Seonbyeong, MOON Jeikwon, Park Jungsoon, PARK Sang Yoon, WON Hui-Jun, YOON In-Ho
Принадлежит: KOREA ATOMIC ENERGY RESEARCH INSTITUTE

The present invention provides a decontamination method including the steps of decontaminating an object containing radioactive contaminated metals or alloys with a chemical decontamination agent comprising sulfuric acid (HSO) and forming a Ba or Sr precipitate by adding Ba or Sr cation and hydroxyl ion or halogen anion salts to the decontamination waste water. 1. A decontamination method comprising the following steps:{'sub': 2', '4, 'decontaminating an object containing radioactive contaminated metals or alloys with a chemical decontamination agent comprising sulfuric acid (HSO) (step 1); and'}forming a Ba or Sr precipitate by adding Ba or Sr cation and hydroxyl ion or halogen anion salts to the decontamination waste water generated in step 1 (step 2).2. The decontamination method according to claim 1 , wherein the chemical decontamination agent of step is an oxidative decontamination agent claim 1 , a reductive decontamination agent claim 1 , or a combined decontamination agent thereof.3. The decontamination method according to claim 1 , wherein the decontamination method additionally includes a step of separating the precipitate (step 3).4. The decontamination method according to claim 2 , wherein the oxidative decontamination agent includes an oxidizing agent and a metal ion.5. The decontamination method according to claim 4 , wherein the oxidizing agent is one or more agents selected from the group consisting of KMnO claim 4 , NaMnO claim 4 , HCrO claim 4 , and HMnO.6. The decontamination method according to claim 2 , wherein the reductive decontamination agent includes a reducing agent and a metal ion.7. The decontamination method according to claim 6 , wherein the reducing agent is one or more agents selected from the group consisting of NaBH claim 6 , HS claim 6 , NH claim 6 , and LiAlH.8. The decontamination method according to claim 1 , wherein the object containing radioactive contaminated metals or alloys in step 1 is generated from the inner system of ...

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Номер записи: 61
02-06-2022 дата публикации

RECEPTOR AND METHOD FOR REMOVING OXOANIONS FROM AQUEOUS PHASE

Номер: US20220169536A1
Автор: Jagleniec Damian, Karbarz Marcin, Romanski Jan
Принадлежит:

A receptor for the simultaneous removal of oxoanions and their counterions from aqueous phase, particularly containing radioactive wastes, containing amide groups specifically coordinating the oxoanions, as well as moieties specifically coordinating cations, according to the present invention is characterised in that it contains within one molecule domains binding oxoanions and domains binding cations, preferably adapting a molecular structure of a general formula: (I) wherein Z this is a group containing crown ether, preferably a benzocrown group, X is any substituent, including the Y-Z grouping, and Y is any substituent or 0 (i.e. a direct bond between N and Z), where the oxoanion binding domain is a squaramide unit coordinating the oxoanions through amide groups, and squaramide contains additional substituents that increase or decrease the acidity of its amide protons, compared to unsubstituted squaramide, whereas the counter ion binding domain is a crown ether of a size adjusted to the type of binding cation, which forms part of at least one of the aforementioned substituents of squaramide, where the receptor has the ability to remove oxoanions and their counterions from aqueous phase to another water-immiscible phase, preferably to organic phase, and has the ability to form soluble complexes in at least one of the aforementioned phases. The invention considers also a method of removing oxoanions in the form of inorganic salts from aqueous phase, using receptors of the invention in the form of organic molecules containing amide groups, according to the invention is characterised in that it uses the aforementioned receptors for simultaneous binding of oxoanions and their counterions in aqueous phase, preferably acidic when using the receptor with substituents increasing acidity of squaramide protons, or alkaline when using the receptor with substituents decreasing acidity of squaramide protons. A sensor for detecting oxoanions according to the invention is ...

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Номер записи: 62
23-04-2015 дата публикации

METHOD FOR THE RADIOACTIVE DECONTAMINATION OF SOIL BY DISPERSED AIR FLOTATION FOAM AND SAID FOAM

Номер: US20150110560A1
Автор: Faure Sylvain, Messalier Marc
Принадлежит:

The present invention relates to a process for treating an earth contaminated by at least one radionuclide such as cesium Cs comprising at least one step of separating said radionuclide by dispersed air flotation foam produced by blowing air bubbles in a suspension comprising said earth and at least one collector. The present invention also relates to the flotation foam obtained by implementing such a process. 1. A process for treating a contaminated earth by at least one radionuclide comprising:at least one step of separating said radionuclide by dispersed air flotation foam produced by blowing air bubbles in a suspension comprising said earth and at least one collector.2. The process according to claim 1 , wherein said radionuclide is selected from the group consisting of tritium H; carbon C; strontium Sr; yttrium Y; cesium Cs; barium Ba; americium Am; a radionuclide from the thorium 232 family; and a radionuclide from the uranium 238 family.3. The process according to claim 1 , wherein said radionuclide is cesium Cs.4. The process according to claim 1 , wherein said process is implemented in a flotation column or a flotation cell.5. The process according to claim 1 , further comprising the steps of:a) injecting air bubbles into a suspension comprising said earth and said at least one collector, to produce a foam; andb) separating at least one part of said foam from the rest of the suspension.6. The process according to claim 5 , wherein said suspension comprises:2-40% by weight of earth to be treated based on a total weight of the suspension;0.005-5% by weight of at least one collector based on the total weight of the suspension; andwater.7. The process according to claim 5 , wherein said collector is selected from a fatty acid claim 5 , a fatty acid salt and a cationic surfactant.8. The process according to claim 5 , wherein said collector is sodium oleate.9. The process according claim 5 , wherein said collector is tetradecyltrimethylammomum bromide (TTAB).10. ...

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Номер записи: 63
20-04-2017 дата публикации

Apparatus for Extracting Radioactive Solid Particles and Method Thereof

Номер: US20170106318A1
Автор: Chang Bing-Hong, Huang Chun-Ping, Wu Jia-Ying
Принадлежит:

An apparatus is provided to extract radioactive solid particles. An extracting nozzle is used to extract radioactive solid particles. Then, a separator is used to separate out the radioactive solid particles into a storing container. The radioactive solid particles are avoided from entering a suction pump. Not only the suction pump is not polluted, but also the secondary waste is not increased. By designing a falling inlet of a suction channel at a position having a specific height, the amount of the radioactive solid particles being extracted is under control. There is a radiation-protection device outside of the storing container to minimize radiation dose. The separator and the storing container can be rapidly detached by remote operation, so that operators are avoided from receiving over-dose radiation. Hence, the present invention improves the level of technology and automation for handling radioactive waste. 1. An apparatus for extracting radioactive solid particles , comprising 'wherein said suction pump has a suction pipe to extract gas;', 'a suction pump,'} 'wherein said separator comprises a body and a chamber; said chamber is surrounded by said body; said body has a suction channel and a gas outlet; said suction channel is located in said chamber and connected with said body; said gas outlet is located on top of said body and adjacent to said suction channel; said gas outlet is connected with said suction pipe of said suction pump; said suction channel has a suction inlet located at an upper section of said suction channel to be protruded out from top of said body; and said suction channel has a falling inlet located at a lower section of said suction channel to be protruded out from bottom of said body;', 'a separator,'} 'wherein said extracting nozzle has an extracting pipe; and said extracting pipe is connected to said suction inlet of the suction channel to extract radioactive solid particles to said separator;', 'an extracting nozzle,'} 'wherein said ...

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Номер записи: 64
28-04-2016 дата публикации

Capture, removal, and storage of radioactive species

Номер: US20160118153A1
Автор: Keith Paul Fruzzetti, Leon Yengoyan, Lisa Maxine Edwards, Paul L. Frattini, Roger Howard Asay, Susan Elaine Garcia
Принадлежит: Electric Power Research Institute Inc

A method of capturing radioactive species from an aqueous solution and removing the radioactive species for disposal, includes: contacting the aqueous solution with a first sequestration resin comprising a sequestration ligand coupled to a sulfonic acid based polymer resin backbone, to allow the first sequestration resin to capture the radioactive species; removing the first sequestration resin with the captured radioactive species from the aqueous solution; and using an acid to lower a pH of the first sequestration resin to release the radioactive species from the first sequestration resin.

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Номер записи: 65
27-04-2017 дата публикации

METHODS AND SYSTEMS TO CONTAIN POLLUTION & HAZARDOUS ENVIRONMENTS (CPHE)

Номер: US20170114513A1
Автор: Dunlap Nathan M., Dunlap Philip S., Muse-Fisher Michael J., Trutanich Nicholas A.
Принадлежит:

A membrane system isolates environment volume, contains hazards, enables detoxification, hazard removal, infrastructure restoration, and substance recovery. The flexible membrane system adjusts to advantageously shape its isolated volume and integrate existing infrastructure. Pollution and hazardous energy, including crude oil, toxic-gas, radioactive fallout, fire, and other hazards are isolated, while concurrently enabling access within the isolated volume. Pod encapsulated and readily deployed, the membrane system operates semi-autonomously, uses selective-filtering, specific gravity and substance differences to channel matter, mitigate hazards, protect the biosphere, preserve infrastructure and capture substances. 1. A membrane system that inhibits the natural spread of the pollution and hazardous energy , by isolating a volume containing the pollution and hazardous environment. The method comprising employment of a membrane-system within the environment.2. The method of wherein the method is comprised of one or more missions to enhance safety of the environment claim 1 , to protect man-made infrastructures claim 1 , and to collect the pollution and preserve its value as a natural resource.3. The method of and wherein the membrane system is comprised of one or more sections or methods that isolate claim 1 , direct claim 1 , channel claim 1 , collect claim 1 , and consolidate pollution and constituent matter to mitigate risks claim 1 , preserve infrastructure and enhance substance value.4. The method of claim 1 , and wherein the membrane-system isolates claim 1 , directs claim 1 , channels claim 1 , collects claim 1 , and consolidates crude oil pollution resulting from oil drilling claim 1 , fissures claim 1 , and subsurface infrastructure leaks in the ocean.5. The method of claim 1 , and wherein the membrane-system isolates claim 1 , directs claim 1 , channels claim 1 , collects claim 1 , and consolidates radioactive particulates resulting from nuclear power ...

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Номер записи: 66
09-06-2022 дата публикации

APPARATUS AND METHOD

Номер: US20220181040A1
Автор: Crabbe Ian
Принадлежит:

An apparatus () is described for removing radioactive contamination, at least in part, from a first article (A) comprising a metal, preferably wherein the metal comprises and/or is a low melting point metal for example lead and/or an alloy thereof. The apparatus () comprises a heated first vessel (A) for melting the metal, at least in part, therein, thereby providing a melt (M) therefrom. The apparatus () comprises casting means () for forming a second article (A), particularly a sheet, having a predetermined thickness (T), from the melt, preferably wherein the casting means () comprises and/or is a rotatable roller () arrangeable to contact the melt (M) to thereby form thereon the second article (A) and a guide () arranged to remove the second article (A) from the roller (). The apparatus () comprises a set of radiation detectors (), including a first radiation detector (A), arranged to detect a first fraction of the radioactive contamination, if present, in a first part (P) of a set of parts of the second article (A), preferably wherein the set of radiation detectors () comprises opposed first and second radiation detectors (A, B) arranged to receive the second article (A) traversing therebetween. The apparatus () comprises a cutter () arrangeable to excise the first part (P) of the second article (A) therefrom. 1. An apparatus for removing radioactive contamination , at least in part , from a first article comprising a metal , preferably wherein the metal comprises and/or is a low melting point metal for example lead and/or an alloy thereof , the apparatus comprising:a heated first vessel for melting the metal, at least in part, therein, thereby providing a melt therefrom;casting means for forming a second article, for example a sheet, a strip or a ribbon, having a predetermined thickness, from the melt, preferably wherein the casting means comprises and/or is a rotatable roller arrangeable to contact the melt to thereby form thereon the second article and a ...

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Номер записи: 67
04-05-2017 дата публикации

METHOD FOR LINING EXISTING ASH BASIN AND LANDFILL SITES

Номер: US20170120314A1
Автор: Zapata Manuel
Принадлежит: Zapata Incorporated

A method for lining an existing waste disposal site having a waste level includes, for example: installing perimeter barrier walls about the site and interior barrier walls within the perimeter barrier walls to define cells; transferring waste, such as for example ash, from one of the cells to one or more other cells to form an empty cell with a layer of contaminated material at the bottom of the empty cell; removing the layer of contaminated material from the empty cell to form a clean cell with a noncontaminated bottom layer; installing a barrier liner layer in the clean cell to form a lined cell; and transferring waste from other cells into the lined cell. 1. A method for lining an existing waste disposal site having a waste level , comprising:installing perimeter barrier walls about the site and interior barrier walls within the perimeter barrier walls to define cells;transferring waste from one of the cells to one or more other cells to form an empty cell with a layer of contaminated material at the bottom of the empty cell;removing the layer of contaminated material from the empty cell to form a clean cell with a noncontaminated bottom layer;installing a barrier liner layer in the clean cell to form a lined cell; andtransferring waste from other cells into the lined cell.2. The method for lining an existing waste disposal site according to claim 1 , further comprising:removing standing water from one of the cells prior to transferring waste from one of the cells to one or more other cells.3. The method for lining an existing waste disposal site according to claim 1 , further comprising:installing a barrier material layer on the noncontaminated bottom layer in the clean cell, wherein the barrier liner layer is installed on the barrier material layer.4. The method for lining an existing waste disposal site according to claim 2 , wherein the barrier material layer is clay.5. The method for lining an existing waste disposal site according to claim 1 , further ...

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Номер записи: 68
24-07-2014 дата публикации

Chelate Free Chemical Decontamination Reagent for Removal of the Dense Radioactive Oxide Layer on the Metal Surface and Chemical Decontamination Method Using the Same

Номер: US20140205052A1
Автор: CHOI Byung-Seon, CHOI Wangkyu, JUNG Chong-Hun, MOON Jeikwon, PARK Jung-Sun, PARK Sang Yoon, WON Hui-Jun, YOON In-Ho
Принадлежит:

A chemical decontamination reagent containing a reducing agent, a reductive metal ion, and an inorganic acid is provided to remove a radioactive oxide layer on a metal surface. The reagent can dissolve the radioactive oxide layer on the metal surface effectively at a relatively low temperature and enables a simple process of contacting the reagent to the radioactive oxide, thus economically effective in terms of cost and time required for the process. Since the decontamination does not use a conventional organic chelating agent such as oxalic acid, but the reducing agent as a main substance, the residuals of the reducing agent remained after decontamination can be decomposed and removed with an oxidizing agent. Due to the easy decomposition with the chemical decontamination reagent, secondary wastes can be minimized and the radionuclides remained in the decontamination reagent solution can be removed effectively. 1. A chelate-free chemical decontamination reagent comprising a reducing agent , a reductive metal ion , and an inorganic acid , for removal of a dense radioactive oxide layer on a metal surface.2. The chemical decontamination reagent according to claim 1 , wherein the reducing agent is one or more selected from the group consisting of NaBH claim 1 , HS claim 1 , NH claim 1 , and LiAlH.3. The chemical decontamination reagent according to claim 1 , wherein the reductive metal ion is one or more selected from the group consisting of Ag claim 1 , Ag claim 1 , Mn claim 1 , Mn claim 1 , Co claim 1 , Co claim 1 , Cr claim 1 , Cr claim 1 , Cu claim 1 , Cu claim 1 , Sn claim 1 , Sn claim 1 , Ti claim 1 , and Ti.4. The chemical decontamination reagent according to claim 1 , wherein the inorganic acid is one or more selected from the group consisting of HBr claim 1 , HF claim 1 , HI claim 1 , HNO claim 1 , HPO claim 1 , and HSO.5. The chemical decontamination reagent according to claim 1 , wherein the radioactive oxide layer on the metal surface is generated from ...

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Номер записи: 69
27-05-2021 дата публикации

PLANT FOR ELECTROCHEMICAL DECONTAMINATION OF METAL RADIOACTIVE WASTE

Номер: US20210158985A1
Автор: NEUPOKOEV Mikhail Alekseevich, SHAROV Aleksandr Nikitovich, SHEVCHENKO Boris Nikolaevich
Принадлежит:

Devices for eliminating radioactive contamination of radioactive waste by providing adaptive processing of the decontamination solution for reuse. The plant for electrochemical decontamination of metal radioactive waste includes a pipe equipped with shut-off valves, a radioactive waste processing module that comprises a unit for electrochemical decontamination connected by a ventilation channel to the ventilation module and pipe for decontamination solution supply and discharge equipped with shut-off valves. The plant is equipped with a decontamination solution preparation module connected with a pipe for decontamination solution supply and discharge, at least one pump, while the module for decontamination solution receiving is equipped with devices for cleaning and pH correction of decontamination solution, and the unit for electrochemical decontamination of metal radioactive waste, the module for decontamination solution receiving and the decontamination solution preparation module are equipped with pH measurement elements. 1. A plant for electrochemical decontamination of metal radioactive waste includes a pipe equipped with shut-off valves , a radioactive waste processing module that comprises a unit for electrochemical decontamination of metal radioactive waste connected by a ventilation channel to the ventilation module and equipped with shut-off valves and pipe for decontamination solution supply and discharge , with the module for decontamination solution receiving. The plant is equipped with a decontamination solution preparation module connected with a pipe for decontamination solution supply and discharge , equipped with at least one pump , with a unit for electrochemical decontamination of metal radioactive waste and a module for decontamination solution receiving , whilst the module for decontamination solution receiving is equipped with devices for cleaning and pH correction of decontamination solution , and the unit for electrochemical decontamination ...

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Номер записи: 70
27-05-2021 дата публикации

Continuous Separation of Radionuclides by Shock Electrodialysis

Номер: US20210158986A1
Автор: Alkhadra Mohammad Ayman, Bazant Martin Z., Conforti Kameron, Gao Tao, Tian Huanhuan
Принадлежит: Massachusetts Institute of Technology

Radioactive nuclides (radionuclides) are separate from an aqueous radioactive liquid by feeding the liquid into a chamber between a porous anode and a porous cathode of a shock electrodialysis device. Meanwhile, an anolyte is fed through the porous anode, and a catholyte is fed through the porous cathode. A voltage is applied to the porous anode and to the porous cathode to create a voltage differential across the chamber. The liquid is passed through the chamber, and cations are selectively driven from the liquid into the cathode by the voltage differential. The voltage differential creates a desalination shock that produces an ion-enriched zone on one side of the desalination shock and a deionized zone on an opposite side. A brine including the radioactive cations is extracted from the ion-enriched zone through a brine outlet, and fresh water is extracted from the deionized zone through a fresh-water outlet. 1. A system for separating radioactive nuclides , the system comprising:a source of an aqueous radioactive liquid including radioactive nuclides;a feed conduit for liquid flow from the source of aqueous radioactive liquid; [ [ the aqueous radioactive liquid, wherein the inlet for the radioactive liquid is in fluid communication with the feed conduit from the source;', 'an anolyte; and', 'a catholyte;, 'respective inlets for, fresh water;', 'a brine that includes the radioactive nuclides;', 'the anolyte; and', 'the catholyte;, 'respective outlets for], 'a chamber including, 'a porous anode contained in the chamber and configured for flow of the anolyte therethrough;', 'a porous cathode contained in the chamber and configured for flow of the catholyte therethrough;', an ion-selective boundary;', 'the anode being configured for ion separation; and', 'the cathode being configured for ion separation, wherein the ion separators are configured to selectively pass at least some cations, wherein a channel for flow of the aqueous radioactive liquid from the feed conduit ...

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Номер записи: 71
19-05-2016 дата публикации

ABYSSAL SEQUESTRATION OF NUCLEAR WASTE AND OTHER TYPES OF HAZARDOUS WASTE

Номер: US20160136701A1
Автор: Germanovich Leonid, Murdoch Lawrence C., Robinowitz Marvin
Принадлежит:

A system and method of disposing nuclear waste and other hazardous waste includes means for, and the steps of, blending a waste stream, which includes either a radioactive waste or a hazardous waste (or both), with a liquid and, optionally, a solid material to produce a dense fluid and pumping the dense fluid into a tubing string of an injection boring. The dense fluid then exits a perforation in a casing of the injection boring and enters a fracture in a rock strata, where it continues to propagate downward until it reaches an immobilization point. The dense fluid may be a slurry formed by a metal and a cross-linked polymer gel or hydrated clay slurry. The metal can be one that has a melting temperature less than the temperature at the bottom of the injection boring. The solid material could also be other nuclear waste or a radionuclide. 1. A system for abyssal sequestration of nuclear waste and other types of hazardous waste , the system comprising:a gravity fracture filled with a fluid having at least one waste selected from the group consisting of a radioactive waste and a hazardous waste, the fluid being denser than a rock formation into which the fluid is to be disposed so as to cause the rock formation to gravity fracture, the fluid propagating downward in the gravity fracture as the gravity fracture propagates downward.2. A system according to wherein the fluid has a density of at least 3.0 g/cm.3. A system according to wherein the fluid is a slurry.4. A system according to further comprising the slurry including a solid material which is blended with the at least one waste.5. A system according to wherein the solid material is a metal.6. A system according to wherein the metal is selected from the group consisting of bismuth claim 5 , iron claim 5 , lead claim 5 , and copper.7. A system according to wherein the solid material contains one or more radionuclides.8. A system according to wherein a liquid component of the slurry is a metal having a melting ...

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Номер записи: 72
19-05-2016 дата публикации

APPARATUS AND METHOD FOR REMOVAL OF NUCLIDES FROM HIGH LEVEL LIQUID WASTES

Номер: US20160141058A1
Автор: Barker Tracy A., Benda Gary A., Braun James L.
Принадлежит:

A method for treating a liquid waste is provided. The method includes supplying the liquid waste to a plurality of cross flow filters from at least one high level waste tank; filtering the liquid waste via the plurality of cross flow filters to form a clarified salt solution; removing at least one radionuclide from the clarified salt solution via a plurality of elutable ion exchange columns filled with an ion exchange media to form an eluate and a decontaminated salt solution; and removing at least one radionuclide from the eluate via a first non-elutable adsorption component to form a dewatered radionuclide sorbent and a decontaminated eluate solution. 1. A method for treating a liquid waste having at least one radionuclide in a salt solution , comprising:supplying the liquid waste to a plurality of cross flow filters from at least one high level waste tank;filtering the liquid waste via the plurality of cross flow filters to form a clarified salt solution;removing at least one radionuclide from the clarified salt solution via a plurality of elutable ion exchange columns filled with an ion exchange media to form an eluate and a decontaminated salt solution; andremoving at least one radionuclide from the eluate via a first non-elutable adsorption component to form a dewatered radionuclide sorbent and a decontaminated eluate solution.2. The method of claim 1 , further comprising disposing the dewatered radionuclide sorbent.3. The method of claim 1 , wherein the plurality of elutable ion exchange columns comprise a lead ion exchange column claim 1 , a lag ion exchange column claim 1 , and a polishing ion exchange column.4. The method of claim 3 , further comprising eluting the lead ion exchange column claim 3 , wherein eluting the lead ion exchange column comprises displacing the lead ion exchange column claim 3 , rinsing the lead ion exchange column with water claim 3 , neutralizing the lead ion exchange column claim 3 , eluting the lead ion exchange column with an ...

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Номер записи: 73
19-05-2016 дата публикации

METHOD AND PLANT FOR THE WET-ROUTE OXIDATION TREATMENT OF HAZARDOUS ORGANIC WASTE, NOTABLY RADIOACTIVE WASTE, CONTAINING MINERAL FILLERS

Номер: US20160141059A1
Автор: CORREA Boris, Mercier Olivier
Принадлежит:

A method and plant for wet-route oxidation treatment of hazardous organic waste products, notably radioactive wastes, which may contain mineral fillers, the waste products being treated in a secure environment. The plant comprises a closed space, with a mechanism for bringing a volume of hazardous organic waste products containing mineral fillers, adding a given quantity of water mixed with a base to the predetermined volume in order to adjust the pH to a determined value so as to make a solution and/or a liquid suspension, with a pressure reactor and with mechanism for transferring the solution and/or liquid suspension into the pressure reactor, and a device for introducing an oxygen atmosphere into the pressure reactor and for pressurizing the atmosphere. A heating mechanism is provided for subjecting the contents of the pressure reactor to heat treatment at a temperature between 150 and 350° C. to complete the wet-route oxidation. 118-. (canceled)19. A method of wet-route oxidation treatment of hazardous organic waste products , notably radioactive wastes , which may contain mineral fillers , the waste products being treated in a secure environment , the method comprising:conveying a certain predetermined volume of the hazardous organic waste products, containing mineral fillers, in a confined manner,adding a given quantity of water mixed with a base, to the predetermined volume, in order to adjust a pH to a determined value and make up a solution and/or a liquid suspension which is transferred into a pressure reactor,pressurizing a content of the reactor, at ambient temperature, with an oxygen atmosphere under a maximum pressure of 60 bars while stirring the solution and/or liquid suspension confined within the reactor,heating the contents of the reactor to a temperature substantially between 150 and 350° C., and preferably temperatures close to 300° C., for a determined time in order to complete the wet-route oxidation, and inducing, via the temperature ...

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Номер записи: 74
07-08-2014 дата публикации

ABYSSAL SEQUESTRATION OF NUCLEAR WASTE AND OTHER TYPES OF HAZARDOUS WASTE

Номер: US20140221722A1
Автор: Germanovich Leonid, Murdoch Lawrence C., Robinowitz Marvin
Принадлежит:

A system and method of disposing nuclear waste and other hazardous waste includes means for, and the steps of, blending a waste stream, which includes either a radioactive waste or a hazardous waste (or both), with a liquid and, optionally, a solid material to produce a dense fluid and pumping the dense fluid into a tubing string of an injection boring. The dense fluid then exits a perforation in a casing of the injection boring and enters a fracture in a rock strata, where it continues to propagate downward until it reaches an immobilization point. The dense fluid may be a slurry formed by a metal and a cross-linked polymer gel or hydrated clay slurry. The metal can be one that has a melting temperature less than the temperature at the bottom of the injection boring. The solid material could also be other nuclear waste or a radionuclide. 2. A method according to wherein the dense fluid after entering the fracture continues a downward travel as the dense fluid drains from the injection boring.3. A method according to wherein the dense fluid after entering the fracture continues a downward travel after becoming detached from any dense fluid remaining in the injection boring.4. A method according to wherein the dense fluid after entering the fracture continues a downward travel and a portion of the downwardly travelling dense fluid remains connected by a thin film to any dense fluid remaining in the injection boring.5. A method according to wherein the dense fluid claim 1 , when in a detached state claim 1 , reaches an immobilization point below the initial entry point of the dense fluid into the rock strata.6. A method according to wherein the immobilization point occurs at a depth in a range of about 2 claim 5 ,000 to 50 claim 5 ,000 feet (about 600 to 15 claim 5 ,000 meters).7. A method according to wherein the dense fluid propagates downward and then curves in a horizontal direction creating a sub-horizontal storage space.8. A method according to further ...

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Номер записи: 75
09-05-2019 дата публикации

COMPOSITION INCLUDING SILICOTITANATE HAVING SITINAKITE STRUCTURE, AND PRODUCTION METHOD FOR SAME

Номер: US20190134599A1
Автор: FUNAKOSHI Hajime, Hirano Shigeru, IZUMI Takeshi, Komatsu Makoto, Sakuma Takashi, SHIMIZU Yoju, TOKUNAGA Keisuke
Принадлежит:

The present invention provides a composition that includes a silicotitanate that has a sitinakite structure, the composition having higher cesium adsorptivity than conventional compositions. The present invention also provides a production method for the composition that includes a silicotitanate that has a sitinakite structure. The production method does not require the use of hazardous or deleterious materials, can generate a product using a compound that is easily acquired, and can use a general-purpose autoclave. Also provided is a silicotitanate composition that has higher strontium adsorptivity than the present invention. Provided is a silicotitanate composition that contains niobium and a silicotitanate that has a sitinakite structure, the composition having at least two or more diffraction peaks selected from the group consisting of 2θ=8.8°±0.5°, 2θ=10.0°±0.5°, and 2θ=29.6°±0.5°. 1. An adsorption method for at least any of cesium and strontium , comprising contacting a medium containing at least any of cesium and strontium with a silicotitanate composition ,wherein the composition comprises a silicotitanate having a sitinakite structure and niobium, and has at least two or more diffraction peaks at X-ray diffraction angles selected from the group consisting of 2θ=8.8±0.5°, 2θ=10.0±0.5°, and 2θ=29.6±0.5°.3. The adsorption method according to claim 1 , wherein the composition comprises a crystalline substance having at least two or more diffraction peaks at X-ray diffraction angles selected from the group consisting of 2θ=8.8±0.5° claim 1 , 2θ=10.0±0.5° and 2θ=29.6±0.5°.4. The adsorption method according to claim 2 , wherein the composition comprises a crystalline substance having at least two or more diffraction peaks at X-ray diffraction angles selected from the group consisting of 2θ=8.8±0.5° claim 2 , 2θ=10.0±0.5° and 2θ=29.6±0.5°.5. The adsorption method according to claim 3 , wherein the crystalline substance is a niobate.6. The adsorption method ...

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Номер записи: 76
17-05-2018 дата публикации

THERMAL VOLUME REDUCTION OF RADIOACTIVE WASTES

Номер: US20180137946A1
Автор: Liu Lu, MILNER Timothy, Poole Scott D.
Принадлежит:

A method for thermal volume reduction of waste material contaminated with radionuclides includes feeding the waste material into a fluidized bed reactor, injecting fluidizing gas into the fluidized bed reactor to fluidize bed media in the fluidized bed reactor, and decomposing the waste material in the fluidized bed reactor. A system for thermal volume reduction of the waste material includes one or more of a feedstock preparation and handling system, a fluidized bed reactor system, a solids separation system, and an off-gas treatment system. The method and system may be used to effectively reduce the volume or radioactive wastes generated from the operation of nuclear facilities such as nuclear power plants including wastes such as spent ion exchange resin, spent granular activated carbon, and dry active waste. The majority of the organic content in the waste material is converted into carbon dioxide and steam and the solids, including the radionuclides, are converted into a waterless stable final product that is suitable for disposal or long-term storage. 1. A method of decomposing dewatered waste material contaminated with radionuclides in a fluidized bed reactor , the method comprising:feeding the dewatered waste material into the fluidized bed reactor;injecting fluidizing gas into the fluidized bed reactor to fluidize bed media and form a fluidized bed in the fluidized bed reactor, the fluidizing gas comprising superheated steam; anddecomposing the dewatered waste material in the fluidized bed reactor;wherein the dewatered waste material is not fed into the fluidized bed reactor in a slurry.2. The method of wherein the dewatered waste material comprises spent ion exchange resin and/or spent granular activated carbon.3. The method of wherein the spent ion exchange resin and/or the spent granular activated carbon have a water content of no more than 70 wt %.4. The method of comprising feeding the dewatered waste material into the fluidized bed reactor using a ...

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Номер записи: 77
09-05-2019 дата публикации

Emergency Method And System For In-Situ Disposal And Containment Of Nuclear Material At Nuclear Power Facility

Номер: US20190139658A1
Автор: Germanovich Leonid N., Murdoch Lawrence C., Robinowitz Marvin
Принадлежит:

A system and method to safely isolate mobile radioactive material during an emergency includes a borehole located in close proximity and at a depth sufficient to safely isolate the material and a man-made vertical-oriented gravity fracture located at the bottom end of the borehole. During an emergency, the mobile radioactive material enters the borehole and then passes from there into the gravity fracture. The mobile radioactive material may have sufficient density to further propagate the fracture vertically downward or a dense slurry or fluid could be mixed with the mobile radioactive material. 2. A method according to claim 1 , further comprising the conveying step to include injecting the mobile radioactive material into the borehole.3. A method according to claim 1 , wherein prior to the emergency claim 1 , the man-made vertical-oriented gravity fracture is made using a slurry containing a weighting material claim 1 , the slurry being denser than the surrounding rock formation claim 1 , the slurry not including the mobile radioactive material.4. A method according to claim 3 , wherein the slurry has an absolute tendency to travel vertically downward in the surrounding rock formation.5. A method according to claim 1 , further comprising conveying additional mobile material into the borehole after conveying the mobile radioactive material into the borehole.6. A method according to claim 1 , further comprising mixing at least a portion of the mobile radioactive material with a weighting material to produce a fluid or a slurry sufficiently dense to cause additional vertical downward propagation of the man-made vertical-oriented gravity fracture.7. A method according to claim 1 , further comprising controlling a cooling rate of the mobile radioactive material as the mobile radioactive material travels past at least a portion of the borehole.8. A method according to claim 1 , wherein the mobile radioactive material includes at least one of a molten material claim 1 , ...

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Номер записи: 78
30-04-2020 дата публикации

Water Vapor Quantification Methodology During Drying of Spent Nuclear Fuel

Номер: US20200135350A1
Автор: Farouk Tanvir, Knight Travis W., Tahiyat Malik
Принадлежит: UNIVERSITY OF SOUTH CAROLINA

Methods and devices for detecting and quantifying water vapor concentration in spent nuclear fuel rods undergoing drying processes for safe storage purposes. 1. A plasma discharge cell water vapor detection system comprising:an inlet feedthrough;an outlet feedthrough;a plasma cell including at least two electrodes forming an inter-electrode gap;an optical emission spectrometer including at least one optical filter;a chamber containing at least the at least one electrode and the optical emission spectrometer;a cathode;an anode; andat least one flange.2. The plasma discharge cell water vapor detection system of wherein the at least one electrode is formed from copper.3. The plasma discharge cell water vapor detection system of including at least two electrodes having an insulation coating on an outer periphery of the at least two electrodes excluding insulation at an end of each electrode that forms an inter-electrode gap between the at least two electrodes.4. The plasma discharge cell water vapor detection system of wherein the chamber comprises a four-way cross chamber.5. The plasma discharge cell water vapor detection system of wherein at least one flange comprises a visualization port.6. The plasma discharge cell water vapor detection system of wherein the system is insensitive to any type of radiation effect from a spent nuclear fuel rod.7. The plasma discharge cell water vapor detection system of wherein the system is installed in a spent nuclear fuel cask.8. The plasma discharge cell water vapor detection system of wherein the system has an inter-electrode separation distance capable of being increased or decreased.9. A method for quantifying water vapor concentration in a gaseous stream comprising:forming a negative pressure via a vacuum pump inside a plasma chamber;injecting a sample into a vacuum chamber thereby mixing the sample with a carrier gas;flowing the mixed carrier gas mixed with sample past at least one mass flow controller;initiating and ...

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Номер записи: 79
14-08-2014 дата публикации

Process and device for reducing radioactive material of object containing radioactive material to safe level in living environment

Номер: US20140228612A1
Автор: Akio Ishido, Hiroshi Hayakawa, Kaname Mogami, Kenji Takeshita, Kunio Yoshikawa, Masanori Aritomi, Saburo Kikuchi, Shigeto Hayafuji, Takeshi Kamaya, Yasuhiro Jinbo
Принадлежит: CDM Consulting Co Ltd, RADWASTE AND DECOMMISSIONING CENTER

Provided are a process and device for reducing radioactive material of an object containing the radioactive material to a safe level in a living environment. Included are a step of performing at least a step of carrying out a heating process on the object, into which radioactive material is absorbed and/or adsorbed from an environment or which absorbs and/or adsorbs radioactive material from an environment, in a state where temperature is less than or equal to the critical temperature of water and pressure is greater than or equal to the saturated vapor pressure of water, or a step of abruptly releasing the pressure; and a step of separating, after the above step, into liquid and solid.

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Номер записи: 80
10-06-2021 дата публикации

DISMANTLING AND DECONTAMINATION SYSTEM AND METHOD OF BIODEGRADABLE CONCRETE OF PWR TYPE NUCLEAR POWER PLANT

Номер: US20210174978A1
Автор: Hong Sung-Hoon, HWANG Seok-Ju, Kim Cheon-Woo, Lee Mi-Hyun
Принадлежит:

A dismantling and decontamination system of biodegradable concrete of a nuclear power plant according to an exemplary embodiment includes: a dismantling device for dismantling an in-core instrument installed under biodegradable concrete to form a lower penetrated part of the biodegradable concrete; a decontamination device inserted inside the biodegradable concrete for decontaminating radioactive waste of the inner wall of the biodegradable concrete; a waste receiving device movable through the lower penetrated part of the biodegradable concrete; and a blocking device for blocking the upper opening of the biodegradable concrete to block an outflow of the radioactive dust. 1. A dismantling and decontamination system of biodegradable concrete of a nuclear power plant , comprising:a dismantling device for dismantling an in-core instrument installed under biodegradable concrete to form a lower penetrated part of the biodegradable concrete;a decontamination device inserted inside the biodegradable concrete for decontaminating radioactive waste of the inner wall of the biodegradable concrete;a waste receiving device movable through the lower penetrated part of the biodegradable concrete; anda blocking device for blocking the upper opening of the biodegradable concrete to block an outflow of the radioactive dust.2. The dismantling and decontamination system of the biodegradable concrete of the nuclear power plant of claim 1 , further comprisinga dust collecting device connected to the dust blocking device and collecting the radioactive dust.3. The dismantling and decontamination system of the biodegradable concrete of the nuclear power plant of claim 2 , whereinthe waste receiving device includes:a receiving unit receiving radioactive waste;a receiving unit size adjusting unit for adjusting the size of the receiving unit; anda moving unit for moving the receiving unit.4. The dismantling and decontamination system of the biodegradable concrete of the nuclear power plant of ...

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Номер записи: 81
10-06-2021 дата публикации

CLEANING COMPOSITION FOR DECONTAMINATING SURFACES, IN PARTICULAR RADIOACTIVE SURFACES, AND RELATIVE DECONTAMINATION

Номер: US20210174981A1
Автор: Marin Adriano
Принадлежит:

The present invention relates to a cleaning composition for the decontamination of surfaces, in particular radioactive surfaces, and the related decontamination method. 1. A method for decontaminating from radioactive residuals components or parts of a plant that come into contact with radioactive material , wherein said method comprises applying to said parts an aqueous composition (C) comprising:i. at least one of citric acid, oxalic acid, tartaric acid, malic acid and the respective sodium or potassium salts, ethylenediaminetetraacetic acid (EDTA), a bisodium salt, other synthetic complexing agents and mixtures thereof; andii. at least one solvent selected from methanol, ethanol, ethyl acetate, propanol and isomers thereof, propylene glycol, butanol and isomers thereof, water soluble low molecular weight esters, methyl acetate, ethyl acetate, ethyl formate, dimethyl carbonate, esters of carbonic acid, and mixtures thereof.2. The method according to claim 1 , wherein said composition (C) further comprises:iii. at least one surfactant selected from coco-glucoside, alkyl polyglucoside, glyceryl oleate, a linear sodium alkylbenzene sulfonate, sodium lauryl sulfate, sodium lauryl ether sulfate, soy lecithin, soy lysolecithin and mixtures thereof.3. The method according to claim 1 , wherein said composition (C) further comprises:iv. at least one apolar solvent, from among limonene or another terpene analog thereof, tetrachloroethylene, carbon tetrachloride, other halogenated solvents, and mixtures thereof.5. The method according to claim 1 , wherein the acid i. is citric acid or the solvent ii. is ethanol or the apolar solvent iv. claim 1 , if present claim 1 , is limonene claim 1 , citral or a mixture of limonene and citral claim 1 , or the surfactant iii. claim 1 , if present claim 1 , is at least one from coco-glucoside claim 1 , soy lecithin or lysolecithin claim 1 , wherein the acid i. is citric acid claim 1 , the solvent ii. comprises ethanol claim 1 , dimethyl ...

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Номер записи: 82
15-09-2022 дата публикации

HAZARDOUS MATERIAL REPOSITORY SYSTEMS AND METHODS

Номер: US20220288658A1
Автор: Muller Richard A.
Принадлежит:

A power generator system includes one or more heat transfer members configured to contact: a heat source in a hazardous waste repository of a directional drillhole that stores nuclear waste in one or more nuclear waste canisters, and a heat sink in the hazardous waste repository; and one or more thermoelectric generators thermally coupled to the one or more heat transfer members and configured to generate electric power based on a temperature difference between the heat source and the heat sink. 1. (canceled)2. A power generator , comprising:one or more heat transfer members configured to contact: a heat source in a hazardous waste repository of a directional drillhole that stores nuclear waste in one or more nuclear waste canisters, and a heat sink in the hazardous waste repository; andone or more thermoelectric generators thermally coupled to the one or more heat transfer members and configured to generate electric power based on a temperature difference between the heat source and the heat sink.3. The power generator of claim 2 , wherein the nuclear waste comprises spent nuclear fuel.4. The power generator of claim 2 , wherein the heat source comprises at least one of one of the nuclear waste canisters or a casing disposed in the drillhole.5. The power generator of claim 2 , wherein the heat sink comprises at least one of the casing disposed in the drillhole or a material that at least partially fills the drillhole.6. The power generator of claim 5 , wherein the material comprises a liquid.7. The power generator of claim 2 , further comprising one or more biasing members configured to urge the one or more heat transfer members into thermal contact with the heat source and the heat sink.8. The power generator of claim 2 , further comprising at least one radiation shield.9. The power generator of claim 8 , wherein the radiation shield comprises tungsten.10. The power generator of claim 2 , wherein the one or more heat transfer members comprises a radiation ...

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Номер записи: 83
21-08-2014 дата публикации

Method and Apparatus for Removing Cesium Ion from Water

Номер: US20140231353A1
Автор: Hata Hideyuki, Ohnishi Noriyuki, Xie Xiaomao
Принадлежит: JNC CORPORATION

The present invention provides a method for efficiently separating cesium ions in a short time from an aqueous solution with the number of human working steps being reduced as much as possible and recovering the cesium ions, and an apparatus therefor. 1. A method for removing cesium ions in an aqueous solution , comprisingpreparing a cesium ion-containing magnetic particle in a cesium-containing aqueous solution and(i) magnetically separating the magnetic particle; or(ii) filtering or magnetically separating the magnetic particle.2. (canceled)3. The method for removing cesium ions according to claim 1 , wherein the cesium ion-containing magnetic particle is prepared after homogeneously reacting cesium ions and a cesium ion-adsorbing substance.4. The method for removing cesium ions according to claim 1 , wherein the cesium ion-containing magnetic particle is prepared by adding a water-soluble ferrocyanide and a water-soluble iron salt to the cesium-containing aqueous solution.5. The method for removing cesium ions according to claim 3 , wherein the water-soluble ferrocyanide is potassium ferrocyanide or sodium ferrocyanide and the water-soluble iron salt is iron chloride claim 3 , iron sulfate or iron nitrate.6. The method for removing cesium ions according to claim 4 , wherein the water-soluble iron salt is a mixture of ferrous chloride and ferric chloride.7. The method for removing cesium ions according to claim 1 , wherein after the preparation of the magnetic particle claim 1 , filtration or magnetic separation is performed by adding a flocculant to the cesium-containing aqueous solution.8. The method for removing cesium ions according to claim 1 , wherein cesium ions in the cesium-containing aqueous solution are removed within 60 minutes in total of the magnetic fine particle preparation step and the filtration or magnetic separation step.9. A removal apparatus for performing the removal of cesium ions in a cesium-containing aqueous solution according to claim 1 ...

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Номер записи: 84
17-06-2021 дата публикации

CLOSURE METHODS FOR MINES

Номер: US20210178437A1
Автор: Alexander Richard Wayne, Simon William, Yost Karl William
Принадлежит:

Treatment technology directed to using mine waste as a raw material to manufacture a mine filling product for use as a suitable precursor product or mine filling product to be used as a backfill material to close a mine. The precursor product or mine filling product retains its metals and is not be able to generate acidity. According to the disclosure, the precursor product or mine filling product, when placed in a mine, may also remove metals from mine fluids in the mine it contacts, and still retain the metals it hosted when it was a mine waste prior to it being used as a raw material to manufacture the precursor stowing backfill product. 117-. (canceled)18. A method for processing mine waste , the method comprising:receiving a mine waste including one or more heavy metals, a first heavy metal leachability, and a first sulfide concentration, wherein the mine waste, when exposed to air or water, forms acid mine drainage that includes at least some of the one or more heavy metals; andtreating the mine waste to produce a product including (i) a second heavy metal leachability less than the first heavy metal leachability and (ii) a second sulfide concentration less than the first sulfide concentration.19. The method of claim 18 , wherein the mine waste has a first acidity and the product has a second acidity less than the first acidity.20. The method of claim 18 , wherein treating the mine waste comprises reducing acid-generating properties of the mine waste.21. The method of claim 18 , wherein the mine waste has a first material strength and the product has a second material strength higher than the first material strength.22. The method of claim 18 , further comprising backfilling at least a portion of a mine with the product.23. The method of claim 18 , wherein the product claim 18 , when exposed to an acidic fluid claim 18 , does not leach the one or more heavy metals.24. The method of claim 18 , wherein the product is configured to remove heavy metals contained ...

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Номер записи: 85
21-08-2014 дата публикации

BERM AND METHOD OF CONSTRUCTION THEREOF

Номер: US20140234028A1
Автор: Dudding Carlton, Pendleton Carrie
Принадлежит: AWT IP, LLC

A berm to increase capacity of an existing landfill comprises a reinforced portion and backfill material. The landfill comprises a waste-receiving recess having a surface and an outer perimeter, an edge surface peripherally adjacent at least a portion of the outer perimeter, and an existing accumulation of waste within the recess having a collective top surface. The reinforced portion has opposing inner and outer surfaces and comprises structural fill material and a plurality of reinforcing members disposed therein. The backfill material has an inner surface adjacent the inner surface of the reinforced portion. The inner surface of the reinforced portion and the adjacent inner surface of the backfill material are substantially non-planar. The backfill material is at least partially encapsulated by an impermeable membrane. At least a portion of the berm sits on at least a portion of the edge surface of the landfill. 1. A combination of an existing landfill and a berm to increase capacity of the existing landfill , the landfill comprising a waste-receiving recess having a surface and an outer perimeter , an edge surface peripherally adjacent at least a portion of the outer perimeter , and an existing accumulation of waste within the recess having a collective top surface , the berm comprising:a reinforced portion having opposing inner and outer surfaces and comprising structural fill material and a plurality of reinforcing members disposed therein; andbackfill material having an inner surface adjacent the inner surface of the reinforced portion, the backfill material being at least partially encapsulated by an impermeable membrane;wherein at least a portion of the berm sits on at least a portion of the edge surface of the landfill;wherein the inner surface of the reinforced portion and the adjacent inner surface of the backfill material are substantially non-planar, andwherein the inner surface of the reinforced portion and the adjacent inner surface of the backfill ...

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Номер записи: 86
02-06-2016 дата публикации

METHOD FOR REDUCING STRONTIUM ION CONCENTRATION

Номер: US20160155524A1
Автор: SUZUKI Takafumi
Принадлежит: KYOWA CHEMICAL INDUSTRY CO., LTD.

A method of reducing the strontium ion concentration of an aqueous solution. 1. A method of reducing the strontium ion concentration of an aqueous solution , comprising the step of:reacting a soluble compound (A) of at least one metal selected from the group consisting of calcium and magnesium with a soluble alkali carbonate (B) in an aqueous solution containing a strontium ion to produce a metal carbonate and incorporate the strontium ion into the metal carbonate.2. The method according to claim 1 , wherein the soluble compound (A) is at least one compound selected from the group consisting of a soluble calcium compound (A1) and a soluble magnesium compound (A2).3. The method according to claim 2 , wherein the soluble calcium compound (A1) is calcium chloride.4. The method according to claim 2 , wherein the soluble magnesium compound (A2) is magnesium sulfate.5. The method according to claim 1 , wherein the soluble alkali carbonate (B) is sodium carbonate.6. The method according to claim 1 , wherein the metal carbonate is at least one selected from the group consisting of calcium carbonate and magnesium carbonate.7. The method according to claim 1 , wherein the amount of the soluble compound (A) is such that the amount of the metal carbonate produced in the aqueous solution becomes 0.2 to 3.0 g based on 100 mL of the aqueous solution.8. The method according to claim 7 , wherein the amount of the soluble calcium compound (A1) is such that the amount of calcium carbonate produced in the aqueous solution becomes 0.2 to 1.0 g based on 100 mL at the aqueous solution.9. The method according to claim 7 , wherein the amount of the soluble magnesium compound (A2) is such that the amount of magnesium carbonate produced in the aqueous solution becomes 1.0 to 3.0 g based on 100 mL of the aqueous solution.10. The method according to claim 1 , wherein the molar amount of the soluble alkali carbonate (B) is 0.9 to 1.1 times the theoretical molar amount that enables it to become a ...

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Номер записи: 87
15-09-2022 дата публикации

DEEP GEOLOGICAL DISPOSAL OF HIGH LEVEL WASTE ONSITE AT NUCLEAR POWER PLANTS

Номер: US20220293292A1
Автор: Crichlow Henry
Принадлежит:

A method for evaluating, selecting, and implementing at existing nuclear surface (or near surface) sites a deeply located high-level nuclear waste (HLW) disposal repository that is located directly vertically below the areal confines of that existing site, within a particular deeply located geologic rock formation. Many of these existing sites are ideal because: they are already legally permitted and/or licensed for using nuclear/radioactive materials, they already have nuclear/radioactive materials onsite that need a long-term safe disposal solution, and many of these existing sites already have onsite useful infrastructure (e.g., roads, buildings, cooling pools, equipment, machinery, personnel, and/or the like). Such existing sites include nuclear power plants (operating or decommissioned), interim spent nuclear fuel rod assemblies (SNF) surface storage sites, and/or near surface SNF storage sites. The deep HLW disposal repository may include a vertical wellbore, a lateral wellbore, and/or a human-made cavern. 1. A system for long-term disposal of radioactive material within a deep geological repository that is located directly vertically below an areal boundary of an existing site that has nuclear waste , wherein the system comprises:at least a terrestrial surface portion of the existing site that has nuclear waste;at least one vertical wellbore that extends from the terrestrial surface of the existing site to the deep geological repository; andthe deep geological repository that is formed within at least a portion of a deep geological formation; wherein the deep geological repository is configured to receive and house a predetermined amount of the radioactive material; wherein the at least the portion of the deep geological formation is located below any water tables that exist below the existing site; wherein the at least the portion of the deep geological formation is located directly vertically below the areal boundary of the existing site.2. The system ...

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Номер записи: 88
07-05-2020 дата публикации

Method for Treating Radioactive Liquid Waste

Номер: US20200143952A1
Автор: HWANG In Tae, Jeong Kyung Hoon, Jeun Joon Pyo, KIM Dong Woo, Kim Tak Hyun, Lee Kang, Lim Seung Joo, Sohn Joon Yong
Принадлежит: KOREA ATOMIC ENERGY RESEARCH INSTITUTE

The present invention relates to a technology for treating radioactive liquid waste containing a hardly degradable compound, and more specifically, to a technology for treating radioactive liquid waste containing a material such as an organic decontamination agent, an inorganic decontamination agent, liquid scintillation counter liquid waste, and the like generated at nuclear power plants, nuclear facilities, facilities at which radiation (radioactivity) is used, and the like. The method for treating radioactive liquid waste of the present invention includes adding two or more selected from the group consisting of a metal ion, an oxidizing agent, air, oxygen, or nitrous oxide, and a semiconductor to radioactive liquid waste to prepare a pre-treatment solution, and irradiating the pre-treatment solution with radiation. 1. A method for treating radioactive liquid waste the method comprising:adding two or more selected from the group consisting of a metal ion, an oxidizing agent, oxygen or nitrous oxide, air, and a semiconductor to radioactive liquid waste, or adding one or more selected from the group consisting of an oxidizing agent, oxygen or nitrous oxide, air, and a semiconductor to radioactive liquid waste containing a metal ion to prepare a pre-treatment solution; andirradiating the pre-treatment solution with radiation.2. The method of claim 1 , wherein the metal ion is a transition metal ion.3. The method of claim 2 , wherein the transition ion comprises one or more selected from the group consisting of a scandium ion claim 2 , a titanium ion claim 2 , a vanadium ion claim 2 , a chromium ion claim 2 , a manganese ion claim 2 , an iron ion claim 2 , a cobalt ion claim 2 , a nickel ion claim 2 , a copper ion claim 2 , a zinc ion claim 2 , a yttrium ion claim 2 , a zirconium ion claim 2 , a niobium ion claim 2 , a molybdenum ion claim 2 , a technetium ion claim 2 , a ruthenium ion claim 2 , a rhodium ion claim 2 , a palladium ion claim 2 , a silver ion claim 2 , ...

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Номер записи: 89
09-06-2016 дата публикации

METHOD FOR TREATING AN ABSORBER PIN CONTAINING CONTAMINATED BORON CARBIDE AND SODIUM

Номер: US20160163405A1
Автор: BLEVIN Gwendal, Gastaldi Olivier, Godlewski Joël, LECLERC Amaud, PELISSET Bruno
Принадлежит:

Disclosed is a method for treating an absorber pin, wherein the pin comprises a cladding in which a sintered boron carbide-based material having cracks is located, the material having porosity less than 1% of the volume of the material, the cracks containing sodium and at least one radioactive material. The method includes contacting the material with a treatment reaction mixture including carbon dioxide and water, in such a manner that the production of sodium carbonate and the expansion thereof cause the opening of cracks and of the sheath from at least one slit provided in the sheath as well as the propagation of the treatment process within the material. The process overcomes the physical-chemical properties of a sintered boron carbide-based material as much as possible. These properties prevent an easy treatment of the sodium and radioactive material contained in the cracks of the material. 19-. (canceled)101) Method for treating an absorber pin , said pin comprising a cladding in which there is a material based on sintered boron carbide whose porosity represents less than % of the volume of the material , the material having cracks that contain sodium and at least one radioactive substance , the method comprising a treatment step in which the sodium is converted to sodium carbonate by a carbonation reaction by contacting the material with a treatment reaction mixture comprising in molar percentage 0.5% to 5% of steam , 5% to 25% of carbon dioxide and 74.5% to 94.5% of a chemically inert gas , in such a way that expansion of the carbonate causes opening up of the cracks and of the cladding starting from at least one slit made in the cladding as well as the propagation of said method of treatment within the material.11) Method of treatment according to claim 10 , wherein claim 10 , before the treatment step claim 10 , a pretreatment step is carried out by contacting the material with a pretreatment reaction mixture comprising in molar percentage 0.5% to 25% of ...

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Номер записи: 90
22-09-2022 дата публикации

ARRANGEMENT AND METHOD FOR DISMANTLING A CONTAINER

Номер: US20220301734A1
Автор: DÄTIG Stefan, LOEB Andreas
Принадлежит: SIEMPELKAMP NIS INGENIEURGESELLSCHAFT MBH

The invention relates to an arrangement for dismantling a container () which comprises a circumferential wall () and an opening surrounded by said wall and which has a screening cover () situated above the opening, and to a dismantling tool for cutting segments () out of the circumferential wall. The dismantling takes place in such a way that regions of the circumferential wall which have a hollow cylindrical geometry are successively cut into segments and these segments are then removed. After a segmented region having the hollow cylindrical ring geometry has been removed, an end edge of the remaining circumferential wall is available. The screening cover has a lowering device for successively lowering the screening cover onto a particular available end edge. 110201426424446535658. An arrangement for dismantling a container () , in particular of a nuclear plant , preferably a reactor pressure vessel , which comprises a circumferential wall () and an opening surrounded by said wall and which has a screening cover () situated above the opening , and a dismantling tool () for cutting segments ( , , , , ) out of the circumferential wall , wherein the dismantling takes place in such a way that regions of the circumferential wall which have a hollow cylindrical ring geometry are successively cut into segments and these segments are then removed , wherein , after a segmented region having the hollow cylindrical ring geometry has been removed , an end edge () of the remaining circumferential wall is available ,characterized in that{'b': 14', '30', '32', '66', '68', '58, 'the screening cover () comprises a lowering device (, , , ) for successively lowering the screening cover onto a particular available end edge ().'}2. The arrangement according to claim 1 ,characterized in that{'b': 30', '32', '66', '68', '30', '68', '14, 'the lowering device (, , , ) comprises at least two supports (, ), preferably three supports offset by 120° with respect to another, which can support ...

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Номер записи: 91
23-05-2019 дата публикации

METHOD FOR PRODUCING SOLIDIFIED RADIOACTIVE WASTE

Номер: US20190156964A1
Автор: KINOSE Yutaka, MIYABE Shinsuke
Принадлежит: Nippon Chemical Industrial Co., Ltd.

The present invention provides solidified radioactive waste into which a titanium-containing adsorbent having a radioactive element adsorbed thereto is vitrified, the solidified radioactive waste being capable of confining a large amount of the titanium-containing adsorbent having a radioactive element adsorbed thereto, and furthermore elution of the radioactive element from the vitrified waste being suppressed. A method for producing solidified radioactive waste of the present invention is characterized by including heat-melting a mixture that includes a titanium-containing adsorbent having a radioactive element adsorbed thereto, a SiOsource, and an MO source (M represents an alkali metal element) to form vitrified waste, and the titanium-containing adsorbent is preferably one or two or more selected from silicotitanate, an alkali nonatitanate, and titanium hydroxide. 1. A method for producing solidified radioactive waste , comprising heat-melting a mixture that comprises a titanium-containing adsorbent having a radioactive element adsorbed thereto , a SiOsource , and an MO source (M represents an alkali metal element) to form vitrified waste.2. The method for producing solidified radioactive waste according to claim 1 , wherein the titanium-containing adsorbent is one or two or more selected from silicotitanate claim 1 , an alkali nonatitanate claim 1 , and titanium hydroxide.3. The method for producing solidified radioactive waste according to claim 1 , wherein the radioactive element is cesium and/or strontium.4. The method for producing solidified radioactive waste according to claim 1 , wherein the SiOsource and the MO source are each an alkali silicate.5. The method for producing solidified radioactive waste according to claim 4 , wherein the alkali silicate is an anhydride.6. The method for producing solidified radioactive waste according to claim 4 , wherein the alkali silicate is sodium metasilicate.7. The method for producing solidified radioactive waste ...

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Номер записи: 92
15-06-2017 дата публикации

METHOD FOR SEPARATING TRITIATED WATER FROM LIGHT WATER

Номер: US20170165590A1
Автор: NAKAMURA Satoshi
Принадлежит: Global Clean Technology Inc.

Provided is an industrially feasible method for separating tritiated water from light water. 1. A method for separating tritiated water from light water , comprising: 'by converting into a gas hydrate consisting essentially of tritiated water and heavy water as the crystal structure under a condition of converting into the gas hydrate of at least one of heavy water and tritiated water, and yet keeping light water in the liquid state, and', 'a step of removing tritiated water and heavy water from light water by adding heavy water to a liquid mixture containing tritiated water and light water,'} by breaking the gas hydrate structure containing tritiated water and heavy water, so as to get a liquid mixture,', 'by converting the liquid mixture containing tritiated water and heavy water into a gas hydrate containing tritiated water in the crystal structure under a condition of converting into a gas hydrate containing tritiated water in the crystal structure and yet keeping heavy water in the liquid condition and then,', 'by breaking the gas hydrate structure of tritiated water, so as to collecting tritiated water in that order., 'a step of separating tritiated water from heavy water'}2. The method for separating tritiated water from light water according to claim 1 , wherein:the liquid mixture containing tritiated water and heavy water obtained by breaking the gas hydrate containing tritiated water and heavy water in the crystal structure may be recrystallized repeatedly for removal or reduction of light water contained in the gas hydrate and then, the liquid mixture containing tritiated water and heavy water may be converted into gas hydrate of tritiated water under a condition of converting into the gas hydrate of tritiated water and yet keeping heavy water in the liquid state. The present invention relates to a method for separating tritiated water from light water.Most of radioactive nuclear species in the contaminated water stored in Fukushima Daiichi Nuclear Power ...

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Номер записи: 93
16-06-2016 дата публикации

METHOD FOR DECONATIMNATING SOIL AND THE LIKE AND SYSTEM FOR DECONTAMINATING SOIL AND THE LIKE

Номер: US20160172064A1
Автор: SUYAMA Yasuhiro, Yoshida Hideo
Принадлежит:

An object to be decontaminated contaminated with radioactive material, e.g., contaminated soil or water, is introduced into eluting solvent and dissolved, and the radioactive material is separated from the object to be contaminated by elution of the radioactive material into the eluting solvent. The eluting solvent containing the radioactive materials dissolved therein and the object to be decontaminated are separated into solid and liquid. The soil after solid-liquid separation and from which the radioactive material is removed is collected, and the eluting solvent after solid-liquid separation and a separated liquid containing contaminated water are introduced into an electrolysis tank and electrolyzed. Metal ions such as those of the radioactive materials are deposited on the cathode in the electrolysis tank. Hydrogen containing tritium generated in electrolysis is collected in the electrolysis tank. The hydrogen is moved to the outside of the electrolysis tank and trapped. 1. A method for decontamination of an object comprising:introducing and dispersing or dissolving an object to be decontaminated, which is contaminated with radioactive materials, into an eluting solvent to separate the radioactive materials from the object to be decontaminated by dissolution of the radioactive materials into the eluting solvent, the object to be decontaminated comprising contaminated soil and contaminated water;separating the radioactive materials dissolved in the eluting solvent and the object to be decontaminated into solid and liquid;collecting the soil after said solid-liquid separation and from which the radioactive materials have been removed by elution;electrolyzing the separated liquid containing the eluting solvent and the contaminated water after said solid-liquid separation by introducing the separated liquid into an electrolysis tank provided with and anode and at cathode;depositing metal ions comprising the radioactive materials on the cathode;collecting hydrogen ...

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Номер записи: 94
29-09-2022 дата публикации

Method And Arrangement Using Buried Tubular Members To Increase The Rate Of Decay Of Radioactive Contamination

Номер: US20220310280A1
Автор: Niemczyk Andrew
Принадлежит:

A method and combination for increasing the rate of radioisotope decontamination in a layer of soil beneath the ground surface, as well as in low growing plants on the ground surface, in which an array of elongated tubular members with defined lengthwise passageways, acting as a passive system, are inserted into vertical holes drilled into the ground. The elongated tubular members are slotted along their lengths to capture positrons emitted into the soil and direct the flow of the captured positrons into the soil layer 1. A method of increasing the rate of decontamination of soil containing radioisotopes from the natural rate of decontamination , including drilling an array of vertical holes downward through the surface of the ground in a region of the ground which is contaminated;installing a respective elongated tubular member in each hole of an array of vertically extending drilled holes; forming said elongated tubular member formed with a cluster of parallel passageways extending along the length of said elongated tubular member main portion;each of said elongated tubular member having a slot extending radially into each passageway along the length of said elongated member so as to allow entry of positrons emitted from radioisotopes and passing through the soil, said positrons redirected vertically upward within each of said passages; andinstalling a cap on an upper end of each of said tubular elongated members blocking said slots on said end thereof and said passages of the upper end of each elongated member main portion to cause said flow of said positrons to be reversed and to be directed down into the continuing upward flow of said positrons to cause flow of said positrons to radially out into said soil and in an upward direction in said soil to cause contact of said positrons with unstable nucleuses of atoms of radioisotopes therein thereby said contact causing stabilization of said nucleuses so as to substantially increase the rate of decontamination to be ...

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Номер записи: 95
08-07-2021 дата публикации

REMOTE DISMANTLING SYSTEM FOR NUCLEAR POWER PLANT AND NUCLEAR POWER PLANT HAVING SAME

Номер: US20210210236A1
Автор: HYUN Dongjun, KANG Shinyoung, KIM Changhoi, LEE Sunguk
Принадлежит:

The present invention relates to a remote dismantling system for a nuclear power plant. The remote dismantling system comprises: a transfer system for transferring an object to be cut, the object being placed on one side of the transfer system; and a parallel vertical rail robot arm system arranged within the movement radius of the transfer system and comprising a cutting robot arm for cutting the object to be cut. The transfer system comprises a gantry crane and a jib crane which are arranged to be movable across each other and transfer the object to be cut. The parallel vertical rail robot arm system comprises a transfer robot arm arranged to be vertically movable along with the cutting robot arm in order to transfer finely cut pieces which have been cut from the object to be cut. 1. A remote dismantling system comprising:a robot arm system provided with a cutting robot arm for cutting an object to be cut and a transfer robot arm for transferring cut pieces, which have been cut from the object to be cut, the cutting robot arm and the transfer robot arm being arranged to be movable in a vertical direction, respectively;a transfer system provided with a gantry crane for transferring the object to be cut and a jib crane for receiving and transferring the cut pieces from the transfer robot arm, the gantry crane and the jib crane being arranged to be movable across each other; anda band saw and turntable system provided with a horizontal support frame, a turntable having the object to be cut placed thereon, and installed on the horizontal support frame to be horizontally movable and rotatable, and a band saw for cutting the object to be cut placed on the turntable.2. The system of claim 1 , wherein the robot arm system comprises:a plurality of vertical rails extending in parallel in a vertical direction and having the cutting robot arm and the transfer robot arm mounted thereto to be movable up and down; andlifting blocks mounted to the plurality of vertical rails to ...

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Номер записи: 96
28-06-2018 дата публикации

Storing Hazardous Material in a Subterranean Formation

Номер: US20180182504A1
Автор: Elizabeth A. Muller, Richard A. MULLER
Принадлежит: Deep Isolation Inc

A hazardous material storage bank includes a wellbore extending into the Earth and including an entry at least proximate a terranean surface, the wellbore including a substantially vertical portion, a transition portion, and a substantially horizontal portion; a storage area coupled to the substantially horizontal portion of the well bore, the storage area within or below a shale formation, the storage area vertically isolated, by the shale formation, from a subterranean zone that includes mobile water; a storage container positioned in the storage area, the storage container sized to fit from the wellbore entry through the substantially vertical, the transition, and the substantially horizontal portions of the wellbore, and into the storage area, the storage container including an inner cavity sized enclose hazardous material; and a seal positioned in the wellbore, the seal isolating the storage portion of the wellbore from the entry of the wellbore.

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Номер записи: 97
28-06-2018 дата публикации

WASTE REPOSITORY FOR THE STORAGE OF RADIOACTIVE MATERIAL AND METHOD FOR ITS CONSTRUCTION

Номер: US20180182505A1
Автор: Diefenbach Reiner
Принадлежит:

The invention relates to a repository () for storing radioactive material in a rock formation, wherein there are at least two cavity systems () which are spaced apart from each other, and wherein a first cavity system () forms a repository chamber () for the radioactive material in containers () and the second cavity system () forms an access system (), wherein the rock formation is a mountain mass (), in which the first and second cavity systems () are connected to each other via connecting passages () at a plurality of transition points, wherein the first cavity system () forms a repository chamber () in which the containers () are free-standing and are accessible and removable, even when the repository chamber () is completely full, and the second cavity system () forms an access system () enabling permanent access and being at a distance from the repository chamber () such that the access system () forms a radiation-free region for access to the repository chamber () at different locations of the first cavity system (). 1. A repository for storing radioactive material in a rock formation , wherein at least two mutually spaced cavity systems are provided , and wherein a first cavity system forms a repository chamber for the radioactive material in containers , and the second cavity system forms an access system ,wherein the rock formation is a mountain mass, in which the first and second cavity systems are connected to each other at a plurality of transition points via connecting passages, wherein the first cavity system forms a repository chamber in which the containers are free-standing and are accessible and removable even when the repository chamber is filled to capacity, and the second cavity system forms an access system enabling permanent access and being arranged at such a distance from the repository chamber that the access system forms a radiation-free region for access to the repository chamber at different locations of the first cavity system.2. The ...

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Номер записи: 98
29-06-2017 дата публикации

Storing hazardous material in a subterranean formation

Номер: US20170186505A1
Автор: Muller Elizabeth, Muller Richard A.
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A hazardous material storage bank includes a wellbore extending into the Earth and including an entry at least proximate a terranean surface, the wellbore including a substantially vertical portion, a transition portion, and a substantially horizontal portion; a storage area coupled to the substantially horizontal portion of the well bore, the storage area within or below a shale formation, the storage area vertically isolated, by the shale formation, from a subterranean zone that includes mobile water; a storage container positioned in the storage area, the storage container sized to fit from the wellbore entry through the substantially vertical, the transition, and the substantially horizontal portions of the wellbore, and into the storage area, the storage container including an inner cavity sized enclose hazardous material; and a seal positioned in the wellbore, the seal isolating the storage portion of the wellbore from the entry of the wellbore. 1. A hazardous material storage bank , comprising:a wellbore extending into the Earth and comprising an entry at least proximate a terranean surface, the wellbore comprising a substantially vertical portion, a transition portion, and a substantially horizontal portion;a storage area coupled to the substantially horizontal portion of the well bore, the storage area within or below a shale formation, the storage area vertically isolated, by the shale formation, from a subterranean zone that comprises mobile water;a storage container positioned in the storage area, the storage container sized to fit from the wellbore entry through the substantially vertical, the transition, and the substantially horizontal portions of the wellbore, and into the storage area, the storage container comprising an inner cavity that encloses hazardous material; anda seal positioned in the wellbore, the seal isolating the storage area of the wellbore from the entry of the wellbore.2. The hazardous material storage bank of claim 1 , wherein the ...

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Номер записи: 99
07-07-2016 дата публикации

METHOD FOR REDUCING THE RADIOACTIVE CONTAMINATION OF THE SURFACE OF A COMPONENT USED IN A NUCLEAR REACTOR

Номер: US20160196889A1
Автор: RAMMINGER Ute, Sempere-Belda Luis, STELLWAG Bernhard
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The invention relates to a method for reducing the radioactive contamination of the surface of a component used in a nuclear reactor, which component is in contact with radioactively contaminated water, in which method a hydrophobic film is produced on the surface of a component by virtue of the surface being wetted with an aqueous solution which contains a film-forming amphiphilic substance. 1. A process for reducing the radioactive contamination of the surface of a component which is used in a nuclear reactor and is in contact with radioactively contaminated water , wherein a hydrophobic film is produced on the surface of a component by wetting the surface with an aqueous solution containing a film-forming amphiphilic substance.2. The process as claimed in claim 1 , characterized in that the hydrophobic film is produced on the interior surface of a component of a water-conducting circuit of the nuclear reactor.3. The process as claimed in claim 2 , characterized in that the hydrophobic film is produced after a part-circuit or full-circuit decontamination of the circuit.4. The process as claimed in claim 2 , characterized in that in the case of replacement of a component by a new component claim 2 , the hydrophobic film is produced on the new component.5. The process as claimed in claim 1 , characterized in that the hydrophobic film is produced at a point in time outside load operation.6. The process as claimed in claim 5 , characterized in that the hydrophobic film is produced during the start-up phase of the reactor.7. The process as claimed in claim 2 , characterized in that at least one depot substance is applied to the surface before production of the hydrophobic film.8. The process as claimed in claim 7 , characterized in that the depot substance is a noble metal.9. The process as claimed in claim 7 , characterized in that the depot substance is a salt of chromic acid.10. The process as claimed in claim 1 , characterized in that it is an inspection process ...

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Номер записи: 100