Система для уменьшения вредных выбросов в атмосферу для промышленной или атомной электростанции

Изобретение относится к системе для уменьшения вредных выбросов в атмосферу из промышленной или ядерной установки (1) в случае аварии. Система содержит следующие компоненты: конструкцию (10) для обеспечения непроницаемости почвы, которая проходит, по меньшей мере, по кольцеобразному участку, окружающему установку (1); множество опрыскивающих вышек (20-22), расположенных вокруг установки (1) и/или на прилегающей территории и выполненных с возможностью разбрызгивания воды в атмосферу, предпочтительно смешанной с химическими, и/или биологическими, и/или минеральными веществами; и периферийную конструкцию (50) для сбора, выполненную с возможностью приема воды, задержанной конструкцией (10) для обеспечения непроницаемости почвы. Техническим результатом является обеспечение возможности локализации загрязнений в случае аварии на ядерных или промышленных установках. 2 н. и 12 з.п. ф-лы, 22 ил.

ПОКРЫВАЮЩЕЕ УСТРОЙСТВО ДЛЯ ЭЛЕКТРОСТАНЦИЙ С КАМЕРНЫМ ЭЛЕМЕНТОМ И НАКРЫВАЮЩИМ ЭЛЕМЕНТОМ С МНОГОКРАТНОЙ ФИКСАЦИЕЙ И СПОСОБ ЗАПИРАНИЯ ПОКРЫВАЮЩЕГО УСТРОЙСТВА

FIELD: power plants incorporating chamber member and multiplelocking covering member. SUBSTANCE: covering device has primaryheader closing member disposed on first chamber and secured for removing by means of load-bearing coupling member designed for sealing inner space of primary chamber member; closing member is joined to chamber member by means of additional coupling members through sealed joint. Chamber member has adjustable-height intermediate members whereon closing member is lowered during their installation on chamber member. Method for locking covering device includes following operations: draw-out of adjustable-height intermediate members; installation of closing member onto intermediate members; partial lowering of intermediate members; turning of closing member relative to chamber member; complete lowering of intermediate members; forced jointing of closing member and chamber member. EFFECT: enhanced reliability. 19 cl, 5 dwg эосгегс пы сэ (19) РОССИЙСКОЕ АГЕНТСТВО ПО ПАТЕНТАМ И ТОВАРНЫМ ЗНАКАМ ВИ "” 2191 306. (51) МПК? 13) С2 Е 16 4 13/06 12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ (21), (22) Заявка: 99107661/06, 19.09.1997 (24) Дата начала действия патента: 19.09.1997 (30) Приоритет: 20.09.1996 ОЕ 19638657.8 (43) Дата публикации заявки: 20.02.2001 (46) Дата публикации: 20.10.2002 (56) Ссылки: Ц$ 4077840 А, 07.03.1978. ЗЦ 1212320 А, 15.02.1986. Ц$ 3895135 А, 22.01.1915. $4 1613756 АЛ, 15.12.1990. $Ц 1078162 А, 07.03.1984. (85) Дата перевода заявки РСТ на национальную фазу: 20.04.1999 (86) Заявка РСТ: ЕР 97/05161 (19.09.1997) (87) Публикация РСТ: М/О 98/12458 (26.03.1998) (98) Адрес для переписки: 129010, Москва, ул. Большая Спасская, 25, стр.3, ООО "Юридическая фирма Городисский и Партнеры", Е_.В.Томской (71) Заявитель: СИМЕНС АКЦИЕНГЕЗЕЛЛЬШАФТ (0Е) (72) Изобретатель: ШОЛЬЦ Манфред (0Е) (73) Патентообладатель: СИМЕНС АКЦИЕНГЕЗЕЛЛЬШАФТ (0Е) (74) Патентный поверенный: Емельянов Евгений Иванович (54) ПОКРЫВАЮЩЕЕ УСТРОЙСТВО ДЛЯ ЭЛЕКТРОСТАНЦИЙ С ...

Способ и установка для извлечения радиоактивных нуклидов из отработанных смоляных материалов

Изобретение относится к способу извлечения радиоактивных изотопов из стоков отработавших смоляных материалов атомных электростанций и к установке для осуществления способа. Способ включает обработку отработавшей смолы органической кислотой или щелочным соединением с целью высвобождения радиоизотопа из отработавшей смолы и получения технологического раствора, содержащего радиоизотоп, при этом отработавшая смола представляет собой ионообменную смолу, выбранную из группы, состоящей из катионо- и анионообменных смол, смешанных ионообменных смол и их смеси, нагруженную радиоизотопом, отделение радиоизотопа из технологического раствора по специфичной к радиоизотопу реакции и получение технологического раствора, обедненного радиоизотопом, при этом специфичную к радиоизотопу реакцию выбирают из группы, включающей физическую реакцию, электрохимическую реакцию, реакцию осаждения и их комбинацию, при этом обедненный технологический раствор содержит органическую кислоту или щелочное соединение и ионные ...

УСТРОЙСТВО НАКОПЛЕНИЯ, ОТОБРАЖЕНИЯ И ОТВОДА ВОЗДУХА ДЛЯ ИСПОЛЬЗОВАНИЯ В ЯДЕРНОЙ ПРОМЫШЛЕННОСТИ

Устройство для накопления, изоляции, отображения и отвода накопленного газа в трубе системы с текучей средой включает в себя основное трубное соединительное устройство, прикрепленное к трубе системы, в которой просверлено отверстие. Вертикальная труба, прикрепленная к трубному соединительному устройству, вмещает в себя магнитный поплавок. Индикатор уровня магнитного поплавка снаружи трубы отображает уровень магнитного поплавка. Клапан, прикрепленный к вертикальной трубе над магнитным поплавком, обеспечивает управляемый отвод газа из вертикальной трубы и, таким образом, из системы трубопроводов. Газ из трубы системы, накапливающийся в вертикальной трубе, удаляется из первичного пути потока текучей среды трубы системы. В вертикальной трубе, по мере снижения поверхности раздела жидкости/газа, поплавок опускается до заданного уровня, при котором пользователь отводит газ из системы трубопроводов, заставляя магнитный поплавок подниматься, отображая, что газ в системе трубопроводов снова находится ...

СИСТЕМА ДЛЯ УМЕНЬШЕНИЯ ВРЕДНЫХ ВЫБРОСОВ В АТМОСФЕРУ ДЛЯ ПРОМЫШЛЕННОЙ ИЛИ АТОМНОЙ ЭЛЕКТРОСТАНЦИИ

... 1. Система для уменьшения вредных выбросов в атмосферу из промышленной или ядерной установки (1) в случае аварии, содержащая:конструкцию (10) для обеспечения непроницаемости почвы, причем конструкция (10) для обеспечения непроницаемости почвы проходит, по меньшей мере, по кольцеобразному участку, окружающему установку (1);множество опрыскивающих вышек (20-22), расположенных вокруг установки (1) и/или на прилегающей к ней территории и выполненных с возможностью разбрызгивания в атмосферу воды, предпочтительно смешанной с химическими и/или биологическими и/или минеральными веществами; ипериферийную конструкцию (50) для сбора, выполненную с возможностью приема воды, задержанной конструкцией (10) для обеспечения непроницаемости почвы.2. Система по п. 1, в которой множество вышек содержит по меньшей мере одну группу вышек из следующих групп: группа опрыскивающих вышек (20, 21), расположенных на прилегающей к установке (1) территории и/или на ее границах, и группа опрыскивающих вышек (22), расположенных ...

KUEHLEINRICHTUNG FUER HILFSSYSTEME EINER NUKLEAREN ANLAGE

PROTECTION DEVICE FOR A FLY-WHEEL

UEberkritischer heterogener Atomkernreaktor vom Druckkesseltyp

TEMPERATURREGELVORRICHTUNG FUER FLUESSIGKEITSANLAGEN

FERNBEDIENTE VERSCHLIESSEINRICHTUNG.

Vorrichtung zum Einsetzen und zum Ausbau von Austauschelementen in einem Kernkraftwerk

EINRICHTUNG ZUM REGELN DES DRUCKS IN EINEM EINEN WAERMEERZEUGER UND EINE GASTURBINE ENTHALTENDEN GESCHLOSSENEN GASKREISLAUF

LIEGENDER SPEISEWASSERBEHAELTER

Combined active and passive reactor cavity water injection cooling system

METHOD OF PURIFYING THE CONDENSATE IN A NUCLEAR POWER PLANT AND PURIFICATION APPARATUS THEREFOR

... 1414179 Degassing and reducing the radioactivity of turbine condensate KRAFTWERK UNION AG 22 Dec 1972 [24 Dec 1971] 59539/72 Heading B1M A method of degasifying and reducing the radioactive content of condensate from a steam turbine in a nuclear power plant having a boiling water reactor arranged to generate steam to drive the turbine, a condenser for condensing exhaust steam from the turbine to form a main condensate, and at least one feed water preheater which is heated by steam which is bled from the turbine and which condenses to form a secondary condensate comprises feeding the secondary condensate into an expansion vessel, introducing the thus expanded secondary condensate into the main condensate from the condenser so as to evaporate and thereby degas the main condensate, the expanded secondary condensate being at a higher pressure and temperature than the main condensate, and conducting the degassed condensate along a flow path to reduce its radioactivity. Fig. 3 shows a degassing ...

MULTI-CABLE CONNECTING DEVICE FOR A NUCLEAR REACTOR

A movable connector plate 1 including a plurality of connectors is moved along a predetermined path towards a fixed connector plate 18 so that complementary connectors 2 of the plates 1,18 are connected, the movement being constrained by linkwork so that the connectors meet in correct alignment. The linkwork may be motor driven. Connectors may be electrical or non-electrical and are used in nuclear reactors. ...

Boiling water type nuclear power plant

To make it possible to more reliably supply cooling water to reactor pressure 102 and containment 101 vessels if a severe accident should occur, a boiling water type nuclear power plant includes a nuclear reactor building 100 and an external building 200 which is installed independently outside the nuclear reactor building and which has an anti-hazard property. The external building houses a power source 202, an operating panel 201, and at least one water injection pump 211/212. Alternative water injection pipes 110-150 allow water injection on at least the reactor pressure or containment vessel in the nuclear reactor building from the water injection pumps, and a valve connected to said pipes makes it possible to perform alternative water injection if a severe accident occurs.

Boiling water type nuclear power plant

Nuclear Power Plant

Protection method and protection system against commercial aircraft crash for nuclear power plant

The present application discloses a system and method for protecting a nuclear power plant against a commercial aeroplane crash. The protection system comprises an anti-plane crash shell 12 provided outside and independent of an inner containment shell 11 of the reactor building 1. A spent fuel storage building 2 and the electrical instrumentation and control building 3 are respectively and symmetrically located on both sides of the anti-plane crash shell, said buildings further comprising local double-layer partition wall configurations (6, Figs 2A & 2B) provided close to a potential impact direction. Two trains of engineered safety feature buildings 4,5, configured to perform a cooling function, and further function supporting buildings are also provided physically separated from the anti-plane crash shell. The reinforced anti-plane crash shell is equally spaced from the inner containment structure and includes a cylindrical wall and a dome with no fixed connections to the inner containment ...

Improvements in or relating to control valves and to plant incorporating control valves

... 910,515. Nuclear reactors. ROLLS-ROYCE Ltd. May 19, 1960 [May 1, 1959], No. 15054/59 Class 39 (4). [Also in Group XXIX] A nuclear reactor 50 has double - walled fuel - element receiving tubes 51 through which steam is fed for cooling purposes and the thin inner walls of which are supported by pressurized helium which is fed to the space between the walls from storage bottles 55 via a control valve 54 (see Group XXIX), which ensures that the helium pressure is equal to or slightly less than the pressure of the cooling steam. A pump 74 continuously circulates the heavy water moderator liquid via coolers 71, 73. The control tubes 80 are fed with heavy water from a dump tank 81 and the level of the heavy water in the tubes is varied, to vary the reactivity of the reactor. A pump 98 draws helium from the bottles 55 to provide a helium atmosphere for the dump tank 81 from which it passes to a weir tank 101, a cooler 103, a flame trap 105 and a recombiner 106. The weir tank 101 is located above ...

Mid-long term cooling method and system for containment and spent fuel pool in nuclear power plant after occurring accident

Heat exchanger

A method of manufacturing a layered structure of a plate-fin heat exchanger comprises stacking (figure 8) a first sheet 150, second sheet 154 and third sheet 152 such that the third sheet 152 is positioned between the first 150 and second sheets 154. Regions of the third sheet 152 are bonded to the first sheet 150 and the second sheet 154. Recesses 162 are provided in an outer surface of the first sheet 150 and/or the second sheet 154 at a position corresponding to the respective regions bonded to the third sheet 152. The first sheet 150, second sheet 154 and third sheet 152 are placed in a die and a hollow structure is formed (figure 9) so that the third sheet 152 spans a gap between the first 150 and second sheets 154 to define channels of the heat exchanger. The first 150 and/or second sheet 154 are pressed against a surface of the die so as to remove the recesses therefrom (see figure 10).

Improvements in or relating to valved ducting

... 857,528. Pipe systems. UNITED KINGDOM ATOMIC ENERGY AUTHORITY. Sept. 4, 1959 [Sept. 30, 1958], No. 31120/58. Class 99(2). In valved ducting for the flow of gases between a nuclear reactor and a heat exchanger 2, an inner pipe 4 providing a duct 5 for the flow of hot gas is coaxial with an outer pipe 3, the annulus between the pipes providing a duct 6 for counter current flow of cool gas. At one end of the duct 6 the pipe 3 is sealed to the pipe 4 as at 18, and outlet pipes 9, 10 communicate with the duct 5, a butterfly type valve 8 being incorporated between the pipes 9, 10 and the pipe 4 to control the flow of hot gas, whilst inlet ports 19, 22 communicate with the duct 6, a butterfly valve 21 being located between the ports to control the flow of cool gas. As shown, the pipes 4, 9, 10 are lined with heat insulation 11 made of laminated steel and vanes 12 are provided across the interior of the pipe 9. The port 19 is formed in an annular part 17 forming an extension of the pipe 3 and the ...

KINETIC ENERGY ABSROBING PAD

SUPPORT ARRANGEMENT FOR VERTICAL PIPING

... 1516438 Supporting pipes WESTING- HOUSE ELECTRIC CORP 7 July 1976 [9 July 1975] 28208/76 Heading F2P A vertical pipe 42 to be supported has a transition member 70 inserted therein. This member has an annular extension 88 with a ledge 86, a straight portion 84 and a tapered portion. Load sensing insulation 90, of a material such as that made of diatamaceous earth and fillers, is located adjacent the extension 88. A clamp 92 completely encircles the load bearing insulation 90. The clamp and insulation 90 are spaced from the pipe by a distance 93. The clamp can be supported from hangers 98 by rods 96 or by floor mounted frames.

RELEASABLE HIGH-PRESSURE SEAL AND METHOD OF FORMING SAME

VERFAHREN ZUR REINIGUNG VON MIT H2O BZW. D2O BENETZTEN TEILEN VON KERNKRAFTWERKSANLAGEN UND VORRICHTUNG ZUR DURCHFUHRUNG DIESES VERFAHRENS

TRAP FUER PIPINGS

LIVE STEAM ENCLOSURE FUER PRESSURIZED WATER REACTORS

PIPING PROTECTION FUER IN PARTICULAR IN NUCLEAR POWER STATIONS ARRANGED PIPINGS

MECHANISM FOR THE AERATION OF LIQUIDS

ROHRLEITUNGSSICHERUNG FUER INSBESONDERE IN KERN- KRAFTWERKEN ANGEORDNETE ROHRLEITUNGEN

ROHRLEITUNGSSICHERUNG FUER INSBESONDERE IN KERN- WERKEN ANGEORDNETE ROHRLEITUNGEN

NUCLEAR INSTALLATION AND PROCEDURE FOR THE OPERATION OF A NUCLEAR INSTALLATION

MECHANISM FOR THE AERATION OF LIQUIDS

TRAP FOR PIPINGS

PROCEDURE FOR THE OPERATION OF A NUCLEAR REACTOR AND NUCLEAR REACTOR

AIR-CLAIMANT UNTERTAGEKÜHLEINRICHTUNG WITH NATURAL LIFT

MECHANISM FOR THE DISTANCE OF HYDROGEN GAS FROM THE REACTOR CONTAINMENT OF A NUCLEAR REACTOR.

WARM MEMORY, IN PARTICULAR FOR THE AIR CONDITIONING OF RAUMEN

OPERATED BY REMOTE CONTROL TUBING LOCKING MECHANISM

Heterogeneous atomic nucleus reactor

THERMAL POWER MEASUREMENT APPARATUS

REMOTE CUTTING APPARATUS

REACTOR CONTAINMENT VESSEL VENT SYSTEM

This nuclear reactor containment vessel vent system (15) comprises: a rare gas filter (23) which is provided to the most downstream section of a vent line (13) and which allows at least steam from a vent gas that is discharged from inside a nuclear reactor containment vessel (1) and released into the atmosphere to pass but does not allow radioactive rare gas from the vent gas to pass; return pipes (24a, 24b) which connect a section of the vent line (13) immediately upstream of the rare gas filter (23) to the nuclear reactor containment vessel (1); and an intermediate vessel (100) which is provided on the return pipes (24a, 24b) and in which a gas comprising a radioactive rare gas that could not pass through the rare gas filter (23) flows and is stored. If a radioactive rare gas of at least a prescribed pressure has stagnated in the section of the vent line immediately upstream of the rare gas filter (23), the stagnated radioactive rare gas flows into the intermediate vessel (100), and thus ...

UPPER VESSEL TRANSPORT

A system for refueling a nuclear reactor is provided. The system includes a lower reactor vessel with a plurality of fuel rods and a plurality of control rods disposed therein, the lower reactor vessel further comprising an upper flange. An upper reactor vessel is provided which encloses a steam generator and a pressurizer, the upper reactor vessel further comprising a lower flange that matingly engages the upper flange of the lower reactor vessel. A transporter surrounds an outer surface of the upper reactor vessel, wherein the transporter is configured to translate the upper reactor vessel vertically toward and away from the lower reactor vessel and also to translate the upper reactor vessel horizontally toward or away from alignment with the lower reactor vessel.

PRESSURIZED WATER REACTOR WITH COMPACT PASSIVE SAFETY SYSTEMS

A nuclear reactor includes a pressure vessel and a nuclear reactor core disposed in the pressure vessel. A subterranean containment structure contains the nuclear reactor. An ultimate heat sink (UHS) pool is disposed at grade level, and an upper portion of the subterranean containment structure defines at least a portion of the bottom of the UHS pool. In some embodiments, the upper portion of the subterranean containment structure comprises an upper dome, which may protrude above the surface of the UHS pool to define an island surrounded by the UHS pool. In some embodiments, a condenser comprising a heat exchanger including hot and cold flow paths is disposed inside the subterranean containment structure; and cooling water lines operatively connect the condenser with the UHS pool.

SELF-CONTAINED EMERGENCY SPENT NUCLEAR FUEL POOL COOLING SYSTEM

An auxiliary system for cooling a spent nuclear fuel pool through a submersible heat exchanger to be located within the pool. In each train or installation, a single loop or series of loops of cooling fluid (e.g., sea water or service water) is circulated. The system is modular, readily and easily installed during an emergency and can be self operating with its own power source. Multiple trains may be used in parallel in order to accomplish the required degree of spent fuel pool cooling required.

HEAT REMOVAL SYSTEM AND METHOD FOR USE WITH A NUCLEAR REACTOR

A nuclear reactor includes a reactor vessel, a containment vessel that surrounds the reactor vessel, and a condenser that receives coolant from within the reactor vessel. The containment vessel and the condenser are at least partially submerged within a common reactor pool of liquid.

LEAK PREVENTION SYSTEM AND METHOD FOR A RETENTION POND

Leak prevention system for a liquid retention pond (10), the system comprising an upper membrane (3) to be covered by the liquid (1), a lower membrane (4) disposed on the bottom of the pond and joined in a sealed manner to the upper membrane on a peripheral portion to thereby form a normally sealed envelope (20) delimited by the two membranes, said envelope being filled with draining material (2), a plurality of passageways (6) disposed substantially horizontally in the draining material, a pumping device (5) adapted to generate an air vacuum in the drains, to suck leakage liquid that may have passed through the upper membrane, such that pollution of the surrounding soil is avoided in the event of a liquid leak through the upper membrane.

Einrichtung zur Abdichtung eines gasförmige Spaltprodukte aus Kernreaktionen enthaltenden Raumes gegen aussen an relativ zueinander bewegten Teilen

Einrichtung zum Abdichten einer ein möglicherweise radioaktiv verseuchtes Arbeitsmittel führenden Maschine

Wellendichtung eines Gebläses

Rohrleitungsanlage mit einem Bestandteil, der mittels fernbedienter Arbeitsgeräte auswechselbar ist

Schiebersteuervorrichtung an einem Wärmeaustauscher zum Steuern des Hin- und Rücklaufes eines Fluidums zum bzw. aus dem Wärmeaustauscher

Vanne

Wärmeabfuhr-Vorrichtung

Vanne

Heterogener Kernreaktor

Appareillage pour la mise en place et le démontage d'éléments d'échangeur dans un réacteur nucléaire

Absperrorgan

Druck- oder Siedewasserreaktor-Anlage mit Dampftrocknung

Fernbediente Verschliesseinrichtung

Aktivitätsmesseinrichtung für radioaktive Flüssigkeiten

Vorrichtung zur Abdichtung der Welle einer Turbomaschine für Wärmekraftanlagen, deren gasförmiges Arbeitsmittel in einem Atomkern-Reaktor erhitzt wird

Dispositif de retrait rapide d'une partie mobile d'une machine

Spule zum Anordnen in Nuten eines Stators oder Rotors einer elektrischen Maschine

Vorrichtung zum Ein- und Ausbauen von Filterkerzen

Luftbetriebene Untertagekühleinrichtung mit natürlichem Auftrieb

DREHKLAPPE.

VERFAHREN ZUM ENTSPANNEN, ENTGASEN UND ABKLINGENLASSEN DER RADIOAKTIVITAET SCHWACH RADIOAKTIVER DAMPFKONDENSATE IN KERNKRAFTWERKEN, SOWIE EINRICHTUNG ZUR DURCHFUEHRUNG DES VERFAHRENS.

REGELEINRICHTUNG FUER DAMPFTURBINEN.

Vorrichtung zum Auswechseln eines Gebläserades in einem Kernreaktor

Kombiniertes Schnellschluss- und Umleitventil

Reibdichtung zum Abdichten einer von relativ zueinander bewegten Grenzflächen begrenzten Öffnung

Fernsteuerbare Vorrichtung zum Bearbeiten von kreisringförmigen Verbindungen zwischen aneinanderstossenden Organen

Vorrichtung zum Überwachen von Lagern

Ventil, insbesondere Absperrventil

Oberflächenkondensator

Verfahren und Einrichtung zur Regelung der Durchflussmenge eines Fluides in einer Rohrleitung eines Kernkraftwerkes

Uberkritischer heterogener Atomkernreaktor vom Druckkesseltyp

Verfahren zur Tiefkühlung von in einer Kernreaktoranlage verseuchtem Kühlgas

Sicherheitsvorrichtung mit einem Ventil, vorzugsweise Absperrventil

Verfahren zur Reinigung von während des Betriebes mit HaeO bzw. DaeO benetzten Teilen von Kernkraftwerksanlagen

Einrichtung zum Absperren von Rohrleitungen

Pipeline safety device for pipelines arranged in particular in nuclear power stations

FLAP CHECK VALVE.

HYDRAULIC OSCILLATION BRAKE, IN PARTICULAR FOR THE APPLICATION IN NUCLEAR POWER STATIONS.

TRAP FOR PIPINGS.

Support device for components of a power station, in particular a nuclear power station

PROCEDURE AND MECHANISM FOR THE AERATION OF LIQUIDS.

Self-contained emergency spent nuclear fuel pool cooling system

An auxiliary system for cooling a spent nuclear fuel pool through a submersible heat exchanger to be located within the pool. In each train or installation, a single loop or series of loops of cooling fluid (e.g., sea water or service water) is circulated. The system is modular, readily and easily installed during an emergency and can be self operating with its own power source. Multiple trains may be used in parallel in order to accomplish the required degree of spent fuel pool cooling required.

Foundation for building in nuclear facility and nuclear facility

To provide a lower foundation 30 provided on a ground 7 and an upper foundation 31 provided above the lower foundation 30 via a seismic isolator 32 . A machine room S in which a machine K can be arranged is provided in the upper foundation 31 . A foundation-side hull structure unit that is configured in a reticular pattern by horizontal reinforcing ribs extended in one direction and vertical reinforcing ribs extended to cross the horizontal reinforcing ribs is provided in the upper foundation 31 , and the machine room S is provided in a space in the foundation separated by the horizontal reinforcing ribs and the vertical reinforcing ribs.

Backup nuclear reactor auxiliary power using decay heat

A nuclear plant auxiliary backup power system that uses decay heat following a plant shutdown to produce electrical power through a dedicated steam turbine/generator set. The decay heat produces a hot operating gaseous fluid which is used as a backup to run an appropriately sized turbine that powers an electrical generator. The turbine is configured to utilize a portion of the existing nuclear plant secondary system and exhausts the turbine exhaust to the ambient atmosphere. The system functions to both remove reactor decay heat and provide electrical power for plant systems to enable an orderly shutdown in the event traditional sources of electric power are unavailable.

CLAMPER AND IN-CHANNEL-HEAD OPERATION DEVICE

A clamper () includes a clamp mechanism () which inserts an insertion portion () into a tube member and clamps the tube member and a lifting and lowering mechanism () which lifts and lowers the clamp mechanism () in the insertion direction of the insertion portion (). The clamp mechanism () includes a clamp body () which has the insertion portion (), a cotter () which protrudes from the insertion portion () to come into friction-contact with the tube member and is movable in a reciprocating manner in the insertion direction of the insertion portion (), a piston rod () which presses the cotter () so that the cotter protrudes from the insertion portion () when being pulled toward the opposite side to the insertion direction of the insertion portion (), and a rod cylinder () which is integrated with the clamp body () and pulls the piston rod (). 1. A damper comprising: a clamp mechanism which inserts an insertion portion into a tube member and clamps the tube member; and a lifting and lowering mechanism which lifts and lowers the clamp mechanism in the insertion direction of the insertion portion , wherein the clamp mechanism includes a clamp body which includes the insertion portion , a cotter which protrudes from the insertion portion so as to come into friction-contact with the tube member and is disposed so as to be movable in a reciprocating manner in the insertion direction of the insertion portion , a piston rod which presses the cotter so that the cotter protrudes from the insertion portion when being pulled toward the opposite side to the insertion direction of the insertion portion , and a rod cylinder which is integrated with the clamp body and pulls the piston rod.2. The damper according to claim 1 , wherein the piston rod passes through the rod cylinder and protrudes toward a rear end portion of the rod cylinder.3. The damper according to claim 1 , further comprising: a pressurizing device which is connected through a tube to a fluid chamber that pulls the ...

Method for filling water into a main circuit of a nuclear reactor, and connection device for implementing said method

The method for filling water into and changing the air of a main circuit of a water-cooled nuclear reactor includes a step of placing a connection and fluid isolation device which is connected to a hot leg of each cooling loop of the main circuit so as to substantially insulate, from inside the vessel, the assembly of hot legs. The method also includes a step of injecting water through an injection circuit on at least one hot leg until each cooling loop is filled with water having changed the air from a steam generator and until the water level in the vessel reaches above the side openings of the vessel that correspond to the loops, after which the connecting device is taken out of the vessel. The connecting device is capable of using telescopic connection elements.

Method and system for supplying emergency power to nuclear power plant

Method and system for supplying emergency power to nuclear power plant, wherein the method includes, providing accumulator battery system, connected to emergency bus, the accumulator battery system is monitored by online monitoring system; in case of power loss of electrical devices of the nuclear power plant, the online monitoring system starts the accumulator battery system to provide power supply to the electrical devices of the nuclear power plant via the emergency bus. The present application is adapt to the key technologies and battery management technologies of million kilowatt-class advanced pressurized water reactor nuclear power plant, facilitating to improve the safety of the nuclear power plant in case of serious natural disasters beyond design working conditions.

ARRANGEMENT AND METHOD FOR PROVIDING AN EMERGENCY SUPPLY TO A NUCLEAR INSTALLATION

The invention relates to a method and an arrangement for providing an emergency supply to a nuclear installation. The arrangement comprises a container () with a plurality of permanently installed devices and at least one motor (), one generator (), one pump (), one fuel tank () and one transformer (), wherein the pump and the generator are functionally connected to the motor in order to activate said motor. 15910. An arrangement to provide an emergency supply to a nuclear installation () with a container () with several integrated facilities , comprising at least{'b': '22', 'a motor (),'}{'b': '26', 'a generator (),'}{'b': '24', 'a pump (),'}{'b': '14', 'a fuel tank (),'}{'b': '34', 'claim-text': whereby the pump and the generator are functionally connected to the motor to actuate said pump and generator,', 'characterized in that', {'b': 14', '10, 'the fuel tank () is situated in the region of the center of gravity of the container ().'}], 'a transformer (),'}2. The arrangement of claim 1 ,characterized in that{'b': 22', '24', '34, 'the motor () is embodied as a Diesel motor, in particular a turbocharged Diesel motor, with two shaft ends, whereby one of the shaft ends is connected to the pump (), preferably embodied as a self-priming pump designed for waste water, in particular a spiral casing pump, and the other shaft end is connected to the generator ().'}3. The arrangement of claim 1 ,characterized in that{'b': 14', '16', '18, 'sup': 3', '3, 'the fuel tank () with a holding capacity of at least 10 m, in particular 15 m, is embodied bullet-proof and in particular is subdivided into relaxation zones by partition plates (,).'}4. The arrangement of at least one of the preceding claims claim 1 ,characterized in that{'b': 10', '36, 'the container () contains a decontamination area () with a shower.'}5. The arrangement of at least one of the preceding claims claim 1 ,characterized in that{'b': 10', '40', '14', '22', '26', '24', '34, 'the container () is at least a ft ...

PASSIVE COOLING SYSTEM OF CONTAINMENT BUILDING AND NUCLEAR POWER PLANT COMPRISING SAME

The present invention discloses a passive cooling system of a containment building, to which a plate-type heat exchanger is applied. A passive cooling system of a containment building comprises: a containment building; a plate-type heat exchanger provided to at least one of the inside and the outside of the containment building and comprising channels respectively provided to the both sides of a plate so as to be arranged dividedly from each other such that the plate-type heat exchanger carries out mutual heat exchange between the internal atmosphere of the containment building and a heat exchange fluid while maintaining a pressure boundary; and a pipe connected to the plate-type heat exchanger by penetrating the containment building so as to form the path of the internal atmosphere of the containment building or the heat exchange fluid. 1. A passive containment building cooling system , comprising:a containment building;a plate type heat exchanger installed on at least one place of an inside and an outside of the containment building, and provided with channels arranged to be distinguished from one another at both sides of a plate to exchange heat between atmosphere within the containment building and heat exchange fluid from each other while maintaining a pressure boundary; anda line connected to the plate type heat exchanger through the containment building to form a flow path of the atmosphere within the containment building or the heat exchange fluid.2. The passive containment building cooling system of claim 1 , wherein the channels are formed in such a manner that a flow resistance of the inlet region is relatively larger than that of a main heat transfer region connected between an inlet region and an outlet region to mitigate flow instability due to two phase flow.3. The passive containment building cooling system of claim 2 , wherein the inlet region is formed with a smaller width than that of the main heat transfer region claim 2 , and formed to extend a ...

MAIN PUMP SHAFT SEAL WATER INJECTION SYSTEM OF A NUCLEAR POWER STATION

A main pump shaft seal water injection system of a nuclear power plant includes a jet pump, a high pressure cooler, a hydrocyclone, valves and a main connection pipeline outside of a main pump, and an auxiliary pump and an internal flow path inside the main pump. Inner and outer flow paths of the main pump are connected with a shaft seal water injection hole and a high temperature water drainage hole. The main connection pipeline is connected between an upper filling water pipeline and a shaft seal water injection hole. A bypass pipeline connected with the jet pump, the high pressure cooler and the hydrocyclone, the main connection pipeline is provided with a normally open main pipeline isolating valve. The bypass pipeline allows low temperature upper filling water in the RCV system to enter the shaft seal water injection hole of the main flange directly. 1. A main pump shaft seal water injection system of a nuclear power plant , comprising: a jet pump , a high pressure cooler , a hydrocyclone , valves and a main connection pipeline arranged outside of a main pump , an auxiliary pump and an internal flow path arranged in the main pump , inner and outer flow paths of the main pump are connected with a shaft seal water injection hole and a high temperature water drainage hole on the main connection pipeline , the main connection pipeline is connected between an upper filling water pipeline of a RCV and a shaft seal water injection hole of a main flange , the jet pump , the high pressure cooler and the hydrocyclone are sequentially arranged on the main connection pipeline , further comprising a bypass pipeline connected in parallel with the jet pump , the high pressure cooler and the hydrocyclone , the main connection pipeline is provided with a normally open main pipeline isolating valve at each end of a parallel section connected in parallel with the bypass pipeline , the bypass pipeline allows low temperature upper filling water in the RCV system to bypass the jet ...

INTEGRATED SYSTEM FOR CONVERTING NUCLEAR ENERGY INTO ELECTRICAL, MECHANICAL, AND THERMAL ENERGY AND METHODS FOR USING THE SAME

Provided is an apparatus for generating electricity, mechanical energy, and/or process and district heat using a gas propellant chamber fueled with fissile material and enclosed in a sealed containment vessel which also contains an operating gas. The system allows for the operating gas to be compressed as it enters the nuclear fuel chamber where it is heated. As the operating gas exits the nuclear fuel chamber, the kinetic energy of the gas is converted to rotational energy by a variety of methods. The rotational energy is further converted to electricity, mechanical energy, and/or process and district heat. The operating gas circulates in the containment vessel and is cooled prior to re-entering the gas propellant chamber. The apparatus thereby provides a simpler and safer design that is both scalable and adaptable. The apparatus is easily and safely transportable and can be designed to be highly nuclear-proliferation-resistant. 1. An apparatus for generating electricity comprising:a gas propellant chamber comprised of an annular body defining first and second ends, the first end of the annular body defining an inlet assembly that is configured to draw operating gas into the gas propellant chamber and the second end defining an exhaust assembly that is configured to expel operating gas from the gas propellant chamber, wherein the gas propellant chamber is housed in a containment vessel, the containment vessel having an inner wall and defining a region between the inner wall and the annular body of the gas propellant chamber, the region forming a bypass for operating gas to pass around the annular body;a nuclear fuel chamber positioned within the annular body of the gas propellant chamber between the first and second ends, the nuclear fuel chamber configured to heat the operating gas;a compressor positioned proximate the first end of the gas propellant chamber, the compressor configured to compress the operating gas prior to entry into the nuclear fuel chamber;a ...

INTEGRATED SYSTEM FOR CONVERTING NUCLEAR ENERGY INTO ELECTRICAL, MECHANICAL, AND THERMAL ENERGY AND METHODS FOR USING THE SAME

Provided is an apparatus for generating electricity, mechanical energy, and/or process and district heat using a gas propellant chamber fueled with fissile material and enclosed in a sealed containment vessel which also contains an operating gas. The system allows for the operating gas to be compressed as it enters the nuclear fuel chamber where it is heated. As the operating gas exits the nuclear fuel chamber, the kinetic energy of the gas is converted to rotational energy by a variety of methods. The rotational energy is further converted to electricity, mechanical energy, and/or process and district heat. The operating gas circulates in the containment vessel and is cooled prior to re-entering the gas propellant chamber. The apparatus thereby provides a simpler and safer design that is both scalable and adaptable. The apparatus is easily and safely transportable and can be designed to be highly nuclear-proliferation-resistant. 1. An apparatus for generating electricity comprising:a gas propellant chamber comprised of an annular body defining first and second ends, the first end of the annular body defining an inlet assembly that is configured to draw operating gas into the gas propellant chamber and the second end defining an exhaust assembly that is configured to expel operating gas from the gas propellant chamber, wherein the gas propellant chamber is housed in a containment vessel, the containment vessel having an inner wall and defining a region between the inner wall and the annular body of the gas propellant chamber, the region forming a bypass for operating gas to pass around the annular body;a nuclear fuel chamber positioned within the annular body of the gas propellant chamber between the first and second ends, the nuclear fuel chamber configured to heat the operating gas;a compressor positioned proximate the first end of the gas propellant chamber, the compressor configured to compress the operating gas prior to entry into the nuclear fuel chamber;a ...

INTEGRATED SYSTEM FOR CONVERTING NUCLEAR ENERGY INTO ELECTRICAL, MECHANICAL, AND THERMAL ENERGY AND METHODS FOR USING THE SAME

Provided is an apparatus for generating electricity, mechanical energy, and/or process and district heat using a gas propellant chamber fueled with fissile material and enclosed in a sealed containment vessel which also contains an operating gas. The system allows for the operating gas to be compressed as it enters the nuclear fuel chamber where it is heated. As the operating gas exits the nuclear fuel chamber, the kinetic energy of the gas is converted to rotational energy by a variety of methods. The rotational energy is further converted to electricity, mechanical energy, and/or process and district heat. The operating gas circulates in the containment vessel and is cooled prior to re-entering the gas propellant chamber. The apparatus thereby provides a simpler and safer design that is both scalable and adaptable. The apparatus is easily and safely transportable and can be designed to be highly nuclear-proliferation-resistant. 1. An apparatus for generating electricity comprising:a gas propellant chamber comprised of an annular body defining first and second ends, the first end of the annular body defining an inlet assembly that is configured to draw operating gas into the gas propellant chamber and the second end defining an exhaust assembly that is configured to expel operating gas from the gas propellant chamber, wherein the gas propellant chamber is disposed in a containment vessel, the containment vessel having an inner wall defining a circulation path for the operating gas to travel from the conversion apparatus to the inlet assembly;a nuclear fuel chamber positioned within the annular body of the gas propellant chamber between the first and second ends, the nuclear fuel chamber configured to heat the operating gas;a compressor positioned proximate the first end of the gas propellant chamber, the compressor configured to compress the operating gas prior to entry into the nuclear fuel chamber;a conversion apparatus positioned proximate the second end of the ...

CONTACT FORCE EVALUATION METHOD

There is provided a contact force evaluation method for evaluating a contact force against a supporting member of a tube bundle positioned in a fluid and supported by the supporting member, including a contact force setting step of setting a contact force of the tube bundle, a probability density function calculation step of calculating a probability density function of a reaction force received by the supporting member from the tube bundle in response to a predetermined input, using a vibration analysis model of the tube bundle and the supporting member, a probability calculation step of calculating a probability that a reaction force equal to or higher than the set contact force occurs, based on the calculated probability density function, and an evaluation step of evaluating the set contact force, based on the calculated probability. 1. A contact force evaluation method for evaluating a contact force against a supporting member of a tube bundle positioned in a fluid and supported by the supporting member , the method comprising:a contact force setting step of setting a contact force of the tube bundle;a probability density function calculation step of calculating a probability density function of a reaction force received by the supporting member from the tube bundle in response to a predetermined input, using a vibration analysis model of the tube bundle and the supporting member;a probability calculation step of calculating a probability that a reaction force equal to or higher than the set contact force occurs, based on the calculated probability density function; andan evaluation step of evaluating the set contact force, based on the calculated probability.2. The contact force evaluation method according to claim 1 ,wherein the probability density function calculation step includes: performing a time history response analysis on the vibration analysis model using the predetermined input to obtain a time history response; calculating an average value and a ...

FINNED STRAINER

The present invention relates to filters used to remove debris from water being sucked into a piping system. It has particular application use in nuclear power plants, which, after a loss of coolant accident, must pump cooling water back into the reactor core from a collection sump. This water may contain various types of debris that must be removed before the water is sent back into the reactor cooling system. There are restrictions on the allowable pressure drop across the strainer and the space available for installing this equipment. The finned strainer of the present invention addresses these issues while maximizing the quantity of debris filtered from the water. 1. A strainer for filtering debris from a fluid:i) an elongated header defining an enclosed flowpath, having an outlet in fluid communication with a suction source and a plurality of inlet apertures disposed along the length of said flowpath, said flowpath exhibiting a pressure drop in the direction of fluid flow;ii) a strainer element disposed in each said inlet aperture for straining debris from fluid entering said flowpath;iii) a flow controlling device for maintaining substantially uniform fluid flow through strainer elements located at different positions along said flowpath.2. The strainer of wherein the flow controlling device comprises an orifice for producing a pressure drop between an inlet aperture and the flowpath located at a position closer to said suction source that is greater than the pressure drop between an inlet aperture and said flowpath located at a position farther from said suction source.3. The strainer of wherein said orifice is in the form of a nozzle for accelerating the fluid entering said flowpath in a direction substantially parallel thereto.4. The strainer of wherein said orifice is formed in a baffle disposed in said header claim 3 , said baffle defining a collection channel enclosing a plurality of apertures.5. The strainer of wherein the header has a generally a ...

FAULT TOLERANT TURBINE SPEED CONTROL SYSTEM

A generator is installed on and provides electrical power from a turbine by converting the turbine's mechanical energy to electricity. The generated electrical power is used to power controls of the turbine so that the turbine can remain in use through its own energy. The turbine can be a safety-related turbine in a nuclear power plant, such that, through the generator, loss of plant power will not result in loss of use of the turbine and safety-related functions powered by the same. Appropriate circuitry and electrical connections condition the generator to work in tandem with any other power sources present, while providing electrical power with properties required to safely power the controls. 1. A turbine speed control system comprising:a generator installed on a turbine, wherein the generator generates electrical power from the turbine; andan electrical connection between the generator and at least one of a speed controller for the turbine and a control room flow controller for the speed controller, wherein the electrical connection permits operation of the speed controller and/or control room flow controller with the electrical power.2. The system of claim 1 , wherein the speed controller and the control room flow controller are remote from the turbine.3. The system of claim 1 , wherein the generator is an AC or DC approximately 200 Watt electrical generator.4. The system of claim 1 , wherein the electrical connection includes at least one isolation diode and filter to condition the electrical power provided to the speed controller and/or control room flow controller.5. The system of claim 4 , wherein the isolation diode prevents current surges to the generator and wherein the filter is a capacitor configured to reduce voltage surges in the electrical connection.6. The system of claim 1 , wherein the turbine is one of a Reactor Core Isolation Cooling (RCIC) turbine and a High Pressure Injection Cooling (HPIC) turbine in a nuclear power plant.7. The system of ...

Boiling Water Type Nuclear Power Plant

To more reliably supply cooling water to a reactor pressure vessel and a reactor containment vessel using a back-up building if a severe accident should occur, a boiling water type nuclear power plant includes a nuclear reactor building including a reactor containment vessel, and an external building, which is installed independently outside the nuclear reactor building and which has an anti-hazard property. The external building has a power source and an operating panel independent of the nuclear reactor building. The boiling water type nuclear power plant includes a water injection pump installed inside the external building, an alternative water injection pipe performing water injection at least on a reactor pressure vessel or the reactor containment vessel in the nuclear reactor building from the water injection pump, and a valve connected to the alternative water injection pipe, making it possible to perform alternative water injection if a severe accident occurs. 1. A boiling water type nuclear power plant comprising:a nuclear reactor building including a reactor containment vessel and a reactor pressure vessel;an external building which is installed independently outside the nuclear reactor building, which includes a power source and an operating panel independent of the nuclear reactor building, and which has an anti-hazard property;a water injection pump installed inside the external building;an alternative water injection pipe configured to perform water injection on at least the reactor pressure vessel or the reactor containment vessel in the nuclear reactor building from the water injection pump; anda valve connected to the alternative water injection pipe.2. The boiling water type nuclear power plant according to claim 1 , wherein a branching-off portion is provided at some midpoint in the alternative water injection pipe; and a hose connection portion allowing connection of a hose of a pumper vehicle is provided in a pipe branching off from the ...

METHODS OF MANUFACTURING STRUCTURES FROM OXIDE DISPERSION STRENGTHENED (ODS) MATERIALS, AND ASSOCIATED SYSTEMS AND DEVICES

Method of fabricating structures, such as parts for use in nuclear power generation systems, are described herein. A representative method of fabricating a part for a nuclear reactor system includes additively manufacturing the part as a monolithic structure from a wire formed of an oxide dispersion strengthen (ODS) material, which includes an oxide material dispersed within a metal material. Specifically, the method can include directing a beam of thermal energy toward the wire to melt the wire, and permitting the melted wire to cool and solidify to form the part such that the oxide material remains substantially dispersed within the metal material. By maintaining the dispersion of the oxide material within the metal material, the ODS material can retain a good creep resistance, wear-resistance, corrosion resistance, and/or other ODS material property at elevated temperatures—even after fabrication. 1. A method of fabricating a monolithic structure , the method comprising: directing a beam of thermal energy toward a wire formed of an oxide dispersion strengthened (ODS) material to melt the wire;', 'depositing the melted wire on a substrate to form a layer of the structure; and', 'permitting the melted wire to cool and solidify on the substrate., 'repeatedly, and in a stack-wise fashion—'}2. The method of wherein the ODS material includes an oxide material dispersed within a metal material claim 1 , and wherein permitting the melted wire to cool and solidify includes preventing the oxide material from coming out of solution from the metal material.3. The method of wherein the ODS material includes an oxide material dispersed within a metal material claim 1 , and wherein permitting the melted wire to cool and solidify includes permitting the melted wire to cool and solidify while the oxide material remains substantially dispersed within the metal material.4. The method of wherein the ODS material is molybdenum-lanthanum oxide.5. The method of wherein the ODS material ...

EMERGENCY METHOD AND SYSTEM FOR IN-SITU DISPOSAL AND CONTAINMENT OF NUCLEAR MATERIAL AT NUCLEAR POWER FACILITY

A method for in-situ subsurface isolation of nuclear material at a nuclear power or nuclear waste facility during an emergency includes a borehole located in close proximity and at a depth sufficient to safely isolate the radioactive material. A gravity fracture in the surrounding rock formation is located at the bottom end of the borehole, with the radioactive material entering the gravity fracture. A dense slurry or fluid could be mixed with the radioactive material to create and propagate the gravity fracture. 1. A method for in-situ subsurface isolation of nuclear material at a nuclear power or nuclear waste facility during an emergency , the method comprising:conveying a radioactive material from a source of the radioactive material into a borehole in proximity to the source of the radioactive material, the borehole being at a depth suitable for safely isolating the radioactive material, a first gravity fracture in a surrounding rock formation located below and in communication with a bottom end of the borehole;wherein the radioactive material is mixed with a slurry containing a weighting material, the slurry being denser than the surrounding rock formation;wherein the radioactive material passes from the borehole into the first gravity fracture; andwherein the radioactive material is not in a containment vessel when entering the borehole.2. (canceled)3. A method according to further comprising using the slurry to create the first gravity fracture.4. (canceled)5. A method according to further comprising conveying the slurry into the borehole after placing the radioactive material mixed with the slurry into the borehole.6. (canceled)7. (canceled)8. (canceled)9. A method according to further comprising extending the first gravity fracture downward as the radioactive material mixed with the slurry propagates downward.10. (canceled)11. (canceled)12. A method according to further comprising controlling a cooling rate of the radioactive material as the radioactive ...

Nuclear Power Generation using a Thorium Molten Salt Reactor with a Compact Thermal Neutron Generator

This patent application is for a process of nuclear power generation with ˜KW output by making the Thorium fuel of LiF+BeF+ThFin a Thorium Molten Salt Reactor (Th-MSR) to undergo fission along the thorium fuel cycle by providing thermal neutrons which were obtained by slowing down of fast neutrons from n external neutron generators with the help of graphite moderators carefully arranged inside the Th-MSR. 1. To maintain the Thorium fuel chain in Thorium Molten Salt Reactors (Th-MSR), neutrons from external neutron generators are supplied into the Th-MSR with Thorium fuel of LiF+BeF+ThFwithout any U-235 mixed into it, in contrast to the conventional method of mixing Uranium-235 in the form of UFmixed into the Thorium fuel to utilize neutrons emitted from the natural decay of U-235 as neutron source. There are three fissionable elements that can be used as nuclear fuel: Thorium, Uranium, and Plutonium. U-235 and Pu-239 spontaneously decay to emit neutrons which sustain the fuel cycle. Thorium-232 does not decay spontaneously. Therefore, it is necessary to supply neutrons to sustain the Thorium fuel chain.In 1965, ORNL mixed Th-232 and U-235 at 80:20 ratio in the Thorium Molten Salt, LiF—BeF—ThF(or UF), so that fast neutrons from the natural decay of uranium generated inside the reactor slowed down into thermal neutrons after passing through moderators carefully distributed and arranged inside the reactor to sustain the thorium fuel cycleand obtained ˜7 MW output until 1970. Then the Th-MSR Program at ORNL was terminated by Nixon Administration and classified as secret until 2005. The thorium fuel cycle develops in the following order: When Thorium-232 encounters a neutron, it becomes Thorium-233, Th-233 with a half-life of 22 minutes becomes Protactinium-233 after beta decay, Protactinium-233 with a half-life of 27 days becomes Uranium-233 after beta decay. When this Uranium-233 collides with a thermal neutron, it causes nuclear fission, splitting into two atoms, ...

IN-CONTAINMENT SPENT FUEL STORAGE TO LIMIT SPENT FUEL POOL WATER MAKEUP

A method and apparatus for extending the period a nuclear steam supply system spent fuel pool can be safely passively cooled by storing the spent fuel offloaded from the reactor, in the containment for one reactor operating cycle. During a refueling the spent fuel that is not to be returned to the reactor and the spent fuel that will be returned to the reactor are stored separately in shielded locations within the containment. After one operating cycle, the spent fuel stored within the containment that was not returned to the reactor just prior to the last operating cycle, is offloaded to the spent fuel pool and replaced by the newly offloaded spent fuel that is being retired. 1. A method of refueling a nuclear steam supply system having a nuclear reactor primary coolant loop enclosed within a hermetically sealed containment wherein the containment comprises a nuclear reactor vessel for supporting and housing a plurality of nuclear fuel assemblies within a core , the nuclear reactor vessel being supported within the containment as part of the nuclear reactor primary coolant loop; a refueling cavity extending above the nuclear reactor vessel within the containment; an in-containment refueling coolant storage tank supported within the containment outside the refueling cavity at an elevation above the core for , upon command , flooding at least a portion of the refueling cavity with a refueling coolant in furtherance of refueling the reactor vessel , the in-containment refueling coolant storage tank having a full level substantially at which a volume of the refueling coolant is maintained during normal reactor operation; and an irradiated nuclear fuel assembly storage tank supported within the containment below a portion of the refueling cavity , the irradiated nuclear fuel assembly storage tank is configured with fuel assembly storage racks for storing irradiated nuclear fuel within the containment outside the core when the reactor vessel is in operation and the ...

MUON-CATALYZED CONTROLLED FUSION ELECTRICITY-GENERATING APPARATUS AND METHOD

A turbine generator for producing electricity is described for use on planets and moons, or corresponding planetary or lunar orbits, where magnetic fields and atmospheres are sufficient low to obtain an adequate ambient flux of cosmic rays and muons for useful micro-fusion. A source of deuterium-containing micro-fusion particle fuel material is supplied via a flue to a columnar reaction volume, where it is dispersed and interacts with incoming cosmic rays and muons. Nuclear micro-fusion products (energetic alpha particles) drive a set of helium-wind turbines arranged around the reaction volume. Electrical generators coupled to the turbines generate electricity to supply nearby habitats and equipment. 1. A micro-fusion-driven turbine generator for producing electricity in the presence of ambient flux of cosmic rays and muons , comprising:a source of deuterium-containing micro-fusion particle fuel material;a reaction volume;a flue coupled to the source and reaction volume for dispersing fuel material into the reaction volume;a set of helium-wind turbines arranged around the reaction volume, wherein cosmic rays and muons entering the volume interact with the dispersed fuel material to cause nuclear micro-fusion events, kinetic-energy containing micro-fusion products driving the helium-wind turbines; anda set of electrical generators coupled to the respective helium-wind turbines to convert mechanical motion of the driven turbines into electricity.2. The generator as in claim 1 , wherein the deuterium-containing fuel material comprises LiD.3. The generator as in claim 1 , wherein the deuterium-containing fuel material comprises DO.4. The generator as in claim 1 , wherein the deuterium-containing fuel material comprises D.5. The generator as in claim 1 , wherein the deuterium-containing fuel material is in solid powder form.6. The generator as in claim 1 , wherein the deuterium-containing fuel material is in pellet or chip form.7. The generator as in claim 1 , wherein ...

Self-contained in-ground geothermal generator and heat exchanger with in-line pump and several alternative applications

A method of harnessing geothermal energy to produce electricity without polluting the environment by using universal portable closed loop systems is provided. The Scientific Geothermal Technology, The Self Contained In-Ground Geothermal Generator; The Self Contained Heat Exchanger; and The IN-LINE PUMP consist of several designs and variations complementing each other and/or operating separately in many different applications in energy sectors. The system can be used for harnessing heat from established lava (tube) flows; harnessing the waste heat from the flame on top of flare stacks; and other situation where a source of heat is difficult to access or is not suitable for relatively heavy equipment of a power plant or power unit. Also, included is an exemplary use for restoration of the Salton Sea which implements the Scientific Geothermal Technology for exchanging water from a salty terminal lake with oceanic water and for production of electricity and fresh water.

Energy storage system

Energy storage system regulating power output of a power generation plant that has a heat exchanger, primary circuit and secondary circuit, primary circuit directs primary fluid flow to components of a primary region and secondary circuit directs a secondary fluid flow to components of a secondary region, the heat exchanger is arranged so the secondary fluid flow is heated from the primary fluid flow. Energy storage arrangement makes a vessel for storing secondary fluid. Fluid transfer arrangement connects the vessel and is connectable to the heat exchanger of the power generation system to arrange the fluid transfer arrangement in fluid communication with the heat exchanger and the vessel. Bidirectional flow arrangement configured to control flow direction of fluid between the vessel and fluid transfer arrangement to selectively store heat energy from the heat exchanger in the vessel, and selectively transfer heat energy stored in the vessel to the heat exchanger.

HEAT TRANSFER METHODS FOR NUCLEAR PLANTS

A method of transferring heat from a nuclear plant may include: connecting a heat transfer system to the nuclear plant; and using the heat transfer system to transfer heat from the nuclear plant. The heat transfer system may include: a piping system that includes first and second connectors; a heat exchanger; a pump; and a power source. The heat transfer system may not be connected to the nuclear plant during normal plant power operations. The power source may be independent of a normal electrical power distribution system for the nuclear plant. The power source may be configured to power the pump. The piping system may be configured to connect the heat exchanger and pump. The first and second connectors may be configured to connect the heat transfer system to a fluid system of the nuclear plant. 117-. (canceled)18. A method of transferring heat from a nuclear plant , the method comprising:connecting a heat transfer system to the nuclear plant; andusing the heat transfer system to transfer heat from the nuclear plant; a piping system that includes first and second connectors;', 'a heat exchanger;', 'a pump; and', 'a power source;, 'wherein the heat transfer system compriseswherein the heat transfer system is not connected to the nuclear plant during normal plant power operations,wherein the power source is independent of a normal electrical power distribution system for the nuclear plant,wherein the power source is configured to power the pump,wherein the piping system is configured to connect the heat exchanger and pump,wherein the first and second connectors are configured to connect the heat transfer system to a fluid system of the nuclear plant, andwherein when the first and second connectors connect the heat transfer system to the fluid system of the nuclear plant, the heat transfer system is configured to receive fluid from the fluid system of the nuclear plant via the first connector, to pump the fluid through the heat exchanger, and to return the fluid to ...

REACTOR PRESSURE VESSEL INCLUDING PIPE RESTRAINT DEVICE, AND/OR PIPE RESTRAINT DEVICE

A reactor pressure vessel includes a reactor pressure vessel body, a nozzle structure connected to the reactor pressure vessel body, a conduit structure connected to the nozzle structure, and a restraint device attached around a portion of the conduit structure. The restraint device includes collar parts that have cross sections corresponding to respective segments of a periphery of the portion of the conduit structure, brackets attached to the nozzle structure, and rods connecting the brackets to the collar parts. The collar parts are connected end-to-end to each other such that a cross section of the collar parts connected to each other corresponds to the periphery of the portion of the conduit structure. The collar parts are pinned to each other. The brackets spaced apart from each other around a periphery of the nozzle structure. 118.-. (canceled)19. A method of attaching a restraint device to a conduit structure connected to a nozzle structure that includes brackets on an outer surface of the nozzle structure , the method comprising:inserting a first end of rods into the brackets such that a remaining part of each of the rods extends from the brackets over a portion of the conduit structure; the collar parts each including a side that defines a threaded hole,', 'the connecting collar parts to the rods including inserting a second end of each of the rods into a corresponding threaded hole among the threaded holes defined by the collar parts; and, 'connecting collar parts to the rods,'}pinning the collar parts to each other end-to-end such that the collar parts pinned to each other wrap around the portion of the conduit structure.20. The method of claim 19 , further comprising:inserting one or more Belleville washers between the portion of the conduit structure and at least one of the collar parts.21. The method of claim 19 , further comprising:inserting engineered-crush material between the portion of the conduit structure and at least one of the collar parts.22 ...

APPARATUS FOR EXPRESSING AND UTILIZING INTELLIGENT GENERAL ARRANGEMENT DRAWING OF ATOMIC POWER PLANT

Disclosed is an apparatus for providing an intelligent general arrangement (GA) drawing of an atomic power plant. The apparatus comprises an intelligent GA drawing utilization unit for providing additional information on an equipment included in a particular level of a particular building of the atomic power plant, on the basis of an intelligent GA drawing in response to an user's input. 1. An apparatus for providing an intelligent GA (General Arrangement) drawing of an atomic power plant , the apparatus comprising:a GA drawing storage unit configured to store at least one or more GA drawings of an image type for a specific level of a specific building of the atomic power plant;an intelligent GA drawing generation unit configured to generate an intelligent GA drawing in which one or more equipments included in the Image-type GA drawing are objectified to be able to be discriminated and information about equipments are stored in a database, on the basis of the Image-type GA drawing;an input unit configured to receive input for generation or utilization of the intelligent GA drawing from a user;an intelligent GA drawing utilization unit configured to provide additional information about equipments included in a specific level of a specific building of the atomic power plant on the basis of the intelligent GA drawing in response to input about utilization of the user; andan output unit configured to display the additional information such that the user can recognize the additional information.2. The apparatus of claim 1 , wherein the intelligent GA drawing generation unit includes:an object converter configured to recognize a figure of a first equipment displayed in a figure section of the image-type GA drawing and to designate a first ID to the recognized first equipment, and to generate information of a text type by recognizing information about the first equipment included in an equipment information section of the image-type GA drawing; anda data processor ...

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

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 , ...

PASSIVE ELECTRICAL COMPONENT FOR SAFETY SYSTEM SHUTDOWN USING AMPERE'S LAW

An electro-technical device includes a circuit including a coil connected to a voltage source for receiving a predetermined current therefrom and connected to an output device. The circuit includes a breakable junction and a photodiode for receiving a light signal from a fiber optic cable. The photodiode receives a light signal from a sensor. A permanent magnet includes a pole end opposing a common pole end of the coil, wherein when the coil receives an increased current from the photodiode, the coil creates an magnetic flux that repels against the common pole of the permanent magnet in order to cause the breakable junction to break and disrupt a connection between the voltage source and the output device. 1. An electro-technical device , comprising:a circuit including a coil connected to a voltage source for receiving a predetermined current therefrom and connected to an output device;the circuit including a breakable junction;the circuit including a photodiode for receiving a light signal from a fiber optic cable receiving a light signal from a sensor; anda permanent magnet having a pole end opposing a common pole end of the coil, wherein when the coil receives an increased current from the photodiode, the coil creates an magnetic flux that repels against the common pole of the permanent magnet in order to cause the breakable junction to break and disrupt a connection between the voltage source and the output device.2. The electro-technical device according to claim 1 , wherein the breakable junction is disposed between the permanent magnet and the coil.3. The electro-technical device according to claim 1 , wherein the breakable junction is made by 3-D printing.4. The electro-technical device according to claim 1 , wherein the sensor includes one of a temperature sensor claim 1 , a pressure sensor claim 1 , or a flow sensor.5. An electro-technical device claim 1 , comprising:a plurality of circuits each including a coil connected to a voltage source for receiving ...

EMERGENCY AND BACK-UP COOLING OF NUCLEAR FUEL AND REACTORS AND FIRE-EXTINGUISHING, EXPLOSION PREVENTION USING LIQUID NITROGEN

A nuclear reactor chamber comprises an inlet portion. The chamber is a part of a nuclear power plant. At least one container contains liquid nitrogen and cold nitrogen vapor and includes an outlet portion. At least one thermally activated release mechanism is respectively connected between one of the at least one container and the inlet portion. Each thermally activated release mechanism is configured to release the liquid nitrogen from a connected container into the inlet portion when a predetermined safety threshold temperature is reached, so that the released liquid nitrogen produces an expanding volume of cold nitrogen vapor within the nuclear reactor chamber. 120-. (canceled)21. A system , comprising:a nuclear reactor chamber comprising an inlet portion, wherein the chamber is a part of a nuclear power plant;at least one container, wherein each respective container of the at least one container contains liquid nitrogen and cold nitrogen vapor and wherein each respective container includes an outlet portion; and,at least one thermally activated release mechanism, wherein each thermally activated release mechanisms of the at least one thermally activated release mechanisms is respectively connected between one of the at least one container and the inlet portion of the nuclear reactor chamber, each thermally activated release mechanism being configured to release the liquid nitrogen from a connected container into the inlet portion when a predetermined safety threshold temperature is reached, so that the released liquid nitrogen produces an expanding volume of cold nitrogen vapor within the nuclear reactor chamber.22. A system as in claim 21 , additionally comprising:a central storage container that stores liquid nitrogen, the central storage container being connected to each of the at least one container, wherein the at least one container can each be independently removed from connection with the central storage container, the central storage container being ...

INTEGRATED SYSTEM FOR CONVERTING NUCLEAR ENERGY INTO ELECTRICAL, MECHANICAL, AND THERMAL ENERGY AND METHODS FOR USING THE SAME

Provided is an apparatus for generating electricity, mechanical energy, and/or process and district heat using a gas propellant chamber fueled with fissile material and enclosed in a sealed containment vessel which also contains an operating gas. The system allows for the operating gas to be compressed as it enters the nuclear fuel chamber where it is heated. As the operating gas exits the nuclear fuel chamber, the kinetic energy of the gas is converted to rotational energy by a variety of methods. The rotational energy is further converted to electricity, mechanical energy, and/or process and district heat. The operating gas circulates in the containment vessel and is cooled prior to re-entering the gas propellant chamber. The apparatus thereby provides a simpler and safer design that is both scalable and adaptable. The apparatus is easily and safely transportable and can be designed to be highly nuclear-proliferation-resistant. 1. An apparatus for generating electricity comprising:a gas propellant chamber comprised of an annular body defining first and second ends, the first end of the annular body defining an inlet assembly that is configured to draw operating gas into the gas propellant chamber and the second end defining an exhaust assembly that is configured to expel operating gas from the gas propellant chamber;a nuclear fuel chamber positioned within the annular body of the gas propellant chamber between the first and second ends, the nuclear fuel chamber configured to heat the operating gas;a compressor positioned proximate the first end of the gas propellant chamber, the compressor configured to compress the operating gas prior to entry into the nuclear fuel chamber;a conversion apparatus positioned proximate the second end of the gas propellant chamber, the conversion apparatus configured to convert kinetic energy of the operating gas exiting the nuclear fuel chamber into rotational energy; anda drive shaft extending axially through the gas propellant ...

PIPE RESTRAINT AND SHIELD

A pipe restraint assembly includes a restraint body configured to be removably attached to a portion of pipe. The portion of pipe is associated with a postulated pipe failure associated with a release of high pressure fluid. A plurality of apertures penetrate through the restraint body and are positioned proximate to a location of the postulated pipe failure. The apertures are configured to provide a number of passageways for the fluid to exit from the location of the postulated pipe failure and to be released outside of the restraint body. One or more restraint devices maintain the position of the apertures relative to the location of the postulated pipe failure. 1. A pipe restraint assembly , comprising:a restraint body configured to be removably attached to a portion of pipe, wherein the portion of pipe is associated with a postulated pipe failure associated with a release of high pressure fluid;a plurality of apertures that penetrate through the restraint body and are positioned proximate to a location of the postulated pipe failure, wherein the apertures are configured to provide a number of passageways for the fluid to exit from the location of the postulated pipe failure and be released outside of the restraint body; andone or more pipe restraints, wherein the one or more pipe restraints are configured to maintain the position of the apertures relative to the location of the postulated pipe failure.2. The pipe restraint assembly of claim 1 , wherein the one or more pipe restraints comprise a pipe protrusion that extends from an exterior surface of the pipe and into the restraint body.3. The pipe restraint assembly of claim 2 , wherein the pipe protrusion comprises a stud that is welded to the exterior surface of the pipe claim 2 , and wherein the stud extends into one of the apertures.4. The pipe restraint assembly of claim 2 , wherein the pipe protrusion comprises a ridge formed along at least a portion of the circumference of the pipe claim 2 , wherein the ...

Alternative air supply and exhaust port for air-operated valve

The present invention is directed to remote operation of an operation valve such as an air operated valve even at the time of power loss. A gas supply apparatus of the present invention includes: an operation valve mounted in some midpoint of a piping for passing at least gas in a plant and operating a valve body by the gas flowing in the piping; an solenoid valve mounted in some midpoint of the piping and allowing/stopping flow of the gas to the operation valve; and a gas supply source for supplying gas to the solenoid valve. A switching valve for switching between exhaust from the solenoid valve and gas supply to the solenoid valve is mounted in an exhaust line of the solenoid valve and, at the time of power loss, the switching valve is switched to connection to the gas supply source for supplying gas to the solenoid valve.

SYSTEM USABLE IN NUCLEAR ENVIRONMENT FOR PROVIDING BREATHING GAS

A system usable in a nuclear environment provides a reservoir of liquefied breathable gas in fluid communication with a deployment system. The deployment system uses a stream of the breathable gas from the reservoir to operate a gas turbine which runs an electrical generator that is mechanically connected therewith to generate electrical power that is stored in a battery bank. The stream of breathable gas then flows from the turbine and is split between a heat exchanger that is situated in heat exchange relation with the interior region of the main control room and an outlet that provides breathable gas to the control room. The portion of the stream that flows through the heat exchanger cools the main control room. The other portion of the stream that provides breathable gas to the main control room also recirculates the atmosphere in the control room 1. A system structured for use in conjunction with an interior region of a nuclear environment , the interior region being at least partially enclosed , the system comprising:a reservoir having stored therein breathable gas whose stored condition is at least one of at a temperature less than the ambient temperature at the exterior of the reservoir and at a pressure greater than the ambient pressure at the exterior of the reservoir, the reservoir being structured to output a stream of the breathable gas responsive to a command;a control apparatus connected with the reservoir and structured to provide the command to the reservoir in a predetermined situation; a heat exchanger comprising a number of flow channels that are in fluid communication with the reservoir and that are situated in heat exchange relation with the interior region, the number of flow channels being structured to receive therethrough at least a portion of the stream and to transfer to it heat from within the interior region, and', 'a generation apparatus comprising a turbine and an electrical generator that are mechanically connected together, the ...

Gas turbine and pressurized water reactor steam turbine combined circulation system

Disclosed is a gas turbine and pressurized water reactor steam turbine combined circulation system, using a heavy duty gas turbine and a pressurized water reactor steam turbine to form a combined circulation system. Heat of the tail gas of the gas turbine is utilized to raise the temperature of a secondary circuit main steam from 272.8° C., and the temperature of the secondary circuit main steam slides between 272.8° C. and 630° C. according to different pressurized water reactor steam yields and different input numbers and loads of the heavy duty gas turbine. The system has a higher heat efficiency than that of the pressurized water reactor steam turbines in the prior art; and as for the electric quantity additionally generated by gas, the heat efficiency of the system is also significantly higher than that of gas-steam combined circulation in the prior art.

Asteroid mining systems facilitated by cosmic ray and muon-catalyzed fusion

Cosmic ray and muon-catalyzed micro-fusion electrical generation provides electrical power for mining operations, including any asteroid habitats and mining equipment. The micro-fusion generator systems deploy deuterium-containing fuel material as a localized cloud interacting with incoming ambient cosmic rays to generate energetic fusion products. Dust or other particulate matter in the fuel material, in the localized cloud, and in the space surrounding the asteroid being mined converts some cosmic rays into muons that also catalyze fusion. The fusion products drive turbines to facilitate the electrical generation.

MULTI-MODULAR POWER PLANT WITH DEDICATED ELECTRICAL GRID

A multi-modular power plant includes a plurality of on-site nuclear power modules that generate a power plant output, and a number of power plant systems which operate using electricity associated with a house load of the power plant. A switchyard associated with the power plant may electrically connect the power plant to a distributed electrical grid. The distributed electrical grid may be configured to service a plurality of geographically distributed consumers. Additionally, the switchyard may electrically connect the power plant to a dedicated electrical grid. The dedicated electrical grid may provide electricity generated from the power plant output to a dedicated service load, and the power plant output may be equal to or greater than a combined load of the dedicated service load and the house load. At least a portion of the power plant output may be distributed to both the power plant systems and the dedicated electrical grid. 1. A multi-modular power plant , comprising:a plurality of on-site nuclear power modules configured to generate a power plant output, a number of power plant systems which are configured to operate using electricity associated with a house load of the power plant; and electrically connect the power plant to a distributed electrical grid, wherein the distributed electrical grid is configured to service a plurality of geographically distributed consumers;', 'electrically connect the power plant to a dedicated electrical grid, wherein the dedicated electrical grid is configured to provide electricity generated from the power plant output to a dedicated service load, and wherein the power plant output is equal to or greater than a combined load of the dedicated service load and the house load; and', 'distribute at least a portion of the power plant output to both the power plant systems and the dedicated electrical grid., 'a switchyard configured to2. The multi-modular power plant of claim 1 , wherein the dedicated service load comprises ...

MOBILE BORATION SYSTEM

A mobile boration system () has a number of components that are mobile and include a water source (), HBOpowder supply (), a mixer to mix the solution () capable of providing a boric acid solution () with minimal air entrainment and optional heat exchanger(s) (), and wherein the system () is capable of transport to a nuclear power plant facility by land, sea or air, rather than being in place in a large vulnerable footprint. 1. A mobile boration apparatus capable of providing nuclear reactor systems with borated coolant that can mix components on-site , to provide borated water , the mobile apparatus comprising:a) a mobile transportation means comprising;b) a connection to a water source;{'sub': 2', '3, 'c) a HBOpowder or other water soluble boron source;'}d) a heater to heat the water;e) a pump or other motive force to move water to a desired location;{'sub': 2', '3, 'f) a mixer configured to provide metered mixing of the water and HBOpowder or other water soluble boron source to generate a preselected, metered appropriate concentration of initial water/boric acid slurry, which slurry during continued mixing provides a relatively hot borated/boric acid water solution;'}g) a dilution tank;h) a hot stream conduit connected between the mixer and the dilution tank for conveying the relatively hot borated/boric acid water solution to the dilution tank;i) a dilution stream conduit connecting the dilution tank to a relative cool dilution stream compared to the relatively hot borated/boric acid water solution; andj) a diluted boric acid output connected to an exit port on the dilution tank for conveying a mixture of the dilution stream and the hot borated/boric acid water solution to a nuclear reactor system.2. The mobile boration apparatus of claim 1 , wherein the heater heats the water/boric acid slurry in the mixer.3. The mobile boration apparatus of claim 1 , wherein the pump of (e) is selected from the group consisting of a positive displacement pump and a centrifugal ...

Source of electricity derived from a spent fuel cask

Apparatus for extracting useful electric or mechanical power in significant quantities from the decay heat that is produced within spent nuclear fuel casks. The power is used for either powering an active forced air heat removal system for the nuclear casks, thereby increasing the thermal capacity of the casks, or for emergency nuclear plant power in the event of a station blackout. Thermoelectric generators or other heat engines are employed using the thermal gradient that exists between the spent nuclear fuel and the environment surrounding the cask's components housing the nuclear fuel to produce the power.

SYSTEM FOR ABATEMENT OF NOXIOUS EMISSIONS IN THE ATMOSPHERE FROM AN INDUSTRIAL OR NUCLEAR POWER PLANT

A system for the abatement of noxious emissions from an industrial or nuclear power plant or the like in the event of accident includes a structure for impermeabilization of the ground, which extends at least in an annular area that surrounds the plant, a plurality of water-sprinkling towers, which are arranged around the plant and sprinkle water in the atmosphere, preferably added with chemical and/or biological and/or mineral substances, and a peripheral collection structure, configured for receiving water withheld by the impermeabilization structure. 11. A system for the abatement of noxious emissions in the atmosphere from an industrial or nuclear plant () in the event of accident , comprising:{'b': 10', '10', '1, 'a structure for impermeabilization of the ground (), the impermeabilization structure () extending at least in an annular area (A1) that surrounds the plant ();'}{'b': 20', '22', '1, 'a plurality of sprinkling towers (-), which are arranged around the plant () and/or in the precincts thereof and operative to sprinkle water in the atmosphere, preferably added with chemical and/or biological and/or mineral substances; and'}{'b': 50', '10, 'a peripheral collection structure (), configured for receiving water withheld by the impermeabilization structure ().'}22021122120221. The system according to claim 1 , wherein the plurality of towers comprises at least one between a series of sprinkling towers ( claim 1 , ) arranged in the precincts of the plant () and/or on its bounds and a series of sprinkling towers () arranged in the annular area (A1) that surrounds the plant () claim 1 , the towers (-) being preferably configured for sprinkling water to a height greater than that of structures of the plant ().32021313240413132404131202230. The system according to or claim 1 , wherein each tower (-) is provided with means ( claim 1 , claim 1 , claim 1 , ) for supply of the water to be sprinkled claim 1 , the supply means ( claim 1 , claim 1 , claim 1 , ) ...

REACTOR PRESSURE VESSEL INCLUDING PIPE RESTRAINT DEVICE, AND/OR A PIPE RESTRAINT DEVICE

A reactor pressure vessel includes a reactor pressure vessel body, a nozzle structure connected to the reactor pressure vessel body, a conduit structure connected to the nozzle structure, and a restraint device attached around a portion of the conduit structure. The restraint device includes collar parts that have cross sections corresponding to respective segments of a periphery of the portion of the conduit structure, brackets attached to the nozzle structure, and rods connecting the brackets to the collar parts. The collar parts are connected end-to-end to each other such that a cross section of the collar parts connected to each other corresponds to the periphery of the portion of the conduit structure. The collar parts are pinned to each other. The brackets spaced apart from each other around a periphery of the nozzle structure. 1. A reactor pressure vessel comprising:a reactor pressure vessel body;a nozzle structure connected to the reactor pressure vessel body;a conduit structure connected to the nozzle structure; and collar parts that have cross sections corresponding to respective segments of a periphery of the portion of the conduit structure,', 'the collar parts being connected end-to-end to each other such that a cross section of the collar parts connected to each other corresponds to the periphery of the portion of the conduit structure,', 'the collar parts being pinned to each other,', 'brackets attached to the nozzle structure, the brackets spaced apart from each other around a periphery of the nozzle structure, and, 'a restraint device attached around a portion of the conduit structure, the restraint device including,'}rods connecting the brackets to the collar parts.2. The reactor pressure vessel of claim 1 , wherein the conduit structure is a pipe.3. The reactor pressure vessel of claim 2 , whereinthe portion of the conduit structure has an outer diameter that is greater than an outer diameter of a different location of the pipe, andthe portion of ...

System Usable In Nuclear Environment For Providing Breathing Gas

A system usable in a nuclear environment provides a reservoir of liquefied breathable gas in fluid communication with a deployment system. The deployment system uses a stream of the breathable gas from the reservoir to operate a gas turbine which runs an electrical generator that is mechanically connected therewith to generate electrical power that is stored in a battery bank. The stream of breathable gas then flows from the turbine and is split between a heat exchanger that is situated in heat exchange relation with the interior region of the main control room and an outlet that provides breathable gas to the control room. The portion of the stream that flows through the heat exchanger cools the main control room. The other portion of the stream that provides breathable gas to the main control room also recirculates the atmosphere in the control room 115-. (canceled)16. A system structured for use in conjunction with an interior region of a nuclear environment , the interior region being at least partially enclosed , the system comprising:a reservoir having stored therein breathable gas whose stored condition is at least one of at a temperature less than the ambient temperature at the exterior of the reservoir and at a pressure greater than the ambient pressure at the exterior of the reservoir, the reservoir being structured to output a stream of the breathable gas responsive to a command;a control apparatus connected with the reservoir and structured to provide the command to the reservoir in a predetermined situation; a heat exchanger comprising a number of flow channels that are in fluid communication with the reservoir and that are situated in heat exchange relation with the interior region, the number of flow channels being structured to receive therethrough at least a portion of the stream and to transfer to it heat from within the interior region, and', 'a generation apparatus comprising a turbine and an electrical generator that are mechanically connected ...

MAIN STREAM FOR REDUCING RELEASE OF RADIOACTIVE MATERIAL TO ATMOSPHERE UNDER SEVERE ACCIDENT

Disclosed herein is a nuclear power plant main steam system, which reduces the atmospheric discharge of radioactive materials generated in an accident, the system including: a decontamination water tank containing decontamination water; and a connection pipe for connecting the decontamination water tank from a main steam pipe which connects a steam generator and a turbine, wherein the connection pipe is connected to the decontamination water tank through a main steam safety valve or a connection valve, wherein the main steam safety valve or the connection valve is configured by a three-way valve and is configured to discharge the generated steam to the air when an accident occurs within a design basis and to transfer the generated steam to the decontamination water tank when a severe accident occurs. A main steam system according to the present invention has an effect of reducing discharge of radioactive materials to the air when a containment bypass accident including a steam generator tube rupture caused by high-temperature steam occurs. 1. A nuclear power plant main steam system , which reduces the atmospheric discharge of radioactive materials generated in an accident , the system comprising:a decontamination water tank containing decontamination water; anda connection pipe for connecting the decontamination water tank from a main steam pipe which connects a steam generator and a turbine,wherein the connection pipe is connected to the decontamination water tank through a main steam safety valve or connection valve,wherein the main steam safety valve or the connection valve is configured by a three-way valve, and is configured to discharge the generated steam to the air when an accident occurs within a design basis, and to transfer the generated steam to the decontamination water tank when a severe accident occurs.2. The nuclear power plant main steam system as set forth in claim 1 , wherein the connection valve are located at one or more positions claim 1 , ...

Corrosion and Wear Resistant Overlay, Method for Forming Corrosion and Wear Resistant Overlay, and Corrosion and Wear Resistant Valve

The present invention is intended is to improve the corrosion resistance of an overlay used in a nuclear power plant, and to reduce dissolution of cobalt from an overlay. The corrosion and wear resistant overlay is formed along a surface of a base by laser lamination modeling, and is configured from a plurality of metal layers of a Co-base alloy. The thickness of carbide eutectics that precipitate in the metal layers is the largest in the metal layer closest to the base, and is gradually smaller in the metal layers farther away from the base. The intensity of the laser beam applied to form layers by laser lamination modeling is adjusted so that the carbide eutectics that precipitate in at least the outermost metal layer have a controlled size of 10 μm or less. 1. A corrosion and wear resistant overlay , comprising:a plurality of Co-base alloy layers formed along a surface of a base;wherein a carbide eutectic that precipitates at a boundary between crystals of a dendrite structure of the Co-base alloy in the plurality of layers has an average thickness that is thicker in layers closer to the base, or thinner in layers farther away from the base.2. The corrosion and wear resistant overlay according to claim 1 , wherein the carbide eutectic that precipitates in at least the outermost layer of the plurality of layers has a maximum thickness of about 10 μm or less.3. The corrosion and wear resistant overlay according to claim 2 , wherein the crystals of the dendrite structure claim 2 , and the carbide eutectic in the outermost layer have a Cr concentration of about 17% or more.4. The corrosion and wear resistant overlay according to claim 1 , wherein the plurality of layers is each formed by heating claim 1 , melting claim 1 , and depositing a Co-base alloy powder discharged through a nozzle.5. The corrosion and wear resistant overlay according to claim 1 , wherein the plurality of layers is each formed by heating and melting a Co-base alloy powder deposited beforehand.6 ...

FACILITY FOR REDUCING RADIOACTIVE MATERIAL AND NUCLEAR POWER PLANT HAVING THE SAME

The present invention provides a facility for reducing radioactive material comprising: a cooling water storage unit installed inside a containment and formed to store cooling water; a boundary unit forming a boundary of radioactive material inside the containment and surrounding a reactor coolant system installed inside the containment to prevent a radioactive material from releasing from the reactor coolant system or a pipe connected with the reactor coolant system to the containment; a connecting pipe connected with an inner space of the boundary unit and the cooling water storage unit to guide a flow of a fluid caused by a pressure difference between the boundary unit and the cooling water storage unit from the boundary unit to the cooling water storage unit; and a sparging unit disposed to be submerged in the cooling water stored in the cooling water storage unit and connected with the connecting pipe to sparge the fluid that has passed through the connecting pipe and the radioactive material contained in the fluid to the cooling water storage unit. 1. A facility for reducing radioactive material , the facility comprising:a cooling water storage unit installed inside a containment and formed to store cooling water;a boundary unit forming a boundary of radioactive material inside the containment and surrounding a reactor coolant system installed inside the containment to prevent a radioactive material from releasing from the reactor coolant system or a pipe connected with the reactor coolant system to the containment;a connecting pipe connected with an inner space of the boundary unit and the cooling water storage unit to guide a flow of a fluid caused by a pressure difference between the boundary unit and the cooling water storage unit from the boundary unit to the cooling water storage unit; anda sparging unit disposed to be submerged in the cooling water stored in the cooling water storage unit and connected with the connecting pipe to sparge the fluid that ...

Small modular reactor power plant with load following and cogeneration capabilities and methods of using

Provided herein is a small modular nuclear reactor plant that can comprise a reactor core comprising a primary sodium comprising cool primary sodium flow and heated primary sodium flow. Heated primary sodium flow can enter one or more IHXs where heated primary sodium exchanges heat with secondary sodium flowing through at least one intermediate sodium loop. Intermediate sodium loop can comprise secondary sodium flow that can transport heat to energy conversion portion via a heat exchanger. Energy conversion portion can comprise a bypass valve. Bypass valve can bypass an energy conversion working fluid (such as S-CO2) away from a turbine during periods of adjustment as discussed herein. The plant may comprise passive load following features along with the ability to provide cogeneration heat.

STEAM GENERATOR

A steam generator comprising a vessel having an inlet and an outlet, and in use a primary fluid flow enters the vessel through the inlet and exits the vessel through the outlet. A plurality of modules are connected in series and at least partially housed within the vessel, and each module comprises at least one tube. The modules are arranged such that at least one tube of one module is coaxial with at least one tube of an adjacent module so as to define a conduit through which a secondary fluid can flow from one module to an adjacent module. 1. A steam generator comprising:a vessel having an inlet and an outlet, in use, a primary fluid flow enters the vessel through the inlet and exits the vessel through the outlet; anda plurality of modules connected in series and at least partially housed within the vessel, wherein each module comprises at least one tube and the modules are arranged such that at least one tube of one module is coaxial with at least one tube of an adjacent module so as to define a conduit through which a secondary fluid can flow from one module to an adjacent module.2. The steam generator according to claim 1 , wherein each module comprises a plurality of tubes and the modules are arranged so that each of the plurality of tubes is coaxial with one of the tubes of an adjacent module so as to form a conduit bundle comprising a plurality of conduits.3. The steam generator according to claim 2 , wherein each module comprises a flange plate positioned at an axial end of the at least one tube claim 2 , and wherein adjacent modules are connected together via the flange plates.4. The steam generator according to claim 3 , wherein each flange plate comprises a plurality of holes claim 3 , each hole receiving one of the plurality of tubes:5. The steam generator according to claim 3 , comprising an intermediate plate positioned between adjacent flange plates of adjacent modules claim 3 , the intermediate plate comprising one or more holes for receiving the ...

HYBRID SAFETY INJECTION TANK SYSTEM PRESSURIZED WITH SAFETY VALVE OF PRESSURIZER

A hybrid safety injection tank system. The system is pressurized with a safety valve of a pressurizer, which functions as a low pressure safety injection tank and as a high pressure core makeup tank of a nuclear reactor emergency core cooling system. The safety valve is configured to be automatically operated in response to a pressure difference and is installed on a pressure equalization pipe that can realize pressure equalization between the low pressure safety injection tank and the high pressure pressurizer in the event of the nuclear power plant station blackout and power outage. 1. A hybrid safety injection tank system pressurized with a safety valve of a pressurizer , comprising:an emergency core cooling water safety injection tank (SIT) charged both with cooling water and with nitrogen gas for cooling a nuclear reactor system;a pressurizer for supplying high pressure steam to pressurize the safety injection tank;a pressure equalization pipe connecting the safety injection tank to the pressurizer so as to realize pressure equalization between the safety injection tank and the pressurizer;a pressure equalization pipe isolation valve installed on the pressure equalization pipe so as to isolate the safety injection tank from the pressurizer;a pressure equalization pipe check valve installed on the pressure equalization pipe in series with the pressure equalization pipe isolation valve so as to prevent a backflow from the safety injection tank to the pressurizer; anda safety valve installed on the pressure equalization pipe in parallel both with the pressure equalization pipe isolation valve and with the pressure equalization pipe check valve so as to isolate the safety injection tank from the pressurizer.2. The hybrid safety injection tank system pressurized with the safety valve of the pressurizer as set forth in claim 1 , further comprising:an emergency core cooling water injection pipe connecting the safety injection tank to the nuclear reactor system;a ...

Small nuclear power generator

Provided is a small nuclear power generator which restores steam to water by applying pressure to the inside of a condenser using a pressurizer disposed over the condenser without condensing steam using cooling water. The small nuclear power generator includes: a nuclear reactor generating high-temperature heat by nuclear fission of a nuclear fuel; a steam generator converting internal water into steam by the high-temperature heat generated in the nuclear reactor; a turbine/generator including a steam turbine rotated by steam generated in the steam generator and a generator connected to an axis of the steam turbine and together rotating to produce electricity; and a condenser restoring steam to water by applying pressure to steam discharged after rotating the steam turbine using two or more pressurizers, again supplying the water into the steam generator, and formed of a titanium (Ti) or an alloy thereof.

POWER CONVERSION SYSTEM FOR NUCLEAR POWER GENERATORS AND RELATED METHODS

Various exemplary embodiments of a power conversion system for converting thermal energy from a heat source to electricity are disclosed. In one exemplary embodiment, the power conversion system may include a substantially sealed chamber having an inner shroud having an inlet and an outlet and defining an internal passageway between the inlet and the outlet through which a working fluid passes. The sealed chamber may also include an outer shroud substantially surrounding the inner shroud, such that the working fluid exiting the outlet of the inner shroud returns to the inlet of the inner shroud in a closed-loop via a return passageway formed between an external surface of the inner shroud and an internal surface of the outer shroud. The power conversion system may further include a source heat exchanger disposed in the internal passageway of the inner shroud, the source heat exchanger being configured to at least partially receive a heat transmitting element. 1. A power conversion system for converting thermal energy from a heat source to electricity , comprising: an inner shroud having an inlet and an outlet and defining an internal passageway between the inlet and the outlet through which a working fluid passes; and', 'an outer shroud substantially surrounding the inner shroud, such that the working fluid exiting the outlet of the inner shroud returns to the inlet of the inner shroud in a closed-loop via a return passageway formed between an external surface of the inner shroud and an internal surface of the outer shroud;', 'a source heat exchanger disposed in the internal passageway of the inner shroud, the source heat exchanger being configured to at least partially receive a heat transmitting element associated with the heat source external to the substantially sealed chamber, the source heat exchanger being further configured to transfer heat energy from the heat transmitting element to the working fluid passing through the source heat exchanger;, 'a ...

EMERGENCY CORE COOLING SYSTEM AND BOILING WATER REACTOR PLANT USING THE SAME

According to an embodiment, an emergency core cooling system has: three active safety divisions each including only one motor-driven active safety system; one passive safety division including a passive safety system; an emergency power source disposed in each of the active safety divisions to supply electric power to the motor-driven active safety system; and an advanced passive containment cooling system disposed in the passive safety division. Only two active safety divisions each includes a low pressure flooder system that is commonly used with a residual heat removal system as the only one motor-driven active safety system. The other active safety division includes an air-cooled injection system as the only one motor-driven active safety system. 1. An emergency core cooling system for a boiling water reactor plant , the plant including:a reactor pressure vessel containing a core a dry well containing the reactor pressure vessel,', 'a wet well containing a suppression pool in a lower part thereof, and a wet well gas phase in an upper part thereof,', 'a LOCA vent pipe connecting the dry well and the suppression pool,', 'an outer well disposed outside of the dry well and the wet well, adjacent to the dry well via a dry well common wall, and adjacent to the wet well via a wet well common wall, and', 'a scrubbing pool storing water, disposed in the outer well,, 'a containment vessel havingthe emergency core cooling system comprising:at least three active safety divisions each including only one motor-driven active safety system;at least one passive safety division each including a passive safety system that does not require any electric motors;an emergency power source disposed in each of the active safety divisions to supply electric power to the motor-driven active safety system; andan advanced passive containment cooling system disposed in the passive safety division including a gas vent pipe, leading end of the gas vent pipe being submerged in water in the ...

SYSTEM AND METHOD FOR REDUCING ATMOSPHERIC RELEASE OF RADIOACTIVE MATERIALS CAUSED BY SEVERE ACCIDENT

Provided are a system and method for reducing the atmospheric release of radioactive materials caused by a severe accident. The system includes a steam generator disposed in a containment building, configured to generate steam by using heat of a coolant heated in a nuclear reactor, and connected to a turbine through a main steam line, a decontamination tank connected to the main steam line through a connection line and containing decontamination water for decontaminating the steam delivered from the steam generator and reducing atmospheric release of radioactive materials when a severe accident occurs, and a depressurizing power generation unit disposed on the connection line and configured to generate emergency power while depressurizing the steam delivered from the steam generator toward the decontamination tank when the severe accident occurs. 1. A system for reducing the atmospheric release of radioactive materials caused by a severe accident , the system comprising:a steam generator disposed in a containment building, configured to generate steam by using heat of a coolant heated in a nuclear reactor, and connected to a turbine through a main steam line;a decontamination tank connected to the main steam line through a connection line and containing decontamination water for decontaminating the steam delivered from the steam generator and reducing atmospheric release of radioactive materials when a severe accident occurs; anda depressurizing power generation unit disposed on the connection line and configured to generate emergency power while depressurizing the steam delivered from the steam generator toward the decontamination tank when the severe accident occurs.2. The system of claim 1 , wherein the depressurizing power generation unit comprises:a turbine including a plurality of blades rotated by the steam delivered from the steam generator; anda power generator configured to generate power through rotation of the turbine.3. The system of claim 1 , wherein ...

FLOATING NUCLEAR REACTOR WITH STABILIZATION ASSEMBLIES

A protection system is provided for protecting a nuclear reactor positioned on a barge which is floating in the water of a tank. The system also includes suspension systems which permits the barge to move downwardly in the tank upon an aircraft, missile strike or earthquake to reduce the impact force of the strike. Each of the suspension systems includes a slack upper chain member, a taut intermediate chain member and a slack lower chain member. A padding material is positioned at the inner sides of the tank. Padding material may be placed of the ends and sides of the barge. 1. A floating nuclear reactor , comprising: (a) a bottom wall having a first end, a second end, a first side and a second side;', '(b) a first end wall, having a first side, a second side, a lower end and an upper end, extending upwardly from said first end of said bottom wall;', '(c) a second end wall, having a first side, a second side, a lower end and an upper end, extending upwardly from said second end of said bottom wall;', '(d) a first side wall, having a first end, a second end, a lower end and an upper end, extending between said first ends of said first and second end walls;', '(e) a second side wall, having a first end, a second end, a lower end and an upper end, extending between said second ends of said first and second end walls;, 'a tank having water therein which includes;'}each of said first end wall, said second end wall, said first side wall and said second side wall of said tank having inner and outer sides;a barge, having a first end, a second end, a first side and a second side, floatably positioned in said tank;an upstanding nuclear reactor positioned on said barge;said nuclear reactor having sides;a plurality of suspension assemblies connecting said barge to said tank;said plurality of suspension assemblies permitting said barge to move upwardly and downwardly with respect to said tank; (a) a vertically disposed guide track, having upper and lower ends, mounted on said ...

FLOATING NUCLEAR REACTOR

A nuclear reactor is positioned on a barge which is floating in a water tank. A plurality of counter weight assemblies interconnect the barge with the tank to create a lifting force to the barge and to maintain the barge in a level position. Structure is also included for limiting horizontal movement of the counter weight of the counter weight assemblies. 1. A floating nuclear reactor , comprising: (a) a bottom wall having a first end, a second end, a first side and a second side;', '(b) a first end wall, having a first side, a second side, a lower end and an upper end, extending upwardly from said first end of said bottom wall;', '(c) a second end wall, having a first side, a second side, a lower end and an upper end, extending upwardly from said second end of said bottom wall;', '(d) a first side wall, having a first end, a second end, a lower end and an upper end, extending between said first ends of said first and second end walls;', '(e) a second side wall, having a first end, a second end, a lower end and an upper end, extending between said second ends of said first and second end walls;, 'a tank having water therein with the tank including;'}each of said first end wall, said second end wall, said first side wall and said second side wall of said tank having inner and outer sides;a barge, having a first end, a second end, a first side and a second side, floatably positioned in said tank;a nuclear reactor positioned on said barge; anda plurality of counter weight assemblies operatively connected to said tank and said barge which are configured to provide a lifting force on said barge thereby increasing the buoyancy of said barge.2. The floating nuclear reactor of wherein each of said counter weight assemblies includes:(a) a pulley rotatably secured to said tank about a horizontal axis;(b) an elongated flexible cable having first and second ends;(c) said first end of said cable being secured to said barge;(d) said cable extending from said barge over said ...

Corrosion and Wear Resistant Overlay, Method for Forming Corrosion and Wear Resistant Overlay, and Corrosion and Wear Resistant Valve

Intended is to improve the corrosion resistance of an overlay used in a nuclear power plant, and to reduce dissolution of cobalt from an overlay. The corrosion and wear resistant overlay is formed along a surface of a base by laser lamination modeling, and is configured from a plurality of metal layers , and of a Co-base alloy. The thickness of carbide eutectics that precipitate in the metal layers , and is the largest in the metal layer closest to the base , and is gradually smaller in the metal layers , and farther away from the base . The intensity of the laser beam applied to form layers by laser lamination modeling is adjusted so that the carbide eutectics that precipitate in at least the outermost metal layer have a controlled size of 10 μm or less. 1. A method for forming a corrosion and wear resistant overlay , the method comprising:a first step of heating and melting a Co-base alloy powder to form a Co-base alloy layer; anda second step of forming another Co-base alloy layer on a surface of the Co-base alloy layer by repeating the first step;wherein an amount of heat input to heat and melt the Co-base alloy powder is reduced in repeatedly forming the Co-base alloy layer in the first step and the second step; andwherein the Co-base alloy layer is formed by heating and melting the Co-base alloy powder deposited beforehand.2. The method according to claim 1 , wherein the Co-base alloy layer is formed such that a carbide eutectic that precipitates at a boundary between crystals of a dendrite structure of a Co-base alloy in at least an outermost Co-base alloy layer has a maximum thickness of 10 um or less in the first step and the second step.3. The method according to claim 1 , wherein the Co-base alloy layer is formed by using any one of a laser beam claim 1 , an arc discharge claim 1 , or a charged-particle beam in a vacuum.4. The method according to claim 1 , wherein the Co-base alloy layer is formed on a surface of a base comprising a carbon steel or a ...

Power Handling In A Scalable Storage System

Systems, methods, apparatuses, and software for data storage systems are provided herein. In one example, a data storage assembly is provided. The data storage assembly includes a plurality of storage drives each comprising a PCIe host interface and solid state storage media, with each of the storage drives configured to store and retrieve data responsive to storage operations received over an associated PCIe host interface. The data storage assembly includes a PCIe switch circuit coupled to the PCIe host interfaces of the storage drives and configured to receive the storage operations issued by a plurality of host systems over a shared PCIe interface and transfer the storage operations for delivery to the storage drives over selected ones of the PCIe host interfaces. The data storage assembly includes holdup circuitry configured to provide power to at least the storage drives after input power is lost to the data storage assembly.

Internal-External Hybrid Microreactor in a Compact Configuration

An exemplary embodiment can include an apparatus including: an internal-external hybrid nuclear reactor, which can include: at least one reciprocating internal engine; and at least one external reactor integrated with said at least one reciprocating internal engine. The reciprocating engine can receive nanofuel (including moderator, nanoscale molecular dimensions & molecular mixture) internally in an internal combustion engine that releases nuclear energy. A method of operating the hybrid nuclear reactor can include operating the reciprocating internal engine loaded with nanofuel in spark or compression ignition mode. A method of cycling the reciprocating internal engine, can include compressing nanofuel; igniting nanofuel; capturing energy released in nanofuel, which is also the working fluid; and using the working fluid to perform mechanical work or generate heat. 1. An apparatus comprising: at least one reciprocating internal engine; and', 'at least one external reactor integrated with said at least one reciprocating internal engine., 'an internal-external hybrid nuclear reactor comprising2. The apparatus according to claim 1 , wherein said reciprocating internal engine comprises said at least one reciprocating internal engine comprising at least one or more of:at least one cylindrical piston;at least one reciprocating engine;at least one reciprocating rotary engine;at least one rotary engine; orat least one reciprocating rotary internal nanofuel engine.3. The apparatus according to claim 1 , wherein said external reactor comprises at least one or more of:at least one plasma core assembly;at least one reflector;at least one Beryllium (Be) reflector;at least one reflector surrounded by at least one solid fuel assembly;at least one solid fuel assembly;at least one solid inverted fuel assembly;at least one core assembly;at least one channel;at least one equivalent annulus;at least one coolant;at least one cladding;at least one gap;at least one fuel;at least one ...

Condensate demineralization apparatus and condensate demineralization method

A condensate demineralization method for a condensate treatment of a nuclear power generation plant, including: passing condensate at a linear flow rate ranging from 20 m/h to 200 m/h through a condensate demineralization apparatus comprising an ion exchange resin layer filled therein wherein the ion exchange resin layer includes a mixed bed of a strongly acidic cation resin and a strongly basic anion resin and a metal doped resin in a volume ratio ranging from 2% to 50% relative to the mixed bed.

Nuclear instrumentation and control system

An Instrumentation and Control (I&C) system for Emergency Diesel Generator (EDG) of the nuclear power plants is provided. The instrumentation and control system is adapted to be divided into two parts: first and second control parts. The first control part includes Safety I&C functions adapted to be controlled by wired logics based on electromechanical relays. Further, the second control part includes Non-safety I&C functions adapted to be controlled by Programmable Logic Controllers (PLCs)/Human Machine Interface (HMI).

Low Temperature Thermal Energy Converter for Use with Spent Nuclear Fuel Rods

According to an aspect, a vapor powered apparatus for generating electric power includes a liquid chamber that contains a working fluid and a first heat exchanger that transfers heat from fluid coming from a heat source to working fluid coming from the liquid chamber, where the transferred heat vaporizes at least a portion of the working fluid to provide a working pressure of the vaporized working fluid. The apparatus includes a pressure motor to convert the working pressure of the vaporized working fluid into mechanical motion for a power generator. The apparatus includes a vapor chamber to capture the vaporized working fluid and a second heat exchanger to use working fluid from the liquid chamber to condense the captured vaporized working fluid. An exchanger fluid system provides the working fluid to the second heat exchanger from a bottom portion of a pool of working liquid in the liquid chamber. 1. A vapor powered apparatus for generating electric power , comprising:a liquid chamber configured to contain a working fluid;a first heat exchanger, in fluid communication with the liquid chamber, configured to transfer heat from fluid coming from a heat source to working fluid coming from the liquid chamber, wherein the transferred heat vaporizes at least a portion of the working fluid to provide a working pressure of the vaporized working fluid;a pressure motor, in fluid communication with the heat exchanger, configured to convert the working pressure of the vaporized working fluid into mechanical motion for a power generator operatively connected to the pressure motor;a vapor chamber configured to capture the vaporized working fluid exiting the pressure motor;a second heat exchanger configured to use working fluid from the liquid chamber to condense the captured vaporized working fluid, returning the condensed working fluid back to the liquid chamber; andan exchanger fluid system within the liquid chamber configured to provide the working fluid to the second heat ...

Residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant

A residual heat removal ventilation system for spent fuel dry storage facility of nuclear power plant includes a natural ventilation apparatus and a forced ventilation apparatus, comprising a cold air intake chamber, a hot air removal chamber, a pipeline, a ventilation heat shield cylinder, a heat removal fan, and an air cooling equipment having certain connecting relationships and being correspondingly arranged in a storeroom, an operating room and a ventilation equipment room. The system doesn't require storing spent fuel in a pool storage manner. The safety of the spent fuel doesn't rely on power equipment, thus not only reducing routine maintenance, saving energy, but also has inherent safety. Furthermore, the system can be used to cool spent fuel storage canisters within spent fuel storage facility of pebble bed high temperature gas-cooled reactor nuclear power plant, and discharge residual heat of spent fuel storage canisters to the external environment.

METHOD AND APPARATUS FOR RECOVERY OF RADIOACTIVE NUCLIDES FROM SPENT RESIN MATERIALS

A process for the recovery of a radioisotope from a waste resin of a nuclear power plant comprises the steps of: a) treating a waste resin loaded with at least one radioisotope with an organic acid or alkaline compound to release the at least one radioisotope and to obtain a process solution containing the at least one radioisotope; b) separating the at least one radioisotope from the process solution through a reaction specific to the radioisotope so as to obtain a treated process solution depleted of the at least one radioisotope, wherein said depleted process solution comprises the organic acid or alkaline compound and optionally a non-reacted radioisotope; c) reacting the organic acid or alkaline compound in the depleted process solution from step b) by thermal and/or photochemical oxidation to form gaseous reaction products; and d) reloading the waste resin with the reacted process solution from step c) to bind the non-reacted radioisotope on the waste resin. Further, an apparatus is provided to carry out the above method. 1. A method for the recovery of a radioactive isotope from a spent waste resin of a nuclear power plant , wherein the waste resin is an ion exchange resin selected from the group consisting of cationic and anionic exchange resins , mixed bed ion-exchange resins , and mixtures thereof , and wherein the spent waste is loaded with at least one radioisotope , the method comprising the steps of:a) treating a waste resin loaded with at least one radioisotope with an organic acid or an alkaline compound to release the at least one radioisotope from the spent waste resin and to obtain a process solution containing the at least one radioisotope;b) separating the at least one radioisotope from the process solution through a reaction specific to the radioisotope so as to obtain a process solution depleted of the at least one radioisotope, wherein said reaction specific to the radioisotope is selected from the group of a physical reaction, an ...

METHOD AND APPARATUS FOR GENERATING ELECTRICITY USING A NUCLEAR POWER PLANT

A method for generating electricity by means of a nuclear power plant and a liquid vaporization apparatus involves producing heat energy by means of the nuclear power plant and using the heat energy to vaporize water or to heat water vapor, expanding the water vapor formed in a first turbine and using the first turbine to drive an electricity generator in order to produce electricity, vaporizing liquefied gas coming from a cryogenic storage in order to produce a pressurized gas, reheating the pressurized gas with a part of the water vapor intended for the first turbine of the power plant and expanding the pressurized fluid in a second turbine to produce electricity. 112-. (canceled)13. A method for the generation of electricity by means of a thermal power plant , and a liquid vaporization apparatus , the method comprising the steps of:a) producing thermal energy by means of the nuclear power plant, and the thermal energy is used to vaporize water or to heat water vapor, the water vapor formed is expanded in a first turbine, and the first turbine is used to drive an electricity generator for the production of electricity;b) vaporizing liquefied gas sourced from a cryogenic storage facility to produce a pressurized gas;c) heating the pressurized gas, andd) expanding the pressurized fluid in a second turbine for the production of electricitywherein in step c) to heat the pressurized fluid, a proportion of the thermal energy produced in step a) is used for the heating of the pressurized fluid, by employing a proportion of the water vapor to be delivered to the first turbine of the nuclear power plant, or a proportion of the heat of the water vapor to be delivered to the first turbine of the nuclear power plant for the heating of the pressurized fluid,wherein a fluid is heated by the heat generated by a nuclear reaction, a proportion of the heat energy of the heated fluid is used to preheat the fluid to be delivered to the second turbine, and a further proportion of the ...

NUCLEAR GRADE AIR ACCUMULATING, ISOLATING, INDICATING AND VENTING DEVICE

A device for isolating, accumulating, indicating and venting gas in a fluid system pipe includes a coupling affixed to a sys tem pipe. The coupling includes an isolation valve. A standpipe is attached to the coupling. The standpipe holds a float trapped between a flow retaining orifice and a closed upper end save for an angled hole therein that allows gas to flow around the float. An indicator exterior to the pipe indicates the float's level in the standpipe regardless of system pressure changes. A vent valve attached above the standpipe allows controlled ventilation of the gas flowing from the system pipe through the standpipe and through the vent valve. Accumulation of gas from the system pipe lowers the float in the standpipe, at which point the user vents the gas, causing the float to rise with the rising fluid level. 1. A device for removing , isolating , measuring and venting gas from an otherwise fluid-filled system pipe , said device comprising:(a) a hollow coupling having a lower end and an upper end, said lower end being attached to and in fluid communication with the interior of a system pipe;(b) a standpipe having a lower end and an upper end, said lower end being attached to said upper end of said coupling so that said standpipe is in fluid communication with said coupling and said system pipe;(c) a float carried within said standpipe wherein, when fluid flows from said system pipe through said coupling and into said standpipe, said float rises with the level of said fluid in said standpipe from said lower end to said upper end, said standpipe having an axis, said upper end having a hole formed therein at an angle with respect to said axis and through which fluid and gas accumulates and flows around said float and through said hole;(d) a vent valve attached to said upper end of said standpipe and in fluid communication with said standpipe for venting accumulated gas above the level of said fluid;(e) a float retaining orifice between said lower flange ...

METHOD AND APPARATUS FOR FILTERING FLUID IN NUCLEAR POWER GENERATION

A filtering apparatus for a fluid intake of a nuclear power generation facility comprise primary and secondary frames. The primary frame defines an enclosed volume having least one inlet opening, and at least one outlet opening in fluid communication with the fluid intake. A primary filter is supported on the primary frame and covers the inlet opening such that fluid passes into the enclosed volume through the primary filter. The secondary frame is located within the volume enclosed by the primary frame. A secondary filter is supported on the secondary frame and defines an enclosed flow passage in communication with the outlet opening, such that fluid passes into the at least one outlet opening through the secondary filter and the enclosed flow passage. 1. A filtering apparatus for a fluid intake of a nuclear power generation facility , comprising:a primary frame defining a primary enclosed volume, at least one inlet opening in fluid communication with the enclosed volume, and at least one outlet opening in fluid communication with the fluid intake;a primary filter supported on said frame and covering said at least one inlet opening such that fluid passes into said enclosed volume through said primary filter;a secondary frame within said primary enclosed volume;a secondary filter supported on said secondary frame defining an enclosed flow passage in communication with said at least one outlet opening, such that fluid passes into said at least one outlet opening through said secondary filter and said enclosed flow passage.2. The filtering apparatus of claim 1 , wherein said secondary filter is wrapped around said secondary frame to enclose said enclosed flow passage.3. The filtering apparatus of claim 1 , wherein said secondary filter circumscribes said enclosed flow passage.4. The filtering apparatus of claim 1 , wherein said secondary filter defines a cylindrical filtering surface.5. The filtering apparatus of claim 1 , wherein said secondary filter defines a ...

Condensate System for Recuperating Energy from a Nuclear Power Plant

A condensation system for recuperating the energy discharge of a nuclear power plant comprises a nuclear power unit, an air intake means, a compressor, a condenser, a water chamber equipped with a sprinkler, an electrical current generator, a pure water pump station, a cooling water pump station, a secondary condensate pool and a turboexpander. The air intake structure is connected to a compressor, which is connected to the condenser, which is connected to the turboexpander, which is supplied with the electric current generator and is connected to the water chamber, which is connected to the secondary condensate pool, which is connected to the pure water pump station, the condenser being connected to the cooling water pump station, wherein the air intake structure is accommodated in the discharge water channel, which is connected to the nuclear power unit and is equipped with a sealing cover. 1. A condensational recuperation system for the energy output of a nuclear power station , comprising a nuclear power plant , an air intake means , a compressor , a condenser , a water chamber provided with a sprinkler , an electric current generator , a pure water pump station , a cooling water pump station , a secondary condensate pool , and a turbo expander , the air intake means being connected to a compressor connected to a condenser connected to the turbo expander provided with an electric current generator and connected to a water chamber connected to the secondary condensate pool , the secondary condensate pool being connected to a pump station of pure water , the condenser being connected to a pump station of cooling water , all of the connections being made in the form of pressure pipelines , characterized in that the air intake means is arranged in the wastewater channel of a nuclear power plant connected to the nuclear power plant , and the wastewater channel is provided with a sealed roof2. The system according to claim 1 , characterized in that the wastewater ...

SOLAR NUCLEAR FUSION DEVELOPMENT

Systems and methods for providing a solar photovoltaic (PV) facility as a source of secondary power for a nuclear power facility in the event of a power failure at the nuclear power facility. The solar PV facility is operably connected to the nuclear power facility by, e.g., a direct connection or through a substation. When a power failure at the nuclear power facility is detected, a switching system connects the solar PV facility to the nuclear power facility to provide a source of backup power to emergency systems. Power may be applied directly to such systems or to batteries at the nuclear power facility. In some implementations, the solar PV facility is physically located proximate to the nuclear power facility. 1. A system , comprising: a solar photovoltaic (PV) facility operably connected to a nuclear power facility , wherein the solar PV facility provides power to the nuclear power facility in the event of a loss of power at the nuclear power facility.2. The system of claim 1 , wherein the solar PV facility is proximate to the nuclear facility.3. The system of claim 1 , wherein he solar PV facility is 5 to 15 miles from the nuclear power facility.4. The system of claim 1 , wherein the solar PV facility produces at or at least 20 MW of electricity.5. The system of claim 1 , wherein the solar PV facility produces at or at least 100 MW of electricity.6. The system of claim 1 , wherein the solar PV facility provides electrical power to a cooling system of the nuclear power facility.7. The system of claim 6 , wherein the cooling system is a spent fuel pool cooling system.8. The system of claim 1 , wherein the solar PV facility provides electrical power to an emergency service water pump of the nuclear power facility.9. The system of claim 1 , wherein the solar PV facility comprising from 120 to 130 inverters.10. The system of claim 1 , wherein the solar PV facility provides power for black start.11. A method of providing a secondary source of power to a nuclear ...

UPGRADING POWER OUTPUT OF PREVIOUSLY-DEPLOYED NUCLEAR POWER PLANTS

Systems and methods for upgrading power output of previously-deployed nuclear power plants are described. Systems and methods may include a base nuclear power plant with a predetermined base power output rating and a predetermined base whole core refueling interval. Systems and methods may also include a power upgrade kit for increasing the base power output rating from the base power output rating to an increased power output rating without a change in fuel charge, reactor structures, or civil structures. 1. A system comprising:a previously-deployed nuclear power plant with a predetermined base power output rating and a predetermined base whole core refueling interval; anda power upgrade kit for increasing the base power output rating from the base power output rating to an increased power output rating without a change in fuel charge, reactor structures, or civil structures.2. The system of claim 1 , wherein the previously-deployed nuclear power plant is a small modular reactor nuclear power plant.3. The system of claim 1 , wherein the predetermined base power output rating is approximately 100 MWe.4. The system of claim 1 , wherein the predetermined base whole core refueling interval is approximately 20 years.5. The system of claim 1 , wherein the increased power output rating is at least approximately double the predetermined base power output rating.6. The system of claim 1 , wherein the increased power output rating is approximately 200 MWe.7. The system of claim 1 , wherein the power upgrade kit comprises an additional energy converter system claim 1 , an additional heat transport loop claim 1 , one or more additional primary pumps claim 1 , and one or more passive decay heat removal heat exchangers.8. The system of claim 1 , wherein the base nuclear power plant comprises a balance of plant zone and a nuclear zone claim 1 , wherein all nuclear safety functions occur in the nuclear zone.9. The system of claim 8 , wherein the balance of plant zone comprises an ...

Method and apparatus for generating electricity and storing energy using a thermal or nuclear power plant

A method for generating electricity by means of a nuclear power plant and a liquid vaporization apparatus involves, during a first period, producing heat energy by means of the nuclear power plant and using the heat energy to vaporize water or to heat water vapour, expanding the water vapour formed in a first turbine and using the first turbine to drive an electricity generator in order to produce electricity, vaporizing liquefied gas coming from a cryogenic store in order to produce pressurized gas, reheating the pressurized gas with a part of the water vapour intended for the first turbine of the nuclear power plant and expanding the pressurized fluid in a second turbine to produce electricity and, during the second period, liquefying the gas to be vaporized.

Tamper prevention system and method for remotely monitored sealing and unsealing of containers

A tamper prevention system comprising a storage unit is described. The storage unit comprises at least one storage compartment for storing a sealing element for containers, and a control system configured for detecting opening of the at least one storage compartment. Each storage compartment comprises a leash element connected to the control system and provided for being connected to a sealing element to be stored in the storage compartment. The control system is configured for monitoring the integrity of the leash element. The tamper prevention system comprises movement protection device configured for detecting and/or preventing movement of the storage unit.

FINNED STRAINER

The present invention relates to filters used to remove debris from water being sucked into a piping system. It has particular application use in nuclear power plants, which, after a loss of coolant accident, must pump cooling water back into the reactor core from a collection sump. This water may contain various types of debris that must be removed before the water is sent back into the reactor cooling system. There are restrictions on the allowable pressure drop across the strainer and the space available for installing this equipment. The finned strainer of the present invention addresses these issues while maximizing the quantity of debris filtered from the water. 1. A strainer element for connection to a collection header , the strainer element comprising:a perimeter frame; anda first fluid permeable screen and a second permeable screen supported on the frame,wherein:said permeable screens have a plurality of corrugations formed of a parallel plurality of peaks and valleys, contact at alternating peaks maintaining the fluid permeable screens in opposed spaced relationship for defining fluid flow channels therebetween through which fluid passing through either one of the fluid permeable screens can be conducted towards the collection header;said fluid flow channels communicate with said header through said perimeter frame,wherein each one of the fluid permeable screens are a perforated metal screen or metal mesh,wherein said fluid flow channels are configured to communicate with said header through openings in a marginal side edge of said perimeter frame.2. The strainer element of claim 1 , wherein the frame comprises claim 1 , at the end of the strainer element for connection to the collection header claim 1 , a cap claim 1 , wherein the cap includes openings through which flow from the fluid passage can communicate with the header.3. The strainer element of claim 1 , wherein the frame comprises a perimeter frame having openings along one side edge of said ...

Device for increase of efficiency and power of transportable nuclear power plant

FIELD: power industry. SUBSTANCE: transportable nuclear power plant includes the nuclear reactor connected to a steam generator, a gas copper boiler superheater, a turbogenerator and an air heater. Also the feedwater heater is added to the first input of which the second output of the boiler superheater is connected which heats the steam up to the temperature 540°C-600°C, and the first output of a heater of feedwater is connected to the second input of an airheater. The second output of the feedwater heater is connected to the steam generator, and the first output of the boiler superheater is connected through the turbogenerator, the condenser, the low pressure heater, the deaerator and the high pressure heater to the second input of the feedwater heater. EFFECT: improvement of efficiency and effective power of transportable, including floating, nuclear power plants. 2 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 550 362 C1 (51) МПК G21D 5/14 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ 2014101675/07, 22.01.2014 (24) Дата начала отсчета срока действия патента: 22.01.2014 (73) Патентообладатель(и): Иванюк Виктор Николаевич (RU) (45) Опубликовано: 10.05.2015 Бюл. № 13 R U Приоритет(ы): (22) Дата подачи заявки: 22.01.2014 (72) Автор(ы): Завьялов Сергей Николаевич (RU), Иванюк Виктор Николаевич (RU), Иванюк Андрей Викторович (RU), Рыжков Вениамин Васильевич (RU) (56) Список документов, цитированных в отчете о поиске: RU2335641 C2, 10.10.2008 . RU2328045 2 5 5 0 3 6 2 R U (54) УСТРОЙСТВО ПОВЫШЕНИЯ КПД И МОЩНОСТИ ТРАСНПОРТАБЕЛЬНОЙ АТОМНОЙ ЭЛЕКТРОСТАНЦИИ (57) Реферат: Изобретение относится к области питательной воды соединен со вторым входом теплотехники, а именно к атомной энергетике. воздухоподогревателя. Второй выход Транспортабельная атомная электростанция подогревателя питательной воды соединен с включает ядерный реактор, соединенный с парогенератором, а первый выход котлапарогенератором, газовый ...

Vanne

燃气轮机压水堆蒸汽轮机联合循环系统

本发明燃气轮机压水堆蒸汽轮机联合循环系统涉及一种大容量的节能、低碳、清洁能源系统；用重型燃气轮机与压水堆蒸汽轮机组成联合循环系统，利用燃气轮机尾气的热量将二回路主蒸汽温度由272.8℃向上提升，随压水堆产汽量的不同和重型燃气轮机投入的台数及负荷的不同，二回路主蒸汽温度在272.8℃至630℃之间滑温运行；燃气轮机压水堆蒸汽轮机联合循环系统的热效率明显高于压水堆蒸汽轮机的热效率；就燃气增发的电量而言，燃气轮机压水堆蒸汽轮机联合循环系统的热效率也明显高于现有技术的燃气‑蒸汽联合循环。

Patent JPS564878B2

Backup nuclear reactor auxiliary power using decay heat

一种核设备辅助备用电源系统，所述核设备辅助备用电源系统在设备停机之后使用衰变热，以便通过专用蒸汽涡轮机/发电机组发电。衰变热产生热的工作气态流体，所述热的工作气态流体用作备用物，以便使得尺寸适当的涡轮机运转，所述涡轮机向发电机提供动力。涡轮机构造成使用现有核设备辅助系统的一部分并且将涡轮机废气排出到环境空气。所述系统用于去除反应堆衰变热并且向设备系统提供电力，以便在不能利用传统电源的情况中使得能够有序地停机。

Passive containment cooling system and nuclear power plant having the same

본 발명은 플레이트형 열교환기를 적용한 피동격납건물냉각계통을 개시한다. 피동격납건물냉각계통은, 격납건물, 상기 격납건물의 내부와 외부 중 적어도 한 곳에 설치되고 압력경계를 유지하면서 상기 격납건물의 대기와 열교환 유체를 서로 열교환시키도록 경계면의 양측에 각각 서로 구분되게 배열되는 채널들을 구비하는 플레이트형 열교환기, 및 상기 격납건물의 대기 또는 상기 열교환 유체의 유로를 형성하도록 상기 격납건물을 관통하여 상기 플레이트형 열교환기와 상기 격납건물을 연결하는 배관을 포함를 포함한다. The present invention discloses a passive containment building cooling system to which a plate type heat exchanger is applied. The passive containment building cooling system comprises at least one of an interior and an exterior of the containment building and is arranged to be separated from each other on both sides of the interface so as to exchange heat between the atmosphere of the containment building and the heat exchange fluid, And a pipe connecting the plate heat exchanger and the containment building through the containment building to form a flow path of the atmosphere or the heat exchange fluid of the containment building.

A kind of passive PWR nuclear power plant reactor coolant loop is arranged

本发明提供一种非能动压水堆核电站反应堆冷却剂环路布置，其包括：反应堆压力容器；所述反应堆压力容器具有接管段筒体；所述接管段筒体设有反应堆冷却剂出口接管嘴和反应堆冷却剂进口接管嘴；蒸汽发生器；所述蒸汽发生器通过主管道与所述反应堆压力容器连接。本发明提供的非能动压水堆核电站反应堆冷却剂环路布置，对反应堆冷却剂回路设计改进，打破大型CAP系列核电站大型主泵技术发展瓶颈难题，在不使用比CAP1000核电站更大功率的主泵，甚至使用比CAP1000核电站还小功率主泵的情况下，实现两个主冷却剂环路，同时比CAP1000核电站功率更大的CAP系列非能动核电站。

Water-spray residual heat removal system for nuclear power plant

提供一种原子能发电站的余热排除系统。该系统包括：空气管，其被配置在安全壳外部；热交换器，其位于所述空气管内部；第一排管，其将所述安全壳内的蒸气产生器所产生的蒸气传输至所述热交换器；第二排管，其将所述热交换器中被冷却及冷凝的冷凝水传输至所述蒸气产生器中。该系统利用流动至所述空气管内部的外部空气来将所述热交换器气冷。上述结构，为了热交换器的冷却不仅通过水冷方式，还通过等候中的循环空气来进行冷却，因此，可不受冷却运作时间限制进行冷却。在核电事故初期可向热交换器喷射冷却水，利用水的蒸发热或是热交换器被淹没在水槽中的状态来进行热排除，且在余热较小的事故后期（事故8小时之后）可转换成气冷方式。

POWER PLANT BASED ON A SMALL MODULAR REACTOR WITH POSSIBILITIES OF LOAD TRACKING AND COMBINED GENERATION OF ELECTRICITY AND HEAT AND METHODS OF USE

РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2019 120 653 A (51) МПК G21D 1/02 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ЗАЯВКА НА ИЗОБРЕТЕНИЕ (21)(22) Заявка: 2019120653, 11.12.2017 (71) Заявитель(и): ЭДВАНСЕД РЕАКТОР КОНСЕПТС ЛЛК (US) Приоритет(ы): (30) Конвенционный приоритет: 11.12.2016 US 62/432,668 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 11.07.2019 R U (43) Дата публикации заявки: 14.01.2021 Бюл. № 2 (72) Автор(ы): УОЛТЕРС, Леон (US), УЭЙД, Дэвид (US) (86) Заявка PCT: (87) Публикация заявки PCT: WO 2018/107170 (14.06.2018) R U (54) ЭНЕРГЕТИЧЕСКАЯ СТАНЦИЯ НА ОСНОВЕ МАЛОГО МОДУЛЬНОГО РЕАКТОРА С ВОЗМОЖНОСТЯМИ СЛЕДОВАНИЯ ЗА НАГРУЗКОЙ И КОМИБИНИРОВАННОЙ ВЫРАБОТКИ ЭЛЕКТРОЭНЕРГИИ И ТЕПЛА И СПОСОБЫ ИСПОЛЬЗОВАНИЯ (57) Формула изобретения 1. Энергетическая станция на основе малого модульного ядерного реактора, содержащая активную зону реактора, которая содержит блок натриевого теплоносителя первого контура, содержащий поток холодного натриевого теплоносителя первого контура; и поток нагретого натриевого теплоносителя первого контура, при этом предусмотрена возможность поступления потока нагретого натриевого теплоносителя первого контура в один или более теплообменников, и возможность теплообмена потока нагретого натриевого теплоносителя первого контура с натриевым теплоносителем второго контура, текущим через по меньшей мере один промежуточный натриевый контур. 2. Энергетическая станция по п. 1, отличающаяся тем, что промежуточный натриевый контур содержит поток натриевого теплоносителя второго контура, выполненный с возможностью транспортировать тепло к блоку преобразования энергии через один или более теплообменников. 3. Энергетическая станция по п. 1, отличающаяся тем, что малый модульный ядерный реактор дополнительно содержит турбину, выполненную с возможностью функционировать как часть блока преобразования энергии с циклом Брайтона. 4. Энергетическая станция по п. 3, отличающаяся тем, что блок преобразования Стр.: 1 A 2 0 ...

A kind of passive PWR nuclear power plant reactor coolant loop is arranged

本发明提供一种非能动压水堆核电站反应堆冷却剂环路布置，其包括：反应堆压力容器；所述反应堆压力容器具有接管段筒体；所述接管段筒体设有反应堆冷却剂出口接管嘴和反应堆冷却剂进口接管嘴；蒸汽发生器；所述蒸汽发生器通过主管道与所述反应堆压力容器连接。本发明提供的非能动压水堆核电站反应堆冷却剂环路布置，通过对反应堆主冷却剂回路的设计改进，打破大型CAP系列核电站主泵技术瓶颈难题，同时打破大型CAP系列核电站主管道热段制造技术瓶颈难题。

Patent RU2019120653A3

ВИ“? 2019120653” АЗ Дата публикации: 21.05.2021 Форма № 18 ИЗПМ-2011 Федеральная служба по интеллектуальной собственности Федеральное государственное бюджетное учреждение ж 5 «Федеральный институт промышленной собственности» (ФИПС) ОТЧЕТ О ПОИСКЕ 1. . ИДЕНТИФИКАЦИЯ ЗАЯВКИ Регистрационный номер Дата подачи 2019120653/07(040334) 11.12.2017 РСТ/О$2017/065634 11.12.2017 Приоритет установлен по дате: [ ] подачи заявки [ ] поступления дополнительных материалов от к ранее поданной заявке № [ ] приоритета по первоначальной заявке № из которой данная заявка выделена [ ] подачи первоначальной заявки № из которой данная заявка выделена [ ] подачи ранее поданной заявки № [Х] подачи первой(ых) заявки(ок) в государстве-участнике Парижской конвенции (31) Номер первой(ых) заявки(ок) (32) Дата подачи первой(ых) заявки(ок) (33) Код страны 1. 62/432,668 11.12.2016 05 Название изобретения (полезной модели): [Х] - как заявлено; [ ] - уточненное (см. Примечания) ЭНЕРГЕТИЧЕСКАЯ СТАНЦИЯ НА ОСНОВЕ МАЛОГО МОДУЛЬНОГО РЕАКТОРА С ВОЗМОЖНОСТЯМИ СЛЕДОВАНИЯ ЗА НАГРУЗКОЙ И КОМИБИНИРОВАННОЙ ВЫРАБОТКИ ЭЛЕКТРОЭНЕРГИИ И ТЕПЛА, И СПОСОБЫ ИСПОЛЬЗОВАНИЯ Заявитель: ЭДВАНСЕД РЕАКТОР КОНСЕПТС ЛЛК, 05 2. ЕДИНСТВО ИЗОБРЕТЕНИЯ [Х] соблюдено [ ] не соблюдено. Пояснения: см. Примечания 3. ФОРМУЛА ИЗОБРЕТЕНИЯ: [Х] приняты во внимание все пункты (см. Примечания) [ ] приняты во внимание следующие пункты: [ ] принята во внимание измененная формула изобретения (см. Примечания) 4. КЛАССИФИКАЦИЯ ОБЪЕКТА ИЗОБРЕТЕНИЯ (ПОЛЕЗНОЙ МОДЕЛИ) (Указываются индексы МПК и индикатор текущей версии) (21) 1/02 (2006.01) 5. ОБЛАСТЬ ПОИСКА 5.1 Проверенный минимум документации РСТ (указывается индексами МПК) 021. 1/02, 1/00 (2006.01 5.2 Другая проверенная документация в той мере, в какой она включена в поисковые подборки: 5.3 Электронные базы данных, использованные при поиске (название базы, и если, возможно, поисковые термины): РУУРТ, Езрасепе, Соозе, боозе Рас РАТЕМТЗСОРЕ, Рабеагсв, КОРТО, ОЗРТО, Уапдех 6. ДОКУМЕНТЫ, ОТНОСЯЩИЕСЯ К ...

External Reactor Vessel Cooling and Electric Power Generation System

본 발명에 따른 원자로용기 외벽 냉각 및 발전 시스템은 소규모의 설비로 원자로용기의 적어도 일부를 감싸도록 형성되고, 상기 원자로용기에서 방출되는 열을 냉각하도록 형성되는 원자로용기 외벽 냉각부, 상기 원자로용기 외벽 냉각부에서 열을 전달받는 유체를 이용하여 전기에너지를 생산하도록 형성되는 소형터빈 및 소형발전기를 포함하는 전력 생산부, 상기 소형터빈을 구동하고 배출되는 상기 유체를 열교환시키고, 상기 유체를 응축시켜 응축수를 생성하도록 형성되는 응축열 교환부 및 상기 응축열 교환부에서 생성되는 상기 응축수를 수집하도록 형성되는 응축수 저장부를 포함하고, 상기 유체는 상기 원자로용기로부터 전달받은 열에 의해 기체로 상변화(phase transition)하는 것을 특징으로 한다. 본 발명에 따른 원자로용기 외벽 냉각 및 발전 시스템은 정상운전 시뿐만 아니라 사고 시에도 지속적으로 작동하여 원자로용기를 냉각하고 비상전력을 생산하여 계통 신뢰성을 향상시킬 수 있다. 본 발명에 따른 원자로용기 외벽 냉각 및 발전 시스템은 소규모 설비로 안전등급 또는 내진설계의 적용이 용이하고, 안전등급 또는 내진설계의 적용으로 신뢰성이 향상된다. The reactor outer wall cooling and power generation system according to the present invention includes a reactor outer casing cooling part formed to surround at least a part of a reactor vessel with a small scale facility and configured to cool heat emitted from the reactor vessel, A power generator including a small turbine and a small generator that are configured to produce electric energy using heat received by the small turbine, heat exchange of the fluid discharged and driving the small turbine, condensation of the fluid to generate condensed water And a condensed water storage part formed to collect the condensed water generated in the condensation heat exchange part, wherein the fluid is phase-transited to gas by the heat transferred from the reactor vessel do. The reactor outer wall cooling and power generation system according to the present invention can continuously operate not only during normal operation but also during an accident so as to cool the reactor vessel and generate emergency power to improve system reliability. The reactor outer wall cooling and power generation system according to the present invention is easy to apply safety grade or seismic design to a small scale facility, and reliability is improved by applying a safety grade or seismic design.

Appliance and method for cleaning heat exchanger inner zone

FIELD: technological processes.SUBSTANCE: invention relates to means of cleaning the inner zone of a heat exchanger, particularly steam generator (2) of a nuclear power plant. In compliance with this invention, high-pressure water jets comprises manipulator (5) installed in vertical corridor passing between bundles of heat exchanger tubes, and elevator (4) connected with said manipulator (5). Manipulator (5) has at least one rotary nozzle (8, 8.1, 8.2, 8.3, 8.4), installed with possibility of rotation around rotary axis (S1, S2, S3, S4), orientation of which is matched with distance between pipes in tube bundle (3) of heat exchanger.EFFECT: technical result is higher efficiency of cleaning heat exchangers with pipes passing in horizontal direction.18 cl, 3 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 692 748 C2 (51) МПК G21D 1/02 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ (52) СПК G21D 1/02 (2019.02) (21)(22) Заявка: 2015122257, 10.06.2015 (24) Дата начала отсчета срока действия патента: Дата регистрации: (73) Патентообладатель(и): Фраматом ГмбХ (DE) 27.06.2019 (56) Список документов, цитированных в отчете о поиске: BE 900716 A, 01.02.1985. RU 123509 (43) Дата публикации заявки: 27.12.2016 Бюл. № 36 U1, 27.12.2012. SU 934187 A1, 07.06.1982. US 4905900 A1, 06.03.1990. (45) Опубликовано: 27.06.2019 Бюл. № 18 2 6 9 2 7 4 8 R U (54) ПРИСПОСОБЛЕНИЕ И СПОСОБ ДЛЯ ОЧИСТКИ ВНУТРЕННЕЙ ЗОНЫ ТЕПЛООБМЕННИКА (57) Реферат: Изобретение относится к средствам очистки по меньшей мере одно поворотное сопло (8, 8.1, внутренней зоны теплообменника, в частности, 8.2, 8.3, 8.4), установленное с возможностью парогенератора (2) атомной электростанции. В поворота вокруг поворотной оси (S1, S2, S3, S4), заявленном изобретении водяными струями ориентация которой согласована с расстоянием высокого давления включает манипулятор (5), между трубами в пучке труб (3) теплообменника. устанавливаемый в вертикальном коридоре, Техническим ...

Method and system for emergency and backup cooling of nuclear fuel and nuclear reactors

1. Система, содержащаякамеру ядерного реактора, имеющую впускной порт и по меньшей мере один резервуар, содержащий жидкий азот, по меньшей мере один резервуар, содержащий выпускной порт, гидравлически соединенный с упомянутым впускным портом камеры ядерного реактора с обеспечением возможности вытекания жидкого азота в камеру по меньшей мере из одного резервуара.2. Система по п. 1, отличающаяся тем, что по меньшей мере один резервуар содержит бор.3. Система по п. 1, дополнительно содержащая термически активируемый клапан, гидравлически соединенный с впускным портом камеры ядерного реактора и выполненный с возможностью управления потоком жидкого азота, выпускаемого в камеру.4. Система по п. 1, отличающаяся тем, что содержит по меньшей мере один малый резервуар, имеющий первый объем,и дополнительно содержит большой резервуар с жидким азотом, имеющий второй объем, превышающий первый объем, при этом большой резервуар гидравлически соединен с упомянутым по меньшей мере одним резервуаром.5. Система по п. 1, дополнительно содержащая оборудование для производства жидкого азота, гидравлически соединенное с по меньшей мере одним резервуаром.6. Система тушения огня без воды, содержащая пожарную машину, снабженную находящимся под давлением контейнером большого объема, наполненным жидким азотом, рукавом для выпускания жидкого азота, имеющим достаточные пропускную способность и длину, и прибором контроля расхода выпускания жидкого азота, характеризующаяся тем, что она выполнена с возможностью направлять выпускаемый жидкий азот в область, пространственно близкую к огню, или на источник огня, с созданием в большом количестве газообразного азота для пр РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (51) МПК G21C 15/18 (13) 2014 136 060 A (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ЗАЯВКА НА ИЗОБРЕТЕНИЕ (21)(22) Заявка: 2014136060, 14.03.2013 (71) Заявитель(и): ЛИН-ХЕНДЕЛЬ Кэтрин (US) Приоритет(ы): (30) Конвенционный приоритет: (72) Автор(ы): ЛИН-ХЕНДЕЛЬ Кэтрин (US) 16.03. ...

Loop system for power station, flushing method of loop system and power station with loop system

本发明涉及一种发电站用回路系统以及具有其的发电站，所述回路系统包括：依次相连的凝汽器、凝结水泵、凝结水处理装置、除氧器液位调节装置、低温加热器、具有除氧水箱的除氧装置、主给水泵、高温加热器、主给水母管，其中：所述回路系统还包括第一流路和第三流路，第一流路连通凝结水泵的出口与高温加热器的入口且设置有第一阀，第三流路连通主给水母管与凝汽器且设置有第二阀，第一流路与第三流路用于形成凝汽器‑高温加热器‑凝汽器的连续内循环。本发明还涉及一种发电站用回路系统冲洗方法。

Apparatus for seperating water and vapour for drying a humid vapour

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Method for active removal of residual heat emission of reactors under conditions of nps blackout

FIELD: nuclear energy. SUBSTANCE: invention relates to methods of removing residual heat emission of the reactor under the conditions of NPS blackout. Additional steam turbine plant 2 continues generating electric power for own needs using steam obtained in the steam generator due to the rector residual heat emission energy. Excess steam generated in the steam generator is directed to steam bypass 8, where it heats the cold water from cold water tank 6. Produced hot water is stored in hot water tank 9. Accumulated in hot water tank 9 hot water is directed to the steam generator. Waste in additional steam turbine plant 2 steam is directed to condenser 4, from where the condensate is drained into cold water tank 6. In the operating mode during the night-time electrical load dropout part of steam from the steam generator is directed to steam bypass 8, where it heats the cold water supplied from cold water tank 6. Produced hot water is stored in hot water tank 9. Drainage of the heating steam after steam bypass 8 is fed into the feed water circuit after high-pressure feed heater 12. EFFECT: technical result is operation for electric power generation to the network in the normal mode of the plants safety increasing, absence of their idle time. 1 cl, 1 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 609 894 C1 (51) МПК G21D 1/02 (2006.01) G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ФОРМУЛА (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ 2016107253, 29.02.2016 (24) Дата начала отсчета срока действия патента: 29.02.2016 Дата регистрации: (72) Автор(ы): Аминов Рашид Зарифович (RU), Юрин Валерий Евгеньевич (RU), Бессонов Валерий Николаевич (RU) Приоритет(ы): (22) Дата подачи заявки: 29.02.2016 Адрес для переписки: 410056, Саратов, ул. Рахова, 103/115, кв. 141, Аминову Рашиду Зарифовичу WO2013176718 A1, 28.11.2013 . US9031183 B2, 12.05.2015. WO1995032509 A2, 30.11.1995. 2 6 0 9 8 9 4 R U (57) Формула изобретения Способ активного ...

Patent RU2015122257A3

ВУ” 2015122257” АЗ Дата публикации: 03.08.2018 Форма № 18 ИЗ,ПМ-2011 Федеральная служба по интеллектуальной собственности Федеральное государственное бюджетное учреждение ж Я «Федеральный институт промышленной собственности» (ФИПС) ОТЧЕТ О ПОИСКЕ 1. . ИДЕНТИФИКАЦИЯ ЗАЯВКИ Регистрационный номер Дата подачи 2015122257/07(034726) 10.06.2015 Приоритет установлен по дате: [Х] подачи заявки [ ] поступления дополнительных материалов от к ранее поданной заявке № [ ] приоритета по первоначальной заявке № из которой данная заявка выделена [ ] подачи первоначальной заявки № из которой данная заявка выделена [ ] подачи ранее поданной заявки № [ ] подачи первой(ых) заявки(ок) в государстве-участнике Парижской конвенции (31) Номер первой(ых) заявки(ок) (32) Дата подачи первой(ых) заявки(ок) (33) Код страны 1. Название изобретения (полезной модели): [Х] - как заявлено; [ ] - уточненное (см. Примечания) ПРИСПОСОБЛЕНИЕ И СПОСОБ ДЛЯ ОЧИСТКИ ВНУТРЕННЕЙ ЗОНЫ ТЕПЛООБМЕННИКА Заявитель: АРЕФА ГмбХ, ПЕ 2. ЕДИНСТВО ИЗОБРЕТЕНИЯ [Х] соблюдено [ ] не соблюдено. Пояснения: см. Примечания 3. ФОРМУЛА ИЗОБРЕТЕНИЯ: [Х] приняты во внимание все пункты (см. П см. Примечания [ ] приняты во внимание следующие пункты: [ ] принята во внимание измененная формула изобретения (см. Примечания) 4. КЛАССИФИКАЦИЯ ОБЪЕКТА ИЗОБРЕТЕНИЯ (ПОЛЕЗНОЙ МОДЕЛИ) (Указываются индексы МПК и индикатор текущей версии) (21) 1/02 (2006.01) 5. ОБЛАСТЬ ПОИСКА 5.1 Проверенный минимум документации РСТ (указывается индексами МПК) 0210 1/02 (2006.01 5.2 Другая проверенная документация в той мере, в какой она включена в поисковые подборки: 5.3 Электронные базы данных, использованные при поиске (название базы, и если, возможно, поисковые термины): РУ/РТ, Езрасепе, Соозе Рав, ]-Р]а( Рав К-РТОМ, КРК, РАТЕМТЗСОРЕ, Рабзеагсй, КОРТО, ТРО, ОРТО 6 ДОКУМЕНТЫ, ОТНОСЯЩИЕСЯ К ПРЕДМЕТУ ПОИСКА Кате- Наименование документа с указанием (где необходимо) частей, Относится к гория* относящихся к предмету поиска пункту формулы № 1 2 3 Хх ВЕ 900716 А ( ...

A kind of active and non-active reactor core water filling heat derivation device combined

本发明属于反应堆设计技术，具体涉及一种能动与非能动相结合的堆芯注水热量导出装置。其结构包括分别与换料水箱和一回路的冷、热管段相连接的安全注入管线，安全注入管线上设有安注泵，安注泵包括两台相互独立的中压安注泵和两台相互独立的低压安注泵，所述的换料水箱设置在安全壳内部堆芯下方地坑位置，安注泵设置在安全壳外部；在安全壳内反应堆的上方还设有非能动的安注箱和堆芯补水箱，所述的安注箱通过设有控制阀门的管道与一回路的冷管段相连接，所述的堆芯补水箱通过设有控制阀门的管道连接在一回路的热管段和冷管段之间。本发明将能动与非能动的优点相结合，具有冗余性、多样性、可靠性高等特点，提高了核电厂的安全性。

Emergency and back-up cooling of nuclear fuel and reactors

본 발명은 핵연료와 원자로의 긴급예비 냉각시스템과 방법에 관한 것으로, 밀도가 가장 높고 운반하기 쉬운 형태의 액체질소와, 용기에서 액체질소가 방출될 때 생기는 찬 질소기체를 연료봉과 원자로의 긴급냉각에 사용한다.

Nuclear plant with ventilation system

FIELD: nuclear power engineering. SUBSTANCE: invention relates to a ventilation system for a nuclear plant and a method for its operation. Ventilation system (2) includes ventilation line (28) leading from the inside of protective shell (8) to the outside, into which wet scrubber (12) of ventilation flow (10) is included. In wet scrubber (12) there is reservoir (26) filled with washing liquid (18), above which, there is gas space (48). Ventilation line (28) includes inlet line (14) for ventilation flow (10) leading to wet scrubber (12), which branches into several exhaust nozzles (38) in distributor (34). Exhaust nozzles (38) are at least partially immersed in washing liquid (18) of reservoir (26). There is at least one separate outlet (56) that allows redistributing ventilation flows, which is connected through outlet line (54, 90) to inlet line (14) or to distributor (34) and is thermally isolated from washing liquid (18) in reservoir (26). EFFECT: possibility, with insignificant hardware and technological costs, to reliably exclude the presence of critical concentrations of flammable gas mixtures in all modes. 20 cl, 8 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 758 899 C2 (51) МПК G21C 13/024 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ (52) СПК G21C 13/024 (2021.08) (21)(22) Заявка: 2019113146, 24.01.2018 (24) Дата начала отсчета срока действия патента: Дата регистрации: (73) Патентообладатель(и): ФРАМАТОМ ГМБХ (DE) 02.11.2021 24.01.2017 DE 10 2017 201 115.7 (43) Дата публикации заявки: 26.02.2021 Бюл. № 6 (45) Опубликовано: 02.11.2021 Бюл. № 31 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 26.08.2019 (56) Список документов, цитированных в отчете о поиске: US 5223209 A1, 29.06.1993. DE 3637795 A1, 11.05.1988. US 20140010340 A1, 09.01.2014. RU 2601285 C1, 27.10.2016. RU 2547836 C2, 10.04.2015. RU 2549544 C2, 27.04.2015. RU 2554071 C2, 27.06.2015. RU 2548011 C2, 10.04.2015. RU 2396211 C1, 10. ...

Wireless transmission of nuclear instrumentation signals

Heating system and method for primary loop cabin during cold test of high-temperature gas cooled reactor nuclear power station

本发明公开了一种高温气冷堆核电站冷试期间一回路舱室加热系统及方法，包括混凝土舱室结构体、压力容器、蒸汽发生器、第一进风管、压力容器舱室热风机装置、压力容器舱室回风管、第二进风管、蒸汽发生器舱室热风机装置及蒸汽发生器舱室回风管，该系统及方法能够保证在高温其冷堆核电站冷试期间维持一回路舱室内温度恒定，提升金属堆内构件温度，满足冷试试验要求。

Maintenance support base station and achievement method applicable to offshore floating nuclear power plant

本发明公布了一种适用于海上浮动核电站的维修保障基地及实现方法，包括浮动平台、通过锚链固定在海床上，在浮动平台上设置有动力定位系统，两个浮动平台的顶部通过高位平台连接，高位平台与两个浮动平台之间构成通道，该通道的一端封闭，另一端设置有一个活动的封闭隔板，在浮动平台的底部设置有可以滑动的底部隔板，两个底部隔板伸出后拼合成整体。本发明尽可能的压缩常规能源的使用时间，将浮动电站的维修保障基地平台靠近平台布置，节省了浮动电站换料和大修的时间，能在相对较短的时间完成浮动核电站的大修和换料，提高了浮动核电站的核能的利用率和经济性。

Method and system for emergency and back-up cooling of nuclear fuel and reactors

FIELD: nuclear physics, nuclear technology.SUBSTANCE: invention relates to a method and system for emergency and backup cooling of nuclear fuel and nuclear reactors. System comprises a nuclear reactor chamber comprising an inlet port and at least one container containing liquid nitrogen, the at least one container comprising an outlet port, in fluid communication with said inlet port of the nuclear reactor chamber, so that liquid nitrogen can flow into the chamber from the at least one container, and a thermally activated valve connected to said inlet port and configured to control the flow of liquid nitrogen. Liquid nitrogen contained in the container can enter the chamber when ambient temperature in the chamber reaches or exceeds a threshold value. Said inlet port of the nuclear reactor chamber is located above the fuel elements of the nuclear reactor.EFFECT: technical result is the provision of efficient, safe and fast cooling of the nuclear reactor under conditions where there is no electrical power or when the nuclear reactor and nuclear fuel are already overheated.5 cl, 5 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 666 790 C2 (51) МПК G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ (52) СПК G21C 15/18 (2006.01) (21)(22) Заявка: 2014136060, 14.03.2013 (24) Дата начала отсчета срока действия патента: (73) Патентообладатель(и): ЛИН-ХЕНДЕЛЬ Кэтрин (US) Дата регистрации: 12.09.2018 (56) Список документов, цитированных в отчете о поиске: US 20120002776 A1 05.01.2012. WO 2003024531 A1 27.03.2003. RU 2082226 C1 20.06.1997. SU 1648209 A1 30.06.1994. 16.03.2012 US 61/611,585 (43) Дата публикации заявки: 10.05.2016 Бюл. № 13 (45) Опубликовано: 12.09.2018 Бюл. № 26 (86) Заявка PCT: C 2 C 2 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 16.10.2014 US 2013/031408 (14.03.2013) (87) Публикация заявки PCT: WO 2013/184207 (12.12.2013) 2 6 6 6 7 9 0 2 6 6 6 7 9 0 Приоритет(ы): (30) Конвенционный ...

Nuclear power station voltage stabilizer and water seal device thereof

本发明提供了一种核电站稳压器水封装置，其包括：上封头本体，设有位于稳压器内侧的内表面堆焊层和位于稳压器外侧的保温层；以及贯穿上封头本体倾斜向上延伸的稳压器安全阀接管，稳压器安全阀接管设有安全端，安全端的一端设有位于稳压器安全阀内的热套管，另一端与稳压器安全阀的工艺管道连接；其中，安全端设有向下倾斜延伸进入稳压器内的弯管，弯管的末端在高度方向上位于弯管上内侧壁最低点的上方。相对于现有技术，本发明采用带热套管的稳压器安全阀接管安全端和弯管连接的水封结构，利用饱和水蒸汽通过稳压器安全阀、稳压器安全阀接管和安全端放热冷凝实现水密封，可防止氢气泄漏，且结构简单、制造方便。

Ceramics heat exchanger

A ceramic heat exchanger includes a heat exchange section that heat-exchanges between two fluids A and B flowing opposite directions to each other. The heat exchange section includes ceramic blocks stacked one on top of another with a seal therebetween. The ceramic blocks have a plurality of parallel lines of flow channels, each line defined by the flow channels through which the same fluid flows, any two adjacent lines being defined by the flow channels through which the different fluids A and B flow respectively. Both ends in the stacking direction of the stack are bound to join and integrate the ceramic blocks with tightening means including end plates and a tie rod. A thermal expansion absorber is disposed on an external surface of the end plates for absorbing thermal expansion in the axial direction of the tie rod.

Trash Rack for Filtering Trash of Power Plant

본 고안은 발전소 취수설비의 이물질 여과용 트래쉬 랙에 관한 것으로, 풍향 또는 취수의 역류에 의해 랙의 설치상태가 불안해져 요동치거나 충격을 받아 파손이 발생하지 않도록 한 발전소 취수설비의 이물질 여과용 트래쉬 랙에 관한 것이다. The present invention relates to a trash rack for filtration of foreign substances in a power plant intake facility. It is about. 본 고안의 구성은 통상의 이물질 걸름용 트래쉬 랙에 있어서, 상기 랙의 하부 바깥쪽 바닥면에 지지부재를 설치한 구조이다. The constitution of the present invention is a structure in which a support member is installed on a lower outer bottom surface of the rack in a conventional trap rack for foreign material filtering. 또한, 상기 지지부재는 하나의 일체화된 절곡된 수평면 및 수직면을 가지는 플레이트로 랙의 폭방향으로 설치될 수 있고, 또한 복수개 마련하여 고정시킬 수 있다. 또한, 지지부재의 고정시에는 절곡된 수평면에 체결수단을 관통하여 지면에 고정되도록 된 구조이다. In addition, the support member may be installed in a width direction of the rack as a plate having one integrated bent horizontal plane and vertical plane, and may be provided and fixed in plurality. In addition, when fixing the support member is a structure that is fixed to the ground through the fastening means to the bent horizontal surface. 이러한 구조를 가지는 본 고안은 별도의 지지수단을 마련하여 태풍과 같은 강한 바람이 불거나 역방향의 바람이 불더라도 트래쉬 랙의 설치상태를 견고하게 유지할 수 있고, 종래와 같이 요동에 의한 충격등에 의해 파손될 우려가 대폭 감소되도록 한 효과가 있다. The present invention having such a structure provides a separate support means to maintain the installation state of the trash rack even if strong winds such as typhoons or reverse winds, and can be damaged by impacts such as swinging as in the prior art. It has the effect of drastically reducing concerns.

A nuclear power plant radioactive liquid waste treatment system

本发明涉及一种核电厂放射性废液处理系统，包括工艺排水处理子系统，化学和地面排水处理子系统，服务排水处理子系统及再浓缩子系统，工艺排水处理子系统包括依次串联的工艺排水收集单元，絮凝注入和离子交换单元，反渗透单元及监测槽；化学和地面排水处理子系统包括化学排水收集单元和地面排水收集单元，还包括均与化学排水收集单元和地面排水收集单元下游依次串联的超滤和反渗透单元，离子交换单元及化学和地面排水复用槽；服务排水处理子系统包括依次串联的臭氧/紫外光处理单元，精处理单元及服务排水复用槽；再浓缩子系统包括浓缩蒸发单元。本发明可实现核电厂氚浓度较高废液的处理后排放以及氚浓度较低或基本不含氚废液的处理后复用的功能。