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.

ПОДЗЕМНЫЙ ЭНЕРГЕТИЧЕСКИЙ ЯДЕРНЫЙ РЕАКТОР С КАМЕРОЙ ГАШЕНИЯ УДАРНОЙ ВОЛНЫ

Изобретение относится к подземному энергетическому ядерному реактору. Ядерный реактор защищен защитной оболочкой, соединенной с удлиненным и пустотелым ударным туннелем, который проходит от одного конца защитной оболочки. Причем на втором конце упомянутой защитной оболочки подвижно расположена дверь, способная перемещаться из нормально закрытого положения в открытое положение, когда в ядерном реакторе возникает взрыв или ударная волна. Ударный туннель образует ударную камеру с множеством расположенных в ней разнесенных дефлекторов обломков. Ударная камера имеет верхнюю стенку с выполненным в ней сводовым проемом, который выборочно закрывается сводом. Техническим результатом является обеспечение возможности гашения ударной волны в случае взрыва подземного энергетического ядерного реактора, а также возможности перемещать реактор из его подземной защитной оболочки для ремонта или замены. 2 н. и 19 з.п. ф-лы, 14 ил.

СПОСОБ ЛОКАЛИЗАЦИИ И ОХЛАЖДЕНИЯ РАСПЛАВА АКТИВНОЙ ЗОНЫ ЯДЕРНОГО РЕАКТОРА

Изобретение относится к способу локализации и охлаждения расплава активной зоны ядерного реактора и может использоваться для обеспечения безопасности атомных электрических станций (далее - АЭС) при тяжелых авариях. В помещении фильтров предварительно устанавливают с кольцевым зазором по отношению к стенке шахты реактора стенку/перегородку высотою, соответствующей минимальному проектному уровню охлаждающей жидкости в шахте реактора как минимум с одним обратным клапаном в нижней ее части, обеспечивающим поступление жидкости из помещения фильтров в указанный зазор. Положение указанной стенки/перегородки в помещении фильтров выбирают таким образом, чтобы обеспечить условия равенства объемов первоначального объема жидкости в первом контуре АЭС, в компенсаторе давления и гидроемкостях системы аварийного охлаждения активной зоны. Над указанным образованным зазором располагают профилированный козырек с наклоном к упомянутой щели. При аварии обеспечивают первоначальное поступление жидкости, истекающей ...

Устройство для фиксации рабочего органа ядерного реактора

Полезная модель относится к устройствам для удержания и извлечения рабочего органа в активной зоне ядерного реактора. Устройство для фиксации рабочего органа ядерного реактора одержит тягу, на одном конце которой расположен механизм сцепления тяги с рабочим органом, а на другом - привод тяги, снабженный корпусом. На внутренней поверхности которого выполнен кольцевой выступ и закреплена обойма с прорезями, а также - разрезная втулка, лепестки которой расположены в прорезях обоймы и подпружинены. Тяга снабжена гайкой и пружиной. Гайка на тяге расположена ниже втулки с возможностью контакта с нижними торцами лепестков втулки. Пружина тяги закреплена на гайке с упором в кольцевой выступ корпуса. На внешней поверхности корпуса установлен магнитный двигатель. В чехле двигателя оппозитно друг другу установлены постоянные магниты с возможностью поворота относительно корпуса. Полезная модель позволяет повысить плотности электромагнитного поля для повышения силы притяжения лепестков разрезной втулки ...

Способ защиты ядерного реактора и предотвращения расплавления его корпуса при тяжелых авариях и устройство для его осуществления

Изобретение относится к средству предотвращения расплавления корпуса ядерного реактора в условиях высокоинтенсивных тепловых воздействий от расплавленных материалов активной зоны при тяжелой аварии. В способе защиты ядерного реактора на верхней поверхности ванны расплава формируют развитую поверхность теплообмена, состоящую из части верхней поверхности ванны расплава и поверхностей теплопроводных элементов, расположенных на верхней поверхности расплава. Устройство защиты ядерного реактора и предотвращения расплавления его корпуса при тяжелых авариях с формированием ванны расплава в корпусе реактора включает развитую поверхность теплообмена, состоящую, по крайней мере, из части верхней поверхности ванны расплава, а эта часть поверхности расплава находится между теплопроводными элементами, имеющими неотрицательную плавучесть в расплаве, и расположены на поверхности этой ванны расплава, и из тех частей внешних поверхностей этих элементов, которые расположены выше уровня этой части верхней ...

СПОСОБ ВОССТАНОВЛЕНИЯ БАРЬЕРОВ БЕЗОПАСНОСТИ В ПУНКТЕ РАЗМЕЩЕНИЯ РАДИОАКТИВНЫХ ОТХОДОВ

Изобретение относится к технологии улучшения или упрочнения грунта с помощью термических, электрических или электрохимических средств. Способ восстановления барьеров безопасности в пункте размещения радиоактивных отходов включает погружение электродов в область образования трещин и полостей в барьерном материале, создание электрического поля между электродами, подачу жидкости-носителя в область, примыкающую к электроду, перемещение жидкости-носителя от одного электрода к другому. Электроды погружают на границах области образования трещины или полости в барьерном материале, обеспечивающем безопасное захоронение твердых радиоактивных отходов. Барьерный материал, смешанный с жидкостью-носителем, подают в первый перфорированный электрод и инжектируют в прианодную область. Создают разность потенциалов между электродами, проталкивают барьерный материал в область образования трещины или полости. Жидкость-носитель, прошедшую между электродами и очищенную от барьерного материала, откачивают через ...

Ядерный реактор интегрального типа (варианты)

Заявлен ядерный реактор интегрального типа (варианты). Теплообменник размещен коаксиально с активной зоной в кольцевом пространстве, образованном между внутренней обечайкой, внутри которой размещены активная зона, входной и выходной коллекторы и защитная пробка, и разделительной обечайкой внутри корпуса реактора, формирующей опускной кольцевой канал и отделяющей нисходящий холодный поток от горячего восходящего потока теплоносителя. Причем теплообменник выполнен витым и секционированным по теплоносителю второго контура так, что трубки секций теплообменника сгруппированы во входных и выходных камерах теплоносителя второго контура, размещенных на патрубках на крышке реактора. Нижняя часть теплообменника размещена выше окон, выполненных во внутренней обечайке, через которые горячий теплоноситель поступает из выходного коллектора активной зоны на вход теплообменника, а холодный теплоноситель из верхней части теплообменника поступает непосредственно в кольцевую буферную емкость с уровнем теплоносителя ...

ЯДЕРНЫЙ РЕАКТОР С УЛУЧШЕННЫМ ОХЛАЖДЕНИЕМ В АВАРИЙНОЙ СИТУАЦИИ

... 1. Ядерный реактор, содержащий бак (4), в котором расположена активная зона реактора, первичный контур для охлаждения реактора, колодец (6) бака, в котором находится бак (4), кольцевой канал (16), окружающий нижнюю часть бака (4) в колодце (6) бака, средства, выполненные с возможностью заполнения колодца бака жидкостью, герметичный корпус (22) реактора, в котором расположены колодец бака и бак, характеризующийся тем, что содержит средства (26) сбора пара, генерируемого в верхнем конце колодца (6) бака, расположенные в герметичном корпусе и образующие объем, отделенный от объема герметичного корпуса (22), обеспечивая появление избыточного давления пара, средства (40), выполненные с возможностью создания принудительной конвекции жидкости в кольцевом канале (16), и средства (32, 42) для приведения в действие средств (40), выполненных с возможностью создания принудительной конвекции, при помощи указанного собранного пара. ! 2. Ядерный реактор по п.1, в котором средства (26), выполненные с возможностью ...

Фильтрующее устройство защиты от мусора системы аварийного охлаждения водо-водяного ядерного реактора

Изобретение относится к фильтрующему устройству защиты от мусора системы аварийного охлаждения водо-водяного ядерного реактора и может быть применено в атомной энергетике, в частности в системах аварийного охлаждения активной зоны водо-водяного ядерного реактора. Фильтрующее устройство (ФУ) содержит фильтрующие модули (ФМ), подсоединенные к коллектору, установленному на приемное отверстие трубопровода аварийной системы охлаждения, и снабжено рамой. Каждый фильтрующий модуль выполнен в виде совокупности типовых фильтрующих элементов, смонтированных параллельно друг другу стопкой с промежутками между соседними фильтрующими элементами на панели фильтрующего модуля. Гидравлически все фильтрующие элементы параллельно соединены друг с другом параллельно, при этом каждый фильтрующий элемент выполнен в виде полой конструкции с плоским каркасом в виде прямоугольной рамки и вмонтированным в каркас патрубком для отведения отфильтрованной жидкости. Толщина рамки определяет толщину фильтрующего элемента ...

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 ...

ПАССИВНЫЙ ВЕРТИКАЛЬНЫЙ И ГОРИЗОНТАЛЬНЫЙ ЗАГЛУШАЮЩИЕ УЗЛЫ ДЛЯ ПРЕДОТВРАЩЕНИЯ УТЕЧКИ (РАЗЛИВА) РАСПЛАВА ВНЕ ГЕРМОЗОНЫ ПРИ ТЯЖЕЛОЙ АВАРИИ НА АТОМНОЙ СТАНЦИИ

... 1. Пассивный вертикальный заглушающий узел для предотвращения разлива расплава по механизму раннего байпаса гермозоны/герметического объема при тяжелой аварии в ядерном реакторе атомной станции, состоящий из вертикальной цилиндрической центральной трубы (1), зафиксированной на стальной плите (2); причем вокруг трубы (1) размещена обсадная труба (3), вмонтированная в бетонную конструкцию (4.b), а в нижней части трубы (1) размещен монолитный биозащитный цилиндр, состоящий из двух полуцилиндрических сегментов (5.1) и (5.2); между двумя указанными сегментами (5.1) и (5.2) проходит центральный отверстие/канал (6), в котором проложен кабель/трос (8), характеризующийся тем, что пространство между обсадной трубой (3) и вертикальной трубой (1) заполнено бетоном (9), а над биозащитным цилиндром из двух сегментов (5.1) и (5.2) установлена заглушающая пробка из двух сегментов (7.1) и (7.2), плотно стянутых между собой скобами (12.1) и (12.2); причем канал (6) также проходит между сегментами (7.1) и ...

DEVICE FOR CLEANING GAS OF ADMIXTURES

Ядерная установка с реактором, охлаждаемым водой под давлением

Kernreaktoranlage mit einem auf dem Prinzip der Kondensation von Dampf beruhenden Druckunterdrueckungssystem

Receptacle construction for water nuclear reactor, uses porous mineral such as ceramic foam between receptacle and reactor core

The reactor includes a grid or horizontal perforated plate for the spreading and dispersion of corium (36). The reactor also contains a basin shaped receptacle (20) separated from the reactor core (12) to allow the circulation of water, the receptacle is partially made of refractory material. The receptacle contains a porous mineral (30) such as a ceramic foam that can lower the temperature of corium.

Improvements in or relating to an integral vapour generator-nuclear reactor

... 1,120,040. Nuclear power plant. BABCOCK & WILCOX CO. 24 May, 1967 [24 May, 1966], No. 24258/67. Heading G6C. In a marine nuclear power plant the pressurized water reactor and the steam generating heat exchangers are mounted in pressure vessels integral with each other. The circulating pumps, which are mounted upon, and outside of, the pressure vessels receive the primary coolant after cooling in the heat exchangers, thereby preventing cavitation and flashing. Vaporizing and superheating are effected in the heat exchangers and an electrically heated pressurizer is mounted within the reactor vessel above the core. In further embodiments, gate valves permit the isolation of defective exchanger tube banks and convection circulation of the secondary coolant is effected by an external drum connected to the exchanger and containing both water and vapour. Radioactive contamination of the secondary coolant is removed in the external drums. The exchanger pressure vessel surrounds the vertical reactor ...

Nuclear Reactor Installation.

... 1,195,165. Reactors. UNITED KINGDOM ATOMIC ENERGY AUTHORITY. 30 May, 1969 [21 May, 1968], No. 24243/68. Heading G6C. A nuclear reactor, in particular a gas-cooled reactor, has means for generating foam for blanketting any reactor component whose failure could result in loss of reactor coolant driving operation. In the case of a gas-cooled reactor the rate of coolant loss is slow and it is possible to erect tempoarary pens of wire mesh to hold the foam where it is required. Permanent pens may however be provided, e.g. the ends of standpipes can be situated in a recessed charge face with a raised surround to form a wall for the foam, or the charge face can be flat with recesses for posts to support a wire mesh structure for penning the foam in place. The recommended foam is a CO 2 -filled high expansion aqueous foam. The foam generating means can be a permanent installation with foam branches, or a portable foam generator. The latter will enable foam to be inserted into ruptured piping to ...

Improvements relating to nuclear reactors

... 958, 629. Nuclear reactors. UNITED KING- DOM ATOMIC ENERGY AUTHORITY. Jan. 9, 1963 [July 5, 1962], No. 25787/62. Heading G6C. A nuclear reactor comprises a reactor vessel, a reactor core within the vessel, a flowpath within the reactor vessel for liquid reactor coolant, and non-return valves within the reactor vessel to counter a reversal of coolanf flow in the flowpath. In the boiling water reactor shown in the drawing, fuel elements (not shown) are housed in fuel tubes 12 through which pressurised water is circulated as a primary coolant. The fuel tubes 12 are situated in a region surrounded by a baffle 13 and contained between a pot 14 and a support plate 27; in a secondary coolant path water flows downwards between the pot and the baffle and upwardly through the core between the fuel tubes. A thermal shield 15 has apertures 16 to permit downward flow of the secondary coolant. The fuel tubes 12 are extended upwards by extension tubes 32 in a heat transfer region 31, and the tubes are ...

Improvements in or relating to nuclear reactors

... 889,200. Nuclear reactors. UNITED KING- DOM ATOMIC ENERGY AUTHORITY. Jan. 12, 1959 [Feb. 12, 1958], No. 4664/58. Class 39(4) In a gas cooled reactor, the gas, when contaminated with fission products, may be passed from the coolant circuit through a reservoir to an external absorbing tower. As shown, the reactor 1 is cooled by the flow of carbon dioxide round the circuit formed of pressure vessel 2, ducts 4, 5, heat exchanger 6 and pump 7. The circuit is connected by the safety valve 9 with the gas-holder 13 and absorbtion tower 14. The latter comprises a continuously recirculated NaOH solution and is coupled to 13 by a closed circuit of pipes 16, 18 and pump 17. To purge the reactor after faulty fuel elements have been removed, the carbon dioxide reservoir 28 is connected to the coolant, circuit by valve 29. Fire extinguishing materials are contained in reservoir 26 coupled to the coolant circuit by valve 27.

GAS-COOLED NUCLEAR REACTOR SYSTEM

... 1519305 Mounting heat exchangers GENERAL ATOMIC CO 7 Aug 1975 [9 Aug 1974] 33036/75 Heading G6C A heat-exchanger tube sheet 11 mounted in an aperture 12 penetrating a pressure vessel 13 is prevented from moving upwards towards the interior of the pressure vessel upon failure of the mounting arrangement by a restraining ring 15. The tube sheet 11 and restraining ring 15 are both supported from an extension 14 of the metal liner 29 in the aperture by independent tubular extensions 17, 39. Sealing rings 51, 53 prevent leakage between the two extensions.

Power breeder reactor

... 857,959. Nuclear reactors. UNITED STATES ATOMIC ENERGY COMMISSION. Oct. 16, 1958 [Nov. 18, 1957], No. 33024/58. Class 39(4). In a fast neutron reactor, the effect of temperature upon reactivity is reduced by restricting the core distortion caused by bowing of the fuel and reflector tubes. Heating of the central section of each tube normally tends to produce bowing towards the reactor central axis which reduces the core diameter, increases reactivity and hence introduces instability. In the invention the mounting of each tube is arranged so that with increasing temperature the bowing produces first a decrease and then an increase in core diameter, thus reducing the reactivity. The reactor core is housed in a neutron flux attenuating shield and immersed in a liquid coolant within a tank. Fuel handling and control is effected vertically through the tank upper cover. The heat developed produces steam for a turbine generator. The fuel core is completely enclosed in reflector zones; the control ...

Nuclear reactor

... 1,035,606. Reactors. UNITED STATES ATOMIC ENERGY COMMISSION. April 30, 1965 [May 28, 1964], No. 18367/65. Heading G6C. In a fast breeder nuclear reactor, the core has a volume of 275 cubic feet, is annular and is cooled by sodium flowing vertically through it to form an upper reflector. The arrangement provides a negative reactivity coefficient upon loss of sodium from the system and upon voiding of the sodium in the reflector. The core is surrounded by an annular outer fertile blanket and encloses an inner, annular blanket. A further annular blanket is disposed below the core. The blankets and core are formed of vertical element tubes which contain assemblies of blanket and fuel pins respectively and which permit the upward flow of coolant through nozzles in the foot of each tube. Control elements pass through the core. The upper reflector is formed by an upper extension of each fuel tube which is filled with coolant sodium. Argon gas separates the upper surface of the coolant and the ...

Process and device to remove certain effects of pressure loss in channels of nuclear reactor traversed by a liquid.

SAFETY DEVICE FUR UNDER PRESSURE STANDING PLANTS

NUCLEAR REACTOR PRESSURE VESSEL

NUCLEAR REACTOR PRESSURE VESSEL

PROCEDURE FOR THE CLEANING FROM GASES AND DEVICE TO IT.

PRODUCTION DE CHALEUR AU MOYEN D'UN REACTEUR NUCLEAIRE

PRESSURE RELIEF AND SECONDARY CONTAINMENT SYSTEM

NUCLEAR REACTOR DOWNCOMER FLOW DEFLECTOR

A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

PRODUCTION DE CHALEUR AU MOYEN D'UN REACTEUR NUCLEAIRE

METHOD AND APPARATUS FOR PROTECTING THE CORE OF A NUCLEAR REACTOR

VVER EMERGENCY COOLING SYSTEM SUMP PROTECTION DEVICE, FILTER MODULE OF SUMP PROTECTION DEVICE.

The present invention pertains to the field of emergency protection systems of nuclear power plants. More particularly, the invention relates to Emergency Core Cooling System under loss-of-coolant accidents, namely, to sump protection device (SPD) in the emergency cooling system of a VVER, to the filter module and filter element of the sump protection device. The purpose of the invention is to protect sumps from accumulation of debris in case of a loss-of-coolant accident. As a solution to the problem, we claim a VVER emergency cooling system sump protection device, comprising a system of filters installed at the intake opening of the upper part of the sump located in the reactor containment bottom and connected to the intake of emergency cooling system pipeline. It consists of header-connected filter modules preventing debris from entering the intake of emergency cooling system pipelines; each filter module has slotted grates on sides and on top, and filter elements arranged inside are ...

INTEGRAL ISOLATION VALVE SYSTEMS AND METHODS OF OPERATING SAME FOR LOSS OF COOLANT ACCIDENT (LOCA) PROTECTION

A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

REACTOR MODULE SUPPORT STRUCTURE

A support structure (100) for attenuating seismic forces in one or more reactor modules (10) housed in a reactor building includes a mounting structure that may be configured to securely connect the support structure (100) to a floor of the reactor building. A receiving area (510) may be sized to receive a lower portion of a reactor module (10), and the support structure (100) may be configured to at least partially surround the lower portion of the reactor module (10) within the receiving area. The support structure (100) may further include a retention system located near a top surface of the support structure (100). The retention system may be configured to contact the reactor module (10) during a seismic event, and an upper portion of the reactor module (10) may extend above the retention system without contacting the support structure (100).

PROCESS AND APPARATUS FOR TREATING A GAS STREAM

A process of treating hydrogen gas liberated from the acid or alkaline dissolution of a metal is provided. The process comprises a step of passing the liberated hydrogen gas through a reactor containing an oxidising agent for oxidation of the hydrogen gas into water, followed by a step of regenerating the oxidising agent. Also provided is an apparatus for carrying out the process, the apparatus comprising a reactor containing the oxidising agent, wherein the reactor is at least partially immersed in an alumina bath.

CLOSURE SYSTEM

Sicherheitsvorrichtung für einen Siedewasserreaktor

Dispositif de mise en communication rapide de deux réservoirs soumis à des pressions différentes

Gruppo costituito da un reattore nucleare ed un generatore di vapore conglobati, particolarmente per applicazioni marine

Sicherheitsventilanordnung für ein verseuchtes Gas, insbesondere in Kernreaktoranlagen

Dispositif de limitation à une valeur prédéterminée du débit d'un fluide réfrigérant dans un tube de force de réacteur atomique

Gehäuseanordnung mit einem Kernreaktor

Atomreaktor

Absperrorgan

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

Sicherheitskühlvorrichtung für den Fall des Einschmelzens des Kerns eines Kernreaktors

Safety device operating system - for press. sensitive units in eg nuclear reactors

Operating assembly in a safety device for a pressure medium influenced unit e.g. a nuclear reactor having >=2 actuable control valves which are incorporated in an operating pipeline to the unit, the pipeline is divided into two parallel pipes while each control valve is provided with two parallel passage ways each connected to one of the parallel pipes. In one position the control valves are open, and in the other position they are switched over in such a way that the inlet of the first passageway is connected to the outlet of the other passageway, and simultaneously, the outlet of the first passageway is connected to atmos. Control valves may be tested without need for shut down of system.

Réacteur nucléaire à modérateur liquide

Verfahren zur Erfassung von Schadensfällen bei Kernreaktoren und Einrichtung zur Durchführung des Verfahrens

Dispositif de limitation de débit, notamment pour réacteur nucléaire

Dispositif électronique de sécurité

Primärkühlmittel-Umwälzvorrichtung für Kernreaktoren

Abgasverbrennungssystem für Kernreaktoren

Dispositif de sécurité pour réacteur nucléaire à tubes de force

Réacteur nucléaire à modérateur et réfrigérant liquides

Kernreaktordruckkessel und Verwendung eines solchen

Gas- oder dampfgekühlter Kernreaktor

Kernkraftanlage

Procédé de production de chaleur utilisant un réacteur nucléaire hétérogène

DRUCKENTLASTUNGS- UND SICHERHEITSEINRICHTUNG.

KERNREAKTORANLAGE.

MECHANISM FOR CLOSING TWO OF COAXIAL PIPINGS OFF.

System for hydrogen injection during start-up/shutdown and boiling water reactors (the SWR) method for.

Die Erfindung betrifft ein System (30) und ein Verfahren zum Einspritzen von Wasserstoff in ein Siedewasserreaktor-(SWR)-Unterstützungssystem in Betrieb während des Anfahrens und/oder Abschaltens des Reaktors, um eine interkristalline Spannungsrisskorrosion (Inter-Granular Stress Corrosion Cracking [IGSCC]) zu verringern. Das System (30) stellt Wasserstoff bei variablen Drücken bereit (einschliesslich relativ höherer Drücke), die auf sich ändernde Betriebsdrücke der Reaktorunterstützungssysteme abgestimmt sind, während der Reaktor durch Anfahr- und Abschaltmodi läuft.

СПОСОБ ЗАЩИТЫ ЯДЕРНЫХ РЕАКТОРОВ И УСТРОЙСТВО ДЛЯ ЕГО ОСУЩЕСТВЛЕНИЯ

... 1. Способ защиты ядерных реакторов, характеризующийся тем, что располагают непосредственно на и/или вблизи от средства, предназначенного для подачи теплоносителя в ядерный реактор, перед входом в ядерный реактор или его активную зону по крайней мере одно средство для создания магнитного поля, предназначенного для взаимодействия с элементарными частицами с магнитным зарядом, и/или по крайней мере одно средство для создания электрического поля, предназначенного для взаимодействия с элементарными частицами с магнитным зарядом. 2. Способ по п.1, отличающийся тем, что в качестве средства для создания магнитного поля используют по крайней мере одну систему постоянных и/или электромагнитов, расположенных вдоль средства, предназначенного для подачи теплоносителя. 3. Способ по п.1, отличающийся тем, что в качестве средства для создания магнитного поля используют по крайней мере одну систему кольцевых постоянных и/или электромагнитов, расположенных вдоль средства, предназначенного для подачи теплоносителя ...

Passive vertical and horizontal plugging assemblies to prevent melt outflow from the containment in case of severe accident in a nuclear power plant

Improvements with the closures of help

Nuclear reactor

NUCLEAR REACTOR WATER-COOLED

Device intended to quickly put in communication two tanks subjected to different pressures

Nuclear reactors with flow forced of the cooling agent

Safety device for tube nuclear reactor of force

Outfall safety device for subsidiary shell engine with water

Installations of pressurized nuclear reactors

Pressurized-water reactor coolant pipe containment

Safety device for use in nuclear power station, has electric source supplying electricity to electric batteries to feed safety elements in control room, where source utilizes electrolytic installation unit for storing oxygen and hydrogen

La présente invention concerne des dispositifs capables d'alimenter les organes de sûreté et de sécurité des centrales ou installations nucléaires en cas de perte totale des alimentations électriques, et donc d'augmenter la sûreté et la sécurité nucléaire. Elle propose notamment au minimum trois circuits de secours électrique. Elle propose aussi l'ajout de matériels (notamment n° 7, 10, 15 de la planche 1 ci jointe) complétant et améliorant la sécurité et la sûreté nucléaire. Le dispositif, objet de l'invention, illustré par la Planchel ci jointe, peut s'appliquer aux centrales et installations nucléaires existantes mais aussi être intégré dans la conception de nouvelles installations.

Breeder engine of power

CLOSURE FOR NUCLEAR REACTOR HAS PREGNANT PRESSURE

Improvements with the generating stations of energy to nuclear reactor

IMAGE ACQUISITION DEVICE AND METHOD FOR ACQUIRING IMAGE IN VAPOR ENVIRONMENT BY USING IMAGE ACQUISITION DEVICE

The present invention relates to an image acquisition device comprising: a camera lens unit which includes a lens; and a moisture and vapor removal unit which removes moisture and vapor from the lens, wherein the moisture and vapor removal unit includes a compressed air forming unit which compresses air to form compressed air, a high-temperature air injection unit which heats the compressed air and injects the heated compressed air onto the lens, a driving unit which supplies driving current to the compressed air forming unit and the high-temperature air injection unit, and a control unit which controls the driving unit. In addition, the present invention relates to a method for acquiring an image in a vapor environment, comprising: a step of compressing air to prepare compressed air; a step of heating the compressed air to prepare heated compressed air; a step of applying the heated and compressed air to the lens; and a step of acquiring an image by using the lens. COPYRIGHT KIPO 2017 ...

A METHOD FOR OPTIMIZING OPERATING MARGIN IN A NUCLEAR REACTOR

Overpower and overtemperature turbine runback and reactor trip (OPDT, OTDT) on pressurized water reactors (PWR) use the temperature difference between hot and cold leg temperature measurements. The difference between filtered hot leg temperature and lead/lag compensated cold leg temperature is used to maximize OPDT margin to hot leg fluctuations, while enabling safety margin increase for secondary accidents mitigated OPDT.

INTEGRAL VESSEL ISOLATION VALVE

A nuclear reactor comprises a nuclear reactor core disposed in a pressure vessel. An isolation valve protects a penetration through the pressure vessel. The isolation valve comprises: a mounting flange connecting with a mating flange of the pressure vessel; a valve seat formed into the mounting flange; and a valve member movable between an open position and a closed position sealing against the valve seat. The valve member is disposed inside the mounting flange or inside the mating flange of the pressure vessel. A biasing member operatively connects to the valve member to bias the valve member towards the open position. The bias keeps the valve member in the open position except when a differential fluid pressure across the isolation valve and directed outward from the pressure vessel exceeds a threshold pressure.

BASKET AND PH ADJUSTING DEVICE

Disclosed is a basket (50) that is disposed within a basket housing vessel, into which a boric acid solution capable of dissolving a pH adjuster flows, and can allow a pH adjusted solution to outflow by the inflow boric acid solution. The basket (50) comprises a plurality of storage parts (71) which are stacked on top of each other in a vertical direction while providing a predetermined first space (L1) therebetween and into which the pH adjuster is stored. Also disclosed is a pH adjusting device comprising the basket (50), a basket housing vessel, in which the basket (50) can be housed and in which cooling water can be stored, and an overflow pipe that, within the basket housing vessel, allows a pH adjusted solution comprising a pH adjuster dissolved in the cooling water to outflow.

SYSTEM FOR HYDROGEN INJECTION FOR BOILING WATER REACTORS (BWRs) DURING STARTUP / SHUTDOWN

A system for injecting hydrogen into Boiling Water Reactor (BWR) reactor support systems in operation during reactor startup and/or shutdown. The system the hydrogen injection system includes at least one hydrogen source, flow control equipment, and pressure control equipment. The pressure control equipment being configured to regulate a pressure of a hydrogen flow between the at least one hydrogen source and the at least one first BWR support system based upon an operating pressure of the at least one first BWR support system. 1. (canceled)2. A system , comprising:at least one first BWR support system; and at least one hydrogen source,', 'flow control equipment,', 'pressure control equipment,, 'a hydrogen injection system fluidly connected to the at least one first BWR support system, the hydrogen injection system including,'}the at least one first BWR support system being a system that operates during a reactor startup mode or a reactor shutdown mode,the pressure control equipment being configured to regulate a pressure of a hydrogen flow between the at least one hydrogen source and the at least one first BWR support system based upon an operating pressure of the at least one first BWR support system.3. The system of claim 2 , wherein the at least one first BWR support system experiences a reactor water fluid flow through the at least one first BWR support system during the reactor startup mode or the reactor shutdown mode.4. The system of claim 2 , wherein the at least one first BWR support system is at least one of a Reactor Water Cleanup (RWCU) return line or a Feedwater Recirculation line.5. The system of claim 2 , wherein the pressure control equipment is further configured to match the pressure of the hydrogen flow to the operating pressure of the at least one first BWR support system claim 2 , the operating pressure of the at least one first BWR support system being variable during the reactor startup mode or the reactor shutdown mode.6. The system of claim ...

ARRANGEMENT FOR PREVENTING SHUTDOWN OF A NUCLEAR REACTOR PLANT WHEN THE LOAD ON A TURBINE DRIVEN THEREBY DECREASES SUDDENLY

Method and apparatus for preventing inadvertent criticality in a nuclear fueled electric powering generating unit

An inadvertent approach to criticality in a nuclear fueled electric power generating unit is detected and an alarm is generated through on-line monitoring of the neutron flux. The difficulties of accurately measuring the low levels of neutron flux in a subcritical reactor are overcome by the use of a microcomputer which continuously generates average flux count rate signals for incremental time periods from thousands of samples taken during each such period and which serially stores the average flux count rate signals for a preselected time interval. At the end of each incremental time period, the microcomputer compares the latest average flux count rate signal with the oldest, and preferably each of the intervening stored values, and if it exceeds any of them by at least a preselected multiplication factor, an alarm is generated. The interval and multiplication factor are chosen such that an alarm is generated early enough in the event to provide adequate time for an automatic system or ...

BASKET AND PH ADJUSTING DEVICE

PROBLEM TO BE SOLVED: To provide a basket and pH adjusting device which can improve the dissolution rate of a pH regulator. SOLUTION: The basket 50 that is placed in a basket accommodation container into which a boric acid solution that can dissolve the pH regulator flows and where the flowing-in boric acid solution allows the pH controlled solution to flow out includes accommodation sections 71 that can accommodate the pH regulator by piling it up in the vertical direction with first prescribed interstices L1. The device also includes the basket 50, the basket accommodation container that is constituted so that it can accommodate the basket 50 inside and can reserve cooling water inside and an overflow pipe for allowing the pH controlled solution where the pH regulator is dissolved in the cooling water to flow out in the basket accommodation container. COPYRIGHT: (C)2010,JPO&INPIT ...

Method for the Pressure Relief of a Nuclear Power Plant, Pressure-Relief System for a Nuclear Power Plant and Associated Nuclear Power Plant

A method and a corresponding device for the pressure relief of a nuclear power plant having an outlet for a relief flow. The relief flow is guided out of a containment into the atmosphere via a relief line provided with a filter system. The filter system has a filter chamber with a filter-chamber inlet and outlet and a sorbent filter arranged therebetween. The relief flow is guided in a high-pressure section of the relief line past the filter chamber, with the latter being heated, and the relief flow is expanded at the end of the high-pressure section and dried. In order for efficient retention of iodine-containing organic compounds, the relief flow is guided through a bed filter, guided in a superheating section past the high-pressure section of the relief line and in the process is heated, guided in this state directly thereafter through the filter chamber having the sorbent filter.

Nuclear power station

A nuclear power station has a containment in which a reactor core is accommodated. According to the invention, an external cooling system for cooling the containment in the event of an accident is associated with the containment. The cooling system in particular has a coolant reservoir that is configured as a lake or is lake-like, and in which the containment is in contact with a coolant or may be brought into contact with a coolant, in particular a liquid coolant, in the event of an accident.

Passive emergency feedwater system

A power module assembly includes a reactor vessel containing a reactor core surrounded by a primary coolant. A containment vessel is adapted to be submerged in a containment cooling pool and to prohibit a release of the primary coolant outside of the containment vessel. A secondary cooling system is configured to remove heat generated by the reactor core. The heat is removed by circulating liquid from the containment cooling pool through the primary coolant.

METHOD OF REDUCING CORROSION OF NUCLEAR REACTOR STRUCTURAL MATERIAL

In a method of reducing corrosion of a material constituting a nuclear reactor structure, an electrochemical corrosion potential is controlled by injecting a solution or a suspension containing a substance generating an excitation current by an action of at least one of radiation, light, and heat existing in a nuclear reactor, or a metal or a metallic compound forming the substance generating the excitation current under the condition in the nuclear reactor to allow the substance generating the excitation current to adhere to the surface of the nuclear reactor structural material, and by injecting hydrogen in cooling water of the nuclear reactor while controlling the hydrogen concentration in a feed water. 1. A method of reducing corrosion of a material constituting a nuclear reactor structure comprising the steps of:applying a substance generating an excitation current and a noble metal to a surface of a material constituting a nuclear reactor structure in advance; and{'sub': 2', '2, 'controlling a concentration of oxidizing chemical species and a concentration of reducing chemical species in a nuclear reactor water so that a molar ratio of H/Ois less than a value of 2 in which a catalytic reaction to recombine the oxidizing chemical species with the reducing chemical species is not accelerated by the noble metal.'}2. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 1 , wherein the substance generating the excitation current is at least one of compounds selected from TiO claim 1 , ZrO claim 1 , ZnO claim 1 , WO claim 1 , PbO claim 1 , BaTiO claim 1 , BiO claim 1 , SrTiO claim 1 , FeO claim 1 , FeTiO claim 1 , KTaO claim 1 , MnTiO claim 1 , SnO claim 1 , and NbO.3. The method of reducing corrosion of a material constituting a nuclear reactor structure according to claim 1 , wherein the noble metal is at least one of elements selected from Pt claim 1 , Pd claim 1 , Ir claim 1 , Rh claim 1 , Os claim 1 , and Ru ...

Nuclear Power Plant

A nuclear power plant has a reactor pressure vessel, a primary containment vessel and a passive pressure suppression pool cooling system. The reactor pressure vessel is installed in the primary containment vessel. A pressure suppression pool filled with cooling water is formed in a lower portion of the primary containment vessel. The passive pressure suppression pool cooling system is provided with a steam condensing pool in which cooling water is filled, disposed outside the primary containment vessel, a steam condenser disposed in the steam condensing pool, a steam supply pipe connecting the reactor pressure vessel to the steam condenser, and a condensed water discharge pipe connected to the steam condenser for discharging condensed water generated in the steam condenser. Another end portion of the condensed water discharge pipe is disposed in the pressure suppression pool. 1. A nuclear power plant comprising:a primary containment vessel; a reactor pressure vessel installed in the primary containment vessel; a pressure suppression pool in which first cooling water is filled for reducing pressure increase in the primary containment vessel, installed in a lower portion of the primary containment vessel; and a passive pressure suppression pool cooling system,Wherein the passive pressure suppression pool cooling system has a steam condensing pool in which second cooling water is filled, disposed outside the primary containment vessel; a steam condenser disposed in the steam condensing pool; a steam supply pipe connecting the reactor pressure vessel to the steam condenser; and a condensed water discharge pipe connected to the steam condenser for discharging condensed water generated in the steam condenser, and;wherein another end portion of the condensed water discharge pipe is disposed in the pressure suppression pool.2. A nuclear power plant comprising:a primary containment vessel; a reactor pressure vessel installed in the primary containment vessel; and a pressure ...

Hydrogen venting device for cooling water of nuclear reactors

A hydrogen venting device for separating and releasing hydrogen gas from a gaseous mixture comprising hydrogen and steam generated in nuclear power plants is disclosed. The method includes providing a chamber at a high point in the cooling water circuit, allowing the collection chamber to cool below a gaseous mixture inlet temperature thereby allowing the hydrogen to collect at a first elevation within the collection chamber and condensate of the steam to collect at a second elevation within the collection chamber below the first elevation, and releasing substantially only hydrogen from the collection chamber at or proximal the first elevation when a threshold temperature, less than the first temperature, is reached.

METHOD FOR DEPRESSURIZING A NUCLEAR POWER PLANT, DEPRESSURIZATION SYSTEM FOR A NUCLEAR POWER PLANT, AND ASSOCIATED NUCLEAR POWER PLANT

A method and a device depressurize a nuclear power plant. A depressurization flow is conducted out of a containment shell into the atmosphere via a depressurization line having a filter system. The filter system contains a filter chamber having an inlet, an outlet, and a sorbent filter. The depressurization flow is first conducted in a high-pressure section, then is depressurized by expansion at a throttle device, then conducted through the filter chamber having the sorbent filter, and finally blown out. To enable an effective retention of activity carriers contained in the depressurization flow, including organic compounds containing iodine, the depressurization flow depressurized by the throttle device is conducted through a superheating section before the depressurization flow enters the filter chamber, in which superheating section the depressurization flow is heated from the not yet depressurized depressurization flow to a temperature that is at least 10 ° C. above the dew point temperature. 1. A method for depressurizing a nuclear power plant including a containment shell for containing activity carriers and having an outlet for a depressurization flow , the depressurization flow conducted out of the containment shell into the atmosphere via a depressurization line being provided with a filter system , the filter system containing a filter chamber having a filter chamber inlet , a filter chamber outlet and a sorbent filter lying there-between , which comprises the steps of:first conducting the depressurization flow in a high-pressure section of the depressurization line;depressurizing the depressurization flow by means of expansion at a throttle device;immediately before the depressurization flow enters the filter chamber, conducting the depressurization flow that has been depressurized by the throttle device through a superheating section, in which the depressurization flow is heated by direct or indirect heat transfer from a not yet depressurized ...

PASSIVE COOLING AND DEPRESSURIZATION SYSTEM AND PRESSURIZED WATER NUCLEAR POWER PLANT

A passive cooling and depressurization system for a pressurized water nuclear plant is provided with a cooling water pool, a steam supply piping, a heat exchanger, a steam supply valve, a coolant return pipe and an outlet valve. The steam supply piping extends from the gas phase of the pressurizer. The heat exchanger exchanges heat between water stored in the cooling water pool and steam flowing through the steam supply piping. The steam supply valve is equipped on the steam supply piping. The coolant return pipe extends from the heat exchanger to a liquid phase of the reactor pressure boundary. The outlet valve is equipped on the coolant return pipe. 1. A passive cooling and depressurization system for a pressurized water nuclear plant having a reactor pressure vessel for containing a reactor core cooled by primary coolant , a steam generator connected to the reactor pressure vessel by a hot leg pipe and a cold leg pipe , and a containment vessel containing the reactor pressure vessel , the steam generator , the hot leg pipe and the cold leg pipe , the passive cooling and depressurization system comprising:a pressurizer connected to the hot leg pipe by a riser for pressurizing an inside of a reactor pressure boundary where the primary coolant flows;a cooling water pool;a heat exchanger installed in the cooling water pool including an upper header, a lower header and a heat exchanger tube;a steam supply piping extending from a as phase of the pressurizer to the upper header of the heat exchanger;a steam supply valve equipped on the steam supply piping;a coolant return pipe extending from the heat exchanger to a liquid phase of the reactor pressure boundary; andan outlet valve equipped on the coolant return pipe,wherein the heat exchanger exchanges heat between water stored in the cooling water pool and steam supplied through the steam supply piping.2. The passive cooling and depressurization system of claim 1 , wherein the steam supply valve is a steam regulator ...

Containment vessel and nuclear power plant

A containment vessel has an inner shell covering a reactor pressure vessel and an outer shell forming an outer well which is a gas-tight space covering the horizontal outer periphery of the inner shell. The inner shell has a first cylindrical side wall surrounding the horizontal periphery of the reactor pressure vessel, a containment vessel head which covers the upper part of the reactor pressure vessel, and a first top slab connecting in a gas-tight manner the periphery of the containment vessel head and the upper end of the first cylindrical side wall. The outer shell has a second cylindrical side wall surrounding the outer periphery of the first cylindrical side wall, and also has a second to slab connecting in a gas-tight manner the vicinity of the upper end of the second cylindrical side wall and the first cylindrical side wall.

ISLAND MODE FOR NUCLEAR POWER PLANT

A nuclear power plant comprises a pressurized water reactor (PWR) and a steam generator driving a turbine driving an electric generator. A condenser condenses steam after flowing through the turbine. Responsive to a station blackout, the nuclear power plant is electrically isolated and a bypass valve is opened to convey bypass steam flow from the steam generator to the condenser without flowing through the turbine. The thermal power output of the PWR is gradually reduced over the transition time interval. After opening, the bypass valve is gradually closed over the transition time interval. A supplemental bypass valve may also be opened responsive to the station blackout to convey supplemental bypass steam flow from the steam generator to a feedwater system supplying secondary coolant feedwater to the steam generator. The supplemental bypass steam flow does not flow through the turbine and does not flow through the condenser. 1. A nuclear power plant comprising:a nuclear reactor comprising a pressurized water reactor (PWR) and a steam generator configured to transfer heat from primary coolant water heated by the PWR to secondary coolant water in order to convert the secondary coolant water to steam;a turbine connected with the steam generator to be driven by steam output by the steam generator;an electric generator connected with the turbine to be driven by the turbine to generate electricity;an electrical switchyard receiving electrical power from the electrical power generator during normal operation of the nuclear power plant;a condenser connected with the turbine to condense steam exiting the turbine; anda turbine bypass system configured to transfer a quantity of steam output by the steam generator to the condenser without passing through the turbine responsive to loss of offsite electrical power to the nuclear power plant wherein the quantity of steam transferred to the condenser without passing through the turbine is effective to (1) allow the nuclear reactor ...

DEFENSE IN DEPTH SAFETY PARADIGM FOR NUCLEAR REACTOR

A nuclear reactor includes a nuclear reactor core disposed in a pressure vessel and immersed in primary coolant water at an operating pressure higher than atmospheric pressure. A containment structure contains the nuclear reactor. A reactor coolant inventory and purification system (RCI) is connected with the pressure vessel by make-up and letdown lines. The RCI includes a high pressure heat exchanger configured to operate responsive to a safety event at the operating pressure to remove heat from the primary coolant water in the pressure vessel. An auxiliary condenser located outside containment also removes heat. The RCI also includes a pump configured to inject make up water into the pressure vessel via the make-up line against the operating pressure. An emergency core cooling system (ECC) operates to depressurize the nuclear reactor only if the RCI and auxiliary condenser are unable to manage the safety event. 1. A method comprising:operating a nuclear reactor disposed in a containment structure and including a nuclear reactor core comprising fissile material disposed in a pressure vessel and immersed in primary coolant water at an operating pressure higher than atmospheric pressure, the operating including maintaining primary coolant water level in the pressure vessel using a reactor coolant inventory and purification system connected with the pressure vessel by make-up and letdown lines; and shutting down the nuclear reactor core by scramming a control rod and', 'dissipating heat generated by the nuclear reactor core after shutting down using a high pressure decay heat removal component of the reactor coolant inventory and purification system that is connected to the pressure vessel by the make-up and letdown lines of the reactor coolant inventory and purification system., 'responding to a safety event by response operations including2. The method of wherein the response operations do not include depressurizing the nuclear reactor.3. The method of wherein the ...

NUCLEAR POWER PLANT AND PASSIVE CONTAINMENT COOLING SYSTEM

According to an embodiment, a nuclear power plant has a core; a reactor pressure vessel; a dry well; a wet well; a vacuum breaker; a containment vessel including the dry well, the LOCA vent pipe, the wet well, and the vacuum breaker; a cooling water pool placed outside the containment vessel; a heat exchanger at least partially submerged in cooling water; a gas supply pipe connected to the inlet plenum of the heat exchanger and the dry well; a condensate return pipe connected to the outlet plenum of the heat exchanger and the containment vessel; and a gas vent pipe connected to the outlet plenum of the heat exchanger and an outside of the wet well so that non-condensable gas inside the heat exchanger is released out of the wet well. The gas vent pipe is not connected to the wet well. 1. A nuclear power plant , comprising:a core;a reactor pressure vessel that houses the core;a dry well that houses the reactor pressure vessel;a wet well whose lower portion houses a suppression pool that is connected to the dry well via a LOCA vent pipe, and whose upper portion includes a wet well gas phase;a vacuum breaker that allows gas inside the wet well gas phase to flow back into the dry well;a containment vessel that includes the dry well, the LOCA vent pipe, the wet well, and the vacuum breaker;a cooling water pool that is placed outside the containment vessel and stores cooling water;a heat exchanger that includes an inlet plenum, an outlet plenum, and a plurality of heat exchanger tubes connecting the inlet plenum and the outlet plenum and being at least partially submerged in the cooling water;a gas supply pipe whose one end is connected to the inlet plenum of the heat exchanger, and whose other end is connected to the dry well so that gas in the dry well is led to the heat exchanger;a condensate return pipe whose one end is connected to the outlet plenum of the heat exchanger, and whose other end is connected to the containment vessel so that condensate inside the heat ...

NUCLEAR TECHNOLOGY PLANT AND METHOD FOR THE PRESSURE RELIEF OF A NUCLEAR TECHNOLOGY PLANT

A nuclear plant has a containment shell and a pressure relief pipe connected thereto in which a blowing device and a Venturi washer placed in a container with a washing liquid are connected in series. Even the finest particles or aerosols carried by air are held in the Venturi washer with a very high degree of reliability and the release thereof in environment is excluded in a particularly reliable manner in the case of decompression even associated with seal failures. For this purpose, the size of the blowing device and the Venturi washer are selected in such a way that during the operation of the blowing device a flow rate of liquid in the Venturi washer flowing to the decompressing pipe is higher than 130 m/sec, preferably higher than 180 m/sec. 1. A nuclear plant , comprising:a containment;a pressure relief line communicating with said containment and enabling pressure relief in said containment by blowing off a pressure relief gas;a blower device and a venturi scrubber connected in series in said pressure relief line, said venturi scrubber being disposed in a container with a scrubbing liquid;said blower device and said venturi scrubber being dimensioned to establish in said venturi scrubber, in an operating state of said blower device, a flow velocity of the pressure relief gas conveyed in said pressure relief line of more than 130 m/s;said blower device connected upstream from said venturi scrubber;said venturi scrubber including a venturi tube being passively fed with a scrubbing liquid due to a negative pressure at the constriction of said venturi tube, andsaid venturi tube is formed with an entry region fed with the scrubbing liquid.2. The nuclear plant according to claim 1 , wherein said blower device and said venturi scrubber are dimensioned to establish a flow velocity of the pressure relief gas of more than 180 m/s in said venturi scrubber.3. The nuclear plant according to claim 1 , wherein said blower device is a radial fan with a rated speed of more ...

SYSTEM AND METHOD FOR IMPLEMENTING UNIFIED COMPUTER-BASED MANAGEMENT OF FIRE SAFETY-RELATED RISK AND COMPENSATORY MEASURES MANAGEMENT IN NUCLEAR POWER PLANTS

A computer-implemented system and method for managing operations in a nuclear power plant generates an electronic request for a permit to perform work in the plant, performs a risk assessment using a rules engine to determine a level of fire risk posed by the work, automatically determines one or more compensatory measures to provide protection against the level of fire risk posed by work, generates a risk score based the probabilistic assessment, and generates electronic authorization for the permit based on the risk score. 125-. (canceled)26. A computer implemented method of determining whether to approve a work permit in a nuclear power plant , comprising:receiving an electronic permit request for a permit to perform work in an area of the plant;determining a quantitative fire risk value associated with the work identified in the permit request;comparing the determined fire risk value to a predetermined quantitative threshold fire risk value associated with the area in which the work will be performed; andgenerating automatic electronic authorization for the permit if the determined fire risk value does not exceed the threshold fire risk value.27. (canceled)28. The method of claim 53 , wherein the step of electronically determining at least one compensatory measure comprises electronically checking the status of fire detection equipment in and adjacent to the area of the plant which the work listed in the permit request will occur.29. The method of claim 53 , wherein the step of electronically determining at least one compensatory measure comprises electronically checking the status of fire suppression equipment in and adjacent to the area of the plant in which the work listed in the permit request will occur.30. The method of claim 53 , wherein the step of electronically determining at least one compensatory measure comprises electronically checking the status of combustible transit permits that have been issued for plant areas adjacent to the area of the plant ...

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 ...

LOSS-OF-COOLANT ACCIDENT REACTOR COOLING SYSTEM

A nuclear reactor cooling system with passive cooling capabilities operable during a loss-of-coolant accident (LOCA) without available electric power. The system includes a reactor vessel with nuclear fuel core located in a reactor well. An in-containment water storage tank is fluidly coupled to the reactor well and holds an inventory of cooling water. During a LOCA event, the tank floods the reactor well with water. Eventually, the water heated by decay heat from the reactor vaporizes producing steam. The steam flows to an in-containment heat exchanger and condenses. The condensate is returned to the reactor well in a closed flow loop system in which flow may circulate solely via gravity from changes in phase and density of the water. In one embodiment, the heat exchanger may be an array of heat dissipater ducts mounted on the wall of the inner containment vessel surrounded by a heat sink. 1. A passive reactor cooling system usable after a loss-of-coolant accident , the system comprising:a containment vessel comprising a wall in direct thermal communication with an external heat sink;a reactor well disposed inside the containment vessel;a reactor vessel disposed at least partially in the reactor well, the reactor vessel containing primary coolant and a nuclear fuel core heating the primary coolant which is circulated between the reactor vessel and a steam generator in a closed primary coolant flow loop;a cooling water tank disposed inside the containment vessel and containing an inventory of emergency cooling water in selective fluid communication with the reactor well via at least one flow control apparatus, the flow control apparatus having a closed position preventing flow of cooling water to the reactor well and an open position providing flow of cooling water to the reactor well; anda heat exchanger attached to an inside surface of the wall of the containment vessel, the heat exchanger in fluid communication with the reactor well and water tank via a closed ...

Venting system for the containment of a nuclear plant and method of operating the venting system

A pressure-relief system for a containment of a nuclear plant has a pressure-relief line which is led through the containment and is closed by a shutoff device, and a wet scrubber being switched into the pressure-relief line lying outside the containment, for the pressure-relief gas flow developing in the pressure-relief operating mode with the shutoff device being open. An effective, reliable operation of the wet scrubber with a compact structural configuration is made possible. This is achieved by a reservoir, arranged in the containment or fluidically connected therewith such that an overpressure, as compared with the outer environment, present in the containment, is transferred to the reservoir, and a feeding line which is led from the reservoir to the wet scrubber and can be closed by a shutoff device, for feeding a liquid active as a scrubbing liquid from the reservoir to the wet scrubber.

Pressure relief system for the containment of a nuclear power facility, nuclear power facility and method of operating a pressure relief system

A pressure-relief system for the containment of a nuclear power facility allows reliable operation of a wet scrubber for the pressure relief flow with a simultaneously compact structural design. The pressure relief system has a pressure relief line guided through the containment and can be closed by a shut-off valve, a wet scrubber arranged in a portion of the pressure relief line located inside the containment, for the pressure relief flow which forms in the pressure-relief mode when the shut-off valve is open, a reservoir arranged inside the containment and is fluidically connected to the remaining inner space of the containment such that any overpressure, with respect to the surroundings outside the containment, prevailing in the containment is transferred at least in part to the reservoir, and a supply line leading from the reservoir to the wet scrubber for supplying the wet scrubber with fluid from the reservoir.

INTEGRAL VESSEL ISOLATION VALVE

A nuclear reactor comprises a nuclear reactor core disposed in a pressure vessel. An isolation valve protects a penetration through the pressure vessel. The isolation valve comprises: a mounting flange connecting with a mating flange of the pressure vessel; a valve seat formed into the mounting flange; and a valve member movable between an open position and a closed position sealing against the valve seat. The valve member is disposed inside the mounting flange or inside the mating flange of the pressure vessel. A biasing member operatively connects to the valve member to bias the valve member towards the open position. The bias keeps the valve member in the open position except when a differential fluid pressure across the isolation valve and directed outward from the pressure vessel exceeds a threshold pressure. 1. A system comprising:at least one coolant pump configured to pump coolant water into or out of an associated nuclear reactor vessel;at least one external coolant conduit connecting said at least one coolant pump with the associated nuclear reactor vessel; anda vessel isolation valve having a mounting flange configured to connect with a mating flange of a vessel penetration through an outer wall of the associated nuclear reactor vessel, the vessel isolation valve fluidly connecting with the at least one external coolant conduit, the vessel isolation valve configured to block outward flow from the pressure vessel when a pressure differential across the valve exceeds prescribed criteria; a valve seat defined in the mounting flange,', 'a moveable valve member movable between an open position permitting flow through the vessel isolation valve and a closed position seating against the valve seat to block flow through the vessel isolation valve, and', 'a biasing member that biases the valve member towards the open position., 'wherein the vessel isolation valve further includes2. The system of claim 1 , wherein the valve member of the vessel isolation valve ...

ORGANIC IODINE TRAPPING APPARATUS AND ORGANIC IODINE TRAPPING METHOD

An organic iodine trapping apparatus and method efficiently traps organic iodine in a nuclear reactor container vessel. A liquid vessel contains a non-volatile liquid (e.g., ionic liquid or interfacial active agent solution) capable of decomposing organic iodine. An introduction pipe introduces a fluid containing organic iodine in the nuclear reactor container vessel to the non-volatile liquid. The non-volatile liquid is heated by heat in the nuclear reactor container vessel or reaction heat of the fluid in the nuclear reactor container vessel. Then, the trapping apparatus decomposes and traps the organic iodine. The organic iodine trapping method includes heating a non-volatile liquid capable of decomposing organic iodine by heat in the nuclear reactor container vessel or reaction heat of fluid in the nuclear reactor container vessel; making the fluid containing organic iodine pass through the heated non-volatile liquid; and decomposing and trapping the organic iodine in the non-volatile liquid. 1. An organic iodine trapping apparatus that traps organic iodine in a nuclear reactor container vessel , comprising:a liquid vessel containing a non-volatile liquid capable of decomposing organic iodine; andan introduction pipe for introducing a fluid containing organic iodine in the nuclear reactor container vessel to the non-volatile liquid, whereinthe non-volatile liquid is heated by heat in the nuclear reactor container vessel or reaction heat of the fluid in the nuclear reactor container vessel, and then decomposes and traps the organic iodine.2. The organic iodine trapping apparatus according to claim 1 , whereinthe liquid vessel is installed in a dry well in the nuclear reactor container vessel, andthe non-volatile liquid is heated by the heat in the nuclear reactor container vessel.3. The organic iodine trapping apparatus according to claim 1 , whereinthe liquid vessel is installed in a wet well in the nuclear reactor container vessel, andthe non-volatile liquid is ...

Inertial energy coastdown for electromagnetic pump

A nuclear reactor is configured with a primary coolant loop for transferring heat away from the nuclear reactor core. In a shutdown event, the primary coolant pump may stop pumping primary coolant through the reactor core, resulting in decay heat buildup within the reactor core. An inertial energy coast down system can store kinetic energy while the nuclear reactor is operating and then release the stored kinetic energy to cause the primary coolant to continue to flow through the nuclear reactor core to remove decay heat. The inertial energy coast down system may include an impeller and a flywheel having a mass. During normal reactor operation, the flowing primary coolant spins up the impeller and flywheel, and upon a shutdown event where the primary coolant pump stops pumping, the flywheel and impeller can cause the primary coolant to continue to flow during a coast down of the flywheel and impeller.

METHOD FOR PROTECTING A NUCLEAR REACTOR AND CORRESPONDING NUCLEAR REACTOR

A method for protecting a nuclear reactor includes reconstructing a maximum linear power density released among the fuel rods of the nuclear fuel assemblies of the core; calculating the thermomechanical state and the burnup fraction of the rods; calculating a mechanical stress or deformation energy density in the cladding of one of the rods by using the said reconstructed maximum linear power density, the calculated thermomechanical states and the calculated burnup fractions, by means of a meta-model of a thermomechanical code; comparing the calculated mechanical stress or the calculated deformation energy density with a respective threshold; and stopping the nuclear reactor if the calculated mechanical stress or the calculated deformation energy density exceeds the respective threshold. 112-. (canceled)13. A method for protecting a nuclear reactor , the nuclear reactor comprising a core having a plurality of nuclear fuel assemblies , each assembly comprising a plurality of fuel rods , each fuel rod comprising a cladding and nuclear fuel enclosed in the cladding , the method comprising the following steps:reconstructing a maximum linear power released among the fuel rods of the nuclear fuel assemblies of the core;calculating the thermomechanical state and the burnup fraction of the fuel rods;calculating a mechanical stress or deformation energy density in the cladding of one of the fuel rods using the said reconstructed maximum linear power, the calculated thermomechanical states and the calculated burnup fractions, by a meta-model of a thermomechanical code;comparing the calculated mechanical stress or the calculated deformation energy density with a respective threshold; andstopping the nuclear reactor if the calculated mechanical stress or the calculated deformation energy density exceeds the said respective threshold.14. The method according to the claim 13 , wherein the step of reconstructing the maximum linear power is carried out using measurements provided ...

ACTUATING A NUCLEAR REACTOR SAFETY DEVICE

A nuclear reactor trip apparatus includes a remote circuit breaker trip device operatively connected to a reactor trip breaker to release a control rod into a nuclear reactor core, an active power source, a passive power source, and a local circuit breaker trip device operatively connected to the reactor trip breaker including a sensor to trigger the local circuit breaker trip device upon sensing a predefined condition. The active power source is electrically coupled to energize the remote circuit breaker trip device under normal operating conditions. The passive power source is electrically coupled to energize the remote circuit breaker trip device based on a loss of the active power source. 1. A nuclear reactor trip apparatus , comprising:a remote circuit breaker trip device operatively connected to a reactor trip breaker to release a control rod into a nuclear reactor core;an active power source electrically coupled to energize the remote circuit breaker trip device;a passive power source electrically coupled to energize the remote circuit breaker trip device based on a loss of the active power source; anda local circuit breaker trip device operatively connected to the reactor trip breaker including a sensor to trigger the local circuit breaker trip device upon sensing a predefined condition.2. The nuclear reactor trip apparatus of claim 1 , wherein the passive power source comprises at least one of a capacitor or a battery.3. The nuclear reactor trip apparatus of claim 1 , wherein the remote circuit breaker trip device comprises a shunt trip coil.4. The nuclear reactor trip apparatus of claim 1 , wherein the local circuit breaker trip device comprises an under voltage trip assembly.5. The nuclear reactor trip apparatus of claim 1 , further comprising a logic device comprising:a first terminal electrically coupled to the remote circuit breaker trip device, anda second terminal electrically coupled to both the active power source and the passive power source.6. ...

REACTOR AND OPERATING METHOD FOR THE REACTOR

Provided are a reactor and an operating method for the reactor, and more particularly, a reactor which may passively cool excessively generated heat without an operation of an operator at the time of abnormality of the reactor, completely passively perform the cooling operation for safety procedures by a structure of the reactor and a change in environmental conditions such as a pressure, etc., without a separate control command, and have a simpler structure than the existing reactor safety system, and an operating method for the reactor. 1. A reactor , comprising:a reactor driving system configured to include a reactor vessel accommodating a reactor core and a steam generator to which a steam pipe and a water supply pipe are connected; anda reactor safety system configured by being divided into an energy release space (ESR) accommodating the reactor driving system, an energy absorbing space (EAS) communicating with the energy release space through a passage formed thereover and accommodating a coolant, and an energy transfer space (ETS) formed to be isolated from the energy release space and the energy absorbing space and having a heat exchange device provided therein to transfer heat released from the reactor driving system to the coolant, the heat exchange device being connected to the energy release space and the energy absorbing space, respectively;wherein the coolant within the reactor safety system is selectively distributed in response to thermal-hydraulic conditions changed depending on a change in pressure within the reactor driving system and whether the coolant is leaked to cool the reactor driving system.2. A reactor , comprising:{'b': '11', 'a reactor driving system configured to include a reactor vessel accommodating a reactor core and a steam generator to which a steam pipe and a water supply pipe are connected; and'}a reactor safety system configured to include a releasing isolation vessel accommodating gas and the reactor driving system, an ...

Systems for debris mitigation in nuclear reactor safety systems

Filtering systems and methods remove debris from coolant in a nuclear reactor setting. One or more filters are installed outside coolant reservoirs specifically where coolant will flow toward the reservoir, such as during a transient or other coolant leak event. Useable filters permit coolant through-flow while catching, straining, diverting, or otherwise removing debris from the coolant without significant interference with the coolant flow. Filters can be installed at any location in a flow path for coolant flowing toward the reservoir, including pipes draining into a suppression pool, floor or personnel platform gratings, areas around main steam legs or steam generators, in a reactor drywell, etc. One or more filters are installed by securing the filter in a coolant flow path into a coolant source. Installation and maintenance can be performed during any maintenance period.

HEAT TRANSFER SYSTEMS FOR NUCLEAR REACTOR CORES, AND RELATED SYSTEMS

A system for transferring heat from a nuclear reactor comprises a nuclear reactor comprising a nuclear fuel and a reactor vessel surrounding the nuclear reactor and a heat transfer system surrounding the nuclear reactor. The heat transfer system comprises an inner wall surrounding the nuclear reactor vessel, first fins coupled to an outer surface of inner wall, an outer wall between the inner wall and a surrounding environment, and second fins coupled to an inner surface of the outer wall and extending in a volume between the outer surface of the inner wall and the inner surface of the outer wall, the outer surface of the inner wall and the first fins configured to transfer heat from the nuclear reactor core to the second fins and the inner surface of the outer wall by thermal radiation. The heat transfer system may be directly coupled to the nuclear reactor vessel, or may be coupled to an external reflector surrounding the nuclear reactor vessel. Related heat transfer systems and systems for selectively removing heat from a nuclear reactor are disclosed. 1. A system for transferring heat from a nuclear reactor , the system comprising:a nuclear reactor comprising a nuclear fuel;{'claim-text': ['an inner wall surrounding the nuclear reactor;', 'first fins coupled to an outer surface of inner wall;', 'an outer wall between the inner wall and a surrounding environment; and', 'second fins coupled to an inner surface of the outer wall and extending in a volume between the outer surface of the inner wall and the inner surface of the outer wall, the outer surface of inner wall and the first fins configured to transfer heat from the nuclear reactor to the second fins and the inner surface of the outer wall by thermal radiation.'], '#text': 'a heat transfer system surrounding the nuclear reactor, the heat transfer system comprising:'}2. The system of claim 1 , wherein the first fins comprise steel.3. The system of claim 1 , wherein the first fins comprise a core and a ...

DEPRESSURIZATION AND COOLANT INJECTION SYSTEMS FOR VERY SIMPLIFIED BOILING WATER REACTORS

Simplified nuclear reactors include depressurization systems or gravity-driven injection systems or both. The systems depressurize and cool the reactor without operator intervention and power. An underground containment building may be used with the depressurization and injection systems passing through the same from above ground. Depressurization systems may use a rupture disk, relief line, pool, and filter to open the reactor and carry coolant away for condensation and exhausting. Injection systems may use a coolant tank above the nuclear reactor to inject liquid coolant by gravity into the reactor through an injection line and valve. The rupture disk and valve may be integral with the reactor and use penetration seals where systems pass through containment. Rupture disks and valves can actuate passively, at a pressure setpoint or other condition, through fluidic controls, setpoint failure, etc. The depressurization system and injection system together feed-and-bleed coolant through the reactor. 1. A simplified nuclear reactor system for commercially generating electricity , the system comprising:a nuclear reactor;a primary coolant loop connecting to the nuclear reactor; anda depressurization system including a rupture disk for the nuclear reactor, wherein the rupture disk is configured to open the reactor at a pressure setpoint below failure of the reactor.2. The system of claim 1 , wherein the rupture disk is integral with the nuclear reactor and wherein the depressurization system further includes claim 1 ,a relief line connected to the rupture disk and configured to carry coolant away from the reactor following opening of the rupture disk.3. The system of claim 2 , further comprising:a containment surrounding the nuclear reactor, wherein the depressurization system includes a pool, and wherein the relief line extends into and opens below a surface of the pool so as to exhaust the coolant into the pool for condensation and/or scrubbing.4. The system of claim 3 ...

Passive nitrogen injecting device for nuclear reactor coolant pump

The present invention relates to a passive nitrogen injecting device for a nuclear reactor coolant pump, the device comprising: a nitrogen supply unit for supplying nitrogen; a pressure control valve for controlling the supply of nitrogen from the nitrogen supply unit according to pressure; an accumulator for filling the nitrogen supplied through the pressure control valve at a set pressure, and supplying the filled nitrogen in the event that an accident involving coolant loss occurs; and an isolation valve for controlling the supply of the nitrogen from the accumulator into a seal housing of a nuclear reactor coolant pump. The present invention uses an accumulator so as to be able to supply nitrogen by using the pressure in the accumulator without the supply of external power in the event of an accident involving coolant loss, and therefore has the effect of being able to improve safety.

Emission monitoring system for a venting system of a nuclear power plant

A nuclear system, in particular a nuclear power plant (), includes a containment () and an associated venting system (), which has a venting line () connected to the containment (), and an emission monitoring system () is provided for the venting system (). A representative measuring sample is taken from the clean gas line of the venting system, and can be tested for aerosol-type decomposition products online in a subsequent analysis system. The emission monitoring system comprises a sampling line () for a sample flow branching off from the venting line () and leading into a sample container (), and a recirculation line () leading from the sample container () to the venting line (). The sample container () contains a wet scrubber () for the sample flow, as well as an ionisation separator () downstream of the wet scrubber () in relation to the sample flow. A liquid removal line () leads from the sample container () to an analysis unit (). 115-. (canceled)16. A nuclear facility comprising:a containment;a venting system associated with the containment including a venting line connected to the containment; a sampling line for a probe flow, the sampling line branching from the venting line and leading into a sample container,', 'a return line leading from the sample container to the venting line, the sample container including a wet scrubber for the probe flow and an ionization separator downstream of the wet scrubber in relation to the probe flow, a liquid-tapping line leading from the sample container to an analyzing unit., 'an emission-monitoring system comprising17. The nuclear facility of claim 16 , wherein the wet scrubber is in a lower part of the sample container and the ionization separator is thereabove in an upper part of the sample container.18. The nuclear facility of claim 16 , wherein the wet scrubber is a venturi scrubber.19. The nuclear facility of claim 18 , wherein the venturi scrubber includes a venturi tube completely immersed in a scrubbing liquid. ...

INTEGRAL VESSEL ISOLATION VALVE

A nuclear reactor comprises a nuclear reactor core disposed in a pressure vessel. An isolation valve protects a penetration through the pressure vessel. The isolation valve comprises: a mounting flange connecting with a mating flange of the pressure vessel; a valve seat formed into the mounting flange; and a valve member movable between an open position and a closed position sealing against the valve seat. The valve member is disposed inside the mounting flange or inside the mating flange of the pressure vessel. A biasing member operatively connects to the valve member to bias the valve member towards the open position. The bias keeps the valve member in the open position except when a differential fluid pressure across the isolation valve and directed outward from the pressure vessel exceeds a threshold pressure. 1. A nuclear reactor comprising:a nuclear reactor core comprising a fissile material;a pressure vessel containing the nuclear reactor core immersed in primary coolant disposed in the pressure vessel; andan isolation valve including a mounting flange secured to a wall of the pressure vessel and a valve body disposed in the wall or in a flange assembly including the mounting flange, the isolation valve closing responsive to a pressure difference across the valve exceeding a threshold pressure difference.2. The nuclear reactor of wherein:the mounting flange includes a valve seat defined in the mounting flange; andthe valve body includes a valving element biased away from the valve seat defined in the mounting flange;wherein a pressure difference across the valve exceeding the threshold pressure difference closes the isolation valve by causing the valving element to seat against the valve seat defined in the mounting flange.3. The nuclear reactor of wherein the isolation valve further includes a rod connected with the valving element and engaging a biasing element.4. The nuclear reactor of wherein the biasing element comprises:a spring engaging the rod.5. The ...

METHODS RELATED TO VALVE ACTUATORS HAVING MOTORS WITH PEEK-INSULATED WINDINGS

A method of operating a nuclear reactor includes operating a valve actuator to open and close a valve in fluid communication with a nuclear reactor fluid control system. The valve actuator includes a motor having windings of magnet wire. The magnet wire includes a layer of insulating material disposed over a conductor. The layer of insulating material comprises polyetheretherketone (PEEK) and has a thickness between about 0.025 mm and about 0.381 mm. A method of replacing a valve actuator motor with such a motor having windings formed of PEEK-insulated magnet wire is also disclosed. A method of coupling a valve actuator with such a motor having windings formed of PEEK-insulated magnet wire to a valve is also disclosed. 1. A method of operating a nuclear reactor , comprising:operating a valve actuator to open and close a valve in fluid communication with a nuclear reactor fluid control system, the valve actuator having an electric motor, the electric motor having windings of magnet wire, the magnet wire comprising a layer of insulating material disposed over a conductor, the layer of insulating material comprising polyetheretherketone (PEEK), the layer of insulating material having a thickness between about 0.025 mm and about 0.381 mm.2. The method of claim 1 , wherein the layer of insulating material is configured to protect the functionality of the motor windings when the motor is exposed to temperatures between about 120 degrees Celsius and about 260 degrees Celsius.3. The method of claim 2 , wherein the layer of insulating material is configured to protect the functionality of the motor windings when the motor is exposed to temperatures between about 120 degrees Celsius and about 260 degrees Celsius for at least about 5 minutes in a 100 percent humidity environment.4. The method of claim 3 , wherein the layer of insulating material is configured to protect the functionality of the motor windings when the motor is exposed to a pressure of at least about 0.344 MPa ...

Organic iodine remover

As an organic iodine remover that removes organic iodine in a containment vessel of a nuclear reactor, an organic agent (for example, an ionic liquid, an interfacial active agent, a quaternary salt, or a phase transfer catalyst) having a function of dissolving and decomposing the organic iodine and retaining iodine is used. The organic iodine remover is a substance composed of a cation and an anion. The organic iodine remover is, in particular, an organic iodine remover in which, in a structure of the cation of the organic agent, carbon or oxygen is bonded to, via a single bond, to a phosphorus element, a sulfur element or a nitrogen element, the number of carbon chains is 2 or more, and an anionic structure is configured with a substance with high nucleophilicity. By using such an organic agent, the organic iodine is removed with an efficiency of 99% or more.

STARTUP/SHUTDOWN HYDROGEN INJECTION SYSTEM FOR BOILING WATER REACTORS (BWRS), AND METHOD THEREOF

A system and a method for injecting hydrogen into Boiling Water Reactor (BWR) reactor support systems in operation during reactor startup and/or shutdown to mitigate Inter-Granular Stress Corrosion Cracking (IGSCC). The system may provide hydrogen at variable pressures (including relatively higher pressures) that match changing operating pressures of the reactor supports systems as the reactor cycles through startup and shutdown modes. 1. A method of injecting hydrogen into a Boiling Water Reactor (BWR) support system during reactor startup and/or shutdown modes to mitigate Inter-Granular Stress Corrosion Cracking (IGSCC) , comprising:fluidly connecting at least one hydrogen source to the BWR support system during at least one of a reactor startup mode and a reactor shutdown mode, the BWR support system being in operation during the reactor startup and shutdown modes;directing a hydrogen flow from the at least one hydrogen source to the BWR support system; andregulating a pressure of the hydrogen flow based upon an operating pressure of the BWR support system.2. The method of claim 1 , wherein the BWR support system experiences a reactor water fluid flow through the BWR support system during the reactor startup and shutdown modes.3. The method of claim 2 , wherein the BWR support system is one of a Reactor Water Cleanup (RWCU) return line and a Feedwater Recirculation line.4. The method of claim 1 , wherein the regulating of the pressure of the hydrogen flow includes matching the pressure of the hydrogen flow to the operating pressure of the BWR support system claim 1 , the operating pressure of the BWR support system being variable during the reactor startup and shutdown modes.5. The method of claim 4 , further comprising:boosting the pressure of the hydrogen flow using a hydrogen booster.6. The method of claim 5 , wherein the hydrogen booster is one of a hydraulically-driven and a pneumatically-driven booster.7. The method of claim 5 , wherein the hydrogen booster ...

CONTAINMENT VENT SYSTEM WITH PASSIVE MODE FOR BOILING WATER REACTORS (BWRS), AND METHOD THEREOF

A system and a method for a passive containment vent system for a Boiling Water Reactor (BWR). The system is capable of venting and scrubbing a gaseous discharge from the primary containment of the BWR over a prolonged period of time leading up to or following a serious plant accident, without the need for monitoring by on-site plant personnel. External electrical power is not required (following initial activation of the system) in order to operate the containment vent system. The system may protect the integrity of primary containment during and following the serious plant accident. 1. A containment vent system , comprising:a containment vent line in fluid communication with primary containment of a Boiling Water Reactor (BWR);one or more containment valves in the containment vent line; andone or more pressure activated devices in the containment vent line, located downstream of the one or more containment valves.2. The containment vent system of claim 1 , further comprising:a discharge point at a distal end of the containment vent line, the discharge point being located in an elevated, remote location from a primary containment boundary of the BWR.3. The containment vent system of claim 1 , wherein the one or more containment valves includes at least one of a ball valve with air-actuator claim 1 , a butterfly valve with air-actuators claim 1 , and a butterfly valve with motor-actuator.4. The containment vent system of claim 1 , wherein the one or more pressure activated devices includes a first pressure set-point rupture disk.5. The containment vent system of claim 4 , wherein claim 4 ,the one or more pressure activated devices further includes a second pressure set-point rupture disk,the second pressure set-point rupture disk being located downstream of the first pressure set-point rupture disk,the second pressure set-point rupture disk having a lower set-point pressure than the first pressure set-point rupture disk.6. The containment vent system of claim 5 , ...

Gas Supply Apparatus and Air or Nitrogen Supply Apparatus of Nuclear Plant

A gas supply apparatus of the present invention includes: an operating valve that is placed in the middle of a piping for letting at least gas in a plant flow and that operates a valve main body by the gas flowing in the piping; a first electromagnetic valve that is placed in the middle of the piping and that opens or closes a flow of the gas to the operating valve; and a gas supply source that supplies the first electromagnetic valve with the gas. A gas discharge line of the first electromagnetic valve has a switching valve placed therein and has a second electromagnetic valve placed between the switching valve and the gas supply source. The switching valve switches between a gas discharge from the first electromagnetic valve and a gas supply to the first electromagnetic valve. When a power source is lost, the switching valve is switched to connection to the gas supply source so as to supply the first electromagnetic valve with the gas. At the time of a normal operation, the second electromagnetic valve opens a gas discharge line side and closes a switching valve side, and when the power source is lost, the second electromagnetic valve opens the switching valve side and closes the gas discharge line side. In this way, even when the power source is lost, an operating valve such as an air-operated valve can not only be operated remotely but also be operated safely by a remote operator. 1. A gas supply apparatus comprising:an operating valve that is placed in the middle of a piping for letting at least gas in a plant flow and that operates a valve main body by the gas flowing in the piping;a first electromagnetic valve that is placed in the middle of the piping and that opens or closes a flow of the gas to the operating valve; anda gas supply source that supplies the first electromagnetic valve with the gas,wherein a gas discharge line of the first electromagnetic valve has a switching valve and has a second electromagnetic valve placed between the switching valve and ...

Emission monitoring system for a venting system of a nuclear power plant and nuclear power plant having the emission monitoring system

An emission monitoring system for a venting system of a nuclear power plant is configured for low consumption of energy while having high reliability of measurement results. The emission monitoring system has a pressure relief line connected to a containment and contains a high-pressure section, a low-pressure section, and a sampling line, which, on the inlet side, opens into the low-pressure section of the pressure relief line and is guided from there to a functional path through which steam flows. An ejector containing a pump fluid connector, a suction connector and an outlet connector is provided. A pump fluid feed line has an inlet side opening into the high-pressure section of the pressure relief line and is guided from there to the ejector and connected to the pump fluid connector. A sample return line is guided from the functional path to the ejector and connected to the suction connector.

INTRINSICALLY SAFE NUCLEAR REACTOR

An improved nuclear fission reactor of the liquid metal cooled type including a core configuration allowing for only two operational states, “Power” or “Rest”. The flow of the primary cooling fluid suspends the core in the “Power” state, with sufficient flow to remove the heat to an intermediate heat exchanger during normal operation. This invention utilizes the force of gravity to shut down the reactor after any loss of coolant flow, either a controlled reactor shut down or a “LOCA” event, as the core is controlled via dispersion of fuel elements. Electromagnetic pumps incorporating automatic safety electrical cut-offs are employed to shutdown the primary cooling system to disassemble the core to the “Rest” configuration due to a loss of secondary coolant or loss of ultimate heat sink. This invention is a hybrid pool-loop pressurized high-temperature or unpressurized reactor unique in its use of a minimum number of components, utilizing no moving mechanical parts, no rotating seals, optimized piping, and no control rods. Thus defining an elegantly simple intrinsically safe nuclear reactor. 1. A liquid metal cooled nuclear fission reactor core , in which the heat created by nuclear fission is utilized to generate thermal energy , comprising:a plurality of nuclear fuel elements in the form of spheres that are of a higher density than the density of the reactor's hot primary cooling fluid, an improvement comprising of a means to hydraulically shuffle the core at the start of each power cycle;a lower core chamber surrounded by neutron absorbers and geometrically shaped so as to hold said fuel spheres in a configuration that will not support fission;an upper core chamber surrounded by neutron reflectors and geometrically shaped so as to hold said fuel spheres in a configuration that will support fission, the improvement comprising of: (a) a method to control the operation of the reactor while upwards primary coolant hydraulic flow maintains said core, (b) making said ...

UNDERGROUND NUCLEAR POWER REACTOR WITH A BLAST MITIGATION CHAMBER

An underground nuclear power reactor having a hollow blast tunnel which extends from one end of a containment member which houses a nuclear reactor, heat exchanger, generator, etc. A hollow blast tunnel extends from one end of the containment member with a normally closed door positioned therebetween. The blast tunnel defines a blast chamber having a plurality of spaced-apart debris deflectors positioned therein. The blast chamber has an upper wall with a roof opening formed therein which is selectively closed by a roof portion. If the reactor needs to be repaired or replaced, the door is opened so that the reactor will pass therethrough into the blast chamber and outwardly through the roof opening. If the reactor explodes, the blast therefrom drives the debris therefrom through the door and into the blast chamber where the deflectors reduce the blast force as the debris passes through the blast chamber. 1. An underground nuclear reactor , comprising: (a) a bottom wall having a first end, a second end, a first side, a second side, an upper side and a lower side;', '(b) an upstanding first end wall having a lower end, an upper end, an inner side, an outer side, a first end and a second end;', '(c) said first end wall extending upwardly from said first end of said bottom wall;', '(d) an upstanding second end wall having a lower end, an upper end, an inner side, an outer side, a first end and a second end;', '(e) said second end wall extending upwardly from said second end of said bottom wall;', '(f) said second end wall of said containment member having a door opening formed therein;', '(g) a door movably positioned in said door opening in said second end of said containment member with said door being movable from a normally closed position to an open position when an explosion or blast occurs in the nuclear reactor;', '(h) an upstanding first side wall having a lower end, an upper end, an inner side, an outer side, a first end and a second end;', '(i) said first ...

Nuclear-reactor control-absorber drive mechanism and corresponding monitoring method and nuclear reactor

A nuclear-reactor control-absorber drive mechanism includes a device for monitoring a potential situation of increase to overspeed of the absorber, configured to measure the number of control steps delivered to at least one of the first, second and third phases of the stator during a time window of preset duration or the number of rotation steps of the rotor during a time window of preset duration. The drive is also configured to compare the number of measured control steps with a preset maximum or the number of measured rotation steps with a preset maximum.

Systems, Devices, and/or Methods for Managing Radiation Shielding

Certain exemplary embodiments can provide a system that comprises a vessel in a nuclear fission system. The vessel defines a flanged access port. The system comprises a door assembly that is constructed to cover the flanged access port. The door assembly constructed to act as a radiation shield. The door is opened and closed via an actuating system. 1. A system comprising:a vessel in a nuclear fission system, the vessel defining a flanged access portal; anda door assembly that is constructed to cover the flanged access portal, the door assembly constructed to act as a radiation shield, a door of the door assembly opened and closed via an actuating system.2. The system of claim 1 , wherein:the actuating system comprises a gas cylinder.3. The system of claim 1 , wherein:the actuating system comprises a hydraulic cylinder.4. The system of claim 1 , wherein:the actuating system comprises a spring.5. The system of claim 1 , wherein:the door assembly comprises an integrated shield mount and a ventilation duct, the integrated shield mount couplable to actuators of the actuating system, the ventilation duct constructed to remove air in proximity to the door assembly.6. The system of claim 1 , wherein:the door assembly is installable as a single piece by one person utilizing integrated bolts that use collapsible thread technology.7. The system of claim 1 , wherein:the door assembly comprises a shield, the shield comprising substantially transparent liquid shielding in a substantially transparent housing.8. The system of claim 1 , wherein:the door assembly comprises a shield, the shield comprising lead, tungsten, or other substantially opaque shielding material.9. The system of claim 1 , wherein:the door assembly comprises a shield, the shield is manufactured as a single or sectional component.10. The system of claim 1 , wherein:the system comprises a plurality of door assemblies and a wye or splitter is installed between doors with interconnecting ventilation ducts; andthe ...

Reactor cooling and electric power generation system

Also, the reactor cooling and power generation system according to the present invention may continuously operate during an accident as well as a normal operation so as to cool the reactor and produce emergency power, thereby improving system reliability. In addition, the reactor cooling and power generation system according to the present invention may facilitate application of safety class or seismic design with a small scale facility, thereby improving the reliability owing to the application of the safety class or seismic design.

REACTOR PRESSURE VESSEL DEPRESSURIZATION SYSTEM AND MAIN STEAM SAFETY RELIEF VALVE DRIVE APPARATUS

According to an embodiment, a reactor pressure vessel depressurization system has: a main steam safety relief valve, a main steam safety relief valve driving gas pipe; a three-way solenoid valve having the first connection port; a driving gas feed pipe connected to the second connection port; and a containment vessel external connection pipe connected to the third connection port and extending to outside of the reactor containment vessel. The three-way solenoid valve is either in the first communication state where the first connection port communicates with the second connection port or in the second communication state where the first connection port communicates with the third communication port. The containment vessel external connection pipe has an open communication section open in normal operation and capable of being unopened, and an external gas receiving section capable of receiving second driving gas. 1. A reactor pressure vessel depressurization system for reducing pressure in a reactor pressure vessel contained in a reactor containment vessel of a nuclear reactor facility , the system comprising:a main steam safety relief valve arranged in the reactor containment vessel to discharge steam in the reactor pressure vessel into the reactor containment vessel in an abnormal condition of the nuclear reactor facility;a main steam safety relief valve driving gas pipe connected at a first end thereof to the main steam safety relief valve to lead first driving gas to the main steam safety relief valve;a three way electromagnetic valve arranged in the reactor containment vessel so as to be connected at a first connection port thereof to a second end of the main steam safety relief valve driving gas pipe opposite to the first end;a driving gas feed pipe connected to a second connection port of the three way electromagnetic valve to supply first drive gas from a supply source thereof; anda containment vessel external connection pipe connected at a third end thereof ...

FLOATING NUCLEAR REACTOR PROTECTION SYSTEM

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 includes one or more cones which are positioned on the upper end of the nuclear reactor which will disintegrate and deflect an aircraft or missile striking the same. The system also includes structure which permits the barge to move downwardly in the tank upon an aircraft or missile strike to reduce the impact force of the strike. 1. A floating nuclear reactor , comprising: (a) a horizontally disposed bottom wall having a first end, a second end, a first side and a second side;', '(b) a vertically disposed 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 vertically disposed 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 vertically disposed 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 vertically disposed 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;said tank being partially or completely buried in the ground;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 at said second end of said barge;said nuclear reactor including an upstanding first containment member having an upper end and a lower end;a first cone mounted on said upper end of said first containment member which extends upwardly therefrom so that the ...

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 ...

Hydrogen Concentration Meter

A hydrogen concentration meter for measuring density of hydrogen in gas, is disclosed having a first electrode and a second electrode. The first electrode is formed of a first metal. The second electrode is formed of a second metal having a work function different from a work function of the first metal. The second electrode faces the first electrode. At least one of the first electrode or the second electrode detects an electrically charged particle generated electrically between the first electrode and the second electrode by a recoil proton generated by an irradiated neutron.

REACTOR SHUTDOWN SYSTEM

A system for use in shutting down a nuclear reactor includes a housing that defines a region therein sealed from an ambient environment and a gate member disposed within the region in a manner such that the gate member segregates the region into a first compartment and a second compartment isolated from the first compartment. The gate member is formed from a material having a predetermined melting point. The system further includes a neutron absorbing material disposed within the first compartment and a dispersion mechanism disposed within the region. The dispersion mechanism structured to encourage the neutron absorbing material from the first compartment into the second compartment. 1. A system for use in shutting down a nuclear reactor , the system comprising:a housing defining a region therein sealed from an ambient environment;a gate member disposed within the region in a manner such that the gate member segregates the region into a first compartment and a second compartment isolated from the first compartment, the gate member comprising a material having a predetermined melting point;a neutron absorbing material disposed within the first compartment; anda dispersion mechanism disposed within the region, the dispersion mechanism structured to encourage the neutron absorbing material from the first compartment into the second compartment.2. The system of claim 1 , wherein the predetermined melting point of the material is around 800° C.3. The system of claim 1 , wherein the gate member comprises a number of heater coils embedded in the material that are structured to melt the material upon actuation by an electrical current.4. The system of claim 1 , wherein the neutron absorbing material comprises a phase change material.5. The system of claim 1 , wherein the phase change material comprises at least one of: an indium/cadmium alloy claim 1 , lithium claim 1 , or boron oxide.6. The system of claim 4 , wherein the dispersion mechanism comprises a porous matrix ...

PRESSURIZED WATER REACTOR DEPRESSURIZATION SYSTEM

A passive cooling system of a pressurized water reactor that relies on a depressurization system to reduce the pressure in the reactor vessel in the event of a loss of coolant accident and vent the steam generated by the decay heat of the reactor core in a post loss of coolant accident stage. The depressurization results in a low pressure difference between the reactor vessel and the containment and enables gravity driven cooling system injection into the reactor vessel. The depressurization system includes a flow restrictor within an orifice in the reactor vessel wall that connects to a vent pipe which forms a flow path between the interior of the reactor vessel and the containment atmosphere when a valve within the vent pipe is in an open position. Preferably, the flow restrictor is a venturi that has a gradual contraction and a gradual expansion in the flow path area. 1. A nuclear power generation system comprising a reactor enclosed within a pressure vessel housed within a containment , the reactor operating at a higher pressure than an area surrounding the reactor in the containment , the reactor including a depressurization system for reducing the pressure within the reactor and venting the coolant within the reactor into the containment , the depressurization system comprising;an orifice within the pressure vessel for venting the coolant within the pressure vessel into the containment; anda flow restrictor in flow communication with the orifice for restricting a critical flow rate of a fluid within the pressure vessel out of the orifice while enabling sufficient flow of the fluid to substantially equalize the pressure within the pressure vessel with the pressure in the area surrounding the reactor.2. The nuclear power generation system of wherein the flow restrictor has a reduced opening compared to openings in other conduits in the depressurization system and the reduced opening is gauged to provide a minimum critical flow required by the depressurization ...

PASSIVE CONTAINMENT COOLING AND FILTERED VENTING SYSTEM, AND NUCLEAR POWER PLANT

A passive containment cooling and filtered venting system includes: an outer well; a scrubbing pool arranged in the outer well; a cooling water pool installed above the dry well and the outer well; a heat exchanger partly submerged in the cooling water; a gas supply pipe that is connected to the inlet plenum of the ruin of the heat exchanger at one end and connected to a gas phase region of the containment vessel at the other end; a condensate return pipe that is connected to the outlet plenum of the heat exchanger at one end, and connected to inside the containment vessel at other end; and a gas vent pipe that is connected to the outlet plenum of the heat exchanger at one end and is submerged in the scrubbing pool at other end. 1. A passive containment cooling and filtered venting system of a nuclear power plant , the plant including:a core,a reactor pressure vessel that accommodates the core, a dry well that contains the reactor pressure vessel,', 'a wet well that contains in its lower portion a suppression pool connected to the dry well via a LOCA vent pipe and includes in its upper portion a wet well gas phase,', 'a vacuum breaker that circulates gas in the wet well gas phase to the dry well, and', 'a pedestal that supports the reactor pressure vessel in the containment vessel via an RPV skirt and forms a pedestal cavity inside,, 'a containment vessel includingthe passive containment cooling and filtered venting system comprising:an outer well that is arranged outside the dry well and the wet well, adjoins the dry well via a dry well common part wall, adjoins the wet well via a wet well common part wall, and has pressure resistance and gastightness equivalent to pressure resistance and gastightness of the dry well and the wet well;a scrubbing pool that is arranged in the outer well and stores water inside;a cooling water pool that is installed above the dry well and the outer well and reserves cooling water;a heat exchanger that includes an inlet plenum, an ...

Reactor Containment Building Spent Fuel Pool Filter Vent

A nuclear containment atmospheric filter including dedicated piping, valves, a control system and a chemical injection system to facilitate the use of a commercial nuclear power plant's Spent Fuel Storage Pool and Spent Fuel Storage Pool Cooling System to filter and cool contaminated air and steam vapor released from within a Reactor Containment Vessel/Building preventing vessel overpressure and radioactive release. 1. A nuclear power generating facility having a containment for housing a nuclear reactor and for confining radiation leaked from the nuclear reactor , the containment having a ventilation outlet for providing a controlled release to the environment surrounding the containment , for an atmospheric pressure buildup within the containment in the event the pressure of an atmospheric effluent within the containment is built up to a level that exceeded a preselected value , and the nuclear power generating facility also having , outside the containment , an associated spent fuel storage water pool , including a filter system for filtering contaminants released from or on route to the ventilation outlet , the filter system comprising:a dedicated piping system connected between an interior of the containment or the ventilation outlet and the spent fuel storage water pool for fluidly communicating any atmospheric effluent to be released from inside of the containment through the spent fuel storage water pool;one or more valves connected to the dedicated piping system for controlling the release of the atmospheric effluent to be released;a chemical injection system configured to release a chemical into the spent fuel storage water pool to facilitate a reaction with the atmospheric effluent to be released to substantially neuter any deleterious environmental impact of the atmospheric effluent to be released; anda control system connected to one or more of the chemical injection systems and/or the one or more of the valves and configured to control the release of ...

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.

DUAL-ALLOY PYROTECHNIC-ACTUATED VALVE ASSEMBLY

A pyrotechnic-actuated valve assembly may include an insert body having an inlet, an outlet, and a flow path extending from the inlet to the outlet. The insert body is formed of a first alloy. A shear structure is bonded to the outlet of the insert body so as to close the flow path. The shear structure is formed of a second alloy. The second alloy of the shear structure is bonded to the first alloy of the insert body so as to form a hermetic seal. The dual-alloy nature of the valve assembly allows a relatively clean shearing of the shear structure during actuation, thus reducing or preventing the occurrence of deformation and/or material fragments in the flow path. 1. A pyrotechnic-actuated valve assembly , comprising:an insert body having an inlet, an outlet, and a flow path extending from the inlet to the outlet, the insert body formed of a first alloy; anda shear structure bonded to the outlet of the insert body so as to close the flow path, the shear structure formed of a second alloy, the second alloy of the shear structure being bonded to the first alloy of the insert body so as to form a hermetic seal.2. The pyrotechnic-actuated valve assembly of claim 1 , wherein the insert body is tapered at the outlet to decrease a contact area with the shear structure.3. The pyrotechnic-actuated valve assembly of claim 1 , wherein the insert body has an outer diameter that is larger than 2 inches.4. The pyrotechnic-actuated valve assembly of claim 1 , wherein the first alloy and the second alloy have different crystal structures.5. The pyrotechnic-actuated valve assembly of claim 1 , wherein the first alloy and the second alloy have different lattice constants.6. The pyrotechnic-actuated valve assembly of claim 1 , wherein the second alloy is harder than the first alloy.7. The pyrotechnic-actuated valve assembly of claim 1 , wherein the second alloy is free of cobalt.8. The pyrotechnic-actuated valve assembly of claim 1 , wherein the second alloy contains at least 0.5 ...

ZIRCONIUM-COATED SILICON CARBIDE FUEL CLADDING FOR ACCIDENT TOLERANT FUEL APPLICATION

The invention relates to a multi-component cladding for a nuclear fuel rod that includes a combination of ceramic and metal components. More particularly, the invention is directed to a cladding that includes a ceramic composite having a zirconium composition deposited thereon to form a zirconium coated ceramic composite. The ceramic composite includes a ceramic matrix and a plurality of ceramic fibers. The cladding is effective to protect the contents of the cladding structure from exposure to high temperature environments during various load conditions of a nuclear reactor. 16.- (canceled)7. A method for preparing a nuclear fuel rod cladding , comprising: a ceramic matrix; and', forming the ceramic composite in a shape such that it has an interior surface, an exterior surface and an inner cavity; and', 'depositing on the exterior surface of the ceramic composite a coating composition to form a coating thereon, the composition comprising a zirconium component composed of zirconium alloy., 'a plurality of ceramic fibers;'}], 'preparing a ceramic composite, which comprises8. The method of claim 7 , wherein preparing the ceramic composite comprises: forming a woven ceramic fiber structure, wherein voids are formed therein; and', 'depositing the ceramic matrix over the woven ceramic fiber structure to at least partially fill the voids., 'obtaining the plurality of ceramic fibers in a form of fiber tows;'}9. The method of claim 8 , wherein the forming a woven ceramic fiber structure employs a process selected from the group consisting of wrapping claim 8 , winding and braiding.10. The method of claim 9 , wherein the winding process comprises winding the ceramic fibers in the form of filaments over a mandrel.11. The method of claim 8 , wherein the depositing a ceramic matrix comprises employing a process selected from chemical vapor deposition claim 8 , chemical vapor infiltration and sol gel infiltration.12. The method of claim 7 , wherein the depositing a coating ...

NUCLEAR REACTOR PROTECTION SYSTEMS AND METHODS

A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs; and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs. 1. A nuclear reactor protection system , comprising:a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs; andone or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.240-. (canceled) This application is a continuation of U.S. Non-Provisional patent application Ser. No. 14/198,891, filed on Mar. 6, 2014, which claims priority under 35 U.S.C. § 119 to U.S. Provisional Patent Application Ser. No. 61/922,625, filed Dec. 31, 2013, the entire contents of which are hereby incorporated by reference.This disclosure describes a nuclear reactor protection system and associated methods thereof.Nuclear reactor protection systems and, generally, nuclear reactor instrumentation and control (I&C) systems provide automatic initiating signals, automatic and manual control signals, and monitoring displays to mitigate the consequences of fault conditions. For example, I&C systems provide protection against unsafe reactor operation during steady state and transient power operation. During normal operation I&C systems measure various parameters and transmit the signals to control systems. During abnormal operation and accident conditions, the I&C systems ...

VALVE ACTUATORS HAVING MOTORS WITH INSULATED WINDINGS AND RELATED METHODS

Systems, devices, and methods include a valve actuator to open and close a valve in fluid communication with a fluid control system. The valve actuator includes a motor having windings of wire. The wire includes insulating material disposed over a conductor. 1. A method of operating a nuclear reactor , comprising:operating a valve actuator to open and close a valve in fluid communication with a nuclear reactor fluid control system; anddriving the valve actuator with an electric motor by rotating an armature of the electric motor with windings of wire of the electric motor, the wire of the windings comprising insulating material disposed over and around a conductor of the wire, the insulating material configured to protect functionality of the windings, the windings of the wire being treated with an impregnating resin after the windings are formed and wound, the impregnating resin configured to act as a binding agent for structural integrity and provide a barrier to at least one harsh environment in order to at least partially reduce degradation of the insulating material.2. The method of claim 1 , wherein the insulating material of the wires of the windings substantially circumferentially and longitudinally envelopes the conductor of the wire.3. The method of claim 1 , further comprising selecting the impregnating resin to comprise a solvent-borne epoxy.4. The method of claim 1 , further comprising selecting the insulating material to comprise a material that substantially does not degrade when exposed to one or more of a boric acid spray or a sodium hydroxide spray.5. A valve actuator for use with a valve of a nuclear reactor fluid system claim 1 , comprising:a rotatable shaft for actuating a portion of a valve in fluid communication with a nuclear reactor fluid control system; and a motor frame;', 'windings of wire disposed in the motor frame, wire of the windings of wire comprising insulating material disposed directly over a conductor of the wire; and', 'an ...

NUCLEAR POWER PLANT

The invention relates to a nuclear power plant including a containment vessel including a reactor pressure vessel for receiving fissionable nuclear fuel, an aerosol filter stage a pressure relief conduit through which a gas volume flow which is filtered in the aerosol filter stage is releasable to ambient through a pass through opening in the containment vessel, and an iodine filter stage through which the gas volume flow that is filtered in the aerosol filter stage is filterable before being released to the ambient, wherein the iodine filter stage is arranged within the containment vessel, characterized in that the aerosol filter stage and the iodine filter stage are connected with one another so that transferring the gas volume flow from the aerosol filter stage to the iodine filter stage is performed essentially at an identical pressure level. 1. A nuclear power plant comprising:a containment vessel includinga reactor pressure vessel for receiving fissionable nuclear fuel,an aerosol filter stage,a pressure relief conduit through which a gas volume flow which is filtered in the aerosol filter stage is releasable to ambient through a pass through opening in the containment vessel, andan iodine filter stage through which the gas volume flow that is filtered in the aerosol filter stage is filterable before being released to the ambient, wherein the iodine filter stage is arranged within the containment vessel,wherein the aerosol filter stage and the iodine filter stage are connected with one another so that transferring the gas volume flow from the aerosol filter stage to the iodine filter stage is performed essentially at an identical pressure level.2. The nuclear power plant according to claim 1 , wherein the aerosol filter stage and the iodine filter stage are connected with one another through a tubular conduit.3. The nuclear power plant according to claim 1 , wherein the aerosol filter stage and the iodine filter stage are arranged within an identical filter ...

PASSIVELY INITIATED DEPRESSURIZATION FOR LIGHT WATER REACTOR

A nuclear reactor is surrounded by a reactor radiological containment structure. Depressurization lines running from the reactor automatically vent the reactor to the containment structure or to a compartment in the containment structure when a low pressure condition exists in the reactor. The depressurization lines include biased-open passive valves and actively actuated isolation valves arranged in series. 1. An apparatus comprising:a nuclear reactor including a pressure vessel containing primary coolant water and a nuclear reactor core comprising fissile material;a radiological containment structure surrounding the nuclear reactor; anda passive pressure vessel depressurization system including a depressurization pipe having an inlet end connected to the pressure vessel and an outlet end, and further including an actively actuated isolation valve and a biased-open passive valve arranged in series along the depressurization pipe between the inlet end and the outlet end, the biased-open passive valve closing responsive to a positive pressure difference between the inlet end and the outlet end exceeding a setpoint value.2. The apparatus of claim 1 , wherein the actively actuated isolation valve is located between the biased-open passive valve and the reactor vessel along the depressurization pipe.3. The apparatus of claim 1 , wherein the biased-open passive valve is located between the actively actuated isolation valve and the reactor vessel along the depressurization pipe.4. The apparatus of claim 1 , wherein outlet end of the depressurization pipe discharges into one of a tank and the radiological containment structure.5. The apparatus of claim 1 , wherein the biased-open passive valve comprises a spring arranged to bias the valve open.6. The apparatus of claim 5 , wherein the biased-open passive valve further comprises a valve disk biased by the spring against a valve seat to close the valve.7. The apparatus of claim 6 , wherein the actively actuated isolation ...

PASSIVE REACTOR COOLING SYSTEM

A nuclear reactor cooling system with passive cooling capabilities operable during a reactor shutdown event without available electric power. In one embodiment, the system includes a reactor vessel with nuclear fuel core and a steam generator fluidly coupled thereto. Primary coolant circulates in a flow loop between the reactor vessel and steam generator to heat secondary coolant in the steam generator producing steam. The steam flows to a heat exchanger containing an inventory of cooling water in which a submerged tube bundle is immersed. The steam is condensed in the heat exchanger and returned to the steam generator forming a closed flow loop in which the secondary coolant flow is driven by natural gravity via changes in density from the heating and cooling cycles. In other embodiments, the cooling system is configured to extract and cool the primary coolant directly using the submerged tube bundle heat exchanger. 1. A method for passively cooling a nuclear reactor after shutdown , the method comprising:heating a primary coolant in a reactor vessel with a nuclear fuel core;extracting the heated primary coolant from the reactor vessel;flowing the heated primary coolant through a tube bundle submerged in an inventory of cooling water in a heat exchanger pressure vessel;cooling the heated primary coolant to lower its temperature; andreturning the cooled primary coolant to the reactor vessel;wherein the primary coolant circulates through a first closed flow loop between the tube bundle and reactor vessel.2. The method according to claim 1 , further comprising:heating the cooling water in the pressure vessel by the primary coolant;converting a portion of the cooling water into cooling water steam;extracting the cooling water steam from the pressure vessel;flowing the extracted cooling water steam through heat dissipater ducts integrally attached to a shell of a reactor containment vessel in a thermally conductive relationship;condensing the cooling water steam in the ...

CONTAINMENT PROTECTION SYSTEM FOR A NUCLEAR FACILITY AND ASSOCIATED OPERATING METHOD

A containment protection system for treating air in a containment of a nuclear facility in the case of accidents involving extensive release of hydrogen and steam is to be able to effectively relieve such conditions in a largely passive manner. Accordingly, the containment protection system has for a circuit, which contains a conduction system and is provided for connecting to the containment, out of the containment and back again for a fluid flow. The system has a recombination device for recombining hydrogen contained in the fluid flow with oxygen to form steam, a condensation device connected downstream of the recombination device for condensing steam fractions contained in the fluid flow with measures for diverting the condensate out of the fluid flow, and a drive device for the fluid flow. A heat exchanger at least partial re-cools the condensation device. 1. A containment protection system for treating an atmosphere disposed in a containment of a nuclear facility in an event of critical incidents with an extensive release of hydrogen and steam , the containment protection system comprising:a line system for connecting to the containment and forming a circuit out of the containment and back again for a fluid stream;a recombination device for recombining the hydrogen contained in the fluid stream with oxygen into steam, said recombination device disposed in said line system;a condensation device disposed downstream of said recombination device, said condensation device condensing steam fractions contained in the fluid stream, said condensation device having means for discharging condensate from the fluid stream;a drive for propelling the fluid stream;a reservoir for an inert gas;a supply line; anda heat exchanger for at least partial recooling of said condensation device, said heat exchanger having an inlet side connected via said supply line to said reservoir for the inert gas effective as a coolant.2. The containment protection system according to claim 1 , ...

PLANT OPERATION SYSTEM AND PLANT OPERATION METHOD

An atomic power plant operation system for assisting the operation of an atomic power generation plant is provided with: an operation monitoring system which monitors and controls the operation of the atomic power generation plant; an abnormality indication monitoring system which, on the basis of an operation history of the atomic power generation plant, monitors an indication of abnormality in the atomic power generation plant; an abnormality diagnosis system which, on the basis of a result of abnormality indication that has been detected, makes an abnormality diagnosis for the atomic power generation plant; and a maintenance system for performing maintenance and management of the atomic power generation plant, wherein the systems are communicably connected, and the abnormality diagnosis system provides the maintenance system with the result of the abnormality diagnosis of the atomic power generation plant. 1. A plant operation system for supporting operation of a plant , the system comprising:an operation monitoring system which monitors the operation of the plant and controls the operation of the plant;an abnormality indication monitoring system which monitors an indication of abnormality of the plant, based on an operation history of the plant which is monitored in the operation monitoring system;an abnormality diagnosis system which performs a diagnosis of abnormality of the plant, based on a result of the abnormality indication which is detected by the abnormality indication monitoring system; anda maintenance system which is used for performing maintenance and management of the plant,wherein the operation monitoring system, the abnormality indication monitoring system, and the abnormality diagnosis system are connected to one another so as to be able to communicate from the operation monitoring system to the abnormality indication monitoring system and the abnormality diagnosis system,the abnormality diagnosis system and the maintenance system are connected ...

RADIOACTIVE IODINE ADSORBENT, AND METHOD FOR TREATING RADIOACTIVE IODINE

Provided is a method for treating radioactive iodine contained in steam discharged from a nuclear power facility, including a filling step of filling an air-permeable container with a granulated radioactive iodine adsorbent of zeolite X, wherein ion exchange sites of the zeolite X are substituted with silver so that a size of minute pores of the zeolite X is suited to a size of a hydrogen molecule, and the radioactive iodine adsorbent has a silver content of 36 wt % or more when dried, a particle size of 10×20 mesh, a hardness of 94% or more, and a water content of 12 wt % or less when dried at 150° C. for 3 h and thereby reduced in weight; and a flow passing step of passing a flow of the steam discharged from the nuclear power facility, through the container filled with the radioactive iodine adsorbent. 2. The method of claim 1 , whereinthe steam discharged from the nuclear power facility contains hydrogen molecules.3. The method of claim 1 , whereinthe steam discharged from the nuclear power facility is superheated steam having a temperature of 100° C. or more.4. The method of claim 1 , whereinin the filling step, the filling density of the radioactive iodine adsorbent is adjusted to 1.0 g/ml or more.5. The method of claim 1 , whereinin the flow passing step, a period of time for which the steam is retained in the container filled with the radioactive iodine adsorbent is set to 0.06 sec or more.6. The method of claim 1 , whereinin the flow passing step, the steam has a pressure of 399 kPa or more.7. The method of claim 1 , whereinin the flow passing step, the container filled with the radioactive iodine adsorbent has a humidity of 95% or more.8. A method for treating radioactive iodine contained in steam discharged from a nuclear power facility claim 1 , comprising:a filling step of filling an air-permeable container with a radioactive iodine adsorbent of zeolite X, whereinion exchange sites of the zeolite X are substituted with silver so that a size of minute ...

Passive reactor cooling system

A nuclear reactor cooling system with passive cooling capabilities operable during a reactor shutdown event without available electric power. In one embodiment, the system includes a reactor vessel with nuclear fuel core and a steam generator fluidly coupled thereto. Primary coolant circulates in a flow loop between the reactor vessel and steam generator to heat secondary coolant in the steam generator producing steam. The steam flows to a heat exchanger containing an inventory of cooling water in which a submerged tube bundle is immersed. The steam is condensed in the heat exchanger and returned to the steam generator forming a closed flow loop in which the secondary coolant flow is driven by natural gravity via changes in density from the heating and cooling cycles. In other embodiments, the cooling system is configured to extract and cool the primary coolant directly using the submerged tube bundle heat exchanger.

DEPRESSURISATION VALVE

A depressurisation valve for a cooling system comprising: a main chamber having a main valve, a pilot line having a secondary valve and a blowdown line; the main valve being located to seal a path of the coolant system of the nuclear reactor. The main chamber is connected to the cooling circuit via the pilot line allowing coolant to enter the main chamber, and the blowdown line allows coolant to escape from the main chamber, the pilot line having a lower fluid resistance than the blowdown line. The pressure of coolant in the main chamber maintains the main valve in a closed position, and under elevated temperature and/or pressure conditions fluid is prevented from entering the main chamber via a closure of the secondary valve on the pilot line and reduce the pressure from the valve, moving it to its open position. 1. A depressurisation valve for a cooling system comprising:a main chamber having a main valve, a pilot line having a secondary valve and a blowdown line; the main valve being located to seal a path of the cooling system,the main chamber is connected to the cooling circuit via the pilot line allowing coolant to enter the main chamber, and the blowdown line allows coolant to escape from the main chamber, the pilot line having a lower fluid resistance than the blowdown line; and wherein the pressure of coolant in the main chamber maintains the main valve in a closed position, and under elevated temperature and/or pressure conditions, with respect to the normal operating conditions, fluid is prevented from entering the main chamber via the closure of the secondary valve on the pilot line, thus reducing the pressure from the main valve and moving it to its open position.2. The depressurisation valve as claimed in claim 1 , wherein the secondary valve on the pilot line is a magnovalve.3. The depressurisation valve as claimed in claim 1 , wherein the secondary valve on the pilot line is a high pressure latching isolation valve.4. The depressurisation valve as ...

Underwater electricity production module

The underwater electricity production module according to the invention includes means in the form of an elongated cylindrical box ( 12 ) in which means are integrated forming an electricity production unit including means forming a nuclear boiler ( 30 ), associated with electricity production means ( 37 ) connected to an external electricity distribution station by electrical cables, is characterized in that the nuclear boiler-forming means ( 30 ) are placed in a dry chamber ( 19 ) of the reactor compartment ( 18 ) associated with the chamber forming a safety water storage reservoir ( 20 ) of the reactor whereof at least the radial wall ( 53 ) is in a heat exchange relationship with the marine environment and in that the dry compartment ( 19 ) of the reactor container ( 18 ) is connected to the safety water storage reservoir chamber ( 20 ) of the reactor by depressurizing means ( 70 ) including means ( 71 ) forming a depressurizing valve placed in the upper portion of the dry chamber ( 19 ) and connected to one of the bubbler-forming means ( 72 ) placed in the lower portion of the storage reservoir chamber ( 20 ).

REACTOR PRESSURE-RELIEVING FILTER SYSTEM

The present disclosure relates to a reactor pressure-relieving filter system having an interior space hermetically enclosed by a pressure-resistant reactor casing, at least one pressure-relieving opening through the reactor casing, and a dry filter for a gas mass flow emerging from the pressure-relieving opening when there is excess pressure in the interior space. The filtering efficiency can depend both on the average dwell time of the gas mass flow in the dry filter and on the temperature difference between the gas mass flow and the respective dew point. A flow channel connects the pressure-relieving opening and the dry filter. A passive orifice plate is provided upstream of the dry filter in the flow channel. 1. A reactor pressure-relieving filter system , comprising:an interior space hermetically enclosed by a pressure-resistant reactor casing;at least one pressure-relieving opening through the reactor casing;a dry filter for a gas mass flow emerging from the pressure-relieving opening when there is excess pressure in the interior space, a filtering efficiency depending both on an average dwell time of the gas mass flow in the dry filter and on a temperature difference between the gas mass flow and a respective dew point;a flow channel for connecting the pressure-relieving opening and the dry filter; anda passive orifice plate is provided upstream of the dry filter in the flow channel.2. The reactor pressure-relieving filter system according to claim 1 , wherein the passive orifice plate is provided directly upstream of the dry filter.3. The reactor pressure-relieving filter system according to claim 1 , comprising:in an entry region of the flow channel, a rupture disc which hermetically seals the flow channel and is configured to rupture when a specified rupturing pressure is exceeded.4. The reactor pressure-relieving filter system according to claim 1 , wherein a region of the flow channel between the passive orifice plate and the dry filter is thermally ...

Atomic Power Plant

[Problem] Provided is an atomic power plant which can be applied to reactors including existing reactors through a simple method and in which a pressure in a primary containment vessel can be restrained from excessively rising in a case where a steam leakage from an exhaust pipe of a stream safety relief valve occurs. [Solution] There are provided a PCV 1 , an RPV 3 , a main stream line 4 , two SRVs 6 , an S/P 8 , an SRV exhaust pipe 9 which is connected to a quencher 10 , a temperature measuring instrument 12 which measures a temperature inside the quencher 10 , an SRV controller 13 which controls opening and closing of the SRVs 6 . After a lapse of predetermined time from when the SRV 6 is opened, in a case where it is determined that a temperature detected by the temperature measuring instrument 12 is equal to or smaller than a predetermined threshold value, the SRV controller 13 causes the SRV 6 to which the temperature measuring instrument 12 detecting the temperature leads, to be closed and to be prohibited from being opened.

SUBMERGED OR UNDERWATER ELECTRICITY PRODUCTION MODULE

The submerged or underwater electricity production module according to the invention, of the type including means in the form of an elongated cylindrical box () in which means are integrated forming an electricity production unit including means forming a nuclear boiler (), associated with electricity production means () connected to an external electricity distribution station by electrical cables, is characterized in that the nuclear boiler-forming means () are placed in a dry chamber () of the reactor compartment () associated with the chamber forming a safety water storage reservoir () of the reactor whereof at least the radial wall () is in a heat exchange relationship with the marine environment and in that the nuclear boiler-forming means () include a pressurizer () connected by depressurizing means () to the safety water storage reservoir chamber () of the reactor. 1. An underwater electricity production module , comprising an elongated cylindrical box in which an electricity production unit is integrated , the electricity production unit comprising a nuclear boiler , associated with an electricity producer connected to an external electricity distribution station by electrical cables , wherein the nuclear boiler is placed in a dry chamber of the reactor compartment associated with the chamber forming a safety water storage reservoir of the reactor whereof at least the radial wall is in a heat exchange relationship with the marine environment and in that the nuclear boiler includes a pressurizer connected by a depressurizier to the safety water storage reservoir chamber of the reactor.2. The underwater electricity production module according to claim 1 , wherein the nuclear boiler includes a primary circuit comprising at least one reactor container claim 1 , a pressurizer claim 1 , a steam generator and a primary pump and a primary backup circuit in parallel on that primary circuit and including at least one primary passive heat exchanger placed in the ...

Nuclear plant with a containment shell and with a pressure relief system

A nuclear plant has a containment shell and a pressure relief line passing out of the containment shell and sealed by a shut-off valve, and through which a pressure relief flow can flow during relief operation, such that it is configured for particularly reliable management of critical scenarios where there is a considerable pressure increase within the containment shell at the same time as the release of hydrogen and/or carbon monoxide. A gas flow treatment device is provided upstream from the respective pressure relief line, and contains a flow duct and has a lower inflow opening and an upper inflow/outflow opening. Catalytic elements for eliminating hydrogen and/or carbon monoxide are arranged in the flow duct above the lower inflow opening. During a critical fault, the flow duct is flowed through from bottom to top by a gas mixture present in the containment shell by the principle of natural convection. 1. A nuclear plant , comprising:a containment shell;a shut-off valve;at least one pressure relief line passing out of said containment shell and sealed by said shut-off valve, and through said pressure relief line a pressure relief flow can flow during relief operation when said shut-off valve is open, said pressure relief line having an inlet mouth;a gas flow treatment device, disposed within said containment shell, and disposed upstream from said pressure relief line on an inlet side, said gas flow treatment device having a lateral casing and a chimney-shaped flow duct, enclosed by said lateral casing, and having a lower inflow opening and an upper inflow and outflow opening formed therein; anda first group of catalytic elements for eliminating at least one of hydrogen or carbon monoxide disposed in said chimney-shaped flow duct above or in a region of said lower inflow opening, and said inlet mouth of said pressure relief line disposed above said first group of catalytic elements and below said upper inflow and outflow opening in said lateral casing such that ...

CONTAINMENT FILTERED VENTING SYSTEM (CFVS) FOR NUCLEAR POWER PLANT

Disclosed is a containment filtered venting system (CFVS) for a nuclear power plant, which may include a filtering and venting container which is configured to store the components of the filtered venting system; an inlet pipe which is connected to the filtering and venting container and a reactor building; combined nozzles which are connected to the inlet pipe and are submerged under a filtering solution filled in part of the filtering and venting container; a cyclone separator which is configured to remove larger size substances in droplets and aerosols mixed with the filtering solution from the combined nozzles and guide to a metal filter; a metal filter which is connected to the top of the cyclone separator and is configured to filer impurities mixed in the residual droplets and aerosols; a molecular sieve which is configured to remove organic iodine from exhaust gas filtered by the metal filter; and an outlet pipe which serves to connect the filtering and venting container and a stack. 1. A containment filtered venting system (CFVS) for a nuclear power plant , comprising:a filtering and venting container which is configured to store the components of the filtered venting system;an inlet pipe which is connected to the filtering and venting container and a reactor building;combined nozzles which are connected to the inlet pipe and are submerged under a filtering solution filled in part of the filtering and venting container;a cyclone separator which is configured to remove larger size substances in droplets and aerosols mixed with the filtering solution from the combined nozzles and guide to a metal filter;a metal filter which is connected to the top of the cyclone separator and is configured to filer impurities mixed in the residual droplets and aerosols;a molecular sieve which is configured to remove organic iodine from exhaust gas filtered by the metal filter; andan outlet pipe which serves to connect the filtering and venting container and a stack.2. The ...

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 ...

LOSS-OF-COOLANT ACCIDENT REACTOR COOLING SYSTEM

A nuclear reactor cooling system with passive cooling capabilities operable during a loss-of-coolant accident (LOCA) without available electric power. The system includes a reactor vessel with nuclear fuel core located in a reactor well. An in-containment water storage tank is fluidly coupled to the reactor well and holds an inventory of cooling water. During a LOCA event, the tank floods the reactor well with water. Eventually, the water heated by decay heat from the reactor vaporizes producing steam. The steam flows to an in-containment heat exchanger and condenses. The condensate is returned to the reactor well in a closed flow loop system in which flow may circulate solely via gravity from changes in phase and density of the water. In one embodiment, the heat exchanger may be an array of heat dissipater ducts mounted on the wall of the inner containment vessel surrounded by a heat sink. 1. A passive reactor cooling system usable after a loss-of-coolant accident , the system comprising:a containment vessel in thermal communication with an external heat sink;a reactor well disposed inside the containment vessel;a reactor vessel disposed at least partially in the reactor well, the reactor vessel containing primary coolant and a nuclear fuel core heating the primary coolant which is circulated between the reactor vessel and a steam generator in a closed primary coolant flow loop;a cooling water tank disposed inside the containment vessel and containing an inventory of emergency cooling water in selective fluid communication with the reactor well via at least one flow control apparatus, the flow control apparatus having a closed position preventing flow of cooling water to the reactor well and an open position providing flow of cooling water to the reactor well; anda heat exchanger attached to an inside surface of the containment vessel, the heat exchanger in fluid communication with the reactor well and water tank via a closed cooling water flow loop in which flow ...

Cooling System of Reactor Suppression Pool

To provide a cooling system of a reactor suppression pool capable of cooling suppression pool water and improving the safety of a reactor in the case where an event surpassing a postulated initiating event occurs, or in the case where cooling of the suppression pool water by a residual heat removable system does not function. A cooling system of a reactor suppression pool according to the present invention includes a heat exchanger for cooling suppression pool water installed in the middle of a suppression pool water cooling line, operating when the temperature of the suppression pool water reaches a given temperature, performing heat exchange with the suppression pool water from a suppression pool water cleanup system suction line to cool the water, and returning the cooled suppression pool water to the suppression pool through a suppression pool water cleanup system discharge line. 1. A cooling system of a reactor suppression pool for cooling a suppression water stored in the suppression pool arranged in a reactor containment vessel in which a reactor is housed , comprising:a suppression pool water cleanup system suction line sucking the suppression pool water from the suppression pool and allows the water to flow;a suppression pool water cleanup system pump installed in the middle of the suppression pool water cleanup system suction line;a fuel pool cooling and cleanup system line one end of which is connected to the suppression pool water cleanup system suction line;a filtration demineralizer installed in the middle of the fuel pool cooling and cleanup system line and cleaning up the suppression pool water flowing in the fuel pool cooling and cleanup system line;a suppression pool water cleanup system discharge line connected to the other end of the fuel pool cooling and cleanup system line and returning the suppression pool water cleaned up by the filtration demineralizer to the suppression pool;a suppression pool water cooling line one end of which is ...

Remote Integrated Monitoring Operation System

A remote integrated monitoring operation system includes: a unit integrated database for sequentially recording a name of each plant unit, a parameter indicating an event that has occurred in the plant unit, a state of the parameter, and warning classification indicated by the parameter and the state; an inter-unit influence degree evaluation database for recording influence of the event on the other plant unit; a restoration response guidance database for defining a response to the event; a per-unit urgency degree determination section for determining a degree of urgency of each plant unit; an inter-unit influence degree determination section for evaluating a degree of influence of the event on the other plant unit; and a priority determination section for determining priorities between the respective plant units from the degree of urgency and the degree of influence.

Cooling Method For Reactor Molten Core Melt And Cooling Control System For Reactor Molten Core

The invention relates to safe operation support systems of nuclear power plants (NPPs) at severe accidents, including methods and systems for cooling and cooling control of the reactors molten core. The invention increases safety of NPP and cooling efficiency of the molten core of a reactor. The invention increases the efficiency of cooling the molten core of a reactor by safely removing the heat load from the molten metal mirror, ensuring the elimination of vapor explosions. The invention changes the principle of cooling the reactor molten core, in that after the molten core destroys the reactor vessel, the conditions for subsequent cooling of the molten metal are determined by the characteristics of the trap casing, but not of the reactor. 1. A method of cooling the molten core of a reactor , which consists in determining , after destruction of the reactor vessel by molten core , the location of the molten core debris and determining the state of core penetration by the information received from temperature sensors , cooling liquid volume supply to molten metal and adjustment , wherein after destruction of the reactor vessel by the molten core , the degree of destruction of the reactor vessel and the start time of the molten metal flowing out of the reactor vessel into the trap is determined , then coolant is supplied into the trap casing with a predetermined time delay , from the inspection chambers of the internal shells and the protective tubes unit of the reactor , after which the conditions for the formation of a sludge cap above the surface of the molten metal mirror are determined , the start time for the formation of a crust on the surface of the molten metal is determined , the completion time of aerosols release is determined , the completion time of vapour sorption and the time of hydrogen formation are determined , the stabilization time of the molten metal cooling processes and the time these processes exit into quasi-stationary mode are determined , ...

Containment Sump Ceramic Drain Plug

The present invention provides a drain plug assembly that prevents significant quantities of corium from entering the drain line. By protecting the drain line, essentially no high-activity fission products would be released to the reactor building or the environment during a severe accident. The ceramic drain plug assembly includes a drain plug base and a drain plug supported by a steel pedestal. The lower surface of the plug has a spherical shape such that the plug can be positioned within the base to block access to the drain opening provided in a central portion of the base. During normal operation conditions, the plug is retained above the base by the pedestal. During a severe accident, when corium comes into contact with the pedestal, it will melt rapidly and the drain plug will drop by gravity, effectively closing the sump drain opening and preventing the flow of corium into the drain line. 1. In a nuclear reactor having a reactor pressure vessel and a sump , a drain plug assembly , comprising:a base coupled to the sump, said base formed of a ceramic material and defining an interior surface;a plug formed of a ceramic material and defining an exterior surface configured to matingly engage said base interior surface; anda fastener interconnecting said base and said plug such that said base and said plug cooperatively define an opening to allow fluid flow therebetween.2. The drain plug assembly of claim 1 , wherein:said base and said plug are formed of a ceramic material having a melting temperature of over 3000° C.; andsaid fastener is formed of a material having a melting temperature with the range of approximately 1300° C. to 1500° C.3. The drain plug assembly of claim 1 , wherein:said base defines a set of holes therein; andsaid fastener includes a set of protuberances positioned within said set of holes.4. The drain plug assembly of claim 3 , wherein:said base defines a plurality of holes therein; andsaid fastener includes a basal ring having a plurality of ...

Containment Building Separation System at a Nuclear Power Plant

The invention is related to safety systems for nuclear power plants (NPP) which can be used in various operational modes, including emergency mode, and is aimed at controlling air flows inside NPP containment buildings. 1. NPP containment building separation system dividing the NPP containment into isolated rooms , comprising the containment building separation system installed on the floor slab between the rooms and located in the circular gap between the floor slab and the containment building wall , and including , at least , one isolating valve to ensure insulation of the airspace in the containment building rooms , and is configured to connect the airspace in the containment building rooms following the pressure drop which may occur , wherein additionally the system contains an air supply unit connected to the manifold ring , being connected to each of the valves in the containment building separation system; wherein each of the valves is designed as an air-inflated shutter aimed at providing insulation of the airspace inside the containment building rooms when inflated and at connecting the airspace when deflated.2. A containment building separation system according to claim 1 , wherein the air-inflated shutters are made of fabric.3. A containment building separation system according to claim 2 , wherein the air-inflated shutters are made of resin-coated fabric.4. A containment building separation system according to claim 1 , comprising support structure elements installed on the floor slab dividing the rooms from each other claim 1 , and the air-inflated shutters are attached to the support structure elements.5. A containment building separation system according to claim 1 , wherein the air-inflated shutters adjoin to each other.6. A containment building separation system according to claim 1 , wherein the vertical service tunnels are arranged between some of the air-inflated shutters.7. A containment building separation system according to claim 1 , wherein ...

VERY SIMPLIFIED BOILING WATER REACTORS FOR COMMERCIAL ELECTRICITY GENERATION

Nuclear reactors have very few systems for significantly reduced failure possibilities. Nuclear reactors may be boiling water reactors with natural circulation-enabling heights and smaller, flexible energy outputs in the 0-350 megawatt-electric range. Reactors are fully surrounded by an impermeable, high-pressure containment. No coolant pools, heat sinks, active pumps, or other emergency fluid sources may be present inside containment; emergency cooling, like isolation condenser systems, are outside containment. Isolation valves integral with the reactor pressure vessel provide working and emergency fluid through containment to the reactor. Isolation valves are one-piece, welded, or otherwise integral with reactors and fluid conduits having ASME-compliance to eliminate risk of shear failure. Containment may be completely underground and seismically insulated to minimize footprint and above-ground target area. 1. A simplified nuclear reactor system for commercially generating electricity , the system comprising:a nuclear reactor;at least one primary coolant loop connecting to the nuclear reactor; andat least one emergency coolant source connecting to the nuclear reactor, wherein the nuclear reactor is integrally isolatable from the primary coolant loop and the emergency coolant source.2. The system of claim 1 , further comprising:a containment, wherein the reactor is inside the containment, wherein the emergency coolant source is outside containment, and wherein the containment is entirely underground.3. The system of claim 2 , wherein the containment has a personnel access point at a top shield accessible from ground.4. The system of claim 2 , wherein the nuclear reactor is a maximum 1000 megawatt-thermal rated boiling water reactor having a height that exceeds its width by a factor of at least 3.9.5. The system of claim 2 , wherein the containment does not include any open coolant pool for emergency cooling.6. The system of claim 1 , further comprising:a plurality ...

VERY SIMPLIFIED BOILING WATER REACTORS FOR COMMERCIAL ELECTRICITY GENERATION

Nuclear reactors have very few systems for significantly reduced failure possibilities. Nuclear reactors may be boiling water reactors with natural circulation-enabling heights and smaller, flexible energy outputs in the 0-350 megawatt-electric range. Reactors are fully surrounded by an impermeable, high-pressure containment. No coolant pools, heat sinks, active pumps, or other emergency fluid sources may be present inside containment; emergency cooling, like isolation condenser systems, are outside containment. Isolation valves integral with the reactor pressure vessel provide working and emergency fluid through containment to the reactor. Isolation valves are one-piece, welded, or otherwise integral with reactors and fluid conduits having ASME-compliance to eliminate risk of shear failure. Containment may be completely underground and seismically insulated to minimize footprint and above-ground target area. 1. A simplified nuclear reactor system for commercially generating electricity , the system comprising:a nuclear reactor;at least one primary coolant loop connecting to the nuclear reactor; andat least one emergency coolant source connecting to the nuclear reactor, wherein the nuclear reactor is integrally isolatable from the primary coolant loop and the emergency coolant source.2. The system of claim 1 , further comprising:a containment, wherein the reactor is inside the containment, wherein the emergency coolant source is outside containment, and wherein the containment is entirely underground.3. The system of claim 2 , wherein the containment has a personnel access point at a top shield accessible from ground.4. The system of claim 2 , wherein the nuclear reactor is a maximum 1000 megawatt-thermal rated boiling water reactor having a height that exceeds its width by a factor of at least 3.9.5. The system of claim 2 , wherein the containment does not include any open coolant pool for emergency cooling.6. The system of claim 1 , further comprising:a plurality ...

MANAGING DYNAMIC FORCES ON A NUCLEAR REACTOR SYSTEM

A nuclear reactor seismic isolation assembly includes an enclosure that defines a volume; a plastically-deformable member mounted, at least in part, within the volume; and a stretching member moveable within the enclosure to plastically-deform the plastically-deformable member in response to a dynamic force exerted on the enclosure. 1. (canceled)2. A nuclear reactor seismic isolation assembly , comprising:an enclosure that defines a volume;a deformable member mounted, at least in part, within the volume; anda stretching member mounted within a bore that extends at least partially through the deformable member, wherein the stretching member is configured to deform the deformable member in a substantially transverse direction to a linear movement of the stretching member into the bore in response to a dynamic force exerted on the enclosure.3. The nuclear reactor seismic isolation assembly of claim 2 , wherein:a first portion of the bore comprises a first diameter substantially equal to an outer dimension of the stretching member;a second portion of the bore comprises a second diameter smaller than the first diameter; andthe second diameter is stretched to substantially equal the first diameter based on the linear movement of the stretching member into the bore.4. The nuclear reactor seismic isolation assembly of claim 3 , wherein the deformable member comprises a first portion mounted within the enclosure and a second portion that extends through a die member to an exterior of the enclosure.5. The nuclear reactor seismic isolation assembly of claim 2 , wherein the stretching member deforms an inside surface of the deformable member.6. The nuclear reactor seismic isolation assembly of claim 2 , wherein the bore at least partially encloses a fluid that dissipates at least a portion of energy generated by the dynamic force exerted on the enclosure.7. The nuclear reactor seismic isolation assembly of claim 6 , further comprising an orifice that fluidly couples the bore to ...

SMALL NUCLEAR REACTOR CONTAINMENT SYSTEM

A nuclear reactor containment system including a nuclear reactor and a container enclosing the nuclear reactor. The container includes a number of heat removal systems each having an active state and an inactive state, wherein the heat removal systems dissipate heat from the container more efficiently in the active state than in the inactive state, and wherein the heat removal systems are structured to switch from the inactive state to the active state based on a temperature of the container. 1. A nuclear reactor containment system comprising:a nuclear reactor; 'a number of heat removal systems each having an active state and an inactive state, wherein the heat removal systems dissipate heat from the container more efficiently in the active state than in the inactive state, and wherein the heat removal systems are structured to switch from the inactive state to the active state based on a temperature of the container.', 'a container enclosing the nuclear reactor, the container including2. The nuclear reactor containment system of claim 1 , wherein the heat removal systems are structured to switch from the inactive state to the active state at one or more predetermined temperatures of the container above temperatures of the container corresponding to normal operation of the nuclear reactor.3. The nuclear reactor containment system of claim 1 , wherein the number of heat removal systems includes a first heat removal system including:fins disposed in an outer portion of the container and forming a plurality of cooling channels; anda plurality of air regulating mechanisms structured to block air from flowing through the cooling channels when the first heat removal system is in the inactive state and to allow air to flow through the cooling channels when the first heat removal system is in the active state.4. The nuclear reactor containment system of claim 3 , wherein at least one of the air regulating mechanisms includes:a bimetallic strip disposed in one of the cooling ...

VALVE ASSEMBLY WITH ISOLATION VALVE VESSEL

Apparatuses for reducing or eliminating Type 1 LOCAs in a nuclear reactor vessel. A nuclear reactor including a nuclear reactor core comprising a fissile material, a pressure vessel containing the nuclear reactor core immersed in primary coolant disposed in the pressure vessel, and an isolation valve assembly including, an isolation valve vessel having a single open end with a flange, a spool piece having a first flange secured to a wall of the pressure vessel and a second flange secured to the flange of the isolation valve vessel, a fluid flow line passing through the spool piece to conduct fluid flow into or out of the first flange wherein a portion of the fluid flow line is disposed in the isolation valve vessel, and at least one valve disposed in the isolation valve vessel and operatively connected with the fluid flow line. 1. An apparatus comprising: an isolation valve vessel;', 'a mounting flange sealing with the isolation valve vessel to define a sealed volume;', 'a fluid flow line in fluid communication with the mounting flange to flow fluid through the mounting flange; and', 'a valve disposed in the isolation valve vessel inside the sealed volume and operatively connected with the fluid flow line., 'an isolation valve assembly including2. The apparatus of claim 1 , wherein the isolation valve assembly further includes a forging including the mounting flange and a second flange to which the isolation valve vessel is secured claim 1 , the forging having a passageway extending between the mounting flange and the second flange through which the fluid flow line passes.3. The apparatus of claim 1 , wherein the valve is a check valve allowing flow out of the mounting flange and blocking flow into the mounting flange.4. The apparatus of claim 1 , wherein the valve includes first and second valves disposed in the isolation valve vessel inside the sealed volume and arranged in series along the fluid flow line.5. The apparatus of claim 1 , wherein the isolation valve ...

NUCLEAR REACTOR CAVITY FLOOR PASSIVE HEAT REMOVAL SYSTEM

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor. 1. A nuclear island comprising:a nuclear reactor including a reactor core comprising fissile material disposed in a reactor pressure vessel;a radiological containment containing the nuclear reactor, the radiological containment including a concrete floor located underneath the nuclear reactor; andan ex vessel corium retention system including flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels.2. The nuclear island of wherein the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation.3. The nuclear island of further comprising:a refueling water storage tank (RWST) disposed inside the radiological containment and connected with the inlet to drain water from the ...

Method and device to prevent severe power and flow oscillations in boiling water reactors

The present invention relates to boiling water reactors (BWR) disclosing a new method and device for preventing the flow and power oscillations from growing to severely large amplitudes thus protecting the reactor from the consequences of instabilities associated with the so-called anticipated transients without scram (ATWS). This invention introduces a new method for preventing boiling water reactor fuel damage due to the growth of unstable density wave oscillations to severely large magnitudes. The method limits the growth of the density waves by limiting the magnitude of the oscillation of the coolant mass flow rate at the inlet of the fuel bundle such that only upward flow is permitted. Further growth of the density wave beyond inlet flow reversal is thus prevented which limits the severity of the coolant flow conditions at the fuel clad surface in the entire fuel assembly such that dryout either does not occur or rewetting of the clad surface occurs every oscillation cycle such that excessive high clad temperatures that may cause its failure do not occur. A device realizing this method is comprised of a screen structure placed inside the lower tie plate of the fuel assembly. The said screen is free to move between two plates with aligned holes. The said screen is lifted upward such that the flow holes are not obstructed allowing upward flow of the coolant as required for normal operation. In the case the flow direction is reversed the hydraulic lifting force vanishes and the screen structure moves to a lower position blocking flow holes and obstructing flow in the reverse (downward) direction. This invention is particularly useful for limiting the consequences of unstable oscillation should the ability of the reactor operator to shut down the reactor power with control rod scram be lost as part of the hypothetical scenario known as anticipated transient without scram (ATWS). It is also useful to mitigate the consequences from a loss-of-coolant accident. This ...

Reactor Containment Vessel Vent System

The invention provides a reactor containment vessel vent system capable of continuously releasing steam generated in a reactor containment vessel to the atmosphere even when a power supply is lost. In the reactor containment vessel vent system (), the noble gas filter () that allows steam to pass through but does not allow radioactive noble gases to pass through among vent gas discharged from the reactor containment vessel () is provided at a most downstream portion of the vent line. An immediate upstream portion of the noble gas filter () and the reactor containment vessel () are connected to each other by the return pipe () via the intermediate vessel (). Further, when the radioactive noble gases having pressure equal to or higher than predetermined pressure stays in the immediate upstream portion of the noble gas filter (), the staying radioactive noble gases flows into the intermediate vessel () by the relief valve (). Thus, the noble gas filter () does not lose steam permeability, and the reactor containment vessel vent system () can continuously release the steam to the atmosphere. 1. A reactor containment vessel vent system that reduces pressure in a reactor containment vessel by releasing gas in the reactor containment vessel to the atmosphere , the reactor containment vessel vent system comprising:a vent line that forms a vent gas flow path through which vent gas is discharged from the reactor containment vessel and released to the atmosphere;a noble gas filter provided at a most downstream portion of the vent line, the noble gas filter allowing at least steam to pass through and not allowing radioactive noble gases to pass through among the vent gas;a return pipe that connects an immediate upstream portion of the noble gas filter in the vent line and the reactor containment vessel; andan intermediate vessel provided on the return pipe, in which gas containing the radioactive noble gases that cannot permeate the noble gas filter flows and is stored.2. The ...

Nuclear Reactor Containment Vessel and Nuclear Reactor

The invention relates to a nuclear power generation plant, and a nuclear reactor containment vessel includes a containment vessel covering a nuclear reactor pressure vessel, an air-cooled heat exchanger which is installed outside the containment vessel and performs heat exchange between steam in the containment vessel and air outside the containment vessel, and a square column-shaped air flow path provided vertically above the air-cooled heat exchanger.

LIQUID NITROGEN EMERGENCY COOLING SYSTEM FOR NUCLEAR POWER PLANTS

A reactor cooling system for cooling a nuclear reactor using nitrogen comprising a refrigeration unit for cooling and compressing nitrogen gas into liquid nitrogen, a liquids storage tank to store liquid nitrogen, the tank in fluid communication with the refrigeration unit, a heat exchanger drop system in fluid communication with the liquids storage tank, adjacent to the nuclear reactor, wherein the nitrogen absorbs heat by becoming gaseous, a tank for receiving and holding nitrogen gas in fluid communication with the heat exchanger and in fluid communication with the refrigeration unit, and where the system is a closed-loop drop system. 1. A reactor cooling system for cooling a nuclear reactor using nitrogen comprising:a. a refrigeration unit for cooling and compressing nitrogen gas into liquid nitrogen,b. a liquids storage tank to store liquid nitrogen, the tank in fluid communication with the refrigeration unit,c. a heat exchanger drop system in fluid communication with the liquids storage tank, adjacent to the nuclear reactor, wherein the nitrogen absorbs heat by becoming gaseous,d. a tank for receiving and holding nitrogen gas in fluid communication with the heat exchanger and in fluid communication with the refrigeration unit, wherein the system is a closed-loop system.2. The system of further comprising a gas-powered generating unit claim 1 , for generating electricity from the nitrogen gas as it expands.3. The system of further comprising a hydraulic system for using the power of the expanding gas from an outlet of the heat exchanger drop.4. The system of wherein the hydraulic system can either be used to restart the nuclear power plant or to provide hydraulic power.5. The system of wherein hydraulic system opens and shuts valves as needed for the safe continued operation of under normal circumstances claim 3 , in the event of a near failure claim 3 , and for emergency shut down.6. The system of further comprising an overpressure relief valve system for ...

Containment Internal Passive Heat Removal System

The invention relates to the nuclear energy field, including pressurized water reactor containment internal passive heat removal systems. The invention increases heat removal efficiency, flow stability in the circuit, and system reliability. The system has at least one cooling water circulation circuit comprising a heat exchanger inside the containment and including an upper and lower header interconnected by heat-exchange tubes, a riser pipeline and a downtake pipeline connected to the heat exchanger, a cooling water supply tank above the heat exchanger outside the containment and connected to the downtake pipeline, a steam relief valve connected to the riser pipeline and located in the water supply tank and hydraulically connected to the latter. The upper and lower header of the heat exchanger are divided into heat exchange tube sections on the assumption that: L/D≦20, L being the header section length, D being the header bore. 1. A pressurized water reactor containment internal passive heat removal system with at least one cooling water circulation circuit , comprising:a heat exchanger located inside the containment and comprising an upper header and a lower header interconnected by heat-exchange tubes,a riser pipeline and a downtake pipeline connected to the heat exchanger,a cooling water supply tank located above the heat exchanger outside the containment and connected to the downtake pipeline, {'br': None, 'i': 'L/D≦', '20,'}, 'a steam relief valve connected to the riser pipeline, located in the water supply tank and connected to the same hydraulically, wherein the upper and the lower headers are divided into heat-exchange tube sections on the assumption thatwhere L is the header section length,D is the header bore,{'sub': 'rs', 'claim-text': [{'br': None, 'i': P', 'gh', 'gh, 'sup': 'c', 'sub': res', 'rs', 'rs', 'he', 'he, 'Δ=Δρ+Δρ,'}, {'br': None, 'i': h', 'P', 'gh', 'g,, 'sub': rs', 'res', 'he', 'he', 'rs, 'sup': 'c', '=(Δ−Δρ)/Δρ'}], 'the riser pipeline ...

Nuclear Power Plant

In view of above problems, an object of the invention is to provide a primary containment vessel venting system having a structure capable of continuously discharging vapor in a primary containment vessel out of the system and continuously reducing pressure of the primary containment vessel without discharging radioactive noble gases to the outside of the containment vessel and without using an enclosing vessel or a power source. In order to achieve the above object, an nuclear power plant of the invention includes a primary containment vessel which includes a reactor pressure vessel, a radioactive substance separation apparatus which is disposed inside the primary containment vessel and through which the radioactive noble gases do not permeate but vapor permeates, a vent pipe which is connected to the radioactive substance separation apparatus, and an exhaust tower which is connected to the vent pipe and discharges a gas, from which a radioactive substance is removed, to the outside. 115.-. (canceled)16. A nuclear power plant , comprising:a reactor containment vessel which includes a reactor pressure vessel;a radioactive substance separation apparatus which is disposed inside the reactor containment vessel, and through which a radioactive noble gas does not permeate but vapor permeates; andan exhaust tower which is configured to discharge a gas from which the radioactive noble gas is removed by the radioactive substance separation apparatus.17. The nuclear power plant according to claim 16 , further comprising:a wet or dry filtered containment venting system between the radioactive substance separation apparatus and the exhaust tower.18. The nuclear power plant according to claim 17 , further comprising:a bypass pipe which is configured to send a gas inside the reactor containment vessel to the filtered containment venting system without passing through the radioactive substance separation apparatus.19. The nuclear power plant according to claim 18 , further ...

INTEGRAL PRESSURE VESSEL PENETRATIONS AND SYSTEMS AND METHODS FOR USING AND FABRICATING THE SAME

Pressure vessels have full penetrations that can be opened and closed with no separate valve piping or external valve. A projected volume from the vessel wall may house valve structures and flow path, and these structures may move with an external actuator. The flow path may extend both along and into the projected volume. Vessel walls may remain a minimum thickness even at the penetration, and any type of gates may be used with any degree of duplication. Penetrations may be formed by installing valve gates directly into the channel in the wall. The wall may be built outward into the projected volume by forging or welding additional pieces integrally machining the channel through the same volume and wall. Additional passages for gates and actuators may be machined into the projections as well. Pressure vessels may not require flanges at join points or material seams for penetration flow paths. 1. A pressure vessel comprising:a wall defining an interior and an exterior of the pressure vessel; anda penetration integral with the wall forming a flow path through the wall, wherein the penetration includes an integral valve openable and closeable from the exterior.2. The pressure vessel of claim 1 , wherein the penetration includes a hub integral with the wall and extending outward toward the exterior claim 1 , and wherein the valve is in the hub.3. The pressure vessel of claim 2 , wherein the hub defines the flow path from the exterior to the interior claim 2 , wherein the valve further includes at least one gate in the flow path configured to open and close the flow path.4. The pressure vessel of claim 1 , wherein the penetration includes no external flange or structure compressed to the valve or the wall.5. The pressure vessel of claim 1 , wherein the flow path extends in at least two different dimensions through the penetration.6. The pressure vessel of claim 1 , wherein the valve further includes at least one gate in the flow path configured to open and close the ...

Nuclear reactor with improved cooling in an accident situation

A nuclear reactor including a vessel configured to hold a reactor core, a primary circuit cooling the reactor, a reactor pit in which the vessel is placed, an annular channel surrounding a lower portion of the vessel in the reactor pit, the channel configured to act as a thermal shield in normal operation and to ascend flow of a liquid in event of an accident, a reserve of liquid capable of filling the reactor pit, a reactor containment, a chamber collecting steam generated at an upper end of the reactor pit, the chamber being separate from the containment, a circulating pump capable of generating a forced convection of the liquid in the annular channel, and a lobe pump or steam piston machine or turbine for actuating the circulating pump and capable of generating forced convection by the collected steam.

Atomic power station having an atomic reactor accommodated in a largely fail-safe fashion

In the presently known atomic power stations, both the atomic reactor and the remaining units belonging to the atomic power station are located above ground at the same level. In the event of a serious reactor accident, it is therefore frequently scarcely to be avoided that the environment is radioactively polluted (contaminated). If the atomic reactor is installed underground in the vicinity of the power station units, an environmental hazard due to pollution by rays can be largely or virtually completely prevented by means of a strong permanent covering or by a covering which is to be applied in the event of an accident. It is also relatively easy to extinguish a burnt-out reactor in the case of a reactor which is accommodated in a pit. The said measures can be applied individually or in combination.

Containment protection system for a nuclear facility and associated operating method

A containment protection system (2) for treating the atmosphere present in the containment (4) of a nuclear facility (6), particularly a nuclear power plant, in case of critical incidents involving extensive release of hydrogen (H 2 ) and steam is to be able to effectively and quickly relieve such conditions in a largely passive manner and where possible without contaminating the environment. According to the invention, the containment protection system (2) has for this purpose a circuit, which comprises a conduction system (10, 72, 120, 128) and which is provided for connecting to the containment (4), out of the containment (4) and back again for a fluid flow, more particularly having the following components fluidically connected in series: a recombination device (20) for recombining hydrogen (H2) contained in the fluid flow with oxygen (O2) to form steam (H2O); a condensation device (74) connected downstream of the recombination device (20) for condensing steam fractions contained in the fluid flow with means for diverting the condensate (94) out of the fluid flow; drive means (18, 180) for the fluid flow, a heat exchanger (96) being provided for an at least partial re-cooling of the condensation device (74) and being connected on the inlet side via a feed line (144) to a storage tank (140) for liquid nitrogen (N2).

NUCLEAR REACTOR.

Nuclear facility

Bei einer kerntechnischen Anlage (1) mit einer Sicherheitshülle (2), an die eine Druckentlastungsleitung (6) angeschlossen ist, in die in Reihe ein in einem Behälter (14) mit einer Waschflüssigkeit (W) angeordneter Venturiwäscher (12) sowie eine Drosseleinrichtung (24) geschaltet sind, sollen im Falle einer Druckentlastung auch feinste luftgetragene Aktivitäten oder Aerosole mit besonders hoher Zuverlässigkeit im Venturiwäscher (12) zurückgehalten werden, so dass eine Freisetzung an die Umgebung mit besonders hoher Zuverlässigkeit ausgeschlossen ist. Dazu sind erfindungsgemäß der Venturiwäscher (12) und die Drosseleinrichtung (24) derart dimensioniert, dass sich bei einer kritischen Entspannung eines in der Druckentlastungsleitung (6) strömenden Luft-Dampf-Gemisches an der Drosseleinrichtung (24) im Venturiwäscher (12) eine Strömungsgeschwindigkeit des Luft-Dampf-Gemisches von mehr als 150 m/s, vorzugsweise von mehr als 200 m/s, einstellt. In a nuclear installation (1) with a safety envelope (2) to which a pressure relief line (6) is connected, into which a Venturi scrubber (12) arranged in a container (14) with a scrubbing liquid (W) and a throttling device ( 24) are connected, in the case of pressure relief even the finest airborne activities or aerosols with particularly high reliability in Venturi scrubber (12) are retained, so that release to the environment is excluded with particularly high reliability. For this purpose, the Venturi scrubber (12) and the throttle device (24) according to the invention are dimensioned such that at a critical relaxation of an air-vapor mixture in the pressure relief line (6) at the throttle device (24) in Venturi scrubber (12) Air-steam mixture of more than 150 m / s, preferably more than 200 m / s, sets.

核电站减压方法、核电站减压系统以及相应的核电站

本发明涉及用于核电站(2)减压的方法和相应装置，该核电站包括封罩放射性载体的安全壳(4)和用于泄压流的出口(10,10')，泄压流通过配设有过滤系统的泄流管道(12,12')从安全壳(4)被导入大气，该过滤系统包括具有过滤室入口(124)、过滤室出口(128)和位于其间的吸附过滤器(18)的过滤室(16)，该泄压流首先在高压部段(70)中导向流动，随后在节流机构(72)处被膨胀减压，随后至少部分被导送经过带有吸附过滤器(18)的过滤室(16)，最后被吹出到大气中。为了能实现对泄压流所含放射性载体的很高效的有效截留，本发明规定，通过该节流机构(72)被减压的泄压流就在其即将进入过滤室(16)之前被引导经过过热部段(80)，该泄压流在过热部段中通过来自在高压部段(70)内的尚未减压的泄压流的直接传热或间接传热被加热到这样的温度，该温度比存在于那里的露点温度高至少10℃，优选高20℃～50℃。

PROCEDURE AND DEVICE FOR THE PRODUCTION OF AN INERTIZATION GAS.

LA INVENCION SE REFIERE A UN PROCEDIMIENTO PARA LA GENERACION DE UN GAS INERTE EN LA ALIMENTACION DE UN RECIPIENTE, EN PARTICULAR UN RECIPIENTE (3) DE SEGURIDAD DE UN REACTOR NUCLEAR. UN GAS (1) INERTE ES MANTENIDO EN ESTADO LICUADO O SOLIDIFICADO EN UN PRIMER ACUMULADOR (4); EN UN SEGUNDO ACUMULADOR (4), SE DISPONE DE SUFICIENTE CALOR PARA EVAPORAR EL GAS (1) INERTE SOLIDIFICADO O LICUADO, ESTANDO DISPONIBLE EN UN MEDIO (2) DE TRANSFERENCIA TERMICA; Y EL MEDIO DE TRANSFERENCIA TERMICA Y GAS (1) INERTE LICUADO O SOLIDIFICADO SON APLICADOS EN CONTACTO TERMICO UNO CON OTRO. SE DESCRIBE TAMBIEN UN DISPOSITIVO PARA LA GENERACION DE UN GAS INERTE. ESTE PROCEDIMIENTO Y DISPOSITIVO SON ADECUADOS PARTICULARMENTE PARA LA GENERACION DE UNA GRAN CANTIDAD DE GAS INERTE, DE TAL FORMA QUE EL RECIPIENTE (3) DE SEGURIDAD DE UN REACTOR NUCLEAR PUEDE SER INERTIZADO DE MANERA RAPIDA GARANTIZADA. THE INVENTION REFERS TO A PROCEDURE FOR THE GENERATION OF AN INERT GAS IN THE FEEDING OF A CONTAINER, IN PARTICULAR A SAFETY CONTAINER (3) OF A NUCLEAR REACTOR. AN INERT GAS (1) IS KEPT IN A LIQUEFIED OR SOLIDIFIED CONDITION IN A FIRST ACCUMULATOR (4); IN A SECOND ACCUMULATOR (4), ENOUGH HEAT IS AVAILABLE TO EVAPORATE THE GAS (1) SOLIDIFIED OR LIQUEFIED INERT, BEING AVAILABLE IN A THERMAL TRANSFER MEDIA (2); AND THE THERMAL TRANSFER MEDIA AND GAS (1) LIQUEFIED OR SOLIDIFIED INERT ARE APPLIED IN THERMAL CONTACT WITH EACH OTHER. A DEVICE FOR THE GENERATION OF AN INERT GAS IS ALSO DESCRIBED. THIS PROCEDURE AND DEVICE ARE PARTICULARLY SUITABLE FOR THE GENERATION OF A LARGE AMOUNT OF INERT GAS, SO THAT THE SAFETY CONTAINER (3) OF A NUCLEAR REACTOR CAN BE INERTIZED QUICKLY GUARANTEED.

Method and device for producing deactivating gas

FIELD: producing deactivating gases for nuclear power and other industries. SUBSTANCE: for producing deactivating gas to be introduced in vessel such as containment 3 of nuclear power plant, inert gas 1 in liquid or solid state is stored in first storage tank 4; second storage tank 5 is used to store heat accumulated in coolant 2 in amount sufficient for evaporating gas 1. Then coolant 2 is brought in contact with liquefied or solidified inert gas. Device for producing deactivating gas is given in description of invention. EFFECT: enlarged quantity of deactivating gas produced for deactivating nuclear power plant containment. 19 cl, 6 dwg АСС ПЧ ГЭ (19) РОССИЙСКОЕ АГЕНТСТВО ПО ПАТЕНТАМ И ТОВАРНЫМ ЗНАКАМ (51) МПК 13) ВИ” 2142 171 Сл С 21С 9/06, А 62 С 3/00 12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ (21), (22) Заявка: 97101489/12, 20.06.1995 (24) Дата начала действия патента: 20.06.1995 (30) Приоритет: 04.07.1994 ОЕ Р 4423400.7 (46) Дата публикации: 21.11.1999 (56) Ссылки: ОЕ 3927958 АД, 22.03.90. $Ц 1829697 АЛ, 09.06.95. ОЕ 2218518 В1, 14.02.14. ОЕ 2627055 АЛ, 20.01.77. ОЕ 4021612 АЛ, 09.01.92. ЕК 2443255 АЛ, 04.07.80. ЕК 2314 73ЗТА, 14.01.7Г. СВ 2090736 А, 21.07.82. СВ 2202440 А, 28.09.88. \МО 9309848 АЛ, 27.05.93. ЕР 0640990 АЛ, 01.03.35. (85) Дата перевода заявки РСТ на национальную фазу: 04.02.97 (86) Заявка РСТ: ОЕ 95/00799 (20.06.95) (87) Публикация РСТ: М/О 96/01477 (18.01.96) (98) Адрес для переписки: 103735, Москва, ул.Ильинка 5/2, Союзпатент, Патентному поверенному Дудушкину С.В. (71) Заявитель: Сименс АГ (0Е) (72) Изобретатель: Бернд Экардт (0Е) (73) Патентообладатель: Сименс АГ (ОЕ) (54) СПОСОБ И УСТРОЙСТВО ДЛЯ ПОЛУЧЕНИЯ ИНЕРТИЗИРУЮЩЕГО ГАЗА (57) Реферат: Изобретение относится к способу для получения инертизирующего газа для введения в резервуар, в частности в оболочку безопасности (3) атомной электростанции. Инертный газ (1) в сжиженной или отвержденной форме запасают в первом накопителе (4) и во втором накопителе (5) готовят достаточное для ...

Nuclear reactor with improved cooling in emergency situation

FIELD: power engineering. SUBSTANCE: nuclear reactor comprises a tank (4), where the reactor core is installed, the primary circuit for reactor cooling, a well (6) of the tank, where the tank (4) is installed, a circular channel (16), which surrounds the lower part of the tank (4) in the well (6) of the tank, a liquid reservoir for filling of the tank well, a tight vessel (22) of the reactor, a chamber (26) for collection of steam generated in the upper end of the tank well (6), separated from the tight vessel (22), a circulating pump (40) and a blade pump or a steam piston machine (32) for actuation of the circulating pump (40). At the same time the channel (16) is designed for performance of the function of a heat shielding screen under normal operation and for provision of upstream circulation of the liquid in case of an emergency, and the circulating pump is made as capable of creating forced convection with the help of the collected steam. EFFECT: increased level of passive emergency protection of a reactor tank against melting through. 24 cl, 6 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) 2 496 163 (13) C2 (51) МПК G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ (21)(22) Заявка: 2010120709/07, 20.10.2008 (24) Дата начала отсчета срока действия патента: 20.10.2008 (73) Патентообладатель(и): КОММИССАРИАТ А Л'ЭНЕРЖИ АТОМИК Э ОЗ ЭНЕРЖИ АЛЬТЕРНАТИВ (FR) (43) Дата публикации заявки: 27.11.2011 Бюл. № 33 2 4 9 6 1 6 3 (45) Опубликовано: 20.10.2013 Бюл. № 29 (56) Список документов, цитированных в отчете о поиске: US 4571323 А, 18.02.1986. RU 2099801 C1, 20.12.1997. RU 2063071 C1, 27.06.1996. US 5825838 A, 20.10.1998. 2 4 9 6 1 6 3 R U (86) Заявка PCT: EP 2008/064088 (20.10.2008) C 2 C 2 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 24.05.2010 (87) Публикация заявки РСТ: WO 2009/053322 (30.04.2009) Адрес для переписки: 109012, Москва, ул. Ильинка, 5/2, ООО "Союзпатент" (54) ЯДЕРНЫЙ РЕАКТОР С ...

Containment filtered venting system having hydrogen reduction unit

본 발명에 따른 원자력발전소의 여과배기장치는 원자로가 격납된 격납건물에서 발생된 수증기 및 핵분열 생성물이 유동 가능하도록 방출배관을 통해 상기 격납건물과 연결되는 원자력발전소의 여과배기장치에 관한 것으로써, 내부에 수용된 세척액 내에 잠겨있도록 배치되는 노즐, 상기 노즐을 통해 방출되어 상기 세척액을 통과한 기체를 외부로 배출시키는 배기배관 및 상기 세척액 및 상기 배기배관 사이에 배치되는 수소저감부를 포함한다. The filtration and exhaust apparatus of a nuclear power plant according to the present invention relates to a filtration and exhaust apparatus of a nuclear power plant connected to a containment building through a discharge pipe so that water vapor and fission products generated in a containment building containing the reactor can flow, And a hydrogen reduction unit disposed between the cleaning liquid and the exhaust pipe. The exhaust pipe may include a nozzle disposed so as to be submerged in the cleaning liquid contained in the cleaning liquid.