УСТРОЙСТВО ДЛЯ РЕКОМБИНАЦИИ ВОДОРОДА В ГАЗОВОЙ СМЕСИ

Изобретение относится к устройству для рекомбинации водорода в газовой смеси, в частности, для атомной электростанции. Устройство для рекомбинации водорода в газовой смеси, в частности, для атомной электростанции, при эксплуатации которого особенно надежно предотвращается нежелательное воспламенение газовой смеси, содержит катализаторную систему, которая в случае эксплуатации расположена в протекаемом в свободной конвекции для газовой смеси корпусе и которой придано в соответствие устройство удержания пламени. При этом в устройство удержания пламени предпочтительно встроен улавливатель осадка так, что выход отделяющихся из катализаторной системы горячих катализаторных частиц против направления течения газовой смеси надежно предотвращается. Технический результат - обеспечение беспламенной эксплуатации устройства рекомбинации. 19 з.п. ф-лы, 4 ил.

ВЫСОКОЭНЕРГЕТИЧЕСКАЯ ПОГЛОЩАЮЩАЯ НАСАДКА ДЛЯ ТЕПЛОВЫДЕЛЯЮЩЕЙ СБОРКИ ЯДЕРНОГО РЕАКТОРА

Устройство предназначено для использования в ядерных реакторах. Высокоэнергетическая поглощающая насадка для тепловыделяющей сборки, в которой используется верхний удлиненный цилиндрический корпус и нижний удлиненный цилиндрический корпус, перемещающийся в верхнем цилиндрическом корпусе. Верхний и нижний корпуса смещены друг от друга посредством множества предельных пружин, которые ограничены продольно перемещающимся поршнем, перемещение которого в верхнем направлении ограничено стопором в верхнем цилиндрическом корпусе. Энергия, сообщаемая насадке регулирующим стержнем, поглощается, в основном, пружинами и гидравлическим действием поршня в насадке. Ударные нагрузки во время аварийного отключения поглощаются без повреждения насадки, тепловыделяющей сборки и/или сборки регулирующих стержней. 3 н. и 12 з.п. ф-лы, 1 ил.

ПЛАВУЧАЯ АТОМНАЯ ЭЛЕКТРОСТАНЦИЯ

Изобретение относится к атомной энергетике и может быть использовано при создании плавучих атомных электростанций средней мощности для эксплуатации в прибрежных зонах. Сущность изобретения: атомная электростанция выполнена в виде несамоходного судна с упрощенными формами корпуса. Электростанция включает платформу, установленные на ней и прилегающие друг к другу прочные корпуса. В корпусах размещены в замкнутых прочных профилированных выгородках энергоблок с реакторной зоной, топливный блок, системы обеспечения безопасности, системы обслуживания и вакуумирования энергоблока, системы вентиляции и энергоснабжения, машинный отсек с турбогенератором, дополнительные системы управления судном и жизнеобеспечения. Платформа выполнена в виде корпуса судна с двойным бортом, двойным днищем и главной, верхней, средней и нижней палубами. Корпус судна разделен поперечными герметичными переборками. Переборки образуют стены прочных корпусов для размещения выгородок. Реакторная зона включает реактор, контуры ...

УСТРОЙСТВО ДЛЯ КОНДЕНСАЦИИ И ОЧИСТКИ ПАРОГАЗОВОЙ СМЕСИ, ПРЕИМУЩЕСТВЕННО ПРИ АВАРИЯХ НА АТОМНЫХ ЭЛЕКТРОСТАНЦИЯХ (ВАРИАНТЫ)

Изобретение относится к области атомной энергетики и может быть использовано в энергетической, химической, металлургической, нефтехимической, газодобывающей и других отраслях промышленности для конденсации и очистки пара или газа, а также их смесей. Устройство, содержащее сливную емкость для конденсата с вертикальной обечайкой в виде тела вращения, внутренняя полость которой сообщается с подводящим паропроводом посредством тангенциальных сопел, дополнительно снабжено второй вертикальной обечайкой, внутренняя полость которой сообщается с внутренней полостью первой обечайки через проем, образованный стыками боковых стенок обечаек, которые образуют в зоне стыков ω-образные профили, а сопла размещены в проеме и направлены на стыки боковых стенок обечаек. Во втором варианте выполнения устройство дополнительно снабжено насадками, установленными перед соплами, запасом охлаждающей воды, в емкости которого находятся эжектирующие и входные каналы насадок. Благодаря использованию нескольких гидравлически ...

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 ...

ПРЕОБРАЗОВАТЕЛЬ НЕЙТРОННОГО ПОТОКА ПРЯМОДЕЙСТВУЮЩЕЙ АВАРИЙНОЙ ЗАЩИТЫ ЯДЕРНОГО РЕАКТОРА

Сущность изобретения: в преобразователе увеличение давления формируется в аварийный сигнал не только при превышении мощности выше заданного уровня, но и при увеличении скорости нарастания нейтронного потока выше установленного предела. Устройство содержит ампулу, заполненную термочувствительным к нейтронам газом, например,3He и размещенную в активной зоне реактора. Ампула соединена капилляром с объемом вне активной зоны, имеющим упругую мембрану, снабженную пустотелой иглой. Объем размещен в дополнительном замкнутом объеме и сообщается с последним через дросселирующее отверстие, образуя диффренцирующее звено сигнала в форме импульса, поступающего из ампулы. Аварийный сигнал в дифференцирующем звене преобразуется в перемещение упругой мембраны с иглой, которое приводит к разрушению плоской мембраны, удерживающей поглотитель нейтронов, например,3He или10BF3 в аккумулируюшщем ресивере. Инжекция поглотителя в полость, размещенную по высоте внутриреакторного пространства, происходит за счет ...

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 содержит тягу 1, на одном конце которой расположен привод 2, а на другом - механизм сцепления 3 с рабочим органом. Привод 2 содержит корпус 4, на внешней поверхности которого закреплен электромагнит 5, а на внутренней - выполнен кольцевой выступ 6. На тяге 1 расположена гайка 7 с пружиной 8. Пружина 8 упирается в выступ 6. Ограничение движения тяги 1 с гайкой 7 от действия пружины 8 осуществляется с помощью разрезной втулки 9, лепестки которой установлены в прорезях 11 и подпружинены.Техническим результатом полезной модели является снижение массогабаритных характеристик привода и быстрота срабатывания устройства при подаче питания на привод. 1 ил.

ПЛАТФОРМА ДЛЯ ТРАНСПОРТИРОВКИ БЛОКА ЗАЩИТНЫХ ТРУБ

Платформа для транспортировки блока защитных труб, содержащая четыре плиты, соединенные между собой шестью стойками, два ползуна, отличающаяся тем, что на второй плите снизу установлены две червячные пары, каждая из которых состоит из червячного колеса, закрепленного на оси ползуна, и червяка, соединенного с приводом червячной пары, причем привод червячной пары установлен на верхней плите платформы.

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 ...

ЯДЕРНАЯ ЭНЕРГЕТИЧЕСКАЯ УСТАНОВКА

... 1. Ядерная энергетическая установка, включающая реактор с жидкометаллическим свинецсодержащим теплоносителем, с размещенными под свободным уровнем активной зоной, состоящей из тепловыделяющих сборок, парогенераторами, насосами и системой защитного газа, узел, содержащий твердофазное средство окисления теплоносителя, растворимое в нем, отличающаяся тем, что узел, содержащий твердофазное средство окисления теплоносителя, выполнен в виде втулки из твердофазного средства, установленной на входе теплоносителя в тепловыделяющую сборку.2. Ядерная энергетическая установка по п.1, отличающаяся тем, что в качестве твердофазного средства окисления теплоносителя использован диоксид свинца.

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 КОНДЕНСАЦИИ АВАРИЙНОГО ПАРА И ОЧИСТКИ ПАРОВОЗДУШНОЙ СМЕСИ ОТ РАДИОАКТИВНЫХ ВЕЩЕСТВ И УСТРОЙСТВО ДЛЯ ЕГО ОСУЩЕСТВЛЕНИЯ

Изобретения относятся к ядерной энергетике и могут быть использованы в энергетической и химической промышленности для конденсации пара и очистки паровоздушной смеси от радиоактивных и токсичных веществ. Предложен способ, заключающийся в формировании потока паровоздушной смеси, подаче в него охлаждающей жидкости, генерировании в части объема охлаждающей жидкости, выделенного вертикальной цилиндрической обечайкой, которая проходит через уровень зеркала охлаждающей жидкости, вертикальной вращающейся воронки, последующем пропускании через вертикальный слой жидкости воронки потока паровоздушной смеси и охлаждающей жидкости, стравливании очищенной паровоздушной смеси в атмосферу и отводе конденсата с охлаждающей жидкостью в бак рециркуляции, расположенный над уровнем зеркала охлаждающей жидкости, обеспечении их оттока в полость вертикальной цилиндрической обечайки и регулировании величины расхода потока паровоздушной смеси перед подачей в него охлаждающей жидкости в зависимости от величины давления ...

СИСТЕМА АВАРИЙНОГО РАСХОЛАЖИВАНИЯ ЯДЕРНОГО РЕАКТОРА

Сущность: система аварийного расхолаживания реактора 3 состоит из нескольких параллельных работающих каналов. Каждый канал включает в себя аварийный теплообменник 4, трубопроводами 5, 6 соединенный с расположенным в корпусе 8 радиатором. К корпусу 8 примыкает воздуховоды 9, 10 с заслонками. Заслонки образованы парными решетками. Трубопровод 5 на вертикальном участке выполнен в виде колонны 11, на которую опирается радиатор. Температурное перемещение радиатора на колонне 11 обеспечивает перемещение решеток. 2 ил.

УСТРОЙСТВО ДЛЯ СБРОСА ДАВЛЕНИЯ В АТОМНОЙ ЭЛЕКТРОСТАНЦИИ С ПРЕДОХРАНИТЕЛЬНОЙ ОБОЛОЧКОЙ

Устройство для сброса давления содержит фильтр, подсоединенный непосредственно к выпускному отверстию предохранительной оболочки. Между фильтром и дымовой трубой расположено устройство дросселирования с предохранительной мембраной. Устройство для сброса давления может дополнительно содержать скруббер, размещенный в одном резервуаре вместе с фильтром. При такой компоновке фильтр эксплуатируется с регулируемым в зависимости от давления в предохранительной оболочке плавно изменяющимся давлением. Технический результат заключается в ускорении процесса сброса давления за счет повышения массового расхода. 10 з.п. ф-лы, 9 ил.

ЯДЕРНАЯ РЕАКТОРНАЯ УСТАНОВКА С УСТРОЙСТВОМ ДЛЯ КОНТРОЛЯ ВЫВОДИМОГО В ТРУБУ ВОЗДУХА

Cущность изобретения: ядерная реакторная установка с устройством для контроля выводимого в трубу воздуха содержит защитную оболочку с ядерным реактором, трубопровод для сброса давления, соединяющий защитную оболочку с вентиляционной трубой, фильтр, установленный на трубопроводе, и устройство для контроля выводимого в трубу воздуха. Установка дополнительно содержит заборный трубопровод с пробоотборником, присоединенный к трубопроводу сброса давления после фильтра, и разжижающую установку, установленную на заборном трубопроводе. Устройство для контроля выводимого в трубу воздуха выполнено в виде измерительной магистрали с измерительным участком, состоящим из фильтров и мониторов. Измерительная магистраль присоединена к заборному трубопроводу после разжижающей установки, а измерительный участок соединен с трубопроводом разгрузки давления при помощи возвратной магистрали. Заборный трубопровод перед разжижающей установкой снабжен нагревателем. Разжижение радиоактивной смеси позволяет использовать ...

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 ...

УСТРОЙСТВО ОГРАНИЧЕНИЯ ТЕЧИ ИЗ ГЛАВНОГО ЦИРКУЛЯЦИОННОГО КОНТУРА ЯДЕРНОЙ ЭНЕРГЕТИЧЕСКОЙ УСТАНОВКИ ВОДОВОДЯНОГО ТИПА

... 1. Устройство ограничения течи из главного циркуляционного контура ядерной энергетической установки водоводяного типа, содержащее патрубок, соединенный с технологическим трубопроводом и состоящий из установленной у патрубка суживающей шайбы, образованной конфузором, суживающим отверстием, диффузором, отличающееся тем, что к суживающей шайбе в сторону технологического трубопровода последовательно установлен по крайней мере один гаситель потока, состоящий из поворотного профиля, сужающего отверстия и диффузора, при этом поворотный профиль состыкован с диффузором суживающей шайбы, а диффузор с поворотным профилем последующего гасителя потока, при этом касательной и образующей поворотного профиля у сужающего отверстия образован острый угол в сторону технологического трубопровода с его осью. 2. Устройство по п.1, отличающееся тем, что суживающие отверстия в суживающей шайбе и гасителях потока выполнены одного диаметра. 3. Устройство по пп.1 и 2, отличающееся тем, что диаметры суживающих отверстий ...

УСТРОЙСТВО ДЛЯ ПАССИВНОЙ ИНЕРТИЗАЦИИ ГАЗОВОЙ СМЕСИ В ЗАЩИТНОМ РЕЗЕРВУАРЕ АТОМНОЙ ЭЛЕКТРОСТАНЦИИ

Предлагается устройство для пассивной инертизации газовой смеси, появляющейся в защитном резервуаре атомной электростанции при аварийной ситуации, основанное на использовании химических веществ, которые при достижении определенной реакционной температуры реагируют или разлагаются с выделением инертизирующего газа или газовой смеси. Устройство особенно пригодно для использования в сочетании с каталитическими рекомбинаторами для удаления водорода за счет окисления с помощью имеющегося кислорода. Выделяющееся при этой протекающей экзотермической рекомбинации тепло может использоваться для подогрева таких химических веществ до необходимой температуры, реакционная температура которых выше температуры (примерно 100°С), которая устанавливается в защитном резервуаре при аварийной ситуации.

СПОСОБ УЛАВЛИВАНИЯ ГАЗО-АЭРОЗОЛЬНОЙ УТЕЧКИ

Способ улавливания газо-аэрозольной утечки относится к области ядерной техники. Цель изобретения - защита персонала и прилегающей территории от аварийной утечки радиоактивных и токсичных веществ при авариях трубопроводов и других устройств, из которых эти утечки возможны. Сущность способа улавливания газо-аэрозольной утечки, включающего наличие сборника-поглотителя, содержащего жидкость, способную связывать вредные продукты утечки, заключается в том, что радиоактивные и токсичные вещества улавливаются разрежением в месте образования утечки и при этом место утечки изолируется от окружающей среды газонепроницаемым материалом, например полиэтиленовой пленкой, а улавливаемые продукты утечки подаются в связующую их жидкость через барбатирующее устройство закрытого резервуара.

СПОСОБ ОСЛАБЛЕНИЯ КОРРОЗИОННОГО РАСТРЕСКИВАНИЯ МЕТАЛЛА

Способ ослабления коррозионного растрескивания металла, заключающийся в образовании на поверхности металла защитного покрытия, выполненного из электроизоляционного материала, отличающийся тем, что покрытие образуют в виде пленки керамической структуры феррита лития в процессе консервации посредством "мокрой" консервации с коррекцией водородного показателя консервирующего раствора до уровня 10,0<рН<10,5 и добавкой гидразин-гидрата и гидрооксида лития, а в процессе эксплуатации поддерживают сплошность этой пленки посредством микродозировок гидрооксида лития при 8,0<рН<9,6 и концентрации ...

ЯДЕРНАЯ ЭНЕРГЕТИЧЕСКАЯ УСТАНОВКА

... 1. Ядерная энергетическая установка, содержащая реактор с жидкометаллическим свинцовым теплоносителем или его сплавами, с размещенными под свободным уровнем активной зоной, парогенераторами, средствами циркуляции и систему защитного газа, отличающаяся тем, что по периметру активной зоны установлены трубки Фильда, центральные каналы которых через запорно-регулирующий клапан сообщены с водяным объемом расположенной выше реактора емкости, к верхней части которой подключены охлаждаемые воздухом теплообменные трубы, объединенные верхним коллектором с предохранительным клапаном, а верхний коллектор сообщен с внешними кольцевыми каналами трубок Фильда. 2. Ядерная энергетическая установка по п.1, отличающаяся тем, что над активной зоной установлены тяговые трубы с более высокими трубами в центральной зоне. 3. Ядерная энергетическая установка по п.1, отличающаяся тем, что золотник запорно-регулирующего клапана соединен с механизмом его перемещения, корпус которого сообщен с полостью емкости, с возможностью ...

ЯДЕРНЫЙ РЕАКТОР НА БЫСТРЫХ НЕЙТРОНАХ С ТЯЖЕЛЫМ ЖИДКОМЕТАЛЛИЧЕСКИМ ТЕПЛОНОСИТЕЛЕМ

Изобретение относится к ядерному реактору на быстрых нейтронах с тяжелым жидкометаллическим теплоносителем. Реактор содержит активную зону, расположенную в полости центральной части корпуса ядерного реактора, и размещенные в полости периферийной части корпуса по меньшей мере один главный циркуляционный насос, один парогенератор и одна выгородка. Выгородка выполнена из двух концентричных обечаек, образующих между собой кольцевой зазор. В нижней части кольцевого зазора установлена перегородка, снабженная отверстиями для прохода тяжелого жидкометаллического теплоносителя. Парогенератор размещен в полости, образованной выгородкой. Техническим результатом является уменьшение пусковой мощности главного циркуляционного насоса за счет формирования свободного уровня теплоносителя на всасе главного циркуляционного насоса, уменьшение колебательного процесса при останове главного циркуляционного насоса, создание термического сопротивления (теплового барьера) между "горячим" теплоносителем, проходящим ...

СИСТЕМА ПАССИВНОГО ОТВОДА ТЕПЛА ЯДЕРНОГО РЕАКТОРА

Изобретение относится к теплообменной технике и может быть использовано в качестве системы аварийного расхолаживания водоводяных ядерных реакторов. В системе пассивного отвода тепла ядерного реактора с водой под давлением, содержащей тракт теплоотвода первого контура с парогенератором и вспомогательный тракт аварийного расхолаживания с теплообменником, охлаждаемым водой атмосферного бака, и устройством регулирования расхода конденсата (мощности теплоотвода), теплообменник погружен в воду атмосферного бака, расположен выше парогенератора и подсоединен трубопроводами так, что выход пара парогенератора соединен с входом теплообменника, а выход теплообменника - с входом питательной воды парогенератора. Саморегулирование эффективности теплоотвода в указанной системе достигается тем, что на трубопроводе отвода конденсата установлен гидравлический дроссель, величина сопротивления которого при известной разности высот расположения теплообменника и парогенератора обеспечивает расход конденсата, ...

Ядерна энергетическа установка

Ядерная энергетическая установка, содержащая реактор с жидкометаллическим свинцовым теплоносителем или его сплавами, с размещенными под свободным уровнем активной зоной, парогенераторами, средствами циркуляции и системой защитного газа, включающей фильтр очистки газа, газовый компрессор, отличающаяся тем, что установка снабжена устройством ввода газовой смеси, выполненным в виде перфорированного кольцевого коллектора, расположенным под свободным уровнем в тракте теплоносителя на входе в главный циркуляционный насос, линия всаса устройства соединена с газовой полостью реактора, с газовым баллоном с восстановительной смесью и с линией напора газового компрессора, причем кольцевой коллектор снабжен круглыми отверстиями, расположенными на поверхности, обращенной к его центру, каждое из которых имеет в верхней части отверстие меньшего диаметра, центр которого смещен по их общей вертикальной оси выше центра круглого отверстия.

ЯДЕРНАЯ ПАРОПРОИЗВОДИТЕЛЬНАЯ УСТАНОВКА

... 1. Ядерная паропроизводительная установка, содержащая реактор, парогенератор, главный циркуляционный насос, циркуляционный трубопровод, емкость с раствором борной кислоты для охлаждения реактора, соединительный трубопровод и отсечную арматуру, отличающаяся тем, что емкость, содержащая раствор борной кислоты, снабжена патрубками, расположенными по высоте, причем каждый патрубок, кроме верхнего, снабжен клапаном. 2. Ядерная паропроизводительная установка по п.1, отличающаяся тем, что клапан включает в себя затвор, противовес, пустотелый шар, соединенный тягой с противовесом.

СПОСОБ АВАРИЙНОЙ ОСТАНОВКИ И ОХЛАЖДЕНИЯ ЯДЕРНОГО РЕАКТОРА

Изобретение позволяет повысить ядерную безопасность за счет повышения надежности системы аварийной остановки и охлаждения ядерных реакторов, упростить техническое обслуживание этой системы, а также уменьшить опасность вредного воздействия системы на персонал. Согласно способу предусматривается избыточное давление в гидроемкости, необходимое для вытеснения рабочей среды в реактор, создавать при возникновении аварийной ситуации, требующей аварийной остановки или охлаждения реактора. Кроме того, в гидроемкости создается избыточное давление различной для каждой конкретной аварийной ситуации величины. Избыточное давление в гидроемкости создается генерированием газа. Изобретение реализуется в системе, состоящей из гидроемкости 1 с рабочей средой, устройства для создания в гидроемкости избыточного давления 2, трубопровода 3, соединяющего гидроемкость с ядерным реактором, и обратного клапана 4.

ЯДЕРНАЯ ЭНЕРГЕТИЧЕСКАЯ УСТАНОВКА КОРПУСНОГО ТИПА

Изобретение относится к ядерной технике и может быть использовано в ядерных энергетических установках корпусного типа. Цель - повышение безопасности ядерных энергетических установок. Ядерная энергетическая установка корпусного типа вертикального исполнения включает корпус с активной зоной, днищем и патрубками, помещенные в бокс, в котором под днищем расположена подвижная поворотная платформа смотровой машины с укрепленной на ней поворотной защитой, обращенной к днищу вогнутой поверхностью, на которой установлена монолитная адиабатная керамическая чаша, у которой диаметр больше диаметра корпуса, а обращенная к днищу внутренняя полость заполнена теплоотводящим материалом, имеющим температуру плавления и кипения меньше температуры плавления днища корпуса и контактирующим свободной поверхностью с днищем корпуса при расплавлении активной зоны в результате перемещения чаши в вертикальном направлении подъемным устройством, установленным на поворотной платформе.

Система пассивной безопасности атомной электростанции

Сущность изобретения: для повышения безопасности АЭС с двойной вентилируемой защитной облолочкой - внутренней 4 и наружной 3 - последняя снабжена вентиляционной системой 5, газодув- ный агрегат 7 которой подключен к турбине 18 дополнительного контура 14 с легкокипящим теплоносителем . В случае аварии с разгерметизацией первого контура и потерей источников электроснабжения с помощью теплообменника 20 выделяющееся под оболочкой 4 тепло передают в парогенератор 17. Конденсатор 13 размещен выше парогенератора в вытяжной шахте 11, за счет чего в контуре 14 обео1ечивается естественная циркуляция теплоносителя Работа газодувного агрегата 7 поддерживает в пространстве между оболочками 3 и 4 некоторое разрежение. Воздух из этого ...

УСТРОЙСТВО ОГРАНИЧЕНИЯ РАСХОДА ТЕПЛОНОСИТЕЛЯ ПРИ АВАРИЙНОЙ РАЗГЕРМЕТИЗАЦИИ КОНТУРА ЯДЕРНОГО РЕАКТОРА

Устройство ограничения расхода теплоносителя при аварийной разгерметизации контура ядерного реактора, содержащее сопло с входным сужающимся, выходным расширяющимся участками и расположенной между ними горловиной постоянного поперечного сечения, отличающееся тем, что, с целью повышения эффективности путем интенсификации процесса парообразования теплоносителя, на внутренней поверхности входного участка выполнены канавки, направленные от входа в сопло к горловине, причем канавки имеют спиральную форму.

СИСТЕМА ОГРАНИЧЕНИЯ ПОСЛЕДСТВИЙ АВАРИИ С ПОТЕРЕЙ ТЕПЛОНОСИТЕЛЯ НА АТОМНЫХ ЭЛЕКТРИЧЕСКИХ СТАНЦИЯХ

... 1. Система ограничения последствий аварии с потерей теплоносителя на атомных электрических станциях, содержащая реактор, по меньшей мере два герметичных помещения, каждое из которых соединено через барботажные трубы с водяным объемом бассейна-барботера и через обратные клапаны, открывающиеся со стороны бассейна-барботера, с воздушным объемом бассейна-барботера, и контур охлаждения, включающий теплообменник, насос, всасывающие и напорные участки охлаждения, воды в бассейне-барботере, отличающаяся тем, что, с целью повышения надежности при аварии с потерей теплоносителя внутри корпуса реактора путем создания давления в зоне локализации меньшего, чем в окружающем пространстве, внутри бассейна-барботера выделена герметичная камера, образующая дополнительный бассейн-барботер, водяной объем которой соединен барботажными трубами с корпусом реактора, а воздушный объем - через обратный клапан и трубу с водяным объемом бассейна-барботера, причем контур охлаждения всасывающим и напорным участками ...

Система ограничения последствий аварии на атомных электростанциях

... 1. СИСТЕМА ОГРАНИЧЕНИЯ ПОСЛЕДСТВИЙ АВАРИЙ НА АТОМНЫХ ЭЛЕКТРОСТАНЦИЯХ по авт. св. СССР № 537389 (авт. св. ЧССР № 177368), отличающаяся тем, что, с целью повышения эффективности локализации аварии, в системе установлены водоуловители и пассивные спринклерные распылители, которые соединены с водоуловителями и барботажйыми конденсаторами. 2.Система по п. 1, отличающаяся тем, что водоуловители и барботажные конденсаторы соединены с активной спринклерной установкой. 3.Система по п. 1, отличающаяся тем, что пассивные спринклерные распылители направлены под углом вверх в сторону барботажных конденсаторов. 5 § (Л 5 1 5 5 СЛ N сх со ...

DEVICE FOR CLEANING GAS OF ADMIXTURES

Преобразователь нейтронного потока прямодействующей аварийной защиты ядерного реактора

Сущность изобретения: в преобразователе увеличение давления формируется в аварийный сигнал не только при превышении мощности выше заданного уровня, но и при увеличении скорости нарастания нейтронного потока выше установленного предела Устройство содержит ампулу, заполненную термочувствительным к нейтронам газом, например, Не и размещенную в активной зоне реактора Ампула соединена капилляром с объемом вне активной зоны, имеющим упругую мембрану, снабженную пустотелой иглой. Объем размещен в дополнительном замкнутом объеме и сообщается с последним через дросселирующее отверстие, образуя диффренцирующее звено сигнала в форме импульса, поступающего из ампулы. Аварийный сигнал в дифференцирующем звене преобразуется в перемещение упругой мембраны с иглой, которое приводит к разрушению плоской мембраны, удерживающей поглотитепь нейтронов, например, Не или BF в аккумулируюшщем ресивере. Инжек- ция поглотителя в полость, размещенную по высоте внутриреакторного пространства, происходит за счет внутреннего ...

Каландр для жидкого замедлителя ядерного реактора

Устройство для очистки газа от примесей

Использование: газоочистка. Сущность изобретения: газоочистное оборудование включает в себя резервуар (1), частично заполненный жидкостью, и газораспределительное средство (6), находящееся ниже Ј поверхности жидкости. Газораспределительное средство (6) состоит из расположенных с наклоном книзу распределительных участков (10) с обращенными вверх впускными отверстиями для приема очищаемого газа и нижним отверстием (12). Каждое из впускных отверстий переходит в сопло (13) типа трубки Вентури, имеющее выпускное отверстие и всасывающие каналы, направленные в жидкость. При работе такого аппарата постоянно поддерживается номинальное рабочее давление, при этом газораспределение осуществляется автоматически с помощью набора рабочих сопловых насадочных средств, впуск газа в которые обеспечивается за счет соответствующего изменения уровня (поднимания и опускания) внутриканальной поверхности жидкости. 3 з.п. ф-лы, 8 ил. (Л С 13 со ю ю 0 ел CJ со фцг.1 ...

Ядерная установка с реактором, охлаждаемым водой под давлением

RADIALE BERSTSICHERUNG FUER VON EINEM BEWEHRTEN BETONMANTEL UMGEBENE DRUCKBEHAELTER, INSBESONDERE FUER REAKTORDRUCKBEHAELTER

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

GEBÄUDE MIT INNENRAUMSCHUTZ

VERFAHREN UND GERÄT ZUM REKOMBINIEREN UND/ODER ZÜNDEN VON WASSERSTOFF, ENTHALTEN IN EINEM H 2-LUFT-DAMPF-GEMISCH, INSBESONDERE FÜR KERNKRAFTWERKE

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.

Nuclear power plant protection - against extraneous actions by conical tent of wire ropes and netting

Protection of nuclear power stations against extraneous disturbances, such as a gas cloud explosion, aircraft crash or sabotage is provided by a deformable safety catch structure which straddles the entire site. This structure has the shape of a multi-cone tent with a network of ropes which is at least partly designed airtight and is calculated to withstand internal and external stresses. This is a simple way of protecting nuclear power plant against external disturbances and of meeting the usually raised demands during the application for approval of a site.

Sicherheitsvorrichtung fuer Atomkernreaktoren

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 ...

SHIELDS FOR NUCLEAR REACTIONS

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 ...

Collecting device for melting fuel rods n a nuclear reactor installation

In order to safely collect the reactor core 3 in the case of a reactor melt-down, wherein the reactor core escapes through the base 5 of the reactor shell 2, there is provided a chute 6 underneath the reactor shell 2. It deflects the dropping parts of the melting reactor sideways and discharges them into a deposition trough 14. The chute 6 and the trough 14 are rotated relative to each other, so that the dropping parts are spread over the deposition trough 14. The chute 6 is made of a thick-walled corrosion-resistant material 10 which is provided with a surface coating 11 of refractorily vitrified porcelain enamel. In order to safely receive the melting fuel rods 4 arriving through the chute 6, the deposition trough 14 comprises a homogeneous mixture of crushed glass, cement, steel cubes (as impact absorbing elements) and barium sulphate (as filler substance). An exhaust tube 25 serves for controlled removal of gases from the deposition region. ...

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 ...

NUCLEAR REACTORS

... 1338898 Fuelling reactors NUCLEAR POWER GROUP Ltd 3 Dec 1970 [14 Jan 1970] 1785/70 Heading G6C A fuelling machine is provided with wedging devices which may be operated to prevent movement of the machine by seismic shock. As shown, the machine flask 1 is supported on a carriage having wheels 9 running on crossmembers 2, 3 forming a gantry, the gantry having wheels 8 running on rails supported by walls 4, 5. Wedging devices 10, 11 prevent movement between the carriage and the gantry and wedging devices 12, 13 prevent movement between the gantry and the walls. Each wedging device comprises a plurality of wedges C, Fig. 3, operated by hydraulic actuators 15, the operation of each actuator causing a movable wedge member A to clamp against a surface D and thereby prevent relative movement between surface D and surface E.

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.

NUCLEAR REACTOR CONTAINMENT SPRAY TESTING SYSTEM

... 1524002 Testing spray system WESTING- HOUSE ELECTRIC CORP 14 April 1977 [22 April 1976] 15511/77 Heading G6C A containment spray system 24, 26 in the containment structure 10 of a nuclear reactor is tested by introducing into the system gas at a temperature different from that of the atmosphere within the containment structure, and scanning the nozzles 26 of the system with an infrared viewing device.

NUCLEAR REACTORS

NUCLEAR REACTOR INSTALLATION INCLUDING A CONTROL VALVE

... 1513775 Fluid-pressure servomotor systems KRAFTWERK UNION AG 18 June 1975 [26 June 1974 26 Sept 1974] 26045/75 Heading G3P [Also in Divisions F2 and G6] A safety valve 8, Figs. 1 and 2, in the live steam line 2 of a nuclear reactor installation, comprises a normally open valve-member 22 controlled by a mechanism sensitive to pressure to rapidly close or almost close the valve when the pressure falls below its normal operating value and to re-open the valve by at most half its fully-open throughflow cross-section if the pressure subsequently rises above a predetermined value in excess of the normal operating value. The valve 8 is located in a safety envelope 3 of the installation together with steam generator 1 and safety valve 13. Outside the envelope the line 2 is associated with a branch 16 having a further safety valve 17 and blow-off regulating valves 20. The valve, Fig. 2, is normally forced fully open by the steam pressure in line 2. However, should the steam-pressure fall, e.g. due ...

VESSEL RESTRAINT DEVICE

... 1491347 Reactor pressure vessels COMBUSTION ENG Inc 27 Feb 1975 [1 March 1974] 8241/75 Heading G6C A pressure vessel has a main cylindrical portion 12, a hemispherical portion 14 integral with and closing an end of portion 12, a removable hemispherical portion 16 secured by bolts to portion 12, and restraining means comprises collars 34, 36, a set of parallel cables 38 extending between the collars, a set of parallel cables 52 extending across the outer surface of portion 14 and each being secured at both ends to collar 36, and a set of parallel cables 44 extending across the outer surface of portion 16 and each being secured at both ends to collar 34. Tensioners may be provided for all the cables. Further cables 60 may be added.

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.

KONDESATIONSROHR FUER BLOW-OFF MECHANISMS, WHICH FOR POSITIVE PRESSURE DELIMITATION WITH NUCLEAR POWER STATIONS, IN PARTICULAR WITH BOILING WATER NUCLEAR POWER STATIONS, SERVING

DEVICE AND PROCEDURE FOR THE RECOMBINATION OF HYDROGEN AND OXYGEN IN A GAS MIXTURE

CATALYST FOR THE RECOMBINATION OF HYDROGEN WITH OXYGEN

ABBLASEEINRICHTUNG ZUR UEBERDRUCKBEGRENZUNG BEI KERNKRAFTWERKEN, INSBESONDERE BEI SIEDEWASSER- KERNKRAFTWERKEN

KERNKRAFTWERK

DAEMPFUNGSVORRICHTUNG ZUR AUFNAHME VON STOSS- KRAEFTEN, INSBESONDERE AN DEN KOMPONENTEN VON KERNREAKTORANLAGEN

KONDESATIONSROHR FUER BLOW-OFF MECHANISMS, WHICH FOR POSITIVE PRESSURE DELIMITATION WITH NUCLEAR POWER STATIONS, IN PARTICULAR WITH BOILING WATER NUCLEAR POWER STATIONS, SERVING

LIFTING DEVICE

BLOW-OFF MECHANISM FOR POSITIVE PRESSURE DELIMITATION WITH NUCLEAR POWER STATIONS, IN PARTICULAR WITH BOILING WATER POWER STATIONS

SAFETY DEVICE FUR UNDER PRESSURE STANDING PLANTS

HEATDAMMING SEALING SYSTEM FOR PRINTING RELIEF PORTS OF PARTITIONS, IN PARTICULAR WITH NUCLEAR REACTOR BUILDINGS.

KONDENSATIONSROHR FUER ABBLASEEINRICHTUNGEN, WELCHE ZUR UEBERDRUCKBEGRENZUNG BEI KERNKRAFT- WERKEN, INSBESONDERE BEI SIEDEWASSER-KERNKRAFT- WERKEN, DIENEN

PROCEDURE FOR THE CLEANING FROM GASES AND DEVICE TO IT.

PROTECTIVE SYSTEM FOR THE REACTOR SAFETY BUILDING IN A NUCLEAR POWER STATION PLANT.

ABBLASEEINRICHTUNG ZUR UEBERDRUCKBEGRENZUNG BEI KERNKRAFTWERKEN, INSBESONDERE BEI SIEDEWASSERKRAFTWERKEN

PRODUCTION DE CHALEUR AU MOYEN D'UN REACTEUR NUCLEAIRE

PRESSURE RELIEF AND SECONDARY CONTAINMENT SYSTEM

SEGMENTAL ENCASEMENT FOR A NUCLEAR REACTOR

CONTAINMENT VESSEL FOR REACTOR TANK

SHOCK ABSORBING STRUCTURE FOR NUCLEAR FUEL ASSEMBLIES

CASE 3766 INDUSTRIAL TECHNIQUE A typical embodiment of this invention provides a hydraulic mechanism for alleviating the effect of seismic forces and other stresses that are applied to a fuel assembly in a nuclear reactor. Illustratively, hollow guide posts protrude into a fuel assembly end fitting grid from biased spring pads. Plungers that move with the spring pads plug one end of each of the respective guide posts. Plates on the end fitting grid that have individual holes for fluid discharge partially plug the other ends of the respective guide posts, thereby providing a hydraulic means for absorbing the longitudinal component of seismic shocks and other anticipated forces.

PRODUCTION DE CHALEUR AU MOYEN D'UN REACTEUR NUCLEAIRE

MEANS FOR COOLING A HEAT-GENERATING DEVICE

A heat- generating member is arranged in a water-filled pressure vessel which is provided with a pressure relief valve or the like. The water of the pressure vessel can be partly evaporated, thereby acting as a heat sink for the generated heat. The walls of an outer vessel surround at least a lower part of the pressure vessel in such a way that a closed, relatively small auxiliary space is formed between the two vessels. The auxiliary space communicates via at least one tube with an open evaporation pool, which is arranged above the cover of the pressure vessel. A tube coil, disposed in an upper part of the pressure vessel, is connected by both ends to the evaporation pool. If a leak should occur in the lower part of the pressure vessel, water from the auxiliary space flows to the evaporation pool via the tube and from the evaporation pool into the tube coil, whereby boiling takes place in the evaporation pool due to the high-temperature steam condensing on the tube coil. The leakege, thus ...

NUCLEAR SAFEGUARD SYSTEM

INSULATING ENCASEMENT FOR A NUCLEAR REACTOR

METHOD AND APPARATUS FOR PROTECTING THE CORE OF A NUCLEAR REACTOR

NUCLEAR POWER PLANT WITH COLLECTOR VESSEL FOR MELTING CORE MASSES

SHOCK ABSORBING STRUCTURE FOR NUCLEAR FUEL ASSEMBLIES

HYBRID PRESSURE VESSEL WITH SEPARABLE JACKET

A pressure vessel is provided including an inner tank formed from a tank liner surrounded by a wound layer of composite filaments. A protective jacket is disposed on the inner tank that facilitates stacking and portability of the pressure vessel and helps to define an air passage for convective heat transfer between the hybrid tank and the environment.

CLOSURE SYSTEM

PROCESS AND DEVICE FOR RECOMBINING AND/OR IGNITING HYDROGEN CONTAINED IN AN H2-AIR-STEAM MIXTURE, PREFERABLY FOR NUCLEAR POWER STATIONS

A first partial flow (f11) of the H2-air-steam mixture (f1) is passed through a first channel (R) and recombined by contact with a first channel wall (r1) having a catalytic coating. In addition to the first partial flow (f11), a second partial flow (f12) of the said mixture is passed through a second channel (z) having a second channel wall (r2) and, in that case, is ignited by being fed past, preferably metallic, ignition elements (z, z1, z2, z3) heated to the H2 ignition temperature on reaching or exceeding the ignition limit. Under these circumstances, the heat liberated during the catalytic reaction in the first channel (R) is, at least partially, transmitted to the second channel (z) for the purpose of preheating it. The novel process functions both at the lower and at the upper ignition limit and recombines even at unignitable H2 concentrations. - A device with which the process can be carried out is also described. FIG 1 ...

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

Verfahren zur Herstellung von N-(5'-Nitro-2'-furfuryliden)-1-amino-hydantoin

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.

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 ...

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 ...

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 ...

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 ...

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 ...

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 ...

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. ...

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 ...

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 ...

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 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 ...

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 ...

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 ...

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 ...

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 ...

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 ...

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 ...

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.

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 ...

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 ...

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.

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 ...

Recirculating drainage channel for the safety circuits of a nuclear reactor

In a nuclear reactor, a drainage channel for the safety injection and spray circuits of the reactor is arranged in the bottom of the annular gap between two substantially vertical walls surrounding the pressure vessel of the reactor, the drainage channel comprising filter panels disposed vertically along each side of a central horizontal solid lower panel at the bottom of the annular gap and between two lateral horizontal solid upper panels so as to bound a central channel having vertical lateral filter walls communicating with two lateral chambers which communicate by substantially horizontal passages with a central collecting chamber disposed below the central lower panel and having an outlet for connection to the main intake of the safety circuit.

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.

원자로용 에너지 격납 구조

복수의 조립체를 포함하는 에너지 흡수 장치가 설명되며, 각각의 조립체는 바람직하게는 복수의 원통형 튜브를 포함하고, 각각의 튜브는 암모늄 카바메이트와 같은 흡열 물질을 수용한다. 조립체는 원자로 격납 구조의 주변을 둘러쌀 수 있는 저장소에 위치 설정된 복수의 세장형 바스켓에서 지지된다. 에너지 흡수 장치는 설계 기준 사고 이벤트에 방출되는 초과 에너지를 흡수한다.

Radioactive material reduction facility and nuclear power plant having the same

The present invention provides a radioactive material reduction facility capable of reducing an exclusion area boundary and a nuclear power plant comprising the same. The radioactive material reduction facility includes: a cooling water storage part which is installed in a containment part and is formed to store cooling water; a boundary part which surrounds a reactor coolant system in order to prevent a radioactive material from being leaked into the inside of the containment part from the reactor coolant system installed in the containment part or a pipe connected to the reactor coolant system; a connection pipe which is connected to the cooling water storage part and the boundary part in order to induce the flow of the radioactive material formed within the boundary part to the cooling water storage part; and an injection part which is connected to the connection pipe to receive the radioactive material from the connection pipe, and is dipped into the cooling water to inject the radioactive material to the cooling water of the cooling water storage part. At least a part of the boundary part prevents occurrence of a coolant loss accident in an area between the boundary part and the containment part by fracture of a penetration pipe penetrating the containment part.

Passive safety system and nuclear reactor having the same

본 발명은 사고 발생시 격납건물 내부의 압력 및 방사선 농도 상승을 순차적으로 억제하는 피동안전설비 및 이를 구비하는 원전을 제안한다. 피동안전설비는, 격납건물 내부와 통하도록 형성되어 상기 격납건물과 압력 평형을 유지하고 사고 시 상기 격납건물 내부의 압력 및 방사선 농도 상승을 1차적으로 억제하도록 상기 격납건물의 압력 상승에 의해 유입된 증기 또는 대기를 수용하는 감압부, 및 상기 감압부와 상기 격납건물 내부의 압력 변화에 따라 피동적으로 작동하며 상기 격납건물 내부의 압력 및 방사선 농도 상승을 2차적으로 억제하도록 상기 감압부에 저장된 냉각수와 상기 격납건물에서 상기 감압부로 유입되어 응축된 응축수를 상기 격납건물 내부로 살수하는 피동격납건물살수부를 포함하고, 상기 피동격납건물살수부는, 상기 냉각수 및 상기 응축수를 상기 격납건물 내부로 살수하도록 상기 격납건물 내부의 기설정된 높이에 설치되는 살수배관, 및 상기 감압부의 압력이 상기 격납건물 내부의 압력보다 커지면, 압력차에 의해 상기 감압부로부터 상기 살수배관으로 냉각수 및 응축수의 유로를 제공하도록 일단이 상기 감압부의 내부에 삽입되고 타단이 상기 살수배관에 연결되는 유체공급배관을 포함한다. The present invention proposes a passive safety equipment for sequentially suppressing pressure and radiation concentration rise in a containment building when an accident occurs and a nuclear power plant equipped with the same. The passive safety equipment is formed so as to communicate with the inside of the containment building to maintain pressure balance with the containment building and to prevent the rise of pressure and radiation concentration inside the containment building And a control unit that operates passively according to a change in pressure inside the depressurizing unit and the containment building and controls the cooling water stored in the depressurizing unit to secondarily suppress pressure and radiation concentration rise inside the containment building, And a passive containment building water spraying unit for spraying the condensed water introduced into the decompression unit from the storage building into the inside of the containment building, wherein the passive containment building water spraying unit is configured to contain the cooling water and the condensed water in the containment building A water spray pipe installed at a predetermined height inside the building, A fluid supply pipe whose one end is inserted into the depressurizing portion and whose other end is connected to the water spray pipe to provide a flow path of cooling water and ...

Tightening apparatus for filter comprising special steel fiber bundle or fiber mat

원자로를 위한 가스 모니터링 시스템 및 방법

가스 모니터링 시스템 및 방법이 제공된다. 일 실시예에서, 가스 모니터링 시스템은 원자로 격납고 환경 내에 있는 가스 모니터링 유닛, 원자로 비격납고 환경에 있는 원자로 격납고 환경, 및 가스 모니터링 유닛을 가스 모니터링 유닛 컨트롤러에 상호 연결하는 고온 또는 산업 규격 케이블을 포함한다. 가스 모니터링 유닛 상의 다양한 센서는 수소 가스 농도를 포함하는 원자로 격납고 환경의 환경을 검출한다.

System for limiting the consequences of a nuclear power plant accident

一种化学式点火装置

本发明提供一种化学式点火装置，所述化学式点火装置包括由金属丝网制成的容纳部,在所述容纳部的上部和下部分别设置密封件,在所述容纳部的外侧包裹用于防止水蒸气和空气透过的膜材料,在所述容纳部内部放置遇水蒸气易燃化学物质并且充入惰性气体。与现有技术相比，本发明具有可以根据外部环境的变化自发地进行点火消氢,无需外部能源输入的有益效果,在出现出现停电或备用电源失效时，仍然能够有效缓解氢气风险。

一种用于核反应堆直接安注情况下的半开式导流装置

本发明公开了一种用于核反应堆直接安注情况下的半开式导流装置，用于与直接安注管嘴的出口对应从而对安全注射水进行导流，所述半开式导流装置包括：连接部件以及导流部件；所述连接部件与核反应堆吊篮筒体外壁固定连接，用于固定并支撑所述半开式导流装置，所述导流部件与所述连接部件固定连接，用于引导所述安全注射水，所述导流部件与所述吊篮筒体形成流通通道，用于供安全注射水流动。本发明引导安全注射水沿吊篮外壁面向下流动，提高安全注射水注入堆芯的流量，便于核反应堆构件在役检查。

Radioactive material reduction facility and nuclear power plant having the same

본 발명은 방사성 물질 저감 설비에 관한 것으로, 격납부; 상기 격납부의 내부에 구비되어, 상기 격납부의 내부공간을 원자로냉각재계통을 수용하는 제1공간과 상기 제1공간과 격납부 사이에 형성되는 제2공간을 구획하고, 사고 시 상기 제1공간 내의 상기 원자로냉각재계통 또는 상기 원자로냉각재계통과 연결된 배관으로부터 방출되는 방사성 물질이 상기 제2공간으로 직접 방출되지 않도록 상기 원자로냉각재계통을 둘러싸는 경계부; 상기 제1공간과 상기 제2공간 사이에 설치되는 냉각수 저장부; 상기 경계부 내 제1공간에 형성되는 방사성 물질의 유동을 상기 냉각수 저장부로 유도하도록 상기 경계부와 상기 냉각수 저장부에 연결되는 연결배관; 및 상기 냉각수 저장부에 저장된 냉각수에 침지되고, 상기 연결배관을 통과한 상기 방사성 물질을 상기 냉각수에 분사하여 상기 냉각수에 상기 방사성 물질 중 수용성 방사성 물질을 포집되게 하고, 상기 냉각수 저장부의 바닥면으로부터 서로 다른 이격 거리를 가지며 상기 제1공간과 상기 제2공간의 압력 차에 대응하여 작동하는 복수개의 분사부를 포함한다. The present invention relates to a radioactive material abatement facility, A first space for accommodating the reactor coolant system and a second space formed between the first space and the compartment are formed in the inner space of the compartment, Wherein the reactor coolant system or the reactor coolant system in the reactor coolant system is not directly discharged into the second space; A cooling water storage unit installed between the first space and the second space; A connection pipe connected to the boundary portion and the cooling water storage portion to guide the flow of the radioactive material formed in the first space in the boundary portion to the cooling water storage portion; And radiating the radioactive material that has passed through the connection pipe to the cooling water so that the radioactive material contained in the radioactive material is captured by the cooling water, And a plurality of jetting portions having different spacing distances and operating in accordance with the pressure difference between the first space and the second space.

Method of restoring safety barriers at a radioactive wastes disposal station

FIELD: technological processes. SUBSTANCE: invention relates to technology of improving or hardening soil using thermal, electrical or electrochemical means. Method of restoring safety barriers in a radioactive wastes disposal station includes immersing the electrodes in the area of formation of cracks and cavities in the barrier material, creation of electric field between electrodes, supply of carrier fluid to the area adjacent to electrode, transfer of carrier fluid from one electrode to another. Electrodes are immersed at the boundaries of the area of formation of a crack or cavity in a barrier material, which ensures safe burial of solid radioactive wastes. Barrier material mixed with the carrier fluid is fed into the first perforated electrode and injected into the anode area. Difference of potentials between electrodes is created, barrier material is pushed to area of formation of crack or cavity. Carrier fluid passed between electrodes and cleaned from barrier material is pumped through second perforated electrode. Dry barrier material is fed through perforation area. EFFECT: invention makes it possible to remotely restore integrity of safety barriers at radioactive wastes disposal sites and to increase safety. 1 cl, 2 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 694 816 C1 (51) МПК G21F 9/00 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ (52) СПК G21C 9/00 (2019.05) (21)(22) Заявка: 2018143958, 11.12.2018 (24) Дата начала отсчета срока действия патента: Дата регистрации: 17.07.2019 (45) Опубликовано: 17.07.2019 Бюл. № 20 2 6 9 4 8 1 6 R U (56) Список документов, цитированных в отчете о поиске: RU 2602615 C2, 20.11.2016. RU 2508954 C1, 10.03.2014. RU 2550367 C1, 10.05.2015. RU 2625329 C1, 13.07.2017. US 5595644 A1, 21.01.1997. (54) СПОСОБ ВОССТАНОВЛЕНИЯ БАРЬЕРОВ БЕЗОПАСНОСТИ В ПУНКТЕ РАЗМЕЩЕНИЯ РАДИОАКТИВНЫХ ОТХОДОВ (57) Реферат: Изобретение относится к технологии отходов. Барьерный материал, смешанный с ...

Multi stage safety injection device and passive safety injection system having the same

본 발명은 원자로용기에 다단으로 냉각수를 주입하도록, 원자로용기의 압력 및 수위가 저하되는 사고 발생시 중력수두에 의해 상기 원자로용기로 주입될 냉각수를 수용하도록 형성되는 안전주입탱크, 상기 원자로용기와 상기 안전주입탱크의 압력평형을 형성하도록 상기 원자로용기와 상기 안전주입탱크에 연결되는 압력평형배관, 및 상기 원자로용기와 상기 안전주입탱크의 압력평형시 상기 원자로용기로 냉각수를 주입시키도록 상기 안전주입탱크와 상기 원자로용기에 연결되고 상기 안전주입탱크의 수위 저하에 따라 상기 원자로용기로 주입되는 냉각수의 유량을 단계적으로 감소시키도록 상기 안전주입탱크에 높이차를 형성하며 연결되는 복수의 안전주입배관을 포함하는 다단 안전주입 장치를 개시한다. The present invention relates to a safety injection tank for injecting cooling water into a reactor vessel in multi-stages, a safety injection tank formed to receive cooling water to be injected into the reactor vessel by gravity head in the event of an accident in which the reactor vessel pressure and water level are lowered, A pressure equalization pipe connected to the reactor vessel and the safety injection tank so as to form a pressure balance of the injection tank, and a safety injection pipe for injecting cooling water into the reactor vessel when pressure balance between the reactor vessel and the safety injection tank And a plurality of safety injection pipes connected to the reactor vessel and connected to each other to form a height difference in the safety injection tank so as to reduce the flow rate of the cooling water injected into the reactor vessel in accordance with a decrease in the level of the safety injection tank Discloses a multistage safety injection device.

Internal Condensation Vapor Exhaust System

순환로의 압력을 줄이기 위하여 상기 순환로로부터 탱크나 풀로 증기를 배출하는 내부 응축 시스템이다. An internal condensation system for discharging steam from the circuit to the tank or the pool to reduce the pressure of the circuit. 내부 응축 도관(20)은 냉각 액체 공급 도관(21)의 인접 상류에 격판(21)을 구비하고, 이 격판(21)의 오리피스를 통과하는 증기 분사는 이 분사주위에 현저한 음압 영역을 생성하여 격판의 하류에서 냉각 액체가 흡인되게 한다. 이렇게 하여 증기가 혼합실(25)에서 응축된다. The internal condensation conduit 20 has a diaphragm 21 upstream of the cooling liquid supply conduit 21, and the vapor injection through the orifice of the diaphragm 21 creates a significant negative pressure region around the diaphragm and thus the diaphragm. Allow the cooling liquid to be aspirated downstream of. In this way, steam condenses in the mixing chamber 25. 적용 가능한 분야는 핵 원자로의 주 순환로에 적용될 것이다. Applicable fields will be applied to the main reactor of nuclear reactors.

Passive containment cooling system and nuclear power plant having the same

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

Cfvs for nuclear reactor

The present invention relates to a containment filtered venting system used for a nuclear reactor. It includes a containment filtered venting vessel; an inlet pipe which is connected to a nuclear reactor building and the containment filtered venting vessel; a combine nozzle which is connected from the inlet pipe and is dipped in a filtered solution filled in the containment filtered venting vessel; a cyclone separator which guides mist and aerosol mixed with the filtered solution leaked from the compound nozzle to a metal filter; a metal filter which is connected to the upper end of the cyclone separator and filters foreign substance from the mist and aerosol; and moleculars which remove organic iodine from exhaust gas filtered by the metal filter.

Depressurizing method for high pressure sealed system

본 발명의 바람직한 일 실시예는 밀폐 시스템 내부에 고압의 증기가 발생할 경우, 미세한 입자를 이용하여 내부의 증기 압력을 일시적으로 또는 지속적으로 감소시켜 밀폐 시스템을 보호할 수 있는 고압 밀폐 시스템 감압방법을 제공하고자 하는 것이다. 본 발명의 바람직한 일 실시예는 밀폐 시스템 내의 압력(P)이 밀폐 시스템의 목표 미세입자 투입압력(P0)이상이 되면, 미세입자를 밀폐 시스템 내에 투입하고, 밀폐 시스템내의 압력(P)이 밀폐 시스템의 목표 미세입자 투입압력(P0) 미만이 되면, 미세입자 투입을 중단하는 고압 밀폐 시스템 감압방법을 제공한다. 본 발명의 바람직한 일 실시예에 따르면, 고압에 노출된 밀폐 시스템 내부를 감압시켜 사고를 예방할 수 있다. One preferred embodiment of the present invention provides a high-pressure sealed system decompression method that can protect the closed system by temporarily or continuously reducing the inside steam pressure using fine particles when high-pressure steam is generated inside the closed system. Is what you want. In a preferred embodiment of the present invention, when the pressure P in the closed system is greater than or equal to the target microparticle input pressure P0 of the closed system, the fine particles are introduced into the closed system, and the pressure P in the closed system is closed. It provides a high-pressure sealing system decompression method to stop the injection of the fine particles, when the target fine particle input pressure (P0) of less than. According to one preferred embodiment of the present invention, an accident may be prevented by depressurizing the inside of the closed system exposed to high pressure.

Convective dry containment filter vent system

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

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

Nuclear installation and method of pressure relief in nuclear installation

FIELD: nuclear technologies; environmental protection. SUBSTANCE: nuclear installation (1) contains containment (2) with pressure relief pipeline (6) connected, into which Venturi scrubber (12) located in tank (14) with washing liquid (W), and throttling device (24) are connected successively. Venturi scrubber (12) and throttling device (24) provide for their installation in Venturi scrubber (12) in case of critical expansion of the vapour/air mixture in pressure relief pipeline (6), at throttling device (24), at a vapour/air mixture flow velocity of over 150 m/s, predominantly over 200m/s. EFFECT: highly reliable elimination of possibility of emissions of even smallest radioactive substances to environment. 2 cl, 3 dwg ÐÎÑÑÈÉÑÊÀß ÔÅÄÅÐÀÖÈß RU (19) (11) 2 324 990 (13) C2 (51) ÌÏÊ G21C 9/004 (2006.01) B01D 45/00 (2006.01) ÔÅÄÅÐÀËÜÍÀß ÑËÓÆÁÀ ÏÎ ÈÍÒÅËËÅÊÒÓÀËÜÍÎÉ ÑÎÁÑÒÂÅÍÍÎÑÒÈ, ÏÀÒÅÍÒÀÌ È ÒÎÂÀÐÍÛÌ ÇÍÀÊÀÌ (12) ÎÏÈÑÀÍÈÅ ÈÇÎÁÐÅÒÅÍÈß Ê ÏÀÒÅÍÒÓ (21), (22) Çà âêà: 2006101984/06, 24.06.2004 (72) Àâòîð(û): ÝÊÊÀÐÄÒ Áåðíä (DE) (24) Äàòà íà÷àëà îòñ÷åòà ñðîêà äåéñòâè ïàòåíòà: 24.06.2004 (73) Ïàòåíòîîáëàäàòåëü(è): ÔÐÀÌÀÒÎÌÅ ÀÍÏ ÃÌÁÕ (DE) R U (30) Êîíâåíöèîííûé ïðèîðèòåò: 25.06.2003 DE 10328773.6 (43) Äàòà ïóáëèêàöèè çà âêè: 10.06.2006 (45) Îïóáëèêîâàíî: 20.05.2008 Áþë. ¹ 14 2 3 2 4 9 9 0 (56) Ñïèñîê äîêóìåíòîâ, öèòèðîâàííûõ â îò÷åòå î ïîèñêå: EP 0285845 A, 12.10.1988. SU 1768242 A, 15.10.2002. US 6280502 B1, 28.08.2001. JP 55067303 A, 21.05.1980. US 6513345 B1, 04.02.2003. (85) Äàòà ïåðåâîäà çà âêè PCT íà íàöèîíàëüíóþ ôàçó: 25.01.2006 2 3 2 4 9 9 0 R U (87) Ïóáëèêàöè PCT: WO 2004/114322 (29.12.2004) C 2 C 2 (86) Çà âêà PCT: EP 2004/006837 (24.06.2004) Àäðåñ äë ïåðåïèñêè: 129010, Ìîñêâà, óë. Á.Ñïàññêà , 25, ñòð.3, ÎÎÎ "Þðèäè÷åñêà ôèðìà Ãîðîäèññêèé è Ïàðòíåðû", ïàò.ïîâ. Ã.Á. Åãîðîâîé, ðåã.¹ 513 (54) ßÄÅÐÍÀß ÓÑÒÀÍÎÂÊÀ È ÑÏÎÑÎÁ ÑÁÐÎÑÀ ÄÀÂËÅÍÈß Â ßÄÅÐÍÎÉ ÓÑÒÀÍÎÂÊÅ (57) Ðåôåðàò: Èçîáðåòåíèå îòíîñèòñ ê äåðíîé óñòàíîâêå ñ çàùèòíîé îáîëî÷êîé, ê êîòîðîé ïðèñîåäèíåí òðóáîïðîâîä ñáðîñà äàâëåíè ...

Emergency and back-up cooling of nuclear fuel and reactors

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

Reactor containment vessel and boiling water reactor power plant

A containment vessel includes a primary containment vessel containing a reactor pressure vessel, an upper secondary containment vessel arranged above the primary containment vessel, and a gas-phase vent pipe linking the primary containment vessel and the upper secondary containment vessel by way of an isolation and connection switching system. The gas-phase vent pipe may be arranged either inside or outside the primary containment vessel and the upper secondary containment vessel. Alternatively, it may be embedded in the wall. An igniter may be arranged in the upper secondary containment vessel. The air in the upper secondary containment vessel may be replaced by nitrogen. A gravity-driven flooding system pool may be arranged in the upper secondary containment vessel and cooling water may be led from the inside of the pool to the inside of the primary containment vessel.

Inherent safety water cooled reactor system for thermal desalination

본 발명은 열담수화를 위한 고유안전 수냉각형 원자로 계통에 관한 것으로, 상세하게는 물을 냉각재 및 감속재로 사용하며, 핵분열을 통해 열에너지를 발생시키는 수냉각형 원자로; 상기 수냉각형 원자로에 연결되어 상기 수냉각형 원자로의 노심에서 발생된 열에너지를 격납용기 내의 열교환기를 통해 간접적으로 전달받아 담수를 생산하는 담수화 계통; 상기 수냉각형 원자로의 핵연료 재장전시 사용되는 재장전수탱크; 상기 수냉각형 원자로 및 재장전수탱크를 포함하는 전체 원자로 계통을 감싸는 강철 격납용기; 및 상기 담수화 계통에서 생산된 담수가 저장되는 담수저장탱크;를 포함하는 열담수화를 위한 고유안전 수냉각형 원자로 계통을 제공한다. 본 발명에 따른 열담수화를 위한 고유안전 수냉각형 원자로 계통은 기존의 수냉각형 원자로와 달리 고유한 안전성을 가짐으로 안전관련 문제가 근본적으로 해결된 매우 안정적인 에너지원으로 활용가능하다. 또한, 고유안전성을 가짐에 따라 기존의 수냉각형 원자로에서 필요로 하는 다양한 안전관련 계통들을 제거할 수 있어 건설비용을 낮출 수 있는 효과가 있으며, 궁극적으로 경제성 있는 담수 생산이 가능하다. 나아가, 본 발명에 따른 원자로 계통이 제공 가능한 온도 범위라면 열담수화 외에도 다른 어떠한 공정에도 활용가능한 효과가 있다.

BREAK DISC FOR A DEVICE FOR PROTECTING AGAINST SURPRESSIONS IN THE INTERIOR OF EQUIPMENT, APPLICATION TO EQUIPMENT INTENDED TO CONTAIN SUCCESSIVELY TWO GASES OF A DIFFERENT CHEMICAL NATURE

Nuclear power plant with containment and pressure release method for containment

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

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

Pressure releasing method for atomic power station, pressure releasing system for atomic power station and respective atomic power station

FIELD: atomic physics. SUBSTANCE: invention relates to a method and a device for releasing pressure of an atomic power station (2), having a protective casing (4) for holding radioactive carriers and an outlet (10, 10') for the discharge stream. The stream is directed using a discharge pipe (12, 12'), fitted with a filter system, from the protective casing (4) into the atmosphere. The filter system comprises a filter chamber (16) with a sorption filter (18). The discharge stream is first directed into a high-pressure portion (70) then undergoes pressure reduction in a throttling device (72), then at least partially directed through the filter chamber (16) with a sorption filter (18) and then finally released into the atmosphere. The discharge stream, in which pressure is reduced using the throttling device (72), is directed through an overheating portion (80) on the high-pressure portion (70) immediately before entering the filter chamber (16). EFFECT: effective retention of radioactive carriers contained in the discharge stream, particularly iodine-containing organic compounds. 36 cl, 5 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (51) МПК G21C 9/004 (13) 2 548 170 C2 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ 2013113041/07, 18.07.2011 (24) Дата начала отсчета срока действия патента: 18.07.2011 Приоритет(ы): (30) Конвенционный приоритет: (72) Автор(ы): ЭККАРДТ Бернд (DE), ЛОШ Норберт (DE), ПАСЛЕР Карстен (DE) 25.08.2010 DE 102010035509.7 (43) Дата публикации заявки: 27.09.2014 Бюл. № 27 R U (73) Патентообладатель(и): АРЕФА ГМБХ (DE) (45) Опубликовано: 20.04.2015 Бюл. № 11 C1, 20.06.1996 . RU2302674 C1, 10.07.2007 . JP3235093 A, 21.10.1991 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 25.03.2013 (86) Заявка PCT: 2 5 4 8 1 7 0 (56) Список документов, цитированных в отчете о поиске: RU2311696 C2, 27.11.2007 . RU2062514 2 5 4 8 1 7 0 R U (87) Публикация заявки PCT: WO 2012/025174 ...

Built-in containment filtering and discharging system

本发明提供一种内置式安全壳过滤排放系统，包括内置水池、射流管、汽水分离器、金属纤维过滤器、银沸石过滤器以及相应的管道阀门，本发明有效地利用了安全壳内部的水池空间，且系统简单、布置形式更加紧凑，取消了复杂的外部布置系统，为海洋浮动平台等小型核动力提供了更好的选择。利用内置水池与纵向射流管结合，有效地利用了水封原理，保证在不同流量下投入孔数的变化，使得射流喷嘴内的流速始终处于最佳状态，从而始终具有较高的气溶胶和碘去除效率。本发明采用限流孔板安装在内置式安全壳过滤排放系统的末端，整个系统处于高压下运行，可以减小设备尺寸，减少了银沸石的装载量，使设备小型化，有利于降低设备成本，提高经济型。

Method of and device for pressure relief at nuclear power station

FIELD: nuclear power engineering. SUBSTANCE: method involves heating of molecular sieve mounted on outlet line by high-pressure relief flow, drying of pressure-relief flow by expanding upon passage through metal-fiber filter, bringing of pressure-relief flow in direct contact with molecular sieve; metal-fiber filter is operated at pressure making up at least 1.2 of pressure on molecular sieve. Pressure-relief device has additional molecular sieve installed in cylindrical tank and having annular shape. Molecular sieve is placed in pressure-relief gas flow region with high pressure and has throttle mounted downstream of metal-fiber filter at molecular sieve inlet. EFFECT: provision for filtering off iodine during pressure relief in reactor containment; reduced cost and failure probability. 11 cl, 5 dwg Г 9бс90с ПЧ ГЭ РОССИЙСКОЕ АГЕНТСТВО ПО ПАТЕНТАМ И ТОВАРНЫМ ЗНАКАМ (19) (51) МПК ВИ” 2 062 514” С1 С 21С 9/00 12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ (21), (22) Заявка: 5010995/25, 16.06.1989 (46) Дата публикации: 20.06.1996 (56) Ссылки: Патент ЕР М 0285845, кл. С 21 С 9/00, 1988. (86) Заявка РСТ: ЕР 89/00678 (16.06.89) (71) Заявитель: Сименс АГ (0Е) (72) Изобретатель: Бернд Экардт[ОЕ] (73) Патентообладатель: Сименс АГ (ОЕ) (54) СПОСОБ РАЗГРУЗКИ ДАВЛЕНИЯ НА АЭС И УСТРОЙСТВО ДЛЯ ЕГО ОСУЩЕСТВЛЕНИЯ (57) Реферат: Использование: в системах разгрузки давления на АЭС и обеспечивает фильтрацию Йода при разгрузке защитной оболочки реактора при одновременном снижении затратных расходов и вероятности отказа. Способ разгрузки давления предусматривает обогрев — разгрузочным потоком высокого давления, установленного на выпускной линии молекулярного сита, осушение — разгрузочного потока путем расширения после прохождения через металловолоконный фильтр, приведение разгрузочного потока в непосредственный контакт с молекулярным ситом, причем металловолоконный фильтр эксплуатируют с давлением, составляющим по меньшей мере 1,2 от давления на молекулярном сите. Устройство для ...

Method and system for diagnostics of maximum load-bearing capacity of prestressed protective shell of nuclear power plant with strands without adhesion to concrete of shell

FIELD: ensuring the bearing capacity of reinforced concrete protective shells of nuclear power plants. SUBSTANCE: invention relates to means of ensuring the bearing capacity of reinforced concrete protective shells of nuclear power plants (NPP). A finite-element model of the NPP protective shell is formed, the required initial load-bearing capacity of the NPP protective shell is determined, taking into account data on the real tracing of strands and the forces in them from prestressing obtained from the sensors of the monitoring system and jacks, as well as taking into account the specified physical and mechanical properties of the shell structures. The sensors provide data on changes in the stress-strain state of the protective shell of the NPP, tension forces in the reinforced ropes, taking into account stress relaxation and the possibility of their slipping in the channel-forming agents, creep of concrete and other factors. Based on the obtained data, using the formed finite element model, the maximum load-bearing capacity of the NPP protective shell is predicted when it is loaded with an internal pressure exceeding the design pressure; next, the sufficiency of the tension forces of the strands or the necessary amount of their tightening is determined, or the conclusion that it is necessary to replace the strands is made. EFFECT: technical result is the possibility of diagnostics with the required reliability of the maximum load-bearing capacity of the nuclear power plant with strands installed without adhesion to the shell concrete, when loaded with internal pressure exceeding the design pressure, and, as a result, improving the safety of the nuclear power plant. 3 cl, 4 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 755 140 C1 (51) МПК G21C 9/00 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ (52) СПК G21C 9/00 (2021.02) (21)(22) Заявка: 2020136903, 10.11.2020 (24) Дата начала отсчета срока действия патента: Дата ...