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Небесная энциклопедия

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

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

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02-02-2017 дата публикации

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

Номер: RU2609670C2

FIELD: nuclear technology; ecology. SUBSTANCE: invention relates to a system for abatement of noxious emissions from industrial or nuclear power plant (1) in event of accident. System comprises following components: structure (10) for impermeabilization of ground, which extends at least in an annular area that surrounds plant (1); plurality of water-sprinkling towers (20–22), which are arranged around plant (1) and/or on adjacent territory and sprinkle water in atmosphere, preferably added with chemical and/or biological and/or mineral substances; and peripheral collection structure (50), configured for receiving water withheld by impermeabilization structure (10). EFFECT: technical result is possibility of localisation of contaminants in case of accidents on nuclear or industrial plants. 14 cl, 22 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 609 670 C2 (51) МПК G21D 1/00 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ФОРМУЛА (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ 2014109422, 27.07.2012 (24) Дата начала отсчета срока действия патента: 27.07.2012 (72) Автор(ы): БЕРТОЛОТТО Антонио (IT) (73) Патентообладатель(и): МАРКОПОЛО ИНЖИНИРИНГ С.П.А. СИСТЕМИ ЭКОЛОДЖИЧИ (IT) Дата регистрации: (56) Список документов, цитированных в отчете о поиске: US 4129627 А, 12.12.1978. US Приоритет(ы): (30) Конвенционный приоритет: 11.08.2011 IT TO2011A000763 (45) Опубликовано: 02.02.2017 Бюл. № 4 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 12.03.2014 (86) Заявка PCT: IB 2012/053858 (27.07.2012) (87) Публикация заявки PCT: 2 6 0 9 6 7 0 (43) Дата публикации заявки: 20.09.2015 Бюл. № 26 4129627 А, 12.12.1978. JP 5087983 A, 09.04.1993. US 4971752 A, 20.11.1990. EP 0491454 A1, 24.06.1992. DE 102008038262 A1, 18.02.2010. US 4056436 A, 01.11.1977. RU 2183682 C2, 20.06.2002. JP 1176988 A,13.07.1989. DE 102008038262 A1, 18.02.2010. R U 02.02.2017 2 6 0 9 6 7 0 R U Адрес для переписки: 191002, Санкт-Петербург, а/я 5, ООО "Ляпунов и партнеры ...

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27-12-2016 дата публикации

ПОДВОДНЫЙ МОДУЛЬ ДЛЯ ПРОИЗВОДСТВА ЭЛЕКТРИЧЕСКОЙ ЭНЕРГИИ

Номер: RU2605762C2
Принадлежит: ДСНС (FR)

FIELD: nuclear engineering. SUBSTANCE: invention relates to underwater nuclear power plant. Electricity production module comprises elongated cylindrical box (12) in which means are integrated forming an electricity production unit including means forming nuclear boiler (30), associated with electricity production means (37) by electrical cables (6). Cables (6) are connected to external electric power distribution station. Nuclear boiler-forming means (30) are placed in dry chamber (19) of reactor compartment (18) associated with chamber (20) forming safety water storage reservoir of reactor. Radial wall (53) of reservoir is in a heat exchange relationship with marine environment. Dry compartment (19) of reactor container (18) is connected to safety water storage reservoir chamber (20) of reactor by depressurising valves (70). Valves are placed in upper portion of dry chamber (19) and connected to bubble chamber placed in lower portion of storage reservoir chamber (20). EFFECT: higher safety of operation of module. 25 cl, 5 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 605 762 C2 (51) МПК G21C 9/004 (2006.01) G21D 3/06 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ 2014133722/07, 18.01.2013 (24) Дата начала отсчета срока действия патента: 18.01.2013 (72) Автор(ы): АРАТИК Жоффрей (FR) (73) Патентообладатель(и): ДСНС (FR) Приоритет(ы): (30) Конвенционный приоритет: (43) Дата публикации заявки: 10.03.2016 Бюл. № 7 R U 18.01.2012 FR 1250501 (45) Опубликовано: 27.12.2016 Бюл. № 36 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 18.08.2014 (86) Заявка PCT: 2 6 0 5 7 6 2 (56) Список документов, цитированных в отчете о поиске: WO2011128581 A1, 20.10.2011; RU2191321 C2, 20.10.2002. US5247553 A, 21.09.1993. US4302291 A, 24.11.1981. 2 6 0 5 7 6 2 R U (87) Публикация заявки PCT: WO 2013/107871 (25.07.2013) Адрес для переписки: 129090, Москва, ул. Б. Спасская, 25, строение 3, ООО "Юридическая ...

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05-06-2018 дата публикации

Номер: RU2016142333A3
Автор:
Принадлежит:

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20-03-2016 дата публикации

ПОГРУЖНОЙ ИЛИ ПОДВОДНЫЙ МОДУЛЬ ДЛЯ ПРОИЗВОДСТВА ЭЛЕКТРИЧЕСКОЙ ЭНЕРГИИ

Номер: RU2014133560A
Принадлежит:

... 1. Подводный модуль для производства электрической энергии, содержащий средство (12) в виде удлиненного цилиндрического кессона, в которые интегрировано средство, образующее электрический энергоблок и содержащее средство (30) в виде кипящего ядерного реактора, связанное со средством (37) производства электрической энергии, выполненным с возможностью соединения электрическими кабелями (6) с внешним пунктом (7) распределения электрической энергии, отличающийся тем, что средство (30) в виде кипящего ядерного реактора расположено в сухой камере (19) реакторного отсека (18), связанной с камерой (20) в виде резервуара для хранения воды защиты реактора, в которой по меньшей мере радиальная стенка (53) находится в состоянии теплообмена с морской средой, и тем, что средство (30) в виде кипящего ядерного реактора содержат компенсатор (33) давления, соединенный при помощи средства (80) сброса давления с камерой (20) в виде резервуара для хранения воды защиты реактора.2. Подводный модуль для производства ...

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20-09-2015 дата публикации

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

Номер: RU2014109422A
Принадлежит:

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

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17-12-2018 дата публикации

СПОСОБ УПРАВЛЕНИЯ ОСТАНОВОМ ВОДО-ВОДЯНОГО ЯДЕРНОГО РЕАКТОРА

Номер: RU2017121052A
Принадлежит: Дснс

РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2017 121 052 A (51) МПК G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ЗАЯВКА НА ИЗОБРЕТЕНИЕ (21)(22) Заявка: 2017121052, 16.12.2015 (71) Заявитель(и): ДСНС (FR) Приоритет(ы): (30) Конвенционный приоритет: 17.12.2014 FR 14 02889 35 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 16.06.2017 EP 2015/080103 (16.12.2015) (87) Публикация заявки PCT: WO 2016/097061 (23.06.2016) A Адрес для переписки: 109012, Москва, ул. Ильинка, 5/2, ООО "Союзпатент" R U (57) Формула изобретения 1. Способ управления остановом водо-водяного ядерного реактора, встроенного в погруженный модуль для производства электроэнергии, при выявлении утечки в первом и/или втором контуре в парогенераторе, снабженном предохранительным клапаном, причем указанный генератор соединен с реактором и связан с аварийным средством охлаждения, отличающийся тем, что способ включает этапы, на которых: 1. обнаруживают (1) утечку первого/второго контура парогенератора; 2. автоматически останавливают (2) реактор и изолируют поврежденный парогенератор; 3. вводят в действие (3) соответствующее аварийное средство охлаждения; 4. контролируют (4) давление в первом контуре; 5. изолируют (5) аварийное средство охлаждения поврежденного парогенератора, как только давление в первом контуре падает ниже давления срабатывания предохранительных клапанов парогенератора, и 6. продолжают (6) пассивное охлаждение реактора с помощью оставшихся парогенераторов и средств охлаждения. 2. Способ управления остановом водо-водяного ядерного реактора по п. 1, отличающийся тем, что обнаружение (1) утечки первого/второго контура осуществляют путем обнаружения одного или нескольких из следующих признаков: 1. повышенная радиоактивность второго контура из-за загрязнения теплоносителем Стр.: 1 A 2 0 1 7 1 2 1 0 5 2 (54) СПОСОБ УПРАВЛЕНИЯ ОСТАНОВОМ ВОДО-ВОДЯНОГО ЯДЕРНОГО РЕАКТОРА 2 0 1 7 1 2 1 0 5 2 (86) Заявка PCT: R U (43) Дата публикации заявки: 17.12.2018 Бюл. № ...

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10-03-2016 дата публикации

ПОГРУЖНОЙ МОДУЛЬ ДЛЯ ПРОИЗВОДСТВА ЭЛЕКТРИЧЕСКОЙ ЭНЕРГИИ

Номер: RU2014133561A
Принадлежит:

... 1. Подводный модуль для производства электрической энергии, содержащий средство (12) в виде удлиненного цилиндрического кессона, в которые интегрировано средство, образующее электрический энергоблок и содержащее средство (30) в виде кипящего ядерного реактора, связанное со средством (37) производства электрической энергии, выполненным с возможностью соединения электрическими кабелями (6) с внешним пунктом (7) распределения электрической энергии, отличающийся тем, что средство (30) в виде кипящего ядерного реактора содержит вторичный контур (36), связанный со средством (37) производства электрической энергии, и вторичный защитный контур (60), параллельно соединенный с этим вторичным контуром и содержащий по меньшей мере один вторичный пассивный теплообменник (61), расположенный снаружи подводного модуля (12) в морской среде.2. Подводный модуль для производства электрической энергии по п. 1, отличающийся тем, что средство (30) в виде кипящего ядерного реактора расположено в сухой камере ( ...

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10-03-2016 дата публикации

ПОДВОДНЫЙ МОДУЛЬ ДЛЯ ПРОИЗВОДСТВА ЭЛЕКТРИЧЕСКОЙ ЭНЕРГИИ

Номер: RU2014133722A
Принадлежит: Дснс

1. Подводный модуль для производства электрической энергии, содержащий средства в форме удлиненного цилиндрического контейнера (12), в которые встроены средства, образующие блок производства электрической энергии, содержащий средства (30), формирующие кипящий ядерный реактор, связанные со средствами (37) производства электрической энергии с возможностью их соединения посредством электрических кабелей (6) с внешним пунктом (7) распределения электрической энергии, отличающийся тем, что средства (30), формирующие кипящий ядерный реактор, размещены в сухом отделении (19) отсека (18) реактора, связанном с отделением (20), образующим резервуар хранения воды системы безопасности реактора, по меньшей мере радиальная стенка (53) которого находится в состоянии теплового обмена с морской окружающей средой, а также тем, что сухое отделение (19) отсека (18) реактора соединено с отделением (20), образующим резервуар хранения воды системы безопасности реактора, посредством средств (70) понижения давления, содержащих средства (71), образующие вентили понижения давления, размещенные в верхней части сухого отделения (19) и соединенные со средствами (72), образующими пузырьковую камеру, размещенными в нижней части образующего резервуар отделения (20).2. Подводный модуль для производства электрической энергии по п. 1, отличающийся тем, что средства (30), формирующие кипящий ядерный реактор, содержат первичный контур (31), включающий в себя по меньшей мере одну камеру (32) реактора, нагнетатель (33) давления, генератор (34) пара, и первичный насос (35), а также первичный контур (54) системы безопасности, параллельный этому первичному контуру и включающий в себя по меньшей мере первичный пассивный теплообменник (55), расположенн РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (51) МПК G21C 9/004 (13) 2014 133 722 A (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ЗАЯВКА НА ИЗОБРЕТЕНИЕ (21)(22) Заявка: 2014133722, 18.01.2013 (71) Заявитель(и): ДСНС (FR) Приоритет(ы): (30) ...

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03-04-2014 дата публикации

Containment-Schutzsystem für eine kerntechnische Anlage und zugehöriges Betriebsverfahren

Номер: DE102012213614B3
Принадлежит: AREVA GMBH

Ein Containment-Schutzsystem (2) zur Behandlung der im Containment (4) einer kerntechnischen Anlage (6), insbesondere eines Kernkraftwerks, befindlichen Atmosphäre bei kritischen Störfällen mit massiver Freisetzung von Wasserstoff (H2) und Dampf soll dazu in der Lage sein, derartige Zustände auf überwiegend passive Weise und möglichst ohne Belastung der Umgebung effektiv und schnell abzubauen. Zu diesem Zweck weist das Containment-Schutzsystem (2) erfindungsgemäß einen ein Leitungssystem (10, 72, 120, 128) umfassenden, zum Anschluss an das Containment (4) vorgesehenen Kreislauf aus dem Containment (4) und wieder zurück für einen Fluidstrom auf, und zwar mit zumindest folgenden strömungsmäßig in Reihe geschalteten Komponenten: eine Rekombinationsvorrichtung (20) zur Rekombination von im Fluidstrom enthaltenem Wasserstoff (H2) mit Sauerstoff (O2) zu Wasserdampf (H2O), eine der Rekombinationsvorrichtung (20) nachgeschaltete Kondensationsvorrichtung (74) zur Kondensation von im Fluidstrom enthaltenen ...

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10-11-1983 дата публикации

VERFAHREN ZUM BETREIBEN VON KERNKRAFTWERKSANLAGEN

Номер: DE0003316037A1
Принадлежит:

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09-03-1972 дата публикации

Номер: DE0002140924A1
Автор:
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05-01-1984 дата публикации

Device for self-checking of a coolant inventory measurement

Номер: DE0003224972A1
Принадлежит:

In a pressurised water reactor, the hydrostatic pressure difference inside a coolant line (2) is measured by a differential pressure transducer (6) via two bleed points (A, B) and a piping system (4, 5). Arranged in parallel with this transducer (6) is a further transducer (8) which measures the pressure difference arising due to the flow resistances in the coolant line (2) between the two bleed points (A, B). As a result, the operation readiness of the entire ''static'' differential pressure measurement, which is required only in the case of coolant loss, is continuously checked. ...

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05-11-2008 дата публикации

Computer-aided fault tracking for industrial plant providing instructions and countdown

Номер: GB0002448894A
Принадлежит:

A computerised system for the monitoring and tracking of faults in an industrial plant such as a nuclear power station, oil or gas rig or refinery, sewage works, water or gas pumping station, etc. providing means of ensuring that plant faults are rectified in a time commensurate with their risk to plant safety. A fault condition is entered into the system, for example, manually by an operator or automatically from plant parameters and the system displays a description of the fault condition, actions for rectifying the fault condition, the time allowed to rectify the fault condition, a countdown of the time remaining to rectify the fault condition and instructions on actions to take if the time to rectify the fault condition is exceeded. A coloured display is used. Audible and visual alarms are provided when the time limit for rectifying the fault is approaching and when it has been exceeded and means are provided to allow an operator to accept an audible alarm. Means to allow an authorised ...

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07-06-1978 дата публикации

NUCLEAR REACTOR INSTALLATION INCLUDING A CONTROL VALVE

Номер: GB0001513775A
Автор:
Принадлежит:

... 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 ...

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15-02-1980 дата публикации

VERFAHREN UND SCHALTUNGSANORDNUNG ZUR ABFUHR DER NACHZERFALLSWAERME EINES DRUCKWASSER- -REAKTORS IM STOERFALL

Номер: ATA930175A
Автор:
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18-10-2012 дата публикации

NUCLEAR REACTOR AUTOMATIC DEPRESSURIZATION SYSTEM

Номер: CA0002821170A1
Принадлежит:

A blocking device for preventing the actuation of an automatic depressurization system in a pressurized nuclear reactor system due to spurious signals resulting from a software failure. The blocking signal is removed when the coolant level within the core makeup tanks drop below a predetermined level.

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29-10-2019 дата публикации

PRESSURIZED WATER REACTOR WITH COMPACT PASSIVE SAFETY SYSTEMS

Номер: CA0002846055C

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

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06-02-2014 дата публикации

CONTAINMENT PROTECTION SYSTEM FOR A NUCLEAR FACILITY AND ASSOCIATED OPERATING METHOD

Номер: CA0002878629A1
Принадлежит:

A containment protection system (2) for treating the atmosphere present in the containment (4) of a nuclear facility (6), particularly a nuclear power plant, in case of critical incidents involving extensive release of hydrogen (H2) and steam is to be able to effectively and quickly relieve such conditions in a largely passive manner and where possible without contaminating the environment. According to the invention, the containment protection system (2) has for this purpose a circuit, which comprises a conduction system (10, 72, 120, 128) and which is provided for connecting to the containment (4), out of the containment (4) and back again for a fluid flow, more particularly having the following components fluidically connected in series: a recombination device (20) for recombining hydrogen (H2) contained in the fluid flow with oxygen (O2) to form steam (H2O); a condensation device (74) connected downstream of the recombination device (20) for condensing steam fractions contained in the ...

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13-08-1976 дата публикации

Номер: CH0000578710A5
Автор:
Принадлежит: KRAFTWERK UNION AG

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15-10-1976 дата публикации

Номер: CH0000580858A5
Автор:
Принадлежит: KRAFTWERK UNION AG

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31-05-1972 дата публикации

Ventil, insbesondere Absperrventil

Номер: CH0000523456A

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30-03-1979 дата публикации

Номер: CH0000610125A5

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15-02-1977 дата публикации

Emergency shutdown valve for reactor steam line - closes if leak occurs and partially reopens to relieve pressure build-up

Номер: CH0000584959A5
Автор:
Принадлежит: KRAFTWERK UNION AG

The Parent Patent describes a nuclear reactor with a rapid shut-off valve in the steam line (2) leading from a steam generator (1) out through the containment shell (towards the right) to a turbine. The valve cross-section is equal to the steam line cross-section. This valve is driven by pressurised fluid to the closure position, as shown in full lines, if a leak occurs in the steam line. The valve is actuated by a two-piston system, connected to the upstream side of the steam line, so that if the steam pressure rises again the valve partially opens to less than half the flow cross-section of the steam line, to permit limited escape of steam without damage. Only one of the pistons is connected rigidly to the valve plate whereas the other piston is coupled to it only over part of the stroke. In the Patent of Addition the pistons associated with the valve plate are actuated solely by a controlled release of pressure. The pistons are pref. both tubular, one inside the other. Used for a pressurised ...

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28-02-1977 дата публикации

Номер: CH0000585456A5
Автор:
Принадлежит: KRAFTWERK UNION AG

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15-08-1977 дата публикации

Номер: CH0000590592A5
Автор:

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31-05-1977 дата публикации

Номер: CH0000588147A5
Автор:
Принадлежит: KRAFTWERK UNION AG

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30-06-2016 дата публикации

VENTILATION SYSTEM AND CORRESPONDING METHOD OF OPERATION FOR USE DURING SERIOUS EMERGENCY ON NUCLEAR-TECHNICAL ENTERPRISE

Номер: EA0201690251A1
Автор:
Принадлежит:

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10-09-2020 дата публикации

VENTILATION SYSTEM AND ASSOCIATED OPERATING METHOD FOR USE DURING A SERIOUS INCIDENT IN A NUCLEAR PLANT

Номер: UA0000122066C2
Автор: HILL AXEL
Принадлежит:

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25-01-2018 дата публикации

VENTILATION SYSTEM AND ASSOCIATED OPERATING METHOD FOR USE DURING A SERIOUS INCIDENT IN A NUCLEAR PLANT

Номер: UA0000116027C2
Автор: HILL AXEL, Hill, Axel
Принадлежит:

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06-01-2017 дата публикации

가압 컨테이너를 위한 수동 감압 시스템

Номер: KR1020170002421A
Принадлежит:

... 가압 컨테이너를 위한 감압 시스템으로서, 일 측이 내부에 가스를 수용하는 가압 컨테이너(1)와 연결되고 반대 측이 대기와 연결된 개방 스프링(10)을 갖는 공압 액추에이터가 메인 밸브(8)에 제공되며, 상기 개방 스프링(10)을 소정의 기계적 압력으로 규정하여, 상기 가압 컨테이너(1) 내부의 압력이 상기 소정의 압력을 초과하는 경우, 상기 메인 밸브(8)가 잠긴 상태로 유지되고, 상기 가압 컨테이너(1) 내부의 압력이 상기 소정의 압력 미만인 경우, 상기 메인 밸브(8)가 열림으로써 가압된 가스를 방출하도록 하는 가압 컨테이너를 위한 감압 시스템이 제공된다. 상기 시스템은 어떠한 어떠한 외부적 파워 공급을 필요로 하지 않고, 사고상황, 심지어 전력을 완전 상실하는 사고상황에서도 적절하게 기능을 수행한다.

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26-06-2018 дата публикации

EMERGENCY POWER PRODUCTION SYSTEM AND NUCLEAR POWER PLANT INCLUDING SAME

Номер: KR1020180070335A
Принадлежит:

The present invention relates to an emergency power production system including: an emergency cooling water storage unit which is used as a heat sink of a passive safety system in a nuclear accident as emergency cooling water is installed inside, has an inner side which is hermetically sealed, and is designed at preset pressure or higher; and a turbine generator which produces emergency power by rotating a turbine with steam supplied from the emergency cooling water storage unit when the inner pressure of the emergency cooling water storage unit reaches the preset pressure or higher. Accordingly, the present invention can stably maintain a unique function of the passive safety system and efficiently produce electricity with a proven method such as a small turbine generator. COPYRIGHT KIPO 2018 (16) Turbine system (17) Water supply system ...

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10-10-1982 дата публикации

SNABBSTOPPANORDNING FOR KERNREAKTORER

Номер: SE0008202178L
Автор:
Принадлежит:

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14-02-2013 дата публикации

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

Номер: WO2013021308A1
Автор: BERTOLOTTO, Antonio
Принадлежит:

A system for the abatement of noxious emissions from an industrial or nuclear power plant (1) or the like in the event of accident comprises: - a structure for impermeabilization of the ground (10), which extends at least in an annular area (A1) that surrounds the plant (1); - a plurality of water-sprinkling towers (20-22), which are arranged around the plant (1) and sprinkle water in the atmosphere, preferably added with chemical and/or biological and/or mineral substances; and - a peripheral collection structure (50), configured for receiving water withheld by the impermeabilization structure (10).

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20-06-2013 дата публикации

METHOD FOR AVOIDING NUCLEAR RADIATION POLLUTION CAUSED BY NUCLEAR LEAKAGE ACCIDENT

Номер: WO2013086656A1
Автор: KUANG, Zhongping
Принадлежит:

A method for avoiding nuclear radiation pollution caused by a nuclear leakage accident comprises the following steps of: (1) moving a nuclear reactor and a safety hood thereof to the underground, the depth being set to be enough to suppress shock waves of a nuclear explosion; (2) setting an elevator raise and arranging a ventilation, pipeline and line pipe for providing the convenience for repair and maintenance; and (3) when an accident is out of control, filling fine sand grains in the reactor along the raise and the pipes, sealing the raise and the pipes by pouring cement concrete, and burying a part close to the ground by using soil.

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29-09-2016 дата публикации

SAFETY SYSTEM FOR A NUCLEAR POWER PLANT AND METHOD FOR OPERATING THE SAME

Номер: US20160284429A1
Принадлежит:

A safety system for a nuclear power plant includes first through fourth sensors; a first division, including a first calculation module that determines first and second calculation results based on signals from the first and second sensors, a first data-sharing module for sharing the first and second calculation results with a second division, and a first voting logic for generating a first safety demand signal based on the first through fourth calculation results; and the second division, including a second calculation module for determining the third and fourth calculation results based on signals from the third and fourth sensors, a second data-sharing module for sharing the third and fourth calculation results with the first division, and a second voting logic for generating a second safety demand signal based on the first, second, third, and fourth calculation results, wherein the first through fourth sensors each monitor the same plant parameters. 1. A safety system for a nuclear power plant , comprisinga first set of safety sensors including a first plurality of sensors providing a first plurality of sensor signals, respectively;a second set of safety sensors including a second plurality of sensors providing a second plurality of sensor signals, respectively;a third set of safety sensors including a third plurality of sensors providing a third plurality of sensor signals, respectively;a fourth set of safety sensors including a fourth plurality of sensors providing a fourth plurality of sensor signals, respectively; a first calculation module configured to determine first and second calculation results based on the first and second pluralities of sensor signals,', 'a first data-sharing module configured to share the first and second calculation results with a second division, and', 'a first enhanced voting logic configured to generate a first safety demand signal based on the first and second calculation results and third and fourth calculation results; and, ' ...

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06-12-2022 дата публикации

Inadvertent actuation block valve for a small modular nuclear reactor

Номер: US0011521757B2

An inadvertent actuation block valve includes inlet and outlet orifices being in selective fluid communication via a chamber. A disc is disposed within the chamber and a bellows is configured to contract at a predetermined pressure differential between reactor fluid entering a reference pressure orifice and control fluid entering the inlet orifice. When the bellows contracts, the disc engages the outlet orifice and isolates fluid communication between the inlet and outlet orifices. The inadvertent actuation block valve prevents inadvertent opening of an emergency core cooling valve when a reactor is at operating pressure that is above the predetermined set pressure range. The inadvertent actuation block valve permits the emergency cooling valves to open and to remain open when reactor pressure is below the predetermined set pressure range. The inadvertent actuation block valve does not impede long term emergency cooling that occurs when the reactor is at low pressure.

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09-04-2014 дата публикации

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

Номер: EP2715735A1
Автор: MEKISKA, Frank
Принадлежит:

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03-12-2014 дата публикации

Nuclear reactor shutdown system

Номер: EP2463864A3
Принадлежит:

To provide a reactor shutdown system (47) that can shut down a reactor (5) by various means in an event of a malfunction in a nuclear facility. A reactor shutdown system (47) that stops nuclear reaction in a reactor (5) in the event of a malfunction in a nuclear facility that includes a reactor (5), a control-rod drive unit (17) that can drive a control rod (16) in pulling and inserting directions with respect to a fuel assembly (15), a power source (31) that can supply power to the control-rod drive unit (17), and a power converter (32) that is provided between the control-rod drive unit (17) and the power source (31), in which when power supply is cut off, the control-rod drive unit (17) inserts the control rod (16) into the fuel assembly (15) to stop nuclear reaction in the reactor (5), and the reactor shutdown system (47) includes a reactor trip breaker (45) provided between the power converter (32) and the control-rod drive unit (17), a safety protection-system device (43) that controls ...

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04-07-1984 дата публикации

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

Номер: EP0000101242A3
Принадлежит:

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

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23-01-1982 дата публикации

PLANT FAULT DETECTING DEVICE

Номер: JP0057013511A
Принадлежит:

PURPOSE: To obtain a plant fault detcting device which is excellent in detecting sensitivity, by providing a means for newly taking a mean value of an input signal, in case when the extent of variation of an input signal has exceeded a preset value. CONSTITUTION: An input signal 1 and output signal 10 of an object system 8 are inputted to a mean value operating circuit 2, and a mean value 3 which is its output is stored in a memory element of a signal variation extent extractor 4. The extractor 4 outputs to a comparing and deciding device 6 the extent of variation of an input signal, which is difference between the latest value of the signals 1, 10 and a value which has been stored in the memory element. The deciding device 6 outputs to the extractor 4 an off-signal and an on-signal whe the extent of variation 5 is within a preset value, and when it is not within a present value, respectively. The extractor 4 inputs a new mean value 3 to the memory element only when the on-signal has been ...

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10-01-2017 дата публикации

ПОГРУЖНОЙ ИЛИ ПОДВОДНЫЙ МОДУЛЬ ДЛЯ ПРОИЗВОДСТВА ЭЛЕКТРИЧЕСКОЙ ЭНЕРГИИ

Номер: RU2606209C2
Принадлежит: ДСНС (FR)

Изобретение относится к подводным АЭС модульного исполнения. Модуль содержит удлиненный цилиндрический кессон (12), в который интегрирован электрический энергоблок в виде кипящего ядерного реактора (30), связанного со средством (37) производства электрической энергии, соединенным электрическими кабелями (6) с внешним пунктом (7) распределения электрической энергии. Кипящий ядерный реактор (30) расположен в сухой камере (19) реакторного отсека (18), связанной с камерой (20) в виде резервуара для хранения воды защиты реактора, в которой, по меньшей мере, радиальная стенка (53) находится в состоянии теплообмена с морской средой. Кипящий ядерный реактор (30) содержит компенсатор (33) давления, соединенный при помощи средства (80) сброса давления с камерой (20) в виде резервуара для хранения воды защиты реактора. Технический результат – повышение безопасности энергоблока. 23 з.п. ф-лы, 5 ил.

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27-03-2014 дата публикации

УСТАНОВКА ДЛЯ ПРОИЗВОДСТВА ЭНЕРГИИ НА ОСНОВЕ ГАЗООХЛАЖДАЕМОГО РЕАКТОРА НА БЫСТРЫХ НЕЙТРОНАХ

Номер: RU2012140437A
Принадлежит:

... 1. Установка по производству энергии, содержащая- первичный контур (10), содержащий первый газ, проходящий через ядерный реактор (12), первый теплообменник (14) и газодувку (16'), приводимую в действие валом (24'),- вторичный контур (17'), содержащий неконденсирующийся газ, проходящий через первый теплообменник (14), турбину (18) и компрессор (22),отличающаяся тем, что газодувка (16'), турбина (18) и компрессор (22) приводятся в действие упомянутым валом (24').2. Установка по п.1, отличающаяся тем, что первичный и вторичный контуры (10, 17') сконфигурированы так, что первый газ и неконденсирующийся газ находятся под одним и тем же давлением.3. Установка по п.2, отличающаяся тем, что клапан соединяет первичный контур (10) с вторичным контуром (17') и сконфигурирован так, что давление во вторичном контуре (17') автоматически регулируется давлением в первичном контуре (10).4. Установка по п.2, отличающаяся тем, что- второй теплообменник (20) размещен во вторичном контуре (17'), причем неконденсирующимся ...

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13-12-1973 дата публикации

Номер: DE0002140924C3

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20-10-1982 дата публикации

Nuclear reactor trip system

Номер: GB0002096809A
Автор: Cook, Bruce Michael
Принадлежит:

Each parameter of the processes of the nuclear reactor and of the components of a power supply which convert the thermal energy generated by the reactor into electrical power is monitored by a set of four like sensors. One each of the unlike sensors which monitor the different parameters is contained in a reactor-trip logic channel. Each such unlike sensor is referred to here as a "local sensor". Each channel is interlocked with the other three channels and receives the signals sensed by the other three sensors, herein called "remote sensors". Each channel also includes means for processing the signals from the local and remote sensors. The apparatus also includes means for tripping the reactor to deenergize or trip the control rod drive and insert the control rods fully into the core so that the reactor stops supplying power. The apparatus normally operates on a "two out of four" configuration. This assumes that all sensors are in normal operating condition. To achieve this purpose, eight ...

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20-09-1978 дата публикации

REMOVAL OF DECAY HEAT ON FAILURE OF A PRESSURISED-WATER REACTOR

Номер: GB0001525020A
Принадлежит:

In the event of operating leaks from the primary to the secondary side in the steam generator (5), of the failure of the main heat sink and/or of breakage of the main steam pipe, the amount of afterheat and steam still produced after shutdown of the reactor (3) is condensed using an emergency circuit (13, 14, 15, 16, 17) and the condensate is resupplied to the steam generator. The emergency circuit is connected via a controllable pressure-reducing valve (10) to the main steam pipes (7), which are connected to the steam generator. It has, inter alia, an afterheat removal heat exchanger (13) with an additional condensate cooler (17), and is connected to the steam generator (5) via an emergency feedwater pump (14), an emergency feedwater control valve (15) and a connecting piece (16). As a result, radioactive contamination of the environment is prevented even in the event of leaks in the steam generator. ...

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14-06-1978 дата публикации

NUCLEAR REACTOR INSTALLATION INCLUDING A CONTROL VALVE

Номер: GB0001514074A
Автор:
Принадлежит:

... 1514074 Fluid-pressure servomotor systems KRAFTWERK UNION AG 29 April 1976 [30 April 1975] 17569/76 Addition to 1513775 Heading G3P [Also in Division F2] The safety valve in the nuclear reactor installation of the parent Specification is modified by the provision of an auxiliary valve in the closure member 7, Fig. 1, of the valve, whereby the opening of the valve by up to one-half of its fully open cross-section is provided by opening the auxiliary valve. The latter comprises a closure member 24 actuated by a piston 27 slidable in a cylinder 22 formed in the connecting member 8 between the closure member 7 and its actuating piston 10. The spaces above and below the pistons 10, 27 are connected to the main flow path via bores 16, 18, 71 each having a respective restrictor 17, 19, 72. Electromagnetically-actuated valves 35, 43, 44, 45, 52, 53 are operated at various pressures to effect the desired opening or closing of the members 7, 24. Each piston 10, 27 has a respective raised portion ...

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27-10-2003 дата публикации

AUTOMATICALLY SCRAMMING NUCLEAR REACTOR SYSTEM

Номер: AU2003228513A1
Принадлежит:

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16-04-1985 дата публикации

METHOD AND APPARATUS FOR AUTOMATIC ABNORMAL EVENTS MONITOR IN OPERATING PLANTS

Номер: CA1185691A

An apparatus and method for automatically monitoring dynamic signals, such as from vibration sensors, in an operating industrial or other plant to identify abnormal events, draw conclusions as to their severity, and indicate action to be taken, utilizing a computer to control the scanning of one or two sensor channels at a time through a matrix of analog switches, and to process one or two channel signals through a signal processor for power spectral density (PSD) analysis (two channel signals for cross PSD analysis). The computer compares spectra with predetermined sets of frequency dependent limits and indicates the abnormal condition of apparatus in the plant associated with the spectra as a function of which set of limits is exceeded. The computer also indicates from a stored table the action to be taken for the abnormal condition found.

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16-04-1985 дата публикации

METHOD AND APPARATUS FOR AUTOMATIC ABNORMAL EVENTS MONITOR IN OPERATING PLANTS

Номер: CA0001185691A1
Принадлежит:

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30-07-2015 дата публикации

NUCLEAR REACTOR PROTECTION SYSTEMS AND METHODS

Номер: CA0002927946A1
Принадлежит:

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.

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27-07-2021 дата публикации

NUCLEAR POWER PLANT AND SAFETY SYSTEM WITH FUSE ELEMENT AND GRAVITY ELEVATOR

Номер: CA2865607C

The present invention relates to a nuclear power plant and safety sys- tem with fuse element (3) and gravity elevator (100), the buildings (6, 7, 9) of the power plant subjected to contamination being buried below sea level and under borated water basins (8), and having a safety system free of electrical and electronic components to act in the event of possible accidents comprising, among others, means for flooding the buildings of the power plant with thermal fuses and gravi- ty elevators for operator evacuation in the event of an emergency.

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25-10-2018 дата публикации

PASSIVE DEPRESSURISATION SYSTEM FOR PRESSURISED RECEPTACLES

Номер: UA0000117963C2
Принадлежит:

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31-03-2014 дата публикации

SYSTEM AND METHOD FOR EMERGENCY SUPPLY NUCLEAR - TECHNICAL POWER PLANT

Номер: EA0201391800A1
Автор:
Принадлежит:

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30-11-2017 дата публикации

VENTILATION SYSTEM AND CORRESPONDING METHOD OF OPERATION FOR USE DURING SERIOUS ACCIDENT IN NUCLEAR PLANT

Номер: EA0201791609A1
Автор:
Принадлежит:

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14-04-1972 дата публикации

EMERGENCY COOLING OF A GAS-COOLED NUCLEAR REACTOR

Номер: FR0002103483A3
Автор:
Принадлежит:

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14-11-2014 дата публикации

TANK COOLING SYSTEM AND EMERGENCY COOLING NUCLEAR POWER PLANT WITH SAID BEARING

Номер: FR0003005523A1
Принадлежит: KOREA ATOMIC ENERGY RESEARCH INSTITUTE

L'invention concerne un système de refroidissement d'un réservoir de refroidissement d'urgence qui comprend un réservoir configuré pour stocker de l'eau de refroidissement, recevant la chaleur transférée d'un réacteur nucléaire ou d'un confinement. Un dispositif d'échange de chaleur exposé à l'extérieur du réservoir pour fonctionner dans l'air, et configuré un échange de chaleur entre le fluide dans le réservoir et l'air et une unité d'ouverture et de fermeture installée au niveau d'une partie supérieure du réservoir configurée pour être ouverte par un écoulement du fluide généré par une évaporation de l'eau de refroidissement, suite à une différence de pression provenant de l'air extérieur supérieure consigne.

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19-10-2012 дата публикации

PASSIVE SYSTEM HAS RESERVE Of INJECTION OF SAFETY HAS High pressure (HPSIT) TO ANSWER HAS a LOSS OF NETWORK OF STATION (SBO) AND HAS ACCIDENTS OF LOSS OF COOLANT (LOCA)

Номер: FR0002974227A1

Système à réservoir d'injection de sécurité à haute pression (HPSIT) qui comprend un réservoir d'injection de sécurité (SIT) (40) qui remplace un réservoir d'appoint du cœur (CMT) et un réservoir d'injection de sécurité (SIT) à basse pression (approximativement 4,3 Mpa ou moins) et qui peut passer à, et fonctionner dans, un mode de fonctionnement à haute pression (approximativement 17 Mpa), pour permettre l'injection de fluide caloporteur de cœur d'urgence (40a) dans un système de réacteur à la fois dans des conditions de basse pression (approximativement 4,3 Mpa ou moins) et de haute pression (approximativement 17 Mpa). A High Pressure Safety Injection Tank (HPSIT) system that includes a Safety Injection Tank (SIT) (40) that replaces a core backup tank (CMT) and a safety injection tank ( SIT) at low pressure (approximately 4.3 MPa or less) and which can switch to, and operate in, a high pressure operating mode (approximately 17 MPa), to allow injection of emergency heart coolant (40a) in a reactor system under both low pressure (approximately 4.3 MPa or less) and high pressure (approximately 17 MPa) conditions.

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29-08-2017 дата публикации

가압수형 핵 원자로의 정지를 관리하기 위한 방법

Номер: KR1020170098222A
Принадлежит:

... 본 발명은, 안전 밸브가 제공되어 있는 증기 발생기에서 일차 및/또는 이차 누출이 검출된 경우에, 잠수형 전력 생산 모듈에 통합되어 있는 가압수형 핵 원자로의 정지를 관리하기 위한 방법에 관한 것으로, 증기 발생기는 원자로에 연결되어 있고 또한 대기(standby) 냉각 수단과 연관되어 있으며, 본 방법은, 증기 발생기의 일차/이차 누출을 검출하는 단계(1); 원자로를 자동적으로 정지시키고 손상된 증기 발생기를 격리시키는 단계(2); 대응하는 대기 냉각 수단을 접속시키는 단계(3); 일차 압력을 모니터링하는 단계(4); 일차 압력이 증기 발생기의 안전 밸브의 설정 압력 아래로 되면, 파손된 증기 발생기의 대기 냉각 수단을 격리시키는 단계(5); 및 나머지 증기 발생기 및 냉각 수단으로 원자로를 계속 수동으로 냉각시키는 단계(6)를 포함한다.

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25-07-2013 дата публикации

SUBMERGED OR UNDERWATER ELECTRICITY GENERATION MODULE

Номер: WO2013107879A1
Автор: HARATYK, Geoffrey
Принадлежит:

The invention relates to an underwater electricity generation module of the type comprising means in the form of an elongate cylindrical casing (12) including integrated means forming an electricity generation unit having nuclear boiler means (30) associated with electricity generation means (37) connected to an external electrical distribution station by means of electric cables. The module is characterised in that the nuclear boiler means (30) are placed in a dry chamber (19) of a reactor compartment (18) associated with a chamber forming a safety water storage tank (20) of the reactor, of which at least one radial wall (53) is in a heat-exchange relationship with the marine environment. The module is also characterised in that the nuclear boiler means (30) comprise a pressuriser (33) connected via blowdown means (80) to the chamber (20) forming the safety water storage tank of the reactor.

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10-01-2017 дата публикации

ПОГРУЖНОЙ МОДУЛЬ ДЛЯ ПРОИЗВОДСТВА ЭЛЕКТРИЧЕСКОЙ ЭНЕРГИИ

Номер: RU2606207C2
Принадлежит: ДСНС (FR)

Изобретение относится к подводным модулям для производства электрической энергии. Модуль содержит удлиненный цилиндрический кессон (12), в который интегрирован электрический энергоблок, содержащий кипящий ядерный реактор (30), связанный со средством (37) производства электрической энергии, соединенный при помощи электрических кабелей (6) с внешним пунктом (7) распределения электрической энергии. Кипящий ядерный реактор (30) содержит вторичный контур (36), связанный со средством (37) производства электрической энергии, и вторичный защитный контур (60), параллельно соединенный с этим вторичным контуром и содержащий по меньшей мере один вторичный пассивный теплообменник (61), расположенный снаружи подводного модуля (12) в морской среде. 24 з.п. ф-лы, 5 ил. РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 606 207 C2 (51) МПК G21C 9/004 (2006.01) G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ФОРМУЛА (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ 2014133561, 18.01.2013 (24) Дата начала отсчета срока действия патента: 18.01.2013 (72) Автор(ы): АРАТИК Жоффрей (FR) (73) Патентообладатель(и): ДСНС (FR) Дата регистрации: (56) Список документов, цитированных в отчете о поиске: US5247553 A, 21.09.1993 Приоритет(ы): (30) Конвенционный приоритет: ;RU2191321 C2, 20.10.2002 ;US4088535 A1, 09.05.1978;US3547778 A1, 15.12.1970. 18.01.2012 FR 1250499 (45) Опубликовано: 10.01.2017 Бюл. № 1 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 18.08.2014 (86) Заявка PCT: EP 2013/050961 (18.01.2013) (87) Публикация заявки PCT: 2 6 0 6 2 0 7 (43) Дата публикации заявки: 10.03.2016 Бюл. № 7 R U 12.12.2016 2 6 0 6 2 0 7 R U Адрес для переписки: 129090, Москва, ул. Б. Спасская, 25, строение 3, ООО "Юридическая фирма Городисский и Партнеры" (54) ПОГРУЖНОЙ МОДУЛЬ ДЛЯ ПРОИЗВОДСТВА ЭЛЕКТРИЧЕСКОЙ ЭНЕРГИИ (57) Формула изобретения 1. Подводный модуль для производства электрической энергии, содержащий средство (12) в виде удлиненного цилиндрического ...

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10-05-2015 дата публикации

УСТАНОВКА ДЛЯ ПРОИЗВОДСТВА ЭНЕРГИИ НА ОСНОВЕ ГАЗООХЛАЖДАЕМОГО РЕАКТОРА НА БЫСТРЫХ НЕЙТРОНАХ

Номер: RU2550504C2

Настоящее изобретение относится к ядерной энергетической установке (ЯЭУ). ЯЭУ содержит первичный контур (10), содержащий газ, проходящий через ядерный реактор (12), через первый теплообменник (14) и через газодувку (16'). Вторичный контур (17'), содержащий неконденсирующийся газ, проходит через первый теплообменник (14), и через турбину (18) и компрессор (22), установленные на одном валу (24'). Газодувка приводится в действие валом. Газ в первичном и вторичном контурах одинаковый, и давление во вторичном контуре автоматически регулируется давлением в первичном контуре. Технический результат - продолжение работы газодувки при аварийном отключении реактора. 5 з.п. ф-лы, 6 ил.

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20-11-2019 дата публикации

СПОСОБ УПРАВЛЕНИЯ ОСТАНОВОМ ВОДО-ВОДЯНОГО ЯДЕРНОГО РЕАКТОРА

Номер: RU2706739C2
Принадлежит: ДСНС (FR)

FIELD: nuclear reactors.SUBSTANCE: invention relates to a method for controlling the shutdown of a water-cooled nuclear reactor. In case of leak detection from first and/or second circuit in steam generator, leakage of first/second circuit of steam generator is detected; automatically shutting down the reactor and insulating the damaged steam generator; the corresponding emergency cooling means is actuated as soon as the pressure in the first circuit falls below the safety pressure of the safety valves of the steam generator, the emergency means of cooling of the damaged steam generator is isolated, and passive cooling of reactor is continued with the help of remaining steam generators and cooling means. First and/or the second circuit in the steam generator is equipped with a safety valve, besides, the generator is connected to the reactor and connected to the emergency cooling facility.EFFECT: technical result is possibility to eliminate leakage of heat carrier from first circuit of steam generator to second without intervention of operator or external power supply in water-and-water reactor with passive safety without using main safety systems, providing for quick depressurisation in the first circuit by opening of protective shell or other intermediate container to minimize contamination of territory, and without emissions into environment.4 cl, 3 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 706 739 C2 (51) МПК G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ ИЗОБРЕТЕНИЯ К ПАТЕНТУ (52) СПК G21C 15/18 (2019.05) (21)(22) Заявка: 2017121052, 16.12.2015 (24) Дата начала отсчета срока действия патента: Дата регистрации: 20.11.2019 R U 16.12.2015 (72) Автор(ы): ГУРМЕЛЬ Венсан (FR), АРАТИК Жоффрей Поль Этьен (FR), ПЮСЕТТИ Фабьен (FR), МАСДЮПЮИ Жан (FR) (73) Патентообладатель(и): ДСНС (FR) 17.12.2014 FR 14 02889 (43) Дата публикации заявки: 17.12.2018 Бюл. № 35 (56) Список документов, цитированных в отчете о поиске: US 5309487 A1, ...

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06-06-2019 дата публикации

Номер: RU2017121052A3
Автор:
Принадлежит:

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04-11-1976 дата публикации

Emergency shutdown valve for reactor steam line - closes if leak occurs and partially reopens to relieve pressure build-up

Номер: DE0002519346A1
Принадлежит:

The Parent Patent describes a nuclear reactor with a rapid shut-off valve in the steam line (2) leading from a steam generator (1) out through the containment shell (towards the right) to a turbine. The valve cross-section is equal to the steam line cross-section. This valve is driven by pressurised fluid to the closure position, as shown in full lines, if a leak occurs in the steam line. The valve is actuated by a two-piston system, connected to the upstream side of the steam line, so that if the steam pressure rises again the valve partially opens to less than half the flow cross-section of the steam line, to permit limited escape of steam without damage. Only one of the pistons is connected rigidly to the valve plate whereas the other piston is coupled to it only over part of the stroke. In the Patent of Addition the pistons associated with the valve plate are actuated solely by a controlled release of pressure. The pistons are pref. both tubular, one inside the other. Used for a pressurised ...

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13-04-1988 дата публикации

Feedwater control in a PWR following reactor trip

Номер: GB0002195815A
Принадлежит:

Excessive primary cooldown and start up of the auxiliary feedwater system in a pressurized water reactor (PWR) following a reactor trip are avoided by switching to primary coolant Tavg programmed, main feedwater flow control in place of normal level programmed control. The Low Tavg set point at which the main feedwater isolation valves are closed is lowered below the no load Tavg temperature set point and the flow rate is decreased gradually to zero at a Tavg between these two set points. In addition, the low-low level set point at which the auxiliary feedwater system is started is lowered if possible, and/or the level signal is filtered, or the start signal delayed, to increase the margin to the low-low set point.

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25-07-2018 дата публикации

Safety system

Номер: GB0002559034A
Принадлежит:

A passive coolant injection system comprising a make-up tank 17 is disclosed, said system for use in a nuclear power plant including a reactor pressure vessel (RPV) 1 in fluid communication with a pressuriser 11 via surge line 13. During normal operation of the system, the make-up tank is maintained at the same pressure as the primary coolant circuit via pressure balance line 23 connecting a tank inlet 19 to the surge line, while a tank outlet 21 is isolated from the RPV by valve 27. Upon detection of a fault condition when the coolant level in the pressuriser drops below a threshold level, e.g. where one or more heaters 15 in the pressurizer are exposed, the injection system switches to isolate the tank inlet from the RPV by closing surge line valve 29, and simultaneously opens outlet line valve 25 so as to allow boronated water to flow into the RPV from the make-up tank, said flow being driven by residual pressure in the pressuriser.

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06-06-2007 дата публикации

Computer aided fault tracking system (limitraK)

Номер: GB0000708442D0
Автор:
Принадлежит:

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26-06-2013 дата публикации

A process for the rapid shut-down of nuclear fission reactions

Номер: GB0002497756A
Принадлежит:

Molten boron trioxide (B2O3) is introduced to a nuclear fission environment resulting in a neutron poison reaction and the eventual full shut down of an active nuclear fission reaction within a nuclear reactor pressure vessel, pressure chamber, or equivalent construction that may be subject to fission reaction events, such as a spent fuel pond or reprocessing facility. The boron trioxide may be added in a pre-melted liquid form, or derived from boric acid (H3BO3) introduced directly into the high temperature nuclear fission environment. The boron trioxide or boracic acid should be arranged to completely fill the reaction vessel or chamber so as to isolate and enclose any exposed or broken off fuel components, seal any breach in the vessel or chamber, and negate the possibility of any combustible gas ignition.

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21-09-1983 дата публикации

PREVENTING CRITICALITY IN NUCLEAR GENERATING UNIT

Номер: GB0008320802D0
Автор:
Принадлежит:

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17-01-2018 дата публикации

Safety system

Номер: GB0201720014D0
Автор:
Принадлежит:

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15-07-1984 дата публикации

VALVE ARRANGEMENT

Номер: AT0000128076A
Автор:
Принадлежит:

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15-02-1980 дата публикации

PROCEDURE AND SWITCHING CONFIGURATION FOR THE REMOVAL THE NACHZERFALLSWAERME OF A PRESSURE WATER - REACTOR IN THE INCIDENT

Номер: AT0000930175A
Автор:
Принадлежит:

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15-07-1984 дата публикации

VENTILANORDNUNG

Номер: ATA128076A
Автор:
Принадлежит:

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06-10-2016 дата публикации

SAFETY SYSTEM FOR A NUCLEAR POWER PLANT AND METHOD FOR OPERATING THE SAME

Номер: CA0002981047A1
Принадлежит:

A safety system (100) for a nuclear power plant includes first through fourth sensors (S1, S4, S7, S10); a first division (110), including a first calculation module (120) that determines first and second calculation results (DA) based on signals from the first and second sensors (S1, S4), a first data-sharing module (123) for sharing the first and second calculation results (DA) with a second division (115), and a first voting logic (126) for generating a first safety demand signal based on the first through fourth calculation results (DA, DB).

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24-10-2013 дата публикации

DEFENSE IN DEPTH SAFETY PARADIGM FOR NUCLEAR REACTOR

Номер: CA0002870859A1
Принадлежит:

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.

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23-01-2014 дата публикации

NUCLEAR POWER PLANT, SAFETY SYSTEM WITH FUSE ELEMENT AND GRAVITY ELEVATOR

Номер: CA0002865607A1
Принадлежит:

The present invention relates to a nuclear power plant and safety sys- tem with fuse element (3) and gravity elevator (100), the buildings (6, 7, 9) of the power plant subjected to contamination being buried below sea level and under borated water basins (8), and having a safety system free of electrical and electronic components to act in the event of possible accidents comprising, among others, means for flooding the buildings of the power plant with thermal fuses and gravi- ty elevators for operator evacuation in the event of an emergency.

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01-09-2011 дата публикации

FACILITY FOR PRODUCING ENERGY FROM A GAS-COOLED FAST REACTOR

Номер: CA0002789172A1
Принадлежит:

The present invention relates to a facility for producing energy comprising a gas primary circuit (10) passing through a nuclear reactor (12), through a first heat exchanger (14), and through a fan (16'). An incondensable-gas secondary circuit (17') passes through the first exchanger (14), and through a turbine (18) and a compressor (22) which are mounted on one and the same shaft (24'). The fan is driven by the shaft. The gases in the primary and secondary circuits are of the same kind, and the pressure in the secondary circuit is slaved to the pressure in the primary circuit.

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22-01-2015 дата публикации

VENTILATION SYSTEM AND ASSOCIATED OPERATING METHOD FOR USE DURING A SERIOUS INCIDENT IN A NUCLEAR PLANT

Номер: CA0002918394A1
Автор: HILL, AXEL, HILL AXEL
Принадлежит:

A ventilation system (2) for an operating room accessible to service personnel in a nuclear plant, in particular a control room (4) in a nuclear power plant (6), is intended to enable a supply of decontaminated fresh air at least for a time span of a few hours in the event of serious incidents involving the release of radioactive activity. In particular, the content of radioactive inert gases in the fresh air supplied to the operating room should be as low as possible. For this purpose, in accordance with the invention the ventilation system (2) is equipped with an air supply line (10) passed from an external inlet (14) to the operating room, with a first fan (12) and a first inert gas adsorber column (e.g. 38) being connected into said air supply line (10), an air discharge line (44) passed from the operating room to an external outlet (72), with a second fan (46) and a second inert gas adsorber column (e.g. 48) being connected into said air discharge line (44), and switchover means for ...

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30-04-2015 дата публикации

RADIOACTIVE IODINE ADSORBENT, AND METHOD FOR TREATING RADIOACTIVE IODINE

Номер: CA0002927657A1
Принадлежит:

Provided is an adsorbent for radioactive iodine, said adsorbent being capable of: adsorbing radioactive iodine more effectively than a conventional adsorbent; and removing hydrogen which is one of the causes of reactor accidents. An adsorbent which is to be used for radioactive iodine and which is obtained by granulating X-type zeolite, wherein: the sizes of micropores of the X-type zeolite are adjusted to the molecular size of hydrogen by replacing ion-exchange sites of the X-type zeolite with silver; the silver content in a dry state is 36wt% or more; the sizes of the particles are 10×20 mesh; the hardness is 94% or more; and the water content is 12wt% or less as determined after the weight reduction by drying at 150ºC for 3 hours.

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09-03-2016 дата публикации

Ventilation system and associated operating method for use during a serious incident in a nuclear plant

Номер: CN0105393310A
Автор: HILL AXEL
Принадлежит:

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22-03-2017 дата публикации

For cooling the reactor core of the passive system

Номер: CN0104919531B
Автор:
Принадлежит:

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07-04-2020 дата публикации

Method for analyzing withdrawal time of power operation loss final heat trap of nuclear power unit

Номер: CN0107863168B
Автор:
Принадлежит:

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14-02-1986 дата публикации

INSTALLATION OF CUT Of a Nuclear reactor

Номер: FR0002503922B1
Принадлежит:

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09-05-1980 дата публикации

CIRCUIT COMPRISED OF INSULATED GATE FIELD EFFECT TRANSISTORS

Номер: FR0002252628B1
Автор:
Принадлежит:

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02-12-2019 дата публикации

Method of Selectiong the Appropriate Abnormal Operationg Procedure for Fire in Nuclear Power Plant Fire Area

Номер: KR0102050883B1
Автор:
Принадлежит:

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06-04-2020 дата публикации

SUBMERGED OR UNDERWATER ELECTRICITY GENERATION MODULE

Номер: KR0102097839B1
Принадлежит: 나발 그룹

본 발명에 따른 침수 또는 수중 전기 생산 모듈은, 일체화되는 전기 생산 유닛을 형성하는 수단인 가늘고 긴 실린더형 박스(12) 형태의 수단을 포함하는 타입이고, 전기 케이블(6)에 의해 외부의 전기 분배 스테이션(7)에 연결된 전기 생산 수단 (37)과 연관된, 핵 보일러(30)를 형성하는 수단을 포함하는 전력 생산 장치를 형성하는 상기 수단, 보일러 핵 형성 수단 (30)은 반응기의 안전 저수 탱크(20)를 형성하는 챔버와 관련된 반응기 컴파트먼트(18)의 건조 챔버(19)에 배치되고, 적어도 반경 벽(53)은 해양 환경과 열 교환 관계에 있고, 핵 보일러-형성 수단(30)은 반응기의 안전 저수 챔버(20)에 감압 수단(80)에 의해서 연결되는 감압기(33)를 포함한다. The submerged or underwater electrical production module according to the present invention is of a type including means in the form of an elongated cylindrical box 12 which is a means for forming an integrated electrical production unit, and external electrical distribution by an electric cable 6 Said means for forming a power generating device comprising means for forming a nuclear boiler 30, associated with electricity production means 37 connected to the station 7, said boiler nucleation means 30 is a safe storage tank of the reactor ( 20) is disposed in the drying chamber 19 of the reactor compartment 18 associated with the chamber forming, at least the radius wall 53 is in a heat exchange relationship with the marine environment, the nuclear boiler-forming means 30 is And a pressure reducer 33 connected by a pressure reducing means 80 to the safety storage chamber 20 of the reactor.

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14-06-1971 дата публикации

Номер: SE0000335889B
Автор:
Принадлежит:

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21-02-2013 дата публикации

Backup nuclear reactor auxiliary power using decay heat

Номер: US20130044851A1
Принадлежит: Westinghouse Electric Co LLC

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

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03-04-2014 дата публикации

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

Номер: US20140093025A1
Автор: Mekiska Frank
Принадлежит:

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

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18-01-2018 дата публикации

APPARATUS AND SYSTEM FOR SIMULATING MAINTENANCE OF REACTOR CORE PROTECTION SYSTEM

Номер: US20180019029A1

A system for simulating maintenance of a reactor core protection system that has at least two or more channels, includes: a simulation signal generation unit for generating a simulation state signal including a normal state or an abnormal state, a communication unit connected to each of the channels of the reactor core protection system to transmit the state signal to the channel, and a control unit for receiving a result signal output from the channel in response to the input simulation state signal and confirming whether the reactor core protection system normally determines a reactor core state by analyzing the result signal. 1. An apparatus for simulating maintenance of a reactor core protection system including at least two or more channels , the apparatus comprising:a simulation signal generation unit configured to generate a simulation state signal including a normal state or an abnormal state,a communication unit connected to each of the channels of the reactor core protection system and configured to transmit the state signal to each of the channels, anda control unit configured to receive a result signal output from each of the channels in response to the input simulation state signal and to confirm whether the reactor core protection system normally determines a reactor core state by analyzing the result signal.2. The apparatus according to claim 1 , wherein the simulation signal generation unit generates the simulation state signal including at least any one of a reactor temperature claim 1 , a reactor pressure claim 1 , a hot leg temperature claim 1 , a pump rotation speed claim 1 , a neutron level claim 1 , a flow rate and a reactor control rod position.3. The apparatus according to claim 2 , whereinthe simulation signal generation unit generates first to fourth simulation state signals for the reactor control rod position, andthe communication unit transmits the first simulation state signal to a first channel, the second simulation state signal to a ...

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25-02-2016 дата публикации

Boiling Water Type Nuclear Power Plant

Номер: US20160055924A1
Принадлежит:

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

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22-02-2018 дата публикации

PWR DECAY HEAT REMOVAL SYSTEM IN WHICH STEAM FROM THE PRESSURIZER DRIVES A TURBINE WHICH DRIVES A PUMP TO INJECT WATER INTO THE REACTOR PRESSURE VESSEL

Номер: US20180053571A1
Автор: GRAHAM Thomas G.
Принадлежит:

In conjunction with a pressurized water reactor (PWR) and a pressurizer configured to control pressure in the reactor pressure vessel, a decay heat removal system comprises a pressurized passive condenser, a turbine-driven pump connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and turbine may be mounted on a common shaft via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) and/or the pressurized passive condenser. A pressurizer power operated relief valve may control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer. 1. A method operating in conjunction with a pressurized water reactor (PWR) including a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel also containing primary coolant water , a pressurizer integral with or operatively connected with the reactor pressure vessel and configured to control pressure in the reactor pressure vessel , and a refueling water storage tank (RWST) , the method comprising responding to a loss of heat sinking of the PWR by operations including:driving a turbine using steam piped from the pressurizer; anddriving a pump using the turbine to suction water from the RWST into the reactor pressure vessel.2. The method of wherein the driving of the pump comprises providing a common shaft mechanically connecting the turbine and the pump whereby the driven turbine rotates the common shaft to drive the pump.3. The method of further comprising:discharging steam piped from the pressurizer into a pressurized passive condenser; andconnecting the suction side of the pump to both the RWST and the pressurized passive condenser ...

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11-03-2021 дата публикации

DOUBLE INCOMING BREAKER SYSTEM FOR POWER SYSTEM OF POWER PLANT

Номер: US20210075208A1
Принадлежит:

The present invention is applied to a power system of a power plant including a three-winding transformer, and relates to a double incoming breaker system, including: a plurality of main circuit breakers respectively connected one by one to the plurality of first non-safety class high voltage buses and the plurality of second non-safety class high voltage buses; a plurality of auxiliary circuit breakers, one of which is connected in series to one of the plurality of main circuit breakers; a first power source measurer installed to correspond to the main circuit breaker and measuring a power source level of a non-safety class high voltage bus applied to the main circuit breaker; a second power source measurer installed to correspond to the auxiliary circuit breaker and measuring a power source level at an installed first point thereof; and a controller that outputs a first open signal to the main circuit breaker when an abnormal situation of the non-safety class high voltage bus is checked through the power source level measured by the first power source measurer, and outputs a second open signal to the auxiliary circuit breaker when it is determined that the main circuit breaker fails through the power source level at the first point measured by the second power source measurer after outputting the first open signal. 1. A double incoming breaker system for a power system of a power plant including a three-winding transformer of which a primary winding is coupled to an output side of a generator or to a switch yard , of which a first secondary winding is coupled to each non-safety class facility through a plurality of first non-safety class high voltage buses , of which a second secondary winding is coupled to each non-safety class facility through a plurality of second non-safety class high voltage buses , and of which the second secondary winding is coupled to each safety class facility through a plurality of safety class high voltage buses , comprising:a plurality ...

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15-03-2018 дата публикации

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

Номер: US20180075935A1
Принадлежит:

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

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23-04-2015 дата публикации

Combustion controller for combustible gas

Номер: US20150108128A1

Provided is a combustion controller for a combustible gas of a pressurized water reactor nuclear power plant, and more particularly, to a combustion controller for a combustible gas installed in a rear end of a filtered vent system outside a containment vessel or an external chimney, configured to convert a combustible gas such as hydrogen, carbon monoxide, or the like, into steam, carbon dioxide, or the like, and simultaneously, operate by itself with no external power supply. Accordingly, the combustion controller for a combustible gas can perform stable combustion control with no probability of explosion of hydrogen through a recombining reaction of the combustible gas, prevent discharge of carbon monoxide, which is a toxic gas, and prevent backward flow of the flame through the quenching mesh.

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28-04-2016 дата публикации

HEAT TRANSFER METHODS FOR NUCLEAR PLANTS

Номер: US20160118148A1
Принадлежит: GE-HITACHI NUCLEAR ENERGY AMERICAS LLC

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

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05-05-2016 дата публикации

Power Plant

Номер: US20160125965A1
Принадлежит:

A power plant includes a steam generator, a turbine driven by steam generated by the steam generator, a condenser which cools the steam discharged from the turbine to form condensate water by using seawater, a condensate water pipe which supplies the condensate water from the condenser to the steam generator, at least one seawater leak detection device which is included in the condensate water pipe and measures water quality of the condensate water to detect a leak of seawater in the condenser, an attemperator spray which connects to the condensate water pipe to be supplied with the condensate water from a connecting point where the attemperator spray connects to the condensate water pipe, and sprays the condensate water to the steam inside the condenser, and a pipe which diverges from the condensate water pipe and supplies the condensate water to the steam generator, wherein if the seawater leak detection device detects a leak of the seawater in the condenser, the power plant stops pouring the condensate water from the connecting point to the steam generator and stops pouring the condensate water to the pipe diverging from the condensate water pipe. 1. A power plant comprising:a steam generator;a turbine driven by steam generated by the steam generator;a condenser which cools the steam discharged from the turbine to form condensate water by using seawater;a condensate water pipe which supplies the condensate water from the condenser to the steam generator;at least one seawater leak detection device which is included in the condensate water pipe and measures water quality of the condensate water to detect a leak of seawater in the condenser;an attemperator spray which connects to the condensate water pipe to be supplied with the condensate water from a connecting point where the attemperator spray connects to the condensate water pipe, and sprays the condensate water to the steam inside the condenser; anda pipe which diverges from the condensate water pipe and ...

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03-05-2018 дата публикации

DIGITAL PROTECTION SYSTEM FOR NUCLEAR POWER PLANT

Номер: US20180122524A1

A digital protection system is provided. The digital protection system may include: a process protection system including at least two channels, each channel including a first bistable logic controller and a second bistable logic controller which are independent and different from each other, the first bistable logic controller and the second bistable logic controller outputting bistable logic results; a reactor protection system including at least two trains, each train including a first coincidence logic controller and a second coincidence logic controller which are independent and different from each other, the first coincidence logic controller and the second coincidence logic controller outputting coincidence logic results by receiving the bistable logic results from the process protection system; and an initiation circuit normally operating or stopping a reactor according to the coincidence logic results received from the reactor protection system. 1. A digital protection system comprising:a process protection system comprising at least two channels, each of the at least two channels comprising a first bistable logic controller and a second bistable logic controller which is independent and different from the first bistable logic controller, the first bistable logic controller and the second bistable logic controller receiving a process parameter and outputting bistable logic results based on the process parameter; anda reactor protection system comprising at least two trains, at least two initiation circuits, and a parallel circuit, whereineach of the two trains comprises a first coincidence logic controller and a second coincidence logic controller which is independent and different from the first coincidence logic controller, the first coincidence logic controller and the second coincidence logic controller outputting coincidence logic results based on the bistable logic results,each of the at least two initiation circuits comprises a serial circuit in ...

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14-05-2015 дата публикации

NUCLEAR POWER PLANT, SAFETY SYSTEM WITH FUSE ELEMENT AND GRAVITY ELEVATOR

Номер: US20150131769A1
Автор: Larrion Javier
Принадлежит: SERBEX TECNOLOGIA Y VALORES, S.L.

The present invention relates to a nuclear power plant and safety system with fuse element and gravity elevator, the buildings of the power plant subjected to contamination being buried below sea level and under borated water basins, and having a safety system free of electrical and electronic components to act in the event of possible accidents comprising, among others, means for flooding the buildings of the power plant with thermal fuses and gravity elevators for operator evacuation in the event of an emergency. 126-. (canceled)27. A nuclear power plant , comprising at leasta containment building inside which a nuclear reactor is located,a power generation building inside which the turbines and other electricity-generating components are located, anda nuclear material building or warehouse for storing nuclear waste or nuclear fuel,all the aforementioned buildings being buried and, except the power generation building, connected by means of cooling pipes with at least one cooling water tank located above them and communicated with the sea and below sea level, such that the water falls due to gravity in the case of needing to cool or flood said buildings and pipes used for steam exhaust coming from at least the containment building and ending at the bottom of the water tank, and', 'means of valve and float systems for keeping the tank at a constant water level., 'further comprising28. The power plant according to claim 27 , wherein the reactor containment building internally has a reactor vessel inside which there is arranged a core vessel which houses the core claim 27 , at least one of the walls of at least the containment building or of the vessels comprising a fuse element connected with a cooling pipe.29. The power plant according to claim 27 , comprising pipes for the exit of steam connecting the inside of the vessels through a fuse element with the cooling water tank and pipes for the exit of steam connecting through security valves the inside of the vessels ...

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25-08-2022 дата публикации

SYSTEMS AND METHODS FOR CONTINUALLY MONITORING THE CONDITION OF NUCLEAR REACTOR INTERNALS

Номер: US20220270772A1
Принадлежит: WESTINGHOUSE ELECTRIC COMPANY LLC

A system configured to monitor the structural health of reactor vessel internals of a nuclear reactor is disclosed herein. The system includes a memory configured to store historical information associated with past performance of the nuclear reactor, and an anomaly detection subsystem including a control circuit configured to receive a signal from a sensor. The anomaly detection subsystem is configured to determine, via the control circuit, a characteristic of a vibrational response of the reactor vessel internals based, at least in part, on the signal; access, via the control circuit, the historical information stored in the memory; compare, via the control circuit, the determined characteristic to the historical information stored in the memory; and determine, via the control circuit, a condition of the reactor vessel internals based, at least in part, on the comparison of the determined characteristic and the historical information. 1. A system configured to monitor the structural health of reactor vessel internals within a nuclear reactor , the system comprising:a memory configured to store historical information associated with past performance of the nuclear reactor; determine, via the control circuit, a characteristic of a vibrational response of the reactor vessel internals based, at least in part, on the received signal;', 'access, via the control circuit, the historical information stored in the memory;', 'compare, via the control circuit, the determined characteristic to the historical information stored in the memory; and', 'determine, via the control circuit, a condition of the reactor vessel internals based, at least in part, on the comparison of the determined characteristic and the historical information., 'an anomaly detection subsystem configured to be communicably coupled to the memory and a sensor, wherein the anomaly detection subsystem comprises a control circuit configured to receive a signal from the sensor, wherein the signal is associated ...

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11-05-2017 дата публикации

PASSIVE DEPRESSURIZATION SYSTEM FOR PRESSURIZED CONTAINERS

Номер: US20170133111A1
Автор: Laborda Rami Arnaldo
Принадлежит:

The depressurization system comprises a main valve () provided with a pneumatic actuator with an opening spring () which can be connected at one side to a pressurized container () housing a gas inside it and at the other side to the atmosphere, defining this opening spring () a predetermined mechanical pressure, so when the pressure inside the pressurized container () is bigger than the predetermined mechanical pressure, the main valve () remains closed, and when the pressure inside the pressurized container () is lower than the predetermined mechanical pressure, the main valve () opens, allowing the pressurized gas from container () be discharged into the atmosphere. 1. Depressurization system for pressurized containers , characterized in that it comprises a main valve provided with a pneumatic actuator with an opening spring which can be connected at one side to a pressurized container housing a gas inside it and at the other side to the atmosphere , defining this opening spring a predetermined mechanical pressure , so when the pressure inside the pressurized container is bigger than the predetermined mechanical pressure , the main valve remains closed , and when the pressure inside the pressurized container is lower than the predetermined mechanical pressure , the main valve opens , allowing the pressurized gas from container be discharged into the atmosphere.2. Depressurization system for pressurized containers according to claim 1 , also comprising at least one solenoid valve connected between the pressure vessel and the main valve.3. Depressurization system for pressurized containers according to claim 1 , also comprising at least one manual valve connected between the pressure vessel and the main valve.4. Depressurization system for pressurized containers according to claim 1 , also comprising a pneumatic line which can connect the output of the main valve with a pneumatic motor of an isolation valve connected to an output of the pressure vessel.5. ...

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07-08-2014 дата публикации

ALTERNATE PASSIVE SPENT FUEL POOL COOLING SYSTEMS AND METHODS

Номер: US20140219411A1
Принадлежит: WESTINGHOUSE ELECTRIC COMPANY LLC

The present invention relates to passive cooling systems and methods for cooling a spent fuel pool in a nuclear power plant in the absence of onsite and offsite power, e.g., in a station blackout event. The systems include a gap formed along the periphery of the spent fuel pool, a heat sink, one or more thermal conductive members, a water supply system for delivering water to at least partially fill the gap and conduct heat generated from the spent fuel pool through the gap to at least one thermal conductive member for transporting heat to the heat sink, and a thermal switch mechanism for activating and deactivating the water supply system. 1. A passive cooling system for a spent fuel pool in a nuclear power plant , to provide cooling in the absence of onsite and offsite power , the system comprising:a gap having a first side and a second side formed at least partially along a periphery of the spent fuel pool;a heat sink;one or more thermal conductive members having a first end connected to the second side of the gap and a second end connected to the heat sink, said one or more members structured to transport heat from the gap to the heat sink; a water source; and', 'a discharge header having a first end connected to the water source and a second end connected to the gap; and, 'a water supply system, comprisinga thermal switch mechanism having an activate position and a deactivate position, structured to deliver water from the water system, into the gap when in the activate position and structured to inhibit the release of water from the water system and into the gap when in the deactivate position.wherein when the thermal switch mechanism is in the activate position, heat generated in the spent fuel pool is conducted to the gap, transported through the one or more conductive members and to the heat sink.2. The passive cooling system of further comprising one or more conductive cooling fins attached to the second end of the one or more members to enhance transport ...

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09-05-2019 дата публикации

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

Номер: US20190139658A1
Принадлежит:

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

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30-04-2020 дата публикации

PASSIVE ELECTRICAL COMPONENT FOR SAFETY SYSTEM SHUTDOWN USING GAUSS' LAW OF MAGNETISM

Номер: US20200135354A1
Принадлежит:

An electro-technical device, includes an input electrical connection supplied with an input signal and electrically isolated from an output electrical connection. A bar magnet is pivotally mounted on a pedicel between the input electrical connection and the output electrical connection. A pair of coils disposed on opposite sides of the bar magnet and each being supplied with an electronic signal from a sensor, the bar magnet being responsive to an electromagnetic filed generated by the pair of coils to cause the bar magnet to contact the input electrical connection and the output electrical connection and complete a circuit and send out a control signal. 1. An electro-technical device , comprising:an input electrical connection supplied with an input signal and electrically isolated from an output electrical connection; anda bar magnet pivotally mounted on a pedicel between the input electrical connection and the output electrical connection; andat least one coil disposed adjacent to the bar magnet and being supplied with an electronic signal from a sensor, the bar magnet being responsive to an electromagnetic field generated by the at least one coil to cause the bar magnet to contact the input electrical connection and the output electrical connection and complete a circuit and send out a control signal.2. The electro-technical device according to claim 1 , further comprising a housing for enclosing the bar magnet claim 1 , the pedicel claim 1 , the input electrical connection and the output electrical connection.3. The electro-technical device according to claim 1 , wherein the at least one coil includes a pair of coils including a first coil connected to one of a temperature sensor claim 1 , a pressure sensor and a flow sensor and a second coil connected to a second one of a temperature sensor claim 1 , a pressure sensor and a flow sensor.4. A fault detection system for a nuclear reactor claim 1 , comprising: an input electrical connection supplied with an input ...

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11-09-2014 дата публикации

Alternative air supply and exhaust port for air-operated valve

Номер: US20140254738A1
Принадлежит: Hitachi GE Nuclear Energy Ltd

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.

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06-06-2019 дата публикации

SAFETY SYSTEM

Номер: US20190172596A1
Автор: IRELAND Benjamin J.
Принадлежит: ROLLS-ROYCE POWER ENGINEERING PLC

A coolant injection system for a nuclear power generation system includes the coolant injection system, and method of operation of the coolant injection system. The nuclear power generation system includes a reactor pressure vessel having a reactor core, a pressuriser in fluid communication with the reactor pressure vessel, and the injection system, which comprises a make-up tank having a tank inlet and a tank outlet. The injection system has an operating condition, and a fault response condition, and is configured to switch between these conditions when coolant level in the pressuriser drops below a threshold level. In the operating condition, the tank outlet is isolated from the reactor pressure vessel such that coolant is retained in the make-up tank, and the tank inlet is in fluid communication with the reactor pressure vessel and the pressuriser. 1. A coolant injection system for a nuclear power generation system including a reactor pressure vessel having a reactor core , and a pressuriser in fluid communication with the reactor pressure vessel , the injection system comprising a make-up tank having a tank inlet and a tank outlet , wherein the injection system is configured to switch between an operating condition and a fault response condition when the coolant level in the pressuriser drops below a threshold level , and wherein:in the operating condition, the tank outlet is isolated from the reactor pressure vessel such that coolant is retained in the make-up tank, and the tank inlet is in fluid communication with the reactor pressure vessel and the pressuriser; andin the fault response condition, the tank inlet is isolated from the reactor pressure vessel, and the tank outlet is in fluid communication with the reactor pressure vessel such that coolant from the make-up tank can flow into the reactor pressure vessel to provide cooling of the reactor core.2. The coolant injection system according to claim 1 , wherein in the fault response condition claim 1 , the ...

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05-07-2018 дата публикации

PLANT OPERATION SYSTEM AND PLANT OPERATION METHOD

Номер: US20180190403A1
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

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 ...

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05-07-2018 дата публикации

RADIOACTIVE IODINE ADSORBENT, AND METHOD FOR TREATING RADIOACTIVE IODINE

Номер: US20180190404A1
Принадлежит:

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

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20-06-2019 дата публикации

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

Номер: US20190189299A1
Принадлежит:

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

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09-10-2014 дата публикации

Underwater electricity production module

Номер: US20140301524A1
Автор: Geoffrey Haratyk
Принадлежит: DCNS SA

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

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26-07-2018 дата публикации

Reactor protection-processor-to-reactor-trip breaker interface and method for operating the same

Номер: US20180211734A1
Принадлежит: Mitsubishi Electric Power Products Inc

A method is provided of operating a reactor trip system, including: determining whether an automated shutdown of the nuclear plant has been demanded; energizing an interface relay to de-energize an undervoltage coil of a circuit breaker when the automated shutdown has been demanded; energizing an interface relay to energize a shunt trip coil of the first circuit breaker when the automated shutdown has been demanded; de-energizing the undervoltage coil when the manual shutdown has been demanded; energizing the shunt trip coil when the manual shutdown has been demanded; performing a check of a cyclical execution of processors in first and second divisions using separate watchdog timers in each division and determining whether watchdog timer signals have been de-energized in the first and second divisions; and de-energizing the undervoltage coil if the first and second watchdog timer signals have both been actuated.

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25-06-2020 дата публикации

PASSIVE TECHNIQUES FOR LONG-TERM REACTOR COOLING

Номер: US20200203029A1
Принадлежит:

In a pressurized water reactor (PWR), emergency core cooling (ECC) responds to depressurization due to a vessel penetration break at the top of the pressure vessel by draining water from a body of water through an injection line into the pressure vessel. A barrier operates concurrently with the ECC to suppress flow of liquid water from the pressure vessel out the vessel penetration break. The barrier may comprise one or more of: (1) an injection line extension passing through the central riser to drain water into the central riser; (2) openings in a lower portion of a central riser to shunt some upward flow from the central riser into a lower portion of the downcomer annulus; and (3) a surge line providing fluid communication between a pressurizer volume at the top of the pressure vessel and the remainder of the pressure vessel which directs water outboard toward the downcomer annulus. 1. An apparatus comprising:a pressurized water reactor (PWR) comprising a pressure vessel containing a nuclear reactor core comprising fissile material;a central riser disposed inside the pressure vessel and defining a coolant circulation path in which coolant water heated by the nuclear reactor core flows upward inside the central riser, exits a top opening of the central riser, and flows downward in a downcomer annulus defined between the central riser and the pressure vessel to return to the nuclear reactor core;a radiological containment structure inside of which the PWR is disposed;an emergency core cooling system configured to drain water from a body of water through an injection line into the pressure vessel in response to a vessel penetration break at the top of the pressure vessel that depressurizes the pressure vessel; andan extension of the injection line disposed inside the pressure vessel and passing through the central riser, the extension configured to operate concurrently with the emergency core cooling system to suppress flow of liquid water from the pressure vessel ...

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02-08-2018 дата публикации

HYBRID SAFETY INJECTION TANK SYSTEM PRESSURIZED WITH SAFETY VALVE OF PRESSURIZER

Номер: US20180218796A1
Принадлежит: KOREA ATOMIC ENERGY RESEARCH INSTITUTE

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

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19-08-2021 дата публикации

The Use of a Stirling Engine to Provide Emergency Heat Removal to the Containment Environment of a Nuclear Reactor Building

Номер: US20210257115A1
Принадлежит: Individual

A Stirling engine provides a means to use the thermal energy in the sealed containment environment of a nuclear reactor building to provide emergency cooling. Acting as the prime mover in a coupled heat exchanger system, a Stirling engine could develop fluid flow thereby resulting in forced convection vice natural circulation and would not rely on an external power source during an unusual accident event where no electric power is available.

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27-08-2015 дата публикации

COOLING WATER SUPPLY TANK HAVING HEAT MIXING PREVENTION FUNCTION AND PASSIVE HIGH-PRESSURE SAFETY INJECTION SYSTEM AND METHOD USING THE SAME

Номер: US20150243384A1
Автор: KIM Kihwan, Kwon Tae-Soon
Принадлежит: KOREA ATOMIC ENERGY RESEARCH INSTITUTE

A passive high-pressure safety injection system includes a compressor which generates high-temperature and high-pressure steam, a cooling water supply tank which supplies cooling water using the compressed steam, a nuclear reactor which receives the cooling water so that the nuclear reactor is maintained in a cooled state, and an internal circulation prevention structure which is provided in the cooling water supply tank and prevents the cooling water from circulating in the cooling water supply tank. 1. A cooling water supply tank having a heat mixing prevention function , the cooling water supply tank supplying cooling water using steam compressed by a compressor , and comprisingan internal circulation prevention structure for preventing the cooling water from circulating in the cooling water supply tank.2. The cooling water supply tank as set forth in claim 1 , wherein the internal circulation prevention structure comprisesat least one cooling water guide preventing the cooling water from moving in the cooling water supply tank, the cooling water guide partitioning at least a portion of an internal space of the cooling water supply tank into a plurality of areas.3. The cooling water supply tank as set forth in claim 2 , wherein the cooling water guide comprises at least one vertical partition.4. The cooling water supply tank as set forth in claim 2 , wherein a height of the cooling water guide is greater than a level of cooling water in the cooling water supply tank when the cooling water supply tank is in a standby state before being operated.5. The cooling water supply tank as set forth in claim 2 , wherein the cooling water guide contains non-corrosive metal.6. The cooling water supply tank as set forth in claim 1 , wherein a cross-section of the internal circulation prevention structure has a shape selected from among a shape having a plurality of polygons claim 1 , a circular shape and a spiral shape.7. A passive high-pressure safety injection system having ...

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25-08-2016 дата публикации

RADIOACTIVE IODINE ADSORBENT, AND METHOD FOR TREATING RADIOACTIVE IODINE

Номер: US20160247588A1
Принадлежит:

Provided is a radioactive iodine adsorbent capable of adsorbing radioactive iodine more effectively than in the conventional art, and removing hydrogen, which is a factor in nuclear reactor accidents. A granulated radioactive iodine adsorbent of zeolite X, wherein ion exchange sites of the zeolite X are substituted with silver so that a size of minute pores of the zeolite X is suited to a size of a hydrogen molecule, and the radioactive iodine adsorbent has a silver content of 36 wt % or more when dried, a particle size of 10×20 mesh, a hardness of 94% or more, and a water content of 12 wt % or less when dried at 150° C. for 3 h and thereby reduced in weight. 1. A granulated radioactive iodine adsorbent of zeolite X , whereinion exchange sites of the zeolite X are substituted with silver so that a size of minute pores of the zeolite X is suited to a size of a hydrogen molecule, andthe radioactive iodine adsorbent has a silver content of 36 wt % or more when dried, a particle size of 10×20 mesh, and a water content of 12 wt % or less when dried at 150° C. for 3 h and thereby reduced in weight.2. The radioactive iodine adsorbent of claim 1 , wherein97% or more of the ion exchange sites of the zeolite X are substituted with silver.3. The radioactive iodine adsorbent of claim 1 , whereinthe ion exchange sites of the zeolite X are not substituted with any material other than silver.4. A method for treating radioactive iodine contained in steam discharged from a nuclear power facility claim 1 , comprising:{'claim-ref': {'@idref': 'CLM-00001', 'claim 1'}, 'a filling step of filling an air-permeable container with the radioactive iodine adsorbent of ; and'}a flow passing step of passing a flow of the steam discharged from the nuclear power facility, through the container filled with the radioactive iodine adsorbent.5. The method of claim 4 , whereinthe steam discharged from the nuclear power facility contains hydrogen molecules.6. The method of claim 4 , whereinthe steam ...

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23-12-2021 дата публикации

ABNORMAL DRIVING STATE DETERMINATION DEVICE AND METHOD USING NEURAL NETWORK MODEL

Номер: US20210397950A1
Принадлежит:

The present invention relates to an abnormal driving state determination device using a neural network model, and the device can comprise: an abnormal driving state data generation unit for generating abnormal driving state data on the basis of information related to an abnormal driving state; an abnormal driving state learning unit which receives the abnormal driving state data so as to visualize the abnormal driving state data through a visualization algorithm, thereby allowing the abnormal driving state data to be learned by a neural network model; and an abnormal driving state determination unit including the neural network model for determining an abnormal state on the basis of the learned abnormal driving state data. 1. An abnormal driving state determination device using a neural network model , the device comprising:an abnormal driving state data generation unit configured to generate abnormal driving state data based on information on an abnormal driving state;an abnormal driving state learning unit configured to receive the abnormal driving state data so as to visualize the abnormal driving state data through a visualization algorithm, thereby allowing the abnormal driving state data to be learned by a neural network model; andan abnormal driving state determination unit provided with a neural network model for determining an abnormal state on the basis of the learned abnormal driving state data.2. The device of claim 1 ,wherein the abnormal driving state learning unit is provided with a first visualization arrangement unit,wherein the first visualization arrangement unit arranges operating variables based on physical locations, thereby collecting and arranging operating variables corresponding to locations where an abnormal driving state occurs.3. The device of claim 1 ,wherein the abnormal driving state learning unit is provided with a second visualization arrangement unit,wherein the second visualization arrangement unit preferentially arranges ...

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02-11-2017 дата публикации

VENTILATION SYSTEM AND ASSOCIATED OPERATING METHOD FOR USE DURING A SERIOUS ACCIDENT IN A NUCLEAR INSTALLATION

Номер: US20170312679A1
Автор: Hill Axel
Принадлежит:

A ventilation system for an operating space accessible to operators in a nuclear installation is intended to allow a supply of decontaminated fresh air for a period of a few hours in the event of serious accidents involving the release of radioactive activity. In particular, the component of radioactive inert gases in the fresh air supplied to the operating space should be as small as possible. For this purpose, the ventilation system has a supply air line that is guided from an external inlet to the operating space, and into which a first fan and a first inert gas adsorber column are connected. An exhaust air line is guided from the operating space to an external outlet, and into which a second fan and a second inert gas adsorber column are connected. A switching device is provided for interchanging the roles of the first and second inert gas adsorber columns. 1. A ventilation system for an operating space accessible to operators in a nuclear installation , the ventilation system comprising:an external inlet;a supply air line guided from said external inlet to the operating space;a first fan connected in said supply air line;a first inert gas adsorber column connected in said supply air line;an external outlet;an exhaust air line guided from the operating space to said external outlet;a second fan connected in said exhaust air line;a second inert gas adsorber column connected in said exhaust air line;a switching device for interchanging roles of said first and second inert gas adsorber columns;a circulating-air line;{'sub': '2', 'a COadsorber column connected in said circulating-air line; and'}a circulating-air fan connected in said circulating-air line, said circulating-air line leading away from and back to the operating space, said second fan being able to be connected into said circulating-air line as said circulating-air fan.2. The ventilation system according to claim 1 , wherein said circulating-air line having an inlet side connected to said exhaust air ...

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01-11-2018 дата публикации

Computer Program for Simulating Nuclear Fuels and Nuclear Fuel Simulation Method Applied Thereto

Номер: US20180315515A1
Принадлежит: KOREA HYDRO & NUCLEAR POWER CO., LTD.

A nuclear fuel simulation method is provided that includes: step (a) for receiving an input of data on the order in which nuclear fuels are moved; step (b) for extracting, from the data, information on nuclear fuels, the coordinates of locations from which nuclear fuels are unloaded, and the coordinates of locations into which nuclear fuels are loaded; and step (c) for simulating the information extracted in step (b) according to a flowchart of the data. The present invention has an advantage in that it is possible to accurately and quickly verify all fuel movement works requiring the unloading and loading of nuclear fuels by receiving an input of a huge amount of data on the order in which nuclear fuels are moved and systematically verify an error that may occur during a simulation according to a flowchart, which enables the workload of about three man-days, required per cycle for each reactor, to be done in three man-hours, thereby achieving a significant reduction in working time. 1. A nuclear fuel simulation method comprising:(a) receiving an input of data regarding movement order of nuclear fuels;(b) extracting, from the data, information on nuclear fuels, the coordinates of a region from which nuclear fuels are unloaded, and the coordinates of a region to which nuclear fuels are loaded; and(c) simulating the information extracted in step (b) according to a flowchart of the data.2. The nuclear fuel simulation method of claim 1 , whereinstep (c) includes a first verification step of detecting an unloading error of nuclear fuels during the simulation; and a second verification step of detecting a loading error of nuclear fuels during the simulation.3. The nuclear fuel simulation method of claim 1 , whereinstep (c) further includes a third verification step of detecting an error that may occur during movement of nuclear cells during the simulation,wherein the third verification step include detecting whether a reactor refueling crane and a wall surface interfere ...

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17-10-2019 дата публикации

Method For Characterising One Or More Faults In A System

Номер: US20190318837A1
Принадлежит:

The invention relates to a method for characterising one or more faults in a system grouping together a plurality of internal physical quantities and delimited by a plurality of boundary physical quantities, the system being modelled by a healthy model establishing relationships linking said internal physical quantities with one another and with the boundary physical quantities in the absence of a fault, a fault being defined as an alteration in the relationships linking said internal physical quantities with one another and with the boundary physical quantities with respect to the healthy model, wherein a fault is characterised by counting a number of iterations having involved said fault in a series of iterations involving a fault matrix. 1. Method for characterising one or more faults in a system grouping together a plurality of internal physical quantities and delimited by a plurality of boundary physical quantities ,the system being modelled by a healthy model establishing relationships linking said internal physical quantities with one another and with the boundary physical quantities in the absence of a fault, a fault being defined as an alteration in the relationships linking said internal physical quantities with one another and with the boundary physical quantities with respect to the healthy model,said system being provided with a plurality of sensors measuring values of internal physical quantities and external physical quantities, determining a vector of the measured measurements of a set of internal physical quantities by reading the values of internal physical quantities measured by the sensors,', 'determining a vector of actual symptoms by the difference between the vector of measured measurements and a vector of expected measurements, said vector of expected measurements grouping together values of internal physical quantities obtained by simulation of the healthy model from the boundary physical quantities,', 'wherein the method further includes ...

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15-11-2018 дата публикации

METHOD FOR MANAGING STOPPAGE OF A PRESSURISED-WATER NUCLEAR REACTOR

Номер: US20180330835A1
Принадлежит:

Disclosed is a method for managing stoppage of a pressurized-water nuclear reactor integrated into a submerged module for producing electrical power, in case of detection of a primary/secondary leak in a steam generator equipped with a safety valve, which generator is connected to the reactor and associated with a standby cooling unit. The method includes: detecting a primary/secondary leak in the steam generator; automatically stopping the reactor and isolating the broken steam generator; bringing the corresponding standby cooling unit online; monitoring the primary pressure and, once the primary pressure has passed below the set pressure of the safety valves of the steam generators, isolating the standby cooling unit of the broken steam generator; and continuing to passively cool the reactor with the remaining steam generators and cooling unit. 1. A method for managing stoppage of a pressurized-water nuclear reactor , integrated into a submerged module for producing electrical power , if a primary and/or secondary leak is detected in a steam generator provided with a safety valve , said generator being connected to the reactor and associated with a standby cooling means , the method comprising:1) detecting (1) a primary/secondary leak of the steam generator,2) automatically stopping (2) the reactor and isolating the damaged steam generators,3) bringing (3) the corresponding standby cooling means online,4) monitoring (4) the primary pressure,5) once the primary pressure has passed below the set pressure of the safety valves of the steam generator, isolating (5) the standby cooling means of the broken steam generator, and6) continuing (6) to passively cool the reactor with the remaining steam generators and cooling means.2. The method for managing stoppage of a pressurized-water nuclear reactor according to claim 1 , wherein the detection (1) of a primary/secondary leak is done by detecting one or several of the following phenomena:1) increased activity of the ...

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15-11-2018 дата публикации

DIGITAL PROTECTION SYSTEM FOR NUCLEAR POWER PLANT

Номер: US20180330837A1
Принадлежит:

A digital protection system includes a process protection system having at least two channels and a reactor protection system having at least two trains. The process protection system includes, in one channel, first and second comparative logic controllers of different types that are mutually independent of each other and that respectively receive process variables as inputs and each outputting comparison logic results. The reactor protection system includes, in one train, first and second concurrent logic controllers of different types that are mutually independent from each other and that respectively receive the comparison logic results as inputs and each outputting concurrent logic results. The reactor protection system includes initiation circuits, each circuit including a plurality of relays connected in series and a plurality of relays connected in parallel. One series-connected relay is controlled by one of the two different concurrent logic results, and one parallel-connected relay is controlled by the other. 1. A digital protection system having at least two channels and at least two trains , the system comprising:a process protection system having, in one channel, first and second comparative logic controllers of different types that are mutually independent of each other, the first and second comparative logic controllers each receiving process variables as inputs and each outputting comparison logic results; anda reactor protection system having, in one train, first and second concurrent logic controllers of different types that are mutually independent from each other, the first and second concurrent controllers each receiving the comparison logic results as inputs and each outputting concurrent logic results, the reactor protection system including at least two initiation circuits, each initiation circuit including a series circuit in which a plurality of relays are connected in series and a parallel circuit in which a plurality of relays are ...

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08-12-2016 дата публикации

Hydraulic control unit and method of shutting down a nuclear reactor using the same

Номер: US20160358676A1
Принадлежит: GE HITACHI NUCLEAR ENERGY AMERICAS LLC

A method of shutting down a nuclear reactor may include compressing a scram gas that is in fluid communication with a scram accumulator. The scram accumulator defines a chamber therein and contains bellows within the chamber. The bellows are configured to hold a scram liquid in isolation of the scram gas. The scram gas exerts a compressive force on the bellows in a form of stored energy. The method may additionally include releasing the stored energy in response to a scram signal such that the scram gas expands into the chamber of the scram accumulator to compress the bellows and expel the scram liquid from the scram accumulator to insert control rods into a core of the nuclear reactor.

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17-12-2015 дата публикации

SOLAR NUCLEAR FUSION DEVELOPMENT

Номер: US20150364951A1
Автор: Crawford Kendal Marie
Принадлежит:

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

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06-12-2018 дата публикации

APPARATUS AND METHOD OF EVALUATING RESPONSE TIME OF PLANT PROTECTION SYSTEM

Номер: US20180350475A1
Автор: LEE Chang Jae, YUN Jae Hee
Принадлежит:

Provided is an apparatus for evaluating a response time of a plant protection system. The apparatus includes: a classification module classifying a design process related to a response time requirement of each channel performing a safety function into four operations of a safety analysis operation, a system design operation, a response time analysis operation, and a response time test operation; and an integrated evaluation module determining whether a response time evaluation is proper based on a time t derived in the safety analysis operation, a time t derived in the system design operation, a time t derived in the response time analysis operation, and a time t derived in the response time test operation. 1. An apparatus for evaluating a response time of a plant protection system , the apparatus comprising:{'b': '1', 'a safety analysis module measuring an analytical response time t that is a response time of a system performing a safety function;'}{'b': '2', 'a system design module measuring a designed response time t representing a total sum of individual response times allocated to each device constituting an instrumentation channel;'}{'b': '3', 'a response time analysis module measuring an estimated response time t representing a response time quantitatively analyzed in each device constituting the instrumentation channel; and'}{'b': '4', 'a response time test module dividing each device constituting the instrumentation channel into n regions and measuring a measured response time t representing a response time measured overlappingly between the n regions,'}{'b': 4', '3', '2', '1, 'wherein the apparatus further comprises an integrated evaluation module determining that a response time requirement is satisfied in each of the safety analysis module, the system design module, the response time analysis module, and the response time test module when t Подробнее

17-12-2020 дата публикации

PWR DECAY HEAT REMOVAL SYSTEM IN WHICH STEAM FROM THE PRESSURIZER DRIVES A TURBINE WHICH DRIVES A PUMP TO INJECT WATER INTO THE REACTOR PRESSURE VESSEL

Номер: US20200395136A1
Автор: GRAHAM Thomas G.
Принадлежит:

In conjunction with a pressurized water reactor (PWR) and a pressurizer configured to control pressure in the reactor pressure vessel, a decay heat removal system comprises a pressurized passive condenser, a turbine-driven pump connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and turbine may be mounted on a common shaft via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) and/or the pressurized passive condenser. A pressurizer power operated relief valve may control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer. 1. A method operating in conjunction with a pressurized water reactor (PWR) including a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel also containing primary coolant water , a pressurizer integral with or operatively connected with the reactor pressure vessel and configured to control pressure in the reactor pressure vessel , and a refueling water storage tank (RWST) , the method comprising responding to a loss of heat sinking of the PWR by operations including:driving a turbine using steam piped from the pressurizer; anddriving a pump using the turbine to suction water from the RWST into the reactor pressure vessel.2. The method of wherein the driving of the pump comprises providing a common shaft mechanically connecting the turbine and the pump whereby the driven turbine rotates the common shaft to drive the pump.3. The method of further comprising:discharging steam piped from the pressurizer into a pressurized passive condenser; andconnecting the suction side of the pump to both the RWST and the pressurized passive condenser ...

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06-02-2014 дата публикации

Containment protection system for a nuclear facility and associated operating method

Номер: WO2014019770A1
Автор: Axel Hill, Norbert Losch
Принадлежит: AREVA GMBH

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

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12-09-2018 дата публикации

비상전력 생산 시스템 및 이를 구비한 원전

Номер: KR101897985B1

본 발명은 내부에 비상냉각수가 저장되어 원전 사고 시 피동안전계통의 열침원으로 사용되고, 내부가 밀폐되며 기설정된 압력 이상으로 설계되는 비상냉각수저장부; 및 상기 비상냉각수저장부의 내부 압력이 기설정된 압력 이상으로 도달 시 상기 비상냉각수저장부로부터 공급되는 증기로 터빈을 회전시켜 비상전력을 생산하는 터빈발전기를 포함하는 비상전력 생산시스템을 제공한다. 이에 의하면, 피동안전계통의 고유기능을 안정적으로 유지하고 소형 터빈발전기와 같은 검증된 방법으로 효율적으로 전기를 생산할 수 있다.

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17-02-2016 дата публикации

Shutdown cooling facility and nuclear power plant having the same

Номер: KR101594440B1
Принадлежит: 한국원자력연구원

The present invention provides a shutdown cooling system including: a steam line connection part which is connected to a steam line to be supplied with cooling water through the steam line connected to an exit of a steam generator; a shut down cooling heat exchanger which is supplied with the cooling water flowing into the shut down cooling system through the steam line connection part, and cools the cooling water heated while circulating in a secondary flow path of the steam generator, and discharges the cooled water to a flow path of the heat exchanger; a shut down cooling pump which forms circulation floating of the cooling water circulating in the steam generator and the shut down cooling heat exchanger by being operated for shut down cooling of a nuclear reactor in case of normal shut down or an accident of the nuclear reactor after first cooling of a nuclear reactor coolant system; and a water supply pipe connection part which is connected to the heat exchange flow path and a water supply pipe in order to supply the cooling water cooled in the shut down cooling heat exchanger through the water supply pipe connected to an entrance of the steam generator.

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09-08-2022 дата публикации

核电厂机组事故工况监测方法和系统

Номер: CN114883021A

本发明公开了一种核电厂机组事故工况监测方法和系统,该方法包括:采集与典型事故工况相关的事故工况特征参数、机组重要安全信号和专设安全设施状态;对事故工况特征参数进行分析处理,筛选出不在预设阈值范围内的异常特征参数;利用多个逻辑计算单元对各种典型事故工况进行并行诊断;在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果。本发明实时监测事故工况特征参数、机组重要参数信号及专设安全设施状态,自动并行诊断机组各个始发事故或叠加事故,并通过人机交互界面予以显示,辅助操纵员对机组事故工况进行判断和处理。

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29-04-2015 дата публикации

核电站dcs可视化运维操作方法和系统

Номер: CN104571016A

本发明公开了一种核电站DCS可视化运维操作方法,该方法包括:建立核电站DCS基础数据库,所述基础数据库包括一次仪表子数据库、执行机构子数据库和运维知识库;通过可视化DCS运维界面获取核电站DCS信号的传输路径信息和电缆端接信息;根据所述传输路径信息和电缆端接信息查找运维关联文件;根据所述运维关联文件和所述基础数据库进行运维操作。本发明核电站DCS可视化运维操作方法是一种可实现准确、易操作的核电站DCS可视化运维操作处理技术。此外,本发明还公开了一种核电站DCS可视化运维操作系统。

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17-01-2013 дата публикации

Fault diagnosis system and method

Номер: KR101223786B1
Автор: 이수일
Принадлежит: 한국수력원자력 주식회사

고장 진단 결과를 필터링하여 우선 순위별로 제공하는 고장 진단 시스템 및 방법을 개시한다. 일 실시예로서, 고장 진단 시스템 및 방법은, 고장 진단 결과에 대한 상관 관계를 운전 전문가의 지식(고장 진단을 위한 로직 기반의 트리 구조, 패턴 분류기의 가중치 부여 등)과 통합하여 고장 진단을 처리해서 운전원에게 우선순위별로 고장 진단 결과를 제공하고, 우선순위별로 고장 진단 결과를 제공하여 고장 진단 결과가 운전원의 경험 지식과 융합해서 운전원이 고장에 대해 올바르게 대처한다. Disclosed are a failure diagnosis system and method for filtering failure diagnosis results and providing the same by priority. In one embodiment, the system and method for fault diagnosis process fault diagnosis by integrating the correlation of the fault diagnosis result with knowledge of a driving expert (logic-based tree structure for fault diagnosis, weighting of pattern classifier, etc.) It provides the operator with fault diagnosis results by priority and the fault diagnosis results by priority so that the fault diagnosis result is integrated with the operator's experience knowledge so that the operator can cope with the fault correctly.

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16-04-2014 дата публикации

Backup nuclear reactor auxiliary power using decay heat

Номер: CN103733267A
Принадлежит: Westinghouse Electric Corp

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

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02-09-2015 дата публикации

Air-water Combined Cooling Passive Feedwater Device And System

Номер: CN104882169A

本文公开了一种水-空气组合式被动给水冷却装置,包括:水冷却热交换器,其连接至安全壳建筑体内部以利用水冷却方法将蒸汽发生器的热冷却;冷却箱,其包括在其中的水冷却热交换器并存储冷凝由蒸汽发生器产生的主蒸汽的冷却水;连接至冷却箱的蒸发蒸汽管,由冷却箱中的水冷却热交换器产生的冷却水的蒸汽流入蒸发蒸汽管内;空气冷却热交换器,其连接至蒸发蒸汽管并冷却及液化流入蒸发蒸汽管内的蒸汽;以及冷凝水收集管,其用于对冷却箱重新填充由空气冷却热交换器液化的蒸汽。

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24-12-2014 дата публикации

Passive containment spray system

Номер: KR101473377B1
Принадлежит: 한국원자력연구원

본 발명은, 순수하게 자연적인 현상만을 이용하여 격납건물의 내부로 냉각수를 살수할 수 있는 피동격납건물살수계통 및 이를 구비하는 원전을 제안한다. 피동격납건물살수계통은, 원자로를 수용하는 격납건물과 통하도록 형성되어 상기 격납건물과 압력 평형을 유지하는 살수냉각수 저장부, 사고 발생시 상기 격납건물 내부의 압력 상승에 의해 상기 살수냉각수 저장부로부터 공급된 냉각수를 상기 격납건물 내부로 살수하도록 상기 격납건물 내부에 설치되는 살수배관, 및 상기 냉각수에 유로를 제공하도록 일단이 상기 살수냉각수 저장부의 내부에 삽입되고 사고 발생에 의해 상기 격납건물의 압력이 상승하여 관 내부로 냉각수의 유동이 형성되면 피동적으로 상기 살수배관에 냉각수를 공급하도록 타단이 상기 살수배관에 연결되는 역U자관을 포함한다. The present invention proposes a floating containment building sprinkler system capable of sprinkling cooling water into the interior of a containment building using purely natural phenomena and a nuclear power plant having the same. The passive containment building sprinkler system includes a sprinkler cooling water reservoir formed to communicate with a containment building containing a reactor and maintaining pressure balance with the containment building, A water spray pipe installed in the containment building to allow the cooling water to flow into the containment building; and a water supply pipe installed inside the water storage cooling water storage part to provide a flow path to the cooling water, And a reverse U-shaped pipe connected at the other end to the water spray pipe to supply cooling water to the water spray pipe passively when the flow of cooling water is formed inside the pipe.

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05-06-2018 дата публикации

PASSIVE PRESSURE REDUCTION SYSTEM FOR PRESSURE CAPACITIES IN NUCLEAR REACTORS

Номер: RU2016142333A
Принадлежит: Асвад Инт, С.Л.

РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2016 142 333 A (51) МПК G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ЗАЯВКА НА ИЗОБРЕТЕНИЕ (21)(22) Заявка: 2016142333, 05.05.2014 (71) Заявитель(и): АСВАД ИНТ, С.Л. (ES) Приоритет(ы): (22) Дата подачи заявки: 05.05.2014 (43) Дата публикации заявки: 05.06.2018 Бюл. № 16 (72) Автор(ы): ЛАБОРДА РАМИ Арнальдо (ES) R U (85) Дата начала рассмотрения заявки PCT на национальной фазе: 05.12.2016 (86) Заявка PCT: ES 2014/070383 (05.05.2014) WO 2015/169975 (12.11.2015) R U (54) ПАССИВНАЯ СИСТЕМА СНИЖЕНИЯ ДАВЛЕНИЯ ДЛЯ ЕМКОСТЕЙ ПОД ДАВЛЕНИЕМ В ЯДЕРНЫХ РЕАКТОРАХ (57) Формула изобретения 1. Система снижения давления для емкостей под давлением, отличающаяся тем, что она содержит емкость (1) под давлением, основной клапан (8), снабженный пневматическим приводом с раскрывающей пружиной (10), который соединен с одной стороны с емкостью (1) под давлением, содержащей газ внутри нее, и с другой стороны с окружающей средой, задавая открывающей пружине (10) заданное механическое давление, так что, когда давление внутри емкости (1) под давлением больше, чем указанное заранее заданное механическое давление пружины, главный клапан (8) остается закрытым, а когда давление внутри емкости (1) под давлением меньше, чем указанное заранее заданное механическое давление, главный клапан (8) открывается и остается открытым, обеспечивая выпуск сжатого газа из емкости (1) в окружающую среду. 2. Система снижения давления для емкостей под давлением по п. 1, включающая также по меньшей мере один электромагнитный клапан (7), присоединенный между емкостью (1) под давлением и основным клапаном (8). 3. Система снижения давления для емкостей под давлением по п. 1, включающая также по меньшей мере один ручной клапан (6а, 6b), присоединенный между емкостью (1) под давлением и основным клапаном (8). 4. Система снижения давления для емкостей под давлением по п. 1, включающая в себя также пневматическую линию (9), которая выполнена с ...

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15-10-2021 дата публикации

Control method and control system for steam turbine of nuclear power station

Номер: CN108877973B
Автор: 尹刚, 曾彬, 王旭峰

本发明属于核电技术领域,提供了一种核电站汽轮机控制方法及控制系统。在本发明中,通过实时检测蒸汽母管的压力,以获取蒸汽母管的压力实测值,并在蒸汽母管的压力实测值与蒸汽母管的压力基准值之间的差值的绝对值大于预设阈值时,调节蒸汽母管压力控制器的输出值,并获取汽轮机的实测功率、目标功率以及当前转速,进而根据汽轮机的实测功率、目标功率、当前转速以及蒸汽母管压力控制器的输出值,调节汽轮机的主调阀,以使汽轮机输出功率恒定。本发明核电站汽轮机控制方法可自动对汽轮机的主调阀进行调节,不需要人为操作,消除了人为操作带来的安全隐患,局限性低和适用性高。

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15-02-2021 дата публикации

Passive depressurisation system for pressurised receptacles in nuclear reactors

Номер: KR102215184B1
Принадлежит: 아스바드 아이엔티, 에스.엘.

가압 컨테이너를 위한 감압 시스템으로서, 가압 컨테이너(1)를 포함하고, 일 측이 내부에 가스를 수용하는 상기 가압 컨테이너(1)와 연결되고 반대 측이 대기와 연결된 개방 스프링(10)을 갖는 공압 액추에이터가 메인 밸브(8)에 제공되며, 상기 개방 스프링(10)을 소정의 기계적 압력으로 규정하여, 상기 가압 컨테이너(1) 내부의 압력이 상기 소정의 스프링 기계적 압력을 초과하는 경우, 상기 메인 밸브(8)가 잠긴 상태로 유지되고, 상기 가압 컨테이너(1) 내부의 압력이 상기 소정의 기계적 압력 미만인 경우, 상기 메인 밸브(8)가 열리고 그것이 개방을 유지하여, 가압된 가스를 방출하도록 하는 가압 컨테이너를 위한 감압 시스템이 제공된다. 상기 시스템은 어떠한 어떠한 외부적 파워 공급을 필요로 하지 않고, 사고상황, 심지어 전력을 완전 상실하는 사고상황에서도 적절하게 기능을 수행한다. A decompression system for a pressurized container, comprising a pressurized container 1, a pneumatic actuator having an open spring 10 connected to the pressurized container 1 on one side and connected to the atmosphere on the opposite side Is provided on the main valve 8, and when the pressure inside the pressurization container 1 exceeds the predetermined spring mechanical pressure by defining the open spring 10 as a predetermined mechanical pressure, the main valve ( 8) is maintained in a locked state, and when the pressure inside the pressurized container 1 is less than the predetermined mechanical pressure, the main valve 8 is opened and it is held open to release the pressurized gas A pressure reducing system is provided for The system does not require any external power supply and functions properly in an accident, even in an accident where power is completely lost.

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18-11-2011 дата публикации

An emergency diesel generator digital excitation system for nuclear power plant using digital triplication controller and duplication rectifier, method for changing operating mode thereof and method for rapidly starting thereof

Номер: KR101085175B1

본 발명은, 특수 입력 신호를 입력받아 발전기 단자전압, 유/무효 전력, 발전기 주파수, 계자 전압과 계자 전류 신호 등으로 가공해서 삼중화 제어기(40)의 입력 모듈에 일치되는 표준신호(0-10[Vdc])를 출력하되, 아날로그 방식으로 구성된 입출력 신호 처리부(40)와; The present invention receives a special input signal is processed into a generator terminal voltage, active / reactive power, generator frequency, field voltage and field current signal and the like standard signal (0-10) matching the input module of the triplex controller 40 [Vdc] and output, the input and output signal processing unit 40 configured in an analog manner; 세 개의 채널이 서로 동기되면서 두 개 채널까지 정지되어도 정상운전이 가능하도록 되고, 비상디젤발전기(GEN)를 운전하는데 필요한 운전 데이터를 디지털 입출력 모듈과 아날로그 입출력 모듈 등을 통해 인가받아 제1 및 제2계자차단기(52-1, 52-2)의 개폐를 지령하며, 상기 입출력 신호 처리부(40)로부터 출력되는 표준신호를 입력받아 제1 및 제2정류기(1-1,1-2)의 싸이리스터 점호신호를 송출해서 직류전원의 크기를 제어함으로써 비상디젤발전기 단자전압의 크기를 제어하는 삼중화 제어기(60); When three channels are synchronized with each other and two channels are stopped, normal operation is possible, and the operation data necessary for operating the emergency diesel generator (GEN) is received through the digital input / output module, the analog input / output module, and the like. The thyristor of the first and second rectifiers 1-1 and 1-2 is commanded to open and close the field circuit breakers 52-1 and 52-2, and receives a standard signal output from the input / output signal processor 40. A triple controller 60 for controlling the magnitude of the terminal voltage of the emergency diesel generator by transmitting a firing signal to control the magnitude of the DC power; 상기 삼중화 제어기(60)로부터 송출되는 싸이리스터 제어 신호를 인가받아 싸이리스터의 게이트 점호 신호로 최종 변환하는 제1 및 제2점호카드(11-1,11-2)와, 상기 제1 및 제2점호카드(11-1,11-2)로부터 출력되는 싸이리스터의 게이트 점호 신호에 의해 PPT로부터 입력받은 3상 교류전원을 직류전원으로 변환하는 제1 및 제2세미 컨버터(10-1,10-2)로 이루어져, 두 개의 채널로 이중화 구성되어 부하를 절반씩 부담하다가 한쪽에 고장이 발생하면 나머지 한쪽이 부하를 전담하도록 된 제1 및 제2정류기(1-1,1-2); First and second call cards 11-1 and 11-2 for receiving a thyristor control signal transmitted from the triplex controller 60 and finally converting the thyristor control signal to a gate ...

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29-07-2019 дата публикации

Device abnormality presensing method and system using thereof

Номер: KR102005138B1
Принадлежит: 한국수력원자력 주식회사

본 발명은 원자력 발전소의 기기 이상징후 사전감지 방법 및 그 시스템을 제공한다. 상기 기기 이상징후 사전감지 방법은 패턴학습장치에서 소정 기간의 과거 기기 감시 데이터로 패턴 학습을 하여 동일 기기별로 유사 패턴을 보이는 감시 변수들을 그룹화하는 단계와 예측값 계산 장치에서 상기 그룹화된 감시변수들의 패턴들을 이용하여 실시간으로 수신된 현재 감시 데이터의 예측값을 생성하는 단계를 포함한다. The present invention provides a method and system for detecting a device abnormality symptom of a nuclear power plant. The method for predicting abnormality of apparatus abnormality includes the steps of grouping surveillance variables that show a similar pattern for each device by performing pattern learning from past device surveillance data for a predetermined period in a pattern learning apparatus and comparing the patterns of the grouped surveillance variables And generating a predicted value of the current monitoring data received in real time using the measured data.

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23-01-2017 дата публикации

Method for nuclear power plant accident management guide considering the disabled main control room

Номер: KR101698334B1
Автор: 김형택, 이영승
Принадлежит: 한국수력원자력 주식회사

The present invention relates to a method for managing an accident in a nuclear power plant. The method for managing an accident in a nuclear power plant includes: step (a) of determining availability of a main control room when a serious accident of a nuclear power plant occurs; step (b) of moving to a second control room in which control of main equipment required to maintain a safety device of a nuclear reactor is permitted when the main control room is disabled; step (c) of confirming an entry condition of management of the serious accident using a monitored variable which can be confirmed in the second control room; and step (d) of performing a management guideline of the serious accident in the second control room. According to the present invention, the method for managing an accident in a nuclear power plant can perform the management guideline of the serious accident using the second control room having the equipment by being separated at predetermined intervals from the main control room in an extreme environment condition in which the main control room is disabled.

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22-04-2015 дата публикации

Containment protection system for a nuclear facility and associated operating method

Номер: CN104541331A
Принадлежит: AREVA NP GMBH

一种反应堆外壳保护系统(2),用于在具有猛烈释放氢气(H 2 )和蒸汽的严重核事故情况下处理核技术设施(6)、尤其是核电站的反应堆外壳(4)内的气氛,该反应堆外壳保护系统(2)应当能够以很大程度上被动的方式并且尽可能没有环境负担地将这种状态有效且快速地排除。为此目的,反应堆外壳保护系统(2)根据本发明具有包括管路系统(10,72,120,128)的、设置用于连接到反应堆外壳(4)上的、针对流体的从反应堆外壳(4)出来并且又回去的循环,更确切地说,至少带有以下按照流动而串联连接的部件:复合装置(20),用于将流体中包含的氢气(H 2 )与氧气(O 2 )复合为水蒸气(H 2 O);连接在复合装置(20)之后的冷凝装置(74),用于将流体中包含的蒸汽部分冷凝,该冷凝装置带有用于将冷凝物(94)从流体流中导出的装置;流体流的驱动装置(18,180),其中为了冷凝装置(74)的至少部分的再冷却而存在热交换器(96),该热交换器在输入侧通过输送管路(144)与液氮(N 2 )的存储容器(140)连接。

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03-08-2016 дата публикации

Containment filtered venting system used for nuclear power plant

Номер: CN105830167A
Принадлежит: FNCTECH

本发明涉及使用于核电站的过滤排放系统,更详细地,涉及使用于核反应堆的过滤排放系统,其包括:过滤排放容器,用于保管过滤排放系统的结构物;入口管,与上述过滤排放容器和核反应堆建筑物相连接;组合喷嘴,与入口管相连接,并浸泡于填满过滤排放容器的一部分的过滤溶液;旋风分离器,去除从上述组合喷嘴脱离,并与过滤溶液相混合的大小大的液滴及气溶胶的大部分之后,向金属纤维过滤器引导;金属纤维过滤器,与上述旋风分离器的上端部相连接,用于过滤残留液滴和气溶胶;分子筛,用于在经由金属纤维过滤器来过滤的废气中去除有机碘;以及出口管,用于连接过滤排放容器和堆栈部。

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09-01-2015 дата публикации

Cooling system of emergency coolling tank and nuclear power plant having the same

Номер: KR101480046B1
Принадлежит: 한국원자력연구원

본 발명은 원자로의 사고 발생시 시간의 경과에 따라 비상냉각탱크로 전달되는 열량이 달라지는 것에 대비하여 냉각수의 보충 없이 장기냉각이 가능한 비상냉각탱크 냉각설비 및 이를 구비하는 원전을 개시한다. 비상냉각탱크 냉각설비는, 내부에 냉각수를 수용하도록 형성되고 원자로의 사고 발생시 상기 원자로 또는 격납건물로부터 상기 냉각수에 열을 전달받는 비상냉각탱크, 대기중에서 작동하도록 상기 비상냉각탱크의 외부에 노출되게 설치되며 냉각수의 보충 없이 상기 비상냉각탱크의 작동을 지속적으로 유지하도록 상기 냉각수와 대기 사이에서 열교환하여 열을 외부로 방출하는 열교환장치, 및 상기 냉각수의 수위보다 높은 위치에 배치되도록 상기 비상냉각탱크의 상부에 설치되며 상기 비상냉각탱크의 냉각 용량을 초과하는 열적 부하의 전달시 상기 냉각수의 증발에 의해 형성된 증기의 일부를 외부로 방출하도록 기설정된 압력보다 높은 압력에서 외부 대기와의 압력차로부터 형성되는 상기 증기의 유동에 의해 개방되는 개폐부를 포함한다. The present invention discloses an emergency cooling tank cooling facility capable of long-term cooling without supplementing cooling water in response to a change in the amount of heat transferred to an emergency cooling tank with the passage of time in the event of an accident of a nuclear reactor, and a nuclear power plant having the emergency cooling tank cooling facility. The emergency cooling tank cooling system includes an emergency cooling tank that is formed to receive cooling water therein and receives heat from the reactor or containment to the cooling water when an accident occurs in the reactor, A heat exchanger for exchanging heat between the cooling water and the atmosphere so as to continuously maintain the operation of the emergency cooling tank without supplementing the cooling water to discharge heat to the outside; Which is formed from a pressure difference with the outside atmosphere at a pressure higher than a predetermined pressure so as to discharge a part of the steam formed by the evaporation of the cooling water to the outside when the thermal load exceeding the cooling capacity of the emergency cooling tank is transferred, And an opening / closing portion that is opened by the flow of the steam.

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20-09-2016 дата публикации

PROTECTIVE PROTECTIVE SHELL SYSTEM FOR NUCLEAR TECHNICAL INSTALLATION AND THE APPROPRIATE METHOD OF OPERATION

Номер: RU2015107005A
Принадлежит: Арефа Гмбх

1. Защитная система (2) защитной оболочки для обработки атмосферы, находящейся в защитной оболочке (4) ядерной технической установки (6), в частности атомной электростанции, при критичных авариях с массивным высвобождением водорода (Н) и пара, содержащая предусмотренную для соединения с защитной оболочкой (4) систему трубопроводов (10, 72, 120, 128), образующую контур циркуляции из защитной оболочки (4) и снова назад потока текучей среды по меньшей мере со следующими включенными последовательно по потоку компонентами:- рекомбинатор (20) для рекомбинации содержащегося в потоке текучей среды водорода (Н) с кислородом (О) в водяной пар (НО),- включенное после рекомбинатора (20) конденсационное устройство (74) для конденсации содержащихся в потоке текучей среды долей пара со средствами для отвода конденсата (94) из потока текучей среды,- приводные средства (18, 180) для потока текучей среды,при этом по меньшей мере для частичного обратного охлаждения конденсационного устройства (74) имеется теплообменник (96), который на стороне входа через подводящий трубопровод (144) соединен с резервуаром (140) для хранения действующего в качестве охлаждающего средства инертного газа, в частности жидкого азота (N).2. Защитная система (2) защитной оболочки по п. 1, в которой теплообменник (96) выполнен в виде испарителя инертного газа, в частности испарителя азота.3. Защитная система (2) защитной оболочки по п. 1, в которой теплообменник (96) на стороне выхода через питающий трубопровод (148, 160) соединен с защитной оболочкой (4), так что обеспечивается возможность применения подводимого для обратного охлаждения конденсационного устройства (74) инертного газа, предпочтительно азота (N), затем для инертизации защитной оболочки (4).4. Защитная РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2015 107 005 A (51) МПК G21C 9/06 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ЗАЯВКА НА ИЗОБРЕТЕНИЕ (21)(22) Заявка: 2015107005, 24.06.2013 (71) Заявитель(и): АРЕФА ГМБХ (DE) ...

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13-07-2017 дата публикации

Passive decompression system for pressurized containers in nuclear reactors.

Номер: JP2017519161A
Принадлежит: Asvad Int SL

加圧コンテナのための減圧システムであって、一方の側がそれの内部にガスを収容する前記加圧コンテナ(1)に接続され、他方の側が大気に接続された、開放ばね(10)を有する空気圧アクチュエータを備える主弁(8)を備えることにより特徴づけられ、前記開放ばね(10)は所定の機械圧力で既定され、前記加圧コンテナ(1)の内部の圧力が前記所定の機械圧力より大きい場合、前記主弁(8)は閉じられたままであり、前記加圧コンテナ(1)の内部の圧力が前記所定の機械圧力より小さい場合、前記主弁(8)が開き、加圧されたガスがコンテナ(1)から大気中へと放出されることを可能にする、減圧システム。それは、それらの主要な動作のために何の外部電源をも要求せず、たとえ電力の全喪失を伴おうとも、事故シナリオにおいてそれらの役割を適切に果たすことを可能にする。

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06-03-2015 дата публикации

Air-Water Combined Cooling Passive Feedwater Device and System

Номер: KR101499641B1
Принадлежит: 한국원자력연구원

The present invention relates to an air-water combined cooling passive feedwater device and to a system thereof. The air-water combined cooling passive feedwater device includes: a water cooling exchanger which cools the heat of a steam generator to be connected to the inside of a containing building by using a water cooling method; a cooling water tank for storing cooling water by condensing main steam generated in a steam generator; an evaporation steam pipe which is connected to the cooling water tank and in which a cooling water steam generated by the water cooling exchanger inside the cooling water tank is inserted; an air cooling exchanger which is connected to the evaporation steam pipe, cools, and liquefies the steam inserted into the evaporation steam pipe; and a condensate collection pipe which refills steam liquefied from the air cooling exchanger to the cooling water tank.

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25-05-2020 дата публикации

Underwater electricity generation module

Номер: KR102115043B1
Принадлежит: 나발 그룹

본 발명에 따른 수중 전기 생산 모듈은, 일체화되는 전기 생산 유닛을 형성하는 수단인 가늘고 긴 실린더형 박스(12) 형태의 수단을 포함하고, 전기 케이블(6)에 의해 외부의 전기 분배 스테이션(7)에 연결된 전기 생산 수단 (37)과 연관된, 핵 보일러(30)를 형성하는 수단을 포함하는 전력 생산 장치를 형성하는 상기 수단, 보일러 핵 형성 수단 (30)은 반응기의 안전 저수 탱크(20)를 형성하는 챔버와 관련된 반응기 컴파트먼트(18)의 건조 챔버(19)에 배치되고, 적어도 반경 벽(53)은 해양 환경과 열 교환 관계에 있고, 반응기 컨테이너(18)의 건조 챔버(19)는, 상기 건조 챔버(19)의 상부에 위치한 감압 밸브를 형성하는 수단(71)을 포함하는 감압 수단(70)에 의해 반응기의 안전 저수 챔버(20)에 연결되고, 감압 수단(70)은 건조 챔버(19)의 상부에 배치된 감압 밸브를 형성하는 수단(71)을 포함하고 저수 형성 챔버(20)의 하부에 위치한 버블러(72)를 형성하는 수단에 연결된다. The underwater electrical production module according to the present invention comprises means in the form of an elongated cylindrical box 12 which is a means for forming an integrated electrical production unit, and an external electrical distribution station 7 by means of an electrical cable 6. Said means for forming a power generating device comprising means for forming a nuclear boiler 30, associated with an electrical production means 37 connected to the boiler nucleation means 30, forming a safe reservoir tank 20 of the reactor Is disposed in the drying chamber 19 of the reactor compartment 18 associated with the chamber, and at least the radius wall 53 is in heat exchange relationship with the marine environment, and the drying chamber 19 of the reactor container 18 is: The drying chamber 19 is connected to the safety storage chamber 20 of the reactor by means of pressure reducing means 70 including means 71 for forming a pressure reducing valve located on top, and the pressure reducing means 70 is a drying chamber ( 19) means for forming a pressure reducing valve disposed on the top and connected to a means for forming a bubbler 72 located at the bottom of the reservoir chamber 20.

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15-02-2017 дата публикации

Method for protecting main pump in nuclear power plant shutdown condition

Номер: CN104347132B
Автор: 唐辉, 李盛杰, 温亮

本发明公开了一种核电厂停堆工况下保护主泵的方法,核电厂的安注系统中设置有通过隔离阀与一回路连接的安注箱,安注箱为下部充有含硼水、上部充满氮气的压力容器,本发明的方法是在停堆工况下,将安注系统中的一个安注箱卸压后,再将安注箱重新与一回路连接,利用安注箱的气水两相来稳定一回路的压力,以提高主泵运行的可靠性。与现有技术相比,本发明核电厂停堆工况下保护主泵的方法通过将隔离后不再执行安全功能的安注箱卸压后与一回路重新连接,保证一回路压力的稳定,在提高主泵运行可靠性的同时,不会对核电厂的安全性造成其他任何不利影响。

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07-08-2020 дата публикации

Nuclear power station primary loop hydrostatic test overpressure protection system

Номер: CN110136852B

本发明涉及一种核电站一回路水压试验超压保护系统,设置在一回路的水压最高点的稳压器、压力采集单元、控制单元、逻辑判断单元及超压保护单元;压力采集单元与稳压器连接、检测一回路最高点的水压试验压力值并输出一回路的水压信息;控制单元与压力采集单元连接、接收并检测水压信息并在水压信息达到预设条件时输出触发信号;逻辑判断单元与压力采集单元和控制单元连接、接收并根据触发信号启动并根据水压信息判断一回路是否超压,并在一回路超压时输出超压保护信号至超压保护单元;超压保护单元与逻辑判断单元连接、接收并根据超压保护信号执行超压保护。本发明提高试验安全水平、减少设备及人力投用、缩短信息传送时间、优化试验实施过程。

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29-07-2008 дата публикации

Digital Security System for Nuclear Power Plant

Номер: KR100848881B1
Принадлежит: 삼창기업 주식회사

본 발명은 발전소의 원자로 관련 장치들의 동작 상태를 나타내는 디지털 신호를 받아서 이상 상태가 발생하면 원자로를 안전하게 정지시키는 디지털 원자로 보호 시스템으로서, 디지털 신호를 받아서 처리하는 채널 4개를 포함하고, 상기 각 채널은 발전소의 원자로 관련 장치들의 동작 상태를 나타내는 디지털 신호를 받아서 처리하되, CPU와 운영체계(OS)를 사용하지 아니하고 각각 다른 디지털신호처리기(DSP)를 사용하여 구성한 바이스테이블 모듈, 동시논리 모듈, 원자로정지 개시모듈, 공학적안전설비 개시모듈을 포함하여 구성함으로써, 운영체계를 사용하는 디지털 원자로 보호 시스템에서 발생하는 소프트웨어 공통 유형 고장 발생을 방지하도록 한 것이다. 원자로, 트립신호, 바이스테이블모듈, 동시논리모듈, 원자로정지개시모듈, 공학적안전설비개시모듈 The present invention is a digital reactor protection system that receives a digital signal representing the operating state of the reactor-related devices of the power plant and safely shuts down the reactor when an abnormal condition occurs, and includes four channels for receiving and processing the digital signal. Receives and processes digital signals indicating the operating status of reactor-related devices in a power plant, but does not use a CPU and an operating system (OS), but uses a different digital signal processor (DSP). By including the start-up module and the engineering safety equipment start-up module, it is possible to prevent the occurrence of a common type of software failure in the digital reactor protection system using the operating system. Reactor, trip signal, vice table module, simultaneous logic module, reactor stop start module, engineering safety equipment start module

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29-08-2017 дата публикации

Stop cooling system and the nuclear facilities with the stopping cooling system

Номер: CN107112059A

本发明提供了一种停止冷却系统,包括:蒸汽管线连接部,其连接到蒸汽管线以通过连接到蒸汽发生器的出口的蒸汽管线接收冷却水;停止冷却换热器,用于接收通过蒸汽管线连接部进入停止冷却系统的冷却水并通过蒸汽发生器的次级通道循环冷却水并通过换热器通道排出该冷却水;停止冷却泵,其当核反应堆正常停止时在核反应堆冷却系统的初级冷却之后或当事故发生时启动以执行核反应堆的停止冷却,并用于形成在蒸汽发生器和停止冷却换热器之间循环的冷却水的循环流动;以及给水管连接部,其连接到换热器通道和与蒸汽发生器的入口连接的给水管,从而将在停止冷却换热器中冷却的冷却水通过该给水管供应到蒸汽发生器。

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13-10-2020 дата публикации

Main pump first sealing low differential pressure operation offline test method and system

Номер: CN110164570B

本发明公开了一种主泵一号密封低压差运行离线测试方法及系统,该方法包括:S1.静态下降压盘动试验;S2.密封升压试验;S3.降压试验;S4.升压试验。本发明的主泵一号密封低压差运行离线测试方法及系统中,通过静态下降压盘动试验,验证在在不同压差下的启动力矩来验证密封之间是否存在液膜;密封升压试验,验证正常工况下初始密封性能;降压试验,验证各个压差下包括低于最小压差下密封特性;升压试验,验证低于最小压差运行后密封性能能够恢复初始状态。能够完成各项试验,保证主泵承压边界密封完整性。降低了因事故发生后缺乏相关经验和理论分析而保守决策带来的机组非计划停堆抢修带来的巨大经济损失,提高了设备可用性的评价分析依据。

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07-09-2022 дата публикации

Remote monitoring of critical reactor parameters

Номер: KR102441539B1
Принадлежит: 뉴스케일 파워, 엘엘씨

백업 전력원, 무선 전송기 및 다양한 시스템 센서들을 포함하는 원자로 계장 시스템 및 방법이 원자로 계장 시스템으로의 정상 전력의 손실을 식별하고 그리고 백압 전력원으로부터 무선 전송기로 전력이 제공되게 하도록 구성되며, 이는 원자로 계장 센서들로부터의 데이터를 상기 무선 전송기를 경유하여 원격 위치로 전송하기 위한 것이다. A reactor instrumentation system and method comprising a backup power source, a wireless transmitter, and various system sensors are configured to identify a loss of normal power to the reactor instrumentation system and cause power to be provided from the back pressure power source to the wireless transmitter, the reactor instrumentation system comprising: for transmitting data from the sensors to a remote location via the wireless transmitter.

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15-12-2020 дата публикации

Method and system for pre-detecting signs of abnormality in nuclear power plant equipment including process for determining equipment importance and alarm validity

Номер: CN112085200A
Принадлежит: Korea Hydro and Nuclear Power Co Ltd

本发明提供一种用于对设备的异常迹象进行早期警报的方法及系统,所述早期警报包括对设备重要度和警报有效性的确定。所述用于对设备的异常迹象进行早期警报的方法包括:第一步骤:通过早期警报处理装置使用基于先前已由操作者分析的用于每个监控变量的重要度数据的加权值来确定设备监控信号值是否超出正常操作范围;第二步骤:当所述设备监控信号值超出所述正常操作范围时,通过所述早期警报处理装置生成警报;以及第三步骤:通过警报确定装置确定所生成的所述警报是否为有效警报,所述有效警报待进行警报分析和追踪。

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08-07-2019 дата публикации

Nuclear power plant safety system

Номер: KR101997638B1
Принадлежит: 두산중공업 주식회사

본 발명은 상호 독립된 서로 다른 종류의 제1 공정 보호 제어기 및 제2 공정 보호 제어기를 포함하고, 원자로 이상 여부 신호를 트레인으로 송신하는, 적어도 2 이상의 채널, 상호 독립된 서로 다른 종류의 제1 이상 판단 제어기 및 제2 이상 판단 제어기를 포함하고, 상기 원자로 이상 여부 신호에 따른 스위칭 제어 신호를 생성하여 스위칭 모듈에 송신하는, 적어도 2 이상의 트레인, 상기 스위칭 제어 신호에 따라, 원자로가 정상인 경우 원자로를 정상 운전하며 원자로가 이상인 경우 원자로를 정지하는, 스위칭 모듈을 포함하는 원자력 발전소 안전 계통 시스템에 관한 것이다. The present invention relates to a control system for a nuclear power plant, comprising at least two channels, a first abnormality controller of a different type, and a second abnormality controller And a second abnormality determination controller for generating a switching control signal according to the reactor abnormality signal and transmitting the switching control signal to the switching module. According to the switching control signal, when the reactor is normal, And a switching module for stopping the reactor when the reactor is abnormal.

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10-01-2017 дата публикации

Submerged energy production module

Номер: RU2607474C2
Принадлежит: Дснс

FIELD: nuclear engineering. SUBSTANCE: invention relates to underwater nuclear power plant. Module (12) in form of an elongated cylindrical box comprises an electricity production unit, including a nuclear boiler (30), associated with electricity production means (37), connected by electrical cables (6) to external electricity distribution station (7). Nuclear boiler (30) is placed in dry chamber (19) of reactor compartment (18), associated with chamber forming a safety water storage reservoir (20) of reactor. In chamber at least one radial wall (53) is in a heat exchange relationship with marine environment. Dry chamber (19) of reactor compartment (18) is connected with compartment (21) for placement of electricity production means, which comprises means (100) for supply of water to flood dry chamber (19). Means (100) are installed in its lower part and comprise water intake (101) for sea water, made in radial wall of module (12), pipeline between said water intake and dry chamber (19) of reactor compartment and valve (102) to flood said chamber. EFFECT: higher safety of power unit during accidents. 24 cl, 5 dwg РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 607 474 C2 (51) МПК G21C 9/016 (2006.01) G21C 15/18 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ФОРМУЛА (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ 2014133729, 18.01.2013 (24) Дата начала отсчета срока действия патента: 18.01.2013 (72) Автор(ы): АРАТИК Жоффрей (FR) (73) Патентообладатель(и): ДСНС (FR) Дата регистрации: (56) Список документов, цитированных в отчете о поиске: WO2011128581 A1, Приоритет(ы): (30) Конвенционный приоритет: 20.10.2011;RU2191321 C2, 20.10.2002;US5247553 A, 21.09.1993;US4302291 A1, 24.11.1981. 18.01.2012 FR 12 50493 (45) Опубликовано: 10.01.2017 Бюл. № 1 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 18.08.2014 (86) Заявка PCT: EP 2013/050946 (18.01.2013) (87) Публикация заявки PCT: 2 6 0 7 4 7 4 (43) Дата публикации заявки: 10.03.2016 ...

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28-06-2012 дата публикации

Reactor shutdown system

Номер: JP2012122907A

【課題】原子力施設の異常発生時において、多様な手段により、原子炉を停止させることができる原子炉停止装置を提供する。 【解決手段】原子炉と、燃料集合体に対し制御棒16を抜差し方向へ駆動可能な制御棒駆動装置17と、制御棒駆動装置17へ電力を供給可能な電源31と、制御棒駆動装置17と電源31との間に設けられた電力変換装置32と、を備えた原子力施設の異常発生時に、原子炉の核反応を停止させる原子炉停止装置47において、制御棒駆動装置17は、電力供給が遮断されると、制御棒16を燃料集合体に差し込んで原子炉の核反応を停止させ、電力変換装置32と制御棒駆動装置17との間に設けられた原子炉トリップ遮断器45と、原子炉トリップ遮断器45を制御して、制御棒駆動装置17への電力供給を遮断する安全保護系設備43と、電力変換装置32を制御して、制御棒駆動装置17への電力供給を遮断するCCF設備44と、を備えた。 【選択図】図2

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30-01-1984 дата публикации

Reactor Trip Device

Номер: KR840000037A
Автор: 엠. 쿠크 부르스

내용 없음

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28-10-2019 дата публикации

Integrated radiological emergency preparedness system with a function to control sheltering and evacuation based on actual geographical features and real-time traffic and meteorological information and forecast

Номер: KR102037204B1
Принадлежит: (주)뉴클리어엔지니어링

An integrated nuclear disaster prevention system of the present invention comprises: an atmospheric information processing unit detecting the radiation concentration of designated space for each time zone using atmospheric diffusion information; a road network information processing unit detecting a location for each time zone of an agent using traffic control information; an integration unit integrating the radiation concentration of the designated space for each time zone and the location for each time zone of the agent to calculate amount of being exposed to radiation for each agent and calculating amount of being exposed to radiation of the entire group for all agents using the calculated amount of being exposed to radiation for each agent; and an information processing unit notifying the agent based on a result calculated by the integration unit.

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29-03-2021 дата публикации

Breakdown type analysis system and method of digital equipment

Номер: KR102232876B1
Автор: 신대근, 예송해
Принадлежит: 한국수력원자력 주식회사

본 발명은 디지털 설비의 고장 분석 시스템 및 분석 방법에 있어서, 디지털 설비의 고장 발생 시 고장정보가 수동 또는 자동으로 입력되는 고장정보 입력부; 분류된 복수개의 고장 유형 중 상기 고장정보가 해당되는 고장 유형을 판단하는 고장 유형 판단부; 상기 고장 유형 판단부로부터 판단된 고장 유형 결과가 저장되는 저장부; 상기 고장정보가 상기 복수개의 고장 유형 중 어느 하나에도 해당되지 않는 경우, 상기 고장정보를 상기 고장 유형 판단부에 새로운 고장 유형으로 추가하는 삽입부; 및 상기 고장 유형 결과가 표시되는 표시부;를 포함하여, 발생된 다수의 고장사건을 고장 유형에 따라 분류함으로써 고장 이력의 저장 및 관리가 가능하고, 저장된 상기 고장 이력을 통해 새롭게 발생하는 고장사건의 고장 유형 및 고장 원인분석이 가능하여 디지털 시스템의 신뢰성을 증진시키고 이를 유지·보수에 이용할 수 있는 것을 특징으로 한다. The present invention provides a failure analysis system and analysis method of a digital facility, comprising: a failure information input unit for manually or automatically inputting failure information when a failure of a digital facility occurs; A failure type determination unit for determining a failure type to which the failure information corresponds among a plurality of classified failure types; A storage unit for storing a failure type result determined by the failure type determination unit; An insertion unit for adding the failure information as a new failure type to the failure type determination unit when the failure information does not correspond to any one of the plurality of failure types; And a display unit for displaying the failure type result; including, by classifying a plurality of generated failure events according to the failure type, it is possible to store and manage a failure history, and failure of a newly occurring failure event through the stored failure history. It is characterized in that it is possible to analyze the type and cause of failure, thereby improving the reliability of the digital system and using it for maintenance and repair.

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13-08-2021 дата публикации

Method for preventing overspeed tripping in full-flow test of auxiliary water supply steam-driven pump of nuclear power station

Номер: CN113257444A

本发明涉及核岛系统技术领域,提供了一种防止核电站辅助给水汽动泵全流量试验超速跳闸的方法,所述给水汽动泵用于向蒸汽发生器供水,所述给水汽动泵包括给水阀门、阀门开度调节结构和阀门复位控制结构,所述阀门复位控制结构用于复位给水阀门开度,所述方法包括:调节阀门开度调节结构将给水阀门的开度设置为100%开度,以使得给水汽动泵启动后给水阀门能够实现自动全开,且阀门复位控制结构启动后,给水阀门的开度在100%开度。给水汽动泵启动后,调节阀门开度调节结构以使得给水阀门的开度达到预设开度。该方法有效避免了给水阀门关闭过程中快速甩负荷导致汽轮机转速快速上升进而导致汽轮机跳闸的风险。

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11-05-2010 дата публикации

Method and system, for providing regional overpower protection of calandria using on-line

Номер: KR20100048563A
Принадлежит: 한국전력공사

가압 중수형 원자로의 노심에서 발생할 수 있는 국부지역의 출력과다(국부과출력)로 핵연료가 손상되는 것을 방지하기 위하여 온라인을 이용한 가압 중수형 원자로의 국부과출력보호를 제공하기 위한 방법 및 시스템이 개시된다. 본 발명에 따르면, 2초마다 원자로 노심상태에 대응하는 국부과출력보호를 위한 정지설정치를 계산하여 ROP 시스템에 제공하거나 계측기 교정값을 제공할 수 있다. 이를 위해 노심상태를 핵설계에 준하여 계산하되 노심상태를 반영하기위한 각종 측정자료를 경계조건으로 하는 핵설계 및 열수력 설계가 가능하며, 실측 정보를 사용하기 위한 국부과출력보호의 정지설정치 평가 방법을 제공한다. 아울러 디지털형식으로 계산된 값을 ROP 시스템이나 계측기 교정시스템에 제공하기 위한 디지털-아날로그 변환기도 개별 계측기마다 구성될 수 있다. 이에, 본 발명은 정상운전 중인 가압 중수로에 적용하는 경우 국부과출력보호 정지설정치를 최소 5%이상 향상 시킬 수 있고 경년열화 가압중수로의 경우 전출력운전 불가 시기를 늦춰 발전소 이용율 증대와 경제성을 향상시킬 수 있으며, 경년열화가 진행되지 않은 중수로의 경우도 핵연료교체시 불필요한 출력 감발을 회피할 수 있어 경제성 이익을 도모할 수 있다. 가압중수로, 국부과출력 보호, 온라인 감시, 실시간 감시, 계측기, 계측기 교정,

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27-09-2012 дата публикации

Detection method of nuclear spill on design basis accident and visible light wireless alarm system

Номер: KR101186424B1
Принадлежит: 한국원자력연구원

PURPOSE: A design basis accident radioactivity leakage detection technique and a visible light wireless alarm system are provided to detect a design basis accident in early stages by using a photo diode detecting gamma rays. CONSTITUTION: A wireless alarm system(100) is composed of a DBA event detection module(130), a DBA event alarm module(140), a DBA event smart transmission module(150) and a signal generation unit(160). A shield block(110) and an illumination block(120) forms the exterior of the wireless alarm system. The wireless alarm system includes a photo diode(131) to detect gamma rays(21). The DBA event smart transmission module transmits a DBA event detection signal to a main controller. [Reference numerals] (130) DBA event detection module; (140) DBA event alarm module; (150) DBA event smart transmission module; (160) Signal generation unit

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22-10-2001 дата публикации

How to alleviate the leakage of pressurized water reactor and steam generator

Номер: KR100300889B1

가압수형 원자로(1)에서 증기 발생기 관(21)의 파열의 영향은 격납용기 연료공급용 냉각수 저장탱크(35)에 잠겨있는 자동 열제거 장치인 열 교환기(37)를 통해, 증기 발생기(11)내의 공급수 고위 레벨에 응답하여 원자로 냉각재를 변환시킴으로써 1차루프내의 압력을 감소시키는 것에 의해 감소된다. 또한, 원자로 냉각재의 적정량은 노심 보충탱크(43)로부터 냉각재를 원자로 냉각재 시스템의 가압기(31)내의 압력으로 1차루프내 에 공급하는 증기발생기 고위 레벨에 응답하여 유지된다. 증기 발생기 고위 레벨은 또한, 개시 공급수 장치(81)와 화학 및 부피 제어장치(85)를 차단하여 증기헤더(27)내로 범람하는 것을 방지하는데 사용된다.

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04-03-2015 дата публикации

DCS for nuclear power station

Номер: CN104392757A

本发明公开了一种核电站用DCS系统,包括用于将仪控系统中数字量信号发至现场仪表的数个DO模块,及用于为DO模块供电的电源装置,电源装置的电源输出端连接于母线,且每一个DO模块通过一个保险端子连接于母线,以获取电源;DCS系统还包括监视报警装置,用于监视DO模块是否失电,并当DO模块失电时发出报警信号。本发明的核电站用DCS系统可监视DO模块是否失电,并当DO模块失电时发出报警信号,这样完善了DCS系统自身监视的不足,保证了核电站DCS系统的稳定性和可靠性,极大的减小了因丢失驱动电源,无法控制就地设备而导致电站跳机和跳堆的概率,有效的提高了设备的使用率,增强了设备使用的经济性,减少了设备损坏、人员伤害的风险。

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26-04-2019 дата публикации

The method and system for preliminary examination nuclear power plant equipment abnormal signs including determining the processing routine of Chemical Apparatus Importance Classification and alarm validity

Номер: CN109690641A

本发明提供一种用于对设备的异常迹象进行早期警报的方法及系统,所述早期警报包括对设备重要度和警报有效性的确定。所述用于对设备的异常迹象进行早期警报的方法包括:第一步骤:通过早期警报处理装置使用基于先前已由操作者分析的用于每个监控变量的重要度数据的加权值来确定设备监控信号值是否超出正常操作范围;第二步骤:当所述设备监控信号值超出所述正常操作范围时,通过所述早期警报处理装置生成警报;以及第三步骤:通过警报确定装置确定所生成的所述警报是否为有效警报,所述有效警报待进行警报分析和追踪。

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07-10-2011 дата публикации

High pressure safety injection tank system for loca and sbo

Номер: KR101071415B1

본 발명에 따른 고압안전주입탱크 시스템은, 노심충수탱크 및 저압(약 4.3Mpa 이하)의 안전주입탱크를 대체한 고압(약 17MPa) 변경이 가능한 하나의 안전주입탱크를 포함하여, 저압(약 4.3Mpa 이하) 및 고압(약 17MPa)에서도 원자로계통에 비상노심냉각수의 주입이 모두 가능하도록 구성된다. 구체적으로, 본 발명은 저압(약 4.3Mpa)의 질소가 충진되고 비상노심냉각수가 수용되며, 원자로용기에 비상냉각수주입관에 의해 연결된 안전주입탱크와, 고압(약 17MPa)수증기가 수용되며, 상기 고압수증기가 배출되는 안전밸브관이 장착된 가압기 및 고압의 상기 가압기와 저압의 상기 안전주입탱크가 압력평형이 되도록, 상기 안전주입탱크의 상부와 상기 가압기의 상부를 연결하면서 선택적으로 개폐되는 압력평형관을 포함하여 구성된다. 원자로계통이 가압되는 사고시에는 상기 비상노심냉각수가 고압의 상기 원자로용기에 주입 가능 하도록, 상기 압력평형관의 개방에 의해 상기 안전주입탱크가 저압(약 4.3Mpa)에서 고압으로 변경되도록 구성된다. 또한, 원자력발전소 비상발전기 고장에 따른 전원 완전상실 사고시에도 배터리 공급 비상전원에 의해 구동되는 밸브를 적용하여 고압안전주입탱크에 의한 비상노심냉각수 주입이 가능한 것을 특징으로 하는 고압안전주입탱크 시스템이다. The high pressure safety injection tank system according to the present invention includes a safety injection tank capable of changing a high pressure (about 17 MPa) in place of a core filling tank and a low pressure (about 4.3 MPa or less) safety injection tank, and thus, a low pressure (about 4.3 Mpa and below) and high pressure (about 17 MPa) are also configured to allow the injection of emergency core coolant into the reactor system. Specifically, the present invention is filled with low pressure (about 4.3Mpa) of nitrogen and the emergency core coolant is accommodated, the safety injection tank connected to the reactor vessel by the emergency coolant injection pipe, and the high pressure (about 17MPa) steam is accommodated, Pressure balance that is selectively opened and closed while connecting the upper portion of the safety injection tank and the upper portion of the pressurizer so that the pressurizer equipped with a safety valve tube through which high pressure steam is discharged and the high pressure pressurizer and the low pressure safety injection tank are in pressure balance. It is composed including a tube. When the reactor system is pressurized, the safety injection tank is configured to be changed from low pressure (about 4.3 Mpa) to high pressure by opening the pressure balance tube so that the emergency core ...

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18-03-2015 дата публикации

Control method and system for overpressure protection of main loop of nuclear power plant

Номер: CN104425043A

一种核电站主回路的超压保护的控制方法及系统,该方法包括在实时温度低于第一预设温度T1时,余热排出系统安全阀进入超压保护状态,稳压器安全阀进入第一超压保护状态;实时温度高于第一预设温度T1时,余热排出系统安全阀进入隔离状态,稳压器安全阀自动取消第一超压保护状态并进入第二超压保护状态。稳压器安全阀的第二超压保护状态能够对高温超压进行保护,且其第一超压保护状态与余热排出系统安全阀的超压保护状态共同实现低温时的超压保护,正常情况下系统低温超压保护由余热排出系统安全阀的开启实现,余热排出系统被隔离时,第一超压保护状态对系统的低温超压进行保护,加强了低温工况下的超压保护,显著降低反应堆压力容器脆性断裂风险。

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01-11-2019 дата публикации

Dual inserting circuit breaker in power system of power plant

Номер: KR102039683B1
Автор: 김윤호, 김한상, 하체웅
Принадлежит: 한국수력원자력 주식회사

본 발명은 3권 변압기를 포함하는 발전소의 전력계통에 적용되며, 상기 복수의 제1 비안전급 고압모선 및 상기 복수의 제2 비안전급 고압모선 각각에 하나씩 연결된 복수의 메인 차단기, 하나의 제1 차단기에 하나가 직렬 연결된 복수의 보조 차단기, 상기 메인 차단기에 대응하여 설치되고 상기 메인 차단기가 설치된 비안전급 고압모선의 상태를 파악하기 위해 상기 메인 차단기에 인가되는 비안전급 고압모선의 전원 레벨을 측정하는 제1 전원측정기, 상기 보조 차단기에 대응하여 설치되고 상기 메인 차단기의 고장 유무를 파악하기 위해 설치된 제1 지점의 전원 레벨을 측정하는 제2 전원측정기, 그리고 상기 제1 전원측정기에서 측정한 전원 레벨을 통해 비안전급 고압모선의 이상을 파악하면 상기 메인 차단기로 제1 오픈 신호를 출력하고, 상기 제1 오픈 신호 출력 후 상기 제2 전원측정기에서 측정한 제1 지점의 전원 레벨을 통해 상기 메인 차단기가 고장이라고 파악하는 경우에 상기 보조 차단기로 제2 오픈 신호를 출력하는 콘트롤러를 포함하는 이중인입 차단기 시스템에 관한 것이다.

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14-01-2022 дата публикации

Diagnostic method, recovery method, diagnostic device and recovery device for sub-criticality key safety function

Номер: CN113936821A
Принадлежит: China Nuclear Power Engineering Co Ltd

本发明提供一种次临界度关键安全功能的诊断方法、恢复方法、诊断装置及恢复装置。方法包括:步骤一:判断功率量程功率是否小于第一设定值,若否,确定高风险,并转步骤一,若是,转步骤二;步骤二:判断中间量程倍增时间是否小于第二设定值,若否,确定高风险,并转步骤一,若是,转步骤三;步骤三:判断源量程仪表是否已通电,若否,转步骤四,若是,转步骤五;步骤四:判断中间量程倍增时间是否小于第三设定值,若否,确定中风险,并转步骤一,若是,转步骤一;步骤五:判断源量程倍增时间是否小于第四设定值,若否,确定中风险,并转步骤一,若是,转步骤一。本发明能够在核电厂事故运行工况下快速有效地诊断并恢复次临界度功能。

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04-06-2019 дата публикации

Passive depressurized system for nuclear reactor pressure container

Номер: CN106663477B
Принадлежит: Aswad Internet Co

压力容器减压系统,其特征在于包括一个加压容器(1),设置有开口弹簧(10)的气动致动器的主阀(8),其一侧连接向所述加压容器(1)壳体里面的气体,另一侧连接大气,限定该开口弹簧(10)的预定机械压力,因此,当加压容器(1)内的压力比规定的弹簧机械压力更大,主阀(8)保持闭合,并且当加压容器(1)内的压力低于预定的机械压力,主阀(8)被打开,并且可以保持打开状态打开,允许容器中的加压气体(1)被排放到大气中它的主要操作的不需要任何外部电源,允许在事故情况下,适当地发挥其作用,甚至电力的全失的情况下。

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29-05-2017 дата публикации

Adsorbent of radioactive iodine and method of processing of radioactive iodine

Номер: RU2620584C1
Принадлежит: Раса Индастриз, Лтд.

Группа изобретений относится к гранулированному адсорбенту радиоактивного йода. Гранулированный адсорбент радиоактивного йода из цеолита X, в котором ионообменные участки цеолита X замещены серебром так, чтобы размер мелких пор цеолита X соответствовал размеру молекулы водорода, и адсорбент радиоактивного йода имеет содержание серебра 36 вес. % или более при высушивании, размер частиц 10×20 меш и содержание воды 12 вес. % или менее, когда сушится при 150°С в течение 3 ч и, таким образом, уменьшается в весе. Также имеется Также имеется способ обработки радиоактивного йода, содержащегося в пару, выпускаемом из атомной энергетической установки. собой перегретый пар, имеющий температуру 100°С или более. Группа изобретений позволяет более эффективно адсорбировать радиоактивный йод и удалить водород, который представляет собой фактор происшествий на атомных реакторах. 2 н. и 8 з.п. ф-лы, 7 ил., 9 табл., 10 пр. РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (13) 2 620 584 C1 (51) МПК G21F 9/02 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ФОРМУЛА (21)(22) Заявка: ИЗОБРЕТЕНИЯ К ПАТЕНТУ РОССИЙСКОЙ ФЕДЕРАЦИИ 2016119398, 22.08.2014 (24) Дата начала отсчета срока действия патента: 22.08.2014 (72) Автор(ы): КОБАЯСИ Тосики (JP), ЭНДО Кодзи (JP) (73) Патентообладатель(и): РАСА ИНДАСТРИЗ, ЛТД. (JP) Дата регистрации: (56) Список документов, цитированных в отчете о поиске: US 4088737 A1, 09.05.1978. RU Приоритет(ы): (30) Конвенционный приоритет: (45) Опубликовано: 29.05.2017 Бюл. № 16 (85) Дата начала рассмотрения заявки PCT на национальной фазе: 23.05.2016 (86) Заявка PCT: JP 2014/072011 (22.08.2014) (87) Публикация заявки PCT: Адрес для переписки: 125009, Москва, а/я 332, ООО "Инэврика" R U 2 6 2 0 5 8 4 (54) АДСОРБЕНТ РАДИОАКТИВНОГО ЙОДА И СПОСОБ ОБРАБОТКИ РАДИОАКТИВНОГО ЙОДА (57) Формула изобретения 1. Гранулированный адсорбент радиоактивного йода из цеолита X, в котором ионообменные участки цеолита X замещены серебром так, чтобы размер мелких пор цеолита X ...

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10-05-2022 дата публикации

Method and device for determining accident procedure setting value of nuclear power plant

Номер: CN113421676B
Принадлежит: Nuclear Power Institute of China

本发明公开了一种核电厂事故规程整定值的确定方法及装置,该方法包括:S1:获取核电厂事故规程的基准整定值信息;S2:确定事故规程整定值的不确定性,根据基准整定值信息和事故规程整定值的不确定性,采用事故规程整定值方法,计算得到安全壳正常情况下整定值信息、安全壳事故工况下整定值信息;S3:进行事故规程整定值的归纳与合并,得到简化后的整定值信息;S4:根据简化后的整定值信息,进行整定值的验证,判断核电厂是否达到预期效果,若没达到则重复执行步骤S1至S4继续调整,直至产生最优化的事故规程整定值信息。本发明提升了事故规程的应对事故、恢复电厂状态的能力。

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27-11-2018 дата публикации

Ventilation system and associated operating method for use during a serious accident in a nuclear installation

Номер: US10137399B2
Автор: Axel Hill
Принадлежит: AREVA GMBH

A ventilation system for an operating space accessible to operators in a nuclear installation is intended to allow a supply of decontaminated fresh air for a period of a few hours in the event of serious accidents involving the release of radioactive activity. In particular, the component of radioactive inert gases in the fresh air supplied to the operating space should be as small as possible. For this purpose, the ventilation system has a supply air line that is guided from an external inlet to the operating space, and into which a first fan and a first inert gas adsorber column are connected. An exhaust air line is guided from the operating space to an external outlet, and into which a second fan and a second inert gas adsorber column are connected. A switching device is provided for interchanging the roles of the first and second inert gas adsorber columns.

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24-06-2016 дата публикации

METHOD FOR CONTROLLING THE STOPPING OF A NUCLEAR REACTOR WITH PRESSURIZED WATER

Номер: FR3030862A1
Принадлежит: DCNS SA

Ce procédé de gestion de l'arrêt d'un réacteur nucléaire à eau pressurisée, intégré dans un module de production d'énergie électrique, en cas de détection d'une fuite primaire/secondaire d'un générateur de vapeur muni d'une soupape de sécurité, raccordé au réacteur et associé à un moyen de réfrigération de secours, est caractérisé en ce qu'il comporte les étapes suivantes: détection (en 1) d'une fuite primaire/secondaire du générateur de vapeur, arrêt automatique (en 2) du réacteur et isolement du générateur de vapeur en avarie, lignage (en 3) du moyen de réfrigération de secours correspondant, surveillance (en 4) de la pression primaire, dès que la pression primaire passe sous la pression de tarage des soupapes de sécurité des générateurs de vapeur, isolement (en 5) du moyen de réfrigération de secours du générateur de vapeur en avarie, et poursuite (en 6) du refroidissement passif du réacteur sur les générateurs et moyens de réfrigération restants. This method of managing the shutdown of a pressurized water nuclear reactor, integrated in an electric power generation module, in the event of detecting a primary / secondary leak of a steam generator equipped with a valve safety device, connected to the reactor and associated with a means of emergency refrigeration, is characterized in that it comprises the following steps: detection (in 1) of a primary / secondary leak of the steam generator, automatic shutdown (in 2 ) of the reactor and isolation of the steam generator in damage, lineage (in 3) of the corresponding backup refrigeration means, monitoring (in 4) of the primary pressure, as soon as the primary pressure passes under the calibration pressure of the safety valves steam generators, isolation (in 5) of the emergency refrigeration means of the steam generator in damage, and continuation (in 6) of the passive cooling of the reactor on the generators and refrigeration means we are left.

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