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Применить Всего найдено 4895. Отображено 100.
20-01-2008 дата публикации

УСТАНОВКА ДЛЯ ОБЕСПЕЧЕНИЯ МАНЕВРЕННОСТИ АТОМНЫХ ЭЛЕКТРИЧЕСКИХ СТАНЦИЙ

Номер: RU0000070312U1

Установка для обеспечения маневренности атомных электрических станций содержит ядерный реактор, парогенератор, паровую турбину, соединенную с электрогенератором, и через конденсатор и конденсатный насос с системой регенеративных подогревателей низкого давления, связанной последовательно установленными деаэратором, питательным насосом парогенератора, подогревателями высокого давления, соединенными с парогенератором, причем подогреватели низкого и высокого давления через конденсатор связаны с паровой турбиной, выходной вал которой соединен с электрогенератором, который связан с реактором для получения кислорода и водорода, за которым установлены емкости для накопления и хранения кислорода и водорода, соединенные с расположенными в технологической последовательности камерой сгорания, паровой турбиной сверхкритических параметров, вторыми конденсатором и конденсатным насосом, связанным через регулирующий клапан с подогревателями низкого давления, деаэратором, питательным насосом камеры сгорания, подогревателями высокого давления, связанными с камерой сгорания, при этом конденсатный насос соединен с резервуаром для воды, связанный с помощью насосов с реактором для получения кислорода и водорода с одной стороны и с камерой сгорания с другой, а подогреватели низкого и высокого давления соединены через второй конденсатор с паровой турбиной сверхкритических параметров, соединенной со вторым электрогенератором. РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) 70 312 (13) U1 (51) МПК F01K 13/02 (2006.01) H02J 9/04 (2006.01) G21D 3/08 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ, ПАТЕНТАМ И ТОВАРНЫМ ЗНАКАМ (12) ОПИСАНИЕ ПОЛЕЗНОЙ МОДЕЛИ К ПАТЕНТУ (21), (22) Заявка: 2007139766/22 , 26.10.2007 (24) Дата начала отсчета срока действия патента: 26.10.2007 (45) Опубликовано: 20.01.2008 (73) Патентообладатель(и): Государственное образовательное учреждение высшего профессионального образования "Кубанский государственный технологический университет" (ГОУВПО "КубГТУ") (RU) Ñòðàíèöà: 1 U ...

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10-09-2009 дата публикации

ЯДЕРНАЯ ПАРОГАЗОВАЯ УСТАНОВКА

Номер: RU0000086783U1

1. Ядерная парогазовая установка, содержащая реактор и, по меньшей мере, две самостоятельных циркуляционных петли, каждая из которых состоит из двух замкнутых контуров, первый из которых включает в себя испаритель с сепаратором, расположенных в страховочном корпусе, а второй контур включает в себя: последовательно расположенные в газовом тракте пароперегреватель высокого давления, пароперегреватель низкого давления, экономайзер, паровую турбину высокого давления с подогревателем высокого давления и паровую турбину низкого давления с подогревателем низкого давления, газовую турбину, конденсатор, электрогенераторы, отличающаяся тем, что установка снабжена сухой градирней, выполненной из купола и аппаратов воздушного охлаждения, соединенных посредством промежуточного контура с конденсатором, под куполом расположены все элементы ядерной парогазовой установки, при этом реактор размещен на центральной оси купола. 2. Ядерная парогазовая установка по п.1, отличающаяся тем, что испаритель выполнен вертикального исполнения, кожухотрубчатого типа, теплообменная поверхность которого состоит из змеевиков, имеющих малый радиус гиба и объединенные в шестигранные теплообменные модули. 3. Ядерная парогазовая установка по п.1, отличающаяся тем, что пароперегреватели высокого и низкого давления и экономайзер имеют теплообменную поверхность, которая состоит из змеевиков с малым радиусом гиба, которые объединены в теплообменные модули плоского типа, вертикального исполнения и разделены на прямоугольные однотипные секции. 4. Ядерная парогазовая установка по п.1, отличающаяся тем, что аппараты воздушного охлаждения имеют теплообменную поверхность, которая состоит из змеевиков с малым радиусом гиба, объединенные в теплообменные модули плоского типа, вертикального исполнения и занимают свободное место между паровыми и газовыми турбинами равномерно. 5. Ядерная парогазовая установка по п.1, отличающаяся тем, что в верхней части газового тракта размещен эжектор, увеличивающий естественную ...

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20-10-2010 дата публикации

ЯДЕРНАЯ ПАРОГАЗОВАЯ УСТАНОВКА

Номер: RU0000098625U1

1. Ядерная парогазовая установка, содержащий реактор, размещенный на центральной оси купола сухой градирни, под которым расположены все элементы ядерной парогазовой установки, имеющей, по меньшей мере, две самостоятельных циркуляционных петли, каждая из которых состоит из двух замкнутых контуров, первый из которых включает в себя испаритель с сепаратором, расположенные в страховочном корпусе, а второй контур включает в себя последовательно расположенные в газовом тракте пароперегреватель высокого давления, пароперегреватель низкого давления и экономайзер, паровую турбину высокого давления и паровую турбину низкого давления, газовую турбину с камерой сгорания и компрессором, электрогенераторы, аппараты воздушного охлаждения, соединенные посредством промежуточного контура с конденсатором, отличающаяся тем, что компрессор выполнен из двух частей в виде компрессора низкого давления и компрессора высокого давления, между которыми установлен промежуточный холодильник воздуха, имеющий два входа и два выхода: один вход соединен с конденсатором, выход - с первым входом введенного деаэратора, второй вход - с компрессором низкого давления, а его выход - с компрессором высокого давления, установка снабжена подогревателем газа, имеющим два входа и два выхода, первый вход которого соединен с трубопроводом топливного газа, а его выход - с камерой сгорания газовой турбины, второй вход соединен с паровой турбиной низкого давления, а его выход - с конденсатором, деаэратор, кроме того, имеет второй вход и один выход, второй вход которого соединен с турбиной низкого давления, а его выход с экономайзером. 2. Ядерная парогазовая установка по п.1, отличающаяся тем, что холодильник выполнен вертикального исполнения кожухотрубного типа, змеевики которого имеют малый радиус гиба и объединены в шестигранные теплообменные модули. 3. Ядерная парогазовая установка по п.1, отличающаяся тем, что подогреватель газа выполнен вертикального исполнения кожухотрубного типа, змеевики которого имеют малый ...

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20-11-2012 дата публикации

ВОДОСБРОС АТОМНОЙ ЭЛЕКТРОСТАНЦИИ

Номер: RU0000122199U1

Водосброс атомной электростанции, включающий сбросные водоводы конденсаторов паровых турбин, сифонные колодцы и переливные стенки, отличающийся тем, что водосброс снабжен транспортной тележкой с гидравлическим домкратом и гидрогенератором, устанавливаемым в сбросном канале. РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (51) МПК G21D 3/04 (13) 122 199 U1 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ (21)(22) Заявка: ПОЛЕЗНОЙ МОДЕЛИ К ПАТЕНТУ 2012122931/07, 04.06.2012 (24) Дата начала отсчета срока действия патента: 04.06.2012 (45) Опубликовано: 20.11.2012 Бюл. № 32 (73) Патентообладатель(и): Открытое акционерное общество "Российский концерн по производству электрической и тепловой энергии на атомных станциях" (ОАО "Концерн Росэнергоатом") (RU) U 1 1 2 2 1 9 9 R U Формула полезной модели Водосброс атомной электростанции, включающий сбросные водоводы конденсаторов паровых турбин, сифонные колодцы и переливные стенки, отличающийся тем, что водосброс снабжен транспортной тележкой с гидравлическим домкратом и гидрогенератором, устанавливаемым в сбросном канале. Стр.: 1 U 1 (54) ВОДОСБРОС АТОМНОЙ ЭЛЕКТРОСТАНЦИИ 1 2 2 1 9 9 Адрес для переписки: 188540, Ленинградская обл., г. Сосновый Бор, Филиал ОАО "Концерн Росэнергоатом" "Ленинградская атомная станция", директору В.И. Перегуде R U Приоритет(ы): (22) Дата подачи заявки: 04.06.2012 (72) Автор(ы): Перегуда Владимир Иванович (RU), Шмаков Леонид Васильевич (RU), Губин Сергей Иванович (RU), Самусев Леонид Ефимович (RU), Комов Александр Николаевич (RU), Судаков Александр Вениаминович (RU), Федорович Евгений Данилович (RU), Шубин Вячеслав Юрьевич (RU), Лаврентьев Павел Викторович (RU), Бусыгин Евгений Николаевич (RU) RU 5 10 15 20 25 30 35 40 45 122 199 U1 Предлагаемая полезная модель относится к области ядерной энергетики, касается в частности водосброса атомной электростанции и может быть использовано на действующих атомных электростанциях с целью получения дополнительной электроэнергии, в частности, в ...

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20-02-2015 дата публикации

УСТРОЙСТВО УПРАВЛЕНИЯ ЭЛЕКТРОМАГНИТАМИ ВОЗДУШНЫХ ЗАТВОРОВ СИСТЕМЫ ПАССИВНОГО ОТВОДА ТЕПЛА

Номер: RU0000150648U1

Устройство управления электромагнитами воздушных затворов системы пассивного отвода тепла, содержащее шины питания положительного и отрицательного потенциалов, дистанционный переключатель и первое, второе и третье реле, отличающееся тем, что в него введены развязывающий трансформатор, входные клеммы которого соединены с входными клеммами дистанционного переключателя, выполненного с возможностью выполнения функций входного автоматического выключателя питающей сети, первый индикатор, первая и вторая клеммы которого соединены, соответственно, с первой и второй выходными клеммами развязывающего трансформатора и с первой и второй клеммами первого реле, второй, третий, четвертый, пятый и шестой индикаторы, первый и второй резисторы, источник питания, первая и вторая выходные клеммы которого являются, соответственно, шинами питания положительного и отрицательного потенциалов, выключатель, конденсатор, измерительный вольтметр, предохранитель, милливольтметр, а также диод, причем второе реле является реле датчика температуры, третье реле является реле контроля изоляции на электромагнитах воздушных затворов системы пассивного отвода тепла, первая и вторая клеммы пятого индикатора соединены, соответственно, с первой выходной клеммой развязывающего трансформатора и с первой выходной клеммой третьего реле, первая и вторая клеммы шестого индикатора соединены, соответственно, с первой выходной клеммой развязывающего трансформатора и со второй выходной клеммой третьего реле, первая и вторая входные клеммы третьего реле соединены, соответственно, с первой и второй выходными клеммами развязывающего трансформатора, первая клемма второго реле соединена с первой клеммой развязывающего трансформатора и с первой клеммой нормально разомкнутого контакта второго реле, вторая клемма которого через первый резистор соединена со второй клеммой второго реле и со второй клеммой развязывающего трансформатора, первая клемма третьего индикатора соединена с первой выходной клеммой развязывающего ...

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10-09-2016 дата публикации

СИСТЕМА ПОВЫШЕНИЯ МАНЕВРЕННОСТИ И БЕЗОПАСНОСТИ АЭС

Номер: RU0000164717U1

Система повышения маневренности и безопасности АЭС, содержащая дополнительную паротурбинную установку, аккумулятор фазового перехода, бак горячей воды, причем дополнительная паротурбинная установка подключена к аккумулятору фазового перехода посредством паропровода, бак горячей воды подключен к аккумулятору фазового перехода посредством двух трубопроводов, аккумулятор фазового перехода подключен к парогенератору посредством трубопровода, отличающаяся тем, что дополнительная паротурбинная установка подключена к парогенератору через быстродействующую редукционную установку посредством паропровода. РОССИЙСКАЯ ФЕДЕРАЦИЯ (19) RU (11) (51) МПК G21D 3/18 (13) 164 717 U1 (2006.01) ФЕДЕРАЛЬНАЯ СЛУЖБА ПО ИНТЕЛЛЕКТУАЛЬНОЙ СОБСТВЕННОСТИ (12) ОПИСАНИЕ (21)(22) Заявка: ПОЛЕЗНОЙ МОДЕЛИ К ПАТЕНТУ 2015153035/07, 09.12.2015 (24) Дата начала отсчета срока действия патента: 09.12.2015 (73) Патентообладатель(и): Бессонов Валерий Николаевич (RU), Аминов Рашид Зарифович (RU), Юрин Валерий Евгеньевич (RU) (45) Опубликовано: 10.09.2016 U 1 1 6 4 7 1 7 R U Стр.: 1 U 1 (54) СИСТЕМА ПОВЫШЕНИЯ МАНЕВРЕННОСТИ И БЕЗОПАСНОСТИ АЭС (57) Реферат: Согласно предлагаемой системе повышения дополнительной ПТУ 9. маневренности и безопасности АЭС в В авариных ситуациях, сопровождаемых эксплуатационном режиме в период ночного обесточиванием дополнительная ПТУ 9 провала электрической нагрузки часть свежего продолжает генерировать электроэнергию на пара направляется после УП1 1 в АФП 6, где пар собственные нужды АЭС, используя пар, отдает часть тепла веществу, заполняющему получаемый в ПГ за счет энергии остаточного аккумулятор и имеющему большую теплоту тепловыделения реактора, при этом избыток пара фазового перехода из твердого состояния в направляется для зарядки АФП 6. Когда энергии жидкое, конденсат свежего пара поступает в БГВ остаточного тепловыделения становится 7. В режиме пиковых нагрузок горячая вода недостаточно, горячая вода из БГВ 7 поступает посредством насоса подачи горячей воды 8 из АФП 6 для ...

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21-06-2012 дата публикации

Control system and method for pressurized water reactor (pwr) and pwr systems including same

Номер: US20120155594A1
Принадлежит: Individual

A pressurized water reactor (PWR) comprises a pressure vessel, a reactor core disposed in the pressure vessel, an integral or external pressurizer, primary coolant disposed in the pressure vessel and heated by operation of the reactor core, and a steam generator disposed in the pressure vessel and configured to convert secondary coolant in the form of feedwater into steam by heat transfer from the primary coolant heated by operation of the reactor core to secondary coolant in the steam generator. A controller is configured to perform a PWR control method including the operations of (i) adjusting one or more parameters of the PWR and (ii) adjusting a pressurizer water level setpoint based on a predicted direction and magnitude of change of a pressurizer water level of the PWR predicted to result from the adjusting (i).

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19-07-2012 дата публикации

Full spectrum loca evaluation model and analysis methodology

Номер: US20120185222A1
Принадлежит: Westinghouse Electric Co LLC

This invention relates to a computational system and method for performing a safety analysis of a postulated Loss of Coolant Accident in a nuclear reactor for a full spectrum of break sizes including various small, intermediate and large breaks. Further, modeling and analyzing the postulated small break, intermediate break and large break LOCAs are performed with a single computer code and a single input model properly validated against relevant experimental data. Input and physical model uncertainties are combined following a random sampling process, e.g., a direct Monte Carlo approach (ASTRUM-FS) and advanced statistical procedures are utilized to show compliance with Nuclear Regulatory Commission 10 CFR 50.46 criteria.

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23-08-2012 дата публикации

Resonance calculation program and analyzing apparatus

Номер: US20120215444A1
Принадлежит: Mitsubishi Heavy Industries Ltd

[Problem to be Solved] To provide a resonance calculation program capable of evaluating a physical quantity such as an effective cross section in a radial direction of a circular region of a fuel rod by making a resonance calculation based on the equivalence principle. [Solution] A resonance calculation program for calculating an effective cross section by performing a resonance calculation based on an equivalence principle includes a radial-distribution calculation step S 8 of calculating a distribution of the effective cross section in a radial direction of a circular region by calculating the effective cross section defined by a predetermined calculation expression including a geographical coefficient for each of a plurality of annular regions while a neutron escape probability in a resonance region is expressed by a polynomial rational expression including the geographical coefficient serving as a factor representing geographical shapes of a plurality of annular regions that are circumferentially annular and that are obtained by radially dividing the circular region that is an axial cross section of a fuel rod at predetermined intervals.

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30-08-2012 дата публикации

Method for assisting in the operation of a nuclear reactor

Номер: US20120219101A1
Принадлежит: AREVA NP SAS

The present invention relates to a method for assisting in the operation of a nuclear reactor, which comprises the steps of: establishing a request using a man/machine interface ( 31 ) interacting with a dedicated operation assistance computer ( 32 ) and using a three-dimensional neutron computation code ( 32 a ) solving the diffusion equation, referred to as the operation assistance code; unidirectionally transmitting, from a system ( 10 ) for monitoring the operation of the reactor core to said operation assistance computer ( 32 ), a set of data ( 13 ) which are representative of the hardware, geometric, and neutron characteristics of the core, as well as the operating conditions of the core, said data ( 13 ) being determined by a three-dimensional neutron code ( 12 ) updating the isotope balance of the core during fuel depletion and periodically solving the diffusion equation online, referred to as the monitoring code, said monitoring code ( 12 ) being installed on a second separate computer, referred to as the monitoring computer, which is dedicated to said monitoring system ( 10 ); determining a change in the behavior of the reactor core using said operation assistance code ( 32 a ), said representative) data ( 13 ) being used as input data for said operation assistance code ( 32 a ).

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20-09-2012 дата публикации

Nuclear power station

Номер: US20120236980A1
Автор: Oliver REDSCHLAG
Принадлежит: Redschlag Holding GmbH

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

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10-01-2013 дата публикации

Apparatus, method and program for monitoring nuclear thermal hydraulic stability of nuclear reactor

Номер: US20130013282A1
Принадлежит: Toshiba Corp

An apparatus for monitoring nuclear thermal hydraulic stability of a nuclear reactor, contains: a calculation unit configured to calculate a stability index of a nuclear thermal hydraulic phenomenon based on nuclear instrumentation signals, the signals being outputted by a plurality of nuclear instrumentation detectors placed at regular intervals in a reactor core; a simulation unit configured to simulate the nuclear thermal hydraulic phenomenon based on a physical model by using information on an operating state of the nuclear reactor as an input condition; a limit value updating unit configured to update a limit value of the nuclear thermal hydraulic phenomenon based on a result of the simulation; and a determination unit configured to determine, based on the stability index and the limit value, whether or not to activate a power oscillation suppressing device.

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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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23-05-2013 дата публикации

CONTROL SYSTEM FOR NUCLEAR FACILITIES

Номер: US20130129028A1
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

Provided are a safety protection system facility outputting a normal actuating signal Si in a case where the safety protection system facility controls actuation of a unit provided in a nuclear facility to a safe side based on an abnormality detecting signal output at the time of occurrence of an abnormality in the nuclear facility, and where this causes the unit to be actuated normally, and a CCF countermeasure facility outputting a CCF-case actuating signal S actuating the unit to a safe side in a case where the CCF countermeasure facility determines from output results of the abnormality detecting signal and the normal actuating signal S that the unit is not actuated normally at the time of occurrence of the abnormality in the nuclear facility. 1. A control system for a nuclear facility comprising: a detecting sensor provided in a nuclear facility and configured to output an abnormality detecting signal at the time of occurrence of an abnormality in the nuclear facility; a main control device for outputting a normal actuating signal when a unit is actuated normally in consequence of controlling the unit provided in the nuclear facility to a safe side based on the abnormality detecting signal; and an auxiliary control device , as an auxiliary of the main control device , for outputting an auxiliary actuating signal to actuate the unit to a safe side in a case where the auxiliary control device determines from output results of the abnormality detecting signal and the normal actuating signal that the unit is not actuated normally for the abnormality in the nuclear facility , wherein the auxiliary control device includes: a NOT circuit connected to an output side of the main control device , and configured to invert input or no input of the normal actuating signal and output an inverted signal; and a first AND circuit configured to output the auxiliary actuating signal based on input or no input of the signal output from the NOT circuit and input or no input of the ...

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30-05-2013 дата публикации

CONTROL SYSTEM FOR PLANT

Номер: US20130136222A1
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

In a control system for a plant controlling a plant such as a nuclear power plant with use of a plurality of digital control devices , the plurality of digital control devices include a plurality of control functions to , and the plurality of control functions to are provided in the plurality of digital control devices in a distributed manner so that the digital control devices may not fall below safety standards preset by safety analyses. This provides a control system for a plant using the plurality of digital control devices and configured to control a plant safely even when a digital control device is failed. 1. A control system for a plant for controlling a plant with use of a plurality of digital control devices , whereinthe plurality of digital control devices include a plurality of control means, andthe plurality of control means are provided in the plurality of digital control devices in a distributed manner so that the digital control devices may not fall below safety standards preset by safety analyses.2. The control system for a plant according to claim 1 , whereinthe plant is a nuclear power plant including a nuclear reactor, a plurality of steam generators connected to the nuclear reactor, and a plurality of main feed water systems configured to supply a coolant to the respective steam generators,the plurality of control means include a plurality of feed water control means controlling the respective main feed water systems, andthe plurality of feed water control means are distributed into the different digital control devices, respectively.3. The control system for a plant according to claim 1 , whereinthe plant is a nuclear power plant including a nuclear reactor,the plurality of control means include an actuation control means controlling actuation of a unit provided in the nuclear power plant and an interlock control means locking actuation control by the actuation control means, andthe actuation control means and the interlock control means ...

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13-06-2013 дата публикации

Mixing system

Номер: US20130148771A1
Принадлежит: Mitsubishi Heavy Industries Ltd

A mixing system including: a backup heater which increases an internal pressure of a pressurizer; a spray valve which decreases the internal pressure of the pressurizer; a pressure sensor which detects the pressure inside the pressurizer; and a pressure control unit which performs a feedback control on a spray valve so that the pressure becomes a target pressure based on a detection pressure value detected by the pressure sensor, wherein the pressure control unit includes a PID control unit which performs a feedback control when heating the coolant by the backup heater and outputs a pressure controller signal and a bias setting unit which sets a bias toward the operation side of the spray valve with respect to the pressure controller signal, and wherein the spray valve is operated based on the pressure controller signal subjected to the setting of the bias.

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20-06-2013 дата публикации

CRITICAL HEAT FLUX PREDICTION DEVICE, CRITICAL HEAT FLUX PREDICTION METHOD AND SAFETY EVALUATION SYSTEM

Номер: US20130156142A1
Автор: Yodo Tadakatsu
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

A critical heat flux prediction device, a critical heat flux prediction method, a safety evaluation system, and a core monitoring system using the safety evaluation system can predict critical heat flux in a core of a reactor with a high degree of accuracy by obtaining a correlation plot distribution representing a relation of critical heat flux on a thermal equilibrium quality based on experimental data, approximating a correlation plot distribution through a logistic function that is a model function in which critical heat flux is expressed by a function of a thermal equilibrium quality, and obtaining a critical heat flux correlation of critical heat flux and a thermal equilibrium quality. 1. A critical heat flux prediction device , comprising:a storage unit for storing experimental data including a thermal equilibrium quality decided based on sampled critical heat flux and an experimental condition;an experimental data plotting unit for obtaining a correlation plot distribution representing a relation of critical heat flux on a thermal equilibrium quality based on the experimental data; anda critical heat flux correlation calculating unit for obtaining a correlation of the critical heat flux and the thermal equilibrium quality by approximating the correlation plot distribution by a logistic function that is a model function in which the critical heat flux is expressed by a function of the thermal equilibrium quality.3. A critical heat flux prediction method , comprising:acquiring experimental data including a thermal equilibrium quality decided based on sampled critical heat flux and an experimental condition;obtaining a correlation plot distribution representing a relation of critical heat flux on a thermal equilibrium quality based on the experimental data; andobtaining a correlation of the critical heat flux and the thermal equilibrium quality by approximating the correlation plot distribution by a logistic function that is a model function in which the ...

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27-06-2013 дата публикации

OPERATION MONITORING APPARATUS OF NUCLEAR POWER PLANT

Номер: US20130163709A1
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

An operation monitoring apparatus is provided with a function integration VDU () including a normal use system VDU () which operates and monitors a nuclear power plant under a normal state, and a safety protection system VDU () which shuts down the nuclear power plant under an abnormal state, and also a safety protection system VDU () which shuts down the nuclear power plant under the abnormal state, and further, a block device which inhibits an actuation of the safety protection system VDU () when the function integration VDU () is under an abnormal state, thereby capable of simplifying an operation manipulation in the nuclear power plant, and ensuring a high level of safety of an operation control. 1. An operation monitoring apparatus of a nuclear power plant , comprising:a function integration manipulation device including a normal use system control function which operates and monitors a plant under a normal state, and a safety protection system control function which brings the plant under an abnormal state into a safety state;a safety protection system manipulation device including a safety protection system control function which brings the plant under the abnormal state into the safety state; anda block device which inhibits an obstruction of an actuation of the safety protection system control function by the function integration manipulation device when the function integration manipulation device is under an abnormal state.2. The operation monitoring apparatus of the nuclear power plant according to claim 1 , wherein the block device is provided to the safety protection system manipulation device.3. The operation monitoring apparatus of the nuclear power plant according to claim 2 , wherein the block device is configured by an AND circuit into which a plant manipulation signal output from the function integration manipulation device claim 2 , and a block signal output from the safety protection system manipulation device claim 2 , are input.4. The ...

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11-07-2013 дата публикации

CONTROL DEVICE AND NUCLEAR POWER PLANT CONTROL SYSTEM

Номер: US20130177119A1
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

A nuclear power plant control system includes control devices, and the control devices and each include arithmetic units that respectively execute arithmetic processing in parallel independently, based on detection results of detection units, and each output a control signal to control a countermeasure unit in accordance with an arithmetic result of the arithmetic processing, a transmission unit that sends out the control signal to the countermeasure unit, when the control signal is outputted from at least one of the arithmetic units, and a system management unit that performs control so as to inhibit the control signal outputted by the arithmetic unit as a test object from being sent out from the transmission unit while maintaining a state where the other arithmetic operation executes the arithmetic processing independently, when a test of either of the control devices is conducted. 1. A control device used in a safety protection system of a nuclear power plant , comprising: a plurality of arithmetic units that respectively execute arithmetic processing in parallel and independently , based on a detection result of a detection unit for detecting a specific event occurring in the nuclear power plant , and each output a control signal to control countermeasure unit for taking countermeasures against the event in accordance with an arithmetic result of the arithmetic processing; a transmission unit that sends out the control signal to the countermeasure unit , when the control signal is outputted from at least one of the plurality of arithmetic units; and a control unit that performs control so as to inhibit the control signal outputted by the arithmetic unit as a test object from being sent out from the transmission unit while maintaining a state where the other arithmetic unit executes the arithmetic processing independently , when a test of one of the plurality of the arithmetic units is conducted.2. The control device according to claim 1 , wherein after the test ...

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18-07-2013 дата публикации

NUCLEAR POWER PLANT CONTROL SYSTEM AND NUCLEAR POWER PLANT CONTROL METHOD

Номер: US20130182809A1
Автор: Fujimoto Hiroshi
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

A nuclear power plant control system () is provided with detection units (to ) which detect phenomena that occurs in a nuclear power plant for each of four systems, a trip control device () which starts, in the case where a signal that indicates an occurrence of the phenomenon is input from at least a predetermined number of signal lines out of signal lines of two systems, processing corresponding to the phenomenon, and majority circuits (and ) which are provided for each signal line of the two systems and each output, in the case where the phenomenon is detected by N or more detection units out of the detection units (to ), a signal that indicates an occurrence of the phenomenon to a corresponding signal line. 1. A nuclear power plant control system , comprising:detection units that detect a phenomenon occurring in a nuclear power plant for each of M systems;a start unit that starts processing corresponding to the phenomenon in a case where a signal that indicates an occurrence of the phenomenon is input from a predetermined number or more of signal lines out of L signal lines; anda majority circuit, provided for each of the L signal lines, each of which outputting a signal that indicates an occurrence of the phenomenon to a corresponding signal line in a case where the phenomenon is detected in N or more systems out of the M systems of the detection units,wherein L is an integer greater than or equal to 1, M is an integer greater than or equal to 2, and N is an integer greater than or equal to 1.2. The nuclear power plant control system according to claim 1 , wherein the majority circuit is combined with a relay or a breaker provided for each of the M systems.3. The nuclear power plant control system according to claim 2 , wherein the relay or the breaker is multiplexed for each of the M systems.4. The nuclear power plant control system according to claim 1 , wherein the start unit is a trip control device of a nuclear reactor.5. A nuclear power plant control ...

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25-07-2013 дата публикации

SYSTEM, METHOD, AND PROGRAM FOR MONITORING REACTOR CORE

Номер: US20130188765A1
Принадлежит: KABUSHIKI KAISHA TOSHIBA

According to one embodiment of a reactor core monitoring system, includes: an information retention portion for retaining a regular cycle and a short cycle as calculation information of reactor core performance data; a signal processing portion for creating heat balance data based on a process signal; a data acquisition portion for acquiring, in a timing of the regular cycle, the heat balance data and reactor core performance data which was calculated in a previous timing of the regular cycle, while acquiring, in a timing of the short cycle asynchronous to the regular cycle, the heat balance data and reactor core performance data which was calculated most recently; and a data calculation portion for calculating new reactor core performance data based on the acquired reactor core performance data and the heat balance data. 1. A reactor core monitoring system , comprising:an information retention portion for retaining a regular cycle and a short cycle as calculation information of reactor core performance data;a signal processing portion for creating heat balance data based on a process signal;a data acquisition portion for acquiring, in a timing of the regular cycle, the heat balance data and reactor core performance data which was calculated in a previous timing of the regular cycle, while acquiring, in a timing of the short cycle asynchronous to the regular cycle, the heat balance data and reactor core performance data which was calculated most recently; anda data calculation portion for calculating new reactor core performance data based on the acquired reactor core performance data and the heat balance data.2. The reactor core monitoring system according to claim 1 , whereinactivation of the short cycle can be started/stopped arbitrarily.3. The reactor core monitoring system according to claim 1 , whereinthe information retention portion also retains storage address information on the reactor core performance data calculated in the data calculation portion.4. The ...

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08-08-2013 дата публикации

CONTROL SYSTEM FOR NUCLEAR FACILITY AND CONTROL METHOD FOR NUCLEAR FACILITY (AS AMENDED)

Номер: US20130202074A1
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

A control system allows controlling a nuclear facility in an evacuation area. The control system includes a control device in ordinary use disposed in a non-evacuation area, an emergency control device—for emergency in the evacuation area, a plant control facility connectable to the control device—and the emergency control device, a signal switching unit that switches from a normal coupling to an emergency coupling based on an emergency switch signal, a first selector switch in the non-evacuation area, a second selector switch-in the evacuation area, an AND circuit configured to output the emergency switch signal to the signal switching unit in the case where the emergency switch signal is input from the first selector switch and the emergency switch signal is input from the second selector switch. 1. A control system for controlling a nuclear facility in an evacuation area , the control system comprising:a first operation control device in ordinary use, the first operation control device being disposed in a non-evacuation area, the non-evacuation area being an area other than the evacuation area;a second operation control device for emergency, the second operation control device being disposed in the evacuation area;a control device connectable to the first operation control device or the second operation control device;a signal switching unit configured to switch from a normal coupling between the control device and the first operation control device to an emergency coupling between the control device and the second operation control device based on an emergency switch signal to be input;a first selector switch disposed in the non-evacuation area, the first selector switch being configured to output an emergency switch signal by a switching operation to the emergency coupling;a second selector switch disposed in the evacuation area, the second selector switch being configured to output an emergency switch signal by a switching operation to the emergency coupling; ...

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10-10-2013 дата публикации

METHODS FOR PROTECTION OF NUCLEAR REACTORS FROM THERMAL HYDRAULIC/NEUTRONIC CORE INSTABILITY

Номер: US20130266107A1
Принадлежит: WESTINGHOUSE ELECTRIC COMPANY LLC

The invention relates to methods for protecting a nuclear reactor core, such as a boiling water reactor core, from fuel and cladding damage due to thermal hydraulic instability in extended operating power flow conditions and, in particular, when an extended power uprate is implemented. The methods employ existing licensed stability methodologies and incorporated minor changes, e.g., to the Average Power Range Monitor (APRM)-based trip system to preclude operation inside the stability vulnerable region of the power/flow map. The APRM-based trip system is modified to set down the APRM flow-biased scram line when core flow is less than a predetermined core flow to prevent the core from entering an unstable region of operation. 1. A method for protecting a nuclear reactor core from fuel damage due to thermal hydraulic instability in an extended operating domain , comprising:calculating a thermal hydraulic limit;identifying an unstable region of operation in a power/flow map;identifying a predetermined core flow level;modifying the APRM-based trip system to set down the APRM flow-biased scram line when core flow is less than the predetermined core flow level; andpreventing the nuclear core from entering the unstable region of operation by causing a APRM scram or a selected rod insert when the set down APRM flow-biased scram line is exceeded.2. The method of claim 1 , wherein a set point for the set down of the APRM flow-biased scram line is determined by employing 3D core analyses of the thermal hydraulic stability limit.3. The method of claim 1 , wherein operation of the nuclear reactor is conducted beyond a MELLLA rod control line to a MELLLA+ rod control line above the predetermined core flow level.4. The method of claim 1 , wherein existing OPRM instrumentation is operable to detect and suppress global and local unstable operation in a MELLLA+ region.5. The method of claim 1 , wherein the predetermined core flow level is defined by the thermal hydraulic stability ...

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07-11-2013 дата публикации

CONTROL ROOM FOR NUCLEAR POWER PLANT

Номер: US20130294560A1
Автор: GRAHAM Thomas G.
Принадлежит:

A reactor control interface includes a home screen video display unit (VDU) displaying blocks representing functional components of a nuclear power plant and connecting arrows that connect blocks that are providing the current heat sinking path for the nuclear power plant. Directions of the connecting arrows represent the direction of heat flow along the current heat sinking path. If the current heat flow path of the plant changes, the connecting arrows are updated accordingly. Additional VDUs include: a mimic VDU displaying a mimic of a plant component; a procedures VDU displaying a stored procedure executable by the plant; a multi-trend VDU trending various plant data; and an alarms VDU displaying side-by-side alarms registries sorted by time and priority respectively. If a VDU fails, the displays are shifted to free up one VDU to present the display of the failed VDU, and one display is shifted to an additional VDU. 1. A reactor control interface comprising: blocks representing functional components of a nuclear power plant including at least (i) blocks representing functional components of a normal heat sinking path of the nuclear power plant and (ii) blocks representing functional components of at least one remedial heat sinking path of the nuclear power plant, and', 'connecting arrows of a first type connecting blocks that are providing the current heat sinking path wherein directions of the connecting arrows of the first type represent the direction of heat flow along the current heat sinking path., 'a home screen video display unit (VDU) configured to display2. The reactor control interface of wherein connecting arrows of the first type do not connect blocks that are not providing the current heat sinking path.3. The reactor control interface of wherein the home screen VDU is further configured to display connecting arrows of a second type visually distinguishable from the first type claim 2 , the arrows of the second type connecting blocks.3. The reactor ...

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14-11-2013 дата публикации

System and Method for Annealing Nuclear Fission Reactor Materials

Номер: US20130301770A1
Принадлежит:

Illustrative embodiments provide systems, methods, apparatuses, and applications related to annealing nuclear fission reactor materials. 1176.-. (canceled)177. A method comprising:adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range;maintaining the temperature deviation from the nominal operating temperature for a period selected to produce a selected annealing effect; andafter the period selected to produce the selected annealing effect, adjusting the reactor coolant system to return to the nominal operating temperature range.178. The method of claim 177 , where the selected annealing effect includes a predicted annealing effect.179. The method of claim 177 , where the selected annealing effect includes a measured annealing effect.180. The method of claim 177 , wherein adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range includes adjusting reactor coolant flow.181. The method of claim 177 , further comprising reversing direction of reactor coolant flow.182. The method of claim 177 , further comprising adjusting a rate of heat generation.183. The method of claim 177 , wherein adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range includes adjusting a rate of heat transferred from the reactor coolant.184. The method of claim 177 , wherein adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range includes adjusting a rate of heat transferred to the reactor coolant.185. The method of claim 177 , wherein adjusting a reactor coolant system to produce a temperature deviation from a nominal operating temperature range includes replacing at least a portion of a first reactor coolant having first heat transfer characteristics with second coolant having second heat transfer characteristics.186. The method of claim 177 , wherein adjusting a reactor ...

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14-11-2013 дата публикации

NUCLEAR POWER PLANT CONTROL SYSTEM AND METHOD OF TESTING NUCLEAR POWER PLANT

Номер: US20130301771A1
Принадлежит: MITSUBISHI HEAVY INDUSTRIES, LTD.

A nuclear power plant control system () includes a control button () which receives an operation for controlling a control target device (), a notification lamp () which notifies that a control signal corresponding to the operation received by the control button () arrives at a predetermined position on a path connected from a control panel () to the control target device (), and a control signal inhibition unit () which inhibits a control signal from arriving at the control target device () in midstream between the predetermined position on the path and the control target device () in response to an operation received by a test permission button (). 1. A nuclear power plant control system comprising: an operation unit which receives an operation for controlling a specific portion of a nuclear power plant; a notification unit which notifies that a control signal corresponding to the operation received by the operation unit arrives at a predetermined position on a path connected to the portion; and an inhabitation unit which inhibits the control signal from arriving at the portion in a position between the predetermined position on the path and the portion.2. The nuclear power plant control system according to claim 1 , wherein the operation unit is provided in a central control room of the nuclear power plant claim 1 , and the notification unit notifies an operator present in the central control room.3. The nuclear power plant control system according to claim 1 , wherein the inhabitation unit inhibits the control signal from arriving at the portion when a signal indicating a state of being on test is received.4. The nuclear power plant control system according to claim 1 , wherein the inhabitation unit is a majority circuit that causes a control signal to arrive at the portion only when a plurality of control signals are received through different paths.5. The nuclear power plant control system according to claim 4 , wherein the inhabitation unit is a majority ...

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14-11-2013 дата публикации

Risk Monitoring Device and Risk Monitoring Method for Use with a Nuclear Power Facility

Номер: US20130301772A1
Принадлежит:

The object of the present invention is to provide a risk monitoring device and a risk monitoring method for use with a nuclear power facility, providing continuous determination of risk associated with the nuclear power facility, based on one-type security model, which can be implemented using simple software and hardware means. The risk monitoring device according to the present invention comprises a memory device for storing at least one minimal fault cutset array (MFC) and probability values for each event in each MFC, and a data input device to input in the risk monitoring device data about status changes of the nuclear power facility facility, wherein the risk monitoring device further comprises a formation unit for forming at least one MFC matrix; a memory device for storing said at least one MFC matrix; a formation unit for forming at least one parameter matrix; a memory device for storing said at least one parameter matrix; a modification unit for modifying elements of said at least one parameter matrix; and a risk evaluation unit. 1. A risk monitoring device for use with a nuclear power facility , comprising:a first memory device for storing at least one minimal fault cutset (MFC) array corresponding to an undesired event associated with said facility, and for storing probability values for each event in each MFC;a data input device for inputting in the risk monitoring device data about status changes of the facility;a first formation unit for forming at least one minimal fault cutset (MFC) matrix, wherein the first formation unit is adapted to form a rectangular MFC matrix based on each MFC array in such manner that events associated with one MFC form one line of the matrix, wherein a horizontal dimension of the matrix is defined by length of the longest MFC, and lines formed by the MFC and having a shorter length than the horizontal dimension of the MFC matrix are complemented with simulated events with probability of 1;a second memory device for storing ...

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02-01-2014 дата публикации

Method and system for supplying emergency power to nuclear power plant

Номер: US20140001863A1
Принадлежит: China General Nuclear Power Corp

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

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20-02-2014 дата публикации

Method of constructing pseudo hot pin power distribution using in-core detector signal-based planar radial peaking factors in core operating limit supervisory system

Номер: US20140050290A1
Принадлежит: Kepco Nuclear Fuel Co Ltd

Disclosed herein is a method for constructing a pseudo hot pin power distribution using in-core detector signal-based planar radial peaking factors in a Core Operating Limit Supervisory System (COLSS). The method includes defining a planar radial peaking factor F xy K based on in-core detector signals in the COLSS, and expanding the planar radial peaking factor F xy K so that the planar radial peaking factor F xy K is suitable for a number of nodes of the COLSS. The planar radial peaking factor F xy K is calculated only for the in-core detector signals using a preset equation, rather than by using table lookup.

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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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21-01-2016 дата публикации

Computer aided design system for pulsed extraction column operation process

Номер: US20160019322A1
Принадлежит: China Institute of Atomic of Energy

A computer aided design system for pulsed extraction column operation process is provided, including a data input module, a data detection module, a computing module, a computing result detection module, and an error data management module. The data input module is connected with the data detection module, the data detection module is connected with the computing module, the computing module is connected with the computing result detection module, and the error data management module is connected with the data detection module and the computing result detection module. The system of the present invention is capable of computing and displaying the concentration changes of the corresponding components during operation process in real time, and reporting abnormal conditions which may arise during operation process. After the system is running, the results themselves are stored in the storage unit, which is convenient for user's analyzing and processing operating results under different conditions.

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21-01-2016 дата публикации

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

Номер: US20160019986A1
Принадлежит: AREVA GMBH

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

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21-01-2016 дата публикации

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

Номер: US20160019987A1
Принадлежит: AREVA GMBH

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

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03-02-2022 дата публикации

METHOD FOR MONITORING NUCLEAR POWER PLANT IN TRANSIENT STATE BY USING SIGNAL CLASSIFICATION

Номер: US20220037045A1
Принадлежит:

The present invention relates to a method for monitoring a nuclear power plant in a transient state, the method comprising the steps of: classifying signals to be monitored of the nuclear power plant into a constant monitoring signal, a primary system monitoring signal, a secondary system monitoring signal, and a monitoring signal during normal operation; constantly monitoring the constant monitoring signal at the time of starting the nuclear power plant; sequentially initiating monitoring of the primary system monitoring signal and the secondary system monitoring signal while monitoring the constant monitoring signal; and initiating monitoring of the monitoring signal during normal operation when it is determined to be operating normally after initiating the monitoring of the secondary system monitoring signal. 1. A method of monitoring a nuclear power plant in a transient state , the method comprising:classifying signals to be monitored of the nuclear power plant into a constant monitoring signal, a primary system monitoring signal, a secondary system monitoring signal, and a normal operation monitoring signal;constantly monitoring the constant monitoring signal at a time of starting the nuclear power plant;sequentially initiating monitoring of the primary system monitoring signal and the secondary system monitoring signal while monitoring the constant monitoring signal; andinitiating monitoring of the normal operation monitoring signal when a normal operation is determined after the initiating of the monitoring of the secondary system monitoring signal.2. The method of claim 1 ,wherein the initiating of the monitoring of the primary system is performed when a primary system initiation signal reaches a predetermined level.3. The method of claim 2 ,wherein the primary system monitoring signal is divided into a plurality of groupswherein initiating monitoring of each of the groups is sequentially performed according to a magnitude of the primary system initiation ...

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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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18-01-2018 дата публикации

Safety critical system

Номер: US20180019030A1
Принадлежит:

According to an example embodiment of the present invention, there is provided a method, comprising defining a task category information element, the task category information element being associated with at least one functional requirement and at least one design principle, associating the task category information element with at least one architecture definition information element, associating each of the at least one architecture definition information element with at least one system-level information element, and verifying the system described by the at least one architecture definition information element and associated system-level information elements is compliant with the at least one design principle 1. A computerized nuclear power station monitoring system , comprising:a memory configured to store a database comprising a task category database comprising plurality of task category information elements comprising a safety function information element and a reactor protection information element, each task category information element being associated with at least one technical functional requirement and at least one technical design principle, each technical design principle being comprised in a technical design principle list, the technical design principle list comprising redundancy, diversity, separation and isolation, each functional requirement being comprised in a functional requirement list, the functional requirement list comprising reactivity control, core cooling and safe shut-down, and an equipment database configured to store at least one equipment information element, andat least one processor configured to, responsive to receipt in the computerized monitoring system of a failure notification concerning a first equipment information element, determine, using the database, a set comprising each technical design principle associated with each task category information element associated, via database relations, with the first equipment ...

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17-01-2019 дата публикации

SHUTDOWN SYSTEM FOR A NUCLEAR STEAM SUPPLY SYSTEM

Номер: US20190019588A1
Принадлежит:

A nuclear steam supply system having a shutdown system for removing residual decay heat generated by a nuclear fuel core. The steam supply system may utilize gravity-driven primary coolant circulation through hydraulically interconnected reactor and steam generating vessels forming the steam supply system. The shutdown system may comprise primary and secondary coolant systems. The primary coolant cooling system may include a jet pump comprising an injection nozzle disposed inside the steam generating vessel. A portion of the circulating primary coolant is extracted, pressurized and returned to the steam generating vessel to induce coolant circulation under reactor shutdown conditions. The extracted primary coolant may further be cooled before return to the steam generating vessel in some operating modes. The secondary coolant cooling system includes a pumped and cooled flow circuit operating to circulate and cool the secondary coolant, which in turn extracts heat from and cools the primary coolant. 1. A nuclear steam supply system with startup primary coolant heating system , the nuclear steam supply system comprising:a reactor vessel having an internal cavity;a reactor core comprising nuclear fuel disposed within the internal cavity and operable to heat a primary coolant;a steam generating vessel fluidly coupled to the reactor vessel;a riser pipe positioned within the steam generating vessel and fluidly coupled to the reactor vessel;a primary coolant loop formed within the reactor vessel and the steam generating vessel, the primary coolant loop being configured for circulating primary coolant through the reactor vessel and steam generating vessel; and an intake conduit having an inlet fluidly coupled to the primary coolant loop;', 'a pump fluidly coupled to the intake conduit, the pump configured and operable to extract and pressurize primary coolant from the primary coolant loop and discharge the pressurized primary coolant through an injection conduit;', 'the ...

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10-02-2022 дата публикации

Method, System, and Apparatus for the Thermal Storage of Nuclear Reactor Generated Energy

Номер: US20220044833A1
Принадлежит:

A method, system, and apparatus for the thermal storage of nuclear reactor generated energy including diverting a selected portion of energy from a portion of a nuclear reactor system to an auxiliary thermal reservoir and, responsive to a shutdown event, supplying a portion of the diverted selected portion of energy to an energy conversion system of the nuclear reactor system. 1127.-. (canceled)128. A method , comprising:providing a first portion of energy from at least one nuclear reactor of a nuclear reactor system to at least one energy conversion system;diverting a selected portion of energy from the at least one nuclear reactor to at least one auxiliary thermal reservoir, the selected portion of energy exceeding operational demand of the at least one energy conversion system; andstoring the diverted selected portion of energy in the at least one auxiliary thermal reservoir in the form of a temperature change or a phase change in at least one heat storage material of the at least one auxiliary thermal reservoir.129. The method of claim 128 , further comprising responsive to a signal regarding a shutdown event claim 128 , supplying at least a portion of the diverted selected portion of energy to the at least one energy conversion system.130. The method of claim 128 , wherein the diverting the selection portion of energy from the portion of the at least one nuclear reactor comprises operating the nuclear reactor at full power.131. The method of claim 128 , further comprising:determining that energy production by the at least one energy conversion system exceeds current grid demand;converting excess electrical power to thermal energy; andstoring the thermal energy in in the at least one auxiliary thermal reservoir in the form of a temperature change or a phase change in the at least one heat storage material of the at least one auxiliary thermal reservoir.132. The method of claim 128 , further comprising providing at least a portion of the diverted selected portion ...

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23-01-2020 дата публикации

FUEL ELEMENT WITH MULTI-SMEAR DENSITY FUEL

Номер: US20200027577A1
Принадлежит:

A fuel element has a ratio of area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis to total area of the interior volume in the cross-section of the tubular fuel element that varies with position along the longitudinal axis. The ratio can vary with position along the longitudinal axis between a minimum of 0.30 and a maximum of 1.0. Increasing the ratio above and below the peak burn-up location associated with conventional systems reduces the peak burn-up and flattens and shifts the burn-up distribution, which is preferably Gaussian. The longitudinal variation can be implemented in fuel assemblies using fuel bodies, such as pellets, rods or annuli, or fuel in the form of metal sponge and meaningfully increases efficiency of fuel utilization. 1. A method of manufacturing a fuel element , the method comprising:modeling fuel strain along a longitudinal axis of a fuel element;modeling a smear density profile along the longitudinal axis of the fuel element to offset the modeled fuel strain such that at least one region of locally decreased strain corresponds to a region of locally increased smear density; andconstructing the fuel element to have a tubular interior volume storing a fissionable composition, the fissionable composition in thermal transfer contact with an interior surface of the fuel element and having a smear density based on the modeled smear density profile.2. The method of claim 1 , wherein the smear density profile increases average burnup at a plurality of locations along the longitudinal axis of the fuel element.3. The method of claim 1 , wherein the modeled smear density approximates an inverted Gaussian shape.4. The method of claim 1 , wherein the smear density of the fuel element is higher at a first end of the fuel element than at a second opposite end of the fuel element.5. The method of claim 4 , wherein the first end of the fuel element is proximate a coolant entry point within a ...

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23-01-2020 дата публикации

FAULT TOLERANT TURBINE SPEED CONTROL SYSTEM

Номер: US20200027595A1
Принадлежит:

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

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23-01-2020 дата публикации

PROCESS SIGNAL CONTROL AND MONITORING SYSTEM

Номер: US20200027597A1
Автор: HAMAYA Yoichiro
Принадлежит: Mitsubishi Electric Corporation

A process signal control and monitoring system, includes: a signal processing device which is installed on an outside of a nuclear reactor containment vessel, an internal electrical power source, an analog-digital conversion part, an internal communication part which transmits the digital signal to the signal processing device, an internal repeater, and an external repeater which transmits the received signal to a communication satellite. When electric power supply from the signal processing device is disconnected, the internal electrical power source supplies electric power which is charged in the rechargeable battery, to the analog-digital conversion part and the internal communication part; and the internal communication part judges whether communication with the signal processing device is continued or disconnected; and when the communication is judged to be continued, the internal communication part continues transmitting the digital signal to the signal processing device. 1. A process signal control and monitoring system , comprising:a signal processing device which is installed on an outside of a nuclear reactor containment vessel,an internal electrical power source which charges a rechargeable battery with electric power, supplied from the signal processing device,an analog-digital converter which converts an analog signal into a digital signal, the analog signal transmitted from a sensor which is installed on an inside of the nuclear reactor containment vessel,an internal communicator which transmits the digital signal, converted in the analog-digital converter, to the signal processing device,an internal repeater which is installed on an inside of the nuclear reactor containment vessel, andan external repeater which is installed on the outside of the nuclear reactor containment vessel and when receiving a signal from the internal repeater, transmits the received signal to a communication satellite,wherein when electric power supply from the signal ...

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23-01-2020 дата публикации

CONTROLLING A NUCLEAR REACTION

Номер: US20200027603A1
Принадлежит:

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction; a riser positioned above the reactor core; a primary coolant flow path that extends from a bottom portion of the volume through the reactor core and through an annulus between the riser and the reactor vessel; a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to generate electric power in a power generation system; and a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions. 1. A nuclear power system , comprising:a reactor vessel that comprises a reactor core mounted within a volume of the reactor vessel, the reactor core comprising one or more nuclear fuel assemblies configured to generate a nuclear fission reaction;a riser positioned above the reactor core;a primary coolant flow path that extends from a bottom portion of the volume below the reactor core, through the reactor core, within the riser, and through an annulus between the riser and the reactor vessel back to the bottom portion of the volume;a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the received heat to generate electric power in a power generation system fluidly or thermally coupled to the primary coolant flow path; anda control rod assembly system positioned in the reactor vessel and configured to position a plurality of control rods in only two discrete positions, such that the plurality of control rods are fully withdrawn from the reactor core in a first discrete position of the only two discrete positions and the plurality of control rods are fully inserted into the reactor core in a second discrete position of the only two discrete positions.2. The nuclear power system of ...

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23-01-2020 дата публикации

CONTROLLING A NUCLEAR REACTION

Номер: US20200027604A1
Принадлежит:

A nuclear power system includes a reactor vessel that includes a reactor core mounted, the reactor core including nuclear fuel assemblies configured to generate a nuclear fission reaction; a riser positioned above the reactor core; a primary coolant flow path that extends from a bottom portion of the volume below the reactor core, through the reactor core, within the riser, and through an annulus between the riser and the reactor vessel back to the bottom portion of the volume; a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the received heat to generate electric power in a power generation system fluidly or thermally coupled to the primary coolant flow path; and a control system communicably coupled to the power generation system and configured to control a power output of the nuclear fission reaction independent of any control rod assemblies during the normal operation. 1. A nuclear power system , comprising:a reactor vessel that comprises a reactor core mounted within a volume of the reactor vessel, the reactor core comprising one or more nuclear fuel assemblies configured to generate a nuclear fission reaction;a riser positioned above the reactor core;a primary coolant flow path that extends from a bottom portion of the volume below the reactor core, through the reactor core, within the riser, and through an annulus between the riser and the reactor vessel back to the bottom portion of the volume;a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the received heat to generate electric power in a power generation system fluidly or thermally coupled to the primary coolant flow path; anda control system communicably coupled to the power generation system and configured to control a power output of the nuclear fission reaction independent of any control rod assemblies during the normal operation.2. ...

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28-01-2021 дата публикации

METHODS AND SYSTEMS FOR FACILITATING THE MANAGEMENT OF REACTOR TRANSIENT CONDITIONS ASSOCIATED WITH REACTORS

Номер: US20210027901A1
Автор: Henry Robert
Принадлежит:

Disclosed herein is a method of facilitating the management of reactor transient conditions associated with reactors. Accordingly, the method may include a step of receiving reactor data associated with a reactor from a reactor computer. Further, the method may include a step of determining a reactor transient condition associated with the reactor based on the reactor data. Further, the method may include a step of receiving reactor design data and measurement data associated with a plurality of reactor components of the reactor from the reactor computer. Further, the method may include a step of analyzing the reactor design data and the reactor measurement data. Further, the method may include a step of generating a notification corresponding to the reactor transient condition based on the analyzing. Further, the method may include a step of transmitting the notification to a user device associated with a user. 1. A system for facilitating the management of reactor transient conditions associated with reactors , the system comprising: receiving at least one reactor data associated with the reactor from the reactor computer;', 'receiving a plurality of reactor design data and a plurality of reactor measurement data associated with a plurality of reactor components of the reactor from the reactor computer;', 'transmitting at least one notification to at least one user device associated with at least one user;, 'a communication device communicatively coupled with a reactor computer associated with a reactor, wherein the communication device is configured for determining at least one reactor transient condition associated with the reactor based on the at least one reactor data;', 'analyzing the plurality of reactor design data and the plurality of reactor measurement data; and', 'generating the at least one notification corresponding to the at least one reactor transient condition based on the analyzing., 'a processing device configured for2. The system of further ...

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31-01-2019 дата публикации

SYSTEM FOR THE HIGHLY AUTONOMOUS OPERATION OF A MODULAR LIQUID-METAL REACTOR WITH STEAM CYCLE

Номер: US20190032519A1
Автор: Vilim Richard B.
Принадлежит:

The invention relates to a nuclear plant in which the power of a nuclear reactor is controlled via demand of a connected electric grid. A naturally circulating nuclear reactor coolant loop is linked to a water/steam loop by means of a steam generator. The water/steam loop consists of an electric power generating unit and a water recirculating and steam control system. The generator is coupled to an external power grid. As power requirements of the grid change, a controller linked to the generator and a three way valve divides steam flow between the expansion turbine and a feedwater heater to boost or retard the power output. Altering the steam flow changes the pressure and temperature in the water/steam system and thus the coolant flow rate. The change in coolant flow allows the reactor core to regulate its reactivity to reach a state of equilibrium to the demand for electric power. 1. A system for regulating nuclear reactor core activity comprising:a naturally circulating nuclear reactor having a nuclear reactor cooling outlet,a nuclear reactor cooling inlet, anda nuclear core with a negative temperature reactivity coefficient;a steam generator having a saturated liquid space displaced above the nuclear reactor cooling outlet, anda steam space;a coolant loop where the coolant loop cycles coolant out through the nuclear reactor coolant outlet, where the coolant loop is in thermal communication with the saturated liquid space of the steam generator, and where the coolant loop cycles coolant in through the nuclear reactor coolant inlet;a steam piping system in fluid communication with the steam space of the steam generator;a three way valve having a valve shaft, in fluid communication at a three way valve inlet port with the steam piping system which leaves the steam generator;an expansion turbine directly fluidly connected to and in fluid communication with the three way valve at a three way valve first outlet port;a condenser in fluid communication with the ...

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17-02-2022 дата публикации

METHOD FOR PROTECTING A NUCLEAR REACTOR AND CORRESPONDING NUCLEAR REACTOR

Номер: US20220051824A1
Принадлежит:

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

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11-02-2016 дата публикации

Methods for Simulating the Flow of a Fluid in a Vessel of a Nuclear Reactor and for Calculating the Mechanical Deformation of Assemblies of a Nuclear Reactor Core, and Associated Computer Program Products

Номер: US20160042822A1
Принадлежит: AREVA NP

A method for simulating the fluid flow in a vessel of a nuclear reactor is provided. The reactor includes a core inside the vessel, the core including a lower plate, an upper plate and fuel assemblies extending between the plates, and having a volume axially delimited by first and second interfaces corresponding to the plates. 114-. (canceled)15: A method for simulating the flow of a fluid inside a vessel of a nuclear reactor , the nuclear reactor comprising the vessel and a core positioned inside the vessel , the vessel including a fluid inlet orifice and a fluid outlet orifice , the core including a lower plate , an upper plate and nuclear fuel assemblies extending in an axial direction between the lower and upper plates , the core having a volume delimited by first and second interfaces in the axial direction , the first and second interfaces respectively corresponding to the lower and upper plates , the fluid being able to flow inside the core between the assemblies , computing, for the core volume, a pressure of the fluid and component(s) of a speed of the fluid, from an initial value of the speed or pressure of the fluid in the first interface and an initial value of the speed or pressure of the fluid in the second interface and using a fluid mass, a movement quantity balance and energy balance equations of the fluid;', 'determining at least one additional volume inside the vessel, the additional volume being outside the core volume and situated at one of the ends thereof in the axial direction, the additional volume being delimited by two interfaces in the axial direction, one of the two interfaces of the additional volume being the first interface or the second interface; and', 'computing, for the additional volume and using the mass balance, movement quantity balance and energy balance equations of the fluid, the pressure of the fluid and the component(s) of the speed of the fluid, from an initial value of the speed or pressure in one of the interfaces of ...

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11-02-2016 дата публикации

METHODS FOR SIMULATING THE FLOW OF A FLUID IN A VESSEL OF A NUCLEAR REACTOR AND FOR CALCULATING THE MECHANICAL DEFORMATION OF ASSEMBLIES OF A NUCLEAR REACTOR CORE, AND ASSOCIATED COMPUTER PROGRAM PRODUCTS

Номер: US20160042823A1
Принадлежит: AREVA NP

A method for simulating the flow of a fluid in a vessel of a nuclear reactor is provided. The nuclear reactor includes the vessel and a core inside the vessel, the core including nuclear fuel assemblies, each one extending in an axial direction, including nuclear fuel rods and a grid for maintaining the rods, and being spaced apart from another by a clearance between the grids in a transverse direction. 114-. (canceled)15. A method for simulating the flow of a fluid inside a vessel of a nuclear reactor , the nuclear reactor comprising the vessel and a core positioned inside the vessel , the core including nuclear fuel assemblies , each assembly extending in an axial direction and including nuclear fuel rods and at least one grid for maintaining the rods , each assembly being spaced apart from another assembly by a clearance between the grids in a transverse direction perpendicular to the axial direction , the fluid being able to flow inside the core ,the method comprising:i) determining head loss coefficients in the core, {'br': None, 'i': 'P=−K×V', '∇'}, 'ii) computing the pressure of the fluid and component(s) of a speed of the fluid in the core, using the following equationwhere ∇ is the order one spatial derivation nabla operator,P is the component of the pressure of the fluid,K is a matrix including the head loss coefficients determined during step i), andV is a vector including the component(s) of the speed of the fluid,wherein, during step i), a transverse head loss coefficient in the assemblies is determined as a function of a transverse Reynolds number in the transverse direction, and an axial head loss coefficient in the clearance is determined as a function of the dimension of the clearance in the transverse direction.16. The method as recited in claim 15 , wherein claim 15 , during step i) claim 15 , the transverse head loss coefficient is determined claim 15 , for a value of the transverse Reynolds number claim 15 , by comparison with a variable ...

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08-02-2018 дата публикации

AIR-COOLED HEAT EXCHANGER AND SYSTEM AND METHOD OF USING THE SAME TO REMOVE WASTE THERMAL ENERGY FROM RADIOACTIVE MATERIALS

Номер: US20180040386A1
Принадлежит:

A system for removing thermal energy generated by radioactive materials comprising: an air-cooled shell-and-tube heat exchanger comprising a shell and plurality of heat exchange tubes arranged in a substantially vertical orientation within the shell, the plurality of heat exchange tubes comprising interior cavities that collectively form a tube-side fluid path, the shell forming a shell-side fluid path that extends from an air inlet of the shell to an air outlet of the shell, the first air inlet located at a lower elevation than the air outlet; a heat rejection closed-loop fluid circuit comprising the tube-side fluid path of the air-cooled heat exchanger, a coolant fluid flowing through the heat rejection closed-loop fluid circuit, the heat rejection closed-loop fluid circuit thermally coupled to the radioactive materials so that thermal energy generated by the radioactive materials is transferred to the coolant fluid; and the air-cooled shell-and-tube heat exchanger transferring thermal energy from the coolant fluid flowing through the tube-side fluid path to air flowing through the shell-side fluid path. 1. A tube-and-shell air-cooled heat exchanger apparatus comprising:a shell having a shell cavity, a primary air inlet at a first elevation, a secondary air inlet at a second elevation, and an air outlet at a third elevation, wherein the second elevation is greater than the first elevation and the third elevation is greater than the second elevation, each of the primary air inlet, the secondary air inlet, and the air outlet collectively forming an airflow passageway through the shell defining a shell-side fluid path; anda plurality of heat exchange tubes that collectively form a tube bundle having a substantially vertical longitudinal axis, the tube bundle located within the shell cavity, a tube-side fluid path collectively comprised of interior cavities of the plurality of heat exchange tubes.2. The tube-and-shell air-cooled heat exchanger apparatus of claim 1 , ...

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19-02-2015 дата публикации

Central control device of nuclear power plant, plant operation support device, and plant operation support method

Номер: US20150049851A1
Автор: Masanori Yokoyama
Принадлежит: Mitsubishi Heavy Industries Ltd

A central control device of a nuclear power plant monitors and controls the nuclear power plant. The central control device includes monitoring operation devices (for example, an alarm VDU, a common system VDU, and a safety system VDU of an operation console) each having both a function as a display unit that displays information on the nuclear power plant (for example, operation information, safety information, and the like) and a function as an operation unit that operates the nuclear power plant, and a control unit that controls these monitoring operation devices. The control unit selects an operation procedure document corresponding to an input operation to a monitoring operation device and displays the operation procedure document on the monitoring operation device.

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18-02-2021 дата публикации

SYSTEM AND METHOD ENABLING SIGNALS FROM REPLACEMENT SELF-POWERED NEUTRON DETECTORS TO BE USED TO GENERATE INPUTS TO LEGACY SOFTWARE

Номер: US20210050123A1
Автор: Heibel Michael D.
Принадлежит: WESTINGHOUSE ELECTRIC COMPANY LLC

A method whereby signals that are output by replacement SPNDs are converted into equivalent signals that would have been detected by legacy SPNDs for input to the legacy software. The replacement SPNDs have a different geometry than the legacy SPNDs and also have a different neutron sensitivity than the legacy SPNDs. The replacement SPNDs are subjected to a neutron flux in a core of a reactor and responsively output a set of signals. The set of signals and the geometry of the replacement SPNDs are employed to create a characterization of the neutron flux in the form of a curve that represents flux as a function of location along the core of the reactor. The legacy geometry of the legacy SPNDs is then employed to find the values on the curve that correspond with the positions where the legacy SPNDs had been located to create inputs for the legacy software. 1. A method of locating a remotely operated vehicle within a workspace , the method comprising:receiving a video feed of the workspace from a video camera;processing the video feed to identify landmarks and features thereof of known physical structures in or near the workspace;determining a correlation between the landmarks and the features identified in the video feed and known physical structures, wherein determining a correlation comprises constructing a three-dimensional reference framework of the workscace based on previously known dimensional information of the known physical structures;calibrating the video feed from the camera to the known physical structures using the correlation;determining the location in the calibrated video feed of a number of fiducial markers on the remotely operated vehicle; anddetermining the position of the remotely operated vehicle within the workspace using the location of the number of fiducial markers in the calibrated video feed.2. (canceled)3. The method of claim 1 , wherein determining the location in the calibrated video feed of a number of fiducial markers on the remotely ...

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25-02-2021 дата публикации

Nuclear driven carbon dioxide sequestration system and method

Номер: US20210053013A1
Принадлежит: INFORMATION SYSTEMS LABORATORIES Inc

A system and method for heat produced at a nuclear power plant as the energy source for carbon dioxide sequestration while simultaneously producing electricity. The system includes a nuclear power plant that differs significantly from conventional designs inasmuch as its design is tightly integrated into the carbon dioxide sequestration system. The system generates electricity and sequesters carbon dioxide at the same time. Instead of simply generating electricity from the nuclear reactor and then using that electricity to run a sequestration process, the method is designed to directly provide the requisite thermal energy to the sequestration process, and simultaneously power an electrical generator. Another feature of the system design is a method of optimizing load balancing between the electrical grid and carbon dioxide sequestration.

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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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15-05-2014 дата публикации

CONTROL ROOM FOR NUCLEAR POWER PLANT

Номер: US20140133618A1
Автор: GRAHAM Thomas G.
Принадлежит:

A control room for a nuclear power plant including two or more nuclear reactor units includes a central workstation providing monitoring capability for both nuclear reactor units, a first operator at the controls (OATC) workstation in front of and to one side of the central workstation providing monitoring and control capabilities for the first nuclear reactor unit, a second OATC workstation in front of and to the other side of the central workstation providing monitoring and control capabilities for the second nuclear reactor unit, and a common control workstation directly in front of the central workstation providing monitoring and control capabilities for systems serving both the first nuclear reactor unit and the second nuclear reactor unit. The central and common control workstations do not provide control capabilities for either nuclear reactor unit. The common control workstation does not include any control capabilities that must be performed by a licensed operator. 1. A control room for monitoring and controlling a nuclear power plant including a first nuclear reactor unit and a second nuclear reactor unit , the control room comprising:a central workstation providing monitoring capability for both the first nuclear reactor unit and the second nuclear reactor unit;a first operator at the controls (OATC) workstation in front of and to one side of the central workstation providing monitoring and control capabilities for the first nuclear reactor unit but not for the second nuclear reactor unit; anda second OATC workstation in front of and to the other side of the central workstation providing monitoring and control capabilities for the second nuclear reactor unit but not for the first nuclear reactor unit;wherein the central workstation, the first OATC workstation, and the second OATC workstation are disposed in the control room.2. The Main Control Room of wherein the central workstation does not provide control capabilities for the first nuclear reactor unit ...

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15-05-2014 дата публикации

Extended operating cycle for pressurized water reactor

Номер: US20140133619A1
Автор: Vince J. Bilovsky
Принадлежит: Babcock and Wilcox mPower Inc

A pressurized water reactor (PWR) includes a pressure vessel containing a nuclear reactor core immersed in primary coolant water, control rod assemblies (CRA's), and control rod drive mechanisms (CRDM's) operating the CRA's. The reactor core has axially varying 235 U enrichment and/or axially varying burnable poison concentration. A CRDM controller controls the CRA's over a burn-up cycle that does not include fuel assembly shuffling and is divided into a plurality of burn-up intervals. The CRDM controller is configured to, for each burn up interval: position the CRA's in accordance with a CRA pattern defining a set of fixed positions for the CRA's except for a sub-set of CRA's designated by the CRA pattern as floating CRA's, and control power level of the PWR by adjusting the floating CRA's without not adjusting the CRA's not designated as floating CRA's. The primary coolant water optionally does not contain soluble neutron poison.

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21-02-2019 дата публикации

METHOD OF OPERATING A NUCLEAR POWER PLANT

Номер: US20190057783A1
Автор: Leblanc David
Принадлежит:

The present relates to the integration of the primary functional elements of graphite moderator and reactor vessel and/or primary heat exchangers and/or control rods into an integral molten salt nuclear reactor (IMSR). Once the design life of the IMSR is reached, for example, in the range of 3 to 10 years, it is disconnected, removed and replaced as a unit. The spent IMSR functions as the medium or long term storage of the radioactive graphite and/or heat exchangers and/or control rods and/or fuel salt contained in the vessel of the IMSR. The present also relates to a nuclear reactor that has a buffer salt surrounding the nuclear vessel. During normal operation of the nuclear reactor, the nuclear reactor operates at a temperature that is lower than the melting point of the buffer salt and the buffer salt acts as a thermal insulator. Upon loss of external cooling, the temperature of the nuclear reactor increases and melts the buffer salt, which can then transfer heat from the nuclear core to a cooled containment vessel. 1. A method of operating a nuclear power plant , the nuclear power plant comprising a molten salt reactor (MSR) to produce heat , a heat exchanger system , and an end use system , the heat exchanger system to receive heat produced by the MSR and to provide the received heat to the end use system , the method comprising steps of:operating the MSR, the MSR comprising a vessel, a graphite moderator core positioned in the vessel, and a molten salt circulating at least in the vessel, the heat exchanger system having an inside portion located inside the vessel and an outside portion located outside the vessel, the inside portion having a plurality of heat exchangers, each heat exchanger having an inlet conduit and an outlet conduit, each inlet conduit and each outlet conduit extending from each respective heat exchanger, through the vessel, toward the outside portion of the heat exchanger system and connecting each respective heat exchanger to the outside ...

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21-02-2019 дата публикации

Nuclear Instrumentation Isolated Output Signal Scaling Method and System Employing Same

Номер: US20190057790A1
Принадлежит: WESTINGHOUSE ELECTRIC COMPANY LLC

A method of determining a core design parameter of a nuclear reactor, includes: calibrating an isolated voltage output from a NIS cabinet associated with the nuclear reactor using a calibrated signal source as an input to the NIS cabinet; recording values of the calibrated signal source used in the calibrating and corresponding values of the output voltage from the calibrating in an as-left cabinet calibration data table; using a computing device connected to the isolated voltage output from the NIS cabinet, converting the voltage output signal to a converted detector signal using at least some of the values in the as-left cabinet calibration data table in an improved signal conversion equation; and using the computing device, employing the converted detector signal to determine the core design parameter. 1. A method of determining a core design parameter of a nuclear reactor , the method comprising:calibrating an isolated voltage output from a NIS cabinet associated with the nuclear reactor using a calibrated signal source as an input to the NIS cabinet;recording values of the calibrated signal source used in the calibrating and corresponding values of the output voltage from the calibrating in an as-left cabinet calibration data table;using a computing device connected to the isolated voltage output from the NIS cabinet, converting the voltage output signal to a converted detector signal using at least some of the values in the as-left cabinet calibration data table in an improved signal conversion equation; andusing the computing device, employing the converted detector signal to determine the core design parameter.2. The method of claim 1 , further comprising comparing the measured core design parameter to a predicted core design parameter to determine if the measured core design parameter is within an acceptable limit.3. The method of claim 2 , further comprising determining from the comparing that the measured core design parameter is not within the acceptable ...

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22-05-2014 дата публикации

Semi Submersible Nuclear Power Plant and Multipurpose Platform

Номер: US20140140466A1
Автор: David W. Richardson
Принадлежит: Individual

Disclosed is an offshore, manned, scalable, modular, floating, moored, nuclear power generating plant and multipurpose platform. The present invention is comprised of a Main Plant Control Deck, Central Plant Deck, and submerged Reactor-Generator Deck(s) integrated into the structure of a Spar or Cell Spar Platform. The Reactor-Generator decks are comprised of a plurality of modular, Naval Nuclear Pressurized Water Reactor Modules. Electricity generated is transmitted via submarine High Voltage Direct Current Cables to shore. Ancillary, co-generated services, e.g. desalinated water, are transmitted to shore via submarine pipelines. Multipurpose Topside Decks house vessel command, crew, and any ancillary and co-generation equipment. The present invention, constructed in a multi-path manufacturing process, provides exceptional economic, environmental, sustainability, security, safety, and operational advantages over the current art of power generation.

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28-02-2019 дата публикации

Shaft sealing structure and primary coolant circulation pump

Номер: US20190063610A1
Принадлежит: Mitsubishi Heavy Industries Ltd

A shaft sealing structure for a rotation shaft, includes a sealing ring having ends formed by removal of its part. The ends abutting each other are continuous in the circumferential direction when the sealing ring is reduced in diameter to a radially inner side. The sealing ring is provided along the circumferential direction of the rotation shaft so as to be contactable with an outer peripheral surface of the rotation shaft. The structure also includes a pressing member configured to be movable between a pressing position and a retracted position; an elastic member configured to bias the pressing member toward the pressing position by elastic force; and a support member configured to support the pressing member at the retracted position against the elastic force, and to allow the pressing member to move to the pressing position at a predetermined temperature.

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17-03-2022 дата публикации

Physics-guided analytical model validation

Номер: US20220084704A1
Принадлежит: UT Battelle LLC

This invention relates to a parameter or response assist filter that ensures that the predictions of a post-validation calibrated physics system simulator will remain within boundaries of a predetermined model validation domain. Embodiments utilize one or more filters to ensure calibrated model parameters {acute over (P)} and/or calibrated responses {tilde over (ϕ)} cause physics simulator model predictions to remain within the boundaries of the model validation domain MVD for a target application. The filters can be constructed prior to use or automatically inferred, or otherwise determined, from available measurements and other renditions of the physics system simulator during operation.

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17-03-2022 дата публикации

METHOD FOR REGULATING OPERATING PARAMETERS OF A NUCLEAR REACTOR AND CORRESPONDING NUCLEAR REACTOR

Номер: US20220084706A1
Принадлежит:

A method regulates operating parameters comprising at least the mean temperature of the core (T), and the axial power (AO) imbalance.The method includes development of a vector (Us) of control values of the nuclear reactor by a supervisor () implementing a predictive control algorithm; development of a vector (u) of corrective values of the nuclear reactor controls by a regulator () implementing a sequenced gain control algorithm; development of a vector (U) of corrected values of the commands of the nuclear reactor, by using the vector (U) of the values of the commands produced by the supervisor () and the vector (U) of the corrective values of the commands produced by the regulator (); and regulation of the operating parameters of the nuclear reactor, by controlling actuators using the vector (U) of the corrected values of the controls. 1. A method of regulating operating parameters of a nuclear reactor , these operating parameters comprising at least the mean core temperature (T) , and the axial power (AO) imbalance , the method comprising the following steps:{'sub': ['U', 'P'], '#text': 'acquisition of a current value of at least one input (D, D);'}acquisition of a current value (Y) of a vector of outputs, the outputs comprising at least the operating parameters;{'sub': ['ref', 'U', 'P'], '#text': 'development of a reference value (Y) of the vector of the outputs, using the current value of the at least one input (D, D);'}{'sub': ['S', 'U', 'P'], 'b': '31', '#text': 'development of a vector (U) of control values of the nuclear reactor by a supervisor () implementing a predictive control algorithm, using at least said current value of at least one input (D, D) and the current value (Y) of the vector of the outputs;'}{'sub': ['K', 'ref'], 'b': '33', '#text': 'development of a vector (u) of corrective values of the nuclear reactor controls by a regulator () implementing a sequenced gain control algorithm, using the current value (Y) of the vector of the outputs and ...

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11-03-2021 дата публикации

NUCLEAR CONTROL SYSTEM WITH NEURAL NETWORK

Номер: US20210074442A1
Автор: Hoover Ryan J.
Принадлежит: Westnghouse Electric Company LLC

A method of controlling a nuclear power plant includes obtaining sensor data from one or more sensors of the nuclear power plant, providing the sensor data and a desired plant response to a neural network, wherein the neural network has been previously trained using a simulated nuclear power plant and is structured to determine at least one control system setting to achieve the desired plant response, determining at least one control system setting to achieve the desired plant response with the neural network, and setting or changing at least one control system setting of a control system of the nuclear power plant to the at least one control system setting determined by the neural network. 1. A method of controlling a nuclear power plant , the method comprising:obtaining sensor data from one or more sensors of the nuclear power plant;providing the sensor data and a desired plant response to a neural network, wherein the neural network has been previously trained using a simulated nuclear power plant and is structured to determine at least one control system setting to achieve the desired plant response;determining at least one control system setting to achieve the desired plant response with the neural network; andsetting or changing at least one control system setting of a control system of the nuclear power plant to the at least one control system setting determined by the neural network.2. The method of claim 1 , wherein the one or more sensors are structured to measure at least one of flows temperatures claim 1 , pressures claim 1 , and valve positions.3. The method of claim 1 , further comprising:controlling one or more components of the nuclear power plant based on the set or changed at least one control system setting of the control system.4. The method of claim 3 , wherein the components include at least one of a pump and a valve.5. The method of claim 1 , further comprising:selecting the neural network from a plurality of neural networks, wherein the ...

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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

METHOD FOR CHECKING EQUIVALENCE OF CODE, NON-TRANSITORY COMPUTER-READABLE STORAGE MEDIUM, COMPUTER PROGRAM PRODUCT AND COMPUTER FOR IMPLEMENTING THE METHOD

Номер: US20180074934A1
Принадлежит:

A method checks the functional equivalence of two pieces of software for control systems, engineered from function block diagram with a plurality of interconnected function blocks. The method includes a) generating a first instance of compilable source code; b) parsing the first instance of source code; c) using the information from step b) to reconstruct a first data flow graph of the first instance of source code; d) generating a second instance of compilable source code from a second plurality of function block diagrams; e) parsing the second instance of source code; f) using the information from step e) to reconstruct a second data flow graph of the second instance of source code; g) comparing the first identified data flow graph and the second identified data flow graph, and checking if the transitive closure of input up to that node in the second instance of source code is the same. 1. A method for checking functional equivalence of two pieces of software for control systems and engineered from function block diagrams with a plurality of interconnected function blocks , which comprises the steps of:a) generating a first instance of compilable source code from a first plurality of function diagrams by virtue of a first code generator obeying to a first set of sequentialization rules;b) parsing the first instance of the compilable source code and identifying a first set of all function block I/O ports and junction points (nodes) and a first set of all connections between them (arcs) of data flow;c) using information from the step b) to reconstruct a first data flow graph of the first instance of the compilable source code in an upstream direction;d) generating a second instance of the compilable source code from a second plurality of function block diagrams by virtue of a same or a second code generator obeying a second set of sequentialization rules;e) parsing the second instance of the compilable source code and identifying a second set of all function block I ...

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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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24-03-2022 дата публикации

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

Номер: US20220093282A1
Автор: Choi Kyunam
Принадлежит:

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

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05-03-2020 дата публикации

CONTROL ROOM FOR NUCLEAR POWER PLANT

Номер: US20200075186A1
Автор: GRAHAM Thomas G.
Принадлежит:

A reactor control interface includes a home screen video display unit (VDU) displaying blocks representing functional components of a nuclear power plant and connecting arrows that connect blocks that are providing the current heat sinking path for the nuclear power plant. Directions of the connecting arrows represent the direction of heat flow along the current heat sinking path. If the current heat flow path of the plant changes, the connecting arrows are updated accordingly. Additional VDUs include: a mimic VDU displaying a mimic of a plant component; a procedures VDU displaying a stored procedure executable by the plant; a multi-trend VDU trending various plant data; and an alarms VDU displaying side-by-side alarms registries sorted by time and priority respectively. If a VDU fails, the displays are shifted to free up one VDU to present the display of the failed VDU, and one display is shifted to an additional VDU. 1. A method operating in conjunction with video display units (VDUs) of a reactor control interface wherein the VDUs include a group of safety VDUs and an additional VDU that is not a safety VDU , the method comprising:detecting a malfunctioning safety VDU, the remaining safety VDUs being functioning safety VDUs;shifting the displays of the functioning safety VDUs to free up one of the functioning safety VDUs wherein the shifting transfers the display of one of the functioning safety VDUs to the additional VDU that is not a safety VDU; andtransferring the display of the malfunctioning safety VDU to the functioning safety VDU freed up by the shifting.2. The method of wherein the group of safety VDUs includes:a home screen VDU displaying a simplified diagrammatic representation of a nuclear power plant;a mimic VDU displaying a mimic of a component of the nuclear power plant;a procedures VDU displaying a stored procedure executable by the nuclear power plant;a multi-trend VDU displaying trends of data acquired from the nuclear power plant; andan alarms ...

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18-03-2021 дата публикации

NETWORK AND INFORMATION SYSTEMS AND METHODS FOR SHIPYARD MANUFACTURED AND OCEAN DELIVERED NUCLEAR PLATFORM

Номер: US20210082591A1
Автор: Cella Charles Howard
Принадлежит:

The systems and methods generally include a nuclear power plant unit assembled in a shipyard from a plurality of structural modules, each of the structural modules having manufactured components for use in power production when moored or fixed to a floor at least one of in and proximal to at least one of an offshore marine environment, a river environment and a coastal marine environment. The nuclear power plant unit is subdivided into at least one arrangement of structural modules that includes an electrical interface for one of transmitting electrical power generated by the nuclear unit and powering a system of the unit, a communications interface for communications internal or external to the unit, a user interface that is configured to permit a user to access a system of the unit, and a network interface for data communications to or from the unit. 1. A system, comprising: a nuclear power plant unit assembled in a shipyard from a plurality of structural modules, each of the structural modules having manufactured components for use in power production when moored or fixed to a floor at least one of in and proximal to at least one of an offshore marine environment, a river environment and a coastal marine environment, wherein the nuclear power plant unit is subdivided into at least one arrangement of structural modules that includes an electrical interface for one of transmitting electrical power generated by the nuclear unit and powering a system of the unit, a communications interface for communications internal or external to the unit, a user interface that is configured to permit a user to access a system of the unit, and a network interface for data communications to or from the unit. The present application is bypass continuation of International Application number PCT/US2018/023663, filed Mar. 21, 2018, published as WO/2018/175663 on Sep. 27, 2018, and entitled Systems And Methods for Shipyard Manufactured and Ocean Delivered Nuclear Platform, which claims ...

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18-03-2021 дата публикации

Programmable logic circuit for controlling an electrical facility, in particular a nuclear facility, associated control device and method

Номер: US20210083671A1
Принадлежит:

A programmable logic circuit () for controlling an electrical facility, in particular a nuclear facility, includes an operating unit (). The operating unit includes a plurality of types of functional blocks (FB, FB, FB), two distinct types of functional blocks being suitable for executing at least one distinct function, at least one processing module suitable for receiving at least one sequence () of functional block(s) to be executed, and at least one internal memory () configured to store at least said sequence (). The programmable logic circuit () includes a single functional block of each type, a given functional block being suitable for being called several times, and an execution module () configured to execute the called functional block(s) in series, according to said sequence (). 118-. (canceled)19. A programmable logic circuit for controlling an electrical facility , the programmable logic circuit comprising an operating unit comprising:a plurality of types of functional blocks, two distinct types of the functional blocks being configured for executing at least one distinct function;at least one processing module configured for receiving at least one sequence of the functional block(s) to be executed; andat least one internal memory configured to store at least the sequence, the programmable logic circuit including a single functional block of each type of the functional blocks, a given functional block being configured for being called several times, and an execution module configured to execute at least one called functional block in series, according to the sequence.20. The programmable logic circuit according to claim 19 , wherein the programmable logic circuit is of a FPGA type.21. The programmable logic circuit according to claim 19 , wherein the execution module is a finite-state machine.22. The programmable logic circuit according to claim 19 , wherein the operating unit further comprises a plurality of parallelizable floating point units.23. The ...

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12-03-2020 дата публикации

Reactor plant and method of operating reactor plant

Номер: US20200082950A1
Принадлежит: Mitsubishi Heavy Industries Ltd

A reactor plant includes a reactor having a reactor core, and a steam circulation system and a bypass system, as a plurality of systems capable of circulating water carrying thermal energy generated by a nuclear fission reaction in the reactor core, and the water as the same heat medium can be circulated in the steam circulation system and the bypass system.

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12-03-2020 дата публикации

NUCLEAR POWER PLANT DEFENSE-IN-DEPTH SAFETY APPARATUS HAVING DIVERSITY

Номер: US20200082952A1
Автор: Bae Byoung Hwan
Принадлежит:

The present invention relates to a nuclear power plant defense-in-depth safety apparatus having diversity, comprising: an analog type first control unit for generating an operation signal for a safety system; a digital type second control unit for generating an operation signal for the safety system; and a device interface module for receiving the operation signal of the first control unit and the operation signal of the second control unit and applying the operation signal of the first control unit to the safety system as a top priority. 1. A defense-in-depth safety apparatus with diversity in a nuclear power plant , the apparatus comprising:an analog-type first controller configured to generate an operation signal for a safety system;a digital-type second controller configured to generate an operation signal for the safety system; anda component interface module (CIM) configured to receive the operation signal from the first controller and the operation signal from the second controller, and apply first the operation signal from the first controller to the safety system as a top priority.2. The apparatus of claim 1 , wherein the operation signal from the second controller comprises an engineered safety feature component control system signal (ESF-CCS) and a diversity protection system signal.3. The apparatus of claim 2 , wherein the safety system comprises at least one of a safety injection system claim 2 , a main steam isolation system claim 2 , a containment spray system claim 2 , or a steam generator auxiliary feedwater system.4. The apparatus of claim 3 , wherein the first controller is not affected from a common mode failure of the second controller.5. The apparatus of claim 3 , wherein the first controller comprises a selector that allows an operator to select operation by the first controller.6. The apparatus of claim 3 , further comprising a selector that allows the operator to select operation by the first controller.7. The apparatus of claim 3 , further ...

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05-05-2022 дата публикации

DEVICES, SYSTEMS, AND METHODS FOR CONFIGURING THE LAYOUT OF UNIT CELL OF A REACTOR CORE

Номер: US20220139582A1
Принадлежит: WESTINGHOUSE ELECTRIC COMPANY LLC

A configurable unit cell of a core of a nuclear reactor is disclosed herein. The configurable unit cell includes a core block material and a plurality of interchangeable components configured to affect a performance parameter of the core of the nuclear reactor. The configurable unit cell further includes a plurality of channels defined within the core block material. Each channel of the plurality of channels is configured to engage an interchangeable component of the plurality of interchangeable components in an operating configuration. Each channel of the plurality of channels is separated from an adjacent channel of the plurality of channels by a predetermined pitch.

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26-06-2014 дата публикации

APPARATUS AND METHOD FOR CONTROLLING GAIN ACCORDING TO RATE OF CHANGE IN WATER LEVEL OF STEAM GENERATOR IN NUCLEAR POWER PLANTS

Номер: US20140177772A1

Provided is an apparatus for controlling a gain according to a water level change rate of a steam generator in nuclear power plants. The apparatus includes a water level variance detector detecting a water level variance of the steam generator, a change rate calculator calculating a water level change rate with respect to the detected water level variance, a compensation value calculator calculating a compensation gain value corresponding to the calculated water level change rate, a reactor power sensor sensing whether or not reactor power corresponds to certain power or less, and a gain compensation controller, when the reactor power corresponds to the certain power or less, outputting a control gain value obtained by combining a general gain value provided to control a proportional-integral (PI) controller with the compensation gain value to the PI controller 1. An apparatus for controlling a gain according to a water level change rate of a steam generator in nuclear power plants , the apparatus comprising:a water level variance detector detecting a water level variance of the steam generator;a change rate calculator calculating a water level change rate with respect to the detected water level variance;a compensation value calculator calculating a compensation gain value corresponding to the calculated water level change rate;a reactor power sensor sensing whether or not reactor power corresponds to certain power or less; anda gain compensation controller, when the reactor power corresponds to the certain power or less, outputting a control gain value obtained by combining a general gain value provided to control a proportional-integral (PI) controller with the compensation gain value to the PI controller.2. The apparatus of claim 1 , wherein the certain power corresponds to a any value from about 10 to about 30% of a full reactor power.3. The apparatus of claim 1 , wherein the gain compensation controller claim 1 , when the reactor power corresponds to the ...

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16-04-2015 дата публикации

Method And System For Determining Power Plant Machine Reliability

Номер: US20150106059A1
Принадлежит: General Electric Co

Disclosed are methods and systems to determine a power plant machine reliability forecast. In an embodiment, a method may comprise obtaining an environmental factor of a power plant machine based on geospatial data of a first area and location data of a second area, obtaining an operating factor of the power plant machine, and determining a reliability forecast based on the obtained environmental factor and the obtained operating factor.

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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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03-07-2014 дата публикации

Power monitoring system for a nuclear reactor

Номер: US20140185727A1
Принадлежит: Toshiba Corp

According to an embodiment, a power monitoring system for a nuclear reactor comprises at least a first system and second system. The first system and the second system respectively comprise a plurality of APRM units, a plurality of FLOW units, and a plurality of OPRM units. The APRM units respectively generate an LPRM signal that indicates the local output of neutrons by the reactor core, and generate an APRM signal indicating the average output of the reactor core, based on the LPRM signal. The FLOW units respectively generate a FLOW signal indicating the flow rate of reactor coolant. The OPRM units respectively are supplied with the LPRM signal and the APRM signal from at least two aforementioned APRM units and are supplied with the FLOW signal from at least one aforementioned FLOW unit; and, based on the supplied LPRM signals, APRM signals and FLOW signals, generate a trip signal for shutting down the reactor.

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21-04-2016 дата публикации

Energy storage system

Номер: US20160109185A1

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

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19-04-2018 дата публикации

CONDITION DETERMINATION SYSTEM, CONDITION DETERMINATION METHOD, DECISION-MAKING SUPPORT SYSTEM, COMPUTER PROGRAM, AND STORAGE MEDIUM

Номер: US20180107934A1
Принадлежит:

A condition determination system includes: an operation condition data obtaining unit that obtains operation condition data indicating an operation condition of a facility; and a determination unit that determines, based on the operation condition data, a level of a phenomenon that occurs due to the operation condition of the facility. 1. (canceled)2. A condition determination system , comprising:an operation condition data obtaining unit that obtains operation condition data indicating an operation condition of a facility;a determination unit that determines, based on the operation condition data, a level of a phenomenon that occurs due to the operation condition of the facility; anda storage unit that stores therein relation data indicating a relation between the operation condition of the facility and the level of the phenomenon, whereinthe determination unit determines the level of the phenomenon, based on the operation condition data and the relation data.3. The condition determination system according to claim 2 , comprising: a display control unit that generates display data claim 2 , based on a result of the determination by the determination unit claim 2 , and causes a display unit to display thereon the display data.4. The condition determination system according to claim 3 , whereinthe operation condition data obtaining unit obtains the operation condition data of each of plural facilities,the determination unit classifies, based on the plural sets of operation condition data, the plural facilities into normal facilities and abnormal facilities, andthe display control unit causes the display device to display thereon the normal facilities and abnormal facilities in different designs.5. The condition determination system according to claim 4 , whereinthe facilities include a storage battery power source,the operation condition data include remaining capacity data indicating a remaining battery capacity of the storage battery power source, andthe display ...

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02-04-2020 дата публикации

SELF-POWERED IN-CORE DETECTOR ARRANGEMENT FOR MEASURING FLUX IN A NUCLEAR REACTOR CORE

Номер: US20200105426A1
Принадлежит:

A self-powered in-core detector arrangement for measuring flux in a nuclear reactor core includes a first in-core detector and a second in-core detector. The first in-core detector includes a first flux detecting material, a first lead wire extending longitudinally from a first axial end of the first flux detecting material, a first insulating material surrounding outer diameters of the first flux detecting material and the first lead wire and a first sheath surrounding the first insulating material. The first sheath includes a first section surrounding the first flux detecting material and a second section surrounding the first lead wire. The first section of the first sheath has a greater outer diameter than the second section of the first sheath. The second in-core detector includes a second flux detecting material, a second lead wire extending longitudinally from a first axial end of the second flux detecting material, a second insulating material surrounding outer diameters of the second flux detecting material and the second lead wire, and a second sheath surrounding the second insulating material. The second sheath includes a first section surrounding the second flux detecting material and a second section surrounding the second lead wire. The first section of the second sheath has a greater outer diameter than the second section of the second sheath. The first section of the first sheath is axially offset from the first section of the second sheath and radially aligned with the second section of second sheath. 1. A self-powered in-core detector arrangement for measuring flux in a nuclear reactor core comprising: a first flux detecting material;', 'a first lead wire extending longitudinally from a first axial end of the first flux detecting material;', 'a first insulating material surrounding outer diameters of the first flux detecting material and the first lead wire; and', 'a first sheath surrounding the first insulating material, the first sheath including ...

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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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28-04-2016 дата публикации

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

Номер: US20160118149A1
Автор: Axel Hill
Принадлежит: AREVA GMBH

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

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27-04-2017 дата публикации

MOVEMENT OF FUEL TUBES WITHIN AN ARRAY

Номер: US20170117065A1
Автор: Scott Ian Richard
Принадлежит:

A method of operating a nuclear fission reactor. The reactor comprises a reactor core, and a coolant tank containing coolant, the reactor core comprises an array of fuel assemblies. Each fuel assembly extends generally vertically and comprises one or more fuel tubes containing fissile fuel. The fuel tubes are immersed in the coolant. The method comprises monitoring and/or modelling fuel concentrations and/or fission rates in each of the fuel assemblies; and in dependence upon results of the monitoring and/or modelling, moving fuel assemblies horizontally within the array, without lifting the fuel tubes from the coolant, in order to control fission rates in the reactor core. A nuclear reactor implementing the method, and fuel assemblies for use in the method are also disclosed. 125.-. (canceled)26. A method of operating a nuclear fission reactor , the reactor comprising a reactor core , and a coolant tank containing coolant , the reactor core comprising an array of fuel assemblies , each fuel assembly extending generally vertically and comprising one or more fuel tubes containing fissile fuel , the fuel tubes being immersed in the coolant , the method comprising:monitoring and/or modelling fuel concentrations and/or fission rates in each of the fuel assemblies;in dependence upon results of the monitoring and/or modelling, moving fuel assemblies horizontally within the array, without lifting any of the fuel assemblies from the array of fuel assemblies, in order to control fission rates in the reactor core;moving a spent fuel assembly to a horizontal periphery of the array without lifting the fuel assembly from the array by moving a row or part row of the array of fuel assemblies towards the periphery of the array; andremoving the spent fuel assembly from the horizontal periphery of the array into a spent fuel storage area within the coolant tank without lifting the spent fuel assembly from the coolant.27. A method according to claim 26 , and comprising:monitoring ...

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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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25-04-2019 дата публикации

MANAGING WATER-SUPPLY PUMPING FOR AN ELECTRICITY PRODUCTION PLANT CIRCUIT

Номер: US20190120087A1
Принадлежит:

A method for assisting with the management of a pumping device capable of supplying a circuit of a power production facility with water taken from a natural watercourse is disclosed here. In particular, at least some parameters relating to a watercourse and having an influence on the quantity of materials liable to clog filters are collected. A statistical model is developed that is at least based on historical data for said parameters relating to the watercourse for which clogging of the filters has been observed. Current parameters relating to at least the watercourse are collected and said statistical model is used in conjunction with said current parameters to assess the risk of an influx of clogging materials, and an alert signal for deactivating the pumping device at a selected time is generated on the basis of the assessed risk. 1. A method for assisting with the management of a pumping device capable of supplying a circuit of a power production facility with water taken from a natural watercourse , the water upstream of the circuit comprising materials liable to clog one or more filters provided at the inlet to the circuit , wherein at least parameters relating to the watercourse and having an influence on the quantity of materials likely to clog said filters are collected , and:during a previous step, a statistical model is developed that is at least based on historical data for said parameters relating to the watercourse for which clogging of the filters has been observed,during a current step, current parameters relating to at least the watercourse are collected and said statistical model is used in conjunction with said current parameters to assess a risk of an influx of clogging materials, andan alert signal for deactivating the pumping device at a selected time is generated on the basis of the assessed risk.2. The method according to claim 1 , wherein the alert signal is generated for the further purpose of deactivating the pumping device for a ...

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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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28-05-2015 дата публикации

Nuclear reactor power regulator

Номер: US20150146836A1
Принадлежит: Toshiba Corp

A reactor power regulator that adjusts an output of a reactor on a basis of an operation pattern or a reactor output target value and a reactor output change rate that are input by a central load dispatching center or an operator, including: a reactor output calculating device that performs computation based on a thermal equilibrium from various necessary plant state quantities to calculate a reactor output signal; a correcting device that corrects a continuously obtained reactor output equivalent signal that is considered to be equivalent to a reactor output at a calculation interval of the reactor output signal, for each calculation interval in the reactor output calculating device so that the reactor output equivalent signal coincides with the reactor output signal calculated by the reactor output calculating device, and calculates a continuous corrected reactor output equivalent signal; a reactor output controlling device that calculates at least one type of reactor output control signal for controlling the output of the reactor, using the corrected reactor output equivalent signal, the reactor output target value, and the reactor output change rate; and a reactor output controller that is actuated on a basis of the reactor output control signal.

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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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09-05-2019 дата публикации

SYSTEM FOR SETTING TOLERANCE LIMIT OF CORRELATION BY USING REPETITIVE CROSS-VALIDATION AND METHOD THEREOF

Номер: US20190139659A1
Принадлежит: KEPCO NUCLEAR FUEL CO., LTD.

A correlation tolerance limit setting system using repetitive cross-validation includes: a variable extraction unit randomly classifying data of an initial DB set into training set data and validation set data at a specific rate and then extracting variables for determining a DNBR limit by optimizing coefficients of a selected correlation; a normality test unit testing normality for a variable extraction result; a DNBR limit unit determining whether data sets have a same population or not depending on normality result and determining DNBR limit from a distribution of 95/95 DNBR; and a controller. 1. A correlation tolerance limit setting system using repetitive cross-validation , the system comprising:{'b': '100', 'a variable extraction unit () randomly classifying data of an initial DB set into training data and validation data at a specific rate, and then matching each of the training data and the validation data with each run ID assigned thereto and storing them in a training initial set and a validation initial set, respectively, thereby extracting variables for determining a departure from nucleate boiling ratio (DNBR) limit by optimizing coefficients of a selected correlation based on the data stored in the training initial set;'}{'b': '200', 'a normality test unit () performing a normality test for data of a training set and data of a validation set after extracting the variables; and'}{'b': '300', 'a DNBR limit unit () determining the DNBR limit by a parametric method or a nonparametric method depending on a result of the normality test,'}{'b': '300', 'wherein the DNBR limit unit () includes{'b': '310', 'an output module () determining whether the data of the training set and the data of the validation set have a same population or not by the parametric method or the nonparametric method depending on a result of the normality test for an M/P of a run ID of the training set and an M/P of a run ID of the validation set derived for each case, and performing the ...

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30-04-2020 дата публикации

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

Номер: US20200135352A1
Принадлежит:

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

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30-04-2020 дата публикации

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

Номер: US20200135353A1
Принадлежит:

An electro-technical device includes a first coil connected to a first sensor for receiving a current therefrom representative of a sensed condition, the first coil being anchored at first and second ends. A second coil is connected to a second sensor for receiving a current therefrom representative of a sensed condition, the second coil being anchored at first and second ends and being adjacent to the first coil. When the first and second coils receive an increased current from the first and second sensors, the first and second coils each create a magnetic flux that repel one another in order to cause at least one of the coils to break so that a shutdown signal can be sent. 1. An electro-technical device , comprising:a first coil connected to a first sensor for receiving a current therefrom representative a sensed condition, the first coil being anchored at first and second ends;a second coil connected to a second sensor for receiving a current therefrom representative of a sensed condition, the second coil being anchored at first and second ends and having the first end adjacent to the first end of the first coil;wherein when the first and second coils receive an increased current from the first and second sensors, the first and second coils each create a magnetic flux that repel one another in order to cause at least one of the coils to break.2. The electro-technical device according to claim 1 , wherein the first coil is connected to the first sensor at the second end and is connected to a first output line at the first end.3. The electro-technical device according to claim 2 , wherein the second coil is connected to the second sensor at the second end and is connected to a second output line at the first end.4. The electro-technical device according to claim 1 , wherein the first and second coils are anchored at the first and second ends by first and second plates claim 1 , respectively.5. An electro-technical device claim 1 , comprising:a first coil connected ...

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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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24-05-2018 дата публикации

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

Номер: US20180144836A1
Автор: Lin-Hendel Catherine
Принадлежит:

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

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15-09-2022 дата публикации

Online monitoring method of nuclear power plant system based on isolation forest method and sliding window method

Номер: US20220291654A1
Принадлежит: Harbin Engineering University

The present disclosure relates to an online monitoring method of a nuclear power plant system based on an isolation forest method and a sliding window method. An isolation forest method used in the present disclosure is an abnormal detection model based on the idea of binary tree division, and has no requirements on the dimension and linear characteristics of monitoring data. In view of the characteristics of strong nonlinearity and high dimension of operation data of the nuclear power plant system, in the process of state monitoring, system abnormalities can be detected more quickly and accurately. In the present disclosure, a sliding window method is used to improve an isolation forest model, so that the improved isolation forest model has the functions of model online updating and real-time state monitoring, and the usability of an isolation forest state monitoring method is enhanced.

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07-05-2020 дата публикации

COOLING FACILITY IN A REACTOR AND ELECTRIC POWER GENERATION SYSTEM

Номер: US20200141351A1
Принадлежит:

A reactor cooling and power generation system according to the present disclosure may include a reactor vessel, a heat exchange section formed to receive heat generated from a core inside the reactor vessel through a fluid, and an electric power production section including a Sterling engine formed to produce electric energy using the energy of the fluid whose temperature has increased while receiving the heat of the reactor, wherein the system is formed to circulate the fluid that has received heat from the core in the heat exchange section through the electric power production section, and operate even during a normal operation and during an accident of the nuclear power plant to produce electric power. 1. A reactor cooling and power generation system , the system , comprising:a reactor vessel;a heat exchange section formed to receive heat generated from a core inside the reactor vessel through a fluid; andan electric power production section comprising a Sterling engine formed to produce electric energy using the energy of the fluid whose temperature has increased while receiving the heat of the reactor,wherein the system is formed to circulate the fluid that has received heat from the core in the heat exchange section through the electric power production section, and operate even during a normal operation and during an accident of the nuclear power plant to produce electric power.2. The system of claim 1 , wherein the electric power produced during the normal operation of the nuclear power plant is supplied to an internal and external electric power system and an emergency battery.3. The system of claim 2 , wherein the electric energy charged in the emergency battery is formed to supply an emergency electric power as an emergency power source during an accident.4. The system of claim 1 , wherein the electric power produced during an accident of the nuclear power plant is formed to be supplied to an emergency power source of the nuclear power plant.5. The system ...

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11-06-2015 дата публикации

Hydrogen Concentration Meter

Номер: US20150160163A1
Принадлежит: Individual

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

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