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Сборка активной зоны экспериментального реактора-размножителя I в Айдахо , США , 1951 г.

Реактор - размножитель представляет собой ядерный реактор , который генерирует более делящийся материал , чем потребляет . [1] Реакторы-размножители достигают этого, потому что их нейтронная экономия достаточно высока для создания большего количества делящегося топлива, чем они используют, путем облучения воспроизводящего материала , такого как уран-238 или торий-232 , который загружается в реактор вместе с делящимся топливом. . Поначалу селекционеры были сочтены привлекательными, потому что они более полно использовали урановое топливо, чем легководные реакторы , но после 1960-х годов интерес снизился, поскольку были обнаружены новые запасы урана [2]а новые методы обогащения урана снизили затраты на топливо.

Топливная эффективность и типы ядерных отходов [ править ]

Вероятности деления выбранных актинидов, тепловые нейтроны в сравнении с быстрыми [3] [4] [5] [6] [7]
Изотоп
Сечение теплового деления		Тепловое деление%
Сечение быстрого деления		Быстрое деление%
Чт-232ноль1 (неделящийся)0,350 сарай3 (неделящийся)
U-23276,66 сарай592.370 сарай95
U-233531,2 сарай892.450 сарай93
U-235584,4 сарай81 год2.056 сарай80
U-23811,77 микробар1 (неделящийся)1.136 сарай11
НП-2370,02249 сарай3 (неделящийся)2.247 сарай27
Pu-23817,89 сарай72.721 сарай70
Pu-239747,4 сарай632.338 сарай85
Пу-24058,77 сарай1 (неделящийся)2.253 сарай55
Пу-2411012 сарай752.298 сарай87
Пу-2420,002557 сарай1 (неделящийся)2.027 сарай53
Ам-241600,4 сарай1 (неделящийся)0,2299 микробар21 год
АМ-242М6409 сарай752.550 сарай94
Ам-2430,1161 сарай1 (неделящийся)2.140 сарай23
См-2425.064 сарай1 (неделящийся)2.907 сарай10
См-243617,4 сарай782.500 сарай94
См-2441.037 сарай4 (неделящийся)0,08255 микробар33

Реакторы-размножители могут, в принципе, извлекать почти всю энергию, содержащуюся в уране или тории , снижая потребность в топливе в 100 раз по сравнению с широко используемыми прямоточными легководными реакторами , которые извлекают менее 1% энергии урана. добывается из земли. [8] Высокая топливная эффективность реакторов-размножителей может значительно снизить озабоченность по поводу поставок топлива, энергии, используемой при добыче и хранении радиоактивных отходов. Приверженцы утверждают, что при добыче урана из морской воды для реакторов-размножителей будет достаточно топлива, чтобы удовлетворить наши потребности в энергии в течение 5 миллиардов лет при общем уровне энергопотребления 1983 года, что сделает атомную энергию эффективно возобновляемой.. [9] [10]

К 1990-м годам ядерные отходы стали вызывать большую озабоченность. В широком смысле отработанное ядерное топливо состоит из двух основных компонентов. Первый состоит из продуктов деления , оставшихся фрагментов атомов топлива после того, как они были расщеплены для высвобождения энергии. Продукты деления состоят из десятков элементов и сотен изотопов, все они легче урана. Второй основной компонент отработавшего топлива - это трансурановые соединения (атомы тяжелее урана), которые образуются из урана или более тяжелых атомов в топливе, когда они поглощают нейтроны, но не подвергаются делению. Все трансурановые изотопы попадают в серию актинидов Периодической таблицы , поэтому их часто называют актинидами.

Физическое поведение продуктов деления заметно отличается от поведения трансурановых соединений. В частности, продукты деления сами по себе не подвергаются делению и поэтому не могут быть использованы в ядерном оружии. Кроме того, только семь долгоживущих изотопов продуктов деления имеют период полураспада более ста лет, что делает их геологическое хранение или захоронение менее проблематичным, чем для трансурановых материалов. [11]

В связи с возросшей обеспокоенностью по поводу ядерных отходов, воспроизводящие топливные циклы снова стали интересными, поскольку они могут уменьшить количество отходов актинидов, особенно плутония и минорных актинидов . [12] Реакторы-размножители предназначены для расщепления актинидных отходов в качестве топлива и, таким образом, их преобразования в большее количество продуктов деления.

После того, как отработавшее ядерное топливо удалено из легководного реактора, оно претерпевает сложный профиль распада, поскольку каждый нуклид распадается с разной скоростью. Из-за физической странности, упомянутой ниже, существует большой разрыв в периодах полураспада продуктов деления по сравнению с трансурановыми изотопами. Если трансурановые элементы останутся в отработанном топливе через 1000–100000 лет, медленный распад этих трансурановых элементов приведет к возникновению большей части радиоактивности в этом отработанном топливе. Таким образом, удаление трансурановых элементов из отходов устраняет большую часть долговременной радиоактивности отработавшего ядерного топлива. [13]

Сегодняшние коммерческие легководные реакторы действительно создают некоторый новый делящийся материал, в основном в форме плутония. Поскольку коммерческие реакторы никогда не проектировались как воспроизводящие, они не преобразуют достаточно урана-238 в плутоний, чтобы заменить потребляемый уран-235 . Тем не менее, по крайней мере, одна треть энергии, производимой коммерческими ядерными реакторами, происходит за счет деления плутония, образующегося в топливе. [14] Даже при таком уровне потребления плутония легководные реакторы потребляют только часть производимого ими плутония и второстепенных актинидов, а неделящиеся изотопы плутония накапливаются вместе со значительными количествами других второстепенных актинидов. [15]

Коэффициент конверсии, безубыточность, коэффициент размножения, время удвоения и выгорание [ править ]

Одним из показателей производительности реактора является «коэффициент конверсии», определяемый как отношение произведенных новых делящихся атомов к израсходованным делящимся атомам. Все предлагаемые ядерные реакторы, за исключением специально разработанных и эксплуатируемых актинидных горелок [16], претерпевают некоторую степень конверсии. Пока в нейтронном потоке реактора есть какое-либо количество воспроизводящего материала, всегда создается новый делящийся материал. Когда коэффициент преобразования больше 1, его часто называют «коэффициентом разведения».

Например, обычно используемые легководные реакторы имеют коэффициент конверсии приблизительно 0,6. Реакторы с тяжелой водой под давлением ( PHWR ), работающие на природном уране, имеют коэффициент конверсии 0,8. [17] В реакторе-размножителе коэффициент конверсии выше 1. «Безубыточность» достигается, когда коэффициент конверсии достигает 1,0 и реактор производит столько делящегося материала, сколько он использует.

Время удвоения - это время, которое потребуется реактору-размножителю для производства достаточного количества нового расщепляющегося материала для замены исходного топлива и дополнительного производства эквивалентного количества топлива для другого ядерного реактора. В первые годы, когда считалось, что урана в дефиците, это считалось важным показателем производительности селекционеров. Однако, поскольку урана больше, чем предполагалось в первые дни разработки ядерных реакторов, и с учетом количества плутония, доступного в отработавшем топливе реактора, время удвоения стало менее важным показателем в современной конструкции реактора-размножителя. [18] [19]

« Выгорание » - это мера того, сколько энергии было извлечено из заданной массы тяжелого металла в топливе, часто выражаемое (для энергетических реакторов) в гигаватт-днях на тонну тяжелого металла. Выгорание является важным фактором при определении типов и количества изотопов, производимых в реакторе деления. Реакторы-размножители по своей конструкции имеют чрезвычайно высокое выгорание по сравнению с обычными реакторами, поскольку реакторы-размножители производят гораздо больше своих отходов в виде продуктов деления, в то время как большая часть или все актиниды предназначены для деления и уничтожения. [20]

В прошлом при разработке реакторов-размножителей основное внимание уделялось реакторам с низким коэффициентом воспроизводства, от 1,01 для реактора Шиппорт [21] [22], работающего на ториевом топливе и охлаждаемого обычной легкой водой, до более 1,2 для советского жидкометаллического БН-350. -охлаждаемый реактор. [23] Теоретические модели размножителей с жидким натриевым теплоносителем, текущим через трубки внутри топливных элементов (конструкция «труба в оболочке»), предполагают, что в промышленном масштабе возможны коэффициенты воспроизводства не менее 1,8. [24] Советский испытательный реактор БР-1 достиг коэффициента воспроизводства 2,5 в некоммерческих условиях. [25]

Типы реакторов-размножителей [ править ]

Производство тяжелых трансурановых актинидов в современных реакторах деления на тепловых нейтронах путем захвата и распада нейтронов. Начиная с урана-238, производятся изотопы плутония, америция и кюрия. В реакторе-размножителе на быстрых нейтронах все эти изотопы могут сжигаться в качестве топлива.

Возможны многие типы реакторов-размножителей:

«Размножитель» - это просто реактор, спроектированный для очень высокой экономии нейтронов с соответствующей степенью конверсии выше 1,0. В принципе, почти любую конструкцию реактора можно изменить, чтобы она стала размножающей. Примером этого процесса является эволюция легководного реактора, тепловой конструкции с очень сильным замедлителем, в концепцию сверхбыстрого реактора [26] , использующего легкую воду в сверхкритической форме с чрезвычайно низкой плотностью для увеличения нейтронной экономии, достаточно высокой для того, чтобы разрешить разведение.

Помимо водяного охлаждения, в настоящее время рассматривается множество других типов реакторов-размножителей. К ним относятся конструкции, охлаждаемые расплавом соли , газом и жидким металлом, во многих вариантах. Практически любой из этих основных типов конструкции может работать на уране, плутонии, многих второстепенных актинидах или тории, и они могут быть разработаны для множества различных целей, таких как создание большего количества делящегося топлива, длительная стационарная работа или активное сжигание. ядерных отходов.

Существующие конструкции реакторов иногда делятся на две широкие категории в зависимости от их нейтронного спектра, который обычно отделяет реакторы, предназначенные для использования в основном урана и трансурановых элементов, от реакторов, предназначенных для использования тория и избегания трансурановых элементов. Эти конструкции:

  • Реактор-размножитель на быстрых нейтронах (FBR), в котором используются быстрые (т.е. немодерированные) нейтроны для получения делящегося плутония и, возможно, более трансурановых соединений из плодородного урана-238. Быстрый спектр достаточно гибкий, чтобы при желании он также мог выделять делящийся уран-233 из тория.
  • Тепловой реактор-размножитель, который использует нейтроны теплового спектра (то есть замедленные) для выделения делящегося урана-233 из тория ( ториевый топливный цикл ). Из-за поведения различных видов ядерного топлива считается, что термический размножитель коммерчески возможен только с ториевым топливом, что позволяет избежать накопления более тяжелых трансурановых элементов.

Повторная обработка [ править ]

При делении ядерного топлива в любом реакторе образуются поглощающие нейтроны продукты деления . Из-за этого неизбежного физического процесса необходимо переработать плодородный материал из реактора-размножителя, чтобы удалить эти нейтронные яды . Этот шаг необходим для того, чтобы в полной мере использовать способность воспроизводить столько или больше топлива, сколько потребляется. Любая переработка может вызвать опасения с точки зрения распространения , поскольку она позволяет извлекать оружейный материал из отработавшего топлива. [27] Самый распространенный метод обработки, PUREX., представляет особую озабоченность, поскольку он был специально разработан для отделения чистого плутония. Ранние предложения по топливному циклу реактора-размножителя вызвали еще большую озабоченность с точки зрения распространения, поскольку в них использовался PUREX для выделения плутония в очень привлекательной изотопной форме для использования в ядерном оружии. [28] [29]

Некоторые страны разрабатывают методы переработки, которые не позволяют отделить плутоний от других актинидов. Например, пирометаллургический процесс электровыделения без использования воды , когда он используется для переработки топлива встроенного быстрого реактора , оставляет большие количества радиоактивных актинидов в топливе реактора. [8] К более традиционным системам переработки на водной основе относятся SANEX, UNEX, DIAMEX, COEX и TRUEX, а также предложения по объединению PUREX с совместными процессами.

Все эти системы имеют немного лучшую устойчивость к распространению, чем PUREX, хотя скорость их внедрения невысока. [30] [31] [32]

В ториевом цикле торий-232 размножается, сначала превращаясь в протактиний-233, который затем распадается до урана-233. Если протактиний остается в реакторе, также образуются небольшие количества урана-232, который имеет в своей цепочке распада сильный гамма-излучатель таллий-208 . Как и в случае конструкций с урановым топливом, чем дольше топливо и воспроизводящий материал остаются в реакторе, тем больше накапливается этих нежелательных элементов. В предполагаемых коммерческих ториевых реакторахбудет допущено накопление высоких уровней урана-232, что приведет к чрезвычайно высоким дозам гамма-излучения от любого урана, полученного из тория. Эти гамма-лучи усложняют безопасное обращение с оружием и конструкцию его электроники; это объясняет, почему уран-233 никогда не использовался для производства оружия, кроме демонстрации концепции. [33]

Хотя ториевый цикл может быть устойчивым с точки зрения распространения в отношении извлечения урана-233 из топлива (из-за присутствия урана-232), он представляет опасность распространения из-за альтернативного пути извлечения урана-233, который включает химическое извлечение протактиния. 233 и позволяя ему распадаться до чистого урана-233 вне реактора. Этот процесс может происходить вне контроля таких организаций, как Международное агентство по атомной энергии (МАГАТЭ). [34]

Уменьшение отходов [ править ]

Актиниды и продукты деления по периоду полураспада
  • v
  • т
  • е
Актиниды [35] по цепочке распадаПериод полураспада
( а )		Продукты деления из 235 U по доходности [36]
4n4n+14n+24n+3
4.5–7%0.04–1.25%<0.001%
228Ra№4–6 a†155Euþ
244Cmƒ241Puƒ250Cf227Ac№10–29 a90Sr85Kr113mCdþ
232Uƒ238Puƒ243Cmƒ29–97 a137Cs151Smþ121mSn
248Bk[37]249Cfƒ242mAmƒ141–351 a

No fission products
have a half-life
in the range of
100–210 ka ...

241Amƒ251Cfƒ[38]430–900 a
226Ra№247Bk1.3–1.6 ka
240Pu229Th246Cmƒ243Amƒ4.7–7.4 ka
245Cmƒ250Cm8.3–8.5 ka
239Puƒ24.1 ka
230Th№231Pa№32–76 ka
236Npƒ233Uƒ234U№150–250 ka‡99Tc₡126Sn
248Cm242Pu327–375 ka79Se₡
1.53 Ma93Zr
237Npƒ2.1–6.5 Ma135Cs₡107Pd
236U247Cmƒ15–24 Ma129I₡
244Pu80 Ma

... nor beyond 15.7 Ma[39]

232Th№238U№235Uƒ№0.7–14.1 Ga

Legend for superscript symbols
₡  has thermal neutron capture cross section in the range of 8–50 barns
ƒ  fissile
m  metastable isomer
№  primarily a naturally occurring radioactive material (NORM)
þ  neutron poison (thermal neutron capture cross section greater than 3k barns)
†  range 4–97 a: Medium-lived fission product
‡  over 200 ka: Long-lived fission product

Nuclear waste became a greater concern by the 1990s. Breeding fuel cycles attracted renewed interest because of their potential to reduce actinide wastes, particularly plutonium and minor actinides.[12] Since breeder reactors on a closed fuel cycle would use nearly all of the actinides fed into them as fuel, their fuel requirements would be reduced by a factor of about 100. The volume of waste they generate would be reduced by a factor of about 100 as well. While there is a huge reduction in the volume of waste from a breeder reactor, the activity of the waste is about the same as that produced by a light-water reactor.[40]

In addition, the waste from a breeder reactor has a different decay behavior, because it is made up of different materials. Breeder reactor waste is mostly fission products, while light-water reactor waste has a large quantity of transuranics. After spent nuclear fuel has been removed from a light-water reactor for longer than 100,000 years, these transuranics would be the main source of radioactivity. Eliminating them would eliminate much of the long-term radioactivity from the spent fuel.[13]

In principle, breeder fuel cycles can recycle and consume all actinides,[9] leaving only fission products. As the graphic in this section indicates, fission products have a peculiar 'gap' in their aggregate half-lives, such that no fission products have a half-life between 91 years and two hundred thousand years. As a result of this physical oddity, after several hundred years in storage, the activity of the radioactive waste from a Fast Breeder Reactor would quickly drop to the low level of the long-lived fission products. However, to obtain this benefit requires the highly efficient separation of transuranics from spent fuel. If the fuel reprocessing methods used leave a large fraction of the transuranics in the final waste stream, this advantage would be greatly reduced.[8]

Both types of breeding cycles can reduce actinide wastes:

  • The fast breeder reactor's fast neutrons can fission actinide nuclei with even numbers of both protons and neutrons. Such nuclei usually lack the low-speed "thermal neutron" resonances of fissile fuels used in LWRs.[41]
  • The thorium fuel cycle inherently produces lower levels of heavy actinides. The fertile material in the thorium fuel cycle has an atomic weight of 232, while the fertile material in the uranium fuel cycle has an atomic weight of 238. That mass difference means that thorium-232 requires six more neutron capture events per nucleus before the transuranic elements can be produced. In addition to this simple mass difference, the reactor gets two chances to fission the nuclei as the mass increases: First as the effective fuel nuclei U233, and as it absorbs two more neutrons, again as the fuel nuclei U235.[42][43]

A reactor whose main purpose is to destroy actinides, rather than increasing fissile fuel-stocks, is sometimes known as a burner reactor. Both breeding and burning depend on good neutron economy, and many designs can do either. Breeding designs surround the core by a breeding blanket of fertile material. Waste burners surround the core with non-fertile wastes to be destroyed. Some designs add neutron reflectors or absorbers.[16]

Breeder reactor concepts[edit]

There are several concepts for breeder reactors; the two main ones are:

  • Reactors with a fast neutron spectrum are called fast breeder reactors (FBR) – these typically utilize uranium-238 as fuel.
  • Reactors with a thermal neutron spectrum are called thermal breeder reactors – these typically utilize thorium-232 as fuel.

Fast breeder reactor[edit]

Schematic diagram showing the difference between the Loop and Pool types of LMFBR.

In 2006 all large-scale fast breeder reactor (FBR) power stations were liquid metal fast breeder reactors (LMFBR) cooled by liquid sodium. These have been of one of two designs:[1]

  • Loop type, in which the primary coolant is circulated through primary heat exchangers outside the reactor tank (but inside the biological shield due to radioactive sodium-24 in the primary coolant)
Experimental Breeder Reactor II, which served as the prototype for the Integral Fast Reactor
  • Pool type, in which the primary heat exchangers and pumps are immersed in the reactor tank

All current fast neutron reactor designs use liquid metal as the primary coolant, to transfer heat from the core to steam used to power the electricity generating turbines. FBRs have been built cooled by liquid metals other than sodium—some early FBRs used mercury, other experimental reactors have used a sodium-potassium alloy called NaK. Both have the advantage that they are liquids at room temperature, which is convenient for experimental rigs but less important for pilot or full-scale power stations. Lead and lead-bismuth alloy have also been used.

Three of the proposed generation IV reactor types are FBRs:[44]

  • Gas-cooled fast reactor (GFR) cooled by helium.
  • Sodium-cooled fast reactor (SFR) based on the existing liquid-metal FBR (LMFBR) and integral fast reactor designs.
  • Lead-cooled fast reactor (LFR) based on Soviet naval propulsion units.

FBRs usually use a mixed oxide fuel core of up to 20% plutonium dioxide (PuO2) and at least 80% uranium dioxide (UO2). Another fuel option is metal alloys, typically a blend of uranium, plutonium, and zirconium (used because it is "transparent" to neutrons). Enriched uranium can also be used on its own.

Many designs surround the core in a blanket of tubes that contain non-fissile uranium-238, which, by capturing fast neutrons from the reaction in the core, converts to fissile plutonium-239 (as is some of the uranium in the core), which is then reprocessed and used as nuclear fuel. Other FBR designs rely on the geometry of the fuel itself (which also contains uranium-238), arranged to attain sufficient fast neutron capture. The plutonium-239 (or the fissile uranium-235) fission cross-section is much smaller in a fast spectrum than in a thermal spectrum, as is the ratio between the 239Pu/235U fission cross-section and the 238U absorption cross-section. This increases the concentration of 239Pu/235U needed to sustain a chain reaction, as well as the ratio of breeding to fission.[16]On the other hand, a fast reactor needs no moderator to slow down the neutrons at all, taking advantage of the fast neutrons producing a greater number of neutrons per fission than slow neutrons. For this reason ordinary liquid water, being a moderator and neutron absorber, is an undesirable primary coolant for fast reactors. Because large amounts of water in the core are required to cool the reactor, the yield of neutrons and therefore breeding of 239Pu are strongly affected. Theoretical work has been done on reduced moderation water reactors, which may have a sufficiently fast spectrum to provide a breeding ratio slightly over 1. This would likely result in an unacceptable power derating and high costs in a liquid-water-cooled reactor, but the supercritical water coolant of the supercritical water reactor (SCWR) has sufficient heat capacity to allow adequate cooling with less water, making a fast-spectrum water-cooled reactor a practical possibility.[26]

The type of coolants, temperatures and fast neutron spectrum puts the fuel cladding material (normally austenitic stainless or ferritic-martensitic steels) under extreme conditions. The understanding of the radiation damage, coolant interactions, stresses and temperatures are necessary for the safe operation of any reactor core. All materials used to date in sodium-cooled fast reactors have known limits, as explored in ONR-RRR-088 review.[45] Oxide Dispersion Strengthened (ODS) steel is viewed as the long-term radiation resistant fuel-cladding material that overcome the shortcomings of today's material choices.

There are only two commercially operating breeder reactors as of 2017: the BN-600 reactor, at 560 MWe, and the BN-800 reactor, at 880 MWe. Both are Russian sodium-cooled reactors.

Integral fast reactor[edit]

Main article: Integral Fast Reactor

One design of fast neutron reactor, specifically conceived to address the waste disposal and plutonium issues, was the integral fast reactor (IFR, also known as an integral fast breeder reactor, although the original reactor was designed to not breed a net surplus of fissile material).[46][47]

To solve the waste disposal problem, the IFR had an on-site electrowinning fuel-reprocessing unit that recycled the uranium and all the transuranics (not just plutonium) via electroplating, leaving just short half-life fission products in the waste. Some of these fission products could later be separated for industrial or medical uses and the rest sent to a waste repository. The IFR pyroprocessing system uses molten cadmium cathodes and electrorefiners to reprocess metallic fuel directly on-site at the reactor.[48] Such systems not only co-mingle all the minor actinides with both uranium and plutonium, they are compact and self-contained, so that no plutonium-containing material needs to be transported away from the site of the breeder reactor. Breeder reactors incorporating such technology would most likely be designed with breeding ratios very close to 1.00, so that after an initial loading of enriched uranium and/or plutonium fuel, the reactor would then be refueled only with small deliveries of natural uranium metal. A quantity of natural uranium metal equivalent to a block about the size of a milk crate delivered once per month would be all the fuel such a 1 gigawatt reactor would need.[49] Such self-contained breeders are currently envisioned as the final self-contained and self-supporting ultimate goal of nuclear reactor designers.[8][16] The project was canceled in 1994 by United States Secretary of Energy Hazel O'Leary.[50][51]

Other fast reactors[edit]

The graphite core of the Molten Salt Reactor Experiment

Another proposed fast reactor is a fast molten salt reactor, in which the molten salt's moderating properties are insignificant. This is typically achieved by replacing the light metal fluorides (e.g. LiF, BeF2) in the salt carrier with heavier metal chlorides (e.g., KCl, RbCl, ZrCl4).

Several prototype FBRs have been built, ranging in electrical output from a few light bulbs' equivalent (EBR-I, 1951) to over 1,000 MWe. As of 2006, the technology is not economically competitive to thermal reactor technology, but India, Japan, China, South Korea and Russia are all committing substantial research funds to further development of fast breeder reactors, anticipating that rising uranium prices will change this in the long term. Germany, in contrast, abandoned the technology due to safety concerns. The SNR-300 fast breeder reactor was finished after 19 years despite cost overruns summing up to a total of €3.6 billion, only to then be abandoned.[52]

India is also developing FBR technology using both uranium and thorium feedstocks.[citation needed]

Thermal breeder reactor[edit]

The Shippingport Reactor, used as a prototype light water breeder for five years beginning in August 1977

The advanced heavy water reactor (AHWR) is one of the few proposed large-scale uses of thorium.[53] India is developing this technology, motivated by substantial thorium reserves; almost a third of the world's thorium reserves are in India, which lacks significant uranium reserves.

The third and final core of the Shippingport Atomic Power Station 60 MWe reactor was a light water thorium breeder, which began operating in 1977.[54] It used pellets made of thorium dioxide and uranium-233 oxide; initially, the U-233 content of the pellets was 5–6% in the seed region, 1.5–3% in the blanket region and none in the reflector region. It operated at 236 MWt, generating 60 MWe and ultimately produced over 2.1 billion kilowatt hours of electricity. After five years, the core was removed and found to contain nearly 1.4% more fissile material than when it was installed, demonstrating that breeding from thorium had occurred.[55][56]

The liquid fluoride thorium reactor (LFTR) is also planned as a thorium thermal breeder. Liquid-fluoride reactors may have attractive features, such as inherent safety, no need to manufacture fuel rods and possibly simpler reprocessing of the liquid fuel. This concept was first investigated at the Oak Ridge National Laboratory Molten-Salt Reactor Experiment in the 1960s. From 2012 it became the subject of renewed interest worldwide.[57] Japan, India, China, the UK, as well as private US, Czech and Australian companies have expressed intent to develop and commercialize the technology.[citation needed]

Discussion[edit]

Like many aspects of nuclear power, fast breeder reactors have been subject to much controversy over the years. In 2010 the International Panel on Fissile Materials said "After six decades and the expenditure of the equivalent of tens of billions of dollars, the promise of breeder reactors remains largely unfulfilled and efforts to commercialize them have been steadily cut back in most countries". In Germany, the United Kingdom, and the United States, breeder reactor development programs have been abandoned.[58][59] The rationale for pursuing breeder reactors—sometimes explicit and sometimes implicit—was based on the following key assumptions:[59][60]

  • It was expected that uranium would be scarce and high-grade deposits would quickly become depleted if fission power were deployed on a large scale; the reality, however, is that since the end of the cold war, uranium has been much cheaper and more abundant than early designers expected.[61]
  • It was expected that breeder reactors would quickly become economically competitive with the light-water reactors that dominate nuclear power today, but the reality is that capital costs are at least 25% more than water-cooled reactors.
  • It was thought that breeder reactors could be as safe and reliable as light-water reactors, but safety issues are cited as a concern with fast reactors that use a sodium coolant, where a leak could lead to a sodium fire.
  • It was expected that the proliferation risks posed by breeders and their "closed" fuel cycle, in which plutonium would be recycled, could be managed. But since plutonium-breeding reactors produce plutonium from U238, and thorium reactors produce fissile U233 from thorium, all breeding cycles could theoretically pose proliferation risks.[62] However U232, which is always present in U233 produced in breeder reactors, is a strong gamma-emitter via its daughter products, and would make weapon handling extremely hazardous and the weapon easy to detect.[63]

There are some past anti-nuclear advocates that have become pro-nuclear power as a clean source of electricity since breeder reactors effectively recycle most of their waste. This solves one of the most-important negative issues of nuclear power. In the documentary Pandora's Promise, a case is made for breeder reactors because they provide a real high-kW alternative to fossil fuel energy. According to the movie, one pound of uranium provides as much energy as 5,000 barrels of oil.[64][65]

FBRs have been built and operated in the United States, the United Kingdom, France, the former USSR, India and Japan.[1] The experimental FBR SNR-300 was built in Germany but never operated and eventually shut down amid political controversy following the Chernobyl disaster. As of 2019, two FBRs are being operated for power generation in Russia. Several reactors are planned, many for research related to the Generation IV reactor initiative.[timeframe?][66][67][68]

Development and notable breeder reactors[edit]

Notable breeder reactors[3][69][70][71]
ReactorCountry
when built		StartedShutdownDesign
MWe		Final
MWe		Thermal
Power MWt		CapacityfactorNo of
leaks		NeutrontemperatureCoolantReactor class
DFRUK1962197714116534%7FastNaKTest
BN-350Soviet Union197319991355275043%15FastSodiumPrototype
RapsodieFrance196719830–40–2FastSodiumTest
PhénixFrance1975201023313056340.5%31FastSodiumPrototype
PFRUK1976199423423465026.9%20FastSodiumPrototype
KNK IIGermany1977199118175817.1%21FastSodiumResearch/Test
SNR-300Germany1985 (partial operation)1991327––––FastSodiumPrototype/Commercial
BN-600Soviet Union1981operating560560147074.2%27FastSodiumPrototype/Commercial(Gen2)
FFTFUS198219930–400–1FastSodiumTest
SuperphénixFrance198519981200120030007.9%7FastSodiumPrototype/Commercial(Gen2)
FBTRIndia1985operating13–40–6FastSodiumTest
PFBRIndiacommissioningcommissioning500–1250––FastSodiumPrototype/Commercial(Gen3)
JōyōJapan1977operating0–150––FastSodiumTest
MonjuJapan19952017246246714trial only1FastSodiumPrototype
BN-800Russia2015operating789880210073.4%–FastSodiumPrototype/Commercial(Gen3)
MSREUS196519690–7.4––EpithermalMolten Salt(FLiBe)Test
ClementineUS194619520–0.025––FastMercuryWorld's First Fast Reactor
EBR-1US195119640.20.21.4––FastNaKWorld's First Power Reactor
Fermi-1US196319726666200––FastSodiumPrototype
EBR-2US19641994191962.5––FastSodiumExperimental/Test
ShippingportUS1977
as breeder		19826060236––ThermalLight WaterExperimental-Core3

The Soviet Union (comprising Russia and other countries, dissolved in 1991) constructed a series of fast reactors, the first being mercury-cooled and fueled with plutonium metal, and the later plants sodium-cooled and fueled with plutonium oxide.

BR-1 (1955) was 100W (thermal) was followed by BR-2 at 100 kW and then the 5MW BR-5.[72]

BOR-60 (first criticality 1969) was 60 MW, with construction started in 1965.[73]

BN-600 (1981), followed by Russia's BN-800 (2016)

Future plants[edit]

The Chinese Experimental Fast Reactor is a 65 MW (thermal), 20 MW (electric), sodium-cooled, pool-type reactor with a 30-year design lifetime and a target burnup of 100 MWd/kg.

India has been an early leader in the FBR segment. In 2012 an FBR called the Prototype Fast Breeder Reactor was due to be completed and commissioned.[74][75][needs update]The program is intended to use fertile thorium-232 to breed fissile uranium-233. India is also pursuing thorium thermal breeder reactor technology. India's focus on thorium is due to the nation's large reserves, though known worldwide reserves of thorium are four times those of uranium. India's Department of Atomic Energy (DAE) said in 2007 that it would simultaneously construct four more breeder reactors of 500 MWe each including two at Kalpakkam.[76][needs update]

BHAVINI, an Indian nuclear power company, was established in 2003 to construct, commission and operate all stage II fast breeder reactors outlined in India's three stage nuclear power programme. To advance these plans, the Indian FBR-600 is a pool-type sodium-cooled reactor with a rating of 600 MWe.[citation needed][needs update]

The China Experimental Fast Reactor (CEFR) is a 25 MW(e) prototype for the planned China Prototype Fast Reactor (CFRP).[77] It started generating power on 21 July 2011.[78]

China also initiated a research and development project in thorium molten-salt thermal breeder-reactor technology (liquid fluoride thorium reactor), formally announced at the Chinese Academy of Sciences (CAS) annual conference in January 2011. Its ultimate target was to investigate and develop a thorium-based molten salt nuclear system over about 20 years.[79][80][needs update]

Kirk Sorensen, former NASA scientist and chief nuclear technologist at Teledyne Brown Engineering, has long been a promoter of thorium fuel cycle and particularly liquid fluoride thorium reactors. In 2011, Sorensen founded Flibe Energy, a company aimed to develop 20–50 MW LFTR reactor designs to power military bases.[81][82][83][84]

South Korea is developing a design for a standardized modular FBR for export, to complement the standardized PWR (pressurized water reactor) and CANDU designs they have already developed and built, but has not yet committed to building a prototype.

A cutaway model of the BN-600 reactor, superseded by the BN-800 reactor family.
Construction of the BN-800 reactor

Russia has a plan for increasing its fleet of fast breeder reactors significantly. A BN-800 reactor (800 MWe) at Beloyarsk was completed in 2012, succeeding a smaller BN-600. In June 2014 the BN-800 was started in the minimum power mode.[85] Working at 35% of nominal efficiency, the reactor contributed to the energy network on 10 December 2015.[86] It reached its full power production in August 2016.[87]

Plans for the construction of a larger BN-1200 reactor (1,200 MWe) was scheduled for completion in 2018, with two additional BN-1200 reactors built by the end of 2030.[88] However, in 2015 Rosenergoatom postponed construction indefinitely to allow fuel design to be improved after more experience of operating the BN-800 reactor, and among cost concerns.[89]

An experimental lead-cooled fast reactor, BREST-300 will be built at the Siberian Chemical Combine (SCC) in Seversk. The BREST (Russian: bystry reaktor so svintsovym teplonositelem, English: fast reactor with lead coolant) design is seen as a successor to the BN series and the 300 MWe unit at the SCC could be the forerunner to a 1,200 MWe version for wide deployment as a commercial power generation unit. The development program is as part of an Advanced Nuclear Technologies Federal Program 2010–2020 that seeks to exploit fast reactors for uranium efficiency while 'burning' radioactive substances that would otherwise be disposed of as waste. Its core would measure about 2.3 metres in diameter by 1.1 metres in height and contain 16 tonnes of fuel. The unit would be refuelled every year, with each fuel element spending five years in total within the core. Lead coolant temperature would be around 540 °C, giving a high efficiency of 43%, primary heat production of 700 MWt yielding electrical power of 300 MWe. The operational lifespan of the unit could be 60 years. The design is expected to be completed by NIKIET in 2014 for construction between 2016 and 2020.[90]

On 16 February 2006, the United States, France and Japan signed an "arrangement" to research and develop sodium-cooled fast reactors in support of the Global Nuclear Energy Partnership.[91]In April 2007 the Japanese government selected Mitsubishi Heavy Industries (MHI) as the "core company in FBR development in Japan". Shortly thereafter, MHI started a new company, Mitsubishi FBR Systems (MFBR) to develop and eventually sell FBR technology.[92]

The Marcoule Nuclear Site in France, location of the Phénix (on the left).

In September 2010 the French government allocated €651.6 million to the Commissariat à l'énergie atomique to finalize the design of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a 600 MW fourth-generation reactor design to be finalized in 2020.[93][94] As of 2013 the UK had shown interest in the PRISM reactor and was working in concert with France to develop ASTRID. In 2019, CEA announced this design would not be built before mid-century.[95]

In October 2010 GE Hitachi Nuclear Energy signed a memorandum of understanding with the operators of the US Department of Energy's Savannah River Site, which should allow the construction of a demonstration plant based on the company's S-PRISM fast breeder reactor prior to the design receiving full Nuclear Regulatory Commission (NRC) licensing approval.[96] In October 2011 The Independent reported that the UK Nuclear Decommissioning Authority (NDA) and senior advisers within the Department for Energy and Climate Change (DECC) had asked for technical and financial details of PRISM, partly as a means of reducing the country's plutonium stockpile.[97]

The traveling wave reactor (TWR) proposed in a patent by Intellectual Ventures is a fast breeder reactor designed to not need fuel reprocessing during the decades-long lifetime of the reactor. The breed-burn wave in the TWR design does not move from one end of the reactor to the other but gradually from the inside out. Moreover, as the fuel's composition changes through nuclear transmutation, fuel rods are continually reshuffled within the core to optimize the neutron flux and fuel usage at any given point in time. Thus, instead of letting the wave propagate through the fuel, the fuel itself is moved through a largely stationary burn wave. This is contrary to many media reports, which have popularized the concept as a candle-like reactor with a burn region that moves down a stick of fuel. By replacing a static core configuration with an actively managed "standing wave" or "soliton" core, TerraPower's design avoids the problem of cooling a highly variable burn region. Under this scenario, the reconfiguration of fuel rods is accomplished remotely by robotic devices; the containment vessel remains closed during the procedure, and there is no associated downtime.[98]

See also[edit]

  • India's three stage nuclear power programme
  • Fast neutron reactor
  • Sodium-cooled fast reactor
  • Integral Fast Reactor
  • Lead-cooled fast reactor
  • Gas-cooled fast reactor
  • Generation IV reactor
  • Reduced moderation water reactor
  • Supercritical water reactor
  • Nuclear fusion-fission hybrid
  • David Hahn

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External links[edit]

  • "Fast Neutron Reactor Plants From Experience to Prospects" (PDF). – on OKBM Afrikantov official pdf(in English)
  • Breeder terminology
  • US Nuclear Program
  • IAEA Fast Reactors Database
  • IAEA Technical Documents on Fast Reactors
  • Reactors Designed by Argonne National Laboratory: Fast Reactor Technology Argonne pioneered the development of fast reactors and is a leader in the development of fast reactors worldwide. See also Argonne’s Nuclear Science and Technology Legacy.
  • Atomic Heritage Foundation – EBR-I
  • The Changing Need for a Breeder Reactor by Richard Wilson at The Uranium Institute 24th Annual Symposium, September 1999
  • Experimental Breeder Reactor-II (EBR-II): An Integrated Experimental Fast Reactor Nuclear Power Station
  • International Thorium Energy Organisation – www.IThEO.org