High-temperature gas-cooled reactor

{{Short description|Type of nuclear reactor that operates at high temperatures as part of normal operation}} {{Use American English|date = February 2019}} {{Use dmy dates|date=June 2024}} [[File:"REFUELING FLOOR" AT ST. VRAIN NUCLEAR POWER PLANT - NARA - 544826.jpg|thumb|250 px|Refueling floor at [[Fort Saint Vrain Nuclear Power Plant|Fort Saint Vrain HTGR]], 1972]]

A '''high-temperature gas-cooled reactor (HTGR)''' is a type of [[gas-cooled reactor|gas-cooled nuclear reactor]] which uses uranium fuel and [[graphite-moderated reactor|graphite moderation]] to produce very high [[nuclear reactor core|reactor core]] output temperatures. (700 to 950 °C)<ref> {{cite web |url=https://inldigitallibrary.inl.gov/sites/sti/sti/Sort_26285.pdf |title=A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor |publisher=Idaho National Laboratory |date=26 August 2020 |author=Evans D. Kitcher |quote="The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures" }}</ref> All existing HTGR reactors use [[helium]] coolant. The reactor core can be either a "prismatic block" (reminiscent of a conventional reactor core) or a "[[Pebble-bed reactor|pebble-bed]]" core. [[China Huaneng Group]] currently operates [[HTR-PM]], a 250 MW HTGR power plant with two pebble-bed HTGRs, in [[Shandong province]], [[China]].

The high operating temperatures of HTGR reactors potentially enable applications such as process heat or [[hydrogen]] production via the thermochemical [[sulfur–iodine cycle]]. A proposed development of the HTGR is the [[Generation IV reactor|Generation IV]] '''very-high-temperature reactor''' (VHTR).

== History == The use of a high-temperature, gas-cooled reactor for power production was proposed in 1944 by [[Farrington Daniels]], then associate director of the chemistry division at the University of Chicago's [[Metallurgical Laboratory]]. Initially, Daniels envisaged a reactor using [[beryllium]] moderator. Development of this high temperature design proposal continued at the Power Pile Division of the Clinton Laboratories (known now as [[Oak Ridge National Laboratory]]) until 1947.<ref name="CLL-HTGCPP">{{cite tech report |title=Summary Report on Design and Development of High Temperature Gas-Cooled Power Pile |last=McCullough |first=C. Rodgers |author2=Staff, Power Pile Division |date=15 September 1947 |publisher=Clinton Laboratories (now [[Oak Ridge National Laboratory]]) |location=[[Oak Ridge, Tennessee|Oak Ridge]], [[Tennessee|TN]], USA |osti=4359623 |doi=10.2172/4359623 |url=https://digital.library.unt.edu/ark:/67531/metadc1026527/ |access-date=25 May 2025}}</ref> Professor [[Rudolf Schulten]] in [[Germany]] also played a role in development during the 1950s. [[Peter Fortescue]], whilst at [[General Atomics]], was leader of the team responsible for the initial development of the High temperature gas-cooled reactor (HTGR), as well as the [[Gas-cooled fast reactor]] (GCFR) system.<ref>{{Cite web|url=http://www.ga.com/peter-fortescue-dies-at-102|title = Peter Fortescue Dies at 102}}</ref>

The United States' [[Peach Bottom Unit 1]] was the first HTGR to produce electricity, and did so very successfully, with operation from 1966 through 1974 as a technology demonstrator. [[Fort St. Vrain Generating Station]] was one example of this design that operated as an HTGR from 1979 to 1989. Though the reactor was beset by some problems which led to its decommissioning due to economic factors, it served as proof of the HTGR concept in the United States (though no new commercial HTGRs have been developed there since).<ref>[[IAEA]] [http://www.iaea.org/inisnkm/nkm/aws/htgr/ HTGR Knowledge Base]</ref>{{Failed verification|date=November 2009}}

Experimental HTGRs have also existed in the United Kingdom (the [[Dragon reactor]]) and Germany ([[AVR reactor]] and [[THTR-300]]), and currently exist in Japan (the [[High-temperature engineering test reactor]] using prismatic fuel with 30 [[MWTh|MWth]] of capacity) and China (the [[HTR-10]], a pebble-bed design with 10 MWe of generation). Two full-scale pebble-bed HTGRs, the [[HTR-PM]] reactors, each with 100 MW of electrical production capacity, have gone operational in China as of 2021.<ref>{{Cite web|url=https://world-nuclear-news.org/Articles/Demonstration-HTR-PM-grid-connected|title=Demonstration HTR PM prepares for grid connection : New Nuclear – World Nuclear News|website=world-nuclear-news.org |date=16 December 2021 }}</ref>

== Reactor design == [[File:HD.6D.755 (13471524583).jpg|thumb|A simplified flow diagram of a 1,100 MWe HTGR.]]

=== Neutron moderator === The neutron moderator is graphite, although whether the reactor core is configured in graphite prismatic blocks or in graphite pebbles depends on the HTGR design.

=== Nuclear fuel === The fuel used in HTGRs is coated fuel particles, such as [[TRISO]]<ref>{{cite conference |last1=Alrwashdeh |first1=Mohammad |last2=Alameri |first2=Saeed A. |title=Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors |book-title=Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering |conference=International Conference on Nuclear Engineering |publisher=ASME |location=Virtual |id= ICONE2020-16838 |isbn=978-0-7918-8376-1 |date=4 August 2020 |doi=10.1115/ICONE2020-16838}}</ref> fuel particles. Coated fuel particles have fuel kernels, usually made of [[uranium dioxide]], however, [[uranium carbide]] or uranium oxycarbide are also possibilities. Uranium oxycarbide combines uranium carbide with the uranium dioxide to reduce the oxygen stoichiometry. Less oxygen may lower the internal pressure in the TRISO particles caused by the formation of carbon monoxide, due to the oxidization of the porous carbon layer in the particle.<ref>{{cite journal | doi = 10.1016/j.jnucmat.2009.01.297 | volume=389 | issue=1 | title=Nuclear fuels – Present and future | year=2009 | journal=Journal of Nuclear Materials | pages=1–22 | last1 = Olander | first1 = D.| bibcode=2009JNuM..389....1O | url=http://engj.org/index.php/ej/article/view/30 | url-access=subscription }}</ref> The TRISO particles are either dispersed in a pebble for the pebble bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks. The QUADRISO fuel<ref>{{Cite journal |doi = 10.1016/j.nucengdes.2010.03.025|title = A novel concept of QUADRISO particles. Part II: Utilization for excess reactivity control|journal = Nuclear Engineering and Design|volume = 240|issue = 7|pages = 1919–1927|year = 2010|last1 = Talamo|first1 = Alberto|url = https://zenodo.org/record/1259279}}</ref> concept conceived at [[Argonne National Laboratory]] has been used to better manage the excess of reactivity.

=== Coolant ===

Helium has been the coolant used in all HTGRs to date. Helium is an [[inert gas]], so it will generally not chemically react with any material.<ref name="IAEA1996HTGRp61">{{cite web|url=http://www.iaea.org/inisnkm/nkm/aws/htgr/fulltext/29026666.pdf |title=High temperature gas cool reactor technology development |access-date=2009-05-08 |date=15 November 1996 |publisher=IAEA |pages=61 }}</ref> Additionally, exposing helium to neutron radiation does not make it radioactive,<ref name="InistHe">{{cite web |url=http://cat.inist.fr/?aModele=afficheN&cpsidt=849696 |title=Thermal performance and flow instabilities in a multi-channel, helium-cooled, porous metal divertor module |access-date=2009-05-08 |year=2000 |publisher=Inist |archive-date=30 January 2012 |archive-url=https://web.archive.org/web/20120130043438/http://cat.inist.fr/?aModele=afficheN&cpsidt=849696 |url-status=dead }}</ref> unlike most other possible coolants.

=== Control === In the prismatic designs, [[control rod]]s are inserted in holes cut in the graphite blocks that make up the core. The VHTR will be controlled like current [[PBMR]] designs if it utilizes a pebble bed core, the control rods will be inserted in the surrounding graphite [[Neutron reflector|reflector]]. Control can also be attained by adding pebbles containing [[neutron absorber]]s.

== Safety features and other benefits == The design takes advantage of the inherent safety characteristics of a helium-cooled, graphite-moderated core with specific design optimizations. The graphite has large [[volumetric heat capacity|thermal inertia]] and the helium coolant is single phase, inert, and has no reactivity effects. The core is composed of graphite, has a high heat capacity and structural stability even at high temperatures. The fuel is coated uranium-oxycarbide which permits high burn-up (approaching 200 GWd/t) and retains fission products. The high average core-exit temperature of the VHTR (1,000 °C) permits emissions-free production of high grade [[process heat]]. Reactors are designed for 60 years of service.<ref>http://www.uxc.com/smr/Library/Design%20Specific/HTR-PM/Papers/2006%20-%20Design%20aspects%20of%20the%20Chinese%20modular%20HTR-PM.pdf Page 489, Table 2. Quote: Designed operational life time (year) 60</ref>

== List of HTGR reactors == ===Constructed reactors=== As of 2011, a total of seven HTGR reactors have been constructed and operated.<ref name=inl>{{cite web |title=High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant |url=https://inldigitallibrary.inl.gov/sites/sti/sti/5026001.pdf |quote=To date, seven HTGR plants have been built and operated |date=April 2011 |publisher=Idaho National Laboratory |author=J. M. Beck, L. F. Pincock}}</ref> A further two HTGR reactors were brought on-line at China's [[HTR-PM]] site, in 2021/22.

{| class="wikitable sortable mw-datatable" |- !Facility name !Country !Commissioned !Shutdown !No. of reactors !Fuel type !Outlet temperature (°C) !Thermal power (MW) |- | [[Dragon reactor]]<ref name=inl/> || [[United Kingdom]] || 1965 || 1976 || 1 || Prismatic || 750 || 21.5 |- | [[Peach Bottom Unit 1|Peach Bottom]]<ref name=inl/> || [[United States]] || 1967 || 1974 || 1 || Prismatic || 700–726 || 115 |- | [[AVR reactor|AVR]]<ref name=inl/> || [[Germany]] || 1967 || 1988 || 1 || Pebble bed || 950 || 46 |- | [[Fort Saint Vrain Nuclear Power Plant|Fort Saint Vrain]]<ref name=inl/> || [[United States]] || 1979 || 1989 || 1 || Prismatic || 777 || 842 |- | [[THTR-300]]<ref name=inl/> || [[Germany]] || 1985 || 1988 || 1 || Pebble bed || 750 || 750 |- | [[High-temperature engineering test reactor|HTTR]]<ref name=inl/> || [[Japan]] || 1999|| Operational || 1 || Prismatic || 850–950 || 30 |- | [[HTR-10]]<ref name=inl/> || [[China]] || 2000 || Operational || 1 || Pebble bed || 700 || 10 |- | [[HTR-PM]]<ref>{{Cite web |title=Advanced Reactor Information System {{!}} Aris |url=https://aris.iaea.org/PDF/HTR-PM.pdf |url-status=dead |archive-url=https://web.archive.org/web/20240225181123/https://aris.iaea.org/PDF/HTR-PM.pdf |archive-date=25 February 2024 |website=International Atomic Energy Agency}}</ref>|| [[China]] || 2021 || Operational || 2 || Pebble bed || 750 || 250 |- | [[HTGR]] ([[Lianyungang]])<ref>{{Cite web |title=China starts construction of innovative nuclear project |url=https://www.world-nuclear-news.org/articles/china-starts-construction-of-innovative-nuclear-project | website=world-nuclear-news.org |date=16 January 2026}}</ref>|| [[China]] || Approved for construction || || 1 || ? || || 660 |- |}

Additionally, from 1969 to 1971, the 3 MW [[UHTREX|Ultra-High Temperature Reactor Experiment]] (UHTREX) was operated by [[Los Alamos National Laboratory]] to develop the technology of high-temperature gas-cooled reactors.<ref> {{Citation | last = Lipper | first = H. W. | year = 1969 | title = Helium symposia proceedings in 1968: a hundred years of helium | publisher = United States | page = 117 | chapter = High-Temperature Gas-Cooled Reactors Using Helium Coolant | quote = Three of these plants, AVR, Peach Bottom, and Fort St. Vrain, are actual electrical generating plants, and two, Dragon and UHTREX, are experimental plants being used primarily to develop the technology of high – temperature, gas-cooled reactors. }} </ref> In UHTREX, unlike HTGR reactors, helium coolant contacted nuclear fuel directly, reaching temperatures in excess of 1300 °C.

===Proposed designs=== * [[Pebble Bed Modular Reactor]] (1994) – reactor proposed for [[Koeberg Nuclear Power Station]], South Africa * [[Gas turbine modular helium reactor]] (1997) – proposed reactor with gas turbine power conversion * [[Next Generation Nuclear Plant]] (2005) – a proposed Generation IV very-high-temperature reactor * [[X-energy]] (2016) – developers of a proposed Generation IV pebble-bed reactor * [[U-Battery]] (2020) – a micro–small modular reactor design effort, discontinued in 2023

==References== {{Reflist}}

==External links== *[http://www.inl.gov/research/very-high-temperature-reactor/ Idaho National Lab VHTR Fact Sheet] *{{Cite web |url=http://gif.inel.gov/roadmap/pdfs/p_grns_june_25-27_presentation_gp32-00.pdf |title=VHTR presentation |access-date=24 November 2005 |archive-url=https://wayback.archive-it.org/all/20090225155637/http://gif.inel.gov/roadmap/pdfs/p_grns_june_25-27_presentation_gp32-00.pdf |archive-date=25 February 2009 |url-status=dead }} (from the year 2002) *[https://www.gen-4.org/gif/jcms/c_9362/vhtr Generation IV International Forum VHTR website] *{{Cite web |url=http://neri.inel.gov/program_plans/pdfs/appendix_1.pdf |title=INL VHTR workshop summary |access-date=21 December 2005 |archive-url=https://wayback.archive-it.org/all/20071129121507/http://neri.inel.gov/program_plans/pdfs/appendix_1.pdf |archive-date=29 November 2007 |url-status=dead }} *{{cite web|url=http://www.raphael-project.org/index.html |title=The European VHTR research & development programme: RAPHAEL |access-date=1 July 2015 |url-status=dead |archive-url=https://web.archive.org/web/20120722104203/http://www.raphael-project.org/index.html |archive-date=22 July 2012 }} *[http://www.nuc.berkeley.edu/pb-ahtr/ Pebble Bed Advanced High Temperature Reactor (PB-AHTR)] {{Webarchive|url=https://web.archive.org/web/20101006155000/http://www.nuc.berkeley.edu/pb-ahtr/ |date=6 October 2010 }} * [http://www.iaea.org/inisnkm/nkm/aws/htgr/ IAEA HTGR Knowledge Base] * [https://web.archive.org/web/20051208220206/http://www.ornl.gov/info/ornlreview/v37_1_04/article_02.shtml ORNL NGNP page] * [https://web.archive.org/web/20060511164737/http://www3.inspi.ufl.edu/icapp06/program/abstracts/6208.pdf INL Thermal-Hydraulic Analyses of the LS-VHTR] * [[IFNEC]] slides from 2014 about Areva's [[SC-HTGR]]: [http://www.ifnec.org/Portals/0/Docs/IDWG%20Meeting%205-8-14/SC%20HTGR%20(Farshid%20Shahrokhi).pdf] {{Webarchive|url=https://web.archive.org/web/20160304042654/http://www.ifnec.org/Portals/0/Docs/IDWG%20Meeting%205-8-14/SC%20HTGR%20(Farshid%20Shahrokhi).pdf |date=4 March 2016 }} * The [[Office of Nuclear Energy]] reports to the IAEA in April 2014: [https://www.iaea.org/NuclearPower/Downloadable/Meetings/2014/2014-04-08-04-11-TM-NPTDS/7_OConnor01.pdf]

[[Category:Nuclear power reactor types]] [[Category:Graphite moderated reactors]]