High-temperature gas-cooled reactor

Last updated

Refueling floor at Fort Saint Vrain HTGR, 1972 "REFUELING FLOOR" AT ST. VRAIN NUCLEAR POWER PLANT - NARA - 544826.jpg
Refueling floor at Fort Saint Vrain HTGR, 1972

A high-temperature gas-cooled reactor (HTGR) is a type of gas-cooled nuclear reactor which use uranium fuel and graphite moderation to produce very high reactor core output temperatures. [1] 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" core. China Huaneng Group currently operates HTR-PM, a 250 MW HTGR power plant in Shandong province, China.

Contents

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 HGTR is the Generation IV very-high-temperature reactor (VHTR) which would initially work with temperatures of 750 to 950 °C.

History

The use of a high-temperature, gas-cooled reactor for power production was proposed by 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. [2] 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. [3]

The Peach Bottom unit 1 reactor in the United States 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). [4] [ failed verification ]

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 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. [5]

Reactor design

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 [6] [7] [8] [9] 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. [10] 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 [11] 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. [12] Additionally, exposing helium to neutron radiation does not make it radioactive, [13] unlike most other possible coolants.

Control

In the prismatic designs, control rods 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 reflector. Control can also be attained by adding pebbles containing neutron absorbers.

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 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. [14]

List of HTGR reactors

Constructed reactors

As of 2011, a total of seven HTGR reactors had been constructed and operated. [15] A further two HTGR reactors were brought on-line at China's HTR-PM site, in 2021/22.

Facility
name		CountryCommissionedShutdownNo. of
reactors		Fuel typeOutlet
temperature (°C)		Thermal
power (MW)
Dragon reactor [15] United Kingdom 196519671Prismatic75021.5
Peach Bottom [15] United States 196719741Prismatic700 - 726115
AVR [15] Germany 196719881Pebble bed95046
Fort Saint Vrain [15] United States 197919891Prismatic777842
THTR-300 [15] Germany 198519881Pebble bed750750
HTTR [15] Japan 1999Operational1Prismatic850 - 95030
HTR-10 [15] China 2000Operational1Pebble bed70010
HTR-PM [16] China 2021Operational2Pebble bed750250

Additionally, from 1969 to 1971, the 3 MW Ultra-High Temperature Reactor Experiment (UHTREX) was operated by Los Alamos National Laboratory to develop the technology of high-temperature gas-cooled reactors. [17] In UHTREX, unlike HTGR reactors, helium coolant contacted nuclear fuel directly, reaching temperatures in excess of 1300°C.

Proposed designs

Related Research Articles

<span class="mw-page-title-main">Nuclear reactor</span> Device used to initiate and control a nuclear chain reaction

A nuclear reactor is a device used to initiate and control a fission nuclear chain reaction or nuclear fusion reactions. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid, which in turn runs through steam turbines. These either drive a ship's propellers or turn electrical generators' shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating. Some reactors are used to produce isotopes for medical and industrial use, or for production of weapons-grade plutonium. As of 2022, the International Atomic Energy Agency reports there are 422 nuclear power reactors and 223 nuclear research reactors in operation around the world.

<span class="mw-page-title-main">Pebble-bed reactor</span> Type of very-high-temperature reactor

The pebble-bed reactor (PBR) is a design for a graphite-moderated, gas-cooled nuclear reactor. It is a type of very-high-temperature reactor (VHTR), one of the six classes of nuclear reactors in the Generation IV initiative.

<span class="mw-page-title-main">Pebble bed modular reactor</span> South African nuclear reactor design

The Pebble Bed Modular Reactor (PBMR) is a particular design of pebble bed reactor developed by South African company PBMR (Pty) Ltd from 1994 until 2009. PBMR facilities include gas turbine and heat transfer labs at the Potchefstroom Campus of North-West University, and at Pelindaba, a high pressure and temperature helium test rig, as well as a prototype fuel fabrication plant. A planned test reactor at Koeberg Nuclear Power Station was not built.

<span class="mw-page-title-main">Rudolf Schulten</span> German physicist (1923–1996)

Rudolf Schulten —professor at RWTH Aachen University—was the main developer of the pebble bed reactor design, which was originally invented by Farrington Daniels. Schulten's concept compacts silicon carbide-coated uranium granules into hard, billiard-ball-like graphite spheres to be used as fuel for a new high temperature, helium-cooled type of nuclear reactor.

Passive nuclear safety is a design approach for safety features, implemented in a nuclear reactor, that does not require any active intervention on the part of the operator or electrical/electronic feedback in order to bring the reactor to a safe shutdown state, in the event of a particular type of emergency. Such design features tend to rely on the engineering of components such that their predicted behaviour would slow down, rather than accelerate the deterioration of the reactor state; they typically take advantage of natural forces or phenomena such as gravity, buoyancy, pressure differences, conduction or natural heat convection to accomplish safety functions without requiring an active power source. Many older common reactor designs use passive safety systems to a limited extent, rather, relying on active safety systems such as diesel-powered motors. Some newer reactor designs feature more passive systems; the motivation being that they are highly reliable and reduce the cost associated with the installation and maintenance of systems that would otherwise require multiple trains of equipment and redundant safety class power supplies in order to achieve the same level of reliability. However, weak driving forces that power many passive safety features can pose significant challenges to effectiveness of a passive system, particularly in the short term following an accident.

<span class="mw-page-title-main">Fort Saint Vrain Nuclear Power Plant</span> Decommissioned nuclear power plant

The Fort St. Vrain Nuclear Power Plant is a former commercial nuclear power station located near the town of Platteville in northern Colorado in the United States. It originally operated from 1979 until 1989. It had a 330 MWe High-temperature gas reactor (HTGR). The plant was decommissioned between 1989 and 1992.

<span class="mw-page-title-main">Nuclear fuel</span> Material fuelling nuclear reactors

Nuclear fuel is material used in nuclear power stations to produce heat to power turbines. Heat is created when nuclear fuel undergoes nuclear fission.

Generation IVreactors are nuclear reactor design technologies that are envisioned as successors of generation III reactors. The Generation IV International Forum (GIF) – an international organization that coordinates the development of generation IV reactors – specifically selected six reactor technologies as candidates for generation IV reactors. The designs target improved safety, sustainability, efficiency, and cost. The World Nuclear Association in 2015 suggested that some might enter commercial operation before 2030.

<span class="mw-page-title-main">Gas-cooled fast reactor</span> Type of nuclear reactor cooled by a gas

The gas-cooled fast reactor (GFR) system is a nuclear reactor design which is currently in development. Classed as a Generation IV reactor, it features a fast-neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reference reactor design is a helium-cooled system operating with an outlet temperature of 850 °C (1,560 °F) using a direct Brayton closed-cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations are being considered based on pin- or plate-based fuel assemblies or prismatic blocks, which allows for better coolant circulation than traditional fuel assemblies.

<span class="mw-page-title-main">THTR-300</span> Thorium nuclear reactor in Germany

The THTR-300 was a thorium cycle high-temperature nuclear reactor rated at 300 MW electric (THTR-300) in Hamm-Uentrop, Germany. It started operating in 1983, synchronized with the grid in 1985, operated at full power in February 1987 and was shut down September 1, 1989. The THTR-300 served as a prototype high-temperature reactor (HTR) to use the TRISO pebble fuel produced by the AVR, an experimental pebble bed operated by VEW. The THTR-300 cost €2.05 billion and was predicted to cost an additional €425 million through December 2009 in decommissioning and other associated costs. The German state of North Rhine Westphalia, Federal Republic of Germany, and Hochtemperatur-Kernkraftwerk GmbH (HKG) financed the THTR-300’s construction.

The Gas Turbine Modular Helium Reactor (GT-MHR) is a class of nuclear fission power reactor designed that was under development by a group of Russian enterprises, an American group headed by General Atomics, French Framatome and Japanese Fuji Electric. It is a helium cooled, graphite moderated reactor and uses TRISO fuel compacts in a prismatic core design. The power is generated via a gas turbine rather than via the more common steam turbine.

<span class="mw-page-title-main">Graphite-moderated reactor</span> Type of nuclear reactor

A graphite-moderated reactor is a nuclear reactor that uses carbon as a neutron moderator, which allows natural uranium to be used as nuclear fuel.

<span class="mw-page-title-main">Dragon reactor</span> UK experimental HTR, operated from 1965 to 1976

Dragon was an experimental high temperature gas-cooled reactor at Winfrith in Dorset, England, operated by the United Kingdom Atomic Energy Authority (UKAEA). Its purpose was to test fuel and materials for the European High Temperature Reactor programme, which was exploring the use of tristructural-isotropic (TRISO) fuel and gas cooling for future high-efficiency reactor designs. The project was built and managed as an Organisation for Economic Co-operation and Development/Nuclear Energy Agency international project. In total, 13 countries were involved in its design and operation during the project lifetime.

<span class="mw-page-title-main">HTR-10</span> Prototype pebble bed reactor, China

HTR-10 is a 10 MWt prototype high-temperature gas-cooled, pebble-bed reactor at Tsinghua University in China. Construction began in 1995, achieving its first criticality in December 2000, and was operated in full power condition in January 2003.

<span class="mw-page-title-main">AVR reactor</span> Prototype nuclear reactor in Germany

The AVR reactor was a prototype pebble-bed reactor, located immediately adjacent to Jülich Research Centre in West Germany, constructed in 1960, grid connected in 1967 and shut down in 1988. It was a 15 MWe, 46 MWt test reactor used to develop and test a variety of fuels and machinery.

A gas-cooled reactor (GCR) is a nuclear reactor that uses graphite as a neutron moderator and a gas as coolant. Although there are many other types of reactor cooled by gas, the terms GCR and to a lesser extent gas cooled reactor are particularly used to refer to this type of reactor.

The Ultra-High Temperature Reactor Experiment (UHTREX) was an experimental gas-cooled nuclear reactor run at Los Alamos National Laboratory between 1959 and 1971 as part of research into reducing the cost of nuclear power. Its purpose was to test and compare the advantages of using a simple fuel against the disadvantages of a contaminated cooling loop. It first achieved full power in 1969.

The HTR-PM is a Chinese small modular nuclear reactor. It is a high-temperature gas-cooled (HTGR) pebble-bed generation IV reactor based on the German AVR design. The technology is intended to replace coal-fired power plants in China's interior, in line with the country's plan to reach carbon neutrality by 2060.

X-energy is a private American nuclear reactor and fuel design engineering company. It is developing a Generation IV high-temperature gas-cooled pebble-bed nuclear reactor design. It has received funding from private sources and various government grants and contracts, notably through the Department of Energy's (DOE) Advanced Reactor Concept Cooperative Agreement in 2016 and its Advanced Reactor Demonstration Program (ARDP) in 2020.

References

  1. ↑ Evans D. Kitcher (26 August 2020). "A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor" (PDF). Idaho National Laboratory. 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
  2. ↑ McCullough, C. Rodgers; Staff, Power Pile Division (15 September 1947). "Summary Report on Design and Development of High Temperature Gas-Cooled Power Pile". Oak Ridge, TN, USA: Clinton Laboratories (now Oak Ridge National Laboratory). doi:10.2172/4359623. OSTI   4359623.{{cite journal}}: Cite journal requires |journal= (help)
  3. ↑ "Peter Fortescue Dies at 102".
  4. ↑ IAEA HTGR Knowledge Base
  5. ↑ "Demonstration HTR PM prepares for grid connection : New Nuclear - World Nuclear News". world-nuclear-news.org.
  6. ↑ Alameri, Saeed A., and Mohammad Alrwashdeh. "Preliminary three-dimensional neutronic analysis of IFBA coated TRISO fuel particles in prismatic-core advanced high temperature reactor." Annals of Nuclear Energy 163 (2021): 108551.
  7. ↑ Alrwashdeh, Mohammad, and Saeed A. Alameri. "Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors." In International Conference on Nuclear Engineering, vol. 83761, p. V001T05A014. American Society of Mechanical Engineers, 2020
  8. ↑ Alrwashdeh, Mohammad, Saeed A. Alameri, and Ahmed K. Alkaabi. "Preliminary Study of a Prismatic-Core Advanced High-Temperature Reactor Fuel Using Homogenization Double-Heterogeneous Method." Nuclear Science and Engineering 194, no. 2 (2020): 163-167.
  9. ↑ Alrwashdeh, Mohammad, Saeed A. Alamaeri, Ahmed K. Alkaabi, and Mohamed Ali. "Homogenization of TRISO Fuel using Reactivity Equivalent Physical Transformation Method." Transactions 121, no. 1 (2019): 1521-1522.
  10. ↑ Olander, D. (2009). "Nuclear fuels – Present and future". Journal of Nuclear Materials. 389 (1): 1–22. Bibcode:2009JNuM..389....1O. doi:10.1016/j.jnucmat.2009.01.297.
  11. ↑ Talamo, Alberto (2010). "A novel concept of QUADRISO particles. Part II: Utilization for excess reactivity control". Nuclear Engineering and Design. 240 (7): 1919–1927. doi:10.1016/j.nucengdes.2010.03.025.
  12. ↑ "High temperature gas cool reactor technology development" (PDF). IAEA. 15 November 1996. p. 61. Retrieved 8 May 2009.
  13. ↑ "Thermal performance and flow instabilities in a multi-channel, helium-cooled, porous metal divertor module". Inist. 2000. Archived from the original on 30 January 2012. Retrieved 8 May 2009.
  14. ↑ 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
  15. 1 2 3 4 5 6 7 8 J. M. Beck, L. F. Pincock (April 2011). "High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant" (PDF). Idaho National Laboratory. To date, seven HTGR plants have been built and operated
  16. ↑ https://aris.iaea.org/PDF/HTR-PM.pdf
  17. ↑ Lipper, H. W. (1969), "High-Temperature Gas-Cooled Reactors Using Helium Coolant", Helium symposia proceedings in 1968: a hundred years of helium, United States, p. 117, 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.
This page is based on this Wikipedia article
Text is available under the CC BY-SA 4.0 license; additional terms may apply.
Images, videos and audio are available under their respective licenses.