The High Temperature Gas-Cooled Reactor
Michael A. Fu ¨tterera, Gerhard Strydomb, Hiroyuki Satoc,F uL id, Eric Abonneaue, Tim Abramf, Mike W. Daviesg, Minwhan Kimh,
Lyndon Edwardsi, Ondrej Muranskyi, Manuel A. Pouchonj, and Metin Yetisirk,aEuropean Commission, Joint Research Centre,
Petten, The Netherlands;bIdaho National Laboratory, Idaho Falls, ID, United States;cJapan Atomic Energy Agency, Ibaraki, Japan;
dINET, Tsinghua University, Beijing, China;eCommissariat à l ’Energie Atomique et aux Energies Alternatives, Paris, France;
fManchester University, Manchester, United Kingdom;gJacobs Clean Energy Limited, Knutsford, Cheshire, United Kingdom;hKorea
Atomic Energy Research Institute, Daejeon, South Korea;iAustralian Nuclear Science and Technology Organisation, Lucas Heights,
Australia;jPaul Scherrer Institut, Villigen, Switzerland; andkCanadian Nuclear Laboratories, Chalk River, ON, Canada
© 2021 Elsevier Inc. All rights reserved.
What is a high temperature gas-cooled reactor? 513
From groundbreaking technology . 513
.to modern characteristics 513
TRISO fuel: Key to performance and safety 514
Which HTR versions were developed? 514
Recent results from test reactors 517
The case for new next generation HTRs 518
Ongoing HTR development 518
Beyond electricity: Emission-free process heat and cogeneration 520
Outlook 520
References 522
Glossary
AGR Advanced Gas-cooled Reactor
AVR Arbeitsgemeinschaft Versuchsreaktor
BISO Bi-Structural Isotropic Fuel
GCR Gas-Cooled Reactor


Outlook 520
References 522
Glossary
AGR Advanced Gas-cooled Reactor
AVR Arbeitsgemeinschaft Versuchsreaktor
BISO Bi-Structural Isotropic Fuel
GCR Gas-Cooled Reactor
GIFGeneration IV International Forum
GT-MHR Gas Turbine Modular Helium-cooled Reactor
HTGR High Temperature Gas-Cooled Reactor
HTR High Temperature Gas-Cooled Reactor
HTR-10 10 MW High Temperature Reactor
HTR-PM High Temperature Reactor –Pebble-bed Module
HTTR High Temperature engineering Test Reactor
HWR Heavy Water Reactor
INET Institute of Nuclear and New Energy Technology (Tsinghua University, Beijing)
JAEA Japan Atomic Energy Agency
LANL Los Alamos National Laboratory
LWR Light Water Reactor
MWe Megawatt electric
MWth Megawatt thermal
NGNP Next Generation Nuclear Plant
ORNL Oak Ridge National Laboratory
PBMR Pebble Bed Modular Reactor
R&D Research and Development
S-ISulfur-Iodine thermochemical process for hydrogen production
SFBR Sodium-cooled Fast Breeder Reactor
THTR Thorium High Temperature Reactor
TRISO Tri-Structural Isotropic Fuel
VHTR Very High Temperature Reactor
512 Encyclopedia of Nuclear Energy, Volume 1 https://doi.org/10.1016/B978-0-12-409548-9.12205-5What is a high temperature gas-cooled reactor?
High Temperature Gas-cooled Reactors (HTR or HTGR) are helium-cooled graphite-moderated nuclear ﬁssion reactors utilizing
fully ceramic fuel. They are characterized by inherent safety features, excellent ﬁssion product retention in the fuel, and high temper-
ature operation suitable for the delivery of industrial process heat, in particular hydrogen production. Typical coolant outlet temper-


fully ceramic fuel. They are characterized by inherent safety features, excellent ﬁssion product retention in the fuel, and high temper-
ature operation suitable for the delivery of industrial process heat, in particular hydrogen production. Typical coolant outlet temper-
atures range between 750/C14C and 850/C14C, thus enabling power conversion ef ﬁciencies up to 48%.
The Very High Temperature Reactor (VHTR) is a longer term evolution of the HTR targeting even higher ef ﬁciency and more
versatile use by further increasing the helium outlet temperature to 950/C14C or even higher ( Gougar, 2011 ).
From groundbreaking technology .
The HTR has evolved from the early gas-cooled reactors (GCRs) that gained widespread popularity for their simplicity and high
power conversion ef ﬁciencies ( Beech and May, 1999 ). The ﬁrst commercial nuclear power plant was a CO 2-cooled graphite-
moderated Magnox reactor (Calder Hall in 1956). In total, 26 Magnox reactors were built (270 –1760 MWth), with the last one
(Wylfa-1, 1971 –2015) shut down at the end of 2015. As a second generation, 14 Advanced Gas-Cooled Reactors (AGRs) were
deployed in seven nuclear power plants at six sites in the UK with a total capacity of approx. 8 GWe. All these AGRs are expected
to remain in operation until 2023 –30, although their life extension required clearance of graphite cracking issues and two power


to remain in operation until 2023 –30, although their life extension required clearance of graphite cracking issues and two power
plants have to run at lower power because of the observation of cracks in boilers. This multi-decade effort in the development andoperation of gas-cooled reactors allowed for collection of a considerable technical background and operational experience, whichthen served as the basis for the development of current HTRs. GCRs have an extremely clean primary cooling circuit (few radiolog-
ical and chemical contaminants) and use a conventional steam cycle ( /C24540
/C14C, same as for coal ﬁred power plants) resulting in
high thermal ef ﬁciencies ( >40%). However, GCRs had to observe a temperature limitation due to the dissociation of CO 2and the
resulting carburization of structural materials and oxidation of graphite at elevated temperatures. Modern HTR are characterized by
increased operating temperature and thermal ef ﬁciency, which could be achieved by two major changes: the designs adopted
helium as the cooling gas along with fully ceramic fuel, which is discussed in more detail in Section “TRISO fuel: Key to perfor-
mance and safety .”
Theﬁrst HTR was proposed in a 1945 design study in the US, but was never realized. It featured a primary circuit (helium at
1.55 MPa, 438 –732/C14C) coupled to a secondary Brayton power conversion cycle (air at 2.9 MPa, 677 –22/C14C) leading to an expected
power rating of 5 MWe.
In 1962 –63, a 3.3 MWth Mobile Low-Power Reactor (ML-1) with 140 (330 nominal) kWe was built in the US with a closed-cycle


In 1962 –63, a 3.3 MWth Mobile Low-Power Reactor (ML-1) with 140 (330 nominal) kWe was built in the US with a closed-cycle
nitrogen turbine. The project was not pursued because it could not ful ﬁll the power output expectations.
In 1964, the Experimental Gas-Cooled Reactor (EGCR) was built at ORNL in the US, but not completed. This was basically
a helium-cooled AGR-type reactor using stainless steel fuel rod clusters. EGCR was expected to produce 85 MWth/25 MWe with
helium at 566/C14C.
Ensuing developments led to conceptual changes in the existing gas-cooled reactors involving, as mentioned, in particular the
use of helium instead of CO 2, and the substitution of metallic fuel clads by fully ceramic fuel, both in view of a further increase of
reactor outlet temperature and improved safety performance.
Theﬁrst tangible step in this direction was made in the UK with the DRAGON reactor (see also Section “Which HTR versions
were developed? ”). With a power of 21.5 MWth, it was an OECD project and operated from 1964 to 1975 primarily as a test bed for
HTR fuel development. It used already early versions of fully ceramic coated particles as its own fuel.
In the US, the Ultra-High-Temperature Reactor Experiment (UHTREX) operated at LANL from 1966 to 1970. Its rated power was
3 MWth using helium at 3.4 MPa (870 –1300/C14C). It used extruded fuel with TRISO coated particles in an annular rotatable core for
on-line refueling.


3 MWth using helium at 3.4 MPa (870 –1300/C14C). It used extruded fuel with TRISO coated particles in an annular rotatable core for
on-line refueling.
More details on the development of HTR technology can be found in a recent authoritative summary ( Kugeler and Zhang, 2019 ).
.to modern characteristics
The following developments led to basic technical characteristics and design features shared by all modern HTR.
can be built with passive safety features up to 625 MWth/core (prismatic block type core) and 250 MWth/core (pebble bed
core); this is the power range of Small Modular Rectors (SMR);
long grace time after an accident (large heat capacity, low power density);
self-stabilization of power transients (negative temperature coef ﬁcient);
low source terms (outstanding ﬁssion product retention in robust TRISO coated fuel particles and structures);
fully ceramic core (fuel and moderator/re ﬂector);
high-purity graphite as moderator/re ﬂector, high thermal inertia;
chemically and neutronically inert helium as primary coolant;
high operating temperatures for high ef ﬁciency, capability for nuclear cogeneration of heat and power, including for bulk
hydrogen production;The High Temperature Gas-Cooled Reactor 513high burn-up capability;
high fuel utilization (good neutron economy and possible use of thorium).
There are two competing HTR designs based on the TRISO coated fuel particle ( Fig. 1 ): the prismatic block core and the pebble bed
core ( Fig. 2 ).
Invented by Peter Fortescue and his team at General Dynamics in the US, the prismatic block core is built from hexagonal


core ( Fig. 2 ).
Invented by Peter Fortescue and his team at General Dynamics in the US, the prismatic block core is built from hexagonal
graphite blocks containing vertical holes. Some of these holes are used for helium cooling, while others receive the fuel in the
form of “compacts ”, which are little cylinders (typically B12.3/C225 mm) pressed from graphite and coated fuel particles
(Fortescue, 1975 ).
The pebble bed HTR was conceived in 1942 by Farrington Daniels in the US ( Daniels, 1944 ). This early vision was later
developed to a power plant design by Rudolf Schulten in Germany, which employed B60 mm fuel spheres made of graphite
and coated fuel particles ( Schulten et al., 1959 ). These pebbles are ﬁlled into the reactor pressure vessel, which is internally lined
with graphite blocks. The resulting pebble bed constitutes the reactor core. The pebble bed can ﬂow and allows discharge and
(re-)injection of pebbles during operation, enabling online refueling.
TRISO fuel: Key to performance and safety
One of the major challenges and key to achieving a fully ceramic reactor core was fuel development ( IAEA, 2010 ). The initially used
UO 2or UC fuel was placed in ceramic clads which showed poor ﬁssion product retention. Coated particle fuel was invented
between 1957 and 1961 by the United Kingdom Atomic Energy Authority (UKAEA) and Battelle, but no patent was granted at
that time. The UO 2fuel kernels were made by external gelation of uranyl nitrate in ammonia and, after a heat treatment, coatings


that time. The UO 2fuel kernels were made by external gelation of uranyl nitrate in ammonia and, after a heat treatment, coatings
were deposited on top of these kernels via pyrolysis of hydrocarbons in a ﬂuidized bed. The next development step was the early
BISO (bi-structural isotropic) particle fuel comprising a buffer layer directly deposited on the kernels and an additional pyrolytic
carbon (PyC) layer on top. Finally, modern TRISO (tri-structural isotropic) particles were given an additional SiC diffusion barrier
leading to con ﬁrmed ﬁssion product retention up to 1600/C14C or even higher ( Gougar et al., 2020 ). These TRISO coated particles,
typically in the order of 1 mm in diameter, are the basis for all modern HTR fuel designs ( Gerczak, 2021 ;Helmreich, 2021 ). As
shown in Fig. 1 , they feature (from inside out) the kernel, a porous PyC buffer to accommodate fuel swelling and ﬁssion gases,
a dense PyC buffer and a dense SiC layer as diffusion barriers against ﬁssion product escape, and a ﬁnal PyC layer (missing in
Fig. 1 ) for better bonding with the matrix graphite into which they will be integrated.
Baked into matrix graphite, the TRISO coated particles can now be given a macroscopic shape ( Fig. 2 ), usually in the form of
thumb-thick cylinders ( “compacts ”) either solid or annular, or in the form of spherical fuel elements ( “pebbles ”). The compacts
are inserted into hexagonal blocks made of graphite, which are then assembled to constitute the reactor core contained in a pressurevessel.


are inserted into hexagonal blocks made of graphite, which are then assembled to constitute the reactor core contained in a pressurevessel.
Typical pebble and compact design characteristics are given in Table 1 :
Which HTR versions were developed?
Based on these characteristics, in the 1960s two different types of reactors were designed and built, primarily to produce electricity.
Experimental HTRs with a prismatic block core and TRISO coated particle fuel were developed in the UK (DRAGON reactor,
Fuel kernel buffer SiC Inner PyC
Fig. 1 SEM picture of a modern TRISO coated particle broken up to visualize the coatings; the top outer PyC layer is still missing on this particle.514 The High Temperature Gas-Cooled Reactoroperated 1964 –1975, 21.5 MWth, an OECD project ( Price, 2012 ) and in the US (Peach Bottom, operated 1966 –1974, 115 MWth/
40 MWe Beck and Pincock, 2011 ). They were followed by the prototype of the Fort St. Vrain Generating Station (operated 1976 –
1989, 842 MWth/330 MWe, Beck and Pincock, 2011 ). This reactor established the technical feasibility of HTRs although it experi-
enced problems ( Rempe, 2021 ) of power ﬂuctuations, jamming of a control rod and leakage of moisture into the core, which ﬁnally
caused its decommissioning for economic reasons.
Over the same period, Germany developed and built an experimental pebble bed reactor (AVR, 46 MWth/15 MWe, Pohl, 2008 )


caused its decommissioning for economic reasons.
Over the same period, Germany developed and built an experimental pebble bed reactor (AVR, 46 MWth/15 MWe, Pohl, 2008 )
at the Jülich Research Centre that successfully operated from 1967 to 1988 and produced valuable feedback on different types ofpebble fuels and overall reactor operation. In particular, it was used for several demonstrations of passive safety performance. After
a water ingress accident provoked by a steam generator leak it could be repaired, dried and returned to service. Following this expe-
rience, a 300 MWe prototype power reactor that aimed at using thorium fuel was built and operated: the Thorium High Temper-ature Reactor (THTR-300, 750 MWth/300 MWe, Baumer and Kalinowski, 1991 ;Dietrich et al., 2019 ). This prototype, however, met
a number of technical dif ﬁculties. Examples of design issues are the direct insertion of the control rods in the pebble bed (causing
Pyrolytic carbonCeramic
kernel
Coated
ParticlePebble
Particles Compacts Fuel ElementsSilicon Carbide
Uranium Oxycarbide kernel
Fig. 2 TRISO coated particle fuel as the basis for hexagonal block and pebble bed core designs ( Gougar et al., 2020 ).
Table 1 Typical examples for nominal characteristic data of German AVR GLE-4 particles and pebbles and US NGNP particles and compacts.
Coated particle AVR pebble NGNP compact
Kernel composition UO 2 UCO
Kernel diameter [ mm] 502 425
Enrichment [U-235 wt%] 16.76 14
Thickness of coatings [ mm]:
bufferinner PyCSiC
outer PyC92


Kernel diameter [ mm] 502 425
Enrichment [U-235 wt%] 16.76 14
Thickness of coatings [ mm]:
bufferinner PyCSiC
outer PyC92
403540100
403540
Particle diameter [ mm] 916 855
Fuel element (FE) Pebble Compact
Dimensions [mm] B60
(spherical)B12.3/C225
(cylindrical)
Heavy metal loading [g/FE] 6.0 1.27
U-235 content [g/FE] 1.00 0.18
Number of coated particles per FE 9560 3175Volume packing fraction [%] 6.2 35Fraction of factory defective SiC coatings 7.8 /C210
/C06<1.2/C210/C05
Matrix density [kg/m3] 1750 1600
Temperature at ﬁnal heat treatment [/C14C] 1900 1850The High Temperature Gas-Cooled Reactor 515pebble damage) and the pebble discharge system, which allowed for jamming. The THTR was closed in 1989 in the aftermath of the
Chernobyl accident after only 3 years of operation.
In the same period, the Power Nuclear Project (PNP-500, 500 MWth, Neef and Weisbrodt, 1979 ) started in Germany aiming at
using nuclear heat to produce hydrogen by steam methane reforming. This project led to development and testing of large modules
of heat exchangers and a steam reformer. It was brought to a halt in 1989 after the Chernobyl accident, which caused a temporarystop of HTR development worldwide.


of heat exchangers and a steam reformer. It was brought to a halt in 1989 after the Chernobyl accident, which caused a temporarystop of HTR development worldwide.
In the 1980s, Interatom/Siemens in Germany developed the 200 MWth HTR-Modul as the ﬁrst modular pebble bed design con-
sisting of a metallic reactor pressure vessel connected to an adjacent steam generator through a hot gas duct ( Siemens, 1988 ). The
concept features a simpli ﬁed design with a size and power rating chosen to enable passive decay-heat removal after a loss-of-
coolant-accident solely by conduction and radiation. No natural or forced convection is necessary ( Reutler and Lohnert, 1984;
Kugeler et al., 2017 ). Although it was never built, the HTR-Modul has served as the basis for the PBMR in South Africa and for
the HTR-10 and HTR-PM reactors in China.
The Gas Turbine Modular Helium Reactor (GT-MHR, LaBar, 2002 ) is a 600 MWth design developed by a group of Russian and
US enterprises, Framatome in France and Fuji Electric in Japan. It was based on the earlier MHTGR-350 design by General Atomics.It employs an annular prismatic core and utilizes a direct helium Brayton cycle for electricity generation with an ef ﬁciency of up to
48% based on a reactor outlet temperature of 850
/C14C. Extensive analysis has shown that this reactor, and more generally most HTR
designs, are particularly suitable for the incineration of excess plutonium which became an issue in the US and in the former USSR


/C14C. Extensive analysis has shown that this reactor, and more generally most HTR
designs, are particularly suitable for the incineration of excess plutonium which became an issue in the US and in the former USSR
for the implementation of the START I disarmament treaty in 1991. Hydrogen production with the Sulfur-Iodine (S-I) process wasalso envisaged. The Preliminary Design of the reactor plant and GT-MHR prototype power plant was completed in 2001. The
GT-MHR regulatory process started in 2002 but was not completed. More recently, the GT-MHR design was proposed by General
Atomics as one of the options for the US NGNP project until the NGNP Alliance expressed in 2012 a preference for the ANTARESconcept (625 MWth) developed by AREVA ( Lommers et al., 2012 ), based on the GT-MHR but with an indirect steam cycle. A
smaller version (SC-HTGR, 350 MWth) equally with indirect steam cycle was proposed by AREVA/Framatome as well ( AREVA,
2014 ). The GT-MHR was also the basis for the Japanese GT-HTR300 designed by JAEA ( Kunitomi et al., 2004 ).
A review summary on the 7 built reactors (Dragon, Peach Bottom, Fort St. Vrain, AVR, THTR, HTTR and HTR-10) can be found in
(Beck and Pincock, 2011 ). The experience of past experimental and prototype HTRs demonstrated their technical viability, however,
they were not given the time to prove their economic competitiveness with LWR for electricity production. No further developmentswere to occur until the late 1990s when the interest in HTRs revived owing to the needs of low carbon high temperature heat supplyfor a variety of industrial processes.


One of these new projects was the Pebble Bed Modular Reactor (PBMR, Matzner, 2004 ) in the Republic of South Africa. PBMR
Pty. Ltd. is a public-private partnership established in 1999 in response to threats of nation-wide power outages in South Africa andto initiate the development of a modular pebble-bed reactor with a rated capacity of 165 MWe. This design featured a thermal power
of 400 MWth and a direct power conversion with a gas turbine operating with a helium outlet temperature of 900
/C14C. In June 2003
the South African government approved a prototype of 110 MWe for the utility Eskom on the site of Koeberg. This prototype wasintended to be put in service in 2014 and expected to precede a ﬂeet of 24 PBMRs so as to make up 4000 MWe out of the
12,000 MWe additional nuclear capacity planned by 2030. Large facilities dedicated to PBMR speci ﬁc technologies testing were built
in 2007: a “Heat Transfer Test Facility ”,a“Helium Test Facility ”,a“Pebble Bed Micro Model ”and an “Electro-magnetic blower. ”A
fuel laboratory developed manufacturing processes and quality assurance testing techniques in collaboration with NECSA andsuccessfully manufactured coated fuel particles with enriched uranium in December 2008.
In 2009 the PBMR project, like other projects of nuclear equipment in South Africa, faced funding dif ﬁculties and had its busi-
ness plan re-oriented towards the supply of industrial process heat, a dif ﬁcult endeavor in a country with large coal reserves and no
CO
2emission limits. The new focus of the PBMR was on onsite power, cogeneration, seawater desalination and direct process heat


CO
2emission limits. The new focus of the PBMR was on onsite power, cogeneration, seawater desalination and direct process heat
delivery. Target process heat applications included coal-to-liquid or gaseous fuels, petrochemicals, ammonia/fertilizer, re ﬁneries,
steam for oil sand recovery, bulk hydrogen for future transportation and water desalination. Thus, PBMR Ltd. started developingoptions for commercial ﬂeets with Sasol (the South African coal liquefaction company), with the utility Eskom for electricity, as
well as with US and Canadian cogeneration end users including oil sand producers. The PBMR project was accordingly revisited
to develop one standard design that meets all requirements for these applications, thus leading to a cogeneration steam plantwith a power of 200 MWth, a helium temperature of 750
/C14C at the core outlet and a steam generator directly placed in the primary
loop. A conventional subcritical steam turbine was selected for ﬁrst generation plants whereas super-critical cycles were envisaged
for next generation plants.
Due to funding issues and problems in the interaction between PBMR and the South African regulator the project was stopped in
2010. This development was analyzed critically in ( Thomas, 2011 ). Another investigation with negative conclusions from opera-
tional performance of HTR in the past with a pessimistic outlook is summarized in ( Ramana, 2016 ).
Since then, the aforementioned technological problems encountered with test reactors (e.g. moisture leakages into the core) have
been solved to a large extent, so that most recent HTR designs could be deliberately geared towards short-term realization with
minimum R&D efforts and development risks. In addition, with a much longer-term view, a number of research organizations


been solved to a large extent, so that most recent HTR designs could be deliberately geared towards short-term realization with
minimum R&D efforts and development risks. In addition, with a much longer-term view, a number of research organizations
cooperate internationally on the Very High Temperature Reactor, which is usually understood to produce heat above 950/C14Ct o
maximize power conversion ef ﬁciency and to enable ambitious process heat applications such as thermochemical hydrogen
production with the S-I cycle. The VHTR is thus a long-term concept requiring new materials and design codes along with fuel qual-
iﬁcation for the higher temperatures. The very signi ﬁcant progress of this cooperation is summarized in ( Fütterer et al., 2014 ).516 The High Temperature Gas-Cooled ReactorRecent results from test reactors
In the 1990s, the Japan Atomic Energy Agency (JAEA) built a research reactor in Oarai, the High Temperature Test Reactor (HTTR,
(Kunitomi, 2013 ),Fig. 3 ). It is a prismatic block type reactor with annular compacts. It was put in service in 1998 and reached its full
design power of 30 MWth in 2001 with a helium outlet temperature of 850/C14C. Subsequent tests until 2010 have demonstrated the
safe behavior of the reactor. This included reactivity insertion as well as partial and complete loss of forced cooling, but not yet at full
power. The HTTR was successfully operated at the design temperature of 950/C14Cﬁrst in 2004, then for 50 continuous days in 2010.
In parallel with tests on the HTTR, JAEA is developing the S-I thermo-chemical process to produce hydrogen ( Fig. 4 ). Aﬁrst demon-


In parallel with tests on the HTTR, JAEA is developing the S-I thermo-chemical process to produce hydrogen ( Fig. 4 ). Aﬁrst demon-
stration of this process was achieved in 2003 when a continuous production of 30 l/h of hydrogen was maintained for several days.
During the March 2011 earthquake, which triggered the Fukushima accident, the HTTR was only slightly damaged. After extensive
inspection, some repair and after the review by the regulator, a restart is planned for 2021, pending a positive outcome of the publichearing. JAEA intends to conduct further safety tests in the frame of an OECD-NEA Loss of Forced Cooling Project.
Fig. 3 External view of the HTTR building in Japan.
Fig. 4 Schematic of HTTR and future heat use facilities.The High Temperature Gas-Cooled Reactor 517The Institute of Nuclear and New Energy Technology (INET) of the Tsinghua University in China has built the experimental
reactor HTR-10 (10 MWth) ( Dong, 2012; Wu et al., 2002 ;Fig. 5 ) that was put into service in 2000. The successful operation of
this reactor demonstrated an updated pebble bed core HTR technology. In particular, it served as a test bed for fuel, components
and for code validation. The HTR-10 was also employed for district heating of the INET campus in the vicinity of the reactor. Withseveral successful demonstrations of its benign safety performance for the public and the licensing authority it paved the way for
scaling up this technology to the High Temperature Reactor –Pebble bed Module (HTR-PM, 210 MWe) project ( Zhang et al., 2016 )
.


scaling up this technology to the High Temperature Reactor –Pebble bed Module (HTR-PM, 210 MWe) project ( Zhang et al., 2016 )
.
Together with their predecessors, HTTR and HTR-10 have signi ﬁcantly contributed to the establishment of the rather high tech-
nology readiness level both for block type and pebble bed HTR designs.
The case for new next generation HTRs
Why have past HTRs not been successful economically and why do we think that this is changing?
GCRs were developed worldwide, but only the AGRs in the UK remain in commercial operation. After reasonable experiences
with the ﬁrst HTR plants in the UK (Dragon), the US (Peach Bottom Unit 1) and Germany (AVR), national HTR programs ended
with no commercial deployments for various reasons. In the UK, the Thatcher government decided to build PWRs essentially
because of absence of con ﬁrmed economic data for other designs, higher perceived ﬁnancial risk of HTR designs compared to
the mainstream PWR, and because of the then unsolved dif ﬁculty to integrate the HTR into a long-term sustainable closed fuel cycle
that included Fast Breeder Reactors and reprocessing. In Germany, AVR was shut down in 1988 due to public opposition to nuclearenergy, shortly after the Chernobyl accident. In the US, poor capacity factors of the Fort St. Vrain demonstration plant led to its
premature shutdown in 1989. This has coincided with the time period of three decades without new nuclear orders in the US start-
ing with the Three Mile Island accident in 1979.
In general, most HTR operational issues were associated, as already mentioned, with leakages, e.g. moisture ingress that resulted


ing with the Three Mile Island accident in 1979.
In general, most HTR operational issues were associated, as already mentioned, with leakages, e.g. moisture ingress that resulted
in corrosion of components, core temperature oscillations caused by coolant ﬂow bypass and in-core behavior of graphite (cracking,
dimensional changes, movement of blocks and distortions, dust formation) ( Beck and Pincock, 2011 ). Most of these were ﬁrst-of-a-
kind operational issues and took a long time to resolve without the bene ﬁt of the broader industry experience that is dominated by
water-cooled reactors. As a result, it led to poor performance in some HTR reactors, most notably the Fort St. Vrain reactor in the US.
On the positive side, however, the operational experiences with HTRs showed excellent fuel performance and demonstrated the
concept ’s inherent safety features. Many lessons learned through past HTR experiences led to improvements in modern HTR
concepts, such as the use of magnetic bearings in the helium circulator, or the use of a steel pressure vessel for improved reliability
instead of a pre-stressed concrete vessel. Passive cooling systems, requiring no pumps or monitoring systems to initiate them, have
been adopted. The excellent performance of TRISO fuel is further improved by recent extensive research programs ( Electric Power
Research Institute, 2019 ), which bene ﬁtted both fuel types, compact and pebble. These developments eliminate major known issues
experienced by early HTRs and further corroborate HTR safety characteristics.
Ongoing HTR development
The last decade has seen signi ﬁcantly growing interest worldwide in Small Modular Reactors, which the IAEA de ﬁnes as units
producing less than 300 MWe. A snapshot of the very dynamic SMR landscape is given in ( IAEA, 2018 ). These reactors are being


producing less than 300 MWe. A snapshot of the very dynamic SMR landscape is given in ( IAEA, 2018 ). These reactors are being
designed by several classical vendor companies and start-ups for ﬂexibility, affordability, for a wide range of users and applications,
Fig. 5 External view of HTR-10 building in China and Control Room.518 The High Temperature Gas-Cooled Reactorand to replace fossil generation plants including in off-grid areas. These advanced reactors are deployable either as single or multi-
module nuclear power plants, and are designed to be built in factory workshops and shipped to utilities for installation as demand
evolves. Fig. 6 shows how a multi-module pack could be con ﬁgured to polygenerate heat, hydrogen and electricity.
Several designs ensure enhanced safety performance through inherent and passive safety features as well as suitability for cogen-
eration and non-electric applications thus opening opportunities for hybrid energy system architectures combining nuclear, fossiland renewable energy carriers. They have reached different stages of development and target near-term deployment with several
vendor companies participating in feasibility and licensing studies.
About 16 of these SMR designs are HTRs with one currently under construction and commissioning in China (HTR-PM). Several
of these reactors are derivatives or evolutions of earlier parent concepts, e.g. the HTR Modul for the pebble bed designs and the
MHTGR-350 (designed by General Atomics) for several prismatic block designs. For the HTR concepts in Table 2 , publicly available
design information can be found in ( IAEA, 2018 ).
R&D efforts as well as cooperation between all stakeholders (vendors, suppliers, regulators, utilities/end-users, investors, poli-


design information can be found in ( IAEA, 2018 ).
R&D efforts as well as cooperation between all stakeholders (vendors, suppliers, regulators, utilities/end-users, investors, poli-
ticians, public etc.) are ongoing and organized at different national and international levels including GIF, IAEA, OECD-NEA, and
are including economic analyses, as well as novel investment options and licensing approaches, e.g. ( Gougar et al., 2020 ;Kalilainen
et al., 2019 ). While the nuclear accident in Fukushima in 2011 has dealt a blow to nuclear energy development for several years, the
ongoing debate about climate change mitigation has created new interest in low-carbon technologies in several countries and specif-
ically awareness of the need to address the massive energy requirements of the process heat market in industrialized countries. As
shown in Table 2 , interest in the inherently safe, highly ef ﬁcient and versatile HTR technology is steadily growing, and new
Fig. 6 Artist ’s view of a 4-pack modular HTR for process heat, hydrogen production and electricity generation (INL).
Table 2 Summary of HTR-type small modular reactor concepts.
Concept Developer
Pebble Bed
HTR-PM Tsinghua University, China
Xe-100 X-energy, USA
HTMR-100 Steenkampskraal Thorium Ltd., South Africa
PBMR-400 Pebble Bed Modular Reactor SOC Ltd., South Africa
AHTR-100 Eskom Holdings SOC Ltd., South Africa
Hexagonal Block
GTHTR300 Japan Atomic Energy Agency, Japan
MHTGR-350 General Atomics, USA
GT-MHR OKBM Afrikantov, Russian Federation
MHR-T Reactor/Hydrogen Production Complex OKBM Afrikantov, Russian Federation


MHTGR-350 General Atomics, USA
GT-MHR OKBM Afrikantov, Russian Federation
MHR-T Reactor/Hydrogen Production Complex OKBM Afrikantov, Russian Federation
MHR-100 OKBM Afrikantov, Russian Federation
SC-HTGR Framatome Inc., USA
MMR-5, MMR-10 UltraSafe Nuclear Corporation, USA
StarCore HTGR StarCore Nuclear, Canada
U Battery U Battery, UKThe High Temperature Gas-Cooled Reactor 519demonstration projects, in particular for the coupling of the nuclear reactor with a process heat end-user installation, are being
implemented to help de-risk (and possibly shorten the time to) industrial deployment.
Beyond electricity: Emission-free process heat and cogeneration
Because HTRs are particularly ﬁt for process heat applications and cogeneration of heat and power, this section is dedicated to non-
power utilization aspects of nuclear energy, which has very signi ﬁcant potential impacts since it reduces fossil fuel consumption in
areas beyond the electric power market, and thus enhances energy security, further increases the reduction of noxious emissions, and
helps mitigating climate change. Already with earlier reactor types, nuclear cogeneration was performed in many countries and withseveral types of reactors including Light Water Reactors (LWR), Heavy Water Reactors (HWR), and Sodium Cooled Fast Breeder
Reactors (SFBR). District heating (80 –150
/C14C) is probably the most widely found application of nuclear heat: 46 reactors in 12
countries, including for instance Slovakia, Switzerland, Russia and China were and are used for this purpose.
Examples for low temperature applications of nuclear heat include seawater desalination (Japan, Kazakhstan), paper and card-
board industry (Norway, Switzerland), heavy water distillation (Canada), or salt re ﬁning (Germany).


Examples for low temperature applications of nuclear heat include seawater desalination (Japan, Kazakhstan), paper and card-
board industry (Norway, Switzerland), heavy water distillation (Canada), or salt re ﬁning (Germany).
The technology options for nuclear process heat utilization with HTRs were already documented quite early ( Schulten, 1976 ). A
survey of two decades of activities in Germany is given in ( Verfondern, 2007a ), and further potential is outlined in ( Verfondern, 2007b ).
The HTR produces heat at a much higher temperature level (exergy) than the LWR. This opens the possibility to replace a large
number of existing industrial cogeneration plants delivering process steam in the 500 –600/C14C temperature range. Very signi ﬁcant
amounts of such process steam are consumed in the chemical and petrochemical sector as well as in the fertilizer industry, wheretoday this steam is mostly produced by gas or coal ﬁring.
For several stakeholders, in particular in those countries where natural gas is expensive, the prospect of hydrogen production
continues to be the main driver for development and potential deployment of the HTR and VHTR. Process heat from an HTRcan be used for several more or less advanced methods of hydrogen production. The most near-term option is steam methanereforming of natural gas with steam at 700
/C14C, 5.5 MPa. Owing to the external heat supply, more than a third of natural gas is saved.
In the 1980s, the necessary components, e.g. heat exchangers or reformers, were developed and tested under nuclear conditions in
Germany and in Japan ( Harth et al., 1990 ).


In the 1980s, the necessary components, e.g. heat exchangers or reformers, were developed and tested under nuclear conditions in
Germany and in Japan ( Harth et al., 1990 ).
Processes and components for allothermal and steam coal gasi ﬁcation processes were also tested in Germany. They require typi-
cally steam in the range of 750 –900/C14C at 0.1 –4 MPa. Although external heat supply makes coal upgrading more ef ﬁcient, these
processes release large amounts of unwanted CO 2.
These activities were brought to a temporary halt in an anti-nuclear climate after the Chernobyl accident, with inexpensive oil
and gas and in absence of CO 2emission restrictions.
As steam methane reforming to produce hydrogen consumes natural gas and generates CO 2emissions in the process, direct
water splitting methods are under investigation in several countries as a clean alternative. HTRs can provide steam for a rather
low temperature process, the copper-chlorine (Cu-Cl) cycle, requiring steam at just over 500/C14C(Rosen et al., 2012 ). Other prom-
inent hydrogen production methods are (i) High Temperature Steam Electrolysis (750 –950/C14C) where a part of the required water
dissociation energy is delivered in the form of heat, and (ii) thermo-chemical cycles such as the Sulfur-Iodine Cycle where one of thethree process steps (SO
3decomposition) requires heat input at 850/C14C(Yan and Hino, 2011 ). This process is particularly suitable
for VHTR operating at 900 –1000/C14. The market for bulk hydrogen is currently very large and growing fast, with distribution networks


for VHTR operating at 900 –1000/C14. The market for bulk hydrogen is currently very large and growing fast, with distribution networks
already in place in several countries. To justify large-scale production of hydrogen, the development of a speci ﬁc“hydrogen
economy ”is not required. Hydrogen uses include upgrading of increasingly heavy oils to lighter fractions, hydrogenation processes,
hydro coal gasi ﬁcation, metal re ﬁning, ammonia production for fertilizers, the synthesis of methanol or synfuel, or the use of
hydrogen in combination with fuel cells as a transport fuel. For some Asian countries, the replacement of coke by hydrogen fordirect iron ore reduction is of particular interest to cut back emissions from steel making. Finally, hydrogen can also play a role
in carbon capture and utilization processes, which would use CO
2together with hydrogen as a feedstock for the fabrication of
a wide array of possible products ranging from plastics or synfuel for aviation to construction materials. A summary of suchprocesses and products is provided in ( Styring et al., 2011 ).
In the context of energy system integration efforts with growing fractions of variable renewable electricity in many countries, it is
of particular interest that the cogeneration capability of HTRs would allow it to contribute to grid stabilization ( “peak shaving ”), e.g.
by modulating the production of (storable) hydrogen depending on the electricity demand in the grid, similar to what is currentlyenvisaged for wind energy ( “power to gas ”).
To further corroborate the incentive for process heat and hydrogen production with nuclear energy, several market research,


To further corroborate the incentive for process heat and hydrogen production with nuclear energy, several market research,
economic analyses, trade studies, and business plans were recently prepared in several countries, some of which are publicly avail-able (e.g. Angulo et al., 2012 ;Bredimas, 2012 ;INL, 2012 ;Konefal and Rackiewicz, 2008 ;Shropshire, 2013 ).
Outlook
The unique capability of the HTR to produce process heat above 600/C14C makes it an ef ﬁcient reactor type to displace fossil fuels in
various applications such as producing electricity, non-conventional hydrocarbon fuels from coal or biomass, and process heat for520 The High Temperature Gas-Cooled Reactorenergy-intensive industries (oil re ﬁning, petro-chemistry, oil sand recovery, chemistry, steelmaking, etc.). Several market studies
conﬁrmed the potential for the HTR system to be used in such applications while the economic boundary conditions (e.g. price
of natural gas, CO 2tax) for market deployment have become clearer. The inherent safety characteristics of the HTR are a precious
asset in contributing convincing answers to today ’s concerns in terms of nuclear safety, energy security, and climate change.
Current research performed within frameworks supported by GIF, IAEA and OECD-NEA, as well as speci ﬁc national programs
address primarily issues related to R&D, licensing, demonstration, and deployment. In particular, the multinational cooperation
within GIF ( GIF, 2018 ) allows sharing efforts to advance the technologies and to accelerate development in view of licensing
and deployment. Currently, cooperation on the VHTR within GIF focuses on development and quali ﬁcation of (i) fuel, (ii) struc-


and deployment. Currently, cooperation on the VHTR within GIF focuses on development and quali ﬁcation of (i) fuel, (ii) struc-
tural and functional materials, (iii) hydrogen production processes and (iv) computer tools. GIF has also produced guidance for (V)
HTR designers, e.g. in the areas of sustainability, economy, reactor safety, non-proliferation questions or energy system integration.
The cooperation is clearly geared towards producing licensing-relevant information across the signatory countries and has recentlyopened to closer interaction with competing designer and vendor companies. Furthermore, the experimental reactors in Japan(HTTR) and in China (HTR-10) offer unique opportunities to qualify technologies and design codes. The next hurdle towards
deployment is being taken by China with the ongoing commissioning of the HTR-PM demonstrator ( Fig. 7 ). Japan will perform
further safety demonstrations on the HTTR.
Since 2002, the bi-annual International Topical Meeting on High Temperature Reactor Technology is the sole international
conference with focus on HTR and process heat applications ( https://htr2020.org/ ).
Although very substantial results were produced, in particular by the signatories of the GIF VHTR System Arrangement, funding
opportunities for a demonstrator coupled with an end-user process will have to be found soon to capitalize on previous invest-
ments. Several such international initiatives are on the way. Their success will depend on how much and where nuclear will be
allowed to contribute to climate change mitigation, be it for political and public acceptance reasons or for economic boundaryconditions (cheap natural gas, CO
2tax,ﬁnancial risk).
See Also: Fuel Design and Fabrication: TRISO Particle Fuel; Pebble Bed Gas Cooled Reactors; Self-Sustaining Breeding in Advanced Reactors:
Characterization of Selected Reactors.


See Also: Fuel Design and Fabrication: TRISO Particle Fuel; Pebble Bed Gas Cooled Reactors; Self-Sustaining Breeding in Advanced Reactors:
Characterization of Selected Reactors.
Fig. 7 Installation of RPV into HTR-PM reactor building in 2016.The High Temperature Gas-Cooled Reactor 521References
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Gen IV Gas-cooled Fast Reactor system PR&PP White Paper
1
GIF-LFR-WP-Rev9 – Limited: GIF
GIF GAS-COOLED  FAST  REACTOR
PROLIFERATION  RESISTANCE  AND
PHYSICAL  PROTECTION  WHITE
PAPER
Proliferation  Resistance  and Physical  Protection
Working  Group  (PRPPWG)
Sodium-Cooled  Fast Reactor  System  Steering
Committee  (SFR  SSC)
April 2021
SAND2022-6859R
Sandia
National
Laboratories
is
a
multimission
laboratory
managed
and
operated
by
National
Technology
&
Engineering
Solutions
of
Sandia,
LLC,
a
wholly
owned
 subsidiary of Honeywell International Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA0003525. Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
Cover page photos: © Delovely Pics/Shutterstock - © Delovely Pics/Shutterstock - © Pyty /ShutterstockDISCLAIMER
This report was prepared by the Proliferation Resistance and Physical Protection
Working Group (PRPPWG) and the Very-High-Temperature Reactor System Steering
Committee of the Generation IV International Forum (GIF). Neither GIF nor any of its
members, nor any GIF member’s national government agency or employee thereof,
makes any warranty, express or implied, or assumes any legal liability or responsibility
for the accuracy, completeness or usefulness of any information, apparatus, product, or
process disclosed, or represents that its use would not infringe privately owned rights.


for the accuracy, completeness or usefulness of any information, apparatus, product, or
process disclosed, or represents that its use would not infringe privately owned rights.
References herein to any specific commercial product, process or service by trade
name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply
its endorsement, recommendation, or favoring by GIF or its members, or any agency
of a GIF member’s national government. The views and opinions of authors expressed
therein do not necessarily state or reflect those of GIF or its members, or any agency
of a GIF member’s national government.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
iPreface to the 2021-2022 edition of the SSCs, pSSCs & PRPPWG white papers on
the PR&PP features of the six GIF technologies
This report is part of a series of six white papers, prepared jointly by the Proliferation Resistance and Physical
Protection Working Group (PRPPWG) and the six System Steering Committees (SSCs) and provisional
System Steering Committees (pSSCs). This publication is an update to a similar series published in 2011
presenting the status of Proliferation Resistance & Physical Protection (PR&PP) characteristics for each of the
six systems selected by the Generation IV International Forum (GIF) for further research and development,
namely: the Sodium-cooled fast Reactor (SFR), the Very high temperature reactor (VHTR), the gas-cooled fast
reactor (GFR), the Molten salt reactor (MSR) and the Supercritical water–cooled reactor (SCWR).
The Proliferation Resistance and Physical Protection Working Group (PRPPWG) was established by GIF to


The Proliferation Resistance and Physical Protection Working Group (PRPPWG) was established by GIF to
develop, implement and foster the use of an evaluation methodology to assess Generation IV nuclear energy
systems with respect to the GIF PR&PP goal, whereby: Generation IV nuclear energy systems will increase
the assurance that they are a very unattractive and the least desirable route for diversion or theft of weapons-
usable materials, and provide increased physical protection against acts of terrorism.
The methodology provides designers and policy makers a technology neutral framework and a formal
comprehensive approach to evaluate, through measures and metrics, the Proliferation Resistance (PR) and
Physical Protection (PP) characteristics of advanced nuclear systems. As such, the application of the
evaluation methodology offers opportunities to improve the PR and PP robustness of system concepts
throughout their development cycle starting from the early design phases according to the PR&PP by design
philosophy. The working group released the current version (Revision 6) of the methodology for general
distribution in 2011. The methodology has been applied in a number of studies and the PRPPWG maintains a
bibliography of official reports and publications, applications and related studies in the PR&PP domain.
In parallel, the PRPPWG, through a series of workshops, began interaction with the Systems Steering
Committees (SSCs) and Provisional Systems Steering Committees (pSSCs) of the six GIF concepts. White
papers on the PR&PP features of each of the six GIF technologies were developed collaboratively between
the PRPPWG and the SSCs/pSSCs according to a common template. The intent was to generate preliminary
information about the PR&PP merits of each system and to recommend directions for optimizing its PR&PP


the PRPPWG and the SSCs/pSSCs according to a common template. The intent was to generate preliminary
information about the PR&PP merits of each system and to recommend directions for optimizing its PR&PP
performance. The initial release of the white papers was published by GIF in 2011 as individual chapters in a
compendium report.
In April 2017, as a result of a consultation with all the GIF SSCs and pSSCs, a joint workshop was organized
and hosted at OECD-NEA in Paris. During two days of technical discussions, the advancements in the six GIF
designs were presented, the PR&PP evaluation methodology was illustrated together with its case study and
other applications in national programmes. The need to update the 2011 white papers emerged from the
discussions and was agreed by all parties and officially launched at the PRPPWG meeting held at the EC Joint
Research Centre in Ispra (IT) in November 2017.
The current update reflects changes in designs, new tracks added, and advancements in designing the six GIF
systems with enhanced intrinsic PR&PP features and in a better understating of the PR&PP concepts. The
update uses a revised common template. The template entails elements of the PR&PP evaluation methodology
and allows a systematic discussion of the systems elements of the proposed design concepts, the potential
proliferation and physical protection targets, and the response of the concepts to threats posed by a national
actor (diversion & misuse, breakout and replication of the technology in clandestine facilities), or by a
subnational/terrorist group (theft of material or sabotage).
The SSCs and pSSC representatives were invited to attend PRPPWG meetings, where progress on the white


subnational/terrorist group (theft of material or sabotage).
The SSCs and pSSC representatives were invited to attend PRPPWG meetings, where progress on the white
papers was discussed in dedicated sessions. A session with all the SSCs and pSSCs was organized in Paris
in October 2018 on the sideline of the GIF 2018 Symposium. A drafting and reviewing meeting on all the papers
was held at Brookhaven National Laboratory in Upton, NY (US) in November 2019, followed by a virtual
meeting in December 2020 to discuss all six drafts.
Individual white papers, after endorsement by both the PRPPWG and the responsible SSC/pSSC, are
transmitted to the Expert Group (EG) and Policy Group (PG) of GIF for approval and publication as a GIF
document. Cross-cutting PR&PP aspects that transcend all six GIF systems are also being updated and will
be published as a companion report to the six white papers.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
iiAbstract
This white paper represents the status of Proliferation Resistance and Physical Protection (PR&PP)
characteristics for the Very-High-Temperature Reactor (VHTR) reference designs selected by the Generation
IV International Forum (GIF) VHTR System Steering Committee (SSC). The intent is to generate preliminary
information about the PR&PP features of the VHTR reactor technology and to provide insights for optimizing
their PR&PP performance for the benefit of VHTR system designers. It updates the VHTR analysis published
in the 2011 report “Proliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy
Systems”, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG)


in the 2011 report “Proliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy
Systems”, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG)
and the System Steering Committees and provisional System Steering Committees of the Generation IV
International Forum, taking into account the evolution of both the systems, the GIF R&D activities, and an
increased understanding of the PR&PP features.
The white paper, prepared jointly by the GIF PRPPWG and the GIF VHTR SSC, follows the high-level paradigm
of the GIF PR&PP Evaluation Methodology to investigate the key points of PR&PP features extracted from the
reference designs of VHTRs under consideration in various countries. A major update from the 2011 report is
an explicit distinction between prismatic block-type VHTRs and pebble-bed VHTRs. The white paper also
provides an overview of the TRISO fuel and fuel cycle. For PR, the document analyses and discusses the
proliferation resistance aspects in terms of robustness against State-based threats associated with diversion
of materials, misuse of facilities, breakout scenarios, and production in clandestine facilities. Similarly, for PP,
the document discusses the robustness against theft of material and sabotage by non-State actors. The
document follows a common template adopted by all the white papers in the updated series.
List of Authors
Tomooki Shiba PRPPWG Japan Atomic Energy Agency
Kiyonobu Yamashita ABC Nuclear
Keiichiro Hori PRPPWG Japan Atomic Energy Agency
Lap Cheng PRPPWG Brookhaven National Laboratory
Benjamin Cipiti PRPPWG Sandia National Laboratory
Michael Fütterer VHTR SSC
Hans Gougar VHTR SSC
Gerhard Strydom VHTR SSC Idaho National Laboratory


Benjamin Cipiti PRPPWG Sandia National Laboratory
Michael Fütterer VHTR SSC
Hans Gougar VHTR SSC
Gerhard Strydom VHTR SSC Idaho National Laboratory
Christial Pohl
Abderrafi Ougouag
Hideyuki Sato VHTR SSC Japan Atomic Energy Agency
Acknowledgements
The current document updates and builds upon the 2011 VHTR PR&PP White Paper. Thanks are due to the
original author of the 2011 SFR PR&PP White Paper, David Moses. The in depth reviews by Giacomo G.M.
Cojazzi (PRPPWG, European Commission Joint Research Centre) and Kevin Hesketh (National Nuclear
Laboratory) are particularly appreciated. A special thanks to the PRPPWG Technical Secretary Gina
Abdelsalam (OECD-NEA) who ably readied the final manuscript for publication. SNL is managed and operated
by NTESS under DOE NNSA contract DE-NA0003525.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
iiiTable of contents
1. Overview of Technology..........................................................................................................................1
1.1. Description of the prismatic VHTR .....................................................................................................1
1.2. Description of the pebble bed VHTR..................................................................................................5
1.3. Current system design parameters and development status.............................................................7
2. Overview of Fuel Cycle(s)........................................................................................................................8
3. PR&PP Relevant System Elements and Potential Adversary Targets ..............................................10
3.1. System elements related to fuel fabrication site for B-VHTR and P-VHTR......................................11
3.1.1. Fresh Fuel fabrication...............................................................................................................11
3.1.2. Fresh Fuel shipment.................................................................................................................12


Figure 14: Movement of the fuel pebbles........................................................................................................ 17
Figure 15: Plutonium build-up in a PBMR fuel element in an equilibrium core ..............................................20
Figure 16: Material Balance Areas and Key Measurement Points of B-VHTR and P-VHTR..............................22
List of Tables
Table 1: Calculated plutonium isotopic fractions for PBMR spent fuel as a function of initial enrichment and
discharge burn-up ...........................................................................................................................................20
Table 2: Summary ............................................................................................................................................21Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
vList of Acronyms
CNEC China Nuclear Engineering & Construction Group
C/S Containment/Surveillance
DIV Design Information Verification
GA General Atomics
GIF Generation-IV International Forum
GT-MHR Gas-Turbine Modular Helium Reactor
HALEU High-Assay Low-Enriched Uranium
HEU Highly Enriched Uranium
HTR High Temperature Reactor
HTR-PM High-Temperature Gas-cooled Reactor Pebble-Bed Module
HTR-TN High-Temperature Reactor-Technology Network
IAEA International Atomic Energy Agency
INET Tsinghua University's Institute of Nuclear and New Energy Technology
JAEA Japan Atomic Energy Agency
KAERI Korea Atomic Energy Research Institute
KI Kurchatov Institute
LEU Low Enriched Uranium
LWR Light Water Reactor
MOX Mixed Oxide
NHDD Nuclear Hydrogen Development and Demonstration
NNSA National Nuclear Security Administration
OKBM Experimental Design Bureau of Mechanical Engineering in Nizhniy-Novgorod
PBMR Pebble Bed Modular Reactor
PP Physical Protection
PR Proliferation Resistance
PR&PP Proliferation Resistance & Physical Protection
PWR Pressurized Water Reactor


PBMR Pebble Bed Modular Reactor
PP Physical Protection
PR Proliferation Resistance
PR&PP Proliferation Resistance & Physical Protection
PWR Pressurized Water Reactor
RCCS Reactor Cavity Cooling System
RDD Radiological Dispersion Device
SC-HTGR Steam Cycle High-Temperature Gas-Cooled Reactor
SSC System Steering Committee
TRISO Tri-Isotopic
UOX Uranium Oxide
VHTR Very-High-Temperature ReactorVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper
vi(This page has been intentionally left blank)Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
11. Overview of Technology
The Very High Temperature Reactor (VHTR) design descriptions, technology overviews and
discussions of issues, concerns and benefits documented in this White Paper establish the
bases to support, as the designs evolve, more detailed assessments of proliferation resistance
and physical protection (PR&PP).
The assessments will be made using the methodology developed for evaluating PR&PP of the
Generation IV reactors [1] with consideration of related reports [2-4]. In April 2017, as a result
of a consultation with all the GIF SSCs and pSSCs a joint workshop was organized and hosted
at OECD-NEA in Paris. The need to update the 2011 white papers [2] emerged from the
discussions and was agreed by all parties and officially launched in November 2017.
Therefore, this white paper was written, based on the status of the six GIF system design
concepts, considering the designs’ evolution in the last decade.
Various versions of the VHTR are under development in several countries that are members


concepts, considering the designs’ evolution in the last decade.
Various versions of the VHTR are under development in several countries that are members
of the Generation IV International Forum (GIF), including the People’s Republic of China,
France, Japan, the Russian Federation, Republic of South Africa, Republic of Korea, Canada,
United Kingdom and the United States of America. The VHTR is a helium-cooled, graphite-
moderated, graphite-reflected, metallic-vessel reactor that can use various power conversion
cycles for electricity production. Co-generation of process steam and high-temperature
process heat for chemical process and hydrogen co-production are additional uses for the
technology. The major VHTR design options that potentially affect PR&PP can be categorized
as follows:
Prismatic versus pebble fuel
Direct versus indirect power conversion cycles
Water versus air cooled Reactor Cavity Cooling System (RCCS)
Filtered confinement versus low leakage containment
Underground versus above-ground nuclear islands
The two VHTR basic design concepts considered here are the Prismatic VHTR and the Pebble
Bed VHTR. Note that a lot of the information described in this section was taken from reference
[5].
1.1. Description of the prismatic VHTR
The safety basis for all the VHTR is to design the reactor to be passively safe, thereby avoiding
the release of fission products under all conditions of normal operation and accidents including
most of the beyond design basis events. This passive safety aspect of the design should make
the VHTR less vulnerable to a significant risk of "radiological sabotage" through malevolent
acts.
There are currently five concepts for the prismatic VHTR under consideration by different GIF


the VHTR less vulnerable to a significant risk of "radiological sabotage" through malevolent
acts.
There are currently five concepts for the prismatic VHTR under consideration by different GIF
countries. The first two of the following have the generic features of low-enriched uranium
(LEU) and plutonium-fuelled block-type cores and are sufficiently developed to be considered
further here as examples for PR&PP assessment. Except for the second concept discussed
below, prismatic VHTRs are being designed assuming the initial use of a once-through LEU
fuel cycle.
United States – Work on the Modular HTGR began with General Atomics (GA) in the 1980s.
The GA concepts include prismatic cores driving either a direct or indirect cycle, an air-cooled
RCCS, filtered confinement, and either a steam cycle (350 MWt MHTGR) or a 600 MWt gas
turbine cycle (GT-MHR) [6-8]. The MHTGR was the subject of a pre-application design review
by the Nuclear Regulatory Commission. GA has ceased development and design efforts but
Framatome (USA), formerly Areva USA, is pursuing a similar concept in the 625 MWt SC-Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
2HTGR. The completion of design and licensing of the SC-HTGR is projected to take at least
10 years. Framatome has also completed some work on a higher temperature HTGR
(designated ANTARES) [9, 10], which began as a collaboration in France with other
EURATOM participants in the High Temperature Reactor-Technology Network (HTR-TN). The


(designated ANTARES) [9, 10], which began as a collaboration in France with other
EURATOM participants in the High Temperature Reactor-Technology Network (HTR-TN). The
ANTARES Modular HTR is also envisioned to be a 600 MWt cogeneration plant; however, the
schedule for completion of research and development depends on end-user engagement.
Smaller (<80 MWt) prismatic concepts are being pursued by the UltraSafe Nuclear and
StarCore Nuclear companies, mainly for off-grid communities and mines in Canada.
Russian Federation – In cooperation with GA and the U.S. Department of Energy (DOE)
National Nuclear Security Administration (NNSA), the Experimental Design Bureau of
Mechanical Engineering (OKBM) in Nizhniy-Novgorod with partners at the Kurchatov Institute
(KI) and the A.A. Bochvar All-Russian Scientific Research Institute for Inorganic Materials
(VNIINM) in Moscow is designing a Russian version of the GA GT-MHR to disposition excess
weapon-grade plutonium; however, OKBM is also analyzing alternative fuel cycles for the
Russian GT-MHR [11]. The deployment of the Russian GT-MHR is subject to DOE/NNSA joint
funding to complete necessary research and development.
Japan – The Japan Atomic Energy Agency (JAEA) continues development work that started
under the former Japan Atomic Energy Research Institute (JAERI) on the Gas Turbine High
Temperature Reactor 300 for Cogeneration (GTHTR300C) [12], which will scale up the
technology from the JAEA 30 MWt High Temperature Engineering Test Reactor (HTTR) into
a 600 MWt configuration. The reactor design is based on a prismatic core and can achieve a


technology from the JAEA 30 MWt High Temperature Engineering Test Reactor (HTTR) into
a 600 MWt configuration. The reactor design is based on a prismatic core and can achieve a
reactor outlet temperature of 950°C.
Republic of Korea – The Korea Atomic Energy Research Institute (KAERI) is pursuing the
Nuclear Hydrogen Development and Demonstration (NHDD) Project; the NHDD reactor is to
be limited to 200 MWt (based on the maximum reactor vessel diameter, 6.5 m, that can be
fabricated in-country) with no decision yet made as to fuel/core type (pebble bed or prismatic)
[13].
United Kingdom – U-Battery Limited is holding the U-Battery project; the U-Battery reactor is
to be categorized as small modular reactor with 20 MWt with prismatic core design. The
strategic goal is to have a first-of-a-kind U-Battery operating by 2028.
Technology summaries can be found for each vendor-proposed design option in the respective
references provided above. SC-HTGR and ANTARES are proposed to be constructed as
modules to be built in sets of four or more modules per site. As indicated above, the baseline
fuel design for the first modules uses LEU as Tri-Isotropic (TRISO)-coated particle fuel in a
once-through fuel cycle. The Russian version of the General Atomics GT-MHR will incorporate
excess weapon plutonium in TRISO-coated fuel particles with the addition of erbium containing
167Er to provide a neutron poison with a thermal neutron capture resonance to guarantee a
negative moderator temperature reactivity coefficient.
 Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper


167Er to provide a neutron poison with a thermal neutron capture resonance to guarantee a
negative moderator temperature reactivity coefficient.
 Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
3Figure 1: Illustration of Coated Particle Fuel in the Prismatic Fuel design [14]
The TRISO-coated particle fuel (see Figure 1) has a small-diameter (nominally 200-500 μm)
spherical ceramic fuel kernel of either uranium oxide or uranium oxycarbide, or mixed oxides
of other actinides. The kernel is coated with four coating layers consisting sequentially of low-
density porous pyrocarbon (buffer), an inner high density pyrocarbon (IPyC), silicon carbide
(SiC)1 and an outer high density pyrocarbon (OPyC) for better contact with the matrix material
which is generally carbon but could also be SiC. The first three coatings on the fuel particles
serve as the primary containment preventing the release of fission products. Plant
configurations and operating conditions are being designed appropriately to limit fuel
temperatures during both normal operations and accident conditions so as to preclude the
release of fission products. The coated particles are loaded into fuel compacts (sticks) held
together by graphitized carbon or silicon carbide. The fuel compacts are loaded into holes in
hexagonal prismatic block fuel elements. Fuel elements are stacked in the reactor core with
fissile and neutron burnable poison loadings tailored so that the power distribution is peaked
toward the top of the core where the inlet cooling gas has the lowest temperature. The power
density is lowest in the bottom of the core where the temperature of the outlet coolant is


toward the top of the core where the inlet cooling gas has the lowest temperature. The power
density is lowest in the bottom of the core where the temperature of the outlet coolant is
highest. The fuel and burnable poison loading patterns are specified so that the peak fuel
temperature will be below the limit for normal operation, which is 1250ºC for TRISO-coated
fuel particles with SiC coatings and more than 1600 ºC in accident conditions.
Spent fuel is retained in cooled storage containers that are embedded underground and
located adjacent to the reactor cavity. Prismatic spent fuel, which is unloaded from the core
during periodic refueling shutdowns, can be tracked remotely by cameras viewing the serial
numbers on the fuel elements during handling and storage operations. Since each fuel element
is loaded with less than 4 kg of LEU, the plutonium content at full burnup (~120 GWD/MT) will
be small (~60-70 g) and isotopically degraded compared to weapon-grade plutonium.
The current concepts for the energy utilization from the prismatic VHTRs are based on:
direct Brayton cycle for electricity generation,
indirect steam generation for process heat and/or electricity generation,
1On-going research focuses on replacing SiC coatings with zirconium carbide (ZrC) coatings to achieve higher
temperature limits (~2000ºC) for fission product retention during accidents and to reduce diffusion of radioactive-
silver.Uranium Oxide or Uranium OxycarbidePorous Carbon BufferSilicon Carbide or Zirconium CarbidePyrolytic Carbon
PARTICLE
SCOMPACTS FUEL ELEMENTSTRISO Coated fuel particles (left) are formed
into fuel rods (center) and inserted into


PARTICLE
SCOMPACTS FUEL ELEMENTSTRISO Coated fuel particles (left) are formed
into fuel rods (center) and inserted into
graphite fuel elements (right).Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
4indirect heat transfer to process heat user (e.g., Hydrogen production).
The vessel configuration for the direct cycle GT-MHR is illustrated in Figure 2, and the reactor
building option for the GT-MHR is illustrated in Figure 3.  Although the GT-MHR is no longer
under development, the plant layout for the Framatome SC-HTGR is very similar.
Figure 2: GT-MHR Reactor, Cross-Duct and PCU Vessels [2]
Figure 3: GT-MHR Fully-Embedded Reactor Building [2]
Power Conversion Unit (PCU)
Reactor VesselVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper
5In many modular VHTRs under development, the reactor vessel and power conversion unit are placed
underground, which enhances physical protection for the plant.
1.2. Description of the pebble bed VHTR
All modern pebble bed VHTR concepts trace their design features to the HTR Module 200
MWt concept developed in Germany in the 1980s. There is currently one national program for
a pebble bed VHTR and one commercial endeavor in the United States.
South Africa – PBMR Pty. Ltd. is a public-private partnership established in 1999 in response
to threats of nation-wide power outages in South Africa and to initiate the development of a
modular pebble-bed reactor (PBMR) with a rated capacity of 165 MWe. This design featured


to threats of nation-wide power outages in South Africa and to initiate the development of a
modular pebble-bed reactor (PBMR) with a rated capacity of 165 MWe. This design featured
a thermal power of 400 MWth and a direct power conversion with a gas turbine operating with
a helium outlet temperature of 900 ºC. Due to funding issues and problems in the interaction
between PBMR and the South African regulator the project was stopped in 2010. However, a
number of research organizations cooperate internationally on the VHTR with a longer-term
view as it requires new materials and design codes along with fuel qualification for the higher
temperatures.
United States – The 200 MWt Xe-100 is a concept under development by the X-Energy
company with some support from the US Government [15-17]. It features a recirculating pebble
bed core driving a steam cycle. Formal conceptual design activities have started, and X-Energy
is also pursuing TRISO fuel manufacturing capability with Centrus. X-Energy is pursuing
deployment of the first commercial reactor by 2030.
People’s Republic of China (PRC) – The China Huaneng Group in a consortium with the
China Nuclear Engineering & Construction Group (CNEC) and Tsinghua University's Institute
of Nuclear and New Energy Technology (INET) has been developing and preparing near-term
(starting in 2010, commissioning completed in 2021) construction of the 2 x 250 MWt, steam-
cycle High-Temperature Reactor-Pebble-bed Module (HTR-PM) [18, 19]; the HTR-PM, which
builds on the success of the Tsinghua University's HTR-10 test reactor [20], is being


builds on the success of the Tsinghua University's HTR-10 test reactor [20], is being
constructed in two module units producing 500 MWt and 210 MWe. Each power plant
comprises two reactor modules with individual steam generators sharing a single turbo-
generator. A 6-module, 600 MWt generating station is undergoing design.  The 6-module plant
is sized to fit into a reactor building roughly that of a large PWR.
The pebble bed reactors share the same passive safety features as the prismatic VHTRs but
have less excess reactivity due to on-line refueling. The LEU fuel for the pebble bed VHTRs is
TRISO-coated particles compacted into tennis ball size spheres, as illustrated in Figure 4.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
6Figure 4: Illustration of Coated Particle Fuel in Pebble Fuel Element [2]
The pebble fuel is usually not tracked individually by serial number as in the prismatic core,
but elements are counted, characterized, and checked following each of multiple re-
circulations until they achieve the target burnup based on radioactivity measurements.
Following several passes of each pebble through the core during on-line pebble recirculation,
when measured pebble activity indicates sufficient burnup, the pebble is transferred to a
storage container with a record kept of the number of pebbles transferred. Once pebble spent
fuel is in the storage container, radiation monitoring is used to quantify by inference the amount
of spent fuel present since, with no more than 0.12 grams of plutonium per pebble, it would
take several tens of thousands of pebbles (or several metric tons by total mass and cubic


of spent fuel present since, with no more than 0.12 grams of plutonium per pebble, it would
take several tens of thousands of pebbles (or several metric tons by total mass and cubic
meters by volume) to be diverted to constitute the basis for recovering a significant quantity of
plutonium. Further, at a burnup around 90 GWD/MT for the HTR-PM or 150 GWD/MTMT for
the Xe-100, the plutonium isotopic composition in the pebble spent fuel is degraded
significantly compared with that of weapon-grade plutonium.
The reactor vessel arrangement for the Xe-100 concept is illustrated in Figure 5, showing the
associated spent fuel storage location to the right of the reactor vessel. The reactor vessel and
vessel arrangement for the 250 MW-thermal steam-cycle PRC HTR-PM are illustrated in
Figure 6, with the steam generator below and to the left of the reactor vessel.
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
7Figure 5: X-Energy Xe-100 [20]Figure 6: 250 MWt HTR-PM Reactor Building
Elevated above Ground Level with Steam
Generator; Spent Fuel Storage Not Shown [2]
1.3. Current system design parameters and development status
The key design parameters for each concept (both prismatic and pebble bed) are presented
in Appendix VHTR.A. The construction of HTR-PM had started in 2012, and commissioning
will continue into 2021 with subsequent connection to the grid. All other concepts require
further development and are at least ten years in the future.


will continue into 2021 with subsequent connection to the grid. All other concepts require
further development and are at least ten years in the future.
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
82. Overview of Fuel Cycle(s)
A comparison of the vendor-proposed VHTR fuel cycle parameters is provided in Appendix
VHTR.B. The information in Appendix VHTR.B is taken either from the references given in
Section 1 or is inferred from these references where no specific information has been provided
by the vendors.
The baseline fuel cycle for the first generation VHTR is the once-through fuel cycle using LEU
fuel enriched to between 8 and 16% U-235. The Russian Federation is simultaneously
pursuing the GT-MHR as a “deep-burn” option for weapon-grade plutonium (Pu) disposition.
The use of highly enriched uranium (HEU) as HTGR fuel, as was done in the past, is no longer
acceptable by many nation states because exporting Special Nuclear Material (SNM), or fissile
production technology, is considered a controlled export.  However, this policy position is not
universally held by all states.  The same is true of separated plutonium, even when considering
a deep-burn fuel cycle as the one currently being considered by the Russian Federation.  Some
regulatory authorities allow for separated plutonium whereas others do not due to their own
domestic policy, export control regulations, or both.  Additionally, under the regulatory
framework of some states, the HEU and separated Pu require heightened safeguards and
security measures, compared to LEU, which incurs added complexity and cost to the fuel cycle.
X-Energy is considering a range of other fuel cycle options for future reactor deployments


security measures, compared to LEU, which incurs added complexity and cost to the fuel cycle.
X-Energy is considering a range of other fuel cycle options for future reactor deployments
including plutonium disposition and transuranic elements (TRU)/MA transmutation and the use
of thorium (Th-232) as a fertile component for high-conversion fuel. Each of these options,
including the so-called deep-burn options, is currently based on an initial once-through
irradiation without recycle, although technologies to reprocess and recycle TRISO fuel are also
under consideration or initial development and were studied extensively in the past at
laboratory and pilot scale for HEU/Th fuels. The ongoing research and development and the
historic experience provide a reasonably sound basis to have confidence in the ability to close
the VHTR fuel cycle in the future, if needed. Note that those alternative fuel cycles are a task
in the GIF VHTR Fuel and Fuel Cycle Project.
The fuel cycle options for VHTRs can be categorized in three ways described below.
First, VHTRs can operate with either pebble or prismatic fuels. Pebble bed reactors operate
with on-line refueling. This enables operation with very low excess reactivity and without
burnable neutron poison, typically only sufficient to overcome the neutron poisoning effects of
xenon that occur following power reductions. Prismatic fueled reactors require periodic
refueling outages and thus operate with substantially higher average excess reactivity
compensated by burnable neutron poison, but allow substantially greater flexibility in fuel
zoning and shuffling.
Second, VHTR fuel cycles can be categorized by the types of fuel particles used, as follows:
LEU fuel particles with or without natural uranium fertile fuel particles.
Pu fuel particles.


Second, VHTR fuel cycles can be categorized by the types of fuel particles used, as follows:
LEU fuel particles with or without natural uranium fertile fuel particles.
Pu fuel particles.
TRU or MA fuel particles.
U-233 fuel particles (or U-233 with U-238).
Thorium (or thorium with uranium) fertile fuel particles.
Pu/Th-232 and/or Pu/U-238 in mixed oxides (MOX).
The first four types of particles contain fissile isotopes that are required to support criticality of
the reactor. The LEU particles also contain the fertile isotope U-238 and in some designs may
contain fertile particles of natural uranium. However, with the VHTR’s thermal spectrum,
thorium has somewhat better properties as a fertile isotope, so, for core designs that add fertile
material, thorium fuel particles may replace the use of natural uranium in the future. This
thorium may be mixed with a small amount of uranium to dilute and “denature” the fissile U-
233 produced by neutron absorption in thorium. In general, it can be expected that future VHTR Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
9reactors will operate with fuels composed of some mix of the six particle types listed above.
Each particle type involves specific technical issues for fabrication, with some being more
challenging than others.
Third, VHTR fuel cycles can be categorized by whether or not the spent fuel is discarded or
recycled. Recycle may occur with either aqueous or pyroprocessing methods, and recycled
materials may be returned to VHTRs or LWRsLWR or sent to fast reactors.  Either method


recycled. Recycle may occur with either aqueous or pyroprocessing methods, and recycled
materials may be returned to VHTRs or LWRsLWR or sent to fast reactors.  Either method
would require a ‘head-end’ process to de-consolidate the coated particles from the graphite
and ‘crack’ the silicon carbide coating so that the heavy metal kernel can be leached. possible
but has not been demonstrated on a commercial scale
Except for the LEU once-through cycle and the historic testing and use of HEU/Th, all other
fuel cycles for the VHTR represent future possibilities given also that there is likely to be a
requirement for several years of effort and a significant financial investment for supporting
research (including irradiation testing of laboratory-scale, pilot-scale and industrial-scale
fabrications of candidate fuels) to qualify the fuel forms for the alternative fuel cycles. Currently,
only LEU fuel is being tested for qualification, so alternative fuel options are likely years away
in development. Regarding the reprocessing of VHTR fuels, the PUREX process can be
applied with specific head end processes to separate the fuel particles from the graphite matrix
and fuel kernels from the coatings, which becomes a strong PR advantage. The process yields
large quantities of 14C-contaminated CO 2 or carbon sludge that must be treated, conditioned,
and disposed safely. Note that the reprocessing technology for irradiated Thorium fuel
(THOREX process, similar to the PUREX process) is ready for application, but its
demonstration at an industrial level has not been carried out yet.
The challenges of realizing such fuel cycles at the commercial level have become major R&D
topics internationally, and many efforts are ongoing. For one of those examples, see the


The challenges of realizing such fuel cycles at the commercial level have become major R&D
topics internationally, and many efforts are ongoing. For one of those examples, see the
reference [22]. In addition, the waste graphite and SiC can be decontaminated to reduce waste
volume. Studies on the subject are ongoing in several countries.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
103. PR&PP Relevant System Elements and Potential Adversary Targets
Although the shape of the fuel is different for the block type very high temperature gas reactor
(B-VHTR) and pebble bed type very high temperature gas reactor (P-VHTR), their safeguards
features and the physical protection features have some similarities because the fuel is made
from a mixture of coated fuel particles with graphite powder that is sintered. Figure 7 shows
sketches of reactors of the B-VHTR and P-VHTR types and their respective fuel elements.
Figure 7: B-VHTR and P-VHTR as well as their fuel elements
In order to retrieve a significant quantity of nuclear material from used VHTR fuels, it is
necessary process metric tons and tens of cubic meter quantities of carbon-encased nuclear
fuel using either grind-leach, burn-leach of electrolysis in nitric acid, the technology for which
is still not matured to industrial level. The cost of removing and storing the large volume of
separated graphite should be considered a proliferation resistance feature.  Such large
quantities are a necessity to retrieve weapons usable fissile material and would be difficult to
conceal by a proliferating state.
The use of LEU is currently planned in both B-VHTR and P-VHTR due to its low


conceal by a proliferating state.
The use of LEU is currently planned in both B-VHTR and P-VHTR due to its low
proliferation characteristics.  For states that own their own domestic enrichment capability, the
raw LEU material for fresh fuel fabrication is more attractive than the fabricated graphite fuel
forms (block or pebble since a lower level of effort would be required for its diversion or
acquisition from the system elements at fuel fabrication sites or product side of reprocessing
sites etc. For states that import the as-fabricated graphite fuels, the attractiveness may be
considered similar between the fresh and spent fuels.  This is because a similar amount of
effort is required to crack the SiC barrier as discussed previously.
It is noteworthy from a security standpoint, IFCIRC/225 (the IAEA Standard on nuclear
security) allows some credit for radioactive source term regarding the degree of physical
protection.  However, once a Category II (i.e., U-235/U<20%) fuel has decayed sufficiently, the
security threat and categorization are the same between fresh and used fuel. The Standard
also prescribes an elevated security posture for High Assay Low Enriched Uranium (HALEU),
10 wt.% ≤ U-235/U < 20 wt.%.  For example, it specifies that HALEU be stored in the facility’s
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
11protected area, as opposed to the limited access area.  It also calls out the need for increased
communication and verification for transport.  Similarly, it elevates the importance of armed
guards (i.e., a dedicated security organization) during transport and storage at facilities.


communication and verification for transport.  Similarly, it elevates the importance of armed
guards (i.e., a dedicated security organization) during transport and storage at facilities.
The "system elements" for B-VHTR and P-VHTR are shown in Figure 8 and Figure 9,
respectively.
Figure 8: B-VHTR System element
Figure 9: P-VHTR System element
The system elements of the both VHTR types are principally the same except for the unloading
and reloading of fuel blocks of the B-VHTR and the recirculating fuel spheres of the P-VHTR.
The common system elements for both VHTRs are discussed in the following.
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
123.1. System elements related to fuel fabrication site for B-VHTR and P-VHTR
3.1.1. Fresh Fuel fabrication
The raw constituents of fresh fuel (Uranium hexafluoride, nitrate, or oxide of LEU, LEU/Pu
(MOX), LEU/Th or Pu / Th(MOX)) are brought into the fuel fabrication facility. Fuel elements
(fuel compacts for block type fuel or fuel spheres) containing TRISO-coated fuel particles
sintered with graphite powder are manufactured and shipped out to reactor sites. LEU is
currently intended for use in B-VHTR and P-VHTR due to its lower proliferation risk, specifically
with respect to material attractiveness. Fuel based on LEU / Th, LEU/Pu (MOX) or Pu / Th may
be used in future VHTRs. Raw material for fresh fuel fabrication is the most attractive target


with respect to material attractiveness. Fuel based on LEU / Th, LEU/Pu (MOX) or Pu / Th may
be used in future VHTRs. Raw material for fresh fuel fabrication is the most attractive target
over the entire set of system elements of B-VHTR and P-VHTR, from fuel fabrication to final
disposal, since it would require the least effort to divert and use for fabrication of NEDs (hence
it will require more attention and protection). However, it should be noted that the material type
will be the same if present in the fuel fabrication facility or in the fresh fuel in terms of the IAEA
safeguards target material. In any case the material will require further processing for use in a
NED unless it is already in suitable form. See the discussion of the section 2 of the reference
[23]. It should be also noted that safeguarding bulk material is more complicated than items.
The fuel kernels of the coated fuel particles are manufactured by dropping uranyl nitrate stock
solution into ammonia water as shown in Figure 10.
Figure 10: Fuel kernel fabrication through dropping uranyl nitrate stock solution [24]
Implementation of adequate measures of Containment and Surveillance (C/S) and physical
protection needs to be enforced over those raw constituents of fresh fuel according to the
grade of nuclear material such as LEU, LEU/Th, LEU/Pu, and Pu / Th.
Every fuel block of B-VHTR is stamped with identification numbers (IDs). On the other hand,
there is no ID on fuel pebbles of P-VHTR, which requires a different safeguards approach as
B-VHTR (item-based safeguards can be applied for B-VHTR). In contrast, quasi-bulk type


there is no ID on fuel pebbles of P-VHTR, which requires a different safeguards approach as
B-VHTR (item-based safeguards can be applied for B-VHTR). In contrast, quasi-bulk type
safeguards are needed for P-VHTR. In the past, however, there have been cases where
safeguards were implemented by assigning IDs to pebbles at the research reactor level, but
not for online monitoring during the re-loading procedure. As one of the ongoing efforts, see
the reference [25]. Fabrication also involves scrap recovery and recycling within the supplier's
fuel fabrication facility. Non-recoverable scrap materials are stored for disposition as low-level
radioactive waste. The isotopes U-235, U-233 and Pu are attractive for adversaries aiming for
manufacturing NEDs. However, once these nuclear materials are encased in graphitized
carbon as the kernel of coated fuel particles of fuel elements of both B-VHTR and P-VHTR,
their use in NEDs poses major difficulties for an adversary. The separation of the kernel from
coated fuel particles is difficult due to the stable chemical and mechanical characteristics of
carbon and SiC layers. Techniques such as grind-leach or burn-leach of electrolysis in nitric
acid are necessary, but they have not yet been matured to industrial level. Also, in order to
acquire significant amounts of nuclear materials, metric tons and tens of cubic meter quantities
of carbon and SiC layers from the coated fuel particles and the graphite matrix surrounding
them must be processed.
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
133.1.2. Fresh Fuel shipment