Datasets:

oMuransky

/

vhtr-1

like 0

2021_MF_VHTR.pdf_0

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_1

2021_MF_VHTR.pdf

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-

2021_MF_VHTR.pdf_2

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_3

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_4

2021_MF_VHTR.pdf

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.

2021_MF_VHTR.pdf_5

2021_MF_VHTR.pdf

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 513high 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

2021_MF_VHTR.pdf_6

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_7

2021_MF_VHTR.pdf

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.

2021_MF_VHTR.pdf_8

2021_MF_VHTR.pdf

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 )

2021_MF_VHTR.pdf_9

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_10

2021_MF_VHTR.pdf

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.

2021_MF_VHTR.pdf_11

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_12

2021_MF_VHTR.pdf

/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.

2021_MF_VHTR.pdf_13

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_14

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_15

2021_MF_VHTR.pdf

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-

2021_MF_VHTR.pdf_16

2021_MF_VHTR.pdf

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 ) .

2021_MF_VHTR.pdf_17

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_18

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_19

2021_MF_VHTR.pdf

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-

2021_MF_VHTR.pdf_20

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_21

2021_MF_VHTR.pdf

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).

2021_MF_VHTR.pdf_22

2021_MF_VHTR.pdf

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 ).

2021_MF_VHTR.pdf_23

2021_MF_VHTR.pdf

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

2021_MF_VHTR.pdf_24

2021_MF_VHTR.pdf

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,

2021_MF_VHTR.pdf_25

2021_MF_VHTR.pdf

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-

2021_MF_VHTR.pdf_26

2021_MF_VHTR.pdf

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.

2021_MF_VHTR.pdf_27

2021_MF_VHTR.pdf

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 Angulo, C., et al., 2012. EUROPAIRS: The European project on coupling of high temperature reactors with industrial processes. Nuclear Engineering a nd Design 251 (2012), 30 –37. AREVA, 2014. HTGR Information Kit. March 2014.Baumer, R., Kalinowski, I., 1991. THTR commissioning and operating experience. Energy 16 (1991), 59 –70. Beck, J.M., Pincock, L.F., 2011. High Temperature Gas-Cooled Reactors dLessons Learned Applicable to the Next Generation Nuclear Plant. INL Report INL/EXT-10-19329 Revision 1, April 2011. Beech, D.J., May, R., 1999. Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering. In: The First In formation Exchange Meeting on Survey on Basic Studies in the Field of High Temperature Engineering. OECD NEA, Paris, France, 27–29 September 1999 . Bredimas, A., 2012. Results of a European industrial heat market analysis as a pre-requisite to evaluating the HTR market in Europe and elsewhere. In: Proc. HTR 2012. Tokyo, Japan, 28 October–1 November 2012 .

2021_MF_VHTR.pdf_28

2021_MF_VHTR.pdf

Japan, 28 October–1 November 2012 . Daniels, F., 1944. Suggestions for a High-Temperature Pebble Pile. MUC-FD-8; N-1668b. Chicago University Metallurgical Laboratory, Chicago, Ill inois. Dietrich G, Michels J, Cleve U (2019) Personal communication 2015 –2020. Dong, Y., 2012. China ’s activities in HTGRs HTR-10 and HTR-PM. In: IAEA Course on High Temperature Gas Cooled Reactor Technology. Beijing, China, 22–26 October 2012 . Electric Power Research Institute, 2019. Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance. Topical Rep ort EPRI-AR-1, May 2019. Fortescue, P., 1975. Advanced HTGR systems. Annals of Nuclear Energy 2 (11 –12). Fütterer, M.A., Fu, L., Sink, C., de Groot, S., Pouchon, M., Kim, Y.W., Carré, F., Tachibana, Y., 2014. Status of the very high temperature reactor syst em. Progress in Nuclear Energy 77 (2014), 266 –281. Gerczak TJ 2021 Irradiation Performance: High-Temperature Gas Reactor Fuels. In: Encyclopedia of Nuclear Energy, vol. 2, pp. 407 –419. GIF, 2018. R&D Outlook for Generation IV Nuclear Energy Systems: 2018. Update, available at. https://www.gen-4.org/gif/jcms/c_108744/gif-r-d-outlook-for-generation-iv-nuclear-

2021_MF_VHTR.pdf_29

2021_MF_VHTR.pdf

energy-systems-2018-update . Gougar, H., 2011. The Very High Temperature Reactor, Nuclear Energy Encyclopedia: Science, Technology, and Applications. Steven Krivit, Editor-i n-Chief. John Wiley and Sons, ISBN 978-0-470-89439-2. Gougar, H., et al., 2020. The US Department of Energy ’s high temperature reactor research and development program dProgress as of 2019. Nuclear Engineering and Design 358 (2020), 110397. Harth, R., Jansing, W., Teubner, H., 1990. Experience gained from the EVA II and KVK operation. Nuclear Engineering and Design 121 (1990), 173 –182. Helmreich, G., 2021. Fuel Design and Fabrication: TRISO Particle Fuel. In: Encyclopedia of Nuclear Energy, vol. 2, pp. 318 –325. IAEA, 2010. High Temperature Gas Cooled Reactor dFuels and Materials. IAEA TECDOC 1645, 2010. IAEA (2018) Advances in Small Modular Reactor Technology Developments d2018 Edition, https://aris.iaea.org/Publications/SMR-Book_2018.pdf . INL, 2012. Energy Development Opportunities for Wyoming. INL Report, INL/EXT-12-26732, November 2012.Kalilainen, J., et al., 2019. High Temperature Gas-cooled Reactors in a European Electricity Supply Environment; Main Outcomes of a Project in PSI. I n: Nuclear Science and Technology Symposium - SYP2019. Helsinki, Finland, 30–31 October 2019 .

2021_MF_VHTR.pdf_30

2021_MF_VHTR.pdf

Technology Symposium - SYP2019. Helsinki, Finland, 30–31 October 2019 . Konefal, J., Rackiewicz, D., 2008. Survey of HTGR Process Energy Applications. MPR Associates report MPR-3181, May 2008.Kugeler, K., Zhang, Z., 2019. Modular High-Temperature Gas-Cooled Reactor Power Plant, 1st edn. Springer, ISBN 978-3-662-57710-3.Kugeler, K., Nabielek, H., Buckthorpe, D., Scheuermann, W., Haneklaus, N., Fütterer, M.A., 2017. The High Temperature Gas-cooled Reactor: Safety c onsiderations of the (V)HTR- Modul. JRC Technical Report JRC107642, EUR 28712 EN, ISBN 978-92-79-71312-5. https://doi.org/10.2760/970340 . Kunitomi, K., 2013. Status of HTTR Project in JAEA. In: TWGGCR Meeting at IAEA. 5 March 2013 . Kunitomi, K., Katanishi, S., Takada, S., Takizuka, T., Yan, X., 2004. Japan ’s future HTR dThe GTHTR300. Nuclear Engineering and Design 233 (2004), 309 –327.

2021_MF_VHTR.pdf_31

2021_MF_VHTR.pdf

LaBar, M.P., 2002. The gas turbine-modular helium reator: A promising option for near-term deployment. In: General Atomics. GA-A23952, 2002.Lommers, L.J., Shahrokhi, F., Mayer III, J.A., Southworth, F.H., 2012. AREVA HTR concept for near-term deployment. Nuclear Engineering and Design 2 51 (2012), 292 –296. Matzner, D., 2004. In: Letcher, T.M. (Ed.), Chapter 14: The Pebble Bed Modular Reactor. Elsevier, Oxford, p. 2008. Neef, J., Weisbrodt, I., 1979. Coal gasi fication with heat from high temperature reactors: Objectives and status of the project “ Prototype Plant for Nuclear Process Heat (PNP) ”. Nuclear Engineering and Design 54 (1979), 157 –174. Pohl P (2008) IRPhE/AVR, High Temperature Reactor Experience, Archival Documentation , OECD/NEA, NEA-1739/02, https://www.oecd-nea.org/tools/abstract/detail/nea-1739 . Price, M.S.T., 2012. The Dragon project: Origins, achievements and legacies. Nuclear Engineering and Design 251 (2012), 60 –68. Ramana MV (2016) The checkered operational history of high temperature gas-cooled reactors, Bulletin of the Atomic Scientists, 72:3, 171 –179, https://doi.org/10.1080/ 00963402.2016.1170395 , 2016

2021_MF_VHTR.pdf_32

2021_MF_VHTR.pdf

00963402.2016.1170395 , 2016 Rempe, J.L., 2021. U.S. Nuclear Reactor Regulation of Two non-LWRs. In: Encyclopedia of Nuclear Energy, vol. 2, pp. 175 –187. Reutler, H., Lohnert, G.H., 1984. Advantages of going modular in HTRs. Nuclear Engineering and Design 78 (1984), 129 –136. Rosen, M., Naterer, G., Sadhankar, R., Suppiah, S., 2012. Nuclear-based hydrogen production with a thermochemical copper-chlorine cycle and super critical water reactor. International Journal of Energy Research 36 (4), 456 –465. March 2012. Schulten, R., 1976. Nukleare Prozeßwärme. Chemie Ingenieur Technik 48 (1976), 375 –380. Schulten R, Bellermann W, Braun H, Schmidt HW (1959) Der Hochtemperaturreaktor von BBC/Krupp (in German) , Die Atomwirtschaft. Shropshire, D., 2013. Integration challenges for nuclear cogeneration coupled to renewable energy systems. In: Proc. Joint NEA/IAEA Expert Worksh op on the Technical and Economic Assessment of Non-Electric Applications of Nuclear Energy (NUCOGEN). OECD Paris, France, 4–5 April 2013 . Siemens (1988) Hochtemperaturreaktor-Modul-Kraftwerksanlage, Kurzbeschreibung , Siemens/Interatom, Germany, November 1988

2021_MF_VHTR.pdf_33

2021_MF_VHTR.pdf

Siemens (1988) Hochtemperaturreaktor-Modul-Kraftwerksanlage, Kurzbeschreibung , Siemens/Interatom, Germany, November 1988 Styring, P., Jansen, D., de Coninck, H., Reith, H., Armstrong, K., 2011. Carbon Capture and Utilisation in the Green Economy, Using CO 2to Manufacture Fuel, Chemicals and Materials. The Centre for Low Carbon Futures. July 2011. Thomas, S., 2011. The pebble bed modular reactor: An obituary. Energy Policy 39 (5), 2431 –2440. Verfondern, K., 2007a. Survey on 20 years of R&D on nuclear process heat applications in germany. In: IAEA Proceedings IAEA-CN-152-16, Intl. Conf. on Non-Electric Applications of Nuclear Power: Seawater Desalination, Hydrogen Production and other Industrial Applications. Oarai, Japan, 16–19 April 2007 . Verfondern, K., 2007b. Potential for nuclear process heat application. In: IAEA Proceedings IAEA-CN-152-59, Intl. Conf. on Non-Electric Applicat ions of Nuclear Power: Seawater Desalination, Hydrogen Production and other Industrial Applications. Oarai, Japan, 16–19 April 2007 . Wu, Z., Lin, D., Zhong, D., 2002. The design features of the HTR-10. Nuclear Engineering and Design 218 (2002), 25 –32.

2021_MF_VHTR.pdf_34

2021_MF_VHTR.pdf

Wu, Z., Lin, D., Zhong, D., 2002. The design features of the HTR-10. Nuclear Engineering and Design 218 (2002), 25 –32. Yan, X.L., Hino, R., 2011. Nuclear Hydrogen Production Handbook. CRC Press, ISBN 978-1-4398-1083-5.Zhang, Z., et al., 2016. The Shandong Shidao Bay 200 MWe high-temperature gas-cooled reactor pebble-bed module (HTR-PM) demonstration power plant: An engineering and technological innovation. Engineering 2 (2016), 112 –118.522 The High Temperature Gas-Cooled Reactor

2022_GIF_VHTR.pdf_35

2022_GIF_VHTR.pdf

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.

2022_GIF_VHTR.pdf_36

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_37

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_38

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_39

2022_GIF_VHTR.pdf

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)

2022_GIF_VHTR.pdf_40

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_41

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_42

2022_GIF_VHTR.pdf

3.1.1. Fresh Fuel fabrication...............................................................................................................11 3.1.2. Fresh Fuel shipment.................................................................................................................12 3.1.3. Fresh Fuel receiving.................................................................................................................13 3.2. System elements related to reactor site of type B-VHTR.................................................................13 3.3. System elements related to reactor site of P-VHTR.........................................................................16 3.4. System elements related to reprocessing site or final disposal site of spent fuel for B-VHTR and P- VHTR 18 3.5. Diversion targets ..............................................................................................................................19 4. Proliferation Resistance Considerations Incorporated into Design..................................................22 4.1. Concealed diversion or production of material .................................................................................23 4.1.1. Diversion of unirradiated nuclear material items ......................................................................23 4.1.2. Diversion of irradiated nuclear material items ..........................................................................23 4.1.3. Undeclared production of nuclear material...............................................................................23 4.2. Breakout ...........................................................................................................................................24 4.2.1. Diversion of existing nuclear material.......................................................................................24 4.2.2. Production of the necessary weapons usable nuclear material ...............................................25 4.3. Pu Production in clandestine facilities ..............................................................................................25 5. Physical Protection Considerations Incorporated into Design .........................................................26 5.1. Theft of material for nuclear explosives............................................................................................26 5.2. Radiological sabotage ......................................................................................................................26 6. PR&PP Issues, Concerns and Benefits................................................................................................28 7. References ..............................................................................................................................................29 APPENDIX 1: VHTR Major Design Parameters ...........................................................................................31

2022_GIF_VHTR.pdf_43

2022_GIF_VHTR.pdf

6. PR&PP Issues, Concerns and Benefits................................................................................................28 7. References ..............................................................................................................................................29 APPENDIX 1: VHTR Major Design Parameters ...........................................................................................31 APPENDIX 2: Summary of PR relevant intrinsic design features.............................................................35Very-High-Temperature Reactor (VHTR) PR&PP White Paper ivList of Figures Figure 1: Illustration of Coated Particle Fuel in the Prismatic Fuel design ........................................................3 Figure 2: GT-MHR Reactor, Cross-Duct and PCU Vessels...................................................................................4 Figure 3: GT-MHR Fully-Embedded Reactor Building ........................................................................................4 Figure 4: Illustration of Coated Particle Fuel in Pebble Fuel Element ................................................................6 Figure 5: X-Energy Xe-100 .................................................................................................................................7 Figure 6: 250 MWt HTR-PM Reactor Building Elevated above Ground Level with Steam Generator; Spent Fuel Storage Not Shown ....................................................................................................................................7 Figure 7: B-VHTR and P-VHTR as well as their fuel elements...........................................................................10 Figure 8: B-VHTR System element ...................................................................................................................11 Figure 9: P-VHTR System element....................................................................................................................11 Figure 10: Fuel kernel fabrication through dropping uranyl nitrate stock solution ........................................12 Figure 11: Movement of fuel blocks in reactor site of B-VHTR .......................................................................14 Figure 12: Door valve and refueling machine ................................................................................................. 15 Figure 13: Neutron detectors and gamma ray detectors installed in the door valve for B-VHTR ...................16 Figure 14: Movement of the fuel pebbles........................................................................................................ 17 Figure 15: Plutonium build-up in a PBMR fuel element in an equilibrium core ..............................................20

2022_GIF_VHTR.pdf_44

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_45

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_46

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_47

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_48

2022_GIF_VHTR.pdf

(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

2022_GIF_VHTR.pdf_49

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_50

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_51

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_52

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_53

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_54

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_55

2022_GIF_VHTR.pdf

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.

2022_GIF_VHTR.pdf_56

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_57

2022_GIF_VHTR.pdf

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.

2022_GIF_VHTR.pdf_58

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_59

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_60

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_61

2022_GIF_VHTR.pdf

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.

2022_GIF_VHTR.pdf_62

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_63

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_64

2022_GIF_VHTR.pdf

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

2022_GIF_VHTR.pdf_65

2022_GIF_VHTR.pdf

them must be processed. Very-High-Temperature Reactor (VHTR) PR&PP White Paper 133.1.2. Fresh Fuel shipment Fuel rods for B-VHTR and fuel pebbles for P-VHTR are put into containers and shipped from fuel fabrication facilities to reactor sites. Adequate C/S system such as sealing and PP need to be applied to containers to ensure continuity of knowledge according the sensitivity grade of the nuclear material being shipped. Note that there are no current domestic or internationally licensed shipping container for transporting large quantities of HALEU fuels. 3.1.3. Fresh Fuel receiving Broken fresh fuel elements should be segregated and must be stored separately by the user for shipment back to the supplier for recycling as un-irradiated scrap. The C/S system for fresh fuel shipment must be confirmed upon fresh fuel receiving. The nuclear material in the broken fresh fuel elements is not attractive because the amounts are small and the material is still in the form of coated fuel particles. 3.2. System elements related to reactor site of type B-VHTR PR of B-VHTR is based on item accountancy. It is possible to imprint an ID on each fuel block, so the safeguards approach has many similarities with the safeguards of LWRs. All system elements related to a reactor site are confined within the reactor building as shown in Figure 11 [26]. All movements of fuel can be monitored by the surveillance cameras. Fuel storage racks of the fresh fuel storage and spent fuel storage areas are sealed after handling fuel therein. Fuel inventory in the reactor core is verified by measuring the fuel flow with detectors in the door valve. Movement of the fuel handling machine is slow due to its mass of more than

2022_GIF_VHTR.pdf_66

2022_GIF_VHTR.pdf

therein. Fuel inventory in the reactor core is verified by measuring the fuel flow with detectors in the door valve. Movement of the fuel handling machine is slow due to its mass of more than 100 tons. This movement can be followed by the surveillance cameras whose data should be continuously transferred to mitigate potential Cyber-attacks. 3.2.1. Fresh fuel storage on site Fuel blocks are assembled by inserting fuel rods into pre-formed holes in the graphite blocks in the reactor building. The on-site movement of fuel blocks of the B-VHTR is shown in Figure 11. The fuel blocks are stored in the fresh fuel storage rack until such time as the blocks scheduled for reloading are returned to the reactor core. An adequate C/S system such as surveillance cameras and PP should be applied to the fresh fuel storage area, the refueling machine, and the spent fuel storage area for continuity of knowledge. Very-High-Temperature Reactor (VHTR) PR&PP White Paper 14Figure 11: Movement of fuel blocks in reactor site of B-VHTR [26] 3.2.2. Refueling Machine for fresh fuel loading and spent fuel discharging This paragraph refers to HTTR as this is considered fully representative of B-VHTR [27]. StandpipeVery-High-Temperature Reactor (VHTR) PR&PP White Paper 15The fresh fuel blocks are taken into the refueling machine from the fresh fuel storage, and then the refueling machine is lifted and moved onto the door valve over the reactor with the crane. The fresh fuel blocks are loaded into the vertical empty space from where the spent fuels have been taken out. The IDs of fuel blocks are confirmed at time of loading of fresh

2022_GIF_VHTR.pdf_67

2022_GIF_VHTR.pdf

blocks are loaded into the vertical empty space from where the spent fuels have been taken out. The IDs of fuel blocks are confirmed at time of loading of fresh fuel. The spent fuel blocks in the reactor are taken into the revolver-rack of the refueling machine and moved to a spent fuel storage facility by the crane before the fresh fuels are loaded. The control rod driving device and the pair of control rods must be removed before refueling. Replaceable side reflectors and fuel blocks are handled using the refueling machine. They are passed through the door valve and the stand pipe at the upper part of the reactor core for any refueling. Fuel reloading in light water reactors (LWRs) is performed in water that provides a radiation shielding effect. However, the coolant of B-VHTR is helium and has no radiation shielding effect. For this reason, the fuel exchange for B-VHTR is performed by remote control of the gripper of the refueling machine, since the fuel cannot be directly viewed. It is also necessary to incorporate a radiation shielding function in the refueling machine because it will contain the spent fuel block in the revolver-rack. For this reason, its mass exceeds 100 tons. When the refueling machine is moved from the upper part of the reactor, the coolant (helium) in the reactor should not be allowed to leak. A door valve is provided between the refueling machine and the standpipe to prevent leakage of the coolant (helium) in the reactor to outside. The position of door valve is shown in Figure 12 [27]. Neutron detectors and gamma ray detectors are attached to the door

2022_GIF_VHTR.pdf_68

2022_GIF_VHTR.pdf

prevent leakage of the coolant (helium) in the reactor to outside. The position of door valve is shown in Figure 12 [27]. Neutron detectors and gamma ray detectors are attached to the door valve, since the door valve is necessary to move out core components (anything such as spent fuel blocks, replaceable side reflectors and irradiated experimental material from the reactor). 3.2.3. Reactor Core The core consists of hexagonal columns of fuel blocks, control rod guide blocks and surrounding replaceable side reflector, constituted of blocks. The permanent reflectors surround the replaceable side reflectors. Fuel blocks are stacked vertically in several stages, and replaceable reflectors are placed above and below them. In order to accommodate the decrease in reactivity associated with fuel depletion as the reactor is operated, by design the reactor core is loaded with adequate excess reactivity at the beginning of operation. Each fuel block is engraved with a unique ID and loaded to a predetermined position in the reactor core. After a certain period of operation, the spent fuel block is taken out through the stand pipe using the refueling machine. The coolant flows through the flow paths in the graphite blocks and is heated. The heated coolant is brought into a hot plenum and guided to outside of the reactor pressure vessel at a temperature of 700 to 950 °C. The control rods are suspended from the control rod drive mechanism in standpipes above the core and inserted into the core or reflector, as needed. Control rod guide columns for inserting control rods are provided in the core. Any undeclared movement of the refueling machine would be detected by surveillance cameras. Furthermore, irradiation of undeclared material is detectable with the neutron and

2022_GIF_VHTR.pdf_69

2022_GIF_VHTR.pdf

control rods are provided in the core. Any undeclared movement of the refueling machine would be detected by surveillance cameras. Furthermore, irradiation of undeclared material is detectable with the neutron and gamma ray detectors attached in the door valve used for introducing and removing materials into and from the core. The combination of neutron and gamma ray detectors, shown in Figure Figure 12: Door valve and refueling machine [27]Very-High-Temperature Reactor (VHTR) PR&PP White Paper 1613 [26] makes it possible to distinguish the nature of materials introduced into the core or removed from it as nuclear materials and non-nuclear materials. Data obtained by both detectors should be continuously transferred to safeguards inspectorates to avoid Cyber- attacks or other tampering. 3.2.4. Spent fuel storage on site The spent fuel blocks are stored for a certain period in racks of the spent fuel storage facility that includes a water-cooling system in order to remove decay heat. The movement of spent fuel blocks can be detected by an adequate C/S system such as sealing the lid on the top of the storage racks and monitoring them with further surveillance cameras. Figure 13: Neutron detectors and gamma ray detectors installed in the door valve for B-VHTR [26] 3.2.5. On-site radioactive waste storage Substances that do not contain nuclear fuel materials, such as activation products, are stored in the on-site radioactive waste storage facility, so their attractiveness from the PR viewpoint is low. However, such materials should be protected from a PP viewpoint. 3.2.6. On-site radioactive waste storage The spent fuel blocks in storage are put in fuel transfer casks for shipping to the final disposal

2022_GIF_VHTR.pdf_70

2022_GIF_VHTR.pdf

3.2.6. On-site radioactive waste storage The spent fuel blocks in storage are put in fuel transfer casks for shipping to the final disposal or to the reprocessing plant after cooling for a certain period in the spent fuel storage on site. Continuity of Knowledge (CoK) is maintained by use of adequate C/S systems, such as sealing transfer casks, and adequate PP is also applied, such as protection by guards. The spent fuel blocks are not attractive as sources of explosive nuclear materials used for NED due to the poor quality of the materials and the great difficulty of reprocessing. But they may be attractive from the view point of “radiological sabotage" due to their high radioactivity content. 3.3. System elements related to reactor site of P-VHTR For safeguards purposes, P-VHTR is regarded as a quasi-bulk type facility. 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. However, it is usually sufficient for safeguards to just count/keep track of the number of fresh fuel and spent fuel pebbles as they are moved from and to their respective storage systems. The operating temperatures and high pressure of the system would make it difficult to divert fuel out of the core. 3.3.1. Fresh fuel storage on site IAEAVery-High-Temperature Reactor (VHTR) PR&PP White Paper 17The containers with fuel pebbles are stored in the fresh fuel storage under an adequate C/S system and PP for P-VHTR. These fuel pebbles are moved to the charging room to be loaded

2022_GIF_VHTR.pdf_71

2022_GIF_VHTR.pdf

17The containers with fuel pebbles are stored in the fresh fuel storage under an adequate C/S system and PP for P-VHTR. These fuel pebbles are moved to the charging room to be loaded into the reactor core. The number of fuel pebbles should be counted if it is possible, and the movement of the fuel pebbles from the fresh fuel storage to the charging room should be observed via surveillance cameras. Diversion or otherwise acquisition of fuel pebbles is not attractive due to the difficulty of recovering the nuclear material from fuel elements and because the amount of nuclear material in them is small. 3.3.2. Recirculation of irradiated fuel pebbles The fuel pebbles have no identification numbers and are loaded randomly into the reactor core. The amount of nuclear material in every fresh fuel pebble is the same (heavy metal loading and uranium enrichment level). If initially fueled entirely with fresh fuel pebbles, P-VHTR cores would become critical with a small total volume of fuel. Therefore, graphite balls and boron balls containing no fuel are loaded into the core along with the fresh fuel in order to maintain the desired height of fuel in the core. With fuel depletion, graphite balls and boron balls are removed, and fresh fuel pebbles are loaded in, as the core evolves from the initial loading core to the equilibrium core. Figure 14 shows the movement of the fuel pebbles in the reactor [28]. Fuel pebbles are taken out from the core through the fuel pebble discharging tube. Failed fuel pebbles are separated

2022_GIF_VHTR.pdf_72

2022_GIF_VHTR.pdf

the fuel pebbles in the reactor [28]. Fuel pebbles are taken out from the core through the fuel pebble discharging tube. Failed fuel pebbles are separated and are stored in the scrap containers. Sound fuel pebbles are led to the dosing wheel where their fuel burnup levels are measured. The fuel burnup is evaluated by measuring the Cesium- 137 gamma ray peak with a gamma spectrometer. However, it has recently been suggested that Cs-137 would not necessarily be a good burnup indicator, and Zr-95, Nb- 95, and La-140 may provide more appropriate burnup instead [29]. Further research is needed. The fuel pebbles that have achieved a predetermined burnup level are discharged through the discharge tube and are led to containers in the discharge compartment as spent fuel pebbles. On the other hand, fuel pebbles that have not reached the predetermined burnup level are transported pneumatically to the upper part of the core and reloaded at the top of the core. This reloading is repeated until the fuel pebble reaches the predetermined burnup level. The number of reloading cycles is typically between 5 to 15. The precise figure depends on the specific design, reloading pattern and target burnup levels. High fuel burnup is achievable due to the highly stable characteristics of coated fuel particles and due to nearly continuous fuel loading. It is higher than the burnup of LWRs as well as B-VHTR. A burnup level of 100 GWd/T is achievable for the spent fuel of P-VHTR, and it results in superior proliferation resistance features due to large isotopic fraction of high content in plutonium that produces a high level

2022_GIF_VHTR.pdf_73

2022_GIF_VHTR.pdf

is achievable for the spent fuel of P-VHTR, and it results in superior proliferation resistance features due to large isotopic fraction of high content in plutonium that produces a high level of decay heat. The physical inventory verification in the reactor core is performed by controlling the number of fresh fuel pebbles loaded and accounting for the spent fuel pebbles discharged Figure 14: Movement of the fuel pebbles for P-VHTR [28]Very-High-Temperature Reactor (VHTR) PR&PP White Paper 18and the number of failed fuel pebbles discharged to the scrap container. Access to the reactor cell will be controlled by an adequate C/S system and PP. 3.3.3. Spent fuel storage on site The spent fuel pebbles in containers are stored for a certain period in the on-site spent fuel storage. The containers are cooled in order to remove decay heat. The movement of a container can be observed using an adequate C/S system, such as sealing the containers and monitoring the storage area with surveillance cameras. The amount of fissile nuclear material (U-235 and Pu-239) in the spent fuel pebbles is small due to high burnup and high content of decay heat-generating Pu isotopes. One of interesting discussions is the treatment of damaged pebbles. In general, the damaged pebbles are added to the spent fuel storage, i.e. there is no separate waste storage of broken pebbles planned for the PBMR design. Damaged pebbles are always to be expected to occur during irradiation in the reactor and cannot be returned for further cycles through the core, so they had to be classified as spent fuel. However,

2022_GIF_VHTR.pdf_74

2022_GIF_VHTR.pdf

pebbles are always to be expected to occur during irradiation in the reactor and cannot be returned for further cycles through the core, so they had to be classified as spent fuel. However, since those pebbles are less burnt, they are potentially more attractive in terms of Pu quality. 3.3.4. Radioactive waste storage on site Substances that do not contain nuclear fuel materials, such as activation products, are stored here, so their attractiveness from the PR viewpoint is low. However, these waste materials still need to be protected from a PP viewpoint. 3.3.5. Spent fuel shipping The spent fuel pebbles in containers will be transferred to the final disposal or to the reprocessing plant after cooling for a certain period in the spent fuel storage area on site. COK is ensured using an adequate C/S system such as sealing the containers and monitoring the movement of the containers with surveillance cameras. The spent fuel pebbles are not attractive from the point of view of nuclear materials for use for NEDs, but they may be attractive from the view point of “radiological sabotage" due to their high radioactivity content. See the section 5.2 for more discussion. 3.4. System elements related to reprocessing site or final disposal site of spent fuel for B-VHTR and P-VHTR The treatment of spent fuel of both B-VHTR and P-VHTR can be divided into (1) direct final disposal and (2) reprocessing. The direct disposal option is attractive because the coatings of coated fuel particles themselves are “containers” for the fission products and the fuel itself possesses high mechanical and chemical stability. Thus, the direct final disposal of the VHTR

2022_GIF_VHTR.pdf_75

2022_GIF_VHTR.pdf

coated fuel particles themselves are “containers” for the fission products and the fuel itself possesses high mechanical and chemical stability. Thus, the direct final disposal of the VHTR fuel has reduced environmental and public impact. Furthermore, the reprocessing of VHTR fuel is not considered attractive. The reason is that metric tons and tens of cubic meter quantities of carbon encasing coated fuel particles would have to be removed using either grind-leach, burn-leach of electrolysis in nitric acid if reprocessing were to be performed. However, these technologies have still not been demonstrated at industrial level. For this reason, spent fuel of VHTR has low attractiveness for diversion / acquisition and / or processing as nuclear material. Spent fuels from the VHTR may potentially still be attractive for radiological sabotage due to their high content in radioactive materials that results from their high fuel burnup levels. The physical robustness of VHTR fuel is favorable in this respect, making it more difficult for a potential adversary to achieve widespread dispersal. The proliferation resistance features corresponding to the reprocessing of the spent fuel of VHTR mentioned-above are valid not only for spent fuels of LEU-fuel, but also for that of LEU / Th, LEU/Pu (MOX), Pu / Th MOX with high burnup. Very-High-Temperature Reactor (VHTR) PR&PP White Paper 193.5. Diversion targets The key proliferation resistance feature of the VHTR is the fuel itself. The extraction of a significant quantity (SQ) of either indirect-use U-235 from LEU (75 kg) or direct-use U-233 and plutonium (both 8kg) from VHTR fuel will require the processing of metric tons and tens of

2022_GIF_VHTR.pdf_76

2022_GIF_VHTR.pdf

plutonium (both 8kg) from VHTR fuel will require the processing of metric tons and tens of cubic meter quantities of carbon encasing coated particles using either grind-leach, burn-leach, or electrolysis in nitric acid. A background report [14] that supported the compilation of the original VHTR white paper (published in 2011) discussed diversion targets for the two fuel forms, prismatic block and pebble. The following discussion is quoted from the background report [14] with some modifications using the PBMR [16] and the GT-MHR [6-8] as example plants for the P-VHTR and the B-VHTR respectively. “Using the PBMR as an example, the diversion of an indirect-use significant quantity (75 kilograms) of U-235 in LEU in fresh pebbles would require, for the equilibrium core with a pebble loading of 9 grams of LEU at 9.6% enrichment, 75,000/(9 * 0.096) = 86,806 pebbles or ~17.4 MT of fuel pebbles, which should be quite readily detectable even over time since that is ~20 percent of a core loading.” “By comparison, for the prismatic core GT-MHR or MHTGR using fuel elements with inscribed serial numbers for visual tracking, the diversion of an indirect-use significant quantity (75 kilograms) of U-235 in LEU in fuel elements containing ~3.43 kilograms of LEU on average at 19.8% enriched would require 75/(3.43 * 0.198) = ~111 fuel elements or 13.5 MT of fuel elements, which would be ~15–16% of a GT-MHR core loading

2022_GIF_VHTR.pdf_77

2022_GIF_VHTR.pdf

fuel elements or 13.5 MT of fuel elements, which would be ~15–16% of a GT-MHR core loading or ~17% of the MHTGR core loading.” “Thus, the mass ratio for the diversion of indirect-use U-235 in LEU between fresh pebbles and fresh GT-MHR fuel elements is 17.4/13.5 = ~1.29 so that 29% more pebbles by mass would have to be diverted to obtain 75 kilograms of indirect- use U-235 in LEU.” “Because the fuel elements of PBMRs are quite difficult to track, the use of LEU-fueled PBMRs has been examined by several researchers from the aspect of the attractiveness for diversion of fully burned spent fuel, one-cycle-irradiated pebbles, and the use of special production pebble. “The calculation results for the plutonium isotopic fractions in the PBMR fully burned spent fuel would likely be very close to those for the prismatic VHTR spent fuel where the prismatic fuel is to be discharged at a burn-up exceeding 100 GWD/MT (or MWD/kg). The PBMR and prismatic VHTR spent fuel will have slightly different plutonium isotopic compositions resulting from differences in the thermal-neutron and epithermal-neutron energy spectra due to a different moderator-to-fissile atom ratio and additional thermal and epithermal neutron self- shielding due to the higher-density fuel compacting used in the prismatic fuel.” …..” It is expected, however, that the spent LEU fuel from both the GA GT-MHR and Areva Modular HTR will have plutonium isotopic fractions very close to the values calculated for the PBM in

2022_GIF_VHTR.pdf_78

2022_GIF_VHTR.pdf

expected, however, that the spent LEU fuel from both the GA GT-MHR and Areva Modular HTR will have plutonium isotopic fractions very close to the values calculated for the PBM in Table 3.6.1.” It appears that the Pu will be of reactor grade in all cases by applying the fissile material type metric of PRPP WG.Very-High-Temperature Reactor (VHTR) PR&PP White Paper 20Table 1: Calculated plutonium isotopic fractions for PBMR spent fuel as a function of initial enrichment and discharge burn-up [Table 4.1 from [14]] “Because the PBMR recirculates a pebble up to six times through the core before it is discharged to spent fuel storage at full burn-up (~92 GWD/MT), the question arises about the diversion of an irradiated pebble after one cycle or the use of special pebbles designed as target elements to produce plutonium.” “The analysis of the PBMR by PBMR (Pty) Ltd. [16] shows in Figure 15 [14] the plutonium build-up per pebble and the relative isotopic content as a function of recirculation.” Figure 15: Plutonium build-up in a PBMR fuel element in an equilibrium core [14] “Figure 15 indicates that at full burn-up each pebble will contain about 0.11 grams of plutonium with the isotopics indicated, and, from this, it can be inferred that, at full burn-up (120 GWD/MT in the GT-MHR), the prismatic fuel elements can be estimated to contain on the order of 60– 70 grams of plutonium of similarly degraded isotopics.” The diversion of 1 SQ of direct-use Pu

2022_GIF_VHTR.pdf_79

2022_GIF_VHTR.pdf

in the GT-MHR), the prismatic fuel elements can be estimated to contain on the order of 60– 70 grams of plutonium of similarly degraded isotopics.” The diversion of 1 SQ of direct-use Pu from pebbles at full burn-up requires 8,000/0.11 = 72,727 pebbles or ~14.4 MT of fuel pebbles. It takes 8,000/65 = 123 prismatic fuel elements to secure 1 SQ of direct-use Pu. “However, the LEU pebble in a PBMR is recirculated up to six times while the fuel element in a GT-MHR or MHTGR is typically reloaded only once. From Figure 15, the plutonium content of a pebble after its initial irradiation is given as ~0.047 grams (~74% Pu-239), whereas for the GT-MHR there are no data quoted for the one-cycle-burned prism, but it is inferred that the plutonium loading would be ~50 grams with less favorable isotopics than in the pebble after Very-High-Temperature Reactor (VHTR) PR&PP White Paper 21one cycle of irradiation. From this, a rough comparison can be made that it would take at least ~1050 pebbles diverted after the first cycle to equal the amount of less favorable plutonium in a prismatic fuel element removed from a GT-MHR after the first irradiation.” Table 3.6-2 shows a summary table indicating the amount of material needed to collect an SQ. Table 2: SummaryDiversion Target U-235 from Fresh LEU Pu from Spent fuel SQ 75 kg 8kg

2022_GIF_VHTR.pdf_80

2022_GIF_VHTR.pdf

a summary table indicating the amount of material needed to collect an SQ. Table 2: SummaryDiversion Target U-235 from Fresh LEU Pu from Spent fuel SQ 75 kg 8kg Equivalent pebbles 86806 (17.4 MT) 72727 (14.4 MT) Equivalent blocks 111 (13.5 MT) 123 (15.0 MT)Very-High-Temperature Reactor (VHTR) PR&PP White Paper 224. Proliferation Resistance Considerations Incorporated into Design The fuel in reactor cores of B-VHTR and P-VHTR is not as accessible and visible as the fuel in an LWR. Therefore, physical inventory verification of nuclear materials in the reactor cores is carried out by measurement of fuel flows into and from the core. Major Material Balance Areas and Key Measurement Points are shown in Figure 16 as an example. Also, adequate C/S is necessary. Adequate counter-measures against cyberattacks are required to maintain CoK by C/S. Figure 16: Material Balance Areas and Key Measurement Points of B-VHTR and P-VHTR Design Information Verification (DIV) and C/S are implemented to avoid concealment of fuel. Direct transfer of the C/S signal to IAEA is recommended to enhance proliferation resistance. As noted previously, the key proliferation resistance feature of the VHTR is the fuel itself. To obtain a significant quantity of either indirect-use U-235 from LEU or direct-use plutonium, one must process metric tons and tens of cubic meter quantities of carbon encasing fuel using either grind-leach or burn-leach of electrolysis in nitric acid. The high burnup of the spent fuel of the VHTRs is also a key proliferation resistance feature

2022_GIF_VHTR.pdf_81

2022_GIF_VHTR.pdf

either grind-leach or burn-leach of electrolysis in nitric acid. The high burnup of the spent fuel of the VHTRs is also a key proliferation resistance feature due to the high isotopic fraction of even plutonium isotopes generating large amounts of decay heat and high dose rate. However, it is controversial. Historically, it has been argued that the technical difficulty of fabricating nuclear weapons depends on the isotopic composition of plutonium, in particular the amount of Pu-240. Although there are several references, the one that summarizes the key points is by Pellaud [30]. For nuclear safeguards verification activities there is no distinction for Pu with less than 80% Pu-238. However, the heat generated by Pu isotopic containing more than a few percent of Pu-238 would substantially increase the technical difficulty related to the fabrication phase (weaponization). Using a set of figures of merit (FOM) for attractiveness, Bathke, et al. [31] estimated that about 8% Pu-238 is required to render the plutonium isotopic unattractive for an unadvanced proliferant state that requires reliably high-yield nuclear devices, however it remains attractive for both technologically advanced states, which can handle it, and subnational groups for which high reliability might not be a requirement. However, these arguments are founded on the assumption that the proliferants demand reliable yield. In the case of unadvanced proliferant or non-state actors who do not pay attention to the yield, high reliability might not be their requirement. With those reasons considered, the current GIF PRPP WG methodology adopts weapon Very-High-Temperature Reactor (VHTR) PR&PP White Paper

2022_GIF_VHTR.pdf_82

2022_GIF_VHTR.pdf

reliability might not be their requirement. With those reasons considered, the current GIF PRPP WG methodology adopts weapon Very-High-Temperature Reactor (VHTR) PR&PP White Paper 23grade, reactor grade, and deep-burn grade for Pu categorization [1]. The fact that Pu in HTGRs’ spent fuel can achieve deep-burn is one of the notable features. 4.1. Concealed diversion or production of material Diversion of large quantities of nuclear materials (U-235, plutonium or U-233) is detectable by spent fuel accountancy based on radiation monitoring or fuel element counting, by C/S on fuel storage, or by recorded reactivity deviations in reactor operations. The VHTR does not produce readily accessible, attractive fissile material. The technologies for reprocessing coated fuel particles are complicated and still require development. 4.1.1. Diversion of unirradiated nuclear material items Once the fuel has been encased within fuel kernel of coated fuel particle and furthermore into fuel elements (such as fuel compacts for B-VHTR or fuel pebbles for P-VHTR), diversion becomes difficult. The latter (fuel elements) consist of coated fuel particles encased within graphitized carbon. Note that fresh fuel fabrication should be performed under surveillance. Once in fuel assembly (compact ball) form, the nuclear material is more difficult to retrieve due to difficulty of separation of nuclear material from large amounts of graphite and of the strength of the coatings of particles. Fabricated fresh fuel can be stored under C/S measures for B- VHTR and P-VHTR. The theft during transportation of fresh fuel can be detected by the C/S.

2022_GIF_VHTR.pdf_83

2022_GIF_VHTR.pdf

of the coatings of particles. Fabricated fresh fuel can be stored under C/S measures for B- VHTR and P-VHTR. The theft during transportation of fresh fuel can be detected by the C/S. The raw constituents are observed under the same C/S applied for fuel fabrication of LWR. 4.1.2. Diversion of irradiated nuclear material items 4.1.2.1. B-VHTR The major irradiated nuclear material items are spent fuel blocks. Diversion of Pu is possible by discharging fuel blocks after a short time of reactor operation and then reprocessing them. The fuel blocks are unloaded through standpipes over the reactor pressure vessel. For such an operation, both the refueling machine and the door valve are required because it is necessary to maintain isolation between the reactor coolant and the outside atmosphere for the B-VHTR. Unexplained or illicit movements of the refueling machine and the door valve by the crane can be detected by surveillance cameras. Also, undeclared movements of prematurely discharged fuel blocks are detected by the neutron detector and the gamma ray detector in the door valve. Any discharged material from the reactor pressure vessel can be identified as nuclear material or not. Nuclear material is indicated when signals of both neutron and gamma ray are detected. If the material is non-nuclear, such as a surveillance sample, then no neutron source is detected. Undeclared discharging of experimental nuclear materials is detected in the same manner as fuel blocks. 4.1.2.2. P-VHTR Diversion of Pu may be possible using the continuous fuel loading feature through early discharging of fuel pebbles from the reactor core before even-mass-number Pu isotopes are

2022_GIF_VHTR.pdf_84

2022_GIF_VHTR.pdf

Diversion of Pu may be possible using the continuous fuel loading feature through early discharging of fuel pebbles from the reactor core before even-mass-number Pu isotopes are accumulated. However, this would be detected by the burnup measuring detectors. Furthermore, it is technically difficult because the reprocessing process of VHTR fuel is still not established and detection of this diversion route is possible if an appropriate C/S system is in place. 4.1.3. Undeclared production of nuclear material 4.1.3.1. B-VHTR Very-High-Temperature Reactor (VHTR) PR&PP White Paper 24Undeclared production of nuclear material may be possible through the irradiation of fertile nuclear material in irradiation holes in the core or replaceable side reflectors of B-VHTR. The materials would be loaded and unloaded through standpipes over the reactor pressure vessel. In the B-VHTR, it is not possible to directly access and to visually observe the fuel in the core or the replaceable side reflectors as would be possible in LWRs where water above the core is used as radiation shielding. For these reasons, a handling machine with radiation integrated shielding function, such as the refueling machine, is necessary for any undeclared production of nuclear material. Any unexplained or illicit movement of handling machines can be detected by surveillance cameras in the reactor building. Moreover, ton quantities of fertile material would need to be loaded illicitly to generate a SQ, and it is difficult to envisage this being practical to achieve without detection. It should be noted that B-VHTR could be used in a mode similar to that of the Magnox reactors

2022_GIF_VHTR.pdf_85

2022_GIF_VHTR.pdf

practical to achieve without detection. It should be noted that B-VHTR could be used in a mode similar to that of the Magnox reactors for producing weapon-grade plutonium. In this case, rod-type Magnox fuel containing metal uranium would be inserted into some cooling holes of the graphite blocks instead of using ordinary B-VHTR fuel rods based on coated particle fuel. In this way the difficulty of reprocessing of VHTR fuel would be avoided, as the reprocessing methods for Magnox fuel are well established. However, the reactors would have to be operated with low reactor coolant outlet temperatures to protect the integrity of the Magnox fuel and ton quantities of Magnox- type fuel would need to be irradiated. This would imply giving up efficient power production, which would be detectable. It might to worth thinking that this mode of operation could be dangerous because of accumulation of Wigner energy in the graphite blocks, but further study is needed. 4.1.3.2. P-VHTR The inlet pipes of fresh fuel pebbles, in the fuel charging room, can be used for loading target pebbles and for the access to the core region of a P-VHTR. However, these pipes cannot be easily used for loading illicit material for the undeclared production of nuclear materials due to the length of, and many curves in, the fuel loading path. Pebbles with diameter of 6 cm could be loaded into the pipe. It is very important to confirm in Design Information Verification that there are no access holes into the pipes except at the fresh fuel pebble loading location, precluding any other pipe access into the reactor core. Irradiation of fertile materials covered

2022_GIF_VHTR.pdf_86

2022_GIF_VHTR.pdf

there are no access holes into the pipes except at the fresh fuel pebble loading location, precluding any other pipe access into the reactor core. Irradiation of fertile materials covered with graphite or carbon that look like fuel pebbles is possible. But such pseudo-fuel spheres may break during movement through the core and would be difficult to remove. Furthermore, such pseudo-fuel without ceramic coatings would release unexpected high radioactivity into the primary cooling system at high temperature operation, which would be detectable. In addition, there would result many operational problems, which would also be detectable and require explanation. Tton quantities of heavy metal would need to be irradiated in the core to generate 1 SQ. Finally, it is important to recognize that the presence of target breeding pebbles in the core will alter the balance between fresh fuel demand and energy production in a way that is detectable long before a significant quantity of fissile material is accumulated [32-35]. In addition, the identification of such breeding pebbles by the gamma measurement is much more difficult than for regular fuel. 4.2. Breakout As mentioned in section 3.4, reprocessing has not yet been demonstrated for the coated fuel particles at industrial scale. In the presence of multi-lateral contractual provisions, for example adhering to the guidance of the international Nuclear Suppliers Group (NSG), for the supply of fresh fuels and the take-back of spent fuels for an exported VHTR, the issue of breakout is further mitigated since there will be either no such material, or limited quantities of material, to be reprocessed in the host states. 4.2.1. Diversion of existing nuclear material Very-High-Temperature Reactor (VHTR) PR&PP White Paper

2022_GIF_VHTR.pdf_87

2022_GIF_VHTR.pdf

be reprocessed in the host states. 4.2.1. Diversion of existing nuclear material Very-High-Temperature Reactor (VHTR) PR&PP White Paper 25As mentioned in Section 3.4, the key proliferation resistance feature is the use of coated fuel particles embedded within a graphite matrix. Therefore, diverting existing nuclear material from VHTR fuels is difficult, lengthy and costly, regardless of the implementation of safeguards and PP for the fuels. Since the reprocessing technology is not developed to industrial level, extraction of nuclear material is significantly difficult. Moreover, the high burnup of spent fuel from VHTRs is also a key proliferation resistance feature due to the presence of plutonium isotopes that produce large amounts of decay heat. Pu in high burnup spent fuel contains considerable even-numbered Pu isotopes, i.e. Pu-238, 240 and 242, whose decay heat negatively affects use as a NED. Note that the diversion of raw material before being coated with carbon and silicon carbide would be the easiest pathway for the processing of nuclear materials to be used in the fabrication of nuclear explosive devices. However, this is not a VHTR-specific problem, but a concern for all types of nuclear reactor systems. 4.2.2. Production of the necessary weapons usable nuclear material As mentioned in section 3.4, the key proliferation resistance feature of VHTRs is the use of coated fuel particles embedded within graphite matrix. It is necessary to process metric tons and tens of cubic meter quantities of carbon encasing the fuel kernels to obtain the amount of nuclear material necessary for production of weapons. 4.3. Pu Production in clandestine facilities

2022_GIF_VHTR.pdf_88

2022_GIF_VHTR.pdf

and tens of cubic meter quantities of carbon encasing the fuel kernels to obtain the amount of nuclear material necessary for production of weapons. 4.3. Pu Production in clandestine facilities High quality graphite with very low impurity levels is used in the technology of the B-VHTR and P-VHTR. This high quality graphite can be used for gas-cooled reactors in which weapons- grade plutonium can be produced from natural uranium. Therefore, the consumption of large amounts of nuclear-grade graphite should be controlled. For that reason, nuclear grade graphite is controlled according to NSG lists part 1. Operation of the clandestine facilities (reactor and fuel reprocessing) could be detected by environmental sampling under the international safeguards regimes.Very-High-Temperature Reactor (VHTR) PR&PP White Paper 265. Physical Protection Considerations Incorporated into Design This section provides a qualitative overview discussion of the aspects of VHTR systems and their design that create potential benefits or problems from the point of view of potential threat by sub-national actors. 5.1. Theft of material for nuclear explosives Plutonium in the spent fuel of LEU cycles and U-233 in that of future LEU/Th cycles are attractive for the NED production. However, these nuclear materials in spent fuels are accompanied with fission products, which are highly radioactive and make it difficult for terrorists to steal the materials. Moreover, the nuclear materials are encased inside the coated fuel particles. In these coated particles, the material of interest would be quite dilute so that the theft of a significant quantity would require the theft of metric tons of contaminated graphite and/or graphitized carbon containing the coated fuel particles. Obtaining access to a significant

2022_GIF_VHTR.pdf_89

2022_GIF_VHTR.pdf

theft of a significant quantity would require the theft of metric tons of contaminated graphite and/or graphitized carbon containing the coated fuel particles. Obtaining access to a significant quantity of plutonium or U-233 in the stolen spent fuels would require substantial effort for reprocessing. Furthermore, plutonium with a high inventory of the plutonium isotopes other than Pu-239 is not attractive for the manufacturing of NEDs (e.g. high decay heat). U-233 with hundreds of parts per millions of U-232 is not attractive due to high radioactivity and to the necessity of further chemical cleaning to remove radioactive decay products. For those reasons, the intrinsic qualities of VHTR spent fuel make it undesirable as a target for theft by a sub-national group for use as nuclear explosive. 5.2. Radiological sabotage VHTRs are designed such that the fuel temperature is maintained below fuel-damaging temperatures under all conditions of normal operations and accident situations, including beyond-design-basis events. The design vision is that, even if the safety-related reactor cavity cooling system were to malfunction, decay heat in the core would still be removed to the external wall of the reactor vessel. As a result, the fuel temperatures in the core do not exceed the levels that would cause the loss of the primary containment provided by the SiC coatings over the fuel kernels. The ultimate radiological sabotage act for reactors is that of an insider or an intruder trying to cause radiological exposure by inducing a large power excursion. For both P-VHTR and B- VHTR designs, appropriate physical protection and controls must be in place to prevent such acts. These designs have several safety benefits from the very high temperature tolerance of

2022_GIF_VHTR.pdf_90

2022_GIF_VHTR.pdf

VHTR designs, appropriate physical protection and controls must be in place to prevent such acts. These designs have several safety benefits from the very high temperature tolerance of the fuel and the strong negative temperature power coefficient. Another relevant discussion is that both VHTRs are extremely resilient to this kind of terrorist attacks because passive heat removal, or reactor cavity cooling system (RCCS), by air cooling, water or a combination of both is available when a loss of coolant happens. The high burnup levels in the spent fuel of both VHTR types is one of the key proliferation resistance features due to high radioactivity. However, spent fuels of VHTRs may be attractive for Radiological Dispersion Device (RDD) due to high radioactivity resulting from the high burnup. Below is the discussion of RDD for both P-VHTR and B-VHTR: In the case of P-VHTR, the quasi-bulk fuel form may be attractive for terrorists when considering the possibility of dispersal of the spent fuel. Protection of spent fuel on the reactor site will be important. This should also be considered in PP when transporting spent fuel by land. In the case of B-VHTR, the item-type fuel allows its PP to be similar to that of LWRs. Moreover, TRISO is considered to be very resistant to scattering and therefore more robust against RDD-type terrorism than LWRs.Very-High-Temperature Reactor (VHTR) PR&PP White Paper 27Finally, some points to be considered for the PP of VHTR are listed referring to the previous VHTR white paper: Quality controls at the fuel fabrication plant in the supplier nation.

2022_GIF_VHTR.pdf_91

2022_GIF_VHTR.pdf

27Finally, some points to be considered for the PP of VHTR are listed referring to the previous VHTR white paper: Quality controls at the fuel fabrication plant in the supplier nation. Proper maintenance, inspection, and protection of (1) the helium supply and the helium supply station to prevent the introduction of corrosive chemicals, (2) the primary coolant contaminant monitoring equipment to detect the introduction of such chemicals, and (3) the helium purification system to remove contaminants. Careful maintenance, inspections, testing and protection of reactivity control systems to assure the capability to achieve safe hot and cold shutdown and, if required, accomplish the same function from a secure remote location. Physical protection is required of and controlled access to fresh and spent fuel storage locations, the inbound and outbound transportation loading systems, the transportation of the fresh fuel from the fuel fabrication facility, and the spent fuel to its processing or disposal facilities.Very-High-Temperature Reactor (VHTR) PR&PP White Paper 286. PR&PP Issues, Concerns and Benefits The key areas of known strengths of the VHTR concept at this time are its robust fuel form, with fissile material strongly diluted in carbonaceous material, high burnup and the use of the once-through LEU fuel cycle, which all make VHTR fuel unattractive for proliferation purposes. When considering PR, B-VHTR will have item-based safeguards applied, while P-VHTR safeguards are quasi-bulk, so differing safeguards approaches will be required is relatively difficult. Regarding PP, typical reactor site protections on the reactor, control systems, and fresh and

2022_GIF_VHTR.pdf_92

2022_GIF_VHTR.pdf

safeguards are quasi-bulk, so differing safeguards approaches will be required is relatively difficult. Regarding PP, typical reactor site protections on the reactor, control systems, and fresh and spent fuel storage will be required. It can be concluded that VHTRs are extremely resilient to terrorist attacks because RCCS is available when a loss of coolant happens. For system designers, program policy makers, and external stakeholders who read this white paper are encouraged to evaluate PR&PP features using the GIF PRPP WG methodology from an early stage of design, and keep updating them as designs progress.Very-High-Temperature Reactor (VHTR) PR&PP White Paper 297. References [1] Evaluation Methodology for Proliferation Resistance and Physical Protection of Generation IV Nuclear Energy Systems, GIF/PRPPWG/2006/005, Revision 6, prepared by the Proliferation Resistance and Physical Protection Evaluation Methodology Expert Group of the Generation IV International Forum (GIF), September 15, 2011. [2] PRPP Working Group and System Steering Committees of the Generation IV International Forum. Proliferation Resistance and Physical Protection of the Six Generation IV Energy Systems. Technical Report GIF/PRPPWG/2011/002, Generation IV International Forum (GIF), 2011. [3] IAEA, Proliferation Resistance Fundamentals for Future Nuclear Energy Systems, IAEA STR- 332, IAEA Department of Safeguards, IAEA, Vienna (2002). [4] Gen IV International Forum, GIF R&D Outlook for Generation IV Nuclear Energy Systems: 2018 Update (2019). [5] M. A. Fütterer, et al., "The High Temperature Gas-Cooled Reactor," Encyclopedia of Nuclear

2022_GIF_VHTR.pdf_93

2022_GIF_VHTR.pdf

Update (2019). [5] M. A. Fütterer, et al., "The High Temperature Gas-Cooled Reactor," Encyclopedia of Nuclear Energy, Elsevier, pp. 512-522, 2021, https://doi.org/10.1016/B978-0-12-409548-9.12205-5 [6] M.B. Richards et al., Part 1 -- H2-MHR Pre-Conceptual Design Report: SI-Based Plant, GA-A25401, General Atomics, Idaho National Laboratory, and Texas A&M University, April 2006. [7] M.B. Richards et al., Part 2 -- H2-MHR Pre-Conceptual Design Report: HTE-Based Plant, GA-A25402, General Atomics, Idaho National Laboratory, and Texas A&M University, April 2006. [8] General Atomics, Gas-Turbine Modular Helium Reactor (GT-MHR) Conceptual Design Description Report, GA Document No. 910720, Revision 1, July 1996, transmitted by letter from Laurence L. Parme (GA) to Raji Tripathi (USNRC), "GT-MHR Conceptual Design Description Report," GA/NRC-337-02, General Atomics, San Diego, CA, August 6, 2002. [9] L. Lommers et al., “AREVA HTR Concept for Near-Term Deployment,” Nuclear Engineering and Design, 251, pp. 292-296, October 2012. https://doi.org/10.1016/j.nucengdes.2011.10.030. [10] Brochure: ANTARES - The AREVA HTR-VHTR Design,

2022_GIF_VHTR.pdf_94

2022_GIF_VHTR.pdf

[10] Brochure: ANTARES - The AREVA HTR-VHTR Design, https://www.yumpu.com/en/document/read/32557580/antares-the-areva-htr-vhtr-design-smr [11] V. Petrunin et al., "Analysis of questions concerning the nonproliferation of fissile materials for low-and medium-capacity nuclear power systems," Atomnaya Energiya 105, Issue 3, pp. 123- 127, September 2008 (in English. pp. 159-164, Atomic Energy 105, Springer, New York, ISSN 1063-4258 (Print), 1573-8205 (Online)). [12] K. Kunitomi, et al., "JAEA's VHTR for Hydrogen and Electricity Cogeneration: GTHTR300C," Nuclear Engineering and Technology 39, pp. 9-20, February 2007. [13] Chang Keun Jo, Hong Sik Lim, and Jae Man Noh, "Preconceptual Designs of the 200MWth Prism and Pebble-bed Type VHTR Cores," PHYSOR-2008, International Conference on the Physics of Reactors “Nuclear Power: A Sustainable Resource,” Casino-Kursaal Conference Center, Interlaken, Switzerland, September 14-19, 2008. [14] D. Moses, “Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP),” ORNL/TM-2010/163, Oak Ridge National Laboratory, August 2010. [15] Presentations by PBMR (Pty) Ltd. to the U.S. Nuclear Regulatory Commission Public Meeting,

2022_GIF_VHTR.pdf_95

2022_GIF_VHTR.pdf

[15] Presentations by PBMR (Pty) Ltd. to the U.S. Nuclear Regulatory Commission Public Meeting, PBMR Safety and Design Familiarization, February 28-March 3, 2006. [16] Johan Slabber, PBMR (Pty) Ltd., "PBMR Nuclear Material Safeguards," Paper No. B14, Proceedings of the Conference on High Temperature Reactors, Beijing, China, September, 22- 24, 2004, International Atomic Energy Agency, Vienna (Austria). [17] E. Mulder, MM. van tStaden et al, “The Coupled Neutronics and Thermo-Fluid Dynamics Design Characteristics of the Xe-100 200 MWth Reactor,” Proceedings of the Conference on High Temperature Reactors, Las Vegas, Nevada, USA, November, 7-10, 2018.Very-High-Temperature Reactor (VHTR) PR&PP White Paper 30[18] Dong Y (2012) China’s activities in HTGRs HTR-10 and HTR-PM. In: IAEA Course on High Temperature Gas Cooled Reactor Technology. Beijing, China, 22-26 October 2012. [19] Z. Zhang et al., “The Shandong Shidao Bay 200 MWe High-Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM) Demonstration Power Plant: An Engineering and Technological Innovation,” Engineering 2, pp. 112-118, March 2016. http://dx.doi.org/10.1016/J.ENG.2016.01.020

2022_GIF_VHTR.pdf_96

2022_GIF_VHTR.pdf

Innovation,” Engineering 2, pp. 112-118, March 2016. http://dx.doi.org/10.1016/J.ENG.2016.01.020 [20] Y. Xu and K. Zuo, "Overview of the 10 MW high temperature gas cooled reactor—test module project," Nuclear Engineering and Design 218, pp. 13–23, October 2002. [21] The image of XE-100 Reactor downloaded from: https://x-energy.com/media/xe-100 [22] M. A. Fütterer, F. von der Weid, P. Kilchmann, A High Voltage Head-End Process for Waste Minimization and Reprocessing of Coated Particle Fuel for High Temperature Reactors, Proc. ICAPP’10, paper 10219, San Diego, CA, USA, 13-17 June 2010. [23] Correct reference for SCWR WP (TBC) [24] Nuclear Fuel Industries, Ltd, “HTGR fuel manufacturing process 1. Fuel particle process,” https://www.nfi.co.jp/product/prod03.html (accessed 2020-7-16). [25] K, Juergen et al., “Upgrading (V)HTR fuel elements for generation IV goals by SiC encapsulation,” Kerntechnik, 77(5), 351-355, 2012 [26] Kiyonobu Yamashita, Fujio Miyamoto, Sigeaki Nakagawa, and Toshiyuki Tanaka, "Safeguards Concept for the High Temperature Engineering Test Reactor Using Unattended Fuel Flow Monitor System," Journal of Nuclear Materials Management, Volume XXV, Number 4, August 1997.

2022_GIF_VHTR.pdf_97

2022_GIF_VHTR.pdf

Concept for the High Temperature Engineering Test Reactor Using Unattended Fuel Flow Monitor System," Journal of Nuclear Materials Management, Volume XXV, Number 4, August 1997. [27] S. Saito, et al., “Design of High Temperature Engineering Test Reactor (HTTR),” JAERI 1332, 1994, https://doi.org/10.11484/jaeri-1332. [28] U. Cleve et al., “The Technology of High-Temperature-Reactors,” Proceedings of ICAPP 2011, Paper 11076, Nice, France, May 2-5, 2011. [29] T. Shiba, et al., “Proliferation Resistance and Safeguardability of Very High Temperature Reactor,” IAEA Symposium on International Safeguards, 5-8 November 2018, Vienna International Centre, Vienna (Austria). [30] B. Pellaud, “Proliferation Aspects of Plutonium Recycling,” Journal of Nuclear Material Management, XXXI, 1, pp. 30-38, 2002 [31] C. Bathke, et al., “The Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various Proliferation and Theft Scenarios,” Nuclear Technology, Vol 179, Issue 1, 2012 [32] A.M. Ougouag and H. D. Gougar, “Preliminary Assessment of the Ease of Detection of Attempts at Dual Use of a Pebble-Bed Reactor,” Transactions of the Winter 2001 Annual Meeting of ANS, Reno, NV, Trans. ANS 85, pp. 115-117, Nov. 2001.

2022_GIF_VHTR.pdf_98

2022_GIF_VHTR.pdf

Reno, NV, Trans. ANS 85, pp. 115-117, Nov. 2001. [33] Ougouag, A. M., S. M. Modro, W. K. Terry, and H. D. Gougar “Rational Basis for a Systematic Identification of Critical Components and Safeguard Measures for a Pebble-Bed Reactor” Transactions of the Winter 2002 Annual Meeting of ANS, Washington, DC, Trans. ANS 87, pp. 367-368, Nov. 2002. [34] A.M. Ougouag, W. K. Terry and H. D. Gougar, “Examination of the Potential for Diversion or5Clandestine Dual Use of a Pebble-Bed Reactor to Produce Plutonium,” Proceedings of HTR 2002, 1st International Topical Meeting on High temperature Reactor Technology (HTR), Petten, Netherlands, April 22-24, 2002. [35] A. M. Ougouag, H. D. Gougar, and T.A. Todd, “Evaluation of the Strategic Value of Fully Burnt PBMR Spent Fuel,” May 2006, Idaho National Laboratory, INL/EXT-06-11272Very-High-Temperature Reactor (VHTR) PR&PP White Paper 31APPENDIX 1: VHTR Major Design Parameters Appendix VHTR.A – VHTR Major Reactor Design Parameters Major Reactor ParametersFramatome SC-HTGRGeneral Atomics GT-MHRX-Energy Xe-100Huaneng Group & CNEC/INET HTR-PMJAEA

2022_GIF_VHTR.pdf_99

2022_GIF_VHTR.pdf

SC-HTGRGeneral Atomics GT-MHRX-Energy Xe-100Huaneng Group & CNEC/INET HTR-PMJAEA GTHTR300COKBM GT- MHRKAERI NHDD Thermal Power (MW-th) 625 600 200 250 600 600 200 Thermal Efficiency (%) in Electricity Generation~40 ~48 40 (inferred) 40 ~50 ~48 None, H 2 production Primary Coolant Helium Helium Helium Helium Helium Helium Helium Moderator High- Temperature GraphiteHigh- Temperature GraphiteHigh- Temperature Graphitized Carbon with Graphite ReflectorHigh- Temperature Graphitized Carbon with Graphite ReflectorHigh- Temperature GraphiteHigh- Temperature GraphiteHigh- Temperature Graphite or Graphitized Carbon with Reflector Power Density (MW/m3) ~6.3 (inferred) 6.3 4.95 (max) ~3.22 5.4 6.3 2.27-3.0 pebble, 5.68 prismatic Fuel Materials LEUO 2 TRISO- coated particlesUC 0.5O1.5 TRISO- coated particles; LEUC 0.5O1.5 (19.8%) fissile and UNatC0.5O1.5 fertileLEUO 2 TRISO- coated particlesLEUO 2 TRISO-