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{"id": "2021_MF_VHTR.pdf_0", "source": "2021_MF_VHTR.pdf", "chunk": "The High Temperature Gas-Cooled Reactor\nMichael A. Fu \u00a8tterera, Gerhard Strydomb, Hiroyuki Satoc,F uL id, Eric Abonneaue, Tim Abramf, Mike W. Daviesg, Minwhan Kimh,\nLyndon Edwardsi, Ondrej Muranskyi, Manuel A. Pouchonj, and Metin Yetisirk,aEuropean Commission, Joint Research Centre,\nPetten, The Netherlands;bIdaho National Laboratory, Idaho Falls, ID, United States;cJapan Atomic Energy Agency, Ibaraki, Japan;\ndINET, Tsinghua University, Beijing, China;eCommissariat \u00e0 l \u2019Energie Atomique et aux Energies Alternatives, Paris, France;\nfManchester University, Manchester, United Kingdom;gJacobs Clean Energy Limited, Knutsford, Cheshire, United Kingdom;hKorea\nAtomic Energy Research Institute, Daejeon, South Korea;iAustralian Nuclear Science and Technology Organisation, Lucas Heights,\nAustralia;jPaul Scherrer Institut, Villigen, Switzerland; andkCanadian Nuclear Laboratories, Chalk River, ON, Canada\n\u00a9 2021 Elsevier Inc. All rights reserved.\nWhat is a high temperature gas-cooled reactor? 513\nFrom groundbreaking technology . 513\n.to modern characteristics 513\nTRISO fuel: Key to performance and safety 514\nWhich HTR versions were developed? 514\nRecent results from test reactors 517\nThe case for new next generation HTRs 518\nOngoing HTR development 518\nBeyond electricity: Emission-free process heat and cogeneration 520\nOutlook 520\nReferences 522\nGlossary\nAGR Advanced Gas-cooled Reactor\nAVR Arbeitsgemeinschaft Versuchsreaktor\nBISO Bi-Structural Isotropic Fuel\nGCR Gas-Cooled Reactor", "characters": 1490, "tokens": 388}
{"id": "2021_MF_VHTR.pdf_1", "source": "2021_MF_VHTR.pdf", "chunk": "Outlook 520\nReferences 522\nGlossary\nAGR Advanced Gas-cooled Reactor\nAVR Arbeitsgemeinschaft Versuchsreaktor\nBISO Bi-Structural Isotropic Fuel\nGCR Gas-Cooled Reactor\nGIFGeneration IV International Forum\nGT-MHR Gas Turbine Modular Helium-cooled Reactor\nHTGR High Temperature Gas-Cooled Reactor\nHTR High Temperature Gas-Cooled Reactor\nHTR-10 10 MW High Temperature Reactor\nHTR-PM High Temperature Reactor \u2013Pebble-bed Module\nHTTR High Temperature engineering Test Reactor\nHWR Heavy Water Reactor\nINET Institute of Nuclear and New Energy Technology (Tsinghua University, Beijing)\nJAEA Japan Atomic Energy Agency\nLANL Los Alamos National Laboratory\nLWR Light Water Reactor\nMWe Megawatt electric\nMWth Megawatt thermal\nNGNP Next Generation Nuclear Plant\nORNL Oak Ridge National Laboratory\nPBMR Pebble Bed Modular Reactor\nR&D Research and Development\nS-ISulfur-Iodine thermochemical process for hydrogen production\nSFBR Sodium-cooled Fast Breeder Reactor\nTHTR Thorium High Temperature Reactor\nTRISO Tri-Structural Isotropic Fuel\nVHTR Very High Temperature Reactor\n512 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?\nHigh Temperature Gas-cooled Reactors (HTR or HTGR) are helium-cooled graphite-moderated nuclear \ufb01ssion reactors utilizing\nfully ceramic fuel. They are characterized by inherent safety features, excellent \ufb01ssion product retention in the fuel, and high temper-\nature operation suitable for the delivery of industrial process heat, in particular hydrogen production. Typical coolant outlet temper-", "characters": 1591, "tokens": 390}
{"id": "2021_MF_VHTR.pdf_2", "source": "2021_MF_VHTR.pdf", "chunk": "fully ceramic fuel. They are characterized by inherent safety features, excellent \ufb01ssion product retention in the fuel, and high temper-\nature operation suitable for the delivery of industrial process heat, in particular hydrogen production. Typical coolant outlet temper-\natures range between 750/C14C and 850/C14C, thus enabling power conversion ef \ufb01ciencies up to 48%.\nThe Very High Temperature Reactor (VHTR) is a longer term evolution of the HTR targeting even higher ef \ufb01ciency and more\nversatile use by further increasing the helium outlet temperature to 950/C14C or even higher ( Gougar, 2011 ).\nFrom groundbreaking technology .\nThe HTR has evolved from the early gas-cooled reactors (GCRs) that gained widespread popularity for their simplicity and high\npower conversion ef \ufb01ciencies ( Beech and May, 1999 ). The \ufb01rst commercial nuclear power plant was a CO 2-cooled graphite-\nmoderated Magnox reactor (Calder Hall in 1956). In total, 26 Magnox reactors were built (270 \u20131760 MWth), with the last one\n(Wylfa-1, 1971 \u20132015) shut down at the end of 2015. As a second generation, 14 Advanced Gas-Cooled Reactors (AGRs) were\ndeployed in seven nuclear power plants at six sites in the UK with a total capacity of approx. 8 GWe. All these AGRs are expected\nto remain in operation until 2023 \u201330, although their life extension required clearance of graphite cracking issues and two power", "characters": 1389, "tokens": 339}
{"id": "2021_MF_VHTR.pdf_3", "source": "2021_MF_VHTR.pdf", "chunk": "to remain in operation until 2023 \u201330, although their life extension required clearance of graphite cracking issues and two power\nplants 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-\nical and chemical contaminants) and use a conventional steam cycle ( /C24540\n/C14C, same as for coal \ufb01red power plants) resulting in\nhigh thermal ef \ufb01ciencies ( >40%). However, GCRs had to observe a temperature limitation due to the dissociation of CO 2and the\nresulting carburization of structural materials and oxidation of graphite at elevated temperatures. Modern HTR are characterized by\nincreased operating temperature and thermal ef \ufb01ciency, which could be achieved by two major changes: the designs adopted\nhelium as the cooling gas along with fully ceramic fuel, which is discussed in more detail in Section \u201cTRISO fuel: Key to perfor-\nmance and safety .\u201d\nThe\ufb01rst HTR was proposed in a 1945 design study in the US, but was never realized. It featured a primary circuit (helium at\n1.55 MPa, 438 \u2013732/C14C) coupled to a secondary Brayton power conversion cycle (air at 2.9 MPa, 677 \u201322/C14C) leading to an expected\npower rating of 5 MWe.\nIn 1962 \u201363, a 3.3 MWth Mobile Low-Power Reactor (ML-1) with 140 (330 nominal) kWe was built in the US with a closed-cycle", "characters": 1585, "tokens": 380}
{"id": "2021_MF_VHTR.pdf_4", "source": "2021_MF_VHTR.pdf", "chunk": "In 1962 \u201363, a 3.3 MWth Mobile Low-Power Reactor (ML-1) with 140 (330 nominal) kWe was built in the US with a closed-cycle\nnitrogen turbine. The project was not pursued because it could not ful \ufb01ll the power output expectations.\nIn 1964, the Experimental Gas-Cooled Reactor (EGCR) was built at ORNL in the US, but not completed. This was basically\na helium-cooled AGR-type reactor using stainless steel fuel rod clusters. EGCR was expected to produce 85 MWth/25 MWe with\nhelium at 566/C14C.\nEnsuing developments led to conceptual changes in the existing gas-cooled reactors involving, as mentioned, in particular the\nuse 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\nreactor outlet temperature and improved safety performance.\nThe\ufb01rst tangible step in this direction was made in the UK with the DRAGON reactor (see also Section \u201cWhich HTR versions\nwere developed? \u201d). With a power of 21.5 MWth, it was an OECD project and operated from 1964 to 1975 primarily as a test bed for\nHTR fuel development. It used already early versions of fully ceramic coated particles as its own fuel.\nIn the US, the Ultra-High-Temperature Reactor Experiment (UHTREX) operated at LANL from 1966 to 1970. Its rated power was\n3 MWth using helium at 3.4 MPa (870 \u20131300/C14C). It used extruded fuel with TRISO coated particles in an annular rotatable core for\non-line refueling.", "characters": 1439, "tokens": 371}
{"id": "2021_MF_VHTR.pdf_5", "source": "2021_MF_VHTR.pdf", "chunk": "3 MWth using helium at 3.4 MPa (870 \u20131300/C14C). It used extruded fuel with TRISO coated particles in an annular rotatable core for\non-line refueling.\nMore details on the development of HTR technology can be found in a recent authoritative summary ( Kugeler and Zhang, 2019 ).\n.to modern characteristics\nThe following developments led to basic technical characteristics and design features shared by all modern HTR.\n\u0081can be built with passive safety features up to 625 MWth/core (prismatic block type core) and 250 MWth/core (pebble bed\ncore); this is the power range of Small Modular Rectors (SMR);\n\u0081long grace time after an accident (large heat capacity, low power density);\n\u0081self-stabilization of power transients (negative temperature coef \ufb01cient);\n\u0081low source terms (outstanding \ufb01ssion product retention in robust TRISO coated fuel particles and structures);\n\u0081fully ceramic core (fuel and moderator/re \ufb02ector);\n\u0081high-purity graphite as moderator/re \ufb02ector, high thermal inertia;\n\u0081chemically and neutronically inert helium as primary coolant;\n\u0081high operating temperatures for high ef \ufb01ciency, capability for nuclear cogeneration of heat and power, including for bulk\nhydrogen production;The High Temperature Gas-Cooled Reactor 513\u0081high burn-up capability;\n\u0081high fuel utilization (good neutron economy and possible use of thorium).\nThere are two competing HTR designs based on the TRISO coated fuel particle ( Fig. 1 ): the prismatic block core and the pebble bed\ncore ( Fig. 2 ).\n\u0081Invented by Peter Fortescue and his team at General Dynamics in the US, the prismatic block core is built from hexagonal", "characters": 1605, "tokens": 382}
{"id": "2021_MF_VHTR.pdf_6", "source": "2021_MF_VHTR.pdf", "chunk": "core ( Fig. 2 ).\n\u0081Invented by Peter Fortescue and his team at General Dynamics in the US, the prismatic block core is built from hexagonal\ngraphite blocks containing vertical holes. Some of these holes are used for helium cooling, while others receive the fuel in the\nform of \u201ccompacts \u201d, which are little cylinders (typically B12.3/C225 mm) pressed from graphite and coated fuel particles\n(Fortescue, 1975 ).\n\u0081The pebble bed HTR was conceived in 1942 by Farrington Daniels in the US ( Daniels, 1944 ). This early vision was later\ndeveloped to a power plant design by Rudolf Schulten in Germany, which employed B60 mm fuel spheres made of graphite\nand coated fuel particles ( Schulten et al., 1959 ). These pebbles are \ufb01lled into the reactor pressure vessel, which is internally lined\nwith graphite blocks. The resulting pebble bed constitutes the reactor core. The pebble bed can \ufb02ow and allows discharge and\n(re-)injection of pebbles during operation, enabling online refueling.\nTRISO fuel: Key to performance and safety\nOne of the major challenges and key to achieving a fully ceramic reactor core was fuel development ( IAEA, 2010 ). The initially used\nUO 2or UC fuel was placed in ceramic clads which showed poor \ufb01ssion product retention. Coated particle fuel was invented\nbetween 1957 and 1961 by the United Kingdom Atomic Energy Authority (UKAEA) and Battelle, but no patent was granted at\nthat time. The UO 2fuel kernels were made by external gelation of uranyl nitrate in ammonia and, after a heat treatment, coatings", "characters": 1526, "tokens": 370}
{"id": "2021_MF_VHTR.pdf_7", "source": "2021_MF_VHTR.pdf", "chunk": "that time. The UO 2fuel kernels were made by external gelation of uranyl nitrate in ammonia and, after a heat treatment, coatings\nwere deposited on top of these kernels via pyrolysis of hydrocarbons in a \ufb02uidized bed. The next development step was the early\nBISO (bi-structural isotropic) particle fuel comprising a buffer layer directly deposited on the kernels and an additional pyrolytic\ncarbon (PyC) layer on top. Finally, modern TRISO (tri-structural isotropic) particles were given an additional SiC diffusion barrier\nleading to con \ufb01rmed \ufb01ssion product retention up to 1600/C14C or even higher ( Gougar et al., 2020 ). These TRISO coated particles,\ntypically in the order of 1 mm in diameter, are the basis for all modern HTR fuel designs ( Gerczak, 2021 ;Helmreich, 2021 ). As\nshown in Fig. 1 , they feature (from inside out) the kernel, a porous PyC buffer to accommodate fuel swelling and \ufb01ssion gases,\na dense PyC buffer and a dense SiC layer as diffusion barriers against \ufb01ssion product escape, and a \ufb01nal PyC layer (missing in\nFig. 1 ) for better bonding with the matrix graphite into which they will be integrated.\nBaked into matrix graphite, the TRISO coated particles can now be given a macroscopic shape ( Fig. 2 ), usually in the form of\nthumb-thick cylinders ( \u201ccompacts \u201d) either solid or annular, or in the form of spherical fuel elements ( \u201cpebbles \u201d). The compacts\nare inserted into hexagonal blocks made of graphite, which are then assembled to constitute the reactor core contained in a pressurevessel.", "characters": 1527, "tokens": 384}
{"id": "2021_MF_VHTR.pdf_8", "source": "2021_MF_VHTR.pdf", "chunk": "are inserted into hexagonal blocks made of graphite, which are then assembled to constitute the reactor core contained in a pressurevessel.\nTypical pebble and compact design characteristics are given in Table 1 :\nWhich HTR versions were developed?\nBased on these characteristics, in the 1960s two different types of reactors were designed and built, primarily to produce electricity.\nExperimental HTRs with a prismatic block core and TRISO coated particle fuel were developed in the UK (DRAGON reactor,\nFuel kernel buffer SiC Inner PyC \nFig. 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 \u20131975, 21.5 MWth, an OECD project ( Price, 2012 ) and in the US (Peach Bottom, operated 1966 \u20131974, 115 MWth/\n40 MWe Beck and Pincock, 2011 ). They were followed by the prototype of the Fort St. Vrain Generating Station (operated 1976 \u2013\n1989, 842 MWth/330 MWe, Beck and Pincock, 2011 ). This reactor established the technical feasibility of HTRs although it experi-\nenced problems ( Rempe, 2021 ) of power \ufb02uctuations, jamming of a control rod and leakage of moisture into the core, which \ufb01nally\ncaused its decommissioning for economic reasons.\nOver the same period, Germany developed and built an experimental pebble bed reactor (AVR, 46 MWth/15 MWe, Pohl, 2008 )", "characters": 1406, "tokens": 352}
{"id": "2021_MF_VHTR.pdf_9", "source": "2021_MF_VHTR.pdf", "chunk": "caused its decommissioning for economic reasons.\nOver the same period, Germany developed and built an experimental pebble bed reactor (AVR, 46 MWth/15 MWe, Pohl, 2008 )\nat the J\u00fclich 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\na water ingress accident provoked by a steam generator leak it could be repaired, dried and returned to service. Following this expe-\nrience, 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\na number of technical dif \ufb01culties. Examples of design issues are the direct insertion of the control rods in the pebble bed (causing\nPyrolytic carbonCeramic\nkernel\nCoated\nParticlePebble\nParticles Compacts Fuel ElementsSilicon Carbide\nUranium Oxycarbide kernel\nFig. 2 TRISO coated particle fuel as the basis for hexagonal block and pebble bed core designs ( Gougar et al., 2020 ).\nTable 1 Typical examples for nominal characteristic data of German AVR GLE-4 particles and pebbles and US NGNP particles and compacts.\nCoated particle AVR pebble NGNP compact\nKernel composition UO 2 UCO\nKernel diameter [ mm] 502 425\nEnrichment [U-235 wt%] 16.76 14\nThickness of coatings [ mm]:\nbufferinner PyCSiC\nouter PyC92", "characters": 1519, "tokens": 391}
{"id": "2021_MF_VHTR.pdf_10", "source": "2021_MF_VHTR.pdf", "chunk": "Kernel diameter [ mm] 502 425\nEnrichment [U-235 wt%] 16.76 14\nThickness of coatings [ mm]:\nbufferinner PyCSiC\nouter PyC92\n403540100\n403540\nParticle diameter [ mm] 916 855\nFuel element (FE) Pebble Compact\nDimensions [mm] B60\n(spherical)B12.3/C225\n(cylindrical)\nHeavy metal loading [g/FE] 6.0 1.27\nU-235 content [g/FE] 1.00 0.18\nNumber of coated particles per FE 9560 3175Volume packing fraction [%] 6.2 35Fraction of factory defective SiC coatings 7.8 /C210\n/C06<1.2/C210/C05\nMatrix density [kg/m3] 1750 1600\nTemperature at \ufb01nal 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\nChernobyl accident after only 3 years of operation.\nIn the same period, the Power Nuclear Project (PNP-500, 500 MWth, Neef and Weisbrodt, 1979 ) started in Germany aiming at\nusing nuclear heat to produce hydrogen by steam methane reforming. This project led to development and testing of large modules\nof 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.", "characters": 1194, "tokens": 365}
{"id": "2021_MF_VHTR.pdf_11", "source": "2021_MF_VHTR.pdf", "chunk": "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.\nIn the 1980s, Interatom/Siemens in Germany developed the 200 MWth HTR-Modul as the \ufb01rst modular pebble bed design con-\nsisting of a metallic reactor pressure vessel connected to an adjacent steam generator through a hot gas duct ( Siemens, 1988 ). The\nconcept features a simpli \ufb01ed design with a size and power rating chosen to enable passive decay-heat removal after a loss-of-\ncoolant-accident solely by conduction and radiation. No natural or forced convection is necessary ( Reutler and Lohnert, 1984;\nKugeler et al., 2017 ). Although it was never built, the HTR-Modul has served as the basis for the PBMR in South Africa and for\nthe HTR-10 and HTR-PM reactors in China.\nThe Gas Turbine Modular Helium Reactor (GT-MHR, LaBar, 2002 ) is a 600 MWth design developed by a group of Russian and\nUS 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 \ufb01ciency of up to\n48% based on a reactor outlet temperature of 850\n/C14C. Extensive analysis has shown that this reactor, and more generally most HTR\ndesigns, are particularly suitable for the incineration of excess plutonium which became an issue in the US and in the former USSR", "characters": 1482, "tokens": 373}
{"id": "2021_MF_VHTR.pdf_12", "source": "2021_MF_VHTR.pdf", "chunk": "/C14C. Extensive analysis has shown that this reactor, and more generally most HTR\ndesigns, are particularly suitable for the incineration of excess plutonium which became an issue in the US and in the former USSR\nfor 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\nGT-MHR regulatory process started in 2002 but was not completed. More recently, the GT-MHR design was proposed by General\nAtomics 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\nsmaller version (SC-HTGR, 350 MWth) equally with indirect steam cycle was proposed by AREVA/Framatome as well ( AREVA,\n2014 ). The GT-MHR was also the basis for the Japanese GT-HTR300 designed by JAEA ( Kunitomi et al., 2004 ).\nA review summary on the 7 built reactors (Dragon, Peach Bottom, Fort St. Vrain, AVR, THTR, HTTR and HTR-10) can be found in\n(Beck and Pincock, 2011 ). The experience of past experimental and prototype HTRs demonstrated their technical viability, however,\nthey 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.", "characters": 1606, "tokens": 391}
{"id": "2021_MF_VHTR.pdf_13", "source": "2021_MF_VHTR.pdf", "chunk": "One of these new projects was the Pebble Bed Modular Reactor (PBMR, Matzner, 2004 ) in the Republic of South Africa. PBMR\nPty. 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\nof 400 MWth and a direct power conversion with a gas turbine operating with a helium outlet temperature of 900\n/C14C. In June 2003\nthe 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 \ufb02eet of 24 PBMRs so as to make up 4000 MWe out of the\n12,000 MWe additional nuclear capacity planned by 2030. Large facilities dedicated to PBMR speci \ufb01c technologies testing were built\nin 2007: a \u201cHeat Transfer Test Facility \u201d,a\u201cHelium Test Facility \u201d,a\u201cPebble Bed Micro Model \u201dand an \u201cElectro-magnetic blower. \u201dA\nfuel laboratory developed manufacturing processes and quality assurance testing techniques in collaboration with NECSA andsuccessfully manufactured coated fuel particles with enriched uranium in December 2008.\nIn 2009 the PBMR project, like other projects of nuclear equipment in South Africa, faced funding dif \ufb01culties and had its busi-\nness plan re-oriented towards the supply of industrial process heat, a dif \ufb01cult endeavor in a country with large coal reserves and no\nCO\n2emission limits. The new focus of the PBMR was on onsite power, cogeneration, seawater desalination and direct process heat", "characters": 1622, "tokens": 391}
{"id": "2021_MF_VHTR.pdf_14", "source": "2021_MF_VHTR.pdf", "chunk": "CO\n2emission limits. The new focus of the PBMR was on onsite power, cogeneration, seawater desalination and direct process heat\ndelivery. Target process heat applications included coal-to-liquid or gaseous fuels, petrochemicals, ammonia/fertilizer, re \ufb01neries,\nsteam for oil sand recovery, bulk hydrogen for future transportation and water desalination. Thus, PBMR Ltd. started developingoptions for commercial \ufb02eets with Sasol (the South African coal liquefaction company), with the utility Eskom for electricity, as\nwell as with US and Canadian cogeneration end users including oil sand producers. The PBMR project was accordingly revisited\nto 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\n/C14C at the core outlet and a steam generator directly placed in the primary\nloop. A conventional subcritical steam turbine was selected for \ufb01rst generation plants whereas super-critical cycles were envisaged\nfor next generation plants.\nDue to funding issues and problems in the interaction between PBMR and the South African regulator the project was stopped in\n2010. This development was analyzed critically in ( Thomas, 2011 ). Another investigation with negative conclusions from opera-\ntional performance of HTR in the past with a pessimistic outlook is summarized in ( Ramana, 2016 ).\nSince then, the aforementioned technological problems encountered with test reactors (e.g. moisture leakages into the core) have\nbeen solved to a large extent, so that most recent HTR designs could be deliberately geared towards short-term realization with\nminimum R&D efforts and development risks. In addition, with a much longer-term view, a number of research organizations", "characters": 1791, "tokens": 371}
{"id": "2021_MF_VHTR.pdf_15", "source": "2021_MF_VHTR.pdf", "chunk": "been solved to a large extent, so that most recent HTR designs could be deliberately geared towards short-term realization with\nminimum R&D efforts and development risks. In addition, with a much longer-term view, a number of research organizations\ncooperate internationally on the Very High Temperature Reactor, which is usually understood to produce heat above 950/C14Ct o\nmaximize power conversion ef \ufb01ciency and to enable ambitious process heat applications such as thermochemical hydrogen\nproduction with the S-I cycle. The VHTR is thus a long-term concept requiring new materials and design codes along with fuel qual-\ni\ufb01cation for the higher temperatures. The very signi \ufb01cant progress of this cooperation is summarized in ( F\u00fctterer et al., 2014 ).516 The High Temperature Gas-Cooled ReactorRecent results from test reactors\nIn the 1990s, the Japan Atomic Energy Agency (JAEA) built a research reactor in Oarai, the High Temperature Test Reactor (HTTR,\n(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\ndesign power of 30 MWth in 2001 with a helium outlet temperature of 850/C14C. Subsequent tests until 2010 have demonstrated the\nsafe behavior of the reactor. This included reactivity insertion as well as partial and complete loss of forced cooling, but not yet at full\npower. The HTTR was successfully operated at the design temperature of 950/C14C\ufb01rst in 2004, then for 50 continuous days in 2010.\nIn parallel with tests on the HTTR, JAEA is developing the S-I thermo-chemical process to produce hydrogen ( Fig. 4 ). A\ufb01rst demon-", "characters": 1630, "tokens": 384}
{"id": "2021_MF_VHTR.pdf_16", "source": "2021_MF_VHTR.pdf", "chunk": "In parallel with tests on the HTTR, JAEA is developing the S-I thermo-chemical process to produce hydrogen ( Fig. 4 ). A\ufb01rst demon-\nstration of this process was achieved in 2003 when a continuous production of 30 l/h of hydrogen was maintained for several days.\nDuring the March 2011 earthquake, which triggered the Fukushima accident, the HTTR was only slightly damaged. After extensive\ninspection, 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.\nFig. 3 External view of the HTTR building in Japan.\nFig. 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\nreactor HTR-10 (10 MWth) ( Dong, 2012; Wu et al., 2002 ;Fig. 5 ) that was put into service in 2000. The successful operation of\nthis reactor demonstrated an updated pebble bed core HTR technology. In particular, it served as a test bed for fuel, components\nand 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\nscaling up this technology to the High Temperature Reactor \u2013Pebble bed Module (HTR-PM, 210 MWe) project ( Zhang et al., 2016 )\n.", "characters": 1554, "tokens": 365}
{"id": "2021_MF_VHTR.pdf_17", "source": "2021_MF_VHTR.pdf", "chunk": "scaling up this technology to the High Temperature Reactor \u2013Pebble bed Module (HTR-PM, 210 MWe) project ( Zhang et al., 2016 )\n.\nTogether with their predecessors, HTTR and HTR-10 have signi \ufb01cantly contributed to the establishment of the rather high tech-\nnology readiness level both for block type and pebble bed HTR designs.\nThe case for new next generation HTRs\nWhy have past HTRs not been successful economically and why do we think that this is changing?\nGCRs were developed worldwide, but only the AGRs in the UK remain in commercial operation. After reasonable experiences\nwith the \ufb01rst HTR plants in the UK (Dragon), the US (Peach Bottom Unit 1) and Germany (AVR), national HTR programs ended\nwith no commercial deployments for various reasons. In the UK, the Thatcher government decided to build PWRs essentially\nbecause of absence of con \ufb01rmed economic data for other designs, higher perceived \ufb01nancial risk of HTR designs compared to\nthe mainstream PWR, and because of the then unsolved dif \ufb01culty to integrate the HTR into a long-term sustainable closed fuel cycle\nthat 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\npremature shutdown in 1989. This has coincided with the time period of three decades without new nuclear orders in the US start-\ning with the Three Mile Island accident in 1979.\nIn general, most HTR operational issues were associated, as already mentioned, with leakages, e.g. moisture ingress that resulted", "characters": 1643, "tokens": 383}
{"id": "2021_MF_VHTR.pdf_18", "source": "2021_MF_VHTR.pdf", "chunk": "ing with the Three Mile Island accident in 1979.\nIn general, most HTR operational issues were associated, as already mentioned, with leakages, e.g. moisture ingress that resulted\nin corrosion of components, core temperature oscillations caused by coolant \ufb02ow bypass and in-core behavior of graphite (cracking,\ndimensional changes, movement of blocks and distortions, dust formation) ( Beck and Pincock, 2011 ). Most of these were \ufb01rst-of-a-\nkind operational issues and took a long time to resolve without the bene \ufb01t of the broader industry experience that is dominated by\nwater-cooled reactors. As a result, it led to poor performance in some HTR reactors, most notably the Fort St. Vrain reactor in the US.\nOn the positive side, however, the operational experiences with HTRs showed excellent fuel performance and demonstrated the\nconcept \u2019s inherent safety features. Many lessons learned through past HTR experiences led to improvements in modern HTR\nconcepts, such as the use of magnetic bearings in the helium circulator, or the use of a steel pressure vessel for improved reliability\ninstead of a pre-stressed concrete vessel. Passive cooling systems, requiring no pumps or monitoring systems to initiate them, have\nbeen adopted. The excellent performance of TRISO fuel is further improved by recent extensive research programs ( Electric Power\nResearch Institute, 2019 ), which bene \ufb01tted both fuel types, compact and pebble. These developments eliminate major known issues\nexperienced by early HTRs and further corroborate HTR safety characteristics.\nOngoing HTR development\nThe last decade has seen signi \ufb01cantly growing interest worldwide in Small Modular Reactors, which the IAEA de \ufb01nes as units\nproducing less than 300 MWe. A snapshot of the very dynamic SMR landscape is given in ( IAEA, 2018 ). These reactors are being", "characters": 1834, "tokens": 394}
{"id": "2021_MF_VHTR.pdf_19", "source": "2021_MF_VHTR.pdf", "chunk": "producing less than 300 MWe. A snapshot of the very dynamic SMR landscape is given in ( IAEA, 2018 ). These reactors are being\ndesigned by several classical vendor companies and start-ups for \ufb02exibility, affordability, for a wide range of users and applications,\nFig. 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-\nmodule nuclear power plants, and are designed to be built in factory workshops and shipped to utilities for installation as demand\nevolves. Fig. 6 shows how a multi-module pack could be con \ufb01gured to polygenerate heat, hydrogen and electricity.\nSeveral designs ensure enhanced safety performance through inherent and passive safety features as well as suitability for cogen-\neration 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\nvendor companies participating in feasibility and licensing studies.\nAbout 16 of these SMR designs are HTRs with one currently under construction and commissioning in China (HTR-PM). Several\nof these reactors are derivatives or evolutions of earlier parent concepts, e.g. the HTR Modul for the pebble bed designs and the\nMHTGR-350 (designed by General Atomics) for several prismatic block designs. For the HTR concepts in Table 2 , publicly available\ndesign information can be found in ( IAEA, 2018 ).\nR&D efforts as well as cooperation between all stakeholders (vendors, suppliers, regulators, utilities/end-users, investors, poli-", "characters": 1769, "tokens": 365}
{"id": "2021_MF_VHTR.pdf_20", "source": "2021_MF_VHTR.pdf", "chunk": "design information can be found in ( IAEA, 2018 ).\nR&D efforts as well as cooperation between all stakeholders (vendors, suppliers, regulators, utilities/end-users, investors, poli-\nticians, public etc.) are ongoing and organized at different national and international levels including GIF, IAEA, OECD-NEA, and\nare including economic analyses, as well as novel investment options and licensing approaches, e.g. ( Gougar et al., 2020 ;Kalilainen\net al., 2019 ). While the nuclear accident in Fukushima in 2011 has dealt a blow to nuclear energy development for several years, the\nongoing debate about climate change mitigation has created new interest in low-carbon technologies in several countries and specif-\nically awareness of the need to address the massive energy requirements of the process heat market in industrialized countries. As\nshown in Table 2 , interest in the inherently safe, highly ef \ufb01cient and versatile HTR technology is steadily growing, and new\nFig. 6 Artist \u2019s view of a 4-pack modular HTR for process heat, hydrogen production and electricity generation (INL).\nTable 2 Summary of HTR-type small modular reactor concepts.\nConcept Developer\nPebble Bed\nHTR-PM Tsinghua University, China\nXe-100 X-energy, USA\nHTMR-100 Steenkampskraal Thorium Ltd., South Africa\nPBMR-400 Pebble Bed Modular Reactor SOC Ltd., South Africa\nAHTR-100 Eskom Holdings SOC Ltd., South Africa\nHexagonal Block\nGTHTR300 Japan Atomic Energy Agency, Japan\nMHTGR-350 General Atomics, USA\nGT-MHR OKBM Afrikantov, Russian Federation\nMHR-T Reactor/Hydrogen Production Complex OKBM Afrikantov, Russian Federation", "characters": 1600, "tokens": 382}
{"id": "2021_MF_VHTR.pdf_21", "source": "2021_MF_VHTR.pdf", "chunk": "MHTGR-350 General Atomics, USA\nGT-MHR OKBM Afrikantov, Russian Federation\nMHR-T Reactor/Hydrogen Production Complex OKBM Afrikantov, Russian Federation\nMHR-100 OKBM Afrikantov, Russian Federation\nSC-HTGR Framatome Inc., USA\nMMR-5, MMR-10 UltraSafe Nuclear Corporation, USA\nStarCore HTGR StarCore Nuclear, Canada\nU 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\nimplemented to help de-risk (and possibly shorten the time to) industrial deployment.\nBeyond electricity: Emission-free process heat and cogeneration\nBecause HTRs are particularly \ufb01t for process heat applications and cogeneration of heat and power, this section is dedicated to non-\npower utilization aspects of nuclear energy, which has very signi \ufb01cant potential impacts since it reduces fossil fuel consumption in\nareas beyond the electric power market, and thus enhances energy security, further increases the reduction of noxious emissions, and\nhelps 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\nReactors (SFBR). District heating (80 \u2013150\n/C14C) is probably the most widely found application of nuclear heat: 46 reactors in 12\ncountries, including for instance Slovakia, Switzerland, Russia and China were and are used for this purpose.\nExamples for low temperature applications of nuclear heat include seawater desalination (Japan, Kazakhstan), paper and card-\nboard industry (Norway, Switzerland), heavy water distillation (Canada), or salt re \ufb01ning (Germany).", "characters": 1778, "tokens": 395}
{"id": "2021_MF_VHTR.pdf_22", "source": "2021_MF_VHTR.pdf", "chunk": "Examples for low temperature applications of nuclear heat include seawater desalination (Japan, Kazakhstan), paper and card-\nboard industry (Norway, Switzerland), heavy water distillation (Canada), or salt re \ufb01ning (Germany).\nThe technology options for nuclear process heat utilization with HTRs were already documented quite early ( Schulten, 1976 ). A\nsurvey of two decades of activities in Germany is given in ( Verfondern, 2007a ), and further potential is outlined in ( Verfondern, 2007b ).\nThe HTR produces heat at a much higher temperature level (exergy) than the LWR. This opens the possibility to replace a large\nnumber of existing industrial cogeneration plants delivering process steam in the 500 \u2013600/C14C temperature range. Very signi \ufb01cant\namounts 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 \ufb01ring.\nFor several stakeholders, in particular in those countries where natural gas is expensive, the prospect of hydrogen production\ncontinues 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\n/C14C, 5.5 MPa. Owing to the external heat supply, more than a third of natural gas is saved.\nIn the 1980s, the necessary components, e.g. heat exchangers or reformers, were developed and tested under nuclear conditions in\nGermany and in Japan ( Harth et al., 1990 ).", "characters": 1615, "tokens": 368}
{"id": "2021_MF_VHTR.pdf_23", "source": "2021_MF_VHTR.pdf", "chunk": "In the 1980s, the necessary components, e.g. heat exchangers or reformers, were developed and tested under nuclear conditions in\nGermany and in Japan ( Harth et al., 1990 ).\nProcesses and components for allothermal and steam coal gasi \ufb01cation processes were also tested in Germany. They require typi-\ncally steam in the range of 750 \u2013900/C14C at 0.1 \u20134 MPa. Although external heat supply makes coal upgrading more ef \ufb01cient, these\nprocesses release large amounts of unwanted CO 2.\nThese activities were brought to a temporary halt in an anti-nuclear climate after the Chernobyl accident, with inexpensive oil\nand gas and in absence of CO 2emission restrictions.\nAs steam methane reforming to produce hydrogen consumes natural gas and generates CO 2emissions in the process, direct\nwater splitting methods are under investigation in several countries as a clean alternative. HTRs can provide steam for a rather\nlow temperature process, the copper-chlorine (Cu-Cl) cycle, requiring steam at just over 500/C14C(Rosen et al., 2012 ). Other prom-\ninent hydrogen production methods are (i) High Temperature Steam Electrolysis (750 \u2013950/C14C) where a part of the required water\ndissociation 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\n3decomposition) requires heat input at 850/C14C(Yan and Hino, 2011 ). This process is particularly suitable\nfor VHTR operating at 900 \u20131000/C14. The market for bulk hydrogen is currently very large and growing fast, with distribution networks", "characters": 1571, "tokens": 374}
{"id": "2021_MF_VHTR.pdf_24", "source": "2021_MF_VHTR.pdf", "chunk": "for VHTR operating at 900 \u20131000/C14. The market for bulk hydrogen is currently very large and growing fast, with distribution networks\nalready in place in several countries. To justify large-scale production of hydrogen, the development of a speci \ufb01c\u201chydrogen\neconomy \u201dis not required. Hydrogen uses include upgrading of increasingly heavy oils to lighter fractions, hydrogenation processes,\nhydro coal gasi \ufb01cation, metal re \ufb01ning, ammonia production for fertilizers, the synthesis of methanol or synfuel, or the use of\nhydrogen 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\nin carbon capture and utilization processes, which would use CO\n2together with hydrogen as a feedstock for the fabrication of\na 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 ).\nIn the context of energy system integration efforts with growing fractions of variable renewable electricity in many countries, it is\nof particular interest that the cogeneration capability of HTRs would allow it to contribute to grid stabilization ( \u201cpeak shaving \u201d), e.g.\nby modulating the production of (storable) hydrogen depending on the electricity demand in the grid, similar to what is currentlyenvisaged for wind energy ( \u201cpower to gas \u201d).\nTo further corroborate the incentive for process heat and hydrogen production with nuclear energy, several market research,", "characters": 1665, "tokens": 340}
{"id": "2021_MF_VHTR.pdf_25", "source": "2021_MF_VHTR.pdf", "chunk": "To further corroborate the incentive for process heat and hydrogen production with nuclear energy, several market research,\neconomic 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 ).\nOutlook\nThe unique capability of the HTR to produce process heat above 600/C14C makes it an ef \ufb01cient reactor type to displace fossil fuels in\nvarious 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 \ufb01ning, petro-chemistry, oil sand recovery, chemistry, steelmaking, etc.). Several market studies\ncon\ufb01rmed the potential for the HTR system to be used in such applications while the economic boundary conditions (e.g. price\nof natural gas, CO 2tax) for market deployment have become clearer. The inherent safety characteristics of the HTR are a precious\nasset in contributing convincing answers to today \u2019s concerns in terms of nuclear safety, energy security, and climate change.\nCurrent research performed within frameworks supported by GIF, IAEA and OECD-NEA, as well as speci \ufb01c national programs\naddress primarily issues related to R&D, licensing, demonstration, and deployment. In particular, the multinational cooperation\nwithin GIF ( GIF, 2018 ) allows sharing efforts to advance the technologies and to accelerate development in view of licensing\nand deployment. Currently, cooperation on the VHTR within GIF focuses on development and quali \ufb01cation of (i) fuel, (ii) struc-", "characters": 1696, "tokens": 377}
{"id": "2021_MF_VHTR.pdf_26", "source": "2021_MF_VHTR.pdf", "chunk": "and deployment. Currently, cooperation on the VHTR within GIF focuses on development and quali \ufb01cation of (i) fuel, (ii) struc-\ntural and functional materials, (iii) hydrogen production processes and (iv) computer tools. GIF has also produced guidance for (V)\nHTR designers, e.g. in the areas of sustainability, economy, reactor safety, non-proliferation questions or energy system integration.\nThe 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\ndeployment is being taken by China with the ongoing commissioning of the HTR-PM demonstrator ( Fig. 7 ). Japan will perform\nfurther safety demonstrations on the HTTR.\nSince 2002, the bi-annual International Topical Meeting on High Temperature Reactor Technology is the sole international\nconference with focus on HTR and process heat applications ( https://htr2020.org/ ).\nAlthough very substantial results were produced, in particular by the signatories of the GIF VHTR System Arrangement, funding\nopportunities for a demonstrator coupled with an end-user process will have to be found soon to capitalize on previous invest-\nments. Several such international initiatives are on the way. Their success will depend on how much and where nuclear will be\nallowed to contribute to climate change mitigation, be it for political and public acceptance reasons or for economic boundaryconditions (cheap natural gas, CO\n2tax,\ufb01nancial risk).\nSee Also: Fuel Design and Fabrication: TRISO Particle Fuel; Pebble Bed Gas Cooled Reactors; Self-Sustaining Breeding in Advanced Reactors:\nCharacterization of Selected Reactors.", "characters": 1879, "tokens": 387}
{"id": "2021_MF_VHTR.pdf_27", "source": "2021_MF_VHTR.pdf", "chunk": "See Also: Fuel Design and Fabrication: TRISO Particle Fuel; Pebble Bed Gas Cooled Reactors; Self-Sustaining Breeding in Advanced Reactors:\nCharacterization of Selected Reactors.\nFig. 7 Installation of RPV into HTR-PM reactor building in 2016.The High Temperature Gas-Cooled Reactor 521References\nAngulo, 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 \u201337.\nAREVA, 2014. HTGR Information Kit. March 2014.Baumer, R., Kalinowski, I., 1991. THTR commissioning and operating experience. Energy 16 (1991), 59 \u201370.\nBeck, 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\nRevision 1, April 2011.\nBeech, 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\non Survey on Basic Studies in the Field of High Temperature Engineering. OECD NEA, Paris, France, 27\u201329 September 1999 .\nBredimas, 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,\nJapan, 28 October\u20131 November 2012 .", "characters": 1321, "tokens": 355}
{"id": "2021_MF_VHTR.pdf_28", "source": "2021_MF_VHTR.pdf", "chunk": "Japan, 28 October\u20131 November 2012 .\nDaniels, F., 1944. Suggestions for a High-Temperature Pebble Pile. MUC-FD-8; N-1668b. Chicago University Metallurgical Laboratory, Chicago, Ill inois.\nDietrich G, Michels J, Cleve U (2019) Personal communication 2015 \u20132020.\nDong, Y., 2012. China \u2019s activities in HTGRs HTR-10 and HTR-PM. In: IAEA Course on High Temperature Gas Cooled Reactor Technology. Beijing, China, 22\u201326 October 2012 .\nElectric Power Research Institute, 2019. Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance. Topical Rep ort EPRI-AR-1, May 2019.\nFortescue, P., 1975. Advanced HTGR systems. Annals of Nuclear Energy 2 (11 \u201312).\nF\u00fctterer, M.A., Fu, L., Sink, C., de Groot, S., Pouchon, M., Kim, Y.W., Carr\u00e9, F., Tachibana, Y., 2014. Status of the very high temperature reactor syst em. Progress in Nuclear Energy\n77 (2014), 266 \u2013281.\nGerczak TJ 2021 Irradiation Performance: High-Temperature Gas Reactor Fuels. In: Encyclopedia of Nuclear Energy, vol. 2, pp. 407 \u2013419.\nGIF, 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-", "characters": 1193, "tokens": 383}
{"id": "2021_MF_VHTR.pdf_29", "source": "2021_MF_VHTR.pdf", "chunk": "energy-systems-2018-update .\nGougar, H., 2011. The Very High Temperature Reactor, Nuclear Energy Encyclopedia: Science, Technology, and Applications. Steven Krivit, Editor-i n-Chief. John Wiley and Sons,\nISBN 978-0-470-89439-2.\nGougar, H., et al., 2020. The US Department of Energy \u2019s high temperature reactor research and development program dProgress as of 2019. Nuclear Engineering and Design 358\n(2020), 110397.\nHarth, R., Jansing, W., Teubner, H., 1990. Experience gained from the EVA II and KVK operation. Nuclear Engineering and Design 121 (1990), 173 \u2013182.\nHelmreich, G., 2021. Fuel Design and Fabrication: TRISO Particle Fuel. In: Encyclopedia of Nuclear Energy, vol. 2, pp. 318 \u2013325.\nIAEA, 2010. High Temperature Gas Cooled Reactor dFuels and Materials. IAEA TECDOC 1645, 2010.\nIAEA (2018) Advances in Small Modular Reactor Technology Developments d2018 Edition, https://aris.iaea.org/Publications/SMR-Book_2018.pdf .\nINL, 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\nTechnology Symposium - SYP2019. Helsinki, Finland, 30\u201331 October 2019 .", "characters": 1273, "tokens": 371}
{"id": "2021_MF_VHTR.pdf_30", "source": "2021_MF_VHTR.pdf", "chunk": "Technology Symposium - SYP2019. Helsinki, Finland, 30\u201331 October 2019 .\nKonefal, 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., F\u00fctterer, M.A., 2017. The High Temperature Gas-cooled Reactor: Safety c onsiderations of the (V)HTR-\nModul. JRC Technical Report JRC107642, EUR 28712 EN, ISBN 978-92-79-71312-5. https://doi.org/10.2760/970340 .\nKunitomi, K., 2013. Status of HTTR Project in JAEA. In: TWGGCR Meeting at IAEA. 5 March 2013 .\nKunitomi, K., Katanishi, S., Takada, S., Takizuka, T., Yan, X., 2004. Japan \u2019s future HTR dThe GTHTR300. Nuclear Engineering and Design 233 (2004), 309 \u2013327.", "characters": 859, "tokens": 309}
{"id": "2021_MF_VHTR.pdf_31", "source": "2021_MF_VHTR.pdf", "chunk": "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 \u2013296.\nMatzner, D., 2004. In: Letcher, T.M. (Ed.), Chapter 14: The Pebble Bed Modular Reactor. Elsevier, Oxford, p. 2008.\nNeef, J., Weisbrodt, I., 1979. Coal gasi \ufb01cation with heat from high temperature reactors: Objectives and status of the project \u201c\nPrototype Plant for Nuclear Process Heat (PNP) \u201d.\nNuclear Engineering and Design 54 (1979), 157 \u2013174.\nPohl 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 .\nPrice, M.S.T., 2012. The Dragon project: Origins, achievements and legacies. Nuclear Engineering and Design 251 (2012), 60 \u201368.\nRamana MV (2016) The checkered operational history of high temperature gas-cooled reactors, Bulletin of the Atomic Scientists, 72:3, 171 \u2013179, https://doi.org/10.1080/\n00963402.2016.1170395 , 2016", "characters": 1148, "tokens": 359}
{"id": "2021_MF_VHTR.pdf_32", "source": "2021_MF_VHTR.pdf", "chunk": "00963402.2016.1170395 , 2016\nRempe, J.L., 2021. U.S. Nuclear Reactor Regulation of Two non-LWRs. In: Encyclopedia of Nuclear Energy, vol. 2, pp. 175 \u2013187.\nReutler, H., Lohnert, G.H., 1984. Advantages of going modular in HTRs. Nuclear Engineering and Design 78 (1984), 129 \u2013136.\nRosen, M., Naterer, G., Sadhankar, R., Suppiah, S., 2012. Nuclear-based hydrogen production with a thermochemical copper-chlorine cycle and super critical water reactor.\nInternational Journal of Energy Research 36 (4), 456 \u2013465. March 2012.\nSchulten, R., 1976. Nukleare Proze\u00dfw\u00e4rme. Chemie Ingenieur Technik 48 (1976), 375 \u2013380.\nSchulten R, Bellermann W, Braun H, Schmidt HW (1959) Der Hochtemperaturreaktor von BBC/Krupp (in German) , Die Atomwirtschaft.\nShropshire, D., 2013. Integration challenges for nuclear cogeneration coupled to renewable energy systems. In: Proc. Joint NEA/IAEA Expert Worksh op on the Technical and\nEconomic Assessment of Non-Electric Applications of Nuclear Energy (NUCOGEN). OECD Paris, France, 4\u20135 April 2013 .\nSiemens (1988) Hochtemperaturreaktor-Modul-Kraftwerksanlage, Kurzbeschreibung , Siemens/Interatom, Germany, November 1988", "characters": 1140, "tokens": 357}
{"id": "2021_MF_VHTR.pdf_33", "source": "2021_MF_VHTR.pdf", "chunk": "Siemens (1988) Hochtemperaturreaktor-Modul-Kraftwerksanlage, Kurzbeschreibung , Siemens/Interatom, Germany, November 1988\nStyring, 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\nMaterials. The Centre for Low Carbon Futures. July 2011.\nThomas, S., 2011. The pebble bed modular reactor: An obituary. Energy Policy 39 (5), 2431 \u20132440.\nVerfondern, 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\nof Nuclear Power: Seawater Desalination, Hydrogen Production and other Industrial Applications. Oarai, Japan, 16\u201319 April 2007 .\nVerfondern, 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\nDesalination, Hydrogen Production and other Industrial Applications. Oarai, Japan, 16\u201319 April 2007 .\nWu, Z., Lin, D., Zhong, D., 2002. The design features of the HTR-10. Nuclear Engineering and Design 218 (2002), 25 \u201332.", "characters": 1150, "tokens": 350}
{"id": "2021_MF_VHTR.pdf_34", "source": "2021_MF_VHTR.pdf", "chunk": "Wu, Z., Lin, D., Zhong, D., 2002. The design features of the HTR-10. Nuclear Engineering and Design 218 (2002), 25 \u201332.\nYan, 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\ntechnological innovation. Engineering 2 (2016), 112 \u2013118.522 The High Temperature Gas-Cooled Reactor", "characters": 485, "tokens": 154}
{"id": "2022_GIF_VHTR.pdf_35", "source": "2022_GIF_VHTR.pdf", "chunk": "Gen IV Gas-cooled Fast Reactor system PR&PP White Paper\n1\nGIF-LFR-WP-Rev9 \u2013 Limited: GIF        \nGIF GAS-COOLED  FAST  REACTOR\nPROLIFERATION  RESISTANCE  AND  \nPHYSICAL  PROTECTION  WHITE  \nPAPER\nProliferation  Resistance  and Physical  Protection  \nWorking  Group  (PRPPWG)\nSodium-Cooled  Fast Reactor  System  Steering  \nCommittee  (SFR  SSC)\nApril 2021\nSAND2022-6859R\nSandia\nNational\nLaboratories\nis\na\nmultimission\nlaboratory\nmanaged\nand\noperated\nby\nNational\nTechnology\n&\nEngineering\nSolutions\nof\nSandia,\nLLC,\na\nwholly\nowned\n 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\nCover page photos: \u00a9 Delovely Pics/Shutterstock - \u00a9 Delovely Pics/Shutterstock - \u00a9 Pyty /ShutterstockDISCLAIMER\nThis report was prepared by the Proliferation Resistance and Physical Protection \nWorking Group (PRPPWG) and the Very-High-Temperature Reactor System Steering \nCommittee of the Generation IV International Forum (GIF). Neither GIF nor any of its \nmembers, nor any GIF member\u2019s national government agency or employee thereof, \nmakes any warranty, express or implied, or assumes any legal liability or responsibility \nfor the accuracy, completeness or usefulness of any information, apparatus, product, or \nprocess disclosed, or represents that its use would not infringe privately owned rights.", "characters": 1454, "tokens": 374}
{"id": "2022_GIF_VHTR.pdf_36", "source": "2022_GIF_VHTR.pdf", "chunk": "for the accuracy, completeness or usefulness of any information, apparatus, product, or \nprocess disclosed, or represents that its use would not infringe privately owned rights. \nReferences herein to any specific commercial product, process or service by trade \nname, trademark, manufacturer, or otherwise, does not necessarily constitute or imply \nits endorsement, recommendation, or favoring by GIF or its members, or any agency \nof a GIF member\u2019s national government. The views and opinions of authors expressed \ntherein do not necessarily state or reflect those of GIF or its members, or any agency \nof a GIF member\u2019s national government.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\niPreface to the 2021-2022 edition of the SSCs, pSSCs & PRPPWG white papers on \nthe PR&PP features of the six GIF technologies\nThis report is part of a series of six white papers, prepared jointly by the Proliferation Resistance and Physical \nProtection Working Group (PRPPWG) and the six System Steering Committees (SSCs) and provisional \nSystem Steering Committees (pSSCs). This publication is an update to a similar series published in 2011 \npresenting the status of Proliferation Resistance & Physical Protection (PR&PP) characteristics for each of the \nsix systems selected by the Generation IV International Forum (GIF) for further research and development, \nnamely: the Sodium-cooled fast Reactor (SFR), the Very high temperature reactor (VHTR), the gas-cooled fast \nreactor (GFR), the Molten salt reactor (MSR) and the Supercritical water\u2013cooled reactor (SCWR).\nThe Proliferation Resistance and Physical Protection Working Group (PRPPWG) was established by GIF to", "characters": 1683, "tokens": 375}
{"id": "2022_GIF_VHTR.pdf_37", "source": "2022_GIF_VHTR.pdf", "chunk": "The Proliferation Resistance and Physical Protection Working Group (PRPPWG) was established by GIF to \ndevelop, implement and foster the use of an evaluation methodology to assess Generation IV nuclear energy \nsystems with respect to the GIF PR&PP goal, whereby: Generation IV nuclear energy systems will increase \nthe assurance that they are a very unattractive and the least desirable route for diversion or theft of weapons-\nusable materials, and provide increased physical protection against acts of terrorism.\nThe methodology provides designers and policy makers a technology neutral framework and a formal \ncomprehensive approach to evaluate, through measures and metrics, the Proliferation Resistance (PR) and \nPhysical Protection (PP) characteristics of advanced nuclear systems. As such, the application of the \nevaluation methodology offers opportunities to improve the PR and PP robustness of system concepts \nthroughout their development cycle starting from the early design phases according to the PR&PP by design \nphilosophy. The working group released the current version (Revision 6) of the methodology for general \ndistribution in 2011. The methodology has been applied in a number of studies and the PRPPWG maintains a \nbibliography of official reports and publications, applications and related studies in the PR&PP domain.\nIn parallel, the PRPPWG, through a series of workshops, began interaction with the Systems Steering \nCommittees (SSCs) and Provisional Systems Steering Committees (pSSCs) of the six GIF concepts. White \npapers on the PR&PP features of each of the six GIF technologies were developed collaboratively between \nthe PRPPWG and the SSCs/pSSCs according to a common template. The intent was to generate preliminary \ninformation about the PR&PP merits of each system and to recommend directions for optimizing its PR&PP", "characters": 1855, "tokens": 364}
{"id": "2022_GIF_VHTR.pdf_38", "source": "2022_GIF_VHTR.pdf", "chunk": "the PRPPWG and the SSCs/pSSCs according to a common template. The intent was to generate preliminary \ninformation about the PR&PP merits of each system and to recommend directions for optimizing its PR&PP \nperformance. The initial release of the white papers was published by GIF in 2011 as individual chapters in a \ncompendium report.\nIn April 2017, as a result of a consultation with all the GIF SSCs and pSSCs, a joint workshop was organized \nand hosted at OECD-NEA in Paris. During two days of technical discussions, the advancements in the six GIF \ndesigns were presented, the PR&PP evaluation methodology was illustrated together with its case study and \nother applications in national programmes. The need to update the 2011 white papers emerged from the \ndiscussions and was agreed by all parties and officially launched at the PRPPWG meeting held at the EC Joint \nResearch Centre in Ispra (IT) in November 2017.\nThe current update reflects changes in designs, new tracks added, and advancements in designing the six GIF \nsystems with enhanced intrinsic PR&PP features and in a better understating of the PR&PP concepts. The \nupdate uses a revised common template. The template entails elements of the PR&PP evaluation methodology \nand allows a systematic discussion of the systems elements of the proposed design concepts, the potential \nproliferation and physical protection targets, and the response of the concepts to threats posed by a national \nactor (diversion & misuse, breakout and replication of the technology in clandestine facilities), or by a \nsubnational/terrorist group (theft of material or sabotage).\nThe SSCs and pSSC representatives were invited to attend PRPPWG meetings, where progress on the white", "characters": 1728, "tokens": 363}
{"id": "2022_GIF_VHTR.pdf_39", "source": "2022_GIF_VHTR.pdf", "chunk": "subnational/terrorist group (theft of material or sabotage).\nThe SSCs and pSSC representatives were invited to attend PRPPWG meetings, where progress on the white \npapers was discussed in dedicated sessions. A session with all the SSCs and pSSCs was organized in Paris \nin October 2018 on the sideline of the GIF 2018 Symposium. A drafting and reviewing meeting on all the papers \nwas held at Brookhaven National Laboratory in Upton, NY (US) in November 2019, followed by a virtual \nmeeting in December 2020 to discuss all six drafts.\nIndividual white papers, after endorsement by both the PRPPWG and the responsible SSC/pSSC, are \ntransmitted to the Expert Group (EG) and Policy Group (PG) of GIF for approval and publication as a GIF \ndocument. Cross-cutting PR&PP aspects that transcend all six GIF systems are also being updated and will \nbe published as a companion report to the six white papers.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\niiAbstract\nThis white paper represents the status of Proliferation Resistance and Physical Protection (PR&PP) \ncharacteristics for the Very-High-Temperature Reactor (VHTR) reference designs selected by the Generation \nIV International Forum (GIF) VHTR System Steering Committee (SSC). The intent is to generate preliminary \ninformation about the PR&PP features of the VHTR reactor technology and to provide insights for optimizing \ntheir PR&PP performance for the benefit of VHTR system designers. It updates the VHTR analysis published \nin the 2011 report \u201cProliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy \nSystems\u201d, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG)", "characters": 1726, "tokens": 381}
{"id": "2022_GIF_VHTR.pdf_40", "source": "2022_GIF_VHTR.pdf", "chunk": "in the 2011 report \u201cProliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy \nSystems\u201d, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) \nand the System Steering Committees and provisional System Steering Committees of the Generation IV \nInternational Forum, taking into account the evolution of both the systems, the GIF R&D activities, and an \nincreased understanding of the PR&PP features. \nThe white paper, prepared jointly by the GIF PRPPWG and the GIF VHTR SSC, follows the high-level paradigm \nof the GIF PR&PP Evaluation Methodology to investigate the key points of PR&PP features extracted from the \nreference designs of VHTRs under consideration in various countries. A major update from the 2011 report is \nan explicit distinction between prismatic block-type VHTRs and pebble-bed VHTRs. The white paper also \nprovides an overview of the TRISO fuel and fuel cycle. For PR, the document analyses and discusses the \nproliferation resistance aspects in terms of robustness against State-based threats associated with diversion \nof materials, misuse of facilities, breakout scenarios, and production in clandestine facilities. Similarly, for PP, \nthe document discusses the robustness against theft of material and sabotage by non-State actors. The \ndocument follows a common template adopted by all the white papers in the updated series.\nList of Authors\nTomooki Shiba PRPPWG Japan Atomic Energy Agency\nKiyonobu Yamashita ABC Nuclear\nKeiichiro Hori PRPPWG Japan Atomic Energy Agency\nLap Cheng PRPPWG Brookhaven National Laboratory\nBenjamin Cipiti PRPPWG Sandia National Laboratory\nMichael F\u00fctterer VHTR SSC\nHans Gougar VHTR SSC\nGerhard Strydom VHTR SSC Idaho National Laboratory", "characters": 1765, "tokens": 384}
{"id": "2022_GIF_VHTR.pdf_41", "source": "2022_GIF_VHTR.pdf", "chunk": "Benjamin Cipiti PRPPWG Sandia National Laboratory\nMichael F\u00fctterer VHTR SSC\nHans Gougar VHTR SSC\nGerhard Strydom VHTR SSC Idaho National Laboratory\nChristial Pohl\nAbderrafi Ougouag\nHideyuki Sato VHTR SSC Japan Atomic Energy Agency\nAcknowledgements\nThe current document updates and builds upon the 2011 VHTR PR&PP White Paper. Thanks are due to the \noriginal author of the 2011 SFR PR&PP White Paper, David Moses. The in depth reviews by Giacomo G.M. \nCojazzi (PRPPWG, European Commission Joint Research Centre) and Kevin Hesketh (National Nuclear \nLaboratory) are particularly appreciated. A special thanks to the PRPPWG Technical Secretary Gina \nAbdelsalam (OECD-NEA) who ably readied the final manuscript for publication. SNL is managed and operated \nby NTESS under DOE NNSA contract DE-NA0003525.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\niiiTable of contents\n1. Overview of Technology..........................................................................................................................1\n1.1. Description of the prismatic VHTR .....................................................................................................1\n1.2. Description of the pebble bed VHTR..................................................................................................5\n1.3. Current system design parameters and development status.............................................................7\n2. Overview of Fuel Cycle(s)........................................................................................................................8\n3. PR&PP Relevant System Elements and Potential Adversary Targets ..............................................10\n3.1. System elements related to fuel fabrication site for B-VHTR and P-VHTR......................................11\n3.1.1. Fresh Fuel fabrication...............................................................................................................11\n3.1.2. Fresh Fuel shipment.................................................................................................................12", "characters": 2109, "tokens": 386}
{"id": "2022_GIF_VHTR.pdf_42", "source": "2022_GIF_VHTR.pdf", "chunk": "3.1.1. Fresh Fuel fabrication...............................................................................................................11\n3.1.2. Fresh Fuel shipment.................................................................................................................12\n3.1.3. Fresh Fuel receiving.................................................................................................................13\n3.2. System elements related to reactor site of type B-VHTR.................................................................13\n3.3. System elements related to reactor site of P-VHTR.........................................................................16\n3.4. System elements related to reprocessing site or final disposal site of spent fuel for B-VHTR and P-\nVHTR 18\n3.5. Diversion targets ..............................................................................................................................19\n4. Proliferation Resistance Considerations Incorporated into Design..................................................22\n4.1. Concealed diversion or production of material .................................................................................23\n4.1.1. Diversion of unirradiated nuclear material items ......................................................................23\n4.1.2. Diversion of irradiated nuclear material items ..........................................................................23\n4.1.3. Undeclared production of nuclear material...............................................................................23\n4.2. Breakout ...........................................................................................................................................24\n4.2.1. Diversion of existing nuclear material.......................................................................................24\n4.2.2. Production of the necessary weapons usable nuclear material ...............................................25\n4.3. Pu Production in clandestine facilities ..............................................................................................25\n5. Physical Protection Considerations Incorporated into Design .........................................................26\n5.1. Theft of material for nuclear explosives............................................................................................26\n5.2. Radiological sabotage ......................................................................................................................26\n6. PR&PP Issues, Concerns and Benefits................................................................................................28\n7. References ..............................................................................................................................................29\nAPPENDIX 1: VHTR Major Design Parameters ...........................................................................................31", "characters": 2986, "tokens": 364}
{"id": "2022_GIF_VHTR.pdf_43", "source": "2022_GIF_VHTR.pdf", "chunk": "6. PR&PP Issues, Concerns and Benefits................................................................................................28\n7. References ..............................................................................................................................................29\nAPPENDIX 1: VHTR Major Design Parameters ...........................................................................................31\nAPPENDIX 2: Summary of PR relevant intrinsic design features.............................................................35Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\nivList of Figures\nFigure 1: Illustration of Coated Particle Fuel in the Prismatic Fuel design ........................................................3\nFigure 2: GT-MHR Reactor, Cross-Duct and PCU Vessels...................................................................................4\nFigure 3: GT-MHR Fully-Embedded Reactor Building ........................................................................................4\nFigure 4: Illustration of Coated Particle Fuel in Pebble Fuel Element ................................................................6\nFigure 5: X-Energy Xe-100 .................................................................................................................................7\nFigure 6: 250 MWt HTR-PM Reactor Building Elevated above Ground Level with Steam Generator; Spent \nFuel Storage Not Shown ....................................................................................................................................7\nFigure 7: B-VHTR and P-VHTR as well as their fuel elements...........................................................................10\nFigure 8: B-VHTR System element ...................................................................................................................11\nFigure 9: P-VHTR System element....................................................................................................................11\nFigure 10: Fuel kernel fabrication through dropping uranyl nitrate stock solution ........................................12\nFigure 11: Movement of fuel blocks in reactor site of B-VHTR .......................................................................14\nFigure 12: Door valve and refueling machine ................................................................................................. 15\nFigure 13: Neutron detectors and gamma ray detectors installed in the door valve for B-VHTR ...................16\nFigure 14: Movement of the fuel pebbles........................................................................................................ 17\nFigure 15: Plutonium build-up in a PBMR fuel element in an equilibrium core ..............................................20", "characters": 2828, "tokens": 391}
{"id": "2022_GIF_VHTR.pdf_44", "source": "2022_GIF_VHTR.pdf", "chunk": "Figure 14: Movement of the fuel pebbles........................................................................................................ 17\nFigure 15: Plutonium build-up in a PBMR fuel element in an equilibrium core ..............................................20\nFigure 16: Material Balance Areas and Key Measurement Points of B-VHTR and P-VHTR..............................22\nList of Tables\nTable 1: Calculated plutonium isotopic fractions for PBMR spent fuel as a function of initial enrichment and \ndischarge burn-up ...........................................................................................................................................20\nTable 2: Summary ............................................................................................................................................21Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\nvList of Acronyms  \nCNEC China Nuclear Engineering & Construction Group\nC/S Containment/Surveillance\nDIV Design Information Verification\nGA General Atomics \nGIF Generation-IV International Forum\nGT-MHR Gas-Turbine Modular Helium Reactor \nHALEU High-Assay Low-Enriched Uranium\nHEU Highly Enriched Uranium\nHTR High Temperature Reactor\nHTR-PM High-Temperature Gas-cooled Reactor Pebble-Bed Module\nHTR-TN High-Temperature Reactor-Technology Network\nIAEA International Atomic Energy Agency\nINET Tsinghua University's Institute of Nuclear and New Energy Technology\nJAEA Japan Atomic Energy Agency\nKAERI Korea Atomic Energy Research Institute \nKI Kurchatov Institute\nLEU Low Enriched Uranium\nLWR Light Water Reactor\nMOX Mixed Oxide\nNHDD Nuclear Hydrogen Development and Demonstration\nNNSA National Nuclear Security Administration \nOKBM Experimental Design Bureau of Mechanical Engineering in Nizhniy-Novgorod\nPBMR Pebble Bed Modular Reactor\nPP Physical Protection\nPR Proliferation Resistance\nPR&PP Proliferation Resistance & Physical Protection\nPWR Pressurized Water Reactor", "characters": 1968, "tokens": 392}
{"id": "2022_GIF_VHTR.pdf_45", "source": "2022_GIF_VHTR.pdf", "chunk": "PBMR Pebble Bed Modular Reactor\nPP Physical Protection\nPR Proliferation Resistance\nPR&PP Proliferation Resistance & Physical Protection\nPWR Pressurized Water Reactor\nRCCS Reactor Cavity Cooling System\nRDD Radiological Dispersion Device\nSC-HTGR Steam Cycle High-Temperature Gas-Cooled Reactor\nSSC System Steering Committee\nTRISO Tri-Isotopic \nUOX Uranium Oxide\nVHTR Very-High-Temperature ReactorVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\nvi(This page has been intentionally left blank)Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n11. Overview of Technology\nThe Very High Temperature Reactor (VHTR) design descriptions, technology overviews and \ndiscussions of issues, concerns and benefits documented in this White Paper establish the \nbases to support, as the designs evolve, more detailed assessments of proliferation resistance \nand physical protection (PR&PP).\nThe assessments will be made using the methodology developed for evaluating PR&PP of the \nGeneration IV reactors [1] with consideration of related reports [2-4]. In April 2017, as a result \nof a consultation with all the GIF SSCs and pSSCs a joint workshop was organized and hosted \nat OECD-NEA in Paris. The need to update the 2011 white papers [2] emerged from the \ndiscussions and was agreed by all parties and officially launched in November 2017. \nTherefore, this white paper was written, based on the status of the six GIF system design \nconcepts, considering the designs\u2019 evolution in the last decade.\nVarious versions of the VHTR are under development in several countries that are members", "characters": 1615, "tokens": 369}
{"id": "2022_GIF_VHTR.pdf_46", "source": "2022_GIF_VHTR.pdf", "chunk": "concepts, considering the designs\u2019 evolution in the last decade.\nVarious versions of the VHTR are under development in several countries that are members \nof the Generation IV International Forum (GIF), including the People\u2019s Republic of China, \nFrance, Japan, the Russian Federation, Republic of South Africa, Republic of Korea, Canada, \nUnited Kingdom and the United States of America. The VHTR is a helium-cooled, graphite-\nmoderated, graphite-reflected, metallic-vessel reactor that can use various power conversion \ncycles for electricity production. Co-generation of process steam and high-temperature \nprocess heat for chemical process and hydrogen co-production are additional uses for the \ntechnology. The major VHTR design options that potentially affect PR&PP can be categorized \nas follows:\n\uf0b7Prismatic versus pebble fuel\n\uf0b7Direct versus indirect power conversion cycles\n\uf0b7Water versus air cooled Reactor Cavity Cooling System (RCCS)\n\uf0b7Filtered confinement versus low leakage containment\n\uf0b7Underground versus above-ground nuclear islands\nThe two VHTR basic design concepts considered here are the Prismatic VHTR and the Pebble \nBed VHTR. Note that a lot of the information described in this section was taken from reference \n[5].\n1.1. Description of the prismatic VHTR\nThe safety basis for all the VHTR is to design the reactor to be passively safe, thereby avoiding \nthe release of fission products under all conditions of normal operation and accidents including \nmost of the beyond design basis events. This passive safety aspect of the design should make \nthe VHTR less vulnerable to a significant risk of \"radiological sabotage\" through malevolent \nacts. \nThere are currently five concepts for the prismatic VHTR under consideration by different GIF", "characters": 1761, "tokens": 361}
{"id": "2022_GIF_VHTR.pdf_47", "source": "2022_GIF_VHTR.pdf", "chunk": "the VHTR less vulnerable to a significant risk of \"radiological sabotage\" through malevolent \nacts. \nThere are currently five concepts for the prismatic VHTR under consideration by different GIF \ncountries. The first two of the following have the generic features of low-enriched uranium \n(LEU) and plutonium-fuelled block-type cores and are sufficiently developed to be considered \nfurther here as examples for PR&PP assessment. Except for the second concept discussed \nbelow, prismatic VHTRs are being designed assuming the initial use of a once-through LEU \nfuel cycle.\nUnited States \u2013 Work on the Modular HTGR began with General Atomics (GA) in the 1980s. \nThe GA concepts include prismatic cores driving either a direct or indirect cycle, an air-cooled \nRCCS, filtered confinement, and either a steam cycle (350 MWt MHTGR) or a 600 MWt gas \nturbine cycle (GT-MHR) [6-8]. The MHTGR was the subject of a pre-application design review \nby the Nuclear Regulatory Commission. GA has ceased development and design efforts but \nFramatome (USA), formerly Areva USA, is pursuing a similar concept in the 625 MWt SC-Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n2HTGR. The completion of design and licensing of the SC-HTGR is projected to take at least \n10 years. Framatome has also completed some work on a higher temperature HTGR \n(designated ANTARES) [9, 10], which began as a collaboration in France with other \nEURATOM participants in the High Temperature Reactor-Technology Network (HTR-TN). The", "characters": 1521, "tokens": 363}
{"id": "2022_GIF_VHTR.pdf_48", "source": "2022_GIF_VHTR.pdf", "chunk": "(designated ANTARES) [9, 10], which began as a collaboration in France with other \nEURATOM participants in the High Temperature Reactor-Technology Network (HTR-TN). The \nANTARES Modular HTR is also envisioned to be a 600 MWt cogeneration plant; however, the \nschedule for completion of research and development depends on end-user engagement.  \nSmaller (<80 MWt) prismatic concepts are being pursued by the UltraSafe Nuclear and \nStarCore Nuclear companies, mainly for off-grid communities and mines in Canada.\nRussian Federation \u2013 In cooperation with GA and the U.S. Department of Energy (DOE) \nNational Nuclear Security Administration (NNSA), the Experimental Design Bureau of \nMechanical Engineering (OKBM) in Nizhniy-Novgorod with partners at the Kurchatov Institute \n(KI) and the A.A. Bochvar All-Russian Scientific Research Institute for Inorganic Materials \n(VNIINM) in Moscow is designing a Russian version of the GA GT-MHR to disposition excess \nweapon-grade plutonium; however, OKBM is also analyzing alternative fuel cycles for the \nRussian GT-MHR [11]. The deployment of the Russian GT-MHR is subject to DOE/NNSA joint \nfunding to complete necessary research and development.\nJapan \u2013 The Japan Atomic Energy Agency (JAEA) continues development work that started \nunder the former Japan Atomic Energy Research Institute (JAERI) on the Gas Turbine High \nTemperature Reactor 300 for Cogeneration (GTHTR300C) [12], which will scale up the \ntechnology from the JAEA 30 MWt High Temperature Engineering Test Reactor (HTTR) into \na 600 MWt configuration. The reactor design is based on a prismatic core and can achieve a", "characters": 1625, "tokens": 375}
{"id": "2022_GIF_VHTR.pdf_49", "source": "2022_GIF_VHTR.pdf", "chunk": "technology from the JAEA 30 MWt High Temperature Engineering Test Reactor (HTTR) into \na 600 MWt configuration. The reactor design is based on a prismatic core and can achieve a \nreactor outlet temperature of 950\u00b0C. \nRepublic of Korea \u2013 The Korea Atomic Energy Research Institute (KAERI) is pursuing the \nNuclear Hydrogen Development and Demonstration (NHDD) Project; the NHDD reactor is to \nbe limited to 200 MWt (based on the maximum reactor vessel diameter, 6.5 m, that can be \nfabricated in-country) with no decision yet made as to fuel/core type (pebble bed or prismatic) \n[13].\nUnited Kingdom \u2013 U-Battery Limited is holding the U-Battery project; the U-Battery reactor is \nto be categorized as small modular reactor with 20 MWt with prismatic core design. The \nstrategic goal is to have a first-of-a-kind U-Battery operating by 2028.\nTechnology summaries can be found for each vendor-proposed design option in the respective \nreferences provided above. SC-HTGR and ANTARES are proposed to be constructed as \nmodules to be built in sets of four or more modules per site. As indicated above, the baseline \nfuel design for the first modules uses LEU as Tri-Isotropic (TRISO)-coated particle fuel in a \nonce-through fuel cycle. The Russian version of the General Atomics GT-MHR will incorporate \nexcess weapon plutonium in TRISO-coated fuel particles with the addition of erbium containing \n167Er to provide a neutron poison with a thermal neutron capture resonance to guarantee a \nnegative moderator temperature reactivity coefficient.\n Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper", "characters": 1610, "tokens": 369}
{"id": "2022_GIF_VHTR.pdf_50", "source": "2022_GIF_VHTR.pdf", "chunk": "167Er to provide a neutron poison with a thermal neutron capture resonance to guarantee a \nnegative moderator temperature reactivity coefficient.\n Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n3Figure 1: Illustration of Coated Particle Fuel in the Prismatic Fuel design [14]\nThe TRISO-coated particle fuel (see Figure 1) has a small-diameter (nominally 200-500 \u03bcm) \nspherical ceramic fuel kernel of either uranium oxide or uranium oxycarbide, or mixed oxides \nof other actinides. The kernel is coated with four coating layers consisting sequentially of low-\ndensity porous pyrocarbon (buffer), an inner high density pyrocarbon (IPyC), silicon carbide \n(SiC)1 and an outer high density pyrocarbon (OPyC) for better contact with the matrix material \nwhich is generally carbon but could also be SiC. The first three coatings on the fuel particles \nserve as the primary containment preventing the release of fission products. Plant \nconfigurations and operating conditions are being designed appropriately to limit fuel \ntemperatures during both normal operations and accident conditions so as to preclude the \nrelease of fission products. The coated particles are loaded into fuel compacts (sticks) held \ntogether by graphitized carbon or silicon carbide. The fuel compacts are loaded into holes in \nhexagonal prismatic block fuel elements. Fuel elements are stacked in the reactor core with \nfissile and neutron burnable poison loadings tailored so that the power distribution is peaked \ntoward the top of the core where the inlet cooling gas has the lowest temperature. The power \ndensity is lowest in the bottom of the core where the temperature of the outlet coolant is", "characters": 1695, "tokens": 364}
{"id": "2022_GIF_VHTR.pdf_51", "source": "2022_GIF_VHTR.pdf", "chunk": "toward the top of the core where the inlet cooling gas has the lowest temperature. The power \ndensity is lowest in the bottom of the core where the temperature of the outlet coolant is \nhighest. The fuel and burnable poison loading patterns are specified so that the peak fuel \ntemperature will be below the limit for normal operation, which is 1250\u00baC for TRISO-coated \nfuel particles with SiC coatings and more than 1600 \u00baC in accident conditions.\nSpent fuel is retained in cooled storage containers that are embedded underground and \nlocated adjacent to the reactor cavity. Prismatic spent fuel, which is unloaded from the core \nduring periodic refueling shutdowns, can be tracked remotely by cameras viewing the serial \nnumbers on the fuel elements during handling and storage operations. Since each fuel element \nis loaded with less than 4 kg of LEU, the plutonium content at full burnup (~120 GWD/MT) will \nbe small (~60-70 g) and isotopically degraded compared to weapon-grade plutonium.\nThe current concepts for the energy utilization from the prismatic VHTRs are based on:\n\uf0b7direct Brayton cycle for electricity generation, \n\uf0b7indirect steam generation for process heat and/or electricity generation,\n1On-going research focuses on replacing SiC coatings with zirconium carbide (ZrC) coatings to achieve higher \ntemperature limits (~2000\u00baC) for fission product retention during accidents and to reduce diffusion of radioactive-\nsilver.Uranium Oxide or Uranium OxycarbidePorous Carbon BufferSilicon Carbide or Zirconium CarbidePyrolytic Carbon\nPARTICLE\nSCOMPACTS FUEL ELEMENTSTRISO Coated fuel particles (left) are formed \ninto fuel rods (center) and inserted into", "characters": 1668, "tokens": 369}
{"id": "2022_GIF_VHTR.pdf_52", "source": "2022_GIF_VHTR.pdf", "chunk": "PARTICLE\nSCOMPACTS FUEL ELEMENTSTRISO Coated fuel particles (left) are formed \ninto fuel rods (center) and inserted into \ngraphite fuel elements (right).Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n4\uf0b7indirect heat transfer to process heat user (e.g., Hydrogen production). \nThe vessel configuration for the direct cycle GT-MHR is illustrated in Figure 2, and the reactor \nbuilding option for the GT-MHR is illustrated in Figure 3.  Although the GT-MHR is no longer \nunder development, the plant layout for the Framatome SC-HTGR is very similar.\nFigure 2: GT-MHR Reactor, Cross-Duct and PCU Vessels [2]\nFigure 3: GT-MHR Fully-Embedded Reactor Building [2]\nPower Conversion Unit (PCU)\nReactor VesselVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n5In many modular VHTRs under development, the reactor vessel and power conversion unit are placed \nunderground, which enhances physical protection for the plant.\n1.2. Description of the pebble bed VHTR\nAll modern pebble bed VHTR concepts trace their design features to the HTR Module 200 \nMWt concept developed in Germany in the 1980s. There is currently one national program for \na pebble bed VHTR and one commercial endeavor in the United States. \nSouth Africa \u2013 PBMR Pty. Ltd. is a public-private partnership established in 1999 in response \nto threats of nation-wide power outages in South Africa and to initiate the development of a \nmodular pebble-bed reactor (PBMR) with a rated capacity of 165 MWe. This design featured", "characters": 1521, "tokens": 374}
{"id": "2022_GIF_VHTR.pdf_53", "source": "2022_GIF_VHTR.pdf", "chunk": "to threats of nation-wide power outages in South Africa and to initiate the development of a \nmodular pebble-bed reactor (PBMR) with a rated capacity of 165 MWe. This design featured \na thermal power of 400 MWth and a direct power conversion with a gas turbine operating with \na helium outlet temperature of 900 \u00baC. Due to funding issues and problems in the interaction \nbetween PBMR and the South African regulator the project was stopped in 2010. However, a \nnumber of research organizations cooperate internationally on the VHTR with a longer-term \nview as it requires new materials and design codes along with fuel qualification for the higher \ntemperatures.\nUnited States \u2013 The 200 MWt Xe-100 is a concept under development by the X-Energy \ncompany with some support from the US Government [15-17]. It features a recirculating pebble \nbed core driving a steam cycle. Formal conceptual design activities have started, and X-Energy \nis also pursuing TRISO fuel manufacturing capability with Centrus. X-Energy is pursuing \ndeployment of the first commercial reactor by 2030.\nPeople\u2019s Republic of China (PRC) \u2013 The China Huaneng Group in a consortium with the \nChina Nuclear Engineering & Construction Group (CNEC) and Tsinghua University's Institute \nof Nuclear and New Energy Technology (INET) has been developing and preparing near-term \n(starting in 2010, commissioning completed in 2021) construction of the 2 x 250 MWt, steam-\ncycle High-Temperature Reactor-Pebble-bed Module (HTR-PM) [18, 19]; the HTR-PM, which \nbuilds on the success of the Tsinghua University's HTR-10 test reactor [20], is being", "characters": 1606, "tokens": 374}
{"id": "2022_GIF_VHTR.pdf_54", "source": "2022_GIF_VHTR.pdf", "chunk": "builds on the success of the Tsinghua University's HTR-10 test reactor [20], is being \nconstructed in two module units producing 500 MWt and 210 MWe. Each power plant \ncomprises two reactor modules with individual steam generators sharing a single turbo-\ngenerator. A 6-module, 600 MWt generating station is undergoing design.  The 6-module plant \nis sized to fit into a reactor building roughly that of a large PWR.\nThe pebble bed reactors share the same passive safety features as the prismatic VHTRs but \nhave less excess reactivity due to on-line refueling. The LEU fuel for the pebble bed VHTRs is \nTRISO-coated particles compacted into tennis ball size spheres, as illustrated in Figure 4.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n6Figure 4: Illustration of Coated Particle Fuel in Pebble Fuel Element [2]\nThe pebble fuel is usually not tracked individually by serial number as in the prismatic core, \nbut elements are counted, characterized, and checked following each of multiple re-\ncirculations until they achieve the target burnup based on radioactivity measurements. \nFollowing several passes of each pebble through the core during on-line pebble recirculation, \nwhen measured pebble activity indicates sufficient burnup, the pebble is transferred to a \nstorage container with a record kept of the number of pebbles transferred. Once pebble spent \nfuel is in the storage container, radiation monitoring is used to quantify by inference the amount \nof spent fuel present since, with no more than 0.12 grams of plutonium per pebble, it would \ntake several tens of thousands of pebbles (or several metric tons by total mass and cubic", "characters": 1671, "tokens": 374}
{"id": "2022_GIF_VHTR.pdf_55", "source": "2022_GIF_VHTR.pdf", "chunk": "of spent fuel present since, with no more than 0.12 grams of plutonium per pebble, it would \ntake several tens of thousands of pebbles (or several metric tons by total mass and cubic \nmeters by volume) to be diverted to constitute the basis for recovering a significant quantity of \nplutonium. Further, at a burnup around 90 GWD/MT for the HTR-PM or 150 GWD/MTMT for \nthe Xe-100, the plutonium isotopic composition in the pebble spent fuel is degraded \nsignificantly compared with that of weapon-grade plutonium.\nThe reactor vessel arrangement for the Xe-100 concept is illustrated in Figure 5, showing the \nassociated spent fuel storage location to the right of the reactor vessel. The reactor vessel and \nvessel arrangement for the 250 MW-thermal steam-cycle PRC HTR-PM are illustrated in \nFigure 6, with the steam generator below and to the left of the reactor vessel.\nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n7Figure 5: X-Energy Xe-100 [20]Figure 6: 250 MWt HTR-PM Reactor Building \nElevated above Ground Level with Steam \nGenerator; Spent Fuel Storage Not Shown [2]\n1.3. Current system design parameters and development status\nThe key design parameters for each concept (both prismatic and pebble bed) are presented \nin Appendix VHTR.A. The construction of HTR-PM had started in 2012, and commissioning \nwill continue into 2021 with subsequent connection to the grid. All other concepts require \nfurther development and are at least ten years in the future.", "characters": 1492, "tokens": 365}
{"id": "2022_GIF_VHTR.pdf_56", "source": "2022_GIF_VHTR.pdf", "chunk": "will continue into 2021 with subsequent connection to the grid. All other concepts require \nfurther development and are at least ten years in the future.\nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n82. Overview of Fuel Cycle(s)\nA comparison of the vendor-proposed VHTR fuel cycle parameters is provided in Appendix \nVHTR.B. The information in Appendix VHTR.B is taken either from the references given in \nSection 1 or is inferred from these references where no specific information has been provided \nby the vendors.\nThe baseline fuel cycle for the first generation VHTR is the once-through fuel cycle using LEU \nfuel enriched to between 8 and 16% U-235. The Russian Federation is simultaneously \npursuing the GT-MHR as a \u201cdeep-burn\u201d option for weapon-grade plutonium (Pu) disposition. \nThe use of highly enriched uranium (HEU) as HTGR fuel, as was done in the past, is no longer \nacceptable by many nation states because exporting Special Nuclear Material (SNM), or fissile \nproduction technology, is considered a controlled export.  However, this policy position is not \nuniversally held by all states.  The same is true of separated plutonium, even when considering \na deep-burn fuel cycle as the one currently being considered by the Russian Federation.  Some \nregulatory authorities allow for separated plutonium whereas others do not due to their own \ndomestic policy, export control regulations, or both.  Additionally, under the regulatory \nframework of some states, the HEU and separated Pu require heightened safeguards and \nsecurity measures, compared to LEU, which incurs added complexity and cost to the fuel cycle.\nX-Energy is considering a range of other fuel cycle options for future reactor deployments", "characters": 1746, "tokens": 374}
{"id": "2022_GIF_VHTR.pdf_57", "source": "2022_GIF_VHTR.pdf", "chunk": "security measures, compared to LEU, which incurs added complexity and cost to the fuel cycle.\nX-Energy is considering a range of other fuel cycle options for future reactor deployments \nincluding plutonium disposition and transuranic elements (TRU)/MA transmutation and the use \nof thorium (Th-232) as a fertile component for high-conversion fuel. Each of these options, \nincluding the so-called deep-burn options, is currently based on an initial once-through \nirradiation without recycle, although technologies to reprocess and recycle TRISO fuel are also \nunder consideration or initial development and were studied extensively in the past at \nlaboratory and pilot scale for HEU/Th fuels. The ongoing research and development and the \nhistoric experience provide a reasonably sound basis to have confidence in the ability to close \nthe VHTR fuel cycle in the future, if needed. Note that those alternative fuel cycles are a task \nin the GIF VHTR Fuel and Fuel Cycle Project.\nThe fuel cycle options for VHTRs can be categorized in three ways described below.\nFirst, VHTRs can operate with either pebble or prismatic fuels. Pebble bed reactors operate \nwith on-line refueling. This enables operation with very low excess reactivity and without \nburnable neutron poison, typically only sufficient to overcome the neutron poisoning effects of \nxenon that occur following power reductions. Prismatic fueled reactors require periodic \nrefueling outages and thus operate with substantially higher average excess reactivity \ncompensated by burnable neutron poison, but allow substantially greater flexibility in fuel \nzoning and shuffling.\nSecond, VHTR fuel cycles can be categorized by the types of fuel particles used, as follows:\n\uf0b7LEU fuel particles with or without natural uranium fertile fuel particles.\n\uf0b7Pu fuel particles.", "characters": 1823, "tokens": 370}
{"id": "2022_GIF_VHTR.pdf_58", "source": "2022_GIF_VHTR.pdf", "chunk": "Second, VHTR fuel cycles can be categorized by the types of fuel particles used, as follows:\n\uf0b7LEU fuel particles with or without natural uranium fertile fuel particles.\n\uf0b7Pu fuel particles.\n\uf0b7TRU or MA fuel particles.\n\uf0b7U-233 fuel particles (or U-233 with U-238).\n\uf0b7Thorium (or thorium with uranium) fertile fuel particles.\n\uf0b7Pu/Th-232 and/or Pu/U-238 in mixed oxides (MOX).\nThe first four types of particles contain fissile isotopes that are required to support criticality of \nthe reactor. The LEU particles also contain the fertile isotope U-238 and in some designs may \ncontain fertile particles of natural uranium. However, with the VHTR\u2019s thermal spectrum, \nthorium has somewhat better properties as a fertile isotope, so, for core designs that add fertile \nmaterial, thorium fuel particles may replace the use of natural uranium in the future. This \nthorium may be mixed with a small amount of uranium to dilute and \u201cdenature\u201d the fissile U-\n233 produced by neutron absorption in thorium. In general, it can be expected that future VHTR Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n9reactors will operate with fuels composed of some mix of the six particle types listed above. \nEach particle type involves specific technical issues for fabrication, with some being more \nchallenging than others. \nThird, VHTR fuel cycles can be categorized by whether or not the spent fuel is discarded or \nrecycled. Recycle may occur with either aqueous or pyroprocessing methods, and recycled \nmaterials may be returned to VHTRs or LWRsLWR or sent to fast reactors.  Either method", "characters": 1593, "tokens": 373}
{"id": "2022_GIF_VHTR.pdf_59", "source": "2022_GIF_VHTR.pdf", "chunk": "recycled. Recycle may occur with either aqueous or pyroprocessing methods, and recycled \nmaterials may be returned to VHTRs or LWRsLWR or sent to fast reactors.  Either method \nwould require a \u2018head-end\u2019 process to de-consolidate the coated particles from the graphite \nand \u2018crack\u2019 the silicon carbide coating so that the heavy metal kernel can be leached. possible \nbut has not been demonstrated on a commercial scale\nExcept for the LEU once-through cycle and the historic testing and use of HEU/Th, all other \nfuel cycles for the VHTR represent future possibilities given also that there is likely to be a \nrequirement for several years of effort and a significant financial investment for supporting \nresearch (including irradiation testing of laboratory-scale, pilot-scale and industrial-scale \nfabrications of candidate fuels) to qualify the fuel forms for the alternative fuel cycles. Currently, \nonly LEU fuel is being tested for qualification, so alternative fuel options are likely years away \nin development. Regarding the reprocessing of VHTR fuels, the PUREX process can be \napplied with specific head end processes to separate the fuel particles from the graphite matrix \nand fuel kernels from the coatings, which becomes a strong PR advantage. The process yields \nlarge quantities of 14C-contaminated CO 2 or carbon sludge that must be treated, conditioned, \nand disposed safely. Note that the reprocessing technology for irradiated Thorium fuel \n(THOREX process, similar to the PUREX process) is ready for application, but its \ndemonstration at an industrial level has not been carried out yet.\nThe challenges of realizing such fuel cycles at the commercial level have become major R&D \ntopics internationally, and many efforts are ongoing. For one of those examples, see the", "characters": 1790, "tokens": 373}
{"id": "2022_GIF_VHTR.pdf_60", "source": "2022_GIF_VHTR.pdf", "chunk": "The challenges of realizing such fuel cycles at the commercial level have become major R&D \ntopics internationally, and many efforts are ongoing. For one of those examples, see the \nreference [22]. In addition, the waste graphite and SiC can be decontaminated to reduce waste \nvolume. Studies on the subject are ongoing in several countries.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n103. PR&PP Relevant System Elements and Potential Adversary Targets\nAlthough the shape of the fuel is different for the block type very high temperature gas reactor \n(B-VHTR) and pebble bed type very high temperature gas reactor (P-VHTR), their safeguards \nfeatures and the physical protection features have some similarities because the fuel is made \nfrom a mixture of coated fuel particles with graphite powder that is sintered. Figure 7 shows \nsketches of reactors of the B-VHTR and P-VHTR types and their respective fuel elements. \nFigure 7: B-VHTR and P-VHTR as well as their fuel elements\nIn order to retrieve a significant quantity of nuclear material from used VHTR fuels, it is \nnecessary process metric tons and tens of cubic meter quantities of carbon-encased nuclear \nfuel using either grind-leach, burn-leach of electrolysis in nitric acid, the technology for which \nis still not matured to industrial level. The cost of removing and storing the large volume of \nseparated graphite should be considered a proliferation resistance feature.  Such large \nquantities are a necessity to retrieve weapons usable fissile material and would be difficult to \nconceal by a proliferating state.  \nThe use of LEU is currently planned in both B-VHTR and P-VHTR due to its low", "characters": 1687, "tokens": 371}
{"id": "2022_GIF_VHTR.pdf_61", "source": "2022_GIF_VHTR.pdf", "chunk": "conceal by a proliferating state.  \nThe use of LEU is currently planned in both B-VHTR and P-VHTR due to its low \nproliferation characteristics.  For states that own their own domestic enrichment capability, the \nraw LEU material for fresh fuel fabrication is more attractive than the fabricated graphite fuel \nforms (block or pebble since a lower level of effort would be required for its diversion or \nacquisition from the system elements at fuel fabrication sites or product side of reprocessing \nsites etc. For states that import the as-fabricated graphite fuels, the attractiveness may be \nconsidered similar between the fresh and spent fuels.  This is because a similar amount of \neffort is required to crack the SiC barrier as discussed previously.  \nIt is noteworthy from a security standpoint, IFCIRC/225 (the IAEA Standard on nuclear \nsecurity) allows some credit for radioactive source term regarding the degree of physical \nprotection.  However, once a Category II (i.e., U-235/U<20%) fuel has decayed sufficiently, the \nsecurity threat and categorization are the same between fresh and used fuel. The Standard \nalso prescribes an elevated security posture for High Assay Low Enriched Uranium (HALEU), \n10 wt.% \u2264 U-235/U < 20 wt.%.  For example, it specifies that HALEU be stored in the facility\u2019s \nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n11protected area, as opposed to the limited access area.  It also calls out the need for increased \ncommunication and verification for transport.  Similarly, it elevates the importance of armed \nguards (i.e., a dedicated security organization) during transport and storage at facilities.", "characters": 1669, "tokens": 372}
{"id": "2022_GIF_VHTR.pdf_62", "source": "2022_GIF_VHTR.pdf", "chunk": "communication and verification for transport.  Similarly, it elevates the importance of armed \nguards (i.e., a dedicated security organization) during transport and storage at facilities.  \nThe \"system elements\" for B-VHTR and P-VHTR are shown in Figure 8 and Figure 9, \nrespectively.\nFigure 8: B-VHTR System element\nFigure 9: P-VHTR System element\nThe system elements of the both VHTR types are principally the same except for the unloading \nand reloading of fuel blocks of the B-VHTR and the recirculating fuel spheres of the P-VHTR. \nThe common system elements for both VHTRs are discussed in the following.\nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n123.1. System elements related to fuel fabrication site for B-VHTR and P-VHTR\n3.1.1. Fresh Fuel fabrication\nThe raw constituents of fresh fuel (Uranium hexafluoride, nitrate, or oxide of LEU, LEU/Pu \n(MOX), LEU/Th or Pu / Th(MOX)) are brought into the fuel fabrication facility. Fuel elements \n(fuel compacts for block type fuel or fuel spheres) containing TRISO-coated fuel particles \nsintered with graphite powder are manufactured and shipped out to reactor sites. LEU is \ncurrently intended for use in B-VHTR and P-VHTR due to its lower proliferation risk, specifically \nwith respect to material attractiveness. Fuel based on LEU / Th, LEU/Pu (MOX) or Pu / Th may \nbe used in future VHTRs. Raw material for fresh fuel fabrication is the most attractive target", "characters": 1444, "tokens": 362}
{"id": "2022_GIF_VHTR.pdf_63", "source": "2022_GIF_VHTR.pdf", "chunk": "with respect to material attractiveness. Fuel based on LEU / Th, LEU/Pu (MOX) or Pu / Th may \nbe used in future VHTRs. Raw material for fresh fuel fabrication is the most attractive target \nover the entire set of system elements of B-VHTR and P-VHTR, from fuel fabrication to final \ndisposal, since it would require the least effort to divert and use for fabrication of NEDs (hence \nit will require more attention and protection). However, it should be noted that the material type \nwill be the same if present in the fuel fabrication facility or in the fresh fuel in terms of the IAEA \nsafeguards target material. In any case the material will require further processing for use in a \nNED unless it is already in suitable form. See the discussion of the section 2 of the reference \n[23]. It should be also noted that safeguarding bulk material is more complicated than items.\nThe fuel kernels of the coated fuel particles are manufactured by dropping uranyl nitrate stock \nsolution into ammonia water as shown in Figure 10. \nFigure 10: Fuel kernel fabrication through dropping uranyl nitrate stock solution [24]\nImplementation of adequate measures of Containment and Surveillance (C/S) and physical \nprotection needs to be enforced over those raw constituents of fresh fuel according to the \ngrade of nuclear material such as LEU, LEU/Th, LEU/Pu, and Pu / Th. \nEvery fuel block of B-VHTR is stamped with identification numbers (IDs). On the other hand, \nthere is no ID on fuel pebbles of P-VHTR, which requires a different safeguards approach as \nB-VHTR (item-based safeguards can be applied for B-VHTR). In contrast, quasi-bulk type", "characters": 1634, "tokens": 375}
{"id": "2022_GIF_VHTR.pdf_64", "source": "2022_GIF_VHTR.pdf", "chunk": "there is no ID on fuel pebbles of P-VHTR, which requires a different safeguards approach as \nB-VHTR (item-based safeguards can be applied for B-VHTR). In contrast, quasi-bulk type \nsafeguards are needed for P-VHTR. In the past, however, there have been cases where \nsafeguards were implemented by assigning IDs to pebbles at the research reactor level, but \nnot for online monitoring during the re-loading procedure. As one of the ongoing efforts, see \nthe reference [25]. Fabrication also involves scrap recovery and recycling within the supplier's \nfuel fabrication facility. Non-recoverable scrap materials are stored for disposition as low-level \nradioactive waste. The isotopes U-235, U-233 and Pu are attractive for adversaries aiming for \nmanufacturing NEDs. However, once these nuclear materials are encased in graphitized \ncarbon as the kernel of coated fuel particles of fuel elements of both B-VHTR and P-VHTR, \ntheir use in NEDs poses major difficulties for an adversary. The separation of the kernel from \ncoated fuel particles is difficult due to the stable chemical and mechanical characteristics of \ncarbon and SiC layers. Techniques such as grind-leach or burn-leach of electrolysis in nitric \nacid are necessary, but they have not yet been matured to industrial level. Also, in order to \nacquire significant amounts of nuclear materials, metric tons and tens of cubic meter quantities \nof carbon and SiC layers from the coated fuel particles and the graphite matrix surrounding \nthem must be processed. \nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n133.1.2. Fresh Fuel shipment", "characters": 1621, "tokens": 363}
{"id": "2022_GIF_VHTR.pdf_65", "source": "2022_GIF_VHTR.pdf", "chunk": "them must be processed. \nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n133.1.2. Fresh Fuel shipment \nFuel rods for B-VHTR and fuel pebbles for P-VHTR are put into containers and shipped from \nfuel fabrication facilities to reactor sites. Adequate C/S system such as sealing and PP need \nto be applied to containers to ensure continuity of knowledge according the sensitivity grade \nof the nuclear material being shipped. Note that there are no current domestic or internationally \nlicensed shipping container for transporting large quantities of HALEU fuels.\n3.1.3. Fresh Fuel receiving  \nBroken fresh fuel elements should be segregated and must be stored separately by the user \nfor shipment back to the supplier for recycling as un-irradiated scrap. The C/S system for fresh \nfuel shipment must be confirmed upon fresh fuel receiving. The nuclear material in the broken \nfresh fuel elements is not attractive because the amounts are small and the material is still in \nthe form of coated fuel particles. \n3.2. System elements related to reactor site of type B-VHTR\nPR of B-VHTR is based on item accountancy. It is possible to imprint an ID on each fuel block, \nso the safeguards approach has many similarities with the safeguards of LWRs. All system \nelements related to a reactor site are confined within the reactor building as shown in Figure \n11 [26]. All movements of fuel can be monitored by the surveillance cameras. Fuel storage \nracks of the fresh fuel storage and spent fuel storage areas are sealed after handling fuel \ntherein. Fuel inventory in the reactor core is verified by measuring the fuel flow with detectors \nin the door valve. Movement of the fuel handling machine is slow due to its mass of more than", "characters": 1750, "tokens": 370}
{"id": "2022_GIF_VHTR.pdf_66", "source": "2022_GIF_VHTR.pdf", "chunk": "therein. Fuel inventory in the reactor core is verified by measuring the fuel flow with detectors \nin the door valve. Movement of the fuel handling machine is slow due to its mass of more than \n100 tons. This movement can be followed by the surveillance cameras whose data should be \ncontinuously transferred to mitigate potential Cyber-attacks. \n3.2.1. Fresh fuel storage on site \nFuel blocks are assembled by inserting fuel rods into pre-formed holes in the graphite blocks \nin the reactor building. The on-site movement of fuel blocks of the B-VHTR is shown in Figure \n11. The fuel blocks are stored in the fresh fuel storage rack until such time as the blocks \nscheduled for reloading are returned to the reactor core. An adequate C/S system such as \nsurveillance cameras and PP should be applied to the fresh fuel storage area, the refueling \nmachine, and the spent fuel storage area for continuity of knowledge. Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n14Figure 11: Movement of fuel blocks in reactor site of B-VHTR [26]\n3.2.2. Refueling Machine for fresh fuel loading and spent fuel discharging\nThis paragraph refers to HTTR as this is considered fully representative of B-VHTR [27].\nStandpipeVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n15The fresh fuel blocks are taken into the refueling \nmachine from the fresh fuel storage, and then the \nrefueling machine is lifted and moved onto the door \nvalve over the reactor with the crane. The fresh fuel \nblocks are loaded into the vertical empty space from \nwhere the spent fuels have been taken out. The IDs \nof fuel blocks are confirmed at time of loading of fresh", "characters": 1676, "tokens": 369}
{"id": "2022_GIF_VHTR.pdf_67", "source": "2022_GIF_VHTR.pdf", "chunk": "blocks are loaded into the vertical empty space from \nwhere the spent fuels have been taken out. The IDs \nof fuel blocks are confirmed at time of loading of fresh \nfuel. The spent fuel blocks in the reactor are taken into \nthe revolver-rack of the refueling machine and moved \nto a spent fuel storage facility by the crane before the \nfresh fuels are loaded. The control rod driving device \nand the pair of control rods must be removed before \nrefueling. Replaceable side reflectors and fuel blocks \nare handled using the refueling machine. They are \npassed through the door valve and the stand pipe at \nthe upper part of the reactor core for any refueling. \nFuel reloading in light water reactors (LWRs) is \nperformed in water that provides a radiation shielding \neffect. However, the coolant of B-VHTR is helium and \nhas no radiation shielding effect. For this reason, the \nfuel exchange for B-VHTR is performed by remote \ncontrol of the gripper of the refueling machine, since \nthe fuel cannot be directly viewed. It is also necessary \nto incorporate a radiation shielding function in the \nrefueling machine because it will contain the spent \nfuel block in the revolver-rack. For this reason, its \nmass exceeds 100 tons. When the refueling machine \nis moved from the upper part of the reactor, the coolant (helium) in the reactor should not be \nallowed to leak. A door valve is provided between the refueling machine and the standpipe to \nprevent leakage of the coolant (helium) in the reactor to outside. The position of door valve is \nshown in Figure 12 [27]. Neutron detectors and gamma ray detectors are attached to the door", "characters": 1631, "tokens": 357}
{"id": "2022_GIF_VHTR.pdf_68", "source": "2022_GIF_VHTR.pdf", "chunk": "prevent leakage of the coolant (helium) in the reactor to outside. The position of door valve is \nshown in Figure 12 [27]. Neutron detectors and gamma ray detectors are attached to the door \nvalve, since the door valve is necessary to move out core components (anything such as spent \nfuel blocks, replaceable side reflectors and irradiated experimental material from the reactor).\n3.2.3. Reactor Core \nThe core consists of hexagonal columns of fuel blocks, control rod guide blocks and \nsurrounding replaceable side reflector, constituted of blocks. The permanent reflectors \nsurround the replaceable side reflectors. Fuel blocks are stacked vertically in several stages, \nand replaceable reflectors are placed above and below them. In order to accommodate the \ndecrease in reactivity associated with fuel depletion as the reactor is operated, by design the \nreactor core is loaded with adequate excess reactivity at the beginning of operation. Each fuel \nblock is engraved with a unique ID and loaded to a predetermined position in the reactor core. \nAfter a certain period of operation, the spent fuel block is taken out through the stand pipe \nusing the refueling machine. The coolant flows through the flow paths in the graphite blocks \nand is heated. The heated coolant is brought into a hot plenum and guided to outside of the \nreactor pressure vessel at a temperature of 700 to 950 \u00b0C. \nThe control rods are suspended from the control rod drive mechanism in standpipes above the \ncore and inserted into the core or reflector, as needed. Control rod guide columns for inserting \ncontrol rods are provided in the core. \nAny undeclared movement of the refueling machine would be detected by surveillance \ncameras. Furthermore, irradiation of undeclared material is detectable with the neutron and", "characters": 1801, "tokens": 373}
{"id": "2022_GIF_VHTR.pdf_69", "source": "2022_GIF_VHTR.pdf", "chunk": "control rods are provided in the core. \nAny undeclared movement of the refueling machine would be detected by surveillance \ncameras. Furthermore, irradiation of undeclared material is detectable with the neutron and \ngamma ray detectors attached in the door valve used for introducing and removing materials \ninto and from the core. The combination of neutron and gamma ray detectors, shown in Figure \nFigure 12: Door valve and refueling \nmachine [27]Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n1613 [26] makes it possible to distinguish the nature of materials introduced into the core or \nremoved from it as nuclear materials and non-nuclear materials. Data obtained by both \ndetectors should be continuously transferred to safeguards inspectorates to avoid Cyber-\nattacks or other tampering. \n3.2.4. Spent fuel storage on site \nThe spent fuel blocks are stored for a certain period in racks of the spent fuel storage facility \nthat includes a water-cooling system in order to remove decay heat. The movement of spent \nfuel blocks can be detected by an adequate C/S system such as sealing the lid on the top of \nthe storage racks and monitoring them with further surveillance cameras.\n                                                       \nFigure 13: Neutron detectors and gamma ray detectors installed in the door valve for B-VHTR [26]\n3.2.5. On-site radioactive waste storage \nSubstances that do not contain nuclear fuel materials, such as activation products, are stored \nin the on-site radioactive waste storage facility, so their attractiveness from the PR viewpoint \nis low. However, such materials should be protected from a PP viewpoint. \n3.2.6. On-site radioactive waste storage \nThe spent fuel blocks in storage are put in fuel transfer casks for shipping to the final disposal", "characters": 1817, "tokens": 371}
{"id": "2022_GIF_VHTR.pdf_70", "source": "2022_GIF_VHTR.pdf", "chunk": "3.2.6. On-site radioactive waste storage \nThe spent fuel blocks in storage are put in fuel transfer casks for shipping to the final disposal \nor to the reprocessing plant after cooling for a certain period in the spent fuel storage on site. \nContinuity of Knowledge (CoK) is maintained by use of adequate C/S systems, such as sealing \ntransfer casks, and adequate PP is also applied, such as protection by guards. The spent fuel \nblocks are not attractive as sources of explosive nuclear materials used for NED due to the \npoor quality of the materials and the great difficulty of reprocessing. But they may be attractive \nfrom the view point of \u201cradiological sabotage\" due to their high radioactivity content.\n3.3. System elements related to reactor site of P-VHTR\nFor safeguards purposes, P-VHTR is regarded as a quasi-bulk type facility. In the past, \nhowever, there have been cases where safeguards were implemented by assigning IDs to \npebbles at the research reactor level, but not for online monitoring during the re-loading \nprocedure. However, it is usually sufficient for safeguards to just count/keep track of the \nnumber of fresh fuel and spent fuel pebbles as they are moved from and to their respective \nstorage systems. The operating temperatures and high pressure of the system would make it \ndifficult to divert fuel out of the core.\n3.3.1. Fresh fuel storage on site \n IAEAVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n17The containers with fuel pebbles are stored in the fresh fuel storage under an adequate C/S \nsystem and PP for P-VHTR. These fuel pebbles are moved to the charging room to be loaded", "characters": 1645, "tokens": 368}
{"id": "2022_GIF_VHTR.pdf_71", "source": "2022_GIF_VHTR.pdf", "chunk": "17The containers with fuel pebbles are stored in the fresh fuel storage under an adequate C/S \nsystem and PP for P-VHTR. These fuel pebbles are moved to the charging room to be loaded \ninto the reactor core. The number of fuel pebbles should be counted if it is possible, and the \nmovement of the fuel pebbles from the fresh fuel storage to the charging room should be \nobserved via surveillance cameras. Diversion or otherwise acquisition of fuel pebbles is not \nattractive due to the difficulty of recovering the nuclear material from fuel elements and \nbecause the amount of nuclear material in them is small.\n3.3.2. Recirculation of irradiated fuel pebbles \nThe fuel pebbles have no identification \nnumbers and are loaded randomly into \nthe reactor core. The amount of nuclear \nmaterial in every fresh fuel pebble is the \nsame (heavy metal loading and \nuranium enrichment level). If initially \nfueled entirely with fresh fuel pebbles, \nP-VHTR cores would become critical \nwith a small total volume of fuel. \nTherefore, graphite balls and boron \nballs containing no fuel are loaded into \nthe core along with the fresh fuel in \norder to maintain the desired height of \nfuel in the core. With fuel depletion, \ngraphite balls and boron balls are \nremoved, and fresh fuel pebbles are \nloaded in, as the core evolves from the \ninitial loading core to the equilibrium \ncore. Figure 14 shows the movement of \nthe fuel pebbles in the reactor [28]. Fuel \npebbles are taken out from the core \nthrough the fuel pebble discharging \ntube. Failed fuel pebbles are separated", "characters": 1562, "tokens": 359}
{"id": "2022_GIF_VHTR.pdf_72", "source": "2022_GIF_VHTR.pdf", "chunk": "the fuel pebbles in the reactor [28]. Fuel \npebbles are taken out from the core \nthrough the fuel pebble discharging \ntube. Failed fuel pebbles are separated \nand are stored in the scrap containers. \nSound fuel pebbles are led to the \ndosing wheel where their fuel burnup \nlevels are measured. The fuel burnup is \nevaluated by measuring the Cesium-\n137 gamma ray peak with a gamma \nspectrometer. However, it has recently \nbeen suggested that Cs-137 would not necessarily be a good burnup indicator, and Zr-95, Nb-\n95, and La-140 may provide more appropriate burnup instead [29]. Further research is needed. \nThe fuel pebbles that have achieved a predetermined burnup level are discharged through the \ndischarge tube and are led to containers in the discharge compartment as spent fuel pebbles. \nOn the other hand, fuel pebbles that have not reached the predetermined burnup level are \ntransported pneumatically to the upper part of the core and reloaded at the top of the core. \nThis reloading is repeated until the fuel pebble reaches the predetermined burnup level. The \nnumber of reloading cycles is typically between 5 to 15. The precise figure depends on the \nspecific design, reloading pattern and target burnup levels. High fuel burnup is achievable due \nto the highly stable characteristics of coated fuel particles and due to nearly continuous fuel \nloading. It is higher than the burnup of LWRs as well as B-VHTR. A burnup level of 100 GWd/T \nis achievable for the spent fuel of P-VHTR, and it results in superior proliferation resistance \nfeatures due to large isotopic fraction of high content in plutonium that produces a high level", "characters": 1645, "tokens": 373}
{"id": "2022_GIF_VHTR.pdf_73", "source": "2022_GIF_VHTR.pdf", "chunk": "is achievable for the spent fuel of P-VHTR, and it results in superior proliferation resistance \nfeatures due to large isotopic fraction of high content in plutonium that produces a high level \nof decay heat. The physical inventory verification in the reactor core is performed by controlling \nthe number of fresh fuel pebbles loaded and accounting for the spent fuel pebbles discharged \nFigure 14: Movement of the fuel pebbles\nfor P-VHTR [28]Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n18and the number of failed fuel pebbles discharged to the scrap container. Access to the reactor \ncell will be controlled by an adequate C/S system and PP.\n3.3.3. Spent fuel storage on site \nThe spent fuel pebbles in containers are stored for a certain period in the on-site spent fuel \nstorage. The containers are cooled in order to remove decay heat. The movement of a \ncontainer can be observed using an adequate C/S system, such as sealing the containers and \nmonitoring the storage area with surveillance cameras. The amount of fissile nuclear material \n(U-235 and Pu-239) in the spent fuel pebbles is small due to high burnup and high content of \ndecay heat-generating Pu isotopes. One of interesting discussions is the treatment of \ndamaged pebbles. In general, the damaged pebbles are added to the spent fuel storage, i.e. \nthere is no separate waste storage of broken pebbles planned for the PBMR design. Damaged \npebbles are always to be expected to occur during irradiation in the reactor and cannot be \nreturned for further cycles through the core, so they had to be classified as spent fuel. However,", "characters": 1627, "tokens": 367}
{"id": "2022_GIF_VHTR.pdf_74", "source": "2022_GIF_VHTR.pdf", "chunk": "pebbles are always to be expected to occur during irradiation in the reactor and cannot be \nreturned for further cycles through the core, so they had to be classified as spent fuel. However, \nsince those pebbles are less burnt, they are potentially more attractive in terms of Pu quality.\n3.3.4. Radioactive waste storage on site \nSubstances that do not contain nuclear fuel materials, such as activation products, are stored \nhere, so their attractiveness from the PR viewpoint is low. However, these waste materials still \nneed to be protected from a PP viewpoint.\n3.3.5. Spent fuel shipping \nThe spent fuel pebbles in containers will be transferred to the final disposal or to the \nreprocessing plant after cooling for a certain period in the spent fuel storage area on site. COK \nis ensured using an adequate C/S system such as sealing the containers and monitoring the \nmovement of the containers with surveillance cameras. The spent fuel pebbles are not \nattractive from the point of view of nuclear materials for use for NEDs, but they may be \nattractive from the view point of \u201cradiological sabotage\" due to their high radioactivity content. \nSee the section 5.2 for more discussion.\n3.4. System elements related to reprocessing site or final disposal site of spent \nfuel for B-VHTR and P-VHTR\nThe treatment of spent fuel of both B-VHTR and P-VHTR can be divided into (1) direct final \ndisposal and (2) reprocessing. The direct disposal option is attractive because the coatings of \ncoated fuel particles themselves are \u201ccontainers\u201d for the fission products and the fuel itself \npossesses high mechanical and chemical stability. Thus, the direct final disposal of the VHTR", "characters": 1680, "tokens": 372}
{"id": "2022_GIF_VHTR.pdf_75", "source": "2022_GIF_VHTR.pdf", "chunk": "coated fuel particles themselves are \u201ccontainers\u201d for the fission products and the fuel itself \npossesses high mechanical and chemical stability. Thus, the direct final disposal of the VHTR \nfuel has reduced environmental and public impact.  \nFurthermore, the reprocessing of VHTR fuel is not considered attractive. The reason is that \nmetric tons and tens of cubic meter quantities of carbon encasing coated fuel particles would \nhave to be removed using either grind-leach, burn-leach of electrolysis in nitric acid if \nreprocessing were to be performed. However, these technologies have still not been \ndemonstrated at industrial level. For this reason, spent fuel of VHTR has low attractiveness for \ndiversion / acquisition and / or processing as nuclear material. Spent fuels from the VHTR may \npotentially still be attractive for radiological sabotage due to their high content in radioactive \nmaterials that results from their high fuel burnup levels. The physical robustness of VHTR fuel \nis favorable in this respect, making it more difficult for a potential adversary to achieve \nwidespread dispersal. The proliferation resistance features corresponding to the reprocessing \nof the spent fuel of VHTR mentioned-above are valid not only for spent fuels of LEU-fuel, but \nalso for that of LEU / Th, LEU/Pu (MOX), Pu / Th MOX with high burnup. Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n193.5. Diversion targets\nThe key proliferation resistance feature of the VHTR is the fuel itself. The extraction of a \nsignificant quantity (SQ) of either indirect-use U-235 from LEU (75 kg) or direct-use U-233 and \nplutonium (both 8kg) from VHTR fuel will require the processing of metric tons and tens of", "characters": 1728, "tokens": 382}
{"id": "2022_GIF_VHTR.pdf_76", "source": "2022_GIF_VHTR.pdf", "chunk": "plutonium (both 8kg) from VHTR fuel will require the processing of metric tons and tens of \ncubic meter quantities of carbon encasing coated particles using either grind-leach, burn-leach, \nor electrolysis in nitric acid. A background report [14] that supported the compilation of the \noriginal VHTR white paper (published in 2011) discussed diversion targets for the two fuel \nforms, prismatic block and pebble. The following discussion is quoted from the background \nreport [14] with some modifications using the PBMR [16] and the GT-MHR [6-8] as example \nplants for the P-VHTR and the B-VHTR respectively.\n\u201cUsing the PBMR as an example, the diversion of an indirect-use significant quantity (75 \nkilograms) of U-235 in LEU in fresh pebbles would require, for the equilibrium core with a \npebble loading of 9 grams of LEU at 9.6% enrichment, 75,000/(9 * 0.096) = 86,806 pebbles or \n~17.4 MT of fuel pebbles, which should be quite readily detectable even over time since that \nis ~20 percent of a core loading.\u201d \u201cBy comparison, for the prismatic core GT-MHR or MHTGR \nusing fuel elements with inscribed serial numbers for visual tracking, the diversion of an \nindirect-use significant quantity (75 kilograms) of U-235 in LEU in fuel elements containing \n~3.43 kilograms of LEU on average at 19.8% enriched would require 75/(3.43 * 0.198) = ~111 \nfuel elements or 13.5 MT of fuel elements, which would be ~15\u201316% of a GT-MHR core loading", "characters": 1437, "tokens": 381}
{"id": "2022_GIF_VHTR.pdf_77", "source": "2022_GIF_VHTR.pdf", "chunk": "fuel elements or 13.5 MT of fuel elements, which would be ~15\u201316% of a GT-MHR core loading \nor ~17% of the MHTGR core loading.\u201d \u201cThus, the mass ratio for the diversion of indirect-use \nU-235 in LEU between fresh pebbles and fresh GT-MHR fuel elements is 17.4/13.5 = ~1.29 \nso that 29% more pebbles by mass would have to be diverted to obtain 75 kilograms of indirect-\nuse U-235 in LEU.\u201d\n\u201cBecause the fuel elements of PBMRs are quite difficult to track, the use of LEU-fueled PBMRs \nhas been examined by several researchers from the aspect of the attractiveness for diversion \nof fully burned spent fuel, one-cycle-irradiated pebbles, and the use of special production \npebble. \n\u201cThe calculation results for the plutonium isotopic fractions in the PBMR fully burned spent fuel \nwould likely be very close to those for the prismatic VHTR spent fuel where the prismatic fuel \nis to be discharged at a burn-up exceeding 100 GWD/MT (or MWD/kg). The PBMR and \nprismatic VHTR spent fuel will have slightly different plutonium isotopic compositions resulting \nfrom differences in the thermal-neutron and epithermal-neutron energy spectra due to a \ndifferent moderator-to-fissile atom ratio and additional thermal and epithermal neutron self-\nshielding due to the higher-density fuel compacting used in the prismatic fuel.\u201d \u2026..\u201d It is \nexpected, however, that the spent LEU fuel from both the GA GT-MHR and Areva Modular \nHTR will have plutonium isotopic fractions very close to the values calculated for the PBM in", "characters": 1506, "tokens": 372}
{"id": "2022_GIF_VHTR.pdf_78", "source": "2022_GIF_VHTR.pdf", "chunk": "expected, however, that the spent LEU fuel from both the GA GT-MHR and Areva Modular \nHTR will have plutonium isotopic fractions very close to the values calculated for the PBM in \nTable 3.6.1.\u201d It appears that the Pu will be of reactor grade in all cases by applying the fissile \nmaterial type metric of PRPP WG.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n20Table 1: Calculated plutonium isotopic fractions for PBMR spent fuel as a function of initial \nenrichment and discharge burn-up [Table 4.1 from [14]]\n\u201cBecause the PBMR recirculates a pebble up to six times through the core before it is \ndischarged to spent fuel storage at full burn-up (~92 GWD/MT), the question arises about the \ndiversion of an irradiated pebble after one cycle or the use of special pebbles designed as \ntarget elements to produce plutonium.\u201d \u201cThe analysis of the PBMR by PBMR (Pty) Ltd. [16] \nshows in Figure 15 [14] the plutonium build-up per pebble and the relative isotopic content as \na function of recirculation.\u201d\nFigure 15: Plutonium build-up in a PBMR fuel element in an equilibrium core [14]\n\u201cFigure 15 indicates that at full burn-up each pebble will contain about 0.11 grams of plutonium \nwith the isotopics indicated, and, from this, it can be inferred that, at full burn-up (120 GWD/MT \nin the GT-MHR), the prismatic fuel elements can be estimated to contain on the order of 60\u2013\n70 grams of plutonium of similarly degraded isotopics.\u201d The diversion of 1 SQ of direct-use Pu", "characters": 1491, "tokens": 384}
{"id": "2022_GIF_VHTR.pdf_79", "source": "2022_GIF_VHTR.pdf", "chunk": "in the GT-MHR), the prismatic fuel elements can be estimated to contain on the order of 60\u2013\n70 grams of plutonium of similarly degraded isotopics.\u201d The diversion of 1 SQ of direct-use Pu \nfrom pebbles at full burn-up requires 8,000/0.11 = 72,727 pebbles or ~14.4 MT of fuel pebbles. \nIt takes 8,000/65 = 123 prismatic fuel elements to secure 1 SQ of direct-use Pu.\n\u201cHowever, the LEU pebble in a PBMR is recirculated up to six times while the fuel element in \na GT-MHR or MHTGR is typically reloaded only once. From Figure 15, the plutonium content \nof a pebble after its initial irradiation is given as ~0.047 grams (~74% Pu-239), whereas for the \nGT-MHR there are no data quoted for the one-cycle-burned prism, but it is inferred that the \nplutonium loading would be ~50 grams with less favorable isotopics than in the pebble after \nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n21one cycle of irradiation. From this, a rough comparison can be made that it would take at least \n~1050 pebbles diverted after the first cycle to equal the amount of less favorable plutonium in \na prismatic fuel element removed from a GT-MHR after the first irradiation.\u201d Table 3.6-2 shows \na summary table indicating the amount of material needed to collect an SQ.\nTable 2: SummaryDiversion Target U-235 from Fresh LEU Pu from Spent fuel\nSQ 75 kg 8kg", "characters": 1357, "tokens": 367}
{"id": "2022_GIF_VHTR.pdf_80", "source": "2022_GIF_VHTR.pdf", "chunk": "a summary table indicating the amount of material needed to collect an SQ.\nTable 2: SummaryDiversion Target U-235 from Fresh LEU Pu from Spent fuel\nSQ 75 kg 8kg\nEquivalent pebbles 86806 (17.4 MT) 72727 (14.4 MT)\nEquivalent blocks 111 (13.5 MT) 123 (15.0 MT)Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n224. Proliferation Resistance Considerations Incorporated into Design\nThe fuel in reactor cores of B-VHTR and P-VHTR is not as accessible and visible as the fuel \nin an LWR. Therefore, physical inventory verification of nuclear materials in the reactor cores \nis carried out by measurement of fuel flows into and from the core. Major Material Balance \nAreas and Key Measurement Points are shown in Figure 16 as an example. Also, adequate \nC/S is necessary. Adequate counter-measures against cyberattacks are required to maintain \nCoK by C/S. \nFigure 16: Material Balance Areas and Key Measurement Points of B-VHTR and P-VHTR\nDesign Information Verification (DIV) and C/S are implemented to avoid concealment of fuel. \nDirect transfer of the C/S signal to IAEA is recommended to enhance proliferation resistance. \nAs noted previously, the key proliferation resistance feature of the VHTR is the fuel itself. To \nobtain a significant quantity of either indirect-use U-235 from LEU or direct-use plutonium, one \nmust process metric tons and tens of cubic meter quantities of carbon encasing fuel using \neither grind-leach or burn-leach of electrolysis in nitric acid. \nThe high burnup of the spent fuel of the VHTRs is also a key proliferation resistance feature", "characters": 1587, "tokens": 376}
{"id": "2022_GIF_VHTR.pdf_81", "source": "2022_GIF_VHTR.pdf", "chunk": "either grind-leach or burn-leach of electrolysis in nitric acid. \nThe high burnup of the spent fuel of the VHTRs is also a key proliferation resistance feature \ndue to the high isotopic fraction of even plutonium isotopes generating large amounts of decay \nheat and high dose rate. However, it is controversial.\nHistorically, it has been argued that the technical difficulty of fabricating nuclear weapons \ndepends on the isotopic composition of plutonium, in particular the amount of Pu-240. Although \nthere are several references, the one that summarizes the key points is by Pellaud [30].\nFor nuclear safeguards verification activities there is no distinction for Pu with less than 80% \nPu-238. However, the heat generated by Pu isotopic containing more than a few percent of \nPu-238 would substantially increase the technical difficulty related to the fabrication phase \n(weaponization). Using a set of figures of merit (FOM) for attractiveness, Bathke, et al. [31] \nestimated that about 8% Pu-238 is required to render the plutonium isotopic unattractive for \nan unadvanced proliferant state that requires reliably high-yield nuclear devices, however it \nremains attractive for both technologically advanced states, which can handle it, and \nsubnational groups for which high reliability might not be a requirement. However, these \narguments are founded on the assumption that the proliferants demand reliable yield. In the \ncase of unadvanced proliferant or non-state actors who do not pay attention to the yield, high \nreliability might not be their requirement.\nWith those reasons considered, the current GIF PRPP WG methodology adopts weapon \nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper", "characters": 1722, "tokens": 364}
{"id": "2022_GIF_VHTR.pdf_82", "source": "2022_GIF_VHTR.pdf", "chunk": "reliability might not be their requirement.\nWith those reasons considered, the current GIF PRPP WG methodology adopts weapon \nVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n23grade, reactor grade, and deep-burn grade for Pu categorization [1]. The fact that Pu in \nHTGRs\u2019 spent fuel can achieve deep-burn is one of the notable features.\n4.1. Concealed diversion or production of material\nDiversion of large quantities of nuclear materials (U-235, plutonium or U-233) is detectable by \nspent fuel accountancy based on radiation monitoring or fuel element counting, by C/S on fuel \nstorage, or by recorded reactivity deviations in reactor operations. The VHTR does not produce \nreadily accessible, attractive fissile material. The technologies for reprocessing coated fuel \nparticles are complicated and still require development. \n4.1.1. Diversion of unirradiated nuclear material items \nOnce the fuel has been encased within fuel kernel of coated fuel particle and furthermore into \nfuel elements (such as fuel compacts for B-VHTR or fuel pebbles for P-VHTR), diversion \nbecomes difficult. The latter (fuel elements) consist of coated fuel particles encased within \ngraphitized carbon. Note that fresh fuel fabrication should be performed under surveillance. \nOnce in fuel assembly (compact ball) form, the nuclear material is more difficult to retrieve due \nto difficulty of separation of nuclear material from large amounts of graphite and of the strength \nof the coatings of particles. Fabricated fresh fuel can be stored under C/S measures for B-\nVHTR and P-VHTR. The theft during transportation of fresh fuel can be detected by the C/S.", "characters": 1665, "tokens": 363}
{"id": "2022_GIF_VHTR.pdf_83", "source": "2022_GIF_VHTR.pdf", "chunk": "of the coatings of particles. Fabricated fresh fuel can be stored under C/S measures for B-\nVHTR and P-VHTR. The theft during transportation of fresh fuel can be detected by the C/S. \nThe raw constituents are observed under the same C/S applied for fuel fabrication of LWR. \n4.1.2. Diversion of irradiated nuclear material items \n4.1.2.1. B-VHTR \nThe major irradiated nuclear material items are spent fuel blocks. Diversion of Pu is possible \nby discharging fuel blocks after a short time of reactor operation and then reprocessing them. \nThe fuel blocks are unloaded through standpipes over the reactor pressure vessel. For such \nan operation, both the refueling machine and the door valve are required because it is \nnecessary to maintain isolation between the reactor coolant and the outside atmosphere for \nthe B-VHTR. Unexplained or illicit movements of the refueling machine and the door valve by \nthe crane can be detected by surveillance cameras. Also, undeclared movements of \nprematurely discharged fuel blocks are detected by the neutron detector and the gamma ray \ndetector in the door valve. Any discharged material from the reactor pressure vessel can be \nidentified as nuclear material or not. Nuclear material is indicated when signals of both neutron \nand gamma ray are detected. If the material is non-nuclear, such as a surveillance sample, \nthen no neutron source is detected. Undeclared discharging of experimental nuclear materials \nis detected in the same manner as fuel blocks. \n4.1.2.2.  P-VHTR\nDiversion of Pu may be possible using the continuous fuel loading feature through early \ndischarging of fuel pebbles from the reactor core before even-mass-number Pu isotopes are", "characters": 1698, "tokens": 368}
{"id": "2022_GIF_VHTR.pdf_84", "source": "2022_GIF_VHTR.pdf", "chunk": "Diversion of Pu may be possible using the continuous fuel loading feature through early \ndischarging of fuel pebbles from the reactor core before even-mass-number Pu isotopes are \naccumulated. However, this would be detected by the burnup measuring detectors. \nFurthermore, it is technically difficult because the reprocessing process of VHTR fuel is still not \nestablished and detection of this diversion route is possible if an appropriate C/S system is in \nplace.\n4.1.3. Undeclared production of nuclear material \n4.1.3.1. B-VHTR Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n24Undeclared production of nuclear material may be possible through the irradiation of fertile \nnuclear material in irradiation holes in the core or replaceable side reflectors of B-VHTR. The \nmaterials would be loaded and unloaded through standpipes over the reactor pressure vessel. \nIn the B-VHTR, it is not possible to directly access and to visually observe the fuel in the core \nor the replaceable side reflectors as would be possible in LWRs where water above the core \nis used as radiation shielding. For these reasons, a handling machine with radiation integrated \nshielding function, such as the refueling machine, is necessary for any undeclared production \nof nuclear material. Any unexplained or illicit movement of handling machines can be detected \nby surveillance cameras in the reactor building. Moreover, ton quantities of fertile material \nwould need to be loaded illicitly to generate a SQ, and it is difficult to envisage this being \npractical to achieve without detection. \nIt should be noted that B-VHTR could be used in a mode similar to that of the Magnox reactors", "characters": 1693, "tokens": 364}
{"id": "2022_GIF_VHTR.pdf_85", "source": "2022_GIF_VHTR.pdf", "chunk": "practical to achieve without detection. \nIt should be noted that B-VHTR could be used in a mode similar to that of the Magnox reactors \nfor producing weapon-grade plutonium. In this case, rod-type Magnox fuel containing metal \nuranium would be inserted into some cooling holes of the graphite blocks instead of using \nordinary B-VHTR fuel rods based on coated particle fuel. In this way the difficulty of \nreprocessing of VHTR fuel would be avoided, as the reprocessing methods for Magnox fuel \nare well established. However, the reactors would have to be operated with low reactor coolant \noutlet temperatures to protect the integrity of the Magnox fuel and ton quantities of Magnox-\ntype fuel would need to be irradiated. This would imply giving up efficient power production, \nwhich would be detectable. It might to worth thinking that this mode of operation could be \ndangerous because of accumulation of Wigner energy in the graphite blocks, but further study \nis needed.\n4.1.3.2. P-VHTR\nThe inlet pipes of fresh fuel pebbles, in the fuel charging room, can be used for loading target \npebbles and for the access to the core region of a P-VHTR. However, these pipes cannot be \neasily used for loading illicit material for the undeclared production of nuclear materials due to \nthe length of, and many curves in, the fuel loading path. Pebbles with diameter of 6 cm could \nbe loaded into the pipe. It is very important to confirm in Design Information Verification that \nthere are no access holes into the pipes except at the fresh fuel pebble loading location, \nprecluding any other pipe access into the reactor core. Irradiation of fertile materials covered", "characters": 1663, "tokens": 362}
{"id": "2022_GIF_VHTR.pdf_86", "source": "2022_GIF_VHTR.pdf", "chunk": "there are no access holes into the pipes except at the fresh fuel pebble loading location, \nprecluding any other pipe access into the reactor core. Irradiation of fertile materials covered \nwith graphite or carbon that look like fuel pebbles is possible. But such pseudo-fuel spheres \nmay break during movement through the core and would be difficult to remove. Furthermore, \nsuch pseudo-fuel without ceramic coatings would release unexpected high radioactivity into \nthe primary cooling system at high temperature operation, which would be detectable. In \naddition, there would result many operational problems, which would also be detectable and \nrequire explanation. Tton quantities of heavy metal would need to be irradiated in the core to \ngenerate 1 SQ. Finally, it is important to recognize that the presence of target breeding pebbles \nin the core will alter the balance between fresh fuel demand and energy production in a way \nthat is detectable long before a significant quantity of fissile material is accumulated [32-35]. \nIn addition, the identification of such breeding pebbles by the gamma measurement is much \nmore difficult than for regular fuel.\n4.2. Breakout\nAs mentioned in section 3.4, reprocessing has not yet been demonstrated for the coated fuel \nparticles at industrial scale. In the presence of multi-lateral contractual provisions, for example \nadhering to the guidance of the international Nuclear Suppliers Group (NSG), for the supply of \nfresh fuels and the take-back of spent fuels for an exported VHTR, the issue of breakout is \nfurther mitigated since there will be either no such material, or limited quantities of material, to \nbe reprocessed in the host states. \n4.2.1. Diversion of existing nuclear material Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper", "characters": 1816, "tokens": 380}
{"id": "2022_GIF_VHTR.pdf_87", "source": "2022_GIF_VHTR.pdf", "chunk": "be reprocessed in the host states. \n4.2.1. Diversion of existing nuclear material Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n25As mentioned in Section 3.4, the key proliferation resistance feature is the use of coated fuel \nparticles embedded within a graphite matrix. Therefore, diverting existing nuclear material from \nVHTR fuels is difficult, lengthy and costly, regardless of the implementation of safeguards and \nPP for the fuels. Since the reprocessing technology is not developed to industrial level, \nextraction of nuclear material is significantly difficult. Moreover, the high burnup of spent fuel \nfrom VHTRs is also a key proliferation resistance feature due to the presence of plutonium \nisotopes that produce large amounts of decay heat. Pu in high burnup spent fuel contains \nconsiderable even-numbered Pu isotopes, i.e. Pu-238, 240 and 242, whose decay heat \nnegatively affects use as a NED. Note that the diversion of raw material before being coated \nwith carbon and silicon carbide would be the easiest pathway for the processing of nuclear \nmaterials to be used in the fabrication of nuclear explosive devices. However, this is not a \nVHTR-specific problem, but a concern for all types of nuclear reactor systems.\n4.2.2. Production of the necessary weapons usable nuclear material \nAs mentioned in section 3.4, the key proliferation resistance feature of VHTRs is the use of \ncoated fuel particles embedded within graphite matrix. It is necessary to process metric tons \nand tens of cubic meter quantities of carbon encasing the fuel kernels to obtain the amount of \nnuclear material necessary for production of weapons. \n4.3. Pu Production in clandestine facilities", "characters": 1715, "tokens": 366}
{"id": "2022_GIF_VHTR.pdf_88", "source": "2022_GIF_VHTR.pdf", "chunk": "and tens of cubic meter quantities of carbon encasing the fuel kernels to obtain the amount of \nnuclear material necessary for production of weapons. \n4.3. Pu Production in clandestine facilities\nHigh quality graphite with very low impurity levels is used in the technology of the B-VHTR and \nP-VHTR.  This high quality graphite can be used for gas-cooled reactors in which weapons-\ngrade plutonium can be produced from natural uranium. Therefore, the consumption of large \namounts of nuclear-grade graphite should be controlled. For that reason, nuclear grade \ngraphite is controlled according to NSG lists part 1.\nOperation of the clandestine facilities (reactor and fuel reprocessing) could be detected by \nenvironmental sampling under the international safeguards regimes.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n265. Physical Protection Considerations Incorporated into Design\nThis section provides a qualitative overview discussion of the aspects of VHTR systems and \ntheir design that create potential benefits or problems from the point of view of potential threat \nby sub-national actors.\n5.1. Theft of material for nuclear explosives\nPlutonium in the spent fuel of LEU cycles and U-233 in that of future LEU/Th cycles are \nattractive for the NED production. However, these nuclear materials in spent fuels are \naccompanied with fission products, which are highly radioactive and make it difficult for \nterrorists to steal the materials. Moreover, the nuclear materials are encased inside the coated \nfuel particles. In these coated particles, the material of interest would be quite dilute so that the \ntheft of a significant quantity would require the theft of metric tons of contaminated graphite \nand/or graphitized carbon containing the coated fuel particles. Obtaining access to a significant", "characters": 1837, "tokens": 372}
{"id": "2022_GIF_VHTR.pdf_89", "source": "2022_GIF_VHTR.pdf", "chunk": "theft of a significant quantity would require the theft of metric tons of contaminated graphite \nand/or graphitized carbon containing the coated fuel particles. Obtaining access to a significant \nquantity of plutonium or U-233 in the stolen spent fuels would require substantial effort for \nreprocessing. Furthermore, plutonium with a high inventory of the plutonium isotopes other \nthan Pu-239 is not attractive for the manufacturing of NEDs (e.g. high decay heat). U-233 with \nhundreds of parts per millions of U-232 is not attractive due to high radioactivity and to the \nnecessity of further chemical cleaning to remove radioactive decay products. For those \nreasons, the intrinsic qualities of VHTR spent fuel make it undesirable as a target for theft by \na sub-national group for use as nuclear explosive. \n5.2. Radiological sabotage\nVHTRs are designed such that the fuel temperature is maintained below fuel-damaging \ntemperatures under all conditions of normal operations and accident situations, including \nbeyond-design-basis events.  The design vision is that, even if the safety-related reactor cavity \ncooling system were to malfunction, decay heat in the core would still be removed to the \nexternal wall of the reactor vessel. As a result, the fuel temperatures in the core do not exceed \nthe levels that would cause the loss of the primary containment provided by the SiC coatings \nover the fuel kernels. \nThe ultimate radiological sabotage act for reactors is that of an insider or an intruder trying to \ncause radiological exposure by inducing a large power excursion. For both P-VHTR and B-\nVHTR designs, appropriate physical protection and controls must be in place to prevent such \nacts. These designs have several safety benefits from the very high temperature tolerance of", "characters": 1795, "tokens": 371}
{"id": "2022_GIF_VHTR.pdf_90", "source": "2022_GIF_VHTR.pdf", "chunk": "VHTR designs, appropriate physical protection and controls must be in place to prevent such \nacts. These designs have several safety benefits from the very high temperature tolerance of \nthe fuel and the strong negative temperature power coefficient. \nAnother relevant discussion is that both VHTRs are extremely resilient to this kind of terrorist \nattacks because passive heat removal, or reactor cavity cooling system (RCCS), by air cooling, \nwater or a combination of both is available when a loss of coolant happens.\nThe high burnup levels in the spent fuel of both VHTR types is one of the key proliferation \nresistance features due to high radioactivity. However, spent fuels of VHTRs may be attractive \nfor Radiological Dispersion Device (RDD) due to high radioactivity resulting from the high \nburnup. Below is the discussion of RDD for both P-VHTR and B-VHTR:\n\uf0b7In the case of P-VHTR, the quasi-bulk fuel form may be attractive for terrorists when \nconsidering the possibility of dispersal of the spent fuel. Protection of spent fuel on the \nreactor site will be important. This should also be considered in PP when transporting \nspent fuel by land. \n\uf0b7In the case of B-VHTR, the item-type fuel allows its PP to be similar to that of LWRs. \nMoreover, TRISO is considered to be very resistant to scattering and therefore more \nrobust against RDD-type terrorism than LWRs.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n27Finally, some points to be considered for the PP of VHTR are listed referring to the previous \nVHTR white paper:\n\uf0b7Quality controls at the fuel fabrication plant in the supplier nation.", "characters": 1635, "tokens": 363}
{"id": "2022_GIF_VHTR.pdf_91", "source": "2022_GIF_VHTR.pdf", "chunk": "27Finally, some points to be considered for the PP of VHTR are listed referring to the previous \nVHTR white paper:\n\uf0b7Quality controls at the fuel fabrication plant in the supplier nation. \n\uf0b7Proper maintenance, inspection, and protection of (1) the helium supply and the helium \nsupply station to prevent the introduction of corrosive chemicals, (2) the primary coolant \ncontaminant monitoring equipment to detect the introduction of such chemicals, and \n(3) the helium purification system to remove contaminants. \n\uf0b7Careful maintenance, inspections, testing and protection of reactivity control systems \nto assure the capability to achieve safe hot and cold shutdown and, if required, \naccomplish the same function from a secure remote location. \n\uf0b7Physical protection is required of and controlled access to fresh and spent fuel storage \nlocations, the inbound and outbound transportation loading systems, the transportation \nof the fresh fuel from the fuel fabrication facility, and the spent fuel to its processing or \ndisposal facilities.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n286. PR&PP Issues, Concerns and Benefits\nThe key areas of known strengths of the VHTR concept at this time are its robust fuel form, \nwith fissile material strongly diluted in carbonaceous material, high burnup and the use of the \nonce-through LEU fuel cycle, which all make VHTR fuel unattractive for proliferation purposes. \nWhen considering PR, B-VHTR will have item-based safeguards applied, while P-VHTR \nsafeguards are quasi-bulk, so differing safeguards approaches will be required is relatively \ndifficult.\nRegarding PP, typical reactor site protections on the reactor, control systems, and fresh and", "characters": 1718, "tokens": 361}
{"id": "2022_GIF_VHTR.pdf_92", "source": "2022_GIF_VHTR.pdf", "chunk": "safeguards are quasi-bulk, so differing safeguards approaches will be required is relatively \ndifficult.\nRegarding PP, typical reactor site protections on the reactor, control systems, and fresh and \nspent fuel storage will be required. It can be concluded that VHTRs are extremely resilient to \nterrorist attacks because RCCS is available when a loss of coolant happens.\nFor system designers, program policy makers, and external stakeholders who read this white \npaper are encouraged to evaluate PR&PP features using the GIF PRPP WG methodology \nfrom an early stage of design, and keep updating them as designs progress.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n297. References\n[1] Evaluation Methodology for Proliferation Resistance and Physical Protection of Generation IV \nNuclear Energy Systems, GIF/PRPPWG/2006/005, Revision 6, prepared by the Proliferation \nResistance and Physical Protection Evaluation Methodology Expert Group of the Generation IV \nInternational Forum (GIF), September 15, 2011.\n[2] PRPP Working Group and System Steering Committees of the Generation IV International Forum. \nProliferation Resistance and Physical Protection of the Six Generation IV Energy Systems. \nTechnical Report GIF/PRPPWG/2011/002, Generation IV International Forum (GIF), 2011.\n[3] IAEA, Proliferation Resistance Fundamentals for Future Nuclear Energy Systems, IAEA STR-\n332, IAEA Department of Safeguards, IAEA, Vienna (2002).\n[4] Gen IV International Forum, GIF R&D Outlook for Generation IV Nuclear Energy Systems: 2018 \nUpdate (2019).\n[5] M. A. F\u00fctterer, et al., \"The High Temperature Gas-Cooled Reactor,\" Encyclopedia of Nuclear", "characters": 1662, "tokens": 379}
{"id": "2022_GIF_VHTR.pdf_93", "source": "2022_GIF_VHTR.pdf", "chunk": "Update (2019).\n[5] M. A. F\u00fctterer, et al., \"The High Temperature Gas-Cooled Reactor,\" Encyclopedia of Nuclear \nEnergy, Elsevier, pp. 512-522, 2021, https://doi.org/10.1016/B978-0-12-409548-9.12205-5\n[6] M.B. Richards et al., Part 1 -- H2-MHR Pre-Conceptual Design Report: SI-Based Plant, \nGA-A25401, General Atomics, Idaho National Laboratory, and Texas A&M University, April 2006.\n[7] M.B. Richards et al., Part 2 -- H2-MHR Pre-Conceptual Design Report: HTE-Based Plant, \nGA-A25402, General Atomics, Idaho National Laboratory, and Texas A&M University, April 2006.\n[8] General Atomics, Gas-Turbine Modular Helium Reactor (GT-MHR) Conceptual Design \nDescription Report, GA Document No. 910720, Revision 1, July 1996, transmitted by letter from \nLaurence L. Parme (GA) to Raji Tripathi (USNRC), \"GT-MHR Conceptual Design Description \nReport,\" GA/NRC-337-02, General Atomics, San Diego, CA, August 6, 2002.\n[9] L. Lommers et al., \u201cAREVA HTR Concept for Near-Term Deployment,\u201d Nuclear Engineering and \nDesign, 251, pp. 292-296, October 2012. https://doi.org/10.1016/j.nucengdes.2011.10.030.\n[10] Brochure: ANTARES - The AREVA HTR-VHTR Design,", "characters": 1139, "tokens": 377}
{"id": "2022_GIF_VHTR.pdf_94", "source": "2022_GIF_VHTR.pdf", "chunk": "[10] Brochure: ANTARES - The AREVA HTR-VHTR Design, \nhttps://www.yumpu.com/en/document/read/32557580/antares-the-areva-htr-vhtr-design-smr\n[11] V. Petrunin et al., \"Analysis of questions concerning the nonproliferation of fissile materials for \nlow-and medium-capacity nuclear power systems,\" Atomnaya Energiya 105, Issue 3, pp. 123-\n127, September 2008 (in English. pp. 159-164, Atomic Energy 105, Springer, New York, \nISSN 1063-4258 (Print), 1573-8205 (Online)).\n[12] K. Kunitomi, et al., \"JAEA's VHTR for Hydrogen and Electricity Cogeneration: GTHTR300C,\" \nNuclear Engineering and Technology 39, pp. 9-20, February 2007.\n[13] Chang Keun Jo, Hong Sik Lim, and Jae Man Noh, \"Preconceptual Designs of the 200MWth Prism \nand Pebble-bed Type VHTR Cores,\" PHYSOR-2008, International Conference on the Physics of \nReactors \u201cNuclear Power: A Sustainable Resource,\u201d Casino-Kursaal Conference Center, \nInterlaken, Switzerland, September 14-19, 2008.\n[14] D. Moses, \u201cVery High-Temperature Reactor (VHTR) Proliferation Resistance and Physical\n Protection (PR&PP),\u201d ORNL/TM-2010/163, Oak Ridge National Laboratory, August 2010.\n[15] Presentations by PBMR (Pty) Ltd. to the U.S. Nuclear Regulatory Commission Public Meeting,", "characters": 1213, "tokens": 377}
{"id": "2022_GIF_VHTR.pdf_95", "source": "2022_GIF_VHTR.pdf", "chunk": "[15] Presentations by PBMR (Pty) Ltd. to the U.S. Nuclear Regulatory Commission Public Meeting, \nPBMR Safety and Design Familiarization, February 28-March 3, 2006.\n[16] Johan Slabber, PBMR (Pty) Ltd., \"PBMR Nuclear Material Safeguards,\" Paper No. B14, \nProceedings of the Conference on High Temperature Reactors, Beijing, China, September, 22-\n24, 2004, International Atomic Energy Agency, Vienna (Austria).\n[17]   E. Mulder, MM. van tStaden et al, \u201cThe Coupled Neutronics and Thermo-Fluid Dynamics Design \nCharacteristics of the Xe-100 200 MWth Reactor,\u201d Proceedings of the Conference on High \nTemperature Reactors, Las Vegas, Nevada, USA, November, 7-10, 2018.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper\n30[18] Dong Y (2012) China\u2019s activities in HTGRs HTR-10 and HTR-PM. In: IAEA Course on High \nTemperature Gas Cooled Reactor Technology. Beijing, China, 22-26 October 2012.\n[19] Z. Zhang et al., \u201cThe Shandong Shidao Bay 200 MWe High-Temperature Gas-Cooled Reactor \nPebble-Bed Module (HTR-PM) Demonstration Power Plant: An Engineering and Technological \nInnovation,\u201d Engineering 2, pp. 112-118, March 2016. \nhttp://dx.doi.org/10.1016/J.ENG.2016.01.020", "characters": 1182, "tokens": 362}
{"id": "2022_GIF_VHTR.pdf_96", "source": "2022_GIF_VHTR.pdf", "chunk": "Innovation,\u201d Engineering 2, pp. 112-118, March 2016. \nhttp://dx.doi.org/10.1016/J.ENG.2016.01.020\n[20] Y. Xu and K. Zuo, \"Overview of the 10 MW high temperature gas cooled reactor\u2014test module \nproject,\" Nuclear Engineering and Design 218, pp. 13\u201323, October 2002.\n[21] The image of XE-100 Reactor downloaded from:\nhttps://x-energy.com/media/xe-100\n[22] M. A. F\u00fctterer, F. von der Weid, P. Kilchmann, A High Voltage Head-End Process for Waste \nMinimization and Reprocessing of Coated Particle Fuel for High Temperature Reactors, Proc. \nICAPP\u201910, paper 10219, San Diego, CA, USA, 13-17 June 2010.\n[23] Correct reference for SCWR WP (TBC)\n[24] Nuclear Fuel Industries, Ltd, \u201cHTGR fuel manufacturing process 1. Fuel particle process,\u201d \nhttps://www.nfi.co.jp/product/prod03.html (accessed 2020-7-16).\n[25] K, Juergen et al., \u201cUpgrading (V)HTR fuel elements for generation IV goals by SiC encapsulation,\u201d \nKerntechnik, 77(5), 351-355, 2012\n[26] Kiyonobu Yamashita, Fujio Miyamoto, Sigeaki Nakagawa, and Toshiyuki Tanaka, \"Safeguards \nConcept for the High Temperature Engineering Test Reactor Using Unattended Fuel Flow Monitor \nSystem,\" Journal of Nuclear Materials Management, Volume XXV, Number 4, August 1997.", "characters": 1206, "tokens": 383}
{"id": "2022_GIF_VHTR.pdf_97", "source": "2022_GIF_VHTR.pdf", "chunk": "Concept for the High Temperature Engineering Test Reactor Using Unattended Fuel Flow Monitor \nSystem,\" Journal of Nuclear Materials Management, Volume XXV, Number 4, August 1997.\n[27] S. Saito, et al., \u201cDesign of High Temperature Engineering Test Reactor (HTTR),\u201d JAERI 1332, \n1994, https://doi.org/10.11484/jaeri-1332.\n[28] U. Cleve et al., \u201cThe Technology of High-Temperature-Reactors,\u201d Proceedings of ICAPP 2011, \nPaper 11076, Nice, France, May 2-5, 2011.\n[29] T. Shiba, et al., \u201cProliferation Resistance and Safeguardability of Very High Temperature \nReactor,\u201d IAEA Symposium on International Safeguards, 5-8 November 2018, Vienna \nInternational Centre, Vienna (Austria).\n[30] B. Pellaud, \u201cProliferation Aspects of Plutonium Recycling,\u201d Journal of Nuclear Material \nManagement, XXXI, 1, pp. 30-38, 2002\n[31] C. Bathke, et al., \u201cThe Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various \nProliferation and Theft Scenarios,\u201d Nuclear Technology, Vol 179, Issue 1, 2012\n[32] A.M. Ougouag and H. D. Gougar, \u201cPreliminary Assessment of the Ease of Detection of Attempts \nat Dual Use of a Pebble-Bed Reactor,\u201d Transactions of the Winter 2001 Annual Meeting of ANS, \nReno, NV, Trans. ANS 85, pp. 115-117, Nov. 2001.", "characters": 1228, "tokens": 376}
{"id": "2022_GIF_VHTR.pdf_98", "source": "2022_GIF_VHTR.pdf", "chunk": "Reno, NV, Trans. ANS 85, pp. 115-117, Nov. 2001.\n[33] Ougouag, A. M., S. M. Modro, W. K. Terry, and H. D. Gougar \u201cRational Basis for a Systematic \nIdentification of Critical Components and Safeguard Measures for a Pebble-Bed Reactor\u201d \nTransactions of the Winter 2002 Annual Meeting of ANS, Washington, DC, Trans. ANS 87, pp. \n367-368, Nov. 2002.\n[34] A.M. Ougouag, W. K. Terry and H. D. Gougar, \u201cExamination of the Potential for Diversion \nor5Clandestine Dual Use of a Pebble-Bed Reactor to Produce Plutonium,\u201d Proceedings of HTR \n2002, 1st International Topical Meeting on High temperature Reactor Technology (HTR), Petten, \nNetherlands, April 22-24, 2002.\n[35] A. M. Ougouag, H. D. Gougar, and T.A. Todd, \u201cEvaluation of the Strategic Value of Fully Burnt \nPBMR Spent Fuel,\u201d May 2006, Idaho National Laboratory, INL/EXT-06-11272Very-High-Temperature Reactor (VHTR)          PR&PP White Paper\n31APPENDIX 1: VHTR Major Design Parameters \nAppendix VHTR.A \u2013 VHTR Major Reactor Design Parameters\nMajor Reactor \nParametersFramatome  \nSC-HTGRGeneral \nAtomics   \nGT-MHRX-Energy      \nXe-100Huaneng \nGroup & \nCNEC/INET \nHTR-PMJAEA", "characters": 1122, "tokens": 376}
{"id": "2022_GIF_VHTR.pdf_99", "source": "2022_GIF_VHTR.pdf", "chunk": "SC-HTGRGeneral \nAtomics   \nGT-MHRX-Energy      \nXe-100Huaneng \nGroup & \nCNEC/INET \nHTR-PMJAEA \nGTHTR300COKBM   GT-\nMHRKAERI   \nNHDD\nThermal Power  (MW-th) 625 600 200 250 600 600 200\nThermal Efficiency (%) in \nElectricity Generation~40 ~48 40 (inferred) 40 ~50 ~48 None, H 2 \nproduction\nPrimary Coolant Helium Helium Helium Helium Helium Helium Helium\nModerator High-\nTemperature \nGraphiteHigh-\nTemperature \nGraphiteHigh-\nTemperature \nGraphitized \nCarbon with \nGraphite \nReflectorHigh-\nTemperature \nGraphitized \nCarbon with \nGraphite \nReflectorHigh-\nTemperature \nGraphiteHigh-\nTemperature \nGraphiteHigh-\nTemperature \nGraphite or \nGraphitized \nCarbon with \nReflector\nPower Density (MW/m3) ~6.3 (inferred) 6.3 4.95 (max) ~3.22 5.4 6.3 2.27-3.0 \npebble, 5.68 \nprismatic\nFuel Materials LEUO 2 TRISO-\ncoated \nparticlesUC 0.5O1.5 \nTRISO-\ncoated \nparticles; \nLEUC 0.5O1.5 \n(19.8%) fissile \nand \nUNatC0.5O1.5 \nfertileLEUO 2 TRISO-\ncoated particlesLEUO 2 \nTRISO-", "characters": 953, "tokens": 351}
{"id": "2022_GIF_VHTR.pdf_100", "source": "2022_GIF_VHTR.pdf", "chunk": "(19.8%) fissile \nand \nUNatC0.5O1.5 \nfertileLEUO 2 TRISO-\ncoated particlesLEUO 2 \nTRISO-\ncoated \nparticlesLEUO 2 \nTRISO-coated \nparticlesPuO 1.8 , \nLEUCO  or \nmixed \nuranium-\nplutonium \noxide (MOX)LEUO 2 \nTRISO-\ncoated \nparticlesVery-High-Temperature Reactor (VHTR)          PR&PP White Paper\n32Appendix VHTR.A \u2013 VHTR Major Reactor Design Parameters (Continued)\nMajor Reactor \nParametersFramatome\nSC-HTGRGeneral \nAtomics\nGT-MHRX-Energy           \nXe-100Huaneng \nGroup & \nCNEC/INET \nHTR-PMJAEA \nGTHTR300COKBM   GT-\nMHRKAERI\nNHDD\nCore Inlet \nTemperature/Pressure \n(\u00baC/MPa)325/6.0 490/7.07 260/~6.1 250/~7.0 587-666/6.9 \n(electrical \nproduction) & \n594/5.1 (H 2 \nproduction)490/7.07 490/~7.0\nCore Outlet \nTemperature/Pressure \n(\u00baC/MPa)750 for \nelectricity \ngeneration)850/7.0 750/~6.0 750/~7.0 850-950/6.9 \n(electrical \nproduction) \n&950/5.1 (H 2", "characters": 842, "tokens": 350}
{"id": "2022_GIF_VHTR.pdf_101", "source": "2022_GIF_VHTR.pdf", "chunk": "generation)850/7.0 750/~6.0 750/~7.0 850-950/6.9 \n(electrical \nproduction) \n&950/5.1 (H 2 \nproduction)850/7.0 950/~7.0\nNeutron Energy Spectrum Thermal \npeaking just \nbelow 0.3 eVThermal \npeaking just \nbelow 0.3 eVThermal peaking \njust below \n0.3  eVThermal \npeaking just \nbelow 0.3 eVThermal \npeaking just \nbelow 0.3 eVThermal \npeaking just \nbelow 0.3 eVThermal \npeaking just \nbelow 0.3 eV\nAppendix VHTR.B \u2013 A Comparison of VHTR Fuel Cycle Parameters\nFuel Cycle \nParametersAreva\nModular HTRGeneral \nAtomics GT-\nMHRX-Energy           \nXe-100Huaneng \nGroup & \nCNEC/INET \nHTR-PMJAEA \nGTHTR300COKBM   GT-\nMHRKAERI\nNHDD\nReactor Thermal \nPower (MW-th)625 600 200 250 600 600 200\nReactor Electrical \nPower (MWe) \nGeneration~250, 186 for \ncogeneration \nwith process \nheat use262 to 286 \n(varied \nassumptions \ndocumented)80 (inferred) 100 per reactor \nin two reactors \nper module274-302 \ndepending on \noutlet T, 87-\n202 depending 262 to 286 \n(varied \nassumptions", "characters": 953, "tokens": 352}
{"id": "2022_GIF_VHTR.pdf_102", "source": "2022_GIF_VHTR.pdf", "chunk": "in two reactors \nper module274-302 \ndepending on \noutlet T, 87-\n202 depending 262 to 286 \n(varied \nassumptions \ndocumented)Only H 2 \nproductionVery-High-Temperature Reactor (VHTR)          PR&PP White Paper\n33on H 2 \nproduction\nFuel type\n -Form\n -Fertile material\n -Fissile materialLEU\nCeramic \ncoated particle\nU-238\nU-235LEU\nCeramic coated \nparticle\nU-238\nU-235LEU\nCeramic coated  \nparticle\nU-238\nU-235LEU\nCeramic \ncoated particle\nU-238\nU-235LEU\nCeramic \ncoated particle\nU-238\nU-235Pu initially\nCeramic \ncoated particle\nNone\nPuLEU\nCeramic \ncoated particle\nU-238\nU-235\nEnrichment (%) ~15 19.8 in fissile \nparticles, 0.7 \n(UNat) in fertile \nparticles10 8.5 in the \nequilibrium \ncore~14 Pure Pu 9.6 pebble, \n15.5 prismaticVery-High-Temperature Reactor (VHTR)          PR&PP White Paper\n34Appendix VHTR.B \u2013 A Comparison of VHTR Fuel Cycle Parameters (Continued)\nFuel Cycle \nParametersAreva\nModular HTRGeneral \nAtomics GT-\nMHRX-Energy           \nXe-100Huaneng \nGroup & \nCNEC/INET \nHTR-PMJAEA \nGTHTR300COKBM   GT-\nMHRKAERI", "characters": 1017, "tokens": 367}
{"id": "2022_GIF_VHTR.pdf_103", "source": "2022_GIF_VHTR.pdf", "chunk": "Xe-100Huaneng \nGroup & \nCNEC/INET \nHTR-PMJAEA \nGTHTR300COKBM   GT-\nMHRKAERI\nNHDD\nSource of Fissile \nMaterial (inputs \nare assumed \nsince not given in \navailable \ndocumentation)U.S. or \nEuropean \nenrichment \nplants \n(inferred)U.S. or \nEuropean \nenrichment \nplants \n(inferred)U.S. or European \nenrichment \nplants (inferredUndefined Undefined Russian \nexcess \nweapons Pu; \nother U and \nPu in later \nversionsUndefined\nFuel Inventory \n(MT) Not given 4.68 initial core, \n2.26 each \nreload~2.0 in \nequilibrium core~2.9 in \nequilibrium \ncoreNot given ~1.8 in \nequilibrium \ncycleNot given\nDischarge Burn-\nup (GWD/MT)150 121 for LEU \ncycle175 90 120 ~120-150 153\nRefueling \nfrequency \n(months)18 18 Continuous on \nlineContinuous on \nline24/18 \n(electrical)/18 \n(H2)18 Pebble \ncontinuous;\nRecycle Approach Baseline is \nonce-throughBaseline is \nonce-throughBaseline is once-\nthroughBaseline is \nonce-throughBaseline is \nrecyclingNo recycle, \ndeep-burnBaseline is \nonce-through\nRecycle \nTechnologyTo be \ndevelopedTo be \ndevelopedTo be developed To be \ndevelopedTo be \ndevelopedNo recycle, -\ndeep-burnTo be \ndeveloped", "characters": 1103, "tokens": 339}
{"id": "2022_GIF_VHTR.pdf_104", "source": "2022_GIF_VHTR.pdf", "chunk": "once-through\nRecycle \nTechnologyTo be \ndevelopedTo be \ndevelopedTo be developed To be \ndevelopedTo be \ndevelopedNo recycle, -\ndeep-burnTo be \ndeveloped\nRecycle efficiency To be \ndeterminedTo be \ndeterminedTo be \ndeterminedTo be \ndeterminedTo be \ndeterminedNo recycle, \ndeep-burnTo be \ndeterminedVery-High-Temperature Reactor (VHTR)         PR&PP White Paper\n35APPENDIX 2: Summary of PR relevant intrinsic design features. Reference IAEA- \nSTR-332. Please refer to IAEA-STR-332, for full explanations and complete \ndefinitions of terms and concepts.\nSummary of PR relevant Intrinsic \ndesign featuresB-VHTR\n(GT-MHR, HTTR)P-VHTR\n(PBMR)\nFeatures reducing the attractiveness of the technology for nuclear weapons programmes\n1. The Reactor Technology needs an \nenrichment Fuel Cycle phaseYes Yes\n2. The Reactor Technology produces  \nSF with low % of fissile plutoniumHigher burnup than LWR SF resulting in \nlow % fissile plutonium.Fully burn pebbles have higher burnup \nthan LWR SF resulting in low % fissile \nplutonium.\n3. Fissile material recycling performed \nwithout full separation from fission \nproductsNo recycling No recycling\nFeatures preventing or inhibiting diversion of nuclear material\n4. Fuel assemblies are large & difficult \nto dismantleYes. Fuel pellets are inserted in holes in \nfuel blocks. There is no fuel assembly in P-VHTR. \nFuel pebbles are small but to acquire 1 \nSQ of U-235 or Pu would require a large", "characters": 1421, "tokens": 359}
{"id": "2022_GIF_VHTR.pdf_105", "source": "2022_GIF_VHTR.pdf", "chunk": "fuel blocks. There is no fuel assembly in P-VHTR. \nFuel pebbles are small but to acquire 1 \nSQ of U-235 or Pu would require a large \nnumber of pebbles (tens of thousands). \n5. Fissile material in fuel is difficult to \nextractTRISO fuel is difficult to reprocess. TRISO fuel is difficult to reprocess.\n6. Fuel cycle facilities have few points \nof access to nuclear material, \nespecially in separated formFuel blocks are replaced after one cycle \nof irradiation, no reprocessing.Fuel cycle facilities mainly involve \npebble handling but no reprocessing, \nand remote operations are required.\n7. Fuel cycle facilities can only be \noperated to process declared feed \nmaterials in declared quantitiesN/A N/A\nFeatures preventing or inhibiting undeclared production of direct-use material\n8. No locations in or near the core of \na reactor where undeclared target \nmaterials could be irradiatedIrradiate target material in moderator \nblocks or control rod is a possibility. \nTon quantities of fertile material \nneeded to generate 1SQ would be \ndifficult to conceal and would affect \nreactor operation. The core is an open cavity filled with \nfuel pebble. There is no space for \ncontrol rod to hide the target \nmaterials. There is no space to hide \ntarget pebbles and no means to \nharvest the target pebbles after \nirradiation. Proliferator has difficulty to \ndistinguish the target materials from \nfuel pebble. Ton quantities of fertile \nmaterial needed to generate 1SQ \nwould be difficult to conceal and would \naffect reactor operation. \n9. The core prevents operation of the \nreactor with undeclared target", "characters": 1599, "tokens": 363}
{"id": "2022_GIF_VHTR.pdf_106", "source": "2022_GIF_VHTR.pdf", "chunk": "material needed to generate 1SQ \nwould be difficult to conceal and would \naffect reactor operation. \n9. The core prevents operation of the \nreactor with undeclared target \nmaterials (e.g. small reactivity \nmargins)The large number of fuel blocks \nrequired to accumulate 1 SQ of Pu \nmakes operation of the reactor with \nundeclared target easy to detect. It \nmight be possible to replace the \ncontrol rods with the target materials.It is easy to detect diversion because \nthe core is designed with little excess \nreactivity. It is possible to introduce U-\n238 pebbles for breeding, but would be \ndifficult to carry-out, owing to the large \nnumber of pebbles involved. Very-High-Temperature Reactor (VHTR)         PR&PP White Paper\n36Summary of PR relevant Intrinsic \ndesign featuresB-VHTR\n(GT-MHR, HTTR)P-VHTR\n(PBMR)\n10. Facilities are difficult to modify for \nundeclared production of nuclear \nmaterialThe particle-fuelled reactor is difficult \nto modify to use other fuel for \nundeclared production of nuclear \nmaterial..The large number of fuel pebbles \ninvolved in any undeclared production \nmakes the activity diifcult to carry out..\n11. The core is not accessible during \nreactor operationNot accessible and very high radiation \nenvironment.Not accessible and very high radiation \nenvironment.\n12. Uranium enrichment plants (if \nneeded) cannot be used to produce \nHEUExpect international safeguards in \nplace to deter HEU production.Expect international safeguards in \nplace to deter HEU production.\nFeatures facilitating verification, including continuity of knowledge\n13. The system allows for \nunambiguous Design Information", "characters": 1631, "tokens": 360}
{"id": "2022_GIF_VHTR.pdf_107", "source": "2022_GIF_VHTR.pdf", "chunk": "place to deter HEU production.Expect international safeguards in \nplace to deter HEU production.\nFeatures facilitating verification, including continuity of knowledge\n13. The system allows for \nunambiguous Design Information \nVerification (DIV) throughout life \ncycleDIV should be straight-forward. DIV should be straight-forward.\n14. The inventory and flow of nuclear \nmaterial can be specified and \naccounted for in the clearest possible \nmannerFuel blocks are amenable to item-\ncounting. Fuel pebbles are treated in bulk for \naccounting. Although it is in a closed \nsystem, nuclear material  in the \npebbles always move due to online \nrefueling through a pipe.\n15. Nuclear materials remain \naccessible for verification the greatest \npractical extentFuel blocks are identifiable by serial \nnumbers. However, since there is no \nwater shielding like LWRs, inspectors \ncannot directly see the fuel block \nloaded in the core.Verification of pebbles may pose \nchallenges.\n16. The system makes the use of \noperation and safety/related sensors \nand measurement systems for \nverification possible, taking in to \naccount the need for data \nauthenticationRadiation monitors and interlocks for \nfuel transfer machinery can also be \nused for safeguards. Measurement systems needed to \ncount and authenticate fuel pebbles \nfor operation can also be used for \nsafeguards. Devices that measure the \nreactivity and the burnup will also be \nimportant for safeguards.\n17. The system provides for the \ninstallation of measurement \ninstruments, surveillance equipment \nand supporting infrastructure likely to \nbe needed for verificationSystem elements are similar to LWRs \nand should be amendable to \ninstallation of safeguards equipment..Though system elements are similar to", "characters": 1758, "tokens": 356}
{"id": "2022_GIF_VHTR.pdf_108", "source": "2022_GIF_VHTR.pdf", "chunk": "instruments, surveillance equipment \nand supporting infrastructure likely to \nbe needed for verificationSystem elements are similar to LWRs \nand should be amendable to \ninstallation of safeguards equipment..Though system elements are similar to \nLWRs fuel accounting is different and \nitem counting is not practical. The \nsystem is a candidate for the \napplication of safeguards-by-design.37THE GENERATION IV INTERNATIONAL FORUM\nEstablished in 2001, the Generation IV International Forum (GIF) was created as a co-operative \ninternational endeavor seeking to develop the research necessary to test the feasibility and performance \nof fourth generation nuclear systems, and to make them available for industrial deployment by 2030. The \nGIF brings together 13 countries (Argentina, Australia, Brazil, Canada, China, France, Japan, Korea, Russia, \nSouth Africa, Switzerland, the United Kingdom and the United States), as well as Euratom \u2013 representing \nthe 27 European Union members and the United Kingdom \u2013 to co-ordinate research and develop these \nsystems. The GIF has selected six reactor technologies for further research and development: the gas-\ncooled fast reactor (GFR), the lead-cooled fast reactor (LFR), the molten salt reactor (MSR), the sodium-\ncooled fast reactor (SFR), the supercritical-water-cooled reactor (SCWR) and the very-high-temperature \nreactor (VHTR).", "characters": 1376, "tokens": 291}