Opportunities for International Collaboration on Modular High Temperature Reactors

Transcription

1 Opportunities for International Collaboration on Modular High Temperature Reactors Matt Richards, 1 Chris Hamilton, 1 Donald Hoffman, 2 Kazuhiko Kunitomi, 3 Min-Hwan Kim, 4 Michael A. Fütterer, 5 Grzegorz Wrochna 6 1 Ultra Safe Nuclear Corporation: 188 Piedra Loop, Los Alamos, NM, 87544, mrichards@ultrasafe-nuclear.com 2 Excel Services Corporation, Rockville Pike, Suite 100, Rockville, MD Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki pref., , Japan 4 Korea Atomic Energy Research Institute, Daedeok-daero, Yuseong-gu, Daejeon, , Korea 5 European Commission, DG Joint Research Centre, Directorate for Nuclear Safety and Security, PO Box 2, NL-1755ZG Petten, The Netherlands 6 National Centre for Nuclear Research, ul. Andrzeja Sołtana Otwock, Świerk, Poland This paper assesses potential opportunities for international collaboration with the East Asia countries of Japan, the Republic of Korea (ROK), and China on deployment and commercialization of Modular High Temperature Reactors (MHRs). All three countries have established MHR programs and could provide strong contributions in all phases of design and technology development for deployment of a commercial-scale MHR demonstration module and first-of-a-kind (FOAK) plant to support follow-on commercialization. In the case of Japan and the ROK, government institutions and industries in both countries have previously collaborated at a high level with their counterpart U.S. government institutions and industries on MHR design and technology development, including participation in the Next Generation Nuclear Plant (NGNP) Project sponsored by the U.S. Department of Energy (DOE). For this reason, and because China is well into the construction phase of its commercial-scale, pebble-bed core HTR-PM demonstration plant, most of the material presented in this paper is focused on potential opportunities with Japan and the ROK for demonstration of a MHR with a fixed block-type core. Opportunities for MHR collaboration with Europe are also discussed. I. INTRODUCTION Liquefied Natural Gas (LNG) is a major fuel source for electricity generation and industrial process heat, as well as an important raw material for industrial and chemical processing in East Asia and other parts of the world. The high cost of LNG, combined with goals for significantly reducing carbon and other emissions, has helped stimulate MHR programs in the ROK, Japan, and China where high fuel costs and MHR capabilities combine to create unique economic and environmental drivers. A number of discussions were held with senior personnel in universities, national laboratories, industry, and government to assess: (1) The nature, objectives, and vitality of the programs; (2) opportunities to perform joint Research and Development (R&D) and thereby reduce U.S. and international MHR development costs; and (3) Interest in potentially participating in a broader, joint, international MHR development and demonstration program which might include a FOAK MHR demonstration module built at a U.S. site. While there are modest existing U.S. MHR collaborations with the ROK and Japan and some information exchanges with China, personnel from all three countries expressed significant interest in expanding international collaboration. ROK and Japanese personnel also encouraged the U.S. to initiate discussions regarding an international development and demonstration program which might include a U.S.-sited MHR demonstration module as a step toward an advanced, higher temperature MHR for nuclear hydrogen production, which are the longer-term focus of their national programs. II. BACKGROUND As indicated above, this paper focuses primarily on potential opportunities for international collaboration with Japan and the ROK. Both countries have established MHR programs and could provide strong contributions in all phases of design and technology development for deployment of a commercial-scale MHR demonstration module and FOAK commercial plant. These countries also have somewhat different aspirations for their programs which results in some different or additional design requirements and technology development needs. Government institutions and industries in both countries have previously collaborated with their counterpart U.S.

2 government institutions and industries on MHR design and technology development, including participation in the recent NGNP Project sponsored by the DOE. China is independently developing MHR technology and has established the High Temperature Reactor Pebble-bed Module (HTR-PM) project (Ref. 1), a commercial-scale MHR demonstration plant. The HTR- PM is based on a pebble-bed core design that is significantly different than the prismatic-block core designs being developed in the U.S., Japan, and the ROK. The HTR-PM project has proceeded to the construction phase, with pouring of first concrete in December 2012 and startup expected in the time frame. Collaboration between the U.S. and China to the level of design, construction, licensing, and startup of a demonstration plant is therefore more challenging, both technically and politically. III. PRISMATIC BLOCK MHR DESIGN The conceptual design developed as part of the DOEsponsored NGNP project (Ref. 2) is a potential basis for developing a common international prismatic-block MHR design. This design is based on legacy MHR concepts going back to the 1980s. The NGNP MHR was designed as a co-generation plant for the production of electricity and process steam for industrial applications. UPPER PLENUM SHROUD UPPER PLENUM REACTOR CORE PERMANENT SIDE REFLECTOR LOWER PLENUM SHUTDOWN COOLING SYSTEM CROSS VESSEL CONTROL RODS REACTOR VESSEL MAIN CIRCULATOR STEAM GENERATOR VESSEL HOT DUCT STEAM GENERATOR The Standard Reactor Module (SRM) is shown in Fig. 1 and includes the Reactor System, Heat Transport System, and Shutdown Cooling System. The core is cooled with purified helium at 7 MPa pressure. The coolant inlet/outlet temperatures are 290/725 C. The core produces a thermal power of 350 MWt at a nominal power density of 5.9 MW/m 3. The steam generator can supply steam to a turbine/generator to produce electricity and also can supply steam for use in industrial applications. Fig. 1. NGNP Standard Reactor Module (350 MWt) The reactor core components include the hexagonal fuel elements, hexagonal graphite reflector elements, startup sources, and reactivity control material. The reactor core components, together with elements of the reactor internal components, constitute a graphite assembly which is supported on the graphite core support assembly and restrained by the metallic core support assembly. The hexagonal fuel elements are stacked in columns that form an active core annulus with columns of hexagonal graphite reflector elements in the central region (inner or central reflector) and surrounding the active core (outer or side reflector), as shown in Figure 2. Fig. 2. Reactor Core and Internal Components

3 The MHR can be designed to operate at higher power levels while maintaining the inherent safety features discussed under Section IV. Fig. 3 shows a 625 MWt MHR concept developed by AREVA (SC-HTGR) with two steam generator loops (Ref. 3). The primary helium and steam point design conditions are nearly identical to those for the NGNP SRM. Annular graphite core with high heat capacity and large surface area for heat transfer Relatively low power density Inert helium coolant, which reduces circulating and plateout activity Negative temperature coefficients of reactivity Multiple barriers to the release of radionuclides, starting with the coated particle fuel 855 µm 845 µm Fig. 4. Coated Particle Fuel Design Fig. 3. AREVA SC-HTGR Concept (625 MWt) A key design feature of the MHR that allows for high-temperature operation is the use of ceramic, coated particle fuel. A coated fuel particle consists of a microsphere ( kernel ) of nuclear fuel (usually in the form of an oxide, carbide, or oxycarbide) that is coated with multiple layers of pyrolytic carbon and silicon carbide. The buffer, inner pyrolytic carbon (IPyC), silicon carbide (SiC), and outer pyrolytic carbon (OPyC) layers are referred to collectively as a TRISO coating. The coating system can be viewed as a miniature pressure vessel that provides containment of radionuclides and gases. Figure 4 shows a typical coated particle fuel design and describes the functions of the fuel kernel and the coating layers. IV. MHR SAFETY DESIGN The MHR is designed with both passive and inherent safety features (Ref. 4). Inherent safety features include: MHR heat removal systems provide active, passive, and inherent safety. The main Heat Transport System (utilizing the steam generator and main circulator) can function as an active system for decay heat removal. The Shutdown Cooling System, consisting of an independent circulator and heat exchangers, is a back-up active system. In the event the main Heat Transport System and Shutdown Cooling System are unavailable, the passive Reactor Cavity Cooling System (RCCS) can remove decay heat using natural convection. Decay heat produced in the reactor core is transferred mostly in the radial direction by conduction and radiation to the un-insulated reactor vessel. The heat is then transferred by mostly radiation across the reactor cavity to panels attached to the reactor building. These panels are part of the RCCS, which includes downcomers and risers to remove the heat via natural convection of air. Figure 5 shows the RCCS integrated into the reactor building. The RCCS has multiple inlets and outlets to prevent flow blockage and is a robust structure designed to withstand natural events (e.g., tornado missiles) and sabotage. The SRM is located below grade level, which reduces seismic response, provides protection against external threats (e.g., aircraft), and allows for RCCS natural convection with a lower plant profile. High temperature, ceramic coated particle fuel

4 A demonstration plant in the U.S. licensed by the U.S. Nuclear Regulatory Commission could provide a valuable carbon-free energy option for the future, both for the U.S. and globally, especially if there is a significant level of international collaboration towards a common design to reduce the technology and other investment risks/costs $/Million Btu Natural Gas Price at Henry Hub Reference Case EIA AEO 2015 Report Fig. 5. RCCS Integrated into Reactor Building V. MARKET ANALYSIS The NGNP Industry Alliance (NIA) consists of a number of companies and organizations that support development and deployment of MHRs. These companies include reactor vendors, utilities, potential industrial end users of MHR process steam/heat, nuclear graphite vendors, and companies with design, technology development, regulatory licensing, and other MHR subject matter expertise. The NIA has been investigating potential markets for MHRs in both North America and globally as part of its business plan development. As discussed in Ref. 5, in the absence of carbon taxes, MHRs are competitive with natural gas when the natural gas price is in the range $6 - $8/million Btu. As indicated in Fig. 6, natural gas prices in the U.S. are presently well below this range and are not expected to reach this range until the time frame (Ref. 6). Japan and the ROK import natural gas in the form of LNG and are the world s No. 1 and No. 2 importers of LNG. Liquefaction and transportation required for LNG adds significant costs. The long-term contract price for LNG in Japan and the ROK is expected to remain above the $6 - $8/million Btu range. The relatively high fossil fuel costs in Europe also make MHRs more economically competitive in the nearer term. Nearer-term market analyses for MHR commercialization should focus on replacement of fossil fuels for both electricity and industrial applications, especially for regions where fossil fuels are presently required but costs are expected to remain relatively high Year Fig. 6. U.S. Natural Gas Price at Henry Hub The market analysis described in Ref. 7 considered both nearer-term applications with MHRs operating with coolant outlet temperatures in the 700 C C range and longer-term applications with Very High Temperature Reactors (VHTRs) operating with coolant outlet temperatures in the 900 C C range. Replacement of just 20% of the LNG imports used for electricity generation only in Japan and the ROK is potentially a large nearer-term market that could account for one hundred or more MHR modules. MHRs providing process steam for industrial applications could further expand the nearer-term market. A large market potential also exists in both countries for using VHTRs to produce hydrogen for industry and possibly for transportation. Between Japan and the ROK, the steel manufacturing industry alone could support approximately two hundred 600-MWt VHTR modules producing both hydrogen for iron-ore reduction and the large quantity of electricity required by the manufacturing process. In both countries, steel manufacturing is the single largest emitter of carbon dioxide (approximately 10% of the total emissions), and conversion to nuclear hydrogen and electricity would essentially eliminate these emissions. Significant opportunities for MHR co-generation applications also exist in Europe. VI. MHR DEVELOPMENT IN JAPAN Japan has been developing MHR/VHTR technology since the 1960s (Ref. 8). The 30 MW(t) High Temperature Engineering Test Reactor (HTTR) is located at the Japan Atomic Energy Agency (JAEA) Oarai

5 Research and Development Center (see Fig. 7), near the town of Oarai in Ibaraki Prefecture. The Japan Materials Testing Reactor (JMTR) and the JOYO sodium-cooled fast reactor are also located at the Oarai Research and Development Center. hydrogen production process. Major design parameters for the HTTR are given in Table I. JOYO Fig. 7. JAEA Oarai Research and Development Center The primary purposes for the HTTR are (1) to establish basic VHTR technology and demonstrate VHTR inherent safety, (2) demonstrate utilization of nuclear heat to produce hydrogen, and (3) irradiation of VHTR fuel and materials under prototypical conditions. The HTTR achieved first criticality in 1998 and is currently the only operational prismatic-block type HTR in the world. Since first criticality, there have been more than a dozen operations of the HTTR, including several rise-to-power tests and safety demonstration tests. In addition, a number of in-service inspections have been performed. In 2010, the HTTR was operated continuously for 50 days at 950 C helium outlet temperature with its 10 MWt intermediate heat exchanger (IHX) on line. During this operation, JAEA successfully collected a comprehensive set of tritium data from the primary helium coolant, secondary helium coolant, water cooling systems, and containment atmosphere. This test program was a collaborative effort between DOE and JAEA, and appropriate quality assurance (QA) reviews were performed to ensure the test data complied with U.S. requirements for computer code validation. Figure 8 shows the overall layout of the HTTR site. Figure 9 shows a cut-away view of the reactor building. The reactor pressure vessel and cooling system equipment are located within a steel containment vessel, which is located below grade. The laboratory building (which includes office space) and development building are located to the west of the reactor building. The development building includes facilities and equipment for development of the Iodine-Sulfur (IS) thermochemical Fig. 8. Overall Layout of the HTTR Site Fig. 9. Cut-Away View of the HTTR Reactor Building VII. MHR DEVELOPMENT IN KOREA Through the Korea Atomic Energy Research Institute (KAERI), the ROK has been working on MHR/VHTR technology development and design and nuclear hydrogen production for over a decade. In 2012, the Nuclear Heat and Hydrogen (NuH2) design project was initiated in collaboration with Korean industry (Ref. 9). In addition, the Korea Nuclear Hydrogen Association (KNHA) was established that includes industry participants. In April 2013, the NIA and KNHA signed a collaborative agreement. A roadmap for the NuH 2 project and supporting technology development has been established. This roadmap shows a phased approach for deployment, starting initially with demonstration of a process heat

6 system with reactor outlet temperature of approximately 750 C before transitioning to a hydrogen production system with reactor outlet temperature > 900 C in the time frame. Table I. HTTR Major Design Parameters Thermal power 30 MW Coolant outlet 850 C /950 C temperature Core inlet temperature 395 C Primary system pressure 4 MPa Fuel Coated particles with low enriched UO 2 (3 to 10 wt%, avg. 6 wt% U-235) kernels Fuel element type Prismatic block Fuel loading Off-load, 1 batch Core diameter 2.3 m Core height 2.9 m Average core power 2.5 MW/m 3 density Core flow direction Downward flow Coolant flow rate 10.2 kg/s (950 C operation) Primary coolant pressure 4.0 MPa Reactor pressure vessel 2¼ Cr 1 Mo Steel material Reactor pressure vessel 13.2 m height Reactor pressure vessel 5.5 m diameter KAERI has been conducting R&D on high temperature materials, including nuclear-grade graphite, high-temperature metals, and carbon/carbon composites. KAERI has also constructed facilities for performing high-temperature testing, including a small-scale nitrogen loop and the 150-kWt Helium Experimental Loop (HELP, Fig. 9). Testing of a printed circuit heat exchanger module in HELP is described in Ref. 10. A Natural Cooling Experimental Facility to test Reactor Cavity Cooling System (RCCS) performance is located in the same building as HELP. This facility is shown in Fig. 10 and was sized to be one-fourth the scale of an RCCS for a 200-MWt reactor module. VIII. PRISMATIC BLOCK CORE ADVANTAGES Two basic designs for MHR cores are (1) a moving bed of spherical graphitic pebble fuel elements with online refueling (e.g., HTR-PM) and (2) a fixed array of prismatic graphite block fuel elements with periodic offline refueling (e.g., NGNP conceptual design). The HTR- PM and other concepts developed in Germany and South Africa have pebble-bed cores. Concepts developed in the U.S., Japan, the ROK, and Russian Federation have prismatic block cores. Both fuel elements contain TRISO-coated fuel particles. Key differences between the two core designs that favor the prismatic block design are discussed below: Fig. 9. KAERI HELP Facility Fig. 10. KAERI Natural Cooling Experimental Facility 1. During limiting design-basis and beyond design-basis events with helium depressurization, the effective thermal conductance of a pebble-bed core is significantly lower than that for a prismatic-block core in the radial direction towards the passive reactor cavity heat removal sink. For this reason and other design limitations discussed below, the core power density in pebble-bed reactors must be lower than that in prismatic-block reactors to limit peak fuel temperatures during accidents. Thus, for equal core volumes, pebble-bed reactors must have lower power ratings than prismatic-block reactors. 2. The overall coolant flow resistance of a pebble-bed core is inherently greater than the overall coolant flow resistance of a prismatic-block core. Thus, the core pressure drop will be higher in a pebble-bed reactor than a prismatic-block reactor for designs having the same core height and coolant flow rates.

7 Consequently, a pebble-bed reactor requires more energy per unit of thermal power output to circulate the coolant, which results in lower plant efficiency. 3. The inherently higher operating power level and efficiency of prismatic-block reactors relative to pebble-bed reactors equates to better overall economics, including lower electricity generation busbar costs for a plants having the same electrical power output. This economic advantage in electricity generation cost translates to an approximately equivalent advantage for the prismatic-block reactor in electricity/process heat cogeneration applications given that a cost measure of the thermal energy utilized as process heat is the value of the electricity that could have been produced had the thermal energy been used for electricity production. 4. The prismatic block designs include an annular core to achieve a high-power rating while maintaining inherent safety. 5. As demonstrated by operating experience in the prismatic-block Fort St. Vrain reactor and in the German pebble-bed AVR, there is much more graphite dust formation in pebble-bed reactors than in prismatic-block reactors. The circulation of large quantities of graphite dust in the primary coolant loop of pebble bed reactors has the potential to adversely affect the operation of a direct-cycle power conversion system (PCS) and/or an IHX, and could potentially preclude use of compact heat exchangers with narrow flow passages. Also, the dust is an excellent medium for enhanced release of fission products during accidents involving depressurization of the primary coolant loop, which could have an adverse impact on reactor building design requirements in terms of fission product retention. 6. Uncertainties associated with pebble-bed core thermal/hydraulic performance could adversely impact licensing and design certification. Although fuel temperatures during normal operation should be lower in a pebble-bed reactor than in a prismaticblock reactor because of the lower core power density and better pebble-to-coolant heat transfer, coolant and fuel temperatures in the AVR were much higher than predicted based on temperature measurements in the core and the results of postirradiation examination (PIE) of AVR fuel. The reasons for these higher-than expected temperatures are not well understood, but they were likely related to power peaking and thermal/hydraulic irregularities at core reflector boundaries or adjacent to the graphite noses in the AVR core; effects that could be enhanced in a pebble-bed annular core. One such anomaly that has been observed experimentally is that pebble flow along reflector surfaces can be two to three times slower than in the interior of the pebble bed core. 7. The prismatic-block reactor refueling approach and fuel element design makes fuel element accountability relatively simple and diversion of nuclear material very difficult. The prismatic-block core design contains at most about 1,000 fuel elements (for a 600 MWt class core) and the core is refueled every 18 to 24 months during regularly scheduled outages. With the on-line refueling of a pebble-bed core, diversion of nuclear material can be a significant risk. The core contains hundreds of thousands of pebble fuel elements that are typically recirculated 10 to 12 times before reaching design burnup. The plutonium composition in low-burnup pebbles can be weapons-grade and some of these pebbles could be diverted before re-insertion into the core. India was able to become a nuclear-weapons state by diverting fuel from a reactor with on-line refueling supplied by Canada. IX. INTERNATIONAL COLLABORATION The U.S. Energy Policy Act of 2005 (EPACT) envisioned a U.S.-based cost-sharing public/private partnership for design, technology development, construction, licensing, and startup of the NGNP demonstration plant. This partnership has not come to fruition for a number of reasons, including the historically low prices for natural gas in the U.S. that significantly increases the risks for U.S. private industry to invest in a new nuclear technology. In contrast, the Chinese government has assumed nearly full responsibility (including financing) for the HTR-PM demonstration plant, which is well into the construction phase. A model based on international collaboration with Japan, the ROK, and Europe could define a commondesign prismatic-block MHR concept that has better economics and more power-range flexibility than China s HTR-PM, along with high confidence with regard to regulatory licensing and safety assessments. More recently, representatives from the NIA, Poland, Japan, JAEA, KAERI, European Union (EU) government institutions and industry, and others have entered into a collaboration to organize an international partnership for a shared near-term deployment of MHR technology, coupled with an R&D program to enable a follow-on deployment of VHTRs for very high efficiency electricity generation, hydrogen production, and other very high temperature non-electric applications. This effort is called PRIME for Polygeneration Reactor with Inherent safety, Modularity, and Economic competitiveness. The PRIME collaboration builds upon collaboration between the NIA and the European Nuclear Cogeneration Industrial Initiative (NC2I) under the Sustainable Nuclear

8 Energy Technology Platform (SNETP, (Ref. 11). The PRIME effort is also aligned with the recent EU GEMINI+ MHR proposal objectives: Define the design basis for a reference MHR system for co-generation o Addressing the needs of European industry o Addressing safety standards, including cogeneration o Converging as much as possible towards a common MHR block-core design To define a licensing frame work for a MHR nuclear co-generation system Demonstration of MHR cogeneration at a specified European industrial site Poland has shown a strong interest in utilizing MHRs for providing electricity and process steam for industrial applications and could be a candidate location for a European demonstration plant. In July 2016, the Poland Minister of Energy appointed a Committee for Deployment of High Temperature Reactors to examine: Analysis of Polish economy needs and export potential Inventory of relevant design and manufacturing capabilities of Polish science and industry Cost estimates, business models, and funding possibilities Analysis of legal framework Establishing international collaboration As shown in Fig. 9, Poland has evaluated candidate industrial sites for MHR co-generation integration. Updated progress on the PRIME initiative will be reported at the ICAPP 2017 meeting. X. CONCLUSIONS The MHR programs in the U.S., Japan, the ROK, and Europe have the goal of developing and demonstrating an inherently safe fission reactor technology that can be used for both electricity production and process heat/steam applications in the industrial and transportation energy sectors. For the latter, only MHRs/VHTRs can provide the high temperatures required for these applications. Globally, nearly 80% of the world s energy demand is consumed in the industrial and transportation sectors, and nearly all of this energy is supplied from the burning of fossil fuels. Using MHRs/VHTRs to address these nonelectric energy requirements can significantly reduce fossil fuel consumption and carbon dioxide emissions while providing a long-term and economically sustainable alternative to burning fossil fuels. MHR for Poland 13 largest chemical plants need 6500 MW of heat at T= C They use 200 TJ / year, equivalent to burning of >5 million t of natural gas or oil Replacing by MHR would reduce CO 2 emission by mln t / year MW t reactor size fits the needs Plant boilers MW ZE PKN Orlen S.A.Płock Arcelor Mittal Poland S.A Zakłady Azotowe "Puławy" S.A Zakłady Azotowe ANWIL SA Zakłady Chemiczne "Police" S.A Energetyka Dwory International Paper - Kwidzyn Grupa LOTOS S.A. Gdańsk ZAK S.A. Kędzierzyn Zakl. Azotowe w Tarnowie Moscicach S.A MICHELIN POLSKA S.A PCC Rokita SA MONDI ŚWIECIE S.A Fig. 9. Candidate Sites in Poland for MHR Co-generation MHRs are an established technology that is well suited for transitioning away from fossil fuels. MHRs/VHTRs have high temperature capability (700 C C outlet temperature), and are designed with inherent safety to preclude the possibility of a fuel failure scenario or significant release of radioactive material. Because no electric power sources are required to prevent core damage and radioactivity release, MHRs are immune to extreme events such as complete station blackout and can be considered walk-away safe. Because of their inherent safety, MHRs can be colocated within new or existing industrial facilities to provide process heat and steam. MHRs also operate with high thermal efficiency, enabling location in areas with very limited supply of cooling water, including in-land locations away from areas that could pose a significant risk from flooding or other natural disasters. Used fuel is not required to be cooled within a water pool, but can be air cooled without motive force required. Significant and economically viable markets already exist in Japan, the ROK, and Europe for MHRs and follow-on VHTRs. Additional markets should develop in the U.S. in the near future. Significant markets may also develop in countries with little nuclear experience but with interest in an inherently safe reactor design to displace expensive fossil fuels. The technical advantages of the prismatic-block design make it an effective competitor to the HTR-PM and the pedigree of the nuclear experience, quality control, and regulatory oversight in the U.S., Japan, and the ROK should be additional marketing advantages.

9 An international collaboration model for MHR development and deployment of a demonstration plant would require establishment of government-togovernment agreements with Japan, the ROK, Europe, and potentially others that go beyond the current basic R&D collaborations and information exchange (e.g., Generation IV International Forum activities). This type of model would also require collaboration among industrial organizations in the partner countries. In September of 2016, the Secretary of Energy Advisory Board (SEAB) submitted its Report of the Task Force on the Future of Nuclear Power (Ref. 12). The Task Force noted other countries are likely to be interested in participating in the program and that international collaboration would be beneficial in terms of technical contributions, cost sharing, and the opportunity to shape future commercial deployments around the world. The Task Force also noted international financial participation would likely be accompanied by requirements for work share, access to intellectual property, deployment rights in their respective countries/regions, and project governance. A project model with international collaboration should be crafted to properly manage any potential complexities associated with these requirements. REFERENCES 1. Y. DONG, HTR Development Progress on China, IAEA Meeting of the Technical Working Group on Gas Cooled Reactors (TWG-GCR), February 25-27, 2015, Vienna International Center, Vienna, Austria. Meetings/2015/ NPTDS/Day1/01- China-DONG_V2.pdf 2. M. RICHARDS, A. BAXTER, C. ELLIS, O. GUTIERREZ, and J. CROZIER, Conceptual Design of the NGNP Reactor System, Proceedings of the 20th International Conference on Nuclear Engineering, July 30 Aug. 3, 2012, Anaheim, CA, paper ICONE20POWER J. MAYER and F. SHAHROKHI, The Steam Cycle High-Temperature Gas-Cooled Reactor, Nuclear News, Vol. 57, No. 13, pp , December M. RICHARDS, Establishing a Safety Standard for Commercial Modular HTGRs, Proceedings of ICAPP 2013, April 14-18, 2013, Jeju Island, Korea, paper FF Modular High-Temperature Gas-Cooled Reactor Technology, An Essential Option for the Global Energy Future, A Business Plan for Commercialization, NGNP Industrial Alliance Ltd., June Annual Energy Outlook 2015, DOE/EIA- 0383(2015), U.S. Department of Energy, Energy Information Administration, April M. RICHARDS, C. HAMILTON, and F. VENNERI, HTGR Economic / Business Analysis and Trade Studies, Task 3, Enhanced Technical and Financial Evaluation of Opportunities for International Collaboration, East Asia: Japan, Korea, and China, Report USNC-NIA-G00003, Ultra Safe Nuclear Corporation, Los Alamos, NM, August M. RICHARDS, Technology Development Plan for Utilization of JAEA Facilities, Data, and Experience to Support the NGNP Project, Report PC , General Atomics, San Diego, CA, December W. LEE, Status of Nuclear Heat and Hydrogen Systems Concept Study, Proceedings of HTR2014, Weihai, China, October 27-31, 2014, Paper HTR C. KIM, et al., High-Temperature Test of 800HT Printed Circuit Heat Exchanger in HELP, Proceedings of HTR 2014, Weihai, China, October 27-31, 2014, Paper HTR M. FÜTTERER et al., The GEMINI Initiative: A Step forward towards Nuclear Cogeneration with HTGR, Proceedings of ICAPP 2015, May 3-6, 2015, Nice, France, Paper Secretary of Energy Advisory Board Report of the Task Force on the Future of Nuclear Power, U.S. Department of Energy, September 22, 2016, 16_SEAB%20Nuclear%20Power%20TF%20Report %20and%20transmittal.pdf