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DOE OSTI2012

MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit 4

Unit 4 of the Fukushima Dai-ichi Nuclear Power Plant suffered a hydrogen explosion at 6:00 am on March 15, 2011, exactly 3.64 days after the earthquake hit the plant and the off-site power was lost. The earthquake occurred on March 11 at 2:47 pm. Since the reactor of this Unit 4 was defueled on November 29, 2010, and all its fuel was stored in the spent fuel pool (SFP4), it was first believed that the explosion was caused by hydrogen generated by the spent fuel, in particular, by the recently discharged core. The hypothetical scenario was: power was lost, cooling to the SFP4 water was lost, pool water heated/boiled, water level decreased, fuel was uncovered, hot Zircaloy reacted with steam, hydrogen was generated and accumulated above the pool, and the explosion occurred. Recent analyses of the radioisotopes present in the water of the SFP4 and underwater video indicated that this scenario did not occur - the fuel in this pool was not damaged and was never uncovered the hydrogen of the explosion was apparently generated in Unit 3 and transported through exhaust ducts that shared the same chimney with Unit 4. This paper will try to answer the following questions: Could that hypothetical scenario in the SFP4 had occurred? Could the spent fuel in the SPF4 generate enough hydrogen to produce the explosion that occurred 3.64 days after the earthquake? Given the magnitude of the explosion, it was estimated that at least 150 kg of hydrogen had to be generated. As part of the investigations of this accident, MELCOR models of the SFP4 were prepared and a series of calculations were completed. The latest version of MELCOR, version 2.1 (Ref. 1), was employed in these calculations. The spent fuel pool option for BWR fuel was selected in MELCOR. The MELCOR model of the SFP4 consists of a total of 1535 fuel assemblies out of which 548 assemblies are from the core defueled on Nov. 29, 2010, 783 assemblies are older assemblies, and 204 are new/fresh assemblies. The total decay heat of the fuel in the pool was, at the time of the accident, 2.284 MWt, of which 1.872 MWt were from the 548 assemblies of the last core discharged and 0.412 MWt were from the older 783 assemblies. These decay heat values were calculated at Oak Ridge National Laboratory using the ORIGEN2.2 code (Ref. 2) - they agree with values reported elsewhere (Ref. 3). The pool dimensions are 9.9 m x 12.2 m x 11.8 m (height), and with the water level at 11.5 m, the pool volume is 1389 m3, of which only 1240 m3 is water, as some volume is taken by the fuel and by the fuel racks. The initial water temperature of the SFP4 was assumed to be 301 K. The fuel racks are made of an aluminum alloy but are modeled in MELCOR with stainless steel and B4C. MELCOR calculations were completed for different initial water levels: 11.5 m (pool almost full, water is only 0.3 m below the top rim), 4.4577 m (top of the racks), 4.2 m, and 4.026 m (top of the active fuel). A calculation was also completed for a rapid loss of water due to a leak at the bottom of the pool, with the fuel rapidly uncovered and oxidized in air. Results of these calculations are shown in the enclosed Table I. The calculation with the initial water level at 11.5 m (full pool) takes 11 days for the water to boil down to the top of the fuel racks, 11.5 days for the fuel to be uncovered, 14.65 days to generate 150 kg of hydrogen and 19 days for the pool to be completely dry. The calculation with the initial water level at 4.4577 m, takes 1.1 days to uncover the fuel and 4.17 days to generate 150 kg of hydrogen. The calculation with the initial water level at 4.02 m takes 3.63 days to generate 150 kg of hydrogen this is exactly the time when the actual explosion occurred in Unit 4. Finally, fuel oxidation in air after the pool drained the water in 20 minutes, generates only 10 kg of hydrogen this is because very little steam is available and Zircaloy (Zr) oxidation with the oxygen of the air does not generate hydrogen. MELCOR calculated water levels and hydrogen generated in the SFP4 as a function of time for initial water levels of 4.457 m, 4.2 m and 4.02 m are shown in Figs. 1 and 2. Water levels increase at the beginning due to the expansion of the water during the heat-up from 301 K to 373 K. Boiling occurs after the water temperature reaches 373 K. The total amount of hydrogen generated is ~2000 kg, this amount includes hydrogen generated from Zr, which is the largest amount (~1580 kg), from stainless steel (~360 kg), and from B4C (~60 kg). In theory, it is possible to generate up to 3.4 kg of hydrogen per assembly (from oxidation of Zr in the fuel cladding and box), or a total of 4,525 kg from the hot 1331 assemblies stored in the SFP4. The hydrogen generated from oxidation of steel and B4C will be additional. So the answers to the questions are YES according to these MELCOR calculations, enough hydrogen (150 kg) could be generated in the SFP4 3.64 days after the earthquake to produce ...

08 HYDROGEN↗

DOE OSTI2017

Global Value Chain and Manufacturing Analysis on Geothermal Power Plant Turbines: Preprint

The global geothermal electricity market has significantly grown over the last decade and is expected to reach a total installed capacity of 18.4 GWe in 2021 (GEA, 2016). Currently, geothermal project developers customize the size of the power plant to fit the resource being developed. In particular, the turbine is designed and sized to optimize efficiency and resource utilization for electricity production; most often, other power plant components are then chosen to complement the turbine design. These custom turbine designs demand one-off manufacturing processes, which result in higher manufacturing setup costs, longer lead-times, and higher capital costs overall in comparison to larger-volume line manufacturing processes. In contrast, turbines produced in standard increments, manufactured in larger volumes, could result in lower costs per turbine. This study focuses on analysis of the global supply chain and manufacturing costs for Organic Rankine Cycle (ORC) turboexpanders and steam turbines used in geothermal power plants. In this study, we developed a manufacturing cost model to identify requirements for equipment, facilities, raw materials, and labor. We analyzed three different cases 1) 1 MWe geothermal ORC turboexpander 2) 5 MWe ORC turboexpander and 3) 20 MWe geothermal steam turbine, and calculated the cost of manufacturing the major components, such as the impellers/blades, shaft/rotor, nozzles, inlet guide lanes, disks, and casings. Then we used discounted cash flow (DCF) analysis to calculate the minimum sustainable price (MSP). MSP is the minimum price that a company must sell its product for in order to pay back the capital and operating expenses during the plant lifetime (CEMAC, 2017). The results showed that MSP could highly vary between 893 dollar/kW and 30 dollar/kW based on turbine size, standardization and volume of manufacturing. The analysis also showed that the economy of scale applies both to the size of the turbine and the number manufactured in a single run. Sensitivity analysis indicated these savings come largely from reduced labor costs for design and engineering and manufacturing setup.

15 GEOTHERMAL ENERGY↗

DOE OSTI2017

Global Value Chain and Manufacturing Analysis on Geothermal Power Plant Turbines

15 GEOTHERMAL ENERGY↗

DOE OSTI2023

Report on the Creation and Progress of the Hydrogen Regulatory Research Review Group

The current U.S. nuclear generation fleet is increasingly recognized by governmental, scientific, public policy, and industrial communities as having a strategic role in support of the ongoing national transition to a clean energy future. Federal incentives and actions are aligning to expand the role of nuclear power as a viable and more flexible contributor to the evolving national clean energy mix through programs and initiatives such as nuclear power loan guarantees; the Inflation Reduction Act’s (IRA) clean nuclear electrical, steam, and hydrogen incentives; the Infrastructure Investment and Jobs Act (IIJA, also referred to as the Bipartisan Infrastructure Act or BIL); and near term Department of Energy (DOE) funding opportunities related to nuclear based hydrogen hubs and nuclear integrated hydrogen demonstration projects. . Additionally, leveraging clean nuclear electricity and steam is being explored by industries desiring to transition away from carbon-intensive energy sources. Even with all these emerging enablers, notable barriers remain for the widespread adoption of these opportunities within the U.S. nuclear fleet, including the following: • Assurance of the markets for alternate products needed to support decision-making for large capital modification investments • Electric utility mindset and business history centered solely on producing electricity • Design change complexity and regulatory uncertainty associated with plant modifications needed to support alternate product streams The DOE Light Water Reactor Sustainability (LWRS) Flexible Plant Operations and Generation (FPOG) Pathway is developing options to help U.S. nuclear power plants (NPP) in all these areas to enable NPPs designed for steady baseload operation to integrate with intermittent wind and solar capacity to assure reliable clean energy for the nation. Current and near-term laboratory research is focusing on the technical, regulatory, safety, demonstration, and economic elements in support of improving nuclear plant flexibility through hybrid production of electricity and other non-electric products such as hydrogen and energy arbitrage.

42 ENGINEERING↗

DOE OSTI2012

The AP1000{sup R} China projects move forward to construction completion and equipment installation

The AP1000 design is the only Generation III+ technology to receive design certification from the U.S. Nuclear Regulatory Commission. This evolutionary design provides the highest safety and performance standards and has several distinct advantages over other designs, including improved operations and reduced construction schedule risks through the use of modern, modular, engineering principles that allow construction and fabrication tasks traditionally performed in sequence to be undertaken in parallel. Since the first granting of Design Certification in 2005 by the NRC, the AP1000 design has been modified to meet emergent NRC requirements such as those requiring the design to withstand the impact of an aircraft crash. Both domestic and foreign utilities have turned to the Westinghouse AP1000 plant design to meet their near - and long-term sustainable energy needs. The first ever deployment of this advanced U.S. nuclear power technology began in China in 2007 with the award of a contract to build four AP1000 units, constructed in pairs at the coastal sites of Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Currently, all four units are at an advanced stage of construction. The commercial operation date for Sanmen Unit 1 is November 2013 followed by Haiyang Unit 1 being operational in May 2014. Construction and equipment manufacture is at an advanced stage. Sanmen Unit 1 equipment that has been delivered includes the reactor vessel, the reactor vessel closure head, the passive residual heat removal heat exchanger, the integrated head package, the polar crane, and the refueling machine. The steam generators are also completed. The RV was installed within the containment vessel building in September 2011. The installation of this major equipment will allow the setting of the containment vessel top head. Haiyang Unit 1 is also achieving significant progress. Significant benefits continue to be realized as a result of lessons learned and experience gained from the first-of-a-kind activities for Sanmen Unit 1 and AP1000 equipment design and manufacturing. For example, the nuclear island basemat at Haiyang Units 1 and 2 and Sanmen Unit 2 was laid in less time than that of Sanmen Unit 1, the ultra-large steam generator and RV forging lead times were reduced for the follow on units, and the fabrication of the auxiliary building module for Haiyang Unit 1 took less time than for the Sanmen first unit. These benefits are also being realized by the United States AP1000 project construction and fabrication modules, and equipment. Some difficulties arise from building this first-of-a-kind (advanced passive) type of plant; however, these difficulties are being overcome and the overall schedule remains achievable. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Aerosol retention during SGTR meltdown sequences: Experimental insights of the effect of size and shape of the breach

This paper summarizes the major insights gained from aerosol retention capability of a tube bundle that simulates the break stage of the secondary side of a failed steam generator under dry SGTR conditions. This scenario is highly relevant in nuclear safety since it affects the potential retention of radioactive particles in case of meltdown sequences with a SGTR. An 8-test experimental campaign has been carried out, extending the current database on the decontamination capability of the steam generator. The effects of the breach features (shape and size) and the particle nature (SiO{sub 2} and TiO{sub 2}) on the collection efficiency have been explored. The results confirmed the strong effect of the physical nature even when tube breaks in a fish-mouth mode. Loose aggregates (i.e. TiO{sub 2}) would be trapped to a limited extent (less than 25%); while single- or few-aggregates (i.e. SiO{sub 2}) would undergo a quite effective removal (i.e. over 75%). For fish-mouth breaches and SiO{sub 2} particles, the breach size has been found to moderately affect retention efficiency. Furthermore, the breach shape does not seem to have any effect on the net collection efficiency within the break stage, no matter the particle type. However, individual tube measurements indicate notably different deposition patterns, although an effect of the facility geometry cannot be disregarded as a key player in this observation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Analysis of the KROTOS KFC test by coupling X-Ray image analysis and MC3D calculations

During a hypothetical severe accident sequence in a Pressurized Water Reactor (PWR), the hot molten materials (corium) issuing from the degraded reactor core may generate a steam explosion if they come in contact with water and may damage the structures and threaten the reactor integrity. The SERENA program is an international OECD project that aims at helping the understanding of this phenomenon also called Fuel Coolant Interaction (FCI) by providing data. CEA takes part in this program by performing tests in its KROTOS facility where steam explosions using prototypic corium can be triggered. Data about the different phases in the premixing are extracted from the KROTOS X-Ray radioscopy images by using KIWI software (KROTOS Image analysis of Water-corium Interaction) currently developed by CEA. The MC3D code, developed by IRSN, is a thermal-hydraulic multiphase code mainly dedicated to FCI studies. It is composed of two applications: premixing and explosion. An overall FCI calculation with MC3D requires a premixing calculation followed by an explosion calculation. The present paper proposes an alternative approach in which all the features of the premixing are extracted from the X-Ray pictures using the KIWI software and transferred to an MC3D dataset for a direct simulation of the explosion. The main hypothesis are discussed as well as the first explosion results obtained with MC3D for the KROTOS KFC test. These results are rather encouraging and are analyzed on the basis of comparisons with the experimental data. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl alloys

A large effort is underway under the leadership of US DOE Fuel Cycle Research program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanical characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Breakaway oxidation of ZIRLO and optimized ZIRLO cladding per 10CFR50.46c criteria

The United States Nuclear Regulatory Commission (NRC) in 10CFR50.46c proposed changes to the loss of coolant accident (LOCA) criteria for LWR licensees including a new requirement to test cladding materials for breakaway oxidation. Cladding materials must be shown to resist breakaway oxidation long enough for the LOCA transient to be terminated. The onset of breakaway oxidation is characterized by a rapidly increasing rate of oxidation and the absorption of hydrogen resulting in a loss of cladding integrity. The proposed rule changes effectively require some manner of periodic and on-going testing to monitor the population of cladding for breakaway resistance. More than 200 lots of production cladding material have been tested in accordance with the draft regulatory guidelines. Testing has been conducted in a steam environment at the critical temperature for breakaway oxidation for durations that exceed the analytical endurance limit required to comply with the breakaway oxidation test criteria. The alloys that have been tested include both ZIRLO cladding and Optimized ZIRLO cladding. Breakaway oxidation testing consisted of three basic steps: preparation of test specimens, high temperature steam oxidation and post-test specimen evaluation. The results show there is a high probability that all ZIRLO and Optimized ZIRLO lots will pass production certification testing. Variations in response to breakaway oxidation testing were observed only in testing that ran well beyond the maximum LOCA duration of 2000 seconds. No indications of early breakaway oxidation were observed in the testing of production material that has been completed thus far. This provides confidence that a sampling approach to breakaway oxidation testing and monitoring remains a sound approach.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Corrosion resistant coatings for zirconium-alloy cladding with improved accident tolerance

The Westinghouse-led accident tolerant fuel (ATF) program is evaluating Accident Tolerant Fuel (ATF) coatings of MAX phase (Ti{sub 2}AlC) and multi-layer Ti{sub x}Al{sub 1-x}N/TiN to enhance the oxidation resistance of zirconium alloy fuel cladding in both normal operation and beyond design-based conditions of light water reactors. A variety of coating technologies have been applied to deposit these coating materials onto Zircaloy-4 or Zirlo sheets and tubes, including cold spray and cathodic arc-physical vapor deposition (CA-PVD). Neutronic analysis using the elemental compositions of the coating materials showed that the coating thicknesses should be less than 30 μm for most coating materials to avoid non-negligible economic penalty. The coated samples were autoclave tested in water at 360 Celsius degrees and 150 bar or in steam at 427 Celsius and 103 bar for corrosion resistance evaluation. The initial Ti{sub 2}AlC coating made by using large particle size MAX phase powder and cold spray deposition did not provide the desired corrosion resistance due to un-optimized microstructure and porosity. Subsequent optimization including reduction in particle size and tuning of cold spray deposition parameters resulted in denser coatings less than 30 microns in thickness. Optimization of process conditions resulted in reduced oxidation kinetics and showed that the coating acts as an effective barrier to prevent the zirconium substrate from oxidizing in accident condition temperatures. High purity Ti{sub 2}AlC powder is needed to further improve oxidation resistance of the Ti{sub 2}AlC coating. The oxidation resistance of pure monolithic MAX phase compound was also found to be remarkable, and the weight gain was only 1.85 mg/cm{sup 2} for 96 hours exposure in 1200 C. degrees steam, which is consistent with the literature. The corrosion resistance of the TiAlN/TiN multilayer coating was excellent. The weight gain was negligible and no oxidation was observed on the surface of zirconium-alloy at the coating-substrate interface. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

CFD Predictions of Severe Accident Natural Circulation Flows in a Combustion Engineering PWR

Computational fluid dynamics (CFD) is used to predict natural circulation flows between a reactor vessel upper plenum and a steam generator during specific severe accidents in a pressurized-water reactor (PWR). The predictions are based on a model for a simplified Combustion Engineering (CE) design. This work is part of a larger study to investigate the potential for severe accident induced steam generator (SG) tube failures in PWRs being conducted by the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research. The CFD model builds upon the lessons learned from previous studies of a generic Westinghouse PWR under similar conditions. The CE SG is significantly larger than the Westinghouse design and geometric differences are expected to reduce the mixing and increase the potential thermal challenge to the SG tubes. The CFD predictions are utilized to improve the basis for simplified flow and mixing models applied in one-dimensional severe accident codes. Parameters are determined to support system code modeling of hot leg and SG inlet plenum mixing, hot leg and SG tube bundle flow rates, and tube inlet temperatures. This updated modeling uses boundary conditions obtained from system analysis code predictions of severe accident behavior in a CE PWR. Model refinements include an improved tube bundle representation and improvements to the thermal boundary conditions. Results indicate less mixing and higher tube inlet temperatures for the CE design compared to previous Westinghouse studies. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Revenue Opportunities through Ancillary Services for Nuclear Power Plants with Integrated Firebrick Resistance- Heated Energy Storage (FIRES)

Deregulation of electricity markets has brought about major changes in the ways investments in and operation of power plants are planned and optimized. Although many NPPs have seen reduced revenue in recent years as power purchase agreement end and wholesale market electricity prices have remained low, there are opportunities to expand the technical capabilities of NPPs, participate in additional market services and increase revenue. This transaction summarizes the opportunities provided by participation in ancillary services markets potentially made possible by integration of thermal energy storage capacity, specifically the Firebrick Resistance-heated Energy Storage System (FIRES), into future nuclear power plants with high-temperature reactors and Nuclear Air-Brayton combined cycle (NACC) power conversion. The restructuring, or deregulation, of electric power systems in most of Europe and North America has had a substantial impact on the economic performance of currently operating nuclear power plants (NPPs) and the prospects for new builds. Owners of NPPs in some markets have expressed concern that low wholesale electricity market prices and insufficient compensation for capacity, reliability, and other services are threatening the financial viability of their plants. However, both current and future NPPs have the potential to increase their competitiveness by increasing the flexibility of their electricity output. Flexible output has previously only been possible in water-cooled reactors through load-following operation, which requires core power maneuvers and operation of the steam turbine at reduced efficiency. Integration of thermal energy storage systems (TESS) provides an alternative, providing the plant the option to either output electricity or store heat for greater later electric output. The plant owner may thus increase sales of electricity during hours when it is most valuable. Even more, some TESS technologies have the potential to greatly increase the net ramp rate capability of the plant. The ability to rapidly change net power output to the grid may make it possible for NPPs to secure additional revenue by providing additional minute-by-minute balancing services to the grid. FIRES is only one technology out of many that might provide additional technical capabilities at NPPs, and this discussion has only highlighted the opportunities in a single market. Analysis of other options, include steam accumulator systems, nitrate salt TESS, phase change material TESS, and opportunities in other markets are subjects of ongoing work. Deregulation of electricity markets has brought about major changes in the ways investments in and operation of power plants are planned and optimized. Although many NPPs have seen reduced revenue in recent years as power purchase agreements end and wholesale market electricity prices have remained low, there are opportunities to expand the technical capabilities of NPPs, participate in additional market services, and increase revenue. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

PGSFR General Arrangement

The Nuclear Island of the PGSFR is constructed upon the seismic isolation structure bearing a common basemat of the Reactor Building and the Auxiliary Building. The Reactor Building is divided into two different functional structures: 1) containment structure comprised of a operating floor, head access area (HAA), annulus area and etc., 2) the other areas accessible to reactor vault area outside of the containment structure via the Auxiliary Building. The arrangement of the Auxiliary Building is composed of symmetric divisional areas where there are two steam generators, heat exchangers of the Decay Heat Removal System (DHRS), facilities of the Intermediate Sodium Purification System (ISPS), the Primary Inert Gas Processing System (PIGPS), a main control room (MCR), electric equipment rooms, and main steam valve houses (MSVH), and etc. The general arrangement of the Reactor Building and the Auxiliary Building of the PGSFR is currently in development. Not only NSSS but supporting systems have not been finalized yet. Moreover, the other facility buildings such as a Turbine Generator Building, a fuel cycle facility, radwaste treatment facilities, and etc. have to be designed with relevant interfaces to the Reactor Building and the Auxiliary Building. We can expect additional modifications to optimize the general arrangement. However, the basic principles such as sodium and non-sodium segregation, convenient accessibility and maintenance, economics of architectural volume would be kept in the design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Overview of PGSFR Project

The long-term plan for advanced SFR development was authorized by Korean Atomic Energy Commission (KAEC) in 2008 and updated in 2011, which will be carried out toward the construction of a SFR prototype plant called PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) by 2028. The primary purpose of PGSFR is to demonstrate the technology reducing radioactive waste of spent nuclear fuel from nuclear power plants (NPP) in Korea. There are 24 NPPs in operation nationwide covering 30% of domestic electricity demand. The spent nuclear fuel is currently stored in the on-site storage pool of NPPs. Then, it has become a critical issue to ensure technical options to handle spent nuclear fuel in the long term basis to secure the sustainable operation of NPPs. PGSFR is a pool type sodium-cooled fast reactor with the power of 150 MWe or 400 MWth. PGSFR uses metallic fuel to make use of inherent safety characteristics of metallic fuel. The spent nuclear fuel will be processed using pyro-process to produce U-TRU-Zr fuel for PGSFR. But the initial core shall be loaded by low enriched U-Zr fuels. PGSFR has two intermediate loops with two intermediate heat exchangers (IHX) for each loop to deliver the generated heat from the core to power conversion system. There a steam generators producing super-heated steam in each loop. There are four decay heat removal circuits; two passive and two active ones. The reactor building is built on a seismic isolation base mat which can significantly mitigate seismic vibration of major components in the reactor building. The reactor building has a containment structure to isolate the reactor building under specific accident conditions and also provide structural integrity at the time of aircraft impact accident. PGSFR project is the national development program of Korea to construct the first SFR in Korea by 2028. The primary goal of PGSFR is the demonstration of reduction of radioactive waste from spent nuclear fuel by incineration in PGSFR. SFRA will flexibly cope with the licensing environment to minimize the uncertainty in licensing and safety concerns. The successful construction of PGSFR will bring Korean nuclear industry a step closer to guarantee the sustainable operation of NPPs nationwide and finally most importantly will solve the spent nuclear fuel issue of Korea. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Plant Operation and Control Strategy for the PGSFR

The plant operating strategy and control system design for the PGSFR plant are being developed to meet functional and design requirements for the plant. The functional requirements are derived from the mission of the reactor while the design requirements are related to the reactor technologies that underlie the PGSFR. This paper describes the operation and control strategy developed to meet these requirements and the implementation of that strategy. Representative results are presented for the dynamic behavior of the nuclear steam supply system (NSSS) during operational transients. A supervisory control system has been designed for the PGSFR for Startup, Reactor Leading, and Load Following modes and it performance assessed for operational transients in the Load Following mode. The control system extends into the balance of plant to include regulation of turbine inlet pressure and steam generator pressure and includes a feedwater pump with prototypic operating characteristics. The results indicate a very stable plant response with minimal deviation from quasi-static conditions for the transients simulated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Pool Boiling CHF Experiment for Cr-alloy-coated Accident Tolerant Fuel Cladding

The severe accident at the Fukushima nuclear power plants forced the nuclear communities to revise and reinforce their safety technology, especially accident tolerance. The main problem of Zr-based alloys is hydrogen generation owing to steam oxidation in a very high temperature environment. Currently, several types of Accident Tolerant Fuel (ATF) claddings that eliminate the steam oxidation problem have been presented as alternatives. Although the ATFs have shown appropriate performances in terms of material engineering, thorough tests including thermal hydraulic concerns are still required to ensure the design compatibilities. In this study, we conducted pool boiling critical heat flux (CHF) experiments using Cr-alloy-coated ATF claddings to verify basic thermal hydraulic performances. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2019

Collaboration for the Management of the Bruce Power Major Component Replacement (MCR) Nuclear Waste - 19630

Bruce Power's Major Component Replacement (MCR) Program will generate large volume waste via steam generators (SGs) and intermediate-level nuclear waste, which will require specialized packaging and handling to safely receive and store it at Ontario Power Generation's (OPG) Western Waste Management Facility (WWMF). The refurbishment of a Candu reactor involves, in part, the removal and replacement of all of a reactor core's fuel channel assemblies. Bruce Power and OPG collaborated to develop nuclear waste management solutions to optimize the use of OPG's existing Steam Generator Storage Building (SGSB) and Retube Component Storage Building (RCSB). By designing specialized Retube Waste Containers (RWCs) and optimizing configurations for the waste to be stored in the buildings, we will reduce the interim nuclear waste storage needs at the OPG WWMF, while improving Bruce Power MCR process efficiencies and reducing the overall costs of the MCR Program (Units 3 through 8). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Energy economic evaluation of hydrogen production by HTGR and solar tower - 18554

Changing the high temperature heat supply of energy intensive industries from today's mostly fossil sources to nuclear or renewable sources offers the opportunity of massive reductions of greenhouse gas emissions. Even though the HTGR as well as the solar tower are technically capable to supply heat at the required high temperature level, it is uncertain whether these technologies can compete with the fossil energy supply. Therefore, the aim of this paper is the assessment of the economic competitiveness of process heat supply by HTGR or solar tower to fossil fired processes using steam methane reforming as industrial example process. Applying a self-developed optimization model, energy economic analyzes are conducted for nuclear, solar and fossil energy supply systems. The results are benchmarked with the IAEA software HEEP and compared to literature values. The analyzes show that hydrogen generation costs of the solar tower system are higher than those of the HTGR system. Both technologies benefit from the support of a fossil heater, as the hydrogen generation costs can be significantly reduced even with a small amount of burned natural gas. Nevertheless, a steam methane reforming process run by HTGR or by solar tower is not yet competitive to the natural gas fired hydrogen production. (authors)

08 HYDROGEN↗