Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen production [DOE 2004] and energy conversion technologies programs are described elsewhere.

A novel approach to fabricating fuel compacts for the next generation nuclear plant (NGNP)

The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures for on-site welding, post-weld heat treatment (PWHT), and inspections will be required for the materials of construction. High-importance phenomena related to the RPV include crack initiation and subcritical crack growth; field fabrication process control; property control in heavy sections; and the maintenance of high emissivity of the RPV materials over their service lifetime to enable passive heat rejection from the reactor core. All identified phenomena related to the materials of construction for the IHX, RPV, and other components were evaluated and ranked for their potential impact on reactor safety.

This report provides supplemental information to the assessment of target markets provided in Appendix A of the 2012 Next Generation Nuclear Plant (NGNP) Industry Alliance (NIA) business plan [NIA 2012] for deployment of High Temperature Gas-Cooled Reactors (HTGRs) in the 2025 – 2050 time frame. This report largely reiterates the [NIA 2012] assessment for potential deployment of 400 to 800 HTGR modules (100 to 200 HTGR plants with 4 reactor modules) in the 600-MWt class in North America by 2050 for electricity generation, co-generation of steam and electricity, oil sands operations, hydrogen production, and synthetic fuels production (e.g., coal to liquids). As the result of increased natural gas supply from hydraulic fracturing, the current and historically low prices of natural gas remain a significant barrier to deployment of HTGRs and other nuclear reactor concepts in the U.S. However, based on U.S. Department of Energy (DOE) Energy Information Agency (EIA) data, U.S. natural gas prices are expected to increase by the 2030 – 2040 timeframe when a significant number of HTGR modules could be deployed. An evaluation of more recent EIA 2013 data confirms the assumptions in [NIA 2012] of future natural gas prices in the range of approximately $7/MMBtu to $10/MMBtu during the 2030 – 2040 timeframe. Natural gas prices in this range will make HTGR energy prices competitive with natural gas, even in the absence of carbon-emissions penalties. Exhibit ES-1 presents the North American projections in each market segment including a characterization of the market penetration logic. Adjustments made to the 2012 data (and reflected in Exhibit ES-1) include normalization to the slightly larger 625MWt reactor module, segregation between steam cycle and more advanced (higher outlet temperature) modules, and characterization of U.S. synthetic fuel process applications as a separate market segment.

The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700 °C. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.

The Next Generation Nuclear Plant (NGNP) Project is tasked with selecting a high temperature gas reactor technology that will be capable of generating electricity and supplying large amounts of process heat. The NGNP is presently being designed as a helium-cooled high temperature gas reactor (HTGR) with a large graphite core. The graphite baseline characterization project is conducting the research and development (R&D) activities deemed necessary to fully qualify nuclear-grade graphite for use in the NGNP reactor. Establishing nonirradiated thermomechanical and thermophysical properties by characterizing lot-to-lot and billet-to-billet variations (for probabilistic baseline data needs) through extensive data collection and statistical analysis is one of the major fundamental objectives of the project. The reactor core will be made up of stacks of graphite moderator blocks. In order to gain a more comprehensive understanding of the varying characteristics in a wide range of suitable graphites, any of which can be classified as “nuclear grade,” an experimental program has been initiated to develop an extensive database of the baseline characteristics of numerous candidate graphites. Various factors known to affect the properties of graphite will be investigated, including specimen size, spatial location within a graphite billet, specimen orientation within a billet (either parallel to [P] or transverse to [T] the long axis of the as-produced billet), and billet-to-billet variations within a lot or across different production lots. Because each data point is based on a certain position within a given billet of graphite, particular attention must be paid to the traceability of each specimen and its spatial location and orientation within each billet. The evaluation of these properties is discussed in the Graphite Technology Development Plan (Windes et. al, 2007). One of the key components in the evaluation of these graphite types will be mechanical testing on specimens drawn from carefully controlled sections of each billet. To this end, this report will discuss the machining of the first set of test specimens that will be evaluated in this program through tensile, compressive, and flexural testing. Validation that the test specimens have been produced to the tolerances required by the applicable ASTM standards, and to the quality control levels required by this program, will demonstrate the viability of sending graphite to selected suppliers that will provide valuable and certifiable data to future data sets that are integral to the NGNP program and beyond.

The U.S. Department of Energy (DOE) has selected the high temperature gas-cooled reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for process heat, hydrogen and electricity production. The reactor will be graphite moderated with helium as the primary coolant and may be either prismatic or pebble-bed. Although, final design features have not yet been determined. Research and Development (R&D) activities are proceeding on those known plant systems to mature the technology, codify the materials for specific applications, and demonstrate the component and system viability in NGNP relevant and integrated environments. Collectively these R&D activities serve to reduce the project risk and enhance the probability of on-budget, on-schedule completion and NRC licensing. As the design progresses, in more detail, toward final design and approval for construction, selected components, which have not been used in a similar application, in a relevant environment nor integrated with other components and systems, must be tested to demonstrate viability at reduced scales and simulations prior to full scale operation. This report and its R&D TDRMs present the path forward and its significance in assuring technical readiness to perform the desired function by: Choreographing the integration between design and R&D activities; and proving selected design components in relevant applications.

The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 ºC to perhaps 1000 ºC. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U. S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present report presents results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made.

Hybrid sulfur cycle flowsheets for hydrogen production using high-temperature gas-cooled reactors

The ceramic fuel and core, inert helium coolant and superior safety characteristics of the modular High-Temperature Gas-cooled Reactor (HTGR) are the bases for its potential to displace the fossil energy resources that are presently dominant in industrial process steam and cogeneration applications. In achieving this potential, a particular challenge is to provide a flexible and efficient coupling between the high-temperature HTGR heat source and various industrial processes. This challenge was addressed in the course of developing the conceptual design of the Steam Cycle-Modular Helium Reactor for the Next Generation Nuclear Plant (NGNP) Project. This paper characterizes the energy requirements of high-potential process steam and cogeneration applications and describes the development of a steam utilization architecture that can serve a wide range of industrial applications with superior flexibility and efficiency.Copyright © 2012 by ASME

Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.

High temperature gas cooled reactor (HTGR) has been considered to be the most promising option to meet energy demands in the future. It has also been selected as the next generation nuclear plant. The primary safety requirement of the next generation nuclear plant design is to limit radioactive material releases to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary. The purpose of this study is to evaluate the safety design of HTGRs in order to fulfill the requirement of the next generation nuclear plant. To achieve this objective, inherent safety features, fundamental safety functions, and confinement functions realized into the design of HTGRs are comprehensively evaluated. It is found that design provisions of HTGRs can fulfill the intention of keeping radionuclides at their original sources. The layers of the coated fuel particles are very robust to retain nuclear fission products for all foreseeable reactivity events. There will be no possibility of radioactive materials to be released even though related safety systems and operator intervention are not involved in the recovery actions. This design has complied with the requirement of the next generation nuclear plant, which is to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary.Keywords: High temperature gas cooled reactor, inherent safety features, fundamental safety functions, confinement functions, next generation nuclear plant

Recently selected for prototyping per the Department of Energy’s Next Generation Nuclear Plant (NGNP) initiative, the very-high temperature reactor (HTGR) is a Generation IV reactor concept which employs helium as a coolant medium at temperatures up to 1000°C. Within the core of prismatic HTGRs, helium flows through thousands of coolant channels which serve as conduits through vertical columns of active hexagonal fuel blocks. Recent studies have focused on experimental and numerical modeling of the lower plenum where coolant jets enter at different temperatures due to non-uniform heat generation in the core and at different velocities due to flow mal-distribution in the coolant channels themselves. This already complex flow is further complicated by bypass flow through the gaps between neighboring hexagonal fuel columns. However, in order to truly quantify the turbulent velocity and temperature fluctuations in a realistic lower plenum environment, both the mean velocity profiles and the spectrum of turbulent fluctuations at the jet inlets need to be provided. These will, in turn, depend on the complicated flow paths upstream of the jet (hundreds of coolant channels collating into a single jet which also interacts with core bypass flow). Before any conceptual HTGR design can be successfully implemented, it is critical that the complex interaction between core bypass and coolant jet flows be fully captured and understood. To this end, the work herein aims to provide empirical data and computational models of the thermal-hydraulics phenomena present upstream of the lower plenum inlet. This investigation will ultimately aid in quantifying the effect core bypass flow has on temperature fluctuations within the support structures present in the lower plenum. This work provides a critical link between two heavily investigated areas of the HTGR (active fuel region and lower plenum) and serves as a benchmark with which to evaluate future numerical thermal-hydraulics models.

Modeling the high-temperature creep behaviors of 7075 and 2124 aluminum alloys by continuum damage mechanics model

As part of the Next Generation Nuclear Plant (NGNP) project, an evaluation of a low-temperature heat-pump interface design for a nuclear-driven high-temperature electrolysis (HTE) hydrogen production plant was performed using the UniSim process analysis software. The low-temperature interface design is intended to reduce the interface temperature between the reactor power conversion system and the hydrogen production plant by extracting process heat from the low temperature portion of the power cycle rather than from the high-temperature portion of the cycle as is done with the current Idaho National Laboratory (INL) reference design. The intent of this design change is to mitigate the potential for tritium migration from the reactor core to the hydrogen plant, and reduce the potential for high temperature creep in the interface structures. The UniSim model assumed a 600 MWt Very-High Temperature Reactor (VHTR) operating at a primary system pressure of 7.0 MPa and a reactor outlet temperature of 900°C. The low-temperature heat-pump loop is a water/steam loop that operates between 2.6 MPa and 5.0 MPa. The HTE hydrogen production loop operated at 5 MPa, with plant conditions optimized to maximize plant performance (i.e., 800°C electrolysis operating temperature, area specific resistance (ASR) = 0.4 ohm-cm2, and a current density of 0.25 amps/cm2). An air sweep gas system was used to remove oxygen from the anode side of the electrolyzer. Heat was also recovered from the hydrogen and oxygen product streams to maximize hydrogen production efficiencies. The results of the UniSim analysis showed that the low-temperature interface design was an effective heat-pump concept, transferring 31.5 MWt from the low-temperature leg of the gas turbine power cycle to the HTE process boiler, while consuming 16.0 MWe of compressor power. However, when this concept was compared with the current INL reference direct Brayton cycle design and with a modification of the reference design to simulate an indirect Brayton cycle (both with heat extracted from the high-temperature portion of the power cycle), the latter two concepts had higher overall hydrogen production rates and efficiencies compared to the low-temperature heat-pump concept, but at the expense of higher interface temperatures. Therefore, the ultimate decision on the viability of the low-temperature heat-pump concept involves a tradeoff between the benefits of a lower-temperature interface between the power conversion system and the hydrogen production plant, and the reduced hydrogen production efficiency of the low-temperature heat-pump concept compared to concepts using high-temperature process heat.