Accident Tolerant Materials for Light Water Reactor Fuels provides a description of what an accident tolerant fuel is and the benefits and detriments of each concept. The book begins with an introduction to nuclear power as a renewable energy source and the current materials being utilized in light water reactors. It then moves on to discuss the recent advancements being made in accident tolerant fuels, reviewing the specific materials, their fabrication and implementation, environmental resistance, irradiation behavior, and licensing requirements. The book concludes with a look to the future of new power generation technologies. It is written for scientists and engineers working in the nuclear power industry and is the first comprehensive work on this topic. - Introduces the fundamental description of accident tolerant fuel, including fabrication and implementation - Describes both the benefits and detriments of the various Accident Tolerant Fuel concepts - Includes information on the process of materials selection with a discussion of how and why specific materials were chosen, as well as why others failed

Declassified 12 Sep 1973. Development of UN in four major areas is presented: 1) powder synthesis and powder metallurgical fabrication, 2) physical and mechanical property testing, 3) fuel-cladding compatibility and barrier development, and 4) irradiation evaluation. Synthesis and sintering methods for producing large quantities of high-densily and high-purity UN are described. Results of experimental studies of the following properties are summarized: melting point, thermal conductivity, thermal expansion, vapor pressure, and thermodynamics, heat capacity, hot hardness, compressive creep, nitrogen self- diffusion, and electrical resistivity. Compatibility tests are described which demonstrated the need for a diffusion barrier. Lithium soak tests for up to 11,000 h at 2200 deg F demonstrated the stability and practicality of vapor- deposited tungsten-lined Nb--1 Zr alloy over the projected life and temperature of SNAP50. Similar static tests of purposely defected simulated fuel pins indicate a relatively high degree of stability of UN towards lithium. Instrumented capsule irradiation tests of simulated Nb--1 Zr alloy - clad fuel pins are described under 2 Mw(t) and 8 Mw(t) SNAP-50 conditions. Under 8 Mw(t) conditions, 20% fission gas release and 2% diametral cladding growth were observed after 2750 h at 2200 deg F (2.0 at.% U burnup). In-pile operation under 2 Mw(t) conditions was achieved for 5940 h at 2000 deg F (1.0 at.% U burnup) and 3360 h at 2200 deg F (1.5 at.% U burnup) while experiencing less than 0.2% fission gas release and less than 0.4% diametral growth. (66 figures) (auth).

Declassified 18 Sep 1973. Uranium nitride was developed as a high- temperature fuel for liquid metal-cooled reactors. The development included fabrication studies, environmental compatibility, thermal stability, hot hardness, and irradiation testing. Besides pure UN, mixtures of UN and ZrN were also investigated. (DLC).

The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E & amp;E) and Chemistry & Material Sciences (C & amp;MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E & E and C & MS Directorates co-sponsored this Laboratory Directed Research & Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. In fact, three out of the six GEN-IV reactor concepts consider using the nitride-based fuel, as shown in Table 1. SSTAR is a liquid-metal cooled, fast reactor. It uses nitride fuel in a sealed reactor vessel that could be shipped to the user and returned to the supplier having never been opened in its long operating lifetime. This sealed reactor concept envisions no fuel refueling nor on-site storage of spent fuel, and as a result, can greatly enhance proliferation resistance. However, the requirement for a sealed, long-life core imposes great challenges to research and development of the nitride fuel and its cladding. Cladding is an important interface between the fuel and coolant and a barrier to prevent fission gas release during normal and accidental conditions. In fabricating the nitride fuel rods and assemblies, the cladding material should be selected based on its the coolant-side corrosion properties, the chemical/physical interaction with the nitride fuel, as well as their thermal and neutronic properties. The US NASA space reactor, the SP-100 was designed to use mono-uranium nitride fuel. Although the SP-100 reactor was not commissioned, tens of thousand of nitride fuel pellets were manufactured and lots of them, cladded in Nb-1-Zr had been irradiated in fast test reactors (FFTF and EBR-II) with good irradiation results. The Russian Naval submarines also use nitride fuel with stainless steel cladding (HT-9) in Pb-Bi coolant. Although the operating experience of the Russian submarine is not readily available, such combination of fuel, cladding and coolant has been proposed for a commercial-size liquid-metal cooled fast reactor (BREST-300). Uranium mono-nitride fuel is studied in this LDRD Project due to its favorable properties such as its high actinide density and high thermal conductivity. The thermal conductivity of mono-nitride is 10 times higher than that of oxide (23 W/m-K for UN vs. 2.3 W/m-K for UO{sub 2} at 1000 K) and its melting temperature is much higher than that of metal fuel (2630 C for UN vs. 1132 C for U metal). It also has relatively high actinide density, (13.51 gU/cm{sup 3} in UN vs. 9.66 gU/cm{sup 3} in UO{sub 2}) which is essential for a compact reactor core design. The objective of this LDRD Project is to: (1) Establish a manufacturing capability for uranium-based ceramic nuclear fuel, (2) Develop a computational capability to analyze nuclear fuel performance, (3) Develop a modified UN-based fuel that can support a compact long-life reactor core, and (4) Collaborate with the Nuclear Engineering Department of UC Berkeley on nitride fuel reprocessing and disposal in a geologic repository.