[en] In the early/mid 1990's Prof. Alvin Radkowsky, former chief scientist of the U.S. Naval Reactors program, proposed an alternate fuel concept employing thorium-based fuel for use in existing/next generation pressurized water reactors (PWRs). The concept was based on the use of a 'seed-blanket-unit' (SBU) that was a one-for-one replacement for a standard PWR assembly with a uranium-based central 'driver' zone, surrounded by a 'blanket' zone containing uranium and thorium. Therefore, the SBU could be retrofit without significant modifications into existing/next generation PWRs. The objective was to improve the proliferation and waste characteristics of the current once-through fuel cycle. The objective of a series of projects funded by the Initiatives for Proliferation Prevention program of the U.S. Department of Energy (DOE-IPP) - BNL-T2-0074,a,b-RU 'Radkowsky Thorium Fuel (RTF) Concept' - was to explore the characteristics and potential of this concept. The work was performed under several BNL CRADAs (BNL-C-96-02 and BNL-C-98-15) with the Radkowsky Thorium Power Corp./Thorium Power Inc. and utilized the technical and experimental capabilities in the Former Soviet Union (FSU) to explore the potential of this concept for implementation in Russian pressurized water reactors (VVERs), and where possible, also generate data that could be used for design and licensing of the concept for Western PWRs. The Project in Russia was managed by the Russian Research Center-?'Kurchatov Institute' (RRC-KI), and included several institutes (e.g., PJSC 'Electrostal', NPO 'LUCH' (Podolsk), RIINM (Bochvar Institute), GAN RF (Gosatomnadzor), Kalininskaja NPP (VVER-1000)), and consisted of the following phases: Phase-1 ($550K/$275K to Russia): The objective was to perform an initial review of all aspects of the concept (design, performance, safety, implementation issues, cost, etc.) to confirm feasibility/viability and identify any 'show-stoppers'; Phase-2 ($600K/$300K to Russia): Continued the activities initiated under Phase-1 with a focus on expanded design and safety analyses, and to address fuel fabrication and testing issues; and, Phase-3 ($300K/$290K to Russia): Focus on thermal-hydraulic testing at Kurchatov for both VVER and PWR lattices

[en] A systematic evaluation has been conducted of the potential for advanced nuclear fuel cycle strategies and options to address the issues ascribed to the use of nuclear power. Issues included nuclear waste management, proliferation risk, safety, security, economics and affordability, and sustainability. The two basic strategies, once-through and recycle, and the range of possibilities within each strategy, are considered for all aspects of the fuel cycle including options for nuclear material irradiation, separations if needed, and disposal. Options range from incremental changes to today's implementation to revolutionary concepts that would require the development of advanced nuclear technologies.

[en] Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U_3Si_5 fuels have the potential to exhibit reactor physics and fuel management performance similar to UO_2. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO_2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO_2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO_2–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U_3Si_5 fuels with Kanthal AF or APMT cladding. The objective of the U_3Si_5 phase in the UN–U_3Si_5 fuel concept is to shield the nitride phase from water. It was shown that UN–U_3Si_5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO_2–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to "1"4N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO_2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in these assessments are preliminary, and that additional data are necessary for these materials, most significantly under irradiation

[en] An accelerator-driven spallation neutron-source target/blanket system has been developed for the production of tritium. The system employs a proton linear accelerator, and a lead neutron-producing target, surrounded by tritium-producing blankets based on the lithium-aluminum technology employed at Savannah River for tritium production since the 1950's. The target/blanket configuration is referred to as the SILC target, for Spallation-Induced Lithium Conversion. In this concept, tritium is produced without the presence of fissionable materials; therefore, no high-level waste is produced, and the ES and H concerns are significantly reduced compared to reactor systems. A preconceptual design has been completed for the SILC target, and the attractive performance and ES and H characteristics demonstrated. Experiments were also performed in support of the target design to confirm materials performance, safety, and to validate the nuclear design methodology

[en] Highlights: • Fully ceramic microencapsulated (FCM) fuel was studied. • Sensitivity and benchmark calculations were performed on lattice level for FCM fuel. • A non-linear reactivity model was vital to represent the reactivity during burn-up. • Reactivity coefficients were determined for FCM versus conventional UO_X fuel. - Abstract: The fully ceramic microencapsulated (FCM) fuel concept is based on the tri-isotropic (TRISO) carbon coated fuel particles. These particles were developed and demonstrated for use in high temperature gas reactors. It has been proposed to use these particles in light water reactors to provide potential operational and safety benefits. The reference fuel in this case assumes TRISO-like particles with a ∼20%-enriched uranium-nitride kernel embedded in a silicon carbide (SiC) matrix. The fuel particles are contained in a “compact” which is then inserted into a cladding. The fuel assembly features the same dimensions as a standard 17 × 17 Westinghouse fuel assembly. FCM fuel requires fission products to traverse several barriers in the proposed fuel design before reaching the cladding. FCM fuel may also reduce fuel-cladding interaction and fuel pellet swelling while enabling higher fuel burn-up. This study is a neutronic evaluation of the use of FCM fuel in an advanced pressurized water reactor (PWR). On the lattice level, the SERPENT Monte Carlo and TRITON deterministic tools were used, while the whole core simulation was based on the three-dimensional PARCS nodal code. This paper presents the results of the lattice-level neutronic study of doubly heterogeneous FCM fuel. Strong agreement was found between the SERPENT and TRITON codes in terms of k-infinity as a function of burn-up, actinide build-up, and “pin” powers. The impact of several simplifying geometric assumptions was considered, such as the use of a square particle lattice within the FCM fuel pins. It was determined that the linear reactivity model does not provide a good estimate of the fuel cycle length, due primarily to non-linear reactivity behavior at high burn-up (>800 effective full power days). To determine cycle length, higher order reactivity models were applied to the lattice results. The calculated cycle lengths are slightly reduced versus a reference uranium oxide case. Finally, the assembly-level reactivity coefficients were calculated as a function of burn-up. The fuel and moderator temperature coefficients were negative for FCM fuel, but reduced in magnitude by approximately 50% versus a reference uranium oxide case

[en] Highlights: • Validated SERPENT–PARCS for analysis of fully ceramic microencapsulated (FCM) fuel. • Modified PARCS to incorporate UN and SiC thermo-physical properties. • Core kinetics simulation of an RIA for a PWR with conventional UOX or FCM fuel. • The results for the core power transient are similar for UOX or FCM fuel. • Sensitivity of the RIA response to FCM thermal model was quantified. - Abstract: The fully ceramic microencapsulated (FCM) fuel is based on the tri-isotropic (TRISO) carbon coated fuel particles. These particles were developed and demonstrated for use in high temperature gas reactors. It has been proposed to use these particles in light water reactors to provide potential operational and safety benefits. The reference fuel in this case assumes TRISO-like particles with a ∼20%-enriched uranium-nitride kernel embedded in a silicon carbide (SiC) matrix. The fuel particles are contained in a “compact” which is then inserted into a cladding. The fuel assembly features the same dimensions as a standard 17 × 17 Westinghouse fuel assembly. FCM fuel requires fission products to traverse several barriers in the proposed fuel design before reaching the cladding. FCM fuel may also reduce fuel-cladding interaction and fuel pellet swelling while enabling higher fuel burn-up. This study is a neutronic evaluation of the use of FCM fuel in an advanced pressurized water reactor (PWR). On the lattice level, the SERPENT Monte Carlo and TRITON deterministic tools were used, while the whole core simulation was based on the three-dimensional PARCS nodal code. The present paper focuses on two of the issues associated with this proposed implementation: specifically the development of a reasonable reference full-core model of an advanced PWR with FCM fuel and the response of the PWR to a reactivity insertion accident (RIA). This work addresses the issues of the increased power density and transients that occur on short time-scales in a PWR. In this case, the RIA takes the form of a control rod ejection for a typical PWR reactor. This results in a sudden increase in power and a corresponding increase in fuel kernel temperature. In the case of a PWR, this response is more demanding than in the case of a gas-cooled reactor, because the kinetic parameters and feedback coefficients of the two reactors are quite different. The parameters for the fuel and matrix material in the PARCS thermal–hydraulic module were modified to reflect the different geometry and materials. Preliminary data for both un-irradiated and irradiated SiC were obtained from the literature and included in the analyses. A super prompt critical RIA produces an average energy deposition (<124.6 J/g) that is estimated for different simplified thermal representations of the FCM fuel pin

[en] This paper presents results and conclusions derived from an integrated analysis of a NERVA based Nuclear Thermal Propulsion (NTP) system. The NTP system is sized to generate a thrust of 70,000 N (15,000 lbf), and have a specific impulse (Isp) of 860 s. This implies a reactor that operates at 350 MWth and has a mixed mean propellant outlet temperature of 2760 K. The integrated analysis will require that self-consistent neutronic/thermal-hydraulic/stress analyses be carried out. The major code packages used in this analysis are MCNP, RELAP, and ANSYS. Results from this analysis indicate that nuclear data will have to be re-generated to cover the wide temperature range, zone loading will be necessary to avoid entering the liquidus region for the fuel, and the effectiveness of the ZrC insulator will have implications for bi-modal applications. These results suggest a path forward in the development of a viable NTP system based on a NERVA reactor should initially concentrate on fuel and structural materials and associated coating development. A series of safety related criticality determinations were carried out addressing water immersion following a launch incident

[en] A new concept; termed ADAPT; for the rapid and virtually complete burning of plutonium is described. ADAPT employs a high current CW linear accelerator (linac) to generate neutrons in a lead/D2O target. The neutrons are then absorbed in a surrounding subcritical (Keff∼0.95) blanket assembly, that holds small (∼0.5 cm diameter) graphite beads containing the plutonium to be burned. The graphite beads are coated and sealed to contain all fission products, including the noble gases. After destruction of virtually all (≥90%) of the original plutonium loading, the fuel beads are discharged and sent to a geologic repository for ultimate disposal. (c) 1995 American Institute of Physics

[en] The homogeneous ThO₂-UO₂ fuel cycle option for a pressurized water reactor (PWR) of current technology is investigated. The fuel cycle assessment was carried out by calculating the main performance parameters: natural uranium and separative work requirements, fuel cycle cost, and proliferation potential of the spent fuel. These performance parameters were compared with a corresponding slightly enriched (all-U) fuel cycle applied to a PWR of current technology. The main conclusion derived from this comparison is that fuel cycle requirements and fuel cycle cost for the mixed Th/U fuel are higher in comparison with those of the all-U fuel. A comparison and analysis of the quantity and isotopic composition of discharged Pu indicate that the Th/U fuel cycle provides only a moderate improvement of the proliferation resistance. Thus, the overall conclusion of the investigation is that there is no economic justification to introduce Th into a light water reactor fuel cycle as a homogeneous ThO₂-UO₂ mixture

[en] Highlights: • Fuel cycle performance of microreactors is not as good as larger reactors. • Heat pipe reactor concepts are best suited to serve as a nuclear battery rather than a centralized power source. • Reducing neutron leakage and parasitic absorption improves fuel cycle performance. - Abstract: The demand for more practical and innovative nuclear reactor designs capable of providing clean energy to meet new global demands is revolutionizing the nuclear industry. In this paper we present the fuel cycle and neutronics analysis of a design expected to be similar to the Westinghouse eVinci™ heat pipe (HP) reactor. The concept considered here uses low enriched urania rods and potassium liquid metal-cooled HPs to remove heat from the core. The fuel and HPs are contained in steel monolith to form the core which is surrounded by an alumina reflector. There is no need for forced circulation to remove heat from the active core region, thus eliminating the use of pumps, valves and tube piping. The concept is designed to operate safely and provide a reliable autonomous power supply in support of off-grid missions. This study shows that the HP reactor design can operate for more than 10 years without refueling. Further, we show that the current design of the HP reactor concept is best suited to serve as a nuclear battery rather than a centralized power source. We also studied the nuclear fuel cycle performance of the HP reactor concept in a once-through fuel cycle. The natural resource utilization, waste output, and environmental impact, as defined in a reference study from the literature, did not perform as well on a GWe per year energy basis compared to light water reactors with less than 5% enriched uranium in a once-through fuel cycle. Parasitic absorption in the steel monolith structure and neutron leakage lower the reactivity of the core, which decreases the discharge burnup of the fuel. Analysis of modifications of the monolith material and size of the reference core configuration showed that the neutron leakage impact on the small sized HP core is the most limiting factor on the fuel cycle performance.