Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “FISSILE”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 271 records · Page 15

Seventy-five years of nuclear criticality safety documents - a bibliography

A collection of nuclear criticality documents was begun in the 1970's at the University of California's Lawrence Livermore Laboratory (now Lawrence Livermore National Laboratory). Documents in this bibliography are dated as early as 1940. The early documents described experimental work that was of vital importance to the safe use of fissile materials at nuclear reactors, the facilities for preparing or transporting nuclear fuel, or its eventual storage. The experiments were of direct value to support safe operations with fissile materials but they were also needed for validating early physics data and calculational methods and then for validating computational methods used in criticality safety evaluations. These documents provided information for our computerized bibliographic database. The bibliography was originally intended for use by Livermore staff members but criticality safety 'pioneers' encouraged us to make the database available to the broader nuclear safety community and even shared copies of their bibliographies and reports. A publication of the bibliography is being prepared and a searchable database is available on the U.S. Department of Energy's Nuclear Criticality Safety Program's NCSP web site. The NCSP web site also includes links from the bibliography to many full-text documents. The database continues to expand--the information in this paper is based on a 'snapshot' of its present status. The 'vintage' hardcopy document collection is preserved and is being scanned into digital format. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Seventy-five years of nuclear criticality safety documents - A bibliography, part II - Links to full-text documents

A collection of nuclear criticality documents was begun in the 1970's at the University of California's Lawrence Livermore Laboratory (now Lawrence Livermore National Laboratory). Documents in this bibliography are dated as early as 1940. The early documents described experimental work that was of vital importance to the safe use of fissile materials at nuclear reactors, the facilities for preparing or transporting nuclear fuel, or its eventual storage. The experiments were of direct value to support safe operations with fissile materials but they were also needed for validating early physics data and calculational methods and then for validating computational methods used in criticality safety evaluations. These documents provided information for our computerized bibliographic database which was originally published in hard-copy format. Our original 'vintage' hardcopy document collection is preserved and is being scanned into digital format. The bibliography is available online as 'LLNL Database' within the U.S. Department of Energy's Nuclear Criticality Safety Program's NCSP web site. Bibliographic information and full-text documents continue to be downloaded from online databases. The online bibliography now contains close to 23,000 records. Links are available to over 3600 full-text documents within NCSP. This capability will be demonstrated at this conference. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Determination of critical experiment correlations for experiments involving arrays of low-enriched fuel rods

Quantifying correlations among critical experiments has become a topic of interest over the last 10-15 years. Correlations arise from the shared use of fissile material, non-fissile materials, fixtures, and procedures in multiple experiments or multiple experiment series. A Monte-Carlo sampling technique proposed as a viable technical method to evaluate these correlations has been implemented in the SCALE 6.2 Sampler sequence. This paper presents an investigation of the use of this new tool and technique for low-enriched uranium pin array experiments with a thermal neutron energy spectrum. The Nuclear Energy Agency's Expert Group on Uncertainty Analyses for Criticality Safety Assessments has been examining methods of calculating the correlation coefficients among experiments from the LCT-007 and LCT-039 series. Results for two scenarios are presented to demonstrate the impact of pin pitch uncertainty modeling on the correlation coefficients. The convergence of the correlation coefficients is also investigated. Additional studies examine the impact of the stochastic uncertainty and the use of continuous-energy Monte-Carlo for individual realizations on the correlation coefficients. Correlation coefficients are also determined for the LCT-042 experiment series. This experiment series differs from other experiments because unique poison panels and fuel array separation distances may reduce correlation of the individual experiments within the series. The effects of pin pitch uncertainty and stochastic uncertainty are also examined for this series. Comparing the results of the two sets of experiments shows potential conclusions that can be drawn regarding viability of the Monte-Carlo sampling technique for determining critical experiment correlations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Determination of critical experiment correlations for experiments involving highly enriched uranium solutions

Quantifying the correlations among critical experiments has become a topic of interest over the last 10-15 years. The correlations arise from the shared use of fissile material, non-fissile materials, fixtures, and procedures in multiple experiments or multiple experiment series. A Monte Carlo sampling technique proposed as a viable technical method to evaluate these correlations has been implemented within the SCALE 6.2 Sampler sequence. This paper presents the investigation of the use of this new tool and technique for highly enriched uranium solution experiments. One of the first attempts to quantify the correlations among critical experiments examined a series of highly enriched solution experiments, including the HST-001 series. These correlations are included in the International Criticality Safety Benchmark Evaluation Project Handbook, and they form a basis for methodology comparisons. Results generated using the Sampler implementation are compared with these previously published results. Additional studies involving the HST-001 experiment series are also presented. A range of different assumptions of the independence of the geometry components, enrichments, and other material compositions are implemented to investigate the impact of these assumptions on the correlation coefficients. These effects are investigated with these experiments because the calculation time is shorter for the solution experiments than for the low-enriched lattice experiments discussed in a companion paper. The magnitude of the effects of different independence assumptions on the correlation coefficients is important to establish which parameters must be clearly identified as shared or unique for accurate correlation determinations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Consistent criticality safety approach to nuclear sites, external transport, and final disposal

A general issue and some selected topics discussed in international projects (ISO, IAEA, OECD) are presented. Long and broad experience in criticality safety is applied to find consistent safety principles for nuclear sites, external transport and final disposal. Support of standards and regulations is a main purpose. An ISO 1709 (criticality safety principles) draft proposes expansion of the scope to transport. Research on the basis for the IAEA Transport Regulations provides insights. Those Regulations are based on criticality safety principles. A general requirement for prevention of criticality during transport is combined with a prescriptive set of requirements on how to get regulatory approval. 'Confinement system' was introduced in 1996 to separate subcriticality from radioactive material containment specifications. The much older intention was to provide technical support, but not safety, for the inter-phases between sites, transport and emergency. The application has changed unintentionally and significantly. A solution is proposed. The principle for determination of a threshold fissile nuclide mass limit for fissile material is used to support acceptability of rounding a Criticality Safety Index value down to zero, similar to determination of the Transport Index. Temperature effects on nuclear properties need to be accounted for. A -40 to 800 Celsius degree temperature range outside a package applies to transport. Burnup credit benchmarks involve high temperatures (depletion) and interpretation of operating reactor measurements for typical water pools (20 to 120 C. degrees). KRITZ critical experiments between 20 and 250 Celsius degrees are currently being evaluated. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

An evaluation of in-line component process modules

The purpose of this paper is to discuss the analysis of abnormal configurations of in-line components, tanks and pumps in support of the integration of Criticality Safety into the Uranium Processing Facility (UPF) at Y-12 National Security Complex (Y-12). In-line components are components that are used in solution piping systems, including instrumentation, valves, pumps, and other general-use equipment. Mounted on self-contained skids/modules, in-line components (instruments, seal pots, level switches, etc.), tanks, and pumps are associated with a wide variety of the UPF processes. Using computer generated models expected to bound abnormal conditions for process module configurations, this report evaluates the impact of loss of spacing between in-line components in in-line component arrays and the effects of abnormal events. All fissile materials are considered to have a {sup 235}U enrichment of 100% and full theoretical density of 18.81 g/cm{sup 3} to conservatively represent the maximum credible enrichment available. Water and concrete were utilized to account for the construction materials associated with the skids and other reflective surfaces. Uranyl nitrate, was evaluated over the range of 0 to 550 g U/L for various normal and abnormal configurations of components mounted on skids. Abnormal events included fissile spills, sprinkler activation, and loss of spacing due to a seismic event. The results from this analysis provide a basis that supports development of nuclear safety design criteria, which generates standardized volume limits expected to bound the actual internal volume of the planned components. The results also provide a basis for consistent Nuclear Criticality Safety (NCS) control of in-line components, pumps, and tanks, for the ultimate development of Criticality Safety Process Studies in the UPF. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Characteristics of concrete reflection of {sup 235}U systems

Water reflection is considered in nuclear criticality safety evaluations because it is readily available and easily forms a close-fitting reflector around fissile material. Concrete is commonly encountered in industrial environments and can be a more effective reflector than water. A comprehensive study using 32 concrete compositions derived from an extensive literature survey was done for the three basic one-dimensional geometries used in radiation transport codes (spherical, infinite length cylindrical, and infinite extent slab) using a {sup 235}U metal/water fissile mixture. The concentration/moderation range over which criticality is possible was spanned, and the change in k{sub eff} per unit change in element atomic number density for various concrete composition elements was computed to determine the relative effect of each element on neutron multiplication. The results show that: -1) the most effective reflector is not the same for each geometry; -2) in general, for each geometry, Δk{sub eff}/ΔN has similar values for {sup 235}U concentrations >0.200 g/cm{sup 3}; and -3) for {sup 235}U concentrations < 0.200 g/cm{sup 3}, the value of Δk{sub eff}/ΔN diminishes toward zero.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Impacts of Recycling Time on Fuel Cycle Transition

The purpose of this work is to show the importance of assumptions regarding the recycling time for time-dependent fuel cycle scenarios that involve transition from the current state of the U.S. fuel cycle to cycles based on continuous recycling via fast reactors. The recycling time is defined here as the total time between when the fuel is discharged from the reactor core and when its recovered heavy metals (HM) are recycled back into the core as part of the fresh/new fuel. The main processes occurring within the recycling time include cooling, separations, and fuel fabrication. The minimum recycling time required for fast reactor fuel largely depends on the minimum cooling time, which will be longer if discharged fuel needs to be moved to a centrally-located separations facility. A longer cooling time is needed for the heat load and fuel characteristics to reach safer levels for handling and transportation. Therefore, if the separation and fabrication facilities were co-located with reactors on-site, then the total recycling time could be reduced by several years. For fuel cycle transition scenarios in which external fissile materials (Pu or TRU) are limiting, these longer recycling times may prolong the transition to the continuous recycling system of fast reactors as well as impact the fuel cycle system performance during the transition. Therefore, in order to demonstrate the quantitative impacts of these recycling time assumptions, an example scenario of transition to the fast reactor fleet with continuous recycle was modeled in the fuel cycle systems code DYMOND. It must be emphasized that this is a point study that demonstrates how a longer recycling time can impact the rate of transition under specific conditions and assumptions. For alternative scenarios, such as those in which there are no fissile material constraints, there are likely no significant impacts due to recycling time. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Limiting Surface Density Method Application to Large Arrays of Highly Heterogeneous Shipping Packages

This work represents the initial phase of an overall effort to reduce the analytical burdens and increase flexibility of the K-Area Complex which provides stewardship of special nuclear materials. A hand calculation method is being extended to apply to large arrays of highly heterogeneous shipping packages. This method has the potential to quickly determine the impact of perturbations on the array. The objective of the particular work reported herein is twofold. First, the analytical results upon which the Limiting Surface Density (LSD) method was originally based were reproduced using the modern computational code KENO-VI in SCALE 6.1. The second objective was to test the classic LSD method on storage of large arrays of highly heterogeneous shipping packages. Specifically, the objective is to determine if the method is adequate for existing arrays of 9975 shipping packages with fissile metal content. The classic LSD method was derived for simple packages of bare spheres of fissile material centered in an air filled cubic unit cell and placed in a cubic-pitch water-reflected array, the properties of which could then be translated to more practical array and package geometries. The classic method did not address highly heterogeneous packages which are close to neutronically decoupled from other identical packages. Therefore, the material based array constant (called c{sub 2}) as defined by the classic method is determined specifically for the packages in question. (authors)

42 ENGINEERING↗

Application of Neutron Detection Third Order Time Correlation

Neutron noise experiments on fissile materials are used to get information such as reactivity, fissile mass, neutron sources, prompt neutron lifetime. One of the more popular methods is based on correlated moments of the neutron counts distribution. The second correlated moment method originally proposed by Feynman was extended to the third order. In this paper we present the application of the third order correlation moments method to SILENE reactor measurements. The third order correlation moments method has been applied to analyse time list mode - subcritical measurements on SILENE reactor in order to evaluate the system subcriticality, The multiplication coefficient is obtained with large statistical uncertainties, but remains consistent with MCNP calculations. To reduce the experimental uncertainties longer measurement time and higher detector efficiency should be necessary. The systematic errors coming from the fission neutron data (moments of the fission neutron distributions) and point model formulas should also be investigated. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Challenges of SLOWPOKE-2 HEU Irradiated Fuel Transportation from Canada to USA - 19123(2)

The SLOWPOKE-2 (Safe Low-Power Kritical Experiment) was designed by AECL (Atomic Energy of Canada Limited) in the late 1960's. The core consists of HEU (highly enriched uranium) fuel with an enrichment of 93%. The fuel cage typically has approximately 300 fuel pins assembled in a triangular pitch between the two plates. The fuel pins are made from uranium-aluminum alloy cladded with aluminum. The SLOWPOKE-2 reactor is rated at a nominal 20 kW. with varying neutron flux, maximum being 1 E12 n/cm{sup 2}/s. The mass of the fuel cage is 5.06 kg. The life of the core can be extended by adding layers of beryllium reflector. The transportation flask for the irradiated fuel is a Type B(U)F. The shipment classification is UN3328 Class 7. To date, 4 SLOWPOKE-2 reactors have been decommissioned. The fuel from all these reactors was transported to the U.S. The CNSC (Canadian Nuclear Safety Commission) joint Canada-United States guide for approval of Type B(U) and fissile material transportation packages (RD-364) satisfies the requirements stipulated in International Atomic energy Agency (IAEA) TS-R-1 1996 edition (revised) and is acceptable to the DOT (Department of Transportation) and NRC (Nuclear Regulatory Commission) for complying with U.S. code of federal regulations 10 CFR Part 71. The export license is issued by the CNSC. The transportation licence is issued by CNSC after approval from TC (Transport Canada) and also by DOT after approval from NRC. The final home of the HEU SLOWPOKE-2 irradiated fuel is in Savannah River Site of DOE (Department of Energy). The transportation licence requires that the Type B(U)F flask is structurally sound in case of a crash. The US DOT provides a CAC (Competent Authority Certificate) which describes the irradiated fuel specifications that can be transported in the designated flask. The U.S. DOT also requires a leak test certification in accordance with ANSI N14.5-2014 to verify a leakage rate not to exceed 1 E-7 std-cm{sup 3}/s. As most SLOWPOKE-2 facilities are not located on the existing approved routes for nuclear fuel transportation, a route survey needs to be performed and submitted to the NRC. The route survey report is assessed by the NRC and issues a route approval for the transportation of the fissile material. This technical paper will discuss the detailed processes required to obtain approvals in order to transport the HEU SLOWPOKE-2 irradiated fuel from Canada to U.S. based upon recent experiences of decommissioning SLOWPOKE-2 reactors in Canada. It will provide details on the requirements of various regulators in Canada and the U.S. that govern and regulate this shipment including transportation, safety and security. It will also present lessons learned from previous experiences that can be utilized for best practice on other similar projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

lNAC's OPTIMUS{sup TM}-L Package - the Next Transformation - 19396

NAC has developed the OPTIMUS{sup TM}-L package, the newest addition to the OPTIMUS{sup TM} packaging product line. OPTIMUS{sup TM}-L is a Type B(U)F-96 transportation package designed for maximum flexibility and cost-efficiency to support shipment of a wide range of challenging low-activity wastes. The OPTIMUS{sup TM}-L package utilizes the same cask containment vessel (CCV) as its predecessor, OPTIMUS{sup TM}-H, packaged inside lightweight outer packaging. The large cavity size of the CCV, which can accommodate a 110-gallon drum, combined with the small size and modularity of the OPTIMUS{sup TM}-L package provides unmatched flexibility for the user. OPTIMUS{sup TM}-L can accommodate more radioactive waste in each drum (up to a Fissile Gram Equivalent (FGE) limit of 395 g Pu-239) than larger packages can in 10 drums. Furthermore, because of its small size and weight, up to ten (10) OPTIMUS{sup TM}-L packages can be accommodated per legal-weight truck shipment. Thus, one Legal Weight Truck (LWT) shipment of 10 OPTIMUS{sup TM}-L packages can accommodate more than 10 times the FGE as a large package containing 10 drums. OPTIMUS{sup TM}-L contents include low level waste, mixed low-level waste, intermediate level waste and contact handled transuranic (CH-TRU) waste. Furthermore, all contents can include aerosol cans with compressed gas and liquefied gas propellants and standard Department of Transportation (DOT) 3E lecture bottles for shipment to treatment facilities. The OPTIMUS{sup TM}-L is also designed to accommodate irradiated fuel debris and other fissile material waste, which may be packaged inside shield inserts if required for supplemental radiation shielding. In this paper, NAC provides a technical overview of the OPTIMUS{sup TM}-L package and identifies the design features and technology advancements that makes the OPTIMUS{sup TM} product line a readily adaptable and flexible solution for packaging DOE and other orphan wastes and enables the integration of these wastes into the waste management system process and disposition infrastructure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Transition to a Thorium-based Fuel Cycle - Choices, Challenges, and Options

The Charter for the Evaluation and Screening (E and S) of Nuclear Fuel Cycle Options requested by the US Department of Energy calls for the identification of promising fuel cycle options to help guide the future directions of research and development in this area. The final report for this study, which was produced by various national laboratory and industry experts, collected the expansive set of fuel cycle options into 40 evaluation groups (EGs). In support of this study, detailed steady-state physics and mass flow data for each of these groups were compiled into a set of Fuel Cycle Data Packages (FCDPs). The data in the FCDPs were then used to quantify the nine criteria detailed in the study to identify the most promising EGs. In addition to supporting the E and S study, data from these FCDPs can be used to benchmark fuel cycle simulation tools for equilibrium calculations. Once benchmarked, their dynamic modeling capabilities can support the continuing effort to determine the most effective approaches for transitioning to proposed advanced future fuel cycles. While steady-state conditions can be calculated directly in some cases, fuel cycle simulation tools allow for the transition scenarios to be simultaneously optimized on multiple parameters such as the fissile material availability, the amount of desirable material in storage, and the time required for transition. In addition to fuel cycle simulations where material availability is considered and is often a constraint to achieving transition, determining the technology needs to enable a transition to take place is also a key consideration (e.g., what technology is needed, when it is needed, and whether it is needed only during the transition and not for the equilibrium fuel cycle). This paper evaluates the material and technology issues and challenges associated with a transition from a uranium-based fuel cycle (such as that operating in the US today, represented by EG01) to a thorium-based fuel cycle, recognizing that thorium requires fissile material to start up. Based on the findings of the E and S report, the two most promising thorium fuel cycle options are considered in this report: (1) EG26, a thermal spectrum molten salt reactor (MSR), and (2) EG28, represented by a sodium fast reactor (SFR, fast spectrum) and a resource-renewable boiling water reactor (RBWR, intermediate spectrum). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Applying understanding of likelihood and consequences to demonstrate post-closure criticality safety of a geological disposal facility - 14377

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of criticality safety under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls (e.g. fissile material limits and/or water barriers) will ensure that waste packages are sub-critical. These controls, plus the containment provided by the engineered barrier system of a geological disposal facility (GDF) will also ensure that criticality is prevented for a long time after disposal. However, over post-closure timescales conditions in the GDF will evolve, and therefore underpinning knowledge is required to demonstrate: under what conditions criticality could occur; what the likelihood of such conditions developing is; and what the consequences would be if a post-closure criticality did occur. These requirements are set out in the UK environment agencies' Guidance on Requirements for Authorisation of a GDF. This paper discusses the disposal of Pressurised Water Reactor fuel, as an example of a waste type that may require disposal. It briefly selects relevant research to combine and summarise results from three recent RWM research projects. The first 'Likelihood of Criticality' considered under what conditions post-closure criticality could occur, whilst a second 'Modelling of Consequences of Hypothetical Criticality' focused on understanding the local physical consequences (temperature rise, power and rock response) of hypothetical 'what-if' post-closure criticality. The paper then explains how this understanding is applied to make safety arguments as to why post-closure criticality is not a significant concern. This is achieved via a third, stream of work that has considered the effects of a potential criticality on pathways that might give rise to a post-closure risk in the future, such as transport of radionuclides in groundwater. Collectively RWMs post-closure criticality safety work allows us to conclude in our generic environmental safety case that a post closure criticality is a low likelihood event, and even if one of these low likelihood events was to occur its consequences would be insignificant to GDF performance. This paper also serves to give context to two companion UK post-closure criticality safety papers which communicate further information on our modelling development and more detailed results underpinning our demonstration of post-closure criticality safety. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nuclear criticality safety of pipes networks containing uranium - 14437

In nuclear facilities, the transfer of fissile material in solution, gas or powder forms between process stages requires many transfer lines (called 'pipes' in this paper) which may be laid out into bundles or networks. In Nuclear Criticality Safety evaluations, it is usual to assimilate such networks as a unique cylinder which area bounds the sum of the surface of pipes making up the network. However, even if this geometric simplification appears intuitively safe, it is not conservative in all cases. The aim of this paper is therefore to explore the validity and limitations associated to the assessment of pipes networks by a unique cylinder. This paper presents the impact on the k{sub eff} value of a detailed modeling of a pipes network compared to a unique cylinder. Variables of interest considered in such calculations are pipes diameter, composition, pitches and layout. Fissile media taken into account are uranium systems enriched from 5 % to 93.5 % in {sup 235}U. Then, the impact on reactivity of pipes material, such as stainless steel or plastic, is discussed and the minimal or maximal safe thickness of material has been determined. Finally, this paper provides guidance on the nuclear criticality safety assessment of pipes networks for uranium systems. (authors)

42 ENGINEERING↗

A system for the measurement of delayed neutrons and gammas from special nuclear materials

The delayed neutron counting (DNC) system at the Royal Military College of Canada has been upgraded to accommodate concurrent delayed neutron and gamma measurements. This delayed neutron and gamma counting (DNGC) system uses a SLOWPOKE-2 reactor to irradiate fissile materials before their transfer to a counting arrangement consisting of six ³He and one HPGe detector. The application of this system is demonstrated in an example where delayed neutron and gamma emissions are used in complement to examine ²³³U content and determine fissile mass with an average relative error and accuracy of -2.2 and 1.5 %, respectively.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in this paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fuel cycle and neutronic performance of a spectral shift molten salt reactor design

Here, the fuel cycle performance and core design of the Transatomic Power liquid-fueled molten salt reactor concept is analyzed. This advanced reactor concept uses configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from intermediate at beginning of life to thermal at end of life. With a harder spectrum during the early years of reactor operation, this spectral shift design drives captures in fertile 238 U. The converted fissile plutonium makes up over 50% of the fissile material in the fuel salt over the last half (~15 years) of reactor operation. A softer spectrum late in reactor life helps drive the fuel to a burnup of 90 GWd/MTU. Continuously changing physics necessitates time-dependent analyses resolved over long timescales (i.e., months to years), as this concept does not meet an equilibrium condition. The spectral shift and molten salt reactor material feeds and removals enable this concept to perform better in fuel cycle metrics, increasing resource utilization by more than 50% compared with a typical light water reactor (i.e., from ~0.6% to ~1%). Finally, these metrics are compared to similar fuel cycles using alternate technologies. Additional core design and analysis challenges associated with the spectral shift and use of molten salt reactor technology are identified and discussed.

20 FOSSIL-FUELED POWER PLANTS↗