Fast Reactor Fuel Cycle Cost Estimates for Advanced Fuel Cycle Studies
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Harrison, Thomas J↗
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Safeguards Challenges and Safeguards Technology Needs Assessment for Leading Thorium Fuel Cycles
Several thorium fuel cycle variants are currently being actively pursued by the global nuclear energy community. These variants have short-, medium-, and long-term deployment pathways in a variety of reactor types, including thorium-fueled molten salt reactors. Those pathways in turn give rise to a variety of fuel designs, fuel cycle facilities, and nuclear material processing requirements. These emerging fuel cycles will impact the technical implementation of safeguards, and they already raise questions about the applicability of current verification technologies. To address this issue, research is being performed to produce a detailed safeguards technology road map. It will include a needs assessment of the detection research and development (R&D;) necessary to transition the current safeguards technology toolkit to meet the verification needs of thorium fuel cycles, and to formulate the scientific basis for building new instrumentation to fill any potential capability gaps. The purpose of this technology road map is to define and inform on the safeguards technology needs for thorium fuel cycles, reflecting only the leading candidate thorium fuel cycles prioritized based on implementation timescales and the current direction of international programs. This work provides a guide to the priorities for future directed R&D; needed to bring the technology readiness levels (TRLs) of safeguards detection solutions in line with the higher TRLs of the most promising thorium fuel cycles. Herein, a summary is presented of thorium fuel cycle options and activities currently under way worldwide, which provides the technical basis for the needs evaluation. The key synergies and differences between these fuel cycles and current conventional uranium- and plutonium-based fuel cycles will be discussed from a nuclear material accountancy and detection standpoint. Reactor inventory calculations, fuel cycle simulations, and conclusions from the safeguards technology needs assessment will be presented, together with plans for experimental validation.
Worrall, Louise G.↗
Insights and Trends from a Literature Assessment of the Thorium Fuel Cycle
The thorium fuel cycle is often perceived to have limited experience, thereby imposing a barrier to its implementation. Thorium fuel cycles do not have the degree of commercial-scale experience associated with the uranium fuel cycle (especially the use of uranium oxide in a once-through configuration in light- or heavy-water reactors); however the abundance of major research campaigns - and the test and demonstration reactor operations that capped these campaigns - illustrate that significant knowledge concerning the use of thorium is available. Part of the reason for thorium's perceived immaturity is that the information is scattered among a multitude of sources and has not previously been consolidated and evaluated on a comprehensive scale. In recognition of this challenge, Vanderbilt University (VU) and Oak Ridge National Laboratory (ORNL) included a thorium fuel cycle literature assessment task as part of their Nuclear Energy University Programs (NEUP) project. The first phase of this literature assessment culminated in a comprehensive report. This paper describes the highlights of the initial version of this literature report, including item quantification, trends by category, and implications. This preliminary literature assessment has identified that a wealth of documented experience with thorium fuel cycles exists, across many varieties and specific technology options. Though a significant share of the literature is comprised of systems studies and/or proposed work, there is also a substantial knowledge base from experimentation and demonstration with thorium fuel cycle technologies. It is intended that this initial literature report serve as a starting point for continued discussions and consolidation of thorium fuel cycle literature. There are likely still records of work from the 1950's to the 1970's (and even more recently) which could be significant but were not adequately catalogued or are not readily available to a broad audience; hopefully, the ongoing discussions that have been spurred by the present work will stimulate making these documents available. The authors strongly encourage those with knowledge of additional data sources to share them with the interested community. Furthermore, the 21. century is experiencing a resurgence in thorium fuel cycle work that continues to develop the knowledge base. Thus, it would be beneficial to produce updated accounts of this thorium literature summary in future years. Ultimately, the items identified during the course of this literature assessment, along with additional items that continue to be identified, will be housed in an information repository. This will take the form of a relational database, with specific functionality to be determined based on user interest. This format is intended to enable future users to quickly find specific topics or sources as well as assess how much information is available about a particular topic. (authors)
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Fuel Cycle Considerations for Uranium, Plutonium and Minor Actinide Partitioning and Transmutation
This paper is written in memorial of the authors' past collaborations with Max Salvatores in the area of advanced nuclear fuel cycle evaluation. In this regard, a summary is provided of the results from two international studies led by Max Salvatores in which the authors participated: (1) the expert group on Homogeneous versus Heterogeneous Recycling of Transuranics in Fast Nuclear Reactors, and (2) the expert group on Potential Benefits and Impacts of Advanced Nuclear Fuel Cycles with Actinide Partitioning and Transmutation. Both of these studies presented the authors with the opportunity to contribute the results of past and ongoing research in the U.S., facilitating engagement and discussion on these subjects with researchers from other countries and organizations, resulting in the creation of broad international views on the relative benefits and challenges of transuranic recycle (NEA, 2012a; NEA, 2012b), a testament to the leadership and capabilities of Max Salvatores. Published by Elsevier Ltd.
Taiwo, Temitope↗
Physics and technology considerations for the deuterium–tritium fuel cycle and conditions for tritium fuel self sufficiency
The tritium aspects of the DT fuel cycle embody some of the most challenging feasibility and attractiveness issues in the development of fusion systems. The review and analyses in this paper provide important information to understand and quantify these challenges and to define the phase space of plasma physics and fusion technology parameters and features that must guide a serious R&D in the world fusion program. We focus in particular on components, issues and R&D necessary to satisfy three 'principal requirements': (1) achieving tritium self-sufficiency within the fusion system, (2) providing a tritium inventory for the initial start-up of a fusion facility, and (3) managing the safety and biological hazards of tritium. A primary conclusion is that the physics and technology state-of-the-art will not enable DEMO and future power plants to satisfy these principal requirements. We quantify goals and define specific areas and ideas for physics and technology R&D to meet these requirements. A powerful fuel cycle dynamics model was developed to calculate time-dependent tritium inventories and flow rates in all parts and components of the fuel cycle for different ranges of parameters and physics and technology conditions. Dynamics modeling analyses show that the key parameters affecting tritium inventories, tritium start-up inventory, and tritium self-sufficiency are the tritium burn fraction in the plasma (f b ), fueling efficiency (η f ), processing time of plasma exhaust in the inner fuel cycle (t p ), reactor availability factor (AF), reserve time (tr) which determines the reserve tritium inventory needed in the storage system in order to keep the plant operational for time t r in case of any malfunction of any part of the tritium processing system, and the doubling time (t d ). Results show that η f f b > 2% and processing time of 1–4 h are required to achieve tritium self-sufficiency with reasonable confidence. For η f f b = 2% and processing time of 4 h, the tritium start-up inventory required for a 3 GW fusion reactor is ~11 kg, while it is <5 kg if η f f b = 5% and the processing time is 1 h. To achieve these stringent requirements, a serious R&D program in physics and technology is necessary. The EU-DEMO direct internal recycling concept that carries fuel directly from the plasma exhaust gas to the fueling systems without going through the isotope separation system reduces the overall processing time and tritium inventories and has positive effects on the required tritium breeding ratio (TBR R ). A significant finding is the strong dependence of tritium self-sufficiency on the reactor availability factor. Simulations show that tritium self-sufficiency is: impossible if AF < 10% for any η f f b , possible if AF > 30% and 1% ≤ η f f b ≤ 2%, and achievable with reasonable confidence if AF > 50% and η f f b > 2%. These results are of particular concern in light of the low availability factor predicted for the near-term plasma-based experimental facilities (e.g. FNSF, VNS, CTF), and can have repercussions on tritium economy in DEMO reactors as well, unless significant advancements in RAMI are made. There is a linear dependency between the tritium start-up inventory and the fusion power. The required tritium start-up inventory for a fusion facility of 100 MW fusion power is as small as 1 kg. Since fusion power plants will have large powers for better economics, it is important to maintain a 'reserve' tritium inventory in the tritium storage system to continue to fuel the plasma and avoid plant shutdown in case of malfunctions of some parts of the tritium processing lines. But our results show that a reserve time as short as 24 h leads to unacceptable reserve and start-up inventory requirements. Therefore, high reliability and fast maintainability of all components in the fuel cycle are necessary in order to avoid the need for storing reserve tritium inventory sufficient for continued fusion facility operation for more than a few hours. The physics aspects of plasma fueling, tritium burn fraction, and particle and power exhaust are highly interrelated and complex, and predictions for DEMO and power reactors are highly uncertain because of lack of experiments with burning plasma. Fueling by pellet injection on the high field side of tokamak has evolved to be the preferred method to fuel a burning plasma. Extrapolation from the DIII-D penetration scaling shows fueling efficiency expected in DEMO to be <25%, but such extrapolations are highly uncertain. The fueling efficiency of gas in a reactor relevant regime is expected to be extremely poor and not very useful for getting tritium into the core plasma efficiently. Gas fueling will nonetheless be useful for feedback control of the divertor operating parameters. Extensive modeling has been carried out to predict burn fraction, fueling requirements, and fueling efficiency for ITER, DEMO, and beyond. The fueling rate required to operate Q = 10 ITER plasmas in order to provide the required core fueling, helium exhaust and radiative divertor plasma conditions for acceptable divertor power loads was calculated. If this fueling is performed with a 50–50 DT mix, the tritium burn fraction in ITER would be ~0.36%, which is too low to satisfy the self-sufficiency conditions derived from the dynamics modeling for fusion reactors. Extrapolation to DEMO using this approach would also yield similarly low burn fraction. Extensive analysis presented shows that specific features of edge neutral dynamics in ITER and fusion reactors, which are different from present experiments, open possibilities for optimization of tritium fueling and thus to improve the burn fraction. Using only tritium in pellet fueling of the plasma core, and only deuterium for edge density, divertor power load and ELM control results in significant increase of the burn fraction to 1.8–3.6%. These estimates are performed with physics models whose results cannot be fully validated for ITER and DEMO plasma conditions since these cannot be achieved in present tokamak experiments. Thus, several uncertainties remain regarding particle transport and scenario requirements in ITER and DEMO. The safety standard requirements for protection of the public and release guidelines for tritium have been reviewed. General safety approaches including minimizing tritium inventories, reducing tritium permeation through materials, and decontaminating material for waste disposal have been suggested.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Reactor and fuel cycle performance of light water reactor fuel with 235 U enrichments above 5%
Recent advances in nuclear fuel materials research, particularly on the topic of accident-tolerant fuels, have brought up potential opportunities for expanding the operating envelope of existing light water reactors. As many of the performance improvements offered by these technologies may be most fully realized by increasing fuel enrichment beyond the standard 5% limit, this paper examines the potential reactor performance and fuel cycle performance of low-enriched uranium oxide fueled light water reactors by generically considering pressurized water reactors with 235 U enrichment from 5 to 7%. Advanced cladding, including accident-tolerant cladding, has the potential to increase fuel burnup limits related to hydrogen in the cladding that coincide with those limits associated with end-of-life reactivity. Therefore, higher enrichment will be necessary in order to realize the higher fuel burnups. This work includes evaluation of the fuel cycle length, discharge burnup, reactivity coefficients, and fuel cycle performance, including radioactive waste and environmental impact metrics per unit energy generated. The analysis was performed using the evaluation metrics from the US Department of Energy Office of Nuclear Energy Fuel Cycle Evaluation and Screening Study. The reactor performance and safety analysis show that enrichments between 5 and 7% would have similar fuel temperature and moderator temperature coefficients. However, the soluble boron coefficient would decrease in magnitude, requiring more corrosive boric acid in the coolant or other methods of reactivity control during the fuel cycle. At these higher enrichments the maximum burnup at the rim of the fuel pellet would increase by almost a factor of two, which is expected to impact the formation of high-burnup structure in the fuel and the corresponding thermo-mechanical fuel properties. The fuel cycle performance assessment shows that increasing enrichment reduces the quantity of high-level waste disposed per unit energy generated, but it increases the natural resource requirements normalized to a gigawatt-electricity-per-year basis. Another impact is the slightly higher discharge burnup, resulting in somewhat different activity levels of the spent nuclear fuel and high-level waste radioactivity at 100 and 100,000 years after fuel discharge. The environmental impacts—including land use, water use, carbon emission, and radiological exposure—are of the same magnitude per unit energy generated. Finally, however, the impacts are distributed differently. Less than 5% enrichment has marginally more impact on the back-end of the fuel cycle, and greater than 5% enrichment has marginally more impact on the front-end of the fuel cycle. Ultimately, no neutronic or reactor safety hindrances to employing light water reactor fuel with enrichments greater than 5% are identified; given the achievable reactor performance benefits with advanced fuels, further practical exploration of increased enrichment fuel is recommended.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Development Plan for the Fuel Cycle Simulator
The Fuel Cycle Simulator (FCS) project was initiated late in FY-10 as the activity to develop a next generation fuel cycle dynamic analysis tool for achieving the Systems Analysis Campaign 'Grand Challenge.' This challenge, as documented in the Campaign Implementation Plan, is to: 'Develop a fuel cycle simulator as part of a suite of tools to support decision-making, communication, and education, that synthesizes and visually explains the multiple attributes of potential fuel cycles.'
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Advancing Inverse Sensitivity/Uncertainty Methods for Nuclear Fuel Cycle Applications
The inverse sensitivity/uncertainty quantification (IS/UQ) method has recently been implemented in the Inverse Sensitivity/UnceRtainty Estimator (INSURE) module of the AMPX cross section processing system [M.E. Dunn and N.M. Greene, “AMPX-2000: A Cross-Section Processing System for Generating Nuclear Data for Criticality Safety Applications,” Trans. Am. Nucl. Soc. 86, 118–119 (2002)]. The IS/UQ method aims to quantify and prioritize the cross section measurements along with uncertainties needed to yield a given nuclear application(s) target response uncertainty, and doing this at a minimum cost. Since in some cases the extant uncertainties of the differential cross section data are already near the limits of the present-day state-of-the-art measurements, requiring significantly smaller uncertainties may be unrealistic. Therefore, we have incorporated integral benchmark experiments (IBEs) data into the IS/UQ method using the generalized linear least-squares method, and have implemented it in the INSURE module. We show how the IS/UQ method could be applied to systematic and statistical uncertainties in a self-consistent way and how it could be used to optimize uncertainties of IBEs and differential cross section data simultaneously. We itemize contributions to the cost of differential data measurements needed to define a realistic cost function.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Evaluation of fuel cycle scenarios on MOX fuel recycling in PWRs and SFRs
Prospects on advanced fuel cycle scenario are considered for achieving a progressive integration of Sodium Fast Reactor (SFR) technology within the current French Pressurized Water Reactor (PWR) nuclear fleet, in a view to benefit from fissile material multi-recycling capability. A step by step process is envisioned, and emphasis is put on its potential implementation through the nuclear mass inventory calculations with the COSAC code. The overall time scale is not optimized. The first step, already implemented in several countries, the plutonium coming from the reprocessing of used Light Water Reactor (LWR) fuels is recycled into a small number of LWRs. The second step is the progressive introduction of the first SFRs, in parallel with the continuation of step 1. This second step lets to prepare the optimized multi recycling of MOX fuel which is considered in step 3. Step 3 is characterized by the introduction of a greater number of SFR and MOX management between EPR reactors and SFRs. In the final step 4, all the fleet is formed with SFRs. This study assesses the viability of each step of the overall scenario. The switch from one step to the other one could result from different constrains related to issues such as resources, waste, experience feedback, public acceptance, country policy, etc.
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Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content
The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)
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Database dedicated to nuclear criticality events that occurred in fuel cycle facilities
Nuclear fuel cycle facilities can encounter events, resulting from human errors, failure of safety organization or technical failures. However, the application of the concept of defense in depth, especially the first two levels (prevention of abnormal situations and detection) requires to ensure safety in such undesirable situations. Furthermore, the periodic safety reviews of each nuclear fuel cycle facility involve the evaluation of the safety of these facilities with regard to the regulation evolution, the best safety practices, the facility evolution (modifications, aging...) and the global feedback. Part of this work is based on the analysis of the feedback and lessons learned from the events that occurred during the past period. This specific work indeed provides lessons about the operating conditions of the facility and highlights malfunctions that have occurred during operations, maintenance or restart after a shutdown. Taking into account this feedback, IRSN developed a database gathering the events which have affected the safety of French fuel cycle facilities (from conversion to waste storage, laboratories and experimental facilities, facilities under dismantling and decommissioning). Among all the recorded events, it is possible to extract in particular those impacting the criticality safety. In order to extend the analysis of these events, the IRSN criticality safety department developed a specific database dedicated to criticality risks. The 'criticality events' database is continuously updated on the basis of the declared events and their analysis by engineers from the criticality safety department. At this time, about 600 events are described and analyzed in this database. (author)
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Standardized verification of the C YCLUS fuel cycle simulator
Many nuclear fuel cycle simulators can analyze transitions from once-through to advanced nuclear fuel cycles. Verification studies compare various fuel cycle analysis tools to test agreement and identify sources of difference. A recent verification study, Feng et al. (2016) established transition scenario test case specifications and accordingly evaluated national laboratory nuclear fuel cycle simulators, DYMOND, VISION, ORION, and MARKAL. This work verifies the performance of C YCLUS , the agent-based, open-source fuel cycle simulator, using the test case specifications in Feng et. al. We report, C YCLUS demonstrates agreement with the results from the previous verification study. Minor differences reflect intentional, detailed material tracking in the C YCAMORE reactor module. These results extend the example results in Feng et al. to further enable future verification of additional nuclear fuel cycle simulation tools.
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Fuel Cycle Assessment: Evaluation and Analyses using ORION for US Fuel Cycle Options
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Sunny, Eva E↗
Quantifying public perceptions on the nuclear fuel cycle - 15395
The nuclear fuel cycle in the United States, specifically the back end portion, remains an unresolved and persistent issue. In comparison with the technical aspects of many different fuel cycles, less attention has been given to public perceptions and attitudes toward certain alternatives. Understanding what the general public values in terms of a nuclear fuel cycle will assist in selecting a nuclear fuel cycle that optimizes the balance between technical feasibility and public acceptance. A multi-criteria decision analysis (MCDA) model was created to incorporate the general public's perceptions on qualified individuals and on important technical and non-technical aspects of any general fuel cycle. Analyzing the pair-wise comparisons for each criterion through the Analytic Hierarchy Process (AHP) yielded the relative importance of each aspect. A major finding is that the general public tend to view environmental scientists as qualified as nuclear engineers and scientists in selecting a method of used nuclear fuel disposal for the United States. (authors)
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Physics challenges for advanced fuel cycle assessment
Advanced fuel cycles and associated optimized reactor designs will require substantial improvements in key research area to meet new and more challenging requirements. The present paper reviews challenges and issues in the field of reactor and fuel cycle physics. Typical examples are discussed with, in some cases, original results.
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Potential External (non-DOE) Constraints on U.S. Fuel Cycle Options
The DOE Fuel Cycle Technologies (FCT) Program will be conducting a screening of fuel cycle options in FY2013 to help focus fuel cycle R&D activities. As part of this screening, performance criteria and go/no-go criteria are being identified. To help ensure that these criteria are consistent with current policy, an effort was initiated to identify the status and basis of potentially relevant regulations, laws, and policies that have been established external to DOE. As such regulations, laws, and policies may be beyond DOE’s control to change, they may constrain the screening criteria and internally-developed policy. This report contains a historical survey and analysis of publically available domestic documents that could pertain to external constraints on advanced nuclear fuel cycles. “External” is defined as public documents outside DOE. This effort did not include survey and analysis of constraints established internal to DOE.
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A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375
Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)
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Rapid methods for radionuclide contaminant transport in nuclear fuel cycle simulation
Here, nuclear fuel cycle and nuclear waste disposal decisions are technologically coupled. However, current nuclear fuel cycle simulators lack dynamic repository performance analysis due to the computational burden of high-fidelity hydrolgic contaminant transport models. The Cyder disposal environment and repository module was developed to fill this gap. It implements medium-fidelity hydrologic radionuclide transport models to support assessment appropriate for fuel cycle simulation in the Cyclus fuel cycle simulator. Rapid modeling of hundreds of discrete waste packages in a geologic environment is enabled within this module by a suite of four closed form models for advective, dispersive, coupled, and idealized con- taminant transport: a Degradation Rate model, a Mixed Cell model, a Lumped Parameter model, and a 1-D Permeable Porous Medium model. A summary of the Cyder module, its timestepping algorithm, and the mathematical models implemented within it are presented. Additionally, parametric demonstrations simulations performed with Cyder are presented and shown to demonstrate functional agreement with parametric simulations conducted in a standalone hydrologic transport model, the Clay Generic Disposal System Model developed by the Used Fuel Disposition Campaign Department of Energy Office of Nuclear Energy.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗