Nuclear Fuel Cycle Technology and Policy Program (NFC) - Technical Reports

Feasibility of Breeding in Hard Spectrum Boiling Water Reactors with Oxide and Nitride Fuels

2019-04-12T11:17:11Z

Feasibility of Breeding in Hard Spectrum Boiling Water Reactors with Oxide and Nitride Fuels Feng, Bo; Kazimi, Mujid S.; Forget, Benoit This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel hardens the neutron energy spectrum and results in higher breeding ratios. The state-of-the-art high conversion light water reactor, the Resource-renewable Boiling Water Reactor (RBWR), served as the template core upon which comparative studies between nitride and oxide fuels were performed. A 1/3 core reactor physics model was developed for the RBWR using the stochastic transport code MCNP. The code was coupled with a lumped channel thermal-hydraulics 5-channel model for steady-state analyses. The depletion code MCODE, which links MCNP with ORIGEN, was used for all burnup calculations. Select physics parameters were calculated and with the exception of the void coefficients, agreed with reported data. The void coefficients of the coupled core were calculated to be slightly positive using two different methods (10% power increase and 5% flow reduction). The standard RBWR assembly designs, which use tight lattice hexagonal fuel rod arrays, with oxide fuel were then replaced with various nitride fuel assembly designs to determine the potential increase in breeding ratio, the potential to breed with pressurized water, and the potential to improve the critical power ratio with a wider pin pitch. Without changing the assembly geometry or discharge burnup, using nitride fuel resulted in a breeding ratio of 1.14. Using single-phase liquid water, the nitride fuel RBWR assembly resulted in a conversion ratio of 1.00. Another nitride fuel assembly design with boiling water maintained a 1.04 breeding ratio while increasing the pitch-to-diameter ratio from 1.13 to 1.20. This modification increased the hot assembly critical power ratio from 1.22 to 1.36, as calculated using the Liu- 2007 correlation. A high-porosity nitride fuel is recommended for high burnup conditions, to accommodate the nitride fuel’s higher swelling and less favorable mechanical properties compared to the oxide fuel. The high porosity allows additional volume for pressure-induced densification, alleviating swelling and subsequent cladding strain. To predict the performance of high-porosity nitride fuel, fission gas and fuel behavior mechanistic models were developed for high burnup and low-temperature conditions. These models were validated with reported irradiation data and implemented, along with fuel material properties, into the steady-state fuel behavior code FRAPCON-EP. Under simulated RBWR conditions, a fuel density no more than 85% of theoretical density is recommended to maintain satisfactory fuel performance.

PROLIFERATION RESISTANT, LOW COST, THORIA-URANIA FUEL FOR LIGHT WATER REACTORS

2019-04-12T20:30:54Z

PROLIFERATION RESISTANT, LOW COST, THORIA-URANIA FUEL FOR LIGHT WATER REACTORS Kazimi, Mujid S.; Driscoll, Michael J.; Ballinger, Ronald G.; Clarno, K. T.; Czerwinski, Kenneth R.; Hejzlar, Pavel; LaFond, P. J.; Long, Y.; Meyer, J. E.; Reynard, M. P.; Schultz, S. P.; Zhao, X. 1. Summary Project Objectives: Our objective is to develop a fuel consisting of mixed thorium dioxide and uranium dioxide (ThO[subscript 2]-UO[subscript 2]) for existing light water reactors (LWRs) that (a) is less expensive overall than the current uranium-dioxide (UO[subscript 2]) fuel, (b) is very resistant to nuclear weapons-material proliferation, (c) results in a more stable and insoluble waste form, and, (d) generates less spent fuel per unit energy production. This project is being conducted in collaboration with INEEL. This annual report presents the MIT progress in the investigations from October 1998 up to June 1999.

High Performance Fuel Design for Next Generation PWRs: 11th Quarterly Report

2019-04-12T20:33:41Z

High Performance Fuel Design for Next Generation PWRs: 11th Quarterly Report Kazimi, Mujid S.; Hejzlar, Pavel; Feng, Dandong; Kohse, Gordon E.; Morra, Paolo; Ostrovsky, Yakov; Saha, Pradip; Xu, Zhiwen; Yuan, Yi; Carpenter, David M.; Feinroth, Herbert; Lahoda, Edward J.; Sundaram, Ramu K.; Hamilton, Holly I. Technical Narrative: The overall objective of this NERI project is to examine the potential for a high performance advanced fuel for Pressurized Water Reactors (PWRs), which would accommodate a substantial increase of core power density while simultaneously providing larger thermal margins than current PWRs. This advanced fuel will have an annular geometry that allows internal and external coolant flow and heat removal. The project is led by the Massachusetts Institute of Technology (MIT), with collaboration of four industrial partners – Gamma Engineering Corporation, Westinghouse Electric Corporation, Framatome ANP (formerly Duke Engineering & Services), and Atomic Energy of Canada Limited. Quarterly Report for Project DE-FG03-01SF22329 April 2004 – June 2004

Flexible Conversion Ratio Fast Reactor Systems Evaluation Final Report

2019-04-10T20:53:39Z

Flexible Conversion Ratio Fast Reactor Systems Evaluation Final Report Todreas, Neil E.; Hejzlar, Pavel; Fong, Chris J.; Nikiforova, Anna; Petroski, Robert; Shwageraus, Eugene; Whitman, Joshua Executive Summary: The goal of this project is to develop the conceptual designs of fast flexible conversion ratio reactors using lead and liquid salt coolants and to compare the results with a gascooled fast reactor developed in an MIT NERI project and a sodium-cooled reactor under development at ANL. To maintain the scope of the study manageable within the 2-year time frame and funding constraints, core designs that fit in the same reactor plant were executed for two limiting conversion ratios: (1) near zero, to transmute legacy waste and (2) near unity, to operate in a sustainable closed cycle. To reap the benefits of economy of scale, a large power rating of 2400MWt was set as the target thermal power for both reactor designs. In addition, the achievement of inherent reactor shutdown in unprotected accidents (without scram) was set as a desirable goal. Project DE-FC07-06ID14733

High Performance Fuel Design for Next Generation PWRs: Final Report

2019-04-10T23:57:36Z

High Performance Fuel Design for Next Generation PWRs: Final Report Kazimi, Mujid S.; Hejzlar, Pavel; Carpenter, David M.; Feng, Dandong; Kohse, Gordon E.; Lee, Won Jae; Morra, Paolo; No, Hee Cheon; Ostrovsky, Yakov; Otsuka, Yasuyuki; Saha, Pradip; Shwageraus, Eugene; Xu, Zhiwen; Yuan, Yi; Zhang, Jiyun; Feinroth, Herbert; Hao, Bernard; Lahoda, Edward J.; Mazzoccoli, Jason P.; Sundaram, Ramu K.; Hamilton, Holly This summary provides an overview of the results of the U.S. DOE funded NERI (Nuclear Research Energy Initiative) program on development of the internally and externally cooled annular fuel for high power density PWRs. This new fuel was proposed by MIT to allow a substantial increase in power density (on the order of 30% or higher) while maintaining or improving safety margins. A comprehensive study was performed by a team consisting of MIT (lead organization), Westinghouse Electric Corporation, Gamma Engineering Corporation, Framatome ANP (formerly Duke Engineering) and Atomic Energy of Canada Limited. The study involved the evaluation of the new fuel in terms of thermal hydraulic, neutronics, fuel performance including first scoping irradiation tests at the MIT reactor, fuel manufacturing and economics. Project DE-FG03-01SF22329

High Performance Fuel Design for Next Generation PWRs Appendices B-I to FY-02 Annual Report

2019-04-11T02:55:10Z

High Performance Fuel Design for Next Generation PWRs Appendices B-I to FY-02 Annual Report Kazimi, Mujid S.; Hejzlar, Pavel; Ballinger, Ronald G.; Carpenter, David M.; Feng, Dandong; Kohse, Gordon E.; Lee, Won Jae; No, Hee Cheon; Ostrovsky, Yakov; Otsuka, Yasuyuki; Stahl, Peter; Xu, Zhiwen; Yuan, Yi; Zhang, Jiyun; Feinroth, Herbert; Hao, Bernard; Lahoda, Edward J.; Mazzoccoli, Jason P.; Sundaram, Ramu K.; Hamilton, Holly B.1.1 VIPRE modeling of PWR core with annular fuel: Optimization studies in the first year used an isolated channel and models for MDNBR analyses. These analyses provided sufficient knowledge of potential thermal hydraulic performance of annular fuels to select the 13x13 array as the most promising configuration. To obtain more realistic and accurate MDNBR, a whole core model is necessary. In particular, the major concern is correct representation of channel flow rate. The earlier models used the core-average mass flux, which does not account for flow rate reduction in the hot channels due to increased pressure drop in this channel as a result of higher subcooled, or possibly, saturated boiling. Therefore, it is expected that the MDNBR obtained from the full core VIPRE-01 model will be smaller than the values obtained from the isolated channel model. Progress Report for Work August 2001 through July 2002

High Performance Fuel Design for Next Generation PWRs 2nd Annual Report

2019-04-12T20:33:41Z

High Performance Fuel Design for Next Generation PWRs 2nd Annual Report Kazimi, Mujid S.; Hejzlar, Pavel; Ballinger, Ronald G.; Carpenter, David M.; Feng, Dandong; Kohse, Gordon E.; Lee, Won Jae; No, Hee Cheon; Ostrovsky, Yakov; Otsuka, Yasuyuki; Stahl, Peter; Xu, Zhiwen; Yuan, Yi; Zhang, Jiyun; Feinroth, Herbert; Hao, Bernard; Lahoda, Edward J.; Mazzoccoli, Jason P.; Sundaram, Ramu K.; Hamilton, Holly The overall objective of this NERI project is to examine the potential for a high performance advanced fuel design for Pressurized Water Reactors (PWRs), which would accommodate a substantial increase of core power density while simultaneously providing larger thermal margins than current PWRs. This advanced fuel employs an annular geometry that allows internal and external coolant flow and heat removal. The project is led by the Massachusetts Institute of Technology (MIT), with the collaboration of four industrial partners – Gamma Engineering Corporation, Westinghouse Electric Corporation, Framatome ANP DE & S (formerly Duke Engineering & Services), and Atomic Energy of Canada Limited. The project is organized into five tasks: 1. Task 1 Assess the thermal hydraulic performance of the internally and externally cooled annular fuel to identify the configuration with the highest potential for power density increase while maintaining ample thermal margins, as well as key aspects of mechanical design to ensure that new fuel will not perform outside established hydraulic and mechanical constraints, 2. Task 2 Determine the neutronic performance of the new fuel, and the design that will minimize fuel cycle cost and assures that reactor physics safety parameters are as good or better than those of current PWRs, 3. Task 3 Explore various methods of manufacturing of this advanced fuel, including new innovative fabrication processes to produce annular fuel elements with the required product characteristics, 4. Task 4 Evaluate fuel cycle cost and capital cost implications of high power density to determine the economic viability of the high-performance fuel, and 5. Task 5 Analyze fuel performance of the new UO2 annular fuel obtained by various production technologies including irradiation testing in the MIT reactor. Progress Report for Work August 2002 through July 2003

Feasibility of Very Deep Borehole Disposal of US Nuclear Defense Wastes

2019-04-12T21:17:28Z

Feasibility of Very Deep Borehole Disposal of US Nuclear Defense Wastes Dozier, Frances E.; Driscoll, Michael J.; Buongiorno, Jacopo This report analyzes the feasibility of emplacing DOE-owned defense nuclear waste from weapons production into a permanent borehole repository drilled ~4 km into granite basement rock. Two canister options were analyzed throughout the report: the canister currently used by the DOE for vitrified defense waste and a reference canister with a smaller diameter. In a thermal analysis, the maximum temperatures attained by the rock surrounding the waste, waste form, canister, liner, and gaps during the post-emplacement period were calculated. From this data, simple analytic equations were formed that can be used to calculate the maximum temperature differences for both defense waste and spent fuel when one does not want to repeat the analysis. Canister corrosion and waste form dissolution analyses were performed using Pourbaix diagrams. Finally, the cost and time for drilling the borehole and emplacing the defense waste were calculated. The temperature change in the granite is 15.1°C for the reference canister and 45.7°C for the DOE Canister. The resulting maximum temperature at the bottom of the borehole is 135.1°C (reference canister) and 165.7°C (DOE canister) for the bounding defense waste. The centerline temperature for the borosilicate glass waste package is approximately 150°C for the reference canister and 207°C for the DOE canister. Because of the thermodynamic properties, overall corrosion resistance, and reasonable cost, pure copper was shown to be the best borehole outer canister material. High-chromium stainless steel could also be a good option for borehole canisters because it has been shown to be highly corrosion-resistant in environments similar to predicted borehole environments. Cesium ion was found to have the highest concentration in the borehole environment. However, the relatively low half life of the most abundant cesium isotope suggests that the cesium would decay before the canister is breached. For the reference canister, the drilling and emplacement costs are not expected to exceed $46/kg of vitrified waste and the total disposal cost was found to be $153/kg of vitrified waste. The total cost of disposal of defense waste in DOE containers is not expected to exceed $53/kg of vitrified waste. Based on these analyses, disposal of vitrified defense waste in deep boreholes is expected to be technically and economically feasible.

Cross Section Generation Strategy for High Conversion Light Water Reactors

2019-04-12T21:17:27Z

Cross Section Generation Strategy for High Conversion Light Water Reactors Herman, Bryan R.; Shwageraus, Eugene; Forget, Benoit; Kazimi, Mujid S. High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of greater than one and assure negative void coefficient of reactivity. This study assesses the generation of few-group macroscopic cross sections for neutron diffusion theory analyses of this type of reactor, in order to enable three-dimensional transient simulations. The goal is to minimize the number of energy groups in these simulations to reduce computational effort. A two-dimensional cross section generation methodology using the Monte Carlo code Serpent, similar to the traditional deterministic homogenization methodology, was used to analyze a single RBWR assembly. Results from two energy group and twelve energy group diffusion analyses showed an error in multiplication factor over 1000 pcm with errors in reaction rates between 10 and 60%. Therefore, the traditional approach is not sufficiently accurate. Instead, a three-dimensional homogenization methodology using Serpent was developed to account for neighboring zones in the homogenization process. A Python wrapper, SerpentXS, was developed to perform branch case calculations with Serpent to parametrize few-group parameters as a function of reactor operating conditions and to create a database for interpolation with the nodal diffusion theory code, PARCS. Diffusion analyses using this methodology also showed an error in multiplication factor over 1000 pcm. The three-dimensional homogenization capability in Serpent allowed for the introduction of axial discontinuity factors in the diffusion theory analysis, needed to preserve Monte Carlo reaction rates and global multiplication factor. A one-dimensional finite-difference multigroup diffusion theory code, developed in MATLAB, was written to investigate the use of axial discontinuity factors for a single RBWR assembly. The application of discontinuity factors on either side of each axial interface preserved multiplication factor and reaction rate estimates between transport theory and diffusion theory analyses to within statistical uncertainty. Use of this three-dimensional assembly homogenization approach in generating few-group macroscopic cross sections and axial discontinuity factors as a function of operating conditions will help further research in transient diffusion theory simulations of axially heterogeneous reactors.

A Drop-In Concept for Deep Borehole Canister Emplacement

2019-04-12T20:31:24Z

A Drop-In Concept for Deep Borehole Canister Emplacement Bates, Ethan A.; Buongiorno, Jacopo; Driscoll, Michael J. Disposal of high-level nuclear waste in deep boreholes drilled into crystalline bedrock (i.e., “granite”) is an interesting repository alternative of long standing. Work at MIT over the past two decades, and more recently in collaboration with the Sandia National Laboratory, has examined a broad spectrum of design aspects associated with this approach. For emplacement, past reports suggest using steel cables to lower each canister into the borehole. This process would require many years to complete and precise control to safely lower the canisters thousands of meters. The current study evaluated a simple, rapid, “passive” procedure for emplacement of canisters in a deep borehole: free-fall release into a water-flooded borehole. The project involves both analytic modeling and 1/5th scale experiments on a laboratory mockup. Experiments showed good agreement and validated the model. Depending on the inputs used for the mass and dimensions of the full scale canister and the viscosity of water, the model predicted terminal velocities of 2.4-2.6 m/s (4.5-5.8 mph). Further experiments showed that this could be reduced by 50% by making the surface hydraulically rough. Based on these predictions and a structural analysis, there seems to be little risk of damage when a canister reaches the bottom of the borehole or impacts the stack of previously loaded canisters. For reference, dropping the canister in air from a height of only 0.3 m (1 ft) would result in an impact velocity of 2.44 m/s. Cost estimates for the conventional drill string based method were developed, and the drop-in method was concluded to reduce emplacement costs and time by a minimum of 70%, down to $700,000 per borehole. It is concluded that a simple drop-in procedure deserves serious consideration for adoption as a standard procedure for borehole loading.

PWR Cores with Silicon Carbide Cladding

2019-04-11T03:09:45Z

PWR Cores with Silicon Carbide Cladding Kazimi, Mujid S.; Dobisesky, J.; Carpenter, David M.; Richards, J.; Pilat, Edward E.; Shwageraus, Evgeni The use in present generation PWRs of fuel clad with silicon carbide rather than Zircaloy has been evaluated as an aid to reaching higher discharge burnups and to operation at higher reactor power levels. A preliminary fuel design using fuel rods with the same dimensions as Westinghouse RFA fuel assemblies but with fuel pellets having 10 vol% central holes has been adopted. The central holes mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide, and the open gap between the fuel and cladding that persists over most of the irradiation. With this fuel design, it has been found possible to achieve 18 month cycles that meet present-day targets for peaking, boron concentration and shutdown margin while allowing average discharge burnups up to 80 MWD/KgU, as well as power uprates of 10% and possibly 20%. For non-uprated cores, the silicon carbide clad fuel has a clear economic advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a fuel cost savings of several million dollars per cycle as long as it does not increase the cost of fabrication by more than 50%, which seems highly unlikely. With 10-20% power uprates, the economics of the fuel cycle will improve, but the total value of such an uprate depends on the cost of needed plant modifications. Modifications to the control rod configuration or absorbing material may also be required to meet the shutdown margin criterion, particularly for the 20% uprate. Silicon carbide’s ability to sustain higher burnups and higher duty than Zircaloy also allows the design of a licensable two year cycle that has a fuel cost comparable to that of the reference 18 month Zircaloy core, and will furthermore reduce the average annual outage time.

A Generalized Optimization Methodology for Isotope Management

2019-04-12T20:32:09Z

A Generalized Optimization Methodology for Isotope Management Massie, Mark Edward; Forget, Benoit This research focuses on developing a new approach to studying the nuclear fuel cycle: instead of employing the trial and error approach currently used in actinide management studies in which reactors are designed and then their performance is evaluated, the methodology developed here first identifies relevant fuel cycle objectives–like minimizing decay heat production in a repository, minimizing Pu-239 content in used fuel, etc.–and then uses optimization to determine the best way to reach these goals. The first half of this research was devoted to identifying optimal flux spectra for irradiating used nuclear fuel from light water reactors to meet fuel cycle objectives like those mentioned above. This was accomplished by applying the simulated annealing optimization methodology to a simple matrix exponential depletion code written in Fortran using cross sections generated from the SCALE code system. Since flux spectra cannot be shaped arbitrarily, the second half of this research applied the same methodology to material composition of fast reactor target assemblies to find optimal designs for minimizing the integrated decay heat production over various timescales. The neutronics calculations were performed using modules from SCALE and ERANOS, a French fast reactor transport code.

Plugging of Deep Boreholes for HLW Disposal

2019-04-11T03:09:45Z

Plugging of Deep Boreholes for HLW Disposal Jensen, K. G.; Driscoll, Michael J. This is a progress report covering work through July 2010 under a Sandia-MIT contract dealing with design and siting/licensing criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. The principal focus is on conceptual design of a m ultilayer plug for our reference case borehole. It is similar to configurations recommended earlier by Swedish and Russian specialists. A secondary set of contributions update previous work in the areas such as multi-branch borehole configurations, use of cast iron canister inserts to resist crushing, and prospects for faster drilling.

Deep Boreholes Attributes and Performance Requirements

2019-04-12T21:17:27Z

Deep Boreholes Attributes and Performance Requirements Driscoll, Michael J.; Jensen, K. G. This is a progress report covering work through mid-May 2010 under a Sandia-MIT contract dealing with design and siting/licensing criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. It consists of additional short technical notes which scope out the performance-related requirements of a deep borehole repository. The most important changes since our last report are reversion to a single-branch vertical borehole (as recommended in the March 15 Workshop), and the consequential recommended adoption of a cast iron canister to alleviate the resulting bottom canister crushing threat. The case is also made that post closure nuclear criticality is not a credible scenario, with a very large margin of safety, even under very conservative assumptions.

Assessment of helical-cruciform fuel rods for high power density

2019-04-11T03:09:44Z

Assessment of helical-cruciform fuel rods for high power density Conboy, Thomas M.; McKrell, Thomas J.; Kazimi, Mujid S. In order to significantly increase the power density of Light Water Reactors (LWRs), the helical-cruciform (HC) fuel rod assembly has been proposed as an alternative to traditional fuel geometry. The HC assembly is a self-supporting nuclear fuel configuration consisting of 4-finned, axially-twisted fuel rods closely packed against one another in a square array. Within the LWR core, HC fuel would in theory possess several inherent advantages over traditional fuel, potentially allowing for operation at a higher power density. Chief among these advantages are a larger surface-to-volume ratio, a shorter radial heat conduction path, and improved mixing characteristics. In previous work, computational models of the HC fuel assembly have been of limited accuracy due to the absence suitable correlations. To address needs within these subchannel analysis models, experimental measurements of rod bundle coolant mixing have been conducted with 4x4 arrays of HC test rods. The tests used the technique of a hot water tracer injection (at 95°C) into a bulk flow of cold water (at 25°C). Downstream temperature measurements were used to judge the rate of lateral cross-flow within the HC rod bundle. These tests were conducted at atmospheric pressure, and encompassed a range of mass fluxes from 1000 kg/m2s to 3500 kg/m2s, HC rod twist pitches of 200cm, 100cm, and 50cm, and different hot water injection velocities and mixing lengths. Data from over 300 tests was analyzed, yielding a best fit correlation for use with any twist pitch, rod length, or coolant flow rate. Compared to the bare rod bundle, this correlation implies an enhancement in the intensity of turbulent interchange of 40% brought about by the HC geometry, and a 1.6% forced diversion of axial flow per subchannel, per quarterturn along the rod length. These parameters fit all data points considered within a standard deviation of 24%. Stochastic error was limited to ±16% by the use of precise temperature sensors. By applying this empirical mixing model to the subchannel representation of a BWR core featuring the HC rod design, a need to increase the flow area of the edge subchannels was demonstrated. This prompted a slight re-design of the HC fuel rod cross-section in order to make room for small spacer protrusions at the duct wall, to increase flow to peripheral subchannels. The modification was accomplished by reducing fin length, but increasing the inner diameter to maintain the reference fuel volume. The water rod region was also adjusted to maintain the reference assembly hydrogen to uranium atom ratio. With this modification, the model predicted a 24% allowable power uprate for the 200cm twist pitch HC core. Inlet and exit enthalpies were maintained from the reference cylindrical-rod core. When applied to a PWR core of HC rods, also with a fixed power to flow ratio, this empirical mixing model predicted an allowable power uprate of 47%, using traditional CHF correlations for cylindrical fuel. In subcooled conditions, CHF is known to be more sensitive to peaked areas of non-uniform heat-flux than in saturated two-phase flow conditions. Therefore power density gains will likely be dependent on the degree to which the rod twist would disrupt of nascent pockets of vapor; this effect should be further investigated experimentally. In order to further ascertain the potential gain in power density for the new design, an experiment must be carried out to obtain CHF data for the HC rod bundle. Two facilities with this aim were designed in great detail for BWR conditions: the first would operate using high pressure water at 7MPa, and the alternate would use a relatively low pressure refrigerant at equivalent conditions. The appropriate scaling laws were applied, which resulted in the choice of R134a as the simulant fluid. The R134a facility was found to be possible to construct at a greatly reduced cost.

Development of a Bayesian Network to Monitor the Probability of Nuclear Proliferation

2019-04-12T21:17:25Z

Development of a Bayesian Network to Monitor the Probability of Nuclear Proliferation Holcombe, Robert; Golay, Michael W. Nuclear Proliferation is a complex problem that has plagued national security strategists since the advent of the first nuclear weapons. As the cost to produce nuclear weapons has continued to decline and the availability of nuclear material has become more widespread, the threat of proliferation has increased. The spread of technology and the globalization of the information age has made the threat not only more likely, but also more difficult to detect. Proliferation experts do not agree on the universal factors which cause nations to want to proliferate or the methods to prevent countries from successfully developing nuclear weapons. Historical evidence also indicates that the current nuclear powers pursued their nuclear programs for different reasons and under different conditions. This disparity presents a problem to decision makers who are tasked with preventing further nuclear proliferation. Bayesian Inference is a tool of quantitative analysis that is rapidly gaining interest in numerous fields of scientific study that have previously been limited to purely statistical methods. The Bayesian approach removes the statistical limitations of large-n data sets and strictly numerical types of data. It allows researchers to include sparse and rich data as well as qualitative data based on the opinions of subject matter experts. Bayesian inference allows the inclusion of both the quantitative data and subjective judgments in the determination of predictions about a theory of interest. This means that contrary to classic statistical methods, we can now make accurate predictions with reduced information and apply this probabilistic method to problems in social science. The problem of nuclear proliferation is one that lends itself to a Bayesian analysis. The data set is relatively small and the data is far from consistent from country to country. There is however, a wide body of literature that seeks to explain proliferation factors and capabilities through both quantitative and qualitative means. This varied field can be brought together in a coherent method using Bayesian inference and specifically Bayesian Networks which graphically represent the various causal linkages. This work presents the development of a Bayesian Network describing the various causes, factors, and capabilities leading to proliferation. This network is constructed with conditional probabilities using theoretical insights and expert opinion. Bayesian inference using historical and real time events within the structure of the network is then used to give a decision maker an informed prediction of the proliferation danger of a specific country and inferences about which factors are causing it.

HLW Deep Borehole Design and Assessment: Notes on Technical Performance

2019-04-12T20:31:27Z

HLW Deep Borehole Design and Assessment: Notes on Technical Performance Jensen, K. G.; Driscoll, Michael J. This is a progress report covering work through mid-April 2010 under a Sandia-MIT contract dealing with design and siting/licensing criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. It consists of a collection of short technical notes which scope out the performance-related requirements of a deep borehole repository. Taken together the results highlight the need to focus on water transport as the dominant phenomenon. In this regard, I-129 is singled out as the likely limiting species because of its high, water chemistry-independent, solubility and long half life. Host rock thermal conditions are also examined, but found not likely to be a limiting constraint. They do, however, argue in favor of using a cluster of shorter multibranch boreholes rather than a much deeper single hole.

A Framework for Performance Assessment and Licensing of Deep Borehole Repositories

2026-01-15T20:37:44Z

A Framework for Performance Assessment and Licensing of Deep Borehole Repositories Jensen, K. G.; Driscoll, Michael J. This is the initial progress report under a Sandia-MIT contract dealing with development of engineering and geological siting criteria for deep borehole disposal of spent nuclear fuel or its separated constituents. Appendix C to this report reproduces the statement of work. The basic conceptual design used as the basis for assessment is presented, followed by screening for features, events and processes of special relevance, using criteria previously developed in the US for mined repository assessment: specifically those identified in the Environmental Impact Statement and Total System Performance Assessment protocols. Transport of radionuclides dissolved in water through highly impervious igneous bedrock (“granite”) is reaffirmed as the dominant mechanism of concern. The important beneficial role of deep-down water chemistry is also highlighted, in that low solubility under reducing conditions, retardation by adsorption, and inhibition of buoyancy and colloid formation by salinity, are all keys to assurance of effective sequestration. These insights are brought to bear to structure our future work scope.

Regional Examples of Geological Settings for Nuclear Waste Disposal in Deep Boreholes

2019-04-10T18:05:12Z

Regional Examples of Geological Settings for Nuclear Waste Disposal in Deep Boreholes Sapiie, B.; Driscoll, Michael J.; Jensen, K. G. This report develops and exercises broad-area site selection criteria for deep boreholes suitable for disposal of spent nuclear fuel and/or its separated constituents. Three candidates are examined: a regional site in the Baltic Fennoscandian shield for the fourteen nation European Repository Development Organization (ERDO) group of small European users; an area in the Arabian shield for newly announced reactor programs in several nations of the Middle East; and, following the same theme, a US site in Minnesota based on exploitation of the Canadian Shield. The criteria applied are restricted to technical, geological aspects and do not address the significant sociopolitical constraints faced by all repository programs. It is concluded that the subject sites all pass first-level technical criteria, and would thus be eligible for in-the-field follow-up, if so desired, by the cognizant organizations.

Impact of Alternative Nuclear Fuel Cycle Options on Infrastructure and Fuel Requirements, Actinide and Waste Inventories, and Economics

2019-04-12T20:31:23Z

Impact of Alternative Nuclear Fuel Cycle Options on Infrastructure and Fuel Requirements, Actinide and Waste Inventories, and Economics Guérin, Laurent; Kazimi, Mujid S. The nuclear fuel once-through cycle (OTC) scheme currently practiced in the U.S. leads to accumulation of uranium, transuranic (TRU) and fission product inventories in the spent nuclear fuel. Various separation and recycling options can be envisioned in order to reduce these inventories while extracting additional energy and sending the ultimate waste to a repository. Choosing one of these options has direct implications for the infrastructure requirements, natural uranium consumption, actinide inventories in the system, waste repository needs and costs. In order to account for the complexity of the nuclear enterprise, a fuel cycle simulation code has been developed using system dynamics (CAFCA). An economic module was added using spreadsheets. Four main advanced fuel cycle schemes are assessed here within the context of the US market: 1) the twice-through cycle scheme (TTC): single-pass plutonium recycling in thermal spectrum LWRs using Mixed OXide (MOX) fuel; 2) Multi-recycling of TRU in sodium-cooled fast spectrum burner cores, characterized by a fissile conversion ratio lower than 1 (FBu); 3) Multi-recycling of TRU in sodium-cooled fast breeders with a conversion ratio of 1.23 (FBr); and 4) A two-tier scenario: a TTC scheme is practiced as a transition scheme to fast reactors. The base case scenario assumes annual nuclear energy demand growth rate of 2.5% from 2020 on. The technologies for plutonium separation as well as MOX fuel fabrication are assumed to be available in 2025 while the first commercial fast reactors, as well as the possibility to recycle their spent fuel, are assumed to be available in 2040. For fast reactors, the cores are assumed to be TRU fueled, and the technology to separate the minor actinides is supposed to be available at the latest 5 years before deployment of fast reactors. Limits are applied on the building rate of reprocessing plants, which are also subject to a 80% minimum life-time loading factor requirement. It is found that, despite its higher cost, at the end of the century, the TTC scheme (single Pu-MOX recycle) does not lead to large improvements in terms of natural uranium consumption (16%), repository needs (considering both fission products and MA from reprocessing facilities, and spent MOX fuel) and TRU inventory reduction (although some shifting of TRU from storage to reactors occurs). This is especially significant because it is the only advanced fuel cycle option that can be deployed in large scale in the next few decades. However, if the primary reason for introduction of the more expensive fast reactors is resource enhancement and/or control of TRU in the nuclear waste, thermal reactor recycling allows the introduction of fast reactors to be delayed by 20-25 years. Moreover, once fast reactors are introduced, their deployment is accelerated compared to a 1-tier FR scenario. However, the two-tier scheme is the most expensive scheme as it combines the requirements of both the MOX technology and the FR technology. Sensitivity analyses were performed in order to assess the impact of secondary parameters. It is found that whatever the growth rate assumed, LWRs remain a significant part of the system at the end of the century, decades after fast breeders are introduced. The reason is the fissile materials required for fabrication of start-up cores considerably affect the rate at which fast reactors can be deployed. As a result, the choice of the core design (compact core vs. large core) may be as significant as the choice of the conversion ratio. For example, the breeder scenario (CR=1.23) may lead to the same cumulative natural uranium consumption reduction (by 2100) as the self-sustaining reactors (CR=1.0) while leading to larger TRU inventory in the system and requiring greater fast reactor fuel reprocessing capacity. Allowing fast reactors to start with uranium only cores was not considered, as it will likely limit resource enhancement benefits of fast reactors. Still, in general, the higher the conversion ratio, the greater the fast reactor installed capacity, hence the greater the savings in natural uranium. Conversely, the best reduction in TRU from the OTC amount is obtained by the lower conversion ratio (45% for a pure burner with conversion ratio 0.0 by 2100). Doubling the minimum cooling time before reprocessing for all fuel types from 5 years to 10 years slows down the deployment of the fast reactors and therefore reduces their share in the total installed capacity. This is almost equivalent to replacing breeders with fast reactors with a conversion ratio of 0.75. Finally, the results show that starting the separation of the TRU 10 years prior to introduction of the fast reactors instead of 5 years provides a mid-term advantage (faster initial deployment) that vanishes within 25 years. In the long term, the fast reactor penetration results are insensitive to the assumed industrial capacity to build reprocessing facilities for the base case or at lower nuclear energy growth rates. However, the assumed industrial capacity can be a real constraint if the nuclear energy growth rates are 4% or higher.

UPDATE ON THE COST OF NUCLEAR POWER

2019-04-10T18:05:19Z

UPDATE ON THE COST OF NUCLEAR POWER Du, Yangbo; Parsons, John E. We update the cost of nuclear power as calculated in the MIT (2003) Future of Nuclear Power study. Our main focus is on the changing cost of construction of new plants. The MIT (2003) study provided useful data on the cost of then recent builds in Japan and the Republic of Korea. We provide similar data on later builds in Japan and the Republic of Korea as well as a careful analysis of the forecasted costs on some recently proposed plants in the US. Using the updated cost of construction, we calculate a levelized cost of electricity from nuclear power. We also update the cost of electricity from coal- and gas-fired power plants and compare the levelized costs of nuclear, coal and gas. The results show that the cost of constructing a nuclear plant has approximately doubled. The cost of constructing coal-fired plants has also increased, although perhaps just as importantly, the cost of the coal itself has spiked dramatically, too. Capital costs are a much smaller fraction of the cost of electricity from gas, so it is the recent spike in the price of natural gas that has contributed to the increased cost of electricity. These results document changing prices leading up to the current economic and financial crisis, and do not incorporate how this crisis may be currently affecting prices.

A Benchmark Study of Computer Codes for System Analysis of the Nuclear Fuel Cycle

2019-04-12T20:31:22Z

A Benchmark Study of Computer Codes for System Analysis of the Nuclear Fuel Cycle Guérin, Laurent; Feng, Bo; Hejzlar, Pavel; Forget, Benoit; Kazimi, Mujid S.; Van Den Durpel, Luc; Yacout, Abdellatif; Taiwo, Temi; Dixon, Brent W.; Matthern, Grechen; Boucher, Lionel; Delpech, Marc; Girieud, Richard; Meyer, Maryan As use of nuclear energy is expected to expand in different parts of the world, several codes that describe the nuclear fuel cycle system are currently under development, featuring a range of capabilities and different levels of flexibility and automation. Such codes model the addition or retirement of reactors, the demand for fresh fuel, and the need for spent fuel storage and recycling facilities as the production of nuclear energy varies with time. The codes enable analysis of various scenarios for the evolution of the nuclear energy system, and the timing of deployment of new facilities. Outputs may also include fuel material mass flows, economic analysis and metrics related to spent fuel or waste assessment. The study reported here is the first attempt for benchmarking the MIT code CAFCA against three independently developed fuel cycle simulation codes. It is also among the first publicly available benchmark exercises. Some reviews of the existing codes were previously reported, but focused mostly on their theoretical capabilities. Benchmarking studies, generally involving two or three codes, have been done over the last few years, but most remain unpublished. The codes included in this study are: CAFCA (developed at MIT), COSI (developed at CEA, France), DANESS (developed at ANL) and VISION (developed by DOE laboratories for the AFCI program). The purpose of this benchmark study is to evaluate the degree of convergence of the current versions of these codes and to compare their basic methodologies. This benchmark is not a comprehensive analysis of all the codes’ capabilities but constitutes a first step towards a more complete benchmark study. In order to compare all 4 codes, only the common capabilities were considered and assessed. Those capabilities are essentially those of CAFCA, as it is the simplest code. Consequently, some of the advanced capabilities of the other codes were disabled, and their complete features were not reflected in this benchmark study. In addition, economic evaluation of the fuel cycles was not considered, even though it is a capability common to the four codes. Furthermore, the initial runs showed that a degree of freedom should be removed to ease the comparison. For that reason reprocessing capacity profiles were provided by CAFCA and used as input by the other codes. Following a description of the codes, the report presents the four scenarios selected as the benchmark cases, including initial conditions. The time period of the simulation covers the 21st century. Those scenarios differ from each other in either the nuclear energy supply growth rate (0%, 1.5% or 3%) or the type of advanced technology introduced in the midterm. The options for advanced reactors were: the “self-sustaining” fast reactor (with a fissile conversion ratio of one), fast burners of transuranics (TRU), or a combination of plutonium recycling (as mixed oxide) in the thermal light water reactors (LWRs) and fast burners. The scenarios, and hence the results, are for benchmarking purposes only and should not be considered realistic for policy studies or forecasts about the future of nuclear power. The set of constraints specified is minimal and only intended to provide a common framework for the simulations. The results are presented and commented on for each case. The first case, characterized by a very constraining zero energy growth rate, shows an excellent agreement among the codes, with identical ratios of fast reactors/thermal reactors over time. This excellent agreement was the iv result of the particular efforts made in order to get very close results (several iterations were performed to allow for adjustments). This case eventually shows how the models can produce very close results if sufficiently tuned to adhere to the same basic assumptions. This case also allowed us to identify a number of minor apparent discrepancies and explain them. In particular, it made obvious differences in results between COSI, which was tuned to track fuel batches (“discrete-flow code”) and CAFCA, DANESS and VISION, which deal with annual mass flows (”continuous-flow codes). The treatment of discrete batches of fuel by COSI, instead of timeaveraged quantities in the other codes, result in somewhat oscillatory flows and inventories of materials. Another factor leading to discrepancies among the codes is the time assumed to exist between the separation of transuranics and the manufacturing of fast reactor fuels. CAFCA speeds up the fuel manufacturing, to avoid the presence of separated transuranics in large quantities, while the other codes do not (as Pu-containing fuel has a very limited shelf-life at the initial fissile content due to Pu 241 decay with a half life of about 14.4 years). Unlike the first cases, there was in the three other cases no attempt, beyond the common set of assumptions, to iterate to get the results of the other cases to converge. Therefore, these three cases are more a reflection of how the codes actually operate and show the level of variation in results that should be considered normal. These three cases were of great interest for comparing the different strategies for fast reactor deployment and their dependence for fuel on available TRU from the operation of light water reactors. Overall, the benchmark shows that, although the codes exhibit reassuring consistency, both internally and among themselves, differences still exist. The fact that reprocessing capacity profiles were externally provided may have disturbed some of the codes designed to internally calculate this variable. Although limitations inherent in the codes exist, the differences in the results generally do not reveal major flaws, but rather reflect differing assumptions and constraints embedded in the methods and approximations of the calculations. This benchmark reveals (or reminds us) that there is no single profile for a fuel cycle scenario but several profiles that depend on industrial practices with regards to manufacturing, storage and reprocessing of nuclear fuels. These practices may aim at differing priorities of reducing stocks of stored spent fuel, avoiding the presence of separated TRU to reduce proliferation risks, ensuring sufficient supply of fresh fuel for advanced reactors and spent fuel for reprocessing plants, and minimizing some costs. Such choices can either be intrinsic to the code (through built-in assumptions) or through user choices (for example, the level of conservatism in the algorithm ensuring fuel supply for fast reactors is a user input). Moreover, the parameters left to the user’s discretion are generally not the same from one code to another, or are expressed in different terms. Finally, complete consistency between the codes is difficult to obtain. Two major conclusions can be drawn from this benchmark. First, the overall results show good consistency and similar trends. Hence, utilization of various codes is likely to lead to similar conclusions. Second, one must not expect the various fuel cycle system simulation codes to provide identical outputs. Therefore, users must keep in mind that, although the results are internally consistent and meet each code’s set of requirements, they do not project unique scenarios for meeting such requirements.

MCODE, Version 2.2: An MCNP-ORIGEN DEpletion Program

2019-04-10T18:05:13Z

MCODE, Version 2.2: An MCNP-ORIGEN DEpletion Program Xu, Z.; Hejzlar, Pavel MCODE Version 2.2 is a linkage program, which combines the continuous-energy Monte Carlo code, MCNP-4C, and the one-group depletion code, ORIGEN2, to perform burnup calculations for nuclear fission reactor systems. MCNP is used as the advanced physics modeling tool providing the neutron flux solution and detailed reaction rates in the pre-defined spatial burnup zones. ORIGEN, in turn, carries out multi-nuclide depletion calculations in each region and updates the corresponding material composition in the MCNP model. The MCNP/ORIGEN coupling follows the predictor-corrector approach. During a burnup timestep, end-of-timestep material compositions are first predicted based on the flux solution at the beginning-of-timestep. Using the predicted end-of-timestep material compositions, an MCNP run is performed to compute the neutron flux and detailed reaction rates, which are then used in a corrector burnup step. The final end-oftimestep material compositions are obtained as the average value of the results from the predictor and corrector steps. As a stand-alone code written in ANSI C, MCODE-2.2 is portable between Windows personal computers (PC’s) and UNIX/Linux machines. There are three utility programs in MCODE-2.2: (1) preproc to pre-process MCNP/ORIGEN libraries; (2) mcode as the console to run steady-state burnup/decay calculations; and (3) mcodeout to collect results from scattered data files under temporary directory and produce a detailed output. Further, there is an auxiliary program called mcnpxs, which is for the purpose of preparing a nuclide summary table of continuous energy MCNP cross section libraries. The routine usage of MCODE-2.2 only requires a tandem running of the three utility codes. The auxiliary code, mcnpxs, is intended to help users during the code installation/setup. Compared to other similar linkage codes, MCODE-2.2 emphasizes functionality, versatility and usability. Several features of the code follow: (1) The execution of MCNP and ORIGEN is in an automatic fashion. (2) All standard nuclear reaction types in ORIGEN2 are considered: capture, fission, (n,2n), (n,3n), (n,p), and (n,α). Therefore, both the nuclear fuel depletion and material irradiation/activation (e.g., boron-10 irradiation) can be handled. (3) A power history can be specified, i.e., power level at each timestep. The default depletion option is constant power depletion. Meanwhile, an iterative robust flux depletion scheme is available. In addition, decay calculations are also possible. (4) With appropriate ORIGEN one-group cross section libraries, users can rely on MCODE- 2.2 to automatically select important nuclides based on absorption ranking from ORIGEN isotope reservoir for MCNP calculations. (5) The enhanced predictor-corrector approach (consistent with CASMO-4) increases the accuracy with negligible computational cost increase. From the user’s point of view, MCODE-2.2 is an extension of normal MCNP criticality (kcode) calculations. The MCNP input inherits the MCODE-2.2 input in the form of a fourth paragraph (added at the end of the MCNP input deck) containing the burnup-related data and MCNP/ORIGEN calculation controls. A user-supplied equilibrium MCNP source file can also be provided, which might save CPU time by reducing the number of initial MCNP inactive cycles. iii The Monte Carlo burnup code has some unique characteristics, one of which is that all results are in nature stochastic. The statistical uncertainty passing through burnup calculations is one concern, which is believed by some people as the weakness or even indication of the Monte Carlo limitations to perform burnup calculations. Using a multiregion Gd-poisoned BWR 8×8 assembly depletion problem, it is shown that the random statistical uncertainties are benign and cancel each other with the burnup. In addition, a single PWR unit cell benchmark problem is documented. Comparison of results against CASMO-4 yields satisfactory agreement.

A System Dynamics Study of the Nuclear Fuel Cycle with Recycling: Options and Outcomes for the US and Brazil

2019-04-10T19:50:00Z

A System Dynamics Study of the Nuclear Fuel Cycle with Recycling: Options and Outcomes for the US and Brazil Busquim e Silva, R.; Kazimi, Mujid S.; Hejzlar, Pavel A system dynamics simulation technique is applied to generate a new version of the CAFCA code to study mass flows in the nuclear fuel cycle, and the impact of different options for advanced reactors and fuel recycling facilities on the inventory and distribution of transuranics (TRU). Several nuclear fuel cycle options are studied for U.S. and Brazil markets, and special consideration is given to potential collaboration between the two countries. This includes the impact of advanced nuclear technologies, under a prescribed growth in demand for nuclear electricity, on demand for uranium resources, uranium enrichment, and fuel reprocessing facilities, and on total cost of nuclear electricity over the next few decades. Introduction of fuel recycling reduces the growing demand for uranium, and the long-term need for storage of radioactive spent fuel. However, the timing of introduction of recycling is important for proper technology development, and the rate of deployment is restrained by the industrial capacity as well as the desire for high utilization factor of the deployed facilities over their life time, and that is reflected in the assessments. The nuclear fuel cycle is modeled as a high level structure diagram, which provides an overview of the interconnections among its blocks without showing all details, and as a structure-policy diagram which details the decision rules applied to the structure. The high level structure diagram represents the nuclear fuel cycle; the fleet of thermal and fast reactors; the separation and reprocessing plants; the waste repository; the spent fuel storage; and the paths for the fuel and waste mass transfer. In addition, an economic model is added to study different cases under the same assumptions. The economic model is based on the forecasted need for advanced reactors and recycling facilities, assuming that all costs are recovered within the nuclear energy system. Different recycling technology options are included in the code: (1) Thermal recycling in LWRs using Combined Non-Fertile and UO2 Fuel (CONFU), (2) Recycling of TRU in fertile-free fast cores of Actinide Burner Reactors (ABR); and (3) Fast recycling of TRU with UO[subscript 2] in self-sustaining Gas-cooled Fast Reactors (GFR). Case studies for different advanced technology introduction dates and for distinct TRU depletion rates are examined. In particular, the code is 3 equipped to simulate the introduction of two recycling technology options with a prescribed allocation of the TRU supply between them. The simulation results show that early introduction of the GFR recycling scheme leads to the most significant reduction in uranium consumption and enrichment requirements, thus delaying the eventual depletion date of uranium ore. The GFR requires less uranium resources due to the use of TRU as recycled fuel and near unity fissile conversion ratio. However, in a nonbreeding reactor system, the consumption of U continues to grow, and the TRU needed to start fast reactors will be growing at a constrained rate. On the other hand, the CONFU recycling scheme keeps the TRU inventory in the entire system well below other schemes, and guarantees equilibrium between the generation and consumption of transuranics without investments in fast reactors. CONFU incinerates more TRU than the GFR and ABR schemes during the simulation period. Also, it reduces the TRU sent to the repository for disposal by orders of magnitude. The ABR scheme does the same but requires the introduction of fast reactors. Nevertheless, the CONFU and ABR schemes have no significant impact on the amount of uranium resources consumption or enrichment requirements. Economic analysis indicates that the CONFU technology is more attractive at current uranium prices, and that fast recycling becomes as attractive as thermal recycling at higher uranium prices. The results also show that if a nuclear fuel cycle state/reactor state collaboration with Brazil is started, there will be a significant impact on the U.S. cumulative TRU inventory at interim storage, enrichment requirements, uranium consumption, and number of advanced fuel facilities. The results show that a nuclear partnership without the introduction of advanced nuclear technologies would not have advantages for the U.S. Furthermore, a nuclear collaboration allows a higher ratio of fast reactors to total installed nuclear electric capacity in the U.S.

Thermal Hydraulic Analysis of a Low Enrichment Uranium Core for the MIT Research Reactor

2019-04-12T20:31:21Z

Thermal Hydraulic Analysis of a Low Enrichment Uranium Core for the MIT Research Reactor Ko, Yu-Chih; Hu, Lin-Wen; Kazimi, Mujid S. The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The objectives of this study are to benchmark the in-house computer code for the MITR, and to perform thermal hydraulic analyses in support of the LEU design studies. The in-house multi-channel thermal-hydraulics code, MULCH-II, was developed specifically for the MITR. This code was validated against PLTEMP for steady-state analysis, and against RELAP5 and temperature measurements for the loss of primary flow transient. The benchmark analysis results showed that the MULCH-II code is in good agreement with other computer codes and experimental data, and hence it is used as the main tool for this study. Various fuel configurations are evaluated as part of the LEU core design optimization study. The criteria adopted for the LEU thermal hydraulics analysis in this study of the limiting safety system settings (LSSS), are to prevent onset of nucleate boiling during steady-state operation, and to avoid a clad temperature excursion during the loss of flow transient. In ranking the LEU core design options, the primary parameter is a low power peaking factor in order to increase the LSSS power and to decrease the maximum clad temperature during the transient. The LEU fuel designs with 15 to 18 plates per element, fuel thickness of 20 mils, and a hot channel factor less than 1.76 are shown to comply with the thermal-hydraulic criteria. The steady-state power can potentially be higher than 6 MW, which was requested in the power upgrade submission to the Nuclear Regulatory Commission.

PWR Transition to a Higher Power Core Using Annular Fuel

2019-04-12T20:31:21Z

PWR Transition to a Higher Power Core Using Annular Fuel Beccherle, J.; Hejzlar, Pavel; Kazimi, Mujid S. The internally and externally cooled annular fuel is a new type of fuel for PWRs that enables an increase in core power density by 50% within the same or better safety margins as traditional solid fuel. Each annular fuel assembly of the same side dimensions as the 17x17 solid fuel assembly has 160 annular fuel rods arranged in a 13x13 array. Even at the much higher power density, the fuel exhibits substantially lower temperatures and a Minimum Departure From Nucleate Boiling (MDNBR) margin comparable to that of traditional solid fuel at nominal (100%) power. The major motivation for such an uprate is reduction of electricity generation cost. Indeed, the capital cost per kWh(e) of a new reactor would be smaller than the standard construction of a new reactor with solid fuel. Elaborating on previous work, we study the economic payoff of an uprate of existing PWRs given the expected cost of equipment and also cost of money using different assumptions. The fate of the already bought solid fuel is investigated. It is demonstrated that the highest return on investment is obtained by gradually loading annular fuel in the reactor core such that immediately before shutting the reactor down for the uprate construction, two batches in the core are of the annular fuel. This option implies running a core with a mixture of both annular and solid fuel assemblies. In order to prove the technical feasibility of such an option, the thermalhydraulics of this mixed core is investigated and the MDNBR is found to be either unaffected or improved. Consequently, a neutronic model is developed to verify and validate the neutronic feasibility of the transition from solid to annular fuel. This involvements assessment of the peaking factors and capability to provide control poisons within allowable concentrations The overall conclusion of this work is that annular fuel is a very promising option for existing reactors to increase their power by 50%, as it enables a significant uprate with an attractive return on investment. We show that, by a smart management of the transition, an internal return on investment of about 22–27% can be achieved.

Sensitivity of Economic Performance of the Nuclear Fuel Cycle to Simulation Modeling Assumptions

2019-04-12T20:31:20Z

Sensitivity of Economic Performance of the Nuclear Fuel Cycle to Simulation Modeling Assumptions Bonnet, Nicephore; Kazimi, Mujid S. Comparing different nuclear fuel cycles and assessing their implications requires a fuel cycle simulation model as complete and realistic as possible. In this report, methodological implications of modeling choices are discussed in connection with development of the MIT fuel cycle simulation code CAFCA. The CAFCA code is meant to find the recycling facilities deployment rate that minimizes the time by which spent fuel in storage today is used up in order to lead to a nuclear fuel cycle with minimum inventory of transuranic elements. The deployment is constrained by the construction capacity of the recycling plants and by the economic requirement that their loading factor never drops below a certain level. First, through a simplified fuel cycle model, it is analytically proven that an optimum solution is to build recycling plants at full speed up to a certain point in time b, then to suspend construction until interim storage is completely depleted. The shape of the optimum solution is injected into an algorithm based on a complete model of the fuel cycle. An iterative process yields the value of b assuring depletion and satisfactory loading factors. Besides providing rigorous optimization, the analytical solution underpinning the CAFCA algorithm is expected to reduce considerably the vulnerability of the results to numerical discontinuities. Degradation of fuel quality with time in interim storage occurs due to the decay of Pu241 into Am241. While an obvious approach to track such effects is to couple the fuel cycle code with a neutronics/decay code (ORIGEN for example), it is more efficient to derive explicit equations from a simplified irradiation and decay model, allowing for analytical tracking of the fuel composition. This approach was implemented in CAFCA. All fuel cycle simulation refinements do not present the same level of importance. One should focus on the dominant parameters, i.e., those contributing most to results sensitivity. The important parameters are determined through a sensitivity study using a novel U.S. thermal recycling scenario called CONFU as a reference case. The CONFU technology is assumed to be commercially introduced 15 years from now, with an industrial capacity allowing the construction of one 1000 MT/year spent fuel separation plant every two years. Additionally, it is assumed that discharged CONFU batches remain in cooling storage for 6 years, reactors have a 60-year lifetime and economic recovery period of 20 years, and are half financed by equity with a rate of return of 15%. It is found that the cost of electricity is most sensitive to the reactors lifetime, since taking it back to a nominal value of 40 years would result in a 44% increase in the cost of electricity. Next in importance is the financing structure of the fleet. The addition of three points to the rate of return on equity would increase the cost of electricity by 14%. While scale effects are locally very beneficial in that they substantially reduce recycling plants operation costs, they prove to be of limited interest from an overall fuel cycle point of view. Using the scale effect model in CAFCA-II, doubling the separation plants capacity yields a 3% reduction of the cost of electricity. The fuel cycle presents good robustness with respect to fuel decay time degradation. Increasing CONFU batches cooling time to 18 years causes a 2% increase in the cost of electricity.

AN EVOLUTIONARY FUEL ASSEMBLY DESIGN FOR HIGH POWER DENSITY BWRs

2019-04-09T18:44:10Z

AN EVOLUTIONARY FUEL ASSEMBLY DESIGN FOR HIGH POWER DENSITY BWRs Karahan, A.; Buongiorno, Jacopo; Kazimi, Mujid S. An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water gap regions with a single large assembly. The traditional BWR cylindrical UO[subscript 2]-fuelled Zr-clad fuel pin design is retained, but the pins are arranged on a 22×22 square lattice. There are 384 fuel pins with 9.6 mm diameter within a large assembly. Twenty-five water rods with 27 mm diameter maintain the moderating power and accommodate as many finger-type control rods. The total number and positions of the control rod drive mechanisms are not changed, so existing BWRs can be retrofitted with the new fuel assembly. The technical characteristics of the large fuel assembly were evaluated through a systematic comparison with a traditional 9×9 fuel assembly. The pressure, inlet subcooling and average exit quality of the new core were kept equal to the reference values. Thus the power uprate is accommodated by an increase of the core mass flow rate. The findings are as follows: - VIPRE subchannel analysis suggests that, due to its higher fuel to coolant heat transfer area and coolant flow area, the large assembly can operate at a power density 20% higher than the traditional assembly while maintaining the same margin to dryout. - CASMO 2D neutronic analysis indicates that the large assembly can sustain an 18- month irradiation cycle (at uprated power) with 3-batch refueling, <5wt% enrichment with <60 MWD/kg average discharge burnup. Also, the void and fuel temperature reactivity coefficients are both negative and close to those of the traditional BWR core. - The susceptibility of the large assembly core to thermalhydraulic/neutronic oscillations of the density-wave type was explored with an in-house code. It was found that, while well within regulatory limits, the flow oscillation decay ratio of the large assembly core is higher than that of the traditional assembly core. The higher core wide decay ratio of the large assembly core is due to its somewhat higher (more negative) void reactivity coefficient. - The pressure drop in the uprated core is 17 % higher than in the reference core, and the flow is 20% higher; therefore, larger pumps will be needed. - FRAPCON analysis suggests that the thermo-mechanical performance (e.g., fuel temperature, fission gas release, hoop stress and strain, clad oxidation) of the fuel pins in the large assembly is similar to that of the reference assembly fuel pins. - A conceptual mechanical design of the large fuel assembly and its supporting structure was developed. It was found that the water rods and lower tie plate can be used as the main structural element of the assembly, with horizontal support being provided by the top fuel guide plate and core plate assembly, and vertical support being provided by the fuel support duct, which also supports the finger-type control rods.

A PWR Self- Contained Actinide Transmutation System

2019-04-12T20:31:26Z

A PWR Self- Contained Actinide Transmutation System Shatilla, Y.; Hejzlar, Pavel; Kazimi, Mujid S. Elements of the new Global Nuclear Energy Partnership (GNEP) initiative in the US call for the expansion of domestic use of nuclear power and the minimization of nuclear waste. To achieve both goals in the short term the transmutation of trans-uranic (TRU) elements in Combined Non- Fertile and Uranium (CONFU) PWR fuel assemblies is evaluated. These assemblies are composed of a mix of standard UO[subscript 2] fuel pins and pins made of recycled TRU in an inert matrix and are designed to fit in currently deployed PWRs. A CONFU-Self-Contained (CONFU-C) assembly is shown to achieve a net TRU destruction in a self-contained TRU multi-recycling system. The system may consist of as little as one currently operating reactor that does not depend on other reactors to supply it with its inventory of recycled TRU. This is considered a major advantage of the new design over its predecessors since it eliminates the need for designating a whole fleet of CONFU reactors to produce recycled TRU for the reactor under consideration. Degradation of fissile content of the multi-recycled TRU is compensated for by drawing from legacy TRU that already comes from standard UO2 spent fuel and the usage of fresh UO[subscript 2] fuel with different enrichments depending on fuel cooling time after discharge. A recycling strategy which uses a 4.5 year period of in core irradiation, followed by one of three cooling periods (6-, 18-, and 32-year) after discharge and reprocessing is considered. Calculations have shown the equilibrium CONFU-C assembly can have a net TRU destruction of approximately 20.6 (for the 6-yr cooling) and 2.7 (for the 18-yr cooling) kg of TRU per TWhe, as compared to 11.0 kg of TRU per TWhe for the CONFU-B with a 6-yr cooling period. This represents a net burning rate of ~13% (6-yr cooling) and 3% (18-yr cooling) of the TRU loaded per assembly compared to 8% for the CONFU-B design. However, Fuel Cycle Costs (FCC) for the equilibrium CONFU-C is shown to be 12.8 (6yr-cooling) and 14.2 (18yr cooling) mills/KWhe and that for the CONFU-B to be 12 mills/KWhe. Due to the relatively long cooling period of the third option (32 yr cooling), a CONFU-C assembly could not be designed to achieve net TRU destruction in a self-contained manner.

Fuel Cycle Options for Optimized Recycling of Nuclear Fuel

2019-04-12T20:31:26Z

Fuel Cycle Options for Optimized Recycling of Nuclear Fuel Aquien, A.; Kazimi, Mujid S.; Hejzlar, Pavel The reduction of transuranic inventories of spent nuclear fuel depends upon the deployment of advanced fuels that can be loaded with recycled transuranics (TRU), and the availability of facilities to separate and reprocess spent fuel. Three recycling strategies are explored in this study: (1) Recycling in thermal Light Water Reactors (LWR) using CONFU technology (COmbined Non-Fertile and UO[subscript 2] fuel), (2) recycling of TRU in fertile-free fast cores of Actinide Burner Reactors (ABR), and (3) recycling of TRU with UO[subscript 2] in self-sustaining Gas-cooled Fast Reactors (GFR). Choosing one strategy over another involves trade-offs. The CONFU, ABR, and GFR strategies differ from each other in terms of T RU loading in the reactor, net TRU incineration, capacities of recycling facilities needed, date for technology option availability, and flexibility. The CONFU and GFR are assumed to achieve zero net TRU incineration, while the ABR is a net consumer of TRU. The TRU loading is greatest in GFR and lowest in CONFU. While both CONFU and ABR require separation (of TRU from U) and reprocessing (recycling of TRUs from fertile-free fuel), the GFR is designed to, in equilibrium, recycle TRU+U after extraction of the fission products. It is assumed that thermal recycling is available in the short-term (2015), as opposed to recycling in fast reactors (2040). Finally, thermal recycling is the most flexible as either CONFU batches or regular LWR uranium batches can be loaded; the issue of running out of TRU fuel is therefore irrelevant for this option. A fuel cycle simulation tool, CAFCA II (Code for Advanced Fuel Cycles Assessment) has been developed. The CAFCA II code tracks the mass distribution of TRU in the system and the cost of all operations. The code includes a specific model for deployment of recycling plants, with certain capacities and investment requirements. These facilities may operate with a minimum target capacity factor during the lifetime of the plant. The deployment of these facilities is also constrained by a user-specified ability to add recycling capacity within a given time interval. Finally, the CAFCA II code includes a specific model for recycling prices which reflects the economies of scale that go with increases in the nominal capacity of recycling plants. Two case studies are presented. The first explores the optimal fuel cycle and recycling plant capacities as a function of the deployment of advanced fuel cycle technologies over the next hundred years, under the assumption of the US demand for nuclear energy growing at a 2.4% annual rate. Key figures for comparison of the strategies are evaluated, including reduction of TRU interim storage requirements, maximization of TRU incineration, minimization of the size of the fleets of recycling plants and fast reactors, fuel cycle cost, and capital cost requirements. We found that it is not possible to minimize the construction rate of advanced reactors and advanced spent fuel recycling facilities simultaneously with the construction rate of separation facilities for UO[subscript 2] spent fuel. The latter was found to be more constraining than the former. Further, we found that reactor technologies with zero net TRU destruction rate can achieve total depletion of TRU inventories in spent fuel interim storage at a lower fuel cycle cost and with fewer recycling facilities than reactor technologies that incinerate TRU. However, the lower fuel cycle cost is achieved at the expense of a smaller reduction of total TRU inventories. Finally, if the construction rate of advanced nuclear technologies is large enough, the later introduction date of fast recycling schemes compared to thermal recycling schemes does not prevent the reduction of TRU inventories in interim storage by 2100. Recently, the potential benefits of multi-lateral approaches to the management of nuclear fuel have been widely discussed. These include cost attractiveness following from economies of scale, proliferation resistance, and more efficient nuclear waste treatment strategies. CAFCA II has been developed to quantify these implications for the back-end of the fuel cycle. Three partnership scenarios have been examined: first, a scenario where the “Fuel leasing/fuel take-back” concept is implemented; second, a scenario with “Limited Collaboration” at the back-end fuel cycle, where spent fuel recycling and advanced fuel fabrication are externalized in countries that have these technologies; and third, a scenario of “Full Collaboration”, under which two regions fully collaborate at the fuel cycle back-end: spent fuel inventories and advanced fuel cycle facilities are co-owned and co-managed. The second case study concentrates on optimizing the choice of (1) fuel cycle option, (2) recycling plant capacities, and (3) partnership scenario by analyzing the implications of these choices for the LWR-CONFU, LWR/ABR, and LWR/GFR strategies. The nuclear fuel cycle is simulated in a two-region context from 2005 to 2100 under the assumption that one region represents the US growing at a 2.4% annual rate and the other region represents Brazil, Indonesia, and Mexico growing at a 7.4% annual rate until 2080, and 2.4% afterwards. Under this scenario, we found that a US partnership with a region representing Brazil, Mexico, and Indonesia could be advantageous to the reduction of TRU storage in both regions if the construction rate of UO[subscript 2] spent fuel separation plants would be larger than one 1,000 MT/yr plant every two years after 2050. From the point of view of the spent fuel recycling industry, use of the largest recycling plants with the lowest construction cost per unit of installed capacity becomes optimal only with multi-national approaches to the fuel cycle back-end.

Actinide Minimization Using Pressurized Water Reactors

2019-04-12T20:31:25Z

Actinide Minimization Using Pressurized Water Reactors Visosky, M.; Kazimi, Mujid S.; Hejzlar, Pavel Transuranic actinides dominate the long-term radiotoxicity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease the long-term burden on the repository. Examining the maximum possible burning of trans-uranic elements in Combined Non-Fertile and UO[subscript 2] (CONFU) PWR assemblies is evaluated. These assemblies are composed of a mix of standard UO[subscript 2] fuel pins and pins made of recycled trans-uranics (TRU) in an inert matrix, and are designed to fit in current or future PWRs. Applying appropriate limits on the neutronic and thermal safety parameters, a CONFU-Burndown (CONFU-B) assembly design is shown to attain net TRU destruction in each fuel batch through at least 9 recycles. This represents a time span of nearly 100 years of in-core residence and out-of-core storage time. In this way, when the TRU is multirecycled, only fission products and separation/reprocessing losses are sent to the repository, and the initial inventory of TRU is reduced over time. Thus, LWRs are able to eventually operate in a fuel cycle system with an inventory of transuranic actinides much lower than that accumulated to date. Three recycling strategies are considered, all using a 4.5-year in core irradiation, followed by cooling and reprocessing. The three strategies involve a short-term cooling (6-year) after discharge, a longer-term cooling (16.5-year) after discharge, or a strategy called Remix. The Remix strategy involves partitioning the Pu/Np after 6-year cooling for immediate recycle, and partitioning the Am/Cm for an additional 10.5-year cooling before remixing it into the next CONFU-B batch. At equilibrium, the CONFU-B can burn approximately 1.5 kg to 10.0 kg of TRU per TWhe depending on the recycle strategy used. This represents a net burning rate of 2- 8% of the TRU loaded per assembly, in addition to burning an amount equivalent to the TRU produced in the UO[subscript 2] pins. However, the highly heterogeneous nature of these assemblies can result in fairly high intraassembly pin power peaking. By design, an IMF pin in the assembly carries the highest power to maximize the TRU destruction. For the initial TRU loading, the highest power peaking in an IMF pin is 1.183. This is compensated by having cooler pins in the immediate vicinity. Even so, the pin peaking distribution in the assembly can result in reduced thermal margins. The assembly mentioned above has an MDNBR of 1.43, instead of 1.62 for the all-UO[subscript 2] assembly, based on a core-wide radial peak-to-average assembly power peaking of 1.50. Use of neutron poisons and tailored enrichment schemes reduces the neutronic reactivity of fresh assemblies, while improving MDNBR to 1.51. In addition, RELAP was used to evaluate the fuel behavior under large break LOCA conditions. CONFU-B performance under these conditions was comparable to the standard all-UO2 assembly. Several options for spent fuel recycling in LWRs are compared economically, and all are found to be more costly than making fresh UO2 fuel from mined ore. However, the CONFU-B strategy is less costly on a mills/kWhe basis than other thermal recycling strategies that recycle the full TRU vector. Given OECD estimates for the unit costs of each fuel type, and assuming 10% carrying charge factor, this cost is 10.0 mills/kWhe for the CONFU-B recycle, compared to 22.2 mills/kWhe for MOX-UE and 5.4 mills/kWhe for all UO[subscript 2]. Note that these FCCs assume the 2 disposal fee collected during power generation of a previous cycle can be invested while the fuel is cooling and provide a credit to the cycle that uses the fuel after reprocessing. The fuel handling challenges of multirecycling TRU in CONFU-B assemblies are compared to other multi-recycling strategies. If we assume that the spent fuel from the seventh recycle in each strategy is no longer recyclable and must be sent to the repository in its entirety, the CONFU-B strategy still places much less total burden on the repository than the once-through cycle, and even less burden than the current MOX cycle. Finally, a methodology for calculating the time integrated proliferation risk of a fuel cycle is introduced. An innovation of this methodology is the discounting of future risks to calculate an overall present value risk of a given cycle. Under this methodology, the CONFU-B presents lower risks than other multi-recycling strategies in the first 100 years. For a 10% rate of discount of risk, the CONFU-B risks are comparable to the once-through cycle. The longer term risk favors recycling due to the limited accumulation of repository risk.

Experimental Determination of Thermal Conductivity of a Lead- Bismuth, Eutectic-Filled Annulus

2019-04-10T18:05:09Z

Experimental Determination of Thermal Conductivity of a Lead- Bismuth, Eutectic-Filled Annulus Carpenter, David M.; Kohse, Gordon E. In order to obtain an accurate prediction of the thermal behavior of an annular fuel assembly (see MIT-NFC-PR-048 for a description of the rods), the thermal conduction of the region from the outside of the fuel capsule to the reactor coolant (within the test assembly) must be known. The effective thermal conductivity of this composite structure is dependent on the interaction of the parts via various physical phenomena, and therefore is difficult to infer accurately from the conductivity of the constituent materials. A mock-up of the annular fuel rod containment thimble was created to allow the conductivity of the annular lead bismuth eutectic-filled gap to be measured. An electric rod heater was used to provide temperatures similar to the in-core environment, and conductivity was determined based on thermocouple temperature readings at various points across the gap. A second series of experiments substituted a steel tube for the aluminum thimble, and used a bucket of stationary water as coolant. The purpose of these changes was to increase the temperature of the eutectic and achieve a larger melted fraction, while at the same time creating a large enough temperature drop across the gap to allow reliable measurements. A third series of experiments refined the setup and were able to produce more precise measurements of the thermal conductivity. The measured conductivities were between 4 and 8 W/m-K, much lower than the reported conductivity of the lead bismuth at about 10 W/m-K. The difference must be attributed to thermal resistances at the eutectic-aluminum and eutectic-steel interfaces. This, and the inherent difficulty of measuring the interface temperature due to the finite width of the thermocouples and the existence of sharp thermal gradients makes it difficult to further reduce the uncertainty in the measured conductivity.

Feasibility of Partial ZrO[subscript 2] Coatings on Outer Surface of Annular UO[subscript 2] Pellets to Control Gap Conductance

2019-04-11T01:22:25Z

Feasibility of Partial ZrO[subscript 2] Coatings on Outer Surface of Annular UO[subscript 2] Pellets to Control Gap Conductance Feinroth, H.; Yuan, Y. The viability of depositing a thin porous coating of zirconia on the outer surface of an annular UO[subscript 2] pellet is investigated experimentally. Such a layer has been proposed to buffer the contact between the fuel and cladding, thus maintaining an appropriate balance of heat transfer from the pellet to the outer and inner cladding. MIT and Gamma Engineering commissioned laboratory studies of the feasibility of depositing a controlled thickness of porous zirconia on an oxide surface. Experiments were conducted at the Thermal Spray Laboratory at SUNY-Stonybrook to produce a thin layer of Yttria Stablized Zirconia (YSZ) on alumina wafers. The experiments concluded that it is possible to use plasma spray guns to produce 50% porous layers of 15–30 micrometer thickness. Measurements were conducted at the Vitreous State Laboratory of the Catholic University of America to determine the thermal conductance of aluminazircaloy and alumina-zircaloy-YSZ sandwiches as a function of applied pressure. A relation is developed to predict the conductance at such surfaces.

Wire Wrapped Hexagonal Pin Arrays for Hydride Fueled PWRs

2019-04-10T18:05:20Z

Wire Wrapped Hexagonal Pin Arrays for Hydride Fueled PWRs Diller, Peter; Todreas, Neil E. This work contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). Core design is accomplished for both hydride and oxide-fueled cores over a range of geometries via steadystate and transient thermal hydraulic analyses, which yield the maximum power, and fuel performance and neutronics studies, which provide the achievable discharge burnup. The final optimization integrates the outputs from these separate studies into an economics model to identify geometries offering the lowest cost of electricity, and provide a fair basis for comparing the performance of hydride and oxide fuels. This work focuses on the steady-state and transient thermal hydraulic as well as economic analyses for PWR cores utilizing wire wraps in a hexagonal array with UZrH[subscript 1.6] and UO[subscript 2]. It was previously verified that square and hexagonal arrays with matching rod diameters and H/HM ratio have the same thermal hydraulic performance. In this work, this equivalence is extended to hexagonal wire wrap arrays, and verified by comparing the thermal hydraulic performance of a single hexagonal wire wrap core with its equivalent square array core with grid spacers. A separate neutronics equivalence is developed, based on the assumption that arrays with matching rod diameters and H/HM ratios will have identical neutronic performance. Steady-state design limits were separated into hard limits, which must be satisfied, or soft limits, which serve to keep the design reasonable. Design limits were placed on the pressure drop, critical heat flux (CHF), vibrations, and fuel and cladding temperature. Vibrations limits on the wire wrap assemblies were imposed for flow induced vibrations (FIV) and thermal hydraulic vibrations (THV). An analysis of the fretting wear of wire wraps indicated that wire wraps outperformed the analogous fretting wear analysis for grid spacers. A CHF study found wire wraps to outperform grid spacers. LOCA and overpower transient analyses were performed for wire wraps. The overpower transient was analyzed over a range of geometries, and found to be more limiting than the steady-state analysis. The LOCA was analyzed for various powers at the reference geometry and another geometry of interest. Through all of these analyses, it was determined that the thermal hydraulic performance of UZrH1.6 and UO2 are very similar. The optimal wire wrap designs were found to have significantly higher maximum powers than the reference core, allowing for uprates up to ~54%. This is due to improved vibrations, pressure drop, and CHF. The steady-state and transient analyses were combined with fuel performance and neutronic studies into an economics model that determines the optimal geometries for incorporation into existing PWR’s. The model also provides a basis for comparing the performance of UZrH[subscript 1.6] to UO[subscript 2] for a range of core geometries. Results presented herein show cost savings for oxide fuel with wire wraps over grid spacers of at least 0.8 mils/kWe-hr, or 4%, due to power increases predicted by the thermal hydraulic analyses. Wire wrap UZrH[subscript 1.6] has a COE savings over UO[subscript 2] of 0.7 mils/kWe-hr, or 4%. Due to the large power uprates possible, cost savings of up to 10.9 mils/kWe-hr, or 40%, can be achieved, with a UZrH[subscript 1.6] wire wrap uprate instead of building a new core.

Thermal Hydraulic and Economic Analysis of Grid-Supported Hydride and Oxide Fueled PWRs

2026-01-15T20:37:11Z

Thermal Hydraulic and Economic Analysis of Grid-Supported Hydride and Oxide Fueled PWRs Shuffler, C.; Trant, J.; Todreas, Neil E.; Romano, A. This report advances the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This effort involves implementing an appropriate methodology for design and optimization of hydride and oxide fueled cores. Core design is accomplished for a range of geometries via steady-state and transient thermal hydraulic analyses, which yield the maximum power, and fuel performance and neutronics studies, which provide the achievable discharge burnup. The final optimization integrates the outputs from these separate studies into an economics model to identify geometries offering the lowest cost of electricity, and provide a fair basis for comparing the performance of hydride and oxide fuels. This report builds on the considerable work which has already been accomplished on the project. More specifically, it focuses on the steady-state and transient thermal hydraulic and economic analyses for pressurized water reactor (PWR) cores utilizing UZrH[subscript 1.6] and UO[subscript 2]. A previous MIT study established the steady-state thermal hydraulic design methodology for determining maximum power from square array PWR core designs. In lieu of a detailed vibrations analysis, the steady-state thermal hydraulic analysis imposed a single design limit on the axial flow velocity. The wide range of core geometries considered and the large power increases reported by the study makes it prudent to refine this single limit approach. This work accomplishes this by developing and incorporating additional design limits into the thermal hydraulic analysis to prevent excessive rod vibration and wear. The vibrations and wear mechanisms considered are: vortex-induced vibration, fluid-elastic instability, turbulence-induced vibration, fretting wear, and sliding wear. Further, the transients investigated are an overpower transient, a large break loss of coolant accident (LBLOCA), and a complete loss of flow accident. In parallel with this work, students at UC Berkeley and MIT have undertaken the neutronics and fuel performance studies. With these results, and the output from the steady-state thermal hydraulic analysis with vibrations and wear imposed design limits, as well as transient thermal hydraulic analysis, an economics model is employed to determine the optimal geometries for incorporation into existing PWRs. The model also provides a basis for comparing the performance of UZrH[subscript 1.6] to UO[subscript 2] for a range of core geometries. Though this analysis focuses only on these fuels, the methodology can easily be extended to additional hydride and oxide fuel types, and will be in the future. Results presented herein do not show significant cost savings for UZrH[subscript 1.6], primarily because the power and energy generation per core loading for both fuels with square arrays supported by grid spacers are similar. Furthermore, the most economic geometries typically do not occur where power increases are reported by the thermal hydraulics. However, preliminary analysis with the lower pressure drop characteristics of wire wraps compared to grids suggest that hexagonal array cores with wire wraps will allow tight ( P over D ≺ 1.25) packing which yield significantly better power performance. This should allow hydride fuel to outperform oxide fuel since this tight core region is not accessible to oxide cores, because of neutronic constraints.

Innovative Fuel Designs for High Power Density Pressurized Water Reactor

2019-04-11T01:22:24Z

Innovative Fuel Designs for High Power Density Pressurized Water Reactor Feng, D.; Kazimi, Mujid S.; Hejzlar, Pavel One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of the pressurized water reactor (PWR) and its related design limits. Those features include: large fuel surface to volume ratio, small fuel thickness, large fuel rod stiffness, low core pressure drop and an open fuel lattice design. Three types of fuel designs are evaluated from the thermal-hydraulic point of view: conventional solid cylindrical fuel rods, internally and externally cooled annular fuel rods, and spiral cross-geometry fuel rods, with the major effort allocated to analyzing the annular fuel. Limits of acceptable power density in solid cylindrical fuel rods are obtained by examining the effects of changing the core operation parameters, fuel rod diameter and rod array size. It is shown that the solid cylindrical geometry does not meet all the desired features for high power density well, and its potential for achieving high power density is limited to 20% of current PWR power density, unless the vibration problems at the coolant higher velocity are overcome. The internally and externally cooled annular fuel potential for achieving high power density is explored, using a whole core model. The best size of fuel rods that fits in the reference assembly dimension is a 13x13 array, since the hot red will have a balanced MDNBR in the inner and outer channels. With proportional increase in coolant flow rate, this annular fuel can increase PWR power density by 50% with the same DNBR margin, while reducing by 1000 ºC the peak fuel temperature. Five issues involving manufacturing tolerances, oxide growth on rod surfaces, inner and outer gap conductances asymmetry, MDNBR sensitivity to changes in core operation parameter and resistance to instabilities were also evaluated. It is found that the main uncertainty for this design is associated with the heat split between the inner and outer channels due to differences in the thermal resistances in the two fuel-clad gaps. Annular fuel is found to be resistant to flow instabilities, such as Ledinegg instability and density wave oscillation due to high system pressure and one-phase flow along most of the hot channel length. Similar power density uprate is found possible for annular fuel in a hexagonal lattice. Large break loss of coolant accident (LBLOCA) for the reference Westinghouse 4-loop PWR utilizing annular fuel at 150% power is analyzed using RELAP, under conservative conditions. The blowdown peak cladding temperature (PCT) is found to be lower because of the low operating fuel temperature, but the flow rate from the safety injection system needs to be increased by 50% to remove the 50% higher decay heat. Loss of flow analysis also showed better performance of the annular fuel because of its low stored energy. The fuel design that best meets the desired thermal and mechanical features is the spiral 3 cross-geometry rods. The dimensions of this type of fuel that can be applied in the reference core were defined. Thermal-hydraulic whole-core evaluations were conducted with cylindrical fuel rod simplification, and critical heat flux modification based on the heat flux lateral non-uniformity in the cross geometry. This geometry was found to have the potential to increase PWR power density by 50%. However, there are major uncertainties in the feasibility and costs of manufacturing this fuel.

NEUTRONIC AND THERMAL HYDRAULIC DESIGNS OF ANNULAR FUEL FOR HIGH POWER DENSITY BWRS

2019-04-12T20:31:11Z

NEUTRONIC AND THERMAL HYDRAULIC DESIGNS OF ANNULAR FUEL FOR HIGH POWER DENSITY BWRS Morra, P.; Xu, Z.; Hejzlar, Pavel; Saha, P.; Kazimi, Mujid S. As a promising new fuel for high power density light water reactors, the feasibility of using annular fuel for BWR services is explored from both thermal hydraulic and neutronic points of view. Keeping the bundle size similar to conventional GE 8×8 solid fuel bundles, annular fuel bundles of 5×5 and 6×6 lattices, that have increased thermal output potential, are explored. The large annular fuel rods allow both external and internal cooling, which increases the fuel surface to volume ratio and significantly reduces the fuel temperature. A methodology is developed and VIPRE code calculations are performed to select the best annular fuel bundle design on the basis of its Critical Power Ratio (CPR) performance. Within the limits applied to the reference solid fuel, the CPR margin in the 5x5 and 6x6 annular fuel bundles is traded for an increase in power density. It is found that the power density increase with annular fuel in BWRs may be limited to 23%. This is smaller than possible for PWRs due to the difference in the mechanisms that control the critical thermal conditions of the two reactors. The neutronic aspects of annular fuel in BWRs, including the reactivity history, power distribution, and burnup characteristics, are investigated. Results are compared to the conventional BWR solid fuel bundle for the same power density and total energy. In general switching to annular fuel implies smaller neutronic differences than in the PWR case. The local peaking factors are found to be similar or slightly better than those of the solid fuel bundle. To maintain the same fuel cycle length, the burnup needs to be increased even for the same bundle energy output, due to reduced fuel loading. These results are based on two dimensional bundle models using the CASMO code, whose validity has been checked using MCNP models. In summary, the annular fuel could be a profitable alternative to the solid fuel due to neutronic and thermal advantages. Further study of the trends observed in this report are needed to increase their certainty.

Alternative Fuel Cycle Strategies For Nuclear Power Generation In The 21st Century

2019-04-09T15:55:26Z

Alternative Fuel Cycle Strategies For Nuclear Power Generation In The 21st Century Boscher, T.; Hejzlar, Pavel; Kazimi, Mujid S.; Todreas, Neil E.; Romano, A. The deployment of fuel recycling through either CONFU (COmbined Non-Fertile and UO[subscript 2] fuel) thermal water-cooled reactors (LWRs) or fast ABR (Actinide Burner Reactor) reactors is compared to the Once-Through LWR reactor system in terms of accumulation of actinides over the next 100 years under the assumption of a growing worldwide demand for nuclear energy. It is assumed that the growth rate is about 2.1% per year up to 2053, with alternative scenarios after that date. The transuranics (TRU) stored in temporary repositories, the TRU sent to permanent repositories, the system cost and a vulnerability index toward proliferation are calculated by the CAFCA code and taken as key figures of merit. Deployment of the ABRs is assumed to occur later (2028) than the CONFU LWRs (2015), whose technology requires less extensive additional R&D. Through 2050 the CONFU strategy performs better than the ABR strategy. The CONFU LWRs in our model yield zero net TRU incineration while the ABRs have a net consumption of TRU. Compared to the Once-Through strategy, by 2050 the CONFU (respectively ABR) strategy reduces by about 35% (respectively 9%) the total inventory of TRU in the system. This reduction corresponds to the TRU production being avoided by CONFU LWRs or being incinerated in ABRs compared to the TRU produced in the traditional LWRs used in the Once-Through strategy. By 2100, the CONFU and the ABR strategies would have reduced the worldwide TRU inventory by 62% compared to the Once- Through case with the CONFU strategy incinerating more TRU than in the ABR strategy. The three strategies are also discussed with regard to uranium ore availability, repository need, and processing plants need. It is interesting to note that with either recycling strategies the total capacity for separation of spent UO2 constituents need 10 to 12 separation plants with a capacity of 2000 MTHM/year. Furthermore, only one TRU recycling plant from fertile-free fuel would be needed at a capacity of 250 MTHM/year up to 2050. The economic analysis shows that both closed fuel cycles are more expensive than the reference Once-Through scheme. The total cost of electricity production is expected to be 3 mills/kWhe, or about 10%, larger than the Once-Through cycle case, if the spent fuel separation is paid off by the electricity sales from the resulting fuel. The timing of collection of fuel cycle costs significantly affects the cost of electricity. Paying for fuel separation by the sales of the electricity producing the spent fuel to be reprocessed later has a smaller effect on the cost of electricity in the advanced fuel cycles (between 1 and 2 mills/kWhe or between 3 and 6%) compared to the cost of electricity in the Once-Through strategy. From a policy point of view, an index of vulnerability toward proliferation is defined and gives an advantage to the advanced fuel cycles. The large amount of heavy metal in the repository and the long life time of this repository penalize the Once-Through strategy. However the results are sensitive to the accessibility factor assigned to the repository which is, as all accessibility factors, a subjective value that is not precisely defined. Moreover, worldwide cooperation to implement the two advanced strategies and the challenges this implementation could face are discussed. The use of a single behaviour mode throughout the world implies an unlikely perfect cooperation between countries that do not have the same capabilities or incentives to choose among the advanced fuel cycle strategies. Revision 1

Combining Thorium with Burnable Poison for Reactivity Control of a Very Long Cycle BWR

2019-04-12T20:31:13Z

Combining Thorium with Burnable Poison for Reactivity Control of a Very Long Cycle BWR Inoue, Y.; Pilat, Edward E.; Xu, Z.; Kazimi, Mujid S. The effect of utilizing thorium together with gadolinium, erbium, or boron burnable absorber in BWR fuel assemblies for very long cycle is investigated. Nuclear characteristics such as reactivity and power distributions are evaluated using CASMO-4. Without thorium, the results show that gadolinium enriched in Gd-157 has the lowest reactivity swing throughout the cycle. However, the local peaking factor (LPF) in the assembly at beginning-of-life (BOL) is high. The erbium case shows more reactivity swing but the LPF is lowest of all three cases. B4C case has the highest reactivity at BOL which would have to be suppressed by control rods. The most important advantage of B4C over others is the saving of uranium inventory needed to achieve the target exposure of 15 effective full power years (EFPY). Further analysis for transient conditions must be performed to ensure meeting all transient limits. Use of thorium in place of some burnable poison makes it possible to save some uranium enrichment while achieving equivalent discharge burnup to the case without thorium, but only by about 1 %. The benefit is small because almost the same amount of burnable poison is always required for suppressing excess reactivity throughout the cycle. Since Th-232 functions more like U-238 than burnable poison, this limits the allowed thorium to extend discharge burnup. Since all fuel assembly designs in this study have the same target exposure of 15EFPY, the economic performance of each design can be compared based on the amount and enrichment of both uranium and burnable absorbers for each fuel design. The B4C-Al fuel is most economical in overall cost even with large uncertainties. The overall cost of gadolinium and erbium cases are concluded to be about the same when large uncertainties are considered.

Optimization of the LWR Nuclear Fuel Cycle for Minimum Waste Production

2019-04-12T20:31:12Z

Optimization of the LWR Nuclear Fuel Cycle for Minimum Waste Production Shwageraus, Eugene; Hejzlar, Pavel; Kazimi, Mujid S. The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to be isolated from the environment for thousands of years. In addition, plutonium and other actinides, after the decay of fission products, could become targets for weapon proliferators. Furthermore, only a small fraction of the energy potential in the fuel is being used. All these concerns can be addressed if a closed fuel cycle strategy is considered offering the possibility for partitioning and transmutation of long lived radioactive waste, enhanced proliferation resistance, and improved utilization of natural resources. It is generally believed that dedicated advanced reactor systems have to be designed in order to perform the task of nuclear waste transmutation effectively. The development and deployment of such innovative systems is technically and economically challenging. In this work, a possibility of constraining the generation of long lived radioactive waste through multi-recycling of Trans-uranic actinides (TRU) in existing Light Water Reactors (LWR has been studied. Thorium based and fertile free fuels (FFF) were analyzed as the most attractive candidates for TRU burning in LWRs. Although both fuel types can destroy TRU at comparable rates (about 1150 kg/GWe-Year in FFF and up to 900 kg/GWe-Year in Th) and achieve comparable fractional TRU burnup (close to 50a/o), the Th fuel requires significantly higher neutron moderation than practically feasible in a typical LWR lattice to achieve such performance. On the other hand, the FFF exhibits nearly optimal TRU destruction performance in a typical LWR fuel lattice geometry. Increased TRU presence in LWR core leads to neutron spectrum hardening, which results in reduced control materials reactivity worth. The magnitude of this reduction is directly related to the amount of TRU in the core. A potential for positive void reactivity feedback limits the maximum TRU loading. Th and conventional mixed oxide (MOX) fuels require higher than FFF TRU loading to sustain a standard 18 fuel cycle length due to neutron captures in Th232 and U238 respectively. Therefore, TRU containing Th and U cores have lower control materials worth and greater potential for a positive void coefficient than FFF core. However, the significantly reduced fuel Doppler coefficient of the fully FFF loaded core and the lower delayed neutron fraction lead to questions about the FFF performance in reactivity initiated accidents. The Combined Non-Fertile and UO[subscript 2] (CONFU) assembly concept is proposed for multirecycling of TRU in existing PWRs. The assembly assumes a heterogeneous structure where about 20% of the UO[subscript 2] fuel pins on the assembly periphery are replaced with FFF pins hosting TRU generated in the previous cycle. The possibility of achieving zero TRU net is demonstrated. The concept takes advantage of superior TRU destruction performance in FFF allowing minimization of TRU inventory. At the same time, the core physics is still dominated by UO[subscript 2] fuel allowing maintenance of core safety and control characteristics comparable to all-UO[subscript 2]. A comprehensive neutronic and thermal hydraulic analysis as well as numerical simulation of reactivity initiated accidents demonstrated the feasibility of TRU containing LWR core designs of various heterogeneous geometries. The power peaking and reactivity coefficients for the TRU containing heterogeneous cores are comparable to those of conventional UO[subscript 2] cores. Three to five TRU recycles are required to achieve an equilibrium fuel cycle length and TRU generation and destruction balance. A majority of TRU nuclides reach their equilibrium concentration levels in less than 20 recycles. The exceptions are Cm246, Cm248, and Cf252. Accumulation of these isotopes is highly undesirable with regards to TRU fuel fabrication and handling because they are very strong sources of spontaneous fission (SF) neutrons. Allowing longer cooling times of the spent fuel before reprocessing can drastically reduce the SF neutron radiation problem due to decay of Cm244 and Cf252 isotopes with particularly high SF source. Up to 10 TRU recycles are likely to be feasible if 20 years cooling time between recycles is adopted. Multi-recycling of TRU in the CONFU assembly reduces the relative fraction of fissile isotopes in the TRU vector from about 60% in the initial spent UO[subscript 2] to about 25% at equilibrium. As a result, the fuel cycle length is reduced by about 30%. An increase in the enrichment of UO[subscript 2] pins from 4.2 to at least 5% is required to compensate for the TRU isotopics degradation. The environmental impact of the sustainable CONFU assembly based fuel cycle is limited by the efficiency of TRU recovery in spent fuel reprocessing. TRU losses of 0.1% from the CONFU fuel reprocessing ensure the CONFU fuel cycle radiotoxicity reduction to the level of corresponding amount of original natural uranium ore within 1000 years. The cost of the sustainable CONFU based fuel cycle is about 60% higher than that of the once through UO[subscript 2] fuel cycle, whereas the difference in total cost of electricity between the two cycles is only 8%. The higher fuel cycle cost is a result of higher uranium enrichment in a CONFU assembly required to compensate for the degradation of TRU isotopics and cost of reprocessing. The major expense in the sustainable CONFU fuel cycle is associated with the reprocessing of UO[subscript 2] fuel. Although reprocessing and fabrication of FFF pins have relatively high unit costs, their contribution to the fuel cycle cost is marginal as a result of the small TRU throughput. Reductions in the unit costs of UO[subscript 2] reprocessing and FFF fabrication by a factor of two would result in comparable fuel cycle costs for the CONFU and conventional once through cycle. An increase in natural uranium prices and waste disposal fees will also make the closed fuel cycle more economically attractive. Although, the cost of the CONFU sustainable fuel cycle is comparable to that of a closed cycle using a critical fast actinide burning reactor (ABR), the main advantage of the CONFU is the possibility of fast deployment, since it does not require as extensive development and demonstration as needed for fast reactors. The cost of the CONFU fuel cycle is projected to be considerably lower than that of a cycle with an accelerator driven fast burner system.

MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL

2019-04-12T20:31:11Z

MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL Long, Y.; Kazimi, Mujid S.; Ballinger, Ronald G.; Meyer, J. E. Fuel performance models have been developed to assess the performance of ThO[subscript 2]-UO[subscript 2] fuels that can be operated to a high burnup up to 80-100MWd/kgHM in current and future Light Water Reactors (LWRs). Among the various issues raised in high burnup fuel applications, the pellet rim effect, fission gas release (FGR), and response to reactivity initiated accidents (RIA) were of special interest in this work. These phenomena were modeled by modifying the NRC licensing codes FRAPCON-3 for normal operation and FRAP-T6 for transient conditions. These models were verified and compared to the results of previous thorium fuel studies and high burnup uranium fuel evaluations. The buildup of plutonium in the outer rim of LWR UO[subscript 2] pellets has been observed to create a region of high fuel burnup, fission gas buildup and high porosity at the fuel rim. The power distribution of the thoria and urania fuel was calculated using a neutronics code MOCUP. Due to the lower build-up of Pu-239 (less U-238 in ThO[subscript 2]-UO[subscript 2] fuel) and flatter distribution of U-233 (less resonance capture in Th-232), thoria fuel experiences a much flatter power distribution and thus has a less severe rim effect than UO[subscript 2] fuel. To model this effect properly, a new model, THUPS (Thoria-Urania Power Shape), was developed, benchmarked with MOCUP and adapted into FRAPCON-3. Additionally a porosity model for the rim region was introduced at high burnup to account for the larger fuel swelling and degradation of the thermal conductivity. The mechanisms of fission gas release in ThO[subscript 2]-UO[subscript 2] fuel have been found similar to those of UO[subscript 2] fuel. Therefore, the general formulations of the existing fission gas release models in FRAPCON-3 were retained. However, the gas diffusion coefficient in thoria was adjusted to a lower level to account for the smaller observed gas release fraction in the thoria-based fuel. To model accelerated fission gas release at high burnup properly, a new athermal fission gas release model was developed. Other modifications include the thoria fuel properties, fission gas production rate, and the corrosion model to treat advanced cladding materials. The modified version of FRAPCON-3 was calibrated using the measured fission gas release data from the Light Water Breeder Reactor (LWBR) program. Using the new model to calculate the gas release in typical PWR hot pins gives data that indicate that the ThO[subscript 2]-UO[subscript 2] fuel will have considerably lower fission gas release beyond a burnup of 50 MWd/kgHM. Investigation of the fuel response to an RIA included: (1) reviewing industry simulation tests to understand the mechanisms involved, (2) modifying FRAP-T6 code to simulate the RIA tests and investigate the key contributors to fuel failure (thermal expansion, gaseous swelling, cladding burst stress), and (3) assessing thoria and urania performance during RIA event in typical LWR situations. ThO[subscript 2]-UO[subscript 2] fuel has been found to have better performance than UO[subscript 2] fuel under RIA event conditions due to its lower thermal expansion and a flatter power distribution in the fuel pellet (less power and less fission gas in the rim region). Overall, thoria has been found to have better performance than urania in both normal and off-normal conditions. However, calculations using the modified FRAPCON-3 showed that the internal pressure and cladding corrosion at the required high burnup of 80- 100MWd/kgHM are not acceptable with the current fuel design. Therefore, advanced fuel designs (including larger gas plenum, larger fuel grains, advanced cladding materials), and carefully designed operating strategy (i.e. decreasing power history) were assessed and the results showed that the targeted high burnup can be achieved. Further investigation of burnup issues is needed, such as the distribution of hydrogen in the cladding for heterogeneous fuels, and response of high pressure fuel pins to a loss of coolant accident, in order to assure satisfactory high burnup behavior.

High Burnup Fuels for Advanced Nuclear Reactors

2019-04-12T20:31:18Z

High Burnup Fuels for Advanced Nuclear Reactors Oggianu, S. M.; Christensen, Holly Colleen No; Kazimi, Mujid S. The goal of this work is to select the best candidate fuel materials to deliver high burnup in advanced light water reactors. Uranium and thorium based fuels are considered. These fuel materials must be able to withstand nearly double the burnup of current LWRs in high irradiation fields. Reactor economics, safety, proliferation resistance, fuel reprocessing and spent fuel disposal are the most important factors to be addressed. High burnup will provide the opportunity for uninterrupted operation over long periods of time, reduction of spent fuel volume and improvement of proliferation resistance. Thus, effective power cycle maintenance and fuel management and reduced fuel storage needs will lead to more economic operation. Several uranium and thorium fuel forms are analyzed to predict their capability to withstand high burnups. Their fuel cycle cost is also considered. To compare the fuel options, simple indices characterizing the behavior of the materials at high burnup are defined. Indices for the thermal stress capability, stored energy and margin for melting are derived from non-dimensional analyses. To evaluate the fuel pin lifetime, a simplified fuel performance analysis code, FUELSIM (FUEL SIMulation code) was developed. The code utilizes the VENSIM simulation system, which allows for great flexibility in the change of governing relations, permits sensitivity analysis, and facilitates graphical outputs. Based on the sensitivity analysis by FUELSIM, dominant parameters are identified and a simplified expression is developed for predicting the increase in the pin internal pressure with burnup. For each material, we obtain a maximum attainable burnup at a given smear density. Cladding strain, internal pressure and fuel melting (or phase-change) temperature are the limiting factors used to obtain these burnups. From neutronic reactivity considerations, the needed [superscript 235]U enrichment can be specified. Thus, the fuel cycle cost for each material and smear density can be estimated. Metals, oxides, carbides and nitrides of uranium and thorium were examined. Although the results show that UN provides the highest potential for attaining high burnup and economic application in once-through cycles, it has limited compatibility with water. UO[subscript 2], at 90- 95% smear density, continues being the most feasible option as a nuclear material. Also, ThO[subscript 2]/UO[subscript 2] seems to offer as good or better potential performance and economics as UO[subscript 2]. However, more reliable data on the irradiation behavior of the different materials is needed before a definitive conclusion can be drawn. Also important in the evaluation of thorium/uranium cycles are attributes that were not considered here. These include the reduction in spent fuel volume, the improvement in proliferation resistance, possible power uprates and allowance for a higher peaking factor that may be possible by taking advantage of the increased margin that results from using fuels with lower stored energies.

A Systematic Study of Moderation Effects On Neutronic Performance of UO[subscript 2] Fueled Lattices

2019-04-11T01:22:23Z

A Systematic Study of Moderation Effects On Neutronic Performance of UO[subscript 2] Fueled Lattices Xu, Z.; Driscoll, Michael J.; Kazimi, Mujid S. This report addresses the physics of reactor cores that can be operated for 10 to 15 years without refueling — inspired by the objective of enhanced nuclear fuel cycle performance with regard to economics and resistance to weapon proliferation. Proliferation resistance is a primary consideration in this design. The long life operation reduces the routine access to the internals of the reactor vessel, therefore reducing the possibility for clandestine production of nuclear weapons. Additionally, reduction of reactor shutdown time can result in improved safety and economics. As a first step, the most promising fuel lattice characteristics to achieve long life from a physics point of view are studied. These studies also define the design tradeoffs involved in conceptualizing such cores. Moderation effects on UO[subscript 2] fueled lattices are analyzed systematically using state-of-the-art computer codes (CASMO-4 and MOCUP). The standard 4-loop Westinghouse pressurized water reactor (PWR) is taken as our reference core and single unit cell analysis is employed. To change the moderator-to-fuel ratio, which is characterized by the hydrogen-to-heavy-metal (H/HM) atom number ratio, various methods are adapted including varying water density, fuel density, fuel rod diameter, and fuel rod pitch. Higher burnup potential as well as longer core endurance (burnup times heavy metal mass) would be desirable. For a given initial enrichment, the results show that higher reactivity-limited burnup is achievable by either a more wet lattice or much drier lattice than normal. However, epithermal lattices are distinctly inferior performers. In terms of longer endurance, current PWR lattice parameters are about the optimum. Higher burnup and endurance can be achieved with higher initial enrichment. Characteristics of the spent fuel from high burnup UO[subscript 2] fueled lattices have been examined. The variation of isotopic mix and quantity of plutonium with moderator-to-fuel ratio for UO[subscript 2] fueled lattices has been studied to clarify the impact on its proliferation resistance. And Np production as a function of H/HM has been computed as a measure of long-term radiological hazard for high level waste disposal. It is shown that Np is mildly affected by the H/HM ratio and the current PWR lattice is close to optimum configuration. However, high burnup is significantly beneficial as a way to make the plutonium isotopic mix less attractive as a weapon material.

ON THE USE OF THORIUM IN LIGHT WATER REACTORS

2026-01-15T20:36:27Z

ON THE USE OF THORIUM IN LIGHT WATER REACTORS Kazimi, Mujid S.; Czerwinski, Kenneth R.; Driscoll, Michael J.; Hejzla, P.; Meyer, J. E. The advantages and disadvantages of the use of thorium bearing fuel in light water reactors have been examined several times from the beginning of the nuclear energy era until the late seventies. The recent motivation for re-examining the use of thorium in light water reactors' once-through fuel cycle is enhancing the cycle proliferationresistance due to reduced plutonium production. Additionally, economic benefits from reducing the initial enrichment needs of high burnup fuels may be obtained. Similarly, it may be possible to rely on the higher melting point and higher thermal conductivity of ThO[subscript 2] to enhance the safety margin of the core. Thorium dioxide is the highest stable oxide form of thorium, which may further improve the spent fuel repository performance. The information obtained in previous studies is reviewed to assess its suitability for application to the current fuel cycle conditions. It is concluded that the thorium fuel experience of the past is insufficient to make a judgement on the feasibility and performance of the thorium bearing fuels in the reactors operating under current conditions. The needs for new research and development efforts in the areas of neutronics, fuel behavior, safety and waste performance are outlined.

FUEL PERFORMANCE ANALYSIS OF EXTENDED OPERATING CYCLES IN EXISTING LWRs

2019-04-12T20:30:54Z

FUEL PERFORMANCE ANALYSIS OF EXTENDED OPERATING CYCLES IN EXISTING LWRs Handwerk, C. S.; Meyer, J. E.; Todreas, Neil E. An integral part of a technical analysis of a core design, fuel performance is especially important for extended operating cycles since the consequences of failed fuel are greater for this operating strategy than for current practice. This stems mainly from the fact that extended cycles offer a unique benefit by running longer without interruption; poor fuel performance, i.e. failed fuel, would degrade this benefit. The issues in this research are assessed only at the steady-state level, as a foundation for the consideration of Anticipated Operational Occurrences (AOOs) and transient conditions, which are certain to present greater challenges to nuclear fuel performance due to their more severe conditions. Even at this preliminary steady state level, extended cycle operation is found to exacerbate several fuel performance issues, resulting mainly from the fact that some fuel in an extended operating cycle is operated at higher powers over part of the core life and does not have the benefit of shuffling. In order to accurately quantify the fuel performance effects of extended cycle operation, a pseudo or "envelope" pin is created, which represents the operating characteristics of the highest power fuel rod in the core at a given pin burnup interval. This envelope pin was created for both extended cycle and current practice, so that extended cycle results could be compared to both existing licensing limits and current practice. While this approach is somewhat conservative, it is the simplest way to evaluate fuel performance in an extended cycle core where the location of the limiting fuel rod changes often and operates at higher powers for prolonged periods of time. The US Nuclear Regulatory Commission's Standard Review Plan's Sections 4.2 and 4.4 are used as the basis for the criteria that should be evaluated in this report, since these are the relevant sections of the document that prescribes the licensing limits and criteria for nuclear fuel design. From this document, ten steady state fuel performance issues are identified: (1) stress and strain, (2) fatigue cycling, (3) fretting, (4) waterside corrosion, (5) axial growth and rod bowing, (6) rod internal pressure, (7) primary hydriding, (8) cladding collapse, (9) cladding overheating, and (10) fuel centerline melt. Of these ten issues, (7) and (8) were found to be not uniquely affected by extended cycle operation. While (9) and (10) are found to not be concerns for extended cycle operation, the higher powers at which extended operating cycles can operate degrade some of the margin for transient effects, which is more of a significant concern for (9). (1) and (5) are predicted to be worse for both BWRs and PWRs when compared to current practice, and (4) and (6) are projected to present greater challenges for PWRs. Additionally, (2) is the only factor that is predicted to actually be better for extended cycle operation in both the BWR and PWR while (4) was predicted to have less of an effect in BWRs, given the comparable operating powers and shorter in-core residence time for the extended cycle case. The effects of the proposed new operating strategy on (3) were uncertain. Of all ten issues, (5) seemed to be the most problematic, as no solution was readily available. Solutions to other issues included improved assembly grid design (3), water chemistry control (4), annular fuel pellets (6), and, potentially, increasing the number of fuel rods per assembly (1,4,6,10).