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Performance of Trasuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Interim Report, Including Void Reactivity Evaluation

TL;DR: In this paper, the physics attributes of transuranice-only Fully-Ceramic Micro-encapsulated (FCM) fuel in an LWR lattice were evaluated using the DRAGON-4 code.
Abstract: The current focus of the Deep Burn Project is on once-through burning of transuranice (TRU) in light water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles would be pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell calculations have been performed using the DRAGON-4 code in order assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells containing typical UO2 and MOX fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Loading of TRU-only FCM fuel into a pin without significant quantities of uranium challenges the design from the standpoint of several key reactivity parameters, particularly void reactivity, and to some degree, the Doppler coefficient. These unit cells, while providing an indication ofmore » how a whole core of similar fuel would behave, also provide information of how individual pins of TRU-only FCM fuel would influence the reactivity behavior of a heterogeneous assembly. If these FCM fuel pins are included in a heterogeneous assembly with LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance of the TRU-only FCM fuel pins may be preserved. A configuration such as this would be similar to CONFU assemblies analyzed in previous studies. Analogous to the plutonium content limits imposed on MOX fuel, some amount of TRU-only FCM pins in an otherwise-uranium fuel assembly may give acceptable reactivity performance. Assembly calculations will be performed in future work to explore the design options for heterogeneous assemblies of this type and their impact on reactivity coefficients.« less

Summary (4 min read)

1. INTRODUCTION

  • Over the past few years, the Deep Burn project has evaluated the prospect of using high temperature reactors (HTRs) for reducing legacy inventories of transuranic (TRU) isotopes from used light water reactors (LWRs) fuel.
  • This reduction is to be achieved by transmuting the undesirable isotopes, primarily through fissioning or “burning” them.
  • Both pebble bed2 and prismatic designs3 were conceptualized and significant design and analysis were performed on them.
  • At present, the focus of the Deep Burn Project has shifted to a once-through burning of the TRU materials in modified LWRs.
  • Because the TRU-only FCM fuel is meant to contain no significant quantities of uranium, it is likely that neutronically it will most closely resemble the fertile-free IMF.5.

2. OBJECTIVES

  • In order to begin assessing the neutronics characteristics of the TRU-only FCM fuel, unit cell calculations were performed.
  • These unit cell calculations can provide information about the neutronic characteristics of a whole core of similar fuel, and also would provide insight into the influence of these types of cells in heterogeneous assemblies containing UO2 pins as well.
  • That other report uses the PASTA code to assess the fuel performance under steady state conditions (including depletion) as well as under the conditions induced by a simulated LOCA transient.
  • Performance of TRU-Loaded FCM Fuel in LWRs: Interim Report with Void Reactivity Evaluation 2 March 2011.

3. METHODOLOGY

  • Section 0 presents information about the assumptions made regarding the design of TRU-only FCM fuel compacts and the LWR assembly into which said fuel is assumed to be loaded.
  • Section 3.2 provides details of the UO2 and MOX fuels analyzed as reference cases.
  • In Section 3.3, information is given describing the lattice code and the solution methods used.

3.1 TRU-Only FCM Fuel Description

  • The specifications chosen for initial analysis approximate the lattice of the AREVA EPRTM because this reactor is expected to be available in configurations that can accept a core of 100% MOX.
  • Table 3-1 shows the dimensions and densities (i.e., specific mass) of the layers of the TRISO particles specified for these initial calculations.
  • Preliminary fuel performance calculations have also been performed to predict the material integrity of the TRISO particles constituents of the FCM fuel.
  • The TRU-Ox fuel is contains primarily neptunium and plutonium with trace amounts of uranium, the vector of which is derived from once-burned LWR fuel.

Performance of TRU-Loaded FCM Fuel in LWRs Interim Report, Including Void Reactivity Evaluation

  • N. Hfaiedh, A. Santamarina, “Detemination of the Optimized SHEM Mesh for Neutron Transport Calculations,” Proc. Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications, Avignon, France, 12 Sept. – 15 Sept. (2005).
  • Barbrault, P., “A Plutonium-Fueled High-Moderated Pressurized Water Reactor for the Next Century,” Nuclear Science and Engineering, Vol. 122, pp. 240–246, 1996.

3.2 MOX and UO2 Fuel Descriptions

  • Unit cells similar to that of the FCM fuel described above containing conventional LWR fuels are also analyzed.
  • The two selected cases are an enriched UO2 cell and a MOX cell.
  • The MOX fuel is assumed to have a density of Performance of TRU-Loaded FCM Fuel in LWRs: Interim Report with Void Reactivity Evaluation 4 March 2011 10.0 g/cm3 and 10 w/o plutonium with the balance of the heavy metal being representative of tails uranium having 0.3 w/o 235U.
  • These values represent a prediction of the average of the once-burned LWR plutonium available in France in 2015.
  • The French design is selected for this case because of data and specifications availability in the open literature.

3.3.1 Neutronics Calculations

  • Calculations were performed using DRAGON-4, an open-source lattice transport code developed and maintained by École Polytechnique de Montréal.13.
  • This code contains multiple solution methods and allows for flexible calculation routes and data manipulation.
  • The code also allows for treatment of the double-heterogeneity of the TRISO particles in the fuel directly using the method developed by Hébert.
  • Ultimately under investigation is how much TRU-only FCM fuel can be used in a LWR core.
  • The results are interpreted with the knowledge that if the TRU-only pin is used in conjunction with UO2 pins or assemblies, the overall behavior would be expected to be the composite result of the effects of the TRU-only FCM unit cells and the UO2 unit cells.

3.4 Estimation of Fuel Temperatures

  • The methodology and assumptions used for this are described in the following paragraphs.
  • The temperature in a fuel pin with dispersed coated fuel particles can be estimated from the pin power density and the temperature at the pin outer boundary.
  • The latter linear power would result in a center fuel temperature of 791 K.

4. RESULTS

  • This section presents the interim results of this study, beginning with those from the UO2 and MOX unit cells.
  • These are then used as a basis for comparison with the TRU-only FCM fuel cases to be studied in subsequent sections.

4.1 UO2 and MOX Calculation Results

  • This section reports on the examination of unit cells of UO2 and MOX.
  • For each of these, the behavior of the key reactivity coefficients is presented as burnup proceeds.
  • In core designs containing 100% MOX assemblies, an increase in the moderator/fuel ratio has been proposed to mitigate the poor reactivity feedback and control worth characteristics of the core, for example by replacing fuel rods with water holes or by increasing fuel rod diameters.
  • This is discussed separately in Section 4.4.
  • Figure 4-1. K versus burnup in EFPD and GWd/tonne for UO2 and MOX unit cells.

4.2.1 Variation of Kernel Size

  • The effects of varying the kernel size on the various reactivity coefficients are analyzed.
  • Because the buffer is carbon, which is found also in the TRISO particle layers and in the SiC matrix, its thickness is a-priori expected to have a minimal impact on the neutronics behavior (although a later study may be needed to assess the ultimate validity of this assumption).
  • Given a particular PF and kernel diameter, the buffer thickness can impact the fuel loading.
  • These correspond to effective plutonium densities in the fuel compact of 0.37, 0.51, and 0.65 g/cm3, respectively.

4.2.2 Variation of Packing Fraction

  • This section repeats the analyses from Section 4.2.1, only now the kernel size is fixed at a diameter of 500 m and the PF is varied incrementally from 20% to 48%.
  • The results are analogous to those of the study of the variation of kernel diameter with fixed PF, and so they are not discussed here in detail.
  • Note that the fuel loading appears to be the primary driver for the burnup and reactivity coefficient behavior of the unit cells.
  • K versus burnup for TRU-only FCM fuel with kernel diameter of 500 m.

4.2.3 TRISO Particle Fuel Distribution Effect

  • To verify that the fuel loading is the primary variable in determining the reactivity coefficient behavior of the TRU-only FCM fuel, cases are analyzed where the amount of fuel is held constant while the PF and kernel diameter are simultaneously varied.
  • A case is also included where the entire fuel compact, matrix and TRISO particles, are homogenized.
  • The effective plutonium density on a per-volume-of-compact basis in the selected case is 0.30 g/cm3.
  • Figure 4-18 shows the reactivity versus burnup for these cases.
  • Some variation is observed between the homogenized case and the FCM fuel with microstructures, but very little difference exists between FCM cases of constant fuel and varying kernel diameters.

4.3 Burnable Poison Effects – Er2O3

  • To enhance the safety characteristics of the feedback parameters, previous designers of full-core uraniumfree inert matrix fuel (IMF) cores proposed that resonant absorbers be used as burnable poisons.
  • The plot against EFPD shows that the reactivity-limited burnup is reduced with increasing Er2O3 content.
  • Figure 4-24 and Figure 4-25 show MTC and void coefficient (10% void) versus burnup, respectively.
  • In contrast, the decrease in TRU content that comes from increasing Er2O3 content is not accompanied by a similar behavior of “less negative” coefficients, instead, as seen in the figures the increase of poison content and concomitant decrease of TRU content are accompanied by an increase in magnitude of the negative coefficients.

4.4 Preliminary Analysis of Complete Coolant Voiding

  • In low-enriched UO2 cores, the voided-core reactivity is low due to the absence of the moderation necessary to the efficient thermalization of neutrons.
  • Pu, the 240Pu effect exceeds the 238U effect and the voided reactivity can be greater than unity.
  • The UO2 case, the MOX case, and some of the FCM cases analyzed to examine the effects of various moderator densities on reactivity.
  • For this, a fixed geometric buckling is assumed (2.2 m-2) and the Eigenvalue without leakage (k*) is plotted.
  • No soluble or burnable poisons are used at this point, as the optimal balance of the two has not yet been explored.

4.5 Nuclide Inventories

  • Heavy metal nuclide quantities in g/pin are given in Table 4-1 for BOI and three different burnup levels: 400, 500, and 600 GWd/tonne.
  • This is done on a per-pin basis because the arrangement of TRU-only FCM fuel pins in the assemblies and core are not yet known.
  • Assuming the intermediate discharge burnup of 500 GWd/tonne, the total plutonium consumption is 42.8 g/pin, or 55% of the initial plutonium.
  • The plutonium destruction performance of this fuel is attractive from the standpoint of the fraction, which is destroyed.

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The INL is a U.S. Department of Energy National Laboratory
operated by Battelle Energy Alliance
INL/EXT-11-21343
Performance of
Transuranic-Loaded
Fully Ceramic Micro-
Encapsulated Fuel in
LWRs
Interim Report, Including
Void Reactivity
Evaluation
Michael A. Pope
Brian Boer
A
bderrafi M. Ougouag
Gilles Youinou
March 2011

INL/EXT-11-21343
FCRD-FUEL-2011-000058
Performance of Transuranic-Loaded Fully Ceramic
Micro-Encapsulated Fuel in LWRs
Interim Report, Including Void Reactivity Evaluation
Michael A. Pope
Brian Boer
Abderrafi M. Ougouag
Gilles Youinou
March 2011
Idaho National Laboratory
Fuel Cycle Research & Development
Idaho Falls, Idaho 83415
http://www.inl.gov
Prepared for the
U.S. Department of Energy
Office of Nuclear Energy
Under DOE Idaho Operations Office
Contract DE-AC07-05ID14517

DISCLAIMER
This information was prepared as an account of work sponsored by an
agency of the U.S. Government. Neither the U.S. Government nor any
agency thereof, nor any of their employees, makes any warranty,
expressed or implied, or assumes any legal liability or responsibility for
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Performance of TRU-Loaded FCM Fuel in LWRs Interim Report, Including Void Reactivity Evaluation
iv March 2011
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Citations
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TL;DR: In this paper, the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR) were evaluated.
Abstract: This study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR l...

24 citations

01 May 2005
TL;DR: In this article, the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials are compared with the results for reference UOX fuel bundles.
Abstract: Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is ‘burning’ the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

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TL;DR: In this article, the fundamental scientific principles governing nuclear fission reactors and the methods used in modern nuclear reactor analysis and design are discussed. But they do not cover the fundamental design of modern power reactors.
Abstract: This comprehensive introduction covers the fundamental scientific principles governing nuclear fission reactors and the methods used in modern nuclear reactor analysis and design. The book is divided into four parts: Part 1 presents a relatively elementary and qualitative discussion of the basic concepts of nuclear fission chain reactions, including a brief review of relevant nuclear physics and a survey of modern power reactors. Part 2 develops a simple model of nuclear reactor behavior, assuming that the neutrons sustaining the fission chain reactions diffuse from point to point in the reactor in such a way that their energy and direction of motion can be ignored (one-speed diffusion theory). Part 3 generalizes this model to develop the primary tool of nuclear reactor analysis, multigroup diffusion theory. Part 4 illustrates the methods of nuclear reactor analysis by considering several important applications in reactor engineering, including a brief introduction to reactor core design. The text is written exclusively in S1 units and features numerous problems and exercises, many requiring digital computation.

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TL;DR: In this paper, a compilation of non-irradiated and irradiated properties of SiC are provided and reviewed and analyzed in terms of application to TRISO fuels, specifically in the high-temperature irradiation regime.

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"Performance of Trasuranic-Loaded Fu..." refers background in this paper

  • ...These were shown to mitigate the challenges of uranium-free core loadings, particularly the 100% void reactivity.(26,27,28) As a preliminary assessment of the effects of the more promising burnable poisons on the unit cell neutronic performance, several loadings of Er2O3 are investigated here....

    [...]