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Study on the Technology for Enhancing Element Value by Neutron Transmutation in Light Water Reactors

Rare elements have been the source of industrial competitiveness. However, the production of the elements has been localized in some area in the world. In addition, they can be politically used as strategic materials. Therefore, projects have been actively ongoing to develop technologies for replacing or efficiently using rare elements. An example of the projects is Elements Strategy Initiative in Japan. Neutron transmutation by existing nuclear reactors can be utilized to produce an element by (n, γ) reaction with appropriate irradiation conditions, and has the potential to enlarge the scope of nuclear industry. Our preliminary analysis examined the ratio of the following parameters between target and transmuted elements to identify elements suitable for the transmutation: (1) Crustal abundance [%], (2) Total material requirement (TMR) [t-TMR], (3) Element price [$/g]. TMR is an index showing the total amount of material required to obtain an element. Several elements were identified by checking whether the ratios exceed 100. One of promising transmutations was Dy to Ho. However, the analysis used one-group constants assuming a typical PWR fuel pin and a neutron flux of 1.0 E14 /cm{sup 2}/s to solve burnup equations for three years irradiation time. Therefore, the purpose of this study is to examine transmutation performance for Dy to Ho by considering actual geometry of a PWR fuel assembly.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SCWR Once-Through Calculations for Transmutation and Cross Sections

It is the purpose of this report to document the calculation of (1) the isotopic evolution and of (2) the 1-group cross sections as a function of burnup of the reference Super Critical Water Reactor (SCWR), in a format suitable for the Fuel Cycle Option Campaign Transmutation Data Library. The reference SCWR design was chosen to be that described in [McDonald, 2005]. Super Critical Water Reactors (SCWR) are intended to operate with super-critical water (i.e. H2O at a pressure above 22 MPa and a temperature above 373oC) as a cooling – and possibly also moderating – fluid. The main mission of the SCWR is to generate lower cost electricity, as compared to current standard Light Water Reactors (LWR). Because of the high operating pressure and temperature, SCWR feature a substantially higher thermal conversion efficiency than standard LWR – i.e. about 45% versus 33%, mostly due to an increase in the exit water temperature from ~300oC to ~500oC – potentially resulting in a lower cost of generated electricity. The coolant remains single phase throughout the reactor and the energy conversion system, thus eliminating the need for pressurizers, steam generators, steam separators and dryers, further potentially reducing the reactor construction capital cost. The SCWR concept presented here is based on existing LWR technology and on a large number of existing fossil-fired supercritical boilers. However, it was concluded in [McDonald, 2005], that: “Based on the results of this study, it appears that the reference SCWR design is not feasible.” This conclusion appears based on the strong sensitivity of the design to small deviations in nominal conditions leading to small effects having a potentially large impact on the peak cladding temperature of some fuel rods. “This was considered a major feasibility issue for the SCWR” [McDonald, 2005]. After a description of the reference SCWR design, the Keno V 3-D single assembly model used for this analysis, as well as the calculated results, are presented. Additionally, the follwing information, presented in the appendixes, is intended to provide enough guidance that a researcher repeating the same task in the future should be able to obtain a vector of nuclei and cross sections ready for insertion into the transmutation library without any need for further instructions: (1) Complete TRITON/KENO-V input used for the analysis; (2) Inputs and detailed description of the usage of the OPUS utility, used to postproces and to extract the nuclei concentrations for the transmutation library; (3) Inputs and detailed description of the usage of the XSECLIST utility, used to postproces and to extract the 1-group cross sections for the transmutation library; (4) Details of an ad-hoc utility program developed to sort the nuclei and cross sections for the transmutation library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Detailed calculations of minor actinide transmutation in a fast reactor

The transmutation of minor actinides in a fast reactor is investigated by a new method to investigate the transmutation behavior of individual minor actinides. It is found that Np-237 and Am-241 mainly contributes to the transmutation rate though the transmutation behaviors are very different.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjusted to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principle be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Transuranic Transmutation and Criticality Calculation Sensitivity to Heterogeneous Lattice Effects - 12391

Using Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies has been researched at length and has shown to provide the benefit of transmutation and targets the amount and toxicity of high level waste needed to be managed. Advanced MOX concepts using enriched Uranium Dioxide (UO{sub 2}) are required for multiple recycling of plutonium. The use of MOX and ordinary UO{sub 2} fuel in the same assembly as well as unfueled rods and assembly edge effects contrasts with the unit cell computational assumption of a uniform infinite array of rods. While a deterministic method of calculating the Dancoff factor has traditionally been employed in fuel assembly analysis due to the lighter computational and modeling requirements, this research seeks to determine the validity of the uniform, infinite lattice assumption with respect to Dancoff factor and determine the magnitude of the impact of nonuniform lattice effects on fuel assembly criticality calculations as well as transuranic isotope production and transmutation. This research explored the pin-to-pin interaction in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel rods through the impact of the calculated Dancoff factors from the deterministic method used in SCALE versus the Monte Carlo method used in the code DANCOFF-MC. Using the Monte Carlo method takes into account the non-uniform lattice effects of having neighboring fuel rods with different cross-sectional spectra whereas the Dancoff factor calculated by SCALE assumes a uniform, infinite lattice of one fuel rod type. Differences in eigenvalue calculations as a function of burnup are present between the two methods of Dancoff factor calculation. The percent difference is greatest at low burnup and then becomes smaller throughout the cycle. Differences in the transmutation rate of transuranic isotopes in the MOX fuel are also present between the Dancoff factor calculation methods. The largest difference is in Pu-239, Pu-242, and Am-241 composition whereas U-238, Pu-242, and Pu-238 composition was not changed by taking into account the non-homogenous lattice effects. Heterogeneous lattice effects do change the calculated eigenvalue and transmutation rate in a non-uniform lattice of MOX fuel rods and UO{sub 2} fuel. However, the uncertainty in the ENDF data used by SCALE in these calculations is large enough that the infinite lattice assumption remains valid. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Americium Transmutation Feasibility When Used as Burnable Absorbers - 12392

The use of plutonium in Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies leads to greater americium production which is not addressed in MOX recycling. The transuranic nuclides (TRU) contribute the most to the radiotoxicity of nuclear waste and a reduction of the TRU stockpile would greatly reduce the overall radiotoxicity of what must be managed. Am-241 is a TRU of particular concern because it is the dominant contributor of total radiotoxicity for the first 1000 years in a repository. This research explored the feasibility of transmuting Am-241 by using varying amounts in MOX rods being used in place of burnable absorbers and evaluated with respect to the impact on incineration and transmutation of transuranics in MOX fuel as well as the impact on safety. This research concludes that the addition of americium to a non-uniform fuel assembly is a viable method of transmuting Am-241, holding down excess reactivity in the core while serving as a burnable poison, as well as reducing the radiotoxicity of high level waste that must be managed. The use of Am/MOX hybrid fuel assemblies to transmute americium was researched using multiple computer codes. Am-241 was shown in this study to be able to hold down excess reactivity at the beginning of cycle and shape the power distribution in the core with assemblies of varying americium content loaded in a pattern similar to the traditional use of assemblies with varying amounts of burnable absorbers. The feasibility, safety, and utility of using americium to create an Am/MOX hybrid non-uniform core were also evaluated. The core remained critical to a burnup of 22,000 MWD/MTM. The power coefficient of reactivity as well as the temperature and power defects were sufficiently negative to provide a prompt feedback mechanism in case of a transient and prevent a power excursion, thus ensuring inherent safety and protection of the core. As shown here as well as many other studies, this non-uniform assembly type successfully addresses the concerns of reduced control rod worth within advanced MOX assemblies because the Am/MOX hybrid fuel rods are on the periphery of the assembly. The radiotoxicity of the high level waste that must be managed from a single use of an Am/MOX hybrid batch is reduced for the first 1000 years in comparison to a regular UO{sub 2} batch. However, there is no heat-load advantage without multi-recycling the Am/MOX hybrid assemblies. As shown through numerous studies, the heterogeneous assembly model used as the basis for this research is capable of multiple recycles while still maintaining negative temperature and power coefficients of reactivity. Multi-recycling would provide an even greater reduction to the radiotoxicity of the high level waste as well as provide a heat-load advantage compared to a once through UO{sub 2} fuel cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Symmetry transmutation and anomaly matching

We explore a situation where a global symmetry of the ultraviolet (UV) theory does not act faithfully on the local infrared (IR) degrees of freedom, but instead acts effectively as a higher-form symmetry. We refer to this phenomenon as symmetry transmutation, where the UV symmetry is “transmuted” into a higher-form symmetry in the IR. Notably, unlike emergent (accidental) symmetries, which are approximate, these symmetries are exact. We illustrate the ubiquity of this phenomenon in various continuum and lattice systems and provide examples where the ’t Hooft anomalies of the UV symmetry are matched by those of the new higher-form symmetry in the IR. We also show that in certain phases and for certain energies, the UV baryon-number symmetry of one-flavor QCD is transmuted into a discrete one-form global symmetry. Finally, we compare our symmetry transmutation to the well-known phenomenon of symmetry fractionalization.

Anomalies in Field and String Theories↗

Neutron-transmuted carbon-14 in neutron-irradiated GaN: Compensation of DX-like center

The transmuted-C related luminescence and net carrier concentration are studied by combining photoluminescence, liquid scintillation, and Raman scattering. GaN single crystal films grown by metalorganic-vapor-phase epitaxy are irradiated with fast and thermal neutrons at fluxes of 3.9 × 10{sup 13} cm{sup −2}s{sup −1} and 8.15 × 10{sup 13} cm{sup −2}s{sup −1}, respectively. Irradiation time is 48 hours. The calculated {sup 72}Ge and {sup 14}C concentrations are 1.24 × 10{sup 18} cm{sup −3} and 1.13 × 10{sup 18} cm{sup −3}, respectively. The transmuted {sup 14}C is detected by the liquid scintillation method to survey β-rays emitted in the process of {sup 14}C decays from {sup 14}N. Tritium ({sup 3}H) is also emitted by a (n,t) reaction of {sup 14}N due to the neutron irradiation above 4.5 MeV. Photoluminescence relating to C, DX-like center of Ge and yellow luminescence band are observed in 1000 °C annealed NTD-GaN. The free electron concentration estimated from Raman scattering is 4.97 × 10{sup 17} cm{sup −3}. This value is lower than that from the transmuted Ge concentration, suggesting the compensation due to the transmuted {sup 14}C acceptors.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Fission Product Transmutation in Mixed Radiation Fields

Work under this grant addressed a part of the challenge facing the closure of the nuclear fuel cycle; reducing the radiotoxicity of lived fission products (LLFP). It was based on the possibility that partitioning of isotopes and accelerator-based transmutation on particular LLFP combined with geological disposal may lead to an acceptable societal solution to the problem of management. The feasibility of using photonuclear processes based on the excitation of the giant dipole resonance (GDR) by bremsstrahlung radiation as a cost effective transmutation method was accessed. The nuclear reactions of interest: (γ,xn), (n,γ), (γ,p) can be induced by bremsstrahlung radiation produced by high power electron accelerators. The driver of these processes would be an accelerator that produces a high energy and high power electron beam of ~ 100 MeV. The major advantages of such accelerators for this purpose are that they are essentially available “off the shelf” and potentially would be of reasonable cost for this application. Methods were examined that used photo produced neutrons or the bremsstrahlung photons only, or use both photons and neutrons in combination for irradiations of selected LLFP. Extrapolating the results to plausible engineering scale transmuters it was found that the energy cost for 129 I and 99 Tc transmutation by these methods are about 2 and 4%, respectively, of the energy produced from 1000MWe.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Clement, J. D.↗

Application of gaseous core reactors for transmutation of nuclear waste

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Schnitzler, B. G.↗

Transmutation Fuels Development in the Advanced Fuels Campaign

Metallic fuels have a history of use that spans the entire nuclear age, especially in association with the development of the LMFBR. Initially, metallic fuels were employed due to their ease of fabrication and high heavy metal density (which led to the most favorable breeding efficiencies in fast reactors). During subsequent decades of use, metallic fuel technology has matured significantly, leading to the realization of major benefits in the areas of fuel reliability and burnup, proliferation-resistant recycling, remote fabrication, and passive reactor safety. Such characteristics continue to make metallic fuels a highly attractive technology, especially for use with future sodium fast reactors. Current research and development activities on metallic fuels in DOE's Advanced Fuels Campaign are focused on their potential use for actinide transmutation in advanced SFRs. In such a mission, metallic fuels would be employed as part of a closed fuel cycle into which plutonium and minor actinides (Am, Np) recovered from LWR spent fuel would be incorporated. All actinides in this metallic fuel would be fissioned or transmuted in a fast reactor to eliminate the extremely long-lived, heat-producing actinides in order to reduce the volume, heat load, and radio-toxicity burden on a future geological repository. Thus, the need exists to demonstrate reliable performance of metallic fuels with Pu and minor actinide constituents to very high burnup. As part of this application, there is also a need for a near zero-loss fabrication process, since losses during fuel fabrication are a potential source of long-lived actinides destined for geologic disposal. This paper presents the current state and future direction of the development of metallic fuels for actinide transmutation in the US. Metallic fuels have had a long history of use in sodium-cooled fast reactors. Continued research, development, and design innovation over a period of decades have demonstrated this fuel technology to be safe and reliable to extremely high burnups over a wide range of compositions. Their simple fabrication has historically given them a significant economic advantage over other fuel forms. Research and development on metallic fuels continues, with particular emphasis on showing that this fuel technology can incorporate significant quantities of plutonium and minor actinides for the purpose of efficiently and economically transmuting these elements in a fast-spectrum reactor. New fabrication methods to improve retention of the volatile fuel constituents during the casting process (and reduce waste generation), are being developed. Irradiation testing of the new metallic fuels has so far indicated that the new compositions, although differing significantly both in the quantities of Pu and minor actinides included and in the Zr alloying addition, perform as a function of fission density very analogously to the historic U-Zr and U-Pu-Zr fuels. In the next few years it is anticipated that a new casting furnace prototype will demonstrate that these new metallic fuel alloys can be fabricated at engineering scale, in a remote environment, with no significant fuel loss or waste generation. Irradiation tests currently underway will provide the performance data needed to assess fuel reliability/burnup capability and identify any unique behaviors that are a result of the new compositions. If these future results are consistent with the intermediate data available at present, metallic fuels are expected to continue to be a strong contender for the fuel technology choice of future fast reactors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

An Introduction to the FCRD Transmutation Fuels Handbook 2015

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. For this reason, the DOE Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) has been performing research on nuclear fuel technologies that could be used in conjunction with a future actinide transmutation mission. One important approach to transmutation involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. It is, therefore, important to understand the properties of U-Pu-Zr alloys. The Idaho National Laboratory has recently published a handbook intended to summarize key properties of these alloys for researchers. This handbook focuses on phases and key thermophysical properties of elements and alloys in the U-Pu-Zr system. It is envisioned as the first of a series, which will serve as a summary of current knowledge about important properties of transmutation fuels. Future volumes will include information about properties and materials not covered in the first edition, as well as updates to information in this edition of the handbook. The handbook is intended to serve two audiences. One audience is researchers, who may find it useful to know what has been measured, where it has been published, and what kinds of information are needed. A second audience is modelers, who need a concise summary of information about specific properties and the accuracy with which they are known. The two audiences have different needs, and the handbook is a compromise intended to be useful to both. Therefore, it includes summaries of research results and needs, but generally does not present the original data. It also includes numerous graphs and, where available, equations that provide convenient approximations to materials properties for modelers. It is important to understand that most of the equations provided in this handbook are derived by fitting polynomials to data and have no physical basis; while they can be used with caution for design studies, fuel modelers will probably want to use the original experimental data to develop or validate their own equations and/or models. Although the handbook includes some citations to research involving development of models, its primary focus is experimental data. Many of the measurements of the U-Pu-Zr system were made before 1980. Where possible, the handbook relies on previous reviews of experimental data rather than presenting new interpretations of the old information. It is expected that researchers will refer to the original literature for details that are not presented in the handbook. Metals and alloys in the U-Pu-Zr system are uniquely difficult to study for a number of reasons. Experimental designs may need to be modified to allow researchers to work with these hazardous materials safely. Kinetics of phase transformations are commonly sluggish, and it is highly unlikely that all of the samples in previous research had equilibrium phases and microstructures. Thermal cycling effects (changes in sample characteristics that are not fully reversed during cooling) are commonly reported and may be significant. Careful control of impurities (particularly oxygen) during sample fabrication and data collection is both crucial and difficult, and concentrations below 100 ppm in samples may significantly affect thermophysical properties. For all of these reasons, differences between measured values for similar alloys are commonly far larger than the estimated errors for individual measurements. Further research on well-characterized materials with careful control of impurities is needed for all of the alloys discussed here. The 2015 FCRD Handbook summarizes currently available information about U, Pu, Zr, and their binary and ternary alloys. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, it attempts to provide information about how well the property is known and how much variation exists between measurements. Although the handbook includes some references to publications about modeling, its primary focus is experimental data. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fast reactor core concepts to improve transmutation efficiency

Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 10 4 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗