Concentrations Of Fission Products Research Articles

Searching for optimal reactivity management using a new burnable absorber for SCLWR with (Th, 233U)O2

This work investigates the efficiency of some new burnable absorbers (BAs) suggested to provide possible improvements in the fuel management of a supercritical light water reactor (SCLWR) assembly fueled with (Th, 233U)O2. Four BAs including gadolinium, Erbium, Iridium and Lutetium are suggested as integral burnable absorber (IBA) rods. A three-dimensional model of SCLWR has been modelled using MCNPX version 2.7. A complete neutronic analysis has been carried out for the suggested BAs and compared with the gadolinium vector which is the common BAs. The IBA rods were distributed through the SCWR fuel assembly in a way that provided a flat power distribution. Various concentrations of the suggested BAs have been investigated to obtain the optimum value that can suppress the excess reactivity at the beginning of the cycle and flatten the infinite multiplication factor. The burnup parameters including the fertile, fissile, BAs and most effective fission product concentration as a function of effective full power days (EFPDs) of the SCLWR have been examined for the suggested BAs. The reactivity temperature coefficients have been studied to ensure the viability of the suggested BAs. A further analysis is performed to study the axial, radial power distribution and local power peaking factor.

Investigating the possibility of using a mixture of thorium with different fissile materials as a fuel in TRISO particles for the PBMR-400 reactor

The increasing demand for nuclear reactors as a source of clean energy forced nuclear researchers to search for ways to enhance the safety operation of nuclear reactors. This paper discusses the viability of using different thorium compounds ((233U, Th)O2, (235U, Th)O2, (RGPu, Th)O2) as an accident tolerant fuel/advanced technology fuel (ATF) for Pebble Bed Modular Reactor 400 (PBMR-400). A comprehensive analysis of thorium-based fuel performance was carried out and compared with UO2 to develop a novel nuclear fuel with the capability to tolerate severe accident conditions. The neutronic characteristics such as the infinity multiplication factor, initial heavy metal concentration, RGPu concentration and fission products concentration were analyzed. In addition to, the main safety parameters have been calculated to distinguish the effect of the proposed fuels on them. The thermal spectrum and thermal output power have been visualized to give a good insight view inside the core about the distribution of the heat. In general, the neutronic analysis demonstrated good behavior of the examined thorium-based fuel in the PBMR-400 and especially in the case of (RGPu, Th)O2. PBMR-400 showed excellent performance in incinerating a large amount of reactor-grade plutonium when it was used in a kernel fuel bed.

Discussing the possibility of using thorium-based fuels as an alternative fuel to uranium dioxide fuel for APR-1400 reactor

This work investigates the possibility of using thorium-based fuels as an alternative fuel for advanced power reactor APR-1400. MCNPX code version 2.7 with cross-section library ENDF.VII has been used to design an APR-1400 fuel assembly. This code has been used to study the neutronic performance of the proposed thorium-based Fuel types (0.944Th, U)O2, (0.955Th, 233U)O2, (0.934Th, rgPu)O2. The fuel burn-up parameters such as infinity multiplication factor (kinf), initial heavy metals concentrations, Minor actinides concentration and fission products concentration have been analyzed during 1500 effective full power days (FPDSs) for thorium-based Fuel types and compared with the common fuel. The analysis of the horizontal thermal power and neutron flux distribution provides valuable insights into the behavior and performance of the suggested Fuel types in the APR-1400 assembly. The analysis of the neutronic results ensures the viability of using the proposed thorium-based fuel types as an alternative fuel to UO2 because they achieved acceptable safety parameter values and provided a good power distribution through the fuel assembly compared to UO2.

Computational study of thermophysical properties of cerium doped UO2: Effect of oxidation states

The interaction of fission products with UO2 fuel, which is generated during the nuclear fission, can amend the fuel behavior. Among different lanthanide (Ln) fission products produced, Cerium (Ce) shows considerable yield. The Ce substitution in UO2 with isovalent Ce4+ cation is interesting due to the absence of oxygen vacancies and no change in the oxidation state of U atoms in the doped system. In this study, the trivalent and tetravalent Ce atom substitution in UO2 and their effect on thermophysical properties is investigated using density functional theory. Further, the effect of Ce doping concentrations in UO2 is assessed by varying the Ce ion concentrations in the lattice (6.25 %, 12.5 %, 25 %, and 50 %) for various charge balancing mechanisms. The lattice volume of Ce doped UO2 structures show higher or lower values compared to undoped UO2 depending upon the oxidation state of Ce and U atoms. The bulk modulus tends to reduce for Ce substituted UO2 as compared to pure UO2. The electronic density of states analysis shows that Ce substitution levels as well as the oxidation state of Ce/U atoms strongly influences the band structure of the UO2 lattice. The specific heat capacity of Ce doped UO2 structures is calculated at different Ce doping concentrations and for disparate charge balancing procedures using the quasi harmonic approximation(QHA). Present results show good agreement with the experimental values reported for similar Ce substituted UO2 structures. Thus, the systematic investigation with varying Ce fission product concentrations and oxidation states of cations (Ce and U) can give insight into the fuel properties of UO2 under reactor conditions.

Contributions to the mechanistic understanding of the microstructural evolution in irradiated U-Mo dispersion fuel

Advanced microstructural characterization techniques, such as scanning electron microscopy (SEM) and scanning transmission electron microscopy - energy dispersive x-ray spectroscopy (STEM-EDS), were used to interpret the fuel microstructure evolution and fission products behavior in U-Mo dispersion fuel irradiated in the Advanced Test Reactor (ATR) as part of the European Mini-Plate Irradiation Experiment (EMPIrE) test. The larger as-fabricated fuel grain size achieved by heat-treating the U-Mo powder resulted in slower high burnup structure (HBS) development and reduced fission gas porosity. Slower HBS kinetics was observed at the fuel kernels’ periphery, which contained smaller and less fission gas bubbles at all fission densities (FDs) investigated and was attributed to a locally reduced damage density and fission products concentration, as corroborated with Monte Carlo simulations. The non-refined grains at the fuel kernel periphery hosted a perfectly ordered fission Gas Bubble Superlattice (GBS) up to 6.3 × 1021 fissions/cm3. Nano-scale STEM-EDS analysis presented in this study provided useful information on the GBS characteristic morphology and evolution in U-Mo fuel. The concentration of fission gas in the GBS progressively increased with FD, pointing to an evolution of the nanobubble pressure status with irradiation. A possible connection between the GBS collapse and HBS onset is proposed for which there exists a threshold in the misorientation of the refined sub-grains above which the GBS stability during irradiation is no longer preserved, resulting in the GBS collapse.

Electrostatic precipitator collection efficiency studies using atmospheric radon progeny as aerosol analogs for nuclear explosion radionuclides

Electrostatic precipitation (ESP) is an attractive low-powered collection mechanism for mobile and fixed aerosol detection of radionuclides (RNs) for Nuclear Explosion Monitoring (NEM). Aerosol samplers deployed in the International Monitoring System use a blower to draw air through a filter media to collect particulates. ESP-based samplers collect aerosols without a filter, which can greatly increase volumetric flow capacity per watt of power consumed. ESP-based collectors may be optimized to perform low-power mobile RN collection or to improve the air throughput of existing monitoring stations. This effort describes the use of unknown concentrations of atmospheric RNs to determine the collection efficiency of a compact ESP design. For this analysis, naturally occurring radon progeny are simultaneously collected by a single stage wire-plate ESP and a filter-based sampler with a known collection efficiency. The activity of resulting samples is measured with gamma-spectroscopy and decay corrected for analysis time offsets. RN collection efficiencies are then derived for an experimental survey of ESP operational parameters that influence the ionization, transit, and collection of aerosols. At volumetric flow rates of 1.5–2 CMM, the optimized collection efficiency was calculated as 21±2%, and slower rates around 0.5 CMM resulted in 55 ±5% collection efficiency. The monitoring performance of the ESP-based collector was assessed for a simplified nuclear explosion source term by calculating the minimal detectable concentrations of short-lived fission & activation products. Results of the study suggest that a low-power ESP is feasible for NEM at distances of 100s of km.

Attractiveness of Fissile Material in Nuclear Wastes from Different Fuel Cycles

Nuclear fuel cycle advancements will result in new types of fissile material, including nuclear wastes, that require security and safeguards. Nuclear wastes may be more vulnerable for diversion by non-state actors, and chemical processing to recover fissile material is not an insurmountable challenge. Previous work has applied a figure of merit (FOM) to assess material attractiveness and security risks. This analysis applies the material attractiveness FOM to wastes produced by fuel cycles from the Fuel Cycle Evaluation and Screening (FCES) study. Two aspects of security risk are studied: (1) the time before the fissile material in the waste becomes attractive and (2) the number of waste packages required to obtain a critical mass of fissile material. Two fuel cycles are presented to highlight detailed results: (1) once-through use of low-enriched U in light water reactors (LWRs) and (2) continuous recycle of Pu in sodium fast reactors (SFRs). Increasing LWR used nuclear fuel (UNF) package loading increases the time to attractiveness, but the larger packages contain enough Pu for multiple critical masses. The high-level waste (HLW) from processing the SFR fuels has similar FOM behavior but longer time to attractiveness due to the concentration of fission products. More HLW packages are required to obtain a critical mass; that number can be further increased by increasing the separation efficiency. Extended to all FCES fuel cycles, the minimum time before attractiveness is generally lower for UNF than for HLW because radioactivity is concentrated in HLW. For nearly all fuel cycles that produce UNF, only one package is required to recover enough fissile material for a critical mass. Notably, some advanced fuel cycles produce HLW, of which only two packages need to be recovered to obtain a critical mass, even when the target fissile material is recycled. Going forward, an assessment of the security risks posed by fissile material in nuclear wastes will need to quantify the challenge posed by separations. Ultimately, the assessment could inform security and response measures; whether any of the observations might affect these measures could be an area for future work. Finally, future analysis could study whether different fuel cycle wastes are more attractive for use in radiological dispersal devices or radiological exposure devices.

Fission products speciation in nuclear fuel: Synthesis and characterisation of mixed oxide (U,Pu)O2 SIMfuel

In the context of the nuclear fuel recycle, U1-yPuyO2±x Mixed Oxides (MOx) are currently the most studied fuels. These materials have to meet accurate physico-chemical properties, known to change drastically during irradiation with the formation of fission products (FP). Investigating the FP behaviour is thus crucial to predict the fuel properties evolution during irradiation. Model uranium-based materials called SIMfuel have been developed in the last decades to overcome the high radiotoxicity of UO2 spent fuel, and to enable separation-of-effect studies on its otherwise overwhelmingly complicated chemistry. In this work, a fabrication route for U1-yPuyO2±x SIMfuel (SIMMOx) has been developed, in order to study the speciation of soluble FP (Ce, La, Nd, Sr, Y, Zr) inside spent MOx fuel, and their interaction with the U1-yPuyO2±x matrix. The resulting material is representative of irradiated MOx fuel with Pu content of 21 g/g% (Pu/(U+Pu)=26 g/g%) and a burnup of 13 at.%.XRD and EPMA show a material representative of real spent fuel (U,Pu)O2 matrix: globally monophasic, and compliant with the homogeneity requirement of irradiated MOx. Raman microscopy confirmed the lack of secondary phases, showing that all the FP successfully entered the (U,Pu)O2 solid solution. Pu-enriched zones (<30 μm) contain higher FP concentration, showing that this phenomenon, already observed in irradiated fuel, is due also to thermodynamics and not only to burnup, as previously assumed.XAS analysis supplied crucial information such as the oxidation states, the O/M ratio, and the prevalent chemical form of each actinide, and FP. A charge compensation mechanism between the (U,Pu)O2 matrix and the FP in solution has been observed.

Accurate prediction of gaseous isotopes Xe and Kr concentrations from the burnup of nuclear fuel using simple regression algorithm.

Mathematical techniques for modeling and simulating dangerous or complex systems, such as nuclear technology systems, often require high-performance computing to process and analyze available data. In this paper, simple and quick method to support studies and research related to nuclear fuel is presented. This reasonably simple method helps to predict different concentrations of actinides and fission products in nuclear fuels without the need for expensive specialized programs and highly-trained researchers. The great importance of this approach is the speed of predicting the components of nuclear fuel concentrations, which in turn leads to quick decision-making, such as the possibility of operating fuel at higher burnup values, predicting the amount of gases resulting from nuclear fission (which may accumulate and cause problems in nuclear fuel such as volume swells), and other important decisions in nuclear fuel technology. The predicted equations have been generalized for higher values of burnup and compared with comparable results from MCNP codes. The equations deduced in calculating the different concentrations of xenon and krypton isotopes resulting from fission in burnup of nuclear fuel showed very precise results with discrepancies (magnitude of an error between the data points and the corresponding predicted ones) less than 2%. The suggested method offers a great advantage for researchers, which are the use one of any simple or common computational programs available to most researchers and do not need much experience to deal with, such as MATLAB, Excel that are easy to use for regression analyses. In this paper, the advantages of the proposed method are explained along with the limitations of its use.

Monte Carlo Analysis of Coolant Stream Impurity Gamma Emissions in Gas-Cooled Fast Reactors

It is well established that a rapid increase in the concentration of fission products in the reactor coolant stream can serve as an early indication of fuel failure. We use Monte Carlo simulations to investigate the feasibility of using several gamma detectors as diagnostic equipment to monitor the presence of major fission product isotopes in high-temperature gas-cooled reactor (HTGR) coolant streams for early detection of fuel failure and therefore the prevention of fuel failure conditions. We model the response of high-purity germanium (HPGe), CdZnTe, NaI(Tl), and LaBr3(Ce) detectors of typical commercial sizes to the gamma emissions from nuclides expected to be found within the coolant stream of the Versatile Test Reactor (VTR) under development by the U.S. Department of Energy. The results indicate that for the 233- and 250-keV gamma rays from 133Xe and 135Xe, respectively, the detection criterion is met in under 1 min using a single HPGe detector. Changes in other spectral lines associated with Xe nuclides are detected within 1 h regardless of the choice of detector.

Scoping Studies for a Lead-Lithium-Cooled, Minor-Actinide-Burning, Fission-Fusion Hybrid Reactor Design

A feasibility study of a subcritical fission-fusion hybrid reactor using lead-lithium eutectic as a coolant and minor actinides (MAs) as fuel is presented. Such a reactor could support the fission community by transmuting MAs and the fusion community by breeding tritium. The feasibility of such a reactor for the burnup of MAs is assessed in terms of burnup performance, tritium breeding, and safety characteristics. Tandem mirrors are a promising neutron source technology, and a deuterium-tritium tandem mirror is considered here for the neutron source with power Psource = 1.13 MW assumed for scoping purposes. Subcritical reactivities from keff = 0.9800 to keff = 0.9950 were considered, representing the initial reference for subcritical reactivity and the assessed upper limit, respectively. Stability analyses indicated the reactivity would be stable under perturbations of fuel, coolant, and inlet temperatures, with a positive reactivity insertion expected during reactor shutdown. This range corresponded to nuclear heating values of 150 to 650 MW and mass burn rates of 53 to 216 kg/year. The upper mass burn rate limit would require 1110 reactor years with a capacity factor of 0.9 to fission the global supply of MAs and could offset the annual U.S. MA production with eight reactors. Tritium breeding was assessed for keff = 0.9800 and 3.795% 6Li enrichment in the coolant, and a tritium breeding ratio of 1.602 0.017 was tallied, suggesting the reactor could, without elevated 6Li enrichment, produce tritium to both sustain operation and supply tritium for other fusion devices. Time-series modeling of fuel burnup was conducted for a four-batch loading scheme and three different fuel residence times at keff = 0.9800, which showed system performance would drop with burnup, and that the rate of this drop was lower for longer fuel residence times, motivating a means of reactivity control. Last, changes in fuel composition with burnup were assessed for relative concentrations of MAs, transmutation products, and fission products. The breeding of plutonium in the blanket was calculated and found to be of minimal proliferation concern.

A HELIOS-Based Dynamic Salt Clean-Up Study Analysing the Effects of a Plutonium-Based Initial Core for iMAGINE

Nuclear technologies have strong potential and a unique role to play in delivering reliable low carbon energy to enable a net-zero society for future generations. However, to assure the sustainability required for its long-term success, nuclear will need to deliver innovative solutions as proposed in iMAGINE. One of the most attractive features, but also a key challenge for the envisaged highly integrated nuclear energy system iMAGINE, is the need for a demand driven salt clean-up system based on the principles of reverse reprocessing. The work described provides an insight into the dynamic interplay between a potential salt clean-up system and reactor operation in a plutonium-started core in a dynamic approach. The results presented will help to optimise the parameters for the salt clean-up process as well as to understand the differences which appear between a core started with enriched uranium and plutonium as the fissile material. The integrated model is used to investigate the effects of the initial fissile material on core size, achievable burnup, and long-term operation. Different approaches are tested to achieve a higher burnup in the significantly smaller Pu-driven core. The effects of different clean-up system throughputs on the concentration of fission products in the reactor salt and its consequences are discussed for general molten salt reactor design. Finally, an investigation into how a plutonium loaded core could be used to provide fuel for future reactors through fuel salt splitting is presented, with the outcome that one Pu-started reactor of the same size as a uranium-started core could deliver fuel for 1.5 new cores due to enhanced breeding. The results provide an essential understanding for the progress of iMAGINE as well as the basis for inter-disciplinary work required for optimising iMAGINE.

Ultra-low-level measurements of airborne fission products from the Fukushima Daiichi reactor accident using high volume collection systems at Savannah River National Laboratory

Ultra-low-level measurements of radionuclides in air have been conducted at the Savannah River National Laboratory (SRNL) to determine the atmospheric concentration of fission products released following the Fukushima Daiichi reactor accident on March 11, 2011. Air filter samples were acquired from two high-volume collection systems (a traditional filter-based system and an electrostatic precipitator-based system) to monitor airborne radionuclide concentrations in the period covering from 2 weeks to 3 years after the disaster. The world-wide spread of low-level concentrations of airborne fission products from the Fukushima event provided a unique opportunity to demonstrate SRNL's electrostatic particle collection technology and other improvements in environmental monitoring developed at the Savannah River Site (SRS). Detecting and analyzing the release allowed a comprehensive test of SRS systems for monitoring environmental radioactivity. Gamma-ray-emitting fission products (131,132I, 134,136,137Cs, and 129,132Te) and cosmogenic isotopes (7Be and 22Na) in air were detected and quantified by high-resolution gamma-ray spectroscopy at concentrations as low as 0.07 μBq per standard cubic meter (SCM) (50 mBq total 137Cs), while plutonium content was quantified by thermal ionization mass spectrometry (TIMS) at concentrations as low as 6.5 × 10−21 g/SCM (3.0 fg 239+240Pu). Isotope concentrations measured at SRNL from gamma-ray spectroscopy were compared to independent measurements from Chapel Hill, NC, located approximately 370 km (230 mi) NE of SRNL. Meteorological modeling was also used to predict radionuclide transport from the location of release to both measurement locations.

Near Infrared Spectroscopy of Fission Products in Alkali Chloride Molten Salt – Structural Interpretation and Analytical Use

Advanced nuclear technologies using high-temperature molten salts are poised to increase the efficiency, flexibility, and acceptability of nuclear energy. To realize these advances, additional understanding of the chemistry and thermochemical and thermophysical properties of the molten salt mixtures is required. Additionally, methods to rapidly and non-destructively monitor and determine the concentration and chemistry of actinides and fission products must be developed. Electronic and vibrational spectroscopy are promising methods to both interpret the structure and speciation of dissolved species in molten salts and to determine concentrations of species of interest. We have determined the molar absorbance coefficients of selected absorbance features for several lanthanides of interest in the ultraviolet, visible, and near infrared ranges and show that some NIR absorbance features, which were previously neglected as a means of concentration determination, are analytically useful. In fact, these NIR absorbance features may have advantages over UV and visible range features for concentration determination in industrial systems. Additionally, we will present our analysis of the effects of the coordination environment on lanthanide electronic structure and transitions, yielding insight into the chemistry and coordination of these elements in molten alkali halide mixtures, which is crucial for understanding the corrosion issues in these salts amongst other requirements.Acknowledgement:This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Programs under award DE-NE0008889 and the US Nuclear Regulatory Commission (USNRC) under contract 31310018M0032. Dr. Kenny Osborne and Ms. Nancy Hebron-Isreal serve as the program managers for the DOE and NRC awards, respectively.This material is based upon work supported by the National Science Foundation Graduate Research Fellowship Program under Grant No. DGE-1447692. Any opinions, findings, and conclusions or recommendations expressed in this material are those of the author(s) and do not necessarily reflect the views of the National Science Foundation, the U.S. Department of Energy, or the United States Government.

Estimation of a Fission Product Concentration Distribution in a Spherical Body from Surface Release Measurements

The diffusion of fission products (FPs) in reactor materials affects the nuclear source term. The diffusion coefficient itself is measured through various techniques. In the release method, it is of interest to know the initial distribution of the FPs in nuclear graphite or other materials from an exterior measurement like mass surface flux or cumulative mass release. In this paper, a Fredholm integral of the first kind is considered, relating the initial distribution to the cumulative release fraction of a diffusant from a spherically symmetric body. The Tikhonov regularization, conjugate gradient least-squares (CGLS) method, and algebraic reconstruction technique (ART) with nonnegativity and conserved mass constraints were compared to fractional release data from a simulated linear profile using data for Cs diffusion in a 0.32-cm sphere NBG-18 at 1090 K. The Tikhonov regularization was shown to provide a better estimation of the initial linear distribution than the CGLS and ART methods. The performance of the Tikhonov regularization was further examined with simulated constant, quartic, and exponential initial distributions. The Tikhonov regularization provided a reasonable recovery of the exponential profile, while the estimation of the linear, constant, and quartic profiles suffers from several issues.

Effect of lanthanide fission product concentrations on the mechanical properties of UO2: A first principle based study

The fission products generated from UO2 during the nuclear fission process alters the mechanical properties of the nuclear fuel. The effect of fission atom substitution in UO2 is a requisite for the safe operation of nuclear reactors. The lanthanide atoms contribute significantly to the mixture of fission products generated. Therefore, in this work, we investigate the effect of lanthanum (La), promethium (Pm), and dysprosium (Dy) doping on mechanical properties of UO2. Further, the relation between doping concentration and different mechanical properties are derived by varying the La, Pm and Dy substitution in UO2. All the calculations are performed with density functional theory with Hubbard U corrections in the DFT + U formalism. Mechanical properties such as bulk, Young’s and shear modulus of UO2 change with the amount of dopant atom in the fuel. The results obtained from our calculations show good agreement with available experimental and theoretical findings. Thus, the present study validates the relationship between fission atom substitution and mechanical properties of UO2, which will help to understand the fuel performance at reactor conditions.

Prediction of the Average Decay Heat per Fission for MOX Nuclear Fuel

MIXED Oxide Nuclear fuel (MOX) contains both uranium and plutonium in oxidized form. It is important to calculate the nuclear decay heat due to the single thermal fission (fission due to 0.0235 eV neutron) for all fissile nuclei in the MOX fuels (U235, Pu239, and Pu241). These fissile nuclei are the main source of the decay heat in MOX fuel. Decay heat calculation of the weighted fissile material content in MOX fuel is also important. A numerical method was used in this work to calculate the concentrations of all fission products due to the individual thermal fission of the three fissile materials as a function of time N(t). The decay heat calculations for the three fissile materials are directly calculated using the summation method by knowing the different concentrations of fission products over time. The average decay heat of the MOX fuel in induced thermal fission is also concluded. The most influential nuclei in the decay heat were also identified. The method used has been validated by several comparisons before, but the new in this work is using the most recent Evaluated Nuclear Data Library ENDF/B-VIII.0. Calculations of decay heat show very common trends for a period of 107 sec after the fission burst of thermal fissions of individual fissile nuclei. Moreover, the code showed high capability in calculating the fission fragments inventories and decay heats due to the decay of fission fragments of 31 fissionable nuclei.

Fission product distribution in irradiated safety-tested and as-irradiated AGR-2 TRISO particles

Two tristructural isotropic (TRISO)-coated nuclear fuel particles were examined by electron probe microanalysis (EPMA) as part of the Advanced Gas Reactor program. The compacts’ average irradiation temperatures ranged from approximately 1260 to 1290 °C. One particle was examined in the as-irradiated condition, while the other was subject to 1600 °C post-irradiation safety testing. This study was undertaken to test a newly-developed EPMA technique to determine fission product masses in TRISO particles on a layer-by-layer basis, and to compare fission product distributions between an as-irradiated and safety-tested particle. Fission product concentration profiles were collected along two radii in each particle, with measured concentrations used to compute the fission product mass in each TRISO particle layer. These measured masses were then compared to those predicted from ORIGEN modeling calculations. Data collected from these measurements show that for these two particles, masses determined via EPMA were within ± 20% of the calculated masses for the rare-earth elements, Mo, Zr, Cs, I, and Pd. Elements that tend to be less homogeneously distributed include Sr, Te, Eu, Ag, and possibly Ba. Measured Ag masses differed by more than 40% from the calculated mass. Lanthanides other than Eu remain primarily within the fuel kernel in the as-irradiated particle but in the safety-tested particle these element masses were divided approximately equally between the kernel and kernel periphery. In both particles, the majority of Sr and Eu accumulated in the carbon-rich kernel periphery, although in the safety-tested particle, Sr and Eu accumulated farther from the fuel kernel than occurred with irradiation alone. A greater mass fraction of mobile elements, such as Cs and I accumulated in the buffer and IPyC in the safety-tested particle as compared to the as-irradiated particle. When fully developed and tested, this mass balance approach to TRISO particle analysis has the potential to provide insight into fuel behavior.

Applicability of the nuclide kinetics fast estimate model for severe accident codes

Modelling of severe accidents (SA) at nuclear power plants (NPP) includes assessing the radioactive material release into the environment (called the «source term»). The accuracy of the source term assessment using integral codes depends on the modeling error of individual phenomena such as the build-up of actinides and fission products (FP) in fuel prior to the accident, evolution of isotopic composition of fuel during the accident, release from fuel, transport, and deposition of actinides and FP in the reactor cooling system (RCS) and containment, their release into the environment.The application of best estimate approach with uncertainty analyses, the deterministic support of probabilistic safety assessment (PSA) level 2, and calculations within emergency response require multiple calculations of SA scenarios, therefore the integral codes have to run pretty fast. For consideration of isotopic evolution both in the irradiated fuel and in chemical compounds released from the fuel a depletion model has to be regularly called during each calculation session. To fit the requirement of smaller runtime, the integral code has to be provided with a fast-running depletion model. The fast performance of such models is payed by the predictive accuracy; therefore, it is important to assess if the accuracy is acceptable for the mentioned issues of SA analyses. This paper presents the results of study of errors in the fast-running depletion model that is intended for use in integral code during multi-variant calculations of SA scenarios.The validated parameters are concentrations of actinides and FP in light water reactor (LWR) spent fuel. Additionally, calculations of the decay heat power are provided. The basis for validation consists of the out-of-pile measurement results provided in Nuclear Energy Agency of Organization for Economic Co-operation and Development (OECD/NEA) database SFCOMPO-2.0, and the ORIGEN-2 code calculation results obtained by Japan Atomic Energy Agency (JAEA) for Units 1–3 of the Fukushima Daiichi NPP.

Depletion capabilities in the OpenMC Monte Carlo particle transport code

A depletion solver has been implemented in OpenMC and is described herein. The depletion solver is implemented in Python and interfaces with OpenMC’s transport solver through a C++ application programming interface, which enables an in-memory transport-depletion coupling. Multiple integration methods for advancing in time have been implemented and exhibit tradeoffs in cost, accuracy, and memory use. For all time integration methods, evaluation of the matrix exponential is performed by using the incomplete partial fraction form of the Chebyshev rational approximation method.Simulations of a pressurized water reactor (PWR) pincell and a sodium-cooled fast reactor (SFR) assembly were carried out with OpenMC and Serpent. For both problems, the use of a high-fidelity depletion chain results in predictions of keff that agree within 20–30 pcm between OpenMC and Serpent. Predicted actinide concentrations were found to agree to a fraction of a percent, and most fission product concentrations were found to agree within 1%. The few cases where larger differences were observed can be attributed either to differences in how the energy dependence of fission product yields is handled or deficiencies in the nuclear data used. OpenMC simulations of the PWR and SFR problems using a simplified 228-nuclide depletion chain demonstrate that it achieves accuracy close to that of the full, high-fidelity depletion chain with respect to the studied responses.