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A method of fuel injection rate shaping of the Diesel engine common rail fuel system with common rail injectors and solenoid control is proposed. The method envisages the impact on control current of impulses applied to the control solenoid valve of the common rail injectors for variation of the injection rate shape. At that, the fuel is supplied via two groups of injection holes. The entering edges of the first group with the coefficient of flow, ??B, were located in the sack volume and the entering edges of the second group (coefficient of flow, ??H) - on the locking taper surface of the nozzle body. The coefficients of flow, ??B, and ??H differ considerably and depend on the valve needle position. This enables to adjust the injection quantity by injection holes taking into account operating conditions of the Diesel engine and hence - by the combustion chamber zones. Using the constant fuel flow set-up, characteristic of the effective cross-section of the common rail fuel system injector holes was investigated. The diameter of injector holes was 0.12 ? 0.135 mm. The excessive pressure at the entering edges varied from 30 to 150 MPa and more and the excessive pressure in the volume behind the output edge - from 0 to 16 MPa.

This paper presents results obtained by reproducing a VVER-1000 lattice physics benchmark using the Serpent2 code. The benchmark data are related to the initial cycles of the Khmelnitskiy nuclear power plant. This study compares results obtained with the Serpent2 code against those calculations published in the benchmark document. Detailed descriptions are provided for two fuel assemblies; one without gadolinium in the fuel pins, and the other with gadolinium. A sensitivity study is performed utilizing different nuclear data libraries, different depletion sequences (including pin by pin burnup) and different number of radial regions within the fuel pins containing gadolinium. The infinite multiplication factor behavior and isotopic concentrations of major isotopes are compared and discussed. It is concluded that Serpent2 provided reasonable results in comparison with the other codes used in the benchmark.

This work applies to recent initiatives at the Radiochemical Engineering Development Center at Oak Ridge National Laboratory to optimize the production of transcurium isotopes in the High F...

Three-dimensional (3D) full-core calculations are an integral part of fuel reload design for today’s light water reactors (LWRs). The current approaches are typically based on the nodal diffusion codes that calculate criticality state points and power distributions as a function of burnup. The CASL VERA (Consortium for Advanced Simulation of Light Water Reactors, Virtual Environment for Reactor Applications) code represents one of the latest advancements in 3D full-core calculation analysis based on transport theory methods employing the Method of Characteristics (MOC) and coupled multi-physics. In this article, a publicly released version of the NuScale reactor core is analysed with the VERA code (version 3.9 and 4.1) and contrasted against some static Serpent and Polaris based simulations. The analysis shows an excellent agreement for the lattice-level calculations as well as with some of the 3D full-core models. However, larger deviations were found in cases with heavy reflector models, whereby the reflector composition was found to impact the differences between the VERA and Serpent results. In this analysis, it was determined that greater than 90% stainless steel content in the heavy reflector leads to higher deviations between the VERA results and the Serpent results. The burnup calculations showed that the presence of the heavy reflector extends the cycle length and also leads to a flatter power distribution in the core, which can generally be interpreted as more efficient.

Safety analyses at the high flux isotope reactor (HFIR) are required to qualify experiment targets for the production of plutonium-238 (238Pu) from neptunium dioxide/aluminum cermet (NpO2/Al) pellets. High heat generation rates (HGRs) due to fissile material and low melting temperatures require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. These simulations are achieved in a fully coupled thermo-mechanical analysis using comsolmultiphysics for four different preliminary target designs using an evolving set of pre- and postirradiation data inputs, and subsequently evolving solution scopes, from the unique pellet and target designs. A new comprehensive presentation of these preliminary analyses is given and revisited analyses of the first prototypical target designs are presented to reveal the effectiveness of evolving methods and input data.

Uncertainty quantification is an indispensable analysis for nuclear reactor simulation as it provides a rigorous approach by which the credibility of the predictions can be assessed. Focusing on propagation of multi-group cross-sections, the major challenge lies in the enormous size of the uncertainty space. Earlier work has explored the use of the physics-guided coverage mapping (PCM) methodology to assess the quality of the assumptions typically employed to reduce the size of the uncertainty space. A reduced order modeling (ROM) approach has been further developed to identify the active degrees of freedom (DOFs) of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. In the current work, a sensitivity study, based on the PCM and ROM results, is applied to identify a suitable compressed representation of the uncertainty space to render feasible the quantification and prioritization of the various sources of uncertainties. While the proposed developments are general to any reactor physics computational sequence, the proposed approach is customized to the TRITON-NESTLE computational sequence, simulating the BWR lattice model and the core model, which will serve as a demonstrative tool for the implementation of the algorithms.

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...

This article presents a benchmark study of a VVER 1000 core simulation of hot zero power (HZP) tests. Measured experimental data from the Temelin Unit 2 nuclear power plant’s operation test during its first reactor startup were compared using MCNP as well as the combination of Serpent + NESTLE lattice-to-diffusion nodal simulation. Results from full-core 3D models included multiplication factor and moderator temperature coefficients (MTC) that were compared to experimental data, which revealed a very good agreement with criticality states during HZP tests, and a reasonable agreement for MTC. Therefore, it is concluded that the Serpent-NESTLE combination and MCNP, both, provided reasonable results for this benchmark. The Serpent-NESTLE code combination was used for the first time in this study and has shown to be capable of providing high quality simulations in an efficient fashion. To encourage future benchmarks by other researchers, this article makes a special effort to document the core design information and details that would facilitate such efforts.

This article describes the light water reactor optimizer (LWROpt), a fuel cycle optimization code originally developed for BWRs, which has been adapted to perform core fuel reload and/or operational control rod management for pressurized water reactors (PWRs) and small modular reactors (SMRs), as well. Additionally, the eighth-core symmetric shuffle option is introduced to help expedite large-scale optimizations. These new features of the optimizer are tested by performing optimizations starting from a base case of an SMR core model that was developed manually and unrodded. The new fuel inventory (NFI) and loading pattern (LP) search in LWROpt was able to eliminate all of the constraint violations present in the initial base solution. However, independent control rod pattern (CRP) searches for the best several LPs found were not successful in generating CRPs without any constraint violations. This indicates that fully decoupling the fuel loading from the CRP optimization can increase the computational tractability of these calculations but at the expense of effectiveness. To improve on the individual search results, a coupled fuel loading (NFI and LP) and CRP search was performed, which produced a better overall result but still with some small constraint violations, emphasizing the fact that optimizing the fuel loading arrangement in a small high-leakage unborated core while concurrently determining its operational rod patterns for a 4-year operational cycle is no easy feat even to an experienced core designer; thus, this process can be greatly aided by employing automated combinatorial optimization tools.

The High Flux Isotope Reactor located at the Oak Ridge National Laboratory utilizes a large cylindrical beryllium reflector that is subdivided into three concentric regions and encompasses the compact reactor core. Nuclear transmutations caused by neutron activation occur in the beryllium reflector regions, which leads to unwanted neutron absorbing and radiation emitting isotopes. During the past year, two topics related to the HFIR beryllium reflector were reviewed. The first topic included studying the neutron poison (helium-3 and lithium-6) buildup in the reflector regions and its affect on beginning-of-cycle reactivity. A new methodology was developed to predict the reactivity impact and estimated symmetrical critical control element positions as a function of outage time between cycles due to helium-3 buildup and was shown to be in better agreement with actual symmetrical critical control element position data than the current methodology. The second topic included studying the composition of the beryllium reflector regions at discharge as well as during decay to assess the viability of transporting, storing, and ultimately disposing the reflector regions currently stored in the spent fuel pool. The post-irradiation curie inventories were used to determine whether the reflector regions are discharged as transuranic waste or become transuranic waste during themore » decay period for disposal purposes and to determine the nuclear hazard category, which may affect the controls invoked for transportation and temporary storage. Two of the reflector regions were determined to be transuranic waste at discharge and the other region was determined to become transuranic waste in less than 2 years after being discharged due to the initial uranium content (0.0044 weight percent uranium). It was also concluded that all three of the reflector regions could be classified as nuclear hazard category 3 (potential for localized consequences only).« less

The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-I, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-I were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.

One of the limiting contributors to the heat load constraint for a long term spent fuel repository is the decay of americium-241. A possible option to reduce the heat load produced by Am-241 is to eliminate it via transmutation in a light water reactor thermal neutron environment, in particular, by taking advantage of the large thermal fission cross section of Am-242 and Am-242m. In this study we employ lattice loading optimization techniques to define the loadings and arrangements of fuel pins with blended americium and uranium oxide in boiling water reactor bundles, specifically, by defining the incineration of pre-loaded americium as an objective function to maximize americium transmutation. Subsequently, the viability of these optimized lattices is tested by assembling them into bundles with Am-spiked fuel pins and by loading these bundles into realistic three-dimensional BWR core-wide simulations that model multiple reload cycles and observe standard operational constraints. These simulations are possible via our collaboration with the Westinghouse Electric Co. which facilitates the use of industrial-caliber design tools such as the PHOENIX-4/POLCA-7 sequence and the Core Master 2 © GUI work environment for fuel management. The resulting analysis confirms the ability to axially uniformly eliminating roughly 90% of the pre-loaded inventory of recycled Am-241 in BWR bundles with heterogeneous target pins. This high level of incineration was achieved within three to four 18-month operational cycles, which is equivalent to a typical in-core residence time of a BWR bundle.

The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear appli...

This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the curre...

At the University of Cincinnati Nuclear and Radiological Engineering Program, a once moth-balled vintage subcritical reactor facility was reassembled to provide a practical laboratory experience for students in nuclear science. The relative simplicity and accessibility of this facility are some of its most desirable features. Recent work had led to supplementing the experimental aspects of this facility with a three-dimensional, full detail, MCNP model of this reactor. This article emphasizes the latest validations of this simulation model against the most recently available laboratory measurements of neutron flux distributions and subcritical neutron multiplication factors, while illustrating a unique feature of this facility which enables experimentalists to carry out step by step measurements, and the corresponding simulations, while the core is being loaded and unloaded. Furthermore, additional simulations are herein included to evaluate the impact of different moderator and reflector arrangements upon core reactivity, including a few combinations of heavy water and/or beryllium, primarily as a venue for researchers to begin studying potential upgrades or changes to the current design.

This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

Various arrangements of beryllium loadings to create an internal neutron reflector in the flux trap region of the Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR) have been investigated. In particular, the impact upon fuel cycle length has been studied by performing calculations using the HFIR MCNP-based model HFV4.0. This study included examining perturbations in reactivity, flux, and power distribution caused by the various beryllium loadings. The HFIR Cycle 400 core configuration was used as a reference to calculate the impact of beryllium loadings upon cycle length. Three different configurations of beryllium loadings were investigated and compared against the Cycle 400 benchmark calculations; Cases 1 through 3 modeled combinations of 12 and 18 beryllium rods loaded into unused experimental sites. Calculated eigenvalues have shown that potential increases in reactivity between 0.56 and 0.79 dollars are attainable, depending on the various beryllium configurations. These results correspond to possible increases in fuel cycle length ranging between 2.3% and 3.3%. On the basis of their practicality, cost versus benefit, and greater potential for implementation, Cases 2 and 3 (both with 18 beryllium rods) were studied further and are herein reported in greater detail. Neutron flux distributions for Cases 2 and 3 were calculated at the horizontal mid-plane of the flux trap region, which showed no significant changes in the thermal flux magnitude and radial profile in comparison to Cycle 400. Likewise, safety analysis related parameters were contrasted, revealing power increments of up to 2% near the inside edge of the inner fuel element, well below the maximum acceptable value of 9%, a standing guideline employed for experiments at the HFIR. Additionally, the average neutron heat generation rate in beryllium rods and the maximum heat generation rate were evaluated to confirm that the design provides adequate coolant flow inside the rod and around the beryllium targets to carry away any excess heat. Case 2 with 18 beryllium rods was recommended for implementation at the HFIR. In fact, funding has been allocated for this project and beryllium rod fabrication is scheduled. The proposed beryllium loading does not involve permanent design changes and is predicted to yield a meaningful increase in fuel cycle length that could translate into potential annual savings in direct fuel costs of approximately $235,000, or of more than five million dollars over the projected life of this reactor.

The determination of the family of optimum core loading patterns for pressurized water reactors (PWRs) involves the assessment of the core attributes for thousands of candidate loading patterns. Fo...

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

Post-irradiation uranium isotopic atomic densities within the core of the High Flux Isotope Reactor (HFIR) were calculated and compared to uranium mass spectrographic data measured in the late 1960s and early 70s [1]. This study was performed in order to validate a Monte Carlo based depletion methodology for calculating the burn-up dependent nuclide inventory, specifically the post-irradiation uranium

The mission of the Reduced Enrichment for Research and Test Reactors (RERTR) Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low enriched uranium (LEU) fuel and targets. Oak Ridge National Lab (ORNL) is reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction of flux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. A current 3-D Monte Carlo N-Particle (MCNP) model was modified to replicate the HFIR Critical Experiment 3 (HFIRCE-3) core of 1965. In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. Foils (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil s activity to the activity of a normalizing foil. The current work consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the normalizing foil. Power distributions were obtained for the clean core (no poison in moderator and symmetrical rod position at 17.5 inches) and fully poisoned-moderator (1.35 g B/liter in moderator and rods fully withdrawn) conditions. The observed deviations between the experimental and calculated values for both conditions were within the reported experimental uncertainties except for some of the foils located on the top and bottom edges of the fuel plates.