Temperature Coefficient Of Reactivity Research Articles

Simulation of the NuScale SMR and Investigation of the Effect of Load-Following on Component Lifetimes

The NuScale small modular reactor (SMR) has been modeled using the Virtual Environment for Reactor Applications multiphysics environment and the results compared with the publicly reported data in the Design Certification Application. The results show an excellent agreement for the compared axial and radial power distributions, temperature coefficients of reactivity, boron and control rod worths, and fast neutron flux. This NuScale model is then used to investigate the effect of different operational modes on reactor components to determine how the flexible load-following operation may affect control rod and reactor pressure vessel (RPV) lifetimes. The control rod degradation is confirmed to primarily affect the silver-indium-cadmium rod tip. The degradation rate is observed to follow a nonlinear function of core power level where the increase in degradation decreases with insertion depth. For the variation in core power levels expected with current load-following schemes, the total control rod degradation is found to be mild, at 5% to 10% of usable life per cycle for a reactor operating at <80% power. Nonetheless, this enables load-following strategies to be confirmed and/or modified to ensure that control rods do not need to be replaced during the 60+ year life of the reactor. The RPV degradation was found to be almost directly proportional to the core power level and was not overly sensitive to flux shape perturbations. Future work is planned using these damage functions to optimize operation over multiple NuScale SMR units and to develop strategies for prognostics and health management.

Preliminary Design and Study of a Small Modular Chlorine Salt Fast Reactor Cooled by Supercritical Carbon Dioxide

Small modular reactors with power below 300 MW have the advantages of small specific mass, long lifetime, and flexible power supply, and they are suitable for providing power support for small and medium-sized towns with small populations and remote areas without grid coverage. In this paper, a small modular S-CO2-cooled molten salt reactor is proposed, and the design of a 10 MW small modular chlorine salt fast reactor (sm-MCFR) with 20 years of operation without refueling is presented. The neutron feasibility of the S-CO2-cooled small modular chlorine fast reactor is analyzed in terms of neutron energy spectrum, reactivity control, temperature reactivity coefficient, and power distribution. A distinctive feature of the sm-MCFR is the use of chlorine salts with high heavy metal solubility and a hard energy spectrum, allowing the core size to be minimized while maintaining the maximum lifetime. The designed core is about 2.44 m in diameter and 2.24 m in height. Meanwhile, the sm-MCFR uses control drum control as the control system, which can effectively achieve reactivity control without increasing the reactor size. The final optimized sm-MCFR has a negative temperature reactivity coefficient, which is necessary to ensure the safe operation of the reactor.

Whole core analysis of the single-fluid double-zone heavy water-moderated molten salt reactor (SD-HWMSR)

Thermal-spectrum molten salt reactor (MSR) concepts usually adopt graphite as a neutron moderator. Although graphite as a moderator has many advantages, it also has drawbacks, including a relatively short lifespan (it has to be replaced), positive temperature feedback, and loss of impermeability due to expansion. Replacing graphite with heavy water in MSRs can effectively solve the problems introduced by graphite. The SD-HWMSR is a Single-fluid Double-zone Heavy Water-moderated Molten Salt Reactor. Optimization of SD-HWMSR’s fuel channel pitch and radius has been conducted in this work. As a result, the SD-HWMSR with a high breeding ratio (1.07465 ± 0.00060), low initial 233U loading (1.43 t), and negative temperature and void reactivity coefficients is put forward. The SERPENT-2 is used to analyze the neutronics parameters of the reactor design. The current work investigates the change in the multiplication factor, breeding ratio, and the accumulation of significant nuclides in the core. The suitable 232Th and 233U refill rates needed to maintain criticality and enable analysis of the whole core of SD-HMMSR are thoroughly determined in this study. Uranium-233 and thorium-232 both have a maximum refill rate of 3.49 and 3.20 kg/d, respectively. While the average refill rate for 233U and 232Th during the 60 years of operation is 1.90 and 2.35 kg/d, respectively. The net production of 233U rises with time, and by the end of the 60 years, it is around 1.98 t. The doubling time for the SD-HWMSR is 31 years, which is consistent with previous results.

Assembly-level analysis on temperature coefficient of reactivity in a graphite-moderated fuel salt reactor fueled with low-enriched uranium

To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors, the influences of geometric parameters on the temperature coefficient of reactivity (TCR) at an assembly level were characterized. A four-factor formula was introduced to explain how different reactivity coefficients behave in terms of the fuel salt volume fraction and assembly size. The results show that the fuel salt temperature coefficient (FSTC) is always negative owing to a more negative fuel salt density coefficient in the over-moderated region or a more negative Doppler coefficient in the under-moderated region. Depending on the fuel salt channel spacing, the graphite moderator temperature coefficient (MTC) can be negative or positive. Furthermore, an assembly with a smaller fuel salt channel spacing is more likely to exhibit a negative MTC. As the fuel salt volume fraction increases, the negative FSTC first weakens and then increases, owing to the fuel salt density effect gradually weakening from negative to positive feedback and then decreasing. Meanwhile, the MTC weakens as the thermal utilization coefficient caused by the graphite temperature effect deteriorates. Thus, the negative TCR first weakens and then strengthens, mainly because of the change in the fuel salt density coefficient. As the assembly size increases, the magnitude of the FSTC decreases monotonously owing to a monotonously weakened fuel salt Doppler coefficient, whereas the MTC changes from gradually weakened negative feedback to gradually enhanced positive feedback. Then, the negative TCR weakens. Therefore, to achieve a proper negative TCR, particularly a negative MTC, an assembly with a smaller fuel salt channel spacing in the under-moderated region is strongly recommended.

Comparative Study on the Neutronic Performance of Thermal and Fast Molten Salt Reactors under Once-Through Fuel Cycle

Once-through molten salt reactors have the advantages of easy availability of fuel, nuclear nonproliferation, and low technical difficulty. It is expected to be the earliest commercial molten salt reactor fuel cycle mode. The current research on the neutronic performance of the once-through fuel cycle mainly focuses on the thermal spectrum range, and broader spectrum research needs to be carried out further. In particular, the fuel utilization characteristics of molten salt fast reactors in once-through fuel cycle are not yet clear. In addition, small modular design is one of the trends of molten salt reactor development, and small modular design under thermal and fast spectrum has different advantages and needs further comparative analysis. In this paper, the single lattice, bare reactor, and core with reflector models are used to compare the neutronic performance in once-through fuel cycle from thermal to fast spectrum, which includes burnup, variation of fuel salt composition and volume with burnup, temperature reactivity coefficient, neutron irradiation lifetime, and small modular size. The results show that the burnup increases first, then decreases, and then increases again with the increase of the average lethargy causing fission (EALF). For small-scale molten salt reactors, the difference in fuel utilization under thermal and fast spectra is small. However, for large-scale reactors, the burnup under fast spectrum is significantly higher than that under thermal spectrum. When the EALF is large enough, and the neutron loss ratio is small enough, the breed-and-burn (BNB) fuel cycle mode can be realized, and then, the fuel utilization will be significantly improved. In the fast spectrum reactors, the HN and volume of molten salt change violently because of the longer burnup period. Based on the infinite single lattice model, the temperature reactivity coefficient increases first, then decreases, and increases again with EALF. It is negative over the entire energy spectrum. In addition, the small modular designs are discussed by the bare reactor model, and the neutron irradiation lifetime of the materials is analyzed by the core model with a reflector.

A dynamic probabilistic risk assessment platform for nuclear power plants under single and concurrent transients

ABSTRACT Dynamic probabilistic risk assessment (DPRA) of nuclear power plants (NPPs) has become one of the most critical research areas, especially in the aftermath of the 2011 Fukushima Daiichi nuclear accident. Uncertainty in NPP behavior is key when considering its safety under different operating conditions. Such uncertainty typically results from operation parameters, system conditions, and modeling assumptions. This study integrates the system dynamics (SD) modelling approach with an uncertainty analysis method to quantify the dynamic probabilistic risk in NPPs. To demonstrate the approach’s applicability, the average fuel temperature is used to estimate the probability of reactor core damage under different transients, representing perturbations in reactivity and steam valve coefficient. A Monte Carlo simulation is employed to investigate the effect of uncertainties associated with the different model parameters. A global sensitivity analysis demonstrates that the total delayed neutron fraction, the heat transfer coefficient from fuel to coolant, the coolant temperature coefficient of reactivity, and the fuel temperature coefficient of reactivity are the primary controllers of the plant response variability under the transients considered. In summary, the integration of SD modelling and uncertainty analysis presents an effective DPRA approach that overcomes the limitations of static counterparts while minimizing the computational resources required.

Investigation on Neutronic Parameters of the KLT-40S Reactor Core with U3Si2-FeCrAl using SCALE Code

From a safety point of view, the fuel-cladding of the current design of the KLT-40S reactor still carries a potential risk in the event of a loss-of-coolant accident (LOCA) allowing the formation of hydrogen gas. The concept of accident tolerant fuels (ATF) offers a variety of new safer fuel-cladding materials, one of which is U3Si2-FeCrAl, a potential fuel-cladding combination according to various research sources. In this research, a study of neutronic parameters (1) cycle length, (2) reactivity feedback coefficient, and (3) reactor proliferation resistance was performed with ATF material U3Si2-FeCrAl as fuel-cladding in the KLT-40S reactor core. Modeling and simulation of the ATF-fueled KLT-40S reactor core were performed using KENO-VI and TRITON modules from SCALE code. The results showed that replacement of the fuel-cladding material with the ATF material in the KLT-40S reactor resulted in a shorter cycle length, and the enrichment required to reproduce the original cycle length was above the safeguard limit. The fuel temperature, moderator temperature, and void reactivity coefficient were negative, although not as negative as the original ones. The spent fuel produced at the end of the cycle had good proliferation resistance, although not as good as the original one.

Feasibility of Using Poisons to Suppress the Positive Temperature Reactivity Coefficient in Hydride-Moderated Reactors

Metal hydrides are being seriously considered for advanced nuclear reactor or microreactor applications due to their solid physical state and high hydrogen density. Using hydrides for autonomous applications poses several research and development challenges, one of which relates to neutron upscattering in the thermal energy regime. These hydrides, including zirconium hydride and yttrium hydride, result in a positive temperature coefficient of reactivity for several advanced reactor designs. In this study, we consider one such design that exhibits positive feedback from metal hydrides and thoroughly investigate the neutronic aspects of the core. Temperature reactivity coefficients for four fuels and two hydride moderator configurations are studied, and the total temperature coefficients are found to be positive for all designs, showing that this issue cannot be resolved simply by material variations. Accordingly, five epi-thermal absorbers were evaluated to demonstrate the feasibility of the excess positive feedback suppression in the core instigating from neutron energy spectrum shift. Following which, two promising burnable poison candidates are selected to investigate further throughout the core discharge. Promising results are shown for this core design, which can be extended to other hydride-moderated remote special-purpose reactor designs.

State of the Art on Doppler Broadening: Modern Developments on the Fuel Temperature Coefficient

The Doppler coefficient represents the primary source of passive and instantaneous negative reactivity feedback to limit peak power excursion during reactivity-initiated accidents as well as a nonnegligible negative reactivity source that changes between cold zero-power and hot zero-power conditions. Furthermore, the mechanism behind the Doppler coefficient may also contribute to an increase in the buildup of Pu under normal operating conditions. As such, its treatment is critical in the design and evaluation of the safety and control of nuclear systems. This paper provides a brief overview of the physical source of the Doppler effect through resonance broadening from first principles as well as an exploration of some recent developments in the treatment of elastic scattering in the Monte Carlo codes Tripoli4® and MCNP. This exploration results in a detailed look at the effect different elastic scattering kernels have on the radiative capture, fission, and elastic scattering rates as they directly tie into the calculation of the Doppler coefficient via the six-factor formula. Also provided is some insight into the propagation of the a priori uncertainty of 238U resonance parameters. This work is performed pursuant to the development of a new experimental program to measure the Doppler coefficient in a zero-power reactor both more accurately and to higher temperatures (1500°C to 2000°C) than has been done in the past at the MINERVE facility at Cadarache.

An assessment of scenarios for transuranic burning and 233U production in a small molten chloride fast reactor

Molten Chloride Fast Reactor (MCFR) attracts more and more attention owing to its significantly hard neutron spectrum and high solubility for actinide trichlorides, making it an appealing option for burning transuranic (TRU) elements discharged from pressurized water reactors (PWRs). Meanwhile, in recent years, the development of thorium fuel cycle has aroused considerable interest worldwide with the purpose of saving uranium reserves and further reducing the production of long-lived minor actinides. A small MCFR incorporating TRU burning and 233U production is proposed in this work in order to evaluate its incineration and production capabilities. By employing an in-house developed tool based on SCALE 6.1, the fuel cycle features of the small MCFR with three fuel loading scenarios, including pure TRU (scenario 1), TRU plus depleted uranium (scenario 2) and TRU plus thorium (scenario 3), are analyzed. The evolution behaviors of main nuclides, reactivity worth, neutron spectra and temperature reactivity coefficient (TRC) are discussed in detail. It is found that the MCFR can burn TRU effectively and the transmutation rates for the three scenarios are 930 kg/GWe/year, 375 kg/GWe/year and 768 kg/GWe/year, respectively, achieving a maximum capability for TRU burning of about 3.7 times the annual TRU production in a 1-GWe PWR with 50 MWd/kgU discharge burnup. Meanwhile, scenario 3 has the highest 233U production capability of about 110.3 kg per year, although it does not offer processing advantages over scenario 1 and scenario 2. The analysis of TRC indicates that the fuel salt TRC and the total core TRC are both strongly negative for all three scenarios owing to the thermal expansion of liquid fuel salt. Finally, a nuclear energy system consisting of both fast and thermal molten salt reactors is proposed and evaluated to establish a closed nuclear fuel cycle with very low radioactive wastes.

Minor Actinides Transmutation and 233U Breeding in a Closed Th-U Cycle Based on Molten Chloride Salt Fast Reactor

Long-lived minor actinides (MAs) are one of the primary contributors to the long-term radiological hazards of nuclear waste, and the buildup of MAs is hampering the development of nuclear power. The transmutation of MAs in reactors is regarded as a potential way to replace direct disposal to reduce the impact of MA on the environment and improve the utilization of fuel. Due to its superior features, such as outstanding neutron economy, no fuel assembly fabrication, high neutron flux, and especially online refueling and reprocessing, the molten chloride salt fast reactor (MCFR) is regarded as one of the potential reactors for MA incineration. In this work, MA transmutation capability and 233U breeding performance for an optimized MCFR have been evaluated in different scenarios. The results show that the MA transmutation capability and 233U breeding performance with online transuranic elements (TRU) and 232Th feeding scenario are improved significantly compared with the case in online 233U and 232Th feeding, when the initial MA loading is 5 mol%, the total mass of MA transmutation and MA incineration is 7160 kg and 1759 kg during the whole 100 years operation under online TRU and 232Th feeding scenario, and the corresponding average annual net production of 233U is 450 kg, however, the MA transmutation amount, MA incineration amount and average annual net production of 233U for online 233U and 232Th feeding scenario is 5298 kg, 1315 kg, and 249 kg, respectively. In addition, the research also shows that the increase in initial loading of MA has no obvious effect on the improvement of the 233U breeding performance but can improve the transmutation efficiency of MA under online TRU and 232Th feeding scenarios. Furthermore, if 233U is continuously extracted online from the core during the operation, the 233U breeding performance will be significantly improved, but it will deteriorate the safety performance, such as the fuel temperature coefficient of reactivity (TCR) and the effective delayed neutron fraction (EDNF), more importantly, it will also put forward higher requirements for the immature online reprocessing technology.

Reactor physics calculation and measurement during first approach to criticality of Apsara-U research reactor

Apsara-U is 2 MW swimming pool type research reactor with low enriched uranium fuelled core. On 10th September 2018, Apsara-U was made critical at 18:41 hrs. After the First Approach to Criticality (FAC) of Apsara-U, low power physics experiments were carried out. The experiments included reactivity worth measurement of all reactivity devices and measurement of the isothermal temperature coefficient of reactivity. Inflated reactivity worth of Shut off Rods as calculated from the power from some of the neutronic channels were obtained on account of shadowing effect of neutron reaching the detector locations. Further the isothermal coefficient of reactivity was observed positive. In this paper, we discuss the mentioned Reactor Physics measurement and related theoretical calculation during FAC that includes solving the benchmark problem for validating the calculation tools, and steady state thermal hydraulic analysis, Reactivity initiated transients for the FAC core configuration for ascertaining the safety of the reactor.

Alternatives to the burnup strategy and fuel material of the EM2 modular reactor

The Energy Multiplier Module (EM2) is a modular high-temperature helium-cooled fast reactor developed by General Atomics (GA). It contemplates a convert-and-burn core, in which fertile fuel is converted to fissile fuel for on-site burning. These features allow the reactor to achieve a core lifetime of 32 years without refueling with different types of fuel materials, like thorium, natural uranium, depleted uranium (DU), and spent fuel. The objective of this study is to evaluate the neutronic performance of the EM2 reactor under four types of fuel material combinations (Low enriched uranium [LEU]/DU, LEU/Th, U-Pu/DU and U/Th) with two burnup strategies. The first strategy was taken from the GA model (configuration 1) and the second was proposed in this paper (configuration 2). The Serpent Monte Carlo code with ENDF/B-VII cross section library was used to perform the calculations. The following parameters were analyzed for each scenario: the effective neutron multiplication factor (keff), the conversion ratio, reactivity coefficients, and the effective delayed neutron fraction (βeff). The fissile inventory and fission rate contribution of the main isotopes were also discussed. The results obtained show that the scenarios with configuration 2 had a higher peak excess reactivity than those with configuration 1. For the scenarios with DU as a fertile fuel, 238U produced ∼20 % of the fission throughout the fuel cycle. In contrast, in the scenarios with thorium as fertile fuel, the 232Th fission rate was negligible. As for the reactivity coefficients, all scenarios presented negative Doppler coefficient and positive void reactivity; both values increased along burnup. On the other hand, the βeff decreased with burnup on and did not vary significantly between both configurations.

Minor Actinides Transmutation Performance in a Closed Th–U Cycle Based on Molten Chloride Salt Fast Reactor

The molten chloride salt fast reactor (MCFR) with a closed Th-U fuel cycle is receiving more and more attention due to its excellent performance, such as high solubility of actinides, superior breeding capacity, and good inherent safety. In this work, the neutronics performances for different minor actinides (MA) loadings and operation modes are analyzed and discussed based on an optimized MCFR. The results indicate that online continuous reprocessing can significantly increase the transmutation performance of MAs. In addition, MA loadings have an obvious effect on the neutronics characteristics of the MCFR, and it is helpful for improving the MA transmutation capability and 233U breeding performance, simultaneously. When MA = 5 mol%, the average annual MA transmutation mass and incineration mass can achieve about 53 kg and 13 kg, respectively, and the corresponding annual net production of 233U is 250 kg. When MA = 33.5 mol%, the annual MA transmutation mass and incineration mass can be about 310 kg and 77 kg, respectively, and the corresponding annual net production of 233U is 349 kg. However, when the MA loadings exceed 10%, the corresponding keff will exceed 1.1 for decades, even if only Th is continuously fed online. The results also indicate that the transmutation ratio (TR) and incineration ratio (IR) of MA increase and reach maximum values in the first decades for all the different MA loadings, which means MA may be fed into the fuel salt to improve its transmutation capability. Moreover, though MA loading will increase the level of radiotoxicity of the core in the early stage of burnup, the radiotoxicity of MA will drop rapidly after a brief rise during the operation. It can also be found that the temperature coefficient of reactivity (TCR) of all different MA loadings can be negative enough to maintain the safety of the MCFR during the whole operation, although it decreases in the beginning of life (BOL) with the increasing MA loading. Furthermore, the evolution of an effective delayed neutron fraction (EDNF) is also researched and discussed, and the EDNF varies most significantly when loading MA = 35.5 mol%, with a range of 273 to 310 pcm over the entire 100 years of operation.

Neutronics simulation of China Experimental Fast Reactor start-up tests using FARCOB and ERANOS 2.1 code systems

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Prevalence and factors associated with sarcopenia in a cohort of patients with chronic pancreatitis

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Conceptual design of a 3 MWth yttrium hydride moderated heat pipe cooled micro reactor

Micro reactors are gaining substantial interest worldwide due to their capabilities to deliver power to remote and decentralized areas and their long fuel cycles, making them more attractive than other energy sources. This paper presents a conceptual design and steady-state neutronics and thermal-hydraulics analyses results for a heat pipe cooled yttrium hydride moderated micro reactor. The reactor was designed for both terrestrial and space application purposes; hence, size and weight were important. The neutronics analysis calculations were performed to evaluate key safety parameters such as control devices reactivity worth, temperature reactivity coefficients, and power distribution of the reactor core. Burnup calculations were performed using Monte Carlo code MCNP6.2 and verified against Serpent code results. The thermal-hydraulics analyses calculations were performed to evaluate the temperature distribution in a representative unit cell in the reactor core using the STAR CCM+ multi-physics computational fluid dynamics (CFD) software by utilizing the power distribution obtained from neutronics calculations. It was demonstrated that this reactor could be operated safely for more than 11 years, generating 3 MW of thermal power.

Liquid-salt-cooled Reactor Start-up with Natural Circulation

The Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR) was modeled using the neutronics analysis code SCALE6.0 and the thermal-hydraulics and kinetics modeling code RELAP5-3D with objective to devise, analyze, and evaluate the feasibility and stability of a start-up procedure for this reactor using natural circulation of the coolant and under the Loss Of Offsite Power (LOOP) conditions. This Generation IV reactor design has been initially developed by Oak Ridge National Laboratory and studied by researchers worldwide for almost a decade. While neutronics and thermal-hydraulics analyses have been previously performed to show the performance of the reactor during normal operation and for shutdown scenarios, no study has heretofore been published to examine the active or passive start-up of the reactor. The fuel temperature (Doppler) and coolant density coefficient of reactivity of the LSVHTR were examined using the CSAS6 module of the SCALE6.0 code. Negative Doppler and coolant density feedback coefficients were calculated. Two initial RELAP5 simulations were performed to obtain the steady-state conditions of the model and to predict the changes of the thermal-hydraulic parameters during the shutdown of the reactor. Next, a series of step reactivity additions to the core were simulated to determine how much reactivity can be inserted without jeopardizing the safety and stability of the core. Finally, a start-up procedure was developed, and the restart of the reactor with natural convection of the coolant was simulated. The results of the simulations demonstrated the potential for natural circulation start-up of the LS-VHTR.

Full Core Criticality Modeling of Gas-Cooled Fast Reactor using the SCALE6.0 and MCNP5 Code Packages

The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals as the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working environment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 with the results of the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analyzed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be useful technique to monitor both fission source convergence (stationarity) and core eigenvalue convergence (keff) to fundamental eigenmode with MCNP5. All calculations were done with ENDF/B-VII.0 library. The obtained results showed high similarity with reference results.

Neutronic and fuel cycle performance of LEU fuel with different means of excess reactivity control: Impact of neutron leakage and refueling scheme

Low enriched uranium (LEU) fuel with different Integral Burnable Absorbers (IBAs) and Burnable Poison Rods (BPRs) in AP1000 assembly was investigated. These include IBAs of Gd2O3, Er2O3, AmO2, PaO2, and BPRs of Al2O3-B4C. The conventional uranium dioxide (UO2) fuel and two accident tolerant fuels (ATFs) namely UN-U3Si2 and U3Si2 were taken as the base fuel materials. Nine fuel compositions of these fuels with and without burnable poison were simulated in the lattice physics code Dragon, version 5. The burnup-dependent infinite multiplication factor was determined and significant variation was found due to the addition of poison. Reactivity was calculated at different levels of neutron leakage (0–5%). More positive reactivity at the early burnup stage and less negative reactivity at end of life were found for PaO2. The Linear Reactivity Model (LRM) was used to calculate the fuel cycle parameters. Cycle burnup and discharge burnup were determined considering different refueling batch number at each leakage scenario. Both parameters are found to decrease with neutron leakage while a large batch number results in a higher discharge burnup. Reduced discharge burnup was observed for each IBA and BPR except PaO2. Cycle length was also calculated for each of the composition and cycle length penalty was positive only for PaO2. The fuel temperature coefficient of reactivity, pin power distribution, and energy-dependent neutron flux were determined using the Monte Carlo code OpenMC. The burnup-dependent isotopic compositions were also studied.