Void Reactivity Research Articles

2D and 3D numerical investigations of sodium boiling in sodium cooled fast reactor with MOX fuel and low sodium void reactivity effect during unprotected loss of flow accidents

• Behavior of SFR under ULOF simulated for Low Sodium Void Worth design. • No core melting simulated and stable boiling behavior predicted by both codes. • Different frequencies and amplitudes of boiling regimes with SIMMER and COREMELT. • 3D modeling performed with COREMELT confirms oscillating behavior and absence of core melting. • If confirmed by enough experimental evidence, could make safety demonstration of SFR more robust.

Analysis of OECD/NEA medium 1000 MWth sodium-cooled fast reactor using the Monte Carlo serpent code and ENDF/B-VIII.0 nuclear data library

This study presents a benchmark evaluation of the new ENDF/B-VIII.0 nuclear data library for the Organization for Economic Co-operation and Development/Nuclear Energy Agency Medium 1000 MWth sodium-cooled fast reactor (SFR). The study presented herein covers both SFR core types, i.e., metallic fueled (MET-1000) and oxide fueled (MOX-1000), simulated using the continuous-energy Monte Carlo Serpent2 code. The neutronics performances of the ENDF/B-VIII.0-based simulations were compared mainly to two libraries: ENDF/B-VII.1 and JENDL-4.0. The comparison includes several neutronics parameters evaluated for the beginning and end of the cycle conditions. These parameters include the effective multiplication factor keff, total effective delayed neutron fraction βeff, sodium void reactivity (∆ρNa), Doppler constant (∆ρDoppler), and control rod worth (∆ρCR). In addition, a sensitivity study was used to reveal the major isotope/reaction pairs contributing to the discrepancy observed in the performance of the three libraries using 33 and 44-energy-group structures.

Evolution of the Safety of Fast Neutron Reactors

In its history, fast neutron reactors (BR) have gone through a long evolutionary period. However in general, BRs have gone the way of the evolutionary development. The function of retention radioactivity in case of an accident was performed by BRs safety vessels. For protection against external influences, new BR projects must be constructed containments. In the function of the reactor core cooling, changes have occurred related to the transition of the emergency cooling from the water circuit to the intermediate sodium circuit. Developments in emergency cooling systems are associated with the placement of an emergency heat exchanger inside the reactor, and the use of natural circulation in sodium and air. In the function on reactivity controlling due to the increase in core sizes, difficulties arise with the suppression of positive reactivity caused by sodium void reactivity effect. Therefore, in BR projects, the use of sodium plenum over the fuel part of the core and / or use of heterogeneous cores are needed. With an increase in the core sizes, the risk of jamming the long shifts of emergency rods increases. Therefore, it is advisable to transfer the safety rods to the hydraulic suspension in a sodium stream.

Accounting for Uranium-234 During the Analysis of Experiments at the Research Fast Reactor with High Enrichment

Paper is an attempt to consider the influence of U-234 presence through the comparison of the calculated assessments of the temperature reactivity effect (TRE) and the sodium void reactivity effect (SVRE) of the fast research reactor CEFR with 64.4% enrichment on U-235 isotope and measured values of the temperature reactivity effect and sodium void reactivity effect. Additional investigation was performed to account for U-234 in the first criticality experiment. Experimental data are provided in frames of IAEA Coordinated Research Project “Neutronics Benchmark of CEFR Start-Up Tests”. The content of U-234 was calculated by us using mathematical model of the quasi-ideal enrichment cascade. Due to the high enrichment, the content of U-234, although still small, has ceased to be vanishingly small. The effect caused by this isotope is appeared not only through the presence of new isotope in the fuel load, but also in a decrease in the mass of the U-238 isotope in the fuel load. It is shown, that accounting for U-234 slightly increases the agreement of the calculated assessments of TRE among themselves (values of deviation shifted toward each other)/Also accounting for the U-234 isotope leads to a small but systematic increase in the values of the effective multiplication factor by 120-150 pcm.

Neutronic evaluation of CANDU-6 core using reprocessed fuels

The spent fuel from a PWR still contains some amount of fissile materials depending on their initial enrichment and the burnup. Thus, spent fuel from PWRs containing about 1.5% of fissile material could be used as fuel for CANDU reactors after some fission products are removed from it. Thus, an important proposal is the DUPIC cycle, where spent fuels from a PWR are packaged into a CANDU fuel bundle using mechanical reprocessing but without the need of chemical reprocessing. When it is refueled with reprocessed fuel, the reactivity of the system increases, and this behavior may affect the safety parameters of the reactor. Therefore, this work studies the neutronic parameters of two reprocessing fuel techniques: AIROX and OREOX, which are evaluated for two different cores configuration. The first one considers heavy water as a moderator and coolant. The second one considers heavy water and light water as moderator and coolant respectively. These studies evaluate the core behavior based on the different number of reprocessed fuels channels and compare them with the reference core. To perform the simulation the MCNPX was used to calculate the effective multiplication factor, fuel temperature coefficient of reactivity, void reactivity coefficient, and neutron flux, which were evaluated at steady state condition for the different cases. The results show that the presence of parasitic absorbers in the reprocessed fuels hardens the neutron spectrum. This behavior provokes an increase in the core reactivity, in the fuel temperature coefficient and in the void reactivity coefficient. Among these parameters, the use of light water reduces the core reactivity but do not improve the fuel temperature coefficient and the void reactivity coefficient.

Sensitivity and uncertainty analysis for the UAM-SFR sub-exercises with linear regression from random sampling

The paper presents solutions for the neutron transport sub-exercises regarding pin cells, fuel assemblies, and supercells of the OECD/NEA UAM-SFR Benchmark. For this, the GRS XSUSA methodology is applied with a recently introduced approach performing linear regression analysis on the output samples, denoted as XSUSA(LR). The main purpose for this is to calculate sensitivity profiles and the main contributions to the total output uncertainties via the first-order uncertainty propagation formula, i.e. by multiplying the corresponding total/partial nuclear data covariance matrices with the total/partial sensitivity vectors. Confidence intervals were estimated by applying Jackknife resampling. Implicit effects are considered. All output uncertainties are compared to the corresponding values calculated with deterministic linear perturbation theory using TSUNAMI from SCALE 6.2.3. Various optimizations were applied with respect to nuclear data and problem geometry. The output uncertainties are significant – around 1.5% for multiplication factors and 5% for Doppler and sodium void reactivity effects.

ЭКСПЕРИМЕНТАЛЬНЫЕ И РАСЧЕТНЫЕ ИССЛЕДОВАНИЯ ТЕПЛООБМЕНА ПРИ КИПЕНИИ ЖИДКИХ МЕТАЛЛОВ НА МОДЕЛЯХ ТВС БЫСТРЫХ РЕАКТОРАХ В АВАРИЙНЫХ РЕЖИМАХ

Studies of the boiling of alkali metals show that, compared with boiling water, the process of boiling liquid metals has significant features. Cooling of a fuel assembly in accident conditions when accident protection is triggered and circulation pumps are switched off (ULOF) leads to the study of fuel elements cooling at reduced coolant flow rates or even tipping circulation in fuel assemblies. There is only limited data on the boiling of sodium in fuel assemblies in these regimes. The results of a series of experiments on heat transfer and stability of circulation during boiling of a sodiumpotassium alloy on single fuel assembly models and in a parallel fuel assembly system with natural coolant circulation performed at JSC “SSC RF - IPPE” are presented. The results of comparing the data of the calculated and experimental studies are presented. The influence of the surface roughness of the fuel rods on the heat transfer and flow regimes during boiling of a liquid metal in bundles is demonstrated. The results of experimental studies of heat transfer during boiling of sodium in natural and forced convection regimes in a fuel assembly model with a “sodium cavity” located above the reactor core intended to compensate for the positive sodium void reactivity effect in emergency situations with sodium boiling are also presented. It is shown that it is possible to provide continuous sodium cooling of fuel element simulators in fuel assemblies under these conditions. The results of the generalization of data on heat transfer during boiling of liquid metals in bundles and a cartogram of the regimes of two-phase flow during boiling of liquid metals in bundles are presented. The objectives of further research are discussed.

INVESTIGATION ON INHERENT SAFETY OF ONE FLUID-MOLTEN SALT REACTOR (OF-MSR) WITH VARIOUS STARTING FUEL

Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

Basic design analysis of a heavy water-cooled thorium breeder reactor

Core design evaluation of heavy water cooled thorium breeding reactor has been investigated based on optimum basic design criteria such as fuel breeding capability and negative void reactivity coefficient value by adopting triangular fuel lattice and hexagonal core arrangement. Core burnup calculation and some thermal hydraulic parameters have been analyzed to evaluate the reactor core performances and to confirm the feasible design region based on the optimum results. Core burnup evaluation has confirmed the feasibility of heavy water-cooled thorium breeder reactor with negative void reactivity. Three batches refueling scheme is employed for 23-month refueling period. This system shows the breeding condition at the end of cycle for average core burnup of 38 GWd/t and it always gives negative void reactivity coefficient during reactor operation. In relation to thermal hydraulic analysis, the system achieves the maximum linear heat rate of 18.2 kW/m, maximum fuel temperature of 1066 °C and friction pressure drop of 0.046 MPa. The reactor has large margins due to the limitation of thermal hydraulic design point of view and some comparable result with the conventional reactor based on the obtained thermal hydraulic properties.

Use of similarity indexes to identify spatial correlations of sodium void reactivity coefficients

ABSTRACT The safety level of Sodium Fast Reactors is directly related with the sodium void reactivity. A low-void effect design has been proposed within the Horizon2020 ESFR-SMART project thanks to the introduction of a sodium plenum above the active core. In order to assess the impact of this core conception on transient analysis, a map with the spatial distribution of sodium void worth can be computed and fed into a point-kinetics-based transient code. Due to the spatial correlations between neighboring zones, the global effect of voiding two different axial or radial regions is not necessarily the sum of both individual contributions. Neglecting those correlations in the void worth map and consequently in the transient analysis may lead to an unrealistic prediction of the transient sequences. In this work, a method based on sensitivity analysis and similarity assessment is proposed for predicting those correlations. The method proved to be able to establish correlations between axial slices of a sub-assembly and was checked against realistic sodium void propagation patterns.

Evaluation of coolant density history effect in RBMK type fuel modelling

The axial heterogeneous void distribution in a fuel channel is a relevant and important issue during nuclear reactor analysis for LWR, especially for boiling water channel-type reactors. Variation of the coolant density in fuel channel has an effect on the neutron spectrum that will in turn have an impact on the values of absolute reactivity, the void reactivity coefficient, and the fuel isotopic compositions during irradiation. This effect is referring to as the history effect in light water reactor calculations. As the void reactivity effect is positive in RBMK type reactors, the underestimation of water density heterogeneity in 3D reactor core numerical calculations could cause an uncertainty during assessment of safe operation of nuclear reactor. Thus, this issue is analysed with different cross-section libraries which were generated with WIMS8 code at different reference water densities. The libraries were applied in single fuel model of the nodal code of QUABOX-CUBBOX/HYCA. The thermohydraulic part of HYCA allowed to simulate axial water distribution along fuel assembly model and to estimate water density history effect for RBMK type fuel.

Conceptual design of Super FR for MA transmutation with axially heterogeneous core

Supercritical water cooled Fast Reactor (Super FR) is featured with the large coolant density reduction by almost 1/10 from the core inlet to the outlet. Since Minor Actinides (MAs) not only have large neutron capture cross sections for thermal neutrons, but also can fission with fast neutrons, MA transmutation performance of Super FR may greatly depend on MA loading positions in the core axial direction of Super FR.However, such investigations have never been conducted. Hence, this study aims to design a Super FR transmutation core concept with the axial configuration of multiple layers of Mixed Oxide (MOX) and blanket fuels, with a focus on the influence of the large axial coolant density change on MA transmutation and core characteristics.With the design criteria of the negative Void Reactivity Coefficient (VRC), the Maximum Cladding Surface Temperature (MCST) < 650 ℃ and the Maximum Linear Heat Generation Rate (MLHGR) < 39 kW/m, three-dimensional neutronics and thermal-hydraulics coupled core burnup calculations have been carried out. Assembly-wise coolant flow rate distribution is determined to attain high core average outlet temperature and the core characteristics of different designs have been evaluated for the equilibrium core after the cores have reached equilibrium states with given fuel shuffling schemes. It has been shown that the MA transmutation amount is limited by deterioration of VRC due to increase of Pu enrichment for compensating the reactivity penalty by MA loading. Moreover, such influence has been found to be more significant in the lower region of the core, where the coolant density is relatively high. Hence, the core design with MA loading to the upper MOX layer is favorable for improving the MA transmutation performance. However, the trade-off relationship between the MA transmutation amount and thermal-hydraulics performance (increase of MLHGR and decrease of average outlet temperature) has been revealed. To overcome the issue, the core radial zoning has been applied and it has been found effective to suppress the trade-off relationship.

Neutronic Analysis of Modular Gas-cooled Fast Reactor for 5-25% of Plutonium Fuel using Parallelization MCNP6 Code

Modular Gas-cooled Fast Reactor (GFR) is one of six advanced reactor concepts set by the generation IV international forum. Modular GFR has the potential for use actinide recycling and closed fuel cycle as well as applying fast reactor, using helium gas as the main coolant, high working temperature and low void reactivity effect. The neutronic analysis of nuclear reactor means behavior study of subatomic particles that interact with matter. In this paper, the feasibility of plutonium fuel in modular Gas-cooled Fast Reactor (GFR) was investigated. The Monte Carlo method has advantages in full-scale and heterogeneous three-dimensional (3D) geometry modeling using Evaluated Nuclear Data File (ENDF/B-VIII.b5) nuclear data but requires a highly computation time. Since the progress of high performance computing, the reactor physicist community began proposing to use Monte Carlo method for nuclear reactor simulation through the parallelization of calculations. The GFR feasibility design study will carried out with plutonium fuel as fuel cycle inputs with 5-25% of fissile contain. The most important neutronic parameters characterizing of GFR core are determined for beginning of life (BOL) and during burnup calculation conditions. The results of calculation series in parallel computing give the good agreement will be faster of calculation time when more threads. Materials of (U-Pu)C and (U-Pu)N fuel are the good candidates to be chosen in GFR research that give keff more than 1.2 in fissile contain 20%Pu. The variation fissile contain gives the linearity with keff. In depletion simulation, the core reactor still in critical during 20 years operation, burn up values linear with operation time and mass evolution of plutonium and uranium from start-up core to equilibrium core.

Extension and preliminary validation of the Polaris lattice physics code for CANDU analysis

Polaris is a 2D lattice physics code available for the study of light water reactors (LWRs). It includes a method of characteristics (MoC) transport solver and the embedded self-shielding method (ESSM). Its advantages include a simplified user input structure, quick performance, and integration into the SCALE package, utilizing its nuclear data libraries, along with other codes such as the ORIGEN depletion code and the Sampler stochastic uncertainty analysis sequence. This work extends the Polaris code to support CANDU lattice transport calculations and performs some preliminary verification and validation, comparing the code to NEWT and KENO, along with to benchmark experimental data. Heavy-water specific self-shielding factors were also compared to those distributed with SCALE. In general, some differences are observed, but these are generally comparable to other code-to-code differences and smaller than nuclear data uncertainties. The generation of self-shielding factors has a noticeable effect, particularly on the benchmark coolant void reactivity calculation. Heavy-water specific self-shielding factors, while not required, can potentially improve the accuracy of ESSM-based calculations, but the choice of parameters can significantly affect the results. Overall, the Polaris transport and self-shielding calculations should be suitable for CANDU analysis.

THE EFFECTS OF TYPICAL THORIUM IMPURITIES ON THORIUM-BASED NUCLEAR FUELS

Natural thorium contains impurities of numerous isotopes. A study was performed to examine the influence of naturally occurring impurities in thorium-based fuels on a few parameters of interest such as: exit burnup, coolant void reactivity (CVR), fuel temperature coefficients (FTC), and the radiotoxicity of the used fuel. Two different fuel bundle designs were modeled: a 43-element bundle containing 2.25% U-233, and a 35-element bundle containing 1.45% U-233. Naturally occurring thorium fuel impurities were applied to both fuel bundle models at various concentrations, from 0% to 100% of the expected maximum. For burnup-averaged k-infinity (k∞) values of 1.050 and 1.030, exit burnup, burnup-weighted CVR, and burnup-weighted FTC were calculated using the neutron transport code WIMS-AECL, and plotted against fraction of full impurity concentration to determine how the impurity levels affect these reactor physics parameters of interest. For the most-realistic (for CANDU) burnup-averaged k∞ of 1.050, both the inhalation radiotoxicity and the production of U-232 were calculated using the fuel depletion code WOBI. Up to the maximum impurity concentrations considered, no effects on the investigated fuel performance parameters were found to be greater than a few percent.

Benchmarking the new ENDF/B-VIII.0 nuclear data library for OECD/NEA medium 1000 MWth sodium-cooled fast reactor

This paper presents the benchmark evaluation of the new ENDF/B-VIII.0 nuclear library for the OECD/NEA Medium 1000 MWth Sodium-cooled Fast Reactor (SFR). There are 2 SFR cores: metallic fueled (MET-1000) and oxide fueled (MOX-1000). The continuous-energy Monte Carlo Serpent2 code was used as the calculation tool. Various nuclear libraries such as ENDF/B-VII.1 and JENDL-4.0 were included to be compared with the newest ENDF/B-VIII.0. The evaluated parameters are k,βeff, sodium void reactivity (∆ρNa), Doppler constant (∆ρDoppler), and control rod worth (∆ρCR).

Basic design parameter optimization on water cooled thorium breeder reactor

Basic study on optimization of heavy water cooled thorium breeding reactor has been investigated to estimate the feasible design region of breeding and negative void reactivity coefficient. Those parametric surveys are moderator to fuel ratio, fuel pin diameter, fuel pellet power density, and void faction. A fuel pin diameter of 14.5 mm shows the optimum feasible fuel breeding and less required U-233 fissile content. The systems require the enrichment less than 8% and it gives some possible area of breeding. The effect of moderator is essential to make the negative void reactivity. Void fraction effect is still good enough to make more negative void reactivity, however, for no more heavy water as moderator, it obtains less negative void reactivity or becomes slightly positive. It has been shown that several significant parameters affect to the feasibility breeding windows and potential negative void reactivity based on heavy water-cooled thorium reactor.

Heterogeneous effects in simulating a fast nuclear reactor on the BFS test facility

Simulating fast neutron reactor cores for comparing experimental and calculated data on the reactor neutronics characteristics is performed using zero power test stands. The BFS test facilities in operation in Russia (Obninsk) are discussed in the present paper. The geometrical arrangement of materials in the cores of the simulated reactors (fuel pins, fuel assemblies, coolant geometry) differs from the simulation assembly on the BFS. This can cause differences between the experimental results obtained at the BFS and theoretical calculations even in the case when homogenized concentrations of all materials of the reactor are thoroughly observed. The resulting differences in neutronics parameters due to the geometry of arrangement of materials with the same homogeneous concentrations are referred to as the heterogeneous effect. Heterogeneous effects tend to increase with increasing reactor power and its size, mainly due to changes in the neutron spectra. Calculations of a number of functional values were carried out for assessing the heterogeneous effects for different spatial arrangements of the reactor materials. The calculations were performed for the following cases: a) heterogeneous distribution of materials in accordance with the design of a fast reactor; b) heterogeneous arrangement of materials in accordance with the capabilities and design features of the BFS test facility; c) homogeneous representation of materials in the reactor core and breeding blankets. The configuration of materials in accordance with the design data for fast reactors of the BN-1200 type was accepted as the basic calculation option, relative to which the effect called the heterogeneous shift of the functional value (HSF) was calculated. The effect of neutron leakage on the HSF obtained as the result of calculations using different boundary conditions was estimated. All calculations were carried out for the same homogeneous concentrations of all materials for all the above three configurations. Calculations were carried out as well for the case when plutonium metal fuel was used in the BFS. The values of the following functionals were calculated for different cases of arrangement of materials: the effective multiplication factor (reactivity), the sodium void reactivity effect, the average energy of fission-inducing neutrons, and the ratios of radioactive capture cross-sections to fission cross-sections for 239Pu. The calculations were performed using the Serpent 2.1.30 (VTT, Finland) Monte Carlo software package for neutronics simulations and ENDF/B-VII.0 and JEFF-3.1.1 evaluated nuclear data libraries. The effects of various options of material arrangement on the values of keff were found to be the greatest (about 1.6%) for the case when fissile material in the form of dioxide is replaced with metal fissile material. Homogenization of the composition reduces the keff value by about 0.4%. The average energy of fission-inducing neutrons depends to a significant extent on the leakage of neutrons and the presence of sodium (the average energy of neutrons increases and reaches in the presence of sodium about 100 keV, that is, it increases by about 11–13%). Replacing fissile material metal with its dioxide in the BFS test facility (while maintaining homogeneous concentrations, including that of oxygen) allows reducing the average energy of fission-inducing neutrons by about 60 keV. The highest values of HSF, reaching 65%, are observed when calculation of sodium void reactivity effect is performed with materials distributed homogeneously; however, HSF is equal to 1.5% when calculation of the reactor mock-up assembled on the BFS is performed. In the absence of neutron leakage (infinitely extended medium), the sodium void reactivity effect becomes positive and the HSF is equal to 4–7%. The heterogeneous effect of α for 239Pu noticeably (6–8%) depends only on the replacement of metallic plutonium with its dioxide (maintaining, of course, the homogeneous concentrations).

ИССЛЕДОВАНИЕ СПОСОБОВ ПОВЫШЕНИЯ КВ В БЫСТРЫХ РЕАКТОРАХ С МОХ-ТОПЛИВОМ СТРЕМЯСЬ К МИНИМАЛЬНЫМ ЗНАЧЕНИЯМ НПЭР

ИССЛЕДОВАНИЕ СПОСОБОВ ПОВЫШЕНИЯ КВ В БЫСТРЫХ РЕАКТОРАХ С МОХ-ТОПЛИВОМ СТРЕМЯСЬ К МИНИМАЛЬНЫМ ЗНАЧЕНИЯМ НПЭР

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.