Sodium Void Reactivity Effect Research Articles

Application of genetic algorithms in optimization of SFR nuclear reactor design

Abstract This work presents a demonstrational application of genetic algorithms (GAs) to solve sample optimization problems in the generation IV nuclear reactor core design. The new software was developed implementing novel GAs, and it was applied to show their capabilities by presenting an example solution of two selected problems to check whether GAs can be used successfully in reactor engineering as an optimization tool. The 3600 MWth oxide core, which was based on the OECD/NEA sodium-cooled fast reactor (SFR) benchmark, was used a reference design [1]. The first problem was the optimization of the fuel isotopic inventory in terms of minimizing the volume share of long-lived actinides, while maximizing the effective neutron multiplication factor. The second task was the optimization of the boron shield distribution around the reactor core to minimize the sodium void reactivity effect (SVRE). Neutron transport and fuel depletion simulations were performed using Monte Carlo neutron transport code SERPENT2. The simulation resulted in an optimized fuel mixture composition for the selected parameters, which demonstrates the functionality of the algorithm. The results show the efficiency and universality of GAs in multidimensional optimization problems in nuclear engineering.

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.

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.

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.

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.

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

Detailed neutronic analysis of a MOX-fueled metal-cooled reactor

With the prominent features of different U-Pu fuel compositions, the MOX-fueled fast spectrum metal-cooled reactors have been of increasing interest for recent years worldwide. Present study invokes a detailed investigation on the calculations for neutronic analysis and it conducts a comparative study for neutron physics parameters of a metal-cooled fast spectrum reactor – the BFS-62-3A. This work is an amalgamation of four tasks. In the first part, a detailed comparative analysis was performed to perform a code-to-code verification using the available results of three Monte Carlo codes including SuperMC, MCNP, and Serpent, and a deterministic one, the DYN3D-MG with the employment of continuous neutron energy cross-sections. The experimental results of BFS-62-3A benchmark were used to assess the potentiality of the aforementioned reactor physics codes. For most of the integral parameters, the SuperMC was found to be on the leading edge. In the second part, the effects of data libraries including ENDF/B-7.1, ENDF/B-7.0, ENDF/B-6.6, JEFF3.2, and HENDL3.0 on the simulations performed using SuperMC code for evaluating k-eff and radial fission rates were all assessed. The third part incorporates the investigation of the fission rates’ deviations in stainless steel radial reflector by changing the density of the reflector. The decrease of density by 5% was found to be in good agreement with the benchmark. In the last part, that has a special importance to the concept pertaining to safety-enhanced Sodium-cooled Fast Reactor (SFR) core, the reactivity of the critical assembly was studied by calculating the sodium void reactivity effect. The simulation results of SuperMC code agreed well with the available experimental and simulation results. The present study has enabled SuperMC code to pass another big milestone on dealing with complex and advanced nuclear systems.

Improvement the value of sodium void reactivity effect of the fast neutron reactor by the instrumentality of the Monte Carlo code

To fulfill safety of fast sodium reactors in a beyond design-basis accident (BDBA) like unprotected loss of flow accident (ULOF), the sodium void reactivity effect (SVRE) should be close to zero. Its value depends on the fuel burnup – the higher burnup the higher value of SVRE. We analyze limitation of the fuel burnup in the core of a large sodium reactor imposed by SVRE.The model of a large sodium-cooled reactor core is chosen for analysis. Two fuel types are considered – MOX and nitride uranium-plutonium. For both we follow the transition of the core from reactor startup to equilibrium reloading state, where the core passes consequently through different stages of fuel burnup. Calculations of maximal and average burnup together with corresponding value of SVRE have been done for a homogeneous model with the codes TRIGEX and MMKKENO. The latter employs transport approximation (the Monte Carlo method) and allows detailed heterogeneous representation of fuel assemblies and safety rods. The results obtained for the MOX fuel show that after the end of second core cycle (maximal burnup about 8%) the refined value of SVRE exceeds two times its maximal acceptable value (0.3% Δk/k). For the nitride fuel this exceeds is found at the end of 3rd fuel reshuffling (maximal burnup is about 8.75%); however it is considerably lower as compared to that found for MOX.

Уточнение величины натриевого пустотного эффекта реактивности в быстрых натриевых реакторах с помощью программ Монте-Карло

Уточнение величины натриевого пустотного эффекта реактивности в быстрых натриевых реакторах с помощью программ Монте-Карло

Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

Inherent safety aspects of metal fuelled FBR

Static and dynamic studies of metal fuelled fast breeder reactors (MFBR) are carried out to verify the passive shutdown capability and its inherent safety parameters. Static calculations are carried out to determine the vested reactivity feedback parameters from the fuel and coolant temperature rise separately. Power reactivity decrement of metal fuel reactor is found to be small as compared to oxide fuel reactor of same size.ULOF analysis of metal (U–Pu–6% Zr) 1000MWe pool type MFBR is studied with a flow halving time of 8s. The study is also made with considering uncertainties on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behaviour of 1000MWe core is benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal System (SGDHRS) capacity before the initiation of boiling. From the study, it is concluded that if the sodium void reactivity is limited (4.6$) then the inherent safety of 1000MWe design is assured, even with 20% uncertainty on the sensitive parameters and also it is found out higher primary pump flow halving time (15s instead of 8s) can avoid cliff edge effects in 1000MWe MFBR transient behaviour.

Studies on Influence of Sodium Void Reactivity Effect on the Concept of the Core and Safety of Advanced Fast Reactor

The paper is devoted to studies on the influence of the sodium void reactivity effect (SVRE) on the safety and technical and economical characteristics of the BN-1200-type reactor. Different core options are considered for application to this reactor. These core options differ in design, dimensions, and, hence, SVRE value. It is shown by the analysis that the most flattened core with sodium plenum at the top assures reactor self-protection under beyond-design-basis accident conditions. Sodium plenum abandonment and core height increase causing an SVRE value increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Self-protection means the possibility to avoid rapid core meltdown under conditions of the above-listed beyond-design accidents. The possibility of controlling beyond-design accidents (for instance, by restoring the power supply of the main pumps in a rather short time) is taken into account. Issues of choosing the optimal core design under these conditions are discussed.

Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500MWe and 1000MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500MWe and 1000MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000MWe design is assured, even with 20% uncertainty on the sensitive parameters.

Studies on Influence of Sodium Void Reactivity Effect on the Concept of the Core and Safety of Advanced Fast Reactor

Studies on Influence of Sodium Void Reactivity Effect on the Concept of the Core and Safety of Advanced Fast Reactor

Innovative Core Design for Generation IV Sodium-Cooled Fast Reactors

The design of innovative cores for Generation IV Sodium-cooled Fast Reactors (SFRs) must chiefly focus on improving safety performance levels and competitiveness, with the aim of achieving all fundamental goals set for Generation IV systems.In this context, one priority is to create a core design such that, in the event of a serious accident, all risks of generalised fuel meltdown are avoided. The core must therefore be optimised to reduce the sodium void reactivity effect whilst acting as a self-converter with respect to the fuel itself.Design options are compared by using the physical sensitivity and parametric studies that were carried out, in order to highlight the most attractive approach for meeting the goals. A very large number of combinations of theses design options are possibleFor a high power EFR-type core (1500 MWe) with Oxide fuel (UO2-PuO2), the goals can be met, with the sodium void effects nearly halved, to a value of around 3 to 4 dollars, by selecting options, such as:- a tight grid assembly design,- introduction of a sodium plenum in the assembly's upper section,- introduction of a moderator material.Use of a denser fuel such as Carbide is shown to be very attractive. Its higher density means that a near-zero Internal Breeding Gain can be achieved. Most importantly, its thermal conductivity, which is significantly higher than that of the oxide fuel allows for optimisation options. Either the overall core volume can be reduced by increasing linear power density, or the margin between fuel operating temperature and melting point can be increased. Some resuls from core simulations are presented.Further studies should focus on a detailed characterisation of the selected cores, in particular involving ULOF and UTOP type accident transient calculations. These analyses alone will be suitable arbitrators to identify the real safety level achieved.

Improved performances of the fast reactor calculational system ERANOS-ERALIB1 due to improved a priori nuclear data and consideration of additional specific integral data

A single consistent scheme of calculational methods and nuclear data called ERANOS-ERALIB1 was produced in 1996 to calculate fast reactor neutronic parameters. It represents a significant improvement on previous schemes such as CARNAVAL-IV, PROPANE and VASCO, each of which were required in order to calculate one specific application. The nuclear data library ERALIB1 has been obtained by a consistent statistical adjustment based on 355 integral data from 71 different systems. The performance of ERALIB1 is excellent, as demonstrated during its validation for which all the k eff SUPER-PHENIX data were reproduced to within 70 pcm. The only restriction on this satisfactory performance is related to the rather poor prediction of the sodium void reactivity effect. This was due to very bad nuclear data for 23Na, and the unsatisfactory methods used to calculate the sensitivity coefficients for the sodium void reactivity variation Δ ρ Na. To improve the performance relative to this point and to enlarge the domain of validation several actions have been undertaken: • a revision of the formalism and algorithms used to calculate the derivatives of Δ ρ Na to the sodium cross section data, • a significant enlargement of the integral data base related to this aspect of the sodium void effect. Compared to the initial data base established in support of ERALIB1, several additional (18) sodium void configurations corresponding to voids of different volumes at different core locations have been studied. In order to broaden the range of application of the improved library, which will be called ERALIB1.A, significant effort has been devoted to additional configurations which have firstly been evaluated, and then if judged suitable, included in the adjustment process. They are related to two specifically targeted experimental programmes: • a study of neutron deep penetration. Several configurations of the JANUS experimental programme (shielding constructed of separate steel, iron and sodium plates) have been analysed. With this complementary information ERALIB 1A becomes applicable for accurate predictions of shielding configurations, • a study of steel reflectors for a fast reactor of the SUPER-PHENIX type. The measurements performed in the MASURCA (the CIRANO experimental programme) and FCA facilities include spectral indices (F25(r)/F25(0),...) at different positions in the reflector. As a consequence of these measurements, important information has been obtained for additional “secondary” structural material isotopes, such as 57Fe, 60Ni and 53Cr. Significant effort has also been devoted to the analysis of 29 β eff experiments. The result of this is an improvement of the uncertainty on ν d( E) which guarantees a prediction of β eff with the required accuracy (3% for critical configurations, and 5% for power reactors). The consistent statistical adjustment method by Gandini et al. (1973) has been completed. Rigorous criteria have been introduced to identify any data which are suspect and/or inconsistent in the integral data base. These data may introduce additional bias in the adjusted library, and for that reason they must be discarded before adjustment.

Reduction of the Sodium Void Reactivity Effect by Using a 99Tc Layer

Analyses were performed by using the U. S. advanced liquid metal reactor (ALMR) core design to determine the feasibility of using it as a 99Tc burner while reducing the sodium void reactivity effec...

A Note on Treatments of Sodium Void Reactivity Effects in a Large Fast Reactor

Fine spectra in large fast reactors of various compositions have been calculated by exactly treating slowing down due to elastic scattering of light elements, and the resulting composition-dependent sets of the effective cross sections have been compared with currently typical standard sets. The following results have been obtained: 1. Effective cross sections depend largely on composition. The dependence is large for groups where predominant resonances exist. A composition-independent set underestimates sodium void effect compared with a composition-dependent set, when variations of fine spectrum are considered. Therefore a composition-independent set like YOM is not adequate, especially for a large reactor.2. In the present standard set, the self-sheilding factors of both elastic removal and elastic scattering cross section are treated under the same principle. This procedure is not valid, especially at the predominant resonances they vary inversely to each other with variations of composition.3. The effective cross section in a blanket region, except close to the interface and boundary, are hardly affected by the buckling assumed, and the effects of buckling on effective multiplication factor as well as the sodium void reactivity effect are small enough to be neglected.

Introductory survey of basic design and fuel-element considerations for large, fast sodium-cooled reactors

Some basic design and fuel element considerations in the development of large, fast breeder power reactors are analyzed on the basis of experimental and calculated results. Nuclear safety and stability, breeding, and economy are three principal requirements for developing large, fast power reactors. For nuclear safety and stability, the Doppler reactivity effect, sodium void (or loss) reactivity effect and structural expansion reactivity effect are analyzed. For breeding, the breeding ratio and gain, and the doubling time of fuel production are presented. Thermal and structural considerations of fuel element design are outlined. For economy, as well as breeding, the development of a ceramic (especially carbide) plutonium fuel system with higher power density and specific power for the large, fast power reactor system is desirable.