•Results of benchmarking TRIPOLI based on the VENUS-2 MOX benchmark are reported.•3-D TRIPOLI calculations were performed using ENDF/B-VI.4, ENDF/B-VII.0 and JEFF-3.•Results are compared with ...measured data and results of other benchmark participants.•TRIPOLI results agree well with experimental data and results from other participants.
The reliability and verification of numerical solutions derived from neutronic codes and the use of nuclear data libraries is a very important issue in nuclear technology. To this purpose, computational benchmarks based on well-defined problems with a complete set of input and a unique solution, are often used. The OECD/NEA VENUS-2 is a widely used MOX benchmark problem for the validation of numerical methods and nuclear data sets. In this paper, the results of benchmarking the TRIPOLI Monte Carlo code based on the VENUS-2 MOX benchmark problem, are reported. 3-D TRIPOLI calculations were performed using the ENDF/B-VI.4, ENDF/B-VII.0 and JEFF-3.1 nuclear data sets. The computational results are compared with measured data, as well as with the results of other benchmark participants. In general the TRIPOLI results agree well with both the experimental data as well as those of other participants.
The safe introduction of Generation IV (Gen IV) reactor concepts into operation will require extensive testing of their components. This must be performed under neutronic conditions representative of ...those expected to prevail inside the new reactor cores when in operation. In a thermal Material Testing Reactor (MTR) such neutronic conditions can be achieved by tailoring the prevailing neutron spectrum with the utilization of a device containing appropriate materials. In this work various materials are investigated as candidate components of a device that will be required in case that a thermal MTR neutron energy spectrum must be locally transformed, so as to imitate Sodium cooled Fast Reactor (SFR). Many nuclides have been examined with respect to only their neutronic behavior, providing thus a pool of neutronically appropriate materials for consideration in further investigation, such as regarding reactor safety and fabrication issues. The nuclides have been studied using the neutronics code TRIPOLI-4.8 while the reflector of the Jules Horowitz Reactor (JHR) was considered as the hosting environment of the transforming device. The results obtained suggest that elements with important inelastic neutron scattering could be chosen at a first level as being able to modify the prevailing neutron spectrum towards the desired direction. The factors which are important for an effective inelastic scatterer comprise density and inelastic microscopic cross section, as well as the energy ranges where inelastic scattering occurs. All the above factors have been separately examined in order to suggest potential device materials, able to locally produce SFR neutron spectrum imitation in a thermal MTR.
The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and ...tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs) by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core.
•ANET is a new neutronics stochastic code.•Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted.•Simulations of thermal, lower epithermal ...and fast neutron fluence rates were performed.•Axial fission rate distributions in standard and MOX fuel pins were computed.
ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement is found with both measurements and independent computations, verifying thus ANET’s ability to successfully simulate important parameters of critical and subcritical systems.
▶ Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. ▶ One-step rod withdrawal/insertion are used for rod worth estimation. ▶ The study showed that when Xe is ...present the rods worth is significantly reduced. ▶ Rod worth variation with burnup depends on rod position in core. ▶ Rod worth obtained with MC code is higher than that obtained from deterministic.
One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.
Knowledge of the efficiency of a control rod to absorb excess reactivity in a nuclear reactor, i.e. knowledge of its reactivity worth, is very important from many points of view. These include the ...analysis and the assessment of the shutdown margin of new core configurations (upgrade, conversion, refuelling, etc.) as well as several operational needs, such as calibration of the control rods, e.g. in case that reactivity insertion experiments are planned. The control rod worth can be assessed either experimentally or theoretically, mainly through the utilization of neutronic codes. In the present work two different theoretical approaches, i.e. a deterministic and a stochastic one are used for the estimation of the integral and the differential worth of two control rods utilized in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system SCALE (modules NITAWL/XSDRNPM) and CITATION is used, while the stochastic one is made using the Monte Carlo code TRIPOLI. Both approaches follow the procedure of reactivity insertion steps and their results are tested against measurements conducted in the reactor. The goal of this work is to examine the capability of a deterministic code system to reliably simulate the worth of a control rod, based also on comparisons with the detailed Monte Carlo simulation, while various options are tested with respect to the deterministic results’ reliability.
Research reactors are used for many applications: material testing; radioisotope production; beam-line applications for material research; nuclear transmutation doping; neutron activation analysis; ...neutron radiography experiments; fuel waste management; and other neutron and nuclear material related quantities, features, and research areas of interest. Each application requires enhanced neutron fluxes in a specific section of the energy spectrum; therefore, appropriate irradiation positions in the core or an appropriate configuration of the beam line need to be chosen. In several cases the required flux exceeds the maximum value that can be obtained in the existing irradiation positions of the operating reactor core, but the desired neutron flux amplification through the reactor power upgrade would require large-scale transformations, high costs, and long shutdown periods. With the creation of a flux trap at a central core position in the open pool Greek Research Reactor (GRR-1), a noticeable local increase of the thermal neutron flux was achieved, compared to the irradiation channels at peripheral core positions. In the present technical note, calculational and measurement results concerning the original core modification are presented, while the possibility of larger sample irradiation at higher thermal neutron flux in the GRR-1 is investigated. The presented results are based on deterministic and stochastic neutronic calculations with numerical models validated using measurements conducted for the original flux trap. The work is completed with a thorough thermal-hydraulic analysis to evaluate the impact of the proposed modifications to reactor operation. The study showed that the flux trap enlargement with complete removal of a central control fuel assembly increases the maximum thermal neutron flux by ~41%, while further removal of the neighboring fuel assembly leads to an average flux increase of ~45%, thus offering capabilities for extended reactor utilization such as additional isotope production.
Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, ...due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for comparison. For this purpose, corresponding reactivity results obtained by detailed geometry/high statistics Monte Carlo calculations, performed by the well documented neutronic code TRIPOLI, were utilized. It was concluded that the assumptions made for the transverse leakages in the one-dimensional cell calculations affect significantly the core reactivity computation. Small modifications in the parameters which determine the transverse cell leakages can induce
k
eff
modifications of the order of 10
3
pcm. Significant factors for the suitability of the adopted assumptions were also shown to be the core size and the inclusion of several components, such as reflector or internal irradiation channels.
•ANET code simulates innovative reactor designs including Accelerator Driven Systems.•Preliminary analysis of thermal hybrid soliton reactor examines breeding capabilities.•Subsequent studies will ...aim at optimizing parameters examined in this analysis.•Breeding capacity could be obtained while preserving efficiency and reactor stability.
Nuclear energy industry asks for an optimized exploitation of available natural resources and a safe operation of reactors. A closed fuel cycle requires the mass of fissile material depleted in a reactor to be equal to or less than the fissile mass produced in the same or in other reactors. In this work, a simple closed cycle scheme is investigated, grounded on the use of a conceptual thermal water-cooled and moderated subcritical hybrid soliton reactor (HSR). The concept is a specific Accelerator Driven System (ADS) operating at lower power than usual pressurized water reactors (PWRs). This type of reactor can be inherently safe, since shutdown is achieved by simply interrupting the accelerator's power supply. In this work a preliminary investigation is attempted concerning the existence of conditions under which the operation of a thermal HSR in breeding regime is possible. For this purpose, a conceptual encapsulated core has been defined by choosing the magnitude of a set of parameters which are important from the neutronic point of view, such as core geometry and fuel composition. Indications of breeding operation regime for thermal HSR systems are sought by performing preliminary simulations of this core. For this purpose, the Monte Carlo code ANET, which is being developed based on the high energy physics code GEANT is utilized, as being capable of simulating particles’ transport and interactions produced, including also simulation of low energy neutrons transport. A simple analytical model is also developed and presented in order to investigate the conditions under which breeding in HSR is possible, which supports the ANET simulation findings.
Assessment of the gamma heating deposited on samples irradiated in material testing reactors is basic safety issue. The GHRRC (Gamma Heating in Research Reactor Cores) code developed in NCSR ...¿Demokritos¿ is used here to estimate the relative importance of the mechanisms contributing to the total gamma heating of the irradiated material, for a core fuelled with Low (LEU) and with High Enrichment Uranium (HEU). The Greek Research Reactor core configuration with a 13% burnup has been assumed for the numerical experiments. The gamma heating of a Fe sample irradiated near the core center is examined. The heating components due to fission gammas, capture gammas and gamma dose build-up in the core are compared with respect to the fuel enrichment. Although individual components are found higher either in the LEU or in the HEU core, the total gamma heating is found comparable for the two fuel enrichments.
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