Demonstration Fast Reactor Research Articles

Gas entrainment in surge tank of CFR600

The Surge Tank in the secondary sodium circuit of China Demonstration Fast Reactor(CFR600)is filled with sodium and argon. The entrainment of gas in the surge tank will be occurred if the operation conditions is not appropriate. Through theoretical analysis and experimental studying it was found that the mechanisms of gas entrainment for surge tank were focused on three points: (1)the vortex of the free surface, (2)the design of the inlet and outlet nozzle, (3)the design of the route for suction- flow. A water experiment was performed using a 1/10 scale model by geometric similarity. It was observed from the experiment that the minimum flow rate required to gas entrainment occurred was about 130% by standard running condition. The experimental critical Froude number is 4.9, when the design value is 2.5. Through the test of the water model we can see that vortex at the free surface of sodium-argon grow up along with the ever-increasing flow velocity. Entrainment of gas was occurred when the intensity of vortex was enough. The results of the experiment confirmed that the vortex is the major factor for gas entrainment of surge tank of CFR600. But anti-vortex baffle can delay the forming process of vortex effective, and raise the critical velocity substantially.

Design, testing, and installation of a shrouded aerosol sampling probe at demonstration fast reactor fuel reprocessing plant

Abstract A shrouded air sampling probe was designed and fabricated to meet the effluent monitoring requirements of an upcoming reprocessing facility at IGCAR. Performance evaluation of the probe was carried out using standard polydisperse aerosols up to 2 μm at the high-efficiency particulate air filter test facility. Its performance was also verified for higher size particles till 10 μm using the “Deposition 2001a” code. Based on the results, the shrouded probe was found to qualify the air sampling requirements and the same was fixed at the sampling point in the exhaust duct. This paper describes the details of design, testing, and installation of the shrouded probe.

Studies and cross-comparisons of severe accident prevention and mitigation capabilities of a SFR and a GFR

CEA has developed Generation IV fast neutron reactors with gas and sodium coolants in the two last tenth of years. Namely these reactor projects were the GFR2400 (gas-cooled fast reactor of 2400 thermal megawatts) and ASTRID (sodium-cooled fast reactor demonstrator of 1500 thermal megawatts). The objective of this paper is to provide a cross-comparison of the severe accident prevention and mitigation capability of these reactor concepts based on the work done during a significant time period and taking benefit of the distance since these studies. This comparison can highlight both generic trends resulting from a common study approach used for both concepts and from very detailed results obtained during their conceptual design studies. Despite their power difference and their fuel element design specificities, the main study results and conclusions are quite generical of the two concepts as explained in the paper. Thus they could be extrapolated to other reactor design providing main reactor features are kept (core materials, coolant, neutron spectrum, etc.). The assessment of core melting prevention relies on the study of the natural behavior of the GFR2400 and of ASTRID when facing the various accident sequence families. Then, the efficiency and the features of the systems to be foreseen for core degradation prevention are presented. As far as mitigation is concerned, all the consequences of core melting are investigated (i.e. the induced loadings in terms of nature and of range) by considering various core degraded states. Based on the magnitude of these loadings, the needs of mitigation means are assessed for each concept. Among other trends, the presented work shows the very good prevention capability of the SFR concept but the necessity to mitigate the fast vaporization and expansion of degraded core materials. Conversely, the limited coolant capability of the GFR concept and its low thermal inertia require a pressurized gas circulation into the core, limiting its prevention capability whereas its core melting should not induce substantial mechanical loadings of the reactor structures. However, for this last concept, thermochemical interactions between the core materials are an issue deeply investigated in order to understand and simulate core degradation.

Comparison of calculation methods for lead cooled fast reactor reactivity effects

This study analyzes three different reactivity feedback coefficients of the Advanced Lead-cooled Fast Reactor Demonstrator (ALFRED) reactor. During the evaluations, the central difference Direct Perturbation Method (DPM) and the Linear Perturbation Theory (LPT) were applied to determine the feedback coefficients and their uncertainties due to cross-section data. The continuous energy SERPENT Monte Carlo code, the multigroup TSUNAMI-3D sequence of the SCALE program package, and PARTISN discrete ordinate neutron transport code coupled with SEnTRi developed by the authors was applied, and the properties of these fundamentally different calculations are presented. In order to validate the methodologies, lead void coefficient calculations were performed for the Comet critical assembly, which confirmed that a good agreement could be reached between the different methods and measurement results. Besides the high-level transport approximation, the calculations indicated that the appropriate spatial resolution has outstanding importance in describing the high flux gradient.

A coupled neutronic and thermal-hydraulic model for ALFRED

A new method of Serpent–OpenFOAM coupling is developed as a multi-physics model for Advanced Lead Fast Reactor Demonstrator. The reactor core is simulated in Serpent, a continuous-energy Monte Carlo reactor physics code for neutronic analysis. A three-dimensional geometry is modeled for the calculation of neutronic parameters of the reactor initial core operation. The calculated parameters are evaluated with a good agreement compared to the available reference. The fuel assembly with the maximum power is pinpointed in order to be used for thermal-hydraulic analysis. A thermal-hydraulic model is developed to perform computational fluid dynamics calculation using OpenFOAM software, with an application of a heat conjugate transfer solver written in C++ language. A symmetric one-six of the fuel assembly with the highest power is considered in order to reduce the time of calculation. A multi-physics approach is adopted to map Serpent and OpenFOAM coupling for neutronic and thermal-hydraulic analysis. Moreover, a procedure is implemented in order to evaluate the convergence while coupling Serpent and OpenFOAM. With the implementation of multi-physics model, the maximum temperature of the fuel and coolant is investigated in order to observe any malfunctions. The results show that the coolant and the fuel temperature limits for the ALFRED’s initial core are well preserved for the fuel assembly with the highest power.

The investigation and calculation of the transmutation paths for the production of 252cf in fast reactors

The generation and transmutation process of the transplutoniums in the conversion chains for 252cf production is calculated and analyzed. The production and conversion efficiency of the chains with different target nuclides and neutron spectrums including fast, thermal and moderated neutron spectrums is computed. The conversion paths for 252cf under different neutron spectrums and neutron flux levels are analyzed. The multi-group depletion program STEP1.0 based on linear chain method is used to calculate the nuclide densities. The model of the reactor is simulated by the transport-burnup coupling system based on the STEP1.0 and MCMG-II for the detailed description of the target designs in the fast reactor. The differences of the conversion efficiency caused by the neutron spectrums can be qualified more accurately with the multi-group cross section calculations. The higher conversion efficiency than the typical thermal neutron spectrum can be obtained by some moderated spectrum optimized for absorption reaction in fast reactor (CEFR China Experimental Fast Reactor) through the irradiation of some transplutoniums (242Pu, 241Am, 244Cm). The neutron flux of large scale demonstration fast reactor or prototype fast breeder reactor is usually higher than the experimental fast reactor (CEFR) or reactor device. China is also developing the demonstration fast reactor (CFR China Fast Reactor) which will have a maximum neutron flux higher than 5.0E15 n/(cm2·s). Therefore, the 252cf production will be more practical and efficient with CFR.

Design and R&D progress of Core Assembly Deformation Test Facility for China Demonstration Fast Reactor

China Demonstration Fast Reactor (CFR600) is under construction, and the Sub-assembly Deformation Analysis Program (SDAP) developed by China Institute of Atomic Energy is used to optimize the core design and operation management of CFR600. The Core Assembly Deformation Test Facility (CADTF), designed and built by Xi’an Jiaotong University, was selected to carry out a series of assembly bending and thermal bowing tests to provide validation problems for the SDAP. The specimen assembly was specially designed for better non-intrusive optical temperature measurement of the inner wrapper tube. The automatic heating control method and a lot of new measurement technology were applied into the tests to gain abundant and accurate test data.

GERMINAL, a fuel performance code of the PLEIADES platform to simulate the in-pile behaviour of mixed oxide fuel pins for sodium-cooled fast reactors

GERMINAL is a fuel performance code developed by the French Commission of Alternative and Atomic Energies (CEA) to simulate the in-pile behaviour of mixed oxide fuel pins for Sodium-cooled Fast Reactors. The code is continuously being improved.GERMINAL was initially designed to simulate the fuel pin behaviour of the PHENIX and SUPER-PHENIX reactors, which were built in France and have been in operation over the last decades. The GERMINAL models were then extended and improved to meet the needs of the design studies of ASTRID, a project of a technological Sodium Fast Reactor demonstrator in France. The goal of this article is to introduce the current modelling implemented in GERMINAL. The code is validated and the validation work is illustrated here by a selection of comparisons between calculations and measurements. Working perspectives for further modelling improvements are finally presented, through more mechanistic approaches sustained by three-dimensional computations or based on extended physical couplings.

Application of Prompt Self-Powered Neutron Detectors to the Lead-Cooled Fast Reactor Demonstrator ALFRED: Validation of the Monte Carlo Model for Selected SPNDs

The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the industrial prototype European lead fast reactor proposed in lead-cooled European advanced demonstration reactor. It has a relatively low power (125 MWe) with a compact design to reduce the cost but maintaining its representativeness and it is cooled by pure lead. One of the open issues is linked to the neutron flux in-core monitoring system because of the harshness of the environment the detectors should be installed in, due to high temperatures, and the neutron-gamma radiation field levels. Monte Carlo simulation is a possible way of facing the problem, reproducing into a virtual world the reactor core, the surrounding environment and radiation interactions. In previous works, neutron spectra and gamma doses at possible detectors' locations in ALFRED were retrieved, with consideration on the applicability of each suitable device currently available. Fission chambers (FCs) were found to be exploited at reactor start-up and intermediate power range. Prompt self-powered neutron detectors (SPNDs) seemed to be the best solution to monitor the reactor full power, becoming the main research target: their effective applicability on field has to be demonstrated. SPND applications do not include reactor control purposes usually. Moreover, their irradiation experience involved thermal and epithermal neutron spectra monitoring, mainly. The lack of data when SPNDs sense fast neutron fluxes in terms of prompt-response pushed the authors to deepen the study in such direction. The work herein shows the mathematical approach based on Monte Carlo simulation of SPNDs by the Monte Carlo N-particle eXtended code (MCNPX), so as to study the capability of the code in reproducing real devices' signals while experimented on field. Such a verification turned out to be the preliminary stage for studying new concepts for SPNDs, in terms of sensitive materials and geometries, envisaging the possibility for designing, prototyping, and testing new devices in suitable fast neutron-flux facilities.

A passive decay heat removal system for the lead cooled fast reactor demonstrator “Alfred”

The generation IV lead cooled fast reactors are of particular interest for the Italian research. At present, one significant European project in progress is LEADER (Lead cooled European Advanced DEmonstrator Reactor) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, ALFRED (Advanced Lead Fast Reactor European Demonstrator). The demonstrator provides technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. In a previous paper, the Authors proposed the EBBSG (External Boiling Bayonet Steam Generator) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. In the present paper, the Authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the system is investigated through a Matlab–Simulink model. The results are satisfactory since, according to simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

Experiences of reprocessing of plutonium-rich mixed carbide fuels

Mixed carbide (70 % Pu, 30 % U) spent fuels up to a burn up of 155 GWd/Te and a cooling time of around 2 years discharged from the Fast Breeder Test Reactor (FBTR) were successfully reprocessed at COmpact Reprocessing of Advanced fuels in Lead shielded (CORAL) facility in India at Kalpakkam. The operational experience accumulated is translated to design of the future reprocessing plants such as Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and Fast Reactor Fuel Reprocessing Plant (FRP). Concurrent R&D activities have been conducted to optimize process steps for reprocessing fast reactor fuels.

Simulation Model of a Passive Decay Heat Removal System for Lead-Cooled Fast Reactors

The generation IV lead cooled fast reactors are of particular interest for the Italian research: several influential companies (Ansaldo Nucleare, ENEA) are involved in these important European R&D projects. At present, one significant European project in progress is lead cooled European advanced demonstrator reactor (LEADER) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, advanced lead fast reactor European demonstrator (ALFRED). The demonstrator has to include technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. The authors have addressed the issue of bayonet-tube steam generators proposing the external boiling bayonet steam generator (EBBSG) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. This is possible thanks to an external feed-water boiling, based on the known Loeffler scheme, coupled to the bayonet tube concept. In the present paper, the authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the proposed DHR system is investigated through a Matlab-Simulink model. The results are satisfactory since, according to the simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

Innovative steam generation system for the secondary loop of “ALFRED” lead-cooled fast reactor demonstrator

The present paper aims to propose an original steam generation system, which resulted suitable for the application to the secondary loops of lead-cooled fast reactors: reaction heat is extracted from the lead by means of a coolant under vapor phase, thanks to both an external feed-water boiling and special bayonet tubes in the primary–secondary fluid heat exchangers.The performance of three different steam cycles equipped with this system were evaluated for the secondary loop of ALFRED 300MWth demonstrator. The bayonet heat exchangers were sized through a coordinated interaction between a 1D finite-difference mathematical model and an automated size optimization process. A structural analysis of the circular plate supporting the bayonets was performed through the ANSYS 11 solver.The proposed steam cycles, 40bar with superheating, 80bar with live steam re-heating and 80bar with double-superheating, show efficiency values of respectively 34%, 40% and 41%. A performance comparison with other cycles proposed in literature is provided.This solution is expected to provide significant safety advantages, for the secondary fluid low pressure and the gaseous flow in the heat exchangers, still keeping satisfactory efficiency values.

Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

Development of a transient thermal-hydraulic code for analysis of China Demonstration Fast Reactor

The transient thermal-hydraulic code THACOS is under development for analysis of China Demonstration Fast Reactor. Applying modular technology, the code contains the core module, the pump module, the sodium pool module and the heat exchanger module and each module could operate separately. It can provide one-dimensional thermal-hydraulic simulation for the primary sodium coolant loop. The point reactor kinetics equations with six-group delayed neutrons have been applied to calculate the core power considering reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of the core. Multiple-channel model is applied to depict the core. Compressible homogenous flow model is used for the two-phase flow of sodium. The calculated results show that sodium boiling will occur quickly under the ULOF accident without any shutdown rods insertion. While, with the insertion of three hydraulically suspended shutdown rods, the core could be shut down safely and boiling will not occur in a short period of time. Obviously, the passive hydraulically suspended shutdown rods could keep the core safe under ULOF accidents.

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator

In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration has been developed meeting the foremost requirements of limiting the reactivity swing over the core lifetime to about 1 $ and flattening the radial power profiles, as demanded by the choice of wrapper-less (i.e. without flow ducts) fuel assemblies and by the stringent technological constraints imposed by the requirement of short-term deployment. Reactivity coefficients and kinetic parameters have been evaluated for the reference beginning-of-life, middle-of-life and end-of-life core configurations. Furthermore, the results of thermal-hydraulic analyses of the primary loop have confirmed that the system can be effectively cooled by natural circulation heat transport, all the technological constraints being respected even when incorporating peaking factors.

Study on integrated TRU multi-recycling in sodium cooled fast reactor CDFR

In view of recently proposed closed fuel cycle strategy which would recycle the integrated transuranics (TRU) from PWR spent fuel in the fast reactors, the neutronics characteristics of TRU recycled in China Demonstration Fast Reactor (CDFR) are studied in this paper. The results show that loading integrated TRU to substitute pure Pu as driver fuel will mainly make the influence on sodium void worth and negligible effects on other parameters, and hence TRU recycling in CDFR is feasible from viewpoint of core neutronics. If TRU is multi-recycled, the variation of TRU composition depends on fuel types and the ratio of TRU and U when recycling. It is indicated that, when TRU is multi-recycled in CDFR with MOX fuel, the minor actinides (MA) fraction in TRU will firstly decrease to ∼7.24% (minimum) within 8 TRU recycle times and then slowly increase to ∼7.7% after 20 TRU recycle times; while when TRU is multi-recycled in CDFR with metal fuel (TRU-U-10Zr), the MA fraction in TRU will gradually approach to an equilibrium state with the MA fraction of ∼3.8%, demonstrating better MA transmutation effect in metal fuel core. No matter 7.7 or 3.8%, they are both lower than ∼10% in PWR spent fuel with burnup of 45 GWd/tU, which presents satisfying effect of MA amount controlling for TRU multi-recycling strategy. On the other hand, the corresponding recycling parameters such as TRU heat release and neutron emission rate are also much lower in metal fuel than those in MOX fuel. Moreover, TRU recycled in metal fuel will bring greater fissile Pu isotopes equilibrium fraction due to better breeding capability of metal fuel. Finally, it could be summarized that integrated TRU multi-recycling in fast reactor can make contributions to both breeding and transmutation, and such strategy is a prospective closed fuel cycle manner to achieve the object of effective control of cumulated MA amount and sustainable development of nuclear energy.

Construction of a High Temperature Grade 91 Sodium Component Test Loop

A large scale sodium test facility of ‘CPTL’(Component Performance Test Loop) for simulating thermal hydraulic behavior of the Korean demonstration fast reactor components such as IHX(Intermediate Heat Exchanger), DHX(Decay Heat Removal Heat Exchanger) and sodium pump under development by KAERI is to be constructed. The design temperature of this test loop is 600°C and design pressure is 1MPa. The three heat exchangers are made of Grade 91 steel. Another sodium test facility of the ‘STEF’(Sodium Thermal-Hydraulic Experimental Facility) will be constructed next to the CPTL facility to simulate the passive decay heat removal behavior in the sodium cooled fast reactor. In this paper, the overall facility features of the CPTL and STEF are introduced and preliminary conceptual design of the facilities are carried out.

Analysis of the PROFIL and PROFIL-2 Sample Irradiation Experiments in Phénix for JEFF-3.1 Nuclear Data Validation

The PROFIL and PROFIL-2 experiments, performed in the Phénix demonstration fast reactor, irradiated 130 small separate samples containing almost pure isotopes. These highly accurate experiments are a very specific and powerful source of information on the nuclear data of major and minor actinides and several fission products. Their analysis is carried out using the ERANOS-2.1 code system associated to JEFF-3.1 cross-section, fission yield, and decay data. The consistency of the results demonstrates the overall good quality of the actinide nuclear data and experimental techniques used and points out where specific improvement is necessary: fission yields of 235U on neodymium isotopes (5% bias) and integral capture cross sections of the actinides 232Th, 234U, 242Pu, 244Cm, 246Cm (more than 10% bias), 233U, 237Np, 241Pu, and 243Am (bias between 5 and 10%). The optimal values of the branching ratios for 241Am capture (0.85 on 242gAm and 0.15 on 242mAm) are consistent with the latest evaluation data in the fast neutron range. A similar analysis characterized the degree of accuracy of the integral capture cross sections of 19 fission products. Two new experiments of the same kind have been irradiated in Phénix and will undergo dissolutions and isotopic analyses. When they are completed, the analysis of the results will provide additional useful data in both a standard and a slightly moderated neutron spectrum.