Development Of Sodium-cooled Fast Reactors Research Articles

Benchmark analysis of the FFTF LOSWOS test #13 with OpenMC and THACS

In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi’an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie’ outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR.

Sodium-cooled Small Modular Reactors

Amid expectations for nuclear power as a carbon-free energy source are growing to prepare for climate change, deploying efforts for small modular reactors (SMRs) are being actively carried out for the application area to which existing large nuclear power plants are not proper to contribute. Sodium-cooled SMRs are also being developed as one of the candidates to lead the SMR market in the future. This manuscript introduces the background and history of the development of sodium-cooled fast reactors, and the current status of development of Sodium-cooled SMRs and microreactors that are being developed based on them.

Development of Sodium-Cooled Fast Reactors Under Modern Conditions: Challenges and Stimuli

The factors that can promote stable development of sodium-cooled fast reactors with mixed uranium and plutonium fuel under conditions of variable demand for atomic energy, types and magnitude of NFC services in the domestic and foreign markets are examined. The possibility of effective control of the balance of plutonium in a two-component nuclear power system with possible change in the demand for electricity and nuclear fuel of reactors built at home and abroad is shown for the example of an analysis of variants for managing fuel in an industrial power-fuel complex with BN-1200 (PEK-BN). It is noted that high-quality plutonium, which can be produced in PEK-BN, for repeated recycling in the VVER system with uranium and mixed uranium plutonium fuel, plays a fundamental role, and promising directions for expanding the fuel business of the State Corporation Rosatom by providing new types of services on the domestic and foreign markets are noted.

Application of Thermal Analysis in Research and Development of Sodium-Cooled Fast Reactor

Application of Thermal Analysis in Research and Development of Sodium-Cooled Fast Reactor

TRIPOLI-4 Simulation of Core Physics Parameters for Two 3600-MW(thermal) Sodium-Cooled Fast Reactors

To support the development of Sodium-cooled Fast Reactors (SFRs) of Generation IV nuclear energy systems and to study the use of the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library on the core neutronics of large fast neutron reactors, in this work two recent Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) computational benchmarks of two 3600-MW(thermal) SFRs were analyzed with the continuous-energy TRIPOLI-4 code. Both a mixed oxide [(U,Pu)O2] core and a carbide [(U,Pu)C] core were investigated. Under two different fast neutron spectra, the reactor physics parameters—Keff, βeff (effective delayed neutron fraction), sodium void worth, Doppler constant, control rod worth, and core power distribution—were calculated for the beginning of equilibrium cycle condition. Both the pin-by-pin heterogeneous and fuel assembly–level homogeneous calculation models were applied in the whole-core simulation in order to evaluate their impact on the calculation results of SFR reactor physics parameters. The ENDF/B-VII.0 data library from the evaluation was also used with TRIPOLI-4 to study its impact on the SFR core reactivity and the boron carbide control rod worth. Using the mesh tally option, the energy deposition tally, and the upgraded display tool of TRIPOLI-4, radial power distribution and core power maps of the two cores were calculated and compared.

Thermal Hydraulics of Sodium-Cooled Fast Reactors: Key Design and Safety Issues and Highlights

This paper discusses key issues and highlighted topics in thermal hydraulics in connection with Japan’s current sodium-cooled fast reactor development efforts, with particular focus on design study and related research of the Japan Sodium-Cooled Fast Reactor (JSFR). Several innovative technologies, i.e., compact reactor vessel, two-loop system, full natural-circulation decay heat removal, and recriticality-free core, have been investigated to reduce construction cost and to achieve a high level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are ongoing. In this paper, the progress of design study is reviewed, and key issues are discussed. Then, research and development activities on the thermal hydraulics are highlighted in connection with the phenomena taking place in natural-circulation decay heat removal and innovative design and enhanced safety features.

High resolution measurement of neutron inelastic scattering cross-sections for 23Na

The neutron inelastic scattering cross-section of 23Na has been measured in response to the relevant request of the OECD–NEA High Priority Request List, which requires a target uncertainty of 4% in the energy range up to 1.35MeV for the development of sodium-cooled fast reactors. The measurement was performed at the GELINA facility with the Gamma Array for Inelastic Neutron Scattering (GAINS), featuring eight high purity germanium detectors. The setup is installed at a 200m flight path from the neutron source and provides high resolution measurements using the (n,n′γ)-technique. The sample was an 80mm diameter metallic sodium disk prepared at IRMM. Transitions up to the seventh excited state were observed and the differential gamma cross-sections at 110° and 150° were measured, showing mostly isotropic gamma emission. From these the gamma production, level and inelastic cross-sections were determined for neutron energies up to 3838.9keV. The results agree well with the existing data and the evaluated nuclear data libraries in the low energies, and provide new experimental points in the little studied region above 2MeV. Following a detailed review of the methodology used for the gamma efficiency calibrations and flux normalization of GAINS data, an estimated total uncertainty of 2.2% was achieved for the inelastic cross-section integrals over the energy ranges 0.498–1.35MeV and 1.35–2.23MeV, meeting the required targets.

Operating experience and prospects for future development of sodium-cooled fast reactors

The main stages of the development of sodium-cooled fast reactors in Russia are described. The statistical information on the deviations from the normal regime of functioning that occur in BOR-60 and BN-600 operation is presented. The important role of BN-600 at the Beloyarskaya nuclear power plant in the validation of sodium-cooled fast reactors is shown. A conclusion is drawn on the basis of a review of the operating experience about the level of mastery of sodium-cooled fast reactors in Russia and the prospects for further development of such reactors are analyzed taking account of this experience.

FEASIBILITY OF AN INTEGRATED STEAM GENERATOR SYSTEM IN A SODIUM-COOLED FAST REACTOR SUBJECTED TO ELEVATED TEMPERATURE SERVICES

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

Stimulation of the development of sodium-cooled fast reactors

Stimulation of the development of sodium-cooled fast reactors