Operation Of Fast Reactors Research Articles

Fuel performance code BERKUT-U to simulate the in-pile behavior of a single oxide or nitride fuel rod for fast reactors

This paper describes the fuel performance code BERKUT-U, which the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) designed as part of the "Codes of The New Generation" subproject of the Proryv Project. The code aims to model oxide or nitride single fuel rod behavior with a gas or liquid metal sublayer under normal and accident conditions of a liquid metal-cooled fast reactor operation. The BERKUT-U code's models, incorporating the MFPR/R code, are grounded in the contemporary understanding of mechanisms governing the fundamental processes in fuel rods under irradiation, which substantially enhances the code's predictive ability in comparison with the engineering analogs. Simulations of the nitride and oxide fuel rod behavior in BN-600 and BOR-60 fast reactors have demonstrated good agreement with the post-irradiation examination data. Further validation is foreseen as the corresponding data are available.

Multi-section fission ionization chamber for measurement of [formula omitted] reaction in fission tagging method

The 239Pu(n,γ) reaction cross section is very important for operation of both thermal and fast reactors, when loaded with MOX fuels. According to the NEA/OECD High Priority Request List the precision of cross section data for this reaction should be improved. The cross section of (n,f) reaction is much higher compared to (n,γ) for this isotope. In such conditions the fission tagging technique could be applied to identify the fission background. In the past, this technique was successfully used for capture measurements at the n_TOF facility at CERN. The multi-section fission ionization chamber was constructed and used in the combination with Total Absorption Calorimeter (TAC) for detecting gamma rays for the precise measurement of 239Pu(n,γ) reaction cross section at the n_TOF facility.

Flow Regime Classification in Hexagonal Wire-Wrapped Fuel Assembly Using Advanced Machine Learning Models

This study presents a new approach to flow regime classification specifically tailored for typical wire-wrapped fuel assemblies in sodium fast reactors. Historically, the definition and understanding of flow regime boundaries have been extensively researched. However, many of these models suffer inaccuracy due to a lack of comprehensive data. In particular, the limited data, with only 36 data points for the laminar-to-transition boundary and 145 data points for the transition-to-turbulent boundary, often result in suboptimal models. Recognizing the critical data gap, this study classified flow regimes based on a robust data set of over 5000 data points. A diverse range of algorithms was used to find the optimal classification model. These included logistic regression, artificial neural networks, support vector classifiers, Naïve Bayes, Gaussian Naïve Bayes, K-Nearest Neighbors, random forest, AdaBoost, GradientBoost, and XGBoost. A comparative analysis of these algorithms provides valuable insights. This study presents a comprehensive set of machine learning algorithms to improve the accuracy and reliability of flow regime classification, which is a critical step in predicting friction factors and the efficient operation of sodium fast reactors.

Influence of operation of thermal and fast reactors of the Beloyarsk NPP on the radioecological situation in river ecosystems

The results of long-term radioecological monitoring of seven rivers in the 15-km zone of influence of the Beloyarsk Nuclear Power Plant have been presented. A comparative analysis of the content of a wide range of natural and artificial radionuclides in the main components of river ecosystems: surface waters, bottom sediments, floodplain soils, macrophytes and ichthyofauna has been made. The influence of waste technological waters of thermal (AMB-100 and AMB-200) and fast (BN-600 and BN-800) reactors of the Beloyarsk NPP on the content of radiologically significant isotopes in the water and bottom sediments of the Pyshma and Olkhovka rivers has been evaluated. It has been studied that the transition from thermal to fast reactors contributed to a significant decrease in the intake of artificial radionuclides into the rivers of the area where the Beloyarsk NPP is located. So, in the water of the Olkhovka river for the period 1978–2019 the decrease in the specific activity of 137Cs was 480, 3H - 36, 90Sr - 3.5 times. The maximum discharge of artificial radioisotopes into river ecosystems was noted during the period of recovery work after emergencies at the AMB-100 and AMB-200 reactors. In recent years, the content of artificial radionuclides in water, macrophytes and ichthyofauna of rivers in the zone of influence of the Beloyarsk NPP, except for the Olkhovka, is at the level of the regional background.

Influence of operation of thermal and fast reactors of the Beloyarsk NPP on the radioecological situation in the cooling pond: Part II, Macrophytes and fish

The influence of waste technological waters of thermal and fast reactors of Beloyarsk NPP (Russia) on the accumulation of 60Co, 90Sr and 137Cs in macrophytes and ichthyofauna of the cooling pond has been studied. Critical radionuclides, routes of their entry into the ecosystem and periods of maximum discharge of radioisotopes into the cooling pond have been determined. It is shown that the technology of electricity generation at the Beloyarsk NPP, based on fast reactors, has a much smaller effect on the release of artificial radionuclides into the environment. Therefore, during the entire period of monitoring studies (1976–2019), the decrease in the specific activity of radionuclides of NPP origin in macrophytes was 13–25800 times, in ichthyofauna 1.5-44.5 times. The maximum discharge of artificial radionuclides into the Beloyarsk reservoir was noted during the period of restoration and decontamination work aimed at eliminating the emergencies at the AMB reactors of NPP. The factors influencing the accumulation of artificial radionuclides in the components of the freshwater ecosystem of the Beloyarsk cooling pond have been determined, including: the physicochemical nature of radioisotopes, their concentration in surface water, the temperature of the aquatic environment, the trophicity of the reservoir, the species of hydrobionts.

Simulating the fuel cycle of a lead-cooled fast reactor

The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined.

Influence of operation of thermal and fast reactors of the Beloyarsk NPP on the radioecological situation in the cooling pond. Part 1: Surface water and bottom sediments

The results of radioecological monitoring of the cooling pond Beloyarsk NPP (Russia) have been presented. The influence of waste technological waters of thermal and fast NPP reactors on the content of artificial radionuclides in surface waters and bottom sediments of the Beloyarsk reservoir has been studied. The long-term dynamics of the specific activity of 60Co, 90Sr, 137Cs and 3H in the main components of the freshwater ecosystem at different distances from the source of radionuclide discharge has been estimated. Critical radionuclides (60Co and 137Cs), routes of their entry and periods of maximum discharge of radioisotopes into the cooling pond have been determined. It is shown that the technology of electricity generation at Beloyarsk NPP, based on fast reactors, has a much smaller effect on the flow of artificial radionuclides into the freshwater ecosystem of the reservoir. During the entire period of monitoring studies, the decrease in the specific activity of radionuclides from NPP origin in surface waters was 4.3–74.5 times, in bottom sediments 10–505 times. The maximum discharge of artificial radionuclides into the reservoir was noted during the period of restoration and decontamination work aimed at eliminating emergencies at the AMB thermal reactors of the first stage of the Beloyarsk NPP.

Соотношение радиационно-обусловленных потенциальных канцерогенных рисков ОЯТ реактора ВВЭР-1000 и РАО реактора БРЕСТ-1200 при выработке 1 ГВт х год электроэнергии. Часть 2. Радиологическая миграционная эквивалентность

Currently it is acknowledged that the problem of disposal of radioactive waste (RW) from the nu-clear reactors largely depends on the use of fast neutron reactors (fast reactors) in nuclear power industry. These reactors are able to burn and transmutate long-lived radionuclides in closed nu-clear fuel cycle (NFC). In an open NFC with thermal neutron reactors (thermal reactors), unpro-cessed spent nuclear fuel (SNF) is buried. In a closed NFC with fast reactors the reprocessed SNF is disposed. In this paper, the feasibility and possibility of closing the NFC in a two-component nuclear power industry with thermal and fast reactors are studied. Potential biological hazard (PBH) to the public from unprocessed SNF from the thermal reactor WWER-1000 and re-processed SNF from the fast reactor BR-1200, were evaluated with account of relevant radionu-clides migration from the deep disposal to the earth surface. PBH values were expressed in terms of committed effective dose and lifetime attributable risk (LAR) of malignant neoplasms, normalized to 1 GW of generated electricity. PBH of RW from BR-1200 is about 100 times lower than risk from unprocessed SNF from WWER-1000. Comparative evaluation of PBH from BR-1200 and unprocessed SNF from WWER-1000 demonstrated that the committed effective doses from the dwell-water used by the public from the time of radionuclides appearance at the earth surface are 0.7 µSv and 80 µSv/year respectively, and LAR values are 6∙10-9 and 7.74∙10-6 re-spectively. At the same time, PBH of reprocessed SNF from WWER-1000, from the time of radio-nuclides appearance on the earth’s surface is an order less than PBH of RW from BR-1200. In accordance with the principle of radiological equivalence of RW disposal, it was established that the RW from BR-1200 satisfies the time limit for achievement of the equality with the PBH of the equivalent mass of uranium raw material will be within the range of 300 years. If PBH is ex-pressed in terms of LAR the radiological migration equivalence is achieved even earlier. Thus, when reprocessing SNF from thermal reactors and operating thermal reactors with simultaneous construction and operation of fast reactors (in a two-component nuclear power industry), it is ad-visable to use a closed NFC and dispose of RW, considering the principle of radiation-migration equivalence.

Оценка радиационной обстановки в районе расположения АО «ГНЦ НИИАР» до начала эксплуатации ИЯУ МБИР. Часть 2. Водные экосистемы

The analysis of the radiation survey results of freshwater ecosystems of three rivers and the Cheremshansky Bay of the Kuibyshev Reservoir in the vicinity of State Scientific Centre “Research Institute of Atomic Reactors” (SSC RIAR, Institute) before the start of the MBIR fast neutron reactor operation was carried out. The content of natural (K-40, Ra-226, Th-232) and artificial (Sr-90, Cs-137) radionuclides in the main components of freshwater ecosystems: surface and underground waters, bottom sediments, floodplain soils and vegetation, and ichthyofauna was estimated. It has been shown that the specific activity of Cs-137 in the waste water of industrial storm sewerage (PLK-1) of the enterprise is 10-70, Sr-90 2-10 times higher compared to the surface waters of reservoirs of its zone of influence. This is due to the removal of radionuclides from the site of local contamination in the sanitary protection zone of SSC RIAR. In Cheremshansky Bay, downstream, the concentration of radioisotopes decreases by 8-50 times due to dilution processes and their sorption by bottom sediments and hydrobionts. It is noted that the sorption properties of bottom sediments are determined by the presence of organic compounds in their composition. The multidirectional nature of the accumulation of artificial radionuclides in floodplain soils and plants is shown. In some cases, it is determined by the distance from the source of gas-aerosol emissions of radioisotopes and affects the secondary pollution of freshwater ecosystems. Monitoring data demonstrate a stable decrease in the content of artificial radionuclides in surface waters. In Cheremshansky Bay in 2005-2020 for Cs-137 it was 5 times, for Sr-90 it was 3 times. It has been noted that in recent years, the specific activity of artificial radionuclides in sewage of SSC RIAR, surface and underground waters, as well as fish in the zone of enterprise influence is at the level of the regional background and is 1-3 orders of magnitude lower than the current radiological standards. The importance of further radioecological monitoring of freshwater ecosystems in the SSC RIAR location area after the commissioning of the MBIR for assessing the impact of the new reactor on humans and the environment is shown.

Simulation of delayed neutron precursor release due to wet rupture in a fuel pin and its detection methodology in PFBR

Detection of dry and wet fuel pin clad rupture is very important to ensure safe operation of fast reactors. An in-vessel concept of delayed neutron detection system has been adopted for detecting wet rupture in a pool type MOX fueled 500 MWe Prototype Fast Breeder Reactor (PFBR) which is currently in the advanced stage of commissioning in India. It has 8 DND modules installed in the hot pool just above the intermediate heat exchangers inlet windows; each module has three 0.2 cps/nv sensitivity high temperature fission chambers which is surrounded by graphite moderator and B4C shield. In this paper, we propose a revised simulation methodology of DN signal estimation involving various steps of neutronics and thermal hydraulics calculations, which is considered to be a pre-requisite for the program of experimental demonstration of bulk DND system. A special DN source sub-assembly simulating a failed MOX pin of 1 cm2 defect area has been manufactured using natural uranium metal fuel slugs with a perforated clad. Prompt recoil model is employed to compute the release of fission products through the perforated clad. The DN source strength is estimated by solving 52 coupled radioactive decay equations for 52 parent and daughter (DN precursor) nuclides. The CFD analysis using PHOENICS code is performed to simulate mixing and decay of DN precursors in the sodium pool during their transport. Neutron transport calculation with distributed DN sources around the DND block is done to estimate the DN signal in the HTFC detectors using the fission equivalent neutron flux concept. The sensitivity of DN signal for different core locations is studied in detail at 20% thermal power level with 50% flow. The DN signal variation with power is also studied. This research work is performed to have a theoretical database prior to the actual calibration experiment of bulk DND system in PFBR after attaining first criticality.

UNCERTAINTY ANALYSIS OF SUB-EXERCISES IN UAM-SFR BENCHMARK WITH THE MCS CODE

Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess the effect of nuclear data uncertainties on parameters of interest in SFR analysis. In this paper, sub-exercises of a medium 1000 MWth metallic core (MET-1000) and a large 3600 MWth oxide core (MOX-3600) are tested by a Monte Carlo code MCS to perform uncertainty analysis. Classical perturbation theory and generalized perturbation theory are used to calculate sensitivity coefficients. Uncertainty is calculated by multiplying the sensitivity coefficients and relative covariance matrix from ENDF/B-VII.1 library.

Possible ways to solve the problems of using MOX fuel in thermal reactors

Modern nuclear power has a number of crucial problems, including the accumulation of spent nuclear fuel (SNF) and the projected shortage of nuclear resources. The conversion to a closed fuel cycle is the most promising solution to these problems. Mixed uranium-plutonium fuel (MOX fuel) has been successfully tested and commissioned under this concept by several European countries. However, its operation in reactors designed for uranium fuel has a number of problems, most of which have been resolved. Nevertheless, the spent MOX fuel has neutronic characteristics that do not allow its reuse in thermal reactors and solve the problem of SNF accumulation. This is due to the accumulation of heavy (even) plutonium isotopes, which are not fissionable in the thermal spectrum. In fast reactors, it is possible to create conditions under which the isotopic composition of plutonium from the spent MOX fuel from thermal reactors changes in such a way that it becomes suitable for reuse in thermal reactors. This process is called improving. The concept for the development of two-component nuclear energy based on the joint operation of thermal and fast neutron reactors and the creation of a closed fuel cycle was adopted in Russia to implement this process and solve crucial problems of nuclear power.

Development of oxygen sensor for pyrochemical reactors of spent nuclear fuel reprocessing

The problem of closing the nuclear fuel cycle is not only related to the development of new types of nuclear fuel and the operation of fast neutron reactors, but also to the complex schemes for the pyrochemical reprocessing of spent nuclear fuel (SNF), which, in turn, require adherence to strict process parameters. In particular, this concerns the operation of the reduction of oxidized SNF mainly by metallic lithium. The paper presents the basic scientific principles and the results of experimental verification of the operation of an electrochemical sensor for measuring oxygen in molten salts in pyrochemical reactors for the reprocessing of spent nuclear fuel. The sensor design consists of two combined electrochemical cells based on the solid electrolyte ZrO2-Y2O3 with a common reference electrode. The sensor allows continuous measurement of the oxygen activity in the oxide-chloride melt and the partial pressure of oxygen in the gas atmosphere above the melt directly during the process of pyrochemical processing. Experimental verification of the sensor performance was performed in a reactor with LiCl-Li2O melts at a temperature of 650 ° C. The resource of continuous sensor operation exceeded 500 hours, and the number of thermal cycles without destruction was at least 20. The sensor readings were found to depend on the specified Li2O content in the LiCl melt.

Structural and Thermodynamic Investigation of the Perovskite Ba2NaMoO5.5.

Neutron diffraction, X-ray absorption spectroscopy (XAS), and Raman spectroscopy measurements of the quaternary perovskite phase Ba2NaMoO5.5 have been performed in this work. The cubic crystal structure in space group Fm3̅m has been refined using the Rietveld method. X-ray absorption near-edge structure spectroscopy (XANES) measurements at the Mo K-edge have confirmed the hexavalent state of molybdenum. The local structure of the molybdenum octahedra has been studied in detail using extended X-ray absorption fine structure (EXAFS) spectroscopy. The Mo–O and Mo–Ba distances have been compared to the neutron diffraction data with good agreement. The coefficient of thermal expansion measured in the temperature range of 303–923 K, using high temperature X-ray diffraction (HT-XRD) (αV = 55.8 × 10–6 K), has been determined to be ∼2 times higher than that of the barium molybdates BaMoO3 and BaMoO4. Moreover, no phase transition nor melting have been observed, neither by HT-XRD nor Raman spectroscopy nor differential scanning calorimetry, up to 1473 K. Furthermore, the standard enthalpy of formation (ΔfHm°) for Ba2NaMoO5.5(cr) has been determined to be −(2524.75 ± 4.15) kJ mol−1 at 298.15 K, using solution calorimetry. Finally, the margin for safe operation of sodium-cooled fast reactors (SFRs) has been assessed by calculating the threshold oxygen potential needed, in liquid sodium, to form the quaternary compound, following an interaction between irradiated mixed oxide (U,Pu)O2 fuel and sodium coolant.

Cross section measurement of 241Am(n,f) reaction at the Experimental Area 2 of the n_TOF facility at CERN: First Results

The condition for the safe design and operation of fast neutron reactors and energy boosters (Generation-IV reactors, ADS systems [1]) is the accuracy of nuclear data. The 241Am isotope (T1/2 = 433 years) is highly present in nuclear waste, accounting for about 1.8% of the actinide mass in PWR UOx nuclear reactors’ waste [2]. In addition, the 241Am isotope is further produced by the β decay of the 241Pu isotope (T1/2 = 14.3 years). Given the high production rate of 241Am isotope, its incineration with concurrent energy production is considered to be of utmost importance for the design and implementation of the recycling of existing nuclear waste. Sensitivity studies of the proposed systems for energy production showed that high-precision measurements of the cross section of the 241Am(n,f) reaction are required. In the present work, the 241Am(n,f) reaction cross section was measured in the Second Experimental Area of the n_TOF facility at CERN, using an array of Micromegas detectors. For the measurement, six targets of 241Am with average activity of 17 MBq per sample were coupled with an equal number of detectors in a common chamber. Additionally two 235U and two 238U samples were coupled with Micromegas detectors utilizing the neutron flux determination. Within this work, an overview of the experimental set-up and the adopted data analysis technique is presented along with preliminary results.

BNcode - УСОВЕРШЕНСТВОВАННЫЙ КОД ДЛЯ НАУЧНОГО СОПРОВОЖДЕНИЯ ДЕЙСТВУЮЩИХ РЕАКТОРОВ БН

The relevance of this work is due to the need of the supervisor (IPPE) to perform work on the control and maintenance of the operation of fast sodium reactors. The purpose of this work is to develop an integrated code of a new generation for the scientific support of existing reactors BN-600 and BN-800 and in the future BN-1200. The code being developed for scientific support of the fast reactor operation serves for current and predictive calculations of reactor characteristics, safety justification during overload, justification of the micro-campaign reactivity balance, compliance with established performance criteria for fuel elements and other elements. Also, this code will be used as an independent calculation tool of increased accuracy for confirming and, if necessary, correcting the results of the operational support received at the station, as well as for analyzing non-standard settlement situations of increased complexity, such as the justification for entering BN-600 and BN-800 irradiating devices for producing isotopes (in particular, 60Co), irradiation of fuel assemblies with a nitride type of fuel.

Ex-Vessel Delayed Neutron Detection Systems for the ASTRID Sodium-Cooled Fast Reactor

Monitoring fuel integrity during sodium-cooled fast reactor operation becomes a crucial issue when sodium uranoplutonate formed consecutively to a clad aperture will foster cooling defaults. Delayed neutron detection (DND) systems are the key tools allowing to prevent and to mitigate such an accident by monitoring the concentration of delayed neutron precursors contained into the coolant. DND systems operate in hard gamma-rays fluxes and low-neutron activity levels. They have to be carefully optimized in order to detect, as early as possible, potentially harmful failures. A discussion is conducted on privileged ways to design DND systems, and some conclusions are stated.

Thermodynamic assessment of the Fe-Te system. Part II: Thermodynamic modeling

A thermodynamic description of the Fe-Te system modeled via the Calphad method is proposed, based on data published in a preceding publication Part I: Experimental study, and that available in literature. End-member formation energies for the phases β, β', δ, δ' and ε, as well as lattice stabilities of FCC and BCC tellurium, have been evaluated via DFT and used in the numerical optimization. The final Gibbs energy models fit thermodynamic and phase diagram data well, and inconsistencies are discussed. The thermodynamic description is then used to evaluate Gibbs energy of formation for selected Fe-Te compounds of interest for the modeling of internal corrosion of stainless steel fuel pin cladding during operation of Liquid Metal-cooled Fast Reactors (LMFR).

Accurate U238 ( n, 2n ) U237 reaction cross-section measurements from 6.5 to 14.8 MeV

The cross section for the $^{238}\mathrm{U}(n,2n)^{237}\mathrm{U}$ reaction has been measured in the incident neutron energy range from 6.5 to 14.8 MeV in small energy steps using an activation technique. Monoenergetic neutron beams were produced via the $^{2}\mathrm{H}(d,n)^{3}\mathrm{He}$ and $^{3}\mathrm{H}(d,n)^{4}\mathrm{He}$ reactions. $^{238}\mathrm{U}$ targets were activated along with Au and Al monitor foils to determine the incident neutron flux. The activity of the reaction products was measured in TUNL's low-background counting facility using high-resolution $\ensuremath{\gamma}$-ray spectroscopy. The results are compared with previous measurements and latest data evaluations. Statistical-model calculations, based on the Hauser-Feshbach formalism, have been carried out using the ${\mathrm{CoH}}_{3}$ code and are compared with the experimental results. The present self-consistent and high-quality data are important for stockpile stewardship and nuclear forensic purposes as well as for the design and operation of fast reactors.

Progress in reliability of fast reactor operation and new trends to increased inherent safety

Abstract The reasons for the renewed interest in fast reactors and an overview of the progress in sodium cooled fast reactor operation in the last ten years are given. The excellent operational performance of sodium cooled fast reactors in this period is highlighted as a sound basis for the development of new fast reactors. The operational performance of the BN-600 is compared and evaluated against the performance of German light water reactors to assess the reliability. The relevance of feedback effects for safe reactor design is described, and a new method for the enhancement of feedback effects in fast reactors is proposed. Experimental reactors demonstrating the inherent safety of advanced sodium cooled fast reactor designs are described and the potential safety improvements resulting from the use of fine distributed moderating material are discussed. One sentence summary: Operating fast reactors have shown excellent in-service behavior within the last 10 years, new designs and methods are available to significantly improve safety.