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This paper shows the verification work of SARAX code system in the reactor core transient calculation based on the simplified EBR-II Benchmark. The SARAX code system is an analysis package developed by Xi'an Jiaotong University and aims at the advanced reactor R&D. In this work, a neutron-photon coupled power calculation model and a spatial-dependent reactivity feedback model were introduced. To verify the models used in SARAX, the EBR-II SHRT-45R test was simplified to an ULOF transient with an input flowrate change curve by fitting from reference. With the neutron-photon coupled power calculation model, SARAX gave close results in both power fraction and peak power prediction to the reference results. The location of the hottest assembly from SARAX and reference are the same and the relative power deviation of the hottest assembly is 2.6%. As for transient analysis, compared with experimental results and other calculated results, SARAX presents coincident results both in trend and absolute value. The minimum value of core net reactivity during the transient agreed well with the reported results, which ranged from −0.3$ to −0.35$. The results verify the models in SARAX, which are correct and able to simulate the in-core transient with reliable accuracy.

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in keff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

A new reactor concept is described that directly couples a supercritical CO2 (sCO2) power cycle with a CO2-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO2 concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (∼20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

Among Generation IV reactors, the sodium-cooled fast reactor (SFR) is attracting attention as a system having great potential for commercial use. Gas entrainment is a thermal-hydraulic issue related to the safety problem of the reactor core in the SFR. Typically, a dipped plate or baffles are installed under the free surface to suppress gas entrainment. However, these approaches can cause gas entrainment in other locations and require many trial-and-error and verifications. In this study, a new strategy using magnetohydrodynamics to suppress gas entrainment in the SFR is proposed. In a counter-flow model, a judgment criterion of gas entrainment occurrence was developed for both water and liquid metal. Moreover, the gas entrainment can be completely suppressed by applying a magnetic field.

To effectively monitor the variety of distributions of neutron flux, fuel power or temperatures in the reactor core, usually the ex-core and in-core neutron detectors are employed. The thermocouples for temperature measurement are installed in the coolant inlet or outlet of the respective fuel assemblies. It is necessary to reconstruct the measurement information of the whole reactor position. However, the reading of different types of detector in the core reflects different aspects of the 3D power distribution. The feasibility of reconstruction the core three-dimension power distribution by using different combinations of in-core, ex-core and thermocouples detectors is analyzed in this paper to synthesize the useful information of various detectors. A comparison of multilayer perceptron (MLP) network and radial basis function (RBF) network is performed. RBF results are more extreme precision but also more sensitivity to detector failure and uncertainty, compare to MLP networks. This is because that localized neural network could offer conservative regression in RBF. Adding random disturbance in training dataset is helpful to reduce the influence of detector failure and uncertainty. Some convolution neural networks seem to be helpful to get more accurate results by use more spatial layout information, though relative researches are still under way.

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the l -plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb–Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

Reactor fuel's power distribution plays a vital role in designing the new generation thermionic Space Reactor Power Systems (SRPS). In this paper, the 1/12th SPACE-R's full reactor core was numerically analyzed with two kinds of different axial power distribution, to identify their impacts on thermal-hydraulic and thermoelectric characteristics. In the benchmark study, the maximum error between numerical results and existing data or design values ranged from 0.2 to 2.2%. Four main conclusions were obtained in the numerical analysis: a) The axial power distribution has less impact on coolant temperature. b) Axial power distribution influenced the emitter temperature distribution a lot, when the core power was cosine distributed, the maximum temperature of the emitter was 194 K higher than that of the uniform power distribution. c) Comparing to the cosine axial power distribution, the uniform axial power distribution would make the maximum temperature in each component of the reactor core much lower, reducing the requirements for core fuel material. d) Voltage and current distribution were similar to the axial electrode temperature distribution, and the axial power distribution has little effect on the output power.

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1 , consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [ 2 ]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were

This work proposes and employs a mixed-complexity modeling approach to creating a 3D computational fluid dynamics (CFD) model to predict the syngas production from Miscanthus briquettes in a two-stage downdraft gasifier operating with different gasification fluids. The study was performed at steady state regime in the Ansys Fluent environment considering the non-premixed combustion model. A probability density function is also included as a tool for the description of the chemical kinetics to predict the syngas composition following the main chemical reactions involved in the gasification process. The goal of this approach is to reduce the computational cost while still providing accurate predictions. In comparison to experimental gasification with air, the model correctly predicts the temperature profile inside the reactor, the composition of the syngas (CO, H2 and CH4), and therefore the lower heating value (LHV). For cases involving the use of saturated steam and oxygen as gasification fluids, the model predicted key species concentrations in the gasification zone and the reactor core and accurately described the significant increase in the LHV of the syngas. This approach opens the possibility of studying the gasification process in moving-bed reactors using different gasification fluids and feedstocks based on their elemental and proximate analysis.

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.

Insulation and magnetic properties of epoxy composites filled with pure iron powders were studied as a function of iron powder volume fraction (R Fe ) in the range of 0-70 vol% to achieve a good balance between insulation and magnetic properties. Frequency-dependent impedance measurement revealed that the relative permittivity, e r ’ increased from 4.2 to 194.5 with increasing R Fe up to 50 vol%. The loss factor, e r " increased from 100.2 to 959.3 and the electric conductivity, σ increased from 5.6×10−5 S/m to 5.3× 10−4 S/m with increasing R Fe up to 70 vol%. The dependence of the σ value on the R Fe value was explained by the equivalent serial CR connection model. The initial maximum relative permeability, μ r_max increased from 3.73 to 8.89 with increasing R Fe up to 70 vol%. These data were compared with those for sample "Reference", which has been in practical use in industrial electric machine applications.

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).

In carrying out quantitative safety assessment at nuclear power plants equipped with pressurized water reactors (the Russian version of which is known as VVER-type reactors), for estimating the risk of boron compound crystallization both in the reactor core and in the spent fuel pool under various emergency conditions of operation, it is necessary to have data on the solubility of these compounds with estimated data accuracy indicators. The article considers the boric acid compounds the data on the solubility of which are of interest in estimating the potential risks of their crystallization with the purpose of retaining the core cooling function during accidents. The forms in which the solid phases of borates exist in the H3BO3/HBO2/B2O3 + Н2О and LiBO2 + H2O systems at elevated temperatures are estimated. Experimental data on the solubility of boric acid stable solid phases and also of lithium borates in water in the temperature range from 273 to 350°С are processed. For the H3BO3/HBO2/B2O3 + Н2О system, the temperature intervals corresponding to equilibrium existence of solid (crystalline, denoted by the subscript “cr”) phases of boric acid Н3ВО3(cr), НВО2(cr), and В2О3(cr) are considered. For each of these phases, formulas for calculating the solubility in water as a function of absolute temperature are obtained. For crystal hydrates of lithium borates, four temperature intervals are identified taking into account different numbers of connected water molecules, and formulas for determining the temperature dependences of solubility are derived for each of them with taking phase transformations into account. For estimating the accuracy of correlations for solubility calculations obtained from processing the totality of experimental data, the average relative error values are determined. It is shown that the above-mentioned formulas give sufficiently accurate replication of experimentally obtained solubility values in the H3BO3/B2O3 + H2O and LiBO2 + H2O systems in the entire range of temperatures considered.

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

At present, most of the widely used system codes for nuclear safety analysis are one-dimensional, which cannot effectively simulate the flow field of the reactor core or other structures. This is true even for the system codes containing three-dimensional modules with limited three-dimensional simulation function such as RELAP-3D. In contrast, the computational fluid dynamics (CFD) codes excel at providing a detailed three-dimensional flow field of the reactor core or other components; however, the computational domain is relatively small and results in the very high computing resource consuming. Therefore, the development of coupling codes, which can make comprehensive use of the advantages of system and CFD codes, has become a research focus. In this paper, a review focus on the researches of coupled CFD and thermal-hydraulic system codes was carried out, which summarized the method of coupling, the data transfer processing between CFD and system codes, and the verification and validation (V&V) of coupled codes. Furthermore, a series of problems associated with the coupling procedure have been identified, which provide the general direction for the development and V&V efforts of coupled codes.

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as 241Am, 137Ba, 244Cm, 238Pu, and 90Y.

This paper discusses an artificial neural inversion approach for non-destructive testing of the graphite moderator bricks that make up the cores of Advanced Gas-Cooled Reactors (AGR). The study employed fully connected feedforward neural networks consisting of four hidden layers and trained with backpropagation approach using Levenberg-Marquardt optimisation algorithm. The approach is based on multi-frequency (MF) eddy current (EC) measurements, and combinations of simulated and measured datasets from a laboratory sample and one of the operating reactor core, along with different regularisation parameters were used to train and test the networks. Various types of artificially generated errors were added to the data during training procedures, which in turn allowed for error tolerance and improved the generalisation of the ANNs to unseen test datasets. First, the ANN was tested using unseen simulated data, followed by the measurements collected from laboratory sample and one of the operating reactor core. The first test from the unseen simulated data showed mean profile error ranging between 1.30% and 8.20%, whereas the profiles estimated from reactor core measurements showed mean profile error ranging between 1.84% and 17.80% when compared with the resistivity measurements from trepanned graphite sample taken out of the reactor core. Further comparison of the network outputs against the profiles estimated using traditional iterative inversion algorithm indicates reasonable agreement between the two approaches with the exception of one case, but the solution time for the ANN was found to be over three orders of magnitude faster than the iterative inversion algorithm.

Considering a hypothetical accident, the deposition of radio krypton (85Kr) has been studied that is released from TRIGA MARK-II reactor by its decay behavior. The measurement of radiological ground concentration leads to the study of the emission process of 85Kr nucleus, and the site-specific data related to this measurement have been analyzed later on. In this work, the radioactivity in the reactor core and release rate as well as Gaussian diffusion factor have also been considered. It is observed from the data analysis that the maximum concentration of 85Kr in ground is 1.115E+3 Bq/m2 in South (S) direction. Here, the nuclear binary fission has been presented with the greater probability of production of magic nuclei where the reactions found to be endoergic where a nuclear reaction occurs with the absorption of energy. Doubly magic nuclei like 4He, 16O, 40Ca, and 48Ca have also been identified as fission fragments that follow the nuclear shell closure. The results of this work will be an important guide in the study of radionuclide splitting into several nuclei as well as for accidental scinerio. GUB JOURNAL OF SCIENCE AND ENGINEERING, Vol 7, Dec 2020 P 57-61

Наведено результати аналізу щодо впливу радіаційної повзучості на розрахункову оцінку формозміни вигородки активної зони реактора ВВЕР-1000 за умов довгострокової експлуатації. Застосовано сучасні моделі радіаційного розпухання і радіаційної повзучості, в яких враховується вплив напруженого стану і накопиченої незворотної деформації на процеси розпухання і повзучості аустенітних сталей, що перебувають під впливом нейтронного опромінення і підвищеної температури. Сформульовано основні положення розрахунку напружено-деформованого стану вигородки та внутрішньокорпусної шахти реактора з урахуванням умов контактної взаємодії. Розрахунковий аналіз виконано у двовимірній постановці для поперечного перерізу вигородки з максимальною за висотою пошкоджуючою дозою і температурою опромінення за умови узагальненої плоскої деформації. Дані щодо формозміни вигородки одержано на основі розв’язання зв’язаної контактної задачі залежно від накопиченої пошкоджуючої дози опромінення. Визначення температурного поля і напружено-деформованого стану виконано з урахуванням перерозподілу температури через порушення проєктних умов протоку теплоносія в зоні контакту вигородки з шахтою. Результати розрахунків одержано з використанням медіанних параметрів температурно-дозової залежності радіаційного розпухання аустенітної сталі 08Х18Н10Т. Встановлено, що урахування радіаційної повзучості сприяє зниженню рівня напружень, проте збільшує рівень розпухання і переміщення, що додає консерватизму до прогнозної оцінки формозміни вигородки порівняно з даними без урахування радіаційної повзучості. The paper presents the results of the analysis of the irradiation creep effect on the calculated assessment of the form change in the WWER-1000 reactor core baffle under long-term operation. The modern models of radia tion-induced swelling and radiation creep are used. They consider the effect of the stressed state and accumulated irreversible strains on the processes of swelling and creep in austenite steels under the neutron irradiation and elevated temperature. The basic tenets of the calculation of the stress-strain state in the core baffle and invessel reactor barrel considering the contact interaction are stated. The calculation analysis is performed in the two-dimensional statement for a cross-section of the core baffle with the maximum (by height) damaging dose and irradiation temperature under the generalized plane strain conditions. The data on a form change in the core baffle are obtained using the solution to the mixed contact task depending on the accumulated damaging irradiation dose. The determination of the temperature field and stress-strain state was made considering the temperature redistribution due to the violation of the design conditions of the heat carrier flow within the contact zone between the core baffle and barrel. The calculation results are determined using the median parameters of the temperature-dose dependence of the free swelling in 08Kh18N10T austenite steel. It is shown that the consideration of the radiation creep facilitates the stress level reduction, however, it increases the swelling and displacement that makes the prediction assessment of the core baffle form change more conservative as compared with the data without considering the radiation creep.

This paper is a comprehensive experimental report on the neutron radiation effects of distributed optical fiber sensors with enhanced Rayleigh scattering profiles in an in-pile environment. Femtosecond laser direct writing was used to inscribe Type-II modifications in standard telecom fibers and radiation-hardened fibers with fluorine-doped cores. Rayleigh backscattering signals were enhanced for continuous 1.5 m. In-pile lead-out sensors tests were carried out at the MIT Research Reactor for two months, which was operated at a nominal power of 5.7 MW with fast neutron (>0.1 MeV) flux of 1.29 × 1014 n/cm2/s and an in-core temperature of up to 560 °C. Using the Optical Frequency Domain Reflectometry technique, the backscattering profiles of fiber sensors were interrogated with a 3-cm spatial resolution to monitor the temperature profile of the reactor. Results show that laser inscribed Type-II modifications in the form of nanogratings are highly stable against extreme temperature and ionizing radiation. Both standard telecom fibers and radiation-hardened fibers with laser-enhanced Rayleigh profiles can continuously perform distributed temperature measurements over the entire duration of the in-pile testing. Temperature coefficients of sensors and spectral shift quality were studied as functions of total radiation fluence. To the best of our knowledge, we present for the first time, the temperature profile of an operating nuclear reactor core with 3-cm spatial resolution, enabled by distributed fiber sensors with laser-enhanced Rayleigh scattering profiles. The high spatial resolution measurements can provide valuable data for the design and validation of digital twin and virtual reality of nuclear energy systems.

The Purdue Advanced Reactor Core Simulator (PARCS) three-dimensional neutron kinetics code and the TRACE nuclear systems analysis code were interfaced. This provides a best-estimate coupled code sy...

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction β for 239Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Development of dry casks is necessary to support the national strategy for management of spent fuels. One of the requirements for the dry cask is shielding performance for neutron emitted by the spent fuels to be stored in the dry cask. The objectives of this study are to determine the emitted neutrons by the spent fuel generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used in dry cask for the spent fuels. The former was carried out using Origen 2.1 software, while the latter using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor were 2.81×103 and 3.32×106 n/s for reactor core power of 15 and 30 MW, respectively. Addition of Cd improves the neutron shielding performance of SS 316L. The evaluation of neutron shielding performance of SS 316L with addition of Cd which is the candidate material for dry cask of the spent fuels from the GAS research reactor can be evaluated using Origen 2.1 software for neutron emission, while the neutron shielding performance was evaluated by the simulation using MNCP 5 software. This study shows the Cd-SS 316L alloy can be used for further study to develop the dry cask design for the GAS research reactor.Key words: Neutron shielding, cadmium, stainless steel, spent fuel.

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).

In this study, the Doppler reactivity coefficient has been investigated for UO2, MOX, and (Th/U)O2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor keff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO2). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

Passive residual heat removal heat exchanger (PRHR HX), which is a newly designed equipment in the advanced reactors of AP1000 and CAP1400, plays an important role in critical accidental conditions. The primary and secondary side coupling heat transfer characteristics of the passive residual heat removal system (PRHRS) determine the capacity to remove core decay heat during the accidents. Therefore, it is necessary to investigate the heat transfer characteristics and develop applicable heat transfer formulas for optimized design. In the present paper, an overall scaled-down natural circulation loop of PRHRS in AP1000, which comprises a scaled-down in-containment refueling water storage tank (IRWST) and PRHR HX models and a simulator of the reactor core, is built to simulate the natural circulation process in residual heat removal accidents. A series of experiments are conducted to study thermal-hydraulic behaviors in both sides of the miniaturized PRHR HX which is simulated by 12 symmetric arranged C-shape tubes. For the local PRHR HX heat transfer performance, traditional natural convection correlations for both the horizontal and vertical bundles are compared with the experimental data to validate their applicability for the specific heat transfer condition. Moreover, the revised natural convection heat transfer correlations based on the present experimental data are developed for PRHR HX vertical and lower horizontal bundles. This paper provides essential references for the PRHRS operation and further optimized design.

A novel ex-core-detector–based core power reconstruction method is presented. The method uses power correlations between fuel regions and can be applied to a real-time small reactor core monitoring...

Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 Triga nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 Triga reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of nuclear reactors using alternative methods for power monitoring.

The aim of this work is to prepare a neutron noise calculator based on the second order of average current nodal expansion method (ACNEM). Generally, nodal methods have the ability to fulfill the neutronic analysis with adequate precision using coarse meshes as large as a fuel assembly size. But, for the zeroth order of ACNEM, the accuracy of neutronic simulations may not be sufficient when coarse meshes are employed in the reactor core modeling. In this work, the capability of second order ACNEM is extended for solving the neutron diffusion equation in the frequency domain using coarse meshes. For this purpose, two problems are modeled and checked including a slab reactor and 2D BIBLIS PWR. For validating of results, a semi-analytical solution is utilized for 1D test case, and for 2D problem, the results of both forward and adjoint neutron noise calculations are exploited. Numerical results indicate that by increasing the order of method, the errors of frequency dependent coarse mesh solutions are considerably decreased in comparison to the reference. Accordingly, the accuracy of second order ACNEM can be acceptable for the neutron noise calculations by using coarse meshes in the nuclear reactor core.

Nuclear characterization of the spent nuclear fuel in a reactor core is essential, especially in case of severe accidents. The radionuclide inventory and its activity can assist in the management of spent fuel handling, transport, or reprocessing. In this paper, the core of the Fukushima Daiichi Unit-1(FD-U1) accident was modeled using the Monte Carlo code (MCNPX 2.7) linked to the depletion calculation code CINDER'90 and ENDF/B-VII.0 cross-section data library. The isotopic inventory and the activity of the radionuclides for the burned fuel were calculated. The input to the code depends on the previous evolution of the reactor core configurations, dimensions and material of the fuel assemblies, initial uranium enrichment, fuel burn-up, and reactor core operational history. The calculations were validated with experimental measurements which were carried out by the Japan Nuclear Energy Safety Organization (JNES) and verified with published results using ORIGEN2-code by Japan Atomic Energy Agency (JAEA). The masses, activities, specific activities, half-lives, and decay schemes for the actinides and fission products were calculated at the time of the accident and after 50 years of cooling time. The calculations showed that the total activity of the burned fuel in the core at the time of the accident was 9.86E+19Bq and after 50 years was 1.89E+17Bq and the higher inventory concentration in the fuel was dominated by the trans-uranic elements. Also, the specific activity in the core at the time of the accident and after 50 years of cooling time was found to be 1.84E+15Bq/g and 5.86E+12Bq/g, respectively.

This work is an experimental study into the hydrodynamics of the coolant flow in the in-vessel pressure duct of a pressurized water reactor. In the article, an experimental testbench and a nuclear reactor model under investigation are described, the measurement methods are set forth, and the operating variables at which the study was conducted and the obtained results are provided. In the experiments, the mixing of the loop coolant flows inside the model of the nuclear reactor downcomer was simulated. The study was conducted on the high-pressure aerodynamic testbench of Alekseev State Technical University in Nizhny Novgorod. The scale model of the nuclear reactor had the structural components characteristic of loop-type reactor units, such as the annular downcomer and the bottom pressure vessel. The experiments were conducted at Reynolds numbers within the 20 000–50 000 range measured in the annular gap of the downcomer of the model. The axial velocity field at the inlet to the reactor core simulator was investigated using a pneumometric probe. The temperature field was recorded in the experiment by the impurity diffusion method, i.e., by introducing of a contrast tracer into one of the loops of the model. The degree of mixing the flows was estimated by the admixture (tracer) concentration at the inlet to the core simulator. Propane was used as the contrast admixture. The study has yielded the spatial distribution of the tracer in the coolant flow in the annular downcomer and in the bottom pressure vessel. The data on the distribution of the contrast admixture are presented in the form of charts. The swirling of the coolant flow in the in-vessel pressure duct has been analyzed. It has been shown that the mixing intensity in the bottom pressure vessel is affected by the central vortex with the central axis. The parameters of mixing the admixture in the model of the in-vessel pressure duct have been estimated.

Spacer grid with mixing vane had been widely used in nuclear reactor core. One of the main feather of spacer grid with mixing vane was that strong swirl flow was formed after the spacer grid. The swirl flow not only changed the bubble generation in the near wall field, but also affected the bubble behaviors in the center region of the subchannel. The interaction between bubble and the swirl flow was one of the basic phenomena for the two phase flow modeling in fuel assembly. To obatin better understanding on the bubble behaviors in swirl flow, full three dimension numerical simulations were conducted in the present paper. The swirl flow was assumed in the cylindral calculation domain. The bubble interface was captured by Volume Of Fluid (VOF) method. The properties of saturated water and steam at different pressure were applied in the simulation. The bubble trajectory, motion, shape and force were obtained based on the bubble parameters captured by VOF. The simulation cases in the present study included single bubble with different size, at different angular velocity conditions and at different pressure conditions. The results indicated that bubble migrated to the center in swirl flow with spiral motion type. The lateral migration was mainly related to shear stress magnitude and bubble size. The bubble moved toward the center with high velocity when the swirl magnitude was high. The largest bubble had the highest lateral migration velocity in the present study range. The effect of pressure was small when bubble size was the same. The prelimenery simulation result would be beneficial for better understanding complex two phase flow phenomena in fuel assembly with spacer grid.

NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and subcomponents can be manufactured in parallel over time frames of weeks rather than months or years. To meet key program initiatives (e.g., timeline, material use), several constraints—including fissile material availability (less than 250 kg of HALEU), component availabilities, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small (less than one cubic meter in volume) cores with near-term viability. The fast-spectrum designs did not meet the fissile material constraint, so the thermal-spectrum systems became the primary design focus. Since significant progress has been made on advanced moderator materials (YHx) under the TCR program, gas-cooled thermal-spectrum systems using less than 250 kg of HALEU that occupy less than 1 m3 are now feasible. The designs for two of these systems have been evolved and matured. In both thermal-spectrum design concepts, bidirectional coolant flow is used. Coolant flows down through YHx moderator elements and is reversed in a bottom manifold and core support structure, and then flows up though or around the fuel elements. The main difference between the two thermal-spectrum design concepts is the fuel elements—one uses traditional UO2 ceramic fuel, and the other uses UN-bearing TRISO fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems are assessed and summarized herein.

The neutron diffracttion studies of the samples of fuel element claddings made of EK-164 and ChS-600 austenitic steels in the initial state and after operation in a BN-600 fast neutron reactor are reviewed. The samples were investigated in a wide range of fast neutron doses and irradiation temperatures, up to ∼80 displacements per atoms (dpa) and 628°C, respectively. Various microstructural characteristics, such as microstresses, dislocation density, and crystallographic texture are determined using full-profile analysis of neutron diffraction data. At high doses studied in this work, the irradiation temperature is the dominant factor responsible for the dislocation density. The procedures of analyzing and revealing defect types were preliminary developed on specially prepared nickel samples and austenitic alloys upon irradiation in the IVV-2M reactor core at a temperature near 80°С, which provided a low thermal mobility of lattice defects. The results obtained in some first diffraction studies on samples after their operation in the reactor core show that neutron data can be used to characterize the microstructures formed due to irradiation and, at least, estimate the mechanical properties of irradiated materials without their destruction and with no radiation risk for personnel.

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.

The power reactor with high-temperature gas-cooled reactor (HTGR) technology uses uranium as the reactor fuel. The energy from fission is converted to electrical energy or used for other needs such as hydrogen production or other research activities at high temperatures of around 700 °C. This operation does not allow the use of metal as the core material for the reactor. The material that fits the requirements as a core structure is graphite. Graphite material has specific characteristics, namely the parameters of the modulus of elasticity, coefficient of thermal expansion, and the volume which changes due to temperature and neutron dose. Because the structure of the reactor core is a vital component in the reactor, this research will develop a method for the design of the reactor core structure with graphite material. The design method is based on Design by Analysis which specifically refers to the strain analysis on each of the reactor core components. The design method developed is based on the finite element method. The object of this research is the side reflector made from the Toyo Tanso IG-110 series graphite. Based on the analysis of heat distribution and heat stress for the material before the effect of neutron exposure, the temperature distribution on the side reflector was found, as well as the displacement and heat stress that occurs. isotropic properties, Young's modulus and Poisson’s ratio values can be verified and estimated. The purpose of this research is to analyze the strain of the reactor core structure by taking into account the uncertainty of the graphite properties.

The heat generated in reactor core is extracted using the sodium water heat exchanger system in fast breeder reactor. In the heat exchanger system there are 8 Steam generators. Steam Generator uses...

Cesium and iodine, which are formed during a fission process in a nuclear reactor, are considered as major fission products responsible for the environmental burden in case of a nuclear accident. From the safety point of view, it is thus important to understand their release mechanism when overheating of the reactor core occurs. This work presents an experimental investigation of the behaviour of caesium iodide and caesium fluoride in fluoride based molten salt reactor fuel during high temperature events. It has been demonstrated that CsF will be retained in the fuel salt and thus its volatility will be significantly reduced, while CsI will not dissolve in the fluoride-based fuel matrix and will thus remain more volatile. The influence of the presence of CsI and CsF on the melting behaviour of the fuel has been investigated using calorimetry, revealing their negligible effects.