The CABRI International Program (CIP) tests irradiated UO2 or MOX fuels submitted to Reactivity Initiated Accidents (RIA) representative power pulses in prototypical PWR thermal-hydraulic conditions. CIP is managed by The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) within a OECD/NEA framework. Experiments are conducted in the CABRI reactor operated by CEA. For CIP, CABRI benefits from a new pressurized water loop. An important refurbishment program enhanced the facility safety and upgraded the experimental equipment such as the non-destructive examination bench IRIS and the Hodoscope on-line fuel motion monitoring system. Specific test devices with appropriate innovative instrumentation follow the test rod behavior during the transient. The successful first test in the pressurized water loop demonstrated the CABRI capability to perform fully instrumented RIA tests in PWR conditions and to provide highly valuable results for RIA phenomena modelling (boiling crisis, post-failure events), for code validation, and for assessing PWR safety criteria.

In this paper, experiments on breakdown voltages of helium gas in a uniform field have been carried out at a separation from 0.25 to 3.02 mm. The pressure and temperature respectively vary from 1 to 7 MPa, 25 °C to 180 °C, which is the working condition for the main helium blower and its Active Magnetic Bearings (AMBs) in the High Temperature Reactor-Pebble-bed Module (HTR-PM). COMSOL Multiphysics and a self-programmed FD-FCT (Finite Difference-Flux Corrected Transport) code with a two-dimensional drift–diffusion plasma model have been used to simulate the breakdown process and the results agree well with the experimental data. This kind of theoretical and experimental work provides valuable references for helium insulation design of electric devices in HTR-PM and other helium gas-cooled reactors. It also helps understand the mechanisms in helium discharge at high temperature and pressure which is quite different from its low counterpart.

We use a novel nonintrusive adaptive Reduced Order Modeling method to build a reduced model for a molten salt reactor system. Our approach is based on Proper Orthogonal Decomposition combined with locally adaptive sparse grids. Our reduced model captures the effect of 27 model parameters on keff of the system and the spatial distribution of the neutron flux and salt temperature. The reduced model was tested on 1000 random points. The maximum error in multiplication factor was found to be less than 50 pcm and the maximum L2 error in the flux and temperature were less than 1%. Using 472 snapshots, the reduced model was able to simulate any point within the defined range faster than the high-fidelity model by a factor of 5×106. We then employ the reduced model for uncertainty and sensitivity analysis of the selected parameters on keff and the maximum temperature of the system.

The critical power limit for a natural circulation-cooled research reactor core is estimated by utilizing a system analysis code RELAP5/MOD3.3 and an in-house companion program RCPP. The variation range of thermal-hydraulic parameters is found from steady-state and transient simulation results. From literature, potential candidate correlations for critical power estimation were selected and the design limit critical heat flux ratio was evaluated for the chosen correlations based on error statistics. A computational fluid dynamic simulation was carried out to observe the power-equilibrium flow relationship and used to prepare the system code input. To ease a rigorous critical power estimation process, an in-house code was developed to automate batch run and output post-processing of system code. The critical power limit was found for each combination of major operation parameters. A critical power map was generated by merging each power limit value, and the effect of the correlation and core states were studied.

The accident occurred in Japan on 11 March 2011 has revived the interest for the analysis of severe accidents. The scarce and sometimes unreliable data concerning boundary conditions, effectiveness of accident management measures and equipment performance, pose a tough challenge in modelling the accident scenarios. Throughout an analysis of the challenges posed by the Fukushima Daiichi Unit 1 data recorded, this paper describes the major postulates proposed by CIEMAT concerning the equipment and component responses, the effectiveness of accident management actions and the MELCOR model applied. Among most influencing assumptions are those related to the reactor pressure vessel (RPV) leaking pathways and failure mode, the water flow rate entering the reactor, the potential leaking pathways and failure mode and location from the primary containment vessel (PCV) to the reactor building, the corium relocation from RPV to the cavity and its distribution in the PCV, the potential stratification of the suppression pool and the hypotheses made a priori concerning fission product release and transport. Based on the postulated scenario and model, a remarkable agreement of the thermal footprints in terms of RPV and PCV pressures during 500 h has been achieved, in which the RPV and PCV leaks/failures as well as venting played a determining role in the short run of the accident and water injection heavily conditioned the long one. As for the scarce data related to fission products (FP), a consistent agreement is found in the suppression chamber, but estimates in the Dry-Well are about an order of magnitude below measurements despite showing the observed trend. A number of factors might affect FP comparisons to data, from the approximate method to derive dose rates (measurements) from FP masses (MELCOR results) to the RPV and PCV postulated failures. Anyway, based on the data available the set of hypotheses and approximations made seem to make up a defensible scenario for Fukushima Daiichi Unit 1. The studies and results presented in this paper have been achieved under the frame of the OECD/BSAF projects through the CSN-CIEMAT collaboration agreement on severe accidents research.

All available experimental data on isotope shifts and absolute frequency measurements in the optical spectra of Ca I and Ca II are re-evaluated, and from them complete tables of isotope shifts and energy levels of all Ca isotopes from 36 to 52 are derived. A global least-squares fitting of these data was performed. From this fit, the field and mass shift constants for all involved transitions and energy levels were derived along with improved values for differences of mean squared nuclear charge radii.

To further understand and monitor the thermal–hydraulic (TH) status of large domain of pressurized water reactor (PWR) core, an applicable engineering approach with high efficiency and high spatial resolution is critical. Traditional engineering computational fluid dynamics (CFD) computation needs too many computing resources to effectively analyze large domain engineering application for PWR core. In this study, a distributed-parallel (DP) CFD scheme is presented. This scheme separates the large domain into some sub-domains in order to optimize computing time and resources. The design completely retains the complex structures and fine-scale CFD mesh. In addition, it also significantly reduces the computing resources and time, and a fine-scale, full-height CFD analysis can be done in hours for all assemblies. Moreover, important design requirements such as energy consumption ratio, relative computing domain, relative assembly number, and the ranking method of important locations are also designed to optimize the applications of DP scheme to satisfy different engineering demands. The performance of the proposed scheme is evaluated using the CFD computation of a representative region from all 121 assemblies in a PWR core. This research serves to advance the development of engineering CFD computation for PWR core.

The corrugated plate is a vital steam-water separation device in nuclear power plants. Research on steam-water separation mechanism and separation efficiency of corrugated plates has been a hot research direction. In this paper, the film lateral movement distance of the film reflux on the corrugated plate wall is measured by PLIF method. Based on experimental results, calculation equation of film lateral movement distance when the film reflux with good applicability is fitted. The degree of liquid film reflux is characterized by the lateral distance between the apex of the liquid film reflux profile and film mainstream. This parameter is defined as distance of liquid film reflux. Based on the law of conservation of mass, theoretical equation of distance of liquid film reflux is derived. Results show that theoretical results of distance of liquid film reflux agree well with experimental results in small and medium Reynolds number regions (Re < 2625).

A large amount of general steel waste is generated during decommissioning and dismantling of nuclear facilities. Very low contaminated radioactive waste, whose radioactivity is below clearance level, generated from the demolition process may be reused for general use. We examined the feasibility of the clearance verification system for uranium waste. The relative error of uranium determination was within 30 % for 1 g of uranium when measuring steel materials (angle bar, channel steel, pipe steel, square steel tube, fragments of steel tube).

One of the core problems of waste canisters layout in high-level radioactive waste repository is the evolution of the temperature field. On basis of the layered thermal analysis model for single waste canister, the expression of temperature increment at any location in surrounding rock in the repository was obtained by the superposition principle. The initially estimated value of the canister spacing (CS) was determined according to the temperature design criterion. Finally, the influence of relevant parameters on the canister surface temperature (CST) was analyzed. The results were drawn as follows: (a) Taking thermal conductivities of 2.4 and 2.8 W/(m × K) for the rock as examples, the appropriate CS is 12.2 and 13.5 m under the tunnel spacing of 40 m, respectively. (b) The greater the CS, the greater the thermal conductivity of bentonite and rock, the smaller the CST would be. (c) The thicker the buffer layer, the less the heat flux inside the canister would spread out.

To determine heat transfer regimes of the pre and post CHF, the SPACE code calculates the wall temperature from a nucleate boiling heat transfer model at the given CHF. It needs iterations and consumes a large amount of computing time. To reduce the calculation time, this paper introduces the application of a machine learning method. Big data of the wall temperature at CHF was built by using the subprogram constructed as is in the SPACE code. Based on that database, the neural network models were trained and two neural network models having different configurations were suggested. The developed neural network models were implemented in the SPACE code and test calculations were performed. The neural network applied SPACE code properly predicted the wall temperature at CHF. In test calculations, the calculation time was also investigated. All suggested neural network models highly enhanced the calculation speed corresponding to a maximum 86% time reduction.

The energy spectrum of the prompt-neutron flux at the Missouri University of Science and Technology Reactor (MSTR) was obtained using an activation method. Foils were irradiated at the bare rabbit tube (BRT) of the reactor core (which has a 120 W configuration). The neutron spectrum was determined using an unfolding method implemented in SAND-II code. The Monte Carlo N-particle (MCNP) model was used to calculate the spectrum at the bare rabbit tube, which was used as the initial guess for input into the SAND-II code. Various MCNP spectra with 620-, 143-, 89-, 50-, 22-, and 12-energy groups were used as the initial guess for input into SAND-II. Seventeen different foils were irradiated at 100 kW. The foil set covers energies from 0.025 eV to 13 MeV for a broad spectrum analysis. The photon counts for the activated foils were obtained using a high-purity germanium (HPGe) detector. The count results were input into the SAND-II code to predict the neutron flux spectrum for the MSTR. The spectra were collapsed into three groups: thermal, epithermal, and fast flux. With the 620-group initial guess, the thermal, epithermal, and fast neutron fluxes were 1.43×1012±2.82×1011n/cm2s, 4.51×1011±2.85×1010n/cm2s, and 5.38×1011±4.85×1009n/cm2s, respectively, giving a total flux 2.42×1012±3.02×1011n/cm2s. Disparities were noted in the distribution of the thermal and epithermal flux predictions, as the number of groups in the initial spectrum guess changed. The 59%/19% distribution of the thermal/epithermal flux, as predicted with the 620-group guess, is inconsistent with the 49%/26% distribution predicted with the 89- and 143-group guesses. The predictions based on the 89- and 143-group guesses are fairly consistent with the predictions obtained with the 22- and 50-group guesses. The 12-group initial guess resulted in the prediction of a fairly even distribution between the thermal (38%) and epithermal (36%) fluxes. Regardless of the number of groups in the initial guess, the SAND-II prediction is fairly consistent in the fast energy range. The fast neutron flux was found to range between 22% and 26%.

Monte Carlo (MC) shielding calculations often use weight windows (WWs) and biased sources formed from a deterministic estimate of the adjoint flux to improve the convergence rate of tallies. This requires a significant amount of computer memory, which can limit the memory available for high-resolution tally output. A new method is proposed for reducing these memory requirements by using singular value decomposition (SVD) in linear or logarithmic space to approximate the adjoint flux. This method’s performance is evaluated using the Shift and Denovo codes for streaming and diffusion base case problems, followed by problems using the Westinghouse AP1000 and the Joint European Torus. The log SVD reduced WW memory requirements by an order of magnitude in all cases without a significant performance penalty. Additionally, the linear SVD reduced biased source memory requirements by an order of magnitude, but further investigation is needed to account for observed limitations.

Steam separator package including swirl vane separator, gravity and dryer is crucial for guaranteeing steam humidity below 0.1% in SG. In this paper, a high-pressure flow system is designed to study the performance of a full-scale steam separator package at a rated pressure of 6 MPa and temperature of 275.6 °C. To obtain separation efficiency, moisture carryover and pressure drop of each stage, the Reynolds number steam and water are 1 × 106–6 × 106 and 8.27 × 105–3.2 × 106, corresponding to 25–145% load. Results indicate that the separation efficiency ranges from 97 to 99%, 0.5 to 2% and 0.2 to 0.75% for separator, gravity and dryer. The moisture carryover ranges from 2 to 16% and 1 to 4% for separator and gravity, and is below 0.1% at dryer outlet. The pressure drop ranges from 2 to 24 kPa and 0.3 to 3 kPa for separator and dryer. Empirical correlations for swirl vanes are proposed, which are responsible for more than 85% and 70% of total separation efficiency and pressure drop.

The passive residual heat removal system (PRHRS) is adopted by the HTR-10 and HTR-PM to remove the decay heat in the reactor core, so that the safety of the reactor can be guaranteed under accident conditions. A multi-scale CFD-system coupled code used to calculate the transient characteristics of the PRHRS of HTR-10 was developed in this study. With this coupled code, the transient characteristics of the PRHRS during the startup process were calculated. The results show that the computational results are in good agreement with the experimental results, which validates the accuracy of the coupled code. The ability of the coupled code to simultaneously calculate the local heat transfer process in the reactor cavity and the global heat removal characteristic of the PRHRS is also shown. Moreover, radiation heat transfer plays a very important role in the heat removal process of the PRHRS.

Helical coil once-through steam generators (HCOTSGs) have been widely used in the design of small modular reactor (SMR) during the past several decades. As the widely accepted system analysis code, RELAP5 underestimates the heat transfer capability and pressure drop of helical coil tube steam generators using the original geometrical parameters of HCOTSG, because the built-in thermal-hydraulic empirical correlations are only suitable for straight tubes. In this study, thermal-hydraulic models for helical coil tubes and tube bundles are implemented in RELAP5 while the original functions of the code are not affected. A new flag for helical component is proposed and heat transfer boundaries for tube side and shell side are developed. The separate-effect validations of friction factor and heat transfer in helical coil tube are carried out based on experimental data and code-to-code verification. Then, the original RELAP5 and developed RELAP5-HCOTSG are used to simulate helical coil tube steam generators of two SMRs. One is the integral reactor IRIS, and the other is MRX. The calculated results are compared to the design parameters. It shows that the HCOTSG module developed in this study can improve the capability of RELAP5 to predict the thermal-hydraulic characteristics in SMR once-through steam generators with helical coil tubes.

This review gives an update on virtual Compton scattering (VCS) off the nucleon, γ∗N→Nγ, in the low-energy regime. We recall the theoretical formalism related to the generalized polarizabilities (GPs) and model predictions for these observables. We present the GP extraction methods that are used in the experiments: the approach based on the low-energy theorem for VCS and the formalism of Dispersion Relations. We then review the experimental results, with a focus on the progress brought by recent experimental data on proton GPs, and we conclude by some perspectives in the field of VCS at low energy.

In this study, a novel assembly design is proposed for a fluoride salt-cooled high-temperature commercial-scale (FHCR) reactor. It employs tristructural isotropic (TRISO) fuel particles nestled within removable cylindrical beryllium carbide (Be2C) moderator blocks, which are further contained within prismatic graphite blocks. As the name implies, the FHCR is a preconception of a 3400 MW(t) commercial power reactor that uses FLiBe (LiF-BeF2) as the primary coolant of choice. This paper uses the SERPENT 2 code to conduct a parametric neutronics study on the two-dimensional lattice assembly design of the FHCR. The calculations examine the effects of various fuel enrichment levels, TRISO particle packing fractions, fuel compacts’ pitch sizes, and moderating materials on the cycle length, while also determining the neutron spectrum and various nuclide inventories. Finally, a preliminary core is modeled based on the study conducted on the assembly design. Based on the negative values of the fuel temperature coefficients (FTC), moderator temperature coefficients (MTC), and coolant temperature coefficients (CTC) obtained, the design is determined to be safe. This new assembly design is also able to achieve keff>1 for approximately 3.55 years, translating to a burn-up of 160.6 MWd/KgU.

The wire-wrapped assembly could enhance the safety and the economy of the reactor, however the complex structure brings some new problems for thermal-hydraulic analysis. In this paper, the hydraulic analysis of the 7-pin wire-wrapped assembly is carried out by computational fluid dynamics (CFD). The structural grid of the entire fluid domain is generated by the commercial grid software ICEM CFD and solved by CFX. Unlike most studies, this article explores the effect of assembly fuel flow by changing the shape of the fuel duct. Considering the influence of the different shapes of the fuel duct, the pressure drop, cross flow, sub-channel flow distribution are studied. Through this study, it can help to improve the flow environment of peripheral fuel rods and optimize the design of assembly.

Point kinetics equations (PKEs) is an significant model used to describe dynamic behaviors of neutron in nuclear reactors. How to cope with its stiffness problems is very important to solve PKEs accurately and efficiently. In this paper, a new semi-analytical algorithm based on Euler-Maclaurin Approximation (SAEMA) is developed to solve PKEs. SAEMA algorithm is applied and tested in different reactors with three typical reactivity insertion cases, including step, ramp and sinusoidal insertions. Firstly, the investigations on the computational stability are performed under different time step sizes. Secondly, the performances on computational accuracy are evaluated by comparing the results by SAEMA algorithm with the ones by analytical method and excellent CATS algorithm. Finally, by comparing the CPU time consumed by SAEMA algorithm with the physical time, the studies on the computational efficiency are also carried out. Just as results shows, SAEMA algorithm is reasonably an attractive way to solve the PKEs.

In this paper, a new approach aimed at the Fault Diagnosis with Full-scope Simulator based on the State Information Imaging (FDFSSII) in NPP is proposed. The FDFSSII approach first constructs a series of gray-image which presents the operating transient (included normal and fault condition) according to the real time monitoring data. Furthermore, the Machine Learning (ML) technology is employed to achieve image feature extraction and classification by analyzing and learning from massive amounts of historical and synthetic gray-image data – the image feature is extracted by the Kernel Principal Component Analysis (KPCA) and classified by the designed classifiers in different learning methods. Finally, diagnosis effect is evaluated by the F1 score. The simulation result shows that the FDFSSII approach has achieved good effect for the fault diagnosis in NPP. Meanwhile, it simplifies the process of nuclear reactor with the large monitoring data and provides useful support information to the operators.

In this paper coupled calculations with the CFD code ANSYS-CFX-19.0 and the Monte Carlo neutronics code MCNP6 were performed to analyze the heat transfer in supercritical water flowing through the typical fuel assembly of the high-performance light water reactor (HPLWR), in order to improve the characterization of the heat transfer phenomena in supercritical water under non-uniform axial heat flux distributions that is characteristic of this type of reactor. To check the capability of the CFX model to predict the thermal-hydraulic behavior of supercritical water, the computational results were compared with two experimental data. The Shitsmańs experiment in the presence of heat transfer deterioration (HTD) using four low-Re turbulence models (SST, k-ω, BSL-k-ω, and ω-Reynolds Stress) and the Wanǵs experiment in absence of HTD, using the low-Re-SST and the scalable-wall-function-SSG turbulence models. In the presence of the HTD phenomenon, results showed the high dependency of the wall temperature with the turbulence model and the turbulent Prandtl number selected. In the absence of HTD, both turbulence models studied adequately predicted the behavior of the wall temperature distribution. For the coupled neutronic/thermal-hydraulic analysis of the typical HPLWR fuel assembly, the low-Re SST turbulence model and the Prt = 1.5 were used. Different axial profiles of heat flux generated in the fuel rods were obtained for the different power values studied. For the analyzed conditions, the presence of HTD in the lower zone of the fuel assembly was observed. In addition, the results showed a strong non-uniformity of the circumferential surface cladding temperature distribution in the sub-channel located at the corner of the fuel assembly; a new curvature radius of the assembly box corner was proposed to obtain a well homogenized circumferential wall temperature distribution.

The rod drop experiment is an important transient in reactor operation. This study focuses on a comparison between experimental reactor physics data and respective calculation. The rod drop experiment was performed at the VR-1 reactor. A full core 3D model was used to calculate this transient using Serpent2. Additionally, a point kinetic solution is presented. The Serpent results were compared with the experiment and with the point kinetic calculation. Two different nuclear data libraries (ENDF/B-VIII.0, JEFF3.3) were used in the data sensitivity analysis. The Serpent 3D kinetics results were almost identical to the point kinetic equation solution. Experimental data differed in the rod worth. This study shows that the new Serpent dynamic toolkit provides an accurate description of the reactor behavior during this transient. The 3D calculation proved that the detector position during the transient has a direct impact on the measured control rod worth.

Extended Kalman filter (EKF) and cubature Kalman filter (CKF) are proposed to estimate the state parameters of an external neutron source driven sub-critical reactor, including power level, reactivity, external neutron source, six-groups of delayed neutrons precursor densities, equivalent fuel temperature, average coolant temperature and nuclear densities of iodine, xenon, promethium, samarium nuclides. Parameter settings and matters needed attention in EKF and CKF are also analyzed, especially the relationship between model prediction covariance matrix and measurement covariance matrix. In order to effectively identify the maneuvering of the external neutron source and reactivity on the uncertainty of the prediction model, two adaptive algorithms are proposed to adjust the covariance matrix of the prediction model online. The results show that these two adaptive algorithms can effectively detect various maneuvering such as neutron source variation and reactivity insertion, and realize the optimal estimation of the reactor state using EKF method. However, CKF has divergence and non-convergence. EKF achieves good results in all parameters estimation.

Flapper Spurious opening during reactor operation is one of the most perilous accidents for pool-type research reactors. In TRR during such event, if none of SCRAMs could shut-down reactor, partially fuel elements burn-out is occurred. To avoid that, a core cooling condition monitoring system is developed. There are also other PIEs, i.e. fuel element coolant inlet blockage or unintended fuel withdrawal, which can be detected by core DP. In flapper opening scenario, the experiments show core DP response is very tangible. About two other scenarios, results indicate core DP is capable to detect the abnormality in the core flow, but maybe some other supplementary data and signal are required to distinguish the accident. Preliminary data from the system exploitation during reactor operation are analyzed to take into account the uncertainties of various normal operation conditions and determine a minimum allowed working window for the system safety signal.

The neutronics calculation of the PB-FHR with the fraction ranging from 5% to 30% are carried out by the CLS method and ERM method. It can be observed that the infinite multiplication factor obtained with the CLS method are consistently smaller than those obtained with the ERM method when the fraction is less than 15%. Moreover, the differences of the two methods basically increases with the increasing burnup level when the fraction is less than 15%. For the fraction up to 30%, the CLS method shows significant bias of 520 pcm with the explicit method. After the CLS method with a modified packing fraction correction is adopted, the largest difference drops to 220 pcm when the fraction is 30%.

In the present paper, CFD simulations related to the operating conditions considered during the experimental campaign on CIRCE-HERO facility are presented, with the aim of investigating the observed temperature stratification phenomena. Calculations are performed using the commercial codes STAR-CCM+ and ANSYS Fluent adopting a RANS approach; the numerical results and the experimental data are compared. Four distinct experimental tests are investigated also performing sensitivity analyses regarding the boundary conditions. In particular, assumptions concerning the heat losses distribution and the shape of the pool inlet were taken into account. The numerical results provide support for further understanding of the involved phenomena, suggesting the possible causes of the thermal stratification observed experimentally inside the pool. Similar trends for the predicted and experimental data were obtained and – even from a quantitative point of view - the observed discrepancies can be considered acceptable, assuming the uncertainties in the experimental boundary conditions and measurement.

A new design integrating heat-pipes into a nuclear cooling system is presented. The heat-pipes are presented as the primary mode of heat transfer. Analyzing the prevailing limits to determine suitability for predicting the performance of the heat-pipe concept. When the limits are determined for the design integration, steady-state behavior needs to be quantified. A model that accounts for 3D behavior is evaluated for use. Using the limits to evaluate the design integration, with sodium, the operating regime would still remain below the predicted limits. With potassium the operating regime would exceed the capillary limit. This is caused by the increase in pressure drop. With the 3D model, a validation shows that conduction can give very good results for both transient and steady-state behavior for sodium. It shows that water has poor transient prediction but accurately predicts the steady-state behavior. Both solutions were close to the reported experimental results for steady-state.

The relativistic multiconfiguration Dirac–Hartree–Fock (MCDHF) method has been employed to calculate the atomic parameters of Mo XVIII, which are important for fusion determination of plasma properties in different conditions. Atomic data, such as energy levels, lifetimes, wavelengths, spectroscopic labels, and transition rates for the transitions among the lowest 112 states of the 3 s23p63d7, 3 s23p53d8, 3s3 p63d8, and 3s23p63d64s configurations are given. To describe the atomic system accurately, electron correlation effects are taken into account. We also present a calculation within fully relativistic frame based on the Flexible Atomic Code (FAC). This, in turn, allowed us to make an intercomparison on the obtained data. Our two sets of results are also evaluated by comparison with the NIST database recommended values. The new calculated energy levels are, on the average, about 1.5% higher than NIST database recommended values and yield oscillator strengths within 5% of the theoretical values. This study provides a substantial amount of updated atomic data, which are essential for fusion applications.

CsI and CdI2 aerosol decomposition rate under irradiation has been quantified at 80 °C and 120 °C in presence of humidity and on different substrate (stainless steel, quartz and Epoxy paint). A model has been developed for the ASTEC-SOPHAEROS code to reproduce the data and help the identification of the gaps remaining in the understanding of iodine volatility in a severe accident of a Nuclear Power Plant (NPP). The current model applied to model the gaseous iodine behaviour in the containment of PHEBUS-FP tests does not fit with the experimental data probably because the nuclear aerosol reaching the containment are much more complex than pure CsI aerosols. It has been clearly shown than the radiolytic oxidation of metallic iodide aerosols into molecular iodine can significantly impact the source term evaluation even if additional experimental data area required to cover the variety and complexity of nuclear iodide aerosols.

Beta+-delayed proton (or α) emission is a typical decay mode of very neutron-deficient nuclei. Valuable information for the ground state in the precursor, such as half-life, spin, and parity, can be obtained by studying the β+-p decay properties. The high efficiency and unique experimental signature for detecting protons allow one to study states in the β +-decay daughter that are not accessible through other means. By measuring the properties of protons emitted to a known state in the daughter, information on the structure of the proton-unbound state can be obtained. The known nuclei that exhibit this decay mode are evaluated to give the recommended values for the nuclear properties of these nuclei. This includes branching ratios, and half-lives. In addition for those nuclei with known resolved proton transitions, proton energies, intensities, and the energies of the proton-emitting states are compiled. A list of experimental references for each β +-p precursor is also given. All papers published prior to June 2019 have been considered in adopting the properties given in this work.

The maximum coolant temperature in the VVR-Ts MAK-2 loop facility for the production of 99Mo was analyzed. The calculations showed that adequate accuracy and temperature 96.3°C are attained using more than 106 grid elements. Taking into account the temperature dependence of the thermal conductivity of the uranium-containing layer increases the maximum temperature in this layer by more than 10%. When boiling, vaporization, and turbulence are taken into account the temperature at the egress from the coolant changes within the error range as compared with the use of a simplified model.

Experiments performed at the Mayak Production Assiciation to determine the 232U content in spent BN-600 fuel are analyzed. Calculations show that to within the limits of error the concentration largely agrees with the measured value, which makes it possible to use the calculations for a preliminary assessment of the 232U content in batches of uranium regenerated from reprocessed spent BN-600 FA.

The design of the TUK-109T large-size container developed at the All-Russia Research Institute of Experimental Physics (VNIIEF) for transporting ampuls with bundles of spent fuel pins is described. The results of full-scale tests and three-dimensional simulation modeling for strength validation for mechanical actions occurring under normal operating conditions and during emergencies are presented. The computational validation of strength was performed in a three-dimensional formulation based on a finite-element model and a validated methodology of modeling dynamic deformation of structures, implemented in the highly parallelized LOGOS software. The validation of the methodology was performed by comparing numerical results with full-scale design tests. The computational prediction is shown to be highly accurate.

The operating time of the IBR-2 reactor depends on the fuel life. Currently, an IBR-2 run is formed at the cost of adding fresh FAs into the core without any reshuffling of the FA. In spite of its technological simplicity, this approach results in suboptimal fuel use because of high burnup non-uniformity. In the present work we propose optimizing a reactor fuel run by reshuffling FA during the run. This will give more uniform fuel burnup and increase the run time by almost 1/4 while limiting the ultimate burnup as compared with the conventional approach to run formation without reshuffling.

The circulation flow in a centrifuge during enrichment of isotopic mixtures of heavy gases is determined. Nickel tetraphosphine Ni(PF3)4 was used as the working substance to study the hydraulic and separative characteristics of a centrifuge. The relative circulation parameter m characterizing the efficiency of the multiplication of the primary effect is determined by comparing the computed dependences with the experimental results obtained in this work.

The results of an analysis of the thermal impact of the circulating cooling water discharged after cooling equipment in the Kalinin NPP on the thermic regime of the cooling reservoir and the climate in the littoral areas are presented. The impact was evaluated by comparing the hydrometeorological characteristics before the commissioning of the nuclear plant (1971–1983) and during its operation at full power (2013–2016) with similar characteristics in the background areas. It is shown that the impact of the Kalinin NPP on the littoral climate is very small and manifests only in a strip 200–300 m from the water edge. The main impact of the plant was to increase the water temperature in the cooling water reservoir and to increase evaporation from the surface of the reservoir.

The exportation of fast reactors under an international nuclear weapons non-proliferation regime is discussed. Russian sodium fast-reactor technology has been successfully demonstrated and is in the commercialization stage. Fast reactors with lead as coolant are being developed in Project Breakthrough. In this connection, the possibility of exporting fast reactors becomes logical. The main criterion for such exports is competitiveness of the electricity produced taking the possibility of offering other, additional services into account. Other criteria that undoubtedly will influence the decisions made about exports are elements associated with the need for strict adherence to the international nuclear weapons non-proliferation regime. The basis of this regime is the nuclear weapons non-proliferation treaty. The IAEA system of safeguards plays a key role as a control element in verifying the obligations of non-nuclear states in the sphere of nuclear non-proliferation.

The trajectory of a radioactive cloud formed above the Dunai Peninsula, Ussuriiskii Zaliv, Primorskii Krai, and the border area of China as a result of a nuclear accident in Bukhta Chazma in 1985 was reconstructed. The population–cloud contact time and the radioactive fallout density of the primary dose-forming radionuclides are estimated and the obtained results are compared with the admissible norms and the background. As a result of radioactive decay, scattering, and wash-out by atmospheric precipitation the volume activity of the dose-forming radionuclides in the ground layer of the atmosphere decreased by a factor of 105–106 over the passage time of the radioactive cloud starting from 1 km away from the site of the accident to the tip of Ussuriiskii Zaliv (60 km) and by a factor of 107–108 up to 400 km. The greatest contamination of Ussuriiskii Zaliv was noted in the uninhabited Cape Sedlovidnyi, located near the coastal radioactive track formed above the Dunai Peninsula after the accident. The radioactive fallout density increased by 3.8 kBq/m2 and the dose rate by 0.03 μSv/h, and in all other locations these indicators did not exceed 0.7–1 kBq/m2 and 0.01 μSv/h, respectively. These values turned out to be lower than the global fallout of 137Cs – 4.4 ± 0.7 kBq/m2 and the natural γ-background 0.14 ± 0.06 μSv/h. On the territory of Primorskii Krai and in the border area of China, the contamination did not exceed 0.01–1% of the background. At all stages of the accident, 60Co determined 95–99% of the long-time contamination of the environment.

The SKP-KhOT system for measuring subcriticality, making it possible to determine Keff of jackets with spent FA VVER-1000 during normal operation as well as design basis and beyond-design basis accidents, has been installed in the KhOT-1 storage facility at the Mining and Chemical Combine. The results of its tests have been recognized as positive; the design basis subcriticality of the studied jackets falls within the confidence interval of the measured value. It is shown that Keff of a jacket does not depend on its environment.

The automated system ASCRO developed and adopted in the nuclear industry for monitoring radiation conditions and on-site monitoring of subsoil status is continually improving, and the Rosatom State Corporation is continually expanding the sphere of activity of these systems. In the present article, the experience gained in the implementation of on-site monitoring of subsoil status in 2017 and the principles of its integration into the complex environmental monitoring system are expounded. Data on the development and integration of information-geoenvironmental packets of nuclear industry enterprises into the complex environmental monitoring system are presented.

The influence of the active-element characteristics on the efficiency of a cadmium-vapor laser pumped by uranium fission fragments was studied experimentally. The 441.6 nm laser radiation of the cadmium ion was studied. An aluminum tube with diameter ~15 mm and an inner coating of U3O8 with thickness 3 mg/cm2 served as the active element. The neutron source was the BARS-6 pulsed, fast, aperiodic, self-extinguishing reactor. A comparison of the efficiency of the laser with similar data previously obtained under the same conditions for a 48 mm in diameter active element with a ~10 mg/cm2 thick uranium metal coating shows that a three-fold reduction of the diameter of the active element and uranium coating thickness results in a 2.5-fold increase of the laser efficiency. This confirms the hypothesis of internal losses of laser radiation in the active element of a fission-fragment pumped laser which are due to the scattering of light by transverse sound waves of the gas density which appear during a pumping pulse because of the radial non-uniformity of the energy input.

An approach to developing the NPP system with the lowest integral risk determined according to the entire life cycle and the conjugate infrastructure is studied. The search for the optimal power series of reactors for the developing NPP system is formalized and presented as a multicriterion optimization problem: reduction of harm from various risk factors and construction cost for the specified system capacity. A solution is found with the aid of an evolutionary algorithm. Recommendations are made on the basis of the calculations for optimizing the power range of reactors that will make it possible to reduce the aggregate risk of NPP operation with different approaches to strategic planning.

In this paper a probabilistic methodology based on core nodalization is proposed to estimate the core power in the presence of xenon oscillation. A time-dependent Monte Carlo neutron transport code named MCSP-NOD is developed for dynamic analysis in arbitrary 3D geometries to simulate xenon oscillations as well as sub-critical condition with feedbacks. The new code is based on the approach adopted in MCNP-NOD which was previously introduced as a tool for core transient analysis using the MCNPX platform. As before, the core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates, leakage ratio are estimated using the MC techniques. However, as a new option the concentration of iodine and xenon are also estimated which enables us to predict the oscillatory behavior of reactor power following poison oscillation. These quantities are then employed within the time-dependent neutron transport equation for each node independently to compute the neutron population. Simulations prove the robustness of the method.

The potential reasons for the corium spreading difference over inert ceramic and reactive concrete channels of VULCANO VE-U7 experiment are investigated using Moving Particle Semi-implicit (MPS) method. A new thermal contact resistance model has been developed for MPS so that influence of the subscale heat transfer between the melt/crust and the substrate on spreading can be considered. The results indicate that the spreading difference is not much influenced by heat loss of the melt to different substrates, but more likely due to gas bubbles in the concrete channel. The most likely responsible gas bubble effect could not be well identified with the single channel analysis, because it could not consider the inflow mass interactions at the stabilization pool. The double-channel analysis with such consideration indicated enhancement of the effective thermal conductivity of the melt as the key influence of the gas bubbles that led to the difference.

Deformation and contact force of the fuel assembly due to thermal gradient are of great concerns to the integrity design and safe operation of the core in the China Fast Reactor (CFR). Up to now, only the 2D deflection measurement of the sub-assemblies has been investigated, and almost no attention is paid to the contact force measurement between the sub-assemblies with a small clearance. In this paper, the measurement method and system are developed for measuring the deformation and contact force of the fuel assembly under thermal gradient with the test fuel assembly simplified based on the stiffness equivalent strategy. In addition, the 3D deformation and contact force are measured through the non-contact industrial photogrammetry system and the thin film pressure sensor or strain gauge, respectively. Free thermal bowing test for a single assembly is performed at first for verifying the reliability of the 3D deformation measurement by comparing with the results of finite element (FE) simulation. Afterwards, the 3D deformation and the contact force are measured for the single assembly restrained thermal bowing test. The accuracy of the proposed contact force measurement is investigated through comparison with the method of spoke-type force sensors. The developed measurement method and system can provide experimental basis for safety design of the CFR due to its potential for comprehensive deformation and contact force measurements.

Suppression pools are an essential passive system for source term attenuation in boiling water reactors during severe accidents, particularly during Station BlackOut (SBO) sequences, as it happened in Fukushima. This paper investigates how uncertain predictions of suppression pools decontamination can be. Based on MELCOR 2.1 calculations of Fukushima Unit 1, a stand-alone version of SPARC-90 (Suppression Pool Aerosol Removal Code) has been used in combination with DAKOTA-6.4, to propagate the uncertainties in the input deck variables affecting the Decontamination Factor (DF). The results indicate that DF uncertainties may spread around two orders of magnitude and the uncertainty margin stays roughly constant over time. In addition, a sensitivity analysis based on the Pearson and Spearman correlation coefficients has been carried out and pointed that uncertainties associated to particle inertia (i.e., particle density and size) and in-pool phase change (i.e., non-condensible gas fraction in the carrier gas) dominate the uncertainties found in the DF for this specific scenario.

The differential operator method (DOM) has been developed to perform sensitivity analyses of generalized responses in the form of reaction rate ratios. In this method, the memory consumption required to store the source perturbation effect will become prohibitively large with a large number of particle histories. This work introduces the superhistory-based differential operator method (SH-DOM) to reduce the memory usage. In the superhistory algorithm, the source perturbation effect is estimated by following the source particle and its progenies over super-generations within a single particle history, which significantly reduces the memory usage. The new method is verified via the Jezebel, Flattop and the UAM TMI PWR pin cell benchmark problems calculated by the collision history-based method and the SH-DOM. Results show that the energy integrated sensitivity coefficients given by the present method agree within 5% with those of the collision history-based method and the SH-DOM can effectively reduce the memory consumption.

Comprehensive knowledge of neutron spectrum is significant in reactor design. Online wide-range neutron spectrum unfolding technology still requires improvement in accuracy and efficiency. In the work, a “two-step” neutron spectrum unfolding method based on artificial neural network (ANN) was developed to unfold spectrum with wide energy range. First, a default spectrum was reconstructed by using the ANN model which had been trained with a large amount of neutron spectra generated from Monte Carlo transport calculation. Second, the default spectrum was optimized by using iteration algorithm. The two-step method was verified with a thermal neutron reactor VENUS-3 and a fast neutron reactor BN-600. Comparison of mean square error (MSE) between this method and the traditional unfolding method showed reduction of 83.4% and 85.6% on VENUS-3 and BN-600 respectively, and average relative deviation (ARD) reduction of 89.3% and 86.1% respectively. Also, comparison of spectrum quality (Qs) showed reduction of 83.4% and 86.0% respectively for the two cases. This work demonstrated that the developed two-step method could obtain the better accuracy than traditional method.

A new method for detecting a gas–water interface in a circular pipe is proposed. In the method, ultrasonic signals are used for non-intrusive measurement and three types of signal analyses are conducted: time-of-flight (TOF), local amplitude, and global amplitude analyses. Horizontal, 45° inclined, and vertical pipe configurations were used to verify the applicability of the proposed detection method. In the case of a horizontal pipe with an acoustic beam directed perpendicular to the water surface, TOF and amplitude analyses detect the water level. In the cases of a horizontal pipe with an acoustic beam directed parallel to the water surface, a 45° inclined pipe, and a vertical pipe, when the pipes were filled with water, TOF analysis was not applicable and amplitude analysis detects the water level. In conclusion, the gas–liquid interface in circular pipes could be analyzed qualitatively and quantitatively through the proposed non-intrusive acoustic method.

In order to give a clear picture of key parameters uncertainties introduced from fuel pebble model in HTR and its effect on the uncertainty propagation, herein different methods of sampling the position of a random dispersed TRISO particle in the fuel region were studied. And total ten different fuel pebble models including high-fidelity models with random dispersed fuel particles and homogenized models were built to quantify the uncertainty of eigenvalue introduced by fuel pebble models. Meanwhile, the effect of fuel pebble model on the uncertainty propagation of nuclear data was also investigated. The numerical results indicate that the fuel pebble models introduce a great model uncertainty to the calculated multiplication factor and also have a significant effect on the uncertainty propagation of nuclear data. However, the uncertainty of the multiplication factors due to the random distribution of TRISO particles is relatively small compared with the uncertainty propagated from nuclear data.

Best estimate plus uncertainty (BEPU) analysis has been widely adopted for safety evaluation of the nuclear reactor, and the sensitivity analysis is an important part of the BEPU methodology. Local sensitivity analysis methods are still widely utilized in the methodology, which is not suitable for complex non-linear nuclear systems. The global sensitivity analysis is essential but the biggest problem is that the computational cost can hardly be accepted. In this study, the moment-independent global method was adopted and optimized, and a low-cost method was obtained and assessed. The sensitivity analysis of a large break loss of coolant accident (LBLOCA) of the LOFT facility was carried out by using the method. Results show that the optimized method can well evaluate the sensitivity indices with very low cost and relatively high accuracy, and the result obtained through the optimized method is much more reliable than that of the local sensitivity analysis method.

This paper presents the verification and validation elements of the UNIST in-house Monte Carlo code, MCS, for the multi-cycle and multi-physics analyses of high-fidelity, large-scale commercial pressurized water reactors (PWRs). Analysis on the neutronic performance with thermal/hydraulic (T/H) feedback is the key to detecting the complex behavior of an operating nuclear power reactor. The MCS solutions with T/H feedback, TH1D, of the consortium for advanced simulation of light water reactors (CASL) virtual environment for reactor applications (VERA) core physics benchmark progression problems 6 and 7 showed excellent agreement in eigenvalues, temperature and power profiles with the MC21/COBRA-IE, MC21/CTF and VERA-CS solutions for the single assembly and whole core of Watts Bar Nuclear 1 (WBN1) Cycle 1 under beginning-of-cycle and hot-full-power condition. Furthermore, the core depletion analysis is one of the most compelling advances for reactor analysis. Therefore, this work is significantly focused on the nuclide depletion simulation coupled with TH1D of the first two cycles of WBN1 to address the VERA core physics benchmark problems 9 and 10. MCS is one of the few Monte Carlo codes that have the capability of depletion calculation for both WBN1 Cycles 1 and 2. The accuracy of MCS simulation of WBN1 Cycle 1 is within 40 ppm and 30 ppm in critical boron concentration (CBC) for all burnup points, compared with the measured data and VERA-CS solutions, respectively. To demonstrate the multi-cycle refueling capability of MCS, the WBN1 Cycle 2 is simulated and compared with the solutions of VERA-CS only, because of the lack of available measured data. MCS shows excellent agreement compared with VERA-CS within 30 ppm in CBC, and the average bias for the entire Cycle 2 is approximately 20 ppm. These results provide confidence in MCS’s capability in high-fidelity, multi-cycle calculations of the practical PWR core.

It is important to ascertain the mechanisms underlying the eutectic reaction between different reactor materials that might be encountered during a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR) since such reactions will affect the accurate progression of a severe accident. In this study, motivated to understand the characteristics of the probable eutectic reaction between fuel debris and the lower head of reactor vessel, a series of simulated experiments has been conducted at the Sun Yat-sen University using a couple of rather lower-eutectic-point (456 K) materials (namely Sn particles and Pb pellet). The experiments were carried out in a self-designed experimental system, which mainly consists of a sample holder and a visible resistance furnace. To acquire a relatively comprehensive understanding, a variety of experimental parameters such as the reaction temperature (463–483 K), contact pressure (0.4–1.2 MPa), Pb pellet diameter (10–25 mm) along with the diameter (0.3–3 mm) and geometry (spherical, cylindrical and droplet-shaped) of the Sn particles have been taken. Through detailed analyses, it is found that the reaction temperature and contact pressure can have noticeable positive impact on the reaction rate. As for the size of Pb pellet and Sn particles, with increasing the diameter ratio of Sn particles to Pb pellet, a non-monotonous effect is observed due to the competing role between the contact area and contact pressure. An evident influence of Sn particle geometry on reaction rate has been verified in accordance with the variation of particle-bed porosity. The analyses in this work also suggest that the reaction rate in previous experiments using block-block samples is generally larger than present experiments using particle-pellet ones, especially at a higher temperature. Knowledge and evidence obtained from this work will be utilized for the design of future high-temperature experiments using actual reactor materials as well as for the improved validations of eutectic-reaction-related models incorporated in fast reactor severe accident codes.

A new Monte Carlo (MC) neutron/photon transport code, called MCS, has been developed at Ulsan National Institute of Science and Technology (UNIST) with the aim of performing the high-fidelity multi-physics simulation of large-scale power reactors, especially pressurized water reactors (PWR). The high-fidelity multi-physics analysis of large-scale PWR is a challenging problem due to two aspects, the first being the difficulty of implementing various state of the art techniques into a single code system, and the other making it feasible to run such simulations on practical computing machines within reasonable amount of memory usage and computing time. In this paper, features implemented into MCS for large-scale PWR simulations are described including but not limited to depletion, thermal/hydraulics coupling, fuel performance coupling, equilibrium xenon, on-the-fly neutron cross-section Doppler broadening, and critical boron search. The efficient memory usage for burnup simulation and the high performance of MCS through various algorithms and optimizations (parallel fission bank, hash indexing) are illustrated on Monte Carlo performance benchmarks. Finally, the large-scale PWR analysis capability is fully demonstrated with BEAVRS Cycles 1 & 2 calculations.

Reactions with radioactive nuclear beams at relativistic energies have opened new doors to clarify the mechanisms of stellar evolution and cataclysmic events involving stars and during the big bang epoch. Numerous nuclear reactions of astrophysical interest cannot be assessed directly in laboratory experiments. Ironically, some of the information needed to describe such reactions, at extremely low energies (e.g., keVs), can only be studied on Earth by using relativistic collisions between heavy ions at GeV energies. In this contribution, we make a short review of experiments with relativistic radioactive beams and of the theoretical methods needed to understand the physics of stars, adding to the knowledge inferred from astronomical observations. We continue by introducing a more detailed description of how the use of relativistic radioactive beams can help to solve astrophysical puzzles and several successful experimental methods. State-of-the-art theories are discussed at some length with the purpose of helping us understand the experimental results reported. The review is not complete and we have focused most of it to traditional methods aiming at the determination of the equation of state of symmetric and asymmetric nuclear matter and the role of the symmetry energy. Whenever possible, under the limitations of our present understanding of experimental data and theory, we try to pinpoint the information still missing to further understand how stars evolve, explode, and how their internal structure might be. We try to convey the idea that in order to improve microscopic theories for many-body calculations, nuclear structure, nuclear reactions, and astrophysics, and in order to constrain and allow for convergence of our understanding of stars, we still need considerable improvements in terms of accuracy of experiments and the development of new and dedicated nuclear facilities to study relativistic reactions with radioactive beams.

Boiling heat transfer over tube bundles has been extensively applied to various industries with a high demand for efficient heat transfer. This work presents a review of recently published studies on pre-CHF boiling heat transfer across plain and enhanced tube bundles. Bundle effect and heat transfer enhancement by modified heating surfaces under various operating and geometric parameters are analyzed. Flow regime maps for boiling two-phase flow in horizontal and vertical bundles are critically described separately. The local boiling heat transfer performance is affected by the non-uniform heat flux distribution in a bundle. A decreasing heat flux distribution along the bundle height can enhance the bundle effect. The effect of the pitch to diameter ratio on bundle effect also depends on the heat flux distribution. Significant influences of the bundle inclination angle and elevation angle on the boiling heat transfer were observed by researchers. Complex bundle effect was found in special shape bundles, such as V-shape, C-shape, and U-shape bundles, which suggests applying different HTC correlations to different regions in a bundle. Moreover, the bundle boiling behaviors under sub-atmospheric and sub-critical pressures have been examined. The heat transfer performance in tube bundles with enhanced surfaces is significantly impacted by the surface characteristics and the imposed heat flux. Bundle effect is still prominent, and the surface enhancement reduces along the bundle height. A mixed bundle with enhanced tubes only in the lower part can achieve the same heat transfer performance as a fully enhanced bundle.

In this paper, the close process of contra-motion check valve in countercurrent pressure difference is studied in nuclear power plant experimental circuit. The transient mathematical model is established. Through the dynamic grid technique, transient flow field in different conditions of valve closing are calculated, and dynamic properties of valve head closing process is analyzed. Through the CFD transient analysis, inherent damping principle and inhibition of water hammer principle are revealed. By writing water hammer analysis, the inhibitation ability of water hammer in typical six conditions is analyzed. At last the performance verification experiments show that contra-motion check valve can effectively inhibit water hammer in the process of parallel double pump switching transition.

To provide a passive cooling system for the reactor containment, Passive Containment Cooling System (PCCS) was adopted in the design of i-POWER nuclear power plant. This study focused on validation tests for condensation heat transfer of the PCCS heat exchanger using the CLASSIC facility. The tests include investigation of the condensation heat transfer in prototypic single tube and bundle tubes. From the single tube experiments, condensation heat transfer model was proposed to reflect the PCCS heat exchanger tube geometry. Experimental results in the bundle tube show consistent trend compared to the proposed heat transfer model from the single tube test. The local condensation heat transfer coefficient of inside tubes was smaller than the average value due to a shadow effect by a larger mass fraction of non-condensable gas, so that design of the PCCS should take into account degradation of the condensation heat removal in the bundle geometry.

UO2 pellets containing 5 vol% diamond particles were irradiated in the Advanced Test Reactor at an LHGR of up to 310 W/cm to the burnup of 7.2 GWd/tHM. Fission gas release, measured during post-irradiation examination, was determined to be 1.08%. Fission gas release modeling performed using BISON fuel performance code showed that undoped UO2 fuel irradiated under the identical conditions, would have had fission gas release of 9.09%. Noting that diamond has the highest thermal conductivity of any known material, these results suggest that doping with a good thermal conductor is an effective means to reduce fission gas release in UO2 fuels.