Liquid droplet radiator (LDR) is a promising cooling system for the high-power spacecraft. The study of the droplet radiation heat transfer mechanism is vitally important for the optimal design of the whole LDR system. Firstly, the radiation heat transfer characteristics of the isolated droplet are analyzed. Then, the Monte Carlo method is used to calculate the angel factor for multiple droplets interacting

The decommissioning of nuclear facilities in nuclear power plants require a rapid dose estimated method. In the past, dose estimation methods include the deterministic and stochastic methods, which have some defects that polarize the computation overhead or accuracy. To address this, we proposed the Monte Carlo–point kernel (MC-PK) coupling method. This paper establishes a nuclear power plant model

In BWR fuel assembly design, the axial enrichments require repeated adjustment for the optimization of the design. Every time the axial enrichments change, a lot of effort has to be spent to perform lattice calculations for generating a new cross section data library. The inconvenient situation can be circumvented by using scoping libraries that enable the two-group constants interpolation functionality

In the safety oriented nuclear engineering world, managing uncertainties on fundamental parameters is crucial. Large uncertainties in the neutron cross sections of materials used in these systems propagate through the modeling process and result in large uncertainties in the predicted behavior of the system. Without reducing the uncertainties on the input neutron cross sections by evaluating new experimental

A serious threat to the integrity of the reactor core, reactor coolant system, or containment is incurred if proper and essential actions to mitigate accidents cannot be taken owing to insufficient information about the internal states of the nuclear power plant (NPP). Therefore, this study was carried out to develop a model capable of mitigating the risk of severe accidents by accurately predicting

This paper describes the effort to improve the safety level of a domestic research reactor and optimize its operation with regard to safety by conducting a probabilistic safety assessment (PSA) under full-power operating conditions. The PSA was undertaken to assess the level of safety for an operating research reactor in Korea, evaluate whether it is probabilistically safe to operate and reliable to

This paper describes steady-state reactor physics measurements and calculations that were performed for the Training Reactor of Budapest University of Technology and Economics (BME TR) with the purpose of benchmarking. Based on the available geometry specifications and material compositions a model of BME TR was created with the well-validated, general-purpose Serpent 2 Monte Carlo code. Uncertain

The source term calculation capability is implemented in the high-fidelity neutronics code. With this high-fidelity tool, the accuracy of the two-step source term calculation method is analyzed in this work, including decay heat, the activation photon source term, and the neutron source term. The accuracy of NECP-X for the source term calculation is demonstrated through the comparison with SCALE6.1

This paper proposed a method for the calculation of assembly bowing reactivity coefficients in sodium fast reactor. Firstly, in the core-analysis nodal method, a fuel assembly and a thin layer of sodium around it are treated as a node in the radial plane. Then the assembly bowing is discretized to radial displacements of the assembly region in each axial segment. Based on perturbation theory, the reactivity

Accurate solution of neutron transport problems in the discrete ordinates (SN) formulations is relevant in many areas of engineering and nuclear science. Several researches have led to a variety of numerical differencing schemes in order to generate even more accurate numerical solutions. However, most differencing schemes seem to become unstable for sufficiently large spatial cell width and negative

The paper presents CORE SIM+, a tool developed for diffusion-based neutron noise simulations. The simulator is based on the 3-dimensional, two-energy group neutron diffusion equation in the frequency domain. The tool includes the necessary solvers to calculate the criticality problems associated with the system and then the response to a variety of perturbations such as absorbers of variable strength

The present paper focuses on the modeling of the CROCUS reactor with the French MORET 5 continuous energy code and compares keff, anti-reactivity effects (variation of keff normalized to keff due to the insertion of an absorber), sensitivity coefficients and kinetics parameters with those provided by other reference codes. The aim is mainly to extend the experimental validation/verification database

The system thermal–hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) and the sub-channel code SubChanFlow (SCF) have been coupled together based on the Interface for Code Coupling (ICoCo) and the SALOME platform. It is a server-client system where TRACE and SCF are computational engines, ICoCo transfers the coupling data, and SALOME supervises the two clients’ synchronization. TRACE and

Fission gas release from high-burnup mixed-oxide (MOX) fuel pellet with Pu fissile contents of 4.1% during reactivity-initiated accident (RIA) was estimated based on radial Xe concentration profiles in fuel pellet using scanning electron microscopy/electron probe microanalysis (SEM/EPMA) before and after pulse irradiation tests called BZ-3 and BZ-4. The tests were conducted at the Nuclear Safety Research

Bayesian adjustment of nuclear cross-section data within their prior uncertainties has been well-established in the neutronic community as the most mathematically-disciplined approach for improving the quality of neutronic calculations. The premise is that nuclear data are believed to contribute the most to the observed discrepancies between code calculations and measurements for key neutronic performance

Fusion Fission Hybrid Systems (FFHS) could have a potential role in the management of fission reactor wastes, with (in principle) some advantage over the comparable Accelerator Driven systems, devoted to the same objective. The validation of the concept poses major challenges in the area of fusion “source” development. However, the physics of the multiplying and transmutation blanket has also to be

Wall thinning rate is an important issue during design of CAP1400 nuclear power plant (NPP) secondary loop. A single-phase flow accelerated corrosion (FAC) test loop was established to simulate the typical working condition of the secondary loop. The wall thinning rate of A335P11 and 20# steel elbow was measured using high precision ultrasonic sensor on the test loop. The results show that elbow part

The Feynman-alpha method is a neutron noise technique that is used to estimate the prompt neutron period of fissile assemblies. The method and quantity are of widespread interest including in applications such as nuclear criticality safety, safeguards and nonproliferation, and stockpile stewardship; the prompt neutron period may also be used to infer the keff multiplication factor. The Feynman-alpha

In this work, a comparison between the capabilities of a spallation source and a fusion source in the burnup of reprocessed fuels and in the regeneration of thorium was carried out. We have performed Monte Carlo simulations of a sub-critical core loaded with thorium and reprocessed fuel under irradiation of two different neutron sources. Two different fuels reprocessed by the GANEX technique were evaluated

Argon covering the free surface of the pool-type sodium-cooled fast reactor may be mixed into the coolant in form of entrainment bubbles due to the fluctuation of liquid surface, which introduces positive reactivity to the core, reduces the heat exchange capacity of the core and intermediate heat exchangers (IHXs), and possibly leads to the cavitation of the primary sodium pump (PSP), threatening the

During the loading and refueling of Pressurized Water Reactor (PWR), Sb-Be secondary neutron source is generally used to induce fission reaction to increase ex-core detector signal, which could eliminate the monitoring blind area. In order to obtain the optimal position of secondary neutron source which would lead a maximum signal, it is necessary to evaluate the contribution of secondary neutron source

Dry storage casks are the most common form of spent nuclear fuel (SNF) storage in the United States. An explosion near a spent nuclear storage facility poses a serious public risk due to the potential of radioactive leakage. In this study, for the first time, the structural performance of vertical steel–concrete-steel sandwich (SCSS) and reinforced concrete (RC) casks under blast loading scenarios

The C-PORCA reactor physics code of the Paks Nuclear Power Plant performs three-dimensional, two-group diffusion calculations applying parametrized group constants generated by the HELIOS code. To improve the accuracy of the calculations, the C-PORCA code was extended with a simplified spherical harmonics module. This paper presents the applied finite-element-based solution algorithm of the SP3 equations

We present a systematic and detailed approach to simulate the University of Utah TRIGA Reactor (UTR) in support of core criticality calculations to profile the entire reactor in detail. In performing this work, we utilized both a 3-D Cartesian SN deterministic code, PENTRAN, and a Monte Carlo code, MCNP6, to calculate complimentary, high accuracy 3-D transport derived neutron flux distributions at

Advanced accident tolerant cladding materials have brought up the potential to delay the deleterious consequences of loos of coolant accidents related to slowing down hydrogen formation from reaction of zirconium with steam in order to minimize the additional heat generation and improve fuel and cladding retention of fission products. The performance improvement offered by these advanced materials

Moderator acts as primary heat sink to arrest the unwarranted heat-up of fuel channels in the scenario of loss of coolant accident along with the unavailability of emergency core cooling system. In addition, severe heat-up of channel (which consists of fuel bundles, Pressure Tube, PT and Calandria Tube, CT) also occurs during any un-called for event of moderator-cooling/make-up system malfunctioning

This work focuses on the rotor dynamic analysis of the vertical hydro-hybrid bearing-rotor system of a two-circuit main loop liquid Sodium pump system. A new modelling method for the vertical hydro-hybrid bearing-rotor system is put forward. Lubrication parameters are compared with the reference, which verifies correctness of the model. Contours pressure of the bearing are presented. Modal analysis

The report presents the novel genetic algorithm (GA) developed to improve fuel cycle performance and in-core fuel management process. The primary purpose was to develop GA dedicated to solving core loading pattern (LP) problem with predefined core operation constraints. Particular focus was put on maximizing the length of the fuel cycle, but presented solutions allow to limit or control the magnitude

In this present work mathematical and numerical analysis are carryout to determine the existence of Ledinegg and dynamic instability phenomena in a rectangular shape of natural circulation loop at supercritical condition using supercritical water as a working fluid. A mathematical model has been developed based on the thermal hydraulic (TH) conversion equations of mass, energy and momentum with and

A correlation is developed to predict mixed convection heat transfer of water flow in rod bundles. The rod bundles are 5 × 5 and the Re number ranges from 1 × 103 to 3 × 104 in the experiment. In the forced convection region, the Dittus-Boelter correlation predicts well with the experimental data. However, in the mixed convection region, the correlations reported in the literature are all deviated

Loss of heat sink (LOHS) accident in a natural circulation system has a complex event sequence scenario. Flow instability phenomena in LOHS accident involves multiple flow patterns and variety instability mechanisms. Through experimental investigation, a methodology analyzing instability under LOHS is established. It is concluded while slug flow occurring in heating channel, the system falls into geyser

The demand characteristics of nuclear power plant (NPP) component parts were analyzed to provide basic data for stable operation of an NPP. KHNP’s material supply and demand plan from 2017 to 2020 was used to obtain the demand characteristics of NPP component parts for various reactor types in Korea. The reactor types include OPR1000, APR1400, CANDU, Westinghouse-design PWR, and Framatome-design PWR

The existing commercial nuclear power plants (PWR and BWR) utilize the oxide fuels, i.e., UO2. This fuel selection is not questionable, where the safety and economy are the top priority in the nuclear industry. In this work, the potential advantages of microencapsulated fuels to System Integrated Modular Advanced Reactor “SMART” has been explored. The UN and UCO have been considered as the candidate

Monte Carlo criticality calculations of large, loosely-coupled problems are long known to suffer from slow convergence of the tally errors due to cycle-to-cycle fission source correlations. In several recent studies, it was suggested that these correlations could be possibly attributed to the neutron clustering phenomenon that is visible in calculations with a small number of neutrons per iteration

An improved depletion coupling scheme is presented, employing reduced-order transport solutions at substeps between high-fidelity solutions. The purpose of the reduced-order solver is to quickly approximate how the one-group flux would change within the depletion interval. This method builds off existing substep methods, where smaller depletion intervals are used to better approximate reaction rates

The coupling of low-energy electron linear accelerators (eLINACs) to a large heavy water reactor is proposed to create an accelerator-driven photoneutron source (ADS). Photoneutron yields of 1012 pn/s per kW of beam power can be achieved in the ADS-CANDU concept where the wide fuel channel spacing of heavy water reactors represents a near-optimal geometry for conversion of accelerator-driven bremsstrahlung

The investigation of neutron noise is key to several applications in nuclear reactor physics, such as the detection of control rod or assembly vibrations and the diagnostic of coolant speed and void fraction. In this paper we will elucidate some aspects of the noise equations in the Fourier domain, for the case of periodic fuel rod vibrations with frequency ω0 in a small symmetrical system in which

The decision-making process in the Nuclear Power Plant (NPP) control room faces some challenges: operator incomplete knowledge, insufficient time for responding to the highly dynamic events, and a large number of indicators to monitor. Because of the complexity of the NPP system, it is hard to pre-plan all the failures/mitigative actions. An intelligent operator support system is vital to mitigate

Megawatt-class heat pipe reactors (HPRs) have been identified as a candidate for the decentralized electricity markets, for instance, remote regions, island communities, and military bases. HPRs exhibit some unique designs, such as the encapsulated solid core with fuel pellets and heat pipes, control drums, and irregular reflectors. The neutronic tools are under development to investigate core characteristics

Frequent and wide-range load changes in small pressurized water reactors applied for nuclear-powered devices are necessary under complex and harsh conditions. However, the traditional power regulation method does not consider the stroke of the control rod, which seriously affects the service life of the control rod drive mechanism. This paper proposes a data-driven reactor power regulation system (RPRS)

Thermal conductivity of buffer material is crucial for the design and performance assessment of high-level radioactive waste (HLW) repository. Large amount of decay heat will increase the maximum temperature in bentonite buffer, which threatens the safety of HLW repository. One possible method to relieve such problem is adding graphite to bentonite for increasing its thermal conductivity. Therefore

Public fear of safety of NPPs reached the highest level due to social insecurity from the Chernobyl and Fukushima NPP accidents. Also, due to possibility of a terrorism or an accident of the radiation containing structures in NPPs, structural design details are being studied in depth by the design engineers to improve its safety from the extreme hazard events. Since PCCV is a last safety barrier in

238Pu can be used as the heat source in radioisotope thermoelectric generators for deep space exploration, while it is currently in shortage because of the suspension of 238Pu production in the USA and Russia for many years. Two possible scenarios for the production of 238Pu as a byproduct in a molten salt fast reactor (MSFR) are assessed in this work. For the first scenario, the MSFR is started with

The capability of performing nuclear data sensitivity and uncertainty analysis of reaction rates has been developed in continuous-energy Reactor Monte Carlo (RMC) code. The sensitivity coefficients of reaction rates are calculated by using new generalized perturbation theory (GPT) formulation which is also implemented in McCARD. The superhistory based generalized perturbation theory (SH-GPT) formulation

This work demonstrates the functionality of the pressurizer using a fast-running three regions, non-equilibrium model and control programs of emulated PLCs. The state variables from an integrated model of primary loop in a representative PWR plant are communicated to the pressurizer’s emulated PLCs using a synchronized data transfer function. In turn, the PLCs communicate back instructions to the pressurizer

Passive methods of nuclear safeguards determine the important parameters of an unknown sample from the statistics of the detection of the neutrons emitted from the item. These latter are due to spontaneous fissions and (α,n) reactions, enhanced by internal multiplication before leaking out. Based on the original work of Böhnel, the methodology of traditional multiplicity counting is based on the first

The application of lead–bismuth small modular reactor (SMR) has unique advantages in offshore engineering. Moving conditions caused by ocean environment are important for the security analysis of nuclear system. In this paper, the models including core neutronics, reactor thermal hydraulics and moving conditions for a natural circulation lead–bismuth reactor with thermal power of 5 MW are established

When 17-4PH martensitic stainless steel serviced at 300 °C for long term in nuclear power plant, the thermal aging embrittlement and hardening will be significant. It will seriously affect the safety and economic operation of nuclear power plant. In this work, Three-Dimensional Atom Probe Technology (3DAPT) was used to characterize the element distribution, and the evolution of Cu-rich precipitate

The DGT (Dense Granular spallation Target) is the crucial component in CiADS (China initiative Accelerator Driven subcritical System), which is the important device in connecting the high power accelerator and the subcritical reactor. In order to keep the stable and reliable running of the coupling system, the temperature control of the DGT should be carefully studied and optimized, where the gravity

High-temperature gas cooled reactor pebble-bed modular (HTR-PM) uses a dry storage system based on steel storage canisters to store and transport spent fuels on-site. For the canister, one of the key equipment, it is necessary to verify the reliability of the structural design in extreme events, especially under the accidental drop. Since there are 40,000 fuel elements in a canister, interfacial coupling

The reverse flow is an important phenomenon occurring in inverted U-tube steam generator under natural circulation, which may result the heat accumulating in the first loop and cause serious accident. Thus, researching the flow performance and its influence factors in steam generator is necessary for the safety of nuclear reactor. In this paper, the criterion of reverse flow judgement and flow distribution

The front-end sensors of the control system are the weak links for a reliable and stable operation of system. Due to insufficient capabilities of the algorithms or methods, there are many control anomalies caused by the failure of the front-end sensors. This paper proposes a front-end redundancy intelligent diagnosis model for control systems, which mainly includes five sub-models: transfinite judgment

In this paper, the benchmark experiment of the lead reactivity worth was introduced in detail, which has been carried out under a super-critical state at the light water reactor of the VENUS-II experimental facility. The reactivity worth of the cylindrical lead target was measured and processed as 142.0 ± 11.3 pcm by the period method. In simulations, the reactivity worth of the lead target was calculated

To study the heat transfer phenomenon in the complex space of small steel containment during the jet flow, the experiments and simulation were conducted with various jet inlet temperatures, velocities, and heights. The results show the flow field and temperature field are quite different from the large containment. There is no obvious thermal stratification phenomenon in small containment during the

Oxygen concentration control is the most promising technique to solve the corrosion problem in LFRs. A numerical study on the oxygen transport characteristics in the solid-phase oxygen control is carried out in this paper. The geometry is created using the Discrete Elements Method (DEM). The mass transfer and boundary layer theory is adopted to model the PbO dissolution process. The dissolution rate

The complex core geometry of Pebble Bed Reactors (PBRs) necessitates multiscale techniques for fast-turnaround design and analysis. This paper describes the multiscale model implemented in the Pronghorn PBR simulation tool and demonstrates application to steady-state analysis of the Mark-1 Pebble Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR). Verification of the pebble model with fully-resolved

Once through type steam generators used in Sodium cooled Fast Reactors (SFR) are prone to flow instability. Operating parameters of the steam generators play vital role in defining a stable operating regime. Towards understanding this, a thermal hydraulic model is developed. The developed model is validated by performing experiments on a 19-tube model of straight vertical steam generator of 5.5 MW

The purpose of this paper is to design an improved container with drainage system for treating and transporting the radioactive wastes under water during decommissioning of nuclear facilities. It is possible to selectively drain water flown into the container body together with the radioactive waste during treatment and transportation of the radioactive waste under water. It is possible to safely lift

Nodal dynamic models allow rapid reduced-order simulations of complex systems. This paper presents a publicly available system dynamic model of a molten salt reactor system with a topology of the Molten Salt Reactor Experiment (MSRE), in MATLAB/Simulink, building upon a model previously published and validated with MSRE data. It adds scaling of nominal power, dynamic representation of decay heat generation

Internal storage is used in fast reactors as it allows burnt fissile assemblies to cool down before being removed from the reactor. These assemblies are usually stored in the neutron shielding zone far from the core to avoid additional induced fissions. The BALZAC-SI experiment performed in the MASURCA Reactor was dedicated to study internal storage characteristics of a SFR-type MOX core. Fission rate