Burnup capability is developed on the ALPHA code with Sanchez-Pomraning MOC (Sanchez-MOC) approach for FCM fuel design. To better model the double heterogeneity (DH) of FCM fuel, ALPHA adopts a combination strategy of macroscopic discretization at pin level and microscopic discretization at particle level for burnup region. The burnup capability is validated by code-to-code comparison with Serpent

In this study, measurement of turbulence intensity, pressure drop and local velocity field in the sodium fast reactor (SFR) wire-wrapped 19-pin bundle have been conducted to compare with the Reynolds averaged Navier–Stokes equations (RANS)-based computational fluid dynamics (CFD) simulations. The refractive indices between test section and working fluid, anise oil, were matched for the flow visualization

The safe operation of Pressurized Water Reactors (PWRs) is highly dependent on the Pressurizer (PRZ) system to stabilize the primary coolant pressure fluctuations during normal operation and maintain the pressure within specified boundaries. The behavior of the PRZ system is inherently non-linear and time-variant. Thus, it may not appropriate and precise enough to model it by the use of transfer functions

During a severe accident in a nuclear reactor, the molten core may relocate in the lower head of the reactor vessel or in the reactor pit in case of vessel failure. In both cases, if water is present, it may lead to the so-called “fuel-coolant interaction”. When the melt jet enters into a water pool, it undergoes fragmentation and simultaneously initiates the rapid vapor generation due to high heat

This study evaluated graphite dust emissions in an HTGR depressurization accident. A simplified experimental platform based on air at ambient temperature was developed to simulate particle resuspension during a depressurization accident and helium blower startup. Then, computational fluid dynamics models were used to calculate the impact, rebound, and deposition of graphite particles to determine the

The effectiveness of the path plan is the main factor of the evacuation affecting the amount of exposure to personnel, however, the effect of the radiation field distribution has not been fully considered in current studies. In this paper, a Minimizing Collective dose based Path-Planning method (MCPP) was proposed. Among which, the accumulative dose based Dijkstra algorithm was proposed to find the

For dismantling nuclear power plants, an underwater cutting study was conducted using a high power fiber laser. With a laser power of 9 kW, cutting tests were performed on stainless steel with a thickness of 50–100 mm. For this study, a waterproof cutting head consisting of a focusing element with a long focal length of 600 mm and a supersonic nozzle was used for high cutting efficiency. For thicknesses

The SBO is defined as the events of Loss of Offsite Power (LOOP) combined by loss of dedicated Emergency Diesel Generators(EDGs) for each unit. The SBO can be also extended when the shared Diesel Generator (AAC-DG) is lost additionally. The Electrical Cross-tie (ECT) is the concept of sharing other unit’s dedicated EDG to the unit under extended SBO in the multi-unit NPP site. We re-evaluate the risk

Scenarios studies have proven the fuel cycle closure feasibility with fleet composed by PWRs and large fast reactors. Nevertheless, large reactors may not be adapted to low or scalable electricity needs (off-grid cities, high renewable share, desalination plant, etc…). Sodium-cooled Small Modular Reactors (SMRs-Na) could be an interesting answer to these needs. However, they tend to be plutonium burners

The aim of this paper is to present the critical verification for TRX and BAPL benchmarks between the ENDF/B-VIII.0 and ENDF/B-VII.0 library. Using DRAGON5.0.6 code with 281-group WIMS-D and DRAG library, the integral parameters for TRX and BAPL benchmarks are compared to the Monte-Carlo code and experimental values. For the verification of DRAGON code, the influence of tracking method, resonance self-shielding

The aim of this paper is the analysis of the neutron diffusion coefficient with 3D approximation of the up-scaled neutron diffusion equation. In this work we present comparisons of 3D ordered and disordered approximations. To this end, we compute the effective diffusion and interfacial neutronic flux coefficients in representative cells of fuel elements composed by several TRISO particles randomly

In this paper, we propose a new non-linear technique for accelerating the solution of the discrete ordinates transport equation. The new method, called Response Matrix Acceleration (RMA), has been designed to complement the Coarse-Mesh Finite Difference method (CMFD) by offering better stability and improved performance in cases where CMFD fails. To accomplish this, RMA uses knowledge of the transport

A real-time research reactor simulator was developed at the Jožef Stefan Institute for education and training of students and future reactor operators, especially in countries with limited or no access to a research reactor. The research reactor simulator (RRS) simulates time behaviour of reactor power, fuel temperature and reactivity by using the 6-group point kinetics equation with feedback. It features

This paper discusses the source convergence of Monte Carlo calculations for α-eigenvalues. Compared with the conventional “k–α method”, the “time-source method” has a strong neutronic coupling even in a system where the fission sources are loosely coupled. The dominance ratio of the time source is smaller than that of the k–α method. The time-source method has a drawback in that it requires a much

The purpose of this study is to investigate the effect of zinc addition to the simulated primary coolant of a PWR on fuel crud deposition behavior. Crud deposition tests were performed in recirculating hydrogenated water containing 2 ppm Li and 1000 ppm B at three different zinc concentrations (0, 20, and 100 ppb) at 325 °C and 120 bar. Crud deposited on the claddings was characterized as cubic nickel

Operating staffs of a nuclear power plant (NPP) are responsible for returning the NPP to a stable state and alert authorities if there is the potential for offsite radiological consequences following an accident. An operator support tool (OST) using deep learning techniques and trained by data from dynamic probabilistic safety/risk assessment (DPSA/DPRA) is proposed to assist the NPP personnel in decision-making

Natural circulation loop systems are promising a reasonable option to provide the emergency decay heat removal from a nuclear reactor during an electrical power outage. The present paper deals with an experimental study on the influence of leakages on the behaviour of a single-phase natural circulation loop. The research shows the thermodynamic and hydraulic processes in a large scale natural circulation

The release of fission products to the environment is one of the concerns with nuclear power. During an accident, the most likely released are the volatile fission products i.e., tellurium. To evaluate the behavior of tellurium in the event of an accident, it was heated under different conditions (oxidizing, inert, reducing; both dry and humidified). The formed vapor was transported to surfaces (aluminum

Non-functional requirement analysis is the most critical factor for safety-critical system construction as it reduces the risk of catastrophic loss of assets by taking measurable actions in the design phase. Reliability and performance analysis are the two main components of non-functional requirement analysis. In this paper, a reliability analysis framework is devised, which maps the Unified Modeling

The strategy for In-Vessel Retention (IVR) of corium is currently considered for several new reactor designs in various countries. One of the issues for the demonstration of the success of this strategy is that there are significant uncertainties in physical modelling of corium in the lower plenum and its transient chemical and thermal interactions with the vessel leading to its significant ablation

For the accurate assessment of the heat generation rate in fast reactors, the gamma library of MC2-3 and the MC2-3 + GAMSOR procedure employing the coupled neutron and gamma heating calculation has been thoroughly verified against Monte Carlo results and validated using the ZPPR-15D gamma dose measurement data. NJOY outputs are post-processed in a consistent way with the NJOY procedure to avoid any

The thermal behavior of reactor channels is of prime concern for a postulated scenario of LOCA along with the un-availability of ECCS as the integrity of the fuel bundle may be challenged and slumping of fuel pins is expected. The present study is aimed to analyze the thermal characteristics of a large capacity Indian PHWR specific fuel bundle with disassembled fuel pins configuration. The configuration

An internal parallel coupling scheme was developed for the coupling between high-fidelity neutronics code NECP-X and sub-channel thermal–hydraulic code CTF. A new coupling iteration scheme is proposed based on NECP-X/CTF in which CTF will run after each NECP-X outer iteration. CTF was involved in NECP-X as a shared library object and driven by NECP-X, NECP-X and CTF have their coupling interface, respectively

In this work, non-modal stability theory is used to study the non-linear dynamics of the Pressurised Water Reactor (PWR), using a zero-dimensional model. Usually, stability analysis for nuclear reactors is performed using modal methods. Instead, the non-modal approach used in the present work offers a different perspective. This approach considers the so-called short-term behaviour following a disturbance

Wrapped wire spacers have been incorporated as the spacing device in advanced nuclear reactors such as the sodium-cooled fast reactor and supercritical water-cooled reactor. The cross flow induced by the wrapped wire affects the subchannel mixing and fuel rod temperature, which is a concerned issue in the thermal-hydraulic design of the fuel assembly. In the present paper, an experiment has been conducted

The Low Rank Approximation (LRA) and Unscented Transform (UT) are integrated to produce a new algorithm having the capability to decrease the time required for the uncertainty quantification during Loss of coolant accident (LOCA) in Pressurized Water Reactors (PWR). The LRA is an efficient technique used in reducing computational cost due to its ability to perform dimensionality reduction by revealing

In-Vessel Retention (IVR) of corium is one of the possible strategies for Severe Accident (SA) mitigation. Its main advantage lies in the fact that, by maintaining the corium within the vessel, it preserves the last containment barrier from corium aggression. One of the issues for the demonstration of the success of this strategy is the evaluation of the behaviour of the corium relocated in the lower

In this paper, a numerical procedure for thermal–hydraulic subchannel analysis of the annular fuel assembly in a PWR using the drift-flux model (DFM) is presented. The system of nonlinear equations was solved with a fully-implicit solution scheme using the Jacobian-free Newton-Krylov (JFNK) method. To improve the computational efficiency and numerical stability of the JFNK method, a semi-implicit physics-based

Recent investigation into designs of molten salt reactors (MSR) requires accurate data of candidate salts. Review of the literature shows lacking thermophysical (TP) property data for coolants and fuel-bearing salts of Advanced High Temperature Reactors (AHTR). Measurement uncertainties of current data are even more deficient, and this affects ongoing research where MSR modeling and testing are being

The present paper is concerned with the extension of the treatment of the scattering anisotropy developed in the AN-SPN (or simply AN) method up to the approximation of the second order. To assess such extension, two 3D examples are considered: the first one is an axial piece of a 17×17 PWR MOX assembly containing the heads of control rods; while the second example is related to a 3D reactor core containing

The Virtual Environment for Reactor Applications (VERA) is being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). VERA is used to perform the Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS), which provides two cycles’ worth of operating power history, along with a full, detailed description of the geometry, and measured data. Cycle 1 and

Systems, Structures and Components (SSCs) have different safety significance for reactor operation, and it is necessary to classify SSCs into different safety classes for design, manufacture and maintenance of a reactor. In order to further identify the safety classes of SSCs in reactors, a new integrated safety classification method was proposed in this study, based on the 50-SG-D1 safety classification

MCNP and the ADVANTG code were used to analyse γ-dose rates for the future Slovenian LILW repository. Using ADVANTG’s variance reduction technique, different detailed calculations of γ-dose rates were achievable in a reasonable time, whereby the analogue MCNP simulations were accelerated up to 9500 times. We analysed the influence of different repository configurations on the annual γ-dose rates up

As a reactor core has time-varying and nonlinear characteristics, it is difficult to achieve an ideal control effect with traditional control theory. To optimize the core power control performance, a compound control scheme is formulated that is composed of multiple controllers by weighting and switching. Firstly, the nonlinear model of a reactor core is transformed into a transfer function model.

The Autonomous Circulation Miniature Integrated nuclear Reactor (ACMIR) has features of compact structure, high power density and high efficiency. The reactivity control strategy is an important aspect in the design of ACMIR, which is essential for reactor safe operation. In this paper, two ex-core reactivity control options, namely, the rotating control drums and sliding reflector segments, were investigated

The eutectic reaction model in JUPITER code was validated against two series of experimental tests that performed by JAEA. An experiment that aimed to evaluate the eutectic reaction between Zircaloy and Stainless steel, was simulated by JUPITER code to validate its reliability on predicting the binary eutectic reaction phenomenon. A comparison of the simulation and experimental results demonstrates

In this paper, we investigate different bias factor methodologies, including the standard bias factor method, the generalized bias factor method, and the product of exponentiated experimental values bias factor method. One new method is proposed—the representativity weighted bias factor method, which is based on weighted statistical dispersion. Finally, an old method, the extended bias factor method

This paper presents a mechanistic wall boiling model for the prediction of flow boiling on a downward-facing oriented heating surface containing sliding slug bubbles on the heating surface at the pre-critical heat flux condition. Essentially, this boiling model is an extension of the well-known RPI boiling model which was initially developed based on dispersed small spherical bubbles on the heated

The present work investigates the attenuation of neutron and photon-induced irradiation damage in a Pressurized Water Reactor (PWR)-like mock-up Reactor Pressure Vessel (RPV) using Tripoli-4® Monte Carlo simulations with a particular emphasis on the presence of a thick stainless steel heavy reflector between the core and RPV. Results show that the photon-induced damage is well described by the exponential

Metakaolin-based Na-Pollucite ceramics (high-temperature sintered Na-pollucite ceramics: HTNP), with a target composition close to Pollucite (CsAlSi2O6), were fabricated at sintering temperatures above 1000 °C to immobilize Cs as waste form material. The sintering behaviors of Metakaolin-based Na-Pollucite ceramics were investigated, in particular with respect to sintering temperature, grain size,

In view of the safety characteristics of the offshore floating nuclear plant (OFNP), an ocean condition available analysis code based on RELAP5 MOD3.3 is developed and the influences of single heaving, single rolling and combined motion of heaving and rolling on a OFNP were analyzed under station blackout accident (SBO). The results indicate that heaving motion accelerates the drop in the reactor pressure

The number of Displacement per Atom (DPA) in the fuel cladding is a key quantity for determining the operation lifetime of fuel assemblies in Sodium-cooled Fast Reactor (SFR). Since the cladding is directly irradiated by Fission Products (FPs), the present work evaluates the FPs-induced atomic displacements in the Fe-14Cr cladding. Three degrees of freedom (the yields of FPs, energy distribution, and

Under very low probability postulated accident like LOCA with unavailability of ECCS, the moderator heat sink maintains the channel integrity by providing adequate cooling. As the cooling is external to channel, the circumferential as well axial temperature gradient exists within the channel internals. An experimental set-up was fabricated to simulate the boil-off condition during a postulated accident

This study deals with the influence of uncertainties in selected cross-section data in a neutronics code while their impact on the results from thermo-hydraulics code is investigated. Particular cross-section data were defined according to the Phenomenon Identification and Ranking Tables (PIRT) for a pressurized water reactor. Uncertainties were identified for the following data: β – delayed neutron

This study proposes an overall methodology that provides in-depth evidence on software reliability. It is used to quantitatively assess the reliability of nuclear power plant (NPP) safety-critical software for the incorporation of digital instrumentation and control systems into NPP probabilistic risk assessment (PRA). The methodology consists of three parts: (1) the relationships among the software

The tracking of fuel inventory is important for safety and for knowledge of the flux distribution. In a research reactor, tracking of the reactor core fuel composition using simulation code is crucial for fuel management strategy, such as by optimizing the core pattern for sample irradiation purposes and for providing an economical fuel cycle length. Operational data history for the McMaster Nuclear

TRISO particle is adopted in the design of various types of reactors and accident tolerant fuel (ATF) due to its excellent accident tolerant performance under high temperature and other severe conditions. This study presents a multidimensional thermomechanical analysis of TRISO particle fuel with SiC/ZrC coating under typical operation conditions in reactor. Key fuel performance phenomena were considered

Optimization problems in the research literature are typically simplified and/or heavily constrained and focus on a single set of physical processes. Real-world nuclear engineering problems feature competing multi-physics phenomena and require equally complex analysis. To prove its usefulness in this area, optimization must demonstrate an ability to handle many competing objectives whilst accurately

Personality traits have been recognized as an important means to mitigate psychological and physiological effects on human performance. In the nuclear industry, personality traits have been utilized only to screen out potential employees of mental illnesses. Although cognitive characteristics are considered in second generation HRA methods, there is no or superficial consideration about personality

One-dimensional two-fluid mass, momentum, and energy conservation equations for gas and liquid droplet and the particle removal rate are formulated to estimate aerosol particle removal rates of venturi scrubbers used in filtered containment venting systems. Compared with existing one-dimensional models, present method enables describing actual two-phase interactions by calculating the local gas and

In case of a large or medium LOCA accident of marine nuclear power plant, the loop pressure drops rapidly. In case of power failure of the whole ship at this time, the active safety injection system will lose its function because the safety injection pump needs AC power, and the core is at risk of melting. A passive marine safety injection system based on high-pressure compressed gas is designed for

The dry storage of spent fuel may foster the radial reorientation of hydrides formed in the cladding. This mechanism reduces the material ductility, being a key aspect to assess the cladding integrity. The present work extends the predictive capabilities of FRAPCON-xt (CIEMAT́s extension to dry storage of the FRAPCON code) to the hydride radial reorientation within the cladding. The validation against

As the core simulator capabilities in the Virtual Environment for Reactor Applications (VERA) have become more mature and stable, increased attention has been focused on coupling Bison to provide fuel performance simulations. This technique has been a very important driver for the pellet-clad interaction challenge problem being addressed by the Consortium for Advanced Simulation of Light Water Reactors

In the whole-core heterogeneous transport calculations with massive parallel, the computational cost of the Coarse Mesh Finite Difference (CMFD) method might become the most time-consuming part if directly using the traditional CMFD method. To accelerate the CMFD calculations, a new three-level CMFD method based on the loosely coupled parallel strategy is developed in the high-fidelity neutronics code

The paper investigates the flow noises caused by the high temperature liquid Sodium in the nuclear main pump system. Three-dimensional unsteady CFD coupled sound field methods are exploited. Spectrum characteristics of sound source load are obtained. Directional sound pressure level distribution based on sound source load of rotor blades and guide vanes are presented and discussed. Distribution of

Leak-before-break analyses on tube microcracks are beneficial for preventing small loss of coolant accident in nuclear power plants. Specially, the tube leakage of steam generator may lead to steam generator tube rupture. Currently, experimental investigations on leak flow rate characteristics of normal temperature water and high pressure subcooled water through axial and circumferential microcracks

Vibrations of tube bundles in heat exchange equipment, such as nuclear steam generators and tube-and-shell type heat exchangers, are often caused by unsteady fluid forces. The study of flow-induced vibration excitation forces on tubes is necessary for understanding the fluid-elastic instability phenomenon. In this paper, a forced vibration method is introduced to investigate unsteady motion-dependent

Fast Neutron Computed Tomography (FNCT) can be applied in identifying induced structural anomalies of Irradiated Fuel Assemblies (IFA). This is because fast neutrons have high penetrability, and FNCT has the capability to avoid the interference of sample radioactivity and reconstruct internal information of sample. Few projections and a high sharpness are required for FNCT. In this study, an Adaptive

To give better reactor physics results, the detailed 3-D diffusion module has been implemented in HTGR engineering simulator system (ESS), which needs more efficient cross-section calculations to meet real-time requirement of ESS. To calculated the HTGR’s cross-sections, more reactor variables are needed with larger scope and more complex non-linear relationships, which makes the polynomial method