Primary Coolant Loop Research Articles

Corrosion protection and failure mechanism of ZrO2 coating on zirconium alloy Zry-4 under varied LiOH concentrations in lithiated water at 360 °C/18.5 MPa

Maintaining a good corrosion resistance in primary coolant loop during normal operations is a prerequisite for ZrO2 as a protective coating on zirconium alloy cladding. Research on corrosion performance of ZrO2 coating in nuclear water chemistry is relatively deficient, and existing reports failed to provide an in-depth explanation for the failure causes of ZrO2 coating. Here, we proposed a detailed corrosion process of ZrO2 coating in lithiated water at 360 °C and 18.5 MPa based on experimental study and molecular dynamics simulation. The protective effect and failure mechanism of ZrO2 coating on Zry-4 under varied LiOH concentrations was further revealed. ZrO2 coating provided a favorable corrosion protection with the occurrence of localized corrosion at low LiOH concentrations. Factors influencing the corrosion resistance include pitting corrosion extension, enhanced Li+ permeation, short-circuit diffusion of O2– and ZrO2 phase transformation. In highly-concentrated LiOH solutions, intergranular corrosion, internal oxidation and perforation result in coating failure. Zr ions were released to the coating surface to form flocculent ZrO2 and ZrO2 clusters due to the strong diffusion and dissolution tendency of α-Zr in the Zry-4 substrate. This work can give some references to understand the service behavior of nuclear coatings under variable water chemistry conditions and promoting in-pile application of ZrO2 coating.

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. To achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.This manuscript has been authored in part by UT-Battelle, LLC, under contract DE AC05-00OR22725 with the US Department of Energy (DOE). The publisher acknowledges the US government license to provide public access under the DOE Public Access Plan (https://energy.gov/downloads/doe-public-access-plan [energy.gov]).

Score-based Bayesian network structure learning algorithms for modeling radioisotope levels in nuclear power plant reactors

Radioactive corrosion products released into the primary coolant loop dominate the final shutdown radiation fields of pressurized water reactors. Thus, reducing the concentration of these corrosion products is a paramount duty in the optimization process of the reactor performance. However, the complexity and uncertainty present in this process make it difficult to predict their evolution in a theoretical way. We propose the application of structural learning of Bayesian networks to discover the complex relations between the corrosion products and the most relevant variables in the primary loop, giving rise to probabilistic models that obtain accurate and reliable predictions of the corrosion products. Our analysis of 5 power plants demonstrates that our approach results in simpler and more reliable models. Additionally, we conclude that the learned structures may represent an interpretable tool for power plant technicians since they reveal useful information that can be directly employed to improve the reactor operation.

Dynamic mass accountancy modeling of a molten salt reactor using equilibrium thermodynamics

A mechanistic-based mass accountancy model in the context of liquid-fueled molten salt reactors was implemented in the dynamic systems modeling software library TRANSFORM by way of coupling with the equilibrium thermodynamics code Thermochimica. Liquid-fueled molten salt reactors present new challenges for mass accountancy because of the dissolution of fuel and evolved fission products, which may be soluble in the salt, off-gas, or precipitate. Two cases of mass loss from the molten salt were addressed: off-gassing and precipitation. The software implementation was tested through a series of increasingly complex demonstration problems, culminating in a model of the primary fuel and primary coolant loops of the molten salt demonstration reactor. Analysis shows that negligible mass was lost from the salt under normal operating conditions, but an overheating event caused by partial loss of fuel loop cooling resulted in release of measurable amounts of uranium (among other elements) via off-gassing. The tools developed here are primarily aimed at capability development but are readily available for use in further modeling of molten salt reactor concepts. These tools have not yet been validated, and future experimental work to perform this validation is recommended.

ATHLET Simulation of PKL I2.2 IB-LOCA Benchmark Test and Quantitative Assessment

The Primary Coolant Loop Test Facility [Primӓrkreislӓufe Versuchsanlage (PKL)] PKL I2.2 Benchmark experiment for an intermediate-break loss-of-coolant accident (IB-LOCA) with a 13% or 17% break at the cold leg was performed in the Organisation for Economic Co-operation and Development/PKL-4 project at PKL in Erlangen, Germany. Analysis of Thermal-Hydraulics of LEaks and Transients (ATHLET) 3.1A was used at Karlsruhe Institute of Technology for its posttest calculations. Crucial predicted parameters were compared with measured data. The calculated trend of the selected parameters fits well with that of the experimental data except for the phenomenon of core heatup and the value of the peak cladding temperature. A fast Fourier transform–based method was chosen to quantify the matching of the parameter trends. According to the quantitative assessment, the IB-LOCA scenario and its detailed phenomena can be predicted well by ATHLET. Additionally, some discrepancies, i.e., insufficient reliable predictions for break mass flow and for reactor pressure vessel collapsed water level, were also observed, possibly deserving another study to undergo deeper scrutiny.

Electrochemical corrosion study on base metals used in nuclear power plants in the HyBRID process for chemical decontamination

Base metal corrosion forms a significant issue during the chemical decontamination of the primary coolant loop in nuclear power plants as it is directly related to the economic and safety viability of decommissioning. In this technical note, potentiodynamic evaluations of several base metals (304 stainless steel, SA106 Grade B carbon steel, and alloy 600) were performed to determine their corrosion behavior during the hydrazine (N2H4)-based reductive ion decontamination (HyBRID) process. The results suggested that N2H4 protected the surface of the base metals in the HyBRID solution, which is primarily composed of H2SO4. The corrosion resistance of the carbon steel was further improved through the addition of CuSO4 to the solution. The corrosion rate of carbon steel in the H2SO4-N2H4-CuSO4 solution was lower than that exhibited in an oxalic acid solution, a commonly used reaction medium during commercial decontamination processes. These results indicate the superiority of the HyBRID process with respect to the base metal stability.

Modeling of Diagnostic Parameter for Assessing the Technical Condition of Sections of Locomotive Cooling System

Current paper presents calculation results of the temperature on the surface of the cooling modules in the initial and operating states; and a new diagnostic parameter for functionality assessment of cooling modules in locomotives. Calculation of the temperature on the surface of the section of the primary coolant loop in the initial state and with different fouling levels (same for all sections) is made. Results, that are obtained from hydraulic, heat and aerodynamic calculations, correspond to the nominal operating mode of a diesel engine, radiator fan and normal atmospheric conditions. At a high level of fouling, there is a marked decrease in surface temperature compared to the clean sections by an average of 3.5 – 4 ºС, what indicates reduction of heat-dissipating capacity of the section. Parameter of the temperature difference of the wall of the tube is used as an informative temperature parameter for assessment of thermal efficiency degree of the section and for convenient presentation of the obtained data. There are obtained 12 possible distributions of surface temperature along the working length of the section tubes on the side of cooling chamber. The influence of the number of fouled modules on performance of locomotive cooling system has been investigated. There are established functional dependencies between technical condition of radiator sections and temperature difference of section tube walls allowing to assess the possibility of the cooling system to provide the required heat dissipation when operating.

Novel Non-Evaporable Getter Materials and Their Possible Use in Fusion Application for Tritium Recovery.

Non-evaporable getters (NEGs) are metallic compounds of the IV group, particularly titanium and/or zirconium-based alloys and are usually used as pumps in vacuum technologies since they are able to sorb, by chemical reactions, most of the active gas molecules, with particular efficacy towards hydrogen isotopes. This work suggests an alternative application of these materials to fusion nuclear reactors, where there is the need to recover small amount of tritium from the large helium flow rate composing the primary coolant loop. Starting from the tritium mass balance inside the primary coolant loop, the amount of coolant to be routed inside the coolant purification system (CPS) is identified. Then a feasibility study, based on the bulk getter theory, is presented by considering three different commercial alloys, named ST707, ST101 and ZAO. The results provide the mass, the area and the regeneration parameters of the three different alloys necessary to fulfill the requirements of the CPS unit. By comparing the features of the three alloys, the ZAO material appears the most promising for the proposed application because it requires the lower amount of material and a lower number of regeneration cycles.

Prognostic systems representation in a function-based Bayesian model during engineering design

Prognostics and Health Management (PHM) systems are usually only considered and set up in the late stage of design or even during the system’s lifetime, after the major design decision have been made. However, considering the PHM system’s impact on the system failure probabilities can benefit the system design early on and subsequently reduce costs. The identification of failure paths in the early phases of engineering design can guide the designer toward a safer, more reliable and cost-efficient design. Several functional failure modeling methods have been developed recently. One of their advantages is to allow for risk assessment in the early stages of the design. Risk and reliability functional failure analysis methods currently developed do not explicitly model the PHM equipment used to identify and prevent potential system failures. This paper proposes a framework to optimize prognostic systems selection and positioning during the early stages of a complex system design. A Bayesian network, incorporating the PHM systems, is used to analyze the functional model and failure propagation. The algorithm developed within the proposed framework returns the optimized placement of PHM hardware in the complex system, allowing the designer to evaluate the need for system improvement. A design tool was developed to automatically apply the proposed method. A generic pressurized water nuclear reactor primary coolant loop system is used to present a case study illustrating the proposed framework. The results obtained for this particular case study demonstrate the promise of the method introduced in this paper. The case study notably exhibits how the proposed framework can be used to support engineering design teams in making better informed decisions early in the design phase.

Fuzzy logic algorithms for classification and parameterization of eddy-current signals

Safety of nuclear power plants operation depends on steam generator safe operation. Russian nuclear power plants with water-water energetic reactors WWER-440 and WWER-1000 use horizontal type steam generators which are cylindrical objects of 3 m diameter and 12 - 15 m length. Steam goes through steam generator tube sheets in the bottom from hot end and passes its energy through heat-exchanger tube system and become colder on the other tube sheet. Reactors WWER-440 have more than 5,000 tubes in one steam generator and reactors WWER-1000 have already 11,000 tubes in one steam generator. Each block of WWER-440 reactor has 6 steam generators, and each block of WWER-1000 reactor has 4 steam generators. A thin-walled steam generator tube can be a threat of nuclear fallout in case of high pressure in primary coolant loop. To be an effective barrier these tubes should be free from corrosion, thinning and cracks.

Effect of wall thinning on deformation and failure of copper‐nickel alloy elbows subjected to low cycle fatigue

Various industrial standards and safety regulations specify the minim number of cycles, which pressure equipment must tolerate during a single seismic event. The compliance with these standards and regulations is necessary in order to ensure safety and strength of critical structural components, eg, pressure pipe system of the primary coolant loop of nuclear reactors or condensed water supply of large hospitals or other large facilities. However, the strength can degrade with time due to flow accelerated corrosion causing wall thinning. Pipe bends are particularly susceptible to the localised corrosion, and, therefore, these structural elements need a special consideration. The current paper studies the effect of localised corrosion on low cycle fatigue behaviour of a copper‐nickel elbows using experiment and finite element method (FEM). The outcomes of the study demonstrate a quite low reduction of low cycle fatigue life even at severe local wall thinning at different locations.

Thermal-hydraulic analysis of the thermoelectric space reactor power system with a potassium heat pipe radiator

The thermoelectric space reactor power system conceptual design incorporates fast reactor, liquid-lithium primary coolant loops that transfer heat to potassium-filled high temperature heat pipes and thermoelectric generators. An integrated system analysis model was developed to study the operating characteristics of the space power system, which consisted of reactor thermal-hydraulic model, neutron kinetics model, thermoelectric energy conversion assembly (ECA) model, heat pipe radiator model etc. One-dimensional thermal-hydraulic model was applied to model the coolant circuit. Quasi two-dimensional analysis models for heat transfer of fuel element, ECA and the heat pipe fin were established. Considering the thermoelectric conversion phenomena, an equivalent electric efficiency model was inserted in the ECA model. Both finite element method and thermal resistance network were applied to simulate the potassium heat pipe system. The normal operation condition and two demonstration accident scenarios (unprotected inadvertent movement of a sliding reflector and loss of heat pipes) were calculated to prove the capabilities of the new system model.

An improved CFD model for a MYRRHA based primary coolant loop

Within the SESAME framework, the Computational Fluid-Dynamics (CFD) modeling of Heavy Liquid Metal (HLM) flows in pool configuration is (mainly) aimed at giving a technical support to the design of innovative GEN-IV nuclear reactors, such as ALFRED, for electricity production, or MYRRHA, as a multi-purpose research installation.This paper presents the CFD model of the primary coolant loop of MYRRHA version 1.4, built and operated with STAR-CCM+ versions 11–13. CFD is the only tool able to capture the 3D flow effects in the plena of a pool reactor. It is however still not feasible to simulate the detailed geometry of an entire reactor core by CFD. Fortunately, for many applications (safety studies, design) the exact geometrical representation of the core, heat exchangers, pumps and bypasses can be replaced by simplified models without loss of relevant information.These parts must be modeled such that they are functionally consistent with their role. Inspiration is taken from the system code techniques like homogenization procedures. A combination of solid and porous volumes is used within the CFD model.As with system code coupling, the homogenization can lead to loss of relevant information. In this case, the flow in the plena suffers from inadequate boundary conditions provided by the over-simplified modeled parts. To reach a sufficient level of maturity, the simplification procedure must be applied only where it is strictly necessary and at the right geometrical scale.First, an overview of the general features of the design and of the most important parameters that must be preserved by the CFD model is given. Then, we concentrate on the numerical model. Its topological structure is illustrated, based on a two-level geometrical and functional decomposition.Next, we focus on the core representation. Each sub-assembly (fuel assembly, dummy assembly, control rod and scram rod) has its own position. The core horizontal cross section is vertically propagated to represent the Above Core Structure (ACS). To obtain the right flow mixing the porosity of the various interfaces is calibrated.Due to the mixture of large and small length scales detailed meshing of the core and the ACS would result in very high mesh numbers. A specific procedure has been employed, to compensate cross sectional variations by physical property changes, maintaining the most important global characteristic. The ACS parameters have then been tuned in order to reduce the hot plenum surface temperature.Bypass flows have been implemented. Their role is to connect the fluid volumes and to avoid stagnation regions, which could cause potential corrosion issues. With the bypass flows and the simplified core/ACS structure models, a versatile CFD model for nominal operation is available.The Volume of Fluids (VoF) method used in the simulation allows the simulation of scenarios with variation of the free surface levels. Transient scenarios can be implemented to simulate operational incidents or accidents.To evaluate the safety of the reactor design, combinations of extreme parameter values are used in the simulations to demonstrate all safety criteria are met under worst case conditions. The simulations can also be used to define technical requirements of the system components (pumps and heat exchangers). Some examples are illustrated and partial conclusions are drawn.

Results of Studies and Life Tests of Russian-Made Monodisperse Nuclear-Grade Ion-Exchange Resins

The results of radiation-thermal tests of an experimental lot of Russian-made nuclear-grade monodisperse ion-exchange resins – strongly basic anion resin A-14chS (TOKEM-845-8) and strongly acidic cation resin K-14chS (TOKEM-145-9.5) – are reported. Data from life tests of ion-exchange resins in the ion-exchange filters of the primary loop of a full-scale prototype stand of a ship nuclear power facility are presented. The results of a comparative analysis of the physical-chemical properties of ion-exchange resins and their foreign-made analogues are reported. The obtained experimental data set gives a basis for recommending the ion-exchange resins A-14chS (TOKEM-845-8) and K-14chS (TOKEM-145-9.5) for use in ion-exchange filters in the primary coolant loop.

Modeling and Simulation for Coolant Activity and Dose Rate Estimation in a Pool-Type Nuclear Research Reactor

Coolant activity in the primary coolant system and reactor pool in an open pool-type research reactor is very important in view of operational and radiological safety considerations. As pool water acts as the shielding medium to minimize the radiation dose at the top of the reactor pool, an estimation of the activities of radionuclides in pool water is essential to establish the safety of the operating personnel and researchers working at the reactor pool top. A system is provided to create a hot water layer (HWL) at the top of the pool by supplying water at a temperature more than that of the pool water so that the dose rate at the reactor pool top can be minimized. This HWL system helps in breaking the natural convection current of reactor pool water by maintaining a higher temperature at this layer so that high-density pool water below this layer cannot replace this low-density HWL. Therefore, pool water that is comparatively more radioactive will not be able to cross this HWL by convection. Hence, diffusion will be the only mechanism by which radioactivity can reach the pool top. So eliminating the convection current keeps the activity at this topmost layer of the pool at a minimum value. The estimation of the activity of the radioactive nuclides is required to assess the radiation field at different locations in the primary coolant loop for designing proper shielding requirements of the system. The radionuclides of interest are the activation products of aluminum (24Na, 27Mg, 28Al) and 41Ar. In this technical note, a transient code is presented for estimating the activity of radioactive nuclides in the coolant loop and reactor pool of a nuclear research reactor. The reduced activity level at the pool top is estimated considering the presence of the HWL at the top of the pool. It is observed that purification flow plays a major role on the activity level of radioactive nuclides in reactor pool water. The variation of the activity dose rate at the reactor pool top with purification flow is also discussed.

Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

Abstract A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the Primarkreislauf-Versuchsanlage (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

Study of tritium transport characteristics in a transportable fluoride-salt-cooled high-temperature reactor

Tritium management is one of the most critical issues that limit the development of fluoride-salt-cooled high-temperature reactor (FHR); therefore, it is important to figure out the tritium transport characteristics in FHRs. In this paper, 3 works concerning about tritium in FHR are conducted: first, the tritium transport characteristics in the primary loop of FHRs are introduced, including tritium production and speciation, the absorption and desorption by graphite, dissolution and diffusion in molten salt, and permeation through structural materials. Second, the physical and mathematical models are established for tritium transport characteristic analysis in a transportable FHR (TFHR). The tritium transport characteristic analysis code (TAPAS) for TFHR is developed and benchmarked. The results prove the fidelity and accuracy of TAPAS. Finally, the tritium transport characteristics in the TFHR are analyzed systematically by TAPAS. Three conclusions are obtained: (1) tritium in the primary coolant loop is mainly in the form of T2; (2) when TFHR operates at steady state, the permeation rate of T2 can be regarded as a constant (9.03 × 109 Bq ⋅ EFPD−1); and (3) 7Li enrichment and redox potential of molten salt have great influence on the tritium distribution. This work might provide contribution to the tritium control in FHRs.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300–500 μm thick cladding.

Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor.Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for accident mitigation.

Qualitative and Quantitative Evaluation of Coupling Approaches for Coupling of RELAP5 and LabVIEW

In the study and design of a nuclear power plant, extensive system modeling is necessary to determine how the reactor will perform in any given situation, not only in the normal performance of the reactor, but also in transients including unanticipated transients without scram and hypothetical accidents. One type of nuclear power plant under study is the hybrid energy system, which uses nuclear power to generate both electricity and heat for facilities. Obviously, the second steam cycle in the nuclear power plant requires several design updates and experiments. Unfortunately, the current versions of the Reactor Excursion and Leak Analysis Program (RELAP) do not allow online data streams from experimental facilities to the computational model of the secondary steam loop. Therefore, this study develops a coupling between RELAP5 and Laboratory Virtual Instrument Engineering Workbench (LabVIEW) to model primary and secondary coolant loops. In this way, the LabVIEW model can easily be connected to an experimental apparatus to provide an online data stream and the online transient behavior of an entire nuclear power plant system. This study shows two different coupling approaches and makes qualitative and quantitative comparisons between these approaches.This paper demonstrates the results of different couplings between the primary and secondary systems of a typical pressurized water reactor (PWR). The primary loop model is a four-loop PWR. The model has been executed with steady state and transients (in this case, a loss-of-coolant accident). The results of both coupling methods have been compared with the typical RELAP5 results.