This is a M. Eng. degree project at Uppsala University carried out at the Forsmark nuclear power plant in Sweden. The purpose of it is to compare the two codes RELAP5 and TRACE during transient changes in mass flow against experiment. The change in mass flow will create a pressure wave and generate pipe loads. RELAP5 is a transient analysis code used to model thermal hydraulic systems. TRACE is an effort to combine the previous codes TRAC-B, TRAC-P, RAMONA and RELAP5. Both RELAP5 and TRACE has been compared to experiments. These comprise two abrupt valve closures, the closure of an inertial swing check valve (a flapper disc which closes when the flow is reversed) and a pump start and stop. Both RELAP5 and TRACE conforms well to the experiments with the abrupt valve closures. The check valve closes faster in the calculations compared to the experiment, both for RELAP5 as with TRACE. The amplitude of the pressure wave from the closure of the inertial swing check valve is lower compared to the experiment in both RELAP5 and TRACE. Numerical disturbances become visual as very high amplitudes in the time history diagram of the force in TRACE. The check valve oscillates between its open and closed position in RELAP5, but not in TRACE. Both RELAP5 and TRACE conforms well to the pump start. The mass flow decreases faster in both RELAP5 and TRACE compared to the pump stop.

Open pool research reactors play a crucial role in industry, medicine, scientific research and training. Ensuring its safety involves the use of widely accepted computer codes, such as RELAP5, that can predict the progression of accidents and evaluate reactor performance during transient events. These codes need a continuous validation process against various accident scenarios to ensure the reliability of the results. Two Loss of Heat Sink events (LOHS) took place previously in the McMaster Nuclear Reactor. One is the Loss of Forced Circulation in the Secondary Side event that happened in the year 2020, and the second is the Pool Temperature Experiment conducted at the McMaster Nuclear Reactor (MNR) in March 2023. These two events became a perfect opportunity to validate the safety analysis tools used by the NOF (Nuclear Operations and Facilities) staff. The focus was on validating the MNR RELAP5 model, particularly on the simulation of a loss of heat sink (LOHS) accident caused by the loss of the secondary pump. This study elucidates the validation results of the RELAP5 code for these two events and also under steady state conditions. A particular finding of this research was that reactor pool cooling transients prior to the start of a loss of heat sink accident (LOHS) can have an impact on the pool heating rate due to the pool thermal stratification. In these cases, the common assumption of an initial homogeneous temperature profile in the pool might not be accurate and could lead to underestimating the core temperature. With the help of CFD simulations it was possible to adjust the RELAP5 model, by providing a stratified temperature profile of the pool to be used as initial condition for the simulations. This led to more accurate estimations of the pool heating rate during the LOHS. Moreover, a sensitivity analysis on the pool nodalization showed that a minimum of two vertical pipes interconnected laterally by cross flow junctions is needed for the accurate analysis of this kind of transients. / Thesis / Master of Applied Science (MASc)

A segurança de instalações nucleares é uma preocupação mundial que tem crescido, sobretudo, após o acidente nuclear de Fukushima. O estudo de acidentes em reatores nucleares de pesquisa tal como o Acidente de Perda de Refrigerante (APR), considerado por muitas vezes um acidente base de projeto, é importante para garantir a integridade da instalação. O APR pode levar ao descobrimento parcial ou total do núcleo do reator e, como condição de segurança, deve-se garantir que haja a remoção do calor de decaimento dos elementos combustíveis. Esse trabalho teve o objetivo de realizar experimentos de descobrimento parcial e total no Elemento Combustível Instrumentado (ECI), construído no Instituto de Pesquisas Energética e Nucleares (IPEN), a fim de estudar os possíveis APRs em reatores de pesquisa. Uma seção de testes, denominada STAR, foi projetada e construída para simular os APRs. O ECI foi irradiado no núcleo do reator IEA-R1 (IPEN) e inserido na STAR, que ficou totalmente imersa na piscina do reator. No ECI, foram instalados termopares para medição das temperaturas do revestimento e do fluido em várias posições axiais e radiais. Foram realizados experimentos para cinco níveis de descobrimento do ECI, um total e quatro parciais, em duas condições distintas de calor de decaimento. Na análise dos resultados, verificou-se que os casos de descobrimento total foram os mais críticos, ou seja, as temperaturas do revestimento foram as maiores quando comparadas com os casos de descobrimentos parciais. Adicionalmente, foi realizada a simulação numérica de dois experimentos com o código RELAP5, cujos resultados demonstraram ótima concordância com os dos níveis experimentais, e temperaturas maiores que as experimentais. As máximas temperaturas do revestimento alcançadas em todos os experimentos ficaram bem abaixo da temperatura de empolamento do combustível, que é de 500°C. Assim, a STAR provou ser um aparato experimental seguro e confiável para a realização de experimentos de perda de refrigerante. / The safety of nuclear facilities has been a growing global concern mainly after the Fukushima nuclear accident. Studies on nuclear research reactor accidents such as the Loss of Coolant Accident (LOCA), considered many times a design basis accident, are important for guaranteeing the integrity of the plant. A LOCA may lead to the partial or complete uncovering of the fuel assemblies and assured decay heat removal is a safety condition. This work aimed to perform partial and complete uncovering experiments in the Instrumented Fuel Assembly (IFA) designed at the Instituto de Pesquisas Energéticas e Nucleares (IPEN) in order to study possible LOCAs in research reactors. A test section for experimental simulation of Loss of Coolant Accident named STAR was designed and built. The IFA was irradiated in the IEA-R1 and installed in the STAR, which was totally immersed in the reactor pool. Thermocouples were installed in the IFA to measure the clad and fluid temperatures in several axial and radial positions. The experiments were carried out for five levels of uncovering of IFA, being one complete uncovering and four partial uncovering, in two different conditions of decay heat. In the results analysis was observed that the cases of complete uncovering of the IFA were the most critical ones, that is, those cases presented higher clad temperatures when compared with partial uncovering cases. Additionally, a numerical simulation of two experiments was carried out by using the RELAP 5 code. The numerical results showed an optimum agreement with the experimental levels results and greater than the experimental temperatures. The maximum clad temperatures reached in all experiments were quite below the fuel blister temperature, which is 500 °C. Therefore, the STAR has proven to be a safe and reliable experimental apparatus for conducting loss of coolant experiments.

A segurança de instalações nucleares é uma preocupação mundial que tem crescido, sobretudo, após o acidente nuclear de Fukushima. O estudo de acidentes em reatores nucleares de pesquisa tal como o Acidente de Perda de Refrigerante (APR), considerado por muitas vezes um acidente base de projeto, é importante para garantir a integridade da instalação. O APR pode levar ao descobrimento parcial ou total do núcleo do reator e, como condição de segurança, deve-se garantir que haja a remoção do calor de decaimento dos elementos combustíveis. Esse trabalho teve o objetivo de realizar experimentos de descobrimento parcial e total no Elemento Combustível Instrumentado (ECI), construído no Instituto de Pesquisas Energética e Nucleares (IPEN), a fim de estudar os possíveis APRs em reatores de pesquisa. Uma seção de testes, denominada STAR, foi projetada e construída para simular os APRs. O ECI foi irradiado no núcleo do reator IEA-R1 (IPEN) e inserido na STAR, que ficou totalmente imersa na piscina do reator. No ECI, foram instalados termopares para medição das temperaturas do revestimento e do fluido em várias posições axiais e radiais. Foram realizados experimentos para cinco níveis de descobrimento do ECI, um total e quatro parciais, em duas condições distintas de calor de decaimento. Na análise dos resultados, verificou-se que os casos de descobrimento total foram os mais críticos, ou seja, as temperaturas do revestimento foram as maiores quando comparadas com os casos de descobrimentos parciais. Adicionalmente, foi realizada a simulação numérica de dois experimentos com o código RELAP5, cujos resultados demonstraram ótima concordância com os dos níveis experimentais, e temperaturas maiores que as experimentais. As máximas temperaturas do revestimento alcançadas em todos os experimentos ficaram bem abaixo da temperatura de empolamento do combustível, que é de 500°C. Assim, a STAR provou ser um aparato experimental seguro e confiável para a realização de experimentos de perda de refrigerante. / The safety of nuclear facilities has been a growing global concern mainly after the Fukushima nuclear accident. Studies on nuclear research reactor accidents such as the Loss of Coolant Accident (LOCA), considered many times a design basis accident, are important for guaranteeing the integrity of the plant. A LOCA may lead to the partial or complete uncovering of the fuel assemblies and assured decay heat removal is a safety condition. This work aimed to perform partial and complete uncovering experiments in the Instrumented Fuel Assembly (IFA) designed at the Instituto de Pesquisas Energéticas e Nucleares (IPEN) in order to study possible LOCAs in research reactors. A test section for experimental simulation of Loss of Coolant Accident named STAR was designed and built. The IFA was irradiated in the IEA-R1 and installed in the STAR, which was totally immersed in the reactor pool. Thermocouples were installed in the IFA to measure the clad and fluid temperatures in several axial and radial positions. The experiments were carried out for five levels of uncovering of IFA, being one complete uncovering and four partial uncovering, in two different conditions of decay heat. In the results analysis was observed that the cases of complete uncovering of the IFA were the most critical ones, that is, those cases presented higher clad temperatures when compared with partial uncovering cases. Additionally, a numerical simulation of two experiments was carried out by using the RELAP 5 code. The numerical results showed an optimum agreement with the experimental levels results and greater than the experimental temperatures. The maximum clad temperatures reached in all experiments were quite below the fuel blister temperature, which is 500 °C. Therefore, the STAR has proven to be a safe and reliable experimental apparatus for conducting loss of coolant experiments.

Choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant (NPP), but also to everyday operation. Current pressurized water reactor steam generators operate on the leak-before-break approach. The ability to predict and estimate a leak rate through a steam generator tube crack is an important safety parameter. Knowledge of the maximum flow rate through a crack in the steam generator tube allows the coolant inventory to be monitored accordingly. Here an assessment of the choking flow models in thermal-hydraulics code RELAP5/MOD3.3 is performed and its suitability to predict choking flow rates through small simulated cracks of steam generator tubes is evaluated based on collected experimental data. Six samples of the data were studied in this work which correspond to steam generator tube crack<br>samples 6-11. Each sample has a wall thickness, channel length (L), of 1.14 mm. Exit areas of these samples, 6-11, are 2.280E-06 m^2, 2.493E-06 m^2, 1.997E-06 m^2, 1.337E-06 m^2, and 2.492E-06. Samples 6-11 have a channel length to hydraulics diameter ratio (L/D) between 3.0-5.3. Two separate pressure differentials of 6.89 MPa and 4.13 MPa were applied across the samples with a range of subcooling from 20℃ to 80℃ and 20℃ to 60℃. Flow rates through these samples were modeled using the thermal-hydraulic system code RELAP5/MOD3.3. Simulation results are compared to experimental values and modeling techniques are discussed. It is found that both the Henry-Fauske and Ransom-Trapp models better predict choking mass flux for longer channels. <br> <br> <br>

The steam generator plays a primary role in the safety and performance of a pressurized water reactor nuclear power plant. The cost to utilities is in the order of millions of Rands a year as a direct result of damage to steam generators. The damage results in lower efficiency or even plant shutdown. It is necessary for the utility and for academia to have models of nuclear components by which research and analysis may be performed. It must be possible to analyse steam generator performance for both day-to-day operational analysis as well as in the case of extreme accident scenarios. The homogeneous model for two-phase flow is simpler in its implementation than the two-fluid model, and therefore suffers in accuracy. Its advantage lies in its quick turnover time for development of models and subsequent analysis. It is often beneficial for a modeller to be able to quickly set up and analyse a model of a system, and a trade-off between accuracy and time-management is thus required. Searches through available literature failed to provide answers to how the homogeneous model compares with the two-fluid model for operational and safety analysis. It is expected to see variations between the models, from the analysis of the mathematics, but it remains to be shown what these differences are. The purpose of this study was to determine how the homogeneous model for two-phase flow compares with the two-fluid model when applied to a u-tube steam generator of a typical pressurized water reactor. The steam generator was modelled in both RELAP5 and in Flownex. A custom script was written for Flownex in order to implement the Chen correlation for boiling heat transfer. This was significantly less detailed than RELAP5’s solution of a matrix of flow regimes and heat transfer correlations. The geometry of the models were based on technical drawings from Koeberg Nuclear Power Plant, and were simplified to a one-dimensional model. Plant data obtained from Koeberg was used to validate the models at 100%, 80% and 60% power output. It was found that the overall heat transfer rate predicted with the RELAP5 two-fluid model was within 1.5% of the measured data from the Koeberg plant. The results generated by the homogeneous model for the overall heat transfer were within 4.5% of the measured values. However, the differences in the detailed temperature distributions and heat transfer coefficient values were quite significant at the inlet and outlet ends of the tube bundle, at the bottom tube sheet of the steam generator. In this area the water-level was not accurately modelled by the homogeneous model, and therefore there was an under-prediction in heat transfer in that region. Large differences arose between the Flownex and RELAP5 solutions due to difference in the heat transfer correlations used. The Flownex model exclusively implemented the Chen correlation, while RELAP5 implements a flow regime map correlated to a table of heat transfer correlations. It was concluded that the results from the homogeneous model for two-phase flow do not differ significantly when compared with the two-fluid model when applied to the u-tube steam generator at the normal operating conditions. Significant differences do, however, occur in lower regions of the boiler where the quality is lower. We conclude that the homogeneous model offers significant advantage in simplicity over the two-fluid model for normal operational analysis. This may not be the case for detailed accident analysis, which was beyond the scope of this study. / Thesis (MIng (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2013.

The steam generator plays a primary role in the safety and performance of a pressurized water reactor nuclear power plant. The cost to utilities is in the order of millions of Rands a year as a direct result of damage to steam generators. The damage results in lower efficiency or even plant shutdown. It is necessary for the utility and for academia to have models of nuclear components by which research and analysis may be performed. It must be possible to analyse steam generator performance for both day-to-day operational analysis as well as in the case of extreme accident scenarios. The homogeneous model for two-phase flow is simpler in its implementation than the two-fluid model, and therefore suffers in accuracy. Its advantage lies in its quick turnover time for development of models and subsequent analysis. It is often beneficial for a modeller to be able to quickly set up and analyse a model of a system, and a trade-off between accuracy and time-management is thus required. Searches through available literature failed to provide answers to how the homogeneous model compares with the two-fluid model for operational and safety analysis. It is expected to see variations between the models, from the analysis of the mathematics, but it remains to be shown what these differences are. The purpose of this study was to determine how the homogeneous model for two-phase flow compares with the two-fluid model when applied to a u-tube steam generator of a typical pressurized water reactor. The steam generator was modelled in both RELAP5 and in Flownex. A custom script was written for Flownex in order to implement the Chen correlation for boiling heat transfer. This was significantly less detailed than RELAP5’s solution of a matrix of flow regimes and heat transfer correlations. The geometry of the models were based on technical drawings from Koeberg Nuclear Power Plant, and were simplified to a one-dimensional model. Plant data obtained from Koeberg was used to validate the models at 100%, 80% and 60% power output. It was found that the overall heat transfer rate predicted with the RELAP5 two-fluid model was within 1.5% of the measured data from the Koeberg plant. The results generated by the homogeneous model for the overall heat transfer were within 4.5% of the measured values. However, the differences in the detailed temperature distributions and heat transfer coefficient values were quite significant at the inlet and outlet ends of the tube bundle, at the bottom tube sheet of the steam generator. In this area the water-level was not accurately modelled by the homogeneous model, and therefore there was an under-prediction in heat transfer in that region. Large differences arose between the Flownex and RELAP5 solutions due to difference in the heat transfer correlations used. The Flownex model exclusively implemented the Chen correlation, while RELAP5 implements a flow regime map correlated to a table of heat transfer correlations. It was concluded that the results from the homogeneous model for two-phase flow do not differ significantly when compared with the two-fluid model when applied to the u-tube steam generator at the normal operating conditions. Significant differences do, however, occur in lower regions of the boiler where the quality is lower. We conclude that the homogeneous model offers significant advantage in simplicity over the two-fluid model for normal operational analysis. This may not be the case for detailed accident analysis, which was beyond the scope of this study. / Thesis (MIng (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2013.

Nuclear power plants rely on the Emergency Core Cooling System (ECCS) to cool down the reactor core in case of an accident. Occasionally, air is entrained into the suction piping of ECCS causing voids that decrease pumping efficiency, and consequently damage the pumps. In an attempt to minimize the amount of voids entering the suction side of the pump in ECCS, a Void Recirculation System (VRS) experiment was conducted for a proof of concept purpose. While many studies have been oriented in studying two-component flow behavior in ECCS, none of them propose a solution to minimize air entrainment. As a consequence, there are no simulation models that use computational fluid dynamics to address gas entrainment solutions in ECCS. The objectives of this thesis are to (1) simulate and investigate two-component air-water flow in a VRS that minimizes the amount of air in piping systems, using RELAP5/MOD3 as the computational tool, and (2) to validate the numerical results with respect to experimental results and observations. A one-dimensional model of the VRS was built in RELAP5, in which eight different scenarios (replicating those from the VRS experiment) were simulated for a period of 150 seconds. Four Froude numbers of 0.8, 1.0, 1.3 and 1.6 were evaluated in two different pipe configurations, and the experimental data obtained from the VRS experiment was used to validate the numerical results obtained from these simulations. It was concluded that air recirculation occurs indefinitely throughout the entire 150 seconds of the simulation for Froude numbers up to 1.3; while for a Froude number of 1.6, air recirculation occurs for approximately 100 seconds and ceases after 125 seconds of the simulation. An average air reduction effectiveness of 90% was found for all simulation scenarios. The VRS model was successfully validated and can be used to investigate the effects of air entrainment in suction piping.

This study explores the behavior of a Boiling Water Reactor (BWR) during a turbine trip scenario initiated by the abrupt closure of the turbine stop valve. The RELAP5/Mod3.4 code is employed to make calculations using the Laguna Verde Nuclear Power Plant input model provided by Innovative Software Systems Company. The event sequences and initial boundary conditions are sourced from the Boiling Water Reactor Turbine Trip 2 Benchmark created by NEA. Results are subsequently compared against the benchmark values. In order to gauge the risk of a turbine trip event leading to elevated power, which could in turn cause Critical Heat Flux (CHF)-related issues in cladding temperature, a best-estimate case is developed. Our findings indicate that the closure of the turbine stop valve (TSV) resulted in a collapse of the void fraction within the reactor core. Although the core power doubled the initial level, the negative feedback mechanism effectively suppressed the power pulse. Throughout the transient phase, the maximum cladding temperature stayed below the CHF threshold, a fact attributable to the fuel's conductivity and the rapid progression of the transient. We further analyzed three hypothetical scenarios to test the computational boundaries of the plant model. The third scenario, which combines conditions from the first two, produced elevated outcomes (6500MW core power, 598K cladding temperature, and 7900kPa dome pressure) as expected. Notably, while the CHF limit remained unbreached in this scenario, literature reviews suggest potential core meltdown risks in subsequent stages of this calculation. Our sensitivity analyses determined that variations in the gamma heating coefficient or the maximum time step of the calculations have little to no impact on core power or peak cladding temperature. Conversely, we noted a significant reduction, approximately 35\%, in the power peak, underscoring the high sensitivity of the parameters to the initial triggering of the SCRAM mechanism. Our results also recommend rapid and early actuation of the BPV as a measure to dampen the pressure wave, consequently decreasing both the power peak and peak cladding temperatures. / Thesis / Master of Applied Science (MASc) / This research investigates the response of the Laguna Verde Boiling Water Reactor to a turbine trip event using the RELAP5/Mod3.4 thermal-hydraulic analysis code. From reactor safety perspective a best-estimate case is evaluated, as well as three additional hypothetical scenarios. Findings are compared with the Boiling Water Reactor Turbine Trip II Benchmark results. Additionally, sensitivity analyses focusing on plant parameters such as shutdown rod behavior, gamma heating coefficient, turbine stop valve, and steam bypass valve characteristics conducted to determine their impact on the results. Insights from these analyses aim to enhance safety protocols and refine best practices in boiling water reactor management.

With the growing capacity of renewable energy production sources, nuclear energy, once a mainstay of power generation, faces challenges due to its limited adaptability to fluctuating energy demands. This inherent rigidity makes it less desirable than the more flexible renewable sources. However, integrating thermal energy storage (TES) systems offers a promising avenue, enabling nuclear power plants (NPPs) to enhance their operational flexibility and remain competitive in an evolving renewable market. A comprehensive ranking methodology has been introduced, delineating the criteria and processes to determine the most synergistic TES/NPP design couplings. This methodology considers the unique characteristics of both current and prospective reactor fleets, ensuring broad applicability across various nuclear technologies. Economic analysis further supports the case for TES integration. Findings indicate that when equipped with TES systems, NPPs can remain price competitive, even with carbon-neutral alternatives like solar power generation. A lab-scale TES system was meticulously designed and constructed to validate these theoretical propositions. For its control, the Python GEKKO model predictive control (MPC) was employed, a decision influenced by the proven efficacy of GEKKO in managing complex systems. Tests conclusively demonstrated the feasibility and efficiency of using GEKKO for MPC of TES systems. A novel methodology for the MPC of a RELAP5-3D input deck has been proposed and elaborated upon. This methodology was rigorously tested at two distinct scales. The initial focus was on a thermal-hydraulic model of the lab-scale TES system. Subsequent efforts scaled up to control a more intricate thermal-hydraulic model, representing a small modular reactor (SMR) paired with an oil-based TES system. In both scenarios, GEKKO exhibited exemplary performance, controlling the RELAP5-3D models with precision and ensuring they met the stipulated demand parameters. The research underscores the potential of RELAP5-3D MPC in streamlining the licensing process for TES systems intended for NPP coupling. This approach could eliminate the need for expensive and time-consuming experiments, paving the way for more efficient and cost-effective nuclear energy solutions.