Validation Studies of Thermal-hydraulic Code for Safety Analysis of Nuclear Power Plants

This thesis gives an overview of the validation process for thermal hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. The cases presented are not exhaustive, but they give a good overview of the work performed by the personnel of Lappeenranta University of Technology (LUT). Large part of the work has been performed in co-operation with the CATHARE-team in Grenoble, France. The design of a Russian type pressurized water reactor, VVER, differs from that of a Western-type PWR. Most of thermal-hydraulic system codes are validated only for the Western-type PWRs. Thus, the codes should be assessed and validated also for VVER design in order to establish any weaknesses in the models. This information is needed before codes can be used for the safety analysis. Theresults of the assessment and validation calculations presented here show that the CATHARE code can be used also for the thermal-hydraulic safety studies for VVER type plants. However, some areas have been indicated which need to be reassessed after further experimental data become available. These areas are mostly connected to the horizontal stem generators, like condensation and phase separation in primary side tubes. The work presented in this thesis covers a large numberof the phenomena included in the CSNI code validation matrices for small and intermediate leaks and for transients. Also some of the phenomena included in the matrix for large break LOCAs are covered. The matrices for code validation for VVER applications should be used when future experimental programs are planned for code validation.

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

Analysis of PWR-PACTEL Small Break LOCA experiment using TRACE

Currently, the power generation is one of the most significant life aspects for the whole man-kind. Barely one can imagine our life without electricity and thermal energy. Thus, different technologies for producing those types of energy need to be used. Each of those technologies will always have their own advantages and disadvantages. Nevertheless, every technology must satisfy such requirements as efficiency, ecology safety and reliability. In the matter of the power generation with nuclear energy utilization these requirements needs to be highly main-tained, especially since accidents on nuclear power plants may cause very long term deadly consequences. In order to prevent possible disasters related to the accident on a nuclear power plant strong and powerful algorithms were invented in last decades. Such algorithms are able to manage calculations of different physical processes and phenomena of real facilities. How-ever, the results acquired by the computing must be verified with experimental data.

Experimental thermal hydraulic studies on the enhancement of safety of LWRs

The safe use of nuclear power plants (NPPs) requires a deep understanding of the functioning of physical processes and systems involved. Studies on thermal hydraulics have been carried out in various separate effects and integral test facilities at Lappeenranta University of Technology (LUT) either to ensure the functioning of safety systems of light water reactors (LWR) or to produce validation data for the computer codes used in safety analyses of NPPs. Several examples of safety studies on thermal hydraulics of the nuclear power plants are discussed. Studies are related to the physical phenomena existing in different processes in NPPs, such as rewetting of the fuel rods, emergency core cooling (ECC), natural circulation, small break loss-of-coolant accidents (SBLOCA), non-condensable gas release and transport, and passive safety systems. Studies on both VVER and advanced light water reactor (ALWR) systems are included. The set of cases include separate effects tests for understanding and modeling a single physical phenomenon, separate effects tests to study the behavior of a NPP component or a single system, and integral tests to study the behavior of the whole system. In the studies following steps can be found, not necessarily in the same study. Experimental studies as such have provided solutions to existing design problems. Experimental data have been created to validate a single model in a computer code. Validated models are used in various transient analyses of scaled facilities or NPPs. Integral test data are used to validate the computer codes as whole, to see how the implemented models work together in a code. In the final stage test results from the facilities are transferred to the NPP scale using computer codes. Some of the experiments have confirmed the expected behavior of the system or procedure to be studied; in some experiments there have been certain unexpected phenomena that have caused changes to the original design to avoid the recognized problems. This is the main motivation for experimental studies on thermal hydraulics of the NPP safety systems. Naturally the behavior of the new system designs have to be checked with experiments, but also the existing designs, if they are applied in the conditions that differ from what they were originally designed for. New procedures for existing reactors and new safety related systems have been developed for new nuclear power plant concepts. New experiments have been continuously needed.

Modeling of Pressurizer Using APROS and TRACE Thermal Hydraulic Codes

Diplomityön tavoitteena on paineistimen yksityiskohtainen mallintaminen APROS- ja TRACE- termohydrauliikkaohjelmistoja käyttäen. Rakennetut paineistinmallit testattiin vertaamalla laskentatuloksia paineistimen täyttymistä, tyhjentymistä ja ruiskutusta käsittelevistä erilliskokeista saatuun mittausdataan. Tutkimuksen päätavoitteena on APROSin paineistinmallin validoiminen käyttäen vertailuaineistona PACTEL ATWS-koesarjan sopivia paineistinkokeita sekä MIT Pressurizer- ja Neptunus- erilliskokeita. Lisäksi rakennettiin malli Loviisan ydinvoimalaitoksen paineistimesta, jota käytettiin turbiinitrippitransientin simulointiin tarkoituksena selvittää mahdolliset voimalaitoksen ja koelaitteistojen mittakaavaerosta johtuvat vaikutukset APROSin paineistinlaskentaan. Kokeiden simuloinnissa testattiin erilaisia noodituksia ja mallinnusvaihtoehtoja, kuten entalpian ensimmäisen ja toisen kertaluvun diskretisointia, ja APROSin sekä TRACEn antamia tuloksia vertailtiin kattavasti toisiinsa. APROSin paineistinmallin lämmönsiirtokorrelaatioissa havaittiin merkittävä puute ja laskentatuloksiin saatiin huomattava parannus ottamalla käyttöön uusi seinämälauhtumismalli. Työssä tehdyt TRACE-simulaatiot ovat osa United States Nuclear Regulatory Commissionin kansainvälistä CAMP-koodinkehitys-ja validointiohjelmaa.

Thermal-Hydraulic Studies on the Safety Of VVER-440 Type Nuclear Power Plants

This thesis includes several thermal hydraulic analyses related to the Loviisa WER 440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transits and development of a calculational model for calculation of boric acid concentrations in the reactor. In the first part of the thesis, in the case of won of boric acid solution behaviour during long term cooling period of LOCAs, experiments were performed in scaled down test facilities. The experimental data together with the results of RELAPS/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. The results of calculations showed that margins to critical concentrations that would lead to boric acid crystallization were large, both in the reactor core and in the lower plenum. This was mainly caused by the fact that water in the primary cooling circuit includes borax (Na)BsO,.IOHZO), which enters the reactor when ECC water is taken from the sump and greatly increases boric acid solubility in water. In the second part, in the case of simulation of horizontal steam generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments, as well as earlier REWET III natural circulation tests, were analyzed with RELAPS/MOD3 Version Sm5 code. The analysis showed that the code was capable of simulating the main events during the experiments. However, in the case of loss of secondary side feedwater the code was not completely capable to simulate steam superheating in the secondary side of the steam generators. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAPSlMODI Eur, RELAPS/MOD3 and CATHARE codes. All three codes were capable to simulate the two selected pump trip transients and no significant differences were found between the results of different codes. Comparison of the calculated results with the data measured in the Loviisa plant also showed good agreement.

VVER-440 Thermal Hydraulics as a Computer Code Validation Challenge

This thesis concentrates on the validation of a generic thermal hydraulic computer code TRACE under the challenges of the VVER-440 reactor type. The code capability to model the VVER-440 geometry and thermal hydraulic phenomena specific to this reactor design has been examined and demonstrated acceptable. The main challenge in VVER-440 thermal hydraulics appeared in the modelling of the horizontal steam generator. The major challenge here is not in the code physics or numerics but in the formulation of a representative nodalization structure. Another VVER-440 specialty, the hot leg loop seals, challenges the system codes functionally in general, but proved readily representable. Computer code models have to be validated against experiments to achieve confidence in code models. When new computer code is to be used for nuclear power plant safety analysis, it must first be validated against a large variety of different experiments. The validation process has to cover both the code itself and the code input. Uncertainties of different nature are identified in the different phases of the validation procedure and can even be quantified. This thesis presents a novel approach to the input model validation and uncertainty evaluation in the different stages of the computer code validation procedure. This thesis also demonstrates that in the safety analysis, there are inevitably significant uncertainties that are not statistically quantifiable; they need to be and can be addressed by other, less simplistic means, ultimately relying on the competence of the analysts and the capability of the community to support the experimental verification of analytical assumptions. This method completes essentially the commonly used uncertainty assessment methods, which are usually conducted using only statistical methods.

Development of a PWR-W and an AP1000® containment building 3D model with a CFD code for Best-Estimate Thermal-Hydraulic Analysis

The simulation of design basis accidents in a containment building is usually conducted with a lumped parameter model. The codes normally used by Westinghouse Electric Company (WEC) for that license analysis are WGOTHIC or COCO, which are suitable to provide an adequate estimation of the overall peak temperature and pressure of the containment. However, for the detailed study of the thermal-hydraulic behavior in every room and compartment of the containment building, it could be more convenient to model the containment with a more detailed 3D representation of the geometry of the whole building. The main objective of this project is to obtain a standard PWR Westinghouse as well as an AP1000® containment model for a CFD code to analyze the thermal-hydraulic detailed behavior during a design basis accident. In this paper the development and testing of both containment models is presented.

CHARACTERIZATION AND THERMAL-HYDRAULIC MODELING OF ENGINEERED BARRIERS FOR A CANADIAN DEEP GEOLOGICAL REPOSITORY FOR USED NUCLEAR FUEL

Within Canada there are more than 2.5 million bundles of spent nuclear fuel with another approximately 2 million bundles to be generated in the future. Canada, and every country around the world that has taken a decision on management of spent nuclear fuel, has decided on long-term containment and isolation of the fuel within a deep geological repository. At depth, a deep geological repository consists of a network of placement rooms where the bundles will be located within a multi-layered system that incorporates engineered and natural barriers. The barriers will be placed in a complex thermal-hydraulic-mechanical-chemical-biological (THMCB) environment. A large database of material properties for all components in the repository are required to construct representative models. Within the repository, the sealing materials will experience elevated temperatures due to the thermal gradient produced by radioactive decay heat from the waste inside the container. Furthermore, high porewater pressure due to the depth of repository along with possibility of elevated salinity of groundwater would cause the bentonite-based materials to be under transient hydraulic conditions. Therefore it is crucial to characterize the sealing materials over a wide range of thermal-hydraulic conditions. A comprehensive experimental program has been conducted to measure properties (mainly focused on thermal properties) of all sealing materials involved in Mark II concept at plausible thermal-hydraulic conditions. The thermal response of Canada’s concept for a deep geological repository has been modelled using experimentally measured thermal properties. Plausible scenarios are defined and the effects of these scenarios are examined on the container surface temperature as well as the surrounding geosphere to assess whether they meet design criteria for the cases studied. The thermal response shows that if all the materials even being at dried condition, repository still performs acceptably as long as sealing materials remain in contact.

Computational evaluation of hydraulic system behaviour with entrapped air under rapid pressurization

The pressurization of hydraulic systems containing entrapped air is considered a critical condition for the infrastructure's security due to transient pressure variations often occurred. The objective of the present study is the computational evaluation of trends observed in variation of maximum surge pressure resulting from rapid pressurizations. The comparison of the results with those obtained in previous studies is also undertaken. A brief state of art in this domain is presented. This research work is applied to an experimental system having entrapped air in the top of a vertical pipe section. The evaluation is developed through the elastic model based on the method of characteristics, considering a moving liquid boundary, with the results being compared with those achieved with the rigid liquid column model.

Computational evaluation of hydraulic system behaviour with entrapped air under rapid pressurization

The pressurization of hydraulic systems containing entrapped air is considered a critical condition for the infrastructure's security due to transient pressure variations often occurred. The objective of the present study is the computational evaluation of trends observed in variation of maximum surge pressure resulting from rapid pressurizations. The comparison of the results with those obtained in previous studies is also undertaken. A brief state of art in this domain is presented. This research work is applied to an experimental system having entrapped air in the top of a vertical pipe section. The evaluation is developed through the elastic model based on the method of characteristics, considering a moving liquid boundary, with the results being compared with those achieved with the rigid liquid column model.

Electric-drive-based control and electric energy regeneration in a hydraulic system

Over the recent years, development in mobile working machines has concentrated on reducing emissions owing to the tightening rules and needs to improve energy utilization and reduce power losses. This study focuses on energy utilization and regeneration in an electro-hydraulic forklift, which is a lifting equipment application. The study starts from the modelling and simulation of a hydraulic forklift. The energy regeneration from the potential energy of the load was studied. Also a flow-based electric motor speed control was suggested in this thesis instead of the throttle control method or the variable displacement pump control. Topics related to further development in the future are discussed. Finally, a summary and conclusions are presented.

Visualization of thermal-hydraulic nodes

The purpose of this thesis was to develop a program that can illustrate thermal-hydraulic node dimensions used in SMABRE simulations. These created node illustrations are used to verify the correctness of the designed simulation model and in addition they can be included in scientific reports. This thesis will include theory about SMABRE and relevant programs that were used to achieve the ending results. This thesis will give explanations for different modules that were created and used in the finished program, and it will present the different problems encountered and provide the solutions. The most important objective in this thesis is to display the results of generic VVER-1000 node dimensions and verify the correctness in the displayed part. The finished program was created using code language Python.