Radiocarbon impact in a nearby tree of a light-water VVER type nuclear power plant, Paks, Hungary

Tree-ring series were collected for radiocarbon analyses from the vicinity of Paks nuclear power plant (NPP) and a background area (Dunaföldvár) for a 10-yr period (2000–2009). Samples of holocellulose were prepared from the wood and converted to graphite for accelerator mass spectrometry (AMS) 14C measurement using the MICADAS at ETH Zürich. The 14C concentration data from these tree rings was compared to the background tree rings for each year. The global decreasing trend of atmospheric 14C activity concentration was observed in the annual tree rings both in the background area and in the area of the NPP. As an average of the past 10 yr, the excess 14C emitted by the pressurized-water reactor (PWR) NPP to the atmosphere shows only a slight systematic excess (~6‰) 14C in the annual rings. The highest 14C excess was 13‰ (in 2006); however, years with the same 14C level as the background were quite frequent in the tree-ring series.

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.

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.

Modelado detallado de una central nuclear CANDU con CITVAP.

La línea de cálculo de INVAP consiste principalmente de los códigos CONDOR y CITVAP. Este último es la versión mejorada del código CITATION II que resuelve la ecuación de difusión neutrónica multigrupo por el método de diferencias finitas. CITVAP es ampliamente usado para estudiar reactores de investigación y reactores de potencia tales como PWR, BWR, VVER y últimamente se implemento nuevas funciones para estudiar una central PHWR tipo Atucha. Siguiendo con la línea de reactores PHWR, en este trabajo se estudian las capacidades y deficiencias del código de núcleo CITVAP para modelar una central nuclear tipo CANDU. Se plantean mejoras a realizar para un manejo mas eficiente desde el punto de vista del usuario, tanto de la gestión de combustibles, movimientos de barras de control y zonas líquidas como mejoras en el modelo termohidraulico. La metodología consiste en validar la línea de cálculo de INVAP, contrastando los resultados con el benchmark IAEA-tecdoc-887. El proceso de validación consiste en cálculos de celda en dos y tres dimensiones usando los códigos CONDOR y SERPENT respectivamente, obtención de secciones eficaces macroscópicas en función del quemado y cálculos de núcleo para distintas configuraciones de los dispositivos de control usando un núcleo fresco y una distribución de quemado en equilibrio. Se analizan las dificultades que se presentan al modelar el núcleo con las capacidades actuales del código y se plantean posibles soluciones a implementar. Para un estudio completo de un reactor CANDU, se estudian tres de la características distintivas de este tipo de reactor: la termohidraulica, la gestión de combustibles y los dispositivos de control de reactividad, distribución de potencia y apagado.

Modelado detallado de una central nuclear CANDU con CITVAP.

La línea de cálculo de INVAP consiste principalmente de los códigos CONDOR y CITVAP. Este último es la versión mejorada del código CITATION II que resuelve la ecuación de difusión neutrónica multigrupo por el método de diferencias finitas. CITVAP es ampliamente usado para estudiar reactores de investigación y reactores de potencia tales como PWR, BWR, VVER y últimamente se implemento nuevas funciones para estudiar una central PHWR tipo Atucha. Siguiendo con la línea de reactores PHWR, en este trabajo se estudian las capacidades y deficiencias del código de núcleo CITVAP para modelar una central nuclear tipo CANDU. Se plantean mejoras a realizar para un manejo mas eficiente desde el punto de vista del usuario, tanto de la gestión de combustibles, movimientos de barras de control y zonas líquidas como mejoras en el modelo termohidraulico. La metodología consiste en validar la línea de cálculo de INVAP, contrastando los resultados con el benchmark IAEA-tecdoc-887. El proceso de validación consiste en cálculos de celda en dos y tres dimensiones usando los códigos CONDOR y SERPENT respectivamente, obtención de secciones eficaces macroscópicas en función del quemado y cálculos de núcleo para distintas configuraciones de los dispositivos de control usando un núcleo fresco y una distribución de quemado en equilibrio. Se analizan las dificultades que se presentan al modelar el núcleo con las capacidades actuales del código y se plantean posibles soluciones a implementar. Para un estudio completo de un reactor CANDU, se estudian tres de la características distintivas de este tipo de reactor: la termohidraulica, la gestión de combustibles y los dispositivos de control de reactividad, distribución de potencia y apagado.

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.

Tritiumin kulkeutumisreitit ja tritiuminventaari Loviisan voimalaitoksella

Tämän diplomityön tavoitteena oli selvittää Loviisan voimalaitoksen vesijärjestelmien tritiuminventaari sekä tritiumin kulkeutuminen eri järjestelmissä. Kulkeutumisreittien kartoittamisessa keskityttiin erityisesti kaasujenkäsittely- ja ilmastointiprosesseihin, ilmapäästöjen lähteen selvittämiseksi. Työssä perehdyttiin tritiumin tuottoon ja päästöihin Loviisan VVER- tyyppisessä ydinvoimalaitoksessa. Lisäksi esiteltiin selvitystä varten tarkasteltavat voimalaitoksen vesiilmastointi- ja kaasujenkäsittelyjärjestelmät. Työssä esiteltiin myös tritiumpäästöjä sekä tritiumin mittausmenetelmiä Loviisan voimalaitoksella. Tritiuminventaarin selvittämiseksi otettiin vesinäytteitä niiden järjestelmien säiliöistä, joiden oli tunnistettu sisältävän tritiumpitoisia vesiä. Tritiumin kulkeutumisreittien selvittämiseksi vesinäytteiden lisäksi otettiin ilmasta ja eri järjestelmistä näytteitä tritiumkerääjillä. Tehtyjen mittausten perusteella määritettiin Loviisan voimalaitoksen tritiuminventaari. Tarkasteltujen järjestelmien sisältämä tritiuminventaari on alle puolet normaalista vuotuisesta tritiumvesipäästöstä eikä sen vuoksi aiheuta toimenpiteitä käytöstä poistoa ajatellen. Ilmapäästöjen tritiumin kulkeutumisreittien määrityksessä tunnistettiin pääreiteiksi TL-ilmastoinnit. Kaasujenkäsittelyjärjestelmien osuus ilmapäästöjen tritiumista oli huomattavasti pienempi.

Analysis of PACTEL small break LOCA experiment using TRACE

A small break loss-of-coolant accident (SBLOCA) is one of problems investigated in an NPP operation. Such accident can be analyzed using an experiment facility and TRACE thermal-hydraulic system code. A series of SBLOCA experiments was carried out on Parallel Channel Test Loop (PACTEL) facility, exploited together with Technical Research Centre of Finland VTT Energy and Lappeenranta University of Technology (LUT), in order to investigate two-phase phenomena related to a VVER-type reactor. The experiments and a TRACE model of the PACTEL facility are described in the paper. In addition, there is the TRACE code description with main field equations. At the work, calculations of a SBLOCA series are implemented and after the calculations, the thesis discusses the validation of TRACE and concludes with an assessment of the usefulness and accuracy of the code in calculating small breaks.

Crystallography of zirconium hydrides in recrystallized Zircaloy-2 fuel cladding lay electron backscatter diffraction

Precipitation morphology and habit planes of the delta-phase Zr hydrides, which were precipitated within the a-phase matrix grains and along the grain boundaries of recrystallized Zircaloy-2 cladding tube, have been examined by electron backscatter diffraction (EBSD). Radially-oriented hydrides, induced by residual tensile stress, precipitated in the outside region of the cladding, and circumferentially-oriented hydrides in the stress-free middle region of the cladding. The most common crystallographic relationship for both types of the hydrides precipitated at the inter- and intra-granular sites was identical at (0001)(alpha) // {111}(delta), with {1017}(alpha) // {111}(delta) being the occasional exception only for the inter-granular radial hydrides. When tensile stress was loaded, the intra-granular hydrides tended to preferentially precipitate in the grains with circumferential basal pole textures. The inter-granular hydrides tended to preferentially precipitate on the grain faces opposite to tensile axis. The change of prioritization in the precipitation sites for the hydrides due to tensile stress could be explained in terms of the relaxation effect of constrained elastic energy on the terminal solid solubility of hydrogen at hydride precipitation.