A three region steam drum model for a nuclear power plant simulator (BRENDA) /

"Date published: August 1981."

Nuclear regulation : efforts to ensure nuclear power plant safety can be strengthened : a report to the Honorable Alfonse M. D'Amato, U.S. Senate /

"B-226879."

Can the Paks-2 Nuclear Power Plant operate without State aid? A business economic analysis

Many studies have investigated the prospects for returns, the budgetary and financing background and energy management effects of the new nuclear power plant units to be built at Paks. This document seeks to complement previous economics-based studies by adding a new criterion. The key question in our analysis is whether the power plant company will be capable of independent operations in an economic sense - or will its survival depend on further additional aid by the owner, i.e. via the central budget, after its commissioning? We shall examine from a corporate perspective in what ways the already disclosed financing terms and conditions may affect the everyday operations of the power plant company.

Life cycle management of TVO nuclear power plant

Tämä diplomityö tutkii eri elinkaarihallinnan menetelmiä ja vertaa niitä TVO:n menetelmiin. Lisäksi TVO:n prosessin ongelmakohdat tunnistetaan ja niihin esitetään ratkaisuja. Vertailukohteina toimii ydinvoimateollisuuden lisäksi vesivoima, fossiiliset voimalaitokset sekä paperiteollisuus. Sähkön hinnan jatkaessa laskuaan on elinkaariajattelusta tullut ajankohtaista myös ydinvoimayhtiöille. Ydinvoimalaitoksien pitkän suunnitellun käyttöiän ansiosta laitoksen elinkaaren aikana voi tapahtua useita asioita, jotka vaikuttavat laitoksen investointitarpeisiin. Turvallisen sähköntuotannon varmistamiseksi eri laitososia on joko muokattava tai uusittava. Elinkaariajatteluun kuuluu tehokas laitoksen kunnon valvonta, laitoksen ikääntymiseen vaikuttavien ilmiöiden tunnistaminen, sekä ikääntymistä hillitsevien toimenpiteiden pitkän tähtäimen suunnittelu. Hyvällä ennakkosuunnittelulla pyritään varmistamaan se, että laitoksella voidaan tuottaa sähköä koko sen jäljellä olevan käyttöiän aikana. Kun tarpeiden tunnistus ja suunnittelu tehdään hyvissä ajoin mahdollistetaan myös investointien optimointi. Paras hyöty pyritään saamaan ajoittamalla oikeat investoinnit oikeaan aikaan.

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.

Effect of Noncondensable Gases on Circulation of Primary Coolant in Nuclear Power Plants in Abnormal Situations

The present study focuses on two effects of the presence of a noncondensable gas on the thermal-hydraulic behavior of thecoolant of the primary circuit of a nuclear reactor in the VVER-440 geometry inabnormal situations. First, steam condensation with the presence of air was studied in the horizontal tubes of the steam generator (SG) of the PACTEL test facility. The French thermal-hydraulic CATHARE code was used to study the heat transfer between the primary and secondary side in conditions derived from preliminary experiments performed by VTT using PACTEL. In natural circulation and single-phase vapor conditions, the injection of a volume of air, equivalent to the totalvolume of the primary side of the SG at the entrance of the hot collector, did not stop the heat transfer from the primary to the secondary side. The calculated results indicate that air is located in the second half-length (from the mid-length of the tubes to the cold collector) in all the tubes of the steam generator The hot collector remained full of steam during the transient. Secondly, the potential release of the nitrogen gas dissolved in the water of the accumulators of the emergency core coolant system of the Loviisa nuclear power plant (NPP) was investigated. The author implemented a model of the dissolution and release ofnitrogen gas in the CATHARE code; the model created by the CATHARE developers. In collaboration with VTT, an analytical experiment was performed with some components of PACTEL to determine, in particular, the value of the release time constant of the nitrogen gas in the depressurization conditions representative of the small and intermediate break transients postulated for the Loviisa NPP. Such transients, with simplified operating procedures, were calculated using the modified CATHARE code for various values of the release time constant used in the dissolution and release model. For the small breaks, nitrogen gas is trapped in thecollectors of the SGs in rather large proportions. There, the levels oscillate until the actuation of the low-pressure injection pumps (LPIS) that refill the primary circuit. In the case of the intermediate breaks, most of the nitrogen gas is expelled at the break and almost no nitrogen gas is trapped in the SGs. In comparison with the cases calculated without taking into account the release of nitrogen gas, the start of the LPIS is delayed by between 1 and 1.75 h. Applicability of the obtained results to the real safety conditions must take into accountthe real operating procedures used in the nuclear power plant.

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.

Failure modes of passive decay heat removing safety systems of modern nuclear power plants

The purpose of this master’s thesis is to gain an understanding of passive safety systems’ role in modern nuclear reactors projects and to research the failure modes of passive decay heat removal safety systems which use phenomenon of natural circulation. Another purpose is to identify the main physical principles and phenomena which are used to establish passive safety tools in nuclear power plants. The work describes passive decay heat removal systems used in AES-2006 project and focuses on the behavior of SPOT PG system. The descriptions of the main large-scale research facilities of the passive safety systems of the AES-2006 power plant are also included. The work contains the calculations of the SPOT PG system, which was modeled with thermal-hydraulic system code TRACE. The dimensions of the calculation model are set according to the dimensions of the real SPOT PG system. In these calculations three parameters are investigated as a function of decay heat power: the pressure of the system, the natural circulation mass flow rate around the closed loop, and the level of liquid in the downcomer. The purpose of the calculations is to test the ability of the SPOT PG system to remove the decay heat from the primary side of the nuclear reactor in case of failure of one, two, or three loops out of four. The calculations show that three loops of the SPOT PG system have adequate capacity to provide the necessary level of safety. In conclusion, the work supports the view that passive systems could be widely spread in modern nuclear projects.

Space charge and dielectric response measurements to assess insulation aging of low-voltage cables used in nuclear power plants

The current design life of nuclear power plant (NPP) could potentially be extended to 80 years. During this extended plant life, all safety and operationally relevant Instrumentation & Control (I&C) systems are required to meet their designed performance requirements to ensure safe and reliable operation of the NPP, both during normal operation and subsequent to design base events. This in turn requires an adequate and documented qualification and aging management program. It is known that electrical insulation of I&C cables used in safety related circuits can degrade during their life, due to the aging effect of environmental stresses, such as temperature, radiation, vibration, etc., particularly if located in the containment area of the NPP. Thus several condition monitoring techniques are required to assess the state of the insulation. Such techniques can be used to establish a residual lifetime, based on the relationship between condition indicators and ageing stresses, hence, to support a preventive and effective maintenance program. The object of this thesis is to investigate potential electrical aging indicators (diagnostic markers) testing various I&C cable insulations subjected to an accelerated multi-stress (thermal and radiation) aging.

Boiling Nuclear Superheater (BONUS) Power Station : supplementary study: extrapolation to large central station intergral nuclear superheat plant /

"January 28, 1960."

Measurements of environmental background radiation in the surrounding area of a coal-fired power plant

Certain materials used and produced in a wide range of non-nuclear industries contain enhanced activity concentrations of natural radionuclides. In particular, electricity production from coal is one of the major sources of increased exposure to man from enhanced naturally occurring materials. Over the past decades there has been some discussion about the elevated natural background radiation in the area near coal-fired power plants due to high uranium and thorium content present in coal. This work describes the methodology developed to assess the radiological impact due to natural radiation background increasing levels, potentially originated by a coal-fired power plant’s operation. Gamma radiation measurements have been done with two different instruments: a scintillometer (SPP2 NF, Saphymo) and a gamma ray spectrometer with energy discrimination (Falcon 5000, Canberra). A total of 40 relevant sampling points were established at locations within 20 km from the power plant: 15 urban and 25 suburban measured stations. The highest values were measured at the sampling points near to the power plant and those located in the area within the 6 and 20 km from the stacks. This may be explained by the presence of a huge coal pile (1.3 million tons) located near the stacks contributing to the dispersion of unburned coal and, on the other hand, the height of the stacks (225 m) which may influence ash’s dispersion up to a distance of 20 km. In situ gamma radiation measurements with energy discrimination identified natural emitting nuclides as well as their decay products (212Pb, 214Pb, 226Ra 232Th, 228Ac, 234Th 234Pa, 235U, etc.). This work has been primarily done to in order to assess the impact of a coal-fired power plant operation on the background radiation level in the surrounding area. According to the results, an increase or at least an influence has been identified both qualitatively and quantitatively.

Does Brazil need new nuclear power plants?

In October 2008, the Brazilian Government announced plans to invest US$212 billion in the construction of nuclear power plants, totaling a joint capacity of 60,000 MW. Apart from this program, officials had already announced the completion of the construction of the nuclear plant Angra III; the construction of large-scale hydroelectric plans in the Amazon and the implantation of natural gas, biomass and coal thermoelectric plants in other regions throughout the country. Each of these projects has its proponents and its opponents, who bring forth concerns and create heated debates in the specialized forums. In this article, some of these concerns are explained, especially under the perspective of the comparative analysis of costs involved. Under such merit figures, the nuclear option, when compared to hydro plants, combined with conventional thermal and biomass-fueled plants, and even wind, to expand Brazilian power-generation capacity, does not appear as a priority. (C) 2009 Elsevier Ltd. All rights reserved.

Education and Training of Future Nuclear Engineers Through the use of an Interactive Plant Simulator

The successful experience of the Jose Cabrera Nuclear Power Plant Interactive Graphical Simulator implementation in the Nuclear Engineering Department in the Universidad Polite´cnica de Madrid, for the Education and Training of nuclear engineers is shown in this paper. The paper starts with the objectives and the description of the Simulator Aula, and the methodology of work following the recommendations of the IAEA for the use of nuclear reactor simulators for education. The practices and material prepared for the students, as well as the operational and accident situations simulated are provided.

New production reactor power plant economic feasibility study : a report to Atomic Energy Commission /

At head of title: Volume I.