Nuclear powerplant standardization : light water reactors.

"April 1981" (v. 1); "June 1981" (v. 2)

Nuclear Power Reactors

Mode of access: Internet.

Uranium reserve, nuclear fuel cycle delusion, CO2 emission from the sea, and electricity supply: reflections after the fuel meltdown of Fukushima nuclear power units

The Great Tohoku-Kanto earthquake and resulting tsunami has brought considerable attention to the issue of the construction of new power plants. We argue in this paper, nuclear power is not a sustainable solution to energy problems. First, we explore the stock of uranium-235 and the different schemes developed by the nuclear power industry to exploit this resource. Second, we show that these methods, fast breeder and MOX fuel reactors, are not feasible. Third, we show that the argument that nuclear energy can be used to reduce CO2 emissions is false: the emissions from the increased water evaporation from nuclear power generation must be accounted for. In the case of Japan, water from nuclear power plants is drained into the surrounding sea, raising the water temperature which has an adverse affect on the immediate ecosystem, as well as increasing CO2 emissions from increased water evaporation from the sea. Next, a short exercise is used to show that nuclear power is not even needed to meet consumer demand in Japan. Such an exercise should be performed for any country considering the construction of additional nuclear power plants. Lastly, the paper is concluded with a discussion of the implications of our findings.

Modelling of a Hydrogen Catalytic Recombiner for Nuclear Power Plant Containment Studies; Case: Siemens FR90/1-150 Recombiner in TONUS OD Code

Ydinvoimalaitokset on suunniteltu ja rakennettu niin, että niillä on kyky selviytyä erilaisista käyttöhäiriöistä ja onnettomuuksista ilman laitoksen vahingoittumista sekä väestön ja ympäristön vaarantumista. On erittäin epätodennäköistä, että ydinvoimalaitosonnettomuus etenee reaktorisydämen vaurioitumiseen asti, minkä seurauksena sydänmateriaalien hapettuminen voi tuottaa vetyä. Jäädytyspiirin rikkoutumisen myötä vety saattaa kulkeutua ydinvoimalaitoksen suojarakennukseen, jossa se voi muodostaa palavan seoksen ilman hapen kanssa ja palaa tai jopa räjähtää. Vetypalosta aiheutuvat lämpötila- ja painekuormitukset vaarantavat suojarakennuksen eheyden ja suojarakennuksen sisällä olevien turvajärjestelmien toimivuuden, joten tehokas ja luotettava vedynhallintajärjestelmä on tarpeellinen. Passiivisia autokatalyyttisiä vetyrekombinaattoreita käytetäänyhä useammissa Euroopan ydinvoimaitoksissa vedynhallintaan. Nämä rekombinaattorit poistavat vetyä katalyyttisellä reaktiolla vedyn reagoidessa katalyytin pinnalla hapen kanssa muodostaen vesihöyryä. Rekombinaattorit ovat täysin passiivisiaeivätkä tarvitse ulkoista energiaa tai operaattoritoimintaa käynnistyäkseen taitoimiakseen. Rekombinaattoreiden käyttäytymisen tutkimisellatähdätään niiden toimivuuden selvittämiseen kaikissa mahdollisissa onnettomuustilanteissa, niiden suunnittelun optimoimiseen sekä niiden optimaalisen lukumäärän ja sijainnin määrittämiseen suojarakennuksessa. Suojarakennuksen mallintamiseen käytetään joko keskiarvoistavia ohjelmia (Lumped parameter (LP) code), moniulotteisia virtausmalliohjelmia (Computational Fluid Dynamics, CFD) tai näiden yhdistelmiä. Rekombinaattoreiden mallintaminen on toteutettu näissä ohjelmissa joko kokeellisella, teoreettisella tai yleisellä (eng. Global Approach) mallilla. Tämä diplomityö sisältää tulokset TONUS OD-ohjelman sisältämän Siemens FR90/1-150 rekombinaattorin mallin vedynkulutuksen tarkistuslaskuista ja TONUS OD-ohjelmalla suoritettujen laskujen tulokset Siemens rekombinaattoreiden vuorovaikutuksista. TONUS on CEA:n (Commissariat à 1'En¬ergie Atomique) kehittämä LP (OD) ja CFD -vetyanalyysiohjelma, jota käytetään vedyn jakautumisen, palamisenja detonaation mallintamiseen. TONUS:sta käytetään myös vedynpoiston mallintamiseen passiivisilla autokatalyyttisillä rekombinaattoreilla. Vedynkulutukseen vaikuttavat tekijät eroteltiin ja tutkittiin yksi kerrallaan. Rekombinaattoreiden vuorovaikutuksia tutkittaessa samaan tilavuuteen sijoitettiin eri kokoisia ja eri lukumäärä rekombinaattoreita. Siemens rekombinaattorimalli TONUS OD-ohjelmassa laskee vedynkulutuksen kuten oletettiin ja tulokset vahvistavat TONUS OD-ohjelman fysikaalisen laskennan luotettavuuden. Mahdollisia paikallisia jakautumia tutkitussa tilavuudessa ei voitu havaita LP-ohjelmalla, koska se käyttäälaskennassa suureiden tilavuuskeskiarvoja. Paikallisten jakautumien tutkintaan tarvitaan CFD -laskentaohjelma.

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.

Licensing Model Development for Small Modular Reactors (SMRs) - Focusing on the Finnish Regulatory Framework

Recently, Small Modular Reactors (SMRs) have attracted increased public discussion. While large nuclear power plant new build projects are facing challenges, the focus of attention is turning to small modular reactors. One particular project challenge arises in the area of nuclear licensing, which plays a significant role in new build projects affecting their quality as well as costs and schedules. This dissertation - positioned in the field of nuclear engineering but also with a significant section in the field of systems engineering - examines the nuclear licensing processes and their suitability for the characteristics of SMRs. The study investigates the licensing processes in selected countries, as well as other safety critical industry fields. Viewing the licensing processes and their separate licensing steps in terms of SMRs, the study adopts two different analysis theories for review and comparison. The primary data consists of a literature review, semi-structured interviews, and questionnaire responses concerning licensing processes and practices. The result of the study is a recommendation for a new, optimized licensing process for SMRs. The most important SMR-specific feature, in terms of licensing, is the modularity of the design. Here the modularity indicates multi-module SMR designs, which creates new challenges in the licensing process. As this study focuses on Finland, the main features of the new licensing process are adapted to the current Finnish licensing process, aiming to achieve the main benefits with minimal modifications to the current process. The application of the new licensing process is developed using Systems Engineering, Requirements Management, and Project Management practices and tools. Nuclear licensing includes a large amount of data and documentation which needs to be managed in a suitable manner throughout the new build project and then during the whole life cycle of the nuclear power plant. To enable a smooth licensing process and therefore ensure the success of the new build nuclear power plant project, management processes and practices play a significant role. This study contributes to the theoretical understanding of how licensing processes are structured and how they are put into action in practice. The findings clarify the suitability of different licensing processes and their selected licensing steps for SMR licensing. The results combine the most suitable licensing steps into a new licensing process for SMRs. The results are also extended to the concept of licensing management practices and tools.

Development of laboratory exercises for basic nuclear thermal hydraulics measurements

The thesis focuses on light water reactors (pressurized water reactors, boiling water reactors) and measurement techniques for basic thermal hydraulics parameters that are used in a nuclear power plant. The goal of this work is a development of laboratory exercises for basic nuclear thermal hydraulics measurements.

New mechanical samples positioning system for irradiations on a radial channel at nuclear research reactor in a full-power continuous operation

This paper describes a new mechanical samples positioning system that allows the safe placement and removal of biological samples for prolonged irradiation, in a nuclear reactor during full-power continuous operation. Also presented herein the materials of construction and operating principles. Additionally, this sample positioning system is compared with an existing pneumatic and automated transfer system, already available at the research reactors. The system consists of a mechanical arm with a claw, which can deliver the samples for irradiations without reactor shutdown. It was installed in the lEA-R1 research reactor at Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, Brazil, and for the past 5 years, the system has successfully operated and allowed the conducting of important experiments. As a result of its introduction, the facility has been in a position to positively respond to the increased demand in studies of biology, medicine, physics, engineering, detector/dosimeter calibrations, etc. It is one example of the appropriated technologies that save energy and resources. (C) 2010 Elsevier Ltd. All rights reserved.

Safety, Security And safeguards In GEN IV sodium fast reactors

This work presents first a study of the national and international laws in the fields of safety, security and safeguards. The international treaties and the recommendations issued by the IAEA as well as the national regulations in force in France, the United States and Italy are analyzed. As a result of this, a comparison among them is presented. Given the interest of the Japan Atomic Energy Agency for the aspects of criminal penalties and monetary, also the Japanese case is analyzed. The main part of this work was held at the JAEA in the field of proliferation resistance (PR) and physical protection (PP) of a GEN IV sodium fast reactor. For this purpose the design of the system is completed and the PR & PP methodology is applied to obtain data usable by designers for the improvement of the system itself. Due to the presence of sensitive data, not all the details can be disclosed. The reactor site of a hypothetical and commercial sodium-cooled fast neutron nuclear reactor system (SFR) is used as the target NES for the application of the methodology. The methodology is applied to all the PR and PP scenarios: diversion, misuse and breakout; theft and sabotage. The methodology is applied to the SFR to check if this system meets the target of PR and PP as described in the GIF goal; secondly, a comparison between the SFR and a LWR is performed to evaluate if and how it would be possible to improve the PR&PP of the SFR. The comparison is implemented according to the example development target: achieving PR&PP similar or superior to domestic and international ALWR. Three main actions were performed: implement the evaluation methodology; characterize the PR&PP for the nuclear energy system; identify recommendations for system designers through the comparison.

Severe Accident Simulation of Small Modular Reactors

Since the Three Mile Island Unit 2 (TMI-2), accident in 1979 which led to the meltdown of about one half of the reactor core and to limited releases of radioactive materials to the environment, an important international effort has been made on severe accident research. The present work aims to investigate the behaviour of a Small Modular Reactor during severe accident conditions. In order to perform these analyses, a SMR has been studied for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. In the thesis will be described in detail the IRIS Small Modular Reactor; the reference reactor chosen to develop the ASTEC input deck. The IRIS model was developed in the framework of a research collaboration with the IRSN development team. In the thesis will be described systematically the creation of the ASTEC IRIS input deck: the nodalization scheme adopted, the solution used to simulate the passive safety systems and the strong interaction between the reactor vessel and the containment. The ASTEC SMR model will be tested against the RELAP-GOTHIC coupled code model, with respect to a Design Basis Accident, to evaluate the capability of the ASTEC code on reproducing correctly the behaviour of the nuclear system. Once the model has been validated, a severe accident scenario will be simulated and the obtained results along with the nuclear system response will be analysed.

Propagation of nuclear data uncertainties in fuel cycle calculations using MONTE-CARLO Technique

The uncertainty propagation in fuel cycle calculations due to Nuclear Data (ND) is a important important issue for : issue for : • Present fuel cycles (e.g. high burnup fuel programme) • New fuel cycles designs (e.g. fast breeder reactors and ADS) Different error propagation techniques can be used: • Sensitivity analysis • Response Response Surface Method Surface Method • Monte Carlo technique Then, p p , , in this paper, it is assessed the imp y pact of ND uncertainties on the decay heat and radiotoxicity in two applications: • Fission Pulse Decay ( y Heat calculation (FPDH) • Conceptual design of European Facility for Industrial Transmutation (EFIT)

On the Extension of the Analytic Nodal Diffusion Solver ANDES to Sodium Fast Reactors

Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermalhydraulics codes such as COBRA-TF and FLICA, allowing transient analysis of LWR at both fine-mesh and coarse-mesh scales. In order to enable also 3D pin-by-pin and nodal coupled NK-TH simulations of SFR, different developments are in progress. This paper presents the first steps towards the application of COBAYA3 to this type of reactors. ANDES solver, already extended to triangular-Z geometry, has been applied to fast reactor steady-state calculations. The required cross section libraries were generated with ERANOS code for several configurations. The limitations encountered in the application of the Analytic Coarse Mesh Finite Difference (ACMFD) method –implemented inside ANDES– to fast reactors are presented and the sensitivity of the method when using a high number of energy groups is studied. ANDES performance is assessed by comparison with the results provided by ERANOS, using a mini-core model in 33 energy groups. Furthermore, a benchmark from the NEA for a small 3D FBR in hexagonal-Z geometry and 4 energy groups is also employed to verify the behavior of the code with few energy groups.

Application of gas-cooled Accelerator Driven System (ADS) transmutation devices to sustainable nuclear energy development

The conceptual design of a pebble bed gas-cooled transmutation device is shown with the aim to evaluate its potential for its deployment in the context of the sustainable nuclear energy development, which considers high temperature reactors for their operation in cogeneration mode, producing electricity, heat and Hydrogen. As differential characteristics our device operates in subcritical mode, driven by a neutron source activated by an accelerator that adds clear safety advantages and fuel flexibility opening the possibility to reduce the nuclear stockpile producing energy from actual LWR irradiated fuel with an efficiency of 45?46%, either in the form of Hydrogen, electricity, or both.

Análisis Integrado de Seguridad de un accidente de SGTR en un reactor nuclear tipo PWR

El accidente de rotura de tubos de un generador de vapor (Steam Generator Tube Rupture, SGTR) en los reactores de agua a presión es uno de los transitorios más exigentes desde el punto de vista de operación. Los transitorios de SGTR son especiales, ya que podría dar lugar a emisiones radiológicas al exterior sin necesidad de daño en el núcleo previo o sin que falle la contención, ya que los SG pueden constituir una vía directa desde el reactor al medio ambiente en este transitorio. En los análisis de seguridad, el SGTR se analiza desde un punto determinista y probabilista, con distintos enfoques con respecto a las acciones del operador y las consecuencias analizadas. Cuando comenzaron los Análisis Deterministas de Seguridad (DSA), la forma de analizar el SGTR fue sin dar crédito a la acción del operador durante los primeros 30 min del transitorio, lo que suponía que el grupo de operación era capaz de detener la fuga por el tubo roto dentro de ese tiempo. Sin embargo, los diferentes casos reales de accidentes de SGTR sucedidos en los EE.UU. y alrededor del mundo demostraron que los operadores pueden emplear más de 30 minutos para detener la fuga en la vida real. Algunas metodologías fueron desarrolladas en los EEUU y en Europa para abordar esa cuestión. En el Análisis Probabilista de Seguridad (PSA), las acciones del operador se tienen en cuenta para diseñar los cabeceros en el árbol de sucesos. Los tiempos disponibles se utilizan para establecer los criterios de éxito para dichos cabeceros. Sin embargo, en una secuencia dinámica como el SGTR, las acciones de un operador son muy dependientes del tiempo disponible por las acciones humanas anteriores. Además, algunas de las secuencias de SGTR puede conducir a la liberación de actividad radiológica al exterior sin daño previo en el núcleo y que no se tienen en cuenta en el APS, ya que desde el punto de vista de la integridad de núcleo son de éxito. Para ello, para analizar todos estos factores, la forma adecuada de analizar este tipo de secuencias pueden ser a través de una metodología que contemple Árboles de Sucesos Dinámicos (Dynamic Event Trees, DET). En esta Tesis Doctoral se compara el impacto en la evolución temporal y la dosis al exterior de la hipótesis más relevantes encontradas en los Análisis Deterministas a nivel mundial. La comparación se realiza con un modelo PWR Westinghouse de tres lazos (CN Almaraz) con el código termohidráulico TRACE, con hipótesis de estimación óptima, pero con hipótesis deterministas como criterio de fallo único o pérdida de energía eléctrica exterior. Las dosis al exterior se calculan con RADTRAD, ya que es uno de los códigos utilizados normalmente para los cálculos de dosis del SGTR. El comportamiento del reactor y las dosis al exterior son muy diversas, según las diferentes hipótesis en cada metodología. Por otra parte, los resultados están bastante lejos de los límites de regulación, pese a los conservadurismos introducidos. En el siguiente paso de la Tesis Doctoral, se ha realizado un análisis de seguridad integrado del SGTR según la metodología ISA, desarrollada por el Consejo de Seguridad Nuclear español (CSN). Para ello, se ha realizado un análisis termo-hidráulico con un modelo de PWR Westinghouse de 3 lazos con el código MAAP. La metodología ISA permite la obtención del árbol de eventos dinámico del SGTR, teniendo en cuenta las incertidumbres en los tiempos de actuación del operador. Las simulaciones se realizaron con SCAIS (sistema de simulación de códigos para la evaluación de la seguridad integrada), que incluye un acoplamiento dinámico con MAAP. Las dosis al exterior se calcularon también con RADTRAD. En los resultados, se han tenido en cuenta, por primera vez en la literatura, las consecuencias de las secuencias en términos no sólo de daños en el núcleo sino de dosis al exterior. Esta tesis doctoral demuestra la necesidad de analizar todas las consecuencias que contribuyen al riesgo en un accidente como el SGTR. Para ello se ha hecho uso de una metodología integrada como ISA-CSN. Con este enfoque, la visión del DSA del SGTR (consecuencias radiológicas) se une con la visión del PSA del SGTR (consecuencias de daño al núcleo) para evaluar el riesgo total del accidente. Abstract Steam Generator Tube Rupture accidents in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are special transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path to the environment. The SGTR is analyzed from a Deterministic and Probabilistic point of view in the Safety Analysis, although the assumptions of the different approaches regarding the operator actions are quite different. In the beginning of Deterministic Safety Analysis, the way of analyzing the SGTR was not crediting the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that operators can took more than 30 min to stop the leakage in actual sequences. Some methodologies were raised in the USA and in Europe to cover that issue. In the Probabilistic Safety Analysis, the operator actions are taken into account to set the headers in the event tree. The available times are used to establish the success criteria for the headers. However, in such a dynamic sequence as SGTR, the operator actions are very dependent on the time available left by the other human actions. Moreover, some of the SGTR sequences can lead to offsite doses without previous core damage and they are not taken into account in PSA as from the point of view of core integrity are successful. Therefore, to analyze all this factors, the appropriate way of analyzing that kind of sequences could be through a Dynamic Event Tree methodology. This Thesis compares the impact on transient evolution and the offsite dose of the most relevant hypothesis of the different SGTR analysis included in the Deterministic Safety Analysis. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP), with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The offsite doses are calculated with RADTRAD code, as it is one of the codes normally used for SGTR offsite dose calculations. The behaviour of the reactor and the offsite doses are quite diverse depending on the different assumptions made in each methodology. On the other hand, although the high conservatism, such as the single failure criteria, the results are quite far from the regulatory limits. In the next stage of the Thesis, the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermohydraulical analysis of a Westinghouse 3-loop PWR plant with the MAAP code. The ISA methodology allows obtaining the SGTR Dynamic Event Tree taking into account the uncertainties on the operator actuation times. Simulations are performed with SCAIS (Simulation Code system for Integrated Safety Assessment), which includes a dynamic coupling with MAAP thermal hydraulic code. The offsite doses are calculated also with RADTRAD. The results shows the consequences of the sequences in terms not only of core damage but of offsite doses. This Thesis shows the need of analyzing all the consequences in an accident such as SGTR. For that, an it has been used an integral methodology like ISA-CSN. With this approach, the DSA vision of the SGTR (radiological consequences) is joined with the PSA vision of the SGTR (core damage consequences) to measure the total risk of the accident.