logo NSS NENE2014 http://www.ijs.si/
http://www.icjt.org/

Preliminary program

This is a HTML version of the abstracts for information only. It can differ from the original,
submitted by the author(s). Special characters are not properly reproduced here.
Please, contact the author(s) or refer to the Book of Abstracts for the correct version.
The content of abstracts is the responsibility of the authors concerned.
The organizer is not responsible for published facts and technical accuracy of presented data.


08.09.2014 14:20

Invited lectures - 102

 

Finnish approach for synergistic effects of research, reliability and safety in nuclear energy

Rauno Rintamaa

VTT - Smart Industry and Energy Systems, Vuorimiehentie 3, Espoo, P.O.Box 1000, FIN-02044 VTT, Finland

rauno.rintamaa@vtt.fi

Energy production and consumption are among the basic functions of society. For industry, energy is an important production factor, especially with respect to many Finnish export products. Since energy consumption per capita in Finland is high and the proportion of indigenous energy sources is low, energy supply, its price and its efficient use are more significant than in most other countries. Industry accounts half of all final energy consumed in Finland. The Finnish manufacturing industry’s heavy consumption of energy is due to its production structure. The forest industry alone accounted for about a half and the metal and chemical industries together for about one third of all industrial energy consumption.

Finland is highly dependent on imported energy, which in 2013 accounted for two thirds of all primary energy consumption in Finland. Although its share has long been in decline, oil has maintained its position as the most important fuel. Therefore, nuclear energy has played a major role in Finnish electricity production since the beginning of the 1980s. In 2013, the proportion of nuclear electricity totalled 25 per cent of total electricity consumption and almost 30 per cent of domestic production.

Finland can be proud of the high load factors of nuclear power plants, low price of  nuclear electricity and  low levels of radioactive emissions. Largely owing to nuclear electricity, Finland can also take pride in low level of carbon dioxide emissions in total electricity generation.

Almost without exception, the average annual load factor has been over 90 per cent since 1983. The operation of the Finnish reactors has been safe and undisturbed. Furthermore, their commercial profitability has been boosted further by extensive modernisations, including the considerable uprating carried out on all units since the late 1990s.

Excellent performance is mainly due to research oriented and open minded attitude prevailing among all nuclear stakeholders, i.e. research, academy, power companies, service industry, regulation and authorities. The basis since early 1980 has been joint R&D programmes where all key stakeholders have been involved.  The focus of nuclear R&D is on the safety and operational performance of the power plants, and the management and disposal of waste. Publicly funded nuclear energy research, on the other hand, provides impartial expertise in nuclear energy issues, contributes to maintaining the necessary personnel and equipment for research and development, and has established a framework for international collaboration. In addition, one of the key objectives of the national research programmes is to train new nuclear experts to meet the requirements for additional human resources. Changes in energy markets and the rapid development of technology will set new challenges with respect to the required knowledge, and this will require a special focus from all parties.

The versatile array of research subjects at research institutes and universities promotes spin-off and spin-in relationships with other industries, either. Spin-off s include simulation technologies, reliability engineering, fracture mechanics, and non-destructive testing, while spin-in benefits have been enjoyed in areas such as human factors, digital automation systems and computational fluid dynamics.

The Finnish public nuclear energy research is organised into national research programmes. The current national research programmes on nuclear fission energy are as follows:
National Nuclear Power Plant Safety Research (safir2014), 2011–2014
Finnish Research Programme on Nuclear Waste Management (kyt2014), 2011–2014

SAFIR2014 is the Finnish public research programme on nuclear power plant safety coordinated by VTT Technical Research Centre of Finland. The programme has been divided into nine research areas: 1. Man, Organisation and Society, 2. Automation and Control Room, 3. Fuel Research and Reactor Analysis,
4. Thermal Hydraulics, 5. Severe Accidents, 6. Structural Safety of Reactor Circuits, 7. Construction Safety,
8. Probabilistic Risk Analysis (PRA) and 9. Development of Research Infrastructure.

KYT2014 is the Finnish public research programme on nuclear waste management coordinated by VTT Technical Research Centre of Finland. The contents of the KYT2014 Research Programme comprise key research subjects in terms of national expertise. These include new and alternative nuclear waste management technologies, research into the safety of nuclear waste management and sociological research related to the issue.

In this presentation specific features of the Finnish energy system and the importance and role of nuclear energy in the system is shown as a background. The governance and operational model of the synergistic effects in the Finnish approach is outlined in more details. The effects are outlined from different stakeholders’ point of view like research, education and training, technology development and innovation, legislation and regulation, The main focus will be how research activities can and have been adopted, implemented and disseminated on improving safety, reliability and operability of NPPs. Highlights of some results and case studies will be given.






08.09.2014 15:00

Nuclear energy and society - 201

 

Nuclear Industry and European BSS

Helena Janžekovič

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

helena.janzekovic@gov.si

 

In the next decades nuclear industry in Europe is expecting very vivid changes. While in some countries a new built of NPPs is announced, in others the intention to phase out the use of nuclear energy is foreseen. Activities related to all kind of ionising sources in the European Union (EU) are regulated in the Member States (MS) implementing also so-called European legislation, i.e. legal acts related to nuclear and radiation safety which must be implemented by the MS. The main aim of these is to harmonise safety on the European level. The requirements of such acts can fundamentally influence the activity of all involved, e.g. operators, radioactive waste management agencies, research institutions as well as regulatory bodies. Such influence can last next few decades.
After nearly 20 years the new basic safety standards for protection against the danger arising from ionizing radiation, i.e. 2013/59/Euratom Directive (BSS Directive) were published in January 2014. The BSS Directive tackles nuclear industry as well as other practices or situations where people or the environment could be affected by ionising radiation. The directive forms a harmonised system related to nuclear and radiation safety in the EU together with two other directives, namely:

• Nuclear Safety Directive, 2009, and
• Spent Fuel and Radioactive Waste Directive, 2011.

The BSS Directive is a comprehensive document repealing altogether five other legal documents. It is a very technical document using more than 1000 physical parameters. It introduces a set of new concepts as well as new or updated physical parameters, such as a new dose limit for a lens of the eye and so-called clearance levels used for solid material released from the regulatory control. Among important concept is requirement to use current state of technical knowledge in the justification process. In addition, a control over building materials as well as natural sources is strengthened. The analysis of more than hundred requirements from the BSS Directive is given taking into account all life phases of an NPP. The analysis shows for example that the licensing of a facility is strengthened, control over discharges is introduced and dilution of radioactive materials is discussed enabling re-evaluation of decommissioning plans of nuclear facilities. The overall structure of the BSS Directive is based on the ICRP 103 (2007) document introducing planned, existing and emergency exposure situation. As a part of post-Fukushima activities emergency preparedness in a case of nuclear or radiological accident is strengthened in the BSS Directive. The MSs must implement the requirements of the BSS Directive in their legislation till 6th February 2018. Challenges of the MS and particularly of nuclear industry when implementing the requirements are discussed in details.






08.09.2014 15:20

Nuclear energy and society - 208

 

Public Opinion about Nuclear Energy – Year 2014 Poll

Radko Istenič, Igor Jenčič

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

radko.istenic@ijs.si

 

Public information is an ongoing activity of the Nuclear Training Centre at the Jožef Stefan Institute. The Information Centre was established within the Nuclear Training Centre more than 20 years ago to inform the visitors about nuclear power and nuclear technology in general and about Krško Nuclear Power Plant.
The main target group of information activity are schoolchildren with their teachers. Most of them are from the 8th and 9th grade of elementary school, age 14 to 15. Every year some 8000 youngsters visit the Information Centre. The visit consists of a live lecture about nuclear technology followed by the demonstration of radioactivity and a guided tour of a permanent exhibition. Since 1993 we monitor the opinion trends by polling about 1000 youngsters every year. The youngsters are polled before they listen to the lecture or visit the exhibition in order to obtain their opinion based on the knowledge from everyday life. In the paper we will present, summarize and comment the trends over the last 21 years.






08.09.2014 15:40

Nuclear energy and society - 203

 

New Training Experiments at the JSI TRIGA Mark II Reactor

Luka Snoj, Sebastjan Rupnik, Anže Jazbec

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

luka.snoj@ijs.si

 

Since the 1990s the Jožef Stefan Institute (JSI) TRIGA reactor has been extensively used for performing practical reactor physics exercises for future nuclear power plant operators, students of physics and nuclear engineering and for participants of various international training courses.
In 2012 we upgraded some of the existing and introduced some new exercises. The pulse mode operation exercise was upgraded by installation of new data acquisition system and development of new graphical user interface (GUI) by using LabVIEW software. The critical experiment exercise was upgraded by adding a new detector. Now we monitor neutron population with two independent fission chambers on different locations. Here as well new graphical user interface (GUI), by using LabVIEW software, was developed. In the past the void reactivity coefficient exercise was performed by inserting Al tube into various positions in the reactor core and measuring the corresponding reactivity changes. In order to make the exercise more realistic, we installed a pneumatic system for generating air bubbles just below the core. The system consists of a system of valves, flow meters and Al tubes for conveying air under the core. The system is operated remotely by a computer running application in LabVIEW. The trainee can adjust the air pressure (proportional to the flow rate) and the location in the core at which the air bubbles are generated. The flow rate at individual locations is measured. The aim of the exercise is to measure reactivity changes versus flow rate and air bubble position. The second new exercise was measurement of water activation. In this exercise we installed special system which pumps the water through the core at a constant flow rate to the reactor platform, where the water activity is measured with a portable GM tube and two spectrometers, a semiconducting HPGe and a scintillating LaBr. The purpose of the exercise is to measure the 16N and 19O gamma line intensity and dose rate versus reactor power. It can be seen that the relationship is linear. Similar system is used at the Krško NPP for primary coolant loop leakage detection. The third new exercise, named in core flux mapping, was performed by measuring the axial fission rate distribution at various radial positions in the core. We used CEA – developed mini fission chambers and a special home developed system for moving the fission chamber in axial direction and measuring the count rate versus FC position. The moving system and the FC response were operated by LabVIEW software running on a portable computer. In the paper we present the new exercises in more details, first results and plan for the future.






08.09.2014 16:20

Severe accidents - 301

 

Iodine Benchmarks in the SARNET2 Network of Excellence

Tim Haste1, Mirco Di Giuli2, Gunter Weber3, Sebastian Weber3

Institut de Radioprotection et de Sureté Nucléaire, Bât. 702 Centre de Cadarache, BP 3-13115 Saint Paul lez Durance, France1

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy2

Gesellschaft für Reaktorsicherheit (GRS), Schwertnergasse 1, 50667 Köln, Germany3

tim.haste@irsn.fr

 

Accurate calculation of the iodine behaviour in the containment is of crucial importance in determining the potential radioactive source term to the environment under light water reactor severe accident conditions. Of particular importance is the behaviour of iodine in the gas phase, particularly organic iodine which is difficult to remove by filtration e.g. in containment venting systems. The iodine behaviour is closely linked with the containment thermal hydraulics that have a major influence on the distribution of iodine throughout the containment atmosphere and sump.
In the EC SARNET2 network of excellence, the predictive ability of current severe accident codes in this important area was assessed through two benchmark exercises. In the first, the codes were assessed against data from the German THAI Iod-11 and Iod-12 integral tests, in which particular attention was paid to molecular iodine transport with atmospheric flows, and the iodine interactions with steel surfaces. In the benchmark of the international Phébus FPT3 integral test all aspects of a severe accident were studied, from core degradation, fission product release, circuit transport and deposition, and containment behaviour. Thermal hydraulic conditions in the containment were simpler, while realistic fission product sources were used and radiolytic interactions of iodine, e.g. with painted surfaces, were studied, an important source of organic iodine in the containment atmosphere. The two benchmarks are thus complementary.In the FPT3 exercise the calculations could predict fairly well the general, well-mixed, thermal hydraulic conditions. For THAI, where there are more detailed measurements, significant differences were noted for atmospheric flows and relative humidities. The iodine results for both cases show also a wide spread in calculated results, again well outside data uncertainties (typically ~20% (1σ)), that indicates the need for model improvements in this area, e.g. for iodine absorption/desorption onto steel in THAI, and for interaction of iodine with paint under irradiation for FPT3. It was necessary in this Phébus case to use an iodine source to the containment based on experimental data, as the calculated source from the circuit was not sufficiently accurate, owing e.g. to lack of models for kinetic effects. Model improvements are under way, based on separate-effects data from independent programs such as OECD/THAI2, BIP2, STEM; EC/PASSAM; and the French national program MIRE, and repeat benchmarks will be proposed to check on progress towards convergence, e.g. under the EC/NUGENIA organisation. Substantial user effects were noted in both exercises, indicating the need for improved user training in phenomenology and optimum code use.






08.09.2014 16:40

Severe accidents - 302

 

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

Alexander D. Vasiliev1, Juri Stuckert2

Nuclear Safety Institute of Russian Academy of Sciences , 52, B. Tulskaya, 115191 Moscow, Russian Federation1

Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany2

vasil@ibrae.ac.ru

 

The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the calculation of QUENCH-LOCA-1 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (Loss of Coolant Accident) nuclear power plant accident sequence in which the overheated up to 1300K reactor core would be reflooded from the bottom by ECCS (Emergency Core Cooling System).The test QUENCH-LOCA-1 was successfully conducted at the KIT, Karlsruhe, Germany, in February 2, 2012, and was the second test for this series. The test bundle was made up of 21 fuel rod simulators which are placed in the square set. Heating was carried out electrically using tantalum heaters. The rod claddings were identical to that used in PWRs. The bundle was electrically heated in steam from 900K to 1300K with the heat-up rate of 5.7K/s. After cooling in the saturated steam the bottom flooding with water flow rate of 100 g/s was initiated. The SOCRAT/V3 calculated results describing thermal hydraulic, hydrogen generation end thermo-mechanical behavior including rods ballooning and burst are in a good agreement with experimental data.






08.09.2014 17:00

Severe accidents - 303

 

Influence of Zirconium Oxidation on Steam Explosion Energetics

Matjaž Leskovar, Vasilij Centrih

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

matjaz.leskovar@ijs.si

 

A steam explosion, in the frame of nuclear reactor safety, is an energetic fuel coolant interaction process, which may occur when the hot reactor core melt comes into contact with the coolant water. Steam explosions are an important nuclear safety issue because they can potentially jeopardize the primary system and the containment integrity of the nuclear power plant. If metallic zirconium is present in the corium melt the oxidation of zirconium may significantly influence the fuel coolant interaction process, as observed in experiments. In the ZREX (ANL, USA) experiments the steam explosion strength was largely increased by the oxidation of the metallic zirconium, whereas in the recently performed OECD SERENA project KROTOS and TROI experiments it seems that the oxidation had an inhibiting effect. To find out the reasons for this qualitatively different behaviour, the experimental results were investigated in comparison and computer simulations were performed.
In the paper, the performed analysis of the influence of the zirconium oxidation on the steam explosion energetics will be presented and discussed. Based on the experimental findings, the hydrogen film hypothesis is proposed, claiming that only a limiting amount of zirconium may be oxidized during the premixing phase in subcooled conditions and that the remaining amount of unoxidized zirconium is available for the oxidation in the explosion phase. Various computer simulations were performed with the MC3D code to support the hypothesis and to get additional insight. It may be concluded that the proposed hypothesis reasonably well explains the observed experimental differences. Suggestions for further analytical and experimental work will also be given.






08.09.2014 17:20

Severe accidents - 306

 

Spatial Distribution of Hydrogen in NEK Containment

Davor Grgić, Tomislav Fancev, Siniša Šadek, Vesna Benčik

University of Zagreb, Faculty of Electrical Engineering and Computing, Unska 3, 10000 Zagreb, Croatia

davor.grgic@fer.hr

 

There is design requirement to limit hydrogen volumetric concentration to 4%. In Safety Analysis Report (SAR) Chapter 6 it is demonstrated that global hydrogen concentration stays below limit using one compartment containment model. It is assumed that due to well connected compartments and large volumes possibility for appearance of high hydrogen local concentrations is very low. It is still possible to see some temporary local increase in hydrogen concentration due to non uniform hydrogen production and hydrogen stratification, especially for Beyond Design Basis Accident (BDBA) conditions. After recent modification NPP Krsko is using Passive Autocatalytic Recombiners (PARs) to control concentration of combustible gases. Two of them are declared as safety related (required according to plant’s design basis) and 20 are there to provide additional protections during BDBAs. In order to calculate spatial hydrogen distribution, and predict peak local concentrations, detailed multi-compartment GOTHIC model of NPP Krsko containment was developed. Real PAR’s locations are taken into account. Two scoping calculations were performed, one for hydrogen sources in containment during and after design basis LOCA accident (long term behaviour), and another (short term) for MAAP calculated hydrogen source after BDBA SBO sequence. In both situations PARs are able to prevent situations where combustible gases can reach flammable concentrations.






08.09.2014 17:40

Severe accidents - 305

 

Analysis of CABRI-BI1 loss-of-flow experiment using ASTEC-Na

Alain Flore Y Flores1, Vaidas Matuzas2, Luca Ammirabile3

Joint Research Centre of the European Commission, Westerduinweg 3, 1755 ZG Petten, Netherlands1

Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, 3 Breslaujos, LT-44403 Kaunas, Lithuania2

European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands3

alain.flores-y-flores@ec.europa.eu

 

The FP7 (Seventh Framework Program) JASMIN (Joint Advanced Severe Accident Modelling and Integration for Sodium-Cooled Fast Neutron Reactors) project aims at developing a robust and modern computer code system, named ASTEC-Na, which will be capable of encompassing all the phases of a hypothetical severe accident in Sodium-Cooled-Fast Reactors (SFR). The Institute for Energy and Transport (IET) in collaboration with the project partners is conducting the validation of the thermo-mechanical fuel-clad and pin behaviour after rupture models implemented in ASTEC Na (Accident Source Term Evaluation Code). A series of experiments performed in the past (CABRI/SCARABEE experiments) were chosen to validate newly developed ASTEC-Na.
One of the in-pile experiments considered during ASTEC-Na validation phase is CABRI-BI1 loss-of-flow transient without transient over power using low burn-up MOX fuel pin. The experiment resulted in channel voiding as a result of sodium boiling and a clad melting. Only some fuel melting took place. The ASTEC-Na calculation was in good agreement with the overall fuel-pin disruption behaviour, which was characterised by a thermal pin-failure mode, observed in the CABRI-BI1 experiment. This paper provides outcome of the analysis using ASTEC-Na as well as findings for future code improvements. The main attention during validation phase was given to modelling of fuel-clad heat transfer, pin mechanical behaviour (creep deformation of the clad, account for direct fuel-clad contact, fuel swelling, and release of fission gases within the fuel), movement of molten materials. The paper will identify main issues and limitations linked to the modelling capabilities of ASTEC-Na.






09.09.2014 08:30

Invited lectures - 101

 

Nuclear energy, an energy for the future

Christophe Behar

CEA-France, Direction de l'Energie Nucléaire, Bât.121, 91191 Gif sur Yvette Cedex, France

brigitte.naulleau@cea.fr

 

France nuclear fuel cycle strategy is driven by the conviction that nuclear energy is to be maintained as a stable pillar of the French energy mix in the long term. Fukushima accident has triggered deep and comprehensive safety and economic assessments which have reinforced the fundamentals of this strategy. Thus sustainability is a major driver of France nuclear strategy, among which fuel cycle policy plays a major role. It is based on the recycling of spent fuel in order to save energy resources and to offer a better management of nuclear wastes. This relies on decades’ feedback experience of the La Hague reprocessing plant, the Melox fuel fabrication facility, and the Mox fuel use in the French nuclear fleet. On the other hand France has also a large feedback experience at industrial size with sodium cooled fast reactor, indispensable tools to burn efficiently the plutonium coming out from light water reactor Mox spent fuel or sodium fast reactor Mox spent fuel. Owing to this experience in La Hague, Phenix and Superphenix, France is developing Generation IV fast reactor along with their related fuel cycle, these systems being the keystone of sustainable energy nuclear development.
In the global context of the increasing demand for energy, while fossil fuel resources gradually reach exhaustion, energy management, as a vital need and a factor of economic growth, is a major challenge for the world of tomorrow, facing with the desire to reduce greenhouse gas emissions. In 2030 the worldwide demand for energy will have doubled under the combined effects of population growth — to 9 or 10 billion by 2050 — and the development of emerging countries. At a time when resources are becoming scarce and it is increasingly urgent to combat climate change, it has become indispensable to ensure competitive low-carbon energy sources.
Nuclear power alongside renewable energy sources will play a fundamental role in this “basket of energy resources” for the future. The advantages of nuclear energy — electricity production without greenhouse gas emissions — make it a promising solution. Although nuclear energy has significant advantages in this regard, it faces safety, resource sustainability, and waste management issues that must be met through continuing technological innovation, to improve its capability to continue satisfying the long-term demand for energy, which is increasing endlessly. Including nuclear power in any sustainable development plan, presupposes the development of a new generation of reactors capable of making the best of these resources; exploiting all of the uranium ore, multi-recycling the plutonium, ensuring that all the operations meet the most stringent safety criteria: these are the challenges for fast neutron reactors included in reactors of “Generation IV”.
With view for France to preparing the sustainable nuclear energy of the future with the development of innovative technologies, in accordance with the French Act of June 28, 2006, the Nuclear Energy Division (DEN) of CEA is responsible for designing the technological reactor ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), based on a strong feedback experience, owing to Phenix on the one hand, and La Hague and Melox on the other hand.
FNRs appear to show the best potential to reach 4th generation criteria for industrial deployment in the middle of the 21st century, or even earlier if needed, based on the accumulated operating experience of more than 400 reactor.years and combining thus essential proven features; offering the highest degree of industrial maturity based on extensive feedback, and clearly identified avenues for progress, they have been adopted today by all countries engaged in the development of fast neutron reactors (Russia, China, India, Japan, South Korea, etc.).
The Astrid reactor, a 600 MWe prototype reactor, is the indispensable step before an industrial deployment. It is representative of the main necessary industrial characteristics and its demonstration capabilities are designed for the qualification of innovative concepts. The ongoing R&D will lead to a selection of particularly advanced options, especially in terms of safety and operability.






09.09.2014 09:10

Multiphysics - 401

 

Analysis of Thermalhydraulic Fluctuations in Trillo NPP with CTF/PARCSv2.7 Coupled Code

Agustin Abarca1, Rafael Miró2, Gumersindo Verdú1

Polytechnic University of Valencia, Department of Chemical and Nuclear Engineering, Camí de Vera sn, 46022 Valencia, Spain1

Universitat Politecnica de Catalunya, C. Jordi Girona, 31, 08034 Barcelona, Spain2

gverdu@iqn.upv.es

 

The low frequency noise is a phenomenon observed in some PWR reactors causing an influence on the normal operational behavior of the power plant. Fluctuations in the neutron flux density, in the low frequency range up to 4Hz, generate noise in the neutron instrumentation (incore and excore neutron detectors) that could affect to the limitation and protection system of the reactor core, even it can activate a scram signal due to high neutron flux.
This activity is included as a complementary study of the joint efforts of CNAT and ISIRYM in the study of the neutronic noise in the PWR reactor cores and their associated instrumentation. In this project has been carried out a deep analysis of the behaviour of a PWR reactor core and the signal of the neutronic instrumentation when a set of inlet thermalhydraulic perturbations are applied.
A set of inlet mass flow perturbations have been simulated with the CTF/PARCSv2.7 coupled code, which provide 3D results in both termalhydraulics and neutronics. The cross sections have been obtained using the SIMTAB methodology and the signal of incore and excore neutron detectors are directly extracted from PARCS through a previously validated modifications incorporated in the code.
The power and outlet temperature fluctuations obtained with CTF/PARCSv2.7 have been analyzed, as well as the signals of the neutronic instrumentation. This results have also been compared with the ones obtained with RELAP5/PARCSv2.7 and SIMULATE-3K for providing an overview of the global results supporting the phenomenology observed in the real PWR reactor cores.






09.09.2014 09:30

Multiphysics - 402

 

Neutronics and Thermal-hydraulics Study on Maximizing In Situ Utilization of U-233 in the Th-U5-fueled HTR-PM

Bing Xia, Fu Li, Chunlin Wei, Fubing Chen

Institute of Nuclear and New Energy Technology, Tsinghua University, Institute of Nuclear and New Energy Technology, Tsinghua University, 100084 Beijing, China

xiabing@tsinghua.edu.cn

 

Thorium is very attractive as sustainable resource in nuclear industry for its large resource inventory, high potential conversion ratio in thermal reactors because of the produced 233U, and less amount of long life MA products. HTR-PM is a demonstration plant of the modular high temperature reactor with two pebble-bed cores of 250 MWth. In a series of previous works, the feasibility of maximizing thorium utilization and minimizing the refueling effort of uranium fissile under the framework of HTR-PM has been investigated. Two types of thorium-based fuelling strategies of in situ utilization in HTR-PM were discussed, namely the Th+HEU (93% enriched) MOX fuel schemes and the Th+LEU SEP (uranium and thorium are loaded into SEParate pebbles) schemes, respectively. In this work, further investigation was implemented, including comprehensive parametric analysis on both strategies mentioned above by implementing coupled neutronics and thermal-hydraulics calculations and transient analysis in accidental conditions. The analysis on MOX schemes reveals that the one can obtain further uranium ore saving with about 9% uranium atomic fraction in the MOX fuels as the C/HM ratio decreases, compared with the traditional LEU loading. This is mainly attributed to the reduction of core leakage along with the increase of heavy metal loading. However, as the C/HM ratio decreases, most of the safety features of HTR-PM are observed to get worse, especially the maximum fuel temperature after DLOFC accident, which restricts the improvement of uranium ore saving. The optimized uranium ore saving is about 20 kg/GWd, corresponding to the thorium loading requirement about 10 kg/GWd, with far insufficient utilization fraction of thorium fuel up to 6% and relatively high-level residual 233U in spent fuels of about 0.2 kg/GWd. On the other hand, one can obtain more superior performance of in situ utilization of thorium in the SEP schemes if the thorium residence time is lengthened sufficiently. A series of SEP schemes of 7g heavy metal loading per pebble are investigated, with varied uranium to thorium ratio and thorium discharge burn-up, as well as fixed uranium discharge burn-up of 90 GWd/tHM. When the ratio of uranium to thorium is set as 12:3 and the thorium discharge burn-up goes as high as 200 GWd/tHM, the uranium ore saving for SEP schemes can reach about 10 kg/GWd, compared with the LEU loading scheme. Although this saving is smaller than the MOX case, the thorium loading requirement decreases dramatically with the increased thorium discharge burn-up, maximally one order of magnitude lower than the MOX schemes, indicating that the cost of thorium fuel can be minimized with relatively considerable uranium loading saving. The superior performance of the SEP scheme can be mainly attributed to the sufficient burning of thorium fuel, with maximum in-core utilization of thorium of more than 20%. It is concluded that the SEP schemes are better choices for in situ utilization of thorium and minimizing the uranium loading requirement than the MOX schemes.






09.09.2014 09:50

Probabilistic safety assessment - 501

 

Evaluation of dependency of human errors in PSA

Jan Prochaska

VUJE, a.s., Okružná 5, 918 64 Trnava, Slovakia

jan.prochaska@vuje.sk

 

Evaluation of human errors dependency plays important role in the PSA field. Importance of human factor including dependency follows from high portion of operator activities that are foreseen to be part of response on particular initiating events or to correct improper activity of safety systems including recovery actions. More and more detailed PSA works imply the need to use an effective, robust and error prone way to cope with enormous cases leading to real or potential dependencies of human errors as well as to perform PSA quantification in such a way which avoids biasing PSA results as whole.
The objective of this paper is to compare used methods to evaluate human errors dependency and analyze implementation how human errors dependency is implemented into PSA quantification process. Paper also introduces approach used by VUJE to evaluate and quantify human error dependencies including self dependency as well as inter personal dependency.






09.09.2014 10:10

Probabilistic safety assessment - 502

 

Benefits Of Robustness Analysis - Robustness Case Studies in European NPP´s

Peter Schimann

AREVA NP GmbH, Koldestraße 16, D-91052 Erlangen, Germany

peter.schimann@areva.com

 

Based on the performed national and EU-wide stress tests AREVA has implemented an additional procedure to define measure of increasing safety after the Fukushima accident. This additional initiated procedure so called robustness analysis. It is (at least it should be) a part of solving process related to the Fukushima-problem in order to perform safety studies before starting plant modifications.
AREVA has alredy performed a wide variety of safety studies, e.g.:

Robustness analysis for German PWRs and NPP Borssele in Netherland, as well
Safety improvement studies for Japanese NPP`s
Studies on bunkered emergency buildings (standard for German NPP`s, similar to planed Hardened Safety Core in France)
Study for implementation of core catcher in existing Japanese BWR`s
Influence of beyond design external hazards, like extreme temperatures, earthquake or flooding
Studies for improvement of SAMG’s

In the paper the effect of performed robustness analysis are be illustrated on the performed safety studies mentioned above.






09.09.2014 10:50

Reactor physics - 601

 

Evaluation of the efectiveness of a simple GFR2400 heterogenous control rod design

Stefan Cerba1, Branislav Vrban2, Jakub Lüley2, Vladimir Nečas1

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia1

Slovak University of Technology Faculty of Electrical Engineering and Information Technology Dep. of Nucl. Physics and Technology, Ilkovicova 3, 812 19 Bratislava, Slovakia2

stefan.cerba@stuba.sk

 

The Generation IV International Forum (GIF) is a cooperative international endeavor that is trying to define and perform research and development needed to establish feasibility and performance capabilities of the next generation of nuclear energy systems. The GIF Technology Roadmap has identified the six most promising reactor concepts. One of them is the Gas-cooled Fast Reactor (GFR). This fast-spectrum reactor is a high temperature He cooled system operated in closed fuel cycle. The GFR2400 reactor is considered to be the conceptual design of the large-scale Gas-cooled Fast Reactor. The GFR2400 design is featuring ceramic fuel and structural materials both allowing high temperatures and efficiency using helium coolant.
One of the biggest concerns in terms of such a challenging system is the reactivity control system. The GFR2400 reactor accommodates a system of control rods with highly enriched boron carbide. This system of 18 control and 13 diverse safety devices should provide appropriate reactivity worth for reactor regulation and shutdown. The GFR2400 reactor is still in development, therefore no final design of the control rods has been selected so far and only homogenous material compositions are available. In the heterogeneous control rod design it is assumed that the absorber material is distributed in form of pins symmetrically placed in several rings within the S/As. Due to the shielding effects of the adjacent absorber pins it is likely that the worth of the heterogeneous control rod is lower than the worth of the homogenous one. This study deals with the proposal of a simple heterogeneous control rod design and with the evaluation of its effectiveness. As the first step, the available homogenous design was studied using both stochastic and deterministic approaches to identify the absolute worth, the integral characteristics and the interferences of the control devices. On the basis of the achieved results a simple control rod design was proposed with an aim to achieve comparable reactivity worth with the homogenous design using exactly the same material composition and volume fractions of materials. In order to save absorber material the introduction of several neutron moderator materials as a part of control rod design was investigated. A special part of this paper is dealing with the introduction of system of movable reflector which can serve as an additional safety system for accidental reactivity removal. The analysis of this system was performed for both homogenous and heterogeneous control rods using the MCNP, SCALE and DIF3D codes.






09.09.2014 11:10

Reactor physics - 602

 

Verification of Power Distribution of real fuel loadings in WWER-440 reactor

Branislav Vrban1, Gabriel Farkas1, Ján Haščík1, Jakub Lüley1, Stefan Cerba2, Vladimir Nečas2, Peter Urban3

Slovak University of Technology Faculty of Electrical Engineering and Information Technology Department of Nuclear Physics and Technology, Ilkovičova 1, 812 19 Bratislava, Slovakia1

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia2

Slovenské elektrárne Mochovce Nuclear Power Plant, 935 39 MOCHOVCE, Slovakia3

branislav.vrban@stuba.sk

 

In pressurized water reactors, the fuel loading pattern has significant meaning in terms of both safety and economics. An optimal loading pattern is defined as a pattern in which the local power peaking factor is lower than a predetermined value during one cycle and the effective multiplication factor is maximized to extract the maximum energy.
This analysis serves as an independent computational assessment of power density spatial distribution in the reactor core after the loading of new fuel type with higher average enrichment of 4.87% 235U. Due to the fact that new fuel type was implemented into the standard operation of NPP Mochovce, the local power peaking factors exceeded the design limits, especially for these assemblies. The origin of this phenomenon can be explained either by higher enrichment of uranium (neutronics) or by the insufficient flow of coolant (thermo-hydraulic). The main objective of this analysis is to determine the power distribution across the core and the results comparison with the data obtained from the on-site power monitoring system in order to identify the principal source of power non-uniformity. The isotopic compositions of the fuel assemblies in the investigated time steps were calculated by SCALE 6.1.1 system and the deterministic NEWT module. The real detailed operational history of each subassembly in the one sixth of the core during the last three campaigns was used in the calculation to achieve the most reliable isotopic composition for the precise MCNP5 stochastic calculation. The incident neutron data libraries of temperature dependent cross sections were prepared by NJOY99.364 code from ENDF/B-VII.0 evaluated data. The analysis of the impact of power level uncertainty given by on site monitoring system to the calculated isotopic composition and the multiplication abilities of loaded assemblies were also carried out. The paper gives a brief description of the geometrical and material models used in the calculations. According to the results, the calculated spatial distribution of the power density correlates with the power density distribution determined by on site power monitoring system. The calculation demonstrates the significant impact of the fuel assemblies with higher enrichment on the power density distribution in their vicinity, which was not sufficiently taken into account at the fuel loading pattern design.






09.09.2014 11:30

Reactor physics - 603

 

Simulation of the NPP Krško startup core with CASL core simulator, VERA-CS

Fausto Franceschini1, Marjan Kromar2, Dušan Ćalić2, Andrew Godfrey3, Jess Gehin3

Westinghouse, Rue Montoyer 10, 1000 Bruxelles, Belgium1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

Oak Ridge National Laboratory, P.O.Box 2008, Oak Ridge, Tennessee 37831-6162, USA-Tennessee3

francef@westinghouse.com

 

The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. The CASL team is a consortium that consists of ten core partners and numerous contributing members, led by Oak Ridge National Lab (ORNL) and with Westinghouse as the Industry vendor representative. The main technology that drives CASL’s modeling and simulations is the Virtual Environment for Reactor Applications (VERA), which incorporates coupled physics and science-based models with state-of-the-art numerical methods, modern computational science and software development practices. The core simulator of VERA, VERA-CS, aims at enabling whole-core pin-by-pin transport coupled analysis. This paper shows the results obtained by applying VERA-CS to the core physics analysis of the Krško NPP, and specifically the startup physics tests for the initial core, showing excellent comparison with the measured data.






09.09.2014 11:50

Reactor physics - 606

 

Determination and validation of fuel cooling with the use of isotopic factors

Dušan Ćalić1, Marjan Kromar2

ARJE, analize in raziskave na področju jedrske energetike, d.o.o., Vrbina 17, 8270, Krško, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

dusan.calic@arje.si

 

The result of the analyses regarding the effect of fuel cooling on isotopic inventory of the NPP Krško fuel [1] were the obtained isotopic factors. Those factors were obtained by applying the effect of fuel cooling and redefining the number densities for each burnup step and finally to correct the isotopic vector for each operational and cooling condition. This was necessary to enhance the accuracy of the isotopic library in the CORD-2 package. In this paper the factors will be presented and validated by using typical fuel cycles in the Krško NPP.
[1] D. Ćalić, A. Trkov, M. Kromar, Effect of Fuel Cooling on Isotopic Inventory of the NPP Krško fuel, Nuclear Energy for New Europe 2013, Bled, Slovenia






09.09.2014 12:10

Reactor physics - 605

 

Computational study of neutron screens performance considering different absorbing materials

Nefeli Chrysanthopoulou1, 2, Gregory-Kyriakos Delipei2, Panayiota Savva2, Melpomeni Varvayanni2

N.C.S.R. "Demokritos" Institute of Nuclear Technology Radiation Protection, 27, Neapoleos Str., 15341 Aghia Paraskevi, USA-Rhode Island1

Aristotle University of Thessaloniki, Faculty of Engineering, School of Electrical and Computer Engineering, Nuclear Technology Laboratory, 54124 Thessaloniki, Greece2

University of Cyprus, Department of Mechanical and Manufacturing Engineering, Computational Science Laboratory UCY-CompSci, 75 Kallipoleos, Nicosia 1678, Cyprus 3

mylonakis@ipta.demokritos.gr

 

The upcoming development of the IV generation fast reactor systems requires the examination of the behaviour exhibited by the structural and fuel materials under the irradiation conditions prevailing in these reactors. The lack of operating fast reactors, hence the lack of capability to perform experiments in such irradiation environments can be compensated by creating the desired irradiation conditions in existing thermal research reactors, such as the Material Test Reactors (MTRs). This can be achieved via neutron shielding materials, the so-called neutron screens, which can be installed in specific irradiation locations of the above mentioned reactors. The effectiveness of various neutron screens in tailoring the neutron spectrum is here investigated, by examining different screen materials and thicknesses. The materials have been selected based on their capability to absorb neutrons of low energies (thermal neutrons, E<1eV) and neutrons of intermediate energies (epithermal neutrons, 1ev<E<1MeV), providing thus irradiation conditions similar to those of a fast reactor regarding the ratio of the fast (high energy neutrons, E>1Mev) to the thermal and epithermal neutron components. Such materials include Boron, Europium, Cadmium, Gadolinium, Molybdenum etc. The Monte Carlo code TRIPOLI is used to simulate the performance of the various neutrons screens, positioned in the reflector area of the Jules Horowitz Reactor (JHR, CEA/Cadarache). The screens are here considered of cylindrical shape with a central hollow which hosts the irradiated target. The present work is a preliminary study intending to highlight the materials that could be effectively used as neutron spectrum tailoring media, with final target to specify the most appropriate neutron screen. This will also require sensitivity studies on the screen thickness (planned for the next stage of the work) since the optimum screen configuration must combine a material with an absorbing capacity as high as possible with a screen thickness as thin as possible.






09.09.2014 14:00

Research reactors - 701

 

Fuel Burnup Calculation in ITU TRIGA Mark II Research Reactor by Using Monte Carlo Method

Mehmet Türkmen1, Üner Çolak2

Hacettepe University, Nuclear Engineering Department, 06800 Beytepe, Ankara, Turkey1

Istanbul Technical University, Energy Institute, Ayazaga Campus, 34469, Istanbul, Turkey2

tm@hacettepe.edu.tr

 

This study is considered as part of an ongoing extensive neutronic research for testing the computer model of ITU TRIGA Mark II research reactor. Validation of the reactor model has been successfully shown in a separate study. Thus, as a subsequent step, the primary purpose of this work is to estimate the burnup value of the fuel rods by using Monte Carlo method and to compare with the recorded reactor data. Operating conditions (e.g., control rod positions and material temperatures) used in the reactor model are thoroughly based on the operation number of No.1599 from the reactor log-book as of March 2013 with a produced total energy of about 272 MWh. All the feedback effects are modeled accordingly. MONTEBURNS2 and TRIGLAV5 codes with a state-of-art temperature-dependent neutron library of ENDFB/V-II.0 which is used for burnup calculations. The analyses are carried out for the cases of fission product poison free and in equilibrium. Effective core multiplication factor, keff, is presented as a function of total energy generated as the reactor operates at a power level of 250 kW and 10 W. This paper also provides a comparison with a previous study performed using TRIGLAV code for a total produced energy of 240 MWh before March 2002. The calculated core-averaged fuel burnup is found to be 0.25 MWd while reactor recorded data gives an average burnup of 0.16 MWd. In the light of current positions of control rods, this disagreement reveals that the current fuel loading have much more burnup than the recorded data suggest. At last, the outcomes are discussed from the viewpoint of refueling strategy of the reactor core.






09.09.2014 14:20

Research reactors - 702

 

Measurements of the in-core neutron flux distribution and energy spectrum at the TRIGA Mark II reactor of the Vienna University of Technology/Atominstitut

Marcella Cagnazzo1, Christina Raith2, Mario Villa1, Helmuth Böck1

Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna, Austria1

Atominstitut, Schüttelstr.115, A-1020 Wien, Austria2

mcagnazzo@ati.ac.at

 

The core of the TRIGA Mark II research reactor at the Vienna University of Technology/Atominstitut has been recently fully refurbished with new fuel, slightly irradiated. This new core configuration needs to be properly characterized in order to support future research activities. Aim of this work is to present the results of the measurements of the in-core neutron flux distribution and energy spectrum performed applying a method based on the synergetic use of the Monte Carlo code MCNP and of a de-convolution technique of activated foils. This method is very flexible and can be applied to characterize nuclear reactors that present a wide variability of core geometries, structural materials’ compositions, fuel composition and neutron energy spectra. The method allows to measure both slow and fast neutron components proving as result a neutron spectrum in 620 energy groups. In the case of the measurements presented in this work, the absolute neutron flux was evaluated within an accuracy less than 10%.






09.09.2014 14:40

Research reactors - 703

 

Thermal power calibration of the TRIGA Mark II reactor

Žiga Štancar1, Luka Snoj2

Jožef Stefan Institute, Reactor Engineering Division, Jamova 39, 1000 Ljubljana, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

ziga.stancar@gmail.com

 

The readings of ex-core nuclear instrumentation enable the safe operation of the reactor, give the operator the ability of precise reactor power monitoring and play a crucial role in the normalisation of reactor calculations. At the IJS TRIGA Mark II reactor the instrumentation is periodically calibrated using the calorimetric method. Though relatively simple and reliable it can be burdened with up to 30 % uncertainty. In the paper a new calibration process using electrical heaters is presented, with which the heat capacity constant of the reactor pool is calculated to be C=19.6 kWh/K ± 0.3 kWh/K and the uncertainty of the thermal power value is significantly reduced with an estimate of 2%.






09.09.2014 15:00

Research reactors - 712

 

Application of best estimate plus uncertainty in review of research reactor safety analysis

Simon Adu1, Marco Lanfredini2, Francesco D'Auria3

1Ghana Atomic Energy Commission, Radiation Protection Institute, P.O. Box LG80, 00233 Accra, Ghana

2Nuclear Research Group of San Piero a Grado, San Piero a Grado, 56122 Pisa, Italy

3University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy

s.adu@gaecgh.org

 

Energy literacy is an understanding of the nature and role of energy in the universe and in our lives. It is the ability to apply this understanding to answer questions and solve problems, especially regarding our energy future. How does nuclear literacy fit into the broader energy literacy concept?
The authors will present one of the possible answers to this question from the perspective of a web-project eWorld (called “eSvet” in Slovene) initiated and developed by energy company GEN energija with expert partners in 2013 and 2014 (and planned to be launched in September 2014).
The paper will bring insights into the communication process of developing the web project with details regarding:
- the multidimensional content structure of the web-page, taking into account the importance of:
o knowledge about energy,
o links between energy and sustainability and
o the necessary elements for discussing our energy future.
- the positioning of nuclear energy related contents in the broader framework of energy literacy and energy-future debate,
- the organisational aspect, i.e. the web-project stakeholder map (developers and potential users of the eWorld web-page), including a discussion of possible combination of future on- and off-line communication activities of GEN energija and partners.
This will be a first-of-the-kind presentation of the new web-page for NENE 2014 conference participants.






09.09.2014 15:00 Poster session

Nuclear energy and society - 204

 

Improving nuclear literacy through e-communication: the “eWorld” web project

Melita Lenošek Kavčič1, Mojca Drevenšek2

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia1

CONSENSUS, komunikacije za odgovorno družbo, d.o.o., Beethovnova ulica 9, 1000 Ljubljana, Slovenia2

melita.kavcic@gen-energija.si

 

Energy literacy is an understanding of the nature and role of energy in the universe and in our lives. It is the ability to apply this understanding to answer questions and solve problems, especially regarding our energy future. How does nuclear literacy fit into the broader energy literacy concept?
The authors will present one of the possible answers to this question from the perspective of a web-project eWorld (called “eSvet” in Slovene) initiated and developed by energy company GEN energija with expert partners in 2013 and 2014 (and planned to be launched in September 2014).
The paper will bring insights into the communication process of developing the web project with details regarding:
- the multidimensional content structure of the web-page, taking into account the importance of:
o knowledge about energy,
o links between energy and sustainability and
o the necessary elements for discussing our energy future.
- the positioning of nuclear energy related contents in the broader framework of energy literacy and energy-future debate,
- the organisational aspect, i.e. the web-project stakeholder map (developers and potential users of the combination of future on- and off-line communication activities of GEN energija and partners.
This will be a first-of-the-kind presentation of the new web-page for NENE 2014 conference participants.




09.09.2014 15:00 Poster session

Nuclear energy and society - 206

 

Travelling Exhibition Fusion Expo

Tomaž Skobe

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

tomaz.skobe@ijs.si

 

The Fusion Expo is a travelling exhibition where visitors could explore fusion energy. Since the beginning of the Fusion Expo support actions under EFDA (European Fusion Development Agreement) in 2008, the Fusion Expo has been the responsibility of the Slovenian Fusion Association (SFA).
Fusion expo is presenting fusion energy as an environmentally acceptable, safe and sustainable energy source. Fusion research, technology and its future use are presented to the citizens of Europe.
Main target group of this exhibition are mediators such as journalists, teachers, decision makers, NGOs, students, taxpayers, voters, primary and secondary school pupils. The paper will present six years of activities of the Fusion Expo support actions under EFDA.






09.09.2014 15:00 Poster session

Nuclear energy and society - 207

 

Nuclear Power Plant Krško 2 Action Plan

Žiga Arnšek, Tomaž Ploj, Jože Špiler

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia

ziga.arnsek@gen-energija.si

 

GEN energija d.o.o., in the context of its business and development plan, carries out activities for the preparation on the construction of a new Nuclear Power Plant (NPP) Krško 2. Construction of new NPP is a large and extensive project which demands a lot of activities to be done in the preparatory phase of the project.
The paper will present the scope and content of NPP Krško 2 Action Plan. It outlines necessary activities to provide technically justified, efficient, transparent and responsible development of the project and it`s successful completion. Action plan provides an overview of important phases of the project and discuss short-term and mid-term activities arising therefrom. The NPP Krško 2 Action plan also provides the strategy of the project management and project implementation.
Important part of the NPP Krško 2 Action plan is project schedule with individual phases, assumptions and the most important activities. It also to defines owner`s responsibilities during different phases and to describe comprehensive and integrated project management plan, project phases and processes. The delineation of project schedule on 5 different phases and descriptions of each phase were prepared as well as description of project implementation schedule assumptions, major steps of project management like strategic decision-making process and establishment of a company for construction phase. Action plan also establish project management system with associated procedures and software support.
Key phases of action plan are planning and preparation phase, site evaluation and preparation, bidding process, construction and commissioning with preparation for operation phase.






09.09.2014 15:00 Poster session

Nuclear energy and society - 209

 

Public opinion and nuclear energy: INPRO methodology and recent findings in the open literature

Rolando Calabrese

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy

rolando.calabrese@enea.it

 

Complex decisions need besides economic utilization and technological practicability, the acceptability of the outcome to the various stakeholders. If the decision is not well accepted by society, it has little chance of successful implementation regardless of its economic and technical merits. Most nuclear–related decision–making processes deal heavily with public opinion attitude. With this regard, different aspects may influence the public support for nuclear: demographic factors, knowledge, risks perception. Moreover, nuclear accidents have always changed public attitudes where a slow recovery was seen following the event. The recent Fukushima accident has had a significant effect on the nuclear policies of many countries whose governments have changed or redirected their investments in nuclear energy.
The methodology of INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) has been developed specifically to determine whether or not a given innovative nuclear energy system (INS) is sustainable. INPRO has established, for each area identified in the methodology, a set of requirements, organized in a hierarchy of basic principles, user requirements and criteria, including indicators and acceptance limits, that must be fulfilled to meet the overall target of sustainable energy supply. Public information, participation, acceptance and political support issues are addressed in the infrastructure area under the statement of user requirement 3.Based on a review of the results published in the open literature, the approach of INPRO to the topics of public opinion and political support in the process of decision–making regarding the use of nuclear energy, is discussed. Finally, moving from a hierarchical approach, feedbacks and inter–relationships in this area are proposed presenting preliminary results obtained by means of the Analytic Network Process.






09.09.2014 15:00 Poster session

Nuclear energy and society - 211

 

Regulatory oversight of activities following the discovery of damaged fuel assemblies during the Krško NPP outage 2013

Tomaž Nemec

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

tomaz.nemec@gov.si

 

In October 2013 the refuelling outage started in the Krško NPP. During inspection of fuel assemblies unloaded from reactor core several fuel assemblies with open defects were found as well as other fuel assemblies with tight defects. The cause of open failure of fuel cladding was vibrations due to baffle jetting at the core baffle plate locations. Corrective measures were implemented prior to NPP core reload to prevent recurrence of such fuel defects in new fuel cycle.
The regulatory body of Slovenia, the Slovenian Nuclear Safety Administration (SNSA), regularly follows the conditions of fuel cladding integrity through a system of safety performance indicators. Open fuel cladding defects were already diagnosed from reactor coolant activities since July 2012, 14 months before the outage 2013. During the outage 2013 activities involving damaged fuel inspections, identification of damage mechanisms and implementation of corrective actions the SNSA inspectors onsite followed closely the findings and actions performed by the fuel provider and the Krško NPP personnel. The SNSA relied also on special expert opinion of authorized institutions (TSOs) that supervised the Krško NPP activities involving fuel assemblies.
The SNSA formed a group of experts to review fuel inspection results, assess fuel conditions and provide an opinion on conclusion of Krško NPP determination of causes for fuel damage and proposed corrective actions. To expand the knowledge base of Slovenian experience with damaged fuel, the SNSA reviewed databases of operational events involving fuel damage in foreign NPPs and contacted also the regulatory body of the USA, the country of origin of the Krško NPP and its fuel assemblies. The SNSA recommended to the Krško NPP management to introduce restricted limits for the fuel leakage in the next fuel cycle with associated actions in case of degradation of fuel assemblies integrity. The SNSA also performed a major task of providing information on all activities and fuel conditions to the Slovenian public, NGOs and international expert organisations. The SNSA continues with monitoring of fuel conditions and assessment of corrective actions in new fuel cycle.






09.09.2014 15:00 Poster session

Nuclear energy and society - 212

 

Sensitivity analyses to support determination of emergency planning zones around the nuclear power plant

Andrej Stritar, Barbara Vokal Nemec, Michel Cindro

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

barbara.vokal-nemec@gov.si

 

In Slovenia the protective measures for population in case of a nuclear accident are determined by National protection and rescue plan for nuclear or radiological accidents that was upgraded last time in 2010. The bases of the plan were prepared three decades ago at the time when Krško NPP started its operation. These bases comprise also the determination of the protective action zones around the NPP as well as predetermined measures for protection of population in these zones.
New recommendations by the IAEA define four zones of preparation of protective measures off the NPP site: PAZ (precautionary action zone), UPZ (urgent protective action planing zone), EPD (extended planing distance) and ICPD (ingestion and commodities planing distance). Current Krško NPP zones OTU and ODU differ from these internation recommendations. Also the NUREG 0396 that is used by the Krško NPP for selection of zone area for protection measures recommends a 10 miles (16 km) OTU zone and 25 miles (40 km) ODU zone. Recommended sizes of UPZ in modern references are larger than those defined in existing National plan (15 to 30 km).The paper will describe analyses that were performed at the Slovenian Nuclear Safety Administration to produce updated assessment of potential threat to the NPP vicinity in case of most severe accidents. With knowledge and tools available we performed simulations of several different scenarii of large radioactivity releases from the Krško NPP. Source terms used as input were selected based on IAEA recommendations. We used generic data of core isotopes inventory and design data on filters decontamination factors of the passive containment filtering venting system. The filtered radioactive release is not reduced for the noble gases but it is 100 times reduced for released Iodine isotopes and 1000 times for the released aerosols (particulates).We used meteorological data of real weather conditions and simulated the dispersion of radioactive material with the RODOS code.
The resuls of performed analyses can be a significant input for revision of bases for the National protection and rescue plan for nuclear or radiological accidents in the vicinity of the Krško NPP. Since this is a rather complex issue to be solved we will need to take into account much wider choice of input data for additional analyses that have to be performed.






09.09.2014 15:00 Poster session

Nuclear energy and society - 213

 

Regulatory Review of the second Periodic Safety Review of the Krško NPP

Davor Lovinčič

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

katarina.kasnar@gov.si

 

The Krško NPP has started its second Periodic Safety Review (PSR2) at the beginning of 2009 following the program which sets out the content, scope and timetable. The first reports were submitted to the Slovenian Nuclear Safety Administration (SNSA) at the beginning of 2012. By April 2014 a total of 41 reports and sub-reports were submitted. The SNSA reviewed all of the reports and provided comments. The review was focused on whether the PSR2 presented a complete review of a certain factor and whether eventually missing topics should be included. In December 2012, several meetings between the SNSA and the Krško NPP were held on individual PSR2 topics. At these meetings the remaining open issues in the final revisions of the reports were discussed and by May 2013 most new revisions of the reports were submitted to the SNSA.
By December 2013, all of the revised reports together with a summary report were delivered to the SNSA. The summary report contained a list of PSR2 recommendations with explanation of prioritization process for the action plan. At the SNSA the prioritization of PSR2 issues was evaluated to confirm the ranking of the issues. Since some of the PSR2 topics were screened-out, the SNSA performed a re-evaluation of the PSR2 issues suggesting additional PSR2 recommendations from which 78 were included into the action plan. PSR2 has proven that the plant safety is in accordance with modern standards and that the plant can operate safely for next 10 years. Implementation of the action plan in next 5 years represents the challenge to plant and SNSA. It will additionally improve the overall plant safety.






09.09.2014 15:00 Poster session

Nuclear energy and society - 214

 

ALARA in Practice

Matjaž Koželj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

matjaz.kozelj@ijs.si

 

Implementation of radiation protection relies on three principles: justification of practice, optimisation of exposures and dose limits. While justification of practice and dose limits are somehow consequences of societal reasoning and decisions, optimisation of exposures is imposed as a requirement that should be satisfied by source users on the case-by-case and day-by-day basis. Optimisation is therefore “living” requirement that must be understood and implemented by users.
In radiation protection this requirement is usually interpreted as ALARA (As Low as Reasonably Achievable) principle which relates to dose. At first glance, there is no problems, since en average user understands this as a requirement that his/her dose should be as low as possible. Of course, there is no explanation on how to reach it, but there must be a way since this is official requirement. Unfortunately, things are not as simple as it looks at first sight. First, ALARA is not about reaching the lowest possible doses, but about reasonably achievable (low) doses. It is a method of optimisation, where dose is not the only factor. Second, ALARA does not necessary mean that each and every worker will have the lowest (reasonably achievable) dose, since ALARA relates also to collective dose. Third, methods used for ALARA optimisation could be different from place to place and could change with time, therefore results of optimisation could differ for the similar/same practice. ALARA is therefore requirement that should not be implemented in copy/paste manner, it must be understood and implementation re-evaluated even for same practice on regular basis.
In paper, author will explain the background of ALARA principle and present an overview of approaches and implementation of ALARA principle for various practices. Author will also present some recommendations for application of ALARA principle in practices where source users do not have direct and continuous support of radiation protection experts.






09.09.2014 15:00 Poster session

Nuclear energy and society - 215

 

Bid Technical Evaluation Process for JEK 2 Project

Aleš Kelhar

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia

ales.kelhar@gen-energija.si

 

A process of bidding is one of the most important tasks that GEN energija will have to carry out before construction of the second unit of nuclear power plant (JEK 2). It consists of several phases. The first phase of the bidding process is the preparation of bid invitation specifications (BIS). The following phases are the evaluation of the bids provided by the bidders and the contracting with the successful bidder. Based on IAEA the evaluation process generally taking not less than 6 months depending on available human resources and their qualification skills.
The paper highlights a bid evaluation process with special attention to a bid technical evaluation process for JEK 2. In addition to the technical (including safety) evaluation, the overall Bid evaluation process comprises economic, financial, contractual and other applicable aspects which have to be considered in the decision-making process of implementing the project and the selection of the supplier(s). This process starts with the receipt of the bids and ends with the issuance of the final evaluation report.
A project organisation and process flow chart for the technical evaluation are presented and discussed. They are based on our own Quality Management System for JEK 2 Project considering the IAEA recommendations and European Utility Requirement group (EUR) experience.






09.09.2014 15:00 Poster session

Severe accidents - 308

 

Analysis of Stratified Steam Explosions

Matjaž Leskovar, Vasilij Centrih

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

matjaz.leskovar@ijs.si

 

When during a severe reactor accident the hot reactor core melt comes into contact with the coolant water a steam explosion may occur. During the steam explosion the energy of the molten corium is transferred to the coolant in a timescale smaller than the timescale for system pressure relief and so induces dynamic loading of surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and thus lead to the radioactive material release into the environment.

In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, are usually disregarded as being incapable to generate strong explosive interactions. The main reason for this assumption of the low energetics of stratified steam explosions is based on the hypothesis that the amount of melt in the premixture formed in stratified configurations is insufficient to produce strong explosions. This hypothesis was based on analytical considerations that interfacial instabilities in a stratified configuration are not efficient in creating an explosive premixuture. It was supported with data from experiments performed with mostly low temperature liquids, which showed rather low energy conversion efficiency and slow explosion propagation. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts a considerable melt-coolant premixing layer formed. In these experiments the energy conversion ratio was even higher than in experiments carried out with the “conventional” melt jet-coolant pool configuration with prototypic materials. In the paper, the study of stratified steam explosions performed with the MC3D code in PULiMS like conditions will be presented. The influence of various parameters/approaches on the simulation results has been analysed, e.g. mixing layer thickness, fraction of phases in premixture, water layer thickness, melt spreading area, triggering location, 2D/3D modelling approach. The calculations were compared to available experimental data. Various simulation results will be presented and discussed. Suggestions for further analytical and experimental work will also be given.






09.09.2014 15:00 Poster session

Severe accidents - 309

 

Simulation of THAI Hydrogen Deflagration Experiments with Upward Flame Propagation in Homogeneous Atmosphere using ASTEC Severe Accidents Code

Ivo Kljenak

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

ivo.kljenak@ijs.si

 

The issue of hydrogen combustion during a severe accident in a nuclear power plant (NPP) came to prominence after the accident at the Three Mile Island (USA) NPP in 1979, and has received new attention since the accident at the Fukushima Daiichi (Japan) NPP in 2011. The Fukushima accidents also highlighted that both in-depth understanding of severe accident sequences and development or improvement of adequate severe accident management measures are essential in order to further increase the safety of NPPs operated in Europe. CESAM (Code for European Severe Accident Management) is a research and development project within the 7th Framework Program of the European Commission that aims in particular at the improvement of the European reference code ASTEC towards usage in severe accident management analyses for NPPs.
Within assessment of the hydrogen combustion modelling in the ASTEC code, nine experiments on hydrogen combustion, which were performed in the THAI experimental facility, were simulated at the Jozef Stefan Institute. The THAI experimental facility, located at Becker Technologies GmbH in Eschborn (Germany), is basically a single-volume cylindrical vessel, with a volume of 60 m3, an internal height of 9.2 m, and an internal diameter of the main part of 3.2 m. In the considered experiments, combustion with upward flame propagation occurred in air-hydrogen and air-steam-hydrogen atmospheres, at different temperatures, pressures and initial hydrogen concentrations. A multi-volume input model of the THAI facility was developed for the ASTEC code, and the experiments were successfully simulated. The calculated pressures, temperatures and flame propagation axial velocities are compared to experimental results and discussed.






09.09.2014 15:00 Poster session

Severe accidents - 311

 

Development of the Post-Accident Strategy after the Nuclear or Radiological Accident

Helena Janžekovič, Andrej Stritar, Darja Slokan Dušič, Marjan Tkavc, Barbara Vokal Nemec

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

helena.janzekovic@gov.si

 

Even today the Fukushima accident is challenging not only the operator and nuclear regulator in Japan but also the entire international and national nuclear safety framework. This framework is actually based on lessons learned from a relatively small number of nuclear accidents with wide consequences. As a result, major developments in various fields, e.g. areas related to prevention as well as mitigation of severe accidents, are under way. One of the fields, which have been somehow put aside in the past, is the development of a national strategy applied in the post-accident phase.
In Slovenia the post-accident strategy has been foreseen already in the National Emergency Response Plan for Nuclear and Radiological Accidents. According to this plan the Slovenian Nuclear Safety Administration (SNSA) should develop the Post-Accident Strategy after the Nuclear or Radiological Accident (the Post-Accident Strategy). The development of the document was finalised in 2013. The Post-Accident Strategy is based on lessons learned from nuclear and other accidents, e.g. flooding, as well as on very rare public documents related to post-accident strategy prepared in other countries, e.g. France. The paper presents main components of the strategy in Slovenia.
According to the document mentioned the Government could formally establish a team of experts, e.g. crisis ministry. One of the main tasks of the team is to prepare so-called Rehabilitation Program reflecting actual situation. The Post-Accident Strategy focuses on seven strategic areas or measures which should take place outside the operator’s site not neglecting strong cooperation with a nuclear operator, if necessary. The areas identified encompass:
• Radiation protection,
• Monitoring of the environment,
• Protective actions,
• Mitigation of consequences of countermeasures,
• Informing,
• Restriction of economical consequences and liability,
• Revitalisation.
All areas must be incorporated in the Rehabilitation Program. Among them radiation protection requires specific attention as uncontrolled radiological situation is a unique characteristic of nuclear or radiological accidents and is strongly connected to all other areas. The Post-Accident Strategy is in line with the ICRP 103 recommendations incorporating flexibility when emergency and existing exposure situations are studied.
For each strategic area the Post-Accident Strategy proposes a leading institution responsible for a specific measure and a list of national databases in order to control a situation on a long term. Example of such database is the “Central register of actions related to radioactive waste”. The document also identifies challenges ahead for institutions which might play a major role after a nuclear or radiological accident in Slovenia. The Post-Accident Strategy is the first attempt to a systematic approach to post-accident management in Slovenia.






09.09.2014 15:00 Poster session

Severe accidents - 312

 

SARNET benchmark on Phébus FPT3 integral experiment on core degradation and fission products behaviour

Mirco Di Giuli1, Tim Haste2, Regis Biehler3

Universita degli Studi di Bologna Nuclear Engineering Laboratory (LIN) of Montecuccolino, Via dei Colli, 16, 40136 Bologna, Italy1

Institut de Radioprotection et de Sureté Nucléaire, Bât. 702 Centre de Cadarache, BP 3-13115 Saint Paul lez Durance, France2

Institut de radioprotection et de surete nucleaire (IRSN), 60-68 Avenue de General Leclerc - BP17, 92262 FONTENAY-AUX-ROSES CEDEX, France3

mircodigiuli@libero.it

 

In the frame of the EU network of Excellence the SARNET2 work package WP8.3 ”Bringing Research Results into Reactor Application” task “Benchmarking of available codes against integral experiments” the PHEBUS FPT3 experiment, has been chosen as the basis for this benchmark. The aim was to assess the capability of computer codes to model in an integral way the physical processes taking place during a severe accident in a pressurised water reactor, from the initial stages of core degradation, the fission product transport through the primary circuit and the behaviour of the released fission products in the containment. The FPT3 Benchmark was supported, with participation from 16 organisations in 11 countries, using 8 different codes. The temperature history of the fuel bundle and the total hydrogen production, also taking into account of the H2 generated by the boron carbide oxidation were well captured, but no code was able to reproduce accurately the final bundle state, using as bulk fuel relocation temperature, the temperature of the first significant material relocation observed during the experiment. Total volatile fission product release was simulated, but its kinetics was generally overestimated, concerning the modelling of semi-volatile, low-volatile and structural material release, needs some improvement, notably for (Mo, Ru) for which it was observed a substantial difference between bundle and fuel release. The retention in the circuit was not well predicted, this was due mainly to the boron blockage formation in the rising line of the steam generator, and the volatility of some elements (Te, Cs, I) could be better predicted. Containment thermal hydraulics are well calculated while as regards the containment aerosol depletion rate only the stand-alone cases provide acceptable results, whilst the integral cases tend to largely overestimate the total aerosol airborne mass. Calculation of iodine chemistry in the containment turned out to be difficult. Its quality strongly depends on the correct prediction of chemistry speciation in the integral codes. The major difficulties are related to the presence of high fraction of iodine in gaseous form in the primary circuit, which affects the iodine behaviour in the containment. In the benchmark a significant user effect was detected, i.e. results achieved with the same code differed considerably. This work reports the benchmark results comparing the main parameters, and summarises the results achieved and the implications for plant calculations.






09.09.2014 15:00 Poster session

Severe accidents - 313

 

Pressurized Water Small and Medium Reactor (SMR) Modelization and Severe Accident Analysis using ASTEC code

Mirco Di Giuli1, Marco Sumini2, Giacomino Bandini3

Universita degli Studi di Bologna Nuclear Engineering Laboratory (LIN) of Montecuccolino, Via dei Colli, 16, 40136 Bologna, Italy1

University of Bologna, Faculta di Ingehneria, Viale Risorgimeno 2, 40136 Bologna, Italy2

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy3

mircodigiuli@libero.it

 

According to classifications adopted by the IAEA, small reactors are characterized by an equivalent electric output of less than 300 MW while medium sized reactors by an equivalent electric power between 300 and 700 MW. Pressurized water small and medium sized reactors (SMR) generally adopt an integral layout of the primary circuit with in-vessel location of steam generators and control rod drives; one compact modular loop-type design features reduced length piping, an integral reactor cooling system accommodating all main and auxiliary systems within a leak tight pressure boundary, and leak restriction devices. In this study, modeling and nodalization of primary loop, secondary loop and passive core cooling system and containment for a SMR were conducted, using the main ASTEC code modules: ICARE module for the vessel, CESAR for the primary circuit, secondary circuit and for the control and safety systems and CPA for the containment. The SMRs as well as the advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts.
Thus, to simulate correctly the main phenomena involved during an accident scenario, the coupling between primary circuit and containment has to be reproduced accurately, by using of different parameter to calculate the condensation rate of steam on the containment walls. Furthermore, given that the containment plays a fundamental role during every accident scenario, it had to be taken into account just as a real safety system.
The main design basis events and one severe accident scenario for the SMR were analysed, and the calculated results were compared with those obtained by Westinghouse using the coupled codes SCDAP-RELAP-GOTHIC in the SMR preliminary safety assessment document.
The aim of this work is to validate the capability of the ASTEC code to simulate the entire accident event and the possible fission product releases in a SMR.


09.09.2014 15:00 Poster session

Severe accidents - 314

 

Transition boiling modelling in the MC3D code

Mitja Uršič1, Renaud Meignen2, Gabrijela Ikovic3

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Institut de Radioprotection et de Sureté Nucléaire, PSN-RES/SAG, BP-17, 92262 Fontenay-aux-Roses, France2

Fakulteta za matematiko in fiziko, Jadranska 19, 1111 Ljubljana, Slovenia3

mitja.ursic@ijs.si

 

The Generation IV International Forum has selected six technologies for future nuclear reactors, including innovative sodium cooled fast reactor. One of the important issues in core melt progression during a severe accident in the sodium cooled fast reactor is the likelihood and the consequences of a vapour explosion. A vapour explosion can occur when the hot core melt comes into contact with sodium. A strong enough vapour explosion in a nuclear power plant could jeopardize the reactor vessel and so lead to a direct release of radioactive material to the environment.
Several experimental programs were launched to help understanding and characterizing the vapour explosion phenomenon during the melt-sodium interaction. These experiments revealed an important effect of the sodium sub-cooling on the behaviour of the melt-sodium interaction. The vapour explosion probability and efficiency is decreasing with the sodium sub-cooling. The physical properties of sodium, which strongly affects the melt-sodium heat transfer, and the melt solidification, which strongly affects the energy efficiency during the explosion, are identified as the reason for the observed behaviour. On the other hand, important analytical activities have been performed to study the vapour explosion phenomenon in light-water reactors. The analytical researches are performed with the use of comprehensive computer codes that are devoted to study the fuel-coolant interaction phenomenon. The models of the key processes that are implemented into the codes were developed manly with the focus on the water. Therefore the applicability of the codes and models for the fuel-sodium interaction must be demonstrated.
In this paper the transition boiling modelling with the MC3D code is analysed. The MC3D code, IRSN, France, is a comprehensive fuel –coolant interaction code that has the potential to simulate the interaction with sodium.
For sodium, it is likely that boiling occurs essentially in the transition boiling regime and then the modelling of this regime is the most important. Indeed, film boiling regime should be reasonably reproduced with the models already existing for water. But the transition boiling regime is difficult to model, because this is essentially an unstable film boiling situation. The purpose of the paper is to propose a model for the transition boiling regime. The first objective is to analyse experiments with water and sodium. In the experiments it was observed that the temperature range of the transition boiling regime strongly depends on the coolant sub-cooling. Additionally, in the experiments with the sub-cooled water, it was observed that a relative constant heat flux exists in the transition boiling region. This is important to consider because the innovative sodium cooled fast reactor will operate at the sub-cooling of 300-400 K. The second objective is to propose a model for a heat transfer and a temperature range for a relative constant heat flux. The model will be compared with the experimental data.






09.09.2014 15:00 Poster session

Multiphysics - 403

 

Optimization of an integrated neutronic/thermal-hydraulic reactor core analysis model

Antonios Mylonakis1, Melpomeni Varvayanni1, Dimokratis Grigoriadis2, Panayiota Savva1, Nicolas Catsaros1

National Centre for Scientific Research “Demokritos”, Institute of Nuclear & Radiological Sciences & Technology, Energy & Safety, Nuclear Research Reactor Laboratory, Agia Paraskevi Attikis, P.O.Box 60037, 153 10 Athens, Greece1

University of Cyprus, Department of Mechanical and Manufacturing Engineering,Computational Science Laboratory UCY-CompSci , 75 Kallipoleos str., 1678 Nicosia, Cyprus2

mylonakis@ipta.demokritos.gr

 

Within the context of an operating nuclear reactor core, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance in reactor safety and design. In this work, the ongoing development of a tool for neutronic/thermal-hydraulic coupled calculations is discussed. OpenMC, a Monte-Carlo 3D neutronic code and COBRA-EN, a thermal-hydraulic code for sub-channel and core analysis, are integrated in a single tool for coupled calculations. This new coupled system (OpenMC/COBRA) is capable of performing coupled neutronic/thermal-hydraulic analysis in both subchannel and core level. As regards the main key parameters concerning the coupled problems, the handling of the involved feedbacks between the two physical processes, the accuracy of the Monte-Carlo calculation and the convergence of such an iterative scheme, are the main ones. Another issue which should be considered carefully is the optimal, in terms of computational time, use of the neutronic Monte-Carlo code, since the main disadvantage of such codes is the high computational cost. This work investigates the role of these parameters in the coupled neutronic/thermal-hydraulic calculations performed by OpenMC/COBRA while it examines in which direction their optimization should move in order to achieve accurate results with reasonable computational cost. The results show that satisfying accuracy can be obtained in reasonable computational time when Monte-Carlo multi-processing is combined with proper selection of the Monte-Carlo parameters as well as the parameters of the coupled iterative scheme.






09.09.2014 15:00 Poster session

Multiphysics - 404

 

Coupling the ATHLET3.0 and the KIKO3DMG multigroup 3D kinetic code developed for the fast spectrum Gen4 reactors

György Hegyi1, Andras Kereszturi1, István Pataki2, Ádám Tóta1, Kiril Velkov3, Ihor Pasichnyk4, Yann Périn4

Hungarian Academy of Sciences, Centre for Energy Research , P.O. Box 49, 1525 Budapest 114, Hungary1

Centre for Energy Research, Hungarian Academy of Sciences , Konkoly Thege M. út 29-33, H-1121, Hungary2

Gesselschaft für Anlagen- und Reaktorsicherheit mbH Forschungsgelände,  D-85748 Garching, Germany3

Gesellschaft für Anlagen-und-Reaktorsicherheit (GRS)mbH, Schwertnergasse 1, D-50667 Köln, Germany4

gyorgy.hegyi@energia.mta.hu

 

Future sustainable nuclear energy production worldwide with closed fuel cycle needs application of fast spectrum reactors cooled by liquid metal or gas. For this reason, several developments of the thermal hydraulic system and 3D reactor kinetic codes are going on. Presently, their coupling became one of the most current tasks to be solved for performing safety analyses.
Such types of coupled codes have been applied mainly for performing analyses of second and third generation reactors. In Hungary, the ATHLET thermohydraulic system code and the KIKO3D neutronics code were coupled and widely used for performing not only safety analyses of the VVER reactors on operation but also for design purposes for new reactor concepts(SCWR). Recently both codes have been further developed in the direction of the fourth generation research.
The ATHLET (Analysis of Thermal-hydraulics of Leaks and Transients) is a well verified thermo-fluid-dynamic system code developed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) in Germany. It offers the possibility of choosing between different models for the simulation of fluid dynamics. The code structure is highly modular and allows an easy implementation of different physical models. Other independent modules (e.g. large models with own time advancement procedure) can be coupled without structural changes in ATHLET by means of a general interface. ATHLET provides a modular network approach for the representation of a thermal-hydraulic system.
The GRS methodology is developed for pseudo 3D TH description of the core with mapping schemes fitted to the 3D neutronically simulated active core. A large number of benchmarks have been successfully solved with the coupled system code and many analyses have been performed for NPPs with different reactor types. An other brand-new development of ATHLET 3.0 makes the thermal analysis of sodium, lead and gas cooled reactors possible.
In MTA EK (Hungary), the new multi-group version of the KIKO3D 3D kinetic code - called KIKO3DMG and based on a new programming structure of the algorithms utilizing the modern multithreaded computer features - is elaborated and verified for the neutronic calculations. In this paper the coupling of these upgraded codes and their verification are presented. The basis of the demonstration of the correct coupling is the calculation of the large oxide fuelled sodium cooled open core (without primary loop) defined as a fast reactor benchmark in the OECD NEA WPRS SFR cooperation. The ATHLET input of the core had been elaborated. The obtained k-eff value of the initial state is reasonably close to the benchmark value defined without feedback by supposing a constant average temperature of the fuel and the coolant. Starting from this stationary state, three “transients” were calculated, namely :- steady state transient for proving the consistency of the stationary and the transient algorithms of both codes,- partial insertion of a control rod in an asymmetric radial position,- partial withdrawal of a control rod in an asymmetric position.
The time dependent behavior of the obtained power, reactivity and temperatures showed that the thermal and the neutronic processes are influencing each other in the expected, reasonable way.






09.09.2014 15:00 Poster session

Multiphysics - 405

 

Peach Bottom Low Flow Stability Analysis with TRACE/PARCS

Consuelo Gómez-Zarzuela1, Teresa Barrachina1, Agustin Abarca1, Rafa Miró2, Gumersindo Verdú1

Polytechnic University of Valencia, Department of Chemical and Nuclear Engineering, Camí de Vera sn, 46022 Valencia, Spain1

Universitat Politecnica de Catalunya, C. Jordi Girona, 31, 08034 Barcelona, Spain2

gverdu@iqn.upv.es

 

Instability events have taken place in several BWR nuclear power plants, during normal operation or experimental test, due to the non-linear dynamic responses of these facilities. There are commonly three instability types observed in BWR: Control system instabilities, channel thermohydraulic instabilities and coupled neutronic-thermohydraulic instabilities. The last type involves two feedback loops, on the one hand, the neutronic feedback, where reactivity changes are caused by variations in void fraction (density), and on the other hand the thermohydraulic feedback, which affects the inlet flow rate. In this case the reactor core becomes unstable and starts to oscillate due to a positive pressure drop feedback because of density wave oscillations. Two modes of oscillations are observed within the core, the core-wide or in-phase oscillation and the regional or out-of-phase oscillation. The out-of-phase oscillation, which is the topic of this work, the power of half core oscillates out-of-phase in relation to the other half.
Since performing empirical test in nuclear power plants means great difficulty, as well as high costs, in order to simulate complex scenarios as those related to BWR instabilities, the use of coupled thermal-hydraulic and neutron kinetics system codes is a very useful option. In this work, the simulations have been performed by using the thermal-hydraulic system code TRACE coupled with the neutron kinetic code PARCS v3.0.The aim of this work is to simulate with TRACE/PARCS coupled code the stability behavior occurring on a point of the exclusion region of the operating power-flow map, called PT_UPV of Peach Bottom Unit 2. In this nuclear power plant, four low-flow stability tests were realized in 1977 at the end of cycle 2 to characterize the unstable behavior of this NPP. Departing from test point 3 and performing a control rods movement, which is represented at the same way as in real nuclear power plants, the point PT_UPV is achieved. The reactor core has been modeled with 72 thermalhydraulic channels, 71 representing the active core and 1 for the core bypass. It has been chosen a mapping based in the lambda modes of the power with which the possibility of conditioning the oscillation pattern is avoided Moreover, it allows an acceptable accuracy without adding significant computational time. Once the simulation has been performed, the results are compared with the obtained using RELAP5-MOD3.3/ PARCS v2.7 coupled code. This comparison allows the validation of the TRACE/PARCS’ models and simulation procedure, as well as the capability of TRACE code to simulate instability scenarios.






09.09.2014 15:00 Poster session

Multiphysics - 406

 

A new coupling CFD/Monte Carlo neutron transport scheme, Application to a single fuel rod problem

Romain Henry, Iztok Tiselj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

romain.henry@ijs.si

 

Multi-physics and Multi-scale modelling is a task particularly challenging for the future. More precisely, the description of phenomena occurring in the core involves a coupling between neutronic and thermal hydraulics through the so called reactivity or thermal feedback. In water cooled reactor, thermal feedback and temperature coefficients of reactivity are important, when temperature change, microscopic cross section change impacting directly the neutron flux and so the thermal power which is an important parameter from an economic point of view. Other feedback such as moderator density effect or Doppler effect can also have a significant impact on core’s performance and inherent safety of the reactor.
Those reasons were naturally leading in the development of coupling method in order to predict correctly the behaviour of the core in normal operation and accidental situation. It exist two way to realise the coupling the first one is called internal coupling in which the code describing one discipline have a module able to solve equations from the other discipline. One classical example is the point kinetic model implemented in a thermal-hydraulics code. One should pay attention that the physical model of the module is a simple model obtain after approximations. The coupling can also be extern, each discipline is described by its own code and a subroutine allows communications between them for data exchange. The work presented here, focuses on the presentation of the coupling scheme developed between the Computational Fluid Dynamic (CFD) code ANSYS CFX and the Monte Carlo neutron transport code TRIPOLI. The scheme is tested on a single fuel rod problem derived from the OECD/NEA and US NRC PWR MOX/UO2 core transient Benchmark. Prediction capabilities of our Coupled system are compared with the mentioned benchmark results.






09.09.2014 15:00 Poster session

Probabilistic safety assessment - 503

 

Overview of Fire PSA and supporting software

Peter Simurka

VUJE, a.s., Okružná 5, 918 64 Trnava, Slovakia

peter.simurka@vuje.sk

 

Continually increasing requirements on nowadays full scope PSA L1 and L2 as whole, which is multiplied by importance of specific data for all modes of operation of nuclear power plant, highlight role of input data used in PSA quantification process. This fact also emphasizes the role of capability to process all necessary information to analyze all nuclear plant modes by appropriate way.
Even if above-mentioned aspects are relevant for all parts of nowadays PSAs, their importance is critical for internal hazards including specific fire analysis (internal fire analysis constitutes one of the most challenging PSA tasks).Application of tailored information system forms one of the ways to speed up analyzing process, enhances manageability and maintainability of particular PSA projects and provides effective reporting mean to document process of work as well as traceable and human readable documentation for customers. This paper provides brief overview of VUJE approach and experience in this area. The paper introduces general purpose of database developed for fire PSA. Paper explains as this basic data source is enhanced by adding several relatively independent tiers to employ all common data for fire PSA purpose. Paper also briefly introduces capability of such system to generate integrated documentation covering all stages of fire analyses, covering all screening stages of fire analysis as well as future plans to enhance this part of work in such a way to be capable to build automatic interface between PSA model and fire database to enable PSA model parameters automatic updating and expansion of fires in combinations of initiating events (for example Fire and seismic event).The final section brings a brief overview of the results of the fire PSA completion of nuclear power plant Mochovce 3 and 4 units, focused on the most significant contributors to the CDF.






09.09.2014 15:00 Poster session

Probabilistic safety assessment - 505

 

PSA analyses of mitigation strategies for extreme external events

Blaže Gjorgiev, Andrija Volkanovski

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

blaze.gjorgiev@ijs.si

 

After the Fukushima nuclear accident nuclear regulatory bodies and nuclear industry in different countries are developing programs in order to strengthen the capability of their nuclear power plants to deal with extreme external events. Recently some of these programs have been discussed and some mitigation strategies have been proposed. In this paper few systems for water injection during severe events are proposed and analyzed. Portable and/or fixed equipment is considered as a part of each system. The initial assumption for each system is that the installed equipment is capable to survive extreme external events. Each system is modeled using the fault tree technique. Probabilistic safety assessment is performed. Comparison is made between the systems. The obtained results are analyzed and discussed.






09.09.2014 15:00 Poster session

Probabilistic safety assessment - 506

 

Assessment of Loss of Offsite Power Initiating Event Frequency

Marko Čepin

Fakulteta za elektrotehniko, Tržaška cesta 25, 1000 Ljubljana, Slovenia

marko.cepin@fe.uni-lj.si

 

The initiating event known under the term loss of offsite power is one of dominant risk contributors in many nuclear power plants. Its event tree with consequent end states and its related event tree station blackout with its consequent end states represent the connection between power system reliability and nuclear power plant safety. The methods for assessing initiating event loss of offsite power frequency are reviewed. The frequency is assessed and the results of the current plant status and power system are compared to the plant status and power system status from years ago. The assessment is evaluated considering the data collected over the years of plant operation. The new features of electric power system include increasingly used distributed sources of electric energy with power dependency on weather parameters. The improvements of the system configurations include new power lines, new hydro power plants in the vicinity of nuclear power plant, improved switchyards connecting nuclear power plant with electric power system.






09.09.2014 15:00 Poster session

Reactor physics - 607

 

Modeling of PWR Biological Shield Boration Using SCALE6.1 Hybrid Shielding Methodology

Mario Matijević, Dubravko Pevec, Krešimir Trontl

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

mario.matijevic@fer.hr

 

The PWR biological shield concrete activation is analyzed using SCALE6.1 hybrid deterministic-stochastic shielding methodology. Since accurate geometry representation is a paramount step for this type of shielding calculation involving heterogeneous three dimensional regions (reactor core, thermal shield, downcomer, pressure vessel, cavity, and shield), the Monte Carlo method is the method of choice. A detailed model of a combinatorial geometry, materials and characteristics of a typical PWR reactor was based on best available input data. The sources of ionizing radiation included fission neutrons and photons originating from HBR-2 benchmark critical core. The activation reduction of reactor biological shield materials was investigated via concrete boration, which is applied when shielding from ionizing neutron radiation is especially important. The concrete boration is common method for activation reduction (i.e. local thermal stress lowering) in nuclear shields since the 10B has especially high thermal neutron cross section which effectively captures thermalized neutrons and decreases their (n,γ) reaction in structural materials. The reduction of secondary gamma emissions from radiative capture of thermalized neutrons in biological shield has been examined via concrete boration with natural boron (B) and boron carbide (B4C). The possibility of neutron flux mitigation in a biological shield as well as impurity isotopes (59Co, 151Eu and 153Eu) activity decrease was especially explored. Satisfactory boron concentration which leads to saturation of neutron flux attenuation has been proposed. The activation levels of impurities in concrete, activated above the limit of IAEA clearance for the free release limit, have been estimated. The obtained results showed that activation over IAEA limits is mostly present in a thin layer of borated shield facing the critical reactor core in cavity, while bulk of the shield is activated below IAEA threshold. The saturation of the neutron flux attenuation was clearly demonstrated for several boron concentrations so even small amounts of boron in biological shield drastically benefit in overall flux reduction.






09.09.2014 15:00 Poster session

Reactor physics - 608

 

Uncertainty Analysis in Reactor Physics Modeling in the Framework of UAM-Benchmark Using SERPENT and SCALE-6.2 for Transport Calculations and TSUNAMI and SAMPLER for Cross Section Perturbation

Antonia Labarile1, Nicolás Olmo2, T. Barrachina3, Agustin Abarca3, Rafael Miró4, Gumersindo Verdú3

Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM) 5K Build- Universidad Politecnica de Valencia, Camino de Vera, s/n, 46022 Valencia, Spain1

Instituto de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Spain2

Polytechnic University of Valencia, Department of Chemical and Nuclear Engineering, Camí de Vera sn, 46022 Valencia, Spain3

Universitat Politecnica de Catalunya, C. Jordi Girona, 31, 08034 Barcelona, Spain4

 

gverdu@iqn.upv.es

 

In recent years, there has been an increasing demand from nuclear safety research for best estimate predictions to be provided with their confidence bounds. That is why it is needed a deep knowledge on Sensitivity and Uncertainty Analysis of Models. The present study, involved in the framework of international research of NEA-OECD estimates confidence bounds of results of simulation and performed sensitivity analysis and uncertainty in real cases of light water reactors.
The proposed technical approach is to establish a benchmarks for LWR modeling, bases of well-defined problems of boiling water reactors (BWR) and pressurized water reactors (PWR), from input specifications and reference experimental data. The objective is to determine and quantify the uncertainty in all steps calculation and propagate uncertainty system in the LWR whole. Calculations are carried out using the programs SCALE-6.2beta2 (TRITON/NEWT modules) as well as Monte Carlo-SERPENT-1.1.19 code, for transport calculations.
The propagation of uncertainties of cross sections for a PWR 15×15 fuel element (with and without control rod) and a BWR 7×7, in two different configurations, and two different states, Hot Full Power (HFP) and Hot Zero Power (HZP), has been performed using the TSUNAMI modules, which use the Generalized Perturbation Theory (GPT) and SAMPLER, which makes use of stochastic sampling techniques of cross sections, fission and decay mechanisms, and geometry. Perturbed values for flux in each energy group are applied and finally the sensitivity coefficients are calculated to discriminate the most sensitive and influential parameters in results of the keff and macroscopic and microscopic cross sections. The results obtained and validated are compared with references results and similar studies presented in the exercise I-1 (Cell Physics) and I-2 (Lattice Physics) of Benchmark UAM.






09.09.2014 15:00 Poster session

Reactor physics - 609

 

Windows interface environment XSUN-2013 for NEA transport and sensitivity-uncertainty computer codes TRANSX-2, PARTISN and SUSD3D

Slavko Slavič, Ivan Aleksander Kodeli, Vladimir Radulović

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

ivan.kodeli@ijs.si

 

A Windows interface XSUN-2013 facilitating the deterministic radiation transport and cross-section sensitivity-uncertainty calculation was developed sponsored by the OECD/NAE. The package was developed to help users in the preparation of input cards, rapid modification and execution of the complete chain of codes including TRANSX, ARTISN and SUSD33D. It allows a user-friendly viewing of results obtained from PARTISN and SUSD3D programs. XSUN can produce 2D color schemes of PARTISN geometries (e.g. x-y, r-Theta, ...) and different 3D plots (e.g. neutron flux distributions).






09.09.2014 15:00 Poster session

Reactor physics - 610

 

Evaluation of the Point Kinetic Model of NEK in APROS

Tadeja Polach1, Ivica Bašić2, Luka Štrubelj3

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia1

APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia2

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia3

tadeja.polach@zel-en.si

 

The Krško Nuclear Power Plant – NEK in Slovenia has a two loop Westinghouse PWR nuclear steam supply system with 1994 MW thermal output power. A model of NEK is being built using APROS – Advanced PROcess Simulation environment [1].The aim of work presented in this paper is to build a computer model of the core and verify and validate it. The data used to describe the properties of the system modelled in APROS, were the data describing NEK and its operational properties after the uprating and the introduction of the 18-month cycle. Basis for data collections was NEK RELAP5\MOD3.3 Engineering Handbook and the 23rd cycle [2]. A model describing the nuclear core was set up, which is composed of the values describing the physical parameters and the nuclear kinetics phenomena. In order to achieve that the APROS Reactor module was used as it includes the point kinetics mathematical model. The physical parameters in the steady state were analysed also at different level rod insertions. And the comparison of NEK responses and APROS core model responses was made.


REFERENCES
[1] APROS Nuclear documentation, VTT and Fortum, 2012.
[2] NEK RELAP5\MOD3.3 Nodalization Notebook, Krško, 2009.






09.09.2014 15:00 Poster session

Reactor physics - 611

 

Scale 6.1 evaluation of the effective cross sections of a LWR assembly-reflector model with application to the NEA TMI-1 PWR Benchmark

Federico Rocchi1, Marco Sumini2, Antonio Guglielmelli3

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy1

University of Bologna, Faculta di Ingehneria, Viale Risorgimeno 2, 40136 Bologna, Italy2

Nuclear Engineering Laboratory of Montecuccolino University of Bologna , via dei Colli 16, 40136 Bologna, Italy3

antonio.guglielmelli@unibo.it

 

The reactor three-dimensional neutronic calculations for the evaluation of the integral physics core parameters (e.g. reactivity, total power, radionuclides inventory as a function of burnup, etc.), despite the enormous increase in the computational power, are still performed with methods that solve the diffusion transport equation. This approach is primarily adopted in an industrial context where it’s necessary to perform sufficiently accurate calculations within affordable computational times. In a PWR, it is seen that this type of calculation is rather sensitive to the assessment of the necessary preliminary evaluation of the condensed and homogenized effective cross section library of the reflector zone. In fact, this zone, which reflects neutrons, acts as a thermal neutron absorber, and causes a high neutron flux gradient at the core/baffle interface, is characterized by a complex geometrical structure; nevertheless, to date only a few studies have been published about approaches and methods for a rigorous calculation of reflector constants and effective cross-sections. For this reason, using the methodology based on the T-NEWT control module of SCALE 6.1, a transport calculation for the evaluation of the effective, condensed, two-group cross sections for the 2-D fuel pin and assembly-reflector geometry of the TMI-1 PWR given in NEA/NSC/DOC (2013)7 benchmark is performed [1]. In detail, the T-NEWT sequence used consists of two steps: the cross sections processing and the transport calculation. The first step has been realized with the functional sequence CENTRM that assesses the cross sections in the non-resolved range with the BONAMI module, and with the CENTRM module in in the resolved range. The functional sequence CENTRM is the most rigorous sequence available in SCALE for the cross section processing. The second step - transport calculations - has been performed with the transport, two-dimensional, discrete ordinate module NEWT (New Esc-Based Weighting Transport) that uses the ESC (Extended Step Characteristic) approach. The cross section library used is the v7-238 one based on the ENDF/B-VII (Release 0) with 238 groups (148 fast and 90 thermal); the cross sections, after the transport calculation, are collapsed to two groups (thermal and fast) and homogenized for the two reflector and assembly zones. The diffusion coefficients have been calculated as one third of the inverse of the transport cross section. The calculation of the effective two-groups cross sections of the reflector zone is performed both for a homogenized geometry and for the exact 2-D one. The effect on the final results of a variation of the boron concentration in the moderator zones is also investigated. The outcomes of the reference case are then compared with those presented in the available literature for the same benchmark problem and obtained with the Montecarlo code MCNP5 and the deterministic codes WIMSD5 and DRAGON. The differences in the numerical values obtained for a few two-groups cross sections obtained with the different codes are discussed.


[1] Nuclear Energy Agency, Nuclear Science Committee “Benchmarks for uncertainty analysis in modelling (UAM) for the design, operation and safety analysis of LWRs”. Volume I: Specification and Support Data for Neutronics Cases (Phase I), Version 2.1 (Final specification); K. Ivanov, M. Avramova, S. Kamerow I. Kodeli, E. Sartori, E. Ivanov, O. Cabellos; 30-May-2013.






09.09.2014 15:00 Poster session

Reactor physics - 612

 

Reactor Physics Analysis of the Krško NPP by JSI and Westinghouse

Marjan Kromar1, Fausto Franceschini2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Westinghouse, Rue Montoyer 10, 1000 Bruxelles, Belgium2

marjan.kromar@ijs.si

 

Jožef Stefan Institute (JSI) and Westinghouse have performed reactor physics analysis of the Krško NPP using their respective core simulator packages, e.g. CORD-2 and NEXUS/ANC 9. The CORD-2 system, developed by the Reactor Physics Department of the Jožef Stefan Institute, is intended for core design calculations of PWRs. It consists of several neutronic codes, utility codes and libraries and is based on the WIMS-D5 lattice code. It includes advanced features for cross-section homogenization and a simple thermohydraulics module so that thermal feedback can be taken into account. The package has been validated for the nuclear design calculations of PWR cores and has been used for the verification of the NPP Krško reload cores since 1990. NEXUS/ANC 9 is the Westinghouse core design system, used for the analysis of commercial PWRs including the AP1000® PWR. The NEXUS code system generates once-through cross-section data for ANC 9, the latest version of the Westinghouse core simulator. This paper shows the performance of each core simulator to predict the plant physics behavior, specifically analyzing the comparison vs. measurements for the critical soluble boron concentration and startup physics tests for a set of representative operating cycles. The results show satisfactory performance from both codes and their adequacy to support the core design and fuel loading optimization for the Krško NPP.






09.09.2014 15:00 Poster session

Reactor physics - 613

 

Fluence Rate Benchmarking of the Stochastic Neutronic Code ANET

Thalia Xenofontos1, Gregory Kyriakos Delipei1, Panayiota Savva1, Melpomeni Varvayanni1, Nicolas Catsaros1, Jacques Mailliard2, Jorge Silva3

National Centre for Scientific Research “Demokritos”, Institute of Nuclear & Radiological Sciences & Technology, Energy & Safety, Nuclear Research Reactor Laboratory, Agia Paraskevi Attikis, P.O.Box 60037, 153 10 Athens, Greece1

Institut du Développement et des Ressources en Informatique Scientifique CNRS, Orsay, France2

Université Pierre et Marie Curie, 75005 Paris, France2

thalia.xenofontos@ipta.demokritos.gr

 

ANET is a new stochastic neutronics code which is being developed based on the high energy physics code GEANT of CERN, for simulating both GEN II/III reactors as well as innovative nuclear reactor designs. ANET has already been successfully tested with respect to criticality computations, while in this work its reliability in computing neutron fluence rates is examined in the framework of the code benchmarking and validation continuation. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 was considered appropriate for the present study, since its core is supplied with purely fresh fuel while fluence rate measurements are available at various characteristic core positions. Following the measurements protocol, thermal, epithermal and fast neutron fluence rates were computed in a 15 cm segment immediately below the fuel mid-height in a critical core. ANET computations were compared with measurements as well as with corresponding results obtained by three different codes, including both deterministic (XSDRN/CITATION) and stochastic (TRIPOLI, MCNP). The comparisons were performed in representative core positions, i.e. standard fuel assemblies, dummy (non-fueled) assemblies, beryllium reflectors, thermal column and free grid positions close to the core. The obtained results show that ANET is capable of satisfactorily simulating the neutron fluence rate in various segments of a Material Testing Reactor, concerning the whole neutron energy spectrum.






09.09.2014 15:00 Poster session

Reactor physics - 614

 

VENUS-2 MOX Core Benchmark Analysis using MCNP5 1.40 with JEFF 3.1.2 and ENDF/B-VII.0 nuclear data sets

Gregory-Kyriakos Delipei, Panayiota Savva, Melpomeni Varvayanni, Nicolas Catsaros

National Centre for Scientific Research “Demokritos”, Institute of Nuclear & Radiological Sciences & Technology, Energy & Safety, Nuclear Research Reactor Laboratory, P.O.Box 60228, 153 10 AGHIA PARASKIEVI, ATTIKIS, Greece

delipei.gregory@hotmail.com

 

Commercial nuclear power plants produce an excess of plutonium during their operation that could be utilized as mixed-oxide (MOX) uranium and plutonium fuel in existing or advanced nuclear reactors. This creates the requirement of reliably simulating such MOX fueled reactor cores and, at the same time, the necessity to validate neutronic codes and nuclear data libraries for such applications. To this purpose, the OECD/NEA launched a widely used blind international MOX three-dimensional benchmark of the VENUS-2 core. The main objective is to calculate the axial fission rates of six fuel pins and to compare the results with measurements and other computations. Besides the above parameters, additional results are provided in the framework of the VENUS-2 benchmark, concerning multiplication factor and reaction rates for each type of fuel pin separately. In this paper, the benchmarking results of the MCNP5 Monte Carlo code based on the 3-D VENUS-2 MOX problem are reported. 3-D MCNP5 calculations were performed using the JEFF 3.1.2 and ENDF/B-VII.0 nuclear data sets. The computational results are compared with measured data, as well as with the results of other benchmark analyzers using different Code/library combinations. In general, the MCNP5 results are in agreement with both the experimental data and other computational analyses of the VENUS-2 benchmark. The multiplication factor calculations show an excellent agreement with both the experimental value and the average of the Monte Carlo based codes. As regards the absorption and fission rates, in most calculations the discrepancies with measurements are found in a reasonable range. Only in some specific positions and for certain isotopes and energy groups the discrepancies become rather high, however similar to corresponding calculations of other analyses of the benchmark, which are attributed to the utilized libraries.






09.09.2014 15:00 Poster session

Research reactors - 704

 

MARIA Research Reactor Thermalhydraulic Calculations With RELAP5 Code.

Eleonora Klara Skrzypek, Krzysztof Gomulski, Maciej Skrzypek

National Centre for Nuclear Research, ul. Andrzeja Sołtana 7, Otwock-Świerk, Poland

eleonora.skrzypek@ncbj.gov.pl

 

The research reactor MARIA is channel type reactor situated in the National Center for Nuclear Research in Swierk and is an important supplier of the pharmaceutical radioisotopes for various uses. Apart from isotopes production, the research reactor it is used to investigate structural materials, which are operating in the irradiated environment and used in nuclear industry.
With the conversion from the HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) and due to safety requirements of International Atomic Energy Agency - IAEA, there was a need for coolant circuit design modifications. The necessity of the modifications were caused by different geometry of LEU fuel comparing to the old design. The HEU fuel was designed as a six plates formed into the shape of cylinder, while the heat transfer area in active region for LEU design was decreased by one cylindrical pipe. Changes in primary circuit resulted in the requirement of the new model of the reactor thermalhydraulic system, for the purposes of the safety analysis report. The unique design of the reactor core demands the experience in modeling using T-H code, but with the acknowledgement of some limitations of the code, which have to be omitted by the user.
Due to the multiplicity and geometrical similarities of LEU (and HEU respectively), some channels are identical and can be merged into zones (peripheral and central), representing different power distributions. This decreases the time needed for the calculations of the scenarios and enables the flexibility to create various core configurations.
The use of programming skills allowed to decrease the amount of actions taken by the code user while defining channels in the core. The automation is made by connections of the pre-modelled channels into zones and finally into core, by the use of dedicated program written in Python.
Created model of research reactor will help to predict possible scenarios during operation and identified accident sequences. The presented results in the paper describe the progression of threatening scenarios: leakage of the coolant from the one of the channel circuits and loss of flow accident.






09.09.2014 15:00 Poster session

Research reactors - 705

 

Calculations to support design and installation of the DT Converter in TRIGA Mark-II

Aljaž Kolšek1, Luka Snoj1, Patrick Sauvan2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Universidad Nacional de Educación a Distancia Ingeniería Energética, Juan del Rosal, 12, 28040 Madrid, Spain2

aljaz.kolsek@gmail.com

 

A deuterium-tritium based thermal-to-fast neutron converter uses the reaction 6Li(n,t)4He to produce tritium, which then fuses with deuterium in the target material and produces a 14.1 MeV neutron. These neutrons are then used for activation studies of fusion relevant materials and measurement of cross sections at 14.1 MeV. The main advantage of such converter is that one can get a 14 MeV component in neutron spectrum relatively easily in an existing nuclear facility.
The Monte Carlo transport code MCNPX and its extension MCUNED, capable of light ion transport using the external evaluated nuclear data libraries, are used to computationally support the design and optimization of the deuterium-tritium converter inside the TRIGA Mark-II research reactor at the Jožef Stefan Institute. Calculations of fast neutron flux and energy spectra inside the converter were performed in order to optimize the device to achieve the highest 14.1 MeV neutron yield with as little fission neutron background as possible. Results also show that the thermal column is the most appropriate irradiation position for the DT converter in the reactor. Additional calculations were done to perform safety assessment of the device, such as analysis of pressure buildup in the device, heating, tritium production, etc. From the potentially suitable active converter materials, 6LiD yields the most 14 MeV neutrons.
The fast neutron flux (10 MeV - 20 MeV) inside the thermal column is increased up to 10 times when using the DT converter.






09.09.2014 15:00 Poster session

Research reactors - 706

 

Validation of the neutron and gamma flux distributions in the JSI TRIGA reactor core

Gašper Žerovnik1, Tanja Kaiba2, Anže Jazbec3, Sebastjan Rupnik3, Loic Barbot4, Damien Fourmentel5, Luka Snoj1, Jean-François Villard5

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Fakulteta za matematiko in fiziko, Jadranska 19, 1111 Ljubljana, Slovenia2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia3

CEA-Cadarache DEN/DER/SPEX/LDCI, Bâtiment 238 - Piece 10, F13108 Saint-Paul-lez-Durance, France4

CEA France, CEN Saclay ORE/SRO, France5

gasper.zerovnik@ijs.si

 

Within the framework of the collaboration between JSI and CEA Cadarache, a series of on-line neutron and gamma flux distribution measurements in the JSI TRIGA reactor core was performed. CEA manufactured fission and ionization chambers were used in this experiment. Axial gamma flux and fission rate distributions were measured in special measuring positions between fuel element in, above and below the reactor core.
The measurements were compared with the Monte Carlo calculations using the JSI TRIGA MCNP5 model. The fission chamber response function in pulse mode operation was compared with the calculated fission rates, while the ionization chamber response in current mode was (after background subtraction) compared to the calculated gamma fluxes. Due to almost equal sensitivities to gamma rays, the fission minus ionization chamber response in current mode was also compared to the calculated fission rates. In general, the agreement is very good which provides additional proof of the quality of the computational model. New set of measurements with fission chambers operated in Campbell mode (fully selective response to neutrons) will be performed in 2014 and will also append the available measurement database.
Apart from the validation of the TRIGA computational model, the main purpose of the experimental campaign is to determine the fission chamber response as a function of control rod positioning, which will later serve as a basis for reactor thermal power measurements.






09.09.2014 15:00 Poster session

Research reactors - 707

 

Determination of Computational Bias for Sub-critical Configuration of VR-1

Jakub Lüley1, Stefan Cerba2, Ján Haščík1, Branislav Vrban1, Vladimir Nečas2, Jan Rataj3

Slovak University of Technology Faculty of Electrical Engineering and Information Technology Dep. of Nucl. Physics and Technology, Ilkovicova 3, 812 19 Bratislava, Slovakia1

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia2

Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, Brehova 7, CZ 180 00 Prague, Czech Republic3

jakub.luley@stuba.sk

 

Criticality safety analysis provided for Nuclear Power Plants in commissioning include investigations where it is required to prove deep sub-criticality of irradiated fuel pool storage, transport cask, spent fuel storage or shut-down core. In the Monte Carlo criticality calculations are in general used the volumetric or point neutron source dispersed in fuel volume. From the computational point of view, it is considered as conservative approach because in a real situations it is impossible that the chain reaction can begin homogenously in whole fuel volume. Otherwise, a neutron source uses in the experimental devices, like experimental reactor, is localized in a specific area, hence this configuration can serve as a special case for validation and verification of computational methodology utilized in sub-critical calculations.
This paper is describing sub-critical experiments performed on zero-power reactor VR-1 in Prague. Measurable states of the core sub-criticality was achieved by method of sub-critical multiplication. Multiplication factor or reactivity was determined by source multiplication method known as Greenspan Method. This method is simple and reliable. Requirement for this method is to know a slope of the reactor power increase from critical state after injection of the neutron source. Measurements were carried out for different power level and position of detectors. For each case, the model for MCNP and SCALE was prepared. Comparison of experimental and calculated values will identify trends and reliance of computational bias to depth of sub-criticality. In special case it can define a requirements on standard benchmark experiment for PWRs. This paper also give a brief description of the experimental reactor VR-1 and equipment used within the analysis.






09.09.2014 15:00 Poster session

Research reactors - 708

 

Physical Analysis of Thermalization in Thermal Column Irradiation Position

Aljaž Kolšek, Luka Snoj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

aljaz.kolsek@gmail.com

 

The TRIGA Mark II research reactor at the Jožef Stefan Institute features several ex-core irradiation facilities that can be used for different applications. One of the irradiation positions is a rectangular port inside the Thermal Column (ThCol port), a graphite stack that extends from the graphite reflector to the outer concrete wall of the reactor and thermalizes the neutrons leaking from the reactor core. Due to the surrounding graphite, the port features a well-thermalized neutron flux having a thermal to fast flux ratios of up to 500 to 1.
In order to thoroughly understand the thermalization process in the ThCol, we analyzed the path of neutrons from the reactor to the irradiation point and importance of various components to the thermalization process. The MCNP6 Monte Carlo transport code and thoroughly validated model of the JSI TRIGA reactor were used to perform calculations. The ratio of thermal to fast neutron flux was calculated throughout the graphite stack and the relative changes due to the inserted heavy water for one of the applications were calculated. In addition the angular neutron flux inside the ThCol was estimated.
The results of the study will be used in planning and designing of irradiation devices that require highly thermalized neutron flux.






09.09.2014 15:00 Poster session

Research reactors - 709

 

Comparison of Measured and Calculated Reactivity Worth of Control Rods in a TRIGA Reactor

Vid Merljak1, Andrej Trkov2, Igor Lengar1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

International Atomic Energy Agency, Wagramerstr. 5, P.O.Box 100, A-1400 Vienna, Austria2

vid.merljak@ijs.si

 

At the TRIGA Mark II research reactor of the “Jožef Stefan” Institute the chain reaction is controlled by four control rods. Their integral and differential reactivity worth curves are commonly measured by the rod-swap and the rod-insertion methods. A comparison of the experimentally measured values with those obtained from numerical simulation by the Monte Carlo method is performed with differences analysed and uncertainties estimated. It was found that the simulation of the rod-swap method gives qualitatively and quantitatively adequate results, while the simulation of the rod-insertion method is less accurate. The deviation between the experiment and simulation for the latter method is found to be due to simplifications in the computational method used. Special care was devoted to the determination of the error in the current computational model of the reactor since experimental measurements revealed insufficient knowledge of the reactor’s control system.






09.09.2014 15:00 Poster session

Research reactors - 710

 

Kinetic Simulation of Control Rod Reactivity Worth Measurement in a TRIGA Reactor by the Rod-Drop Method

Vid Merljak1, Andrej Trkov2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

International Atomic Energy Agency, Wagramerstr. 5, P.O.Box 100, A-1400 Vienna, Austria2

vid.merljak@ijs.si

 

Reactivity worth of a control rod can be determined by various experimental methods. During such measurements by the rod-drop method or by the rod-insertion method kinetic effects play an important role. Difference between static and dynamic reactivity values may exceed ten percent. Thus one cannot properly determine the reactivity worth using only static computer simulations – kinetic simulation code is needed. In this paper preliminary results from such kinetic simulation of a rod-drop experiment with a newly updated deterministic computer code GNOMER are presented.






09.09.2014 15:00 Poster session

Research reactors - 711

 

Analysis of neutron flux and reaction rates of TRIGA research reactor using Monte Carlo code Serpent

Dušan Ćalić

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

dusan.calic@ijs.si

 

Serpent is a three- dimensional continuous –energy Monte Carlo reactor physics burnup calculation code used for various reactor simulations. One of the applications is the full- core modeling of research reactors. In this paper the computational results of the Serpent code against the experimental and MCNP results will be presented for the TRIGA research reactor. Detailed geometrical model of the reactor was constructed using Serpent code. This code has an advantage over the MCNP code because the burnup calculations can be performed. However in this paper the validation of the model will be performed by comparing the neutron flux and reaction rate distribution in the reactor core.






09.09.2014 15:00 Poster session

Radioactive waste and decommissioning - 808

 

Optimization of the Vrbina LILW Repository Preliminary Design

Boštjan Duhovnik1, Janja Špiler2

IBE, d.d., Hajdrihova 4, 1000 Ljubljana, Slovenia1

ARAO – Agencija za radioaktivne odpadke, Celovška cesta 182, 1000 Ljubljana, Slovenia2

bostjan.duhovnik@ibe.si

 

In 2009, a Preliminary Design of the Vrbina LILW Repository was prepared. The Design solutions followed the guidelines given in the Comparative Multilateral Study of Alternatives where the solution of disposal in below-ground silos was proved to be the most suitable one for the Vrbina site. According to the structures, systems and components extent the Design solutions also complied with the then applicable regulatory requirements and program documents from the field of radwaste management. The Preliminary Design represented one of the bases for adoption of the National Spatial Plan of the Repository at the end of 2009.
In order to reduce the costs of investment and operation, and increase technical feasibility and safety, an approach to optimizing the Design solutions was initiated immediately after the Preliminary Design had been completed. The opinions of the IAEA missions and several independent external reviewers as well as modified starting points for preparation of a new NEK decommissioning program revision were taken into account. At the same time, real-time knowledge in the field of safety analyses and conclusions regarding the required waste acceptance criteria were taken into account in the optimization procedure. The optimization was mainly concluded in 2010 and 2011 in form of several independent studies. The priority solutions indicated by the optimization procedure are mostly as follows: repository operation mode changing from a continuous to an occasional one with an standstill period; disposal of a revised-reduced LILW quantity in optimized disposal containers; solution of the disposal silos by abandoning the access shaft and the inspection gallery and by modified closing-up approaches after the end of operation; waste treatment for disposal in NPP Krško and waste transport by the existing and reconstructed public spaces. At the moment, optimization of the Repository non-disposal part is taking place. The optimization solutions depend on a precise definition of individual Repository activities implementation delimitation between ARAO, NPP Krško and other operators.
The optimized Design solutions of disposal and other plants and relating systems will be re-examined from the standpoint of safety while the optimized Preliminary Design relating to the disposal container and waste treatment for disposal will be re-examined from the standpoint of waste acceptance criteria for disposal. The re-examined optimized Design solutions will represent a basis for the next stages of the design documents implementation.






09.09.2014 15:00 Poster session

Radioactive waste and decommissioning - 810

 

The Solidification of Liquid Radioactive Waste – New Technique and Knowledge Applied in Institutional Waste Management in Slovenia

Marko Kostanjevec, Simona Sucic

ARAO – Agencija za radioaktivne odpadke, Celovška cesta 182, 1000 Ljubljana, Slovenia

simona.sucic@arao.si

 

There are many challenges in institutional radioactive waste management and especially in countries with small nuclear programs, where almost every waste stream is unique. In countries with small nuclear programs providers of institutional radioactive waste management have accepted different methods and developed various procedures for waste streams that are more frequent and those actions are preformed routinely. However, occasionally they are also faced with waste streams that do not meet the waste acceptance criteria for storage and that triggers the need to introduce a new method and procedure, which is often not easy.
Institutional radioactive waste management in Slovenia is handled by state public service which also includes the operation of the Central storage facility (CSF). Since the CSF is licensed for the storage of solid radioactive waste, wastes that are not in solid form must be treated prior to its acceptance into storage. Radioactive waste in liquid state represent clear example of waste that does not meet the acceptance criteria. However, liquefied radioactive waste in Slovenia is produced in small quantities and mostly coming from research activities in medicine. Also, this waste stream includes mixture of scintillation fluids, solvents, organic and aqueous solutions. Due to presence of tritium this waste belongs to the category of so called problematic waste stream. Another significant difficulty is related with the relatively incomplete information which is obtained from the waste producers about the composition and proportions of the individual waste components. On the implementation level, the main challenge for the performer of public service is to find an appropriate method of solidification, which will take into account the problematic waste stream, the aspect of waste amount and resources (human, economic and infrastructural). New technologies enabled treatment of complex waste streams and in Slovenian case the solidification by polymers constitutes a suitable treatment solution. However, technology is not everything and human factor can not be neglected. Competent and motivated workforce must be able to proficiently manage day-to-day uncertainties, always improving its ability to do so, and measuring the effect of an action, with feedback to adjust the next action, thus improving skills and learning by doing. Within this paper, benefits and lessons learned from solidification by polymers together with other practical issues from the institutional radioactive waste management are presented.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 907

 

Code-to-code comparison of the VVER 440/V213 containment response to the DBA

Ján Kubačka1, Peter Juriš1, Martin Gajdoš2

VUJE, a.s., Okružná 5, 918 64 Trnava, Slovakia1

Slovenské elektrárne a.s., NPP Mochovce, 935 39 Mochovce, Slovakia2

jan.kubacka@vuje.sk

 

Recently, the new model of VVER440/V213 containment has been developed for the APROS 5.11.08 code. The paper presents detailed description of the model validation in code to code comparison against MELCOR 1.8.5 model of the same containment, which is used standardly for diverse safety analyses, including licensing ones in the Slovakia. Overall five validation cases have been selected, all focused to the containment response to various initiating events.
As the MELCOR 1.8.5, unlike APROS, does not simulate flashing of the superheated water released into the containment atmosphere, it was expected that the results of APROS and MELCOR codes could significantly differ. To more specific evaluation of the impact of the flashing phenomenon, additional comparison of MELCOR 1.8.5 to the newest MELCOR 2.1 with built-in flashing model was performed, based on a simplified three-volume confinement model to postulated mass and energy source. As the MELCOR 2.1 is able to provide results utilizing both enabled and disabled flashing option, totally three code-specific variants were considered for additionally analyzed LOCA scenarios. The APROS containment model development project has been initiated in 2013 for Slovenské elektrárne, a.s., in cooperation of VTT and FORTUM (Finland) and VUJE (Slovakia). The project resulted in the full scope model of the VVER 440/V2013 containment capable to analyze large spectrum of normal, abnormal operational states and design basis accidents of the VVER440 plant. The developed containment model represents the state-of-art model for simulations of the containment response to any arbitrary DBA in Slovak republic.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 909

 

RELAP5 analysis of Krško nuclear power plant abnormal event from 2011

Andrej Prošek, Marko Matkovič

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

andrej.prosek@ijs.si

 

Measured plant data are of great importance for thermal-hydraulic system code validation, because they are full scale and have true geometry. Typically, real plant data are limited mostly to operational events. The purpose of this study was both validation and better understanding of the plant response to abnormal event, which occurred on 23 March 2011 in Krško nuclear power plant (NPP). The abnormal event was called “Reactor trip and actuation of safety injection system at the loss of external power“. The sequence of events started with spurious activation of 400 kV bus differential protection in the NPP’s 400 kV switchyard, followed by the disconnection of circuit breakers in the bays, resulting in loss of off-site load. Automatic stabilization at house load operation using the turbine control system and steam dump system was not successful to stabilize main steam line pressure. Safety injection actuation and reactor trip occurred on low pressure signal in steam generator 1 due to divergent oscillation of pressure in steam generators.
For the analysis the latest RELAP5/MOD3.3 Patch 04 best-estimate thermalhydraulic computer code and standard Krško NPP input deck for RELAP5 have been used. Loss of external load has been modeled with artificial turbine control as external perturbation, which is provided in the standard input deck. The turbine external power function was determined from measured reference temperature. Due to manual operation of main feedwater the measured feedwater flow was used as boundary condition. Steam flow was also measured, therefore two cases were performed: a) total steam flow as boundary condition; b) total steam flow modelled by steam dump system and turbine flow (by estimating turbine flow from available measured data because in the standard Krško NPP input deck the secondary side is modelled up to the turbine only).In the paper the results of the RELAP5 calculated plant response to abnormal event will be presented. Comparison of the calculated data with plant measured data suggests that RELAP5 code can accurately simulate such abnormal events. It will be also shown that performed calculation provides additional insights into the plant response, including how the pressure change influences the safety injection signal on low pressure in steam generator, which is compensated signal.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 910

 

Supporting Deterministic T-H Analyses for Level 1 PSA

Slavomir Bebjak1, Tomas Kliment2

VUJE, a.s., Okružná 5, 918 64 Trnava, Slovakia1

VÚJE Trnava Engineering, Design and Research Organization, Ltd., Okružna 5, 91864 Trnava, Slovakia2

slavomir.bebjak@vuje.sk

 

The main concern of PSA Level 1 is to identify the accidents in event trees which lead to core damage and to quantify their frequency of occurrence. The core damage occurs, if the success criteria of safety system are not fulfilled.
The success criterion of safety system is defined as minimum degree of required ability to ensure safety function (e.g. number of trains in operation). This criterion is evaluated for each particular scenario separately and the criterion is fulfilled when the core damage does not occur. Success criterion of safety systems evaluation is based on the assessment of peak cladding temperature. Fulfilment / Failure of success criterion of safety systems applied for PSA Level 1 is assessed on the basis of results of T-H analyses.
This paper introduces deterministic T-H analyses that have been elaborated as a support for PSA Level 1 of Mochovce 3 and 4 units. Safety systems configuration in each analyzed scenario was defined on the basis of PSA Level 1 requirements. Paper provides overview of VUJE approach and experience in the area of T-H analyses performed by the RELAP5 computer code. Paper introduces the methodology considered within calculation, explains the choice of applied approach and describes initial and boundary conditions as well as operator action. Finally, paper shows the results of analyzed scenarios that were performed for full power and for shutdown reactor states and provides evaluation of success criterion evaluation in deterministic analysis.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 911

 

Steam Generator Modeling in Apros

Klemen Debelak1, Tadeja Polach2, Ivica Bašić3, Luka Štrubelj1

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia1

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia2

APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia3

klemen.debelak@gen-energija.si

 

GEN-energija is developing best estimate model of nuclear power plant in Krško (NEK) with simulation code APROS. A very important part of this model is Steam Generator (SG), which is responsible for heat transfer from primary to secondary system. NEK has two SG with heat transfer capacity of 1000 MW each. Data for model description was taken from NEK RELAP5\mod3.3 Engineering Handbook and manufacturer design data summarized in the Siemens-Framatome Reports.
The goal was to create a model that would give accurate results in simulation of the plant steady state operations and various transients. The basis for SG in APROS is module called Advanced Steam Generator, which was modified with additional heat structures and volumes to achieve proper performance. In order to maintain volumes and heights, another volume in form of node and branch was created to represent a part of steam dome. All the material properties were considered in heat structures in a form that was consisted with advanced Steam Generator module. The model also includes regulation of level and pressure in the steam generator.
The APROS SG model was verified comparing the following data: volumes, heights, volume vs height and mass of heat structures. Validation of the separated SG model on design initial and boundary conditions (hi T-average, 0% tube plugging, etc.) shows some minor deviation due to limitation of model, but generally the results were good and within acceptable tolerances. The SG model was then integrated with entire NEK model. After the adjustments of connected systems and solving of issues, the performance of SG was tested. Also the limitations of the model were recognized and the validation of recirculation inside SG was made. The results (mass flows, temperatures and heat transfer) were compared with the results of RELAP5/mod3.3 and NEK data for 23rd cycle.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 912

 

Modeling of Low Pressure Injection System of NEK in computer code APROS

Jure Jazbinšek1, Ivica Bašić2, Luka Štrubelj3

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia1

APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia2

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia3

jure.jazbinsek@zel-en.si

 

The Residual Heat Removal (RH) System, functions to remove heat from the core and Reactor Coolant System (RC) during plant cooldown and refueling operations. In addition, RH system serves as a part of Emergency Core Cooling System (ECCS) providing a Low Pressure Injection (LPI) to the cold legs of RCS and assures long-term heat removal following a Loss of Coolant Accident (LOCA).
LPIS model was built in program Apros based on the existing NEK RELAP5 engineering handbook and associated nodalization. Verification of LPIS model was performed taking into account the data from the mentioned RELAP engineering handbook and associated NEK documentation. Some parts of the model like RHR Pump settings take into account measured parameters, obtained during system surveillance testing in NEK. Surveillance test results also served for the validation of LPIS model, where model boundary condition parameters were set accordingly to the NEK test precautions and limitation.
After the LPIS model was constructed and verified, simulations were run and only minor adjustments of some model components were made, to adjust difference between models and lack of some parameters needed. The flow and pump head in different modes of operation are the main results of the simulation. Comparison of simulation results and NEK measurement data gives great results, and proves suitability of APROS code for these kinds of simulations. Validated LPIS system was at the end adjusted to suitable (steady state) parameters and will be later integrated into NEK complete model of reactor coolant system and used for simulations of various transients.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 914

 

Capability of TRANSURANUS code to reproduce LOCA integral test results

Oleksandr Lisovyy

University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy

o.lisovyy@gmail.com

 

The present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis loss-of-coolant accident (LOCA) scenario. The analysis refers to LOCA integral test results of Argonne National Laboratory (ANL) performed with special focus on embrittlement effect of various cladding alloys. Four tests were conducted with high-burnup fuel rod samples from Limerick fuel rods: ICL#1 test was conducted in argon up through ballooning and burst to provide data on ballooning strain and minimum wall thickness; ICL#2 was conducted in steam with a 300-s hold time at 1204oC, cooling at 3oC/s to 800oC, and slower cooling to RT; ICL#3 was conducted under the same conditions as ICL#2 through cooling to 800oC, followed by quench and rapid cooling from 800oC to 460oC; and ICL#4 was conducted through the complete LOCA sequence with quench and rapid cooling from 800oC to 100oC.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 915

 

TRACE Three Dimensional Pressure Vessel Model Development for Krško NPP

Ovidiu-Adrian Berar, Andrej Prošek, Borut Mavko

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

adrian.berar@ijs.si

 

The transient and accident analysis of reactor systems has been one of the main activities in nuclear research. In the specific case of some Design Basis Accidents the analysis requires multidimensional prediction of the thermal-hydraulic conditions in order to realistically simulate the accident. The TRAC/RELAP Advanced Computational Engine (TRACE) is an advanced, best-estimate reactor systems code developed by the U.S. Nuclear Regulatory Commission for analyzing light water reactors, and it has the capability of solving the fluid-dynamics equations in three dimensional space using the specialized VESSEL component. The objective of this study was to develop a TRACE three-dimensional model of the Krško NPP Reactor Pressure Vessel (RPV) needed for coupled thermal-hydraulic and neutronic calculations. For the Krško RPV TRACE model development the VESSEL component has been used employing the cylindrical geometry model discretization in axial levels, radial rings, and azimuthal sectors. Specific physical geometry of RPV such as the downcomer, lower plenum, core section, and upper plenum were taken into account for the three-dimensional RPV model. The fuel assemblies are modeled using Heat-Structures connected to the VESSEL component. The three-dimensional model of the RPV has been separately tested first and then it has been incorporated into a one-dimensional TRACE input model of Krško NPP, which has been obtained from an existing RELAP5 input deck through semiautomatic conversion and manual input. The TRACE code calculation of the NPP model response to plant nominal conditions using three-dimensional RPV will be presented. In addition, a sensitivity study assessing the influence of the RPV nodalization schemes (downcomer, lower plenum and core region) on the predicted steady-state conditions will be presented and discussed.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 916

 

Simulation of flooding waves in vertical air-water churn flow using Neptune_CFD 2.2.0 code

Matej Tekavčič, Boštjan Končar, Ivo Kljenak

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

matej.tekavcic@ijs.si

 

In safety analyses of nuclear reactors, many complex multiphase flow phenomena that occur during thermal-hydraulic processes could be modelled using computational fluid dynamics (CFD) approach. Flooding or counter-current flow limitation is one of such phenomena that are of particular interest for safety analyses of the loss-of-coolant accident in pressurized water reactor where steam-water flows are encountered. After the reflux condenser mode of cooling is established during such an accident, the upward flow of steam in the central region of a vertical pipe can limit the downward flow of the water film on the pipe wall. Flooding occurs when the liquid film flow reverses and cannot penetrate further downwards into the reactor primary system. Even for the simplest cases, the prediction of onset of flooding conditions is still very uncertain and more thorough understanding of the triggering mechanisms is needed. Numerical models using accurate local interface tracking methods can face that challenge. However, they should be supported by reliable experiments with local measured data.
The focus of this paper is the simulation of an isothermal counter-current air-water flow in the churn flow regime of a vertical pipe, which is an example of such flows. The churn flow regime can be viewed as a transitional regime between slug flow and annular flow and is often related to the onset of flooding mechanism. Large waves of liquid travelling upwards can typically be observed in the churn flow regime.
Neptune_CFD 2.2.0 is the latest version of general purpose computational fluid dynamics software developed by EDF and CEA that is available at JSI within the European NURESAFE project. The code is specifically well suited for modelling of multiphase flows that occur in nuclear reactor systems. In the latest version, surface tension model and interface sharpening algorithm were added.
The above-mentioned new models will be used for a transient simulation of isothermal air-water churn flow in a vertical pipe. The gas and liquid phases are considered immiscible and incompressible with no mass transfer between them. Interfacial momentum transfer is modelled using the Separated Phases and Large Interface Model for comparison. Turbulence is modeled using the SSG (Speziale-Sarkar-Gatski) Reynolds Stress model. The numerical domain consists of an axis-symmetric wedge with the porous wall inlet region representing a vertical pipe experimental test section. A constant fluid mass flow rates for gas and liquid inlets are taken from the experimental data (from the literature).Simulation results are compared with the reported experimental data for churn flow regime in a vertical pipe under different flow conditions. In the experiment, the frequency and velocity of the waves are reported. In the results of the present numerical simulation, these flooding wave properties are obtained from the calculated evolution of liquid (or gas) volume fraction with tracking of the interphase surface. Simulations are performed at different mesh densities. For each flow condition, a cascade of several waves is simulated to evaluate uncertainties of calculated wave properties.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 917

 

Dynamic Simulation of NEK Reactor Coolant Pump with Best Estimate Full Scale Model in APROS

Dejan Slovenc1, Samo Fürst2, Ivica Bašić3, Luka Štrubelj2

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia1

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia2

APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia3

dejan.slovenc@zel-en.si

 

This paper presents the simulation of the Krško NPP (NEK) Reactor Coolant Pump (RCP)) start-up test by means of independent off-site power source and RCP coast down test on loss of offsite power signal. The simulation model was developed in APROS - Advanced PROcess Simulation environment. APROS is a multifunctional software for modelling and dynamic simulation of various physical processes in power plants which enabled a coupled simulation of thermal-hydraulics, electrical and regulation systems including reactor kinetics.
The goal was to create a full scale simulation model which enabled simulation of the transient process occurring at the RCP start-up and coast down tests. The entire model consists of NEK primary system and a part of secondary systems. RCP pump model, as a part of reactor coolant loop, is supplied from electric power distribution systems connected to external power grid. The RCP moment slip and current slip characteristics were performed and compared with RCP manufacturer and vendor data to confirm validity of the APROS pump model. It is shown that, by using the available characteristic data (no-load, blocked rotor and coast down tests of the pump as well as the inputs obtained from the motor data sheet, accurate dynamic simulation of RCP start-up is possible. Furthermore, paper discusses the simulation of RCP motor start-up current transient and RCP coast down of primary mass flow after RCP trip.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 918

 

Upper plenum temperature calculations: comparison of TRANSURANUS with a 2–D model under steady–state conditions

 

Rolando Calabrese1, Arndt Schubert2, Paul Van Uffelen2, Luka Vlahovic3, Csaba Gyori4

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy1

European Commission Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz Platz 1, D-76344 Eggenstein-Leopoldshafen , Germany2

Istanbul Technical University, Energy Institute, Ayazaga Campus, 34469, Istanbul, Turkey3

NucleoCon, Slovakia4

rolando.calabrese@enea.it

 

The model for the evaluation of the upper plenum temperature (pressure) of the TRANSURANUS code is further developed to predict with better accuracy the role played by this parameter under steady–state and accident conditions (LOCA, RIA), rather than providing this parameter on input. At present, for the gas contained in the upper plenum volume, the code has a “low temperature” and a “high temperature” model. The former assumes that the plenum temperature coincides with the coolant temperature while the latter adopts a weighted value of the cladding inner temperature and the fuel central temperature evaluated in the uppermost zone of the fuel stack.
To refine the accuracy of the code predictions, the adopted strategy was to improve the description of the plenum volume sub–system by means of 2–D models. For this purpose, both a transient heat transfer model and a FEM model by using the commercial software COMSOL Multiphysics were developed. In addition, ENEA implemented in TRANSURANUS the FRAPCON–3 plenum temperature model for comparison with the two other models.
The paper presents the comparison of the code results for a PWR fuel pin under steady–state conditions. In addition, thanks to the capability of the code to account for the change in the plenum gas composition, preliminary calculations on the effect of fission gas release on the plenum temperature are also presented. Finally, we discuss the plenum spring gamma heating and the clad–to–coolant heat transfer coefficient under transient conditions in the light of the FK–1 test (NSRR) and the results published in the open literature.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 919

 

Investigation of coolant temperature fluctuations circulating in the primary circuit of VVER-440 reactors

 

Sándor Kiss, Sándor Lipcsei

Centre for Energy Research, Hungarian Academy of Sciences , Konkoly Thege M. út 29-33, H-1121, Hungary

lipcsei.sandor@energia.mta.hu

 

The neutron noise caused by core inlet temperature fluctuations carries diagnostic information, and it has been investigated since the seventies up to these days. Most of the investigations analyze neutron noise with known temperature fluctuations. In this paper properties of the temperature fluctuations in measurement data are investigated. In order to accomplish that we have used the transfer properties of the main components of the primary circuit and the circulation time of the coolant. We have identified the source of the temperature fluctuations are identified furthermore we have analyzed their propagation. The frequency dependences of fluctuation part ratios originated from the different sources are provided both for the hot and the cold leg.






09.09.2014 15:00 Poster session

Thermal-hydraulics - 920

 

Numerical simulations of a turbulent flow in a fuel assembly

Blaž Mikuž, Iztok Tiselj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

blaz.mikuz@ijs.si

 

Core of pressurized water reactor consist of fuel assemblies which contain thin and long fuel rods fixed in place by spacer grids. Cooling water is flowing along the rods and carrying away heat that is being released in a fission reactions inside fuel rods. The flow is so turbulent that the turbulence dominates all other flow phenomena including mixing and heat transfer. For this reason, the turbulence alone is put in a focus here without heat transfer. Our approach is using Computational Fluid Dynamics (CFD) and the results are assessed against measurements from MATiS-H experiment performed at KAERI in 2011.
The turbulent flow in the geometry of 5×5 fuel rod assembly is single-phase with Reynolds number about 50000. These conditions are less turbulent than the typical conditions in a real PWR core, but at the same time still too demanding for the present supercomputers to be able to calculate it with direct numerical simulation (DNS). Therefore other approaches, such as Reynolds Averaged Navier-Stokes simulations (RANS) and Large Eddy Simulations (LES), must be used for numerical simulations which introduce modelling of turbulence on certain scales. Both approaches have their limitations in accuracy, however LES simulation is expected to be more accurate since it employ less modelling. It is known that RANS models have their limitations in accuracy for certain flow situations, however they have shown their strength essentially for wall-bounded flows. For this reason the case here does not contain any flow defector, such as mixing vane.
In the present paper the capability of two low-Reynolds RANS turbulence models (isotropic k-omega SST and anisotropic v2f) are compared together with the Smagorinsky LES model. Since all of them aim to resolve the boundary layer profile, the mesh requirements are similar. Furthermore, the effect of different mesh elements (e.g. hexa, tetra and prisma) on the results will be investigated too.






09.09.2014 15:00 Poster session

Nuclear power plant operation - 1001

 

Krško NPP 2nd Periodic Safety Review Lessons Learned

Aleksandra Antolovič1, Bruno Glaser1, Ivica Bašić2, Ivan Vrbanić2

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia1

APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia2

aleksandra.antolovic@nek.si

 

The Periodic Safety Review is a systematic safety reassessment, which has attained international recognition as a primary means to assess the cumulative effects of plant aging and plant modifications, operating experience, technical developments and siting aspects. In the current, internationally accepted, safety philosophy periodic safety reviews (PSRs) are comprehensive reviews aimed at the verification that an operating NPP remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain an acceptable level of safety.
NPP Krško (NEK) undertook the 1st PSR following approximately twenty years of plant operation, starting in 2001. As a result, it was found that NEK could safely operate, as a minimum up to the completion of the next Periodic Safety Review.
The 2nd NEK PSR project has been initiated per the requirements of national Ionizing Radiation Protection and Nuclear Safety Act, considering also the requirements from the Regulation JV 9, as well as practical directives issued by Slovenian Nuclear Safety Administration (SNSA). The review methodology used in the 2nd PSR was defined in the associated NEKs Program, NEK ESD TR 09/09, based on the IAEA Draft Safety Guide DS426, approved by IAEA CSS Committee as of September 2011, determining all relevant elements of the second PSR that were agreed with and confirmed by the SNSA. The 2nd PSR project for Krško NPP covered all changes from current national and/or international safety standards/practices or plant design and operational arrangements and history from the 1st PSR to the so called freeze date: December 31st, 2010.
After the Fukushima accident NEK has immediately initiated relevant activities at the plant level and issued the addendum to the original scope of the 2nd PSR program, requiring review and evaluation of the NEK response to Fukushima accident, taking into consideration special NEK reports issued following Fukushima accident and, in particular, the decrees that were issued by SNSA as a response to Fukushima accident. The post-Fukushima activities performed in NEK until June 30th, 2012, were reviewed, as well as the documents produced by NEK, in response to above mentioned decisions and SNSA letter 3570-11/2011/9.
The 15 PSR safety factors have been selected in accordance to the IAEA Draft Safety Guide DS426. These 15 safety factors also correspond to the requirements of SNSA Regulation JV9, and Practical Directives from SNSA Letter No. 018-4/2006/52, and they are in accordance with WENRA principles postulated in the Reactor Safety Reference Levels document.
The methodological approaches used in the implementation of the 2nd NEK PSR were defined in the above mentioned program NEK ESD TR 09/09, compliant to the methodology given in the IAEA Draft document DS 426.
The 2nd NEK PSR activities were divided into two basic phases:
• Safety factors review phase;
• Ranking / prioritization, implementation plan development and global assessment phase.
In the review phase, the assessment of each of the relevant safety factors was focused on finding the differences in:
• Review elements defined by DS246, in comparison with previous definitions, or
• Modern standards / norms / requirements as compared to those evaluated in the 1st PSR, or
• Plant processes / programs / procedures / SSCs as assessed and evaluated in the 1st PSR.
The outcome of this step was a list of potential safety issues for each relevant safety factor. In accordance with the 2nd NEK PSR program, the overall process of PSR issues ranking was divided into two major steps:
1. Broad ranking, preliminary and final;
2. Detailed ranking.
Broad ranking was done in two phases: preliminary and final.
Paper summarizes the NEK 2nd PSR Lessons Learned.






09.09.2014 15:00 Poster session

Nuclear power plant operation - 1005

 

Historical Review of Exposure due to the C-14 Discharges from the Krško NPP

Milko Janez Križman1, Matjaž Stepišnik2

upokojenec, , Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

matjaz.stepisnik@ijs.si

 

Carbon-14 in the global environment is a result of continuous interactions of cosmic rays with the upper atmosphere, and also of the past atmospheric nuclear weapons tests. The third global source of C-14 is a result of the current operation of nuclear facilities (mostly reprocessing plants) all over the world. Locally, its environmental concentrations are slightly increased in the very vicinity of NPPs due to their gaseous and liquid releases.
In general C-14 is produced in NPPs mostly by neutron activation. Carbon-14 is also a ternary fission product, but the amount produced in this way is negligible. In case of the Krško NPP regular monitoring of the C-14 radionuclide in the plant environment was considered not being important regarding to other radionuclides from the beginning of the plant operation. For the first twenty years of operation its most important transfer pathway (ingestion) and related radiation exposure have not even been taken into account. A similar approach could be noticed in numerous NPPs.
Evaluation of public exposure due to gaseous releases of C-14 from the Krško NPP shows some essential changes of estimated dose levels of the public during the current plant lifetime. For a certain period in the twenties relevant dose levels were estimated at a flat rate, compared to other PWR plants. Over the recent decade, when the monitoring programme was expanded, the dose estimations have been based on measurements of selected vegetal samples taking into account the local reference values and the time of the power plant outage in relation to the growing season. Exposure assessment due to C-14 in liquid releases was based on discharged activity and on a campaign of grab sampling of the Sava River, performed in the recent year. These results undoubtedly indicated that the levels of exposure due to ingestion of fish could be dominant over the dose contributions of other radionuclides in liquid releases, such as tritium, fission and activation products.
The aim of this paper is to present the historical development of monitoring of C-14 discharges (gaseous and liquid) and the assessment of related radiation exposures due to these discharges. C-14 gaseous and liquid releases from nuclear installations are not the subject of operational limitations, but during normal operation their radiological impact usually prevails over the relevant impact of other discharged radionuclides.






09.09.2014 15:00 Poster session

Nuclear power plant operation - 1007

 

Post-Fukushima Assessment of the AP1000® Plant

Bryan Friedman1, Adam Malinowski2, Ernesto Boronat De Ferrater2

Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Twp 16066, USA-Pensylvania1

Westinghouse, Rue Montoyer 10, 1000 Bruxelles, Belgium2

friedmbn@westinghouse.com

 

The AP1000 plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, and safety with reduced plant costs. The AP1000 passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at the Fukushima Dai-ichi nuclear power station without compromising core and containment integrity. Without AC power, using passive safety technology, the AP1000 plant provides cooling for the core, containment and spent fuel pool for more than 3 days without the need for operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. With the re-supply of fuel oil the coping time may be extended for an indefinite time. Connections for a few, small, easily transportable components provide a diverse backup means of extending passive system operation after the first 3 days. As a result, the AP1000 design provides very robust protection of public safety and the utility investment.
Following the accident at the Fukushima Dai-ichi nuclear power station in Japan, several initiatives were launched worldwide to assess the lessons learned. These include, but are not limited to, the European Nuclear Safety Regulators Group (ENSREG) stress tests, the Office for Nuclear Regulation (ONR) Final Report, the International Atomic Energy Agency (IAEA) Expert Mission Report, and the U.S. NRC Near-Term Task Force Recommendations. The AP1000 design has been assessed against these initiatives and lessons learned.
The purpose of this paper is to describe:
• How the accident at the Fukushima Dai-ichi nuclear power station was evaluated and translated into conclusions and recommendations for nuclear power plants worldwide
• How the AP1000 plant was evaluated in light of the recommendations resulting from the various post-Fukushima assessments
• The key conclusions resulting from the post-Fukushima evaluation of the AP1000 design






09.09.2014 15:00 Poster session

Nuclear power plant operation - 1008

 

Quality Assurance Oversight of Important Plant Functional Areas

Igor Fifnja

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

igor.fifnja@nek.si

 

Krško NPP integrated management system is a set of interrelated and interacting elements that establishes policies and objectives and which enables those objectives to be achieved in a safe, efficient and effective manner. It integrates the principles of Quality Management, Quality Assurance and Quality Control and ensures that safety is not compromised by considering the implications of all actions. Safety is the paramount element in the management system, overriding all other demands. The integrated management system is defined in the Quality Assurance Plan (QD-1). The system is bringing together in a coherent manner all requirements for managing the organization.
The approach to oversight of important plant processes / plant functional areas was augmented during the last year. In accordance with the latest revision of internal plant procedure ADP-1.0.008, Nuclear Oversight Activities, the responsible QA engineers were nominated for the oversight of selected important plant processes, such as: operations, maintenance, modifications process, nuclear fuel, chemistry, radiological protection, industrial safety, environment protection, purchasing, training, emergency preparedness, and fire protection. Oversight of important plant functional areas and assessment of their effectiveness are performed on continuing basis. Various inputs are used for such assessment, including results of internal audits, open actions from previous audits, observations, inspections, walk downs, corrective action requests, status of procedures and programs, self-assessments, organizational changes during last period, and other. The results of important plant processes / functional area oversight and overall assessment of their effectiveness are gathered in the Quality and Nuclear Oversight (QNOD) annual report. The QNOD annual report is provided to the plant management and presented to the Quality Assurance (QA) Committee. The conclusions of the QA Committee and resulting corrective actions are tracked through the Plant Corrective Action Program. Based on the Krško NPP experience, the QA involvement in plant processes has fulfilled its important role and expectations in achieving overall quality goals. Krško NPP will continue to improve internal Quality Assurance processes and activities in the future. The most important objective of the entire organization – to ensure safe and efficient power plant operation, will continue to be the most important goal of Krško NPP QA Program.






09.09.2014 15:00

Nuclear power plant operation - 1010

 

Neutron noise analysis in the NPP Krško

Marjan Kromar1, Bojan Kurinčič2, Urban Simončič3

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia2

Institut "Jožef Stefan", Jamova 39, 1001 Ljubljana, Slovenia3

marjan.kromar@ijs.si

 

Due to the construction of a Pressurized Water Reactor and its core components, flow-induced mechanical vibrations can occur, which most likely lead to vibrations of other core internals. Those vibrations are difficult to detect but usually they are of minor importance for the reactor operation. However, noise diagnostics can be used to detect them. NPP Krško initiated the neutron noise monitoring program in which few measurements of the nuclear instrumentation signals (incore and excore detectors signals) are performed during the cycle. With the help of noise diagnostic methods, the vibrational frequencies of the reactor internals and fuel components as well as the axial distribution of the neutron noise are determined.






09.09.2014 15:00 Poster session

Nuclear power plant operation - 1011

 

Reactor Vessel Closure Head Replacement at Krško NPP

Franc Škrabec1, Frank Courtney2

NUMIP inženiring, montaža, vzdrževanje in proizvodnja d.o.o., Knezov štradon 92, 1000 Ljubljana, Slovenia1

Westinghouse Energy Systems U.S.A.,  USA2

franci.skrabec@numip.si

 

In response to the industrial experience with the Alloy 600 material and with plant life-time extension and potential further upgrades in mind, Krško NPP decided to replace the existing Reactor Vessel Closure Head (RVCH) during the 2012 outage. The head is a Westinghouse 2-loop design type, with 40 penetrations through the dome and 48 bores through the flange part for closure head studs.
Westinghouse Electric Company, as a main contractor, won the contract for the supply and installation of the RVCH and associated equipment in 2009. The contract also included scope for other significant upgrades including a Simplified Head Assembly (SHA) which is designed to expedite the refuelling process, neutron shielding, Control Rod Drive Mechanism components, and thermal insulation. Westinghouse then subcontracted NUMIP, as a leading Slovenian nuclear services provider, for a substantial portion of the on-site installation scope. This paper is jointly authored by NUMIP and Westinghouse and provides a high level description of every phase of the project, from planning and preparation all the way to the close-out after the successful head replacement. The paper further describes the importance of technical expertise, modern project management practices, experience in managing other involved partner companies, as well as of QA and QC aspects on the course of the project. Some key lessons learned are also presented.
Proposed presentation type: Verbal presentation preferred over the Poster Presentation
Proposed topic of the Paper: Nuclear Power Plant Operation






09.09.2014 15:00 Poster session

Nuclear power plant operation - 1013

 

RJET – 111: Autonomous Mobile Robot for Water Intake Channel Maintenance in Power Plant Cooling Systems

Matko Orsag1, Zdenko Kovacic1, Stjepan Flegaric2, Kristijan Brkic2, Borislav Balac2

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia1

Inteco Robotics, Jaruscica 7, 10000 Zagreb, Croatia2

matko.orsag@fer.hr

 

This paper presents a state-of-the-art, highly specialized, autonomous mobile robot for water intake channel maintenance, in power plant cooling systems. Thermal power plant in Plomin (TE Plomin), with a total capacity of 330 MW cools down using sea water which is rerouted via 2.5km long channel shown in Fig.1. Due to the abundance of algae present in the seawater, which spread throughout the channel and clog it, the water flow reduces and the power plant efficiency drops. In present situations, in order to eliminate the clog, the power plant needs to be shutdown, so that workers can manually clean it. The proposed robotic solution isable to clean the channel while the power plant is still operational, providing both money savings and human safety. RJET – 111 is a result of an ongoing collaboration between the University of Zagreb and Inteco Robotics in research and development of mobile robots for hydrodynamic surface treatment. So far this research yielded a modular design of a lightweight mobile robot for hydrodynamic treatment of concrete and metal surfaces [1]. Building upon the results from this project we developed the proposed robot keeping in mind the highly specialized nature of its design, replacing the former modular approaches with a specific design approach. The shier size of the intake channel, and the necessity for the robot to carry the high pressure pump and the power system dictate that both robot's size and mass have to be significantly larger thanour previous, modular designs. Furthermore, robot's construction was tailored for the specific dimensions of the channel, so that its tool fits inside the channel, and its body can drive on top of it and safely navigate, localize and steer throughout its entire 2.5 km long path.
Figure 1 R JET 111 design, simulation, construction and real world implementation at the TE Plomin, Croatia The paper presents a detailed mechanical design of the robot that encompasses: 1 degree of freedom high pressure cleaning tool tailored for the channel profile; differential drive supported with two additional passive wheel pairs; and body capable of carrying full load of the high pressure pump, power unit, drives, electrical motors and electronics. The design and implementation of necessary electronics for both communication and control is also disseminated in the paper. Finally, we present the control system capable of steering the robot throughout the 2.5 km long channel with centimetre precision, starting from mathematical modelling and simulation, to real world implementation.
[1] Kovacic, Z.; Balac, B.; Flegaric, S.; Brkic, K.; Orsag, M. Light-weight Mobile Robot for Hydrodynamic Treatment of Concrete and Metal Surfaces“, The 1st International Conference on Applied Robotics for the Power Industry CARPI 2010, Montreal, Canada, 2010.






09.09.2014 15:00 Poster session

Nuclear fusion - 1101

 

Treatment of Inconel alloy from COMPASS tokamak with O2 and H2 plasma at high temperature

Alenka Vesel1, Aleksander Drenik1, Miran Mozetič1, Marianne Balat-Pichelin2, Jan Stockel2, Jozef Varju2, Radomir Panek3

Institut "Jožef Stefan", Jamova 39, 1000 Ljubljana, Slovenia1

CNRS-PROMES, Laboratoire Procédés, Matériaux & Energie Solaire, 7 Rue du Four Solaire, 66120 Font-Romeu Odeillo, France2

Institute of Plasma Physics, Czech Academy of Science, Za Slovankou 3, 18200 Prague, Czech Republic3

alenka.vesel@ijs.si

 

Oxidation of Inconel 625 alloy (Ni60Cr30Mo10Ni4Nb1) from COMPASS tokamak upon treatment with plasma at different temperatures up to about 1600 K was studied. Samples were treated for different periods up to a minute by oxygen or hydrogen plasma created with a microwave discharge in the standing-wave mode at a pressure of 40 Pa and a power 500 W. Simultaneous heating of the samples was realized by focusing concentrated solar radiation from a 5 kW solar furnace directly onto the samples. The temperature of the samples was measured by infrared pyrometer. The morphological changes upon treatment were monitored using SEM (scanning electron microscopy), compositional depth profiling was performed using AES (Auger electron spectroscopy), while structural changes were determined by XRD (X-ray diffraction). Treatment in oxygen plasma caused formation of metal oxide clusters of three dimensional crystallites initially rich in nickel oxide with the increasing chromium oxide content as the temperature was increasing. At about 1100 K iron and niobium oxides prevailed on the surface. Simultaneously the NiCr2O4 compound started growing at the interface between the oxide film and bulk alloy and the compound persisted up to temperatures close to the Inconel melting point. Intensive migration of minority alloying elements such as Fe and Ti was observed at 1600 K forming mixed surface oxides of sub-micrometer dimensions. The treatment in hydrogen plasma with small admixture of water vapor did not cause much modification unless the temperature was close to the melting point. At such conditions aluminum segregated on the surface and formed well-defined Al2O3 crystals.






09.09.2014 15:00 Poster session

Nuclear fusion - 1102

 

TRIGA Irradiations of Mn foils and TLD as Potential Tritium Production Monitors for Fusion Applications

Ivan Aleksander Kodeli1, Vladimir Radulović1, Darko Kavsek1, Władysław Pohorecki2, Tadeusz Kuc2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

AGH - University of Science and Technology,  30-059 Cracow, Poland2

ivan.kodeli@ijs.si

 

During the analysis of the Tritium Breeder Modules (TBM) which were tested in the scope of the EC ITER fusion activities we observed large similarities between the sensitivity profiles of the tritium production in 6Li and those of the 55Mn(n,gamma)56Mn reaction in the TBMs. This suggested that the latter reaction could be used as a tritium production monitor, at least for short term monitoring the half life being 2.579 h. However, in spite of the recent progress the Mn capture cross sections still include relatively large uncertainties and further improvements and validations are needed to meet the required accuracy. Manganese foils were therefore irradiated in a TRIGA thermal reactor, together with the Au foils and TLD (LiF) dosimeters and LiPb as monitors of tritium production, with the principal objective to study the energy response of the 55Mn(n,gamma)56Mn reaction.
Several experimental campaigns were performed at different irradiation channels in the JSI TRIGA research reactor, i.e. in the reactor thermal column, IC40 and pneumatic transport channel. Irradiations in different neutron spectra provide complementary information for the data validation. Measured results are analysed using M/C code MCNP, including evaluation of uncertainties involved in the measurements and the calculations.
The experiment was performed in the scope of the F4E supported project of the EC. The results of the measurements as well as the preliminary analysis and comparisons with the calculations will be presented in the paper.






09.09.2014 15:00 Poster session

Nuclear fusion - 1105

 

Towards a theory of collisionless-discharges with imbalanced global charge-outfluxes

Nikola Jelić1, Leon Kos2

Association EURATOM-ÖAW, Technikerstrasse 25, A-6020 Innsbruck, Austria1

University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia2

leon.kos@lecad.fs.uni-lj.si

 

Conventional Tonks & Langmuir collission-free plasma-sheath theory, which for decades serves as an archetype in formulating and solving a coupled set of Newton and Poison equations in a bounded laboratory, technology-oriented and fusion-related plasmas, is intrinsically limited to relatively simple physical scenarios in which internal electric curent are carried out exclusively by charged particles, which is clearly inaplicable for modelling the transient phenomena during a discharge development and decay. This seems to be quite a satisfactory approach when such a system is supposed to be a good prototype for e.g., resolving theoretically or in kinetic simulations the plasma-sheath boundary under various assumptions of volume plasma production and electron density distributions as functions of local plasma potential, of course, providing that such a problem is under given particular assumptionssolvable at all, by either of above methods. Namelly the plasma-sheath problem is still not only a strongly non-linear one, but depends on a considerable numbers of parameters of which just a single one, such as ion-source velocity distribution function, i.e., its particular shape and width (temperature) prevents generalizing kinetic results into a user-friendly terminology of statistically averaged, i.e., fluid quantities, exept under e.g., approximation of pure Maxellian or simpler ion-sources [L. Kos et al., Phys. Plasmas 16, 093503 (2009); N. Jelić et al., Phys. Plasmas 16, 123503 (2009)]. However, these difficulties could not be an excuse for a prior disregarding all possible solutions in which the rates of positive and negative charges leaving the discharge-domain is not ballanced, for e.g., a simple reason that one of charged population could be better/worse confined than the oposite one. In fact, this physical scenario should be regarded as a rule while a strict steady discharge is, actually a true exception which we are able to realize just on expense of a prior relaxed boundary conditions, which, in fact resut in a low plasma density. Within present investigation we perform a series of CPU-expensive numerical simulations towards obtaining periodically-relaxing time-dependent solutions that a dense plasma still can persist as a steady (properly averaged in time) discharge. A theoretical approach with first results regarding relevant plasma parameters (such as time-averaged plasma-profile density and critical ion-velocity at the plasma edge) is presented.






09.09.2014 15:00 Poster session

Nuclear fusion - 1106

 

On the Role of the Source Term in a Fluid Model of the Sheath Formation in an Oblique Magnetic Field

Tomaž Gyergyek1, Jernej Kovačič2

University of Ljubljana, Faculty of Electrical Engineering, Tržaška 25, 1000 Ljubljana, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

tomaz.gyergyek@fe.uni-lj.si

 

Sheath formation in front of a planar negative electrode immersed in magnetized plasma where magnetic field forms an arbitrary angle with the surface of the electrode is a very important subject for understanding edge fusion plasmas, especially in divertors. The scrape-off layer of a tokamak has low density plasma and strong magnetic fields. Further, the angle of incidence of the magnetic field lines onto the divertor plates spans nearly the whole range from perpendicular to parallel. So a fluid model of a magnetized sheath is very appropriate for the study of such conditions. In this paper the structure of the magnetic sheath and pre-sheath is investigated using a simple fluid model. The key equations of this model are the continuity equation and the momentum exchange equation. The electrons are assumed to be Boltzmann distributed and the potential is determined by a one-dimensional Poisson equation. Since there are 3 velocity components the momentum equation is decomposed into 3 equations. Together with continuity and Poisson equation a system of 5 differential equations for 5 unknown functions (3 velocity components, ion density and potential) is obtained. Three types of source terms are inserted into the continuity and momentum equation: 1) the zero source term, 2) the constant source term and 3) the exponential source term. The zero term means no charged particle production, the constant source term means uniform particle production in space and the exponential source term means spatially distributed particle production proportional to the local electron density, which is given by Boltzmann law. It is shown that the selection of the source term has strong impact to the solutions. For the constant and exponential source terms a very small initial ion velocity can be selected and the exponential source term results in a larger pre-sheath length than the constant term. The zero source term on the other hand means qualitatively different behaviour of the solutions. The initial velocity must be increased for several orders of magnitude in order to obtain similar pre-sheath length as with the other two source terms. In addition the solutions become very sensitive on the exact value of the initial velocity. Also at certain conditions the solutions exhibit strong oscillations. The nature of these oscillations is not yet clear, very probably this is a numerical issue.






09.09.2014 15:00 Poster session

Nuclear fusion - 1107

 

PIC Simulations of an Ion Energy Analyzer

Lojze Gačnik1, Jernej Kovačič1, Tomaž Gyergyek2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

University of Ljubljana, Faculty of Electrical Engineering, Tržaška 25, 1000 Ljubljana, Slovenia2

tomaz.gyergyek@fe.uni-lj.si

 

We modified the xpdp1 1d3v particle-in-cell plasma simulation code, adding 2 conductive grids to the system. Leaving one grid to the floating potential, and driving the potential of the other grid directly, we simulate a retarding-field ion energy analyzer (RFA), and directly compare the measured and real plasma energy distribution. Setting the opacity of the plasma-facing grid to near-1, we reduce the disturbance the RFA causes, approximating its small area compared to the plasma container surface in experiments. We examine the cases of a planar plasma source and volume ionization, both with a Maxwellian distribution, and compare the RFA characteristic if Monte-Carlo charged-neutral interactions with the background gas and simplified secondary emission from the plasma vessel walls and RFA grids are included in the simulation.






09.09.2014 15:00 Poster session

Nuclear fusion - 1109

 

Visualization schema for fusion data structures

Leon Kos1, Girish Ramesh2

University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia1

The University of Manchester, Materials Performance centre Materials Performance Centre, School of Materials, PO Box 88, M60 1QD Manchester, United Kingdom2

leon.kos@lecad.fs.uni-lj.si

 

3D scientific visualization in HPC environments is a topic that ranges from post-processing (on dedicated visualization clusters) to in-situ code instrumentation. Often, 3D visualization is based on multi-layered data access frameworks that need custom plugins to be developed for specific codes. Interfacing fusion codes in EUROfusion Code Development for Ingerated Modelling is based on Consistent Physical Objects (CPOs). CPOs are standardized data structures that describe various physical aspects of fusion experiments and are designed to be suitable for use with simulation codes and experimental data. Integration with CPOs thus brings a common data model to integrated simulations that allows direct comparison with experiment, use of experimental data as an input or mixed approaches. To facilitate change and to support different programming languages, the data structure is described by a XML schema definition (XSD) from which visualization schema in XML is generated and included in datastructure description.






09.09.2014 15:00 Poster session

Nuclear fusion - 1110

 

The influence of nitrogen co-deposition in mixed layers on deuterium retention and thermal desorption

Anže Založnik1, Sabina Markelj1, Iztok Čadež1, Primož Pelicon1, Corneliu Porosnicu2, Cristian P. Lungu2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

National Institute for Laser, Plasma and Radiation Physics, P.O. Box MG36, Magurele-Bucharest, Romania2

sabina.markelj@ijs.si

 

Due to different conditions inside the fusion devices plasma-facing components consist of different materials. The most appropriate materials chosen for ITER are beryllium and tungsten for the first wall and divertor, respectively. Because of the plasma-surface interactions during the operation of the fusion device, erosion and re-deposition of the material occur, leading to formation of mixed material deposits on various surfaces inside the vessel.
Nitrogen or noble gas seeding is considered in ITER for the reduction of power loads on inner wall and to improve the confinement of the core plasma [1]. Seeding gas interacts with plasma-facing components, leading to incorporation of seeding impurities into the mixed material deposits. For the estimation of the overall fuel retention, the retention in these mixed material deposits needs to be determined.
The influence of nitrogen co-deposition in mixed W:Al and W:Be layers (Al being used as a Be substitute) on deuterium uptake was studied in situ by Nuclear Reaction Analysis (NRA) and mass spectroscopy during linear heating of the samples. Samples were prepared using Thermionic Vacuum Arc (TVA) method [2]. Mixed W:Al layer without nitrogen was used as a control sample. Layers were prepared on silicon wafers and were 1µm thick in the case of W:Al and 100nm thick in the case of W:Be. All samples were exposed to deuterium atom beam at 100°C for 20 hours and then linearly heated to around 800°C. Deuterium depth profiles were measured by NRA using 3He ion beam at different ion beam energies before and after heating. During the heating the same beam at single energy (2.5MeV) was used together with the quadrupole mass spectrometer continuously monitoring the release of deuterium and other gas species.
We have observed significant increase of the uptake of deuterium in the layers with co-deposited nitrogen. Main release of deuterium from the mixed layers took place between 350°C and 600°C. During the heating severe layer modifications were observed on W:Al samples using the Rutherford Backscattering Spectroscopy (RBS) technique.
[1] M. Oberkofler et al., J. Nucl. Mater., 438 (2013), S258-S261
[2] C. P. Lungu et al., Phys. Scr. T128 (2007) 157–161.






09.09.2014 15:00 Poster session

Nuclear fusion - 1111

 

Fabrication and characterization of W-based composites

Saša Novak, Aljaž Ivekovič

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

sasa.novak@ijs.si

 

Present design studies for high heat load, plasma facing components make use of the high temperature strength and good thermal conductivity of tungsten, a primary candidate for structural applications in future fusion power plants. The most critical issue of tungsten materials for structural applications is their brittleness at low temperature and recrystallization and creep at high temperature.
Therefore, in this work strengthening of tungsten by addition of low activation reinforcing phases was investigated. (Nano)particle-reinforcement of W-based composites (W-Si-C, W-TiC, W-W2C and W-Ti3SiC2) or reinforcement with SiC fibres was proposed to enhance mechanical performance of W-based materials, with sufficient retention of initial thermal properties in order to enable application of the material in high heat load applications, even after exposure at temperatures exceeding the tungsten recrystallization temperature. The effect of additive type and volume fraction on microstructure, mechanical properties and their influence on thermal conductivity as well as reaction kinetics at elevated temperatures were investigated.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1208

 

Uncertainties in Fatigue Life Assessment of Pipes due to Random Thermal Loads

Oriol Costa Garrido, Samir El Shawish, Leon Cizelj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

oriol.costa@ijs.si

 

Fluids at different temperatures turbulently mixing in pipes are known to be the cause of thermal fatigue of surrounding material. Fluid temperature histories in the fluid-wall interface have been shown, experimentally and numerically, to resemble random signals of multi-frequency content with higher amplitudes in the low frequency band which is characteristic of the inertial sub-range. However, fatigue life of components under this type of thermal loads is usually estimated with sinusoidal temperature signals of a single frequency and amplitude that represents the fluids’ temperature difference. Fatigue lives are computed for various frequencies by intersecting derived surface stress amplitudes with S-N design fatigue curves.

In this paper, uncertainties in fatigue life assessment are studied by means of random temperature signals of different statistics compared to single frequency sinusoidal. The frequency power content of the generated random signals is controlled in order to evaluate parameters such as most harmful frequencies and signals’ time length. In these cases, structural damage is accounted by rainfall counting and linear damage accumulation rule. The results show that variability in fatigue life estimation using random temperatures is reduced with an increase of the signal time. Higher variance and lower frequency content of the temperature fluctuations reduce the estimated fatigue life.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1209

 

DFT study of uranium and plutonium oxides in gas phase: structures and thermodynamic properties

Marta Cerini, Giuseppe Dia, Elena Macerata, Eros Mossini, Marco Giola, Mario Mariani, Carlo Cavallotti

Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

marta.cerini@mail.polimi.it

 

Recent significant advances in computational methods have allowed the actinide chemistry to be modeled for the first time, suggesting the possibility to circumvent expensive experimental studies. The impact of such progresses on research areas devoted to nuclear materials and the development of Gen IV nuclear systems could be fundamental in determining their successful evolution.
The present work would represent an effective contribution to this scenario by exploring the application of the density functional theory combined with the spin-polarized generalized gradient approximation to study the equilibrium structures, the thermodynamics properties and energetics of uranium and plutonium oxides in gas phase. DFT-GGA calculations have been successfully performed to investigate UO, UO2, UO3, PuO, PuO2 and PuO3. The bond lengths and the harmonic vibration frequencies of the ground states of UO, UO2 and UO3 are in good agreement with the available experimental data, while for PuO, PuO2 and PuO3 molecules no reference data were found. For each compound several initial geometric configurations and different spin multiplicities have been systematically studied in order to determine the minimum energy configuration of the molecule. Entropy and specific heat have been determined by the calculated vibrational frequencies, obtaining for UO, UO2, UO3 and PuO3 encouraging results in complete agreement with experimental data. Starting from the energy information, it has been determined the bond dissociation energy and the formation enthalpy for all the compound considered. The bond dissociation energy of the ground state of UO, UO2 and UO3 are in agreement with the experimental data. For PuO, PuO2 and PuO3 molecules, the obtained values are coherent with the only values available in literature and derived by calculations themselves. Negative enthalpies of formation for UO, UO2, UO3, PuO, PuO2 and PuO3 in gas phase have been obtained by the DFT calculations with a good agreement with the experimental data, where available. In conclusion, the theoretical approach considered has been validated on the uranium oxides by comparison with experimental values and then successfully applied to the study of plutonium oxides.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1210

 

Development of the probe for the Ultrasonic
examination of tube butt welds with tube wall thicknesses less than 8 mm

Leonardo Trupinić, Nikola Pavlović, Renato Hohnjec

INETEC-Institute for Nuclear Technology, Dolenica 28, 10 250 Zagreb, Croatia1

leonardo.trupinic@inetec.hr

 

Based on INETEC’s Customer specific needs for ultrasonic examination of boiler alloyed steel tubes having relatively small diameters and with wall thicknesses less than 8 mm, manually butt welded with expected manufacturing defects (surface cracks in weld toe and root regions, porosities, weld inclusions and slags, lack of fusion flaws, weld root imperfections), INETEC performed a series of experimental investigations and initial testing trials with purpose to find and determine the optimal solution for fulfillment of the examination objectives and Customers examination requirements.
During initial testing trials and UT analysis of results obtained from characteristic welded samples with nominal tube wall thicknesses less than 8 mm, when the main effort was made to avoid dead zone and near field effects and provide high detection, resolution and sizing capabilities of potential defects, while keeping the level of sound field attenuation acceptably low, INETEC adopted the new concept of the dual element angled beam probe generating 70° transversal wave refracted angle beam. The developed and manufactured probe type is characterized by 6.2 MHz nominal frequency, dual element 70° transversal wave refracted beams in alloyed steel, while having contoured wedge design with relatively short beam exit-point offset thus enabling maximal approach to weld crown and direct weld root examination.
In order to evaluate the theoretical presumptions set up for the developed probe efficiency and to verify its examination capabilities in regard to welding defect detection and sizing, the test samples containing welding defects were manufactured by the Customer and made available to INETEC. The extensive ultrasonic examination of the test samples was performed at INETEC laboratory, in accordance with the applicable examination procedure by group of UT examiners independently. Based on obtained testing results acquired during performed probe evaluation and reasonable comparison of obtained UT results with the Customer provided RT results, the following was concluded:
The examination of available test samples demonstrated that the UT system applying the newly developed dual element 70° transversal wave probe may detect and differentiate the variety of embedded welding flaws (surface cracks in weld toe and root regions, porosities, weld inclusions and slags, lack of fusion flaws, weld root imperfections).The application of newly developed probe provided an adequate angle-beam examination, full coverage of required inspection volume as well as good detection, resolution and sizing capabilities of potential welding flaws throughout the entire tube-wall thickness.
Comparison of the results obtained from the two different examination methods provided the solid proven base for application of ultrasonic examination of tube butt welds with tube wall thicknesses less than 8 mm as the less time and resources consuming examination method, while ensuring satisfactory flaw detection, sizing and classification capabilities.
The capability of the newly INETEC developed ultrasound probe to detect, differentiate and assess found welding flaws was during the evaluation process proven and verified.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1211

 

Fuel-coolant chemical interaction in LFR systems: a preliminary thermodynamic approach

Maddalena Negrin, Eros Mossini, Elena Macerata, Marta Cerini, Marco Giola, Mario Mariani, Laura Pellegrini

Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

marta.cerini@mail.polimi.it

 

Within the development of Gen IV nuclear systems, safety is one of the most relevant objectives to pursue. In particular, for Lead-cooled Fast Reactors the limited experience on the chemical interaction between MOX fuel and lead, together with the not negligible probability of fuel de-cladding that could lead to fuel-coolant contact, have forced to debate the chemical issues for safety considerations. Being apprised of the system behavior under normal operation or accidental conditions is paramount for defining the actions to undertake in order to avoid serious consequences.
In the present work this issue has been faced for the first time leading to the evaluation of the system composition at equilibrium, by considering the ELSY reactor as reference system for the definition of geometrical characteristics and initial chemical composition. Systems composed of Pb and both metallic and oxide fuels have been considered. After a thorough bibliographic research, a database of thermodynamic data has been completed for all the species that are present or could be formed following the contact. Enthalpies, entropies and specific heats are fundamental information and, if missing, they have been estimated. Therefore, starting from the thermodynamic database and by minimization of the Gibbs free energy, the composition of the systems at equilibrium has been evaluated in the range of temperature from 500 to 1700K. This preliminary research activity showed as the systems having metallic fuels are really very reactive with formation of intermetallic compounds already at the lower temperatures, while those with oxide fuels seem to be stable up to 1200K. The work done has highlighted the importance of knowing which could be the evolution of an event of contact between nuclear fuel and Pb, and how the applied approach could be useful to perform a first preliminary analysis of the problem. Obviously, much more interesting considerations could be drawn by introducing in the system the fission products and the cladding constituents. Unfortunately, the thermodynamic data necessary to study such complex systems are not available in literature up to now, therefore studies aiming at evaluating such properties by theoretical methods are already in progress, exploiting in particular semi-empirical and ab-initio theories.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1212

 

Formation Enthalpies estimation by a semi-empirical approach for chemical considerations in Lead-cooled Fast Reactors

Alberto Tosolin, Elena Macerata, Eros Mossini, Marta Cerini, Marco Giola, Mario Mariani

Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3, 20133 Milano, Italy

marta.cerini@mail.polimi.it

 

Within the development of Gen IV Lead-cooled Fast Reactors, a great interest is growing up for the issue of chemical interactions between MOX fuel and coolant. A better knowledge of such chemical aspects will enable to foresee the system behavior in case of a de-cladding event under normal operation or in accidental conditions. For this purpose it is paramount to have available the thermodynamic data for all the species that could be formed from chemical reactions between mixed U and Pu oxides and Pb. Since a very limited experience on Pb-cooled reactors has been collected in the last decades, no thermodynamic data involving Pb and nuclear materials are available in literature. The present work would like to contribute to the compiling of a thermodynamic database containing enthalpies, entropies, specific heats, ecc., all information needed to assess the composition of a system at thermodynamic equilibrium as well as the formation of stable compounds by exothermic reactions.
For this purpose, the Miedema model, a well-established semi-empirical method, has been considered in order to estimate the enthalpies of formation of all the Pb-containing binary compounds of interest for LFRs. Binary compounds referring to the Pb-fuel, Pb-cladding and Pb-Fission Products sub-systems have been considered. An extension of the model to the elements of group 16, developed by a research group at Paul Scherrer Institute, has been used to estimate enthalpies for O-containing compounds. Moreover, in order to extend the Miedema’s approach to multicomponent systems in particular to ternary systems, the Gallego and Ray methodologies have been identified and compared: the Gallego proposal has been chosen as the most suitable for the case in point and combined with the extension to the oxygen group. The calculated and experimental values of formation enthalpies are in good agreement, considering the semi-empirical nature of the model. In addition, new values for Miedema electronegativity and hybridization term for Pb have been proposed obtaining a better agreement with experimental data. In conclusion, the results obtained has showed that the extended Miedema’s approach could be successfully applied to the estimation of formation enthalpies of compounds of interest to investigate the chemical interaction between fuel and coolant in LFRs, making available thus the first set of data necessary to formulate preliminary thermodynamic evaluations about the fuel-coolant interaction event. More generally, the present work has demonstrated that semi-empirical methods could be powerful tools to easily provide thermodynamic information in order to study even more challenging scientific and technological issues.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1213

 

The corrosion behaviour of dissimilar steel weld due to synergy of several effects: Field case study

Marjan Suban1, Robert Planinc2, Borut Bundara1, Robert Cvelbar1

Inštitut za metalne konstrukcije, Mencingerjeva 7, p.p.3410, 1001 Ljubljana, Slovenia1

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia2

marjan.suban@imk.si

 

Various combinations of carbon/low alloy steel and stainless steel in welded joints are widely applied in the process, construction and nuclear industry, as well. In practice it can happen that the welded joints or construction parts are exposed to the unexpected conditions that accelerate their degradation. The paper presents the results of the study focused on strictly localized and accelerated corrosion of flexible hose material pairs of carbon steel and austenitic stainless steel that has been witnessed in practice. Flexible hose as a part of Component Cooling System for Reactor Coolant Pump Motor Bearing was subjected to exceptional environmental conditions during its operation, which lead to unexpected corrosion. Results show that synergy of several non-connected effects caused localized corrosion of just one part of less noble coupled carbon steel and significantly increased corrosion rate comparing to the other parts made of same carbon steel. The weight-loss of metal during corrosion is mainly caused by galvanic corrosion, but also other corrosion processes and environmental conditions are not negligible and can significantly contribute to overall corrosion rate.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1215

 

High Sensitivity Ultrasonic NDE Method for Early Detection of Creep Damage in Alloy Steel Steam Systems in Power Plants

Marko Budimir1, Channa Nageswaran2, Liudas Mažeika3

INETEC-Institute for Nuclear Technology, Dolenica 28, 10 250 Zagreb, Croatia1

TWI Ltd, United Kingdom2

Kaunas University of Technology, K. Donelaicio 73, LT-44239 Kaunas, Lithuania3

marko.budimir@inetec.hr

 

Creep is the time-dependent, thermally assisted deformation of a component operating under stress. Metal pressure components such as boiler tubing, headers, and steam piping in power plants might operate at thermal conditions conducive to causing creep damage over the operating life of the component. To ensure safe and reliable operation of such components in service, utilities periodically use non-destructive evaluation (NDE) techniques to inspect these components for damage. These inspections are largely targeted at detecting late stage creep damage in which cracking is active in the component and provides qualitative rather than quantitative data. Recent advances in NDE technology have provided enhanced capabilities for incipient creep failure detection. In this work we seek to develop a high sensitivity NDE system that will apply advanced signal processing ultrasound testing techniques that have already shown a capability to identify early stage creep damage, and to produce a library of defects with the aim of providing inspection limits and the probability of detection for the technique and thus enable accurate life cycle prediction for components under inspection. For this purpose, an analysis of specimens with a range of creep induced damage has been performed with the aim to generate a specimen set with representing Type IV creep damage in an early stage of damage. The nature of the creep strength enhanced ferritic steel material prone to creep induced damage is presented along with a description of the types of failure, which are classified according to the position with respect to the weldment where the failure occurs. Modelling and experimental validations were performed that determine the amplitudes of the ultrasound signals reflected by small diameter reflectors with dimensions down to 5µm. The experiments have shown that small cracks with dimensions close to several micrometres, caused by creep damage, should be detectable using ultrasonic focusing transducers with frequencies 25-50MHz. Future advanced signal analysis work and mechanical scanning prototype system design for in-situ testing are envisaged and presented.






09.09.2014 15:00 Poster session

Materials, integrity and life management - 1217

 

A Methodology to Evaluate Stress in Surge Line from Outer Surface Temperature for Fatigue Assessment

Younjung Kim, Sunyeh Kang, Geeseok Kim, Hyunmin Kim, Daehee Lee

KEPCO-E&C , 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353, South Korea

yjkim@kepco-enc.com

 

The pressurizer surge line is one of locations of high usage factors and most easy-to-fail considering the effects of LWR environments. Meanwhile, it is known that usage factors at many locations could be shown to be acceptable by refined analysis using fatigue monitoring based on actual metal temperature. This paper discusses the methodology to evaluate the stress from the measured outside wall temperature by means of inverse heat conduction problem (IHCP) and direct transfer function method. The result of IHCP was divergent with the low sensitivity coefficient. The sequential IHCP method was suggested to solve this problem. As a result, it was possible to predict the inner surface temperature of pipe based on the outer surface temperature and to evaluate the thermal stress through the calculated inside temperature. In conclusion, the proposed methodology is effective in the evaluation of inner surface temperature and stress time history at the surge line.






09.09.2014 16:40

Radioactive waste and decommissioning - 805

 

MCNP calculations in connection to the decommissioning scenario

Amine Bouhaddane1, Gabriel Farkas1, Vladimír Slugeň2

Slovak University of Technology Faculty of Electrical Engineering and Information Technology, Department of Nuclear Physics and Technology, Ilkovičova 1, 812 19 Bratislava, Slovakia1

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava, Slovakia2

amine.bouhaddane@stuba.sk

 

This paper is focused on the detailed look on the radionuclide inventory determination as an essential part of the decommissioning selection process and as a prerequisite for the radioactive waste management. Based on the radionuclide inventory calculation, it is possible to optimize the time frame and to choose the appropriate dismantling procedure during the disposal of reactor internal and external components in the decommissioning phase of a nuclear power plant (NPP). MCNPX code enables precise three-dimensional modelling of these components. Due to high technological and financial demands in the case of in situ measurements or sampling, it is important to use validated code systems to precisely estimate the radionuclide inventory. Required precision of modelling depends on the expected decommissioning method. Essential part of a correct calculation is the collection of reactor operation data. In this paper, ability of MCNPX code in this field is presented on the example of calculation for reactor internal components used in VVER-440 reactor. As parts next-to the reactor core these are the most radioactive components besides the spent nuclear fuel.






09.09.2014 17:00

Radioactive waste and decommissioning - 802

 

A Well Established System for the Dry Storage of Spent Fuel

Juergen Skrzyppek, Stefan Fopp, Michael Koebl

GNS - Gesellschaft für Nuklearservice mb, Hollestrasse 7a P.O. Box 101 253, 45127 Essen, Germany

eva.troedel@gns.de

 

For several decades in the use of nuclear energy the predominant solution for the interim storage of spent nuclear fuel (SNF) have been spent fuel pools. As integral part of the original construction of the plants the fuel pools were designed to accommodate – at the most - all the spent nuclear fuel assemblies accumulated during the originally intended plant life.

While internationally there still are no final repositories for high-level waste in operation or at least in construction, many plants are reaching the end of their originally intended life cycle as well as capacity limits of their fuel pools. A prerequisite for the in many cases aspired plant live extensions therefore is a feasible way for removing the fuel assemblies from the pools to additional storage facilities. In addition to these rather practical and economical issues, the interim storage in spent fuel pools has undergone a substantial re-evaluation in the aftermath of the Fukushima Accident. Therefore systems for dry interim storage in dual-purpose casks become increasingly import.
In contrast to most other countries using nuclear energy, Germany’s nuclear industry already in the 1970ies pursued feasible ways for dry interim storage in dual-purpose casks. With the political decision of the year 2000 to quit reprocessing, on-site interim storage facilities had to be erected at all the NPP’s sites and the demand for casks rose to unprecedented numbers.
The complete life cycle of the spent fuel cask is covered by the German company GNS Gesellschaft für Nuklear-Service mbH, since it is not only the designer and manufacturer of the casks, but is also responsible for the loading, transport and storage of the casks. GNS today looks back on more than 30 years of operational experience with dual-purpose casks. Following customer demands, GNS developed two different cask types for SNF – the CASTOR® and the CONSTOR® cask type. While the CASTOR® type is optimized for high thermal loads – which allows loading after extremely short cooling times and/or high burn-up of the SNF – the CONSTOR® type is cost-optimized for the cost-efficient storage of large quantities of SNF.
The GNS casks can be stored with or without storage buildings depending on national regulations. For heat removal from the storage buildings, a passive cooling system by natural convection is sufficient. This makes the interim storage of the casks a completely passive system. The building provides additional protection from environmental influences and reduces radiation exposure to the public.
Operational experiences gained during several hundred loadings mostly carried out by GNS’ own staff, have consequently been fed back to the cask designers. The result today is a cask design which offers easy handling and guarantees minimum turnaround times within the reactor unit.
By now almost 1,300 GNS-casks are in operation worldwide. In Germany alone, more than 1000 CASTOR® casks are stored with individual storage periods of up to 30 years. Taking into account the additional casks that have to be manufactured, loaded and stored during the final years of the German Nuclear Phase-Out, there will be 2000 casks by GNS in operation worldwide.
The presentation will give an overview over several national and international projects and show the bandwidth of customized solutions by GNS.






09.09.2014 17:20

Radioactive waste and decommissioning - 811

 

A global used fuel solution (Closing the nuclear fuel cycle for a utility or country with a limited fleet and no confirmed new build program)

Peter Breitenstein, Cecile Evans

AREVA, 1 place Jean Millier, 92084 Paris La Defense, France

peter.breitenstein@areva.com

 

The closed fuel cycle, as it is being practiced in countries with a large nuclear program, typically involves many nuclear power plants, including older reactors shutting down and newer ones starting up. Valuable materials (uranium, plutonium) recovered from treatment of used fuel discharged from older reactors can be recycled as new fuel in the current fleet, thus leaving no used fuel and only conditioned residues to manage. Taking into consideration the long cycle time (a typical fuel loop takes about 15 years), it may happen that recycled fuel such as MOX and ERU is used for power production in different reactors than the reactor that originally generated the nuclear materials. And some utilities cannot or may not recycle MOX in their reactor(s).
When looking at countries or utilities with only one or a few operating reactors and no upcoming new build program, the scheme has to be adapted to match the different constraint fields. Thus, in order to allow utilities to keep their options open, AREVA proposes a wide range of flexible reprocessing/recycling solutions that can be tailored to the specific needs of each individual utility/country.
Based on a current example, the paper will demonstrate how the nuclear fuel cycle is and/or may be closed for such countries or utilities with only vitrified and compacted universal waste canisters to manage as a final output at the end of their nuclear cycle. The paper will illustrate the technical, logistic, legal and economic aspects of such sustainable solutions.






09.09.2014 17:40

Radioactive waste and decommissioning - 804

 

Natural and Engineering Barriers – the Safety Concept Basis for LILW Repository in Vrbina, Krško

Sandi Viršek, Tomaž Žagar

ARAO – Agencija za radioaktivne odpadke, Celovška cesta 182, 1000 Ljubljana, Slovenia

sandi.virsek@gov.si

 

From 2009 Slovenia has the approved site Vrbina Krško for the future low and intermediate level waste (LILW) repository. The site is situated about 300 meters eastwards from Nuclear power plant Krško. For the combination of the site properties and radioactive waste inventory in Slovenia, the disposal concept was developed. Waste will be disposed in near surface silos, excavated from the surface. The good properties of the surface and underground radwaste repositories are joined in this concept and they are reasonably adapted to the site.
In the paper the disposal concept is shortly presented, with emphasize on the concept robustness. The passive disposal barriers are then presented and their combination that presents the multi barrier system that assure the long term safety of the repository. The planned design fulfils also the requirements about multiple safety functions and defence in depth. Such planned and accomplished facility present negligible impact on the human and the environment.






10.09.2014 08:30

Thermal-hydraulics - 901

 

Mitigation Strategy for Extended Blackout Power Condition

Andrej Prošek, Andrija Volkanovski

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

andrej.prosek@ijs.si

 

The accident at Fukushima Dai-ichi showed the significance of the challenge presented by a loss of safety-related systems following the occurrence of a beyond-design-basis external event. In the case of Fukushima Dai-ichi, the extended loss of alternating current (AC) power condition caused by the earthquake and consequential tsunami led to loss of core cooling and a significant challenge to containment.
In response to the accident at the Fukushima Dai-ichi nuclear power plant a set of measures have been proposed and implemented in the nuclear power plants. Those measures consist of adding diverse and flexible mitigation strategies that will increase defense-in-depth for beyond-design-basis scenarios to address the extended loss of AC power and loss of normal access to the ultimate heat sink occurring simultaneously at all units on a site. Proposed measures include utilization of portable equipment that provides means of obtaining power and water to maintain or restore key safety functions for all reactors at a site. Those measures include actions to maintain power to the instrumentation for the key reactor parameters.
In this paper results of the analysis of the utilization of portable equipment for mitigation of the extended blackout condition, when all power in pressurized water reactor (PWR) is lost, are presented. For deterministic calculations the RELAP/MOD3.3 Patch 04 computer code was used. The following general boundary conditions were used in the analysis:
• Reactor was initially operating at power and is successfully shut down when required (i.e., all rods inserted, no Anticipated Transient Without Scram).
• Steam generators (SGs) are available.
• Connection for portable pump to feed required SGs are available.
• Alternate water supply is available for portable pump
• Power to essential instrumentation and control systems is lost
The results of the analysis will provide answers to the following questions:
- Is it possible to maintain effective core cooling with mobile pump during the extended station blackout condition
- What is the available time for utilization of the mobile pump
- Is it possible to assess flow rates of the mobile pump from the deterministic model of the nuclear power plant
- What are the probabilities of the analyzed station blackout accident sequences
The obtained results will show if and to what extent the proposed strategy is effective for mitigation of the extended blackout condition in a PWR. The assessed probabilities of the analyzed accident sequences will be compared to other dominant sequences of the given plant.






10.09.2014 08:50

Thermal-hydraulics - 908

 

Simulations of condensation induced slug formation and slug acceleration in partially filled elbow shaped geometry

Jure Oder, Matej Tekavčič, Iztok Tiselj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

jure.oder@ijs.si

 

Slug flow is a two-phase flow in which gaseous phase is in the form of large bubbles that are separated by liquid slugs. The slugs are associated with pressure oscillations in a pipe and under certain conditions, the formation and acceleration of the slug can lead to a hydraulic shock. This occurs when a rapid change of fluid velocity in a pipe causes a change in pressure. For example, when a slug with high velocity reaches the pipe end, the fluid of the slug is forced to rapidly decelerate and that results in a pressure surge.
In horizontal pipes, the slug can arise as a by-product of condensation. Hot gaseous phase, which inflows into a pipe that is partially filled with cold liquid, causes shear on the surface. This can lead to a wavy gas-liquid interface that may eventually form a slug. The gas in front of the slug is thus isolated and because of condensation the pressure drops. This pressure imbalance accelerates the slug which is then hurdled into the pipe ending.
The hydraulic shock or water hammer is an unwanted phenomenon. It causes damage to piping or to other equipment. The pressure surge can be very short-lived, which results in it not being released through the safety valves that are installed to protect the piping, but still be strong enough to break welds. The usual technique to reduce the effects of hydraulic shock involve reducing the lengths of straight parts of pipes by introducing elbows as well as by installation of accumulators, expansion tanks or tank drums.
In this paper we present preliminary results of simulations of condensation induced slug formation and slug acceleration in partially filled elbow shaped geometry. This process is modelled with incompressible two-fluid model up to the point of void collapse. The time between the formation and the void collapse is in the order of seconds. The velocity of sound is therefore not relevant. At the time of void collapse, the liquid inertia becomes the dominant cause for the slug propagation.






10.09.2014 09:10

Thermal-hydraulics - 903

 

Analysis of CET and PCT during a SBLOCA. Application to a scaled-up model.

Andrea Querol, Sergio Gallardo, Gumersindo Verdú

Polytechnic University of Valencia, Department of Chemical and Nuclear Engineering, Camí de Vera sn, 46022 Valencia, Spain

sergalbe@iqn.upv.es

 

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions, which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Temperature (CET) thermocouples to detect core overheating by considering that superheated steam flows in the upward direction when core uncover occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant time delay and temperature discrepancy in the superheat detection by the CET thermocouples from core heat-up can be produced. CET is important for initiating the AM action and the safety concern is that such AM measures could be so delayed that recovery actions would be less effective.
It was noted in Test 6-1 of the OECD/NEA ROSA Project, which was performed in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). This experiment simulates a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9 % of the cold leg area under the assumption of total failure of High Pressure Injection System (HPIS). Experimental results showed that the core uncover started significantly early before the CET thermocouples indicated superheating and that the temperature increase rate was higher in the core than in the CET. A time delay of some seconds for the CET thermocouples readings to reach 623 K (criterion to start AM action) was observed in the test.
In this work, the thermal hydraulic code TRACE5 is used to clarify responses of CET thermocouples versus PCT and to study if the same physical phenomena is reproduced in two TRACE5 models with different geometry. These both TRACE5 models correspond to LSTF, a Full Height Full Pressure (FHFP) facility, and a scaled-up LSTF model obtained applying the power-to-volume scaling criterion. Due to the fluid exhibits the same properties at full pressure, the power-to-volume scaling criterion is applied to preserve time, power and mass.
Results obtained shown the relation-ship between CET and PCT and how the delay between the core uncover and the CET excursion is produced in Upper Head SBLOCA scenarios reproduced in small scaled facilities such as LSTF and in a scaled-up PWR plant using the thermal hydraulic code TRACE5.






10.09.2014 09:50

Thermal-hydraulics - 904

 

Best Estimate Calculation and Uncertainty Evaluation of LBLOCA of APR+ Considering New Safety Features

Young Seok Bang, Sweng Woong Woo, Min Jeong Hwang, S.K. Sim

Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

k164bys@kins.re.kr

 

Advanced Power Reactors Plus (APR+) Standard Design has several safety design features such as full four trains of Safety Injection System (SIS), ECCS Core Barrel Duct (ECBD) inside Reactor Vessel (RV) and Safety Injection Tank with Fluidic Device (SIT/FD). The reactor power of APR+ is expanded to 4290 MWth and thus, its RV is enlarged and SI pump capacity is increased while water amount of SIT is the same as the previous APR1400 design. To confirm the effect of such design features adopted in APR+ on safety, Best Estimate (BE) calculation with uncertainty evaluation for a Large Break loss-of-coolant accident (LBLOCA) is conducted. The transient system thermal-hydraulic response is calculated using MARS-KS code and the noding scheme which has been extensively validated at the previous APR1400 calculation is used. One of the major features of noding scheme is a modeling of two parallel channels for RV Upper Guide Structures (UGS) and Upper Plenum (UP), which leads to a conservative approximation by exclusion of the temporal quenching during blowdown (blowdown quenching) instead of the complex process for the evaluation of uncertainty related to the prediction of the phenomena. Especially, the SIT/FD which has a strong impact on core reflood behavior is specifically modeled, in which the SIT vessel, standpipe, connecting pipes in FD, and mixing chamber are represented by PIPE component instead of ACCUM component. Dynamic flow resistance model is applied to the standpipe, which was validated by the SIT blowdown test of the SKN Unit 3. Totally 23 parameters are considered as source of uncertainty important to Peak Clad Temperature (PCT). Among them, fuel parameters such as fuel gap conductance, core heat transfer correlations, and system parameters such as break discharge coefficient, pump two-phase multiplier, and SI design parameters are included. Three parameters for the flow path resistance connecting ECBD and RV downcomer are considered. Core reflood phenomena related parameters such as Forslund-Rohsenow correlation and Ishii-Mishama entrainment correlation are also considered. In the present study, ranges of the major uncertainty parameters are determined based on the previous studies such as OECD BEMUSE program and the previous experience. For the remaining parameters, wide uncertainty ranges are selected conservatively. The result shows that the reflood PCT in 95 percentile upper limit and 95 percent confidence level is 1363 K and higher than the blowdown PCT of 1275 K due to the conservatively wide range of uncertainty for reflood related parameters.






10.09.2014 10:10

Thermal-hydraulics - 905

 

Instantaneous Heat Transfer Characteristics of Multiple Impinging Jets

Martin Draksler1, Boštjan Končar1, Leon Cizelj1, Bojan Ničeno2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Paul Scherrer Institut (PSI), Nuclear Energy and Safety Department, OVG/421, CH-5232 Villigen PSI, Switzerland2

martin.draksler@ijs.si

 

Jet impingement cooling technique is recognized as one of the most efficient cooling techniques among single-phase convection flows. As such, it is widely used in many industrial applications and the method is also proposed for cooling of the divertor, a component of the future fusion reactor DEMO. Despite that topic has been extensively analyzed, complete understanding of the heat transfer mechanisms is still absent.
In this paper, the instantaneous flow and heat transfer characteristics of highly turbulent impinging jets in hexagonal configuration are numerically studied by the means of Large Eddy Simulation. The study is conducted in order to identify the key physical phenomena that govern efficient heat transfer of impinging jets. Instantaneous flow field is examined, with emphasis laid on the propagation of the coherent structures and their influence on the instantaneous heat transfer characteristics. The results show that already intense and rapidly changing heat transfer is additionally enhanced by the impact of these large-scale structures upon the target wall. A thorough analysis of three-dimensional, time-dependent velocity and temperature field is presented and discussed in the paper.






10.09.2014 10:30

Thermal-hydraulics - 906

 

NEK Small Break LOCA Simulation by APROS code

Tadeja Polach1, Ivica Bašić3, Luka Štrubelj2

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia1

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia2

APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia3

miha.lenic@gen-energija.si

 

The subject of the paper is usage of APROS system code for simulation of the Small Break Loss of Coolant Accident (SBLOCA) on a model of Nuclear Power Plant Krško (NEK). The paper briefly describes basic characteristics of the model developed in APROS, nodalization and performed steady-state verification and validation, upon which further accident scenarios will be simulated. Proper verification and validation process of the plant model (including the code developer validation of code itself) is essential to assure accurate calculation of steady state and/or postulated accident scenarios.
The APROS thermal-hydraulic model of NEK encompasses the primary system, part of the secondary system and simplified containment model. The primary circuit model is very detailed including all the main components of reactor coolant (RC) system. The model also includes all the main control systems (i.e. reactor control system, primary circuit pressure and pressurizer level control, turbine power control, steam generator pressure and level control) , protection systems (i.e. Reactor Trip Protection (RT), and Engineered Safety Feature Actuation (ESFAS) Systems). Model includes the verified and validated safety systems as Emergency Core Cooling System (Low Pressure Injection System (LPIS), High Pressure Injection System (HPIS)), and Auxiliary Feedwater (AF) System to accommodate requirements for performance of an SBLOCA accident scenario simulation. For analysed SBLOCA scenario, the break opening of 4”˝ located on the primary system cold leg was assumed. The paper describes various aspects of the SBLOCA simulation by APROS code and presents the NEK SBLOCA APROS simulation and its comparison with the identical scenario and similar NEK model performed with a system code RELAP5.






10.09.2014 11:10

Nuclear power plant operation - 1012

 

Autonomous Mobile Robots for Deep and Shallow Hydrodynamic Treatment of Concrete and Metal Surfaces

Zdenko Kovacic1, Matko Orsag1, Borislav Balac2, Stjepan Flegaric2, Kristijan Brkic2, Ante Mamic2

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia1

Inteco Robotics, Jaruscica 7, 10000 Zagreb, Croatia2

zdenko.kovacic@fer.hr

 

treatment of concrete and metal surfaces. The robot is moving on two tracks with a robot arm able to carry various water jet tools. The size of a robot allows easy transportation and manoeuvring in very confined spaces. Besides two tracks, the on-board system controls also a beam, a rotor tool, lance swing and pitch angles, and a tool-to-surface distance deviation. During operation, a tool is automatically adjusted towards a treated surface. Addition of a vacuum unit allows suction and storage of all water and material during treatment. Robot tests in real exploitation conditions have confirmed all advantages of the described robot design for possible water-based decontamination of metal and concrete surfaces in nuclear power plants.
A hydrodemolition mobile robot described in [1] has all characteristics needed to be an autonomous/teleoperated decontamination robot (Figure 1). It is equipped with advanced control functions and wireless remote control to execute deep & shallow high pressure water treatment (1000 – 3000 bar) of concrete and metal surfaces. Robot characteristics are suitable for collection of contaminated surface deposits, removal of contaminated epoxy coatings or for deep and surface decontamination of contaminated concrete.
Figure 1: Modular construction of R Jet 06 hydrodemolition robots
[1] Kovacic, Z.; Balac, B.; Flegaric, S.; Brkic, K.; Orsag, M. Light-weight Mobile Robot for Hydrodynamic Treatment of Concrete and Metal Surfaces“, The 1st International Conference on Applied Robotics for the Power Industry CARPI 2010, Montreal, Canada, 2010.






10.09.2014 11:30

Nuclear power plant operation - 1002

 

Capacity and Reliability of Slovenian Transmission Grid in Light of a Future Reasonable Production Increase from Krško NPP Units

Pavel Omahen

ELES, d.o.o., sistemski operater prenosnega elektroenergetskega omrežja , Hajdrihova 2, 1001 Ljubljana, Slovenia

pavel.omahen@eles.si

 

Actual and planned capacity of the Slovenian electrical power transmission grid is shown in light of feeding domestic consumers as well as huge Italian electrical energy deficit. In parallel with the interconnected grid transmission capacity its reliability course is shown in order to fulfill the required specifications for the Krško Nuclear Power Plant (NPP) external electrical power supply reliability.
After 2x400-kV line Beričevo-Krško construction Slovenian transmission grid has actual capacity, in sense of N-1 outages security criteria, to supply all domestic needs as well as to transmit up to 1500 MW power for cross border Italian customers. ELES, Slovenian transmission system operator has intention to fulfill the adequate Italian operator, Terna, plan to increase that capacity of power transfer up to 2000 MW to the end of year 2020. The main deficiency of such Slovenian transmission grid capacity increase lies in fact that there is no adequate surplus of stable electrical energy production neither inside Slovenian power system neither in the closest neighborhood. That fact pull out a need for a large distance energy transfer from surplus production sources in Balkan region and from north-east regions of Austria. Large distance energy transfer is no good neither in technical sense (grid stability decrease) neither in economic sense (loses and huge transport costs). Considering that huge energy production deficit in vicinity of Slovenia together with its huge transmission grid capacity there is no reasonable economic sense that second unit of Krško NPP is still not in construction process. Economy of Slovenia has not even close so great other chance to increase its gross income and pull out the country from its economic crisis. If we are talking of existing Krško NPP unit life-time extension as well as a great opportunity for second unit construction, we should not neglect the reliability of external electricity supply for those nuclear units. An approximate estimation of reliability course of the external power system energy supply is given from 1980 up to present time in order to define the future existence of necessary sources of electrical energy for that purpose.






10.09.2014 11:50

Nuclear power plant operation - 1003

 

Deploying Experiments to Support Internal Hazard Management

Harri Tuomisto

FORTUM, Power Division, P.O. Box 100, FIN-00048 FORTUM, Finland

harri.tuomisto@fortum.com

 

Hazards resulting from extreme natural events have been subject to very high interest since the Fukushima accident. Severe Accident Management programs have been studied in detail to check and complement them with regard to the external hazards. While this additional focus has added very important new insights, it is crucial to keep in mind that the essential content of nuclear safety still rests on internal hazard management.
Internal hazards and accident progression are dependent on the specific features of the given plant. Therefore, development of hazard management strategy necessitates proper plant-specific analyses and even specific thermal hydraulic experiments. In order to address these plant specific features and to obtain case-specific data, thermal hydraulic experiments were deployed by the utility to support management of various internal hazards and resolve various issues raised for the Loviisa VVER-440 units. The Loviisa plant configuration is quite unique, since the original VVER-440 design has been amended with ice condenser containment, specific reactor coolant pumps and many other features. After commissioning of the plant, many internal hazards were identified, many of them based on operating experience from the other plants as well as from the own experience.
A typical feature of internal hazards and a special concern of accident progression are that they challenge simultaneously more than one functional level of the defence-in-depth concept or penetrate more than one of the physical barriers against fission product releases. Internal hazards can themselves be initiating events, such as common cause failures, internal fires or floods, missiles, inhomogeneous boron dilution, or primary-to-secondary leakage accident. Internal hazards can also be created during the accident progression such as pressurized thermal shock, loop seal issue, boron crystallization, containment sump clogging, or inherent boron dilution mechanisms.
Cases of thermal hydraulic experiments and the internal hazards addressed with them will be discussed in the paper:
pressurized thermal shock - thermal mixing experiments
in adequate cooling during loss-of-coolant accidents due to
 - loop seal issue - loop seal experiments
- counter-curent flow limitation - CCFL experiments
boron crystallization - VEERA experiments
sump clogging - various sump screen experiments
non-uniform ice sublimation - ice condenser experiments

This paper forms an integral continuation to paper "Leveraging Specific Plant Features to Manage Internal Hazards" that the author presented in NENE2013.






10.09.2014 12:10

Nuclear power plant operation - 1004

 

ATHLET/KIKO3D results of the OECD/NEA benchmark for coupled codes on KALININ-3NPP measured data

György Hegyi, Andras Kereszturi, István Trosztel

Hungarian Academy of Sciences, Centre for Energy Research , P.O. Box 49, 1525 Budapest 114, Hungary

gyorgy.hegyi@energia.mta.hu

 

The reliable prediction of the global and local reactor parameters is important issue to achieve a high validation stage of the coupled neutron-physics/thermal-hydraulics system codes which nowadays are considered to be the state of art by performing of accident analysis. It is similarly important to be able giving an answer and assurance of the following questions:
• how good and correct are the applied models in the simulation tools,
• the accuracy of the numerical methods incorporated in the codes,
• the completeness of the methodology to describe the multi-physical processes,
• the correctness of the required input data to model the processes.

The paper gives a short summary of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) VVER-1000 (Kalinin-3) 'Switching-off of one Main Circulation Pump (MCP)' transient benchmark for evaluating coupled system codes. The large number of available real plant experimental data made these benchmark problems very valuable. The experiment is very well documented, its uncertainties are known and the measurements were performed with a quite high frequency. In the first exercise instead of the complicate reactor-physical model the ATHLET 2.1 code with its point kinetics method (its parameters given in the benchmark specification) was used. Then in the second exercise, the core and the vessel are modelled. The inlet and outlet core transient boundary conditions and appropriate parameterized cross section data are provided by the benchmark team. In the paper, our solution of the above detailed benchmark problem is presented and the results are compared to documented experimental data.






11.09.2014 08:30

Invited lectures - 103

 

The European Fusion Roadmap

Francesco Romanelli

JET-EFDA Culham Science Centre, OX14 3DB, Abingdon, United Kingdom

francesco.romanelli@jet.efda.org

 

With the reduction of CO2 emissions driving future energy policy, fusion can start market penetration around 2050 with up to 30% of electricity production by 2100. This requires an ambitious, yet realistic roadmap towards the demonstration of electricity production by 2050.

This talk describes the main technical challenges on the path to fusion energy. For all of the challenges candidate solutions have been developed and the goal of the programme is now to demonstrate that they will also work at the scale of a reactor.

The roadmap has been developed within a goal-oriented approach articulated in eight different Missions. For each Mission the critical aspects for reactor application, the risks and risk mitigation strategies, the level of readiness now and after ITER and the gaps in the programme have been examined with involvement of experts from the ITER International Organization, Fusion for Energy, EFDA Close Support Unites and EFDA Associates. High-level work packages for the roadmap implementation have been prepared and the resources evaluated.

ITER is the key facility in the roadmap and its success represents the most important overarching objectives of the EU programme.

A demonstration fusion power plant (DEMO), producing net electricity for the grid at the level of a few hundreds MW is foreseen to start operation in the early 2040s. Following ITER, it will be the single step to a commercial fusion power plant.

Industry must be involved early in the DEMO definition and design. The evolution of the programme requires that industry progressively shifts its role from that of provider of high-tech components to that of driver of the fusion development. Industry must be able to take full responsibility for the commercial fusion power plant after successful DEMO operation. For this reason, DEMO cannot be defined and designed by research laboratories alone, but requires the full involvement of industry in all technological and systems aspects of the design.

Europe should seek all the opportunities for international collaborations. Some of the ITER parties have a very aggressive programme in fusion and Europe can clearly benefit by the participation in the design, construction and operation of their facilities. Already the Broader Approach with Japan is a good example of a positive collaboration that can give further advantages on the time scale considered here.

The talk will also address the needs in the area of education and training and basic research.






11.09.2014 09:10

Nuclear fusion - 1112

 

Laboratory Experiments of Ammonia Production in Plasmas of H2 – N2 Mixtures

Aleksander Drenik, Rok Zaplotnik, Gregor Primc, Miran Mozetič, Anže Cigoj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

aleksander.drenik@ijs.si

 

In next-step thermonuclear fusion devices, a large fraction of the energy stored in the plasma will have to be dissipated through radiation, in order not to exceed the power-handling capabilities of the plasma facing components. Impurity seeding to promote radiation in the plasma edge has thus far yielded promising results. However, as the list of seeded impurities includes nitrogen, it gives rise to additional chemical and physical processes which are not yet fully understood. The main concern is the formation of ammonia, which could cause operational issues, and also issues linked to in-vessel fuel retention.
Experiments with nitrogen seeding and fusion devices indeed indicate that production of ammonia takes place, however the technological and operational constraints of a fusion device prevent an in-depth research of the processes that contribute to ammonia formation. To evaluate the role of the surface in ammonia production, a series of laboratory experiments were performed in an inductively coupled RF plasma reactor. At pressures between 30 and 50 Pa, samples of Fe, Ni and W were exposed to the downstream afterglow of a N2 – H2 mixture plasma, at varying reactor power and N2 concentration in the gas mixture. The reaction products were analyzed with a differentially pumped mass spectrometer. Concentrations of ammonia in the downstream atmosphere reached up to 1 vol % and exhibited strongest dependence on the RF generator power, and the N2 concentration in the gas mixture.






11.09.2014 09:30

Nuclear fusion - 1108

 

Sensitivity and Uncertainty Pre-analysis of the FNG Copper 14 MeV DT Benchmark for Fusion Relevant Cross-Section Validation

Ivan Aleksander Kodeli1, K. Kondo2, A. Klix2, Ulrich Fischer3, M. Angelone4,

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany2

Forschungszentrum-INR, Postfach 3640, 76021 Karlsruhe, Germany3

ENEA Fusion Division, Via E. Fermi 45, 1-00044 Frascati, Rome, Italy4

ivan.kodeli@ijs.si

 

At the Frascati neutron generator (FNG) of ENEA Frascati a copper mock-up will be irradiated in 2014 with 14 MeV neutrons. The objective of the benchmark is to provide the experimental data base required for the validation of the European nuclear data libraries JEFF, with the focus on the experimental validation of ITER design calculations, including the related uncertainties.
The planned Copper mock-up consists of a block of copper with a cross-sectional area of 60 cm x 60 cm and a depth of 70 cm. Several detectors will be introduced at 7 to 8 locations up to about 60 cm in the block. In addition to those originally planned also the Mn and Cu reactions as possible candidates for fast and thermal neutron spectra monitoring were studied.
The Copper benchmark is the last in the series of benchmarks performed at FNG in the scope of the EC fusion activity since the 1990’s, such as FNG Bulk Shield benchmark, FNG Streaming, FNG SiC, FNG Tungsten, FNG-HCPB tritium breeding module, and FNG-HCLL tritium breeding module. Like these previous benchmarks, also the preparation of the FNG-Cu experiment started with the pre-analyses which have been performed both by deterministic and Monte Carlo codes.
The pre-analysis includes the simulation of the planned copper integral benchmark experiment in order to determine the sensitivity of the reaction rates to be measured in the mock-up to the underlying cross sections, and the associated uncertainties. This permits to optimize the geometry, the detector positions and the choice of activation reactions, and in the post-analysis phase to interpret the results of the measurements and the calculations and to conclude on the quality of available nuclear data, such as the recent JEFF evaluations. The pre-analysis was accomplished by means of cross-section sensitivity and uncertainty analysis using the deterministic code SUSD3D (at JSI) and the Monte Carlo code MCSEN5 (at KIT). Cumulative reaction rate integrals, their sensitivity to the cross sections, as well as the uncertainties were estimated for selected detector working positions in the copper block, in particular for deep penetrations.






11.09.2014 09:50

Nuclear fusion - 1103

 

Characterization of the neutron field for the JET torus

Igor Lengar1, Luka Snoj1, Aljaž Čufar1, Brian Syme2, Paola Batistoni3, Sean Conroy4

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Culham Science Center JET-EFDA, OX14 3DB, Abingdon, United Kingdom2

ENEA Fusion Division, Via E. Fermi 45, 1-00044 Frascati, Rome, Italy3

Uppsala University Department of Physics and Astronomy EURATOM-VR Ass., Box 516, SE-75120 Uppsala, Sweden4

igor.lengar@ijs.si

 

A new Deuterium-Tritium campaign (DTE2) is planned at JET in 2017, with a proposed 14 MeV neutron budget nearly an order of magnitude higher than any previous DT campaigns. At the expected plasma performance, the neutron flux on the first wall achieves levels comparable to those expected in ITER between the blanket and the vacuum vessel (> 10^17 n/s•m2). The neutron fluence will be up to 10^20 n/m2. This level of neutron flux/fluence will offer the opportunity to irradiate samples of functional materials used in ITER diagnostics, and of materials used in the manufacturing of the main in-vessel ITER components, to assess the degradation of the physical properties and the neutron induced activities, respectively.
A number of neutronics experiment and measurements on irradiated samples are planned during the DTE2 campaign. Extensive neutron transport calculations are required in support of these experiments which are performed with the MCNP code and a verified JET model.
The purpose of the present work is to characterize the neutron and gamma ray field inside the JET device during DT plasma operations. An analysis of the neutron/gamma ray flux, energy spectrum and dose rate levels is performed at selected irradiation locations, such as the neutron activation irradiation ends, the new long term irradiation stations located inside the vessel and inside a circular horizontal port, where samples would be exposed to the maximum neutron flux or fluence. The neutron flux and spectrum at different irradiation ends are calculated and compared. The study will offer a unique opportunity for comparing results of calculations with measurements, when the latter become available after the DTE2 campaign.






11.09.2014 10:10

Nuclear fusion - 1104

 

Hydrogen atom recombination study on W by vibrational spectrometer in differential pumping mode

Anže Založnik, Sabina Markelj, Iztok Čadež

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

anze.zaloznik@ijs.si

 

Detailed knowledge of surface processes such as sputtering, reflexion and recombination on plasma facing components is of high importance for understanding and modelling edge plasma [e.g. 1]. This requires information on the kind of emitted particles as well as their state of excitation. While this was to more extent studied for carbon materials much less is known for metals (ITER relevant - W) and especially at higher temperatures relevant to divertor operation and transient heat load phenomena.
The emphasis of our work is on the study of production of vibrationally excited molecules by hydrogen atom (H and D) recombination. For this purpose we have designed and built a spectrometer for determination of vibrational excitation of hydrogen molecules applicable to all its isotopologues. The diagnostic method is based on detection of low energy H- and D- ions produced by electron impact on hydrogen molecule in the 0-20 eV energy range. We have recently upgraded the spectrometer by single-stage differential pumping thus separating the detector and reaction chamber. In this case the hydrogen molecule detector and recombination zone are connected only through a narrow collimating slit allowing molecules emitted from the sample surface to be directly observed by spectrometer while lowering the detection background. The new, high resolution electron gun is also tested in order to allow better separation of individual contributions to detected ion yield. The recombination set up is beam-like, where the atoms from an atom beam source are hitting the surface under certain angle and molecules produced by recombination on the surface are detected under other angle. The differentially pumped spectrometer is used for the studies of H and D recombination on W and determination of the temperature dependence of production of excited molecules in the temperature range from 300 – 600 K. Influence of seeding impurities, nitrogen in particular, on hydrogen atom recombination will be studied as well. The new results of measurements on the upgraded set-up will be presented and compared to previous measurements.

[1] Schneider et al.: Plasma edge physics with B2-Eirene, Contrib. Plasma Phys. 46 (2006) 3.






11.09.2014 10:50

Materials, integrity and life management - 1201

 

TEM Foils Preparation from Irradiated Austenitic Stainless Steel - An applied methodology to attain 1mm samples

Hygreeva Kiran Namburi1, Petra Bublíková1, Vít Rosnecký1, Jan Michalička1, Eliška Keilová2, Jan Kočík2, Miroslava Ernestová2

Research centre Rez, Hlavni 130, 250 68 Husinec-Řež, Czech Republic1

ÚJV Řež, a.s., Hlavní 130, Řež, 250 68 Husinec, Czech Republic2

nab@cvrez.cz

 

After FUKUSHIMA disaster, the concern regarding necessity for structural materials with improved properties and safe operation of the Nuclear Power Plants (NNPs) has increased enormously. This kept great challenging tasks for researchers to assess current and develop new materials that are more appropriate for structural stability of NNP’s. Structural and System diagnostics department at Research Center Rez focuses to study the irradiated and un-irradiated materials behavior and their ageing aspects. This includes analysis of structural and core internal materials.
In the surveyed paper, we present work that is related to post irradiation examination of 300-series austenitic stainless steel taken from Reactor Vessel Internals (RVI) of PWR. High neutron irradiation dose in NNP’s leads to a degradation of microstructure of the material in a nano-metric scale. Due to the effect of complex radiation damage the RVI material may turn out to be hardened, swelled and/or to be sensitive to Irradiation Assisted Stress Corrosion Cracking (IASCC). Hence, there is prominent research interest to characterize the irradiated materials that are subjected to the long life span conditions and to understand the physical basis of the degradation mechanisms. Microstructural characterization of neutron irradiated radioactive materials by TEM requires enhanced sample preparation methodologies, which commonly needs general improvements regarding a particular experiment to be performed. In our study, we have had to develop and apply tailored methodology specialized in 1 mm TEM thin foil preparation from a deformed shank of a broken miniaturized tensile specimen with shank diameter 2 mm. The uncommon TEM foil size was chosen firstly because of the small shank diameter, secondly because of high radioactivity of the studied material, which was irradiated in a nuclear reactor. The reduction of the TEM foil radioactivity to minimum is crucial when an EDX chemical analysis is performed. The radioactivity of classic 3 mm TEM foils from the same material is supposed to be too high for EDX measurements and also it may decrease the EDX detector life-time rapidly.
The paper describes whole process from bulk sample handling, including e.g. remote-controlled material cutting in shielded hot-cells and TEM disk polishing in glow-boxes, up to the main final procedure of electrolytic-polishing of electron transparent 1 mm TEM foils. The chapter of sample preparation is followed by the results of TEM microanalysis of radiation-induced defects.






11.09.2014 11:10

Materials, integrity and life management - 1202

 

Impact on Neutronic Calculation of Thermomechanical Expansion of Reactor Vessel Internals

Cecile-Aline Gosmain1, Sylvain Rollet2, Michel Tommy-Martin1, Isabelle Rupp1, Jean-Luc Flejou1

EDF - R&D, 1, avenue du General de Gaulle, 92141 Clamart Cedex, France1

EDF/SEPTEN, 1-14 Avenue Dutrievoz,, F-69629 Villeurbanee Cedex, France2

cecile-aline.gosmain@edf.fr

 

In the framework of surveillance program dosimetry, one of the main parameter of the Reactor Pressure Vessel (RPV) integrity assessment is the determination of the fracture toughness. This parameter is derived from the fast neutron fluence on the RPV. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel.
The reference calculation of the neutron flux over 1MeV for a 900MWe reactor is based on a 3D modelisation of one eighth of the reactor from the core to the primary concrete, including internals and capsules. It considers the geometrical values of pins, fuel lattice and internals at cold shutdown. The water in the bypass region is divided in four sectors between two formers, one uniform water temperature is considered in each sector. The axial distribution of water temperature in the bypass region is defined to match the coolant inlet and outlet temperature. This paper presents a study on the sensitivity of the neutron flux over 1MeV on a 900MWe reactor RPV and at surveillance capsules when considering a refine mesh of water temperature in the bypass calculated by □ and n energy deposition and the calculated thermomechanical expansion of the internals.
To carry out this study, a four steps methodology has been developed :
- the first step of the methodology consists in considering in the reference calculation the pin by pin power calculated by COCCINELLE (deterministic code) of a given campaign to evaluate the □ and n energy deposition in solids and fluids by a 3D stochastic code TRIPOLI4;
- the second step of the methodology is dedicated to the calculation of the temperatures of fluid (water) and solid (internals) by Code_Saturne and SYRTHES codes based on □ and n energy deposition previously obtained ;
- the third step consists in the calculation of thermomechanical expansion of the internals by code Aster based on the temperature fields provided by Code_Saturne ;
- the last step of this methodology is dedicated to the calculation of the neutron flux over 1MeV at RPV hot spot and at surveillance capsules and reaction rates of activation and fissile dosimeters considering the fuel pins and lattice expansions in the core, the thermohydraulic load in a refine mesh in front of RPV and surveillance capsules and the thermomechanical expansion of the internals.

The consideration of a refine mesh of thermohydraulic load in the bypass region calculated by code_Saturne by taking into account the energy deposition in fluid and solid leads to an impact on neutron flux over 1MeV at RPV hot spot and surveillance capsules far less than 1%.The cumulative impact of the consideration of the thermohydraulic load in the bypass region and the thermomechanical expansion of the fuel pins, lattice and the internals on the calculations of the neutron flux over 1MeV is limited to less than 2% at the RPV hot spot and surveillance capsules.
These results allow us to conclude that the reference 3D modelisation of one eighth of the reactor from the core to the primary concrete, including internals and capsules which considers the geometrical values of pins, fuel lattice and internals at cold shutdown and a four sectors mesh of water temperature between two formers is sufficient at this level of refinement to calculate the neutron flux over 1MeV at RPV and surveillance capsules.






11.09.2014 11:30

Materials, integrity and life management - 1203

 

K/J value estimation of specimen containing dissimilar metal welds

Igor Simonovski1, Oliver Martin1, Gangadhar Machina1, Szabolcs Szavai2, Robert Beleznai2

European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands1

Bay Zoltan Nonprofit Ltd. for Applied Research, Institute for Logistics and Production Engineering, Igloi ut 2, H-3519 Miskolc, Hungary2

igor.simonovski@ec.europa.eu

 

In 2012 a MULTIMETAL research project was started within the 7th Framework Programme for Research and Technological Development of the European Union. The aim of the project is to develop a standard for measuring the fracture resistance of multi-metallic specimen and the development of harmonized procedures for dissimilar metal welds (DMWs) brittle and ductile integrity assessment. DMW connect ferritic and austenitic stainless steels and are present in a number of primary piping systems in light water reactors. They contain a number of different material zones, i.e. ferritic zone, austenitic zone, weld material and heat affected zones which differ significantly in their material properties and fracture behavior. The MULTIMETAL project involves an extensive test program in which standard geometry specimens (CT, SENB, SENT) are cut from mock-ups containing DMWs resembling real DMWs from NPPs in terms of geometry, material and weld procedure. These specimens contain different material zones as described above. A part of the project includes performing numerical analyses to estimate fracture parameters such as stress intensity factors K, J-integral values and extraction of eta factors for different crack positions and crack lengths. In this paper the results of the numerical analyses performed by JRC so far in multi-metallic specimens are presented.






11.09.2014 11:50

Materials, integrity and life management - 1206

 

Automated Motion of the Forerunner Mobile Manipulator Over the Tube Sheets

Tomislav Tomašić1, Andrej Jokić2, Ante Bakic1

INETEC-Institute for Nuclear Technology, Dolenica 28, 10 250 Zagreb, Croatia1

Faculty of Mechanical Engineering, Laboratory for Fluid Dynamic and Thermodynamics, Aškerčeva 6, 1000 Ljubljana, Slovenia2

tomislav.tomasic@inetec.hr

 

This paper describes the algorithm for automated motion over the tube sheets of steam generators implemented in the Forerunner manipulator. Forerunner belongs to a group of lightweight mobile manipulators for eddy current testing and repair actions of heat exchangers, primarily vertical steam generators in nuclear power plants of PWR type. Its task is to move along the tube sheet and position different tools at the entrance of the desired pipe. The motion is achieved by the two main axis, rotation and translation, while fixation to the tubes is accomplished by four pneumatic grippers. Additional two axes are employed for auxiliary axis manipulation with a quick tool exchanger at the end. During the motion its position and orientation are constantly changing. In order to facilitate the manipulator navigation, the high level of autonomy is integrated in the system so that operator only selects a group of tubes to be tested while the algorithms for automated motion decompose the command on sequential series of moves that must be made based on the current position. Thereby plugged tubes are avoided for gripping and possible collision with a wall or another manipulator is taken into account. As secondary position verification, machine vision system is implemented to independently count the actual position of the manipulator and alerts operator in case of any irregularities. Although its motion finally looks very intuitive, searching for the optimal path among several thousand tubes represents a real challenge from the mathematical point of view. The robustness of the algorithm can be summarized in the following statement: If there is a solution to make the manipulator positioning in a given group of tubes, the algorithm will find it.






11.09.2014 12:10

Materials, integrity and life management - 1205

 

Extended Crystal Plasticity Finite Element Approach for Neutron Irradiated Austenitic Stainless Steels

Samir El Shawish, Leon Cizelj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

samir.elshawish@ijs.si

 

During long-time neutron irradiation exposure in pressurized water reactors, significant changes of the mechanical behaviour of materials used in reactor core internals (made of 300 series austenitic stainless steels) are observed, including irradiation induced hardening and softening, loss of ductility and toughness, as well as formation of plastic instabilities. These evolutions are commonly ascribed to the formation of high density nano-sized irradiation defect clusters, mainly interstitial Frank loops in stainless steel. Ductility loss is mainly ascribed to increased plastic strain localization. In many cases, shear bands develop in the form of clear channels where irradiation defect clusters are progressively removed by interaction with mobile dislocations.
In this work, a micromechanical crystal plasticity model developed by CEA, France is implemented in Abaqus subroutine UMAT to simulate the nonlinear mechanical response of austenitic stainless steels subjected to neutron irradiation. The proposed model, based on dislocation dynamics inferred mechanisms and finite strain theory, is able to capture the irradiation-induced hardening followed by softening due to the formation of defect free channels on each slip plane during plastic deformation. Results for polycrystalline simulations of non-irradiated and irradiated AISI 304 steels show a good adequacy to reproduce the tensile behaviour observed at different doses.





Login form
- Login
- Create an account
- Forgot your password?

 

Dates to remember
30 May   Abstracts submittal
30 June   Abstracts acceptance
31 July   Reduced registration fee
31 August   Full length papers
30 September   Results
of peer review
30 October   Camera ready papers
 

Sponsors

Username: Password: