Martina Adorni

The Fukushima Event: The Outline and the Technological Background

The paper deals with the evaluation of the Fukushima-Daiichi Nuclear Power Plant (NPP) accident in Units 1 to 4: an attempt is made to discuss the scenario within a technological framework, considering precursory documented regulations and predictable system performance. An outline is given at first of the NPP layout and of the sequence of major events. Then, plausible time evolutions of relevant quantities in the different Units, is inferred based on results from the application of numerical codes. Scenarios happening in the primary circuit and containment (three Units involved) are distinguished from scenarios in spent fuel pool (four Units involved). Radiological releases to the environment and doses are approximately estimated. The event is originated by a natural catastrophe with almost simultaneous occurrence of earthquake and tsunami. These caused heavy destruction in a region in Japan much wider than the land around the NPP which was affected by the nuclear contamination. Key outcome from the work is the demonstration of strength for nuclear technology; looking at the past, misleading Probabilistic Safety Assessment (PSA) data and inadequacy in licensing processes have been found. Looking into the future keywords are Emergency Rescue Team (ERT), Enhanced Human Performance (EHP), and Robotics in Nuclear Safety and Security (RNSS).

Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents), a primary-to-secondary leak, and a parametric study (natural circulation test) aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

Assessment of Transuranus fuel performance code against Studsvik Inter-Ramp BWR database

The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation - International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for model development and code validation. This database includes the data set of the Studsvik Inter-Ramp BWR Project. The objectives of the project are to establish the failure-safe operating limits and the failure mechanism and associated phenomena, during power ramp tests, by varying the design parameters (i.e. cladding heat treatment, gap thickness and fuel density). The experimental data are used for the assessment of the Fission Gas Release (FGR) models implemented in the TRANSURANUS code versions “v1m1j07” and “v1m1j08”. The starting point of the activity is the availability of a “new” transient fission gas release model, the “TFGR model”, specifically implemented in the last code version, to cover power ramp conditions. The paper presents the complete set of simulations of all twenty rods irradiated in the R2 research reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of geometric parameters and the choice of the different code options, relevant to model the FGR, on results.© 2009 ASME

Modeling Large Break-LOCA: In Reactor Fuel Bundle Materials Test MT-4 and MT-6A

When creating power from nuclear fission, the fuel matrix and its cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well.Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. In this connection, OECD/NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation (IFPE)”. This database includes the data set of the projects MT-4 and MT-6A analyzed in the current paper.The MT-4 test bundle simulated a 6×6 section of a 17×17 3% enriched, full-length non-irradiated PWR fuel assembly. There were 20 non-pressurized guard fuel rods to isolate the 12 central, pressurized tests rods; the four corner rods were deleted. In the MT-6A test, the 20 guard rods used in the previous tests were replaced with 9 pressurized thus, a total of 21 test rods were in MT-6A. Only limited destructive post irradiation examination was performed on these two tests.The objective of the activity is the validation of TRANSURANUS “v1m1j09” code in predicting fuel and cladding behavior under LOCA conditions using the experimental databases MT-4 and MT-6A. It is pursued assessing the capabilities of the code models in simulating the phenomena and parameters involved, such as: pressure trend in the fuel rod, cladding creep, α to β-phase phase transformation, oxidation, geometry changes and finally failure prediction. The analysis is aimed at having a comprehensive understanding of the applicability and limitations of the code in the conditions of the experiments. Finally, probabilistic calculations are performed to complete the analysis.The objective of the activity is fulfilled addressing the behavior of two equivalent full lengths fuel rods, one for each test., suitable for the assessment of TU code versions “v1m1j09”.Copyright

Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis.This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient.A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.Copyright

A Procedure to Address the Fuel Rod Failures During LB-LOCA Transient in Atucha-2 NPP

Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LBLOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

Selected Papers from the International Conference on Nuclear Energy for New Europe 2007

We are proud to introduce this special issue of the journal “Science and Technology of Nuclear Installations,” which is devoted to selected papers from the International Conference “Nuclear Energy for New Europe 2007.” The conference was organized by the Nuclear Society of Slovenia and Jožef Stefan Institute, and took place from September 10 to 13, 2007, in Portorož, Slovenia. Although it began modestly as the yearly meeting of the Nuclear Society of Slovenia, this annual conference has gradually become a truly international meeting of professionals, researchers, academics, members of regulatory bodies, and others involved in the peaceful use of nuclear energy. The International Conference “Nuclear Energy for New Europe 2007” was the 16th in the series. The conference was attended by 175 registered participants from 22 countries. Altogether, 112 papers were presented: 48 orally during plenary sessions and 64 as posters. The conference thus proved again to be an international forum for the exchange of ideas from various topics related to nuclear energy. This is the first time that extended versions of selected papers from the conference are published in a special issue of an international journal. Although the idea of publishing such a special issue has been discussed for some years, it was not realized until now due to the lack of opportunity. The journal “Science and Technology of Nuclear Installations” proved to be a suitable publication for such an undertaking. After the closing of the conference, authors of papers, presented at the conference, were invited to submit extended papers for publication in the special issue. Papers that were submitted then went through an extensive peer-review process to ensure a high quality of the publications. As a result, 13 papers are included in this special issue. The papers belong to various topics that were considered at the conference. We take this opportunity to express our gratitude to the conference organizers, to the members of the program committee who performed the first selection of papers and elaborated the conference program, and the reviewers who contributed to the final form of this special issue. Last but not least, we would like to thank Hindawi Publishing Corporation for offering the possibility for this special issue and the members of its editorial staff for all the assistance they have provided.

Analysis of a Total Flow Blockage of a Fuel Assembly in a Typical MTR Research Reactor by RELAP5/MOD3.3

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool Type Research Reactor [1]. The reactor is core cooled and moderated by downward forced circulation of light water. The transient herein considered is the related to partial and total obstruction of a single Fuel Assembly (FA) cooling channel. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FA integrity. Two cases are analysed to emphasize the severity of the accident. The first one is a partial blockage of a single FA considering four different obstruction levels: 50%, 75%, 85% and 95% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FA. This study constitutes the first step of a larger work which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation.Copyright