The Findings from the OECD/NEA/CSNI UMS (Uncertainty Method Study)

Within licensing procedures there is the incentive to replace the conservative requirements for code application by a “best estimate” concept supplemented by an uncertainty analysis to account for predictive uncertainties of code results. Methods have been developed to quantify these uncertainties. The Uncertainty Methods Study (UMS) Group, following a mandate from CSNI (Committee on the Safety of Nuclear Installations) of OECD/NEA (Organization for Economic Cooperation and Development / Nuclear Energy Agency), has compared five methods for calculating the uncertainty in the predictions of advanced “best estimate” thermal-hydraulic codes. Most of the methods identify and combine input uncertainties. The major differences between the predictions of the methods came from the choice of uncertain parameters and the quantification of the input uncertainties, i.e. the wideness of the uncertainty ranges. Therefore, suitable experimental and analytical information has to be selected to specify these uncertainty ranges or distributions. After the closure of the Uncertainty Method Study (UMS) and after the report was issued comparison calculations of experiment LSTF-SB-CL-18 were performed by University of Pisa using different versions of the RELAP 5 code. It turned out that the version used by two of the participants calculated a 170 K higher peak clad temperature compared with other versions using the same input deck. This may contribute to the differences of the upper limit of the uncertainty ranges. A ‘bifurcation’ analysis was also performed by the same research group also providing another way of interpreting the high temperature peak calculated by two of the participants

CIAU Method for Uncertainty Evaluation for System Thermal-Hydraulic Code Calculations

Best-Estimate calculation results from complex thermal-hydraulic system codes (like Relap5, Cathare, Athlet, Trace, etc..) are affected by unavoidable approximations that are un-predictable without the use of computational tools that account for the various sources of uncertainty. Therefore the use of best-estimate codes within the reactor technology, either for design or safety purposes, implies understanding and accepting the limitations and the deficiencies of those codes. Uncertainties may have different origins ranging from the approximation of the models, to the approximation of the numerical solution, and to the lack of precision of the values adopted for boundary and initial conditions. The amount of uncertainty that affects a calculation may strongly depend upon the codes and the modeling techniques (i.e. the code’s users). A consistent and robust uncertainty methodology must be developed taking into consideration all the above aspects. The CIAU (Code with the capability of Internal Assessment of Uncertainty) and the UMAE (Uncertainty Methodology based on Accuracy Evaluation) methods have been developed by University of Pisa (UNIPI) in the framework of a long lasting research activities started since 80’s and involving several researchers. CIAU is extensively discussed in the available technical literature, Refs. [1, 2, 3, 4, 5, 6, 7], and tens of additional relevant papers, that provide comprehensive details about the method, can be found in the bibliography lists of the above references. Therefore, the present paper supplies only ‘spot-information’ about CIAU and focuses mostly on the applications to some cases of industrial interest. In particular the application of CIAU to the OECD BEMUSE (Best Estimate Methods Uncertainty and Sensitivity Evaluation, [8, 9]) project is discussed and a critical comparison respect with other uncertainty methods (in relation to items like: sources of uncertainties, selection of the input parameters and quantification of their uncertainty ranges, ranking process, etc.) is presented

Methodology for Pressurized Thermal Shock Analysis in Nuclear Power Plant

The relevance of the fracture mechanics in the technology of the nuclear power plant is mainly connected to the risk of a catastrophic brittle rupture of the reactor pressure vessel. There are no feasible countermeasures that can mitigate the effects of such an event that impair the capability to maintain the core covered even in the case of properly functioning of the emergency systems. The origin of the problem is related to the aggressive environment in which the vessel operates for long term (e.g. more than 40 years), characterized by high neutron flux during normal operation. Over time, the vessel steel becomes progressively more brittle in the region adjacent to the core. If a vessel had a preexisting flaw of critical size and certain severe system transients occurred, this flaw could propagate rapidly through the vessel, resulting in a through-wall crack. The severe transients that can lead the nuclear power plant in such conditions, known as Pressurized Thermal Shock (PTS), are characterized by rapid cooling (i.e., thermal shock) of the a part of the internal reactor pressure vessel surface that may be combined with repressurization can create locally a sudden increase of the stresses inside the vessel wall and lead to the suddenly growth of the flaw inside the vessel thickness. Based on the long operational experience from nuclear power plants equipped with reactor pressure vessel all over the world, it is possible to conclude that the simultaneous occurrence of critical-size flaws, embrittled vessel, and a severe PTS transient is a very low probability event. Moreover, additional studies performed at utilities and regulatory authorities levels have shown that the RPV can operate well beyond the original design life (40 years) because of the large safety margin adopted in the design phase. A better understanding and knowledge of the materials behavior, improvement in simulating in a more realistic way the plant systems and operational characteristics and a better evaluation of the loads on the RPV wall during the PTS scenarios, have shown that the analysis performed during the 80’s were overly conservative, based on the tools and knowledge available at that time. Nowadays the use of best estimate approach in the analyses, combined with tools for the uncertainty evaluation is taking more consideration to reduce the safety margins, even from the regulatory point of view. The US NRC has started the process to revise the technical base of the PTS analysis for a more risk-informed oriented approach. This change has the aim to remove the un-quantified conservatisms in all the steps of the PTS analysis, from the selection of the transients, the adopted codes and the criteria for conducting the analysis itself thus allow a more realistic prediction. This change will not affect the safety, because beside the operational experience, several analysis performed by thermal hydraulic, fracture mechanics and Probabilistic Safety Assessment (PSA) point of view, have shown that the reactor fleet has little probability of exceeding the limits on the frequency of reactor vessel failure established from NRC guidelines on core damage frequency and large early release frequency through the period of license extension. These calculations demonstrate that, even through the period of license extension, the likelihood of vessel failure attributable to PTS is extremely low (≈10-8/year) for all domestic pressurized water reactors. Different analytical approaches have been developed for the evaluation of the safety margin for the brittle crack propagation in the rector pressure vessel under PTS conditions. Due to the different disciplines involved in the analysis: thermal-hydraulics, structural mechanics and fracture mechanics, different specialized computer codes are adopted for solving single part of the problem. The aims of this chapter is to present all the steps of a typical PTS analysis base on the methodology developed at University of Pisa with discussion and example calculation results for each tool adopted and their use, based on a more realistic best estimate approach. This methodology starts with the analysis of the selected scenario by mean a System Thermal-Hydraulic (SYS-TH) code such as RELAP5 [2][3], RELAP5-3D [1], CATHARE2 [4][6], etc. for the analysis of the global behavior of the plant and for the evaluation of the primary side pressure and fluid temperature at the down-comer inlet. For a more deep investigation of the cooling load on the rector pressure vessel internal surface at small scale, a Computational Fluid Dynamics (CFD) code is used. The calculated temperature profile in the down-comer region is transferred to a Finite Element (FE) structural mechanics code for the evaluation of the stresses inside the RPV wall. The stresses induced by the pressure in the primary side are also evaluated. The stress intensity factor at crack tip is evaluated by mean the weight function method based on a simple integration of the stresses along the crack border multiplied by the weight function. The values obtained are compared with the critical stress intensity factor typical of the reactor pressure vessel base material for the evaluation of the safety margin

BEPU-FSAR: A new paradigm in Nuclear Reactor Safety

Nuclear safety has become a technology following extraordinary industrial investments since the 50’s. Events in the last decades occurring in the Three Mile Island Unit-2, Chernobyl Unit-4 and Fukushima Units 1-3 have challenged the sustainability of nuclear technology and undermined the trust of the public, of the decision makers and even of the scientific community toward nuclear safety. Innovative ideas and proposals are possibly needed to restore the confidence and escape the irreversible loss of competence which also feeds the further degradation of sustainability for the technology. Efforts have been completed by the technological community following each of the disasters and ended-up in reinforcements of the Engineered Safety Features (ESF) and of Safety Barriers. The FSAR shall be seen as the compendium of information concerned with the safety of the specific NPP and includes the demonstration of acceptability of the NPP against the rules and related criteria established for the Country. The Safety Analysis is part of the licensing process and is documented in the FSAR. The role of safety analysis as a key part of Nuclear Reactor Safety Technology (NRST) is discussed first and the innovation connected with the words ‘independent’ and ‘systematic Best Estimate Plus Uncertainty’ is presented, as well as the propose of a BEPU-FSAR. To perform an entire FSAR based on BEPU, a homogenization of the analysis is proposed. The first step towards BEPU-FSAR requires identification and characterization of the parts where numerical analyses are needed (so-called BEPU topics). The next step is to create a list of key technological areas, so-called key disciplines and their related key topics and then an overview of the currently computational activities in each technological area. Based on the finalized BEPU applications one can conclude that this methodology is feasible, which encourage to extended its range of use to the other technological areas of FSAR, and therefore to demonstrate the industrial worth and interest. The future step of this work will mainly be focused on the propagation of this expertise into the remaining technical areas of FSAR, adding new knowledge and therefore creating coherent and rigorous background of the BEPU-FSAR methodology. The purpose of the present paper is to provide the background and the framework for the proposed approach

International Training Program in Support of Safety Analysis: 3C S.UN.COP – Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the ‘user effect’ and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Eleven seminars have been held at University of Pisa (two in 2004), at The Pennsylvania State University (2004), at the University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006), in Buenos Aires, Argentina (October 2006), requested by Autoridad Regulatoria Nuclear (ARN), Nucleoelectrica Argentina S.A (NA-SA) and Comisión Nacional de Energía Atómica (CNEA), at the College Station, Texas A&M, (January-February 2007), in Hamilton and Niagara Falls, Ontario (October 2007) requested by Atomic Energy Canada Limited (AECL), Canadian Nuclear Society (CNS) and Canadian Nuclear Safety Commission (CNSC), in Petten, The Netherlands (October 2008) in cooperation with the Institute of Energy of the Joint Research Center of the European Commission (IE-JRC-EC), at the Royal Institute of Technology, Stockholm (October 2009) and in Petten, The Netherlands (October 2010) in cooperation with the Institute of Energy of the Joint Research Center of the European Commission (IE-JRC-EC). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed ‘permanent’ stepwise approach to user training. The 3D S.UN.COP 2010 at IE-JRC was successfully held with the attendance of 23 participants coming from more than 10 countries and 20 different institutions (universities, vendors and national laboratories). More than 30 scientists (coming from more than 10 countries and 20 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. The eleventh seminar has been held (March 2011) in Wilmington, North Carolina, involving more than 30 scientists between lecturers and code developers (http://www.nrgspg.ing.unipi.it/3dsuncop/)

International Standard Problem No 50 – The University of Pisa contribution

The present paper deals with the participation of the University of Pisa in the last International Standard Problem (ISP) focused on system thermal hydraulic, which was led by the Korean Atomic Energy Research Institution (KAERI). The selected test was a Direct Vessel Injection (DVI) line break carried out at the ATLAS facility. University of Pisa participated, together with other eighteen institutions, in both blind and open phase of the analytical exercise pursuing its methodology for developing and qualifying a nodalization. Qualitative and quantitative analysis of the code results have been performed for both ISP-50 phases, the latter adopting the Fast Fourier Transfer Based Method (FFTBM). The experiment has been characterized by three dimensional behavior in downcomer and core region. Even though an attempt to reproduce these phenomena, by developing a fictitious three-dimensional nodalization has been realized, the obtained results were generally acceptable but not fully satisfactory in replicating 3D behavio

Main Results of the OECD BEMUSE Programme

The BEMUSE (Best Estimate Methods Uncertainty and Sensitivity Evaluation) Programme promoted by the Working Group on Analysis and Management of Accidents (WGAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI) represents an important step towards reliable application of high-quality best-estimate and uncertainty and sensitivity evaluation methods. The methods used in this activity are considered to be mature for application, including licensing processes. Skill, experience and knowledge of the users about the applied suitable computer code as well as the used uncertainty method are important for the quality of the results

Capabilities of TRANSURANUS Code in Simulating BWR Super-Ramp Project

After one-two years of normal operation in a LWR, the fuel–cladding gap may close, as a result of as a result of several phenomena and processes, including the different thermal expansion and swelling of both the fuel and the cladding (Pellet Cladding Interaction). In this equilibrium state, a significant increase of local power (like a transient power ramp, i.e. power increase in the order of 100kW/m-h), induces circumferential stresses in the cladding. In presence of corrosive fission products (i.e. iodine) and beyond specific stress threshold, material dependent, cracks typical of stress corrosion may appear and grow-up: this phenomenon is called stress corrosion cracking (SCC). The cracks of the cladding may spread out from the internal surface, causing the fuel failure. The objective of the activity (performed in the framework of the IAEA CRP FUMEX III), is to validate the TRANSURANUS models relevant in predicting the fuel failures due to PCI/SCC during power ramps. Focus is given on the main phenomena, which are involved or may influence the cladding failure behavior. The database selected is the Studsvik BWR Super-Ramp Project, which belongs to the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database” by OECD/NEA. It comprises the data of sixteen BWR fuel rods, that have been modeled and simulated with suitable input decks. The burn-up values range between 28 and 37 MWd/kgU. Eight rods, of KWU standard type, are subjected to fast ramps, the remaining rods experience slow ramps and are of standard GE type

Application of Best Estimate Plus Uncertainty (BEPU) Methodology in a Final Safety Analysis Report (FSAR) of a Generic Plant

The licensing process of a nuclear power plant is motivated by the need to protect humans and the environment from ionizing radiation and, at the same time, sets out the basis for the design and determining the acceptability of the plant. An important part of the licensing process is the realization of accident analysis related to the design basis, which should be documented in the Final Safety Analysis Report (FSAR). There are different options on accidents calculation area by combining the use of computer codes and data entry for licensing purposes. One is the Best Estimate Plus Uncertainty (BEPU), which considers realistic input data and associated uncertainties. Applications of BEPU approaches in licensing procedures were initiated in the 2000s, first to analysis of Loss of Coolant Accident (LOCA), and then to the accident analysis as a whole, documented in Chapter 15 of the FSAR. This work has as main objective the implementation of BEPU methodology in all analyses contained in FSAR, through the homogenization of the analytical techniques and identification of key disciplines and key topics in the licensing process