Objective of the Courses

The MMARS-3 Courses will provide a transfer of experience and know-how from recognized experts for the application of Models and Methods adopted for the advanced analysis of Nuclear Power Plants.
Eight different courses consisting of 35 hours each are offered:

  • Advanced RELAP5 Training: ITF and NPP Safety Analysis

The hands-on training course is directed toward advanced RELAP5 users with system thermal-hydraulics background. The course will provide information on the nodalization techniques of components in Integral Test Facility (ITF) and on the qualification process of a system thermal-hydraulic calculation, including the qualitative and quantitative evaluation of the accuracy. The so called Kv-scaled calculation of a NPP to the selected test in a ITF is also part of the course as well as the identification of simple and complex errors in a NPP input nodalization. Finally the course provides with lectures which give an overview of the code assessment process and of the method to evaluate the uncertainty of system code calculations.



  • Methods and Codes for Cross Section Generations (Deterministic Methods)

The training is intended for nodal reactor physics code users who need to prepare their own cross-section for nodal LWR reactor core analysis. The course will provide an overview of the methods and codes for cross-section generation, and in depth description of requirements and procedures for nodal cross-section generation. The focus of the course is the reactor physics core analysis and the requirements of coupled codes analysis. Participants will practice hands-on cross-section generation with Serpent and HELIOS codes for LWR core modelling with reactor physics code PARCS.



  • Preparation and Review of Safety Related Documentation (FSAR)

The objective of the seminar is to develop practical skills required fo the preparation and review of the safety related documentation. Target audiences are staff of the regulatory bodies, technical supporting organizations and plant personnel involved in the process of the preparation and review of the safety documentation. Concept of the safety assessment process is discussed including the relevant safety issues, such as defense in depth, graded approach, basic safety functions etc. The seminar primarily focuses on preparation and review of design basis deterministic safety analyses and includes practical exercises on review of selected parts of the SAR of existing PWR and/or BWR. Simplified plant simulator calculations are used during the exercises to enhance the development of review and evaluation skills. The safety assessment requirements practiced during the seminar are based on IAEA safety standards.



  • Practical and Theoretical Training on Level-1 PSA for Internal Initiating Events

The training is directed toward beginners in probabilistic safety assessment (PSA). The training programme is developed in the way that the participants obtain sufficient knowledge on Level-1 PSA basic concepts and methodology and practical experience in the development of the PSA models using selected PSA Software. The major part of the course will be spent on hands-on training on the development of accident sequences and system models for simplified LWR. It is expected that after training participants will be able to continue PSA activity using PSA software themselves.



  • Thermal-Hydraulics Core Analysis – COBRA Genesis Codes

A detailed modeling of the core is becoming more important in response of the industry toward higher utilization factor. Fuel cycles increased from 12 months to 18 months and more recently to 24 months over the last three decades and average discharge burnup almost doubled in the same time period. New fuel degradation phenomena have been discovered and, as result, regulatory requirements evolved to reflect the new knowledge gathered. A detailed analysis of the core component is now typical for most scenarios, both Anticipated Operation Occurences (AOOs) and Design Basis Accidents (DBAs). Moreover a detailed core thermal-hydraulic model is required in various disciplines associated with core engineering (core design, fuel rod design, subchannel analysis, etc.). The trend is now to develop detailed core models in the framework of multiphysics tools. The objective if this introductory course is to review the model needs with focus on the core component and the approach taken for various scenarios. The course will provide an overview of the computer codes used to perform safety analyses and address core engineering problems. Students will learn about the purpose and various uses of these methods. A version of COBRA will be used training. The syntax and the input structure of the code and plotting tools will be covered. Hands-on training on simple modeling is provided. In the last day advanced and future applications of are also presented.


 

  • Fuel Behavior Analysis by TRANSURANUS

The aim of the training is to provide both practical and theoretical insights on nuclear fuel behavior. Nuclear fuel undergoes continues changes while it is irradiated. Such changes affect the thermo-mechanical fuel characteristics and hence the fuel behavior and response both under normal and off-normal conditions. A series of relevant phenomena will be illustrated and discussed within the theoretical part and addressed into the practical sessions by hands-on training on suitable examples. The course will provide an overview of the computer code including the syntax, the preparation of the inputs and the analysis of the results.


 

  • Severe Accident Analysis: Phenomenology and Computational Tools

The severe accidents at Three Mile Island (TMI), Chernobyl, and Fukushima are a reminder that commitment to nuclear power includes a commitment to public safety. The nuclear industry recognized early the potential hazards of nuclear power. Severe accident has acquired an increasing relevance from the point of view of licensing and some severe accident are now recognized and included in the design basis accidents. Features to prevent, contain, and otherwise protect the public from reactor accidents were applied from the outset. As the industry has evolved, so has safety in the form of design features and strategies to both prevent severe accidents and mitigate consequences should they occur. This course presents both historical and technical information regarding severe accidents in the design and safety assessment of nuclear power plants. It is divided into daily morning theory and afternoon practice sessions. Theory aspects address phenomenology, accident progression, challenges to containment integrity, and radiological release and transport, computational tools. Practice aspects address licensing, computer codes applications, deterministic and probabilistic evaluation methods, and modeling.

 

  • Important Elements of Risk Quantification and PSA

The objective of the course is to give to the participants the essential knowledge of those elements and aspects of the probabilistic safety analysis (PSA) and risk quantification which are most important but are usually only briefly touched at the training courses or workshops based on particular PSA software tool and held either by a vendor or a user of particular tool. The purpose is to make the participants understand the important elements of quantitative risk modeling regardless of the software to be used and of the facility or system to be modeled, and to enable them, thus, to build the quantitative risk model by means of any software tool and for any industry or particular facility. Many of the PSA training courses begin with event trees (ET) and fault trees (FT) which are the skeleton of the risk model for any complex system. In engineering practice (and nuclear safety engineering in particular), however, no fault tree for a system or function should be developed by an analyst who does not have the essential knowledge in the topics which include reliability engineering (with underlying knowledge of probability theory), human reliability analysis (HRA) techniques, quantitative parameters estimate and quantitative treatment of uncertainty, to name some of them. Likewise, any member of a PSA / risk quantification project or group should understand the principles of the risk curve and the roles of deterministic safety margin analyses and probabilistic risk analyses in the design verification process. The course purposefully does not involve work on or presentation of any PSA software tool. However, it includes practical hands-on exercises on all relevant aspects of risk quantification by means of the elementary tools such as spreadsheets or “manual” calculations, which provide the first-hand experience. Who should come to this training course:
 • PSA newcomers who want to / need to have an understanding of
   PSA / quantitative risk analysis as a whole;
 • Specialists in specific PSA tasks such as data analysts, HRA, ET/FT
   model developers and others who want to / need to have an integrated
   understanding of PSA;
 • PSA or safety analysis / assessment managers who, by definition of their
   positions, need to have an integrated understanding of PSA;
 • Deterministic design basis assessors who need to understand the
   risk assessment;
 • Just any analyst or assessor (either from utility or regulatory body or
   industry or any other organization) involved in the development, verification
   or review of design of systems or functions in industrial or societal facilities.

The overall course is divided into the nine main topical areas covered by nine lectures.

 


 

ORGANIZATION

The Network of Nuclear Engineering and Energy Services (NNEES) has organized Hands-on Training Courses directed toward engineers with advanced expertise in System and Core Thermal-Hydraulic Codes, Reactor Physics and Fuel Behavior Codes, Severe Accidents as well as with enough background in Probabilistic Safety Analysis and Preparation and Review of Safety related documentation (i.e. Final Safety Analysis Report).

The Hands-on Training Courses will take place in Lucca (Italy) from 13th to 17th November, 2017.

Further information about participation and registration as well as useful practical information can be obtained from Alessandro Petruzzi at the following email address: alessandro.petruzzi@nnees.sk.
For Accommodation and Transportation see the Venue page



EXPECTED PRODUCT

The Courses will provide a transfer of experience and know-how from recognized experts in the respective fields. It will thus contribute to maintaining and increasing technical competence and to ensuring the sustainable development of nuclear technology. CDs containing all lectures will be distributed to the participants.


 

OORGANIZING COMMITTEE

M. KristofNNEES, Slovakia
M. ModroNINE, Italy
A. Petruzzi NINE, Italy


 

 

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