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NuComBiP

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History and Experience
NINE experience is based upon the NINE’s personnel support to IAEA SAET (Safety Assessment and Education Training) in development and conduction of competence building program for embarking countries since 2009 for Poland, Vietnam, Malaysia, Jordan, Bangladesh, Bulgaria, Romania and UAE. In addition, since 2014 NINE’s personnel organizes twice a year “Inspector and Plant Walkdown” trainings under the guidance of US NRC inspectors at abandoned NPP at Zwentendorf, Austria.


Objective
The NuComBiP program is based on the modular approach allowing for flexibility in training structure depending on the current situation and needs of customer. It consists of all essential elements of the nuclear safety including:
•   Essential nuclear safety knowledge
•   Assessment of engineering aspects important to safety
•   Deterministic safety assessment
•   Probabilistic safety assessment
•   Plant inspection

 

Expected Products

The expected outcomes of NuComBiP are:

•   Support to draft/review Feasibility studies
•   Review of the existing infrastructure and gap analysis
•   Familiarization of the Government and other stakeholders of a national nuclear programme with the
     key components of the technical infrastructure
•   Support implementation of a competence building program
•   Delivery of the competence building program
•   Assessment of the progress and advice in development of the long-term sustainable national
     training program

 

 

Course's Subjects

 

Fundamental Knowledge                         A – Introduction to Safety Assessment
      1.    Fundamental Safety Principles and overview of IAEA Safety Standards
      2.    Basic Safety Concepts
      3.    Scope of Safety Assessment

B - Fundamentals of Safety Analysis
      1.    Scope of safety analysis
      2.    Preparing for safety analysis
      3.    Criteria for Judging Safety and Acceptance Criteria
      4.    Scope and Overview of Deterministic Safety Analysis Methods
      5.    Scope and Overview of Probabilistic Safety Analysis Methods
      6.    Use of Computer Codes
      7.    Integrate Risk Informed Decision Making
      8.    Overview SA Applications (Licensing Analyses, Development of EOPs and SAMGs, …)

C – Basic Nuclear Technology Courses
      1.    Reactor Physics
      2.    Thermal Hydraulics
      3.    Nuclear Power Reactor Design
Assessment of Engineering
Aspects Important to Safety   		  A – Crosscutting Topics
      1.    Implementation of defence in depth
      2.    Operational experience
      3.    Radiation protection
      4.    Classification of structures systems and components
      5.    Equipment qualification
      6.    Aging and wear-out mechanisms
      7.    Human factors in NPP design and operation
      8.    Protection against internal fire and explosions
      9.    Protection against internal hazards other than fire and explosions
     10.    Protection against earthquakes
     11.    Protection against external events excluding earthquakes

B – Site Evaluation
     1.    General aspects of the site evaluation
     2.    Impact of the site on the installation
     3.    Site characteristics and the potential effects of the nuclear installation in the region
     4.    Monitoring of hazards

C – Safety Assessment of the Design of the Main Systems
     1.    Reactor Core
     2.    Reactor coolant system and associated systems
     3.    Reactor containment systems
     4.    Emergency power systems
     5.    Fuel handling and storage systems
     6.    Supporting and auxiliary systems
     7.    Instrumentation and control systems
Deterministic Safety Assessment       A - Deterministic Analysis
     1.    Deterministic Safety Assessment
     2.    Scope of Deterministic Analysis
     3.    Deterministic Analysis: Summary of Technical Areas and Overview of Codes
     4.    Fundamentals of Nuclear System Modelling

B - Design Basis Analysis
     1.    Intro to Design Basis Analysis
     2.    Basic Code Modelling
     3.    Code Verification and Validation
     4.    Separate Effects Tests Modelling
     5.    Integral Effects Tests Modelling
     6.    Sensitivity Analysis
     7.    Fundamentals of Conservative vs. Best Estimate Analysis
     8.    Nuclear Power Plant Modelling
     9.    Accident Analysis, Uncertainty Evaluation and Interpretation of Results

C - Beyond Design Basis Analysis
     1.    Intro to Beyond Design Basis Analysis
     2.    Intro to BDBA Experimental Data Base
     3.    Primary Circuit Analysis
     4.    Containment Analysis
     5.    Fission Product Release and Dispersion Analysis
     6.    Interpretation and Use of Results
Probabilistic Safety Assessment   A – Probabilistic Safety Assessment
     1.    Basic Concepts
     2.    System Modeling and Analyses

B - Level 1 PSA
     1.    Intro to Level 1 PSAs
     2.    Exercises in Level 1 PSAs
     3.    Safety Assessment and Verification with Level 1 PSAs
     4.    Risk Monitors

C - Level 2 PSA
     1.    Intro to Level 2 PSAs
     2.    Exercises in Level 2 PSAs
     3.    Safety Assessment and Verification with Level 2 PSAs

D - Level 3 PSA
     1.    Intro to Level 3 PSAs
     2.    Exercises in Level 3 PSAs
     3.    Safety Assessment and Verification with Level 3 PSAs
Plant Inspection   A – Introduction to Plant Safety
     1.    Personnel safety and Equipment
     2.    Radiological Safety and Equipment
     3.    Fire Protection
     4.    Security
     5.    Emergency Preparedness
     6.    Housekeeping
     7.    Plant Status, Systems and Equipment

B – Plant Design and Operation
     1.    Basic NPP designs and operations
     2.    Plant Tour

C – Plant Walk-down
     1.    Planning a Walk-down
     2.    Preparing for a Walk-down
     3.    Detailed Plant a Walk-down of the designated Areas
     4.    Evaluation of Issues

D – Plant Inspection
     1.    Fundamentals of Inspection
     2.    Plant Operation and Documentation Familiarization
     3.    Conduct of Inspection
     4.    Evaluation of Inspection
     5.    Reporting
Development Thermal-Hydraulics Code Skills  

A – General Aspects
     1.    Introduction to Deterministic Safety Analysis
     2.    Features and Limitations of System-Thermal-Hydraulic codes
     3.    Overview of Safety Analyses with System Computer Codes: Conservative and BE Approach
     4.    Validation of system computer codes on integral test facilities

 

B – Code Syntax
      1.    Overview of the RELAP5 code architecture and structure, models; input, output, rstplt files
      2.    Minor Edits, Major Edits and Time Step Controls
      3.    Hydrodynamic Components in RELAP5
      4.    Heat Structure Components in RELAP5
      5.    Special Components in RELAP5
      6.    Logic Trips and Control Variables in RELAP5

 

C – Physical Code Models
      1.    The Hydrodynamic Models & The Heat Transfer Models
      2.    Closure Relationships
      3.    Other Special Models

 

D – Code Numerics
      1.    Numerical Methods and RELAP5 Equation
      2.    Solution Algorithms in RELAP5
      3.    Numerical Effects in RELAP5 Applications

 

E – Hands-on Training
      1.    Familiarization with plotting tools (APT plot, etc..)
      2.    Modeling a simple pipe
      3.    Edwards Pipe Problem
      4.    Valve Sizing
      5.    Blow-Down Problem
      6.    Steam Generator & Pressurizer Models
      7.    Boiling Channel Problem
      8.    Effect of Time Step and Spatial Discretization

 

F – Qualification Procedures for System Thermal-Hydraulics Calculations
      1.    Procedure for Developing  and Qualifying Nodalizations
      2.    Procedure for Nodalization Qualification
      3.    Quantification of Accuracy of a Code Calculation
      4.    Origin of Uncertainties in SYS-TH Calculations
      5.    Approaches to perform Uncertainty Analysis

 

G – Advanced Hands-on Training
      1.    Achievement of Steady State
      2.    Developing an Integral Effect Facility Nodalization
      3.    Qualification of a System Code Calculation of a ITF
      4.    Developing a NPP Nodalization
    5.  Qualification of a System Code Calculation of a NPP: The Kv Scaled Calculation
      6.    Identifying Simple and Complex Input Error

 
Development Reactor Physics Code Skills  

A – General Aspects
      1.    Features and Limitations of nodal core simulator codes
      2.    Procedures and codes for cross-section generation
    3.    Nodal cross-section requirements for static, transient and depletion analysis
      4.    Overview of the diffusion code models; input and output files

 

B – Code Syntax
      1.    Nodal cross-section generation
      2.    PAMXS and table formats
      3.    PARCS Syntax

 

C – Hands-on Training
      1.    Modeling of static core with internal feedback
      2.    Generating PMAXS cross-sections
      3.    Modeling of transient PWR core (control bank withdrawal)
      4.    Using PMAXS stand-alone steady-state calculations
      5.    Using PMAXS for transient calculations
      6.    Coupling to RELAP5
      7.    Coupling PARCS to RELAP5 PWR model
      8.    Coupling to TRACE
      9.    Coupling PARCS to TRACE PWR model
      10.  Using PMAXS for coupled calculations (Control Rod Ejection)
      11.  Fuel cycle analysis capabilities of the PARCS
      12.  Using PMAXS for depletion calculation (Core Loading Analysis)