EMERSONWestinghouse E 20 Training From Normal Operation to Se lectric Belgium 14 Catalog vere Accident Management

THE 2014 WESTINGHOUSE TRAINING PROGRAM 1

 TRAINING SCOPE 

In addition to our engineering and site-service 

activities, Westinghouse Electric Belgium 

provides training services to customers in the 

nuclear industry. 

Starting from basic courses describing the 

Pressurized Water Reactor (PWR), the training 

scope extends to the plant operation field and 

Emergency Operating Procedures (EOP’s). 

Another area of expertise is represented by the 

training courses on core damage mitigation and 

severe accident management. 

The courses contained in the 2014 training 

catalog are arranged in five categories: 

 Operation and Accident Related Courses 

 Maintenance Courses 

 Engineering Courses 

 Accident-Based Seminars 

 AP1000 

Particular effort is made to eliminate 

redundancies and repetition between courses, 

thus avoiding loss of time for attendees 

participating in several courses. 

 TEACHING MATERIAL 

The materials used to prepare and teach the 

courses are based on regularly updated 

Westinghouse documents issued by the 

Westinghouse Training Organization. 

During preparation of the courses, the 

instructors have free access to any information 

available in the Belgian Engineering Department 

or in the US, which assures the high quality of 

the technical content of our courses. 

The teaching material includes PowerPoint 

presentations but also the equipment mock-ups 

located at the Field Services, and computer 

software available in the Engineering 

Department, when needed. 

 COURSE TEACHERS 

The instructors used by the Belgian Training 

Organization are drawn mostly from our Belgian 

Engineering Department, from the staff of 

engineers of the Field Services, or are from the 

US training organization. All have demonstrated 

instructional skills and practice. Their technical 

knowledge is based on years of experience in 

engineering and/or on-site jobs, as well as on 

specific instructional training courses. 

The excellent feedback received from attendees 

attests the high level our instructors have attained

 GENERAL COURSE ORGANIZATION 

Our generic courses are provided on a yearly 

basis, in English, and are open to all. 

All courses are given in our Nivelles offices, in 

Belgium. 

Most of courses run for 6 hours a day, starting at 

9:00 am. 

In addition to our generic courses we also 

provide plant specific courses, on site, according 

to customer request or needs. This allows the 

customer to enroll a large number of attendees 

in the classroom which results in a price-perattendee reduction. In addition, the Belgian 

Training Organization has a multi-language 

capability which is to the benefit of attendees 

who are not totally fluent in English. 

 INFORMATION AND REGISTRATION 

Information forms are provided at the end of this 

catalog. You may use these now or later in 2014 

to receive additional information about a course. 

Registration forms are also provided at the end 

of the catalog. You may use them now or later in 

2014 to reserve one or several seats on a 

course. 

 CONTACTS 

For any queries about our training program and 

capabilities, please contact: 

Nathalie Dessars 

Manager 

Safety Analysis & Risk Applications 

Tel. 32-67.28.82.11 

e-mail: dessarn@westinghouse.com

Annick Colignon 

Training Administrator 

Safety Analysis & Risk Applications 

Tel. 32-67.28.82.19 

CALENDAR FOR 2014 COURSES 5

The following agenda is liable to modification in case of coincidence between a generic course and a 

customer specific training session. The courses for which no date is indicated will be organized upon 

customer request. 

Identification Title Dates

OPERATION AND ACCIDENT

OP 141 PWR Plant Systems Description and Operation 3-7 March 

15-19 September 

OP 142 PWR Plant Chemistry Upon request 

OP 143 Transient and Accident Analysis 6-17 October 

OP 144 Emergency Response Guidelines 20-24 October 

OP 145 Core Damage Mitigation and Severe Accident Management 29 September - 

3 October 

OP 146 Dedicated TSC/STA Training for EOP Support 1-5 December 

MAINTENANCE

MAIN 141 Reactor Coolant Pump Seal Maintenance and Inspection Upon request 

ENGINEERING

ENG 141 ASME Code Familiarization Upon request 

ENG 142 US-Nuclear Standards, Rules and Regulations Upon request 

ENG 143 Fracture Mechanics Applications Upon request 

ACCIDENT-BASED SEMINARS

ABS 141 Loss of Coolant Accident and Loss of All AC Power Accidents Upon request 

ABS 142 Steam Generator Tube Rupture Upon request 

ABS 143 Pressurized Thermal Shock Upon request 

ABS 144 Fission Products Behavior During a Severe Accident Upon request 

ABS 145 Emergency Plan Implementation During a Severe Accident Upon request 

AP 1000 AP 1000 Upon request 

PWR Accident Analysis and Mitigating Core Damage SNE560 15 

Shift Technical Advisor (STA) Operational Applications SNE570 3 

Shift Technical Advisor (STA) Training Seminars SNE580 

Station Nuclear Engineer (SNE) Applications SNE594 15 

Advanced Station Nuclear Engineer (SNE) Refresher SNE714 

Online Station Nuclear Engineer (SNE) SNE800 12 weeks 

INSTRUMENTATION & CONTROL TRAINING 

Title US Code # Days

Inadequate Core Cooling Monitor (ICCM-86) NIC275 10 

RVLIS Hydraulics NIC276 5 

Common Q Platform Overview NIC290 3 

7300 Process Instrumentation NIC320 10 

7300 Process Instrumentation Refresher NIC321 5 

Nuclear Instrumentation System (NIS) NIC335 5 

Nuclear Instrumentation System (NIS) Refresher NIC336 5 

ONLINE – Nuclear Instrumentation System NIC337 10 weeks 

Nuclear Instrumentation System Troubleshooting NIC338 5 

Incore Instrumentation System NIC340 5 

Solid State Protection System (SSPS) NIC350 10 

Solid State Protection System (SSPS) Refresher NIC351 5 

ONLINE – Solid State Protection System (SSPS) NIC352 10 weeks 

Solid State Protection System (SSPS) Troubleshooting NIC353 5 

Rod Control System (RCS) NIC355 10 

Rod Control System (RCS) Refresher NIC356 5 

ONLINE – Rod Control System (RCS) NIC357 10 weeks 

Rod Control System (RCS) Troubleshooting NIC358 5 

Digital Rod Position Indication System (DRPI) NIC360 5 

7300 Process Instrumentation Scaling NIC380 5 

Ovation System Overview OPC100 - 

 PWR PLANT SYSTEMS DESCRIPTION 

AND OPERATION (OP 141) 

Course Objectives 

This course is designed to review the design 

basis and layout of the major PWR plant 

systems. The interactions between the different 

systems and the overall plant integration are 

particularly enhanced. 

Exercises are foreseen in order to understand 

systems operation. 

Course Outline 

DAY 1 

 Course Introduction 

 Introduction to Reactor Theory 

 Neutron physics principles and reactor 

theory 

 Core feedback 

 Thermal hydraulics of the PWR core and 

heat exchanger 

DAY 2 

 Reactor Coolant System 

 Reactor vessel and core 

 Steam generator 

 Reactor coolant pump 

 Pressurizer 

 Chemical and Volume Control System 

 Make-up 

 Boron recycle 

 Boron thermal regeneration 

 Residual Heat Removal System 

DAY 3 

 Balance of Plant Systems 

 Main steam and turbine 

 Main feedwater and condenser 

 Instrumentation & Control Systems 

 Resistance temperature detectors 

 Incore and excore nuclear instrumentation 

 Pressure, DP and level measurements 

 Rod control system 

 Steam dump 

 Pressurizer level and pressure control 

 Steam generator water level control 

DAY 4 

 Plant Operation 

 Reactor Trip and Protection System 

Actuation Logic 

 Engineered Safety Features Description 

 Emergency core cooling system 

 Auxiliary feedwater 

 Containment systems 

DAY 5 

 Accident Analysis 

 FSAR 

 Technical Specification 

 Introduction to Emergency Operating 

Procedures (EOP’s) 

 Introduction to Severe Accident Management 

Guidelines (SAMG’s) 

 Questions and Answers 

 PWR PLANT CHEMISTRY (OP 142) 

Course Objectives 

The purpose of this course is to explain the 

chemistry of the PWR, the problems that arise 

during operation and the actions to be taken to 

minimize the effects caused by these problems. 

After an initial refresher on the basic principles 

of nuclear power, the course proceeds with an 

examination of the controls required for safe 

long term operation and leads to a set of 

specifications and actions aimed at maintaining 

the required conditions. 

On completion of the course, attendees will have 

a full understanding of chemistry requirements 

and how they are met. 

Course Outline 

DAY 1 

 Introduction and presentation of course 

 PWR chemistry and corrosion 

 PWR conventional and radiochemistry 

 Reactivity control, chemical shim and boron 

chemistry 

 Reactor coolant crud 

 Tritium 

DAY 2 

 Fission product poisoning 

 Primary chemistry specifications 

 Secondary side chemistry 

 SG primary water stress corrosion cracking 

 SG secondary side corrosion 

 Secondary side specifications 

DAY 3 

 Secondary side specifications 

 Shutdown chemistry and radiochemistry 

 Operating experiences 

 Chemistry during abnormal conditions 

 Course summary and closure 

Evaluation and feedback

 TRANSIENT AND ACCIDENT ANALYSIS 

(OP 143) 

Course Objectives 

The purpose of this course is to give the 

attendees the feel for plant behavior during 

normal, abnormal and accident transients. 

Course Outline 

FIRST WEEK 

DAY 1 

 Course Introduction 

 Radiological Aspects of Core Damage 

 Fundamentals of Reactor Theory 

DAY 2 

 Fundamentals of Reactor Theory (continued) 

 Introduction to Accident Analysis 

DAY 3 

 Introduction to Accident Analysis (continued) 

 Reactivity Addition and Power Distribution 

Anomaly Accidents 

DAY 4 

 Increased Heat Removal by the Secondary 

System Accidents 

 Reduced Heat Removal by the Secondary 

System Accidents 

DAY 5 

 Reduced Reactor Coolant Flow Accidents 

 Loss of Reactor Coolant Accidents 

SECOND WEEK 

DAY 1 

 Loss of Reactor Coolant Accidents 

(continued) 

 Steam Generator Tube Rupture Accidents 

DAY 2 

 Introduction to Mitigating Core Damage 

 Critical Safety Function: 

 Subcriticality 

 ATWS 

DAY 3 

 Critical Safety Function: 

 Core Cooling 

 Inadequate Core Cooling 

 Critical Safety Function: 

 Heat Sink 

 Loss of Secondary Heat Sink 

DAY 4 

 Critical Safety Function: 

 Primary Integrity 

 Pressurized Thermal Shock 

Critical Safety Function: 

 Containment 

 Severe Accident Phenomenology 

DAY 5 

 Severe Accident Phenomenology 

(Continued) 

 Accident Response Instrumentation

 EMERGENCY RESPONSE GUIDELINES 

(OP 144) 

Course Objectives 

The purpose of this course is to explain the 

background of the ERGs Rev. 2, and their use. 

Emphasis is placed on understanding of 

phenomena and recovery actions rather than 

pure description of procedures. 

Course Outline 

DAY 1 

 Philosophy and structure of the ERGs 

 E-0 procedure and subprocedures 

DAY 2 

 LOCA concerns 

 E-1, E-2 and subprocedures 

DAY 3 

 Steam Generator Tube Rupture 

 E-3 and Subprocedures 

DAY 4 

 ATWS 

 Inadequate Core Cooling 

 Loss of Feedwater 

DAY 5 

 Pressurized Thermal Shock (PTS) 

 Containment Integrity 

 RCS Inventory 

 Total Loss of AC Power 

 Questions and Answers 

 CORE DAMAGE MITIGATION AND 

SEVERE ACCIDENT MANAGEMENT 

(OP 145) 

Course Objectives 

This course is designed to familiarize plant 

operation personnel and staff members with 

severe accident phenomena and accident 

scenarios highlighting the recovery and 

mitigation actions to prevent and limit core 

damage, maintain containment integrity and 

minimize the fission product releases. The 

Severe Accident Management Guidelines 

(SAMGs), developed by the Westinghouse 

Owners Group, are presented and their link with 

the Emergency Operating Procedures and the 

Site Emergency Plan is explained. 

Course Outline 

 Introduction 

 Definition of a severe accident 

 Description of Chernobyl, Three Mile 

Island, and Fukushima Accidents 

 Tools for the Study of the Severe 

Accidents 

 PSA Terminology and Scope 

 PSA Example Results 

 PSA Uses 

 PSA Decision Making Criteria 

 Severe Accident Simulation Models 

 Example Severe Accident Sequence 

 Introduction to Severe Accident 

Management 

 Plant behavior prior to Core Damage: 

Initiating Events, Emergency Operating 

Procedures (EOPs) 

 Anticipated Transient Without Scram 

(ATWS) 

 Loss of Coolant Accidents / Inadequate 

Core Cooling (ICC) 

 Loss of Feedwater / Loss of Heat Sink 

(LOHS) 

 Loss of AC Power 

 Severe Overcooling / Pressurized 

Thermal Shock (PTS) 

 Response of Instrumentation to Core 

Uncovery 

 Plant behavior during and after core damage: 

in vessel phase 

 Behavior up to core uncovery 

 Core melt progression 

 Hydrogen generation 

 Natural circulation and creep failure 

phenomena 

 Reactor vessel failure 

 Importance of EOPs and operator actions 

 Plant behavior during and after core damage: 

ex- vessel phase 

 Containment design 

 Debris dispersal 

 Direct containment heating 

 Vessel thrust 

 Steam explosions 

 Debris coolability 

 Core concrete attack 

 Hydrogen behavior in containment 

 Containment fragility and failure modes 

 Radiological Aspects 

 Fission product inventory 

 Fission product release from fuel 

 Fission product transport 

 Source terms 

 Severe accident mitigation hardware 

 Filtered containment venting 

 Emergency containment spray system 

 Hydrogen control systems 

 Severe Accident Management Guidance – 

WOG SAMG Overview 

 Background 

 Scope and philosophy 

 Technical basis 

 Goals 

 Structure of SAMG 

 Interface with EOPs and E-plan 

 Control room SAMG 

 TSC SAMG 

 Instrumentation 

 Phenomenology 

 Computational aids 

 Design variations 

 Summary 

 DEDICATED TSC/STA TRAINING 

FOR EOP SUPPORT (OP 146) 

Course Objectives 

The purpose of this course is to provide the 

necessary information to Technical Support 

Center (TSC)/Shift Technical Adviser 

(STA)/Plant Engineering Staff (PES) such that 

they can provide adequate and effective support 

to the operators during an accident recovery. 

Course Approach

The teaching approach combined classic 

presentations where the physical aspects of the 

most important EOP recovery strategies are 

presented and explained with Case Studies, 

which put the attendees in situations for which 

they have to come up with answers to operator 

initiated questions or advises. The Plant 

Engineering Staff (PES) Case Studies in the 

following agenda are intended to provide 

information on the possible evaluations that a 

Technical Support Center (TSC), Shift Technical 

Adviser (STA) or Plant Engineering Staff (PES) 

would have to perform to support control room 

operators in case of accidents. This includes 

evaluations concerning: 

 RHR suction alignment 

 Need to transfer to hot leg recirculation 

 Establishing RCS letdown or not 

 Venting RV head or not 

 Post-SGTR cooldown method 

 Control of sump pH 

 SG overfill 

 Which SG to use for cooldown 

 Reinitiation of feed to a dry SG 

 RCP status 

 Long term plant status 

Course Outline 

DAY 1 

 Symptom-Based Emergency Operating 

Procedures – Introduction 

 Diagnostic 

 SI Termination & Reinitiation Criteria 

 EOP Evaluations by TSC or Plant 

Engineering Staff 

 Loss of Coolant Accident Physics in Relation 

with Break Size 

DAY 2 

 RCP Trip/Restart Criteria 

 SI Reduction Criteria 

 Case Study 1 – Small Break LOCA 

 Case Study 2 – Large Break LOCA 

DAY 3 

 Shutdown LOCA 

 Case Study 3 – Stuck Open Safety Valve 

 Case Study 4 – LOCA Outside Containment 

 Pressurized Thermal Shock Aspects – FR-Ps 

DAY 4 

 E-3, Steam Generator Tube Rupture 

 SGTR Contingencies 

 Case Study 5 – SGTR 

 Case Study 6 – Faulted & Ruptured SG 

 Return to power & ATWS 

DAY 5 

 Case Study 7 – Anticipated Transient 

Without Trip 

 Total Loss of Feedwater & Bleed and Feed 

 Case Study 8 – Loss of Secondary Heat Sink 

 Case Study 9 – Degraded and Inadequate 

Core Cooling 

 Instrument Response in Accident Conditions 

 MAINTENANCE COURSES 

MAIN 141: Reactor Coolant Pump Seal Maintenance and Inspection 

 REACTOR COOLANT PUMP SEAL 

MAINTENANCE AND INSPECTION 

(MAIN 141) 

Course Outline 

1. Theory and functional description of the 

RCP 

 - 3 days: theoretical 

 - 2 days: practical 

 General description of the pump and 

motor components 

 Description of the seal area 

 Flow paths 

Internal 

External - Interface with CCW 

 - Injection water 

 Auxiliary equipment 

 Pump side: filters, standpipe 

 Motor side: oil lift system 

 Bearing system 

2. Walk through the total inspection program 

(classroom phase) 

 Dismounting technique of seals 

 Discussion of 

 Why to inspect 

 How to inspect 

 Inspection intervals 

 Inspection technique 

 Inspection criteria 

 Monitoring of seals 

3. Practical training in dismounting/erection 

 Practical demonstration of seals 

dismounting/erection following the 

manual 

The purpose is to apply the theoretical 

aspects reviewed on the first day. 

The exercise will be done on the ESC 

mock-up using the adequate tools and 

equipment. 

4. Practical performance of inspection 

program 

This session is dedicated to the practical 

performance of dismounting/remounting 

and seals inspection on the ESC mock-up. 

Each person will have the opportunity to 

handle the equipment under the lead of 

instructors. 

5. Theoretical description of motor to pump 

alignment 

 Centering of pump (on Mock-Up) 

 Swing check 

 Oil lift 

 Axial end play 

 Discussion of tooling used and 

modifications 

 Conclusion and open discussion 

ENGINEERING COURSES 

ENG 141: ASME Code Familiarization 

ENG 142: US-Nuclear Standards, Rules and Regulations 

ENG 143: Fracture Mechanics Applications 

 ASME CODE FAMILIARIZATION 

(ENG 141) 

Course Objectives 

The purpose of the course is to present the 

background and history of the ASME code for 

boiler and pressure vessel design and 

construction. Specific attention is also paid to 

the organization and use of the code. At the end 

of the course the attendees will be able to 

manipulate the ASME code for their own 

application. 

Practical workshops are foreseen to enhance 

code utilization. 

Course Outline 

DAY 1 

 What is the ASME code? 

 General introduction to the ASME B&PV 

Code 

 Comparison and relationship with other 

US and European Codes: ANSI, ASTM, 

AWS, ASNT, French RCCM, German 

KTA 

 Administration of the ASME Code and 

organization of the Code editions and 

addenda. 

 Why was the Code introduced and how is it 

applied in the nuclear industry? 

 Significant events in the USA 

 References to US Code of Federal 

Regulations (CFR) and NRC Regulatory 

Guides 

 Design Basis and the Code 

 Adaptation and use of the ASME Code in 

European countries. Position of the 

respective National Regulatory Bodies. 

 How is the ASME code organized? 

 Detailed content of the ASME B&PV code 

 Definitions 

 Organization in sections 

 Organization in subsections and articles 

 Interpretations and Code cases 

 Relationship between the different parts of 

the Code 

 Nuclear Power Plant components: 

definition of and relationship between 

ANSI Safety Classification and ASME 

Code Class 

 Workshop on the general Code structure and 

use. 

DAY 2 

 Specific presentation of the Code sections of 

main interest to the Nuclear Power Industry 

 Section II Material Specifications 

(Ferrous, Nonferrous and Welding 

Materials) 

 Section III Subsection NCA 

 Scope of each Subsection 

 Structure in Articles (1000 to 8000) 

 Subsection NB: Class 1 Components 

 Article 2000 (Material) 

 Article 3000 (Design) 

 Article 4000 (Fabrication and 

Installation) 

 Article 5000 (Examination) 

 Article 6000 (Testing) 

 Subsections NC, ND: Class 2 and 3 

Components 

 Subsection NF: Component Supports 

 Subsection NG: Core Support Structures 

 Specific presentation of the Code section III 

Article NB-3000 (Class 1 Components 

Design) 

 This presentation will address the 

concepts and background applied in 

NB-3000: General Design, Design by 

Analysis (i.e. NB-3200, stress categories 

and respective limits), Design by Rules 

(e.g. NB-3600 Piping) 

 Purpose, meaning and details of specific 

equations will be presented (e.g. NB-3600 

Class 1 fatigue analysis) 

 Workshop on the use and application of 

ASME III subsection NB. 

DAY 3 

 Specific presentation of the Code sections 

 Section V: Nondestructive Examination 

NDE methods and actual records will be 

presented to the attendees 

 Section VIII: Pressure Vessel, relation and 

difference with Section III 

 Section IX: Welding and Brazing 

Qualifications 

 Introduction to Fracture Mechanics concepts 

and their application in ASME III Appendix G, 

ASME XI, and 10CFR50 Appendix G 

 Section XI: In-service Inspection of NPP 

Components. 

 Exercises on Section XI application 

 Summary overview of the Code usage 

through the component life (design, 

inspection, repair, fatigue monitoring) 

 Current development of the Code, NRC 

position, European Regulators position

 US-NUCLEAR STANDARDS, RULES 

AND REGULATIONS (ENG 142) 

Course Objectives 

This course will provide an overview of the 

nuclear rules, regulations and standards, 

currently applicable in the US and followed in 

many other Western countries. 

Their application in licensing during construction 

and operation of a nuclear power plant will be 

discussed. 

Aspects of safety classifications, qualification, 

quality assurance programs, maintenance and 

inspection, and rules of backfitting are included 

in the program. 

The US rules presented will be compared 

samplewise with other international regulations. 

This course is addressed to people looking for 

an introduction to Western nuclear standards. It 

can serve as a preparation for specialized 

courses (for example on the ASME-Code). 

Course Outline 

DAY 1 

 Safety objectives 

 Licensing process 

 Introduction to US rules and regulations 

 Policy statements 

 Regulations 

 Regulatory guides and standard review plan 

 National standards 

 Industry practice 

 Other NRC documents 

 Compliance with regulatory requirements 

 Safety classification 

DAY 2 

 Seismic classification 

 QA classification 

 Classification for electrical and I&C 

equipment 

 Application of safety classes and ASME code 

for mechanical equipment 

 Application of safety classes and IEEE 

standards for electrical equipment 

 Qualification of electrical and I&C equipment 

 Software qualification 

 QA program requirements 

 Backfitting and upgrading