ML031200267
| ML031200267 | |
| Person / Time | |
|---|---|
| Issue date: | 05/01/2003 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Klingler, G R, NRR/DIPM/IIPB, 415-3077 | |
| Shared Package | |
| ML031200212 | List: |
| References | |
| CN 03-014 | |
| Download: ML031200267 (50) | |
Text
Appendix C-2 Reactor Engineering Inspector Technical Proficiency Training and Qualification Journal Issue Date: 05/01/03 APP C2-1 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-2 Issue Date: 05/01/03
Table of Contents Introduction............................................................... 5 Required Reactor Engineering Inspector Training Courses......................... 5 Post Qualification Engineering Inspector Training Course......................... 5 Required Refresher Training.................................................. 5 Engineering Individual Study Guide............................................ 7 (ISA-ENG-1) Capability of Safety Systems to Perform Intended Safety Functions..... 9 (ISA-ENG-2) NRCs Review of Temporary and Permanent Plant Modifications...... 13 (ISA-ENG-3) Evaluations of Changes, Tests, and Experiments (10CFR 50.59)...... 17 (ISA-ENG-4) Basic Codes, Standards and Regulatory Guides for Engineering Support....................................... 21 Engineering On-the-Job Activity.............................................. 31 (OJT-ENG-1) Safety System Design and Performance Capability................ 33 (OJT-ENG-2) Permanent Plant Modifications................................ 37 (OJT-ENG-3) Inspection of Licensee Changes, Tests, and Experiments (10 CFR 50.59)............................. 41 (OJT-ENG-4) Security Plan and Implementation............................ 43 (OJT-ENG-5) Radiation Protection Program and Implementation................ 45 Reactor Engineering Technical Proficiency Level Signature Card and Certification........................................ 47 Form 1: Reactor Engineering Technical Proficiency Level Equivalency Justification.............................................. 49 Issue Date: 05/01/03 APP C2-3 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-4 Issue Date: 05/01/03
Introduction Do not begin the activities or complete the courses in this qualification journal until you have completed the Basic Inspector Certification Journal. You may complete the General Proficiency requirements contained in Appendix B together with the Technical Proficiency requirements outlined in this journal.
Before signing up for any course, be sure that you have checked and have met any l
prerequisites.
l l
Required Reactor Engineering Inspector Training Courses l
(These courses have the completion of Appendix A as a prerequisite.)
l l
Reactor Full Series (either BWR or PWR) l BWR Series = R-304B, R-504B, and R-624B or l
PWR Series = R-304P, R-504P, and R-624P l
Basic Reactor Operations for alternate reactor type l
If you completed the BWR series then you must take R-200P l
If you completed the PWR series then you must take R-200B l
l (This course DOES NOT require the completion of Appendix A but you must meet course l
prerequisites.)
l l
Power Plant Engineering (self-study) l l
l Post Qualification Engineering Inspector Training Course l
This course IS NOT required for initial qualification. Attendance at this course will be l
determined by your supervisor and is dependent on your work experience and planned l
inspection activities.
l l
PRA Technology and Regulatory Perspectives (P-111) l l
l Required Refresher Training One of the following is required once every three years:
BWR or PWR systems refresher OR Simulator refresher OR EOP refresher If you completed your qualification before the requirement for a full series was added, you l
should alternate between R-200B and R-200P. Your selection should be coordinated with l
your supervisor.
l Issue Date: 05/01/03 APP C2-5 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-6 Issue Date: 05/01/03
Engineering Individual Study Guide Issue Date: 05/01/03 APP C2-7 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-8 Issue Date: 05/01/03
Engineering Individual Study Guide TOPIC:
(ISA-ENG-1) Capability of Safety Systems to Perform Intended Safety Functions PURPOSE:
The purpose of this guide is to acquaint you with the actions taken by the NRC in the review of safety systems to determine their capability to perform their intended safety function(s) and to discover any performance issues that hinder that capability. As a reactor engineering inspector, you will be required to understand how the inability of one or more systems to perform as intended causes increased risk for core damage and increased likelihood that the plants inherent redundancy may not be able to mitigate the loss of safety functions of those systems.
COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:
32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />
REFERENCES:
- 1.
IP 71111.21, Safety System Design and Performance Capability
- 2.
IP 93801, Safety System Functional Inspection
- 3.
NUREG-1275, Volume 14, Causes and Significance of Design-Based Issues at U.S. Nuclear Power Plants
- 4.
NUREG/CR-5640, Overview and Comparison of U.S.
Commercial Nuclear Power Plants
- 5.
Regulatory guide 1.186, Guidance and Examples for l
Identifying 10 CFR 50.2 Design Basis (ADAMS Accession l
Number: ML003754825) l l
- 6.
NEI 97-04, Design Basis Program Guidelines, Appendix B, l
(ADAMS Accession Number ML003771698) l l
- 7.
10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants
- 8.
Inspection Manual Chapter 2515
- 9.
Part 9900 10 CFR GUIDANCE 10 CFR 50.59 Changes, Tests, and Experiments Issue Date: 05/01/03 APP C2-9 1245
- 10.
NRC staff Safety Evaluation Report for a specific plant for original operating license.
- 11.
NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants EVALUATION CRITERIA:
Upon completion of the tasks in this guide you will be asked to demonstrate your understanding of the NRCs periodic review of a sample of safety systems at a given nuclear power plant and the NRCs continuing role in determining system performance deficiencies which impact a licensees quality assurance program by performing the following:
1.
State the NRCs inspection objectives for the reviews of samples of safety systems on a periodic basic and the basis for their importance.
2.
Discuss what the purposes are for a Safety System Design Inspection (SSDI:
- a. overall objective for each inspection
- b. number of inspectors and their areas of expertise
- c. duration of each inspection and the allocated resources
- d. how the inspection is risk-informed 3.
State the basis for selecting systems for a SSDI and the various methods for that selection process. The emphasis should be on some measure which can analytically determine or approximate the loss of a safety function.
4.
Define the following terms and state how samples of each are developed and assembled/paired with others for review.
State whether the reviews of these samples are considered a vertical slice inspection or has now become a horizontal inspection (scope of inspection expanded):
- a. Inspection attributes
- b. System needs
- c. System condition and capability
- d. Inspection activity
- e. Component inspection activity 5.
State the factors that cause a vertical slice inspection to become a horizontal inspection and whether the original vertical slice inspection can be resumed or not.
1245 APP C2-10 Issue Date: 05/01/03
6.
Define the contribution of each of the following documents to a SSDI, and what is its benefit in determining the functional capability of one or more systems.
- a. design basis documents
- b. licensing basis
- c. calculations and analyses
- d. Technical Specifications
- e. design changes and modifications
- f. operator training manual
- g. maintenance procedures
- h. surveillance and inservice test procedure results
- i. applicable vendor manuals
- j. generic communications ( IN, BL, and GL) 7.
For the listed documents in previous question, state how each provides insights into the assessment of a licensees quality assurance program at least in regard to the design and functional capability of safety systems.
8.
Develop a set of criteria that establishes a reasonable likelihood about the functional and operational capability of safety systems. For that set of criteria determine subsets with the minimum criteria that must be present to derive a similar reasonable affirmation.
9.
Define which reactor-oversight program cornerstones are verified by the reviews of safety systems via a SSDI.
TASKS:
1.
Read the references in sufficient detail to perform adequately in accordance with the requirements of the Evaluation Criteria.
2.
Meet with your supervisor, or the person designated to be your resource for this activity and discuss the answers to the questions listed under Evaluation Criteria.
DOCUMENTATION:
Engineering Proficiency Level Qualification Signature Card Item ISA-ENG-1.
Issue Date: 05/01/03 APP C2-11 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-12 Issue Date: 05/01/03
Engineering Individual Study Guide TOPIC:
(ISA-ENG-2) NRCs Review of Temporary and Permanent Plant Modifications PURPOSE:
The purpose of this activity is to acquaint you with the actions taken by the NRC in the review of both temporary and permanent plant modifications of power reactor facilities. As a reactor engineering inspector, you will be required to understand how design changes resulting in hardware modifications or different operating requirements of a facility can potentially impact the plants design basis and licensing basis, and also the performance capability of safety systems and components.
COMPETENCY AREAS:
INSPECTION LEVEL OF EFFORT:
32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />
REFERENCES:
1.
NUREG-1397, An Assessment of Design Control Practices l
and Design Reconstitution Programs in the Nuclear Power l
- Industry, l
l
- 2.
IP 71111.17, Permanent Plant Modifications l
l 1.
IP 71111.23, Temporary Plant Modifications l
l EVALUATION CRITERIA:
Upon completion of the tasks in this guide you will be asked to demonstrate your understanding of permanent and temporary plant modifications and the NRCs continuing role in the monitoring of design changes to power reactor facilities through the review of these types of modifications by successfully performing the following:
1.
State the NRCs inspection objectives for the reviews of both permanent and temporary plant modifications and why important.
2.
Discuss the typical format of both permanent and temporary modifications (major sections and purpose of each).
3.
Discuss how licensees control modifications both prior and post implementation including affected design documents and plant procedures.
4.
Define the following terms:
Issue Date: 05/01/03 APP C2-13 1245
a.
Configuration management b.
Current Licensing Bases c.
Design d.
Design Bases e.
Design-Bases Document f.
Design Change g.
Design Control h.
Design Margin i.
Design Output j.
Engineering Design Bases k.
Essential Design Documents l.
Fully Documented and Auditable Design 5.
Justify why the NRC is concerned about agreement between the design change of a modification and the safety evaluation contained in the modification package. Be able to address the following items - outside design basis and requirements for a license amendment for a design change.
6.
State at least five of the types of changes that comprise the category permanent plant modifications and what is the reason for each ones inclusion.
7.
State at least five of the types of changes that comprise the category temporary plant modifications and what is the reason for each ones inclusion.
8.
State which reactor-oversight program cornerstones are verified by the independent reviews of permanent and temporary plant modifications.
9.
List the following:
a.
types of design documents that may be affected by modifications b.
types of plant procedures which could be affected by modifications TASKS:
1.
Read the references in sufficient detail to perform adequately in accordance with the requirements of the Evaluation Criteria.
2.
Meet with your supervisor, or the person designated to be your resource for this activity and discuss the answers to the questions listed under Evaluation Criteria.
1245 APP C2-14 Issue Date: 05/01/03
DOCUMENTATION:
Engineering Proficiency Level Qualification Signature Card Item ISA-ENG-2.
Issue Date: 05/01/03 APP C2-15 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-16 Issue Date: 05/01/03
Engineering Individual Study Activity TOPIC:
(ISA-ENG-3) Evaluations of Changes, Tests, and Experiments (10CFR 50.59)
PURPOSE:
The purpose of this activity is to acquaint you with how to review safety evaluations which are used to determine if the power reactor facility change, test, or experiment, requires NRC approval prior to implementation. As a reactor engineering inspector, you will be required to understand how design changes resulting in hardware modifications or different operating requirements of a facility can potentially impact the plants design basis and licensing basis, and also the performance capability of safety systems and components.
The purpose is to:
1.
Familiarize you with the NRC regulations governing changes, tests, and experiments for commercial nuclear power facilities and 2.
Enable you to demonstrate an ability to conduct a 50.59 inspection in accordance with inspection procedure (IP) 71111.02 COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
REFERENCES:
10 CFR 50.59; Changes, Tests, and Experiments NRC Regulatory Guide 1.187; Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments NEI 96-07, Revision 1; Guidelines for 10 CFR 50.59 Implementation NRC Inspection Manual Part 9900, 10 CFR Guidance 50.59.CFR Regulatory Issue Summary (RIS) 2001-03; Changes, Tests, and Experiments RIS 2001-09; Control of Hazard Barriers (guidance on 50.59 applicability to barriers).
Current Regional or Office guidance for processing potential violations of 10 CFR 50.59 Issue Date: 05/01/03 APP C2-17 1245
IP 71111.02; Evaluations of Changes, Tests, or Experiments IP 71152, Identification and Resolution of Problems IP 71111.15, Operability Evaluations EVALUATION CRITERIA:
At the completion of this activity, you should be able to:
1.
State the criteria for when the licensee may make changes to the facility or procedures or perform tests or experiments without getting prior NRC approval.
2.
State the meaning of key terms used in this regulation:
UFSAR, changes, facility, procedures, tests or experiments.
3.
Describe when provision 10 CFR 50.65(a)(4) of the maintenance rule should be used instead of 10 CFR 50.59.
4.
Describe the applicable NRC regulation governing when a licensee may make changes to the fire protection program of a facility.
5.
Evaluate example changes, tests, or experiments for whether the licensee may perform them without prior NRC approval. Also, evaluate the example changes for their affect on operability.
6.
Draft a notice of violation against 10 CFR 50.59 and prepare a 50.59 enforcement review panel package.
TASKS:
1.
Review the references listed above.
2.
Complete the NRC WEB site training on 10 CFR 50.59.
1245 APP C2-18 Issue Date: 05/01/03
3.
Review at least three recently documented examples of violations of 10 CFR 50.59 and at least one 50.59 review panel package.
4.
Meet with your supervisor and demonstrate your understanding of 10 CFR 50.59, including your ability to satisfy the above evaluation criteria.
DOCUMENTATION:
Engineering Proficiency Level Qualification Signature Card Item ISA-ENG-3 Issue Date: 05/01/03 APP C2-19 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-20 Issue Date: 05/01/03
Engineering Individual Study Activity TOPIC:
(ISA-ENG-4) Basic Codes, Standards and Regulatory Guides for Engineering Support PURPOSE:
The purpose of this activity is to familiarize you with the basic NRC codes, regulatory guides (RGs) and associated industry standards commonly used by Engineering Inspectors. This activity will also acquaint you with the requirements (codes), guidelines (RGs) and accepted methodologies (industry standards) for licensees to accomplish various safety related activities. And lastly this activity will prepare you to determine an individual licensees commitment to RGs and standards.
COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:
40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />
REFERENCES:
See attached listings of general and discipline related references EVALUATION CRITERIA:
At the completion of this activity, you should be able to:
1.
State the general code sections commonly used by Engineering inspectors and discuss the topics included in these sections.
2.
Discuss the relationship between RGs and industry standards.
3.
Identify the RGs and associated industry standards which address the 10 CFR 50 Appendix B, Quality Assurance Criteria.
4.
Discuss the topics included in the RGs and industry standards associated with your engineering discipline.
5.
Discuss application of these references to engineering inspection activity.
TASKS:
1.
Read 10 CFR 50, Appendix B and review a selected l
licensees Quality Assurance Manual. A sample of licensee l
implementing procedures should be reviewed (such as those l
associated with engineering inspections - Design Control l
Issue Date: 05/01/03 APP C2-21 1245
l and Corrective Action) with an Evaluation Criteria to explain l
how a typical licensee meets the requirements.
l 2.
Review the references in the attached list of general references as well as those listed for your specific discipline.
3.
Locate on the NRC external web page the listing of Regulatory Guides.
4.
Review a plant specific UFSAR to identify what the licensees commitments are to particular RGs and standards.
5.
Discuss with experienced inspectors any questions you have concerning the topics of the references or the application to inspection activities 6.
Meet with your supervisor and demonstrate your familiarity with the applicable references and discuss the applications of these references to engineering inspection activities.
DOCUMENTATION:
Engineering Proficiency Level Qualification Signature Card Item ISA-ENG-4.
1245 APP C2-22 Issue Date: 05/01/03
REFERENCES for ISA-ENG-4 General 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants 10CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants 10 CFR 50.55a, Codes and Standards 10 CFR 50.65.
Requirements for monitoring the effectiveness of maintenance at nuclear power plants Regulatory Guides ANSI STANDARDS (10 CFR 50 Appendix B) 1.28 ASME NQA-1 l
1.33 ANSI 18.1 l
1.37 ANSI N45.1 l
1.38 ANSI N45.2.2 - 1972 l
1.39 ANSI N45.2.3 l
1.30 ANSI N45.2.4 l
1.94 ANSI N45.2.5 l
1.116 ANSI N45.2.8 l
1.54 ANSI N101.4 l
l ANSI/ANS 58.14 Safety and Pressure Integrity Classification Criteria for LWR l
Regulatory Guide 1.100, Seismic Qualification of Electrical and Mechanical Equipment for Nuclear Power Plants Regulatory Guide 1.26 Quality Group Classifications and Standards for l
Water-,Steam-,
and Radioactive-Waste-Containing l
Components of Nuclear Power Plants l
Regulatory Guide 1.29, Seismic Design Classification Regulatory Guide 1.155, Station Blackout l
Regulatory Guide 1.186, Guidance and Examples for Identifying 10 CFR 50.2 Design l
Bases (ADAMS Accession Number: ML003754825) l Issue Date: 05/01/03 APP C2-23 1245
l NEI 97-04, Design Basis Program Guidelines, Appendix B, (ADAMS Accession Number l
of Appendix B of NEI 96-04 : ML003771698)
Electrical Regulatory Guide 1.6, Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems (ADAMS ML0037739924 Regulatory Guide 1.30, Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment (ANSI N45.2.4/IEEE 336)
Regulatory Guide 1.32, Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants (IEEE 308 )
Regulatory Guide 1.40, Qualification Tests of Continuous-Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants (IEEE 334)
Regulatory Guide 1.47, Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems Regulatory Guide 1.53, Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems (IEEE 279 and IEEE 379)
Regulatory Guide 1.63, Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants (IEEE 317)
Regulatory Guide 1.73, Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants Regulatory Guide 1.75, Physical Independence of Electric Systems (IEEE 384)
Regulatory Guide 1.81, Shared Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants Regulatory Guide 1.89, Qualification of Class 1E Equipment for Nuclear Power Plants (IEEE 323)
Regulatory Guide 1.106, Thermal Overload Protection for Electric Motors on Motor-Operated Valves Regulatory Guide 1.128, Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants (IEEE 484) 1245 APP C2-24 Issue Date: 05/01/03
Regulatory Guide 1.129, Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants (IEEE 450)
Regulatory Guide 1.131, Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-Cooled Nuclear Power Plants (IEEE 383) l Regulatory Guide 1.180, Guidelines for Evaluating Electromagnetic and Radio-l Frequency Interference in Safety-Related Instrumentation and l
Control Systems l
l Instrumentation & Control Regulatory Guide 1.11, Instrument Lines Penetrating Primary Containment Regulatory Guide 1.12, Instrumentation for Earthquakes Regulatory Guide 1.30, Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment (ANSI N45.2.4/IEEE 336)
Regulatory Guide 1.32, Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants (IEEE 308)
Regulatory Guide 1.47, Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems Regulatory Guide 1.53, Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems (IEEE 279 and IEEE 379)
Regulatory Guide 1.62, Manual Initiation of Protective Actions Regulatory Guide 1.63, Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants (IEEE 317)
Regulatory Guide 1.75, Physical Independence of Electric Systems (IEEE 384)
Regulatory Guide 1.89, Qualification of Class 1E Equipment for Nuclear Power Plants (IEEE 323)
Regulatory Guide 1.97, Instrumentation for Light -Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident Regulatory Guide 1.105, Instrument Set points (ISA S67.04)
Regulatory Guide 1.151, Instrument Sensing Lines (ISA S67.02)
Issue Date: 05/01/03 APP C2-25 1245
l EPRI TR-102348 Guideline on Licensing Digital Upgrades (ADAMS l
ML02080169) l l
IEEE 7-4.3.2-1993 IEEE Standard Criteria for Digital Computers in Safety l
Systems of Nuclear Power Generating Stations l
l IEEE 1050-1996 Guide for Instrument and Control Equipment Grounding in l
Generating Stations l
l IEEE 338-1987 IEEE Standard Criteria for Periodic Testing of Nuclear Power l
Generating Station Class 1E Power and Protection Systems l
Mechanical l
Regulatory Guide 1.1, NPSH for ECCS & Containment Heat Removal System Pumps l
(safety guide 1) l l
Regulatory Guide 1.9, Design, Qualification & Testing of EDGs used as Class 1E l
Onsite Electrical Power Systems l
l Regulatory Guide 1.26 Quality Group Classifications and Standards for Water, Steam, l
and Radioactive Waste Containing Components of Nuclear l
Power Plants l
l Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants l
l Regulatory Guides 1.84, l
1.85, and l
1.147 ASME Code Case Applicability.
l l
Regulatory Guide 1.82 Water Sources for Long Term Recirculation Cooling Following l
a Loss of Coolant Accident l
l Regulatory Guide 1.102 Flood Protection for Nuclear Power Plants l
l Regulatory Guide 1.116, QA Requirements for Installation, Inspection and Testing of l
Mechanical Equipment and Systems l
l Regulatory Guide 1.137, Fuel Oil Systems for Standby Diesel Generators l
l Regulatory Guide 1.140, Design, Inspection, & Testing Criteria for Air Filtration and l
Adsorption Units of Normal Atmosphere Clean Up Systems l
Regulatory Guide 1.160 Monitoring the Effectiveness of Maintenance at Nuclear Power l
Plants l
l ASME OM CODE-(year), Code for Operation and Maintenance of Nuclear Power Plants l
(Section ISI, Rules fro Inservice Testing of LWR) Subsections l
ISTA, General Requirements l
ISTB, IST of Pumps, l
Appendix I, IST of Pressure Relief Devices l
Appendix II, IST of Check Valves) 1245 APP C2-26 Issue Date: 05/01/03
ASME OM-S/G-(year),
Standards and Guides for Operation and Maintenance of l
Nuclear Power Plants l
l NUREG 1482 Guidelines for Inservice Testing at Nuclear Plants l
l l
Civil l
l ASME OM CODE-(year), Code for Operation and Maintenance of Nuclear Power Plants l
(Section ISI, Rules for Inservice Testing of LWR) Subsection, l
ISTD, Inservice Examination and Evaluation of Snubbers l
l AWS DI.1 Structural Welding Code l
l ACI 311 Recommended Practice for Concrete Inspection l
l ACI 318 Building Code Requirements for Reinforced Concrete l
l ACI 349.3 Evaluation of Existing Nuclear Safety Related Concrete Structures l
l ACI 214-77 Recommended Practice for Evaluation of Strength Test Results of l
Concrete (1983) l l
ACI 304R-89 Guide for Measuring, Mixing, Transporting and Placing Concrete l
l ACI 309R-87 Guide for Consolidation of Concrete l
l ACI 347R-88 Guide to Formwork for Concrete l
l Regulatory Guide 1.12, Nuclear Power Plant Instrumentation for Earthquakes, Rev. 2 l
l Regulatory Guide 1.35, Inservice Inspection of Ungrouted Tendons in Prestressed l
Concrete Containments, Rev. 3 l
l Regulatory Guide 1.35.1, Determining Prestressing Forces for Inspection of Prestressed l
Concrete Containments, Rev. 0 l
l Regulatory Guide 1.59, Design Basis Floods for Nuclear Power Plants, Rev. 2 l
l Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear l
Power Plants, Rev. 1 l
l Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, l
Rev. 0 l
l Regulatory Guide 1.76, Design Basis Tornado for Nuclear Power Plants, Rev. 0 l
l Regulatory Guide 1.102, Flood Protection for Nuclear Power Plants, Rev. 1 l
l Regulatory Guide 1.117, Tornado Design Classification, Rev. 1 l
l Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic l
Design of Floor-Supported Equipment or Components, Rev. 1 l
Issue Date: 05/01/03 APP C2-27 1245
l Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with l
Nuclear Power Plants, Rev.
l l
Regulatory Guide 1.132, Site Investigations for Foundations of Nuclear Power Plants, l
Rev. 1 l
l Regulatory Guide 1.136, Materials, Construction, and Testing of Concrete Containments l
(Articles CC-1000, -2000, and -4000 through -6000 of the l
Code for Concrete Reactor Vessels and Containments), Rev.
l 2
l l
Regulatory Guide 1.138, Laboratory Investigations of Soils for Engineering Analysis and l
Design of Nuclear Power Plants, Rev. 0 l
l Regulatory Guide 1.142, Safety Related Concrete Structures for Nuclear Power Plants, l
Rev. 2 l
l Regulatory Guide1.165, Identification and Characterization of Seismic Sources and l
Determination Safe Shutdown Earthquake Ground Motion, l
Rev. 0 l
l Regulatory Guide1.166, Pre-Earthquake Planning and Immediate Nuclear Plant l
Operator Postearthquake Actions, Rev. 0 l
l Regulatory Guide1.167, Restart of a Nuclear Power Plant Shut Down by a Seismic l
Event, Rev. 0 l
l ANSI N14.6 Special Lifting Devices for Shipping Containers Weighing 10000 l
pounds or more l
l ANSI N45.2.5 Supplemental QA Requirements for Installation, Inspection, and l
Testing of Structural Concrete and Structural Steel l
l ASME B&PV Section III, V, IX, XI l
ANSI/ANS 58.2 Design Basis for Protection of LWR Power Plants Against the Effects of Postulated Pipe Rupture l
Metallurgical / Welding l
l AWS DI.7 Structural Welding Code l
l Regulatory Guide 1.31 Control of Ferrite Content in Stainless Steel Weld Metal l
l Regulatory Guide 1.43 Control of Stainless Steel Weld Cladding of Low-Alloy Steel l
Components l
l Regulatory Guide 1.44 Control of the Use of Sensitized Stainless Steel l
l Regulatory Guide 1.50 Control of Preheat Temperature for Welding of Low-Alloy Steel l
Reglulatory Guide1.150 Ultrasonic Testing of Reactor Vessel Welds During Preservice l
and Inservice Examinations 1245 APP C2-28 Issue Date: 05/01/03
l Regulatory Guide1.71 Welder Qualification for Areas of Limited Accessibility l
l Regulatory Guide 1.54 Service Level I, II, and III Protective Coatings Applied to l
Nuclear Power Plants l
l Regulatory Guide 1.150 Ultrasonic Testing of Reactor Vessel Welds During Preservice l
and Inservice Examinations l
l Regulatory Guide1.178 An Approach For Plant-Specific Risk-informed Decisionmaking l
Inservice Inspection of Piping l
l Regulatory Guide 1.84 Design and Fabrication Code Case Acceptability l
l Regulatory Guide 1.85 Materials Code Case Acceptability l
l Regulatory Guide 1.147 Inservice Inspection Code Case Acceptability l
l EPRI PWR Steam Generator Examination Guidelines l
l EPRI Steam Generator Integrity Assessment Guidelines l
l ASME B&PV Section III, V, IX, XI l
l GL90-05 Temporary Non-Code Repair of ASME Code Class 1,2,3 Piping l
l Industry Standards Industry Standard endorsed by the above Regulatory Guides ASME Boiler and Pressure Vessel Code, Sections III, V, and VIII Issue Date: 05/01/03 APP C2-29 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-30 Issue Date: 05/01/03
Engineering On-the-Job Activity Issue Date: 05/01/03 APP C2-31 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-32 Issue Date: 05/01/03
Engineering On-the-Job Activity TOPIC:
(OJT-ENG-1) Safety System Design and Performance Capability PURPOSE:
The purpose of this activity is to:
1.
Familiarize you with activities commonly performed by an inspector while participating as a team member of a SSDI team.
2.
Observe and perform portions of the SSDI, as assigned by team leader, using Inspection Procedures (IPs) 71111.21 and 71152 (Identification and Resolution of Problems) 3.
Provide you the opportunity to locate and identify the design and licensing base requirements for a safety system and determine if those requirements are met and maintained.
COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:
40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> in-office preparation 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> onsite inspection
REFERENCES:
1.
IP 71111.21, Safety System Design and Performance Capability 2.
IP 71152, Identification and Resolution of Problems 3.
MC 1245, On-the-Job Activity 4, Inspection 4.
MC 0612, Power Reactor Inspection Reports 5.
Site Specific Inspection Plan (provided by team leader) 6.
Site Specific Design Basis Documents (DBDs, system descriptions, calculations, accident analyses, etc.)
7.
Site Specific Licensing Basis (UFSAR, TS, Licence Amendments, and LARs) 8.
Licensee provided preparation information, i.e. lists for applicable calculations, equipment history, problem reports, engineering evaluations, modifications, and procedures EVALUATION Issue Date: 05/01/03 APP C2-33 1245
CRITERIA:
Complete the activities as outlined in this guide and meet with your supervisor to discuss any questions you may have as a result of completing this activity. Upon completion of the tasks in this guide, you will be asked to demonstrate your understanding of the baseline inspection procedure 71111.21.
1.
Demonstrate your ability to conduct inspection activities as applied to a SSDI (IP 71111.21).
2.
Demonstrate your ability to locate and identify the design base and licensing base information.
3.
Demonstrate your familiarity with the design and licensing bases for the system(s) selected by the SSDI team inspection plan. Identify critical parameters and performance criteria.
4.
Demonstrate your ability to identify critical equipment required to achieve the design basis function for the selected system(s).
5.
Demonstrate your ability to develop an individualized inspection plan for the discipline/system/equipment you are assigned from the team inspection plan.
6.
Discuss your conclusions regarding the capability of your assigned equipment/system(s) to achieve their design and licensing base functions. Provide the bases for that conclusion, i.e. evaluations, testing, performance history, etc.
7.
Demonstrate your capability to document your inspection findings consistent with MC 0612.
8.
Demonstrate your familiarity with the Significance Determination Process (SDP), Group 1, 2, and 3 questions of MC 0612 for an actual or simulated finding.
1245 APP C2-34 Issue Date: 05/01/03
TASKS:
1.
Perform the tasks listed in Entry Level On-the-Job Activity 4, Inspection Activities, as applied to an inspection focused on IP 71111.21.
2.
Review IP 71111.21 and IP 71152 for overview of SSDI activity 3.
Review previous SSDI reports to improve your understanding of the implementation of IP 71111.21.
4.
Review site specific design and licensing bases documentation, provided during preparation week, to become familiar with the design and licensing bases for the systems selected for review in the team inspection plan.
5.
Develop an individualized inspection plan for the system(s)/equipment you are assigned.
6.
Identify specific critical equipment required for the safety system to achieve its design and licensing bases functions.
7.
Review available information to determine if equipment is capable of achieving and maintaining its design function.
This includes vendor manuals, specification documents, maintenance and testing documents, problem identification reports, etc.
8.
Make an assessment based on your inspection activity if the system/equipment is capable of meeting its design function.
9.
Perform a walkdown of accessible portions of the selected l
systems and equipment.
l 10.
For at least one observed or simulated finding, process the issue through the significance determination process (SDP).
11.
Meet with your supervisor or a qualified inspector designated by your supervisor and discuss the result of your activities.
DOCUMENTATION:
Engineering Proficiency Level Qualification Signature Card Item OJT-ENG-1.
Issue Date: 05/01/03 APP C2-35 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-36 Issue Date: 05/01/03
Engineering On-the-Job Activity TOPIC:
(OJT-ENG-2) Permanent Plant Modifications PURPOSE:
The purpose of this activity is to:
1.
Familiarize you with activities commonly performed by an inspector while inspecting permanent plant modifications.
2.
Observe and perform portions of an inspection of permanent plant modifications using Inspection Procedure (IP) 71111.17 COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:
40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />
REFERENCES:
1.
NRC IP 71111.17, Permanent Plant Modifications 2.
10 CFR 50 Appendix B, Criterion III, Design Control 3.
IMC 1245 Entry Level On-the-Job Activity 4, Inspection Activities 4.
MC 0612, Power Reactor Inspection Reports 5.
ANSI Standard N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants EVALUATION CRITERIA:
Complete the activities as outlined in this guide and meet with your supervisor to discuss any questions you may have as a result of completing this activity. Upon completion of the tasks in this guide, you will be asked to demonstrate your understanding of how to conduct an inspection of plant modifications using the baseline inspection procedure 71111.17.
1.
Demonstrate your ability to satisfy the evaluation criteria of Entry Level On-the-Job Activity 4, Inspection Activities, as applied to an inspection focused on IP 71111.17 2.
Demonstrate your ability to select modifications for review which are risk significant.
Issue Date: 05/01/03 APP C2-37 1245
3.
For selected modifications demonstrate your ability to identify the design safety function of the structure, system, or component (SSC) and the design requirements 4.
Discuss for each modification, how the licensee assured the modification did not adversely impact the design, availability, reliability, or functional capability of the SSC.
5.
Demonstrate an understanding of potential risk significant plant configurations that could occur during modification implementation. How does the licensee address this?
6.
Demonstrate your capability to document your inspection findings consistent with IMC 0612.
7.
Demonstrate your familiarity with the Significance Determination Process (SDP), Group 1, 2, and 3 questions of IMC 0612 for an actual or simulated finding.
TASKS:
1.
Perform the tasks listed in Entry Level On-the-Job Activity 4, Inspection Activities, as applied to an inspection focused on IP 71111.17 2.
Discuss with Regional PRA specialist, which systems or equipment modifications have the highest risk significance.
3.
For modifications selected, determine the intended safety function and design requirements for the applicable SSC.
4.
For modifications selected, determine the motivation for the change, i.e. problem report, equipment failure, etc. and verify that the modification resolved the problem.
5.
Review post-modification testing and inspection documentation and verify the testing was adequate to assure the functional capability or design function of the SSC was not degraded.
6.
Review the plant configuration for modification implementation and testing. Review licensees actions to assure the plant was not placed in a risk significant configuration.
7.
When possible, perform a field walkdown of the SSC modified and determine if final condition was as designed by the modification documentation.
1245 APP C2-38 Issue Date: 05/01/03
8.
For change or substitution of component parts via the procurement or modification process, review equivalency evaluations that validate the adequacy of the replacement part.
9.
For at least one observed or simulated finding, process the issue through the significance determination process (SDP).
10.
Meet with your supervisor or a qualified inspector designated by your supervisor and discuss the result of your activities.
DOCUMENTATION:
Engineering Proficiency Level Qualification Signature Card Item OJT-ENG-2.
Issue Date: 05/01/03 APP C2-39 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-40 Issue Date: 05/01/03
Engineering On-the-Job Activity TOPIC:
(OJT-ENG-3) Inspection of Licensee Changes, Tests, and Experiments (10 CFR 50.59)
PURPOSE:
The purpose of this activity is to:
1.
Familiarize you with activities commonly performed by an inspector while inspecting licensee changes, tests, and experiments to determine if they may be done prior to NRC approval.
2.
Observe and perform portions of an inspection of changes, tests, and experiments using inspection procedure (IP) 71111.02.
COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:
40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />
REFERENCES:
NRC Inspection Procedure (IP) 71111.02; Evaluations of Changes, Tests, or Experiments Engineering Individual Study Activity ISA-ENG-3 on 10 CFR 50.59 Entry Level On-the-Job Activity 4, Inspection Activities EVALUATION CRITERIA:
Complete the activities as outlined in this guide and meet with your supervisor to discuss any questions you may have as a result of completing this activity. Upon completion of the tasks in this guide, you will be asked to demonstrate your understanding of the baseline inspection procedure 71111.02 by:
1.
Demonstrating your ability to satisfy the evaluation criteria of Entry Level On-the-Job Activity 4, Inspection Activities, as applied to an inspection focused on IP 71111.02.
2.
Describing the changes, tests, or experiments that you reviewed and your evaluation of why the licensee may (or may not) perform them without prior NRC approval and why they did not (or did) affect operability.
Issue Date: 05/01/03 APP C2-41 1245
TASKS:
1.
Perform the tasks listed in Entry Level On-the-Job Activity 4, Inspection Activities, as applied to an inspection focused on the topic of 10 CFR 50.59 (IP 71111.02).
2.
For at least one observed or simulated finding, relating to 10 CFR 50.59, process the finding through the significance determination process (SDP).
3.
Meet with your supervisor or a qualified inspector designated by your supervisor and discuss the result of your activities.
DOCUMENTATION:
Engineering Proficiency Level Qualification Signature Card Item OJT-ENG-3.
1245 APP C2-42 Issue Date: 05/01/03
Engineering On-the-Job Activity TOPIC:
(OJT-ENG-4) Security Plan and Implementation PURPOSE:
The purpose of this activity is to familiarize you with a typical security plan for a nuclear facility.
COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
REFERENCES:
1.
Security Plan for a selected facility 2.
Technical Specifications for the selected facility 3.
10 CFR Part 73.55 EVALUATION CRITERIA:
Upon completion of the tasks, you should be able to:
- 1.
Generally describe how the site security force maintains access control of the owner-controlled, protected, and vital areas.
2.
Demonstrate the appropriate procedures for escorting visitors into and out of the protected and vital areas.
3.
Explain the site specific methods used to detect intruders.
4.
Explain the need for maintaining classification of certain safeguards material.
TASKS:
1.
Review the references listed above, as appropriate, to develop an understanding of the site security system.
2.
Conduct a walkdown of the protected and vital areas to identify the various types of intruder detection equipment used.
3.
Tour the Central and Secondary Alarm Stations. Discuss the duties and responsibilities of personnel stationed in those facilities with the watchstanders and the security shift supervisor.
Issue Date: 05/01/03 APP C2-43 1245
4.
Discuss inspector responsibilities related to site security and safeguards with your supervisor or a qualified Operations or Physical Security inspector. Your discussion should include practical circumstances that you may encounter such as loss of security badge or identification of an inattentive guard. In addition, discuss any questions that you may have as a result of this activity.
5.
Meet with your supervisor or a qualified inspector designated by your supervisor and discuss the result of your activities DOCUMENTATION:
Engineering Inspection Proficiency Level Qualification Signature Card Item OJT-ENG-4 1245 APP C2-44 Issue Date: 05/01/03
Engineering Inspector On-the-Job Activity TOPIC:
(OJT-ENG-5) Radiation Protection Program and Implementation PURPOSE:
The Radiation Protection Program and implementing procedures are intended to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine nuclear reactor operation. As Low As Is Reasonably Achievable (ALARA) program, external exposure, internal exposure, respiratory protection, posting and labeling, survey, and reporting requirements are addressed in licensees procedures and in 10 CFR Parts 19 and
- 20. This activity will provide you a general understanding of the applicable regulatory requirements, the licensees radiation protection program, and implementing procedures.
COMPETENCY AREA:
INSPECTION LEVEL OF EFFORT:16 Hours
REFERENCES:
- 1.
Licensee procedures addressing the implementation of NRC inspections in the Radiation Protection program.
2.
Plant Technical Specifications, Plant Updated Final Safety Analysis Report, and 10 CFR Parts 19 and 20.
3.
Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas.
4.
Radiation Work Permit (RWP) used for NRC inspection activities.
EVALUATION CRITERIA:
Upon completion of the tasks, you should be able to:
- 1.
Generally describe the following terms and provide examples of each term.
a.
Controlled area b.
Radiological restricted area c.
Radiation area d.
Very high radiation area g.
Hot spots h.
Airborne radiation area Issue Date: 05/01/03 APP C2-45 1245
2.
Explain the ALARA concept and how it is applied to performance of radiological work at your site.
3.
Describe the plants overall administrative procedures for control of external exposure, internal exposure, and airborne exposure and how the procedures are implemented during NRC inspections.
4.
Describe physical and administrative controls for radiation areas, high radiation areas, very high radiation areas, and airborne radioactivity areas.
TASKS:
1.
Locate the listed references for a selected facility.
2.
Review the references and licensees procedures to develop an overall understanding of the regulatory requirements and how the radiation protection program is being implemented.
Review the Radiation Work Permit (RWP) which allows a visiting NRC inspector to complete their assigned inspection.
3.
During a plant tour, identify at least one of each of the following: radiation area, high radiation area, very high radiation area, hot spots area, and an airborne radioactivity area and observe how each access is controlled in accordance with regulations and the licensees requirements.
4.
Review at least one completed radiation survey results and explain how you will incorporate the survey results into your inspection effort.
5.
Review the licensee procedures for radiation control. Review the actions required of an individual when contamination is detected prior to exiting the Radiation Controlled Area (FCA).
6.
Meet with your supervisor or a qualified Engineering inspector to discuss any questions that you may have as a result of these activities and demonstrate that you can meet the evaluation criteria.
DOCUMENTATION:
Engineering Inspector Proficiency Level Qualification Signature Card Item OJT-ENG-5 Reactor Engineering Technical Proficiency Level Signature Card and Certification 1245 APP C2-46 Issue Date: 05/01/03
Inspector Name: ___________________________________
Employee Initials/ Date Supervisors Signature/Date A. Training Courses Power Plant Engineering (Self Study)
Reactor Full Series (either BWR or PWR)
Basic Reactor Operations for alternate reactor type B. Individual Study Activities ISA-ENG-1 Capabilities of Safety Systems to Perform Intended Functions ISA-ENG-2 NRCs Review of Temporary and Permanent Plant Modifications ISA-ENG-3 10 CFR 50.59: Changes, Tests and Experiments ISA-ENG-4 Codes and Standards for Engineering Support C. On-the-Job Training Activities OJT-ENG-1 Safety System Design OJT-ENG-2 Permanent Plant Modifications OJT-ENG-3 Inspection of Licensee Changes, Tests and Experiments (10 CFR 50.59)
OJT-ENG-4 Security Plan and Implementation OJT-ENG-5 Radiation Protection Program and Implementation Supervisors signature indicates successful completion of all required courses and activities listed in this journal and readiness to appear before the Oral Board.
Supervisors Signature: _____________________________ Date: _______________
This signature card and certification must be accompanied by the appropriate Form 1, Reactor Engineering Technical Proficiency Level Equivalency Justification, if applicable.
Copies:
Inspector HR Office Supervisor Issue Date: 05/01/03 APP C2-47 1245
PAGE INTENTIONALLY BLANK 1245 APP C2-48 Issue Date: 05/01/03
Form 1: Reactor Engineering Technical Proficiency Level Equivalency Justification Inspector Name: ___________________________________
Identify equivalent training and experience for which the inspector is to be given credit Training Courses Power Plant Engineering (Self Study)
Reactor Full Series (either BWR or PWR)
Basic Reactor Operations for alternate reactor type Individual Study Activities ISA-ENG-1 Capabilities of Safety Systems to Perform Intended Functions ISA-ENG-2 NRCs Review of Temporary and Permanent Plant Modifications ISA-ENG-3 10 CFR 50.59: Changes, Tests and Experiments ISA-ENG-4 Codes and Standards for Engineering Support Issue Date: 05/01/03 APP C2-49 1245
On-the-Job Training Activities Identify equivalent training and experience for which the inspector is to be given credit OJT-ENG-1 Safety System Design OJT-ENG-2 Permanent Plant Modifications OJT-ENG-3 Inspection of Licensee Changes, Tests and Experiments (10 CFR 50.59)
OJT-ENG-4 Security Plan and Implementation OJT-ENG-5 Radiation Protection Program and Implementation Supervisors Recommendation Signature / Date_________________________
Division Directors Approval Signature / Date_________________________
Copies to:
Inspector HR Office Supervisor 1245 APP C2-50 Issue Date: 05/01/03