Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety 7.3.2. Regulation and Standard Related to Nuclear Quality Assurance in Various Countries United States In July 1970, the United States Atomic Energy Commission (USAEC) first published and released the nuclear quality assurance program requirements, 10 CFR 50 Appendix B. The detailed guide and standard such as USNRC Regulatory Guide, ASME code Section Ⅲ, NCA‐4000, ANSI N45.2, ANS 15.8 and others have been developed and applied. The parts of ASME codes and ANSI standards that are not in conformity in details have been complemented, and ASME codes and ANSI standards are unified into ANSI/ASME NQA‐1, which was first published in 1979 and is still being used today after continuous revisions. The pressure boundary component quality assurance program requirements based on 10CFR50 Appendix B are specified in the Subsection NCA‐4000 of ASME code sectionⅢ. Since 10 CFR 50 Appendix B consists of comprehensive and recondite sentences, many detailed standards have been released to complement it; NQA‐1 has been released in 1979, NQA‐2 followed in 1983, and in 1990, the unified edition of both, the new NQA‐1, has been published. Thereafter, the NQA‐1 was endorsed in Regulatory Guide 1.28, revision 3 published on August 1985. Canada The outstanding feature of quality assurance standard of Canada is that quality assurance requirements are separately specified for each stages of nuclear power plant project. ￭ CSA N286.0: QA Program Requirements for Nuclear Power Programs (NPP) ￭ CSA N286.1: Procurement QA for NPP ￭ CSA N286.2: Design QA for NPP ❙ 424 ❙
￭ CSA N286.3: Construction and Installation QA for NPP ￭ CSA N286.4: Commissioning QA for NPP ￭ CSA N286.5: Operation QA for NPP ❙ 425 ❙ 7. Quality Assurance ￭ CSA N299.0: Guide for Selecting and Implementing the CAN3‐Z299‐85 QAP Standard ￭ CSA Z299.1: QAP Category 1 ￭ CSA Z299.2: QAP Category 2 ￭ CSA Z299.3: QAP Category 3 ￭ CSA Z299.4: QAP Category 4 Japan The JEAG 4101‐Quality Assurance Guides for Construction of Nuclear Power Plants, the quality assurance standard of Japan, which was firstly established in 1972 as an Electricity Technical Standard by the Electricity Technical Standard Survey Committee of the Japan Electricity Association, consists of 13 standards at present in accordance with the IAEA safety series. ￭ JEAG 4101‐1985: Quality Assurance Guide for Nuclear Power Plants ￭ JEAG 4102‐1986: Criteria for Formation of Guide for Quality Assurance Program for Nuclear Power Plants ￭ JEAG 4104‐1986: Design Management Guide for Nuclear Power Plants ￭ JEAG 4105‐1986: Purchasing Management Guide for Nuclear Power Plants ￭ JEAG 4106‐1986: Manufacturing and Installation Guide for Nuclear Power Plants ￭ JEAG 4107‐1987: Operational and Repair Guide for Nuclear Power Plants ￭ JEAG 4108‐1985: Quality Record Control Guide for Nuclear Power Plants ￭ JEAG 4109‐1985: Quality Inspection Guide for Nuclear Power Plants
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1. Nuclear Reactor Principles · 1
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1. Nuclear Reactor Principles Regio
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Table 1.2. Average cross sections f
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1. Nuclear Reactor Principles ❙ 7
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energy of the fuel nuclei. ❙ 9
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❙ 11 ❙ 1. Nuclear Reactor Princ
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1 Bq = 2.7x10 -5 µCi = 2.7x10 -8 m
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1.1.5. Natural radioactive series
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238 U 92+ 1 n0 -----> 135 Te 52 + 9
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σ t = σ s + σ a = σ elastic +
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FIG. 1.10. And here in the 1keV to
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1.2.3.2. Neutron emission in fissio
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Three possible states could then be
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TABLE 1.3. Slowing down efficiency
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1 UNGG/AGR CANDU PWR/BWR TABLE 1.5.
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✼✼✼ REFERENCES ✼✼✼ ❙
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2.1. Introduction C O N T E N T S 2
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2. Radiation Protection 2.1. Introd
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2.2.1. Exposure ❙ 61 ❙ 2. Radia
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❙ 63 ❙ 2. Radiation Protection
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2.2.4. Effective dose ❙ 67 ❙ 2.
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schematically in Fig. (2.7). ❙ 77
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The great utility of the G‐M tube
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protection conditions. These factor
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2.6.2. Effects of dose, dose‐rate
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2.7. Principles of Radiological Pro
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❙ 115 ❙ N 0 = N 2 1 t / t 1/ 2
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ln 10 = μ t1/10 ≈ 3 · ln 2 = 3
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C O N T E N T S | Table of Contents
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Fig. 1 Committees for IAEA Safety S
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Regional Basic Professional Trainin
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❙ 221 ❙ 4. DESIGN OF NUCLEAR RE
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4.1.1.2. The concept of defence in
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4.1.2. Requirements for management
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administrative control. ❙ 231 ❙
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Design basis accidents ❙ 237 ❙
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Single failure criterion ❙ 239
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Deterministic approach The determin
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accidents, with account taken of th
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Fuel elements and assemblies Fig. 4
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Reactor shutdown ❙ 251 ❙ 4. DES
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functional in the event of a severe
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4.1.5.4. Instrumentation and contro
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4.1.5.9. Radiation protection Gener
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Operation at a power level < 1 MW.
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of these exchangers, the water is c
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Fig. 4.11. HANARO Core Pool Fig. 4.
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4.2.6.2.3 Reactor Systems ❙ 295
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4.2.6.5. Neutron Activity Analysis
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❏ physical separation. ❙ 303
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5. Siting Considerations Regional B
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5.4.2. Sources of risk 5.4.3. Evalu
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6. Safety Classification Of Structu
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Safety functions necessary to preve
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6.1.3. Principal Safety Classes 6.
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Page 475 and 476: C O N T E N T S 8.1. INTRODUCTION |
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❙ 475 ❙ 8-A. Deterministic Acci
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guidelines for this categorization.
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Table 8‐3 Typical Initial Conditi
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8.5.3. Decrease in Reactor Coolant
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or by operator error, often during
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8-B. Deterministic Accident Analysi
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| List of Tables | Table 8.7‐1 Ru
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❙ 509 ❙ 8-B. Deterministic Acci
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SG Containment RCP RCP RCP SIT SIT
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8.2.5. Evaluation Model ❙ 521 ❙
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Figure 8.7‐7 Flow chart of KREM
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Cladding Temperature (K 1300 1200 1
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8.8.1.6. Loop Seal Formation and Cl
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to ECCS injection time using CEFLAS
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Table 8.8‐2 SBLOCA Break Spectrum
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F TEMPERATURE, o 0 100 200 300 400
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8.8.2.2.3 RELAP5‐sEM Methodology
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9. Probabilistic Safety Analysis Re
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9.4. LEVEL 2 PSA 9.4.1. Objectives
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❙ 627 ❙ 10-A. Limiting Conditio
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performance, correct set points, an
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injecting borated water into the pr
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Positive reactivity insertion rates
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10.4.3.3. Canadian type NPP ❙ 655
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2) Safety Limits and Safety System
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the technical specifications is not
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10-B. Technical Specifications for
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10.7.7.1. Monitoring Systems 10-B.
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10.8.1. Reactor Core Parameters 10-
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10.8.6. Emergency Power 10-B. Techn
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authorized by responsible authority
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11. Human Performance: A Perspectiv
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Figure 11.1 Model of violation Figu
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12. Surveillance Programmes Regiona
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12.4. IMPLEMENTATION 12.4.1. Genera
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13.Maintenance Management in KHNP R
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15. Operational Safety 15.1. IAEA S
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and modified if required. ❙ 837
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15.2.2. Organization and functions
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− Temporary and permanent plant m
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Operating procedures and instructio
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Off‐normal conditions are easily
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15.2.6. Work authorizations ❙ 859
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conventional hazards at the point o
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15.2.7. Accident management ❙ 863
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commissioning process. ❙ 869 ❙
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Event reports and root cause analyz
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Material regarding evaluation of ex
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Monitoring of initiatives in progre
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16. In‐Plant Accident Management
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❙ 895 ❙ 16. In‐Plant Accident
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16.2.2. Safety Functions ❙ 899
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is to place the plant in a safe, st
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each a final stable and controlled
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16.3.5. Review of Korean SAMG by KI
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this stage. Control rod degradation
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Pressurization of the containment
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Fractional Relase Fractional Relase
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19. Decommissioning Regional Basic
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| List of Tables | Table 1. Stages
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20.1. Introduction C O N T E N T S
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21. Safety Culture Regional Basic P
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21. Safety Culture 21.1. DEFINITION
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FIG. 21.1. Basic elements of safety
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❙ 1059 ❙ 21. Safety Culture Ult
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Obtaining useful information from o
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terms of what they do. There is an
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Consequently, people also understan
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organization e.g. contractors. ❙
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21.4. MANAGEMENT OF SAFETY 21.4.1.
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subdivided in 4 sub‐objectives, w
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21.4.4. Specific safety management
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21.4.4.5. Self assessment ❙ 1089
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AtomCARE Regional Basic Professiona
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AtomCARE 1. OVERVIEW OF NATIONAL RA
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❙ 1121 ❙ AtomCARE The RETAC of
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❙ 1123 ❙ AtomCARE secondary rad
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❙ 1125 ❙ AtomCARE actions by pr
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2.1.4. Component Modules (Configura
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Classification Major Functions Tech
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❙ 1131 ❙ AtomCARE containment b
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❙ 1133 ❙ AtomCARE b) Monitors a
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a) Calculates the distribution of t
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❙ 1137 ❙ AtomCARE telephone or
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Introduction Of e‐FAST and Exerci
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| List of Figures | Fig. 1.1 Schema
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C O N T E N T S 1.1. OVERVIEW OF TH
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❙ 1181 ❙ Nuclear Safety Informa
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1. Overview of OPiS C O N T E N T S
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Regional Basic Professional Training Course in Korea

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