Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety Capability for containment pressure tests The containment structure shall be designed and constructed so that it is possible to perform a pressure test at a specified pressure to demonstrate its structural integrity before operation of the plant and over the plant’s lifetime. Containment leakage The reactor containment system shall be designed so that the prescribed maximum leakage rate is not exceeded in design‐basis accidents. The containment structure and equipment and components affecting the leak‐tightness of the system shall be designed and constructed so that the leak rate can be tested at the design pressure after all penetrations have been installed. Determination of the leakage rate of the containment system at periodic intervals over the service lifetime of the reactor shall be possible, either at the containment design pressure, or at reduced pressures that permit estimation of the leakage rate at the containment design pressure. Adequate consideration shall be given to the capability to control any leakage of radioactive materials from the containment in the event of a severe accident. Containment penetrations The number of penetrations through the containment shall be kept to a practical minimum. All penetrations through the containment shall meet the same design requirements as the containment structure itself. They shall be protected against reaction forces stemming from pipe movement or accidental loads such as those due to missiles, jet forces and pipe whip. Adequate consideration shall be given to the capability of penetrations to remain ❙ 258 ❙
functional in the event of a severe accident. Containment isolation ❙ 259 ❙ 4. DESIGN OF NUCLEAR REACTORS Each line that penetrates the containment as part of the reactor coolant pressure boundary or that is connected directly to the containment atmosphere shall be automatically and reliably sealable in the event of a design‐basis accident in which the leak‐tightness of the containment is essential to preventing radioactive releases to the environment that are above acceptable limits. These lines shall be fitted with at least two adequate containment isolation valves arranged in series (normally with one outside and the other inside the containment, but other arrangements may be acceptable depending on the design), and each valve shall be capable of being reliably and independently actuated. Each line that penetrates the primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one adequate containment isolation valve. This valve shall be outside the containment and located as close to the containment as practicable. Adequate consideration shall be given to the capability of isolation devices to maintain their function in the event of a severe accident. Containment air locks Access by personnel to the containment shall be through airlocks equipped with doors that are interlocked to ensure that at least one of the doors is closed during reactor operations and in design‐basis accidents. Internal structures of the containment The design shall provide for ample flow routes between separate compartments inside the containment. The cross‐sections of openings between compartments shall be of such dimensions as to ensure that the pressure differentials occurring during pressure
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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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Page 227 and 228: ❙ 221 ❙ 4. DESIGN OF NUCLEAR RE
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Page 245 and 246: Single failure criterion ❙ 239
Page 251 and 252: Deterministic approach The determin
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Page 287 and 288: of these exchangers, the water is c
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Page 301 and 302: 4.2.6.2.3 Reactor Systems ❙ 295
Page 305 and 306: 4.2.6.5. Neutron Activity Analysis
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Page 311: 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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function; 6. Safety Classification
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the safety function would be requir
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6.3.2. Description of Quality Group
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6.4.3. Seismic category 1: 6.4.3.1.
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7. Quality Assurance Regional Basic
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7.5.1. Manufacture Quality Assuranc
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C O N T E N T S 8.1. INTRODUCTION |
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❙ 471 ❙ 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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❙ 831 ❙ 15. Operational Safety
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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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❙ 1067 ❙ 21. Safety Culture 21.
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organization e.g. contractors. ❙
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❙ 1079 ❙ 21. Safety Culture All
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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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