Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety of experimental programs or economic considerations. The safety reassessments of old research and test reactors provide an opportunity to examine their overall safety and to look at the behavior of the equipment and the elements which are important for safety in more detail on the basis of past incidents and lessons learnt from operating experience. 4.2.5. Some severe accidents which have occurred in the past 21 August 1945 ‐ Los Alamos (USA) Criticality accident: when piling up the reflector blocks around a sub‐critical fuel assembly, the last block fell on the fuel assembly. The person carrying out the experiment received a dose equivalent of 5.1 Sv and died 28 days later. The guard of the building received an equivalent dose of 0.5 Sv. 21 may 1946 ‐ Los Alamos (USA) Similar criticality accident to the previous one and in the same installation: caused by accidentally bringing a hollow beryllium shell too close to the fuel when demonstrating how to induce a critical state in a fuel assembly. Eight people received equivalent doses of between 0.37 Sv and 2.1 Sv. The person carrying out the experiment died 9 days later. 12 December 1952 ‐ NRX ‐ Chalk River (Canada) Power excursion resulting from the regulating rods being inadvertently removed and failure of the safety rods to drop. This resulted in destruction of the core and release of ❙ 288 ❙
❙ 289 ❙ 4. DESIGN OF NUCLEAR REACTORS 3.7x1014 Bq that spread into the basement of the building with approximately 4000 m 3 of reactor coolant water. 29 November 1955 ‐ EBR ‐1 (USA) Power excursion during an experiment to measure the reactivity coefficient of the reactor. The reactimeter was not connected to the safety system. This accident caused the meltdown of around 40% of the core. 15 October 1955 ‐ Vinca (Yugoslavia) Uncontrolled criticality of the reactor due to an inadvertent increase in the level of the heavy water following a mistake made by an operator: six people received equivalent doses of between 4 Sv and 11 Sv. One person died and the other 5 had a bone marrow transplant in Paris. 03 January 1961 ‐ SL1‐ Idaho Falls (USA) Power excursion due to human error during the manual removal of the central regulating rod: two operators were killed outright by the explosion that followed this power excursion. A third person died two hours later from wounds to the head. 30 December 1965 ‐ Venus ‐ Mol (Belgium) Power excursion due to human error during the manual removal of a regulating rod: the operator received 5 Sv to the chest and 40 Sv to the foot (amputation of the leg). 07 November 1967 ‐ SiloeE ‐ Grenoble (France)
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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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Page 243 and 244: Design basis accidents ❙ 237 ❙
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Page 251 and 252: Deterministic approach The determin
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Page 293: ❙ 287 ❙ 4. DESIGN OF NUCLEAR RE
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Page 301 and 302: 4.2.6.2.3 Reactor Systems ❙ 295
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Page 311: 5. Siting Considerations Regional B
Page 314 and 315: 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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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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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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