Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety Trip Ctrl : To control the reactor and turbine trip manually Data Change : To change the data by the user Setpoint : To change the set point of NPP by the user ESFAS Setpoint : To change the set point of ESFAS (Engineered Safety Features Actuation Signal) Quit Sim : To exit Malfunction list for the abnormal & accident simulation is as follows; 1.1 Rod Control Mechanism: drop one rod 2.1~6 Auto CEA Control System: uncontrolled slow(fast) withdrawal(insertion) 3.1~4 Reactor Coolant Pumps: trip RCP no.1~4 3.5 Reactor Coolant Pumps: trip all (RCP) 4.1~3 Pressurizer Pressure Transmitter: fail low/high, normal 5.1~4 Pressurizer Main Spray valves: fail open/close/stuck, normal 6.1~2 Pressurizer Proportional Heaters: non‐operative, normal 7.1~3 Pressurizer Level Transmitter: fail low/high, normal 8.1~3 CVCS Letdown Lind: unisolable break, isolable break, normal 9.1~3 CVCS Automatic Makeup: water source blocked, boron source blocked, normal 10.1 Reactor Scram 11.1~2 Turbine Load Rejection: shed all load, shed part load 12.1~6 Atmosph. Dump Valve: fail open/shut/stuck, open/close 10pct then fail stuck, normal 13.1 Grid frequency perturbation ❙ 1160 ❙
14.1~2 Main Steam Line Break Inside/Outside Containment 15.1~3 SG Feed Control, Steam Flow Signals: one/all lost, normal 16.1~3 SG Feed Control, Feed Flow Signals: one/all lost, normal 17.1~2 main Feedwater Pumps: trip, normal ❙ 1161 ❙ Introduction Of e‐FAST and Exercise 18.1~6 Feedwater Control Valve: fail open/shut/stuck, open/close 10pct % fail stuck, normal 19.1~2 Feedwater Preheaters : bypassed/normal 20.1~2 Off‐Siter Power: blackout/normal 21.1~3 Steam Generator Tube Rupture: ont/multi tube, corrected 22.1~2 LOCA: small break in Discharge Cold/Hot Leg 22.3 LOCA: very small break in Suction Cold Leg 22.4 LOCA: small break in Discharge Cold Leg, no SI 22.5 LOCA: break corrected 2.2.3 Graphic Screen Control Function Mimics This represents each systems of NPP (Nuclear Power Plant) that e‐FAST is simulating, by 11 dynamic graphic screens. It displays main components such as reactor, steam generator, pump, valve and the value of specific variables. RCS : Reactor Coolant System CVCS : Chemical & Volume Control System SIS : Safety Injection System SCS : Shutdown Cooling System
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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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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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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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Page 1116 and 1117: Regional Basic Professional Trainin
Page 1121: AtomCARE Regional Basic Professiona
Page 1125 and 1126: AtomCARE 1. OVERVIEW OF NATIONAL RA
Page 1127 and 1128: ❙ 1121 ❙ AtomCARE The RETAC of
Page 1129 and 1130: ❙ 1123 ❙ AtomCARE secondary rad
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Page 1141 and 1142: a) Calculates the distribution of t
Page 1143 and 1144: ❙ 1137 ❙ AtomCARE telephone or
Page 1145: Introduction Of e‐FAST and Exerci
Page 1148 and 1149: | List of Figures | Fig. 1.1 Schema
Page 1185 and 1186: C O N T E N T S 1.1. OVERVIEW OF TH
Page 1187 and 1188: ❙ 1181 ❙ Nuclear Safety Informa
Page 1199: ❙ 1193 ❙ Nuclear Safety Informa
Page 1203: 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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