Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety 8.5.6. Decrease in Reactor Coolant Inventory For a PWR, the initiating events that can result in decrease in reactor coolant inventory are as follows: • Inadvertent opening of pressurizer relief valve or safety valve • Steam generator tube rupture (SGTR) • Postulated pipe breaks within reactor coolant pressure boundary Among the events listed above, the first is an AOO and the remaining two are accidents. As aforementioned, the last is defined as the LOCA and is discussed in another chapter. The major concern of SGTR is a radioactive material release bypassing the containment to the offsite, due to a direct release of RCS coolant to the secondary system. A typical sequence of the SGTR, assuming the double ended rupture of a SG tube is presented here. A SGTR results in release of RCS coolant to the secondary system and potentially to the atmosphere if the secondary system pressure rises sufficiently to open the SG safety and/or relief valves. If normal operation of various plant control systems is assumed, the following sequence of events for significant break flow exceeding charging capability is expected: The pressurizer low pressure and low level alarm are actuated, and charging pump flow increases in an attempt to maintain the pressurizer level. On the secondary side, steam and feed flow mismatch occurs as feedwater flow to the affected SG is reduced as a result of break flow to the SG secondary side. Decrease in RCS pressure due to continued loss of reactor coolant inventory leads to a reactor trip signal. ❙ 496 ❙
❙ 497 ❙ 8-A. Deterministic Accident Analysis (Non‐LOCA) RCS cool‐down, following reactor trip, leads to a rapid decrease in pressurizer water volume and the SI signal is initiated on low pressurizer pressure. The SI signal automatically terminates normal feedwater supply and initiates auxiliary feed water system (AFWS) actuation. The SG blowdown liquid monitor and/or the condenser off‐gas air ejector radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system and will automatically terminate SG blowdown. The reactor trip automatically trips the turbine; if offsite power is available, steam dump valves open permitting steam to the condenser while in a loss of offsite power steam dump valves close. Following turbine trip, the steam flow rapidly decreases and the SG pressure rapidly increases, resulting in the steam discharge to the atmosphere through the SG relief and safety valves. Following reactor trip, continued action of the auxiliary feedwater supply from the condensate storage tank (CST) and borated SI flow from the refueling water storage tank (RWST) provides a heat sink to absorb the decay heat. For the SGTR, it is expected that the accident diagnostics and isolation procedure can be completed within 30 minutes of event initiation, considering the indications provided at the control board, together with the magnitude of break flow. However, it is assumed in the analysis that the operator action initiates 30 minutes after initiation of the accident, isolating the faulted SG, followed by reduction of the RCS temperature and pressure to the shutdown cooling or residual heat removal (RHR) condition. 8.5.7. Radioactive Release from a Subsystem or Component For a PWR, the initiating events that can result in readioactive release from a subsystem
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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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❙ 547 ❙ 8-B. Deterministic Acci
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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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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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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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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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