Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety leg break was additionally calculated. As listed in the Table, the peak cladding temperature was 1187℉ for 0.5 ft 2 pump discharge pipe break and other acceptance criterion including the maximum oxidization were satisfied. Table 8.8‐1 SBLOCA Analysis Condition Parameter Value Reactor Power (102% of nominal power) 2,871 MWt Average Linear Heat Generation rate (LHGR) 5.5 kW/ft (180.4 W/cm) Peak LHGR at Hot Rod 15.0 kW/ft (492.1 W/cm) Gap Conductance at Peak LHGR ❙ 542 ❙ 1,801 Btu/hr‐ft 2 ‐℉ (8,794 kcal/m 2 ‐hr‐℃) Fuel Centerline Temperature at Peak LHGR 3,634 ℉ (2001 ℃) Fuel Average Temperature at Peak LHGR 2,259 ℉ (1237 ℃) Gas Pressure of Hottest Rod 1,111 psia (78.1 kg/cm 2 A) Moderator Temperature Coefficient at Initial Density 0.0 ∆ρ/℉ (0.0 ∆ρ/℃) Axial Power Profile Top‐skewed RCS Flow Rate (Total) 121.5x10 6 lb/hr (55.1x10 6 kg/hr) Core Flow Rate 117.9x10 6 lb/hr (53.5x10 6 kg/hr RCS Pressure 2,250 psia (158.2 kg/cm 2 A) Core Inlet Temperature 564.5 ℉ (295.8 ℃) Core Exit Temperature 622.0 ℉ (327.8 ℃) Reactor Trip Setpoint on Pressurizer Low Pressure 1,555 psia (109.3 kg/cm 2 A) Safety Injection Setpoint 1,555 psia (109.3 kg/cm 2 A) SIT Pressure 585 psia (41.1 kg/cm 2 A) High Pressure Safety Injection Pump Shutoff Head 1,600 psig (112.5 kg/cm 2 ) Low Pressure Safety Injection Pump Shutoff Head 158.5 psig (11.1 kg/cm 2 ) Number of SG U‐tubes (un‐pluged/pluged per SG) 7,635/705
Table 8.8‐2 SBLOCA Break Spectrum Analysis for KSNP ❙ 543 ❙ 8-B. Deterministic Accident Analysis (LOCA) Break Size/ Location Peak Clad Temperature* Max. Oxidization** Hydrogen at Hottest Rod, (ft %*** 2 /cm 2 ) ℉(℃) Elevation ft (m) % Elevation ft (m) 0.50 (465)/PD 1,187 (642) 11.25 (3.43) 0.0157 11.25 (3.43) < 0.0023 0.35 (325)/PD 986 (530) 11.25 (3.43) 0.0025 11.25 (3.43) < 0.0004 0.20 (186)/PD 1,020 (549) 10.625 (3.24) 0.0021 11.25 (3.43) < 0.0003 0.05 (46.5)/PD 1,033 (556) 10.625 (3.24) 0.0007 11.25 (3.43) < 0.0002 0.02 (18.6)/PD 1,042 (561) 11.25 (3.43) 0.0010 11.25 (3.43) < 0.0003 0.03 (27.9)/HL 1,029 (554) 10.625 (3.24) 0.0008 11.25 (3.43) < 0.0002 * Acceptance Criteria: *2,200℉, ** 17%, *** 1.0%. Figures 8.8‐10 to 8.8‐12 show behaviors of RCS pressure, core level, and cladding temperature for 0.5 ft 2 pump discharge pipe break, respectively. Fig. 8.8‐13 shows a break spectrum analysis result. INNER VESSEL PRESSURE, psia 2400 2000 1600 1200 800 400 0 0 0 40 80 120 160 200 TIME, seconds Figure 8.8‐10 RCS Pressure of 0.5 ft 2 (PD) Break 168.72 160 140 120 100 80 60 40 20 INNER VESSEL PRESSURE, kg/cm 2 A
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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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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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❙ 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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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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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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