Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety Regulations on Technical Standards for Nuclear Reactor Facilities, etc. prescribe technical standard as regards structure, installations, performance, operation and quality assurance of nuclear reactor and related facilities. In Article 30, Emergency Core Cooling System, ECCS with sufficient capability shall be installed to meet each of the following requirements following loss of residual heat removal capability or loss of reactor coolant accidents, and such system shall meet the requirements determined and publicly notified by the Minster of Science and Technology: (1) Cladding temperature shall not exceed an acceptable design value (2) Oxidation and hydrogen generation in cladding shall be limited to and allowable level (3) Deformation of fuel and internal structures shall not significantly reduce the effective core cooling (4) Core cooling shall be ensured for a sufficient time. And it shall have the design feature of redundancy, leak detection, isolation, and containment capabilities to perform the safety function with sufficient reliability under the assumption of loss of offsite or onsite power (single failure). ECCS acceptance criteria are based on meeting the relevant requirements of the following regulations: Article 13, External Event Design Bases, Article 15, Environmental Effects Design Bases, Article 16, Sharing of Structures, Systems, and Components, Article 24, Electric Power System, Article 25, Control Room, Article 28, Reactivity Control System, Article 37, Overpressure Protection, and Article 41, Testability, Monitorability, Inspectability, and Maintainablity. As provided in Article 30 of Regulations on Technical Standards for Nuclear Reactor Facilities, MEST (Regulatory authority) Notice No. 2008‐16 requires that ECCS must be designed so that its calculated cooling performance following postulated LOCAs conforms to the criteria set forth in the followings [4]: (1) The calculated maximum fuel element cladding temperature shall not exceed 1204ºC (2) The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation (3) The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical ❙ 512 ❙
❙ 513 ❙ 8-B. Deterministic Accident Analysis (LOCA) amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel were to react (4) Calculated changes in core geometry shall be such that the core remains amenable to cooling (5) After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long‐ lived radioactivity remaining in the core. It includes the ECCS evaluation model and licensee’s reporting for any changes or error in an acceptable evaluation model. ECCS cooling performance shall be calculated in accordance with one of the acceptable evaluation models (conservative or best‐estimate models) and shall be calculated for a number of postulated loss‐of‐coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss‐of‐coolant accidents are calculated. ECCS cooling performance shall be calculated in accordance with one of the following acceptable evaluation models and shall be calculated for a number of postulated loss‐of‐ coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss‐of‐coolant accidents are calculated. (1) Conservative model The conservative model must provide the model and calculation method whose conservatism is justified and approved for use. A conservative ECCS evaluation model may be developed in conformance with the required and acceptable features of KINS Technical Guidance, KINS/GT‐N007‐1, corresponding to 10 CFR Part 50 Appendix K ECCS Evaluation Models [5]. (2) Best estimate model The best estimate model in conformance with the required and acceptable features of KINS Technical Guidance, KINS/GT‐N007‐2 [6], corresponding to US Regulatory Guide
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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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Page 468 and 469: Regional Basic Professional Trainin
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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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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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