Regional Basic Professional Training Course in Korea

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Regional Basic Professional Training Course (BPTC) on Nuclear Safety characteristic features of a good safety culture. It is the task of safety management to improve all three of them, in order to obtain an ever‐higher safety culture. The shaping of attitudes and behaviour is a much more demanding task than the – equally necessary – task of improving competence and skills of individuals. It is the real core of the “management of safety” and, in spite of all doubts, it can be met successfully. We do perform this task in NPPs all over the world, and we do have some success in our endeavours. But it is my personal impression that, in a large number of cases, the success could be increased by applying a more systematic approach, or with some outside assistance. We at WANO are preparing for such assistance to our members. 21.4.3.5. Improving feedback on safety performance To facilitate an effective feedback system, safety performance must be measured routinely. The measures should have the capability to highlight whether the safety performance of the organization is being maintained or improved, and they should also allow the underlying causes of any performance deficiency to be identified. This is essential, if appropriate counter‐measures are to be identified. Measuring the safety performance is both easy, and delicate, simultaneously. I will come back to that problem in No. 4.4. After measures have been defined, audits and reviews should be performed on a regular basis. They should cover all safety‐related areas and should be carried out either internally or through independent institutions on a peer basis. IAEA, WANO and INPO offer corresponding services. In response to findings in audits and reviews, appropriate corrective actions have to be identified and implemented. The safety management has to ensure that this is done systematically, and includes an assessment of the effectiveness of the corrective actions after their implementation. ❙ 1086 ❙
21.4.4. Specific safety management issues 21.4.4.1. Flexibility ❙ 1087 ❙ 21. Safety Culture The description of the “management of safety” given above is neither complete nor generally applicable. In any case, due allowance should be given for adjustments to different cultures in different countries, and in different companies. It is the overall effectiveness that counts, not the specific composition. 21.4.4.2. Effectiveness The “management of safety” sets the framework for all endeavours to enhance safety culture. But it should be clearly pointed out that even the most perfect framework can never be a substitute for the commitment of the individuals concerned. A more perfect framework is a help to reach the goal with a higher degree of probability, but it can never be a warranty. To enhance safety culture, the “right” individual attitude of the senior managers is indispensable. If the senior managers do not feel it necessary to improve the safety culture, no framework and no system, however perfect, can be successful. Another problem with perfect systems holds even for “willing” people: Whenever a system is very perfect, people tend to perceive this system as a means that automatically achieves the goal, without the need for strenuous individual effort; the “system” solves the problem, not individual strain and engagement. This subjectively felt sense of relief from individual challenge and responsibility is a common phenomenon in case of perfect systems and also holds for the “management of safety”. A good system is essential; a perfect system can create new problems. Finally, to reach a high effectiveness level, all employees, managers and staff, have to accept the goal of high safety performance and to understand the processes and procedures the organization has chosen to pursue that goal. They have to do what is right,
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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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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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❙ 799 ❙ 13.Maintenance Manageme
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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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Page 1061 and 1062: 21. Safety Culture 21.1. DEFINITION
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Page 1139 and 1140: ❙ 1133 ❙ AtomCARE b) Monitors a
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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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