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Timestamp: 2020-07-04 14:41:10
Document Index: 392621952

Matched Legal Cases: ['art 2', 'art 3', 'art 13', 'art 16', 'art 2', 'art 3', 'art 13', 'art 16']

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Jong C. Jo and Howard H. Chung
This chapter describes the nuclear regulatory organizations, aspects of the regulatory authority including licensing, status of nuclear installations, Korean nuclear reactor regulatory framework and regulations referring to domestic and/or international industrial codes and standards in the area of pressure vessels and piping, and status of Korean Electric Power Industry Codes (KEPIC) that covers standards for design, manufacturing, operation, maintenance, and testing and inspection of nuclear and non-nuclear mechanical components, including pressure vessels and piping. A comparative assessment of the U.S. and Korean codes is also AQ1 addressed. Besides, a comparison between the KEPIC code and its reference to codes of other foreign countries is also provided. The dependency of electricity production on nuclear energy is expected to increase continuously worldwide during next few decades, at least until some available and economically competitive alternative energy sources and/or production technologies are secured. Also, in the near future, the nuclear energy may take up an important position as a safe, environment-friendly, economically affordable, and sustainable energy source owing to various efforts to develop new evolutionary reactors that are underway in several countries. Since 1970s, Korea has been promoting the nuclear energy industry to produce electricity needed for the rapidly expanding industry and enhancing quality of human life. As a result of the intensive national program for nuclear energy promotion, at present there are 20 nuclear power reactors in operation, 4 reactors under construction, and another 4 reactors under consideration for construction in Korea. The reactor types and vendors are diverse and especially the first three units, Kori Units 1 and 2 of pressurized water reactors (PWRs) by Westinghouse Co. and Wolsung Unit 1 of pressurized heavy water reactor (PHWR) by Atomic Energy of Canada Limited (AECL), were supplied in the form of a turnkey system. To cope with this expected regulatory environment, the Korean nuclear regulatory authorities have been making effort to improve nuclear safety regulatory framework to establish new regulatory requirements, as necessary, and to streamline regulations. In the early stage of the introduction of reactors into Korea, as there was no well-established domestic regulatory framework for the safety regulation of operating reactors, the technical safety requirements and safety standards of the countries from which a reactor was sourced were applied to the reactor concerned, as necessary. Since then, the Korean regulatory authority has made an effort to establish its own effective and streamlining regulatory framework by making its own rules and regulations applicable to domestic nuclear reactors and by amending them. As a result of this effort, Korea has developed and issued, as of December 2005, the following rules and regulations, the number of articles contained in each of these rules and regulations are shown in parenthesis: · · · · Atomic Energy Act (122) Enforcement Decree of the Act, Presidential Decree (337) Enforcement Regulation of the Act (137); Enforcement Regulation Concerning the Technical Standards of Reactor Facilities (101) · Enforcement Regulation Concerning the Technical Standards of Radiation Safety Management (122) · 26 cases of Notices by the Minister for Science and Technology for the reactor regulation Some Notices of the Minister of Education, Science and Technology (MEST) concerning the reactor regulation have been prepared by adopting or referring to the U.S. NRC's technical safety requirements as well as the Nuclear Safety Standards of the International Atomic Energy Agency (IAEA) . As is well known, the detailed technical requirements of the reactor regulation are referred to or endorsed to some proper industrial codes and standards such as ASME B&amp;PV codes in the United States, RCC-M codes in France, JSME B&amp;PV codes in Japan, and others. Korea applied, as needed, the technical safety requirements and safety standards of the countries from which the reactors and/or technologies-related design, construction, and operation were imported. Since the type and suppliers of these reactors are
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diverse and the supplying countries are different, it was hard to consistently apply same detailed regulatory requirements to these reactors. This caused the designers, builders, and operators of nuclear power plants (NPPs) in Korea to be perplexed. To resolve this issue, a set of KEPIC for the application to the design, construction, and operation of electric power facilities including NPPs, has been developed on the basis of the prevailing U.S. codes and standards by Korea Electric Association (KEA). The first edition of KEPIC was published in 1995 and the Notice of the Minister of Education, Science and Technology &quot;The Guidelines for the Application of the Korea Electric Power Industry Code Issued by the KEA as the Technical Standards Related to the Construction and Operation of Nuclear Power Reactor and Related Facilities,&quot; defined in Articles 12 and 22 of the Atomic Energy Act, was notified publicly in 1996. KEPIC was initially started to be applied to the design and construction of Ulchin NPP Unit 5 and 6 in 1997 and its application has been extended to all NPPs either in operation, or in construction, or in contemplation of construction in Korea. Accordingly, since the execution of the Agreement on Technical Barriers to Trade (TBT) in 1995, the international trend of industrial codes and standards has been changing. To line up with the tendency of changing international environment, KEA has been putting a great deal of effort into the project of improving and expanding KEPIC. As a part of this effort, the 2005 Edition of KEPIC has been published to be retrofitted according to the ISO/IEC Guide 21 for the adoption of international standards as regional or national standards.
The authority and responsibilities of the governmental regulatory body MEST are prescribed in the Atomic Energy Act and the Enforcement Decree of the National Government Organization Act as follows: · To issue, amend, and revoke licenses for the construction and operation of nuclear installations and to take the necessary regulatory enforcement actions on the cases where regulatory requirements are not met or regulations are violated. · To conclude agreements with other domestic governmental or nongovernmental bodies and to delegate tasks to other organizations, where such delegation is directly essential for the performance of the regulatory body's responsibilities. · To obtain documents and opinions from public or private organizations or persons, which are necessary and appropriate. · To maintain contact with foreign regulatory bodies and relevant international organizations. · To access, at any time, premises of any nuclear installations licensed or under review. MEST is responsible for the establishment of the acceptance criteria for constructing and operating nuclear installation and technical standards for operational safety measures and for the assurance of compliances with regulations at every stage of the selection of sites, design, construction, commissioning, operation, and decommissioning of nuclear installations.
AQ2 AQ3
Nuclear safety regulatory organizations of Korea are mainly composed of the MEST with the Nuclear Safety Commission (NSC) as a safety regulatory authority and Korea Institute of Nuclear Safety (KINS) as a safety regulatory expert body, as shown in Fig. 69.1. MEST, the nuclear safety regulatory body, has full independent authority and responsibility for the safety regulations, including the issuance of permits and licenses for nuclear installations. The Minister, as an official member of the Atomic Energy Commission (AEC), participates in making decisions on major national policies related to the development and utilization of nuclear energy. As shown in Fig. 69.1., NSC, under the jurisdiction of the Minister of Education, Science and Technology, is responsible for deliberating and making decisions on important matters concerning nuclear safety. The Vice Minister and the Director General in charge of the Atomic Energy Bureau are on a vertical organization under the Minister. KINS pursues matters on nuclear safety regulation as entrusted by MEST in accordance with &quot;Atomic Energy Act.&quot; KINS also bears responsibility for various activities such as the development of nuclear safety regulatory technology, technical support to MEST for policy development and radiation protection, information management on safety regulation, and the monitoring and evaluation of environmental radioactivity.
NSC is established under the jurisdiction of MEST to deliberate and decide on important matters concerning nuclear safety, pursuant to the Atomic Energy Act. NSC is chaired by the Minister of Education, Science and Technology and consists of nine members, including eight members appointed or commissioned by the Minister. NSC organizes the Special Committee on Nuclear Safety to technically investigate and deliberate issues and concerns under its jurisdiction. The Commission deliberates and decides on the following matters: · Consolidation and coordination of matters concerning nuclear safety control · Matters concerning the regulation of nuclear materials and reactors · Matters concerning the protection against hazards due to radiation exposure · Matters concerning the plan for estimation and allocation of expenditures for nuclear safety control · Matters concerning the formulation of tests and research for nuclear safety control · Matters concerning the fostering and training of researchers and engineers in the area of nuclear safety control · Matters concerning the safety management of radioactive waste · Matters concerning the measures against radiation accidents · Other matters deemed important by the chairman
The primary mission of MEST is to ensure adequate protection of the public health and the environment against radiation hazards that are associated with the peaceful use of nuclear energy.
KINS was founded in December 1981 and initially operated under the name of &quot;Nuclear Safety Center&quot; (NSC), which had been attached to Korea Atomic Energy Research Institute (KAERI). It started to function as an independent expert organization
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 3
n tio ua al ev ults its s re bm l Su ca ni ch te w s st vie ue re eq
Is su es lic pe en rm se it o r
Ap pl ie s fo lic r p en er se mi
Applies for inspection Nuclear Industries Inspection Korea Institute of Nuclear Safety (KINS)
WORKING MECHANISM OF NUCLEAR SAFETY REGULATION
in February 1990 according to &quot;Korea Institute of Nuclear Safety Act.&quot; The following main technical activities for the MEST's nuclear safety regulations have been entrusted to KINS: · Safety reviews in relation to the licensing and approval of nuclear installations · Regulatory inspections during manufacturing, construction, and operation of nuclear installations · Research and development of the technical standards of safety regulation for nuclear installations · License examinations for the handling of nuclear materials and radioisotopes, and the operation of nuclear installations · Receive and process notifications relevant to licensing formalities · Quality assurance examination and inspection
The main legislation governing the safety of nuclear facilities in Korea is shown in Fig. 69.2. The figure shows the associated statutory provisions of the Atomic Energy Act, the Enforcement Decree, the Enforcement Regulations, and the Notice of the Minister of Education, Science and Technology. The Atomic Energy Act enacted in 1958 (as amended) defines fundamental issues and the general principles including requirements concerning the development, utilization, and safety regulation of nuclear energy. All provisions of nuclear safety regulation
and radiation protection are prescribed in the Atomic Energy Act that was established as the main law concerning safety regulations of nuclear installations. It also includes provisions of the AEC, the NSC, nuclear energy promotion program, Construction Permit (CP) and an Operating License (OL) of nuclear installations, and others. Table 69.1 shows the contents of the Atomic Energy Act of Korea. The Enforcement Decree of Atomic Energy Act (presidential decree) provides administrative matters and specific (technical) requirements necessary to enforce Atomic Energy Act. The Enforcement Regulation of the Act (the MEST Ordinance including Enforcement Regulations Concerning the Technical Standards of Reactor Facilities, and the Radiation Safety Management) provides particulars including the detailed procedure, format of documents, and technical standards, as entrusted by the Atomic Energy Act and the Enforcement Decree. The Notice of the Minister of Education, Science and Technology prescribe the detailed rules and procedural requirements for regulatory actions, specific regulations, technical codes and standards, and regulatory guidelines on the implementation of regulatory requirements, as entrusted by the Atomic Energy Act, the Enforcement Decree, and the Enforcement Regulation. Table 69.2 shows the list of 26 Notices of the Minister of Education, Science and Technology applicable for the design, construction, operation, inspection, testing, and so on of reactor facilities. The Notices of the Minister of Education, Science and Technology and Regulatory Guidelines may endorse or incorporate the industrial codes and standards developed by professional societies, such as KEPIC Code, ASME B&amp;PV Code, CSA/CAN3
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The Act provides the bases and the fundamental matters concerning the development and utilization of atomic energy and safety regulations
Enforcement Decree of the Act (Presidential Decree)
The Decree provides the particulars entrusted by the Act, which are necessary for the enforcement of the Act
Enforcement Regulation of the Act Enforcement Regulation Concerning the Technical Standards of Reactor Facilities, etc. Enforcement Regulation Concerning the Technical Standards of Radiation Safety Management, etc. Notice of the Minister of Education, Science and Technology The Regulation provides the technical standards and particulars entrusted by the Act and the Decree such as detailed procedures and format of documents
The Notice provides detailed particulars for the technical standards and guidelines
Codes and Standards for materials, design, test, and inspection of components and equipment
LEGISLATION SYSTEM [1, 2]
(Canadian Standards Association/ CANDU3), as detailed technical requirements for the design, operation, and inspection and testing of reactor pressure vessels and piping components. Finally, the Regulatory Guidelines on safety reviews and regulatory inspections developed by KINS and then endorsed by MEST for their use for regulatory purpose provide advice for the preparation of a license application and present acceptable implementation methods of regulations, regulatory evaluation techniques for specific problems, and data needed by the KINS staff in its regulatory safety reviews.
licensing process. The SDA system will ensure the validation of approved standard design without imposing additional regulatory requirements for the multiple CP applications of the same design of NPP applied within 10 years of the SDA issuance, and the same portions of NPP design that are approved as standard design will be excluded in the process of safety review for the following CP applications. 69.4.1.2 Early Site Approval To obtain an early site approval (ESA) with which a limited construction work on a proposed site can be started prior to the issuance of a CP, an application for the ESA with a site survey report and a radiological environmental report shall be filed with the Minister of Education, Science and Technology. Based on the results of the safety review by KINS of the ESA application, the Minister will give an official approval. The objective of the safety review is to evaluate the adequacy of the proposed nuclear site and the radiological impacts on the environment surrounding the nuclear installation. The Ministry of Environment (MOE) is in charge of reviewing nonradiological environmental impacts. 69.4.1.3 Construction Permit for Nuclear Installation To obtain a CP for nuclear installation, a CP application with the radiological environmental report, the preliminary SAR, and the quality assurance program for design and construction shall be filed
LICENSING SYSTEM AND SAFETY ASSESSMENT
The licensing process of nuclear installations consists of two steps including issuance of a CP and an OL, which are prescribed in the Atomic Energy Act; the early site approval system is considered as a preparatory step of the CP application, as shown in Fig. 69.3. 69.4.1.1 Standard Design Approval Standard Design Approval (SDA) system can be applied to a standard design of NPP with an enhanced level of safety to improve the efficiency of
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 5
CONTENTS OF THE ATOMIC ENERGY ACT [2, 3]
Title Chapter 1 Chapter 2 Chapter 3 General Provisions Atomic Energy Commission and Nuclear Safety Commission Establishment and enforcement of the overall nuclear energy promotion program, research and development, etc., of nuclear energy Nuclear energy research and development fund Construction and operation of nuclear power reactors and related facilities
Major Contents Purpose of this Act and definitions of the terminology used in this Act Establishment, functions, and composition of the Atomic Energy Commission and the Nuclear Safety Commission Establishment and enforcement of the comprehensive promotion plan for nuclear energy, nuclear energy research and development institution, burden of cost for nuclear energy research and development work
Chapter 3-2 Chapter 4
Establishment, management, and operation of the fund Criteria for permit (license), licensing procedures, license application documents to be submitted, regulatory inspection, records and keeping, appointment (dismissal) and obligation of responsible persons for nuclear reactor operation, notification of suspension or disuse of operation, transfer and inheritance, measure for suspension, decommissioning and penalty surcharge Criteria for permit (license), licensing procedures, license application documents to be submitted, regulatory inspection, records and keeping, appointment (dismissal) and obligation of responsible persons for nuclear reactor operation, notification of suspension or disuse of operation, transfer and inheritance, measure for suspension, and decommissioning Criteria for permit (license), licensing procedures, license application documents to be submitted, regulatory inspection, records and keeping, appointment (dismissal) and obligation of responsible persons for nuclear reactor operation, notification of suspension or disuse of operation, transfer and inheritance, measure for suspension, and decommissioning Criteria for permit (license), licensing procedures, license application documents to be submitted, regulatory inspection, records and keeping, appointment (dismissal) and obligation of responsible persons for nuclear reactor operation, notification of suspension or disuse of operation, transfer and inheritance, measure for suspension, and decommissioning Criteria for permit (license), licensing procedures, license application documents to be submitted, and regulatory inspection Criteria for permit (license), licensing procedures, license application documents to be submitted, and regulatory inspection Criteria for permit (license), licensing procedures, license application documents to be submitted, and regulatory inspection Criteria for permit (license), licensing procedures, and regulatory inspection Permit for construction and operation of disposal facilities, and regulatory inspections Registration of personnel dosimetry service and regulatory inspection License examination and certificate of license Establishment of exclusion area and preventive measures against radiation hazards Conditions for permit or designation, approval of report on specific technical subjects, hearing, protection for the individual in charge of safety management, education, and training Penal provisions, fine for negligence, and joint penal provisions Enforcement date, transitional measures, and relations with other laws
Construction of nuclear power reactors and related facilities
Operation of nuclear power reactors and related facilities
Construction and operation of nuclear research reactors, etc.
Chapter 6 Section 1 Section 2 Chapter 7 Chapter 8 Chapter 9 Chapter 10 Chapter 11 Chapter 12
Nuclear fuel cycle enterprise and use, etc., of nuclear materials Nuclear fuel cycle enterprise Use of nuclear materials Radioisotopes and radiation generating devices Disposal and transport Personnel dosimetry service License and examination Regulation and supervision Supplementary provisions
Chapter 13 Addenda
with the Minister of Education, Science and Technology. KINS reviews the CP application with submittals, including Preliminary Safety Analysis Report. Then the CP is issued by the Minister after NSC deliberation of the KINS review results.
The KINS review of the CP application is performed to confirm if the site and the preliminary design of the nuclear installation comply with the relevant regulatory requirements and technical guidelines. It addresses the design principle and concept of the
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NOTICES OF THE MINISTER OF EDUCATION, SCIENCE AND TECHNOLOGY APPLICABLE TO REACTOR FACILITIES [2, 4]
Notice no. 00-08 05-08 05-04 05-19 04-17 01-38 01-39 01-40 04-15 01-43 05-07 01-46 01-47 05-09 02-05 04-13 04-14 02-21 05-03 03-11 03-12 05-10 03-19 03-20 06-05 07-18
Title Technical Standards of the Location, Structures and Installation of Reactor Facilities Regulation on Other Facilities Related to Nuclear Reactor Safety Guidelines for the Application of the Korea Electric Power Industry Codes to the Technical Standards of Reactor Facilities Regulation on Preparation, etc. of Radiation Environmental Report of Nuclear Power Utilization Facilities Regulation on Survey and Evaluation of Environmental Radiation in the Vicinity of Nuclear Power Utilization Facilities Technical Standards for Safety Valve and Relief Valve of Reactor Facilities Standards for Performance of Emergency Core Cooling System of the Pressurized Water Reactor Pressure Integrity Test Criteria for Major Components of Reactor Facilities Standards for Leak Rate Test of Reactor Containment Regulation on Disposition and Management of Inspection Findings from Nuclear Facilities Regulation on the Reporting and Public Announcement of the Accident and Incident for Nuclear Facilities Standard for Preparation of Operational Technical Specification Detailed Standards for Quality Assurance of Reactor Facilities Regulation on Pre-Service Inspection of Reactor Facilities Regulation on the First Review Schedule for the Periodic Safety Review of Reactor Facilities Regulation on In-Service Inspection of Reactor Facilities Regulation on In-Service Testing of Safety-Related Pumps and Valves Regulation on Safety Classifications and Applicable Codes &amp; Standards for Reactor Facilities Material Surveillance Criteria for Reactor Pressure Vessel Technical Standards for Investigation and Evaluation of the Meteorological Conditions of Reactor Facility Sites Technical Standards for Investigation and Evaluation of the Hydrological and Oceanographic Conditions of Reactor Facility Sites Regulation on Items and Method of Periodic Inspection for Reactor Facilities Regulation on Establishment and Implementation of Fire Protection Program Technical Standards for Fire Hazards Analyses Subjects to be Discussed According to Installation of Industrial Facilities etc. around the Nuclear Facilities Guidelines on the Application of Technical Standards for Evaluating the Continued Operation of Nuclear Facilities
Numbers of Amendments 1 3 2 4 3 1 1 2 1 4 4 2 0 1 0 3 0 1 3 0 0 0 0 0 0 0
nuclear installation, the implementation of the regulatory criteria, the evaluation results of the effects of construction on the environment, and a proposal for minimizing those effects. A radiological environmental report submitted for the CP application or the ESA application should contain the public opinions from residents of the area surrounding the nuclear installation through a public hearing, if necessary. 69.4.1.4 Operating License (OL) for Nuclear Installation To obtain an OL for a nuclear installation, the OL application should have the operational technical specifications, the final SAR, the quality assurance program for operation, and the radiological emergency plan. The application shall be filed with the Minister of
Education, Science and Technology. Based on the KINS review results of the OL application and preoperational inspections, the Minister issues the OL after deliberation by the NSC. The safety review of the OL application is conducted to confirm if the final design of the nuclear installation complies with the relevant regulatory requirements and technical guidelines and if the nuclear installation can continue to operate throughout its lifetime. 69.4.1.5 Amendment to the OL for Nuclear Installation To make modifications to the specifics for which the OL has been given, such as changes in the operational technical specifications or in the design that may affect the safety of operating nuclear installations, it is necessary to obtain approval from the Minister of
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 7
Korea Hydro &amp; Nuclear Power Co. Ltd. Company (KHNP)
Apply for early Site approval (ESA) Construction permit Stage Operating license stage Commercial operation stage
Request safety review for ESA
Conduct safety review
Launch foundation work
Grant ESA
Submit safety review report
Apply for construction permit (CP)
Request safety review for CP
Issue CP &lt;Nuclear Safety Commission Review&gt;
Apply for preoperational inspections (POIs) Apply for operating license (OL)
Request POIs
Conduct POIs
Request safety review for OL
Submit safety review report Load fuel Issue OL &lt;Nuclear Safety Commission Review&gt; Start-up test Submit POIs report (construction, performance) Submit POI report (Start-up test) Inform the POI results Commence commercial operation
Reload fuel and apply for periodical inspections (PI)
Conduct PI
Inform PI results
Submit PI report
REACTOR LICENSING AND REGULATION SYSTEM
Education, Science and Technology. The procedure for the approval for an amendment to the OL is the same as the application for an OL. A safety evaluation is performed to confirm if the amendment to the OL affects the operational safety of nuclear installation. 69.4.1.6 Approval for Decommissioning of Nuclear Installation In case where the operator (licensee) of a nuclear installation intends to decommission it, to obtain an approval of the Minister of Education, Science and Technology, a decommissioning plan shall be submitted.
KINS conducts a safety review for the application of decommissioning approval. The review includes safety evaluations of the radiation protection during decommissioning, the radiological impacts on the environment surrounding the nuclear installation after decommissioning, and the proposal for minimizing the impacts.
Regulatory inspections for a nuclear installation include the preoperational inspection for nuclear installations under construction, the periodic inspection for operating nuclear installations, the quality assurance audit, the daily inspection by resident inspectors, and
8 · Chapter 69
Submitting application document for inspections
Receiving Application document and reviewing
Preparing inspection plan
Reviewing and adjusting inspection plan
Receiving inspection plan and preparing inspections
Informing inspection plan
Attending premeeting
Holding Premeeting (inspection items and method)
Resident Inspector attending premeeting to comment
Reporting findings of inspections
Receiving and reviewing Findings of inspections
Requesting corrective actions
Reviewing corrective actions results
Requesting review of corrective action results
If unsatisfactory, re-action
Informing whether results are satisfactory
Preparing and submitting review report
Receiving and reviewing review report
Planning corrective actions and reporting results
Formally informing results including any corrective actions
Closing after confirmation
REACTOR INSPECTION PROCESS
the special inspection pursuant to the Atomic Energy Act. The general inspection procedure is given in Fig. 69.4. 69.4.2.1 Preoperational Inspection for the Nuclear Installations under Construction The preoperational inspection for the set up of nuclear installations is conducted to verify if the nuclear installation is properly constructed in conformity with the conditions of the CP and if the constructed nuclear installation can be operated safely throughout its lifetime. It is
conducted for the construction quality and the operational performance of the facilities by means of a document review and a field inspection. 69.4.2.2 Periodic Inspection for Operating Nuclear Installations The periodic inspection for a nuclear installation in operation is conducted to confirm if the nuclear installation has been properly operating in conformity with the OL conditions, if it can be maintained to be continuously competent for
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 9
the high pressure and radiation conditions or other operating environments, and if the performance of the installation maintains license-based conditions. To do this, a document review and a field inspection are performed for a PWR plant during the refueling outage period and for a PHWR plant during the periodic maintenance. 69.4.2.3 Quality Assurance Audit The quality assurance audit is performed to verify if all activities affecting the quality at every stage of the design, construction, and operation of a nuclear installation are being performed in conformity with the quality assurance program approved by the regulatory body. It is performed periodically for nuclear installations in operation. 69.4.2.4 Daily Inspection by Resident Inspectors The main purpose of the daily inspection is to daily check the nuclear installations either under construction or in operation. It consists of a field inspection on the surveillance tests, an investigation on the measures taken when the reactor comes to reach an abnormal state, and a verification of the adequacy of the operator's activity regarding the radiation control. 69.4.2.5 Special Inspection When a serious or potentially serious safety issue is identified or encountered, the special inspection consisting of an examination of the issue and an in-depth field investigation is performed to obtain a resolution for preventing any potential accidents or recurrence of the occurred accidents.
It is prescribed in the Atomic Energy Act that any violation of the relevant provisions specified in the Act shall cause a penalty and/or a fine according to its seriousness of violation.
LOCATIONS OF THE NUCLEAR POWER PLANTS IN KOREA
The status of the nuclear installations in Korea is shown in Fig. 69.5. As of January 2008, there are 20 units of NPPs in operation, which produce about 40% of the total electricity generation, 4 units under construction, and another 4 units under consideration for construction. The 20 operating units consist of 16 PWR-type units and 4 PHWR-type units, while all 8 units under construction or under consideration for construction are of PWR type. Kori Unit 1, the first NPP in Korea, started its commercial operation in April 1978. The reactor types, total installed capacity, and reactor suppliers of operating NPPs in Korea are shown in Table 69.3.
INTRODUCTION TO THE NOTICES OF THE MINISTER OF THE EDUCATION, SCIENCE AND TECHNOLOGY RELATED TO NUCLEAR POWER REACTOR BOILER AND PRESSURE VESSELS IN KOREA
If the safety review results confirm that the CP application complies with the relevant requirements, the Minister of Education, Science and Technology will issue a CP. The Minister may impose additional conditions on the issuance of CP to the minimum, if judged necessary to secure safety. If any violation is identified in the process of the regulatory inspection, the Minister may order the license holder to take appropriate corrective measures in accordance with the Atomic Energy Act. The Minister of Education, Science and Technology is authorized to order the operators (licensees) to submit the needed documents concerning their business and supplemental materials, if considered necessary for the enforcement of the regulations. The Minister may also conduct a regulatory inspection to verify if the documents comply with the field conditions and order the operator to take appropriate corrective measures, if any, on the basis of the inspection results. The Minister of Education, Science and Technology may order the revocation of the permit (or license) or the suspension of business during a period of not exceeding one year, if one or more of the following cases are applicable: · The case where the installer or operator has modified any matters concerning the permit (or license) without approval. · The case where the installer or operator has failed to meet the criteria for permit (or license). · The case where the installer or operator has violated an order of the Minister of Education, Science and Technology issued to take corrective measures on the basis of the results of regulatory inspection for the construction or operation of a nuclear installation. · The case where the installer or operator has violated any of the permit (or license) conditions or regulations on safety measures during the operation of a nuclear installation.
Among the 26 Notices of the Minister of Education, Science and Technology applicable for the design, construction, operation, inspection, testing, and so on of reactor facilities, the following 5 Notices are related to the reactor boiler and pressure vessels, including piping, pumps, valves, and heat exchangers. These endorse or refer to publicly authorized industrial codes and standards such as detailed technical requirements or standards. · Notice No. 02-21, &quot;Regulation on Safety Classifications and Applicable Codes &amp; Standards for Reactor Facilities&quot; (issued in 1994 and amended in 2002) applied the ASME B&amp;PV Section III for the PWR plants and the CAN3/CSAN285.0 and N285.1 for the PHWR plants in prescribing the safety classification for the structures, systems, and components important to safety and the applicable codes and standards in accordance with Article 12(1) of the Regulations on Technical Standards for Nuclear Reactor Facilities, etc. · Notice No. 00-08, &quot;Technical Standards of the Location, Structures and Installation of Reactor Facilities&quot; (issued in 1983 and amended in 2000) applied the ASME B&amp;PV Code, Section III, Subsections NCA, NB, NC, and ND as the standards for safety valves and relief valves installed at the nuclear facilities, which was later replaced with the Notice No. 01-38, &quot;Technical Standards for Safety Valve and Relief Valve of Reactor Facilities&quot; (issued in 2000 and amended in 2001). · Notice No. 05-03, &quot;Material Surveillance Criteria for Reactor Pressure Vessel&quot; (issued in 1992 and amended in 2000, 2003, and 2005) referred to the term &quot;reference nil-ductility transition temperature (RTNDT)&quot; defined in the ASME B&amp;PV Code, Section III, NB-2330. · Notice No. 04-13, &quot;Regulation on In-Service Inspection of Reactor Facilities&quot; (issued in 1995 and amended in 1998, 2002, and 2004) applied the inspection standards provided in
10 · Chapter 69
: in operation : under construction : under consideration of construction China
Seoul Yellow Sea Daejeon Daegu Yonggwang 1,2,3,4,5,6 Gwangju Wolsong 1,2,3,4 Shin-Wolsong 1,2 Ulchin 1,2,3,4,5,6 Shin-Ulchin 1,2
Kori 1,2,3,4 Shin-Kori 1,2,3,4 Busan
FIG. 69.5
LOCATIONS OF COMMERCIAL NUCLEAR POWER PLANTS IN KOREA (AS OF JANUARY 2008)
the ASME B&amp;PV Code, Section XI, Division 1, &quot;Rules for InService Inspection of Nuclear Power Plant Components for the PWR Plants&quot; and the CAN/CSA-N285.4, &quot;Periodic Inspection of CANDU Nuclear Power Plant Components&quot; and CAN/CSA-N285.5, &quot;Periodic Inspection of CANDU Nuclear Power Plant Containment Components&quot; for the PHWR plants. · Notice No. 04-14, &quot;Regulation on In-Service Testing of Safety-Related Pumps and Valves&quot; (issued in 1995 and amended in 1998, 2002, and 2004) applied the testing standards provided in the ASME B&amp;PV Code, Section XI, Subsections IWP (for pumps) and IWV (for valves) or ASME Operation and Maintenance Code, ISTB (for pumps) and ISTC (for valves). Prior to the issue of each of the above five Notices, the regulations and safety standards of the countries from which a reactor was supplied were applied to the subject reactor, as necessary. The Notices endorsed or incorporated the ASME B&amp;PV Code for the PWR plants and the CAN/CSA and ASME B&amp;PV Codes for the PHWR plants until the Notice of the Minister of Education, Science and Technology No. 05-04 &quot;Guidelines for the Application of the Korea Electric Power Industry Codes to the Technical Standards of Reactor Facilities&quot; became effective. Since the Notice No. 05-04 of the Minister of Education, Science and Technology was issued in 1996 for the first time and
amended in 2000 and 2005, the Notices of the Minister of Education, Science and Technology, which endorsed or referred to the ASME B&amp;PV and/or the CAN/CSA Codes as discussed above, are being amended to incorporate the KEPIC instead of the corresponding ASME B&amp;PV Codes or alternatively to apply the ASME B&amp;PV and/or the CAN/CSA Codes.
GUIDELINES FOR THE APPLICATION OF THE KOREA ELECTRIC POWER INDUSTRY CODES TO THE TECHNICAL STANDARDS OF REACTOR FACILITIES (NOTICE OF THE MINISTER OF EDUCATION, SCIENCE AND TECHNOLOGY NO. 05-04)
The full text of the Notice No. 05-04 &quot;Guidelines for the Application of the Korea Electric Power Industry Codes to the Technical Standards of Reactor Facilities&quot; amended in 2005 is discussed below: Notice of the Minister of Education, Science and Technology No. 2005-04 (MEST Reactor 021) The Guidelines for the
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 11
COMMERCIAL NUCLEAR POWER PLANTS IN KOREA (AS OF JANUARY 2008)
Station name Kori-1 Kori-2 Kori-3 Kori-4 ShinKori-1 ShinKori-2 ShinKori-3 ShinKori-4 Ulchin-1 Ulchin-2 Ulchin-3 Ulchin-4 Ulchin-5 Ulchin-6 Wolsong-1 Wolsong-2 Wolsong-3 Wolsong-4 ShinWolsong-1 ShinWolsong-2 Yonggwang-1 Yonggwang-2 Yonggwang-3 Yonggwang-4 Yonggwang-5 Yonggwang-6
Reactor Type PWR PWR PWR PWR PWR PWR APR APR PWR PWR PWR PWR PWR PWR PHWR PHWR PHWR PHWR PWR PWR PWR PWR PWR PWR PWR PWR
Capacity Operator (MWe) (Owner) 573 637 963 967 960 960 1400 1400 939 937 994 993 994 991 578 684 682 685 1000 1000 945 939 985 988 987 993 KHNP
NSSS Supplier WH WH WH WH DOOSAN DOOSAN DOOSAN DOOSAN FRAM FRAM KHI/KAERI/CE KHI/KAERI/CE DOOSAN DOOSAN AECL AECL/KHI AECL/KHI AECL/KHI DOOSAN DOOSAN WH WH KHI/KAERI/CE KHI/KARI/CE DOOSAN DOOSAN
Issuance of Initial Construction Criticality Permit
1972-05-31 1977-06-19 1977-06-26 1978-04-29 in operation 1978-11-18 1983-04-09 1983-04-22 1983-07-25 1979-12-24 1985-01-01 1985-01-22 1985-09-30 1979-12-24 1985-10-26 1985-11-15 1986-04-29 2005-07-01 2010-08-01 2010-12-31 2005-07-01 1983-01-25 1983-01-25 1993-07-16 1993-07-16 1999-05-17 1999-05-17 1978-02-15 1992-08-28 1994-02-26 1994-02-26 2007-06-04 2007-06-04 1981-12-17 1981-12-17 1989-12-21 1989-12-21 1997-06-14 1997-06-14 1988-02-25 1989-02-25 1997-12-21 1998-12-14 2003-11-28 2004-12-16 1982-11-21 1997-01-29 1998-02-19 1999-04-10 1986-01-31 1986-10-15 1994-10-13 1995-07-07 2001-11-24 2002-09-01 2011-08-01 2011-12-31 1988-04-07 1989-04-14 1998-01-06 1998-12-28 2003-12-18 2005-01-07 1982-12-31 1997-04-01 1998-03-25 1999-05-21 1986-03-05 1986-11-11 1994-10-30 1995-07-18 2001-12-19 2002-09-16
under construction construction planned
1988-09-10 in operation 1989-09-30 1998-08-11 1999-12-31 2004-07-29 2005-06-01 1983-04-22 1997-07-01 1998-07-01 1999-10-01 under construction 1986-08-25 in operation 1987-06-10 1995-03-31 1996-01-01 2002-05-21 2002-12-24
AECL, Atomic Energy of Canada Limited; CE, Asea Brown Boveri-Combustion Engineering; DOOSAN, Doosan Heavy Industries Co.; FRAM, Framatom; KAERI, Korea Atomic Energy Research Institute; KHI, Korea Heavy Industries Co.; KHNP, Korea Hydro &amp; Nuclear Power Co.; WH, Westinghouse Electric Co.
Application of the Korea Electric Power Industry Code issued by the Korea Electric Association as the technical standards related to the construction and operation of nuclear power reactor and related facilities, defined in Articles 12 and 22 of the Atomic Energy Act, are hereby notified publicly as follows: March 22, 2005 Minister of Education, Science and Technology Guidelines for the Application of Korea Electric Power Industry Code (KEPIC) as Technical Standards of Nuclear Reactor Facilities (Table 69.4) Article 1 (Purpose) The purpose of this notice is to prescribe the necessary requirements related to
scope, method, etc., in applying Korea Electric Power Industry Code as the technical standard for the nuclear power reactor and its related facilities (hereinafter referred to as &quot;nuclear reactor facilities&quot;) defined in Articles 12 and 22 of the Atomic Energy Act. Article 2 (Definitions of Terms) Definitions of the terms used in this notice shall be as follows: 1. The term &quot;technical standards for nuclear reactor facilities&quot; means the technical standards used to confirm safety of nuclear reactor facilities by the Minister of Education, Science and Technology, which are based on the standards of construction permit and operating license provided in Articles
12 · Chapter 69
KEPIC TO BE APPLIED AS TECHNICAL STANDARD OF NUCLEAR REACTOR FACILITIES
Category QAP
Title Nuclear Quality Assurance Authorized Inspection Certificate of Registered Professional Engineer General Requirements Class 1 Components Class 2 Components Class 3 Components Class MC Components Component Supports Core Support Structures Appendices General Requirements Class 1 Components Class 2 Components Class 3 Components Class MC and CC Components Class 1, 2, 3, and MC Component Supports Requirements for Class CC Concrete Components Appendices General Requirements In-Service Testing of Pumps In-Service Testing of Valves In-Service Testing of Pressure Relief Devices In-Service Testing of Snubbers Performance Testing of Closed Cooling Water Systems
Reference Standards ASME NQA-1 (`1994 Edition 1995 Addenda) ASME QAI-1 (1995 Edition 1996 Addenda) ASME Appendix XXIII (1996 Addenda) ASME III NCA (1995 Edition 2000 Addenda) ASME III Div. 1 NB (1995 Edition 2000 Addenda) ASME III Div. 1 NC (1995 Edition 2000 Addenda) ASME III Div. 1 ND (1995 Edition 2000 Addenda) ASME III Div. 1 NE (1995 Edition 2000 Addenda) ASME III Div. 1 NF (1995 Edition 2000 Addenda) ASME III Div. 1 NG (1995 Edition 2000 Addenda) ASME III Div. 1 NZ (1995 Edition 2000 Addenda) ASME XI Div. 1 IWA (1995 Edition 2000 Addenda) ASME XI Div. 1 IWB (1995 Edition 2000 Addenda) ASME XI Div. 1 IWC (1995 Edition 2000 Addenda) ASME XI Div. 1 IWD (1995 Edition 2000 Addenda) ASME XI Div. 1 IWE (1995 Edition 2000 Addenda) ASME XI Div. 1 IWF (1995 Edition 2000 Addenda) ASME XI Div. 1 IWL (1995 Edition 2000 Addenda) ASME XI Div. 1 Appendix (1995 Edition 2000 Addenda) ASME OM-ISTA (1995 Edition 1999 Addenda) ASME OM-ISTB (1995 Edition 2000 Addenda) ASME OM-ISTC (1995 Edition 2000 Addenda) ASME OM-Appendix I (1995 Edition 2000 Addenda) ASME OM-lSTD (1995 Edition 2000 Addenda) ASME OM-Part 2 (1994 Edition 1999 Addenda)
QAI QAR MNA MNB MNC
KEPIC shall be prior KEPIC shall be prior
MN (Nuclear mechanical components)
MND MNE MNF MNG MNZ
KEPIC shall be prior for MNA
Table 2 shall be applied
MI (In-service inspection of nuclear power plant components)
MIA MIB MIC MID MIE MIF MIL MIZ
MO (In-service testing of nuclear power plant components)
MOA MOB MOC MOD MOE MOF
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 13
TABLE 69.4 Area Category MOG MOH Title Vibration Testing of Piping Systems Performance Testing and Monitoring of Power-Operated Relief Valve (PORV) Assemblies In-Service Testing and Maintenance of Diesel Devices Code Cases General Requirements Functional Qualification of Active Pump Assemblies Functional Qualification of Active Valve Assemblies General Requirements
Continued Reference Standards ASME OM-Part 3 (1994 Edition 2000 Addenda) ASME OM-Part 13 (1994 Edition 2000 Addenda) Remarks
ASME OM-Part 16 (1994 Edition 1999 Addenda) ASME OMN Code Case (1998 Edition 2000 Addenda) ASME QME-1 Section QR (1997 Edition 1998 Addenda) ASME QME-1 Section QP (1997 Edition) ASME QME-1 Section QV (1997 Edition 2000 Addenda)
MON MFA MF (Functional qualification of mechanical equipment used in nuclear power plants) EN (Nuclear electrical components) MFB MFC ENA ENB
END ENF SNA SN (Nuclear structures) SNB SNC SND STA
ST (Structure general) STB
ANSI/ANS 51.1-1983 (R1988) Design IEEE 279 (1971, R78), 308 (1991), 352 (1987, R93), 379 (1994, 2000), 384 (1992, R97), 420 (1982), 494 (1974, R9), 497 (1981), 577 (1976, R92), 603 (1998), 7-4.3.2 (1993), 1023 (1988, R95), ANSI/ISA S67.04 (1994, 2000) Qualification IEEE 323 (1983, R96), 344 (1987), 420 (1982), 627 (1980, R96) Periodic Surveillance Testing IEEE 338 (1987, R93) General Requirements ASME III, NCA (1995 Edition 2000 Addenda) Concrete Containment ASME III, Div.2 CC (1995 Edition 2000 Addenda) Concrete Structures ACI 349 (1997) Steel Structures AISC N690 (1994) Design Loads ASCE 7 (1998) Load criteria and interpretation for architecture issued by the Architectural Society of Korea (2000) Seismic Analysis ASCE 4 (1986), IEEE 344 (1987), ASME QME-1 (1997), ANSI/ANS-2.2 (1988)
KEPIC shall be prior
12 and 22 of the Atomic Energy Act and the technical standards provided in Articles 3 through 85 of the Regulations on Technical Standards for Nuclear Reactor Facilities, etc; 2. The term &quot;Korea Electric Power Industry Code&quot; (KEPIC) means the industrial technical standards developed and maintained by the Korea Electric Association for applying to the electrical industry; and
3. The term &quot;reference standards&quot; means the technical standards used as technical basis in developing KEPIC as defined in the column of &quot;reference standards&quot; of Table 1. Article 3 (Scope of Application) 1. This notice may be applied to the safety-related facilities classified in accordance with &quot;Regulation on Safety
14 · Chapter 69
Classification and Applicable Codes and Standards for Nuclear Reactor Facilities.&quot; 2. The whole or partial portion of the KEPIC defined in this notice may be applied optionally to the specific nuclear reactor facilities. In this case, the details such as applicable date, scope of application, etc. of the KEPIC shall follow the licensing conditions or the related regulations for the facilities. 3. KEPIC applied as the technical standards of nuclear reactor facilities in accordance with this notice shall be limited to those corresponding standards in KEPIC 2000 Edition, 2001 Addenda, 2002 Addenda and 2003 Addenda to the reference standards of Table 1. However, Paragraph 2 may apply to the technical standards, which are not included in Table 1. Article 4 (Application Method) Method of application of the KEPIC to the technical standards for nuclear reactor facilities shall follow Articles 5 through 8. Article 5 (Interpretation of Standard) If there is an argument among the related parties on the interpretation of KEPIC as technical standard for nuclear reactor facilities, the interpretation of the Minister of Education, Science and Technology shall prevail. Article 6 (Application of Technical Contents Verified) The KEPIC contents, which have been verified as suitable to assure the safety of domestic or foreign nuclear reactor facilities shall be applied as the technical standards of nuclear reactor facilities as follows: 1. The reference standards shall be applied in case there are any differences in technical contents between KEPIC and reference standards. Provided, that KEPIC shall be applied when the priority of KEPIC is stipulated in Remarks of Table 1 or when the Minister of Education, Science and Technology deems KEPIC as applicable; and 2. The technical contents of reference standards which are not included in KEPIC may be applied as the technical standards
of the nuclear reactor facilities when the Minister of Education, Science and Technology deems it necessary for the safety of nuclear reactor facilities. Article 7 (Application of Technical Contents Unverified) 1. The contents of KEPIC which have not been used in domestic or foreign nuclear reactor facilities, nor verified as suitable to assure the safety may be applied only in case the Ministry of Education, Science and Technology approves its suitability. 2. The limitations defined in Table 2 shall be followed for the application of KEPIC (Table 69.5). Article 8 (Report) 1. The Korea Electric Association shall report the status of development and maintenance of KEPIC, operation status of certification system, etc. semi-annually to the Minister of Education, Science and Technology. The Minister may request the corrective action for the reported contents. 2. The Korea Electric Association shall take a corrective action for the request of Paragraph 1 unless there is any special reason. Addenda Article 1 (Enforcement Date) This notice shall be effective on the date of its promulgation. Article 2 (Relationship with Other Existing Notices) Relationship between this notice and the other existing notices shall be as follows, in spite of the enforcement of this notice. 1. The KEPIC issued before this notice may be applied only if such KEPIC meets requirements of this notice. 2. The detailed requirements for &quot;Safety Classification and Applicable Codes and Standards,&quot; &quot;In-Service Inspection,&quot; and &quot;In-Service Test&quot; shall be defined independent of this notice. Article 3 (Repeal of Notice) Notice of the MEST No.2000-17 &quot;Guidelines for the Application of KEPIC as the Technical Standards of Nuclear Reactor Facilities&quot; is repealed at the time this notice becomes effective.
LIMITATIONS FOR APPLICATION OF KEPIC 2000 EDITION AND 2001/2002/2003 ADDENDA
Limitations (1) The requirements of KEPIC, which are not consistent with the quality assurance requirements of the Atomic Energy Law and its subordinate law, shall not be applied. (1) Weld leg dimensions Licensees shall not apply Paragraph MNB-3683.4(3)(A), the equation of footnote (7) in Figure MNC-3673.2-1 and Figure MND-3673.2-1. (2) Seismic design Licensees shall use Articles MNB-3200, MNB-3600, MNC-3600, and MND-3600 up to and including the 1995 Edition. Licensees shall not use these Articles in the 1996 Edition through the latest edition and its addenda. (3) Independence of inspection in Licensees shall not apply MNA 4200.10(1). AQ4 (1) Licensees shall not apply the reference standards MIA-1600-1 in Table. (2) Licensees shall apply re-certification period of 3 years only for levels I and, II instead of 5 years AQ5 for levels I, II, and III defined in MIA 2314. (3) The authorized inspector in MIA 4410, which allows the other procedure of welding material control in case of acceptance by the authorized inspector, shall be changed to &quot;regulatory agency.&quot; (4) The sentence of exemption of the periodic system pressure test for the penetrating piping of containment vessel defined in MIA 5110(3) shall be deleted. (continued)
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 15
Limitations (5) The in-service inspection program for steam generator tubing is governed by the relevant requirements in the technical specifications for operation, etc. (6) Examination of concrete containment structure. For the applications of MIL of KEPIC 2001/2002/2003 Addenda, the licensee shall apply the following additionally. (A) For Class CC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report required by MIA-6000 (1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation. (2) An evaluation of each area and the result of the evaluation. (3) A description of necessary corrective actions. (B) Personnel who examine containment concrete surfaces and tendon hardware, wires, or strands shall meet the qualification provisions in MIA 2300. The &quot;owner-defined&quot; personnel qualification provisions in MIL 2310(4) are not approved for use. (7) Examination of metal containments and the liners of concrete containment licensees applying Subsection IWE, KEPIC 2001/2002/2003 Addenda, shall satisfy the followings additionally. (A) For Class MC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report as required by MIA-6000: (1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation. (2) An evaluation of each area, and the result of the evaluation. (3) A description of necessary corrective actions. (B) The following requirements may be used as an alternative to the requirements of IWE-2430: If the examinations reveal flaws or areas of degradation exceeding the acceptance standards of Table MIE-3410-1, an evaluation must be performed to determine whether additional component examinations are required. For each flaw or area of degradation identified, which exceeds acceptance standards, the licensee shall provide the following in the ISI Summary Report required by MIA-6000: (1) A description of each flaw or area, including the extent of degradation, and the conditions that led to the degradation. (2) The acceptability of each flaw or area, and the need for additional examinations to verify that similar degradation does not exist in similar components. (3) A description of necessary corrective actions. (4) The type and number of additional examination to verify the similar degradation. (C) A general visual examination must be performed once each period. When performing remotely the visual examinations, the maximum direct examination distance specified in Table MIA- 2210-1 may be extended and the minimum illumination requirements specified in Table MIA- 2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination. (D) VT-1 and VT-3 examinations must be conducted in accordance with MIA-2200. Personnel conducting examinations in accordance with the VT-1 or VT-3 examination method shall be qualified in accordance with MIA-2300. The &quot;owner-defined&quot; personnel qualification provisions in MIE-2330(1) for personnel who conduct VT-1 and VT-3 examinations are not approved for use. (E) The VT-3 examination method shall be used for the examinations in Items E1.12 and E1.30 of Table MIE-2500-1, and the VT-1 examination method shall be used for the examination in Item E4.11 of Table MIE-2500-1. An examination of the pressure-retaining bolted connections in Item E1.11 of Table MIE-2500-1 shall be conducted once each period using the VT-3 examination method. The &quot;owner-defined&quot; visual examination provisions in MIE2310(1) are not approved for use for VT-1 and VT-3 examinations. (F) Containment bolted connections that are disassembled during the scheduled examinations in Item E1.11 of Table MIE-2500-1 shall be examined using the VT-3 examination method. (continued)
16 · Chapter 69
Limitations Flaws or degradation identified during the VT-3 examination must be examined using the VT-1 examination method. The criteria in the material specification or MIB-3517.1 shall be used to evaluate containment bolting flaws or degradation. As an alternative to performing VT-3 examinations of containment bolted connections that are disassembled during the scheduled performance of Item E1.11, VT-3 examinations of containment bolted connections may be conducted whenever containment bolted connections are disassembled for any reason. (G) The ultrasonic examination acceptance standard specified in MIE-3511.3 for Class MC pressure-retaining components shall also be applied to metallic liners of Class CC pressure-retaining components. (H) Following items shall be examined additionally: (1) Circumferential welds of flued head and bellows seal penetration shall be examined in addition to the item E1.10 of Table MIE 2500-1. (2) Sealants, gaskets, dissimilar metal welds, and bolt connections shall be examined in accordance with items E5.10, E5.20, E7.10, and E8.20 in Table MIE 2500-1 of KEPIC 2000 Edition, respectively. Class 1 piping Licensees may not apply MIB-1220, &quot;Components Exempt from Examination,&quot; of KEPIC and shall apply IWB-1220, 1989 Edition of ASME Code Section XI. Underwater Welding The provisions in MIA-4660 &quot;underwater welding&quot; of KEPIC MI are not approved for use on irradiated material. Flaws of Class 3 piping ASME Code Case N-513 (Rev. 0), &quot;Evaluation Criteria for Temporary Acceptance of Flaws in Class 3 Piping,&quot; and N-523-1, &quot;Mechanical Clamping Devices for Class 2 and 3 Piping&quot; may be applied. For the applications of Code Case N-523-1, the licensee shall apply all the requirements of this Code Case. For the applications of Code Case N-513, the licensee shall apply all the requirements of this Code Case on the following conditions. (A) For the applications of Code Case N-513, specific safety factors of Article 4.0 shall be met. (B) Code Case N-513 may not be applied in the following cases: (1) Components other than pipe and tube, such as pumps, valves, expansion joints, and heat exchangers. (2) Leakage through flange gasket (3) Nonstructural seal-welded threaded connections to prevent leakage (integrity of thread shall be maintained even thought leak path of seal weld is not a structural flaw). (4) Failed socket weld. MIZ, Appendix VIII personnel qualification All personnel qualified for performing ultrasonic examinations in accordance with Appendix VIII shall receive 8 h of annual hands-on training on specimens that contain cracks. Training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility. Training time requirement may not be applied in case that the personnel continuously performs ultrasonic examination continually. MIZ, Appendix VIII specimen set requirements (A) When applying Supplements 2, 3, and 10 to Appendix VIII, the following examination coverage criteria requirements must be used. (1) Piping must be examined in two axial directions, and when examination in the circumferential direction is required, the circumferential examination must be performed in two directions, provided access is available. Dissimilar metal welds must be examined axially and circumferentially. (2) Where examination from both sides is not possible, full coverage credit may be claimed from a single side for ferritic welds. Where examination from both sides is not possible on austenitic welds or dissimilar metal welds, full coverage credit from a single side may be claimed only after completing a successful single-sided Appendix VIII demonstration using flaws on the opposite side of the weld. Dissimilar metal weld qualifications must be demonstrated from the austenitic side of the weld and may be used to perform examinations from either side of the weld. (continued)
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 17
Limitations (B) The following provisions must be used in addition to the requirements of Supplement 4 to Appendix VIII: (1) Related to Paragraph 3.1 (Detection acceptance criteria) of Supplement 4, personnel are qualified for detection if the results of the performance demonstration satisfy the detection requirements of Appendix VIII, Table VIII-S4-1 and no flaw greater than 0.25 in. through wall dimension is missed. (2) Related to Paragraph 1.1(5) (Detection test matrix) of Supplement 4, flaws smaller than the 50% of allowable flaw size, as defined in MIB-3500, need not be included as detection flaws. For procedures applied from the inside surface, use the minimum thickness specified in the scope of the procedure to calculate a/t. For procedures applied from the outside surface, the actual thickness of the test specimen is to be used to calculate a/t. (C) When applying Supplement 4 to Appendix VIII, the following provisions must be used: (1) A depth sizing requirement of 0.15 in. RMS must be used in lieu of the requirements in Subparagraphs 3.2(1) of Supplement 4, and a length sizing requirement of 0.75 in. RMS must be used in lieu of the requirement in Subparagraph 3.2(2). (2) In lieu of the location acceptance criteria requirements of Subparagraph 2.1(2) of Supplement 4, a flaw will be considered detected when reported within 1.0 in. or 10% of the metal path to the flaw, whichever is greater, of its true location in the X and Y directions. (3) In lieu of the flaw-type requirements of Subparagraph 1.1(5)(a) of Supplement 4, a minimum of 70% of the flaws in the detection and sizing tests shall be cracks. Notches, if used, must be limited by the following: (a) Notches must be limited to the case where examinations are performed from the clad surface. (b) Notches must be semielliptical with a tip width of less than or equal to 0.010 in. (c) Notches must be perpendicular to the surface within ± 2 deg. (4) In lieu of the detection test matrix requirements in Paragraphs 1.1(5)(b) and 1.1(5)(c) of Supplement 4, personnel demonstration test sets must contain a representative distribution of flaw orientations, sizes, and locations. (D) The following provisions must be used in addition to the requirements of Supplement 6 to Appendix VIII: (1) With regard to Paragraph 3.1(Detection Acceptance Criteria) of Supplement 6, the following provisions must be met for the detection qualification of personnel: (a) No surface-connected flaw greater than 0.25 in. through wall has been missed. (b) No embedded flaw greater than 0.50 in. through wall has been missed. (2) With regard to Paragraph 3.1(Detection Acceptance Criteria) of Supplement 6, all flaws within the scope of the procedure are detected for procedure qualification. (3) With regard to Paragraph 1.1(2) of Supplement 6, flaws smaller than the 50% of allowable flaw size, as defined in MIB-3500, need not be included as detection flaws. Flaws that are less than the allowable flaw size, as defined in MIB-3500, may be used as detection and sizing flaws. (4) Notches are not permitted. (E) When applying Supplement 6 to Appendix VIII, the following provisions must be used: (1) A depth sizing requirement of 0.25 in. RMS must be used in lieu of the requirements of Subparagraphs 3.2(1), 3.2(3)(b), and 3.2(3)(c) of Supplement 6. (2) With regard to the location acceptance criteria requirements in Subparagraph 2.1(2) of Supplement 6, a flaw will be considered detected when reported within 1.0 in. or 10% of the metal path to the flaw, whichever is greater, of its true location in the X and Y directions. (3) In lieu of the length sizing criteria requirements of Subparagraph 3.2(2) of Supplement 6, a length sizing acceptance criteria of 0.75 in. RMS must be used. (4) In lieu of the detection specimen requirements in Subparagraph 1.1(5)(a) of Supplement 6, a minimum of 55% of the flaws must be cracks. The remaining flaws may be cracks or fabrication-type flaws, such as slag and lack of fusion. The use of notches is not allowed. (5) With regard to Paragraphs 1.1(5)(b) and 1.1(5)(c) detection test matrix of Supplement 6, personnel demonstration test sets must contain a representative distribution of flaw orientations, sizes, and locations. (continued)
18 · Chapter 69
Limitations (F) The following provisions may be used for personnel qualification for combined Supplements 4 and 6 to Appendix VIII. Licensees choosing to apply this combined qualification shall apply all of the provisions of Supplements 4 and 6 including the following provisions: (1) For detection and sizing, the total number of flaws shall be at least 10. A minimum of 5 flaws shall be those from Supplement 4, and a minimum of 50% of the flaws shall be those from Supplement 6. At least 50% of the flaws in any sizing must be cracks. Notches are not acceptable for Supplement 6. (2) Examination personnel are qualified for detection and length sizing when the results of any combined performance demonstration satisfy the acceptance criteria of Supplement 4 to Appendix VIII. (3) For examination of reactor pressure vessel nozzle-to-shell welds conducted from the outside of the vessel. (a) The clad to base metal interface and the adjacent metal to a depth of 15% T (thickness measured from the clad to base metal interface) must be examined from one radial and two opposing circumferential directions using a procedure and personnel qualified in accordance with Supplement 4 to Appendix VIII, as modified by (B) and (C) of (12) in this limitations, for examinations performed in the radial direction, and Supplement 5 to Appendix VIII, as modified by (J) of (12) in this limitations, for examinations performed in the circumferential direction. (b) The examination volume not addressed by [(K)(3)(a)] in this paragraph must be examined in a minimum of one radial direction using a procedure and personnel qualified for single-sided examination in accordance with Supplement 6 to Appendix VIII, as modified by (D), (E), (F), and (G) of (12) in this limitations. (4) Table VIII-S7-1, &quot;Flaw Locations and Orientations,&quot; of Supplement 7 to Appendix VIII, shall be modified as follows: Table VIII S7-1 Flaw Locations and Orientation Group Inner 15 percent OD Surface Subsurface Parallel to weld Perpendicular to weld O O O O
(L) As a modification to the requirements of Supplement 8, Subparagraph 1.1(3) to Appendix VIII, notches may be located within one diameter of each end of the bolt or stud. (M) When implementing Supplement 12 to Appendix VIII, only the provisions related to the coordinated implementation of Supplement 3 to Supplement 2 performance demonstrations are to be applied. (13) MIZ, Appendix VIII single side ferritic vessel and piping and stainless steel piping examination. (A) Examinations performed from one side of a ferritic vessel weld must be conducted with equipment, procedures, and personnel who have demonstrated proficiency with single side examinations. To demonstrate equivalency to two-sided examinations, the demonstration must be performed to the requirements of Appendix VIII as modified by this paragraph and (B) through (G) of (12) in this limitations, on specimens containing flaws with nonoptimum sound energy reflecting characteristics or flaws similar to those in the vessel being examined. (B) Examinations performed from one side of a ferritic or stainless steel pipe weld must be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single side examinations. To demonstrate equivalency to two-sided examinations, the demonstration must be performed to the requirements of Appendix VIII as modified by this paragraph and (A) of (12) in this limitations. (14) Certification of NDE personnel (A) Level I and II nondestructive examination personnel shall be recertified on a 3-year interval in lieu of the 5-year interval specified in MIA-2314 (1) and (2). (continued)
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 19
Limitations (B) Paragraph MIA-2316 may only be used to qualify personnel that observe for leakage during system leakage and hydrostatic tests conducted in accordance with MIA-5211(1) and (2). (C) When qualifying visual examination personnel for VT-3 visual examinations under Paragraph MIA-2317, the proficiency of the training must be demonstrated by administering an initial qualification examination and administering subsequent examinations on a 3-year interval. Alternative nondestructive examination methods The provisions for the substitution of alternative examination methods, a combination of methods or newly developed techniques defined in MIA-2240 in KEPIC 2000 must be applied. The provisions in MIA-2240 of KEPIC 2001 Addenda (including its later addenda) are not approved for use. The provisions in MIA-4520(3), allowing the substitution of alternative examination methods, a combination of methods or newly developed techniques for the methods specified in the Construction Code are not approved for use. System leakage tests When performing system leakage tests in accordance with MIA-5213(1), a 10-min hold time after attaining test pressure is required for Class 2 and 3 components that are not in use during normal operating conditions. No hold time is required for the remaining Class 2 and 3 components provided that the system has been in operation for at least 4 h for insulated components or 10 min for un-insulated components. Table MIB-2500-1 examination requirements (A) The provisions of Table MIB-2500-1, Examination Category B­D, Full Penetration Welded Nozzles in Vessels, Items B3.120 and B3.140 (examination plan B) in the KEPIC 2000 Edition, must be applied when using the KEPIC 2002 Addenda (including its later edition and addenda). A visual examination with enhanced magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length criteria in Table IWB-3512-1, may be performed in lieu of an ultrasonic examination. (B) The provisions of Table IWB-2500-1, Examination Category B­G-2, Item B7.80 in the ASME 1995 Edition, are applicable only to reused bolting. (C) The provisions of Table IWB-2500-1, Examination Category B­K, Item B10.10, in the ASME 1995 Addenda must be applied. Motor-operated valve testing shall be performed in accordance with the requirements of in-service operating test for category A, B defined in MOC 4200, KEPIC 2000, or the requirements of valve testing defined in MOC 3500, KEPIC 2001/2002/2003 Addenda. And a program to ensure that motor-operated valves continue to be capable of performing their design basis safety functions shall be established. Code Cases except MON-1 &quot;Alternative Requirements for pre-service and in-service testing of motor-operated valve assemblies used in nuclear power plants&quot; may be applied through such procedures as approval of relief request. When applying Appendix I, &quot;Check Valve Condition Monitoring Program&quot; of the MOC, following requirements shall be satisfied. (A) Valve opening and closing functions must be demonstrated when flow testing or examination methods (nonintrusive, or disassembly and inspection) are used. (B) The initial test interval may not exceed two fuel cycles or 3 years, whichever is longer. Any extension of this interval may not exceed one fuel cycle per extension with the maximum interval not to exceed 10 years. Trending and evaluation of existing data must be used to reduce or extend the time interval between tests. (C) If the condition monitoring program is discontinued, then the test requirements of MOC 4510 through 4540, KEPIC 2000 Edition, or MOC 3510/3520/3540/5221, KEPIC 2001/2002/2003 Addenda must be applied. MOE &quot;In-service testing for Snubbers&quot; instead of the requirements of Snubbers defined in MIF 5200(1), (2) and MIF 5300(1), (2) may be applied by making appropriate changes to their technical specifications or licensee-controlled documents. Manual valves shall be tested at 2 year interval instead of 5 year interval defined in MOC 3540, KEPIC 2002 Addenda. (1) STA 2000, STA 3270, STA 4000s for design load, which are standards for commercial facilities, are not approved for use in nuclear reactor facilities.
ST (Structure general)
20 · Chapter 69
INDUSTRIAL CODE IN KOREA: KOREA ELECTRIC POWER INDUSTRY CODE (KEPIC) [5, 6]
Background and Status of KEPIC Development
Since 1970s, Korea has been promoting the nuclear energy industry to secure a stable, economical, and environment-friendly energy needed for the rapidly growing industry for increasing the standard of living, which demands a lot of energy. In the early stage of 1970s, when Kori Unit 1 and Wolsong Unit 1 were imported and constructed on a turnkey basis, Korea had neither a matured technology needed for the design, construction, and operation of electric power plant nor had their own detailed regulatory technical codes and standards. During 1980s, as an engineering self-reliance program for electric power plant systems and components was established, Korea firstly accomplished the system standardizations of 500 MWE class for fossil power and 1000 MWE class for nuclear power. In the process, some action plans were devised and one of them was to develop Korean codes and standards with a detailed approach toward the engineering self-reliance in power plant materials and components. Korea Electric Power Corporation started a feasibility study of the program for developing a set of Korean codes and standards for power plant materials and components in 1987, at the request of the Korean Government. As a result of the feasibility study, the title of the Korean codes and standards under consideration for development was proposed as Korea Electric Power Industry Code (KEPIC) and the following key principles were suggested to be applied as the bases of the code development. · The scope of KEPIC should cover the safety and reliabilityrelated materials and components. · The part of KEPIC for the safety-related system of nuclear power unit should be based on those applied for the construction of Younggwang nuclear power units 3 and 4. · The part of KEPIC for the fossil power system should be based on those applied for the construction of Tae-An fossil power Units 1 and 2. · The part of KEPIC for the nonsafety-related system of nuclear unit should be applicable for the fossil power facilities. · The technical requirements of KEPIC are basically the same as those of the authorized foreign and/or international industrial codes and standards selected as references such as ASME B&amp;PV Code, IEEE Code, and so on. · The administrative requirements of KEPIC should be suitable for the existing administrative elected systems of Korea by modifying the authorized foreign and/or international industrial codes and standards that are selected as references. · The structure of KEPIC concerning electricity and power facilities shall match with the existing codes and standards that had been endorsed or referred in the regulations. In 1992, the development of KEPIC was initiated on the basis of the above key principles. The first edition of KEPIC Codes composed of five parts in 33 volumes was published in 1995 by KEA. KEA was founded in 1965 for the purpose of promoting and advancing the technologies in the fields of electric power generation, equipment manufacturing and construction, and electrical safety.
In accordance with the Notice No. 05-04 of the Minister of Education, Science and Technology, which was issued in 1996, part of KEPIC Code was initially applied for the construction of Ulchin Nuclear Units 5 and 6 from the year of 1997. The Notice was amended in 2000 and 2005 to endorse the application of later editions of KEPIC Code to all phases of nuclear power project such as the subsequent constructions of NPPs as well as to the designs of new reactor systems such as the SMART-P, an integraltype PWR. KEPIC is also being applied to the design, construction, and operation of fossil power plants. Pursuant to the Agreement on Technical Barriers to Trade (TBT) in 1995, the international trend of codes and standards has been changing rapidly. To keep abreast of such international environment, the KEA has been trying to amend the KEPIC Code AQ6 continuously (Table 69.6).
Contents of the KEPIC
The KEPIC consists of two main areas, technical requirements and administrative requirements. Technical requirements are based on the U.S. codes and standards. They are same as the corresponding standards except for the use of KEPIC's own numbering system. The administrative requirements of each standard have been developed by modifying the ASME Section III, NCA General Requirements to be suitable for the industrial circumstances and situation of each technical sector in Korea. The ISO 9001 quality management system and the authorized inspection system are adopted for the non-nuclear safety sector. The contents of KEPIC developed so far, with the corresponding reference codes on which each KEPIC provisions are based, are given in Table 69.4. While the subparts for nuclear structures, systems, and components of KEPIC are applicable to all items and activities related to nuclear safety, those for non-nuclear structures, systems, and components of KEPIC are applicable to nonsafety items of NPPs or fossil power plants. Both nuclear- and non-nuclear-related subparts can be identified respectively with the second letter, &quot;N or G if any,&quot; of two capital-lettered symbols representing the titles of KEPIC subparts (area). For example, MN denotes the subpart for nuclear mechanical components, whereas MG denotes the nonnuclear mechanical components. Other subparts of the KEPIC Code, such as MD (materials), ME (nondestructive examination), MQ (welding and brazing qualification), SW (structural welding), and ST (extra provisions for structure) are applicable to either nuclear or non-nuclear items that are common to N or G. The subpart ET for the transmission, transformation, and distribution standards are partially applicable to switchyard items at power plants.
69.8.3
Development Procedure of the KEPIC
The first step of the KEPIC development process is to prepare a preliminary draft of code, which is conducted by the working groups comprised of technical experts from the industry pertaining to corresponding areas. The draft is reviewed by the subcommittees, consisting of experts from the industry, academia, research sectors, and authorities; opinions and comments on it from the relevant industry are also sought. The corresponding subcommittee revises the draft by considering or reflecting the review comments and opinions and refers it to the relevant technical committee. The final revised draft becomes effective on approval by the technical committee. Figure 69.6 shows the organization chart of KEPIC committees.
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 21
COMPARISON BETWEEN KEPIC CODE AND REFERENCED FOREIGN OR INTERNATIONAL CODES AND STANDARDS FOR THE MECHANICAL COMPONENTS AND STRUCTURES (1/2)
Area QA (Quality assurance)
Category QAP QAI QAR MNA MNB MNC MND MNE MNF MNG MNZ MGA MGB MGC MGD MGE MGF MGG MGH MGI MCN MCF MHA MHB MHD MDF MDN MDW MDP MEN MQW MQB MIA MIB MIC MID MIE MIF MIL MIZ MOA MOB MOC MOD MOE MOF MOG MOH MOI MIA MIB MIC MBB MTG SNA SNB SNC SND
Title Nuclear Quality Assurance Authorized Inspection Certificate of Registered Professional Engineer General Requirements Class 1 Components Class 2 Components Class 3 Components Class MC Components Component Supports Core Support Structures Appendices General Requirements Pressure Vessel Heat Exchanger Storage Tank Piping Pump Valve Condenser Feed water Heater Crane for Nuclear Power Plants Crane for Fossil Power Plants General Requirements Air Cleaning and Conditioning Testing Procedure Ferrous material Non ferrous metal Welding Rods, Electrodes, and Filler Metals Allowable Stress Non destructive Examination Welding Qualification Brazing Qualification General Requirements Class 1 Components Class 2 Components Class 3 Components Metallic Containment &amp; Metallic Liner Class 1, 2, 3, and MC Component Supports Class CC Concrete Containment Structure Appendices General Requirements Pump Valve Pressure Relief Device Snubber Cooling System Performance Test Piping System Vibration Test Motor-Operated Pressure Relief Valve Diesel-Driven Equipment General Requirements Active Pump Assembly Active Valve Assembly Boiler Turbine/Generator General Requirements Containment Structure Steel-Concrete Structure Steel Structure
Reference Standards ASME NQA ASME QAI-1 ASME Appendix. XXIII ASME Section .III NCA ASME Section III, Div. 1 NB ASME Section III, Div. 1 NC ASME Section III, Div. 1 ND ASME Section III, Div. 1 NE ASME Section III, Div. 1 NF ASME Section III, Div. 1 NG ASME Section III, Div. 1 NZ ASME Section III NCA ASME Section VIII, Div. 1 HEI, TEMA API 650 ASME B 31.1 HI ASME B 16.34 HEI HEI ASME NOG-1 CMAA-70 ASME AG-1, Div.I ASME AG-1, Div.II ASME ASME Section II, Part A ASME Section II, Part B ASME Section II, Part C ASME Section II, Part D ASME Section V ASME Section IX, Part QW ASME Section IX, Part QB ASME XI, Div. 1 IWA ASME XI, Div. 1 IWB ASME XI, Div. 1 IWC ASME XI, Div. 1 IWD ASME XI, Div. 1 IWE ASME XI, Div. 1 IWF ASME XI, Div. 1 IWL ASME XI, Div. 1 Appendix ASME OM Subsection ISTA ASME OM Subsection ISTB ASME OM Subsection ISTC ASME OM Mandatory Appendix 1 ASME OM Subsection ISTD ASME OM S/G Part 2 ASME OM S/G Part 3 ASME OM S/G Part 13 ASME OM S/G Part 16 ASME QME-1 Section QR ASME QME-1 Section QP ASME QME-1 Section QV ASME Section I RRC-TA ASME Section III NCA ASME Section III Div. 2 ACI 349, ACI 318 AISC-N690 (continued)
MG (General mechanical components)
MC (Crane) MH (Air cleaner and conditioner) MD (Material) ME (NDE) MQ (Welding) MI (In-service inspection of NPP components)
MO (In-service testing of NPP components)
MF (Functional qualification of nuclear mechanical components) MB (Boiler) MT (Turbine/Generator) SN (NPP Structure)
22 · Chapter 69
Area SG (Structure general) ST (General rules of structure) SW (Structure welding) FP (Fire protection) ND (NPP Design) NF (Nuclear fuel) NR (Radiation) KEPIC-E (Electrical)
Category SGA SGB SGC SGD STA STB SWS SWT FPC FPN FPF NDA NFA EN EM EE EC ET
Title General Requirements Steel-Concrete Structure Steel Structure-Allowable Stress Design Method Steel Structure-Load Resistance Coefficient Design Method Design Loads Seismic Analysis Steel Structure Thin Steel Plate Structure Common Requirements Fire Protection of Nuclear Power Plants Fire Protection of Fossil Power Plants Design of Nuclear Power Plants Nuclear Fuel Radiation Class 1E Equipment Measuring &amp; Control Equipment Electric Equipment Cables &amp; Raceways Transmission, Transformation, and Distribution
Reference Standards ASME Section III NCA ACI-318 AISC-ASD AISC-LRFD ASCE 7 ASCE 4 AWS D 1.1 AWS D 1.3 NFPA 10, 11, 12, 13, 14, 15, 24 NFPA 20, 803, 804 NFPA 850 ANS-51.1 RCC-C ANS IEEE, ANSI, ISA, etc. IEEE, ISA, IEC, etc. NEMA, IEC, ANSI, etc. ASTM, NEMA, IEEE, etc. IEC, IEEE
RRC-TA, French NPP Code-Turbine &amp; Generator; RCC-C, French NPP Code- Nuclear Fuel.
69.8.4
It is required in the KEPIC that qualified organizations and individuals should perform their appropriate functions to achieve the safety and reliability goals of the NPPs. This conformity assessment system includes accreditation for nuclear safety-related organizations and the qualification for personnel, such as
authorized inspectors /supervisors, registered Professional Engineers (RPEs), and NDE personnel. The nuclear certification system including the N-type certificate, authorized nuclear inspection and pressure relief testing laboratory, and so on. is also very similar to that of ASME except that KEPIC includes the organizations related to class 1E items
Quality Assurance Nuclear Mechanical Technical Committee Technical Committee Technical Committee
Structural Electrical Technical Committee Technical Committee
Fire Protection Technical Committee
Subcommittee (1) Quality system
Subcommittee (3) NPP design Nuclear fuel Radiation
Subcommittee (8) Nuclear mechanical components Boiler &amp; pressure vessel Materials Welding NDE T/G Fluid components Auxiliary components
Subcommittee (4) Concrete containment Reinforced concrete structure Steel structures Seismic design
Subcommittee (5) Nuclear electrical Induced &amp; rotating equipment I&amp;C Breakers &amp; isolators Cables &amp; raceways
KEPIC COMMITTEES
COMPANION GUIDE TO THE ASME BOILER &amp; PRESSURE VESSEL CODE · 23
ASME Section.III, Div.1
KEPIC - MN
Components 1 Parts &amp; appurtenance 3 Installation 2 2NC 3NP 1N
FIG. 69.7 CODE SYMBOL STAMPS OF ASME AND KEPIC
and seismic category I structures. KEPIC specifies that every organization including the owner, designer, manufacturer, installer, and material organization shall obtain a certificate from KEA according to the general requirements of each nuclear standard. Especially, the nuclear mechanical items need the code data report and stamping. The code symbol stamps of ASME Section III, Div.1 and KEPIC-MN are shown for comparison in Fig. 69.7. KEPIC-MN (nuclear mechanical) and KEPIC-SNB (concrete containment) require that pressure-retaining items shall be inspected at both phases of manufacturing and site installation by authorized nuclear inspectors, who are affiliated with the authorized inspection agency accredited by KEA in accordance with the requirements of KEPIC-QAI (authorized inspection). Both authorized nuclear inspectors and authorized nuclear inspector supervisors shall be qualified by KEA. Pressure relief devices, such as safety valves, relief valves, rupture disks, and so on, shall comply with the requirements for overpressure protection of KEPIC-MN, and shall be tested at the place, wherever testing facilities, methods, procedures, and authorized observers are required to meet the requirements of ASME PTC-25. KEA has certified Framatome-ANP in Germany as a KEA's designee and pressure relief device testing laboratory. The design drawings and all test results of pressure relief devices shall be submitted to KEA or any of KEA's designees for review and acceptance.
industrial codes and standards in the area of pressure vessels and piping. This chapter has also addressed the status of Korean Electric Power Industry Codes (KEPIC), which covers standards for design, manufacturing, operation, maintenance, and testing and inspection of nuclear and non-nuclear mechanical components including pressure vessels and piping.
1. Jo, J. C., Kim, H. J., Oh, K. M., and Cho, D. Y., Current Status of Inservice Testing Program Development and Implementation in Korea. The 5th NRC/ASME Symposium on Valve and Pump Testing, 1998, U.S. Nuclear Regulatory Commission: NUREG/CP-0152: 3B/253B/37. 2. Cho, C. W. et al., The 3rd National Report for the Convention on AQ7 Nuclear Safety, The Republic of Korea, 2004. 3. The Statute Book including the Atomic Energy Act, Enforcement Decree of Atomic Energy Act, Enforcement Regulation of the Act, Enforcement Regulation Concerning the Technical Standards of Reactor Facilities, and others., and the Enforcement Regulation Concerning the Technical Standards of Radiation Safety Management, and so on., Korea Institute of Nuclear Safety, 2006. 4. The Collection of Notices of the Minister of Education, Science and Technology, Korea Institute of Nuclear Safety, 2005. 5. Kim, J. H. and Kim, N. H., Status of Korean Nuclear Codes and Standards, Proceedings of the 18th KAIF/KNS Annual Conference, 2003. 6. Park, T. J. and Ahn, Y. T., KEPIC Development Status and Application of ISO/IEC Guide 21, Proceedings of the KSME Spring Conference, KSME 05S001: 3189-3194, 2005.
This chapter has covered the Korean nuclear regulatory organizations, aspects of the regulatory authority including licensing, status of nuclear installations, nuclear reactor regulatory framework and regulations that pertain to domestic and/or international
C.W. Cho, Y.S. Kim, S.B. Kim, H.D. Chung (Ministry of Education, Science and Technology); Y.S. Eun, Y.W. Park, C.B. Kim, K.S. Choi, D.K. Park, H.S. Chang, S.W. Kim, W.S. Kim, S.N. Choi, W.T. Kim, S.H. Yang, G.T. Kim, C.H. Hyun, J.B. Lee, B.S. Lee, D.I. Kim, S.H. Lee (Korea Institute of Nuclear Safety); Y.S. Park, H.B. Cho, K.N. Kim and B.R. Park (Korea Hydro &amp; Nuclear Power Co., Ltd.); K.H. Chang (Korea Power Engineering Company); S.K. Kim, J.D. Kim (Doosan Heavy Industry Co., Ltd.); S.C. Chang (Korea Atomic Energy Research Institute),
AQ1: Please check the change made for correctness. AQ2: The paragraph &quot;This chapter addresses a brief-------------piping has also been addressed&quot; is a repetition of the first paragraph of the chapter. Hence, it has been deleted. Please check. AQ3: The sentence &quot;A comparison between the KEPIC code and its reference to codes of foreign countries is also provided.&quot; has been added as the last line to the first paragraph of the chapter. Please check for appropriateness. AQ4: Please check whether the Table no. is missing. AQ5: Please check whether the intended meaning has been retained after edits. AQ6: Tables have been cited and renumbered. Please check for correctness. AQ7: Please provide the list of all the authors.