Ill XA9950501 International Conference on Strengthening of Nuclear Safety in Eastern Europe Vienna, June 14-18,1999
Dukovany Nuclear Power Plant Safety
by
collective of the authors
from Dukovany Nuclear Power Plant and Nuclear Research Institute
261 1. Introduction
The safety of operation in our plant - Dukovany - has an overriding priority and we are glad to present achieved results within framework of this conference. We would like to share with you our goals for future as well. We are enjoyed that this conference concerns all safety aspects and it is not focused only on close complex of concerns. On the other side it is possible in such case to address individual concerns to certain depth only. Number of information on the Dukovany NPP safety is described in the report for "Nuclear Convention" thus repeating the facts given in this report would be waste of time only. Our presentation does not cover all recommended areas in uniform way. We shall try to address such themes which have been resolved with satisfactory conclusions thus our experience could be useful for other operators in the area of nuclear industry. In addition, our presentation will be focused also on concerns, which need to be still improved and modified. We believe that results concerning the Dukovany NPP safety would be of interest for all participants and that such assessment of safety will be provided under the IAEA (Agency) guaranty also for all other reactors from different countries.
2. Safety management in Dukovany NPP
2.1 Safety The safety presents set of activities serving for protection of persons. In case of nuclear power plant it concerns the protection of employees as well as other individuals in plant surroundings. According to the nature of risk against which the particular safety provides protection the following types of safety are recognised:
• radiation safety • nuclear safety • security • emergency preparedness • health protection at work • fire protection • environmental protection • chemical safety • technical safety
The level of the safety is determined by a number of factors and it covers in complex manner all activities in nuclear power plant. The general principles for the safety management follows from the following picture:
262 Fig. n. 1
The basic components, hierarchy and mutual relationships of safety management are shown in the following Figure n. 1. It is necessary to keep safety (with regard to all components) within demarcated area. In case those costs should be exceeded the plant becomes incapable to be competitive; when legislative conditions should not be fulfilled the operating license can not be awarded. These are the bases for the text hereinafter which describes methods and mutual relationships concerning safety management in the Dukovany NPP.
263 Fig. n. 2
Dukovany
The CEZ Safety Strategy The Dukoa NPP Safety Strategy \ Safety Control Indicators to yr Rules Assessment Inspections ^s^ By Inde- Cor itrol Safety Data Events Analyses External Own External Docum entation Instructions sources Guaran- pendent tee EDU
/O 00 w inter-
FE T extern. Illllllllll Illllllllll § > AL I internal hnica l s i-i H a l prep ION- e mergenc y cologica l ratio n - chemic a VI V) * BOZP = Safety and Health Protection at Work Brief description of individual parts and mutual relationships with references to up to now issued Guidelines follows in the next Sections 2.1.1 The EEZ joint-stock company Safety Strategy The CEZ joint-stock company Safety Strategy has been issued in 1995. It concerns the general and essential statement through which the management of the entire company declares observance and enforcement of safety principles. Among others the following is described there: "The inseparable part of the safety shall be implementation of declared principles within framework of the company's life including provision of all necessary resources (human, financial, material) in such manner that the safety will be at optimum level". 2.1.2 The Dukovany NPP Safety Strategy The Dukovany NPP Safety Strategy follows closely "the CEZ joint-stock company Safety Strategy" and it is slightly modified according to the local conditions observing all general principles and requirements. 2.1.3 Safety Management The Safety Management in the Dukovany NPP is provided according to the above mentioned strategies. The Safety is controlled through indicators and several other tools: • "Rules" • "Evaluation" • "Inspections" 2.1.4 Rules The rules determine duties, procedures and rules in the area of safety. Concerning the safety management the safety principles are described in the following basic types of documents: • the 1st level standards , interfacing Guidelines and Programming Processes • Control documentation • Safety Instructions and the so called "The Ten Commandments" These documents contain requirements from the Czech and World legislature. At the same time also expectations of the Dukovany NPP management are involved there for each individual area of activities. 2.1.4.1 Control documentation It contains the general principles of Safety Management, responsibilities, relations, and processes. The Safety principles are involved in essence within all 26 areas of the control documentation The general methodical instructions for each individual area are developed for: 265 • Nuclear Safety • Radiation Safety • Emergency preparedness • Fire Protection • Ecological Safety • Security (Physical Protection) • Quality • Safety Culture • Safety and Health Protection at Work • Technical Safety The principles determined by the control documentation are detailed in operating procedures and safety instructions. 2.1.4.2 Safety Instructions The particular requirements, expectations and commands for activity of employees in the Dukovany NPP are described in Safety Instructions for each area affecting safety. The principles for development are given in Guideline (prepared for issue). The documents are developed for the following areas: Radiation Safety, Nuclear Safety, Fire Protection, Safety and Health Protection at Work, Ecology, Waste, Safety Culture, Quality and Company's Culture. Employees are obliged to follow these instructions. Brief summary of the general requirements is given in the so called "The Ten Commandments". 2.1.5 Assessment The assessment allows determining what are the correct "Rules" for the area of safety. For this assessment the general feedback loop is valid. - See Fig. n.3. Fig. n. 3 Deviation Remedy Data determination from Data arrangement measures Assessed area j- acquisition the required i (indicators) determination . J condition Determination of Remedy 1 priority for remedy Plan of Actions measures I—| measures implementation | implementation 266 For the area of nuclear power plant safety the following sources of information and data are important: DUKOVANY NPP SURROUNDINGS External Data I SAFETY EVENTS lANALYSES' sources tuu Indicator evaluation, deviations determination, remedy measures determination Priorities Plan of Actions Implementation The loops shown in the Fig. n. 3 mean evaluation of individual data, events, their mutual comparison and assessment according to the general process in the Fig. n. 3. 2.1.5.1 Data The group "Data" is the greatest group of data about the Dukovany NPP Safety. Because of their amount the data are evaluated mainly using statistics by groups and based on their trends. The evaluation is subsequently performed at the work level in particular responsible departments. Also responsible departments control completion of the loop including implementation of remedy measures. In case that the margins of competence of appropriate department are exceeded as for the scope of remedy measures and their significance, the appropriate manager shall solve the concern at: • Failure Commission (PK) • Technical Board (TR) • Technical - safety group (TBS) • Management meeting (PV) For individual group of data, their monitoring, evaluation and remedy measures implementation the Programming Processes are developed (according to the responsibility of appropriate department). It is obvious that any large modification and complementation will 267 depend on the ISE (the Information System of Plant) and its implementation. The groups of data are as follows: • Results and frequency of In-Service Inspections (In-Service Inspection Programme) • Frequency and results of equipment checks • Small failures of individual components • Frequency and duration of operative maintenance • Frequency of the same type of operative maintenance • Frequency and duration of preventive maintenance • Frequency of the same type of preventive maintenance • Training of operators results on simulator (detail) • Mistakes in procedures • Deviation of standards in the area of radiation safety • Shortcomings in investment activities • Outage schedule fulfilment • Unavailability of safety systems • Tightness of three barriers • Personnel dose equivalent • Effluent into environment • Shortcomings in waste management • Injuries (when there are none events) • Shortcomings in Fire Protection • Shortcomings in fuel loading proposal, refuelling • Deviations from chemical mode Other possible data. 2.1.5.2 Event The processing of internal events is described in the Plant Guideline No. 09/101. The internal events are divided in: • near miss • beyond scale — are investigated within working routine • under the scale (INES "0") - are investigated together with results and remedy measures at Failure Commission meeting, in the case of more significant events the "investigative group" is called. • INES 1 - the investigative group is called every time; for selected cases "the Extraordinary Failure Commission" shall be called. • abnormal events - the Shift Emergency Staff for operative resolution of situation Evaluation of repeated events; group of events, group of root causes is performed yearly. Such evaluation is performed as follows: • in working routine for events classified by INES < 0 • within the Failure Commission - any event classified by INES "0" and higher • within the Technical Safety Group and/or management meeting - for events classified by INES 1 and higher and for annual evaluation • by the Emergency Staff and at management meeting — abnormal events 268 Also priorities as well as the "Action Plans" for implementation are accepted. 2.1.5.3 Safety Analyses The results of analyses could affect remedy measures for safety enhancement in significant way (see for example "Dukovany NPP Modernisation Programme"). As the source of information about problems the results from the following can be utilised: • The Dukovany NPP Operational Safety Review • Safety Analyses • Lifetime management • Standards • Special requirements of regulatory bodies • Analyses from beyond design basis accident The results of these analyses are evaluated and processed within appropriate responsible departments. In case that the need exceeds capability of an appropriate department, the results are submitted to the Technical Board and when appropriate to the Technical Safety Group together with approvals for further progress concerning priority determination. 2.1.5.4 External Sources Rules for transfer of lessons learned from external sources are processed in the Guideline "External Sources". • information from the CEZ company plants • information from the Bohunice and Mochovce NPPs • information from the Paks NPP • information from VVER 440 NPPs • information from manufacturers (Skoda, Sigma, Vitkovice, ..) including equipment supplied out of the area of NPP • information from WANO, IAEA, Internet, Saint Albain, Obrigheim .... Information from external sources shall be reasonably filtered and processed. The Failure Commission performs determination of correct remedy measures. WANO summary reports (other when appropriate) are discussed at the Technical safety Group meetings. 2.1.5.5 Complete Evaluation The complete evaluation with all attributes is summarised in annual report of the Nuclear Safety Section. This report is negotiated by the Technical Safety Group and by the Management meeting. Conclusions accepted here are the main tasks for the next year concerning area of safety. Evaluation is performed based on indicators characterising the following areas: 1) Impact on environment 2) Safety barriers tightness 3) Regulatory bodies activity 4) Perspective of further operation 5) Reliable, safe and for response ready operation 269 2.1.6 Inspections (Checks) It shall be proven by the inspections that "Rules" are applied in correct way. These inspections are described in Guideline. They have several forms: Internal • performed by direct supervisors within area of their competence • performed by methodical guarantee for appropriate area • independent internal check External • national - state regulatory bodies audits • international - IAEA WANO other 3. CEZ, a.s. strategy of Quality Assurance General meeting of the CEZ, a.s., shareholders has approved concept of the business activity which is derived from the analysis of external and internal business environment, and establishes both the subject of business and strategic initiatives of the company which must be fulfilled to maintain the CEZ, a.s., mission and to achieve the business targets. The Company's Quality Assurance Concept, one of the tools instrumental in achieving the business targets and maintaining the company mission, has been published as a decision of the Board of Directors on December 21, 1995, it enumerates the company quality assurance related activities as follows: • to create conditions for satisfaction of customers, shareholders and employees, • to document the quality assurance system and to use it as a system instrument within the company, • to create conditions for quality system of all processes and their unceasing improvement, • to assure mutual and effective co-operation of all organisational units of the company, • to achieve a high quality standard by increasing competence and motivation of all employees in the environment of the corresponding culture, • to assure co-ordinated responsible approach to the environment, • to assure controlled selection and evaluation of suppliers. In 1997, the quality concept was extended to include an obligation of the Board of Directors to improve systematically the CEZ, a.s., profile as a company which consistently fulfils its obligations toward the environment, managers on all levels are responsible for realisation of quality concept. All employees fulfil their duties in accordance with the quality system with permanent effort to improve all processes by implementing quality system. 270 One of the most significant tools for achieving the company's goal is application of quality system. In 1995, the Board of Directors, besides the quality concept, has decided on the time schedule of its implementation. The CEZ Quality System is a comprehensive set of principles and requirements representing quality related philosophy of the company. Quality system comprises methods and tools for the quality control, it is developed in accordance with the Czech Republic regulations and takes into account the requirements included into ISO 9000 and 14000 standards, as well as recommendation of the IAEA. The implemented quality system permits to fulfil requirements of both the internal and external environment. Created quality system has all features of a complex quality controlling tool and is aimed at three basic spheres: • customers - to fulfil their requirements and expectations, • internal need of the company - to provide efficient management instruments, including those for evaluation of potential risks and benefits of the business, and mitigation of the environmental impacts, etc., • suppliers - to build a base of the competent suppliers. For this purpose the company provides the appropriate human, material, financial and information resources. 3.1 Quality Assurance programmes at all stages of nuclear installations service life In July 1995, the CEZ, a.s., set out the Safety Strategy for nuclear activities. This strategy incorporates safety obligation of the CEZ management for the area of utilisation of nuclear energy, and defines company's safety goals. All company documents are conform to its strategy for the area in question. The Top Quality Assurance Programme (TQAP) is the main document of the company quality system with universal validity. This document, approved by the company management in January 1996, has been revised with respect to requirements of the internal and external environments, changes in the CEZ, a.s., system of management and development of the quality system itself. The objective of the TQAP is to establish general principles for management of individual processes and for development of hierarchically arranged controlling documents, divided in the Quality Assurance Programmes, Rules, control and quality assurance procedures and follow-up documentation. Quality Requirements established in the TQAP are binding, and their application is in compliance with the importance of a process, item or service. Aspects of nuclear safety, radiation protection, emergency preparedness and physical protection are described in the document "Safety Rules for the Area of CEZ, a.s., nuclear activities". These rules are binding for all personnel employed in nuclear area, they define authorities and responsibilities with respect to the interface between the company Headquarters and its individual organisational units. As to the Atomic Act requirements, the CEZ, a.s., has an approved Quality Assurance Programme for the Area of Nuclear Activities. Both nuclear power plants, Dukovany and 271 Temelin have developed their own quality assurance programmes for individual stages of each plant's service life. 3.2 Application and evaluation of quality assurance programmes efficiency The CEZ, a.s., has established responsibilities for the quality control and verification for all processes, and on all levels. Responsibilities with respect to equipment quality and processes verification are described in the relevant documents which are an integral part of the documented quality system. Quality Assurance Section is responsible for development and co- ordination of the quality system implementation within the company as a whole, as well as for evaluation of the quality system's effectiveness. Responsibilities for the actual implementation bear all company managers. Each employee is responsible for quality of his/her "nuclear" work. Individuals who perform control and verifications have adequate authority to identify discrepancies and, if necessary, to require that remedial measures are taken. Required quality is verified by the employees who do not perform control and verification activities. All company employees are encouraged to submit proposals on upgrading and modifications of the quality system. Regular quality oriented training of company employees is understood as an investment into quality. Employees on levels of management are trained in accordance with the unified training programme. Training programme for the management and other employees, focused on quality, is based on the CEZ, a.s. quality concept. Objective of the training programme is to achieve that all company employees will become engaged in the quality assurance and enhancement process, and thus participate in the development, implementation and improvement of quality system. At the end of each calendar year, effectiveness of the quality system is assessed and the system is subsequently updated. Managers on the all levels of management periodically evaluate all processes and procedures (subjects of their authority) with the objective to review their actual condition and efficiency. Quality systems of nuclear power plants Dukovany and Temelin are evaluated quarterly. External audits of the quality systems at the suppliers as well as internal audits, performed by qualified auditors in accordance with the written procedures, are an important part the company control system. Company managers apply results of quality audits to take necessary corrective, preventive and remedial measures. Final evaluation of the quality system effectiveness performs the Quality Assurance Section - for the company Executive Board, and quality department of the organisational unit - for the Directors of those units, always by the end of a year. Comparison with trends in the processes and procedures behaviour, is a part of the regular evaluation of the quality system effectiveness. Evaluation criteria are always measurable parameters which provide information on the process conditions and their output from the aspects of safety, reliability, efficiency and environmental impacts. The evaluation criteria of the quality system are measurable information on the units of quality system within a given system or procedure. Results of the quality system evaluation within the organisational units and verification of processes are described in the quarterly reports. 272 4. The Dukovany NPP staff training provision including training on simulator There are three educational centres within framework of the CEZ joint-stock company providing professional training of employees within area of nuclear activities of the company. It concerns the Education and Training Centre in Brno on the level of the CEZ - HQ (Headquarters), the education centre as the part of the Dukovany NPP and the education department as the part of the Temelin NPP. The Education and Training Centre Brno fulfils the role of the methodical and professional guarantee of professional training within area of nuclear activities determining conception, system and objectives of the Dukovany NPP personnel professional training and provides the basic training of employees performing activities important from the point of view of nuclear safety and radiation protection (including selected workers). Identification of the real training needs as well as requirements for professional qualification of employee is provided by the education Centre of the Dukovany NPP; this centre co- operates with training guarantee in case of submission of application for license issued by the state regulatory bodies in compliance with the Atomic Law (including co-operation in submission of prescribed documentation) and it fulfils the statute of the training power plant in area of on-job training implementation and implementation of selected parts of the theoretical training ( staff training days, training of employees performing activities unspecified from the point of view of nuclear safety or from the point of view of radiation protection). Both the Dukovany NPP personnel training and the training of supplier staff fall within competence of this centre. Currently the full-scope simulator located in the Education and Training Centre of VUJE (Research Institute for Nuclear Safety) in the Slovak Republic is utilised in order to provide training of the operational managerial personnel (shift supervisor, safety supervisor, unit shift supervisor, primary circuit operators and secondary circuit operators). The training on the full- scope simulator is performed in form of the general training (the part of general training before acquirement of the ,,Authorisation for job performance" from the SONS), in form of the periodical training (in case of employees performing operational managerial function) and in form of repeated qualification training (is the part of training in case of assignment to other operational managerial function). The duty to participate in this training is given by the applied legislature of the Czech Republic (the SONS Decree No. 146/1997 Law Digest) The contents and scope of the training is given by the training programmes developed in the Education and Training Centre VUJE and approved by experts of the Dukovany NPP in such manner that they shall take into account requirements for final level of the professional qualification of trained employees as well as equipment capability to provide simulation. These programmes are regularly updated and completed according to requirements and needs of the Dukovany NPP the same way as in the case of the training scenarios for simulated tasks in the area of nominal, failure and emergency conditions which are updated by the feedback from operational events with influence of the human factor. In September 1998 the significant improvement in the area of training on simulator occurred since the multifunctional simulator has been delivered to the Dukovany NPP developed in co- operation with the EU within framework of the PHARE Programme. This simulator was 273 supplied by the consortium comprising the Corys SA, the Siemens and the Belgatom and the work on this project has been commenced in 1995. Also experts from the Dukovany participated permanently in the development of simulator and during its tests. Currently the multifunctional simulator (MFS) is under trial operation and the Instructor Training as well as pilot training of operators under surveillance and management of foreign experts is implemented within framework of the PHARE Project instructor Training". In addition work focused on acquisition of the SONS license (the State Office for Nuclear Safety - one among regulatory bodies) are performed in order to achieve involvement of the simulator into regular staff training. From the end of 1997 the full-scope simulator is developed for the Dukovany NPP as the tool for the highest level of the on-job training of staff. This simulator will be supplied by the Czech firm ORGREZ joint-stock company with its subcontractors the GSE Systems (the USA) and the REGULA Praha. We assume that the full-scope simulator will be assigned to the user (Training Centre in the Dukovany NPP) after installation and tests in the 4th quarter of 2000 and the training of operating personnel shall be commenced following the training of service personnel and after fulfilment of all legislative requirements in 2001. The new building of the Training Centre where both simulators, training rooms as well as offices of lectors and instructors will be located is directly in the site of the Dukovany NPP and its construction will be completed in September this year. It is expected that both the full-scope as well as multifunctional simulators should provide the modern system enabling education of the unit control room staff. The staff shall be acquainted with special knowledge, skills, experience, habits and opinion concerning individual as well as team work for independent performance of its job in the NPP. Furthermore the training on the simulator is concerned for the most effective way of training of team behaviour of operators and for training in communication in the real environment of the unit control room. 5. Emergency Operating Procedures NPP Dukovany recognised that the event oriented procedures is not sufficient and it was necessary to try out something better. NPP compared Emergency Operation Procedures (EOPs) of some NPPs in Europe. After that, NPP decides for symptom oriented procedures developed by Westinghouse. Project was divided to several phases. NPP Dukovany want an Approval letter, that confirmed, that the Westinghouse procedures are applicable for eastern unit VVER. Key material was the entry analysis of behaviour WER unit during basic types of the accidents. There were issued the Comparison material that compares western PWR and VVER, and a Strategy Report, which describes, how to solve different basic accidents and how to use VVER devices. Five years co-operation between Westinghouse and Dukovany experts was based on Contract with Westinghouse. The whole scope of work was divided to two teams. The Westinghouse team explain to Dukovany team members strategy of all procedures to all details. The Dukovany team explain precisely to the Westinghouse team many technical details of WER 213 NPP Dukovany units. Each part of the procedure was tailored to VVER unit step by step. Preliminary results was checked by the second team. When all members of both teams agreed with draft report, the basic material was written. It was checked again and consult with Dukovany operators. All comments and notes was incorporated to the basic material. The important procedures was validated by the 274 set of analysis. All book of EOPs procedures was translated to Czech language. After that the final meeting was done and revision 0 material was issued. The material was validated in the full scope simulator in Trnava and multifunctional simulator in Dukovany. The training courses for all operators started three years ago. Three group of shift and unit supervisors had a basic lessons in Brussels. From this group was formed a team responsible for next education and simulator training, validation team and future safety engineers. Practically every step of the EOPs was explained to each operator during regular lessons, because the new procedures change strategy and philosophy of the EOPs. Usage of new two column procedure is totally new and therefore all crews have been trained on full scope simulator for two years. Due to different philosophy of procedure is necessary to train using of the EOPs very well. The operators had a plant examinations of the new EOPs, and from this year is a question from the new EOPs a part of the licence examinations. Now is finished a year work on background materials. There are detail explanations of accident description without operator control and comparison with the same transient of accident with operator actions in this material. After that is explaining of the main strategy steps. At the second part of each background documents detail description of all steps, cautions and notes. Independent part of the EOPs are the Critical Safety Function Restoration guidelines. Safety Engineer will check six Status Trees and also he will communicate with Technical Support Centre and Emergency Staff. The current EOPs have to be changed. All emergency status was deleted and a new part was added. This procedure is now called the abnormal operating procedure. From the last year we started a long term Maintenance Programme with Westinghouse for a possibility to incorporate all technical changes to EOPs due to Dukovany Plant, feedback from training and also to incorporate last changes from Westinghouse Owners Group. For this reason we maintain English and Czech version of EOPs. We will start to use the new symptom oriented EOP in autumn this years after the last refuelling in all four units in the same day. For this day will be prepared: • new EOPs including backgrounds materials in the control room • all crews in control room will have a licence of the new EOPs • there will be the upgrade of the abnormal operating procedure in control room 5.1 Severe Accident Management Guidelines It will be necessary to have also Procedures for solution of a Beyond Design Bases Accident to increase nuclear safety in the future. NPP Dukovany will probably continue with co- operation in developing of this type of procedure with Westinghouse. Between EOPs and SAMGs are many connections both of them are written in similar two column format and the same philosophy. Before developing severe accident management guidelines we have: • results and all analyses from PHARE project 4.2.7.a/92 Beyond Design Basis Accident Analysis And Accident Management, finished last year • results PHARE 2.06 and 2.07/94 Filtered Venting and Hydrogen Control, finished this year • Co-operation with Research Institute REZ - analyses, presentations, lessons for safety engineers, Technical Support Centrum members and emergency staff members • Preparation in advance some modifications that are supposed to help us in BDBA situation 275 5.2 Modifications Many modifications were done as a recommendations from WESE or for better performance of EOPs and also were supported by PSA studies: • Installation Pressuriser Power Operated Relief Valve - it is third PZRZ safety valves. First two valves was the Sempell design, new one is different design and principle by Sultzer. By analysis PORV is designed for steam flow rate approximately of 56 mVhour on 13,8 MPa, the valve is qualified for steam, water, mixture steam-water and gases. • Hydrogen control - 16 passive recombiners, they are placed in the containment on the top of SG, reactor, hydroaccumalators, PRZR and other devices in containment and in the RCP's room. After the end of PHARE project 2.07/94 some additional recombiners may be added. Each recombiner has a concentrationsensor and thermometer, with an indication and a recorder in the control room. By indication of increasing temperatures in the containment the operator can estimate the place of break during LOCA accident. • Pressure in containment - new measurement with wide range from underpressure cca 50 kPa to overpressure 450 kPa, this span covers accidents beyond design. • Level in containment - to measure flooding during LOCA situation. With using of this measurement we can indicate interfacing LOCA, and prevent to put the overfiooded water to reactor cavity in the first hours of emergency situation. Range is 0 to 2 metres. Both measurements (pressure and level) have two trains based on different principle. • Pressure in hydroaccumulators - we decrease pressure in HA. Research Institute 0ED investigates behavior of different pressure in HA in many significant accident. Optimal pressure was different for two HA injecting bellow core and for two HA injecting above core. This solution with different pressures was not optimal for operations. So, we use the second optimal solution and we decrease pressure in HA from 6 MPa to 3,5 MPa. Reason was not only, that maximum pressure in HA was higher then opening setpoint of SGSV, but also negotiation during LOCA accident - main reason was decreasing maximum fuel cladding temperature. • Auxiliary feed water pump - increase shutoff head pressure to 6,5 MPa, the reason is a possibility to cool down primary circuit from secondary side if SG pressure is about SG valve setpoint. • Core exit temperature - range from 0°C to 400°C was increased to 1000°C - for better temperature checking during accident with high temperature of fuel cladding • Containment sump - protection before clogging by insulation during recirculation phase from the containment sump. We start to change shape of grid, the surface will be increased. 5.3 Modifications planning in the near future • We are planning a large reconstruction after year 2000 in NPP, including changes in I and C, especially reactor protections, ESFAS, ELS • Emergency venting from reactor vessel head and SG collectors - possibility of venting in high pressure mode, especially for venting gas voids • SPDS panel • Emergency feed water collector - crash protection of piping on secondary side before penetration to the confinement and replace piping from +14m to higher floor • Steam dump to atmosphere - add new two steam dumps placed before quick operated valve on steamline from SG 2 and 5 on reactor hall. 276 • Technical Support Centre - will be placed in shelter (bunker) close to the room with emergency staff. TSC will advise strategic decisions during accident and control mitigation during severe accident. 5.4 Benefits from the new symptom oriented EOPs: • Westinghouse type of procedure are used by almost a half of PWR in the whole world • This procedures cover large spectrum of emergency situations, combination of accidents, even low probability accident • Solutions have priorities • Independent checking of Critical Safety Functions • Several levels of diagnoses correct errors of operators (human factor) • Several levels of diagnoses allow to correct solution, when condition are changed because of progress of accident 6. Emergency preparedness in the Dukovany NPP The director for technique is responsible for the area of emergency preparedness in the Dukovany NPP. Administratively this area is covered by the section of emergency preparedness from the Department for Nuclear Safety and Fire Protection. The duty to provide emergency preparedness of nuclear facilities follows from the Law No. 18/97 Law Digest (Atomic Law) applied in July the 1st, 1997. This duty is more detailed through the set of decrees, in particular the Decree No. 219/97 Law Digest ,,..on details for emergency preparedness assurance.." Based on this legislature, on the recommendations of international Agencies and on experience gained from operation the company's internal legislature has been reviewed and after approval of the Dukovany NPP director for technique as well as approval of the State Office for Nuclear Safety this legislature concerning emergency preparedness was issued under the title ,Jnternal Emergency Plan of the Dukovany NPP". This legislature is applied since July the 1st 1998 after an extraordinary training of all staff (employees and other individuals) performing work activities within the Dukovany NPP site. Currently based on this control documentation the emergency instructions are developed what are ,,step by step" described subprocesses for members involved within the internal emergency arrangement in case of each particular type and grade of abnormal events regarding above mentioned documentation. Since April 1996 the Shift Emergency Staff has been established in the Dukovany NPP because of fast availability assurance and capability to manage internal emergency arrangement in case of occurrence, duration and mitigation of consequences of abnormal events . In course of abnormal events and during their consequences mitigation the Shift Emergency Staff is the main controlling body bearing, after its initiation, responsibility for situation solution. Within framework of technology this Shift Emergency Staff provides recommendations for shift supervisor as well as for unit control room staff based on the available data, on its experience and on SW equipment enabling to identify prognosis of an abnormal event development. This Shift Emergency Staff is on the continuous alert; in case of initiation and a for long-term abnormal event it works in cycle with 4 shifts. Since implementation of the Dukovany NPP internal emergency plan the number of 24 exercises of the Shift Emergency Staff has been carried out. This way the activities of individual members of the emergency 277 arrangement given in the control documentation were verified and based on these activities the above mentioned ,,on activity" focused processes (so called emergency instructions) are currently developed. Training of employees and other individuals on described theme is carried out with regard to the state as well as company's internal legislature. The period of this training and examination is determined to be 1 year. Respective contents of training and form of examination depend on involvement of employees and other persons within categories regarding to the emergency preparedness (the employees and other persons are classified into 5 categories by the Internal Emergency Plan). Demonstrable acquaintance with the themes of emergency preparedness, as one among other conditions for fulfilment of duty towards to the State Office for Nuclear Safety, is provided either by tests or by oral examination when appropriate. In consequence with training of the internal emergency arrangement members respectively of all workers within the Dukovany NPP site the regular exercises are carried out aimed on verification of theoretical knowledge , skills and overall preparedness to cope with real abnormal events. Furthermore the evaluation of such training and exercises perform also the service for enhancement of the plant emergency preparedness level. In addition to exercising of human potential also regular tests of technical tools necessary for emergency preparedness assurance are carried out, in particular communication tools, emergency notification and alerting system in region of the emergency planning. In autumn this year the complex exercises focused on sheltering and evacuation of persons from the Dukovany NPP site should be implemente 7. ,,Year 2000,, problem 7.1 Dukovany NPP situation The Dukovany NPP is one part of the EEZ joint-stock company thus the ,,YEAR 2000,, problem solution will be resolved within framework of working team established at the EEZ joint-stock company level. The team is divided to the central part and to the on-site parts that are in each appropriate power plant. The central part of the working team shall • define methodical procedures • co-ordinate work of the on-site parts within framework of whole company • provide financial resources for project The ,,on-site,, working team is responsible for technical as well as administrative resolution in its own plant. The central database of the team ,,YEAR 2000,, generally accessible within information system is utilised in order to • share information • store common documents • record concerning endangered equipment This work was commenced in the Dukovany in June 1997 in advance of the central team establishment. 278 7.2 Work progress The Dukovany NPP follows the Guidance for Achieving Year 2000 Readiness issued by the IAEA in January 1999. Furthermore the best experience gained from foreign countries is utilised, for example Health and Executive Health and Safety Executive and/or Institution of Electrical Engineers recommendations. 7.2.1 Main tasks • to search and to identify all equipment using time in form of variable in their operation • to identify possible risks and to define kind of these risks resolution • to apply individual solution including its trial operation • to implement defined measures 7.2.2 Team work milestones 1. To identify all endangered objects as well as all kind of T: 28.02.1998 danger 2. To develop and to implement particular solution including T: 30.06.1999 tests of equipment and processes endangered by the time shift to the year 2000 3. To develop emergency plan for the time period of the turn of T: 30.09.1999 years 1999 to 2000 including staff training 7.2.3 Project status as on the 25th March 1999 Resolved by the central team 5% Resolved 50% Solution under way 43% Not commenced yet 2% 7.2.4 Other activities of team Team ,,Year 2000,, activities leading to problem elimination are carried out in co-ordination with other activities ongoing currently in the plant (modification, new Information system). The results should be provision of warranty for turn over the year 2000 in case of any equipment. In addition within framework of co-ordinated work the list of endangered equipment was checked by appropriate administrators in order to be supplemented with other endangered equipment. To the relevant tasks of team belongs among other performance of work with real costs. For this reason some negotiations were held with suppliers resulting in total costs reduction. An essential task for the Dukovany NPP operation is the elimination of external risks and solution of external unpreparedness. Approach of the Dukovany NPP has been assessed also in conclusions of the State Office for Nuclear Safety inspection held in the Dukovany in February 1999. Furthermore it was stated that the generally serviceable system of detailed 279 monitoring for resolution of the ,,Year 2000,, problems has been developed by the CEZ joint- stock company within its area of industry. The company was ready in time and as for the professional aspect this problem is resolved in an appropriate manner. 7.3 Experience To our best-learned lessons belongs up to now an early test and equipment verification. This way we succeed in identification of faults where they were expected. Furthermore relatively accurate estimation of costs was reached. During subsequent negotiation and consultation with suppliers it was reached their better understanding as for individual problems resolution. We concern for relevant the list of endangered equipment to be released within plant information system. Such information is then accessible for all plant employees. The process of work was continuously published in plant internal news (ATOMIX). The purpose of this was to engage not only working team but also other employees. We concern for beneficiary also an early commencement of plan development for period of turn between 1999 and 2000 and development of our own transition to the year 2000 (see next paragraph). 7.4 Emergency schedule for transition to the year 2000 (Emergency measures) The schedule for the transition period is now in the stage of its development and it will be detailed concerning risk elimination and safe plant operation. Currently the process of emergency schedule negotiation is under way. Within framework of work for year 1999 also the following steps are involved concerning plan of financial resources for unpredictable events, preparation of spare parts and equipment procurement, development of measures as well as recommendations in order to provide operation of relevant equipment during transition to the year 2000, staff preparation and training. 8. Inspections and lifetime management In all Dukovany NPP inspection activities the following important catchword is applied : „ The inspection at the right place and at the right time = improvement of the safety". 8.1 Inspection fields existing at Dukovany NPP : • non-destructive testing : • visual, penetrant, magnetic, X-ray, ultrasonic, detection of leakages, dimension inspections • diagnostics: • thermovision of electrical equipment • vibration of rotating machines • monitoring of vibrations of primary circuit components • monitoring of lost parts in primary circuit • revisions of equipment: • supervision of ,,classified items" (equipment important from the point of view nuclear and technical safety - pressure, electrical, gaseous and lifting equipment) • performing of pressure tests of primary and secondary circuits equipment 280 • metrology of measuring: • supervision of metrological reassigning of meters and measurements • practical metrology in accredited metrology laboratories in the fields of temperature, pressure, electrical and physical-chemical quantities and ionising radiation • chemistry: • chemical and radiochemical control of technological circuits • chemical control of chemicals to be used • chemical control of waste waters • control and management of chemical modes of technological circuits • monitoring and evaluation of corrosion stage of technological equipment • lifetime management: • monitoring of degradation effects of main components of primary circuit and monitoring of erosion corrosion damage of secondary circuit components 8.2 Stage of inspections at Dukovany NPP 1/ non-destructive testing • using contractors • mechanised inspections of reactor pressure vessels from inside and outside - SKODA Plzen ( ultrasonic and eddy current) • mechanised inspections of distinguished welds of main primary piping ( 500 mm) - connecting welds to the steam generators, to the main circulation pumps and main closing valves - SKODA Plzen (ultrasonic) • mechanised inspections of heat transfer tubes of steam generators - Vitkovice, VUJE, Eddy test (eddy current) • by Dukovany NPP personnel • routine manual inspections 2/ diagnostics • by Dukovany NPP personnel 3/ revisions • by Dukovany NPP personnel 4/ metrology • using contractors ( specialised firms) - instruments and calibration references and standards their calibration at Dukovany NPP would be non economic • by Dukovany NPP personnel 5/ chemistry • by Dukovany NPP personnel • using contractors in case there are no needful instruments 6/ lifetime management • data acquisition by Dukovany NPP personnel 281 • evaluation by external contractors or by using of soft wares developed by research institutes 8.3 Scope of inspections activities Non-destructive testing • primary circuit and another 60 types of equipment (pipelines, pressure vessels, heat exchangers, pumps, parts of hermetic systems) Diagnostics • parts of primary circuit and selected rotating machines (pumps etc.) Lifetime management • main primary circuit components and selected secondary circuit pipelines 8.4 Range of inspections Non-destructive testing Focused on detection of inside and outside defects in the welds even in the basic material of the main components of primary circuit and another equipment. The inspection scope was appointed by producers of equipment in so called individual quality assurance programs" approved by authority (SUJB). The program is continuously up-dated according the our own experience or according the experience of other NPPs (for example the events posted in the WANO net). The inspections are performed on reactor pressure vessels, main circulation pipelines, main closing valves, main circulation pumps, steam generators and pressuriser. Furthermore the inspections are performed on another 60 types of components of primary circuit and main components of secondary circuit. The scope, the methods and the frequency of inspections depend on the importance of component from the point of view of integrity preservation or the functionality of the component. The inspections of hermetic compartment is the part of inspection program. Very important part of the program are the dimension controls during tightening of flanges and during settlement of assembled parts. Quality assurance of supplied new equipment and spare parts requires the performing of input inspection (including the inspection of equipment itself and inspection of necessary documentation). Recently the inspection activities are still more focused on inspections directly in manufacturing companies. This fact has a positive influence on enhancement of quality during the manufacturing period of equipment. Nowadays the inspection qualification process according the ENIQ and IAEA methodologies was started at Dukovany NPP. Diagnostics of rotating machines This inspection is still more and more important with the connection with intended predictive maintenance of rotating machines at Dukovany NPP. At present the stage of data collection and evaluation of correlation between the measurement results and real wearing of important parts was established. After good managing of this problem the decreasing of the scope of preventive dismantling and their replacement by predictive action according the diagnostics 282 inspection results is expecting. This way the scope of maintenance activities should be depressed and the hazard of equipment damage due to frequent dismantling of equipment should be decreased. Thermovision Thermovision diagnostics is broadly used at detection of defective connections in electrical distributors during operation. Vibration monitoring Vibration monitoring system of primary circuit components and system for detection of lost parts serve as the safety systems for detection of anomaly situation in primary circuit during operation. Chemistry Control of chemical and radiochemical parameters of technological systems with following evaluation. The aim of these activities is to set up and keep the optimal chemistry mode and minimise the corrosion attack on the material of the equipment. Methods of measurement: • discontinuous - sampling and following sample treatment and analysis in laboratory • continuous - using continuous measurement and transfer of data to the computer net Metrology The basic purpose of metrology is to ensure of metrological reassuming which is prescribed by governmental laws. Next the supervision over the adhering of prescribed legislative in the fields, which are not directly provided by NPP Dukovany metrology laboratories. In the special case we provide the more accurate and comparative measurements according the requirements of other NPP Dukovany departments. Lifetime management The critical places of main component of primary circuit are identified. The main components from this point of view the inconvertible or hardly exchangeable components (reactor pressure vessel, steam generator, pressuriser, main circulation pump, main circulation pipeline and the pipeline of the first safety rank in which we follow the degradation effects caused by operating modes of equipment. The databases of characteristic values of used materials are elaborated. The calculations of remaining lifetime of above mentioned components are conducted. The erosion corrosion problems focused on carbon steel pipelines of secondary circuit are monitored by using of soft ware CHECKWORKS. This soft ware is able to predict the necessary maintenance and inspections activities. 283 9. Dukovany NPP PSA Projects 9.1 Brief Contents 1. Information on status and results of Living PSA-1 2. Information on status of Shutdown-PSA 3. Information on status and results of PSA-2 4. Information on status of risk monitoring and further development 9.2 Information on status and results of Living PSA-1 The ,,Living PSA" is the form of PSA which is maintained in actual status . In order the status of PSA will be consistent with the factual status on operated NPP all changes are modelled into the PSA. To the modelled changes belong performed modifications of equipment, changes in operational procedures, maintenance changes and changes of equipment tests, changes resulting from new analyses or changes based on new experience with equipment etc.: In 1997 and 1998 were modelled these following most significant changes: 1. Actualisation of equipment data based on evaluation of reliability data for the period from 1992 to 1996 (from ,,Technical Specifications" database, SIS, maintenance database etc.) 2. Modifications of events interfacing LOCA (release from PC into intermediate RCP cooling circuit TF 10, release into intermediate cooling circuit for reactor control rods drives TF30) based on new analyses. The analyses showed that the course of such accidents is more favourable in comparison to similar LOCAs with direct release into hermetically sealed area. 3. Involvement of results of qualification tests of seal assembly of MCP in the case of loss of cooling. Large scale tests of MCP for Mochovce NPP on test facility in Russia showed capability to provide sealing for 24 hours after loss of cooling ; tests of sealing assembly performed in NRI 0eD showed resistance up to the temperature of 240°C. 4. Involvement of Operational Guide No. 36/96 - Measure in the case of break of circulating water pipeline in turbine hall and turbine hall flooding. 5. Involvement of performed design changes of equipment: • implementation of automated back switch from the sump suction in SG compartment to the suction from ECCS LP tank • assembly of relief valve on pressuriser • reconstruction of protection of MSH rupture In addition other performed modifications were involved into model (replacement of sectional 0,4 kV distributors, partial changes of operational procedures, etc.) but their influence on the PSA result is insignificant, this is why they are not referenced. The sum of all modelled changes resulted into the substantial reduction of core damage frequency (roughly to one half). The most impact resulted from the change under item 3, significant impact was achieved through the changes under 2 and 5 also. The new core damage frequency is CDF = 9,93 E-5/reactor -year 284 The value is in fact smaller than the limit l,0E-4 recommended by the IAEA for operated NPPs. Overview of initiating events whose contributions to the CDF belong to the most one is presented in the following table: No. Initiating IE title CDF (I/year) Contribution event to the CDF (%) 1 Tl Break of MSH (+ flooding on the PoE+ 14,7m) 4,99 E-05 50,3 2 F Fires 1,67 E-05 16,8 3 T4 Break of Main Feedwater Header 9,20 E-06 9,3 4 T5 Loss of external power supply 5,33 E-06 5,4 5 L0 Compensable LOCA 0-10 mm 4,33 E-06 4,4 6 LI(TFIO) Release into intermediate circuit TF10 4,14 E-06 4,2 7 L3 LOCA 60-100 mm 1,64 E-06 1,7 8 T9 Loss of circulating water pumps 1,35 E-06 1,4 9 SGCR Rupture of SG primary header 1,31 E-06 1,3 10 T12 Rupture of cooling water pipeline in turbine 1,27 E-06 1,3 hall 11 Til Loss of SG feedwater (loss of electrically 1,05 E-06 1,1 driven feedwater pumps) The total contribution of all remaining initiating events is smaller than 5% . In the last year also new emergency -operational procedure (EOP) was modelled in PSA 1. It was verified whether all sequences with probability greater than E-8 are included in the EOP (determined by the Westinghouse methodology) and such way the PSA was used also for independent verification of the EOP. After the new EOP will put into force (assumed in half of next year) the new core damage frequency will be CDF =7,93E-5 /reactor - year Implementation of several modifications will be substantial for further reduction of the Core Damage Frequency close to the value 1,0 E-05 : • displacement of Ultimate Emergency Feedwater pipelines • assembly of qualified electrical drives and I&C at the level + 14,7 m • assembly of steam pipeline and feedwater pipeline whip restraints at the level + 14,7 m • Fire extinguishing equipment assembly in the room of electric drives of MCPs - A301 • implementation of drainage train from A3 01 to A201 which will be opened in the case of release into A 301 • modification of automatics on the discharge train of ultimate emergency feedwater pump All these proposed modifications concern protection of equipment placed in PoE + 14,7 m and on the MCPs deck. According to the conclusions of PSA -1 these rooms are the rooms with the highest risk. Achievement of the value 1,0 E-05 belong to the objective of the most of operated NPPs and this reduction could be expected also from the IAEA (the limit 1,0 E-05 is today recommended value for newly constructed NPPs). 285 The mission IPERS was held from the 2nd to the 13th November 1998; by this mission the check of PSA-1 concerning in particular applied approach was carried out. The PSA was submitted from our site in the status corresponding to the end of 1997. The final report from this IPERS mission will be sent to Dukovany NPP in February of 1999. The probability of reactor core damage Development of results of PSA-1 model for Dukovany NPP Year Core Damage Frequency ( I/year) 1995 1,77 E-04 1996 1,84 E-04 1997 1,09 E-04 1998 9,93 E-05 -2010 1 to 2 E-05 9.3 Information concerning Shutdown PSA -1 Shutdown PSA (SPSA) includes conditions during shutdown of unit as well as conditions with low power during start of unit operation and its shutdown. As the point of break between the operation with power and the SPSA was chosen the point with value of 55% of the nominal power. The PSA Project was commenced in 1996 and originally its completion was planned in the last year. Regarding the delay in the PHARE Project 2.09/95 ,,Low power and Shutdown PSA for Bohunice V2 NPP „ also our Project SPSA was delayed. To our objectives belongs utilisation of all knowledge and analyses from this Project for our SPSA and also the results could not be too different between Bohunice and Dukovany NPPs. Preliminary results are till enough conservative and they present CDF comparable with operation at full power. The results of necessary thermohydraulic analyses are still not available and the detailed modelling of human factor is not performed. In majority of these cases we expect also comparison with the Bohunice NPP. To the main contributors to the CDF belong the following: • LOCA during pressure and leak tests. • LOCA during test of relief valve of pressuriser • loss of natural circulation • regular maintenance of electric equipment for emergency power supply of the 1st and the 2nd category of reliability of appropriate system combined with random failures on remaining systems. • loss of cooling of spent fuel storage pool 9.4 Information concerning PSA - 2 The PSA-2 provides probability of radioactivity release from hermetically sealed area and data concerning the size of such release. Different categories of releases according to the size and timing of release are determined in the PSA-2 Project and for each individual category the size and the probability of release is computed. Large early release frequency (LERF) is the most interesting one from the point of view of safety evaluation. 286 The PSA-2 study was elaborated by the American firm SAIC (Science Application International Corporation) with co-operation of NRI 0ED in the form of grant of the US government for the SONS. It concerns the PSA-2 study with limited scope - it is in the same scope as in the case of review of US NPPs (the scope in compliance with IPE - Internal Plant Evaluation, issued by the US NRC). The limited scope relies in the fact that detailed sensitivity analyses are not performed and only limited number of group of radionuclides is concerned. As the base for PSA-2 study the revision of PSA-1 carried out also by the SAIC in 1994 was used. Thermohydraulic computations were performed by programme MELCOR. Final report on PSA-2 was submitted in May 1998. Totally, eleven categories of releases were concerned in the PSA-2 study for Dukovany NPP. Containment behaviour in the course of accident with reactor core damage was analysed. Positive features of the Dukovany NPP containment: • due to the evacuated bubbler condenser system the small overpressure is maintained (small pressure drop for release of radioactivity into environment) • large volume of containment in comparison with the size of reactor core (low pressure in containment can be maintained; small concentration of hydrogen) Determination of containment damage modes belongs to the results of these analyses. • an early large failure (rupture, large break) -12,5 % of cases • an early small release - 8,9 % • a late large failure (rupture, large break) - 0,9 % • a late small release -14,6 % • no failure (an operational leakage is concerned) - 63,0 % The probability of an early large failure of containment is similar as in the case of BWR , but greater than the majority of PWR in the US. Computed frequency for an early large release of radioactivity : LERF = 6,8E-6 Rough estimate taking into account successful qualification of MCP seals as well as implementation of ECCS automatics is LERF = 3,3E-6 Proposal of measure for radioactivity release mitigation in the case of reactor core damage creates an integral part of the PSA-2 study. These possible measures were presented during presentation of conclusions of the PSA-2 study. • depressurisation of primary circuit in the time period between beginning of reactor core damage and damage of RPV bottom. • protection of door on the bottom of the reactor pit against thermal and pressure impacts. • control of hydrogen concentration in hermetically sealed area The fact that the PSA-2 is based on the PSA-1 carried out by SAIC in 1994 could be concerned for shortcoming. Since 1994 number of modification were implemented and also new analyses were performed and they are involved into Living PSA for year 1997. Further prepared measures for nuclear safety enhancement as well as application of new emergency 287 procedure will lead to further substantial reduction of reactor core damage frequency. All prepared modifications were preliminary assessed by PSA-1 methodology and now the reduction by one order concerning the value of CDF was calculated. This fact corresponds to the conclusions presented by SAIC for PSA-1. According to the SONS Decision No. 197/95 the proposals for measures for reduction of radioactive release into environment resulting from PSA-2 study shall be submitted within 6 months since these results were received by the Dukovany NPP. Implementation of such measures which could lead to reduction of core damage frequency shall be concerned for overriding . These measures are presented hereinafter in the text containing information about Living PSA. From the point of view of the PSA-2 these measures could be concerned for preventive, it means such measures which can reduce the probability of reactor core damage as much as possible thus also the possibility of radioactivity release will be reduced. If the CDF will be reduced by such measures by one order it can be expected that also the probability of radioactive release will be reduced by one order, in particular in the category ,,early large release of radioactivity". Therefore these measures are concerned for essential and with the highest priority. However, in order to perform detailed assessment and evaluation of measures which have the greatest impact on improvement of PSA-2 study results it is necessary to perform upgrade of PSA-2 results on the base of actual model of Living PSA-1, respectively on the base of PSA-1 model together with new emergency procedure. This upgrade of the PSA-2 was carried out by NRI 0eD and the final report including proposed measures was submitted at the end of 1998. Further category contains the measures for mitigation of radioactivity releases in the case the reactor core was damaged. We concern these measures for complementary and possible implementation some of them will be concerned within framework of detailed preparation of guidelines for management of severe accidents (Severe Accident Management Guidelines - SAMG). Development of the SAMG will be commenced by department 4190 when the validation and verification of the new EOP will be completed. PSA-2 recalculation taking into account actual Living PSA-1 , respectively release involving new EOPs (condition corresponding to the half of 1998) was performed by NRI 0eD. Final report was submitted in December 1998. This report contains also assessed proposals of preventive measures and selected measures for mitigation. The results show that even implementation of the preventive measures will reduce frequency of large radioactivity releases roughly by one half in comparison with the results presented by SAIC. Should we provide implementation of selected mitigation measures, the frequency of radioactivity releases about E-06 / year could be achieved. Regarding the term in which this report was submitted we are still getting acquainted with its conclusions. 9.5 Information about risk monitoring system and its further development The monitor of risk ,,SAS" (supplied by SAIC and financed from the US DOE resources) is used in Dukovany NPP from 1995. With assistance of this monitor of risk (the SONS property) the analyses of equipment unavailability of individual units were performed for month's period from 1995 to 1997. 288 This risk monitor makes possible operative evaluation of core melting probability at the nominal arrangement of unit (unit at nominal power operating both turbines). Impact of unavailability is reviewed both by increase of core melting probability in comparison with the basic level (all equipment available) and by cumulated risk (product of core melting frequency increase against the basic level and unavailability period). Contributions to the total cumulated risk from each individual equipment were evaluated yearly followed by development of proposals for cumulated risk reduction for next year. In order to provide optimisation of planned repairs the table containing combination of actual unavailability of safety related equipment (allowed by procedure A04 - Technical Specifications) together with their contributions to the risk was prepared. Using the risk monitor also period of equipment unavailability allowed from the point of view of the SONS can be determined. In such way several times an application of Dukovany NPP for exception from Technical Specifications submitted to the SONS was successfully justified, in order to perform equipment repair without need of unit outage. In the last year the utilisation of SAS risk monitor was stopped, since actual Living PSA-1 differs significantly from the PSA model applied in the SAS, thus also results of SAS could be unreliable. Since upgrade of SAS would be cost demanding and moreover other studied PSA (Shutdown PSA, PSA-2) can not be involved into the SAS the risk monitor supplied by NUS/SCIENTECH will be implemented in Dukovany NPP for future. This monitor called ,,Safety Monitor" was purchased by the CEZ join stock company for Temelin NPP with extension of user licence for Dukovany NPP also. This Safety Monitor makes possible application of PSA-1 study , Shutdown PSA as well as PSA-2 study and its upgrading will be performed in parallel with the Living PSA Project. Implementation of this monitor will be commenced this year with the aim to obtain risk monitor with capability to review differences between individual units operated in real time within computer network. NPP Dukovany Living PSA 8 1,00E-03 3 j§| 1,00E-04 4 S 1,00E-05 rrp : 11! ^ 1,00E-06 o 1,00E-07 JIM II a XII VIII XII VIII 1993 1995 1996 1997 Total value 'ATWS Floods 289 An example of Dukovany plant risk profile evaluated using SAS (the second quarter of 1997 at Unit 4 ) 1E-3 :• 8E-4v 7E-4-- 6E-4 •• 5E-4 •• 4E-4 ••1 3E-4 2E-4 1E-4 i-H r+i -t-i r+i r+n 1.4. 6.4. 11.4. 16.4. 21.4. 26.4. 1.5. 6.5. 11.5. 16.5. 21.5. 26.5. 31.5. 5.6. 10.6. 15.6. 20.6. 25.6. 30.6. Unavailable diesel generator QW (maintenance of SW pipeline) 1 Maintenance of water makeup pump TK20D02 2 Unavailable diesel generator QX (maintenance of SW pipeline) 3 Failure of intermediate cooling pump TF11 DO 1 4 Simultaneous unavailability of water makeup pump TK40D02(maintenance) and high pressure ECCS pump (tests) 5 Unavailable 6 kV buses BL and BM (failure of 110 kV reserve line) 290 Table 1: The allowed combinations of the simultaneous equipment unavailabilities and their risk \~ o D0 1 I D0 1 Q )D0 1 o •o 5 5 ra (n QL to CO S S O CO i 5. S 5. o o CO TO 8 csi z o Vi _J O iZ X S X 1 X TS' A * A * * • DG 8 SD 72 13 11 18 15 18 17 14 17 18 19 12 15 ** the second system only A TS: TF11(12,13)D01 20 SD 13 72 18 26 31 47 45 29 43 42 44 33 33 TS: TF1 i A* TS: Forb SHNC 68 8 17 43 27 53 71 72 72 59 72 iclon 72 72 72 Acceptable risk » Forb TS: Forb HNC 65 8 18 42 27 51 72 72 72 60 iclon 72 72 idon 58 * TS: CDVIMPa 70 9 19 44 27 54 72 72 72 64 72 72 240 72 61 Low risk ** AA Forb TS: SHNCiaCDV1MPa2 34 5 12 33 23 48 49 72 72 43 72 Iden 72 72 55 ft * * A TS: Test SOB 44 8 15 33 23 31. 72 72 72 72 72 58 61 55 72 10. Integrity of NPP Dukovany equipment 10.1 Reactor Pressure Vessels Integrity Reactor pressure vessels for Czech NPPs - Dukovany and Temelin - were manufactured in SKODA Nuclear Machinery in accordance with a strict requirements to content of detrimental impurities (P, Cu, As, Sb, Sn) to decrease their susceptibility to radiation damage. Integrity of these vessels are assured by a system of activities realized either in NPPs or in co-operated organisations which for NPP Dukovany represents: Low-leakage core configuration is permanently applied to decrease neutron flux/fluence on pressure vessel wall, Ex-vessel neutron fluence measurements to precise values of real neutron fluence on vessel wall and its trend, Standard surveillance specimens programme has been practically finished and its data have been analyzed. Charpy impact specimens have been reconstituted and tested for static fracture toughness determination, Supplementary surveillance programme was designed and inserted into all four vessels. This programme fully satisfies requirements of ASTM with low lead factor as well as irradiation temperature monitoring. The programme contains archive materials as well as cladding materials and specimens after irradiation and annealing to ascertain re-embrittlement rate, if such a requirement will arise as a result of plant life extension actions, Fracture mechanics properties of vessel materials have been collected to represent these type of materials (from Russia, Czech Republic, Slovak Republic, Hungary and Finland), additional data (mainly on arrest fracture toughness) are being gained and within the PHARE project new design fracture toughness curves will be issued, Re-calculation of pressurised thermal shock regimes and their impact on vessel integrity started several years ago taking into account new definitions of regimes as well as IAEA methodology. Most severe regimes have been analysed and maximum allowable critical temperature of brittleness has been found close to the designed one. These calculations are going on with concentration on LOCA, SLOCA and primary- to-secondary circuit leakage, Vessels are fully inspected every four years, SKIN manipulator from inner surface and USK-213 from outer surface are used for these inspections, as well as other special equipments (for bolts, nuts, nozzles etc.), Preparation for qualification of NDE methods, equipments and operators have started in accordance with ENIQ requirements, within PHARE projects several mock-ups have been prepared and finished. All further inspections will be realized only by organisations which fully fulfill these requirements, Instrumented hardness measurements are part of in-service inspection programmes - measurement of cladding as well as of outer surface is included, SONS Requirements for reactor pressure vessel integrity assessment were prepared and approved for the use. Procedure for lifetime and integrity evaluation of VVER pressure components have been also prepared by NRI and under Czech Association of Mechanical Engineers (ASI) is now under final stage of approval, 292 Periodic assessment of reactor pressure vessel integrity under these new requirements and procedure in now under way and will include all existing data of vessels (including their defectness, embrittlement, neutron fluence etc.). Reactor pressure vessels in NPP Temelin are still not yet in operation, but the following activities have been realized to support their integrity evaluation and safety : A wide experimental programme was performed which included mainly determination of special mechanical properties and radiation as well as corrosion mechanical damage, both necessary for integrity evaluation, Standard surveillance programme (designed by OKB Gidropress) was fully re- designed because of its disadvantages (irradiation temperature,, neutron fluence determination etc.) and special holders were welded into the vessel. Specimens will be located in special containers which also fulfills requirements of ASTM (low lead factor, irradiation temperature equivalent to beltline region), All other actitivities remain identical with NPP Dukovany, i.e. ex-vessel neutron fluence monitoring, low-leakage core planning, four years period of ISI and qualification of NDE, instrumented hardness measurements of cladding and outer vessel wall, PTS re-analysing, collection and analysing of fracture mechanics data is under preparation, SONS requirements and ASI Procedure for vessel integrity and lifetime evaluation will be also fully applied. 293 10.2 Steam Generators integrity 10.2.1 Evaluation of the steam generator tube damage The first tube plugging occurred at Dukovany NPP after first three years of operation, afterwards, however, the rate of plugging has been stabilized. Plugging is applied on tubes with indications higher than 60-75 %. Activities performed: - In-service measurements on tubes, carried out by the eddy current method - Water chemistry monitoring and hide out return evaluation - Crevice water chemistry evaluation by means of MULTEQ computer code - Modelling experiments, stress corrosion cracking tests - Analysis of pulled out tubes - Leak limit determination - Plugging limit determination For assessment of the steam generator aging an evaluation system has been prepared, enabling to perform the above mentioned analyses for concrete conditions of WWER steam generators. Principal features of the system are: Relations for critical crack size and permissible and measurable leak through the defect Introduction of a hideout return evaluation procedure. At NPP Dukovany analyses have been carried out to identify chemical conditions in crevices A knowledge of the local crevice environment, based on a HOR analysis by the MULTEQ code, a temperature analysis and a model of dynamics of concentration. In case of occurrence of short-term deviations in water chemistry, these changes can be taken into account. A model of damage initiation and kinetics of SCC crack growth. It will be experimentally determined on a structurally identical material under conditions simulating the local environment Time to defect initiation is to be measured by the method of constant loading of C-rings. By slow loading of CT-specimens (RDT test) a threshold value of the stress intensity factor for crack growth should be determined, together with the dependence of the growth rate on the J-integral. 10.2.2 Leak Limits The basic approach to the determination of primary-to-secondary leak permissible limits is based on measurement on tubes with through-wall cracks. The tests were carried out on equipment which enabled simulation of the primary and secondary circuit temperatures and pressures of WWER 440 and 1000 steam generators. Leakage of medium through an axial crack was measured in time dependence of exposure to the pressure inside the tube pressurized by the primary water from the experimental reactor water loop. - On the basis of graphical evaluation using lower bound curve of the measured values the permissible leak rates have been determined, as 8 l.h"1 for crack length of 11.3 mm for WWER 440 and 10.2 mm for WWER 1000. 294 - The leak rate value of 5 l.h"1 is recommended to keep secondary water radioactivity below a sufficiently low level under operating conditions. 10.2.3 Tube Plugging Limits - Are based on experimental measurement of the critical pressure at the rupture of a tube with a crack. - Burst test were carried out at pressures up to 100 MPa and at temperatures of 290° and 320° C. - The critical pressure was measured on tubes with artificial corrosion defects prepared by EDM method and on the pulled out tubes of WWER stream generators. - The resulting limit for steam generator tube plugging of 67.6 % was established. The plugging limit value of 70 % has been suggested with respect to WWER 440 operating experiences. 10.3 Piping integrity NPP piping systems are designed in accordance with valid codes and standards (Russian Design Code, ASME III ...) and initial piping quality preservation is assured by maintenance codes (ASME XI) involving in-service inspection, repairs etc. These Codes ensure basic safety level. For ensuring higher level of the nuclear safety it is currently supposed that pipe ruptures could result from random events induced by unanticipated conditions. The assumption requires to perform other safety evaluations and measures. Two main methods of evaluation are used: • Leak-Before-Break concept • Evaluation of the effects of postulated pipe rupture 10.3.1 LBB concept application During the LBB evaluation of all Czech NPPs knowledge was created especially concerning material data and components failure behaviours. Material database were collected containing tensile, fracture, corrosion, fatigue, fatigue corrosion, SCC properties for whole range of operating conditions. Large scale experiments and finite element method calculations enabled us to solve components with complex geometry. Experiments were performed on safe-ends, elbows, T-pieces, reactor nozzles. NPP Temelin was evaluated in following scope: all piping with diameter larger then 100 mm inside the containment. They are main circulating line, surge line, high pressure ECCS, low pressure ECCS, accumulator lines, residual heat removal system, piping to bubble condenser tank, coolant purification system. Some pipelines did not meet the requirements due to either pipe size or conditions decreasing the leakage. The LBB application was not successful on two pipelines of low pressure ECCS (due to high seismic load on long rising pipe), high pressure ECCSs, coolant purification system, piping to bubble condenser tank (due to small pipe size). The LBB application on piping of small diameters is well known problem caused by large ratio of postulated crack to circumference. 295 Three independent leak detection systems enabling leak detection of values down to the value of 3.8 1/min are planned to be installed. The systems are: acoustic emission, leakage sensor tube and humidity monitoring. The LBB evaluation of NPP Dukovany was limited only to main circulating piping and surge line. The application was successful in both cases. Although calculation part of the LBB is completed, leak detection systems suitable for the LBB are not installed. The pipe whip and pipe whip restraints analysis of NPP Dukovany is now underway. The same analysis will be performed on non LBB piping systems of NPP Temelin. 11. Modernisation, equipment innovation and safety upgrading programme. 11.1 Introduction This information is aimed at presentation of ongoing Programme of Equipment Innovation and Modernisation of Dukovany NPP and to show reasons leading to its origin, its supposed progress as well as its influence on nuclear safety. The Programme was entitled ,,MORAVA". We would like to express by this title our attitude to the region and environment surrounding our NPP. The title of Programme was created using initial letters of the following English words: J|DERNISATION i.e. modernisation (equipment and systems) ECONSTRUCTION i.e. reconstruction (equipment and systems) NALYSES i.e. analyses (safety cases) IQLID ATION i.e. validation (results and process correctness verification) The purpose of this connection presents also the mainframe of the Programme: Modernisation and Reconstruction through Analyses and Validation The Dukovany NPP belongs today to well safe and reliable power source of the Czech Republic. Implementing this Programme the Dukovany NPP will meet also increased requirements on safety, determined by the International Atomic Energy Agency in Vienna, in order to be classified among the other top quality nuclear units of the same type and age. Our effort is directed to allay inappropriate doubt resulting from NPP operation and any impact on its surroundings. 11.2 Programme Morava - main objectives The main objectives can be summarised in three areas: • to obtain an operational licence for Dukovany NPP operation up to year 2025 • to reach an economical competitiveness on deregulated market • to achieve a safety level comparable with the best NPPs 296 11.3 An Approach to the area of modernisation and safety enhancement The interest of Dukovany NPP was focused already from the beginning of operation of each unit on the area of safety enhancement. Condition of NPP designed in the seventies and commenced from 1985 to 1987 was continuously upgraded based on needs and requirements of operation as well as on the actual condition of plant including new requirements for safety. Such review was based on the actual condition of plant for given period. During work on design of nuclear power plant the original Russian regulative was used for the area of safety. In the course of review performed by the State Office for Nuclear Safety (SONS) , IAEA, external as well as internal audits and other reviewers, the condition of plant was on the higher level in comparison with condition in the year of units start-up. The Dukovany was reviewed by the IAEA mission as well as by the technical audit from the point of view of valid standards issued by the IAEA, safety regulative in force in the Czech Republic as well as internationally accepted concepts, practice and generalised national standards. Standards of this type are developed (and revised) by the IAEA. This Project called NUSS (Nuclear Safety Standards) was commenced in half of the seventies. However the safety rules which were in force in the country of original designer was used. In such way, implementing existing and by the Dukovany NPP accepted recommendations, the Dukovany NPP reaches step-by-step further qualitatively higher grade of nuclear safety assurance. It should be achieved that the Dukovany NPP could be concerned for acceptable not only for European respectively world-wide technical public but also for general public as it was so far. 11.4 Individual steps of modernisation after commercial operation commencement a) History of construction and commercial commencement of operation The primary part of Dukovany NPP units was designed and completed according to the design from the former USSR and the secondary part according to the design prepared by the former CSSR. Prevailing part of main primary equipment (reactor, steamgenerators, pressuriser) as well as all the equipment of the whole secondary part were manufactured in the former CSSR By this fact a controlled and documented quality on the relatively good level has been guaranteed what established good basis for positive operational results. The technical design of the Dukovany NPP was prepared on the basis of a treaty between the governments of the USSR and the CSSR on the 30th April 1970 (a treaty concerning co- operation during NPP construction) and according to the contract between ATOMENERGOEXPORT (supplier) and SKODAEXPORT (customer). In the order the so called design basis accident was defined also (taking into account the instantaneous cross rupture of cold leg on one primary coolant loop with nominal diameter 500 mm in the inseparable part). The main designer of the primary part was the company LOTEP from the former USSR; the main designer of the secondary part including authorised supervision on the entire construction was the company ENERGOPROJEKT Praha; the main supplier of buildings was the company ,,Prumyslove Stavby Brno" and the main supplier of technology was the company ,,SKODA Praha". 297 History of NPP units 1st unit 2nd unit 3rd unit 4th unit commencement of construction 01/1979 01/1979 03/1979 03/1979 connection to grid 24.2.1985 30.1.1986 14.11.1986 11.6.1987 trial operation from 3.5.1985 21.3.1986 20.12.1986 19.7.1987 commercial operation 3.11.1985 21.9.1986 20.6.1987 19.1.1988 end of approval process 12.12.1988 15.12.1988 14.6.1989 28.5.1990 b) Situation after commencement of the Dukovany NPP under commercial operation Based on the international knowledge it was decided about plant modernisation within framework of so called ,,NPP Backfitting". What does the ,,NPP Backfitting" meant? On the 20th November 1986 the Government Decree of the CSSR no. 309 was accepted which enacted the so called ,,Dukovany NPP Backfitting", i.e. implementation of the set of investment activities with the main objective- nuclear safety enhancement". The preliminary design was developed in 1990 and in 1991 its implementation was started. Today the main measures are implemented. Completed activities selected from the ,,NPP Backfitting" are for example: 0 improvement of electronic fire signalling systems as well as systems outdoor of the main block 0 haloid fire extinguish equipment for electrical equipment of NPP units 0 back-up of the 4th system of the power supply of the 1st category of reliability for home consumption 0 cooling of the steel roof construction in turbine hall 0 provision of the central oil plant by fire extinguish equipment 0 hydrogen elimination in atmosphere of the hermetically sealed area in the case of accident with release of coolant 0 system of teledosimetry Other identified and resolved problems after start-up of units Naturally such complex technical plant, what the nuclear power plant can be considered, requires solution of several problems after units start-up. To these belong in particular: • solution of the back part of the fuel cycle; it means solution of handling with spent fuel assemblies from reactor. They were particularly the following: 0 to make the refuelling pool more compact 0 construction of Interim Spent Fuel Storage • Other continuous upgrading implemented on the basis of needs and requirements of operation, for example: 0 Fire protective coats of cabling 0 cooling tower reconstruction 0 Signalling system for inhabitants in surroundings in the case of accident 298 During time the need of further modernisation upgrading became evident based on the comparison of nuclear safety standards of advanced countries as well as based on the operational experience. The work performed so far resulted in the development and implementation of the complex programme of equipment renewal under the name ,,MORAVA". 299 11.5 Technical assessment performed from 1991 to 1997 After 1990 the real level of safety of WWER 440/213 type of reactors was concerned and reviewed. The question was discussed whether these reactors can be left under operation. An international review of safety (initiated by the German government) known under the name ,,Green Book" performed by the GRS for the 5th unit of the NORD NPP (near Greifswald in the former DDR) revealed some differences of this type of plant. It was shown necessary to perform thorough assessment of actual condition of plant also in the case of the Dukovany NPP. Number of analyses and supporting programmes were performed for our NPP both by the experts from the Czech Republic and within the framework of international activities. These data were used for development of Equipment Reconstruction Programme of the Dukovany NPP. It concerns in particular the following: • ,,Operational Safety Report" after ten year operation of the Dukovany NPP • technical audit of the Dukovany NPP (internal, external - see below) • IAEA missions (OSART, ASSET assessing teams and team for safety enhancement) • state regulatory body requirements • PSA level 1 (Probabilistic Safety Assessment) • supporting analyses, PHARE (EU) as well as IAEA Projects • exchange of operational experience within the frame of WANO • common activities of WWER 440/213 units operators • analyses performed within framework of ,,Emergency Operational Procedure" development • review of conclusions of the above mentioned ,,Green Book" Technical Audit of the Dukovany NPP The mapping of actual condition of Dukovany NPP plant served as the initial step for development of modernisation work which was carried out within framework of the internal technical audit, (i.e. assessment of the plant condition by ten groups composed from more than 100 experts from Dukovany NPP, NRI 0eD, 3E Praha Engineering and the State Office for Nuclear Safety). The audit was divided into two parts (internal as well as external audit). Results of the internal audit give among others the assessment of actual condition of plant according to the following point of view: • reliability of equipment and its influence on nuclear safety • failure rate and its impact on unit availability (impact on loss of production) • demands for maintenance • equipment remaining lifetime, stock of spare parts and possibility for their provision in future • other impacts (economy, ecology, fire protection etc.) 300 The overview of issues both from the point of view of nuclear safety and from the point of view of reliability of NPP operation follows from this internal technical audit. The Consortium EN AC was chosen within framework of PHARE Project to be an external reviewer - auditor. This Consortium plays also the role of consultant for European Union. An independent view was reached through performance of the Evaluation Report concerning the Dukovany NPP safety. Final Report presented by the external auditor served at the same time as confirmation of correctness of approach of the Dukovany NPP concerning the prepared equipment modernisation. An additional step concerning verification of the Dukovany NPP approach to the prepared modernisation was the IAEA Mission on safety enhancement. IAEA Mission on safety enhancement Based on the Dukovany NPP application (through the State Office for Nuclear Safety) the IAEA mission was carried out on the Safety Enhancement Issues. What is the proper sense of this mission? It concerns assessment of safety enhancement what means an extensive review of selected issues in nuclear power plant design as well as current condition of safety enhancement at nuclear power plant. This mission was performed by the international group of experts co-ordinated by the IAEA based on the technical documentation and discussion with organisations involved in design, construction and operation of plant. The scope of review presents status of solution of Design Safety Issues of given NPP with emphasis on those whose were identified in previous IAEA Reports relevant for this NPP. The mission provides review of the following technical areas based on the document IAEA- EBP-WWER-03 Safety Issues and Their Ranking for WWER 440/213 Nuclear Power Plants : 1. General Issues (assessment and qualification of components and equipment with an particular attention for qualification in the case of emergency conditions) 2. Reactor Core 3. Components Integrity, i.e. review of integrity observation under different conditions (whip restraints of primary circuit, non-destructive inspection and implementation of the LBB (leak before break) concept with possible focus on selected issues concerning RPV (reactor pressure vessel) integrity. 4. Systems including the system of primary circuit coolant and methods of mitigation of consequences of primary circuit rupture, problems concerning reliability of emergency core cooling system, SGfeedwater supply system and service water system. 5. I&C system 6. Electric power supply system 7. Hermetically sealed area 8. Internal hazards 9. External hazards excluding seismic events 10. Accident analysis 301 This mission was carried out in Dukovany from the 25th September to the 13th October 1995. Following preparatory workshops in the Dukovany NPP the final meeting of experts was held from the 2nd to the 6th October in Vienna and then from the 9th to the 13th October in the Dukovany NPP also. In addition to participants from Dukovany NPP (65) and NRI 0eD (Nuclear Research Institute) there were involved also 5 IAEA experts as well as experts from France, Russia, The Slovak Republic, Spain and the US one expert from each country. Also one expert from each of the following Czech organisation was involved : SVUSS Bichovice, EGU Praha, EGP Praha, and 3E Praha Engineering. The set of NUSS Standards and Guides issued by the IAEA, safety regulative valid in the Czech Republic as well as internationally accepted concepts, practices and general national standards were used as the initial bases for review of condition of the Dukovany NPP. Complementary also the safety rules valid in the country of initial WWER designer (it is in Russia) were applied. The Final Report was reviewed at the end of 1995 and complemented in 1996. This report is available in the Dukovany NPP and in addition also in the SONS. The report presents further relevant input for modernisation development. Conclusions from all the assessments are presented in the following areas: • Safety Issues in the Dukovany NPP design was assessed by calculations based on probabilistic analyses. This assessment follows from the Operational Safety Analysis Report (OSAR) and in addition from the IAEA recommendations, i.e. Safety Issues resulting from extensive evaluation of WWER 440/213 design performed by the IAEA. This assessment contains 74 Safety Issues divided into three categories according to their relations to nuclear safety as well as 13 operational recommendations. Their fulfilment in Dukovany NPP was verified by the IAEA mission in 1995. Resolved activities with higher priority are for example: containment spray system sumps protection (TQ - Safety Systems), modifications at the level +14,7 m in the intermediate building, renewal of I&C systems etc. • consequences of lifetime exhaustion of individual equipment or their components respectively in relation to the lifetime given in design or by manufacturer or estimated from operational experience (technical audit). The most relevant problem in this area is the I&C equipment, furthermore the emergency power supply of the 1st category of reliability, selected valves and pumps etc. • impact of equipment failure rate, cost demands for maintenance of operated units (technical audit), for example I&C equipment, reconstruction, equipment of dieselgenerator station etc. • area of supporting analyses, programmes, upgrade of design base documentation. For example equipment qualification, check of pipeline integrity, programme of probability of core melting (PSA level 1, PSA level 2, Living PSA , risk monitor and other) large set of analyses were performed also within framework of the new procedure development for ,,Accident Management". Results of the PHARE Project focused on solution of safety issues have contributed to this area in significant way. 302 11.6 Activities of Equipment Upgrading Programme The above mentioned activities resulted in the set of recommendation, topics and requirements for measures leading to fulfilment of Equipment Upgrading Programme objectives. The set of activities within framework of the Programme MORAVA was established based on the analysis and the following determination of priority of individual activities according to the ,,Rules for modification management of Dukovany NPP property" which have the nature of Equipment Renewal Programme. (ERP) The objectives of the Programme of Equipment Upgrading ,,MORAVA" in Dukovany Nuclear Power Plant are as follows: • to provide safe operation also in future 0 technical equipment of Dukovany NPP shall be further operated in accordance with increased requirements of legislature as well as requirements of SONS (State Office for Nuclear Safety - the main regulatory body) or with requirements which could be valid in the close future (for example IAEA recommendations); 0 to enable capability of the Dukovany NPP to achieve operational licence (achievement of regulatory body approval) after year 2000 up to year 2025 0 to achieve core damage frequency with value 10 E -05 for reactor-year 0 to be in compliance with requirements of European Union (EU) (achievement of international safety standards for smooth transition of nuclear industry of the Czech Republic into the European Union) • the operation must be economical 0 to provide operation of all four units of Dukovany NPP up to the end of their planned lifetime (it means to year 2015) 0 to create preconditions for lifetime extension of Dukovany NPP (achievement of operational licence up to year 2025) 0 to utilise design reserve of units of nuclear power plant (power increase) 0 to provide competitiveness both within the EEZ company and on the market with electricity in the Czech Republic. 0 to enable further operation and maintenance of equipment together with an early supply of spare parts and provision of reliable service in order to reduce unplanned loss of production due to equipment failure. The set of activities involved in the Equipment Renewal Programme is concerned to be an open programme enabling involvement of additional requirements arising in future. Determination of priorities is in compliance with an approved process. An important part of the ERP containing safety enhancement was named ,,Programme of Dukovany NPP Modernisation". (MOP) 303 Examples of activities selected from the Equipment Upgrading Programme MMORAVA". 1. Group of activities required by the SONS as well as IAEA Safety Issues, for example: • I&C equipment upgrading • Suction sumps protection in SG box • Equipment modification on the level +14,7 m in the intermediate building • Displacement of the sectional header of ultimate emergency feed water pump • Equipment qualification • Continuous measurement of level and pressure in SG compartments 2. Group of activities required for assurance of operation, for example: • Extension of Interim Spent Fuel Storage • Reconstruction and extension of diagnostic systems • Full scope simulator of main control room • Replacement of material of main condenser tubes • Uninterruptible power supply of safety systems • Replacement of electric equipment and I&C of DG station • Reconstruction of fan room equipment 3. Group of activities for improvement of operational economy, for example: • Utilisation of design reserve of NPP units • Boric acid recycling (H3BO3) 4. Group of other activities, for example: • Water activity measurement in coolant polishing system SVO -3 tanks • Fire water and drinking water distribution 5. Group of activities with the nature of analysis and safety cases (activities with intangible outputs), for example : • Analyses and safety cases 0 probabilistic safety assessment - the 2nd level of PSA (PSA-2) 0 HP pipeline integrity evaluation 0 analyses necessary for equipment lifetime extension 0 thermohydraulic analyses and analyses of modes of operation 0 safety and functionality cases of equipment 0 remaining lifetime of equipment assessment 304 11.7 Progress of the main results of probabilistic assessment of the Dukovany NPP Safety The part of the Safety evaluation of Dukovany NPP creates the probabilistic assessment resulting from international methodology which allows comparison with other NPPs. (Probability of reactor core damage frequency /CDF/; PSA level 1 results model development for Dukovany NPP) The table below presents successive steps of development of probabilistic assessment. 1995 - PSA for internal initiating events (including consequences of HP CDF = l,77E-4 /reactor-year piping rupture) and for internal flood 1996 - PSA for internal initiating events, for internal floods and for CDF = l,84E-4 / reactor-year internal fires 1997-Living PSA (internal initiating events, floods and fires) CDF = l,09E-4 / reactor-year 1998 - Living PSA (internal initiating events, floods and fires) CDF = 9,93E-5 / reactor-year 2010-based on the Living PSA (internal initiating events, floods and CDF = 1E-5 - 2E-5/ reactor- fires) - estimation after implementation of all modifications within year Programme «MORAVA" Currently the core damage frequency value is 9,93 . 10'3 / reactor-year (according to the above table 9,93 E-3 /reactor-year) By implementation of the proposed modernisation steps in the period from 2000 to 2010 the Dukovany NPP will reach through its core damage frequency values the same level at the best operated nuclear power plants in whole European Union. 11.8 The Dukovany NPP modernisation Programme ,,MORAVA" from the point of view of safety practice in EU countries. An effort of the Czech Republic toward to be involved within economical and political structures of the EU is one among generally shared political objectives. A significant role within this process will concern readiness of the Czech Republic in the area of operated nuclear power plants. In the recent two decades the quantitative as well as qualitative increase occurred world-wide in area of requirements on safe operation of nuclear power plants. On one hand safety requirements for new constructed plants were established what, on the other hand, provided initiation for solution of essentially complex problem which relies on the safety enhancement of nuclear power plants operated so far. This problem is particularly significant in the case of power plants designed according to the safety standards being in force at the beginning of seventies. 305 In connection with the above mentioned it is necessary to provide answer for several essential questions: • How the safety of the older nuclear power plants can be reviewed (it means choice of an appropriate set of criteria) • Does exist any international consensus concerning methodical (review) as well as practical (technical upgrading of plant) approach to review of safety of nuclear power plant ? • What role is to be played by international organisations such as IAEA, OECD/NEA, CEC, WANO? • What is the influence of other national specific factors which are to be concerned? It can be stated that currently a formalised approach to solution of given problems exists. It is generally known under the name ,,Periodic Safety Review (PSR)". This PSR in the Dukovany NPP was entitled as the ,,Operational Safety Review after ten year operation" and it was performed for all four units. At the same time it must be noted that the required legislature in order to perform PRB was not brought in harmony within EU so far. For example in France and Sweden the process for PRB follows directly from law, in Belgium, in the Netherlands and in Germany (in some cases of nuclear power plants) the implementation of PRB creates condition for operational licence and in Spain and in Switzerland the PRB shall be performed based on regulative of an appropriate regulatory body. In contrast to the standard safety report which requires fulfilment of conditions formulated in legislature of respective country the PRB is the safety analysis of operated nuclear power plant from the point of view of actual internationally recognised safety standards. Such standards are developed (and step by step reviewed) based on the IAEA initiative. This project which began in the 2nd half of the seventies is known under the name NUSS (Nuclear Safety Standards). The Modernisation Programme of Dukovany NPP is based on an extensive assessment of nuclear safety in Dukovany NPP. In the course of this assessment the current world trends described above were used in a large scale. 1. Safety evaluation for Dukovany NPP was, in fact, carried out in compliance with the methodology typical for periodic safety review. Assessment of current status of Dukovany NPP safety was carried out in the first phase. In the second phase deficiencies and design reserve were reviewed at the experts level. In the third phase the depth analysis of identified shortcomings was carried out and the choices of benefit of proposed remedy measures were assessed. 2. Safety evaluation was administratively divided to the preparatory phase, provided by NPP employees in co-operation with domestic organisations and to the finalising phase carried out by foreign organisations naturally with intensive co-operation of NPP as well as other organisations. 3. Ranking criteria were compiled based on the NUSS standard completed with other common recognised standards (for example IEEE standards, ASME codes etc.) 4. Safety evaluation was in fact completed at the same time within the framework of the PHARE Programme (technical audit of Dukovany NPP) and within the IAEA 306 Extrabudgetary Project ,,Safety Evaluation of the NPPs with WWER 440/213 type of reactor". 5. Both above mentioned Programmes are mutually complementing. In the PHARE Programme the safety evaluation was based on the complex set of criteria given under item 3). This evaluation was covered by formulation of the set of safety issues respectively the groups of similar problems. In certain sense the IAEA Project was solved in inverse order. This Project was developed by international working group of experts. The work of this team led to the formulation of the number of 87 generic safety issues (74 +13 operational) which are typical for given type of nuclear power plants. Based on the generic issues the approach of Dukovany NPP was verified by the IAEA mission. Respective assessment relies on the determination of relevance of given issues to the actual condition of Dukovany NPP and on the formulation assessing the approach of the Dukovany NPP. Based on the summary of above mentioned approach of Dukovany NPP it is evident that this process of safety evaluation corresponds to the actual practice applied in EU countries. The graph presented below shows the process of fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of the ,,MORAVA" - Equipment Upgrading Programme. Prehled plneni pozadavku MAAE (IAEA requirements fulfilment overview; prubezne = continuously; rocne -yearly; surname = in total; zbyvajici = remaining). 307 11.9 Conclusion Presented facts can be summarised in two main conclusions: • the Programme ,,MORAVA" developed by the Dukovany NPP is complex in nature and is developed in accordance with European safety practice • implementing this Programme in period from 2000 to 2010 the Dukovany NPP will achieve the level of safety and reliability comparable with well operated NPPs within EU countries. 308 12. Appendix 1 A) Completed Constructions and Activities in technological part of equipment of units in the Dukovany NPP Number Title Implementation from to 1 Heat-up of volume of emergency boric acid storage tanks 2 Rearrangement of algorithm of the MCP - the Is' part 1991 3 SG level measurement reliability enhancement (incl. 1991 WSNo.3212) 4 Automated check of chemical modes (WS No.2138 1992 incl. WSsNo. 3060,3470) 5 Radiation monitoring installation in hermetically 1996 sealed area (incl. WS No.ZL 4070) 6 Level measurement in Rad. Monit. Station beyond the retention basin 7 Hydrogen elimination in case of BDBA (WS 1997 No.2357) 8 Computer network reconstruction for radiation monitoring (incl. WSNo.4230) 9 Replacement of HP compressors for control of fast 1994 1996 acting valves 10 Backup of the 4th system of the power supply of the 1st 1993 1997 category of reliability (incl. WSs No.4014,4313) 11 System for teledosimetry (WS No.2145 incl. WS No. 1997 3818) 12 Provision of air overpressure in rooms ECR,RPS, 1995 MCR and comp. 13 Equipment for fire hazard workplaces of oil plant (WS 1995 No. 2127) 14 Compactness enhancement for spent fuel storage pool 1993 1994 (incl. WS No. 1493) 15 DIAMO adaptation (WS No.2500,2501) 1993 1994 16 Generator switches modernisation (WS No.2179) 1996 17 Cold supply plant reconstruction — freon replacement 1996 1997 (WSNo. .2101 incl. WS4298) 18 Water Treatment Plant Modernisation (WS No.2131 1997 incl. WSs. ZL 3187,4303) 19 Replacement of water supply route for IaC sensors 1997 washing (incl. WS No.2870) 20 Replacement of water and oil coolers in DG plant 1996 21 Main condensers reconstruction - replacement 1996 22 DG EL. equipment reconstruction (stage Al - 1996 1997 excitation) 23 0,4 kV Sectional switchboards reconstruction for 1995 1997 reliability improvement 24 Replacement of MC pumps for sealless pumps 1997 1998 25 Rearrangement of outdoor transformer standstills 1996 1998 (incl. WSNo. 3122) 26 Pressuriser RV reconstruction (completion of the 1997 second safety valve) 309 27 Replacement of coolers for DG plant II 1996 28 TQ pumps replacement for sealless pumps 1997 1998 29 The 4th system of the UPS - replacement of 1995 1997 accumulators 30 Coating of cabling in outdoor cabling channels 31 Rearrangement of DG signalling in MCR (stage A2) 1997 32 TF pumps replacement for sealless pumps 1997 1998 33 Continuous measurement of chemical mode in 1996 secondary circuit, part for twin unit II (incl. WS No. 2795) 34 Sectional switchboards service supply leads 1996 1997 35 Modification from UCPTE connection - pressure value signallling lead-out 36 Modification from UCPTE connection - blackout mode 37 Coating of crucial cabling in the intermediate building and in the turbine hall of the 2nd unit Number Title Implementation from to 38 Protective cap sealing surface smoothing 39 Electric hoists B 301 40 Heavy current distribution of reactor 41 RCH sampling from ECCS tanks 1997 42 Signalling and blocking of hermetically sealed doors 1997 in A,B0032/l 43 Signalling of valve 0.04.112.1.2 condition 44 Oil escape into rain water drainage 45 Ph and conductivity measurement 46 Electric drive of circulating cooling water pump 47 Draining route from RY30-21 to 1RY20B01 48 Temperature measurement labelling change in the system ZD on the 1st to 4th unit 49 Replacement of sensor type for TQ-058 measurement on the 4th unit 50 Replacement of used penetrations during overhaul 51 IaC replacement for HVAC systems, syst.S0J-Z8O- 00 52 DMER - Diferrential pressure gauge modification 53 PZR Water temp, measurement lead-out to ECR 54 Water level meas. in 0(7)TW10B02-6-TW 27 change of principle 55 Temperature measurement in UPS room 56 Isolated supply leads 6kV replacement (incl. WS No. 3113) 57 Loading test of 2501 crane in A,B 501 58 Signalling of pressure drop in high pressure air distribution system 59 IaC reconstruction on circulating water pumps in central pumping station II 60 Displacement of LP and HP nitrogen pipeline on the 3rd unit 61 Underpressure sensor in compartment A,B 305 62 Accumulators modernisation 310 63 Signalling cancellation on the panel B5 64 Temperature measurement in pure condensate tanks 65 Complex of C14 measurement in stacks 1,2 (icnl. WS 1997 No. 3638) 66 Cabling coating - fire-protective measure (crossing places in intermediate building) 67 Duplex, oil .filtr on TG- oil distribution 68 Modif connection of SKODA transf- 250MVA2,3,4AT01,02 69 Fitting of isolating valves into circulating cooling water routes 70 IaC of Circulating water cooling pumps in the Circulating pumping station I - total innovation 71 220 MW Gen. power supply controller 72 Installation of silencer on exhaust of PZR RV from the 1st to 3rd unit 73 Rennovation of spherical surfaces of caps on PZR and on AT (accumulator tank) 74 S02 DN250 valve replacment 1994 1995 75 TIS to LAN connection 1994 1995 76 Involvement of units into frequency regulation 77 Application of clarified water for reconditioning of the catex upper layer 78 Replacement of coolers manufactured in the Sowiet Union 79 DUS system extension for newly connected switchboards 80 RSS 111 Exposure rate monitoring modernisation 81 Gammaspectrometry modernisation 82 Tritium measurement modernisation 83 Condensate cooler 84 Equipment for alpha - radiation measurement reconstruction 85 TLD Readers (thermoluminscent dosimetry) Number Title Implementation from to 86 Waste water measurement innovation 87 Duplex.oil. filtr on TG - lubricating oil 88 Unit protection system and TG protection system innovation (incl. WS 3035) 89 Cabling routes reconstruction for loops and Re temperature measurements 90 Tubes replacement and change of summer coolers on the 1st unit 91 Completion of the 4th system of the UPS 92 Bolts inspection on SG, PZR, AT 93 Equipment for stator water bypassed filtration 94 Air chamber (type ARF) fitting on fast acting valve 95 Common home consumption 96 Coating of additionaly layed cabling 97 Automated switch-over from the suction from sump in 1997 compartment to suction from LP ECCS 98 Completion of expansion receptacle to transformers 311 99 Mech. coupling of tap switching of transformers 100 Improvement of cooling of penetrations on 15,75 kV transformers 101 Modification of common switching station 102 UNICOP manipulator 103 Coating of crucial cabling in the intermediate building 1996 and in the turbine hall of the 1st unit 104 Coating of crucial cabling rooms of the Is' and the 2nd unit 105 Signalling lead-out from non-standard to IVS 106 Measurement of air flow in stacks (incl. WS 3627) 1997 107 Coating of cabling for switchboard plant 9CH703 108 CSP reconstruction - other contracts- pipelines 109 CSP reconstruction - other contracts- HVAC of the 1st unit 110 CSP reconstruction - other contracts- HVAC of the 2nd unit 111 CSP reconstruction - other contracts- HVAC of the 3rd unit 112 CSP reconstruction - other contracts- HVAC of the 4th unit 113 CSP reconstruction - other contracts- HVAC common part 114 TV cameras connection of the GEMINI system 115 TH 2.4 U01 blocking 1997 116 TH 2.7 U01 blocking 1997 117 H2SO4 acid level measurement in water treatment plant 118 Transfer of raw water level measurement in water reservoir to CHEMIS B) Other completed construction and activities in the Dukovany NPP affecting nuclear safety (incl. Fire Protection, Safety and Health Protection at Work), Radiation Protection, Security and Emergency Preparedness or affecting the human factor in plant or in region. Number Title Implementation from to 1 Radioactive waste treatment (incl. WSs No.2788, 1994 4012) 2 AKOBOJE (NPP automatic security guard complex) system 3 EFS system improvement for the 1st to the 4th unit 1993 1996 (incl. WSNo.3173) 4 Halonic spraying extinguishing equipment for the Is1 1993 1996 to the 4th unit (incl. WS No.2340) 5 Cooling of roof steel construction for the 1st to the 4th 1994 unit 6 Fire training area for NPP Dukovany Fire Brigade 7 Training channel for Fire Potection 312 8 Health Service Centre extension (part used by the Inf. Centre , incl. WS No. ZL 2953,3087) 9 Oil plant extension 10 Special Civil Defence Building , building 1,2,3 11 Retention tanks on the Skryje creek 12 Fire Brigade garage extension 13 Radioactive waste disposal 14 Lift for punts 15 Warning and notification of inhabitants - the 1st stage (WS No. 2180) 16 AKOBOJE reserve entry 17 Interim Spent Fuel Storage - constr. part incl. Fire 1991 continued Prot. (WS No. 2374) 18 AKOBOJE system optimisation (WS No.2664 incl. 1997 WSs. No. 2119,3558) 19 Main porter's lodge modification 20 Dukovany NPP computer network 21 Warning and notification of inhabitants - the 2nd stage 1992 (WS No. 2180) 22 Identification card sensors innovation (WS No. 2835) 1994 1995 23 Intermediate ceiling in rooms PPR and common 1996 1997 control room (CCR) 24 Construction of new telephone central office 1996 25 Replacement of HERKUL-S in system of warning and notification of inhabitants 26 Completion of acoustic and luminous signalling in water purification station JIHLAVA 27 Countemoise measures in compressor plant 28 Slant of floor in intermediate building + 14,7 m 29 Fire water manhole No. VII 30 Fire water manhole No. VIII 31 Fire water manhole No.IX 32 Fire water manhole No. X 33 Built-in platform in compartment B0060/2 34 Partition wall in room 447-PT.E.-the 2nd unit 35 Partition walls in corridor 410- the 3rd unit 36 Protective roofs over switchboards 37 Fire signalling system in water purification station JIHLAVA 38 Service road in outdoor switchboards 39 Control of hot water supply for heater OVL 40001 40 Mechanical gate with electric opening in both twinns 41 Access road to Central Pumping Station II 42 Unified time in computer network 43 Fire protective equipment -P0E, corridor 219 +5,40 44 Cooling towers - el. distribution around towers 45 Drainage A,B 521/1,2 a A,B 522/1,2 46 Fire protective equipment - turbine hall,"wall A" twin 1,11 47 AKOBOJE external barrier resistance enhancement 48 Measuring instrument AQUAMER 49 HVAC engine room inP0E 50 Computer network power supply provision 51 Sprinkling room for screens - cooling water plant No. 3 and No. 7 313 52 Ground road - cooling water plant No. 3 and No. 7 53 Waste piping from HVAC units 54 Emergency sprays connection in operational building III 55 Cooling tower No. 3 and 7 modernisation 56 FLUCOMAT- equipment for identification of oil substances 57 Reinforced road at Cooling water plant 58 Accessibility of cabling room at longitudinal corridor - 3,6 m 59 Building 529/1-16.2 - modification (superstructure) 60 Fire Signalling Centrale completion with information system 61 Submersible wall handling at Pumping Station Jihlava 62 Container room for storage unacceptable substances 63 Roof entry at common switchboards 6 kV 64 Warning illumination of towers 65 Cooling tower No. 1 and No. 6 reconstruction 66 Lifting equipment in Radioactive Waste Plant 67 Cooling tower No. 2 and No. 6 reconstruction 68 Building Security Central Desk Pult - the 1st stage 69 Venetian blind in intermediate building - transversal corridor 70 Spectrometry analyser power supply modification 71 Shielding walls in reactor hall A,B 101 1997 72 Cacthing roofs over IaC switchboard 73 Interim Cacthing roofs Installation 74 HP compacting of solid Ra waste - el. junction 75 Hydrogeological wells in Dukovany NPP surroundings 76 Electric installation for medical purposes 77 Displacement of horns (Inhabitants warning and notif. system) - Oslavany,Stanovi§ti C) Activities implemented by department of radiation safety Number Title Implementation from to 1 Rad. Monitoring blowers replacement 1995 under way 2 Rad. Monit. Sampling valves from hermetically sealed 1996 area - replacement 3 TriCarb - 3H Measuring instrument innovation 1996 4 Gamma spectrometry equipment innovation 1997 5 Gamma spectrometry equipment innovation 1994 6 3H TriCarb instrument innovation 1994 7 Strontium measuring instrument Berthold - 1995 innovation 8 TLD reader innovation 1995 D) Activities implemented by department of nuclear safety Number Title Implementation from to 314 j 1 I Emergency Staff workplace equipment" 1996 E) Activities implemented by department of hard ware supervision An overview of activities implemented in the Dukovany NPP which were financed from maintenance budget. Number Title Implementation from to 1 Cooling tower overhauls 2 Penetrations for MCPs, PZR replacement 3 Isolated conductors 6kV replacement 4 Upper feedwater distribution nozzle assemblage; one- armed SG blowdown assemblage 5 Separator reheater revision for innovation 6 Coating of cabling by KS1 material 7 Repair of floors and walls in turbine hall and in reactor hall 8 DG oil cooler tubes replacement 9 TK, TJ pump electric drives feplacement for domestic one 10 11 12 13 14 15 F) Activities implemented by department of radiochemistry Number Title Implementation from to 1. Detectors for continuous gama - spectrometric 1986 monitoring 2. PDP (minicomp.) network + spectrometry analysis 1986 3. VAX (computers) innovation 1990 4. Tragbar gama-spectrometry 1986 5. alpha -spectrometry 1994 6. gama -spectrometry in stack No. 2 1996 7. detectors innovation 1995 1996 8. remote modules of gama -spectrometry 1995 1996 9. UV spectrophotometric instrument M80 1988 10. capilary electrophoresis 1996 11. AAS (atomic spectrometry) 1994 12. microwave decomposition of samples 1993 13. ion - chromatographic equipment 1995 14. FTIR spectrometer 1996 315 13. Appendix 2 Main Areas of Modernisation involved within ,,MORAVA" Programme 316 No Title .,.,„• Objective ; Benefit Term, of ; impl. 1. Steam generator sump protection Upgrading of safety system (suction in Safety level of design 15.10.99 compartment) according to the advanced IAEA improvement requirements and the SONS requirements 2. Equipment modification at level + 14,7m in intermediate Restriction of mutual effects of pipelines in the case Safety level of design 15.10.04 building (the set of activities increasing resistance of equipment of their rupture improvement and piping systems) 3. Instrumentation and Control system reconstruction [&C tools replacement according to the advanced Safety level of design 31.12.09 IAEA requirements and the SONS requirements improvement 4. Displacement of sectional collector of emergency feedwater Restriction of actual damage of emergency Safety level of design 08.10.99 pumps feedwater system by missiles improvement 5. Interim Spent Fuel Storage extension (ISFS) Provision of spent fuel storage capacity for whole Resolution of the 07.05.04 OJ plant lifetime back part of fuel cycle 6. Utilisation of unit design reserve (power increase) Enhancement of available power output Economy 06.10.06 enhancement 7. Full scope control room simulator Staff training quality increasing Preparedness of staff 11.10.00 enhancement 8. pH in the secondary circuit increasing Reduction of corrosion - erosion wear of plant Costs for equipment 31.12.00 (set of activities - replacement of main condensers and related maintenance activities enabling change of chemical mode) reduction 9. Reconstruction of electrical part of DG station - I, II stage Reliability of emergency power supply source Safety level of design 20.09.02 Improvement improvement 10. Replacement of emergency feedwater pumps Reliability of feedwater supply into SGs during Safety level of design 30.07.99 transients improvement 11. Elimination of hydrogen during accidents Hydrogen removal from atmosphere of hermetically Safety level of design 30.06.99 sealed area in the course of severe accidents improvement No Title Objective Benefit Term, of intpl. 12. Completion of powered valves on the special drainage from Loss of coolant reduction during accidental water Safety level of design 09.08.02 main circulating pump deck (A, B 301) discharge from primary circuit improvement 13. Upgrading of valve automatics on emergency feedwater pump Reliability of feedwater supply into SGs during Safety level of design 09.06.00 injection lines emergency mode improvement 14. Reconstruction of protection acting on the signal ,,main steam Enhancement of level of equipment technological Safety level of design 16.10.98 header rupture" protection improvement 15. Relief valve of pressuriser - completion of protection against Upgrading of safety system (suction from Safety level of design 15.10.99 cold overpressure compartment) according to the advanced IAEA improvement requirements and the SONS requirements 16. Signalling of basement flood under turbine hall Staff information enhancement in emergency Safety level of design 29.09.00 situations improvement 17. TV camera control for check of reactor internals Maintenance automation enhancement of nuclear Decrease of radiation 06.07.01 oo power facility burden of staff 18. Fire protection (assembly of spraying fire-extinguishing Provision of continuous monitoring of fire exposed Safety level of design 05.10.01 equipment, electrical fire detection signalling etc.) rooms of nuclear power plant. improvement 19. Activity necessary for new emergency operational procedure Technology upgrading in compliance with Emergency 14.09.01 requirements of new EOP preparedness enhancement 20. Cabling rooms spray coating Fire resistance enhancement of cabling rooms in Safety level of design 01.12.00 NPP improvement 21. Reconstruction of subdistribution boards Replacement of equipment with exhausted lifetime Operational features 29.09.06 restoration 22. Change of connection of 110 kV power supply reserve Disponibility of reserve power supply for home Safety level of design 07.02.03 consumption improvement 23. Technical and supporting Centre Creation of control centre for accident management Emergency 16.12.03 and mitigation according to the advanced IAEA preparedness requirements and the SONS requirements enhancement No ,,-."-jt _,v*r->T «*"<<-• < -• fitlfe"' '"--'*••' "' *"'*'"** * "-""'* "' ; ' y « Objective Benefit Term, of impl. 24. Activity resulting from LBB Project (leak before break) Upgrading necessary for LBB statut. Nuclear safety of 24.10.03 plant enhancement 25. Activity resulting from operational experience, audit and from Technical safety enhancement to the level Safety level of design 14.09.01 assessment of reactor core melting probability (PSA) acceptable in EU countries improvement 26. Reconstruction of the air measurement in the stack Enhancement of level of releases monitoring Environment 26.07.02 according to the requirements resulting from new protection legislature for environment protection enhancement 27. Provision of habitability of main and emergency control rooms Provision of safe working place for plant control Safety level of design 07.10.05 - venting systems under normal and emergency conditions improvement 28. Fast acting valves - VELAN DN 450 Reliability enhancement for segregation of Additional restriction 27.09.02 hermetically sealed area in emergency conditions of Ra-substances release into environment 29. Source term in radiation surveillance Computation model of radioactive substances Emergency 01.08.03 spreading in hermetically sealed area in the case of preparedness radiation incidents. enhancement 30 Diagnostic systems Provision of continuous monitoring of relevant Reduction of scope of 08.08.03 (the set of activities enabling more extensive equipment components technical condition necessary repairs monitoring) 31 Intangibles Development of complete set of safety analyses Level of acceptability 31.12.08 (the set of analyses and safety cases required by the IAEA and the according to the IAEA and the SONS advanced enhancement SONS) requirements Involvement of Modernisation Areas within Equipment Reconstruction Programme ,,MORAVA" Equipment reconstruction Programme MORAVA contains the set of requirements for Dukovany NPP equipment modification, providing safe, reliable and economical operation at acceptable level up to end of NPP lifetime. This programme creates assumption both for lifetime extension in economically profitable period and for fulfilment of conditions of the Czech Republic entry to the EU in the area of nuclear industry. The Programmme of equipment reconstruction is an opened file which will be continuously completed. For the reason of this documentation the Programme was internally divided into following groups of activities: • group ,,S" (activities required by SONS, IAEA, PSA, EOP etc.) • group ,,P" (activities required for reliable operation) • group ,,E" (activities with significant economical benefits) • group ,,O" (activities which are not in direct relation to operation, they are directed to the area of supporting activities) • group ,,N" (intangibles, i.e. activity with the nature of analyses and safety cases). Modernisation Programme is the set of activities for equipment reconstruction ,,M0RAVA" which are in direct relation to the fulfilment of the SONS requirements (resulting from decision) as well as IAEA requirements (Safety Issues). In practice it contains the activities from the group ,,S", part of the group N and selected activities from groups ,,P" and ,,E". Intangibles - creates the set of analyses and safety cases directly interfacing fulfilment of the SONS requirements, the IAEA requirements as well as recommendation from audits. They are involved in the group ,,N".